Dresden Station Units 2 & 3 - Answers to AEC Questions - Amendment No. 7 and Amendment No. 8.~-... . .' ' . . .. __ . .. _. . 4 y I
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AMENDMENT NUMBER s FOR_ UNIT· 3
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UNITS 2 AND 3
AMENDMENT NUMBER 8 FOR UNIT 3
Received. w/Ltr Dated 8--~o-P~ • .. . • ~ ·-L.>'- • ,... ... .._
Commonwealth Edison·· Company
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2.:..7 '.',.
.J •• _ •••• ~-;,,'"· 2.)3.' 2;14 2.15 2.16 2. 17 2.18
3. 3.i 3.2 3:3 3.4
4. 4.1 4.2 4:3 4.4 4.5
l: 4.G \·: 4.7 .l 4~8
. i· 4.9 4. 10 4.11
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Subject
QUALITY ASSURANCE AND CONTHOL Quality Assurance of Conti·actors Quality Assurance Records . Class A Ve.ssel Listing ·vessel Electroslag Welding CRD Stub Tube Design . Vessel Stainless Steel Listing
SEISMIC DESlGN Design Loads Integrity of Internals ·Effect of Internals Irradiation
· · Instantaneous Recirculation Rupture . Identification 9f Clas·s.I Equipinent
· ECCS Compartmen.t"Flooding SuCtion Header Seismic . Detail of Torus Co_nnections Piping Seismic Design Torus Vent Header Supports Drywell Shell Filler Material Gore Shroud and Jet Pump Design Vessel Pedestal Design·
· Dynamic Model of Structures Building Design Criteria ClaE;s I Seismic ~eismic Inputs Seismic AcQelerations.
TORNADO DESIGN · Tornado Structure Failure
Integrity Against Missiles . .Max. Depressy_rization, React0r Bldg. Missile PI'otec1J~n ·
- REACTOR CHARACTERISTICS
Calculations Model . ,,.
Thermal Margin Evaluation Fuel Assembly Orificing Cladding Strains
. Peaking Factors Transients From Rod Block T-G Trip Transients Excess· Reactivity Flux Shaping · Load Transients Viberation Damage on Internals
Later Later L 3-' 1.: Later 1. 5-1 i.6':::1
2.1-1 2 .. 2-1 Later 2. 4.-1 2.5-1 2.6-l
. · 2. 7-1 ;.:'.·2:. 8-.l . 2 .. 9-1 2.10-1 2.11-1 2. 12-1
. 2.13-1 2.14-1 Later
.2. 16..:.1 2.17-1 2. 18-1
3.1-1 3. 2-1 3.3-1 3.4-1
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7.
8.
9.
.5, 4· 5.5 5.6 5.7 5.8 5.9 5.iO 5.11 5.12 5.13 5.14 5. 1.5
.5.16 5.17 s.i8 5.19 5.20 5. 21. s: 22. 5·:.23
6.1 6.2 6.3 6.4
8 .. 1 8.2.
9. 1 . 9.2
: flPCI Control System HPCI Isolation Signals
. 'HPCI Performance Data and Analytical Meth9ds · .. HPCI Water Slugs. and ·system' Failur.e Analysis.
HPCI Moisture Removal System Design Basis HPCI Line Breaks Analyses . HPCI Depressurization Effect and Sequencing
. HPCI Pre_;Op Tests · HPCIS Radiation Monltoring
· LPCIS .Control i,oglc Design Containment. Spray Controls
.. Shroud and jet Pµmp Leakage During. bPCIS Operation · Jet Pump Operation in Two Phase .Environment · Core Spray Test Program ECCS Leakage Detection ECCS Break Size· Analyses
.·Predicted .Peak Clad Teri:nperature · 'Smail· Leak Analysis· . ECCS Ope:r;ation with Off Site Power ECCS ComponentCooling·
.. Pipe Whip Criteria .. E.CCS Components Environmental Tests
. . . .
co~tROL ROD DRIVE HOUSING SUPPORTS .. Design Load Combination Buckling Hesista~ce Single· Failure Performance Capability Pre-Op and Surveilence. Tests
. . .
OPERAT_iONAL MATTERS Periodic Inspection Reactivity Anomalies Pr.i mary System Leakage Safety System Adequacy Shared System Faults Vessel Surveillance Area Monitoring Plan
·.Operating Procedure Review
. 5. 8-1 5.9-1
. 5. 10-1 5. 11-1 Later · 5.13-1 5.14-1 5:15-1 5;16-1 5,1Fl 5.18-i 5.19-1 5. 20-1 5.21-1
.. ·. 5. 22-1 ·. 5. 23-l
6.1-1 Later 6.3-1 6.4-1
. L1-1 • Later· 7>. 3-1
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Subject ·.·
·Monitoring of Release Limits Procedure for Loss of Power .
· Ope1~atihg Procedures . Operator Training . Ite1i1s Requiring Further Analysis Vital Systems Tests
. EMERGENCY POWEH SYSTEM Design Change Consideration
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Page
10: 1-1
1. i Quality Assurance of Contractors 1. 1-1
1. 2 Quality Assurance Records 1. 2-1
1. 4 Vessel Electroslag Welding 1. 4-1
2. SEISMIC DESIGN
2. 15 Building Design Criteria 2. 15-1
5. EMERGENCY CORE COOLING SYSTEM
5. 12 Containment Spray Controls 5. 12-1
6. CONTROL ROD DRIVE HOUSING SUPPORTS
6.2 Buckling Resistance 6.2-1
• 9. OPERATIONAL MATTERS
9. 12 Operating Procedures 9. 12-1
9. 15 Vital Systems Tests 9. 15-1
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1. 1 Question
Provide a listing of all organizations, contractors and subcontractors, with major onsite
functions in the construction, inspection, or testing of essential structures, systems, or
components. For the major products supplied to site, identify the supplier and the organi­
zation responsible for assessing the supplier's control of quality. Indicate the specific
responsibilities of each organization for assuring adequate quality, and discuss the program
and procedures used by each organization to implement its responsibilities. Include for
each a functional organization chart, indicating the titles of personnel performing quality
program functions. Specifically indicate the degree of participation of the Commonwealth
Edison Company in assessing and reviewing the effectiveness of the control of quality by
each organization.
Ans·wer
This answer is contained in the Dresden 2, 3 Quality Assurance Report submitted as an
appendix to the FSAR .
1. 2 Question
Identify the records and data used and maintained as evidence of the control of quality in
each onsite ·activity in the construction, inspection, or testing of essential structures,
systems, and components. For the major products supplied to the site, identify the records and data used in assessing the supplier's control of quality. Indicate, for the
records and data identified for each activity and major product, the nature of the observa­
tions recorded, the number or extent of the observations made, and the applicable code
or standard, if appropriate. Indicate the geographical location and availability of any
essential records and data not maintained onsite.
Answer
This answer is contained in the Dresden 2, 3 Quality Assurance Report submitted as an
appendix to the FSAR .
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·, 1. 3. Question·
Pro,vide a tabulation of all the nuclea~ pressure vessels designated as ASME
ci~ss :A for the facility. The tabulation shoul.d include a notation of the ..
. . ··. e~tenf·to which each .of the vessels complies with each of the 34 supple-,
meritiry criteria 'in "Tentative Regulatory Supplementary C'dteria for .
. ASME Cod.e_::Gonstructed .Nuclear Pressure Vessels,'' issued. by AEC P.ress
Release No .. IN-817~ dated August 25, i967:
·.Answer
The pressure vessels in the Dresden 2 and 3 facility .that are designated as
ASME :Class A ·are the two Reactor Pressure Vessels and the tu~e sheets·
of the Isolation Condenseri:;, There are two tube sheets per isolation ···
condenser .•.
· The degree of co~pliance .to the 34 supplementary criteria in "Tenta.tive
Regulatory Supplementary Criteri.a for ASME: Code Constructed Nuclear . . - . . . .
Pressure Vesselsu issued _by AEC Press Release No; IN-817; dated
Augu_st 2'5, .1967, IS covered in a document submitted to Harold L .. Price,.
USA.E ~nMarch 13; 1968, by George:StathaJds, GE-Ai>Eri. ··The titleof ·. . ~. .. .
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1. 4 Question
As indicated in our letter to you dated December 21, 1967, and in our 'meeting on March 21; 1968, ,additional iilformation is required concerning the electroslag welding
process employed in the fabrication of the reactor pressure vessels for Dresden Units 2
and 3. In addition to the information requested in the above letter, provide sufficient test
data from actual production weld specimens from similar reactor_ vessels to permit a correlation with ASME test plate data and to establish the variability of the process. The
test specimens should include both tensile and Charpy V-notch specimens from the center
of production welds. The Charpy V-notch specimens should be tested at.a sufficient num­
ber of temperature to demonstrate the upper and lower plateaus and each temperature
· point should include at least 12 data points. Include information defining the chemical
composition of each plate from which a weld specimen is, taken and of the weld rod material ,
used in the process. The average va,lue and range should be given for the chemical com­
position. Correlate this data with the chemical composition for the plates and weld rod
material used in the vessels for Dresden Units 2 and 3. Also provide photographs of
macroetched speCimens taken in the longitudinal direction across the center plane from
one heat affected zone to the other of production welds.
Answer
This answer is contained in the Dresden 2, 3 Electroslag Welding Report submitted as an
amendment to the docket.
1. '5 Question
We understand that the control rod stub tube for the Dresden 2 and 3
, .
. Creek Nuclear Power Plant. Accordingly, a detailed evaluation of the
design in terms of allowable stresses and conditions for stress corrosion.
should be provided. An analysis of the expected fabrication stresses
from radial contractions induced by the field welding, and from differ­
ential contractions following ih:--shop annealing of the pressure vessel,
should be included. Describe the quality control practices which will be
followed during the field welding. Explain your rationale for the selec­
. tion of any liquid solutionEf to .which stub tube materials will be exposed
prior to operation.
. Answer
The CRD stub tubes used in the Dresden 2 and 3 reactors have the follow:..
ing .characteristi1::s: . .
a. · The material will he fuconel 600, per Specification ASME-SB-167.
b. The attachment to the vessel bottom head will be means of welds
of·the type allowed by Section III of the ASME Pressure Vessel
Code. As currently planned, the stub tubes in Dresden 2 wiil be . .:
set into counter-bores illustrated in Figure 1. 5. 1, while ih '
Dresden 3 the tubes are currently planned to be of the set-on
design as illustrated in Figure 1. 5. 2 .. The design differs from
that used in reactors currently under construction principally
in the material used (fuconel instead of Type 304SS) and in the
location 'of the field weld (4-1/4'' minimum height above the shop
weld, compared to 15/16").
SH .NO.
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.. DRIVE HOUS\NG.·~\ AS0E-'-SA 3\2. · ... · \ TP.304 ·.. . \
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Detailedanalysis of stresses in the designs have not been completed;·
however, based on experience with similar designs and rough calcula-'
tions, stresses will be within the limits of Section III of the ASME Pres­
sure Vessel Code .. An analysis and description of a similar stub tube
design are included in a Topical Report, which will be available about . . ' . . . . ; . . ~ . . . , .
September 15, 1968. This report also includes. a description of the
Quality Control Procedures to be followed in the fabrication and assembly
of these components.
The field weld will also be of the type allowed by Section III specifically . .
· illustrated in Figure . Based on experience with other welds of
similar' design shrinkage is expected to cause 2-3% strain and residual
stresses above the yield point of the ma~erial. Strains and stresses of
this magnitude are present in essentially all weldments, are not amenda­
ble to calculation, and are of the "shakedown" type .which do not .exist
after one or two operation cycles.
The fluids to which the stub tubes will be exposed consist of the following:
. 1. Cutting fluid --' a water soluble chloride and sulphur free emulsion
approved for Navy reactor components ..
· 2. Dye penetrant and remover - Turco Dy..:Chek or Magnaflux SKL-S.
They contain no chlorides ~r sulphur except as trace impurities.
3. Acetone and/or methyl alcohol.
4. Deionized water, with· and without TSP in low concentrations
( ..... 500 ppm).
5. Condensation.
. The rationale for their selection is as follovrs: .
a. . They_, or their equiva;ents - are llfCessary or ~navoidable. . .": ·~ • ~ .; •t .: . . ~:l
b. . Each .has been dern9n~trated by lab,pratory tests)or by experi.:. '
ence to be incapable of causing da~:1age to stub materials. TSP
has. b.een shown to inhibit stre~s _corrosiori,cracking of furnace.
sensitized 304 SS.
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Provide a list of all stainless steel mat.erials,. iit~lu.c:lirtg specification,
grade, condition, and vendor' present in the r~actor pressure vessel .
and its attachments and which have b~en subj~cted to the stres.s-relieying
heat. treatment of the vessel. · ·
Answer·
Stainless' steel material in the Dresden Unit 2 and 3 reactor vessels . . . . .
. which. have been subjected to stress-relieving heat treatment is indicated
in Table 1 .. 6-i.
FURNACE SENSITIZED STAiNLESS STEEL IN DRESDEN 2. AND 3 VESSELS
· · Description·
Safe End~- J~t Pump Inst. Noz~le. . .
&i:fe End .:.. Recirc. Out Noizte
. Safe End -ReCi~c. In. Nozzle
Safe End- Is(). Cond. Nozzie
Safe End - Cor_e Spray Nozzie . . . . ' . . .
Safe End - Gore 6P Nozzle ·. ·. :· . . ·.· .
Safe ~nd -: CRD Hyd. Ret. Nozzle
· . Steam Dryer Suppt. Brkt'. ·.
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SA-182, GRF-304 _
SA-:1S2;. GRF-304
SA-182, GRF-'304
SA-336, .CL-F8 ...
.*I Mclmies Steel· . _-4 -Allegheny Ludlum _ · .2 · Davidsmi . · ... 5 Alloy Flange & Fitting
3 ·33 & w . " 6 . cann & Sati.le
~:::c:;::*-*A .. ::Alinealed · . · ~ ·-~- _:-~Q··~:.WAfer_;Qiienched ·
VENDOTI* Condition**.
1 _A
2 .WQ
3-· A
3: A·
specimen' Holder Bracket .
. ,Specimen Roider Bracket .
Iriternal Cladding***
• ·.'.!:Y.Qg ..
· 4 . Allegheny Ludlum
3 B &W ..
5 Alloy Flange & F)tting .6 Cann & Saule.
• ***Appli~d by the subm~rged arc process usi11g Arcosite S-4 fluz •.
. VENDOR*
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• .2.1 . Question ·
For all Clas~ I structures, systems and components provid~ the design . . • . . J \ . .
basis load combinations and the stress and deformation.limits for each . . \
combination.
Answer· I ..
. All Class I structures, systems and components are desig~ed to accommodate ·
• the load c~nditfons ~nd sfress c~ite~ia was presented in paragr_~ph 12. i. I. 3 (Z9' 7; I .
of.the FSAR. - -z9, 2-4-Z) .
Jdentlfy ali specific reactor ·inte,rnals whose 9ontinueci integrity 3:nd/ or ··
operability is essential to prevent an a~cident anc.i/ or mitigate the conse- .
· quences. of a~ accident should it occur. Pro";ide calculated or tested
maxlinum limits o.f deformation or stresses; at which inabilify to func- ..
tion occurs, for. each component identified. Also, supply the qalculated · .· ..
or. tested maximµm design valu~ and the expected defor~~tiqn or stress.
In all .cas~s ;,iden~ify th~ applicable loading 'combinaticm and· state the .
proposed margin of safety~ ..
The reactor1nternals which must maintain their functional integrity to
. assure s'afe. shutdown following" the various postul~ted accidents are the
. . foll()Wing~:
~· Fuel _cha~el-core support coniplex.· .·
. • 2. Control rod co~trol assembiies. I ·,_.,
B. · Stand..'.by Liquid Cont~ol • ·
2. : Emergency Core Cooling Systems;
A. LPCI System ~- ',
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' 1. Gore shroud and b~fle, relative to the ability to maintain
water level in the cqre .
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'2. Jet pump structure relative to the ability to .intro·d~c~ and
maintain a water ievel in'the.core.
34, 14
C~ Core Spray System
1. Core spray piping and Sparger jn the Reactor Pressure
·vessel. .
2 · · Arrangement of the core support cc>"mplex~ relative to its.
ability to accept water from· th~ core spray .
· Based oh analyses of the reactor internals during bbth normal and . I
accident conditions, it was determined that ~tresses in the indfvidual
components. are limiting, except .in .the follO\ying areas ~here deforma­
tion-is the controlling parameter:
1. ··. Defl~ction of the.fuel channels under accident pressure conditions, . . .
to an amount substantially less than would prevent control rod drive : . ·: .
·· insertion. The maximum ftictionai force exerted on the fuel : . . . ·.. '
channel by the control rod· (as a result of interference), during a . . . l . . ~ . .
design basis.accident is less than: 100 lbs. The minim\im force
exerted by the control rod d~i.ve on a co~trol blade is 3000· lbs.
Therefore, control blades. can be fully inserted against the forces
··of fuel channel deflection under the ~ost severe accident conditions.
Under the above control rod insert conditions, the fui1 b~dles which
. weigh about 700 lbs. will not be lifted due to the resist!~ insertion ·
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2. . Horizontal deflection of the Control Ro.ct Drive housings is limited . ~ . : ' .
to a value which through test has been demonstrated not to impede . . .
. control rod insertion.
DRESDEN 2, 3 · 2;2-3
3. D·~fle~tion of the core plate and lower. gr;id assembly is Hmited under
normai operation to preClude taking up \iertical clearance between the ·. . . . . . . ~ ..
core plate and control rod guid_e tubes SI~ that the core bypass leakage
. -flow can be predicted: This re~ults ih ~tresses that' are below yield
: even during accident c'onditions. The mp.x:i.murr1 deflectiOn of the core
plate under iccident conditions. is limite~ fo 0.12.5 inches,-which
represents a considerable factor of safe~y·below the deflection .at
.. which the. core plate and guide tube coul~· co~e into contact. '.
The -~aximum; value of primary stress in reactor internal components ..
. "generally results from the large pressure difference created when·either. . . . . . :., ' .. . .
the recirculation line. or the steam lin~. are qompletely severed. A
discussion of these two accidents is gjven -in· some detail in paragraphs ·
3 ~ 6. 3 •. 1-. 2 and 3. 6. 3 .1. 3 and values of calculated pressure difference·
.versus desigh' pressure capability for ~ajar reactor internal co~pqnents . ~~e included' in .Tables ·3. 6. 2 .and 3. 6. 3. of the Dt.esden 2, 3 FSAR to show .
.. . . ·. ,, . . ·the margin .of safety that exists below the A_SME Se~tion IIIHmits.
The margin ·of safety for these· components which actuaiiy eXists, based
up~n the APED desigri criteria* for'reactor 'tnternals, Js equal tC> or
gr~·ater than the·· margiil specifi~d in the tabl~s. The loading co~binations, ·
arid stress and deformation limits 'for r~actc»~ internal components are . - ~· . . . . . . .
· · also discussed _in these c~iteria.
The sen~itive point within the reador press4re v:essel which is most·· : : . . . : : . ~ ·. . . . . . . . . . ..
affected by operation 9f the emergency core cooling systems (HPCI
~nd LPCI) is. in the area of the jet pump to b~fle plate jqint. The stress . . ' . . ' . .
. ··. ~Qd fatigue evaluation of this location is discussed 1n the answer,,to
qu~stions. 2 .. 12 and 7. l. · :,.
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2. 3 Question
"" For reactor internals, provide information that will permit evaluation of the effect of irradiation on the material properties and on the proposed deformation limits.
Answer
The reactor internal component which receives the maximum irradiation is the shroud
at the inside surface opposite the midpoint ofthe core where the·total integrated neutron
flux at end of life ts 2. 7 x 10 20 nvt (> 1 Mev). All reactor internal structural members
located in high flux regions, including the shroud, are constructed of 304 stainless steel
which does .not suffer from irradiation embrittlement. · It does experience hardening and
an apparent loss in uniform elongation but not a loss in reduction of area, Since the reduc­
tion in area is the property which relates to tolerable local strain, it can be concluded
that the effects of irradiation can generally be ignored. However, even on the basis of
changes .in the total elongation, one would conclude that this material at 2. 7 x 1020 nvt
integrated flux would be capable of about 15 to 20% elongation .
34.17
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. State ff the recirculation line rupture, described on Page 3. 6-7 'of the
• FSAR, ,was assumed to be instantaneous'.
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Identify all Class I structures~ systems and. components .located within
Class II structures. Please describe ,and justify the provisions taken to
e~sure th~t Class I requirements are not compromised.
Answer
. . - ' . . . '
. ; were def?igned. to Clas.s II requirements and have been investigated to assure
·that the integrity of the Class I items is not compromised. Class I struc- . . . . . . . '
tures, systems and components located in Class II structures include the
· . control r:ooni, standby gas treatment system, and the· standby e.lectrical
power syste'ms comprising of the station batteries, diesel generators,
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· ·.· 2. 6 · QuestiOii · · . :· .,,.:
A .;_umber of motor""'.pump units are l~cated on the io~er levels of the reactor·
b~ilding. ·Many of these sys~ems a~e \tirai to safe shutdbwn arid dontain- ·. .
merit in the· event of an 3.ccident .. What. steps have. been htken to provide· .
· . · protection for these mofur pump units. from flooding through possible feak- ·
··· · · ·. agefr6tn the t9rus or from .vilrious piping sy~tems? . .. . · ..
Answer .. ·...,.:
The torus,. torus header, and associated piping systems are considered
19gi~al e,xtensions of tll.e pri~ary containmEmt. and, and ~uch, must meet
tlie same de.sign; surieillance·and: testing criteria as the prUnary contain- ..
. . ' . . . . . ~ .
. . firmly supported from the 'lower portion of th~ suppression chamber at
. , fifteen positions' ar.ound the ctrcumference. · Maximum protection of-the·
. ring header and. piping is afforded by its phys~cal locatio.n:; Le. , adjacent
·to· the suppression cham~erin a_ reinforced c9ncrete room containing no
high pr~ssure piping or inec~ical equi~ment. Be~ause of the above design . . . . . . . . . .
·considerations, -ruptur~ ofthi_~ lo~ pr~ssure_ system is considered
· incredible. . . . .. : /
Iii order to remove postulated-Water lerumge from vaive stems, flanges,
etc., the ~eactor building hasbeen equipped.with two floor drain sump pumps·.
· ·.each_ having a c~pacity of 50 gpin fot a tota~ r.emov~l Gapadty of 100 gp~'. -··
Leakage of 100 gpin correspbnds to a syste·:m rupture equivalent tO-a one . .·· I ·, ' , '
i~~J:i diameter hole. PhysicaJjnspectionof the ~on.is. area is made approxi- .
•· · <_ Il1ately: every eig~t hours._ Excess operation of the_ sunip pumps woUld ~lso .· .
·. b~. Ii.oted in the :r~aWaste facility. In additiOn,_·.torlis water'levelis "con~~_:;.., .. ·.··
.. ··. '· . ;: ··.· :: . ·.. . .. ·.:· ·-···, .
. ·, .. · .. " ~· .·, ..
. . ·:. ·; _~.,: .. .., ·· .. ... : ·: . ·<·
·· ... :, .. _ . .. . ·:· .. ·'·.
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· 2.7 Question · . . .
A suction .header is used with the torus for the pressure suppression
system.· Jnforination is requested on.the dynamic analysis of the suctiori...­
header-torus system' and the details that would ensure _that difficulties
with this system will not occur un:~er earthquake lo.ading .• ·· . ' ~ . . .. .
Answer
· ·.· A dynamic analysis of the suppression chamber ring header was performed
. to confirm that the response of the he.ader with the recommended restraints .
will remain within the permissible limits.
The spectral acceleration used for the various modes of vibration of the
header were obtained fro~· an acceleration response spectrum for 0. 5%
damping generated for the suppres~ion chamber supporting the suction
header.
For analysis, the· ring header was separated . into segments and two seg-
. ments were investigated due to symmetry. The two segments were·
mathematically modeled to determine the natural periods of vibration and
c~rresponding _mode. shapes. · Using' the· spectural accelerations from the
response ~pectrum equivalent static seismic induced forces, displacements
and resulting stresses were. det~rmined. The analytical m~thod used in
calcul,ating the three-di~rnnsional stiffness takes int~ account the atfects_ .
· . of' flexural, torsional,. shear and axial deformations .
. ··The analysi~ indicates twelve hydra~lic snubbers are required._
The results of the seismic analysis are as follows:.·
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2. 8 Question .
The point of attachment of the suction header to the torus in Dresden Units ·
2 and 3 appears to be at a significantly higher elevation thah has been pr~-··
viously observed. Details should be provided with regard to the. elevation· . \ .
· • of ·the attachment pf the suction header with respect' to.the tor.us, .particularly. . . . '· . . . .
with respect to internal structures in the torus' such as ring stiffeners.
The details of•the T-seetion in the suction header and of the reinforcement
employed at the attachment point to the toru~ also should be provided •
Answer
Details of the. attachment of the section header to the torus are .given in
sketches 2. Sa~ b, and c.
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' .
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2.9 · Que~tion
.Provide a desc·riptiori of the procedure by Whifh the seismic ~esign of the
pit»ing was carried out including a description of the treatment of the various
sfag~s of analysis, the manner in which the lq1cation of snubbers was decided,.
and how valves were treated in the analysis. ·· . - ..
Answer . . . . . . , . I
A dynamic analysis was performed on the ma:i,n steam lines, f eedwater lines
and recirculation piping and a description of the procedure and method of . . ,·. '
' analysis is as follows·:
Method of Analysi.s
Each pipe loop was idealiz~d as a mathematical model consisting of lumped .
·masses separated by .elastic m~mbers. LumI?ed niasses were located at
. selected critical p~ints as required to adequately represent the dynamiC and .
elastic characteristics of the pipe s;sterr:r. Using th~ elastic properties of
the pipe qetween successive mass points the flexibility matrix for each
pipe loop ~as determined. The flexibility cal_culations inducted the: effects
of torsional, bending, .shear and axial deformations. Also included was the
.~hange in flexibility due to the curved members ... The methods, for calcufa- .
'ting these curvature effects are discussed later in this report.
.. After the flexibility arid mass matrices of the mathematical model were.
obtained, the frequencies a'nd mode shapes for the first seven modes of
vibration were determined .. Seven modes were. used in this analysis because
the effeCt. of higher modes was f?und tobe negligible. The mode. s~a·pes .
,.· ..
.in which:
= frequency for the n mode ' ' ' ' . th .
- mode shape matrix of then rriode
2.9-2
After the frequency was det.ermined for each mode, the corresponding ·
spectral acceleration was read from the floor response. sp~ctra for the
reactor building. · These floor response· spectra were given for two direc­
tions of earthquake 'motion for each floor. level in the building.· These
spectra were considered representative for the Dresden StatiOn since the ·
two plants have equivalent reactor buildings, foundation conditions and
ground motions. The response spectrum used for this analysis was for_ ~ ' . . .
Mass· 6 N-S direction with 0.005 damping ratio. This spectrum was chosen
. because. it resulted in the maximum response for the pipe systems .. Using
these spectral accelerations, the response for each mode was found by.
solving the following equation:
·' . . th = response of the n mode
part:lcipation factor for th~ nth mode = ' th . .·
= spectral acceleration for the n mode . I , ~ • •
D = earthquak:e direction matrix.
. . 2 EM. </J. ',. .. 1 in
•
·....... /.
' . . ' .
,Using t~ese results the maximum di'spla9emei;its for each mode were
_dete:rmined for, each m~ss. by the following relatfonship':
V. = (/). Y inax in in n
in which:
v .' = -maximum displacement of mass i .·for mode n in . . . .
The total response for each mass was determined by taking the square root . . . . . . . ' .
· of the sum of the squares of the maximum ·deflection for each mode:
.·in which.:
-·
Vi' = inaxirm.im displacement of mass i due to seven modes
· The iilerfaa forces for each dfrection of earthquake were then determined
from:
v - . maximum displacement matrix . . ' '
The internal.forces, moments, .and stresses for the pipe loop~ can be
:. }
' .
·Computation of Stresses
All val~es f~r forces and mometits .are for the global coordinate syst~m. ·.
After cori.verting these va.lues t9 the .member coordinate system, the pipe
stresses are determined i.n accordance with the following equation given
in Reference 2: ·1 ·.
Mt z B
equivalent resultant stress for pipe
resultant bending stress at joint = B M~/Z. resultant torsional. stress. at joint = M/2 Z
. resultant bending moment . ·
resultant torsional moment
Effect of Curved Members
The flexibility mat;rix of the pipe system l.ncludes the effects of curved
members. a~d elbows .. · This curvature effect depends on the bend character­
istic as given in Reference 2 and shown b~low:
in which:
r mean radius of pipe m R = radius of bend
h tR 2· r .
. .
Th~ flexibility factor gives the change in flexibility due to the curved .
members and is given by:
K 1. 65' -·--
. .
The stresses in curved pipes will differ from those calculated for straight . . . .· I
pipes. with eqlla.l bending moments. This stress increase is given by the
~tress .intensification factor:
in which:.
DescripUoil of Computer Program·
· An of the calculations outlined above were perf ormeci with the aid· of a . . .
. dig~tal computer.· Th~ computer program employed· has been writt~n specifically for the analysis of three dim.ensional piping systems. . . .. .
The input_ data for this program consists 1of the coordinates of all c:fitical
joints and valves in the pipe system' inclu~tng the coordinates of the joints
.selected as ltimped masses. · Additional input consists of the thickness,
.· diameter, weight and elastic modulus of the pipe loops. The computer
34, 44-
then calculates the stiffness and force transformation matrices, mode . .
shapes, frequencies and inertia forces. Using the inertia forces as loading
c.onditions, the internal forces and moments, displacements and stresses
.are then computed and p;inted out •
. For the dynamic response of the pipe l~ops, ~n analysis was made for both
the X and Z pirection earthquake. For each ?f these loading conditions, a
constant vertical Y Direction acceleration of 0. 067 gravity (two-thirds ground
· motion) was combined with the horizontal.
The results are in the form of coordinates, intetnalforces, moments, dis-,. ~- ' . . .
placements, reactions at intermediate supports and stresses .. The results
are given for four different cases:
L Combined X and Y Direction Earthquake . . . .
2. . Combined Z and Y Direction Earthquake
3. .Comb.med X and Y Direction.Earthquake
4; Combined. Z and Y Direction Earthquake
·. Other class I p!ping systems were investigated for. seismic support~ by
the following method:
Lateral deflection and force evaluation curves have been develope.d based
.· on 'the. natural period of. the piping system. as a fonction of .pipe size
(diameter and schedule) and span. There are three criteria.that must be
• ·. satisfied: .·
L Piping systems lateral~supports a.re designed to avoid the resonant
. range of the supporting structure. Knowing the period ~f the struc­
ture, the period of the piping system' is established to determine ; . . . . .
whether it is rigid, .fleX:i.ble, or resonant. Adjusting the span between
34.4S
DRESDEN 2, 3 2.9-7
lateral supports adjusts the natural periqd of the piping system to the
desired period range.
. . . . . . . .
2.. The span between lateral supports is checked agaihst a deflection limit L .
of 480 . This deflection limit restrictio!J. may govern in small pipe
sizes. ·
3 .• The span between lateral supports is checked against the allowable span ' . .
. for various pipe .sizes subjected to a horizontal load of 0. 5g. The
. seismi,c stress is not to exceed 1500 psi.
Additional cU:rves,have.been developed to determine the reaction on the
supports. Considerat.ion is given to magnification of response due to in.:.
stailed elevation, valves, branch lines, and bends. ,
. \ . . · Location of snubbers is based upon engineering judgment satisfying stress.
· . and deflection limitations before the analysis is performed. If the original
snubber location is pr'oved to be unsatisfactory' additional snubbers are
added ·at points of maximum deflection or snubber locations are charged in·
order to bring the stresses. '\Vithin the criteria .. ·'
· .... - .
•
2.10 Question
Indicate the means by which lateral support is provided for the torus vent
header from ,which the downcomers originate and which is connected to
· the draft tubes leading into the drywell.
Answer
·Each draft tube is.welded to the drywell and then passes through a larger . . . . . . . . ..
diameter; pipe to the torus vent header to w?icb it is welded. The design
of the 'd~ywe}lconsiders the vent system (v~nt pipe, vent header, and
downcomers) as an appendage to the drywell. ·The seismic induced lateral
34/r1
. ' . ' . . . . . ,. .
loads are taken by the dryweU. The l~rger diameter pipes through which the . . .
draft tubes pass ii re welded to the torus.. A bellows seal is then used to . . . .
connect the draft tube to the larger diameter pipe. The bellows permit
lateral niov~ment of the draft.tubes and act as part of the ccmta,innient . . ·, ' . . .
pressure. envelope.
dimensions, and procedures employed in fHli~ the space between the
reactor drywell and the concrete shield surrounding the dryweU.
·Answer
2.11-1
The combination of materials used to fill the ?ry\vell expansion gap was
generally defined in the Dresden U~i~ 2, .3 FSAR on page 5 .. 2-28 beginn.ing
with the last paragraph. , Following is a specific description for eacp ·of
the ~ate~ials utilized and a narrative of the procedures employed during
the installation of the materials around the drywell ..
Material Description
1. Polyurethane Foam: This material is a polyester base flexible
. polyurethane foam manufactured to exacting controls frorri refined
raw materials to ·produce .a quality foam suitable for use in az:eas of
high radiation. Sheets used conform to the following requirements:
a. Base Specification: MIL-PPE-200F.
b ~ Chemistry: Isocyanate fo·am formed by reaction of polyisocyanates
with polyester polyols.
d. Thermal Value: . 26 K factor.
e .. Service Temperature: 285° F.
f. Physical Properties:
(2) Elongation - 100%
(3) Compressibility - 35%at 1.0 psi maximum.
(4) Compr.ession Set _: 10% .at 50% compressibility.
g~ Sheet Size·: 2-1/4" x 2' x 8 I with tolerances as specified by .
MIL-C-26861.
. . .
of brushing consistency. Application of tpe cement was made to the
drywell shell over the entire contact aref for each foam sheet at the
thickness recommended by the manufacturer.
3. Sealing Tape for Foam Sheets: Epoxy impregnated fiberglass tape of
a width not less than 3" wide. This tape was· installed over all joints
. of cover panels as the panels ·were placed, with the tape· centered.on . ' . ' . "
the joints and with a lap of not .less than l" at all ends of ~he tape.
4. Fibrous Glass - Epoxy pre-molded coven panels: These panels are I
made of fibrous glass in chopped fiber fo~m with fibers 3/4" to 1" in
leIIgi:h with an isophtallic polyester resin' as a binder. · Properties of
the mix are as follows:
a. Flexural Strength - 16, 500 psi.
b. Fl~xural Modulus of Elasticity..:, 5.8x105 psi.
c. Tensile Strength - 8, 00 o psi.
.. d •. · Barcol Hardness ..,. 50
e.-. Thickness - 1/4" mini~um and 3/8" maximum; these panels were
·shop fabricated in sections using field measured moids for each of
the cylinder, kz:iuckle/, and spherical portions of the drywelL
5. Steel Anchor Fasteners: These were 4" x 4" x 1/4" stee_lplates with
1/2' diameter steel studs welded to face of piates. Studs were placed
at 24" centers in both directions_ and_ the length of the studs was
sllificiertt to projec_t 1-1/2" from the front face.of the cover panels.
Installation Procedures
••
.
' .
fiberolass cover· panel.·
FiOi.lre 2. II. I
fibergla~s cover panel·
• • ~I
'···· ,r
I
1. . Apply adhesive cement to the drywell shell ..
. 2. · Apply polyurethane foam sheets over adh~sive cement in sheets
approximately 2' x 8', pre-cut as requir~d to fit around penetrations.
The polyurethane foam sheets were buttep up tight against ~ach othe.r
to provide a continuous foam covering arpund the vessel. At the end
3.
of each day, all exposed polyurethane fo~m was covered with polyethye­
lene sheeting to protect it from the weatqer.
Apply a 4" strip of masking tape over the polyurethane foam at the
·location of joints of the cover panels .. The masking tape was useq as
.ah extra precaution to prevent any epoxy in the joint from seeping
into the polyurethane foam.
4. Apply the adhesive cement to the back fa9e of the cover panel leaving ' ' '
'a 2;' ~trip all around th~ edges free of cepient.
5~ Place the. cover pan~ls on the polyurethar:i.e foam 'leaving 1/4" to 1/2''
open between each of panels.
6. Fasten the panels together temporarily u~ing steel straps attached to
the form studs. These steel straps 'serv~d to hold the new panels in . ' - ' . . . . .
place while the joint was completed. See Figure 2.11.,1.
' '
7. . Using a special T-shaped tool, place the 3" strip of epoxy impregnated·
·fiberglass tape behind the panel joint.
8. · Fill the. joint with epoxy and place a seco~d 3" strip of epoxy impreg- , I .
! .;
'\ ·~.
' )
\ }
/
'
DRESDEN 2, 3 2.11-4
in addition to the steps followed in the standal'C( procedures outHned above,
the following special prec~utions were taken ~t the junction of·the gap filler
and pipe penetrations. These procedures are as follows:
l. · Appiy the polyurethane foam sheets. on the drywell shell up tight . . . . . . .. . . . .
against the penetration.·
. .
2. Place t~e penetration pipe sl~eve on the penetration stopping at the
p~lyurethane foam sheet (i.~. 2-1/41' from dryweli shell).
. 3. place cover panels to within approxilllatefr '114" of the sleeve and .
caulkthe joint between cover paneis and the sleeve with epoxy caulking.
4~ Apply epqxy· and fiberglass tape to join th~ sleeve with the cover panels.
. , ... :.
. ~- . . ,
2 .12 Question . .
With regard to· the design of the core shroud and jet pumps, present an ' . . . . . . .
analysis to show that relative motion will not result in damage to these ·
components.
Answer
_A discussion of the jet pump assembly and its' relationship to the vessel
and to the othe.r reactor internal components during steady state and
. transient operation is included in the jet .pump 'topical report entitled
"besigri and Performance of G. E. BWR Jet Pumps", APED-5460,
Sept. 1968. Section 4. 3~ 2 .. 2 of the report describes the stress analysis
that was performed to demonstrate the adequate of the structurai design
of the jet pump assembly and the core shroud durfog operation of the.
emergency core cooling system, which is the condition of maximum.
stress for tll.e. jet pump-core- shroud assemblies.
•
2. 13 Question
Present an analysis of the manner in which the supports carry the vertical
loads from the react~r vessel down through the pedestal into the lower
structure' and also the manner in which lateral forces are carried through .
to the drywell and exterior supports.
Answer
Vertical loads from the reactor pressure vessel (RP:V) are transmitted to
the foundation through t.he RPV skirt, RPV support girder, and RPV
support pedestal. Lateral loads are transmitted to the building'through . -
RPV stabilizers. The RPV stabilizers are attached near the top thir<;i of . . .
the RPV and are connected to the top of the concrete' and steel shield wall.
·.The shielc,i wall in turn is anchored at the base to the top of the RPV.
pedestal and restrained at the top by a horizontal tub.ular truss system.·.
The lateral loads are transmitted through. the truss system to the drywell
shear lug mechanism .. This shear lug mechanism permits vertical movement . . .
of the drywell, but restricts rotational movement. However, lateral loads
&re transmitted through the shear lug mechanism to the heavy concrete
envelope around the drywell which is part of th~ reactor building. A portion . !
of the. lateral loads are transmitted- from the reactor pressure vessel to the
RPV pedestal and thence to the foundation.
I
2.14 Question·
The conceptual models employed in the dynl}miC analysis of the drywell, . .
reactor building, and turbine building are i~completely described.· Specific
. questions· follow~
With regard to the model pictured in Fig. 5. 2.17 arid again in Fig. 12.L8,
clarification is required as to the significa~ce of the. spring numbered 14 in
Fig. 12. 1. 8 which it is presumed is the sarpe as the connecting link num­
bered 15 in Fig. 5. 2.17 .. What is the sign~icance of the solid li~age
numbered 15 in Fig. 12. 1. 8 which appears to be ·similar to the connecting " • J • •• : • •
lirlk numbered 16 in Fig~ 5-;-2'.17 and shown with a spring there. Cla·riHca- . ' . .
tion of the nature of this coupling of the systems is requested, along with
values of the spring constants.
Answer
The spring numbered 14 in Fig. 12.1. 8 or ~5 in Fig. 5. 2.17 are three
14WF167 connecting the turbine building superstructure to the concrete
portion of the reactor building having a value of 105, 000 · Kpis/Ft .
. The coruiecting link n.umber 15 in Fig. 12. 1. 8 or 16 in Fig. 5. 2. 17 is the
. rigid connection of the reinforced concrete operating floor of the turbine
building to the. reactor building and is not considered tO act as a. spring:.·
2.14b Question
In addition, although the differences in values are ·fiot great, comment is
requested on the differ~nce in mass values sh~wn on the two figures cited.
Also what are the sprl.ng values assumed for the dcywell, reactor building
. arid turbine building?
Theniathematical model of the reactor.'...turbine buiiding (Fig. 12.1. 8)
includes the mass of the drywell for the seismic analysis .. In the analysis
?f the drywell, the mass and properties of the drywell are taken out of the
reactor-:turbine building model (Fig. 5. 2. 17) .. ·Thus making a difference in .
the masses of the reactor-turbine building in the two analysis.
The drywell lumped mass model was cons.idered fixed at elevation .
500 '-0 5/8" and laterally supported at elevation 572'-211 •
The spring value between the reactOr building and the turbine building is
105, 000 kips/ft.
2.·14c Question
Is vertical excitation of these units considere(j? If so, Hst the. applicable . . .· . .
. mass arid stiffness values. Also provide typical accelerations, load and
deflection results.
Answer
Class I structures and systems a,re designed to resist a constant vertical
acceleration equal to 0.067g acting simulataneously with horizontal design .
accelerations.
2. l 4d In Fig. 12 .1.11, it is. noted that in the turbine building an acceleration as
high .as' 2. 4g is noted at about the 580 ft. elevation level. Discussion of
/
Answer
The in~ximum shears and moments produce~ by the acceleration (2.4g)
were distributed arriong the turbine building superstructure columris. A
further seismic study with 59% of the .crane load acting oh one turbine
. buildihg frame indicates a maximum acceleration ,of 0. 80g acting at the.
elevation of the crane. The period of.the system is 0. 54 sec.
Note that the turbi_ne ·building is a Class II· str_ucture and. that the degree
of analysis which was performed is not required.
· 2. i4e Question
Additional. comment concerning the moments shown in Fig. 5. 2. 25 is
required i'n order to gain an understanding of the mode of behavior. From
the shear diagrams and accompanying sketches of the model it is assumed
that the drywell is held against lateral motiOn at elevation 57 5 and at. the
base as shown in Fig. 5, 2.17. The .moment diagram corresponds_ roughly·
to 'that which ~.ould be associated with a free standing cantilever fixed q.t
the base.and subjected to a unif~rm loading. W~th some degree of ffxity .·
at an uppe~ level, .the mome~t diagram would be expected to irregular . . .. . . .
with possibly a sign reve~sal, but such behavior is riot evident fo Fig_. . . . - . .
9.2.25.·. It is rioted in Fig. 5.·2.23 that the dryWell undergoes some
· ·.· .. displacement at the 575 ft .. elevation level which would appear to· reduce . \
··the tendency for reversal in moments. However, the smali displacement · . . . . .· ' . . . . . , ..
. shown there, on the. order of 60 ~ils, ,seems urilikely to lead to ~major · relaxation in moment.. Clarification is required ..
Answer
The m~del of the drywell is considered fixed at elevation 5()0 '-0 5/8" and . . . .
laterally supported at elevation 572 1.:.211 • The displacements of the dry-
/
.. ..:..:
'
'
DRESDEN 2, 3 2.14-4
rocking. The results of the seismic analysi~ of the drywell are envelopes .. . . . . . : . .
of maximum.shears and moments regardles~ of sign. ·.·•·
The drywell is embedded integrally with the concrete mass substructure ..
The reactor pressure vessel pedestal w~s cast integrally with the ma::;s
concrete above th~ bottom of the drywell. ·.The relative mass and stiffness
of the c~n<:!rete substructure assures the fixity of both the drywell and
reactor pressure vessel.·
DRESDEN 2, 3 2.15-1
1 2. 15 Question
.Comments on the general design criteria summarized in Section 12. 1. 1. 3 for the primary containment, the reactor building, primary pressure vessel supports and the reactor
primary vessel internals and emergency cooling systems are as follows:
(a) The piping loading and stress design criteria are requested for all systems compris­ ing the primary pressure boundary and for all engineered safety features.
(b) With regard to the loading combination involving the maximum earthquake for each of the items just cited, additional information .is requested as to the implementation of
the criteria employed in the analysis and design. This discussion of implementation
of criteria is requested for each major component listed in Section 12. 1. 1. 3 and for
piping systems also. For example, under "primary containment (including penetra­
tions)," it is noted that if the total stress exceeds yield, an analysis is made to deter­
mine that the energy absorption capacity exceeds the energy input from the earthquake.
An example of this type of calculation is desired.
For the reactor primary vessel internals it is noted that the strains are limited to
preclude failure by deformation, etc. An example of the implementation of this
criteria is requested. Evidently this same criterion is followed for piping and
comments on the application of this criterion to the design of such systems is
requested.
Answer
(a) Design basis loads and design criteria for Class I piping are outlined in the answer to
Question 2. 1. In addition, shutdown capability from the standpoint of pipe integrity
is evaluated under maximum earthquake. Since the earthquake load is the only
significant component increased over design, and this component is a small part of
the total load, the maximum stress does not markedly exceed code allowable. The
application of special criteria to preclude failure by deformation is not required.
(b) All Class I structures and systems have been investigated for maximum earthquake
and only in insignificant isolated locations iri the reactor building's superstructure
were the yield or ultimate stresses slightly exceeded. Because of the redistribution
of forces and the additional structural systems, such as the diaphragm action of the
roof truss, the design is considered adequate with the slight over stress in isolated
locations and an energy absorption analysis is not warranted.
The fuel channel and core plate are items where deflection is limiting. The answer
to Question 2. 2 discusses these two items in detail.
2.16
Question
For major categories of Class I items, a r~~port on the seismic analysis
-· _details, giving as 'a minimum the following information, preferably
through the use of specific examples' is requeste.d.
2. 16a Question
·.The analyticai model us~d;. the location of the lumped masses, the values
of parameters used, and identification of ~q.pport conditions. For the
piping or equipmenLprovid~ justification fof the boundary conditions as­
sumed ·in the· analysis.
Answer
. Representative mat~ematical models a~e shown in the following FSAR.
figures:
,5.2.19
procedures are used.
i•'·
Answer··
The natural periods of vibration for the var:ious modes. are given in the
-following F~ARparagr~phs: .· . . · • I . . •
·:.·
r:.:l rorl'-~'.'.." (.;. • .--..... °"' I
REACTOR PRESSURE VESSEL .
l{f ;;J, ) ;: /Q r "'/;;)
EL. 548'-11 11
• D!lESDEN 2, 3
Th~ participation factors are not involved irt the analysis because eith~r a time history analysis was used as outline in answer to question 2.16,
or a modal analysis; taking the square root of the sum of the equares
with each mode participating equaily as outlined in answer to question 2. 9
was used.
The in ode shapes are not plotted in the analysis.·
· "· · 2. 16c .Question . . .
The method employed in combining the m~dtil values to obtain the design . . . ~ . .
valuesof the acceler~tion, seismic force, ~hears and/or moments.
Answer
. . . . .
The response of each mass for each mode oonside.red at each increment
of time is retained in the computations and total response for each incre­
ment of time is oqtained through the algebrl'}ic sum of each mass points
model contribution at that particular instant of time. Once displacement
·and inertia f~rce are: calculated, time histories ~re established, a time
·.history of shears, moments, displacements, and accelerations is deter­
mined. These records are then scanned to deter~ine the maximum values.
Model Analysis
. . . .
The total. response for each mode is determ.ined by taking the square root
· • o~ the ·sum of the· squares of the maximum deflection for each mode.
.. .... ;
'
,
2.16d Question . . . . . .
The resultant design values actually used ane;t e.xplanation of the differences
.from.analytical .results if any.
Answer .. •.,.
The results of the seismic analysis were us~d in the design of the associ­
atetj Class· I strwctures, systems and components .
2. 16e · Question
Justification f~r use' of, arid selection of, values of the seismic coefficient ~ / ...
if b. arid c. · steps were not ~erformed in the analysis.
Answer ·. · .. : . ·.. .. . .· .··. . . . . . . .
Where a dynamic analysis' was not performed the,horizontalseismic coef-:-
ficients for rigid Class I equipment in the reactor turbine building is· equal
.. to or greater than the building acceleration at the installed elevation .. The
vertical seismic coefficient is equal to 2/3 ground acceleration or O. 067g:
Flexible and rigid Class I piping systems are analyzed as described in the . . . ~ . . .
answer to question 2.~ 9.
2.16f . Question
A description or the support conditions and inner connections where com­
plex systems were considered, and. a discus~ion of the analysis leading to
the decision that the systems could be decoupled for purposes of analysis
and design.
34,&3 ..
Answer.
The mathematical model of the reactor pres1l'ure vessel, Figure 2.16.1 ·
·gives the support conditions which are fixity at the base of the pedestal,
lateral support at the stabilizer elevation CO\'.ffiecting the reactor pressure
vessel to the sacrifical shield and a horizontr-i pipe truss system coru.iect:'-
ing the sacrifical shield to the buildiilg. · i.
. .
· The reactor pressure vessel seismic analysls is affected by seismic mo- . .
tions at the support points only.' The~efore, the reactor pressure vessel
' could be d~cot.tpled from the rest of the building and analyzed separately
.·taking into consideratlon the effects of input piotibn at all supports and
conservatively combining. the individual affeqts.
2.J6g Question .
.• . The application ·of the stress and deformatio:µ crite.ria to th~ actual pro­
. portioning, espeCially for the cases involvinf!; the maximum earthquake.
Answer· \.
The application of the crit'eria can be found qy referring to paragraph
. 12.1.1. 3 for the Criteria .and Basis of Desig;n.
34,&4
-:d
· 2.17 · . Question
In cases where, equipment·or it~ms are SUPPprtea, in or on a building or
other structural system and the input rnotio:rv to the equipment is assumed
to be that of the structure ~t th_e point of supp~rt .of the .equipment, discuss
the reaso~s fol'. using the relative· motion of the building to ground as input . . . . ! .
motions to the equipment rather than the abs;olute motion in space of. the . . ' . . . \ . .
support points of the equipment.
Answer
For the seismic analysis of .equipment absolute acceleration is used at - . ·- . .
. the points of support.
2.18 .Question
It is noted in Fig. 12. 1.19 that accelerations at various levels in the . .
control room are given. However, in comparing these acceleration
values with those shown in Fig. 12.1..11 for the turbine building, it is
obvious that the·tw6 levels of acceleration are substantially different,
_which suggest that the input may be from a different source or that the
systems are not interconnected. Clarification is requested as to the
interconnection, if any, between the turbine building and the control room
and the reasons for the large differences in,.~qceleration values that . :~ .. ~:~.·
. appear to exist under tP,e earthquake excitation.
Further~ information is requested as to the procedures that are being . .
tak~n to insure that the instrumentation located in the control room
complex, including the instrumentation required for shutdown, can with­
. stand the dynamic excitation.
Answer
The analysis of the reacto:i;- turbine building was a coupled time history
analysis. The accelerations. are maximum absolute accelerations with
respect to height ..
. . . . . .
. The analysis of the control room .was modeled and dynamically analyzed
, independently as described in paragraph 12.1. 2. 2.
A comparison of accelerations for the concrete portion of the structur~
is as follows:
• .<'• DRESDEN 2, ;3
E.levation · .Control Room
549.0 . 0 .19.g
'
The difference in accelerations of the superstruCture is due to the model.,..
ing assumptions. and method of analysis of the superstructure and.distribu­
tion of the 9rane loads .. A furthei; exJ)lanation of this is given in answer
. to qu~stion 2. 14d.
The i:>roteqtive sytem instrumentation ana its supporting panel or cabinet.
· · , 19~ated in th~ control ~oom will be analyzed, tested or investigated to
coqfir~ that it will wit~st5ffid the interaction effects resulting from the
_,,, .... . .. ~·
/
/
. . .
The str~ctural frame. of the reactor. building superstructure is stated to
.. be at yield stress in the event of a300 mph tornado. Since this fra~e . . .
. supports the bridge crane and other equipni~nt; discuss the consequences
of deformation a~d yielding of the st~~l franie iri t~rms of potential · . . . . .
missile damage to fuel assemblies ln the core and the storage pool during
refoeling status. -
The strubtural system_ of the reactor building superstructure consists of · ,, - .
a horizontal roof truss which transmits local loads from each bay to
vertically braced bays at. eithe:r end of the buil~ing. In. the design of this .
. · systefu, ·.ioads caused by a 300 mph.tornado were reduced to design loads
by a factor .of22/36 which is the ratio of working stress t9 yield stress
.for A36 s~ee~. As a re.sult of the truss system, members were sized
·. •. primarily as c~mpression member.s •.. The factor of safety for ~ typicai ..
c.ompression member varied b~tween 1.67 and 1.92, or an.·average ~f 1. 85 (which compares ~ith _36/22, or 1. 63 fi:-oni whichthe loads were
determined).: .This factor~ ~~upled with the fact that reference loads for · . ! . . . .
compressive failure are always below yield, demonstrates that the
principle loa~ carrying members will al~ays be stressed below yield aild,
. therefore no yield deformation can take place.
' The crane ran supports are rigidly fastened to the reactor building super-
. structure. Each of the four c:rane trolleys is equipped witli two safety lugs
. which prevent the troileys from leavirig the' rails. When not in use' the
crane is parked at the center of the building, equi-distant from the two
reactors, and the two storage pools.
•
•
•. DRESDEN 2, 3 3. 1-1
SUPPLEMENTAL INFORMATION FOR QUESTION 3.1
The original answer states the actual stresses are below yield and that "no yield deformation can
take place," even though the design criteria permitted stresses up to yield .
· .. : ... 34,b9 ··-··.· ..
.'':"!•" • ... :: :.,, ~ r ·. ./.. ··~·, ::
I·.
i
' .. i ;.
·REACTOR PRoTECTION SYSTEM (SCRAM) INSTRuMENTATION REQUIREMENTS
' '· MOdes in which Function Minimum No. of ; .. Inst. Channels per must be operable
UntripJ>.!3d Logic Trip Function Trip Level Setting Hot
Channel Refuel Standby Run. Action
1 Manual Scram x x x
3 -High Flux IBM :S i20/125 of Full Scale x x A ....
. .
2 High Reactor Pressure :Sl060 psig x x x A
2 High DryweUPressure :S 2 psig x x x A
' 2 Reactor Low Water Level ~l inch** x x x A
2 Scram_ Dischg. Vol. High Level :S 50 gallons x x x A
2. .. Turbine Condenser Low ~ 23 in. Hg Vacuum x x c
Vacuum
2 Main Steamline High . :S 7 X normal full x x x c Radiation power background
'·· .. >4 .• Main Stea~lhie lsoi:ation :S 10% vaJ.ve closure x x c
Valve Closure . .. .. ::i,.
: 2 : GeneratOr Load ·Rejection ·*** x c
2 ~rbine Stop Valve Closure :S 10% valve closure x c ..
. ,, ··Notes: L ·Bypassed. in Startup/Hot Standby when teacto~ pressure is <600 psig.
. . .
A~tion to be taken if first' column cannot be niet:
A .
. .. Iiise~ all rOcls immediately: .
B. ·Reduce power level to IBM range and pla~e mode switch in the Startup/Hot Standby position.
·c. Reduce tutbbie load and close isolation valves within 8 hours.
"'. An APRM Wfll be consider'ed inoperable if there are less than 2 LPRM inputs per level or there are less than 50% of the normal compliment of LPRM's to an APRM. ·
•• ~ inch on the w~ter level -instrumentation is 143" above the top of the active fuel.
. ;, . ~ . . . . .
. . . .
The FSAR indlcates that the l '-6" wall of. tP.e reactor building could be
pierced by certain potential. missiles.· ·Please provide diagrams which . . . . . '·"
· show the lqcation of all essential s"ystems and components near. each
outer wall and discuss the provisions taken to protect these components
from missile dainage.
/.
Either 0,f th·e two types of .mis~iles listed on Page 12.1-11 of the FSAR ' • r • , ,
· · in~y penetrate the l' -6" wall based on. theoretical assumpti9ns and . . . - -. . .
calculations. By the use of the modified Petri formula it is established
·. _that in penetrating the. wall", the energy bf th~ missile is dissipated to ,
the extent that the missile will remain embedded in the wall. · Tlierefore .
no essentiai ~ompon~nt~ near the ~uter walls .could be damaged by ·. . . - . •. ·. ' .
. missiles,·
Figures 12:1~29.througli i2.L35 in the FSAR sho~·the iocation of the , . , .. , ' . . : , , . . I
essential components. As can be seen in:these drawings, the essential
. c~mpo.nents fcir each {rnit are separated such that. even i,Ilternal missiles
·. could not ~lisable an entire system .
34-, 7o
SUPPLEMENTARY INFORMATION FOR QUESTION 3. 2
We have reviewed the ORNL document NSIC 22, which suggests a preference for the Army
Corps of Engineers' formula over .the Modified Petri formula because the former is morE! conserva­
tive. It is noted in NSIC 22 that the Corps of Engineers' formula is applicable only within a limited
range, i.e. , that of ballistic missiles. Our potential missiles do not fall within this range. The
Modified Petri represents the best available formula for our missile penetration design. This
formula indicates that at worst missiles will merely eml;>ed themselves in the wall. In view of
the fact that there. is no Class I equipment including electrical wiring mounted on the inside of
external walls, even these embedded missiles cause no damage. In addition, Class I redundant
systems are physically separated .
3 •• 3 · Question
. . . . ~ . ,• . . .
. t~e maxirrm_m wali differelltials, the resultirjf wall stresses, and the.
criteria. assumed for struetural failure. 1 :... • .. · · · . · · ·
Answer
.. The design of the reactor buiiding concrete \Vails is based o~ a differential ·
pressure of 900 pou'nds per square footwithQut exceeding the riormal ~ork- ·
ing str~sses of the r~.irtforc_~ng steelor concrete~ . Thispressure is equiv"'" · • , • _._:.-·--~LX:..f' '". .. "• ·. ". , , ,
alent to 6. 3 poilnds per square foot or more than twice the three poilnds
per square inch differential pressure· associf!,ted with the tornado criteria. ' . . ' . . . ~. . . .
. As for the. 'structural steel super-structure~ 'tlie structural steel frame,
crane girder~, bra.eking, and girts are designed for a 300 m.p.h. wind.
The relief panels :are designed to blow off w~en a wind pressure -of 70 psf · . . . ' .
1s reached .. The removal of these relief pai\elS rapidly reduces the
internal pressure by ·venting •. Reynolds(l) r¢ports that a vent area of· . . ,. . .
. . I
1 sq. ft. for every_ 1,000 cu. ft. of c~ntaine~ space should reduce the
. pressure differential to a sa:fe leveL The Dresden reactor building:relief
panel area provi_des approximately 4. 8. sq·. ft~ for_ every 1, 000 cut. .ft~. ?f
contained space for a safety ratio ofalmost 6:1. . . . . .
. •;
SUPPLEMENTARY IN FORMATION FOR QUESTION 3. 3
The design of the exterior concrete walls is such that the reinforcing steel quantities are the same in both the inside and outside faces. Therefore, the walls are capable of withstanding equal
internal and external pressures. As stated in the original answer, our calculations show that the reactor building concrete walls can withstand a differential pressure of 900 pounds per square foot,
or the equivalent of 6. 3 pounds per square inch. The building as a whole can withstand a signif­ icantly higher external pressure.
34,73
· :_Describe. the compartmenthatch covers as~mmed llfted during the above
·. depressurization calculation and discuss t~eir inissile potential and··
pj:-ovi~ions to prevent damage to Class I es.sential system components.· . . . . - . . . .
· · Ariswer
, The ctjteria ~sed in the design is su9h that no hatch cover will lift loose
. :
·. ,·
.~- ';-.
SUPPLEMENTARY lliFORMATION FOR QUESTION 3. 4
In Clarification of our original answer, it was intended to convey that those hatch covers which could be lifted loose would not act as potential missiles. The lighter weight hatch covers .are
located over areas with a small volume of air, thus precluding large upward air flow from the covered area which could cause the hatch covers to become potential missiles. For larger hatch
covers, sufficient venting is provided throughout the building in the form of stairwells, an elevator
shaft, and open floor grating to preclude the possibility of lifting the covers and creating missile
action. Furthermore, it would be physically impossible to lift some of the larger concrete hatch covers with the small differential pressures present.
34., 7.:{'
4. 1 . Question . . . .
Provide a basis for the design flow control line in. terms of reactivity
·,!
· varfables on the shape qf the control line. Indicate the probabie acc-uracy
. of the ca1cul~tional model with power-to-flow data from applicable test_s . ' . . . . . . .
. ·in other. boilirig water reactors ..
Answer
. The design flow control line is the quas~ steady state loci of operating ·
points of power vs. flow obt~inable in the absence of control rod motion . . . '• . . ' . '. '.
starting from 100% power and 100% flow. The flow control line is evaluated ·
at the beginning of core 1 life (BOL) (s~e Figure 4. 1. 1) and at the end of , . . . . . . . . .
the- first cyde (EOC) (see Figure 4.1. 2). The normal design assumption . . . . . . . . .
is that ~enon concentrations remain constant over the flow control range.
and that power distribution ~ithinthe core is maintained within thermal
. limits.
.. ; ' ..
The flow control system min~mizea the power shape ~hift which occurs in a
load change thus. minimizing power shape relocation from xeno~ .. In this . . .
respect flow control is more optimum than subcooiing or.inlet temperature
control as used in Dresden 1, ,Gundremmingen' and Garigliano. ·Both flow
'control and subcooiing control affect the po~er shape relatively uniformly . l '
in the horizontal planes of the reactor and ~ini1nize power distribution ..
relocation and asso~iated xenon. redistributi~n effects. Local feedback at
each point in the -re~ctor is provided by tlie boiling void and Doppler
reactivity coefficients~
()
•: t-;t_;<::·~EL. ,.s, F:;;St!R <G.
·''
.;:.' · ... ·
.·'•:··
. hon-uniform axial power coefficient and non-unifor~ xenon microscopic . .
cross .section: are important and make the BWR respond much the same
as a conventional automobile shock absorber in· damping axial xenon re_:
distributions during power changes. The n~n:-linear axial trends .in a BWI:t
. controlled with flow result in axiai redistribution being damped in (one cycle
almost independent of core length. These effects haye been illustrate<;} by·
,,. measurements at the Dresden 1 Station an.d by analyses of large 241 long·. . . . . ·. /.· !
BWR's .
. . The_refore, in large boiling water power reactors both of the coolant control ·
. systems are more optimum than use of control ro~s for load following. ·
· As discussed in section 3. 2, 3 of the FSAR, control rods are moved as
necessary to maintain the operating point of the plant on or l;>elow the flow . . . . . .
control line. ,
The reactor core pressure for Dr~sden 2 and 3 varies under flow contro,1 .
as shown in Figure 4. 1. 3. · The difference in flow vs. power relationship
between a constant core p~essure and the pre_ssure shown inF:lgure 4. L 3
is included in Figure 4. 1. 1 and 4. 1. 2. ·".
Operating plant data ha~ been utilized to establish the probable' accuracy of · . . . . . . .
the ca;lculational model. The results of this comparison are shown in ~ ' . . . .
Figure 4.1. 4: The ~perating data was obtained from a dual cycle plant . .
, operating under single cycle conditions and is riormalized at 25 kw/£ ' which.
was the highest power density achievable.in the single ~ycle mode of
operation.
··:r'i-'1·'1ll,/11 1;.\1 ·m.I .,. 11 11 ·111-t '-I!' 'Ill li\i '''I ii·!'·:•·'· •\ 1·1 ii'' '·i! ·''·' 'r 1 i , .. ,'II' ... , .. :1·1 1 'Iii ·11·11··.r- ''II •:11 1111 1: 1 1
' .. , .. , .• 1'·11i· 1,·.·1·,' ·!··.,I'· ., - ·j '1' 11. :111 l·i 11;1 l:'! ,,1., i111. ;111' !ii' 1.,,, i'•·, :.1: i!rl' i. l !ii· 11 !,I, i,11., '1·:1. J.1' i·!.
I · ..Ll..;_! I ' .. j.: I µ,J ! 11-' r!:. · , ! '1 ! I! ; ! Ii I';~ I I 1· I"' L' I I "" · 11 1 1 r ! · 1-i I " :~ ,L ., '..W ... ,.. .< · " ~ Tii; -,-, 11- Hi 1· i
1 • i ~ 1
1 '·
1
~ - .
DRESDEN 2; 3 4.1-3
The plant response to flow control wiil be det~rmirted cfurii:lg the s~rtup
test program and can be observed durihg ali toad maneuvers employing
flow control. The maximum flow control line will always be normaiized
at 100% power a~d 100% flo\\'. Experimental and analytical data indicates . . . . . . ' . ~
that the most th~rmally li~iting I>oint in-the f~ow control. line is at' 100% · .. ··
power a~d '100% how.
·'·
•'.
· 4. 2 Question
Provide an evaluation of the nominal steady-state thermal margins available I . .
as,a function of 'recirculation fiow rate. Since operation with natural cir-
culation flow is not precluded, include an evalµation of .the margins avail­
able in this mode of operation. As a part of the evaluation, provide curves
showing the parameters listed below as a function of flow, assuming
·reactor ope~tion along the (a) flow control line and the (b) natural Circu:....
lation line. Assume for these curves that the design peaking factor exists ·
at 100% power and 100% flow.
(a) Reactor pressure
<c> (d)
Hot channel heat flux
Minimum CriHcal Heat Flux 'Ratio (MCHFR)
Answer
The thermal margins available under steady state operating conditions have
been analyz~d. These analyses include operating conditions along .the flo~
control line and under natural circula~ion conditions~ The thermally most
limiting conditions for the design peaking factors occur at 100% power and
100% flow. Reduction in flow from these starting conditions increase the . . . . .· .
thermal margin. See Figures 4.1. 3 in question·4. l, 4. 2.1, 4. 2. 2, 4. 2. 3,
4. 2. 4,. and 4. 2. 5. These data were obtained assuming that the design
· peaking factors existed at 100% power and 100% flow.
. ,.
·Lio~_ 10 X IO"TO 112 INCH 46 1323-n c 7''x 10 INCHES . UADE:Ut u.:s~A . •.
KE:UFFEi... & ESSER CO ..
-~;.
: I
H::>NI t/1 OJ. 01 X Ol
.~ ~;; ·.-· .• :·
-..--- ---:--· :~f--f--L·.'
10 X 10 TO 112 INCH 46 1323 7 X 10 INCHES 11°'0( IN l!. ·5.A,
KEIJFFEL I\ ESSER CO.
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