ML20236B440
| ML20236B440 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon, 05000000 |
| Issue date: | 10/16/1974 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20236A877 | List:
|
| References | |
| FOIA-87-214 NUDOCS 8707290118 | |
| Download: ML20236B440 (261) | |
Text
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Ort ober 16, 1974
!?AFETY EVALUATIOT EY THE D_lRECTORATL OF, LICENSING .
U. S. A10MIC ENERGY COMMISSION IN THE MATTER OF i
PACITIC CAS AND ELECTRIC COMPAN) j DI ABLO CANYON NUC1. EAR POWER STATION, UNITS 1 AND 2 SAN LUlb OBISPO COUNTY, CALIFURNIA DOCKE1 NOS. 50-275 AN1' -323 i
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i TABLE OF CONTENTS PAGE i
viii ABBREVIATIONS . . . . ....................... l
)
1-1
1.0 INTRODUCTION
........................ ........... . . 1-1 1.1 General Background .... .
1-4 1.2 General Plant Description . . ............. 1-8 1.3 Comparison with Similar Facility Designs ...... . 1-8 1.4 Identification of Agents and Contractors . 1-9 1.5 Summary of Principal Review Matters . . . . . . . . .
1.6 Facility Modifications Required as a Consequence 1-11 !
of Regulatory Staff Review .....
............. 2-1 2.0 SITE CHARACTERISTICS ....... . . ......... .... 2-1 2.1 Geography and Demography 2-1 l 2.1.1 Site Location . . . ............
.............. . 2-1 2.1.2 Site Description . .
2-2 2.1.3 Population and Population Distribution ... . 2-4 .
2.1.4 Uses of Adjacent Lands and Waters . . . . . . .
l 2.2 Nearby Industrial, Transportation and Military 2-6 Facilities .....................
............. 2-7 2.3 Meteorology ... .... . 2-7 !
Regional Climatology ... ..........
2.3.1 2-7 2.3.2 Local Met eorology . . . . . . . . . . . . . . .
.. 2-8 2.3.3 Onsite Meteorological Measurements Program 2-10 2.3.4 Short-Ters (Accident) Diffusion EstLastes . . . 2-11 2.3.5 Long-Ters (Routine) Diffusion Estimates . . . . . 2-11 2.3.6 Conclusions . . . . ............. 2-12 2.4 Hydrology . . . . . . . . . ..... ......... 2-12 2.5 Geology, Seismology, and Foundation Engineering . . . .
3.0 DESIGN CRITERIA - STRUCTURES, COMPONENTS, EQUIPMENT, AND 3-1 SYSTEMS 3-1 3.1 Conformance with AEC General Design Criteria .. . ..
3.2 classification of Structures, Components and 3-1 Systems . . . . . . . . . .......... 3-1 3.2.1 Seismic Classification ... ... ..... 3-3 3.2.2 System Quality Group Classifications .. 3-4 3.3 Wind and Tornado Design criteria ......... .. 3-6 3.4 Water Level (Flood) Design Criteria . . . . . . . . .. 3-7 3.5 Hissile Protection Criteria . . .........
3.6 Protection Against the Dynamic Ltfects Associated with 3-9 the Postulated Rupture of Piping ..... ... .. 3-12 Seismic Design ...... .... .......... ,
3.7 i
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.i TABLE OF CONTENTS (Cont'd) 3.6 Design of Category I Structures....................... 3-13 i 3.8.1 Con c r e t c Con t a i nmer t .' . . . . . . . . . . . . . . . . . . . . . . . . . 3-1M.
3.8.2 Concrete and Structural Steel Internal Structures ................................. 3-1 s 3.8.3 Other Seismic Category I Structures .......... 3-17 3.8.4~ Foundations-and Concrete Supporta............ 3-14 l' 3.9 - Mechanical Sys t ems and Component s. . . . . . . . . . . . . . . . . . . . . 3-2:
3.9.1 Dynamic System Analysis and Testing. . . . . . . . . . . 3-24 3.9.2 J ASME Code Class 2 and 3 Components. . . . . . . . . . . . 3-24 3.10 Seismic Qualification of Category 1
-Ins trumer. tat ion and Elec trical Equipment. . . . . . . . . . . 3-25 4.0 REACT 0R..................................................... 4-1 4.1 Summary Description................................... 4-1 4.2 Mechanical Des 1gn..................................... 4-1 4.2.1 Fue1.......................................... 4-1
-4.2.2 Reactor Vessel Incern41s........ ............. 4-7 4 Nuclear 0esign........................................ 4-k 4.3.1 Genera 1....................................... 4-b 4.3.2 Power Distribution............................ 4-6 4.3.3 Res-tivity Coefficients.... .................. 4-9 4.4 Thermal . and Hyd raulic Design. . . . . . . . ............... 4-It' 5.0 REACTOR COOLANT SYSTEM...................................... 5-1 5.1 S umma ry De s c ri p t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Integrit; of Reactor Coolant Pressure Boundary. .. . . . . . 5-1 5.2.1 Design of Reactor Coolant Pressure Boundary Components................................. 5-J 5.2.2 Overpressu riza tion Protect ion. . . . . . . . . . . . . . . . . 5-3 5.2.3 Ceneral Mat erial Censidera tions. . . . . . . . . . . . . . . 5-3 5.2.4 Fracture Toughness.,.......................... 5-5 5.2.5 Pump Flywhee1................................. 5-6 5.2.6 P ump 0v e r s peed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-7 5.2.7 RCPB Leakage Detec tion Syst em. . . . . . . . . . . . . . . . . 5-7 5.2.8 Inservice Inspection Pr 3 gram. . . . . . . . . . . . . . . . . . 5-b 5.3 Reactor Vessel Integrity.............................. 5-9 5.4 Loose Parts Monitor................................... 5-11 5.5 Res idual Hes t Removal Sys t em. . . . . . . . . . . . . . . . . . . . . . . . . . 5-11 6.0 . ENGINEERED S AFETY FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 6.2 Genera 1...............................................
Containment Systems...................................
6-1 -
6-2 6.2.1 Containment Functional Design................. 6-2 )
6.2.2 Containment llea t Removal Sy s t ems . . . . . . . . . . . . . . 6-7 6.2.3 Containment Air Purification and Cleanup ] '
Systems.................................... 6-9 I
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111 TABLE OF CONTrNTS (Cont'd) ;
6-11 '
Systems................. 6-12 6.2.4 Containment Isolation Systems............... 6-13
-.2.5 Combustible Gas Control 6-14 i
Containment Le akage Testing P rogram. . . . . . . . . . .
6.2.6 6.3 Emer genc y Core Cooling System (ECCS) . . . . . . . . . . . . . . 6-14 ....
f-16 6.3.1 Design Bases..................................
Design................................. 6 6.3.2 System 6.3.3 Pe r f ormanc e Ev aluation. . . . . . . . . . . . . . . . . . .6-19 Inspections......................... .....
6-22 6.3.4 Tests and 6-22 6.3.5 Conclusion....................................
6.4 Hab it ability Sy s t ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7"'I 7-1 1 7.0 INSTRUMENT I ATIP AND CONTR0LS .* ..=. . .***. . . . . 7-. . ..........
7.1 Genera 1....................... ...- ..*....................
7.2 Reac tor Trip Sy s tes. . . . . . . . . . . . . . .-
7.2.1 Cencral. ..... - - . - - .
Reactor Itip )., a - - - . . -
7,2.;
7 2.2.1 Pl. v :
7.2.2.2 Electr ual is .
7-Seisma ?ualib. 4tson............... 7-4 7.2.2.3 7.2.2.4 Conclusions......................... 7 f' P ro ce ss Analo g Sy s tem. . . . . . . . . . . . . . . . . . . . . . . . . 7-6 7.2.3 7-7 7.2.4 Testability of Protection Systems. . . . . . . . . . . . .
7.2.5 Anticipated Transients Without Scram (ATWS)... 7-7 7.2.6 Con c lu s ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7-8 Engineered Saf ety Feature Actuation Systems. . . . . . . . . . . 7-8 7.3 7.3.1 General...........
........................... 7-8 I
ESF Actuation Logic........................... 7-8 7.3.2 7.3.3 Accumulator Isolation Valves . . . . . . . . . . . . . . . . . .
7.3.4 ~ Changeover f rom Injection to Recirculation 7-9 Mode....................................... 7-10 7.4 Systeme Required f or Saf e Shutdown. . . . . . . . . . . . . . . . . . . . 7-11 Safety Related Display Information.................... 7-12 7.5 7-12 7.6 RHR Overpressure e P rotection Interlocks . . . . . . . .
Safety...............
Control Systems Not Required for 7-12 7.7 7-14 7.8 Environmental and Seismic Quali?tcation.. . . . . . . . . . . . . .
7.9 Conclusion............................................ 8-1 8-)
8.0 ELECTRIC POWER ..............................
8.1 Genera 1............................................... 8-1 8.2 O f f s i t e P ow e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .8-3 ........
8.3 Ons i t e P ow e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .8-3 B-5 A-C P ow e r Sy s t ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.3.1 8~b D- C P ow e r Sy s t ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
8.3.2 8.3.3 Conclusions...................................
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i iv TABLE OF CONTENTS (Cont'd) j i
l 8.4 Physical Independence of Electrical Equipment and I Circuits.............................................. 8-6 l 9.0 AUXILIARY SYST mS........................................... 9-1 9.1 Genera 1............................................... 9-1 9.2 Fuel Storage and Handling............................. 9-2 9.2.1 New and Spent Fuel Storage.................... 9-2 9.2.2 Spent Fuel Pool Cooling and Cleanup System.... 9-3 9.2.3 Fuel Handling System.......................... 9-3 9.3 Water Systems........................~................ 9-4 9.3.1 Auxiliary Saltva ter Sys tas. . . .. . . . . . . . . . . . . . . . . 9-4 9.3.2 Component Cooling Water Sys tes. . . . . . . . . . . . . . . . 9-5 9.3.3 Mateu p Wa t e r Sy s t em. . . . . . . . . . . . . . . . . . . . . . . . . . . 9-7 9.3.4 Ultimate Heat Sink............................ 9-7 9.4 Process Auxiliaries................................... 9-8 9.4.1 Chemical and Volume con trol Sys tan. . . . . . . . . . . . 9-8 9.5 Air-Conditioning, Heating, Coolias, and Ventilation Systems............................................ 9-9 9.5.1 Control Area.................................. 9-9 9.5.2 Auxiliary kilif r.g (,F.xcluding the Fuel Hand li ng A;ces) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-10 9.5.3 Fuel Handling Area of the Auxiliary Building................................... 9-11 9.5.4 Intake Structure (Auxiliary Saltwater Pump Compartments).............................. 9-11 9.5.5 Diesel Cenarstor Compartment.. .............. 9-12 9.6 O t h e r Aux il iary Sy s t ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-13 9.6.1 Fire Protection System........................ 9-13 9.6.2 Diesel Generator Auxiliary Sys tems . . . . . . . . . . . . 9-14 9.6.3 Diesel Generator Fuel 011 Systesa. . . . . . . . . . . . . . 9-15 10.0 STEAM A D POWER CONVERSION SYSTEM........................... 10-1 10.1 Sunusary Description................................... 10-1 ;
10.2 Turbine 0enerator..................................... 10-1 '
10.3 Main S t eam S upply Sys t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2 10.4 Other Features........................................ 10-4 10.5 Auxiliary Feedwater System............................ 10-7 )
11.0 RADIOACTIVE WAST E MANAGEMENT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.1 Summary Description................................... 11-1 11.2 Liquid Waste System................................... 11-2 l
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v TABLE OF CONTENTS (Cont'd)
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11.2.1 System Description............................ 11-2 l 11.2.2 Sys t em Ev alua t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-3 ]
11.2.3 Liquid Waste System Evaluation Findings... .. .. 11-6 11.3 Cascous Waste System.................................. 11-7 11.3.1 System Description................s........... 11-7 l 11.3.2 System Evaluation............................. 11-8 1 11.3.3 Caseous Waste System Evaluation Findings. . ... . 11-11 11.4 Process and Effluent Radiological Nnitoring System... 11-12 11.4.1 System Description............................ 11-12 i 11.4.2 System Evaluation............................. 11-12 11.4.3 Process and Effluent Radiological Honitoring Evalua tion Findings . . . . . . . . . . . . . . . . . . . . . . . . 11-14 11.5 Solid Waste System.................................... 11-14 11.5.1 System Description............................ 11-14 11.5.2 System Evaluation............................. 11-15 11.5.3 Solio Waste System Evaluation Findings........ 11-17 12.0 RAD I ATION PRC EC T I 0N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-1 12.1 shielding............................................. 12-1 12.2 Ventilation........................................... 12-2 12.3 Dea l th Phys ics Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-3 12.4 Rad ioac tiv e Ma terials Saf ety. . . . . . . . . . . . . . . . . . . . . . . . . . 12-4 12.5 Area Monitoring....................................... 12-4 13.0 CONDUCT OF OPE RATIONS . . . . . . . . . . . . . . : . . . . . . . . . . . . . . . . . . . . . . . . 13-1 13.1 Organization and Qualifica tions . . . . . . . . . . . . . . . . . . . . . . . 13-1 13.2 Training Program...................................... 13-3 13.3 Emergency Planning.................................... 13-4 13.4 S a f e ty Rev iew and Aud i t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-5 13.5 P lan t P ro c ed u re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-6 13.6 Industrial Security................................... 13-6 14.0 INITIAL TESTS AND 0PERATION................................. 14-1 ,
I 15.0 AC C I DENT ANALY S ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 1 15.1 Cenera1............................................... 15-1 15.2 Design Bas is Accident Assumptions . . . . . . . . . . . . . . . . . . . . . 15-3 15.2.1 loss-of-Coolant Accident (Containment Leakage) 15-3 15.2.2 Fuel Handling .".ccident........................ 15-5 ,
15.2.3 Cas De cay Tank Ru p ture . . . . . . . . . . . . . . . . . . . . . . . . 15-5 l' 15.2.4 Con trol Rod Ej ec tion Ac cident. . . . . . . . . . . . . . . . . 15-6 15.2.5 Hyd ro g en P u rg e Do s e . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-7 1
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TABLE OF CONTENTS (Cent'd) 4 16-1 16.0 TECHNICAL SPECIFICATIONS.... ...............................
1 17.0 QUALITY ASSULANCE........................................... 17-1 17.1 Genera 1............................................... 17-1 17.2 Organization.......................................... 17-1 17.3 Quality Assuran:e Program............................. 17-4 17-6 17.4 Conclusion............................................
18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 18-1
( AC RS ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...................
10-1 19.0 COMMON DE FENSE AND S ECUR1TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
20-1 20.0 FINANCIAL QUALIFICATIONS...................................
21-1 21 0 FINANCI AL PROTECTION AND INDEHNITY REDUI REMENTS. . . . . . . . . . . .
21.1 Preoperational Storage of Nuclear Fue1................ 21 'A i 21.2 Operating License............... ..................... 21-2 21.3 Co n c l u s i o .1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21-3 22-1 22.0 CO N C LU S I ON S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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.l .'_A G.E..
IFPENDlX t A - Chronoloxy in t he R..d iological Review. . . ...... .. .... .4- 1 APPENDIX S - Pacific jas and EJ eet ric Company - Financial Analysis. . .. 3-1 M'PI NDlX C Bibliography.... .................................... C-1 FABLES Table i 2 - Fuel Mechanical Design Comparison........................ 4-17 i
Table 4.2 - Range of Design Par amet er Experience. . . . . . . . . . . . . . . . . . . . . . 4-16 f abl e 4. 3 - Thermal and Hyd raul i c Des ign Pa ramet ers. . . . . . . . . . . . . . . . . . . 4-19 Table '1.1 - Design Parameter.4 of Principal Cowponents Considered in Liquic Fadwante Evaluation.......................... 11-18 Table 11.2 - Design Parameterr of Principal Cor.ponents Considered in Gaseous Radwaste Evaluation......................... 11-19 Table 11.5 - Process and Effluen- Monitoring.......................... 11-20 ra5le 15.1 - Potential Of f atte Ooses Due to 9e 41gn Basis Accidentr . . ... .................................... 15-8 t'I.C.U.R..E.S Fir,ure 2.: - Diablo Canyon Pinct General Site Location................ 2-13 1 1
Figure 2.2 - Location i.i Atte h<iondary and Principal Structures....... 2-14 F f gure 2. ') - Cumulat . ve Popu;it 8an Distribution....................... 2-15 Fy,ure 2... - Bereau o' Econon.c w.alysis - Water Resources Subarea # 1807. ...................................... 2-16 ,
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viii ABBREVIATIONS a-c alternattog current ACI American Concrete Institute ACES Advit.ory Cormittee on Reactor Safeguards
/IC United States Atomic Energy Commission AISC American Institute of Steel Construction !
ALAP as low as practicable ANS A+erican Nuclear Society ANSI American National Standards Institute ASCE American Society of Civil EngiLeers ASME Ae rican Society of Mechanical Engi.ieers ASTM American Society for Testin?. and Materials ATVS anticipated transients without scran DEA Bure.2u of Economic Analysis B[T boron injection tank BOL beginning of life Bits boron recycle system BTU British Thermal Units BTU /hr British Thermal Units per hour BTU /hr-ft British Thermal Units per hour per square foot BWR boiling water reactor cA wake factor
ix cal /gm calories per gram i
CAM continuous air monitor i cc cubic centimeter CEA control element assembly cfm 4.ubic feet per minute efs cubic feet per second CFR Code of Federal Regulations Ci Curies Ci/yr Curies per year cm/sec centimeters per second CO 2 carbon dioxide CP construction permit CVCS chemical and volume control system DEA design basis accident d-c direct current DF decontamination factor DKB departure from nucleate boiling DNER departure from nucleate boilir.g ratio DOT Department of Transportation ok,ap reactivity change AT temperature change or differencre ECCS emergency core cooling system l
ECL end of life ESF engineered safety features j
^
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GFAS eng:neered uattt y t e.it ure actu it ion .vstems
'F degrees Farenheit FLECl!T f ull length emergency coeling heat transfer F peaking facter q
FSAR final Safety Analysis Report it feet t FES Final Enviror. mental Statement n .
tt- square :eet it3 cubic feet it/sec feet per second d gravitational acceleratica, 32.2 feet per secor.d per secend GDC AEC General Design Crittria for Nuclear Power Plants 3-M Getgcr-Mueller JUSPR & AC Ceneral Of fice Nuclear Plant Review and Audit Committee
- pd gallons per day aPD gallons per minute J'iP S
- ;aseous w.nste processing system iLTA high efficiency particulate air hi s hours
- 7CE Institute of Electrical and Electronics Engineern l
.n inch r,
kiloteter kilovelt 44' kilowatt
< 4h kilewatt-bours i
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'# $ '.1%l : " i r '7 ie; ;
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11- t
' i, te n! , i.. i . . . - d c .. . e * .-
J b .' t t - pounds pet cuti fout ) ,
I b .Ihr pt undr [+r h a r
!.a.. A lons-of-roola t accident ;
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1 1'. l inear p iver A nnit y lc t.P.: los p pulativ z le i 1.W15 liquid waste treatment ayst em j J
meter m- square t.e urs r:pi. mlJer. pet hou:
m / t.cc met ers per se;;nd M..' W mean lover low water M SI mean sea levei
& mepawatts j 1
MW/MT .negawatt-davn per met ric ton MWe megawat ts ele rical E't megawat t s tber:ul mr.d one t housandt6. cf a rad ;
mren one thousandth of a rad equivalent mari mrem /yr one thousaniltn of a rad equivalent man per year NaOH sodium hydroxide ,
NDTT nil ductility transition temperature NPSif net positive suction head l NSSS nuclear steac; .upply system nyt neutron fluence, neut rons per square centimeter i
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xii OBE operating basis earthquake PC&E Pacific Cas and Electric Company pH expression of acidity or alkalinity on a scale of 0-14 P(F probable maximum flood PHP probable maximum precipitation PNAC President's Nuclear Advisory Comm16, ae PP* Parts per million f
PSAR Preliminary Safety Analysis Report psf pounds per square foot i psi pounds per square inch pais pounds per square inch absolute Paig pounds per square inch gauge PSRC Plant Staff Review Committeo PWR pressurized water reactor QA quality assurance j QC quality control RCS reactor coolant system RHR residual heat removal l
I RHRS residual heat removal system RCPB reactor coolant pressure boundary j RPS reactor protection system rea rad equivalent man R&D research and development rpm revolutions per minute s
xiii RWST refueling water storage tank scfm standard cubic feet per minute sec/m3 seconds per cubic meter SER Safety Evaluation Report i
SCBTS steam generator blowdovn treatment system SI safety injection 1 l
SIAS safety injection actuation signal l SIS safety injection system SSE safe shutdown earthquake std standard TBS turbine bypass system TDC thermal diffusion coefficient (
TMD Transient Mass Distribution U-235 uranium 235 U-238 uranium 238 002 uranium dioxide USAS !
United States of America Standard USGS United States Geological Survey v/o volume percent w/o weight percent
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wg water gauge X/Q relative concentration yr ye ar 10 CFR AEC, Title 10 Code of Federal Regulations Part 1 Statement of Organization and General Information Part 2 AEC Rules of Practice
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I Part 20 A':C Stands.rds tor Prote tts.. Against Pa diat ion !
Part 50 AEC L?censine, of Production end l'tilirst2on Iacilitle- i I rt M operai>i* : i c er.se t !
1.u t il P.n a.a r i tip. c.1 1,ao luu t . vt Lit e .a1 16 ; "i t ein npo r t and Transpor ta t. i at, of f. adios, tIve Mater':al Under Cet t a t n Cw.! s t f ont. t Part ! Physical Prot a t ioni :s' 8p' eta: hw le
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- 1. 0 INTRODUCTION 1.1 General Background The Pacific Gas and Electric Co .pany (PC&E, and bereinaf ter referred to as the applicant) filed with the Atomic Energy Commission (Commission) appli-cations dated January 16, 1967 and June 28, 1968, and as subsequer.tly amended, for licenses to construct and operate pressurized water reactors, identified as Units 1 and 2. respectively, of the Diablo Canyon Nuclear Power Station. These Units are being constructed at the Diablo Canyon site which is located on the central California coast in San Luis Obispo County, approximately twel>e (12) miles west-southwest of San Luis Obispo, the County Saat. Unf t 1 is being constructed under AEC Construction i Permit CPFR-39 issued on April 23, 1968, and Unit 2 under CPPR-69 issued on December 9, 1970. On July 10, 1973, the applicant filed an applica-tion for operating licenses to operate Units 1 and 2 of the Diablo Canyon Plant. Included as part of this application var a Final Safety Analysis Report (FSAR) as required by 10 CFR 50.34(b). The application for operating licenses was docketed on October 2,1973.
The application requests operating licenses of 3338 and 3411 thermal messwatts (HWt) fo: Units 1 and 2, respectively; these levels are equiv-alent to net electrical oscputs of 1084 and 1106 electrica). megawatts (MWe) for Units 1 and 2, respectively. The slight difference in output for the two units is due to the upgrad2d turbine generstor design for Unit 2. The Unit 1 thermal power level (3338 MWt) is slightly higher J
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than the 3250 value given in the Prelimina:y Safety Analysis Report (PSAR) which was tendered in 1967. The applicant states that the expected ultimate therusi outputs for Units 1 and 2 are 3488 and 3568 MWe, i respectively; the corresponding net electrical outputs for these values
.are 1131 and 1156 MWe.
The radiological safety review with respect to a decision concerning issuance of operating licenses for Diablo Canyon Units 1 and 2 has been based on the applicant's Final Safety Analysis Report and subsequent Amendments 1 through 17, all of which are available for review at the Atomic Energy Commission's Public Document Room at 1717 H Stt. et N.W.,
Washington, D. C... and at the San Luis Obispo County Library, 888 Horrow Street, San Luis Obispo, California 93406.
During the course of the safety review of the material submitted, we held a number of meetings with representatives of the applicant, his consultants, and Westinghouse Electric Corporation to discuss the plant design, construction, proposed operation an/ performance under postulated accident condations. During our revsew, we requested the applicant to provide additional information that we needed for our evaluation. This additional information wea provided in Amendments to the application.
As a result of ou: review, a number of changes were made in the facility design and proposed operating practices; these changes are described in the applicant's Amendments to the FSAR and aJe discursed in appropriate sections of this report. Section 1.6 provides a listing of the principal design changes which were made. A chronology of the principal actions
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1-3 relating to the processing of the application is attached as Appendix A to this Safety Evaluation Report (SER).
This Safety Evaluation Report sumanarizes the results of the radio-Ingical safety review of Diablo Canyon Units 1 and 2 that was performed by the Commission's Regulatory staff. The review and evaluation of the f acilities for operating licenses is only one stage in the continuing review by the staff of the destga, construction, and operating features of Units 1 and 2. The proposed design of these facilities was reviewed i before construction permits were issued. Construction of these f acil-ities has been sonitored in accordance with the iespection program of the Commission's Regulatory staff. At this, the operating license application phase, we have reviewed the final design to determine that all of the Commission's safety requirements have been met. If operating licenses are granted, the facilities will be operated only in accordance with the terms of the operating licenses and the Commission's regulations, and vi i be subject to the continuing inspection program of the Regulatory staff.
In addition to our review, the Advisory Cosedttee on Raactor Safe-guards (ACRS) is reviewing the application and has met and will meet with both the applicant and the Regulatory staff to discuss the facilities.
The ACE report to the Commission on the Unit 1 and 2 f acilities wirl be provided in a supplement to this Safety Evaluation Report.
The conclusions reached as a result of our evaluation of PG&E's 1 application to operate Diablo Canyon Units 1 and 2 are presented in Section 22 of this Safety Evaluation Report.
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l 1.2 General Plant Description ,
Units 1 and 2 of the Diablo Canyon Nuclear Powar Station are located on a 750 acre site in San Luis Obispo County, California. We site is adjacent to the Pacific Ocean and is roughly equidistant from San Francisco and Los Angeles. )
l Units 1 a d 2 are substantially identical pressurized water nuclear I L
power units, each consisting of a Nuclear Steam Supply System (W.1SS) 16 a 4-loop reactor coolant system, turbine generator, auxiliary equipment, ;
I and controls and instrumentation. For each unit, the principal structures include the containnient, the turbine building, and the auxiliary building (which includes the control room, the fuel handling areas and the venti-lation areas). The ultiaste beat sink for rejectien of heat from the Diablo Canyon Units is the Pacific Ocean.
ne NSSS for each unit consists of a pressurized water reactor, reactor coolant system. e..G ssociated auxiliary fluid systems. The reactor coolant system consists of four parallel reactor coolant loops, ;
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l each containing a steam generator and a reactor coolant pump. A '
t pressurizer is connected to the hot leg of one reactor coolant loop. i The basic fuel design is a 17x17 matrix of fuel rods in each fuel assembly.
The reactor core is composed of an array of 193 fuel assemblies, each containing 264 fuel rods. These rods are composed of uranium dioxide pellets enclosed in Zircaloy tubes with welded end plugs. All fuel rods are pressurized with helius during f abrication to reduce I stress and increase fatigue life. The reactor core will initially con- ,
i tain 3 regions of slightly different enrichments of U-235.
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l 1-5 The reactor will be controlled by control element movement and The full regulation of the boron concentration in the reactor coolant. i length rod cluster control assemblies are etainless steel tubes contain-ing a silver-indium-cadmium absorber, and are positioned by drive mechanisms of the magnetic istch type. Part length control rods are A soluble also provided for use in controlling axial power distribution.
poison (boron) is introduced into the reactor coolant to cospensate for long term reactivity ct.anges.
The reector vessel and reactor internals contain and support the fuel and rod cluster control assemblies. The vessel is cylindrical with hemispherical heads and is clad with stainless steel.
The pressurizer is a vertical cylindrical pressure vessel with hemispherical heads and is equipped with electrical heaters and spray nozzles for system pressure control. The steam generators are vertical U-tube type heat exchangets with Inconel tubes. Reactor coolant flows r
inside the tubes; steam is generated in the shell and flows through the main steam lines to the turbine. Integral moisture separating equipment reduces moisture content of the steam at the turbine thrott?.e to 0.25 percent or less. The reector coolant pumps are vertical, single-stage, centrifugal units equipped with controlled leakage shaf t seals.
Auxiliary systems are provided to charge the reactor coolant system q l
i and add makeup water, to purify reactor coolant water, to provide I chemicsis for corrosion inhibition and reactor control, to cool system !
components, to remove residual heat when the reactor is shut down, to 1
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l 1-6 cool the spent f uel storage pool, to sample reactor coolant water, to provide for emergency safety injection, and to vent and drain the reactor coolant system.
The engineered safety features provided for the Diablo Canyoa Units have suf ficient capacity and redundancy to protect the health and safety of the public by keeping exposures below the limits set forth in 10 CFR Part 100 for any postulated malfunction or accident, including the most severe ices of coolant accf dent.
An instrumentation and control system provides auton.atic protection against unsafe and improper reactor operation during both steady state and transient conditions. The entire operation of the plant is monitored and controlled by operators in the control roce4 'thich is located in the I
auxiliary building. !
The electrical systems generate and transmit power to the applicant's high voltage system, distribute power to the auxiliary loads, and provide control, protection, instrumentation and annunciator power supplies for f
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the units. Of f site a-c power is available f rom two 230 kV transmission l 4
lines and three 500 kV transmission lines. The 230 kV line serves the i Onaite a-c emergency standby /startup transformers for both Units 1 and 2.
Two power is supplied by redundant and independent diesel pe,erators.
diesel generators are dedicated to each unit and a fifth generstor can i serve either unit. Onsite d-c power f or each unit is available f rom three 125 v batteries.
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'l-7 The steam and power conversion system is designed to receive the heat absorbed by the reactor coolant system during nc.ms1 power operation, as well as following an amergency shutdown of the turbine generator from full load. Heat rejection under the latter condition is accomplished by steam bypass to the condenser and pressure relief to tha . atmosphere.
Auxiliary systems are supporting systems included in the facility. l some of which are required to perform cucain functions during emergency or accident conditions. Included are the cooling water systems, the heating and ventilating systers, the fire . protection systes, the procens auxiliaries, the compressed air system, the diesel generator fuel oil system, the communication systems, an:J the lighting systems.
Certain facilities and equipment are shared between Units 1 and 2.
In terms of structures, the two units share a common ruxiliary building vhere the asjor portion of the radioactive waste treatment equipment 'is shared by tt.e two units. Also, the plant is provided with a central control room located in the auxiliary building. physical separation of j i'
control panels eliminates interaction of Unit 1 and 2 control syst(ms.
The turbir.e building for Unit 2 is an extension of the Unit 1 building.
The two units also share a common raw water storage reservoir, fire pumps, fire water storage tank, diesel fuel oil storage tanks and transfer pumps, auxiliary boiler, makeup water system, plant air system -
and Imbricating oil storage system. Details of the sharing of these systems are discussed in appropriate sections of this report. l t
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1-8 1.3 Comparison with Similar Facility Designs
! Many features of the d-sign of the Diablo Canyon Units 1 and 2 are l similar to those that we have evaluated and approved previously for other nucleat power plants now under construction or in operation, particularly i
the Zion Units 1 and 2 (Docket Nos. 50-295 and 50-304). The principal 1 l
dif ference between the Diablo Ccnyon and Zion Units is he 17x17 Nel j l
assembly design which is planned for Diablo Canyon. In additint to Zion. l l
the Diablo Canyon Units are quite similar to the Trojan Nuclear Plant (Docket No. 50-344), whose operating license appifcation is currently under review by the staf f. To the extent that is feasible and appro-priate, we have made use of these previous evaluations in conducting our review of the Diablo Canyon Units. Our Safety Evaluation Reports for these other f acilities ti:st have been appreved have also been pub-l lished and are available for public inspection at the Atomic Energy ,
Commirsion's Public Document Room in Washington, D. C.
1,4 Identification o LAyents__and Contractors Pacific Gas and Electric Company (PG6E) is the sole applicant for operating licenses for Units 1 and 2 of the Diablo Canyon Nuclear Power ;l i
Station. PC&E is the architect-engineer, constructor, operator, and owner of the Diablo Canyon Nuclear Power Station, and as such assumes i
full responsibility and authority for the design, construction, startup, j and operation of Units 1 and 2. The applicant has engaged the services l of Westinghouse Electric Corporation (hereinaf ter referred to as .
Westinghouse) to design and manuf acture the NSSS for both Units.
1-9 Westinghouse is also resocnsible for fabrication of tF- nuclear fuel for the initial cores for Units 1 and 2. In addition to We .tingnouse as the contractor for the NSSS, the applicant has retainen numerous consultants who performed investigations and submitted reports to PC6E on a wide range of subjects. A list of these consultants and their subject areas is provided in Table 1.4-1 of the FSAR.
1.5 Summary of Principal Review letters ;
Our evaluation included a technical review of the information sub-adtted by the applicant, particularly with regard to the' following principal matters:
(1) We evaluated the population density and land use charseteristics of the site environs, and the physical characteristics of the site, including seismology, meteorology, geology and hydrology to estahlieh l that these characteristics have been determined adequately and have been given appropriate consideration in the final design of the plant, and that the site characteristics are in accordance with the Commiscion'. siting criteria (10 C7R Part 100), taking it.to conridera-tion tre design of the f ac.lities including the engineered scfety featur(s provided.
I (2) We evaluated the design, f abrication, construction, and testing ano I performance characteristics of the plant strettures, systems, and compor ents important to safety to determ',ne that they are in accordance with the Commission's General Design Criteria (GDC), Quality Asper-ance Criteria, Regelatery Guides and other appropriate uies, codes
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1-10 and stands-ds, and that any departures f rom these criteria, codes and standards have been identified and justified.
(3) We evaluated the expected response of the facilities to various anticipated operating transients and to a broad ' spectrum of accidents, and determined that the potential consequences of a few highly unlikely postulated accidents (design basis accidents) would exceed those of all other accidents considered. We performed conservative analyses of these design basis accidents to determine that the calculated potential offsite radiation doses that might result in the very unlikely event of their occurrence would not exceed the Commission's guidelines for site acceptability given in 10 CFR Part 100.
(4) We evaluated the applicant's engineering and construction organiza-tions, plans for the conduct. of plant operations, including the pro-i posed organization, staffing and training program, the plans for )
industrial security, and the plans for emergency actions to be taken 1
in the unlikely event of an accident that might affect the general' public, to determine that the applicant is technically qualified to safely operate the urits.
(5) We evaluated the design of the systems provided for control of the radiological effluents f rom the plant to determine that these systems are capable of controlling the release of radioactive wastes from the facilities within the limits of the Commission's regulations, and that the equipment provided is capable of being operated by the applicant in such a manner as to reduce radioactive releases to l
1evels that are as low as practicable.
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j (6) We evaluated the financial position of the applicant to determine 1 l
that the applicant is financially qualified to operate Diablo Canyon l l
Units 1 and 2. 4 1.6 Facility Modifications Required as a Consequence of Regulatory Stqff Review _
l As a consequence of the staff review, a number of design changes t I
These modifications are discussed were or will be made to Units I and 2.
The i
in greater detail within the body of this Safety Evaluation Report.
principal changes are as follows:
(1) Upgrading of the onsite meteorological program with regard to control room monitoring of certain parameters (see Section 2.3.3).
(2) Upgrading of the post-tornado availability of certain Category I systems and structures (sce Section 3.5).
(3) Design measures taken for protection against the dynamic effects i associated with pipe ruptures outside containment (see Section 3.6).
(4) Augmentation of the seismic instrumentation program (see Section 3.7).
(5) Installation of a loose parts monitoring system (see Section 5.4).
(6) Installation of appropriate chlorine protection devices in the control room (see Section 6.4).
(7) Improvement of the response time testing program for components of the reactor protection system (see Section 7.2.4).
(8) Provision for auto .ic opening of accumulator isolation valves when reactor coolant pressure exceeds a preselected value (see Section 7.3.3).
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(9) Provision for automatic tripping of the residual heat removal pumps when the water in the refueling water storage tank reaches a specified low level (see Section 7.3.4).
(10) An increase in the diversity of interlocks for the residual heat removal system to prevent possibic over-pressurf sation of this system (see Section 7.6). fi l
(11) Tentative commitment that spent fuel will not be stored in the spent l fuel pool in locations where it could be struck by a drepped cask (see Section 9.2.3). f 1 4
(12) Installation of expansion joint sleeves around each expansinn joint in the circulating water system to preclude flooding of safety related equipsent in the turbine building (see feetion 10.4).
(13) Consnitment to provide a flood wall in the turbine tmilding i
to prevent flooding in the event of a water box failure in the circulating water system (see Section 10.4).
(14 Revision of the operator requalification program (see Section l's.2).
(15) Requirements for arming of the plant security guards (see Section 13.6).
(16) Modification of the QA program for operations to provide for an independent qualified inspection staff reporting to the QA Director (see Section 17.3).
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i 2-1 2.0 SITE CHARACTERISTICS 2.1 Geography and Demography 2.1.1 Site Location The Diablo Canyon site is adjacent to the Pacific Ocean in San Luis Obispo County, California, and is approximately 12 miles west-southwest ,
of the city of San Luis Obispo, the County Seat. The site is roughly equidistant from San Francisco and Los Angeles. Figure 2.1 locates the site on a map of west-central California.
2.1.2 Site Description The site consists of approximately 750 acres near the mouth of Diablo Creek. The parcel immediately south of the creek consists of i
585 acres and has been leased to the applicant for a term of 99 years with an option to renew for an additic .al 99 years. The 165 acre parcel on the north side of the creek is owned by the applicant. The site boundary and the location of principal structures are shown in Figure 2.2. The locations of the gaseous and liquid effluent release points i
are also given. As shown in Figure 2.2, a portion of the site is bounded by the Pacific Ocean. The distance from either reactor to the nearest site boundary on land is one-half mile (approximately 800 meters) which is the exclusion distance. The minimum distance from either reactor to the ocean (mean high water) is 600 feet (approximately 200 meters).
Da land there are no activities unrelated to plant operation within the exclusion area. The exclusion area is not traversed by public highway or railroad. The applicant has stated in the FSAR that within the 585
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acre parcel leased to the Company, it has the right to use {
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2-2 excavated materials in the construction of the plant. All other mineral rights are retained by the lessor, although the applicant stated that these retained rights cannot be exercised in a manner inconsistent with the applicant's use of the land.
The of fshore area (below the mean high water line) within the exclu-sion distance is not under the applicant's routine control, and is at times entered by count ref al or sports fishing boats. The shoreline of the site is rough and precipitous and the land area below the maan high water line could be occupied only with great difficulty. Portions of the exclusion area which extend offshore will be controlled by the U.S. Coast Guard in the event of an emergency.
2.1.3 Population and Population Distribution Population data taken from the applicant's FSAR have been compared with the 1970 Census and with projections prepared by the Bureau of Economic Analysis (BEA), U.S. Department of Commerce. The nearest resi-dence is 1-3/4 miles north-northwest of the site and is occupied by two persons. The applicant has selected a distance of 6 miles to be the low population zone as defined in 10 CFR Part 100. The 1970 population within this distance is estimated to be 18. We have investigated the possibility of greatly increased residential growth within the low population zone.
There is a possibility that a large condominium development, including c resort hotel with recreational facilities, will be located within the low population zone at a distance of one to six miles south of the plant. The maximum number of residents anticipated for this development
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2-3 would be 2760 persons. Plans for this development, st.111 tentative, call for cowpletion by 1984 and require approval both by the San Luis Obispo County Planning Commit lon and the California Coastal Environment l
Commission. The population center (as defined in 10 CFR Part 100) has been selected to be the city of San Luis Obispo, a distance of 12 miles away f rom the plant, with a 1970 population of 28,036. The distance to the nearest beundary of the population cer;ter is more than one and one-third times the low population zone radius, in compliance sith 10 CFR Part 100 guidelines. Figure 2.3 shows the 1970 cumuistive population surrounding the plant as a function of distance from the plant out to l
50 miles. The cumulative population corresponding to a moderately populated area of 400 people per square adle is also shown. Comparison of the curves in the figure ebows that the site area is not heavily populated.
U.S. Census data indicate that tl.c city of San Luis Obispo showed l an increase of 37% in popuistion from 1960 to 1970, while the county of {
San Luis Obispo showed a gain of 30% over the same period. For the area I
within a 50 mile radius of the pla..; the applicant projects a population growth of 65% in the 20 year period from 1970 to 1990 and a 148% increase l
l in the 40 yr. period from 1970 to 2010. We have compared the applicant's projections with those prepared by the BEA, U.S. Department of Commerce, for Water Resources Subtrea No.1807. Figure 2.4 shows a circle of 50 mile radius around the Diablo Canyon site in relation to Water Resources Subarea No. 1807. The BEA projects a population growth of 40% for this l
I 2-4 area over the 20 year period from 1970 to 1990, and an.831 growth over the 40 year period from 1970 to 2010.
In addition to the resident population, there is a seasonal influx
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of transient vocation and weekend visitors, primarily to the beach areas, and especially during the susumer months. Within the low population zone the maximum number of persons at a single cime is estimated to be 5000.
This corresponds to the maximum daytime use of the Montana de Oro State Park, whose area of principal use is along the beach, between 4 and 5 f I
miles northwest of the site. Overnight use is considerably less, an estimated maximum of 400 persons. We concur with the applicant that f evacuation of these numbers of persons f rom the park could be accomplish.ed expeditiously and without injury in the event of an emergency.
On the basis of the Part 100 definitions of the population center distance (12 miles), the exclusion radius (one-half mile), and LPZ distance (6 miles), our analysis of the onsite . meteorological data f rom which dilution f actors were calculated (Section 2.3 of this report), and the calculated potential radiological dose consequences of design basis accidents (Section 15.0 of this report), we have concluded that the exclusion area radius and the LPZ distance are acceptable. The site meets the requirements of 10 CFR Part 20 with respect to the restricted area. )
2.1.4 Upes of_ Adjacent Lands and yaters l The San Luis Range, which attains a height of approximately 1800 feet, dominates the region between the site and U.S. I'oute 101. This upland country is used to a limited extent for grazing beef cattle and, to a very minor extent dairy cattle.
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San 1.uis Obispo County has relatively little level land, except for i
a few small coastal valleys where farm.ng in t he predominant activity. J Principal crops include vegetables, poult ry, and grain. The county's leading agricultural product is liveu.ock, which constituted over 40 1
pe rcent of the gross valut of farm products sold in 1970.
The nearest dairying activity is between 7 and 8 m'les northeast of the site. There are two small operations yielding a total of approxi- l mately 1500 gallons per day. One additional small dairy located about 10 miles east of the site produces about 800 gallons per day. Tne l
largest dairy within a 15 mile radius is some 12 miles north of the 1
site, and produceu 2200 gallons per day.
The ocean area adjacent to the plant is a small part of the larger coastal fishing grounds extending from slightly north of Point Buchon to Point San Luis. In 1966 the total annual combined sport and commer-l cial catch f rom this fishery was estimated to be 621.000 pounds of abalone, 61,000 pounds of rockfish, and 21,000 pounds of other species.
Diablo Cove, where the circulating water discharge structure is located, is est imated to have contributed, in its undisturbed state, about one percent of the fish catch mentioned above.
There are no public water supply gr~md water basins within 10 miles I
of the site. Property owners north and south of the site capture eurf ace '
water f rom small intermittent streams and springs f or private domestic use.
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2-6 2.3 Nearby Indust rial. Transportation and Military Facilities On the basis of information supplied by the applicant, there are no airports, military or industrial f acilities, gas pipelines, major high-ways or railroads within 5 miles of the Diablo Canyon site.
Significant transportation routes in the vicinity include U.S. Highway 1 l
101 which passes about 10 miles east of the site. State Route 1 which passes about 10 miles north of the site, and the Southern Pacific Railroad which roughly parallels U.S. Highway 101 and passes abocat 10 f 1
l miles east of the site. The nearest operational airport is the San Luis l Aside fron l Obispo County airport located 12 miles east of the site.
fishing vessels, the nearest marine traffic is that associated with I local coastal tankers approaching no closer than 5 to 10 miles of the plant to load and unload petroleum products. These products are stored I' at Avila Beach and Morro Bay, which are located at distances of 8 and IC miles f rom the site, respectively.
l The largest industrial complex near the site is Vandenberg Air Force Base, loceted about 35 miles south of the site. Missiles fired to the Western Pacific Missile Range from this base are not directed toward Diablo Canyon.
Because of the abser.ce of industrial, transportation or military f acilities in the area of the site, we conclude that safe operation of the plant at the Diablo Canyon site will not be adversely affected.
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2-7 3.3 Meteorology
- 3. 3.1 Regional Climatolog The climate of the Diablo Canyon site, located along the central coast of California about 200 miles west-northwest of Los Angeles, is maritime, resulting in modified temperature extremes and high humidities. l The central coast of California is along the eastern portion of the Pacific high pressure system, which generally dominates the circulation patterns of the region and keeps the principal tracks of low pressure i systems north of the area. The Pacific high pressure system weakens i and moves southward in the winter, allowing Pacific cold fronts and storm centers to move inland. This results in the well-defined rainy )
season f rom November to March.
3.3.2 _ Local Meteorology Meteorological date from several local areas have been examined.
These areas are the plant site, the city of San Luis Obispo (12 miles i
east of the site), Oceana (16 miles coutheast of the site), and Santa Maria (29 miles southeast of the site). Mean 'nonthly temperatures in this locale may be expected to range f rom about 50'F in January to about 62*F in July, August, and September. Annual average precipitation is .
about 16 inches, wit.h about 85% of this occurring from November to March. Wind data f rom the 25-foot level on the "E" tower at the Diablo Canyon site for the period July 1967 to October 1969 indicate a predom-inance of winds from the northwest (25%) and from the west-northwest (18%). The prevciling wind direction at Santa Maria for a 21 year
2-3 period of record is west-northwest . The mean wind speed at the plant site is about 11 miles per hour (mph), and at Santa Maria this value is
' aph.
Severe weather is not courxin f r. t he area. Thunderstorms can be expected on only about 3 days per year. One of the mort severe t ropfi al storms en record along the Southern California coast occurred on S ptember 24-25, 1939. This stora noved north along the coast an!
curved inlano near Los Angeles. The most severe ef fects were observed in the Los Angeles area and south. The storm had little effect on the site. The number of days having high air pollution potential average.-
about 9 per year at the site.
! .1. 3 onsi t e Met eorologi.ca l, fleas,u.reme_nt,s Progr_a,m Topography at the site is extremely complex, and it- is dif ficult to l I
document representat ive atmospheric dispersion characteristiet,. Six meteorology towere have been used by the applicant in order to repre-sent sit e meteorology . The 250-foot "E" tower will remain as t he pernanent l facility. 1his tower is located on a relatively flat plate between the rujor plant st ructures and t he coastal bluf f. Inst rument at ien on t hin tower consists of the following: (1) temperature sensors at the 25 ,
150 , and 250-foot levels; (2) men.uren nt of wind speed and directson at the .'s- and 250-f oot levels; (3) bivines taeasurin,; horizental and vert iul wind flut t uat ior s at the 25- and 250-f oor levels; and (4) mead-urement of dewpoint t emperature at the 25-foot level. The prir:ury data recording, system ut ilizes magnet ic tape, wi t h st r ip chart s as the secon-dary system.
2-9 The applicant has presented in Section 2.3. 3 of the FSAR a proposed This orogram f or control room monitoring of meteorological parameters.
, . gram is currently being reviewed, but the applicant must provide more detailed information before the evaluation can be completed. Resolution of this it em will be reported on in a supplement to this Safety Evaluation Report.
The applicant has submitted joint frequency distributions of wind speed and direction by atmospheric stability (as defined by vertical I l
temperature gradient between the 25- and 250-foot icvels) for the 25-and 250-foot levela. Data for both of these levels have been supplied for the period July 1967 to October 1969. Additional joint frequency l distributions of wind speed and direction by atmospheric stability (as defiwed by vertical angle fluctuations) for the 25-foot level for the perious October 1969 to March 1971 and April 1972 to September 1972 were I
suomitted by the applicant, as well as similar joint frequency distribu-tions f or these saac periods with atmospheric stability defined by azimuth angle fluctuations. All joint f requency distributions were sub-mitted in accordance with the requirements of Regulatory Guide 1.23, 1
"Onsite .'let=orological Programs," and are acceptable.
We have examined all available onsite meteorological data, and have concluded that the meteorological assumptions selected by the applicant ti, estinate expected accident dispersion conditieis f rom buildings and vents art dertuately conservative; these as,umptions were taken from Regulatory Guide 1.4, " Assumptions Used for Evaluat int, the Potential Radiological Consequences of a Loss of Coelant Ac ' dent for Pressurized l
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l Water Reactors." Annual average dispersion conditions for releases l from buildings and vents have been evaluated using wind speed and direction i measurements from the 25-foot level, and with stability defined by l
vertical temperature gradient between the 25- and 250-fcot levels for the period July 1967 to October 1969. ,
2.3.4 Short-Ters (Accident) Diffusion Estimates In the evaluation of short-tern (0-2 hours at the exclusion distance and 0-8 hours at the boundary of the LPZ) accidental releases from buildings and vents, a ground-level release with a building wake f actor, {
cA, of 800 square meters (m ) was assumed. The relative concentration (X/Q) for 0-2 hours, which is exceeded 5% of the time, was calculated i I
to be 5.3 x 10 ' seconds per cubic meter (sec/m3 ) at the exclusion
~
distance of 800 meters (m). This calculation was performed using the model and assumptions described in Regulatory Guide 1.4. This relative !
concentration is equivalent to dispersion conditions produced by Pasquill Type F stability with a wind speed of 1.0 meter per second (m/sec). The relative concentration, which is exceeded 5% of the time, at the outer boundary of the LPZ (9654 m from the reactors) was calculated to be
~
- 2. 4 x 10 sec/m for the 0-8 hour time period. The corresponding estimated relative concentration at the LPZ for the 8-24 hour time
-6 period is 4.8 x 10 sec/m . For the 1-4 day and 4-3G day time periods,
-6 ~
these concentrations are 1.5 x 10 and 3.4 x 10 sec/m , respectively.
t __ __________a
1 2-11
)
2.3.5 tong-Tern (Routine) Dif fusion Estimates i l
ne highest overland off site ennual averc.ge relative concentration j
-6 for releases from buildings and vents was calculated to be 7.2 x 10 sec/m at a location on the minimum site boundary 800 m north of the l
1 containment structures.
2.3.6 Conclusions j We have concluded that the applicant's meteorological assumptions taken from Regulatory Guide 1.4 are adequately conservative, and that they provide an acceptable design basis for the calculation of relative concentrations for accidental releases to the atmosphere f rom buildings and vents.
We will require further discussion with the applicant regarding his proposed program for control room tonitoring of meteorological parameters.
Issues yet to be resolved include the method to be used for rapid assess-ment of atmospheric stability, and the control room display cf pertinent parameters. The applicant crast submit to the staf f for evaluation any j proposed modification of the present Meteorological Program. No changes ;
l to this program may be implemented without prior staff approval.
The applicant must submit at least one additional year of onsite data.
These data should preferably be taken from the most recent 2-3 year compila-tion of onsite meteorological data, and should be submitted in the form of joint frequency distributions of wind speed and direction by atmospheric 1
stability (as defined by vertical temperature gradient) in accordance with l
b l \
c 1
2 L _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ ]
l 2-12 the recommendations of Regulatory Guide 1.23. These data should be submitted in a timely manner in order to permit verification of the l relative concentration values.
I 2.4 Hydrology Because of delays by the applicant in the submittal of required in for mation, the staff's review of site hydrology has not yet been completed. Our evaluatioa of this area vill be contained in a supplement to this Safety Evaluation Report.
2.5 Geolo gy , Se ismology , and Foundation Engineering Because of delays by the applicant in the subrittal of required i
information, the staf f's review of geology, seismology, and foundation I
engineering has not yet been completed. Our valuation of these areas will be contained in a supplement to this Safety Evaluation Report .
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3-1 CAPONE.
1 Tji,,J10lPKNT, AND SYSTFy.S I
- 3. 0 DESIGN CRITERIA - STRUCIUREf3
- l 3.1 Conformance with AEC General Desip Crlteria i The Diablo Canyon Units 1 and 2 were designed and are being con- l structed on the basis of the presosed General Design Criteria (CDC),
which were published on July 11, 1967. Since February 20, 1971. when I
has the AEC published the CDC for Na:1 ear Power Plants, the applicant attempted to comply with the inter.t of the newer criteria to the extent
.ommitments. Any exceptions j that is practical, recognizing previous design i to the 1971 CDC which have been taken becauss of earlier design or con-struction commitments are identified ir. the FSAR in the discussion of the As a result, corresponding criterion (see Appendix 3.lA of the FSAP.).
our review assessed the plant against the General Design Criteria now in effect, and we have concluded that the plant design conforms to the intent of these newer criteria.
3.2 Classification of Structures, Caroor.ents, and_ Systems 3.2.1 Sc_ismic Classification Structures, systems and connonents important to safety that are j l
required to be designed to withstand the ef f ects of a safe shutdovn carthquake (SSE) and remain futctional have been properly classified as I l i Seismic Category 1 (Applicant's Design Class 1) items. These plant j
f eatures are those necessary te assure (1) the integrity of the reactor k coolant pressure boundary (RCTS), (2) the capability to shutdown the reactor and maintain it in a safe shutdown condition, or (3) the i
i i l
- t 1
)
3-2 capability to prevent or mitigate the consequences of accidents which could result in potential of fsite exposures comparable to the guide-line exposures of 10 CFR Part 100.
All other structures, systems and components that may be required for operation of the facilities are designed to other than Seismic Category I (Applicant's Design Classes II and III) requirements. Included in this classification are those portions of Category I systems which are not l
required to perform a safety function. Structures, systems and components I important to ssfety that are designed to withstand the effects of an SSE and remain functional have been identified in an acceptable manner in Tat ie 3.2-4 of the FSAR and on system piping and instrumentation diagrams, Figures 3.2-01 through 3.2-27 of the FSAR.
l l The basis for acceptance in our review has been conformance of the 1
applicant's designs, design criteria, and oesign bases for structures, 1
l systems and components igorcant to safety with: (1) the Commission's regulations as set forth in AEC General Design Criterion No. 2; (2) t.he positions set forth in Regulatory Guide 1.29, " Seismic Design Classification";
I and (3) industry standards.
We have concluded that structures, systems and components important to safety that are dewigned to withstand the effects of a safe shutdown l
l earthquake and remain functional have been properly classified as Seismic Category I items in conformance with the Commission's regulations, the applicable Regulatory Guide and industry standards. Design of these
1 l
)
3-3 I
items in accordance with Seismic Category I requirements provides i reasonable assurance that the plant will perform in a manner providing adequate safeguards for the health and safety of the public.
3.2.2 System Quality Group Classifiestions Fluid system pressure-retaining components important to safety will j
be designed, fabricated, erected and tested to quality standards commen- l The f surate with the importance of the safety function to be performed.
f applicant has applied a classification system (Code C1seses 1. II and III) !
to those fluid containing components which are part of the reactor I coolant pressure boundary and other fluid systems igortant to mafety where reliance is placed on these systems, in order to: (1) prevent or 1 mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary; (2) permit shutdown of the teactor and maintain it in the safe shutdown condition; and (3) contain l radioactive material. These fluid systems have been classified in an acceptable manner in Tables 3.2-2, 3.2-3 and 3.2-4 and ce systein piping )
and instrumentation diagrams, Figures 3.2-01 through 3.2-27 of the FSAR.
The basis for acceptance in our review has been conformance of the applicant's designs, design criteria, and design bases for pressure-retaining components such as pressure vessels, beat exchangers, storage )
tanks, pumps, piping and valves in fluid systems important to safety l
)
with: (1) the Commission's Regulations as set forth in AEC Ceneral Design ru 1
3-1 Criterian L. 1; ( .' ) the requirensnts of t he ( odes specified in f ect ion i
- 50. 55a of 10 tTK Par t 50; (3) t he positi ons set fort h in Regulatory Luide 1. 26. "Qu.il i t y Group Cl.v;s i: i t .it lone, ans! St andards"; and (I.)
l industi) atuidard.
l Ws have i oncluded t hat f luid syst em pi c% ore-ret aining component s it:por t ant to safety that are designeJ, f abricat ed, erected and testeJ to quality standards in conformance wit h the Comission's Regulat ions, the applicable Regulatory Guide, and industry standards are acceptable.
Conf ormans e wit h t hese requirement s provides reasonable assurance that the plant will perform in a manner providing adequate safeguards for the healt h and saf et y of the publis.
- 1. J Wind and Tornado Deshn, Criteria 1 1
All Seismic Cat egory 1 (also referred to as Category I) structures ;
l i
exposed to wind forces have been designed to withstand the ef fects of the
{
design wind. The design wind specified has a velocity of 80 mph based on a recurrecte int erval of 100 years. 'Ibe procedures used to transf orm t he j 1
wind velocity into pressure loadings on st ruct ures and the assoc iated l
vert ical dist ribut ion of wind pressures and gust f actors are in accordance will. t he laternat ional Conferent e of Building Of ficials "L'niform Building i
imh - 19'J l d i t t un. "
i l Alt heugh a t ornade design ( riter ion was not required as a condition l
f or e ne grant ing os const r uct ion permit s for t hese units, a review of the f i
tornaJo resisting capabilities c' Category I and certain non-Category I i
structures has been undertaken by the applicant at the request of the ;
4 j
f.
/l
3-5 staff. The objective of the review was to establish capabilities of the Category I structures, as designed and constructed, to withstand tornadic wind pressure and the associated atmospheric pressure drop and tornado borne missile effects. The results of this review indicate that the maximum postulated tornadic wind velocity (300 mph) will not cause a loss of coolant accident or structural damage which would impair con- i tainment integrity. All other Category I structures or structures housing Category I compor.ents are capable of withstanding the wind ,
l effects of at least a 225 mph tornado without f ailure of major structural ;
1 elements. Their resistance to the hypothetical missiles corresponds to 1
a 150 mph wind velocity.
l The procedures used to transform the tornado wind velocity into l
l pressure loadings are similar to those used for the design wind loadings l as discussed above. The pressure drop associated with the tornado is treated as a static uniform load applied on vertical and horizontal projected areas of the structures. Tornado missile ef fects have been l l determined using procedures which are discussed in section 3.5 of this l
l l report. The total effect of the tornado on Category I structures was l
determined by the appropriate combination of the individual ef fects of the tornado wind pressure, pressure drop and associated missiles.
Structures are arranged on the plant site and protected in such nanner that the collapse of structures which do not have tornado resisting capability will not affect those which must withstand the tornado ef fects.
1 -
l
3-6 The criteria used in the design of Category 1 structures to account
~
for the ' loadings due to specific winds postulated to occur at the site, and the methods used in determining those loads provide a conservative ba61s for the plant design. The use of these loading criteria provides reasonable assurance that, ic the event of wind (or a tornado), the structural integrity and safety function of Seismic Category 1 structures will not be impaired by the specified environmental forces.
We have concluded that conformance with these criteria is an accept-able basis for satisfying the requirements of AEC General! Design-Criterion No. 2.
3.4 Water Level (Flood) Design Criteria The design flood level resulting f rom the most unfavorable condition or combination of conditions that produce the maximon water level of the site is discussed in Sect.on 2.4 of the FSAR (Hyd ulogy). .This dis- l cussion indicates that it is not possible to develop suf ficient ponding to flood safety related buildings. Thus, the depth of water at the plant location for the Probable Maximus Flood is sero.
The auxiliary saltwater systen is the only safety related system 1 that has components that would be affected by the design combination l of tsunami-storm wave activity. The intske structure is designed so that the auxiliary saltwater pumps are housed != separate water-tight compart-ments, which assures that the pumps will eterste during the storm. The auxiliary saltwater system intake structure is also provided with a ventilation system that is above the postulated flood level.
I 4
3-7 ,
i On the basis of our review,.we have concluded that the design of the ,
i intake structure to withstand the effects of tsunami-storm wave activity provides reasonable assurance that the safety related equipm6nt is adequately prc ' ected and may be expected to perform its required safety functions. Furthermore, we conclude that the design satisfies the requirements of AEC General Design Criteria Nos. 2 and 4 as related to the environmental design bases for safety related structures.
3.5 Missile Protection Criteria The design of essential structures and vital equipment has taken into account the effects of a spectrum of tornado-borne missiles and internally generated missiles associated with component overspeed failures and missiles that could originate from high-pressure system ruptures. The j design assures that there will be no loss of function of a Seismic Category I structure or of essential system or component functions as a result of missiles. The missilos applicable to each of these structures )
I and pieces of equipment have been adequately identified, and their ]
characteristics defined.
Initially, the tornado resisting capability of all Seismic Category _
1 structures was calculstad for a limited spectrum of tornado-borne missiles (see Page 3.3-5 of the FS..u). In response to our request, the applicant increased this missile spectrum to include the following items l that could be present at the site or dislodged from structures by tornadic winds to become missiles: (1) a utility pole 13.5 inches in l
9 8
3-8
)
l l
diameter and 35 f eet long, with e density of 43 pounds per cubic foot '
(ib/ft ); (2) a one-inch diameter solid steel rod, 3 feet long with a density of 490 lb/ft ; and (3) pieces of 6 and 12 inch schedule 40 pipe, each 15 feet long with a density of 490 lb/ft . in general, essential components contained in Seismic Category I structures are inherently protected by virtue of the fact that the seismic and other design requirements result in structures that are tornado resistant.
j The analysis of structures, shields and b'arriers to determine ti.e
)
ef fects of missile impact was accomplished in two steps. In the first step, the potential damage that could be done by the missile in the immediate vicinity of impact was investigated. This was accomplished by estimating t he depth of penetration of the missile into the tapacted s t ruc t ure . In the second step of the analysis, the overall structural I
response of the target when impacted by a missile was determined using established methods of impactive analysis.
The design procedures used to determine the ef fects and Joading on Seismic Category I structures by design basis missiles selected for the plant provide a conservative basis for engineering design to assure adequate protection f rom the ef fects of missila impacts. The use of this information provides reaso:able assurance that, in the event of design basis missiles striking Seismic Category I structures, the struc-toral integrity of structures will not be impaired or degraded to an extent that will result in a loss of required protection. Seismic l
3-9 Category I systems and components located within these structures are, i therefore, expected to be adequately protected against the effects of l
rd ssiles. Conformance with these missile protection design procedures l
is an acceptable basis for satisfying the requirements of AEC General Design Criterion No. 4.
Based on the results obtained f rom the revised tornado missile analysis, the applicant evaluated the plant layout to determine what systems or components necersary for safe reactor shutdown could be damaged by potential tornado missiles. As a result of this evaluation, the applicant has proposed design modifications to f urther iciprove the post-tornado availability of components in the emergency power system.
These modifications consist of tornado missile barriers that protect the louvers in the diesel generator rooms. We find these modifications to be acceptable. With the exception of these modifications, the appli-cant's analysis indicated that existing shielding or equipment separation provided adequate protection to assure that safe shutdown of the reactor can be achieved and maintained. We have eviewed the information sub-mitted by t!ae applicar: and concur with his conclusions.
- 3. 6 Protection Against the Dynamic Effects Associated with the Postulated Rupture of P_ipinJ i
f The applicant's criteria which were used for identifying high energy 1 fluid piping and for postulating pipe break locations, break orientations and break flow areas are equivalent to the criteria for piping inside
(
?
l l
1 3-10 l I
containment set forth in Regulatory Guide 1.46, " Protection Against pipe Whip Inside Containment," and are also consistent with the criteria I
stated in the A. Cinmbusso letter of December 18, 1972 for piping out- '
side of containment. The provisions for protection against the dynamic ef f ects associated with pipe rupture (inside containment) and the I
resulting discharging coolant provide acceptable assurance that, in the event of the occurrence of the combined loadings imposed by an earth-quake of the magnitude specified for the safe shutdown earthquake (SSE) and a concurrent single pipe break of the largest pipe at any one of the design basis break locations, the following conditions and safety i
functions will be accommodated and assured:
(1) The magnitude of the design basis loss-of-coolant accident cannot i 1
be aggravated by poter.tial multiple f ailures of piping; 1 l
(2) The reactor emergency core cooling systems can be expected to per-form their intended function; and (3) Structures, systems and components important to safety will be appropriately protected.
The analytical 'nethods and procedures that were used to determine pipe motion subsequent to rupture and the pipe-whip restraint dynamic interaction appropriately consider the structural characteristics of the system. The pipe-whip restraints are designed to withstand the resultant loadings in accordance with acceptable criteria.
On the basis of our review, we have concluded that the criteria used fo: the identification, design and analysis of piping systems where -
- P
3-11 postuisted breaks may occur constitute an accep.able design basis in meeting the applicable requirements of AEC General Design Criteria Nos.
1, 2, 4, 14, & 15, and are :ensistent with Regulatory staff positions for plants currently under review for operating licenses.
With regard to piping outside the containment, the appitcant has 1 submitted a final report which presents the results of the investigations l
[
conducted to determine the consequences of postulated ruptures of high l Included in this report energy fluid piping outside the containment.
l are definitions of criteria and methods employed in the analyses, the identification of high energy fluid piping outside containment, and the structures and systems required for safe shutdown of the reactor following postulated ruptures of this piping. A summary of the analysis results for breaks in the main steam piping between the containment and the turbine stop valves, and breaks in the feedwater piping between the containment and the feedwater pues, including proposed design modifi-cations, has been reviewed. Pipe break effects analysed included jet impingement, pressurization of compartments, water flooding, and the environmental effects of pressure, temperature, and humidity.
We have reviewed the material presented and have found it to be in accord with our requirements. For the breaks in the main ste.am and feed-water piping that were analysed, we conclude that the applicant has developed an adequate design so that postulated pipe breaks outside of containment will not prevent safe shutdown of the reactor. The results 9
4 ,
4
3-12 of our f dnal evaluation of the applicant's report on pipe break outside containment will be documented in a supplemeat to this Saf ety Evaluation i Report.
3.7 Seismic Desip I The seismic design response spectra curves were presented in the PSAR and approved prior to the issuance of the construction persits f or Units 1 and 2 of the Diablo Canyon Plant. The modified earthquake time i
histories used for component equipment design are adjusted in amplitude i and f requency to envelope the response spectra specified for the site.
We conclude that the seisaic input criteria proposed by the applicant provide an acceptable basir, for seismic design.
Hodal response spectrum, multi-degree-of-freedom, and normal mode- l time history methods are used for the analysis of all Seismic Category I structures, systems and components. The vibratory motions and the associated mathematical models account for the soil structure interaction
! and the coupling cf all Category 1 structures and plant equipment.
Governing response parameters have been combined by the square root of the sum of the squares to obtain the modal maximums when the modal response E spectrum method was used. Tne absolute sum of responses is used for closely spaced frequencies. Horizontal and vertical floor spectra l inputs used for design and test verification of structures, systems ano components were generated by the normal mode-time history method.
Torsional loads have been adequately accounted for in the seismic analysis l
l
I 3-13 of tFe Category ! structures. Vertical ground accelerations were assumed to be 2/3 of the horizontal ground accelerations, and the horizontal and vertical effects were combined simultaaeoosly. Constant vertical load factors were esployed only where analysis showed suf ficient vertical rigidity to preclude significant vertical amplifications in the seismic system being analyzed.
We have reviewed the FSAR and applicable Amendments and find the seismic system and subsystem dynamic analysis methods and procedures proposed by the applicant to be acceptable.
l The type, number, location and utilization of strong motion accelero-1 graphs to record seismic events and to provide data on the f requency, 1
l map 11tude and phase relationship of the seismic response of the contain-l ment structure corresponds to the recommendations of Regulatory Guide 1.12. " Instrumentation for F.arthquakes." Supporting instrumentation will be installed on Category I structures, systems, and components in order l
to provide data for the verifier. tion of the seismic responses determined 1
analytically for such Seismic Category I items.
~
l We conclude that the seismic instrumentation program proposed by the applicant is acceptable.
- 3. 8 Design of Category I Structures 3.8.1 Concrete Containagnt The reactor coolant system is enclosed in a reinforced concrete containment as described in Section 3.8.2 of the FSAR. The containment l
5 f
R 3-14 ctructure has been designed in accordance with applicable subsections of the ASME Boiler and Pressure Vessel Code,Section III, and ACI 318 to resist various combinations of dead loads, live loads, environmental loads (including those due to wind, the OBE, and the SSE), and loads genersted by the design basis accident (including pressure. temperature ar.d associated pipe rupture effects). The effects associated with postu-1sted rupture of high-energy pipes such as reaction and jet impingement fcrces and impact effects of whipping pipes have also been considered in the design.
The static analysis for the containment ehell and base utilises methods that have been previously applied. Likewise, the liner design for the contain=nt employs methods similar to those previously accepted.
The choice of the materials, the arrangement at the anchors, the design criteria and design methods are similar to these evaluated for previously I licensed plants. Materials, construction methods, quality assurance and quality control sensures are covered in the FSAR, and, in general, are i
sisdlar to those used for previously accepted facilities.
I Prior to operation, the containment structure will be subjected to cn acceptance test in accordance with the recommendations given in !
i Regulatory Guide 1.18, " Structural Acceptance Test for Concrete Primary 1 Reactor Containments," During this test the internsi pressure will be i
1.13 times the em sinment desigr: pressure of 47 psig.
Although the ccatainment is considered to be a prototype, the appit- l cant took exception to paragraph C5 of Regulatory Guide 1.18 in that the 2 3
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3-15 i
strain measurements will be taken near inside and outside f aces of the containment walls onty, omitting the midpoint location. We find this to be acceptable for an operating license.
The criteria used in the analysis, design and construction of the concrete conte.inment structure to account for anticipated loadings and 1
postulated conditions that may be imposed upon the structure during its !
service lifeties are in canformance with established criteria, codes, I
standards, and specifications acceptable to the P.egulatory staff. The use of these criteria as defined by applicable codes, standards and specifications; the loads and loading combinations; the design and I
analysis procedures; the structural design criteria; the materials, quality control and special construction techniques; and the testing
. requirements provide reasonable assurance thaz, in the event of winds, tornadoes, earthquakes and various postulated accidents occurring inside 1 and outside the containment, the structure may be expected to withstand the specified design conditions without impairment of its structural integrity and safety function. Conformance with these criteria, codes, specifications, and standards constitutes an acceptable basis for i j
satisfying the requirements of AEC General Design Criteria Nos. 2, 4 16, and 50.
3.8.2 Concrete and Structural Steel Internal Structures _
The containment interior structure consists of a shield wall aromid the reactor, secondary shield walls, and other interior walls, compartments i
e
. . . . . . J
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I 3-16 and floors. The principal code used in the design of concrete internal structures is ACI 318-63, " Building Code Requirements for Reinfor(ed Concrete." For steel internal structures the AISC document, "Specif1-cation for the Design, Fabrication and Erection of Structural Steel j
f or Buildings," was used. The containment concrete and steel internal structurcs have been designed tc resist various combinations of dead 1
and live loads, acciuent induceu loads (including pressure and jet l I
loads), and seismic loads. The load combinations used cover those cases likely to occur and include all loads which umy act simultaneously. The design and analysis procedures that have been used for the internal structurce are identical with those approved for previously licensed plants and, in general, are in accordance with procedures delineated in the ACI 318-63 Code and in the AISC Specification for concrete and steel structures, respectively.
The containment internal structures have been designed and propor-tioned to rem 11n within limits established by the Regulatory staf f under the various load combinations. These limits are, in general, based on the ACI 318-63 Code and on the A15C Specification, modified as appropriate for the specific loac combinations. The materials of construction, and their fabrication, construction and installation, are also in accordance I with the previously mentioned codes and specifications.
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3-17 The criteria that have been used in the analysis, design and construc-tion of the internal structures of the containment, to account for anti- '
cipated loadings and postulated conditions that may be isposed upon the structures during their service lifetime are in accordance with established criteria, codes, standards, and specifications acceptable to the Reg ory staf f. The use of these criteria as defined by applicable ed e, standards and specifications; the loads and loading combinations; the design and analysis procedures; the structural acceptance ,
criteria; trd the materials, quality control and special construction techniques; provide reasonable assurance that, in the event of earthquakes !
and various postulated accidents occurring inside the containment, the interior structures may be expected to withstand the specified design conditions without impairment of their structural integrity and safety function. Conformance with these criteria, codes, specifications and standards constitutes an acceptable basis for satisfying the require-ments of AEC Ceneral Design Criteria Nos. 2 and 4.
3.8.3 Other Seisric Category I Structures Seismic Category I (also referred to as Category 1) structures other than containment and its interior have been built from structural steel and reinforced concrete members. The structural components consist of slabs, walls, beams and columns. The design method for reinforced concrete followed that specified in the ACI 318-63 Code. Structural steel components were designed in accordance with the AISC specifiestions.
The concrete and steel Category I structures have been designed to resist
i 3-18 )
various combinations of dead loads, live loads, environmental loads I l
(including those dus te wind, the OBE and the SSE), and loads generated j by postulated ruptures of high energy pipes (such as reaction and jet impingement forces, compartment pressures, and impact effects of whipping pipes). The design and analysis procedures that have been used f or j l
these Category I se ructures are the same as those approved on previously j l
licensed plants. The various Category I structures have been designed j 1
and proportioned to remain within limits established by the Regulatory ]
l staff under the various load combinations. These limits have been modified j as appropriate for load combinations that are considered extreme. The l materials of construction, and their fabrication, construction and installation are in accordance with previously mentioned codes and specifications, i.e., ACI 318-63 and AISC specifications. l l In order to provide assurance that the function of Category I equipment located in the turbine building and the intake structure l
(both Seismic Category II structures) will not be adversely affected in the event of a safe shutdown earthquake, these structuren have been reviewed fcr that earthquake to assure that they would not collapse.
These analyses have shown that some yiciding would occur in the turbine building, but that this yielding would be limited to safe values. All l 1
t, tresses in the intake structure would be less than yield. For components located in both of these structures, the occurrence of an SSE will not impair the capability of S.:ismic Category I equipment to perform its safety related function. .
4' 4 I
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The criteria used in tre analysis, design and construction of all !
I the plant Category I structures to account for anticipated loadings and ]
d l
postulated conditions that may be imposed upon each structure during l 4
its service lifetime are in conformance with established criteria, J l
codes, standards, and specifications acceptable to the Regulatory staff.
The use of these criteria as defined by applicable codes, sr.andards and )
specification; the loads and loading combinations; the design and analysis procedures; the structural acceptance criteria; the materials, quality control and special construction techniques; provide reasonable assurance that, in the event of winds, tornadoes, earthquakes and various postulated occidents occurring within these structures, they l
may be expected to withstand the specified design ponditions without impairment of theit structural integrity and safety function. Confo rm-ance with these criteria, codes, specifications, and standards consti- i tutes an acceptable basis for satisfying the requirements of AEC General j Design Cr ':ria Nos. 2 and 4.
3.8.4 Foundat. and Concrete Supt. orts Foundations of Category I structures are described in Sections 3.8.1 and 3. 5. 2 of the FSAR. Primarily, these foundations are reinforced con.-
crete of the est type. The major code used in the design of these con-crete mat foundations is ACI 318-63. These concrete foundations have <
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3-20 live loads, been designed to resist various combinations of dead loads, environmental loads (including those due to wind, the OBE and the "iE), l The ;
and loads generated by postulated ruptures of high energy pipes. I i
design and analysis procedures that have been used for these Category 1 1 1
foundations are, the same as those approved on previously licensed plants and, in general, are in accordance with procedures delineated in the ACI 318-63 Code. The various Category I foundations have been designed and proportioned to remain within limits established by the Regulatory 1
staf f under the various load combinations. These limits h&ve been modified as appropriate for load combinations that are considered extreme.
The materials of construction, and their fabrication, construction and installation, are in accordance with the ACI 318-63 Code. l The criteria used in the analysis, design and construction of plant Category I foundations to account for anticipated loadings and postulated conditions that may be imposed upon each foundation during its service lifetime are in conformance with established criteria, codes, standards, The use of l and specifications acceptable to the Regulatory staf f.
these criteria as defined by applicable codes, standards and specifica-tions; the loads and loading combinations; the design and analysis procedures; the structural acceptence criteria; and the materi11s, quality control and special construction techniques; provide reasonable assurance that, in the event of winds, tornadoes, earthquakes and various 1
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21 postulated accidents, the foundations may be expected to withstand the specified design conditions without impairmant of their structural integrity and safety function. Conformance with these criteria, codes, specifications, and standards constitutes an acceptable basis for sa:is- 4 fying the requirements of AEC Ceneral Design Criterir. Nos. 2 and 4.
3.9 Mechanical Systems and Components 3.9.1 Dynamic Systes_ Analysis and Testina l
In order to assure that the vibration of piping systems is within 1 acceptable le els, the applicant will conduct a piping vibration opera-1 tional test program. The preoperational vibration dynamic effects test program that will be conducted on safety related ASME Class A and Class B piping systems and piping restraints during startup and the initial operating conditions is acceptable t.o the staff. These tests will pro-vide adequate assurance that the piping and piping restraints of the systems have been designed to withstand vibrational dynamic effects due I
to valve closures, pump trips, and operating modes associated with the design operational tran::ients. The tests, as planned, will develop loads similar to those experienced during reactor operation, and are l
consictent with recent Regulatory staff positions concerning preopera-i tional piping dynamic ef fects test programs for other pl nts. Compliance with this test program constitutes an acceptable basis for partial fulfill-ment of the requirements of AEC General Design Criterion No. 2.
With regard to flow-induced vibrational testing of reactor internals i for Diablo Canyon Units 1 and 2, the applicant has identified Indian l
e
l 3-22 Point Unit 2 as the prototype plant for Unit 1, and has established Trojan as the prototype plant for design verification of Diablo Canyon Unit 2. Two prototypes have been designated because Unit I has a thermal shield while Unit 2 utilizes neutron rads. These designations have been made in accordance with the provisions of Regulatory Guide 1.20
" Vibration Hessurements on Reactor Internals." The applicant will perform additional confirmatory vibration testing and subsequent visas 1-inspection es part of the Diablo Canyon preoperational tests to provide added confirmation of the capability of the structural elements of the reactor internals to sustain flow-indaced vibrations. The proposed program is consistent with the positions stated in Regulatory Guide 1.20.
We have reviewed the preoperational vibration test program proposed by the applicant for verifying the design adequacy of the reactor inter-nals under loading conditions that will be comparable tc those experienced during operation. We find this program to be acceptable. The combination of tests, predictive analysis, and post-test inspection will providt adequate assurance that the reactor internals can be expected, duri g their service lifetime, to withstand the flow-induced vibrations resulting from reactor operation, without loss of structural integrity.
We have reviewed the preoperational vibration test program that will be performed in accordance with Regulatory Guide 1.20 for assurance tha:
1 1
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it constitutes an acceptable basis for demonstrating the design adequacy I
of the reactor internals in satisfying the applicable requirements of AEC General Design Criteria Nos. 2 and 14. We find that the progras does constitute such an acceptable basis. If the test results on the prototype reactors mentioned previnusly are not found to be acceptable, i
we will require that the test programs for Diablo Canyon Units 1 and 2 l l
be expanded to obtain the appropriate test data.
The applicant has performed a dynamic system analysis ot' the reactor. .q internals and of the broken and unbroken piping loops. This analysis i
provides an acceptable basis for confirming the stnctural design adequacy of the react or internals and the unbroken piping loops to with-stas.d the combined dynamic effects of the postulated occurrence of a l
design basis accident (LOCA) and a safe shutdown earthquake. The analysis l providen, adequate assurance that the combined stresses and strains in the components of the reactor coolant system and reactor internals will not exceed the allowable design stress and strain limits for the materials of construction, and that the resulting deflections or displacements of any structural elements of the reactor internals will not distort the reactor internals geometry to the extent that core cooling is impaired.
The asrurance of structural integrity cf the reacter internals under the postulated SSE and the most severe LOCA conditions provides added confidence that the design can be expected to withscand a spectrum of lesser pipe breaks and seismic loading combinations.
9
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I We have concluded that the applicant's design procedures and J
(
analytical techniques represent acceptable bases for structural design I
of the reactor internals for Df ablo Canyon Units-1 and 2.
ASME Code Class 2 and 3 Components j 3.9.2 i f
All Category 1 pressure retaining systems, components and equipment 1 i
outside of the reactor coolant pressure boundary (RCPB) are designed to
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sustain norac1 loads, anticipated transients, the operating basis earth-quake, and the safe shutdown earthquake within design limits which are comparable to those outlined in AEC Regulatory Guide 1.48, " Design Lir.its and Loading Combinations for Seismic Category I Fluid Systems Components."
The specified design basis combinations of loadings as applied to the derign of the safety related ASME Code Class 2 and 3 pressure-retaining components in systems classified as Seismic Category 1 provide reason-able assurance that in the event (1) an earthquake should occur at the site, and (2) other upset, emergency or f aulted plant transients should occur during normal plant operation the resulting combined stresses imposed on the system components may be expected not to exceed the allowable design stress and strain limits for the materials of construc-tion. Limiting the stresses under such loading combinations provides a conservative basis for the cesign of the systes components to withstand the most adverse combinat ions of loading events without gross loss of !
st ructural integrity. The applicant's design load combinations and I
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3-25 t associated stress and deformation lia.its specified for all ASME Code
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Class 2 and 3 components constitute an acceptable basis for design in !
1 satisfying AEC General Design Criteria Nos. 1, 2, and 4, and are con-sistent with recent Regulatory staff positions. 1 The applicant has conducted component test programs, supplemented by j l
analytical predictive methods, which provide adequate assurance that the j capability of ASME Code class 2 and 3 active pumps and valves (1) to withstand the igosed loads associated with normal, upset, emergency, I
and faulted plant conditions without loss of structural integrity, and (2) to perform the " active" function (i.e., valve riosure or opening),
is confirmed under conditions and combinations of conditions comparable to those expected when a safe plant shutdown is to be effected, or the consequences of an accident are to be mitigated.
We have concluded that the design and analytical procedures used by i
the applicant provide reasonable assurance of pump and valve operability. '
3.10 Seismi_c Qualification of___ Category I Instrumentation and Electrical l Equipment Instrumentation and electrical components required to perform a safety function are designed to meet Seismic Category I requirements. l These requirements established by the seismic syscem analysis have been incorporated into equipment specifications to assure that the equipment i purchased or designed meets seismic requirements equal to or in excess of the requirements for Category I components; this assurance can be provided either by appropriate analysis or by qualification testing. )
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The applicant has implemented a seismic qualification program for .
Category I instrumentation and electrical equipment and the associated l c ,'
supports for that equipment to provide assurance that such equipment Ll l
can be expected to function properly, and that structural integrity of
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the support s will not be impaired during the excitation and vibratory
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forces imposed by the safe shutdown earthquah and the conditions of post-accident operation. The applicant has based his instrumentation and equipment qualification test program on Topical Report WCAP-8021,
" Seismic Testing of Electrical and Control Equipment (PG6E Plants)," .
~i i II and has also referenced IEEE Std 344-1971. This WCAP report is pres- ;
i, ently under review by the staff (see Section 7.8 of this report). This !
qualification prograu, when completed, will constitute en acceptable k basis for satisfying staff requirements and ALC Ceneral Design Criterion No. 2. Resolution on this item will be reported on in a supplement to
[
this Saf ety Evaluation Report. ,
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4.0 REACTOR
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l 4,1 Sunenary Description The Wucicar Steam Supply System design for Diablo Canyon is similar to that reviewed and approved for the Zion Nucisar Power Station n s U it 1 and 2 and D. C. Cook Nuclear P16nt Units 1 and 2 except for the following:
(1)
Diablo Canyon Units 1 and 2 have initial core power levela of 3318 and 3411 st, respectively, which are 2.7 and 5% higher than the Zion and D. C. Cook ratings of 3250 W t.
The power levels for the Diablo Canyon facility apply to either the 15x15 or the 17x17 feel assemblies, as discussed in Section 4.4 of this report; (2)
Diablo Catoyon will use a 17x17 fuel assembly design as compared to ue 33x15 design for the Zion and D. C. Cook facilities; (3)
DDblo Canyon has eliminated the loop isolation valves ere which w
. included in the Zion reactor coolant system design; (4)
'De design core inlet reactor coolant temperatures for o Diabl Cusyon thics 1 and 2 are 544.4 and 545'F, respectively, as coe pred te> the Zion Unit 2 and D. C. Cook values of 530.2 and 536 3*L . ,
retpectiveIy.
- 4. 2 RechanienL Detsin ,
.i.2.1 Fuel The JMoh3"tidyotr fuel assembly consists of 264 fueled s, 24 guide rod thimble, and 1 instrumentation thimble plus ancillary hardware rar.ged ar in a Ikl7 array.
The instrumentation thimble is at the center of the
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4-2 .
l -Y assembly and facilitates the insertion of neutron detectors. The guide thimbles provide channels for inserting various reactivity controls. The.
fuel rods contain enriched uranium dioxide hermetically clad in ;
Zircaloy-4. The assembly is supported at both ends by stainless steel l 1
nozzles. Alignment and transverse spacings are maintained by 8 spacer grido equally spaced along the axis of the fuel assembly.
The Diablo Canyon fuel assembly (17x17) is mechanically similar to ,
the 15x15 Westinghouse assemblies used previously in the D. C. Cook and >
Zion reactors. Those oechanical aspects which are different are ex- {
1 hibited in Table 4.1 where the comparison is made with the D. C. Cook reactors. The dif ferences are essentially geometric, resulting in a [
lower linear power density and other increased safety margins, as dis- -
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-'i cussed later.
The evaluation of the Westinghouse fuel mechanical design is based upon mechanical tests, in-reactor operating experience and engineering ;
analyses. Additionally, the in-reactor performance of the design will be e subject to the continuing surveillance programs to be conducted by q Westinghouse and the individual utilities. These programs continually provide confirmatory and current design performance information.
4 In our evaluation of the fuel thermal performance we assumt- that .
j i
densification of uranium dioxide fuel pellets may occur during irradia-tion in power reactors. The initial density of the fuel pellets, and }l the size, shape and distribution of pores within the fuel pellet in-l fluence the densification phenomenon. The effects of densification +
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on the iuel rod will increase the stored energy, increase the linear thermal output, increase the probability for local power spikss, and i decrease the thermal conductance. The primary effects of densification 5 l on the fuel rod mechanical design are manifested in calculations of -
,i time-to-collapse of the cladding and fuel-cladding gap conductance. l Ii Time-to-collapse calculations predict the time required for unsup- I ported cladding to become dimensionally unstable and to flatten into ll !
ar. axial gap caused by fuel pellet densification. Gap conductance calculations predict the decrease in thermal conductance due to opening '
of the fuel-clad radial gap.
The engineering methods used by Westinghouse to analyze the fuel '
+
thermal performance have been previously submitted in the Westing- }
house Topical Report WCAP-8218, " Fuel Densification, Experimental .
Results and Model for Reactor Application," dated October 1973, and have been reviewed by the Regulatory staff. The results of our review .
were reported in " Technical Report on Densification of Westinghouse }
PWR Fuel," issued by the Commission on May 4,1974. On the basis of -
)
our review we have concluded thct the applicant has considered the effects of densification on the Diablo Canyon fuel assemblies in a - t manner which adequately describes the fuel behavior.
All fuel rods will be internally prepressurized with helium during final welding to minimize cladding compressive stresses during service. _
The level of prepressurization is designed to preclude any cladding --
- tensile stresses throughout operation due to total internal pressure.
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4-6 Substantially all of CN n-reactor operating experience with Westing-hou:e fuel rods an/ F-Z am +vtlies is applicable to the Diabic Canyon fuel since the 177 "" ' simbly is a slight mechanical extrapolation f rom the 15x15 aJWy. The range in design parameters for which this in-reactor expertenec is applicable is tabulated in Table 4.2. The assemblies referrst to in this Table have been irradiated up to 6 years and have had pejk asposures of 30,000 megawatt-days / metric ton (MWD /MT),
totaling more ;ths 70 million megawatt hours of pewer generation.
During thiis: puer reactor service a small fraction of the fuel rods have experiencedWacts. There has been no instance where cladding aetects hast threarasd either the plant or the public safety. Cladding, defects were; r;ae,d by either excessive emnufacturing f Nurities, ex-cessive coolhhtanse-flow velocities and/or fuel pellet densification.
These causestWaheen amended by modifications to both the manufacturing procedures a6dtst plant coolant system. The fuel related modifications required readjsm:4a of the magnitude of a design characteristic rather than a'Wesign of the fuel assembly. Fuel rods and assemblies identical to 'tMiaM^ Canyon design have not yet experienced power reactor service. H o.n tt, the current use of similar fuel roce and assed;ies has yielkhoperating experience that provides confidence in the acceptablaerformance of the 17x17 fuel rods and assemblies.
Out-of-reas nechanical tests have been performed on typical 15x15 fuct asdlies. These tests demonstrate acceptable mechanical performance ote,15x15 fuel assembly. Since the 17x17 assembly is 4
gf '
4-5 a slight mechanical extrapolation from the 15 x 15 assembly, we expect '
the mechanical behavior of the two assemblies r.o be similar; therefore, we have concluded that the 17 x 17 fuel assembly is acceptable.
Verification tests on the 17 x 17 assemblies have been completed and reported in WCAP-8278, "flydraulic Flow Test of the 17 x 17 Fuel Assembly;" and WCAP-8236 " Safety Analysis of the 17 x 17 Fuel Assembly f or Combined Seismic and Loss-of-Coolant Accident." The staff has reviewed these two topical reports and concluded that t ey are acceptable. ~)
The grid tests were made on assemblier with 7 grids; Westinghouse will ,
document in a topical report the justification for applying the test ,
r results to 8 grid assemblies. Also, the first phase of the single rod burst tests has been completed and will also be documented. We will review i the documentation and will report the results of our evaluation in a
- l supplement to this Safety Evaluation Report prior to a decision concerning the issuance of operating licenses for Diablo Canyon Units 1 and 2. .l l
Performance of the fuel during operation will be indirectly monitored >1 by measurement of the activity of both the primary and the secondary Il 1 l coolant for compliance with technical specification limits. The first ;
available irradiated 17 x 17 fuel assemblies and rods will undergo an extended surveillance program following each cycle of operation. Onsite - ,
examinations will include fuel rod integrity, fuel rod and fuel assembly a
dimensions and alignment, and surface deposits. Details of the surveillance i
program will be reported on in a supplement to this Safety Evaluation Report.
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In addition, Region III of the Diablo Canyon core will have one removable fuel rod assembly conceptually similar to removable rod ast,emblies :
used in Zion-1, Surry-3, Point Beach, San Onof re and others. Eighty-eight 13 (88) fuel rods will be removable to facilitate interim and end of life i
l l (E0L) fuel evaluation as a function of exposure. The removable fuel rods ; ) '
i l are identical to the other fuel rods in the core and will provide direct R inspection results on fuel which performed under the combined thermal, 1
hydraulic and nuclear conditions. expected during normal operation. The ;l !
1: I technical specification requeements will result in surveillance of *l t j these 17 x 17 irradiated fuel rods which have been precharacterised. ;j We have concluded, subject to confirmation of the previously cited ;c required documentation, that based on (1) operating experience with ;}
t similar fuel, (2) the results of out-of-reactor tests on an assembly of !,
- ! I similar design, (3) the increased thermal margins which the 17 x 17 fuel lg l 1 l provides. (4) the technical specification requirements to monitor and j l l
limit off-gas and ef fluent activity, and ($) the existence of a continuing fuel rod surveillance program which includes destructive and non-destructive post irradiation examinations, the cladding integrity of j the 17 x 17 fuel will be maintained and significant amounts of radio- ;'
activity will not be released. Furthermore, on the basis of our review
^
i of the Westinghouse Topical Report WCAP-8236, we have also concluded that f neither accidents or earthquake induced loads will result in either an inability to cool the fuel or interfer:.nce with control rod insertion. e
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4-7 4.2.2 Reactor Vessel Internals the design and vibrational test programs for the reactor vessel internals are discussed in Section 3.9.1 of this report.
We have reviewed the selection of materials for the reactor vessel internals required for distributing the coolant flow to achieve acceptable heat transfer perf ormance for all modes of reactor operation.
The major containment and support member of the reactor internals is the lower core support structure. For the Diablo Canyon Plant, the full thermal shield in Unit I was replaced on Unit 2 by a neutron panel assembly.
The principal material for the reactor internals is Type 304 stainless steel.
Type 316 stainless steel and Incone) are specified for small parts such as bolts, dowel pins, and inserts. Type 403 stainless steel is 8Pecified for parts requiring a yield strength above 90,000 psi, i
All materials used are compatible with the reactor coolant, and have performed satisfactorily in similar applications. They meet the ASME Boiler and Pressure Vessel Code requirements (1973). Undue susceptibility to intergranular stress corrosion cracking will be prevented by avoiding the use of sensitized stainless steel as reconsended in Regulatory Guide !
1.44, " Control of the Use of Sensitized Stainless Steel."
The use of materials proven to be satisfactory by actual service i experience, and avoidance of sensitization by the methods recommended in Regulatory Guide 1.44 will provide reasonable assurance inat the !
reactor vessel internals will not be susceptibic to failure by corrosion l
or stress corrosion cracking 1
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. I 4-8 f 4.3 Nuclear Design .
I 4.3.1 Ceneral j '
Diablo Canyon Unita 1 and 2 are similar to several Westinghouse designed 4 loop reactors that will use the 17 x 17 fuel design and that are currently being reviewed for operating licenses. The staff has recently completed a 1 l
l generic review of the Westinghouse 17 x 17 fuel design (described in 1' 1
WCAP-8185. " Reference Core Report 17 x 17"), and has concluded that the l t
nuclear design is acceptable. We have also concluded that the nuclear }
design applies directly to the Diablo Canyon reactors, and is therefore <
acceptable. The nuclear characteristics of the 17 x 17 fuel assemblies are j) essentially the same as those of the previous 15 x 15 Westinghouse design. h
)
As a result, there are no changes in control regi.irements, control rod .
patterns and reactivity worths, and zenon stability. The analytical k i
methods employed in the design of this core are the same as those used sor j V
Westinghouse reactors in recent years, and are acceptable. 3 Ii The staff has performed certain independent analyses to support its ;i conclusions of acceptability. The results of these analyses are-discussed 1,
in the following sections. s
]2 4.3.2 Power Distribution {!
b4 The ptimary feature of the 17 x 17 fuel design reinting to physics and f' h
power distribution monitoring considerations is the increase in the number 3 1
of linear feet of fuel provided. For a given core power rating and peaking , {
4 f actor (F ), this leads to a reduction in the averaga rnd peak linear f 1 {
f i =
1 7 i ;
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t i
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3 4-9 power densities (LPD) . Thus, the average LPD for Diablo Canyon Unit 2 m de.:reased f rom 7.03 kW/f t to 5.44 kW/f t with the change f rom 15 x 15 to 17 x 17 fuel. At 102 percent of full power, the peak LPD associated with an F of 2.32 is now 12.9 kW/ft; by way of comparison, the LPD which would produce a 2200*F clad temperature in a IDCA analysis performed ,<
in accordance with the Final Acceptance Criteria is 15.7 kW/f t. The l Unit 1 average LPD is 5.33 kW/f t which is corresponding 1/ more conservative.
The applicant has proposed to take credit for correct normal operator action in determining the peaking f actor used to define initial conditions for accident analyses. The information presented indi:stes Fq would to limited to 2.32 (including fuel densit Acation power spikes), whereas with the previously used axial offset relation, the F limit would be about 2.5. The applicant's plan would eliminate most of the xenon transient effects on F that otherwise would occur in load following.
q
. 3.3 Reactivity Coefficients The staff has made an independent comparfson of the beginning of life (BGL) moderator and Doppler reactivity coefficients for the 15 x 15 and 17 x 17 'uel assemblics using methods equivalent to those employed by f the applieu.t. The calculated BOL isothermal moderator temperature f 2
coefficient in the operating temperature range is 0.1 to 0.2 x 10 -4 /*F 1
a more negative with th< 17 x 17 fuel assembly than with the 15 x 15 design. l The moderator tempt catu e r C ' tient varies irom about 0 to -3.5 x 10 /*F over the first cycle. De calco ated BOL it thermal Doppler coefficient in 3
the operating temperature rang is approximately 2 percent more negative %
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with the 17 x 17 fuel assembly that with the 15 x 15 design. These effects are due to the slightly hignea resonance absorption in U-238 in the 17 x 17 fuel assembly lattice.
The reactivity coefficients of cores using the 15 x 15 fuel assembly have been determined in recent reactor startup test pr ograms, e.g., Surry 1
Units 1 and 2, and compare favorably with predictions.
Moderator temperature coefficients as a function of temperature and soluble boron concentration 6bnwn in Figure 4.3-30 of' the FSAR for the
,17 x 17 fuel assembly core are almost identical to those previously reported for the 15 x 15 design. The Doppler coefficient for the 17 x 17 assembly core in Figure 4.}-27.of the FSAR contains the x-y spatial power shape weighting and is not directly comparable to.the pointwise Doppler coefficient reported for the 15 x 15 fuel assembly (the pointwise value agrees well with the staf f's calculation). However, the x-y power shape weighting factor appears reasonable, and this form of the data is appropriate for use in the study cf axial power shapes using the PANDA code. We conclude that the reactivity coef ficients are acceptable for use in control and safety analyses related to the 17 x 17 l fuel assembly design.
4.4 Thermal and Hydraulic Design .
The reactors for Diablo Canyon Units 1 and 2 are designed to operate at core power levels of 3338 and 3411 MWt, respectively, which correspond to net electrical outputs of 1084 and 1106 MWe. The thermal and hydraulic p
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4-11 '
design has been evaluated on the basis of 3411 Mk't.
Although the'Diablo J Canyon Plant will utilize a 17 x 17 fuel assembly, whereas the original Diablo Canyon submittal incorporated . x 15 fuel assembly,-the following . ,
thermal and hydraulic parameters have remained unchanged: ,
Unit 1 Unit 2 (1) Core power (1) Core power I (2) System pressure .(2) Vessel loop flow rate (3) Coolant inlet (3) System pressure temperature (4) Open lattice fuel (4) Coolant inlet temperature rod array .
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(5) Core and vessel average and '
exit coolaat temperatures I
(6) Open lattice f uel rod array.
The basic modification to the Diablo Canyon core desig: is an increase in the number of fuel rods and the resulting reduction in peak and average .
linear heat ge eration rates and heat flux.
The principal criterion for the thermal and hydraulic design of a j i
reactor is to prevent fuel rod dacege by providing adequate heat transfer )
J for the. various core heat generation patterns occurring during normal l operation, operational transients, and transient conditions resulting f rom f aults of moderate frequency. The fuel damage littts and thersc1- !
hydraulic criteria used to evaluate the fuel perfor:ance of the Diablo t i
Canyon reactors are the same for the 17 x 17 design as for the previously g s
l l
4-12 l 1
These damage ifnits f or normal operation, proposed 15 x 15 design.
conditions arising f rom faults operational transients, and any t ransient of moderate f requency are (1) departure f rom nuclear bsfling will not 1
)
occur on at least 95 percent of the limiting f uel rods at a 95 percent l
confidence icvel (95/95 criterion); (2) the maximum f uel temperature aball be less than the melting temperature of 00 3 ; (3) at'least 95.5 percent of the thermal flow will pass through the fuel rod region of the core; and (4) the permitted modes of operation shall not lead to I
1 i hydrodynamic instability. In order to show compliance with these criteria, the applicant performed L'NB and f uel temperature calculations as well as flow distribution and flow instability analyses. Some of the. ;
f important thermal and hydraulic parameters of the 17 x 17 design and I the re.;ults of the DNB calculations are presented in Table 4.3. For j
comparison purposes, corresponding information representative of the former Diablo Canyon 15 x 15 and Zion designs has also been included in Table 4.3.
The Diablo Canyon 17 x 17 design has approximately the same DNB margin It as the 15 x 15 design, along with an increased fuel temperature margin.
should be noted that the effects of fuel densification were included in l
the 17 x 17 calculations, while these effects were missing from he 15 x 15 calculat ions. Also, the Diablo Canyon 17 x 17 DNB calculations included 1 l
approximately a 14 percent margin on DNB for the following reasons:
(1) to allow incorporation of the final results of the DNB and mixing tests; i
i l
4-13 ;
(2) to allow incorporation of the final results of the hydraulic testa (D-loop ter.ts); and (3) to allow for any f abrication tolerances larger than those presently used.
The first part of the DNB tests, utilizing uniformly heated rods, was completed and reported in WCAP-b296 "Effect of 17 x 17 Fuel Assembly Geometry on DNB." The results indicate the following: (1) the previously used DNB correlation (W-3 correlation with a modified space factor) must i be multiplied by 0.88 in order to show agreement with the 17 x 17 data; (2) the use of a thermal diffusion e.oefficient (TDC) of 0.038 is conservative; and (3) a 11NBR value of 1.275 corresponds to etw 95/95 criterion. Since only data with uniformly heated rods were considered, it is uncertain at the present time whethea further adjustments in the correlation or in the DNBR corresponding to the 95/95 criterion are needed to cover the expected range in axial power shapes. Additional DNB tests with non-uniform axial heating are planned for December, 1974.
The results of these tests, together with those reported in WCAP-8296, must be used to set technf.;al specification lirits for the Diablo Canyon 17 x 17 fuel assembly assign.
Although Westinghouse does not expect further changes in the DNB correintion or lei the statistical evaluation of this correlation, based on our review of critical heat flux correlations, considering both uniform and non-uniform axial heat flux data, we censider that changes are possible. If the results of the non-uniform DNB tests are not available 1
4-14 when the technical specifications for Diablo Canyon are finalized, we -
will tequire that the minimum allowable DNBR be increased 5 percent above that required to satisfy the 95/95 criterion.
The Westinghouse Topical Report WCAP-8185 itemized the presently The sources available DNB margin in the reference 17 x 17 fuel design.
and amounts of these margins for a four loop plant are as follows:
Source DNBR_ Martin (%)_
DNB calculations used a multiplier of 0.86 while data justify a multi-2 plier of 0.68.
A DNBR of 1.3 was used in lieu of the '
95/95 criterion. Data justifies a 2
. DNBR of 1.275.
A TDC value of 0.051 was used in the data reduction while a value of 0.038 1.4 was applied in the analysis.
DNB tests were pet formed with 26 in, grid spacing while the design utilised 5
20.5 in. spacing.
l Thus, the Diablo Canyon design of fers a total UntA' margin of approximately l We find l
10 percent beyond the requirements of the Westi.nghouse critaria. l j
this margin to be suf ficient to cover uncertainties due to the present l
state of incomplete rceults of the 17 x 17 tests and unavailability of as-built tolerances. We therefore have concluded that the thermal ble for and hydraulic performance of the Diablo Canyon design is accepta the design conditions shown in Table 4.3. ;
.a i
4-15 i
l The reactors for Diablo Canyon Units 1 and 2 were designed to opetate I
at a higher heat output and a higher inlet coolant temperature than Zion Unit 2 (see Table 4.3). This performance increase was partially based on the use of the THINC Code which permitted a more detailed analysis of the thermal and hydraulic characteristics of the core. The j 1
THINC co!.e was developed to consider crossflow between adjacent assen:blies )
in the core as well as thermal diffusion between adjacent subchannels in the assembly. The effect of changes in local power distribution on flow redistribution is also considered. As a result of these considerations, 1
the THINC code permits the computation of more realistic power shapes i than those that had been available f rom previously used computer codes.
These power shapes are especially important at the design overpower conditions.
Certain elements of the THINC verification test program have been performed at the Zion facility this year. W will review the results of these tests and analyses as they become available. In the event that sufficient verification cannot be obtained from the combined test and analytical programs. estri.tions will be imposed on the operation of the Diablo Canyon reactors. Changes to the technical specifications will be made to maintain required astgins to fuel rod damage during normal operation, as well as during anticipated transients.
Another paramet er that influences the thermal-hydraulic design of the core is rod-to-rod bowing with fuel assemblies. Experimental data
4-16 1
on the extent of bowing in the 17 x 17 design are not yet available. l l
The 17 x 17 futi performance surveillance program should provide this e
information. In the meantime, the design of the core is based on predicted values of bowing deriwd from measurements made on incore 15 x 15 fuel assemblies. Westinghouse recently submitted two topical ,
reports, WCAF-8346, "An Evaluation of Fuel Rod Bowing." and WCAP-8176, "Ef fect of a bowed Rod on DNB," that describe the analytical techniques used to predict bowing and the methods used for assessing the ef feet of bowing on thermal performance. We are presently reviewing these l
Westinghouse Topical Reports and will report the results of our evalua- !
l tion in a supplement to this Safety Evaluation Report. I We have concluded, subject to satisfactory resolution of the matters discussed above, that the thermal and hydraulic design of the Diablo 1
Canyon reactors is acceptable, and that these res.ctors can operate at core power levels of 3328 and 3411 Wt for Units 1 and 2, respectively. Resolution of the outstanding items disussed in this section will be reported im in a supplement to this Safety Evaluation Report prior to a decision concerning the issuance of operating licenses for Diablo Canyon Units 1 and 2.
l O
of
4-17 Table 4.1 Fuel Mechanical Desian r~sarison l I.
Design Parameter Westinghouse Westinghouse Diablo Canyon Plant ._D. C. Cook Plant FUEL ASSEMBLY Rod Array 17x17 15x15 Number of Fueled !
Rods 264 204 Number of Spacer Crida 8 Number of Guide 7 Thimbles 24 Inter-rod Pitch 20 Average Theras1 .496 in .563 in -
Output (4 locp) 5.4 kW/ft. 7.0 kW/ft FUEL PELLETS Density (theoretical)
~
95% 94%
Fuel Weight / Unit Length
.364 lbs/ft .462 lbs/ft I
\
Outside Radiun Thickness .187 in .211 in Radius / Thickness Ratio .0225 in .0243 in 8.31 8.68 I
l O
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4-18 Table 4.2 kange Of Design Pcrameter Experience P_ARAMETER RANCE ON POWER REACTOR EXPERIENCE Fuel Rod Array 14x14, 15x15 Rod Assembly 179 to 204 !
Cuide Thimbles / Assembly 16 to 20 Assembly Envelope 7.76 in to 8.43 in Inter-rod pitch .556 in to .563 in Plenum length 3.27 in to 6.69 in Prepressurization 14.7 psia to 34C0 psia Diametral gap .00b5 in to .0075 in Spacer Crids/ Assembly 7 to 9 i
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A N A N H P (
5-1
.i 5.0 REACTOR COOLANT SYSTEM 5.1 Summary Description The reactor coolant system (RCS) for the Disblo Canyon Units consists !
of four coolant loops connected to the reactor pressure vessel Each loop contains a reactor coo 3 ant pump, steam generator and associated piping and valves. Reacto' :oolant system pressure will be maintained by a pressurizer connected to one of the coolant loop hot legs. All of these components are located within the containment building.
5.2 Intearity of Reactor Coolant Pressure Boundary 5.2.1 Desian of Resetor Coolant Pressure Boundary Cm-snents i
We have reviewed the pressure-retaining components within the reactor coolant pressure boundary (RCPB)'. In accordance with Section 50.55a of 10 CFR Part 50, these components have been identified and classified into three catescries: (1) ASME Section III, Code Class A; (2) ANSI B31.7, Class 1; and (3) ANSI B16.5. These components were designed and constructed in accordance with the requirements of the applicatie codes and addenda as specified by the rules of 10 CFR Part
)
50, Section 50.55a, Codes and Standards. The specified ASME Section III i code cases, whose requirements have been applied in the construction of pressure-retaining, Code Class A components within the RCPB (Quality Crcup Classification A), are acceptable.
We have concluded that compliance with the requirements of the code cases mentioned above, in conformance with the Commission's regulations, provides reasonable assurance that the resulting quality standards and
+
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i 5-2 i l
classifications are commensurate with the importance of the safety function of the reactor coolant pressure boundary. Therefore, we con-clude that the applicant's program regarding code cases applicable to the RCPB is acceptable. ,
I The design loading combinations specifIed for RCPB componer ts are comparable to the plant conditions currently identified as normal, upset, emergency or faulted. The design limits used by the applicant for these plut conditions are comparable to the criter2a recommended in Regulatory )
i Guide 1.48. Use of these criteria for the design of the RCPS components l
provides reasonable assurance that, in the event (1) an earthquake j should occur at the site, or (2) a system upset, emergency or f aulted -
transicut should occur during normal plant operation, the resulting combined stresses imposed on the system components may be expected not to exceed the allowable denign stresses and strain limits for the materia}s of construction. Limiting the stresses and strains under such loading combinations provides a basis for the design of the nystem components to withstand the most adverse loadings postulated to occur during the service lifetime without gross loss of the system's structural integrity. The load combinations and associated stress and deformation j limits considered in the design of RCPB components constitute an acceptable basis for design in satisfying the related requir oents of AEC General Design Criteria Nos. 1, 2, and 4.
I t
5-3 5.2.2 Overoressurization Protection i i
Overpressurization protection in accordance with the ASME Doller and j Pressure Vessel Code, Section 111 Article 9, 65th Edition, is provided by pressure relief of the RCS using three pressurizer safety valves mounted on the pressurizer nozzles. The pressurizer safety valves discharge to j I
the pressurizer relief tank. Each of the safety valves is rated to carry 420,000 lbs/hr., which is greater than one-third of the total rated ;
4 capacity of the system. The maxitsum pressure transient imposed on * -
I RCS results from a power imbalance caused by a turbine trip from saxism.s overpower conditiona.
l l
We find the method of overpressurization protect, ion to be acceptable.
5.2.3 General Material Considerations l We have reviewed the materials of construction for the reactor coolant pressure boundary 1sure that the possibility of serious corrosion or stress corrosion 1:- .cimized. All materials used are compatible with the expected environment, as proven by extensive testing and satisf actory ser-vice performance. The possibility of intergranular stress corrosion in austenitic stainless steel used for components of the RCPB vill be inini-mized because sensitization has been avelded and adequate precautions l have been taken to prevent contaminar ton during manufacture, shipping, storage, and construction. The measures taken to avoid sensitization l
are in general conformance with the retotenendations of Regulatory Guide l
l 1.44, and include controls on compositions, heat t reatnent s, weldic.g procesr.cs.
l l
i l
l
5-4 ar.d cooling rates.
The use of materials with satisfactory service experience, and confor: nance to the recoramendations of Regulatory Guide 1.44 provide reasonable assurance that austentic otainless steel components will be compatible with the expected service environments, and that the probability of loss of structural integrity is minimized.
further protection against corrosion probicas will be provided by control of the chemical environment. The composition of the reactor coolant vill be controlled, and the proposed maxima contaminant levels, as well as the proposed pH, hydrogen overpressure, and boric acid ton-centrations, have been shown by' tests and service experience to be ade-quate to protect against corrosion and s.:ress corresion problems.
We have reviewed the controls proposed to prevent hot cracking I
(microfissuring) of austenitic stainless steel welds. These precautions.
include control of weld metal composition and welding processes to ensure that the filler metals contain f rois 5 to 15 percent delta ferrite, as calculated from a Schaef fler constitution diagram. The interpass temper-ature of all welding methods will be limited to a maximum temperature of f
350*F. The proposed welding materials, procedures, and methods comply i
with Sections III (Subs?ction NB2432) and II vi the .'.SME Beiler and "ressure Vessel Code. The use of materials, procewes, and test methods that are in accordance with these requirements and recommendations will provide reasonabic assurance that loss of integrity of sustenitic : stainless steel welds caused by either hot cracking or sensitization during welding will not occur.
1 1
l
1 l*
5-5 ]
1 i
1
,t j 5.2.4 Fracture Toughness We have revieveo .he materials selection, roughness requirements, and the extent of materials testing proposed by the applicant to provide assur-I' a=:e that the ferritic materials used for pressure retaining components of 1 the RCPB will have adequate tought.ess under test, normal operation, and i 1
l transient conditions. The ferritic materials will meet the test require- 'l J
nents and acceptance standards of the ASME Code,Section III, Subsection NB2300 of the Summer 1972 Addenda. These materials also meet the require-i ments of Appendix G of 10 CFR Part 50. The fracture toughness tests and i
procedures required by Section III of the ASME Code as augmented by Appen- 1 l
dix G of 10 CFR Part 50 for the reactor vessel provide reasonable assurance j that adequate safety margins sgainst the possibility of nond,ctile behavior l i
or rapidly propagating f racture can be established for all pressure retain- I ing components of the RCPB.
j The reactor will be operated in a manner to minimize the possibility l 1
of rapidly propagating f ailure, in accordance with Appendix G to Section i
III of th.* ASME Code, Summer 1972 Addenda, and Appendix G of 10 CFR Part l
- 50. Additional conservatism in the pressure-temperature limits used for heatup, cooldown, testing, and core operation will be provided. These will be determined assuming that the beltline region of the reactor vessel has already been irradiated. The use of Appendix G of the ASME Code as a guide for establishing safe operating limitations, and the use of results of frar.ture toughness te :ts performed in accordance with the code and AEC
! regulations, will ensure adequate safety margins during operating, testing, l
l l l s
s!
5-6 s)
O maintenance, and postulated accident conditions. Compliance wit h chesc code provisions and AEC regulations constitutes an acceptable basis for 'i' I
- 9 -*
satisfying the requirements of AEC General Design Criterion No. 31. y The toughness properties of the reactor vessel beltline caterial ,'
r.'
will be nonitored throughout the service life with a material surveil-r lance program that vill meet the requirements of Appendix H of 10 CFR 2 Part 50 (July 17,1973). The specimen orientation and number of speci- i i
mens per capsule conform to ASTM E 185-70 for Unit 1; this standard was in ef fect when the vessel was manufactured. For Unit 2, the specimen ,
orientation, number, selection procedure, and removal schedule conform to ASTM F 185-73. Changes in the fracture toughness of material in the i reactor vessel beltline caused by exposure to neutron irradiation will be assessed properly, and adequate safety nargins against the possibility of vessel failure can be provided if the material surveillance requirements af ASTM E 185-70, ASTM E 185-73, and Appendix il to 10 CFR Pm 50 are met.
I Compliance with these documents will ensure that the survelliance program constitutes an acce', tabic basis for nonitoring irradiation induced changes in the fracture toughness of the reactor vessel material, and will satisfy the requirements of AEC General Design Crit erion No. 31.
5.2.5 .Puml J11whe"1 The probability of a loss of pung flywheel integrity has been mini- j mized by the use of suitable materi::!, adequate design, and inservice inspect ion. The applic ant has stated that the integrity of the reactor !
i i
coolant pump flywheel it. assured by compliance with Regulatory Guide 1.JP, ,
3
<a. ,,
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)
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1 i
5-7 j l
l
" Reactor Coolant Pump Flywheel Integrity." The use of suitable caterial, 1 and adequate desigt. and intervice inspection for the flywheels of reactor l
coolant pump motors, as specified in the FSAR, provi?.e reasonable assurance s (1) that the structural integrity of flysheels is adequate to withstand the l q
forces imposed in the event af pump design overspeed transient without j
loss of function, and (2) that their integrity will be verified periodically !
in service to assure that the required level of soundness of the flywheel material is adequate to preclude failure. Compliance with the recommenda- 1 tions of AEC Regulatory Guide 1.14 constitutes an acceptable basis for i
I satisfyirg the requirements of AEC General Design Criterion No. 4. f j
5.2.6 Pun.p oversfeed 1 l
l The Regulatory staff is investigating, on a generic basis, the consequences of an unlikely rupture of a reactor coolant pipe which in I l
certain locations might result in reactor coolant punp overspeed. If this study indicates that additional protective measures are warranted under specific circumstances in order to prevent significant pump overspeed or to limit the potential effects on safety related equip-ment, the staf f will review the circumstances applicable to the Diablo Canyon Units to determine what modifications, if any, are required to assure that an acceptable level of safety is maintained.
i.2.7 RCPB 1,cakage Detection System Adequate provisions have been made to detect leakage of reactor coolant to the containment. The major compenents of the system are: containment atmosphere particulate radioactivity nonitors, radiogas monitors, 1cvel I
-_____._____.__w
Uh !
i 5-8
,: l 2
t r
indicators on the containment sumps, and a water temperature monitor on the containment air recirculation and cooling unit. The system has suffi- !
cie-t sensitivity to measure small leaks, will identify the leakage source l
)
to the extent practicable, will include suitable control alarms and read-outs, and conforms with the f unctional requirements recommended in Regula- 4 l
tory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems." In addition, iadirect indications of leakage can be obtained f rom the containment humidity, press , and temperature indicato-s. !
Significant intersystem leakage will be indicated by abnormal readings from the radioactivity monitors used to detect failed fuel, and indirectly, j
by the coolant flow and level measuring instrumentation provided for nornal i operational control of the system. i The leakage detection r ystems have detection capabilities which conform with those recommended in Regulatory Guide 1.45, and provide reasonable assurance that any structural degradation resulting in leakage auring service will be detected in time to permit corrective actions. This constitutes an acceptable basis for satisfying the requirements of AEC Ceneral Design Criterion No. 30.
5.2.8 Inservice Inspe_ction Program The program for the inservice inspection of the reactor coolant system will be conducted, to the extent practical, in compliance with 1971 Edition, the ASME Boiler and Pressure Vessel Code, Section 7.1, including the Winter 1972 Addenda. The program will also conform to Regulatory Guide 1.51, " Inservice Inspcction of ASME Code Class 2 and 1
1 a_____.
i 5-9 1
i i
3 Nuclear Power Plant Components," for the inspection of Class 2 systems. )
Remote access methods and data acquisition, methods have been developed to facilitate the inspection of those areas of the reactor vessel not :
1 1
readily accessible to inspection personnel. The conduct of periodic j i
inspections and hydrostatic testing of selected welds and weld heat- I J
af fected zones of the pressure retaining components in the RCPB in accordance with the requirements of Section XI of the ASME Code provides reasonable assurance that evidence of structural degradation or loss of leaktight-integrity occurring during service will be detected in time to permit corrective action before the safety function of a componeet is compromised. Compliance with the inservice inspections required by this Code constitutes an acceptable basis for satisfying the requirements of AEC General Design Criterion No. 22. I l
5.3 teactor Vessel Integrity l
We have reviewed all factors contributing to the structural integrity I of the reactor vessels and conclude that there are no special considerations j that make it necessary to consider potential vessel failure for Units 1 l and 2 of the Diablo Canyon Plant. The bases for our conclusion are that desiga, material, fabrication, inspection and q':ality assurance require-l nents conform to the rules of the ASME Code, Fection III,1965 Edition, j including the Summer 1966 Addenda for Unit 1 and the 1968 Edition for l
1 Unit 2. L l
Although the applicable code cases were not listed in the FSM(, the j applicant has stated that an effort was maintained during construction l i
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1 i
_ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ . . . _ _ _ . _ _ _ . b
5-10 1
to continually update and apply AEC Quality Standards to the requirements j
of the vessels. We find this to be acceptable. l The stringent fracture toughness requirements of the ASME Code, Section 111, 1971 Edition, and the 1972 Summer Addenda will be met.
Also, operating limit ations on temperature and pressure will be estab-11shed for these plant s in accordance with Appendix C (' rtetection Againsc j l
Non-Ductile Failure") of the 1972 Summer Adder.da of Section III of the ASME Code.
l 1 The integrity of the reactor vessels is assured because the vessels:
l
- 1. Were designed and f abricated to the high standards of quality required by the ASME Boiler and Pressure Vessel Code and pertinent code cases.
- 2. Were made f rcm materials of controlled and demonstrated high quality.
- 3. Were inspected and tested to provide substantial assurance that the vessels will not fail because of material or fabrication deficiencies.
- 4. Will be operated under conditions and procedures and with prote:tive-devices that provide assurance that the reacter vessel design condi- i tions will not be exceeded during normal reactor operation or during i
most upsets in operation, and that the vesrels will not fail under the conditions of any of the postulated accidents.
l
- 5. Will be subjected to monitoring and periodic inspection to demonstrate that the high initial qualit.y of the reactor vessels has not deteriorated significantly under the service ce ditions.
- 6. May be annealed to restore the material toughness properties if this becomes necessary.
5-11 1
5.4 Loose Parts Monitor Occasionally, miscellaneous items such as nuts and bolts have become )
loose parts within reactor coolant systems.
In addition to causing oper-ational inconvenience, such loose parts can damage other components within the system or be an indication of undue wear or vibration. For such reasons, 1
the staff has encouraged applicants over the past several years to support pregrams designed to develop effective, on-line loose parts monitoring.
For the past few years we have required each applicant for an operating license of a PWR plant to initiate a program, or to participate in an on-I going program, the objective of which is the development of a functional, loose parts monitoting system within a reasonable period of time.
Recently, prototype loose parts monitoring systems have been developed !
and are presently in operation or being installed at several p<1 ants. The applicant has selected an appropriate available monitoring system, and has made a commitment to install this syctes prior to plant operation .
This commitment to provide a loose parts monitor is acceptable to the l staff.
5.5 Residual Heat Removal System :
1 The residual heat removal system (RHRS) is designed to remove decay heat and sennible heat from the RCS and core during the latter staBes of cooldown.
The aystem also maintains the reactor coolant temperature during refueling, provides the means for filling and draining the refueling cavity, and as a secondary function, serves as part of the ECCS during the injec-tion and recirculation phases of a LOCA.
The RHRS consists of two heat 1
e 65
5-13 exchangers, two pumps, and associated piping, valves, and instru-mentation required for operational control.
The EHRS is placed into operation approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after initation of plant shutdown when the temperature and pressure of the RCS are below 350*F and 425 psig, respectively. Assuming operation of the two pumps and two heat exchangers, and that each heat exchanger is supplied with cogon-ent cooling water at design flow and temperature, the RHRS is designed to reduce the RCS temperature f rom 350*F to 140*F wi Ain 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. Since the RHRS is provided with two pumps and two heat exchangers arranged in separate independent flow paths, if one of the two pumps or one of the two heat exchangers is not operable, safe cooldown of the platt is assured, but the time required for cooldown is extended. (See Section 7.6 of this report for a discussion of RHRS overpressure interlocks.)
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6-2 6.2 Containment _ Systems
_ l 6.2.1 Containment Functional Design, The containment systems for each of the Diablo Canyon Units include a reactor containment structure, containment heat removal systems, con-tainment isolation systeins, and a combustible gas control system.
The containment is a steel-lined, reinforced concrete structure with net free volume of 2,630,000 cubic feet. The containment structure houses the nuclea steam supply system, including the reactor, steam generators, reactor coolant pumps and pressurizer, as well as certain components of the plant's engineered safety feature systems. The structure is d tsigned for an internal pressure of 47 peig and a temperature of 271*F.
The applicant has described in the PSAR the methods used to analyze the containment pressure response for a spectrum of design basis loss-of-coolant accidente, and the results of these analyses. Various break locations and sizes were evaluated to determine that the doubic-ended pipe rupture at the pump suction of the reactor coolant system results in the highest containment pressure. l l
The containment pressure response from postulated loss-of-coolant l l accidents was analyzed in the following manner. Mass and energy release l
rates to the containment were calculated and then used as inputs to the C0C0 computer program to calculate the containment pressure response. !
l The SATAN V computer code was used to determine the mass and energy addition rates to the containment during the blowdown phase of the accident; i.e.. the phase of the accident during which most of the 1 ,
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i energy coctained in the reactor coolant system including the prir9ry coolant, metal, and core stored energy, is released to the containment.
To obtain a conservatively high energy release rate to the containment 4 during the blowdown phas<c. the core was assumed to remain in nucleate boiling _for an extended period of time so that the energy release rate from the core would be maximized. Under this assumptien, more heat a l
l is transferred from the core to the containment than would be predicted by a calculation suitable for core heatup and an emergency co;e cooling performance evaluation. This additional energy release from the core increases the calculated containment pressure and therefore assures a i
margin of conservatism in the analysis. The SATAN V computer code has been accepted by the staff for calculating energy released during a LOCA.
During the core reflood phase of the accident, when the core is again filled with water, mass and energy release rates were calculated using a hydraulic model and an energy balance model. The hydraulic mod?1 determines the core flooding rate and the entrainment fraction. The energy balance model calculates the core e:,it conditions and the energy addition from the steam generator. The analysis of the reflood phase of the accident is important with regard to pipe ruptures of the reactor coolant system cold legs, since the steam and entrained. liquid carried cut of the core for these break locations pass through the steam generators which constitute an additional energy source. The steam and
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I entrained water leaving the core and passing through the steam generators will be evaperated and/or superheated to the temperature of the steam )
generator secondary fluid. 1 l
Results of the FLECHT experiments indicate that the carryout fraction l of fluid leaving the core during reflood is about 80% of the incoming I j
flow to the core, and continues until the fuel is recovered with water i f
to about the B-foot elevation, at which time the fuel clad temperature 1
l transient ceases (quenching occurs). The applicant has conservatively wc assumad quenching of the core at the 10-foot elevation for tne l
ment pressure calculations.
The rate of energy release to the containment during the reflood The rupture of phase is proportional to the flow rate into the core.
the cold leg at the pump suction results in the highest mass flow through f We have compared the f the core, and thus through the steam generators. I l mass and energy releases to the containment during the reflood phase of l thr. accident, as calculated with our FLOOD computer code, with those l
l values predicted by the applicant. The results of this comparison indicate l
equivalent predictions of energy release. Therefore, we have accepted the l applicant's computer models as a method of computing core reflood for this plant.
The applicant has included consideration of a possible additional energy release to the containment during the post-reflood phase of the i
large break accident. This phase begins after ttse core has been conr-pletely covered with water. During this phase, dacay heat generation l
6-5 will produce boiling in the core, and a' two-phase mixture of stean and water will exist. 'In calculaticcs performed by the FROTil code, this two-phase mixtura was assumed to rise above the core and enter the steam generators.
By this precess the remainder of the a'ailable d
steam generator energy was removed by boiling of the water entrained ,
in the two-phase mixture and carried into the containment as steam.
In calculating the rate of energy removeJ from the steam generators.
the maximum steam flow based on the, hydraulic resistaace of'the sys tem was used. A portion of the stean that flows through the unbroken loops through the ECCS injection points is assumed to be quenched before exiting to the containment. In the post-reflood period the ECCS system was conservatively assumed to be functioning at one half 1
{
capacity; this assumption minimizes the condensing effects of quenching 3
and therefore increases the energy released to the containment. We therefore conclude that the applicant has calculated the energy release 1 from the steam generator in a conservative manner.
1 The COCO code was used to calculate the containment pressure from the mass and energy data discussed above. The peak containment pressure of 43.8 psig was calculated by the applicant assumirg maximum operation of the safety injection system and a single-active failure cf one spray pump.
We have calculated a peak containment pressure of 44.6 psig using the CONTDiPT code. We therefore conclude that the mariem c containment pressure is conservatively calculat:ed to be belev the design pressure of 47 psig.
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The applicant has also analyzed the containment pressure respunst.
due to a postulated f ailure of a main steam line within containment.
The maximum calculated containment pressure is about 42 psig which is below the design value for the containment.
The applicant used the TMD (Transient Mass Distribution) code with augmented flow correlation to analyze the transient pressure response I of the containment interior compartments. The IMD code was developed by i
Westinghouse to analyze the short-term pressure response in the containment I subcompart ments. TMD calculates the critical flow of a two-component, <
two-phase fluid (air, stea, and water), assuming a thermal equilibrium condition. However, a correction factor, which was determined by West-inghouse frcm data obtained f rom sma'.1 scale experiments, is then app ad to the calculated critical flow. The correction factor as used in the code increases the critical flow up to 20% through the compartments as the quality of the fluid decreases. Westinghouse refers to this in-1 e creased critical flow as " augmented" flow. 1he net effect results in a l lower subcompartment dif ferential pressure when compared to the pressure calculated without the augmented flow correlation. The use of the augmented flow f actor is therefore less conservative than the use of the thermal equilibrium assumption in calculating subcampartment pressure transients.
5 During the course of our review of the FSAR, we informed the applicant that we would base our evaluation of the containment subcompartment pres-sure response on the analysis without the augmented critical flow correla-tion, and that the subcompartment pressure response should be reanalyzed 4
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$ with t he non-augmented flow correlation. In response to our request ,
h the applicant is reanalyzing the subcompartment pressure responses con-f I sidering a non-augmented flow correlation. The applicant has provided us ;
i I with an analysis of the pressure response within the pressurizer enclosure and loop compartments using the non-augmented vent flow correlations. We i are currently reviewing the applicant's analyses of these subcompcrtments. j
[ The applicant has not completed the analysis of the reactor conlant pipe l I
l annulus, reactor ussel annulus, and lower reactor cavity. We expect to {
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receive the results of these analyses at a later time, and will report the results of our e aluat.fon in a supplement to thir Safety rvaluation }cport. l
[ Ue have reviewed the design of the components and substructures inside the containment, and subject to favorabic resolution r ubcompartment I pressure analysis item discussed previously, we find that trils design satisfit.s the requirements of AFC General Design Criteria Nos. 16 aad 50, l j and is acceptable.
l 6.2.2 Containmen_t,_Ileat Eemo_v,a1 pstems The containment heat removal systems include two redundant containment spray traine and five containment fan-cooling units.
The containment spray system serves only as an enEi neered safety l
l feature and performs no normal operating function. It is a Seismic Category I system consisting of redundant piping, valves, pumps and spray headers. All active components of the system are located out-
- side the containacnt. Missile protect Mn is provided by direct shield-r ing or physical separation of equipment. The containment sump intakes
1 6-8 i
for the spray pumps are covered by a screen assembly designed to prevent debris that could clog the spray nozzles f rom entering the sprav sy tem.
A high-high containment building pressure on two of four sensors will )
eause the ESF actuation syntes to automatically plac3 the containment l l sprays in operation. The spray pumps and valves can al'so be operated manually ft,m the control room. The spray pumps will initially take suction f rom the refueling water storage tark (RWST). Then the water in the RWST reaches a low-low level, the spray pump suction is transferred l to the containment sump to initiate the spray recirculation phase. The
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l j applicant's analysis indicates that suf ficient water will have been i delivered to the containment at that time to provide the required NPSH to the spray pumps. The system is designed to provide adequate net pop-itive suction head to the system pumps, considering the water temperatures j and containment pressure calculated to be present during the accident.
We conclude that the design meets the intent of Regulatory Guide 1.1,
" Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps."
The containment fan-cooler system consists of five equal capacity fan-cooler units. Each fan-cooler contains a set of cooling co1Js, HEPA filters and a two-speed fan. Cooling water for the units is supplled Irom the component cooling water system. During normal- opera-l t ion, four of the five fan-coolers operating at high speed are required to provide sufficient cooling. Upon receipt of a safety injection actuation signal, the idle fan-cooling unit is automatically started on
6-9 the low speed setting. Simultaneously, the running units are switched from high speed to low speed operation. The containment fan-cooling system is a Seismic Category-I system. The fan-cooling units are located outside the secondary concrete shield for missile protection, and are accessible for periodic testing and inspection during normal plant operation.
We have reviewe d the materials selection for the containment heat re-moval and ECCS systems, in conjunction with the expected chemistry of the cooling and containment spray system water. The applicant has stated that the use of sensitized stainless steel will be avoided, and that the pro-posed chemistry will not cause stress corrosion cracking of austenitic stainless steel under conditions that would be present during accident conditions.
We have concluded that the controls on material and cooling water chemistry proposed will provide reasonable asteurance that the integrity of components of these systems will not be inpaired by corrosion or stress corrosion.
We have reviewed the containment heat removal systems for conformance with AEC Ceneral Design Criteria Nos. 38, 39, and 40, and have found them to be acceptable.
2.3 Containment Air Purification and Cleanup System _s The containment air purification and cleanup system 9 consist of:
(1) the normal containment preaccess filtration system; (2) the normal containment purge system; (3) the containment fan-cooler system; and (4) the containroent spray additive system.
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6-10 The preaccess filtration system consists of two fan filter traina, each capable of processing containment atmosphere at 12,000 cfm through a filter bank consis, ting, of a pre-filter, REPA filter, and a charcoal filter to reduce airborne activity so as to permit safe and continuous access to the containment. This system is not required for post-accident operation.
The normal containment purge system supplies outside air to the containment where it is circulated and then exhausted though pre-filters It is designed for use during normal plant operation and SEPA filters.
and serves no post-accident function.
The containment fan-cooler systes is provided with HEPA filters to Undsr reduce the airborne particulate fission products following a 1.0CA.
norn.a1 conditions there is no treatment of the air within the containmen other than cooling. Upon receipt of a safety injection actuation signal the dampers on the fan-cooler units will be positioned to allow the air-steam mixture to circulate through the HEPA filters, in additfore to the l
cooling coils. The system components are Seismic Category 1 and are designed to conform to the requirements of Regulatory Guide 1.52, " Design,j I Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water Conir? Nuclear Plants."
In addition to its heat removal function, the containment spray system also is used for iodine removal from the contain:aent atmosphere following i Sodium hydroxide is added to the containment spray a postulated LOCA.
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, j solution by the spray additive system to enhance the lodine scrubbing i
function of the system. The spray additive system consists of the spray j auditive tank, eductors,, valves and connecting piping.
A sufficient gaantity of Nat H will be injected to raise the ' equilibrium pH in the containment stump to a m.nimum value of E.5. We have evaluated the containment spray and spray additive systems and found them ef fective for removal of elemental iodine, and fodine absorbed on airborne particulate satter.
The first order removal coef ficients for elenental and particulate iodine are 10 and 0.45 (hrs-I), respectively, in an estimated effective I
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3 volume of 2.16 x 106 gg . The minimum sump pH of 8.5 is considered ade-a quate to achieve and maintain a decontamination factor (DF) of 100 for the elemental iodine.
We hase reviewed the containment air purification and cleanup systens i
for conformance with AEC General Design Criteria Nos. 41, 42, 43, and 46, and have found them to be acceptable.
6.2.4 Containment Isolat ion Systems The containment isolation systems are designed to isolate the contain-ment atmosphere f rom the outside environment under accidcnt conditions.
Double barrier protection, in the form of closed systens and isolation valves, is provided so that no single valve or piping failure results in '
loss of containment integrity. Reactor building penetration piping up l to and including the external isolation valve is designed to Seismic Cater,ory I requirements, and is protected against missiles that could be generated undct accident conditions. i a
6-82 Reactor building isolation will occur automatically upon receipt of a containment isolation signal. All fluid penetrations not required for operation of the engineered safety features equipment will be isolated.
Remotely operated isolation valves will have position indication in the I
control room.
We have reviewed the containment isolation systems Dr conformance with AEC General Design Criteria Nos. 55. 56 and 57 and have f ount them I to be acceptable.
6.2.5 Combust ible Cas ,Conirol_.hste_as, p
Following a LOC \. hydrogen may accumulate inside the containment.
The major sources of hydrogen generation include: (1) a chemical reaction.
between the zirconium fuci rod cladding and water; (2) corrosion . f mater-ials of construction; and (3) radiolysis of aqueous solutions in the l
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reactor core and the containment sump. The applicant's analysis of post-LOCA hydrogen generation, which is consintent with the guidelines of Regulatory Guide 1.7. " Control of Combustible Gas Concentrations in Containment Following a Loss-Of-Coolant Accident." indicates that the hydrogen concentration in the containment would not reach the lower flammability limit of 4 volume percent (v/us until about 40 days after the postulated LOCA. We have performed a similar analysis of hydrogen I generation in the containment following a LOCA and our results are in agreement with the applicant's.
The containment f an cooler system will provide mixing of the con-tainment atmosphere following a LOCA so as to prevent possible problems L
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6-13 associated with hydrogen stratification. Tac containment hydrogen purge system, consisting of two tedundant purge supply routes, is provided to limit the hydrogen concentrations to below the guideline values given in Regulatory Guide 1.7. The system incorporates several design features that are intended to assure tl.e capability of the system to be operable l
in the unlikely event of an acciuent. These features include Seismic CateEory I design, and redundance to the extent that no single component failure disables the system. Redundant monitoring systems are provided to allow periodic sampling and acalvsis of the hydrogen concentration in the contalisment.
Based on our review of the systems provided for combustibic gas control following a post 41sted 1DCA, we con. lude tha' these systems will.
meet the recommendations of Regulatory Guide 1.7, are in conformance with AEC Cencral Design Criteria Nos. 41, 42 and 43, and are, therefore, acceptable.
6.2.6 C_ont_a1.nment,Lenkag T t estini P,roit,am o
The proposed containment design includes provisions and feature:
to pei Mt leakage testing in accordance with the requirements of Appendix ,
I J of 1. CFR Part 50. The design of containment penetrations and itolation I valves permits individual periodic Icakage rate testing at the nressure specified in Appendix J. Included are those penetrations that have resil-l 1ent scals and expansion bellows, i.e. , airlocks, emergency hatches. re-l fueling tube blind flanges, hot process line penetrations, and electrical penetrations. !
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6-14 The proposed reactor containment leakage testing program complies with the requirements of Appendix J of 10 CFR Part 50.
Such com11ance provides adequate asserance that containment leaktight integrity can be verified periodically throughout service lifetime on a timely basis to maintain such leakages within the limits of the technical specifications.
Maintaining conta innwant leakage rates within such limits provides reasonable assurance that, in the event of any radioactivity releases within the containment, the loss of the containment atmosphere through leak paths will not be in excess of acceptable limits specified for the site.
Compliance with the requirements of Appendix J constitutes an acceptable basis for satisfyinP. the requirements of AEC Ceneral Design Criteria Nos. 52, 5 3. and 54.
6.A r,an,rgency, Core, Coolinz Syst e,n ,(ECCS) t,. 3.1 Den,1 3 n tasca, ,
The Diablo Canyon ECCS has been designed to provide ercrgency core cooling during those postulated *
' dent conditions where it is assumed that mechanical f allures occur in the reactor coolant piping resulting in loss of coolant f rom the reactor vessel greater than the available coolant makeup capacity using normal operating equipment. The ECCS is also designed to provide cooling in the event of a suin steam line break.
The design bases are to prevent fuel cladding damage that would int erf ere with adequate ecw rgency core cooling and to mitigt.te the amount i
ot elad-water reaction f or any size break up te and including a double '
ended rupt ure of the largest primiry coolant line. These requirements i
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6-15 i i
are intended to be met even with minimum engineered safeguards available, j
such as the loss of one emergency power source together with the unavail-ability of offsite power.
The ECCS subsystems provided are stated to be of such number, diversity, reliability and redundancy that no single f ailure of ECCS equipment , occur-ring during a LOCA, will result in inadequate cooling of the reactor core.
Each of the ECCS subsystems are design.i to function over a specific range of reactor coolant piping system break stres, up to and including the flow area associated with a postulated double-ended break in the largest reactor coolant pipe (9.14 f t2 is the double-ended area). 1 The proposed design of the ECCS for the Diablo Canyon Plant is the amme as that previously reviewed and approved for the Zion Plant. However, on the basis of nur evaluation of the application of the single failure criterion to the functional design of the ECCS, we have identified certain locations where a single incorrta.tly positioned valve could result in total loss of the intended safety function (See Section 7.3.4 of tnis report for additional discussion of this item). A tabulation of the valves in question i can be found in item 7.12 of a request for additional information j
eent to the applicant (See Items 43 and 45 in Appendix A of this report).
Resolution of this issue can be accomplished by either: (1) including in the technical specifications provisions to remove power from the electrical system to lock certain motor-operated valves in their preferred safety positions; or (2) providing the necessary design modifications to preclude the loss of capability to perform a specified safety function . We have not yet reached agreement with the applicant on this, and will report on the n
6-It>
resolution of the problem in a supplemer:t to this Safety Evaluation Report.
t> . 3. 2 System Design The ECCS proposed for the Diablo Canyon Plant consists of accumulator tanks and high pressure injection and low pressure injection systems, with provisions for recirculation of the oorated coolant af ter the end of the injection phase. Various combinations of these systems assure core cooling for the complete range of postulattu break sizes.
Following a postulated LOCA, the ECCS will operate initially in the passive accumulator injection mode and the active high pressur . injection m>de, then in the active low pressure injection mode, and subsequently in the recirculation mode. Each of the four accumulator tanks has a total volume of 1350 cubic feet with minimum water volume of 850 cubic feet and 500 cubic feet of nitrogen gas at a normal operating pressure of 650 psig.
Each tank is connected to one of the cold legs of the reactor coolant system by a line with two check valves and a normally open motor operated isolation salve in series. A more detailed discussion of these valves is i
presented in Section 7.3.3 of this report. The high pressure injection l I
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mode of operation, upon actuation of a safety injection signal, consists I of the opere ion of two centrifugal charging pumps (rated at 150 gym each at a design head of 5800 f t) which provide high pressure injection of borte acid solution (via the boron injection tank maintained at 21,000 !
ppe boron concentration) into the reactor coolant system. Also part j of the high pressure injection mode are two safety injection pumps (rated .it I 425 gpm each at a design head of 2500 f t), which take their suction initia .ly j f rom the refueling water storage tank (350,000 gallons) with a boron !
concentration of 2000 ppia.
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low pressure injection consists of two residual heat removal pumps (rated at 3000 gpm each at a design head of 375 f t) which take their f
aut Lion f rom the refueling water storage tank. For both the high and low head pumps, the system is designed to provide adeq ate NPSH, considering the water temperatures and containment pressure calculated to be present during the accident.
When a predetermined amount of water in the refueling water storage tank has been injected, auction will be transf erred manually to the con-tainment sump for the recirculation mode of operation. Then the FCf S will provide the long-term core cooling requirements by recirculating the spilled reactor coolant collected in the containment sump back to the reactor vessel via the reactor coolant cold legs after it has been cooled in the RHR heat exchangers.
The materials selection for the ECCS is discussed in Section 6.2.2 of this report.
6.3.3 Performance Evaluation The emergency core cooling system has been designed to deliver fluid to the reactor coolant system in ordtr to control the calculated l
cladding temperature transient following a postulated pipe break, and for removing decay heat in the recirculation mode.
On January 4,1974, the Atomic Energy Commission published its decision in the rulemaking proceeding (Docket No. RM-50-1) concerning j
acceptance criteria for emergency core cooling systems for light-water '
cooled nuclear power reactors. This decision included the new amendment to 10 CFR Part 50 which incorporates the ruling. The new ruling specifies
6-18 that boiling and pressurized light-water nuclear power reactors fueled with uranium oxide pellets within cylindrical Zircaloy cladding that are licensed after December 28, 1974 shall be provided with an emergency core cooling system (ECCS) whict shall be designed such that its calculated cooling per-l f ormance following a postulat ed loss-of-coolant accident conforms to the criteria set forth in subparagraph (b) of Section 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear F wer Reactors," 10 CFR Part 50. The new criteria include the following limits:
(1) The calculated maximum f uel element cladding temperature does not exceed 2200* F.
(2) The calculated total oxidation ot the cladding does not exceed 0.17 times the total cladding thickness before oxidation. ,
(3) The calculated total amount of hydrogen generated from the chemical reaction of the claddir g with water or steam does not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding s rrounding the plenum valume, were to react. 3 (4) Calculated changes in core geometry are such that the core remains amenable to cooling.
(5) Af ter any calculated sucerssful initial operation of the ECCS, the calculated core temperature is maintained at an acceptable low value and decay heat is removed for the extended period of time required by the long-lived radioactivity renaining in the core.
In addition, paragraph 50.46 states:
ECCS :ooling performance shall be calculated in accordance with an I
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acceptable evaluation model, ind shal; be calculated for a number !
I of postulated loss-of-coolant accidents of dif ferent sizes, locations, j
and other properties sufficient to provide assurance that the entire i spectrum of postulated loss-of-coolant accidents is covere Appendix K of 10 CFR Part 50, EECS Evaluaticn ibdels , sets forth certain re-1 quired and acceptable features of evaluation models. f i
This decision on the Final Acceptance Criteria is applicable te Diablo Canyon Units 1 and 2. Accordingly, the applicant has analyzed the per- j f ormance of the emergency core cooling system in a::ordance with .he criteria q l
set forth in Sectior 50.46 and Appendix K to 10 CFR Part 50. This analysis 1
was submitted in Amendment 15 ? a the Fr3 dated August 2,1974. The l l
adequacy of the emergency core cooling system is being evaluater' by the )
l staff in light of the Final Acceptance Criteria. Our evaluatio1 of the performance of the energency core cooling system, and of the applicar.t 's 1 evaluation model, will be reported in a supplement to this Safety Ev.aluation Report.
6.3.4 Tests and Inspections, The applicant vill derestrate the operability of the ECC', by subjecting all components te preoperational tests, periodic testint,,
and inoervice testing and inspections. The preoperational cests performed f all into three categories:
(1) System actuatien tests to verify: (a) tre operability cf all ECC . valves initiated by the safety injection sicnal, the pnsse A containment isolatsen signal, arc the phase B ccntainnant iso-lation signal; (b) the operability of all safegeard pu:23 circui ty
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down through the pump breaker control circuits; and (c) the proper operation of all valve interlocks.
(2) Accumulator injection tests to check the accumulator injection lines to verify that the lines are free of obstructions and that l the accumulator check valves operate correctly. The applicant will perform a low pressure blowdown of each accumulator to con-l firm that the line is clear, and to check the operation of the i
(3) Operational testing of all the major pumps. These pumps consist l
l of the charging pumps, the residual heat removal pumps, the con- )
I tainment recirculation pumps, and the safety inject! ion pumps. I The applicant will use the results of these tests to evaluate l
l the hydraul.ic and mechanical performance of these pumps delivering through the flow paths for emergency core cooling. These pumps will operate under both miniflow (thrcugh test lines) and full flow (through the actual piping) conditions. By measuring the flow in each pipe, the applicant will make the adjustments neces-sary to assure that no one branch has an unacceptably low or high resistance. System caecks will be ma:!e to ascertain that total j line resistances are sufficient to prevent excessive runout of the pump. The applicant will be required to show that the minimum acceptable flows as determined for the FSAR analysis are met by the ceasured total pump flow and relative flow between the branch lines.
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6-21 The applicant must demonstrate that all ECCS components meet or exceed FSAR analyses values. In addition, in response to a request from the staff, the applicant has evaluated his proposed compliance with the positions stated in Kegulatory Cuide 1.79, "Preoperat ional Testing of Emergency Core Cooling Systems for P essurized Water Reactors." With the exception that the auxiliary feedwater pumps will not inject water into the steam generators, the applicant will comply with Regulatory Guide 1.79. We find this to be acceptable.
The applicant will perform routine. periodic testing of the ECCS components and all necessary support systems with the plant sta at power. Valves that are required to operate after a LOCA will be operated through a complete cycle; pumps will be operated individually in this testing on their miniflow lines except the charging pumps, which will be tested by their normal charging function.
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l Test circuits will be used to periodically check for leckage of reactor coolant back through the accumulator discharge line check valves to ascertain that these valves seat whenever the reactor coolant system pressure is above a preset value. This periodic systec testing will also include a visual inspection of pump seals, valve packings, flanged connections, and relief valves to detect leakage.
The inserv* e inspection program for the ECCS Cluid carrying components is included as part of the ASME Code Class 2 and 3 inspection program. This program is described in Section 5.2.8
_ _ _ , _ _ _ _ _ . , _ _ _ _ _ . - _ _ _ - - - - - - - - - - - " - - - - - - - - - - - - ^
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J of this report.
6.3.5 _ Conclusion i,
As stated in Sections 6.3.1 and 6.3 2 of this report , the acceptability of the emergency core cooling system is still being evaluated. Specifically)
(1) the applicant must agree to either lock-out power to certain motor-operated valves or modify the design to render locking out of power i
unnecessary; and (2) the applicant 's ECCS evaluation model and analysis I results have not been found to be acceptable. We will report our con-elusions regarding acceptability of the ECCS in a supplement to this Safety Evaluation Report. I 6.4 liabitability Syctems j The applicant proposes to meet the intent of AEC General Design a g
Criterion No. 19 by use of adequate concrete shielding and by installing a ,
2100 cfm recirculation charcoal filter (having redundant active components
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in the control room ventilation system.
In addition, the construction
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I details of the control room are such as to result in low air infiltration into the ecntrol room under isolated conditions. The charcoal filter will >
l be automegically activated upon an iccident signal, high radiation signal, or high chlorine signal. ble have calculnted the potential radiation doses ,
to control room personnel following a LOCA; the resultant doses are within /,
the guidelines of CDC No. 19.
o i The applicant has indicated that chlorine will be stored onsite for water treatment purposes. Because the effects of an accidental release of '
chlorine could have an adverse effect on control room habitability, the applicant will provide appropriate chlorine protection devices- such as
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I quick acting chlorine detectors and celf contained breathing apparatus.
q, We conclude that these protection devices are acceptable.
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7-1 7.0 INSTRUKEN1 ATION AND CONTROLS 7.1 General The instrumentation and control systems for the Diablo Canyon Nuclear Power Plant , Units 16 2, have been evaluated utilizing: (1) the Commission's General Design Criteria (CDC) as published in July 1971; (2) Institute of Electrical and Electronics Engineers (IEEE) !
i Standards, including IEEE Criteria for Protection Systems for Nuclear Power Generating Stations (IEEE Std 279-1971); and (3) applicable j l
Regulatory Guides for power reactors. These criteria have been us>id as the bases for evaluating the adequacy of these systems.
The designs of the instrument and control systems for the Diablo Canyon Plant are functionally the same as those for Commonwealth Edison's Zion Station. The evaluation of the Diablo. Canyon designs l concentrated on the equipment qualification, system 14plementation, 1
applicability of previous generic evaluations, and concerns unique to !
the Diablo Canyon Plant itself.
A major desigt change from that presented during the CP review was implemented in the actuation systems for the reactor trip and engineered safety features. The review of these systems is presented in Sections i.2 and 7.3 of this report.
The results of our review of selected logic and schematic diagrams, and the findings of our initial site visit are reflected in this Safety Evaluation Report. However, additional selected drawings, technical specifications,
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i 7-2 and several unresolved items still must be reviewed; these will be !,
reported on in a supplement to this Safety Evaluation Report.
7.2 Reactor Trip System .
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7.2.1 Genera _1 Our review covered the aspects of the protection system which 1
init lat ea, monitors. bypasses, testa 4.nd controls the reactor trip system, including field implementation, logic diagrams and detailed l sehematics.
l 7.2.2 Reactor Trip Sgtys Actuat ion _ Logic.
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l Functionally, the reactor trip system for Diablo Canyon is exsen-i tfally the same as that evaluated during the CP review. However, the l 1
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electro-mechanical relay logie has been replaced bv solid state logtr.
I The solid state protection system generates signals to open the reactor )
l trip breakers when the required ecebinat ion of int ut signals from the analog process system occurs. The design of the solid state protect ion syst em was evaluated as a West inghouse generic iter during the review However , ac cept ance of thia Donald C. Cook Nuclear I'lant , tinits 1 & 2. >
of the design f or utiwr plants is conditioned on verification of the following: (1) set smic and environment a1 eiual I f lcat lon; (2) that the prot ei t ive f unct ions provided meet t he wafet y crit eria f or t he part i. w!ar pl .uit : and (3) implement at Ion of the syst.m requirement s f or pre ,crvirs t*"
the inlepenitenee of redundant port ion . i.f the prot ce t ion svotcri.
e conduc t ed a detailed resiew of these arew , nr.d supplement ed our te wit h a wit. visit to exan,ine the phr , f r.il installat ion of t he elet t r! O t*tif f ((Gefit .
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l 7.2.2 1 Physical _ Separation During the site visit, we found that the physical separation in the :
solid state protection system racks was inadequate. The input and output The wire bundles terminate at a cousson connector of the isolation board.
center pins of the connector are not used to maintain physical separation; however, there are ne physical b rriers or protection to separate the )
wire bundles as they are routed f rom the connectors. The applicant has agreed to provide physical barriers to separate the input and output wire bundles. Finsi resolution of this item will be reported on in a l
supplement to this Safety Evaluation Rept rt.
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! 1.2.2.2 F.lectrical isolation The photodiode isolators used to electrically isolate the safety signsis irom the non-safety f unctions, as implemented in the solid ,
state protection system, have not been qualified as ace?ptable isolation devices. We informed the applicant that we require tests be performed j
l to verify that the photodiode isolators will meet the design basis requirements for system isolation. The applicant has indicated that he will provide the test procedures and results to substantiate the capability and reliability of the photodiodes as isolation devices. f Final resolution of this item will be discussed in a supplement to l this Safety Evaluation Report.
1.2.2.3 Seismic Qualification We have concluded from our review that the results of the reismic I I
qualification tests of the solid state protection system are unacceptable.
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I I I During the seismic testing of the system, output relays had momentary contac* closure, and in some instances, had a permanent I-change of contact status due to mechanical latching. We have informed the applicant that the staff's position requires that all safety related electrical equipment be designed to withstand the effects of 4
the naft shutdown earthquake without malfunction before, during, or subsequent to a seismic event. We require that the applicant meet the requirements for seismic qualification of the solid state protection system and other safety related electrical equipment (see Section 7.8 of this report).
7.2.2.4 Conclusions We have concluded that the solid state protection system will be acceptable, providing that: (1) the test procedure and results for l I
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qualifying the photodiode isolation devices are acceptable; (2)
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adequate separation or barriers are provided to separate the input and output wiring: and (3) the seismic qualification program for functional electrical opei ability of safety related electrical equipment conforms to our requirements (Section 7.8 of this report).
7.2.3 Process Analeg System
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I Functionally, the process analog system for Diablo Canyon is ;
essentially the same as that evaluated during the CP review. This i
system has been reviewed f or the Zion Station, Units 16 2, and for Trojan. The staf f has concluded that the functional design and implewntation of the process analog system is ac:eptable for l
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these plants. However, as a result of our initial site visit, we have %
determined that the physical separation between the protection and con- ,
i trol circuits within the protection system process analog racks is unacceptable. The protection system input viring to the modular isolation aglifier and the out sut wiring to the control racks are routed in the same utreways. Tne electrically isolated outputs are j 4
also in physical contact with other protection system internal wiring i
of that chant.el in the process racks. These outputs terminate in the q control racks. The redundant outputs of protectisn channels, in j some instances, terminate in a consnon control rack. A failure in the control portion of the system could negate protective actions due to i lack of physical separation between the inputs and outputs of the isolation aglifiers, and result in loss of protective functicn.
The CP Safety Evaluation Report for Diablo Canyon Unit 1 (issued in January 1968) noted that the physical and electrical isolation of protection and control instrumentation was not adequate. The CP Safety Evaluation Report for Unit 2 (iesued in Noveder 1969) t references the same item. The applicant stated, during the CP safety i
review of Unit 1, that the protection system would be designed to i
the proposed IEEE Standards at that time (IEEE Std 279-1968). The (
i implementation of the design presented in the FSAR does not meet the requirements of Sections 4.2, 4.6 and 4.7 of IEEE Std 279-1968. Tne j
applicant further states that the protection system in the FSAR for Units 1 and 2 meets all requirements of IEEE Std 279-1971; however, i
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7-6 the staff's observations during the site visit indicate that the implementation of this present design does not meet theic requirements.
We require that the applicant either (1) andify the present system to provide a minimum physical separation of six inches, or provide barriers between the control outputs of the isolation amplifiers and the protection system circuitry, including the inputs to the isolation amplifiers; or (2) qualify the present system, as implemented, by testing. The test scope will include all tests performed to qualify the individual isolation amplifiers, and aAso monitoring of tne other protection system input and output channels for signal interf erence or degradation. We will report the resolution of this item in a supplement to this Safety Evaluation Report.
7.2.4 Testability of Protectio _n Systems l During our review, we considered the adequacy of the proposed testing of the response time of the protection systems. The applicant considered that a suf ficient testing program was provided by preopera-tional tests, together with the response time tests conducted whenever a component af fecting response time was replaced. The staff concluded that, until experience with the Diablo Canyon design or other identical designs demonstrates that the protection sysces response times, including sensor response times, do not change over long intervals of operating experience, the response time testing should be repeated periodically.
The applicant subsequently agreed to repeat the system response time tests during each ref t.eling outage, but no less f requently _ than every -
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7-7 l 18 months. We have concluded that this is an acceptabic testing program.
The technical specifications will include a requirement specifying this test program.
1.2.5 Anticipated Transients Without Scram (AWS_ )_
The Regulatoiy staff's requi'rements with respect to ATWS are provided in the staff's technical report, " Anticipated Transients Without Scram for Water-Cooled Power Reactors," WASH-1270, dated '
September 1973. As applied to the Diablo Canyon Nuclear Plant, Units 1 & 2, these requirements state that changes should be provided to I J
make ATWS consequences acceptable. Unit I has been classified by the staff as a "1.C." facility; for this Unit, the applicant will submit i
t an analysis describing and evaluating the consequences of a postulated ATWS. Unit 2 has been classified as a "I.B." facility; for this Unit (
the applicant has been requested to implement a program to incorporate any design changes necessary to assure that the consequences of anticipated transients would be acceptable in the event of a postulated failure to scraa in accordance with Section II.B of Appendix A of WASH-1270. The applicant documented the information required by I WASH-1270, for Units 1 and 2, on October 3,1976. The staff evaluation of this information will be contained in a supplement to this Safety Evaluation taport.
7.2.6 Conclusion Subject to favorable resolution of the items indicated in Sections 7.2.2.1, 7.2.2. 2, 7.2.2. 3. 7.2. 3, and 7.2. 5, we have concluded that w-________________________________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
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7-8 ti.e design of the reactor trip system meets the Commission's require-ments and will be acceptable.
7.3 Engineered Safety Feature Actuation Systems 7.' 3.1 Ceneral Our review covered those aspects of the protection system which initiates, monitors, bypasses, tests, and controls the engineered _
safety feature. (ESF) systems and their auxiliary supporting systems, including field implementation, logic diagrams and detailed schenstic drawings. The following sections identify those areas of the design that were changed by the applicant as a result of our review. 4
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l 7.3.2 o ESF Actuat_i,on_L goc, i 6
The ESF actuation logic is part of the solid state protection system discussed ia Section 7.2 of this report. The staff's evaluation of the ESF actuation system desigt.s is presented in the Safety Evaluation Report for the Donald C. Cook Nuclear Plant. Subject to the resolution of the items discussed in Section 7.2 of this report, we conclude that the results of our previous evaluations are applicable to the .Diablo Canyon Units, and that the ESF actuation logic system is acceptable.
7.3.3 Accumulator Is_olation Valves The design proposed by the applicant for the control circuits for the notor-operated accumulator isolation iralves included provisions for the safety injection signal to automatir. ally open the isolation valves.
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1 We concluded that the proposed design of the control circuits did not provide auequate assurance that the accumalater isolation valvos would be open when reanired. The staf f's position, which has been applied to recently-licensed plants, is that in ardition to the safety-injection .
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signal, the design should include provitices to automatically open the 'j isolation valves when reactor coolant pressure exceeds a preselected value. The applicant has modified the control circuit design to conform I
with our' position regarding automatic opening of the valves. We conclude i
that this modified design is acceptable. J
. 3.4 Changeover from Injection to Recirculation Mode The designs of the ECCS and containment spray systems require '
1 operator action to change over from the injec'. ion mode of operation to l the recirc21ntion mode before the refueling water storage tank (RWST) J
'1 is completely emptied following a loss-of-coolant accident. The applicant's original design provided two Icvel instrument s to be used to provide indication of the tank level in the control room. Depending on the failure mode of the level instrument, this could result in the operator either tripping both pumps prematurely or failing to trip the pumps.
As a result of our review, the applicant will revise his design to provide automatic tripping of the lov pressure injection (RRR) pumps ;
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vhen the RWST level decreases to a specified low level (see Section 5.5 of this report for a description of the R3 system). The automatic tripping of the low pressure injecticn uumps significantly increases the time available for the reactor operato? to complete the changeover from the injection to the recirculation code of operation, without degrading
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the required performance.of the core cooling systems b l level, e ow an acceptable -
The revised design will also include three level instruments .
i We have informed the applicant that the instrumentation used to provide the signals for automatic tripping of the low pressure injection pumps. -
and to provide the information that the operator needs to complete the changeover, must meet the requirements of IEEE Std 279-1971 .
We have not reached agreement with the applicant concerni ng the I locking-out of power to the motors of valves, where ncorrectly a single i positioned valve could result in the total loss of the safetyon. functi The difficulty is that some of these valves are presently i ncluded in the procedure for changeover to the recirculation mode , and that the applicant believes that the probability of an incorrectlyoned- positi valve is sufficiently low so as to be considered incredible(See .
Section 6.3.1 of this report for additional discussion of thi ,
We will report s item).
the resolution of this item in a supplement Evaluatiort Report. to this Safety 1
i We have concluded that, subject to the satisfactory resol ti u on of the above mentioned concerns, the operator will have u s ffi i c ent: time to perform the actions required to change over to the recirculation n mode of operation, and the design will meet the Commission's ements requir and is, therefore, acceptable.
7.4 Systems Required for Safe Sh6tdcwn We have reviewed the instrumentation and controlesystems ng bi provided for safe shutdown, including the design provisions c evitg for a hi
7-11 hot and cold shutdown in the event that access to the main control room is restricted or lost. We have concluded that these systems conform to the Commission's requirements and are, therefore, acceptable.
7.5 Safety Related Display Information 1 The safety related display information provides . data to enable the operator to perform the required manual safety functions, and provides information for post-accident surveillance. The instrumentation provided is similar to that for the Zion Nuclear Plant , except for the physical configuration. The applicant has identified the parametere to be monitored during and af ter an accident or abnormal occurrence. The instrumentation is redundant, qualified for the environment, powered by redundant l
vital a-c sources, and a minimum of one channel is recorded. We have reviewed the drawings and verified the implementation during the site J visit. However, the applicant has not provided a description of the bypass and inoperable status indication in the FSAR. The applicant has 'i indicated that he will provide this description. Resolution of this item will be discussed in a supplement to this Safety Evaluation Report.
We have concluded, subject to documentation by the applicant of the description of the bypass and inoperable status indication, that the
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safety related display provides the required information indicated above, i
and that it conforms to the Commission's requirements and is, therefore, acceptable.
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7.6 RHR Overpressure Protection Interlocks The initial design of the interlocks for the motor operated isolation i
valves on the suction side of the RRR pumps, which are provided to l i
preve it overpressurization of the RHR system by the reactor coolant system, did not meet our criteria. We require interlocks of diverse principles in order to prevent opening of these valves when the reactor coolant system pressure is greater than approximately 425 psig, and to automatically close the valves when the system pressure exeseds approximately 600 psig. The applicent has modified the design to conform with these criteria, and, theref ore, we conclude that the design g
is equivalent to those provided on other recently licensed plants, and is acceptable. l 7.7 Control Systems Not Required for Safety The design of these control systems provided for the Diablo Canyon ,
Plant is similar to those for other recently licensed plants, including the Zion Nuclear Plant, except for the following differences:
Zion has analog rod position indication and 50% load rejection capability 5
while Diablo Canyon has digital rod position indication and 100% net load g rejection. We have determined from our review that these design differences do not affect the safety of the plant. We conclude that the Diablo Canyon design is acceptable. ,
7.8 Environmental and Seismic Qualification ['
We have not completed our review of the environmental qualification of safety related electrical equipment for Diablo Canyon. A major portion )
of the equipment was qualified by test programs documented in Topical
7-13 Report WCAP-7744, " Environmental Testing of Engineered Safety Features Related Equipment (NSSS Standard Scope)." This topical report was reviewed and found to be unacceptable by the Regulatory staf f. We therefore conclude that the environmental qualification testing of the NSSS supplied safety related electrical equipment for the Diablo Canyon Plant .ic unacceptable. The applicant has been informed that we require a response to the concerns indicated in the staff's review of WCAP-7744.
The resolution of these concerns will be provided in a supplement to '
this Safety Evaluation Report.
We have not completed our review of the seismic qualification of saf ety related electrical e quipment that must operate before, during, and subsequent to a seismic event. A major portion of the equipment was qualified by test programs documented in WCAP-8021. Based on the results of these tests, we have concluded that the electrical functional capability of the equipment tested, in that it did not operate as was 4
designed, is unacceptable. The applicant has been informed that we require either a test program and results indicating that all safety related s
electrical equipment will operate as designed, without malfunction, before, )
during and subsequent to a seismic event, or else the results of a failure i
mode and effects analysis of all possible combinations of abnormal {
s equipment operations which demonstrates that the safety of the plant is y not compromised during a seismic event. The resolution of this item will be provided in a supplement to this Safety Evaluation Report. .
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i 7.9 conclusion Subject to the resolution of the items indicated in Sections 7.2.2.1, 7.2.2.2, 7.2.3, 7.2.5, 7.3.2, 7.3.4, 7.5 and 7.8, we have 1 concluded that the design of the electrical instrumentation and controls meet the Commaission's requirements, and is, therefore, acceptable.
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S-1 8.0 ELECTRIC POWER 8.1 General The Commission's General Design Criteria (GDC) 17 and 18, IEEE Standards including IEEE Criteria for Class IE Electric Systems for Nuclear Power Generating Stations (IEEE Std 308-1971), and Regulatory Guides 1.6,1.9,1.32 and 1.41, served as the bases for evaluating the adequacy of the electric power. systems af the Diablo Canyon !!uclear 1
l Plant, Units 1 & 2.
l 8.2 Offsite Power l
The Diablo Canyon Plant will be intr. connected to PG6E's electric grid system via two 230 kV and three 500 kr lines emanating from their respective switchyards; these yards are physically and electrically separated and independent of each other. The two incoming 230 kV transmission lines, which share common towers, provide power to the 230 kV switchyard which is a doubic bus arrangement with the capability of feeding the two start-up/ standby transformers (one per Unit) from either source or bus, isolating the other source. The 230 kV power system provides an immediate source of of fsite power. The three incoming
- - kV transmission lines, located on twc rights-of-way, provide power to the 500 kV switchyard which is a breaker-and-a-half configuration with the capability of bacsfeeding through the main transformer of either unit. This pource of offsite power is made available by manually initiating a motor operated disconnecting link which is an integral part l
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1 of the generator's isolated phase bus. This disconnect link is operated f rom the contrcl room and is interlocked to prevent opening under load. ,
1 The applicant states that thit; source of offsite power can be made available by manual initiation in approximately M seconds, providing a delayed sourco of offaite power. The 230 kV and 500 kV switchyards i
have primary and backup relaying control and independent d-c control l l
l power sources for their respective switchyard breakers.
A single line f rom the 230 kV switchyard supplies the standby /
start-up transformer (230 kV/12 kV) of each unit. The low voltage side of these transformers supplies the 12 kV buses and the primary side of another standby / start-up transformer (12 kV/4.16 kV) for each unit. 'he low voltage side of these transformers supplies the 4.16 kV and the three EST buses of each unit.
1 The 500 kV switchyard can supply power by backfeeding through the '
main transformers (500 kV/25 kV) of each unit to the unit auxiliary transformers. One auxiliary transformer (25 kV/12 kV) per unit sup-l plies the 12 kV buses and a second auxiliary transformer (25 kV/4.16 l kV) per unit supplies the 4.16 kV buses and the three F.Sr buses of each unit. Either the standby / start-up or unit auxiliary transformers l and their attendant distribution systems have sufficient capacity to supply shutdown and emergency load requirements.
The applicant has performed an electrical grid stability analysis which indicates that the loss of any single generator on the grid, 1
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-including that for Diablo Canyon Units 1 or 2, while operating at full load, will not adversely affect the stability of the remainder of the transmission grid or the abilit) to provide offsite power to the Diablo Canyon Plant. Furthermore, Diablo Canyon Units 1 and 2 have the capacity of 2001 ner load rejection without reactor trip.
The design of the offsite electrical power system has provisions for periodic inspection and testing to demonstrate that all Class IE electrical systems are capable of perforcing their safety functions.
We have concluded that a coe'. ', nation of either 230 kV circuit and one of the three 500 kV tircuits provides sufficient assurance that redundant and indeper. dent sources of offsite power are provided, as required by AEC Ceneral Design Criterion No.17, and that the design of the off site power system meets AEC Ceneral Design Criteria Eos.17' and 18, IEEE Std 308-1971, and Regulatory Guide 1.32, and is, therefore, acceptable.
8.3 Onsite Power 8.3.1 A-C Power Systems The engineered safety feature loads are divided among three 4.16 kV ESF busew per unit. The ESF loads are assigned to these buses such that operability of any two 4.16 kV ESF buses and their attendant distribution.
systems will supply the minimum safety requirements for that unit. Two sources of of f site power are provided to the three ESF buses per unit, as discussed in Section 8.2 of this report.
The onsite power source for Units 1 and 2 is supplied by five 4.16 kV, 2600 kW diesel generators. Four diesel generators are separately assigned to four ESF buses, two per unit. The fifth diesel generator is assigned
I 1 8-4 4 I to be shared betveen the third ESF bus f rom each unit. The shared l diesel generator swings to the unit having a safety injection signal.
1 Undervoltage of an ESF but, will start its respective d!esel generator !
and a safety injection signal vill start all of the diesel generators.
The diesel generators for Diablo Canyon have been previously qualified for use in nuclear power plant applications. and are being utilized as onsite y wer sources at several nuclear plants. The onsite system has the capability to provide power for redundant ESF equipment for one unit in an accident condition, and safely shutdown the other unit. The system also has the capability to provide the minimum ESF equipment power requirements for one unit in an accident condition, and safely shut down the other unit, assuming a false accident signal or a single failure. No communication is required between the operators I
of both units to maintain each unit in a safe condition.
The diesel generators are individually located in separate Seismic Category I compartments on the getund floor of the turbine building.
l These compartments are protected from fire, flooding, and external and internal missiles. The fuel oil system has two underground fuel l all storage tanks which are shared by both units. The fuel oil syster is designed to remain operable af ter sustaining a single failure of either an active or passive component. The total onsite fuel oil storage capacity provides a minimum of seven days of fuel for diesel generator operation supplying the power rea"irements of: (1) the minimus ESF equipment for one unit, and (2) the equipment required for maintaining the second unit in a shutdown condition.
8-5 The 120 volt a-c vital instrumentation bus system supplies power for instrumentation, control, protection and annunciation. The vital a-c system consists of separate and electrically independent buses and attendant distribution systems, with each supplying their respective redundant load groups. Each bus is served by a separate inverter which is supplied f rom a 125 volt d-c system or from one of the 480 voir a-c ESF buses; this satisfies the sinnie failure criterion. An additional 120 volt a-c source, fed f rom a 480 volt ESF bus, is available, through a manual transfer switch, to provide power to the vital instrument buses when an inverter is out of service.
8.3.2 D-C Power Systems The onsite d-c emergency power systems for each unit consist of 125 and 250 volt systems. Three 125 volt batteries are provided for each unit (batteries 11,12 and 13 for Unit 1, and batteries 21, 22, and 23 for Unit 2). Each unit has one of these 125 volt batteries assigned to a separate and independent bus (batteries 13 and 23 for Units 1 and 2, respect ively) . Batter) 11 is in series with battery 22 providing a 250 volt d-c system for Unit 1, and battery 12 is in series with battery 21 providing a 250 volt d-c system for Unit 2. The 125 volt d-c sources are shared as described above, between Units 1 and 2; however, there is no interconnection between redundant syst ems wit hin each unit.
The 125 volt d-c systemt provide power for control, instrumentation, annunciat ion and emergency light ing. The 250 volt d-c systems provide power for the turbine lubricating systems. Eact, of the 125 volt d-e u
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8-6 j buses has a battery charger assigned to it. The two buses shared with the other unit have a backup swing charger that can be manually connected to either bus. The third 125 volt d-c bus has a redundant charrer. Each of the five charFers per unit are capable of maintaining the required j d-c voltage on the system and recharging the batteries. The battery chargers are supplied by the onsite standby power system in the event j of loss of station and of feite power. Separate battery rooms and switchgear rooms are provided for each of the 125 volt d-c distribution I systems.
j 8.3.3 Conclusions We have reviewed the ucsign of the onsite a-c and d-c power distribution .
systents, and have determined that the design meets AEC Ceneral Design Criteria Nos.17 and 18, IEEE Std 308-1971, and Regulatory Guides 1.6, 1.9, and 1. 32. The applicant will conduct preoperational tests to confirm the independence of redundant onsite power distribution systems, their controls, and loads. These preoperat* inal tests will meet the recommendations of Regulatory Culde 1.41. We conclude that the design of the electrical power systems meets the Consnission's requirements, and is, therefore, acceptable. 1 8.4 Physical Indepe_ndence of E...etria_al Equipment and Circuits )
i Section 8.3.3 of the FSAR pr>vides a description and eTalysis of the i
applicant'n criteria and procedarty for providing physical independence i t,f safety related circuits and equipment. We have reviewed the applicant's criteria and procedures regarding sef aration, and verified their implementation during our initial sit e visit. The applicant has not 2
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i responded to our request to provide a description and analysis of his criteria for protection of Class IE cabling and equipment in hazardous i and missile prone areas. We will report the resolution of this item in a supplement to this Safety Evaluation Report. With a satisfactory response to this item anJ resolution of the staf f's concerns regarding j separation in the process analet system (see Section 7.2.3 of this report),
we conclude th the separation criteria meet the Commission's requirements.
and are, therefore, acceptable.
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- 9.( AUX 1tl.%RY SYSTEMS 2e 9.! General In the course of our review of the auxiliary systems, we have directed
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our atten' ion to the design of the systems with respect to the safety
- related objectives of these systems, and to the manner in which these objectives are achieved.
The auxiliary systems necessary to assure reactor safety are: '1) j p
auxiliary saltwater system, (2) component cooling water system, (3) 4 g
portions of the makeup water v* stem, (4) condensate storage f acilitie. 4 m be, c>njunctica vith the auxiliary saltwater (5) ultimate heat si y
system and the intake strueta ?), (6) control room heating, ventilation, i
" and air conditioning system, (?) auxiliary building heating, ventilation,
," and air conditioning system, (b) fuel handling area heating and ventilation 2
I system, (9) fire protection system, and (10) diesel generator auxiliary l
-I systems. The systess necessary to assure safe handling and adequate cooling Y-of the spent fuel include the fuel handling systems and the spent fuel
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4 pool cooling and cleaning system.
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- We have also reviewed those auxiliary systems whose failure would h5 prevent safe reactor shutdown, but could, either directly or indirectly, not 4
e be the potent.lal source of a radiological release to the environment.
3 These systems include the floor drainage and sampling system. From our :
review of the proposed design of the floor drainage and sampAing systems
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for the Diablo Canyon Units, we find that they are comparable in design l and function to other PWR facilities that have been previously reviewed and approved. On this basis, we have concluded that these auxiliary systems are in compliance with the applicable rules and regulations and are acceptable. .
We have reviewed those systems and components to be shared by both Units, and find that their designs meet the requirements of AEC General Design Criterion No. 5, and are acceptable.
9.2 Fuel Storage and Handling 9.2.1 New and Spent Fuel Storage Fuel handling and storage facilities for each unit are provided for storage and transfer of new and spent fuel. Spent fuel will be stored underwater in the spent fuel storage pool and new fuel will be stored dr:>
in the new fuel storage area. The new and spent fuel storage racks are de-signed to prevent assemblics from being placed in other than prescribed locations. The new fuel storage area accommodates one-third of a core and the spent fuel pool accommodates a full core plus the normal quantity of spent fuel assemblics from the reactor during refueling (usually one-third of a core). The fuel is stored in a vertical array with sufficient center-to-center distance between assemblies to assure thatek rt will not exceed 0.90, even if unborated water is used to fill the pool.
We conclude that the designs of the new and spent fuel storage facilities are acceptable.
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9-3 9.2.2 Spent Fuel Pool Cooling and Cleanup System The spent f uel pool cooling sys'.en is designed to remove heat generated by stored spent fuel elements in the spent fuel pool. A l
secondary function of this system is to clarify and purify spent fuel j pool water, transfer pool water, and refueling water. The system is 1
designed and constructed to Seismic Category 11 requirements; however.
demineralized makeup water can be added to the pool from Seismic Cate- l gory 1 sources. Each Unit has an independent cooling and cleanup I
system for its spent fuel pool. The system dissipates decay heat from j stored fuel to the component cooling unter system. The system is designed I
to handle the decay heat generated by one-third of a core in the pool, and the temperature of the spent fuel pool water under this condition s.11 not exceed 120*F. If it is necessary to remove a complete core from the reactor while the spent fuel assemblies from the previous refuelieg remain in the pool, the system can maintain a pool water temperature at or below 150*F.
I We have reviewed the design of the spent fuel pool cooling and elean-up system, and conclude that it meets the intent of the positions set forth in Regulatory Guide 1.13. " Fuel Storage Facility Design Basis," and AEC i
General Design Criterion No. 61. Therefore, the design is acceptable.
9.2.3 Fuel Handling System For the Diablo Canyon design, the cask loading area is in the spent fuel pool. During cask handling operations, unacceptable damage re-sulting from a spent fuel cask drop will be prevented by limiting the
9-4 travel of the spent f uel cask over areas that contain no safety related equipment or stored fuel. The travel of the cask bridge crane is limited by mechanical stops and limit switches. The subject of cask i i
movement over spent f uel stored in the pool has been discussed with f i
the applicant. PC6E has tentatively agreed that spent fuel will not I i
be stored in the spent f uel pool in locations where it could be struck by a dropped cask, assuming the worst tipped position for the dropped ,
cask. However, final resolution of this item is still pending, and will be reported on is d supplement to this Safety Evaluation Report.
We have considered in our review the design safety features of all {
the fuel handling equipment, and judge that the system design meets i u d" the intent of the positions stated in Regulatory Guide 1.13 and the i requirements set forth in General Design Criterion No. 62. We conclude, !
subject to f avorable resolution of the dropped cask ites, that the design of the fuel handling system is acceptable, q f
9.3 Water Systems 9.3.1 Auxiliary Saltwater Systen The auxiliary saltwater system provides cooling water from the Pacific Ocean to the component cooling water heat exchangers. For each of the Diablo Canyon Units, two separate pump compartments are located in the intake structure which is located on the Pacific Ocean. Each of the pump compartments contains a full capacity pump. Each unit is l
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9-5 provided with two pueps, which are headcred into two separate traine,.
The double valved, nomily open interconnection between the two treins is provided with remote operating capability. A normally closed esotor-I ope rated alve provides separation between the Unit i and Unit 2 aux-111ary saltwater discharge headers. I i
The entire auxiliary saltwater system, including the intake structure, la designed to withstand the ef fects of the safe shutdown earthquake.
Furthermore, the system is, designed such that a single failure of any component. of the system, or the onsite power system, will not prevent a safe shutdown of the plant.
We have reviewed the design of the auxiliary saltwater system and have concluded that it meets the requirements of AEC Ceneral Design Criterion No. 44, and is acceptable.
9.3.2 Component Coolina Water System The component cooling water system is designed to remove residual and sensible heat froa. thee reactor e .alant system via the residual heat removal system during plctic shutdown, to cool the spent fuel pool water, to cool the letdown flow to the chemical and volume control j i
system during power operation, and to provide cooling to dissipate f i
waste heat from various primary plant components. The system design provides means for detection of radioactivity entering the system f rom the reactor coolant system and associated auxiliary systems, and includes provisions for isolation of system components as required. l l
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9-6 During normal full power operation, two of the three component cooling water pumps and one of two component cooling water heat ex-changers acconsodate the heat removal loads. The standby pump and heat exchanger provide backup during normal operation. Both component' cooling water heat exchangers provide removal of residual and sensible heat during a normal plant shutdown. Failure of one of these heat exchangers increases the time required for shutdown, but does not affect the safe shutdown of the plant.
In the event of a loss of coolant accident, one pump and one Seat exchanger are capabis of fulfilling system requirements. The system design provijas three main cooling headers, two isolable headers that supply cooling water to essential safety equipment, and one header which supplies cooling water to the other plant auxiliaries. Under this arrangement, long-term cooling of the engineered safety features under accident conditions is assured, assuming either active component f ailures or the development of excessive leakage in one header of: the component cooling water system. Cooling water for the component cooling water heat exchangers is supplied from the auxiliary saltwata.r system, thereby assuring a continuous source of cooling under all conditions.
I The component cooling water system is a Seismic Category I system which is required for post-accident removal of decay heat f rom the reactor. As such, the system is designed to meet the single failure criteria with two completely independent, parallel trains available, 9
9-7 each containing one pump and one heat exchanger. The applicant pre-sented in the FSAR the results of a failure mode and effects analysis of the system pumps, heat exchangers, and valves; we found the results-of this analysis te be acceptable.
We have reviewed the design of the component cooling water system and have concluded that it meets the requirements of AEC General Design Criterion No. 44, and is acceptable.
9.3.a dakeup Water Systen The makeup water system utilises distilled seawater to supply 150 gpm of distillate to storage tanks, makeup demineralizers, and reser-voirs. This syste. is designed to provide a source of water for make- '
up to the reactor coolant loop, supply and makeup for the fire protec- !
tion sys'.f.in and the spent fuel pool, and supply for the component i
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cooling water system. The makeup water system, from the condensate storage tanks to the auxiliary feedwater pumps, and from the conden-sate storage tanks through the makeup transfer pumps to the component cooling water systen, is Seismic Category 1 in order to meet the de-sign requirements with respect to safety related systems.
We have reviewed the design of the makeup water system and con-clude that it is acceptable.
9.3.4 Ultimate Heat Sink The ultimate heat sink consists of the Pacific Ocean. This body of water, together with the intake structure and the discharge system O
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f in conjunction with the auxiliary saltwater system, provides a means of supplying cooling water for reactor equipment to use as a heat sink.
the intake structure is located directly on a small inlet cove of the ocean, thereby assuring the source of cooling cater (see Figure 2.2 of this report). With regard to low water level conditions in the cove, t.e most severe oceanographic phenomenon would be a tsunami (see Section 2.4.6 of the FSAh). tlc expected downsurge from the tsunami during short periods of time would be to 9 feet below MLLW (mean lower l
low wa,er) . The arrangement of the intake channel and the design of the 1 auxiliary saltwater puwps allows operation down to 17.4 feet below HLLW.
Thus, the operation of the Seicmic Category I auxiliary saltwater pumps For reference, NLLW equals ,
is assured during extreme tsunami drawdown.
MSL (mean sea level) minus 2.6 feet. MSL is ground elevation zero.
Based on uur evaluation of the ultimate heat sink, we conclude J
that the design meets the positions set forth in Regulatory Guide 1.27, !
" Ultimate Heat Sink for Nuclear Power Plants," and is acceptable.
9.4 Process Auxiliarie_s 9.4.1 Chemical and Volume Control System .
The chemical and volume control system (CVC',) is designed to ad-just the concentration of chemical neutron absorber (boron) in the reactor coolant for reactivity control, maintain the proper water in-ventory and concentration of corrosion inhibiting chemicals in the reactor coolant system, provide required seal water injection to the .
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9-9 reactor coolant pump seals, and remove corrosion products and fission products from the reactor coolant. Two separate and independent flow paths are availt.ble for reactor coolant boration, i.e. , the charging line and the reactor coolant pump seal injection. In the event of loss of of f site power, the safety (borativa) function of the system would I be maintained, in that power to the charging pumps and associated valves would be available from the diesel generators. To evaluate i I
the safety of the system, f ailures or malfunctions were assumed con-current rich a loss-of-coolant accident, and the consequences evaluated.
J The results of the failure analyses were found to be acceptable.
On the basis of the similarity of the design of the Diablo Canyon chemical and volume control system to that for previously reviewed and approved plants, such as Zion Units 1 and 2. We have concluded 1
that the system is acceptable.
9.5 Air Conditioning, Heating, Cooling, and Ventilation Systems 9.5.1 Control Area l The control area air conditioning, heating, and ventilation sys- ]
tems are designed to provide a controlled environment for the control l r:>om, the computer room, and the instrument safety-feature rooms for 1
each Unit. These systems are designed to Seismic Category 1 require- j ments, and are required in order to assume continuous occupancy of the ]
' i control room under normal and accident conditions, j i
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9-10 As a result of our review, the applicar.t has added chlorine detec-tora in the control room supply duct, with local. and remite alarms (sec.
Section 6.4 of this report).- The remote alarm will be used for auto-matic initiation of an internal. recirculation operational mode, the same as would be initiated for a high activity alarm in the supply duct. The systema also have features for protection f rom smoke gen-ersted either inside or outside the control room area. The systems ;
have two full capacity units of equipment for each primary element.
The source of power for each electrically powered primary unit of equip-ment is from a vital bus.
We have reviewed the design of the control area air conditioning, j heating, and ventilation systems, and conclude that they are acceptable.
9.5.2 Auxiliary Building (Excluding the Fuel Handling Area) ,
f The auxiliary building heating and ventilation system has the primary function of maintaining the temperature of the engineered safety feature pump motors within acceptable limits during their opera-tion. The secondary function of this system is to provide heating and ventilation to the auxiliary building.
The system is designed, built and installed.to Seismic Category I requirements. The power for t5e fans and initiating (logic) circuitry is taken from vital buses. All dampers are designed to assume the position required for emergency operation upon failure of damper motive power. All primary active co nponents of the system, including initia-tion circuitry, are redundant. ;
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l We have reviewed the system design with resocct to its ef fects on i the operation of equipment essential for safe shutdown of the reactor, ,
and conclude that it is acceptable. 4 9.5.3 Fuel Handling Area of the Auxillag Building The principal function of the fuel handling area heating and venti-lation system is to sweep radiolytic gases from the surface of the spent fuel pool, and to treat the exhaust air in order to remove most f of these gases. The purpose of treating the exhaust air is to reduce i the offsite dose to acceptable levels in the event of a fuel handling !
accident. The systeta is designed and built to Seismic Category 1 re-quirements, j i
The fuel handling area for each of the Diablo Canyon Units is j physically isolated from the rest of the auxiliary building. The heating and ventilation system consists of redundant supply fans, re-dundant exhaast fans, and redundant HEPA and charcoal filter banks.
l A third fuli capacity exhaust f an and HEPA tilter bank train is pro- I vided for normal operation, and is either automatically or manually 1
I prevented from operating when exhaust cleanup through the charcoal fil-ters is required.
We have reviewed the design of the fuel handling area heating and ventilation system, and conclude that it is acceptabic.
9.5.4 Intake Structure (Auxiliary Saltwater Pump Compartments)
Units 1 and 2 are each provided with two redundant auxiliary saltwater pumps. Each punp is installed in a separate watertight compartment in the e
9-13 intake structure.
Proper ventilation of these compartments is assured by providing each compartment with a separate ventilation system consisting of intake louvers, an exhaust fan and an exhaust duct. The: power sources for -l i
these fans are supplied by ESF buses. The function of this system is i
to maintain the temperature of the auxiliary saltwater pump motors-within acceptable limits during their operation. The system is designed l
and built to Seismic Category I requirements. Minimum saltwater pump-ing requirements are met, assuming a single. f ailure in the ventilation system.
We have reviewed the design of the ventilation system for these i compartments, and conclude that it is acceptable.
9.5.5 Diesel Generator Compartments The ventilation systems for the diesel generator compartments i have the function of maintaining the air temperature of the compart- 8 ments within acceptable limits during diesel operation and providing 1 cooling air for the radiator of the diesel generator closed loop jacket water system. i l
Each diesel generator compartment has its own Seismic Category I !
ventilation system. Ventilation :I the compartments is assured when the diesels are operating, since the ventilation for each compartment !
1 is provided by the same direct engine-driven fan which provides cooling 1 air to the radiator. No special provisions for heating the diesel gen-erator are required since the diesel jacket water and lubricating oil .
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are kept warm by thermostatically controlled heaters during periods when the diesel generators are not operating.
We have reviewed the design of the diesel generator compartment ventilation systen, and conclude that it is acceptable. l 6 Other Auxilla,rd v6 tens l
.6.1 Fire Protection System The fire protection system has been designed to: (1) provice l 1
automatic or manual fire extinguishing capability; (2) provide redundant I
detection equipment throughout the plant; (3) provide automatic suppression l 1
systems in area . where hazardous materials are stered, when the malfunction of these systems will not interf ere with the function of essential equipment; and (4) comply with the standards of the National Fire Protection Assoclatfon.
l Two reliable water supplies, a 300,000 gallon storage tank and a 4.5 million gallon raw water storage reservoir, are provideu as sources of wrter for fire protection. Water is piped to all levels of all f acility buildings (both units), with the supply lines sectionalized by valves for isolation in the event of damage to any section of the line. Water spray systems are provided at high hazard locations, with hand hoses and CO2 devices at all other points. In order to minimize the potential of fire throughout the plant , non-combustibic and heat resistant materials have been used wherever practicable.
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On the basis of our review, we have concluded that the fire pro-tection system conforms with the intent of the requirements of AEC General Design Criterion No. 3 and, therefore, is acceptable.
9.6.2 Diesel Generator Auxiliary Systems )
l Each diesel generator has two independent air starting systems The original design provided powered by separate d-c power sources.
automatic switching of d-c power sources if the primary source failed.
j We concluded that this was unacceptable since a single failure in the l automatic switching could result in the loss of two independent d-c power sources. The applicant modified the design providing for manual l 1
switching of the standby d-c power source for starting. We have con- .
cluded that the modified design meets our requirements and is acceptable.
Each air starting system consists of an air receiver which provides 45 seconds of continuous engine cranking, an air compressor, and two air starting motors. The system is Seismic Category 1. Diesel combustion air is taken in through Seismic Category I filters from the west wall j
of the building, and the engine exhaust is directed out the north side of the building; this assures that the exhaust will be removed without diluting the combustion air. The lubricating oil system for each engine 1
is entirely contained on the engine's base plate. The oil is cooled in This closed cooling water the jacket cooling water heat exchanger.
Thermostatically system is in turn cooled by the engine air radiator.
controlled inversion heaters keep the jacket water and lubricating oil I
j a In j warm for fast starting when the engine is in a shutdown conditien. l i l N
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a f the event of a high energy line break in the terbine building, it is possible that the air flow from the turbine building to the generator 4
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. .a compartmer ts could be diluted slightly to a steam air mixture for a t,
's short period of time. This would not sf fect the engine combustion air, q and would not be expected to affect operation of the engine generator
{ since the engines that will be in use at Diable Canyon are routinely l subjected to a high humidity environment.
9.6.3 Diesel Generator Fuel Oil Syvstem s
A 40,00U gallon fuel oil storage tank for each unit is the source I
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]a of fuel oil supply for the diesel generators. Each tank can supply 1
( one diesel with enough oil to run it for seven days at full load. Two j
.j diesel fuel oil transfer pumps, capable of taking suction from either storage tank, feed a 550-gallon day tank on each of the five diesel l
7 generators. This 550-gallon capacity is suf ficient for two and one- ;
}N half hours of full load operation. The day tek is provided with
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9 level indication and overflows. The entire diesel engine fuel oil y
system is designed to Seismic Category I requirements. During loss of of fsite power conditions, only one diesel generator is required for i
1 each unit.
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, We have reviewed the design of the diesel generator fuel oil sys-
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- tem, and conclude that it is acceptable.
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10-1 I
i 10.0 STEAM AND POVER CONVERSION SYSTEM l
10.1 Sunenary Description l
The steam and power conversion system for Units 1 and 2 is of con-ventional design, similar to those of previously approved plants. Each f system is designed to remove heat energy f rom the reactor coolant in four steam generators and convert it to electrical energy by a turbine ;
i driven generator. The condenser transfers unusable heat in the cycle to the condenser cooling water. The entire system is designed for the maxi-mun expected energy from the nuclear steam supply system. Upon loss i
of full load, the system is capable of dissipating the energy in the !
l reactor coolant through bypass valves to the condenser or through {
l power operated relief valves, dump valves, and safety values to the l I
atsosphere.
l 10.2 Turbine-Generator The main turbine-generator and auxiliary systems are designed for l l
steam flows corresponding to 3496 and 3592 MWt for Units 1 and 2, I
respectively. The Unit 2 turbine-generator has a higher power rating, thereby causing the subsequent higher rating of the Unit 2 NSSS. The turbine generator consists of a four casing, tandem-compound, six flow exhaust unit with a design speed of 1800 rpm. The turbine has one double-flow l high-pressure element in tandem with three double-flow low-pressure {
elements. Moisture separation and reheating of the steam is provided i
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f 10-2 I between the high- and loepressure turbines by six horizontal-axis, f I
two stage reheat cylindrical-shell combined moisture-separator-reheater {
assemblies.
The turbine electro-hydraulic control system will control the speed of the turbine by modulating the turbine inlet steam control valves to control the steam flow to the turbino. The control system regulates turbine speed prior to the time that the generator is synchronized, and controls unit output when the generator is connected to the grid. The 1
turbine control system will be designed to automatically trip the tur-bine under the following conditions: turbine overspeed, low condenser l vacuum, thrust bearing failure, low bearing oil pressure, high-high steam generator water level or safety injection, generator electrical trips, and reactor trip. The turbine overspeed trip system co sists of a mechanical trip which functions at 110 percent overspeed, and an electrical solenoid trip which is actuated at 111.5 percent overspeed.
We have reviewed the turbine generator control systems and pro- i j
tective devices, and conclude that they are acceptable.
10.3 Main Steam Supply System The steam generated in each of four steam generators will be routed to the turbine by means of the main steam lines. F,ach steam line contains a flow restrictor to limit maximum flow and the resulting thrust loading caused by a steam line rupture. Each line will contain one main steam isolation valve (stop valve). The four main steam lines
10-3 will join together in a pressure and temperature equalization header from which the steam is normally routed to the high pressure turbine.
Steam dissipation capability in the event of a turbine and/or reactor trip will be as follows:
(1) The turbine bypass system will have the capacity of dumping 40 percent of full flow steam to the main condenser. This, in conjunction with a 10 percent step load change capability of i
the nuclear steam supply system, permite a 50 percent turbine generator load rejection without a reactor trip, turbine trip, or safety valve actuation; (2) The power operated relief valves will have the capacity of dumping 35 percent of full flow steam to the atmosphere.
These valves will be used for controlled cooldown during loss of offsite power conditions; (3) The steam generator safety valves, having a capacity in excess of 100 percent full steam flow, will open as a last resort.
They will be set to relieve at higher pressures than the power relief valves. .
l The maili steam isolation valves (back to back check valves) will j i
be designed to provide positive isolation against forward or reverse steam flow. The main steam line isolation valves will close if I there is evidence of a main steam line rupture, as indicated by signals of containment high pressure or high steam flow coincident !
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10-4 Closing ,
with low steam line pressure or low average tergerature. 1 time will be 5 seconds or less. The valves can' be operage 2 f rom the control room. The applicant is presently performing an analysis of l
.the ability of the main steam check velves to remain functional Our evale.atica et following a steam line break upstream of the valve. .I this analysis will be reported os in a supplement to this Safety Evaluation Report.
l We have reviewed the main steam system and isolation valve designs, together with the testing program, and we conclude, subject to '
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f avorsble resolution of the item discussed above regarding integrity
.j of the check valves, that these designs are acceptable.
10.4 Other Features This section discusses subsystems of the steam and power conversion system that are used during the process of converting thermal energy to These include the main condenser, turbine bypass J electrical energy.
system, circulating water system, and the condensate and feedvc:er systems. Other subsystems of the steam and power conversion system that have been reviewed but not discussed in detail are the main '
condenser evacuation system, the turbine gland sealing system, and These system the condensate and feedwater chemical isjection system.
designs have been found to be siellar to those of previously approved facilities, and are acceptable. The steam generator blowdown system it discussed in Section 11 of this report.
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The main condenser has been designed to serve as a heat sink for '
the turbine exhsust steam, turbine bypass steam and other flows. The
condenser hotwells are sized to provide adequate storr>ge of water to <
allow for losses to the atmosphere and shrinkage on a full load trip.
During operation, air is removed from the condenser by steam jet air i
ejectors and discharged to the plant vent. The discharged air is continuously monitored for radiozetivity.
l The turbine bypass system (TBS) discharges main steam directly to
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I the condenser during the emergency condition of a sudden load rejection
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by the turbine generator or t'urbine trip, and during plant startup and i
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shutdown. The turbine bypass system has been designed for a total i
steam flow capacity equivalent to 40 percent of the full-load main )
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steam flow. There are twelve power relief valves which take steam from i
the dump header; these valves dia, charge into spray distribution headers in the condenser. Four of these twelve valves are used during cooldown of the reactor. ,
The circulating water system furnishes the main condenser with
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l cooling water from the Pacific Ocean. For optinum turbine-condenser ,
performance, the circulating water system utilizes two pumps to supply I
876,000 gpm of salt water to the condenser of each unit. The water l
passing through the condenser is returned to the Pacific Ocean via the discharge structure.
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' 1 10-6 Because the circulating watt ? sycten is not Seismic Category 1, the staff has considered the p;,tecital effects of flooding due to a break in this systea. In pa r ,1. 4) ar ,. we have reviewed the consequences of the rupture of c circul. og wa*. r line expansion joint. In order to preclude iloodina - e a e t ,, . Inted equipment in the turbine building due to the rupture of this pint, the applicant has installed an 3 This sleeve provides expansion joint sleeve aroand each expansion joint.
an essentially watertight barrier. Weep holes on each expansion joint We have sleeve are provided for indication of expansion joint failure.
concluded that by the use of these sleeves, an expansion joint failure will not adversely effect the safe shutdown of'the reactor.
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With regard to the remaining exposed portions of the circulating water pipe, we have requested that the applicant supply design modifica-I tions to the turbine building to ensure that rupture of this pipe
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' would not impair safe shutdown of the reactor. Specifically, in order to prevent water from entering the compartments containing the diesel generators, the applicant has proposed a two foot high door separating the entrance to the emergency generator hallway from the main turbine l building floor. The staff is currently reviewing this proposed design modification. Resolution of this item will be discussed in a supplement i
to this Safety Evaluation Report. !
The condensate and feedwater system has been designed to process {
l the condensate to maintain the required quality of feedwater and provide l I
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I the required amount of feedwater to the steam generators at the ,
,l associated feedwater temperature and margin of flow to acconnodate al'. l' f
anticipated transient conditions. i
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From our review of the steam and power st*5 systems, and subject to favorable resolution of the item discussed above regarding modificatfor.s : 1 i
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to the turbine building, we conclude that the design of these subsystema i is acceptable. yI h
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10.5 Auxiliary Feedwater System 5
The auxiliary feedwater system supplies water to the steam (l 1
generators for reactor decay heat remaval if the normal feedwater [
sources are unavailable due to loss of offsite power or other mal- ff 3'
function. Each unit is equipped with one full capacity turbine-driven j and two half capacity motor-driven auxiliary feedwater pumps. Steam ii
.5 for the turbine-driven pump is taken f rom two of the four main steam q lines upstream of the steam generator isolation valves. Separate j isolation valves are provided for these branch connections. The fi motor-driven pumps receive power from the vital buses.
The turbine-driven auxiliary feed pump, rated at 930 gpm and 3000 we 5,)
4 l feet discharge head, and the motor-driven auxiliary feed pumps, rated Y!
7l at 490 gpm and 3000 feet discharge heat, take suctian f rom the {
tr condensate storage tank. Each auxiliary feedwater pump discharges ,
to all four of tne steam generators. Feedwater flow is controlled from fp
.>i the control room by renotely operated flow control valves in the supply lines to each steam generator. (
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10-8 The turbine-driven pump automatically starts upon loss of offsite power by opening the turbine stop-start valve in the steam supply line.
The motor-driven pumps start automatically during conditions of loss of of fsite power or loss of the turbine-driven feed pump system. The pumps can be started f rom the control room. The system has been designed and built to Seismic Category I requirements.
The condensate storage tank, from which the auxiliary feed pumps ;
l take suction, is also Seismic Category I. The tank has a capacity of 425,000 gallons, of which 170,000 gallons is required for the auxiliary feedwater system during plant cooldown. The tank is adequately i
protected f rom the ef fects of flooding, and has a tornado resisting l capability (combined wind and missile) for winds up to 150 mph. Missile damage to the tank below the 170,000 gallon level could impair its safety related function. The tank has suf ficient capacity to cool the reactor coolant to 350*F and reduce system pressare to 350 psig. The 1
4.5 million gallon raw water reservoir serves as a backup source of l water.
We have reviewed the auxiliary feedwater system capability in accordance with the criteria set forth in A. Ciambusso's letter of ,
December 1972 with respect to high energy piping system breaks outside containment (see Section 3.6 of this report). Based ca this review, we have determined that if a high energy line fails outside of containment,
10-9 ]
1 and assuming a concurrent single active component f ailure in the auxiliary feedwater systes, that the minimum feedwater flow from one of the motor-driven pt.aps (490 spa at 3000 f eet discharge head) is sufficient for removal of reactor decay heat.
We have reviewed the auxiliary feedwater system ano the condensate 4
storage facilities, and conclude that their designu'are acceptable.
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11-1 11,0 RADI0 ACTIVE WASTE MANAGEMENT 11.1 Susanary Description The radioactive waste systems, which will be shared by the two units, consist of the liquid, gaseous, and solid waste systems.
The liquid waste system will process waste liquid streams such as equipment drains, leakage, blowdown domineralizer regenerant waste, decontamination and laboratory waste liquids, and laundry and shower waste water. The trested liquid waste will be recycled for reuse if is the plant waste balance requires makeup and if the water quality adequate. The liquid waate system will utilize evaporation, dominera11sation, and filtration for removal of radioactive material, chemical impurities, and particulate.
Caseous wastes will be generated during the operation of the plant as a result of desassing of the primary coolant, vents from equipment handling radioactive materials, and leskage f rom systems and components containing radioactive material. The gaseous waste system will remove radioactive materials from gaseous streams by filtration, adsorption, and holdup for radioactivity decsy. The treated gas streams will be released to the environment.
Solid wastes will siso be generated during plant operation: The vastes vill consist of such contaminated material as clothing, evaporator bottoms, demineralized resins, and discarded radioactive components and tools. Treatment will consist of solidification, packaging, and I
shipping to a licensed burial site.
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1 11-2 In the Final Environmental Statement (FES) for Diablo Canyon, 'I Units 1 and 2, dated May 1973, the staff performed an evaluation to determine the quantities of radioactive matcrisla that will' be released in the liquid and gaseous plant effluents, and that will be shipped offsite as solid wastes for burial. In that evaluation, waste flows, waste activities, and equipment operating performance that are consistent with normal plant operation were considered, including anticipated operational occurrences over the life of the - 1 l
Plant. The parameters used in the FES evaluation, along with their. I bases, are given in Appendix B to WASH-1258. Modified versions of '
the ORICEN and STEFFEC Codes, which were the liquid and ~ gaseous cal- 4 I
culational models that were used, are discussed in Appendix C to l
.l WASH-1258. Our evaluation of the system decontamination factors, along with a listing of plaat dependent parameters, is given in Table 3.5 of the FES.
11.2 Liquid Waste System 11.2.1 System Description The liquid radwaste treatment system is designed to collect and process wastes based on the chemical purity, relative to the primary coolant, as determined by the origin of the waste in .the plant. The boron recycle system (BRS) will process shim bJeed and equipment drain vaste, collected inside the reactor containment, by means of evaporation and demineralization. The liquid waste treatment system (LWTS) will process, by evaporation and demineralization, equipment drain vastes and
11-3 tank overflow wastes from components outside reacter containment, I waster from laboratories and sampling drains, and demineralized regeneration solutions and mi .laneous low purity wastes which have !
been collected in floor drains and building semps. The LWTS will also i
process detergt-nt wastes and/or turbine building floor drain vastes I should radiation measurements indicate higher than expected radio-activity levels. These wastes will normally be filtered and released .
l without treatment, af ter monitoring for radioactivity. The steam j generator blowdown treatment systen (SGBTS) will process blowdown wastes by mixed-bed demineralization. The BRS and LWTS are shared 1
l between Units 1 and 2, while the SCBTS is a separate system for 1
each unit. The principal components making up each of these systers, along with their capacities, are listed in Table 11.1.
11.2.2 System Evaluation j
l In cur evaluation of the liquid radwaste system we have considered the following criteria: (1) the capability of the system to reduce radioactive releases to "as low as practicable" levels based on expected redwaste inputs over the life of the plant; (2) the capability of the system to maintain releases below the limits in Appendix B of 10 CFR
? art 20 (ste Table 2, Column Il for periods c f fission product leakage at design levels from the fue]); (3) the capability of the system to meet the processing demands of the plant durirg anticipated operational occurrences; (4) the quality group classification and seismic category l
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I 11-4 applied to the system design; and (5) the design features incorporated I
to preclude uncontrolled releases of radioactive materials due to tank /
overflows. The process and t fluent monitoring design capabilities are considered in Section 11.4 of this report.
Our evaluation of the liquid radwaste treatment system for normal operation is given in the Sinal Environmental Statement .or Diablo Canyon. In the FES we have determined that the proposed liquid radwaste i
treatment systems will be capable of reducing the relesse of radiactive i 1
materials in liquid ef fluents to approximately 5 Ci/yr/ reactor, excluding tritium and dissolved gases and 350 Ci/yr/ reactor for tritium. An isotopic listing of our calculated liquid source term is given in Table 3.6 nf the FES. Based on that evaluation, we have found that the release of radioactive materials in liqtid effluents will not result in whole body or critical organ doses in excess of 5 ares /yr at or beyond the site boundary, and that radioactive materials released in liquid ef fluents, exclusive of rritium and dissolved gases, will not exceed l 5 Ci/yr/ reactor. We have reviewed the ef fects of reactor operation ,
with one percent of the operating fission product inventory ,in the core being released to the primary coolant. We have determined that under these conditions, the concentrations of radioactive materials in liquid effluents will be a small f raction of the limits given in Appendix B of 10 CFR Part 20.
The design capacities of the BRS and LWTS evaporators are each 21,000 gallons per day (gpd). We calculate the average es.pected waste
11-5 flows to the hts and th LWST to be 2880 and 3430 spd <u pectively,
}
for the combined wastes from both units. The differences between the '
expec'ed flows and design capacities provide adequate reserve for l
processing surge flows. In addition, the design allows wastes to be processed interchangeably between the two systems in the event of equipment downtime. We conclude that the system capacities and designs are adequate for meeting the demands of the plant during anticipated operational occurrences.
The SCBTS is a separate system for each reactor. Each SCBTS consists of two mixed-bed demineralizers, capable of passing up to 65 spa in series flow. The system design also allows for untreated blowdown discharges to be monitored. Automatic closure of an isolation valve terminates releases if' a preset radiation level is reached. We calculate the average expoeted blowdown rate to be approximately 24,000 spd/ unit.
The capacity of each SGBTS is greater than 86,000 spd. We conclude, l I
therefore, that the system capacity and design are adequate for meeting i the demands of the plant during anticipated operational occurrences.
The liquid radwaste systems are located in a Saismic Category I struc ture. The failure of system components would not result in radionuclides concentrations in excess of the limits in 10 CFR Part 20 in the nearest potable water supply or at the nearest surface water supply. The quality group designations of the equipment in the liquid' radwaste systems are listed in Table 11.1 The systems are designed to
11-6 I
preclude the uncontrolled release of radioactive materials due to over-flows from indocr and outdoor tanks by providing level instrumentation !
q which will alarm in the control room, and by means of curbs and Seismic Category i vaultr to collect 'iquid spillage and retain it for process-ing. .ie consider these provisions to be capabic of preventing the uncontrolled release of radioactive materials to the environment.
11.2.3 Liquid Waste System Evaluation Findings The liquid radwaste system includes the equipment and instrumentation te control the release of radioactive materials in liquid ef fluents. The scope of our review included: (1) the capability of the system to reduce
! releases of radioactive materials in liquid effluents to "as low as l practicable" Icvels in accordance with 10 CFR Parts 20 ar.d 50.36a, considering normal operation and anticipated operational occurrences:
(2) the design provisions incorporated to preclude uncontrolled releases I of radioactive materials in liquids due to leakage or overflows in 1
acem *ance with AEC T.eneral Design Criterion No. 60; and (3) the quality group classification and seismic design criteria. The review has included an evaluat-ion of ef fluent releases based on the proposed treat- i ment processes, and was based on information obtained from piping and instrumentation diagrams, schematic diagrams, and descriptive inforina-tion in the FSAR. The basis for acceptance in our review has been conforr.ance of the applicant's designs, design criteria, and design bases for the liquid radwaste system to the Commission's Regulations i.
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l and to applicable Regulatory Guides, as well at to staf f technical 3 l
l positions and industry standards. Bassed on the foregoing tvaluation, we conclude that the proposed liquid radwaste system is acceptable.
11.3 Caseous Waste System j
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I 11.3.1 System Description The gaseous radwaste treatment system is designed to process vastes i
based on the origin of the vastes in. the plant and their expected 1 activity levels. The gaseous waste processing system (GWPS) will process gasen stripped from the primary coolant boron recycle systes gas stripper !
l and miscellaneous tank cover gases by means of continuous recirculation ]
I through pressurized storage tanks (waste gas decay tank system). The GWPS is shared between Units 1 and 2. Radioactive gases from the main I condenser steam air ejector will be released to the plant vent without i l
specialized treatment. Radioactive gases vented from the SGBTS will be routed to the main condenser. When the untreated portion of the steam menerator blowdown system is in operation., as discussed in Section j i
11.2.2 of this report, gases will be vented and the discharges monitored.
l Automatic closure of an isolation valve will terminate discharges if a preset radiation level is reached. Ventilation exhausts f rom the rad-
"este, fuel handling, and safeguards areas will be processed through HZPA (high efficiency particulate air) filters and ch:trcoal adsorbers trior to release. The auxiliary building ventilation system normally exhausts to the plant vent through HEPA filters, but an alternate route
11-8 will permit passage through charcoal adsorbers. In addition, the containment building atmosphere will be recirculated through filters and charcoal adsorbers prior to purging. The turbine building '
ventilation exhausts will be released without treatment. The principal j
components in the CWPS, along with their capacities, are listed in
{
Table 11.2.
11.3.2 System Evaluation In our evaluation of the gaseous radwaste system we have considered the following criteria: (1) the capability of the system to reduce radioactive releases to "as low as practicable" levels based 43 expected gaseous waste inputs and radioactive leakage rates over the life of the l Plant; (2) the capability of the system to maintain releases below the limits in Appendix B of 10 CFR Part 20 (see Table 2,. Column ! for periode i
of fission procuct leakage at design levels from the fuel); (3) the capability of the system to meet the processing demands of the plant during anticipated operational occurrences; (4) the quality group classification and seismic category applied to the system design; and (5) the potential for gaseous releares due to hydrogen explosions.
The process and effluent monitoring design capabilities are considered !
in Section 11.4 of this report.
Our evaluation of the gaseous radwaste treatment system for normal operation is given in the FES for Dicbic Canyon. In the FES, we have determined that the proposed gaseous radwaste treatment systems will be 1
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11-9 capable of reducing the release of radioactive materials in gaseous effluents to approximately 3700 Ci/yr/ reactor of noble gases and 0.28 Ci/yr/ reactor of iodine-131. An isotopic listing of our calculated l gaseous source term is given in Table 3.7 of the FES. Based on that !
evaluation, we have found that the release of radioactive materials
)
in gaseous effluents will not result in an annual air dose, at or beyond the site boundary, in excess of 10 mrad for gamma radiation and 20 mrad for beta radiation; the annual thyroid dose to an individual j will not exceed 15 mres, considering the location of the nearest cow 1
(9.5 miles east of the reactors); and the annual quantity of iodine-131 '
released will not exceed 1 Ci for each reactor at the site. We have reviewed the effects of reactor operation with one percent of the i
operating fission product inventory in the core being released to the primary coolant. We have determined that under these conditions, the concentrations of radioactive materials in gaseous effluents will be a small fraction of the limits given in Appendix B of 10 CFR Part 20.
Operating with three 40 scfm compressors (one unit in continuous use and the others as backup), and six 705 f t3gas decay tanks (each of which is capable of being isolated from all the others), the system has adequate capacity to allow operation during periods of equipment downtime. We conclude that the system capacity and design are adequate 1
for meeting the demands of the plant during both normal operation and' l anticipated operational occurrences.
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Most of the gas entering the CVPS during normel operations will be f cover gas displaced from the baron recycle holdup tanks as they fill l
with liquid. This gas will consist primarily of nitrogen and hydrogen. 4 To prevent oxygen buildup in the system, the vent header is designed to f In addition, gas sampics will operate at a slightly positive pressure. i J
be periodically drawn from the tanks discharging to the vent header, and l
from the decay tanks being filled. An alarm will be activated if the In oxygen content of any sample exceeds two percent by volume (v/o).
this manner, the potential for explosive hydrogen / oxygen mixtures vill be mitigated.
The Caseous wastes from the main condenser will not be treated. !
system releases will be proportional to the rate of primary to secondary system leakage and the priary coolant activity. In the event of 1
excessive primary to secondary leakage, the affected steam generator (s) will be isolated before radioactive material concentrations in main condenser of f gas releases exceed the limits in the technical specifications.
The plant ventilation systeu are designed to induce air flows from potentially less radioactively contaminated areas to areas having a greater potential for radioactive contamination. The ventilation system has adequate capacity to limit radioactive material concentrations in areas within the plant t. hat are accessible during operation to below the limits in 10 CFR Part 20.
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I 11.3.3 Caseous Waste System Evaluation Findings i
I The gaseous radwaste system includes the equipment and instruments-tion to cot. trol the release of radioactive raterials in gaseous effluents. The scope of our review included: (1) the capability of the system to reduce releases of radioactive materials in gaseous effluents to "as low as practicable" levels in accordance with 10 CFR Part6 20 and 50.36a, considering normal operation and anticipated operational occurrences; (2) the design provisions incorporated to reduce the potential for hydrogen explosions; and (3) the quality group classification and seismic de: sign criteria. The review has included an evaluation of effluent releases based on the proposed treatment processes, and hos considered pathways due to process vents and leak, age affecting building ventilation systems.- The review was also bnsed on information- obtained f rom piping and instrumentation diagrams, schematic diagrams, and descriptive information in the PSAR. The basis for acceptance in our review has been conformance of the applicant's designs, design criteria, and design bases for the gaseous radwaste system to the Commission's Regulations and to applicable Regulatory Guides, as well as to staff technical positions and industry standards. Based on the foregoing evaluation, we conclude that the proposed gaseous radwaste system is acceptable.
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11.4 Process and Ef fluent Radiological Monitoring System 11.4.1 System Description The process and effluent radiological monitoring system is designed i to provide information concerning radioactivity levels in systems throughout the plant, indicate radioactive leakage between systems, monitor equipment performance, and monitor and control radioactivity levels in plant discharges to the environs. Scintillation detectors will be used for particulate monitoring in gaseous effluents. Geiger-Mueller detectors will be used for monitoring liquids and for monitor-ing radioactive gases in vent effluents. Caseous iodine will be collected on replaceable, impregnated charcoal adsorbers which will be continuously monitored by scintillation detectors while in use. Systeme which are not amenable to continuous monitoring, or for which detailed isotopic analyses are required, will be periodically sampled and analyzed in the plant laboratory.
Table 11.3 indicates the proposed locations and types of continuous monitors. knitors on ef fluent release lines will automatically terminate -
discharges should radiation levels exceed a predetermined value.
11.4,2 System Evaluation in our evaluation of the process and ef fluent monitoring system we i
have considered the capability of the system to: (1) monitor all normal and potr tial pathways for release of radioactive materials to l
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11-13 I the environment; (2) control the release of radioactive materials to the j i
environment; and (3) monitor the performance of process equipment and )
detect radioactive material leakage between systems.
We have reviewed the locations and types of ef fluent and process monitoring provided. Based on the plant design and on the continuous monitoring and intermittent sampling locations, we have concluded that all normal and potential release pathways, excluding the turbine building vent, will be monitored. Due to the high potential for exfiltration from the turbine building, which is a relatively open structure, it is not practical to monitor the potential gaseous releases from the turbine building. The ciesign includes provisions for automatically terminating effluent releases in the event radiation levels in discharge lines exceed a predetermined level. We have also determined that the sampling and monitoring provisions are adequate for detectir.g radioactive material leakage. to normally uncontaminated systems, and for monitoring plant q 1
processes which affect radioactivity releases. On this basis, we consider that the monitoring and sampling provisions meet the require-ments of AEC General Design Criteria Nos.13, 60 and 64 and the guide-lines set forth in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents fros Light-Water-Cooled Nuclear Power Plants."
_ _ - - _ _ - - _ - _ _ - _ . _ _ _ _ _ _ . _ - _ _ - - _ _ _ _ . _ _ - _ _ _ - _ - _ _ _ _ - _ _ - - - _ . - . _ . - _ _ _ . - _ . _ . _ . . _ . . - - - _ - _ . - - _._.__.--__b
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11.4.3 Process and Ef fluent Radiological Monitoring Evaluation Findings
]
i The provisions for process and effluent radiological monitoring include the instrumentation and controls for monitoring and controlling i j
the releases of radioactive materials'in plant effluents, and the I
monitoring of the level of radioactivity in process streams. . The scope .. l of our review included the provisions for monitoring and controlli.ng the release of radioactive materials.in plant effluents in accordance i with AEC General Design Criteria Nos. 60 and 64 and ' Regulatory Guide 1.21, and for monitoring radioactivity levels within the plant in process streams in accordance with AEC Ceneral Design Criterion No.13. The basis for acceptance in our review has been conformance of the applicant's design, design criteria, and design bases for the process and effluent ,.
-l monitoring systems to the Connaission's Regulations, as set forth in the General Design Criteria, and to the applicable Regulatory Guide, as well as to staff technical positions and industry standards. Based on the foregoing evaluation, we conclude that the proposed provisions for monitoring process and ef fluent streams are acceptable.
11.5 Solid Waste System ;
11.5.1 System Description The solid radwaste treatment system is designed to collect-and Process vastes based on their physical form and need for solidification Prior to packaging. " Wet" solid wastes, consf sting of spent demineralized resins, evaporator bottoms, filter sludges, and chemical drain tank
11-15 effluents, will be combined with a cement-vermiculite mixture to form a solid matrix and sealed in 55-gallon drums. Dry solid wastes, consisting of ventilation air filters, contaminated clothing and paper, and miscellaneous items such as tools and glassware, will be compacted ;
into 55-gallon drums using an industrial baling machine. The solid I
waste system is shared between Units 1 and 2.
11.5.2 System Evaluation In' aur evaluation of the solid radvaste treatment system we have i
considered: (1) the system design objectives in tems of expected types, volumes, and activities of wastes processed for shipment offsite; (2) the design capacities of system components, methad c,f l
operation, and capability of meeting the demands of the plant due to anticipated operational occurrences; (3) waste packaging and conformance to applicable Federal packaging regulations; (4) provisions for controlling potentially radioactive airborne dasts during baling operations; (5) ic design and quality group classification; and (6) provisions for snsite storage prior to shipping.
Our evaluat. ion of the solid radwaste treatment system fer normal operation is given in the FES for Diahlo Car"on. In the FES ve determined that the expected so1.id waste volumes and activities, shipped offsite frem each unit will be 250 drums /yr of " wet" solid weste containing an average of 20 C1/ drum, and 500 drums /yr of "dr""
solid waste containing less than .i Ci total.
)
11-16 Drum filling operations will be controlled remotely from consoles located outside the drum fill area. Drumming operations will have interlock features to prevent opening of filling valves when a drum In addition, the is not properly positioned in the filling station.
equipsent is designed so that any spills will be collected in a drain pan and prevented from dripping on the floor. Baling of dry Wastes will wastes will be carried out inside a closed dust shroud. I t
be packaged in 55-gallon steel drums that meet DOT requirements, and will be shipped to a licensed burial site in accordance with AEC and DOT regulations.
Storage f acilities for up to 2000 drums of solid radioactive wastes are provided in the fuel handling area of Unit 2, plus a solid radwaste storage area for 28 drums (high level), 300 drums (intermediate level) and 300 boxes (low level) of packaged solid radioactive wastes. Based on our estimate of 750 drums /yr/ reactor, we find the storage capacity adequate for meeting the demands of the s.1an t.
The spent derineralizer resin storage tank is designed as Quality Group C. Waste transfer piping is designed to Quality Cro p D and non-Seismic Category I requirements. Since the quantity of radioactive f 1
materials in the piping vill not have a significant potential for uncontrolled release to the euvirons, we consider this design to be acceptable.
i l
11-17 11.5.3 Solid Waste System Evaluation Findings The solid radwaste Lysten includes the equipment and instrumentation for solidifying and packaging radioactive vastes prior to shipment offsite for burial. The scope of our review included: (1) capability of the system f or processing the types and volumes of wastce expected during normal operation and anticipated operational occurrences, in accordance with AEC General Design Criterion No. 60; (2) the provisions for handling wastes with regard to the requirements of 10 CFR Parts 20 and 71, and 49 CFR Parts 170-178; and (3) the quality group classification and seismic design criteria. The review has included the provisions for onsite storage and for the centrol of airborne dusts during dry waste compaction, and was based on information obtained from piping and instrumentation diagrans, schematic diagrams, and descriptive information in the PSAR. The basis for acceptance in our review has been conformance of the applicant's designs, design criteria, and design bases for the solid radwaste system to the Commission's Regulations and to applicable Regulatory Guides. as well as to staff technical positions and industry standards. Based on the foregoing evaluation, we conclude that the proposed soil.' radwaste system is acceptable.
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'i 4 Table 11.1 Design Parameters Of Principal Cceponents Considered In Liquid _Radwaste Evaluation Qualirr Capacity, ea._ Croup System Component Numbe_r _
Recycle Holdup Tanks 5 83,200 gal. C BRS 4 30 gpm C BRS Evaporator Feed Demin.
15 gpm C Evaporator 1 BRS 2 30 gym C BRS Evap. Cond. Decin.
100] gal. D LVTS Chemical Drain Tank 1 Regenerant Receiver ranks 2 15,000 gal. D Lk'TS Lk7S Equipment Drain Receiver 2 15,000 gal. D 1anks Lk'IS Aux, Bldg. Sump 1 3600 gal. D ,
)
Lkis Floor Drain Receiver 2 15,000 gal. D !
j Tanks 1 Misc. Equip. Jrain Tank 1 2700 gal. l IXTS 2 15,000 gal.
LkTS Waste Condensate Tat.ks Waste Evaporator 1 15 gpm l LhTS l 50 eptr, C j LkTS Mixed Bed Demin. 1 i
IXTS Laundry and Hot Shower Analysis Tank 2 1000 gal. D 2030 gal. D LLTA Waste Cor. Holding Tank 1 65 gpm D 50618 Mixed-Bed Dettineralizers 2 SCBTS Steam Cencrator Flash D 2 200 gal.
Tank i:BTS Steam Generator B1m.down l Tank 2 4160 gal. D
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Spent Resin 5torare Tank 2 200 1t C i
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TABLE 11.2 Design Parameters Of Principal Components Considered In Gaseous Radwaste Evaluation
' System Quality Component Number Capacity. Ea. Group
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I WGPS' Compressor 3 40 scfm C '
WGPS Decay Tanks 6 705 ft 3' C' .
WGPS Surge Tanks 2 14 ft C I l
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[
e 11-20 Table 11.3 Process And Effluent Monitoring Steam Detector Monitored j Type -
Containment Purge exhaust (particitlate and iodine) y Scintillation (gas) Geiger-Mueller (G-M)
Plant Vent (particulate) (2 channels) y Scintillation (gas) (2 channels) G-M l Residu; Best Removal Exh. (particulate) y Scintillation Liquid Ra3 waste Release y Scintillat.cc.
Gas Decay Tank Discharge (gas) G-M Control Room Air (particulate) y Scintillation Equiyent Drain Receiver Recirculation y Scintillation Condenser Air Ejector (gas) G-M Component Cooling Heat Exchranger Outlet (2 :hannels) y Scintillation 1 Steam Ge:erator - secondary side liquid phase _y Scintillation 5
Steam Generator Blowdown Tank Vent (gas) G-M 1
12-1 12.0 RADIATION PROTECTION 12.1 Shielding Shielding of a nuclear power facility for radiological protection during normal operation has two objectives: (1)'to ensure that radiation limits to operating personnel and the general public, as set forth in 10 C51 Parts 20 and 50, are upheld; and (2) to ensure that radiation ex-posures to operating personnel during refueling, maintenance and inspec- l I
tions are maintained as low as practicable (ALAP), based on applicable i provisions of Regulctory Guide 8.8, "Information Relevant 'to Maintaining Occupational Radiation Exposure as Low as Practicable [ Nuclear Reactors)."
We have reviewed the Diablo Canyon shielding design to determine if the design fulfills the generic objectives described above.
Radiation exposures at Diablo Canyon are intended to be maintained ALAP by classifying all plant areas into radiation zones based on expected frequency and duration of occupancy. Thus, the design of the radiation shield was carried out with consideration of the dose rate criterion for each zone. Shielding analysis was made with accepted computer codes. Consistent with the design, the applicant addressed the -
i steps he took to assure that low dose rate zones were not likely to be compromised by inadvertent increases in radiation levels. Hence, pipes i
carrying radioactive liquids including piping penetrations, ducts, reach ;
rods, tube withdrawal spaces, etc. were designed to be located in a
properly shielded compartments. When the tanks within compartments f
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f 12-2 contain significant quantities of radioactivity, they are shielded from each uther. Thus, each component or tank within a compartment is '
isolated to allow for maintenance, inspection and some non-routine opera-tions with ALAP radiation interferences from other components (or tanks).
Movable shiciding and means for its utilization will be available for use where permanent shielding is impractical.
Based on the above design and operating philosophy of the applicant, we conclude that adequate consideration has been given to the shielding of facilities and components to keep exposures to operating personnel and the public within the applicable limits of 10 CFR Part 20, and to reduce unnecessary exposure during normal operation of the plant, as described by Regulatory Guide 8.8. The effectiveness of the shield will, however, be evaluated by means of radiction surveys during initial startup and full power operation.
12.2 Ventilation The ventilation r ystems for the Diablo Canyon Units are designed to provide a suitable radiological environment for personnel and equipment.
The path of the ventilation air is f rom areas of low radioactivity toward areas of higher activity to ensure contamination control. Various com- :
partments throughout tha plant are provided with roughing and high {
ef ficiency particulate air (HEPA) filter banks, with charcoal filters added at selected locations, to preclude a buildup of airborne contami-nation. The ventilation system is also desigt.ed to limit doses at the !
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site boundary to within as low as practicable guidelines of proposed ,
f)
Appendix I of 10 CFR Part 50. The design criteria of the systems, the t x
description of the systems, and the operating procedures discussed in .)
4 the FSAR provide reasonable assurance that adequate consideration has been given to ventilation design for protection of in-plant i I
personnel from airborne radioactivity har.ards. 4
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))
physics program are to limit radiation expopures to personnel to as low as practicable, and to comply with the appropriate reguistions ir 10 CFR j Part 20. Educating the individual on radiation control standards and i]
.? i e
procedures to ensure achievement of these objectives is the responsibility of the health physics personnel at-the Diablo Canyon Plant. ]
1{
Company policy for radiation protection at the Diablo Canyon Units is 4J El ce.rried out based on appropriate AEC regulations. Consistent with this I policy, programs and procedures will be adapced which are consistent with ;
Regulatory Guide 8.8. Among these are personnel dosimetry by film badges; f, respiratory protection including a respiratory fitting program; personnel 1 '
.* j 5'
protective clothing and personnel decontamination procedures; signs, tags, ropes and other access control measures to preclude unauthorized entry into high radiation areas; special work permits and procedures; testing {
and calibrating monitoring instrumentation; ar.d r intenance of radio- jl I
logical reports and records. j 4
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Monitoring instrumentation and counting room equipment will be 4 operated by the health physics staf f. Portable radiological survey in-f strument inventory is satisf actory and of state-of-the art quality. ,
Self-reading dosimeters are maintained by this group for recording daily a.xposures on exposure estimate cards. A routine bioassay program consisting of urinalysis and whole body counting will be performed by an
)
outside contractor, and will provide supporting data on the effective-3 ness of the air monitoring program.
We conclude that the applicant's program for inplant radiation safety, as reficcted by the health physics progras and the concern of management for radiation control standards, as stated in the FSAR, is adequate to x
limit occupational exposures to wichin the Ilmits set forth in 10 CFR ;
/
Part 20 and Regulatory Guide 8.8. ;
i, 12.4 Radioactive Materials Safety i.
The personnel qualifications, facilities, equipment and procedures f I
for handling the byproduct, source and special nuclear material sources ;
utilized for reactor startup and equipment calibration were reviewed.
Based on the information provided in the FSAR and amendments, we con- ;
s, clude that there is reasonable assurance that these sources will be f7 stored, handled, end used in a manner to meet the applicable radiation j t'
protection provisions of 10 CFR Parts 20 and 30. l 12.5 Area MonitorinE 1 The radiological monitoring system is designed to continuously 4
12-5 i
measure the radiation levels at fif teen selected locations within the plant. Each instrument of the systen will have a local alarm and readout j feature at the fixed location and in the control room. Dose rate Icvels I will be recorded in the control room.
Airborne radioactivity monitoring will be performd by eleven fixed gas and particulate am*ttors located throughout the plant. These are supplemented by particulate and iodine-111 mobile continuous air monitors (CAM's) which will be used in specific locations, such as the fuel handling and radwaste areas, during operations that may cause air-I borne radioactivity. A routine grab sampling program will also be main- ,{
tained as part of L.e air monitoring program.
We conclude that the area monitoring program, with its concomitant .
instrument calibration rechniques, will provide satisf actory radiological j
l protection to in-plant personnel. ;
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13.O CONDUCT OF OPERATIONS
'13.1- Organization and Qualifications The plant staf f proposed by the applicant for one unit operation consists of 62 full-time employees, In addition to security personnel.
The plant staff functions in the following groups: operations, main-tenance and technical support The staff will expand to 80 employees on-site for two unit operation. The plant superintendent is responsible for all onsite activities in connection with the safe operation and maintenance of the plant. He reports to the manager of steam generation, who in turn reports to the vice president - electric operations.
The supervisor of operations directs day-to-day operation of t.ie plant and is responsible to the plant superintendent for the operation of the facility. Normal shif t operation is under' the direc't control of the ahif t foreman who is directly responsible te the supervisor of operations. The normal snif t complement for single unit operation will consist of six employees, including a licensed senior reactor operator and three licensed reactor operators. For two unit operation, the i
( l applicant originally proposed a crew of eight, including one senior reactor operator. Such a shif t crew is not in accordance with the Regulatory staff position which etates that the number of licensed senior l
reactor operators onsite should not be less than the number of reactors that are operating. The applicant has now proposed that the crew staff during Unit 2 startup will number nine, and duing comunercial operation i
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13-2 l
of both units, eight, including two licensed senior reactor operators and three licensed reactor operators. These shift crew compositions are in accord with the established staff position.
Normal plant maintenance is accomplished by a maintenance staf f of 24, under the direction of the supervisor of maintenance. Maintenance personnel from the applicant's other thermal plants may be used to supple-ment plant forces as necessary for major repairs. The technical support section will have 14 personnel under the power plant engineer, who is responsible to the plant superintendent. The technical sut pc provides services in the areas of reactor engineering, chemistry. 3J 0--
chemistry, radiation protection, reactor plant performance, and instrument and control systems. Qualifications for plant employees will meet the criteria set forth in ANSI N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel." The key managenent, supervisory and technical cositions in the plant will initially be filled by employees whose s.cperience in the nuclear power field ranges from eight to sixteen years.
The manager of steam generation functions as the engineer-in-charge l l
(as defined in ANSI N18.1) of technical support activities for the plant staff. He has a technical staf f of 14 in addition to the resoarces of technical specialists from other company departments. All of the engineering and scientific disciplines necessary to provide operatior.nl technical support are available within the applicant's organization so that PG6E does not plan for outside consultants to have assigned
T 13-3 responsibility for any specihc area of technical support. However.
consultants will be usad to usist on special problems. -l 13.2 Training Program o _ i i
The applicant has arranred a training program for operating personnel which was tailored to meat its needs of each man.with respect to his previous background and job responsibilities. the progran included a series of lectures provided by the nuclea steam-system supplier q
covering the function, design description, control and instrumentation, normal and abnormal operation, and maintenance of the pressurized water reactor system of the Diablo Canyon reactors. Key individuals in the initial plant organization were assi ned b to an opriating PWR to observe and participate in operations, with the length of assignment lasting about a month. Candidates for cold AEC licenses will a*. tend a two-week train-ing program at the reactor simulator of the nuclear eteam supplier.
Selected members of the plan: technical support groupt. have completed fermal training specifically criented to their assigned responsibilities.
The applicant has described his proposed operator requalification program which will be placed into effect within three months after issun.ce i
of the Unit 1 operating license. The program includes a lecture series; on-the-jcb training with p? ant centrol manipt.lation; review cf all design, procedure and license changes; semi-annu:1 review of all abnormal and emergency procedures, and both written and oral examinations. We have re-viewed the information submitted by the applicant, and have concluded that certain revisions must be made in the opera Mr requalifiution program in oroer to meet the requirements of Section 50.54(1-1) of 10 CFR Part 50 1
13-4 and Appendix A of 10 CFR Part 55. Resolution of this item will be l
discussed in a supplement to this Safety Evaluation Report. -!
Subject to f avorable resolution of the operator requalification l program, we conclude that the proposed organization, .he training and retraining, sad the qualifications of the Diablo Canyon- staff are adequate f
to provide acceptable staff and technical support for safe operation o I the plant.
13.3 Emergency Planning The emergency plan submitted by the applicant for Diablo Canyon Units 1 and 2 includes California's radiological emergency plan of the Bureau of Radiologichi Health of the Department of Public Health, dated !
j January 1971. The plan also includes the San Luis Obispo County Sherif f's I l
Department interim evacuation plan, dared June 1974, which contains l specif f e instructions covering evccuation of people in the environs of the Diablo Canyon Plant.
In the event of an emergency, the normal operating crew is qualified 7
to and responsible for making an initial evaluation of the incident, performing nny immediate operations which are necessary, and placing l
The seniv appropriate Partions of the emergency plan into ef fect.
operating crew essumes the position of member cf the normal shif t emergency coordinator,u. til a senior member of plant staf f crrives. :
J The energency plan includes provisions for primary and alternate .
erergency control centers, notification of offsite state and federal agencies with responsibilities during an emergency, an emergency com-munications network , a description of onsite first aid and decontamination s
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13-5 f acilities, and provisions for emergency transportation. The Sierra Vista llospital in San Luis Obispo has agreed to accept and treat contaminated patients. The plcn describes a spectrum of accidents with specific action levels for protective measures. The plan provides for annual training drills including checks of communications with local !
I age,icies. j We have reviewed the emer8cncy planning program for the Diab 3o 1
Canyon Plant, and we find that the program conforms with . Appendix ;; of :{
10 CFR Part 50, and is acceptable.
13.4 Safety Review and Audit i
The safety review and audit functions for the Diab?o Canyon Plant will be performed by three separate and independent groups - the Plant Scaf f Review Committee, the General Of fice Nuclear Plant F-"'-* and Audit Comittee, and the President's Nuclear Advisory Comit s s .a Plant Staff Review Comittee is composed of senior plant staf' personnel. It ,
I is advisory to the plant superintendent and will review all proposed tests and design modifications, and operating, maintenance and test procedures and changes thereto involving safety related aspects of safety related equipment. The General Office Nuclear Plant Review and Audit Comittee is composed of individuals without line responsibility for the reactors. It provides reviews of audits, actions and practices bearing on nuclear safety, and initiates audits independent of the personnel directly responsible for the activity being audited. The President's Nuclear Advisory Committee is composed of four senior l
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managers in the PG&E organization. Its primary function is to examine i and report to the president on the activities of the General Office I
Nuclear Plant Review and Audit Committee. The charters of both indepen-dent review and audit committees have been designed to confone to the guidance in ANSI N18.7-1972, " Standard for Administrative Controls for Nuclear Power Plants."
We conclude that the provisions for review and audit of plant opera-tions are acceptable. -
13.5 Plant Procedures All safety related operations are to be performed in accordance with written and approved operating and emergency procedures. These l
procedures are incorporated into a Diablo Canyon Plant Manoal, and include administrative, operating, emergency, maintenance, surveillance and test, radiation protection, chemical and radiochemical, and security procedures. The procedures confona to the guidance in Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)," and are based on the requirements and recommendations of ANSI N18.7-1972.
We conclude that the provisions for preparation, review, appreval, and use of written procedures are satisfactory. 1 13.6 Industrial Security The applica. t's initial proposal for industrial security for the Diablo Canyon Nuclear Plant was considered by the staff to be deficient in certain areas. Revisions to the security plan were submitted and J
o.
13-7 reviewed by the staff.
In particular, the applicant has agreed to the arming of plant security guards. The security plan is now in accordance with the positions of Regulatory Guide 1.17, " Protection of Nuclear Power Plants Against Industrial Sabotage," and ANSI N18.7-1973, " Industrial Security for Nuclear Power Plants," and conforms to the requirements of 10 CFR Part 50, Section 34(c) and .10 CFR Part 73. Section 40. The plan was submitted biit withheld from public disclosure as provided for in Section 2.790 of 10 CFR Part 2.
We conclude that the industrial security program provides reason-able assurance that the risks associated with potential acts of sabotage that would lead to a significant threat to the public health and. safety are' acceptably low.
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14-1 14.0 INITIAL TESTS AND OPERATIO)]
Preoperational teoting, fuel loading, initial criticality and approach to full power operation will be performed by the station operating personnel under the direct control of PC6E. A PC&E resident startup enFineer vill direct th< pror, ram and will have sign-off respon-sibility, with the concurrence of the plant superintendent, for approval of test procedures and f or evaluation of the completed test results.
j The Nuclear Stear ' system Supplier (Westinghouse) will furnish technical l
l advice during the startup program. During hot functional testing.
Westinghouse wil! provide a reactor coolant pump specialist, a ch mist, and a quality assurance specialist for Internals, inspection. DurinE .
s fuel loading, a physicist, a chemist, and a fuel handling specialist will be on hand. During startup tests to power, a nuclear test engineer, a L
chemis t , a tr msient analyst, and a reactivity computer instrumentatten !
specialist are scheduled to be onsite.
The appli: ant has listed the t ests to be pei f orened .os .'ng the test program, "ith a statement of the objective of t ach test . We have .
s f
i reviewed the program and conclude that it conform to tia positions stated )
in Pegulatory Guidt 1. 68. "Preoperational and ! nit ial st ar tup Ter,e inu-We conclude that L7e pregram j grams for Water Cooled Power Keat tors."
l describ(d by the applicant will provide ar adequate basis to confirrr 4
the safe apr:ation of the { lant, a"61 i G detTptN JC.
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15-1 15.0 A{CIDENT ANALYSES 15.1 C_eneral The staff and the applicant have evaluated the offsite radiological consequences for a number of postulated design basis accidents. These accidents are the same as those analyzed for previously licensed PWR plants and include a loss-of-cool. nt accider.t (including leakage from ESF components outside of containment), fuel-handling accident, a hydrogen purge dose (post-LOCA) accident, rod ejection accident, and rupture of a radion:tive gas storage tank in the gaseous radioactive waste treatment system. All accidents have been evaluated at a core power level or 3423 HWt, with the exception of such accidents as the LOCA and those analyses for which adequacy of the containment and engineered safety features must be demonstrated. These accidents were evaluated at a core power level of 3580 MWt.
The of f site doses calculated by the staf f 1 r these accidents are t presented in Table 15.1 ef this report, and the assumptions used in these calculations are listed in Section 15.2. All pot en*.4. ' affsite doses cciculated by the applicant and the staff for the postulated accidents l are within the guideline values of 10 CFR Part 100.
As part of the loss-of-coolant accident (LOCA), the staff and the applicant have also evaluated the consequences of leakage of containment sump water containing radioactive fission products which is circulated by the RHR system outside the containment af ter a postulated LOCA. We have assumed 1
4
l 15-3 the sump water to contain a mixture of iodine fission products in agree- 1 ment with' Regulatory Guide 1.7. About 0.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a postulated LOCA l 1
this water is pumped into the auxiliary building to be cooled by the f 1
residual heat renoval, heat exchsinger. If a source of leakage should develop, such as a failure of the RHR pump seal, some water would Icak into the auxiliary building. A portion of the iodine would become gaseous and would exit to the outside atmosphere af ter passhg through the charcoal filters in the auxiliary building. The offsite doses aesulting f rom such a sequence of events depends upon the tem-perature and magnitude of the assumed leakage. If the leakage occurred
)
when the water temperature was below 212*F, a leak rate of less than 10 spa over a period of one-half hour would result in doses (without l
f11 tere) which could exceed the guideline values of 10 CFR Part 100.
If the leakage occurred when the fluid is near its peak temperature of 248'F, then part of the leaking water would flaah to steam, leading to additional iodine release. In this case, less than 1 gpm leakar ,' .
I for one-half hour would result in doses (without filters) which could l I
exceed Part 100 guidelines. j If the charcoal filters in the auxiliar, building, are assumed to be ef f ective in removing lodine, the of f aite doses would be within the guidelines of 10 CPR Part 100, even for substantial anounts of laalage.
A. a result of the analyses discussed above, we will recuire that the auxiliary building filters conform to the requiremntn of engineered safety features (l'cf) to the extent that electric heatert, for humidity control vill be required, and to the extent that the failure of any i
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15 ' i 4
single active component can be tolerated. The staff has agreed that elements of the f11ter system such as the charcoal beds, need not be redundant. The appifcant has been informed of our concern re-garding potential RHR leakage, and is expected to propose appropriate design modifications. Resolution of this ites will be reported on in l a supplement to this Safety Evaluation Report. j On the basis of our experic::,ce with the evaluation of the steam line break and the steam generator tube rupture accidents for PWR plants of similar design, we have concluded that the consequen:cs of these accidents j can be controlled by limiting the permissible primary and secondary coolant system radioactivity cox entrations so that potential offsite i doses are small. We will include appropriate limits in the technical !
i specifications on primary and secondary coolant activity concentrations. ]
1' Similarly, we will include appropriate limits in the technical specifications on gas decay tank activity so that a single failure (such as sticking ani lif ting of a relief valve) does not result in doses that are more than a small f raction of the 10 CFR Part 100 ; uide- l 1
line values.
15.2 Des., Basis Accident Assumptions L5.2.1 Lons-of-Coolant Accident (Containment leakage)
The assumptions used by the Regulatory staf f in calculations of off site doses from a LOCA were:
- 1. Power level of 3580 MWt.
_____.______a
15-4 ,
4
- 2. Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors," Revision 1. June 1973.
- 3. Design containment leak rate of 0.10% per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.05% per day thereafter.
- 4. lodine removal by the containment quench spray system utill ed the following parameters:
6 3 Primary Containment Vo'une 2.67 x 10 ft Spray Fall Height 128 ieet i
2600 spe Spray Flow Rate Elemental Mass Transfer Velocity 4.70 cm/sec Organic Mass Transfer Velocity 0.0 cm/see Spray Drop Diameter 1500 microna Spray Terminal Velocity 480 cm/sec Fraction of Primary Containment 17%
Spray Reduction Limits:
bJ Elemental Organic 1.0 1000 Particulate Spray Removal Rates (for sprayed region only)
Elemental 10.0 hrs"
~I Organic 0.0 hr
-I Particulate 0.45 hr l
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- 5. Crouno level release with Pasquill Type "P' conditions and wind speed of 1.0 meter per second for short-term releases.
Our evaluation of the iodine removal effectiveness of the containment sprays is discussed further in Section 6.2.3 of this report. 1 1
1 15.2.2 Fuel Handling Accident The assumptions used by the Regulatory staf f to calculate offsite doses from a fuel handling accident were: .
- 1. Rupture of all fuel rods in one assembly.
f the rods, assumed to be 10% of the noble gases
- 2. All gap activity and 10% of the iodine (with a peaking factor of 1.65), is relassed. ,
- 3. The accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af ter shutdown. 1 4 99% of the iodine is retained in the pool water.
- 5. Iodine above the pool is 75% inorganic and 25% organic species.
- 6. Standard ground release meteorology and dose conversion f actors. .
1
- 7. lodine removal factors of 90 and 70% for the charcoal filters for elemental and organic todines, respectively.
15.2.3 Cas Decsy Tank Rupture !
The assumptions used by the Regulatory staff to calculate the offaite doses from a gas decay tank rupture were: i
- 1. Cas decay tank contains one complete primary coof, ant loop inventory of noble gases resulting from operation with 1% failed fuel (75,000 curies of noble gases).
- 2. Standard ground level release meteorology and dose conversion f actors.
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15-6 i l
l 15.2.4 control Rod Ejection Accident.
l The assumptions used by the Regulatory staf f to calculate of fsite l doses from a control rod ejection accident were:
Case 1 ;
l
- 1. Power Lev el of 3580 Nt. l l
- 2. 10% fuel failed in transient. l i
- 3. 10% of iodinc and noble gas inventory in gap of failed fuel. l 4 Release of total gap activity in failed fuel to containment building.
- 5. 50% plate-out of radioactive iodinen.
- 6. Containmen' building sprays are not initiated.
- 7. Containment building leak rate of 0.10% per day for 24 h< ars and one-half of this value thereafter.
- 8. Standard ground level release meteorology and dose conversion f actors.
Case II
- 1. Power level of 3580 Nt.
- 2. 10% fuel failed in transient.
- 3. 10% of iodine and noble gas activity in gap of failed fuel.
- 4. Release of total gap activity in failed fuel to primary coolant.
- 5. Primary to se:ondary coolant operational leakage is 1.0 gpz .
I
- 6. Loss of of fsate power so that steam is released f rom secondary side ;
relie. valve. j
- 7. Primary-secondary coolant equilibrium reached at 30 minutes af ter the accident.
1
- 8. Standard steam line release meteorology and dose conversion factors. l l
L____________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
j
15-7 15.2.5 Hydrogen Purge Dose 1 l
The assumptions used by tha Regulatory staff to calculate the exclusion boundary doses due to post-loss-of-coolant accident hydrogen purging were:
1
- 1. Power level of 3580 Mwt .- !
6 3
- 2. Containment Volume of 2.67 x 10 ft
- 3. Purge Rate: 300 cfm (duration varies f acs 2 hrs per day at 28 days to I hr per day at 100 days).
- 4. Holdup Time Prior to Purging: 28 days. )
- 5. Sodium Hydroxide Spray Reduction Factor for Iodine: 20.2
- 6. Charcoal filter efficiencies of 90 and 70% for elemental and organic i
iodine, respectively. i
- 7. X/Q Value: 4-30 days (1.75 x 10-5 sec/m3 ).
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15-8 I
TABLE 15.1 l
l Potential Of f site Doses Due To Design Easis Accidents Two Hour Course of Accident Exclusion Boundary Low Population Zone
' (800 Meters) (9600 Meters)
Whole Body Thyroid Whole Body Thyroid (Rem) (Rem) (Rem)
Accident (Rem) 151, 7.5 21. <1 Losu-of-Coolant *
(Containment leakage)
Post-LOCA Hydrogen 21.4 0.96 Purge Dose 2.6 <1 <1 Fumi Handling 23, Cas Decay Tank Negligible 2.6 Negligible <1 Rupture Rod Ejection **
Case 1 19 <1 3.8 <1 1 3 <1 Case II 19 1
- The iodine fractions assumed to be released in the postulated loss-of-coolant accident. are: elemental, 91%; organic, 4%; and particulate, 5%.
- See Section 15.2.4 of ttis report for the dif ferent assumptions used for Cases I and 11.
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16-1 i
16.0 TECINICAL SPECIFICATIONS j The technical specifications in a license define certain featt es, characteristics, and conditions governing operation of a facility that cannot be changed without prior approval of the AEC. We review'd the proposed technical specificaticas in detail and hr.se held a n; .,er of meetings with the applicant to discuss their contents. Modifications to the proposed technical specifications submitted by the applicant were made to describe more clearly the allowed conditions for plant operation. The finally approved technical specificatir ns will br made part of the operating licenses. Included w#11 be sections coveri z safety limits and limiting safety syste:n settings, limiting conditions for operation, surveillan e re ,uiremer.ss, design features, and admin-1strative controls. On the basis of our review, we ha,e conclu.ied that normal plant operation within the limits of the technical specifi-cations will not result in potential offsite exposures in excess of the 1
10 CFR Part 20 limits. Furthermore, the limiting conditions for operatiet.
]
and surveillance requirements will assure that necessary engineered safety features will be available in the event of malfunctions withtn the plant.
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17-1 i 17.0 QUALITY ASSURANCE 17.1 General j
1he description of the Quality Assurance (QA) Program for the f operations phase of the Diablo Canyon Nuclear Power Plant, Units 1 and 2, is contained in Section 17.2 of the FSAR, as. amended. Our evaluation of the QA Program for the operations phase is bcced on a review of this description, and on detailed discussions conducted with the applicant to determine if Pacific Las and Electric Company's (PG6E's)
QA Program for Diabro Canyon complies with the requirements of Appendix l
B of 10 CFR Part 50. l Our review of the applicant's QA Program included:
(1) A detailed evaluation of the QA Program contained in Section 17.2 of the FS AR, as amended.
(2) A meeting and d:scussions with PCLE representatives which resulted {
in revisions to the prograt description; these revisions were l submitted in Ar.endments 4 thru IC of the USAR.
(T) Discussions wit' tiie Directorate cf Regulatory Orcrations, Region V, relating to ti t'.r review of PC6E's OA acticities, and the implement i-
)
tion cf n QA I re;ran which cora[.12es with Appendix B ot' 10 CFR Part 50.
J[ggn i zat ion l 1 he P0;f'. org<tniza tion f or Tic l u-r power generat ion, as it t elates te l
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17-2 presents the company-wide organization and the Nuclear Mant Review and Aud it Committee.
The President and chief operating of ficer has the responsibility for the overall management of the plant. The Senior Vice President (Engineering and Construction) provides technica: support to the Elect ric Operat ionc Department for plant modifications and other related activities, ard is responsible for the development of the Quality Assurancy Program. The Quality Assurance Prograr: .is directed and managed by the Director of the Quality Assurance Department.
l 1
i A group of three committees is also assigned quality assurance 1
l responsibilities at the managerial level. These are: (!) the Plant Staff Review Committee (PSRC); (2) the General of fice Nuclear Plant 4 i
Review and Audit Cormittee (CONPR 6 AC); and (1) the President's l
l l Nuclear Advisory Commit tee (PNAC). These are identified in Figure !
l l 17.2-1 of the FSAR.
l The P5RC la made up of the Plant Superintendent and his staff.
I The iunctions of LSis Committee are to review any change or test proposed for the plat.., to review platt operating experience and to r ecommend d isposit ion of najor non-conf ormances. Minutes of the meetings are fowarded to the CONPR 6 AC.
The GONPR 6 AC has represent 'ives f rom the Engineering, Steam Generat ion, Station Construction, and Quality Assurance Departments.
'ihe Committee reviews proposed changes in the facility cr license ,
as well as PSRC reports of violations of procedures and
3 .!
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i deviations. The GONPR & AC reguarly reviews the auditing activities i
of cthers and audits all plant activities. The Committee reports to management and the PNAC.
I The PNAC members are the Manager of Claims and Safety (Chairman),
the Senior Vice President of Engineering and Construction, the Assistant General Counsel, and the Quality Assurance Manager. The Committee will examine and report the activities of the CONPR & AC, review other audit activities and make independent investigations related to nuclear safety.
The Quality Assurance Department is organizationally independent of Power plant operation. In the original submittal of the Diablo Canyon FSAR, the involvement of the QA Department in plant operations was limited to auditing. The Plant Superintendent was assigned respon-sibilities for conducting the quality control activities, whereby he would assign appropriate members of the plant staff to conduct the inspections. We questioned the applicant as to how adequate confidence of control of quality was obtained when the quality. assurance functions are assigned to the same organization which executes the work. The j applicant responded to our concern by changing theso responsibilities to the Quality Assurance Engineer, and by providing the Quality Assurance Engineer with an inspection staff to be supervised at the plant.
Furthermore, the applicant has stated that the QA Department has the
)
authority to stop work, except in instances where stopping the work would
)
1 6
9
17-4 involve changing power level or separating a generating unit f rom the comapny's system grid. PC&E's reply also included the assignsent of responsibilities to the QA Engineer and inspection staf f to inspect modifications, repairs, replacement items, and reworked items. The inspectors on the site QA staf f vill have no other collateral duties-and will be independent of personnel performing the work being intpected. The minimum qualifications and experience for the position cf Quality Assurance Director have been described in the FSAR.
Based on our review of the QA organization of PG&E, as it af f ects l
the Diablo Canyon operations, we f ind adequate responsibility assf.gned i i
In <,onnect 'on to both onsite and _of f site QA personnel and organization.
with this respon.41bility, adequate authority has been assigned to the to assure implementation QA organization and to the Plant Superintendent of an ef f ective QA Program. We conclude that sufficient organizational 1
freedos exists for the QA engineer and inspe . ion staf f and for the Director of the QA Departwnt to assure organizational independence and l
the objective implementation c,f QA activitice,.
17.3 QLality_ Assurance Prograr: 1 4
The applicant's original FSAR 11d not adegoately dereribe the QA
^
Program f or Opci ations. The dcacriptien su limited in the areas of i
t raining, noncanformance cont rol, inspection, welding starp r.ontrol, cc::ci t tent to Resulatory guide ., procurement document control, channi Following eliv. u-sis wit h t h(
tuat rol, and inst rom nt c allbrat ' w.
- ,t a : 1, s'r.si amenL o t ' ei r FRJ t o provi.fc .in .oirquate de- riptive af I
tl.( QA Program f or operatlone.
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_ _-_a
17-5 The QA Program for Operations is documented in the Quality Assurance Manual applicable to plant operations. The Manual contains written policies and procedures which cover all phases of plant operations, '
including operation maintenance, repair, modification testing, refueling, and procurement. '1 The QA Manual is prepared, revised and distributed by the QA Department. In additson, the Plant Superintendent prepares, revises, and controls plant procedures which implement the Quality Assurance Manual appifcable to plant opdrations. A list of the structures, systems, and components controlled by the QA Program has !
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been identified in the FSAR.
A particular concern of the staff was the assignment of responal-bility for the domign and development of modifications of safety related systems, structures, and components (with the exception of mejor design) to the Manager of. Steam Generation. PC6E responded to '
this concern with the establishment of a procedure which provides a practical and standardized means for asking the determination, based.
{
on couple.-ity of design together with reviews and checks required by l
(
the QA Program. I The applicant has described a design control system to l
i be utilized by the Manager of the Steen Generation Department to apply control measures f or compatibility f materials, accessibility for inservice inspection, repair, and delineation of acceptance criteria.
{
i These measures will provide for review of applicable Regulatory standards, design bases, and licensing requirements.
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17-6 Our concern for the lack of an independent qualified plant {
1 inspection staf f was resolved in Amendment 8 to the PSAR. A staff of {
i inspectors, supervised by the Quality Assurance Engineer, will now insp3ct, to instructions specified in an inspection plan, material and ]
I l
equipment received at the power plant, as well as the modifications, 1
l repairs, replacement items, and reworked items.
lhe applicant has deveribed a procurement document control system supported by written procedures in the Quality Assurance Manual. The description of these procedures includes necessary measures to assure that the applicable Regulatory requirements, design bases, and reviews have been included in the procurement doctasents, except that these measures-initially did not include a review by QA personnel. The FSAR has been amended to provide review of procurement documents for quality assurance provisions by the onsite QA Engineer. A QA record system has been described in the FSAR which requires that records of all significant activities be properly stored, retrieved, and retained at l
the plant. j PL&E has identified the implementation of a comprehensive system of 1
planned and periodic audit functions. The QA Department, under the authority of the QA Director, has the assigned responsibility to verify )
I compliance with the QA Program. Audits of plant operations, vendors, l
l and suppliers are conducted in accordance with written procedures.
Provisions have been made to utilize the assistance of personnel from
I 17-7 ,
..i other departments, including persons from the Operating Department, provided they do not have direct involvement with the work being sisdited. Audit reports are submitted to the Senior. Vice President (Engineering and Construction), the General Office Nuclear Plant Review and Audit Committee, the Manager of ti e Steae C oeration Departwnt., and the Plant Superinterwent . haagement audits are conducted by the General office raclear Plant Review and Audit Committee. The President's Nuclear Advisory Committee will also make investigations at the request of corporate management.
l A significar.t concern to the staf f initially was the lack of consitment by the applicant to the applicable Regulatory Guides.
However, in Amendment 8 to the FSA.1, PG&E stated that they had confctmed to the guidance provided in the document, " Guidance on i
Quality Assurance Requirements During Operations Phase of Nuclear- .
Power Plant," (Orange Book) dated October 2t>,1973, and the
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Regulatory Guides and Standards referenced therein. We find this to be acceptable.
In a Commission Memorandum and Order dated December 7,1973 I concerning the LaSalle County Nuclear Station, t! nits 1 and 2, the Regulatory staff was directed to determine, for facilities under construction and for construction applications under review, if quality assurr.nce personnel have auf ficient authority and organir.a-tional freed:>m to perform their critical functions ef fectively and without reservation.
17-8 .
^
Based on our evaluation of the information documented by PG&E in their letter dated February 6,1974, and in Amendment 10 to the FSAR, we conclude that suf ficient authority and organizational f reedom exist within PG&E and Westinghouse to enable QA personnel to perform their i
critical functions effectively and without reservation for the design and construction of Diablo Canyon Units 1 and 2.
Bes.d on our review of the QA Program controls which are being j imposed on the operation of Units 1 and 2, and on our reviev of how these controls are being implemented, we conclude that the QA Program, I
as described in the FSAR, complies with the requirements of Appendix B of 10 CFR Part 50 and is acceptable for the operations phase of Units 1 and 24 including operation, mai.ntenance, modification, repsf r.
and refueling.
17.4 Conclusion As a result of our detailed review and evaluation of PG6E's CM Program description contained in Section 17.2 of the FSAR and a series of discussions and acetings with the applicant, we conclude that the QA organization of PG6E has sufficient independent and authority to r
ef f ectively conduct the QA Program without undue influence from those organizational elements responsible for cost and schedules. In addition, the QA Program description contains adequate QA provisions, requirements dnd controls demonstrating CoEpliance with Appendix B of 10 CFR Part 50, l and is, therefore, acceptable for controlling the operational phase of Units 1 and 2 of the Diablo Canyon Nuclear Plant.
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18.0 EEVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEC !
The report of the ACRS on the review cf the application for operat-ing licenses for Diablo Canyon Units 1 and 2 will be placed in the !
Camaission's Public Document Room, and will be published by the Regula l
tory staff in a' supplement to this Safety Evaluation Report. The '!
supplement will be published prior to the final determination regarding ,
l' issuance of operating licenses for the two units. I j
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- l 15 . 0 CONON DEFENSE AND SECURITY The application reflects that the activities to be conductal vill
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be within the jurisdiction of the United States and that all of the I directors and principal officers of the applicant are United States 1 j
citizens. The applicant is not owned, dominated, or controlled by an !
- alien, a foreign corporation, or a foreign government. The activities to be conducted do not involve any restricted data, but the applicant g
F has agreed to safeguard any such data which might become involved in s
accordanc, with the requiremen's of 10 CFR Part $0. The applicant will l7 rely upon obtaining fuel as it is needed from sources of supply avail-able for civilian purposes, so that no diversion of special nuclear j material required for cilitary purposes la involved. For those reasons, and in 7 the absence of any information to the contrary, we have found that the ig t
l activities to be performed will not be inimical to the common defense j and security, i e
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4n 20-1 I El i D
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.M f!NANCIA1. QUALIMCATIGIS I j j The Commission's regulations which r elate to financial data and i j i M information required to establish financial qualifications pofcant an ap l 4 for operating licenses are Section 50.33(f) and Appendix C of 10 CT
{
Part 50. i 7 We have reviewed the financial information presented in the 6 application and the associated amendments regarding finaccial quali tions.
N, Based on this review and consideration of financial data generally available to the financial analyst, we have concluded that 4
% Pacific Cas and Electric Company possesses or can obtainrythe neces f(
] funds to meet the requirements of 10 CFR 50.33(f) to operate Units 2
y and 2 of the Diablo Canyon Nuclear Power Plant, and if necessary ,
4 1
- '(( shut dcwn the facilities and maintain them in a safe sh .
n Funds to cover the estimated cost of operating the facilities are It l j ex;>ected to be derived from sales of electric energy ac rates which w
- Q j
cover all costs of production plus a reasonable return on invested capital.
4 j
The applicant's estinstes of the annual cost of operating Units and 1
2 are presented below.
7 Unit costs (mills per kWI) are based en the
.{ following capacity factorn for Unit 1:
? 1975 (commercial operation j
.d assumed to begin s
9/1/75) - 32%; 1976 - 65%; 1977 - 75%; 1978 - 81%; I
}- 1979 - 83%; and 1980 - 83%.
Capacity factors estimated for Unit 2 are as follows: (
1976 (commercial operation assumed to begin 6/1/76 - 47%;
1977 - 66%; 1978 - 79%; 1979 - 83%; and 1980 - 83%.
Maximum net elec-trical outputs assumed for Units 1 and 2 are 1131 and 1156 e, HW 1
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r respectively, and according to the applicant, are the expected outputs ;
corresponding to the ultimate reactor outputs of 3488 and 3568 MWt, respectively, as discussed in Section 1.1 of the FSAR.
r '1' Unit 1 Unit 2
! Mills per Total Cost Mills per Total Cost
[
(millions) kW (millions) _ _ kW i
1975 $ 0.9 0.3 85.2 13.2 $26.1 5.5 1976 88.0 11.8 66.9 10.1 1977 90.7 11.3 69.6 8.7 j 1978 93.7 11.4 72.7 8.6 1979 1
1980 95.8 11.6 75.4 9.0 f
Averaging the estimaced operating costs for Unit I during the years
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1976-80 and for Unit 2 during the years 1977-80, when each unit will k
be in operation during the whole calendar year, produces the following
)
t
- amounts grouped by cost elements (A. + C. = administrative and general; 0, + M. = cperation and maintenance),
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Average Annual Cost (millions) Mills per kW l
Unit 1 Unit 2 1976-80 1977-80 Unit 1 Unit 2 Nuclear fuel expense $23.4 $24.0 3.0 3.0 Other nuclear power 2.3 1.2 0.3 0.2
) gener.ition expense 0.1 r Transmission expenses 0.2 --- ---
f A. + G. expenses 2.2 1.3 0.3 0.2 )
- Total 0, + M expenses 28.1 26.6 3.6 3.4 Dep eciation 13.8 10.2 1.8 1.3
/
j Taxes other than income taxes 5.1 3.7 0.7 0.5 ) '
Income taxes - Federal 7.3 4.0 1.0 0.5 Income taxes - other 1.8 1.1 0.2 0.1 Ret urn @ 9.50% 34.6 25.5 4.5 3.2 Total cost $30.7 $71.1 11.8 9.0 i
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The 9.50% rate of return is the applicant's expec:ed weighted cost of i f- j capital for new investments. System-wide sales of electric energy are
} cxpected to cover the costs shows above, which are equivaler.t to the
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" annual revenue requirements" of the subject facilities. The average annual costs of 11.8 and 9.0 mills per kW fot Units 1 and 2, respectively, as shov:: above, are below the' applicant's average revenue of 38.3 mills per kW received during 1973.
1 The applicant has estimated the total cost of decommissioning the subject facilities at $4.5 million for both units on the basis of 1974 dollars. Salvage value of useable equipment could partially offset the i' cost. The following activities are included in the cost estimate: (1) flushing and sealing of auxiliary systems outside the containment; (2) disposal of liquids and gases containing radioactive materials; (3) dis-posal of resins, filters, and miscellaneous radioactive naterials; (4) post cleanup radiation surveys; and (5) sealing of the containment.
Removal of spent fuc1 was considered part of the normal operating expenses and, therefore, was not included as a decor.nissioning cost.
The applicant has indicated that essentially all of the above activities could be carried out by the normal operating crew of the two units in approximately one year. The applicant estinates that disposal of raJio-active materials associated with decommissioning is equivale.nt to about i
three years of contracted waste disposal services during normal opera- '
tion, and has included the disposal costs ir. the $4.5 million estimated 4
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j total cost of decommissioning. The applicant considers the decommissioning j 1
costs to be recoverable operating costs according to state regulatory policies, io that special reserves are not required.
The annual cost of maintaining the shutdown facilities in a safe condition is estimated at $160,000 on the basis of 1974 dollars, and provides for a permanent full time security force at the site plus periodic inspections and radiation surveys. 'Ihe applicant considers the costs of maintaining the shutdown facilities in a safe condition to be recoverable operating costs according to state regulatory policies, so that special reserves are not required.
Information presented in Pacific Gas and Electric Company's annual report for 1973 indicates that operating revenues totaled-$1490.2 million.
Operating expenses were stated at $1172.6 million, of which $158.3 million represented depreciation. Interest on long-term debt was earned 2.8 times. Het income totaled $243.6 million, of which $149.7 million was distributed as dividends to stockholders, with the remaining $93.9 million retained for use in the business. As of December 31, 1973, the i
company's assets totaled $5471.1 million, mest of which was invested in j
utility plant ($5109.9 million) . Retained earnings amounted to $884.4 million. Financial ratios computed from the 1973 financial statements indicate an adequate financial condition, e.g., long-term debt to total capitalization - 51%, and to net ottlity plant - 50%; net plant to capitalization - 1.03; the operating ratio - 79%; and the rates of return 4
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on common equity - 11.4%, on stockholders' investment - 10.0%, and on total investment - 7.1%. I The record of the company's operations during 1971-73 shows that !
i operating revenues increased f rom $1260.3 million in 1971 to $1490.2 l
million in 1973; net income increased from $167.7 million to $243.6 million; net investment in utility plant increased from $4330.9 sillion l j
to $5109.9 million; and the number of times interest on long-term debt was earned increased from 2.7 to 2.8. Moody's Investors Service rates
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the company's first and refunding mortgage bonds ac Aa (high grade bonds) l
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The company's Dun and Bradstreet rating is 5A1, the highest rating.
Recent published data (Moody's Public Utility News Reports dated May 10 1974) indicated that operating revenues increased from $1375.9 million for the 12 month period ended March 31, 1973 to $1532.1 million for the l
12 month period ended March 31, 1974; net income increased from $220.3 '
I million to $250.7 million; carnings available for connon stock increased from $187.2 million to $213.0 million; and earnings per average common
)
share outstanding increased from $3.07 to $3.27. l According to the " statement of aggregate net earnings" prepared in !
connection with the May 21, 1974 issuance of Series 74A, 9-1/8%, First l and Refunding Mortgage Bonds, aggregate net earnings af ter income taxes for the 12 months ended April 30, 1974, amounted to $569.990,000 and vera 3.40 times (versus the required 1.75) the $167,588,715 of annual interest charges on aggregate bonded indebtedness, including the
_ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - ^ ^ ~
?O-6 Series 74A bonds, as defined in the mortgage securing the bonds. As of December 31, 1973, the company had unreimbursed capital expenditures of $2.28 billion, of which $200 million was utilized as the basis for the issuance of the Series 74A bonds.
A cummary analysis reflecting selected financial ratios and other pertint.nt data for Pacific Cas and Electric Company is included as l
Appendix B to this report.
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21-1 1
21.0 FINANCIAL PROTECTION AND INDEMNITY REQUIRDtENTS l Pursuant to the financial protection and indemnification provisions of the Atomic Energy Act of 1954, as amended (Section 170 and related l sections), the Conunission has issued regulations in 10 CFR Part 140.
These regulations set forth the Conunission's requirements with regard
)
to proof of financial protection by, and indemnification of, licensees for facilities (such as power reactors) that are licensed under 10 CPR Part 50.
21.1 Preoperational Storage of Nuclear Fuel The Commission's regulations in Part 140 require that each holder of a construction permit under 10 CFR Part 50, who is also to be the holder of a license under 10 CFR Part 70 authorizing the ownership and possession for atorage only of special nuclear material at the reactor l
- c. construction site for future use as fuel in the reactor (af ter issuance of an operating license under 10 CFR Part 50), shall, during the interim storage period prior to licensed operation, have and maintain financial protection in the amount of $1,000,000 and execute an indemnity agreement 1 with the Conssission. Proof of financial protection is to be furnished prior to, and the indemnity agreement executed an of, the effective '
1 date of the 10 CFR Part 70 license. Payment of an annual indemnity fee 1
2s required.
The applicant, with respect to Diablo Canyon Units 1 and 2, is subject to the foregoing requirements, and will take the following steps, as required.
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21-2 i
l The applicant will furnish to the CommLssion proof of financial l l
protection in the amount of $1,000,000 in the form of a nuclear energy licbility insurance policy.
Further, the applicant will execute an Indemnity Agreement with the Commission as of the ef fective date of its pertinent preoperational fuel ,
I storage license. The applicant will pay the annual indecnity f ee appli- l cr.ble to preoperational fuel storage.
j Operating License l Under the Commission's regulations,10 CFR Part 140, a license l l
l cuchorizing the operation of a reactor may not be issued until proof of l financial protection in the amount required for such operation has been executed. The amount of financial protection which must be maintained for reactors which have a rated capacity of 100.000 electrical kilowatts or more is the maximum amount available from private sources, i.e., the combined capacity of the two nuclear liability insurance pools, which is
$110 million. l l
l Accordingly, no licenses authorizing operation of the Diablo Canyon Units 1 and 2 will be issued until proof of financial protection in the requisite amount has been received and the requisite indemnity agreement or amendment- 1 I
l executed.
l We expect that, in accordance with the usual procedure, the nuclear liability insurance pools will provide, in advance of anticipated issuance of the operating license documents, evidence in writing on behalf of the j
31-3 applicant, that the prescat coverage itse been appropriately amended and -
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that the poli:y limits have been increased to an amount that meets the j requirements of the Commission's regulations for reactor operation.
Similarly, no operating licenses will be issued until an appropriate amend- !
ment to the present indemnity agreement has been issued. The applicant -
I will be required to pay an annual fee for operating license indeamity as provided in the AEC regulations.
1.3 Conclusion On the basis of the Maove considerations, we conclude that the pres-ently applicabic requirements of 10 CFR Part 140, concerning preoperational ;
l storage of fuci, are being satisfied and that. prior to issuance cf an)
I operating licenses, the applicant will be required to comply with the pro-visions of 10 CFR Part 140 applicab'.c to vperating licenses, including those as to proof of financial protection in the requisite amount and to execution cf an approprintc indemnity agreement or amendment thereto I with the Commission.
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CONCLUSIONS Based on our evaluation of the application as set forth above, it is our position that, upon f avorable resolution of the outstanding matters as J
described herein, we will be abic to conclude that:
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(1) The application for f acility licenses filed by Pacific Gas and Electric $[
N$
Company dated October 2,1973, as amended (Arendments Nos. 1 through $
9 17), complies with the requirements of the Atomic Energy Act of 1954, .
I as amended (Act), and the Cossaission's regulations set forth in 10 -j e
CFR Part 1; and
_ir (2) Construction of Units 1 and 2 (the f acilities) has proceeded and ; &
there is reasonable assurance that it will be substantially completed, ^?
A F
in onformity with Construction Permit Nos. CPPR-39 and CPPR-69 i the application as amended, the provisions of the Act, ar.d the rules If i
-lh and regulations of the Connission; and
.d (3) The facilities will operate in conformity with the application as amended, the provisions of the Act, and the rules and regulations ,
9 is of the Commission; and g s
(4) There is reasonable assurance (a) that the activities authorized by -
p a
the operating dcenses can be conducted without endangering the M
- Bt y
health and safety of the public, and (b) that such activities will i -4 be conducted in compliance with the regulations of the Commission p
-y set forth in 10 CFR Part 1; and d NE 7
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BLANK PAGE t.
1 22-2 I
(5) The applic a".t is technically and financially qualified to engage e
in the activities authorized by these licensei, in accordance with the regulations of the Commission set forth in 10 CFR Part 1; ,.
and (6) The issuance of these . censes will not be inimical to the common a
defense and security or to the health and safety of the public.
Before operating licenses will be issued to Pacific Gas and Electric ,
Company for operation of Units I and 2, the units must be completed in k conformity with the provisional construction permits, the appifcation, the Act, and the rules and regulations of the Commission. Such complete-ness of construction as is required for saf e operation at the authorized power levels must be verified by the Commission's Directorate of Regulatory Operatiens prior to issuance of the licenses. i.
3 Further, before operating licenses are issued, tne applicant will be required tc satisfy the applicable provisions of 10 CFR Part 140.
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A-1 APPENDIX A CHRONOLOGY OF THE RADIOLOGICAL REVIEW
- 1. July 10,1973 Application containing the FSAR tendered by Pacific Cas and Electric Company.
- 2. August 13, 1973 Applicant notified that the PSAR portion of the application is not sufficiently complete for docketing.
- 3. August 15, 1973 Initial site visit by LPM.
- 4. August 21, 1973 Heeting with applicant to discuss the deficiencies I s
in the FSAR.
- 5. September 26, 1973 Revised application tendered by PC6E.
- 6. September 28, 1973 Applicant notified that application is suf ficiently emplete, and to file the appropriate documents as required by Section 50.30(c) of 10 CFR Part 50.
- 7. October 2, 1973 Application docketed.
- 8. October 10, 1973 Letter to applicant disclosing staff position regarding ATWS.
- 9. October 19, 1973 !
Notice of opportunity for hearing published in Federal Register (38 FR 29105). j
- 10. October 25, 1973 Site visit and meeting related to geology and seismology.
- 11. November 5, 1973 letter to applicant reminding him of his responsibility to maintain the local Public 1
Document Room. '
- 12. November 14, 1973 Site visit and meeting related to meteorology, hydrology, radiological assessment, and accident analysis.
i
- 13. November 19, 1973 i Submittal of Amendment No. 1 consisting of miscellaneous revised and additional pages of '
the FSAR.
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'14. November 19, 1973 Staff notified by USGS of the discovery of-possible offshore faults in the vicinity of Diablo v. myon.
- 15. December 11, 1973 Meeting with applicant to discuss electrical and instrumentation and control systems.
- 16. December 13, 1973 Submittal of preliminary Scological information related to slope stability.
- 17. December 18, 1973 Meeting with potential intervanors in San Luis Obispo, California.
- 18. December 21, 1973 latter to applicant confirming the safety review schedule for Diablo Canyon.
- 19. December 28, 1973 Request No. I to applicant for additional infor-mation concerning the site and certain radiolog-ical aspects of the plant.
- 20. January 4,1974 Request Hv. 2 to applicant for additional information concerning the site and certain radiological aspects of the plant.
- 21. January 7, 1974 Response received f rom applicant to letter request ,
of October 10, 1973 regarding ATWS. '
- 22. January 8, 1974' Meeting with applicant and USGS in Menlo, Park regarding offshore f aults.
- 23. January 16, 1974 Request No. 3 to applicant 'or additional informa- i tion concerning compliai.ce with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50.
l
- 24. January 17, 1974 Submittal of Amendment No. 2 consisting of l revised and additional pages of the FSAR, and i providing responses to several items contained in the staf f's acceptance review letter of August 13, j 1973. j
- 25. January 21, 1974 Site visit and meetin;; related to ECCS.
- 26. January 22, 1974 Letter informing applicant of the Conmission's Memorandum and . der dated December 7,1973, con-cerning the LaSa.:.le Count / Nuclear Station, Units i 1 and 2, and requesting information regarding the Quality Aasurance Program.
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- 27. January 25, 1974 ASLB Order indicating that an Operating 1.icense Hearing will be held for Diablo Canyon Units 1 d and 2.
l
- 28. February 7,1974 Request No. 4 to applicaat for additional infor-l i
mation concerning Radioactive Materials Safety. l
- 29. February 12, 1974 Response received from applicant to letter request of January 22, 1974 regarding authority and $5 I l
organtr.ational f reeden in the Quality Assurance Program. ,
- 30. February 19, 1974 Submittal of Amendment No. 3 consisting of partial '
response to the staff's requests for additional ,
information dated December 26, 1973 and January 4 '
1974.
- 31. February 20-22, 1974 Site visit and meeting related to electrical and -
instrumentation and control systems. '
- 32. !'ebruary 22, 1974 Site visit and meeting related to geology and slope stability. f
- 33. March 4, 1974 Submit tal of Amendment No. 4 consisting of partial d(b response to the staff's requests for additional information dated December 28, 1973, January 4 E 1974 and January 16, 1974.
- 34. March 15, 1974 Letter to applicant requesting notification as to when information on outstanding items will be submitted.
- 35. March 18, 1974 Submittal of the Diablo Canyon Industrial Security Plan.
- 36. March 19, 1974 Submittal of Amendment No. 5 updating Chapters 4 ;h\ I and 15 of the PSAR concerning the 17x17 fuel design. - t A
3'. March 20, 1974 letter to applicant requesting information on y
Bergen-Patterson snubbers which may be installed j ki on any safety related systems.
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- 38. March 20, 1974 Letter to applicant confirming that the Industrial Security Plan will be withheld from public dis-l closure in conformance with Section 2.790(d) of 10 CFR Part 2. .
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- 39. March 25, 1974 Submittal of the Diablo Canyoi Site Emergency Plan.
- 40. March 26-27,1974 Beginnint of the first OL Prehearing Conference.
- 41. March 29, 1974 Submittal of Amendment No. 6 consisting of miscellaneous revised and additional pagen, of the FSAR.
- 42. April 12, 1974 Letter to applicant requesting additional infor-mation en the Industrial Security Plan.
- 43. April 12, 1974 Request No. 5 to applicant summarizing previously requested information for which acceptable responses have not been received.
1
- 64. April 15, 1974 Submittal of Amendment No. 7 consistinR of I toiscellaneous revised and additional pages of !
I the FSAR. l 1
- 45. April 15, 1974 Request No. 6 to applicant summarizing previously l l requested information for which acceptable l responses have not been received. !
- 46. April 16, 1974 Letter from applicant responding to the staf f's letter of March 15, 1974 regarding outstanding l items. ]
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- 47. April 24, 1974 Meeting vith applicant to discuss electrieni and instrumentation and control systems.
- 48. April 25,1974 Meeting with applicant to discuss the Quality Assurance Program, the Industrial Security Plan, and the Site Emergency Plan, j
- 49. April 26, 1974 letter from applicant indicating when the infor-l mation requested in our letters of April 12 and j
- April 15, 1974 vill be provided. l l
- 50. April 26, 1974 Meeting with applicant to discuss of f shore geology ,d seismology.
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- 51. April 3 N kiy 1, 1974 Conclusion of the first OL Prehearing Conf erence. l l
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- 52. May 7, 1974 Letter to applicant requesting additional finan- !
cial information. j l
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- 53. May 13, 1974 Submittal of Amendment No. P. consisting of partial response to the staff's requests for-additional information dated April 12 and April 15, 1974.
- 54. May 17,1974 Letter from applican. requesting extension of the completion dates shown in the construction ;
permits for Units 1 and 2 (CPPR-39 and CPPE-69, respec t ively) .
- 55. May 21, 1974 Site visit and meeting related to pipe break out- ;
side containment.
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- 56. Ma/ 23, 1974 Submittal of additional infomation on the !
Industrial Security Plan that was requested on April 12, 1974.
- 57. May 24, 1974 Letter to applicant requesting additfonal informa-
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tion regarding the estension of the construction permits for Units 1 a.d ' 2.
- 58. May 28, 1974 Submittal of additional electrical and instrumen-tation drawings.
- 59. May 31, 1974 Submittal of knendment No. 9 consisting of partial response to the staff's requests for additional information dated April 12 and April 15, 1974..
- 60. May 31, 1974 Letter to applicant confiming that the revised Security Plan which was received on May 23, 1974 vill be withheld f rom public disclosure in con-formance with Section 2.790(d) of 10 CFR Part 2.
- 61. June 4, 2 974 Meeting with applicant to discuss offshcre geology and seismology
- 62. June 4, 1974 Submittal of Amendment No. 10 consisting of additional information requested by the staff in connection with its review of the 17x17 fuel design.
- 63. June 6,1974 Meeting with applicant regarding tsunami wave calculations.
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- 64. June 17,1974 Letter from applicant providing additional justification for their request for extension of construction permits CPPR-39 and CPPR-69.
- 65. June 17, 1974 Letter f rom applicant providing the additional financial information that was requested on May 7, 1974.
- 66. June 17,1974 Request No. 7 to applicant for additional information on the 17x17 fuel design.
- 67. June 18,1974 Letter to applicant requesting additional informa-tion regarding the preoperational testing program for the emergency core cooling system.
- 68. June 20,1974 Submittal of Appendix 1 to the Site Emergency Plan consisting of the San Luis Obispo County Sheriff's Department Interim Evacuation Plan.
- 69. June 25,1974 Response from applicant to our letter of March 20, 1974 concerning the use of Bergen-Patterson snubbers.
- 70. June 26, 1974 Letter to applicant concerning the scheduling of forthcoming operator and senior operator cold examinations for Unit 1 of the Diablo Canyon Nuclear Plant.
- 71. J une 27, 1974 Submittal of Amendment No.11 consisting of partial response to the staf f's requests for additional information dated January 4, April 12, and April 15, 1974. Of particular importance in this amendment are the initial responses to questions on offshore geology and seismology.
- 72. June 28,1974 Letter to applicant extending the latest dates for completion of construction for Units 1 and 2.
- 73. July 1,1974 Submittal of four appendices which supplement the applicant's final report on the potential effects of pipe break outside containment.
- 74. July 2, 1974 Submittal of Amendment No. 12 consisting primarily of the applicant's final report on the potential ef f ects of pipe break outside containment.
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- 75. July 5,1974 Heeting with applicant to review the progress of new geological field investigations related to offshore faults in the vicinity of the Diablo Canyon site.
- 76. July 5,1974 Submittal of Amendment No. 13 consisting primarily of responses to the staff's request for additional-information dated June 17, 1974 regarding the 17x17 fuel design.
- 77. July 12. 1974 Letter to applicant requesting additional informa-tion on the Industrial Security Plan.
- 78. July 16,1974 Response from applicant to our letter of June 18, 1974 regarding preoperational testing of the ECCS.
- 79. August 2, 1974 Submittal of Amendment No. 14 consisting of l
miscellaneous revised and additional pages of the FSAR.
- 80. August 5, 1974 Submitt,a1 of Amendment No.15 consisting pr'-
marily of the applicant's evaluation of com pliance with the final ECCS acceptance criteria.
- 81. August 13, 1974 Response frem applicant to our letter of July 12, 1974 requesting additional changes in the Industrial Security Plan.
- 82. August 16, 1974 Submittal of Amendment No. 16 consisting pri-marily of a partial response to the staff's request for information on tsunami waves caused by near-shore generators (see letter to appli-cant dated January 4, 1974). Amendment 16 l
71so providen additional information in response to our letter of June 18, 1974 regarding pre-operational testing of the EC.:S.
- 83. September 3, 1974 Submittal of Amendment No.17 consisting of miscellaneous revised and additional pages of the FSAR.
- 84. September 12, 1974 ACRS Subcommittee meeting emphasizing Eeology and seismology and ECCS - Appendix K evaluations.
- 85. September 18, 1974 Initial meeting regarding the Westinghouse -
Standard Technical Specifications.
- 86. September 19, 1974 Site visit emphasizing itms still outstanding in the safety review.
7 A-8 k
- 87. October 3, 1974 Response from applicant to our letter of Oc:ober 10, 1973, regarding ATWS.
- 88. October 10, 1974 Request No. 8 to applicant for addition.*1 information on the operator requalificati;n program.
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APPENDIX B i PACIFIC CAS AND ELECTklC COMPANY FINANCIAL ANALYSIS DOLKET NOS. 50-275 AND 50-323 (dollars in millions)
Calendar Year Ended December 31 1973 1972 1971 Long-term debt k Utility plant (net) $2,539.0 $2.390.0 $ 2, 301.0 Ratio - debt to fixed plant 5,109.9 4.715.6
.50 4.330.9 Utility plant (net) .51 .53 Capitalization 5.109.4 4,71 5.6 Ratio of net plant 4, 965.4 4,564.9 4.330.9 to capitalization 4.287.5 1.03 1.03 i Stockholders' equity 1.01 Total assets 2.426.4 2,175.0.
i Proprietary ratio 1,986.5 \
5.4 71.1 4,993.1
.41 4,630.8 4 1
.44 Eoinings available to commen equity 43 Common equity 206.9 184.2
! 1,811.5 167.7 Rate of carnings on common equity 1,610.0 1,521.5 i
Net income 11.4% 11.4% 11.01 Stockholde'rs' equity 243.6 215.3 2,426.4 167.7 Rate of earnings on stockholders' equit y 2,175.0 1,986.5 10.0% 9.9%
Net income before interest 8.4%
Liabilities and capital 337.3 34 5.3 Rate of earnings on total investment 5.471.1 309.8 4,993.1 7.1% 4.630.8 Net income before Interest 6.9% 6.72 3 Interest on 1cng-term debt 387,3 No. 345.3 309.P of t imes long- term lat erest c u ... o 138.0 127.6 2.8 114.4 Net income 2.7 2.7 Total revenues 243.6 215.3 Net income r2tio 1.559.9 167.7 1,415.0 1,303.2 ;
.lb .15 i
Total utility operating expenses .13 I lotal utility operating revenues 1,172.6 1,069.7 993.4 Operating ratio 1,490.2 1,350.6 I 1,260.3 Utility plant (gross / .79 .79 .79 Utility operating revenues 6,748.4 6,233.8 1,490.2 5.734.3 Ratio of plant investment to revenue. 1, 150.6 1,260.3 4.53 4.62 4.55
.aliza!!on: 1973 1972 Amount : of Total tong-term debt Amount of Tot a l
$ 2, *. 3 9. 0 Preferred stock 51.11 52,390.0 t14.9 52.4%
Common st ock 6 surplus 12.4 564.0 Total _13 01.5 12.4
%.5 1,610.0 35.2
' 4, 0f6. 4 100.0%
Moody's 13ond Rat iny,: Aa
$4.564.9 100.0T Dun & liradst reet tredit it im W 4
C-1 APPENDIX C LIBLIOCRAPHY (Docum:nte referenced in or used to prepare the Safety Evaluation Report for the Dicbio Canyon Nuclear Power Station, Units 1 and 2) generel
- 1. Preliminary Safety Analysis Reports with Amendments for the Diablo Canyon Huelcar Power Station, Units 1 and 2 (Docket Nos. 50-275 and 50-323).
- 2. Final Saf ety Analysis Report with Amendments 1 through 17 for the Diablo Ccnyon Nuclear Power Station, Units 1 and 2 (Docket Nos. 50-275 and 50-J23).
3 Unitec States Atomic Energy Commission Rules and Regulations,10 CFR:
Part 1, Statement of Organization and General Information Part 2, Rules of Practice Part 9. Public Record Part 20, Standards for Protection Against Radiation Part 50, Licensing of Production and Utilization Facilities Part 55, Operators' Licenses Part 71, Packaging of Radioactive Material for Transport and Transportation of Radioactive Material Under Certain Conditions Part 73, Physical Protection of Special Nuclear Material Part 100, Reactor Site Criteria Part 140, Financini Protection Requirements and Indemnity Agreements Part 170, Fees for Facilities and Materials Licenses Under the Atomic Energy Act of 1954, as Amended.
l4. United States Ato nic Energy Commission Regulatory Guides.
l 5. National Environmental Policy Act of 1969 (NEPA).
l Meteorology l6. Harris, M.F., " Effects of Tropical Cyclones Upon Southern California."
- lbster of Arts Thesis, Department of Geography, California State University, Northridge Ca., 1973.
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- 7. Holeworth, G.C. " Mixing Heights, Wind Speeds, and Potent f al for Urban Air Pollution Throughout the Contiguous United States AP-101".
Environmental Protection Agency. Of fice of Air Programs, Research Triangle Park, North Carolina,1972.
- 8. Huachke, R.E., "Clossary of Meteorology." American Meteorological I Society Boston, Mass., 1959.
- 9. Korshover, J., " Climatology of Stagnating Anticyclones East of the Rocky Mountains, 1936-1970." NOAA 'lechnical Memorandum ERL ARL-34, Silver Spring, Md., 1971.
- 10. Srgendorf, .I., "A Prograr for Evalcating Atmospheric Dispersion from a Nuclear Power Station." NOAA Technical Memorandum ERL ARL-42, Idaho Falls Idaho, 1974.
- 11. SELS Unit Staff, National Severe Storms Forecast Center, " Severe Local Itorm Recurrences, 19 D-1967." ESSA Technical Memorandum WBIM FCST 12, Of fice of Meteorological Operations. Silver Spring, Md.,1969,
- 12. Slade, D.ll. (ed.), " Meteorology and Atomic Energy - 1968." TID-24190, Naticnal Technical Information Service, Springfield, Va.,1968.
I 'l . Smith, M.E. (ed.), " Recommended Guide for the Prediction of the Dispersion of Airborne Elfluents." The American Society of Mechanical Engineers, New York, N.Y., 1968.'
- 14. Thom, ii.C.S., " Tornado Probabilities." Monthly Weather Review, October-December 1963, pp. 730-737,1962.
- 15. Thom, H.C.S. "New Distributions of Extreme Winda in the United States."
Journal of the Structural Division, Proceedings of the American Society cf Civil 1.ngineers - July 1968, pp.1787 -1801,1968. )
- 16. Turner, D.It., " Workbook of Atmospheric Dispersion Estimarts." Public ilcalth Service Publication No. 999-AP-26, Cincinnati Ohio, 1970.
I
)
- 17. United Stater, At omic Energy Conaission, Regulatery Guide 1.4, "Assump-l tjons Used for Evaluating the Potential Radiological Consequences of a )
1."
Los:. of Coolant Accident for Pressurized Water Reactors-Revision l L'SAEC. Directorate of Regulatory Standards Washington, D.C.,1973.
18 United States Atomic Energy Commission, Regulatory Guide 1.23, "Onsite Meteorolor,1 cal Programs." USAIC, Directorate of Regulatory Standards, Washington, D.C., 19 D .
i I
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C-3
- 19. United States Atomic Energy Commission, Regulatory Guide 1.42, " Interim Licensing Policy on As low As Practicable For Caseous Radiciodine Releases From Light-Water-Cooled Nuclear Power Reactors." USAEC, Directorate of Regulatory Standards, Washingten D.C.,1973.
- 20. United States Department of Commerce, Environmental Data Service: " local Climatological Data, Annual Summary with Comparative Data - Santa Maria, California." Published annually through 1972.
- 21. United States Naval Weather Service, " Worldwide Alifield Summaries, Volusse VIII Part 1 United States of America (West Coast, Western Mountains and Creat Basin)." Federal Clearinghouse for Scientific and Technical Information, Springfield, Va., 1969. i Structural Engineering
- 22. American Institute of Steel Construction, " Specification for Design.
Fabrication & Erection of Structural Steel for Buildings," (Sixth Edition) 101 Park Avenue, New York, New York, 10017, 1969.
I
- 23. American Concrete Institute, " Building Code Requirements for Reinforced Concrete (ACI 316-1963)," P.O. Box 4754, Redford Station, Detroit, j Michigan 48219, 1963. '
- 24. American Society of Mechanical Engineers, "ASME Boiler and Pressure l Vessel Code,"Section III, and Addenda, United Engineering Center, 345 East 47th Street, New York, New York 10017.
- 25. " Wind Forcer on Structures", Final Report of the Task Committee on Wand Forces of the Committee on Load and Stresses of the Structural Division, (
Transactions of the American Society of Civil Engineers, 345 East 47th -
Street, New York, N.Y.,10017, Paper No. 3269, Vol.126, Part II, p.
1124-1198. 1961. .
j Missile Protection
- 26. Owaltney, R.C., "Miselle Generation and Protection in Light-Water-Cooled Power Reactor Plants," USAEC Report ORNL-NSIC-22, September 1968.
- 27. Williamson, R.A. and Alvy, R.R., " Impact Effeet of Fregments Striking Structural Elements," Holmes and Narver (Revised Edition),1973.
Seismic Instrumentation
- 28. Kapur, Kanvar K., " Seismic Instrumentation for Nuclear Power Plants,"
ANS Topical Meeting on Water-Reactor Safety, Salt Lake City, Utah, March 1973. .
C-4
- 29. United States Atomic Energy Commission, Regulatory Guide 1.12. "Instru-mentation f or Earthquakes." USAEC, Directorate of Regulatory Standards.
1974.
Washington, D. C., ;
Component Design
- 30. American Society of Mechanical Engineers, " Criteria of the A$ME Boiler and Pressure Vessel Code for Design by Analysis in Sections III and VIII, Division 2." United Engineering Center, New York, New York,1969.
- 31. American Society of Mechanical Engineers, " Nuclear Power Piping," USA Ctandard B 31.7, United Engineering Center, New York, New York,1969.
- 32. American Society of Mechanical Engineers, "Powet Piping," USA Standard B 31.1.0, United Engineering Center, New York, New York, 1967.
- 33. American Society of Mechanical Engineers, " Steel Pipe Flanges and Flanged Fittings," USA Standard B 16.5, United Engineering Center, New York, New York, 1968.
- 34. American Society of Mechanica! Engineers, " Wrought Steel and Wrought Iron Pipe," USA Standard B 36.10, United Engineering Center, New York, New York, 1959.
- 35. American Society of Mechanical Engineera, " Wrought Steel Buttwelding '
Fittings," USA Standard B 16.9, United Engineering Center, New York, New York, 1964
- 36. " Manufacturers Standardization Society Pipe Hangers and Supports - Materials and Design," MSS Standard Practice SF-58, Arlington, Virginia,1959.
- 37. Manufacturers Standardization Soe #'y, " Pressure-Temperature Ratings for Steel Butt-Welding End Valves," Med Standard Practice SP-66, New York,1959.
- 38. American Water Works Association, " Standard for Steel Tanks Standpipes, Reservoirs, and Elevated Tanks for Water Storage," AWA D 100-1967 New York New York, 1967.
I
- 39. " Standards of Tubular Exchange Manuf acturers Association," Fif th Edition.
- 40. American National Standards Institute. " Design Basis f or Protection of Nuc1 car Power Plants Against the Ef f ect of Postulated Pipe Rupture," ANSI N 176 Draft, April 5, .1974.
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C-5 pterials Engineering
- 4. American National StsAards Institute, 4eakage-Rate Testing of Contain-ment Structures for Nuclear Reactors," ANSI N45.4-1972, New York, New York.
March 16, 1972.
- 42. American Society of Mechanical Engineers, "ASME Boiler and Pressure Vessel Code," Sections II, III and XI, with Addenda through Summer 1973, United Engineering Center, New York, New York. 1
- 45. American Society of Mechanical Engineers, " Methods and Definitions for Mechanical Testing of Steel Products," ASME-SA-370-71b, ASME Boiler and Pressure Vessel Code,Section II, Part A, Ferrous, 1971 Edition, with Addenda through Suzaer,1972, New York Eew York.
4t . American Society for Testing and Materials, " Copper-Copper Sulphate- l Sulfuric Acid Test for Detecting Susceptibility to Intergranular Attack 1 in Stainless St(els " Annual Book of AStK Standards, Part 31. Philadelphia, Pennsylvania, July 1973.
4! . American Society for Testing and Materials, " Notched Bar Impact Testing Of Metallic Materials," ASTM-E-23-72, Annual Book of ASTM Standards, Part 31, Philadelphia, Pennsylvania, July 1973. j
- 40. American Society for Testing and Materials, " Standard Method for Conducting Dropweight Tect to Determine Nil-Ductility Transition Temperature of Ferritic Steels," ASTM-E-208-69 Annual Book of ASTM Standards, Part 31 Philadelphia, Pennsylvania, July 1973.
- 47. American Society for Testing and Materials, " Surveillance Tests on Structu-ral Materials in Nuclear Reacters," ASYM-E-185-73, Annual Book of ASTM Standards, Part 31, Philadelphia, Pennsylvania, July 1973.
- 48. United States Atomic Energy Commicsion R* ales and Regulations,10 CFR Part 50, and Appendices A, G, H, and J, and Paragraph 50.55a.
- 49. United States Atomic Energy Commission, Regulatory Guide 1.14. " Reactor Coolant Pump Flywheel Integrity," USAEC, Directorate of Regulatory Standards, Washington, D. C., October 1971.
- 50. United States Atomic Energy Commissica, Regulatory Guide 1.31, " Control of Stainless Steel Welding - Revision 1," USAEC, Directorate of Regulatory .
Standards, Washington, D. C. , June 1973.
f 9* t
C-6 51.
United States Atomic Energy Commission, Regulatory Guide 1.34, " Control of Electro-Slag Weld Properties," USAEC, Directorate of Regulat my l Standards, Washington, D. C., December 28, 1972.
52.
United States Atomic Energy Commission, Regulatory Guide 1.36, "N(cmetallic Thermal Insulation for Austenitic Stainless Steels," USAEC, Directorate of 23, 1973.
Regulatory Standards Washington, D.C., February 53.
United States Atoeic Energy Commission, Regulatory Guide 1.43, " Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components,"
C., May USAEC, 1973.
Directorate of Regulatory Standards, Washington, D.
54.
United States Atomic Energy Commission, Regulatory cuide 1.44, " Control of the Use of Sensitized Stainless Steels," USAEC, Directorate of Regulatory Standards, Washington, D. C., May 6,1973.
55.
United States Atomic Energy Commission, Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," CSAEC, Directorate of C., May 1973.
Regulatory Standards Washington, D, 56.
United States Atomic Engery Commission, Regulatory Guide 1.50, " Control of Preheat Temperature for Welding of Low-Alloy Steel," USAEC, Directorate of Regulatory Standards Washington, D.C., May 1973.
57.
United States Atomic Energy Commission, Regulatory Guide 1.51, " Inservice Inspection of ASME Class 2 and 3 Nuclear Power Plant Components, "USAEC, Directorate of Regulatory Standards Washington, D. C. May 1973.
- 58. United S stes Atomic Energy Commission Cuideline Document, " Inservice Inspectica Requirements for Nuclear Poser Plants ConstructedC., with Limited Accessibility for Inservice Inspections," USAEC, Washington, D.
January 31, 1969. ;
Reactor
- 59. Clasatone, S. and Sesonski, A., " Nuclear Reactor Engineering," D. Van Nostrand Company, Inc., Princeton, New Jersey,1963.
- 60. Westing' ause Electric Corporation, " Reference Core Report 17 x 17," WCAP-8185 Pittsburgt., pennsylvania, September 1973. l
- 61. Tong, L.S., " Boiling Heat Transf er and Two-Phase Flow," John Wiley and I Sons, Inc. , New York,1967. l t
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- 62. United States Atomic Energy Commission, " Safety Evaluation Report of the ]
Zion Nuclear Power Station, Units 1 and 2 (Docket Nos. 50-295 and 50-304)," i USAEC, Directorate of Licensing, Washington, D. C., October 6,1972. l l
- 63. United States Atomic Energy Commission, " Safety Evaluation Report of the l Trojan Nuclear Power Station (Docket No. 50-344)," USAEC, Directorate i of Licensing, Washington, D. C., October 1974.
- 64. Westinghouse Electric Corporation, " Anticipated Transients Without Reacter Trip in Westinghouse Pressurized Water Reactors," WCAP-8096, Pit taburgh, Penasylvanic, April 1973.
- 65. Barry, R.F., et al, "The Panda Code " WCAl-7759, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, April 1967. ;
i
- 66. Westinghouse Electric Corporation, " Safety Analysis of the 17 x 17 Fuel l Assembly for Ccabined Scismic and Loss-of-Coolant Accident," WCAP-8278 (Propriat:ay), Pittsburgh, Pennsylvania, February 1974.
- 67. Westinghouse Electric Corporation, " Hydraulic Flow Test of the 17 x 17 Fuel Asseshly," W:AP-8236 (Proprietary), Pittsburgh, Pennsylvania, December 1973.
l
- 68. United States Atomic Energy Cossaission, " Technical Report on Densification of Westinghouse PWR Fuel," USAEC, Regulatory Staf f, Washington, D.C.,
May 14, 1974.
- 69. Westinghouse Electric Corporation, " Overpressure Protection for Westing-house Pressurized Water Reactors," WCAP-7769, Revision 1 Pittsburgh, Pennsylvania, 1972.
Engineered Eafety Features
- 70. Allen, A. O., "The Radiation Chemistt:y of Water and Aqueous Solutions,"
Van Nostrand Company, 1961.
- 71. American Nuclear Society, " Decay Energy Release Ratee Following Shut-down of Uranium-Fuel Thermal Reactors (DRAFT)," ANS Standard ANS-5.1, i liinsdale, Illinois, October 1971. l
- 72. Brodrick, J. R., Burchill, W. E., and Lowe, P. A., "1/5 Scale Intact Loop Post-LOCA Steam Relief Tests," CENPD-63, (Rev.1), Combustion Engineering, Inc., March 1973.
l
1 C-8
- 73. Coward, H. F., and' Jones, C. W., "Limf t s of Flammability of Gases and !
Vspors, Bureau of Mine Bulletin 503, 1952.
j q )
- 74. DiNunno, J . J . , Ande rson, F. D . , take r , R. E . , and Wate r f ield , R. 4 . ,
" Calculation of Distance Factors for Power and Test Reactor Sites," l TID-14844, USAEC, Washington, D. C., March 23, 1962. !
- 15. "nD0D/ MOD 001 - A Code to Determine the Core Reflood Rate .for a PWR Plant with 2 Core Vessel OUTLET Legs and 4 Core Vessel Inlet Legs,"
Interim Report, Aerojet Nuclear Company . November 2.1972.
i Lowe, P. A., Brodrick, J. R., and Burchill, W. E. , " Steam-Water Mixing 76.
Test Program. Task D: Formal Report for Task A: 1/5 Scale Broken loop,"
CENPD-65 (Rev.1), Combustion Engineering, Inc. , March 197 3. ,
j :
I 77. Moody, F. J., " Maximum Flow Rate of a Single Component, Two-Phase Mixture,"
Journal of Heat Transfer, Vol 87, p.134, February 1965.
j 78. Paraly, L. F., " Design Considerations of Reactor Contaitunent Spray Systema.,
Part VI, the Heating of Spray Drops in Air-Steam Atmospheres," USAE Report
{ ORNL-IN-2412, Oak Ridge, Tennessee, January 1970.
- 79. Rit tig. W. H., Jayne, G. A. , Moore, K. V. , Slater, C. E. , and Uptmer, M. L. ,
i "RE1.AP 3 - A Computer Program for Reactor Slowdown Analysis," 1N-1321, Idaho Nuclear Corporation, June 1970, t
[ 80. Richardson, L. C. , Finnegan, L. J., Wagner, R. J. , and Waage, J. M.,
l " CONTEMPT - A Computer Program for Predicting the Containment Pressure f Temperature Response to a Loss-of-Coolant Accident," IDO-17220, Phillips Petroleum Company, June 1967.
}
- 81. Schmitt, R. C. Eingham, G. E., and Norberg, J. A. , " Simulated Design Basis l
Accident Tests of the Carolina Virginia Tube Reactot Containment," Final
' Report, IN-1403, Idaho Nuclear Corporation, December 1970.
- 82. Slaughterbeck, D. C., " Comparison of Analytical Techniques Used to Determine Distribution of Mass and Energy in the Liquid and Vapor Regions
) of a P'eiR Containment following a Loss-of-Coolant Accident," Special
[ interim Report, Idaho Nuclear Corporation, September 1970.
j
- 83. Slaughterbeck, D. C., " Review of Heat Transfer Coefifients for Condensing Steam in a Containment Building Following a Loss-of-Coolant Accident,"
1N-1388, Idaho Nuclear Corporation, September 1970. '
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- 84. Tagami, T., " Interim Report on Safety Assessments and Facilities Establish-ment Project in Japan for Period Ending June 1965 (NJ. 1)," Prepared for the National Reactor Testing Station, February 28, 1966_(Unpublished work).
l
- 85. Uchida, H., Oyama, A., and Toga, Y. , "Evaluatica of Post-Incident Cooling l
( Systems of Light-Water Power Reactors," Proceeding of the Third Inter-l national Conference on the Peaceful Uses of Atorde Energy Held in Geneva August 31 - September 9,1964. Volume 13, Ses.icn 3.9, New York: United Nations 1965, (A/ Conf. 28/P/436) pp.93-104, May 1964. !
- 86. Wagner, R. J., and Wheat. L. L., " CONTEMPT-LT Urers Manual " Interim Report 1-214-74-12.1 Aerojet Nuclear August 1973.
Instrumentation, Controls and Electric Power Systems l l
l 87. Fir.a1 Safety Analysis Report for the Donald C. Cook Nuclear Plant Units
~
)
1 and 2 - Sections 6, 7 and 8 (Docket Nos. 50-3?.5 and 50-316) .
1 88.~ Final Safety Analysis Report for the Zion Nuclear Power Station, Units 1 and l 2 - Sections 6, 7 and 8 (Docket Nos. 50-295 and 50-304) .
- 89. Institute of Electrical and Electronic Engineers, " Criteria for Protection Systems for Nuclear Power Generating Stations," IEEE Std. 279-1971, New York j New York.
L
) 90. Institute of Electric and Electronic Engineers, " Criteria for Class IE Electric Systems for Nuclear Power Cec + rating Stations," IEEE Std. 308-1971.
- 91. Institute of Electr'c and Electronic Engineers. "IEEE Trial-Use Standard:
General Guide for qualifying Class I Electronit Equipment for Nuclear Power ,
( Generating Stations," IEE 5+.d. 323-1F1, New Ycrk, New Yor k. '
- 92. Institute of Electric and %ctronic Engineers, " Trial Use Criteria for f the Periodic Testing of % elear Power Generating Station Protection Systems,"
IEEE Std. 336-1971, Ne York, New York.
- 93. Institute of Electric and Electronic Engineers, "IEEE Guide for Seismic p Qualification of Class I Electric Equipment for Nuc1 car Power Gcnerating l Station Protection Systen." IEEE Std. 344-1971, New York, New York.
i l 94. United States Atonic Energv Corrissic ., Regulatory Guide 1.6, " Independence Between Redundant Standby Ontite) Pet:r Four c; and Between Their Distri-but ion Systems," USAEC, Dit cc t orate c# Repulatrry btandarde. Washington, D. C., March 1971.
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- 95. United States Atomic Energy Commission, Regulatory Guide 1.9., " Selection of Diesel Generator Set Capacity for Standby Power Supplies " USAEC, Directorate of Regulatory Standards, Washington, D. C., March 1971.
- 96. United States Atomic Energy Commission, Regulatory Guide 1.22. " Periodic Testing of Protection System Actuation Functions " USAEC, Directorate of Regulatory Standards, Washington, D. C., February 1972.
- 97. United States Atomic Energy Commission, Regulatory Guide 1.32, Use of IEEE Std. 308-1971, " Criteria for Class IE Electric Systems for Nuclear Power Generating Stations," USAEC, Directorate of Regulatory Standards, Washington, D. C., August 1972.
- 98. United States Atomic Energy Commission, Regulatory Guide 1.40 " Qualification
(
Tests of Continuous Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants," USAEC, Directorate of Regulatory Standards, Washington, D. C., March 1973.
- 99. United States Atomic Energy Commission, Regulatory Guide 1.41, "Preopera-tional Testing of Redundant Onsite Electric Power Systems to Verify Proper Load Group Assignments " USAEC, Directorate of Regulatory Standardc, Washington, D. C., March 1973.
100. United States Atomic Energy Commission Regulatory Guide 1.47, "By-passed and Inoperable States Indication for Nuclear Power Plant Satety Systems," USAEC, Directorate of Regulatory Standards, Washington, D. C. ,
May 1973.
101. United States Atomic Energy Commission, " Safety Evaluation Report of the Donald C. Cook Nucicar Power Station, Units 1 and 2 (Docket Nos. 50-315 and 50-316)," USAEC, Directorate of Licensing, Washington, D. C., September 10, 1973.
102. United States Atomic Energy Commission " Safety Evaluation Report of the Zion Nuclear Power Station, Units 1 and 2 (Docket Nos. 50-295 and 50-304),"
USAEC, Directorate of Licensing, Washington, D. C., October 6,1972, 103. Westinghouse Electric Corporation, " Seismic Testing of Electrical and Control Equipment (High Seismic Plants) " WCAP-8021, Pitteburgh, Pennsylvania, December 1971.
104. Westinghouse Electric Corporation, " Environmental Testing of Engineered Safety Features Related Equipment (NSSS-Standard Scope)," WCAP-7744, Pittsburgh, Pennsylvania, August 1971.
C-11 105. Westinghouse Electric Corporation, " Solid State Logic Protection System Design," WCAP-7672, Pittsburgh, Pennsylvania, June 1971.
106. Westinghouse Electric Corporation, " Engineered Safeguards Final Device or Actuator Testing," WCAP-7705, Pittsburgh, Pennsylvania, March 1973.
Conduct of Operations 107. American National Standards Institute, " Industrial Security for Nuclear Power Plants," ANSI N 18.17-1973 (ANS 3.3), New York, New York,1973.
108. American National Standards Institute, " Standard for Administrative Controls for Nuc) oar Pcwer Plants," ANSI N 18.i !"72 (ANS 3.2), New York, New York, 1972.
109. American National Standards Institute, " Standards for Selection and Training of Personnel for Nuclear Power Plants," ANSI N 18.1-1971, New York, New York, i 1972. l 110. United States Atomic Energy Commission, Regulatory Guide 1.8, " Personnel Selection and Training," USAEC, Directorate of Regulatory Standards.
Washington, D. C., March 10, 1971.
111. United States Atomic Energy Comuniasion, Regulatory Guide 1.17 " Protection of Huclear Power Planto Against Industrial Sabotage," USAEC, Directorate of Regulatory Standards, Washington, D. C., June 1973.
112. United States Atomic Energy Commission, Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)", USAEC, Directorate of Regulatory Standards. Washington, D. C., November 3,1972.
113. International Atomic Energy Agency, " Planning for the Handling of Radiation )
Accidents," IAEA Safety Serica No. 32 Vienna, Austria, Noven:bar 1969.
114. United States Atomic Energy Commission, " Guide to the Preparation of Emergency Plans for PrcJoction and Ut111:stion Pacilities," USAEC, Washington, D. C., December 1970.
115. United States Atomic Energy Coussission, " Utility Staffing and Training for Nuclear Power," WASH-1130 (Revision 1), USAEC, Washington, D. C., June l
1973. i l
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l
i C-12 Initial Tests and Operation 116. United States Atomic Energy Countission, " Guide for the Planning of Initial Startup Programs," USAEC, Washington, D. C., December 17, 1970.
117. United States Atomic Energy Commission, " Guide for the Planning of Pre-operational Testing Programs," USAEC, Washington, D. C., December 7, l 1970.
Rediation Protection 118. American National Standards Institute, " Guide for Administrative Practices in Radiation Monitoring, " USAS N 13.2-1969, New York, New York,1969.
119. American National Standards Institute, " Guide to Sampling Airborne Radio-active Materials in Nuclear Facilities," ANSI N 13.1-1969 New York, New York,1970.
120. Environmental Instrumentation Group, " Survey of Instrumentation for En-vironmental Monitoring," Volume 3 - Radiation Lawrence Berkeley Laboratory, Berkeley, California October 1973.
W:ste Management 121. United States Atomic Energy Coundesion, " Final Environmental Ste sment i Concerning Proposed Rule Making Action: Numerical Guides For Design l Objectiws and Limiting Conditions To Meet The Criterion "As Low As Practicable" For Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents," WASH-1258, USAEC, Washington, D. C., July 1973, i
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