Waterford, Unit 3 - Control Element Assembly Drop Times ...Control Element Assembly (CEA) drop times associated with Technical Specification 3.1.3.4 for Waterford Steam Electric Station

Proprietary Information -Withhold From Public Disclosure Under 10 CFR 2.390The balance of this letter may be considered non-proprietary upon removal of Attachment 4.

--En tergy Entergy Operations, Inc.17265 River RoadKillona, LA 70057Tel 504 739 6660Fax 504 739 6678

Michael R. ChisumSite Vice PresidentWaterford 3

W3FI1-2015-0062

September 23, 2015

U.S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, DC 20555

SUBJECT: Control Element Assembly Drop Times Submittal Request for AdditionalInformationWaterford Steam Electric Station, Unit 3Docket No. 50-382License No. NPF-38

REFERENCES: 1. W3F1-2015-0040, License Amendment Request to ReviseControl Element Drop Times, July 2, 2015 [ADAMSAccession Number ML15197A1 06].

2. W3F1-2015-0061, Supplement to Revise Control ElementAssembly Drop Times Associated with TechnicalSpecification 3.1.3.4, August 13, 2015 2015 [ADAMSAccession Number MLI15226A346].

3. NRC CEA Drop Time Submittal Request for AdditionalInformation, August 26, 2015, [ADAMS Accession NumberML1 5232A275].

Dear Sir or Madam:

On July 2, 2015, Entergy Operations, Inc. (Entergy) requested an amendment to revise theControl Element Assembly (CEA) drop times associated with Technical Specification3.1.3.4 for Waterford Steam Electric Station Unit 3 (Waterford 3) [Reference 1]. On August13, 2015, Waterford 3 submitted a supplement [Reference 2] to provide additional accidentscenario results to bound a possible delay in the control element assembly holding coildecay time. Subsequently, the Nuclear Regulatory Commission (NRC) has requestedadditional information to aid in their review [Reference 3]. This letter provides theresponse to the NRC request for additional information.

This submittal does not alter the no significant hazards consideration or environmentalassessment previously submitted by Waterford 3 in letter W3F1 -201 5-0040 [Reference 1].

MTR1

Cross-Out

W3FI1-2015-0062Page 2

Waterford 3 letter W3F1-2015-0040 [Reference 1] submitted a proprietary fuel thermalconductivity degradation evaluation. This letter provides a non-proprietary and proprietaryversion of that fuel thermal conductivity degradation evaluation.

Attachment 4 is proprietary in its entirety, as it contains information that is proprietary toWestinghouse Electric Company (Westinghouse). Attachment 3 contains a redacted non-proprietary version. Attachment 2 contains the Proprietary Information Affidavit. Thepurpose of Attachment 2 is to withhold the proprietary information contained in Attachment4 from public disclosure. The Affidavit, signed by Westinghouse as the owner of theinformation, sets forth the basis for which the information may be withheld from publicdisclosure by the Commission and addresses with specificity the considerations listed inparagraph (b)(4) of § 2.390 of the Commission's regulations. Accordingly, it is respectfullyrequested that the information proprietary to Westinghouse be withheld from publicdisclosure in accordance with 10CFR 2.390.

If you have any questions or require additional information, please contact John Jarrell,Regulatory Assurance Manager, at 504-739-6685.

I declare under penalty of perjury that the foregoing is true and correct. Executed onSeptember 23, 2015.

Sincerely,

M RC/J PJ/wjs

Attachments:

1. Response to NRC Request for Additional Information2. Fuel Thermal Conductivity Degradation Evaluation Proprietary Affidavit3. Non-Proprietary - Fuel Thermal Conductivity Degradation Evaluation4. PROPRIETARY - Fuel Thermal Conductivity Degradation Evaluation

cc: Mr. Marc L. DapasRegional AdministratorU. S. NRC, Region [email protected]

NRC Senior Resident Inspector for Waterford [email protected] (SRI)Chris.Speer~nrc.gov (RI)

NRC Project Manager for Waterford [email protected]

Louisiana Department of Environmental QualityOffice of Environmental ComplianceSurveillance [email protected]

Attachment I to

W3FI1-2015-0062

Response to NRC Request for Additional Information

Attachment 1 toW3F1 -201 5-0062Page 1 of 21

1.0 DESCRIPTION

On April 22, 2015, a Category 1 public meeting was held between the U.S. NuclearRegulatory Commission (NRC) staff and representatives of Entergy Operations, Inc.(Entergy) and Westinghouse Electric Company (Westinghouse) at the NRC Headquarters.The purpose of the meeting was to discuss Entergy's proposed license amendmentrequest (LAR) regarding changes to Technical Specification (TS) 3.1.3.4 (CEA Drop Time)and Updated Final Safety Analysis Report (UFSAR) Chapter 15 (Accident Analyses).Reference 3.1 provides the meeting summary information and Reference 3.2 provides themeeting presentation information.

As discussed in the public meeting and pursuant to 10 CFR 50.90, Entergy requested anamendment to revise the Control Element Assembly (CEA) drop times associated withTechnical Specification 3.1.3.4 for Waterford Steam Electric Station Unit 3 (Waterford 3)[Reference 3.3]. The submittal identified that additional information would be useful to theNRC during their review. The NRC staff reviewed the Waterford 3 submittal andconcluded that this same additional information was necessary to enable the NRC staff tomake an independent assessment regarding the acceptability of the proposed amendment[Reference 3.4].

On August 13, 2015, Waterford 3 submitted a supplement [Reference 3.51 to provideadditional accident results due to a possible cause identified in the apparent causeevaluation. Subsequently, the Nuclear Regulatory Commission (NRC) has requestedadditional information to aid in their review [Reference 3.6]. Section 2 provides theresponse to the NRC request for additional information.

2.0 REQUEST FOR ADDITIONAL INFORMATION (RAI)

NRC RAI #1In Attachment 2, page 4, the following statement is made:

The analysis margin for the axial power distribution was reduced from an ASI [axialshape index] of +0.3 to +0.2, which is conservative to the COLR [core operatinglimits report] limit of +0.16.

Table 15.0-4 of the FSAR shows the axial shape index used is: -0.2 :5 ASI s +0.2. Providethe bases for the current margin for the ASI of +0.3 and explain why this is different fromthe ESAR value.

Waterford 3 RAI #1 ResponseW3FI-2015-0040 Attachment 2 page 4 is for UFSAR Section 15.1.2.3 (Increased MainSteam Flow). UFSAR Table 15.0-4 is for the general overview of Chapter 15 initialconditions.

UFSAR Section 15.0.3.2 states:The range of values of each of the principal process variables that were consideredin analyses of all incidents discussed in this section are listed in Table 15.0-4. It is

Attachment 1 toW3F1 -2015-0062Page 2 of 21

emphasized that no plant operational or safety problems have been identified foroperating conditions outside of the range shown in Table 15.0-4. This range merelyrepresents a range of expected normal reactor operation.

UFSAR Table 15.0-4 is based upon the Technical Specifications, Core Operating LimitsReport (COLR), and reload design allowed values. The specific analyses may usedifferent values provided that they are conservative and provide bounded results.

Letter W3F1-2003-0074 [Reference 3.9] for the extended power uprate was approved inNRC Technical Specification Amendment 199 [Reference 3.10]. Letter W3F1-2003-0074[Reference 3.9] Attachment 5 Section 2.13.0.2 described the range of initial conditions(similar to UFSAR Table 15.0-4) as follows:

The range of initial conditions evaluated in the non-LOCA transient analyses islisted in Table 2.13.0-2. Values beyond this range are used in certain analyses toprovide additional margin.

This is the same approach used for UFSAR Table 15.0-4. UFSAR Section 15.0.3.2 will beupdated to contain similar wording to that contained W3FI-2003-0074 Attachment 5Section 2.13.0.2 to better describe that more bounding inputs may be used to provideconservative results.

NRC RAI #2In Attachment 2, pages 4, 6, 16, and 17, the following statement is made:

Analysis margin for the least negative Doppler reactivity was reduced from -0.00113to -0.0013 Ap/h/°K.

Provide the bases for the least negative Doppler coefficients of -0.00113 to -0.0013Ap/U/K. Also, justify adequacy of the use of the value of -0.0013 Ap/PPK in the reanalysisused to support the proposed increase in the limits of the CEA drop times in TS 3.1.3.4.

Waterford 3 RAI #2 ResponseFigure 2-1 documents the fuel temperature coefficient (FTC) ranges from the mostnegative (-0.0026 Ap/h/°K) to the least negative (-0.0013 Ap/•/°K) for the current WaterfordUnit 3 operating fuel cycles. These bounding cycle independent values are confirmedevery reload cycle.

A conservative value of -0.00113 Ap/P/°K was selected during the Waterford Unit 3extended power update analyses with the intention of bounding a wider range ofsubsequent reload cycles. As part of the reanalysis to support the revised CEA drop timecurve for this License Amendment Request, this extra analysis conservatism wasremoved.

Figure 2-1 shows the FTC in units of IAp/°F for the -0.00113 Ap/h/°K, -0.00130 Ap//°K,and -0.0026 App/a/K curves.


Figure 2-1Cycle Independent Fuel Temperature Coefficient Range

and Safety Analysis Limits

Fuel Temperature, °F-4.00E-05

-3.50E-05

-3.00E-05

-2.50E-05

• -2.00E-05

-1.00E-05

-5.00E-06

.F,, - ,w.

- _ __ _ _ __ _ _ _

0.00E+00 ~___5' iC00 15b30 20 30 25K30 30

-- -- -0.00260 rho/(SQRT-K) ....- 0.00130 rho/(SQRT-K) - -0.00113 rho/(SQRT-K)

NRC RAI #3In Attachment 2, page 7, the following statement is made:

Plant changes since the extended power uprate have been incorporated into theanalysis under the 10 CFR [Title 10 of the Code of Federal Regulations] 50.59process. The analysis has been updated to account for the replacement steamgenerators (SGs) and the NGF [next generation fuel] DNBR [departure fromnucleate boiling ratio] correlation. The revised evaluation started with the analysis ofrecord and only revised CEA drop time to determine the impact.

Clarify the difference between "analysis" and "analysis of record." Does the revisedevaluation include the replacement steam generators and NGF DNBR correlation?

Waterford 3 RAI #3 Response"Analysis" and "Analysis of Record" are used synonymously in this instance. The revisedanalysis does include the replacement steam generators and Next Generation Fuel DNBRcorrelation.

Attachment 1 toW3F1-201 5-0062Page 4 of 21

NRC RAI #-4In Attachment 4, page 4, the Figure 2, "Cycle 15 through Cycle 20 CEA Insertion Times,"shows the repeated measured CEA insertion times for Cycle 20. The measured insertiontimes for Cycle 20-2 are significantly shorter than that of Cycle 20-1.

Discuss the measurement methods for the Cycle 20-1 and Cycle 20-2 data, includinguncertainties, identifying the causes for the reduction of the insertion times frommeasurements for Cycle 20-1 to Cycle 20-2, and justify the adequacy of the Cycle 20-2data by showing that the reduced insertion times are not contributed from randommeasurement errors. Was anything done to improve the drop time between Cycle 20-1and Cycle 20-2, or was the test simply repeated?

Waterford 3 RAI #4 ResponseNo changes were made between performance of the Cycle 20-1 and Cycle 20-2 CEA droptime tests.

The CEA Drop Time Test (CDTT) software records all the CEA positions every 50milliseconds [CENPSD-388 Revision 01 - CEA Drop Time Test Software User's Manual forWSES-3]. The CDTT software will always round time up to the next 50 millisecond intervalensuring the rod drop times delivered will be conservative. This results in the CDTTsoftware having a CEA drop time uncertainty of 100 milliseconds (0.1 seconds).

The Cycle 20-1 initial test was performed with the average time being 3.024 seconds. TheCycle 20-2 second test was performed with the average time being 2.967 seconds. Thedifference between the Cycle 20-1 and Cycle 20-2 test results were within the CDTTsoftware uncertainty band.

NRC RAI #5Table 4.0-1 in Attachment 1, page 4, provides the new CEA drop times. Other than thechange at 90% CEA insertion (from 3.0 to 3.2 seconds), what is the basis for the otherchanges to the CEA drop time curve? Was any plant data used to determine the newcurve? If so, provide plant data and demonstrate that the new curve isconservative. Provide the reactivity vs. time (or CEA position) curve that was presentedduring the pre-application public meeting on April 22, 2015.

Waterford 3 RAI #5 ResponseThe basis for the revised rod drop position versus time was based on plant data and aconservative analysis approach. Reviewing the time it takes the rods to reach 90%inserted for each cycle reveals that for Cycles 5 through 17, the average drop time is 2800milliseconds while the average drop time for Cycles 19 and 20 is 2950 milliseconds. Tobound the step change, the analysis rod drop time was shifted by 200 milliseconds or 0.2seconds for all rod positions from 10% to 90% inserted. To maintain continuity in thecurve, the 5% insertion position time was increased by 0.15 second.

Attachment 1 toW3F 1-2015-0062Page 5 of 21

The shape of the revised drop time curve is the same as that utilized in the current UFSARChapter 15 Non-LOCA safety analysis, but shifted to the right (slightly slower) to accountfor the impact of major plant modifications on the reactor coolant system overtime. Therefore, the shape of the rod drop time curve remains constant in the revisedanalyses supporting the Technical Specification change request. Confirmation of therevised rod drop position versus time curve will occur as part of the Cycle 21 rod droptesting.

Figure 5-1 provides the rod drop position (% withdrawn) versus time for the current safetyanalysis and the revised rod drop time.

Figure 5-1Rod Drop Position versus Time

120

100

80~

"C 60~

40~

20

00 1 2 3

Ttu., seconds4

-- -- Curnte Rod DropTiuw -.. R-- ied Rod Drop Time

Per the response to Question 1, the actual full power axial shape index (ASI) range is ±0.2.The normalized reactivity versus time value of +0.3 ASI was conservatively used duringthe Waterford Unit 3 extended power uprate analyses with the intent of bounding a widerrange of subsequent reload cycles. As part of the reanalysis due to the revised rod dropcurve, this extra analysis conservatism was removed. Figure 5-2 provides normalizedreactivity versus time for both a +0.3 and +0.2 ASI.


0.2

-O.2

4 .0.4

-1

-1.2

Figure 5-2Normalized Reactivity vs. Time

0123

Timne, seconds

4

--- .O.3 M Ciurnt Drp Time -.-. •0*.3 ASS ul Domp Time .0,.2 ASS Rte~isel Drop Time

NRC RAI #6Attachment 1, page 12, calls out references 7.40, 7.41 and 7.42, however, these threereferences are not provided in the References section (pages 13-15). It appears thatsome of the references are improperly numbered. For example, page 12 of Attachment 1states "The NRC approved this request in NRC Technical Specification Amendment 158[Reference 7.42]." However, this appears to be Reference 7.37 on page 15. Update thereference numbers in text so they are consistent with the reference list.

Waterford 3 RAI #6 ResponseThe reference issue is limited to letter W3F1-2015-0040 Section 6. All references areincluded in the reference section. The following is the correct reference numbers.

Waterford 3 letter W3P89-3094 was listed as reference 7.37 and should have beenreference 7.32.Waterford 3 Technical Specification Amendment 58 was listed as reference 7.38and should have been reference 7.33.ANO letter 2CAN080701 was listed as reference 7.39 and should have beenreference 7.34.ANO Technical Specification Amendment 275 was listed as reference 7.40 andshould have been reference 7.35.


St. Lucie letter L-2009-127 was listed as reference 7.41 and should have beenreference 7.36.St. Lucie Technical Specification Amendment 158 was listed as reference 7.42 andshould have been reference 7.37.

The corrected W3FI-2015-0040 Section 6 is provided.

6.0 PRECEDENCE

Waterford 3 previously requested a Technical Specification change to the CEA dropmethodology in letter W3P89-3094 [Reference 7.32]. This submittal was approvedin NRC Technical Specification Amendment 58 [Reference 7.33]. This changerequest is similar in scope to that previously requested and approved.

Arkansas Unit 2 requested a Technical Specification change to increase theindividual CEA drop time in letter 2CAN080701 [Reference 7.34]. The NRCapproved this request in NRC Technical Specification Amendment 275 [Reference7.35]. This change was necessitated by the transition to Next Generation Fuelsimilar to Waterford 3's submittal request.

St. Lucie Unit 2 requested a Technical Specification change to increase the CEAdrop time in letter L-2009-127 [Reference 7.36]. The NRC approved this request inNRC Technical Specification Amendment 158 [Reference 7.37].

NRC RAI #7Attachment 1, pages 2 and 3, discuss apparent and potential causes for increased roddrop times. Provide the cycles as to when plant modifications were made (i.e.,replacement steam generators were installed between Cycles XX and YY).

Waterford 3 RAI #7 ResponseAppendix K Power Uprate was implemented in Cycle 12.Extended Power Uprate was implemented in Cycle 14.Alternate Source Term was implemented in Cycle 14.Next Generation Fuel (NGF) - Cycle 16 implemented a partial core of NGF and Cycle 17

implemented a full core of NGF.Control Element Assembly Replacement - The outage prior to Cycle 18 (between Cycle

17 and Cycle 18) replaced the CEAs.Steam Generator Replacement - The outage prior to Cycle 19 (between Cycle 18 and

Cycle 19) replaced the steam generators.Reactor Head Replacement - The outage prior to Cycle 19 (between Cycle 18 and Cycle

19) replaced the reactor vessel head.Control Element Drive Mechanism Replacement - The outage prior to Cycle 19 (between

Cycle 18 and Cycle 19) replaced the CEDMs.


NRC RAI #8Attachment 1, Page 6, Paragraph 3 states that:

In order to aid the NRC review, the relevant Waterford 3 licensing basis history is provided.The last major plant change that submitted the accident and transient analyses to the NRCfor review was the extended power uprate [Reference 7.15] and alternate source termimplementation [Reference 7.18]. Two additional major plant changes that have beenimplemented since the extended power uprate are the use of Next Generation Fuel (NGF)and Steam Generator replacement. These changes were addressed by NRC approvals(Table 4.0-2) and the 10CFR50.59 process. The Westinghouse reload analysismethodology has been applied to implement these Waterford 3 changes.

Please confirm that the stated "Westinghouse reload analysis methodology" waspreviously approved by the NRC and that no changes were made to the approvedmethodology since it was used to implement the use of NGF and Steam Generatorreplacement. If changes were made to the NRC-approved reload analysis methodology,identify and justify the changes for the use in support of the proposed TS regarding theCEA drop times.

Waterford 3 RAI #8 ResponseThe CEA drop time analyses did not require any changes to the methodologies describedin the Update Final Safety Analysis Report [Reference 3.8].

Westinghouse reload analysis methodology is not a specific topical that was approved bythe NRC. The Westinghouse reload analysis methodology is a term used to describe theaccumulation of those methodologies that have been approved by the NRC and used inthe reload process.

The UFSAR methodologies were submitted in the extended power uprate request[W~aterford 3 letter W3F1-2003-0074 Reference 3.9]. Letter W3FI-2003-0074 Attachment5 Section 2.6 (Reactor Systems), Section 2.12.3.1 (LBLOCA Methodology), Section2.12.4.1 (SBLOCA Methodology), Section 2.12.5.1 (LOCA Long Term CoolingMethodology), and Section 2.13.0.1 (Non-LOCA Methodology) provided the analysismethodologies utilized by Westinghouse. Methodology changes since extended poweruprate that could be applicable to this change were listed in letter W3FI-2015-0040[Reference 3.3] Table 4.0-2 (NRC Amendments of Interest).


N RC RAI #9Attachment 2, Page 5, "Section FSAR 15.1.2.3 Increased Main Steam Flow," states that:

The increase in peak secondary pressure is based on the loss of condenservacuum results which showed an increase in peak secondary pressure of less than1 psi.

ESAR Section 15.1.2.3 does not discuss a loss of condenser vacuum. Explain why a lossof condenser vacuum was used for this case.

Waterford 3 RAI #9 ResponseThe limiting events with respect to peak primary and secondary pressures are the loss ofcondenser vacuum and feedwater line break accidents. These are the limiting eventsbecause they have the closest approach to the Technical Specification Section 2.1.2(Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressure limitof 1210 psia (110% of design pressure). UFSAR Section 15.1 contains the increase inheat removal by the secondary system events. These events are less adverse than theloss of condenser vacuum with respect to peak primary and secondary pressure.Specifically, loss of condenser vacuum peak secondary pressure is 1181 psia [UFSARSection 15.2.1.3.3.3] whereas the increased main steam flow event peak secondarypressure is 1102 psia [UFSAR Table 15.1-8A].

The increased CEA drop time impacts the power reduction post-trip and the amount ofenergy deposited into the primary coolant system. The slightly longer drop time means aslight increase in the amount of energy added to the system. The assessment provided inW3FI-2015-0040 used the loss of condenser vacuum event because it produced thelargest post-trip primary and secondary pressure spike due to losing its secondary heatremoval capability. Since, loss of condenser vacuum event produces a more adversepressure transient, the slight energy deposition increase would be expected to have themost adverse impact on this event. Thus, taking the loss of condenser vacuum eventresults and applying them to the increased main steam flow event was considered aconservative approach to ensure the change was bounded.

Attachment 1 toW3FI1-2015-0062Page 10 of 21

NRC RAI #10Attachment 2, page 6, states:

The increase in peak primary pressure was based on the loss of condenser vacuumresults which showed an increase in peak primary pressure of less than 1 psi[pounds per square inchJ when the pressurizer pressure exceeded the pressurizersafety valve opening setpoints (2575 psia [pounds per square inch absolute]). Theincrease in peak secondary pressure was based on the loss of condenser vacuumresults which showed an increase in peak secondary pressure of less than 1 psi.

FSAR, Section 15.1.2.4 does not discuss loss of condenser vacuum. Explain why loss ofcondenser vacuum was used for this case.

Waterford 3 RAI #10 ResponseThe limiting events with respect to peak primary and secondary pressures are the loss ofcondenser vacuum and feedwater line break accidents. These are the limiting eventsbecause they have the closest approach to the Technical Specification Section 2.1.2(Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressure limitof 1210 psia (110% of design pressure). UFSAR Section 15.1 contains the increase inheat removal by the secondary system events. These events are less adverse than theloss of condenser vacuum with respect to peak primary and secondary pressure.Specifically, loss of condenser vacuum peak primary pressure is 2711 psia and secondarypressure is 1181 psia [UFSAR Section 15.2.1.3.3.3] whereas the inadvertent operating ofan atmospheric dump valve event peak primary pressure is 2583 psia and secondarypressure is 1117 psia [UFSAR Table 15.1-8C].

The increased CEA drop time impacts the power reduction post-trip and the amount ofenergy deposited into the primary coolant system. The slightly longer drop time means aslight increase in the amount of energy added to the system. The assessment provided inW3F1-2015-0040 used the loss of condenser vacuum event because it produced thelargest post-trip primary and secondary pressure spike due to losing its secondary heatremoval capability. Since, the loss of condenser vacuum event produces a more adversepressure transient, the slight energy deposition increase would be expected to have themost adverse impact on this event. Thus, taking the loss of condenser vacuum eventresults and applying them to the inadvertent operating of an atmospheric dump valve eventwas considered a conservative approach to ensure the change was bounded.


NRC RAI #11Attachment 2, pages 4 and 5, FSAR, Section 15.1.2.3 considers two cases: (1) typicalcase and (2) worst departure from nucleate boiling performance case. This section doesnot discuss which case was analyzed. Clarify if both cases were considered in the updatedanalysis.

Waterford 3 RAI #11 ResponseUFSAR Section 15.1.2.3 describes the increased main steam flow with a concurrent lossof offsite power event. The event analyzed was the worst DNB performance case. Onlythe worst DNB performance case was evaluated because it bounds the typical case interms of fuel failure.

NRC RAI #12The table in Attachment 2, page 6, contains a "*" after 2584 psia. Provide the significanceof the "*".

Waterford 3 RAI #12 ResponseThe "*" should have not have been in the table. It has no significance.

NRC RAI #13

Attachment 2, page 26, states, in part, that changes to the updated analysis include:

..the use of the actual initial thermal margin reserved in the LCO.

Provide the parameter that was changed and the original and updated values.

Waterford 3 RAI #13 ResponseU FSAR Section 15.9.1.1 describes the Asymmetric Steam Generator Transient (ASGT).The revised CEA drop time analysis changed the AOR requirement for the actual initialthermal margin reserved. The limiting AOR scenario evaluated was the 95% powerclosure of the main steam isolation valve #2. The AOR initial thermal margin is defined asRequired Over Power Margin (ROPM) and was 119.9% whereas the revised CEA droptime analysis ROPM was 120.23%. The ROPM that was already preserved in COLSS at95% power is 123%, thus the ASGT event used additional initial thermal margin but waswithin the initial thermal margin already preserved so no COLSS changes were required.The ASGT analysis demonstrated that the initial margins were adequate to ensure that theASGT event does not violate the DNBR (> 1.24) and Fuel Centerline TemperatureSAFDLs.


NRC RAI #14In FSAR 15.9.1.1, "Asymmetric Steam Generator Transient," there are four eventsconsidered including: Loss of Load to One Steam Generator (LLI1SG), Excess Load toOne Steam Generator (EL/i1SG), Loss of Feedwater to One Steam Generator (LF/1 SO)and Excess Feedwater to One Steam Generator (EF/1SG). In the updated analysisdescribed in Attachment 2, pages 25-26, verify if all four cases were analyzed or if it wasassumed the limiting case was still limiting after the rod drop time change.

Waterford 3 RAI #14 ResponseThe limiting initiating event for the Asymmetric Steam Generator Transient analysis is theclosure of the Main Steam Isolation Valve (MSIV) to steam generator #2 (LL/1SG). Onlythe limiting case was evaluated with the revised CEA drop time.

NRC RAI #15In Attachment 2, the results of the analysis for a loss of normal feedwater (LONE) eventwere used to support the adequacy of the dose releases for the following events:

Page 10 - Steam System Piping Failures: Pre-Trip Power Excursion AnalysisPage 18 - Single Reactor Coolant Pump (RCP) Shaft Seizure / Sheared ShaftPage 22 - CEA EjectionPage 23 - Primary Sample or Instrument Line BreakPage 24 - Steam Generator Tube RupturePage 24 - Loss of Coolant Accident

Provide justification for each event above for the use of the LONE results to support theradiological releases.

Waterford 3 RAI #15 ResponseWater-ford 3 letter W3F1-2015-0040 [Reference 3.3] generically described that the W3F1-2015-0040 Attachment 2 Loss of Normal Feedwater Flow event (ESAR 15.2.3.2) waschosen to evaluate the transient characteristics with respect to energy deposition andassociated steam releases which would be applicable to all the events. The analysisshowed that the differences in primary and secondary system energy after reactor trip areinsignificant. As time increases farther past the time of CEA rod insertion, the differencesof the impact of the revised CEA drop time become negligible. The radiological releasesdue to steam release and break flow would remain the same. In addition, the W3FI-2015-0040 Attachment 2 results for each of the individual events demonstrate that the fuelfailure limits remain unchanged which mean the radiological source terms remain thesame. The W3F1-2015-0040 conclusion was there is no change to the radiologicalresults.

The increased CEA drop time impacts the power reduction post-trip and the amount ofenergy deposited into the primary coolant system. The slightly longer drop time means aslight increase in the amount of energy added to the system. For the loss of normalfeedwater event, Figure 15-1 shows the energy added to the reactor coolant system fromtime of trip (43.7 seconds) to 50 seconds for both the AOR and revised CEA drop time


analysis. Figure 15-2 shows the loss of normal feedwater event AOR and revised CEAdrop time analysis energy deposition from the time of trip to 1800 seconds. The depositedenergy between the reactor trip time and 1800 seconds is 161485.7 Mwt-sec for the AORand 162228.1 Mwt-sec for the revised CEA drop time analysis. This equates to an energychange of 0.0046 or 0.46% [(162228.1 - 161485.7)1/161485.7] which would have anegligible change on the radiological consequences.

Figure 15-1. Deposited Energy

4000

3500 r

3000.,

2500

I2000

1500

1000

500 --

0-43 44 45 46 47 48

Time (seconds)

-I,-Revised CEA Drop Time Analysis

49 50

-4Analysis of Record


Figure 15-2. Deposited Energy

3500 i

3000

2500 i

1500

*

*

*1000-

5001 L0

0 200 400 600 800 1000 1200 10 60 10

Time (seconds)

-4.-Analysis of Record -U1-Revised CEA Drop Time Analysis

1400 1600 1800

Waterford 3 letters W3F1-2004-0053 [Reference 3.13] and W3F1-2004-0071 [Reference3.16] submitted the Alternate Source Term (AST) analyses to the NRC. NRC OperatingLicense amendment 198 [Reference 3.14] approved the use of AST for Waterford 3.W3FI1-2004-0053, W3F1 -2004-0071, and Amendment 198 provide additional details foreach of the radiological events. The Waterford 3 AST analyses follow the guidanceprovided in NRC Regulatory Guide 1.183 [Reference 3.15]. The AST analyses areevaluated for the Exclusion Area Boundary (EAB) dose for a 2 hour duration [RegulatoryGuide 1.183 Section 4.1.5], Low Population Zone (LPZ) dose for the accident duration[Regulatory Guide 1.183 Section 4.1.6], and control room dose for the accident duration[Regulatory Guide 1.183 Section 4.2.6]. Since, the EAB, LPZ, and control room doses arecalculated for 2 hours and beyond, the use of the loss of normal feedwater eventinformation is appropriate for the time duration. In addition, Figure 15-2 shows a negligiblechange (0.46%) in deposited energy at 1800 seconds. As time increases to 2 hours andbeyond, the deposited energy difference will continue to decrease. Thus, the radiologicalconsequences are not adversely impacted by this change.

The NRC had requested a more detailed description of the release pathways. Thefollowing provides the references to more detailed release information for each of theevents requested.

Attachment 1 toW3FI-201 5-0062Page 15 of 21

Steam System Pipingi Failures: Pre-Trip Power Excursion AnalysisW3FI-2004-0053 [Reference 3.13] Attachment 2 Section 7.0 describes the pre-tripsteam line break event. The NRC Alternate Source Term Safety Evaluation Report[Reference 3.14] Section 2.1.2 also describes the pre-trip steam line break event.

Single Reactor Coolant Pump (RCP) Shaft Seizure I Sheared ShaftW3F1-2004-0071 [Reference 3.16] Attachment 1 Section 4.0 describes the reactorcoolant pump sheared shaft I seized rotor event. The NRC Alternate Source TermSafety Evaluation Report [Reference 3.14] Section 2.1.4 also describes the reactorcoolant pump sheared shaft I seized rotor event.

CEA EjectionW3FI-2004-0053 [Reference 3.13] Attachment 2 Section 10.0 describes the CEAEjection event. The NRC Alternate Source Term Safety Evaluation Report[Reference 3.14] Section 2.1.5 also describes the CEA ejection event.

Primary Sample or Instrument Line BreakW3FI-2004-0071 [Reference 3.16] Attachment 1 Section 7.0 describes the letdownline break event. The NRC Alternate Source Term Safety Evaluation Report[Reference 3.14] Section 2.1.6 also describes the letdown line break event.

Steam Generator Tube RuptureW3F1-2004-0053 [Reference 3.13] Attachment 2 Section 8.0 describes the CEAEjection event. The NRC Alternate Source Term Safety Evaluation Report[Reference 3.14] Section 2.1.7 also describes the CEA ejection event.

Loss of Coolant AccidentW3FI-2004-0053 [Reference 3.13] Attachment 2 Section 5.0 and 6.0 describes theCEA Ejection event. The NRC Alternate Source Term Safety Evaluation Report[Reference 3.14] Section 2.1.8 and 2.1.9 also describes the CEA ejection event.


NRC RAI #16Attachment 2, page 17, discussed the analysis of the single reactor coolant pump (RCP)shaft seizure/sheared shaft events. The reanalysis showed that the peak primary pressurewas increased by 20 pounds per square inch (psi) to 2442 pounds per square inchabsolute (psia) and the peak secondary pressure was increased by 1 psi to 1118psia. The increase in the peak primary pressure was based on the LONE long term resultsand the increase in the peak secondary pressure was based on the loss of condenservacuum (LOCV) results.

Please provide justification for the use of two different events results, LONF and LOCV, toderive the peak primary and secondary pressure, respectively, for the RCP shaftseizure/sheared shaft events.

Waterford 3 RAI #16 ResponseThe limiting events with respect to peak primary and secondary pressures are the loss ofcondenser vacuum (LOCV) and feedwater line break accidents. These are the limitingevents because they have the closest approach to the Technical Specification Section2.1.2 (Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressurelimit of 1210 psia (110% of design pressure). Since the LOCV peak primary pressure isgreater than the opening setpoint of the pressurizer safety valves (PSVs), this is onlyapplicable to the events that open the PSVs. The loss of normal feedwater event results ina primary pressure increase of less than 20 psi and has peak primary pressure that is lessthan the opening pressure of the PSVs. Thus, the loss of normal feedwater event wouldprovide bounding primary pressure results for those less bounding events that do notexceed the PSV setpoints.

The single reactor coolant pump (RCP) shaft seizure/sheared shaft event (SR/SS)(UFSAR Table 15.3-3) has a peak primary pressure less than the PSV opening setpoints.The SR/SS event peak primary pressure was increased by 20 psi.

Peak secondary pressure increases are based on the LOCV events because it providesthe limiting secondary pressure results. The LOCV event results in a secondary pressureincrease of less than I psi for all secondary pressure cases. The SR/SS event peaksecondary pressure was increased by I psi.


NRC RAI #17The fourth paragraph in Attachment 2, page 7 states, in part, that:

The steam generator blowdown is not impacted by the revised CEA drop time...

The steam generator (SG) blowdown rate was dependent on the SG pressure, however,the impact of the revised CEA drop time on the SG pressure response was not sufficientlydiscussed in the LAR. Provide a more detailed discussion on the impact of the revisedCEA drop time on the SG pressure response.

Waterford 3 RAt #17 ResponseThe steamline break events are analyzed at hot full power (HFP), hot zero power (HZP),and Mode 3 and 4 conditions. The main steam system pipe break causes an increase insteam flow resulting in an increase in energy removal from the affected steam generator(SG). The increased affected SG energy removal causes a decrease in the overall reactorcoolant system (ROS) temperatures and pressure. In the presence of a negativemoderator temperature coefficient (MTC), the RCS cooldown causes positive reactivity tobe added to the core.

The initial SG pressures are based upon the beginning operating conditions. The poststeam line break SG pressures fall together until a Main Steam Isolation Signal (MSIS) isactuated and the Main Steam Isolation Valves (MSIVs) close. Once the MSIVs close, theaffected SG continues to blowdown whereas the unaffected SG pressure recovers ormaintains (refer to UFSAR Figure 15.1-37 as an example). The affected SG blowdownduration is dominated by the SG water inventory, initial temperatures, and break size. Theincreased CEA drop time impacts the power reduction post-trip and the amount of energydeposited into the primary coolant system. The slightly longer drop time means a slightincrease in the amount of energy added to the system. The slight increase in energywould have a minimal impact on the SG pressure and no change to the eventcharacteristics.

NRC RAI #18The second to last paragraph in Attachment 2, page 9, states:

The loss of feedwater flow event demonstrated that the impact of the revised CEAdrop time on long term parameters is insignificant. Hence there is an insignificantimpact on the plant cooldown to shutdown cooling conditions post-trip.

Provide justification for the use of the results of the loss of feedwater flow event to thesteam line break long term cooldown.

Waterford 3 RAI #18 ResponseThe steamline break events are analyzed at hot full power (HFP), hot zero power (HZP),and Mode 3 and 4 conditions. The main steam system pipe break causes an increase insteam flow resulting in an increase in energy removal from the affected steam generator(SG). The increased affected SG energy removal causes a decrease in the overall reactor

Attachment 1 toW3FI-201 5-0062Page 18 of 21

coolant system (RCS) temperatures and pressure. The post steam line break SGpressures fall together until a Main Steam Isolation Signal (MSIS) is actuated and the MainSteam Isolation Valves (MSIVs) are closed. Once the MSIVs are closed, the affected SGcontinues to blowdown whereas the unaffected SG pressure recovers or maintains (referto UFSAR Figure 15.1-37 as an example).

The increased CEA drop time impacts the power reduction post-trip and the amount ofenergy deposited into the primary coolant system. The slightly longer drop time means aslight increase in the amount of energy added to the system. The Loss of NormalFeedwater Flow event (FSAR 15.2.3.2) was chosen to evaluate the transientcharacteristics with ,respect to energy deposition and associated steam releases whichwould be applicable to all the events. The analysis showed that the differences in primaryand secondary system energy after reactor trip are insignificant. As time increases fartherpast the time of CEA rod insertion, the differences of the impact of the revised CEA droptime become negligible. NRC RAI#15 Figure 15-2 shows the energy deposition fromreactor trip time to 1800 seconds for the loss of normal feedwater event. For the long termresponse, the energy deposition change between the AOR and the revised CEA drop timeanalysis was 0.46% for the loss of normal feedwater event which is considered negligiblein the long term response.

NRC RAI #19The fourth paragraph in Attachment 2, page 16, states, in part, that:

The increase in peak primary pressure was based on the loss of feedwater long termresults which showed an increase in peak primary pressure of less than 20 psi...

Provide justification for the use of the loss of feedwater long term results to the total loss offorced reactor coolant flow event in determining the peak primary pressure.

Waterford 3 RAI #19 ResponseThe limiting events with respect to peak primary and secondary pressures are the loss ofcondenser vacuum (LOCV) and feedwater line break accidents. These are the limitingevents because they have the closest approach to the Technical Specification Section2.1.2 (Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressurelimit of 1210 psia (110% of design pressure). Since the LOCV peak primary pressure isgreater than the opening setpoint of the pressurizer safety valves (PSVs), this is onlyapplicable to the events that open the PSVs. The loss of normal feedwater (LOFW) eventresults in a primary pressure increase of less than 20 psi and has peak primary pressurethat is less than the opening pressure of the PSVs. Thus, the LOFW event would providebounding primary pressure results for those less bounding events that do not exceed thePSV setpoints.

The total loss of forced reactor coolant flow event (LOF) (UFSAR Table 15.3-1) has a peakprimary pressure less than the PSV opening setpoints. Thus, the LOF event peak primarypressure was increased by less than 20 psi.


NRC RAI #20The second to last paragraph in Attachment 2, page 18 states, in part, that:

The loss of condenser vacuum event peak primary and secondary pressureincreases would bound that expected for the uncontrolled CEA withdrawal fromsubcritical conditions.

Further, the second to last paragraph in Attachment 2, page 19, states, in part:

The loss of condenser vacuum event peak primary and secondary pressureincreases would bound that expected for the uncontrolled CEA withdrawal from lowpower condition.

/Provide justification for the use of the loss of condenser vacuum results to both cases ofthe CEA withdrawal events in determining the peak primary and secondary pressureincreases due to the CEA drop time changes.

Waterford 3 RAI #20 ResponseThe limiting events with respect to peak primary and secondary pressures are the loss ofcondenser vacuum and feedwater line break accidents. These are the limiting eventsbecause they have the closest approach to the Technical Specification Section 2.1.2(Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressure limitof 1210 psia (110% of design pressure). UFSAR Section 15.4 contains the reactivity andpower distribution anomalies. These events are less adverse than the loss of condenser

•vacuum with respect to peak primary and secondary pressure. Specifically, loss ofcondenser vacuum peak primary pressure is 2711 psia and secondary pressure is 1181psia [UFSAR Section 15.2.1.3.3.3] whereas the CEA withdrawal event peak primarypressure is 2287 psia [UFSAR Table 15.4-7] and secondary pressure is 1085 psia [UFSARTable 15.4-3].

The increased CEA drop time impacts the power reduction post-trip and the amount ofenergy deposited into the primary coolant system. The slightly longer drop time means aslight increase in the amount of energy added to the system. The assessment provided inW3F1-2015-0040 used the loss of condenser vacuum event because it produced thelargest post-trip primary and secondary pressure spike due to losing its secondary heatremoval capability. Since, loss of condenser vacuum event produces a more adversepressure transient, the slight energy deposition increase would be expected to have themost adverse impact on this event. Thus, the loss of condenser vacuum event resultswould continue to bound the CEA withdrawal event pressure transients.

NRC RAI #21Attachment 1, page 2, of the August 14, 2015, supplement provides the CEA drop timecurve for the 0.8 second holding coil decay time (Curve 3). This curve has the initial CEAmotion delayed 0.2 seconds over the revised curve {Curve 2). However, by 1.15 seconds,both curves are identical. Given that the CEAs insert via gravity, justify how the CEAs that


start dropping at 0.8 seconds (Curve 3) are at the same position as CEAs that starteddropping at 0.6 seconds (Curve 2) by 1.15 seconds (1 0 percentinserted).

Waterford 3 RAI #21 ResponseLetter W3FI-2015-0061 [Reference 3.5] presented the following CEA drop time curves.

W3F1-2015-0061 Table 3.0-1. CEA Drop TimesCEA Insertion Curve 1 Curve 2 Curve 3

(%) AOR Time Revised Time Increased Holding(Seconds) (Seconds) Coil Time

(Seconds)0 0.00 0.00 0.000 0.60 0.60 085 0.80 0.95 1.00_______

10 0.95 1.15 1120 1.25 1.45 1430 1.55 1.75 1740 1.80 2.00 2.00______

50 2.05 2.25 2260 2.3 2.50 2.5070 2.535 2.75 2.7580 2.75 2.95 2.9590 3.0 3.20 3.20100 3.5 3.50 3.50

Curve 1 was submitted to the NRC in letter W3P89-3094 [Reference 3.11] andsubsequently approved in NRC Technical Specification Amendment 58 [Reference 3.12].Curve 1 was based upon letter W3P89-3094 Figure A1-2.3 which was generated to boundthe measured data from the CEA drop time testing performed at the beginning of Cycle 3.Letter W3P89-3094 Figure A1-3.1 and Appendix Al-A provided the Cycle 3 CEA drop timetest data.

Curve 2 and Curve 3 are both based upon the Curve 1 shape. Curve 2 was generatedassuming a failure mode that increased the CEA insertion time with no change to the initialCEA drop start time. Curve 3 was generated assuming a failure mode that increased theCEA holding coil decay time which delayed the initial CEA drop start time. Curve 2 andCurve 3 both reach the same CEA drop time at 10% insertion. This is because Curve 2conservatively assumed the CEA slowed down by 0.2 seconds over Curve 1 by the 10%CEA insertion and Curve 3 conservatively used an additional initial delay of 0.2 secondsover Curve 1.

Curve 2 and Curve 3 are not intended to match the expected test data curves but to boundthe test data to ensure conservative analysis results.


3.0 REFERENCES

3.1 Entergy Pre-submittal Meeting Summary for the Revised Control ElementAssembly Drop Times [ADAMS Accession Number ML151 17A503].

3.2 Entergy Pre-Submittal Meeting Revised Presentation Slides for Control ElementAssembly Drop Times [ADAMS Accession Number ML15113A787].

3.3 W3F1-2015-0040, License Amendment Request to Revise Control Element DropTimes, July 2, 2015 [ADAMS Accession Number ML15197A106].

3.4 NRC Letter, Regarding License Amendment Request to Revise Control ElementAssembly Drop Times, Unacceptable with Opportunity to Supplement, August 3,2015 [ADAMS Accession Number ML15205A306].

3.5 W3F1-2015-0061, Supplement to Revise Control Element Assembly Drop TimesAssociated with Technical Specification 3.1.3.4, August 13, 2015 2015 [ADAMSAccession Number ML15226A346].

3.6 NRC CEA Drop Time Submittal Request for Additional Information, August 26,2015, [ADAMS Accession Number ML15232A275].

3.7 Waterford Nuclear Generator Station Unit 3, Technical Specifications, ThroughAmendment 245.

3.8 Waterford Nuclear Generator Station Unit 3, Update Final Safety AnalysisReport, Revision 308.

3.9 W3F1-2003-0074, Extended Power Uprate License Amendment Request,November 13, 2003 [NRC ADAMS Accession Number ML040260317].

3.10 NRC Safety Evaluation Report, Waterford Steam Electric Station, Unit 3 -Issuance of Amendment 199 RE: Extended Power Uprate, April 15, 2005 [NRCADAMS Accession Number ML05I1030082].

3.11 Waterford 3 Letter W3P89-3094, Technical Specification Change Request,Average Control Element Assembly Drop Time, August 14, 1989.

3.12 NRC Waterford 3 Technical Specification Amendment 58, Average ControlElement Assembly Drop Time, October 31, 1989 [ADAMS Accession NumberML021 760257].

3.13 W3F1-2004-0053, Alternate Source Term License Amendment Request, July 15,2004 [NRC ADAMS Accession Number ML042020294].

3.14 NRC Safety Evaluation Report, Waterford 3 Amendment 198, Full ScopeImplementation of an Alternative Accident Source Term, March 29, 2005 [NRCADAMS Accession Number ML050890248].

3.15 NRC Regulatory Guide 1.183, Alternative Radiological Source Terms forEvaluating Design Basis Accidents at Nuclear Power Reactors.

3.16 W3F1-2004-0071, Supplement to Alternate Source Term Submittal, August 19,2004 [NRC ADAMS Accession Number ML042360712].

Attachment 2 to

W3F1 -2015-0062

Fuel Thermal Conductivity Degradation Evaluation Affidavit

Affidavit to Withhold from Public DisclosureProprietary Information

Under 10 CFR 2.390

As Attachment 4 contains information proprietary to Westinghouse Electric Company LLC,it is supported by an Affidavit signed by Westinghouse, the owner of the information. TheAffidavit sets forth the basis on which the information may be withheld from publicdisclosure by the Commission and addresses with specificity the considerations listed inparagraph (b)(4) of Section 2.390 of the Commission's regulations.


W estin houseWestinghouse Electric Company

W estin house1000 Westinghouse DriveCranbenry Township, Pennsylvania 16086USA

U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643Document Control Desk Direct fax: (724) 940-85601 1555 Rockville Pike e-mail: greshaja~westinghouse.comRockville, MD 20852

CAW-15-4272

September 11, 2015

APPLICATION FOR WITHHOLDING PROPRIETARYINFORMATION FROM PUBLIC DISCLOSURE

Subject: CE-15-284-P, Revision 1, Attachment 2, "Fuel Management Adjustment to [Radial Fall-offf•cto Reserve Margin for Thermal Conductivity Degradation." (Proprietary)

The Application for Withholding Proprietary Information from Public Disclosure is submitted byWestinghouse Electric Company LLC (Westinghouse), pursuant to the provisions of paragraph (b)( 1) ofSection 2.390 of the Commission's regulations. It contains commercial strategic information proprietaryto Westinghouse and customarily held in confidence.

The proprietary information for which withholding is being requested in the above-referenced report isfurther identified in Affidavit CAW-15-4272 signed by the owner of the proprietary information,Westinghouse Electric Company LLC. The Affidavit, which accompanies this letter, sets forth the basison which the information may be withheld from public disclosure by the Commission and addresses withspecificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Conmmission'sregulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Entergy Operations,Inc.

Correspondence with respect to the proprietary aspects of the Application for Withholding or theWestinghouse Affidavit should reference CAW-I15-4272, and should be addressed to James A. Gresham,Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive,Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

amnes A. Gresham, Manager

Regulatory Compliance


CAW- 15-4272September 11, 2015

AFFIDAVIT

COMMONWEALTH OF PENNSYLVANIA:

SS

COUNTY OF BUTLER:

I, James A. Gresham, am authorized to execute this Affidavit on behalf of Westinghouse ElectricCompany LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and

correct to the best of my knowledge, information, and belief.

"/James A. Gresham, Manager

Regulatory Compliance


2 ~CAW-1 5-4272

(1) 1 am Manager, Regulatory Compliance, Westinghouse Electric Company LLC (Westinghouse),

and as such, I have been specifically delegated the function of reviewing the proprietary

information sought to be withheld from public disclosure in connection with nuclear power plant

licensing and rule making proceedings, and am authorized to apply for its withholding on behalf

of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.3 90 of the

Commission's regulations and in conjunction with the Westinghouse Application for Withholding

Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) 1 have personal knowledge of the criteria and procedures utilized by Westinghouse in designating

information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragr'aph (b)(4) of Section 2.390 of the Commission's regulations,

the following is furnished for consideration by the Commission in determining whether the

information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held

in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not

customarily disclosed to the public. Westinghouse has a rational basis for determining

the types of information customarily held in confidence by it and, in that connection,

utilizes a system to determine when and whether to hold certain types of information in

confidence. The application of that system and the substance of that system constitute

Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several

types, the release of which might result in the loss of an existing or potential competitive

advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component,

structure, tool, method, etc.) where prevention of its use by any of


3 ~CAW-] 5-4272

Westinghouse's competitors without license from Westinghouse constitutes a

competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or

component, structure, tool, method, etc.), the application of which data secures a

competitive economic advantage, e.g., by optimization or improved

marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his

competitive position in the design, manufacture, shipment, installation, assurance

of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or

commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer fu~nded

development plans and programs of potential commercial value to Westinghouse.

(1) It contains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system which include the

following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive

advantage over its competitors. Itt is, therefore, withheld from disclosure to

protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such

information is available to competitors diminishes the Westinghouse ability to

sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by

reducing his expenditure of resources at our expense.


4 CAW-l 5-4272

(d) Each component of proprietary information pertinent to a particular competitive

advantage is potentially as valuable as the total competitive advantage. If

competitors acquire components of proprietary information, any one component

may be the key to the entire puzzle, thereby depriving Westinghouse of a

competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of

Westinghouse in the world market, and thereby give a market advantage to the

competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and

development depends upon the success in obtaining and maintaining a

competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the

provisions of 10 CFR Section 2.390, it is to be received in confidence by the

Commission.

(v) The information sought to be protected is not available in public sources or available

information has not been previously employed in the same original manner or method to

the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is

appropriately marked in CE-15-284-P, Revision 1, Attachment 2, "Fuel Management

Adjustment to [Radial Fall-off~c' to Reserve Margin for Thermal Conductivity

Degradation" (Proprietary), for submittal to the Commission, being transmitted by

Entergy Operations, Inc. letter and Application for Withholding Proprietary Information

from Public Disclosure, to the Document Control Desk. The proprietary information as

submitted by Westinghouse is that associated with Fuel Performance, Safety Analysis

and the associated Thermal Conductivity Degradation methodologies, and may be used

only for that purpose.


5 CAW-15-4272

(a) This information is part of that which will enable Westinghouse to:

(i) Perform Reload Fuel and Safety analyses.

(b) Further this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers

for the purpose of performing reload fuel and safety analyses.

(ii) Westinghouse can sell support and defense of industry guidelines and

acceptance criteria for plant-specific applications.

(iii) The information requested to be withheld reveals the distinguishing

aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the

competitive position of Westinghouse because it would enhance the ability of

competitors to provide similar technical evaluation justifications and licensing defense

services for commercial power reactors without commensurate expenses. Also, public

disclosure of the information would enable others to use the information to meet NRC

requirements for licensing documentation without purchasing the right to use the

information.

The development of the technology described in part by the information is the result of

applying the results of many years of experience in an intensive Westinghouse effort and

the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical

programs would have to be performed and a significant manpower effort, having the

requisite talent and experience, would have to be expended.

Further the deponent sayeth not.


PROPRIETARY INFORMATION NOTICE

Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRCin connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning theprotection of proprietary information so submitted to the NRC, the information which is proprietary in theproprietary versions is contained within brackets, and where the proprietary information has been deletedin the non-proprietary versions, only the brackets remain (the information that was contained within thebrackets in the proprietary versions having been deleted). The justification for claiming the informationso designated as proprietary is indicated in both versions by means of lower case letters (a) through (f')located as a superscript immediately following the brackets enclosing each item of information beingidentified as proprietary or in the margin opposite such information. These lower case letters refer to thetypes of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a)through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE

The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted tomake the number of copies of the information contained in these reports which are necessary for itsinternal use in connection with generic and plant-specific reviews and approvals as well as the issuance,denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license,permit, order, or regulation subject to the requirements of 10 CFR 2.3 90 regarding restrictions on publicdisclosure to the extent such information has been identified as proprietary by Westinghouse, copyrightprotection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC ispermitted to make thle number of copies beyond those necessary for its internal use which are necessary inorder to have one copy available for public viewing in the appropriate docket files in the public documentroom in Washington, DC and in local public document rooms as may be required by NRC regulations ifthe number of copies submitted is insufficient for this purpose. Copies made by the NRC must includethe copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Documents

Waterford, Unit 3 - Control Element Assembly Drop Times ...Control Element Assembly (CEA) drop times associated with Technical Specification 3.1.3.4 for Waterford Steam Electric Station