ML20126F197
| ML20126F197 | |
| Person / Time | |
|---|---|
| Site: | 05000495 |
| Issue date: | 04/30/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0076, NUREG-76, PB-265-597, NUDOCS 8103100476 | |
| Download: ML20126F197 (150) | |
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related to the preliminary design of tile Office of Nuclear Reactor Regulation a.
,4 SWESSAR-Pl s
Id Standard PWR Referenc3 Docket No. STN 50-496 9
E Nuclear Power Plant
"*'ch 1977 f.
(and its relationship to the RESAR-3S j
Standard Reference System) h
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7,184.loGRAPHAC D ATA
- 1. Arport No.
2.
- 3. Recipient's Aceesssen No.
$DEET NUREG-0076
- s. asperi oue i
- 4. Taa..d s. nema April 1977 8
l Safety Evaluation Report: SWESSAR-P1
- a. Festerissag Organisaties nepc.
- 1. Amtmes(s)
No. NUREG-nn76
- 80. Protect / Task /tock Una No.
- f. Fetteesmaag Organsassaos Name and Address U.S. Nuclear Regulatory Comission Office of Nuclear Reactor Regulation it. C u,act/crea u Washington, D.C.
20555
- 13. Trpo el RePert a Period tz. 5pensersag O<gnasauien Name sad Address Ceeered Safety Evaluation Report i
Same as 9 above 4
14.
IS. Sepplesmentary Notes Pertains to Docket No. 50-495 A Safety Evaluation Report of the application by Stone & Webster Engineerino l
. lo. Abstracts i
Corporation for a Preliminary Design Approval for a reference system design of a i
balance-of-plant design, designated as SWESSAR-P1, for a pressurized water reactor j
nuclear power plant utilizing the Westinghouse RESAR-35 nuclear steam supply system t'
design has been prepared by the Office of Nuclear Reactor Regulation of the U.S.
5 Nuclear Regulatory Comission. This application was submitted in response to Option 1 of the Comission's Standardization Policy. AWASH-1341, "Programatic Infont.ation for the Licensing of Standardized Nuclear Plants". Option 1 allows for l
the review of a " reference system" that involves an entire facility desiori or major fraction of a faciltiy design outside the context of a license application. The li review of the SWESSAR-Pl application was similar to that of a construction pemit review except that it was limited to only those features within the scooe of
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,the SWESSAR-P1 appliiation,'plus safety-related interfaces between SWESSAR-P1 and the RESAR-3S nuclear steam supply system design and site and utility apolicant j
i The Staff has concluded that the SHESSAR-P1 related design and operation aspects.
design is acceptable and may be incorporated or referenced in construction permit applications. Therefore, a Preliminary Design Approval will be issued "rather than l
l a construction permit.
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}J 17b. Identifiers /opea.Zaded Terr.s a
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17s. CosATI Fieldicroup
- 19. secusity Cans ilhas
- 21. :.o. es r ases
- 15. Avast.Dility lHateinen:
"'%"),. ur e,re, Mo Restrictions on Disi.ribution
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11 NUREG-0076 f,.
March 1977 E
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b SAFETY EVALUATION REPORT 2
BY THE I
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IN THE MATTER OF
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STONE & WEBSTER ENGINEERING CORPORATION I
PRESSURIZED WATER REACTOR k
RE ERENCE NUCLEAR POWER PLANT SAFETY ANALYSIS REPORT h
AS RELATED TO RESAR-35 DOCKET NO. STN 50-495 l'
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i TA8tE cr corTENTS
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1.0 INTRODUCTION
AND GENERAL DISCUS $10N............................................
11 i
1.1 I n t roduc t i o n.............................................................
1-1 1.2 General Plant Description................................................
14 1.3 Compa ri son wi th $1 mil a r Facil i ty Desi gns.................................
1-10 i
1 1.4 Identi fication of Agents and Contractors.................................
1-10 1.5 Requi rements f or Further Techni cal In f orma tion...........................
1 10 1.6 Suma ry o f Pri nci pal Review. Ma tters......................................
1 10 1.7 Ou t s ta nd i ng i s s u es.......................................................
1-11 j
l 1.8 Interfaces...............................................................
1-12 1.8.1 SWE55AR.Pl Interfaces with RESAA-35..............................
1-13 s
1.8.2 SWESSAR-P1 Interfaces with Site and Utility Appilcant............
1 16 3j 1.8.3 Conclusions......................................................
1-17 1
2.0
$ ! TE CHARACT E R I ST I CS........ :...................................................
2-1 L
2.1 Geog ra phy a nd Demog ra phy................................................
21 8
2.2 hearby Ind:.strial, Transportation and Military Facilities................
2-1
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2-1
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2.3 Neteorology..........
4 2-1 2.3.1 Regional Climatology....
2.3.2 Local Meteorology................................................
2-2 2.3.3 Ons i te Me teorol og ic a l Meas uremnts Program.......................
22 2.3.4 Short.Ters (Accident ) Dispersion Estimates.......................
2-2 Long-Term (Routi ne ) Di s pe rs i on Es ti ma tes.........................
2-4 2.3.5 2.3.6 Conclusions......................................................
2-4 6
2.4 Hydrology................................................................
2-5 2.4.1 Floods..............
2-5 f
2.4.2 Log Water Consideration and Ultimate heat $1nk...................
2-5
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2.4.3 Groundwater....................................................
2-6 G
2.5 Geology and Scismology................................................
2*6
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3.0 DC51M CRiftRI A F0st STRUCTURIS. SYSTDts 20 Cor0NINT3.........................
31 3.1 Conf ormance wi th General Desi gn Cri teri a.................................
31 3.2 Classification of Str*xtures. Systems and Caeponents.....................
31 i
3.2.1 Se i smic C1 as s i fica tt oe...........................................
31 3.2.2 Sys tem Qual i ty Group Cla ssi f ica ti on..............................
32 3.3 Wind and Tornado Desi p Cri teri a.........................................
32 3.3.1 W i nd Des i gn Cri teri a.............................................
3-2 e
3.3.2 Torna do Des i gn C ri teri a..........................................
33 l
I 3.4 Wa te r Lev el ( F l ood ) Des i gn...............................................
34 j
3.5 Missile Protection.......................................................
35
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l 3.5.1 Mi s sil e P rotection Cri teri a......................................
35 j
3.5.2 Ba rri er Des i gn P rocedures........................................
37 i
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i 3.6 Protection Against Dynamic [ffects Associated with the Postulated Rupture g
j ofPiping................................................................
37 3.6.1 Pos tula ted P ipe R4tu re Ins i de Con ta lfinent.......................
37 I
3.6.2 Pos tul at ed P ipe Reture (ktsi de *.on ta irunen t......................
38 8
l 3.7 Seismic Design..............................................
3 10 l
I 3.7.1 Seismic Input....................................................
3 11 j
3.7.2 smic Systats Analysis.....................................
3 11 f
3.7.3 Se ismi c Instrueenta tion Pro gras..................................
3 13
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3.7.4 Seismic Interface Reautrements...................................
3-13 3.8 Des i gn o f se i smi c Ca tegory I 5 tructures..................................
3 13 3.8.1 Re a c to r Con t a i rre nt..............................................
3 13 I
3.8.2 Concrete and Steel loternal 5t uctures...........................
3 14 i
3.8.3 Othe r Se i sr.i c Ca tegory I s truc tures.............................
3 15 i
3 16 l
3.8.4 F o u n da t i o n s....................................................
f 3.8.5 S t ruc tu ra l I n te rf ace Requi reren ts................................
3 17 4
3 17 3.9 Mecha n i c a l Sy s t ems a nd Ccrconen ts.......................................
3.9.1 Dynamic Sys tem Analysi s a nd Tes ting..............................
3 17 i
3.9.2 ASME Code Class 1, 2 and 3 Ccr9ements and Cortonent Supports....
3 19 11 l
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3.10 Setssic balification of Setssic Category I Instrumentation and j
r 3-21 j
El ect ri cal Equi pment.....................................................
i 4-1 l
4.0 REACT 0R........................................................................
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5-1 1
s 5.0 REACT 0a C00LM T SYSTEM AND CQo ECTED SYSTD6...................................
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5-1 2
5 5.1 Genera l I n f orma ti on......................................................
5-1
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5.2 In tegri ty o f Reactor Cool ant Pres s ure Bounda ry...........................
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5-1 9
I 5.2.1 Genera l Ma teri a l Cons i dera ti ons..................................
5-2 Reactor Coolant Pressure Boundary Ledage Detection. System.......
5.2.2 52 5.2.3 Inservice Inspection Program.....................................
5-3 f
!.3 Containment Bu t 1 d i ng Polar Crane.........................................
6-1 6.0
[NGINEERED SAFETY FEATURE5.....................................................
6 I
6-1
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6.1 Design Considerations....................................................
6-1 l
6.2 Containment Systems......................................................
61 6.2.1 Contalement Functional 0esign....................................
6-6 6.2.2 secondary Conta i nment Functi onal 0esi gn..........................
i 6-10 6.2.3 Contairment Heat Removal System..................................
6-11 I
s 6.2.4 Conta i nment Ai r Cl eanup 5ys tem...................................
I 6-12 l
6.2.5 containment Isolation 5ystem..............
6-14 l
6.2.6 Combustible Gas Control System...................................
6-15 0
6.2.7 Ccr.ta i nment Lea ka ge Tes ti ng Program.............................
i 6 16 6.3 Emergency Core Cooling 5ys t em............................................
l i
6-18
- 6. 4 Control Room Habitability........................
6-20 6.5 Engineered Safety Features Air Filtration Systems.......................
6-20 6.5.1 Sume ry De s cri p t i on..............................................
6-20 6.5.2 Supplewentary teak Collection and Release System.................
t 6-21 j
6.5.3 Control Room Pressurization System.............................
I' Co.iustoes.....................................................:
6-n 6.5..
6-22 f
Engi neered Sa f e ty F ea tures Ma terials.....................................
6.6 7-1 i
7.0 INSTRMNT ATION AND CONTROL 5..............................................
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7-1 7.1 General..............
71 7.2 Reactor Trip System.........
- 2 7.3 Engineered Safety features Qstems.........
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7.3.2 Main Steam Isolation....................
7-3 7.3.3
.Szil l a ry Feedwa ter 5ys t em.......................................
7-4 7.3.4 Nuclear Steam !upply System Interface Requirements for Safety l
Systems........................................................
7-5 7.3.5 Periodic Testing of Engineered Safety Features Systems...........
77 1
i 7.4 Systens Requi red f or Sa f e 5hutdown.......................................
77 i
7.5 Sa f e ty Re l a ted O l s pl ay I ns trumenta ti on...................................
7-8 t
i 1.5.1 Bypassed and Inoperable Status Indicatica for Safety Systems.....
7-8 7.5.2 Post-Accident and Incident Monitoring System.....................
78 j
g 7.5.3 Conc 1951ons......................................................
7-9 j
l 7.6 Other Instrumentation Systems and Requirements for Safety................
79 l
1 7.6.1 Environmental Qualification of Class IE Electrical Equipment.....
79 7.6.2 Independence and Identification of Safety Related Equipment......
7-10 7.6.3 Manual Ini tiation of Protective Actions..........................
7-10 i
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7.7 Control Sys tems not Requi red for 5a fe ty..................................
7-10
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7.8 Instrumentation and Controls Interface Requirements......................
7-11 8.0 ELECTRIC P0WER.................................................................
8-1 8-1 8.1 General..................
8.2 Offsite Powe* System.....................................................
81 8.2.1 Of fsi te Power System Inteef ace Requirements......................
8-2 8.3 Onsite Power Systems.....................................................
8-3 8.3.1 Al t e rna ting Current Power 5ys tem.................................
B-3 8.3.2 Di rec t C urrent Power Sys t em.....................................
8-7 1
8.4 Interface Requirements for Electric Power 5ystems........................
8-8 i
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9.0 AUXILIARY SYSTEMS.............
91
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l 9.1 Fuel Storage and Handling..................
92 n
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I 9.1.1 New Fuel Storage.,.....................................
9-2 9-2 6
9.1.?
Spent Fuel Storage.............
9.1.3 Spent Fuel Cooling and Cleanup System.........
9-3 9.1.4 Fuel Handling System...........
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l 9.2 Water 5ystems............................................................
95
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9.2.1 Se rv i ce Wa te r 5ys t en.............................................
95 9.2,2 Cooponen t Cool i ng Wa t e r 5ys t em..................................
96 9.2.3 Ultimate Heat Stat..........
4...................................
99 l
9.3 Proces s A;;x 111 e r i e s......................................................
99 9.4 Air. Conditioning. Heating, Cooling and Ventilation $ystas...............
99 I
r 9.4.1 Control Bui l di ng Venti ta ti on 5ys tems.............................
99 9.4.2 Fuel Bui l di ng Ventil a ti on Systs11...........................
9 12 9.4.3 Engineered Safety Features Venti?stion Systen....................
9 13 o
9.5 Other Aux i l i a ry 5 ys tems.................................................
9 13 9.5.1 F i re P ro tect i o n Sys t r.........................................
9 13
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9.5.2 Diesel Generstor Fuel Storage and Transfer System................
9 15 9.5.3 Di ese l Gene ra tor Auxi l i a ry 5ystems...............................
9 16 h
9.5.4 S tora 9e o f Coev res s ed Gas es.....................................
9 13 h
10.0 ST EM MD POWER CONV ERSION 5YSTEM.............................................
10 1
%4 10.1 Summa ry De s c ri pt i on......................................................
10 1 l.a 10.2 Tu rt> t ne Ge ne r a to r........................................................
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10.3 Fa i n 5 t e am Supp l y Sys t ers................................................
10 2
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Ift. 4 Circulating Water 5ystem.................................................
10 3 h.
10.5 Auxiliary Feedwater 5ystem...............................................
10 3 i
20.5 ste, m Conside,ations.................................................
iO.5 7
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i 11.1 Summary Descript1on......................................................
11 1 11.2 Liquid Radwaste Treatment 5ystes.........................................
11 2 a
4
- 11. 3 Ga seous Ra dwas te T rea tnen t 5ys tem........................................
11 6 3
11.4 Soli d Ra dwe ste T re a tant 5ys tem..........................................
11 7 11.5 Process and E f fluent Radiological Monitoring.............................
11 8 T,
11.6 Conclus ons......................................................
11 8 a
F-12.0 RAC I AT! CM P ROT E CT I ON........................................................
12 1 o
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12.1 Ra d i a t i on S h i e l d i n g...................................................
12 1 12.2 A r e a Ra s t a t i o n Moni to ri ng............................................
12 2 l
12.3 Lose Assessment............................
12 2 l
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13.0 CONDUCT OF OPERATIONS..........................................................
13 1 2
I 14.0 IMITIAL TESTS AND OPERAT!0N$...................................................
14-1 15.0 ACCIDENT ANALYSES..............................................................
15-1 6
i 15.1 I n trod ucti on.............................................................
15-1 15.2 Rad i ol og t eal Cons equences of Ace t den ts...................................
15-1 i
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,i 15.2.1 Genere1..........................................................
15-1 1
15.2.2 Los s -of *.ce l an t Acci den t.........................................
15 3 I
i
- 15. 2. 3 Hydrogen Pu rge Dose Analysi s.....................................
15-7
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15.2.4 Fue l Ha Mi i ng Acci dent...........................................
15-9 s
15.2.5 Rod EJ ecti on Ac c i dent............................................
15 9 I
15.3 Anticipa ted Trans t ents bi thou t Scram.....................................
15 9 i
1 is.O TECm ! CAL srE Ci r ! CAT!cas......................................................
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- 17. 3 QUAL I TY A$$URMCE.............................................................. 17 1 1
17.1 Genera1..................................................................
17-1 i
1 7. 2 O rga n i z a t i on.............................................................
17 1 t
- 17. 3 Qua l i ty Assurance Program................................................
17 3 e
17.4 I g l emen ta t i on...........................................................
17-4 17.5 Conclusion...............................................................
17-4 18.0 REVIEW BY THE ADy!$0RY C01MITTEE ON REACTCR SAFEGLACS....
18 1 6
- 19. 0 C0NCL U$ ! CN S...................................................................
19-1 L
APPEM0!cES I
APPEMCII A:
N0bST AM0ARD AMD OPT 10 MAL $Y STD45 I N SWE15AR.P 1.......................
A.1 h
A 9.0 AUXILIARY SYSTEMS..................
A-1 A 9.3 Process Aunt 1taries.........................................
A1 A 9.3.1 B o ron R e co ve ry Sy s tem.......................................
A.1 A 11.0 RADICACT I VE WA$TE MMAGEMENT................................
A2 A 11.2 L i qu i d Rad.a s t e Trea tmec t Sys tem............................
A-2 i
A 11.2.1 Laundry haste system........................................
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B-1 CHROM 0 LOGY OF REVIEW OF $WES$AR-P1 REFCtEhCE SAFETY ANALY$15 REPORT E.
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i C-1 APPEMDIX C:
BIBL10GAAPHT..........................................................
I REPORT BY THE ADVISORY COM41TTEE ON REACTOR SAFECUARDS Olt $WES$AR-APPEADIX D:
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(j, STONE & WEBSTER ENGINEERING CDRPORATION. TRANCE OF PLANT DESIGN......
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FIGURE 1-1 SE SSAA-P1 LAYWT OF STRUCTAES Me 8U I LD I NGS.........................
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TA8LE 1-1 MMOR ITDt5 TO K AcrAE5$ED ST UT:LITY APPLICANT REFERENCING i.e suss4R-P1............................................................
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I TABLE 14 INTERFACE MATTER $ FOR SW5 sam P1/RE$AR-35 DE$1GN COM5! NATION 1-14 r
70 at ADORESSED BT UTILITT APPL 1 CANT...................................
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TABLE 2-1 MA11MJM ATMOSPHERIC DISPER$1Cet FACTORS (0 2 HOURS 1/Q VALUES) FOR 2-3 e
5dE55AR-P 1 CONT A!MDif CQhCLPT5.......................................
TABLE 3-1 TORNADO GENERATED EXTERM L MIS $1L C FCR SWE$5AR.Pl....................
3-6 nN g
TABLE 3-2 CESIGN AND ANALT515 RESPONSIBILITT FOR REACTOR COOLANT SYSTDt.
3-9 COPPCNENT AND SUPF 0RTS................................................
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TA8LE 9-1 FUEL POOL COOLING ST3 TEM PERFORMANCE OF SWESSAR-P1/RESAR-35 DESIGN....
9-4 s
i TABLE 1!-1 DE5tGM PAMMETERS OF PRINCIPAL COMPONENTS FOR LIQUID. GASE0V$ AND11-4 P
SOL I D RADWAST ES ST 5Ti.MS...............................................
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a 11-9 TABLE 11 2 MONITURING OF PROCESS AND EFFLUENT STREAMS..
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4 TABLE 15-1 POTTNTIAL RADIOLOGICAL C0d$!CUEMCEs or OESIGN BA$t$ ACCIDENTS AND LIMITING ATM00PHERIC DISPER$10N FACTCR$ FOR SWE55AR.P1/RESAR-35 t4 15-2 OE51GN...............................................................
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a L0$! OF.C0C. ANT ACC10ENT A55LMPTIONS AND INPUT PARAMETERS TO 1
TABLE 15-2 DETER.MINC T iE ' IMITING ATMOSPHERIC O!!PERSION FACTOR $(1/Q VALUES).....
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TABLE 15-3 LOSS */-COOLANT ACC17hT Ctr.fA!PEMT LEAXAGE Air *TIONS.............
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(J TA8715-4 HY".?%EN PUMf VG**10M A55LMPTICRS AN3 INPUT PARAMETERS TO EST! MATE f
15-8 6 5:TE C0~.ES......
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TAsti 15-5 Fu!* MNDLING ACCT:,EMT AS$'JMPTIONS AND thPUT PARMETERS TO ESTIMATE 15-10
'diblTE 00sES FOR S at$$,*a.P1/PiSAR-35 DEJIGN.........................
L 7A8t! 15-6 R00 EJECTION ACCIDENT A55 UMP 0.'tS AND INPUT PARMETERS TO ESTI" ATE 15-11 0FF5 ' TE CCSE S F G4 $wt 57-i s / uiAR-15 CC $1 GN..........................
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1.0 INTRODUCTION
AND CCMERAL 0150U$$10M 1.1 Introduction The Stone & Webster Engineering Corporation (hereinafter also referred to as Stone 4 Webster) filed on April 25. 1974, with the United States nuclear Regulatory Comission i
(the Coassission) then known as the United States Atomic Energy Comission, a proposed
- rreliminary standard design for the balance.cf-plant portic,n. designated as $WESSAR.Pl.
of a pressurized water reactor nuclear power olant. This submittal ins in the fom of an application for a Preliminary Design Approval by the Commission in response to Option 1 of the Comnission's standardization policy. WA$H.1341. "Programatic Informs.
tion for the Licensing of $tand.rdized huclear Power Plants.* Option 1 allows for the review of a ' reference system shat involves an entire facility design or major fraction of a facility design outside tha context of a license appilcation. The application was j
f docketed on June 3,1974, under Docket No. STN $0 495, The initial Comission policy statement on standardization of nuclear power plants was issued on 1 ril 28,1972. This policy statement provided the tapetus to the nuclear 4
industry and the Comission to initiate active planning in their respective areas.
Thet is, it provided a method whereby the benefits of standardization could be realized while maintaining the Comission's standards for protecting the health and safety of the puolic and for protecting the envircrenent. On March 5.1973. the Comission announced its intent to implement a standardization policy for nuclear power plants, in August 1974 the Comission issued its standardization program plan. WASH.1341.
Amendmant I to WASH.1341. discussing " options
- ar.d " overlaps
- was issued January 16 1975. The regulations governing the sutmittal and review of stand rd designs under the j
" reference system
- option are stated in Appendix 0 to Part 50 and Section 2.110 of Srt i'
2 of Title 10 of the Code of Federal Regulations (CFR).
A Safety Analysis Report, entitled " Pressurized Water Reactor Reference Nuclear Pcuer Plant Safety Analysis Report. SWESSAR.Pl.* was submitted with the application, and is referred to in this report as SWE$$AR.Pl. The information in SWES$AR.P1 has been supplemented by /vnendments 1 through 31. $WE55AR.P1 and copies of the amendments are svallable for public inspection at the Nuclear Regulatory Ccruission's Public Docenent Room.1717 H $treet. N.W., Washington. D. C.
The SWES$AR.P1 $afety Acalysis Report describes the preliminary standard design ar.d analyses of structures, systems, and cc ponents that comprise the balance-of. plant portio's of a standard pressurized water reactor nuclear power plant. The WE55AR.P1 design does not includo a nuclear steam supply system, but the app 11 cation includes by reference the following standard pressurizad water reactor nuclear steam supply
[
systems, unicn have been sutmitted to the Cynissioa in separate applications, for 11 I
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.c Preliminary Design Approvals in accordance with the Commission's standardization 9
policy.
(1) RESAR.41(DocketNo.STN50.a80): a design by the Westinghouse tiectric Corpora.
tion (hereinaf ter referred to as liestinghouse) with a core thermal output of 3800 megasatts. The SW55AA-P1 design information related to the A15AR.41 design was t
I i
included in the application docketed on June 28,1974 The Commission's Safety
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Evaluation Report for RESAR.41 (NUREG.75/103) was issued in. December 1975. and a
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Preliminary Design Approval for the design was issued on Decaneer 31. 1975.
(2) CESSAA System 80 (Docket No. STN 50 470): a design by Combustion Engineering.
j Incorporated (hereinafter referred to as Combustion Engineering) with a core
}
thermal output sf 3800 megawatts. This system is also referred to as the CESSAR j
design. The SWE55AR.PI design information related to the CIS$AR design was included in the Safety Analysis Report by Amenchment 3 on October 21, 1974 The l
l Commission's safety Evaluation Report for C!$$AA (NUREG.75/112) was tssued in l
j December 1975, and a Preliminary Desfon Approval was issued on December 31. 1975.
l (3) 8 5AR 205 (Docket No. STN 50 561): a design by the Babcock & Wilcos Company (hereinaf ter referred to as Babcock & W11cos) with a core thermal output rating of 3000 megawetts. The SWESSAR.Pl design information related to the B-5AR 205 design was included in the Safety Analysis Report by Ameneent 19 on December 19 1975. The application for a Prelisinary Design Approval for the design is presently under review by the Cocnission.
(4) RESAR.35 (Docket No. $TN 50 545): a design by Westinghouse witn a core thernal output rating of 3411 megawatts. The SWE55AR.P1 design information related to the RESAR 35 design was included in the Safety Analysis Report by Annendment 17 on October 2.1975. The Comission's 59fety Evaluation Report for RISAR.35 (NUREG.0104) was issued in Decembe? 1976. and a Preliminary Design Approval for the design was issued on December 20. 1976.
The proposed SE55AR.P1 standard balance.of plant design can be uttilted with each of the above standard nuclear steam supply systems. Those portions of the SWISSAA.P1 design that are different for each or ary of the refe*enced systems are specifically identified 'n the $WESSAR.Pl Safety Ar>1ysis Report.
This Safety Evaluation Report (the repo.t) presents our evaluation of the SWESSAR.Pl standa d balance.of. plant design and its relationship to the RE!AA 35 standard nuc? ear I
{
steam supply system design. Those portions of our evaluat'on of tre S ESSAR.PI design and analyses that specifically relate to Lee SiiESSARJIMESAR-35 desigt. Combination i
are identified in the margin of this re:4rt. The identificatten extends, as a minirum.
~
over the entire sentence addressing suc9 a spesif 6c Sn55AR.P1/RESAR.35 relationship.
de wl11 prepare separate Safety Evaluatico Reports for the ceroination of the ShE55AbP1 standard balance.of.olant design with tre other standard nuclear steam supply systers using the format of this report.
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The Office of Nuclear Reactor Regulation of the Cosmission issued a Report to the Advisory Committee on Reactor Safeguards on the $WE$$AR P1 application (and its rela.
tionship to the CESSAR design) dated June 1976, staunarizing the results of the review by the Commission's staff at that ties. Subsequently, the Advisory Committee on Reactor $afeguards considered the SWE$$AR P1 application at a meeting on August 12 1976, as discussed further in Section 18.0 of this report. Copies of the Report to the Advisory Committee on Reactor Safeguards are available for public inspection at e
the Commission's Public Document Room 1717 H Street, N.W., Washington, D.C.
This $afety Evaluation Report samarizes the results of the technical evaluation performed by the Ceanission's staff, of the proposed SWE$$AR.P1 standard balance-of.
e plant 4esign and its relationship to the RESAR-35 standard nuclear steam supply system f
design. The report delineates the technical matters considered in our evaluation of I
the radiologicel safety aspects of the $WE$$AR P1 design, a'.# esses the cassents made l
by the Advisory Committee on Reactor $4feguards in its report of August 18. 1976, and addresses the resolution of outstanding issues previously identified during oor review.
The $WES$AR.P1 application is not related to a specific site for the construction of the SWES$AR-P1 plant and does not include specific site information. We, therefore, have not performed an environmental review of the $WES$AR41 design and have not written cui envirefwental impact statement. We will evaluate the environmental is9act of the'$WES$AR P1 design on a specific site during our review of a construction perett j
application referencing the $WES$AR-P1 design.
Based on our evaluation, we conclude that the proposed $WES$AR PI preliminary design I
of a standard balance-of-plant can be cortined with the RISAR 15 standard nuclear steam supply system design, can be incorporated by reference in a construction permit application, and can be constructed without e"ndangering the health and safety of the public. We conclude that a Preliminary Design Approval for the proposed design can be granted. Our detailed conclusions are presented in Section 19 of this report.
A future utility applicant referencing the $WES$Ae.P1 design siust also reference one of the standard nuclear steam supply system designs referenced in the $Wi$$AR P1 application. We will need to conclude for each application that the utt11ty applicant.
along with its contractors, is technically coepetent to manage, design construct and operate a nuclear power p* ant prior to issuance of a construction permit.
g The review and evaluation presented in this report is only the first stage of a continalng review by the Consission of the design, construction, and operating features of the $WIS$AR-P1 standard balance of-plant design. Prior to the issuance of an operating license for any appiteation referencing the Siit$$A.'t-P1 dest;n. we mill review the $WE$%P1 final design to determine that all of the Corviission's safety requirenents have been met in accordance with 10 CFR Part 50 requirerents. The facility may then be operated only in accordance with the terms of the operating license for that facility and the Cenmission's replations under the continued sur-veillance of the Corrission's staff.
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In the course of our safety review of the $WES$4P1 design, to held numerous meetings
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with representatives of $ tone a Webster to discuss the plant design. analysis, and gj performance. During our review, we requested $ tone & Webster to provide additional.
g information needed for our evaluation. This additional inforretion was provided in 4
amendments to the $WIS$AR P1 applicatien. As a result of our revise, a number of changes were made in the facility desten. These changes are described in the amend.
monts to the $WES$AR-P1 appilcation and are discussed in appropriate sections of this.
6 report.
+
A chronology of the principal actions relating to the processing of the $WES$AR-P1 application is (*1uded as Appendia B to this report. The bibliography for this report is enclosed as Appendiz C.
The Report of the Advisory Committee on Reactor Safeguards is enclosed as Appendia D.
The $WES$AR.P1 standard balance-of-plant design doe. not include all parts of a nuc? car peer plant fad'ity. The design includes by reference the above identified i
standard nuclear steam supply tystem designs. Thi.s. the $WESSAR.P1 design is based on safety reisted interface requirments for the balance.cf-plant dest;n as estabitshed by the nuclear steam supply system designs. As a starwi4rd design SWESSAA.P1 does not l
include those portions of the design and analyses of an entire nuclear powr plant l
that are related to the characteristics of a specific site for the facility and to the I
utility ppitcant referencing the $WE$$AR-P1 design. Stone & Webster has therefore e
established in SWE$$AR-P1 specific safety related interface requirements for those f
systems or programs that are not within the scope of SWESSAR.P1 and which must be i
addressef. by a utility applicant that references the SWESSAR-Pl design in its construc-tion permit application.
Stone & Webster has taten encantons ',o some of the systems and coreponents within the scope of a standard nuclese steam supply system design in order to make the $WE$$AR.P1 desipn more adaptable to all the standard nuclear steam supply system designs refer-enced in SWE$$AR-P1 and to optistre the overall plant design. Stone & Weister has Identified these exceptions in $WES$AR-P1 and has stated that the changes in design to accomodate the exceptions have been reviewed and approved t;y Westinghouse for the
$WES$AR-P1/RESAR-35 design combinatiot We have identified and evaluated these design changes in SWESSAR-P1 to the RESAR.35 design in the appropriate sections of this report.
1.2 General Plant Deser4 tion 9
The $WESSAR.P1 standard talance.of-plant design discussed in this re ort corplements the RE$AR 35 standard nuclear steam supply system which is incorporated by reference e
inco the $4ES$AR.P1 design. With the exception of site related systems (e.g. ultimate l
e tat sin 6.) and with the exception of utility related aspects (e.g. pre. operational e
test progen). the combination of the $WE$$AR P1 design with the RESAR-3$ design results in a complete nuclear po.er plant.
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4 y The proposed SWES$AR.P1/RESAR.35 standard balance.of plant design appifcation is for a Preliminary Design Approval for a plaat with a masimum core thennel power level of j
3411 megawatts, resulting in a pet electrical output of approximately 1180 negavetts l
from the plant. The analysis of the engineered safety features within the scope of SME5NP1 has been performed for a nazimum cor' thermal power of 3565 segawatts.
These power levels are consistent with the maximum design and appilcation power levels of 3636 megawatts and 3411 megawatts, respectively, for the kN35 standard nuclear.
i steam supply system.
I The REh35 design for the nuclear steam supply system portion of a pressurized water t
reactor nuclear power plant encompasses the reactor coolant system, caergency core cooling system, reactor control and protection systems, engineered safety features l
. j actuation s) stem, chemical and volume control system, boron recycle system, residual l
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heat removal system, feel hand 11ag equipment. and related systems and features. The RESAR.35 de.ign uses a two train concept for the engineered safety features. The 6
design requires two diesel generators and four 125 Volt direct current supplies as independent on-site emergency power sources, to supply the power for each of the two a
redundant engineered safety features trains and each of the four channels of the reactor protection system, respectively. The interface requirements of REW35, which must be met by a balance.of. plant desten utilizing the RI5AR.35 design, have been identified in the SWE5Wp1 Safety An.slysis Report. The subject of interfaces f
is discussed further in Section 1.8 of this report.
The SWESSAR-p1 design includes all structures for an entire nuclear power plant.
engineered safety features in addition to those within the scope of REW35, instru.
mentation and control systems for systems within the scope of SWE5Wpl. inputs to the REN35 reactor protection system. the electrical power systems. the steam system beyond the steam generators, the turtf u generator systee, radioactive waste systems.
and availlary systems. A detailed listing of all structures, systems and corponents within the scope of the SWE5Wp1 design is provided in Table 1.81 and Table 3.2.51 of ShE5@pl.
The layout of the major structures of the proposed SWESSAR-P1 plant is shown in Figure 1 1.
The containment building is a reinforced concrete. steel. lined structure, which will house the nuclear steam supply system and portions of the engineered The containnent building all) be provided with flitration safety features systems.
'and spray systems to limit and contain radioactive material that could be released in the r *ikely event of an accident.
l The annulus building, to be constructed on the same contieuous base mat as the con-tainment butiding, surrounds the containment building along its entire periphery and to about one. half of its overall height. The annulus building houses portions of the The corponents for each of engineered safety features systec.s and auntliary systems.
the trains of the engiceered safety features systens in the containment building and the annulus building are separated front each other by distance and physical barriers.
15
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SOLID WASTE &
9 AREA TANK DECONTAMINJ. TION 90lLOING
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DIESEL GENERATOR
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CONTROL CONTAINMENT
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AND mg CHEMISTRY LAn NORMAL AUXlLIARY 80lLER ADMINISTRATION SWITCH-BUILDING GEAR TURBINE a
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The annulus building is designed as a partial secondary containment for the reactor j
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' containment butiding. It will be served by a suppiamentary leak collection and i
release system, which collects and processes the leakage of radioactive material from
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the contairment late the annulus butiding prior to (Ls release to the environment.
following a postulated design basis accident.
1 2
l In addition to the partial secondary contatruent as the basic design concept. $ tone &
Webster has included in $WE$$4P1 e full secondary contatteent building which com.
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plately encloses the primary containspat building to further reduce the direct leakage 8
from the prisery containment to the environment. This secondary enclosure building
,I l can be selected by a uttitty appiteant referencing the $WES$4P1 design for a site
^
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with less favorable meteorological characteristics (atmospheric dispersion factors) t than those appitcable for the partial secondary contatament concept. The full second.
ary enclosure building is provided in SWES$4P1 (1) without mixing and (2) with mix.
ing of the atmosphere in tie enclosure building, which are identified as Design g
g Optioa A and Design Option 8. respectively.
g The steam and power conversion system will be designed to reuove the heat from the RESAR.3$ nuclear steam supply system in the steam generators and to convert it into I
7 electrical energy by means of the steam turbine generator located in the turbine building. The high energy main steam and feedwater piping for the four RE$AR.35 ste.m generators will be routed in two pairs from the containment building (one pair each j
from the near and far side with respect to the turbine building) over the roof of the annulus building and through tunnels into the turbine butiding. This arrangement provides for physical separation of high energy piping from safety related systems and j
equipment. The circulating water systes, which util remove waste heat from the rein j
steam conde'nser to a heat sink.~is not within the scope of the $wES$AR.Pl design and
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will bs provided by a uttitty applicant referencing the $bESSAR.P1 design.
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The reactor protection system, which will automatically initiate appropriate corrective action whenever a plant parameter monitored by the system appreaches pre estibitshed f
limits, is within the scope of RESAR.3$. Plant parameters within the scope of
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$WESSAA.Pl that are input to the reactor protection system include reactor coolant pump undervoltage, reactor coolant pupp underfrequency, and turbine trip. The reactor j
6 protection system and the engineered safety features actuation systm will act to shut l
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down the reactor, close isolation valves. and initiate operation of the engineered
! I safety features should any or all of these actions be required.
l The service water system will provide cooling water for all components necessary for safe shutdtwn. The system consists of two independent trains for the $wf.5$AR.Pl/
RESAR.3$ design combinatici, one of which is required to omvide cooling water for f
Safe plant snutdown in the unliktly event of an accident. The source of cooling water for the service water system for nor el operation and energency operation (i.e..
ultimate heat sink) is not within the scope of the $WE$$AR.Pl design and will he provided by a ultiity applicant referenctng tre $WESSAA pl design, 14 i
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The $1ES$AA P1 design will include tuo fast startinj diesel generators for the i
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$WE55AA-P1/RESAA 35 design combination located it. separate rooms in the diesel I
generator building, with their associated independent safety feature busses to provide f
adequate power to each of the engineered safety features trains for a safe shutdown under accident conditions with a concurrent lost of offsite power. Direct current 4
power to the four channels of the reactor protection system will be provided by four redundant 125-Volt busses and their associat.d battery banks.
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l The balance-of plant design described in the SWESSAA P1 Safety Analysis Report is for I
e single-unit nuclear power station. Stone & Webster has proposed in $WE55AR-P1 trat f
the same design can also be used for a dual unit station without any chanves in the j
layout shown in Figure 1-1 for a single unit station. In general, each unit is a j
[i separate entity and only a limited amount of sharing will be required in a dual unit f
s tation. The structures and systems to be shared by both units will include the t
f011owing:
l (1) Service butiding (machine shop, warehouse, etc.).
l (2) Administration building (offices).
I l
(3) Auxiliary boiler providing steam to process equipment and the plant heating systen.
(4) Water treatment fact 11ttes.
j (5) Parts of the fire protection system.
I (6) Hydrogen recombiners for the combustible gas control system of each unit.
I With reference to Figure 11, the second unit would be located imediately to the
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right and in the same orientation 45 the unit shown in the figure. Our evaluations f'
and conclusions presented in this report are for a $1ngle unit station only. We will require a utility applicant referencing the $WESSAR-P1 design for a multi-unit station to provide additicnal information in its preliminary safety analysis report regarding the multi unit concept, for example, the site related layout for the entire station as related tc, the potential damage from postulated turnine missiles.
4 5 tone & Webster has identified in SW($$AR Pl those systems, components, and operational prograet for an entire nuclear power plant that are dependent on the characteristics of a specific site or on the operation of a utility applicant referencing the SWC$$AR-P1 deign. Table 1 1 lists the rajor items that are not within the scope of the $WESSAA-P1 application and that must be addressed by the utility applicant in its safety analysis report.
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MMOR ITDt$ TO BE ADORtsstD BY UTILITY APPt! CANT REFERIMC19G j
j swE55M-P1 M51GN j
(1) Site evaluation (infomation on geography. demography, and nearby industrial.
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transportation and military facilities; verification that site characteristics are within envelope defined in SWESSM-P1).
(2) On. site meteorological measurements program.
(3) Ultimate beat sink.
($) Service water system (intake structure at ultimave heet sink and water chemistry control).
i 4
r (6) Water treatment system.
(7) Putable and sanitary water system.
I (8) Offsite power sources.
(9) Sedtch yard.
(10) Hydrogen reccznbiner (selection and testing of casoonent).
j (11) Spent fuel cask.
(12) Emergency diesel generators including auxiliary systems (utility applicant will select manufacturer).
(13) Steam turbine. generator (utility applicant will select manufacturers SWE55AA-P1 includes option of Westinghouse or General Electric Casany system).
i
'14) Laundry waste system (system described in SWESSM.P1 is optional).
(1k) Fireprotectionsystem(sourceofwaterforsystem).
(16) Health physics program.
(17) Preoperational test program.
(18) Inservice inspection program.
(19) Technical specifications.
(20) Emergency plan.
(21) Industrial security plan.
(22) RESAR.35 nuclear steam supply syster.
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1.3 Cosearison With $1milar Facility Desions The $ESSAR P1 design as a standard balance-of plant design utilizes different nuclear I
steam supply systems and will be used for sites with different characteristics.
Therefore, the NES$AR P1 design differs from a custom balance-of-plant design for a
?
utt11ty applicant. The speific arrangement of major building structures, the methods l.
E of separation for engineered safety features trains, and the routing of high energy 1
piping used in other nuclear power plants differ from those that will he used in the
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$FE$5AR P1 design.
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Stone & Webster has pr9vided in tabular form a comparison of the desipi and performance parameters of the balance-of-plant features for the $ESSAR P1 design with those for the Millstone fluclear Power Station Unit 3 (Docket Museer 50-423), for u4ich a Construc-tion Permit was issued by the Cosmission in August 1974, and which presently is under construction.
j The engineering design aspects, the design criteria and procedures, and the eethods of analysis for individual structures, systems and components that will be used in the
$WE55AR P1 design are in many aspects similar to those we have evaluated and previously approved for other nuclear power plants. To the extent feasibts and appropriate, we have made use of our previous evaluations during our review of those features that are sinitar in the $WE55AR Pl design. Where this has been done, we have identi* fed in this report the specific Safety Evaluation Reports involved. These Safety Evaluation Reports are available for public inspection at the Nuclear Regulatocy CornissioVs Pubile Document Room at 1717 H 5treet, N.W., Washington, D.C.
1.4 Identif; cation of Acents and contractors The Stone & Webster Enginesring Corporation ulli design the $WE$5AR-P1 standard balance-of-plant, and is on its cwn behalf the applicant for a Preliminary Des'Ign Approval of the design. There are no other agents or contractors associated with the
$WESSAR P1 application. While Stone & Webstte has incorporated by reference the RE$AR-33 (and other) standard nuclear steam supply system design into the $WE15AR-P1 design, $ tone & Webster does not act in any form as a representative for Westinghouse with regard to the RESAR-3$ design or in any other matter.
t 1.5 Requirewents for Further Tecnnical Information We have concluded in Section 19.0 of this report that the information orovided in
$WESSAR-Pl is sufficient for issuance of a Preliminary Design Approval. No further technical information is needed for such issuance.
1.6 Su rury of Principal Review Matters i
Our technical review and evaluation of the information sutcitted by Store & '.'ebster in support of the SWE5SAR Pl appitcation considered the principal ratters s.rsnarized below.
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We enviamed the msical site cnaracteristics, including seismology, hydrology and
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esteerology to determine that appropriate comiderction had been given to these o,'.
characteristics in the developeant of the site parameter envelope which util be used a
for the siting of a nuclear poner plant utilizing the 5455ANP1 desip.
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l We reviamed the design criteria and the espected perforumco characteristics of the 1
facility structures, systems and components important te safety to deterwine whether
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they are in accord with the Commiission's General Design friteria. Quality Assursnew
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f; Criteria. Regulatory Guides, and other appropriate codes and standards, and that ary Y
departures from these criteria, codes and standards have been identified and justified.
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h We reviewed the response of the structures, systems nad components within the scope k
of SkE55AR-P1 to certain anticipated operating transients and postulated accidents.
j We considered the potential consequences of a few highly unlikely postulated accidents d
l (design basis accidents). We performed conservative analyses of these design basis i;
i accidents and deterwined that the calculated potential offsite doses that sight result in the very unlikely event of their occi.rrence would be within the Commission's
, guidelines for site acceptability as given in 10 CFR Part 100 for the site envelope conditions identified in SWESSAR.Pl.
l i
We evaluated the plans and measures described in SWE55AR-71 regarding the industrial.
f security aspect of the balance-of-plant design to deterwine that they can be incor.
f b,
porated into the industrial security plan by the utility applicant referencing f
SWE55AR-Pl.
We evaluated the design of the systesrs provided for control of the radioactive efflu-
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ents from the facility to determine that these systems can control the release of radioactive wastes from the facility within the limits of the Comission's regulations.
10 CfR Part 20.
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1.7 Outstandine issues v
f.:
In our Report to the Advisory Comittee on Reactor Safegards, dated June 1976. on
).
the SWE55AR-P1/RESAR 35 design combination, we identified three outstanding issues.
Two of these issues directly relate to two outstanding items in our review of the RESAR-35 application. Since that time we have resolved these issues for RESAR 35 and have incorporated these resolutions in our review cf the SWE55AR.P1/RESAR 35 design costination. The re.olution of these two items is disca:ssed in Section 7.3;a of this report. For the third issue, the main steam line break accident. Stone & Webster h
provided additional information in accurdance with our requests. Th? resolution of f
tPis item is discussed in Section 6.2.1 of this report.
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The outstanding issues identified in our Report to A5 for the SiiE55AR-P1/RESAR-35 design contination have been resolved in a manner acceotable for issuance of a Preliminary Design A; proval.
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in our effort to develop a consistent and reasonable policy for handling the interfaces 7
between various portions of a nuclear power plant design, as for exagle between a 7
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standant nuclear steam supply system design and a standard balance-of plant design. or between a standard bstance-of-plant design and a utility appilcant. we have held fj 7
numerous staff meetings and have set on maay occasions with nuclear steam supply I
system vendors, architect engineers. including Stond & idebster, and utility appitcants.
l II-iI f.
In order to approve any overall design of a nuclear power plant we must have the
!4
?
opportunity to discuss and review the dusign and the manner of integrating sojor
!k cogonents and systems into the overall design concurrently with the major parties participating in the development of the design. Zaperience with the integration of a standard nuclear steam supply system. for which a preliminary Design Approval has been
=
issued. into an overall plant design has indicated that it is not possible te approve I
g an integrated design without joint discussions with and appropriate doctmentation from all parties involved in the joint design. Esperience has also demonstrated that the integration of a standard design of a major system, for which an unconditioned
[
l Preliminary Design Approval has been issued. is a relatively simple task.
['
For the proposed preitsinary design of the SWE55AR-pl/RESAR 35 combination we have not
(
had this opportunity to discuss and review the integrated design with Stone & Webster and Westinghouse in jolat conference, e.or do we have available documented verification by Westinghouse of its review and approval of the manner of integrating its RES R-35 nuclear steam supply system into the SWE55AR-p1 balance-of plant design. The utility applicant referencing the SWE55AR-P1 standard balance-of-plant design and the RISAR-35 standard nuclear steam supply system design will. therefore, be responsible for decon-strating under oath or affirmation that the necessary interchanges for its design of a total ruclear power plant will be implemented among all parties participating to assure that all gortfors of the design will be properly integrated. At suen time we vill be able to determi e that the integration will be accomplished in an acceptable e
e manner.
[
The technical informatten presented in a preliminary safety analysis report by a util-ity applicant in suppSrt of its aMi* stion for a construction permit for a custom dasign reflects interaction between the utility applicant, who assumes full responsi-3 bility for the preliminary design, with the vendor of the nuclear steam supply system and with the architect engineer who generally designs the balance-of plant portion of the er. tire facility. As stated 4 Section 1.4 of this report. Stone & Webster as the j
applica91 for a Preliminary Design
% val of the SWESSAR-p1 standard balence-of-plant design, is acting entirely on its own behalf in this application. Based on our review of the 5WE55AR-P1 application, we determined that the appropriate eschanges of design 4Ad aaelysis information between the organizations and the development of specific programs related to the overall design, which take place during the prelimina y design of a nuclear power plart unde-the direction et tht utility appiteant, are not suf f t-c.stly reflected in the SWESSAR.Pl application.
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- Stane & Webster has identified in SWE55AR.P1 certain asnects of the propoied SWE55AR-P1 I
design and supporting analyses thereto which will require enchanges of infomation f
among the utility applicant. Stone & Webster and Westinghou;e to assure the compati.
y bility of the proposed $WE55AR P1 design with the overall preliminary design of the i
J nuclear pouer pleet. We have listed those matters in Table 1 2 and have identified fJ W
each item in the appropriate section of this report. These issues are laterface k
matters that mest t.e addressed by the utility applicant in its construction permit
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sppilcation in order to support the proposed SWE5LAR.P1 preliminary design for the 3
helance-of. plant design.
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c 1.8.1 SWEs5AR.P1 faterfaces with RESAR 35 j
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' 1 In our Safety Evaluation for RESAR 35, w concluded that the interface requirements k
established by Westinghouse and the staff are sufficiant to determine the compatibility g
of the safety related systems and components within the scote of RESAR.35 with a j
y l
balance-of. plant design referencing RESAR-35. We also concluded t W the interface
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inforention provided in RESAR.35 and in our safety Evaluation for RESAR 35 are adequate 5;
to determine the validity of the RESAR.35 accident analyses when RESAR 35 is referenced Y
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in a balance-of.plaat application.
1 Stone & Webster has identifled throughout $WES$AR-P1 the safety related interface
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j requirements estabsished for the RESAR-35 and other standard nuclear steam supply U
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a systee designs and the corresponding SWESSAR P1 interface infomation to demonstrate j
the compatibility between the te designs. The identification is provided in a sep-i arete interface subsection for each SWESSAR.P1 system with cross references to appro-i priate sections, figures and tables in teth RESAR 35 and SWE55AR.P1, including summary i
fk tables in Chapter 1 of the $WE55AR-P1 Safety Analysis Report. Interface points in connecting lines of fluid systems are clearly idee'ified using alphanuneric SWE55AR-P1 symbols and cross referencing the coreesponding RESAa.35 syntiols.
?
In some instances the analysis of a SWESSAR.P1 systes has not teen performed in accord-L ance with the appropriate requirements identified in RESAR.35. For exa gle, the SWESSAR.P1 input parameters are not consistent with tt.2 Westinghouse siniensa contain.
h ment back pressure calculation which is part of the emergency core cooling system performance evaluation. We have identified, in the appropriate sections of this
^
report, the interface discrepancies in the analyses. These issues must be addressed
[4 j
in the construction permit application of a utility appilcant referencing the i
$WE55AR-P1/RE$AR.3$ design combination.
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As stated in Section 1.1 of this report. Stene li Wetister has made sore changes to the l
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RE*,AR.15 stanoard nuclear steam supply system design as it is utilized in the SWE55AR.P1 standard balance-of. plant design. These thanges involve the identification and loca.
tion of interfaces between a RESAR45 system and a $WES$AR-P1 system in order that the sam SWESSAR.P1 system can be used with all standard nuclear steam supply systems a
r utilized in SWES$AR.Pl. For example, the RESAR 35 design indicates a connection t,etween the reacter coolant system (RISAR 35 scope) and the prieery grade water system 1
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e IRTERFAct MATTTPS F0'. SWf 55AA-P1/Rf5AR-35 Of510's CCsetNATION t
TO et Acca 55fD ef UTILITY Aprt1CANY
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The folloutng matten require escranges of information aseng the utility appiteant. Stone &
A Wettster end laesttnpeese to assure the cos9atibility of the proposed SWl55AR-P1 design trith the overall crel'i'aary design of the nuclear p:ver plant. A utt11ty applicant referencing the j'
SWESSAA-P1/RESAR 15 design comeination in its construction permit application shall address the follower.g inta-face matters and provide sufflctent Information to enable the staff to Complete its rolew of these late-f :e matters. Tis section(s) of r.nts report where these matters are F.
discussed in detatt Pave Deen identitled in perenthese9 after each item.
L (1) Envelope for short tem atectpheric dispersion factors (2.3.4).
i a
h (2) Ultimate bet sina cspact11ty (7.4.2. 9.2.3).
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I; (3) Imp 1:sentation of design responsiblitty for reactor coolant system and supports (3.6.1.3.9.1).
L F
(4) Developumat of preoperat!onal vibration test pro; ram (1.'.1).
(5) Development of seisste qualification progria (3.8.1;.
L (6) Design provisions to achieve cold shutdown 6 sit 1 cely safety-grede systants (5.1).
(7) Evaluation of reovirement for single failure ocoof polar crew (5.3).
(8) Containment pressure analysis for main steam Mne break accident with appropriate Westinghouse data (6.2.1).
(9) Preoperational test program for chemical additive system (6.2.4).
E (10) Definitice of proposed containment purge operations (6.2.5).
(11) Design details for the hydrogen recostiner subsystem (6.2.6).
(12) Provisions for self-contained breathing apoaratus (6.4).
I i
(13) Verification of lWESSAR-P1/PE5AA 35 netal water reaction assumption fo? hydrogen generatie (6.2.6).
l (14) Minimum contatrument tackpressurt calculation for SWE55AR P1 in accordance with Westing-house eeergency core cooltag evaluation (6.3).
(15) Implementation of IEEE Std 279-1971 forreactortripsysteminests(7.2).
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TABLE 1-2 (Continued)
(16) Environmental qualification of Class it equipment to meet objectives of IEEE Std 5
3231974(7.6.1).
(17) Provisions for ensuring that the independence of the safety related 4.16 kilovolt i
emergency twses are retained (8.3.1).
(18) laplementation of IEEE Std 279 1971 forelctricalpenetrationconductors(8.3.1.2).
(19) Specification s. fire stops and seals, quality assurance program, test procedures.
andquellficationtesting(8.3.1.4).
k (20) Qualification of reactor coolant psamps or provision for.mnent cooling mater in i
accordance with $W[$$AR.P1 interface requirements (9.2.2).
(21) Design provisions for the radiation area monitoring systen (12.2).
(22) Occupational radiation dose assessment (12.3,12.4).
3, (23) Program for initial tests and operations (14.0).
t (24) Exclusion boundary and low population zone distances (15.2.2).
(25) Offsite doses due to the postulated post loss.of-coolant accident hydrogen purge (15.2.3).
(26) Raciological consequences o' the rod ejectien accident as a function of primary to secondarysteamseneratorneakage(15.2.5).
(27) SWESSAR.P1 design changes on basis of RISAR 35 analysis of anticipated transients without scram (15.3).
i 1 15
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($WE$$AR.P1 scope) Gutside of the containment (RESAA.35. Figure 5.1 1). Stone &
Webster has moved the connecting point inside the containneet so that only a single containment penetration of t% primary grade water system is required for all prirary grade water needs inside containment ($WESSAA.Pl. Figure 5.1 1). We have reviewed these changes and conclude that they have been identified in a concatible and accept.
able manner.
i i
The boven recovery system. In acccrdance with the guidelines of WASH.1341.' Amendment 1 is within the scope of a standard nuclear steam supply system design and. accordingly has been identified la RESAR 35 as a Westingtouse system. In SWES$AA.Pl this entire system, which is not safety related, has been replaced by a $ tone & Webster designed boron recovery system as discussed in Appendis A of this report. We have reviewed this system within the context of this safety evaluation but. as we did with other standard design applications that include systems which are outside the guidelines of WASH.1341 Amendment 1 for such a standard design the Preliminary Design Approval for the $WE$$AR.Pl design will not encon9 ass the boron recovery system.
If the systems and cosmonents discussed in Appendis A to this report are submitted as topical reports with no change in the technical design of the system. then our evalua.
tions contained in Appendix A should remain valid. These eveluations may then be t
l utilf red as part of our review of utility applications referencing standard plants.
l Stone & Webster has identified in $WESSAA.Pl the design changes and replacewent made by the $WE15AR.Pl systems to the RESAR-35 design. Stone & Webster has also statcJ that for all changes to the RESAR.35 design a technical evaluation and approvst has been cotained fran Westinghouse. We have reviewed the design changes as documented in
$WES$AR.Pl and concl6de that the $WESSAR-P1 design can be interfaced with the RESAR.35 desig"n as described in $WES$AA.Pl. However, as stated p'eviously, we will require a utility applicant referencing the $WESSAR.P1/P'SUt.35 design to identify these changes in its construction permit appilcation and to demonstrate the compatibility of the two desig*.s.
We have reviewed the $WESSAA.P1 method and detail of presenting RESAR.35 interface requirements and the corresponding $WE$$AA.P1 Sterface information and conclude that it is an acceptable approach to demonstrate adecuate compatibility of the two designs with regard to specific systems for the purpose of a prs 11minary Design Approval. but as discussed in Section 1.8 above is not acce: table for a construction permit. A Preliminary Design Approval for the $WIS$AR.Pl design will apply only to the telance.
of. plant portion of tre SWE$$AR.P1/RESAR.3$ c:lrtined plant design and not to the RI$AR.35 nuclaar steam supply system in that design con 61 nation.
1.8.2 Swi$$AA.pl in*erfaces with $1te and Utility A plicant Stone & 'mebster has identified interface requirewnts of the $WES$AA.Pl design that are related to the siting and cperation of a nuclear power plant facility uttilzing the $WES$AA.Pl standa'rd balance of. plant design. These interface recuirenents need to 1 16
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l be addressed by the utt11ty appilcant for the facility. In addition. Stone & Webster lies identified in Section 1.8 of SE55AA P1 those sections of a safety analysis report to which the utility applicant referencing the $' 55AA-P1 design eut provide input to 4
ensure the completeness of the report for cur review. In Section 1.2 of this report,
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we have identified the major items that must be provided by the utility applicant.
We conclude that the SWESSAR-71 interface requirements regarding the siting and opere-tion of a SWESSAR-71 nuclear power plant are acceptable.
1.8.3 Conclusions We have reviewed the interface inforsetton p m idad by Stone & Webster in SWESSAA-P1, through Amendment 31, with respect to the RESAR 35 interface requirements. We conclude that the interface requirements of RESAR-35 have been acceptably addressed in the
%E55AA-P1 app 11 cation and that the SWE55AR-P1 standard balance-of-plant design can l
support in a compatible sonner the safety related systems and components of the RESAR 35 j
steMard nuclear steam supply system design. We will require the utility applicant to i
demonstrate the compatibility of the SE55AA-P1 design with the RESAR 35 design when integrated by a utility applicant in an entire nuclear power plant facility.
We have also reviewed the interface requirements of the $ESSAR-p1 design as related to the siting and operation of a nuclear power plant facility based on the SWE55AA.P1 l
desir. We conclude that this information is sufficient to determine the compatibility of the SWE55AR-P1 design with the site and operation related aspects of the facility.
i We conclude that the interface information provided in SWE55AR-p1 and in this Safety Evaluation Report is acceptable for the issuance of a Preliminary Design Approval for _
the SWE55AA-p1 standard balance-of plant design in its relationship to the RESAR-35 standard nuclear steam supply system design.
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2.0 SITE CHARACTERf571CS e
The $WE5NP1 appilcation does not include the characteristics for a specific site.
This information will be provided by a utility applicant in its safety analysis report that references the SE55AR.P1 design. However. Stone & Webster has established in SWE15AR-P1 an envelope of astoorological. hydelogical and seisselegical site con.
ditions for the SW15AR P1 design. The conditions provide an indication. in advance of the esamination of a particular site. of the type of site for which the SWESSAR P1 design is suitable. Stone & Webster has summarized the coWittons as interface require-ments for a specific site in SWE55AR P1. Section 2.5.
The conditions are discussed in Section 2.3 through 2.5 of this report. We will i
evaluate the characteristics of a specific site selected by a utility applicant l
referencing SWE55AR-p1 as described in its i:eastruction pomst appilcation, in order j
to confirm that the site characteristics fall within the envelope of conditions reviewed and evaluated for the SWE55AR PI application.
1 2.1 Geography and Demooraphy I
This subject is not within the scope of SWESSAR.PIs it will be eddressed by the utility appiteant.
2.2 Nearby industrial. Transportation and Military Facilities This subject is not within the scope of SWESSAR-Pls it will be addressed by the utility applicant.
L 2.3 Meteorology 2.3.1 cooloaal climatolocy f
Stone & Webster has provided an adequate description of the w teorological conditions l
used as the bases for the safe design and siting of a nuclear power plant referencing SWE55 & Pl.
The design basis tornado. with a eximum wind speed of 360 miles per twr casisting p
i of a maaimum rotational wind speed of 290 miles per hour and a maxisun tr. slational wind speed of 70 miles per hour, a masinum pressure drop of 3.0 pounds per square inch. and a maximum pressure drop rate of 2.0 pounds per square inch per second.
j conforms to the reconneWattons of Regulator / Guide 1.76 and is sufficient for all regions of the contiguous United States. The plant design for protection against tornada damage 1$ considered to be nore than adequate for protection against water-Spouts. The operating basis sustained wind speed (* fastest mile') for the plant t-1 7
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I design is 120 miles per hour at a heignt of 30 feet above ground level, with a l'
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,1 recurrence period of 100 years. This is adequate for appresleetely 90 percent of the 4
contiguous United States. Areas in wL his design wind speed neuld be expected to be exceeded are generally restricted to A61 antic coastal sections free Maryland southmerd through the Carolinas, the coastal portions of Florida from Cape Canaveral l
co Cape $abel, and offshore locations along both the Atlantic and Gulf coasts.
The $WES$ Alt P1 design basis for snow accoulation. In terus of snow load on the I
ground. is 80 pounds per square foot. This lead is generally expected to be exceeded j
only in asuntainous areas of the United States and in some areas near the Great Lakef. In about 90 percent of. the contiguous United States, samtaum snouloads would not exceed the lead c,f 80 pounds per square foot specified in $WES$AR-Pl.
i The design einfanas dry-bulb air temperature of -40 degrees Fahrenheit is not expected to occur more frequently than 1 percent of the time anywhere in the contiguous United 1
states. The design maximum dry-bulb air temperature of 100 degrees Fahrenheit is not expected to occur more frequently than 1 percent of the time in approximately 80 to 85 percent of the contiguous United States. Areas in which the frequency of occur-rence of temperatures of 100 degrees Fahrenheit ney exceed 1 percent include much of central and southern California, the desert southmest, the southern Great Plains from the Rio Grande north to the Platte River. and desert areas of the northwestern United States.
2.3.2 Local Meteorology This subject is not within the scope of $WE$$AA-Pl; it will be addressed by the e
uttitty applicant.
I i
2.3.3 Onsite Meteoroloaical Measurements progras This subject is not within the scope of SWE$$AA Pls it will be addressed by the utility applicant.
2.3.4 Short-Term (Accident) Ofspersion Estimates The $wtsiAR.P1 standard balance-of plant design appilcation does not include the atmospheric dispersion characteristics for a specific site. An discussed in Section 6.2.2 and Section 15.2.2 of this report. the $WES$A2 P1 application inC'f udes three separate secondary contairv:ent design concepts. Stone & Webster has evaluated the suitability of each of these concepts for potential sites by determining the maxisue short term (0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) atmospheric dispersion f actor (1/0 value) for each concept such that the resultant of fsite deses (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, esclusion boundary) mill meet the guidelines of Regulatory Guide 1.4.
Tre nastrum values determined by Stone & Webster are listed in Table 2-1.
The table incluces also the results of our Independent evaluation of the maxi un short term (0-2 Pours) atrospheric dispersion factors wh8ch is discussed in Section 15.2.2 of this report. On tLe basis of our evaluation we 2-2 e
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MAXIMM ATMO5PHERIC Of5Ptits!0N FACTOR $_
r (0 2 HOURS I/O VALUE51 o
FOR SWE55AR.P1 CONTAl*EMT CONCEPTS _
3 Maxisus f/0 Value (sec/m )
Contaivament Conceot.
l SWESSAR.P1 NRC 4.5 x 10~4 4.3 x 10 (1) Base Design (partial dual containment) t DesignOptionA(completedualcontain.
9.0 x 10~4 1.4 x 10~3 (2) mentwithoutmining)
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(3) Design Option B (complete dual contain-1.7 x 10~3 2.1 a 10~3 I
l mentwithmixing)
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t Abtreviations used in table:
Sec/m ~ seconds per cubic meter Nuclear Regulatory Consission staff NRC 1
1/Q Atmospheric dispersion factor
,f Based on analysis of offsite radiological conseasences at 4100 megawatts thermal III reactor core power.
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t conclude that the $WES$AR.P1 design is acceptable for sites with values that fall within the envelope as specified by us in Table 2 1.
Accordingly we define this as t
an interface requirement.
We have reviewed the short ters atmos,7heric diffusion uudel in SWES$AR.Pl to be used in the' dose calculations for the 30 day period following a postulated accident.
The model is based on the recommendations of Regulatory Guide 1.4 and we conclude.
therefore. that it is acceptable. This dose calculation consists of a sumation of doses for separate time intervals during the 30 day period, each interval having its own and independent atmospheric dispersion factor. Any combination of these factori as a set rather than an individual factor for a particular time interval will deter.
eine the atmospheric dispersion characteristics. Therefore, as discussed in Section 15.2.2 of this report. a specific site envelope for the atmospheric dispersion characteristics during the entire 30 day period cannot be determined. A utility f
applicant referencing the $WESSAR.P1 design will provide in its safety analysis
[
report the atmospheric dispersion data associated with the accidental releases from the plant buildings and vents and as related to specific distances to the exclusion l
boundary and the low population zone. We will evaluate this information to verify i
that the site dispersion characteristics so determined are acceptable.
2.3.5 Lono.Ters (Routine) Discersion Estimates t
f Stone & Webster has provided in $WESSAR.Pl. Figure 2.3.5 7. annual average atmo.
I spheric dispersion factors (X/Q values) as a function of distance from the plant for the routine releases of airborne effluents from tha plant. 'he graphs, which repre.
f sent the maximum annual average value expected in any direction from the plant, are based upon various annual average values calculated by us in previous reviews of 42 sites for nuclear pover plan?.s. Based on previous evaluations completed by us. the majority of sites within the contiguous United States would be expected to,have maximm annual average atmospheric dispersion factors that fall between the 5 and 95 l
percentile curves of the graphs for the 42 sites presented by Stone & Webster.
Ii The annual atmospheric dispersion characteristics presented in SWES$AA.Pl do not l j constitute a site envelope of these characteristics but indicate the potential range l
of long term diffusion estimates for the $WE55AR.PI design. We conclude. therefore, f)l that an evaluation of annual average atmospheric dispersion factors for routine
~
j releases will have to be made for each site by the utility applicant, using meteoro.
logical data collected onsite in accordance with the recomendations of Regulatory l
Guide 1.23 over at least one annual cycle i
2.2.6 Conclusions We conclude that the meteorological conditions of the $lte envelope for the $WE$$AR.P1 standard balance.of plant design have been analyzed in accordance with the recomenda.
tions of Regulatory Guides 1.4 and 1.76. and therefore, are acceptable. We will evaluate the meteorological conditions for a specific site. to be submitted by a 24 w
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I utt11ty-appiteant referencing the $W$$AA-PI design, to determine that the conditions fall within the envelope described in SWES$AR Pl. We will evaluate these conditions
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e with regard to the 10 CFR Part 100 guidelines for dose calculations and with regart' to
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the safe design and operation ef the plant in combination with the distancas for the esclusion aree and the low population zone for the specific site.
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2.4 Hydroloey k
The SW$$AR-P1 plant is designed to be located on a site with access to a natural body j
of water to provide the plant makeup water for safety related systems at all times.
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Detailed hydrological characteristics for the site will be provided by the utility applicant in its safety analysis report and will be evaluated by us.
J 1
2.4.1 Floeds Tu yard grade for safety related structures will be set above the water level that can
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be reached by the design basis flood, including coincident wind-generated wave effects.
as identified in Regulatory Guide 1.59. We consider that these bases are in accordance
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with position 1 of the guide and comprise acceptable criteria. The utility applicant j
referencing the $WES$AR-P1 design will provide the J.st.iiled hydrological conditions for
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the specific site end will determine the design basis flood based on these conditions.
l Additional protection for the plant. Such as external flood barriers. must be previded by the utility applicant should the design basis (1 cod for the specific site be at an elevation above the yaid grade (see also Section 3.4 of this rer. ort). We will review the additional information for the specific $1te selected by the utility applicant and
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described in its construction pemit application.
I Protection of the safety related structures against locally heavy precipitation will be provided by yard grading and drainage systems which carry surface runoff from the
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localized probable maximum precipitation away from the site without flooding safety related structures, systems or components. The rc.ofs of safety related structures will not have parapets and will be slope <1 to preverit ponding. We find the flood criteria 1
acceptable.
l l
l 2.4.2 Low Water Considerations and Ultimate Heat $1nk f
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This subject is not within the scope of the $W$$AR P1 design and will be addressed by the utility applicant. Stone a Webster has specified as an interface requirv ent of the $WES$AR P1 service water system (see Section 9.2.1 of this report) on 1% ultimate beat sink that it must be capable of removing the heat loads specified by the nu:lta*
steam supply system and by the $W$$AR P1 systems for all modes of operation. Stone &
l-Webster has also specified that the maximum allowable service water systee inlet I
temperature at the ultimate heat sink sna11 be 100 degrees fahrenheit. This interface requirement is further discussed in Section 9.2 of this repert.
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Our evaluation of nee was types and geographic locations of ultimate heat sinks indicates that the SIESSAA-71 service water system tegurature Itait of 100 degrees tj Fahrenheit could limit the type of ultiente heat sink for many geographic locations l
where once-through cooling mater cannot be provided for t e service wate system. We
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h advised Stone & Webster of our conclusions. Stone & Wei,aar ' concluded,,on the basis b
of its experience as an architsc. engineer in she design of nuclear power plants at
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different geographic lecations, that the above temperature limit will not E m lude 4
the use of the $1ESSAA-P1 standard balance-of-plant design for sost geographic l
e locations.
Ifl u
r3 We will evaluate the type of ultimate heat stat and the imperature limitations of the ultimate heat sink with respect to the service water tagerature limitation for j
the time period and environmental conditions described in Regulatory Guide 1.27 on j
the basis of site related information to be provided by a utility app 11sant referencing f'
the SWES$AR-P1 in its construction pemit application.
i) y 2.4.3 Groundwater t
The deterufnation of 0.6 groundwater level is not within the scope of the SWE55AR-P1 appliestica. However, safety related structu es will be protected against natural groundwater as discussed in Section 3.4 of this recort. The structures will be L
designed for a hydrostatic loading equivalent to a floodwater elevation at yad L
grade. We consider this to be a conservative and acceptable design basis. We will evaluate tl.e groundwater level for a specific site based on the information to be provided in the construction permit application of a utility applicant referencing the SWCS$AR-P1 design.
L 2.5 Geology and Seismology a
Geology, seismology and foundation engineering characte istics are not within the scope of the $WE15AR P1 application and will b* prov9ed for a specific site by the r
utility appiteant referencing $WESSAR Pl. Howedr. the plant design is based on the following envelope of site characterls-ics:
(1) The safe shutdown earthquake horizontal ground acceleration for seissfc design is not la ger than 0.3 g (g is acceleration due to gravity).
(2) The operating basis earthquake horizontal ground acceleration for seismic design is net larger than 0.15 g.
d (3) There is no surface faultf ag on and in tre vicinity cf the site that rust be considered in the plant design.
(a) The snear modulus of geological material v4erlying the plant foundations fall within a sco;e of 6.000 to 1.000,000 pour.h per square inch.
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These conditions are specified by Stone & Webster for a site as interface require.
ments. We find these conditions acceptable. In particular, the sefssic design value of 0.3 g for the safe shutdown earthquake is adequate for about 70 percent of poten-tial sites east of the Rociy Mountains. We will evaluate the seismologtui and geological characteristics of a specific site selected by a utility app 1tcant and described in its construction permit application to confirm that the site character-i istics fall within the above envelope conditions for the $WE55AR.Pl design.
We will also evaluate the geology. seismology, and foundation engineering characteristics, as
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required by Appendia A to 10 CfR Part 100. of each Individual site for which the
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$WE55AR-P1 design will be used.
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3.0 DistGn CalTERf A FOR STRUCTtmES. SYSTEM 5 AND Cm i
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3.1 Conformance with General Desian Criteria k
n Stone & Webster has stated that the structures, systans and components of the l
SK55AA-P1 standard balance-of-plant will be deshmed in accordance with the j
h Commission's General Des.gn Criteria for nuclear poner alangs and has discussed, in h
Section 3.1 of SW55AA-P1 the cogliance with each criterion applicable to the 3
design. On the basis of our review, we have concluded that the SW55AR-P1 standard
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balance-of-plant can be designed to meet t?.e reqvfruments of the General Design Criteria of Appendtx to 10 CFR Part 50.
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3.1 Classification of structurri. Systems and Components 3.2.1 Seismic Classiftcation j
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l' Criterion 2 of the General Design Criteria r% quires that nuclear power plant struc-l
(
tures, systems and components isportant to safety be designed to withstand the i
effects of a safe shutdown earthquake and remain functional (see also Section 2.5 of f
p this report). These plant features are those necessary to assure (1) the integrity i
I of the reactor coolant pressura boundary. (2) the capability to shut down the reactor l
and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the ccnsequences of accidents which could result in potential offsite i
exposures compreole to the exposure guidelines of 10 CTR Part 100.
We have reviewed.he structures, systems and cosporents important to safety that are I
within the scope 1,f SWE55AR-P1 and that will be desired to withstand the effects of a safe shutdown t<.rthquake and remain functional. They have teen identified in an
(
accsptable manner as seismic Category I items in SWE55AA-Pl. Table 3.2.5-1.
All other structurcs, systems and ccuponents that may te required for the operation of the f*cility will be designed to other than seksic Category I requirements. Included in this classification are those portions of selssic Category I systems which are not required to pe form a safety function.
We cor.clude that structures, systens and cogonents within the scope of SWE55AA-P1 that are important to safety will be designed in accordance with sefseic Category !
requirements. The basis for our acceptance is tre conformance of the appitcsnt's designs, design criteria, and design bases for structures. systems and compor.ents issportarit to safety with the Cossilssion's 19ulati:ms as set forth in Criterion 2 of
=
the General Design Criteria, and with the recormendations of Regulatory Guide 1.29.
staff technical positions and industry standards.
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3.2.2 System Quality Group Classification
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+j Criterion 1 of the General Design Criteria requires that nuclear power plant systems W
and components important to safety shall be designed, fabricated, erected and tested i
f to quality standards ccamensurate with the importance of the safety function to be l
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j 1
performed.
8 We have reviewed the classification system as presented in SWES3AR P1 for pressure-
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retaining components within the scope of the SWESSAR-P1 design such as pressure j
h vessels, heat exchangers. storage tanks. pumps, piping and valves in fluid systems
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b.
Important to safet), and the assignment of quality groups to those sections of j
E systems required to perfore safety functions, j
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Stone & Webster has applied the classification systen of the American Nuclear Society j
f consisting of Safety Classes 1, 2. 3 and tels (non. nuclear safety) which corresponds j
to the Cosmission's Quality Groups A. 8. C and D in Regulatory Guide 1.26 to those j
pressure-retaining components which are part of fluid systems important to safety.
Reliance is placed on these sve'--- (1) to prevent or mitigate the consequences of accidents and malfunctions originating rithin the reactor coolant pressure boundary.
4 (2) to permit shutdown of the reactor and maintenance in the safe shutdow.: condition.
I and (3) to cool other safety sys'
. The fluid systems that are within the scope of the $WESSAR-P1 design have been Alessified in an acceptable manner and on system piping and instrumentation diagrams in the SWES$AR Pl Safety Analysis Report.
3 The basis for our acceptance'is the conformance of the Stone & Webster designs.
design criteria. and design bases for pressure retaining components such as pressure
(-
y vessels, heat exchangers. storage tanks, pumps, piping and valves in fluid syste=5 important to safety with the Comission's regulations as set forth in Criterion 1 of the General Design Criteria, and with Regulatory Guide 1.26. staff technical post-S l
tions. and i:adustry standards.
L t
3.3 Wind and Ti,rnado Uestan Criteria P
3.3.1 Wind Desion Criteria All seismic Category I structures exposed to wind forces will be designed to with-(
j stand the effects of the design wind. The specified design wind has a velocity of l
120 miles per hour based on a recurrence of 100 years (see also Section 2.3.1 of this report). The wind velocity will be transfonned into pressure loadings on structures
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and into the associated vertical distribution of wind pressures and gust factors in accordance with ANSI A58.1-1972 of the Anerican futional Standa ds Institute (AN51).
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which we find acceptable.
vocedures that all) be utilized to determine the loadings on seismic Category [
"ruct res induced by the design wind specified for the plant are acceptable since t3ese procedures provide a conservative basis for engineering design to assure that the structures will withstand fuch environmental forces.
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l The use of these procedures provides reasonable assurance that in the event of design N
basis winds, the structural integrity of the plant seismic Category I structures will J
1 not be impaired and, in consequence, seismic Category I systans and components located withis these structures will be adequately protected and will perfom their
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intended safety functions if needed. Conformance with these procedures is an accept.
U' able bests for satisfying the appilcable regirements of Criterion 2 of the General t
4 Design Criteria.
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3.3.2 Tomado Deston Criteria w
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All seismic category I structures uposed to tomado forces will be designed to re.
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sist a tornado of 290 miles per hour tangential wind velocity and a 70 miles per hour f
I translational wind velocity. The simultaneous atmospheric pressure drop will tr f
"i assumed to be 3 pounds per square inch in 1.5 saconds (see also Section 2.3.1 of chis l
p report).
4 i
The procedures that util be used to transfom *.he tornado wind velocity into pressues tI I
i l
loadings are similar to those used for the design wind loadings as discussed in Section 3.3.1 of tels report.
T'.e tornado missile effects will be determined using l
procedures discussed in Section 3.5 of this report. ~he total effect of the design tornado on seismic Category 1 structares will be determined by appropriate comeina.
4i tions of the individual effects of the tornado wind load, tornado differential (I
pressure load and the tornado missile load, d
3 Structures will be arranged on the plant site and protected in such a manner that I
collapse of structures not sesigned for the design basis tornado will not affect the L
ability of safety related structures. systems and coseonents to perfom their func.
[
tion to achieve a safe shutdown condition of the plant.
The procedure lized to determine the loadings on structures induced by the design basis torNdo...cif f ed for the plant are acceptable since these procedures provide t
a conseewative basis for efigineering design to assure that the structures withstand y,
h such environmental forces.
e The use of these procedures provides reasonable assurance that in the event of a design basis tornado the structural integrity of the plant structures which have to P
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be designed for tor 9adoes will not be impaire4 and, in consequence, safety related
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systems and comComents located within trese structures will be adequately protected j
and can be expected to perform necessary safety functions as required. Confo.siance i
with these procedures is an acceptacle bests fcr satisfying the aopilcable rew ice.
rents of Criterion 2 of the General Cesign Criteria.
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- t, The desly flood level resulting from the most unfavorable condition or combination 1
of conditions that produces tre maximum water level at the site is discussed in 6
Section 2.4 of this report. The hydrostatic and buoyancy effects of the flood will be considered in the desip of all selsric Category I structures exposed to the water
(
head. All seismic Category I structures will be protected from the damaging effects y
of flood water up to the elevation ef the station yard grade. which is the desip k
bests flood level for the SW55AR.P1 desip. Flood protection. such as dikes or U
other enternal structures. will be wovided for a design basis flood level above yard h
grade elevatfon. The design of such additional flood protection is *ot within the E
scope of SW55AR-P1 and will be provided for a specific site ty the utility applicant in its safety analysis resort.
Yh All safety relarj systems and components requiring flood protection will be located (g
in seismic Category I structures. All exterio-building openings which come,nicate t
with safety related components will be located above the station yard grade eleva.
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tion. Structures contaleing safety related components will be protected from ground.
E water wit % a waterproof membrane as an addittoral protection for sites where the Y
groundmater tatle is above the foundation mat. The moerane will be sealed to the outside of the annulus building at an elevation of 5 feet above the maatmum ground-f water level for the specific site (see Section 2.4.3 of this report). Idater-stops will be provided in all construction joints in exterior walls and slabs.
I The proc &dures utilfred to determine the loadings on seismic Category I structures induced by the design flood or highest groundwater level specified for the SWESSAR-P1 design provide a conservative basis for engineering design to assure that the st vc-E I
tures will withstand such environmental forces and. therefore are acceptable.
I R.
The use of th:se procedures provides reasonable assurance that in the event of floods.
up to the elevation of the station yard grade or high grounesater. the structu al a,
j integrity of the plant setsric Category I structures will not be impaired and, in f
I consequence, essential systems and cogonents located within these structures will be adequately protected and may be espected to perform necessary safety functions, as required. We ccnclude that conformance with these design procedures is an acceptable basis for satisfying the applicable requir%ents of Criterion 2 of the General Design b
Criteria. Acceptability of the additional protection against a design basis flood above yard grade elevation will be determined during our review of a specific site on the basis of the hydrological characteristics of tre site and the design of the soecific additional flood protection as described by a utility applicant in its safety analysis report referencing S'aiE55AR.Pl.
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3.5 Missile Protection 3.5.1 Missfie Protection criteria The SWESSAR-P1 plant will be designed so that missiles from internal sources and from outside of the contairment will not cause an accident or increase the severity of an accident.
Safety related systees will be protected against loss of function due to internal missile impett. Pressurized components and rotating machinery are potential internal missile so m es. These include retaining bolts. control rod drive assemblies, valve I
bonnets. sad valve stems. Protection against such potential missiles includes proper orientation of the potential missile source, missile barriers, and physical separa-tion o'/ redundant safety systems and components.
?
We conclude that the design criteria and bases are in conformance with Criterion 4 of the General Design Criteria as it relates to structures housing essential systems and to the capability of the systems of withstanding the effects of internally generated '
missiles. Regulato:y Guide 1.13 as it relates to protection of spent fuel pool
' systems and spent fuel assemblies from internal missiles, and Regulatory Guide 1.27 as it relates to the design of the intake structure to withstand the effects of internal missiles. The design criteria and bases are, therefore. acceptable.
The SWE55/Jt-P1 application is for a single-unit nuclear power station as discussed in Section 't.2 and shown in Figure 1-1 of this report. In the SWE55AR-P1 design, the turbin<r generator will be oriented radially with respect to the containment and annulus buildtrg in a peninsular arrangenent. Safety related systems will not be withf.n the strike zone of potential low trajectory missiles. Based on the informa-tion provided,in SWE55AR P1 for a dul unit station, se note tnat with the peninsular l
errangement and with adjacent anrf parallel turbine buildings. a portion of the con.
tainment of each unit will be within the strike zone of potential low trajectory missiles from some of the low pressure end wheels of the turbine of the adjacent We will evaluate this issue during our review of the construction permit unit.
app 1! cation for a dual unit station by a utility applicant referencing SWESSAR.Pl.
)
since the extent of potential interaction depends on the arrangement of the units on a specific site.
l Safety related systems and structures will be designed to withstand various tornado generated missiles. Stone & Webster provided at our request a spectrum of tornado gen-l ersted missiles against which safety related systems and structures will be designed.
We find these missiles. listed in Table 3-1. acceptable for the design of the SWESSAR-P1 plant.
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l TAatt 3-1 70anA00 ctNtaAfto t 7 tem missitts ran svtssAn.pt l
Tree of Misstie Weieht (th)
Velocity (fos) i I
(1) used plank 4 In. a 12 in s 12 ft 200 420 (2) Steel pipe 3 in, diameter.
78 210 10 ft long, schedule 40 (3) Steel rod 1 In. diameter.
8 310 3 ft long l
(4) Steel pipe 6 in diameter.
285 210 j
15 ft long.
i schedule 40 t
I (5) Steel pipe-12 in, diameter.
743 210 15 ft long.
F schedule 40 (6) Uttifty 13.5 in. diameter.
1.490 210 pole 35 f t long 2
(7) Automoblie 20 f t frontal area 4.000 100 The above missiles are considered as stdking a structure or comument from any direction and at any elevation.
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l Abbreviations used in table:
Ib pound fps feet per second in, inch ft foot ft square feet 3-6 y
y-y-*>
y
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6 3.5.2 Barrier Desian Procedures j
i 4
The analysis of structures. shields and barriers to deteneine the. effects of missile igact will be accoglished in two steps as described in SWEssAA.P1. Section 3.5.4.
l In the first Step the potential damage that could be done by the missile in the
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immediate vicinity of teact is investigated. This is accoglished by estimating the
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depth of penetration of the missile into the impacted structure. Secondary missiles i
i will be prevented by designing the structure thickness well above that determined for l
penetration. In the second step of the analysis the overall structural response of l
the target when impacted by a missile is determined using estabitshed methods of 5
i I
ispective analysis. Missile barriers will be designed to result in ductility factors less than or equal to 10. Technical justifications will be Provided for barrier design cases in which ductility factors greater than 10 will be used. The equivalent loads of missile impact, whether the missile is environmentally generated outside the plant or is accidentally generated within the plant. are co@ined with other app 1t=
cable loads as is discussed in Section 3.8 cf this report.
The procedures that will be utilized to determine the effects and loadings on seismic Category I structures and missile shields and barriers, induced by design basis missiles selected for the plant. are acceptable since these procedures provide a conservative basis for engineering design to assure that the structures or barriers N
are adequately resistant to and will withstand the effect of such forces.
We conclude that the use of these procedures provides reasonable assurance that in the event of design casts missiles striking seismic Category I structures or other missile shields and barriers. the structural integrity of structures, shields, and barriers will not be impaired or degraded to an extent that will result in a loss of required protection. Seismic Category I systems and coeonents protected by these structures are. therefore, adec uately protected against the effects of missiles.
Conformance with these procedures *$ an acceptable basis for satisfying the require.
ments of Criterion 4 of the General Design Criteria.
3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.1 Postulated Pipe eupture inside Containment Stone & Webster has identified in SWESSAR.Pl. Section 3.6. the criteria that will be applied in the $WE$$AR.P1 design for postulating rupture of piping and protection i
against associated dynamic effects. The criteria will apply to all safety related i
systees which are listed in SWCSSAR.P1 Table 3.6 2.
The criteria are censistent with the recomendations of Regulatory Guide 1.46.
For the primary coolant loop specific break types, sizes and locations have been identified in SkESSAR.Pl with reference to the Westinghouse analysis of the systec tased on RE$AR.35 compenent supports. $ tone & Webster has identified in Section 1.8 of SWE$$AR.P1 the responsiDility of Westinghouse and Stone & Webster for the design
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- t and analysis of the reactor Coolant systm 45 listed in Table 3-2 of this report..
Idestinghouse as the nuclear steam supply system vendor has the overall responsibility for the design and analysis of the reactor coolant system. The necessary change of information between the nuclear steam supply system vender and Stone & Webster to confirm total capability of the reactor coolant system will be implemented by 8 -
g utility appilcant in its appilcation for a construction perelt referencing the g
$W$$AR-P1/RESAR 35 oesign combination. ~ 5tene & Webster has identified as an inter-j face requirement. in Ta41e 5.1-2 of $WES$AR-Pl. that the utility applicant sust j
verify the implement 4tlet of the information enchange to assure total compatibility
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of the reactor coolant system.
t Based on the division of design and enslysis responsibility and the interface require.
ment to verify the enchange of the necessary information, we conclude that the proposed design of the reactor coolant system is acceptable for a Pre 11einary Design Approval. We will evaluate the isolementation of the design and analysts responsibil.
ity during our review of a construction permit application by a utility applicant referencing the S ESSAR-P1/RESAR-35 design combinatio.a.
l The proposed piping arrangements and design considerations for high and moderate energy fluid systems inside containment. that are within the scope of SESSAR-Pl.
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will provide adequate assurance that those systems important te safety which are in.
close proxielty to the systems in which postulated pipe fattures are assuend to occur, will be protected. The design will acconnodate the consequences of a pipe break so that the reactor can be safely shut down and meintained in a safe shutdown condition in the event of a postulated failure of a pipe carrying a high or endefate energy fluid.
On the basis of our review we conclude that the criteria and analysical methods that will be used to design and evaluate the piping systems for postulated ruptures and associated dynamic effects will provide an acceptable basis to meet the recutrements of Criteria 1, 2 and 4 of the General Design Criteria and, therefore. are acceptable.
3.6.2 Postulated Pioe Ruoture Outside Contairument The criteria that will be used in the $EssAR-P1 design and layout of high energy j
piping systens outside the containment to determine postulated break types and locations and measures for protection against pipe whip and jet ispingement have been identified in $Wi$5AR P1. Section 3.6.
These criteria are consistent with the I
acceptable criteria identified in our Branch Technical Position MiB 31. " Postulated f
o Break and Leakage Location in Fluid System Piping Outside Containment.' and therefore j
j are acceptable.
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The SWE5SAR P1 design ulli accom odate the effects of postulated pipe breaks and cracks in high energy fluid piping systees outside contairment with respect to pipe ship. jet inpingement and resulting reaction forces. and environmental effects. The means used to protect safety related systems and components will include physical i
l 3-8 r
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DESIGli AfC ANALYSIS RESP 08t$19tL!rY FOR l-
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j acAcron coauwr syss. comers A,e supers i
Design of Piping Stress component J
System Layout Analysis supports I
(1) Reactor coolant prienry loop including W
W W
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a l-reactor pressure vessel twactor coolant
- i pump. hot leg. cold leg. crossover leg.
4 and steam generator.
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4 (2) Reactor vessel support (shield tank)
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$W
$W (3) Pressurizer vessel and pressurizer SW W
W surge line (4) All branch connections to reactor SW SW SW coolant primary loop
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l Abbreviations used in taDie:
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W Westinghouse 9;
SW 5 tone & Webster i
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separation. enclosure in suitably designed structures or compartmen.a. physical pipe enclosures. pipe whip restraints, and equipment shields.
4 The protection to be provided against pipe failure outside containment will be in conformance with the criteria contained in the Commission's ietter of July 12, 1973.
" Protection Against Postulated Events and Accidents outside Containment" and in, i
l conformance with our tranch Technical Position APC58 3 I. *irotection Against Postu-i I
lated Piping Failures in a Fluid System outside Contalement." Stone & Webster will analyse high energy piping systems for the effects of pipe whip, jet impingement, and environmental effects on safety related systems and structures. For anderate e
f energy systems, the jet and environmental effects due to critical cracks will be considered.
l The plant design will. clude the ability to sustain a high energy pipe breat h
accident coincident with a single active failure and retain the capability for safe h
cold shutdown. For postulated pipe failures. the resulting environmental effects will not preclude the habitability of the control room or the accessibility of other I
areas that have to be manned during and following an accident or cause the loss of i
function of electric power suoplies, controls and instrumentation needed to complete I
a safety action.
We have evaluated the criteria for the facility design and the protective design ___
l features to be provided for protection against dynamic effects associated with postulated pipe failure for individual systees. We will review piping layout l
drawings and other pertinent information at the operating license stage of an app 11-cation referencing the SWES$AA-P1 desigm.
i f
Based on our review we conclude that the design criteria, desigi Deses and the ana-lytical eethods for the piping systees outside the containment with respect to the protection against postulated pipe failures outside contalament are in conformance with our high energy line Branch Technical Position APC$8 3-1 and, therefore. are acceptable.
3.7 Seismic Design Criterion 2 of the General Design Criteria requires that structures. systems and
,?
cogonents important to safety shall be designed to withstand the effects of natural i
phenomena such as eartnquakes. We reviewed the $WE15AA P1 structures systems and
)
components important to safety and identified as seismic Category 1 stri.ctures.
systems and components, to determine their ability to withstand the effects of the operating basis earthquake and the safe shutdown eart% Quake.
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]t 3.7.1 seismic fnout The cessess spectra input for the operating basis and safe shutdown earthquate applied la the seismic design of seismic Category I structures, systers and con-ponents comply with the recommendations of Regulatory Guide 1.60. The specific percentages of critical desping values to be used in the seismic analysis of setssic Category I structures. systans and components are in conforwence with Regulatory Guide 1.61.
4-Aj The synthetic time history to be used for the seismic design of seismic Category I 81 aglitude ord frequsney plant structures, systems and cesponents is adjusted 4
content to ottain response spectra that envelop the rcscones spectra rpecified in SE55AA-Pl.
Conformance with the recommendations of Regulatt 4 % $ 1 60 A 1.61 provides j
l reasonable assurance that for an earthquate intenshy of 0.15 e #
'he operating ction 2.5 of l
basis earthquake and of 0.30 g for the safe d.utoc.
'rthW -
4 this report). the seismic inputs to seismic Category 1.*
's and l
components will be adequately defined and mill assure t>',t a ac u t...ciae basis will be used for the design of such strv'ctures. Systems snd Loser 'p.es tr wl hstand the consequent seismic loadings.
3.7.2 Seismic Systsu Analysis i
k The scope of our review of the seismic system and sutsystem analysis for the plant included the seismic analysis nothods for all seismic Category I structures systems and com::enents 45 destritd in SESSAR-Pl. Section 3.7.2.1. It included review of procedures for riodeling, seismic soll-structure interaction developient of floor response spectra, inclusion of torsional effects, evaluation of seismic Category I The review included structure overturning, and determination of composite damping.
the design criteria and procedures for evaluating the interaction of non-seismic Category I structures and piping with seisele Category I structures and piping and The review also included effects of parameter variations on floor response spectra.
criteria and seismic analysis procedures for seismic Category I piping buried outside the contairment.
Model The system and subsystem analyses stil be performed on an elastic basis.
response spectrum nultidegree of freedom and tim history methods will form the bases for the analyses of all major seiselc Category I structures, systems and hhen the sedal response spectrum sethod is used. governing response components, the square root of the sum of the squares rule, parameters will be combined of ff However, the absolute swa of the redal responses will be used for redes with closely The square root of the sum of the squares of the manipun I
spaced frequencies.
codirectional responses will be used in accounting for three comoonents of the Floor earthquake rction for both the time history and response spectrum methods.
3-11 4
i L
- Mwya-a eg..,
e w
,n--
w
- ww a-
%rg w
+
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.c--
)
6 l
me e
-. L n. o oe
..n
..,-e.
.e o. wm.
a
.--,e I
6 spectra inputs to be used for design and test verifications of structures, systems
]
)
and components are to be generated from the time history method, taking into account l
vartetten of parameters by peak widening. A vertical seismic systes dynesic analysis will be employed for all structures systaan and cowonents where analyses show significant structural amplification in the vertical direction. Torsi m ) effects j
and stability against overtuming will be considered.
j 1
)
j With respect to the generation of design envelopes for seismic responses of seismic
]
Category I structures, systems and components, dynamic analyses will be performed
. y j
I for structures founded on subgrades with the following values for the shear j
modulus. G:.
l Gy=
6.000 pounds per square inch t
G2=
24.000 pounds per square inch I
G3=
300.000 pounds per square inch l
G4 = 1.000,000 pounds per square inch f
l For the coatainment structure. it is further assumed that the containment concrete shell may be cracked due to intemal pressure. This condition is comeined with the softest subgrade preperty (i.e. G =6.000poundspersquareinch). Thus,five j
l dynamic models (one for each value of subgrade) will ba used to obtain five different
~
floor response spectra associated with each direction of excitation and eact. 019e I
of damping.
l f
Apprcpriata envelope response data will be generated froe the above described site conditions. Detailed analyses will be performed by the utility applicant referencing the S'."ISSAR.P1 design in its application. Mastrum loads or stresses for all parts of I
l the plant will be deteretned and presented together with the envelope response data.
{
To the extent that the envelope response data envelop and encompass the site. dependent 1
seismic response data. the particular site related design will be accepted.
The finite element approach will be used to evaluate soll structure interaction and l
l structure-to-structure interaction effects. Stone & Webster will perform the finite i
I alement analysis with the computer codes TRIA1 and PLAILY.a. Appropriate nonlinear I
stre:5. strain and daging relationships for the underlying soil will be considered i
in the analysis. Stone & Webster previded, at our request, additional technical bases for the computer codes in SWE55AR.Pl. Ba ed on our review, we conclude that f
the TRIAX and PLAILY.4 codes are acceptable for design use in the $WE$$AR.P1
?
appilcation.
I We conclude that the seismic system and subsystes analysis procedures and Criteria prcposed by the applicant provide an acceptable basis for the seismic design.
1 31T L
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.r s+
me
!-1'9'
i prr - s e -i
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+.
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3.7.3 Seismic festnamentation Proeras 4
The type, mater. location and utilization of strong motion accelerographs to record i
I seismic events and to provide data on the frequency, amplitude and phase relationship of the seismic response of the contaffmant structure as described in Section 3.7.4 of l
SES$AA-p1 will comply with the recommendations of Regulatory Guide 1.12. Supporting j
Instrvmontation will be installed in seismic Category I structures, systans and compo-l nonts in order to provide data for the verification of the seismic responses deter-l.
i mined analytically for such seismic Category I items.
a N
h 1
f The installation of the specified salmelt instrumentation la the reactor containeent j
structure and at other selsmic Category I structures. systems and components will constitute an acceptable program to record data on seismic input of ground action as well as data on the fregency and amplitude relationship of the seismic response o*
major structures and systems. A prompt readout of pertinent data in the control room can be expected to yield sufficient information to guide the operator on a timely basis for the purpose of evaluating the seismic response in the event of an earthquate. Data obtained from such installed seismic instrumentation will te sufficient to determine t
(1) whethe-the seismic analysis assunettons and the analytical model used for the design of the plant are adewate. (2) that allowable stresses were not exceeded and i
(3) whether operation can be enntinued or resumed. The provisions for such seismic
{'
instrumentation is is accordance with our recommendations in Regulatory Guide 1.12 and.
therefore, ace accestchie.
t 3.7.4 5,eismic Intteface Recuf rements Stone & Webster, at our request. has included in SkESSAR Pl. as Table 3.7.6-1. a p'
description of the seismic interf ace rewirements for the $WE51AR-pl/RESAR-3$ design e
combination. This information is usistent with the seismic interface requirements 5
i e
described in the RESAR-35 appiteation and in Section 3.5.1 of our Safety Evaluation e
Report for RESAA-35. We have evaluated this information and conclude that the seismic e
interface requirerents presented for the SWESSAR-Pl/RESAR-35 design in Table 3.7.61 e
of SWESSAA-P1 are acceptable.
e 3.8 Design of Seismic Cate1ory I Structures 3.8.1 Reactor Containment The reactor coolant system will be concletely enclosed in a seismic Category I cen-crete structure. This structure will have cylindrical walls with a hemispherical dome I
and a base met which extends beyond the containment structure and also supports the surrounding annulus building. A welded steel liner will be attached to the interior l
)
of the contalwnt to form a leak tignt barrier. The functiceal performance of the
)
contairwent is discussed in Section 6.2 of this report.
The containment structure will be designed in accordance with appitcable Sussections of the Aserican Society of Mechanical Co1 r.eers Boi!&r and Pressure Yessel Code. Section f
1 i
3-13 l
ha - -
..,.._,,s,...
-~,
w*--m---6 g
-+-P-y wmit.
.-,4 m.
,e-,-
w.
.--e
-y
m
~~ -.
. ~
-.. ~.
(.... (. s.--
g ur 7
~
i T
!!!. Division 2 (the A9E Code). including modifications as requested by the staff to resist various ccabinations of dead loads. live loads environmental loads including j
those due to wind. toenadoes, the operating basis earthquake, the safe shutdown earth-quote ani loads generated by the design bests accident. including pressure, temperature and asseclated pipe rupture effects, i
I.
The design and analysis procedures that will be used for the contatusat are in 1
accordance with procedures delineated in the A9E Code. The containment structure will be designed and proportioned to remain within elastic limits under the various postulated load coeinations. The criteria for allowable stresses and strains are f'
In accordance with those delineated in the A$ME Code. The materials of construction.
'I the quality control procedures and the fabrication and construction requirements also will be in accordance with requiroments delineated in the A9E Code. Testing of materials during construction and of the completed containment prior to operation will be in acco.4ance with the ASME Code, and with Regulatcry Guide 1.18.
1 The criteria that will be used in the analysis, design, and construction of the concrete containment structure to account for anticipated loadings and postulated conditions that may be imposed upot the structure during its service 1tfetime, will l
be in conformance with established criteria, ccdes. standards, guides, and specifica.
l tions acceptable to the staff.
i The use of these criteria as defined by applicable codes, standards, guides.. sad specifications; the loads and loading coeinations; the design and analysis pro-cedures; the structural acceptance criteria; the materials. Quality control prograss, and special construction techniques; and the testing anj inservice surveillance requirements provide reasonable assurance that. in the event of winds tornadoes, earthquakes, and various postulatt.' sccidents occurring within and outside the containment. the structure will withstand the speciffed design conditions without impatisent of structural integrity or safety function. We have concluded that con.
formance with these criteria constitu;es an acceptable basis for satisfying the L
requirements of Criteria 2. 4.16 and 50 of the General Design Criteria.
3.8.2 Concrete and steel internal structures i
The containment interior structures will consist of a shield well around the reactor, secondary shield walls and other interior walls. Cogartments and floors. The principal code used in the design of concrete internal structures will be the ACI 31841 Code of the Arerican Concrete Institute (ACI). For steel internal structures the A15C Specification. *$wecification for the Design. Fabrication and Erection of structural Steel for Buildings.' of the Aferican Institute of Steel Construction (AISC) will be used. For equipment supports. Subsection AF of the ASME Code will be used.
The cortainctnt cc. crete and steel internal structures will be designed to resist vertui,s cortinaticas of cead and live loads accident induced loads, including 3 14
..-w s.w.
,,,,a 2.~
e wy.. -
..m.w, wry
.v--,,-w.m.,-e-
._,,m.,
,..e,, 4 e.
r ex-<
r v 2;v.,.=
m i,;-w. w m w
- s. wwn y m 1.
u u a;.m mu-w b
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j pmssure and jet loads, and seismic loads.. Thuse load costinations are principal i
architectural and engineering design criteria and are discussed in Section 3.8.4 of
.I SWE55AR-P1. The load combinations to be used cover all loads which may act simul-l
?J taneously. The design and analysis procedures thst will be used for the internal t ;
If; structures are in accordance with procedurva delineated in the.*f.! 318 71 Code and f
X in the A15C Specification for concrete and steel structures, respectively. with ij I
modifications requested by the staff.
}
t F.
j 4
f The containment internal structures will be designed and proportioned to rimain-
{;l, within limits established by the staff under the various load con 61natio%. These limits are, in general. based on the ACI 318-71 Code and on the AISC Specification for concrete and steel strvetures, respectively, modified as appropriate for Iced combina-tions thet are considered extreme. The materials of construction. including their l
1 fabrication, construction and installation, will be in accordance with the ACI 318 71 i
(
Code and the AISC Specification for concrete and steel structures, respectively.
t I-I We conclude that the c.*iteria to be used in the design, analysis, and construction of the containment internal structures to account for anticipated loadings and postulated conditions that any be imposed upon the structures during their service lifetime are I_
j h: conformance with estabitshed criteria and with codes. Standards, and specifica.
tions, and, therefore. are acceptable.
)
The use of these criteria as defined by applicable codes, standards, and specifica-tions; the loads and loading conbinattunst the design and arilysis procedures & tM structura1 acceptance criteria; the materials. quality control programs, and special construction techniquest and the testing and inservice surveillance reqdirements pro.
vide reasonable assurance that. in the event of earthquakes and various postulated
[
accidents occurring within the containment the interior structures will withstand tne specified design conditions without inesirment of structural trategrity or the per.
formance of required safety functions. We have concluded that conformance with tnett criteria constitutes an acceptele basis for satisfying the requirements of Criteria 2 and 4 of the General Design Criteria.
3.8.3 Other Seismic Cateoory 1 Structures
~
All seismic Category I structure. other than the containment and its interior struc.
tures will be of structural st.el and concrete as discussed in SWESSAR pl. Section 3.8.4.
The principal code is be used for the design of cor. crete seismic Category I structures is the ACI 316 71 Code, and for steel seismic Category I structures the A15C
~
l Specification (see Section 3.8.2 cf this report).
The concrete and steel seismic Category I structures will be designed to resist various combinations, as appitcable, d dead loads itwe loads, environ =ntal loads including winds, torriadoes, the operating basis earthquake, the safe shutdown earthquake, and loads goersted by postulated ruptures of high energy pipes such as reaction and jet irpingenent forces. corpartrierit pressures, and impact ef fects of whipping pipes.
3-15 I'
.m.--.
~..
A 1
F
-4g J
hh 9
3 The design and analysir procedures that will be used for dese seismic Category I Q
structures are in accordance with procedures delineated in the ACI 318-71 Code and in the A!$C Specification for concrete and steel structures, respectively.
j t
I The various seismic Category I structures will be designed and proportioned to reasin within limits established by the staff under the various load combinations. These k
M limits are in general. beseef on the ACI 318 71 Code and on the AISC Specification for -
1 concrete and steel structures, respectively, modified as appropriate for load coebina-40 tions that are considered estreme. The materials of construction, including their fabrication, construction and installation, will M in accordance with the ACI 31841 l
ff Code and the A!$C Specification for concrete and steel structures, respectively.
l l*
i The use of these criteria as defined by appitcable codes, standards and specifica.
!2 tions, the loads and loading continations, the design and analysis procedures, the
'j structural acceptance criteria, and the naterials quality control and specdal construc-tion techniques provide reasonable assurance that in the event of winds. tornadoes.
earthquakes and various postulated accidents occurring within the structures, these structures will withstand the specified design conditions without irpairment of struc.
i j
tural integrity or the performance of required safety functions. We hava concluded that conforiaance with these criteria, codes, specifications, aid standards in the 7
desten of seismic Ca(egory I structures other than the containmer.t structure constitutes an acceptable basis for satisfying the requirements of Criteria 2 and 4 of the 'iereral 2
Design Criteria.
n 3.8.4 foundations l
The foundations of seismic Category I structures for the fESSAR-PI design will be reinforced concrete. primarily of the mat type design. The principal code for the design of these concrete mat foundations will be the ACI 31s-71 Code. These concrete foundations will be designed to resist various continetters of dead loads live 1
i u
Ioads, enviroivnental loads including winds, tornadaes. the operating basis earthquake.
p l
the safe shutdown earthquake, and loads generated by postulated ruptures of high
)
energy pipes as discussed in Section 3.8.5 of SWE55AR-Pl.
B The design and analysis procedures that will be used for these,elsnic Categori I founda-j tions are in accordance with procedures delineated in the ACI 318-71 Code. The warfous j
seismic Category I foundations will
- e designed and proportioned to remain within 15tts established by the staf f under the various laad combinations. Thest limits are, i
in general, based on the ACI 318-71 Code rodified as wropriate for load canbinations
]
r
{
that are considered entreme. The materials of construction and their fabricatioq.
construction and installation, will be in accordance with the ACI 318 71 Code.
The criteria used in the analysis, design, and c nstruction of 511 the plant seismic Category I foundations to accouct for anticipated ioadings and postulated ennditions that may be imposed upon each foundation during its service lifetime are in confor-mance with establishes criteria, codes, standards, and specifications, and therefore, are acceptable to the staff.
3-16 1
1
_.m. _..
J 1
,,a j
t
. -,.. ~,
-,g-
.3t W
.j p.$
t I
r;4 gv w
l The use of these criteria as defined by r.pplicable codes, standards, and specifica-C tions, the loads and toeding combinatir.as. the design and analysts procedures, the j
structurs) acceptance criteria. and t'e materials quality control and special con-j struction techniques, provide ressa.dble assurance that in tan event of winds, torna.
does. earthquakes, and various pratulated events, the seismic Category 1 foundations f
1 will be able to withstand tha, specified design cormalons without twiruent of struc-l tural intecetty :.3 stability or the performancs d required safety functions. We p
conclude that conforsence with these criteria, codes, specifications. and standards const'tutes an acceptable bests for satisfying the requirements of Critaria 2 and d ef l
'the General Design Criteria.
.j h
3.8.5 Structural Interface Requirements
]
W Stone & Hobster, at our request. has included in SWES$AR.P1. as Table 3.8.6-1. a description of the structural interface requirements for the SWE55AR.P1/PE5AR 35 design cos61 nation. This information is consistent with the structural interface i
requirements described in the RE5AR 35 appilcation which we found acceptable in 7
Section 3.6.3.6 of our Safety [ valuation Report fer RESAR 35. We have evaluated 4
this information and conclude that the structural interface requirements presented l
['
for the SWE55AR-P1/RESAR-35 design in Table 3.8.61 of SWE55AR P1 are acceptable.
l 6
)A 3.9 Mechenical Systems and Coseonents l
3.9.1 Dynamic System Analysis and Testino 4
vhe definitica. of the preoperational vibration test program for ASMC Code Class 1. 2
[
and 3 pig.ing systems and the organization of the staff to conduct the program are
$f not within the scope of SWE55AR Pl. but are the responsibility of a utility applicant referencing the SWE55AR P1 design as stated in Section 3.9 and 14.0 of SWE55AR-Pl.
j
?
5 tone & Webster has provided in Section 3.9 of SWESSAR.P1 the criteria for the pr,o-
{
A preoperational vibration test progree will be conducted under simulated y
gras.
transients that are considered to be credible within the normal and upset operating L'
modes of the systees to check the performance of piping systems important to safety.
h Corrective actions will be taken if excessive vibrations are observed. The pre.
[
operational test program will be conducted during startup and inital operation on
~
all safety related wiping systems, restraints, components. and component supports classioed as A54 Code Class 1. 2. and 3.
The tests will provide adequate assurance
[.
that the piping and piping restraints of the system have been designed to withstand vibrational dynamic effects due to valve closares, pro trips. and other operating modes associated with the design basis operaticeal transients. Tr.e planned tests will develop loads similar to those espected to be emperienced during reactor operation.
Core 11ance with such a test program wt:1 constitute an acceptable basis 'or fulfillbg the requirements of criterion 15 of the General Design Criteria.
I 3-17
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. ~....
,. ~., - -,
g
s.
5
{
We find the criteria provided in SWE11AA-pl to be reasenchle and consistent with j
y those used in actual practice for vibretton test programs. The responsibility of 1
the wtility applicant for the development of the actual program as identified in p
$kES$AA p1 in an interface requirement to be met by the utility applicant la its f
constrwction pemit applicattor.. Our conclusion; on the accepta4111ty of the actual j
program will be based on our review of the program to be provided in the utility appifcant's construction posit application.
[1 i'-p Stone & Webster has identitled criteria for the selssic qualification of selssic S
, Category I safety related mechanical equipment and has described the principal y
methods that will be isolamented to verify tile equipment operability during and g
after postulated seismic occurrences up to and including the safe shutdoun eerthquake.
i The details of tre qualification progree will be described in the safety analysis E
report of the utility applicent referencing $WE15AA.pl. This is an interface matter
~
to be addressed by the utilisy app 1tcant in the construction pemit application.
V t
We conclude that a selssic quellfication progree based on the criteria specified in SWES$AA-pl util provide adequate assutaara that the seisatc Category I mechanical j
equipment will function properly under the vibratory fones and actions japosed by
- i the operettng basis and the safe shutdown earthquake and, therefore, will constitute y
an acceptable basis for satisfying the applicable requirements of Criteria 2 and 14 of the General Design Criteria.
F i
Stone & Webster will perform a dynamic analysts on those pirtions of the reactor
{
coo! ant system that are within the scope of SWE$1AA-pl (see Table 3-2 of this report) to confim the design adequacy for the corditions of a combined loss-of-coolant I
accident and safe shutdown earthquake. The analytical methods for the dynamic
/
analysis have been presented in 14Ws;Jt-pt. Stone & Weester will generate floor response spectra using the apprortate Site related characteristics of the safe
{
shutdown earthquake. These sped ra utli be presided to the nuclear steam supply
{
systen vendor for its verification that its interface requirewnts have been met.
The detailed dynamic model. Including the mathematical modele forcing Nctiot.s and procedures, and a summary of representative results will be provided by the utility rh, appitcant referencing SWES$AA.pl in its operating license application.
k I
The resronsibility for the analytte.il justification of the primary coolant loop E i j
f under postulated occurrences of the Caettned loadings from a lossacf-Coolant accident and safe shutdown earthquake has been identified in $wtssAA.pl and RESAR-35 (see b '
)
also Section 3.6.1 of this report). Assurance of the reactor coolant system integrity' d
under these faulted condittos.s will require considerable input from and exchange of fnfermation between Stone & Webster and hestinghouse, who will saintain overall I
l design responsibility for the primary coolant loop and will perfore the reactor Inteesals analysit under faulted conettions. Stone & Weoster has identified as an Interface requirement la Table 5.1-2 of Sht15AA-Pl. that a utility appitcant 3 18 i
L
.w d
7 gy e
w-we
.-yes-p---.m n-
-ia.-,
c.
g.-
sy- -,
y-g-e--y
-e
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y ywc--..-
-,y
,-yy, y-,, - -
,%,7
- - - - -, --wrr-
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.... - ~,
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referencing the SE55AR-P1/RESAR 35 design combination most verify the implementa-e tion of the informatica eschenge to assure total compatibility between the SWE55AA-P1 e
and RESAR 35 designs. We will evaluate the detailed dynamic model and the represents.
e tive results during our rvview of an application for an operating license by a i
utility applicant.
The dynamic system analysis will provide an acceptable 6451s to confirm the structural I
design adequacy of the unbroken piping
- oops to withstand the cousined dynamic loads
(
wstulated loss-of-coolant accidents and the safe shutdonsi earthquake. The load from this dynamic analysis will be combined with other couponent loads to provide assurance that the conLined stresses and strains in the components of the reactor e
coolant system will not exceed the allowable design stress and strain Ilmits for the o
materials of construction, and through exchange of load inforination with Westinghouse.
that the resulting deflections or displacements of any *tructural components of the reactor internals assembly will not distort the reactor internals geometry to the extent that core cooling may be impaired. The methods used for component analysis e
have been found to be compatible those used for the system analysis. The prvposed combinations of component and system analyses are, therefore, acceptable. The assur-ance of structural integrity of the reactor coolant system under loss-of-coo 1&nt ac-cident conditions for the most adverse postulated loading event provides added con.
fidence that the design will withstead a spectrum of lesser pipe breaks and seismic loading events. Accomplishment of the dynamic system analysis constitutes an accept-able basis for satisfying the applicable requirements of Criteria 2 and 4 of the General Design Criteria.
3.g.2 ASME Code Class 1. 2. and 3 Components and Conconent 5corts Safety related ASME Code Class 1. 2. and 3 comonents, suprorts and systems, includ-ing piping systems, within the scope of Sh;55AR-P1. will be designed to provide assurance that the structural integrity of these components Supports, and systems will be maintained under normal, upset emergency and faulted plant conditions.
These conditions are consistent with those outlined in Replatory Guide 1.48.
Design bests loading combinations will conservatively envelop anticipated transient conditions provided by Westirghouse end will apprors'ately consider the operating basis earthquake and safe shutdown earthquake thrrey analyses that are commensurate with the stress limits applied to each type of cLgonent. support., and piping system.
5 tone & Webster, at our request, provided additional clarification and oefinition in
~
l SWE55AR-P1 with respect to the design and analysis responsibilities between Stone &
Webster and Westinghouse for the cogonent1 of the priedry reactor coolant systee and their supports as Itsted in Table 3-2.
The supports for the primary coolant system are within the scope of Rt5AH 35 except for the supports of branch lines connecting to the primary coolant system. The reactor vessel support assertily.
designed by Westinghouse. will be supported by the reactor vrssel shield tank, which will be designed by Stone & Webster as indicated in Table 3-2 of this report. As
]
stated in Section 1.8 of 5WE55AR-Pl. Stone & Webster has tne responsibility for 3-19
.se %y, h
4 h
ge coordinating all activities necessary to assure overall design cometibility of the c
$WES$AR-pl/RESAR 3$ design coelnation in an appilcation by a utility appilcant e
referencing the design combination.
e The specified design basis for combinations of loadings as applied to safe stated ASE Code Class 1. 2 and 3 pressure-retaining cowonects in systems and their supports designed to meet seismic Category 1 requirements provides reasonable assurance that in the event of an earthquake or an upset, amargency, or faulted plant transient occurt Ing during normal plant operation, the resulting costined stresses imposed on tystem components and supports will not exceed allowable stress and strain ilmits for the materf'Is of construction. Limiting the stresses under such loading coelnattoms will provik a cor.servative basis for the design of system components and supports to l
withstand the most adverse combination of loading events without loss of structrral Integrity. We have concluded that the design load combinatio#M and associated stress and deformation limits specified for A$NE Code Class 1. 2 and J components and supprets constitute an acceptable basis for the design to meet the applicable require-monts of Criteria 1. 2 and 4 of the General Design Criteria and are acceptable.
The operabil.ty qualification program will demonstrate operability of A$ME Code Class 1. 2 and 3 active pumps and valves under strulated leads representing a maxi-mise con 61 nation of operating, safe shutdown earthquake, and dynamic system loads.
The criteria for quellfying the different toes of pumps and valves entall various analytical and test methods that can be suitably matched to the structural, func-tional and installation characteristics of each particular piece of equipment.
The operability quellffcation program for active punos and valves is described in i
Section 3.9.2.4 of $WE$$AA-Pl. The qualification method essentially will consist of tests and analyses. For determining the response during the safe shutdown earth-l quake, all motors and electrical components will be tested and mechanical coseonents will be tested or analyzed. The criterta used to determine whether a cowonent will be tested or analyzed statically or dynamica11y have been provided. Testing is the principal method of qualification. $tatic analysis will be performed when the com-ponent can be described by a singic mass cnd spring. Dynamic analysis will be
)
performed when the component is sere complex. Additionally, programs will be imple.
mented as applicable based upon the recommendations of the knerican National $tandards Institute AM5! N 45 comittee standards under development.
Systems with their appropriate class designation have been identified in $VES$AR-Pl.
The detailed listing of ASME Code Class 1. 2. and 3 active pugs and valves and the specific quellf fcation methods to be emoloyed will be provided in the app 1tcation f
for an operating Ilcense by a utt11ty appilcant referencing $WE$$AR pl.
On the basis of our review we conclude trat the prope'.ed operability qualification program meets the requirements of Criteria 1, 2 and 3 of the General Design Criteria as related to the operability of seismic Category I active pues and valves end.
3-20
.... -.. ~... _..
v
-c-
r. _ _.
~~
yeq i
l I
I l
T I,
I.
4 the efore, is acceptable for a Preliminary Design Approval. We will require a
'l vtil'ty appitcant referencing the SWES$Mt-P1/ItESAR-35 design coebination to provide in its appUtation for an operating Ilcanse a detailed listing of ASME Class 1, 2 and 3 active pwes and valves and to identify the specific qualtffcation methods to be employed.
The criteria to be used for the design analysis and installation of A$ pee Code Class
- 1. 2, and 3 safety and relief valves are consistent with the recommendations of Regulatory Guide 1.67 for open discharge systseat for closed systeme Stone & Webster has provided an outline for a conservative #namic analysis of the complete system-The calculational procedures and methods to be used in the analysis include the development of transient hydraulic forcing functions and their application to de-termine the dynamic responses of the system. T1e time history analysis will account for valve opening time.nd water slug effects where loop seals are part of the system.
Using these criteria in developing the design and mounting of ASME Code Class 1. 2 l'
and 3 ssfety and relief valves will provide adequate assurance that, under discharg.
ing conditions, the resulting stresses will not exceed the allowable design stress and strain ilmits for tre materials of construction. Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices will provide a conservative basis for the design of the systes components to with-stand these loads without loss of structural integrity and impairment of the over-pressure protection function.
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We conclude that the application of these criteria in the design, analysis and installation of ASME Code Class 1. 2 and 3 overpressure relief devices constitutes an acceptable basis for meeting the applicable requirements of Criteria 1, 2, 4,14 and 15 of the General Design Criteria.
Seismie Qualification of setsric Cateery 1 Instrumentation and Electrical Equipment 3.10 Instrumentation and electrical components required to perform a safety fur < tion will Stone & Webster has pro-be designed to meet the seismic Category I design criteria.
posed a seismic qualification program that will be imp 1*eented for seismic Category I instrumentation and electrical equipment and the associated supports for this equip-ment to provide assurance that such equipment can be expected to function properly and that structural integrity of the supports will not be tspelred during and after the excitation and vibratory forces fsposed by the safe shutdown earthquake.
The seismic qualification program provides for appropriate analysis and/or testing cualification for all off the shelf and designed equipment classified as seismic
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The seismic qualtitcation of Category I electrical equipment wit! be in l
Category 1.
accordance with IEEE Std 3441975,
- Seismic Qualification of Class ! Electrical Equipment for Nuclear power Generating Stations
- of the Institute of Electrical and 3
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The seismic qualtffcation testing program for seismic Category I instrumentation and electrical equipment provides adequate assurance that such equipment allt function properly during the escitation from vibratory fcrees imposed by the cafe shutdom earthquake and under the ccNitions of post. accident operation. We have com,1uded that this program constitutes an acceptable basis for satisfying the appilcable requirements of Criterion 2 of the General Desip Criteria.
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i The reactor, as part of the nuclear steen supply system, is not within the scope of i
the $E$$AR-71 standard balance-of-plant design. $WE$$AR P1 includes by reference j
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i the niestinghouse standard nucleer steam supply system design. RC$AR 35. Our evalue.
i tion of the RCSAR-35 design is discussed in our Safety Evaluation Report for RESAR-31.
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3 5.0 REACTOR COOLANT $Y$TDI AND CONNfCTED $Y$ TEM 5.1 General Inr rustion o
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The reactor coolant system in its entirety is designed and analyzed by Westinghouse l
as described in RE$AR 35 and is not within the scope of SWES$AR Pl. Stone 8 Webster f
has provided supplemental information for the systen in Chapter 5.0 of $WES$AR-Pl.
g As stated in section 3.6.1 of this report. Stone & Webster has identified in E
SWESMR-P1 the respoesibilities of Westinghouse and Stone & Webster with regard to the design, analysis and layout of the primary cooleet system. The verification of the necessary exchange of information between these organtrations is the responsl.
bility of the utility applicant and has been identified as an interface' requirement I
in Section 5.1 of SWES$AR-P1. The criteria for material selection and construction of systems that are within the scope of $WESSAR-P1 and that 6re related or connected to the reactor coolant systes pressare boundary are identified in $WES$AR-Pl.
We are presently considering on a generic bests the question of whether capability should be provided for transferring heat from the reactor to the environment from a
i nomal reactor operating conditions to cold shutdown using only safety-grade systems, with only offsite or onsite power available, and ass'. ming the most limiting single a
failure. If we determine that this capablitty should be provided, we util require that the RE$AR-35 design and the designs of the balance-of plant portions of e
$WES$AR-P1/RESAR.35 be modified accordingly. We have determined that such modifics-I e
tions are technically feasible and conclude that this matter can be lef t for post-preliminary design approval stage consideration.
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5.2 Integrity of Reactor Coolant Pressure Boundary 5.2.1 General Material Considerations The materials for the reactor coolant pressure boundary are specified by Westinghouse in Rf5AR-35 and are not within the scope of the SWESSAR-P1 design.
e l hl The materials for systems connected to the reactor coolant pressure boundary and for f
related auxiliary systems are identified in SWE$$AR Pl. They will be procured in accordance with the requirements of the Arnerican Society of Mechanical Engineers (ASME) Botier and Pressure vessel Code. Section 111 (the A'J12 Code). The contrs:s to i
be imposed upon components to be constructed of austenttic stainless steel used in the reactor coolant pressure boundary conform to the recomendations of Regulatory l
Guide 1.31. The testing of qualification welds to demonstrate nonsensitization will be in accordar.ce with the recomendations of Regulatory Guide 1.44 Components of
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hardened stainless steel. or hardenable mortensitic stateless steel meterials having j
i a yleid streng*.h greater than 90.000 pounds per square inch.
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The fabrication and other processing requirements.for enterials in the auxiliary systems will be compatible with the thersel insulation and will be in conformance i
with the recomendations of Regulatory Guide 1.36. The procedures to be used to assure that the components will be suitably cleened and protected against contam.
inants capable of causing stress corrosion cracking conform to the recommendations of Regulatory Guides 1.37.1.38 and 1.39.
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The material selection and construction of the systems within the $WE55AR P1 scope and connected to the reactor coolant systes pressure boundary satisfy the interface i
requirements of the RESAR 35 design. The controls to be imposed on the fabrication and construction of components in the related auxiliary and connected systems to the i
j reactor coolant pressure boundary constitute acceptable bases'for meeting the re-quirements of Criteria 1 and 14 of the General Design Criteria. The design of SWE55AR-Pl. and in particular the design of the reactor vessel shield tank 45 the reactor vessel support system. Its cooling system. insulation and surrounding struc-tures. will not limit any in place annealing of the reactor vessel that may be needed, i
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i 5.2.2 Reactor Coolant Pressure Boundary testage Detection System i
Reactor coolant leakage within the containment may be an indication of a small I
through-wall flaw in the reactor coolant pressure boundary. The leakage detection system proposed will include diverse leak detection methods. will have sufficient sensitivity to measure small leaks will fdentify the leakage source to the extent practical, and will be provided with suitable control roce alarms and readouts. The system will consist of the containment radiation monitors (gas and particulate). sump i
level measuring system. and the recirculation fan cooler system. Indirect indication
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of leakage will be obtained from the contatteent humidity, pressure and temperature l
monitors. Intersystem leakage will be detected by abnormal readings from radioactivity monttors in the secondary system. The leakage. detection systems proposed to detect leakage from components and piping of the reactor coolant pressure boundary are in accordance with the reccuseendations of Regulatory Guide 1.45 and provide reasonable assurance that any structural degradation resulting in leakage during service will be detected in time to permit corrective actions. Conformance with the reconmendations of Regulatory Guide 1.45 constitutes an acceptable basis for satisfying the require-j eents of Criterion 30 of the General Design Criteria.
5.2.3 Inservice laspection Program To ensure no deleterious defects develop during service. Selected welds ar1 weld heat.af fected zones will be inspected periodically. The reactor pressure va sel.
System piping, pumps. valves. and components =hich require inservice inspecbon, as defined by the ASME Code.Section II. will be designed, fabricated, and erected in 52
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accordance with the requirements of Section II and its addenda in effect on the date of docketing for a construction permit appittation by a utility applicant as required I
by 10 CFR Port 50. paragraph 50.55a(g). Remote inspection equipment will be used to inspect those areas not readily accessible to inspection personnel. Details of the Inservice inspection program and equipment are outside the scope of SWESSAR pl and will be submitted in the safety analysis report of the utility applicant referencing the SWESSAR-P1 design.
The conduct of periodic inspections and hydrostatic testing of pressure retaining camponents in the reacter coolant pressure boundary, in accordance with the require-monts of the ASME Code.Section II, provides reasonable assurance that evidence of structural degradation or loss of integrity occurring during service will be detected in time to permit corrective action before the safety function of a component is compromised. Compliance with tie inservice inspections required by this Code consti-tutes an acceptable basis for satisfying the requirements of Criterion 32 of the f
General Design Criteria.
5.3 Contairment Sullding Polar Crane The RESAR 35 design uses the containment polar crane during refueling and mal..**aance The reactor vessel head. the upper internals and the control rods will operations.
a be lif ted individually by the polar crane which is within the scope of the SWE55AR P1 design.
Durirq w review of RESAR-35. we requested Westinghouse to provide analyses to i
dnenstrate the acceptability o' the conseqt,ences that might result from dropping the eacctor vessel head from the spectfled maximum lift height. At the time we completed our RESAR 35 review we concluded that additional information was needed to confirm the conclusions obtained from the Westinghouse analyses. The following paragrsph is a
taken from section 9.1.2 of our Safety Evaluation Report for R(SAR 35:
- Westinghouse has advised us that they are ana yz ng this accident.
l i However, until we are able to determine that the consequences of this accident are acceptable, we have imposed an interface req;irement trat e
applicants referencing RESAR 35 provide an overbecd reactor vessel head assembly handling system that is designed so that the connected load would not rail in the event of a single failure or malfunction.
This hand 11pg system, or single failure proof crane shall be designed as a safety system and shall be designed, fabricated. Installed.
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inspected. tested and operated in accordance with the Auxiliary and l
Power Conversion System Branch Technical Position APC58 9-1. *0ver-f head Handling Systems for Nuclear Power Plants.* which is contained in f
Standard Review Plan 9.1.4 If Westingnouse provides additional information which demonstrates acceptable consequences from the postu-1ated dropping of the reactor vessel head while it is enroute to the l
head laydown area. a single failure proof crane will not be required."
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1 cation. Accordingly, we have defined it as an interface setter that mast be f
addressei by a utitur g;1' cant referencing the SWi$$ta-P1/RESAR-35 design cosMna-tion in its construction r att applicatten.
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.1 6.0 ENGINEERED $AFETY FEATUR'S i
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The purpose of the various engineered safety features is to provide protection for li the plant personnel and the public by limiting the radiation exposure that could result frem a major accident in the plant. In this section, we discuss the engineered safety features systems' proposed for the SWES$AR-P1 balance.cf plant design. Certain of these systems or parts of these systems will have functions for
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normal plant operation as well as serving as engineered safety features.
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We have reviewed the proposed systems and components designated as engineered safety.
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I features. These systems and components will be designed to be capable of assuring safe shutdown of the reactor under the adverse cuiditions of the various postulated design basis accidents. They stil be designed, therefore, to seismic Category 1 requirements and must function with complete loss of offsite power.
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i Components and systems well be provided in sufficient redundancy so that a single failure of any component or system will not result in the loss of the capability to achieve safe shutdown of the reactor. These design requirements are in accordance j
i with the General Design Criteria.
h 6.2 Containwnt Systems The containment systems for the SWLS$AR.P1 plant will include a reactor contairvient structure 45 the primar-containment, a secondary containment (partial or full secondary containment). a supplementary leak collection and release system, contain.
ment heat removal systems, a containment air cleanup system, a contairment isolation system a containment combustible gas control system. and provisions for contairsient L
leakage testing.
6.2.1 Contain= nt runctsonal Desian a
The containment will be a steel. lined reinforced concrete structure with a net free volume of about 3.060.000 cubic feet. The contairment structure will house the
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nuclear steam supply system, which includes the reactor vessel, reactor coolant piping, reactor coolant pumps, prtssuriter, and steam generators, as well as certain The containment will be components of the plant engineered safety features systems.
designed for an internal pressure of 48.0 pounds per square inch. gauge and a tever.
ature of 280 degrees Fahrenheit.
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Stone & Wat> ster has described the analytical model for the SWESSAR-PT containment k;
pressure analysis, including the assumptions made regarding the availability of heat removal systas and structural heat sinks. For the SWIS$AR-pl/RESAR-35 design com-t binatica Stone & Webster has analyred reactor coolant system pipe break accidents for
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a spectrue of break locations and sizes. The Stone & Webster LOCTIC computer code j
was used for the cretainment pressure response analysis. The postulated dooble-ended l
t rupture at the pump suction in the cold leg of the reactor coolant system resulted in h
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the highest calculated containment pressure of 3g.8 pounds per square inch-gauge.
g Stone &Webstercalculatedthemaximumpressurebasedonaninitial(pre-accident)
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'cmetainment pressure of 16.2 pounds per square inch-absolute. The contairment l
I i desty pressure (48.0 poeds per square inch-gauge) provides a margin of 20.6 percent l l absve the maxisua calculated pressure (39.8 pounds per square inch gauge), which is
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in excess of the minimum 10 percent mergin which we require to be available in a
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The pressure was calculated using the mass and preliminary safety analysis report.
j energy release rates provided in RESAR 35 for the postulated break which are based on As a Westinghouse assumed contairement pressure of 47.0 pounds per square inch-gauge.
stated in our Safety Evnluation Report for RESAR-35. these mass and energy release E,
rates will be acceptable if the calculated containment pressure is less than the
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assumed pressure of a7.0 pomds per scuare inch-gauge. We, therefore, found the mass and energy release rato provided in RESAR 35 to be acceptable for calculating the I
$WES$AR-Pl/ RESAR-3$ saximum containment pressure due to a postulated rupture of the i
reactor coolant systan.
We have performed a confirmatory analysis using the COMTEMPT 24 code, the mass and F.
enerv;y release rate data given in RESAR 35, and data provided by Stone & Webster for s[
Based on our contairement structural heat sinks and heat removal system performance.
V confit natory analysis, we conclude that the Stone & Webster analysis to determine the i
maximum containment pressure due to a postulated reactor coolant system rupture is
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,f based on conservative assusgtions and, thttef-3re is acceptable.
For the postulated main steam line break accident Stone & Webster identified the
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double-ended rupture of a main steam line occurring at 102: of full power, with one j
of two fan cooler trains inoperable, as the accident resulting in the highest contain.
ment temperature (362*F) and a limited displacement rupture of a main steam line at h
30% ? full po er, with one of two spray trains inoperable, as the accident resulting in the highest containment pressure (38.5 psig). We have performed a confirmatory analysis using the CONTEMPT-26 code and have calculated corresponding pressures and temperatures that are in good agreement with the Stone & Webster results.
In our Report to ACRS for SWIS$AR Pl/R~$AR-35, we indicated that Stone & Webster had analyzed a spectrm of RJn steam line breaks and postulated various single active
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Our failures which could occur coincident with a suit steam line break accident.
review of the postulated single f ailures indicated that stone & Webster had not adequately justified that the fatture of a diesel generator to start would not repre.
Sent the limiting case. On Aly 9.1976, a reeting was held with the MC staff and 6-2
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1 reprtaantatives of Stone & Webster to discuss this metter. As documented in a meeting l
simmerf dated July 15.1976 (Itas 7). the NRC staff agreed that for the issuance of a l
Prelle nary Design Approval we would accr$t a qualitative disc.ssion by Stone t 0
h Webster as to why the failure of the diesel generator would not be the ifalting single failure for the main steam Ifne break accideet. This rationale was to be h
based o't information supplied to Stone & Webster by Ostinghouse. At tiat time w
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Indicated that our final acceptance of tHs justification sould be dapecent upon our review of a Westinghouse topical reoort describing mass and energ/ releases for main
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staan line breeks to be sutaitted to the NRC. We indicated that we sculd require vs.
p this analysis to be provided in a construction permit application submitted by a h
utility applicant. The rationale for justifying the failure of the diesel generator k.
k-as not being the most limiting single failure, as wii a* sther information requ*sted from Stone & Webster in our July 15,1976 =*eting s'amary was receiv9 in Amendment 28 to the SWE55AR-p1 application.
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In our discussion *, with Stone & Webster, we identifled certain instrumentation changes that have been made by Westirghouse which enerate an Solatisn.lgnal ti the main 0
steam and feedwater isolation valves 6.0 seconds arter.* Ilmited displacement rup.
ture, i.e.. Ca.fgn basis main steam line break resulting in peak conta$ ment pressure, f,*
This signal is generated by the Westinghouse steam line break instrumentation, i.e..
T 1ead lag compensated low steam ifne pressure signal. In a July 21. 1976 telecon with 4
Westinghouse, we requested tnat Westinghouse prevfde on the RESAR 35 application an k
interface requirement that Westinghoust.trovide to the bala9ce-of-plant designer the
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characterization of the lead. lag circuit that will ensare that the compensated low E
l steam Itne pressure signal will in!tiate the engineered santy features in the time
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frame established fee the accident a9alysis. Westinghouse amplied with this request and in Amendment 12 to RESAR 35 Taile 10.11 provided a cormitment to supply this fnformatica to a utility applicant. In our review of a construction permit application by a utility applic.nt for the 5WE55AR.P1/RESAR 35 design combination, we will require verification that this infor1 nation i.es been peevided and f; documented. Stone &
f Webster has incorporated this comitment into Table 10.12 of the SWES3AR-P1 applica-A tion in Ameneent 211. Wa will pursue our review of the lead. lag compensated low
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steam line pressure signal, and its justification, during our review of a construction permit applicaticn referencing the $WE55AR-Pl/RESAR 35 design combination.
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The mass and perfy release rates to the contaltunent for the main steam lire break
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accident are calculated using the MARYEL code which.s under review by the NRC staff 4
and TRANFLO which will be submitted by Westinghouse for NRC staff review. The applicant has agreed to incorporate the results of this generic review in the analysis F
for the $WE55AR.P1/RESAR.35 design combination containment. We find this comitment to be acceptable.
On this basis, we are unable to conclude that the main steam Ifne break analysis for the SWE55AR-pl/RE5AR 35 design combination is acceptable. Consequently as an fnterface mattee far the SWE55AR-Pl/RESAR.35 design combination. we will require 6-3
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y a utility applicant in a constructics. permit application combinathe to provide the
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following:
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(1) Analytical justification to demonstrate that the failure of the diesel generator is not the limiting single failure for the main steam IIne break analysis.
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(2) Analyses of a coeplete sGJctrue of main steen line breaks using mass and energy fj release data based upon models dich have been approved by the SC staff.
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(3) Verification that Westinghouse has provided to the utt11ty applicant (balance.
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of-plant designer) the characterization of the lead-lag circuit that will ensure g
that the compensated low steam line pressure signal will initiate the engineered
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safety features in the time frame established for the accident analysis, f
E' Stone 1 Webster has analyted the pressure response of compartments inside the contain-h, ment to postulated high energy line breaks identified in RESAR-35. The Compartments
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investigated are the steam generator compartner t th pressurizer compartment and the
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reactor cavity. The mass and energy release rate data frae RISAR 35 were used in the p
analysis. The data were input to the Stone & Webster THREED code and the RELAP-4
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code to calculate the compartment pressure responses. The THRIED code accounts for the flow of a steam-water air mixture through the vents, but does not consider the Q
inertial effects which may be significant when the vent flow is subsonic. The RELAP-D 4 code considers inertial effects but does not include the effects of air mining wit O
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the flowing steam-wter misture. For cases where the THREED code predicted subsonic 1
flow during most of the transient. Stone & Webster also used the RILAP-4 code to 4
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investigate the importance of inertial effects on the pressure response of the i
compartment. The results from the code predicting the higher differential pressures
,,j' were used in the design of the corpartment walls.
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For each compartment. Stone & Webster calculated a uniform pressure response based on Ij; l
a single node model, and a pressure profile based on a multi. node model. The calcu-
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f lated uniform I assure or asyvmetric loading resulting from the pressare profile i
lU rotated to an.ngle in the plan view, whichever is more limiting, uns used to h
establish the design differential pressure across the compartment wils. The pres-f sure profile will be used to determine the design differential pressures for the j
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component supports in accordance with tre design and analysis responsibility discussed 1
in Section 3.6.1 of this report. Stone & Webster has specified a 40 percent riargin o
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for the component and compartment design differential pressures for use in the struc.
tural loadirg equations, which we find acceptable, ie For the steam generator coccartment pressure analysis, a single-ended rupture in the i
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hot leg and in the pwp discharge line of the cold leg have teen postuisted. The
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8 single-endM rupture in the punp discharge line yields the highest preuvre within l
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the cospartment. We have performed confirmatory analyses using the R(LAP 3 coguter h
code, and based on our results, ne find the calculated pressures in SWESSAR.pl k
acceptable for the $WES$AA-pl/R($AA.35 design combination.
5-5 For tM pressurtaer cosportment pressure analysis, double-ended breats in the pres.
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suriser spray and surge lines have been postulated. Based on our review of the information and our confirestory analysis of the pressurlaer compartment, we conclose
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that the Stort & Webs *.or analysis aad the design differential pressures f=- the i
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Pressuriser compartment are acceptable for the SWESSAA-pl/RESAR.35 design combination.
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For the reactor cavity. Stone & Webster perfo-med a pressure analysis for a poste-1ated limited displacement rupture in the pop discharge line of the cold leg. This break type has been adequately justified in $WE55AA pl ss the most severe one for the a
reactor cavity subcompartment analysis. 84 Sed on our review and our confirectory ahalysis, we conclude that the Stone & Webster analysis of the reactor cavity pres-I
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sure response and the design differential pressures for the reactor cavity for the
$WE55AR pVRESAR.35 de%1gn combination are accept *ble.
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5 tone & Webster has investigated the consequences of an inadvertent actuation of the containment spray system in the contaltuneet. The initfal conditions of the contain.
ment atmosphere were conservatively assumed to be a temperature of 105 degrees L
Fahrenaett, a pressure of 18.7 pounds per squere inch absolute and a relative heidity l
of 100 percent. The spray water from the refueling water storage tank was assmed to i
he at a tosperature of 32 degrees rahrenheit. Stor_a & Webster calculated a maximum i
enternal differential pressure of 2.9 pounds per square inch. We have performed k
stallar analyses and our results confirm the Stone & Webster results. The contalment enternal differential design pressure is 4.0 pounds per square Inch, which we conclude 7
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is adequate.
We have evalu ted the proposed containment systne functional design for conformance a
r with the General Design Criteria, in particuler Criteria 16 and 50. We conclude that i
the proposed containner.t design pressure of 48.3 pounds per square inch-gauge provHet an acceptable margin unen comoared to the mesima calculated pressure of 39.8 po m L
per square inch-Swge due to a postulated rvpture of the reactor coolant system. We will, however. require a utility applicant refc. enc:ng the $WESSAR41/RESAR.35 design comotnation in its construction permit app 1' cation to reenalyse the masinum calculated containment pressure and temperature for a complete spectron of main steen line breaks using Westinghouse mass and energy release rates which have been reviewed and approved by the staff to demonstrate that a margin of 10 percent is available.
I In addition. based on our confirmatory calculations and the 40 percent margin
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Speelf f ed for the subcegartment design pressure differentials. we find that the prcposed subcompartment c'esign pressure differentials are accepteble. Therefore, me have concluded t?.at the contatronent functional design conforst with the require =ents of Criteria 16 and 50 of the General cesign Criteria and is acceptable.
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6.2.2 Secondary CentsNent Fvactional Oesten t
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The SW55AR-p1 appiteation includes three different secondary contairement concepts.
i The base design concept is the annulus building as a partial secondary cont 41rument.
A, full seconcery containment is inc1wied as Design Option A (full enclosure batiding withoutmining)andDesignOptionI(fullenclosurebutidingwithmining). As discussed in Section 2.3.4 and 15.2.2 of this report. these options can be used by a utility applicant referencing the SESSAR-p1 design for specific sites with less favorable meteorological conditions than these for the base design contatraent concept. Our evalual. son of each of these concepts is presented in separate sub.
Sections below.
6.2.2.1 partial Secondary Containment in the proposed SWESSAR pt plant design, the annulus butiding will serve as a partial secondary containment and, in conjunction with the supplementary leak collection and release system is designed to reduce the release of radioactive material to the erwirorument following a postulated loss-of coolant accident.
The annulus building, which surrounds the primary contatrument building along its entire periphery and to about one-half its height, will enclose approatmately 45 preent of the containment steel liner surface area ard 65 percent of the linear length of all steel Itner welds. All fluid Ifnes penetrating the primary containment will terutnote in the annulus building with the exception discussed below.
The supplementary leak collection and release systes is designed to collect the leakate f-on the primary containment into the annulst and process such leakage prior to its release to the environment as discussed in Section 6.5.2 of this report.
Following a loss-of-coolant accident. the supplementary leak collection and release system will establish and maintain a ne9ative pressure of 0.25 inches of water in the annulus butiding and other contiguous plant areas. Tile above negative pressure will be reached within 38 seconds following the accident. We have reviewed the analytical model and assumptions for the pressure response, including assumptions regarding inleakage to the plant areas served by the supplementary leak collection and release system. We conclude that the analysis is acceptable. The suriplementary leak collec-tion and release ustem will be tested periodically to verify its functional capa-biltty and to deterutne that the prescribed negative pressure of 0.25 inches of water can be reached in the prescribed time of 38 seconds and that this Condition can be maintained.
The design leakage rate for the primary contatreent is specified in SWE55AR-P1 as O.20 weight percent per day of the primry containment atmosphere. For the deter-miutton of the offsite radiological consegances following a postulated loss of.
cociant accident. Stone & Webster has assumed tut 50 percent of the design leakage rate (i.e. 0.10 weight percent per day) is released from the primry containment directly to the atmospheret the remaining 50 percent is assuned to consist of (1) leakage through valves and penetrations into tr.e annutus tuilding which =t11 oe 6-6
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cellected and filtered prior to its release to the atmosphem, and (2) bypass leakage through fluid lines not terminating within the anslus building which is not filtered and is released directly to the atmosphere.
A design leakage rate of 0.20 weight percent per day requires, in accordance with 10 CFR Part 50. Appendia J. that the integreted leakage rete. to be measured in a Type A test be Ilmited to 75 percent of this value f.e. 0.15 wesght percent per day. The integrated leakage test. however. will not provide any quantitative essessment regarding the Stone & Webster leakage partitlening assustions of direct leakage to the atmosphere as used in the Stone & Webster dose calculations. In order te justify l
the assimettons. Stone & Webster leered. at our request. tM limit for the integrated leakage rate test from 0.15 to 0.10 weight percent per day.' to establish an upper bound for the direct leakage that could occur through the upper containment shell.
t During our review of the leakage rate test requirements. Storm & Webster had proposed to increase the 0.10 weight percent per day limit by a leakage rate LC'. designated as the se of local leakage rates through specific Itnes which teminate in the annulus butiding and which could be measured conservatively during the integrated leakage rate test. We agree with this approach in principle, however, me cannot conclude that the allowable integrated leakage rate can be increased to account for j
the collected leakage Lc'.
We, therefore, conclude that the alloweble measured inte-greted leakage rate test acceptance level be specified as 0.10 weight percent per day as stated above. We will evaluate a leakage test program that can directly quantify the uncollected leakage if such program is progesed by a utility appitcant.'
i Stone & Webster has evaluated. In accorfance with our Branch Technical Position CS8 6 3. ' Determination of Bypass Leakage Paths in Dual Containment Plants.* 1eakage paths that potentially could bypass the annulus building and therefore bypass the supplementary leak collection and release system. A total of 14 potential bypass leakage paths have been identified with a combined bypass leakage rate estimated as l
1.15 percent of the primary containment design leakage rate, t.e.
0.0023 weight percent per day. These bypass leakage paths will be tested in accordance with the requirements of 10 CFR Part 50. Appendix J. for the Type 8 and Type C tests. Based en our review of the infomation in SWtssAA.Pl. we conclude that this bypass leakage is an acceptable assugtion for the dose caleviations. We will evaluate the details of the bypass leakage test program during our review of the operating Itcense app 11 I
cation by a uttltty applicant referencing 5'IS$AA.P1.
a In conclusion, we have evaluated the proposed design. Iayout, and analysis of the j
annulus building as a partial secondary containment. and the potential leakage paths from the primary contairment. We conclude that the asseption of 50 percent direct.
uncollected leakage and 50 percent collected and bypass leakage is acceptable for the dose model on the basis that the total reasured integrated containment leakage rate is Ilmited to 0.10 weight percent per day as stated in the proposed technical specf.
fication for the partial dual containrent cmcept of the $ht$5AR.P1 design. Our
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evaluation of potential offsite doses from a postulated design basis accident in ac.
I cordance with these leakage rate assumptions is presented in Sectico 15.2 of this l
report.
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Based on our review of the information provided in SWE$$AR Pl. we conclude that the functional design and analysts of the annulus twilding in conjunction with the sup-plemetery leak collection and release system as a partial secondary contairment is acceptable.
6.2.2.2 Full Enclosure Buildine Without Minino (0eston Option Al i
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Stone & Webster has provided. in Appendia A of the SWES$AR-P1 Safety Analysis Report.
I a design for an enclosure building without provisions for alAlag, which is identified as Design Option A of the SWES$AA-P1 appilcation. Option A is an addition to the annulus butiding and together with the ar.nulus building forms a full secondary containment. Design Option A reduces further the potential for direct leakage, following a postulated loss of coolant accident frac the primary containment to the j
I envirorment. Thus the $WE$$AR P1 design with Design Option A can be used by utility f
applicants for sites with meteorological conditions less favorable than the limiting conditions for the partial secondary contairseent (see Section 1$.2).
The enclosure building is a cylindrical structure completely surrounding the primary contalruent abuve the roof of the annulus building and is designed as a structural steel framework with metal stdtng. The building is designed to withstand the safe shutdown earthquake and will remain functional under all applicable loading con.
ditions except for the design basis tornado loads. During the postulated design basis tornado, the metal siding will fail, however, the structural steel framework l
1s r signed to remain intact. The' enclosure butiding is designed to form a con-l tinuous seal with the annulus butiding so that a tight enclosure around the entire I
primary containment will complete the full secondary contatrument.
The supplementary leak collection and release system, discussed in Section 6.2.2.1 and 6.5.2 of this report will also serve the enclosure building in the same manner as the annulus building. The capacity of the supplementary leak collection and release system is increased so that. following a postulated lost of coolant accident.
a negative pressure cf 0.25 inches of water can be achieved and maintained in the enclosure building. The system is designed with the capability to isolate the enclosure building if necessary. The supplemer.tary leak collectica and release system allt be tested periodically to verify its functional capability and to de-termine that the prescribed negative pressure of 0.25 inches ef water can be reached within the prescr bed time of 38 seconds and that this condition can be maintained.
I We have reviewed the analytical model and assumotions for the pressure response in the annulus building and the enclosure building. including assurGtions 79arding inleakage to the plant areas served by the suppleavntary leak collection and release Based on our evaluation, we conclude thaf the analysts is acceptable. Se system.
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capability of the supplementary leak collection and release system as an air flitra.
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tion system for the processing of potential leakage from the primary containment into the secondary contatrumet is discussed in Section 6.5.2 of this repo*t. The design 1eekage rate for the primary containment is 0.20 weight percent per day. In accordance with 10 CTR Part 50. Appendia J. the acceptance level for the integrated leakage rate On the basis of test (Type A test) util be specified as 0.15 welght percent per day.
our evaluation of potential leakage paths discussed in the previous section. we also conclude that, with the exception of the combined bypass leakage rate of 0.0023 teelght percent per day, the primary containment leakage vl11 be collected in the annulus and the enclosure buildings and util be processed by the supplementary feet collection and release systes prior to its release to the envirorment.
We have reviewed the adequacy of the proposed design criteria and bases for the enclosues butiding without mining. Design Option A.
We conclude that the combination of the enclosure building with the annulus building forms an acceptable full secondary f
contalement concept, i
6.2.2.3 Full f aclosure Building stith Minina (0esten Cotton 8)
Stone I Webster has provided, in Appendia B of the SWISSAR Pl Safety Analysts Report, a design for an enclosure building with provisions for mining, which is identified as Design Option B of the SWE55AR P1 application. Option B is an addition to the annulus building and together with the annulus buildtr,g forms a full secondary contatraent.
Design Option B reduces further the potential for direct leakage. fellowing a postu-lated loss of-coolant accident. from the primary contalement to the environment.
Thus, the SWE55AR-P1 design with Design Option B can be used by utility applicants for sites with metecrological conditions less favorable than the limiting condittens for the partial secondary containmont (see Section 15.2).
The design for the enclosure butiding of Design Option B is identical to and wl11 seet the same requirements as the enclosure butiding for Detign Option A discussed in Section 6.2.2.2 atove.
The enclosure bt11 ding for Design Option. B in combination with the annulus building is a full secondary containma..t design. Following a loss of coolant accident the fuel tiellding ventilation system discuned in Section 9.4.2 of this report. will be used to reduce and ratetain a negative pressure of 0.25 inches of water in the annulus butiding and other contiguous areas, thereby collecting the potential leakage from the primary contaltsent into the annulus building. For Design Option 8. the fuel butiding ventilation system will be designed as a seismic Category I and a j
Safety Class C systen. The discharge from the fuel building ventilation system ts directed into the enclosure building. For Design Option B. an air mising system inside the enclosure butidtrg provides a holdup for fission product releases from the The systee prirary contairrent by mining the atnesphere in the enclosure building.
g consists of sin fans and related duct work. The syste'n is designed as a safety 69 L
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related system and la accordance with the requirummets for seisste Category I system.
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g The supplementary leak collection and release system, discussed in Section 6.2.2.1 I
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and 6.5.2 of this report is designed to natntain a nogettre pressure of 0.25 inches of water in the enclosure betiding she a process the air flou prier to its discharge i
f to the enviroment.
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Ide have reviewed the analytical model and esem pttens for the pressure response la the annulus building ard the enclosure butiding. including assumptions regarding
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tnleekage to the plaat areas served by the fuel building ventilation system and the f
Based on our evoluettoa. we con-supplementary leak collection and release system.
1 clude that the analysis is acceptable. The design leakage rete for the primary f
i containment is 0.20 weight Sorcent per day. In accordance with 10 CFR part 50 Appendts J the acceptance level for the integrated leakage rate test (Type A test) j i
will be specified as 0.15 weight percent per day. On the basis of our evaluation of 5
potential leakage paths discussed in Section 6.2.2.1 of this repeet. we also conclude that, with the exception of the combined bypass leakage rate of 0.0023 weight percent i
per day, the primary contatment leakage will be collected in the afumelus and enclosure buildings and will be processed by the supplementary leek collection and reloose system prior to its release to the enviroment.
We have reviewed the adequacy of the proposed design criteria and bases for the enclosure but1 ding utth mining. Design Option B.
We conclude tnat the combination of the enclosure building with the annulus but1 ding forms an acceptable full secondary contaf ramat concept.
6.2.3 Contalment Heat Renoval System The contalment beat removal system will consist of the containment spray system and the containment atmosphere recirculation systen. These systems will reduce the The contaiment pressure following a postulated high energy Itne break accident.
contalruent atmosphere recirculation system will also be used during normal plant operation. whereas the containment spray system will not have a normal operating function.
Tuo emergency The containment spray system will consist of two redundant trains.
All active compeaents of the system seps will be provided inside the contairment.
e will be located outside the containment to facilitate maintenance operettons.
Protection against internally generated missiles will be provided by direct shielding or physical separation of equipment. The system will be classifted as seismic The containment spray pump recirculation intake in each of the contain.
Category 1.
ment emergency sumos st11 be enclosed by a screen assembly to prevent the entry of debris which could clog the spray nonles. The protective screen assembly design f s consistent with the guidelines of Regulatory Guide 1.82.
A hig% contairret pressure signal frca the engineered safety features actuation The system design system will autcriatically actuate the contaireent spray systen.
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l The spray will perwit manual operation of pumps and valves from the control room.
When the tank l
will 1.nitia11y take suction from the refueling water storage tank.
reaches a low level, a switchover from idection to recirculation will be initiated f
autametically as described in Section 7.3.1 of this report.
i Sufficient not positive suction head will be ave 11able to the spray pues for both The evaluation of the not the iM ection and recirculation modes of operation.
The positive suction head is consistent with the guidelines of Regulatory Guide 1.1.
containment atmosphere recirculation system will consist of four equal capacity fan costers. The system components and equipment rwquired to remain operable following f
an accident will be located outside the secondary concrete shield for missile protec-l f,
tion at en elevation that precludes floodtg. and will be of seismic Category I i
design. A high containment pressure signal or a safety idection signal from the l
engineered safety feature actuation system will automatically actuate the contairment j
The system design will permit remote manual opera.
atmosphere recirculation system.
l tion from the control room. Cogonent test data to demonstrate the performance capability and reliability of each cogoneet wl11 be provided in an operating license i
f We find this application by a utility applicant referencing the $WE$$AR-p1 design.
approach acceptable.
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Based on our review of the containment heat removal systems, we conclude that the systems will be designed in accordance with the requirements of Criteria 38. 39. and i
40 of the General Design Criteria and are, therefore. acceptable.
6.2.4 rentainment Air Cleanuo Tystem In addition to its heat removal fwnction, the containment spray system will be used to reduce the airborne concentration of fission products inside the containser.t followie; a postulated accident. For this purpose, soditan hydroxide will be added to j
ine containment spray solution to enhance the lodine absorption effectiveness of the f
The containment spray system will be actuated automatically on a high solution.
contairment pressure signal. The system initially will take suction from the refuel-s f
On a low water level signal from ing water storage tant containing borated water.
I the refueling water storage tank, the pump suction will automatically be switched to the recirculation mode of operation as described in Section 7.3.1 of this report.
At our request. Stone & Webster revised the chemical additive systes for the two i
train contaifunent spray system of the Sht$$AA.P1/RISAR.35 design continatlod as
$odlum hydroxide from the chemical addition tank stil be injected into the follows.
I borated water drawn from the refueling water storage tank by a separate eductor, one for each of the two redundant spray system trains to produce a pH value of approsf-mately 9.3 during the injection phase coeration of the contatteent spray system (the pH value is a measure of the hydrogen ton concentration and f adicates the degree of acidity or alkalinity of a solutton). $xium hydroxide injection will continue af ter the automatic switchover of the containnent spray system purts to the recirculation mode of operation untti a minimum pH value of 8.5 is achieved in the contairrent 6 11 L
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smo. The contatraent spray system is designed such that the pH of the spray solu-tion during this mode of operation will not exceed a value of 11.0.
9 Installed eductor systems of this type have demonstrated sensitivity to the layout of the piping for the sodium hydroalde addition system. mainly at a result of friction i
losses in the piping. It is essential, therefore, that full-flow preoperation of tests be perforund under all undes of operation of the contalement spray system and l
l of th emergency cort cooling system to verify that the contairamat spray system will l
produce the sodium hydroxide concentrations required to achieve the previcutly stated pH values. Stone & Webster has included in SWF.55AR.P1 information regarding shop
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tests of the eductors as well as the necessary preoperational tests. The performance j
of this preoperational tests, homever, is not within the scope of the SWESSAR.P1 f
appilcation as stated in Section 14.0 of this report, but is the responsibility of the utility applicant referencing the SWESSAA P1 design. This is an interface setter l
to be addressed by the utility applicant in its construction permit application.
The containment spray system is designed to achieve a long ters pH value of 8.5 in j
the containment sump 50 that the equilibrim partitioning of the elemental iodine i
between the 11guld and the gas phase will maintain a decontaminatto.1 factor of 100 in f
i the containment atmosphere. For the two-train system for the SWES$AA-P1/RESAA.35 l
design coelnation we have calculated removal coefficients of 10 per hour and 0.60 per hour for the elemental and particulate forms of iodine, respectively, based on t
an estimated effective spray coverage of 85 percent of the total free contsirunent volume. These values have been used in our evaluation of offsite doses discussed in j
5ection 15.2 of this report.
Stone & Webster revised, at our request. the proposed 5WC55AA.P1 design of the con.
I tatraent spray system by providing only one type of spray not:1e for all ring f
l headers. This util result in an improved fodine removal effectiveness of the spray as compared to the earlier provision for two different types of spray nor:1es.
We have reviewed the proposed two-train contatraent spray system for the SWI55AR.T1/
RESAR 35 design. including the additional information provided by Stone & Webster. We I
have reviewed the revised design and find that the above concentrations can be main.
tained with the specified flow rates for the two train contalrment spray system design of the $WE55AA P1/Rt5AA 35 design combination. Based on our review we conclude that the system is acceptable for a Preliminary Design Approval. it will require in accord.
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ance with our interface requireeent. 4 utill'y appilcant referencing the 5455AR-P1 l
design to perform the necessary preoperational test to demonstrata that the required sodium hydronide concentration can be achieved with the system.
i 6.2.5 Contairment Isolation sviten l
The containment isolation system w411 be designed to automatically isolate the con.
tairment atmosphere frcm the outside ervironment under accident conditions. Double barrier protection. in the form of closed systems and isolation valves, will be 6 - 12 t
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provided to assure that no single failure will result in the loss of containment l
fntegrity. The contatraent isolation provisions will be designed as seismic Category
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I equipment and will be protected against potential missiles. The $WE55AR.Pl design j
of the containment isolation system will incorporate the provisions for certain
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system lines and isolation signals described in RCSAA 35. We have reviewed the inter-l face requirements in RISAR-25 and conclud'e that they will be met by the 5WES$AA.P1 f
l f
design.
i We have reviewed the closure times for the isolation valves. Valve closure tiews l
f are established on the basis of minleiting the release of contaltment atmosphere to l
j the environment under accident conditions. to mitigate the offsite radiological l
j consegsences, and to assure that the emergency core cooling system effectivenes; 15 l
l I
not degraded by a reduction in containment backpressure. We conclude that the closure times for the isolation valves are acceptable.
4 Based on our review, we have concluded that the design of the contairment isolation system conforms to Criteria $4. 55. $6 and 57 of the General Design Criteria and to the recommendations of Re9ulatory Culde 1.11 and is acceptable. We will require a utt11ty applicant referencing the $WtisAR-P1 design to deconstrate in its construction permit appilcation that the potential offsite esposure'. resulting from a postulated failure as described in Regulatory Guitie 1.11 mill be consistent with the rectFulenda*
tions of the guide.
The contatraent purge system for the proposed $WESSAR.P1 plant is designed with the I
objective to reduce airborne radioactivity in the contalement. to limit radiation esposure to operating personnel, and to provide outside air to the contalruient during entended periods of occupancy. The systee will be operated only during hot and cold reactor shutdown and, therefore. the isolation valves will te closed during f
all other modes of operation.
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Plant operating experience of seny plants has shown, as described in our Branch Technical Position C$8 4 'Contairmenv Purging During Normal Plant Operations.* a need for purging the contairment during normal plant operations. Purging of this type is required. for example, to alleviate excess air leakage into the containment from pneumatic controllers, to reduce airborne activity within the containment to facil-t ttate personnel access during reactor peer operetton, and to control the contJrnent atmosphere pressure, toeperature and relative he ldity. Should a lo; of-coolant accident occur during such containment purge operations, an open path would be j
Provided for the release of radioactivity from the contairvnent to the environment.
l Stone & Webster provided, at our request, additional inforviation in support of the
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$Wi$$tA.Pl position that purging of the contairraent during the operation redes of i
The con-reactor startup to power operation and hot standby will not be necessary.
trol of temperature, pressure and h eldity inside the $WES$AR.Pl contairment will not be accceplished by the purge syston but t v the proper des 11' of other containment wcatation system caring norval operations.
syste s such as the contairsent atmosphere -
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The containment instrument air systan will be located inside contaimant including the
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etr intake and discharge, and consequently there will be no pressure builhp within the contalrument from the operation of this systan. Tlur containment atmos.Aere flitre.
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tion system will be used to control the airborne radioactive fodine, town access to the containment is required during periods of normel operation.
Based on these considerations. Stone & Webster concludes that purging of the contain-ment during reactor operation will not be necessary. Stone & Webster, hauever, has included at our request tne spare penetrations in P.he proposed contalment design that l
could be used by a utility appitcent for the installation of a supplemmatal on.1ine purge systam.
Based on our review, me conclude that the SWESSMt pl containment purge system in
' j conjunction with its proposed modes of operation is acceptable. The subjact cf contain.
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sent purging during normal operations is an interface setter that sust be addressed by a utility applicant. A utility appitcant referencing the SWESSAR.Pl desips cuuld commit to the proposed modes of operation in its construction permit applir.ation, and
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i we will include in the plant technical specifications at the operatirg Itcease stage the requirement that the contafrument purge system be isolated durlog reactor operation.
The utility applicant could also elect to use an on-line purge system. utiliaing the two spare penetrations. We will evaluate this interface matter during our review of a construction permit applicatte. of a utility appilcant referencing the M55Mt.P1 l
design.
6.2.6 Combustible Cas Co trel Syste9 following a loss.cf-coolant accident. hydrogen sisy accumulate inside the contairunent as a result of (1) a chemical reaction between the fuel rod cladding and the steam resulting from vaporization of energency core cooling water. (2) corrosion of construction materials by the alkaline spray solution, and (3) radiolytic decomposition of the cooling water in the reactor core and the contalrunent sumps.
In order to mitigate the consequences of hydrogen accumulation in the con *.altunent, two redundant hydrogen recomoiner systens to be locateel outside the containment, and a backup purge systes will be provided. Each of the 100 percent capacity reconoiners and the backup purge system will be capable of processing the containment atsosphere at a rate of 50 standard cubic feet per minute. The recombiner system will incerterate several Jesign features to assure the capability of the system ta remain ocerable in the event of an accident. Anong these are: (1)seismicCategcryIdesign.(2) pro.
tection from missile end jet impingement, and (3) redundancy to the extent that no single cceponent failure will disable both recombiner systeess.
The type of hydrogen recroiner for the Corbustible gas Control system is ect within the scepe of SiitSSAR.pl as stated in Section 1.2 of this report. The 6tility applicant referencing SiiE55AR.PI will use either electric hydrogen recortiners that we Pave i
f previously found acceptable, or will provide a Cor"Olete description and analysis of 6 14 4
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the hydrogen rocoetner selected and will conduct tests to demonstrate the functional I
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laside contafruhent will be permanently installed in each of the suction lines to the y
recombiners.
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hf Stone & ' Webster has performed an analysis of the post accident production and occasua f
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14 tion of hydrogen within the containment which is consistent with the guidelines of Regulatory Guide 1.7 as modified by Branch Technical Position CS8 6-2 ' Control of
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j Coeustible Gas Concentrations in Contafsuant following Loss-of-Coolant Accident.*
i The detailed determination of the metalmter reection was performed by Westinghouse in accordance with the requirements of 4pendia K to 10 CfR Part 50 for the RESAR-35 E
t design.
f Our evaluation of this analysts is presented in Section 6.3.4 of our safety e
Evaluation Report for RESAA-35.
Stone & Webster has assumed a metal-uster reaction value of 1.5 percent for the 5WE55AR Pl/RISAR-35 design. (M the basis of our experi-f ence with such calculations, we consider this to be a conservative design value.
However, we will require a utility applicant to confirm this design value for the I
SESSAA-Pl/RESAR 35 design coetnation when tr.e detailed determination of the setal.
i water reaction will be provided by the utility applicant in accordance with 10 CfA Part 50. Appendix K for the actual emergency core cooling system evaluation. This Is an interface matter to be addressed by the utility applicant in its construction permit application.
The analyses provided in SWE55AR P1 Indicates that the hydrogen concentration in the containment will not resch the lower flassability limit of 4 volwne percent until about 35 days af ter a postulated loss-of-coolant accident. Hydrogen recombiner operation.
l however, will be initiated 15 days af ter en accident occurs to maintain the hydrogen concentratton below the lower fleemability limit. If the recombiners are not func-tioning, the backup purge system operation will be initiated 30 days af ter en accident or sooner if required by the indicated hydrogen corcentration.
Our confiruetory anaisses verify the acceptability of the hydrogen generation analysts presented in SESSAR.Pl. and the effectiveness of the coeustible gas control syttems to maintain the hydrogen concentration wf thin the containment below acceptable Ilmits.
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We have reviewed the proposed combustible gas control system with regard to the require-ments of Criteria 41, 42, and 43 of the General Design Criteria and the reccemendations of Regulatory Guide 1.7.
We conclude that an acceptable system as described in SWE55AR.Pl can be provided for cor$ustible gas control following a loss-of-coolant accident. The hydrogen recombiner subsystem is an interface ratter to be addressed by a utt11ty
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applicant referencing the 5455AR-pl design. We will review the design of the system
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in the construction perett application of a utility appifcant referencing SWE55AR-Pl.
6.2.7 Containrent teskage Testtm Procram I
The containment design will include the provistens and features to satisfy the,
f testing requirements of Appendix J to 10 CFR Par *. 50. The design of the containtnent l
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penetrations and isolation valves will posit periodic leak rate testing at the c
h pressure specified in Appendia J.
Included are those penetrations that have gasketed i
seals and electrical smetrations.
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the proposed reactor contairment leakage testing program will comply with the re-1 1
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qukements of Appendia J to 10 CPR Part 50. Such compilance will provide adequate l
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assurance that contatruant leaktight integrity can be verified throughout the service
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lifetime and that the leakage rates will be periodically checked during service on a I
h,
, timely basis to maintais such leakages within the specified limits.
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y Maintaining contairmant leakage rates within such limits provides reasonable as-
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surance that, in the event of arty radioactivity releases within the containment, the
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loss of the contairement etacsphere through leak paths will not be in excess of i
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acceptable limits. Compliance with the requirements of Appendis J constitutes an f.
I acceptable basis for satisfying the requirements of Criteria $2. 53, and 54 of the j
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General Design Criteria.
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i 6.3 Emeroency Core Coolino $rstem I
The design and analysis of the emergency core cooling system is not within the scope
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e of the SWES$AR.P1 appilcation but is within the scope of the RFSAR.33 design. Our i
e evaluation of the emergency con cooling system design and analysis for the kESAR 35
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e design is presented in Section 6.3 of our Safety Evaluation Report for RESAR.35.
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However, cerwen aspects of the minimum contairement pressure calculaticn as related
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e to the emergency core cooling system evaluation are within the scope of SWES$AR-P1 as
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e discussed below.
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I Appendia K to II, CTR Part 50 of the Consnission's regulations requires tf.P the r
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effect of the operation of all installed pressure reducing systems and processes be included in the emergency core cooling system evaluation. For this evaluation it is conservative to minimite the containment pressure since this will increase the resistance to steam flow in the reactor coolant loops and reduce the reflood rate in b
the core. Following a loss-of-coolart accident the pressure in the containment f
butiding will be increased by the addittoi of steam and water l' rom the pr1%4ry P
reactor systen to the containment atmosphere. After initial blowdown. heat transfer from the core, primary metal structures, and steam generators to the emergency core
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cooling system water, will produce additional steam. This steam together with any emergency core cooling system water spilled from the primary system will flow through the postulated break and into the containment. This energy will be released to the containment during both the blowdown and later energency core cooling system opera.
tional phasest f.e,. reflood and post eflood.
(nergy reoval occurs within the containment by several means. Steam condensation on
- he contairment walls and internal structures serves as a passive heat sink that becomes effective early in the blowdown transient. $wbsequently the operation af the Conteircent heat removal systems such as contairment sprays and fan coolers mill 6 16 E** -
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remove steam from the sentairment atmosprare. When the steen removal rate esceeds the rate of steam addition from the primary system, the Containment pressure will
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decrease from its maximum value.
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The emergency core couling system containment pressure calculations were perfocued
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for the RE$AR.3$ design on that app 1tcation with the >stinghouse energency core f
cooling system evaluation model. As stated in Section 6.3.4 ef our Safety Evaluation
(
Report for RESAR.35. we have reviewed this model and concluded that it is acceptable for the RESAR.35 emergency core cooling system evaluation. The minimum pressure F
calculation by Westinghouse included assumptions for thw containment net free volume.
passive heat sinks, and operation of the containment heat removal systems with regard
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to conservatism for the seergency core cooling system analysis. The data for the passive heat sinks are conservative in comparison with our recommendations in our Oranch Technical Position C58 61 " Minimum Containment Pressure Medal for PWR i:C$
Performance Evaluation."
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We concluded in Section 6.3.4 of our Safety Evaluation Report for RESAR 3$. that the
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plantdependent input information fe the emergency' core cooling system containment pressure analysis 11 RESAR.3$ is.reaseably conservative and that, therefore, the g
calculated contalement pressures are in accordance with Appendfx K to 10 CFR Part 50 t
of the Commission's regulations. We also concluded that each appilcation utilfring h
the RESAR-35 emergency core cooling system evaluation must demonstrate that the significant containment Urameters for the balance-of-plant design are conservative k
k when compared with those used in RESAR-3$.
j Stone & Webster references in SWES$AR.Pl the amargency core cooling system evaluation G
f model by Westing %use and states that the minime containment backpressure will be
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calculated for the $WE$$AR-P1/RESAR.35 design combination as part of the evaluation.
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To determine the applicability of this model for the SWESSAR-P1/RISAR 3$ design.
i appropriate containment paraAters for the $WESSAR.Pl containment have been included in $WE55AR-P1 t
We have reviewed the plant dependent input parameters provided in SWESSAR.Pl and y
determined that the pssive heat sink data are not consistent with the guidelines in
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our Branch Technical Position CSB 61. In addition the input parameters used in the r-RESAR.35 analysis are signif feantly different from those provided in $WESSAR.Pl.
For 2
emp1r. the $WESSAR-P1/RETAR.35 plant design is based on a higher spray ' low rate.
1arger containment net free volume 10wer spray solution temperature, and loner b
outside temperature.
These differences mould result in a lower mirimum containment backpressure than the minimum backpressure calculated in RI$AR-35.
i We, therefore.
conclude that the emergency core cooling system analysis provided in Ri$AR 35. is not applicable to the $WE$$AR P1/RESAb35 plant design. We have advised Stone & Febstte t-of our conclusions that the SWESSAR-P1 analysts does not comply with the assumptions 5pecified by Westinghoute for its emergency core cooling system.
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q mergency core cao11ag syste etnimum contairment backpressure analysis, consis.ent j
with the guidelines of our tranch Technical Positten CSS 6-1 and in accordance with f.
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the Comission's regulations (10 CTR Part 50. Appendia L) will be provided in the j'
construction pevisit application referencing the $lESSAR-71/RE$AR-3$ design combina.
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h tion. Tae utility applicant must also demonstrate that the SlESSAR-P1 plant-dependent t
input parameters are not less conservative than those used in the RESAR.35 evaluation, t*
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This is an interface matter to be addressed by the utility applicant in its construc-
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j tion permit appilcation.
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g 6.4 Control Room Hobitability l
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.I Stone 8 Webster has proposed to meet the control room habitability regulNuents of
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E Criterion 19 of the General Design Criteria by use of concrete shielding and by installing an eme gency outside air systum with dual inlets. $1nce the actual
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j.1 location of the inlets is site related. Stone 8 Webster has stipulated in SWESSAR-Pl.
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as an Interface requirement to be met by the utility applicaat refere1cing $WES$AR-Pl.
h that the two inlets be separated by a distance of 1.000 ta 1.400 feet and that they
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be located in an appromiente orientation of 180 degrees with respect to the control t
building. Each of the tual emergency inlets will Contaid a 2.000 cubic feet per minute charcoal filter train and # n. to be located in the control rom building, J.,
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The emergency mode will ee initiated autoestically in the event of a'contai eet
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isolation signal and also upon signals frue the normal control room air intake L
radiation or chlorine detectors. The novieil air intake will be closed and the pi control room ventilation system switched to the recirculation mode. The emergency r
dual inlet system will automatically selc c an uncontaminated in1(t to admit up to
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2.000 cubic feet per einute of filtered on,tside air to anlitain the ;ontrol room f
atmosphere at a pressure of 0.25 inches of water above atmospheric pressure ano thus f.
avoid inleakage into the control room.
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Each of the dual amergency intets will be provided with two dampers in series to h
assure isolation of the inlet in case of unfavorat,le air conditions at its location, b
Siace tSe proposed design does not assure the availability of air from the remaining inlet, asselr.g the closure of one of its two dmoers as a result of a single active i
failure, we requested Stone & Webster te provide an naluation of the damper con-figuration and control of the remote air intake system demonstrating that talet air
'I i flow as well as isolatinn can be accomplishe<l assaing a single active failure.
[f Stone & Webster provided adottlonal information which states that manual corrective Jction will be taken if one of the valves in tte uncontaminated air intake falls to pj The <alves All bt !cc&ted in the control build.ng, a seismic Category I open.
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structure, with acequate access to the velv 5.
'he valves can be opened with a lever of hand wheel. Stone & leb%ter also it.cluded, at our re%est. the following criteria
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i for the d? sign specifications and installation cf the valves or dampers in the eergency air intates:
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f (1) The need for manual manipulations of the failed valve or deper will not be i
C recurrent during the course of the accident. Manipulation will not be more.
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them once during the accident. Adjustment or realtgment of ether parts of
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the system will be possible from the control room with the failed valve or dauper in a fixed position.
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(2) 43ropriate control room instrumentation will be provided for clear indication l
and annonication of valve or damper malfunction.
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(3) The valve or damper components will be identified as to which are internal k
(non-repatrable) and which are saternal (repeirable). These will be designed as follows:
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Internal valve components (components that are diffleult to repair manually without opening the ductuork) will have an entremely low probability of failure.
b.
External valv? components (conponents, including motors and power supplies, that are assumed to be repatrable or removable) will be designed to ensure
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that the failed salve component can be bypassed easily and safely and that the valve can be manipulated into an acceptable position. The electronic componefits must be isolated from other equitment to assure that the repair operations de not result in further equipment failure.
l (4) The location and positioning of the valve or damper wils permit easy access from the control room for cenwilent repair, especially unde applicatie design basis accident conditions.
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(5) Periodic manipulation of the valve or damper by control room operators should be required for training purposes and to verify proper manual operability of the l
valve or daewr. This aspect is within the scope of the utility applicant referencing the SWE55AR.P1 design.
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We have independently determined the additional operator exposure resulting from a k
postulated valve failure in the non-contaminated emergency air intake. We postulated that the valve does not respond to the 'coen* signal and manual action is, tPerefore.
reoutred to open the valve. Our evalu tten is based on the asswnotfon that the
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control room will not be pressurized for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 35 minutes af ter the postulated accident (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> delay in Nnual action. 5 minute repair time Msed on information {n SWE$$42-pl and 33 minute margin). We conclude that the operator esposure during the
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course of t'e coritrolling design basis accident, including the contribution from non.
L pressart:4'.'on of the control room for 2 houra and 35 minutes. aculd be t'elow the f
guideliws of Criterion 19 of the Ceeeral Design Criteria.
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' He have reviewed the proposed provisions for manual action and conclude that they l
provide en acceptable approach to meet the single fatture criterion. Our evaluation j
of the filtration system for the control room pressurtaatten system is presented tr.
Section 6.5.3 of this report. and the entire control room air-conditioning systen is discussed in Section 9.4.1.1 of this report.
Stone & listster has spectfled as an interface requirement, that a attitty applicant referencing $E$$AA pl susst provih pwtable self-contained breathing units utth an att supply for at least sin hours. These units will be used in case of a texte gas release or smoke conditions that potentially could close the normel and both emergency air intakes. In this case the control room will be completely tse14ted without pressurization. Ide find these provisions acceptable and stil? evaluate full compliance r.f th the recommendation of Regulatory Guide 1.78 during our review of the construction pcruit appittation of a utility app 1tcant referencing $W5$AA fl.
l 6.5 peered safety Features Air Filtration Systems 1
6.5.1 Swatery Desertetton I
l The engineered safety features air filtration systems for the $wt35AR P1 design util consist of process equipment and instrumentation designed to control the release of I
radioactive materials in gaseous effluents (radioactive lodine and particulate metter) f following a casign basis accident. Two filtration systens util be provided for this l
purpose, the supplementary leak collection and release system, and the control room pressurtaatton tysten.
6.5.2
$spolenentary teak Cettection aad Rel;ase System The function of the supplementary leet collection and release system is to control the release of radioactive materials in gaseous effluents from the plant. The supplementary leak collection and release system will collect and r ocess the seakage from the primacy containment into the annulus butiding follaring a fuel t.<ndling l
accident. The supplementa*y leak col 1Jction and release system also serves the enclosed main steam and feedwater valve areas located on the roof of the annulus l
f building and the electrical tunnels in the annulus but1 ding. T>e system will be j
tesigned to natatain a partial vacuum of 0.25 inches wate. gauge in the annulus l
butiding. fuel building. and all areas enntiguous to the contatteent structure following a loss.cf. coolant accident. The supplementary leak collection and release system is a redundant system. [acn of tPe two trains has a distga capacity of 15.000 l
cuote feet per minute of air and includes the followirig ccepocentst demister, elec-tric neating coll. prefilter, hip efficiency particulate air filters. carb.,n adsort.er and fan. The equipment and components mili be designed to Owltty Growo C and setssic Category 1 requiree. tats. De $4plemntary leak collection and release system util te located in tre arnulbs balldteg. a sensetc Category 1 Strature.
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following a less-of-coolant accident both trains of the system will be started on the t,
receipt of a containment isolation signal. Following a fuel hand 11ag accident, both l
trains of the systes will be started on a signal from the area radiation monitors located la the fuel building exhaust air duct. Since only one train of the system is required to maintain the negative pressure the operator asy terminate cp-ration of i
one of the tuo trains. We have determined that the supplementary leak collectfun and release system is designed in accordance with the guidelines of Regulatory Guide 1.52 and is capable of controlling the release of radioactive meterials in gaseous effluents 1
to the environment following a less-of-coolant accident or fuel handling 3Ccident.
lie. therefore, conclude that the system is acceptable.
6.5.3 Contret Room Pressurirstion System The function of the control room pressurtration system is to supply clean, non.
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radioactive outside air to the control room af ter a desf n tests accident or cther
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outside emergency conditions and to pressur13e the control room to a ninfeum of 0.25
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taches of water. The system is a part of the control room air co-sttioning system l
i discussed in Section 9.4.1.1 and is designed to permit operating personnel to remain l
in the control room after a design basis accident or under other emergency conditfons as discussed in Section 6.4 of this report.
I The control room pressurization system is a redundant system with the tw emergency air intakes to be located remotely from tre plant. [ach train has a design capacity of 2.000 cubic feet per minute of air and will include the following components:
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i desistere electric heating cell, prefilter. high efficiency particulate air filters.
l carton adsorter and fan. The equipment and components allt be oesigned to Quality Group C and.atssic Category I requirements. The system will be, located in a seismic j
I Category I structure, following a loss of-coolant accidevit or fuel handlfeg accident the pressurfiation systen is automatically initiated as described in Section 6.4 of this report. The system may also be inttfated manually.
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We have determined that the control room pressarlastion system will be destgr id in I
accordance with the guidelines of Aegulatory Guide 1.52 and is capable of controllits the air from the emergency intakes to maintain a suitable can'.rol room anyl onment folloulng a loss-of. coolant accident. We. therefore, find the system accep;sble.
6.5.4 Coaciusions Our review of the engineered safety features air filtration systers incitded an evaluation of these systems with res;ect to the guidelines cf Regulstory Guide 1.52.
he have leviewed the system descriptions and design criteria for the supolementary test collection and release system and the control rocn peessurf tstion system. The basis for our acceptance is the conforvunce of t% 5'm155AA.P1 design, d:stgn criteria, and design basis for the air filtration units to applfcable regalaticits and galdes a*4 to staf f pcsitloes and indastry standards. Based on our evalwatton. we conclude tnt the pec:osed,atr filtration units for the hoolerentary leak collection ard r:1 esse system and tPe control recri pretsurf tett:n syste9 are accectable.
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6.6 Eneincered Safety Features Materials j
The materials for the emergency core cooling system are specified in RESAR 35 and are not within the scope of the $ESSAR-pl appilcation. Tk asterials used for the
' engineered safety features associated with the containment systems will be selected te satisfy the requirements of tre A$lt teller and Pressure vessel Code. Section !!!.
Appendia I. aad our positten that the yield strength of cold worked stainless steels shall not escoed 90.000 pounds por square inch.
l The controls on the pH value of the reactor containment spray and coolant water are adequate te ensure freedom free stress corresten cracking of the austenttic stataloss steel components and welds. The controls on the use and fabrication of Austenttic stainless steel in the system satisfy the requirements of Repslatory Guide 1.1 and conform with our tranch Technical petition Mitt 6-1. " Control of pH for Emergency Coeling Water." The control of the pH value of the spray and coolant water. in conjunction with controls on selection of contairment materials, are in accordance with the rocamendations of Regulatory Guide 1.7. and provide assurance that the sprey and coolant water will net give rise to excessive hydrogen gas evolution by corrosion of contatraent metal, or cause serious deterioration of the contairment.
$ tone & liebster has committed in SWt$5AR P1 to comply with the recommendations of Regulatory Guide 1.44 with respect to testing of quellfication pmedure welds for sensittaa tion.
Conformance with the ASME Code and the recommendations of the Regulatory Guides mentioned above, with our reevirment on the allowable mestaum yield strength of cold worked austenttic stainless steel, and with our requirement on the minimum pH value of contaltnent spray a ut coolant water, constitutes an acceptable basis for meeting the s eguirements of Criteria 35, 38. and 41 of the General Design Criteria.
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7.i Gen.rai The $WESSAR P1 instroentation and control systems have been revleied utill Ing l
the Commission's General Design Criteria, the Institute of Electrical and Electronics Engineers (IEEE) Standards, as listed in itas to of Appendia C and f acluding IEEE Std 279-1971
- Criteria for Protection $ystems for Nuclear Power $tations.* applicable Regulatory Guides for power reectors. and staff technical positions as the bases for evaluating their adequacy.
The SWC$$AS-PI Safety Analysis Report references appropriate portions of the Westing-l house RESAR 35 Safety Analysis Report for systems that are not within the scope of the
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SWESSAR P1 design. Accordingly, we perforved our review also with regard to interface
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requirements that have been identified for RE$AR 3$ and dich sust be met by the
$WESSAR P1 balance of plant design.
7.2 Reacter Trip System The reactor trip system is within the scope of the RE$AR 35 design except for the following three trip inputs which are within the $WE$$AR Pi Rope:
(1) Undervoltage trip for the reactor coolant pop.
(2) Underfrequency trip for the reactor coolant pe p.
(3) Turbine trip.
The reactor coolant pump undervoltage and underfrequency trips are required for reactor coolant system low fles protection in the RE$AR 3$ accident analyses. Westinghouse l
requires that the conformance of these two trips with IEEE Std 279 1971 and seismic criteria be discussed in the preliminary safety analysis report of a balance-of-plant design. In our Safety Evaluation Report for RESAR 35. we require that any inputs to the reactor trip system. including those which are outside the RISAR 35 scope. Should not in any way result in a degradation of the overall reactor trip system. We.
a therefore, require that the undervoltage and underfrequency trip inputs. including the sensors, be designed to satisfy all requirements of lEEE Std 279 1971 without exception.
Stone & Webster has provided, at o*se request, additional information regarding the reactor coolant pump undervoltage and underfrrquency trips including the following:
+
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(1) identification of specific sections of !EEE $td 279-1971 that are applicable to those portions of the trips within the scope of $VESSAR-Pl. (2) delineation of responsibility for hardware and design tetween $ tone & Webster. Westinghouse. and a l
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utt11ty applicant, and (3) design criteria for the components that generate the j
undervoltage and underfrequency trips. We have reifewed the additional infocustion j
and conclude that the proposed $WES$AR.pl design ft r the reactor coolant pup under-voltage and underfrequency trip inputs are consiste t with the interface requireents spectfled in our Safety Evaluation Report for RESAA-A and satisfy the requirements f
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of IEtt $td 279-1971 and.' therefore. are acceptable.
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J The potential transformers. Including cabinet. and the rescar costant pop under.
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voltage and underfrequency relays, will be designed for Class It service, and all a
system components from the input source to the reactor trip system cabinet will be l
j located within a seismic Category I structure. To assure that the proposed design j
meets the requirements of 1[ft $td 779-1971 in totality. Stone & Webster has committed in $WES$AR.pl to provide additional design details for the reactor trip syste Inputs.
The information will be sutimittse in the application for a construction permit by a utility applicant referencing $WE$$AR.P1 and will include as a minium drawings e
related to the physical location ef the potential transformers. including cabinet.
a the associated buses and connections to the reactor coolant pump power cables. the undervoltage relays, the underfrequency relays and the associated cable routing for the redundant reactor trip system channels from the sensing point to the reactor trip system cabinet. This aspect of the reactor trip system has been defined as an a
taterface metter to be addressed by the utility applicant. We will review the laple-mentation of IEEE std 279-1971 for tr4 proposed design during our review of a construc-tion permit application by a utility applicant referencing $WE$$AA.Pl.
The turbine trip is an input to the reactor trip system and is within the scope of l
$W($$AA.pl. In addition to the logic diagram for this input. Stone & Webster has provided in $Wi$$AR pl the interface requirements for the turbine trip input by i
I coswitment to the appropriate R[$AA.35 sections. Based on our review of the additional information. we cenclude that the proposed design for the turbine trip input within the l
Scope of $WES$AA.Pl meets our requirerents as identified in Section 7.1 of this report.
l Including the Rf 5AA.35 interface requirements and therefore. is acceptable.
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7.3 (noineered Safety Features Systems The engineered safety features actuation systee,is within the scope of the RISAR.35 design. The system will include the instrweetation and controls used to detect a a
plant condition that requires the initiation. operation and control of an engineered I
safety features systee. Stone & Webster has crovided le. Section 7.3 of Shi$$AR-P1 f
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supplemental information for those engineered safety features systems and their I
supporting ausiliary systems that are within the scope of $WES$AR.Pl.
We have reviewed the design of the engtreered safety features systems within the scope of the $Wi$$AA.Pl design. including functional logic diagrams testing provisions.
design criteria and design bases and the analysis provided by Stone & Webster on tPe adequacy of these criteria. tases and interf ace reautrements for the $W($$AA.Pl/R[sta.3%
design cortination. he conclude LNt the electrical systems. Instreentation and 12 e
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d controls associsted with the evneered safety foetures syste in the SWE55AA-P1 scope of design satisfy our requirements identified in Section 7.1 of this report and.
therefore. are acceptable.
7.3.1 Contaiment Sorav trstem a
The contalment spray system is an engineered safety features system which is entirely within the scope of the SWES$AA.Pl design. 'the system will serve as a contaiment teet l
removal system for contalment depressuritetton following a loss-of-coolant accident (see Section 6.2.3 of this report) and will be used to reduce the airborne radioactivity concentration of fission products inside containment following a loss.of-coolant acci.
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l dont(seeSection6.2.4ofthisreport).
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The concalement spray systes for the $WE55AR.P1/RESAR.35 design combination will be a The two train system conHtible with the RESAR.35 emergency core cooling system.
systes will be initiated automatically on a algh high contairmaent pressure signal f
generated by four pressuee transmitters with a two.out.cf.four logic and associated circuitry. The system can also be inttleted manually at the system level. Change.
over from the injection mode to the recirculation mode occurs automatically when the refueling water storage tant led reaches a low level set point (one-out-of tuo). in e
conjunction with the engineered safety features sump level reaching the high set point (one-out-oftwo). Upon reaching this logic condition. the containment spray pw p suc.
tion valves from the refueltag water storage tank close and the suction valves from the i
engineered safety features seps are opened. The changeover function can also be Initiated manually at the system level.
We will review the ietalls of the instrwentation and controls for the containment l
spray system cbngeover functions during our review of an operating Ilcanse application by a utility applicant referencing the SntSSAR.pl design to ensure conformance to IEEE Std 279 1971 requirements.
We have reviewed the dtsign description of the contairrent spray system 'ncluding functional logic diagrams. design criteria and design bases and the analysis regarding the adequacy of these criteria and bases. We con-lude that the lastrumentation and controls associated with the centairvnent spray system will satisfy our requirements a
Identified in Section 7.1 of this report and, therefore, are acceptable.
7.3.2 Main 5tese (Solation The main steam system, encept the steam generators and portions of the associated f
instrumentation and control system, is wit %tn the scope of the 5'al55A2-P1 desfgn.
This includes the piping and valve arrangement as wil as the instruaentation and controls provided for isolation of the min steem lines followieg a Nin stem line break accideet.
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t The analysis of the rupture of the main steam line is presented in RESAR.35. and i
the following interface rewire ents are identified in RELAR 35.
t (1) The electrical instrumentation and controls for the poner operated relief I
valves must be independent and destened such that no single fativre can cesse opening of more than one pomer operated relief vstve.
t (2) Any single failure in the electrical instrumentation and controls for the main l
steam isolation valves should not cause a failure of valees downstream of the r
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l (3) Fa lure in any single valve in either the upstream or domstrees side of the main steam isolation valves should not result in steam flors in excess of the ammant established in the RESAR 35 accident analysts.
l t-Stone A *bster has included these regulWJ egarding single failure in Table 7.8 3 of SW155AA-Pl. based on our review of the additional information i
provided in SWISSAA-pl. we conclude that the proposed design for the electrical.
instrumentation and controls pertelning to the main steam system valves satisfies the RESAA-35 interface requirements and, therefore, we fine the design acceptable.
j To mitigate the consequences of a steam line br%k accident. Wstinghouse has taken j
credit for proper functioning of certain equipment and circuits, most of which are j
in the scope of the SWE55A2 Pl design. Stone & Waster has prsvlied additional j
Information for the instroentation and controls pertaining to the main steam system equ;saent Itsted in Table 15.4 7 0F RISAa.35. The instrumentation and controls for this l
equisment will be designed to the requirements of IItt Std 279-1971 ar.d Ittt Std 308 1971. We conclude that these criteria provide an acceptable basis to validate the assumptions made in RESAA.35 with regard to the main steam line break accident analysis and conclude that this aspect of the SWISSAR.P1 design is acceptable.
7.3.3 Aus114ary Feedwater Srstem j
I The auxiliary feedwater system for the SWI55AR Pl/RESA2 35 design comelnation is f
totally within the SWE55/R-91 scope, it consists of two motor operated pwp trains and one steam turbine-driven ow.o train. Power for each motor-driven pop train and its associated motor operated valves is supplied from a separate emergency alterriating current Ns. Stone & Webster has documented that the traf n supplied by the turbine driven pump does not rely on any alternating current power source and the turbine i
control systes la powered from the Class It direct current power system.
The motor-operated stop valve in the steam line to the turbine, the retor.ooerated uodulating valve, anu the contairment isolation va1ve in the turbine driven pep discharge line will be prwered from en emergency alternating current bus. These valves are norNlly in the open sc *non. The inadvertent closure of any one of the r
above three motor. operated valves will not negate the assuStlons made in the loss of feed =ater flow / loss of all alternating current power accident analysis.
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i We have revleued the proposed design for the electrical. Instrumentation. and Sentrols for the ausillary feodwater system and have concluded that it satisfies
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our regstrements identified in section 7.1 of this report and is, therefore.
l a:ceptable.
I i
7.3.4 muclear Steam Supots 51 stem Interface Nuirements for Safety Systems j
In Section 7.1.1 of our Safety Evaluation Report for R!$AR.35 we stated that the laterface Information and criteria contained in RESAR.3$ as supplemented by the j
addittoman laterface regairements included in our Safety Evaluation Report for RESAR 35.
{
provide ressenable assurance that the balance.cf-plant design can be accomplished j
f in a manner that util validate the assumptions in section 15.0 of RESAR 35. Our j
laterface acceptance criteria for specific R15AR.3$ systems are listed in Table 7 2 I
l of ouc Safety Evaluation Report for RESAR-35.
The RESAR.35 reference design identif tes eleven interlocks between valves of redundant
[
engineered safety feature trains, which are listed under item 16 of Appendia 7A of j
the RISAA.3$ application. In response to the staff's req' west for additional informa-i tion. Stone & Webster supplanented Section 7.8 of SWESSAR-Pl to include a description of the laterfaces and division of responsibility between Westinghouse and Stone &
l Webster for the valves. The Westinghouse scope of responsibility includest the design of the valvest the physical arrangement (including separation) of their switches; and the circuitry design. The Stone & Webster responsibility is limited to the field wiring between these valves. This wiring will meet the separation require.
[ !
ments of Regulatory Guide 1.75.
I
! I We have reviewed this information and conclude that it adequately defin(s the scope of l
responsibility between $ tone & Webster and Westinghouse and that Stone & Webster's I-commitment to incorporate these interface requirenents is in accordance with Regula-tory Guise 1.75 and, therefore. is acceptable.
In Section 1.3.5 of our Report to ACR5 for SWE55AR P1/RESAR 3$ we stated that f
seseral manually-controlled electrically-operated valves agloyed in the RE1AR 35 emergency core coo 11og system did not meet our single failure criterion. We indicated that these valves did not meet the single failure criterton in that electrical mal-g ;
functions could result in spurious valve movements to undesirable positions and thereby result in loss-of-capability of the emergency core cooling system to perform i
its intended safety function and were, unacceptable. We indicated that in lieu of design changes that also may be acceptable. it is acceptable to lock out power to these valves according to the prowlsions of Electrical. Instrumentation and Control Systems Branch Position 18.
- Application of the Single Failure Criterion to Manually.
Controlled Electrically-Cperated Valves." In leend*nt 12 to RESAR 35. Westinghouse Included the capability to lockout and restore motive power to the affected valves from the main cuntrol -oom and to provide in the main control room redundant indication of the positions of these valves. We have determined that with this nodification the design reets the single failure critecton and conclude, therefore. that it is accept-cble as discussed in Section 7.3.1 of our Safety Evaluation Arport for RESAR-35, 7.$
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In accordance with the resolution of this f ten for the RESAR.35 application, we have 7
~ determined that there are no additional inteiface requirments placed upon the
$1ESSAR.pt/RESAR.35 design contbination as part of this resolutton. On this basis.
this setter is considered resloved.
l I
a In "estion 7.4.1 of our Report to ACR$ for $ES$AR.pt/At$AR.3$ w stated that the l
residual heat ruoval system proposed for Rf5AR.35 did n6t meet the single failure
(
criterion in that one of the tue isolation valves in each of the tus trains mes pouered from the same source end therefore, uns unacceptable. h further s% d g
~
- that in accordence with the resolution of this outstanding itas in RESAR.35, any i
requirement placed upon the belance.ot.;, lent sust be incorporated into the
$W$$AR.Pl/RESAR.3$ design combination prior to the issuance of a preliminary Design
' Approval for the $MES$AR.pt/RESAR.35 design combination.
In Amendment 13 to RESAR.35. Westinghouse specified design criteria for the balance.cf.
plant. We stated the bast? for our acceptability of these criteria for the belance.of.
l plant in Section 7.,4.1 and 7.6.4 of our Safety Evaluation Report for RESAR.35. On j
January 21. 1977 a meeting uns held with Stone & Webster in which a design concept
'utill Ing a temporary p wr s pply arrangement to supply motive power to the inoperable o
u isolation valves (Meeting $ussiary dated January 26, 1977) uns presented for our review. This concept uss found to be unacceptable since it could not be designed to acet the single failure criterion. Subsequently. In Ameneont 31 to $WES$AR.P1/RESAR.3$
r in Table 8.4 4 and Figure 8.41 Stone & Webster described a means of supplying motive j
power to the residual heat removal system suction isolation valves by means of pro.
widing tuo Class IE power sources in addition to the two exsisting Class it pouer l
sources in the $WE$$AR.Pl/R($AR.3$ design. This design provides tn.tn the independent 1
and redundant power sources. cross connections, and isolation for the overpres;ure
)
protection interlocks for the residual heat removal system suction isolation valves in a menner that satisfies the single failure criterton for both system operation and isolation. We heve reviewed this design and have concluded it meets the interface requireets for RI$AR 35 and the basis for our acceptability of these interface
)
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require =entt for the balance.cf. plant described in Section 7.4.1 and 7.6.4 of our j
$4fety Ev:luation Report for RESAR.3$. and General Design Cetteria 19 and 34 and.
j therefore, is acceptable.
[
1 We have revleued the engineered safety features systems for the $Wi$$AR.P1/RESAR.31 design combination with respect to the above interface requirspents and conclude that the proposed $wt$$AR.pl design conforms to these requirements end therefore, is acceptable.
We will evaluate the detailed Implementation of these aspects of the design during our review of an operating license application by a utility applicant referer.ctng the 1
$WES$4R pt/RESAR 3$ design combination.
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Y 7.3.5 Periodte Testino of Enetasered safety reatures Systems f
The Ariodic testing of those portices of the protection system within the SESSAR.P1 Stone scope will be in confomance with the recosamendattens of Regulatory Calde 1.22.
(
& Webster provided additional infomation in 5455%I.Pl which identifies 1*n protection j
j functions that will only be partially tested during power operation and iditch provides 3
the bases for their esclusion from complete testing. The bases stated are in confor.
J l
P mence with ths receanendations of Regulatory Guide 1.22.
i
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Based on our review of the additional infomation. we conclude that the criteria for
]
the periodic testing of protection systems within the SE55AR.P1 Scope satisfy our j
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requirements identified in Section 7.1 of this report and, therefore, are.*cceptsble.
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.1 We will reptre a utility applicant referencing the SK55AR.Pl design to sutueit. In its y
j' appitcation for en opersting Itcense. a program for system and sensor response time p' i testing of those porti)ns of the reactor trip system and the engineered safety features
]
c actuation system of the SESSAR P1/RESAR.15 design combination that are within the scope of SE55AR.Pl. The scope of this test program will include safety related j
systems and sensors witnin the scope of SWE55AR-P1. ini tuding those for reactor coolant pump undervoltage and underfrequency, containment pressure, refueling water storage tank level and engineered safety features sump leve).
l l
t Based on our review of the additional information provided in SWE55Ad.Pt. we conclude
~
( l that the criteria for periodic testing of safety systems within the $KSSAR.Pl scope
~
with regard te their response time testing satisfy our requirements and are, therefore, acceptable. We will review the adequacy of the test procadures for periodic respcnse time testing during the operating Itcense stage review of an applicrtion referencing
?
SESSAR-Pl.
T.4 Systems Required for Safe shutdown Stone & Webster has referenced the RESAR 35 safety Analysts Report for information en tystans required for safe shutdom. In addition, supelnmeata) information for 5 K 55AR.P1 systems re e tred for safe shutdown.n been inciuded in SWE55AR.Pl..To
(
l j
meet the requirements of Criterion 19 of the General Destgr. Criteria, the SWC55/Jt rl design includes provisions to control vital systeers required for hat shutdow of the f
reactor from control points outside the control room. An aust11ery shutdow panel This panel will contain all the will be provided in the emergency switchgear etna.
controls and indicators as required in RESAR.35. Additionally. controls and instrwen.
f i
tation for the aval11ary feedwater systen, atmospheric steam dump system. pressuciter heaters (beckup group) and boric acid transfer pumps. Charging flow control valves Other ecutsent and lettown ortf tce isolation valves util be provided in this panel.
required for hot shutdown (typically controls for c:rooment cooling water acd service ester systems) util have controls at adjacent locations to the shutdown panel.
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Af ter tot shutdoem ceMitions have been achieved, a cold shutoc,wn condition will te i
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S able to be accomplished with the controls and instrumestation provided on the easil-lary penet and elseders throughout the plant.
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noe have revleued the electrical systems, instrumentation and controls associated E
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with the systees reestred to achieve a safe shutdoun condition of the plant from outside the asia centrol room., lie conclude that the design of the electrical.
f-
- festrumentation and control systems within the scope of $ESSAR pl for the $E$$AA41/
4 1
RE$AR 35 design cambination cieforms to the reestrements identified in Section 71 l
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ee this repors
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7.5 Safety Related Disolar instrumentatim, k
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The safety related display instrumentation will provide the operator with iaforsiation on the status of the plant to enable his to perform appropriate manual ;,4fety functions
. and for post-accident and incident surveillance. ide revisend the safety related l
display instrumentation for the monitoring of safety related systems within the scope I
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of the SW$$AA41 and for post-accidat and incident surveillance.
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7.5.1 sysessed and Inocerable status Indication for safety system J
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v The safety systems and their ausiliary supporting systems which are included in the f
scope of the bypassed and inoperable status indication system are identified in
$E S$4A pl. The implementation of the recommendations of Regulatory Guide 1.47 has f
been discussed.
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ide have reviewed the information and conclude that the proposed design for the bypassed
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Y and inoperable status indication system is consistent with the recessendations of
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Rep 14 tory Guide 1.47 and. therefore, is acceptable.
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I 7.5.2 Post. accident and incident Monitorino System
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4-The proposed SM$$AR pl design will include a post-accident and incident monitoring system that will provide the operator with the followiry information:
(:f (1) Contairment atmosphere pressure.
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(2) [ngineered safety features swp level indication.
(3) Ausiliary femsster system storage tank level indication.
(4) Contairment isolation indication by stans of the contalment isolation valve position indicator lights.
r (5) Contairmient atmoschere temperature.
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These perameters will be monitored with redundant channels and at least one channel D.Wg,9 W*C.4. '
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will he recorded continuously. The redundant channels will maintain their physical and" t "'S d electrical li,
ce and will be powered from the onsite emergency poner supplies. [.' - Mll l [E'L
[] % f.,.o We hose revieued the criteria for the proposed instruentatten of the post.accidmet and,,;,
facident monitoring system and conclude that the proposed design is in conformance with
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our eewirements and therefore, is acceptable.
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We have revleued the design description design criteria and the Stone & Websber n
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analysts of the manner in dich the design of the safety related display testfemen.. m - gu
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tattee conforms to the propoM criteria. We conclude that the porposed destge of the~t W g
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safety related system display instrumentation conforms to the requirements.identiffN
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_ $51 W4 In Section F.1 of this rsport end. therefore, is acceptable.
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. ;; g., ;N g,g. p. e p" F. 6 Other f astrumentation Systems and Reouf rements for Safety p
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J F.6.1 Enviewmental Qualification or Class It (1ectrical Equipment r('
Stone & Webster has stated in $WE55AR.pl that instrumentation. controls and electricals. _
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-0 eg ipment frportant to safety and within the scope of the SE S$AR.P1 design will be - " ' ~ W p. ".
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purchesed by referencing ![tt 5td 3231974. *Itt Standard for Qualifying Class It j ', n@.,.<cy.(.%M.a
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[gwimment for Aucleer poner Generating Stations.* in the purchase spectficatlaws... N. -
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Stone & Webster also has convirtted in SWE55AR pl to participate in the orderly V_. -?. a.f e.y.
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development of an industry wide gaallfication program under the sponsorship and,,j g; g, Q (..g.,.
direction of a recognized technical sorte f or stellar control body. Homever,if f " g.'hV~r? ',L j9 y
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k problems amerge in specific instances, when attempting te isolament the agtet iregsire-gj. 'Dh hai%
monts of IItt Std 323 1974 one of the followtri methods, singularly or in combination.,.
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will salidate the quellfication for that equissent:
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l (1) Analyses based upon enviroruental tests.
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D; (2) Operating enerience (taking into consideration inservice inspection, perlodle-
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(3) Type tests uttitzing qualitettve aging techniques (e.g.. environr.sntal cW11ag. -
operational cycling. elevated stres, tecimiques).
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(4) >. going or pacing tests.
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u Based on our review of te se coruttaents, we cow 1ude that tre proposed criteria for c
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the saalification of Class It eesipment within the scope of tst5$AR.pl can facilitate devel:pment of a qualification program consister,t with the objectives established in
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IIII itd 323 1974 and that the above urnitments provide an acceptable basis for the f
i Preliminary Design Approval of the Clus !! ewtpment qualification progree. The L ^
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detatis for the developnent of the progran. Including the accep'.ance criteria. have F.9 P
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n toen deffnod as en faterface matter to be addressed le the appitcation for a c A.,
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75 23-a.' /
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^ 7.6.2 fc_Aae and identification of Safete Related faufseent
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y ei.y.L At our repsest Stone & Webster provided additiosol inforestion is 5E55AR41 with l,
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y fdestificatios cf redundest safety related equipneet.
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jn Ihk' > ;N $tane & W6 ster has deleted as seriter Interpretstles of Asplatory Guide 1.75 test
rM,7</..w+.M.;.F, J' non-Class !! -frcuits of 600 Volt classificaties, mPich share eMosures wi b.
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dg @n'$ M ('l cfrewits, can te exampted from the requirements of associated circuits. lastead the y
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..g desty of the electrical systen will comply in all respects with the recuumendations,,
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of Replatery Gufde 1.75. W conclude that the proposed criterta fo* the physical g
j g c.g; p ~yi 1.ndependence of electrical systene fa the $sE55AA41 desip are acceptable.
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We heee reviased the desip description and desip criteria for unintaining ic-_ Me s
and identification of safets related equipment for the SE55AR-P1/RESAR 35 design
'n combination and have concluded the criteria are consistent with the roccumundations
~lI in Regulatory Gefde 1.75 and are. therefore. acceptable.
e
! 4" 7.6.3 Nanust fattfation of Protectiv_e Actions x
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- W Stone & Webster interpreted our recomessadatfons is tregulatory Guide 1.62 regarding the i f i
manual fattfatfos at the systan level to mean that not more than three operator actions
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'.dll be required to result in the actuation of at least one traff sf a systas, inceud-1 6
fag ausf11ery supportion systems. Stone a Webster has prevised additional clartffca-I tin on the manual initiation of protacti/e actior.s. stating that designs requiring,
l i
more than two operator actions per train are not used h the SESta* PI design. On 4,
the basis of this clarification. me conclude that the proposed design for manual initiation of protective act(on within the scope of SE55AR-P1 meets obr ret.utrements identified in Sectioc 7.1 of this report, and threfore, is acceptable.
]:'
d 7,7 Control Systems Not #*outred For. Safety q
Controls for the following systems not rsgaired for safety us (Jeatiffeu in ShtSNPl. Supplemnting the information provided in a corresponding section in Af5AR.35:
(;) Porttoves of the reactor plant component cooling meter system providing water to various noe. safety related syste d*sring c,rmal pl.nt operation.
(2) Contairement leaQ9e monitoring,$ystem.
be have concluded that fgitures in these Control systees are not espected tt degraft p
the capabilf tfes of the plant safety systests in day significant degree or lead %*
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plant conditions more severe than'those for which the safety systems will be designed a
to protect against. We conclude that the control and testrumentation systems satisfy our requirements in Section 1.1 of this rep 0rt. and. therefore, are acceptable.
I 7.8 Instrumentation and Contrn1s Interf ace Reoutrecents i
4 Stone & Webster. at our regaest, has included in SWE55AR-Pl. as Tab 1e 7.8 2. an interface design criteria applicability matria for the instrumentation and control systems for SWESSAR P1/RISM 35 systems. This metria is consistent with the interface j
criteria for instrumentation and controls identified for RESAR-35 systems at the 4'
i boundary of the nximar steam supply system in our Safety Evaluation Report for
,e RE5AR 35. The conforsence of the criteria between SWES$AR-P1/RESAR-35 systems at j
their interfaces facilitates validation of the assmettons made in t!w RESAR 35 l
accident analysis and provides reasonable assurance that the total instrumentation l
and controls for a specific plant appit*ation referencing the SWE55AR-P1/RESAR 35 l
1 design can satisfy our requirements identified in Section 7.1 of this report. We 1
have evaluated this information and conclude that the laterface criteria for the instrumentation and controls for the SWE55AR-Pl/RE5AR-35 design presented in Table 7.8 2 of $WE553-P1 are acceptable.
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8.0 EttefRIC POW 0t 8.1 General r
The electric power systems of the $WES$Ap-P1 balance-of-plant design have been evale.
~
ated with regard to their adequacy on the bases of Criteria 1*/ and 18 of the General Design Criteria. standards cf the lastitute of Electrical and Electronics Engineers (IEEE) as listed in item 20 of Appendia C. Including IEEE Std 3081971 " Criteria for Class IE Electric Systems for Nuclear Generating Stations". and Regulatory Guides 1.6 1.9 and 1.32.
In addition we reviewed the proposed SWE$$AR.P1 design with regard to RESAR 35 interf ace requircunts for the electric power systems.
8.2 Offsite Power $rstem The of fsite power system is outside the scope of the SWISSAR.P1 design and will be presented in the Safety Analysis Ftpo*t of a utility applicant referencing the SWE$$AR-P1 design. However. Stone & Webster has included in SWESSAR-P1 a design description of the connection of the offsite power system to the $WES$AR-P1 plant alternating current distribution systen, and appropriate interface r&quirements with regard to the offsite power system to sieet certain requirements imposed by the RESAR-3$ design.
Three independent 69 kilovolt offsite power connections are required from the uti1Py I
applicant's offsite power system. Two of these lines will feed a separate 69 kilo-I volt /4.16 kilovolt emergency transformer and each transformer in turn will feed one of the two engineered safety features buses of the SWESSAR.P1/RESAR-35 design. The third 69 kilovolt offsite power line feeds a 69 k11ovolt/4.16 kilovolt emergency transformer not utilized during normal plant operations. The offsite power connections wf11 also provide power to non-safety related loads on 13.8 kilovolt buses through 69 kilovolt /13.8 kilovolt reserve station transfomers for startup and shutdown condi-tions of the plant.
For normal plant operation, the par for the non-safety related ".8 kilovolt buses will be derived through the unit auxiliary transformer from the main generator. On unit trip conditio*s. these non-safety related buses automatically will be fransferred to the reserve station transformer for power feed from the offsite source. the loads on t'e engineered safety features buses normally wi'l be fed directly from the offsite l
power system. thus eliminating the dependency of taese loads for power on the plant
=
tureine-generator unit availability and preventing interruption of power to the engineered safety features loads OA a.Mt vii 8-1
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N 8.2.1 offsite Power System Interface Reevirements f
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The following interface requirements on the offsite power system of a utility appli.
cant have been identified in $WE55AR P1 to meet certain functional requPements of i
RESAR-35 safety systens:
}
(1) One of the two offsite power sources must be available to the Class !E system i
(assJuing onsite power is not available) in order for the engineered safety featums to meet the minimes required level of functional performance.
(2) The credible grid decay rate shall not be greater than 5 Hert2 per 14cond.
I 1
l We will evaluate, therefore during our review of a utility application referencing l
the $WESSAR.P1/RE$AR.35 design combination the following i,
(1) The configuration of the incoming transmission lines to the switchyard of the j
plant and the arrangement of the buses and breakers in the switchyard to assure j
that the proposed design will satisfy the offsite power circuits requirements of i
the SWESSAR P1/RESAR 35 design consistent with the requirements of Criterion 17
{
of the General Design Criteria.
l (2) The grid stability analyses to assure that the worst case frequency decay rate in the grid will not exceed the stipulated value of 5 Hert! per second.
The design of t;.e offsite power distribution system within the scope if the SWE55AR.
P1 design will be in accordance with the requirements of Criteria 17 and 18 of the General Design Criteria and the re'conmendations of Regulatory Guide 1.32.
Stone & Webster has specified as an interface requirement it SWESSAR-P1 that the utility applicant's grid decay rate shall not be greater than 5 Herts per second.
l This criterion is in conformance with the requirements specified in RESAR.35 Section 8.0.
We are currently evaluating the Westinghouse design bases for the correlation between the grid frequency decay rate and the Ilmiting underfrequency trip setpoint for the reactor to assure adequate reactor coolant pump coastdowt, capability. Upon completion of our revis, if it is determined that design changes are required, we will require that these changes be identified and teolemented in o appilcation referencing the SWE$$AR.P1/RESAR.;5 design conibination. We conclude that the proposed offsite power system within tha SWi$$AR-p1 score is in conformance with the RESAR-35 interface requirements and satisfies our ret,uirements identified i n Section 8.1 of
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this report and. therefore, is acceptable.
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.j 8.3 Onsite Powee Systems j
8.3.1 Alternatine Current Power System The proposed alternating current emergency onsite power system for the SWE55AR-Pl/
RESAR 3$ design will have two redundant and independent engineered safety features distribution systems. which will normally receivit peer from the of fsite power system.
i on the loss of offsite power each of the redundant engineerud safety features distri-bution systems will receive power from a completely independent diesel generator f
unit. Each distribution system will include 4160, 400 and 120 Volt load centers to provide power to the various safety loads. Each of the redundant load groups will
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consist of the complement of safety equipment needed to achieve safe plant shutdown and/or to mitigate the consequences of a design basis accident.
The two emergency diesel generators to be used in the proposed SWE55AR P1/RESAR-35 design combination will be selected by the utility applicant referencing this design (see also Section 1.2. 9.5.2 and 9.5.3 of this report). A continuous rating of 6000
{
kilowatts for each diesel generator is specified in SWESSAR-p1 based on the currently established load demand. However, as the setailed design of the SWE55AR-P1 plant 3
progresses, the demand and hence the diese1' generator rating may be revised. The final sizing of the diesel generators will be based on a continuous rating that will be consistent with the recommendations of Regulatory Guide 1.9.
For those diesel generator units which have not been previously 1tcensed for a nuclear power plant cop 11 cation, the following prototype quellfication tests will be performed:
(1) At least tuo full load and maagin tests on each diesel generator unit to denen-Strate the start and load capability of the unit with some mergin in excess of the design raquirements.,
(2) At least 300 valid start and load tests prior to initial fuel loading, with the failure rate not exceeding one per hundreo. A valid start and load test normily consists of a start from cold ambient condittoas with sequential loading of the generator to at least 50 percent of its continuous rating within the required time intervs) and continued operation until equilibrium temperature is attained.
(3) Break-in runs on each unit for the length of *.ime required for passing through the initial failure period of the unit.
(4) Other onsite ;ests as detailed in IEEE 5td 387-1972.
- Criteria 'for ofesel Genera-tor Units Applied as Standby Power Supplies for Nuclear power Generating $tations."
t Each diesel generator unit will be housed in a separate seismic Category I room with an inda,erdent ventilation systcm. Each diesel engine will have redundant, independent 83
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1 a?r starting systans, indepene.L air intake and exhaust systens, and independent and separate seismic Category I fuel oli storage ar.d transfer systems (see also Sections i
9.4.1. 9.5.2 and 9.5.3 of this report). The storage capacity of the system will pro-vide sufficient fuel oil for each diesel generator system to operate continuously at a maalaum rated load for seven days.
During our review of the SWESSAR.P1/RESAR-35 design coe61 nation, we requested Stone &
'l Webster to provide additional information to demonstrate how the two 4.16 kilovolt i
emergency buses are isolated from the spare emergency transformer to which they could I
botn be connected. Our concern was the potential loss of both emergency buses due to r
a single failure. In response to our concern. Stone & Webster modified the design I
such that the circuit breakers connecting the two emergency buses wl13 ret be installed during normal plant operations. This configuration prevents inavvertent connection of the buses or tying the buses to a common transformer. We have reviewed this l
- ~
modified design and have concluded that it meets the requirements of Criterion 17 of the General Design Criteria. IEEE Std 308 1971 and.the recoemendations of Regulatory j
Culdes 1.6 and 1.75 and is. therefore, acceptable. We will review the provisions for f
ensuring that these circuit breakers can be removed during our review of a construc-tion peref t application submitted by a utility applicant referencing the $WESSAR.P1/
RESAR-35 design combination.
+
We have reviewed the design description. design criteria. design bases. logic diagrams for the alternating current onsite power system and the analysis provided by Stone &
Webster regarding the adequacy of these criteria and bases. The proposed design of the alternafing current onsite power system confoms with the requirements of Criteria 17 and 18 of the General Design Criteria. IEEE $td 308 1971 and the reconmendations of Regulatory Guides 1.6 and 1.9.
We conclude that the proposed design of ths onsite alternating current power system is compatible with the RESAR 3$ requirements. that the system will meet our requirements identifled in Section 8.1 of this report and.
therefore, is acceptable.
i 8.3.1.1 Electrical Protective Trios for Enoineered $afet F
t eatures $vstems and touf ment furing our review of the $WESSAR P1 design we requested Stone & Webster to provide additions) informatten for the electrical protective devices with regard to the potentia! ef these devices to spuriously trip out engineered safety features systems or equipantat at a time when they are required to mitigate the effects of an accident o
in the plant. We also stated our position that all electrical protective trips for engineered safety features systems and equipment. which remain operative for accident conditions shall meet the following:
(1) All thermi, everload protective trips retained for accident conditions shall be tested every 3 months.
(2) All other trips retained for accident conditions shall be tested every year.
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7 The objectives of these periodic tests are to verify the trip set point to ascertain the trip setpoint drift. if arty, and to establish the repeatability of the trip at its set veius.
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Stone & Webster has provided additional information in SWESSAR-P1 which states that the protective trios utilised for the diesel generators. the 4160 Volt and 480 Volt load center rotors and ali other Class IE circuit protective devices that are retained
.l for accident conditions will be tested annually. The thermal overload trips provided
- f for the 460 Volt continuous and intermittent duty arters will be bypassed for the
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accident conditions and the bypass circuitry will be designed to meet the requirements i
l of IEEE Std 279 1971. We have reviewed the additional information and conclude that
.A the proposed periodic testing of Class IE protective devices satisfies our requirements and 15. therefore, acceptable.
8.3.1.2 Protection of Electrical Penetration Conductors J
l The SWE55AR.Pl design of the electrical penetration assem611es in the contairment structure util be in compliance with the roccamendations of Regulatory Guide 1.63.
Additional design information for the electrical penetration conductors, including a discussion of how the penetration circuits overloed protection will meet the criteria of IEEE Std 279-1971. will be submitted by a utility applicant referencing the SWESSAbP1 design in its appilcation for.the construction peralt. This is an inter.
l face requirement to be addressed by the utility applicant. We conclude that the j.
comitment by Stone & Webster in SWE55AR-P1 to meet these requirements is acceptable for a Preliminary Design Approval for SWE55AR.P1.
r 8.3.1.3 Class TE Underornund Cables The $WE55AR.P1 balance-of-plant design will include underground cable systems from t'a control building to the service water ptswps and to the cooling tower fans. At our request. Stone & Webster has provided additional information in SWE55AR.P1 on the Class IE underground cable system, including the following:
( j) The criteria for Class IE underground cable systou will be the same as those used for Class IE cable systems within the plant.
i I
(2) Redundant cables for Class !E systems will be enclosed in separate reinforced j
concrete duct banks.
t (3) The duct banks will be missile protected.
(4) Fire stops for cables in underground duct banks and provisions for precluding the accumulation of water withir. the duct banks will be provided.
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We lieve reviewed the additfor.41 information and conclude that the criteria for the "j
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design of Class IE underground cable systems meet the requirements of Criteria e, 2
- 3. 4 and 17 of the General Design Criteria and of IEEE Std 308 1971 Section 5.2.1
.j and, therefore, are acceptable.
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- 8. 3,1. 4 Fire Stoos and Seals for Cable Systems l
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The fire detection and protection system for the 5WE55AA-P1 design is addressed in
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5ection 9.5.1 of this repo t.
The design includes fire stops and seals to control fires in electrical cable systaus and to assure that a fire in one system will not w
propagate to another redundant system. Stone & Webster has provided design provisions and criteria for fire stops and seals in cable systes. The design criteria for the
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fire stops include the followirg:
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4 (1) The fire rating of the material shall be consistent with the fire rating require-s monts of the penetrated wall, floor, or ceiling.
(2) Sultability to penetration geometry and arrangewnt.
(3) Compatibility with cable and insulation materials.
(4) Ability to withstand mastman required pressure on either side of the penetration.
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The installation procedure is depe: dent on the type of material to be used.~ The type
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of material selected and the installa'fon procedure will be included in the construc-tion permit application by a utf *h applicant referencing the SWE55AR.P1 design.
l Transition points at which fire stops will be installed have been identified. Addf-tional fire stops and seals will be provided where necessary in specific plant designs.
Their location wl11 be determined in the application for a construction permit by a utility applicant referencing 5WE55AR.P1. These matters are defined as design inter-faces that must be addressed by a utility app 1tcant. The following information, therefore, will be required to be included in the construction permit application by a utility applicant referencing the SWE55'R+pi designs (1) Fire stop materials, flamability, and fire rating.
(2) nuality assurance program and test procedures used to verify that penetration fire stops and seals have been properly installed.
(3) Qualification testing of tne fire stops and seals to demonstrate adequacy over f
the lif* of the plant.
i (4) Administrative procedures and controls that will be followed when it becomes necessary to t, reach a comleted fire stop or seal for the addition or rencval of Cables.
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and seals.
- J i5 We will review and evaluate the details of the above referred information in the l
l construction permit app 1tcation by a utility applicant referencing the $WE55AR-P1
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dtsign. We sanclude that the criterie with regsrd to cable fire stops and seals provide adequate bases for the Preliminary Design Approval of the SWESSAR-P1 design.
8.3.2 Direct Current Power $rstes
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The onsite direct current power system of the $WES$AR-P1 balance.cf plant design will
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consist of four redundant and indeoendent 125 Volt direct current supplies, each
,j consisting of a bettery with its own charger and direct current bus. Standby chargers I
will be provided to backup eacN of the main chargers and supply the 125 Volt direct current power requirements at all times. Including maintenance periods for the main l
chargers. The independence of redundant direct current systems will be maintained by 14 housing the redundant system components in separate rooms in the seismic Category I
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control building. Each battery room will be provided with seoarate ventilation f}
systems.
The ampere hour capacity of each 125 Volt direct current battery will be suitable for supplying all safety related loads for a sinimus of two hours without the use of the l
battery charger. Non-safety related loads will not be shared with the Class lE 125 Volt direct current system.
The proposed 125 Volt direct current power system is in conformance with Criteria 17 and 18 of the General Design Criteria, with the requirements of IEEE Std 308-1971 f
and with the reconnendations of Regulatory Guides 1.6 and 1.32.
Four redundant 120 Volt alternating current vital bus systems will be provided to lI supply power to plant protection system instrianentation and related circuits. Each vital bus will be fed from an independent static inverter which in turn normally will be fed through a static battery charger free a 480 Volt emergency bus. Sheuld the normal power source fall, the static inverter automatically will be powered from its associated battery. The 120 Volt alternating curre.nt vital bus systems will be designed in accordance with IEEE Std 308-1971.
i We have reviewed the design description, design criteria, design bases and single j
line diagrams ice the direct current onsite power system and the 120 Volt alternating
(
current vital bus system and tne analysis regarding the adequacy of these criteria and bases. We have concluced that the proposed design for the direct current onsite I
power and 120 Ys.It alternating current vital bus system for the SWESSAR-PVRESAR-35 design coebination meets our requirements identified in Section 8.1 of this report i
and, tharefree, is acceptable, t
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I interface Requirements for Electric Power Systees 8.4
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i Stone & Webster nas included in SWES$AR P1 the electricai system interface rcquire.
ments for electric power systems in the SWE55AA P1/RESAA 35 design costination. This i
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information is presented in Table 8.41 of SWES$AR Pl. The interf ace design criteria I
f for electric power systems is presented in Table 8.4 2 of SWES$AR P1 as an interface design criteria applicability astrix for SWE55AR P1/RESAR-35 electric power systems.
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i We have evaluated this information er.d conclude that the interface criteria for the electric power system for SWESSAR-P1/RESAA-35 design preser.ted in Tables 8.4-1 and 8.4-2 of SWE$$AR P1 are acceptable.
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9.0 AUXfLIAAY SYSTDt5 i
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i ine auxiliary systems within the scope of SE55tR-P1 that are necessary to assure i
safe reactor operation or shutoown will inc1wie the service eter system (except I
intake structure and chemistry control) unsponent cooling inter systs, ultimate f
heat sink (service noter systa requirments only), portions of the chemical and l
volme control syste, ventilation and air-conditicMag systems that are safety related, fire protection system, diesel generafw fuel til storsge aM transfer I
system, and the diesel generator auxillary syste. We have reviemri dose systes to j
determine their conformance to the applicable remirements of the General Design l
Criteria and Regulatory Guides. Discussfon of the designs of these systems are
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provided in the following sections snd include the safety related objectives of the systems and the manner in which tnese objectives will tw achieved. In additf5n we reviewed these systems with respect to the interface requirements established in RC5AA-35.
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The systems necessary to assure safe handling of fuel and adequate cooling of the spent fuel will include new and spent fuel storage fac!11 ties, the fr1 pool cooling and purification systam, the fuel handling facilities and a portion of the fuel j
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handling building ventilation system.
We have also Mviewed those auallf ary systents or portions of the systens whose failure would not prevent safe shutdown but could, either directly or indtrectly, be The condensate a potential source of a radiological release to tha envircement.
stcrage facility, the dominere11 ed water makeup system, potable and sanitary water system. primary grade water system, chilled water system. cir-condit(cning chilled water system. compressed air system, equipment and floor draltage system. boron recovery system, and the equipment vent system are additional reactor au.sfliary sys-tems that are not safety related and will not be designed to seismic Category !
In our review, we determined that (1) seismic Category 1 isolation l
requirements.
valves will be provided to physically separate the non. essential portfor.3 from the essential system or component at system interfaces or connecticns to a seismic Cate-gory I system or component, and (2) the failure of systems or cortions of a system pot designed to seismic Category I requirements will not prec1Je the operation of
- l On this basis we safety related systems or components located in close proxinity.
have concluded that these systens are acceptable. The water treatment system the circulatory water system, and the potable and sanitary water system are related to a specific site and are.iot within the scope of SW$$AR-Pl. Pese systems will be described by the utility applicant referencing tne SESSAR.P1 design.
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i g.1 fuel Storaos and Handifne The fust butiding will house the new fuel area, spent fuel oort. shipping cask area.
3 fuel pool cooling and cleanup system, and fuel handling systm. The location of the i
fuel building is show in Figure 11 of this report. The port!an of the fuel building thst houses safety related components will be located on the annulus building founda.
tion met and will be designed to seismic Category I requirements. The entire building i
i will be tornado protected.
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9.1.1 New Fuel $torane_
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The new Twel storage area in the fuel lullding will provide the capability for dry storage of one-third ef a RESAA $ core. The new fuel storage area. located in the j
annulus portion of the fuel building, will be designed to seismic Category I require-The new ments. The storage racks for the r.sw fuel arn within the scope of RESAR.3$.
fuel storage facility will be designed to accommodate the fuel racks such that a safe degree of suberiticality can be maintained for the condition of optimum moderation.
Based on our review, we conclude that the design criteria ai I bases for the new fuel storage facilities supplied within the scope of SWE$$AR-P1 meet the requirements of p
Criterion 62 of the General. Design Criteria and the appropriate guidelines of Regula.
tery Guide 1.13 including the positions on seisele design and missile protection.
ar$ also meet the RESAR-35 interface requirements and are, therefore, acceptable.
9.1.2
$oent Fuel $toran t
i Spant fuel'will be stored underwater in the spent fuel storage pool located in the j
annulus portion of the fuel building. The pool will provide storage capacity for a l
mintam of one and one third RESAR 35 spent cores. The spent fuel storage pool will be design 6d to seismic Category I requirerants. The storage racks for the spent fuel j
are within the scope of RESAR 3$.. The spent fuel storage facility will be designed I
to prevent the storaga racks from being scunted in any other than their prescribed locations, and a safe degree of suberiticality can be maintained. for the condition of cptime moderation. The fuel pool will be of reinforced 6oncrete construction and will be provided with a stainless steel ifner. The embedmerits for the spent fuel r
storage racks will be designed to withstand the splif t force of the spent fuel pool bridge hoist. The facility will be designed to prevent the cask handling crane from traveling over, or in the vicinity of the pool, thereby precluding damage to the stored spent fuel in the event of a drc; ped cask (see Section 9.1.4 of this report).
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Based on our review. we conclude that the design eriteria and bases for the spent t
fuel storage facilities are in confornance with the requirements of Criterion 62 of the General Design Criteria and the recoroiendations of Regulatory Guide 1.13, includ-ing the recormendations on selsnic design, missile pectection, design compatibility
- f with th-* handling o' the fuel cask in tre spent fuel pool area, and also meet the 1l RESAR.33 interface requirements and are, tnerefc*e. acceptable.
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t 9.1g Soont Fuel Coo 11ne end Cleanue System The spent fuel cooling and cleanup system will be designed to saintain the water '
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quality and clarity of the pool unter and to remove the decay heat generated by the stored spent fuel afsamblies.
s The cooling systen will be designed to seismic Category I requirements and will consist of taso 50 percent capacity trains, each including a pump, heat exchanger, 2
associated piping, valves and instrumentation. The capability to supf.y makeup F
water to the pool will be provided by permanently installed connections from the t
primary grade unter system and the refueling water storage tank. In addition tuo q
trains of the ssrvice water system (see Section 9.2.1 of this report) will be con-nected to the fuel pool cooling system via a spool piece in each train and, therefore.
will provide an assurus seismic Category I makeup water supply. The speci piece will be stored along-side iu point of insertion. The expected time requi ed for its insertion will be considerably less than the pool hat-up rata to bothng in the avent that im1 pool cooling is lost.
Stone & Webster has calculated the fuel Mol temperatures resulting from the st rage of one-third of a RESAR 35 core and of one and one-third RESAR-35 cores using the heat loads listed in RESAR-35. The rssults are listed in Table 9 1 for the condi-tions of one and two-train operation of the spent fuel pool cooling system. We I. ave i
independently evaluated i,he heat loads and inter temperatures and find that the fuel pool water temperatures listed in Table 9-1 can be maintained for the storage condi.
l 113ns described in Table 9-1 and are, therefore, acceptable.
The spent fuel cooling system is designed to withstand the effects of a single i
active failure. No failure wiP result in the loss of all makeup water during expected storage conditions.
The cleanup system is not a safety reested system anc will not be designed to seismic Category I requirements, The pump, piping and valves of the system will be physically.
Independent from the essential seismic Category I spent fuel pool cooling sys' tem and cross connections are not provided. In addition, a failure of this system will act adversely affect sny safcty related equipment.
Based on our review, w conclude that the criteria and' bases for the design of the spent fuel pool cooling and cleant/ system are ir, confomance with ihe ra;omendations of Re*julatory Guide 1.13, ' Slut'irg the recomendations on seismic design, mi site
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f protection. and availab.;ity of tne assured makeup systems, and the requirements of Criterion 62 of the General Cesign Criteria; the design also Mets the RISAR 33 system interface requirements. We, therefore, find the spent fuel cooling.nd cleanup system acceptable.
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TABLE 9-1 i
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1 FUEL POOL C00(ING SYSTEM PERFORMANCE _
OF ShTSSAR-P1/RESAR-35 DESIGM_
l 1/3 Core 1 1/3 Cores Storage Storage t
6 6
14.0 x 10 43.5 a 10
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Design Heat Load (Stu/hr)
Maxisam Pool Tamperature ('F) 133 192 1 train operation f
120 148 l
f 2 train operation I
f Abbreviations used in table:
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n, The fuel handling system will provide the means of transporting and handling fuel from the time it reaches the plant th an untrradiated condition until it leaves the plant after it has been removed from the reactor. Mejor portions of the fuel han-dling systan. including the components required for transferring fuel free the reactor to the spent fuel pool, are withia the scope of the RE5Ab35 design. The equipment within the scope of SWESSAR-P1 are the fuel twiloing crane with auxiliary hoist, decontamination equipment, and a spent fuel cask lifting beem.
The spent fuel cask loading aree will be located adjacent to th) spent 51 pool.
The area will be separated from the fuel pool by reinforced concrete wils. Unac-cantable damage to stored fuel due to a spent fuel cask drop will be prevented by limiting the travel of the spent fuel cask to en area which contains no safety related equipment or stored fuel. The travel of the sask bridge trane will be limited by mechanical ste,. and 11elt switches. Furthermore, if the spent fuel cask were dropped, it would be prevented froe rolling or toppling into the spent futi pool by physical separation between the spent fuel pool and cask loading pool. Our evalud tion of the SWESSAR.P1 polar crane inside the containment to be used for th2 removal of the RESAR-35 reacter vessel heti assembly is presented in Section 5.3 of this l
report.
Based on our review, we conclude that the fuel 'iandling system design r*tteria and bases are in conformance with the recomerations of Regula'.ory Guide 1.13, including the recommendation regarding protection of the spent fuel storage facility from the impact of unacceptable heavy loads carried oy overhtad cranes. h.. therefore, find the design of the systen acceptable.
9.2
' Witer systems 9.2.1 Service Water System The service water system will provide r )
.ater to ttt tc y related plant systems for nomat operation, cold shu for the ueve.ition and mityatun of postulated accidents. The saevice wace.
..em pumps will te located in t,e titate structure which 1. part of the ultimate heat sink, dfscussed in Etir ' 9.2.1 of t.ais report. The service water system will be designed to stis,ite ra a gry 1 reouirements.
The service water system for the 5WESSAR-F1/pESAb35 design w'11 censist of two phyr.ically separated safety trains with two se='vice water psips in each train, iach
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train will serve tw coepownt cooling wat.r heat exchangers, a diesel generatcr l
cooler, an auxiliary feecwater system m.kup line, a unit cooler for tae supplementary 1eet collection and release system area, a unit ceoler for the chargt'ig pump and engineered safety feature area, ar.d a control butiding water chiller. Both nf the trains wil? also saply rakeup water to t e fuel pool cooling system as descMbed in Sec'. ion 9.1.3 of this report. Addittor. ally, a third control butiding water chiller 95
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and a third unit cooler for the charging pap area are cross connected to hoth service In the event a control building water chiller, a unit cooler for I
water system trains.
the charging pop area or, a service water system train is out of service the reduMont component, i.e., the third control building mater chiller and the third unit cooler for the charging pwp area. Can be supplied with cooling mater fran the service wates sys.
1 trcia with the inoperable component or by tre unaffected service wter systen train.
tem The independence between the safety related serv'ce meter system trains is maintainedj E nh of the trains will be by two sets of locked closed manual valves in series.
powered from an indef,endent engineered safety featufts bus.
J During anticipated modes of operation only one pump in each of the two safety trains Normal l
is required to operate to provide cooling to the reactor aufliary componerts.
shutdown operations will fvquire both safety trains. Howewr, fr. the event of a single failure in one of the operating trains, a safe shutdown condition can be achieved
+
The time required to achieve a cold shutdown condition woulf with the remaining train.
During a be extended over a longer period of time fAan if both trains were available.
Toss of-coo 14't accident one train will always be available to provide cooling to emergency core cooling system cinponents in the event of a single fatture.' The heat loads of the service water system are based on the requirements of RESAR-35 and
$WE$$AR-P1 systems that are associated with the conditions of nomal operation, shut.
down, design basis accident and loss of off site power.
The systen imposes some interface requirements on the intake structure of the system i
which is not within the scope of SWESSAR P1 and which will be discussed in the safety!
The SWES$4P1 design requires the intake i
analysis report of a utility app!! cant.
structure to be a seismic Category I structure. to be tornado missile protected, and to f
be protected from the design basis flood. In addition, each service water pwp is to i
The separrting be located in a separate compartment within this intake structure.
walls between compartments are to be internal missile proof and the pap casings arc to I
The chasistry control of the be located in a sump to assure adequate s@ergence.
service water system is also site related aM will be discussed in the safety analysis report of a utility applicant.
I Based on our review, we conclude that the service water systes design criteria and j
bases are in conformance with the requirements of Criterion 44 of the General Design I
Criteria regarding the ability to transfer neat from safety related conponents to the i
i They are also in con.
ultimate heat sink and regarding the single failure criterion.
formanc2 with the requirements of Cr.teria 45 and 46 of the General Design Criteria regarding the system design for periodic tests and inspections, including functional j
The service water tyste1 testing and confimation of heat transfer rapabilities.
We conclude design also meets the RESAR-3$ interface rfquirenents for this system.
3 i
that the system is acceptable.
9,2.2 C y pwent Cooling Water System inc component cooling water system for t% reactor plant will be a closed cooling water 4
system which will transfer heat to the sea nice water system frcrs coraporents unich 9-6 I
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1 contain radioactive fluids. The system will be designed to function during nomal I
plant operation, cold shuthwn, and postulated accident conditiets. The component l
cooling water system for the SWESSAR P1/Ri,$AR 31 design will consist of tte physically separated and redundant safety trains. Each train will contain two compenent cooling mater pumps and two heat exchangers to transfer heat from the c:ssponent cooling water system to the service water system. Each train will provide cooling for one residual j
heat removal exchanger and pep cooler, one safety injection pwsp cooler, one centri.
fugal charging pep cooler one contairment spray pump cooler, and one containment atmosphere recirculation cooler. In addition. the cooling loads'from other components such as the reactor coolant pops, hydrogen recombiner cooler, excess letdown heat exchanger, and fuel pool coolers, will also be removed by the component cooling water
]
system.
The safety related portion of the system will be oesigned to seismic Category I tequire-ments. The non-seismic portion of the system will be remotely isolated in the event of
[
a malfunction. Makeup will be provided by the deminerallred makeuo water system. Each I
component cooling water train will be powered by an independent engineered safety C
features bus.
During all anticipated modes of operation, only one pump in each of the two sat 0ty I
trains will be required to provide sufficient cooling to the auxiliary components.
l Normal shutdown operations will utilite both safety trains. However, in the event of l
a single failure in one of the trains, a safe 'hutdown condition can be achieved with the remaining train. The time required to achieve a cold shutdown condition would be extended over a longer period of time than if both trains were available. During a I
]
loss of-coolant accident, only one train is required and will be available to provide adequate cooling to the components of the emergency core cooling system. The cornponent cooling water system heat loads for the conditions of normal operations, shutdown.
design basis accident, and loss of offsite power, as identified in RESAR 31. are l
included in the total heat load of the component cooling water system. Component cooling water will be sus,, lied at 105 degrees Fahrenheit during normal operation and a
below 120 degrees Fahrenheit during accident and shutdown conditions as required by l
RELAR 35.
j The proposed design of the component Cooling water system will provide two $Jpply and two return lines, one each for two of the four RESAR 35 reactor ccolant pumps.
l Each of the lines wl11 contain one motor-operated valve for con'ainrent isolation.
k The rotors, seals, and bearings of the RESAR-35 reactor coolant puros reoutre con.
tinuous cooling, provided by the SWE55AR-P1 cor'ponent cooiing water system during A
normal operation, anticipated transients, and following postulated accidents, so that g
the safety function of the pumps will not be precluded. Inadvertent failure or closure of any one of the above motor-operated valves would terrinate the coolent flow to two of the pumps, thus potentially leading to fuel damage durirg an antic-ipatera transient, due to a multi-pump locked rotor incident. We, therefore, reqaired 4
97
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I Sto.4 & Webster to modify this portion of the component cooling inter system so that the following criteria are eet g
i (1) A moderste energy leakage crack or a single failure in the component cooling mater system shall not result in f,al damage or damage to the reactor coolant l
System pressure boundary caused by an extended loss of cooling to the reactor coolant pumps. Single failure includes operator error, spurious actuation of motor-operated valves, and loss of component cooling water peps. Moderate leakage cracks should be detemined in accordance with the guidelines of Branch Technical Position APC$B 31 ' Protection Against Postulated failures in a Fluid System Outside Containment."
1 (2) An accident that is inittated from a failure in the component cooling water piping shall not result in excessive fuel damage or a breech of the reactor coolant system pressure boundary when an exter.ded loss of cooling to the reactor coolant pwps occurs. A single active failure shall be considered when evaluating j
the consequences of this accident.
j Stone & Webster provided, at our request, safety grade instrumentation to detect the loss of component cooling water to the reactor coolant pumps and to alarm the operator l
in the control room. The entire it.strumentation system, including audible and visual l
Status indicutors for loss of component cooling water, will meet the requirements of l
IEEE Std. 279-1971. To seet the criteria above. Stone & Webster has identified as an l
interface requirement, to be met by the utility applicant referencing the SWE$$AR.P1/
RESAR 35 design., that the reactor coolant pumps shall ba demonstrated to be capab'le to j
operate with loss of cooling for longer than 20 minutes without loss of function l
and the need for operator corrective action. The safety grade instrumentation f
discussed above will enable the operator to initiate protective action for the plant in sufficient time.
Alternately. If it cannot be demonstrated that the reactor coolant pumps will operate longer than 20 minutes without loss of function or reactor operator corrective action.
then the utility applicant shall meet either of the following two SWES$AR-P1 interface l
requirements; (1) Safety grade instrumentation consistent with the criteria for the protection f
l system shall be provided to initiate automatic protection of the plant.
]
l (2) The component cooling water supply to the pumps shall be capable of withstanding a single active failure or a moderate energy line leakage crack as defined in I
j our Branch Technical Position APC$B 3 1.
hsed or our review, we conclude that the component cooling water system design criteria and bases are in confomance with the requirements of Criterion 44 of the (eneral Design Criteria regarding the abilf ty to transfer heat from safety related components to the ultimate heat sink under normal and accident conditions and to meet 9-8 L
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the single failn criterion. We further, conclude that the system design criteria and bases meet the requirements of Criteria 45 and 46 of the feneral Design Criteria regarding system design for periodic inspections and tests, including functional j
testing and confirmation of heat transfer capabilities. The design criteria and bases for the SE55AR-P1/RESAA-35 component cooling uter systen also wet the RISAR-35 requirements for the system. We conclude that the system is acceptatls.
9.2.3 Ultimate Her.t sink The citimate heat sink is related to a specific site and therefore is not within thf.
scope of SWESSAR Pl. (see also Section 2.4.2 of this report). Stone & Webster has
}
specified, as an interface requirement, that the ultimate heat sir:k must be capable of removing the heat loads from the $WESSAR-p1 service water system as discussed in Section 9.2.1 of this report for all modes of operation. Stone & Webster has specified a service water systee inlet temperature at the ultimate heat sink of 95 degrees Fahrenheit for normal modes of operation, and a maximun inlet temperatu t of 100 degrees Fahrenheit following a design basis accident. The infomation and the evaluation of the ultimate heat sink, including the considerations of low water conditions, will be provided by a utility applicant referencing the SESSAR-P1 design in its construction permit application. We will evaluate this infomation to determiae that the ultimate heat sink will be capable of removing the specified heat loads for all modes of plant operation.
9.3 Process Auxiliaries Our evaluation of the boren recovery system is discussed in Appendix A of this report.
l 9.4 Air-Conditioning. Heatino. Coo 11no and lentilation Systems 9.4.1 Control Building ventilation systems The control building will house the control room, the emergency switchgear rooms, the cable spreading rooms, the battery rooms, the relay rooms, and the equipment rooms for the heating, ventilation, and air-conditieing systems for the control building.
These systems will serve these rooms and the electrical cable tusels connected to the control building. Each of the diesel generator rooms, located adjacent to the control building, will be provided with an independent ventilation system. The function of the control building ventilation systems 1: to maintain these areas within the therinal and air quality limits required fo. operation of plant controis and uninterrupted safe occupancy of required manned areas during normal operation, shutdown and post accident conditions.
The control bailding ventilation systems include separate air conditioning systems for the control room, the control building refrigeration equipment rooms, and the emergency switchgear area. In addition, other auxiliary ventilation systens are provided for equipment important to safety. The control building ventilation systems are evaluated in the following paragraphs.
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9.'4.1.1 control Room Air-conditionina System The control room air coeditionf.4 system will consist of two redurvant trains eMh powered from a separate engineered safety features bus. Normally one train will be operating and the other will be on standby. The physical separation criteria between j
i t
redundant safety related trains for protection against dynamic effects associated,
with postulated rupture of piping are discussed in section 3.6 of this report. The system will be designed to neet seicaic Category I requirements. The control room'
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air-conditioning system will be designed to maintain the control room under sito>tiy
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l positive pressure during normal operations so that outlaakage can be maintained.
?
Two 100 percent capacity seismic Category I control building air-conditioning chilled water systems will remove the heat from the cooling coils of the air-conditioning units so that the control room environment can be maintained at a temperature of 75 degrees Fahrenheit and at a relative humidity below 60 percent. A standby water chiller can supply chilled water to either system train. The service water system will remove the heat from the chill;d water systems.
3 l
During a design basis accid)nt or if the presenci of outdoor smoke :hlorine. or f
other toxic gases is detected, the air intake :nd the exhaust air isolation valves are closed and the control rra air is recirculated through the charcoal filter banks. The control room will be pressurited to a minimum pressure of 0.25 inches of water by the control room pressurization system to prevent inloakage of contaminated air. The control room habitability under these conditions is evaluated in Section 6.4 of this report.
The control room pressurtration s; stem includes two 100 percent capacity seismic j
Category 1 air supply trains with separate emergency outside air intakes. The intakes will be physically separated by a distance of 1.000 to 1.400 feet and will be l
located 180 degrees apart. The exah loeition and description of these remote erergency air intakes are site relattd and this inferration will be proviced in the j
l safety analysis report of a utility applicant. however interface requirements are I
established in $WESSAR-PI. The air intakes are to be designed as seismic Category I structures. to be tornado missile protected, and if necessary, to be protected from the design basis flood. The intakes are to te located remote to the location of toxic gases or any flammable ilquids and gases potentially to be stored on or near the site. The intakes are to be located 90 degrees away " rom the direction of the prevailing winds and they are to be positioned such that potential building wake l
I effects are avoided.
l Based on our evaluation. we have determined that the design of the control building air conditioning and ventilation systems contains sufficient component redundancy and physical separation to meet the single fatture criterion so that dir-conditioning and wentilation elll be assured as required for anticipated eperating conditions.
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Based on our review, we cont,1ude that the system design criteria and bases are in 0
1 conformance with the requirements of Criterion 19 of the General Design Criteria I
j regarding the capability to operate the plant from the control room during normal an f
[i accident conditions. and that appropriate interface requirements are identified in We, therefore.
I k
SWE55AR-P) which are to be iglemented by a utility applicant.
.h conclude that the system is acceptable.
'k Refriceration Ecuipment Room Air-conditionino systen_
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The air-conditicting system for the refrigeration equipment rooms in the control j
h building will consist of separate 100 percent capacity air-conditioning trains. one j
The entire ld each for the three control butiding refrigeration equipment rooms.
i system will be destptd to seismic Category I requirements and each train of the f
system will be powered from the same engineered safety featur-s bus as the refriger i i tion equipment 1*. the room it air-conditions. The loss of any one air-condit on ng
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dd system will not prevent the equipment in the other rooms from performing their int
(
The control building air-conditioning chilled wter systes will j
ll be safety function.
j remove heat free the air-conditioning units so that the equipment rooms wi maintained at a tagerature of 80 degrees F#hrenheit.
Based on the review of the design criteria and bases for the system, we conclude th h
the system (sactional performan,e meets the requirenents of Criterion 19 of t e General Design Criteria. and therefore, is acceptable.
Emereenev $witchgear Area Air Conditionino $vstem_
9.4.1.3 The air conditioning system for the entrgency switchgear area will provide cooled itchgear to the emergency switchgea? rooms, safety battery rooms, and the eme air-conditioning equipment rooms.
seismic Category I trains, eacf *o be powered from a separate engineered safety The al. will be pW 4ed elirectly to all equipment rooms with the The air for the safety battery rooms will be taken features bus.
exception of the battery rooms.
fran their adjacynt torridor by the battery room exhaust fans which will draw the cooled air into tne rooms.
the The control building air conditioning chilled water system will remove heat from cooling coils of the units so that all equipment rooms will be maintained at an f
f ambient temperature of 80 degrees Fahrenheit which is below the design temperatu the equipmeat.
Based on our evaluation, we hcve determined that the design of the air-condi*io fficient and ventilation systems for the emergency switchgear area will contain su iterion so component redundancy and physical separation to meet the $1ngle failure cr I
ting
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that air-conditioning and ventilation will be assured during anticipated opera Based on the review of the design ceittria and bases for the systems i
conditions.
functional performance, we conclude that the systems are acteptable.
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9.4.1.4 other Auxiliary Ventilation Systant j
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Other auxiliary ventilation systems designed to provide a suitable envirormental condition for equipment isportant to safety include the cable spreading room wectila-i tion system, electrical tunnel ventilation system, and the diesel generator room ventilation systan. Syf, tan redundancy is provided so that the systems or components j
l cooled by the ventilation systems are capable of performing their safety function r
i considering a single active failure.
i The cable spreading room ventilation system ullt be powered from a separate engineered safety features bus. The diesel generator room ventilation system will be powred from the esergency diesel generator in the room it cools.
I i
Tt.e cable spreading room and the electrical tunnel ventilation system will be da'aaned to ruir.tain an aselent tenperature of 110 degrees Fahrenntit, while the diesel genera-tor room ventilation system will maintain a esalmum ambient temperature of 1% oegrees Fahrenheit, which are below the design rated temperatures of the equipment Based on our review, we conclude that the dssign criteria and bases for the ventilation systems functional performance are acceptable.
9.4.2 Fuel Building vent 11ation System The fuel building ventilation system will be designed to maintain the fuel building atmosphere within acceptable temperature and humidity limits for personriel and equipsumt. to maintain the building at a negative pressure, and to mitigate the con-sequences of a fuel handling accident by filtration of the exhaust air. The system consists of a normal supply and exhaust system; for energency conditions the supple-mentary leak collection and release system is used (see also Section 6.5.2 of this report).
The exhaust from the fuel handling area during nomal operation will be discharged either through the station vent by the normal exhaust system without filtration or will be filtered through the *nnuluf building air exhaust system. A slight negative pressure will be maintained in the fuel building. The supplementary leak collection and release system as the fuel building emergency ventilatico system, will be j
designed to mitigate the consequences of the fuel handling accident. This system will te designed to meismic Category I requirements. In the event of a fuel hand 11r.g accident, a high radiatior, signal from the radiahon monitors in the exhaust air duct will automatically acutate the system. Pbtor operited dampers in the normal ventila-i f
tion system will direct contaminated exhaust through the redundant charcoal filter t>anks in the supplewntary leak collectica and release system prior to discharge to the atrosphere through the station vent. A negative pressure of 0.25 inches water gauge will be reacted in approximately 35 seconds. Based on nur independ, tnt evalua-tion of the systen design we find that following a fuel handling accident, the supplementary leak collection and release system will start before excessive airborne radioactive materials will be released.
9-12 4
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Based on uur nview, we conclude that the design criteria and bases for the fuel building ventilation systas saet the reccamendations of Regulatory Guide 1.13, and, j
therefore, are acceptable.
i 9.4.3 Eneneered Safety Features Ventilation $rstem l
The ventilation and sir-conditioning systems for engineered safety festare and other essential equipment rooms will be designed to provide an cdequate supply of cooled
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air to safety related and emergency equipment that is required to remain operable during a design basis accident and to be capable of fuactioning during post-accident conditions. The rooms le; sing the emergency core cooling system peps and those for the camponent cooling water pumps, will be designed to be serviced by such systans.
Air-conditioning and ventilation systems for the engineered safety featums areas will be redunoant, seismic Category I systems, will be pcwered from the emergency l
Based on our buses, and will also be protected from the effects of toado missiles.
evaluation and the results of our failure sede and effwts valy;is, we hava deter-mined that the design of these safety room ventilatian and air-conditioning systems meets uur single failure criterion and that the required ventilation c9n be provided during anticipated emergency operating conditions.
{
Based on our review, we conchde that the design critaria and bases for the afr-conditioning and ve6titation systems of the engineere(safety features and other l
f essential equipment rooms are acceptable.
k 9.5 Other auxiliary $vstems 9.5.1 Fire Protection System Stone & Webster has incorporatsd design features throughout the SWE15AR-P1 plant design to prevent the occurrence of a fire within the plant. Design bases for fire protection have been 16 entitled with respect to the type, characteristics and loca-tion of a potential fire, with respect to tuilding and systems arrangements and structural features, and with respect to the requirements for seismic design, water supply and plant construction. For exa@le, to reduce the effects of a potential fire in the flant, flame retardant materials will be used throughout the plant.
Safety related components and recras will be protacted by walls ar.d floors with a three hour fire rating. Openings in the fire walls and floors will have fire doors The storage of combustible material in safety n1sted areas and in i
aM dampers.
ccnstruction artar will be controlled in accordance with the recomendations of I
Regulatory Guide 1.39. Warebses w111 be located remotely from safety rel4*.ed areas.
Fire protectio't considerations for ele <tMcal systeers are discussed in detail in These considerations include the physical separation of
. Section 8.3 uf this report.
engineered safety features systems ir.cluding the physical separation of the cabling for each engineered safety features trefn.
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a The exception to this is the cabling within the cable spreading roomi, where barriers
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I will be used to maintain the integrity between redundant Class IE control and instru-ment cables and spe:ial fire fighting systems will be provided. Class !E electric l
1 equipment will be separated fram non-class !E equipment. Cable insulation will be l
either flane-retardant or non. combustible. Fire stops of noncumbustible sealing material will be placed where cable runs pass through mells and floors. Separate metal conduits or enclosed raceways, enclosed cable troughs, or covered trays will also be used for the protection of cables.
The fire protection system will be designed in compliance with applicable codes and 1
standards, including those of the National Fire Protection Association and the Nuclear Enedy Liability and Proparty Insurance Association. The system will include a fire detection system and an automatic and manual fire extinguishing systen using water and carton dioxide.
?
The fire detection system will utilize fixed rate.cospensated, therwelly actuated f
detectors, and ionization smoke detectors. The thermally actuated detectors will be 1
l used for automatic system actuation while the ionization smoke detectors will annunct.
i ate. A fire detection annunciator located in the control room will alars upon actua.
l tion of a fire detector and indicate if any automatic extinguishing system has been
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actuated. Complete separaticn of detection systems and alarus sill be used for those
)
systems esployed to actuate extinga.1shing systems. Safety related areas will be I
provided with at least two detectors.
The following water systems will be used for fire protection of safety related and non-safety related areas: standpipe and hose systems, wet pipe sprinkler systems, pre-action sprinkler systems, and automatic and manual water spray systems. In safety related areas the water system will be manually operated and deluge type valves will be used to reintain dry lines to preclude adverse effects on the safety related equipment.
The water supply system, including the station yard distribution fire loops, for the
, fire protection system is not within the scope of the SWESSAR-p1 design. The system will be provided by a utility applicant and will be described in its safety analysis report. A requirement for a 600,000 gallon water storage tank is identified in SWESSAR-p1 as an interface requi % nant. The fire protection system will utilize one 2,500 gallons per minute motor.oriven pump and one 2,500 gallons per minute diesel.
- l engine driven pur@. The water storage and ptreing capacities are in accordance with l
Nuclear Energy Liability and Property Insurance Association requirements. In addi.
f tion to the yard hydrants, the water supply system will supply the standpipe and hose systems, sprinkler systems. and special hatard uster spray system.
Automatically and manually actuated total flooding type carbon dicxide systems will be provided for the electrfcal penetration areas and tunnels, cab' spreading rooms, l
emergency switchgear rooms and the diesel generator fuel oli storage tanks and ptrp house. Liquid carbon dioxide will be stored in tanks at a pressure of 300 pounds per i
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square inch-gauge ami at a temerature of O degrees FaWheit or less. Activation
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of any carbon disside subsyste will automatically clost all fire despers in the effected room to seal the area. To protect personnel woeting in these areas, safety f
lockout switches will be installed in the control circuitr to preclud. inadvertent system operation. A time delay between the sounding of aa alarm and system actuation j.
I will be provided to allow personnel to exit from the room.
tic The following safety related areas will be served by the fire protec*lon system:
diesel generatar rooms (automatic pre-action sprinklers), the diesel genernor fuel l
oil storage tank rooms (automatically actuated carbon dioxide system). the anrralus building (pre-action standpipes and hose system with remote manual delug? valve j
}
trip), and the cable spreading rooms, emergency switchgear rooms, and electrical I
penetration areas and tunnels (automatic or manual discharge of carbon dioxide). The operation of any of these systems will not affect the overall safety related function of systems served by the fire protection system since physi al separati e of redun-dont safety related componr** s will be maintained, i
The fire protection system will not be designed to seismic Category I requirements.
g The supports for the fire protection equipment ar.d piping located in safety related am as are seismically analyzed. Based on our review of the fire protection system.
i we conclude that safety related structures, systeras snd cogonents allt not be damaged by flooding deluge or any other adverse effect from a faOure of any part of the fire protection system. As a result of investicatiuns and tvaluations preset *y being conducted by the Commission's staff on fire protection systems, further requt e-ments may be imposed on the SWES$AR.pl design to further iqmve the capability of the fire proter. tion system to prevent unacceptable da%ge that may result from a i
fire.
Based on our review, we conclude that the criteria and bases for the fire protection system design are in conformance with Criterion 3 of the General Design Criteria regarding the design of safety related areas and systems to minimize the probability j
and effect of fire and regarding the design of fire fighting systems to assure that their rupture or inadvertent operation will not significantly igale the safety capability of structures and systems. Therefore, we conclude that the proposed fire protection system design criteria and bases are acceptable for a preliminary Cesign Approval.
9.5.2 Diesel. Generator F,el storace and Transfer system The diesel generator fuel storage and transfer system will be designed to provide suff 41ent storage of fuel oil to allow continuous operation for a minimum of seven days of each of the two eurgency diesel generators for the SWESSAR.P1/RESAR 35 design combination. Accordingly, ead d the diesel generators will have its own inde;endent fuel oli storage and transf:r u,um and auxiliary systems. Each system will incluje an 80,000-9411o4 fuel oli storage tank. The fuel storage and transfer system wili I
l The fuel 01) transfer purts sol be designed to meet seismic Category I requirecents.
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L be powered from separete emergency buses. There will be no cross connections in the j
systems between the storage tanks and the diesel fuel oli transfer pues.
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l The diesel generators will be housed it, separate rooms located in the seismic Cate-i gort I, tornado protected diesel generator building. Each diesel generator room is l
l 1solated from the adjacent room by a wall w!th a three-hour fire rating. The diesel f
~l generators will also be flood Protected against the maximum flood level established j
I for the SWEU4p1 design as discussed in Section 2.4 of this report. The diesel generator fuel oil storage and transfer system for each of the diesels will be housed l
in a separate adjacent vault. These vaults also will be designed to seismic Category I
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requirements and will be tornado missile end flood protected.
2 Based on our independent evaluation, we have determined that the integrated design of
(
4 the diesel generator fuel storage and transfer system wl11 seet the single failure criterion. Based on our review, we conclude that the diesel generator fuel oil storage and transfer system design criteria and bases provide adequate assurance that the systems will be designed to perform their designated safety functions, and are.
therefore, acceptable.
9.5.3 Diesel Generator Auxiliary Systems t
l The diesel generator auxiliary systems will consist of the diesel generator closed cooling 9ater system. the diesel generator air starting system, and the diesel genera-tor lubt f estic.. ":ystem. The diesel generator auxiliary systems will be an integral part of the diesel gencretor which is not within the scope of the $WESSAR-P1 design, but will be selected by a utility applicant and will be described in its application.
However, the SWESSAA-P1 design will be capable of providing cooling water and electric power to these diesel generator auxiliary systems where required. These capabilities have been identified as interface design information in $WE55AR-Pl.
9.5.4 Storaae of Compressed Gases in the SWESSAR P1 design nitrogen and hydrogen will be used in the reactor plant gas supply system. These gases will be stored under pressure in tanks that supply sepa-rate manifolds of the system. Stone & Webster h.s considered these containers as a potential missi' > Orurce since their failure could affect safety related systems or cogonents. Protw. son within the facility frun potertial missiles is based on that following: (1) relief valves will be provided on tinks and the set points will be below the design pressures of the tanks; (2) tanks will be located in limited eccess areast(3)tanksandcylindersutl1beanchoredsothattheywillnotbecomemissiles themselves following the failure of attached piping; and (4) the safe location of gas storage facilities in relation to equipment essential for initiating and maintaining a safe reacto* shutdown will preclude the possibility of interaction in the event cf an incident.
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are, therefore, accep M ie.
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10.0 $ TEAM AND POWER CONVER510N $Y$ TEM 10.1 Sumary Description i
The steam and power conversion system will be of conventional design operating on r
l' Heat a modified Rankine cycle with moisture separation and malti stage reheat.
addition to the cycle will take place by generating steam in the four RESAR-35 steam generators. Output will be in the form of electrical energy from the steam turbine driven generator. Heat rejection from the cycle will be to the circulating l
water system uten will take place in the condenser. The entire system will be designed for the maxim a licensed thermal output from the RESAR 3$ nuclear steam I
The stcan and power conversion system will provide the load follow.
supply system.
ing capability require t by the RESAR-35 nuclear steam supply system and will be designed witnin the 1 Sits to be specified by the turbine generator manufacturer, d
The $WES$AR p1 design of the steam and power conversion system will accormodate 1
either a Gereral Electric Company or Westinghouse turbine generator unit (Westing.
bse turbine generator is not within the scope of RESAR 3$). The utility applicant I
referencing SWES$AR P1 will select one of these specific marufacturers and address the details in its safety analysis report. The utility applicant will also address l
l the circulating water syst.m required to remove the heat rejected in the condenser.
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10.2 Turhire Generator _
The following evaluatinn is equally Uplicable to the use of a llestingheise or General E'ectric Cnmpany turbine generator in the SWE$$1 SPI balance-of plant
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j design. If a utility applicant referencing $WE$$AR-p15 elects a turbine generator from a dl%-ant manufacturer than provided for in the above options. the appropriate r
design information snd anclyses applicable to that manufacture
- will be provided in its safety analysis report.
The turbine generator will be a tandem compound type consist ng of one double flow i
The )tational speed will high-pressure turbine and three 1 > pressure turbines.
The turbine electro-rydraulic control system will be 1800 revolutions per minute.
control the steam flow through the turbine by modulating tre turbine inlet steam The turbine control system will be designed to trip the turbine fg turbine overspeed, condenser low vacuum, excessive control valves.
under the following conditions:
l 9 thrust bearing wear, reactor trip, electric generator trip. low bearing oil pres-
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sure, low hydraulic fluid pressure, or manual trip. Overspeed protection for the turbine generator will be provided by ?.vo independent systems, the electro hydraulle I
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s centrol' system and the sechanical overspeed protection system. The electro-hydraulic control system will rapidly close the governor and interceptor valves if
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103 percent cf rated speed is exceeded. If 111 percent of rated speed is reached, the mechanical overspeed sensor will trip all steam valves (throttle, governor.
rehest stop and interceptor valves) to main.;ain the speed below 120 percent o.
rated speed. As a backup, an electro-asgnetic speed sensor (separate from the i
notati speed sensor) will also trip all valves at 111 percent of rated speed.
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i Based on our review of the design, des 8gn criteMa and bases of the turnee
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8 generator overspeed protection system, we have concluded that this system can meet k
its designated safety functions and is, therefore, acceptable.
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10.3 Main Steam Supply System f
The steam generated in the four steam generators of the SWE55AR-P1/RESAR 3$ design combination will be routed to the turbine by the main steam lines. Each main steen line will contain five safety valves, one air-operated relief valve and one main steam isolation valve. The main steam supply system will be designed to
. f41ssic Category I requirements up to and including the main steam isolation valves, which are housed in separate enclosures in the annu!vs building.
The valves wi1* be designed to close within ten seconds after a maior steam line break. Since the closure signal will reach the actuator within five seconds, the valves will be designed to close in five seconds upon receipt of a signal from the RESAR 35 min steam flow, pressure and steam generator level instrumentation for protection and control of the system. The valves will be designed to close for the condition of the maximum mass flow rate in the event of a double ended steam Failure of one m'in steam isolation valve to line break in either direction.
a i
close, coincident with a steam Ifne break, will not result in unenntrolled flow from more than one cteam generator. Additional aspects of the main steam supply system isolation capability are discussed in Section 7.3.2 of this report.
The main steam isolation valves safety valves, atmospheric dump valves, and the f
steam supply line up to and including the auxiliary feed pumps will be physically j
separated and housed in the safety related cubicles located on the roof of the annulus building which will be designed to withstand the failure of high-energy lines. The relief valves connected to the unaffected steam lines can be manually operated to decrease primary and secondary plant pressure at a rate that is corw i
patible with initiation of the residual heat removal system which will then be i
utilized to remove the decay heat, Based on our review, we conclude that the main steam supply system design criteria and bases are in conformance with the single failure criterion, the position of Regulatory Guide 1.29 as related to seismic design, and valve clesure time require-ments, and are, therefore, acceptable. We have evaluated the separation criteria l
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regarding protection against dynamic effects from piping failure outside contalmnent.
< and we will review the final system layout drawings as discessed in Section 3.6.2 of this report.
10.4 Circulatino Water Systne The circulating water system, which is not safety related, will remove the heat rejected by the condenser. The system is not within the scope of the SWESSAR-P1 design, but will be provided by a utility appilcant referencing the $WESSAR P1 design aM will be addressed in its safety analysis report. The utility applicant will determine the type of circulating water system, the heat load to be removed frem the condenser by the system, and the need for a vacwe priming system. Stone & Webster has identified the interfaces betmeen the circulating water system and the $WES$AR P1 condenser and terhine plant wat w *ystems. Stone & Webster has also specified as an interface requirement for this system that it be physically isolated and remote from the ultimate heat sink end the reactor plant service water system. We find these requirements acceptable for the design of the systas. We util evaluate the system durirg our review of a construction permit application by a utility appilcant referene.
ing the $WES$AR-P1 design.
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l 10.5 Auxiliary Feedwater System.
j The auxiliary feedwater system will be designed to 5:4 ply water to the steam generators for sensible and decay heat removal when the main feedwater system is not available.
The system will also be utilized during certain periods of nonnel startup and shutdown.
in the event of malfunctions such as loss of offsite power and following certain postulated accidents. The system will be designed as a scismic Category I system and will be protected from. tornado missiles.
The auxiliary feedwater system for the SWISSAR P1/USAA-3$ design combination will consist of two redundant 100 percent capacity trains. Each train will serve two of the four steam generators. Each train will be capable of delivaring a minimum flow of 470 gallons per minute to the steam generators as rewlaed by the RESAA-35 design.
Each train will include an electric motor-driven pwo that will be powered from a separate alternating current engineered safety features bus. A steam turbine-driven pump with a minimum flow of 940 gallons per minute is cross connected to both motor.
driven trains of the aux 111ery feedwater system 50 t*at it is capable of providing flow to any of the four steam generators. The steam supply for the turbine will be provided by steam lines from tie of the rain steam lires upstream of the main steam isolation valve.- ihe control system for this train will be operated by direct current power, independent of the alternating current power as discussed in Section 7.3.3 of this report.
l The system piping will be designed so that the design function of each train is independent of other trains when considering either a pipirg failure, a component e
failure, or a power supply or control r 1 function. Erossover lines between trale.s 10-3 L
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will contain two normally ope.1, motor-operated valves. The auxiliary fee 6sater pumps and associated piping will be physically separated from each other and located in ihdtvidual weler tight co@artserds. The $WE55AR-P1 application includes the results of ta analysis which demonstrates that adequate decay heat removal will be obtained with one pump and two steam generators. We have reviewed this analysis and concur a
with the conclustons. The system will be designed to assure that at least one pug supplying tuo steam generators will be available assuming the failure of a single a
active component concurrent with the failure of a high energy line.
The three auxiliary feeduster pumps will start automatically on a safety injection signal. All pumps will also be started automatically on two-out of four low-low wem 1evel signals fNa any one of the four steam generators. Manual ctatrol will be possible both from tie control room aM the auxiliary shutdown panel. The actuation logic for the auxilla y feedwater system will be designed to be consistent with the RESAR-3$ requirements. Westinghouse requires in RISAR 3$ that the aualliary feedwater flow to the steam generators be established within 60 seconds from the time 'he system actuation logic actuates the safety system. Based on SWESSAR-P1 requirements for the diesel generators, the auxiliary feedwater system pumps all) be capable of delivering j
the reguired minimun flow in less than 40 seconds. This tieg is well within the tine required by the RESAR 3$ design.
l The auxiliary feedwater will be supplied from the auxiliary feedwater siorage tank.
l Each auxiliary feedwater puno will be provided with an individual supply line from t>e l
storage tank and the entire supply system will be designed to seismic Category I
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j requirements and protected from tornado missiles. The aualliary feedwater storage j
j tank will have a capacity of 214.000 gal?ons which is in accordance with the R[$AR 35 l
Ir.terface requirement. This quantity of water will W sufficient to d!ntain the I
plant at het shutdown for two hours. f,110wed by cooldown at a rate of $0 degrees Fahrenheit per hour to a condition at which the residual kst removal system can be initia ted. The long term emergency water supply will be provided by the reactor plant service water system by cross connection lines to the suction side of the pumps. A spool piece in each service water system line will be provided to preclude inadvertent contamination of the steam generators under normal meses of operation.
The spool piece is approstrately 1 foot long and will be stored in the innediate I
vicinity of the connection. The expected time required for the installation is one hour or less as compared to the seven-hsur supply provided by the auxiliary feedwate storage tant. We, therefore, conclude that sufficient tire will be available for the l
9 installation of the spool piece into the service water system to assure a conthous i
long-term supply of auxiliary feedwater to the steam generatus.
Based on our review, we conclude that the system design Criteria and tales are in accordance with our positicns including diversity of peser sources. system flexibil-ity, and redundancy including the cortination of single active and nigh energy 10-4 2
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,j ments regarding minimum delivered flow rate, required pug head, and actuation logic. We conclude that the system design criteria and bases are acceptable.
We are currsntly evaluating design and operating conditions that could result in damage *.o feedwater system piping as a consequence of pressure waves (mater hamer) i resulting from f?ow instabilities in the feedwater system. The results of this Investigation may result in further requirements being igosed on the RESAR-35 standard nuclear steam supply system design and/or on the SWEs$AR-p1 design of the feedwater system so that unacceptab.e damage will not result from potential feed.
water hasser. We will require that the $WE55AR P1 design be modified if the resolution of this design aspect so dictates and we will review any such required changes during our review of an application for an operating license by a utility applicant referencing the $WE55AR-P1 design.
10 6 Materials considerations The mechanical properties of materials selected for Class 2 and 3 ccenponents foa the steam and feedwater systems of the SWESSAR-P1 design satisfy the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (theASMECode),Section!!!, Appendix!. The fracture toughness properties of ferritic materials will satisfy the requirements of subsections NC 2300 and ND.
2300 of the ASME Code. The mini::un service temperatures for these materials will be set at 100 degrees Fahrenheit above the applicable reference nil-ductility temperature.
The controls to be imposed upon austenitic stainless steel are in confomance with the reconnendations of Regulatory Guide 1.31. The testing qualification of pro-cedu"e welds for sensitization will comply with the recormendations of Regulatory Guide 1.44. The controls to be placed upon concentrations of leachable impurities in nonmetallic thennal insulation used on austenitic staletess steel cegonents of the steam and feedwater systems are in conformance with the recomendations of Regulatory GuiJe 1,36.
The welding procedures to be usM in limited access areas conform to the recomenda-tions of Regulatory Guide 1.71. The ensite cleaning and cleanliness controls to be applied during f abrication satisfy the recomendations given in Regulatory Guide 1.37, and the requirements of AN$! Standard N45.2.1-1973 of the American Hattonal Standards Institute. The precautions to be taken in controlling and monitoring IN prthest and interpass temperatures during welding of carbon and low alloy steel cocponents conform to the recomendations of Regulatory Guide 1.50.
We have concluded that conformacts with the codes, standards, and applicable Rugulatory Guides constitutes an acceptable basis for assuring the integrity of the steam and feedaater systems, and for meeting the requirvents of Criterien 1 of the General Design Criteria, 10-5 l
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j 11.0 RAD 10ACT1vt WASTE MAmasottri
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11.1 Summary Description i
The radioactive waste systes for the SWE55AA-P1 standard balance-of-plant design i
will consist of the liquid, gaseous and solid waste systems. The systen will be l
designed to provide for controlled handling and treatment of all liquid, gaseous and l
solid wastes.15-evaluated the radioactive waste system for a single unit station.
$eparate systees will be provided for each unit of a multi unit station.
The following aspects of the radioactive waste sy'. tem were not considered in our review because they are dependent on the characteristics of a specific site:
1.
The capability of the liquid and gaseous waste systems to meet the dose design objectives of Appendix ! to 10 CTR Part 50.
The cost-benefit analysis required by Appendix 1 to 10 CFR Part 50.
2.
The consequences of a component failure that could result in the release of 3.
radioactive liquids to site related potable water supplies and nearby surface water.
We will evaluate these aspects during our review of a construction perzit application by a utility applicant referencing the $dESSAR-P1 design.
The radioactive waste system. Including process and effluent radioactive monitors.
will be designed to process and control the radioactive waste materials and flow-rates from the nuclear steam supply system and specified in RESAR 35 as interface The RESAR 35 application includes a boron recycle system which has requirements.
been replaced by the boron recovery system in SWE$5AR-p1 (see Section 11.2 and Appendix A of this report). During our review of the SWESSAR P) radioactive waste system we have determined that the interface requirements of the RESAR-35 design have been identified in the SWESSAR P) application and are met by the SWESSAR-Pl design.
The Itquid waste sysiem will process wastes from equiprent and floor drains, and decontamination laboratory and laundry wastes. The gasecus waste system will provide delay capacity to decay short-lived noble gases st*ipped from the primary coolant and treatment of ventilation exhaust air throJ;h high cfficiency particulate air filters and charcoal adsorbers. The solid waste system will provide for the solidification, packaging, and storage of radioactive wstes nenerrted durina station operation prior to shiceent 'o* offsite burial. Solid packacec wastes will De shipped to a liceased facility for burfa).
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In our evaluation of the waste managment systems we have considered (1)the i
capability of the systens to control the levels of radioactive materials in liquid '
eff1rnts based on expected radweste inputs ove* the life of the plant. (2) the capability of the systes to maintain releases below the limits in 10 CFR Part 20 Section 20.106 during periods of fission product leakage at design levels from the fuel. (3) the capability of the systeres to meet the processing demands of the station during anticipated operet. anal cccurrences. (4) the quality group and seismic design classification applied to the syates design. (5) the design features incorpor4ted to preclude uncontrolled releases of radioactive materials due to tant overflows. and (6) the potential for gaseous release due to hydrogen explosion in the gaseous I
redweste systan.
4 I
e In our evaluation of the solid radweste trestrent system, we have also considered (1) system design objectives in terms of expected types, volumes and activities of waste processed for shipment offsite. (2) waste packaging and conformnce to appli '
cable Federal packaging regulations, and provisions for controlling potentially radioactive airborne dusts during baling operations, and (3) provisions for onsift storage prior to shipping.
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In our evaluation of the process and effluent mnitorhg system. we have considered the systea's capability (1) to control the release of radioactive materials to the environment. (2) to monitor all normal and potential rathways for release of raale-active materials to the environnent, and (3) to monitor the perforvance of process equipment and detect radioactive m terial leakage between systems.
We have detemined the quantitles of radioactive sterials tMt will be released in the liquid and gaseous effluents and the quantity of material that will be shipped offsite as solid waste for tasrial during norm 1 operations including anticipated operational occurrences.
In making these determinations, we considered waste flows and activites and equiprrent performance consistent with expected normi plant operation, including anticipated operational occurrences, over an assumed 30-year life of the planc. Liquid and gaseous sourte tems were calculated using the PWR-GALE code described it. NUREG-0017 I
' Calculation of Releases of Radioactive Material in Effluents from pressurized Water Reactors (PWR)." April 1976 including the principal parameters and their bases used in these calci.leHons.
6 j
Dased on our evoluction as described in detail belcw. we find the above aspects of the proposed liquid. gaseous, and solid radwaste systems and associated process and effluent monitoring systems to be acceptable.
11.2 Liquid Radwaste Trs*_?.9t Systm The 11guld radioactive waste treatment system will consist of process equiprent and lostrumentation necessary to collect, process, ronitor and recycle or dispose of 11-2 j
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r liquid radioactive wastes. The Itquid radioactive waste will be processed on a "atch c
basis for opt'.un control of releases. Prior to being released, samples will be y
analyzed to determine the types and amounts of radioactivity present. Based on the results of the analysis, the waste ul11 be retained for further processing, recycled for eventual use in the pisnt, or released under controlled conditions.
The liquid radweste treatment system will collect and process wastes based on the chemical purity relative to the primary cuotant, as determined by the origin of the l
weste in the plant, and consists of two subsystens, the boron recovery system and the radioactive liquid waste systas.
j In addition to the above systems, the condensate polishing system and the steam generator blowdown system were considered in our evaluation. The condansate polish-irg system will use domineralitation to process secondary system condensate which t
becomes radioactive due to primary to see ndary leakage. The regeneration of the demineraliters will result in liquid waste wLich wl11 be proc:ssed in the radio-active liquid waste systen. The steam generator blowdown system will be used in conknction with the fee <hseter systes and the conder.sete polishing system to control the concentration of radioactivity and solidt in the steam generators. Design parameters of prin;:1 pal components considered in our evaluation of the liquid rah active waste system are Ifsted in Tele 11-1.
The boron recovery system will process shim bleed from the reactor coolant system and I
from equipment drain waste (see also Appendia A of this report). Wastes will be processed through the radioactive geseous waste system degasiffer for stripping of f
noble gases from the liquid. The liquid wastes will then he processed through l
I demineralizers and a boric acid evaporator to recover boric acid. The evaporator distillate will be routed to the primary grade water storage tank, if 11 is necessary to discharge the distillate to maintain the pl&nt water balance or to c;ntrol tritium concentration, the distillate will be rout 6d to the radioactive liquid waste systen for release to the envircraent. We estimate tha.10 percent of these wastes wiil be discharged.
The radioactive liquid waste system will 5. ocess miscellaneous low-purity wastes collected in floor and laboratory drafns, and in bui1 ding sumps, by flitration, evaporation and deminere11 ration. We estimate that 100 percent of the evaporator distiilate from these wastes will be discharged. The system will also process, by evaporation, chemical regenerant waste solut'ons from the condensate polishing system in a separate processing train and will route the evaporator distillate to tSe condenscr. We estimate that, as a result of anticipated operat.onal occurrences.10 percent of the evaporator distillate sill be discharged to the environment. The laundry waste system is discussed in Appendix A of this report as an optional system i
in the TE %AR.pl design.
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3 DE$1GN PARAMEfrPS OF PRINCIPAL COMPONENTS t
FOR L10VID. GA$E0US: AND $0LIO RAtWAS.E SYSTEMS
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' Capacity halftg Number fach Group i
'f Boron Recovery System l
s Baron recovery tanks 2
150.000 gal D
l Cesium removal lon enchangers 2
J5 ft3 D
l Boron evaporator 1
25 gpa D
I Boron domineralizers 2
35 ft3 D
Steam Generator Blowdown System l
l Flash tank 1
380 gpm D
i flash tank pumps 2
380 gpm D
Radienctive liquid Waste System High level waste drain tanks 2
25.000 gal D
Low level waste drain tanks 4.000 gal D
Waste test tanks 2
18.000 gal D
Waste evaporator 1
25 gpm D
j Regenerant ches.fcal evaporator 1
25 gpm D
'4aste domineraliter 1
35 f t3 D
Padioactive Gaseous Waste System
}-
Process gas conpressors 2
3 scfm C
Degazifier 1
150 gpm C
+
Charcoal bed adso-bers 2
11.500 lb charcoal C
Oegasiffer recirculation Pumps 2
150 gpm.
C
+
j Radioactive Solid Waste System
$ pert resin hold tank 1
3.200 gal D
$ pent resin surge tank 1
500 gal D
Evaporato. tott: ens tank 1
4.000 gal 0
Waste sludge ta..k 1
2.250 gal D
I*IQuality Group D design criteria include additional quality assurance provisions in accordance with Branch Technical Position ETSB.11.1 Revision 1.
Abbreviations used in table:
gal gallons gpm gallons per minute ft3 cubic feet i
4 scfm standa d cubic feet per minute lb pounds i
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i Turbine building floor drain discharge wasts ul11 normally be monitored and dis-charged without treatment. If the radioactivity exceeds a predeteraired level, the stream will be processed through the radioactive liquid waste system.
The design flow capacity of the boron recruery system evaporator, the radioactive l
liquid waste system waste evaporator and the radioactive liquid waste system regenerant I
evaporator is 36.000 gallons per day for each. We have calculated the average
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expectaJ waste flow to each of these thw evaporators as 4820,1400 and 3400 gallons per day, respectively. The difference between the design and expected flow capacity for each evaporator will provide a6quate re'.arve capacity for processing surge flows. We consider the design and the capacity pf the radioactive liquid waste system to be adequate for meeting tt ? demands of the plant during any anticipated operational occurrences.
i The st9am generator blowdowa system will blow down water to a flash tank where the steam will be souted to the feedwater heaters anel the condensate will be routed to the aatn condenser hotwell. The tuo blowdown flash tank pumps will each have a i
j design capacity af 547,200 gallons per day. The average expected blowdown rate will be appronmately 86,400 gallons per day. We consider the syster design capacity to
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be adequate for meeting the needs of'the plant, t
l The Itquid radweste systems will be located on a seismic Category I foundation. The l
seismic and quality group classification. listed in Table 11-1, and the seismic classification of the equipment are consistent with Branch Techttin31 Position EST8 r
11 1, Rev.1.
- Design Guidance for Radioactive Waste Management tystems Installed in I
Light-Water-Cooltd Nuclear Reactor Power Plants." The systn will also be datigned to preclude the uncontrolled release of radioactive matfr > ss due to overf!r.rs from indoor and outdoor tanks. Level instrumentation will alarm in the control room, and curb; and retention walls will collect liquid spillage and will route it to the liquid radwaste system. We consider these provis'es to be capable of preventing the
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uncontrolled release of radioactive materials to the envircrvnent. We conclude that l
the applicant's proposed system design is acapaole in accordance with our Branch TechnicalpositionETSBJ1-1,Rev.1. W2 W1 requiie a utility applicant referencing the $WES$AR-P1 design to derenstrata in f ts construction permit application that the doses, associated with the postulated fv:1ure of non-selimic Category I components of I
the liquid radweste systems, will nct exceed the limits set furth in 10 CFR part 20.
We have determined that during normal operation the proposed liquid radwaste treatment
- g systems for the $WES$AR-pl design utilizing a RESAR-3$ nuclear steam supply system will be capable of reducing the release of radioactive materials in the liquid l
effluents to approximately 0.2 curies per year < excluding tritium and dissolvel gases, and 720 curies per year of tritium.
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f 11.3 Gaseous Radweste Trattment System Gaseous radweste treatment systems will be designed to process gaseous plant wastet l
based on the origin of the wastes in the plant and the expected activity leveis. The I
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gaseous waste treatment system will consist of a radioactive gaseous waste system, a main condenser evacuation systen and building venttistion systems that will coatrol 2i the release of radioactive materials in effluents to the environment.
{w The radioactive gaseous waste system trill degasify primary coolant letdown as well as l
liquids collected in the reacMr coolant drain tank and the primary drains transfer tank. These gases will be dehumidified, passed through ambient temperature charcoal bed adsorbers to provide holdup time for decay of noble gases and to adsorb rsdio.
iodine. and will then be discharged to the environment. Cover gas from aerated tanks f
and equipment will be passed Lhrough high efficiency particulate dir filters and released via the plant vent. Design parpeters of princiod components considered in f
our evaluation of the radioactive gasecus vaste system are Ilsted in Table 111.
f I
The design capecity of the degasifier will be 150 gallons per minute. The average i
l l
expected input flow to the degasifier will be 75 gallons per minute. The difference i
cetween the expected flow and design flow capacity will provide adequate reserve for L
processing surge flows.
Doerating with redundant degasif f er recirculation panps, redundant compresso-s and dehumidifiers, and the provision for two trains of charcoal adsorber teds. assurC that the radioactive gasecus waste system all) have adequate capacity to allow operation during periods of equipment downtime. We consider the system capacity and j
the system design to be adequate for meeting the darands during normal operations and anticipated operational occurrences.
The system design will include redundant oxygen analysers which will initiate an 4
6 alarin if oxygen concentrations exceed the design concentration limits. !n this j
T l
manner the potential for explosive hydrogen.cxygen eintures will be minimized. In M
addition, appropriate portions of the systea will te designed to withstand the j
i effects of a hydrogen esplosion. The systei will be designed to Quality Group C and
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seismic Category I requirements and will be located in a seismic Category I struc.
ture. We find the system quality group ar4 seismic design classifications, and the
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-d design provisions incorporated to reduce the potential of hydrogen explosion to be 5
I acceptable.
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The offgas from the main condenser evacuation systen will be processed through high j
efficiency particulate air filters for particulate removal if the radioactivity ex.
ceeds a predetermined level, hoble gases and radiciodine will not be affected by the j
treat.9ent pr0vided. In the event that concentrations of radioactive raterial in the j
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offgas from the main condenser exceeds predetermined linits, the af fected generator (s) j wiil be isniated to reduce tne release rate.
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s Ventilation euhaust from the tanulus butiding and from the solid weste and decoo-taminationbuilding(whenradioactivereleasesexceedapredeterminedlimit)willbe processed through high efficiency particulate air filters and charcoal adsorbers prior to release to the envirorment. The wr%1ne to11dirg veettistion exhaust will be released to the envirorment without treatment.
The $WE55AR-P1 proposed containment purge system will be isolated during reactor aration as discussed in Section 6.2.5 of this report. Therefore, in accordance wit 'UtfG-0017. we have calev1sted the radioactive releases from the contairment based un four purges per year during hot and cold shutdoet. Containrent purges during plant shutdom will be processed through high efficiency particulate air filters and charcoal adsorbers prior to purging to the ventilation exhaust 5-stem at r9ector shutdown. As discussed in Section 6.2.5 of this report the pre;esed SWESSAR-P1 containment design includes two spare penetrations that can be used for the instal-j lation of a supplemental on-line purge systen. We will evaluate the rt.ed for the i
supplemental on-line purge system during our review of a construction permit applica-l tion referer.*;ing the $WE55AR-P1 design. We will determine the additional ridtoactive I
releases from a purge of this type if we Jetermine t!.at the system is required.
We have determined that the proposed gaseous radmasie treatment systems and plant vc.stilation system for the $WE55AR-p1 design utilizing a RESAA-35 nuclear steam c
supply system will be capable of reducirg the release of radioactive materials in gaseous effluents to approximately 450 curies per year of noble gases and 0.03 curies per year of iodine-131.
11.4 Solid Royste Treatet System The radioactive solid waste system will be designed to collect and process wastes based on their physical form and need for solidification prior to packaging. ' Wet
- f' solid mastes, consisting of spent demineralizer resins and evaporator bottoms, will be cornbined with an urea forraldehyde solidification agent and catalyst to form a i
solid matria and will be sealed in shipping cortainers of 50 c41c foot capacity.
- Dry
- solid wastes, consisting of ventilation air filters. contaminated clothing and paper, ed miscellaneous items such as tools and glassmare, will be compacted into steel dr
- ith a capacity of 55 gallons. Miscellaneous solid wastes, such as irradiat
,irimary system components will te handled on a case-by-case basis coesid-ering their size and activity. Expected solid waste volumes and activities shipped offsite for each reactor will be 360 contairers per yee of " wet
- solid waste contain-l ing an average of 6 curies per container and $50 dres per year of " dry" solid waste containing less than 5 curies total activity. Design paraceters of the solid radio-active waste system considered in our evaluation are listed in Table 11 1.
i Drum filling operations will be controlled remotely from consoles located outside the drue fill area. Druming operations will have interlock features to prevent i
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e overfilling of containers. In addition, the building will be designed so that any I
spills will be collected in curbed cubicles. M11ng of
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- mestes will be carried out in an area.which is exhausted through a high efficiency particulate air filter and then to the plant vent.
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1 i
The portion of t M radioactive solfJ waste system containing radioactive liquids will be located on a seismic Category I foundation. The quality group classifications of the equipment, which is consistent with our guidelines, are listed in Table 11-1.
I Storage facilities for up to 50 costainers of solid radioactive westes will be provided in the solid easte and decontainat.on building. Based on our estimate of 360 contelnors per year, te find the storage capacity adeg. ate. Wastes will be f
packaged in accordance with ret.frements of 10 CFR Part 20.10 CFR Part 71 and 49 CFR Parts 170-178 am shipped to a licensed burial site in accordance with regulations of the Commission and the Department of Transportation.
l 11.5 Process and f ffluert Radiolacical Monitorino Tte process and effluent radiological monitoring syste1 will be designed to provide information concerning radioactivity levels in systmas throughout the plant, indicate radioactive leakage between systems, monitor equipment performance, and monitor and control radioactivity levels in plant discharges to the environs. Stone & Webster l
has identified in SVE51AR-P1 the liquid and gaseous streams to be e:onitored as listed in Table 11-2. Monitors on (1) the gaseous effluent release lines of the coetainment purge system and of the radioactive gaseous waste syste and (2) the liquid effluent release lines of the tnrbine building drain and of the radioactive liquid waste system will automatically terminate discharges should radiation levels exceed a predetermined value. Systems which are not anenable to continuous monitoring, or for.
which detailed isotopic analyses are required, will be periodically sampled and analyzed in the plant laboratory.
I We have reviewed the locations and types of effluent and process monitoring to be I
provided. Based on the plant design and on the continuous monitoring l eations and intermittent sampling locations, we have concluded that all normal and poteatial i
release pathways will be monitored. We have also determined that the secoling and monitor % povisiens will be adequate for detecting radioactive material leakage to normally uncontaminated systems and for monitoring plant processes which affect 1
radioactivity releases. On this basis we consider that the monitoring and sampling provisions meet the requirements of Criteria 60, 63, and 64 of the General Design Criteria and tha guidelines of Regulatory Guide 1.21 and, therefore, are acceptable.
11.6 Conclusions 1
I Dur review cf the radioac+tve vaste r.anagement systems included: (i)systemcapa-1 f
bilities to process t& types and volumes of wastes expected daring nonnal opera-l tions and during anticipated o.,erational occurrences, (2) the design provisions 11-8
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i TA8tf 11 2 M0u1TORING Q(QTE55 AND EFFLUENT STREAM 5 A.. Lieuld Acactor plant component cooling water Liquid wasta release Plant discharge line Turbine buf1 ding drain discharge Service water Steam gewrator blowdown sasole s
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incorporated to control releases of radioactive materials in accordance with Crf.
te* ion 60 of the General Design Criteria, and O) the conforr.:nce of the quality group and seismic design classification of the systans with the guideline 6 of Branch Technical Position ETSB 11-1. Rev.1. We have reviewed the system descriptions, process flow diageams, piping and instrumentation diagrams and design criteria for the ccuponents of the radmaste treatment systems. We have performed an independent calculation of the releases of radioactive materials in liquid and gaseous effluents.
Our review of the radiological monitoring systass inCbded: (1)theprevisionsfor sampling and monitoring all plant effluents in accordance with Criterion 64 of the General Design Criteria. (2) the provisions for automatic terwination of effluent releases and assuring control over discharges in accordance with Criterion 60 of the General Lasign Criteria and Regulatory Guide 1.21. (3) the provisions for sampling and monitoring plant waste process streams for grocess control in accordance with Criterion 63 of the General Design Criteria. (4) the provisions for conducting sempling and analytical programs in accordance with the guidelines in Regulatory Guide 1.21. and ($) the provisions for monitoring process and effluent streams during postulated accidents. The review included pioing and instrunentation dia-grams, process flow diagrams and interface requirements for the liquid, gaser 15. and l
Solid redweste systeir.s and ventilation system, and the location of s:onitoring points relative to effluent release points on the site plot diagram.
The basis for acceptance has been confomance of the esign, design criteria and design bases for the radioactive maste treatment aa.1 monitoring system to the appli-cable regulations and guides referenced above as well as to staff technical pesi-tions end industry standards.
We find the radioactive liquid waste system and the radioactive gaseous waste system will be capable of maintaining concentrations of mat rials released in liquid efflu-ents durig periods of eculpment downtime and design bases fuel leakage within the limits of 10 CFR Part 20. Section 20.106. The proposed seismic and quality group classifications of the systems are in accordance with our Branch Technical Position ET58 11-1. Sev. 1. and the design of the systeins contain adequate provisions to control release of radioactive materials. These aspects of the systems are acceptable.
The capability of the proposed radioactive 11guld waste system and radioactive gase-ous waste systeri to meet the dose design ot:1ectiv% of Appendia 1 to 10 CFR Part 50 is site dependent and will be reviewed for indivikal appilcaticns that reference the
$'mT$$AR PI design.
f We find the radioactive solid waste e stem will have adequate capacity to handle radioactive solid wastes produced durieg norral reactor operation including 49tici-pated o;erasional occurrences, in accordance with triterion 60 of the General resign Criteria, at.d 15. therefore, acceptable. he find the radiological ronitoring systees f
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capable of monitoring all major process effluent pathmays, in accordance with Criteria 13 aed 64 of the General Cesign Criteria and of controlling suitably the release of radioactive raterials in 11ould and gaseous effluents. in accordance with Criterion 60 of the General Design Criterte and, therefore, acceptable.
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12.0 RA !ATION r8TTCT104 l
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nae have evaluated the proposed radiation protect 4 destp considerations, includleg radiatica shielding, as described in $431AA.Pl. The shielding for the fdi1AA-P1 facility will be designed with the objective to uswJ that the criuria sxcilled la 10 CTR Part 20 will be net that exposure of personnel is as low as is reasonab?e achievatle during normal c5eration, and tut the requirements specified in 10 CTR Part 100 ull) be met during accident conditions. TN $411AR.71 plant will be desired to maintain occapational radiation esposures as low as is reasonably achfewable is j
conforma. ace with Position C3 of Retstatory Guide 8.8 and with 10 CTR Part 20. Section 20.1(c). Stone & Webster has proviced radiatten desi r dose rates for different radiation zones proposed for the %IS$AA.71 plaat layout. he have confirmed tut the I
I proposed desip has been reviewed by competent radiation rotectica specialists, as recomended in 8eplatory kide 2.8. and find tre desip ocjectives accertatie.
The shielding desip and ardlysis will te sis 11a. to tut descrited la t)e Stone &
Webster Topical Report RP-EA.
- Radiation Shieldirs Dest e and Analysis Asproach For Light ' ater Resctor Power Plat:ts*. umica mas reviewed and accepted by tr.: staff.
Stone & Webster t.as stated that everical values for dose rates for the typical pressurized water reacter euclear power plant, described in 14 tegical resort. are indicative of the general radiation levels in a Seti$AA.Pl 5 ast. TPe nasimes snielded 1
i desip cose rate fer eacn tone vill cot te exceeded w*en all sources o. rafiation l
including penetrations in the sateles are consicered. frecause of the dif ference in l
power levels tetween the $455AA-P1 and t*e topical report naclear power plank f.*e
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radiation levels in tre unsatelied areas of tre IdE.S$AA-71 plast weald gewr:11y te l
aboat 50 percent higner. The 5155AA-71 snieldirg desip, hchever, will te based om j
t*e nazisua radionuclide concentrations la all $4%AA-71 eqwipneet. Other assep-j ticas used la the shieldir1 anal; sis result in a stfeld design t*.st is c:eservative.
l Stone & metster has also reviewed infarnatice paired at eseratieg statices of a desip sisiler to the $411Mt.Pl *asyn ta impreve t*e ratNr.atical accels. he fl4 these asproaches to t*e sPield desip to te consistent with tte twi$ellres in l
Regulatery k tse 8.8.
Der acce;tance of t*e plant radiatics tonteg is t>ated on results of the d:se assessrent discussed in Se: tics 12.3 telcw.
I i
1 Stone & meoste e.as presented in 54stAa.71 a ce*. ailed descristica, fc11ovirg t*e j
forvat of Position C3 sf Reguist:ry hice 8.8. cf t*4 facility ar4 e%i; ment gesip i
j featvets tut relate to linitirg the rastation espasste to as los as is reasseacly l
achievaDie levels. The recome*daticas cf tre dice e*, used to devetes tw f acility aM iesip triteria. Ike arnelas Nilding c::mcent cf the ! 15$1A.Pl design provides a new a;;roat? to access coateol aM facility *.*af f'c ratteers. Persawel 12-1
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access to the annulus butiding, fuel building, weste building and containment will be through the health physics building, except when certain large shipping doors will be open. Most amas with a potential for signtficant radiation esposuru can be reached by travel paths in radiation zones designated as three or lowr with a radiation level of less than 5 milltres per hour, where rem is the unit of radiat*on dose equivalent. Included in the design features are the area and airborne radiation monitoring equipment. We find that the fact 11ty and equipment design features are consistent with the critaria of Regulatory kide 8.8 and are, therefore, acceptable.
The shielding design for areas containing high level radiation sources has been described in detail in SWC15AR.Pl. Additional rationale has been provided, at our request. regarding the design of these areas to assure that occupational radiation exposures will be as low as is reasonably achievable. The discussion provides 1
acdttional evidence of spprooriate consideration by Stone & Webster to the as low as is reasonably achievable caposure principles, and we conclude that the shield design features are, therefore acceptable.
The health physics program. including the design for the health physics area, through which access to the annulus building is to be controlled, is not within the scope of SWI11AR.Pl. At our request Stone & Webster developed co ceptual layouts which indicate that sufficieet sise and space orientation will be available to provide an appropriate and detailed design of this area. We find this provision acceptable for j
a Preliminary Design Approval. he will require a utility applicant referencing the
$.Ti$AA.P1 design to provide the appropriate design for the kalth physics area in its construction permit appilcation.
1 l
12.2 Area Radiation % nitorinq l
i The design features of the area radiatt% monitoring system include renotely operated local check sources, readout in the control room local audible and visual alare capability, and an independent powr supply located in the control room. Radiation monitors will be located in areas where operating personnel are excetted to remain for estended periods of time. Eighteen area monitors will provide adequate coverage of potential radiation areas. The guidelines in Regulatory Lide 8.8 and ether staff reccrsiendations were used to develop the SWISSAR-P) radiation eenitoring system. he will evalt,ata the radiatt e sonitoring System of a utility application referencing the SWI55AR-71 cesign to deterette test the system will te operated in the r4nnec described in $WISSAR.Pl. he conclude that cri*eria and systes design features for the SWI55AR 71 radiation monitoring system are acceptable.
i i
12.3 Dose assestret
$ tone & betster has provided. at our req <est. estirates of ractation exposures fer selected nor-a1 eperations in the SWE55AR.Pl plant cesign. The esttrates for routl*
cperations were devele;ed using radiation fore designation levels ratMr tran the 12 2 h
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actual levels, which are expected to be a fraction of the zone designation levels.
An exposure of 74 man rem per year (rem is the unit of radiation dose equivalent) is estimated from these operations. Based on our experience, me find this estimate to be consistent with other radiation design exposure estimates which we have performed.
Stone & Webster has provided, in SWE55M-#1. Section 12.1.6. additional bases and appropriate assumptions to demenitrate that 500 man rem per year is conservative ter the long tern radiation, exposure from all operations in the SWE55AR P1 plant.
Considering the bases for this estimate w find the value of 500 man rem per year acceptable. Wt will evaluate the dose assessment, to be provided by a utility applicant referencing the $WE55AR-P1 design, to assure that the estimate of the occupational radiation exposure will be based on the manner in which the SWESSAR-P1 plant will be operated. This estimate is to include routine maintenance, instrunent calibration. fuel handling. and inservice inspection operations, which are not included in tne estimate provided in SWE55AR-Pl.
12.4 Ventilation 1
j The ventilation systems fo* the ShE55AR-P) plant will be designed to maintain a i
suitable envirorument for personnel and equipment. hng the design objectives of f
these systems are the protection of operating personnel from possible airborne l
radioactivity and the assurance that manisua expected airborne radioactivity concen-trations will be maintained within the Itaits of 10 CFR Part 20 and as low as is reasonably achievable. The functional performance of the proposed ventilation systems i
l is discussed in Section 9.4 of this report. Stone & Webster, et our request. provided additiW1 infonnation in SWE55AR-p1 for the ventilation systems regarmg the pal i
and principle that occupational radiation exposures will be as low as is reasonuly achievable. The recoariendations for the ventilation systems as described in Regulatory Guide 8.8. Position C31 (design for easy access and Service) and Fosition C3 j (control of airborne contaminarts), the minha air changes per hour in poten-tia11y contaminated working space 4 including the need for portable ventilation equipment and the need for containment purging and fuel building ventilation, under conditions of high airboras radioactivity have been addressed in SWE55AR-Pl. On the basis of our review of the additional infomation provided by Stone & Webster, ne j
find the ventilation systems for the proposed SWE55AA-P1 design acceptable.
i i
The $wESSAR-P1 cesigrt will include airborne radiation penitoring instruments in-j 4
stalled throughout the plant to eesure that inferration on airborne radioactivity l
1evels will be centinuously eval 14ble to the plant operating personnel. The type of I
detutor (gas. p.rticui.te) its range and sensitivity have been spuified for the f
various locatters throughout the plant. Based on our review. we conclude that the airborne radiation monitoring system is acceptable.
g Stone & Wettster has provided estirates of the inhalation ene whole body dose rates in the major buildings of the proposed SWE55AR-P1 design. These expected cose rates are teased on the espected airborne activity levels in the contalment, turbine, annulus and fuel buildings. The personnel occupancy factors fer shesa buildings have not 12 3 "M$__'6,q_&A2
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been specified in $ES$AR-Pl. We will evaluate the estimates of the inhalation and i
whole body doses during our review of a construction permit application referencing I
the $E$$AR-P1 design based on the utility applicant's estimte of occupancy factors to determine that the resultant doses are as low as is reasonably achievable, i
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13.0 CONOUCT OF OPERATIONS l
I Information relating to the conduct of operations is not within the scope of the SWE55AP-P1 application. This information will be provided in each application that references the SWE55AR-P1 design.
We have reviewed the infor1 nation in SWESSAR P1 related to industrial security. Stone
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& Webster has described design considerations for seco,ity in building layout. in.
j l
terior arrangement of equipment, physical control of access and persennel circulation routes. Stone & Webster has also described the separation of redundant safety l
systems and provisions for bypass indication that will alert the operator in the i
f event of an intentional act that could imair the performance of vital equipment.
The building areas to be designated as vital areas in accordance with the definition of 10 CFR Part 73. Section 73.2 have been listed and the protective features of such i
structures have been described. Specifications of detection systems, access control systems, site-related design features and-assinistrative controls are within the scope of a specific utility applicant and will.be discussed in its safety analysis i
r: port.
We conclude that the Stone & Webster design and arrangements of the SWE15AA-P1 plant for protection against acts of industrial sabotage are acceptable. We will review the detailed industrial security plan of a utility applicant referencing the
$WES$AR-P1 design.
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14.0 INIT1At TtSTS *MD DrteAtitms The definition of the program for initial tests and operation is rot within the sccpe of the $WE$$AR-P1 application for a Preliminary Design Approval but is the responst-bility of the utility applicant referencing the $WES$AR-P1 design. Accordingly.
Stone & neebster has not addressed the recommendations of Regulatory Guide 1,68 Houever, at our request.
regarding the preoperational and startup test programs.
$ tone & Webster provided test abstracts in $WES$AR Pl. inc?uding the test objectives, for safety-related systems and design features that are within the scope of the
$WES$AR-P1 design. A preoperational test for the instrument air system will be conducted in accordance with the requirements of 49u1 story Guide 1.80.
We find the test abstracts in SWIS$AR-P1 acceptable based on the information provided in Chapter 14.0 and in appropriate sections for the safety-related systems and design features of $WES$AR-P1. We will require a utility applicant referencing tM $WES$AR P1 design to define its program for initial tests'and operations in its preif ainary safety analysis r port and to provice the details for the program in its final safety analysis report.
lased on the above. we conclude that the information provided in $WIS$AR P1 with i
t regard to the program for initial tests and operatica is acceptable.
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l 15.0 ACCIDENT Atal,Y3E5 l
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Introduction.
Our evaluation of the capability of the RISAR 33 nu. lear steam supply syster to w.tlw stand aimerms) operational transients and postulated accidents is presented in
)
Section 15.0 of our Safety Evaluation Report for RE5AR-35. Therefore the discussico l
below is limited to radiological consequences of accidents related to the SWE55AR-Pl/
[
n RESAR-35 design combination.
)
15.2 Radioloalcal Consecuences of Accidents 15.2.1 General Stone & Webster has analysed the offsite radiological consequences for the postulated j
loss-of-coolant acef dent based on a manieun core thermal power level of 4100 megawatts j
as an upper bound analysis for the SWE55AR.P1 balance-of-plant design. Consequently, j
Stone & Webster did not perform a specific analysis for the postulated loss-of-l coolant accidert at the msninse
- ore thermal power level of 3636 r egawatts for the SWE55A2-P1/RESAR 35 design comnination. The offsite radiological consequences for j
all other postulated accidents were analyzed based on.he maximan de,1gn power level of 3635 megawatts for the RESAR-35 nuclear steam supply system. As Stated in Section 1.2 of this report. the SWE55AR-P1/RESAR-35 appilcation is for a thermal power level of M11 siegawatts. We have reviewed the accident analyses presented in SWES$AR-P1 and hate performed independent calculations of the offsite raciological consequences resulting from a loss-of coolant accident. a hydrogen purge operation of the contain-meet following a loss of coolant accident, a fuel hardlirg accident, and a rod ejection accident. These evaluations are discussed in separate subsections oeiow and the j
results are presented in Table 15-1. We also considered in our evaluation (Se offsite I
doses resulting from leakage from the energency core cooling system following tr.e postulated loss-of-coolant at:ident.
On the basis of our experience in evaluating steam line break and steam generator I
tube rv;ture accidents for pressurtred water reactor nuclear power plants, we have concluced that the radiological consequences of these accidents can be cortrolled in l
a SWE$ TAR P) plant, as they are in other pressurized water reactor plants, by limiting
~
the permissible radioactivity concentrations in the primary and secondary coolant system in a reasonable way so that the potentia) offsite doses will be small. During f
our review of an operating Itcense application referencing the SWE$$AR P1 design, we will include in the technicai specirtestions for the piant appropriate limits on tne l
primary and secondary coolant activity concentrations.
j l
The SW115AR-P1 process gas charcoal adsorber beds in the radioactive gaseous waste I
Therefore, the total system will be designed as seismic Eategory 1 CocConents.
failute of these beds is su ricientiy trecrobable snat to crR part 100 guicetime cases e
are at:1tcable. On tM basis of our experience in evaluating tee rustare of 15-1 i
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TABLE 15-1 d
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POTENTIAL RADint0GICAL CONSE0urNCES OF DESIGN BASIS ACCIDENTS l
AND tlMITING ATMOSPHERIC DISPERSION FACTORS (I/Q VALUES) j.'
i FOR 5WESS.'!J1/RESAR-35 DESIGN t
i Option A. Compiete Opt 1on B. Complete Pattial Dual Dual Contatnment Dual Contelament A.
Accidents and Doses Contalment Without Mining With Mining Thyrold Whole Body Thyroid Whole Body Thyroid Whole Body
[
(Ren)
(Rem)
(Rem)
(Rem)
(Rem)
(Rem) l l
(1) Loss of Coolant *I 150 6.1 83 20 110 20 I
(2) Post-LOCA Hydro en Purge (b) 10
<1 10
<1 10
<1 (3) Fuel Hand 1tn 6
<1 21 1
30 2
a G
(4) Rod Ejection 'I 16
<1 7
<1 12 1
(
l 8.
ilmittna Conditions 3
4.3 x 10'4 1.4 x 10-3 2.1 x 10'3 (1) Lietting I/Q (sec/m )
for LOCA (item A1)
(2) Dose Limiting Organ Thyroid Whole Body lihole Body Abbrevletions used in tables notes:
LOCA loss-of-coolent accident (a) doses (2-hour) based on limiting X/Q value in item 81 I
-6 3,gy,3 in Table 15-4 X/Q atmospheric disperston factor (b) doses (30-day) based on assumed X/Q
- 5.0 x 10 sec/m seconds per cubic meter Rem Roentgen equivalent men I
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I charcoal adsorber beds, we conclude that the dose produced as a result of a potential rupture of this system would be well within the 10 CFR Part 100 guidelines.
Our model for the calcuhtton of whole body doses for judstng acceptability with respect u 10 CTR Part 100 has recently been revised. The revision has resulted in reductions of the calculated whole body doses by about a factor of two. compared to the former model. The revised model considers the dose from low energy beta radiation as a skin dose, rather than as a contributor to the whole body <lose, as the former model did. This change is based upon recent recommendations by the National Council
(
on Radiation Protection that skin and whole body doses be distinguished, and that only that radiation which penetrates the body to a depth of 5 centimeters can be considered as a dose to the whole body.
15.2.2 Loss-of-Coolant Accident t
The nuclear steam supply systan will be housed inside the contalment a reinforced l
concrete structure with a steel liner, which has a design leakage rate of 0.2 weight percent per day. The SWE55AA-P1 application includes the following three containment f
concepts, which are discussed in detail in Section 6.2.2 of this report:
i t
(1) Base design - partial dual containment.
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6 (2) Design Option A - full dual contalnuent without provisions for afsing.
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(3) Design Option B - full dual containnent with provisions for mixing.
Each of the concepts has different characteristics for the leakage from the contain-t l
ment and/or for reduction of fission products resulting from a postulated loss-of-coolant accident.
A containment spray system, with a chemical additive to incre( se the iodine removal capability of the spray, will be used to reduce the concentration of radioactive lodine inside contatraent following a postulated loss-of-coolant accident. Sections 6.2.3 and 6.2.4 of this report discuss the operation of the contafruent spray ard f
spray additive systees.
- I The SWESSAA-P1 standard balance-of. plant application is not related to a specific g
site and uees not include site specific information required to calculate the potential o'fsitt deses (i.e.. distances and appitcable atmespheric dispersion characteristics for the exclusion boundary and the low population gone). Instead j
of calculating potential offstte doses for a specific site. Stone & Webster has evaluated the suitability of the SWE55AR P1 design for future sites. In the evaluation Stone & Webster has determined the maximun atmospheric dispersion factors i
(X/Q values) that will limit the potential offsite doses to the guidelines of Regulatory Guide 1.4. i.e.
150 rem thyroid dose or 20 rem whole bcdy dose (rea =
The Roentgen equivalent man) for each of the a$ove three cor.tairnent concepts.
15-3
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i limiting I/Q values determined by Stone & Webster for each of the contatreent con-capts are listed in Table 21. which also includes the values calculated by the i
staff.
I We have made an independent determination of the lietting I/O value at the exclusion
'i boundary for each of the contatament concepts resulting from a postulated loss-of-coolant accident. The assuntions used in our dose evaluation for the loss-of-coolant accident are sianmarized in Table 15-2 inc1Aing the assantions for spray j
effectiveness as discussed in Section 6.2.4 of this report. Our asswuptions for the i
contaivuont leakage paths for each of the contaitunant concepts are listed in Table
[.
15-3 which is based on our evaluation presented in Section 6.2.2 of this report.
4 The results of our evaluation are listed in Table 15-1 for the three contairment I
concepts. The table also identifies the organ (thyroid or whole body) that limits
!I the doses to the values listed in the table. The limiting atmospheric dispersion l
factors in Table 15-1 will determine the mininas exclusion boundary distance required for a specific site such that the resultant potential offsite doses from a postulated l
loss-of coolant accident will not exceed the guidelines of Regulatory Guide 1.4 I
We have not evaluated the limiting diso*M'an characteristics for the SWE55AR-p1 plant regarding potential offsite doses at the low population zone distance during the 30 day period following the postulated loss of-coolant accident. These character-1stics are determined by different dispersion factors for four consecutive time intervals during the 30 day period, such that the combination of these factors as a set rather than any individual factor for a particular time interval will estabitsh j
the lwiting characteristics. However, we do not expect these seteorological conditions to be more limiting than the short ters atmospheric dispersion factors listed is Table 15-1.
I i
g The enclosure building for design Options A and 8 has a design leak rate in excess of 100 persent per day, which potentially can cause enfiltration from the enclosure building under some atrespheric conditions. Our evaluation of the offsite doses and the limiting atrescheric dispersion factors listed in Table 15-1 has not taken into e
account any exfiltrattoa effects. These effects partially depend upon actual meteoro-logical conditions for a specific site and partially depend upon SWE55AA-p1 design and plant georJetry features which suy affect tWilding leak tightness and the occurrence of negative pressure regions at certain locations on the exterior surface of the enclosure building. We have performed an independent enfiltration analysis assianing that the secondary contaituneet ul11 be maintained at a negative pressure of 0.25 inch of water. Our evaluation indicates exf titration commences at sufficiently high wind speeds (about 5 to 8 meters per second) and at such small rates that. when the limiting atmospheric dispersion factors are appropriately reduced for the j
increased wind speed, the design basis accident doses with enfiltration are less than the doses calculated without erfiltration (Table 15-1). We will require, in the plant teche' cal specifications of an application referencing the SWE55AR.P1 b
15-4 i
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~ - - ~ 3-6 TA8tf 15-7, L0$5-Or-C00UWT ACCTOENT ASSUMPTIONS AMO INPUT PARAMETERS TO DETERMINE LIMITING ADIOS /NERIC DISPERSION FACTORS (1/0 VALUES)
I (1) power Level 3636 segana6.s thermal (2) Operating Time 3 years (3) Fraction of Core Inventory Available for Leakage:
todines 25 percent Notle Gases 100 percent (4) Initial lodine Composition in Contairment:
Elemental 91 percent Organic 4 s arcent 0
Particulate 5 percent (5) Effective con cituent Vol m (85 percent Of total net free volume) 2.601.000 cubic feet (6) Contairment Spray System Effectiveness Decontarination Factor, Elemnt:1 lodine 100 Removal Coefficients:
(
Elemental lodine 10 per hour Particulate lodine 0.60 per hour Organic lodine 0
a (7) Supplementary Leak Collection and Release
+
system Filter Efficiency. Iodine Remova!
95 percent i
l 7
(8) Primary Contairment Leak Rate (see also Table 15-3) t 0.2 percent per day 0-24 hours after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.1 percent per day t
i 1
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i TABLE 15-3 LC$$-Or.COOLANTACCIDE9, CONTADMOff LEAKAGE ASStMPT10h$
f Containment Leakage Itate
( eakage Concept _
(percent per day)_
,TE A.
Tiet 0-38 seconds (1) All Concepts 0.2 direct unfiltered 8.
Time: 38 seconds - 2 *eurs 1
(1) Base Design 0.1000 direct, unfiltered (partialcual) 0.0023 by.ss, unfiltered i'
0.0977 1Nirect. filtered (2) Base Design Option A 0.0023 b pass, unfiltered (completedual 0.1977 indirect, filtered withoutmixing) r (3) Design Option B 0.0023 typass, unfittered (completedual 0.1977 indirect, filtered with mixing) 1 f
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design, perioele testing of the secondary contatuent to provide assurance that the secondary contalement can achieve and maintain the necessary negative pressure.
As part of the loss-of-coolant accident anelysis. we and Stone & Webster have also cons!dered the radiological consequence, uf 'eakage inside the annulus building of contalrnent sump water containing redl04ctive fission products which is circulated by the emergency ccre cooling systam outside the containment af ter the postulated accident. The supplementary leak c 11ection and release system, an engineered safety features system #:iscussed in Section 6.2.2 and C.5.2 of this report will process such potential leakage, inside the annulus building, from postulated fattures of passive components in systems carrying the post accident recirculation water 2
outside of ccattairment. This conforms to our requirements discussed in Section 15.7.2 of the RESAR-35 Safety Evaluation Report regarding the location of emergency core cooling systen equipment. We conclude that doso resulting from the postulated leakage of post-accident recirculation water from these systems would be low and, when added to the accident doses, result in total doses that are within the guideline values (,f 10 CTR Part 100.
We will require a utility applicant referencing the SWESSAR-P1 design in a construc-tion permit application to provide the appropriate exclusion boundary and low population zone distances and the meteorological characteristics appilcable for the I
specific site. We will evaluate the potential offsite doses for the postulated loss of-coolant accident during our review of such a construction permit application on the basis of the specific site characteristics and on the basis of the limiting I
g conditions established in $WE$$AR-P1 and in Table 2-1 of this report.
15.2.3 Hydrocen Purge cose analysis The proposed $WE15AR-P1 design will include two redundant hydrogen recombiner sub-systems for the purpose of controlling the post accident hydrogen concentration inside the contalrunent that can be geriersted as a result of a postulated loss of.
j coolant accident as described in Section 6.2.6 of this report. In the event of f ailure of both hydrogen recombiner sut' systems, the concentration will be controlled by a backup purge operat')n. We have evaluated the additional dose that an i
individual might receive during the course of a 30-day containment purge operation f
to be initiated 35 days after a design basis loss of-coolent accident basad on the lower flansnability concentration for hydrogen of 4 volume tercent (see Section 6.2.6 of this report). Our assumptions are listed in Table 15-4 and typical <foses for a system withus purge filters are listed in Table 151. These doses are based on an assumed 4Lespheric dispersion factor of 5.0 x 10 seconds per cubic seter which is a reasonable value and based on our experience of previously reviewed sites, i
~
i We will evaluate the cffsite hydrogen purge dose during our review of a construction I
permit application referencing the $WI55AR-P1 design utiliging the appropriate dis-persion factors for the specific site to assure that the coribined offsite doses doe f'
to the postulated loss-of-coolant accident and the purge operation are within the exposure guidelines of 10 CFR Part 100.
15-7 1
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Tutt 15-4 HYDROCEM PJRGE OPERATION A350*T10M5 49 fMNT PARMETERS TO ESTfMATE CrT5ffE tesES (1) Power Level 3636 megawatts thermal (2) Containrent Volume 3.060.000 cubic feet
~
(3) Holdup Time in Containment Prior to Purge Initiation 35 days (4) Purge Da ation 30 days (5) Purge Rate 22.5 standard cubic feet per minute (6) Atmospheric Dispersion Factor 5.0 a 10-6 secorwis per cubic meter l
(Iow population sone distance.
course of accident) i h
i 1
15 8
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15.2.4 Feel Mandline Accident i
I he have evaluated the radiological consequences of a fuel handling accident in the fuel building in accordance with the assimuptions of Regulatory Guide 1.25.
We assiased that a spent fuel asseatly is dropped in the fuel pool during refueling operations and that as a result all of the spent fuel rods in the assembly are damaged thereby releasing the volatile fission gases from the fi,41 rod gaps into the pool. The radioactive material that escapes from the fuel pool is assmed to be released to the environment over a two-hour period with the iodine activity belf.g reduced by filtration through the supplementsry leak collection and release systes.
' The filters used in the system to sitigate the consequences of the fuel handling
~
accident emet the requirements of Regulatory Guide 1.52 and. therefore, we have used a removal efficiency of 95 percent for all forms of iodine (see Section 6.5.2 of this report). For the calculation of the offsite doses (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. exclusion boundary), we used the limiting atmospheric dispersion factors for the loss-of-coolant accident as listed in Table 15-1 of this report. Other assantions are listed in Table 15-5.
The results of our evaluation are listed in Table 15-1.
l We conclude that the radiological consequences of the postulated fuel handling acci-dent in the !WE55AA-P1 plant design are less severe than those from a loss-of-coolant accident and therefore are not limiting.
15.2.5 and Hection Accident We have evaluated the radiological consequences of a RESAR 35 rod ejection accident for the containment leakage mode. In our evaluation. =c used the same asswiptions with regard to contalruent leakage, ef fectiveness of the supplementary leak collection and release system and limiting atmosseeric dispersion factors as for the 1 ass-of-coolant accident discussed in Section 15.2.2 of this report ana IIsted in Table 15-1. Other assumpticas are Itsted ir. Tat >1e 15 6. The results of our evaluation are listed in Table 15-1.
we conclude that the radiological consequences of a rod ejection accident are less severe than those resulttag from a postulated loss-of-coolant accident and are within the guidelines of Regulatory Guide 1.4 and, therefore, of 10 CfA Part 100.
We have not analyzed the consequences of the leakage of fission products to the Secondary systee af ter a rod ejection accident. Badiological consequences of this l
accident may be limiting in tems of limits for primary to secondary steam gene-nor 8 a construction perm t 1eaLage. We will evaluate this as;ect during cur review 0 application that teferences the SW155AR-P1 design.
15.3 anticipated Transients Without Scr=3 Stone & Webster did not directly address the retter of anticipated transients without scram in its SoI5!AR-P1 a;pilcatien. In our Safety (valuation Report for RISAR.35 15-9 h ( W N. ? Q Q.nyys g wa.g g r y
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A553FT108t$ A.v0 TUUT PAtMETT?$
70 ESTT4 TT OFT 5fTE toits FW $WES$AR 81/7t$A8-15 Of51GN j
1 (1) Pour Level xx seemetts trem 1 (2) lbseer of Fuel nods Dameged 264 (3) Total thseer of Fuel Rods ta Core 50.952 (4) Pouer Feating Factor of Cameged Fuel 1.65 (5) Shstdown Time 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (6) Inven a ry teleased from Oaaeged Rods 10 perce:rt (todites and hotle Gases)
F 7
i (7) Pool Detectestnation Factors lodines 100 hable cases 1
(8) Iodine Fractions Above Pool 1
Elemental 75 percent 25 percent Organic l
(9) Filter Efficiencies for fodine Rasoval Elemental 96 percent Orgamir 95 percent (10) Atsspheric 01spersion factor see Table 15-1
)
(2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, exclusion toedary) 4 3
b 3
9 15 10
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i TABLE 15-6 i
R00 EJECT!0N ACCICENT A$$ls@TIONS AND INPUT PARAMETERS TD E$71 MATE OFFSITE 00$($
FOR $WE$$AR-P1/ef1AR-3$ CE$fCN i
l (1) Power Level 3636 segawatts thermal (2) Fuel Failed in Transient 10 percent (3) Inventory in Gap of Failed Fuel:
todines 10 percent Noble Gates 10 percent (4) Release of Inventory in Cap:
lodines 100 percent hable Gases 100 percent
+
(5) Plate-cut of Iodines in Containment 50 percent (6) Contairment $ pray System Operation not initiated (7) Contaffusent Leak Rate see Table 15-3 (8) Atmospheric Dispersion Factor see Table 15-1 i
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1 (Section 15.5.7).== indicated the current stat;as of anticipated traestests without locause our eenetic review is et scram for the Rf!AA-3$ suclear stein supply sistent.
j completed, se have determined that it mid be premature to estan115A appropriate inter-face requirements for t.*e RE3AA-33 design. Potential plant andifications f avolve l
j balance-ef-plant responsibilities aM. ttus, mst te addressed by $tane & beester for We have defined tw matter of esticipated treasier.ts without the 3d$$AA-71 ersign.
scrum,1esofar as it tapacts the belasce-of-plant design, as as laterface matter and will rewire any changes that need be ande on the bests of aprmed analyses to be 5
facorporated into the $ 153AA-71/EC$AA 35 des 19e comminatica is a tisely aswwr.
De We will issas a Prelistaary resign Approval far $415AA-P1/RI$AA-3$ on this basis.
conclude that this interf ace requirement provides se acceptable besit for a Pre 11sinary Design Approval for tm U.S$AA-71/tE$AA-35 desige castimtion.
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16.0 TfCHNICAL $PICfFICATIONS o
The technical specifications in an operating Ilcense define certain features, charac
- teristics, and conditions governing operntion of a facility tMt cannot be changed without prior approval of the Commission. Final technical specifications will be developed and evalu.ted at the final design review stage. However, in accordance with Appendis 0. paragraph 3. of 10 Cf R Part 50 an application for a Preliminary l
Design Approval of a standard design is required to include preliminary technical specifications. The regulations require an identification and justification for the selection of those variables conditions, or other items which are dctermined, as a result of the preliminary safety analysis and evaluation, to be probable subjects of technical specifications, with special attention given for those Itans which may.
i significantly influence the final design.
We have reviewed the proposed technical specifications presented in Chapter 16 of the SWESSAR-pl Safety Analysis Report in conjunction with our review of Chapters 1 l
through 15 of SWES$AR.P1 with the objective of identifying those items that woula requiis special attention at the pec11minary design review stage to preclude the necessity for any significart change in design to support the final techalCa? SpeCI-fications. The proposed technical specifications are similar to those'being developed or in use for nuclear power plants.
Several of the technical specifications proposed in SWE55AR.P1 have been modified as a result of our review and are addressed in the sections of this report where the particular aspect is evaluated.
On this basis, we have concluded that the proposed preliminary technical specifica.
tions are acceptable.
4 I
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17.0 OUallTY A$$U9ANCE 4
I i
17.1 General l
Chapter 17 of the $WES$AR.P1 Safety Analysis Report describes the qualttv assurance l
program of the Stone & Webster Engineering Corporation by reference to the Stone &
We j
Webster topical report SW5QAP 174 *$tandard Nuclear Quality Assurance Program."
The Stone #
j previously evaluated this topical report and found it acceptable.
6 Webster quality assurance program covers safety related equipmen* from decign through I
procurement and construction activities which are within the $ tons & Wettster scope of work for nuclear power plants.
17.2 Oroanization i
l The Stone & Webster organization responsible for design, procurement, and construc=
tion activities is shown in Figure 17-1. The President of the Stone & Webster Engineering Corporation has estabi'shed the quality assurance policies for the He has delegated the authority for the development of the quality corporation.
assurance program to the Vice President of Quality Assurance, who is responsible for the overall Stone & Webster quality assurance program. As Figure 17 1 shows. the
{
Vice President of Quality Assurance is independent of and has organizational authority equal to the other Stone & Webster Vice Presidents.
The Manager of Quality Assurance, who :eports to the Vice President of Quality Assurance, is responsible for adninistering and managing the quality assurance The Chief Engineer. Engineering program for procurenent and construction activities.
Assurance Division of the Engineering Department. is responsible for administering and managing the quality assurance program for engineering and design.
Major quality assurance activities which are carried out by the quality assurance and engineering assurance organizations of $ tone & Webster aret The review and approval of design, procurement manufacturing, inspection.
(1) construction. and test documents.
Inspection and audits within $ tone & Webster. at suppliers' facilities, and (2) construction sites.
Quality assurance and engineerino assurance personnel have the authority and freedom to identify cuality problems, to initiate, recortnend or provide solutf ons, and to control further processing delivery, cr installation of a nonconforming item until ke e,onclude that the proper disposition of the nonconformance has been approved, 17-1
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SENeGR W3CE PRf $4DENE wtCE PoetstDEfui vtCE PRf $10f 45 ett 20 56 NI T
yggg psig$sogget AND AND OOAttif A35URANCS AseO MANAGER OF
' NC'NE l RINO MANAGEN peRECTOR OF PROJECTS g
CONSTRtsCIION MANAGE 4 OUALITV AtassAGER 05 k
ASSURANCE PURCHA$ tNG
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- AS4NO LONSTRetCilON Rf16Df M T SUPERvesons E NOtNEER l
6.,.ee t# 1 COfwlPANY ORGANtIAleON FOR OUALITY ASSullANCE setne.t A vnte$$ $ P (9sCINE FRIhG CfsHtDRAIRelf i
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Stone & Wettster organization provides sufficient direc;lo o and control of the quality J
assurance and quality control function and that personnel performing these functions j
fn the organt:ation have sufficient authority and organizationni freedom to perform I
their c/itical functions effectively and without retervation.
17.3 gua_1,itz Assurance % ram The quality assurance program appites ta all safety related structures. systems and components within the scope of Sht$$AR.P1. The program commits Stone & Webster to comply with the requirements of Appendia B to 10 CFR Part 50 and to follow the guld.
ance provided by the Cosmission in (1) %idance on Quality Assurance Requirements During Design and Procurunent Ptiase of Nuclear Power Plants.* WASH 1283. Rcrision 1 Maf 24,1974, and (2) %idence on Quality Assurance Requiruments During the Censtruc.
tion Phase of Nuclear Power Plaets.' WASH.1309. hay 10.1974.
The qu:lity.:nrance program for engineering includes the review of applicable an 6 se..- N 't % 1ctions procedures.,pecifiestions and drawings to assure the s* **/1e% remors are citarly, act srately and adequately stated. The program 3
j
.e t
rcqu.re; ths'. desir,n merk be verifie J or reviewed by individuals within the engineer.
W organization not responsible for originating the design and that a deterrination is made that the engineerW3 specifications, procedures, instructions, ae J 3rawings
[
comply with regulatory requireme.cs and design bases.
b i
For procurement control. the quality assurance measures provide fe* the review of procurement documents to assure that the stated quality requirments are adequate.
I for supplier qa lification. and for approval of the suppliers' quality assurance program. The program provides for inspection. surveillance and audit of the suppliers' safety related structures systems and components to assure compliance with precurerent requirements.
I j
During construction the program provides for onsite quality assurance involvement including inspection, nondestructive testing, retention of records. ar.d processing of j
deficiencies. nonconforiaances and design changes. The quality control engineers, f
inspectors and nondestructive testing personnel are organizationally separate and
(
independent 'rca the construction organtration.
I The quality assurance program also provides for a comprehensive system of detailed audits to be performed by the Stone & Webster quality assurance organization. The audits encompass the review and evaluation of all Quality related activities asso.
ciated with the quality assurance program and invcive procedures, work areas. hard.
were, activities, and reconts. The program ret, ires that the audits be conducted in accordance with reeestablished pmcedures by qualified personnel not haring diregt
[
responsibilities in the area being audited. The results are documented and distribut.
ed to the appropriate levels of management.
17 3
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laelementation l.
The Office of Inspection and Enforcement of the Nuclear Regulatory Casatission has
[
I conducted inspecticers to examine the laplementation cf the qualtry assurance program commitments made by Stone & Webster in the 5WE55AR-p1 application to ascertain their l
conformance with the requirements of Appendix 8 to 10 CFR Part 50. These examina.
tions focused on quality assewnce activities related to the design. procurenant. and 1
1 manufacture cf structures. wstems and components for nuclea power plants. and j
'nrie:te ? rpview of establishef proct tures and instructions and the esecution of I
i 1
"[
previsions contained therein.
I f
l Based thereon, the Office of Inspection and Enforcement has determined that there are 1
l no substantive unresobed issues relating to the implementation of the quality 2t f
assurance program which require further identification and followup at this time with I
regard to the SWESSAR-Pl design. We conclude that the implementation of the commit.
monts cf the Stone & hebster quality assur.nce program as described in SWESSAR.PI (by i
l reference to the topical report SW50AP 174) is consistent with ongoing activities in i
l the Stone & Webster Engineering Corporation, i
l Continuing acceptability will be cc:stingent upon Stone & Webster maintaining a sustained satisfactory level of program implementasion which will be vertfled thmugh en ongoing program of periodic inspections by the Office of Inspection and Enforcement.
I 1
17.5 Ccnelusion_
I We find that the quality assurance program described in the Stone & Webster topical i
report SW50AP 174, which is incorporated by reference in the SWE55AR.P1 appitcation.
provides for a comprehen' ave system of planned and systematic controls.
Those controls adequately der-astrate the Stone & Webster ability and cometitment to comply with each of the eighteen criteria of Appendix 8 to 10 CFR Part 50. In addition, we f
have determined that the Stone & Webster quality assurance personnel have sufficient authority. organizational freedom, and independence to perform their quality assurance i
l functions effectively and without undue influence from those organizational elements directly respor.sibit for cost and schedules.
We conclude that the Stone & Webster quality assurance program complies with the requirements of Appendix 8 to 10 CFR Part 50 and is, therefore. acceptable.
l i.
t 17-4
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~18.0 REVIEW 97 tHE hDVl50PY CC994tTTEE ON REACTOR SAFEGUARDS I
i 4
The Advisory Committee on Reector Safeguards reviewed the Stone & Webster appilcation i
for a Preliminary Design Approval for the SWE55AR P1 standard balance-of-plant design f
at its 190th meeting on February 5 7. 1976, in Washington. D. C.
A Subecumitte had l
t previously reviewed the application in asetings on August 1.1975. In Chicago.
)
I Illinois, and on January 22. 1976 in Washington D. C.
At its 194th meeting, on June 3-5.1976 the Committee reviewed SWE55AR-P1 and its relationship to the CESSAR j
nuclear steam supply systes. At its 196th meeting on August 12 14. 1976, the Consit-j tee reviewed SWE55AR-P1 and its r6httonship to the RESAR-35 nuclear stese supply
]
system, i.e.. the SWESSAR-P1/RESAR-35 des 12n combination. The Comittee's report for the SWESSAR P1/RE5AR-35 design coseinstion, dated August 18. 1976 is included in this report as Appendis D.
The report contains certain cassents and reconsendations to the Consission regarding the SWE55AR-P1 application. The report concludes that if due consideratfo9 is given to these coements and racosmendations a Preliminary Design Approval for the EWESSAR-P1 standard bahnce-of-plant design, to be used in conjunc-tion with the RESAR-35 standard nuclear steam supply system design. can be granted.
l We have transmitted the report by the Committee to Stone & Webster for its considera.
tion in proceeding with the SWE55AR P1 design.
The actions we have taken and additional actions we plan to take in response to the j
comments and recommendations identified by the Comittee in its report of August 18 1976, are described in the paragraphs below.
(1) physical separation for Protection Acainst Connon Mode Faflures.
The Cosmittee states in its report the following:
- The arrangenent af SWESSAR-P1 provides extensive physical separation of critical safety related equipment to protect against comon mode failures associated with fires or other operational contingencies. However. complete design details for SWE55AR P1 have not been developed and the concept has not yet been applied to a complete nuclear power plant design. Consecuent'f. further review of the physical separation arrangenent should be swde prior to the Final Design Approval or when SWE55AR-P1 is proposed for a nuclear power plant for which a construction permit is being sought. The conmittee wishes to be kept informed.'
s The staff has investigated the circumstances and consequences of a fire which was experienced on March 22, 1975, at the B1wns Ferry Nuclear Power Plant near Decatur. Alabama. The investigation includes an evaluation of the fire protection systen with respect to potential comon mode failures of safety related systems n
and components. The results of this investigation are presented in a report by l
l 18-1 1
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a special review group of the Consission. entitled 'Recomendations Related to l
Browns Ferry Fire.' dated February 1976. As a sesult of this investigation the
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l staff has developed Standard Review Plan Section 9.5.1.
- Fire Protection System" and Branch Technical Position APC58 9.5-1.
- Guidelines for Fire Protection for Nuclear Power Plants.' tising these new guidelines. the staff soy impose additional requirements on the proposed fire protection system and the physical separation
)
arrangement to further improve the capability of the proposed SWES$AR-Pl design
,I
. to prevent unacceptable damsge that may result from a postulated fire. We have discussed the matter of incorporating these new guidelines into the $WESSAR-P1 j
4 4.1911 cation with Stone & Webster. The NRC staff and Stone & Webster have agreed j
that on the upcoming review of SWES$AR-P1 and its relationship to 85AR-205
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i.e.. the Babcock & Wilces standard nuclear steam supply system. that we will rereview the fire protection system for the SWE55AR-P1 balance-of plant design to the new guidelines and report to the Committee on the r=sults of our review In October 1977. In addition we will eva!uate the integration of the SWE$5AR-P1 fire prttaction system design into the overall fire protection systus for a nJclear power plant fact 11ty during our review of a construction permit applica-tion referencing the SWE15AR-P1 design. Finally we intend to act in accord with the Committee's recommendations and will review further the provisions for physical separation prior to a Final Design Approval and during our review of a construction permit applic6 ton refenncing the SWESSAR-Pl/RESAR 35 design contination.
(2) !atety Related Interfsees The Cosmittee states in its report the following:
- A matter of major concern in the NRC Staff's review has been the safety related i
Interfaces between the SWESSAR-71 BCP design and the RESAR-3$ NS$5 design, on one hand, and the custom-designed site-related structures and compotrus, on the other hand. The responsibilities and requirements related to the $WL55AR-P1/
RESAR-35 interfaces havc been partialle defined in the Safety Analysis Reports for these two standardized designs. TN Cormittee believes that these 'nterfsce requirements are satisfactory for a Preliminary Design Approval, but expects the NRC Staff and the Applicant to continue to examine them further in onection 3
with the proposal to use these designs for a specific plant when 10 4 reviewed a
e for a construction permit. The interfaces between $WESSAR-P1 and the site-related features are defined in the SWESSAR-P1 Safety Analysis Report, but have not yet been subjected to the test of a complete design for a nuclear power e
plant. The NRC Staff should review these interfaces in greater depth when a construction permit appilcation is received."
.{
We have recognised the importance of defining the safety related interface Information necessary to establish compatibility of the $WE$$AR P1 standard e
l talar.ce-of-plant design with the REsAR 35 standard nuclear steam supply system design as well as with the site and utility applicant related features. As 18-2 1
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discussed in Section 1.8. we have concluded that this interface information is I
acceptable for a Prellsinary Design Approval for the $WE51AR-71 eesign. homever.
as part of car long range effort to isprove the lawlementation of the Commission's Standardization policy, we have initiated a dialogue with the nuclear 1;.dustry in an effort to dwelop improved procedures for defining interface requiraments l
I for standard plant designs. Through this effort and additional experience that will be gained in our evaluation of standard plants during the Preiteinary Design Approval stage and during the Final Design Approval stage, we will te able to assure functions 1 compatibility between SWE55AA-P1 and eISAA 35. We will evaluate the implementatier. of the site and utility appittant related interface requirements identified in the SWE15AR-P1 %f'ety Analysis Report during the construction pemit review stage of an application to assure the functional compatibility between all systans and ccaponents of the en? ire nuclear pewt piant.
We may impose additional interface requirements on the $VE15AA-P1 design at that time to assure overall cascatibility. We will report the results of our evalua.
g tica in our Safety Evaluation Report on a constrvction permit application by a utility.ppittant referencing the SWE$$AR-P1 design.
i The NRC staff has pursu~j the matter of the development of safety related inter.
faces between the balance-of-plant and the nuclear steam supply system on a continuing basis. This was evidenced in car Report to the Advisory Comt* tee on Reactor Safeguards for $WE$1AA-P1/Pi$AR-35 in which we f rentifiaj two areas where interface requirements have been developed by Stone & Webster and Westing.
house with NRC staff guidance, i.e.. Structural Interfaces and Seismic Inter-I faces. These interfaces have been developed as a result of the NRC staff gaining esperience with the standardization process and have been irplemented into our review in a timely asmer. We will continue to do so.
(3) Interdisciplinary trstem Analyses The Cosmittee states in its report the following:
i
- The Committee roccimends that, during the design, procurement, construction.
e and startup, timely and appropriate interdisciplinary s/sts analyses be per-formed tJ assure coeplete functional coppatibility acrvss each ir.terface for the entire spectrws of anticipated cperations and pestulated design basis accident conditions."
s The SWE35AR P1 application identifies the specific design responsibilities of l
nuclear stese supply system vendors. $ tone & Weoster and a utility applicant for
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certain systems o* a nuclear power plant based on the SWES$4P14esign.
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$WE55AA-P1 furte.er states tnat Stone & Webster retales the responsibility for
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ensuring the ir.tegration of the various systeas into a functioral plant.
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We will require a utility applicant referencing the SWE$$AR-pl/RESAR-3$ dasign combination to provide in its appilcation for a construction permit and/or operating license the appropriate analyses of the entire facility incorporating j
the SWE$$AR-pl/RESAR-35 design combination to assure the complete functional j
compatibility for the entire spectrum of anticipated operations and postulated l
design basis accident conditions. The responsibility. in fact. rests with the uttitty app 11 cant do may elect to assign to Stone & Weter certain integration I
tasks relating to that responsibilit).
l (4) Interdependent instruentation and Controls i
The Committee states in its report the following:
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- The coordination of interdependent instrumentation and controls in the nuclear i
island and in the balance of plant will require attention ist the time when SWESSAR p1 is used as a portion of a nuclear power plant license application.
These metters should be included in the NRC Staff's standard review plans.*
i i
l As stated in itan (2) above, we have initiated an effort to develop improved l
procedures for defining and isolamenting interface requirements between various i
I portions 4f a nuclear poser plant. typically a standard nuclear steam supply system. a standard balance-of-plant (esign and site and utility app 11: ant related design and operation aspects. The results of the effort may lead to specific l
review procedures for standard designs. As an initial result of our efforts we f
developed. during cur review of the RISAR 3$ appitcation, a matrix for the
)
applicability of design criteria at the interfaces of instroentation and control systems in RISAR-3$. which is included in our AISAR 35 Safety Evaluation Report I
At our request. Stone & Weoster incorporated in an appropriate manner this i
matrix in the $WESSAR-pl appilcation. We consider that these matters are already included in a general way in our standard review plans. We intend to review I
these matters specifically in the context of our reviews of standard plants.
Appropriate revisions to our standard review plans will be made on the basis of I
OLr esperience and need.
($) Turbice Generator Orientation fo, A;1tiple Unit Power plaats
)
i The Corinittee states in its report the following:
"The proposed orientation of the turbine-generator with respect to the nuclear l
lsland is suttable for a single unit installation, for multiple unit power plants, the location and orientation of the units should be such as to y. eld acceptably low probabilities of damage by low-trajectory turbine-generator sass 11es, or suitable missile shielding should be provided."
Inis report presents our safety evaluation fo 2 5'al55AR p1 single unit nuclear power plant facility in accordance with the ShT55)R-P1 appilcation. We wi'1 18-4
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evaluate the aspects of a sulti-unit nuclear power plant facility during the review of an application by a utility applicant referencing the $WE55AR-P1 f
design for a multi unit facility, la particular we will evaluate at that time 1
the infomation provided by the utility applicant regarding the location and j
orientation of the units to assure that the probability for daange from lon e
trajectory turbios generator missiles is acceptably low.
e (6) Reliabiltty Assessment of Two-Track Continuous Duty Systems e
The Committee states in its report the following:
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- The $WE55AR-P1 and the RESAR-35 M555 designs. as do seny others. vt111re the concept of two-track continuous duty systems such as ventilation and service unter which perform critical service functions. In some cases the probability of failure of one o< these systems is not low. The failure of the second system e
to start or conting to operate may cause progressively damaging consequences.
The Committet -ecuenends that failures of this kind be.rvaluated to determine if the necessary reliability exists for these systems and whether remedial measures e
are appropriate.'
+
In a letter dated July 14. 1976 from D. W. Moeller. Chairman of the Advisory Committee on Reac+.or Safeguards to L. V. Gossick. Esecutive Director for Operations for the U.S. Nuclear Regulatory Cornission, this subject of reliability assessacnt e
of continuous duty two-track systems is discussed as item (3) of the letter.
This letter and the recoseendations contained therein have been reviewed by the f
Office of Nuclear Regulatory Research and the Office of Nuclear Reactor Regulation and have been responded to in a letter dated July 28. 1976 from L. V. Gossick.
e Executive Director for Operations for the U.S. Nuclear Regulatory Coenission to D. W. Moeller Chairman of the Advisory Cosnittee on Peactor Safeguards. This 1etter outilnes the cooperative plans of the Office of Nuclear Regulatory Research and the Office of Nuclear Reactor ttegulation in developing a program to test e
ways in which the reliability methodology used in the Reactor Safety 5tudy.
e WASH-1400. can be used to improv. the NRC regulatory process, taking into account the need to strengthen the presently limited capabilities of the NRC staff and industry to apply this methodology on an efficient and consistent l'esis. Various e
apprtaches 4re being considered, along with the identification of specific e
topics (including reliability assessment of continuous duty two-track systems) that could be investigated using the methodology. A program of this importance e
and magnitude must be carefully planned and the cc snents and recorrnendations e
made by the Connittee on this subject provides confirvation of tne importance to g
be attached to this effort. The NRC staff will keep the ACR$ informed o# the
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e secpe and direr.tlen of this program as it is developed.
The matter of reliability as can te related to operability ard maintainability e
has been discussed with Stor'e & Webster. StMe & Webster's stated plant availa-
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bility goal for tie SW155AR.P1 design is 66 perce t.
To achieve this goal.
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I Stone 81debster has revleued their balance-of-plant design for equipment layout j
to assure adequate space exists for maintenance including complete reseval and
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replacemen' of equipment if required, as well as shielding requirements which l
t would influence plant meintainability during normal plant operation. While the l
NRC staff has not required nor have we revleued any reliability analysts for the j
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$WE5!AR-P1 design. Stone & Webster has stated that balance cf-plant safety I
related fluid systems have been analyzed for derndability. i.e., the probability I
that a system will function successfully for a specified period of time when called upon at a rIndom point in time. Stone & liebstet believes that an inadequate data base currently exists with dich to ferfore deperdability analyses which result in a high degree of confidence. Their analyses have generally indicated that the inherent system redundancy of safety related fluid systems (e.g..
separate flow paths. redundant active components) which provide parallel success s
paths to achieve a safety
- unction is a dominant factor in the achievement of i
their dependability goals based upon the methanatical models they have utilized.
l l
(7) Protection Acainst Industrial $abotage The Committee states in its report the follo.ing:
f
- Although $WESSAR-P1 and RE$AR-35 include provisions for protection against i
industrial sabotage. the Consittee believes that further steps can be taken beyond those provided. prior to the use of SWEs$AR-Pl/RESAR-3$ as a portion of l
an application for a nuclear power plant license tM utility-applicant should be required to demonstrate that acceptable industrial sabotage provisions will
[
be incorporated into the plant design."
t Studies concerning possible modes of sabotage at nuclear power plants are being conducted under contract for the Office of Naclear Regulatory Research of the i
Comission. Any recomendations resulting from these studies. regarding addi-tional design features to protect against acts of industrial sabotage, will be considered by the staff for implementation in the SWE55AR-P1 design. We also will require a utt11ty applicant referencing the SWE$$AR-PI design in a construe.
tion permit application to demonstrate that acceptatle provisions for protection against industrial sabotage are incorporated into the plant design.
9 (8) Maintenance. f aspection and Ocarational Ne+es The Coemittee states in its report the following:
v "The SWE15AR.Pl design innudes some provitices which anticipate the rateteeance.
i l
Inspection, and operational needs of the plant througNut its service 11 including cleaning and decontamination of the primary coolant system, and eventual deconmissioning.
Ha ever, when Sm[51AR-P1 is used as a portion of a w
nuclear power plant Itcense app 11 cation. the C:rritttee telieves that the hRC staff and the utility applicant should further review methods and procedures for 16-6
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removing accumulated radioactive contamination whereby maintenance and inspection p.ograms and ultimate decomissioning can be more effectively and safely carried out."
During the past year, the staff has been revbtng radioactive contamination i
problems. including data on occupational radiation esposures related to activa.
tion products and methods and procedures for preventing or reducing and removing accumulations of radfoactive contamination in 3Ae reamr coolant system of Ilght water reactors. Information gathered at conferences on decontamination and decomissioning and from specific technical sources In industry have resulted in the staff examining this issue in more detail during the review of nuclear oower plant applications.
The draft working paper for Regulatory Guide 8.8. Revision 2. "Information l '
l Relevant to Assuring that Occupational Radiation Exposure at Nuclear Power i
Stations Will 8e As Low As !s Reasonably Achievable." which was reviewed by the industry in April 1975. has incorporated all of the staff's findings to date on methods and procedures that are effective in reducing radioactive exposure related to activation products. N areas we have identified where information is lacking or insufficient, are the costs associated with various measures and the quantitative benefits in reduction of occupational radiation exposure associated with the measures. Because we do not have this information, we do not require that additional design features for exposure reduction be imple.
mented in plants presently ur. der review.
We are continuing our study of the problems associated with decomissioning, but i
as yet we do not require specific usign provisions for decomissioning. A few reactors have been decomissioned and w know from this experienr* that the resultant exposures can be kept witMr. acceptable bounds Because the expertence in this area has been acceptable to date, we plan to continue our investigations further into this ir.atter before we require that any special design features related to decomissioning associated radiation exposure be incorporata duri.3 a
the design and construction of a plant.
(9) Generic Problem Related to tare Water Reacters The Comittee states in its report the following:
- Generic problems related to large water reactors are discussed in the Comittee's f
report dated April 16. 1976. Those problems relevant to SWE55AR.P1 and RESAb35 should be dealt with appropriatelf by the NRC Staff and the utfif ty appittant as solutions are found. The relevant items are: Group !!. Items I. 2. 3. 4, 5 6, 7. 9, 10. 11; Group !!A - Items 1. 4. 5. 6. 7. 8; Group I!8 - Item 2; Group
!!C. Items 1. 2. 3. 4. 5, 6. 7."
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The generic problems in general identif'ed by the Cassittee in its report of l
g April 16,1976, and the specific itens identified in particular are'the subject of ongoing evaluations and analyses by the staff, various reactar vendors. and cther associated Industries and organizations. They are the subject of con-i tinuing attention by the staff. Based on the conclusions of thes6 ongoing l
l studies we may impose, to an extent that is both reasonable and practicable, t
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additional requirements on the SWES$AR.P1 design. The safety evaluation report I
at the operating license review stage will document any departures, and the i
bases for these departures. free the generic resciutions.
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e 19.0 CONCLU5109t$
i Based on our analysis of the proposed SWESSAR-P1 standard balance-of-plant design.
we have reached the following conclusions, subject to the conditions discussed in I
this report, for the portion of a nuclear power plant design covered by the SWE55AR-P1 design:
f (1) The Stone & Webster Engineering Corporation has described. analyzed and evaluat.
ed the proposed $WE55AR-P) design including, but not limited to. the principal architectural and engineering criteria for the design; the interface information necessary to assure compatibility between the submitted SWES$AR-P1 standard balance-of-plant design and (1) a RE$AR-3$ standard nuclear steam supply design and (11) the site and utt11ty applicant related design aspects; the envelope of site perameters pastulated for the design; t'.e quality assurance program to be applied to the design, procurement and fabrication of safety related features of the SWESSAR-P1 design; the design features that affect plans for coping with emergencies in the operation of the reactor or major portion thereof; and has identified the major features and components incorporated therein for the pro-tection of the health and safety of the public.
(2) Such further technical or design information as may be required to conelete the safety analysis will be 'upplied prior to or in the final design application.
(3) On the basis of the foregoing there is reasonable assurance that: (1) safety questions will be satisfactorily cesolved at or before the issuance of the operating license for the first nuclear power plant uttllting the SWE55AR-P1/
RESAR-35 design combination; and (ii) taking into consideration the site criteria contained in 10 CFR Part 100, a facility can be constructed and operated without a
undue risk to the health and safety of the public, provided the site character.
istics conform to the site parameters specided in SWESSAR-P1 as discussed above and otherwise conform to the 10 CFR Part 100 requirements, and provided further that the site and utility applicant related systems of the nuclear powve plant are properly designed and constructed in conformity with the interface requirements specified in SWESSAR-P1 and in this report, as discussed above.
(4) The Stone & Webster Engineering Corporation is technically qualified to design the proposed SWESSAR-P1 standard balance-of. plant.
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l APPENDIX A i
MON-$TANDARD AND Orff0NAL $YSTEMS IN $WES$AR-P1 1
Sections in this appendix are net,ered in the same manner as appropriate sections in the main body of this report.
A 9.0 AUXtLiApri $YSTEMS 1
A 9.3 Process Auxiliarfes i
A 9.3.1 Boron Recovery System Amendment 1 to WASM-1341. "Prograsmatic Informatu for the Licensing of Standardized Nuclear Power Plants.* defines the standaid scope of standard design applications.
The boron recovery system is identified as a system within the scope of a standard nuclear steam supply system design application. Stone & Webster has replaced the PisAR-3$ boron recoyny system (RE$AR-35. Section 9.3.6. Boron Recych $ystes) by the SWESSAR-P1 boron recovery systen ($WESSAR-Pl. Section 9.3.6. Boron Recovery System).
The boron recovery system s non-safety related system, will collect and process the reactor coolant letdown flow from the chemical and volume control systas, a RESAR 35 f
system, for the recovery of the reactor coolant. and reuse or disposal of boric acid.
In order to accorsodate the Stone & Webster designed system and the radioactive gaseous waste system, which is also within the $WESSAR P1 scope of design. Stone &
Webster has modified the appropriate interface connection points between the RESAR 35 cassical volume and control system and the $WES$AR-P1 boron recovery system and radioactive gaseous waste system.
Stone & Webster has identified in Table '. 3.6-5 of $WES$M-P1 the specific require.
ments of the RE$AR 35 boron recovery system and has addressed these as interface requirements for the $WESSAR-P1 system.
We have reviewed the $WES$AR-P1 boron recovery system including the aspects of the replacement of the RE$AR 3$ boron recovery system by the SWES$AR-P1 system. We
+
conclude that the design bases and criteria for the $WEs5AR-71 bcron recovery system, including design changes to the RESAR 35 chemical volume and control system.
have no adverse effect on the RESAR 35 designated safety functions of the RESAR 35 We also conclude that the interface requirem nts established by the RESAR 35 system.
chemical volume and control system on the boron recovery system and the radioactive gaseous waste system will be met by the $WESSAR Pl design.
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A 11.0 RADI0 ACTIVE WASTE MANAGEMENT j
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I Stone & Wanter included in.the SWE$$AR.P1 design (Sectie 11.2.2.1 of SK$5AR-pl) provisions for processing laundry wastes. The syste is non-Safety related and j
is provided as an option in SWESSAR.pl that may be selected by a utility appilcant i
referencing the SWESSAR.pl design.
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Laundry wastes will be collected in the laundry weste drain tank and will be pro-i cessed by the radioactive liquid waste system (see Section 11.2 of this report) in a separate processing train using a laundry waste evaporator. The design flow capacity for the evaporator is 2900 gallons per day (2 gallons p r minute) as listed in Table A 11-1, which also lists the capacity of the laundry watte drain tank and the quality classification for these components. We have calculated the average expected waste flow to the evaporator to be 450 gallons per day. The difference between the design and expected flow capacity for the evaporator will provide adequate reserve capability for processing of surge flows. We estimate that 100 percent of the evaporator distillate from the laundry wastes will 14 discharged. We consider the design and the capacity of the radioactive liquid weste systen with respect to the laundry wastes to be adequate for meeting the demands of the plant during any anticipated operational occurrences.
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i DESIGN PARAMETERS F(4t PRtHCIPAL CtwPMNTS OF LAUNDRY WASTE SYSTEM _
Capacity Quality )
Grouo(a M etee_
tach Systen Radioactive Liquid Waste System l
1 4000 gal D
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(a) Quality Group D design criteria include additional quality assurance provisions in accordance with Branch Technical Position (T58-11-1. Revision 1.
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6 AP9EN01x a f
CHRON0t0GY OF RfVIEd 0F.
SWESSAR-P1 RETTRENCE SAFETY ANALYSIS #[ PORT This appendix presents a chmnology of the principal actions during the processing of the I
I SWESSAR P1 appilcation with respect to all nuclear steam suoply systems referenced in i
SWE55AR-P1. The following abbreviations are used in this appendix:
?
5 tone & Webster Engineering Corporation.
SW United States Nuclear Regulatory Commission.
NRC Office of Nuclear Reactor Regulation, NRR Advisory Committee on Reactor Safeguards.
i ACRS Westinghouse Electric Corporation.
W t
Combustion Engineering. Incorporated.
f CE BW Babcock & Wilcox Company, t
B0P Belarce-Of Plant.
Nuclear Steaa Supply System.
N555 Preliminary Design Approval.
PDA SIR Safety Evaluation Report.
October 30. 1973 Meeting with S&W: standardigstion policy. S&W intent to sutait standard B0P design.
i I'
December 20, 1973 Meeting with $&W: arrangement and layout of proposed Standard BOP design.
March 13, 1974 Meeting with S&W: seismic design and structural analysis of proposed standard 60P desfgn.
April 4.1974 Meeting with S&W: content of proposed standard BOP design.
review procedures.
i SWE55AR P1 application consisting of standard BOP etign April 25,1974 and interfaces with RESAR-41 standard N555 design tendered by S&W for acceptance review.
Aprl) 26. 1974 Letter to S&W: acknowledgement of appilcation and schedule for acceptance review.
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Mays.1974 Meeting with SW: presentation of SE55AR-P1 concept by SW.
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i' May 7. 1974 Letter from SW: copyright statement f e 5WE55AR-Pl.
June 3. 1974 Letter to SW: $WE55AR-P1 application acceptable for docketing, request for infocustion based on acce">tance l
review.
June 21. 1974 Meeting with SW: discussion of MRC requests for additional information in letter of June 3.1974.
June 28,1974 Letter from SW: sutuilttal of SESSAR-Pt application i
for PD4 review.
June 28, 1974 SWESSAR-P1 appilcation docketed.
June 28. 1974 Letter from SW: schedule for response to items in NRC letter of June 3.1974 July 11, 1974 Letter to SW: notification of docketing and copy of Federal Register notice.
July 30,1974 Submittal of Amendment 1: partial responses to items in NRC letter of June 3.1974 and interface information 'or SWE55AR P1/8-5AR-241.
August 2. 1974 Letter to S&W: review schedule for SWE55AR-P1/RESAR 41 i
(Schedulef).
August 9. 1974 Letter from S&W: revision of response schedule in $&W 1etter of June 28, 1974 August 30,1974 Submittal of Amendment 2: additional responses to items in NRC letter of June 3.1974 September 12. 1974 Letter to 5&W: requests for additional information and response to staf f positions.
t September 13. 1974 Letter from S&W: revisions to Asendment 2.
l 9
September 20, 1974 Letter to S&W: requests for additional information.
l l
October 1.1974 Meeting with S&W: discussion of items in fiRC letter 4.f Septercer 12, 1974.
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October 10. 1974 Meeting with $ W: discussion of items in NRC letter of
- 1 October 11,1974 September 12. 1974, discussion of forwat for interface 1
information.
October 11. 1974 Letter from S W: schedule for responses to NRC letters P
September 12. 1974, and September 20,1974.
i r
October 21. 1974 Submittal of Amendment 3: interface information for SESSM-P1/C15$AR. changes and additional information
)
to SE SSAR-P1.
j October 25. 1974 Meeting with $W: need, type and forest of interface information, scope of standard NS$$ and 80P designs, j
i November 1.1974 Sutalttal of Amendment 4: responses to mejority of iters I
in MAC letters of September 12. 1974, and September 20. 1974.
November 4.1974 Letter to S W: review schedules for SESSAR-Pi/8-5AR-241 (Schedule 11) and $WESSAR-P1/CESSAT (Schedule !!!).
November 12. 1974 Letter to SW: potential slip for SWESSAR-P1/RESAR-41 Schedule 1.
I December 2. 1974 Submittal of Amendment 5:. completion of responses to NRC letter of June 3.1974. additional responses to NRC letten of September 12,1974, and September 23. 1974 Decewber 17. 1!74 Meeting with $W: discussion of SW responses in Amendmnts 4 and 5. discussion of additional reqsests for information.
December 18. 1974 January 7.1975 Meetin9 with S&W: discussion of additional requests for information.
January 15,1975 Letter fran SW: request ta continue review of SWE55AR-P1 with respect to BW standard NS$$ design (8 SAR 241 review was terminated by NRC on request by 8 W.
B 5AR-205 to be submitted).
Ameneent 1 to WASH-1341 defining scope of January 16. 1975 tetter to S W:
standard NS$$ and BOP designs in stardard safety analysis l
f reports.
l January 17. 1975 Letter to S&W: request for additional infonnation and response. to staff positions suspension of Schedule !! for I
revien of f455AR-P1/8-SAR 241
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January 23. 1975
$utnittal of knendment 6: update of !'s!5$AR-P1.
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January 27. 1975 Letter to S&W: request for a'dditional information and response to staff positions.
i February 6.1975 Letter to S&W: request for additional information and response to staff positions, revision of review for $WESSAR-Pl/RE$AR-41 (Schedule 1).
Fetwwary 7,1975 Lett3r to $&W: request for revision of SWES$AR.P1 in accordance with Regulatory Guide 1.49 regarding maximse licensed power level.
February 7.1975 Letter fr a S&W: acknowledgement of MRC le'.ters of January 17. 1975, and January 27,1975.
s February 13. 1973 Letter from $4W: request for exemption to WASH.1341 Amendment 1.
March 5. 1975 Sutaittal of Amendment 7: response to majority of requests in NRC letters of January 17. 1975 January 2'. 1975, and February 6. 1975, schedule for remaining responses.
March 31. 1975 Sutaittal of Amenenent 8: additional responses to requests in NRC letters of January 17. 1975 January 27,1975 and j
February 6.1975. revision of Sk"$5AR-P1 for 3800 MWt licensed power update of SWESSAR-Pl.
April 14. 1975 Letter from $1W: schedule and status of additional infonne-tion on contairment analysis requested in NRC letters of j
5eptember 12. 1974, and January 27. 1975.
i' April 16.1975 Meeting with $&W: discussion of status for containment
[
analysis responses slip in SWESSA2-P1/RESAR-41 schedule.
status of Staff review.
l l
April 21. 1975 Letter from $&W: advance copy of partial Amenenent 9
.(contairment analytis).
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April 25, 1975 Letter to S&W: request for additional information and response l
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to staff positions regarding electrical. Instrumentation and i
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control systems on SWESSAR-P1/CESSAR relationship.
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April 30.1975 Submittal of benenent 9: inclusion of information sutnitted l
by letter of April 21. 1975, additional interface information on SWESSAR.P1/CESSAR. update information.
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April 30.1975
' Letter from $&W: advance copies of responses to remaining requests for information regarding containment analysis.
May 1. 1975 Letter to S&W: request for additional informatica egarding industrial security.
May 5. 1975 Letter from $&W: projection of additional standaro N555 desi9ns to be included in SWE55AR."1.
May 5. 1975 Letter to SW (r.rsoonse to 5&W 1etter of Fe'>ruary 13.1975):
remsting interface informati:n on WASH-1341. Amendment 1 i
optional items.
4 May 8. 1975 Meeting with 5&W: discussion cf outstanding items in review of electrical, instrimentation. and control systems.
May 14. 1975 Submittal of Ariendmens 10: additional responses on contalmeent analys's items partial responses to items in NRC letter of April 25. 1975 update of SWESSAR-P1/CESSAR information, j
May 21. 1975 Meeting with S&W: discussion of interface information on flow diagrams.
May 29. 1975 Meeting with S&W: discussion of additional requests for f
-- information on contairment analysis for $WE55AR.P1/RESAR.41.
June 2. 1975 Suisnittal of Amenement 11: remaining responses to requests in NRC letter of April 25. 1975. update of SWESSAR.Pl.
- f..
Joe 3, 1975 Letter from S&W: additional information on industrial I
security in respoise to NRC letter of May 1.1975.
k June 10.1975 Letter from S&W: response to NRC letter of May 5.1975 regarding exraptlw to WASH-1341, mendment 1. list of SWE55Alt.P1 interface changes for options, i
June 10,1975 Letter from S&W: request for additional evaluation of supplementary leak collection and release systen. In SWESSAR-pl.
I June 10.1975 Letter to S&W: req *et f ar addittoral inforeation and response to staff positions re%. ding containment analysis itms j
discussed in meeting of N y 29,1975.
June 13. 1975 Sutmittal of Ameronent 12: update of interface inforvaticn.
j partial respor.ses to items discussed in sieeting of May 8.1975 S&W position on future updating of interface information.
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h ou 4 M. 1975 Letter frta SW: advance responses to itms in imC letter of r
June 10.1975.
Jul) 3.1975 Sutaittal of Ameet 13: update of SE55AR-P1 flow diagrams i
and interface points.
i i
Aly 10. 1975 Meeting with SW: gener4 discussion of interfaces with HrJt management. S W proposal for task force on interfaces.
L N < 17, 1975 Satuitta) of Amenteent 14: update of SE55AR-71 and responses.
to outstanding issues discussed in meeting of May 8.1975.
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August 1.1975 First ACR5 Subcamoittee seeting on SWE55AA-P1/RESAR-41.
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August 5.1975 Meeting with SW: status of response te outstanding items i
discussed in meeting on May 8. 1975.
August 7,1975 Letter to SW: request for additional infomation regarding fire protection.
August 8.1575 Sahmittal of Awormt 15: update of SESSARJ1. partial response to ovutanding items. partial responses to requests for iM 'mation in RAC 1etter of June 10, 1975.
i August 8.1975 Letter to SW: request for additions) inforv.ation regarding opticnal enclosure buildings for SE55AA-P1 design.
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August 15. 1975 Letter to S W : list of items discussed in meeting on j
August 5.1975. reovest for response schedule, rewest for
~
SWtSSAA-P1 interface infomation update.
Ausust 18, 1975 Letter to SW (response to SW 1etter of June 10. 1975):
l reqwest for identification and oescription of SWE55AR-71 itens f,
that replace corresponding RESAR-41 items.
I 8
August 20,1975 Letter to SW: request for additional infomatiJn and response l
to staff requests regarding electrical. Instrumercation and I
control systees on SWE55AA-71/C(55AA.
Agust 22, 1975 Letter to S W: evaluatica of prepesed technical specifications for containnent leak rate testing.
l Asgust 29, 197$
Sutaittal of Amentsment 16: InfCmation regarding open items identified by staff at AC15 Subc:rwittee reeting on August 1 l
1975, and identified in eeeting on Asgust 5.1975; partial I
respenses to itees in UC letter of Agust 7 and S.1975.
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. September 8.1975 Letter fmm $W: additional inforestion on reactor cavity pressure analyst:.
September 10. 1975 Letter fran S W : forthcoming submittal of amendment to include Westinghouse standard NS$$ design RESAR.3$ in S E S$AR.P1 applicationt additional information on containment leakage testing and identification of bypass leakage.
September 15. 1975 Letter from S W: responses to NPC requests of August 20 1975 for additional information regarding electrical, Instrumentation and control systems for $WESSAR.P1/CE$$AR design combinationt additional information regarding subcampartment pressure analysis.
I October 2.1975 Submittal of Amenenent 17: interface information for i
SWE$$AR.P1/RESAR.35 design combinatios,t additional information on SWES$AR.Pl design.
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October 3.1975 Letter from $W: response to NRC letter of August 18. 1975 4
reoarding identification and description of SWES$AR.P1 l
itsets that replace corresponding RESAR.41 items.
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Octotee 9,1975 NRC letter to ACR5: transmittal of advance copies of j
Deport' to ACRS on SWES$AR-P1/RESAR.41.
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l October 9. 1975 Letter from SW: information regarding update of interface infs:Ntion in $WE55AR.P1 as related to RI5AR 41 and CESSAR.
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October 10. 1975 Report to ACRS on SWESSAR.P1/RESAR 41 issued by NRC.
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October 16, 1975 Letter to SW: transmittal of Report to ACRS on f
SWES$AR.P1/RESAR.41.
I October 30, 1915 Meeting with Su: discussion of $W and hRC por.itions on l
SWESSAR-P1 partial secondary containment.
t October 31, 1975 Submittal of Arnenenent 18: additional information on k
various issues of $WES$AR.P1 design.
i October 31, 1975 Letter to SW: Identification of outstanding issues in Report to ACR$ and request fo* additional information on these issues.
Noverber 7.1975 Letter frta SW: additional infomation on cornputer codes used for dynmic analysis of soil structure interaction.
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november 11. 1975 Letter fran 54sf: forthcoming subsittal of amenement to include SW i
standard NSS desi n B-SAR-205 is SE55AR P1 :;;,iitation; request 4
9 for review schedule for $ES$AR-71 application.
j i
Novutne 17, 1975 Letter free SW: additional information on partial secondary coretairme.nt requested by NRC in meeting of October 30. 1975.
Nov a ber 17. 1975 Meeting with SW: discussion of outstanding issues mov eber 18. 1975 (electrical. Instruentatioe ed control systems) identified j
in Report to ACR5.
i
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4 Noveber 21, 1975 Letter fram 5 m : response to anc letter of October 31. 1975 inc1 Wing responses to outstanding issues (1) through (19) j in Report to ACR5. Section 1.9.
j hovember 26. 1975 Letter from Sau: response to pitC letter of Octoter 31. 1975 regarding outstanding issues on electrical. instrumentation l
l and control systems in Sections 7.0 and 8.0 of Report to ACR$.
j December 5.1975 Meeting with SW: discussion of leakage test requirements for $ESSAR-P1 partial secondary containment.
Dec eber 16, 1975 Letter from S&W: additional generic informatitn on computer codes regarding soil-structure interact'on.
I i
December 19. 1975 Sutaittal of Amendment 19: interface information for j
SWE55AR-P1/8-5AR-205 tesign combination which supersedes
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previous SESSAR Pl/8-5AR 241 ccubination; inctnlon of I
i information from S&W 1etter of hovember 26. 1975 into SE SSAR-P1 SAR.
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January 16. 1976 Supplement No.1 to Report to ADt3 on SWE55AR-P1/RESAR-41 issued by NRC.
January 16. 1976 Letter to S&W: transmittal of review schedule for j
SWESSAR-P1/RISAR-35 design can.biretion.
January 16, 1976 Letter from $&W: revision of coroonent cooling water systen f
for SWE55AR-Pl>RI5AR-41 design ccobination in response to l,
outstanding issue in Report to Acts; revision of auallf ary feed =ater system design for $W 55AR.Pl/Ct55AR.
January 22, 1976 second ACR5 59beernittee meet'au; on SWE55AR-P1/RESM-41.
January 26. 1976 Submittal of Are~snent 20: ecsttional information for SWE55AR-P1/8-5AR-205; documentatten of 5&W r sponses to seae outstanding issats in Report to *425.
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February 11,1976 Letter from ACR$: Report by the ACRS on SWESSAR-P1 (See Appendfa C).
j 1
i February 17. 1976 Ls'.ter to S&W: trenilkistal of Report by the ACR5 on 1
1 SWESSAR-P1.
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February 19. 1976 Letter feta S&W: proposed revision to SWE55AR-71 to provide increased design load rejection capability.
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j February 24, 1976 Submittal of Amendment 21: additional documentation of S&W responses to outstanding issues in Report to ACR$;
SWE55AR-P1 changes related to RISAR-35 and B SAR-205 designs.
1 February 24,1976 Letter to S&W: transmittal of requests for additional l
information on the 5WE55AR/RESAR-35 design combination.
March 5.1976 Letter frem 5&W: responses to NRC requests of February 24 1976. for additional information on SWE55AR-P1/RESAR-35.
I March 12. 1976 Letter from W : requirements for review by ACR5 of
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SWE55AR-P1/CESSAR design combination.
]
Letter from S&W: transmittal of information on harch 12. 1976 l
SWE55AR-P1/RESAR-41 energency boration system and contalsment backpressure calculation.
Mare,h 17,1976 Meeting with S&W: discussion by S&W and NRC management of schedule for issuance of SER and PCA for SWESSAR P1/RESAR-41 i
and SWE55AR-P1/CE55AR degign combinations.
March 18.1976 Submittal of Mendment 22: documentation of j
SWESSAR-P1/RESAR 35 responses in letter of March 5.1976; I
revisions for SWE55AR-P1/RESAR-41 and SWIS$AR P1/CESSAR design combinettons, l
Nrch 19.1976 Letter to S&W: confimation of schedules discussed in i
~
l meeting on March 17, 1976; identification of outstanding l
issues on SWE55AR-P1/RESAR 41 and SWE55AR-P1/CE55AR design combinations.
1 I
l March 26, 1976 Meeting with S&W: discussion of open items in SWE55AR/RESAR-35 review.
March 30,1976 Sutenittal of Me+ent 23: responses to outstanding issues ideritified in NRC letter of March 19. 1976.
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gil 8,1976 Letter fras S&W: status of resolution of outstanding issues j
identified in NRC letter of March 19, 1976.
I i
i-e April 9.1976 L a ter to S&W: change of review schedule for SWE55AR-P1/Rl5AR 35.
t i
April 16,1976 Letter frem 56d: withdrawal of proposed revision in S&W 1etter i
of February 19, 1976 for SWE55AR-P1 increase design load j
i rejection f apability.
i i
s; April 19,1976 Letter from S&W: proposed revision of SWE55AR-pl/CE55AR l
contairment spray systen in response to NRC letter of l
i March 19,1976, and as discussed in maeting with S&W on March 26,1976.
j l.
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April. 22. 1576 Meeting with 5&W: discussion of issues related to SER for I
SWE55AR-P1/RESAR-41, of component esoling water system design for reactor coolant peps, and of ultimate heat sink interface requirements.
April 28. 1976 Sutaittal of Ameneent 24: documentation of additional information requested by NRC on SWE55AR-F1 as ref at*f. to RESAR-35 (main steam line break analysis, structural interfaces.
contatrument spray systan) and CESSAR (identification of reactor l
coolant system breaks in reactor cavity, analysis of roo ejection accident. contalment spray system); additional infor-mation on containment tailding polar crane.
April 30,1976 Submittal rf teendment 25: documentation of additional information on SWE55AR-PI/RE5AR-41 as requested by NRC and discussed 11 meeting on April 22. 1976.
May 5. 1976 HR issuance of Preliminary Design Approval No. PCA 4 for SWE55M P1 design and its relationship to RESAR-41.
May 5.1976 NRR issuance of Safety Evaluation for SWE55AR-P1 design and its relationship to RESAR-41 design.
May 5.1976 Letter to S&W: transmittal of PDA-4 and SER (**
SWE55AR P1
~
design and its relationshp to Rl5AR-41 cesign.
May 1'8. 1976 Letter to S&W: request for additional information for the SWESSAR P1/CESSAR review.
June 2. 1976 Suomittal of teendment 26: additional documentation of S&W response to centainnent spray system design, combustible gas control, and reactor cavity analysts.
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June 3,1976 ACR$ meeting on $WES$AR-P1/CESSAR.
E June II.1976 Letter f on ACR$: Report by the ACR$ on SWE$5AR 81/CESSAR.
i June 30. 1976 Submittal of Anenenent 27: additional information for SWE15AR P1/
RESAR 35 main steam line break analysis. Staff report to ACR$ on SM SSAR-Pl/RE$AR 33 issued.
July 9.1976 Meeting with $4W: discussion of outstanding issues for SWE$1AR-Pl/
RESAR-3$ (main steam Ifne break analysis) and SWE$5AR-Pl/CES$AR I
(reactor cavity analysis).
i August 6. 1976 Submittal of Amendment 28 add'tional documentation for SWE15AR-Pt/
CE$$AR (reactor cavity analysis) and $WE55AR-P1/RE$AR 35 (main steam linebreak).
August 12. 1976 ACR$ meeting on SWE$$AR-Pl/RESAR 33.
L August 18.197C NRR issuance of Preliminary Design Approval No. PDA-6 for SWESSAR P1 design and its relationship to CES$AR design.
4 August 18,1976 NRR issuance of Safety Evaluation for $WES$nR P1 design and its f
relationship to CIS$AR design.
August 18, 1976 Letter frw ACR$: Report by ACR$ on SWE$$AR-P1/RESAR 35.
September 10. 1976 Letter to $4W: transatttal of $WE55AR-Pl/BSAR 205 review schedule.
Septem*>er 21,1976 Letter from $4W: ' requesting reevaluation of the $WE$$AR-P1/BSAR-205 review schedule.
September 21. 1976 Letter to $8W: design options for $WESSAR Pl. Sutnission of the boron recovery system as a topical report.
September 30, 1976 Letter to $4W: reevaluation of fire protection provisions for
$WES$AR-P1 October 5, 1976 Letter to $8W: reevaluation of review schedule for SWES$AR P1/85AR-20$
after review of Round One Responses.
October 29, 1976 Submittal of Amendment 29: additional information for SWESSAR Pl/
RESAR 35 review and update of $WE5$AR-Pl/B$AR 205 interfaces.
fiovember 5,1976 Lette-frce S&W: schedule for submittal of fire hazards analysis for $WESSAR Pl.
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December 10, 1976 Let kr to 5&W: Round One Questions for SWE55AR.P1/85AR 205.
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January 20,197/
Letter from $1W: response to Industrial Security and Esercency Planning Branch Round One 4.astions.
.}
January 21. 1977 14eeting with S&W: ' discuss 5&W's proposed resolution for the loss of wwer to the RWt suction valves for SWE55AR.Pl/RESAR-35.
February 10. 1977 Submittal of Amendment 30: responses to Round One Questions for SWE55AR P1/85AA.205 and revised Section 9.5.1 for thn fire
. I protection reevaluation, e.,.
'f; <
February 11,1977 Letter from S&W: design for the supply of power to the RHR suction valves for SWE55AR.P1/RESAR-35.
F February 23, 1977 submitta) of Amendrent 31: design for supply of power to the RMt suction. valves for SWESSAR.P1/RESAR 35.
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e ll APPENorr C
! i 8!8tt0GAAPHY
'j StwCruRAL ENstNEEaths
.{
1.
American Institute of Steel Construction (A!$C). *Speelfication for Design. Fabrication l.
and Erection of Structu*al Steel Buildings.* Sixth Edition. 1969, 101 Park Lvenue.
j i
i 2.
American Concrete institute (ACI) " Building Code Requirements for Reinforced Concrete.*
ACI 318-1971. P. O. Bos 4754. Redford Station. Detroit. Michigan.
l 3.
American National Standards Institute (ANSI). *American National Standard Buildi Code Requirements for Minimus Design Loads in Buildings and Other Structures.* A58.1 972.
4.
A. Am4rikian. " Design of Protective Structures." Bureau of Yards and Docks. Publication No. NAV00CK5 P-51. Department of the Navy. Washington D. C.. August 1950.
l 5.
Williamson. R. A and Alvy. R. R., *!seact Effects of Fragments Striking Structural Elanents." Holmes and Nerver. Revised Edition 1973.
MATERIAL $ ENetNEERING 6.
American Society of Mechanical Engineers. ASME Boller and Pressure Yessel Code.
Section !!.1971 Edition includMg Addenda through Summer 19'3. United Engineering Center.
345 East 47th Street. New York. New York.
7.
American Society of Mechenleal Engineers. ASME Boller and Pressure Vessel Code.
Section !!! 1974 Edition including Addende through Winter 1974. United Engineering Center. 345 East 47th Street. New York, New York.
8.
American Society of Mechanical Engineers. " Protection Aga!nst Non Ductile Failure,"
ASME Boiler and Pressure Yessel Code. Section !!!.1974 Edition. Appendia G. United Engineering Center 345 East 47th Street. New York. New York.
9.
American Society of Mechanical Engineers. " Methods and Definitions for Mechanical Testing of Steel Products " ASME Boiler and Pressure vessel Code. Section !!!. Part A-Ferrous.
1971 Edition. Sumer and Winter 1972 Addenda. United Engineering Center. 345 East 47th Street. New York. Ntw York.
- 10. American Society of Mechanical Engineers. ASME Boiler and Pressure Yessel Code.
Section Ill,1971 Edition including Addenda through Surner 1973. Subsection NA 2110 NA-2140. NS-2121. NS 2122. NC 2300 and NO-2300, and Appendix !. " Specification for Materials.* and Appendis IV. ' Approval of New Materials." United Engineering Center.
345 East 47th Street. New York. New York.
- 11. American Society of Mechanical Engineers. ASME Boiler and Pressure Vessel Code.
Section XI,1974 Edition. United Engineering Center 345 East 47th Street, New York, New York.
- 12. American Society of Mechanical Engineers.
- Rules for Inspection and Testing of Components of Light-Water Cooled Plants." ASME Boiler and Pressure vessel Code.Section II 1971 Edition. Division 1. United Engineering Center. 345 East 47th Street New York, New York.
- 13. Westinghouse Electric Corporation. Process Specification 84201 NW. " Corrosion Testing of l
Wrought Austenttic Stainless Steel Alloy "
l
- 14. American Society for Testing Materials. " Standard Method for Conducting Dropweight fest to Determine Nil.Nettlity Transition Temperature of Ferritic Steels." ASTM.E.208-67. Annual Book of ASTM Standards. Part 31. July 1973.1916 Race Street. Philadelphta. Pennsylvania.
l C-1 t
hb h
EMI UYb 2 b
- D 4
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^^
5 e
i l
l l
t t
(
-x.m.
a-
- ,~;2. r_~
n. m -- -- - -
y.3 i
l 1
- 15. American Society for Testing Materials. " Notched Bar Impact Testing of Metallic
}1i Materials." STM E 23 72. Annual Book of ASTM Standards. Part 31. July 1973 1
1916 Race street. Philadelphia. Pennsylvania.
l }
- 16. American Society for Tessing Materials. " Copper-Copper Selfate-sulfuric Acid Test for 8
Detecting Susceptibility to Intergranular Attack in stainless Steels.* ASTM A 262-70 Annual Book of ASTM Standards. Part 3, 1973. 1916 Race $treet. Philadelphia, Pennsylvania.
- 17. American Society for Testing Materials. ' Methods and Definitions for Mech'anical Testing i
of Steel Products.' ASTM A 3070-72. Annual Book of ASTM Standards. Part 31. July 1973 1916 Race Street. Philadelphia, Pennsylvania.
1 J
- 18. American Society for Testing Materials. *$urveillance Tests on Structural Materials in l
Nuclear Reactors." ASTM E 185-73. Annual Book of ASTM Staimaards. Part 30, 1916 Race Street. Philadelphia, Pennsylvania.
j
- 19. American Society for Testing Materials. *Recomended Practice for Conducting Acidified Copper Sulfate Test for Intergranular Attack in Austenitic Stainless Steel." A$TM A393-63 Annual took of ASTM 5tandards. Part 3. 1963. 1916 Race Street. Philadelphia Pennsylvania.
I ELECTRIC!1. IN5TatMENT AND CONT ***, $Y$ TEM $
- 20. Standards of the Institute of Electrical and Electronic Engineers (IEEE). United Engineer-ing Center. 345 East 47th Street New York, New York:
a.
1EEE Std 2791971
- Criteria for Protection Systems for Nuclear Power Generating
{
5tations."
I' b.
IEEE Std 3081971 ' Criteria for Class IE Electric Systems for Nuclear Power Generating Stations."
c.
IEEE Std 3171972 " Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations."
d.
"!EEE. Standard for Qualifying Class IE Equipment for Nuclear Pouer Generating Stations."
e.
IEEE Std 334-1974 " Type Test of Continuous Duty Class IE Motors for Nuclear Power Generating $tations."
f.
- Installation. Inspection and Testing Requirements for Instru-mentation and Electric Equipment During the Construction of Nuclear Power Generating Stations."
s g.
IEEE Std 338-1971 " Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems."
i h.
IEEE Std 344-1975 *$elsmic Qualification of Class i Electrical Equipment for
)
Nuclear Power Generating $tations.*
1.
- Trial-Use Guide for Application of the Single failure Criterion to Nuclear Power Generating Stations Protection Systems."
- j. IEEE Std 3921972 " Trial Use Guide for Type Test of Class ! Electric Valve Opera-
)
tors for Nuclear Power Generating Stations."
j k.
IEEE Std 3831974 " Type Test of Class IE Electric Cables. Field Splices and Connec-tions for Nuclear Power Generating Stations."
j i
1.
IEEE Std 384-1974 " Trial-use Standard: Criteria for separation of Class IE Equipment and Circuits.*
m.
IEEE Std 3871972 " Trial Use Standard: Criteria for 01esel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating $tations."
n.
IEEE Std 4501972 "Recomended Practice for Maintenance. Testing. and Replacerent of Large Stationary Type Power Plant and Substation Lead Storage Batteries."
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METEOR 0t0GY Solsenga. S. J., " Great Lakes Snow Depth Probability Charts and Tables." U. S. Lake f
21.
Survey Research Report No. 5-2. Corps of Engineers. U. $. Department of the Aruty.1967 j
Detroit Michigan.
l Brower. W. A.
Meserve. J. M.. and Quayle. R. G.
" Environmental Guide for Seven U. $.
22.
Ports and Harbor Approaches." U. S. Envirormental Data service. NOAA. National Climatic l
5 Center.1972. Asheville. North Caroline.
1' Brewer. W. A Moserve. J. M.. and Quayle. R. G.. "Envirorseental Guide for the U. S. Gulf 23.
Coast." U. S. invironmental Data Service. NOAA. National Climatic Center.1972. Asheville.
North. Carolina.
Dunlap. D. V., "The climate of the Northeast Probabiliti", of Extreme Snowfalls and Snow 24.
Depths." Bulletin 821. New Jersey Agricultural Experiment $tation. Rutgers University.
l l
1970. New Brunswick. New Jersey.
4
- 25. Golden. J. H., *$cale !nteraction Implications for the Waterspout Life Cycle." Journei of Applied Meteorology. Volume 13. No. 6. September 1974, pp. 693 709.
i
]
- 26. Golden. J. H., "The Life Cycle of Florida Keys' Waterspouts." Journal of Applied Meteo-l rology. Volume 13. No. 6. September 1974, pp. 676 692.
- 27. Hmsing and Home Finance Agency. *$now Load studies." Housing Research Paper 19. 1952.
Olvision of Housing Research. Washington D. C.
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Marley Company. " Engineering Weather Date." (Reprinted with the per,sissics of the U. $.
29.
Departments of the, Army. the Nevy, and the Air Force). The Marley Comparty, Mission.
6 l
y Kansas.
I 4
I
- Meteorological Characteristics of the Probable Maximum Hurricane. Atlantic and Gulf 30.
Coasts of the United P.ates." Memorandum HUR 7-97, 1968. Interim Report. Hydrometeorolog-1 teal Branch. Office el Hydrology. U. $. Weather 8ureau to the Corps of Engineers _._,
" Asymptotic and Peripheral PNssures for Probabu Maximum Hurricanes." l'.amorandum HUR 31.
7 97A.1968, Hydrological 8tweh. Of fice of Hydrblogy U. S. Weather Bureau to the Corps of Engineers.
- 32. Riordan. P.
' Extreme 24 Hour $nowfalls in the United States: Accunulation. Distribution.
and Frequency." Special Report ETL 5R-73-4.1973. U. $. Army Engineer Topographic Labora-tories. Fort Belvoir. Virginia.
Sangendorf. J. F., "A Program for Evaluating Atmospheric Dispersion from a Nuclear Power J3.
Station." NOAA Technical Memorandum ERL ARL-42.1974 Air Resources Laboratory NOAA.
Tattleman. P. and Gringorten. I. !.. " Estimated Glaze Ice and Wind Loads at the Earth's f
34 Surface for the Contiguous United States." 1973. Air Force Cambridge Reseerth Labora-tories. Bedford. Maine.
Them. H. C. $.. " Design Snowloads for the Contiguous United States.* 1969. Of fice of 35.
Climatology. U. S. Weather Bureau, Washington. D. C.
3.. "New Olstributions of Extrcee Winds in the United States." Jaurnal of the Thom. H. C 36.
structural Olvision. Proceedings of the Anerican Society of Civil Engineers July 1968, pp. 1787 1801.
U. 5. Department of Ccmerce. "C11 matte Atlas of the United States.* Envircrvnental Dats
~
37.
Service,1968. Envirornental Science Services Administrat.un. Washington D. C.
U. 5. Department of Corinerce. " Frequency of Maximum Water Eculvalent of March Snow Cover 38.
in Ncrth Central United States.* Technical Paper No. 50. 1964 Office of Hydrology. U. $.
Weatrer Bureaa. Washington. D. C.
t U. $. Naval Weather Service, " Worldwide Airfield turnaries. United States of Aurica.
39.
" Volume vill.1969. Federal Clearinghouse for Scientific and Technical Information.
Springfield. Virt, inia.
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Ct3AlletENT AND ENGINEEpED SAFITY FIATURES SYSTEMS
- 40. Allen. A. 0.. *The Rodiation Chemistry of Water and Aqueous Solutions." Van Nostrand s
Company,1961.
i
- 41. American Nuclear Society.
- Decay Energy Release Rates Following shutdown of Uranium-Fuel Thermal Reactors (DRAFT).* AN5 Standard ANS-5.1. Hinsdale Illinois. October 1971.
l
- 42. Comerd. H. F., and G. W..bes.
- Limits of Flasusbility of Gases and Vapers." Sureau of Mirwt Bulletin 503. 1952.
J I
- 43. Moody. F. J., 'Maxime Flow Rate of a Single Component. Two-Phase Mixture." Vol. 87 p.134. Journal of ueet Transfer. February 1965.
- 44. Persly. L. F., " Design Considerations of Reactor Contatsmient spray Systems Part VI." The Heating of Spray Drops in Alr-Steam Atmospheres. USAEC Report ORNL-TM-2412. Oak Ridge.
Tennessee January 1970.
l 8
- 45. Wettig. W. H., G. A. Jayne. K. V. Moore. C. E. Slater, and h. L. Uptmor. *RELAP3 - A Computer Program for Reactor Blowdown Analysis. *!N-1321. Idaho Nuclear Corporation.
l3 June 1970.
.t '
- 86. Richardson. L. C.. L. J. Finnegan. R. J. Wagner, and J. M. Waage.
- CONTEMPT - A Computer Program for Predicting the Containment Pressure-Temperature Response to a Loss-of. Coolant Accident." 100-17220 Phillips Petroleum Company. June 1967
- 47. $ctaitt. R. C.. G. E. Bingham. J. A. Norberg. *$1mulated Design Basis Accident Tests of the Carolina Virginia Tube Reactor contairment." Final Report. IN.1403. Idaho huclear Corporation. December 1970.
6
- 48. $1wghterbeck. D. C.. "Concarison of Analytical Techniques Used to Determine Distribution of Mass and Energy in the Liquid and Vapor Regions of PWR Containment following a Lots-of-Coolant Accident." Special Interin Report. Idaho Nuclear Corporation. January 1970.
- 49. Slaughterbeck. D. C..
- Review of Heat Traefer Coefficients for Condensing Steam in a Containment Butiding Followirig 4 Loss of-Coolant Accident." IN.1388. Idaho Nuclear Corporation. September 1970.
- 50. Tagami. T.
' Interim Report on Safety Assessments and Fact 11ttes Establistment Project in Japan for Period Ending June 1955 (No.1).* Prepared for the Nationcl Reactor Testing Station. February 28,1966(Unpubitshedwork).
Uchida. H., A. Oyama, am! Y. Toga. " Evaluation of Post Incident Cooling Systems of Light-51.
Water Power Reactors." in Proceeding of the Third International Conference on the Peaceful Uses of Atomic Energy held in Geneva. Augast 31 - September 9. U64. Volume 13. Sesvion 3.9. New Yorkt United Nations 1965. (A/ Conf. 28/P/436) (May 1964) pp.93-104. Division of Technical Review. U. 5. Nuclear Regulatory Cosmission. Washington. D. C.
- $2. Westinghouse Electric Corporation, " Westinghouse Mass and Energy Release Data for Contain-ment Design." WCAP.8312 A Rev. 2 (Non-Proprietary), August 1975.
- $3. Westinghouse Ele:tric Corporailon.
- Westinghouse Mass and Energy Release Data for Contrin-msnt Design." WCAP.8264P A Rev.1 (Proprietary). August 1975.
i
[.M.c 54 " Standard Nuclear Quality Assurance Program - SW50AP.* Topical Report. December 1974 a
Stone s Webster Engineering Corporation. Boston Massashusetts.
- 55.
- Radiation Shielding Design and Analysis Approach for Light water Reactor Power Plants -
RP 84." Topical Report. May 1975. Stone & Webster Engineering Corporation. Boston.
American National Standards Institute.
- Nuclear Safety Criteria for the Design of 56.
5tationary Pressur12ed heter Reactor Plants." ANS! N18.2. Hinscale. [11tnois. August 1974.
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United States Atomic Energy Coseission Guideline Document. " Inservice Inspection Require-f I
- 57. ments of Nuclear Power Plants Constructed with Limited Accessibility for Inservice t
January 1969.
United States Atomic Energy Commission.
- Guidance on Quality Assurance Requ 58.
May 1974 United States Atomic Energy Commission. *Cuidance on Quality Assurance Requirements During the Construction Pnese of Nuclear Power Plants.* WASH-1309. My 1974.
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59.
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RE'nPT BY THE
-t ADVISORY C0tHITTEE ON REACTOR SAFEGUARDS i
}
i ON SWESSAR P1 l
STONE & k'EBSTER ENGINEERING CORPORATION I
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~12-14,1976,ahe_ Advisory 0:nutittee on -
- At' *; SIC 196th meetini,dALagust'reviewi:the agplicationlof the' Stone ard Webster Engi.
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? : neerf.-[S atior#Eciedt selwmy Design Approval of,r.its SElSSAR-P1, y yz 3,,,
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- tialance-of-. plant 1(BCP) design that would' interface l
., wit $ sing 3dnit'. Westinghouse Electric.Corpo-atimLRESARG pressurize (,
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water?ni'acleaf steamtmpply"syFtSE ysSSggshe 3eSSAR-P1 BOP design was J
l reviewed'by.the Ocummittee irr.xelatiori to the toestie== RESAR-41..
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. fits ard s iiiporilc'oridedian Petdsary 1$1976.q'the description of i
og 5 NSSSc
SWESSit-P1fprbrided in't!,er Pstinary 11",*1976 report is applicable to t - c 1 RE:.>AR:;3Si'thflattec%es' reviewed runs a report prov.ided by the Ccssaitteer.'
on Duty,.14#1976.* Du' ity,its revia;(, the Caesnittee had the tenefit of
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.T.discussionsMth hesentstiin of the' Stone al WeMter EngMrW,
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the. Nuclear Regulatory cuneission @9.%.Sta g,W Q Corporation"
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' criUcal; safety;related~ ecpipment'to frotect against ocsumon mode failu:es qs
-'asedeiated dithi firesJoe ot'ier. operational carstinge'ncies.. However, cort-
' MeM design details'foCSDElSSAh-P1 have not besn developed ard the mocep,-
haisi not!' pet beerf agpliini to a oreplete' darp@ ant M@ - N i
4 eseguently! further reviebt of.thit girysick1; sept. ration arrangement, should
'be made prioritx,.the Final Design Approval, or when SWESSAR-P1 As pro;osed fod aliucnar pdwer;plasit foc idrich a construction' permit is being s:nght.
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t, Honorable Mercus A. Ilcuden August 18, 1976 A metter of major concern in the IWC Staff's rwimw has been the safety-related interfaces between the SNEBEMEP1 IEP design and the 1ESAR-35 NBBS design, on one had, and the customdesigned site-related structures amt oogenents, on the other hand. She responsibilities and reqtirements related to the SEBEAR-P3/RESMF3B interfaces have been partially defined s
in the Safety Aulysis Reports for these tuo standardised designs. The Com-mittee believes that these interface requirements are se*M-R r for a Preliminary Design Approval, but expects the IEC Staff and the utility applicant to corptinue to esamine then further in connection with the proposal to use these designs for a specific plant when it is revisued for a con-struction permit. The interfaces between SleE5AIFP1 and the site-related features are defined in the 9ESSAR-P1 Safety Analysis Report, but have not yet been subjected to the test of a cegales design for a nuclear power plant. The NIIC Staff should review these interfanes in greater depth when a
a construction permit application is received.
the Caesaittee recommerris that, durirg the design, procureaant, constrac-tion, and startup, timely and appropriate interdisciplfnary system analyses be performed to assure couplete fmetional cxupatibility across each interface for the entire spectra of anticipated operations and postulated design basis accident corditions.
1 The coordination of interdependent instrtsmentation and controls in the nuclear island and in the balance of plant will require attention at the time when SNESSAR-P1 is used as a portion of a nuclear power plant license application.
j These matters should be incisied in the IGC Staff's Standard Review Plan.
I j
ihe gW orientation of the turbine-generator with respect to the nuclear island is suitable for a sitgle tmit installation. Por sultiple satit Iower plants, the location and orientation of the units should be such as to yield acaptably low probabilities of dams 3e by low-trajectory turbine-generator' missiles, or suitable missile shielding should be provided.
The SWElSSAR-P1 and the RESAR-3S lESS desigras, as do many others, utilize the concept of two-track continuous duty systems such as ventilation and service i
water which parform critical service functions. In some < - the probability I
of failure of one of these systens is not low. The failure of the second systen to cart or continue to operate may cause progressively dasaging con-seqJenms. The Conmittee reconnends that failures of this kind be evaluated to determine if the necessary reliability exists for these systems and whether renedial measures are appropriate.
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i-i Bonorable Marcus A. Ituden 3-August 18, 1976 Although SWISBMk-P1 and HESAD 38 include prorisions for pra*=e+1m against industrial satotage, the Consittee believer that ftrther steps can be taken beyond those provided. Prior to the use of SEBSAR-P3/RBSAR-35 as a portion of an application for a nuclast power plat ifrunaa, the utility applicant should be required to demonstrate that aaa=r*=h1= industrial sabotage pro-visions will be incorporated into the plant design.
'the SIESSMbP1 design includes some provisions eh anticipate the main-l tenance, inspection, ard operatimal needs of the planc throughout its service life, including cleaning ad decontamination of the pris.ry-coolant system, and eventual d= = =mianianing. Bouever, edssa SIEBEhI> F1 is used as a portion of a nuclear power plant license application the Cossaittee halteves that the let staff and the utility applicant should further review methods ad gracedures for removing am-lated radioactive contamination whereby mainta-nasce and inopoetion programs and ultiasta de==ia=ioning can be more effec-tively and safely performed.
a Generic problems related to large water reactors are diaa= mad in the CGamit-tee's report dated April 16, 1976. 'those probless relevant to SM:SSAR P1 and RESAR-3S should be dealt with aRropriately by the lac Staff and the utility aR111 cent as solutions are found. 'the relevant itses are: Group II -
Itses 1, 2, 3, 4, 5, 6, 7, 9, 10, 11; Gro g IIA - Itaas 1,4,5,6,7,8; Geo g IIB - Itan 2; Group IIC - Items 1,2,3,4,5,6,7.
'Ihe Advisory Cbustittee on Reactor Safeguards believes that the itsas asn-
~
tioned above can be resolved during t5e stardardized plant licensing process l
and that, if due consideration is given to the foregoing and to the recxzenen-(
dations in the Consmittee's report of July 14, 1976 on RESAR-3S, Preliminary l
Design Appecval for 9ESSAR-P1 to be used in conjunction with RESAR-3S can I
be granted in momed with the spirit and purposes set forth in the Cattais-sion's policy statement en standardization of nuclear power plants as de-scribed in 18WiE-1341, "Progranaatic Information for the Licensing of Stand-ardized Nuclear Power Plants" and in conformance with the Regulations of c
Appendix 0 to Part 50 and &ction 2.110 of Part 2 of Title 10 of the Code of Pederal Regulations.
Sincerely yours, i
[
Dede W. Itx11er i
3 Chairman 1
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Pressurised Mater Beactor Reference.pr pq=ytql.*y t,q~&*t' W.9 -97 p'~~
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'l Analyels asport.(mm8EMk F1h and-1.through 27. NH'4fi.F%..;*a.8 'y r$
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I,51 Stone and unbetar aurinnerin@g Corporation letterssa.i s.*M@,W.$tS/$$ 1,& #
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a.., Pehruary.18i19M. Desist Imed asjection CagebilityT1,ht#%Atwu;.M.". ;q d3
- b..tapril.13,19M -;Withdrasal of Design Emed asjectionfropanal d.t0'fE'h..OM&M4:W jiS t.4/..s $ # 4seS9r.L fFJ r yi @ 6 ;y P Y,{,:d$!? e./1., -
. 4) 3.
Report to the' Advisory Onemittee on13mecher Sadeguardsln,:the Matter a " -
"L 7
?.
i of stone ans mheter agineering Corporation sw saMy Analysis Report met asference thwdaar 30uer Plant WWEntf1 (and its relatime-ship to the EMBW-35'stamniard Mus. Design) Dochet us. SIN'50 495,7, ";#.>.
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Publishedt June-19M, U. S. helaar aspalatory Commission, fWien of Nuclear asector megulation.,y? -%g gfyy3*giggypCy..,, ' a
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Stone and Webster Engineering %d.ica letter dated March 3,1976, transmitting additional fa*w==tica av= weening meBSMMESna-38 Design.
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