ML20246P119
| ML20246P119 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 10/11/1974 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| References | |
| NUDOCS 8907200112 | |
| Download: ML20246P119 (117) | |
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- i October 11,1974 SAFETY EVALUATION RE MRT BY THE DIRECTORATE OF LICENSING
, U. S. ATOMIC ENERGY COMMISSION IN THE MATTER OF DUQUESNE LIGHT COMPANY TOLEDO EDISON COMPANY PENNSYLVANIA POWER COMPANY BEAVER VALLEY POWER STATION UNIT 1 DOCKET NO 50-334 !
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, TABLE OF CONTENTS PAGE-
11.0 INTRODUCTION
AND GENERAL DESCRIPTION OF PL ANT. . . . . . . . . . . . . . . . . . . . . 1 -1 1.1~ . Introduction.....,.......................................... 1 1.2 Gen era l Pl a n t De s cri p ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . 1-2 1.3 Comparison with Similar Facility Designs..................... 1-3' 1.4 Identi fication of Agents and Contra ctors . . . . . . . . . . . . . . . . . . . . 1-3 1.5 Summa ry of Princi pal Review Matte rs . . . . . . . . . . . . . . . . . . . . . . . . . .1-4 1.6 Facility Modifications as Result of Regulatory. Staff Review 4 2.0 S I T E C HARA CTE R I S T I CS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1. Geogra phy an d Demo gra phy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.2 L Nearby Industrial, Transportation and Military Facilities... 2-4 2.3 Meteorology........................... ..................... .2-5 2.4 Hydrology................................................... 2-7 2.5 Geol ogy and Se i smol ogy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-10 3.0 DESIGN CRITERIA FOR STRUCTURES, CDMPONENTS. EQUIPMENT AND SYSTEMS '3-1 3.1 Con formance with AEC General Design Criteria. . . . . . .. . . . . . . . . 3-1 3.2 Classification of Structures, Components, and Systems....... 3-1 3.3 Win d and Tornado Loadi ngs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2' 3.4 Wa ter Level - ( Fl ood) De s ign . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2
'3.5 Missile Protection.......................................... 3-3 3.6 ' Protection Against Dynamic Effects Associated with the-Postulated Rupture of Piping.............................. 3-4 3.6.1 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Inside Containment. 3-4 3.6.2 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Outside Containment', 3-4 3.7 Se i sm i c 0e s i g n . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-6 3.8 Design of Sei smi c Category I $tructures . . . . . . . . . .. . . . . . . . . . . 3-7 3.9 Mechanical Systems and Components.................... ...... 3 -8 3.10 Seismic Qualification of Category I Instrumentation and Electrical Equipment...................................... 3-10 4.0 REACTOR.......................................................... 4-1 4.1. Genera 1...................................................... 4-1'
-4.2 Fuel Mech an i c al De si gn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.3 Reactor Vessel Internals............................. ....... 4-4 4.4 The rma l and Hyd rauli c Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4.5 N uc l ea r De s i g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-9 5.0 REACTOR COOLANT SYSTEM........ ................................... 5-1 o 5.1 S ummary Descri pti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Integrity of the Reactor Coolant Pressure Boundary........... 5-1 5.3 Leakage Detection System...... .............................. 5-5 5.4 I nse rvice Inspecti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 5.5 P ump Fl ywhee l Inte gri ty. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 5.6 Loose Parts Monitor.......................................... 5-6 5.7 Pump 0verspeed.............. ................................ 5-6 9
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11 6.0 ENG I NEERED S AFETY FE ATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 Genera 1..................................................... 6-1 6.2 Containment Systems.................... .................... 6-1 6.2.1 Contai nment Functional Design. . . . . . . . . . . . . . . . . . . . . . 6-1 6.2.2 Con tai nment Hea t Removal Sys tems . . . . . . . . . . . . . . . . . . . . 6-3 6.2.3 Supplementary Leak Collection and Rele&se System.... 6-4 6.2.4 Containment Isolation Systems....................... 6-4 6.2.5 Combustible Gas Control Systems............. ....... 6-4 6.2.6 C ontainment Leakage Testing Program. . . . . . . . . . . . ... 6-5 6.2.7 Containment Air Purification and Cleanup Systems.... 6-6 6.3 Dnergency Core Cooli ng Sys tem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6 . 3.1 Design 8ases....................... ................ 6-6 6.3.2 System Design........... ........................... 6-6 6.3.3 Pe rformance E val uati on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 j i
6.4 Habi tabi l i ty Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 J 7.0 INSTR UMENTATION AND C DNTR0LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.1 I n tr od uc ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.2 Reactor Trip System......................................... 7-1 7.3 Engineered Safety Features Actuation and Control .... . . . .. . . . 7-1 7 . 3.1 Transfer to Recirculation Mode. . . . . . . . . . . . . . . . . . . . . . 7-1 7.3.2 Reactor Coolant Loop Isolation. . . . . . . . . . . . . . . . . . . . . . 7-2 7.3.3 Accumulator Isolation Valves............... ........ 7-2 7.3.4 Hydrogen Recombiner System and Supplementary Leak Collection and Release System. . . . . . . . . . . . . . . . 7-2 7.3.5 Vital Supporting Systems for Engineered Safety Features........................... .............. 7-2 7.4 Systems Requi red for Safe Shutdown. . . . . . . . . . . . . . . . . . . . . . . . . . 7-2 7.5 Safety Related Display Instrumentation. . . . . . . . . . . . . . . . . . . . 7-3 7.6 Overpressure Protection Interlocks for Residual Heat R emo v a l Sys t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-3 7.7 Environmental and Seismic Qualifications..... .. .......... 7-3 7.8 Cable Separation and Identification Criteria. . . . . .. . . . . . . . . 7-4 7.9 Control Sys tems Not Required for $afety. . . . . . . . . . . . . . . . . . . 7-4 7.10 Anticipated Transients Wi thout Scram. . . . . . . . . . . . . . . . . . . . . . . . 7-4 8.0 ELECTRIC P0WER............................... ... ............. . 8-1 8.1 I n t roduc ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 8.2 O f f s i t e Pow e r Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 8.3 Dnsi te P ower Systens . . . . . . . . . . . ........................... 8-2 8.3.1 A-C Power System................................,... 8-2 8.3.2 D -C 'P ow e r Sys t e ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 9.0 AUXILIARY SYSTEMS.... ............. ........... ............. .. 9-1 0
9.1 Chemi cal and Vol ume Control Sys tem. . . . . . . . . . . . . . . . . . . . . . . 9-1 9.2 Fuel Storage and Handling............................ ...... 9-1 9.3 Wa ter Sys tens . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-3 9.3.1 Camponent Cooling Water System.......... ........... 9-3 9.3.2 Residual Hea t Removal System. . . . . . . . . . . . . . . . . . 9-4 9.3.3 River Water System............... .. ............... 9-4 9.3.4 Ultimate Heat Sink................ ................. 9-5 9.4 Other Au xi l i a ry Sys tens . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-5 9.4.1 Compress ed Ai r Sys t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-5 9.4.2 Equipnent and Floor Draina ge System. . . . . . . . . . . . . . . . . 9-6 9.4 3 Air Conditioning, Heating, Cooling and Ventilation System.............................................. 9-6
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9.4.4, Diesel Generator Fuel'011 Storage and Transfer System......................................... 9-7 9.4.5 -Diesel Generator Auxiliary Systems.............. 9-7 9.4.6 F i re Protecti on Sys tem. . . . . . . .' . . . . . . . .. . . . . . . . . . . '9-7 10.0 ' STEAM AND POWER CONVERSION SYSTEM............................. 10 10.1- Summary Description.................................... 10-1 10.2 Turbine Generator...................................... 10-1
'10.3 Main Steam Supply System............................... 10-1 10.4 'Other Features......................................... 10-2 10.5 - Auxili ary Feedwa ter Sys tem. . . . . . . . . . . . . . . . . . . ... . . . . . . . . 10-3 11.0 RADIOACTIVE WASTE MANAGEMENT....................... .......... 11-1 11.1 Desi gn Objective and Criteri a . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.2 Li qui d Was te Systems .' . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . 11-2 11.3- Gaseous Waste Systems.................................. '11-3 11-5 11.4 ' Solid Radwaste.........................................
11.5 Design...'.............................................. 11-5 11.6 . Process and Area Radiation Monitoring Systems.......... 11-5 11.7- Conclusions............................................ 11-6 i
RADIATION PROTECTION.......... ............................... 12 -12.0 12.1 Shielding. .s............................................ 12-1 12.? Heal th Physi cs P rog ram. . . . . . . . . . . . . . . . . . . . .'. . . . . . . . . . 12-1 12.3 V en t 1 1 a t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-2 12.4 Area Monitoring........................................ 12-2 13.0 CONDUCT OF OPERATIONS......................................... 13-1 13.1 Plant Organization. Staff Qualifications and Training.. 13-1 13.2 Safety Review and Aud1t................................ 13-1 13.3- Plant Procedures and Records............................. 13-2 13.4 Emergency Planning.............. ...................... 13-2
'13.5 Industrial Security.................................... 13-3 INITIAL TESTS AND OPERAT!0NS.................................. 14-1 14.0 15-1 15.0 ACCIDENT ANALYSIS.............................................
Genera 1................................................ 15-1 15.1 Derign Basis Acci dent Assumptions . . . . . . . . . . . . ... . . . . . . . . 15-1 15.2 15.2.1 Loss-of-Coolant Accident. . . . . . . . . . . . . . . . . . . . . . 15-1 15.2.2 Fuel Handling Accidents....................... 15-2 15.2.3 Gas Decay Tans, Rupture.............. ......... 15-2 16-1 16.0 T ECHN ICAL SPECI FICATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
QUALITY ASSURANCE,..................... ....................... 17-1 17.0 Quality Assurance Program for Operations. . . . . . . . . . . . . . 17-1 17.1 17.1.1 Genera 1....................................... 17-1 17.1.2 Org an i z a ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17-1 17.1.3 Quality As surance Program. . . . . . . . . . . . . . . . . . . . . 17-1 17.1.4 Conclusions...................................
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Quality Assurance Program for Construction............ 17-3 17.2 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 18-1 ( ACRS) 18-1 18.1 Construction Permit Review............................
18.2 Operating License Review.............................. 18-l 4
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iv 19.0 _00fHON DEFENSE AND SECURITY................................... 19 1 l 20.0 FINANCIAL QUALIFICATIONS.. ....... .................... ..... 20-1 21.0 FINANCI AL PROTECTION AND INDEMNITY REQUIREMENTS . , . . . . . . . . . . . 21 1 21.1 General............ ............ ...... ............... 21-1 21.2 Preopera tional Storage of Nuclear Fuel . . . . . . . . . . . . . . . . . 21-1 21.3 O p e ra ti ng L i c e n s e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21-1
22.0 CONCLUSION
S.... ........ ..................................... 22 1 APPENDICES APPENDIX A Chronology of Radiol ogical Review. . . . . . . . . . . . . . . . . . . . . A-1 APPENDIX B Financial Analysis.................................... B-1 APPENDIX C Bibliography................... ...................... C-1 LIST OF TABLES TABLE 4.2 1 Comparison of Fuel Mechanical Design of D. C. Cook and Bea ver Val ley Un i t 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 TABLE 4.2 2 Range of Design Parameter Experience. . . . . . . . . . . . . . . . . . 4-3 TABLE 4.4 Canparison of Thermal and Hydraul:. nesign Parameters for the Beaver Valley Unit 1 and Surry Plants....... 4-6 TABLE 15.1 Potential Offsite Doses Due to Design Basis Accidents 15-1 LIST OF FIGURES FIGURE 2.1 Beaver Valley Power Sta tion Site . . . . . . . . . . . .. . . . . . . . 2-2 FIGURE 2.2 Cumulati ve Populati on Di stribution. . . . . . . . . . . . . . . . . . . 2-3 FIGURE 17.1 Duquesne Light Company Quality Assurance Organization 17-4 o
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s ABBREVIATIONS 7- a-c alternating current ACI American Concrete Institute
,ACRS Advisory Consnittee on Reactor Safeguards -
AEC United States Atomic Energy Consnission AISC - American Institute of Steel Construction ANS American Nuclear Society ANSI American National Standard Institute
.ASME American Society of Mechanical Engineers ASTM American Society for Testing and Material ATWS' anticipated transients without scram -!
2 British themal units per. hour per square foot-Btu /hr-ft CVCS chemical volume anti control system !
cfs cubic feet per second'
-CFR Code of Federal Regulations Ci/yr curies per year i Ci/yr/ unit curies per year per unit DBA . design basis accident DBE design basis earthquake dc direct current DNB departure from nucleate boiling DNBR departure from nucleate boiling ratio DLC Duquesne Light Company i
ECCS emergency core cooling system-ESF engineered safety features f[
f ultimate strength of concrete yield strength of steel y
FSAR Final Safety Analysis Report
'F degrees Fahrenheit ft foot 2 square feet ft ft3 cubic feet-g gravitational acceleration. 32.2 feet per second per second gal gallons gpm gallons per minute hr hour
- I-1 31 iodine 131 IEEE Institute of Elec. a. and Electronics Engineers in inch.
kV kilovolt kW kilowatt kW/ft kilowatts per foot Ib pound LOCA loss-of-coolant accident LPD linear power density h.
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vi ABBREVIATIONS Cont'd l
LPZ low population zone m meter m2 square meters l MM modified Mercalli mph miles per hour m/sec meters per second MSL mean sea level MWe megawatts electrical MWt megawatts thermal mrem one thousandth of a Roentgen equivalent man OBE operating basis earthquake PMF probable maximum flood ppm parts per million PSAR Preliminary Safety Analysis Report :
psi pounds per square Inch psig pounds per square inch qauge I psia pounds per square inch absolute PWR pressurized water reactor QA quality assurance
. rad radiation absorbed dose RCPB reactor coolant pressure boundary RCS reactor coolant system Rem Roentgen equivalent man j RHR residual heat removal )
RTS reactor trip system i scfm standard cubic feet per minute sec/m3 seconds per cubic meter SSWS standby service water system SWS service water system SLCR$ supplementary leak collection and release system SSE safe shutdown earthquake i U-235 uranium 235 I UO2 urapium dioxide USGS United States Geological Survey l W Westinghouse Electric Corporation )
0 X/Q relative concentration l 10 CFR AEC. Title 10. Code of Federal Regulations Part 2 AEC Rules of Practice Part 20 AEC Standards for Protection Against Radiation Part 50 AEC Licensing cf 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 100 AEC Reactor Siting Criteria 9
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1 1-1
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction The Duquesne Light Canpany Toledo Edison Company, and Pennsylvania Power Company (hereinafter referred to as the applicants) filed with the Atomic Energy Comission (Commission) an application dated January 13, 1969 and as subsequently
. amended, for a license to construct and operate a nuclear pswer plant, identified as the Beaver Valley Power Station Unit 1 (Beaver Valley Unit I or Unit I).
The plant is located on a site adjacent to the Shippingport Atomic Power Station in Shippingport Borough, Beaver County, Pennsylvania, which is about 25 miles north-west of Pittsburgh, Pennsylvania. Beaver Valley Unit 1 is being constructed under Construction Permit CPPR-75 issued on June 26, 1970. Be:ver Valley Power Station Unit 2 (Beaver Valley Unit 2 or Unit 2) is also being constructed on the same site as Unit 1 under Construction Permit CPPR-105 issued on May 3,1974.
The current application which was docketed on October 18. 1972 requests an operating license for Beaver Valley Unit 1 for 2652 thermal megawatts (MWt) reactor output power. This is equivalent to a net electrical output of 851.9 electrical megawatts (MWe) and is the same power level that was requested in the initial application. A Final Safety Analysis Report (FSAR) was filed with the application as required by 10 CFR 50.34(b). The information in the FSAR has been supplemented by Amendments I through 11. The FSAR and copies of these amendments are available f or public inspection at the J. 5. Atomic Energy Commission, Public Document Room, 1717 H Street, N. W., Washington, D. C. and at the Beaver Area Memorial Library, 100 College Avenue, Beaver, Pennsylvania.
In the course of the safety review, we held a number of meetings with represen-tatives of the applicants Stone and Webster Engineering Corporation, and Westing-house Electric Corporation to discuss the plant design, construction, proposed op-eration and performance under postulated accident conditions. During our review we requested the applicants to provide additional information that we needed for our evaluation. This additional information was provided in amendments to the application. As a result of our review, a number of changes were made in the facility design and proposed operating practices. These changes are listed in Section 1.6 of this report and are described in appropriate sections of the report.
A chronology of the principal actions relating to the processing of the application is attached as Appendix A to this report.
o This safety Evaluation Report sumarizes the results of the radiological safety review of Unit 1 performed by the Commission's Regulatory staff and delineates the scope of the technical details considered in evaluating the radiological safety aspects of the final design and proposed operation of Unit 1. The review and evalua-tion of the facility for an operation license is only one stage in the continuing review by the staff of the design, construction, and operating features of Unit 1.
The proposed design of Unit 1 was reviewed before a construction permit was issued.
The Commission reported the results of the radiological safety review for the con-struction permit in its Safety Evaluation Report dated April 24, 1970. Construction
S 1-2 of Unit I was and is being monitored in accordance with the inspection program of the Cbission's Regulatory staff. Subsequent to the issuance of an operating license, Unit 1 may then be operated only in accordance with the Conmission's regulations and the terms of the operating license under the continued surveillance of the Commission's Regulatory staff.
In addition to our review, the Advisory Committee on Reactor Safeguards (ACRS) l is reviewing the application and will meet with both the applicants and the Regulatory staff to discuss the final design and proposed operation of Un't 1. The ACRS report to the Commission will be provided in a supplement to this report.
The environmental impact considered in the review of Unit 1 in accordance with Ap-pendix D to 10 CFR 50 is discussed in the Connission's Final Environmental Statement dated July 1973.
Based on our evaluation of the application to operate Beaver Valley Unit 1, we have concluded that Unit 1 can be operated as proposed without endangering the health and safety of the public. Our detailed conclusions are presented in Section 22.0 of this report.
1.2 General Plant Description Beaver Valley Unit I utilizes a nuclear steam supply system which consluts of a
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pressurized water reactor and a three-loop reactor coolant system with a reactor coolant pump and a steam generator in each reactor coolant loop. The nuclear core is composed of fuel rods nade of enriched uranium dioxide pellets enclosed in zircaloy tubes with welded end plugs that an! grouped and supported in assemblio. The mechan-ical control rods consist of clusters of stainless steel-clad silver-indium-cadmium alloy neutron absorber rods that are inserted into zircaloy guide tubes located within the fuel assemblies. Water will serve as both the neutron moderator and reactor coolant, and will be circulated through the reactor vessel, core, and reactor coolant loops by the react 0V coolant pumps located in the cold leg of each reactor coolant loop.
The reactor will be controlled by the regulation of boric acid concentration in the reactor coolant and by movement of the mechanical control rods. The control rods, whose drive shafts penetrate the top head of the reactor pressure vessel, will be moved verti-cally within the core by individual control rod drive mechanisms, i
The reactor and reactor coolant system will operate at a pressure of 2250 psia wito a reactor coolant inlet temperature of 542.5 'F and a reactor coolant outlet temperature of 612.8'F at the reactor design output power of 2652 MWt. The reactor coolant water will be circulated through the tube side of the three steam generators to produce satur-ated steam. An electrically heated pressurizer connected to the hot leg of one of the reactor coolant loops will establish and maintain the reactor coolant pressure and O provide a surge chamber and water reserve to accommodate reactor coolant volume i change during reactor operation. The steam that is generated in the three steam genera-tors will be used to drive a tandem-compound turbine coupled to an electric generator and will be condensed in a twin-shell, single pass, divided water box surface condenser.
Heat rejected in the surface condenser is transferred by a closed loop circulating water system to a natural draft cooling tower. Make-up water for the cooling tower will be drawn from the Ohio River through the river water system and raw water system, t
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The reactor is protected by a reactor protection system that automatically initiates .
appropriate _,, action whenever a plant condition nonitored by the system approaches pre-establishedlimits. This reactor protection system will act to shut down the reactor, close isolation valves, and actuate the engineered safety features should any or all of these actions be required.
The nuclear steam supply system is housed in a steel-lined reinforced concrete con-tainment structure which is maintained at subatmospheric pressure. The containment,
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including its penetrations, is designed to limit the radioactive material that could be I
released in the event of an accident. The areas contiguous tn the containment, such as the safeguards area and the main steam valve enclosure, will confine most of the leakage that might occur from the penetrations. This leakage will be filtered and exhausted to the atmosphere by the supplementary leak collection and release system.
The emergency core cooling system consists of accumulator tanks, high pressure safety injection charging system, and low head safety injection system with provisions for recirculation of borated coolant after the end of the it jection phase of operation.
Various combinations of these systems assure core cooling for the complete range of postulated pipe break sizes in the reactor coolant system.
An auxiliary building is located near the containment structure and houses the waste treatment facilities, components of engineered safety features, and various related auxiliary systems for the reactor unit. Separate fuel handling facilities will contain the spent fuel pool for storing spent fuel and the new fuel storage facility.
The plant is capable of being supplied with electric power from offsite sources through at least two independent transmission lines and is provided with independent and redundant onsite emergency electrical power supplies which are capable of providing the power necessary to shut down the plant safely or to operate the engineered safety features in the event of an accident and a loss of offsite pcwer.
1.3 Comparison with Similar Facility besigns The principal features of the design of Beaver Valley Unit I are similar to those we have evaluated and approved previously for other nuclear power plants now under con-struction or in operation, especially Surry Power Station Units 1 and 2 (Docket Nos.
50-280 and 50-281), which is presently licensed for operation. To the extent fecsible and appropriate, we have made use of our previous evaluations of those features that were shown to be the same as those previously considered. Comparisons to Surry Units 1 and 2 are contained in Table 1.4-1 of the FSAR. Our evaluation reports for these other facilities have been published and are available for public inspection at the Atomic Energy Comission's Public Document Room at 1717 H Street, N. W., Washington. 0.C.
1.4 Identification of Agents and Contract _o,rs, The joint applicants, previously cited, assume complete responsibility for Beaver Valley Unit 1. The Duquesne Light Company is managing the construction of the plant and will operate the plant for the applicants. The applicants have retained the Stone and Webster Engineering Corporation to perform architectural and engineering work and to supervise the construction. The Westinghouse Electric Corporation was contracted to design, manufacture, and deliver to the site the nuclear steam supply system and the initial core for Beaver Valley Unit 1 Westinghouse has also provided technical assist- j ance during the erection of the nuclear steam supply system and will provide technical assit,tance for core loading, startup, and preoperational plant testing. The turbine generator was purchased from Westinghouse.
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applicants also utilized the advice ~ and assistance of four consulting firms, as required, 'in spet ta112ed areas; for example, Hansen,'Holley and Biggs assisted in struc-tural designs and analyses;.NUS Corporation assisted in site meteorology, ecology, and radiological monitoring; Weston Geophysical .Research, Inc. assisted in seismology and .
seismic surveys; and Whitman and Rand assisted in soil dynamics.
1.5 : Sumery of Principal Review Natters
- The evaluation performed by the staff included a review of the information submitted by the applicants, particularly with regard to the following matters:
(1) We evaluated the population density and use characteristics of the site environs, and the physical characteristics of the site, including seismology, meteorology, geology and hydrology to establish that these characteristics had been determined adequately and had been given appropriate consideration in the final design of the plant, and that the site characteristics are'in accordance with the Canmission's siting criteria as stated in 10 CFR Part 100,' taking into consideration the design of the facility and the engineered safety features provided in the design.
(2) We evaluated the design, fabrication, construction, and testing and performance of the plant to determine that they are in accordance with the Commission's Gene-ral Design Criteria, Q)ality Assurance Criteria, Regulatory Guides, and other .
appropriate rules, codes and standards, and that any departures from these criteria, codes, and standards have been identified and justified.
(3) We evaluated the expected response of the plant to certain anticipated operating transients and postulated accidents. We judged that the potential consequences of a few highly unlikely postulated accidents (design basis accidents) would ex-ceed those of all other accidents considered credible. We performed conservative.
analyses of these design basis accidents to determine that the calculated potential -
offsite doses that might result in the very unlikely event of their occurrence would not exceed the Coninission's guidelines for site acceptability as given in 10 CFR Part 100.
.(4) We evaluated the applicants' organizational structure and qualifications of opera-ting and technical support personnel, plans for the conduct.of plant operation, the measures taken for industrial security, and the plans for emergency actions to be taken in the unlikely event of an accident that might affect the general public, to determine that the applicants are technically qualified to safely operate the plant.
(5) We evaluated the design of the systems provided for control of the radioactive ef-fluents from the plant to determine that these systems are capable of controlling the release of radioactive wastes from the plant within the limits of the Commis-sion's regulations, and that the equipment provided is capable of being operated
- by ths applicants in such a manner as to reduce the radioactive releases to levels that are as low as practicable.
(6) We evaluated the financial data and the information provided by the applicants as required by the Coninission's regulations (Section 50.33(f) of 10 CFR Part 50 and Appendix C to 10 CFR Part 50) to determine that the applicants are financially qualified to operate the facility.
1.6 Facility Modifications as Result of Regulatory Staff Review During the course of the review a number of technical and administrative changes were made as a result of staf f requirements. These changes are described in the amendments
1-5 to the application. We have listed below the more significant modifications that have resulted from our review. Included are references to the sections of this report where each matter is discussed more fully. The principal changes which were made are as follows:
(1) Installation of an auxiliary river water system to provide the capability for safe shutdown of the plant in the event of a postulated barge impact and explosion at the main intake structure (see Section 2.2 and Section 9.3.4).
(2) Relocation / reburial of oil pipelines which traverse the site to a minimum depth of four feet (see Section 2.2).
(3) Development of plant operating procedures and technical specifications limiting operation of the plant during flood conditions and providing for flood protection of safety-related equipment (see Section 2.4).
(4) Densification of additional soils in the area of the intake structure to remove the ootential for soils liqu faction in the event of an earthquake (see Section 2.5).
(5) Design measures for protection of safety-related equipment in the service building, auxiliary building and in the intake structure (see Section 3.4).
(6) Design measures for protection against the dynamic effects associated with pipe ruptures (see Section 3.6).
(7) Augmentation of the seismic instrumentation program (see Section 3.7).
(8) Modifications to the ventilation system of the intake structure to provide pro-tection againM wave action during flood conditions (see Section 2.4).
(9) Modifications 6f control circuits to provide automatic opening of the accumulator iso?ation valves when the primary coolant pressure exceeds a pre-selected value (see Section 7.3.3).
(10) Installation of an administratively controlled status board to indicate the availability of systems (see Section 7.5).
(11) Modification of the controls for the residual heat removal system to provide protection against overpressurization (see Section 7.6).
(12) Modifications of the offsite power system to protect against a single failure by providing a second redundant protective relay scheme (see Section 8.2).
(13) Modifiest;on of the dual feeder breakers which power redundant loads te protect against single failure of the onsite a-c power system (see Section 8.3.1).
(14) A revised loading sequence for the emergercy diesel generator units to minimize undervoltage and under frequency transients (see Section 8.3.1)
(15) Modification of the redundant emergency a-c power buses to protect against cc.m mode failure by removal of the single tie breaker (see Section 8.3.1).
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' 2.0 SITE CHARACTERISTICS
(: + 2.1 Geography and Demography q The Beaver Valley site is a 449 acre tract of land located on the south bank of the Ohio River in Beaver County, Pennsylvania. The site is approximately one mile from Mid-land, Pennsylvania (population 5,271), five miles from East Liverpool, Ohio (population 20.020) and 'S miles from Pittsburgh, Pennsylvania (population 520,117). The plant is q adjacent to the Shippingport Atomic Power Station. Beaver. Valley Unit 2 for which a construction permit was Lissued on May 3,1974, is also being constructed on the same .
site.
The applicants have specified a circular site exclusion area with a radius of 610 meters (2,000 feet) for Beaver Valley Unit 1. The applicants have specified a low popu-lation zone with a' radius of 5795 meters (3.6 miles). The nearest occupied residence is approximately 650 meters (2,100 feet) from Beaver Valley Unit I reactor. Figure 2.1 shows the exclusion area for the site.
The topography of the Beaver Valley site consists of a fairly level river terrace on the south bank of the Ohio River at an approximate elevation of 730 feet above mean sea level-(MSL). Hills rise to approximately 1,100 feet above MSL on both sides of the Ohio River which is about 1.300 feet wide in the vicinity of the site.
Figure 2.2 shows the present and projected cumulative population surrounding the Beaver Valley site. The population within the low population zone is about 14,000 '
based on the 1970 census. The nearest population center (as defined in 10 CFR Part 100) with a present population exceeding 25,000 is Weirton, West Virginia (1970 population of 27,000), and its nearest boundary is 15 miles from the plant. the applicants have, however, specified East Liverpool, Ohio (1970 population of 20,000)' as the nearest population center and its nearest boundary is five miles from the plant. Since this-distance is at least one and one-third times the low population zone radius of 3.6 miles, the applicants' selection of the population center distance is in compliance with 10 CFR Part 100 guidelines.
At the present time, the land bordering the Ohio River is highly industrialized, whereas the inland areas are of a rural nature. The Ohio River is used for transportation and source of water for the industries and towns in the river valley. Pleasure boating and sport fishing take place during the warm months of the year. There is no extensive conenercial fishing on the Ohio River. The nearest downstream municipality using the
' Ohio River for drinking water is Midland, approximately 1.3 miles downstream and on the e opposite side of the Ohio River from the Beaver Valley site. The public facilities with-in the low population zone include six schools and one jail.
The major agricultural land use in the vicinity of the Beaver Valley site during 1965 was for field and forage crops, as reported by the applicants. According to the 1969 Census of Agriculture issued by the U.S. Department of Conrierce, this held true for surrounding counties in Ohio and West Virginia as well. The predominant acre-age was in corn, wheat and hay. Vegetables account for a minor portion of the land use.
The Census also reported a reduction of about 18% in the number of milk cows between
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1964 and 1969 in Beaver County. The nearest location considered suitable for grazing of a dairy cow is approximately 0.38 miles northeast of the Berver Valley Unit 1 contain-ment building. , Recreational land and water use in the area of the Beaver Valley site consists of hunting, fishing, and boating. According to the applicants Raccoon Creek State park, located eight miles south of the site, has varied recreational facilities and has an annual attendance in excess of 500,000.
We have concluded that land uses have been acceptably considered and are not criti-j cal with respect to the operation of the plant.
l On the basis of the applicants' specified population center distance, minimum exclu-sion area and low population zone, our analysis of the onsite meteorological data from which atmospheric dilution factors were calculated (Section 2.3 of this report), and the calculated potential radiological dose consequences of design basis accidents discussed in Section 15.0 of this report, we have concluded that the exclusion area, low populatbn tone, and population center distance meet the guidelines of 10 CFR Part 100 and are acceptable.
2.2 Nearby Industrial, Transportation and Military Facilities The closest industry to the Beaver Valley site is the Crucible Steel / Mpany steel mill at Midland, about one and one-half miles downstream on the opposite bank of the Ohio River, where approximately 6,000 people are employed. In Shippingport Borough, there is one industrial operation which employs 60 people in deep coal mining and washing activities.
The Ohio River is .used for barge shipment of coal, coke, petroleum products, sand, gravel, steel products, and chemicals. Records indicate approximately 6,600 commercial lockages per year at Montgomery Dam, located three miles upstream from the Beaver Valley site.
Across the river from the site, the Penn Central railroad follows the north bank of the Ohio River. There is a Penn Central railroad right-of-way on the plant site which is controlled by Ouquesne Light Company and is limited to servicing the Shippingport Atomic Power Station and the Beaver Valley Power Station.
The state highways which provide access to the site and surrounding areas are shown in Figure 2.1. There are no airports, military installations or missile sites within ten miles of the Beaver Valley Power Station. Five oil pipelines and one natural gas pipeline cross the Beaver Valley site.
Data provided by the epolicants indicated that during a five year perind, three gasoe line or oil barges have broken loose on the Ohio River in the vicinity of the site.
Based on this data, we concluded that the probability of a gasoline barge impact at the river water intake structure, with resulting explosion of the barge, would be high enough O
to consider this kind of an accident in the design of the intake structure. Since this kind of accident could result in a loss of function of the safety-related pumps and piping located in the intake structure, the applicants have proposed an auxiliary river water system (further discussed in Section 9.3.4 of this report) to provide the cooling requirements for safe plant shutdown in the event that such an accident causes a loss of river water flow from the pumps located in the intake structure.
We have reviewed the applicants' design of the auxiliary river water system and have concluded that for this design to be acceptable, the applicants must modify the design to provide for automatic actuation of the system on loss of the normal river 4
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2-5 w ter system flow. The applicants propose to demonstrate by analysis that sufficient time is available for manual actuation. We will review the applicants' analysis prior to granting an operating license to confirm whether manual actuation would be accept-able or whether automatic actuation would still be required. The applicants propose to complete the installation of the auxiliary river water system within one year after attaining full commercial power operation. We find this time schedule to be reasonable and therefore acceptable.
During our review at the construction permit stage for Oeaver Valley Unit 2, we re-quired the applicants to evaluate the effects on safety-related structures of a postu-lated rupture of the natural gas pipeline which traverses the site. We concluded that we could not concur with the applicants' evaluation provided in the PSAR for Beaver Valley l' nit 2. We performed our own evaluation and, based on the results of that evaluation, we informed the applicants that we require relocation of the pipeline to a minimum distance of about 1,000 feet from the nearest sa'ety-related structurcs of Beaver Valley Unit 2. The applicants have committed to relocate the natural gas pipeline to meet our requirements for minimum separation distance from the Unit 2 safety-related structures. Since the present routing of the pipeline, as well as the planned re20cated route, is more than 1000 *eet from safety-related structures of Beaver Valley Unit 1, we have concluded that the present location, as well as the planned relocateo route is acceptable with respect to Unit 1.
We have also required the applicants t9 evaluate the effects on plant cooling sys-tems of a oostulated release of oil from the oil pipelines which cross the Ohio River in the vicinity of the plant intake structure. We have reviewed the results of the applicants' evaluation and concur with the applicants' conclusion that the release of oil will not affect essential plant cooling systems and, therefore, would not affect plant safety.
During our review of the Beaver Valley Unit 2 at the construction permit stage we had also evaluated the consequences of a postulated rupture of the oil pipelines which I traverse the site in the vicinity of safety-related structures and had concluded that plant safety will not be endangered, provided that the oil lines are buried to a minimum depth of about four feet. Since the applicants have committed to bury the oil pipe-lines to a minimum depth of four feet, we find this acceptable.
The applicants have also analyzed the consequences of rail and truck cargo explosions on plant structures and determined that plant s'tfety will not be impaired. W? concur with these analyses and conclusions.
2.3 Me teorology Western Pennsylvania has a humid, continental climate that is modified only slightly by the Great Lakes to the northwest and the Atlantic Ocean to the east. The predominant air mass type over the area during nost of the year is continental polar of Canadian I
origin. In the summer, however, there are fiequent intrusions of warm, humid maritime tropical air masses moving northward from origins over the Gulf of Mexico. Precipita-tion is well distributed throughout the year. During the winter months there is, clima-otologically, a 50 percent chance of measurable precipitation (>0.01 inch) on any day and one fourth of the precipitation occurs as snow. Thundershowers provide much of the pre-cipitation during the sumar. High air pollution potential (atmospheric stagr,stion) is expected to exist, on the average, four days during the year. Atmospheric dispersion rates are expected to be about average for all sites in the United States.
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2-6 The-0hio River Valley, on the south bank of the Ohio River, is approximately one mile wide in the vicinity of the site and is rather sharply defined, with the tops of ,
the nearby ridges generally being 400 to 500 feet above the valley bottom. The steep sided valley tends to affect the local meteorological conditions by channeling the wind
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so that the predominant wind directions are oriented along the axis of the valley. At J night, cold air drains down the valley walls and collects on the valley bottom, resulting -
in a relatively high frequency of inversion conditions and fog formation in this portion of the Ohio River Valley.
High wind occurrences in the vicinity of the site are as'.0ciated mainly either with severe thunderstorms or large winter storm systems, although dissipating tropical storms or hurricanes occasionally affect the area. During the period 1871 through 1972, eight tropical storms or hurricanes have passed within 50 miles of the site. Six te'nadoes were reported within the one degree latituda-longitude square containing the site during the period 1955-1967, giving a mean annual frequency of 0.5 and a computed recurrence interval
- of 2575 years. The predominant wind flow over the site is from the northwest quadrant.
An onsite meteorological measurements program was initiated in April 1969. The pro-gram consisted of the installatica of and measurements from a 150-foot tower constructed on the site about 800 feet east-northeast ot the Beave, Valley Unit 1 containment build-ing, Temperature instruments were located at the 50 atd 150-foot levels on the tower and a set of wind measuring instruments was located at the 150-foot level. Tro sets of wind meas #ng instruments were installed at the 50-foot level; one is a fast-rerponse system
- Wactured by Packard-Bell (now Taledyne-Geotech) ard the other is a more rugged system manufactured by Bendix-Ft icz that, heuever, is less sensitive and had a higher starting speed.
He examined the data from both of these sets of wind measuring instruments and con-cluded that acceptable correlation existed between mea?urements made from the two sets.
Therefore, since the joint data recovery rate for the fast-response wind set (which could better represent onsite meteorological dispersion conditions) was only 67".. the applicants supplemented these cata with wind data from the Bendix-Friez instruments for those periods of time during which the Packard-Bell instruments were inoperable in order to obtain data with at least a 90% joint data recovery rate. We concur with the use of these data from both sets of wind measuring instruments.
The applicants have submitted a one year period of data record (September 1970 to September 1971), in joint frequency form, to previoe a basis for our evaluation of atmospheric dispersion conditions. For releases from the 158-meter (518-foot) above grado vent installed on the cooling tower, the joint frequency o distribution of wind direction and speed measured at the 150-foot level, and vertical temperature difference (delta-T) between the 150- and 50-foot levels ' u used. For nur estimates of routine releases from buildings and other vents, the composite joint frequency distribution of wind direction and speed measured by the two sets of instruments at the 50-foot level, reduced to represent wind speed at the 3.1-foot level (the height assumed for ground-level release calculations), and vertical temperature difference between the 150 and 50-foot levels was used. The joint
- The recurrence interval is computed by the method described in the paper by H.C.S. Thom,
" Tornado Probabilities," Mor.thly Weather Review. October-December 1963, pp 730-737.
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dan red,very durinq the one year period of record was 82 percent at the 150 foot level l' t.nd 97 percent. for the composite data, at the 50-foot level.
Utilizing standard staff practices. we evaluated the meteorological diffusion charac-1
.teristics of the site for both accident analysis and routine release analysis purpotes.
l's The evaluation of the calculated offsite doses resulting from radioactive releases l due to postulated' accidents requires calculations of the relative concentration (X/Q) for' the first 30 days following an assumed accident. The impact of routine radioactive releases requires calculations of an annually averaged X/Q.
Accident dose analyses utilize calculated X/Q values which vary with time. The staff uses its most conservative assumptions when calculating the X/Q values for the first II eight hours following an assumed accident. Additional credit is given for diffusion and
. spread of the gaseous plume for time periods beyond the first eight hours.
In our evaluation of diffusion rates for calculating short-term (0-2 hour at the i- site boundary and 0-8 hour at the low population zone) accidental releases from the buildings and vents, a ground level release with a building wake factor (cA) of_800 i square meters was assumed. Using the diffusion model described in AEC Regulatory
' Guide 1.4. the relative concentration (X/Q) for the 0-2 hour time period which is exceeded 5% of the time was calculated to be 1.3 x 10-3 seconds / cubic meter at the minimum site boundary distence of 610 meters, This relative concentration is equivalent to dispersion conditions produced by Pasquill Type F stability with a wind speed of 0,5 meters /second. The relative concentration which is exceeded 5% of the time at the outer boundary of the low population zone (5795 m) was calculated to be 1.3 x 10~4 seconds / cubic meter fo' the 0-8 hour time period. At the low population zone the estimated relative concentration. 'in seconds per cubic meter,' is 1.3 x 10-5 for the 8-24 hour time period. 6.7 x 10-6for the 1-4 day period, and 2.4 x 10-6 for the 4 30 day period.
The applicants' relative concentration (X/Q) estimates were less conservative than those of the staff by a factor of 1.67 for the 0-2 hour time period. For the other time periods, the applicants' values were less conservative than those of the staff by about a factor of two to three. These differences are attributed mainly to the applicants' use only of the wind data collected using the Packard-Bell wind system (which had a diata -
recovery rate of only 67 percent) and clso to the use of different mathematical and meteorological models.
Long-term diffusion estimates were made using the procedure described in Appendix 8 to AEC Regulatory Guide 1.42. The highest offsite annual average relative concentration of 2.6 x 10-5 seconds per cubic meter for releases from vents, other than the vent on the cooling tower, would occur at the site boundary northeast of the reactor complex. For releases from the 158 meter (518-feet) above crade vent on the cooling tower, the highest offsite annual average relative concentration of 1.4 x 10-6 seconds per cubic meter would occur at a distance of two miles north-northeast of the cooling tower. A slightly lower annual average relative concentration of 1.3 x 10-6 seconds per cubic meter would occur southeast of the cooling tower at the same distance.
We have concluded that the meteorological data presented in the FSAR provide an acceptable basis to make conservative estimates of atmospheric diffusion for accidental and routine (,aseous "eleases from the plant.
2.4 Hydrology The site is on the south bank floodplain of the Ohio River. The approximately 23.000 square mile upstream drainage area is regulated by several reservoirs located on 9
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2-8 the mainstream of the Ohio River and its tributaries. Plant grade is at elevation 730
) feet above mean sea level (MSL) or above. The nonnal river level is approximately 664 =
- feet above MSL. . The only safety-related structure subject to potential flooding is the l
river water intake structure, which is on the bank of the river.
' Although historical flood records date back to the year 1762 in the region, contin-uous records in the site area date back only to the early 1930's. The worst historical
- i. flood of record occurred in March 1936 and produced an estimated maximum runoff rate of about 623,000 cubic feet per second (cfs), which corresponds to a water level of about 703 feet above MSL. The lowest instantaneous flow of record occurred in August, 1930 at a flow rate of 1,250 cfs. Both high and low flows are now regulated by an extensive system of dams that provide considerable flood control and low flow augmentation capability.
Three sources of flood potential were analyzed by tne app' licants, runoff-producinq floods on the Ohio River, runoff floods from site drainage, and seismically-induced floods. In addition, we also considered the potential for ice indaced flooding.
A precipitation-induced probable maximum flood (PMF) estimat2 on the Ohio River was developed for our ceview of Beaver Valley Unit 1 at the construction permit stage and found acceptable. The PMF was estimated to produce a maximum runoff rate of 1,500,000 cfs and a water level of 730 feet above MSL. This estimate was again reviewed during our current review at the operating license stage and was found to be acceptable.
However, adequate provision for wind-generated waves that could occur coincidentally with a PMF had not been considered in estimating the design basis flood. Consequently, at our request, the applicants estimated that the maximum coincident wave would be five feet high. The maximum wave is defined as tie statistical average of the highest one percent of the waves t' expected in a tyr, cal wave train.
The only safety-related structure thr4 would be affected by wave action would be the river water intake structure. In the analysis of the requirea flood protection for the additional wave-induced increment, the applicants determined that the wave action would not exceed the structural design bases for the intake structure. The applicants have, however, provided permaner4 reinforced concrete ventilation air intake extensions on top of the intake structure and portable ventilation exhaust chimneys for attach-ment to the exhaust slots inside the intake structure. The implementation of flood pro-tection also requires the closure of flood doors and inflation of pneumatic door seals in the intake structure pump compartments.
Floods which could be caused by seismically-induced upstream dam failures have been analyzed by both the applicants and the Regulatory staff. The applicants have concluded that the worst case of such an arbitrarily assumed failure, even when postulated coin-cident with a flood about half as severe as a PMF, would not produce water levels as o severe as a PMF. We have considered flood waves which could result from arbitrarily assumed seismically-induced dam failures coincident with both a flood producing about half the PMF runoff rate, and the simultaneous failure of more than one dam coincident with a relatively frequent flood, using the criteria suggested in Appendix A to Regulatory Guide 1.59. In both cases we concluded that a PMF would establish the controlling Ohio River flood level at the site.
The possibility of rapid river rises, during floods less severe than the PMF, was also considered. Such rises could be caused by dam failures or, more likely, by heavy rainfall over only portions of the upstream drainage area after general basin rainfall.
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.. 4 2-9 To provide assurance that flood protection will be effective for the river water intake structure-(the only safety-related structure that would be affected) we have taken the position that technical specifications limiting plant operation would provide for such situations. Accordingly, the applicants have proposed a technical specification which provides for flood protection in the event of rapid river rises as well as for runoff floods up to and including a PMr. He find this acceptable.
Site drainage includes the hillside drainage south of the plant, the plant area it-self, and Peqqs Run which parallels the highway road fill just east of the plant between the highway and cooling tower area. Although the design basis selected for site drain-age is substantially less severe than would be produced by local probable maximum preci-pitation, the ground in the plant area slopes toward the Ohio River and Pegns Run, and' runoff in excess of stom drainage inlet and piping capacity is not expected to cause water levels greater than a few inches above the ground surface. The applicants have concluded, and we concur, that such levels should not constitute a flood threat to safe-ty-related facilities.
Peggs Run is cc,nstricted in a deeply incised channel between the highway embankment asd the cooling tower area at elevations as low as about 670 feet above MSL. A struc-ture across Peggs Run abnut 700 feet east of the proposed location of Beaver Valley Unit 2 containment provides the base for the railroad track strean crossing servir.g the plant area. The railroad crossing and grade in the area is sufficiently below plant grade such that Peggs Run flooding from severe storms, even with the railroad structure waterway blocked, should not reach safety-related structures located to the west and on hiaher ground.
Another potential source of plant fluoding of safety-related features is associated with the failure of the roofs of buildings. At our request. the applicants have pro-vided design bases for roof drainage systems that will prevent rainfall accumulations from exceeding the structural design bases of the roofs of safety-related buildings during stoms as severe as a local probable maximum storm.
We have also analyzed the potential for ice flooding cased upon historical recoros in the region 8.id consideration of local topography. Although such flooding is possible at the site, the PMF is the controlling flood and was used in establishing the design flood level for the site.
Water will'be drawn from the Ohio River through the intake structure for plant cool-ing purposes. This includes make-up to the cooling tower basins for plant operation (non-safety-related).andforsafety-relatedpurposes.
A sill at elevation 646 feet above MSL will limit the river water level for.which water can be supplied to the intake strue.ture sisnp for plant use. The applicants have proposed a minimum design river level at elevation 649 feet above MSL based on a 4,700 o
cfs minimum river flow coincident with an arbitrarily postulated downstream dam failure.
The minimum safety requirements are an emergency river water system flow of 9,000 gallons per minute (20 cfs) coincident with fire pump demand of 2,500 gallons per minute (6 cfs). 4 In examining the analysic of potential low river flows, it was detemined that the 4,700 cfs minimum river flow assumed by the applicants was predicated on the ability of upstream reservoirs to augment low river flows. Since the dependability of such augmen-tation is based on others providing reservoir storage for approximately a drought of l record, we and our consultant (Nunn. Snyder and Associates) analyzed the potential river flow that could be expected under very stvere drought conditions.
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2+10 We-estimated that the minimum river flow rate could be as low as 800 cfs. This flow rate is considered adequate for safety-related plant water supply. We had originally )
determined that such a low flow could result in a river level well below the minimum '
design level selected by the applicants. On this basis we informed the applicants that we required an analysis to demonstrate the long-term residual heat renoval capability for river flows as low as 800 cfs, the lowest river flow considered reasonable possible of occurring. We have reviewed and concur with the applicants' analysis which shows that under reasonably severe river conditions that could be expected with flows as low as 800 cfs the river water level can be maintained above the minimum design level selected by the applicants.
The potential for channel diversion or blockage of the Ohio River such that safety-related water supplies would not be available is not considered credible because of the lack of adverse floodplain topography along the river where diversion or blockage could adversely affect the plant.
Operating procedures employing the use of technical specifications to limit plant operations are required for severe floods to assure the operability of safety-related l equipment. At the request of the staff, the applicants have provided the bases for a technical specification that contains the following elements: (1)provlsionsfora flood alert at river levels of 690 feet MSL or above, (2) provisions for plant operating ,
personnel to maintain contact with operators et upstream dams. (3) immediate plant shut-down to be undertaken and protection of required safety-related equipment to be initiated at rising river levels above 695 feet MSL. and (4) a detailed emergency procedure to be developed fer such situatic,ns.
The site is on a predominantly permeable sand and gravel terrace, underlein by bed-rock at about elevation 630 feet above MSL. Although some ground water migrates in bed-ding and joint planes and some permeable seams in the bedrock, the major movement of ground water at the site is through the permeable surface terrace materials toward a direct hydraulic connection with the river. Ground water levels, because of the direct hydraulic connection. can te expected to be slightly higher than river levels, except where influenced by well pumping. The on'y wells within the influences of the operating plant are two wells at the adjacent Shippingport Atomic Power Station and two temporary Beaver Valley plan't construction wells.
We have concluded that acceptable flood design bases have been pro ided, that an acceptable water supply can be assured for safety-related purposes, and that ground water flow is not intercepted by any wells beyond the control of the applicants before teaching the Ohio River. Acceptable bases for technical specifications which limit plant operation during high river water levels have been incorporated in the t.pplication, and acceptable hydrologically-related design bases for the auxiliary river water system (furthur discussed in Section 9.3.4 of this report) have been established.
2.5 Geoloqy and Seismology We and our advisor, the U.S. Geological Survey (USGS) reviewed the geology of the site as presented in the PSAR and its amendments for Unit ? at the construction permit stage of our review, and corTared this information with the available literature. The USGS stated, and we concurred, that the anelysis appeared to be carefully derived and to present an adequate appraisal of those aspects of the geology that would be pertinent to an engineering evaluation of the site.
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r I: We and' our USGS advisor both concluded that there are no known active faults or other geologic Ttructures in the area that could be expected to localize seismicity in the
- j. inmediate vicinity of the site. ' A ground acceleration of 0.12 g was used for the safe shutdown earthquake, based on an intensity VII earthquake on the Modified Mercalli scale. An acceleration of 0.06 g was used as the operating basis earthquake.
During the construction permit stage of our review' of Unit 2 (construction permit -
issued on May 3,1974) which was conducted concurrent with this review at the operating license stage for Unit 1, we again reviewed the geology and seismology of the Beaver Valley site. On the basis of that review we confirmed our previous conclusion that the -
geological and seismological aspects of the site are acceptable.
With regard to foundation engineering, the plant structures, except for the river water intake structure and river water system lines, are founded on the sand and gravel of the older terrace deposits. Cherever the upper, looser granular soils of the older terrace deposits, or tree younger deposits of silt and clay were encountered at or below-founding levels, they were removed and replaced with an engineered granular. fill to founding grade. We determined that the upper terrace materials will be stable under all static and dynamic design conditions appropriate for the site.
The river water intake structure and river water system lines for the plant are located in the low terrace which is recent alluvial valley fill. The low terrace is made up of soft clays and clayey silts near the surface, underlain by loose to medium dense sands and gravels down to bedrock. D the course of our review of Beaver Valley Unit 1 at the construction permit stage, we concluded that the low terrace materials have the potential to liquefy under safe shutdown earthquake vibration. To mitigate the effects of liquefaction on the river water system lines, we required the applicants to densify the foundation soils under the lines to 75% relative density, which wa con-cluded would be adequate to remove the potential for liquefaction of the low teNa materials in the event of a safe shutdown earthquake.
During our review of Unit 2 (at the construction permit stage of review) we deter-mined that, although the densified soils under the service water lines would remain stable in the event of a safe shutdown earthquake, the non-densified soils in the general area, if in a state of liquefaction, could flow in such a manner as to block the intake structure and/or clog the river water pumps.
As a result of our review, we required the applicants to either (1) demonstrate that the low terrace soils, if tt ey do liquefy, do not pose a hazard to the continued operation of the river water ' stem in the event that a safe shutdown earthquake occurs, or (2) modify the proposed design to remove this hazard. The applicants informed us that, while they did not expect liquefaction to lead to the conditions postulated, they h ve densified additional soils in the area of the intake structure to protect the o intake structure from such a condition. We have reviewed the modification to the soils and the method of densification, and have concluded that densification is an appropriate method to remove the potential for liquefaction.
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% l l 3-1 i l 1 3.0 DESIGN CRITERIA FOR STRUCTURES, COMPONENTS, EQUIPfiENT AND SYSTEMS 3.1 Confonnance with AEC General Design Criteria Beaver Valley Unit 1 was designed and is being constructed on the basis of the pro-l posed General Design Criteria which were published in the Federal Register on July 11, 1967. Design and construction were initiated and proceeded to a significant extent 3 based upon the criteria proposed in 1967. Since February 20, 1971, when the Conmission published the General Design Criteria for Nuclear Power Plants. Appendix A to 10 CFR Part 50, the applicants have attempted te comply with newer criteria to the extent practical, recognizing previous design coannitments. As a result, our review assessed the plant against the General Design Criteria now in effect, and we have concluded that the plant j design confonns to the intent of these newer criteria.
3.2 Classification of Structures. Components, and Systems The applicants have identified in Table B.1-1 of the FSAR those seismic Category 1 structures, systems and components important to safety that are designed to withstand the effects of the safe shutdown earthquake and remain functional. These features are those necessary to assure (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintair, it in a safe shutdown condition, and (3) the capability to prevent er mitigate the consequences of eccidents which could result in potential offsite exposures comparable to the guideline a:xposures of 10 CFR Part 100.
All other structures, systems and components that may be required for operation of Unit 1 are designed to other than seismic Category 1. Included in this classification are chose portions of Category 1 systems which are not required to perform a safety function. Structures, systems and components important to safety that are designed to withstand the effects of the safe shutdown earthquake and remain functional have been identified in an acceptable sanner. We conclude that the design of these items as seismic Category 1 provides reasonable assurance that, in the event of an earthquake, Unit I will perform in a panner providing adequate safeguards to the health and safety of the public.
The applicants have applied a Quality Group Classification System to those water and steam containing components which are part of the reactor coolant pressure boundary and other fluid systems important to safety where reliance is placed on these systems: (1) to prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary. (2) to permit s'tutdown of the reactor and main-tenance in the safe shutdown conditions, and (3) to contain radioactive N terial.
For those fluid systems identified in the applicants' classification groups, we and the apf.'.: ants are in general agreement on the application of the Quality Group Classi-fication System. The applicants have identified in the applicable FSAR subsections those fluid systems or portions of fluid s.vste.m.s important to safety and the industry codes and standards applicable to each pressure-containing component in the systems. Piping and Instrtsnentation Diagrams in the FSAR identify the boundary limits of each piping classi-fication group within the fluid systems. Pressure-retaining components in fluid systems I
i 4
, t, 3-2 within the boundaries of the applicants' piping Group Classification Q1, Q2, and Q3 are bui H to meet the requirements of the applicable codes. Conformance with such codes is an acceptable basis for meeting the requirements of Criterion No.1 of the AEC General Design Criteria and provides reasonable assurance that the plant will perform in a manner providing acceptable safeguards to the health and safety of the public.
3.3 Wind and Tornado Loadings All the plant seismic Category I structures that will be exposed to wind were designed for a wind velocity of 84 mph, based on a 100-year recurrence interval. The containment structure, the auxiliary building which encloses the control room, the fuel building and all other seismic Category I structures needed for the safe shutdown of the plant were designed to resist a tornado of 300 mph tangential wind velocity and a 60 mph transnational wind velocity. The simultaneous atmospheric pressure drop was assumed to be 3 psi in 3 seconds. In addition, appropriate postulated tornado missiles have also been factored in the design (see Section 3.5). ' Wind pressures, shape factors, gust factors, and variations of wind velocities with height were determined in accordance with the American Society of Civil Engineers Paper No. 3269, " Wind Forces on Structures."
Structures are arranged on the plant site (or protected) in such a manner that a collapse of structures not designed for tornadn forces will not affect those designed for tornado forces.
The criteria used in the design of seismic Category I structures to account for the loadings due to specific winds and tornadoes postulated to occur at the site, and the method used to determine those loads provided a conservative basis for the plant design.
The use of these loading criteria provides reasonable assurance that, in the event of wind or tornadoes, the structural integrity and safety function of seismic Category I structures will not be impaired by the specified environmental forces.
We have concluded that conformance with these criteria is an acceptable basis for satisfying the requirements of Criterion No. 2 of the AEC General Design Criteria.
3.4 Water Level (Flood) Design All seismic Category I structures were designed to withstand a localized probable maximum precipitation (PMP) stom as well as that condition or combination of conditions that produces the maximura water level at the site. Section 2.4 of this report presents the details of assessing the site flood level which includes the probable maximum flood level (PMF) of 730 feet above mean sea level (MSL). All seismic Category I structures are designed for buoyancy and hydrostatic pressures associated with the IMF flood level.
The use of these design loading criteria provides reasonable assurance that, in the event of flooding, the seismic Category I structures can be expected to withstand the specified environmental forces without impairment of their structural integrity and safety function, o We have concluded that conformance with these criteria is an acceptable basis for satisfying the requirements of Criteria Nos. 2 and 4 of the AEC General Design Criteria.
With regard to the flood protection of safety-related equipment in the intake struc-ture, service building, and the auxiliary building, we have informed the applicants of our concerns about the adequacy of the design in the following areas:
(1) The adequacy of the pneumatic door seals for the sliding flood doors in the intake structure. We have informed the applicants that we require additional assurance that the design is adequate. We have suggested that additional tests and inspections be performed during the life of the plant as well as during the
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initial installation. These tests should be perfomed to determine the i-
' acceptable air.teakage rate from the pneumatic seals as a, function of the air '
l supply inventory and the amount of time that the seals will be required to operate during extreme flood conditions.
(2) The provisions for testing the electrical penetrations in the plant below the PMF elevation of 730 feet above MSL. We have infomed the applicants that we require a program which prcvides for periodic testing / inspection of these penetrations at periodic intervals throughout the life of the piant as well as during the initial installation.
(3) The qualification and testing of electrical cables installed in ducts below the PMF elevation of 730 feet above MSL, particularly the cables between the diesel-generator structure and the service building, and the cables that provide power -
to safety-related equipment in the intake structure. We have-informed the applicants that we require a program of testing / inspection to verify that these .
cables, including any splices in the cable runs, are qualified for the intended serv' ice and maintain their integrity throughout the plant life.
We will review the applicants' program for testing and inspection regarding the items discussed above, and will report the results of our review in the supplement to this report.
The applicants have made provisions to protect vital equipment in the service building by providing for curbing and sump pumps to collect and remove normal amounts of seepage and leakage. The applicants have also provided for sump pumps (powered from the emergency buses) in the cubicles at the intake structure which house satety-related equipment, to remove normal amounts of seepage and leakage past the flood doors.
We find that these design modifications provide sufficient assurance that normal amounts of leakage and seepage under extreme flood conditions can be adequately removed.
3.5 Missile Prottetton The seismic Category I structures and components are shielded from, or designed for, the various postulated missiles, including tornado generated missiles and containment
-internal missiles, such as those associated with a loss-of-coolant accident. (Contain-ment internal missiles are discussed in Section 5.2.6.1 of the FSAR). The critical tornado-generated missile used in the design of seismic Category I structures was a wooden utility pole 35 feet long, traveling at 150 miles per hour. The design of structures for missile impact was performed in accordance with Bureau of Yards and Docks, Department of the Navy, NAVDOCKS, P-51,1950 " Design of Protective Structures",
by A. Amirikian.
The use of these criteria for protection from postulated missiles provides assurance that resulting loads and effects will not impair the structural integrity of seismic Category I structures, or result in any loss of function of safety related systems and components contained in such structures.
We have concluded that conformance with these criteria is an acceptable basis for satisfying Criteria Nos. 2 and 4 of the AEC General Design Criteria.
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- 4 3-4 3.6 Protection Aaainst Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.1 Protection Aqainst Dynamic Effects Associated with the Postulated Rupture of Piping Inside Containment The design criteria for protection against the dynamic effects of pipe ruptures and the resulting discharging fluid provide that, in the event of the occurrence of the combined loadings imposed by an earthquake of the magnitude specified for the safe shutdown earthquake and a concurrent single pipe break of the largest pipe at one of the design basis break locations inside containment, the following conditions and safety functions will be accommodated and assured:
(1) The magnitude of the design basis loss-of-coolant accident cannot be aggravated by potentially multiple failures of piping.
(2) The reactor emergency core cooling systems can be expected to perform their intended function.
Pipe motion subsequent to rupture and the pipe restraint dynamic interaction was analyzed by the use of an elastic-plastic lumped mass beam element model, sufficiently detailed to reflect the structural characteristics of the piping system.
The design criteria used for identifying high energy fluid piping and for postu-lating pipe break locations (inside containment) and flow areas are consistent with the criteria set forth in AEC Regulatory Guide 1.46, On the basis of our review, we have concluded that the criteiia used for the identification, design and analysis of piping systems inside containment, where postu-lated breaks may occur, constitute an acceptable design basis for satisfying the appli-cable requirements of Criteria Nos.1, 2, 4,14, and 15 of the AEC General Design Criteria.
3.6.2 Protection Aaainst Dynamic Effects Associated with the Postulated Rupture of Piping Outside Containment In December 1972, we requested the applicants to assess the consequences of postu-lated failuret of high energy fluio piping outside of containment, including failure of the m&ln steam and feedwater lines. The applicants have conducted an assessment for Beaver Valley Unit I utilizing criteria and guidelines provided by the staff. The basic criteria require that:
(1) Protection be provided for equipment necessary to shut down the reactor and main-tain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of protection equipment, from all effects resulting from ruptures in pipes carrying high-energy fluid, up to and including a double-ended rupture of such pipes, where the temperature and pressure conditions of tne fluid exceed 200'F and 275 psig. Breaks should be assumed to occur in those locations specified e in the staff pipe whip criteria. The rupture effects on equipment to be considered include pipe whip, structural (including the effects of jet impingement), and environmental.
(2) Protection be provided for equipment necessary to shut down the reactor and main-tain it in a safe shutdown condition, assuming a concurrent and unrelated single active failure of protection equipment, from the environmental and structural effects (including the effects of jet impingement) resulting from a single open crack at the most adverse location in pipes carrying high-energy fluid routed in 9
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3-5 the vicinity of this equipment, where the temperature and pressure conditions o' f
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the fluid exceed 200'F and 275 psig. The size of the cracks should be assumed to be 1/2 the pipe diameter in length and 1/2 the wall thickness in width.
The applicants responded to our request by submitting Appendix D to the FSAR on September 20. 1973, which describes their findings and the resultant plaat modifica-tions that will' assure that Beaver Valley Unit 1 <can accommodate the effects of postu-lated high energy piping breaks. The applicants have identified all high energy piping and indicated the routing of piping through the plant, have identified the equipment required for safe shutdown and described their location with respect to high energy piping; have identified the ventilation equipment necessary to cool the safety related equipment; and have committed to the following actions as a result of their assessment:
(1) Six inch curbs were installed on all floor penetrations' at the 735.5-foot floor elevation of the service building.
'(2) Pipe whip restraints together with energy absorbing material have been installed 'i J
in the pipe chase area in the service building.
(3) Temperature monitoring equipment has been installed in the pipe tunnel at the cable vault area as well as alarms.in the control room to aid the operator in identifying and terminating flow due to a steam generator blowdown line break.
(4) Spray shields have been provided for piping in the area of safety-related electri-cal equipment'in the auxiliary building.
(5)_Inthevalvecubiclearea.thefollowingwillbeprovided:
.(a) A missile proof roof to allow existing openings to be fitted with louvered panels to provide adequate vent area.
-(b) Main steam line restraints at the cubicle wall along with structural modifica-tions to restrain valve impact on the wall.
-(c) The leak exhaust in the valve cubicle room has been reinforced to prevent failure in the event'of a break in the room.
(d) Stiffening of the side walls of the cubicle against jet impingement.
(e) Feedwater line restraints.
(f) Flood dams in penetrations at the 751-foot floor elevation.
(g) Temperature monitoring equipment and alarms in the auxiliary feedwater pump room area.
As a result of our review of Appendix 0 to'the FSAR. we required the following additional verification and/or justification of the design:
(1) Verification that the material used to make the honeycomb segments of the main steam and feedwater line axial restraints is capable of withstanding the cyclic .
temperature variations that it will be subject to over the life of the restraint.
(2) Verification that all safety related electrical equipment that must function .
O following postulated events and which will be subject to steam environments will be environmentally qualified by test prior to installation in the plant.
(3) A description of the metns ti which the functional ability of required temperature detectors will be assured over ts.e life of the plant.
(4) Verification that cooling wat*r lines and all related wiring to safety features have been included under the heading of " targets" in the analysis.
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3 (5)' Justification that jet impingement forces on the external north wall of th'e valve-ewbicle, due to a circumferential rupture of the main steam line at the wall' will ,
not cause structural collapse of the wall,' including consideration of the possi-bility that the whipping main steam line may cause the failure of an adjacent feedwater line. thereby imposing additional . loading on the wall.
In a letter dated January 16, 1974; and in Amendment 7 to the FSAR, the applicants-submitted statements in response to the above additional requests. On the basis of our review we.have concluded that the design of the plant will be in accordance with our s
' required design criteria for hit,h energy piping systems, and will provide acceptable protection of safety-related equipment from the offects of high energy pil,e ruptures outside containment.
The protection provided against the dynamic effects of postulated pipe breaks and discharging fluids in piping systems containing high energy fluids and located outside the containment is acceptatie to prevent damage to structures,' systems and components to the extent considered necessary to assure the maintenance nf their structural integri ty. Such protection provided reasonable assurance that the' safe shutdown of the reactor can be accomplished and maintained, as needed.
The criteria used for the identification, design and analysis of high energy fluid lines where postulsted breaks nay occur (outside containment) is consistent with the criteria contained in the December 18,1972 and January 31,1973 Commission letters to .
the applicants, and therefore constitutes an acceptable design basis for. satisfying the applicable requirements of Criterion No. 4 of the AEC General Design Criteria.
3.7 Seismic Design The seismic design response spectra curves were presented in the PSAR and' approved prior to the issuance of the construction permit for Beaver Valley Unit 1. The modified earthquate time-histories used for component and equipment design were adjusted in .
amplitude and frequency to envelope the response spectra specified for the site. We conclude that the seismic input criteria used by the applicants provide an acceptable basis for seismic design.
Modal response spectrum multi-degree of-freedom and normal mode-time history methods were used for the analysis of all seismic Category I structures, systems and components.
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 spectrum method was used.
The absolute sun of responses was used for closely spaced frequencies. A hori,tontal and vertical floor spectrum method was used. The absolute sum of responses was used for those responses having closely spaced frequencies. Horizontal and vertical floor spectra inputs used for design and test verification of structures, systems and compon-g ents were generated by semi-empirical methods and confirmed by the nonnal mode-time history nethod. Vertical ground accelerations were assumed to be two-thirds of the horizontal ground accelerations for items rigidly attached to structures. Constant vertical load factors were employed only where analysis showed sufficient vertical rigidity to preclude significant vertical amplifications in the seismic system being analyzed.
We have reviewed the information presented in the FSAR and have concluded that the )
seismic system and subsystem dynamic analysis methods and procedures used by the appli-cants are acceptable. l
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The type, number, location and utilization of strong motion accelerographs to record seismic events and to provide data on the frequency,' amplitude' and phase relationship of.'
the seismic response of the containment structure conforms to the recommendations of-AEC Regulatory Guide 1.12.
At our request, the applicants have aareed to provide multielement seismoscopes for measuring the input response spectra at selected locations including one in the basement.
- of the reactor containment structure. Supporting instrumentation will be installed on Category I. structures,' systems, and components to provide data for the verification of the selsmic responses determined analytically for such Category I items.
He conclude that the seismic instrumentation program es* ' *1shed by the applicants '
is acceptable.
3.8 Design of Seismic Category I Structures The reactor coolant system is enclosed in a reinforced concrete containment, as.
described in Section 5.2 of the FSAR.
The containment has a flat base, o cylindrical wall and a hemispherical dome. A-steel liner is attached to the inside of the containment vessel. The containment structure was designed in accordance with applicable sections of the ACI-318 code for concrete, and the ASME Boiler and Pressure Vesul Code, Section.III, for steel parts.
The containment was designed to resist various combinations of dead loads, live loads, environmental loads, including those due to wind, tornadoes, and the operating basis earthquake and safe shutdown earthquake, and loads generated by the design basis loss-of-coolant accident-including pressure, temperature and associated pipe rupture effects.
The structural design criteria that were used are the same as those previously reviewed and found acceptable.
The static analysis for the containment shell was based on the thin shell theory with elastic material behavior. Results from the base slab analysis wert used as boundary conditions for the shell analysis. The liner design for the containment is similar to that reviewed and approved in previously licensed plants. The choice of the materials, the arrangement of the anchors, the design criteria and design methods are similar to those evaluated for previously licensed plants. The stresses computed by the applicants are below the code allowable. - Materials, construction methods, quality.
assurance and quality control measures were acceptably covered in the FSAR and, in general, are similar to those used for previously accepted. facilities.
Prior to operation, the containment will be subjected to an acceptance test in accordance with AEC Regulatory Guide 1.18 during which the internal pressure will be 1.15 times the containment design pressure of 45 psig. The liner welds have been tested in accordance with AEC Regulatory Guide 1.19.
AEC Regulatory Guides 1.10 and 1.15 have also been applied in the construction or !
o testing of the structural parts of the plant. The use of these design criteria defining the applicable codes, the loads and loading combinations, the design and analysis pro-cedures, the materials, quality control and the testing provide reasonable assurance that, in the event of winds, tornadoes, earthquakes and various postulated accidents (occurring within containment), the seismic Category I containment structure will with-stand the specified conditions without impairment of its structural integrMy and safety function. We have concluded that conformance with these criteria constitutes an acceptable basis for satisfying the requirements of Criteria Nos. 2, 4,16, and 50 of the AEC General Design Criteria.
l4 , fej H l- 3-3 Ibe containment interior structures consist of a shield wall around the reactor, secondary shield wall a d other. interior walls, compartments.and floors. The interior-structures were designed in accordance with the ACI-318 Code.for: concrete and the AISC J specifications for structural steel.
The applicants have considered all the loads which nay act on the structure during .
its lifetime..sach as dead and live loads, accident induced loads'(including pressure and jet loads) and seismic loads. The load combinations used cover ell cases likely to occur and include all loads which may act simultaneously.- In the design of concrete interior structures, the ultinate strength design method was used. An elasto-plastic analysis was conducted 'on the steam generator cubicle walls to assure that they have sufficient ductility to absorb the energy of the jet impingement forces induced by : -
postulated pipe ruptures.
The use of these design procedures and criteria provides reasonable assurance that the containment internal structure will withstand all the specified design loads-(includ-ing those due to orthquakes and various postulated accidents occuring within the con-tainment)withoutimpairmentofthestructuralintegrityandsafetyfunction.
The foundation of the containment is a circular met and was analyzed to determine the effects of various combinations of loads expected during the life span of the plant.
Analysis has been accomplished by means of a digital computer taking into account the bending moments, shears, and soil pressures for a symmetrically loaded circular plate .
on an elastic foundation. The foundation was designed in accordance with the ACI-318.
Code, E
i Conformance with these criteria constitutes an acceptable basis for satisfying the I" requirements of Criteria Nos, 2, 4,16 and 50 of the AEC General Design Criteria.
l Seismic Category I structures other than containment and its interior structures l were built from structural steel and reinforced concrete members. The structural com-ponents consist of slabs, walls, beams and columns. The design method for reinforced '
concrete followed that specified in the ACI-318 Code. Structural steel components were l
desi5ned in accordance with the AISC specifications.
- The various conditions used in the design of these seismic Category I structures (other than containment) include appropriate combinations of loads likely to occur -
. during normal operation or shutdown, and during postulated accidents and earthquakes.
These combinations are similar to those previously reviewed and found acceptable.
Foundations of other major structures,'such as the fuel building, auxiliary building and main control area in the service building consist of reinforced concrete mats and l
were desige d in accordance with the ACI-318 Code.
The use of these design criteria provided reasonable assurance that seismic Category I structures will withstand all the specified loads without impairment of their structural integrity and safety functions. Conformance with these requirements constitutes an . acceptable basis for satisfying the requirements of Criteria Nos. 2 and .
4 of the AEC General Design Criteria.
3.9 Mechanical Systems and Components The applicants will conduct a piping vibration operational test program on all ASME Class 1 and Class 2 piping systems and piping restraints during startup and initial operating conditions to verify that the piping and piping restraints of the system have been designed to withstand vibrational dynamic effects due to valve closures, pump trips, i.-
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3-D and operating modes associated with the design operational transients. The tests, as planned 7will develop loads similar to those experienced during reactor operation and are consistent with recent Regulatory staff positions concerning preoperational piping dynamics effects test programs. We have concluded that compliance with this test pro-gram constitutes an acceptable basis for fullfilling the requi-ement of Criterion No. 2 of the AEC General Design Criteria.
The dynamic testing and analysis procedures, which have been implemented to confirm that all seismic Category I mechanical equipment will function during and after an earthquake of magnitude up to and including the safe shutdown earthquake, and that all equipment support structures are adequately designed to withstand seismic disturbances are acceptable.
The process of subjecting the equipment and its supports to these dynamic testing and analysis procedures provides reasonable assurance that in the event of an earthquake at the site, the seismic Category I mechanical equipment will continue to function during and after a seismic event. and the combined loading imposed on the equipment and its supports will not exceed applicable code allowable design Stress and strain limits.
Limiting the stresses of the supports under such loading combinations provides an acceptable basis for the design of the equipment supports to withstand the dynamic loads associated with seismic events, as well as operational integrity.
We have concluded that implementation of these dynamic testing and analysis pro-cedures constitutes an acceptable basis for satisfying the requirements of Criteria Nos. 2 and 14 of the AEC General Design Criteria.
All seismic Category I systems, components and equipment outside of the reactor coolant pressure boundary have been designed to sustain normal loads, anticipated transients, the operating basis and the safe shutdown earthquake within design limits which are consistent with those outlined in AEC Regulatory Guide 1.48.
The specified design basis combinations of loading, as applied to the design of the safety-related ASME Code Class 2 and 3 pressure-retaining components in systems classi-fied as seismic Category I, provide reasonable assurance that in the event (1) an earthquake should occur at the site, or (2) other upset, emergency or faulted plant transients should occur during 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 construction. Limiting the stresses under such loading combinations provides a conservative basis for the design of the system compon-ents to withstand the most adverse combinations of loading events without gross loss of structural integrity.
We have concluded that the design load combinations and associated stress and deformation limits specified for all ASME Code Class 2 and 3 components are consistent 0 with recent Regulatory staff positions and consHNte an acceptable basis for design in satisfying Criteria Nos. 1, 2 and 4 of the AEC General Design Criteria.
The applicants have stated that all ASME Class 2 and 3 active valves were designed to function at normal operating conditions maximum design conditions and design basis acci-j dent conditions. The applicants have provided acceptable assurance of the capability of I ASME Code Class 2 and 3 active valves to withstand the imposed loads associated with normal, upset, emergency and faulted plant conditions without loss of structural integ-rity and to perform the " active" function (i.e., valve closure or opening) under i
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3-10 conditioEs and combination 4 - onditions comparable to those expected when a: safe plant shutdown is to be effecteo i consequences of an accident are to be mitigated.
The maximum full discharge loads resulting from the opening of ASME Code Class 2 safety af 4 relief valves were calculated by either an equivalent static analysis or a time response dynamic analysis of.the system. The maximum stress intensities and stresses resultinq from these loads were calculated in accordance with Subsection.
'NC-3600~of the AS!1E. Code Section 111.- In the case of open safety or relief valves, i- mounted on a common header and full discharge occurring concurrently, the additional,
[ stresses induced in the header were combined with.previously computed local and primary membrane stresses to obtain the maximum stress intensity.
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.The criteria used in developing the design and mounting of the safety and relief valves of ASME Code Class 2 systems provides acceptable assurance that,' under discharg--
ing conditions, the resulting stresses are expected not to exceed the allowable design stress and strain limits for the materials of construction. Limiting the stresses'under the loading combinations associated with the actuation of these pressure relief devices .
provides a conservative basis for the design of the system components to withstand these loads without loss of structural integrity and impairment of the overpressure protection function.
i The criteria used for the design and installation of overpressure relief devices in ASME Code Class 2 systems constitute an acceptable design basis in meeting the applicable requirements of Criteria Nos.1, 2. 4,14 and 15 of the AEC General Design Criteria and are consistent with recent Regulatory staff positions.
3.10. Seismic Qualification of Category I Instrumentation and Electrical Equipment A seismic qualification program was implemented to confim that (1) all seismic.
Category I instrumentation and electrical equipment will function properly when sub-jected to the excitation and vibratory forces imposed by a safe shutdown earthquake and the conditions following a postulated loss-of-coolant accident, and (2) their. ,
support structures are adequately designed to withstand the seismic disturbances.
The program, as specified by the applicants in the FSAR, constitutes an acceptable basis for satisfying staff requirements and the applicable requirements of Criterion -
No. 2 of the AEC General Design Criteria.
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4.0 . REACTOR'
.4,1 General
' The reactor' design for. Beaver Valley Unit.1 is similar to that reviewed and approved for Surry Nuclear Plant Units -1 and 2 (Docket Nos. 50-280 and 50-281)
^' except for the following: (1) the design core thermal power level for' Beaver.
Valley Unit 1 is 2652 MWt compared to 2411 MWt for Surry and (2) Beaver Valley Unit 1 will use a 17'x 17 fuel assembly compared ko the 15 x 15 fuel assembly which was used for Surry.
Beaver Valley Unit 1 is compared to Surry Units 1 and 2 because both facilities are three loop plants with essentially identical physical configuration an' d reactor coolant flow rate.
4.2 Fuel Mechanical Design-The Beaver Valley fuel assembly consists of 264 fueled rods, 24 guide thimbles, one instrumentation thin.ble plus ancillary hardware arranged in a 17 x'17 array. The instru-mentation' thimble is at the center of the assembly and facilitates the insertion of-neutron detectors. ' The guide thimbles provide channels for inserting various reactivity control s. The fuel rods contain uranium dioxide pellets hermetically clad in Zircaloy-4.
The assembly is supported at both ends by stainless steel nozzles. Alignment and trans -
verse spacings are maintained by eight spacer grids located axially equidistant.
The Beaver Valley Unit 1 fuel assembly (17 x 17) is mechanically similar to the D.C. Cook Nuclear Plant fuel assembly (15 x 15). Those mechanical aspects which differ ~
~
from the previously used 15 x 15 fuel assembly are exhibited in Table 4.2-1. Fuel .
Mechanical Design Comparison. The differences are essentially geometric, resulting in a lower linear power density and other increased safety margins.
The evaluation of the Westinghouse fuel' mechanical design has been based on mechan-ical tests, in-reactor operating experience and engineering analyses. Additionally, the in-reactor' performance of the design will be subject to the continuing surveillance programs at Westinghouse and at individual utilities. The programs will provide con-firmatory and current design performance information.
In our evaluation of the fuel thermal perfonnance, we assume that densification of uranium doxide fuel pellets may occur during irradiation in power reactors. The initial density of the fuel pellets and the size, shape and distribution of. pores within the fuel pellet influence the densification phenomonen. The effects of densification on the fuel rod will increase the stored energy, increase the linear thermal output, increase the probability for local power spikes, and decrease the thermal conductance.
- The primary effects of densification on the fuel rod mechanical design are mant-fested in calculations of time-to-collapse of the cladding and fuel-cladding gap conductance. Time-to-collapse calculations predict the time required for unsupported cladding to become dimensionally unstable and to flatten into an axitl gap caused by fuel pellet densification. Gcp conductance calculations predict the 03 crease in thermal conductance due to opening of the fuel-clad radial gap.
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4-2 TABLE 4.2-1 FUEL MECHANICAL DESIGN COMPARISON Westinghouse Westinghouse Design Parameter 8eaver Valley Unit 1 D. C. Cook Plant FUEL ASSEM8LY Rod Array 17 x 17 15 x 15 Number of Fueled Rods 264 204 Number of Spacer Grids 8 7 Number of Guide Thimbles 24 20 Inter-rod Pitch 0.496 inches 0.563 inches Average Thermal Output 5.4 kW/ft 7.0 kW/ft FUEL PELLETS Density (theoretical) 95% 94%
Fuel Weight / Unit Length 0.364 lbs/ft 0.462 lbs/ft FUEL CLADDING Outside Radius 0.187 inches 0.211 inches Thickness 0.0225 inches 0.0243 inches Radius / Thickness 8.31 8.68 The engineering methods used by Westinghouse to analyze the fuel thermal performance have been previously submitted in Westinghouse Topical Report WCAP-8218. " Fuel Densifi-cation Experimental Results and Model for Reactor Application," dated October 1973.
The results of our review were reported in a document entitled " Technical Report on Densificatici of Westinghouse PWR Fuel" issued by the Commission on May 14, 1974 On the basis of our review we have concluded that the applicants have considered the effects of densification on the Unit I fuel essemblies in a 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 pressurization is designed to preclude any cladding tensile stresses throughout operations due to total internal pressure.
Substantially all of the in-reactor operating experience with Westinghouse fuel rods and 15 x 15 fuel assemblies is applicable to the Unit I fuel assemblies since the 17 x 17 fuel assembly is a slight rechanical extrapolation from the 15 x 15 fuel assemblies. This in-reactor experience has been accumulated for a range of design parameters which are tabulated in Table 4.2-2.
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4-3 TABLE 4.2-2 RANGE OF DESIGN PARAMETER EXPERIENCE PARAMETER RANGE ON POWER REACTOR EXPERIENCE Fuel Rod Array 14 x 14,15 x 15 Rods Assembly 179 to 204 Guide Thimbles / Assembly 16 to 20 Assembly Envelope 7.76 inches to 8.43 inches Inter-rod Pitch 0.556 inches to 0.563 int.hes Plenum length 3.27 inches to 6.69 inches Prepressurization 14.7 psia to 400 psia Diametral Gap 0.0065 inches to 0.0075 inches Spacer Grids / Assembly 7 to 9 The fuel assemblies referred to in Table 4.2-2 have been irradiated up to six years and had peak exposures of 30 megawatt-days per metric ton totaling more than 70 million megawatt hours of power generation. During this power reactor service a small fraction of the fuel rods have experier. red defects. There has been no instance where cladding defects have threatened either tne plant or the public safety. Cladding defects were caused by excessive manufactur'ng impurities, excessive coolant cross-flow velocities and fuel pellet densification. This problem has been rectified by modifications to both the manufacturing procedures and the plant coolant system.
The fuel related modifications required readjustments to the magnitude of a design characteristic rather than a redesign of the fuel assembly. Fuel assemblies identical to the Beaver Valley Unit I design have not yet experienced power reactor service.
However, the current use of similar fuel assemblies have yielded operating experience that provides confidence in the acceptable performance of the 17 x 17 fuel assemblies.
Out of reactor mechanical tests have been performed on typical 15 x 15 fuel assemblies.
Those tests demonstrate acceptable mechanical perfonnance of the 15 x 15 fuel assemblies.
Since the 17 x 17 fuel assembly is a slight mechanical extrapolation from the 15 x 15 fuel assembly, we expect the mechanical behavior of the 17 x 17 fuel assembly to be similar and therefore have concluded that it is acceptable.
Hydraulic and structural verification tests on the 17 x 17 fuel assemblies have been completed and are reported in Westinghouse Topical Reports WCAP-8279, " Hydraulic Flow Tests of the 17 x 17 fuel Assembly' and WCAP-8288, " Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident." We have reviewed these two topical reports and evaluated them as acceptable for inclusion in the license application by reference. The Grid Tests were made on assemblies with seven grids and o Westinghouse will document in a topical report the justification for applying the test results to eight grid assemblies. Also, the first phase of the Single Rod Durst Test have been completed and will also be documented. We will review the documentation and will report the results of our evaluation in a supplement to the Safety Evaluation Report prior to a decision concerning the issuance of an operating license.
Performance of the fuel during operation will be indirectly ronitored by measurement of the activity of both the prinary and the secondary coolant for compliance with 9
4-4 technbl specification limits. The first available irradiated 17 x 17 fuel assemblies and rods will undergo an extended surveillance program following each cycle of operation.
On-site examinations will include fuel rod integrity, fuel rod and fuel assembly dimen-sions and alignrent, and surface deposits.
We have concluded, subject to confirmation of the above cited documentation, that based on (1) operating experience with similar fuel, (2) the results of out-reactor test on an assembly of similar design. (3) the increased thermal margins which the 17 x 17 fuel has, (4) the technical specification requirements to monitor and limit off-gas and effluent activity, and (5) the existence of a continuing fuel rod surveil-lance program which includes destructive and non-destructive post-irradiation examint-tions, the cladding integrity of the 17 x 17 fuel will be maintained and significant amounts of radioactivity will not be released. On the basis of our review of Westing-house Topical Report WCAP-8288, " Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," we have also concluded that accidents or carthquake induced loads will result in neither an inability to cool the fuel nor interfere with control rod insertion.
4.3 Reactor Vessel Internals The integrity of the reactor vessel internals in service is essential to assure that all fuel assemblies remain in place to permit unimpaired operation of the control rod assemblies.
l We have reviewed the selection of materials for the reactor vessel internals . The {
materials are compatible with the reactor coolant, and have performed satisiactorilf
{
in similar applications. Undue susceptibility to intergranular stress corrosion crack- 1 ing will he prevented by avoiding the use of sensitized stainless steel in accordance with methods recorrnended in AEC Regulatory Guide 1.44.
The use of materials proven to be satisfactory by actual service experience, and avoidance of sensitization by tha methods reconinended in AEC Regulatory Guide 1,44 will provide reasonable assurance that the reactor vessel internals will not be susceptible tc failure by corrosion or stress corrosion cracking.
The applicants have described the measures that will be taken to assure that dele-terious hot cracking of austenitic steel welds is prevented. All weld filler metal will be of selected composition, and welding processes will to controlled to produce welds with at least 5% delta ferrite, in conformance with the recocinendation in AEC Regulatory Guide 1.31.
Following these recommendations will provide reasonable assurance that no deleter-ious hot cracking will be present that could contribute to loss of integrity or functional capability.
Wiu regard to flow-induced vibrational testing for the Beaver Valley Power Station Unit 1, the applicants have designated the ti. B. Robinson plant (Docket No. 50-261) as the prototype plant for a three-loop plant from which reoperation 61 vibration test resaits for the 15 x 15 fuel assembly configuration are applicable in evaluating the design adequacy of the reactor internal structures of Beaver Valley Unit 1. The response characteristics of vibratory strain, acceleration, and pressure signals were analyzed in terms of major frequency components for obtaining modal contributions and O
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to define the dynamic behavior under flow induced excitations. These test results
. were reported in Westinghouse Topical Report WCAP-7765-L-AR, " Westinghouse PWR Internals -
Vibration Sumnary - 3 Loop Internals Assurance," October 1973.
l, 4 We have reviewed and approved this report for referencing in license applications subject to the condition that a'sunmary of the analysis to determine forcing functions, ~
gnd a description of the dynamic analysis methods and procedures is submitted for l ~41uation on a case-by-case basis.
The condition set by us is acceptably fulfilled for Beaver Valley Unit I by the information presented in Westinghouse Topical Report WCAP-8303, " Prediction of the _
Flow Induced Vibration of Reactor ilnternals by Scale Model Tests," March 1974, in conjunction with data presented by Westinghouse that demonstrated good correlation between scale model testing as described in WCAP-8303 and the results of measurements nede at the H. B. Robinson plant. We find that the test program and correlation with the data from the H. B. Robinson plant acceptably fulfill the criteria reconsnended in AEC Regulatory . Guide 1.20. Therefore, only a confirmatory preoperational vibration test in accordance with AEC Regulatory Guide 1.20 will be conducted on Beaver Valley Unit 1.
Beaver Valley Unit I will use 17 x 17 fuel assemblies in place of the 15 x 15 assemblies currently employed at H. B. Robinson. Consistent with AEC Regulatory -
Guide 1.20, prototype test data are taken with the core support structures unloaded i.e. . without fuel assemblies. The acceptability of this approach is based on data reported from the full scale tests indicating that the mass and damping added by the fuel assemblies far outweighs any contribution the response of the individual assemblies may have to total response of the core support structures. The use of 17 x 17 fuel assemblies is not. therefore, inconsistent with the acceptance of the prototype tests conducted at H. B. Robinson an sufficient for Beaver Valley Unit 1. A testing program to assure acceptable vibration characteristics of the individual 17 x 17 fuel assemblies is reported in Westinghouse Topical Report WCAP-8279, " Hydraulic Flow Test of the 17 x 17 Fuel Assembly" and is discussed in Section 4.2 of this report.
We have concluded that the confirmatory preoperational vibration test specified for Beaver Valley Unit 1 is consistent with the reconsnendations of AEC Regulatory Guide 1.20 for a non-prototype plant and is acceptable.
The applicants have performed a dynamic system analysis of the reactor internals and the broken and unbroken piping loops (postulated to occur in a loss-of-coolant accident) to provide an acceptable basis for confirming the structural design adequacy of the reactor internals and the unbroken piping loops to withstand the combined dynamic effects of the postulated occurrence of a LOCA and a safe shutdown earthquake.
We have reviewed the analytical methods and find that the results of the analysis show that there is reasonable assurance that the combined stresses and strains in the o components of the reactor coolant system and reactor internals will not exceed the allowable design stress and strain limits for the materials of construction as specified in Appendix F to the ASME Boiler and Pressure Vessel Code Section III, and that the resulting deflections or displacements of any structural elements of the reactor inter-nals will not distort the reactor internals geometry to the extent that core cooling can be impaired.
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4-6 T,he as',urance of structural integrity of the reactor internals under the postulated safe shutjown earthquake and the most severe LOCA conditions provides added confidence that the design can be oxpected to withstand a spectrum of lesser pipe breaks and seism'c loading combinations.
We have concluded that the use of the cited analytical techniques provides an acceptable structural design for the Beaver Valley Unit i reactor internals.
4.4 Thermal and Hydraulic' Design The thermal-hydraulic design of the Unit I reactor has been evaluated on the basis of design full power operation at 2552 MWt. The Unit 1 reactor will use a 17 x 17 fuel assembly, and when compared to the Surry reactor using a 15 x 15 fuel assembly, the thermal-hydraulics parameters are found to be similar in many respects as shown on Table 4.4 The principal differences are in the reduced linear power generation rates and ,
heat fluxes resulting from the increated number of fuel pins and heat transfer area in Beaver Valley Unit 1.
The principal thermal-hydraulic design criterion for a reactor is the prevention of fuel damage by providing adequate heat transfer for the prevailing core heat generation distribution scheme during normal operation and operational transients and during transients resulting from faults of moderate frequency.
TABLE 4.4 COMPARISON OF THERMAL AND HYDRAULIC DESIGN PARAMETERS FOR THE BEAVER VALLEY UNIT I AND SURRY PLANTS BEAVER VALLEY SURRY Reactor Core Heat Output (Mwt) 2652 2411 System Pressure, Nominal (psia) 2250 2250 Minimum DNBR for Design Transients >l.30 >l.30 Total Thermal Flow Rate (lb/hr) 100.9 x 10 6 100.7 x 10 6
Effective Flow Rate for Heat Transfer (ib/nr) 96.3 x 10 6 97.0 x 10 6
Average Velocity Along Fuel Rods (ft/sec) 14.4 14.2 6
Average Mass Velocity (lb/hr-ft2) 2.32 x 10 2.31 x 10 6
Coolant Temperature ('F)
Design Nominal Inlet 542.5 543.0 Average Rise in Core 70.3 65.5 Active Heat Transfer Surface Area (ft ) 48,700 42,460 Average Heat Flux (Btu /hr-f t2 ) 181,400 191.100 Maximum Heat Flux (Btu /hr-ft ) 453,500 524.100 Peak Fuel Central Temperature at 100% Power 3400 4050 Peak Fuel Central Temperature at Maximum Overpower (*F) 4150 4300 Peak Linear Power at 100% Power 13.0 17.3 Maximum Linear Power at Maximum Overpower kW/ft 18.0 20.6 Average Linear Power at 100% Power 5.2 6.2 e
':..' s,
'a-7 The fuel damage limits and thermal-hydraulic criteria used by the reactor designer
. (Westinghouse) to evaluate the performance of. the fuel are the same for the 17 x 17 design as for the 15 x 15 design. - These damage limits for normal operation, operational transients, and any transient condition arising from faults of. moderate frequency are -
(1) departure from nucleate boiling will not occur on at least 95 percent of the limiting fuel rods at a 95 percent confidence level, (2) the maximum fuel temperature shall be less than the melting temperature of UO 2
, (3) at least 95.5 percent of the thermal flow will pass through the fuel rod region of the core, and (4) the pemitted modes of oper- .
ation shall not lead to hydrodynamic hstability.
l To show compliance with these criteria, Westinghouse perfomed DNB and fuel temperature calculations as well as flow distribution and flov instability analyses.
The new design has approximately the same DNB margin as the earlier design and an in-<
ceased fuel temperature margin to hydrodynamic instability.' The effect of fuel densification was included in the 17 x 17 calculations. ' Also, the 17 x 17 DNB calcula-tions included a 15 percent margin on DNB to (1) to incorporate final results of DNB and mixing tests. (2) incorporate final results of hydraulic tests, and (3) allow for
.any fabrication tolerances larger than presently used.
The first part of the DNB tests, utilizing unifomly heated rods, was completed and reported in Westinghouse Topical Report WCAP-8?96 "Effect of 17 x 17. Fuel Assembly 8296 Geometry on DNB " March 1974. The results indicate that (1) the previously used DNB correlation (W-3 correlation with modified spacer factor) must be multiplied by D.88 in order to show agreement with the 17 x 17 data. (2) the use of a themal diffusion coefficient of D 03B is conservative, and (3) a DNBR value of 1.275 corresponds to the 95/95 criterion.
Since only data with uniformly heated rods were considered, it is uncertain at the present time whether 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 heat are planned for December 1974. The results of these future tests together with those reported in UCAP-8296 will be used to set technical specification limits for the 17 x 17 designs. Although Westinghouse does not expect further changes in the correlation, or in the statistical evaluation of the correlation, from our review of critical heat flux correlations, considering both unifom and non-unifom axial heat flux data, we consider changes are possible. If the results of the non-uniform DNB tests are not available when the technical. specifications are finalized for the Beaver Valley Unit 1 plant, we will require a five percent DNBR margin above and beyond the 95/95 criterion.
The Westinghouse Topical Report WCAP-8185. " Reference Core Report 17 x 17 " itemized the presently available DNB nergin in the reference 17 x 17 fuel assembly design. The e sources and amounts of these margins for a three-loop plant are as follows:
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4-8 Source DNBR Margin (percent)
DNB calculations us9d a multiplier of 0.86 while data justify a multi-plier of 0.88. 2 A DNBR of 1.3 was used in lieu of the 95/95 criterion. Data justifies a DNBR of 1.275 2 A TDC value of 0.051 was used in the data reduction while a value of 0.038 was applied in the analysis. 1.4 DNB tests were performed with 26 in, grid spacing while the design utilized 20.5 in. spacing. 5 Thus, the 17 x 17 fuel design offers a DNBR margin of approximately 10 percent beyond the requirements of the Westinghouse criteria. We find this margin to be sufficient to cover uncertainties due to the unavailability of as-built tolerances.
The criterion for over power protection requires that the maximum fuel centerline temperature be less than that of the fuel melting temperature at a core power generation rate of 14.0 kW/ft during modes of operetton associated with Condition I and Condition II events. In fulfillment of this objective, the applicant has selected a calculated fuel centerline temperature limit of 4400'F for the beginning of life conditions.
Limiting of the fuel centerline temperature of 4400'F is accomplished by the following two steps. By assuring the presence of nucleate boiling in the rod bundle with a DN9R of 1.3 or more, as previously described, the maintenance of a clad temperature close to that of the coolant is assured. By use of an appropriate analytical-empirical thermal model which includes the effects of gap conductance, fuel thermal conductivity and power distribution, the fuel centerline temperature can be calculated. This model is the same as that previously approved in our operating license reviews of Point Beach Unit 2 (Docket No. 50-301) and Zion 1 and 2 (Docket Nos. 50-295 and 50-304).
On the above basis, we have concluded that there is reasonable assurance that the fuel centerline temperature will be below that of the fuel melting temperature at fuel rod heat generation rates up to 14.0 kW/ft, and is acceptable, for recently reviewed Westinghouse designed reactors, the THINC computer code has been used to calculate core thermal-hydraulic performance characteristics. The THINC code considers cross-flow between adjacent assemblies in the core and thermal diffusion between adjacent subchannels in the assemblies. The effect on local power distributions is also considered. As a result of these considerations, the THINC code permits the O computation of more realistic power shapes, which are especially important at design over-power conditions than had been available from previously used computer codes.
Certain elements of the THINC verification test program have been performed at the Zion facility this year. We 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, restrictions will be imposed i,n the operation of Beaver Valley Unit 1. Changes to the technical specifications will l
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be made to ma.intain required margins to fuel rod damage during normal operation, as well as during anticipated transients.
Another parameter that influences.the thermal-hydraulic design of the core.is rod to rod bowing with fuel assemblies. Experimental data on the extent of bowing in the 17 x 17 design is not yet,4vailable. The 17 x 17 fuel performance surveillance program should provide this information. In the meantime. the design of the core is based on predicted values of bowing derived from measurements made on incore 15 x 16 fuel-assemblies.' Westinghouse recently submitted two topical reports (HCAP-8346 "An Evaluation of Fuel Rod Bowing" and WCAP-8176. "Effect of Bowed Rod on DNB") that describe the analytica1' techniques used to predict bowing and the method used for assessing the effect of bowing on thermal performance. We are presently reviewing the above Westinghouse Topical Reports and will report the results of our evaluation prior to a decis On concerning the issuance of an ' operating license.
We have concluded subject to satisfactory resolution of the matters discussed above, that the thermal and hydraulic design of Beaver Valley Unit 1 is acceptable for operation at a core power level of 2652 megawatts thermal.
We will report the resolution of the matters discussed a kve in a supplement to the Safety Evaluation Report prior to a decision concerning the issuance of an operating license, 4,5 - fluclear_ D%icn Deaver Valley Unit 1 is similar to several Westinghouse designed reactors specifying the 17 x 17 fuel design and currently being reviewed for operating licenses. We have recently completed a generic review of the Westinghouse 17 x 17 fuel design for both three and four loop reactors '(described in WCAP-8185) and have concluded that the nuclear design is acceptable. We have also concluded that the nuclear design applies directly to Beaver Valley Unit 1 and is therefore acceptable. The nuclear characteristics of the 17 x 17 fuel rod bundles are essentially the same as those of the previous 15 x 15 fuel rod design. As a result, there are no changes in control requirements, control rod patterns and reactivity worths, and xenon stability. The analytical methods employed in the design of this core are the same as those used in previously reviewed l' and licensed plants using Westinghouse reactors and are acceptable, j
We have performed certain independent analyses which are discussed below.
The primary feature of the 17 x 17 fuel design relating to physics and power dis.
l tribution monitoring considerations is the increase in the number of linear feet of h
l fuel provided. For a given core power rating this leads to a reduction in the average linearpowerdensity(LPD). Thus, the average LPD for the Beaver Valley Unit 1 design decreased from 6.7 kW/ft to 5.2 kW/ft with the change from 15 x 15 to 17 x 17 fuel. At o
102% of full power, the peak LpD associated with an overall peaking factor (Fg ) of 2.32 is now 12.3 kW/ft.
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The applicants have proposed to take credit for correct normal operator action in determining the peaking factor used to define initial conditions for accident analyses.
1 L The information presented indicates Fg would be limited to 2.32. whereas with the The applicants' previously used axial of fset relation, the Fg limit would be about 2.5.
plan would eliminate m'ost of the xenon transient effects on Fg that otherwise would occur in load following. This will be accomplished by maintaining operation within R
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_15% in-flux difference (top minus bottom excore detector readings) above 90% power.
Below this level the flux difference will be allowed to be out of the 15% band one hour out of 24. but must remain within 112% below 90% power or the power level must be reduced to 50% of rated power with the high neutron flux setpoint at 55%. Two computer.
alarms will be provided, one to indicate violation of the 15% flux difference band, and the other violation of the one hour limit below 90% power. We find these provisions acceptable to assure maintenance of Fg <_2.32.
We have made an independent comparison of the beginning-of-life (BOL) moderator and Doppler reactivity coefficient for the 15 x 15 fuel assembly and the 17 x 17 fuel assembly using methods equivalent to those used by the applicants. Our results show that the calculated BOL isothermal moderator temperature coefficient in the operating temperature range is 0.1 x 104 /'F to 0.2 x 10'4 more negative with the 17 x 17 fuel assembly than with the 15 x 15 fuel assembly. The moderator temperature coefficient varies from about 0 to -3.5 x 104 /'F over the first cycle. The calculated BOL isothermal Doppler coefficient in the operating temperature range is approximately two percent more negative with the 17 x 17 fuel assembly than with the 15 x 15 assembly. These effects are due to the slightly higher 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 programs (e.g., Surry 1 and 2) and compare favorably with predictions.
Moderator temperature coefficients as a function of temperature and soluble boron concentration shown in Figure 4.3-30 of the F5AR for the 17 x 17 fuel assembly core are almost identical to those previously reported for a 15 x 15 fuel assembly core. The Doppler coefficient for the 17 x 17 assembly core in Figure 4.3-27 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 staff's calculations.) However, the x-y power shape weighting factor is considered reasonable and this fonn of the data is appropriate for use in study of axial power shapes using the PANDA code. We have concluded that the reactivity coefficients are reasonable and acceptable for use in control and safety analyses.
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.- e, 5-1 5.0 REACTOR COOLANT SYSTEM 5.1 Sunrnary Description The reactor coolant system for Beaver Valley Unit 1 includes a reactor pressure vessel and three coolant loops connected in parallel to the reactor pressure vessel.
Each of the coolant loops is equipped with a reactor coolant pump and steam generator.
There are isolation valves in each loop to isolate the steam generator and reactor scoolant pump from the reactor pressure vessel. An electrically heated pressurizer is connected to a hot leg in one of the reactor coolant loops.
The design of the reactor coolant system is the same as that previously reviewed and approved for the Surry Power Station Units 1 and 2.
5.2 Integrity of the Rebetor Coolant Pressure Boundary We have reviewed the reactor coolant system and the pressure retaining components within the reactor coolant pressure boandary. The ASME Code Class 1 components within the reactor coolant pressure boundary will be designed, fabricated and inspected in accordance with the requirements of the applicable codes delineated in Section 4.0 of the FSAR. We have determined that the applicable codes, code editions and addenda, comply with the requirements of 10 CFR Part 50. Section 50.55a.
Compliance with these codes and rules provides reasonable assurance that the resulting componer.t quality level are Edequate to safely withstand the plant loading conditions and combination of design loadings which the systems may experience over their service lifetime, without loss of structural integrity. We have concluded that conformance to these rules constitutes an acceptable basis for satisfying the applir.able requirements of Criteria Nos, 1. 14 and 30 of the AEC General Design Criteria.
The design loading combinations specified for ASME Code Class I components of the reactor coolant pressure boundary have been appropriately categorized wit' respect to i the plant condition identified as normal, upset, emergency or faulted. The design limits proposed by the applicants for these plant conditions are consistent with the criteria reconvnended in AEC Regulatory Guide 1.48. Use of these criterit, for the design of the components of the reactor coolant pressure bounriry will provide reasonable assurance that (1) in the event an earthquake should occur at the site, or (2) other system upset emergency or faulted conditions should develop, the resulting combined stresses imposed on the system components will n:t exceed the allowabl . design stresses and strain limits for the raterials of construction.
Limiting the stresses and strains undet such loading combinations provides a basis for the design of the system components for the most adverse loadings postulated to occur during the service lifetime without loss of the system's structural integrity.
The design load combinations and associated stress and deformation limits specified for ASME Code Class 1 components constitute an acceptable besis for design in satisfying
- the applicable requirements of Criteria Mos.1, 2 and 4 of the AEC General Design Criteria.
The applicants have stated that all ASME Class 1 active valves were designed to function at normal operating conditions, noximum design conditions and design oasis accident conditions. We have concluded that the applicants have provided reasonable assurance of the capability of ASME Code Class 1 active valves to withstand the imposed loads asseriated with normal, upset. emergency and faulted plant conditions without loss 9
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5-2 i of structural irtegrity and to perform the " active" function (i.e. valve closure or opening) 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 reviewed the materials selection, toughness requirements, and extent of materials testing proposed by the applicants to provide assurance that the ferritic materials used for pressure retaining coaponents of the reactor coolant boundary will have adequate toughness under test, normal operation, and transient conditions. All (
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. ferritic materials met the toughness requirements of the ASME Boiler and Pressure l Vessel Code,Section III (1968 Edition for the reactor vessel and 1965 Edition for the pressurizer and steam generators). In addition, materials for the reactor vessel met -
3 the additional testing and acceptance criteria of the Sumner 197: Addenda to the extent possible.
The fracture toughness tests and procedures required by Section III of the ASME Code for the reactor vessel and other ferritic components provide reasonable assurance that accept 6ble safety margins against the possibility of nonductile behavior or rapidly propagating fracture have beel established for the pressure retaining components of the reactor coolant boundary.
The reactor will be operated in a manner that will minimize the possibility of rapidly propagating failure, in accordance with Appendix G to Section III of the ASME Boiler and Pressure Vessel Code, Sumner 1972 Addenda, and Appendix G,10 CF2 50.
Additional conservatism in the pressure-temperature limits used for heatup, cooldown, testing, and core operation will be provided because these will be determined assuming thct the beltline region of the reactor vessel has already been irradiated. The use of Appendix G of the Code as a guide in establishing safe operating limitations, and results of the fracture toughness test performed in accordance with the Code and AEC regulations, will assure acceptable safety margins during operating, testing, maintenance, and postulated accident conditions. He have concluded that compliance with these Code orovisions and AFC regulations, constitutes an acceptable basis for satisfyino the requirements of Criterion No. 31 of the AEC General Design Criteria.
We have reviewed the materials of construction for the reactor coolant pressure boundary to anure that the possibility of serious corrosion or stress corrosion is minimized. All materials used are compatible with the expected environment, as proven ;
by extensive testing and satisfactory service performance. The applicants have stated that the possibility of intergranular stress corrosion in austenitic stainless steel used for components of the reactor coolent pressure boundary was minimized because sensitization will be avoided, and adequate precautions will be taken to prevent contariination durinc manufacture, shipping, storage, and construction. The plans to l
0 avoid sensitization are in substantial conformance with Regulatory G#de 1.44 and include controls on compositions, heat treatments, welding processes, and cooling rates. l We have concluded that the use of materials with satisfactory eervice experience, and i j' substantial conformance v:ith Regulatory Guide 1.44 provide reasonable assurance that
!- austenitic stainless steel components are compatible with the expected service environ-l ments, and that the probability of loss of structural Integrity is minimized, l
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Further protection against corrosion problems will be provided by control of the chemical environment. The composition of Be reactor coolant will be controlled; and the proposed maximum contaminant levels, as well as the proposed pH, hydrogen overpressure, and boric acid concentrations, have been shown by test and service experience to be adequate to protect against corrosion and stress corrosion problems. We have evaluated the proposed requirements for the external insulation used on austentic stainless steel components, and conclude that it will be in conformance with Regulatory Guide 1.36. The controls on chemical composition that will be imposed on the reactor coolant, and the use of external thermal insulation in conformance with Regulatory Guide 1.36, provide -
reasonable assurance that the reactor coolant boundary materials will be acceptably protected from conditions that would lead to loss of integrity from stress corrosion, tie have reviewed the controls proposed to prevent hot cracking (fissuring) of austenitic *tainless steel welds. These precautions included control of weld metal composition and welding process to assure adequate delta ferrite content in the weld metal . With the exception described below, the proposed methods comply with Section III of the ASME Code, and are in essential conformance with Regulatory Guide 1.31.
During field welding, cracks were discovered in the safe-end of the main coolant nc221e (Field weld No. 25). Investigation revealed that the safe-end weld buildup had not been qualified as a strength weld as prescribed by the ASME Code.Section III. The applicants and their contractors have repaired the defective area, and have prepared a mock-up of the joint to perform the necessary qualification test. We will review the results of the qualification test and will report our evaluation of the safe-end investigation, qualification test, and repairs in a supplement to the Safety Evaluation Report. Augmented inservice inspection of these areas may be required, depending upon the results of the mock-up tests.
The maximum full discharge loads resulting from the opening of ASME Code Class I safety and relief valves were calculated by either an equivalent static analysis or a time response dynamic analysis of the system. The maximum stress intensities and stresses resulting from these loads have been calculated in accordance with Subsection fB-3600 of the ASME Code,Section III. In the case of open safety or relief valves mounted on a cocnon header and full discharge occuring concurrently, the additional stresses induced in the header have been combined with previously computed local and primary membrane stresses to obtain the maximum stress intensity.
The criteria used in developing the design and mounting of the safety and relief valves of ASf1E Code Class 1 systems provides reasonable assurance that, under discharging conditions, the resulting stresses are expected not to exceed the allowable design stress and strain limits for the materials of construction.
Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices provides a conservative basis for the design of the system components to withstand these loads without loss of structural integrity and impairment of the overpressure protection function. The criteria used for the design and installt. tion of overpressure relief devices in ASME Code Class 1 Systems constitute an acceptable design basis in meeting the applicable requirements of Criteria Nos.1, 2.
4,14, and 15 cf the AEC General Design Criteria, and are consistent with recent Regula-toru pcsitions.
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5 's. We have reviewed all factors contributing to the structural integrity of the reactor vessel and conclude that there are no special considerations which make it necessary to consider potential vessel failure for Beaver Valley Unit 1.
The bases for our conclusion are that the design, material, fabrication, inspection, and quality assurance requirements conform to the rules of the ASME Boiler and Pressure Vessel Code 'Section III,1%8 Edition and all addenda through Winter.1968, and selected Code Cases.
The stringent fracture toughness requirements of the ASME Code Section III,1971 Edition, and the 1972 Stryner Addenda will be met. Also, operating limitations on-temperature and pressure will be established for this plant in accordance with Appendix G of the 1972 Summer Addenda of the ASME Boiler and Pressure Vessel Code Section III.
The integrity of the reactor vessel is' assured because the vessel (1) ,lill be designed and fabricated to the high standards of quality required by the ASME Boiler and Pressure Vessel Code and pertinent Code Cases listed above.
(2) Will be made from materials of controlled and demonstrated high quality.
(3) Will be inspected and tested to provide substantial assurance that the vessel will not fail because of material or fabrication deficiencies.
(4) Will be operated under conditions and procedures and with protective devices which provide assurance that the reactor vessel design conditions will not be exceeded during normal reactor operation or during most upsets in operation, and that the vessel 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 quality of the reactor vessel has not deteriorated significantly l' 'under the service conditions.
(6) May be annealed to restore the material toughness properties if this becomes necessary.
The toughness properties of the reactor vessel beltline material will.be monitored throughout service life with a material surveillance program that meets all the requirements of ASTM E-185-73 and Appendix H to 10 CFR Part 50 (July;47,1073). Chances in the fracture toughness of material in the reactor vessel beltline, caused by exposure to neutron radiation will be assessed properly, and acceptable safety margins against l-the possibility of vessel failure are provided by a material surveillance program I which meets the requirements of ASTM E-185-73 and Appendix H to 10 CFR Part 50.
Compliance with these documents assures that the surveillance program constitutes an acceptable basis for monitoring radiation induced changes in the fracture toughness of the reactor vessel material, and satisfies the requirements of Criterion No. 31 0 of the AEC General Design Criteria.
Although the use of controlled composition material for the reactor vessel beltline will minimize the possibility that radiation will cause serious degradation of the toughness properties, the applicants have stated that should results of test indicate-p that the toughness is not adequate, the reactor vessel can he annealed to restore the toughness to acceptable levels. We have concluded that the methods proposed are l
feasible.
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5.3 teakage Detection System The major components of the system are the containment atmosphere particulate radio.
activity nonitor, radiogas monitor, and level indicator on the containment sump. The system has sufficient sensitivity to measure small leaks, identify the leakage source to the extent practicable, includes suitable control room alarms and readouts, and is in confomance with the functional requirements reccmended in AEC Regulatory Guide 1.45.
In addition, indirect indications of leakage can be obtained from the containment humidity, pressure, and temperature indicators. Significant intersystem leakage will be l- indicated by abnomal readings from the radioactivity mcnitors used to detect failed fuel, and indirectly, by the coolant flow and level measuring instr 6 ?ntation provided for normal operational control of the system.
We have concluded that the leakage detection system provides reasonable assurance that any structural degradation resulting in leakage during service will be detected in time to permit corrective actions, satisfies the requirements of Criteriun No. 30 of the AEC General Design Criteria and therefore, is acceptable.
$.4 Inservice Inspection To assure that no deleterious defects develop during service, selected welds and -
weld heat-affected zones will be inspected periodically. The applicants have stated i that the design of the reactor coolant system incorporates provisions for direct or remote access for inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. Remote access methods are under development to facilitate the inspection of those areas of the reactor vessel not readily accessible to the inspection personnel.
We have evaluated the access provisions and planning for inservice inspection and conclude that they meet the provisions of the AEC Guideline, " Inservice Inspection Requirements for Nuclear Power plants Construc;ed with Limited Accessibility for Inservice Inspection," (January 31,1969) and are acceptable. )
The conduct of periodic inspections and hydrostatic tetting of pressure retaining components in the reactor coolant pressure boundary in accordance with these require-ments provides reasonable assurance that evidence of structural degradation or loss of leaktight-integrity occuring during service will be detected in time to permit corrective action before the safety function of 3 component is compromised, and constitutes an acceptable basis for satisfying the requirements of Criterion No. 32 of the AEC General Design Criteria.
5.5 Pump Flywheel Integrity The probability of a loss of pump flywheel integrity has been minimized by the use of suitable material, adequate design, and inservice inspection. We have evaluated the o integrity of the reactor coolant pump flywheel and have concluded that its integrity is provided by compliance with AEC Regulatory Guide 1.14. The use of suitable material, and adequate design and inservice inspection for the flywheels of reactor coolant pump motors as srecified in the FSAR provide reasonable assurance (1) that the structural integrity of flywheels is adequate to withstand the forces imposed in the event of pump design overspeed transient without loss of function, and (2) that their integrity will be verified periodically in service to assure that the required level cf soundness of 4
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5-6 the flydel material is adequate to preclude failure. Compliance with the recorren-dations of AEC Regulatory Guide 1.14 constitutes an acceptable basis for satisfying the requirements of Criterion No. 4 of the AEC General Design Criteria.
5.6 Loose Parts Manitor Occasionally, miscellaneous items such as nuts and bolts have beco:3e loose parts within reactor coolant systems. In addition to causing operational inconvenience, such loose parts can damage other components within the system or be an indication of undue wear or vibration. For such reasons, we have encouraged applicants over the past several years to support programs designed to develop systems for 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 ongoing program, the objective of which is the development of a fur.ctional, loose parts monitor-ing system within a reasonable period of time. Recently, prototype loose parts monitor-ing systems have been developed and are presently in operation or being installed at several plants. The applicants are presently evaluating available systems for loose parts monitorit., and have made a comiitment to install the appropriate available system prior to fuel loading. This commitment to provide a loose parts monitor is acceptable -
to the staff.
5.7 Ptrip Overspeed The Regulatory staff is investigating, on a generic basis, the consequences of an unlikaly rupture of a reactor coolcnt pipe which in certain locations might result in reactor coolant pump overspeed. If this study indicates that additional protective' measures are warranted under specific circumstances in order to prevent significant pimip overspeed or to limit potential consequences to safety-related equipment, the Regulatory staff will review the circumstances applicable to Beaver Valley Unit I to determine what modifications, if any, are needed to assure that an acceptable level of safety is maintained, i
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6-1 6.0 ENGINEERED Sl,iETY FEATURES 6.1 General --
Engineered safety features is the designation given to those systems which are provided for the protection of the public and station pertonnel against the postulated release to the environment of radioactive products from the reactor coolant system, parti-cularly as a result of a postulated loss-of-coolant accident. This section discusses the emergency core cooling system, the reactor containment system, containment heat removal system, containment isolation system, combustible gas control system, and the supple-mentary leak collection and release system.
Systems and components designated as engineered safety features are designed to be capable of assuring safe shutdown of the reactor under the adverse conditions of postulated design basis accidents. They are designed, therefore, as seismic Category I and must function even with complete loss of offsite power. Components and systems are provided in sufficient redundancy 50 that a single failure of any component or system will not result in the loss of capability to achieve and maintain safe shutdown of the reactor. The instrumentation systens and emergency power systers are designed to the same seismic and redundancy requirements as the systems they serve. These systems are discussed in Sections 7.0 and 8.0 of this report.
6.2 Containment Systems _
6.2.1 Containment Functional Design The containment structure for Beaver Valley Unit i utilizes the subatmospheric type concept. The containment is a steel-lined, reinforced concrete structure with a net free volume of about 1,800,000 cubic feet. The structure houses the reactor coolant system, including the reactor, pressurizer, and reactor coolant pumps, as well as cer-tain components of the plant engineered safety features systems. The structure is de-signed for an internal pressure of 45 psig and a temperature of 280'F.
The applicants have analyzed various postulated lost-of-coolant accidents including a spectrum of hot leg and cold leg breaks, up to and including the double-ended rupture of the largest reactor coolant line to determine the subsequent containment pressure responses . The LOCTIC computer code was used by the applicants to calculate the pres-sure response of the containment to postulated loss-of-coolant accidents. The LOCTIC code calculates the mass and energy addition to the containment during the blowdown phase.
ttass and energy addition during the core reflood and post-reflood phases of the accident were based on separate analyses.
We have reviewed the applicants' calculational method and conclude that the LOCTIC code conservatively calculates the mass and energy release to the containment during the blowdown phase of a postulated loss-of-coolant accident. The applicants' analytical models for predicting the mass and energy released to the containment during the core e
reflood and post-reflood phases of a LOCA include the following assumptions:
(1) Core Reflood The energy released to the containment was maximized by assuming:
(a) Suffir.. ent water remained in the reactor vessel at the e..d of blowdown to reach the bottom of the core.
(b) Full operation of the safet) injection systems to FBXimite the core reflooding rate.
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6-2 (c) The release of core and core internals sensible heat during this period.
(d) ""A steam-liquid carryout correlation based on FLECHT data.
(e) All steam and water passing through the steam generators is evaporated at the temperature of the steam generator secondary water.
(f) No steam quencMng.
(g) Liquid carryout begins at the two-foot level in the core and continues until-the ten-foot level is reached.
(2) Post Reflood The applicants used a calculational model for predicting the release of l
secondary system stored energy to the containment for this phase of the postu- l lated accident that is baseo on a model developed by the Westinghouse Electric Corporation, and which has been previously accepted by us in our review of similar plants. The model assumes that after the core has been completely covered with water, decay heat will produce a two-phase mixture that rises above the core and enters the steam generators, covering the steam generator tubes.
The remainder of the available steam generator energy would then be removed by evapc6dion of the two-phase mixture. In calculating the rata of energy removed from the steam generators the applicants have used the maximum steam flow permitted by tho hydraulic resistance of the reactor coolant system and the driving head generated by the density difference between the liquid in the downtoner and the two-phase mixture in and above the core.
The peak containment pressure calculated by the applicants was about 38 psig and occurred about 15 seconds after accident initiation. Essentially the same results were obtained for the postulated double-ended rupture of a cold leg, pump suction pipe and a hot leg pipe. Winter conditions (lowest river water temperature) and the availability of normal engineered safety features, except for minimum quench spray system operation were assumed for the postulated cold leg break; winter conditions and the availability of minimum engineered safety features were assumed for the hot leg break. The longest containment depressurization time was calculated to be about 55 minutes and occurred for the postulated cold leg breaks; summer conditions (highest river water temperature),
nomal ECCS operation and minimum quench and recirculation system operation were assumed.
We have used the CONTEMPT computer code to perform confirmatory containment pressure response analyses of tre above postulated pipe breaks. The results of our analyses in-dicate reasonable agreement with the applicants' calculated peak containment pressure and depressurization time. Since the containment design pressure of 45 psig provides a margin of about 18% above the peak calculated pressure, we have concluded that the containment design pressure is acceptable. In addition, the containment depressurization O time is less than one hour, which is the time used to essess the radiciogical consequences of the postulated loss-of-coolant accident.
The applicants have also analyzed the containment pressure response to a postulated main steam line failure. The analysis was done in an acceptably conservative manner by assteing conservatively high energy release rates to the containment. The peak contain-ment pressure calculated by the applicants was also about 38 psig. The containment de-sign pressure of 45 psig provides a margin of about 18% above the peak calculated pres-sure.
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o The applicants have analyzed the transient pressure response of the containment in-
.terior compa'rtments enclosing the reactor _ vessel, pressurizer, and steam generators. The
- applicants used the blowdown data calculated by the LOCTIC code and in some cases the -
reactor vendor's blowdown data as input-to the CUPAT computer code to calculate the pressure response of an interior compartment.
The applicants have calculatedpeak differential pressures of 146 psi for the upper.
reactor cavity, 88 psi for the annular volume between the reactor vessel and,the reactor cavity shield wall,17 psi for the lower reactor cavity,24.6 psi for a steam generator compartment, 937 psi for the reactor cavity shield wall pipe penetration,13 psi for the lower pressurizer cubicle 17 psi for the upper pressurizer cubicle, and 1.7 psi for the.
pressurizer cubicle superstructure. He have performed similar analyses using our own .
computer codes and have confinned the applicants' results for these interior compart-ments. . The structural design of the containment interior structures was based on dif-ferential pressure loadings greater than those stated above.
We have evaluated the containment system functional design in accordance with Cri-teria Nos.16 and 50 of the AEC General Design Criteria. We conclude'that the appli-cants have provided an acceptably conservative analysis 'of the containment transient response following a postulated loss-of-coolant accident, and the containment design pressure of 45 psig provides an acceptable margin above the maximum calculated contain-ment pressure. We have also concluded that the applicants' calculations of the maximum interior compartment differential pressures are acceptable.
6.2.2 Containment Heat Removal Systems Two spray systems are provided to depressurize the containment following a loss-of-coolant accident. They are the quench spray system and the . recirculation spray system.
The recirculation spray system is designed to maintain a substmospheric pressure in the containment after it is depressurized following a postulated loss-of-coolant accident.
Redundant quench spray subsystems, each with 100% capacity, will draw water from the refueling water storage tank. The water in the 425,000 gallon tank will be maintained at a temperature of about 45'F. The quench spray system will be designed to automati-cally spray the cold borated water into the containment about 60 seconds after accident initiation.
The recirculation spray system, which will be actuated about four minutes after the i quench spray system, is designed to provide additional depressurization of the contain-ment and will be used to maintain the containment at subatmospheric conditions in the long term following an accident. The four redundant 50% capacity recirculation spray subsystems will take suction from the containment sump, which will be enclosed by a protective screen assembly. The recirculation spray water will flow through recircula-tion coolers and be cooled by river water. Two of the recirculation pumps and all o recirculation coolers are located inside the containment.
Provision has been made in the design of the quench and recirculation spray systems j to permit periodic testing of the pumps and spray nozzles. The spray nozzles will be individually air tested.
We conclude that the heat removal systems are designed to meet the intent of Cri-teria Nos. 38, 39, and 40 of the General Design Criteria, and are acceptable, j
m fey 644-6.2.3 Supplementary Leak Collection and Release System The-supplementary leak collection and release system (SLCRS) is provided to control l the atmosphere in areas outside the containment that may contain airborne radioactivity'y?
The SLCRS will be operated continuously during plant operation. The system was designed to maintain a negative pressure of 0.25 inch-water gauge in the areas served by the
, system. The system is seismic Category I and incorporates redundant exhaust fans, filter banks, and dampers where required.
Following a loss-of-coolant accident, the SLCRS will process the atmosphere in those areas into which the containment may leak; 1.e., the fuel building and areas contiguous to the containment which include the personnel access hatch area, the purge duct area, the main steam valve area, the cable vault and rod control building, the pipe tunnel, .
and the engineered safeguards areas. . All containment penetrations, except the equipment hatch are within areas served by the SLCRS.
The applicants maintain that the SLCRS will filter and reduce the amount of fission products. that could potentially be released to the environment-during the time the con-tainment is being depressurized (less than one hour) following a postulated loss-of-coolant accident. The applicants also maintain that a conservative estimate of the s fraction of the technical-specification limit for total containment leakage that would-terminate within areas served by the SLCRS is 50%. However, the applicants have not-satisfactorily justified the leakage measuring techniques that would be used to demon-strate that at least 50% of the total containment leakage would be filtered. There fore, the analysis of the radiological consequences of a loss-of-coolant accident are based on direct leakage to the environment at the design leak rate for the containment, i.e.,
0.1% per day, during the period of containment pressurization following a postulated loss-of-coolant accident. This is a conservative assumption since it is reasonable to expect that a significant amount of the total leakage will be into areas served by the SLCRS and would be filtered so that the resultant doses would be less than that calcu-lated (See Section 15.2.1).
On the basis of our review we conclude that the SLCRS is acceptable.
6.2.4 QnpinmentIsolationSystems The containment isolation system was designed to automatically isolate piping systems that penetrite the containment to prevent out-leakage of the containment atmosphere following a postulated loss-of-coolant accident. Double barrier protection, in the fom of closed systems and isolation valves, is provided to assure that no single active failure will rc. ult in the loss of containment integrity. Containment penetration piping, including the isolation valves, are designed to seismic Category I. Contain-ment isolation will automatically occur on a safety injection, containment isolation Phase A, containment isolation Phase B, or steam line isolation signal.
.., We have reviewed the containment isolation system for confonnance to Criteria Nos.
54, 55, 56 and 57 of the General Design Criteria and AEC Regulatory Guide 1.11. We conclude that the system design meets the intent of the applicable General Design Criteria and Regulatory Guide, and is acceptable.
6.2.5 Combustible Gas Control Systems Following a postulated loss-of-coolant accident, hydrogen may accumulate inside the containment. The major sources of hydrogen generation include (1) a chemical reaction 9
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6.2.7' Containment Air Purification and cleanup' Systems
,The quench spray system and the recirculation spray system will also be used for iodine removal following a postulated loss-of-coolant accident. We estimate that the quench spray systems, which become effective approximately one minute after the acci-dent will achieve an iodine removal effectiveness equal to a first order removal con-
'stant (1) of 10 per hour, in an effective volume of approximately 910,000 cubic feet of the total volume of 1.800.000 :.ubic feet of the containment. The recirculation spray
- system, which does not become effective until five minutes after the accident also reaches an effective removal coefficient of 10 per hour, in an effective volume of-740.000 cubic feet. According to our estimates. 150.000 cubic feet of the containment -
volume are not covered by either quench or recirculation spray. This region of the-containment is treated as an unsprayed volume in our analytical model which we use to evaluate the performar e of the containment spray systems for iodine removal.
Sufficient sodium hydroxide will be injected via the quench spray system to support the assumption that the containment spray systems will achieve an overall partitioning of the elemental iodine such that inital iodine concentration in the containment atmos.
' phere is reduced by a factor of 30.
6.3 Emergency Core Coolina System 6.3.1 Design Bases The emergency core cooling system (ECCS) is designed 6o provide core cooling for postulated mechanical failures in the reactor coolant system piping resulting in loss-of-coolant from the reactor vessel greater than the available coolant makeup capacity using normal operating equipment.
The design bases are to prevent fuel and clad damage that would interfere with ode-quate emergency core cooling, and to mitigate the amount of clad-water reaction for any break, up to and including a double-ended rupture of the largest primary coolant'ptpe.
There requirements are intended to be met even . Ith minimum engineered safety features available, such as the loss of one emergency power bus, together with the unavailability of offsite power.
The ECCS subsystems are of such number, diversity, reliability and redundancy that no single failure of ECCS equipment occurring during a loss-of-coolant accident wfIl result in inadequate cooling of the reactor core. Each of the ECCS subsystems is designed te function over a specific range of reactor coolant piping system break sizes, up to and including the flow area associated with a postulated double-ended break in the largest reactor coolant pipe (8.25 square feet).
The design of the ECCS for Beaver Valley Unit 1 is the same as that previously reviewed and approved for the Surry Power Station.
6.3.2 System Desion The Beaver Valley Unit 1 ECCS consists of accumulator tanks, high pressure injec-4 tion and low pressure injection systems with provicions for recirculation of the borated coolant after the end of the injection phase. Various combinations of these systems will assure core cooling for the complete range of postulated break sizes.
Following a postulated LOCA. the ECCS will operate initially in the passive accumu-lator injection mode and the active high pressure injection mode, then in the active low pressure injection mode, and subsequently in the recirculation mode.
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- t. evaluated by the staff in light of the new Acceptance. Criteria.' Our evaluation of the
[ performance of the emergency core cooling system will be reported in a supplement to this .
Safety Evaluation Report.
. 6.4 Habitability Systems I' The control room has been designed with adequate concrote shielding to protect the.
control. room occupants. The control room ventilation will be provided by seismic Category I equipment. During accident conditions the system will be isolated from out-side air and will operate.in a closed recirculation mode. During the first hour follow-
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ing the postulated accident the control room will be pressurized by an emergency bottled -
air supply system.
Upon depletion of the bottled air supply, the contro' room area will be kept above atmospheric pressure by pressurization fans which will take outside air and discharge thrcugh'a charcoal filter bank into the air conditioning system.
We have reviewed the applicants' calculations of radiation doses to control room personnel under accident conditions. For postulated loss-of-coolant accident the cal-c'ulated doses will be within the limiti specified in Criterion No.19 of the AEC' General Design Criteria.' We have determined also that the .4 'ation doses to control' room .
personnel at Beaver Valley Unit I due to a postulated accident at the adjacent Shipping.
port Atomic Power Station will be within the limits specified in Criterion No.19.
We have recently completed an analysis of potential chlorine releases on reactor.
- sites that store substantial amounts of chlorine for water treatment purposes. As a result of this analysis, we have concluded that protection of control room personnel from chlorine releases will be required at Beaver Valley Unit'l. We have informed the applicants of our conclusion and will review the required modifications .to the ventila-tion system.'and will report the results in a supplement to the Safety Evaluation Report.
On the basis of our review, and subject to the satisfactory resolution of the requirement to protect against chlorine releases, we conclude that the control room habitability systems are acceptable.
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- 7,0 . INSTRUMENTAL!ON AND CONTROLS.
- 7,1 Introduction.
The Corviission's Gerral Design Criteria. IEEE Standards including IEEE Criteria for Protection Systems for Nuclear Power Generating Stations (IEEE-279-1971), and applicable -
AEC Regulatory Guides for. Power Reactors have been utilized as the bases for evaluating the adequacy of the protection and control systems. The review of the protection and control systems was accomplished by comparing the designs with those of the hrry Units 1 and 2 previously reviewed and licensed. Our review concentrated on those areas of design .
(1) which are unique to Beaver Valley Unit 1. (2) for which new information has been .
received, or (3) which have remained as continuing areas of concern during this and prior reviews of similarly designed plants. During the course of our evaluation, we had sev--
eral meetings with the applicants and its contractors. In addition, members of the staff -
conducted an engineering drawing review in Decenter 1973 and at the plant site in July 1974.' Our safety evaluation findings reflect the results of the engineering drawing
. review and site visit which was intended to ascertain that the design' criteria are
. properly being put into effect in the implementation of the instrumentation, control and engineered safety feature circuits.
- 7.2 ' Reactor Trip System The reactor trip s),cem (RTS) design described in the FSAR is functionally the same to that of the Surry Units 1 and 2, with the exception that in Beaver Valley Unit 1 the.
solid state logic system instead of the relay logic system is used. The Westinghouse solid state protection system has been introduced in recent plants starting with D. C.
. Cook Plant. We have reviewed its incorporation in the Beaver Valley Unit I design and we have concluded that the RTS mects the IEEE-279-1971 requirements The RTS contains a number of anticipatory trips for. which no credit is taken in the safuy analysis. The applicants have been advised that the introduction of these trips
.in the RTS should not in any way result in the degradation of any primary trips. The applicants have documented in the FSAR that all trips are considered part of the RT$.and as such are designed to meet the requirements of IEEE 279-1971. Periodic testing pro-visions for the RTS trip channels are provided. 81ased on the results of our review, we conclude.that the RTS design is acceptable.
7.3 Engineered Safety Features Actuation and Control The engineered safety features actuation system is functionally comparable to that of Surry Units 1 and 2. Our review encompassed all aspects of the protection system that initiate and contro' ts operation of the engineered safety features systems and their vital auxiliary supporting estems, including fur.ctional logic diagrams, testing capabil-ities and control of bypasses. The following sections identify those aspects of the design that were not acceptable to us and that were changed as a result of our review.
' Also, they discuss those pending items of concern that must be satisfactorily resolved before this system will be considered acceptable.
7.3.1 Transfer to Recirculation Mode Changeover from injection to the recirculation mode of operation following a postu.
lated loss-of-coolant accident is accomplished by the operator in accordance with estab-lished procedures which include a series of inanual actions. We have requested A
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7-2 the applicants to consider design changes aimed at reducing the number of manual opera.
tions that the reactor operator has to perform in order to switch over from injection to recirculation mode. The applicants will submit these design changes for our evaluation and we will report the results of our evaluation in the supplement to this Safety Evaluation Report.
7.3.2 Reactor Coolant loop Isolation Reactor coolant loop step valve position interlocks and extensive administrative controls are used to constitute the steamline break actuation circuits in a manne to recognize reactor operation with one loop out of service. Our review of the des'gn revealed that during operation with a loop isolated the steamline break isolation protective logic for the active loops would be effectively bypassed (high steam line differential pressure trip) or changed to two-out-of-two (high steamline flow, steam line pressure, and low-low temperature average) which does not meet the single failure criterion. We have concluded that the proposed administrative controls do not provide sufficient assurance that the protection of the active loops is maintained during opera-tion with a loop isolated. We require that the procedures be modified to include om positive means of assuring that full protection is maintained in the active loops during operation with one loop out of service. The applicants intend to submit modified proce-dures. We will report our evaluation in a supplement to the Safety Evaluation Report.
7.3.3 Accumulator Isolation Valves Our review of the control circuits for automatic opening of the accumulator isoletion valves revealed that the design did not conform to our criteria with regard to providing the feature for automatic opening of these valves when the primary coolant pressure exceeds a pre-selected value. The applicants have agroed, at our request, to incorporate this feature in the design. We concluded that, with this modification the design is acceptable (see Section 6.3.2).
7.3.4 Hydrogen Recombine
- System and Supplementary leak Collection and Release System The hydrogen recombiner system and those portions of the supplementary leak collection and release system which are required to function following a postulated loss-of-coolant accident have been identified in the FSAR as engineered safety features systems. Accordingly, the applicants at our request, have documented in the FSAR that the instrumentation, control ar.d electrical equipment pertaining to these systems will be designed in accordance with the requirements of IEEE 279-1971. We find this commit-ment acceptable. He will review the final design and assure that the recommendations of Regulatory Guide 1.7 are met. We will report on this item in a supplement to the Safety Evaluation Report.
0 7.3.5 Vital Supporting Systems for Engineered Safety Features The applicants have identified all of the supporting systems essential to the proper functioning of the engineered safety features. It has documented in the FSAR that the instrumentation and controls for these vital supporting systems are designed to the same criteria as for ESF systems that they support, including conformance with IEEE 279-1971.
Based on our review we have concluded that this design is acceptable, 7.4 Systems Required for Safe Shutdown We have reviewed the instrumentation, control and electrical systems provided for safe shutdown as well as the design provisions to place ud keep the plant in a safe e
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. shutdown condition in the event that access to the main control room is restricted or lost. Weliave concluded that the designs conform to our criteria and cre acceptable.
7.5 Safety Related Display Instrumentation We have reviewed the design of the instrumentation systems that provide information (1) to enable the operator to perform required manual safety functions and (2) for post-accident monitoring. For each plant parameter required for these functions, )
two instrumentation channels with at least one recorded should be provided. Also, they i should be powered from the emergency power buses. The applicants have indicated their intention to provide two instrumentation channels with at least on recorded. We will j repor6 aur evaluation of this item in a supplement to this Safety Evaluation Report. l Our review of the design provisions for indicating operational bypasses revealed that there were no provisions for accomplishing that at the system level. The applicants agreed to provide an administrative 1y controlled status board indicating to the operator the a"ailability of systems. The indications will be in the form of manually actuated back-lit panels which display, at the system level, any system and train which is not operable as a result of bypassed or deliberately made inoperable equipment. We find this design acceptable for Beaver Valley Unit 1.
7.6 Overpressure protection Interlocks for Residual Heat Removal System Our review of the residual heat removal (RHR) motor-operated suction valve inttrlocks uted to prevent over pressurization of the RHR system by the reactor coolant systen revealed that the design did not satisfy our criteria with regard to providing inte* locks of diverse principles to prevent opening and automatic closure of these valves. The applicants have agreed, at our request, to include in the design the feature of diversc interlocks. We cencluded that this design is acceptable.
The applicants hue identified the RHR discharge lines as anoth*r point of inter-face between this system and the reactor coolant system. Each of 1 two discharge lines is isolated from the reactor coolant system by a motor-operai.a valve in series with a check valve. Administrative controls will be employed to ensure that the motor-operated valve will remain closed when required.
We have concluded that the proposed administrative controls do not provide suffi-cient assurance that these valves will be closed and remain closed when required to prevent overpres:urization of the RHR system in the event of the check valve failure.
Therefore, we have required and the applicants have agreed to modify the design so that the RHR discharge motor-operated valves will be interlocked to prevent valve opening whenever the reactor coolant system pressure is above the RHR system design pressure, and automatically close whenever the reactor coolant system pressure exceeds the RHR system design pressure. This design is consistent with our position on meeting the criteria for high pressure to low pressure interfaces and is acceptable.
7.7 Environmental and Seismic Qualifications The applicants have identified and stated that all instrumentation, control and electrical equipment inportant to safety has been environmentally and seismically qualified. Prototypes of all vital instruments, motors, cablet and per.trations located within the containment have been tested, where required, under similated LOCA conditions of combined pressure, temperature, and containment spray. Further, resistance 8
to radiation damage has been verified up to at least 10 rads. The applicants have described these tests, sumarized the test results, and identified the test documentation.
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7-4 tee applicant's 'have also stated that the safety related electrical and instrumen-tation equipment have been ' seismically qualified by prototype test or analysis as documented in. Westinghouse Topical Report WCAP-7817 with Supplements 1, 2 and 3 which have been reviewed and found acceptable by the Regulatory staff.
Ou ' review of the plant protection system instrumentation requiring qualification -
revealed that the reactor coolant temperature detectors would not be qualified. The I applicants have agreed, at our request,' to seismically and environmentally qualify
.these detectors. We have requested the applicants to provide information to verify that .
the detectors will be qualified for the conditions expected at Beaver Valley Unit 1.
We will review the information and report the results of our evaluation a a supplement.
to the Safety Evaluation Report.
Subject to the' satisfactory resolution of the outstanding item cited above, we conclude that the seismic and environmental qualification for safety related equipment, is acceptable, 7.8 Cable Separation and Identification Criteria The applicants' criteria governing separation and installation of safety related -
cables were reviewed and found acceptable for Beaver Valley Unit 1. He have also reviewed the means proposed for identification of safety related equipment, cables and cable trays, and concluded that they are acceptable.
7.9 Control Systems Not Required For Safety The applicants have stated that the functional designs of the major control systems for this plant are the same as that for Surry Units 1 and 2. We find this commitment' acceptable.
7.10 Anticipated Transients Without Scram The applicants have reviewed the staff report, WASH-1270, " Technical Report on Anticipated Transients Without Scram (ATWS) for Water-Cooled Power Reactors". Unit 1 -
has been classified by the staff as a Class I.B. facility, and the applicants have 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 scram in accordance with Section II.B of Appendix A'of WASH-1270.
The applicants have documented the information required by WASH-1270 on September 24, 1974, by referencing Westinghouse Topical Report WCAP-8330, " Westinghouse ATWT Analysis,"
and WCAP-7706, "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients." The applicants state that no hardware modifications are required to mitigate the consequences of AlWS. The staff evaluation of this infonnation will be contained in a supplement to this Safety Evaluation Report.
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0-1 8.0 ELECTRIC POWER 8.1 Introduction-Criteria Nos.17 and 18 of the AEC General Design Criteria, IEEE Standards including IEEE Criteria for Class IE Electric Systems for Nuclear Power Generating Stations (IEEE 308-1971), and AEC Regulatory Guides for Power Reactors including AEC Regulatory Guides 1.6 and 1.9 served as the bases for evaluating the adequacy of the electrical power system.
8.2 Offsite Power System Deaver Valley Unit I will be interconnected to the electrical grid system through four 345 kV and five 138 kV transmission lines emanating from their respective switch-yards. Each type of high voltage transmission lines converges on its respective switchyard by means of two or more separate and independent routes. Both switchyards are arranged in a double bus configuration and interconnected through an autotransformer.
Power from the Beaver Valley Unit 1 generator will be supplied to the 345 kV switchyard.
The adjacent Shippingport Atomic Power Station supplies power to and receives it direct-ly from the 138 kV switchyard. The Beaver Valley Unit 1 generator will also supply power to two half-size unit auxiliary transformers, each having two secondary windings.
Offsite power to Beaver Valley Unit 1 will be supplied from two separate feeders emanating from each bus of the 138 kV switchyard. Each ieeder line, associated breakers, and station service transformer have sufficient capacity to handle all required normal and emergency load demands. All of the high voltage circuit breakers in both switchyards are provided with primary and backup relaying circuits powered from independent supplies.
There are four separate main 4160 V bus sections, each being fed from one of the two secondary windings of one of the two unit auxiliary or startup transformers. Each redundant emergency bus is connected to a main 4160 V bus through two series isolation t rea kers . During normal operation, the emergency buses can be powered from the unit auxiliary transformers, or the station service transformers, or a combination of both.
Upon loss of the selected normal supply, power is made available automatically to these buses from the alternative supply (i.e. the unit auxiliary cr the startup transformer).
Each one of the station service transformers and related distribution systems have sufficient capacity to meet shutdown and emergency load requirements.
The applicants have conducted electrical grid stability analyses showing that the loss of the largest generating unit in the grid, or the most critical transmission line, will not adversely effect the stability of the remainder of the transmission system or the ability to provide offsite power to Beaver Valley Unit 1. q Our review of the offsite power system revealed that in the event of an electrical fault, a single failure in the protective relaying circuits of the breakers connecting the 138 kV switchyard to the adjacent Shippingport Atomic Power Station transformer f w uld result in the simultaneous loss of both offsite power circuits to Beaver Valley o
Unit 1. The applicants ..,resa - . 9 'v the design by providino a second redundant protective relay scheme including redundant relays, current transformers and d-c control power. This design change commitment was found acceptable. We will verify its proper implementation and report our conclusions in a supplement to this Safety Evaluation Report. We have concluded that the offsite power system satisfies the requirements of I Criteria Nos.17 and 18 of the AEC General Design Criteria and IEEE 308-1971, and is acceptable. ;
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.. e 8-2 8.3 Onsite Power Systems 8.3.1 A-C power System The a-c emergency onsite power system is comprised of two redundant distribution systems, each powered by one of the two redundant diesel generators. Each distribution system includes 4160, 480, 240 and 120 volt load centers to accorrncdate the voltage requirements of the safety loads.
The safety loads for the unit are distributed evenly between the two distribution systems with the exception of those loads that provide extra redundancy such as the high pressure injection pump and river water pump. Each of these loads can be powered from either distribution system through separate breakers. The selection of the power feed for these loads is accomplished manually through key-interlocked bus-transfer switches which prevent interconnection of the power supplies.
In addition, the design includes the capability for disconnecting selected loads from the emergency buses that are not required to operate during the containment isolation phase B of the accident which encompasses spray actuation. The applicants have stated that this capability is provided to protect against diesel generator overloading.
Our review of the emergency onsite power distribution system showed that it was subject to single failure of one of the dual feeder breakers used to power redundant loads, that will result in interconnecting the two redundant emergency 4160 V buses.
We advised the applicants that this design defic 9cy was unacceptable and should be removed. The applicants have modified the desh, so that only one of the dual breakers will be installed in its cubicle at a time, the other breaker will be physically removed and located elsewhere. We conclude that this modification is ac:eptable.
Each diesel generator wii be automatically started on an undervoltage signal from its respective 4160 V emergency bus, on safety injection signal, or on opening of either of the two series isolation breakers through which offsite power 1: being supplied to the emergency buses. If offsite pnwer is not available, the 4160 V emergency buses will be automatically isolated from all supply sources and all outgoing feeder breakers will be tripped. The diesel generators will be connected automatically to their respective 4160 V emergency bus and the safety loads will be autorstically connected in a predetermined sequence to their respective diesel generator.
The diesel generator units used by Beaver Valley Unit I are of similar design as those used in previously reviewed and approved Fitzpatrick Nuclear Power Plant (Docket No. 50-333) and we have found these qualified for nuclear power plant service. The originally proposed loading sequence resulted in a first load step of about 1500 kW.
We found that this would result in unnecessarily severe initial voltage and frequency drops and we required the applicants to rearrange the loads and loading sequence so as o to minimize undervoltage and underfrequency transients. The applicants agreed to submit a revised loading sequence and the corresponding voltage frequency transient curves for our review. We will review the revised loading sequence to determine that the revised design satisfies the recommendations of AEC Regulatory Guide 1.9 and will report the results of our evaluation in a supplement to the Safety Evaluation Report.
The diesel generator units will be located in separate seismic Category I structures.
Each unit will have independent auxiliary systems and separate seismic Category 1 9
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- v underground fuel storage tanks. The total onsite fuel oil storage capacity will provide for at least seven days of diesel generator operation at full rated load. ' Our *eview .
of the a-c emergency onsite power system also revealed that the independence of the redundant emergency buses would be compromised through a single bus tie breaker connect-ing the redundant 4160 V load center buses, thus' making the design vulnerable to common failure mode. The applicants have modified the design by removing the single bus tie breaker and the bus disconnect links. We concluded that this design change is acceptable.
. Subject to the satisfactory resolution of the item mentioned above, we have concluded that the a-t emergency onsite power system will satisfy the requirements of Criteria Nos. 17 and 18 of the AEC General Design Criteria. IEEE 308-1971 and -
Regulatory Guides 1.6,1.9 and 1.22. and is acceptable.
8.3.2 D-C Power System Onsite d-c emergency power is derived from the station battery / charger system. The '
station battery / charger system is compromised.of five serar. ate 125 volt battery bank-charger units and related distribution system. Four of the battery / chargers are arranged in a configuration of. two redundant and separate trains providing d-c power to safety related loads. The fifth battery / charger is used to power non-safety related loads.
Each distribution system will normally be supplied by the battery charger and backed up by the floating battery bank which is sized to carry all connected loads for two hours .
upon the loss of the normal supply. Each pair of safety related chargers will be sup-plied from separate 480 V emergency buses. The non-safety related battery charger is also powered from one of the 480 V emergency buses. Each of the five battery bank units
.is located in a separate seismic Category I room.
- Air scpply to each of the five separate battery rooms is through a separate seismic Category I exhaust duct. The five separate battery room exhaust ducts are connected to a seismic Category I duct header teminating at the suction of two exhaust fans. each powered from separate emergency buses. We reviewed this design with respect to the possibility of. single events in the coninon exhaust duct causing redundant batteries to .
1 be inoperable. As a result of our review, we have concluded that the design meets the single failure criterion, and is acceptable.
Four redundant 120 volt vital a-c distribution buses are provided to supply power to the plant protection system instrumentation and associated circuits. Each a-c vital
} bus is supplied separately from an inverter. Each pair of inverters is normally l supplied from separate 480 i emergency buses and upon loss of normal supply, each one of the four inverters is automatically fed from its respective one of the four safety related battery units. There are also two redundant 480/240/120 volt distribution transformers, one per pair of vital buses, and each transformer is Connected to separate 480 V emergency buses. These transformers are used as an alternate source of power to the vital buses in the event of inverter failure.
The switchyard battery system consists of two separate batteries and related distribution systems. These power sources provide control power to all switchyard breakers.
On the basis of our review, we have concluded that + d-c emergency onsite power system satisfies the requirements of AEC General Design Criteria Nos.17 and 18. IEEE 308-1971, and the reconinendations of AEC Regulatory Guides 1.6 and 1.22 and is acceptable.
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- ,, *( 9-1 9.0 AUXILIARY SYSTEMS
,In the course of our' review, we have' directed our attention to the design of these auxiliary systems, including any safety related objectives of the respective systems, and the manner in which these objectives are achieved.
1The auxiliary systems necessary to assure safe plant shutdown include portions of the chemical and volume control system. residual heat removal system, the component
. cooling water. system, the river water system, the control room air conditioning system, the diesel generator auxiliary systems and certain ventilation systems. ' The systems.-
necessary to assure safe handling and adequate cooling of the spent fuel include the fuel handling system and the spent fuel cooling system. These systems are discussed further in the following sections.
We have also reviewed those auxiliary systems whose failure would not prevent safe reactor shutdown but could, directly or indirectly, be the potential source of a -
radiological release to the environment. These systems include the auxiliary building -
ventilation system, the compressed air system and the sampling system. From our review of the design of these systems, we find they are comparable in des $gn and function to other PWR facilities. that have been previously reviewed and approved. On this basis, we conclude that these auxiliary systems are in compliance with the applicable. rules and regulations and are acceptable.
9.1 Chemical and Volume Control System The chemical and volume control system is provided to adjust the concentration of chemical neutron adsorber (boron) in the reactor coolant for reactivity control, main .
tain the proper water inventory and concentration of corrosion inhibiting chemicals in.
the reactor coolant' system, provide required seal water injection to the reactor coclant '
pump seals,and remove corrosion products'and fission products from the reactor coolant.
Portions of the system also supply high pressure injection for emergency core cooling '
and is designed as seismic Category 1. To evaluate system safety, failures or malfunc-tions were assumed concurrent with a postulated loss-of-coolant accident and the con-sequences evaluated.
On the basis of the similarity of the design of the Beaver Valley Unit I chemical and volume control system to that of previously reviewed and approved systems, we have concluded that the design is acceptable.
9.2 Fuel Storace and Handling The handling of new and spent fuel will take place in the seismic Category I con-tainment and the adjoining fuel and decontamination structures. Within the containment structure, handling and transfer operations will be accomplished under borated water utilizing the reactor cavity manipulator crane, fuel upender and fuel transfer tube.
The borated water serves multiple purposes in that it assists in conjunction with the control rods in keeping the core approximately 10 percent ak/k subcritical during the
'a refueling operations in addition to serving as a transparent shielding material during fuel handling operations and a cooling mdium for the spent fuel assemblies.
New fuel will be stored within the fuel and decontamination building. The assemblies will be stored vertically in steel racks such that K,ff will be limited to 0.90 even if 1 the assemblies were flooded with unborated water. We conclude the design is acceptable.
Spent fuel assenblies will be stored in the spent fuel storage pool in the fuel and
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decontamination buildings. There they will be stored vertically in floor mounted stain-less steel storage racks having 21 inch center-to-center spacing between fuel asserblies. '
The pool ie-a seismic Category I reinforced concrete pool lined with stainless steel plate. ,It has been designed to accommodate a total of one and two-thirds cores plus 11 i spare fuel assemblies.
Whereas the water in the refueling cavity, transfer canal and spent fuel storage pool will contain approximately 2000 ppm of baron, the 21 inch center-to-center spacing of the spent fuel racks is sufficient to assure that the K,ff would not exceed 0.90 even if the pool were to be filled with unborated water. Therefore, we conclude the spent fuel storage design is . acceptable.
l The pumps, heat exchangers, piping and valves, as well as the component cooling water lines to the heat excharjers of the spent fuel pool cooling system are designed to seismic Category I. Additional assurance that cooling water will be available to remove the heat has been provided from the seismic Category I river water system or the engine driven fire ptsnp. Further, the heat exchangers have been designed such that they can be isolated to permit temporary pool water cooling connections to accomplish circu-lation through the heat exchanger by means of a temporary pump taking suction from the pool.
The fuel pool water purification system has been provided to maintain the water quality of the spent fuel ' pool and to process the refueling water during refueling opera tions. The system is independent of the spent fuel pool cooling system and has been designed as non-seismic Category I. We have reviewed the system arrangement and have determined that failure of the system would not drain the fuel pool or affect safety related systems.
We conclude that spent fuel pool cooling and cleanup svetems will perform their intended functions and are acceptable.
In addition to the fuel handling equipment util' ired in transporting and placement of individual fuel assemblies, the containment structure contains an overhead polar bridge having a capacity of 200 tons, plus two trolleys. One trolley has a 130 ton hoist and the other trolley has a 15 and 130 ton capacity hoists.
The applicants have indicated that if during the refueling process a crane handling accident should occur where the redundant seals behen the reactor vessel flange and refueling cavity were to fail, all essential electrical equipment in the compartments below would be above the flood level which would result from the release of all the ,
refueling water above the seals being released to the containment basement.
A 125 ton overhead handling system has been provided for lifting and transporting the spent fuel cask. Its bridge travels the length of the decontamination building which is oriented such that the cask travel does not pass over the spent fuel pool.
Further, raising and lowering of the cask in its laydown area will be accomplished in a steps to limit the possible drop height.
The separate cask loading pool is such as to permit the fuel assemblies to move under water from the spent fuel storage pool to the spent fuel cask through a 24 inch slot separating the two pools. The depth of the slot is above the stored fuel assemblies, thereby providing assurance that the loss of water via the slot could not uncover the fuel.
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- _ , 's 9-3 The physical arrangement of the fuel Nol, cask loading pool and the cask handling equipment is such that a cask drop ar 1 dent would not cause damage to the stored spent I- fuel assemblies. We have review the overhead handling system and have determined that its design, in con.ie.alon with the pool arrangement, will prevent a dropped fuel cask from enter',..g the spent fuel pool or affect other safety related systems. Our deter-mination k contingent upon acceptable arrangements by the applicants to perform 6 pre-opewonal test of the overhead hanuling system in accordance with the requirements of industrial standards and manufacturers recommendations. In addition, a periodic testing program requirement will be incorporated in the technical specifications to provide for testing and maintenance of the crane prior to each planned period of crane handling opera-tions. Based on the above, we conclude the design of the fuel and cask handling systems are acceptable.
9.3 Water Syst ms 9.3.1 Component Cooling Water System The primary component cooling water system is composed of the following three 4 sub*vstems; component cooling water subsystem, chilled water subsystem and neutron shield tank cooling water subsystem. Only a portion of the component cooling water system is required for safe shutdown. The other systems have been reviewed and we have determined that acceptable isolation valving has been provided to isolate these systems from the main system in the event of their failure and that their failure will not affect other safety related systems.
The component cooling water system is a closed intermediate cooling system composed of three equally sized parallel pumps joined by a common header to three parallt' heat exchangers to remove the heat loads from the plant equipment including reactor coolant pump thermal barriers, react? coolant system and steam generator blowdown coolers, seal water heat exchangers, residual heat removal pump seal coolers, residual heat removal heat exchangers, boron recovery system equipment, containment penetration cooling coils, and containment air recirculation cooling coils.
As stated above, the component cooling water system will normally supply water to some safety related items required for safe cold shutdown but not for accident purposes.
In the event of a loss of the component cooling water the reactor would be brought to a safe shutdown with all reactor coolant pumps tripped. This would be accomplished by maintaining the reactor system at hot standing conditions indefinitely by decay heat mmoval via main conder.ser and main .a -
or, if necessary, with the auxiliary feed system and atmospheric dump valves. To attain a cold safe shutdown the following component cooling water system heat removal components are required; residual heat removal pump seal coolers, residual heat removal heat exchangers, primary component cooling water pumps and primary component water heat exchangers. Therefore, to attain 0 a cold safe shutdown the portion of the component cooling water system supplying the cooling water to the above components must remain operable.
The applicants state that if gross failure were to occur in the system, that repair procedures, equipment, and material will be maintained onsite to affect any repairs required. In addition, plant personnel will be trained to perform any required repairs.
We agree with the applicants that the required portions of the component cooliag water system can be maintained operable for normal shutdown requirements and that if a e
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failure in the common pump headers were to occur, that sufficient time is available to affect repairs.
He conclude that the safety related portions of the component cooling water system I
are . acceptable.
9.3,2 Residual Heat Removal System lhe residual heat removal system (RHR) has been designed to remove residual and l sensible heat from the core once the reactor coolant has been reduced to approximately 350'F and 400 to 450 psig by the steam and power conversion system. This is normally expected to be accomplished in approximately four hours. Following this, the residual heat removal system will be placed in operation and it will be capable of reducing the reactor coolant from 350'F to 140'F in approxituately 16 tours.
The RHR system consists of two pumps and two heat exchangers. If either one of the pumps and/or one of the two heat exchangers is inoperative , safe operation is not affected but the cooldown time will bo extended. The heat will be transferred to the river water via the intermediate cmponent cooling water subsystem.
The pressure barriers of the RHR system has been designed to 600 psig. To protect the system from over-pressurization, both of the RHR system reactor coolant system inlet and outlets have been provided with redundant valves for isolation, plus the feature that should the RCS pressure rise to approximately 600 psig while the RHR system is in operation, isolation valves will automatically be closed as discussed in Section 7.6, and cannot be opened unless RHR system pressure is below 600 psig.
He conclude that the design provides sufficient assurance that it Will be capable of meeting its intended functional requirements, and acceptable isolation capability has been provided to protect the system from overpressurization, 9.3.3 River Water System The river water system is functionally designed to fulfill hoth operational and accident heat loads. It consists of three motor driven vertical wet pit pump assemblies individually located in separate intake structure compartments at the river's edge.
Each pump is sized to deliver 9,000 gpm while the minimum total flow of 8660 gpm will be required for a postulated loss-of-coolant accident. The minimum total flow is required for at least one of the three charging pump lube oil coolers and seal coolers, one of the two control room river water cooling coils, and at least one of two emergency diesel generator cooling system heat exchangers. The system design and its above requirements assumes 86'F river water at either its high or low level (elevation 649 or 730 feet).
During a loss of off-site powe , one pump will be powered by one emergency bus, the second pump will be powered by the other emergency bus. The third pump is not normally connected to either bus but can, if necessary, be connected to either bus but not both o emergency buses at the same time.
The river water system has been designed as a seismic Category I system and is tornado and missile protected up to where the discharge from the various heat exchanger discharges enter the turbine building.
In addition, all components, pumps and heat exchangers can be individually isolated during normal operation to permit repair and maintenance. A containment isolation phase B signal following an accident causes the river water flow to be diverted from its normal operating heat load of the primary componert cooling water heat exchangers
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- to the recirculation spray coolers. In addition, following an accident a safety i - injection signal will initiate' the flow of river water to the diesel generator cooling f . system heat exchangers.
We have reviewed the system to determine that no single failure will preclude the' system from delivering the required cooling water to the essential plant equipment.
Redundancy in pumps, critical. valving and piping have been provided such that a single L failure will not disable the system.
On the basis of our review, we conclud; that the system is acceptable.
s 9.3.4 Ultimate Heat Sink
- The intake structure and river water system has been designed to serve as the ultimate heat sink. - To provide additional assurance that the system will remain operational, the applicants have modified the original intake structure design by erecting an additional concrete structure above the standard project flood level (elevation 705 feet) to protect the essential equipment housed within separate -
compartments from the probeble maximum flood level (elevation 730 feet).
In response to the staff's concern that developed during the review relating to the adequacy of the present design to withstand the impact and coincident explosion of a gasoline barge, the applicants have indicated that extensive additional modifications would be required to withstand the postulated event.
Rather than carry out these modifications, the applicants will provide an alternate river water system which will provide river water in the event of failure of the main .
structure. To support this revision to the application and the above concern, the applicants presented criteria that will form the specific system design requirements of the alternate system.
We require that the alternate intake structure contains a river water pumping system that provides for redundant headered pumps and valves. In addition, we require that the system be capable of being automatically started and aligned to the plant upon sensing a low river water flow in the main system and ttst the alternate river water pumps be capable of being powered or driven from on-site power supply sources' .
The applicants have provided information as to the location, type, design description, capacity and compatibility of the alternate ultimate heat sink with the remainder of the plant. We will review this information to determine that the design is adequate to supply plant ecoling requirements following a gasoline barge collision and explosion with th: main intake structure and thereby enable the plant to safely attain and main-tain a cold safe shutdown, and will report the results of our evaluation in a supplement to the Safety Evaluation Report.
9.4 Other Auxiliary Systems 9.4.1 Compressed Air System The facility has two separate compressed air systems to provide air as required a for normal station service and instrumentation. These systems are the station air system and the containment air system. The actuation of a number of wklves in the reactor coolant pressure boundary and other seismic Category I systems utilize compressed air. However, the applicants state that operation of either system is not required for station safety but are necessary for station operation. Therefore, each system has been provided with 100 percent design capacity redundant non-lubricating 9
9 -6 type compressors and compressor power sources. Each compressor is supplied with an intakeTilter and after-cooler.
Based on our review of the system, we conclude that it is not required for engineered safety functions, that the failure of the system will not affect safety related systems, and therefore is acceptable.
9.4.2 Equipment and Floor Drainage System The information contained in the application regarding the drain system stressed the protective measures taken to collect potentially radioactive fluids that could possibly be released from various systems throughout the plant. In addition to this possibility, due to the elevation of seme spaces within the plant with respect to the elevation of the probable maximum flood (730 feet) we also evaluated the potential for essential areas to become flooded via the drainage system.
The applicants ' indicate that the method of handling and disposing of the collected liqulds is dependent upon whether they are high or low level activity fluids and if they are aerated or non-aerated. Aerated fluids will be collected via sumps or other collection points throughout the plant and directed to either the high or low level liquid waste disposal drain tanks depending upon the activity level of the fluid.
Non-aerated reactor coolant fluid will be rotated to the boron recovery system for proc-essing and recovery. Sumps located in critical areas are provided with a double full-sized pump arrangement where the operation of the pumps will be periodically alternated.
One pump is on automatic service and the other is on standby. In the event of a failure of the first pump or a leakage exceeding the pump capacity, the sump level would raise to a point where the second pump would be started. Further, a high level alarm is provided.
We conclude that the measures taken to collect and control the radioactive liquids are acceptable 5 and that adequate measures have been taken to assure that spaces housing essential equipment will not become flooded and thereby disable or degrade essential equipment.
9.4.3 Air Conditioning, Heating Cooling and Ventilation System The ventilation systems have been designed to provide a suitable environment for personnel and equipment, and prevent the spread of radioactive contamination. The exhaust of areas subject to radioactive contamination are capable of being diverted through the main filter banks in the supplementary leak collection and release system when required. In addition, the required flow rates will be sufficient to remove heat from warm areas and to prevent the accumulation of condensation or toxic vapors. For these areas the rate of air supplied will be such that slight negative air pressure will be maintained so as to assure only inleakage.
The supplementary leak collection and release system has been provided for areas o subject to radioactive contamination to replace the normal ventilation system, upon being activated by radiation monitor signals, and process the flow through the rain filter banks in the system. Its primary purpose is to assure that radioactive leakage from equipmtnt, radioactive release due to a fuel handling accident or a release in the waste gas storage areas is collected and filtered for iodine removal prior to discharge to an elevated atmospheric vent.
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!s. *\ 9 g To assure the habitability of the main control room area following a postulated loss-of-coolant accident, the applicants have supplied an emergency pressurization system capable of providing 400 cfm of bottled air'for one hour (see Sections 6.2.1 and 6.4).'
9.4.4 Diesel Generator Fuel Oil Storace and Transfer System The emergency diesel engine fuel oil system consists of two buried storage tanks, four fuel suction strainers, four 100 percent design capacity (each 10 gpm) transfer pumps, two 550 gallon capacity day tanks (2-1/2 hours at full load), two 550 gallon capacity engine mounted fuel oil tanks, two engine-driven and two motor-driven positive displacement pumps that take suction from the engine mounted fuel oil tanks and discharges 1through four duplex type fuel filters to the engine fuel injectors.
The fuel system, including transfer pumps, storage tanks, piping and valves are designed as seismic Category I. The tanks have been protected from external corrosion.
All fuel lines are protected from external corrosion by being within enclosed heated buildings, encased in concrete or, for those lines directly buried in ground, coated with bitumastic enamel and wrapped with coal tar saturated asbestos.
The combined capacity of the two 20,000 gallon fuel oli storage tanks is adequate to supply fuel for more than seven days of full load operation. The discharge 'of the four fuel transfer pumps are cross connected between the two diesel engines such that it is possible for fuel to be drawn from either storage tank for either diesel engi_ne.
In addition, there are six major oil distributors within four miles of the site.
The applicants have stated that the fuel oil storage will be periodically checked for water, sediment, contamination and deterioration in order to avoid the possibility of all fuel being in an unacceptably degraded condition.
We conclude that the diesel fuel oil storage and transfer system is acceptable.
9.4.5 Diesel Generator Auxiliary Systems The capacity of the, air starting system has been selected such that it is capable i of five engine starts without outside power. To ennance the "first try" starting
! reliability of the diesel engines, the engine's c*osed cooling water will be heated which in turn will heat the engine's lubricating oil cooler, heating the oil to 125 to IS5*F. In addition, the compartment ambient air temperature will be limited to a l minimum temperature of 65'F by means of electric space heaters. Each compartment has
! its own outdoor air intake with redundant damper assemblies whose motors are energized by emergency power to provide assurance that engine combustion air and _ compartment cooling air will be available during operation.
During engine operation heat from the lubricating oil and closed cooling water -
System will be rejected to the river water system via the cooling water heat exchangers.
, Redundant river water supply piping has been provided from the river water pumps in the
) intake structure to both of the diesel engine cooling water heating exchangers to pro.
L vide protection against a single failure.
a We conclude that the design of the auxiliary systems for the onsite emergency power supply system can perfonn their intended functions and therefore are acceptable.
9.4.6 Fire protection System The fire protection system is designed to provide the capability to detect and extinguish one or a probable combination of fires that may occur. In accomplishing this purpose, the system:
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9 -8 (1) Complies with the standards of the National Fire Protection Association.
(2) Is also based upon the recommendations of the Nuclear Energy Property insurance Association.
(3) Supplies wet-pipe sprinkler systems, deluge systems, carbon dioxide systems and hand operated (manual) fire fighting devices throughout all potentially hazardous areas of the plant.
The primary water supply is the Ohio River, from which water will be drafted by means of either an electric motor driven fire pump or a diesel engine driven fire pump.
Supply lines to all areas I,erved are located in the ground, below maximum frost penetra-tion. Exposed piping is not circuited through an unheated area. Water sprinkler fire protection is provided in areas where combustibles are located, while spray heads are used in conjunction with electrical transfonners, turbine room oil auxiliaries and the main exhaust charcoal filters. In addition to the " wet" systems, there are two carbon dioxide sub systems. One of these is used for protection of the high-pressure turbine enclosure. with the other portion serving the diesel generators and cable vault.
Activation of any system will also cause annunciation of the event in the main con-trol room. A local audio signal will also be activated to warn personnel of the dis-charge of carbon dioxide gas.
In addition to the above, the applicants erphasire that the fundamental basis of the design is the control at all times of the amount of combustible material allowed to accumu'f ate and to minimize the amount of combustibles used in the construction of the plant.
To safeguard the protection systems (1) dual fire pumps, each of 100% capacity, are provided (2) heat detection devices will be periodically tested, using pcetable equip-ment. (3) waterflow tests will be conducted on a yearly basis. (4) outdoor systems will be allowed to discharge to confirm their operability, and (5) all hand or portable extinguishing equipment will be periodically inspected and tested.
On the basis of our review we conclude that the fire protection systems conform with the intent and requirements of Criterion No. 3 of the AEC General Design Criteria and are therefore acceptable.
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10.0. STEAM MD POWER COWERSION SYSTEM 10.1' S m ary Description t M N4 The steam and power conversion system is of conventional' design, simiiar'to those of-previously approved pressurized water reactor plants. The systen is designed to remove heat energy from the reactor coolant by three steam gencrators and convert it to electri-cal energy by the turbine driven generators. The condenser will transfer unusable heat in the cycle to the condenser coolirg water, ,The entire system is designed for the maximum expected energy from the nuclear stene supply system.
Upon loss of load,' the system is capab4 of dissipating the energy in the reactor coolant through bypass valves to the contcenser and through' dump valves, and safety valves to the atmosphere.
10.2 T'urbine Generator The turbine generator is a tendem compound arrangement of & double flow high pressure turbine and two double flow low pressure turbines driving a direct coupled generator at .
1800 rpm.
The electro-hydraulic' control system will control the speed of. the turbine by modu-lating the turbine inlet steam control valves to control the steam flow to the 'high t
pressure turbine, in addition the control system will trip the turbine upon sensing an overrspeed condition.1ow condenser vacuum. generator faults. turbine support system faults. steam generator high-high water level and reactor trips.
Any of these conditions will cause the control. system to close the turbine high pres-sure inlet stop valves and control valves and the intercept valves in the reheater sys-tem between the high pressure turbine and the low pressure turbines.
The stop and control valves will be exercised periodically by closing and opening
'them to detect possible valve stem sticking. Also during normal station startup the turbine overspeed trip point will be checked by running up the turbine speed. In addition, a device has been provided for testing the turbine overspeed trip mechanism during normal operation without actually overspending the turbine. The testing interval for the valve tests and the overspeed control systems test will be specified in the technical specifications.
We conclude that the turbine generator and its control syt, tem are acceptable.
10.3 Main Steam Supply System The main steam piping will pass the steam generated in the three steam generators to
.the turbine. the atmosphere steam dump lines, safety valves and the turbine-driven steam generator auxiliary feedpump. The following portions of the main steam system have been designeo ?s seismic Category It steam generators and supports, piping from steam generators t6 the main steam isolation and non-return valves, piping from main o steam lines to the decay heat release valve as well as the lines to the turbine-driven steam generator auxiliary feedpump.
Each 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 a main steam iso-lation and non-return valves to isolate the system during a steam line break accident.
The three main steam lines are headered in the turbine building to equalize steam temperature and pressure. The system then takes four steam lines from the header and delivers the steam to the high pressure turbine. A turbine stop valve is located in h
10-2 each ef the four lines to rapidly stop steam flow to the turbine if a turbine or reactor trip occurs.
Bt, sed on our review of the system, we conclude that the design is acceptable.
10.4 Other Features A twin shell, single pass, divided water box condenser is provided to condense the exhaust from the two low pressure turDines, the discharge from the 18 turbine steam by-cass valves and miscellaneous drains. The total hotwell storage capacity is equivalent to approximately four minutes of full load operation. Two twin element, two-stage steam jet air ejector units plus inner and after condensers remove ncn-condensable gases.
The effluent f rom the air ejectors is monitored for radiation. We conclude that the main condenser and evacuation system is acceptable.
A turbine bypass N tem has baen provided to dump excess steam generated by the nu-clear steam supply system to the condenser without relieving it to the atmosphere for a variety of operating modes, such as a large step load decrease, normal shutdowns, hot standby, and startup and physics testing. The system consists of two 24-inch turbine steam bypass lines, each having nine bypass valves that discharge into the condenser.
The total discharge capacity of the 18 valves is the equivalent of 85 percent of full steam flow. Those valves are prevented from opening upon the loss of condenser vacuum.
In the unlikely event such a situation should arise, the excess steam will be relieved to the atmosphere through the three atmosphere steam dump valves or the fifteen main steam safety valves. We conclude that the turbine bypass system is acceptably designed to perform its intended function and is acceptable.
The circulating water system, a non-safety related system, has bean designed to dis-sipate to the atmosphere the heat obtained from the main surface condenser of the turbine generator unit. It is composed of one natural draft cooling tower, four circulating water pumps, piping and associated valves required for maintenance. The four pumps de-velop a total flow of 507,400 gpm. The circulating water piping is made up of two 103 inch diameter conc. rete pipes.
To provide makeup to the cooling tower for evaporative and drif t losses as well as to replace that lost through blowdown, water will be supplied to the closed loop circu-lating water system by discharging the turbine plant raw water and river water system flow into the circulating wa;er condenser discharge lines.
The applicants state that in the unlikely event of a circulating water system failure in the turbine building (a) the turbine building does not contain any equipment that is safety related, (b) there are no possible flow paths for flood waters below elevation 707.5 feet, and (c) the maximum elevation that the flood water can attain within the turbine building will not exceed elevation 702.8 feet. We conclude that the circulating a water system is acceptable.
Condensate will be withdrawn from the hotwell by two half-canacity (9,700 gpm each) motor-driven pumps which discharge into a single header supplying condensate to the two parallel trains of feedwater heaters. The discharge of the tro trains join to form a common supply for two half-capacity steam generator feedptanps. They in turn discharge through two half-sized high pressure feedwater heaters. Manual valves have been provided to permit isolation of one train for maintenance without a station shutdown.
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10-3 The portion of the feedwater system between the containment isolation valves outside of contja nment and the steam generators has been designed as seismic Category I.
Also, this portiere of the system is acceptably protected from missiles, pipe whip and environmental conditions.
We cone.lude that the condensate and feedwater systems are acceptably designed to perform their intended functions and are, therefore, acceptable.
[ 10.5 Auxiliary Feedwater System The auxiliary feedwater system supplies water to the steam generators for reactor decay hnt removal if the normal feedwater sources are unavailable due to loss of offsite power or cther malfunction. The system includes one 100% capacity turbine driven and two 50% capacity motor driven auxiliary feed pumps. Steam for the turbine driven pump is taken from each of the three mcin steam lines upstream of the steam generator isola-tion valves. The motor driven pumps receive power from the 4 kV vital buses.
The turbine driven auxiliary feed pump, rated at 700 gpm, and 2696 feet total dif-ferential head and the motor driven auxiliary feed pumps, rated at 350 gpm and 2696 feet total differential head, receive suctt from the 140.000 gallon primery plant demineral-ized water storage tank or the river wate
- system. Both supply sources and the entire auxilisry feed.ater system are designed to seismic Category I.
Both the motor driven pumps and the turbine driven pump discharge to the three steam genera tors. Feedwater flow will be controlled from the control room by remotely operated flow control valves in the supply lines to each steam generato.s The turbine driven pump automatically starts upon any one of the following signals:
(1) two out of three low-low steam generator level signals (mot.nted on each generator) from any one of the three steam generators. (2) two out of three under voltage signals iMicating a loss of power, (3) manual start from control room. (4) manual start from local station.
The two motor driven auxiliary feed pumps, also part of the auxiliary feedwater system. are initiater oy any one of the following signals: (1) two out of three low-low steam generator level signals (mounted on each generator) from any one of the three steam generators. (2) a trip signal from both of the two steam generator main feed pumps. (3) safety injection signels, (4) blackout signal. (5) manual start from control room (6) manual start from local station.
We have reviewed the auxiliary feedwater capability with respect to piping system breaks outside of containment (high energy). Based on this review we have detemined that if a high energy line fails outside of containment, and assuming a concurrent single active component failure in the auxiliary feedwater system, that a minimum auxiliary feedwater flow of 350 gpm is available to two of the three steam generators
'ir removal of decay heat.
On the basis of our review we have concluded that the auxiliary feedwater system 0 will perform its intended function and is acceptable.
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11-1 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1 Design Ot4ective and Criteria The radioactive waste management system for Beaver Valley Unit 1 is des 1 Dned to provide for the controlled handling and treatment of radioactive liquid, gaseous and solid wastes. The applicants design objective for these systems is to restrict the amount of radioactivity released from normal plant operation to unrestricted areas to within the limits set forth in 10 CFR Part 20.
The technical specifications issued as part of the operating license will require the applicants to maintain and use existing plant equipment to achieve the lowest practicable releases of radioactive materials to the environment in accordance with the requirements of 10 CFR Part 20 and 10 CFR Part 50. The applicants will also be required to maintain radiation txposures to inplant personnel and the general public "as low as practicable" in conformance with the requirements of 10 CFR Pa-t 20.
Our evaluation of the design and expected peri rmance of the waste managoent sys-tem for Beaver Valley Unit 1 is based on the following design objectives:
Liquids (1) Provisions to treat liquid waste to lic it the expected releases of radioactive ma-terials in 11guld effluents to the environment to less than 5 Ci/yr/ unit, excluding tritium and noble gases.
(2) The calculated annual exposure to the whole body or any organ of an individual at or beyond the site boundary not to exceed 5 mrem for expecteo releases.
(3) Concentration of radioactive materials in liquid efflur nts not to exceed the limits in 10 CFR Part 20, Appendix B. Table II, Column 2. for the expected and design releases.
Gaseous (1) Provisions to treat gaseous waste to limit the expecteo release of radioactive ma-terials in gaseous effluents from principal release points so that the annual aver-age exposure to the whole body or any organ of an individual at or beyond the site boundary not to exceed 5 mrem.
(2) Provision to treat expected radiciodine released in gaseous effluent from principal release points so that the annual average exposure to the thyroid of a child through the pasture-cow milk pathway be less than 15 mrem.
(3) Concentration of radioactive materials in gaseous effluents not to exceed the limits in 10 CFR Part 20 Appendix B. Table II, Column 1, for thu expected and design releases.
Solid (1) Provisions to solidify all liquid waste from normal operation including anticipated operational occurrences prior to shipment to a licensed burial ground.
, (2) Containers and me+ hod of packing to meet the requirements of 10 CFR Part 71 and applicable Department of Transportation regulations.
The following sectionr. present our evaluation of the liquid, gaseous, and solid waste treatment systems, the design codes and quality assurance criteria, and the radiation monitoring of process effluents and of inplant creas. The liquid, gaseous, and solid waste systems are designed to accomcdate the waste produced durinD simultaneous opera-tion of Units 1 and proposed Unit 2. In addition to that provided for Unit 1, a separate chemical and volume control system will be provided for proposed Unit 2. The boron 9
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11-2 recycle system, steam generater blowdown treatment system, the liquid waste disposal system, most of the waste gas processing system, and the solid weste processing system will be shared by Unit 1 and Unit 2 when it becomes operational.
Our evaluation and estimates of releases are based on our model which is adjusted to this plant. The model uses somewhat different values for the parameters than those of j the applicants. Our calculated effluents are, therefore, different than the applicants.
- 11.2 Liquid Waste Systems The liquid waste treatment system is divided into the chemical and volume control system, boron recovery system, deaerated system, aerated system, and the steam generator blowdown system. The Ber;er Valley Power Station will have an integrated liquid waste processing system, which will serve Unit 1 and proposed Unit 2. A separate blowdown treatment system will be installed within Unit 2 for use of both units.
Treatment of the waste is dependent on the source, activity and composition'o? the particu*.e liquid weste and on the intended disposal procedure. Cross connection between the subsystems provides flexibility for processing by alternate methods. Treated wastes will be handled on a batch basis es required to permit optimum control and release of radioactive waste. Prior to the release of any treated liquid wastes, samples will be analyzed to determine the type and amount of radioactivity in a batch. Based on the analytical results, these wastes will either be recycled, reprocessed, or released.
Radiation monitoring equipment will automatically trip a valve on the discharge pipe terminating the release of liquid waste if the levels of activity are above a pre-det m ined value.
The chemical and volume control system (CVCS) consists of mixed t'ed demineralizers for the removal of cation and anion radios, tive products and cation bed demineralizers for the removal of cation radioactive prodt ;ts which,during normal operation, processes a reactor coolant letdown stream of about 60 ppm, tends it to the volume control tank and then returnsit to the primary coolant system.
The boron recovery system consists of two 75 gpm degassifiers to strip gases from the total CVCS letdown stream before the stream is sent to the volume control tank. The ba-lance of the system consists of a train of two cation exchangers, two 190,000 gallon coolant recovery tanks, two 15 gpm evaporators to remove and concentrate radionuclides, and two 12,000 gallon test tanks. This portion of the system will process a small part of the CVCS letdown to reduce the boron concentrations during core life. Af ter processing, it will be routed to the primary water storage tank for reuse or will be discharoed. The l evaporator bottoms containing concentrated boron and radioactive prcducts will be sent to l
the evaporator bottoms hold tank or boric acid hold tank for reuse or to the solid radioac-l tive waste system for solidification and offsitt shipment.
The deaerated liqu'Id waste system consists of two primary drain tanks which collect
- drains and leakoffs from various primary system equipment. The drain tanks will be pumped to the boron recovery system for processing through the cation exchangers and evaporator.
l Our evaluation of the boron control system and deaerated waste system assumed that 720 gal / day / unit of deaerated wastes from the primary drain tank and 860 gal / day / unit from the shim bleed for boron control will be processed by the boron recovery system and that 90% will be recycled and 10% discharged.
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~ ' 11 -3 The aerated waste system consists of two 5,000 galbn and two 2,000 gallon waste drain' talks (two 7,500 ' gallon ta'nk's will be added to this stream in conjunction with Unit 2 construction), a G gpm waste disposal evaporator, two 3,000 gallon distillate test '
< tanks, and a polishing mixed-bed demineralized. This system will process. liquid waste from the containment and auxiliary buildings, laboratory drains and sampling sources,ind demineralized sluice. The wastes ney be reprocessed or discharged.1 The evaporator bottoms will be sent to the solid radioactive waste system for solidification and offsite -
shipment.
Our evaluation assumed 885 gal / day / unit of aerated wastes will 'be processed by"the evaporato* ano polishing demineralized and 100% of the distillate will be discharged.
The blowdown treatment system for Unit I consists of a flash tank which will flash L 90% of the blowdown to the main condenser and the remainder will be processed by the-aerated waste system. When Unit 2 is installed, the blowdown from Units 1 and 2 will' be processed by a separate system consisting of two 50,000 gallon blowdown hold tanks after the flash tanks, two 20 gpm blowdown evaporators, two 12,000 gallon test tanks, and a polishing demineralized. The processed blowdown will then be sent to the primary -
water storage tanks for reuse, or discharged.
Our evaluation assumed a blowdown rate of 20,000 gal / day / unit, before flashing,~ pro-cessed by evaporation and demineralization by either of the treatment systems with 90%
recycled and 10% discharged.
In addition to the sources listed above, low activity turbine' building drains and laundry system drains will be released without treatment. We accept-these negligible release systems without treatment.
We estimate that a total of about 0.3 C1/yr, excluding tritiun, will be discharged '
when Unit'l only is operating and 0.4 Ci/yr/ unit when Unit 1 and 2 are sharing the sys-tem. The applicants have estimated 0.014 C1/yr/ unit will be discharged when Unit l' only
, ts operating and when Unit I and 2 are operating. Based on operating experience of other pressurized water reactors, we esticate that approximately 350 C1/yr/ unit of tritium will be disch&rged. The applicants,have estimated that 506 Cijyr/ unit of tritium will be released.
We calculate the whole body and critical organ dose to individuals to be less than 5 mrem /yr. The applicants also calculate these doses to be less than 5 mrem /yr.
Based on our evaluation that the liquid radweste system will process the expected wastes to less than 5 C1/yr/ unit, with a resultant dose of less than 5 mrem /yr from both l
units, we conclude that the system is acceptable. We conclude that the system is capa-ble of processing redwastes in accordance with 10 CFR Part 20 and '10 CFR Part 50, and to levels which are "as low as practicable",
11.3 Gaseous Waste Systems O
The gaseous waste treatment system is comprised of a gaseous waste recycle system, gaseous waste disposal system, containment internal recirculation and purge system, a sweep gas system for liquid tanks. and an air ejector treatment system.
The gaseous waste recycle system consists of four charcoal delay beds for each unit containing 1.5 tons of charcoal, a gas surge tank for each unit (50 cubic feet,100 psig),
and three gas decay tanks for both units (132 cubic feet each,100 psig). Gases strip-ped from the primary coolant letdown flow in the chemical and volume control system, and 0
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11-4 from the deaerated wastes and shim bleed in the boron recovery system, will be processed
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by this system. The gases will pass through the charcoal dhcay tanks where the krypton and xenon gases will be delayed and the iodine retained. T'e gases will then pass to the surge tank from where most of the gases will be recyclel. A portion of the stream will be removed from the surge tank to one of the three decay tanks for holdup before processing through the gaseous waste disposal. system (charcial adsorbers and HEPA fil-
~ters) and released to the process vent above the cooling'tontr.
Our evaluation considered a gas flow of 500 cubic feet /dey/ unit through the charcoal beds and 5 cubic feet / day / unit to ti.e shared gas decay tanks. We calculate that the '
charcoal delay beds will provide 30 days decay for xenon and).5 days decay for krypton and that the gas decay tanks will provide storage time in exce;s of 30 days.
The condenser air ejector treatment systems consist of processing through the gaseous waste disposal system (charcoal adsorbers and HEPA filters). After Unit 2.is in opera-tion, the air ejector gases from Unit 1 and 2 will be processed through charcoal delay ^
beds containing 11 tons of charcoal. then to the gaseous waste disposal system and re-leased to the process vent.
The auxiliary and turbine building effluents will be released to the atmosphere un -
treated. The ventilation system for the auxiliary building has been designed to ensure -
that air flow will be from areas of low potential to areas having a greater potential for the release of airborne radioactivity. During abnormal operations.the auxiliary -
building ventilation may be processed through the supplementary leak collection and re-
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lease system (charcoal adsorbers and HEPA filters) and to the containment vent.
The containment ventilation and purge system consists of two 4.000 cfm internal re-circulation filters (charcoal adsorbers and HEPA filters) to reduce the iodine and par--
ticulate activity before purging through another train of charcoal adsorbers and HEPA filters, and discharged to the containment building vent. :Our estimate of releases from the containment purge assumes a 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation of the recirculation system before -
purging.
The sweep gas system consists of fresh air replacement in liquid tanks such as the waste drain tanks, coolant recovery tanks, and test tanks to prevent these gases from diffusing into buil' ding atmospheres, processing them through' the gaseous waste disposal
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system (charcoaladsorbersandHEPAfilters),andreleasetotheprocessvent.
The steam generator blowdown flashtank and reboiler vent will be routed to the main I~ condenser eliminating this effluent stream, Main steam will be used for the gland seal-ing system and the gland seal leakoff steam will be routed to the gland seal condenser.
l-The noncondensibles will pass through a charcoal adsorber and be vented to the roof vent and the condensibles drained to the condensate storage tank. We estimate the gland seal e
gaseous releases to be negligible.
We estimate that a total of approximately 1150 Ci/yr of noble gases and 0.09 Ci/yr of iodine-131 will be released when only Unit 1 is in operation. We estimate 1150 Ci/yr of noble gases and 0.00 C1/yr of iodine-131 will be released from Unit 1 when both Unit i and 2 are operating. The applicants estimate approximately 2250 Ci/yr 01 noble gases and 0.0064 C1/yr of iodine-131 will be released when Unit 1 only is in operation and '
2020 C1/yr of noble gases and 0.0064 C1/yr of iodine-131 from Unit I when Unit I and 2 are in operation. The applicants considered less auxiliary building leakage than we have.
and has considered higher charcoal filter efficiencies.
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11-5 We calculate the whole taiy dose due to noble gases to be less than 5 mrem /yr to an indlyidual at the site boundary and beyond. We calculate the dose to a child's thyroid due to the pasture-cow-milk chain to be less than 15 mrem /yr. The applicant also calcu-lates a whole body dose of less than 5 mrem /yr and a dose to a child's thyroid of less than 15 mrem /yr.
Based on our evaluation that the gaseous radwaste system will process the expected noble gases to a calculated whole body dose of less than 5 mrem /yr to an individual, and process the expected iodine-131 to a calculated child's thyroid dose via the pasture-cow-milk chain of less than 15 mrem /yr, we conclude the system is acceptable.
11.4 solid Radwaste The solid radwaste system will be designed to collect. monitor, process, package, and provide temporary storage for radioactive solid wastes prior to offsite shipment and disposal in accordance with applicable regulations.
Spent demineralized resins from the various treatment systems will be transferred to a resin hold tank. The resins will then be dewatered. The resin sluice water will be processed by the aerated waste system.
Evaporator concentrates fro;;. the waste disposal evaporator. and the boron evaporators will be pumped to an evapurator concentrates shipping container where it will be mixed with a solidifying absorbent.
Expended filter cartridges will be placed into a cask or drum depending on the acti-vity. Other dry solid wastes consisting of contaminated rags, paper, protective cloth-ing and miscellaneous contaminated items will be packaged in drums or other ,uitable containers for disposal.
Containers will be filled and sealed by remote control when the radiation levels so require. f.11 containers will be contained and shipped in accordance with AEC and Departmer.t of Transportation regulations.
The staff estimates that approximately 5,000 C1/yr of solid wastes will be shipped offsite from Unit 1. The applicants have not made a curie estimate for solid waste.
We conclude that the solid radioactive waste system will be able to package and store radioactive wastes according to existing regulations. The system will allow handling of the wastes 50 that doses to operators of the system will be in accord with existing regu-lations cid is, therefore, acceptable.
11.5 Design The radwaste systems are designed to meet the existing codes and standarde.. The decay tanks, surge tank, and charcool beds in the waste gas decay system are dA sed for Quality Group C in accordance with AEC Regulatory Guide 1.26 and seismic Category I in ,
accordance with AEC Regulatory Guide 1.29. The liquid radwaste system is designed for l Quality Group D in accordance with AEC Regulatory Guide 1.26. The auxiliary building which houses the liquid radwaste systems is a seismic Category I structure.
- We conclude that the radwaste systems design are in accordance with appropriate codes and standards and are acceptable, 11.6 process and Area Radiation Monitorino Systems The process radiation monitoring system is designed to provide information on radio-activity levels of systems throughout the plant, on leakage from one system to another, and on levels of radioactivity released to the environment. The system will consist of 9
11-6 monitors for ventilation vent particulate and gas, elevated release particulate and gas, auxiliary building exhaust, fuel building exhaust, containment purge exhaust, leak col-
. lection area gas, component cooling water. condenser air ejector discharge, steam genera-'
tor blowdown tank discharge and composite sample, reactor coolent letdown, auxiliary steam condensate, gaseous waste disposal system gas and particulate. waste gas tank vacit ventilation, liquid waste contaminated drain monitor, liquid waste effluent moni-tor, corrponent cooling / recirculation spraf heat exchangers discharge, primary plant com-ponent cooling heat exchanger discharge, waste gas decay tanks, and four spare channels.
The gross activity of the reactor coolant will be monitored by a low range detector and a high range detector located in the reactor coolant letdown line. We believe that this system is capable of detecting for gross failed fuel t,nd is acceptable.
The area radiation monitoring system is designed to provide information on radio-activity fields in various areas. The system will consist of monitors for containment particulate and gas and a multisample particulate and gas for monitoring 12 stations in the fuel building, decontamination facility, solid waste handling area, tunnel, contain-ment area waste drain storage and handling area, laboratcry area, main control room, auxiliary building, fuel pit bridge area, new fuel pit area, decontamination area, incore instrumentation equipment area, and manipulator crane area.
The system will detect, indicate, annunciate and/or record the levels or fields of -
activity to verify compliance with 10 CFR Part 20 and keep the radiation levels as low as practicable. We conclude that the plant is acceptably provided with process and area monitoring equipment.
11.7 .Cnnelusions Based on our calculational model and assumptions, we calculate an expected whole bcdy and organ dose of less than 5 mrem /yr to an individual from gases and less than 5 mrem /yr from liquids at or beyond the site boundary. We calculate the potential dose to a child's thyroid from the pasture-cow-milk chain to be less than 15 mrem. Solid-radwastes will be contained and shipped in accordance with AEC and Department of Transportation regulations. Therefore, we conclude that the liquid, gas, and solid waste treatment systems meet the requirements of "as low as practicable" and use of state-of-the art equipment in accordance with 10 CrR Part 20,10 CFR Part 50, and Regulatory Guide 1.42; and therefore, the radwaste systems are acceptable.
We also conclude that the system is designed in accordance with acceptable codes and standards 3 and that t! 3 process monitoring system is adequate for monitoring ef fluent discharge path $ as specified in Criterion 64 of the AEC General Design Criteria.
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12-1 12.0 RADIATION PROTECTION 12.1 Shielding -
The plant radiation shielding is designed 50 that dose rates for all controlled access areas of thf plant, coupled with anticipated occupancy time, correspond to a maxi-mum total body exposure consistent with limits specified by 10 CFR Part 20. Shield wall thickness around radwaste equipment and processing systems, in addition to provisions for temporary shielding, is designed so that applicable limits of 10 7FR Part 20 would not be exceeded for the radioactivity expected in the individual system elements and for the maximum realistic occupancy of the areas outside the shield walls.
The reactor vessel and the primary loop components are shielded both by internal structures and the reactor building shell. Radiation levels in those areas designed for continuous occupancy outside the shell will be less than 1.0 mrem per hour. Areas of the auxiliary building which contain radioactivity are shielded. Access for maintenance purposes without unnecessary exposure to adjacent equipment is provided by the shielding of the tanks, pumps, filters, demineralizers and piping containing contaminated materials, as far as is practicable.
Based en our evaluation, we conclude that acceptable consideration has been given to shielding design, equipment layout bnd r9diation protection to keep exposures within applicable limits and to reduce radiation exposures during normal operation of the plant, in general agreement with AEC Regulatory Gaide 8.8.
12.2 Health Physics program The objective of the health physics program is to assure that radiation exposure to -
station personnel is as low as practicable. The health physics program at Unit 1 has and will be prepared by health physicists acting directly under the Nuclear Services Superintendent. Implementation of the program falls under the responsibility of the plant Superintendent acting through various individuals including the reactor control chemist, the radiation control foreman and the chemical and radiation technologists.
Personnel protection will be accomplished through administrative controls and pro-cedures, through the use of protective equipment and verified by personnel monitoring.
All work in contrulled areas will require an appropriate radlation work permit which requires determination and evaluation of the radiological hazards associated with the job before issuance. Exposures in-plant will be reduced by rotating personnel as-signed to tasks in high exposures areas and by prejob training and practice runs. Ex-tension tools will be used where feasible and equipment decontamination will be effected if feasible and/or portable shielding will be provided. Pemanent shielding is provided for waste treatment components as described above.
Special protective equipment includes covering garments, shielding and self-contained air-breathing units. A change room and personnel decontamination facilities are also
, provided.
Personnel monitoring will normally be accomplished by film badges or thermolumine-sent dosimeters in unrestricted areas and either one, or a combination of pocket cham-bers, wrist badges, self-reading dosimeters and pocket high radiation alarms within re-stricted areas as required. Bioassay and medical programs are under the direction of the Duquesne Light Company Medical Manager.
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12 We conclude that the applicants plan to impiement' a health physics program of suffi-7 cient scope to maintain in-plant exposures of personnel within applicable limits. Plut design criteria and health physics related equipment and procedures indicate the appli-cants intent to minimize in-plant personnel exposure, and are in general agreement with Regulatory Guide 8.8.
12.3 . Ventilation in addition to the health physics program, the ventilation systems are designed to
. provide a safe habitable environment for operations personnel, and to prevent the spread of radioactive contamination. Clean and potential radioactively contaminated areas are served by separate systems with potentially contaminated areas maintained at a negative pressure with respect to adjacent clean areas. The design criteria of the systems, the physical description of the systems as provided in the FSAR and the description of plan-ned operations provide reasonable assurance that adequate consideration has been given to ventilation design for the protection of personnel from airborne radioactivity hazards.
12.4 . Area Monitoring -
The area radiation monitoring system was designed to conform with 10 CFR Part 20 L
and Criterion No. 64 of the AEC General Design Criteria, to indicate radiation levels that may exist at various locations throughout the station where personnel are most likely- 1 to be exposed. Eighteen separate detectors are located in fifteen plant areas with three of these being in the auxiliary building. The location and range of sensitivity of these instruments is given in Table 11.3-3 of the F$AR. The monitoring equipment proposed by the applicants provides reasonable assurance that radiation levels within the plant will be acceptably monitored.
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13.0 . CONDUCT OF OPERATIONS j 13.1 ' Plant Organization, Staff Qualifications and Training f
The Beaver Valley Unit 1, plant staff will . consist of approximately 70 full time : i employees, not including clerical. help and security force personnel. The plant is under the onsite supervision of the Station Superintendent who reports to the General Superin-tendent Pmver Stations Department, who in turn reports to the Vice President Operating i
Division. The Station Superintendent is responsible for the safe and reliable operation of the plant. The plant staff consists of a Chief Engineer in charge of operations (ap-proximately 25 people) and maintenance (approximately 25 people), a Results Coordinator (approximately 15 people), and an Office Manager.
The Station Operating Supervisor, who reports to the Chief Engineer, directs the day-to-day operation of the facility and is responsible for all station operation. Reporting to him are the plant operatir.g shifts. The normal shift complement for Unit 1 operation is one Shift Supervisor, licensed as a Senior Reactor Operator; one Station Operating Foreman, nonaally licensed as a Senior Reactor Operator; two Stetion Operators, licensed as Reactor Operators; and two Attendants. The Maintenance Supervisor, who reports to the Chief Engineer, is responsible for the maintenance of all station equipment including instrumentation. .The Results Coordinator has a Reactor Control Chemist reporting to him
' ho is responsible for plant chemistry, health physics, and radiation control, and a Tech-naal Supervisor responsible for plant reactor engineering and plant serfonaance, n.s applicants have conducted a training program for most operat%g personnel which consisted of at least eight phases, including for eSst personnel a Shippingport Training Course Phase I, Shippingport Training Course Phase II, Reactor Training at Rensselaer Polytechnic Institute, PWR Simulator Training Design Lectures, Radiation Protection Training, On-site Operator Training, and Preliminary Operations and Testing. Selec-ted members of the plant staff technical support groups completed fonnal training speci-fically oriented to their assigned responsibilities.
The qualifications of key supervisory personnel with regard to educational background, experience, training and technical specialties have been reviewed and conform to those defined in AEC Regulatory Guide 1.8.
Technical support for the plant staff is available from the Technical Services Sec-tion, Maintenance Section. Operating Section, Personnel and Records Section (directs overall training of station personnel), and Nuclear Services Section of the Power Sta-tions Department at the Duquesne Light Company. Assistance can also be obtained from the various Engineering and Construction Division Departments.
We have concluded that the organizational structure, the training and qualifi-cations of the staff for the Beaver Valley Unit I are adequate to provide an acceptable operating staff and technical support for the safe operation of the facility. Addf-tional technical support during the startup test program will be provided primarily by the Duquesne Light Company Engineering and Construction, and Operating Divisions.
13.2 Safety Review and Audit The safety review and audit for Beaver Valley Unit I will be conducted by the Onsite 1 Safety Comittee and the Offsite Review Comittee. The Onsite Safety Comittee is advi-sory to the Station Superintendent and will review ali proposed test procedures, changes in plant operating procedures and design modifications. The Offsite Review Comittee l
13-2 provides corporate management with a review ar' audit capability to verify that organiza-tional cIIecks and balances are functioning to assure continued safe operation and design adequacy of the plant. The Offsite Review Committee will function in accord with Regula-tory Guide 1.33 (ANSI N18.7. " Standard for Administrative Controls for Nuclear Power Plants". Sections 4.1 through 4.4). Detailed features of the review and audit program will be incorporated in the Administrative Controls Section of the Technical Speci fications.
We conclude that the provisions for the review and audit of plant operations are acceptable.
13.3 Plant Procedures And Records Plant operations are to be performed in accordance with written and approved opera- i ting and emergency procedures. Areas covered include nonnal startup operation and shut-down, abnormal conditions and emergencies, refueling, safety related maintenance, sur-veillance and testing, and radiation control. All procedures, and changes thereto will be reviewed and approved by the Onsite Review Comittee prior to implementation.' Plant records to document appropriate station operations and activities will be maintained by the applicants. Plant procedures and record keeping have been reviewed against AEC Regu- !
latory Guide 1.33.
We conclude that the provisions for preparation, review. approval, and use of writ-ten procedures and reco*d keeping are satisfactory. Detailed features regarditq plant procedures and the retc. d management program will be incorporated in the Ackninist: ative Controls Section of the technical specifications.
13.4 Emergency Planning An emergency plan for the Beaver Valley Power Station has been prepared which des-cribes the applicants' organization, including responsibil! ties. and delegation of j authority in emergency situations. Agreements. liaison and consnunications have been made with appropriate agencies that have responsibilities for coping with emergencies.
Written agreements have been concluded wi th the Coninonwealth of Pennsylvania. Department of Environmental Resources; Pennsylvania Power Company. Department of the Army. District Engineer; City of East Liverpool. Board of Public Utilities; the Municipal Authority of the Borough of Midland; Penn Central Transportation Company; the Pittsburgh and Lake Erie Railroad Company; Mayfair Ambulance Company; Aliquippa Hospital; and the United !
States Atomic Energy Consnission. The plan includes protective action guides, based on the Pennsylvania State Department of Environmental Resources Protective Action Guides, that describe protective measures to be implemented in the event of an emergency that affects persons offsite. Arrangements have been made to provide for medical support in the event of a radiological incident or other emergencies. Provisions for the training of both plant personnel and offsite emergency organizations have been included o in the plan.
The applicants have stated that they are considering the means to provide additional protective measures for control room personnel at the Shippingport Atomic Power Station (SAPS) to assure that adequate protection is provided SAPS control room persnnnel in l the event of an accident at the Beaver Valley %er Station. We informed the applicants that we will require assurance that the essential personnel required to maintain SAPS in a safe condition be adequately protected from radiation in the event of an accident i at the Beaver Valley Power Station so that performance of safety functions relative to SAPS can be assured.
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3- I We have reviewed the emergency plan and find that it conforms with the requirements
. of Appendix E, to 10 CFR Part 50 and therefore is acceptable. contingent upon assurance '
-that essential personnel at Shippingport Atomic Power Station are adequately protected.
-in the event of an accident at the Beaver Valley Power Station.
13.5 -Industrial Security
.The applicants have submitted a description of the' industrial security program for'-
the protection of the Beaver Valley Power Station from industrial sabotage. The infor- ,
mation was submitted as proprietary information pursuant to Section 2.790 of 10 CFR I Part 2.
We have reviewed the program and have concluded that'it complies with the require -
ments of Section 50.34(c) of '10 CFR Part 50 and Section 73.40 of 10 CFR Part 70, and . .;
conforms to the Regulatory Position stated in AEC Regulatory Guide.1.17. except for the surveillance provisions described in Section-3.3.3 of ANSI N18.17-1973. We consider <
ti the security program acceptable, subject to the condition that.the program acceptable. ;
subject to the condition that the program be modified to provide acceptable surveillance-provisions to meet the requirements of-Section 3.3.3 of ANSI N18.17-1973.
.The applicants have informed us ' that they intend to' comply with our t::quirement' and. s will revise their security program accordingly. We will report the results of our I evaluation in a supplement to the Safety Evaluation Report. I j
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14.0 ' INITIAL' TESTS AND OPERATIONS, Duquesne Light Company has overall responsibility for the supervision and performance:
of the test and startup program. Coordination and control. of the organizations involved
-in this program will be perfomed by a Joint Test Group. The Joint Test Group will be -
made up of a representative of Duquesne Light Company, Stone and Webster, and Westinghouse Electric Company The Chairman of the Joint Test Group is the Beaver Valicy Station .
Superintendent. All plant systems operations required by the preoperational and startup
' test program are perfomed by the regular plant staff The preoperational and startup test procedures are initially draf ted by the Beaver Valley Power Station Test Section engineers and reviewed by the Operations Group and approved by the Onsite Safety Committee (where appropriate) ar4 the Joint Test Group'.
Test results will be evaluated and . approved by Duquesne Light Company and the Joint Test Group.
We have reviewed the preoperational startup and test program' described by the appli-cant and conclude that it is in accord with the AEC publication, " Guide for the Planning of Preoperational Testing Programs", and the " Guide for the Planning of initial Startup Programs", and is acceptable, e
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15.0' ACCfDENT' ANALYSIS' 15.1 General --
- The applicants have analyzed reactor performance for normal steady-state plant opera-tion-and for anticipated operational. transients on the basis of the initial core power 1evel of 2.660 eegawatts thermal; however, for evaluating the radiological: consequences, the accident analyses were perfomed for a higher nower level of 2766 megawatts thema1.
the ultimate power achievable. ~!'
We and the applicants have evaluated the offsite radiological consequences for pos-y tulated design basis accidents. These postulated accidents are the same as those :
~ analyzed for previously ' licensed PWR plants and include a steam line break accident, a steam ' generator tube rupture accident, a loss-of-coolant accident, a fuel handling accident. and a rupture of a radioactive gas storage tank in the gaseous radioactive .
waste treatment system.
. The applicants have evaluated the loss-of-coola'nt accident, the fuel handling acci -
dent and sne radioactive gas decay tank rupture using asstanptions that are substantially
.the same as' those used by the Regulatory staff. The offsite doses we calculated for these accidents are presented in Table 15.1 of this report, and the assumptions we used are listed in Section 15.2.1 of.this report. All potential doses calculated by the ap.
- plicants and by us for the postulated accidents are within the 10 CFR Part 100 guideline values.
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On the basis of our experience with the evaluation of the steam line' break and the steam generator tube rupture accidents for PWR plants of similar design, we have concluf deo that the consequences of these accidents can be controlled by limiting the pemissi-ble primary and secondary coolant system radioactivity concentrations so that potential offsite doses are small. We will include appropriate limits in the technical specifica-tions on primary and secondary coolant activity concentrations. . Similarly, we will.
include appropriate limits in the technical specifications on gas decay tank activity.
- 15.2 Design Basis Accident Assumptions 15.2.1 Loss-of-Coolant Accident The assumptions used by the Regulatnry staff in calculations of offsite doses from a LOCA were:
(1) Power level of.2766 MWt.
(2) Regulatory Guide No.1.4. " Assumptions used for Evaluating the Potential Radiolo-gical. Consequences of a Less-of-Coolant Accident for Pressurized Water Reactors."
June 1973.
TABLE 15.1 Potential Offsite Doses Due to Design Basis Accidents 0 Two Hour Course of Accidents Exclusion Boundary Low Population Zone L (610 meters) (5.794 meters)
Accident Thyroid Whole Body Thyroid Whole Body ,
Loss of coolant 210 10 21 >l fuel Handling 24 7 4 1. 1 Gas Decay Tank negligible 2 negligible >l Rupture t
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8 15-2 (3) Design containment leak rate of 0.1% for the first 60 minutes.
(4) Idine removal by the containment quench spray system was based on an iodine remo-val effectiveness equal to a first order removal constant of 10 per hour, in an effective volume of approximately 820,000 ft . 3The quench spray system was assumed to become effective one minute af ter the accident. Iodine removal by the recircu-i lation spray system, which does not become effective until five minutes af ter the accident, was based on an effective removal constant of 10 per hour, in an effective volume of 980,000 ft3 of the total containment volume of 1,800,000 3ft . A total of 150,000 ft3 of the containment volume was modeled as an unsprayed region. No credit was given for elemental iodine removal by the containment spray af ter the initial elemental iodine concentration has beer, reduce by a factor of 30.
(5) No credit is given for leakage which may occur into the penetration room and which exhausts through the supplementary leak collection and release system.
(6) Ground level release with Pasquill type "F" condition with wind speed of 0.5 meters per second for short term releases based on the meteorological data discussed in Section 2.3 of this report.
The rem dose does not include any credit for leakage which may occur into the pene-tration room and other areas contiguous to the containment which will be maintained un-der a negative pressure by the supplementary leak collection and release system, and which will be exhausted to the atmosphere through the HEPA and charcoal filters of the supplementary leak collection and release system (see Section 6.2.3). The dose rate would be significantly less than the calculated dose because it is reasonable to expect that a significant amount of the total leakage will be into the penetration room, and would therefore be collected and processed through the filters of the supplementary leak collection and release system.
15.2.2 Fuel Handling Accidents The assumptions used to calculate offsite doses from a fuel handling accident were:
(1) Rupture of all fuel rods in one assembly.
(2) All gap activity in the rods, assumed to be 10% of the noble gases and 10% of the iodine (with a peaking factor of 1.65), is released.
(3) The accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.
(4) 99% of the iodine is retained in the pool water.
(5) Standard ground release meteorology and dose conversion factors.
(6) Iodine removal factor of 95% for 2 charcoal filters in series.
15.2.3 Gas Decay Tank. Rupture The assumptions used to calculate the offsite doses from a gas decay tank rupture were:
(1) Gas decay tank contains one complete prir.ary coolant loop inventory of noble gases
- resulting from operation with 1% failed fuel (12,000 curies of noble gases).
(2) Gas decay tanks are located in Unit I area with exclusion area radius of 610 meters.
(3) Standards ground level release meteorology and dose conversion factors.
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.; .s y 16-1 16.0 TECHNIC 4. SPECIFICATIONS The technical specifications in a license define certain features, characteristics, and conditions governing operation of a facility that cannot be changed without prior approval of the AEC. The finally approved technical specifications will be made part of the operating license. Included are sections covering safety limits and limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls.
We intend to use the " Standard Technical Specifications for Westinghouse Pressurized Water Reactors" (dated September 1,1974) for Beaver Valley Unit 1, and are currently working with the applicants to develop the final technical specifications for Unit 1. 1 On the basis of our review to date, we conclude that nomal plant operation within f the limits of the technical specifications will not result in potential offsite exposures in excess of the 10 CFR Part 20 limits. Furthermore, the limiting conditions for operation and surveillance requirements will assure that necessary engineered safety features will be available in the event of malfunctions within the plant. ,
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-' j 17.0 - QUnLITY ASSURANCE !
17.1 QualityJssurance Program for Operations 17.1.1 -. General -
-i The description of the Quality Assurance (QA) Program for Operation for Beaver Valley : .
Unit 1 is contained in Appendix A of the FSAR, as amended. . .Our evaluation of the des- I cription.of the QA Program is based on a review of this information and related' discussions with the ap.plicants to determine the ability of Duquense Light Company (DLC) to comply with the requirements ~of Appendix B to 10 CFR Part 50 and AEC Regulatory Guide 1.33.
Our review of the DLC QA Program has included:
R (1) ~ A detailed review of the FSAR, as amended.
- (2). Meetings with DLC Management in which the proposed program was explained in depth.'
) (3) 01scussions with Regulatory Operations Region I relating.to RO's acceptance of the QA Manual and their observations of DLC's. ability to comply with the'18 criteria of Appendix B t' 13 CFR Part 50.
17.1.2 Organization DLC's organization (See Figure 17.1) comprises an executive management departments consisting of the President, the Vice President of Design and Construction and the Vice President of Operations. The Quality Assurance Manager, responsible for establishing
-and implementing the QA Program for Operation and Design and Construction, report di-rectly to the Nuclear Servics Superintendent who is on an equal reporting level as the Beaver Valley Power Station Super'i ntendent, both reporting to the Power Stations Department' Manager. The Power Stations Department Manager reports directly to the Vice Presider.t of Operations. The QC Supervisor has the additional authority of reporting significant quality matters to the QA Manager.
Since the QA Manager reports to the Vice President of Engineering and Construction and has the authority to report quality problems to the Vice President of Operations and/or the President, we find that the QA Manager has sufficient independence and authority to effectively carry out the QA Program during plant operation, maintenance, modification / repair, and refueling without undue influences from the Station Operations organization which is responsible for costs and schedules. Sufficient independence also exists between the QA Supervisor and the Station Superintendent such that the execution of the QC Program and the attention to quality problems and there resolutions will not become subordinated. We conclude that the QA organization as identified and described in the FSAR is acceptable and complies with Criterion I of Appendix B to 10 CFR Part 50.
17.1.3 Quality Assurance Program DLC was required to significantly upgrade their QA Program from that originally pre-sented in the FSAR for Unit 1, and our final review and evaluation is based on this revised and updated QA Program as amended in the FSAR.
DLC's QA Program description contained in the FSAR provides for controlled written policies, procedures, and instructions governing the implementation and control of quali-ty related activities associated with the operation of Unit I which includes maintenance, modification / repair, and refueling. The QA Program requires that indoctrination and training programs be established and conducted for those personnel performing quality related activities to assure they are knowledgeable of the QA Program procedures and requirements and become proficient in implementing these procedures.
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17-2 The QLProgram provides that the inspection and verification operations for the maintenance, modification, and repair of safety related structures, systems, ard l components including refueling operations be properly inserted in controlled documents. {
The QA Program requires that the quality verification and inspection of quality q related activities be performed by individuals or groups independent of those individuals {
or groups directly responsible for performing the work being verified or inspected.
DLC's QA Program also provides for the collection and retention of records that define and attest to the quality of safety related structures, systems, and components through-out operations, maintenance, modification / repair, and refueling.
DLC requires comprehensive scheduled audits to be performed by qualified QA person-nel independent of those individuals or groups in the area being audited. The audits are required to be in accordance with preestablished written procedures and include the verification and evaluation of procedures and quality related activities to assure they are meaningful and effective. The QA Program requires audit results and corrective actions to be documented and reported to responsible management. The DLC QA Department audits are supplemented by audits conducted by the Offsite Review Committee and the QC Supervisor. The Offsite Review Committee performs audits within the Operations Division to assess thL technical adequacy of procedures as well as their implementation.
The QC Supervisor is responsible for auditing the activities to evaluate and deter-mine the effectiveness of the implementation of the QC Program. In addition, the DLC QA Program will be audited annually by the Vice President of Design and Construction and the Vice President of Operations to assess the . status and adequacy of the Program.
As a result of a detailed review and evaluation of the DLC QA Program description in the FSAR, as amended, we conclude that the Program provides sufficient procedural requirements and controls necessary to demonstrate compliance with each of the criteria of Appendix B to 10 CFR Part 50.
17.1.4 Conclusions Based on our review of the QA Program we conclude that:
(1) DLC's QA Program complies with the 18 criteria of Appendix B to 10 CFR Part 50 and Regulatory Guide 1.33 an:: is therefore acceptable for the operation of the Beaver Valley plant.
(2) The documented policies and procedures of DLC's QA Program provide an acceptable basis for the implementation of a QA Pt ogram that meets the requirements of Cri-terion 11 in Appendix 0 to 10 CFR Part 50.
(3) A QA staff has been provided with adequate authority, responsibility, and indepen-dence to implement the DLC QA Program.
(4) DLC's organization structure and delineation of responsibilities to qualified QA personnel to effectively carry out the QA/QC functions meets the requirements of o
Criterion I in Appendix B to 10 CFR Part 50.
(5) The audit system of DLC conforms with the requirements of Criterion MIII in Appen-dix B to 10 CFR Part 50.
Accordingly, we find that DLC has described a QA Program for Operation that complies with Appendix B to 10 CFR Part 50 and AEC Regulatory Guide 1.33 and is therefore accept-able for control of the operation, maintenance, modification / repair, and refueling acti-vities of Beaver Valley Unit 1.
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Implementation of the Quality Assurance Program will. be ver.ified prior to licensing, by tha Commission's Directorate of Regulatory Operations who will also monitor the -
' adequacy of the QA Program for the term of the operating license. If deviations or' deficiencies in the QA Program are found, they will be identified and the necessary measures will be taken to correct them and to prevent their recurrence.
17.2 Quality Assurance Program for Construction The Quality Assurance Program for Beaver Valley Unit 1 originated before 'the'18 criteria of Appendix 'B to 10 CFR Part 50 were' published 'as proposed criteria'on April 17, 1969. During the construction permit stage of our review of Unit I we evaluated the'.
program and, as a result, the applicants stated that written procedures would be pre--
pared to control quality related activities in accordance with the Consnission's criteria.
The 18 criteria of Appendix D to 10.CFR Part 50 were published on June 27,1970 with -
an effective date of 90 days after ' publication. The construction permit for Beaver Valley Unit I was issued on June 26, 1970. The applicants have made efforts to upgrade
'the Quality Assurance Program.
- With regard to the implementation of the QA Program for construction, the Directorate of Regulatory Operations has found numerous violations. Our review of the enforcement
' history for Beaver Valley Unit I has revebled three major, areas of concern: (1) the large number of violations. (2) the repetition of certain items of violations, and-(3)-
the length of time required to correct violations. The Directorate of Regulatory Operations has issued a Notice of Violation by letter dated May 24, 1974 to Duquesne.
Light Company, and has discussed these concerns with Duquesne Light Company at a meeting '
on June 20, 1974 As a result of these violations, the ' Directorate of Regulatory Operations' has increased its inspection and surveillance of the Quality Assurance Program
- implementation.
These violations, and other outstanding items of deficiencies, must be resolved to the satisfaction of the Regulatory staff, such that the staff can meM +he. requisite findings under 10 CFR Part 50.57 that the health and safety of the pubi. 411 be protected, prior to a decision concerning the issuance of an operating license.
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18.0- REVIEW BY THE ADVISORY CO M ITTEE ON REACTOR SAFEGUARDS (ACRS)'
L -18.1 Construction Permit Review The ACRS completed the construction pemit review of Beaver Valley Unit 1 at its 119th meeting on March 5-7. 1970. . In its report dated March 12, 1970 the ACRS discussed several matters which were to be resolved during the construction of.. Unit 1. These matters are sumarized below and discussed in the appropriate sections of this report.
The applicants have resolved the problem associated with the relatively poor atmospheric dispersion conditions at the site by acquiring additional property to increase the exclusion area radius from 1,200 feet to 2.000 feet (see Section 2.1 and
.2.3).
'Several matters related to seismic design were resolved to the satisfaction of the staff. These matters included the proposed response spectra (see Section 3.7), analyti-cal methods for Class 1 piping (see Section 3.9), and the soil density.under the river water system piping (see Section 2.5).
~ The applicants have described a method of annealing the pressure vessel if this becomes necessary to restore the properties of the vessel material because of neutron irradiationdamage(seeSection5.2).
A suitable p. operational vibration testing program will be employed for reactor coolant piping and reactor internals (see Sections 3.9 and 4.2). Also, the applicants will install instrumentation for in-service monitoring of loose parts in the reactor coolant system (see Section 5.7).
The testing programs for the emergency diesel generators have been found acceptable bythestaff(seeSection14.0).
The matters related to core themal and hydraulic design have been satisfactorily resolved. The applicants will use the 17 x 17 fuel assembly design (see Section 4.4).
The matters related to anticipated transients without scram are to be resolved following our review of the applicants response to the request for information required by the staff report. WASH-1270, " Technical Report on Anticipated Transients Without Scram (ATWS) for Water-Cooled Power Reactors," (see Section 7.10).
. The control of hydrogen buildup in containment following a postulated loss-of-coolant accident has been resolved by the use of redundant hydrogen recombiners (see Section 6.2.5).
With regard to the detection of failed fuel, the applicants have provided radiation detectors in the reactor coolant let down line (see Section 11.6).
The applicants have indicated their intention to provide safety-related display instrumentation to monitor the course of events in the unlikely event of a loss-of-coolant accident (see Section 7.5).
18.2 Operating License Review 0 The report of the ACRS on the review of the operating license application for Unit 1 will be placed in the Comission's Public Document Room and will be published by the Regulatory staff in a supplement to this Safety Evaluation Report. The supplement will be published prior to the final determination regarding issuance of an operating license.
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'j k '.19.0 ' COMIN DEFENSE AND SECURITY .
The Application reflects that the' activities to be conducted will be within the jurisdiction of the United States and that all of the directors and principal officersi
' of the applicants are United States citizens. . The applicants are 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 applicants have agreet to safeguard any such data which might become involved in accordance with the. requirements of 10 CFR Part 50. The applicants will rely upon obtaining fuel as it is needed from sources of supply available for civilian purposes, so that no diversion of special nuclear -
material' for military purposes is involved. For these reasons and in the absence of any infonnation to the contrary, we have found that the activities to be perfomed will, ,
not be inimical to the connon defense and security.
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20.0 ' FINANCIAL QUALIFICATIONS '
The Connission's regulations which relate to financial data and information required .
to establish financial qualifications for applicants for operating licenses are Paragraph- i 50.33 (f) of 10 CFR Part 50 and Appendix C to 10 CFR Part 50. We have reviewed the financial information presented in the application and Amendnent No.1 thereto. Based on this review, we have concluded that Duquense Light Company, Ohio Edison Company, and Pennsylvania Power Company possess or there is reasonable assurance they can obtain the necessary funds to meet the requirements of 10 CFR 50.33 (f) to operate the Beaver Valley
~
Unit 1 and if necessary permanently shut down the facility and maintain it in a safe shutdown condition.
Beaver Valley Unit I will be used'to meet projected load growth'in the area served by
. the applicants (part of western Pennsylvania and eastern and northern Ohio). Operating revenues will provide the funds to cover. cost of operations. Fuel, other operation and maintenance expenses, and depreciation expense during the first five years of commercial operation of the facility (1976 - 1980) are presently estimated by the applicants to be (in millions of dollars) $30.1; $29.1; $28.2; $28.7 and $28.9 in that order. Applicants will share facility operation and maintenance expenses in proportion to their undivided ownership interest, as.noted below, with adjustments in the event full entitlement to .
- the capability and output of the facility is not taken by any tenant in comon.
Duquesne Light Company 47.55 Ohio' Edison Company 35.0 Pennsylvania Power Company 17.5 100.0%
Also, the agglicants will share in proportion to their undivided ownership interests, as stated above, the cost of permanently shutting down the vacility and maintaining it in a safe shutdown condWon. Assuming shutdown measurus comparable to those authorized by the Connission for the retirement of the Hallam Nuclear Power Facility, the applicants estimate that the cost of retirement of the subject facility, based on 1972 dollars and technology, would not exceed $6.5 million. An additional $60 thousand annually would be set aside to cover the cost of round-the-clock surveillance of the facility and periodic maintenance to fences and barriers. Sources of funds to cover these costs are estimated to be in the same relative ratio as funds obtained for construction purposes. For years subsequent to 1979, internally generated funds (retained earnings plus depreciation, deferred taxes, and other accruals) are expected to provide 18% to 30% of total funds for construction purposes and therefore the same portion of the' total cost of permanent shutdown of the facility and its maintenance in a safe condition.
Applicants state that uranium will be aquired by lease from private sources for the first core and first two reload regions. The necessary fuel enrichment will be obtained by toll enriching arrangements with the Comission.
o We have examined the financial infonnation submitted by the applicants to detemine whether they are financially qualified to meet the above estimated costs. The infor-mation presented in Duquesne Light Company's annual report for 1972 indicates that operating revenues totaled $220.8 million. Operating expenses were stated at $161.7 million, of which $20.9 million represented depreciation. Irterest on long-term debt was earned 2.6 times. Net income totaled $46,0 million, of which $34.6 million was distributed as dividends to stockholders with the remaining $11.4 million retained for l
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20-2 use in the business. As of December 31, 1972, the Company's assets totaled $1,084.2 millioiE most of which was invested in utility plant ($1,008.6 million). Retained earnings amounted to $119.7 million.
Financial ratios computed from the 1972 statements indicate an adequate financial condition, e.g., long-term debt to total capitalization-55%, and to net utility plant -
54%; net plant to capitalization - 1.02; the operating ratio - 73%; and the rates of return on :onrnon equity - 11.3% on stockholders' investment - 10.5%, and on total investment - 7.0%. The record of the Company's operations during 1970-72 shows that operating revenues increased from $167.0 million in 1970 to $220.8 million in 1972; net income increased from $34.4 million to $46.0 million; and net investment'in utility plant from $744.7 million to $1,008.6 million. However, the number of times interest earned declined from 2.9 to 2.6.
Moody's Investors Service rates the Company's first mortgage bonds as As (high grade bonds) and its debentures as A (upper middle grade obligations). The Company's current Dun and Bradstreet rating is sal, the highest rating.
Infomation presented in Ohio Edison Company's annual report for 1972 indicates that operating rever.ues totaled C343.2 million. Operating expenses were stated at $264.6 million, of whien D/.7 million represented depreciation. Interest on long-tem debt was earned 2.6 times. Net income totaled $53.9 million, of which $44.3 million was distributed as dividends to stockholders with the remaining $9.6 million retained for use in the business. As of December 31, 1972, the Company's assets totaled $1,272,5 million. (
most of which was invested in utility plant ($1.173.2 million). Retained earnings amounted to $127.2 mill bn. Financial ratios computed form the 1972 statements indicate an adequate financial condition, e.g. long-term debt to total capitalization - 57%, and to net utility plant - 54%; net plant to capitalization - 1.06; the operating ratic - 77%;
and the rates of return on comon equity - 13.6% on stockholders' investment - 11.2%
and on total investment - 7.0%. The record of the Company's operations during 1970-72 shows that operating revenues increased from $281.8 million in 1970 to $343.2 million on 1972; net income increased from $50.1 million to $53.9 million; and net investment inutility plant from $928.8 million to $1.173.2 million. However, the number of times interest etrned declined from 3.5 to 2.6. Moody's Investors Service rates the Company's first mortage bonds as Ass (nighest quality). T.he Company's current Dun and Brodstreet rating is SA1, the highest rating.
Information included in Pennsylvania Power Company's annual report for 1972 indicates l that operating revenues totaled $45.8 million. Operating expenses were stated at $36.7 million, of which $5.2 million represented depreciation. Interest on long-term debt was earned 2.0 times. Net income totaled $6.1 million. Dividends to stockholders totaled
$7.5 million, of which $6.4 million represented comnon dividends paid to Ohio Edison 0 Company, the Company's parent at d owner of all of its outstanding connon stock. As of December 31, 1972. the Company's assets totaled $188.6 million, most of which was invested in utility plant ($1/5.1 million). Retained earnings amounted to $9.8 million.
Financial ratios computed from the 1972 statements indicate an adequate financial con-d; tion, e.g., long-term debt to total capitalization - 58%, and to net utility plant -
58%; net, plant to capitalization - 1.00; the operating ratio - 80%; and the rates of return on conpon equity - 9.3%, on stockholders' investment - 8.3%, and on total in-l l vestment - 6.5%. The record of the Company's operations during 1970-72 shows that l
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20 operating revenues im reased from $38.0 million in 1970 to $45.8 million in 1972; net income increased from $5.6 million to $6.1. million; and net investment in utility plant from $121.8 million to $175.1 million. However, the number of times interest earned declined from 2.6 to 2.0. Moody's Ir.vestors Service rates the Company's first mortgage bonds es Aa(high grade bonds). The Company's current Dun and Bradstreet rating is 5Al, the highest rating.
Summary analyses reflecting these ratios and other pertinent data for each a3plicaiit are attached as Appendix B to this report.
We have requested the applicants to submit additional financial data to update the infomation in the application. We will report the results of our evaluation of these data in a supplement to the Safety Evaluation Report.
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. 'e 21-1 21.0 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS 21.1 ' General ._
Pursuant to the financial protection and indemnification provisions nf the Atomic Energy Act of 1954. as amended (Section 170 and related sections), the Commission has issued regulations in 10 CFR Part 140. These re;ulations set farth the Comission's requirements with regard to proof of financial protection by. a d indemnification of.
licenses for facilities such as power reactors under 10 CFR Part 50.
21.2 Preoperational Storane of Nuclear Fuel The Comission's regulations in Part 140 require that each holder of a construction permit under 10 CFR Part 50. who is also the holder of a license under 10 CFR Part 70 authorizing the ownership and possession for storage only of special nuclear material at the reactor 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 agreemer.t with the Comission. Proof of financial '
protection is to be furnished prior to. and the indemnity agreennt executed as of, the effective date of the 10 CFR Part 70 license, payment of an annual indemnity fee is required.
The applicants have furnished to the Comission proof of financial protection in the amount of $1.000.000 in the forn of a Nuclear Energy Liability lasurance Association Policy (Nuclear Energy Liability Policy, facility form No. NF-226). Further. the applicants have executed Idemnity Agreement No. B-73 with the Comission 83 of August
- 12. 1974 the effective date of its preoperational fuel storagt licer.se. SNM-1472.
The applicants have paid the annual idemnity fee applicsble to preoperational fuel storage.
21 .3 Operating License Under the Commission's regulations,10 CFR Part 140. a license authorizing the operation of a reactor may not be issued until proof of financial protection in the amount required for such operation has been furnished, and an indemnity agreement covering such operation (as distinguished from preoperational tuel storage only) has been executed. The amount of fin'ancial protection which must be maintained for Beaver Valley Power Station Unit 1 (which has a rated capacity in excess of 100.000 electrical kilowatts) is the maximum amount available from private sources. i.e.,
the combined capacity of the two nuclear liability insurance pools, which amount is currently $110 million.
Accordingly, no license authorizing operation of Beaver Valley Power Station. Unit I will be issued until proof of financial protection in the requisite amount nas been received and tbt requisite indemnity agreement executed.
We expect that, in accordance with the usual proced tre. the nuclear liability insurance 8 pools will provide, several days in advance of anticipated issuance of the operating license document, evidence in writing, on behalf of the applicant, that the present coverage has been appropriately amended so that the policy limits have been increased, to meet the requirements of the Commission's regulations for reactor operation. Similarly, no operating license will be issued until an appropriate amendment to the present indemnity agreement has been executed. The applicants will be required to pay an annual fee for operating license indemnity as provided in our regulations, at the rate of $30 !
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<: s 21-2 per each thousand 6110 watts of thermal capacity authorized in its operating license.
On the basis of the above considerations, we conclude that the presently applicable.
requirements of 10 CFR Part 14f have been satisfied and that, prior to issuance of the operating licanses, the applic ants will be required to comply with the provisions of 10 CFR Part 140 applicable to operating licenses, including those as to proof of financial protection in the requisite amount and as to execution of an appropriate indemnity agreement with the Commission.
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22.0 CONCLUSION
S Based on our evaluation of the application as set forth above (and assuming favorable resolution of outstanding matters discussed in Sections 2.2, 3.4, 4.2. 4.4, 5.2, 6.2.5, 6.3. 3, 6.4. 7. 3.1, 7. 3.4, 7.5, 7.7. 7.10, 8.2, 8. 3.1, 9. 3.4, 13.4, and 13.5) we na ve concluded that:
(1) The application for a facility license filed by the Duquesne Light Company Toledo Edison Company, and Pennsylvania Power Company, dated January 13, 1969, and as sub-sequently amended, complies with the requirements of the Atomic Energy Act of 1954, as amended (Act), and the Comission's regulations set forth in 10 CFR Chapter 1.
(2) Construction of Beaver Valley Power Station Unit 1 (the facility) has proceeded and there is reasonable assurance that it will be substantially completed, in conformity with Provisional Construction Permit No. CPPR-75, the application as amended, the provisions of the Act, and the rules and regulations of the Comission.
p (3) The facility will operate in conformity with the application as amended, the pro-visions of the Act, and the rules and regulations of the Commission.
(4) There is reasonable assurance (a) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and (b) that such activities will be conducted in compliance with the regulations of the Commission set forth in 10 CFR Chapter 1.
(5) The applicants are technically and financially qualified to engage in the activities authorized by this license, in accordance with the regulations of the Comission set forth in 10 CFR Chapter 1.
(6) The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public.
Before an operating license will be issued to Duquesne Light Company et al., for operation of the Beaver Valley Power Station Unit 1, the unit must be completed in conformity with the provisional construction pemit, the application, the Act, and the rules and regulations of the Commission. Such completeness of construction as is required for safe operation at the authorized power level must be verified by the Commission's Directorate of Regulatory Operations prior to license issuance.
Further, before an operating license is issued, the applicants will be required to satisfy the applicable provisions of 10 CFR Part 140.
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A-1
_. APPENDIX A CHRONOLOGY OF REGULATORY RADIOLOGICAL REVIEW OF BEAVER VALLEY POWER STATION, UNIT 1 March 30,1972 FSAR submitted for preliminary review April 10,1972 Letter to applicants regarding the draft industrie.
security criteria April 20,1972 Meeting with applicants to discuss results of pr6liminary review.
April 27,1972 Letter to applicants concerning establishment c' public document room July ll, 1972 Letter fm3 applicants advising that revised FSAR to be filed Septemoer 1,1972 September 17, 1972 FSAR sub.nitted for second pmliminary review.
September 18, 1972 Submittal of Application for operating license September 2G, 1972 Letter to applicants requesting analysis of results of failure of non-Category I (seismic) equipment.
October 18, 1972 Letter to applicants stating that application acceptable for docketing. Application docketed.
October 25, 1972 Letter from applicants in response to letter of September 26, 1972.
October 31, 1972 Meeting with applicants to discuss results of second preliminary review.
November 2, 1972 Issuance of Amendment No. I to Construction Permit No.
CPPR-75.
November 20, 1972 Letter to applicants requesting an analysis of the con-sequences of fuel der.sification.
November 22, 1972 Letter from applicants providing additional infomation c9ncerning request of September 26, 1972.
December 18, 1972 Letter to applicants requesting an analysis of the con-sequence, of pipe failures outside containment.
December 20, 1972 Letter from applicants transmitting Industrial Security Plan as proprietary information.
Decenter 29, 1972 Letter to applicants transmitting review schedule.
o January 2, 1973 Letter from applicants in response to request of November 20, 1972.
January 31, 1973 Letter to applicants transmitting errata sheet related to request of December 18, 1972.
February 8,1973 Meeting with applicants to discuss first-round questions, review schedule, and status of exemption request.
February 16, 1973 Letter to applicants requesting additional information.
March 16, 1973 Letter to applicants requesting additional information.
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' March 26 1973 Letter from applirants concerning request of March 16 1973.-
March 30,'1973 Letter to applicants requesting additiona1'information.
. April 16, 1973 Submittal of Amendment No.1, consisting of response to request of February 16, 1973, and other replacement-pages.
April 17,'1973 Letter to applicants regarding letter of March 16, 1973.
April 25,'1973 Meeting with applicants to discuss review schedule and .
questions concerning electrical, instrumentation and con- -
trol systems.
tiay 1,1973 Letter from applicants advising of Change in fuel design .
and change of fuel loading date to January 1975.
May 7, 1973 Submittal of Anendment No. 2, consisting of responses to requests of March 17. and March 30. 1973, pnd replacement -
pages.
May 7, 1973 Letter to applicants requesting additional information.
May 8, 1973. Meeting with applicants to discuss impact of changes in fuel design and review schedule, and discuss request for electrical drawings.
. May 15, 1973 Letter from applicants transmitting proprietary and nonproprietary electrical drawings.
May 23,1973 Letter from applicants transmitting schedule forrespond-ing to request of May 7,1973.
May 30-31, 1973 Site visit to review flood protection, operational safety and emergency planning.
June 7,1973 Meeting with applicants to discuss containment design.
June 12, 1973 Meeting with applicants to discuss quality assurance June 26, 1973 Submittal of Amendnent No. 3, consisting of additional response to request of May 7, 1973, and replacement pages.
June 27,1973 . Letter to applicants requesting updated financial infomation.
June 28, 1973 Letter from applicants requesting extension of ron-struction completion dates for Construction Permit No. CPPR-75.
July 19,1973 Letter to applicants requesting additional information and response to list of staff. positions.
July 27,1973 Letter to applicants requesting additional informat 1.
August 7,1973 Meeting with applicants to discuss iodine removal by
.O containment spray systems, I'
i August 16, 1973 Meeting with applicants t1 discuss security plan and emergency preparedness plan.
August 28, 1973 Meeting with applicants to discuss containment pressure .
response analysis and iodine removal by containment spray systams.
August 29, 1973 Issuance of Order extending Completion date to June 30 l 1975.
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'A-3 August 30.-1973- Submittal. of Arrendment No.1 to License Application. and replacement pages in response to request of June 27. 1973.
Septembe$4.1973 ' Submittal of Amendment No.' 4. in response to requests -
, of June 19 and June 27. 1973 and replacement pages.
Septanber 12. 1973 Letter to applicants stating that security plan will be withheld from public disclosure.
October 8. 1973 Submittal of Amendment No. 5.~ consisting of replacement pages.
October 9.1973 Letter to applicants regarding Regulatory staff report on anticipated transients without scram.
October 12, 1973 Letter from applicants transmitting Revision No. I to-Industrial Security Plan as Proprietary information.
October. 15. 1973 Meeting with applicant to discuss responses to review -
questions and certain staff positions.
November 9. 1973 Letter from applicants transmitting revised pages to Industrial security plan as proprietary information.
November 27, 1973 Letter from applicants transmitting Revision No. 3 of Industrial Security Plan as proprietary information.
November 28, 1973 Meeting with applicants to discuss design for protection against effects of postulated high-energy line ruptures outside contair. ment.
December 3. 1973' Submittal of Amendment No. 6. consisting of replacement pages.
December 12. 1973 Letter to applicants stating that revisions to Industrial submitted October 12, 1973 and November 9.1973 will -
be withheld from public disclosure, and returning copies -
of December 20. 1972 submittal.
December 12-14. 1973 - M eting with applicants to conduct review of safety-related schematic diagrams.
December 26. 1973 Letter from applicants transmitting an interim reply to request of October 9.1973.
January 3,1974 Letter to applicants stating that Industrial Security Plan Revision No. 3 submitted November 27.1973. will be withheld from public disclosure.
January 16. 1974 Letter from applicants concerning high energy line analysis outside containment.
l February 5.1974 Letter to applicants statin 0 that insufficient information l provided to .iustify withholding of proprietary report on electrical drawings sub.vitted May 15. 1973.
February 21. 1974 Letter to applicants requesting information regarding byproduct source, special nuclear matcrials.
February 21. 1974 Letter te applicants requesting information regarding quality assurance organization.
l February 25. 1974 Letter from ipplicants providing information in support of May 15.1773 letter.
I March 1,1974 Letter to applicants transmitting revised review schedule.
March 4. 1974 Letter to applicants requesting additional infomation and statement of Regulatory staff positions.
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A-4 1 March 12.1974 Letter from applicants in response to request of February 21, 1974 ';
March 25. 1974 Submittal of Amendment No. 7, consisting of response to request of March 4,1974, and replacement pages.
April 17, 1974 Submittal of Amendment No. 8, consisting of additional response of March 4,1974, and other replacement pages.
April 29,1974 Reconstitution of ASLAB - Farrar, ack, Johnson (Published 5/3/74. 39 FR 15522)
May 20, 1974 Letter to applicants requesting additional infonnation and statement of additional Regulatory staff positions.
June 19, 1974 Submittal of Amendment No. 9 consisting of information .
related to the fuel design, and other changes.
June 20, 1974 Letter from applicants concerning ECCS evaluation.
July 1-3, ?C4 Site visit to conduct review of reactor protection system, engineered safety features, and flood protection.
July 1. 1974 Submittal of Amendment No.10, consisting of response to request of May 20,197, and other replacement pages.
July 31, 1974 Letter to applicants stating that proprietary report submitted May 15, 1973, will be withheld from public 4 disclosure.
August 5, 1974 Submittal of Amendment No,11, consisting of -eplacement pages.
August 7. 1974 Letter from applicants requesting a meeting to discuss Regulatory staff positions.
August 12. 1974 Letter from applicants transmitting information on hydro-gen recombiners and emergency shutdown panel.
August 23, 1974 Letter from applicants regarding nonproprietary report submitted May 15, 1973.
August 28, 1974 Letter from applicants regarding realignment of ECCS from injection to recirculation mode and automatic startup of the auxiliary river water system.
August 30, 1974 Letter to applicants requesting additional information and statement of additional Regulatory staff positions.
September 10, 1974 Telegram to applicants concerning construction schedu'e, September 12, 1974 Letter from applicants in response to telegram of (
September 10, 1974.
September 24. 1974 Letter from applicants in response to our letter of October 9, 1974, containing information related to antici-pated transients without scram, 6eptember 30, 1974 Submittal of Amendment No.17, consisting of response to j request of August 30, 1974, and replacement pages.
i October 7, 1974 Meeting with applicants to discuss LCCS realignment to recirculation mode of operation, and auxilisry river water system.
October 9. 1974 Submittal of Amendment No. 2 to License Application, con-sistino of updated financial information.
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B-1 APPENDIX B OUQUESNE LIGW COMPANY (Consolidated)
FINANCIAL ANALYSIS DOCKET No. 50-334 (dollars in millions)
Calendar Year Ended December 31 1972 1971 1970 Long-term debt $ 542.6 $ 484.8 $ 415.6 Utility plant (net) 1.008.6 854.3 744.7 #
Ration -debt to fixed plant .54 .57 .56
' Utility plant (net) 1.008.6 854.3 744.7 Capitalization 981 6 843.2 732.2 Ratio of net plant to capitalization 1.02 1.01 1.02 Stockholders' equity 439.0 358.4' 316.6 Total assets 1.084.2 914.2 . 803.9.
Proprietary ratio .40 .39 .39 Earnings available to cocenon equity 39.0 34.7 29.3 Comon equity 345.5 282.4 240.6 Rate of earnings on common equity 11.3% 12.3% 12.2%
Net income 46.0 39.7 34.4 Stockholders' equity 439.0 358.4 316.6 Rate of earnings on stockholders' equity 10.5% 11.1% 10.9%
Net income before interest 76.0 64.8 55.5 Liabilities and capital 1,084.2 914.2 803.9 Rate of earnings on total investment 7.0% 7.1% 6.9%
Net income before interest 76.0 64.8 55.5 Interest on long-term debt 28.7 24.2 19.4 No. of times long-term interest earned 2.6 2.7 2.9 Net income 46.0 39.7 34.4 Total revenues 237.7 208.9 180.1 Net income ratio .19 .19 .19 Total utility operating expenses 161.7 144.1 124.6 Total utility operating revenues 220.8 196.7 167.0 Operating ratio .73 .73 .75 Utility plant (gross) 1.255.7 1.085.7 960.9 Utility operating revenues 220.8 196.7 167.0 Ratio of plant inv?stment to revenues 5.f9 5.52 5.75 1972 1971 Capitalization: Amount : of Total Amount % of Total Long-term debt $ 542.6 55.3% $ 484.8 57.5%
Preferred stock 93.5 9.5 76.0 9.0 Corenon stock & surplus 345.5 35.2 287.4 33.5 s
Total 3' 91F.T p_ 1747.7 T5T.TiY Moody's Bond Rating: Mortgage Aa. Debenturn A Durt & Bradstreet Credit Reting: SA) 1 1
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L l= OHIO EDISON COMPANY h (Consolidated)
FINANCIAL ANALYSIS DOCKET NO. 50-334 (dollarsinmillions)
Calendar Year Ended December 31 1972 1971 1970 Long-term debt $ 629.1 $ 557.1 $ 475.1 Utility plant (net 1,173.2 1.047.2 928.8 Ratio - debt to fixed plant .54 .C3 .51 Utility plant (net) 1,173.2 1,047.2 928.8 Capitalization 1,109.1 991.1 897.1 Ratio of net plant to capitalization 1.06 1.06 1.04 Stockholders' equity 480.0 434.0 421.9 1,272.5 1,146.7 1,024.1 Total assets .41 Proprietary ratio .38 .38 Earnings available to common equity 49.1 4C.8 47.3 Coninon equity 360.1 350.5 344.2 Rate of earnings on common equity 13.6% 13.1% 13.7%
Net income 53.9 48.5 50.1 480.0 434.0 421.9 Stockholders' equity Rate of earnings on stockholders' equity 11.2% 1 1 . 21: 11.9%
Net income before interest 83.7 '77.9 70.8 Liabilities and capital 1.272.5 1.146.7 1.024.1 Rate of earnings on total envestment 7.05 6.8% 6.9%
Net income before interest 89.7 77.9 70.8
. Interest on long-term debt 34.1 28.5 20.0 No. of times long-term interesi earned 2.6 2.7 3.5 Net income 53.9 48.5 50.1 Total revenues 354.3 318.6 287.9
.15 .15 .17 Net income ratio Total utility operating expenses 264.6 240.7 217.1 Total utility operating revenues 343.2 380.8 2'11 . 8 Cperating ratio .77 .78 .77 Utility plant (gross) 1.568.6 1.412.1 1,278.0 Utility operating revenues 343.2 308.8 281.8 Ratio of plant investment to revenues 4.57 4.57 4.54 1972 1971 Capitalization: Amount % of Total Amount % of Total Long-tenn debt $ 629.1 55.7% $ 557.1 56.2%
119.9 10.8 83.5 8.4 Preferred stock 350.5 35.4 Conmon stock & surplus 360.1 32.5 0 Total 51,109.1 100.0% TTWT 100.0%
Moody's Bond Rating: Aaa Dun & Bradstreet Credit Rating: 5A1 l
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8-3 PENP.5YLVANIA POWER COMPANY
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FINANCIAL ANALYS!$
DOCKET N0. 50-334 (dollars in millions) .
P Calendar Year Ended' December 31 1972 1971 1970 Long-term debt $ 101.6 :$ 89.6L $ 67.6 Utility plant (net) 175.1 152.3 .121.8
- Ratio . debt to fixed plant .58 .59 . .56 Utility plant (net)
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175.1 152.3 121.8 Capitalization '175.1 159.5 121.4~
Ratio of net plant to capitalization 1.00 .95 1.00 Stockholders' equity' 73.5 69.9 53.8 Total assets 188.6 .172.4 130.8
- Proprietary ratio .39 .41 41 Earnings available to connon equity ' 5.0 5.9 5.0 Conrion equity 53.6 50.0 39.7 Rate of earnings on connon equity ' 9.3% 11.8% 12.6%'
Net income 6.1 6.6 5. 6 '
Stockholders' equity . 73.5 69.9 53.8 Rate of earnings on stockholders' equity 8.3% 9.4% 10.4%
liet income before interest 12.2 11.3 9.0
!.fabilities and capital .188.6 172.4 130.8 Rate of earnings -on total investment ' 6.5% '6.6% 6.9%
Net income before interest 12.2 11.3 9.0 Interest on long-term debt ' 6.1 4.7 3.4 No. of. times long-tenn interest earned 2.0 2.4 2.6
. Net income ' 6.1 6.6 5.6 Total revenues 48.9 45.3 39.7 Net income ratio .12 .15 .14 Total utility operating expenses 36.7 34.0 30.7 Total utility operating revenues 45.8. 42.5 38.0 Operating ratio .80 . .80 .81 Utility plant (gross) 226.2 198.9 164.7
-Utility operating revennes 45.8 42.5 38,0 Operating ratio 4.94 4.68 4.33 1972' 1971 Capitalization: Amount % of Total Amount .% of Total Long-term dest $ 101. 6 .58.0% $ 89.6 56.2%'
Preferred stock 19.9 11.4 19." 12.5 Connon stock & surplus 53.6 30.6 50.i 31.3 Total $M TUlD Y TET 1FM Moody's Bond Rating: Aa Dun & firadstreet Credit Rating: SA1 l
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C-1 Appendix C BIBLIOGRAPHY (Documents referenced in or used to prepare the Safety Evaluation Report for Beaver Valley Power Station Unit 1)
General
- 1. Preliminary Safety Analysis Report with Amendments 1 through 11 for Beaver Valley Unit 1 (Docket No. 50-334).
- 2. United States Atomic Energy Commission Rules and Regulations,10 CFR:
Part 1 Statement of Organization and General Infomation Part 2, Rules of Practice Part 9 Prklic Record Pert 20, Standards for Protection Against Radiation Part 30. Licensing of Production and Utilization Facilities Part 100. Reactor Site Criteria
- 3. United States Atomic Energy Commission Regualtory Guides.
Meteorology
- 5. Alaka, M. A., " Climatology of Atlantic Tropical Stoms and Hurricanes," ESSA Technical Report, WB-6. Techniques Development Laboratory, Silver Spring, Maryland, 1968.
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- 15. Memorandum HUR 7-97. " Interim Report - Meteorological Characteristics of the Probable Maximum Hurricane, Atlantic and Gulf Ccests of the United States," 1968. From the Hydrometeorologocal Branch. Office of Hydrolog/, U. S. We ,ther Bureau to the Corps of Engineers.
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Weather Bureau to the Corps of Engineers.
l 9
a Q .
_'C-2 17 Pasquill. F. and Smith. F B.. "The Physical and Meteorological Basis for the Estimation of Dispersion,".1970. Paper presented at the Second International.
Union of Air Pollution Prevention Associations Washington D.' C.
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Coastline.: NOAA Technical Memorandum NWS SR-58 Southern Region, National Weather Service, Fort Worth.' Texas,1971.
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- 28. U. S. Department of Comerce Environmental Data Service: " Local Climatological Data. Annual Summary with Comparative Data - Pittsburgh, Pa.,' City Office" Published annually through 1972,
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- 30. The Computer program used Program to calculate the X/Q values Nuclear Power Station Evaluation (in FORTRAN code , J.may)be referenced as follo R. Sagendorf, ARL/NOAA programmer. Program available at the USAEC, Directorate of Licensing Bethesda, Maryland, or at the Air Resources Laboratory, NOAA Field Research Office, Idaho Falls. Idaho, o Hydrologic Engineering
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V C-3 Structural Engineering
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Missile Protection 34 Amirikian, A., " Design of Protective Structures," Bureau of Yards and Docks, Pu-blication No. NAVDOCKS P-51, Department of the Navy, Washington, D. C., August, 1950.
Seismic Design
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1
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'Maserials Engineering
,, 50. American National Standards institute, " Leakage-Rate Testing r.f. Containment L .
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Materials " ASTM-E-23-72. Annual Book of ASTM Standards, Part 31, July 1973, Philadelphia, Penns @ nia.
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~ 58. United States ' Atomic Energy ComIssion Regulatory Guide 1.14. " Reactor Coolant Pump Flywheel Integrity," USAEC, Directorate of Regulatory Standards, Washington, D. C.', October,1971.
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- 62. United States Atomic Energy Comission Regulatory Guide 1.43, " Control of Stain--
less Steel Weld Cladding of Low-Alloy Steel Components," USAEC, Directorate of Regulatory Standards, Washington, D. C., May, 1973.
1
- 63. United States Atomic Energy Comission Regulatory Guide 1.44, " Control of the Use
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64, ly dtad States Atomic Energy Comission Regulatory Guide 1.45, " Reactor Coolant
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W .y . 1 -; .
- 66. . United States Atomic Energy Comission Regulatory Guide 1.51, " Inservice' Inpection -u of1tSME Class 2 and A Nuclear Power Plant Components," USAEC, Directorate of - '
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Reactor !
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L 73. Westinghouse Electric Corporation, " Hydraulic Flow Test of the 17 x 17 Fuel i Assembly," WCAP-8279, Pittsburgh, Pennsylvania, Februery, 1974. :
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74.' Westinghouse Electric Corporation, " Safety Analyses of the 17 x 17 Fuel Assembly -1
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Engineered Safet.y Features j
i
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ORNL-TM-2412, Oak Ridge, Tennessee, January, 1970. -!
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- 81. Rittig W. H. , Jayne, G. A. , Moore, K. V. , Slater, C. E. , and Uptmer, M. L. , j "RELAP3 - A Computer Program for Reactor Blowdown Analysis," IN-1321 Idaho ;
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, 83. ' Schmitt, R. C), Bingham. G.' E. .' Norberg J. A.', '" Simulated Design Basis Accident Tests of the Carolina Virginia Tube Reactor Containment,' . inal Report. IN-1403 Idaho Nuclear Corporation,. December,1970.
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84; Slaughterbeck. D. D . "Comperison of Analytical Techniques Used to Determine zo
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Distribution of Mass and Energy in the Lt1uid and Vapor Regions of a PWR Contain-ment Following a loss-of-Coolant Acciden+," Special Interim Report, Idaho Nuclear Corporation, January, 1970.
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4 Instrumentation. Controls and Electric Power Systems
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91.' Institute of Electric and Electronic Engineers, "IEEE Standard for Electrical Penetration Assemblies in Containment Structures for Nuclear Fueled Power Generating Stations," IEEE Std. 317-1971, New York..New York.
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General Guide for Qualifying Class I Electric Equipment for Nuclear Power Generatin9 Stations." !EEE Std. 323-1971 New York, New York.
- 93. Institute of Electric and Electronic Engineers, "IEEE Standard Installation.
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Periodic Testing of Nuclear Power Generating Station Protection Systems," ;d IEEE Std. 330-197i New York, New York,
- 95. Institute of Ele:tric and Electronic Engineers. "lEEE Guide for Seismic Qualification of Class ! Electric Equipment for Nuclear Power Generating Station Protection Systems." IEEE Std. 344-1971. N.3w York. New York. f D 96 United States Tomic Energy Comission Regulatory Guide 1.6, " Independence Between Redue t Standby (Onsite) Power Sources and Between Their Distribution Systems." USAEC, Directorate of Regulatory Standards, Washington D. C.,
,- March. 1971.
- 97. United States Atomic Energy Comission Regulatory Guide 1.9. " Selection of Diesel Generator Set Capacity for Standby Power Supplies," USAEC. Directorate of Regulatory Standards Washington, D. C. , March.1971.
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- 98. United States Atomic Energy Comission Regulatory Guide 1.11. " Instrument Lines Penetrating Primary Reactor Containment." USAEC, Directorate of Regulatory Standards, Washinton D. C. , March,1971.
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100. United States Atomic Energy Comission Regulatory Guide 1.32. Use of _fEEE Std, 308-1971. " Criteria for Class IE Electric Systems for Nuclear Power Generating Stations," USAEC. Directorate of Regulatory Standards, Washington, D. C.. April, 1972.
101. United Stated Atomic Energy Comission Regulatory Guide 1.47. " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systets." USAEC Directorate of Regulatory Standards. Washington D. C. , May 1973.
102. United States Atomic Energy Comission. "Safet; Evaluation Report - Surry Power Station. Units 1 and 2." Docket Nos. 50-280 and 50-281. October,1971, 103. Virginia Electric and Power Company,, " Final Safety Analysis Report - Surry Power Station, Units 1 and 2," January 21. 1970.
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106. American National Standards Institute. " Standards for Selection and Training of Personnel for Nuclear Power Plants." ANSI N 18.1 1971, New York, New York, 1971.
107. United States Atomic Energy Commission Regulatory Guide 1.8. " Personnel Selection and Training." USAEC, Directorate of Regulatory Standards.
Washington, D. C. , March 10, 1971, 108. United States Atomic Energy Comission Regulatory Guide 1.17. " Protection of Nuclear Power Plants Against Industrial Sabotage." USAEC. Directorate of Regulatory Standards. Washington, D. C., June 1973.
109. United States Atomic Energy Commission Safety Guide 17. " Protection Against Industrial Sabotage." USAEC. Directorate of Regulatory Standards. Washington.
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Inittel Tests and Operation 1
110. United States Atomic Energy Commission. " Guide for the Planning of Initial {
Startup Programs," USAEC. Washington, D. C.. December 17. 1970. 4 111. United States Atomic Energy Comission. " Guide for the Planning of Pieoperational e Testing Programs." USAEC. Washington, D. C.. December 7.1970.
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