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R. Michael GloverH. B. Robinson SteamENER
Electric Plant Unit 2Site Vice President
Duke Energy Progress3581 West Entrance Road
Hartsville, SC 29550
0:843 857 1704F: 843 857 1319
Mike. Glo vert duke-energy.com,
Serial: RNP-RA/15-0021
APR 01 2015U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2DOCKET NO. 50-261 1 RENEWED LICENSE NO. DPR-23
SUPPLEMENTAL RESPONSE TO 120-DAY RESPONSE SUBMITTAL TO REQUEST FORADDITIONAL INFORMATION ASSOCIATED WITH LICENSE AMENDMENT REQUEST TOADOPT NATIONAL FIRE PROTECTION ASSOCIATION (NFPA) STANDARD 805
REFERENCES:
1. Letter from W. R. Gideon (Duke Energy Progress) to U. S. Nuclear Regulatory Commission(USNRC) (Serial: RNP-RA/1 3-0090), License Amendment Request (LAR) to Adopt NFPA805 Performance-Based Standard for Fire Protection for Light Water Reactor GeneratingPlants (2001 Edition), dated September 16, 2013, ADAMS Accession No. ML1 3267A21 1
2. Letter from Martha Barillas (USNRC) to Site Vice President, H. B. Robinson Steam ElectricPlant (Duke Energy Progress), H. B. Robinson Steam Electric Plant, Unit 2 - Request forAdditional Information on License Amendment Request to Adopt National Fire ProtectionAssociation Standard 805, Performance-Based Standard for Fire Protection (TAC No.MF2746), dated October 23, 2014, ADAMS Accession No. ML14289A260
3. Letter from R. Michael Glover (Duke Energy Progress) to U. S. Nuclear RegulatoryCommission (USNRC) (Serial: RNP-RA/14-0122), Response (60-Day) to Request forAdditional Information Associated with License Amendment Request to Adopt National FireProtection Association (NFPA) Standard 805, dated November 24, 2014
4. Letter from R. Michael Glover (Duke Energy Progress) to U. S. Nuclear RegulatoryCommission (USNRC) (Serial: RNP-RA/14-0134), Response (90-Day) to Request forAdditional Information Associated with License Amendment Request to Adopt National FireProtection Association (NFPA) Standard 805, dated December 22, 2014
5. Letter from R. Michael Glover (Duke Energy Progress) to U. S. Nuclear RegulatoryCommission (USNRC) (Serial: RNP-RA/1 5-0006), Response (120-Day) to Request forAdditional Information Associated with License Amendment Request to Adopt National FireProtection Association (NFPA) Standard 805, dated January 22, 2015
6. Letter from Martha Barillas (USNRC) to Site Vice President, H. B. Robinson Steam ElectricPlant (Duke Energy Progress), H. B. Robinson Steam Electric Plant, Unit 2 - Request forAdditional Information on 60-Day Response to License Amendment Request to AdoptNational Fire Protection Association Standard 805, Performance-Based Standard for FireProtection (TAC No. MF2746), dated March 26, 2015, ADAMS Accession No.ML15057A403
SDDo
U. S. Nuclear Regulatory CommissionSerial: RNP-RA/15-0021Page 2
Dear Sir/Madam:
By letter dated September 16, 2013 (Reference 1) Duke Energy Progress, Inc. submitted a licenseamendment request to adopt a new risk-informed performance-based fire protection licensing basisfor the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2).
During the week of September 22, 2014, the NRC conducted an audit at HBRSEP2 to supportdevelopment of questions regarding the license amendment request. On October 23, 2014 theNRC provided a set of requests for additional information regarding the license amendment request(Reference 2). That letter divided the requests for additional information into 60-day, 90-day, and120-day required responses. The Duke Energy Progress 60-Day, 90-Day, and 120-Day responseswere conveyed to the NRC Document Control Desk via letters from R. Michael Glover onNovember 24, 2014 (Reference 3), December 22, 2014 (Reference 4), and January 22, 2015,respectively. The NRC and Duke Energy Progress agreed per telecom on March 25, 2015 that theresponse to the Clarification RAIs, PRA RAI 05.a.01, 15.01, 18.01, 21.01, 23.d and 24.01, FM RAI01 .b.01 and 01 .c.01, SSA RAI 07.01 and LAR Attachments A, B and L would be submitted to theNRC by April 3, 2015. Enclosed as agreed are the Duke Energy Progress responses to therequests for additional information conveyed via Reference 6.
Please address any comments or questions regarding this matter to Mr. Richard Hightower,Manager - Nuclear Regulatory Affairs at (843) 857-1329.
There are no new regulatory commitments made in this letter.
I declare under penalty of perjury that the foregoing is true and correct. Executed on2015.
Sincerely,
R. Michael GloverSite Vice President
RMG/jmw
Enclosure
cc: Mr. V. M. McCree, NRC, Region IIMs. Martha C. Barillas, NRC Project Manager, NRRNRC Resident Inspector, HBRSEP2Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)
U. S. Nuclear Regulatory CommissionEnclosure to Serial: RNP-RAI15-0021201 Pages (including this cover page)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING VOLUNTARY FIREPROTECTION RISK INITIATIVE
REQUEST FOR ADDITIONAL INFORMATION
VOLUNTARY FIRE PROTECTION RISK INITIATIVE
DUKE ENERGY PROGRESS
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2
DOCKET NO. 50-261
Safe Shutdown Analysis (SSA) Request for Additional Information (RAI) 07.01
In a letter dated March 16, 2015 (ADAMS Accession No. ML15079A025), the licensee responded to SSA
RAI 07, item 5, and stated that current transformers (CTs) located in switchgear or components that arenot credited for safe shutdown (SSD) were excluded from the CTs' open secondary circuit analysis. The
purpose of evaluating the potential for causing an open circuit in a CT circuit is to determine the
potential for a secondary fire in a different fire area than the originating fire. The concept for asecondary fire due to an open circuited CT applies to any CT circuit, regardless of its use or credit forSSD. Therefore, the licensee's disposition that, "CTs located in switchgear or components that are not
credited for SSD were excluded," is not applicable.
Provide a response that addresses the potential fires caused by open circuited CTs, regardless
of whether or not the CT has been credited for SSD.
Response:
In the initial response to RAI-SSA-07, item 5, we stated "CTs located in switchgear or components that
are not credited for SSD were excluded".
The evaluation in EC 93120 of Current Transformers (CT) with turns ratios greater than 1200 to 5 usedthe statement for systems 5040 (Generator System) and 5120 (Switchyard and Transformer), that basedon assumption 4 (EC93120) these CTs will not cause damage to any Safe Shutdown relatedcomponents". The basis for this statement is; should an OPEN secondary circuit be created by a fireevent along the cable route, the cable breakdown due to the high voltage created by the CT would alsooccur at the point where the insulation is fire damaged. Should the secondary fire be caused bycatastrophic failure of the CT itself, as postulated in the same reference, the damage would then becontained within the CT enclosure and would not propagate outside the enclosure or damage anyadjacent equipment. (Ref section 6.2.3 of the PIRT Panel's review NUREG CR-7150, Volume 1, which is
attachment 4 of EC 93120).
Design consideration number 5 in the original RAI-SSA-07 will be revised to be more specific on thereasons which allow this group of CTs to be evaluated. It will read as follows:
5. The majority of the CTs installed in the plant with ratios >1200:5, are designed such that they
supply electrical circuitry (protective / indication) that is enclosed within the switchgear. Forthese cases, no "external to the switchgear" cables are utilized in the design.
Page 2 of 10
CTs with turns ratios >1200:5 located in switchgear or components that are not credited for SSD
but with secondary wiring that does extend into fire areas containing SSD equipment will not
cause damage to any Safe Shutdown related components. The basis for this statement is;should an OPEN secondary circuit be created by a fire event along the cable route, the cable
breakdown due to the high voltage created by the CT would also occur at the point where the
insulation is fire damaged. Should the secondary fire be caused by catastrophic failure of the CT
itself, as postulated in the same reference, the damage would then be contained within the CT
enclosure and would not propagate outside the enclosure or damage any adjacent equipment.(Ref section 6.2.3 of the PIRT Panel's review NUREGCR-7150, Volume 1, which is attachment 4 of
EC 93120).
Fire Modeling (FM) Request for Additional Information (RAI) 01.b.10
In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to FMRAI 01.b and explained how the effect of the increased heat release rate (HRR) due to fire propagationin cable trays was accounted for in the hot gas layer (HGL) and multi-compartment analysis (MCA)calculations. In its response, the licensee stated "Fire spread in each tray is assumed to be offset by theburnout."
Provide technical justification for this assumption as it is not consistent with the flame spread rates forthermoplastic and thermoset cables recommended in Appendix R of NUREG/CR-6850, "EPRI/NRC-RESFire PRA Methodology for Nuclear Power Facilities: Summary and Overview" and Chapter 9 ofNUREG/CR-7010, "Cable Heat Release, Ignition, and Spread in Tray Installations During Fire(CHRISTIFIRE)."
In addition, the response does not address the potential effect on the zone of influence (ZOI) from theadditional HRR of the cable trays. Explain how the effect of the increased HRR due to fire propagation inthe cable trays was accounted for in the ZOI calculations; or provide technical justification for ignoringthis effect.
Response:
b. Based on the amount of combustibles (generally cable trays in the ZOI), the fire growth is
estimated using guidance from NUREG/CR-6850, Appendix R. The fire propagation among a
stack of vertical cables trays follows this general timeline:
TIMELINE TIMELINE DESCRIPTION
T, Time to build ignition source fire to scenario HRR = 12 minutesT2 Time to ignite the first target (tray) based on NED-M/MECH-1009
T, Time to ignite second tray = T2 + 4min
"1"4 Time to ignite third tray = T3 + 3min
T5 Time to ignite forth tray = T4 + 2min
T6 Time to ignite fifth tray = T5 + 1min
T, Time to ignite X tray = previous tray + 1min
The following properties are assigned to the horizontal cable fire growth for RNP based on
NUREG/CR-6850 and NUREG/CR-7010:
Page 3 of 10
Cable Tray Width 0.61 m Typical tray width
HRR per unit area 250 kW/m 2 NUREG/CR-7010, section 10.1
for thermoplastic cables
Using the values listed above, the heat release rate for the cable trays is calculated as the
surface area of the tray multiplied by the heat release rate per unit area. The angle of 35'described in Appendix R.4.2 of NUREG/CR-6850 is used for determining the length of the cable
trays in the stack above the ignition source so that the appropriate burning surface for each tray
is determined.
The total fire growth is based on adding the source fire HRR plus each tray HRR per unit time.
Fire spread in each tray is assumed to be offset by the burnout. If the fire grows large enough to
support a HGL, the time to HGL can be estimated. An adjustment was made to the process of
calculating the time to HGL by using cumulative HRR by comparing the energy required to
produce an HGL to the total energy produced by the fire. The MCA was performed in the same
manner as the HGL analysis.
"Fire spread in each tray is assumed to be offset by the burnout." means that as the progression
of the fire extends outwards, the burning region remains somewhat constant. In other words,
the flame spread outward along the cables will be equal to flame extinguishment along the just
consumed cables. This is shown in Figure 9-2 of NUREG/CR-7010, "Cable Heat Release, Ignition,
and Spread in Tray Installations During Fire (CHRISTIFIRE)."
If there are significant intervening combustibles, the total HRR of the fire will grow, expanding
the ZOI. The most probable increase in damage will be above the fire. The vertical ZOI would be
extended to the ceiling if multiple cable trays exist in the vertical ZOI for the source.
The process for determining the heat release rate associated with intervening combustibles
(e.g., those cable trays within the zone of influence of an ignition source) is summarized in
section 5.6.2 of Fire PRA calculation RNP-F/PSA-0094. The heat release rate for the cable trays
that are identified during walk downs is calculated assuming:
1. A cable tray width of 2 feet (0.61 m). Typical cable tray widths range from 6-inches to 3 feet
wide. A 2 foot wide tray was selected to represent for all RNP fire zones. This is consistent
with the average cable tray size as presented in NUREG/CR-6850, section R.4.2.1.
2. Initial cable tray burning length of approximately 3 feet (1 m). This value is representative of
the fire diameter for the ignition source, which is also the length of the first tray burning in
the stack.
3. A heat release rate per unit area of 250 kW/m 2 . This heat release rate value is
recommended in NUREG/CR-7010 for Thermoplastic cables.
Using the values listed above, the heat release rate for the cable trays is calculated as the
surface area of the tray multiplied by the heat release rate per unit area. The angle of 350
described in Appendix R of NUREG/CR-6850 is used for determining the length of the cable trays
in the stack above the ignition source so that the appropriate burning surface for each tray is
determined.Page 4 of 10
FM RAI 01.c.01
In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to FM
RAI 01.c and explained that intervening combustibles within the ignition source ZOI were identified in
the walkdowns, and that the HRR contribution from these combustibles were incorporated in the fire
modeling analyses. It appears that the ZOI calculations were only performed for 69 kW, 211 kW, 317
kW, and 702 kW fires. This would imply that the licensee may not have accounted for the additional
HRR from non-cable intervening combustibles on the ZOI.
Confirm that the effect of the increased HRR from non-cable intervening combustibles was indeed
accounted for in the ZOI calculations; or provide technical justification for ignoring this effect.
In addition, the licensee did not provide any details on how the HRR of non-cable intervening
combustibles was calculated. Describe themethodology that was used to estimate the HRR from non-
cable intervening combustibles in the ZOI, HGL, and MCA calculations.
Response:
c. To ensure that intervening combustibles (including non-cable intervening combustibles and
cables that are not targets in the Fire PRA) are properly accounted for in the fire modeling
analysis supporting the Fire PRA, walk down project instructions were followed:
" FPIP-0200, Rev. 8, Fire PRA Walk down Instructions. This procedure provides specific
guidance on dealing with intervening combustibles. The guidance consists of identifying the
intervening combustibles within the zone on influence and capturing them in the walk down
forms. Possible intervening combustibles included cables, trays, batteries/chargers, panels,
and equipment, etc.
* FPIP-0208, Rev. 5, Scoping Fire Modeling. This procedure provides guidance to consider
cable trays above the ignition sources up to the enclosure ceiling to prevent missing the
contribution from these cable trays to the heat release rate.
The practical implications of the guidance included in the project instructions listed above are that the
RNP Fire PRA includes the contribution from two types of fire scenarios from the perspective of this RAI.
These fire scenarios are:
" Those fire scenarios affecting the ignition source only. That is, there is no fire propagation
outside the ignition source. No propagation outside the ignition source is due to either no
intervening combustibles within the zone of influence, or credit to passive fire protection
features such as solid bottom trays.
* Those fire scenarios where fire propagates throughout the zone of influence. If during walk
downs, intervening combustibles were identified within the zone of influence, the heat
release rate contribution from these combustibles was included as part of the heat release
rate profile characterizing the fire scenario. The zone of influence extends to the ceiling of
the physical analysis unit.
i. Non-cable intervening combustibles were not found during RNP walkdowns, so there was no
effect on the scenario HRR.
Page 5 of 10
ii. Non-cable intervening combustibles were not found during RNP walkdowns, so there was no
effect on the ZOI, HGL, and MCA calculations.
Probabilistic Risk Assessment (PRA) RAI 05.a.01
In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to PRARAI 05.a and explained that, as part of the integrated analysis provided in response to PRA RAI 3,scenario development of fire propagation from electrical cabinets greater than 440V will be based ondraft Frequently Asked Questions (FAQ) No. 14-0009. The U.S. Nuclear Regulatory Commission (NRC)staff has provided comments on the treatment of motor control centers (MCCs) in "NRC Comments onMCC Treatment White Paper August 29, 2014" (ML14245A133) on the draft FAQ 14-0009. These NRCstaff comments recognize that the basis for the 0.1 multiplier is a 0.19 estimate of the probability of firebreaching a well-sealed cabinet (based on an evaluation of operating experience) and a 0.45 factor oflikelihood of damage to targets outside of the cabinet (based on a phenomenological fire modelingevaluation of the geometry of fires within MCC cabinets). There is no provision to change the 0.19 valuebut the method allows for re-evaluation of the 0.45 factor using phenomenological modeling of plant-specific cabinet geometry and ignition sources.
1. Provide the technical justification for any revised values used to represent the probability of
damage to targets located outside the MCC cabinet that deviate from accepted values. If thebasis for the revised values relies on phenomenological fire modeling that has not beenreviewed by the NRC, include a detailed discussion of the modeling that was performed and the
results that support the revised values.
2. Describe how these fires will be modeled in the integrated analysis provided in response to PRA
RAI 3.
Response:
With respect to the NRC staff comments (dated August 29, 2014, ML14245A133) on an earlier draftFPRA FAQ 14-0009, draft H of FPRA FAQ 14-0009, clarifies the scope of panel breaching due to energeticarcing faults to be limited to "well-sealed" MCCs greater than 440V, provides an expanded basis forusing a breaching factor of 0.21 and a severity factor of 0.454 to obtain a 0.095 multiplier that a firebreaches a well-sealed MCC and damages nearby (within six inches) external targets, and presents a firemodeling discussion based on NUREG/CR-6850 models to support credit for fire growth and propagationdamaging subsequent targets. While Fire PRA FAQ 14-0009 has not yet been accepted by the NRC, draftH represents the best available guidance produced by considerable collaborative efforts.
1. In the RNP FPRA, well-sealed MCC electrical cabinets greater than 440V were treatedconservatively relative to the guidance in draft H of FPRA FAQ 14-0009. This treatment of arcingfaults, which are not to be confused with HEAFs, is not applicable to non-MCC electricalcabinets. Although draft H of FPRA FAQ 14-0009 permits a case-by-case re-evaluation of the0.454 severity factor when the distance between the closest external target and the MCC ofinterest is more than six inches, the resultant multiplier would be proportionally less than the0.1 which was used by the RNP FPRA.
2. In the response to PRA RAI 3, these fires are included as part of the base model.
Any subsequent change in the accepted version of FPRA FAQ 14-0009 would be addressedthrough the normal PRA maintenance process.
Page 6 of 10
PRA RAI 15.01
In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to PRA
RAI 15 and explained that inadequate breaker fuse coordination is accounted for in the Fire PRA. Theresponse explains that the "cables causing issues for power supplies" were identified, along with theirrouting, and used "to create an assumed failures list for breaker coordination." Though it appears thatcommon power supply failures were modeled in the PRA, it is not clear whether possible ignition of
affected cables (common enclosure) is considered. The Fire PRA modeling should be consistent withNUREG/CR-6850, "Fire Probabilistic Risk Assessment Methodology for Nuclear Power Facilities,"guidance, Section 3.5.4.2, Step 4.2, which states that, "[i]n evaluating the adequacy of cable thermalprotection, the criteria for acceptance should be based on a secondary fire concern and not simply
exceeding the continuous or overload thermal limit for the cable."
Clarify how the risk associated with secondary fires from cables between uncoordinated breakers isaddressed and describe how these fires will be modeled in the integrated analysis provided in responseto PRA RAI 3. Alternatively, complete breaker coordination work prior to self-approval, updateAttachment S of the LAR as necessary, and discuss how it would be reflected in the Fire PRA for the
response to PRA RAI 3.
Response:
Breaker coordination was reviewed for Robinson. In some cases breaker coordination was achieved by
crediting cable length for the load, in others this credit could not be applied for the NFPA-805 analysis
due to the location of the fire. The upstream power supply was assumed to trip, causing the loss of
power to additional equipment as a result. Cable protection against overload and short circuit is a
consideration in the general design criteria for protective device selection, however is limited in
application. The Fire PRA currently models the upstream power trip, but does not model secondary fires
resulting from a lack of adequate protection.
PRA RAI 18.01
In a letter dated March 16, 2015 (ADAMS Accession No. ML15079A025), the licensee responded to PRA
RAI 18 and explained that treatment of self-ignited and cutting-and-welding fires is consistent with FAQ
No. 13-0005 for one fire compartment, and for other compartments, these fires were screened out
because they were determined to have "insignificant impact." In apparent contrast to this, the response
PRA RAI 08 states "cable fires due to cutting and welding are assigned no target sets because a
continuous fire watch with an extinguisher is required by procedure to be present during hot work
activities and is assumed to extinguish such a fire before it can spread beyond the original tray." The
cited responses to PRA RAI 08 and PRA RAI 18 appear to be inconsistent.
Clarify this apparent inconsistency and identify how these cable fires will be modeled in the integrated
analysis provided in response to PRA RAI 3.
Page 7 of 10
Response:
The statement in the response to PRA RAI 08 (Reference Robinson Letter RNP-RA/14-0122) is consistent
with Fire PRA FAQ 13-0005, as it assumes damage is limited to a single tray and no additional target sets.
For Robinson, a bounding approach was initially used to assess all cable fires due to cutting and welding
by conservatively applying the limiting source scenario CCDP or CLERP (not limited to a single tray) for
the applicable compartment to the cutting and welding scenario, effectively assuming the cutting and
welding scenarios had a target set equal to the most limiting scenario. If the bounding treatment
resulted in significantly high results, then a tray-by-tray assessment was performed per Fire PRA FAQ 13-
0005 to identify the limiting tray CCDP and CLERP for the compartment. The tray-by-tray assessment
was only needed for Fire Compartments 250 (Turbine Building) and 70 (Lower Hallway). This treatment
of the cable fires is included in the quantification results presented in the response to PRA RAI 3.
PRA RAI 21.01
In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to PRA
RAI 21 and based on the response, fires that are not believed to cause an automatic trip are assigned a
conditional probability of manual trip that reduces the likelihood of the associated fire scenario.
1. Discuss whether your review of the fire-induced initiating events is consistent or
conservative compared to the review steps described in NUREG/CR-6850 guidance in
section 2.5.3, "Step 3: Identify Fire-induced Initiating Events Based on Equipment Affected."
2. Discuss any Fire PRA model changes after completing these evaluations that will be included
in the integrated analysis provided in response to PRA RAI 3.
Response:
1. The review of fire-induced initiating events is consistent with the guidance in Section 2.5.3
of NUREG/CR-6850 in that a reactor trip was assigned as the initiator for fires that are not
believed to cause an automatic trip. Additionally, although the guidance in section 2.5.3
does not require the assignment of an initiating event for fire compartments where none of
the three conditions is judged to occur (i.e., no automatic, manual, or LCO forced trip), a
conditional trip probability is conservatively applied to account for operator discretion to
perform a manual trip even for a fire in an area containing no equipment important to plant
operations.
2. Consequently, no related Fire PRA model change has been included in the integrated
analysis provided in response to PRA RAI 3.
Page 8 of 10
PRA RAI 23.d
Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the
NRC. Section 2.4.4.1 of NFPA-805 further states that the change in public health risk arising from
transition from the current fire protection program to an NFPA-805 based program, and all future plant
changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on
CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to
the plant's licensing basis and describes a general framework to determine the acceptability of risk-
informed changes. The NRC staff review of the information in the LAR has identified the following
information that is required to fully characterize the risk estimates.
Section W.2.1 of the LAR provides some description of how the change-in-risk and the additional risk of
recovery actions associated with VFDRs is determined but not enough detail to make the approach
completely understood. Provide the following:
d) A description of the type of VFDRs identified, and discuss whether and how the VFDRs
identified, but not modeled in the Fire PRA, impact the risk estimates. Include any qualitative
rational for excluding VFDRs from the change-in-risk calculations.
Response:
As per Attachment C of the LAR, VFDRs are characterized as two types. Type 1 VFDRs are forunprotected cables, where damage to these identified cables would cause failure to meet thedeterministic requirements of NFPA 805. Type 2 VFDRs are the dedicated shutdown recovery actions
used in the event the control room is abandoned due to a fire. The actions taken at a remote shutdownlocation that does not meet the definition of a primary control station are considered VFDRs.
Some Type 1 VFDRs identified in Table B-3, "Fire Area Transition," of Attachment C were dispositionedas "Not a VFDR." The VFDRs dispositioned as "Not a VFDR" are not included in the Fire PRA. TheseVFDRs were determined to be "Not a VFDR" within the Nuclear Capability Assessments (NSCA) with thefollowing justifications:
* There is a redundant component available to support the ability to achieve and maintain thenuclear safety performance criteria.
* Failure of the component can be recovered by action(s) taken in the MCR.
* Fire damage to the component will not prevent the component from performing its nuclear
safety function.
Excluding VFDRs dispositioned as "Not a VFDR" from the change-in-risk calculations were not based on
qualitative rationale.
The VFDRs dispositioned as "Not a VFDR" will be removed from LAR Attachment C. An updated LAR
Attachment C will be provided with response to PRA RAI 03.
Page 9 of 10
PRA RAI 24.01
In a letter dated March 16, 2015 (ADAMS Accession No. ML15079A025), the licensee responded to PRARAI 24 and in apparent contrast to the response to PRA RAI 24.e, the response to PRA RAI 01.f indicatesthat there are actions taken in the plant at the remote shutdown locations to recover equipment
affected by fire not associated with main control room (MCR) abandonment that are credited in the FirePRA. Attachment G and the response to PRA RAI 24.e seem to indicate that these actions aredesignated as defense-in-depth actions. Actions taken in the plant at the remote shutdown locations to
recover equipment affected by fire not associated with MCR abandonment may be recovery actions,
since command and control is not established at the remote shutdown panel.
Clarify whether these actions should be considered recovery actions as discussed in Regulatory Guide1.205, or discuss the rationale for their designation as defense-in-depth. If these actions should be
recovery actions, discuss how the additional risk of recovery actions for the applicable scenarios in
which the MCR is not abandoned will be modeled in response to PRA RAI 3.
Response:
The response to PRA RAI 24.e (Reference Robinson Letter RNP-RA/14-0122) has been updated to clarifythat RA-DIDs were categorized as such due to low risk significance. The process for determining thisoutcome is based on assessing the delta risk contribution from modeled actions in the FPRA.
For the Robinson NFPA 805 LAR, recovery actions are identified for operator actions performed in theplant at locations other than in the Main Control Room or at primary control stations, which are
identified in Attachment G. The primary control stations (PCSs) are the Dedicated Shutdown DieselControl Panel, Secondary Control Panel on the Turbine Deck, and the Charging Control Panel in the
Charging Pump Room. Therefore, operator actions identified in Attachment G that are credited in theFPRA and performed at locations other than the MCR or PCS are classified as "recovery actions". Itfollows that actions taken at the remote alternate control stations, for non-abandonment, are "recoveryactions", given that they are activities outside of the MCR to achieve the nuclear safety performance
criteria.
The change in risk (delta risk) for a recovery action is modeled by the difference between CDF/LERF forthe recovery actions based on the non-compliant (or variant) base case, which applies the nominal
human error probability (HEP), and the compliant case, which has the HEP set to zero (guaranteedalways successful).
The categorization of RA or RA-DID for recoveries modeled in the FPRA is based on the magnitude of the
change in risk of the recovery action.
The results of this analysis are provided in LAR Table W-5.
Page 10 of 10
Duke Energy Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
A. NEI 04-02 Table B-1 Transition of Fundamental Fire ProtectionProgram & Design Elements
79 Pages Attached
HBRSEP LAR Rev I Page A-I II
HBRSEP LAR Rev 1 Page A-1 I
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.1 General
Chapter 3 Requirement: 3.1* General.
This chapter contains the fundamental elements of the fire protection program and specifies the
minimum design requirements for fire protection systems and features. These fire protection program
elements and minimum design requirements shall not be subject to the performance-based methodspermitted elsewhere in this standard. Previously approved alternatives from the fundamental protectionprogram attributes of this chapter by the AHJ take precedence over the requirements contained herein.
Compliance Statement
N/A
Compliance BasisN/A - General statement; No technical
requirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.2 Fire Protection Plan
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A - General statement; No technical
requirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.2.1 Intent
Chapter 3 Requirement: 3.2.1 Intent.
A site-wide fire protection plan shall be established. This plan shall document management policy and
program direction and shall define the responsibilities of those individuals responsible for the plan's
implementation. This section establishes the criteria for an integrated combination of components,
procedures, and personnel to impleme t all fire protection program activities.
Complance statement
Complies
.ompliance 3as sNo Additional Clarification
Reference Document Doc Detwl
HBRSEP LAR Rev 1 Page A-2
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
OMM-002,Fire Protection Manual ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.2.2 Management Policy Direction and Responsibility.
Chapter 3 Requirement: 3.2.2* Management Policy Direction and Responsibility.
A policy document shall be prepared that defines management authority and responsibilities and
establishes the general policy for the site fire protection program.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document DoDetals
OMM-002,Fire Protection Manual Section 3
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.2.2.1 [Management Policy on Senior Management]
Chapter 3 Requirement: 3.2.2.1*
The policy document shall designate the senior management position with immediate authority and
responsibility for the fire protection program.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document DocDetails
OMM-002,Fire Protection Manual Section 3.1 & 3.2
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.2.2.2 [Management Policy on Daily Administration]
Chapter 3 Requirement:
Compliance Statement
3.2.2.2*
The policy document shall designate a position responsible for the daily administration and coordination
of the fire protection program and its implementation.
Compliance Basis
HBRSEP LAR Rev 1 Page A-3
Duke EnergyComplies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
No Additional Clarification
Reference Document Doc Details
OMM-002,Fire Protection Manual Section 3.7
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.2.2.3 [Management Policy on Interfaces]
Chapter 3 Requirement: 3.2.2.3*
The policy document shall define the fire protection interfaces with other organizations and assignresponsibilities for the coordination of activities. In addition, this policy document shall identify the
various plant positions having the authority for implementing the various areas of the fire protection
program.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document Doc Details
OMM-002,Fire Protection Manual Section 3
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.2.2.4 [Management Policy on AHJJ
Chapter 3 Requirement: 3.2.2.4*
The policy document shall identify the appropriate AHJ for the various areas of the fire protection
program.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document DocDetamls
OMM-002,Fire Protection Manual Section 4.1.7
l BlTable B-i1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.2.3 Procedures
HBRSEP LAR Rev 1 Page A-4
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Chapter 3 Requirement: 3.2.3* Procedures.
Procedures shall be established for implementation of the fire protection program. In addition to
procedures that could be required by other sections of the standard, the procedures to accomplish the
following shall be established:
(1) * Inspection, testing, and maintenance for fire protection systems and features credited by the fire
protection program.
Compliance Statement
Main Header: Complies
Section (1):License Amendment Required
Compliance BasisMain Header: No Additional Clarification
Section (1): See Attachment L for
Surveillance Optimization.
Reference Document Doc Details
OMM-002,Fire Protection Manual ALL
FP-012,Fire Protection Systems Minimum Equipment and ALL
Compensatory Actions
FP-013,Fire Protection Systems Surveillance Requirements ALL
Chapter 3 Requirement: 2) * Compensatory actions implemented when fire protection systems and other systems credited by
the fire protection program and this standard cannot perform their intended function and limits on
impairment duration.
Compliance Statement Compliance Basis
Section (2): Complies Section (2): No Additional Clarification
Reference Document Doc Dels
FP-012,Fire Protection Systems Minimum Equipment and ALL
Compensatory Actions
OMM-002,Fire Protection Manual Section 8.13.2
Chapter 3 Requirement: (3) * Reviews of fire protection program - related performance and trends.
Compliance Statement Compliance Basis
Section (3): Complies Section (3): No Additional Clarification
Reference Document DocDetaih
EGR-NGGC-0008,Engineering Projrams ALL
EGR-NGGC-0010,System & Compcnent Trending Program and Section 1.1 & Enclosure 1
System Notebooks
Chapter 3 Requirement: (4) Reviews of physical plant modifications and procedure changes for impact on the fire protection
HBRSEP LAR Rev 1 Page A-5
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
program.
Compliance Statement
Section (4): Complies
Compliance Basis
Section (4): No Additional Clarification
Reference Document
OMM-002,Fire Protection Manual
EGR-NGGC-0003,Design Review Requirements
EGR- NGGC-0005,Engineering Change
PRO-NGGC-0204,Procedure Review and Approval
EGR-NGGC-0102,Safe Shutdown/Fire Protection Review
REG-NGGC-0010,10 CFR 50.59 AND SELECTED REGULATORY
REVIEWS
Doc Dgtails
Section 3
ALL
ALL
ALL
ALL
ALL
Chapter 3 Requirement: (5) Long-term maintenance and configuration of the fire protection program.
Compliance Statement Compliance Basis
Section (5): Complies Section (5): No Additional Clarification
Reference Document DocDetaols
EGR-NGGC-0102,Safe Shutdown/Fire Protection Review ALL
EGR-NGGC-0003,Design Review Requirements ALL
EGR-NGGC-0005,Engineering Change ALL
Chapter 3 Requirement: (6) Emergency response procedures for the plant industrial fire brigade.
Compliance Statement Compliance Basis
Section (6): Complies Section (6): No Additional Clarification
Reference Document Doc Detall,
FP-001,Fire Emergency ALL
AOP-041,Response to Fire Event ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3 Prevention
Chapter 3 Requirement: 3.3 Prevention.
A fire prevention program with the goal of preventing a fire from starting shall be established,
documented, and implemented as part of the fire protection program. The two basic components of the
HBRSEP LAR Rev 1 Page A-6
Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design Elementsfire prevention program shall consist of both of the following:
(1) Prevention of fires and fire spread by controls on operational activities.
Compliance Statement
Main Header: Complies
Section (1): Complies
Compliance Basis
Main Header: No Additional Clarification
Section (1): No Additional Clarification
Reference Document DcDeils
OMM-002,Fire Protection Manual Section 8.4
FIR-NGGC-0003,Hot Work Permit ALL
Chapter 3 Requirement: (2) Design controls that restrict the use of combustible materials
The design control requirements listed in the remainder of this section shall be provided as described.
Compliance Statement Compliance Basis
Section (2): Complies Section (2): No Additional Clarification
Reference Document DoDetals
OMM-002,Fire Protection Manual Section 8.3
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND ALL
IGNITION SOURCE CONTROLS PROGRAM
EGR-NGGC-0005,Engineering Change Attachment 3, Section A3.0.7 & Attachment 2,
Section A3.3.24
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.1 Fire Prevention for Operational Activities.
Chapter 3 Requirement: 3.3.1 Fire Prevention for Operational Activities.
The fire prevention program activities shall consist of the necessary elements to address the control of
ignition sources and the use of transient combustible materials during all aspects of plant operations.
The fire prevention program shall focus on the human and programmatic elements necessary to prevent
fires from starting or, should a fire start, to keep the fire as small as possible.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND
IGNITION SOURCE CONTROLS PROGRAM
DocDetaoll
ALL
Section 8.4, ALLOMM-002,Fire Protection Manual
HBRSEP LAR Rev 1 Page A-7
Duke EnergyFIR-NGGC-0003,Hot Work Permit
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.1.1 General Fire Prevention Activities.
Chapter 3 Requirement: 3.3.1.1 General Fire Prevention Activities.
The fire prevention activities shall include but not be limited to the following program elements:(1) Training on fire safety information for all employees and contractors including, as a minimum,
familiarization with plant fire prevention procedures, fire reporting, and plant emergency alarms.
Compliance Statement Compliance Basis
Main Header: Complies Main Header: No Additional Clarification
Section (1): Complies Section (1): No Additional Clarification
Reference Document Doc Details
OMM-002,Fire Protection Manual Attachment 10.1 & Section 3.16
GNI0008N,Initial General Employee Training - Contractors Computer Based Training (CBT)
GETSSG,General Employee Training Self Study Guide ALL
GNBO1N,Initial General Employee Training - Progress Energy Computer Based Training (CBT)
Personnel
Chapter 3 Requirement: (2) * Documented plant inspections including provisions for corrective actions for conditions where
unanalyzed fire hazards are identified.
Compliance Statement Compliance Basis
Section (2): Complies Section (2): No Additional Clarification
Reference Document Doc Details
FP-010,Housekeeping Controls Section 8.2 Attachment 10.2
Chapter 3 Requirement: (3) * Administrative controls addressing the review of plant modifications and maintenance to ensure
that both fire hazards and the impact on plant fire protection systems and features are minimized.
Compliance Statement Compliance Basis
Section (3): Complies Section (3): No Additionrl Clarification
Reference Document
EGR-NGGC-0005,Engineering Change
Doc Details
Attachment 3, Section A3.0.7 & Attachment 2,
Section A3.3.24
HBRSEP LAR Rev 1 Page A-8
Duke EnergyOMM-002,Fire Protection Manual
WCP-NGGC-0300,Work Request Initiation, Screening, Prioritization,
and Classification
EGR-NGGC-0102,Safe Shutdown/Fire Protection Review
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsSection 3.8.2
Section 9.2.1.h
ALL
Table B-I NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.1.2 Control of Combustible Materials
Chapter 3 Requirement:
Compliance Statement
Main Header: Complies
Section (1): Complies
3.3.1.2* Control of Combustible Materials.
Procedures for the control of general housekeeping practices and the control of transient combustibles
shall be developed and implemented. These procedures shall include but not be limited to the following
program elements:
(1) * Wood used within the power block shall be listed pressure-impregnated or coated with a listed fire-
retardant application.Exception: Cribbing timbers 6 in. by 6 in. (15.2 cm by 15.2 cm) or larger shall not be required to be fire-
retardant treated.
Compliance Basis
Main Header: No Additional Clarification
Section (1): No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.1.8
IGNITION SOURCE CONTROLS PROGRAM
FP-010,Housekeeping Controls ALL
Chapter 3 Requirement: (2) Plastic sheeting materials used in the power block shall be fire-retardant types that have passed
NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, large-scale
tests, or equivalent.
Compliance Statement Compliance Basis
Section (2): Complies Section (2): Complies
Reference Document Doc Details
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.1.9
IGNITION SOURCE CONTROLS PROGRAM
Chapter 3 Requirement:
Compliance Statement
(3) Waste, debris, scrap, packing materials, or other combustibles shall be removed from an area
immediately following the completion of work or at the end of the shift, whichever comes first.
Compliance Basis
HBRSEP LAR Rev 1 Page A-9
Duke EnergySection (3): Complies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Section (3): No Additional Clarification
Reference Document DoDals
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.1.2
IGNITION SOURCE CONTROLS PROGRAM
Chapter 3 Requirement: (4) * Combustible storage or staging areas shall be designated, and limits shall be established on the
types and quantities of stored materials.
Compliance Statement
Section (4): Complies
Compliance Basis
Section (4): No Additional Clarification
Reference Document Doc Details
FP-010,Housekeeping Controls Attachments 10.1 -10.3
Chapter 3 Requirement: (5) * Controls on use and storage of flammable and combustible liquids shall be in accordance with
NFPA 30, Flammable and Combustible Liquids Code, or other applicable NFPA standards.
Compliance Statement Compliance Basis
Section (5): Complies Section (5): No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.4
IGNITION SOURCE CONTROLS PROGRAM
FAQ 06-0020,Identification of "applicable NFPA standards" ALL
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1 B-9
Chapter 3 Requirement: (6) * Controls on use and storage of flammable gases shall be in accordance with applicable NFPA
standards.
Compliance Statement
Section (6): Complies
Compi2:ance Basis
Section (6): No Additional Clarification
Reference Document DoDetals
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND ALL
IGNITION SOURCE CONTROLS PROGRAM
FP-006,Handling of Flammable Liquids and Gases ALL
FAQ 06-0020,ldentification of "applicable NFPA standards" ALL
Table B-1 NFPA 805 Ch.3 Transition Details
HBRSEP LAR Rev 1 Page A-1 0
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Chapter 3 Reference: 3.3.1.3 Control of Ignition Sources
Chapter 3 Requirement: 3.3.1.3 Control of Ignition Sources
Compliance Statement Compliance Basis
N/A N/A - General statement; No technical
requirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.1.3.1 [Control of Ignition Sources Code Requirements]
Chapter 3 Requirement: 3.3.1.3.1*
A hot work safety procedure shall be developed, implemented, and periodically updated as necessary
in accordance with NFPA 51B, Standard for Fire Prevention During Welding, Cutting, and Other Hot
Work, and NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations.
Compliance Statement Compliance Basis
Complies COMPLIES: No Additional Clarification
Complies with Clarification COMPLIES WITH CLARIFICATION:
Compliance with NFPA 241 is by
clarification and is addressed through
compliance with NFPA 51B. NFPA 241,
2009 edition, as referenced by NFPA 805-2001 ed., Section 5.1.1, with respect to hot
work, states "Responsibility for hot work
operations and fire prevention precautions
, including permits and fire watches, shall
be in accordance with NFPA 51B,
Standard for Fire Prevention During
Welding, Cutting, and Other Hot Work."
Reference Document DocDetals
FIR-NGGC-0003,Hot Work Permit ALL
NFPA 241 ,Standard for Safeguarding Construction, Alteration, and Section 5.1
Demolition Operations, 2004 Edition
NED-M/BMRK-0001 ,Code Compliance Evaluation for NFPA 51B, ALL
Standard for Fire Prevention during Welding, Cutting, and Other Hot
Work- 1999 Edition
Table B-1 NFPA 805 Ch.3 Transition Details
HBRSEP LAR Rev 1 Page A-1 1
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Chapter 3 Reference: 3.3.1.3.2 [Control of Ignition Sources on Smoking Limitations]
Chapter 3 Requirement: 3.3.1.3.2
Smoking and other possible sources of ignition shall be restricted to properly designated and
supervised safe areas of the plant.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document DocMlls
NO-80-169,Revision to the Administrative Controls for Fire Enclosure No. 4, Page 2
Protection, 2/1/1980
FP-010,Housekeeping Controls Section 5.3
FIR-NGGC-0003,Hot Work Permit ALL
Table B-I NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.1.3.3 [Control of Ignition Sources for Leak Testing]
Chapter 3 Requirement: 3.3.1.3.3
Open flames or combustion-generated smoke shall not be permitted for leak or air flow testing
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document DoDetail
FIR-NGGC-0003,Hot Work Permit Section 6.15
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.1.3.4 [Control of Ignition Sources on Portable Heaters]
Chapter 3 Requirement: 3.3.1.3.4*
Plant administrfative procedure shall control the use of portable electrical heaters ih the plant. Portable
fuel-fired heaters shall not be permitted in plant areas containing equipment important to nuclear safety
or where there is a potential for radiological releases resulting from a fire.
Compliance Statement
CompliesCompliance Basis
No Additional Clarification
HBRSEP LAR Rev 1 Page A- 12
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements
Reference Documentc Deails
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.1.11
IGNITION SOURCE CONTROLS PROGRAM
FIR-NGGC-0003,Hot Work Permit Section 6.13
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.2 Structural.
Chapter 3 Requirement: 3.3.2 Structural.
Walls, floors, and components required to maintain structural integrity shall be of noncombustible
construction, as defined in NFPA 220, Standard on Types of Building Construction.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1.B-6
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.3 Interior Finishes
Chapter 3 Requirement: 3.3.3 Interior Finishes.
Interior wall or ceiling finish classification shall be in accordance with NFPA 101®, Life Safety Code®,
requirements for Class A materials. Interior floor finishes shall be in accordance with NFPA 101
requirements for Class I interior floor finishes.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document
NFPA 101,Life Safety Code, 2009 Edition
CPL-)(XXX-W-005,Nuclear Powir Plant Protective Coatings
L2-C-007,Field Coatings
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
FPP-RNP-900,Fire Hazards Analysis
Doc Details
(a) Sections 10.2.3.4 & 10.2.7.4
ALL
ALL
HBRSEP LAR Rev 1 Page A- 13
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFirn Prnfr~rtinn Prngrnm P Q. 1"aain Ilementnihikp Fnir0Y
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference. 3.3.4 Insulation Materials
Chapter 3 Requirement: 3.3.4 Insulation Materials.
Thermal insulation materials, radiation shielding materials, ventilation duct materials, and
soundproofing materials shall be noncombustible or limited combustible.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1 .B-6
CPL-HBR2-M-025,Heating, Ventilation, and Air Conditioning (HVAC) Section 11-2.01
Main Plant Fabrication and Installation
CPL-HBR2-M-028,Specification for RHR Pump Pit to HVE-5 Exhaust Section 11-2.01
Tie-In Fabrication and Installation
L2-M-039,Piping and Equpment Thermal Insulation Section 4.4.1.4
GID/87038-0014,Fire Barrier System ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.5 Electrical.
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A - General statement; No technical
requirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.5.1 [Electrical Wiring Above Suspended Ceiling Limitations]
Chapter 3 Reauiremet: 3.3.5.1 1
Wiring above suspended ceiling shall be kept to a minimum. Where installed, electrical wiring shall be
listed for plenum use, routed in armored cable, routed in metallic conduit, or routed in cable trays with
solid metal top and bottom covers.
HBRSEP LAR Rev 1 Page A-14
Duke EnergyCompliance Statement
License Amendment Required
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsCompliance Basis
NRC approval is being requested in
Attachment L for electrical wiring above
suspended ceilings that may not comply
with the requirements of NFPA 805.
Reference Document
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
Doc Details
Appendix 9.5.1B-7
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.5.2 [Electrical Raceway Construction Limits]
Chapter 3 Requirement: 3.3.5.2
Only metal tray and metal conduits shall be used for electrical raceways. Thin wall metallic tubing shall
not be used for power, instrumentation, or control cables. Flexible metallic conduits shall only be used
in short lengths to connect components.
Compliance Statement Compliance Basis
License Amendment Required NRC approval is being requested in
Attachment L for electrical raceway
construction at HBRSEP that may not
comply with the requirements of NFPA
805.
Reference Document
HBR2-0B060 Sht D6,Electrical Installation Practices, Notes and
Details
HBR2-0B060 SH D2,Electrical Installation Practices, Notes and
Details
HBR2-0B060 SHC3,ELECTRICAL INSTALLATION PRACTICES,
NOTES AND DETAILS
DBD/R87038/SD62,Design Basis Document Cable and Raceway
System
DALDetals
ALL
ALL
ALL
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.5.3 [ElectricalCable Flame Propagation Limits]
Chapter 3 Requirement: 3.3.5.3*
Electric cable construction shall comply with a flame propagation test as acceptable to the AHJ.
HBRSEP LAR Rev 1 Page A-1 5
Duke Energy
Compliance Statement
Complies with Clarification
Complies via Previous NRC Approval
Complies via Engineering Evaluation
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Compliance BasisCOMPLIES WITH CLARIFICATION: FAQ
06-0022 evaluates currently recognized
flame propagation tests to the IEEE 383-
1974 Standard, the US NRC minimum test
standard, and acceptance criteria for cable
flame propagation tests. Table 2 in
'Summary of Results' section of FAQ 06-
0022 provides a summary of the testing
methods that are more severe than IEEE
383-1974. Non-IEEE-383-1974 qualified
cables used at RNP are IEEE-383-1974
equivalent since they meet the cable
standards identified in Table 2 of FAQ 06-
0022 except for some original PVC
jacketed cabling. Depending on when the
PVC jacketed cables were installed they
might not have met the requirements of
IEEE-383-1974 or equivalent. These
original cables not meeting the
requirements of IEEE-383-1974 or
equivalent were coated with fire retardant
material which meets or exceeds the
original cable coating requirements to
prevent propagation of fire.
COMPLIES VIA PREVIOUS NRC
APPROVAL: In the SER dated 2/28/78,
the NRC stated the following:
"4.8 Electrical Cables
In the plant areas outside the containment,
cable jacket and insulation material is
polyvinyl chloride. Inside the containment,
cable insulation is silicone rubber. The
flame test standard for cables IEEE 383
was not in effect at the time electrical
cables were purchased and installed at H.
B. Robinson. Cables in critical areas,
inside and outside containment will be
coated with a flame retardant coating.
Detailed discussion of these areas can befound in Section 5.0 of this report."
Section 5.0 listed the following areas
where cables would L`e coated:
Safety Injection Pump Room (Fire Area
No. 3)
Component Cooling Water Pump Room
(Fire Area No. 5)
Aux Feedwater Pump room (Fire Area No.
HBRSEP LAR Rev 1 Page A-16
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements
7)Cable Vaults (Fire Areas No. 9 and No. 34)
Aux Building Hallway, Lower Level (Fire
Area 10A, 10B, 10C)
Aux Building Hallway, Upper Level (Fire
Areas 14A thru 14G, except 14D)
Unit 2 Cable Spreading Room, Computer
Room (Fire Area No. 18)
Electrical Equipment Area (Fire Area No.
19)
Rod Control Room (Fire Area No. 20)
In the evaluation dated 2/21/80, the NRC
stated:
"3.1.4 Fire Retardant Cable Coating
Fire retardant coating will be applied to
cables located in 13 different fire areas of
the plant (4.8)."
By letter dated December 5, 1978, the
licensee stated that the flame-retardant
coating would be applied in accordance
with manufacturer's recommendations and
that the manufacturer would be consulted
to determine alternate application methods
for situations not covered by the
manufacturer's standard
recommendations.
We accept the licensee's proposal."
Per the 2/21/80 evaluation, the status of
Fire Retardant Coating was "Complete".
Proposed modifications were evaluated
and implemented per the SERs, where
applicable, to fulfill the intent of this
requirements.
There have been no plant modifications or
other changes that would invalidate the
basis for approval.
COMPLIES VIA ENGINEERING
EVALUATION: Engineering evaluations
EE-84-0043, EE-90-0037, EE-92-0090,
and NED-I3/BOP-1001 are applicable to
electric cable construction at HBRSEP.
Reference Document Doc Details
HBRSEP LAR Rev 1 Page A-17
Duke EneNLU-78-7A,License Amendment 31
NLU-80-106,RFI and Requirements to Resolve Issues Concerning
Fire Protection
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
FAQ 06-0022,Acceptable Electrical Cable Construction Tests
DBD/R87038/SD62,Design Basis Document Cable and Raceway
System
EE-84-0043,Qualification of Rockbestos Fire Zone R Cable to IEEE-
383 Vertical Flame Test
EE-90-0037,Evaluation Of Use Of Non-IEEE 383, Vertical Flame
Test Cable Proposed By Modification M-1001
EE-92-0090,Evaluation Of Abandoned Cables (Belden # 8424) Inside
Containment (General Area)
NED-B/BOP-1001,Comparative Analysis of Fire Propagation
Characteristics of UL-910 and IEEE-383 Qualified Cables
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsSection 4.8 & 5.0
Section 3.1.4
Section 9.5.1.4.4.4.2
ALL
Sections 3.5.1.3.3 and 3.5.1.3.5
ALL
ALL
ALL
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.6 Roofs.
Chapter 3 Requirement: 3.3.6 Roofs.Metal roof deck construction shall be designed and installed so the roofing system will not sustain a
self-propagating fire on the underside of the deck when the deck is heated by a fire inside the building.Roof coverings shall be Class A as determined by tests described in NFPA 256, Standard Methods ofFire Tests of Roof Coverings.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
Doc Details
Appendix 9.5.1B-6
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.7 Bulk Flammable Gas Storage.
Chapter 3 Requirement:
Compliance Statement
3.3.7 Bulk Flammable Gas Storage.
Bulk compressed or cryogenic flammable gas storage shall not be permitted inside structures housing
systems, equipment, or components important to nuclear safety.
Compliance Basis
HBRSEP LAR Rev 1 Page A-1 8
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Complies No Additional Clarification
Reference Document Doc Details
FP-010,Housekeeping Controls Section 8.1.7
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-8
EGR- NGGC-0005,Engineering Change ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.7.1 [Bulk Flammable Gas Location Requirements]
Chapter 3 Requirement: 3.3.7.1
Storage of flammable gas shall be located outdoors, or in separate detached buildings, so that a fire orexplosion will not adversely impact systems, equipment, or components important to nuclear safety.NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, shall be followed forhydrogen storage.
Compliance Statement Compliance Basis
Complies via Engineering Evaluation HBRSEP complies with NFPA 50A asevaluated in RNP-M/BMRK-1015.
Reference Document Doc Details
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-8
RNP-M/BMRK-1015,Code Compliance Evaluation for NFPA 50A, ALLStandard for Gaseous Hydrogen Systems at Consumer Sites - 1999Edition
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.7.2 [Bulk Flammable Gas Container Restrictions]
Chapter 3 Requirement: 3.3.7.2Outdoor high-pressure flammable gas storage containers shall be located so that the long axis is notpointed at buildings.
Compliance Statement
Complies
Compliance BasisNo Additional Clarification
Reference Document
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
Doc Details
Appendix 9.5.1B-8
HBRSEP LAR Rev 1 Page A-1 9
Duke EnergySAF-SUBS-00023,Compressed Gases
MCP-NGGC-0402,Material Management (Storage, Issue, and
Maintenance)
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Section 5.J.8
Section 9.1.6
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.7.3 [Bulk Flammable Gas Cylinder Limitations]
Chapter 3 Requirement: 3.3.7.3
Flammable gas storage cylinders not required for normal operation shall be isolated from the system.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
SAF-NGGC-2172,1ndustrial Safety Section 9.14
SAF-SUBS-00023,Compressed Gases Section 5.g.5
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.8 Bulk Storage of Flammable and Combustible Liquids.
Chapter 3 Requirement: 3.3.8 Bulk Storage of Flammable and Combustible Liquids.
Bulk storage of flammable and combustible liquids shall not be permitted inside structures containing
systems, equipment, or components important to nuclear safety. As a minimum, storage and use shall
comply with NFPA 30, Flammable and Combustible Liquids Code.
Compliance Statement
Complies via Engineering Evaluation
Compliance Basis
HBRSEP complies with NFPA 30 asevaluated in RNP-M/BMRK-1002.
Reference Document Doc Details
RNP-M/BMRK-1002,Code Compliance Evaluation NFPA 30 - Unit 2 ALLDiesel Fuel Oil Storage Tanks
FP-010,Housekeeping Controls Section 8.1.7, Attachment 10.1
UFSAR,HBR 2 Updated Fifal Safety Analysis Report (FSAR) Appendix 9.5.1B-9
FP-006,Handling of Flammable Liquids and Gases ALL
Table B-1 NFPA 805 Ch.3 Transition Details
HBRSEP LAR Rev I Page A-20
Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design Elements
Chapter 3 Reference: 3.3.9 Transformers.
Chapter 3 Requirement: 3.3.9* Transformers.
Where provided, transformer oil collection basins and drain paths shall be periodically inspected to
ensure that they are free of debris and capable of performing their design function.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document DoDetals
OST-642,Main Transformer Deluge System Flow Test (Refueling ALL
Interval)
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.10 Hot Pipes and Surfaces.
Chapter 3 Requirement: 3.3.10* Hot Pipes and Surfaces.
Combustible liquids, including high flashpoint lubricating oils, shall be kept from coming in contact with
hot pipes and surfaces, including insulated pipes and surfaces. Administrative controls shall require the
prompt cleanup of oil on insulation.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document Doc Details
FP-010,Housekeeping Controls Section 8.1
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.4.1.1
IGNITION SOURCE CONTROLS PROGRAM
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.11 Electrical Equipment
Chapter 3 Requirement: 313.111 Electrical Equipment
Adequate clearance, free of combustible material, shall be maintained around energized electrical
equipment.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
HBRSEP LAR Rev 1 Page A-21
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Reference Document Doc Details
FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.4.12IGNITION SOURCE CONTROLS PROGRAM
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.3.12 Reactor Coolant Pumps.
Chapter 3 Requirement: 3.3.12* Reactor Coolant Pumps.For facilities with non-inerted containments, reactor coolant pumps with an external lubrication systemshall be provided with an oil collection system. The oil collection system shall be designed and installedsuch that leakage from the oil system is safely contained for off normal conditions such as accidentconditions or earthquakes. All of the following shall apply.(1) The oil collection system for each reactor coolant pump shall be capable of collecting lubricating oilfrom all potential pressurized and nonpressurized leakage sites in each reactor coolant pump oil
system.
Compliance Statement Compliance Basis
Complies via Previous NRC Approval No oil collection system is provided for thereactor coolant pumps at HBRSEP.
Section (1): Complies via Previous NRCApproval By letter NLS-85-176 (3/7/1985), in
response to HBRSEP request for
exemption for requiring reactor coolantpump oil collection systems, the NRCstated the following:
"The containment contains three reactorcoolant pumps (A, B and C). These arelocated in bays (A, B and C). These baysalso contain safety related cabling for thereactor coolant loop instrumentation. BaysA and B share a common ceiling; Bay C isisolated from Bays A & B to some extent.The bays are covered by removable
concrete blocks. These blocks will causethe plume from an unmitigated fire to bediverted through the steam generator area.This area contains safety related steam
flow instrumentation sensing lines.
Oil spilled in Bay A, will be confined to BayA; however, oil spilled in Bays B and C canflow to adjacent areas. The foundation forthe reactor coolant pumps is at the237.000' level. The foundation for the
HBRSEP LAR Rev I Page A-22
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements
steam generators is at the 238.33' level.
The reactor coolant pump is located
between the pressurized portion of the oil
system and the steam generator supports,
and serves to shield the steam generator
supports in the event of an oil system
rupture.
The major combustible in each bay is the200 gallons of oil in each reactor coolant
pump.
The existing fire detection system in each
reactor coolant pump bay is a two-zone
detection system. One zone consists of asingle infrared flame detector; the other
zone consists of a 325°F fixed-
temperature heat detector. Activation of
one zone of detection sends an alarm to
the control room; activation of the second
zone of detection alarms in the controlroom and also opens the preaction water
deluge valve to the bay. Both detectors
are wall mounted.
The existing fire suppression system for
each bay, is a preaction sprinkler system.
Each bay has its own deluge valve, supply
header, and a ring
header that encircles the reactor coolant
pumps at elevation 239 feet 4 inches.
Each of the five risers off the ring header
have three 220°F closed head side wall
sprinklers at approximately 240 feet, 245
feet and 252 feet. elevations. These
systems are designed to meet theminimum residual pressure and flow
requirements of NFPA-Std-15.
The suppression system ring header
piping in Bay A is designed to withstand an
SSE, while Bays B and C are designed
such that a seismic event would not impact
safety related equipment due to
suppression system rupture The risersare restrained to withstand tt e nozzle
reaction forces. These forces are greater
than those anticipated from a seismic
event.
The existing containment spray system
HBRSEP LAR Rev 1 Page A-23
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elementswould be used as an emergency back-up
to the bay suppression system if
necessary to cool the operating level and
containment annulus outside of the RCP
bays.
By letter dated June 7, 1983, the licensee
proposed to:
(1) Provide additional ceiling mounted heat
detectors to meet the spacing and location
requirements of NFPA-STD-72E,
"Standard on Automatic Fire Detectors.
(2) Replace existing closed head
sprinklers with special open water spray
nozzles and manual actuation from the
control room.
(3) Construct 6 inch dikes at the 231 feet
elevation in Bay B and Bay C.(4) Revise operating procedures for the
containment spray system to allow its
operation as a back up fire suppression
system with the sodium hydroxide valves
out.
By letter dated October 5, 1983, the
licensee committed to maintain an
automatically actuated closed-head
preaction system in lieu of a manually
actuated open-head system.
We have evaluated the fire protection for
the reactor coolant pump lube oil system
and conclude that the effects of a fire in an
RCP Bay will not prevent safe shutdown
capability. There are no components within
the RCP Bay that are required for safe
shutdown. The effects of any fire within an
RCP Bay will be prevented from affecting
the safe shutdown equipment outside the
RCP Bay by the suppression system
inside the RCP Bay and the ContainmentSpray System outside the Bay.
It is the staff's conclusion that: 1)installation of a re ctor coolant pump oil
collection system ih this facility would not
significantly enhance fire safety, and 2) the
existing fire protection system in the
Reactor Coolant Pump Bays with the
addition of the proposed modifications
HBRSEP LAR Rev I Page A-24
Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design Elementsprovides an acceptable level of safety to
that achieved by compliance with the
requirements of Section 111.0 of Appendix
R to 10 CFR 50. Therefore, the licensee's
request for an exemption should be
granted."
Proposed modifications were evaluated
and implemented per the SERs, where
applicable, to fulfill the intent of this
requirements.
There have been no plant modifications or
other changes that would invalidate the
basis for approval.
Section (1): No oil collection system is
provided for the reactor coolant pumps at
RNP. See Section 3.3.12 above for
discussion of acceptability for lack of oil
collection system.
Reference Document Doc Details
NLS-85-176,RCP Oil Collection Exemption ALL
Chapter 3 Requirement: (2) Leakage shall be collected and drained to a vented closed container that can hold the inventory ofthe reactor coolant pump lubricating oil system.
Compliance Statement Compliance Basis
Section (2): Complies via Previous NRC Section (2): No oil collection system isApproval provided for the reactor coolant pumps at
HBRSEP. See Section 3.3.12 above fordiscussion of acceptability for lack of oil
collection system.
Chapter 3 Requirement: (3) A flame arrestor is required in the vent if the flash point characteristics of the oil present the hazardof a fire flashback.
Compliance Statement Compliance Basis
Section (3): Complies via Previous NRC Section (3): No oil collection system isApproval provided for the reactor coolant pumps at
HBRSEr. See Section 3.3.12 above fordiscuss on of acceptability for lack of oilcollection system.
HBRSEP LAR Rev 1 Page A-25
Duke EnergyC-hapter 3Requirement:
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
(4) Leakage points on a reactor coolant pump motor to be protected shall include but not be limited to
the lift pump and piping, overflow lines, oil cooler, oil fill and drain lines and plugs, flanged connections
on oil lines, and the oil reservoirs, where such features exist on the reactor coolant pumps.
Compliance Statement
Section (4): Complies via Previous NRCApproval
Compliance Basis
Section (4): No oil collection system is
provided for the reactor coolant pumps at
HBRSEP. See Section 3.3.12 above for
discussion of acceptability for lack of oil
collection system.
Chapter 3 Requirement: (5) The collection basin drain line to the collection tank shall be large enough to accommodate the
largest potential oil leak such that oil leakage does not overflow the basin.
Compliance Statement Compliance Basis
Section (5): Complies via Previous NRC Section (5): No oil collection system is
Approval provided for the reactor coolant pumps atHBRSEP. See Section 3.3.12 above for
discussion of acceptability for lack of oil
collection system.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4 Industrial Fire Brigade.
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A - General statement; No technical
requirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.1 On-Site Fire-Fighting Capability.
Chapter 3 Requirement: 3.4.1 On-Site Fire-Fighting Capability.
All of the followi g requirements shall apply.
(a) A fully staffed, trained, and equipped fire-fighting force shall be available at all times to control and
extinguish all fires on site. This force shall have a minimum complement of five persons on duty and
shall conform with the following NFPA standards as applicable:(1) NFPA 600, Standard on Industrial Fire Brigades (interior structural fire fighting)
HBRSEP LAR Rev 1 Page A-26
Duke EnergyCompliance Statement
Section (a): Complies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Section (a): No Additional Clarification
Section (a) (1): Complies via Engineering
Evaluation
Section (a) (1): HBRSEP complies with
NFPA 600 as evaluated in NED-M/BMRK-
0002.
Reference Document Doc Details
OMM-002,Fire Protection Manual Section 8.6
NED-M/BMRK-0002,CODE COMPLIANCE EVALUATION FOR ALLNFPA 600, STANDARD ON INDUSTRIAL FIRE BRIGADES, 2000EDITION
Chapter 3 Requirement: (2) NFPA 1500, Standard on Fire Department Occupational Safety and Health Program
Compliance Statement Compliance BasisSection (a) (2): N/A Section (a) (2): NFPA 1500 is not
applicable to HBRSEP as the site utilizes afire brigade, not an organized fire
department. Fire Brigade requirements arereviewed using NFPA 600.
Chapter 3 Requirement: (3) NFPA 1582, Standard on Medical Requirements for Fire Fighters and Information for FireDepartment Physicians.
Compliance Statement Compliance Basis
Section (a) (3): N/A Section (a) (3): NFPA 1582 is notapplicable to HBRSEP as the site utilizes afire brigade, not an organized fire
department. Fire Brigade requirements arereviewed using NFPA 600.
Chapter 3 Requirement: (b) * Industrial fire brigade members shall have no other assigned normal plant duties that wouldprevent immediate response to a fire or other emergency as required.
Compliance Statement Compliance Basis
Section (b): Complies Section (b): No Additional Clarification
Reference Document Doc Details
OMM-002,Fire Protection Manual Section 3.12 & 8.6
Chapter 3 Requirement: (c) During every shift, the brigade leader and at least two brigade members shall have sufficient training
and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on
HBRSEP LAR Rev 1 Page A-27
Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design Elementsnuclear safety performance
Exception: Sufficient training and knowledge shall be permitted to be provided by an operations advisor
dedicated to industrial fire brigade support criteria.
Compliance Statement
Section (c): Complies
Compliance Basis
Section (c): No Additional Clarification
Reference Document Doc Details
OMM-002,Fire Protection Manual Section 8.6.1
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL
Chapter 3 Requirement: (d) * The industrial fire brigade shall be notified immediately upon verification of a fire.
Compliance Statement Compliance Basus
Section (d): Complies Section (d): No Additional Clarification
Reference Document Doc Details
AOP-041,Response to Fire Event Section 2 Step 4
Chapter 3 Requirement: (e) Each industrial fire brigade member shall pass an annual physical examination to determine that he
or she can perform the strenuous activity required during manual fire-fighting operations. The physical
examination shall determine the ability of each member to use respiratory protection equipment.
Compliance Statement Compliance Basis
Section (e): Complies Section (e): No Additional Clarification
Reference Documentc D±amls
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.4.1
OMM-002,Fire Protection Manual Section 8.6.3
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.2 Pre-Fire Plans.
Chapter 3 Requirement: 3.4.2* Pre-Fire Plans.
Current and detailed pre-fire plans shall be available to the in qustrial fire brigade for all areas in which
a fire could jeopardize the ability to meet the performance criteria described in Section 1.5.
Compliance Statement Compliance Basis
Complies No Additional Clarification
HBRSEP LAR Rev 1 Page A-28
Duke EnergyReference Document
FIR-NGGC-0008,NFPA 805 Pre-Fire Plans
OMM-002,Fire Protection Manual
HBR2-11937,Fire Pre-Plan Drawings, Sh 1-60
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Doc Detanis
Sections 9.2 and 9.5
Section 8.8
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.2.1 [Pre-Fire Plan Contents]
Chapter 3 Requirement: 3.4.2.1*The plans shall detail the fire area configuration and fire hazards to be encountered in the fire area,
along with any nuclear safety components and fire protection systems and features that are present.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document
FIR-NGGC-0008,NFPA 805 Pre-Fire Plans
Doc Details
Section 9.2
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.2.2 [Pre-Fire Plan Updates]
Chapter 3 Requirement: 3.4.2.2 Pre-fire plans shall be reviewed and updated as necessary.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Section 4.0
AP-043,RNP Procedure Biennial Review Process Section 8.1
EGR-NGGC-0005,Engineering Change Section 4.5
RNP/94-1890,PROPOSED CHANGE TO QUALITY ASSURANCE ALL
PROGRAM
Table B-1 NFPA 805 Ch.3 Transition Details
HBRSEP LAR Rev 1 Page A-29
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDukeater 3Referencen 3.4.2.3 [Pre-Fire Plan Locations]
Chapter 3 Requirement: 3.4.2.3*
Pre-fire plans shall be available in the control room and made available to the plant industrial fire
brigade.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Section 9.5
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.2.4 [Pre-Fire Plan Coordination Needs]
Chapter 3 Requirement: 3.4.2.4*
Pre-fire plans shall address coordination with other plant groups during fire emergencies.
Compliance Statement Compliance Basis
Complies with Clarification HBRSEP has procedure FP-001, "Fire
Emergency" which is not specifically a fire
pre-plan, however FP-001 provides
specific instructions for actions required
from key groups at HBRSEP supporting
the fire brigade/fire emergency actions.
There are detailed response coordination
actions specified for Control Room, RC,
and the Security group. Any other
coordination actions would be initiated by
the Control Room personnel as needed for
any plant emergency.
Reference Document Doc Details
FP-001 ,Fire Emergency ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.3 Training and Drills.
Chapter 3 Requirement: 3.4.3 Training and Drills.
Industrial fire brigade members and other plant personnel who would respond to a fire in conjunction
HBRSEP LAR Rev 1 Page A-30
Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design Elementswith the brigade shall be provided with training commensurate with their emergency responsibilities.
(a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.
(1) Plant industrial fire brigade members shall receive training consistent with the requirements
contained in NFPA 600, Standard on Industrial Fire Brigades, or NFPA 1500, Standard on Fire
Department Occupational Safety and Health Program, as appropriate.
Compliance Statement
Main Header: Complies
Section (a) (1): Complies via EngineeringEvaluation
Compliance Basis
Main Header: No Additional Clarification
Section (a) (1): HBRSEP complies with
NFPA 600 as evaluated in the applicable
portions of NED-M/BMRK-0002.
Reference Document DDetails
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL
NED-M/BMRK-0002,CODE COMPLIANCE EVALUATION FOR ALLNFPA 600, STANDARD ON INDUSTRIAL FIRE BRIGADES, 2000EDITION
Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(2) Industrial firebrigade members shall be given quarterly training and practice in fire fighting, including radioactivityand health physics considerations, to ensure that each member is thoroughly familiar with the steps to
be taken in the event of a fire.
Compliance Statement Compliance Basis
Section (a) (2): Complies Section (a) (2): No Additional Clarification
Reference Document DoDetals
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL
GNR01 N,Plant Access Annual Requalification, CBT ALL
Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(3) A writtenprogram shall detail the industrial fire brigade training program.
Compliance Statement Compliance BasisSection (a) (3): Complies Section (a) (3): No Additional Clarification
Reference Document Doc Details
FIR-Nt 3GC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL
Chapter 3 Requirements (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(4) Written
records that include but are not limited to initial industrial fire brigade classroom and hands-on training,
refresher training, special training schools attended, drill attendance records, and leadership training for
industrial fire brigades shall be maintained for each industrial fire brigade member.
HBRSEP LAR Rev 1 Page A-31
Duke Energy
Compliance Statement
Section (a) (4): Complies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Compliance Basis
Section (a) (4): No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL
TAP-404,Training Documentation and Records ALL
Chapter 3 Requirement: (b) Training for Non-Industrial Fire Brigade Personnel. Plant personnel who respond with the industrial
fire brigade shall be trained as to their responsibilities, potential hazards to be encountered, and
interfacing with the industrial fire brigade.
Compliance Statement
Section (b): Complies with Clarification
Compliance Basis
Section (b): Guidance for non-industrial
fire brigade members is found in FP-001.
The procedure defines the actions needed
to be taken by personnel discovering a fire,
security personnel actions, and duty health
physics contact actions.
Reference Document DoDetamls
FP-001 ,Fire Emergency ALL
Chapter 3 Requirement: (c) * Drills. All of the following requirements shall apply.
(1) Drills shall be conducted quarterly for each shift to test the response capability of the industrial fire
brigade.
Compliance Statement Compliance Basis
Section (c) (1): Complies Section (c) (1): No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.10.3.3.a
Chapter 3 Requirement: (c) * Drills. All of the following requirements shall apply.(2) Industrial fire brigade drills shall be
developed to test and challenge industrial fire brigade response, including brigade performance as a
team, proper use of equipment, effective use of pre-fire plans, and coordination with other groups.
These drills shall evaluate the industrial fire brigade's abilities to react, respond, and demonstrate
proper fire-fighting techniques to control and extinguish the fire and smoke conditions being simulated
by the drill scenario.
Compliance Statement Compliance Basis
Section (c) (2): Complies Section (c) (2): No Additional Clarification
HBRSEP LAR Rev 1 Page A-32
Duke EnergyReference Document
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Doc DetailsSection 9.10
Chapter 3 Requirement: (c) * Drills. All of the following requirements shall apply.(3) Industrial fire brigade drills shall be
conducted in various plant areas, especially in those areas identified to be essential to plant operationand to contain significant fire hazards.
Compliance Statement
Section (c) (3): Complies
Compliance Basis
Section (c) (3): No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.10.2
Chapter 3 Requirement: (c) * Drills. All of the following requirements shall apply.(4) Drill records shall be maintained detailing thedrill scenario, industrial fire brigade member response, and ability of the industrial fire brigade to
perform as a team.
Compliance Statement
Section (c) (4): Complies
Compliance Basis
Section (c) (4): No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL
TAP-404,Training Documentation and Records ALL
Chapter 3 Requirement: (c) * Drills. All of the following requirements shall apply.(5) A critique shall be held and documented
after each drill.
Compliance Statement Compliance Basis
Section (c) (5): Complies Section (c) (5): No Additional Clarification
Reference Document Doc Details
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.10.7
TAP-404,Training Documentation and Records ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 References 3.4.4 Fire-Fighting Equipment.
Chapter 3 Requirement: 3.4.4 Fire-Fighting Equipment.Protective clothing, respiratory protective equipment, radiation monitoring equipment, personal
HBRSEP LAR Rev 1 Page A-33
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements
dosimeters, and fire suppression equipment such as hoses, nozzles, fire extinguishers, and other
needed equipment shall be provided for the industrial fire brigade. This equipment shall conform with
the applicable NFPA standards.
Compliance Statement Compliance Basis
Complies with Clarification Per FAQ 06-0020, the following guidance
applies as to which NFPA standards
referenced in Chapter 3 are applicable:
"Where used in NFPA 805, Chapter 3, the
term, "applicable NFPA Standards" is
considered to be equivalent to those NFPA
standards identified in the current license
basis (CLB) for procedures and systems in
the Fire Protection Program that are
transitioning to NFPA 805."
Firefighting equipment is provided. A
monthly inspection/inventory "of Fire
Protection equipment and supplies located
in the Fire Equipment Staging areas to
meet the demands of the site Fire
Brigade..." is conducted per OST-639.
Personnel dosimeters are issued in
accordance with the plant radiationprotection program and DOS-NGGC-
0002. HP personnel, who provide fire
brigade support, provide radiation
monitoring equipment in accordance with
FP-001.
HBRSEP makes use of Duke Energy fleet
procedure AD-EG-ALL-1531, Selection,
Care and Maintenance of Fire Fighting
Ensembles which follows the guidance
found in NFPA standards associated with
firefighting Personal Protective Equipment.
Standards and requirements associated
with installed fire protection equipment
such as hoses, nozzels, fire extinguishers
and other equipment are maintained in
accordance with NFPA standards as
described elsewhere in Chapter 3 and
evaluated for compliance under the
various NFPA Code Compliance
Calculations listed for those sections of
Chapter 3.
Reference Document Doe Details
HBRSEP LAR Rev 1 Page A-34
Duke EnergyFAQ 06-0020,Identification of "applicable NFPA standards"
OST-639,Fire Equipment Inventory (Monthly)
FP-001,Fire Emergency
OMM-002,Fire Protection Manual
DOS-NGGC-0002,Dosimetry Issuance
AD-EG-ALL-1531 ,Selection, Care, and Maintenance of Fire FightingEnsembles
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
ALL
ALL
Section 3.7
Section 3.27
ALL
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.5 Off-Site Fire Department Interface.
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A - General statement; No technical
requirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.5.1 Mutual Aid Agreement.
Chapter 3 Requirement: 3.4.5.1 Mutual Aid Agreement.
Off-site fire authorities shall be offered a plan for their interface during fires and related emergencies on
site.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
PLP-007, Robinson Emergency Plan Attachment 6.2
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.5.2 Site-Specific Training.
Chapter 3 Requirement: 3.4.5.2* Site-Specific Training.
Fire fighters from the off-site fire authorities who are expected to respond to a fire at the plant shall be
HBRSEP LAR Rev 1 Page A-35
Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design Elementsoffered site-specific training and shall be invited to participate in a drill at least annually.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document
FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM
Doco Dta9is
Section 9.7.1
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.5.3 Security and Radiation Protection.
Chapter 3 Requirement: 3.4.5.3* Security and Radiation Protection.
Plant security and radiation protection plans shall address off-site fire authority response.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
OMM-002,Fire Protection Manual Section 8.10
PLP-007,Robinson Emergency Plan Table 5.3.2-1 Notes
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.4.6 Communications.
Chapter 3 Requirement: 3.4.6* Communications.An effective emergency communications capability shall be provided for the industrial fire brigade.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
PLP-007,Robinson Emergency Plan Attachment 6.1
OST-639,Fire Equip ent Inventory (Monthly) ALL
Table B-1 NFPA 805 Ch.3 Transition Details
HBRSEP LAIR Rev 1 Page A-36
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Chapter 3 Reference: 3.5 Water Supply
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A - General statement; No technical
requirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.1 [Water Supply Flow Code Requirements]
Chapter 3 Requirement: 3.5.1
A fire protection water supply of adequate reliability, quantity, and duration shall be provided by one of
the two following methods.
(a) Provide a fire protection water supply of not less than two separate 300,000-gal (1,135,500-L)
supplies.
(b) Calculate the fire flow rate for 2 hours. This fire flow rate shall be based on 500 gpm (1892.5 L/min)
for manual hose streams plus the largest design demand of any sprinkler or fixed water spray
system(s) in the power block as determined in accordance with NFPA 13, Standard for the Installation
of Sprinkler Systems, or NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection. The fire
water supply shall be capable of delivering this design demand with the hydraulically least demanding
portion of fire main loop out of service.
Compliance Statement
Complies via Engineering Evaluation
Compliance Basis
HBRSEP complies with Section (b) ofNFPA 805 Ch. 3 Section 3.5.1 as detailedin RNP-M/BMRK-1011, RNP-M/MECH-1727, and RNP-M/MECH-1728.
Reference Document
RNP-M/BMRK-1 01 1,Code Compliance Evaluation for NFPA 15,Water Spray Fixed Systems
RNP-M/MECH-1 727,Hydraulic Analysis of the Hydrogen Seal OilWater Spray System
RNP-M/MECH-1 728,Hydraulic Analysis of the Auxiliary & Start-UpTransformer Water Spray System
NLU -78-71, License Amendment 31
Code Section 3012
Section 5
Section 5
Section 4.3.1.1
Table B-1 NFPA 805 Ch.3 Transition Details
HBRSEP LAR Rev 1 Page A-37
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDukeapter 3Reference: 3.5.2 [Water Supply Tank Code Requirements]
Chapter 3 Requirement: 3.5.2*The tanks shall be interconnected such that fire pumps can take suction from either or both. A failure in
one tank or its piping shall not allow both tanks to drain: The tanks shall be designed in accordance with
NFPA 22, Standard for Water Tanks for Private Fire Protection.
Exception No. 1: Water storage tanks shall not be required when fire pumps are able to take suction
from a large body of water (such as a lake), provided each fire pump has its own suction and both
suctions and pumps are adequately separated.
Exception No. 2: Cooling tower basins shall be an acceptable water source for fire pumps when the
volume is sufficient for both purposes and water quality is consistent with the demands of the fire
service.
Compliance Statement
Complies with Clarification
Compliance BasisFire water is obtained directly from Lake
Robinson via a single intake structure.
Two physically separated automatic fire
pumps are provided with separate suction
lines.
As such, HBRSEP complies with
Exception No. 1
Reference Document DoDetals
NLU-78-71 ,License Amendment 31 Section 4.3.1.1
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.3 [Water Supply Pump Code Requirements]
Chapter 3 Requirement: 3.5.3*
Fire pumps, designed and installed in accordance with NFPA 20, Standard for the Installation of
Stationary Pumps for Fire Protection, shall be provided to ensure that 100 percent of the required flow
rate and pressure are available assuming failure of the largest pump or pump power source.
Compliance Statement Compliance Basis
Complies via Engineering Evaluation HBRSEP complies with NFPA 20 as
evaluated in RNP-M/BMRK-1012, RNP-
M/MECH-1725, an• RNP-M/MECH-1610.
Hydraulic analysis 4emonstrates the ability
of one pump to provide required flow rate
to the largest system.
Reference Document Doc Details
HBRSEP LAR Rev 1 Page A-38
Duke Energy
RNP-M/MECH-1725,Evaluation of NFPA 13 Code ComplianceVariances
RNP-M/BMRK-1012,Code Compliance Evaluation NFPA 20 -Centrifugal Fire Pumps
FP-012,Fire Protection Systems Minimum Equipment andCompensatory Actions
RNP-M/MECH-1610,Hydraulic Analysis - Main Transformers
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
ALL
Conclusions (Page 9) & Attachment 6-Code
Section 33-Page 4 of 62
Section 8.2
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.4 [Water Supply Pump Diversity and Redundancy]
Chapter 3 Requirement: 3.5.4
At least one diesel engine-driven fire pump or two more seismic Category I Class IE electric motor-
driven fire pumps connected to redundant Class IE emergency power buses capable of providing 100
percent of the required flow rate and pressure shall be provided.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification.
Reference Document Doc Details
RNP-M/BMRK-1012,Code Compliance Evaluation NFPA 20 - Body of CalculationCentrifugal Fire Pumps
FP-012,Fire Protection Systems Minimum Equipment and Section 8.2Compensatory Actions
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.5 [Water Supply Pump Separation Requirements]
Chapter 3 Requirement: 3.5.5Each pump and its driver and controls shall be separated from the remaining fire pumps and from therest of the plant by rated fire barriers.
Compliance Statement Compliance Basis
Complies via Previous NRC Approval In a submittal dated 6/23/77, HBRSEPprovided the following information:
"The Unit 2 fire protection water supplysystem meets the intent of the
requirements (BTP ABCSB 9.5-1) either
by itself or by virtue of cross-connection to
HBRSEP LAR Rev 1 Page A-39
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy the Unit 1 system, except that the Unit 2 Fire Protection Program & Design Elements
supply pumps are not separated from
each other or remaining pumps by three-
hour rated fire walls. However, the various
fire pumps are out-of-doors and separated
by distance as well as intervening
equipment."
In the SER dated 2/21/1980, the NRC
stated:
"The staff does not agree with the
licensee's contention that the arrangement
of the propane storage tank and other
equipment is satisfactory in relation to
safety-related equipment on the intake
structure for the following reasons....
Therefore, we will require the licensee to:
-Replace the propane engine with a diesel
engine, or
-Replace the propane engine-driven fire
pump and associated equipment to alocation substantially remote from any
safety-related equipment. "
The original propane-fueled engine driver
on one of the two 100% capacity fire
pumps was changed to a diesel-fueled
driver per Modification M-445P to address
concerns raised by the NRC over the
propane storage location and
arrangement.
Proposed modifications were evaluated
and implemented per the SERs, where
applicable, to fulfill the intent of this
requirements.
There have been no plant modifications or
other changes that would invalidate the
basis for approval.
Reference Document Doc Details
NLU-80-106,RFI and Requirements to Resolvp Issues Concerning Section 3.2.3
Fire Protection
NG-77-704,Fire Protection Program Review Question 15
M-445P,Fire Pump Engine Replacement & Propane Tank Relocation ALL
HBRSEP LAR Rev 1 Page A-40
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Table B-1-NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.6 [Water Supply Pump Start/Stop Requirements]
Chapter 3 Requirement: 3.5.6
Fire pumps shall be provided with automatic start and manual stop only.
Compliance Statement Complsance Basis
Complies via Engineering Evaluation Fire pumps are provided with automatic
start and manual stops, as detailed in the
applicable portions of the NFPA 20 code
compliance evaluation RNP-M/BMRK-
1012.
Reference Document nD taifi
RNP-M/BMRK-1012,Code Compliance Evaluation NFPA 20 - Code Sections 515 (Attachment 6), Code Section
Centrifugal Fire Pumps 9-5 (Attachment 7), and Code Section 9-5
(Attachment 8)
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.7 [Water Supply Pump Connection Requirements]
Chapter 3 Requirement: 3.5.7
Individual fire pump connections to the yard fire main loop shall be provided and separated with
sectionalizing valves between connections.
Compliance Statement Compliance Basis
Complies via Previous NRC Approval In the SER dated 2/28/78, the NRC stated:
"4.3.1.3 Fire Water Piping System
The two fire pumps have a common
discharge through a twelve inch
underground main into a ten-inch
underground fire water loop which
encircles the plant.
All yard fire hydrants, automatic water
s+ppression systems and interior fire hose
lines are supplied by the fire loop.
Sectionalizing valves of the post indicator
type are provided on the fire loop to permit
partial pipeline isolation without
interruption of service to the entire system.
HBRSEP LAR Rev 1 Page A-41
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy The licensee will install isolation valves at Fire Protection Program & Design Elements
the connection of the ten inch fire line from
the Unit I fire loop to the Unit 2 fire loop,
and provide separate headers for
automatic sprinkler systems to be installed
in the reactor auxiliary building. The
licensee will also provide barriers around
all hydrants and post indicator valves to
protect against vehicular damage.
Electrical supervision, to monitor the
position of fire water system controlvalves, is not provided. A means of sealing
these valves open will be provided, and
this in combination with administrative
controls and periodic inspections will be
used to assure that valves are maintained
open.
We find that, subject to the implementation
of the above described modifications, the
fire water piping systems satisfy the
objectives identified in Section 2.1 of this
report and are, therefore, acceptable."
As seen in drawings HBR2-08255-Sheets
1 and 2, isolation valves were installed at
the connection of the Unit 1 fire loop lines
to the Unit 2 fire loop and separateheaders for automatic sprinkler systems
were installed in the reactor auxiliary
building.
Proposed modifications were evaluated
and implemented per the SERs, where
applicable, to fulfill the intent of this
requirements.
There have been no plant modifications or
other changes that would invalidate the
basis for approval.
Reference Document Doc Details
NLU-78-71 ,License Amendment 31 4.3.1.3
HBR2-08255 Sh 2,Fire Protection System Flow Diagram ALL
HBR2-08255 Sh 1 ,Fire Protection System Intake Structure Flow ALL
Diagram
HBRSEP LAR Rev 1 Page A-42
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalI 11i1w I--l~rp1 I-|rw wrrTinn Irn•prn m & I np.inn R-I ments
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.8 [Water Supply Pressure Maintenance Limitations]
Chapter 3 Requirement: 3.5.8
A method of automatic pressure maintenance of the fire protection water system shall be provided
independent of the fire pumps.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.4.2.7
RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, Code Section 661
Standpipes and Hose Stations
HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow ALL
Diagram
APP-044,Fire Alarm Console (FAC) C55, C58
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.9 [Water Supply Pump Operation Notification]
Chapter 3 Requirement: 3.5.9
Means shall be provided to immediately notify the control room, or other suitable constantly attended
location, of operation of fire pumps.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document Doc Details
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.4.3.4.1.c
APP-044,Fire Alarm Console (FAC) C55, C58
Table B-1 NFPA 805 Ch.3 TLansition Details
Chapter 3 Reference: 3.5.10 [Water Supply Yard Main Code Requirements]
HBRSEP LAR Rev 1 Page A-43
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFirp Prnt.etionn,2P-mlam_&JDesiq•,Fpemrs01ukeErie=
Chapter 3 Requirement: 3.5.10
An underground yard fire main loop, designed and installed in accordance with NFPA 24, Standard for
the Installation of Private Fire Service Mains and Their Appurtenances, shall be installed to furnish
anticipated water requirements.
Compliance Statement
Complies via Engineering Evaluation
Compliance BasisHBRSEP complies with NFPA 24 as
evaluated in RNP-M/BMRK-1013.
Reference Document DoDetamis
RNP-M/BMRK-1013,Code Compliance Evaluation NFPA 24 - ALL
Standard for Outside Protection
RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code Section 4.1
Compliance Variances
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.11 [Water Supply Yard Main Maintenance Issues]
Chapter 3 Requirement: 3.5.11
Means shall be provided to isolate portions of the yard fire main loop for maintenance or repair without
simultaneously shutting off the supply to both fixed fire suppression systems and fire hose stations
provided for manual backup. Sprinkler systems and manual hose station standpipes shall be connected
to the plant fire protection water main so that a single active failure or a crack to the water supply piping
to these systems can be isolated so as not to impair both the primary and backup fire suppression
systems.
Comp•liance Statement Compliance Basis
Complies via Previous NRC Approval In the SER dated 2/28/78, the NRC stated:
"4.3.1.3 Fire Water Piping System
The two fire pumps have a common
discharge through a twelve inch
underground main into a ten-inch
underground fire water loop which
encircles the plant.
All yard fire hydrants, automatic water
suppression systems and interior hose
lines are supplied by the fire loop.
Sectionalizing valves of the post indicator
type are provided on the fire loop to permit
partial pipeline isolation without
interruption of service to the entire system.
The licensee will install isolation valves at
HBRSEP LAR Rev 1 Page A-44
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elementsthe connection of the ten inch fire line from
the Unit 1 fire loop to the Unit 2 fire loop,
and provide separate headers for
automatic sprinkler systems to be installed
in the reactor auxiliary building. The
licensee will also provide barriers around
all fire hydrants and post indicator valves
to protect against vehicular damage.
Electrical supervision, to monitor theposition of fire water system control
valves, is not provided. A means of sealing
these valves open will be provided, and
this in combination with administrative
controls and periodic inspections will be
used to assure that valves are maintained
open.
We find that, subject to the implementation
of the above described modifications, the
fire water piping systems satisfy the
objectives identified in Section 2.1 of this
report and are, therefore, acceptable."
Proposed modifications were evaluated
and implemented per the SERs, where
applicable, to fulfill the intent of this
requirements.
There have been no plant modifications or
other changes that would invalidate the
basis for approval.
Reference Document Doc Details
NLU-78-71 License Amendment 31 Section 4.3.1.3
HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow ALL
Diagram
HBR2-08255 Sh 2,Fire Protection System Flow Diagram ALL
HBR2-08255 Sh. 6,Fire Protection System Flow Diagram ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.12 [Water Supply Compatible Thread Connections]
Chapter 3 Requirement: 3.5.12
Threads compatible with those used by local fire departments shall be provided on all hydrants, hose
HBRSEP LAR Rev 1 Page A-45
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
couplings, and standpipe risers.
Exception: Fire departments shall be permitted to be provided with adapters that allow interconnection
between plant equipment and the fire department equipment if adequate training and procedures are
provided.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
Doc Details
Section 9.5.1.4.2.5
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.13 [Water Supply Header Options]
Chapter 3 Requirement:
Compliance Statement
N/A
3.5.13
Headers fed from each end shall be permitted inside buildings to supply both sprinkler and standpipe
systems, provided steel piping and fittings meeting the requirements of ANSI B31.1, Code for Power
Piping, are used for the headers (up to and including the first valve) supplying the sprinkler systems
where such headers are part of the seismically analyzed hose standpipe system. Where provided, such
headers shall be considered an extension of the yard main system. Each sprinkler and standpipe
system shall be equipped with an outside screw and yoke, gate valve or other approved shutoff valve.
Compliance Basis
No headers at HBRSEP are fed from each
end.
Reference Document
HBR2-08255 Sh 2,Fire Protection System Flow Diagram ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.14 [Water Supply Control Valve Supervision]
Chapter 3 Requirement:
Compliance Statement
3.5.14*
All fire protection water supply and ire suppression system control valves shall be under a periodic
inspection program and shall be supervised by one of the following methods.
(a) Electrical supervision with audible and visual signals in the main control room or other suitable
constantly attended location.
Compliance Basis
HBRSEP LAR Rev 1 Page A-46
Duke EnergyComplies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
HBRSEP complies with Section 3.5.14 bya combination of (b) & (c).
Reference Document Doc Details
OST-602,Unit No. 2 Fire Water System Flowpath Verification ALL
(Monthly) and Valve Cycling (Annual)
OST-652,Unit 2 Containment Fire Water System Valves ALL
NLU-78-71,License Amendment 31 Section 4.3.1.3
Chapter 3 Requirement: (b) Locking valves in their normal position. Keys shall be made available only to authorized personnel.
Compliance Statement
Complies
Compliance Basis
HBRSEP complies with Section 3.5.14 by
a combination of (b) & (c).
Reference Document Doc Details
OST-602,Unit No. 2 Fire Water System Flowpath Verification ALL(Monthly) and Valve Cycling (Annual)
OST-652,Unit 2 Containment Fire Water System Valves ALL
NLU-78-71 ,License Amendment 31 Section 4.3.1.3
Chapter 3 Requirement: (c) Sealing valves in their normal positions. This option shall be utilized only where valves are locatedwithin fenced areas or under the direct control of the owner/operator.
Compliance Statement Compliance Basis
Complies HBRSEP complies with Section 3.5.14 bya combination of (b) & (c).
Reference Document Doc Details
OST-602,Unit No. 2 Fire Water System Flowpath Verification ALL(Monthly) and Valve Cycling (Annual)
OST-652,Unit 2 Containment Fire Water System Valves ALL
NLU-78-71 License Amendment 31 Section 4.3.1.3
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.15 [Water Supply Hydrant Code Requirements]
Chapter 3 Requirement: 3.5.15Hydrants shall be installed approximately every 250 ft (76 m) apart on the yard main system. A hose
HBRSEP LAR Rev 1 Page A-47
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements
house equipped with hose and combination nozzle and other auxiliary equipment specified in NFPA 24,
Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be provided
at intervals of not more than 1000 ft (305 m) along the yard main system.
Exception: Mobile means of providing hose and associated equipment, such as hose carts or trucks,
shall be permitted in lieu of hose houses. Where provided, such mobile equipment shall be equivalent to
the equipment supplied by three hose houses.
Compliance Statement Compliance Basis
Complies via Previous NRC Approval COMPLIES VIA PREVIOUS NRC
APPROVAL: In the SER dated 2/28/78,
Complies via Engineering Evaluation the NRC stated:
"Five yard fire hydrants are provided at
approximately 250 foot intervals around
the exterior of the plant except at the north
end of the plant, where the distance
between hydrants is somewhat larger. A
hose house located near each fire hydrant
contains 2-1/2 inch diameter fire hose and
other manual firefighting tools. A sixth
hose house is centrally located between
the reactor and Turbine Building, a
seventh is located at the intake structure,
and one hose house is located outside the
Unit 2 fence east of the auxiliary building.
Standard fire hose threads are used on allfire protection equipment, and the threads
are compatible with those used by the
local public fire departments.
We find that, subject to the implementation
of the above described modifications, the
fire water piping systems satisfy the
objectives identified in Section 2.1 of this
report and are, therefore, acceptable."
Proposed modifications were evaluated
and implemented per the SERs, where
applicable, to fulfill the intent of this
requirements.
There have been no plant modifications or
other changes that would invalidate the
basis for approval.
COMPLIES VIA ENGINEERING
EVALUATION: HBRSEP complies with the
applicable portions of NFPA 24 as detailed
in RNP-M/BMRK-1013.
HBRSEP LAR Rev 1 Page A-48
Duke EnergyReference Document
NLU-78-71 ,License Amendment 31
RNP-M/BMRK-1013,Code Compliance Evaluation NFPA 24-
Standard for Outside Protection
RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code
Compliance Variances
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDocDetalil
Section 4.3.1.3
ALL
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.5.16 [Water Supply Dedicated Limits]
Chapter 3 Requirement: 3.5.16*
The fire protection water supply system shall be dedicated for fire protection use only.
Exception No. 1: Fire protection water supply systems shall be permitted to be used to provide backup
to nuclear safety systems, provided the fire protection water supply systems are designed and
maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by
the applicable analysis.
Exception No. 2: Fire protection water storage can be provided by plant systems serving other
functions, provided the storage has a dedicated capacity capable of providing the maximum fire
protection demand for the specified duration as determined in this section.
Compliance Statement
License Amendment Required
Compliance Basis
NRC approval is being requested in
Attachment L for the use of the fire
protection water supply system for
purposes other than fire protection.
Reference Document
HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow
Diagram
HBR2-08255 Sh 2,Fire Protection System Flow Diagram
HBR2-08255 Sh 3,Fire Protection System Containment Flow
Diagram
HBR2-08255 Sh. 6,Fire Protection System Flow Diagram
OMM-002,Fire Protection Manual
AOP-014,Component Cooling Water System Malfunction
AOP-022,Loss of Service Water
EDMG-001,Extreme Damage Event Early Actions and Response
Determination Criteria
EDMG-002,Refueling Water Storage Tank (RWST)
EDMG-003,Condensate Storage Tank (CST)
Doc Details
ALL
ALL
ALL
ALL
Section 8.15
ALL
ALL
ALL
ALL
ALL
HBRSEP LAR Rev 1 Page A-49
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
EDMG-005 ,Containment Vessel (CV)
EDMG-01 1 ,Spent Fuel Pool Casualty
EDMG-012,Core Cooling Using Alternate Water Source
EDMG-013,Airborne Release Scrubbing
ALL
ALL
ALL
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.6 Standpipe and Hose Stations.
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.6.1 [Standpipe and Hose Station Code Requirements]
Chapter 3 Requirement: 3.6.1
For all power block buildings, Class III standpipe and hose systems shall be installed in accordance
with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems.
Compliance Statement
Complies via Engineering Evaluation
Complies via Previous NRC Approval
Compliance Basis
COMPLIES VIA ENGINEERING
EVALUATION: HBRSEP complies with
NFPA 14 as detailed in RNP-M/BMRK-
1010 and RNP-M/MECH-1709.
COMPLIES VIA PREVIOUS NRC
APPROVAL: HBRSEP has Class II
standpipes in lieu of Class I1l. In the
License Amendment dated 2/28/78, the
NRC stated:
"4.3.1.4 Interior Fire Hose Stations
A total of 24 interior hose stations, each
presently equipped with 50 feet of 1-1/2
inch diameter hose, have been provided
throughout all portions of the plant except
containment. There are presently several
safety-related areas containing
combustible materials that are beyond the
reach of the existing hose lines. The
HBRSEP LAR Rev 1 Page A-50
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy licensee will provide additional hose Fire Protection Program & Design Elements
stations or additional lengths of hose at
existing stations so that sufficient hose
reach is provided to protect all the areas of
the auxiliary building.
Hose racks originally used for unlined
hose are being used to store rubber lined
hose. The licensee has committed to
replace these with suitable hose reels or
hose racks designed and sized for lined
hose.
The nozzles on the interior hose lines are
1-1/2" spray nozzles. In areas with
electrical hazards, "electrically safe" hose
nozzles have been provided on the hose
station nearest these areas.
We find that, subject to the implementation
of the above described modifications, the
interior fire hose stations satisfy the
objectives identified in Section 2.1 of this
report and are, therefore, acceptable."
Proposed modifications were evaluated
and implemented per the SERs, where
applicable, to fulfill the intent of this
requirements.
There have been no plant modifications or
other changes that would invalidate the
basis for approval.
Reference Document Doc Details
RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, ALL
Standpipes and Hose Stations
RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code ALL
Compliance Variances
NLU-78-71 ,License Amendment 31 4.3.1.4
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.6.2 [Standpipe and Hose Station Capability Limitations]
Chapter 3 Requirement: 3.6.2
A capability shall be provided to ensure an adequate water flow rate and nozzle pressure for all hose
stations. This capability includes the provision of hose station pressure reducers where necessary for
the safety of plant industrial fire brigade members and off-site fire department personnel.
HBRSEP LAR Rev 1 Page A-51
Duke Energy
Compliance Statement
Complies via Engineering Evaluation
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Compliance BasisHBRSEP complies with NFPA 805requirement 3.6.2 as detailed in theapplicable portions RNP-M/BMRK-1010.
Reference Document Doc Details
RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, ALLStandpipes and Hose Stations
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.6.3 [Standpipe and Hose Station Nozzle Restrictions]
Chapter 3 Requirement: 3.6.3The proper type of hose nozzle to be supplied to each power block area shall be based on the area firehazards. The usual combination spray/straight stream nozzle shall not be used in areas where thestraight stream can cause unacceptable damage or present an electrical hazard to fire-fightingpersonnel. Listed electrically safe fixed fog nozzles shall be provided at locations where high-voltageshock hazards exist. All hose nozzles shall have shutoff capability and be able to control water flowfrom full open to full closed.
Compliance Statement Compliance BasisComplies via Engineering Evaluation HBRSEP complies with NFPA 805
requirement 3.6.3 as detailed in theapplicable portions of RNP-M/BMRK-1010.
Reference Document Doc Details
RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, Code Section 451 & 452Standpipes and Hose Stations
NLU-78-71 ,License Amendment 31 Section 4.3.1.4
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.6.4 [Standpipe and Hose Station Earthquake Provisions]
Chapter 3 ReqLirement: 3.6.4 1
Provisions shall be made to supply water at least to standpipes and hose stations for manual firesuppression in all areas containing systems and components needed to perform the nuclear safetyfunctions in the event of a safe shutdown earthquake (SSE).
HBRSEP LAR Rev 1 Page A-52
Duke EnergyCompliance Statement
Complies with Clarification
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Seismic standpipes are not an original
commitment for HBRSEP.
The Federal Register notice that
promulgated adoption of NFPA 805 makes
the following statement:
"A commenter noted that Appendix A to
BTP APCSB 9.5-1 did not require
seismically qualified standpipes and hose
stations for operating plants and plants
with construction permits issued prior to
July 1, 1976. NRC agrees that Appendix A
to BTP APCSB 9.5-1 made separate
provisions for operating plants and plants
with construction permits issued prior to
July 1, 1976, and did not require
seismically qualified standpipes and hose
stations for those plants. Therefore, the
requirement in Section 3.6.4 of NFPA 805
is not applicable to licensees with
nonseismic standpipes and hose stations
previously approved in accordance with
Appendix A to BTP APCSB 9.5-1."
There have been no plant modifications or
other changes that would invalidate the
basis for approval.
Reference Document
69 FR 33356,Final Rule - NFPA 805
DocDetails
Page 33544
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.6.5 [Standpipe and Hose Station Seismic Connection Limitations]
Chapter 3 Requirement: 3.6.5
Where the seismic required hose stations are cross-connected to essential seismic non-fire protection
water supply systems, the fire flow shall not degrade the essential water system requirement.
COm~lanceSttment
N/A
Compliance Basis
HBRSEP is not committed to havingseismic standpipes. See the ComplianceSection for 3.6.4
HBRSEP LAR Rev 1 Page A-53
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.7 Fire Extinguishers.
Chapter 3 Requirement: 3.7 Fire Extinguishers.
Where provided, fire extinguishers of the appropriate number, size, and type shall be provided inaccordance with NFPA 10, Standard for Portable Fire Extinguishers. Extinguishers shall be permitted tobe positioned outside of fire areas due to radiological conditions.
Compliance Statement
Complies via Engineering Evaluation
Compliance Basis
HBRSEP complies with NFPA 10 as
evaluated in RNP-M/BMRK-1001.
Reference Document Doc Details
RNP-M/BMRK-1001,Code Compliance Evaluation NFPA 10 Portable ALLFire Extinguishers
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.8 Fire Alarm and Detection Systems.
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A - General statement; No technicalrequirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.8.1 Fire Alarm.
Chapter 3 Requirement:
Compliance Statement
3.8.1 Fire Alarm.Alarm initiating devices shall be installed in accordance with NFPA 72, National Fire Alarm Code®.Alarm annunciation shall allow tle proprietary alarm system to transmit fire-related alarms, supervisorysignals, and trouble signals to th control room or other constantly attended location from whichrequired notifications and response can be initiated. Personnel assigned to the proprietary alarm stationshall be permitted to have other duties. The following fire-related signals shall be transmitted:(1) Actuation of any fire detection device
Compliance Basis
HBRSEP LAR Rev 1 Page A-54
Duke Energy(Main Header): Complies via Engineering
Evaluation
Section (1): Complies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
(Main Header): HBRSEP complies with
NFPA 72 as evaluated in the applicable
portions of RNP-M/BMRK-1014, 1005,
1006.
Section (1): No Additional Clarification
Reference Document
RNP-M/BMRK-1014,Code Compliance Evaluation NFPA 72 National
Fire Alarm Code
RNP-M/BMRK-1005,Code Compliance Evaluation NFPA 72D
RNP-M/BMRK-1006,Code Compliance Evaluation for NFPA 72E
RNP-M/MECH-1697,Evaluation of NFPA 72E Code Compliance
Variances
APP-044,Fire Alarm Console (FAC)
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
Doc Details
ALL
ALL
ALL
ALL
ALL
Appendix 9.5.1B-14
Chapter 3 Requirement: (2) Actuation of any fixed fire suppression system
Compliance Statement Compliance Basis
Section (2): Complies Section (2): No Additional Clarification
Reference Document Doc Details
APP-044,Fire Alarm Console (FAC) ALL
Chapter 3 Requirement: (3) Actuation of any manual fire alarm station
Compliance Statement Compliance Basis
Section (3): Complies Section (3): No Additional Clarification
Reference Document Doc Details
APP-044,Fire Alarm Console (FAC) ALL
Chapter 3 Requirement: (4) Starting of any fire pump
Compliance Statement Compliance Basis
Section (4): Complies Se tion (4): No Additional Clarification
Reference Document Doc Details
APP-044,Fire Alarm Console (FAC) ALL
HBRSEP LAR Rev 1 Page A-55
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design ElementsChapter 3 Requirement: (5) Actuation of any fire protection supervisory device
Compliance Statement
Section (5): Complies
Compliance Basis
Section (5): No Additional Clarification
Reference Document Doc Details
APP-044,Fire Alarm Console (FAC) ALL
Chapter 3 Requirement: (6) Indication of alarm system trouble condition
Compliance Statement Compliance Basis
Section (6): Complies Section (6): No Additional Clarification
Reference Document Doc Details
APP-044,Fire Alarm Console (FAC) ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.8.1.1 [Fire Alarm Communication Requirements]
Chapter 3 Requirement: 3.8.1.1
Means shall be provided to allow a person observing a fire at any location in the plant to quickly and
reliably communicate to the control room or other suitable constantly attended location.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
FP-001,Fire Emergency Section 5.2
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Sections 9.5.2
Table B-I NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.8.1.2 [Fire Al rm Prompt Notification Limits]
Chapter 3 Requirement: 3.8.1.2
Means shall be provided to promptly notify the following of any fire emergency in such a way as to allow
them to determine an appropriate course of action:
(1) General site population in all occupied areas.
HBRSEP LAR Rev 1 Page A-56
Duke EnergyCompliance Statement
Section (1): Complies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Compliance Basis
Section (1): No Additional Clarification
Reference Document DoDetaols
AOP-041, Response to Fire Event Section 2
Chapter 3 Requirement: (2) Members of the industrial fire brigade and other groups supporting fire emergency response
Compliance Statement Compliance Basis
Section (2): Complies Section (2): No Additional Clarification
Reference Document Doc Details
AOP-041,Response to Fire Event Section 2
Chapter 3 Requirement: (3) Off-site fire emergency response agencies. Two independent means shall be available (e.g.,
telephone and radio) for notification of off-site emergency services
Compliance Statement Compliance Basis
Section (3): Complies Section (3): No Additional Clarification
Reference Document Doc Details
PLP-007,Robinson Emergency Plan Attachment 6.1
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.8.2 Detection.
Chapter 3 Requirement: 3.8.2 Detection.
If automatic fire detection is required to meet the performance or deterministic requirements of Chapter
4, then these devices shall be installed in accordance with NFPA 72, National Fire Alarm Code, and its
applicable appendixes.
Compliance Statement
Complies via Engineering Evaluation
Compliance BasisHBRSEP complies with NFPA 72 as
evaluated in the applicable portions ofRNP-M/BMRK-1014, 1005, 1006.
Reference Document
RNP-M/BMRK-1014,Code Compliance Evaluation NFPA 72 National
Fire Alarm Code
Doc Details
ALL
HBRSEP LAR Rev 1 Page A-57
Duke Energy,RNP-M/BM RK-1005,Code Compliance Evaluation NFPA 72D
RNP-M/BMRK-1006,Code Compliance Evaluation for NFPA 72E
RNP-M/MECH-1697,Evaluation of NFPA 72E Code Compliance
Variances
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
ALL
ALL
ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference; 3.9 Automatic and Manual Water-Based Fire Suppression Systems.
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A - General statement; No technicalrequirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.9.1 [Fire Suppression System Code Requirements]
Chapter 3 Requirement: 3.9.1*
If an automatic or manual water-based fire suppression system is required to meet the performance or
deterministic requirements of Chapter 4, then the system shall be installed in accordance with the
appropriate NFPA standards including the following:
(1) NFPA 13, Standard for the Installation of Sprinkler Systems
Compliance Statement Compliance Basis
Section (1): Complies via Engineering Section (1): HBRSEP complies with NFPA
Evaluation 13 as evaluated in the applicable portions
of RNP-M/BMRK-1009.
Reference Document Doc Details
RNP-M/BMRK- 1009,Code Compliance Evaluation for NFPA 13, ALL
Sprinkler Systems
RNP-M/MECH-1725,Evaluation of NFPA 13 Code Compliance ALL
Variances
Chapter 3 Requirement: (2) NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection
Compliance Statement
Section (2): Complies via Engineering
Evaluation
Compliance BasisSection (2): HBRSEP complies with NFPA
15 as evaluated in the applicable portions
HBRSEP LAR Rev 1 Page A-58
Duke Energy of RNP-M/BMRK-1011.Attachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design Elements
Reference Document
RNP-M/BMRK-1011,Code Compliance Evaluation for NFPA 15,
Water Spray Fixed Systems
RNP-M/MECH-1726,Evaluation of NFPA 15 Code Compliance
Variances
RNP-M/MECH-1 727,Hydraulic Analysis of the Hydrogen Seal Oil
Water Spray System
RNP-M/MECH-1728,Hydraulic Analysis of the Auxiliary & Start-Up
Transformer Water Spray System
Doc Details
ALL
ALL
ALL
ALL
Chapter 3 Requirement: (3) NFPA 750, Standard on Water Mist Fire Protection Systems
Compliance Statement Compliance Basis
Section (3): N/A Section (3): N/A - No Water Mist FireProtection Systems are installed at
HBRSEP.
Chapter 3 Requirement: (4) NFPA 16, Standard for the Installation of Foam-Water Sprinkler and Foam-Water Spray Systems
Compliance Statement
Section (4): N/A
Compliance Basis
Section (4): N/A - No Foam-Water
Sprinkler or Foam-Water Spray Systems
are installed at HBRSEP.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.9.2 [Fire Suppression System Flow Alarm]
Chapter 3 Requirement: 3.9.2
Each system shall be equipped with a water flow alarm.
Compliance Statement Compliance Basis
Complies via Engineering Evaluation COMPLIES VIA ENGINEERINGEVALUATION: Several of the automatic
water-based fire suppressio' systems do
not have water flow alarms. ;These
systems have less than 20 sprinklers and
are not required to have water flow alarms.
The systems discussed are in the CCW
Pump Room, Turbine Generator
HBRSEP LAR Rev 1 Page A-59
Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design ElementsMonitoring Room (TGMR), and the Rad
Waste Building. Per RNP-M/BMRK- 1009,
"Code Compliance Evaluation NFPA 13-
Standard for Installation for Sprinkler
Systems", and RNP-M/BMRK-1011,
"Code Compliance Evaluation NFPA 15 -
Water Spray Fixed Systems", all
requirements from NFPA 13 and NFPA 15
for water flow alarms are met.
Reference Document Doc Details
RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Code Sections 3-16.2, 3-17.2
Sprinkler Systems
RNP-M/BMRK-1011,Code Compliance Evaluation for NFPA 15, Code Section 2124
Water Spray Fixed Systems
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.9.3 [Fire Suppression System Alarm Locations]
Chapter 3 Requirement: 3.9.3
All alarms from fire suppression systems shall annunciate in the control room or other suitable
constantly attended location.
Compliance Statement Compliance Basis
Complies via Engineering Evaluation HBRSEP complies with NFPA 805
requirement 3.9.3 as detailed in the
applicable portions of RNP-M/BMRK-
1009.
Reference Document Doc Details
RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Code Sections 5-3.5.2, 3-17.3.3, 3-17.6.2
Sprinkler Systems
RNP-M/BMRK-1011,Code Compliance Evaluation for NFPA 15, Code Section 8041
Water Spray Fixed Systems
APP-044,Fire Alarm Console (FAC) ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.9.4 [Fire Suppression System Diesel Pump Sprinkler Protection]
HBRSEP LAR Rev 1 Page A-60
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Enerqy 3Fire Protection Program & Design ElementsCDueterr3fRepuirement:o t.9ks
Diesel-driven fire pumps shall be protected by automatic sprinklers.
Compliance Statement
Complies via Engineering Evaluation
Compliance BasisCOMPLIES VIA ENGINEERING
EVALUATION: The diesel-driven fire
pump is installed outdoors at the Intake
Structure on Lake Robinson. Per RNP-
M/BMRK- 1012, "Code Compliance
Evaluation NFPA 20- Centrifugal Fire
Pumps", all requirements from NFPA 20-
1978, for outdoor diesel-driven fire pumps,
are met.
Reference Document Doc Details
HBR2-11937,Fire Pre-Plan Drawings, Sh 1-60 Sheet 45
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.9.5 [Fire Suppression System Shutoff Controls]
Chapter 3 Requirement: 3.9.5
Each system shall be equipped with an OS&Y gate valve or other approved shutoff valve.
Compliance Statement Compliance Basis
Complies via Engineering Evaluation HBRSEP complies with NFPA 805
requirement 3.9.5 as detailed in the
applicable portions of RNP-M/BMRK-1009
and RNP-M/BMRK-1011.
Reference Document Doc Details
RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Sections 3-13.1.1, 3-14.1.1
Sprinkler Systems
RNP-M/BMRK-101 1 ,Code Compliance Evaluation for NFPA 15, Section 2080
Water Spray Fixed Systems
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.9.6 [Fire Suppression System Valve Supervision]
Chapter 3 Requirement: 3.9.6
All valves controlling water-based fire suppression systems required to meet the performance or
HBRSEP LAR Rev 1 Page A-61
Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental
Fire Protection Program & Design Elementsdeterministic requirements of Chapter 4 shall be supervised as described in 3.5.14.
Compliance Statement
Complies
Compliance Basis
No Additional Clarification
Reference Document Doc Details
OST-602,Unit No. 2 Fire Water System Flowpath Verification ALL
(Monthly) and Valve Cycling (Annual)
OST-652,Unit 2 Containment Fire Water System Valves ALL
NLU-78-71 ,License Amendment 31 Section 4.3.1.3
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10 Gaseous Fire Suppression Systems.
Chapter 3 Requirement: N/A
Compliance Statement Compliance Basis
N/A N/A - General statement; No technical
requirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10.1 [Gaseous Suppression System Code Requirements]
Chapter 3 Requirement: 3.10.1
If an automatic total flooding and local application gaseous fire suppression system is required to meet
the performance or deterministic requirements of Chapter 4, then the system shall be designed and
installed in accordance with the following applicable NFPA codes:
(1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems
Compliance Statement
Section (1): Complies via Engineering
Evaluation
Compliance BasisSection (1): HBRSEP complies with NFPA
12 as evaluated in the applicable portions
of RNP-M/BMRK-1007.
Reference Document I
RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, CarbonDioxide Extinguishing Systems
RNP-M/MECH-1708,Evaluation of NFPA 12 Code ComplianceVariances
Doc Details
ALL
ALL
HBRSEP LAR Rev I Page A-62
Duke Energy
HBR2-11992 SH00001,EDG HIGH PRESSURE C02 FIRE
EXTINGUISHING SYSTEM
HBR2-11992 Sh 02,EDG High Pressure C02 Fire Extinguishing
System
HBR2-11992 Sh 03,EDG High Pressure C02 Fire Extinguishing
System
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
ALL
ALL
ALL
Chapter 3 Requirement: (2) NFPA 12A, Standard on Halon 1301 Fire Extinguishing Systems
Compliance Statement Compliance Basis
Section (2): Complies via Engineering Section (2): HBRSEP complies with NFPA
Evaluation 12A as evaluated in the applicable
portions of RNP-M/BMRK-1008.
Reference Document DocDetaols
RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A, ALL
Halon 1301 Systems
Chapter 3 Requirement: (3) NFPA 2001, Standard on Clean Agent Fire Extinguishing Systems
Compliance Statement Compliance Basis
Section (3): N/A Section (3): Clean Agent Fire
Extinguishing Systems are not utilized at
HBRSEP.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10.2 [Gaseous Suppression System Alarm Location]
Chapter 3 Requirement: 3.10.2
Operation of gaseous fire suppression systems shall annunciate and alarm in the control room or other
constantly attended location identified.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon
Dioxide Extinguishing Systems
RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A,
Halon 1301 Systems
Section 1452 & 1-8.5
Section 1-8.4
HBRSEP LAR Rev 1 Page A-63
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Enery
APP-044,-ire Alarm Console (FAC) ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10.3 [Gaseous Suppression System Ventilation Limitations]
Chapter 3 Requirement: 3.10.3
Ventilation system design shall take into account prevention from over-pressurization during agent
injection, adequate sealing to prevent loss of agent, and confinement of radioactive contaminants.
Compliance Statement Compliance Basis
Complies via Engineering Evaluation HBRSEP complies with NFPA 805requirement 3.10.3 as detailed in the
applicable portions of RNP-M/BMRK-1007
and RNP-M/BMRK-1008.
Reference Document Doc Detanis
RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon ALL
Dioxide Extinguishing Systems
RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A, ALL
Halon 1301 Systems
RNP-M/MECH-1708,Evaluation of NFPA 12 Code Compliance ALL
Variances
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10.4 [Gaseous Suppression System Single Failure Limits]
Chapter 3 Requirement: 3.10.4*
In any area required to be protected by both primary and backup gaseous fire suppression systems, a
single active failure or a crack in any pipe in the fire suppression system shall not impair both theprimary and backup fire suppression capability.
Compliance StatementN/A
Compliance Basis
No areas at HBRSEP are required to beprotected by both primary and backupgaseous fire suppression systems.
Table B-I NFPA 805 Ch.3 Transition Details
HBRSEP LAR Rev I Page A-64
Attachment A - NEI 04-02 Table B-1 Transition of Fundamental
Chapter 3 Reference: 3.10.5 [Gaseous Suppression System Disarming Controls]
Chapter 3 Requirement: 3.10.5
Provisions for locally disarming automatic gaseous suppression systems shall be secured and under
strict administrative control.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
OMM-002,Fire Protection Manual Section 8.13.6
OP-809,Diesel Generator Carbon Dioxide Suppression System ALL
OPS-NGGC- 1308,Plant Status Control ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10.6 [Gaseous Suppression System C02 Limitations)
Chapter 3 Requirement: 3.10.6*
Total flooding carbon dioxide systems shall not be used in normally occupied areas.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document DoDetals
HBR2-9717,Fire Area/Zone Locations ALL
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10.7 [Gaseous Suppression System C02 Warnings]
Chapter 3 Requirement:
Compliance Statement
Complies with Clarification
3.10.7
Automatic total flooding carbon dioxide systems shall be equipped with an audible pre-discharge alarm
and discharge delay sufficient to permit egress of personnel. The carbon dioxide system shall be
provided with an odorizer.
Compliance BasisSee proposed modification pertinent to
NFPA 805 Chapter 3, Section 3.10.7
compliance in Attachment "S", Table S-2
of the Transition Report.
HBRSEP LAR Rev 1 Page A-65
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy
Reference Document Doc Details
RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon Code Sections 122 & 1-6.2
Dioxide Extinguishing Systems
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10.8 [Gaseous Suppression System C02 Required Disarming]
Chapter 3 Requirement: 3.10.8
Positive mechanical means shall be provided to lock out total flooding carbon dioxide systems during
work in the protected space.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document DoDetails
OP-805,Carbon Dioxide Suppression System Section 8.3
OP-809,Diesel Generator Carbon Dioxide Suppression System Section 8.3
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.10.9 [Gaseous Suppression System Cooling Considerations]
Chapter 3 Requirement: 3.10.9
The possibility of secondary thermal shock (cooling) damage shall be considered during the design of
any gaseous fire suppression system, but particularly with carbon dioxide.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Malls
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.1
BTP APCSB 9.5-1,Guideline for Fire Protection for Nuclear Power Appendix A Section E.4 & E.5
Plants Docketed Prior to July 1, 1976
Table B-1 NFPA 805 Ch.3 Transition Details
HBRSEP LAR Rev 1 Page A-66
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements
Chapter 3 Reference: 3.10.10 [Gaseous Suppression System Decomposition Issues]
Chapter 3 Requirement: 3.10.10Particular attention shall be given to corrosive characteristics of agent decomposition products on
safety systems.
Compliance Statement Compliance Bas• s
Complies No Additional Clarification
Reference Document Doc Details
BTP APCSB 9.5-1,Guideline for Fire Protection for Nuclear Power Appendix A Section E.4 & E.5
Plants Docketed Prior to July 1, 1976
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.1
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.11 Passive Fire Protection Features
Chapter 3 Requirement: 3.11 Passive Fire Protection Features.This section shall be used to determine the design and installation requirements for passive protection
features. Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire
dampers, and through fire barrier penetration seals. Passive fire protection features also include
electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical
components and equipment from the effects of fire.
Compliance Statement
N/A
Compliance Basis
N/A - General statement; No technicalrequirements.
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.11.1 Building Separation.
Chapter 3 Requirement: 3.11.1 Building Separation.
Each major building within the power block shall be separated from the others by barriers having a
designated fire resistance rating of 3 hours or by open space of at least 50 ft (15.2 m) or space that
meets the requirements of NFPA 80A, Reconmended Practice for Protection of Buildings from Exterior
Fire Exposures.
Exception: Where a performance-based analysis determines the adequacy of building separation, the
requirements of 3.11.1 shall not apply.
HBRSEP LAR Rev 1 Page A-67
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design ElementsCompliance Statement Compliance Basis
Complies COMPLIES: No Additional Clarification
Complies via Engineering Evaluation COMPLIES VIA ENGINEERING
EVALUATION: Design and installation
deviations pertaining to passive fire
protection features are evaluated on a fire
zone basis at HBRSEP in calculations
RNP-M/MECH-1672 through RNP-
M/MECH-1696 or EC packages as
installed in the plant.
Reference Document DocDetails
RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 1
RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 2
RNP-M/MECH-1674,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 3
RNP-M/MECH-1675,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 4
RNP-M/MECH-1676,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 5
RNP-M/MECH-1677, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 6
RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 7
RNP-M/MECH-1679,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 8
RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 9
RNP-M/MECH-1681 Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 10
RNP-M/MECH-1682, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 11
RNP-M/MECH-1684, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 13
RNP-M/MECH-1685, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 14
RNP-M/MECH-1686, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 15
RNP-M/MECH-1687, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 16
RNP-M/MECH- 1688, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 17
HBRSEP LAR Rev 1 Page A-68
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke EnerlRNP-MECH- 1689Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 18
RNP-M/MECH- 1690, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 19
RNP-M/MECH-1683, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 12
RNP-M/MECH-1691 ,Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 20
RNP-M/MECH-1692, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 21
RNP-M/MECH-1693,Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 22
RNP-M/MECH- 1694, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 23
RNP-M/MECH-1695, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 24
RNP-M/MECH-1696, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 27
NLU-78-71 ,License Amendment 31
FPP-RNP-900,Fire Hazards Analysis
HBR2-9717,Fire Area/Zone Locations
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Sections 4.11 & 4.14
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Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.11.2 Fire Barriers.
Chapter 3 Requirement: 3.11.2 Fire Barriers.
Fire barriers required by Chapter 4 shall include a specific fire-resistance rating. Fire barriers shall be
designed and installed to meet the specific fire resistance rating using assemblies qualified by fire tests.
The qualification fire tests shall be in accordance with NFPA 251, Standard Methods of Tests of Fire
Endurance of Building Construction and Materials, or ASTM E 119, Standard Test Methods for Fire
Tests of Building Construction and Materials.
Compliance Statement
Complies
Complies via Engineering Evaluation
Compliance BasisCOMPLIES: No Additional Clarification
COMPLIES VIA ENGINEERING
EVALUATION: Design and installation
deviations pertaining to passive fire
protiction features are evaluated on a fire
zone basis at HBRSEP in calculations
RNP-M/MECH-1672 through RNP-
M/MECH-1696 or EC packages as
installed in the plant.
HBRSEP LAR Rev 1
Page A-69
HBRSEP LAR Rev 1 Page A-69
Attachment A - NEI 04-02 Table B-1 Transition of Fundamentalkire Prnrtion-Empn_DUke a Pnrn,,
Reference Document
RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 1
RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 2
RNP-M/MECH-1674,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 3
RNP-M/MECH- 1675,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 4
RNP-M/MECH- 1676,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 5
RNP-M/MECH- 1677,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 6
RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 7
RNP-M/MECH-1679,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 8
RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 9
RNP-M/MECH- 1681, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 10
RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 11
RNP-M/MECH-1683,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 12
RNP-M/MECH-1684,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 13
RNP-M/MECH-1685,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 14
RNP-M/MECH-1686,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 15
RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 16
RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 17
RNP-M/MECH-1689,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 18
RNP-M/MECH-1690,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 19
RNP-M/MECH- 1691, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 20
RNP-M/MECH-1692,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 21
RNP-M/MECH-1693,Evaluation of Non-Standard Fire Barrier
DALDL lls
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HBRSEP LAR Rev 1 Page A-70
Duke EnergyPenetration Seals in Fire Zone 22
RNP-M/MECH-1 694,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 23
RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 24
RNP-M/MECH-1696,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 27
NLU-78-71 ,License Amendment 31
FP-014,Control of Fire Barrier Penetrations
RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration
Seals
EE-87-0166,Evaluation Of Concrete Brick "Rubble" Fire Barrier
Penetration Seals
ESR-97-405,Evaluation of Pyrocrete Fire Barrier Designs
EE-93-0095,Evaluation of Reduced Pre Soak Time for Grouting
ESR-94-1003,Discrepancy Resolution For RNP2-M-063
FPP-RNP-900,Fire Hazards Analysis
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
ALL
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Sections 4.11 & 4.14
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Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.11.3 Fire Barrier Penetrations.
Chapter 3 Requirement: 3.11.3* Fire Barrier Penetrations.
Penetrations in fire barriers shall be provided with listed fire-rated door assemblies or listed rated fire
dampers having a fire resistance rating consistent with the designated fire resistance rating of the
barrier as determined by the performance requirements established by Chapter 4. (See 3.11.3.4 for
penetration seals for through penetration fire stops.) Passive fire protection devices such as doors and
dampers shall conform with the following NFPA standards, as applicable:
(1) NFPA 80, Standard for Fire Doors and Fire Windows.
Compliance Statement
Section (1):Complies via Engineering Evaluation
Compliance Basis
Section (1):Design and installation deviationspertaining to passive fire protectionfeatures are evaluated on a fire zone basisat HBRSEP in calculations RNP-M/MECH-1672 through RNP-M/MECH-1696 or ECpackages as installed in the plant.
HBRSEP complies with NFPA 80 asevaluated in the applicable portions ofRNP-M/BMRK-1003.
HBRSEP LAR Rev 1 Page A-71
Duke EnergyReference Document
RNP-M/MECH-1672, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 1
RNP-M/MECH-1673, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 2
RNP-M/MECH-1674, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 3
RNP-M/MECH-1675, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 4
RNP-M/MECH-1676,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 5
RNP-M/MECH-1677, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 6
RNP-M/MECH-1678, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 7
RNP- M/MECH- 1679 Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 8
RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 9
RNP-M/MECH-1681 ,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 10
RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 11
RNP-M/MECH-1683, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 12
RNP-M/MECH-1684, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 13
RNP-M/MECH-1685, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 14
RNP-M/MECH-1686,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 15
RNP-M/MECH-1687, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 16
RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 17
RNP-M/MECH-1689,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 18
RNP-M/MECH-1690,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 19
RNP-M/MECH-1691 Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 20
RNP-M/MECH-1692, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 21
RNP-M/MECH-1693,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 22
HBRSEP LAR Rev 1
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
Doc Dtals
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Page A-72
Duke Energy
RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 23
RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 24
RNP-M/MECH-1696,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 27
RNP-M/BMRK-1003,CODE COMPLIANCE EVALUATION NFPA 80
STANDARD FOR FIRE DOORS AND WINDOWS
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration
Seals
RNP-M/MECH-1 670,Evaluation Of Concrete Hatch Covers
EE-90-0104,Generic Evaluation Of HVAC Fire Damper And Fire
Door Installation Discrepancies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
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Appendix 9.5.1A-7
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Chapter 3 Requirement: (2) NFPA 90A,
Compliance Statement
Section (2): Complies via Engineering
Evaluation
Standard for the Installation of Air-Conditioning and Ventilating Systems.
Compliance Basis
Section (2):
Design and installation deviations
pertaining to passive fire protection
features are evaluated on a fire zone basis
at HBRSEP in calculations RNP-M/MECH-
1672 through RNP-M/MECH-1696 or EC
packages as installed in the plant.
HBRSEP complies with NFPA 90A as
evaluated in the applicable portions of
RNP-M/BMRK-1004.
Reference Document
RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 1
RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 2
RNP-M/MECH-1674,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 3
RNP-M/MECH- 1675,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 4
RNP-M/MECH-1676,Ev luation of Non-Standard Fire Barrier
Penetration Seals in FireZone 5
RNP-M/MECH-1677, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 6
RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 7
Doc Details
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HBRSEP LAR Rev 1 Page A-73
Duke Energy
RNP-M/MECH-1679, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 8
RNP-M/MECH- 1680,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 9
RNP-M/MECH- 1681 ,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 10
RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 11
RNP-M/MECH-1683,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 12
RNP-M/MECH-1684,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 13
RNP-M/MECH-1685,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 14
RNP-M/MECH-1686, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 15
RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 17
RNP-M/MECH-1689, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 18
RNP-M/MECH- 1687 Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 16
RNP-M/MECH-1690, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 19
RNP-M/MECH- 1691 Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 20
RNP-M/MECH-1692,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 21
RNP-M/MECH-1693, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 22
RNP-M/MECH-1694, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 23
RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 24
RNP-M/MECH-1696, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 27
RNP-M/BMRK-1004,CODE COMPLIANCE EVALUATION FOR
NFPA 90A 1976 & 1985 EDITIONS AIR CONDITIONING
UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)
RNP2-M-063 Selection of 3 Hour Fire Rated Barrier Penetration
Seals
ESR-94-1003,Discrepancy Resolution For RNP2-M-063
EE-90-0104,Generic Evaluation Of HVAC Fire Damper And Fire
Door Installation Discrepancies
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
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Appendix 9.5.1A-7
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HBRSEP LAR Rev 1 Page A-74
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalIii 11W-Pperav E-re P-rotecton P-roprlms JC. I IPCIFl -ments
Chapter 3 Requirement: (3) NFPA 101, Life Safety Code
Exception: Where fire area boundaries are not wall-to-wall, floor-to-ceiling boundaries with allpenetrations sealed to the fire rating required of the boundaries, a performance-based analysis shall berequired to assess the adequacy of fire barrier forming the fire boundary to determine if the barrier will
withstand the fire effects of the hazards in the area. Openings in fire barriers shall be permitted to beprotected by other means as acceptable to the AHJ.
Compliance Statement
Section (3): Complies with Clarification
Compliance BasisSection (3): HBRSEP complies with
clarification with NFPA 101. HBRSEP
complies with NFPA 101 with regards to
fire rated door assemblies since NFPA
101, Section 8.3.3.1 refers to NFPA 80,
which is evaluated in RNP-M/BMRK-1003.
HBRSEP complies with NFPA 101 with
regards to rated fire dampers since NFPA
101, Section 9.2.1 refers to NFPA 90A,
which is evaluated in RNP-M/BMRK-1004
Reference Document Doc Details
RNP-M/BMRK-1003,CODE COMPLIANCE EVALUATION NFPA 80 ALL
STANDARD FOR FIRE DOORS AND WINDOWS
RNP-M/BMRK-1004,CODE COMPLIANCE EVALUATION FOR ALL
NFPA 90A 1976 & 1985 EDITIONS AIR CONDITIONING
NFPA 101,Life Safety Code, 2009 Edition Sections 8.3.3.1 & 9.2.1
Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.11.4 Through Penetration Fire Stops.
Chapter 3 Requirement: 3.11.4* Through Penetration Fire Stops.
Through penetration fire stops for penetrations such as pipes, conduits, bus ducts, cables, wires,
pneumatic tubes and ducts, and similar building service equipment that pass through fire barriers shall
be protected as follows.
(a) The annular space between the penetrating item and the through opening in the fire barrier shall be
filled with a qualified fire-resistive penetration seal assembly capable of maintaining the fire resistance
of the fire barrier. The assembly shall be qualified by tests in accordance with a fire test protocol
acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period.
Compliance Statement Compliance Basis
Section (a): Section (a):
Complies COMPLIES: No Additional Clarification
Complies via Engineering Evaluation COMPLIES VIA ENGINEERING
HBRSEP LAR Rev 1 Page A-75
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements
EVALUATION: Design and installationdeviations pertaining to passive fireprotection features are evaluated on a firezone basis at HBRSEP in calculationsRNP-M/MECH-1672 through RNP-M/MECH-1696 or EC packages asinstalled in the plant.
Engineering evaluations were developedto analyze the acceptability of the typicalpenetration seal designs utilized atHBRSEP.
Reference Document Doe Details
RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 1
RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 2
RNP-M/MECH-1674,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 3
RNP-M/MECH-1675,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 4
RNP-M/MECH-1676,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 5
RNP-M/MECH-1677,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 6
RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 7
RNP-M/MECH-1679, Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 8
RNP-M/MECH-1680, Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 9
RNP-M/MECH-1681 Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 10
RNP-M/MECH-1682, Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 11
RNP-M/MECH-1683,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 12
RNP-M/MECH-1684, Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 13
RNP-M/MECH-1685,Evaluation kf Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 14
RNP-M/MECH-1686,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 15
RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier ALL
HBRSEP LAR Rev 1 Page A-76
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design ElementsPenetration Seals in Fire Zone 16
RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 17
RNP-M/MECH-1689,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 18
RNP-M/MECH-1690, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 19
RNP-M/MECH-1691 Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 20
RNP-M/MECH-1692, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 21
RNP-M/MECH-1693, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 22
RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 23
RNP-M/MECH-1695, Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 24
RNP-M/MECH-1696,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 27
HBR2-09716,Fire Barrier Penetrations ALL
GID/R87038/0014,Design Basis Document; Fire Barrier System Sections 4.1.3 & 4.1.4
FP-014,Control of Fire Barrier Penetrations ALL
RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration ALL
Seals
EE-93-0095,Evaluation of Reduced Pre Soak Time for Grouting ALL
ESR-94-1003,Discrepancy Resolution For RNP2-M-063 ALL
EE-90-0025,Past And Present Operability Of Steam Generator ALL
Blowdown Line Penetration Seals (Penetrations CP-2674 And CP-
5612)
EE-93-0043,Evaluation Of Temporary Fire Barrier Penetration Seals ALL
Between Fire Zone 11 And 24
ESR-98-0221, Penetration Seals Containing Copper Piping and ALL
Tubing
RNP-M/MECH-1671, Evaluation Of Large Bore Piping Penetrations ALL
Chapter 3 Requirement: (b) Conduits shall be provided with an internal fire seal that has an equivalent fire-resistive rating to that
of the fire barrier through opening fire stop and shall be permitted to be installed on either side of the
barrier in a location that is as close to the barrier as possible.
Exception: Openings inside conduit 4 in. (10.2 cm) or less ir diameter shall be sealed at the fire barrier
with a fire-rated internal seal unless the conduit extends greater than 5 ft (1.5 m) on each side of the
fire barrier. In this case the conduit opening shall be provided with noncombustible material to prevent
the passage of smoke and hot gases. The fill depth of the material packed to a depth of 2 in. (5.1 cm)shall constitute an acceptable smoke and hot gas seal in this application.
HBRSEP LAR Rev 1 Page A-77
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design ElementsCompliance Statement Compliance Basis
Section (b): Complies Section (b):
COMPLIES: No Additional Clarification
Complies via Engineering Evaluation
COMPLIES VIA ENGINEERING
EVALUATION: Design and installation
deviations pertaining to passive fireprotection features are evaluated on a fire
zone basis at RNP in calculations RNP-
M/MECH-1672 through RNP-M/MECH-
1696 or EC packages as installed in the
plant.
Reference Document DocDtails
RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 1
RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 2
RNP-M/MECH-1674,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 3
RNP-M/MECH-1675,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 4
RNP-M/MECH-1676,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 5
RNP-M/MECH-1677,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 6
RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 7
RNP-M/MECH- 1679,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 8
RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 9
RNP-M/MECH-1681 Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 10
RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 11
RNP-M/MECH-1683,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 12
RNP-M/MECH-1684,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 13
RNP-M/ME!•H- 1685,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 14
RNP-M/MECH-1686,Evaluation of Non-Standard Fire Barrier ALL
Penetration Seals in Fire Zone 15
RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier ALL
HBRSEP LAR Rev 1 Page A-78
Duke EnergyPenetration Seals in Fire Zone 16
RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 17
RNP-M/MECH-1689,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 18
RNP-M/MECH-1690, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 19
RNP-M/MECH-1691,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 20
RNP-M/MECH- 1692, Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 21
RNP-M/MECH- 1693 Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 22
RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 23
RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 24
RNP-M/MECH- 1696,Evaluation of Non-Standard Fire Barrier
Penetration Seals in Fire Zone 27
GID/R87038/0014,Design Basis Document; Fire Barrier System
CTL# CRE093-4324,Conduit Fire Test of One Hundred One
Electrical Conduit Penetrations
RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration
Seals
ESR-94-1003,Discrepancy Resolution For RNP2-M-063
ESR-94-0930,Fire Barriers
ESR-94-1103,Evaluation for 12" of Dow Corning Foam For Conduit
Seals
NED-B/MECH- 1001, Fire Resistance of Capped Conduits
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements
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Section 2.2.2.1, Table A.4.0-1
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Table B-1 NFPA 805 Ch.3 Transition Details
Chapter 3 Reference: 3.11.5 Electrical Raceway Fire Barrier Systems (ERFBS).
Chapter 3 Requirement: 3.11.5* Electrical Raceway Fire Barrier Systems (ERFBS).ERFBS required by Chapter 4 shall be capable of resisting the fire effects of the hazards in the area.
ERFBS shall be tested in accordance with and shall meet the acceptance criteria of NRC Generic Letter
86-10, Supplement 1, "Fire Endurance ITest Acceptance Criteria for Fire Barrier Systems Used to
Separate Safe Shutdown Trains Within the Same Fire Area." The ERFBS needs to adequately address
the design requirements and limitations of supports and intervening items and their impact on the fire
barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits
free of fire damage for a specific thermal exposure, barrier design, raceway size and type, cable size,
fill, and type shall be demonstrated.
HBRSEP LAR Rev 1 Page A-79
Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements
Exception No. 1: When the temperatures inside the fire barrier system exceed the maximum
temperature allowed by the acceptance criteria of Generic Letter 86-10, "Fire Endurance Acceptance
Test Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Training Within the
Same Fire Area," Supplement 1, functionality of the cable at these elevated temperatures shall be
demonstrated. Qualification demonstration of these cables shall be performed in accordance with the
electrical testing requirements of Generic Letter 86-10, Supplement 1, Attachment 1, "Attachment
Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Barrier Systems During
and After Fire Endurance Test Exposure.
Exception No. 2: ERFBS systems employed prior to the issuance of Generic Letter 86-10, Supplement1, are acceptable providing that the system successfully met the limiting end point temperature
requirements as specified by the AHJ at the time of acceptance.
Compliance Statement Compliance Basis
Complies No Additional Clarification
Reference Document Doc Details
GID/R87038/0014,Design Basis Document; Fire Barrier System Section 4.4
Table B-1 NFPA 805 Ch.3 Transition Details
HBRSEP LAR Rev 1 Page A-80
Duke Energy Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
B. NEI 04-02 Table B-2 - Nuclear Safety Capability Assessment -Methodology Review
96 Pages Attached
HBRSEP LAR Rev I 7age B-1
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical componentsrequired to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
This section discusses a generic deterministic methodology and criteria that licensees can use to3 Deterministic perform a post-fire safe shutdown analysis to address regulatory requirements. The plant-specific
Methodology analysis approved by NRC is reflected in the plant's licensing basis. The methodology described in
this section is also an acceptable method of performing a post-fire safe shutdown analysis. This
methodology is indicated in Figure 3-1. Other methods acceptable to NRC may also be used.
Regardless of the method selected by an individual licensee, the criteria and assumptions provided inthis guidance document may apply. The methodology described in Section 3 is based on a computer
database oriented approach, which is utilized by several licensees to model Appendix R data
relationships. This guidance document, however, does not require the use of a computer database
oriented approach.
The requirements of Appendix R Sections III.G.1, III.G.2 and ill.G.3 apply to equipment and cables
required for achieving and maintaining safe shutdown in any fire area. Although equipment and
cables for fire detection and suppression systems, communications systems and 8-hour emergency
lighting systems are important features, this guidance document does not address them.
Additional information is provided in Appendix B to this document.
Applicability Comments
Applicable
AlignmentAlignment Basis
Statement
Robinson Nuclear Plant's (HBRSEP) Safe Shutdown Methodology was reviewed against theAligns requirements of Appendix R Sections IlI.G, liI.J, and 1l1.L as required by 10CFR50.48(b). NRC review
and approval of the HBRSEP safe shutdown methodology is contained in a series of Safety
Evaluation Reports.
For the re-validation of the Safe Shutdown Analysis (SSA) performed prior to and in conjunction with
the transition process, NEI 00-01, Revision 1 (which has since been updated to revision 2) was oneof the references used in developing the circuit analysis procedure, which is now captured in FIR-
NGGC-0101, Revision 2. Except as noted in this document, the plant's methodology meets the
guidelines of NEI 00-01.
Comments
HBRSEP LAR Rev 1 Page B-2
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuk••Energy Capabil . dolommReview
Reference Document
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability
Assessment (NSCA)
NLS-84-516, Fire Protection Rule - Alternate Safe Shutdown
Capability - Sections III.G.3 and III.L of Appendix R to 10 CFR 50 -
H.B. Robinson Steam Electric Plant Unit No. 2
NLS-85-732, Supplemental Safety Evaluation for Appendix R to 10
CFR 50, Items IlI.G.3 and Il.L; Alternate Safe Shutdown Capability -
H.B. Robinson Steam Plant, Unit 2 - TAC No. 60106
Doc Detail
Section 3.2, 3.34
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE IRef NEI 00-01 GuidanceThis section discusses the identification of systems available and necessary to perform the required
3.1[Itro]n Safems safe shutdown functions. It also provides information on the process for combining these systems
Shutdown Systems into safe shutdown paths. Appendix R Section III.G.1.a requires that the capability to achieve and
and Path maintain hot shutdown be free of fire damage. It is expected that the term "free of fire damage" will
Development be further clarified in a forthcoming Regulatory Issue Summary. Appendix R Section IIl.G.1.b
requires that repairs to systems and equipment necessary to achieve and maintain cold shutdown be
completed within 72 hours. It is the intent of the NRC that requirements related to the use of manual
operator actions will be addressed in a forthcoming rulemaking.
[Refer to hard copy of NEI 00-01 for Figure 3-11
Applicability Comments
Applicable
AlignmentSttment
Aligns
Alignment Basg
RNP-E/ELEC-1216 identifies the systems and components necessary achieve and maintan safeshutdown.
For the re-4lidation of the Safe Shutdown Analysis (SSA) performed prior to a4d in conjunction withthe transition process, NEI 00-01, Revision 1 was one of the references used in developing the circuitanalysis procedure, which is now captured in FIR-NGGC-0101, revision 1. Except as noted in thisdocument, the plant's methodology meets the guidelines of NEI 00-01.
HBRSEP LAR Rev 1 Page B-3
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety CapabilityAssessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.0Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI00-01 Ref NEI 00-01 GuidanceThe goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and
3.1 [B, Goals] Safe components remains free of fire damage for a single fire in any single plant fire area. This goal isShutdown Systems accomplished by determining those functions important to achieve and maintain hot shutdown. Safeand Path shutdown systems are selected so that the capability to perform these required functions is a part ofDevelopment each safe shutdown path. The functions important to post-fire safe shutdown generally include, but
are not limited to the following:
Reactivity controlPressure control systemsInventory control systemsDecay heat removal systemsProcess monitoringSupport systems- Electrical systems- Cooling systems
These functions are of importance because they have a direct bearing on the safe shutdown goal ofbeing able to achieve and maintain hot shutdown which ensures the integrity of the fuel, the reactorpressure vessel, and the primary containment. If these functions are preserved, then the plant will besafe because the fuel, the reactor and the primary containment will not be damaged. By assuringthat this equipment is not damaged and remains functional, the protection of the health and safety ofthe public is assured.
Applicability Comments
Applicable
AlianmentAlignmentBasis
HBRSEP LAR Rev 1 Page B-4
Duke Energytatement
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
RNP-E/ELEC-1216 identifies the safe shutdown functions.Aligns
Comments
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.1
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
0-01Ref NEI 00-01 Guidance
In addition to the above listed functions, Generic Letter 81-12 specifies consideration of associated3.1 [C, Spurious circuits with the potential for spurious equipment operation and/or loss of power source, and the
Operations] Safe common enclosure failures. Spurious operations/actuations can affect the accomplishment of the
Shutdown Systems post-fire safe shutdown functions listed above. Typical examples of the effects of the spurious
and Path operations of concern are the following:
Development
- A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup
capability
- A flow loss or blockage in the inventory makeup or decay heat removal systems being used for the
required safe shutdown path.
Spurious operations are of concern because they have the potential to directly affect the ability to
achieve and maintain hot shutdown, which could affect the fuel and cause damage to the reactorpressure vessel or the primary containment. Common power source and common enclosure
concerns could also affect these and must be addressed.
Applicability CommentsApplicable
AlignmentStateent
Alignment Basi
AlignsHBRSEP has considered spurious operation, common power sources, and common enclosure
concerns that would cause a circuit to be considered an associated circuit.
RCS isolation valves (such as the RHR Pump suction valves) are defined as high/low pressure
HBRSEP LAR Rev 1 Page B-5
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewinterface boundary valves if their spurious operation could lead to the rupture of low pressure piping
or a loss of RCS inventory that exceeds the RCS makeup capability. Such interface boundary valves
are subject to more stringent circuit analysis criteria, and are identified in FSSPMD by the HLP flag.
This high/low pressure interface boundary valve definition is conservative with respect to that in in
Appendix C of NEI 00-01 and NFPA-805 FAQ 06-0006.
During the re-validation, the definition from the previous SSA was carried forward for conservatism.Thus, some components are classified as high-low interfaces which do not meet the above definition
since their spurious opening will not result in a rupture of downstream piping and a subsequent
intersystem LOCA. Robinson may choose to remove the classification of these components as high-
low interfaces at a future date.
Comments
Reference Document DoDetal
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.4, 3.34
Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
The following criteria and assumptions may be considered when identifying systems available and3. 1.1 Criteria/ necessary to perform the required safe shutdown functions and combining these systems into safe
Assumptions shutdown paths.
Applicability Comments
Applicable
AlignmentAligmentAlignment BasesStatement
This is generic introductory information and contains no specific requirements.N/A
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-6
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI00-01Ref NEI 00-01 Guidance[BWR] GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths For The
3.1.1.1 [GE BWR BWR" addresses the systems and equipment originally designed into the GE boiling water reactors
Paths] (BWRs) in the 1960s and 1970s, that can be used to achieve and maintain safe shutdown per
Section III.G.1 of 10CFR 50, Appendix R. Any of the shutdown paths (methods) described in this
report are considered to be acceptable methods for achieving redundant safe shutdown.
Applicability Comments
Not Applicable
Alignment
Statement
HBRSEP is a PWR. This guidance is specific to BWRs.N/A
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical componentsrequired to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEL00-IRef
3.1.1.10 [Manual I
Automatic Initiation
of Systems]
NEI 00-01 Guidance
Manual initiation from the main control room or emergency control stations of systems required to
achieve and maintain safe shutdown s acceptable where permitted by current regulations or
approved by NRC; automatic initiatio r of systems selected for safe shutdown is not required but may
be included as an option.
ApplicabilityApplicable
HBRSEP LAR Rev 1 Page B-7
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
AlignmenttAligimitAlignment Basis
Statement
Reliance on the automatic logic for safe shutdown systems is not required, but if credited needs to beAligns appropriately evaluated as being free of fire damage. The only automatic logics evaluated at
HBRSEP are the Emergency Diesel Generator (EDG) Automatic Sequencing Logic and the Fast Bus
Transfer logic. These logics are incorporated in the overall SSD fault tree for HBRSEP.
Comments
Reference Document DocDetaal
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.4.3 and 2.2.4.4
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE[D00-1Ref NEI 00-01 Guidance
Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and3.1.1.11 [Multiple maintain safe shutdown for each affected unit must be demonstrated.
Affected Units]
Applicability Comments
Not Applicable
AlignmentAlignment BasisStatement
Robinson is a single unit site.N/A
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805
Requirement
A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
event shall be developed. The equipment list shall contain an inventory of those critical components
HBRSEP LAR Rev 1 Page B-8
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewrequired to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE[ 0-1Ref
3.1.1.2 [SRVs / LP
Systems]
NEI 00-01 Guidance
[BWR] GE Report GE-NE-T43-00002-00-03-R01 provides a discussion on the BWR Owners' Group(BWROG) position regarding the use of Safety Relief Valves (SRVs) and low pressure systems
(LPCI/CS) for safe shutdown. The BWROG position is that the use of SRVs and low pressure
systems is an acceptable methodology for achieving redundant safe shutdown in accordance with the
requirements of 10CFR50 Appendix R Sections III.G.1 and lII.G.2. The NRC has accepted the
BWROG position and issued an SER dated Dec. 12, 2000.
A4pplicabilit Comments
Not Applicable
Alignment Alignment Basis
StatementHBRSEP is a PWR. This guidance is specific to BWRs.
N/A
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI0 Ref NEI 00-01 Guidance
[PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be3.1.1.3 [Pressurizer maintained without the use of pressurizer heaters (i.e., pressure control is provided by controlling the
Heaters] makeup/charging pumps). Hot shutdown conditions can be maintained via natural circulation of the
RCS through the steam generators. The cooldown rate must be controlled to prevent the formation of
a bubble in the reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well
as steam release must be controlled.
Applicability Comments
Applicable
AlignmentStatement
Aliganment Basis
HBRSEP LAR Rev 1 Page B-9
Duke Energy
Aligns with intent
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
In most fire areas, HBRSEP does not rely on the use of pressurizer heaters to maintain hot
shutdown.
RCS pressure is controlled during hot shutdown and cooldown by controlling the rate of charging to
the RCS. Pressurizer heaters and/or auxiliary spray reduces operator burden. Neither component is
required to provide adequate pressure control if charging is available. Pressure reductions are made
by allowing the RCS to cool/shrink, thus reducing Pressurizer level/pressure. Pressure increases are
made by initiating charging/makeup to maintain Pressurizer level/pressure. Manual control of the
related pumps is acceptable.
Use of the SI Pumps in lieu of the Charging Pumps may be required for certain shutdown scenarios.
The RCS would have to be depressurized to less than operating pressure of the SI pumps.
Pressurizer heaters are credited to stabilize pressure transients when SI pumps are operated
intermittently.
The NEI guidance does not prevent the use of pressurizer heaters, but only serves to note that they
are generally not required.
Comments
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.1.2 (2)Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE[D00-01 R NEI 00-01 GuidanceThe classification of shutdown capability as alternative shutdown is made independent of the
3.1.1.4 [Alternative selection of systems used for shutdown. Alternative shutdown capability is determined based on anShutdown inability to assure the availability of a redundant safe shutdown path. Compliance to the separationCapability] requirements of Sections III.G.1 and III.G.2 may be supplemented by the use of manual actions to the
extent allowed by the regulations and the licensing basis of the plant, repairs (cold shutdown only),exeml~tions, deviations, GL 86-10 fire hazards analyses or fire protectionl design change evaluations,as apl ropriate. These may also be used in conjunction with alternative shutdown capability.
Applicability CommentsApplicable
HBRSEP LAR Rev 1 Page B-10
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Alignment
StatementAlignment Basis
AlignsThe plant's alternate and dedicated safe shutdown systems and strategies were reviewed and
approved in the supplemental SER.
Comments
Reference Document Doc Detail
NLS-85-732, Supplemental Safety Evaluation for Appendix R to 10
CFR 50, Items Ill.G.3 and ll.L; Alternate Safe Shutdown Capability -
H.B. Robinson Steam Plant, Unit 2 - TAC No. 60106
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEID0-1Ref NEI 00-01 Guidance
At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains)are assumed operable and available for post-fire safe shutdown. Systems are assumed to be
Conditions] operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc. in progress.
The units are assumed to be operating at full power under normal conditions and normal lineups.
Applicability Comments
Applicable
Alignment Basis
Statement
This is a basic assumption for all safe shutdown analyses and applies to HBRSEP.Aligns
Comments
Reference Document DoDetail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.1 (2)
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-1 1
Duke EnergyNFPA 805 Section
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805
Requirement
A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE-I00-0Ref NEI 00-01 Guidance
No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident,3.1.1.6 [Other earthquake), single failures or non-fire induced transients need be considered in conjunction with theEvents in fire.Conjunction withFire]
Applicability Comments
Applicable
Alignm entAlg m n BAliganment Basis
StatementThis is a basic assumption for all safe shutdown analyses and applies to HBRSEP.
Aligns
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.1 (6,7,8)Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
3.1.1.7 [ Offsite
Power]
NEI 00-01 Guidance
For the case of redundant shutdown, offsite power may be credited if demonstrated to be free of firedamage. Offsite power should be assumed to remain available for those cases where its availabilitymay adversely impact safety (i.e., reliance cannot be placed on fire causing a loss of offsite power ifthe consequences of offsite power availability are more severe than its presumed loss). No creditshould be taken for a fire causing a loss of offsite power. For areas where train separation cannot be
HBRSEP LAR Rev 1 Page B-12
Duke Energy
Applicability
Applicable
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
achieved and alternative shutdown capability is necessary, shutdown must be demonstrated both
where offsite power is available and where offsite power is not available for 72 hours.
Comments
AlignmentStatement
Aligns
Alignment Basis
For fire areas that use redundant shutdown capabilities offsite power is credited unless the fireimpacts equipment required to support offsite power. If the fire impacts offsite power, at least oneonsite power source is available to provide the required power.
For areas that use alternative / dedicated shutdown, a LOOP is assumed.
In the analysis the LOOP is not credited for preventing or terminating spurious operations orpositioning SSE in its required position. Steps in the procedures insure that the appropriate actionsare taken to line up SSE and deal with potential spurious equipment operations.
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE[O0-1Ref NEI 00-01 Guidance
Post-fire safe shutdown systems and components are not required to be safety-related.3.1.1.8 [Safety-
Related Equipment]
Applicability CommentsApplicable
AlignmentStateme~nt
Alignment Basis
This is a basic assumption for all safe shutdown analyses and applies to HBRSEP.Aligns
HBRSEP LAR Rev 1 Page B-13
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
Comments
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.2Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor3.1.1.9 [72 Hour scram/trip. Fire-induced impacts that provide no adverse consequences to hot shutdown within thisCoping] 72-hour period need not be included in the post-fire safe shutdown analysis. At least one train can
be repaired or made operable within 72 hours using onsite capability to achieve cold shutdown.
Applicability Comments
Applicable
AlignmentAligmentAlignment BasisStatement
NFPA 805 does not require a plant to transition to cold shutdown in the event of a fire. The fire area-Aligns by-fire area assessment documents the method of accomplishment of the NFPA 805 performance
goals, including an optional transition to cold shutdown. For all fires at HBRSEP, the systems andequipment required to place the plant in a safe and stable condition are available following a fireoccurring while the plant is at power without regard to a specific mission time or event copingduration.
Comments
Reference Document DocDetail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 1.5.1 and 1.5.2Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section
HBRSEP LAR Rev 1
2.4.2.1 Nuclear Safety Capability System and Equipment Selection
Page B-14
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review
NFPA 805
Requirement
A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI -0Ref NEI 00-01 Guidance
The following discussion on each of these shutdown functions provides guidance for selecting the3.1.2 Shutdown systems and equipment required for safe shutdown. For additional information on BWR systemFunctions selection, refer to GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths
for the BWR."
Applicability Comments
Applicable
AlignmentAligmentAlignment BasisStatement
This is an introductory section with no specific requirements. The GE information does not apply toAligns HBRSEP.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE[D00-01Ref
3.1.2.1 Reactivity
Control
Applicability
Applicable
HBRSEP LAR Rev 1
NEI 00-01 Guidance
[BWR] Control Rod Drive System
The safe shutdown performance and design requirements for the reactivity control function can be
met without automatic scram/trip capability. Manual scram/reactor trip is credited. The post-fire safe
shutdown analysis must only provide the capability to manually scram/trip the reactor.
[PWR] Makeup/Charging
There must be a method for ensuring that adequate shutdown margin is maintained by ensuring
borated water is utilized for RCS makeup/charging.
Comments
Page B-15
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Aligment Alignment BasisStatement
Reactivity control is provided by insertion of control rods via a reactor trip. Long term reactivityAligns control is provided by boron addition via the charging pumps or safety injection pumps taking suction
from the Reactor Water Storage Tank.
Comments
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.2.2.5
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical componentsrequired to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 0-01Rf NEI 00-01 Guidance
The systems discussed in this section are examples of systems that can be used for pressure control.3.1.2.2 Pressure This does not restrict the use of other systems for this purpose.
Control Systems
[BWR] Safety Relief Valves (SRVs)
The SRVs are opened to maintain hot shutdown conditions or to depressurize the vessel to allow
injection using low pressure systems. These are operated manually. Automatic initiation of the
Automatic Depressurization System is not a required function.
[PWR] Makeup/Charging
RCS pressure is controlled by controlling the rate of charging/makeup to the RCS. Although utilization
of the pressurizer heaters and/or auxiliary spray reduces operator burden, neither component is
required to provide adequate pressure control. Pressure reductions are made by allowing the RCS to
cool/shrink, thus reducing pressurizer level/pressure. Pressure increases are made by initiating
charging/makeup to maintain pressurizer level/pressure. Manual control of the related pumps is
acceptable.
Applicability Comments
Applicable
AlignmentStatemetnt
Alignment Bas
The Reactor Coolant Pressure Control function uses the same components as the RCS Inventory
HBRSEP LAR Rev 1 Page B-16
Duke EnergyAligns
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Control function. RCS Pressure is controlled by controlling the rate of charging to the RCS. Pressure
reductions are made by allowing the RCS to cool/shrink, thus reducing Pressurizer level/pressure.
Pressure increases are made by initiating charging/makeup to maintain Pressurizer level/pressure.
Pressurizer heaters are credited for pressure control when the SI pumps are used for makeup. When
the SI pumps are utilized for makeup, the RCS is de-pressurized to the SI Pump operating pressure
by cycling a Pressurizer PORV.
Comments
Reference Document Doc DetailRNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.2.1.2
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
[BWR] Systems selected for the inventory control function should be capable of supplying sufficient3.1.2.3 Inventory reactor coolant to achieve and maintain hot shutdown. Manual initiation of these systems is
Control acceptable. Automatic initiation functions are not required.
[PWR]: Systems selected for the inventory control function should be capable of maintaining level to
achieve and maintain hot shutdown. Typically, the same components providing inventory control are
capable of providing pressure control. Manual initiation of these systems is acceptable. Automatic
initiation functions are not required.
Applicability CommentsApplicable
AlignmentStatementf
Aligns
Alignment Basis
The RCS Inventory Control function is required to restore and maintain RCS integrity and reactor
coolant makeup capability to compensate for RCS fluid losses (i.e. RCP seal leak-off) and shrinkage
during cooldovn. Reactor Coolant Inventory Control is accomplished by the following actions:
- RCS Isolation - RV Head Vents and Pressurizer PORVs (RCS)
- Normal Letdown Isolation (CVCS)
- Excess Letdown Isolation (CVCS)
- RHR Isolation (RHR)
HBRSEP LAR Rev 1 Page B-17
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review- Charging/makeup from the RWST via the charging pumps or SI pumps. The SI pumps require the
CCW system for cooling. (CVCS, SI, CCW, SW)
- Use of the SI Pumps for RCS Makeup requires RCS depressurization via the Pressurizer PORVs.
- RCP Seal Cooling via Seal Injection and/or Thermal Barrier Cooling (CVCS, CCW, SW)
Comments
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.2.1.2
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
[BWR] Systems selected for the decay heat removal function(s) should be capable of:3.1.2.4 Decay Heat
Removal - Removing sufficient decay heat from primary containment, to prevent containment over-
pressurization and failure.- Satisfying the net positive suction head requirements of any safe shutdown systems taking suction
from the containment (suppression pool).
- Removing sufficient decay heat from the reactor to achieve cold shutdown.
[PWR] Systems selected for the decay heat removal function(s) should be capable of:
- Removing sufficient decay heat from the reactor to reach hot shutdown conditions. Typically, this
entails utilizing natural circulation in lieu of forced circulation via the reactor coolant pumps and
controlling steam release via the Atmospheric Dump valves.
- Removing sufficient decay heat from the reactor to reach cold shutdown conditions.
This does not restrict the use of other systems.
Applicability Comments
Applicable
AlignmentStatement
Alignment Basis
HBRSEP uses the Auxiliary Feedwater System (AFW) and the Main Steam System (MS) to remove
decay heat from the reactor through the steam generators for Hot Shutdown. The Residual HeatAligns
HBRSEP LAR Rev 1 Page B-1 8
Duke EnergyAttachment 8 - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology ReviewRemoval (RHR) System is available at temperatures below 350F and pressures less than 375 psig to
remove decay heat and continue the reactor cooldown to cold shutdown.
Comments
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.1.3 and 2.2.1.4
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEL 01Ref
3.1.2.5 ProcessMonitoring
NEI 00-01 Guidance
The process monitoring function is provided for all safe shutdown paths. IN 84-09, Attachment 1,Section IX "Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems(10CFR50 Appendix R)" provides guidance on the instrumentation acceptable to and preferred by theNRC for meeting the process monitoring function. This instrumentation is that which monitors theprocess variables necessary to perform and control the functions specified in Appendix R SectionIII.L.1. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 listof process monitoring is applied to alternative shutdown (III.G.3). IN 84-09 did not identify specificinstruments for process monitoring to be applied to redundant shutdown (III.G.1 and III.G.2). Ingeneral, process monitoring instruments similar to those listed below are needed to successfully useexisting operating procedures (including Abnormal Operating Procedures).
BWR- Reactor coolant level and pressure- Suppression pool level and temperature
- Emergency or isolation condenser level
- Diagnostic instrumentation for safe shutdown systems- Level indication for tanks needed for safe shutdown
PWR- Reactor coolant temperature (hot leg / cold leg)- Pressurizer pressure and level- Neutron flux monitoring (source range)- Level indication for tanks needed for safe shutdown- Steam generator level and pressure- Diagnostic instrumentation for safe shutdown systems
HBRSEP LAR Rev 1 Page B-19
Duke Energy
ApplicabilityApplicable
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
The specific instruments required may be based on operator preference, safe shutdown procedural
guidance strategy (symptomatic vs. prescriptive), and systems and paths selected for safe shutdown.
Comments
AlignmentStatement
Aligns
Alignment Bas'
The process monitoring function is capable of providing direct reading of those process variablesnecessary for plant operators to perform and/or control identified safe shutdown functions. Plantmonitoring instrumentation, in the context of post-fire safe shutdown operation, consists of thoseminimal instrument channels or local gauges/indicators necessary to monitor the operation of primary
shutdown components and systems, and the operation of those components or systems that providerequired support functions. The parameters to be monitored during post-fire shutdown operations,along with the credited instruments, are summarized on Table 2-2 of RNP-E/ELEC-1216.
Comments
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.1.8, 2.2.2.12, and Table 2-2
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI01 Ref NEI 00-01 Guidance
[Blank Heading - No specific guidance]3.1.2.6 Support
Systems
Notlicbi CommentsNot Applicable
N/A
Comments
Alignment Basis I
Support system requirements will be addressed under the corresponding NEI 00-0 1 sub-section.
HBRSEP LAR Rev 1 Page B-20
Attachment B - NEI 04-02 Table B-2 Nuclear Safetyf'gn.hilifw A . .. m- - f KA~fhn-nlný I D i-WniL, C:-~~r~
-J =.l~l.l -_ L:- , llLy • O O l•lL V•AV•.IrI •
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
VV
NiE[00-01 Ref
3.1.2.6.1 Electrical
Systems
ApplicabilityApplicableI
NEI 00-01 Guidance
AC Distribution System
Power for the Appendix R safe shutdown equipment is typically provided by a medium voltage systemsuch as 4.16 KV Class 1E busses either directly from the busses or through step downtransformers/load centers/distribution panels for 600, 480 or 120 VAC loads. For redundant safeshutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, powermay be supplied from either offsite power sources or the emergency diesel generator depending onwhich has been demonstrated to be free of fire damage. No credit should be taken for a fire causinga loss of offsite power. Refer to Section 3.1.1.7.
DC Distribution System
Typically, the 125VDC distribution system supplies DC control power to various 125VDC controlpanels including switchgear breaker controls. The 125VDC distribution panels may also supply powerto the 120VAC distribution panels via static inverters. These distribution panels typically supplypower for instrumentation necessary to complete the process monitoring functions.For fire events that result in an interruption of power to the AC electrical bus, the station batteries arenecessary to supply any required control power during the interim time period required for the dieselgenerators to become operational. Once the diesels are operational, the 125 VDC distributionsystem can be powered from the diesels through the battery chargers.[BWR] Certain plants are also designed with a 250VDC Distribution System that supplies power toReactor Core Isolation Cooling and/or High Pressure Coolant Injection equipment.The DC control centers may also supply power to various small horsepower Appendix R safeshutdown system valves and pumps. If the DC system is relied upon to support safe shutdownwithout battery chargers being available, it must be verified that sufficient battery capacity exists tosupport the necessary loads for sufficient time (either until power is restored, or the loads are nolonger required to operate).
Comments
AlignmentStatement
Alignment Basis
AlignsThe Electrical Distribution System provides 4160VAC, 480VAC, 120VAC and 125VDC power from
off-site (115KV Grid) and onsite sources (EDGs and DSDG) to safe shutdown electrical loads. The
HBRSEP LAR Rev 1 Page B-21
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewfuel oil systems associated with the onsite power supplies are also included in the analysis.
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.1.7, 2.2.2.11Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 GuidanceHVAC Systems
3.1.2.6.2 CoolingSystems [HVAC] HVAC Systems may be required to assure that safe shutdown equipment remains within its operating
temperature range, as specified in manufacturer's literature or demonstrated by suitable testmethods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxicgases, and gaseous fire suppression agents).HVAC systems may be required to support safe shutdown system operation, based on plant-specific
configurations. Typical uses include:- Main control room, cable spreading room, relay room- ECCS pump compartments
- Diesel generator rooms
- Switchgear rooms
Plant-specific evaluations are necessary to determine which HVAC systems are essential to safe
shutdown equipment operation.
Applicability CommentsApplicable
AlignmentStatementt
ý ligns
Alignment Basi
Plant ventilation systems are required fun tional to provide environmental conditions that supportcontinuous occupancy or safe shutdown dquipment operation in the following area:
- EDG Rooms- Main Control Room- Motor Driven AFW Pump Room
HBRSEP LAR Rev 1 Page B-22
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
- SI Pump Room Capability Assessment Methodology Review
The Main Control Room ventilation system includes refrigerant units that are cooled by the SW
system.
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.1.9, 2.2.2.13, Table 2-3Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 0-01Ref NEI 00-01 GuidanceVarious cooling water systems may be required to support safe shutdown system operation, basedon plant-specific considerations. Typical uses include:
Systems [Main - RHR/SDC/DH Heat Exchanger cooling waterSection] - Safe shutdown pump cooling (seal coolers, oil coolers)
- Diesel generator cooling
- HVAC system cooling water
Applicability Comments
Applicable
AlignmentAlignment BStatement
The Component Cooling Water and Service Water Systems provide cooling to the various safeAligns shutdown loads.
Comments
Reference Document DocDetailRNP-E/ELEC-1216, The Fire Safe Shutdown Analysis fir H.B. Sections 2.2.1.5, 2.2.1.6Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-23
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown3.1.3 Methodology systems and developing the shutdown paths.
for Shutdown
System Selection The following methodology may be used to define the safe shutdown systems and paths for an
Appendix R analysis:
[Refer to hard copy of NEI 00-01 for Figure 3-2]
Applicability Comments
Applicable
Alignment
StatementThis is an introductory statement and provides no requirements. The sub-paragraphs with specific
N/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE[D00-01Ref
3.1.3.1 Identify safe
shutdown functions
NEI 00-01 Guidance'
Review available doumentation to obtain an understanding of the available plant systers and the
functions required to achieve and maintain safe shutdown. Documents such as the following may be
reviewed:
- Operating Procedures (Normal, Emergency, Abnormal)
HBRSEP LAR Rev 1 Page B-24
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review- System descriptions
- Fire Hazard Analysis
- Single-line electrical diagrams
-Piping and Instrumentation Diagrams (P&IDs)
[BWR] GE Report GE-NE-T43-00002-00-01-R02 entitled "Original Shutdown Paths for the BWR"
Applicabllity
Applicable
Comments
Alignment Alignment BasisStatement
Using Operating Procedures (Normal, Emergency, and Abnormal), System Descriptions, Fire Hazard
Aligns Analyses, Piping and Instrument Diagrams (P&IDs) and Single Line Diagrams for each system
comprising the safe shutdown paths, the mechanical or electrical equipment required for the
operation of the system and the equipment whose spurious operation could affect the performance of
the safe shutdown systems were identified.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.1
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.2.6
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI00-01Ref
3.1.3.2 Identify
Combinations of
Systems that
Satisfy Each Safe
Shutdown Function
Applicability
HBRSEP LAR Rev 1
NEI 00-01 GuidanceGiven the criteria/assumptions defined in Section 3.1.1, identify the available combinations ofsystems capable of achieving the safe shutdown functions of reactivity control, pressure control,inventory control, decay heat removal, process monitoring, and support systems such as electricaland coolin systems (refer to Section 3.1.2). This selection process does notlrestrict the use of othersystems. In addition to achieving the required safe shutdown functions, consider spurious operationsand power supply issues that could impact the required safe shutdown function.
Comments
Page B-25
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Applicable
AlignmentStatment
Alignment Basis
AlignsIn accordance with the provisions of 10CFR50, Appendix R, Section lIl.G, at least one means of
achieving and maintaining safe shutdown conditions must remain available in the event of a fire in
any fire area. In developing an appropriate shutdown equipment complement to support this
requirement, it is necessary to categorize equipment into logical train-oriented groupings, identified as
shutdown categories; these categories are further defined as Alternate A and Alternate B. Although,
in many cases, the equipment complement selected corresponds closely to safety-related train
divisions, it should not be construed that Alternate A and B equipment automatically corresponds to
safety-related Train A and B equipment. In general, the Alternate A division constitutes the
equipment credited for safe shutdown outside the control room (i.e. dedicated shutdown) and the
Alternate B division constitutes the equipment credited for control room shutdown scenarios.
Reference Document Doc Detail
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.2.1
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI0-01 Ref NEI 00-01 Guidance
Select combinations of systems with the capability of performing all of the required safe shutdown3.1.3.3 Define functions and designate this set of systems as a safe shutdown path. In many cases, safe shutdown
Combinations of paths may be defined on a divisional basis since the availability of electrical power and other support
Systems for Each systems must be demonstrated for each path.
Safe Shutdown
Path
ApplcabCommentsApplicable
AlmanmentStatement
Algnment BI
AlignsComponents have been grouped into Appendix R fire-safe shutdown systems according to theAppendix R fire-safe shutdown function they support. Each Appendix R fire-safe shutdown system is
HBRSEP LAR Rev 1 Page B-26
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewdivided into one or more "success paths". Each success path represents a functionally independent
method of accomplishing a unique fire-safe shutdown function. Each success path is divided into two
or more redundant Appendix R fire-safe shutdown trains or "success path trains". Each success
path train will often be comprised of components from different plant systems necessary to
accomplish an Appendix R fire-safe shutdown function. These combinations are reflected in the safe
shutdown fault tree developed during the safe shutdown re-validation project.
Reference Document Doc Detail
RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.1.2.1 - 2.1.2.4, 2.2.4.1Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI.0-1Ref NEI 00-01 GuidanceAssign a path designation to each combination of systems. The path will serve to document the
3.1.3.4 Assign combination of systems relied upon for safe shutdown in each fire area. Refer to Attachment 1 to thisShutdown Paths to document (NEI 00-01) for an example of a table illustrating how to document the variousEach Combination combinations of systems for selected shutdown paths.of Systems
Applicabfility Comments
Applicable
AlignmentAlignment BasisStatement
Components have been grouped into Appendix R fire-safe shutdown systems according to theAligns Appendix R fire-safe shutdown function they support. Each Appendix R fire-safe shutdown system is
divided into one or more "success paths". Each success path represents a functionally independentmethod of accomplishing a unique fire-safe shutdown function. Each success path is divided into twoor more redundant Appendix R fire-safe shutdown trains or "success path trains". Each successpath train will often be comprised of components from different plant systems necessary toaccomplish an Appendix R fire-safe shutdown function. These combinations are reflected in the safeshutdown fault tree developed during the safe shutdown re-validation project.
Comments
HBRSEP LAR Rev 1 Page B-27
Duke EnergyReference Document
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.
Robinson Nuclear Plant
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Sections 2.2.2.1 - 2.2.2.4, Table 2-1
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
The previous section described the methodology for selecting the systems and paths necessary to3.2 Safe Shutdown achieve and maintain safe shutdown for an exposure fire event (see Section 5.0 DEFINITIONS for
Equipment "Exposure Fire"). This section describes the criteria/assumptions and selection methodology for
Selection identifying the specific safe shutdown equipment necessary for the systems to perform their Appendix
R function. The selected equipment should be related back to the safe shutdown systems that they
support and be assigned to the same safe shutdown path as that system. The list of safe shutdown
equipment will then form the basis for identifying the cables necessary for the operation or that can
cause the maloperation of the safe shutdown systems.
Applicability Comments
Applicable
Alignment BasisStatement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805Reqluirement
A comprehensive list of systems and equipme t and their interrelationships to be analyzed for a fire
event shall be developed. The equipment list hall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
HBRSEP LAR Rev 1 Page B-28
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
NEI 00-01 Ref NEI 00-01 GuidanceConsider the following criteria and assumptions when identifying equipment necessary to perform the
3.2.1 Criteria / required safe shutdown functions:Assumptions
Applicability Comments
Applicable
Agnment Alignment BasisStatement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
3.2.1.1 [PrimarySecondaryComponents]
NEI 00-01 Guidance
3.2.1.1 Safe shutdown equipment can be divided into two categories. Equipment may be categorizedas (1) primary components or (2) secondary components. Typically, the following types of equipmentare considered to be primary components:- Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc.- All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbinespeed indicator, pressure indicator, level recorder)- Power supplies or other electrical components that support operation of primary components (i.e.,diesel generators, switchgear, motor control centers, load centers, power supplies, distributionpanels, etc.).
Secondary components are typically items found within the circuitry for a primary component. Theseprovide a supporting role to the overall circuit function. Some secondary components may provide anisolation function or a signal to a primary component via either an interlock or input signal processor.Examples of secondary components include flow switches, pressure switches, temperature switches,level switches, temperature elemenis, speed elements, transmitters, converters, controllers,transducers, signal conditioners, hand switches, relays, fuses and various instrumentation devices.
Determine which equipment should be included on the Safe Shutdown Equipment List (SSEL). As anoption, include secondary components with a primary component(s) that would be affected by firedamage to the secondary component. By doing this, the SSEL can be kept to a manageable size
HBRSEP LAR Rev 1 Page B-29
Duke Energy
ApplicabilityApplicable
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
and the equipment included on the SSEL can be readily related to required post-fire safe shutdown
systems and functions.
Comments
Alignment ~ nmm~Alignment Basis
Statement
Components are not identified as primary or secondary. Components providing a "secondary"
Aligns function are either identified as safe shutdown components and included in the safe shutdown
equipment list, or have their applicable cables assigned to the primary component.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEL 00-01Ref NEI 00-01 Guidance
3.2.1.2 Assume that exposure fire damage to manual valves and piping does not adversely impact3.2.1.2 [Fire their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping
Damage to materials, including tubing with brazed or soldered joints, are not included in this assumption). Fire
Mechanical damage should be evaluated with respect to the ability to manually open or close the valve should
Components (not this be necessary as a part of the post-fire safe shutdown scenario.
electrically
supervised)]
Applicability Comments
Applicable
AlegnmentStatement
Aligns
Alignment Basis
Due to the substantial nature of equipment and the nature and location of combustibles, fire will not
not impact the pressure boundary function. A fire does not cause a manual valve to change itsposition, Manual stroking of a valve once the fire is extinguished is evaluated as part of the manualaction feasibility study.
HBRSEP LAR Rev 1 Page B-30
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 9.1.3, 9.4.1
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
N.EI00-01 Ref NEI 00-01 Guidance
Assume that manual valves are in their normal position as shown on P&lDs or in the plant operating3.2.1.3 [Manual procedures.
Valve Positions]
Applicability Comments
Applicable
Alignment Basis
Statement
This is a basic assumption for all safe shutdown analyses and applies to HBRSEP.Aligns
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.1
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.2
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805
Requirement
A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireevent shall be developed. The equipment list shall contain an inventory of those critical components
HBRSEP LAR Rev 1 Page B-31
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewrequired to achieve the nuclear safety performance criteria of Section 1.5T Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
Assume that a check valve closes in the direction of potential flow diversion and seats properly with3.2.1.4 [Check sufficient leak tightness to prevent flow diversion. Therefore, check valves do not adversely affect the
Valves] flow rate capability of the safe shutdown systems being used for inventory control, decay heat
removal, equipment cooling or other related safe shutdown functions.
Applicability Comments
Applicable
AIflgnment Alignment Basis
Statement
FIR-NGGC-001 identifies that properly oriented check valves credited as system boundaries shouldAligns be included in the SSEL, and that those in the flow path need not be included. Check Valves
credited as boundaries are included in the SSEL, but the assumption that they are leak tight is
inherent in the analysis.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.1.3
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref
3.2.1.5 [InstrumentFailures]
Applicability
Applicable
HBRSEP LAR Rev 1
NEI 00-01 Guidance
Instruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flowtransmitters) are assumed to fail upscale, midscale, or downscale as a result of fire d mage,whichever is worse. An instrument performing a control function is assumed to provide an undesiredsignal to the control circuit.
Comments
Page B-32
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Canability Assessment Methodology Review
Alignment
Statement
Instruments exposed to the fire are assumed to fail. It is a generally accepted practice (that can beAligns verified based on a review of the fire area analysis) that instruments are assumed to fail to their worst
case position.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 9.3.2 and 9.4.1
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEL 00-0 ef NEI 00-01 Guidance
Identify equipment that could spuriously operate or mal-operate and impact the performance of3.2.1.6 [Spurious equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1
Components] of RIS 2004-03 during the equipment identification process.
Applicability Comments
Applicable
AlignmentAlignment Basis
Statement
FIR-NGGC-0101 states, "Electrically operated or controlled valves or dampers in the flow pathsAligns whose spurious operation could adversely affect system operation shall be included on the SSEL."
This is affirmed in the Section 2.1.2.5 of RNP-E/ELEC-1216.
RIS 2004-03 was a reference for the procedures used in the safe shutdown re-validation, and is also
referenced in FIR-NGGC-0101.
Comments I
Reference Document
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability
Assessment (NSCA)
DocDetailSection 2.1, 9.1.3
HBRSEP LAR Rev 1 Page B-33
Duke EnerqyRNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.Robinson Nuclear Plant
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Sections 2.1.2.5, _3.
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI00-1Ref NEI 00-01 Guidance
Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a3.2.1.7 [Instrument result of fire. Determine and consider the fire area location of the instrument tubing when evaluating
Tubing] the effects of fire damage to circuits and equipment in the fire area.
Applicability Comments
Applicable
Alignment Alignment B
Statement
FIR-NGGC-0101 provides direction for evaluating the fire effects on instrument tubing and theAligns potential impact on spurious operation. FSSPMD documents tubing routing to ensure the impact of
this issue is evaluated.
Comments
Reference Document DoclDtail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.1.7
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.2.7
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805Requirement
A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear
HBRSEP LAR Rev 1 Page B-34
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEL00-01Ref
3.2.2 Methodologyfor EquipmentSelection
Applicability
Applicable
NEI 00-01 Guidance
Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdownequipment.
Use the following methodology to select the safe shutdown equipment for a post-fire safe shutdownanalysis:
[Refer to hard copy of NEI 00-01 for Figure 3-3]
Comments
Alignment Alignment BassStatement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Guidance
Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each3.2.2.1 Identify' the shutdown path. Refer to Attachment 2 for an example of an annotated P&ID illustrating this concept.System Flow Pathfor Each ShutdownPath
Applicability Comments
Applicable
AlignmentStatement
Alignment Basis.
The safe shutdown flow paths at Robinson are depicted on the HBR2-11390 Series of drawings.Aligns
HBRSEP LAR Rev 1 Page B-35
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.1.3
Assessment (NSCA)
HBR2-11390, Appendix R and Station Blackout Safe-Shutdown
Analysis Flowpath/Boundary Diagrams
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI•00-01ReJ f NEI 00-01 GuidanceReview the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to
3.2.2.2 Identify the assure that all equipment in each system's flow path has been identified. Assure that any equipment
Equipment in Each that could spuriously operate and adversely affect the desired system function(s) is also identified. If
Safe Shutdown additional systems are identified which are necessary for the operation of the safe shutdown system
System Flow Path under review, include these as systems required for safe shutdown. Designate these new systems
Including with the same safe shutdown path as the primary safe shutdown system under review (Refer to
Equipment That Figure 3-1).
May Spuriously
Operate and Affect
System Operation
Applicability Comments
Applicable
Alignment Alignment Basis
Statement
Using Operating Procedures (Normal, Emergency, and Abnormal), System Descriptions, Fire HazardAligns Analyses, Piping and Instrument Diagrams (P&IDs) and Single Line Diagrams for each system
comprising the safe shutdown paths, the mechanical or electrical equipment required for the
operation of the system and the equipment whose spurious operation could affect the performance of
the safe shutdown systems were identified.
Reference Document
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability
Assessment (NSCA)
Doc Detali
Section 9.1.3
HBRSEP LAR Rev 1 Page B-36
Duke EnergyRNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.
Robinson Nuclear Plant
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Section 2.1.2.5
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 0-01Ref NEI 00-01 GuidancePrepare a table listing the equipment identified for each system and the shutdown path that it
3.2.2.3 Develop a supports. Identify any valves or other equipment that could spuriously operate and impact the
List of Safe operation of that safe shutdown system. Assign the safe shutdown path for the affected system to
Shutdown this equipment. During the cable selection phase, identify additional equipment required to support
Equipment and the safe shutdown function of the path (e.g., electrical distribution system equipment). Include this
Assign the additional equipment in the safe shutdown equipment list. Attachment 3 to this document provides an
Corresponding example of a (SSEL). The SSEL identifies the list of equipment within the plant considered for safe
System and Safe shutdown and it documents various equipment-related attributes used in the analysis.
Shutdown Path(s)
Designation to
Each.
Applicability Comments
Applicable
Aignment BasisStatement
System and component identification is discussed in section 2.2.2, and section 2.2.3 refers to theAligns SSEL maintained in FSSPMD. Attachment 24 of is a printout of the SSEL. This includes valves and
pumps whose spurious operation may impact a safe shutdown system from performing its function.
Information in FSSPMD includes the component's power supply, fire zone location, normal and
required positions, required cables, and associated circuits.
The components and the safe shutdown function(s) they support are depicted in the safe shutdown
fault tree. Development of the fault tree is described in FIR-NGGC-0101, revision 2.
Comments
Reference Document
FIR-NGGC-0101, Fire Protection Nuclear Safety CapabilityAssessment (NSCA)
Doc DetailSection 9.2
HBRSEP LAR Rev 1 Page B-37
Duke EneryFSSPMf, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.
Robinson Nuclear Plant
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Sections 2.2.2, 2.2.3, and Att. 24
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NE IRef NEI 00-01 Guidance
Collect additional equipment-related information necessary for performing the post-fire safe shutdown3.2.2.4 Identify analysis for the equipment. In order to facilitate the analysis, tabulate this data for each piece of
Equipment equipment on the SSEL. Refer to Attachment 3 to this document for an example of a SSEL.
Information Examples of related equipment data should include the equipment type, equipment description, safe
Required for the shutdown system, safe shutdown path, drawing reference, fire area, fire zone, and room location of
Safe Shutdown equipment. Other information such as the following may be useful in performing the safe shutdown
Analysis analysis: normal position, hot shutdown position, cold shutdown position, failed air position, failed
electrical position, high/low pressure interface concern, and spurious operation concern.
Applicability Comments
Applicable
Alignment Alignment BasisStatement
The information identified as needed for performing safe shutdown analysis on the componentsAligns identified on the SSEL is contained in the FSSPMD. This can be verified on a component basis
through reports that can be generated through the FSSPMD.
Information in FSSPMD includes the component's power supply, fire zone location, normal and
required positions, required cables, and associated circuits.
Comments
Reference Document Doc Detail
FSSPMD, Fire Safe Shutdown Program Mana er Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Inalysis for H.B. Attachment 24
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-38
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection
NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire
Requirement event shall be developed. The equipment list shall contain an inventory of those critical components
required to achieve the nuclear safety performance criteria of Section 1.5. Components required to
achieve and maintain the nuclear safety functions and components whose fire-induced failure could
prevent the operation or result in the maloperation of those components needed to meet the nuclear
safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI00-01Ref NEI 00-01 Guidance
In the process of defining equipment and cables for safe shutdown, identify additional supporting3.2.2.5 Identify equipment such as electrical power and interlocked equipment. As an aid in assessing identified
Dependencies impacts to safe shutdown, consider modeling the dependency between equipment within each safe
Between shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram
Equipment, (SSLD). Attachment 4 provides an example of a SSLD that may be developed to document these
Supporting relationships.
Equipment, Safe
Shutdown Systems
and Safe Shutdown
Paths.
Applicability Comments
Applicable
Alignment
Statement
Power supplies are identified and documented in the FSSPMD. Cables that are associated with aAligns component because of interlocks or permissive are documented with the component in the FSSPMD.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2
Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805Requirement
2.4.2.2.1 Cilcuits Required in Nuclear Safety Functions. Circuits required for th_ nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
Page B-39HBRSEP LAR Rev 1
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology ReviewThis will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01Ref
3.3 Safe Shutdown
Cable Selection and
Location
NEI 00-01 Guidance
This section provides industry guidance on the recommended methodology and criteria for selecting
safe shutdown cables and determining their potential impact on equipment required for achieving and
maintaining safe shutdown of an operating nuclear power plant for the condition of an exposure fire.
The Appendix R safe shutdown cable selection criteria are developed to ensure that all cables that
could affect the proper operation or that could cause the maloperation of safe shutdown equipment
are identified and that these cables are properly related to the safe shutdown equipment whose
functionality they could affect. Through this cable-to-equipment relationship, cables become part of
the safe shutdown path assigned to the equipment affected by the cable.
Applicability Comments
Applicable
Alignment Alignment Basis
Statement
This is an introductory statement and provides no requirements. The sub-paragraphs with specific
N/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
tl e operation, or that result in the maloperation of the equipment ide tified in 2.4.2.1. This evaluation
sl}all consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
HBRSEP LAR Rev 1 Page B-40
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewtheir impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref NEI 00-01 Guidance
To identify an impact to safe shutdown equipment based on cable routing, the equipment must have3.3.1 Criteria /cables that affect it identified. Carefully consider how cables are related to safe shutdown equipment
Assumptions so that impacts from these cables can be properly assessed in terms of their ultimate impact on safe
shutdown system equipment.
Consider the following criteria when selecting cables that impact safe shutdown equipment:
Applicability Comments
Applicable
AlignmentAlnment Basis
Statement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
encl'sure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
theirlimpact on the ability to achieve nuclear safety performance criteria!
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
HBRSEP LAR Rev 1 Page B-41
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
coordinated with the downstream protection device. Capability Assessment Methodology Review
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref
3.3.1.1 [Cable
Selection]
ApplicabilityApplicable
NEI 00-01 Guidance
The list of cables whose failure could impact the operation of a piece of safe shutdown equipment
includes more than those cables connected to the equipment. The relationship between cable and
affected equipment is based on a review of the electrical or elementary wiring diagrams. To assure
that all cables that could affect the operation of the safe shutdown equipment are identified,
investigate the power, control, instrumentation, interlock, and equipment status indication cables
related to the equipment. Consider reviewing additional schematic diagrams to identify additional
cables for interlocked circuits that also need to be considered for their impact on the ability of the
equipment to operate as required in support of post-fire safe shutdown. As an option, consider
applying the screening criteria from Section 3.5 as a part of this section. For an example of this see
Section 3.3.1.4.
Comments
Alignment
Statement
FIR-NGGC-01 01 Section 9.3.1 provides direction for assigning cables to components. This processAligns is documented in Section 3.0 of RNP-E/ELEC-1216 and in the FSSPMD.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2
Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.0
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are recuired for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
HBRSEP LAR Rev 1 Page B-42
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 0-1Rf NEI 00-01 Guidance
In cases where the failure (including spurious actuations) of a single cable could impact more thanone piece of safe shutdown equipment, include the cable with each piece of safe shutdown
Affecting Multiple equipment.
Components]
Applicability Comments
Applicable
AigmnAlignment BasisStatement
Circuit analysis is performed independently on individual components, so cables affecting more thanAligns one component will be identified with each applicable component.
Comments
Reference Document DocDDtail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3
Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.0
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
INFPA 895Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Funclions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
HBRSEP LAR Rev 1 Page B-43
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology ReviewThis will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01Ref NEI 00-01 Guidance
Electrical devices such as relays, switches and signal resistor units are considered to be acceptable3.3.1.3 [Isolation isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in
Devices] the loop to determine that an acceptable isolation device has been installed at each point where the
loop must be isolated so that a fault would not impact the performance of the safe shutdown
instrument function.
Applicability Comments
Applicable
Aliginment Alignment Basis
Statement
Isolation devices are defined in FIR-NGGC-0101, Section 3.Aligns
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 3, Item 43.
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 1.3.2
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
HBRSEP LAR Rev 1 Page B-44
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref NEI 00-01 GuidanceScreen out cables for circuits that do not impact the safe shutdown function of a component (i.e.,
3.3.1.4 [Identify annunciator circuits, space heater circuits and computer input circuits) unless some reliance on these"Not Required" circuits is necessary. However, they must be isolated from the component's control scheme in such aCables] way that a cable fault would not impact the performance of the circuit.
Applicability Comments
Applicable
Alignment Alignment BasisStatement
In FSSPMD cables that are not required for safe shutdown have an "A" or an "NA" entered in theAligns FMEA section of the circuit information form in FSSPMD. The "A" indicates that the component
"achieves" its safe shutdown function even if that cable is damaged by fire. The "NA" indicates that
the cable is not part of a safe shutdown circuit.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2 and Attachment 1Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capa ility Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. CircUits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
HBRSEP LAR Rev 1 Page B--45
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewshorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achievethe nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref
3.3.1.5
[Identification of
Power Supplies]
Applicability
Applicable
NEI 00-01 Guidance
For each circuit requiring power to perform its safe shutdown function, identify the cable supplyingpower to each safe shutdown and/or required interlock component. Initially, identify only the powercables from the immediate upstream power source for these interlocked circuits and components(i.e., the closest power supply, load center or motor control center). Review further the electricaldistribution system to capture the remaining equipment from the electrical power distribution systemnecessary to support delivery of power from either the offsite power source or the emergency dieselgenerators (i.e., onsite power source) to the safe shutdown equipment. Add this equipment to thesafe shutdown equipment list. Evaluate the power cables for this additional equipment for associatedcircuits concerns.
Comments
AlignmentStatement
Aligns
Alignment Bas*
The power cables for individual components are listed in the circuit analysis for that component ifpower is needed for the component to perform its safe shutdown function. Power supplies are linkedto their components in FSSPMD in the "Power Supplies, Related, Auxiliary, and Other ImportantCircuits" portion of the Circuit Information Form. A standard note "A" entered for a power supply inthis section indicates that the power supply is required for the component to perform its safeshutdown function. The power supply requirement is modeled in the fault tree.
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear S fety Capability Section 9.3.2 and Attachment 1
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.1
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-46
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01Ref
3.3.1.6 [ESFAS
Initiation]
Applicability
Applicable
NEI 00-01 Guidance
The automatic initiation logics for the credited post-fire safe shutdown systems are not required to
support safe shutdown. Each system can be controlled manually by operator actuation in the main
control room or emergency control station. If operator actions outside the MCR are necessary, those
actions must conform to the regulatory requirements on manual actions. However, if not protected
from the effects of fire, the fire-induced failure of automatic initiation logic circuits must not adversely
affect any post-fire safe shutdown system function.
Comments
AlignmentStaement
Aligns
Alignment Basis
Reliance on the automatic logic for safe shutdown systems is not credited at HBRSEP. Althoughautomatic ESFAS signals are not credited, they have been included in the SSA to assure that anactuation of the logic does not cause any adverse consequences.
The Fast Bus Transfer scheme and EDG Auto Sequencing fault trees have been modeled so thatthey maA be credited when available. I
Comments
Reference Document Doc Detail
HBRSEP LAR Rev 1 Page B-47
FR-NGC 0101, Fire Protection Nuclear Safety Capability
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.Robinson Nuclear Plant
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Sections 9.2.3, 9.3.1, 9.3.2, and 9.3.7
Sections 2.1.3.2 and 2.2.4.4
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref
3.3.1.7 [Circuit
Coordination]
Applicability
Applicable
NEI 00-01 GuidanceCabling for the electrical distribution system is a concern for those breakers that feed associatedcircuits and are not fully coordinated with upstream breakers. With respect to electrical distributioncabling, two types of cable associations exist. For safe shutdown considerations, the direct powerfeed to a primary safe shutdown component is associated with the primary component. For example,the power feed to a pump is necessary to support the pump. Similarly, the power feed from the loadcenter to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordinationis not provided, the same cables discussed above would also support the power supply. Forexample, the power feed to the pump discussed above would support the bus from which it is fedbecause, for the case of a common power source analysis, the concern is the loss of the upstreampower source and not the connected load. Similarly, the cable feeding the MCC from the load centerwould also be necessary to support the load center.
Comments
AlignmentStatement Alignment Basis
HBRSEP LAR Rev 1 Page B-48
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
The guidelines that will be used in the evaluation of the common power supplies are as follows (Ref.Aligns FIR-NGGC-0101):
- Using the single-line drawings, a list of the safe shutdown power supplies to be reviewed for
electrical coordination will be developed.
- For each safe shutdown power supply, the existing short circuit calculations, load studies,
coordination calculations, protective device setting sheets, and time current curves as appropriate to
confirm proper coordination between upstream and downstream protective devices will be reviewed.
- In reviewing coordination, electrical system line-ups credited in the safe shutdown analysis will be
considered.
- For cases in which coordination between series protective devices cannot be demonstrated, a
common power supply associated circuit will be assumed to exist. These circuits will be dispositioned
by one of the following means:
1) Demonstrate coordination by refining the available short circuit current and/or trip device
characteristics.
2) Demonstrate that the lack of coordination does not adversely affect safe shutdown (e.g.,
equipment located in same fire area as power supply).
3) Identify readily achievable protective device setting changes (including changes in fuse size and/or
clearing characteristics) that will establish coordination.
4) Incorporate the Associated Circuits and Cables into the safe shutdown analysis as 1OCFR50
Appendix R safe shutdown when protection devices do not provide the desired coordination.
5) Existing short circuit and coordination calculations will be updated as necessary to fully document
where coordination is credited for 10CFR50 Appendix R safe shutdown.
The loss of control power or fire induced cable damage to a 4KV breaker's trip circuit prior to a fault
on the same breaker's load cable would prevent the breaker from tripping on over-current and could
result in loss of the 4KV power supply (i.e. switchgear). Therefore, all non-safe shutdown 4KV
breakers that are associated with a safe shutdown 4KV power supply will be included in the 4KV
power supply circuit analysis as Associated Circuits and Cables.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.6
Assessment (NSCA)
FPP-RNP-200, 10CFR50, Appendix R, Section Ill.G, Associated
Circuits Analysis
RNP-E-8.005, 10CFR50 Appendix R Associated Circuit, Common
Power Supply Analysis
RNP-E-8.053, Non-Safety Overcurrent Protection Coordination
RNP-E-9.021, 10CFR50 Appendix R Fuse Analysis for DS Bus
RNP-E/ELEY-1216, The Fire Safe Shutdown Analysis for H.B. Secti n 3.1.2.1
Robinson N ýclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-49
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
3.3.2 Associated
Circuit Cables
Applicability
Applicable
NEI 00-01 Guidance
Appendix R, Section III.G.2, requires that separation features be provided for equipment and cables,including associated non-safety circuits that could prevent operation or cause maloperation due tohot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achievehot shutdown. The three types of associated circuits were identified in Reference 6.1.5 and furtherclarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhut, Reference6.1.6. They are as follows:- Spurious actuations- Common power source- Common enclosure.
Comments
AlignmentAligmentAlignment BasisStatement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Ta le B-2 Nuclear Safety Capability Assessment Methodology Revie
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
HBRSEP LAR Rev I Page B-50
Duke Energy
NFPA 805
Requirement
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properlycoordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside ofthe immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
3.3.2 [A] Associated
Circuit Cables -
Cables Whose
Failure May Cause
Spurious Actuations
NEI 00-01 Guidance
Safe shutdown system spurious actuation concerns can result from fire damage to a cable whosefailure could cause the spurious actuation/mal-operation of equipment whose operation could affectsafe shutdown. These cables are identified in Section 3.3.3 together with the remaining safeshutdown cables required to support control and operation of the equipment.
Applicability Comments
Applicable
AligmentAlignment BasisStatement
Cables that can cause an undesired spurious actuation are identified by an "S" in the FMEA code of
Aligns the circuit information form in FSSPMD. They are evaluated in the SSA in the same manner as"required" cables. RNP-E/ELEC-1216 evaluates throughout for spurious operation of valves, pumps,
and breakers.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.0, 9.1.3, 9.3.2, and Attachment 1
Asse sment (NSCA)
FSS MD, Fire Safe Shutdown Program Manager Database
RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.1.2.3
Robinson Nuclear Plant
HBRSEP LAR Rev I Page B-51
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke EnergNTable 0- Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref
3.3.2 [B] Associated
Circuit Cables -
Common Power
Source Cables
Applicability
Applicable
NEI 00-01 Guidance
The concern for the common power source associated circuits is the loss of a safe shutdown powersource due to inadequate breaker/fuse coordination. In the case of a fire-induced cable failure on anon-safe shutdown load circuit supplied from the safe shutdown power source, a lack of coordinationbetween the upstream supply breaker/fuse feeding the safe shutdown power source and the loadbreaker/fuse supplying the non-safe shutdown faulted circuit can result in loss of the safe shutdownbus. This would result in the loss of power to the safe shutdown equipment supplied from that powersource preventing the safe shutdown equipment from performing its required safe shutdown function.Identify these cables together with the remaining safe shutdown cables required to support control
and operation of the equipment. Refer to Section 3.5.2.4 for an acceptable methodology foranalyzing the impact of these cables on post-fire safe shutdown.
Comments
AlignmentStatement
Aligns
Alignment Basis
The guidelines that will be used in th• evaluation of the common power supplies are as follows (Ref.
FIR-NGGC-0101): I
- Using the single-line drawings, a list of the safe shutdown power supplies to be reviewed for
electrical coordination will be developed.
- For each safe shutdown power supply, the existing short circuit calculations, load studies,
HBRSEP LAR Rev 1 Page B-52
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewcoordination calculations, protective device setting sheets, and time current curves as appropriate to
confirm proper coordination between upstream and downstream protective devices will be reviewed.
- In reviewing coordination, electrical system line-ups credited in the safe shutdown analysis will be
considered.
- For cases in which coordination between series protective devices cannot be demonstrated, a
common power supply associated circuit will be assumed to exist. These circuits will be dispositioned
by one of the following means:
1) Demonstrate coordination by refining the available short circuit current and/or trip device
characteristics.
2) Demonstrate that the lack of coordination does not adversely affect safe shutdown (e.g.,
equipment located in same fire area as power supply).
3) Identify readily achievable protective device setting changes (including changes in fuse size and/or
clearing characteristics) that will establish coordination.
4) Incorporate the Associated Circuits and Cables into the safe shutdown analysis as 1OCFR50Appendix R safe shutdown when protection devices do not provide the desired coordination.
5) Existing short circuit and coordination calculations will be updated as necessary to fully document
where coordination is credited for 10CFR50 Appendix R safe shutdown.
The loss of control power or fire induced cable damage to a 4KV breaker's trip circuit prior to a fault
on the same breaker's load cable would prevent the breaker from tripping on over-current and could
result in loss of the 4KV power supply (i.e. switchgear). Therefore, all non-safe shutdown 4KV
breakers that are associated with a safe shutdown 4KV power supply will be included in the 4KV
power supply circuit analysis as Associated Circuits and Cables.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.6
Assessment (NSCA)
RNP-E-8.005, 10CFR50 Appendix R Associated Circuit, Common
Power Supply Analysis
RNP-E-8.053, Non-Safety Overcurrent Protection Coordination
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.1.2.1
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified . This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evlluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
HBRSEP LAR Rev 1 Page B-53
Attachment B - NEt 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
NE[I•000IRef NEI 00-01 Guidance
The concern with common enclosure associated circuits is fire damage to a cable whose failure could3.3.2 [C] Associated propagate to other safe shutdown cables in the same enclosure either because the circuit is not
Circuit Cables - properly protected by an isolation device (breaker/fuse) such that a fire-induced fault could result in
Common Enclosure ignition along its length, or by the fire propagating along the cable and into an adjacent fire area. This
Cables fire spread to an adjacent fire area could impact safe shutdown equipment in that fire area, thereby
resulting in a condition that exceeds the criteria and assumptions of this methodology (i.e., multiple
fires). Refer to Section 3.5.2.5 for an acceptable methodology for analyzing the impact of these
cables on post-fire safe shutdown.
Applicability Comments
Applicable
AlignmentAliganment Basis
Statement
The following guidelines were used in the evaluation of common enclosure associated circuits (Ref.Aligns FIR-NGGC-0101):
- Perform an evaluation of the common enclosure associated circuits by reviewing design and
installation criteria for cable and electrical penetrations. Confirm that cables are adequately protectedagainst short circuits and will not propagate a fire from one fire area to another. In evaluating
common power supply circuits the acceptance criteria shall not be limited to standard cable damage
temperatures, which are based on not degrading cable insulation (typically 2500 C for thermoset
cable). Rather, the criteria will be based on not exceeding temperatures at which self ignition or
damage to surrounding cables could occur.
- If a common enclosure associated circuit is determined to exist, the concern shall be resolved by
one of the following means:
1) Demonstrate by analysis that the cable does not pose a risk to cables within the common
enclosure under fault conditions (i.e., the cable exceeds its recommended tempera ure rise but does
not represent a hazard to surrounding cables),2) Demonstrate that the lack of fault protection does not adversely affect safe shutdown,
3) Identify readily achievable protective device setting changes (including changes in fuse size and/or
time-current characteristics) that will establish cable protection without affecting other performance
requirements, or
HBRSEP LAR Rev 1 Page B-54
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review4) Incorporate the cables of concern into the safe shutdown analysis as post-fire safe shutdown
cables for the affected power supply.
5) Existing short circuit and electrical protection calculations will be updated as necessary.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.6
Assessment (NSCA)
FPP-RNP-200, 10CFR50, Appendix R, Section Ill.G, Associated Section 4.0
Circuits Analysis
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.1.2.2
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
N-EL00-01IRef
3.3.3 Methodologyfor Cable Selectionand Location
NEI 00-01 Guidance
Refer to Figure 3-4 for a flowchart illustrating the various steps involved i selecting the cablesnecessary for performing a post-fire safe shutdown analysis.Use the following methodology to define the cables required for safe shutdown including cables thatmay cause associated circuits concerns for a post-fire safe shutdown analysis:
[Refer to hard copy of NEI 00-01 for Figure 3-4]
HBRSEP LAR Rev 1 Page B-55
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology ReviewApplicability Comments
Applicable
Alignment Basis
Statement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref
3.3.3.1 IdentifyCircuits Requiredfor the Operation ofthe Safe Shutdown
Equipment
NEI 00-01 Guidance
For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical
diagrams including the following documentation to identify the circuits (power, control,
instrumentation) required for operation or whose failure may impact the operation of each piece of
equipment:
- Single-line electrical diagrams
- Elementary wiring diagrams
- Electrical connection diagrams
- Instrument loop diagrams.
For electrical power distribution equipment such as power supplies, identify any circuits whose failure
may cause a coordination concern for the bus under evaluation.
If power is required for the equipment, include the closest upstream power distribution source on the
safe shutdown equipment list. Through the iterative process described in Figures 3-2 and 3-3,
HBRSEP LAR Rev 1 Page B-56
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewinclude the additional upstream power sources up to either the offsite or the emergency power
source.
Applicability Comments
Applicable
Alignment BasisStatement
FIR-NGGC-01 01 Section 9.3.2 provides direction for assigning cables to components. This processAligns is further documented in Section 3.0 of RNP-E/ELEC-1216 and in the FSSPMD.
Comments
Reference Document Doc Detail
FIR-NGGC-01.01, Fire Protection Nuclear Safety Capability Section 9.3.2Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.0Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safetyRequirement functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, andshorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.This will ensure that a comprehensive population of circuitry is evaluated.2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properlycoordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achievethe nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concerh is that the effects of a fire can extend outside ofthe immediate fire area due to fire-induced electrica/faults on inadequately protected cables or viainadequately sealed fire area boundaries.
NE[ 00-0 ef
HBRSEP LAR Rev 1
NEI 00-01 Guidance
In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes,
Page B-57
Duke Energy3.3•.3.2 Identify
Interlocked Circuits
and Cables Whose
Spurious Operation
or Mal-operation
Could Affect
Shutdown
ApplicabiletyApplicable
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
cables and equipment. Assign to the equipment any cables for interlocked circuits that can affect the
equipment.
While investigating the interlocked circuits, additional equipment or power sources may be
discovered. Include these interlocked equipment or power sources in the safe shutdown equipment
list (refer to Figure 3-3) if they can impact the operation of the equipment under consideration.
Comments
Alignment
Statement
Aligns with intent
Alignment Basis
As an alternative to adding the interlocked equipment to the SSEL, it is acceptable to include the
cables that are required for the interlocking function (or that could cause the spurious actuation) with
the main component that was originally under consideration. Adding them to the components may
ease the development of a suitable mitigating strategy in areas where the interlocked cables may be
damaged by the fire. Interlocked circuits were either included in the analysis, or the interlocked
contact or relay was assumed to be in its worst-case position. Associated circuits identified for each
component are either included in the main circuit analysis, or are included by listing the applicable
circuit in the "Power Supplies, Related, Auxiliary, and Other Important Circuits" on the Circuit
Information Form.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2
Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other cýircuits that share common power supply and/or common
enclosure with circuits required to achiev• nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
HBRSEP LAR Rev 1 Page B-58
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewcoordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside ofthe immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
NEI 00-01 Ref
3.3.3.3 Assign
Cables to the Safe
Shutdown
Equipment
NEI 00-01 Guidance
Given the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that
may result in maloperation of each piece of safe shutdown equipment.
Tabulate the list of cables potentially affecting each piece of equipment in a relational database
including the respective drawing numbers, their revision and any interlocks that are investigated to
determine their impact on the operation of the equipment. In certain cases, the same cable may
support multiple pieces of equipment. Relate the cables to each piece of equipment, but not
necessarily to each supporting secondary component.
If adequate coordination does not exist for a particular circuit, relate the power cable to the power
source. This will ensure that the power source is identified as affected equipment in the fire areas
where the cable may be damaged.
Applicability Comments
Applicable
AinntAlignment Basis
Statement
The circuit analysis results for each electrically operated safe shutdown component are contained inAligns the FSSPMD. FSSPMD contains various forms and reports for presenting SSD cables and
associated circuits. Refer to Progress Energy procedure FIR-NGGC-0101 for a description of the
circuit analysis nomenclature used in the FSSPMD.
Comments
Reference Document Doc DetaflFIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.2.1.1
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.
NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be
Requirement identified.
NEIBR -L0 Revf
HBRSEP LAR Rev 1
NEI 00-01 Guidance
Identify the routing for each cable including all raceway and cable endpoints. Typically, this
Page B-59
Duke Energ3.3.3.4 Iycentify
Routing of Cables
Applicability
Applicable
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
information is obtained from joining the list of safe shutdown cables with an existing cable and
raceway database
Comments
igmnt AlignmentB
Statement
Equipment location and Cable raceway routing information (i.e. cable raceway to fire area/zone
Aligns correlation) was migrated in 2004 from the safe shutdown analysis of record at the time (FPP-RNP-
150, Revision 7A) to the FSSPMD. Additional cables were added to FSSPMD based on revised
component selection.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.9
Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 4.0
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.
NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be
Requirement identified.
NEI 00-01 Ref NEI 00-01 GuidanceIdentify the fire area location of each raceway and cable endpoint identified in the previous step and
3.3.3.5 Identify join this information with the cable routing data. In addition, identify the location of field-routed cable
Location of by fire area. This produces a database containing all of the cables requiring fire area analysis, their
Raceway and locations by fire area, and their raceway.
Cables by Fire Area
Applicability Comments
Applicable
AlignmentAlignment Basis
Cable to raceway infbrmation is contained in the Cable Information Forms of the FSSPID. RacewayAligns and endpoint locations for all required cables are also contained in FSSPMD.
Comments
HBRSEP LAR Rev 1 Page B-60
Duke EnergyReference Document
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability
Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.
Robinson Nuclear Plant
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
DoctDeti9l
Section 9.3.9
Section 4.0
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
0-0 NEI 00-01 Guidance
By determining the location of each component and cable by fire area and using the cable to3.4 Fire Area equipment relationships described above, the affected safe shutdown equipment in each fire area
Assessment and can be determined. Using the list of affected equipment in each fire area, the impacts to safe
Compliance shutdown systems, paths and functions can be determined. Based on an assessment of the number
Assessment and types of these impacts, the required safe shutdown path for each fire area can be determined.
The specific impacts to the selected safe shutdown path can be evaluated using the circuit analysis
and evaluation criteria contained in Section 3.5 of this document.
Having identified all impacts to the required safe shutdown path in a particular fire area, this section
provides guidance on the techniques available for individually mitigating the effects of each of the
potential impacts.
Applocabfflet Comments
Applicable
Alignment Basis
Statement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805
Requirement
HBRSEP LAR Rev 1
Fire Area Assessment. An engineering analysis shall be performed in accordance with the
requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
Page B-61
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review
4 for methods of achieving these performance criteria (performance-based or deterministic).
NE[D00-01Ref NEI 00-01 Guidance
The following criteria and assumptions apply when performing fire area compliance assessment to3.4.1 Criteria / mitigate the consequences of the circuit failures identified in the previous sections for the requiredAssumptions safe shutdown path in each fire area.
Applicability Comments
Applicable
Almommit Alignment BasisStatement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with theRequirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter4 for methods of achieving these performance criteria (performance-based or deterministic).
NEI.0001Ref NEI 00-01 Guidance
Assume only one fire in any single fire area at a time.3.4.1.1 [Number ofPostulated Fires]
Applicability Comments
Applicable
AligenmentStatement
RNP-E/ELEC-1216 postulates only one fire occurring at a time.Aligns
Comments
Reference Document I
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.
Robinson Nuclear Plant
Doc Detail
Section 9.4.1
Section 1.4
HBRSEP LAR Rev 1 Page B-62
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
NEI.00-01Ref NEI 00-01 Guidance
Assume that the fire may affect all unprotected cables and equipment within the fire area. This3.4.1.2 [Damage to assumes that neither the fire size nor the fire intensity is known. This is conservative and bounds the
Unprotected exposure fire that is required by the regulation.
Equipment and
Cables]
Applicability Comments
Applicable
AlignmentAlignment BasisStatement
RNP-E/ELEC-1216 postulates all electrical equipment and cables in a given fire area are damagedAligns and unavailable unless NRC exemption or appropriate evaluation (GL-86-10) has been completed.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.1
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 5.0
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
act vities on the ability to achieve the nuclear safety performance crite ia of Section 1.5. [See Chapter4 fr methods of achieving these performance criteria (performance-b sed or deterministic).
NEI 00-01Ref
3.4.1.3 [Assess
HBRSEP LAR Rev 1
NEI 00-01 Guidance
Address all cable and equipment impacts affecting the required safe shutdown path in the fire area.
Page B-63
Duke Energy
Impacts to Required
Components]
ApplicabilityApplicable
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
All potential impacts within the fire area must be addressed. The focus of this section is to determine
and assess the potential impacts to the required safe shutdown path selected for achieving post-fire
safe shutdown and to assure that the required safe shutdown path for a given fire area is properly
protected.
Comments
Aignment Alignment Basis
Statement
All potential impacts of the fire are identified in the fault tree. Potential damage to equipmentAligns required to show success in each area is addressed with an appropriate compliance strategy. The
results are documented in FSSPMD and in RNP-E/ELEC-1216.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 5.0
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
NEIO0-1Ref NEI 00-01 Guidance
Use manual actions where appropriate to achieve and maintain post-fire safe shutdown conditions in3.4.1.4 [Manual accordance with NRC requirements.
Actions]
Applicability Comments
Applicable
Alignment
Statement
Aligns
AlignmentBasis
Manual actions credited in the shutdown analysis are sumrrmarized on a fire area basis in Attachment
26 of RNP-E/ELEC-1216. RNP-E-8.050 documents the feasibility of the manual actions. The
current regulatory guidance, as reflected in FAQs 06-0012 and 07-0030 was used as the basis for
determining the acceptability of the manual actions.
Comments
HBRSEP LAR Rev 1 Page B-64
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review
Reference Document Doc DetailFIR-NGGC-0101, Fire Protection Nuclear Safety Capability Attachment 2Assessment (NSCA)
RNP-E-8.050, Appendix R Transient Analysis and TimelineEvaluation for H.B. Robinson - Unit No. 2
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 5.2 and Alt. 26Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with theRequirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter4 for methods of achieving these performance criteria (performance-based or deterministic).
NEI 00-01 Ref NEI 00-01 GuidanceWhere appropriate to achieve and maintain cold shutdown within 72 hours, use repairs to equipment
3.4.1.5 [Repairs] required in support of post fire shutdown.
Applicability Comments
Applicable
AlignmentAligmentAlignment BasisStatement
Repairs are identified where necessary for cold shutdown equipment. A list of credited repairAligns procedures can be found in RNP-E/ELEC-1216.
No repairs are required to achieve safe and stable conditions.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.15 and 9.4.2Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 1.7.2.2Robinson Nuclear Plant
Table •-12 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
HBRSEP LAR Rev 1 Page B-65
Duke Energy
NFPA 805Requirement
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Fire Area Assessment. An engineering analysis shall be performed in accordance with the
requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
EIEL00-01Ref
3.4.1.6 [Assess
Compliance with
Deterministic
Criteria]
Applicability
Applicable
NEI 00-01 Guidance
Appendix R compliance requires that one train of systems necessary to achieve and maintain hotshutdown conditions from either the control room or emergency control station(s) is free of firedamage (lll.G.l.a). When cables or equipment, including associated circuits, are within the same firearea outside primary containment and separation does not already exist, provide one of the followingmeans of separation for the required safe shutdown path(s):- Separation of cables and equipment and associated non-safety circuits of redundant trains withinthe same fire area by a fire barrier having a 3-hour rating (lll.G.2.a)- Separation of cables and equipment and associated non-safety circuits of redundant trains withinthe same fire area by a horizontal distance of more than 20 feet with no intervening combustibles orfire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed inthe fire area (llI.G.2.b).- Enclosure of cable and equipment and associated non-safety circuits of one redundant train withina fire area in a fire barrier having a one-hour rating. In addition, fire detectors and an automatic firesuppression system shall be installed in the fire area (lll.G.2.c).For fire areas inside noninerted containments, the following additional options are also available:- Separation of cables and equipment and associated non-safety circuits of redundant trains by ahorizontal distance of more than 20 feet with no intervening combustibles or fire hazards (llI.G.2.d);- Installation of fire detectors and an automatic fire suppression system in the fire area (llI.G.2.e); or- Separation of cables and equipment and associated non-safety circuits of redundant trains by anoncombustible radiant energy shield (llI.G.2.f).Use exemptions, deviations and licensing change processes to satisfy the requirements mentionedabove and to demonstrate equivalency depending upon the plant's license requirements.
Comments
AlignmentStatement
This section of NEI 00-01 repeats the requirements of Appendix R III.G.2. RNP-E/ELEC-1216Aligns documents how each fire area has adequate systems to comply with these requirements and the
requirements of NFPA 805 Sections 4.2.3 and 4.2.4 for the post-transition configuration.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.2Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 5.0Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-66
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
NEI00-01 Ref NEI 00-01 GuidanceConsider selecting other equipment that can perform the same safe shutdown function as the
3.4.1.7 [Consider impacted equipment. In addressing this situation, each equipment impact, including spurious
Additional operations, is to be addressed in accordance with regulatory requirements and the NPP's current
Equipment] licensing basis.
App1icability Comments
Applicable
Alignment Alignment Basis
Statement
This consideration is not clearly stated but is inherent in performing a safe shutdown analysis. RNP-Aligns E/ELEC-1216 only documents the systems and components that were actually selected and not
those that were considered but not necessary.
Any plant system that supports meeting the safe shutdown performance goals may be considered for
inclusion in the SSEL. However, the intent is to minimize the systems and components identified in
the SSEL for configuration control purposes. If the system or component cannot directly assist in
demonstrating compliance with the deterministic requirements of NFPA 805, its inclusion in the SSEL
may not be warranted.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 9.1 and 9.4
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An ek.gineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
HBRSEP LAR Rev 1 Page B-67
Duke Ene81RefNEI 00-1Re
3.4.1.8 [ConsiderInstrument TubingEffects]
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyNEI 00-01 Guidance Capability Assessment Methodology Review
Consider the effects of the fire on the density of the fluid in instrument tubing and any subsequent
effects on instrument readings or signals associated with the protected safe shutdown path in
evaluating post-fire safe shutdown capability. This can be done systematically or via procedures
such as Emergency Operating Procedures.
Applicability Comments
Applicable
AlignmentSatemnt Alignment BasisStatement
RNP-E/ELEC-1216 documents the consideration of fire effects on instrument tubing for HBRSEP.Aligns
Comments
Reference Document DoDetael
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.1.7
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.1.2.7, 2.2.2.12, Att. 22, and Att. 25
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
NEI 00-01 Re NEI 00-01 GuidanceRefer to Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area
3.4.2 Methodology assessment.
for Fire Area Use the following methodology to assess the impact to safe shutdown and demonstrate Appendix R
Assessment compliance:
[Refer to hard copy of NEI 00-01 for Figure 3-5]
Applicability Comments
Applicable
augmo n
Staement
N/A
Alignment Basis
This is an introductory statement and provides no requirements. The sub-paragraphs with specific
requirements are addressed separately as required.
HBRSEP LAR Rev 1 Page B-68
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
NEIA0-1ef NEI 00-01 Guidance
Identify the safe shutdown cables, equipment and systems located in each fire area that may be3.4.2.1 Identify the potentially damaged by the fire. Provide this information in a report format. The report may be sorted
Affected Equipment by fire area and by system in order to understand the impact to each safe shutdown path within each
by Fire Area fire area (see Attachment 5 for an example of an Affected Equipment Report).
Applicability Comments
Applicable
AlignmentAlignment Basis
Statement
RNP-E/ELEC-1216 lists the components potentially affected in each fire area. These reports areAligns available in the FSSPMD.
Comments
Reference Document
FSSPMD, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Attachment 22
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria o0 Section 1.5. [See Chapter
4 for m. thods of achieving these performance criteria (performance-based or deterministic).
NEI 00-01 Ref
3.4.2.2 Determine
HBRSEP LAR Rev 1
NEI 00-01 Guidance
Based on a review of the systems, equipment and cables within each fire area, determine which
shutdown paths are either unaffected or least impacted by a postulated fire within the fire area.
Page B-69
Duke Energythe Shut-down Paths
Least Impacted By
a Fire in Each Fire
Area
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Typically, the safe shutdown path with the least number of cables and equipment in the fire area
would be selected as the required safe shutdown path. Consider the circuit failure criteria and the
possible mitigating strategies, however, in selecting the required safe shutdown path in a particular
fire area. Review support systems as a part of this assessment since their availability will be
important to the ability to achieve and maintain safe shutdown. For example, impacts to the electric
power distribution system for a particular safe shutdown path could present a major impediment to
using a particular path for safe shutdown. By identifying this early in the assessment process, an
unnecessary amount of time is not spent assessing impacts to the frontline systems that will require
this power to support their operation.
Based on an assessment as described above, designate the required safe shutdown path(s) for the
fire area. Identify all equipment not in the safe shutdown path whose spurious operation or mal-
operation could affect the shutdown function. Include these cables in the shutdown function list. For
each of the safe shutdown cables (located in the fire area) that are part of the required safe shutdown
path in the fire area, perform an evaluation to determine the impact of a fire-induced cable failure on
the corresponding safe shutdown equipment and, ultimately, on the required safe shutdown path.
When evaluating the safe shutdown mode for a particular piece of equipment, it is important to
consider the equipment's position for the specific safe shutdown scenario for the full duration of the
shutdown scenario. It is possible for a piece of equipment to be in two different states depending on
the shutdown scenario or the stage of shutdown within a particular shutdown scenario. Document
information related to the normal and shutdown positions of equipment on the safe shutdown
equipment list.
A~ppicabilt CommentApplicable
AlignmentStatemenlt
Algnment ais
AlignsRNP-E/ELEC-1216 identifies both the equipment that is selected for a given fire area and the
equipment that is not selected. This shows that the division selected for a given safe shutdown
system is the train that was generally that least affected by the fire.
The use of a fault tree in the analysis helps to ensure that components that may have support
systems affected, such as cooling or power supplies, are not credited without taking these failures
into account.
Reference Document Detal
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Att. 22Robinson Nuclear Plant
Table B-2 Nuclear SaIfety Capability Assessment Methodology Review
NFPA 805 Section
HBRSEP LAR Rev 1
2.4.2.4 Fire Area Assessment.
Page B-70
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
NFPA 805Requirement
Fire Area Assessment. An engineering analysis shall be performed in accordance with the
requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
NEI NEI 00-01 Guidance
Using the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine3.4.2.3 Determine the equipment that can impact safe shutdown and that can potentially be impacted by a fire in the fire
Safe Shutdown area, and what those possible impacts are.
Equipment Impacts
Applicability Comments
Applicable
Alignment BStatement
RNP-E/ELEC-1216 identifies the equipment used for safe shutdown and what the potential impact ofAligns the fire on the safe shutdown equipment. This information is also contained in FSSPMD.
Comments
Reference Document DocDetail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4
Assessment (NSCA)
FSSPMD, Fire Safe Shutdown Program Manager Database
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Attachment 22
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter4 for methods of achieving these performance criteria (performance-based or deterministic).
NE[lI00-01Ref
3.4.2.4 Develop a
Compliance
Strategy or
Disposition to
Mitigate the Effects
Due to Fire
Damage to Each
HBRSEP LAR Rev 1
NEI 00-01 Guidance
The available deterministic methods for mitigating the effects of circuit failures are summarized asloollows (see Figure 1-2):
Provide a qualified 3-fire rated barrier.- Provide a 1-hour fire rated barrier with automatic suppression and detection.- Provide separation of 20 feet or greater with automatic suppression and detection and demonstratethat there are no intervening combustibles within the 20 foot separation distance.- Reroute or relocate the circuit/equipment, or perform other modifications to resolve vulnerability.
Page B-71
Duke Energy
Required
Component or
Cable
ApplicabilityApplicable
Attachment B - NE 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
- Provide a procedural action in accordance with regulatory requirements.
- Perform a cold shutdown repair in accordance with regulatory requirements.
- Identify other equipment not affected by the fire capable of performing the same safe shutdown
function.
- Develop exemptions, deviations, Generic Letter 86-10 evaluation or fire protection design change
evaluations with a licensing change process.
Additional options are available for non-inerted containments as described in 10 CFR 50 Appendix R
section Ill.G.2.d, e and f.
Comments
AlignmentStatment
Aligns
Alignment Basis
RNP-E/ELEC-1216 verifies that appropriate separation is used for redundant cables. This can take
the form of 3-hour fire barriers, 1-hour fire barriers with suppression and detection, or 20 feet of
separation with suppression and detection. In some cases, exemptions have been requested from
and granted by the NRC for configurations that did not meet these requirements. Also some fire
protection evaluations have determined that the protection in place provides adequate separation for
the hazards of the area.
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Attachment 22
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
NE[OI 0-1Ref
3.4.2.5 Documentthe ComplianctStrategy or IDispositionDetermined toMitigate the EffectsDue to Fire
HBRSEP LAR Rev 1
NEI 00-01 Guidance
Assign compliance strategy statements or codes to components or cables to identify the justificationor mitigating actions proposed for achieving safe shutdown. The justification should address thecumulative effect of the actions relied upon by the license to mitigate a fire in the area. Provideeach piece of safe shutdown equipment, equipment not in the path whose spurious operation or mal-operation could affect safe shutdown, and/or cable for the required safe shutdown path with a specificcompliance strategy or disposition. Refer to Attachment 6 for an example of a Fire Area AssessmentReport documenting each cable disposition.
Page B-72
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology ReviewDamage to EachRequiredComponent orCable
Applicability Comments
Applicable
AlignmentStatmentAlignment Basis
Statement
The fire area by fire area separation reports contained in Attachment 22 of RNP-E/ELEC-1216Aligns identify all the analyzed circuits and components and the credited compliance strategies.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Attachment 22
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safetyRequirement functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, andshorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properlycoordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and wl ose fire-induced failure could cause the loss of therequired components shall be identified. The cZoncern is that the effects of a fire can extend outside ofthe immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
HBRSEP LAR Rev 1
NEI 00-01 Guidance
Page B-73
Duke Energy
3.5 Circuit Analysis
and Evaluation
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
This section on circuit analysis provides information on the potential impact of fire on circuits used to
monitor, control and power safe shutdown equipment. Applying the circuit analysis criteria will lead to
an understanding of how fire damage to the cables may affect the ability to achieve and maintain
post-fire safe shutdown in a particular fire area. This section should be used in conjunction with
Section 3.4, to evaluate the potential fire-induced impacts that require mitigation.
Appendix R Section Il.G.2 identifies the fire-induced circuit failure types that are to be evaluated for
impact from exposure fires on safe shutdown equipment. Section III.G.2 of Appendix R requires
consideration of hot shorts, shorts-to-ground and open circuits.
Applicability Comments
Applicable
AlignmtAlignment BasisStatement
This is an introductory statement and provides no requirements. The sub-paragraphs with specific
N/A requirements are addressed separately as required by NFPA 805 Sections 4.2.3 and 4.2.4 for the
post-transition configuration.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-i duced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
NE-•00-01RQef
3.5.1 Criteria /
HBRSEP LAR Rev 1
NEI 00-01 Guidance
Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations.
Page B-74
Duke EnergyAssumptions
Applicability
Applicable
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
Comments
Ali nment Alignment Basis
Statement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
3.5.1.1 [Circuit
Failure Types andImpact]
NEI 00-01 Guidance
Consider the following circuit failure types on each conductor of each unprotected safe shutdowncable to determine the potential impact of a fire on the safe shutdown equipment associated with that
conductor.
- A hot short may result f[om a fire-induced insulation breakdown between conductors of th• samecable, a different cable or from some other external source resulting in a compatible but undesired
impressed voltage or signal on a specific conductor. A hot short may cause a spurious operation of
safe shutdown equipment.
- An open circuit may result from a fire-induced break in a conductor resulting in the loss of circuitcontinuity. An open circuit may prevent the ability to control or power the affected equipment. An
open circuit may also result in a change of state for normally energized equipment. (e.g. [for BWRs]
HBRSEP LAR Rev 1 Page B-75
Duke Energy
ApplicabilityApplicable
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
loss of power to the Main Steam Isolation Valve (MSIV) solenoid valves due to an open circuit will
result in the closure of the MSIVs). Note that RIS 2004-03 indicates that open circuits, as an initial
mode of cable failures, are considered to be of very low likelihood. The risk-informed inspection
process will focus on failures with relatively high probabilities.
- A short-to-ground may result from a fire-induced breakdown of a cable insulation system, resulting
in the potential on the conductor being applied to ground potential. A short-to-ground may have all of
the same effects as an open circuit and, in addition, a short-to-ground may also cause an impact to
the control circuit or power train of which it is a part.
Consider the three types of circuit failures identified above to occur individually on each conductor of
each safe shutdown cable on the required safe shutdown path in the fire area.
Comments
AlignmentStatement
Aligns
Alignment Basis
The safe shutdown circuit analysis shall be reviewed and updated as necessary for credible circuit
failures as a deterministic analysis utilizing the Current Design Method (CDM). These failures include:
- Multiple shorts to ground or grounded conductor.
- Multiple open circuits.
- One hot short per affected component or multiple hot shorts for high/low pressure interface
components.
- Cable-to-cable shorts are postulated to occur.
Reference Document DocDetail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.4
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.1.1
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to actIieve the nuclear safety performance criteria, including spurious op ration and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
HBRSEP LAR Rev 1 Page B-76
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the [oss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref NEI 00-01 Guidance
Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal3.5.1.2 [Circuit mode/position of the safe shutdown equipment as shown on the schematic drawings. The analyst
Contacts and must consider the position of the safe shutdown equipment for each specific shutdown scenario when
Operational Modes] determining the impact that fire damage to a particular circuit may have on the operation of the safe
shutdown equipment.
Applicability Comments
Applicable
Alignment BasisStatement
Components are assumed to be in their normal position at the time of the fire. This includes electricalAligns contacts and switches.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 9.3.2 and 9.4.1
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated./
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
HBRSEP LAR Rev 1 Page B-77
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Reviewa power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI000Ref NEI 00-01 Guidance
Assume that circuit failure types resulting in spurious operations exist until action has been taken to3.5.1.3 [Duration of isolate the given circuit from the fire area, or other actions have been taken to negate the effects of
Circuit Failures] circuit failure that is causing the spurious actuation. The fire is not assumed to eventually clear the
circuit fault. Note that RIS 2004-03 indicates that fire-induced hot shorts typically self-mitigate after a
limited period of time.
Appolicability Comments
Applicable
AlignmentAlignment Basis
Statement
"Hot Short" duration is considered to exist until action has been taken to isolate the given circuit fromAligns the fire area, or other actions as appropriate have been taken to negate the effects of the spurious
actuation. HBRSEP does not postulate that the fire will eventually clear the "Hot Short."
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.3.6
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
HBRSEP LAR Rev 1 Page B-78
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodoloqy Reviewa power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
N.EI0-01 Ref NEI 00-01 Guidance
When both trains are in the same fire area outside of primary containment, all cables that do not3.5.1.4 [Cable meet the separation requirements of Section II.G.2 are assumed to fail in their worst caseFailure configuration.Configurations]
Applicability Comments
Applicable
Alignment Alignment Basisstatement
The following damage is assumed to occur due to the postulated fire:Aligns a. Fire damage occurs throughout the fire area under consideration.
b. Fire damage results in an unusable cable that cannot be considered functional with regard toensuring proper circuit operation.
Electrical equipment located in a fire area is assumed to fail as a result of the postulated fire in thefire area, and is considered unavailable to ensure completion of safe shutdown functions unless it
meets the separation criteria of 10 CFR 50 Appendix R or is shown to be acceptable as-is based onan approved exemption. This electrical equipment includes motors, instruments, UiP converters,controllers, switches, MCC's, switchgear, transformers, generators, batteries, panel boards, etc.
Comments
1Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.1
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 89L5
Require, ent
2.4.2.2.1 Circuits Required in Nuclear Safety Funclions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
HBRSEP LAR Rev 1 Page B-79
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology ReviewThis will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref
3.5.1.5 [A, Circuit
Failure Risk
Assessment
Guidance]
NEI 00-01 Guidance
The following guidance provides the NRC inspection focus from Bin 1 of RIS 2004-03 in order to
identify any potential combinations of spurious operations with higher risk significance. Bin 1 failures
should also be the focus of the analysis; however, NRC has indicated that other types of failures
required by the regulations for analysis should not be disregarded even if in Bin 2 or 3. If Bin 1
changes in subsequent revisions of RIS 2004-03, the guidelines in the revised RIS should be
followed.
Applicability Comments
Not Applicable
Alignmet Alignment Basis
Statement
This is an introductory statement and provides no requirements. The sub-paragraphs with specific
N/A requirements are addressed separately as required.
Comments
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safetyfunctions shall be identified. This includes circuits that are required for operation, that could preventthe operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, andshorts to ground, to identify circuits that ;re required to support the proper operation of componentsrequired to achieve the nuclear safety p rformance criteria, including spurious operation and signals.This will ensure that a comprehensive population of circuitry is evaluated.2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.
HBRSEP LAR Rev 1 Page B-80
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref
3.5.1.5 [B, Cable
Failure Modes]
NEI 00-01 Guidance
For multiconductor cables testing has demonstrated that conductor-to-conductor shorting within the
same cable is the most common mode of failure. This is often referred to as "intra-cable shorting." It
is reasonable to assume that given damage, more than one conductor-to-conductor short will occur
in a given cable. A second primary mode of cable failure is conductor-to-conductor shorting between
separate cables, commonly referred to as "inter-cable shorting." Inter-cable shorting is less likelythan intra-cable shorting. Consistent with the current knowledge of fire-induced cable failures, the
following configurations should be considered:
A. For any individual multiconductor cable (thermoset or thermoplastic), any and all potential spurious
actuations that may result from intra-cable shorting, including any possible combination of conductors
within the cable, may be postulated to occur concurrently regardless of number. However, as a
practical matter, the number of combinations of potential hot shorts increases rapidly with the number
of conductors within a given cable. For example, a multiconductor cable with three conductors (3C)has 3 possible combinations of two (including desired combinations), while a five conductor cable
(5C) has 10 possible combinations of two (including desired combinations), and a seven conductor
cable (7C) has 21 possible combinations of two (including desired combinations). To facilitate aninspection that considers most of the risk presented by postulated hot shorts within a multiconductor
cable, inspectors should consider only a few (three or four) of the most critical postulated
combinations.
B. For any thermoplastic cable, any and all potential spurious actuations that may result from intra-
cable and inter-cable shorting with other thermoplastic cables, including any possible combination of
conductors within or between the cables, may be postulated to occur concurrently regardless of
number. (The consideration of thermoset cable inter-cable shorts is deferred pending additional
research.)
C. For cases involving the potential damage of more than one multiconductor cable, a maximum of
two cables should be assumed to be damaged concurrently. The spurious actuations should be
evaluated as previously described. The consideration of more than two cables being damaged (and
subsequent spurious actuations) is deferred pending additional research.
D. For cases involving direct current (DC) circuits, the potential spurious operation due to failures of
the associated control cables (even if the spurious operation requires two concurrent hot shorts of the
proper polarity, e.g., plus-to-plus and minus-to-minus) should be considered when the required
source and target conductors are each located within the same multiconductor cable.
E. Instrumentation Circuits. Required instrumentation circuits are beyond the scope of this
associated circuit approach and must meet the same requirements as required power and control
circuits. There is one case where an instru ent circuit could potentially be considered an associated
circuit. If fire-induced damage of an instrument circuit could prevent operation (e.g., lockout
permissive signal) or cause maloperation (e.g., unwanted start/stop/reposition signal) of systems
necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an
associated circuit and handled accordingly.
HBRSEP LAR Rev 1 Page B-81
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology ReviewApplicability Comments
Not Applicable
AlinmentAlignment BStatement
Section 9.3.3 of FIR-NGGC-0101 describes the specific cable failure modes to be considered inAligns conducting a circuit analysis. Section 3.0 of FIR-NGGC-0101 contains general definitions for the
cable failure modes. Configurations considered included the following, and were applied to all safe
shutdown cables.
a) For any multiconductor cable (including thermoset, thermoplastic, and armored), any and all
potential spurious actuations that may result, including possible combinations of conductors within the
cable, may be postulated to occur concurrently regardless of the number.
b) Inter-cable shorting of thermoset cables, or thermoset and thermoplastic cables, are considered
to be credible events.
c) Compatible polarity hot shorts for DC circuits were considered to the degree specified in the cases
below:
- Case 1 - Intra-Cable Shorts within a Single Cable
For this case, a single cable must contain both a source and target conductor for both polarities. It is
postulated that intra-cable shorts within the cable will result in compatible polarity connections for
both polarities (e.g., a plus-to-plus and a minus-to-minus connection for a DC control circuit). Given
the relatively high probability of intra-cable conductor-to-conductor shorting, this failure mode was
considered.
- Case 2 - Intra-Cable Shorts on Separate Cables
For this case, two independent but coincident hot shorts of the proper polarity (without grounding) in
separate cables must occur. Given the relative high probability of intra-cable conductor-to-conductor
shorting, this failure mode was considered.
- Case 3 - Inter-Cable Shorts on Separate Cables
For this case two independent but coincident hot shorts of the proper polarity (without grounding)must occur. This case differs from Case 1 and 2 in that one or both of the hot shorts must involve
inter-cable shorting. Given the low likelihood of coincident proper polarity shorts combined with the
low likelihood of inter-cable hot shorting, this failure mode was only considered for components
identified as "high-low pressure interface" or Fire PRA "high consequence equipment."
In the plant's review of multiple spurious actuations, the following were considered.
a. Multiple spurious operations resulting from a fire-induced circuit failure affecting a single conductor.
b. Multiple fire-induced circuit failures affecting multiple conductors within the same multi-conductor
cable with the potential to cause a spurious operation of a component were assumed to exist
concurrently.
c. Multiple fire-induced circuit failure affecting separate conductors in separate cables with the
potential to cause a spurious operation of a component must be assumed to exist concurrently when
the effect of the fire-induced circuit is sealed-in or latched. There was no specific limit to the number
of cables that were considered to be damaged.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.0, 9.3.2, 9.3.3, 9.3.10, 9.4.3, 9.4.6
Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 1.1.2.3
HBRSEP LAR Rev 1 Page B-82
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.4 Fire Area Assessment.
NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the
Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression
activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter
4 for methods of achieving these performance criteria (performance-based or deterministic).
NE-IL•DD•Re• NEI 00-01 Guidance
Determination of the potential consequence of the damaged associated circuits is based on the3.5.1.5 [C, examination of specific NPP piping and instrumentation diagrams (P&IDs) and review of components
Likelihood of that could prevent operation or cause maloperation such as flow diversions, loss of coolant, or other
Undesired scenarios that could significantly impair the NPP's ability to achieve and maintain hot shutdown.
Consequences] When considering the potential consequence of such failures, the [analyst] should also consider the
time at which the prevented operation or maloperation occurs. Failures that impede hot shutdown
within the first hour of the fire tend to be most risk significant in a first-order evaluation. Consideration
of cold-shutdown circuits is deferred pending additional research.
Applicability Comments
Applicable
Alignment BasisStatement
RNP-E/ELEC-1216 limits the evaluation of multiple spurious operations, implementing the designAligns with intent strategy of any and all potential spurious operation, on a one at a time basis. As part of the manual
action feasibility study, two concurrent spurious operations were evaluated.
Multiple spurious operations (MSO) were considered for a variety of scenarios by the MSO Expert
panel. Components were identified for consideration and possible inclusion in the Safe Shutdown
Analysis and the Fire PRA. Any MSOs that were determined to be risk-significant by the PRA were
analyzed accordingly.
Comments
Reference Document Doc Detail
RNP-E-8.050, Appendix R Transient Analysis and Timeline
Evaluation for H.B. Robinson - Unit No. 2
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-83
Duke EnergyNFPA 805 Section
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
NE[.0_0.IRef
3.5.2 Types ofCircuit Failures
Applicability
Applicable
NEI 00-01 Guidance
Appendix R requires that nuclear power plants must be designed to prevent exposure fires fromdefeating the ability to achieve and maintain post-fire safe shutdown. Fire damage to circuits thatprovide control and power to equipment on the required safe shutdown path and any other equipment
whose spurious operation/mal-operation could affect shutdown in each fire area must be evaluatedfor the effects of a fire in that fire area. Only one fire at a time is assumed to occur. The extent of firedamage is assumed to be limited by the boundaries of the fire area. Given this set of conditions, itmust be assured that one redundant train of equipment capable of achieving hot shutdown is free offire damage for fires in every plant location. To provide this assurance, Appendix R requires that
equipment and circuits required for safe shutdown be free of fire damage and that these circuits bedesigned for the fire-induced effects of a hot short, short-to-ground, and open circuit. With respect tothe electrical distribution system, the issue of breaker coordination must also be addressed.This section will discuss specific examples of each of the following types of circuit failures:- Open circuit- Short-to-ground- Hot short.
Comments
Aliianment
Statement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
HBRSEP LAR Rev 1 Page B-84
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 00-01 Ref
3.5.2.1 Circuit
Failures Due to an
Open Circuit
NEI 00-01 Guidance
This section provides guidance for addressing the effects of an open circuit for safe shutdown
equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of circuit
continuity. An open circuit will typically prevent the ability to control or power the affected equipment.
An open circuit can also result in a change of state for normally energized equipment. For example, a
loss of power to the main steam isolation valve (MSIV) solenoid valves [for BWRs] due to an open
circuit will result in the closure of the MSIV.
NOTE: The EPRI circuit failure testing indicated that open circuits are not likely to be the initial fire-
induced circuit failure mode. Consideration of this may be helpful within the safe shutdown analysis.
Consider the following consequences in the safe shutdown circuit analysis when determining the
effects of open circuits:
Loss of electrical continuity may occur within a conductor resulting in de-energizing the circuit and
causing a loss of power to, or control of, the required safe shutdown equipment.
In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or
0oher device. This loss of power may change the state of the equipn ent. Evaluate this to determine
iftequipment fails safe.
Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in
secondary damage.
HBRSEP LAR Rev 1 Page B-85
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology ReviewFigure 3.5.2-1 shows an open circuit on a grounded control circuit.
[Refer to hard copy of NEI 00-01 for Figure 3.5.2-1]
Open circuit No. 1:
An open circuit at location No. 1 will prevent operation of the subject equipment.
Open circuit No. 2:
An open circuit at location No. 2 will prevent opening/starting of the subject equipment, but will not
impact the ability to close/stop the equipment.
CommentsApplicability
Applicable
AlignmenStatement
Alignment Basis
AlignsCircuits are evaluated using both the Current Design /Method (CDM) and Revised Design Method
(RDM). Fire induced circuit failures (CDM and RDM) to be considered are described in Progress
-Energy Procedure FIR-NGGC-0101. Fire induced circuit failure analysis utilizing CDM includes
multiple open circuits.
An evaluation considering the potential for secondary fires resulting from an open circuit on all CT
secondary circuits at RNP has been performed and is documented in EC 93120. This evaluation
concludes that there is little or no potential for adverse impacts to Safe Shutdown equipment resulting
from this postulated failure mode.
Comments
Reference Document Doc Detail
EC 93120, Evaluation of Possible Secondary Fire Caused by Open
Circuited CT at RNP
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.3, 9.3.4Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.1.1.4Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equ pment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open .circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
HBRSEP LAR Rev 1 Page B-86
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI ~00-1Ref
3.5.2.2 CircuitFailures Due to aShort-to-Ground [A,General]
ApplicabilityApplicable
NEI 00-01 Guidance
This section provides guidance for addressing the effects of a short-to-ground on circuits for safeshutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation systemresulting in the potential on the conductor being applied to ground potential. A short-to-ground cancause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases whereproper coordination does not exist.
Consider the following consequences in the post-fire safe shutdown analysis when determining theeffects of circuit failures related to shorts-to-ground:- A short to ground in a power or a control circuit may result in tripping one or more isolation devices(i.e. breaker/fuse) and causing a loss of power to or control of required safe shutdown equipment.
- In the case of certain energized equipment such as HVAC dampers, a loss of control power mayresult in loss of power to an interlocked relay or other device that may cause one or more spuriousoperations.
Comments
AlignmenttAliganment Basis
Statement
This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.
Comments
Table 8-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
HBRSEP LAR Rev 1
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
Page B-87
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
NEI0-01Ref NEI 00-01 Guidance
This section provides guidance for addressing the effects of a short-to-ground on circuits for safe3.5.2.2 Circuit shutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation systemFailures Due to a resulting in the potential on the conductor being applied to ground potential. A short-to-ground canShort-to-Ground [B, cause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-
Grounded Circuits] ground may affect other equipment in the electrical power distribution system in the cases where
proper coordination does not exist.
Short-to-Ground on Grounded Circuits
Typically, in the case of a grounded circuit, a short-to-ground on any part of the circuit would present
a concern for tripping the circuit isolation device thereby causing a loss of control power.
Figure 3.5.2-2 illustrates how a short-to-ground fault may impact a grounded circuit.
[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-21
Short-to-ground No. 1:
A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to
the control circuit. This will result an inability to operate the equipment using the control switch.
Depending on the coordination characteristics between the protective device on this circuit and
upstream circuits, the power supply to other circuits could be affected.
Short-to-ground No. 2:
A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch
is closed. Should this occur, the effect would be identical to that for the short-to-ground at location
No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop
control switch, the equipment will still be able to be opened/started.
Aplical_•bilat Comments
Applicable
Alignment
HBRSEP LAR Rev 1 Page B-88
Duke Energy
Statement
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewAlignment Basis
Circuits are evaluated using both the Current Design /Method (CDM) and Revised Design Method
(RDM). Fire induced circuit failures (CDM and RDM) to be considered are described in Progress
Energy Procedure FIR-NGGC-0101. Fire induced circuit failure analysis utilizing CDM includes
multiple shorts to ground.
Aligns
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.3Assessment (NSCA)
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.1.1.4Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safetyRequirement functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
3.5.2.2 Circuit
Failures Due to a
Short-to-Ground [C,
Ungrounded
Circuits]
HBRSEP LAR Rev 1
NEI 00-01 Guidance
Short-to-Ground on Ungrounded Circuits
In the case of an ungrounded circuit, postulating only a single short-to-ground on any part of the
circuit may not result in tripping the circuit isolation device. Another short-to-ground on the circuit or
another circuit from the same source would need to exist to cause a loss of control power to the
circuit.
Page B-89
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology ReviewFigure 3.5.2-3 illustrates how a short to ground fault may impact an ungrounded circuit.
[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-3]
Short-to-ground No. 1: A short-to-ground at location No. 1 will result in the control power fuse
blowing and a loss of power to the control circuit if short-to-ground No. 3 also exists either within the
same circuit or on any other circuit fed from the same power source. This will result in an inability to
operate the equipment using the control switch. Depending on the coordination characteristics
between the protective device on this circuit and upstream circuits, the power supply to other circuits
could be affected.
Short-to-ground No. 2:
A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch
is closed. Should this occur, the effect would be identical to that for the short-to-ground at location
No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop
control switch, the equipment will still be able to be opened/started.
CommentsApplicability
Applicable
AlmawmntStatement
Alignment Basis
AlignsA single ground fault on an ungrounded AC or DC control circuit has no immediate functional affect.
Thus, ungrounded systems are more resilient to functional failures. Nonetheless, multiple ground
faults are credible and must be considered. For ease of analysis, an existing - but unspecified -
ground fault from the same power source will be assumed when analyzing ungrounded systems.
Furthermore, multiple shorts-to-ground are to be evaluated for their impact on ungrounded circuits.
As noted in FIR-NGGC-0101, it is likely that over the course of a fire at least one conductor from
each polarity of a circuit (positive and negative polarity) will eventually become grounded. Thus, the
circuit analysis should not try to take credit for a circuit remaining functional simply because two
conductors must short to ground to render the circuit inoperable (i.e., blow the fuse or trip the circuit
breaker).
Comments
Reference Document Doc Qetail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.3Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
HBRSEP LAR Rev 1
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
Page B-90
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of components
required to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
NEI 0-01Ref NEI 00-01 Guidance
This section provides guidance for analyzing the effects of a hot short on circuits for required safe3.5.2.3 Circuit shutdown equipment. A hot short is defined as a fire-induced insulation breakdown between
Failures Due to a conductors of the same cable, a different cable or some other external source resulting in an
Hot Short [A, undesired impressed voltage on a specific conductor. The potential effect of the undesired impressed
General] voltage would be to cause equipment to operate or fail to operate in an undesired manner.
Consider the following specific circuit failures related to hot shorts as part of the post-fire safe
shutdown analysis:
- A hot short between an energized conductor and a de-energized conductor within the same cable
may cause a spurious actuation of equipment. The spuriously actuated device (e.g., relay) may be
interlocked with another circuit that causes the spurious actuation of other equipment. This type of
hot short is called a conductor-to-conductor hot short or an internal hot short.
- A hot short between any external energized source such as an energized conductor from another
cable (thermoplastic cables only) and a de-energized conductor may also cause a spurious actuation
of equipment. This is called a cable-to-cable hot short or an external hot short. Cable-to-cable hot
shorts between thermoset cables are not postulated to occur pending additional research.
Applicability Comments
Applicable
AAgenment
Statement
The Current Design Method is the safe shutdown circuit analysis method used for applying failures toAligns circuits. One key atribute of this method is that a hot short is applied independent of the cable
configuration and is applied as a hot probe. The probe's power is postulated to be present, and its
source is not identified.
Unless otherwise documented, all cables at HBRSEP were assumed to be thermo-plastic. Hot shorts
are postulated to occur regardless of the cable insulation type.
HBRSEP LAR Rev 1 Page B-91
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.17 and 9.3.3Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safetyRequirement functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, andshorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.This will ensure that a comprehensive population of circuitry is evaluated.2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside ofthe immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.
3.5.2.3 CircuitFailures Due to aHot Short [B,Grounded Circuits]
NEI 00-01 Guidance
A Hot Short on Grounded Circuits
A short-to-ground is another failure mode for a grounded control circuit. A short-to-ground asdescribed above would result in de-energizing the circuit. This would further reduce the likelihood forthe circuit to change the state of the equipment either from a control switch or due to a hot short.Nevertheless, a hot short still needs to be considered. Figure 3.5.2-4 shows a typical groundedcontrol circuit that might be used for a motor-operated valve. However, the protective devices andposition indication lights that would normally be included in the control circuit for a motor-operatedvalve ha e been omitted, since these devices are not required to understany the concepts beingexplainec# in this section. In the discussion provided below, it is assumed that a single fire in a givenfire area could cause any one of the hot shorts depicted. The following discussion describes how toaddress the impact of these individual cable faults on the operation of the equipment controlled bythis circuit.
[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-4]
Page B-92HBRSEP LAR Rev 1
Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety
Capability Assessment Methodology Review
Hot short No. 1:
A hot short at this location would energize the close relay and result in the undesired closure of a
motor-operated valve.
Hot short No. 2:
A hot short at this location would energize the open relay and result in the undesired opening of a
motor-operated valve.
Applicability
Applicable
Comments
Alignment Alignment Basis
Statement
Hot shorts on grounded circuits were considered. Cables susceptible to grounds were identified withAligns the associated equipment.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.3
Assessment (NSCA)
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis
NFPA 805
Requirement
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety
functions shall be identified. This includes circuits that are required for operation, that could prevent
the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation
shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and
shorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.
This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common
enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for
their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of
a power supply required to achieve the nuclear safety performance criteria shall be identified. This
situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly
coordinated with the downstream protection device. I(b) Common Enclosure Circuits. Those circuits that share enclosLres with circuits required to achieve
the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the
required components shall be identified. The concern is that the effects of a fire can extend outside of
the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via
inadequately sealed fire area boundaries.
HBRSEP LAR Rev 1 Page B-93
Attachment B - NEI 04-02 Table B-2 Nuclear Safetyn 6 =LuI eI~~ LndedUIIL Capain I iissent ivietii u IUUIU Rxeview~
NEL0-01Ref
3.5.2.3 Circuit
Failures Due to a
Hot Short [C,
Ungrounded
Circuits]
Applicability
Applicable
NEI 00-01 Guidance
A Hot Short on Ungrounded Circuits
In the case of an ungrounded circuit, a single hot short may be sufficient to cause a spuriousoperation. A single hot short can cause a spurious operation if the hot short comes from a circuit fromthe positive leg of the same ungrounded source as the affected circuit.
In reviewing each of these cases, the common denominator is that in every case, the conductor inthe circuit between the control switch and the start/stop coil must be involved.
Figure 3.5.2-5 depicted below shows a typical ungrounded control circuit that might be used for amotor-operated valve. However, the protective devices and position indication lights that wouldnormally be included in the control circuit for a motor-operated valve have been omitted, since thesedevices are not required to understand the concepts being explained in this section.
In the discussion provided below, it is assumed that a single fire in a given fire area could cause anyone of the hot shorts depicted. The discussion provided below describes how to address the impactof these cable faults on the operation of the equipment controlled by this circuit.
[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-5]
Hot short No. 1:A hot short at this location from the same control power source would energize the close relay andresult in the undesired closure of a motor operated valve.
Hot short No. 2:A hot short at this location from the same control power source would energize the open relay and
result in the undesired opening of a motor operated valve.
Comments
AAlAlignment B
Statement
Hot shorts on ungrounded circuits were considered. Cables susceptible to grounds were identifiedAligns with the associated equipment. Hot shorts between two cables were considered credible in the RNP
analysis.
Comments
Reference Document Doc Detail
FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.3
Assessment (NSCA)
RNP-E/ELE4_-1216, The Fire Safe Shutdown Analysis for H.B. Sectiin 3.1.1
Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-94
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.
NFPA 805
Requirement
Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be
identified.
NE[I0001Ref
3.5.2.4 Circuit
Failures Due to
Inadequate Circuit
Coordination
NEI 00-01 Guidance
The evaluation of associated circuits of a common power source consists of verifying proper
coordination between the supply breaker/fuse and the load breakers/fuses for power sources that are
required for safe shutdown. The concern is that, for fire damage to a single power cable, lack of
coordination between the supply breaker/fuse and the load breakers/fuses can result in the loss of
power to a safe shutdown power source that is required to provide power to safe shutdown
equipment.
For the example shown in Figure 3.5.2-6, the circuit powered from load breaker 4 supplies power to a
non-safe shutdown pump. This circuit is damaged by fire in the same fire area as the circuit
providing power to from the Train B bus to the Train B pump, which is redundant to the Train A pump.
To assure safe shutdown for a fire in this fire area, the damage to the non-safe shutdown pump
powered from load breaker 4 of the Train A bus cannot impact the availability of the Train A pump,
which is redundant to the Train B pump. To assure that there is no impact to this Train A pump due
to the associated circuits' common power source breaker coordination issue, load breaker 4 must be
fully coordinated with the feeder breaker to the Train A bus.
[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-6]
A coordination study should demonstrate the coordination status for each required common power
source. For coordination to exist, the time-current curves for the breakers, fuses and/or protectiverelaying must demonstrate that a fault on the load circuits is isolated before tripping the upstream
breaker that supplies the bus. Furthermore, the available short circuit current on the load circuit must
be considered to ensure that coordination is demonstrated at the maximum fault level.
The methodology for identifying potential associated circuits of a common power source and
evaluating circuit coordination cases of associated circuits on a single circuit fault basis is as follows:
- Identify the power sources required to supply power to safe shutdown equipment.
- For each power source, identify the breaker/fuse ratings, types, trip settings and coordination
characteristics for the incoming source breaker supplying the bus and the breakers/fuses feeding the
loads supplied by the bus.
- For each power source, demonstrate proper circuit coordination using acceptable industry methods.
- For power sources not properly coordinated, tabulate by fire area the routing of cables whose
breaker/fuse is not properly coordinated with the supply breaker/fuse. Evaluate the potential for
disabling power to the bus in each of the fire areas in which the associated circuit cables of concern
are routed and the power source is required for safe shutdown. Prepare a list of the following
information for each fire area:
- Cables of concern.
- Affected common power source and its path.
- Raceway in which the cable is enclosed.
- Sequence of the raceway in the cable route.
HBRSEP LAR Rev 1 Page B-95
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy - Fire zone/area in which the raceway is located.
For fire zones/areas in which the power source is disabled, the effects are mitigated by appropriate
methods.
Develop analyzed safe shutdown circuit dispositions for the associated circuit of concern cables
routed in an area of the same path as required by the power source. Evaluate adequate separation
based upon the criteria in Appendix R, NRC staff guidance, and plant licensing bases.
CommentsApplicabilityApplicable
AlignmentStatement AlignmentBasi
AlignsPower cables for Safe Shutdown equipment have been selected for evaluation for all components
that are required to change states. Coordination of electrical breakers and fuses assure that other
power cables from loads on the same electrical bus or distribution center will not adversely impact
safe shutdown equipment. FIR-NGGC-0101 provides guidance on verifying that circuit coordination
exists, as well as methods for addressing cases where coordination is not readily apparent.
Circuit breaker and fuse coordination are verified by calculations.
Comments
Reference Document
FIR-NGGC-0101, Fire Protection Nuclear Safety CapabilityAssessment (NSCA)
RNP-E-8.005, 1OCFR50 Appendix R Associated Circuit, CommonPower Supply Analysis
RNP-E-8.053, Non-Safety Overcurrent Protection Coordination
RNP-E-9.021, 1OCFR50 Appendix R Fuse Analysis for DS Bus
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.Robinson Nuclear Plant
Doc Detail
Section 9.3.6
Section 3.2.2
Table B-2 Nuclear Safety Capability Assessment Methodology Review
NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.
NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be
Requirement identified.
I
3.5.2.5 CircuitFailures Due toCommon Enclosure
HBRSEP LAR Rev 1
NEI 00101 Guidance
The common enclosure associated circuit concern deals with the possibility of causing secondary
failures due to fire damage to a circuit either whose isolation device fails to isolate the cable fault or
protect the faulted cable from reaching its ignition temperature, or the fire somehow propagates along
the cable into adjoining fire areas.
Page B-96
Duke EnergyConcerns
Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review
The electrical circuit design for most plants provides proper circuit protection in the form of circuit
breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature
is reached. Adequate electrical circuit protection and cable sizing are included as part of the original
plant electrical design maintained as part of the design change process. Proper protection can be
verified by review of as-built drawings and change documentation. Review the fire rated barrier and
penetration designs that preclude the propagation of fire from one fire area to the next to
demonstrate that adequate measures are in place to alleviate fire propagation concerns.
Applicability
Applicable
Comments
AlgnmentStatement
Circuit breaker and fuse coordination are verified by calculations. Adequate coordination exists to
Aligns assure that a common enclosure issue is not credible.
Comments
Reference Document DocDetall
FPP-RNP-200, 1OCFR50, Appendix R, Section lIl.G, Associated Section 4.0Circuits Analysis
RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.2.2.3Robinson Nuclear Plant
Table B-2 Nuclear Safety Capability Assessment Methodology Review
HBRSEP LAR Rev 1 Page B-97
Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
L. NFPA 805 Chapter 3 Requirements for Approval10 CFR 50.48(c)(2)(vii)
13 Pages Attached
HBRSEP LAR Rev I Page L-1 IIHBRSEP LAR Rev I Page L-1 I
Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
Approval Request 1
NFPA 805 Section 3.3.5.1
NFPA 805 Section 3.3.5.1 states:"Wiring above suspended ceiling shall be kept to a minimum. Where installed,electrical wiring shall be listed for plenum use, routed in armored cable, routed inmetallic conduit, or routed in cable trays with solid metal top and bottom covers."
HBRSEP has wiring above suspended ceilings that may not comply with therequirements of this code section.
Suspended ceilings are noncombustible and exist only in the Control Room (FZ 23),Inside AO Office and old Turbine Building RCA Entrance (FZ 25A). Combustibles inconcealed spaces are minimal.
The three areas currently with suspended ceilings inside the NFPA 805 defined powerblock are in the Control Room (FZ 23), Inside AO Office and old Turbine Building RCAEntrance (FZ 25A). The Inside AO Office and old Turbine Building RCA Entrance (FZ25A) are not risk significant. Neither of the rooms nor the cables are safety-related.
Most electrical wiring above the Control Room partial suspended ceiling is in conduitexcept for short flexible connectors to lighting fixtures. There is one eight-foot length ofeight-inch diameter UL approved flexible air duct with flame spread rating of 25 or less.The quantity of cabling above the suspended ceilings in the Control Rooms is very lowand results in limited combustible loading. The existing fire detection capability and/orthe Control Room Operators who are continuously present in the area would identify thepresence of smoke. In addition, no equipment important to nuclear safety is located inthe vicinity of these cables.
Video/communication/data cables that have been field routed above suspended ceilingsare low voltage. Existing cables for video, communication, and networking may not beplenum rated, but are not generally susceptible to shorts that would result in a fire.
Basis for Request:
The basis for the approval request of this deviation is:" All electrical wiring above the control room partial suspended ceiling is in
conduit except for short flexible connectors to lighting fixtures. According toFAQ 06-0021, cable air drops of limited length (-3 feet) are consideredacceptable.
* No equipment important to nuclear safety is located in the vicinity of thesecables.
* Minimum amount of cables exist above the Control Room ceiling, which resultsin limited combustible loading.
* Smoke Detectors are installed both above and below the partial suspendedceiling in the Control Room.
" The Inside AO Office and old Turbine Building RCA Entrance (FZ 25A) are notrisk significant. Neither of the rooms nor the cables are safety related.
I
HBRSEP LAR Rev I Page L-2 I
Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
* Existing fleet procedures will be used to ensure that changes moving forwardare considered for NFPA 805 impacts. (FIR-NGGC-0010)
Acceptance Criteria Evaluation:
Nuclear Safety and Radiological Release Performance Criteria:
The location of wiring above suspended ceilings does not affect nuclear safety. Powerand control cables comply with this section. No equipment important to nuclear safetyis located in the vicinity of these cables. Therefore, there is no impact on the nuclearsafety performance criteria.
The location of cables above suspended ceilings has no impact on the radiologicalrelease performance criteria. The radiological release review was performed based onthe manual fire suppression activities in areas containing or potentially containingradioactive materials and is not dependent on the type of cables or locations ofsuspended ceilings. The location of cables does not change the radiological releaseevaluation performed that potentially contaminated water is contained and smokemonitored. The cables do not add additional radiological materials to the area orchallenge system boundaries that contain such.
HBRSEP LAR Rev I Page L-3 IIHBRSEP LAR Rev I Page L-3 I
Duke Energy Aftachment L - NFPA 805 Chapter 3 Requirements for Approval
Safety Margin and Defense-in-Depth:
Power and control cables meet the requirements of this requirement. The use of thesematerials has been defined by the limitations of the analytical methods used in thedevelopment of the FPRA. Therefore, the inherent safety margin and conservatisms inthese methods remain unchanged.
The three echelons of defense-in-depth are 1) to prevent fires from starting(combustible/hot work controls), 2) rapidly detect, control and extinguish fires that dooccur thereby limiting damage (fire detection systems, automatic fire suppression,manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protectionfor systems and structures so that a fire will not prevent essential safety functions frombeing performed (fire barriers, fire rated cable, success path remains free of firedamage, recovery actions). The prior introduction of non-listedvideo/communication/data cables routed above suspended ceilings does not impact fireprotection defense-in-depth. Echelon 1 is maintained by the current cable installationprocedures documenting the requirements of NFPA 805 Section 3.3.5.1. The controlroom is a continuously manned area of the plant. The introduction of cables abovesuspended ceilings does not affect echelons 2 and 3. The video/communication/datacables routed above suspended ceilings does not result in compromising automatic firesuppression functions, manual fire suppression functions, fire protection for systemsand structures, or post-fire safe shutdown capability.
Conclusion:
HBRSEP determined that the performance based approach satisfies the followingcriteria:
* Satisfies the performance goals performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release
* Defense in Depth
" Safety Margin
HBRSEP LAR Rev I Page L-4 I!HBRSEP LAR Rev I Page L-4 I
Duke Energy Affachment L - NFPA 805 Chapter 3 Requirements for Approval
Approval Request 2
NFPA 805 Section 3.3.5.2
NFPA 805 Section 3.3.5.2 states:"Only metal tray and metal conduits shall be used for electrical raceways. Thin wallmetallic tubing shall not be used for power, instrumentation, or control cables. Flexiblemetallic conduits shall only be used in short lengths to connect components."
The use of PVC piping for underground embedded conduit is permitted by HBRSEP perHBR2-0B060 Sht D6 for electrical raceway installations. Polyvinyl Chloride (PVC) orHigh Density Polyethylene (HDPE) type ducts (conduits) are permitted when embeddedin compacted sand or reinforced concrete. In addition, some PVC conduit was found inreinforced concrete wall. The PVC/HDPE conduit is embedded within a noncombustibleenclosure which provides protection from mechanical damage and from damageresulting from either an exposure fire or from a fire within the conduit impacting othertargets.
Basis for Request:
" The PVC/HDPE conduit, while a combustible material, is not subject toflame/heat impingement from an external source which would result in structuralfailure, contribution to fire load, and damage to the circuits contained withinwhere the conduit is embedded in concrete or compacted sand.
" Failure of circuits within the conduit resulting in a fire would not result in damageto external targets.
Acceptance Criteria Evaluation:
Nuclear Safety and Radiological Release Performance Criteria:
The use of PVC/HDPE conduit in embedded locations does not affect nuclear safety asthe material in which conduits are run within an embedded location is not subject to thefailure mechanisms potentially resultant in circuit damage or resultant damage toexternal targets. Therefore there is no impact on the nuclear safety performance criteria.
The use of PVC/HDPE conduits in embedded installations has no impact on theradiological release performance criteria. The radiological release review wasperformed based on the manual fire suppression activities in areas containing orpotentially containing radioactive materials and is not dependent on the type of conduitmaterial. The conduit material does not change the radiological release evaluationperformed that concluded that potentially contaminated water is contained and smoke ismonitored. The conduits do not add additional radiological materials to the area orchallenge systems boundaries that contain such as the PVC/HDPE conduits areembedded.
HBRSEP LAR Rev I Page L-5
Duke Energy Affachment L - NFPA 805 Chapter 3 Requirements for Approval
Safety Margin and Defense-in-Depth:
The PVC/HDPE conduit material is embedded in a non-combustible configuration. Thematerial is protected when embedded from mechanical damage and from damageresulting from either an exposure fire or from a fire within the conduit impacting othertargets. The areas with PVC/HDPE conduit have been analyzed in their currentconfiguration. The precautions and limitations on the use of these materials do notimpact the analysis of the fire event. Therefore, the inherent safety margin andconservatisms in these analysis methods remain unchanged.
The three echelons of defense-in-depth are 1) to prevent fires from starting(combustible/hot work controls), 2) rapidly detect, control and extinguish fires that dooccur thereby limiting damage (fire detection systems, automatic fire suppression,manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protectionfor systems and structures so that a fire will not prevent essential safety functions frombeing performed (fire barriers, fire rated cable, success path remains free of firedamage, recovery actions). The use of PVC/HDPE conduits in embedded installationsdoes not impact fire protection defense-in-depth. The PVC/HDPE conduit in embeddedinstallations does not affect echelons 1, 2, and 3. The PVC/HDPE conduits do notdirectly result in compromising automatic fire suppression functions, manual firesuppression functions, or post-fire safe shutdown capability.
Conclusion:
HBRSEP determined that the performance based approach satisfies the followingcriteria"
* Satisfies the performance goals performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release
* Defense in Depth
" Safety Margin
HBRSEP LAR Rev I Page L-6 IIHBRSEP LAR Rev 1 Page L-6 I
Duke Energy Aftachment L - NFPA 805 Chapter 3 Requirements for Approval
Approval Request 3
NFPA 805 Section 3.5.16
NFPA 805 Section 3.5.16 states:"The fire protection water supply system shall be dedicated for fire protection use only.Exception No. 1: Fire protection water supply systems shall be permitted to be used toprovide backup to nuclear safety systems, provided the fire protection water supplysystems are designed and maintained to deliver the combined fire and nuclear safetyflow demands for the duration specified by the applicable analysis.Exception No. 2: Fire protection water storage can be provided by plant systems servingother functions, provided the storage has a dedicated capacity capable of providing themaximum fire protection demand for the specified duration as determined in thissection.
The review of plant flow diagrams show no hard connections to other plant systems,besides those for fire protection use. It should be noted that although there are no hardpipe connections to other plant systems, there are procedures that utilize the fireprotection water supply. They are as follows:
" AOP-014 - Loss of CCW
" AOP-022 - Loss of Service Water
* EDMG-001 - Extreme Damage Event Early Actions
* EDMG-002 - Refueling Water Storage Tank (RWST)
* EDMG-003 - Condensate Storage Tank (CST)
" EDMG-005 - Containment Vessel (CV)
" EDMG-01 1 - Spent Fuel Pit Casualty
* EDMG-012 - Core Cooling Using Alternate Water Source
* EDMG-01 3 -Airborne Release Scrubbing
" SAM-1 - Inject into the Steam Generator
* SAM-3 - Inject into the RCS
• SAM-4 - Inject into Containment
" SAM-6 - Control Containment Conditions
" SAM-8 - Flood Containment
The use of the fire protection water for these non-fire protection system water demandswould have no adverse impact on the ability of the fire protection system to providerequired flow and pressure. OMM-002, Section 8.15, details restrictions and allowancesfor use of the fire protection water supply system at HBRSEP.
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Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
Basis for Request:
The use of the fire protection water for these non-fire protection system water demandswould have no adverse impact on the ability of the fire protection system to providerequired flow and pressure. This is based on how fire water usage is restricted (CR 99-01247), in the following ways:
1. Fire service related activities (emergency, testing and training).
2. When the use of fire water is specifically called out in approved plantprocedures (i.e., AOPs).
3. During plant emergencies when fire water is needed to protect safety relatedequipment.
4. When usage is deemed necessary AND sufficient justification is provided toshow that the use of the fire water system for the proposed activity does notcause the fire water system to be in a condition outside of its design basis(i.e., the quantity of water needed for the proposed activity does not dropsupply and pressure below that required/defined in UFSAR Section 9.5.1).Permission shall have the approval of the Shift Manager (CR 96-00729 andCR 96-00730).
The water supply system is capable of maintaining the pressure in the main plant loopat 70 psi or higher with the largest deluge system in operation and with the systemsupplying an additional 1000 gpm to hoses.
Acceptance Criteria Evaluation:
Nuclear Safety and Radiological Release Performance Criteria:
The use of fire protection water for non-fire protection plant evolutions is an occurrencethat requires Shift Manager review and concurrence. The flow limitations to those non-fire protection functions ensure that there is no impact in the ability of the automaticsuppression systems to perform Therefore, there is no impact on the nuclear safetyperformance criteria.
The use of fire protection water for plant evolutions other than fire protection has noimpact on the radiological release performance criteria. The radiological releaseperformance criteria is satisfied based on the determination of limiting radioactiverelease (Attachment E), which is not affected by impacts on the fire protection systemdue it's use for non-fire protection purposes.
Safety Margin and Defense-i n-Depth:
The use of the fire water system, including the use of hydrants and hose, for non-fireprotection uses does not impact fire protection defense-in-depth. The fire pumps havethe excess capacity to supply the demands of the fire protection system as well as thenon-fire protection uses identified above. This does not compromise automatic ormanual fire suppression functions, fire suppression for systems and structures, or thenuclear safety capability assessment. Since both the automatic and manual firesuppression functions are maintained, defense-in-depth is maintained.
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Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
The methods, input parameters, and acceptance criteria used in this analysis werereviewed and found to be in accordance with NFPA 805 Chapter 3. The methods, inputparameters, and acceptance criteria used to calculate flow requirements for theautomatic and manual suppression systems were not altered. Therefore, the safetymargin inherent in the analysis for the fire event has been preserved.
Conclusion:
HBRSEP determined that the performance based approach satisfies the followingcriteria:
" Satisfies the performance goals performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release
" Defense in Depth
* Safety Margin
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Duke Energy Aftachment L - NFPA 805 Chapter 3 Requirements for Approval
Approval Request 4
NFPA 805 Section 3.2.3(1)In accordance with 10 CFR 50.48(c)(2)(vii), "Performance-based methods," the fireprotection program elements and minimum design requirements of Chapter 3 may besubject to the performance-based methods permitted elsewhere in the standard.
In accordance with NFPA 805 Section 2.2.8, the performance-based approach to satisfythe nuclear safety, radiation release, life safety, and property damage/businessinterruption performance criteria requires engineering analyses to evaluate whether theperformance criteria are satisfied.
In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shalldetermine that the performance-based approach utilized to evaluate a variance from therequirements of NFPA 805 Chapter 3:
A. Satisfies the performance goals, performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release;
B. Maintains safety margins; andC. Maintains fire protection defense-in-depth (fire prevention, fire detection, fire
suppression, mitigation, and post-fire nuclear safety capability).Duke Energy, HBRSEP requests formal approval of performance-based exception tothe requirements in Chapter 3 of NFPA 805 as follows:
NFPA 805, Section 3.2.3(1)"Procedures shall be established for implementation of the fire protection program.In addition to procedures that could be required by other sections of the standard,the procedures to accomplish the following shall be established:Inspection, testing, and maintenance for fire protection systems and featurescredited by the fire protection program."
Duke Energy, HBRSEP requests the ability to utilize performance-based methods toestablish the appropriate inspection, testing, and maintenance frequencies for fireprotection systems and features required by NFPA 805. Performance-based inspection,testing, and maintenance frequencies will be established as described in Electric PowerResearch Institute (EPRI) Technical Report TR-1006756, "Fire Protection SurveillanceOptimization and Maintenance Guide for Fire Protection", Final Report, July 2003.
Basis for Request:NFPA 805 Section 2.6, "Monitoring," requires that
"A monitoring program shall be established to ensure that the availability andreliability of the fire protection systems and features are maintained and to assessthe performance of the fire protection program in meeting the performance criteria.Monitoring shall ensure that the assumptlons in the engineering analysis remainvalid."
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Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
NFPA 805 Section 2.6.1, "Availability, Reliability, and Performance Levels," requiresthat "Acceptable levels of availability, reliability, and performance shall beestablished."
NFPA 805 Section 2.6.2, "Monitoring Availability, Reliability, and Performance,"requires that "Methods to monitor availability, reliability, and performance shall beestablished. The methods shall consider the plant operating experience andindustry operating experience."
The scope and frequency of the inspection, testing, and maintenance activities for fireprotection systems and features required in the fire protection program have beenestablished based on the previously approved Technical Specifications / LicenseControlled Documents and appropriate NFPA codes and standard. This request doesnot involve the use of the EPRI Technical Report TR-1006756 to establish the scope ofthose activities as that is determined by the required systems review identified inAttachment C
This request is specific to the use of EPRI Technical Report TR-1006756 to establishthe appropriate inspection, testing, and maintenance frequencies for fire protectionsystems and features credited by the fire protection program. As stated in EPRITechnical Report TR-1 006756 Section 10.1, "The goal of a performance-basedsurveillance program is to adjust test and inspection frequencies commensurate withequipment performance and desired reliability." This goal is consistent with the statedrequirements of NFPA 805 Section 2.6. The EPRI Technical Report TR-1006756provides an accepted method to establish appropriate inspection, testing, andmaintenance frequencies which ensure the required NFPA 805 availability, reliability,and performance goals are maintained.
The target tests, inspections, and maintenance will be those activities for the NFPA 805required fire protection systems and features. The reliability and frequency goals will beestablished to ensure the assumptions in the NFPA 805 engineering analysis remainvalid. The failure criterion will be established based on the required fire protectionsystems and features credited functions and will ensure those functions are maintained.Data collection and analysis will follow the EPRI Technical Report TR-1 006756document guidance. The failure probability will be determined based on EPRI TechnicalReport TR-1 006756 guidance and a 95% confidence level will be utilized. Theperformance monitoring will be performed in conjunction with the Monitoring Programrequired by NFPA 805 Section 2.6 and it will ensure site specific operating experience isconsidered in the monitoring process. The following is a flow chart that identifies thebasic process that will be utilized.
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Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for ApprovalDuke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
Program Framework
Identify Target Tests and Inspections
Establish Reliability and Frequency GoalsSet Failure Criteria
Assess Licensing Impact and Other Constraints
Data Collection and Evaluation
Establish Data Collection Guidelines
Collect Required Surveillance Data
Assemble Data in Spreadsheet or DatabaseAnalyze .Data to Identify Failures
Reliability and Uncertainty Analysis
Compute Failure ProbabilitiesCompute Uncertainty Limits
Confirm That Reliability Supports Target Frequency
Program ImplementationModify Program Documents
Revise -Surveil lance ProceduresConduct Ongoing Performance Monitoring
Refine and Modify Frequencies as Appropriate
rP RI TR-1 006756 - Figure 10-1Flowchart fo&Performance-Based Surveillance Program
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Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
Duke Energy, HBRSEP does not intend to revise any fire protection surveillance, test orinspection frequencies until after transitioning to NFPA 805. Existing fire protectionsurveillance, test and inspection will remain consistent with applicable station, Insurer,and NFPA Code requirements. HBRSEP's intent is to obtain approval via the NFPA 805Safety Evaluation to use EPRI Technical Report TR1 006756 guideline in the future asopportunities arise. Duke Energy, HBRSEP reserves the ability to evaluate fireprotection features with the intent of using the EPRI performance-based methods toprovide evidence of equipment performance beyond that achievable under traditionalprescriptive maintenance practices to ensure optimal use of resources while maintainingreliability.
Nuclear Safety and Radiological Release Performance Criteria:Use of performance-based test frequencies established per EPRI Technical ReportTR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, willensure that the availability and reliability of the fire protection systems and features aremaintained to the levels assumed in the NFPA 805 engineering analysis. Therefore,there is no adverse impact to Nuclear Safety Performance Criteria by the use of theperformance-based methods in EPRI Technical Report TR-1006756.
The radiological release performance criteria are satisfied based on the determination oflimiting radioactive release. Fire Protection Systems and Features may be credited aspart of that evaluation. Use of performance-based test frequencies established per theEPRI Technical Report TR-1 006756 methods combined with NFPA 805 Section 2.6,Monitoring Program, will ensure that the availability and reliability of the fire protectionsystems and features are maintained to the levels assumed in the NFPA 805engineering analysis which includes those assumptions credited to meet theRadioactive Release performance criteria. Therefore, there is no adverse impact toRadioactive Release performance criteria.
Safety Margin and Defense-in-Depth:Use of performance-based test frequencies established per EPRI Technical ReportTR-1006756 methods combined with NFPA 805, Section 2.6, Monitoring Program, willensure that the availability and reliability of the fire protection systems and features aremaintained to the levels assumed in the NFPA 805 engineering analysis which includesthose assumptions credited in the Fire Risk Evaluation safety margin discussions. Inaddition, the use of these methods in no way invalidates the inherent safety marginscontained in the codes and standards used for design and maintenance of fireprotection systems and features. Therefore, the safety margin inherent and credited inthe analysis has been preserved.
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Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval
The three echelons of defense-in-depth described in NFPA 805 Section 1.2 are1) to prevent fires from starting (combustible/hot work controls),2) rapidly detect, control and extinguish fires that do occur thereby limiting
damage (fire detection systems, automatic fire suppression, manual firesuppression, pre-fire plans), and
3) provide adequate level of fire protection for systems and structures so that afire will not prevent essential safety functions from being performed (firebarriers, fire rated cable, success path remains free of fire damage, recoveryactions).
Echelon 1 is not affected by the use of the EPRI Technical Report TR-1 006756methods. Use of performance-based test frequencies established per EPRI TechnicalReport TR-1006756 methods combined with NFPA 805 Section 2.6, MonitoringProgram, will ensure that the availability and reliability of the fire protection systems andfeatures credited for defense-in-depth are maintained to the levels assumed in theNFPA 805 engineering analysis. Therefore, there is no adverse impact to echelons 2and 3 for defense-in-depth.
Conclusion:NRC approval is requested for use of the performance-based methods contained in theElectric Power Research Institute (EPRI) Technical Report TR-1006756, "FireProtection Equipment Surveillance Optimization and Maintenance Guide", Final Report,July 2003 to establish the appropriate inspection, testing, and maintenance frequenciesfor fire protection systems and features required by NFPA 805. As described above, thisapproach is considered acceptable because it:
A. Satisfies the performance goals, performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release;
B. Maintains safety margins; andC. Maintains fire protection defense-in-depth (fire prevention, fire detection, fire
suppression, mitigation, and post-fire safe shutdown capability).
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