ML19182A200

Safety Evaluation

ML19182A200
Person / Time
Site:	FitzPatrick
Issue date:	11/20/1972
From:	US Atomic Energy Commission (AEC)
To:	Power Authority of the State of New York
References
Download: ML19182A200 (174)

Text

SAFETY EVALUATION BY THE DIRECTORATE OP LICENSING U.S. ATOMIC ENERGY COMMISSION IN THE HATTER OF POWER AUTHORITY OF THE STATE OF NEW YORIC JAMES A. PITZPATRICK NUCLEAR POWER PLANT DOCKET NO. S0-333 msKrN M rms ooc*.l'Mlirr ,s IASTER ,MJM.

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i ABBREVIATIONS a-c alternating current ACI Amercan Concrete Institute ACRS Advisory Co111111ittee on Reactor Safeguards ADS Automatic Depressurization System AEC United States Atomic Energy Commission AISC American Institute of Steel Construction ANS American Nuclear Society ANSI American National Standard Institute ASCE American Society of Civil Engineers ASNB American Society of Mechanical Engineers ASTN American Society for Testing and Materials A'IWS Anticipated Transient Without Scram AWS American Welding Society BBL Brookhaven National Laboratory Btu/hr-ft2 British thermal units per hour per square foot Btu/1b British thermal units per pound BWR Bo11ing Water Reactor cf* Cubic feet per llinute cf* Cubic feet per aecond Ci/aec Curies per aecond css Core Spray System DBA Deaian Ba*i* Accident DBE D**ian Basia Earthquake

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ii d-c direct current

( l ECCS Emergency Core Cooling System

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. l ESP Engineered Safety Feature ESW Emergency Service Water System

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• p degrees Fahrenheit l ft 2 square feet ft 3 cubic feet

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• A PSAll Final Safety Analysis Report

.. acceleration, 32.2 feet per second per second J

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f. ' ABC General Desian Criteria for Nuclear Power P lants

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,i General Electric Coapany a.J gallons per minute Hi.gh Effi.ciency, Parti.culate, Air

!I t IIBPA IIPCIS High Pressure Coolant Injecti.on System IDE Inatitute of Electrical and Electronics Bnaineers 1.11 inch 3Al'RPP Jaaea A. PitzPatrick Nuclear Power Plant 4k/k reacti.vi.ty change kV ki.lovolta

• , kV/fc kilowatt* per foot LSD lake aurvey datua 1b pound 1b/br pounda/hr

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iii LOCA Loss-of-Coolant Accident LPCIS Low Pressure Coolant Injection System LPZ Low Population Zone

!EHFR Minimum Critical Heat Flux Ratio m meters i

I mph miles per hour mrem millirem

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MSL mean sea level MSLIV Main Steam Line Isolation Valve lllD/ton megawatt days pr ton Mie megawatts electrical MWt megawatts thermal NMPC Niagara Mohawk Power Corporation NOAA National Oceanic and Atmospheric Administration

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NPSH net positive auction head NSSS Nuclear Steam Supply System OBE Operating Basis Earthquake

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PASNY Power Authority of the State of New York l ,,

PHS Public Health Service PMP probable maximua flood PSAR Preliminary Safety Analysis Report psi pounds per square inch paid pounds per square inch differential psig pounds per aquare inch gauge

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iv QA Quality Assurance QC Quality Control R&D Research and Development RCICS Reactor Core Isolation Cooling System RCPB Reactor Coolant Pressure Boundary Rem Roentgen equivalent man Residual Heat Removal System RPS Reactor Protection System sec second SGTS Standby Gas Treatment System SORC Site Operations Review Conmittee SUB Site Review and Audit Board x/Q atmospheric diffusion factor (sec/m 3 )

w/ o weight percent 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 of Production and Utilization Facilities Part 100 AEC Reactor Site Criteria

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V TABLE OF CONTENTS Page ABBREVIATIONS. * * * ** * * ** * ** * * * **************** *** ***** * * ********

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1.0 INTRODUCTION

. * **** * * * ******* * * * *** *** ************************* ** 1-1 2.0 SITE************************************************************ 2-1 2.1 Site Description ******************************************* 2-1 2.2 Demography and Land Use*.**..****.*.******************.**** 2-4 2.3 Hydrology ************************************************** 2-7 2.4 Geology Seismology *************************************** 2-13 2.5 Meteorology ************************************************ 2-13 2.6 Environmental Radiological Monitoring********************** 2-17 3.0 REACTOR.................*......*..**......*****.*.*.**...**....*. 3-1

3. 1 General**************************************************** 3-1 3.2 Nuclear Design. * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 3-2 3.3 Core Thermal and Hydraulic Design. * * * * * * * * * * * * * * * * * * * * * * * *

• 3-7 3.4 Reactivity Control****** * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 3-9 3.5 Reactor Internals******** * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 3-12 3.5.1. Design Criteria. * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 3-12 3.5.2 Dynamic System Analysis for Seismic. Operational and LOCA Loadings******************************** 3-14 3.5.3 Vibration Control********************************** 3-14 4.0 REACTOR COOLANT SYSTEM. * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 4-1 4.1 General**************************************************** 4-1 4.2 R eactor Coolant Pressure Boundary Design and Evaluation**** 4-1 4.3 Fracture Toughness***********.*****.***********.*********** 4-3 4.4 Reactor Recirculation System*.**************.*************. 4-4 4.5 Pressure Relief System***.**.*****************.*********.** 4-5 4.6 Residual Heat Removal System. * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 4-6 4.7 Leakage Detection System*******************************.**. 4-9 4.8 Reactor Coolant Auxiliary Systems *** * * * * * * * * * * * * * * * * * * * * * *

• 4-10 4.9 Sensitiaed Stainless Steel********************************* 4-1 2 '

i 4.10 Reactor Vessel Material Surveillance Program** * * * * * * * * * * * *

• 4-13 I 4.11 In*ervice Inspection Program******************************* 4-13  !

s.o CONTAINMENT SYSTEMS****** * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 5-1 5.1 General*****.******** * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 5-1

'I 5.2 Primary Containment. * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 5-1 5.2.1 De*ign ********************************************* 5-1 5.2.2 Missile and Pipe Whip Protection******************* S-4

vi TABLE OF CONTENTS (Cont'd)

Paae 5.2.3 Containment Iaolation ********************.*****.** 5-1 5.2.4 Leakage Testing Program********************.****** 5-7 5.3 Secondary Containment ************************************* 5-8 6.0 ENGI NEERED SAFETY FEATUKES**************************************

6.1 lleral. ***************************************************

6-1 6-1 t

6.2 Eaergency Core Cooling System (ECCS) ********************** 6-1 g 6.2.1. High Pressure Coolant Injection System (HPCIS) **** 6-4 i 6.2.2 Automatic Depressurization System (ADS} *********** 6-5 6.2.3 Low Pressure Coolant Injection System (LPCIS) ***** 6-5 l 6.2.4 Core Spray System (CSS) *************************** 6-6 6.2.5 Discussion 0£ ECCS Review************************* 6-6 6.3 Standby Gas Treatment System ****************************** 6-9 6.4 Other Engineered Safety Peatures ************************** 6-9 7.0 INST1lUNENTATI0N 9 CONTROLS AND ELECTRICAL SYSTEMS **************** 7-1 7.1 General *************************************************** 7-1 7.2 Plant Protection and Control Systems ********************** 7-2 7.2.1 Comparison of Protection Systems ****************** 7-2 7.2.2 Comparison of Control Systems ********************* 7-2 7.2.3 Protection System Testability ********************* 7-3 7.2.4 Bypass Indication for Plant Protection System and 7-3 Enaineered Safety Feature Equipment************* 7-3 7.2.S APRM Reactor Trip at 15 Percent Power************* 7-3 7.2.6 Condenser Low Vacuum Trip************************* 7-4 7.3 Independence of Redundant Plant Protection System Channels 7-4 7.4 Incident and Accident Surveillance Instrumentation ******** 7-5

7. 5 Anticipated Transients Without Scram (ATWS) *************** 7-5 7.6 Radiation and Environmental Qualification***************** 7-6 7.6.1 Radiation Qualification*************************** 7-6 7.6.2 Environmental Qualification *********************** 7-6 7.7 Electric Power Systems ************************************ 7-7 7.7.1 Offsite Power************************************* 7-7 1

I 7.7.2 Onsite Power************************************** 7-9 8.0 AUXILIAllY SYSTDIS *********************************************** 8-1 l 8 .1 General.....*..**...**..*.***.**....*****..**.*..*.**..*.. 8-1

• 8.2 Radioactive Waste llanaaement ****************************** 8-1

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vii TABLE OF CONTENTS (Cont'd)

Page 8.2.1 Liquid Radwaste System**************************** 8-2 8.2.2 Gaseous Radwaste System*************************** 8-6 8.2.3 8.2.4 Solid Radwaste System ****************************

Radwaste System Structural Evaluation *************

8-10 8-11 I 8.3 Proceas and Area Radiation Monitoring Systems ************* 8-12 I

I 8.4 Fuel Handling and Storage********************************* 8-13 8.5 Cooling and Service Water Systems************************* 8-14 8.6 Control Room Ventilation System*************************** 8-15 l

8.7 Circulating Water System********************************** 8-16 9.0 STRUCTURES AND EQUIPMENT************************************** 9-1 9.1 Structural Classification********************************* 9-1 9.2 Structural Design Criteria******************************** 9-1 t.

9.2.1 Wind and Tornado Criteria************************* 9-1 I 9.2.2 Flood Design Criteria***************************** 9-2 9.2.3 9.2.4 Kissi1e Protection Criteria***********************

Seismic Design Criteria***************************

9-3 9-3 I*i 9.3 Evaluation of Class I Structures and Equipment *****..***** 9-5 10.0 ACCIDENT ANALYSIS********************************** * * * * * * * * * *

• 10-1 cf 10.1 General*********************************************** 10-1 i 10.2 Loas-of-Coolant Accident****************************** 10-2 10.3 Refueling Accident************************************ 10-3  :-

10.4 Control Rod Drop Accident***************************** 10-4 10.S Hain Steamline Break Accident************************* 10-8 j 10.6 Conclusion********* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 10-9 t 11.0 CONDUCT OF OPE'RATIONS ***************************************** 11-1 11.1 Station Organization and Staff Qualifications********* 11-1 11.2 Teat and Startup Proara******************************* 11-3 -

11.3 Plant Procedures************************************** 11-4 11.4 Safety Review and Audit******************************* 11-4 l 11.s Emergency Planning************************************ 11-S 11.6 Industrial Security*********************************** 11-6 ..

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12.0 QUALITY ASSURANCE********************************************* 12-1

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viii TABLE OF CONTENTS (Cont'd)

Page 13.0 TECHNICAL SPECIFICATIONS******************** * * * * * ** * ** ** * * * * * **

• 13-1 14.0 CONFORMANCE WITH GENERAL DESIGN CRITERIA****************** * * ** *

• 14-1 15.0 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUAIU>S********** 15-1  :]

16.0 COMMON DEFENSE AND SECURITY ***** * * * * * * * * * * * * * * * * * * * * * * *** * * * * * *

• 16-1 17.0 FINANCIAL QUIFICATIONS**************************************** 17-1 17.1 Financial Protection and Indemnity Requirements *********** 17-1 17.2 Preoperational Storap of Nuc*lear Fuel..*....*.*.**.**.... 17-1

17. 3 Operating License. ...........*....*......*.*.*..**.*..*.*. 17-2 17.4 Concluaion************************************************ 17-3

18.0 CONCLUSION

S..................................................... 18-1 APPENDIX A Chronology of Regulatory Review................. ** * * *

• A-1 APPENDIX B Report of Coastal Enaineering Research Center..*...... B-1 APPENDIX C Report of the Departaent of the Interior - Geological SuY-'7ey ********************************************** C-1 APPENDIX D Report of Nathan M. Newmark Conaulting Engineering Services **************** * * * * * * * * * * * ** * * * * * * * * ** ** * * ** D-1 APPENDIX E Financial Qualificationa****************************** E-1

ix LIST OF TABLES Page TABLE 2-1 Population vs. Distance********* * * * * ** * * ** **** ** ****** ***

• 2-4 TABLE 2-2 Oswego County Population Trends *************************** 2-5 TABLE 2-3 Communities and Population Centers in the Vicinity of Plant ............... * . * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

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X LIST OF FIGURES Page FIGURE 2-1 Site Location********************************************* 2-2

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1 FIGURE 2-2 Arrangement of Joint Nine Mile Point - FitzPatrick Site*** 2-3 1

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1-1 1.0 Introduction This report is the Atomic Energy Commission's Safety evaluation of the application by the Power Authority of the State of New York (PASNY) and Niagara Mohawk Power Corporation (NMPC) for a license to operate the James A. FitzPatrick Nuclear Power Plant. The plant is owned by PASNY and will be operated by NMPC under contract with PASNY. It utilizes a single cycle, forced circulation, General Electric Boiling Water Reactor of similar design to several which have been previously approved for operation.

The plant rated thermal power level is 2436 MWt corresponding to a net electrical output of 821 MWe for which the .plant will be licensed. This safety review and evaluation were based on the plant design power level of 2SSO MWt, corresponding to a net elec trical output of 853 MWe, which is the ultimate desired power level at some future time. The plant occupies a 700 acre site on the shore of Lake Ontario in the Town of Scriba, Oaweao County, New York. It is located adjacent to and approximately 3000 ft east of the Nine Mile Point Nuclear Station of the Niagara Mohawk Power Corporation which has been in commercial operation aince 1969.

However, the two plants are independent and ahare no facilities or component* except for offaite power lines. Purther,an accident at one cannot initiate an accident in the other. The aite i* 36 ailea

1-2 northwest of Syracuse. The nearest population center, as defined by 10 CFlt Part 100, is the City of Oswego, seven miles west of the site, which had a population of about 24,000 in 1970.

On December 30, 1968, PASNY filed a construction permit application for the James A. PitzPatrick Nuclear Power .Plant.

A review of this application was made by the AEC's regulatory staff and by the Advisory Committee on Reactor Safeguar ds.

Both concluded that the*facility could be constructed without undue riak to the health and aafety of the public. Following a public hearing on March 31, 1970, construction of the facility was authorized by Construction Permit No. CPPR-71 issued by the Co111Di8sion on May 20, 1970. The application for operating license and the Pinal Safety Analysis ltaport were submitted on June 4, 1971. An Environmental Report was submitted on May 21, 1971.

In accordance with the provisions of the National Environ mental Policy Act of 1969, the AEC has made an independent and comprehensive review of all environmental issues and impacts associated vi.th the construction and operation of the PitzPatrick Plant. The AEC's Draft Environmental Statement was issued in November 1972. A &UDID8ry notice of the availability of the appli cant'* Environmental Report and the Draft Environmental Statement '

wa* publi*hed in the Pederal ..ai*ter. Thi* notice reque*ted conaant* from intere*ted person* on the propo*ed action and on

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• -*111",L 1-3 the Draft Environmental Statement. The Final Environmental State ment, which will consider all problems and objections raised by Pederal, State, and local agencies and officials, private organi zations and individuals, and the disposition thereof, is scheduled to be issued in April 1973.

This report summarizes the results of the safety evaluation performed by the Co111Dission's regulatory staff. Our evaluation included a review of information and data aubmitted by the applicant with emphasis on the following principal matters:

1. We evaluated the population density and land use character ietics of the eite environs, and the physical characteristics of the site, including aeismology, meteorology, geology and hydrology to determine that these characteristics had been determined adequately and had been given appropriate coneideration in the plant design, and that the aite characteristics were in accordance with the Commission's aiting criteria (10 CPI\ Part 100) taking into conaideration the design of the facility including the engineered aafety feature* provided. We also conaidered the concurrent operation of th* PitzPatrick and Nine Mile Unit 1 plants with reaard to r adioloaical and environmental impact* on the aite environs.

2. We evaluated *the desian, fabrication, conatruction, and testina criteria, and expected performance characteristic* of the plant atructurea, ayateaa, and components important to aafety to

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1-4 determine that they are in accord wi.th the Commission's General Deeign Criteria. Quality Aasurance c.riteria 9 and other appropriate guides. 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 facility to various anticipated operating transients and to a broad spectrum of postulated accidents. and determi.ned that the potential conse quences of a few highly unlikely postulated accidents (design basis accidents) would exceed those of all other accidents cobaidered. We performed conservative analyses of these design baeia accidents to detera:lne that the calculated potential off site doses that might result in the very unlikely event of their occurrence would be well within the Comnission's guidelines for site acceptability given :l.n 10 CPR Part 100.

4. We evaluated the applicant*' plans for the conduct of plant operationa 9 the oraaniaational etructure. the technical qualificationa of operating an4 technical support personnel, the measure* to be taken for industrial security, and the planning for emergency actions to be taken in the unlikely event of an accident that a:lght affect the general public, to deterai.ne that the applicant* are technically qualified to operate the plant and have eatabliahed effective organisations and plana for continuing ae operation of the facility.

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5. We evaluated the design of the systems provided for control of the radiological effluents from the plant to determine that these systems will be able to control the release of radioact ive wastes within the limits of the Commission's regulations (10 CPR 20) and that the plant will be operated in such a manner as to reduce radioactive releases to levels that are as low as practicable in accordance with the Commission's regulations in 10 CPR Part 50.

6.. We evaluated the financial qualifications of the applicants. and the protection and indemnity agreements made for the plant.

Our technical evaluation was accomplished with the assistance o coneultants in certain disciplin... The reports of our consultant* on hydrology. aeology and structural design are in cluded in this report as Appendices B 9 and C and D 9 respectively.

During our review of the information eublllitted in the PSAR we requested that PASNY provide additional iinforaation needed to com plete our evaluation. Thia additional inforaation vaa provided in eupple..nta to the PSAll. In the cour*e of our review we aleo held

..etinaa with PASNY and its representatives to discuss and clarify the technical information eublllitted. Aa a reeult of our reviillf 9 we requested a number of technical and adlllinietrative chana***

which are deecribed in various amendments to the oriainal appli cation. These chanae* are diecuaaed in appropriate aectiona of

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1-6 this report. A chronology of principal milestones related to our review of this application is attached as Appendix A. The chronology continues the review milestones from those previously listed in the Safety Evaluation Report on our review of the application to construct the FitzPatrick Plant.

Many features of the facility design are similar to those we i have evaluated and approved previously for other reactors now under construction or in operation. FitzPatrick uses a high power density. BWR-4 product line type of reactor similar to the recently reviewed Browns Ferry and Peach Bottom reactors and has an average power density of about 51 kW/liter and a peak fuel linear heat generation rate of 18.S kW/ft. To the extent feasible and appro priate. we have made use of our previous evaluations to expedite our review of the features that were shown to be substantially I'

the same as those previously considered. The application. as amended. together with the FSAR. as supplemented. and other I' i:

pertinent documents are available for public inspection at the i.

U.S. Atomic Energy Commission's Public Document Room. 1717 H i.

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Street. N.w .

• Washington. D.C. and at the Oswego City Library.

120 East Second Street. Oswego. New York.

The applicant has submitted a document entitled Industrial Security as proprietary information. The Commisa{on has deter mined that this information is of the type that may properly be

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1-7 withheld from public disclosure pursuant to sections 2.790(d) and 9.S(a)(4) of the Commission's Rules and Regulations.

Accordingly, this document will be withheld from disclosure except in accordance with the prvisions of section 9.10 of the Commission's Rules and Regulations.

Based on our evaluation of the application to operate the plant, subject to satisfactory resolution of certain items

_identified herein, we have concluded that the James A. Pitz Patrick Nuclear Power Plant can be operatd as proposed without endangering the health and safety of the public. f

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2-1 2.0 SITE

2. 1 Site Description The 700 acre James A. FitzPatrick site is located on the shore of Lake Ontario, in the Town of Scriba, Oswego County, New York. Figure 2-1 shows its location within the State, and with respect to some principal cities. The plant site is about seven miles east of Oswego, 36 miles northwest of Syracuse, 135 miles east-northeast of Buffalo, and 64 miles northeast of Rochester.

The location of the FitzPatrick plant within the joint Nine Mile Point-FitzPatrick (NMP-JAF) site is shown in Figure 2-2 The FitzPatrick portion of the joint site is owned by PASNY. The plant itself is located about 3,000 feet east of NMPC's Nine Mile Point Nuclear Station, Unit 1, which has been in coDDDercial operation since 1969.

The distance from the plant to the nearest property line other than the lake is 3,200 feet or 975 meters east. This has been specified as the radius of the exclusion area. The applicants have the authority to control all activities within the exclusion area, including removal of personnel and property, except for that portion which extends some 2,000 feet into the lake. Arrangements have been made with the Coast Guard to control access to this area of the lake in the event of emergency. The restricted area corresponds to the eite property 1ines and does not include any portion of the lake.

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i 2-4 The low population zone has been specified by the applicant to have a 3.4 mile radius. The nearest population center is the City of Oswego.

We conclude that the site meets the requirements of 10 CFR Part 100 with respect to exclusion area, low population zone and popu lation center distance. The site meets the requirements of 10 CFR Part 20 with respect to the restricted area.

2.2 Demography and Land Uae Table 2-1 shows the cumulative population as a function of distance out to 5 miles from the NMP-JAF site. The applicant states that the population enclosed within his specified low population zone of 3.4 miles is 1350.

Table 2-1. Population vs. Distance

-Distance 0-1 0-2 0-3 0-4 0-5 Population 18 36 1 833 1 765 3145 The City of Oawego, the nearest population center, had a 1970 population of 23,844, a gain of 7.6% over the 1960 census population of 22,155. Oswego County ahowed a population growth of 1 7.2% in the decade from 1960 to 1970, and the Town of Scriba grew from 2489 to 3619, an increase of* 45.4%.

Table 2-2 *hows the population trend for Oswego County and its lllinor civil divisions *ince the 1940 census. Table 2-3 identifies

2-5 Table 2-2. O.wego County Population Trends 1970 1960 1950 1940 OSWEGO COUNTY 100,897 86,118 77,181 71,275 Albion Town 1,452 1,125 1,036 1,094 Altmar Village 448 277 299 304 Amboy Town 557 524 482 493 Boylston Town 276 293 302 365 Constantia Town 3,547 2,730 1,947 1,538 Cleveland Village 821 732 555 440 Fulton City 14,003 14,261 13,922 13,362 Grandby Town 4,718 3,704 2,775 2,220 Hannibal Town 3,165 2,673 2,230 2,010 Hannibal Village 686 611 501 437 Baa tinge Town Brewerton (U) (Part) 6,042 438 --

4,457 3,063

-- 2,361 Central Square Village 1,298 935 665 568 Mexico Town 4,174 3,435 3,035 2,710 Mexico Village 1,555 1,465 1,398 1,348 Minetto Town 1,688 1,290 1,025 1,052 New Haven Town 1,845 1,478 1,259 1,194 Ozwell Town 836 663 752 806 Oawego City 23,844 22,155 22,647 22,062 Oawego Town 3,583 2,796 2,106 1,972 Palermo Town 2,321 1,663 1,397 1,148 Pariah Town 1,782 1,439 1,264 1,199 Pariah Village 634 567 574 521 Redfield Town 386 388 418 517

, Richland Town .5.,324 4,554 4,067 3.848 Pulaeki Village 2,480 2,256 2,033 1,895 Sandy Creek Town 2,644 2,506 2,354 1,821 .

Lacona Village 556 556 540 413 Sandy Creek Village 731 697 708 646 Schroeppel Town 7,153 5,554 4,037 3,219 Pheonix Village sand llidae (U)

Scriba Town 2,617 1,109 3,619 2,408 2,489 1,917 2,248 1,757 2,184 Volney TOlfll 4,520 3,785 3.106 2.659 Weat Monroe Town Breverton (U) (Part) 2,535 346 --

1,417 1,002 731 Willi... tOIIII Town 883 739 707 710

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2-6 communities and population centers by distance and direction from the plant, and shows their 197*0 census population. The distance from the plant to the nearest boundary of the City of Oswego is 5 miles.

Table 2-3. Conmunities and Population Centers in the Vicinity of Plant Miles From Total Place Plant Direction Population Oswego (City) 7 Southwest 23,844 Mexico (Village) 10 East 1,555 Pulaski (Village) 14 East 2*,480 Fulton (City) 14 South 14,003 Phoenix (Village) 23 South 2,617 Baldwinsville (Village) 26 South 6,298 Syracuse (City) 38 South 197,208 Watertown (City) 41 Northeast 30,787 Auburn (City) 42 South 34,599 llome (City) 51 Southeast 50,148 Kingston, Canada (City) 53 North 59,004*

Rochester (City) 64 Southwest 296,233 Utica (City) 65 Southeast 9,164 Population figures from 1970 U.S. Census.

• 1966 Census Dominion Bureau of Statistics.

The 1969 Census of Agriculture shows that 26.1% of land in Oswego County is devoted to farming, a decrease from the 33.9% quoted in the 1964 Census of Agriculture. Principal crops harvested are vegeables and grains, with 1351 acres devoted to orchard land.

Although dairying is signifLcant in the County, with 475 farms I

. l reporting a total of 12,573 milk cows for the most recent census, ,j the 1964 census showed 17,314 cows reported by 896 farma. Thia

2-7 represents a decrease of about 27 percent in the five year period.

The applicant states that the nearest farmland presently in use is a pasture about 1.2 miles south of the plant.

Several industrial establishments are located along the lake shore between Oswego and the plant. However, the two largest industrial firms in the immediate area are located in Fulton.

There are no defense, air traffic, lake traffic or industrial facil ities in the immediate area which would affect the safe operation of the FitzPatrick Plant or be adversely affected by its operation.

2.3 Hydrolop The ground elevation in the vicinity of the site is elevation 272 feet above Lake Survey Datum of 1935 (LSD - which is 1.23 feet higher than 1955 International Great Lakes Datum). The annual average Lake Ontario level is elevation 245.2 feet LSD and is controlled between elevations 243 and 249 feet LSD by dams operated by the applicant on the St. Lawrence River under a regulation plan devised by the International Joint Commission for Lake Ontario and the St. Lawrence River.

There are no streams in the vicinity of the aite. Surface drain age and groundwater movement are toward the lake. 'The sandy glacial till ia interspersed with ailts and clays which result in low permeabilities. No groundwAter *supply users are down aradient of the site, i.e. water passina beneath the site is not intercepted by any wells in its alow movement toward and into the lake.

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2-8 The applicant has reported 15 public water supplies within 30 mi.les of the site. two of which obtain water from the lake, the others using wells. The applicant has surveyed private groundwater u se in the area and has found approximately 53 wells within two miles of the site. Forty-eight of these wells are to the south and east of the site. The applicant has calculated the radiological consequences of a spill of the contents of outside radioactive storage tanks. The spillage would be discharged into the lake via the sewer main. The resulting whole body dose to an individual consuming his total daily water ration from the nearest lake water supply for the duration of the contandnation passage would be a few hundredths of a mrem. A more detailed discussion is included in Section 8.2.

Variations in Lake Ontario levels are caused by both runoff and wind aenerated waves and surges. Although PMF runoff was considered.

the level of the very large lake is not considered to fluctuate as greatly due to runoff as it is to wind aenerated waves and surges.

To estimate the maximum lake level at the site which could occur due to wind generated surges and waves. the applicant has utilized one- and two-dimenional models to calculate the consequences of the worst wind field reasonably possible over the long axis of Lake Ontario. west of the site. The one-dimensional model considers wind induced effects on the lake along a single line wh1.le the two dimensional model considers the response of the entire lake.

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2-9 Both models were used to reconstitute the January 1972 surge that resulted in recorded lake level fluctuations of less than a foot due to winds in excess of 50 mph over much of the lake surface. Once calibrated using the analysis of the January 1972 surge. the surge that would be produced by the worst wind field considered reasonably poaaible (called the probable maximum wind field) was computed to eatimate the probable maxilllUlll surge level at the site. The applicant estimated that the probable maximua wind field over the lake would have a aaximua wind speed of 88_ mph_ and would produce a aurge height of about 4.S feet above a conservatively high lake level of elevation 248 LSD at the aite to an elevation of 252.5 feet LSD. At the requeat of the ataff. the applicant has analyzed the surge effects that would be produced by a wind field having aaximua winds of 100 mph inatead of the 88 mph value previously assumed. The applicant has

.. t1..ted that the reaulting *urge level .would reach elevation 254 feet LSD at the aite.

Superiaposed on the *urae are wincl-aenerated wavea. Por safety related atructurea which are exposed to wave action the coabination o! *ur.. and wave effect* auat be conaidered to detend.ne whether adequate protection is provided. Por the intake acreenhouae atructure, vhich ia back froa the lake on a bluff and not directly aubject to wave action, the wave effect* would be vrtually daapened by the lona intake conduit and only the *urae level tran..ttted to the *creenbouae *UIIIP 1* of pri.aary iaportance *

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1 2-10 The safety-related pumps and equipment in the screenhouse are above elevation 255 feet LSD, one foot above the applicant's estimated surge level in the lake (which would be transmitted to the screenhouse sump via the intake conduit) which would be produced by the wind field which the staff considers most severe (100 mph I

1 maximum wind speeds).

The applicant has estimated the maximum wave runup of 7.5 feet l' at the site. The lowest elevation of exposed safety related j

jj s.tructures other than the interior of the screenhouse which 1.a di.s 1 1l cusaed above, that could be exposed to wave action is above elevation 272 feet LSD (ground level around the 1.ntake screenhouae), and thus well above wave effects. The intake structure on the bottom of the lake would not be exposed to such waves.

The applicant has stated that the plant yard and roof drai.nage :il

• of aafety related structures will safely di.scharge the probable 1i maximwn precipitation.

Cooli.ng water is to be taken from an intake structure about l

900 £eet from the ahoreline and conveyed to the acreenvell atructure Il through 1150 feet of 14 foot diameter conduit. The 1.nvert of the

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lake intake 1* at elevation 232.8 feet LSD. Diacharae froa the l l

l open loop cooling ayatem wi.11 be v1.a a diacharae tunnel about 1400 -

I:i feet from the *creenwell, and throuah a diaharae diffuser

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2-11 with six diffuser heads, three on each branch of a tee, with a distance of 150 feet between each of the heads.

Once-through heated water and processed liquid radwaste will be disposed of via the six diffusers some 270 feet north-northeast of the intake. The dominant current pattern in the lake is counterclockwise, i.e. from west to east at the site. Storm con ditions can, however, reverse the current pattern. The total flow of cooling water effluent of about 825 cubic feet per second (cfs) is expected to be dispersed and diluted with the water carried past the site by the lake currents by factors varying from 4 x 102 to greater than 10 3 , the lesser values being associated with higher ambient lake currents. Water travel times have also been estimated to vary from a minimum of about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to more than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> between the site and the Oswego Public Water Supply Intake and Selkirk Shore State Park.

The control over low lake levels exerted by the applicants' dams on the St. Lawrence ver, and the potential for wind gener ated low lake levels were considered in determinina whether an adequate water supply can be_assured over the intake atructure.

The applicant ha estimated that lake levels could gradually decline t o elevation 240.6 feet LSD in the event of an arbitrarily aasUJDed failure of its St. Lawrence River dalDS before natural

controls would prevent a further decline. By superimposing the effects of a probable maximum wind field which blows water to the western end of the lake, a probable minimum surge level can be I

! estimated at the site. Superimposing this negative surge on the minimum lake level that could result due to the postulated St.

Lawrence River dam failures, the resulting water level has been considered by the applicant to result in the minimum design lake lev*1 of elevation 236.5 feet LSD. Comparision of this level with the top and bottom elevations of the lake intake structure (elevations 232.8 and 222.8 feet LSD, respectively) indicates that an adequate water supply should be available from the lake.

We and our consultant, the U.S. Army Coastal Engineering Research Center (CERC), have concluded there is little or no likelihood that the plant can be flooded from wind induced conditions on Lake Ontario, and that the location and elevation of the intake assure a water supply from the lake under similar wind induced atorm conditions. The report from CERC i* included as Appendix B.

We agree with the app1icant that there is no potential for flooding due to the overflow of local atreama, and little likelihood of contamination of both aurface and groundwater.

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2-13 2.4 Geology - Seismolo&Y During the operating license stage safety evaluation a minor fault disclosed during excavation of the circulating water intake and diecharge tunnels was investigated by our g eological con sultant, the U.S. Department of the Interior - Geological Survey.

The inveetigation concluded that the fault is a minor geological

.' J' feature which has no eignificant eafety effect. The consultant's

'i I report ia inluded as Appendix C.

Nn other aeological or aeisa>logical investigations were deemed necessary beyond those completed and reported during the construction permit aafety evaluation. We conclude, based on our

/

previoua review of the PSAR and the earlier revi.ews of the USGS and NOAA, that the aeological and aeismolog ical investigations are adequate and that the design value* of 0.08g and 0.15g for the Operatina Baaia and Deaian Baaia Earthquakes, respectively, are appropriately conaervative.

2.5 Na&eorolo&Y The climate of northern New York State is quite variable because the area 1* aituated near the Jllljor atora track along which low preaaure ayate.. travel. Therefore, prolonaed period* of ataoapheric dffuaion rate* are a*nerally hiah

• Th* terrain at the aite ri*** to about 200 feet above plant arad* altout three 11111** aouth of the plant. The airflow i* aoatly

2-14 from the southwesterly sector extending from aoutheast to west with a strong preference for light winds (i.e., <10 mph) to flow from south through west. Therefore, the wind direction duri ng the moat adverse diffusion conditions tends to carry effluents j out over the lake.

An onsite meteorological program was initiated in 1962 to provide data for evaluation of the adjoining Nine Mile Point Nuclear Station. The program consisted of installation and measurements from a 200-foot tower situated in a cleared area near the shoreline. The tower has wind instruments at the 31- and 203-foot levels and temperature instruments at the JO-. 65-. 107- and 202-foot levels. The applicant has submitted a two year period of data records (1963-1964) in joint frequency distribution form. as deacribed in Safety Guide 23, to provide a baais for the staff'* evaluation of ataoapheric diffusion conditions. Por buildina and vent releases, the joint frequency distribution of wind direction and speed 111eas1.:.re, at the 31-foot level and vertical temperature difference (T) between the 202- and 30-foot levels waa used.

For releases from the 385-ft stack. the joint frequency diatribution of wind direction and speed aeaaured at the 203-foot level and 4T between the 202- and 30-foot levels was uaed. The joint data recovery durina the two year period of record was 90% for t he law level ¥1.nda and 97% for the hiah level winds.

2-15 In evaluating diffusion of short term (0-2 hr. at the site boundary and 0-8 hr. at the LPZ) accidental releases from the buildings and vents, a ground level release with a building wake factor, cA, of 1123m 2 was assumed. The relative concentration (X/Q) for onshore flow conditions which is exceeded 5% of the time waa calculated to be 1.3 x 10-4 sec/m 3 at the miniaum site boundary distance of 975 m. This relative concentration is equivalent to dispersion conditions produced by Pasquill type F acability with a wind speed of 3 meters/second. The relative concentration for comparable meteorological condition* at the outer boundary of the low population zone (5470 a) was calculated to be 1.s x 10-s aec/m 3

• For periods of 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.at the LPZ the relative concentration was calculated to be 3.2 x 10-6 aec/m 3

• In evaluatina accidental releaaea froa the 117 ..tar atack, an elevated point aource aoclified for terrain h*iaht ..... auaed.

l'uaigation condition* were aaauaed to exiat durina the first four hour* of Che accidental release period. The relative 0-2 hr. con centracion (X/Q) at the aite boundary va* eatiaated to be 5 .1 x 10- 5

• • c/m 3 at the neareat aite boundary (97Sa) ..auaina fUllliaation condition*

with a wind apeed of 2 aetera/aecond. At th* low population aone diacance, a relative concentration of 2.6 x 10-s aec/a 3 ... cal-culated ..auaina fuaiaation tbrouah a depth of 50 * (reduced froa full atack beia ht due to the ri** in terrain) and a wind apeed of

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2-16 2 meters/second for the first four hours after an accident. For the time period from 4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident the relative concentration from an elevated (50 m) release was calculated to be 6.2 x 10- 6 sec/m ?. . For time periods from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 30 days after the accident, the values presented in Safety Guide 3 for an effective stack height of 50 m were used.

The applicants' relative concentration estimates for abort time periods are generally higher than those calculated by the staff by at least a factor of two due to the applicant's use of wind fluctuation* measured at the 200-foot level for establishment of at1DOapheric atability and diffusion rates. However, the staff conclude* that the T*method of atability classification with the attendant diffusion rates is adequately conservative for use in the ataff's evaluation of doses.

Computation* of annual average offsite relative concentration for the atack release, considering topography and plume rise as a function of wind apeed, ahowed a aaximuaa value of S.S x 10-e aec/m 3 aouth of the atack at a distance of 4800 meters. The hiahe*t offaite annual averaae relative concentration of 1.6 x 10-6 aec/a 3 for vent release* occur* at the 975 aeter aite boundary diatance eaat of the reactor complex.

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2-17 We conclude that the meteorological data presented in the FSAR provides an acceptable basis for estimating atmospheric diffuaion for accidental and routine gaseous effluent releases from the plant.

2.6 Environmental Radiological Monitoring The FitzPatrick - Nine Mi.le Point area has een a subject of substantial radiological environmental study since 1967, two years I

prior to the startup of the Nine Mile Point-1 Station. The applicant propoa** to uae the preoperational studies carried out for the Nine Mile Point-1 Station as well as its operational environmental results to provide information on the radiological environment of the FitzPatrick Plant prior to its start-up. The operational apprach for enviro111Dental monitoring will be an augmented joint effort of the two adjacent plants.

The proposed environmental radiation monitoring program includes the collection and analysis of airborne particulates and halogens, .

precipitation, milk, fish, clams, shrimp, and surface water. Also,

.... ure..nta are made of ambient gawna radiation levels, both onaite and offsite, at 15 locations. The proposed frequency of sample analyei* rans** from weekly for*airborne particulate* and halogens to **1111-annually for aampl** associated with the aquatic environaent.

Nore detailed information on the program appears in Section 2 of the J

Pinal Safety Analy*i* Report.

2-18 We have evaluated the monitoring program proposed by the appli cant and find that it is generally acceptable. However, we will require that the program be augmented to include soil, vegetation, and lake bottom sediment samples. We will also require that the program be conducted on a continuous basis, as opposed to the applicant's proposal to implement sampling only when predetermi.ned radionuclide release rates are exceeded. The specific requirements for the entire monitoring program will be addressed in the Technical Specifications for the plant. The Techni cal Specifications will also define the specific surveillance program to monitor potentially contaminated milk from nearby cows and the associated administrative and corrective measures.

• ,*ra*** ........

3-1 3.0 REACTOR 3.1 General The nuclear steam supply system includes a General Electric Company (GE) boiling water reactor (BWR) which generates steam for direct use in the steam-driven turbine generator. The design of the FitzPatrick reactor is similar to the Cooper, Browns Ferry, Hatch and other reactors which have been evaluated at both the construction permit and operating license atages. The reactor core, containing nuclear fuel elements and control rods, i* aupported in a domed, cylindrical ahroud inside the reactor veaael. Steam aeparatora are mounted on the shroud dome. Two external, motor-driven recirculating puaps inject high-velocity water into 20 jet pumps which are located in the annulus between the ahroud and the reactor vessel. The high velocity water £rom the jet nozzles entrains and imparts energy to additional water

£rom the annular region. The combined liquid flow (about 3 t:Laes that 0£ the high-velocity water £low) enters the bottom 0£ the reactor core. Thia fluid becomes a eteaa-water lllixture as it paaaea through and cools the reactor core. The steam eaergea from the ateam aeparatora and dryers and enters £our 24-inch diameter pipes leading to the turbine-generator.

Reactor power ie controlled either by aoveaant of control rod* or by chanaing the apeed 0£ the two extern al recirculation puapa * .. actor power operation 1* tend.natad (reactor ehutdown) i'

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3-2 by inaerting control rods into the core. A standby liquid control syatem is provided as a backup system for reactor shutdown and operates by pumping a sodium pentaborate solution into the reactor.

3.2 Nuclear Design The initial core to be used for the reactor will consist of three types of fuel assemblies. Two of the three types contain gado linia in some of the assembly fuel rods. which acts as a burnable poion to control the core excess reactivity throughout the operating cycle. The reference design described for the construction permit safety evaluation employed boron-steel curtains which now have been replaced by gadolinia. Type I fuel assemblies will contain an average uranium-235 enrichment of 1.1% and the fuel pins for this aasembly will contain no gadolinia-uranium fuel pins. The type I fuel aaaembliea will no longer be used after approximately 10,000 Jan>/ton average exposure at which time they will be reaoved from the core and replaced with type II and III fuel aaamabliea. The type II and III fuel assembli.ea will contain an average uran111111-235 enrichment of 2.5%. Type II nd III fuel aaaaabliea will contain fuel pins with four different urani.um-235 enrichments to reduce the local power peaking factors. Three fuel pin* in each of the type II and III fuel asaeabliea contain full length gadolinia-uraniua fuel pine. In addition to the full lenath aadolinia-uraniua fuel pins, the type rr and III fuel asaemblea will contain one and two partial lenath gadolinia bearina fuel pin*,

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3-3 respectively. which provide axial flux shaping throughout the fuel cycle. The fuel rod cladding is evacuated and backfillea with an atmosphere of helium prior to sealing.

The end fittings of each fuel assembly are designed to assure that a higher enrichment fuel pin cannot be positioned in a lower enrichment fuel pin location. In addition. five of the high enriched fuel assemblies will have two removable fuel pins to facilitate interim fuel pin inspection during their expected core life. Quality assurance audits on fuel enrichment are per formed on fuel pellet groups prior to loading into fuel rods. on fuel pellet and fuel rod groups during pellet loading. and on fuel *.-

rod groups within a bundle during and after assembly into the fuel bundle. In addition. 100% of the fuel rods other than the segmented apacer rod will have been subjected to gamma scanning to confirm proper enrichment distribution.

The unit cell of the core consists of a. repeating array o f three type II or III fuel assmebliea and one type I fuel assembly.

The type I assemblies have flow re*tricting orifices to provide proper flow distribution to meet thermal hydraulic lilllita. Details of the phy*ical dimen*ion* of the fuel pin deaign are aiven in Section 3.2 of the FSA1l.

. * *. . , - \,"

3-4 Our evaluation of the nuclear design indicates the characteristics of this gadolinia controlled core will be similar to the gadolinia cores used in the Browns Ferry. Quad-Cities and Dresden 2 reactors. which were extensively reviewed and evaluated during our operating license review of these facilities. The adequacy of the calculations to predict gadolinia controlled initial core reactivities has been confirmed by critical experiments. Dresden I tests. and Quad-Cities start-up tests. The design criteria used in the Quad-Cities core design have been applied to the FitzPatrick core design. Even after burnout of the highly absorbing gadolinium isotope. the power density in a gadolinia fuel pin is sufficiently lower than the power density in a peak uranium-oxide fuel pin so that thermal-mechanical limits such as MCHFR. centerline melting and one percent clad strain are limited by the uranium oxide fuel pins. The extent of reduction of these limits in the gadolinia fuel pins resulting from changes in the thermal conductivities and lower melting point are the same as thoae evaluated in our review of the Quad-Cities core. The principal difference in the use of gadolinia between the Quad-Cities and FitzPatrick designs is a stronger degree of axial power shaping for the FitzPatrick core. This increased axial flux shaping capability is expected to improve the end of cycle power distribution.

' ** I -... * - "

3-5 The design limits for normal operation, MCHFR 1.9 and maximum linear heat generation rate of 18.5 kW/ft, are unchanged from the reference design evaluated for the plant construction permit application. The peak power densities and peak assembly power distributions used in the LOCA analysis remain within the range studied in General Electric Company Report, "Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors," NED0-10329, Supplement 1 (April 1971). From our review of the power distribution information we conclude that power distribution, as monitored by incore instrumentation, will maintain adequate safety margins.

Other nuclear design parameters which are important for abnormal operational transient analyses, such as as moderator and Doppler reactivity coefficients and the scram reactivity function, have undergone relatively minor changes as a direct result of the changes in the gadolinia core.

In Supplement No. 8 to the PSAR, PASNY has presented revised transient analyses which take into account the improvements in core design involving control augmentation using gadolinia and enrichment diatribution, changes in fuel design, nuclear parameters, acram reactivity function, and calculational models.

3-6 Our review of this supplement indicates that the calculated value of MCHFR, the thermal limit parameter, is still greater than 1.0 for the worst anticipated transient analyzed, i.e., the case of the single recirculation pump seizure from rated power and flow. In addition, based on our review of the similar fuel design for the Browns Ferry Nuclear Plant (Docket Nos. 50-359/360/396), only for a misorientation of the high-powered, high enriched assembly (a higly unlikely event) would the calculated MCHFR for the worst anticipated transient be less than 1.0, and then for only a few rods .

We consider thid acceptable based on our review of the Browns Perry Plant.

The major effect on transient results occurred from the change in the scram reactivity function. The General Electric Company performed parametric studies on the moderator void and Doppler coefficients for the more important transients. For these transients the changes in the nuclear parameters produced relatively minor changes in the results. We have concluded from our review that the proposed quality assurance plan for the control of fuel enrichments and burnable poison is satisfactory. and that the thermal limit values estimated for the moat critical transients and fuel loading errors analyzed are acceptable. We have concluded from our review that, in general. aatisfactory parameters have been uaed in the normal and transient analysis of the core deaian.

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The core thermal and hydraulic desian bases are formulated to limit the local power density within the core to levels which

.. sure maintenance of acceptable heat flux limit* such that the fuel rods do not overheat during normal operation or operational tranaienta.

The controllina mechanism that could cause fuel damaae during reactor tranaienta is aevere overheatina of the fuel cladding caused by inadequate coolina if critical heat flux condition* in the core are exceeded. The critical heat flux is defined as that which occur* at the fuel claddina at the onaet* of th* tranaition from nucleate boilina to fila boilina and below which fuel d...ae doe*

not occur. For deaian purpo*** the critical heat flux i* conaerva tively ue*d a* a fuel theraal lilllit althouah actual fuel d ... ae ..Y not occur until well into the film boilina rea1... The preaent critical heat flux lilllit* are calculated uaina th* correlation reported in th* GE topical report APED-5216, "Deaip Ba*i* for

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3-8 Critica1 Heat F1ux Conditions in Boiling Water Reactors." issued in 1966. This correlation is based on experimental data taken over the range of conditions representative of BWRs. The minimum critical heat flux ratio (MCHFR) is defined as the ratio of the critical heat flux correlation value at the corresponding fluid conditions to the actual maximum calculated heat flux occurring at a given point in the fuel assembly at any time during operation.

including reactor anticipated transients. A MCHFR >1.0 conserva tively assures that cooling of the fuel is maintained through nucleate boiling heat transfer.

The current design basis for normal operation is that the MCHFR calculated for any point is greater than 1.9 during normal operation and greater than 1.0 during anticipated transients. These liad.ts provide considerable margin between expected conditions and those required to cause fuel clad damage aince the critical heat flux correlation preaented in APED-5286 is conservatively based on a limit line drawn below all of the available experiaental data pointa. The maxi11111111 linear heat aeneration rate reached durina noraal rated power operation is not e xpected to exceed 18.5 kW/ft, correapondina to a MCHPR of 1.9.

We have conaidered the experi.. nt* planned at the AEC'* Power Burat Facility which are deaianed to provide inforaation reaardina

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3-9 fuel failure thresholds and the potentials for failure propagation resulting from loss-of-flow, flow blockage, local hot-spot generation and reactivity transients. We believe that this informa tion will assist us to verify the adequacy of the design safety margins for the FitzPatrick plant and provide assurance that individual fuel failures will not result in failure propagation or otherwise unacceptable releases of fission products.

We have reviewed the methods used to calculate the thermal limits, the experimental basis for the calculations, their validity a* damage limits and the applicant's analyses of normal operation and anticipated transients for this plant and previously reviewed reactors, and conclude that the design provides adequate margin to protect the core against fuel damage.

3.4 Reactivity Control Reactor power can be controlled either by movement of control rod* or variation in reactor coolant recirculation aystea flow rate.

A *tandby liquid control *Y*t** is al*o provided as a backup

• hutdown *Y*t***

There are 137 control rods which are used to brina t he reactor throuah the full ranae of power (fro* ahutdown to full power operation).

to *hape the reactor power diatribution. and to coap*n*ate for chanae*

in reactivity reaultina fro* fuel burnup. Each control rod drive baa

• • parat* control and rapid inaertion (a crAII) devic*** A colllllOn hydraulic pr***ure *ourc* for normal operation and a co111110n duap r

' a ' * -. , , * * "' , " * , . " ** *

• 3- 10

• I volume for scram operation are used for the drives. On the basis of our review of the drive system design and the aupporting evidence accumulated from operation of similar systems in other General Electric reactors, we conclude that the installed system will meet the functional performance requirements for this reactor in a safe manner.

The operating lifetime of the control rods will be governed by the loss of control effectiveness due to depletion of the boron-10 isotope, buildup of gas pressure by absorption of neutrons in boron-10 to produce helium, and change* in the yield strength of structural

..mbera due to fast neutron and gamma irradiation. Analytical methods have been developed to predict the need for control rod replacement baaed on the above considera.tions. Normally the control rods will be replaced durina a refueling outage in accordanc with the calcu lated replace.. nt requirements.

Durina operation at power level* between aero to 101 of the rated power, control rod reactivity worth* have been liaited by the rod worth lllinimiaer (RWM), a device which utiliaea a coaputer to restrict control rod patterns auch that the total worth of any in-sequence rod that can be aoved will be no aore than 1% 6. Por reactor power level*

in axe*** of 10% of the rated power, RWN operabilty h .. not been reuired. The AEC ha* been reevaluatina th* aodelina and conaequenc**

of th* poatulated Control Rod Drop Accident<*** Section 10). The

3-1 1 staff has concluded that modifications are required which would provide means to augment the RWM so that the probability of occur rence of the postulated accident is negligibl y low and/or that the consequences are consistent with the guidelines of 10 CFR Part 100. The applicant is considering a Rod Sequence Control System for thi* purpose, similar to that described in Amendment 43 to the Browns Ferry Application. The staff is reviewing the acceptability of this syatea aa a backup to the RWM.

A control-rod-ejection accident is precluded by a control rod hou*ina *upport *tructure located below the reactor pressure vessel, ailllilar to that inatalled on the other large General Electric reactors.

Thia *tructure lillllta the distance that a ruptured control rod drive houaina could be displaced. The applicant concluded, and we aaree, thac the control rod diaplacement would be ao *aall in thi* event that any re*ultina nuclear tran*ient could not be *ufficient to cauae fuel rod failure. Rapid control rod withdrawal 1* prevented b y the control rod velocity 11111.ter which liaita the free fall of a rod to 5 fc/aec but do** not retard acr .. action.

Reactor power can al*o be controlled throuah chana** in the pri.. ry coolant recirculation flow rate. The recirculation flow i

control ayat.. 1* th* normal control ..chod uaed to adjuat reactor  !

power level to atation load de..nd whenever the reactor

• operatna f

3-12 between approximately 60% and 100% rated power. The recirculation flow control system is designed to allow either manual or automatic control 0£ reactor power. These methods 0£ BWR control have been well established in other BWR facilities such as Dresden Units 2 and 3 9 Monticello and Millstone I.

The standby liquid control system is manually initiated from the control room to pump a sodium pentaborate solution into the reactor 1£ for any reason. the reactor cannot be abut down or maintained abut down with the control rods. It is designed to bring the reactor to a cold shutdown condition from the full power steady atate operating condition at any time in the core life with adequate capability to compensate for the effects of xenon burnup and decay.

On the basis of our previous review of similar design* and of aatiafactory operating e xperience with similar syste.. in other operattna BWRa. we conclude that the reactivity control features of the reactor are acceptable. aubject to the acceptability of the backup to the rod worth llinillliaer.

3.5 Reactor Internal*

3.S.1 Deaian Criteria Por all operatina loadina cateaori** (normal, upaet, e11er1ency and faulted), the deatan of the reactor internal* 1* in accordance

3-13 with Section Ill of the ASME Boiler and Pressure Vessel Code.

All internal components are designated as seismic Class I items.

as defined in Safety Guide 29. and are designed to withstand. in addition to normal and upset operating load categories. loads resulting from a Design Basis Earthquake. a Loss-of-Coolant Accident and the combination of these hypothetical events.

Calculated primary stresses in the reactor internals under the above loading combinations are within the acceptable emergency and faulted stress limits specified in Section III of the 1971 ASME Code.

Deflections of the fuel channels. control rod housings. and the core

• upport structure under the above loading combinations were limited to assure control rod operation and the preservation of core cooling geometry.

A fatigue analysis was performed using the ASME Boiler and Pres

• ure Ve*sel Code.Section III. as a guide. The 11M>St *ignificant fatigue loadina occurs in the jet pwap-shroud-shroud *upport area of the internals. The re*ulting usage factor* are well within accepted al lovab lea The fore** and atrea*e* reaulting from thermal ahock followina refloodina of the inner volume durina a LOCA or ..in at*** line break will be within the *y*t .. capabilitiea.

We find that the desian loadina conditions. d **ian atre** limit*.

def lection limit** and deaian fatiaue analy*** .. applied to he

l 3-14 ,'. .;

i J; \

reactor vessel internals are acceptable: and that adequate margins of safety are available to provide reasonable assurance that core flooding and cooling capabilities will not be impaired under the most severe loading conditions.

3.5.2 Dynamic System Analysis for Seismic, Operational and LOCA Loadings Seismic loading on the core support structure has been determined by means of a multimass dynamic analysis using a lumped mass mathe matical model of the reactor pressure vessel, internals and core support structure coupled with the contai nnaent building soil-structure model. We find this procedure acceptable.

Design loadings for the postulated Loss-of-Coolant Accident (LO CA) have been determined by computing the response of each struc tural member to the calculated peak pressure differential applied as an equivalent static load. In re*ponae to our concerns regarding the validity of this atatic analysis. the applicant has stated that l i

the natural frequency of the BWR internal structures is a>re than 1 ten ti..* the calculated forced frequency of the LOCA load** thus aaauring no sianificant dynaaic amplification. On the baaia of the inforaation *ubaitted by the applicant. we find this analytical

.. thod acceptable.

3.5.3 Vibration Control The J*..* A. PitsPatrick Nuclear Power Plant h** been daaignated aa the prototype plant for the 218 in 1.0. pressure ve**el *1** by

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3-15 G. E. and the applicant, since it will be the first of this kind to be tested. The applicant has made provisions for conducting a preoperational test program in accordance with Safety Guide 20, "Vibration Measurements of Reactor Internals." The applicant has etated that a post test surface inspection will be made of the core components which include such items as jet pump components, incore guide tubes, shroud components and welds, and core spray lines.

Where there are many of a similar component, the applicant has stated that the most severely stressed component will b.e examined. The inspection will be performed after the cold flow test. Subsequent meaaurements will be made during the hot flow test to verify that the cold flow test adequately represents the forcing functions and response characteristics of the hot flow test.

In addition, the applicant will submit, prior to testing, a preoperational teat plan that includes test procedures, instrumentation location*, analytical predictions and acceptance criteria in accordance with Safety Guide 20. The adequacy of the applicants' vibration Ii monitoring proaram will be resolved prior to licensing.

4-1 4.0 REACTOR COOLANT SYSTEM 4.1 General The principal equipment or system items to be discussed in this section are the reactor pressure vessel, the reactor recirculation system, the main steam and feedwater lines, and the pressure relief system. These items form the major components of the reactor coolant pressure boundary. The pressure boundary also contains various auxiliary and backup systems such as the reactor core isolation cooling system, residual heat removal system and reactor water clean up system. The portions of the power generation cycle o utside of the reactor building (i.e., turbines, the main condenser, and the feed water system) are not considered for the purposes of our safety evaluation as portions of the pressure boundary, as provisions will be made for their rapid isolation, when necessary, from the coolant systems within the reactor building.

All of the components of the reactor coolant pressure boundary were designed and built to the appropriate codes in effect at the time of order. In our review of the application, we have determined that the codes and code editions used comply with the rules of 10 CJl'll Part 50, Section 50.55a, "Code* and Standards."

4.2 Reactor Coolant Pressure Boundary Desian and Evaluation The reactor coolant system was designed as a Class I ( seismic) sy*tem to withstand normal design loads of mechanical, hydraulic and

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4-2 thermal origin. including anticipated transients and the Operating Basis Earthquake within the acceptable stress limits of the appli cable codes.

Additional analyses of the reactor coolant system have confirmed that the stress levels calculated under loads from the ,sign Basis Accident. the Design Basis Earthquake and the combination of these events are within the acceptable emergency and faulted stress limits of current component codes.

The reactor pressure vessel was designed. fabricated and inspected to the Class A requireaents of Section III of the ASME Boiler and i

i i

Pressure Vessel Code. 1965 Edition including published addenda through Winter 1966.

The reactor coolant system piping was designed, fabricated and inspected to the USAS B31.l.O Power Piping Code.

We find that the component codes used in the desian of the reactor coolant system are acceptable.

The applicant has stated that an acceptable method will be provided for measuring the vibration aaplitudes of those syste.. and coaponente that respond to the preoperational transients, such as PUIIP trips and valve cloaures, iaposed on the pipina ayate.. to verify their structural integrity. Aa a first step, the pipina ayste .. will be viaually observed durina testina. Those ayate.. and coaponents which appear to have aianificant responae will be fitted with inatru

..ntation to me .. ure the aaanitude of their reaponae. 'lbe applicant

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4-3 wi.11 submit, prior to testing, the criteria for visual observation and the acceptance criteria used to confirm the structural integrity of the piping and piping restraints.

4.3 Fracture Toughness To aasure that ferritic materials of pressure-retaining compo uents of the reactor coolant pressure boundary will exhibit adequate fracture toughness under normal reactor operating conditions, during system hydrostatic teats, and during transient conditions to which the ayatem may be aubjected, we have reviewed materials testing and operating limitations proposed by the applicant.

The applicant has stated in the FSAll, Amendment Noa. 7, 13 and 14 that acceptance testing for ferritic materials has been performed in accordance with the requirement* of the ASME Boiler and Pressure Veaael Code,Section III (1965 Edition, including Addenda through Winter 1966). Dropweight NDT data have been obtained for the reactor veaael plate aaterial opposite the core.

To eatablish operating preaaure and temperature *1i111itationa dur

\

ina atartup, ahutdown, and hydroatatic teating of the reactor coolant t

ayacea, the applicant has agreed to follow the reco11111endationa of Appendix G, "Protection Aaainat Non-Ductile Failure," of the recently revi*ed ASHE Code,Section III, 1972 su....r Addenda fracture touah-n*** rule* and ha* atated that reviaed heatup, cooldown and hydroatatic teat lilllitation* will be **tabliahed and docu.. nted by December, *1972.

4-4 This commitment is adequate and we conclude that the planned operation of the reactor coolant system will assure adequate safety

t margina.

4.4 Reactor Recirculation System

' ! The reactor recirculation system is similar to that which we have reviewed on other recent BWRa. It consists of two 28 in. maxi DIWD diameter loops located totally within the primary containment.

Each loop contains a variable speed, motor driven pump and two motor operated gate valves and provides the driving flow for 10 jet pumps.

With the exception of the pump motor drive equipment, the entire syatem is de*igned to seismic Class I requirements. The jet pump deaign aaaures the refloodability of the core to the level of the jet puaap auction inlet followina any recirculation line break.

Recent studies by GE have indicated that it llliaht be possible for the recirculation puap to drve its motor up to excessive rotational apeeda followina a loss-of-coolant accident. Thia could re*ult in the formation of hiah velocity missiles fro m pump and 111Dtor breakup. GE has proposed, in NED0-106.7.7, "Anl!lyaia of Recirculation Puap Overap .. d in a Typical General Electric Boilina Wacer .. actor," dated October 1972, to modify the recirculation ayacea on FiczPatrick and *illlilar BWRa by addina a clutch arranae

.. nt between the motor and puap to decouple the two and thereby preclude d...a1n1 overspeed of the motor. Miaailea froa puap I

4-5 overspeed will require additional protective means to preclude the possibility of impeller fragments being ejected from the open end of the aevered recirculation line and damaging the containment. GE has suggeated the use of additional line restraints or impact plates.

7he details of the propoaed modifications are presently under review by the ataff.

4.5 Pressure llelief System Overpressure protection of the reactor vessel is provided by a combination of aafety and relief valves similar to'thoae used at other recently licensed BWR plants. All are mounted on the main steam line* within the drywell. The two aafety valves are a balanced, apring-loaded type and discharge directly to the drywall interior. Their aetpoint is 1240 paia. The nine relief valves are dual purpoae. They are self-actuated at the aet relieving preaaure range 0£ 1080 to 1100 psis to limit the preasure riae and to avoid safety valve actuation. They can be operated by remote manual control* at pressures below the set point to control coolant ayatea preaaure. Six of the nine relief valve* contain pneuaatic diaphra .. I actuator* and open automatically in response to autoaatic depreaauri zation initiation signals. Thia automatic depreasurization *Y*t**

function is one 0£ the emergency core coolina ayate118 and 1* deacribed in Section 6.2. All relie£ valves diacharae throuah pipina directly to the auppreaaion pool. The preasure relief *Y*t** provi.dea adequate

4-6 protection of the reactor coolant boundary against overpressurization under the most severe operational overpressure *transient which is caused by sudden closure of all main steam line isolation valves.

l' The plant is assumed to be operating at design conditions and the I' .

j direct isolation valve position scram fails to occur thus requiring reactor shutdown by the backup, indirect, high pressure scram.

The applicant's analyses show that the peak coolant system pressure under these conditions is limited to 1289 psig, an 86 psi margin belw the ASME Code allowable of 1375 psig.

The applicant has submitted a "Summary Technical Report 0£ Reac-

/ j l l' tor Veasel Overpressure Protection" which was prepared in accordance '

l i with Section Ill 0£ the ASME Boiler and Pressure Vessel Code.

Thia report describes the design of the valves and presents an analy*ia of their pressure limiting capabilities. We have reviewed this report and conclude that the design and capacity of the pressure rele£ ayatem meets the intent of the ASME Code and is acceptable.

4.6 ***idual Heat Removal System The lle*idual Heat Removal System (RHRS) is designed for four 1111.jor modes of operation besides the low pressure coolant injection (LPCI) mode which is discussed in Section 6.2. The ays*tem consists I

of two heat exchanaera, four main aystem pumps, service water pumps, and a*aociated valves, piping, controls and instrumentation. All functional components are designed to satisfy aei*mic Class I

4-7 requirements. Each loop, consisting of one heat exchanger, two RHR pumps in parallel and ancillary equipment, is physically separated from the other. However, a cross connection by a single header makes it possible to supply either loop from the pumps in the other loop. Provision also exists for pumping RHR service water directly into the containment or the reactor vessel if necea sary. The RHRS operational modes are described briefly below. The 1lH1l pump deaign is controlled by the low preasure coolant injection mode deacribed in Section 6.

During reactor isolation the RHRS, in the condensing aode, can be_operated in conjunction with the reactor core isolation cooling ayatem (BCICS). With the reactor isolated, decay heat normally i* duaped to the auppression pool via the relief valves and the RCIC turbine exhaust. However, the suppression pool temperature under these condition* is liai.ted to about 130 ° P *uch that the temperature rise due to a aubsequent desian basis loas-of coolant accident would not cause the pool temperature to exceed 170 ° P durina the reactor blowdown. The conden*ina aode of lUUtS operation relieve* the burden on the auppre*aion pool by transferrina a portion of the decay heat to the IlHR *ervice water. Reactor

• teaa 1* taken to the *hell *ide of the 11HRS heat exchanaera and tranafer* heat to the aervice water in the tubea. The condenaate 1* either duaped to the *uppr***ion pool or returned

> J

4-8 to the reactor vessel through the suction of the RCIC pump.

Shortly after shutdown, both heat exchangers are* used to handle essentially all of the decay heat. After about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, one heat exchanger is adequate and the other is transferred to the pool

.1 cooling mode.

The suppression pool cooling mode utilizes the RHRS heat exchangers to cool the suppression pool water by transferring heat to the RHR service water. This can be used in conjunction wi.th the condensing mode or to provide long term suppression pool cooling following a loss-of-coolant ac cident blowdown.

The shutdown cooling and reactor vessel head spray mode is operated during normal shutdown and cooldown. Reactor water is diverted from one of the recirculation loops with the RHRS pumps through the RHRS heat exchangers where heat is transferred to the RHR service water *nd the water returned to the reactor vessel via the recir culation loop. Part of the flow is diverted to a reactor head spray nozzle where it maintains saturated conditions by conden sina the steam aenerated by the hot vessel 1illlls and internals.

The containment apray aode is initiated manually after the LPCI requireaents are aatisfied and aids in reducing post-LOCA drywell presaure. The RH1l puapa tranafer water from the suppression pool through the RHRS heat exchanaers where it is cooled by the RHll service 111111ter. The cooled water enters the containaent through header* in the drywell and above the auppression pool and reducea

4-9 the pressure by condensing existing steam. The spray collects in the bottom of the drywell until it overflows into the vent lines and drain* back to the auppression pool.

We conclude that the design of the RHRS is acceptable.

4.7 Leakaae Detection System Coolant leakage within the reactor containment may be an indication of a amall through-wall flaw in the reactor coolant pre**ure boundary. The leakage detection *ystem includes diverse leak detection .. thods. The major components of the ayatea are the contaimnent atmosphere particulate and gaseous radioactivity ' '

monitor*. and level indications on the containment aump. Indirect indication of leakage can be obtained from the contaimaent hUlll:l.dity.

pre**ure and teaperature indication*. Leakage detection outside the primary contaimnent is accoaplished by monitoring drain* or temperature* in apecific compartment* and rooma. There are at least two diverse leakage detection aethoda for each ayatea and area within the coolant boundary.

The limit* for identified and unidentified leakage rate* are 20 gpm and S o*. re*pectively. The applicant baa calculated that the ad.nimua leak rate froa any crack larae enouah to propaaate rapidly to a brittle failure 1* 1SO apa. The ayatea will have aufficient

• en*itivity to ..aaure ...11 leak* and ¥1.11 be provided-with auitable control room alar1111 and readouts. We conclude that the lealc.*a*

4-10

! detection system provides acceptable redundancy, diversity and

.,i i sensitivity to detect leakage from small through-wall flaws in the J

• t reactor-coolant pressure boundary. It ia consiatent with those of

"*. }r I

' ; previously approved BWll plants and is acceptable.

4.8 Reactor Coolant Auxiliary Systems The reactor coolant auxiliary systems include the reactor core isolation cooling system (RCICS), the reactor water cleanup

' f ayatem, and the main steamline and feedwater piping. These systems are designed to seismic Class I requirements except for the reactor water cleanup system external to the primary contaimaent and the main *team1ines and feedwater lines external to the primary contai111Dent. All piping in theae systems is de*igned and fabricated to the requireaents of the Power Piping Code, USAS B]l.1.0-1967.

We have reviewed the design, fabrication and inspection requirements for these aystems and their major component* and found them to be acceptable for their respective applications.

The *cICS has a similar deaign and function as in other recently reviewed BWll plants. It takes auction from the condensate atoraa* tank* or from the auppresaion pool and delivers about 40011Pm of water to the reactor vea*el throuah a feedwater line.

The pump ia powered by a ateaa driven turbine. I.ta pr:l...ry purpoae is to maintain aufficient water in the reactor v.. **l

• 1 ',

4-11 during periods of shutdown and isolation or in case of a loss of feedwater flow. The RCICS can be actuated manually or will start automatically on a reactor vesse1 low water level signal.

The Reactor Water Cleanup System provides continuous purification of about 100,000 lb/hr of the recirculation flow as a mean* of maintaining reactor water purity so as to limit chemical and corrosive action. It also removes corrosion products to limit aources for neutron act*ivation. The systea consists of pump*, regenerative and non-regenerative heat exchangers, filter deaineraliaers and other supporting equipment. '.

The main steam lines and feedwater piping aystea provide for the routing of reactor ateam to the turbine generator and the delivery of 11&keup water to the reactor vessel. The aain ateallllines are equipped vi.th flow restrictors to limit the rate of coolant lo** in the event of a line rupture outside of the primary containaent. Main ateamline isolation valves both inside and outaide the primary contaimaent provide a redundant means of terminating ateam blowdown in the event of auch an accident.

The applicant atates that the PitzPatrlck feedwater aparaer ha* been aodified aiad.larly to that of the Dresden Nuclear Power Station Unit 2 to eliminate flow aaldistribution aa a possible cauae of £1ux aayr etriea which had been experienced on earl:f.e!!"

planta.

4-12 We have reviewed the reactor coolant auxiliary systems on the basis of their similarity of design and function to others which we have previously reviewed and approved. We conclude that these systems are acceptable.

4.9 Sensitized Stainless Steel Stainless steel that has been sensitized has an increased susceptibility to stress corrosion cracking. The applicant has shown in Amendment No. 4 that all sensitized austenitic stainless

,i steel has been replaced on the t*itzPatriclr reactor pressure vessel except for the jet pump riser pads and recirculation inlet i

thermal sleeve attachment buildups. Significant sensitization of I these items has bean precluded by materials selection and control j of heat input and interpass temperatures during welding operations.

In addition. they were fabricated from weld metal with controlled ferrite content (at least 5 percent) and are not subject to stress corrosion cracking.

Auatenitic atainless steel used in other component parts of the Reactor Coolant Pressure Boundary. including relief and safety valve*. is fully annealed to preclude aenaitization. Hicrofiasurea have been avoided by utilizina weld metal with controlled ferrite content of at least 5%.

We conclude that the plannina to avoid lllicrofisaurea and aenai tization of auatenitic stainless ateel during the fabrication period 1* acceptable.

4-13 4.10 Reactor Vessel Material Surveillance Program A material surveillance program is required to monitor changes in the fracture toughness properties of the re.actor vessel materia1 as a result of neutron ir ra diation. The applicant has shown in Amendment No. 4 that the material surveillance program will comply with the proposed AEC § 50.SSa Appendix_ H, Reactor Vessel Material Surveillance Program Requirements," and ASTM E 185-70. The program specification is acceptable with respect to the number of capsules, number and type of specimens, withdrawal schedule, and retention of archive material.

We conclude that the proposed program will adquately monitor neutron radiation induced changes in the fracture toughness of the reactor vessel beltline material.

4.11 Inservice Inspection Program Selected welds and weld heat-affected zones must be inspected periodically to assure continued integrity of the reactor coolant pressure boundary during the service lifetime of the plant. . \

The applicant has stated, in the FSAR, Appendix F and in .<

Amendment No. 4, that the inservice inspection program for the "J reactor coolant pressure boundary will generally comply with Section XI of the ASME Boiler and Pressure Vessel Code, Rules for Inservice Inspection of Reactor Coolant Systema" 1970 edition, to the extent practical.

4-14 The applicant has stated in Amendment No. 12 that access to the Group B and C fluid systems such as the engineered safety systems, reactor shutdown systems, cooling water systems and the radioactive waste treatment systems outside the limits of the reactor coolant pressure boundary will be provided for visual inspection. The applicant has also made a commitment to keep abreast of developments in automated equipment which would permit inspection at presently inaccessible locations.

We conclude that the access provisions and planning for inservice inspection are acceptable as the provisions of the AEC Guideline, "Inservice Inspection Requirements for Nuclear Power Plants Constructed with Limited Accessibility for Inservice Inspection,"

(January 31, 1969) have been aatisfied.

/ . '

..1 .* * . * ,. *.** .. *., *-

5-1 5.0 CONTAINMENT SYSTEMS 5.1 General The containment systems include the primary containment which utilizes the pressure suppression concept and the secondar y containment which is formed by the low-leakage reactor building surrounding the primary containment. The reactor building has an air recirculator system and a Standby Gas Treatment System (SGTS) to mix and filter primary containment leakage prior to its discharge to the environment via the in stack.

5.2 Primary Containment 5.2.1 Design The primary containment is a typical "lightbulb" pressure suppression system consisting of a drywell, pressure suppression chamber (torus), and a connecting vent system. The drywell has a steel spherical lower portion 65 feet in diameter, and a steel cylindrical upper portion 35 feet 7 inches in diameter. The overall height of the drywell is about 111 feet. The pressure suppression chamber is a steel torus located below and encircling the drywell, with a centerline diameter of approximately 108 feet and a cross-sectional diameter of 29.5 feet. Eight vent pipes lead from the drywell to a header inside the torus* and 96 - 24 '.,i; inch diameter downcomer pipes project downward from the header l

..., .. '.,.i

' ** *-.i

f 5-2 and terminate approximately 4 feet below the water surface of the pressure suppression pool. The free air volumes in the drywell and torus are approximately 150,000 ft 3 and 114,000 ft 3 , respectively.

The pressure suppression pool contains about 105,600 ft 3 of water.

In the event of a reactor coolant system pipe rupture within the drywell, the released steam passes through the vent pipes, torus header,and downcomer pipes into the suppression pool water where it is condensed. This transfer of energy into the pool w ater reduces the peak accident pressure that otherwise would be experienced by the primary containment.

The applicant has calculated that the peak pressures that might be reached as a result of the design basis loss-of-coolant accident, using the GE Topical Report NEDO 10320 and Supplement 1 thereto, are 45 paig in the drywell and 26 psig in the pressure suppression chamber. These pressures were calculated assuming a hypothetical instantaneous break of one recirculation loop pipe. The analytical methods used are similar to those used and accepted on other recently reviewed BWR plants.

The primary containment is designed for an internal pressure of 56 paig coincident with drywell and suppression chamber temperatures of 309 ° F and 220 ° F, respectively. The design leak rate for the containment is 0.5% per day at the design pressure. In accordance

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• l

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"'1 5-3 with Section Ill of the ASME Boiler and Pressure Vessel Code, maximum d rywell pressures up to 62 psig are permissible for this design.

Combinations of live, dead, and seismic loads in conjunction with thermal stresses have been considered in the design analysis. The design also considered the jet forces that might act on the contain ment consequent to a pipe severance. Adequate strength has been provided to prevent failure of the containment wall as a result of direct jet impingement, and crit,ical penetrations have been provided with restraints and auxiliary stops to limit pipe movement and prevent failure of the containment.

The primary containment was designed to sustain the combination of loads resulting from the design basis loss-of-coolant accident, earthquake, and the conventional live and dead loads within the stress limits defined in Subsection Ill B of the ASME Boiler and Pressure Vessel Code 1968 and applicable addenda in effect as of June 1968. We find the design stress limits for the primary containment system to be acceptable. /

Containment piping penetrations satisfy the criteria of the ASME Boiler and Pressure Vessel Code noted above. Lines connected to the reactor coolant system incorporate a sleeve to extend the drywall to the outer isolation valve, thus containing the effluent I*'

. *.7

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5-4 in the event of a pipe break. Hot lines which must sustain large thermal and mechanical stresses are designed with combinations of penetration sleeves and flued fittings.

Based on our review of the information contained in this application and similar designs we conclude that the primary containment design basis is acceptable.

5.2.2 Missile and Pipe WhiE Protection Several locations on the main steam lines and feedwater lines are not restrained to prevent pipe whip in the event of pipe failure at these locations. The applicant has stated that the physical layout withirt the drywell precludes restraints at these points. For all other lines and locations, restraints have been provided where a break could result in containment impact. The applicant has identified the unrestrained high stress areas in these lines where breaks could result in pipe whip such that the pipe could impact the primary containment wall. At those locations which are accessible the applicant has provided 1-1/4 inch thick impact plates as sup plementary protection for the drywe11. In addition, he has agreed to perform augmented inservice inspection of these weld locations during each inspection period. At the remainder of these identified areas the physical layout precludes installation of impact plates.

Here, the applicant will perform augmented inservice inspection of

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the welds during each inspection period. The requirements of this augmented inspection will be set forth in the Technical Specifications and will call for 100% rather than 25% inspection during each period.

The applicant has also considered the effects of pipe whip on the emergency core cooling systems. The systems are redundant and physically separa.ted such that a ruptured pipe could impact and l affect only one of the redundant ECCS. The remaining ECCS components were shown to limit peak fuel clad temperature to 1370 ° F following the most severe postulated break sequence.

The applicant has considered the effect of missiles ranging in size from nuts and bolts to valve bonnets* and conc*ludes that no missile would have sufficient energy to penetrate the containment.

In addition, where possible. components are arranged so that the direction of flight of potential missiles is away from the contain ment wall.

The effects of pipe whip and steam jet impingement on the shield and vessel support structue resulting from a LOCA occurring within the sacrificial shield area were analyzed and found to be acceptable.

We conclude that the applicant has provided adequate measures to protect against the occurrence and consequences of missiles and pipe whi.p.

5.2.3 Containment Isolation The ability to isolate the primary containment provides the necessary integrity between the coolant system pressure boundary.

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5-6 l or the containment atmosphere, and the environs in the event of accidents or other non-nominal conditions. Isolation is accomplished by means of valves. The numbers, types and locations of these valves in the various lines depend on the manner in which the lines penetrate the reactor vessel and the containment. Where necessary, the valves are equipped with operators and close automatically when sensors detect certain accident or fault conditions.

The consequences of postulated pipe failures both inside and outside of the containment have been evaluated and are described in Section 10. The isolation valves and their control systems have been reviewed to assure that no single failure can result in a loss of containment integrity. An exception exists in the case of instrument lines connecting to the reactor coolant system which penetrate the containment and dead-end at instruments located in the reactor building. Such lines are provided with manually operated.isolation valves and excess flow check valves, both of which are outside the containment. A break in the line between the containment and the outer check valve would result in blowdown directly into the reactor building.

The applicant has installed 1/4 inch diameter orifices in each

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5-7 of these lines inside the primary containment to prevent over pressurization of the reactor building and limit offsite doses to substantially below the 10 CFR Part 100 values in the event of the postulated instrument line break. Based on our review of the design we conclude that the provisions for instrument lines penetrating the primary containment are adequate and satisfy the supplement to Safety Guide 11.

Leakage through the closed main steam line isolation valves following a postulated LOCA presently relies on the low leakage characteristic of the valves. The acceptability of present leakage limits and the need for an auxiliary sealing system are under study by the staff. There is nothing in the existing design which would preclude incorporation of an additional sealing feature if such is determined to be necessary. The applicant will continue to study developments in this area.

Based on our review we conclude that the primary contain ment isolation provisions are adequate.

5.2.4 Leakage Testing Program i Leakage testing of the reactor primary containment and I

associated systems is intended to provide initial and periodic verification of the leaktight integrity of the containment.

The applicanc has stated in Amendment Nos. 4 and 5 that the primary reactor containment and its components have been designed

.... .,.' ., . ., /* ..

5-8 so that periodic integrated leakage rate testing can be performed at the calculated peak pressure. Penetrations, including personnel and equipment hatches and airlocks, and isolation valves, have been designed with the capability of being individually leak t*ested at calculated peak pressure.

We conclude that the containment system will permit contain-ment leakage rate testing in compliance with the AEC proposed "Reactor Containment Leakage Testing for Water Cooled Power Reactors," 10 CFR § 50.54(0), Appendix .J, and therefore is acceptable.

In addition to agreeing to meet the requirements of proposed

.J Appendix .J, PASNY has agreed to perform a leak test of drywall to suppression chamber piping, headers, downcomers and vacuum breaker valves at each refueling outage. They will also determine acceptable bypass leakage limits and other teat criteria and will be required to perform frequent surveillance testing of the vacuum breaker*. We have not completed our review of the details of the te*t and surveillance program. However, the applicant has indicated his intention to base it on the recently approved Brown* Perry leak check program * . We find this collllllitment acceptable pending completion of our review.

S.3 Secondary Containment The reactor building, toaether with the Standby Gaa Treac.. nt Sy*tem (SGTS) and the main *tack, form the **condary

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5-9 containment. It is a seismic Class I structure completely enclosing the reactor and primary containment. It also houses refueling and reactor servicing equipment and other auxiliaries to support reactor operation. The reactor building is .a reinforced concrete atructure up to the refueling floor. Above this level it is of steel frame construction with insulated metal aiding and aealed joints. The roof is insulated metal deckina.

During normal operations; the reactor buildina atmosphere is monitored and exhausted to the environment through the roof vent. In the event of an accident, the reactor building isolation and control aystea would isolate the reactor buildina and start the SGTS. The reactor building is desianed for a maxiaua inleakage rate of 100% of its volume per day at a negative preesure of 1/4 inch of water while the SGTS is operating.

Thia provide* for control*, proceasing and discharge throuah the 11Ain etack of any radioactive aaaes that aiaht be released into the buildina froa the priaary containaent durina or followina an accident. The SGTS ia diacuaaed further in Section 6.

A teat proaraa will be conducted prior to operation and per1odcal1y thereafter to deaonetrate the leakaae rate and

• j

i.

operab111ty of the varioua ay*t*-* Baaed on our review of Chia and.previoualy approved ailllilar BWll aecondary containaenta

• • conc1ud* that the deaian and teetina are acceptable.

• • -*' _j -,-. - --.. ; .. ' ' !

6-1 6.0 ENGINEERED SAPETY FEATURES 6.1 General There are a large number of diversified enaineered aafety features I

provided in the design of the PitzPatrick Plant. Many of these are

! discrete and have been included in this section of the report, while other* are closely related to other aysteas or aspects of the plant and are more logically discussed elsewhere. Aside from the

... raency core cooling eyste11&, which are of areatest interest, other* described herein include the *tandby gas treatment aystem,

...raency **rvice water system, containment atmosphere control ayatem, and contaimaent inerting *ysteaa.

6.2 B..raency Core Coolin& SyatelD8 (ECCS)

The BCCS *ubayste11& provide ...raency core coolina during those po*tated acc:ldent cond:ltiona where :lt 1* aaauaed that mechanical fa:llur** occur :ln the primary coolant *Y*t** piping reaulC:lna in a 1oa* of coolant from the reactor veaael areater than the ava:llable coolanc rekeup capac:lty uaina noraal operat:lna equipaent. The BCCS aubayac... are provided of *ucb nuaber. divera:lty 9 reliability. and redundancy that, even 1.f any act:lve coaponent of the BCCS fail*

dur:I.Da a 1oa*-of-coolant acc:ldent. inadequate cool:lna of th* reactor core v:111 not reau1t.

The *-r..ncy core*cooli.na *Y*C*- cona:l.*C of cvo hf.ah pr***ure ayac... : ch* bi.ah pr***ure coolanc :l.njecC:lon *Y*t- (BPCIS) and

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6-2 the automatic depreasurization system (ADS); and two low pressure ayste111&: the low pressure coolant injection systea *(LPCIS)

• which is one mode of operation of the Residual Heat Reaoval Sys** (IUlllS).

and the core spray system (CSS).

The ECCS for the FitzPatrick Plant are functionally similar to those of other GE 1967 product line BWll facilities for which operating licenaea have already been iaaued. 'ftley are designed aa Claa* I (aeialllic) ayate.. and all included pipina i* d**ianed and fabricated in accordance with AIISI B31.1.o._

(1969) "Power Pipina."

The various ayate.. are initiated by a hiah dzyvell preaaure aignal or a low preaaure va***l water le,,.l aianal. except for the ADS which require* coincidence of both aianala. 'Die ECCS 1*

deaiped to provide adequate core coolina and to lia:lt the peak fuel claddina teaperatur* for the coaplete accident apectrua up to and includina the deaip b .. 1. loe* of coolant accident. 'lh*

deal b .. i* break for PitaPatrick h .. an area of about 4.2 aquare feet wich 1* obtained by au 1na the ar** of a completely aevered auction line to a recirculation puap (3.7 fc2 ) and the effective area of ten jet PUIIIP no**l** (0. 5 ft 2 ). 'ftle aces alao fuactiona co lillit peak fuel claddina t...erature* in Che event of a aain ate .. line break inaide the dr,vell.

6-3 A loss of offsite power wil1 not prevent ECCS operation and all of our evaluations of the system have been made assuming that only onsite electrical power is available. In addition, the performance capability has been shown to be adequate assuming a failure 0£ any active component within the ECCS. This single failure criterion has been applied coincident with the assumed loss of offsite power.

The applicant ana1yzed the availability of adequate net positive suction head (NPSH) for all ECCS pumps in conformance with Safety Guide No. 1 which requires that there be no reliance on calculated increaaea in containment pressure. The most liting case occurs during the long term transient following the design basis LOCA when one core apray and one RHR pump will be runnina continuously. The analyai* ahows that a containment overpressure of less than 2 psi would be required for a period of about 25 mi.nutes during the transient to aasure adequate NPSH. A containment overpressure of 7 psi s aval able durina this period. Althouah the deaign does not fully meet the auidelinea of the aafety auide we have concluded that the applicant's analysia 1* conservative and that there ahould be adequate NPSH to the BCCS pumps even in the unlikely event of a LOCA.

The BCCS puapa are located in a ain1le creacent-ahaped rooa adjacent to the preaaure auppreaaion chaaber. The applicant ha*

inacalled a partition to phyaically **parate the RCICS froa the HPCIS and redundant coaponenta of the LPCIS and CSS and preclude

6-4 a sing1e f1uid 1ine break from flooding both loops. Spillways are provided which divert flood water from either of the two compartments of t he crescent shaped room to the area beneath the suppression chamber before the water level reaches the top of the partition. We have reviewed this design and conclude that it provides adequate flood protection for redundant ECCS components.

6.2.1 High Pressure Coolant Injection System (HPCIS)

The HPCI system provides and maintains an adequate coolant inventory and limits fuel clad temperatures in the event of small breaks in the Reactor Coolant Pressure Boundary (up to about 0.1 ft2 in liquid lines and 1.3 ft 2 in steam lines). A high pressure system is needed for such breaks because the reactor vessel depressurizes slowly, thus delaying operation of the low pressure systems. The HPCIS includes one 100% capacity steam turbine driven pump which injects 4250 gpm of water through one of the feedwater lines into the reactor vessel. Steam for the turbine is supplied from one of the main steam headers in the drywell. Exhaust steam is discharged to the suppression pool through a submerged pipe.

Initially the HPCIS takes suction from a common header fed by a 100,000 gallon reserve in each of the two condensate storage tanks. Should this supply be inadequate, suction is transferred automatically or manually to the suppression pool.

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• 6-5 6.2.2 Automatic Depressurization System (ADS)

The ADS reduces reactor vessel pressure during accident situations in which the HPCIS fails to function properly. The system uses six of the 9 pressure relief valves to relieve high pressure steam to the suppression pool and depressurize the reactor coolant system rapidly enough so that the low pressure cooling systems can operate before excessive fuel clad temperatures result.

The ADS actuates upon coincident signa1s of reactor low water leve1 9 high drywell pressure and discharge pressure indication of any LPCIS or CSS pwap. after a two minute delay. The delay provides time for the operator to cancel the signal -if he determines that it is false or that the system is not required.

6.2.3 Low Pressure Coo1ant Injection System (LPCIS)

The LPCIS mode of the RHR system provides rapid flooding of the reactor vesse1 in the event of large coo1ant line breaks (from 0.2 to 4.2 ft2 ). LPCIS and CSS protection also extend to a sma11 break stuation in which the control rod drive hydrau1ic pwnps.

HPCIS and RCICS are a11 unab1e to maintain the reactor vesse1 water 1eve1 and the ADS has operated to depressurize the vessel. The LPCIS includes four 33 1/3% capacity pumps. each of which is designed to deliver 7 9 710 gpa of water from the suppression pool into one of the two recirculation loops. The water enters the vessel through the jet puaps. A recent redesign has been inco.rporated in the loop J

6-6 selection logic to eliminate a postulated single failure following a LOCA, wherein the auction valve would close in addition to the dis charge valve in the breached recirculation loop, thus bottling up the break and interfering with proper operation of the ECCS systems.

The redesign permits closure of only the discharge valve in the intact loop following a LOCA. The auction valves can be closed only manually. This logic has been reviewed and found to be acceptable

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. 6.2.4 Core Spray System (CSS)

The core spray system provides a high volume spray to the reactor core under the same conditions as does the LPCIS. It consists of two 100% capacity redundant subayste1DS, each capable of delivering 4,625 gpm of water from the suppression pool through spray apargers above the core. The CSS is capable of cooling the core independent of the LPCIS.

6.2.5 Discussion of ECCS Beview The performance of the ECCS was analyzed using the assumptions and caculational techniques described in the Co111Diseion's Interim Policy Statement providina interim accepance criteria for the per formance of ECCS in light water nuclear paver plants aa published in the _Federal Beaister on June 29 9 1971. The analysis applied the AEC

)

} aaauaptions with no deviations. Break sizes froa 0.02 ft2 to 4.2 ft 2 were treated in the analysis. Various single failure aasu111ptione were made to determine the situation that resulted in the maxilllUIII fuel clad temperature.

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6-7 For the LOCA with the largest break size, (the design basis accident), the calculated peak clad temperature is 2095 ° F, assuming a failure of the LPCIS injection valve that renders both LPCIS inoperative and leaves only the two core spray loops. Thi$ was the single failure that results in the maximum peak clad temperature.

The corresponding metal-water reaction was calculated to be less than 0.12%. An assumed failure of one of the two station diesel generator systems resulted in a calculated peak clad temperature of 1900 ° F. In this latter analysis, two LPCIS pumps and one core spay pump were assumed operable.

The applicant will calculate and submit to us the effects o f the most severe postulated fuel loading, orientation and gadolnia concentration errors on the peak fuel cladding temperatures under LOCA conditions. However, our review of a sim:llar error ana1ysis for the Browns Ferry Nuc1ear Power Plant showed that the resulting temperature increases still y6eld peak c1ad te111peratures be1ow the 2300 ° F criterion. (:

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'l Analyses for the entire break spectrum, up to and inc1uding a L i i double ended aeverence of the largest pipe of the reactor coolant JI

}

aystem (the DBA) ahowed a continuous decrease in the peak c1ad f]

1.

i.

i temperature and percentage of aetal-water reaction as the br-k size waa decreased from the largest break size to about 0.5 ft 2

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• For breaks smaller than 0.1 ft 2 the peak clad temperature and percent metal-water reaction again decreased. These analyses neglect the operation of the HPCIS. Thus, the 0.1 ft 2 break size was found to result in the maximum clad temperature and metal-water reaction for a break in the intermediate size range.

The calculated peak clad temperature for a 0.1 ft 2 break was i620 ° F with a corresponding metal-water reaction of less than 0.12%.

The limiting assumption for this break size was the failure of one diesel generator system to operate. An assumed failure of the LPCIS injection valve resulted in a lower maximum clad temperature of 1530 ° F.

We conclude that *the design of the FitzPatrick emergency core cooling system is acceptable and meets the requirements of the AEC interim acceptance criteria based on analyses using the G.E. evaluation model acceptable to the Commission and described in Appendix A, Part 2 of the Commission's Interim Policy Statement which shows that the consequences of the loss-of-coolant accident are such that (a) the calculated maximum fuel rod cladding temperature does not exceed 2300 ° P, (b) the amount of fuel rod cladding that reacts chemically with water or steam does not exceed 1% of the total amount of cladding in the reactor, (c) the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling, and before the cladding is so embrittled as to fail during or after quenching, and (d) the core temperature is reduced and decay heat is removed for an extended period of time *.

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6-9 6.3 Standby Gas Treatment System

'!be Standby Gas Treatment System (SGTS) whose purpose is des cribed in Section 5.3 consists of two identical, parallel, physically separated air filtration trains . Each has a full capacity design flow rate of 6 ,000 cfm and will exhaust 100% of the reactor building free volume per day while maintaining a negative pressure of 0.25 inches of water. Each train consists of a demister, prefilter, heater, high efficiency particulate absolute (HEPA) filter, activated charcoal adsorber and HEPA after filter. The particulate filters each have an efficiency of at least 99% for particles of 0.3 micron diameter and larger. The activated charcoal adsorber has a mimimum iodine removal efficiency of 99%. At least one of the trains will start automatically upon receipt of a high radiation signal from the reactor building or refueling floor ventilation exhaust monitors. An HPCIS initiation signal automatically starts both trains. A test program will be conducted before reactor operation and periodically thereafter to demonstrate its adequacy.

Based on our review of this and other similar systems, we conclude that the design of the SGTS is acceptable.

6.4 Other Engineered Safety Features An emergency service water system has been designed to provide cooling water to vital equipment required for safe reactor shutdown in the event of a LOCA or loss of offsite power. The system consists

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6-10 of two* independent, redundant loops and is designed to seismic Class I criteria. The pump in each loop takes suction from lake water in the screen pumphouse and is connected to a separate emergency bus. We conclude that the design is acceptable.

As an operational technique to preclude flammable gas concentra tions, the primary containment will be operated with an inert nitrogen atmosphere. The system will maintain the oxygen content of the containment atmosphere below 5 volume percent and we find it acceptable.

Following a loss-of-coolant accident hydrogen gas could be generated and released into the primary containment from a chemical reaction between the fuel rod cladding and steam (metal-water reaction). In addition, both hydrogen and oxygen would be generated as a result of radiolytic decomposition of the recirculating water. If a sufficient amount of hydrogen is generated and oxygen is available in stoichiometric quantities, the subsequent reaction of hydrogen and oxygen can occur at sufficiently rapid rates to create a significant pressure increase in the containment. This could damage the contain ment such that it fails to maintain its low leakage integrity.

General Design Criterion 41 of Appendix A to 10 CFR Part 50 requires that systems to control hydrogen, oxygen, and other substances which may be released into the primary containment be provided as necessary to control their concentrations following*

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6-11 postulated accidents to ensure that containment integrity is maintained. In accordance with the guidelines of the Supplement to Safety Guide 7, "Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant Accident," PASNY has proposed a Containment Atmosphere Dilution (CAD) system.

Basically the CAD concept involves the maintenance of an oxygen deficient (inert) containment atmosphere during the post-LOCA period. ,,

This would be accomplished by addition of nitrogen gas from an .,

external nitrogen makeup and supply system. As nitrogen is added, the containment pressure would rise in the post-LOCA period. However, even assuming a zero containment leakage rate in the post-LOCA period, the containment pressure would not reach the design pressure of 56 psig until 75 days following the accident. Under this condition, containment purging under long-term controlled conditions would be necessary to prevent excess pressure rise and to allow the introduction of nitrogen to maintain the hydrogen-oxygen balance below the flammable limits. The resultant radiological doses would be well within the 10 CFR Part 100 guideline values. If the containment is assumed to leak at a rate of 1.5 volume percent per day the pressure would peak at 33 psig 125 days following the accident. Consequently , the implementa tion of the CAD system as conceived should allow the control of conbustible gases t be accomplished in the post-LOCA period, while not increasing the radiological consequences of the LOCA.

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6-12 All of the final design details and responses to our questions were not received in time to complete the evaluation for this report

• However, the applicant has informed us that the design bases and objectives of his CAD system are the same as those for the Browns Ferry Nuclear Plant which the staff has recently reviewed and found to be satisfactory. The applicant has also stated that the CAD system for the FitzPatrick Plan t will be ready for service at the end of the first normal refueling period.

On the basis of our review of a similar system proposed for Browns Ferry we conclude that the CAD system is a satisfactory method for controlling the containment atmosphere following a LOCA.

Our conclusions concerning its implementation in the FitzPatrick Plant will be finalized when the evaluation is complete.

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7-1 7.0 INST1lUMENTATI0N 1 CONTKOLS AND ELEClRICAL SYSTEMS 7.1 General Our review of instrumentation and controls encompassed the reactor protection and control systems, and the engineered safety r.

l feature ayatems. The Commission's General Design Criteria (GDC) and IEEE Std 279-1968, "Criteria for Protection Systems for Nuclear Power Generating Station*" served as the bases for evaluating the adequacy of these designs.

The evaluation of the FitzPatrick plant was accomplished by comparing its design with that of the previously evaluated Vermont Yankee plant. We have also evaluated the information peculiar to FitzPatrick in the areas of radiation and environmental qualification, protection aystem testability, incident and post-accident monitoring inatrumentation, indication of reactor protection and engineered safety feature bypasses, ATWS, AP:RM reactor trip at 15% power, con .

denser low vacuum trip, and independence of redundant plant protec tin ay*tem channels.

We have reviewed various *chematic diagrams to confirm confor mance with the de*ign and have reviewed the installation at the *ite.

The Colllllliaaion's General Desian Criteria 17 and 18, AEC Safety Guide* 6 and 9, and IEEE Std 308-1971, "Criteria for Cla** IE Electric S y*te1111 for lluclear Power Generatina Sy*te.. " served as the bases for judaina the acceptability of the electrical power *Y*te118.

7-2 7.2 Plant Protection and Control Systems 7.2.1 Comparison of Protection Systems The applicant indicated that the designs of the FitzPatrick reactor protection systems (RPS) and engineered safety features (ESF) are essentially identical to those of Vermont Yankee. Several changes were made to provide more complete circuit separation between redun dant equipment and improve testability. The changes are listed on pages Q.7.3-3 through -5 of the FSAR. We have found that the changes improve safety and make the design consistent with recently approved designs. The remaining portions of the RPS and ESF systems were found to be essentially the same as those of the Vermont Yankee plant and are acceptable.

The applicant has stated that the plant can be shut down and maintained in a safe shutdown condition from outside the control room. 'Dlis capability will be demonstrated during the startup and power test program.

7.2.2 Comparison of Control Systems The applicant has stated that the major control systems for this plant are aenerally identical to the silllilar systeas of the Vermont Yankee plant with the few lllinor diff erences listed on pages Q.7.3-9 through -11 of the FSAR. We have found that these minor differences have not changed the functional deaian nor degraded the safety of the plant. We conclude that the control syste.. are acceptable.

7-3 7.2.3 Protection System Testability The applicant included additional circuitry* and features in the desi gn to permit testing of the plant protection systems during power operation. Our review of this additional circuitry, not included for the Vermont Yankee plant, confirmed that the plant protection system and engineered safety feature system are testable during power opera tion. We conclude that this design is acceptable.

7.2.4 Bypass Indication for Plant Protection System and Engineered Safet7 Feature Equipment The desi gn of the instrumentation and controls for the plant pro*tection system and engineered safety features includes control room indication to identify reduction in system redundancy which could result from operator action. Our review has determi.ned that reasonable annunciation and indication are included at the system level for these redundant safety systems. We conclude that the bypass indication systems are equivalent to those of previously licensed plants and are acceptable.

7.2.5 APRM Reactor Trip at 15% Power The design includes an APRM reactor trip at 15 percent power while operating in the startup mode. Our review has found that this feature satisfies the requirements of IEEE 279 and we find it accept able. This trip has the same function as the IRM trip. However, at le .. t for the time being, the applicant proposes to retain the IBM

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7-4 trip. If, in the future, the applicant proposes to disable the IBM trip, we will require analysis to justify this deletion.

7.2.6 Condenser Low Vacuum Trip The condenser low vacuum reactor trip has been deleted. An j

l additional circuit which closes the main steam isolation valves on low condenser vacuum has been provided to assure that steam flow is l l

j restricted from the main condenser during a leak. This valve closure I

initiates a reactor trip and steam flow is diverted to the torus through the relief valves. Our review has determined that this cir cuitry satisfies the requirements of IEEE-279 and is acceptable.

7.3 Independence of Redundant Plant Protection System Channels Our review of the FSAR revealed that the applicant's criteria are accptable and the elementary diagrams indicated that the criteria are properly implemented. Our visit to the plant site revealed that discrepancies which were identified and corrected on previous BWRs were being corrected. These items are: 1) connection of redundant protection channels to single switches and terminal boards in the control room panels; 2) installation of redundant protection system switche* on control room panels within a few inches of each other and their wiring bundled and routed together, and 3) installation of redundant protection aystem instruments on a colllllOn rack outside the control room.

7-5 We have reviewed the cable installation design, routing and identification criteria relating to the preservation of the indepen dence of redundant channels. We have found that these criteria and I

their implementation are acceptable.

7.4 Incident and Accident Surveillance Instrumentation The applicant has provided a list identifying the redundant instrument channels whose readouts are available to the operator for assessing plant conditions during and subsequent to accident and operational occurrences. Our review has found that the systems are comparable to recently approved BWRs and we conclude that the incident and accident surveillance instrumentation is acceptable.

7.5 Anticipated Transients Without Scram (ATWS)

The applicant stated that provision will be made to include the function of tripping the recirculation pumps. This is described in the General Electric Company report NED0-10349, "Analysis of Anticipated Transients Without Scram," dated March 1971, and has been proposed for the Browns Ferry Plant with one exception. This exception is that the recirculation pumps will be tripped on high reactor pressure only. The report (NED0-10349) proposed tripping the pumps on coincidence of high neutron flux and high reactor pressure. The applicant has indicated that they will propose a concept similar to that of Browns Perry. General Electric has not completed the details of the final design of this aeneric item. We will review the desian prior to installation in the PitzP atrick plant. Howeve no decision has been made to require this improvement on a backfit basis, and we have concluded that it is acceptable to operate the PitzP atrick Plant prior to installation of the pump trip.

7-6 7.6 Radiation and Environmental Qualification 7.6.1 Radiation Qualification The crieria for radiation qualification of safety related devices have been reviewed. The materials used in fabrication were selected to assure that margi n exi sts over the ex pected integrated dose plus the dose resulting from a LOCA. In addition, power and control cable has been tested. We conclude that these criteria and their implementation are acceptable.

7.6.2 Environmental Quali fication The applicant has identified the safety related equipment located inside the containment which must operate during or followi ng a DBA.

The equipment will be capable of functioning under the post accident temperature, pressure and humidity conditions for the time periods required. 'Dle speci fic parameters are listed in Table Q7.3-3 of the 1

I FSAR. 'Ibis capability has been demonstrated by testing of all except the limi.torque valve operators. solenoid valves for the safety/relief valves. and the main steam isolation valve position switches. A test program on these itellB is presently in progress.

All required control room equipment has been demonstrated to be operable over the range of possible environmental conditions.

we*conclude that the environmental quali fication of safety related

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equipment is acceptable pending successful completion of the present test program.

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7-7 7.7 Electric Power Systems 7.7.1 Offsite Power The FitzPatrick Plant will be interconnected to the New York Power Pool transmission grid through 345 kV and 115 kV transmission sys-tems by two 345 kV and two 115 kV transmission lines. A 345 kV-115 kV sectionalized switchyard is provided, each section arranged with a bus connected to its two transmission lines through separate breakers.

The 115 kV bus has, in addition, a bus sectionalizing electrically operated disconnect. The two 345 kV transmission lines are used only for supplying power to the grid. The two 115 kV transmission lines emanate from the station on towers located on separate and independent rights-of-way. One line extends southward to Lighthouse Hill Station and one westward to the Nine Mile Point switchyard which receives power from the Oswego steam station. Two paths, one overhead and one underground, each containing a reserve station service transformer bring power from the switchyard to the plant.

Each of the two 4 kV buses is supplied by one of these pths.

• Either of these separate and independent paths is capable of sup plying minimum accident loads from either of the two incoming lines.

Al.l switchyard breakers have two trip coils for breaker protection, and electrical disconnects have two control circuits. Protective relaying and separate redundant d-c control circuits are provided.

'.i 7-8 ii\ I The 345 kV transmission 1ine to the Nine Mi1e Point switchyard crosses over both incoming 115 kV offsite power transmission 1ines.

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' I The two 115 kV 1ines are under adjacent spans of the 345 kV 1ine such that a broken 345 kV conductor in either span could disab 1e no more than one 115 kV 1ine. If a 345 kV conductor were to become discon nected from its insulator at the tower between the spans, the con ductor would not sag 1ow enough to come in contact with the 115 kV lines Protection against failure of both 115 kV lines due to a 345 kV conductor parting at the insulator between spans is provided by ground wires run above the 115 kV lines. To prevent a falling 345 kV tower from causing 345 kV conductors to disable the 115 kV lines, the applicant has provided an extra strength design for the three 345 kV towers in the vicinity of the 115 kV lines by locating towers closer together, using two independent sets of supporting insulators and using special heavy vertical loading criteria. We have concluded that these measures provide reasonable assurance that offsite power will not be lost as the result of failure of the 345 kV -line.

The applicant has completed transient stability studies that simulated 345 and 115 kV transmission line faults, the loss of the FitzPatrick generaor and the loss of the largest generator on the grid. The results have shown that loss of offsite power would not occur under these conditions.

7-9 Our revi.ew has determined that the applicant's offsite power system is in accordance with General Design Criteria 17 and 18 and is acceptable in all but one area. Our concern results from a Nine Mile Point Station requirement that in the event of a total system blackout, the Lighthouse Hill line be used to supply power from a nearby hydro unit to Nine Mile Point. This line is further required to be isolated from the grid and other loads. To meet this require ment, automatic circuitry is provided to isolate FitzPatrick from offsite power. Criterion 17 would not be met if a single failure could isolate FitzPatrick inadvertently and the offsite power could not be retored within the time required by Criterion 17. We will require either: (1) that the applicant demonstrate that no single failure can disconnect the FitzPatrick Plant from the grid, or (2) that power can be restored following inadvertent isolation in a time period consistent with Criterion 17. The applicant's studies.

of these items will be subndtted by December 1972.

7.7.2 Onsite Power The engineered safety features and safe shutdown loads are*

divided between two independent and separate 4 kV emergency buses, either capable of supplying mi.nimum engineered safety features or safe shutdown equipment. Each of these two buses is capable of receiving power from the normal station service transformer, a re serve station service transformer or a diesel generator unit. There

7-10 are four 600 volt emergency buses, two supplied from one 4 kV emer gency bus and the other two supplied from the other 4 kV emergency bus. Separation and independence of these redundant power systems has been maintained.

Two emergency diesel generator units provide the onsite power supply, one unit for each emergency bus. Each diesel generator unit is started automatically on loss of both normal and reserve a-c power to the plant or low reactor water level or high drywell pressure.

The accident. loads are automatically sequenced on each 4 kV emer gency bus.

Each diesel generator unit and associated auxiliaries are housed in separate seismic Class I structures. Each diesel generator unit 1 l

is a self sustaining entity with its own independent lube oil, fuel l oil, cooling water and control systems. Two diesel fuel oil storage tanks are provided, one for each dieel unit. Each tank has suf ficient fuel for operating its associated diesel unit at full load for seven days.

Each diesel generator unit consists of two diesel generators rated at 2600 kW continuous and 2850 kW for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. Both diesel generators of a unit are started simultaneously, force I

l para1leled and then connected to the emergency bus. A total of 5200 kW is available for each redundant emergency bus* which is more than required to supp1y the safety 1oads during an accident. The

7-11 applicant has successfully completed a rigorous and extensive factory test program on each unit, since this design has not been provided on any previously licensed nuclear plant. The completed tests included: (1) starting, force paralleling and step loading with 5200 kW, fifty times, (2) starting, force paralleling and sequen tially loading, at 10 second intervals, loads of 1000 kVA, 2000 kVA, and a 1250 hp motor, fifty times, and (3) starting, force paralleling and step loading with 2850 kW,*ten times. These same units will be tested in the plant by starting, force paralleling and sequentially loading of the following loads: 1000 kVA, a second 1000 kVA load, and a 1250 hp motor one hundred five times on each unit and starting, force paralleling and sequentially loading all in-plant ECCS loads twenty times on each unit. Expected limits have been established from the results of the factory test program and will be used to determine acceptability of in-plant testing. Lindts*are established on (1) time to synchronous speed, ( 2) time from engine start signal to rated voltage, (3) percent voltage drop, frequency drop, voltag e recovery time and frequency recovery time after motor load appli cation, and (4) load sharing ratio under transient and steady-state conditions. These limits are within those suggested in Safety Guide 9.

Sixty of the one hundred ten factory teats have demonstrated that these units are capable of supplying larger stepped loads* than

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7-12 those to be sequenced in the plant. This has indicated, and will be reaffirmed during in-plant testing, that a large margin exists in these units. The reliability of the diesel generator units will have been demonstrated by the 110 starting and loading tests which have been successfully performed in the factory without a failure i

• and the 250 onsite tests which remain to be performed.

We conclude that with no more than three failures in the remaining two hWldred fifty in-plant tests, the test program will 1

assure that a reliability of 0.99 exists in each unit. On this basis I 1

l we will find the diesel generator Wlits acceptable. i The applicant stated and we have confirmed that the emergency power distribution systems are split in accordance with Safety Guide 6, with the exception of the motor operated valves associated with LPCI and the ADS valves. The power supply to these valves is auto matically transferred between redundant buses. We have determined that this design satisfies the single failure criterion and is con sistent with.previously approved 1967 product line BWR plants. We have concluded that it is acceptable.

The 115 volt a-c systems provided for safety are arranged with two independent reactor protection system buses and two independent engineered safety _feature system instrumentation buses.

'!be 125 volt d-c power supplies consist of two independent 125 volt systems. Each system has a 125 volt battery with its own

7-13 I

charger and distribution panel. The battery charger is capable of keeping the battery fully charged and supplying the d-c system loads simultaneously. Each 125 volt battery is located in a separate room with an independent ventilation system. Each battery is sized to supply essential loads for a period of two hours on loss of its battery charger during any plant operating or incident condition.

This d-c power supply arrangement provides for adequate separation and independence of redundant supplies. We believe the onsite power system satisfies the appropriate General Design Criteria, IEEE 308 and Safety Guides 6 and 9 and conclude that it is accep.table pending successful completion of the in-plant diesel generator test program.

8-1 8.0 AUXILIARY SYSTEMS 8.1 General The systems described in this section. with the exception of the circulating water system, generally are process systems which provide plant aervices auxiliary to the production of power. In the course of our review, we have directed our attention to their safety related objectives and criteria. Major cons iderations included appropriate neas of the seismic design classification and the use of suitable codes, standards and specifications for the design. fabrication. _and inspection of the piping and other components within each- system.

The safety related items which received special attention in the course of the review are discussed in the following sections.

The applicant has made an analysis which determined that failure of any non-seismic Class I system or equipment could not cause flooding of safety related equipment.

8.2- Radioactive Waste Manaaement The app1icant'a design objective* and operational procedures for the liquid radwaate system are to proceas and reuse the 1111jority of the liquid waate* within the plant *o that liquids diacharged from the plant satiafy the 10 CPR Part* 20 and 50 requireaenta aa we11 as thoae of proposed Appendix I to 10 CPR Part 50. The deeian objective for the gaaeous raclwaate eyatem is to utilize the equipaent-ao that a**eoua radioactive vaetea fro* the plant alao aatiefy the require-1'eDCe of 10 CPR Parts 20 and 50 and the propoaed Appendix I to 10 CPll

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8-2 Part 50. The design objective of the solid radwaste system is to package solid waste in accordance with 10 CFR Part 71 and protect plant personnel in accord with 10 CFR Part 20.

8.2.1 Liquid Radwaste System The liquid radwaste system is designed to collect, monitor, process. store and dispose of radioactive liquid wastes. These wastes are classified, collected, and treated as high purity. low purity, chemical, detergent, and sludge wastes. Cross connections between the subsystems provide flexibility for processing by alter nate methods.

High purity, low conductivity liquid wastes 11 be collected in the 30,000 gallon waste co11ector tank from the reactor coolant cleanup system, the residual heat removal system, equipment drains in the drywell, reactor, radwaste and turbine buildings, decantates from resin phase separator tanks, and centrifuge eff1uents. These wastes will be processed by filtration and deep-bed dea:lnera1ization.

After processing they will be sent to the waste saap1e tanks (two at 14,000 gallons each), saapled and analyzed, and if satisfactory for reuae they will-be transferred to the two 200.000 aa1lon condenaate atoraae tanks. If the water doe* not aeet apecification* for reuae.

it will be returned to the ayatea for additional treataent or dia charaed to Lake Ontario via the diacharae canal after dilution with circu1atina and aervice water. We eati1111te that 26.000 aallon* per

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8-3 day wi.11 be processed through this system and that 10% of this water will be discharged and diluted with the 370,000 gpm flow in the dis charge canal. We estimate 4.4 Ci per year wil1 be discharged from thia atream.

Low purity, high condµctivity 1iquid wastes wi11 be co1lected in the floor drain tank (8,500 gal.) from the drywell, reactor, rad waate, and turbine building floor drains, and decant from the waste aludge tank. These wastes wi11 be processed through a floor drain filter and transferred to the floor drain sample tank (two at 17,000 aallons each) for sampling and analysis. If the concentration of radioactive contaminants is sufficiently low, the batch may be sent to the diacharge canal. If the waste needs further treatment, it will be aent to the chemical waste system for evaporation or to the high purity aystem for demi.neralization, depending on its quality.

Our an alyais asawned 5 9 700 gallons per day will be processed through thi* ayatem and sent to the chea:Lcal waste aystem for evaporative treat..nt. We estimate 1 mCi per year will be discharge from this atreaa.

The chea:lcal w.. tea from condensate dellli.neralizer reaeneration aolution*, non-deteraent decontaa:lnation aolutiona. and laboratory draiaa vi.11 be collected and neutralized in the waste neutralizer taoka (tvo at 17,000 aallona each). After neutralization. theae vaae* will be evaporaed by one of the two 20 apa w.. te concentra tor*. The diatillate will be *ent to the vaete collector tank in the

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• and the condensate aent t_o the high purity system for demin eralization and reuse. Prom this system we assumed 10% will be diluted and discharged via the discharge canal. We estimate 1 mCi per year will be discharged from this stream.

The detergent aystem will collect laundry. personnel decontamina tion. and other detergent washdown wastes. These wastes are normally *,

low in radioactivity and after *sampling and an alysis will be dis charaed through a filter. I neceasary. the 0.'8 gpm evaporator will be uaed to concentrate these wastes. Our evaluation assumed all detergent w.. t** will be discharged without treatment and that the radioactivity content will be nealiaible.

The 11 9 000 aallon waste aludae tank will collect waste filter,

£loor drain filter. and fuel-pool filter backwash and sludge. After aettlina and decanting to the 19¥ purity aystea, the sludge will be aent to the centrifua*** The phase aeparator tanks (two at 4,000 aallon* each) will collect backwash froa the reactor water cleanup filter dea:l.neraliaer precoat, decant the vaate to the hiah purity ayat.. and aend the aludae to the centrifuaea. The 3 9 000 aallon

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8-S spent resin tank will collect spent resins from the condensate deaineralizer and waste dem:lneralizer and wi11 send the res:l.ns to the centrifuges. An ultrasonc resin cleaner ¥1.11 be used to clean the condensate deaineralizer during operation and is expected to result in a lower quantity of regeneration solutiqns for processing than is now estimated. Excess water from the centrifuges will be sent to the high purity system and the sludges sent to the dr111 1ng station for solid waste disposal.

In addition to the waste treatment systems described above. the Reactor Water Cleanup Systea continually processea a portion of the coolant flow to maintain high purity reactor water and thus a lower activity in the liquid waste by reaoving radioactive aaterial.a through a Powdex systea. 'lbe Condensate Cleanup Systea proc.....

the full reactor condensate flow through deep bed dealneraliaere to aaintain feedwater purity. The fuel pool coolina and cleanup ayatea aaintaina the water purity by the uae of filter-dea:lneraliaera.

We calculate that about 4.4 Ci per year. excludina tritiua h the liquid radioactive waate ayatea. will be dtacb.arpd. We ..tiaate that 20 Ci per year of tritiua will be diacbarpd b**d on preaent operatina reactor experience. With the dilution flow of 370,000 &PIii*

th* averaae concentration of a:lxed nuclid... excludilla triti1111. du hara*d to the lake would thua be 6 x 10-9 Ci/cc. Dae corr..pondina tritiua concentration would be 2.7 x 10-* Ci/cc. n.... valuea are

,.' .J 8-6 well below the corresponding Maximum Permissible Concentrations (MPCw) 1 1

'j as defined in 10 CFR Part 20. The applicant has calculated that 6.2 Ci/yr, excluding tritium, and 14 Ci/yr of tritium, w ill be released from the liquid radwaste treatment system. The differences between the applicant's release estimates and ours are due to differences in assumptions and calculational techniques. The differences are not considered to be significant. We have used our values because we believe they are based on more reasonable assumptions and models.

We estimate the whole body dose to individuals from the.liquid effluents to be less than 5 mrem/yr. We conclude, on the basis of our evaluation, that the liquid radwaste system is acceptable and should be capable of processing liquid effluents so as to aatis£y the requirements of 10 CFR Parts 20 and 50 and the numerical values listed in proposed Appendix I to 10 CPR Part 50.

The combined liquid effluents from Nine Mile Point, Unit 1, and FitzPatrick have been evaluated in the FitzPatrick Draft Environmental Statement.* Nine Mile Point, Unit 1, has committed to the installation  !

I of auamented effluent treatment *Y*tems which will result in the combined effluent from both plants meetina the intent of 10 CPR Part* 20 and 50.

8.2.2 Gaaeoua Radw-te Sy*t**

The major aource of aaeeou* waste activity durina normal *tation operation will be the off-gas from the ..in a team conden**r air ejec tor*. Other aource* of gaseous va*t* include off-a.. ** fro* the

,oil,

8-7 mechanical vacuum pump, ventilation air released from the radwaste, reactor, and turbine building exhaust systems, and purging of the drywell and suppression chamber. The turbine gland seal system will use a clean source of steam and is not e xpected to be a source of gaseous activity. The source terms and detailed assumptions are discussed in the Commisaion's Draft Environmental Statement and are based upon a failed fuel source of 71,500 Ci/sec after a 30 minute holdup time.

Off-gases from the main condenser air ejector will be processed through catalytic recombiners where the hydrogen and oxygen gases are recoined to form steam, thereby reducing the volume of gases which must be treated. The *team wi11 be condensed and the condensate will be di*charged to the main condenser. The non-condensible gases will be delayed for five hours in a holdup pipe to allow for the decay of abort-lived radioactive noble gases and activation products, fil tered throuah hiah efficiency particulate filters an held up for further decay in an ambient*teaaperature charcoal adaorber **Y*tea con*i*Cina of 17.4 tona of charcoal in twelve beds. Prior to dis chara* throuah the main *tack, the off-a**** are aaain filtered throuah hiah efficiency particulate filters to reaove particulate*

formed by the decay of noble aa*e* and any charcoal fin** which may be carried over in the vent *treaa. We calculate thi* charcoal adaorber *yatea to provide a 7.5-hour decay of krypton and a 4.4-day

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8-8 decay of xenon. It is assumed that essentially all of the radio iodine which may be present in the off-gases from the main condensers wi11 be removed in the charcoal beds. We estimate 200,000 Ci per year of noble gases and 0.3 Ci per year of iodine 131 will be re1eaaed from this source. This estimate includes a 10 day per year down time for the treatment train due to lack of redundancy and during this time the off-gases are held up for 30 minutes only. The applicant has ca1culated 50,000 Ci/yr of noble gases and negligible iodine 131 from this source. The applicant's calculations assumed that the charcoal adsorber system provides a 13 hour1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> decay of krypton and a 240 hour0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> decay of xenon and did not consider a down time factor.

The aechanical vacuum pump, used during startup, exhausts air an d radioactive gases from the main ateam condenser. Off-gases from thia ayatem will be discharged to the gland seal holdup line and re1eaaed to the main stack. We estimate 1700 Ci per year of noble aaa .. will be released from this source. The applicant estimates less than 50 Ci/yr of nob1e gases from this source.

aource teraa rather than the applicant's for both thia and the main We have utilized our j

condenaer offaaa source diacuaaed above because we believe them to be more conaervative and baaed on more reaaonable aeewaptions.

The reactor building, radwaste building, and turbine building ventilation ayeteaa will uae once-through ventilation with air passing rom clean areas to those with higher radioactivity potential. In

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8-9 the event of abnormal air activity le vels, the react or building air will be recirculated and exhausted through the standby gas treatment system (prefilter, HEPA, charcoal adsorber, and HEPA in series) prior to being released to the atmosphere through the 385 foot main stack.

Normally this air will be discharged to the reactor building vent.

We estimate 0.013 Ci per year of iodine 131 will be released from this source and that noble gas release will be negligible. The applicant estimates 1.5 Ci of noble gases and 0.04 Ci of iodine 131.

The radwaste building will be exhausted through prefilters and HEPA filters and released via roof mounted fans. The releases from the radwaate building are expected to be small. The applicant estimates negligible noble gases and 0.009 Ci of iodine 131. Ventilation air from the turbine building will be exhausted to the atmosphere without treatment. We estimate 1200 Ci per year of noble gasee and 0.57 Ci per year of iodine 131 released from the turbine buildina. The applicant eetimates 1800 Ci of noble gases and 0.3 Ci of iodine 131.

We have ueed our source ter1D11 for buildina ventilation rel***** becauae we believe they are baaed on more reasonable aeeumptiona.

The primary containment (drywell) is nolly a eealed volume.

However, durina periods of refueling, operation, and ...intenance, it ie neceeaary to purge the drywell and auppreeeion chaaber and, when thi* occurs, the potential exist* for the releaae of airborne radio activity to the environ.. nt. The *Y*tem will be arranaed auch that he purge exhaust can be directed to the etandby a** treat.. nt *Y*tem

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,) 1 .. . .. . . ..

• L, *

• 8-l.0 if significant radioactivity is present. The rel.eases through this syatem are estimated by the applicant to be 3 Ci/yr of noble aases and 0.09 Ci/yr of iodine 131. We agree with the appl.icant's estimate.

We estimate the average whole body exposure rate due to noble gasea to be less than l.O mrem/yr at the excl.usion radius and that due to iodines to be 11 mrem/yr to the thyroid from consumina milk from the nearest dairy farm. We concl.ude that the gaseous radwaste system is acceptabl.e and ahould be capabl.e of proceasina effluents to aatiafy 10 CPR Parts 20 and SO. We eatimate the direct radiation dose from the pl.ant due to N-16 to be about 0.04 mrera/yr at the site boundary and conclude that this is acceptabl.e. The combined gaseous discharges from FitzPatrick and Nine Mil.a Point. Unit 1. are evaluated in the Draft Environmental. Statement and are expected to meet the intent of 10 CPR Part* 20 and 50 once augmented aaseous effl.uent treatment systems are instal.l.ed at Nine Nile Point. Unit 1.

8.2.3 Solid Radwaate System Solid waates from the plant operation vil.l be coaposed prmarily of spent resins from the various system delllineral.izers. various tank sludges. evaporator concentrates. and precipitates or inaol.ubl.e aat ter that comes from the backwash of filters. After centrifugina they wil.1 be Dd.xed with concrete. pl.aced in drU11111 9 and stored before ship-pina. Dry sol.id waste consiatina of fil.tera and a:lscel.l.aneoua paper and raa* vil.1 be coapacted in druaa and pl.aced in a atoraae area.

• -**:-.-----....--..,. *- ,.,,.1 .... ,_.-.... - . .- ... ---- . .,, ..... ... .. ,* -.. . '*'_,,..,*-,--.. -*-...... - .* ,. ------**-, ............ ... ____ ... *, *'"'-- *** . - .. _.... ,..

-i 8-11 All solid wastes will be packaged and shipped to a licensed burial around in accordance with AEC and Department of Transportation reaulationa.

We estimate an annual disposal of 900 drums of resins. filters.

and evaporators bottoms at an activity concentration of about 3 Ci/

drum. and 600 drums of dry and compacted waste per year with activity leas than 5 Ci/drum. The atoraae and packaging facilities described are aiailar to those previously reviewed. We conclude that the solid radwaate aystem is acceptable.

8.2.4 llaclvaate Syatem Structural Evaluation The radioactive waste treataent ayateaa are deaipled in accordance with acceptable code* and atandarda. All radwaste tanks are in aeiamic Clas* I atructures except the aample tank* which are located in the turbine buildina whoae floor drains ao to the radwaate buildina. We conclude that the buildina*. equipment. and piping deaiana are acceptable.

The applicant ha* evaluated the consequences of failure of out aide tank* contanina activity. Th*** include the two condenaate atoraa* cank* and Che waate aura* tank. The conaequence* of condenaate acoraa* tank failure were diacuaaed in Section 2.3. The reaultina concentration at the drinkina water intake ia well below tba NPC **

defined in 10 Cit Part 20. Th* v.. t* aura* tank 1* aurrounded by a l:

J.:;

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8-12 Claas I dike which will contain the spill and drain it into the waste collector tank for subsequent treatment.

8.3 Process and Area Radiation Monitorina Systems The process radiation monitorina aystem will provide information on radiation levels in certain systems, transport of activity from one ayatem to another, and activity levels released to the environ ment. The monitoring will include: main ateam line; air ejector off gaa; main stack (gas, iodine, and particulate); process liquid; RHR service water; reactor, turbine and radwaste building exhausta; refuelina floor exhaust; and control room ventilation aupply. The aain ateaa line, and reactor-turbine and radwaste building aonitor ayate.. will initiate cloaure of the primary ayatem and building ventilation ayat.... reepectively. upon reachina a preaet radiation level. The reactor buildina unit will additionally atart the standby aa* treat.. nt ayatea and **cure the priaary containaent purae and exhauat. The liquid proc*** aonitor will abut off the diacharae vaive OD a preaet effluent level.

Th* area adiation aonitorina ayatea will provide information on radiation field* in varioua ar*** of the plant for peraonnel protec tion and for *ualitative infon111tion on a ayatea'* condition. auch ..

a failed deaineraliaer. Thi* ayatea conai*t* of 30 channel* and aoni

&ora varioua are.. in th* adlliniatration. turbine. radwaat** and reac tor ltuildina**

8-13 These monitoring systems will detect, indicate, annunciate and/or record the levels or fields of activity to verify compliance with 10 CFR Part 20 and help to maintain radiation levels as low as practicable. We conclude that the plant is adequately provided with process and area monitorina equipment.

8.4 Fuel Handlin& and Storaae Seismic Class I facilities are provided for the ator_aae and tranafer of new and apent fuel. New fuel is atored dry in apaced racks within a reinforced concrete vault designed to preclude I flooding. The rack geometry and dry atoraae aerve to prevent the attainment of criticality. Durina refuelina, the drywall and reactor veaael heads are reaoved and the cavity over the reactor ia filled with water. Spent fuel 1* then tranaferred underwater throuah the refuelina canal to the atainl*** ateel lined concrete apent fuel atoraae pool which ia water filled to provide for ahieldina and heat reaoval. No inlet*. outlec* or draina are provied that aiallt allow the pool to be drained below about 10 feet above th*

top of _the atored fuel.

All with previoua awa plant** we have conaidered Che capability of Che apent fuel atoraa* pool o vithacand an inadvertent drop of the fuel ahippina ca*k ldthout auatainina ...... 1'h1ch could r**ult in a loa* of water 1'hich could uncover the atored fuel. The appli cant baa provided adlllini*trative and ***tan feature- co.al.ailld. .. the

'II:

I

8-14 occurrence and consequences of such an event. The physical arrange ment precludes traversing the cask over any atored spent fuel assam bliea. The access of the loaded 111&in hoist is restricted to the limited fuel cask storage area. This area has the auxiliary protection of an 18 inch thick aluminum honeycomb energy abaorbing pad encased in a stainleas ateel container.

We conclude that the handling and storage p*rovisiona are accept able and that the means for protecting the integrity of he spent fuel atoraae pool are adequate.

8.S Coolina and Service Water Syatell8 The reactor building and turbine building closed loop cooling water ayate.. provide coolina to equipaent in the reactor. turbine and radioactive waate buildina* durina normal operation and aerve

.. an intermediate barrier between the priaary equipaent and the lake water. Each ayat** con*i*t* of 3 puap* and 3 heat exchanaers aQd deliver* deaineralised water to cool the deaianated equipaent it-. ..actor buildina equip-nt whi.ch 11111at function for aafe reactor ahutdown followina accident* or tranaient* 1* eerved by a croaati* with th* ...raency aervic* water ayat.. (a.. Section 6.4).

Th* noraal aervice water ayat.. provide* a continuoua flow of lake water to the aeconclary aide of th* reactor and turbine buildina cloaed loop coolina water beat exchan .. r* and provide* the heat aillk for th*** *Y*t*.. durina nor.. 1 operation.

..., ......-.._ ..._ . .,. ...,...... - .. ,-,Ill'**,* *,_.., . ..... . *' .i-:c_,.,.,,c;-.... ,. . . ... .....,_ ______ .,.__,...__,.,.

8-15 The reaidual heat removal service water system is desianed to se1amic Class I requirements and provides coolina to the RHll heat exchanaers during normal reactor cooldown as well as under post accident conditions. The syste111 can also provide post-accident contaimaent floodina. It consists of two 100% capacity redundant loop*, each powered from a aeparate eaeraency bus.

We conclude that the cooling water systems are adequate.

8.6 Control Room Ventilation Syatem Normal control room ventilation air is drawn from outaide duct*

located near the roof of the administration buildina. Hiah activity level* in the reactor or turbine buildinaa cauae the roof exhaust V9nt* on tbeae buildings to be automatically iaolated, thus el1111inatina a eource of discharae activity which could be drawn into the control room intake ducts.

Hiah radiation levels in the control room air intake duct* or within the control room itaelf activate an alara in the control rooa. Thia alara aiplals a procedure which include* 1111111ual or reaote .cloaure of the intake and exhauat daapera and actuation of one of two redundant filter traina. Bach train conaiata of a pref.1lter, HEPA filter, charcoal filter and aecond HBPA filter and provide* filtered outaide air for breathina, while ..1ntainina a poaitive pre**ure in the control rooa. Th* recirculation ayatea contain* no filter*.

8-16 We have calculated the potential radiation doses to control room personnel following a LOCA and other design basis accidents using suitably pessimistic assumptions. The resulting doses are well within the limits of General Design Criterion No. 19 provided the emergency filter train is actuated within about 10 minutes of the accident. On this basis we conclude that the control rqom ventilation ayatem ia acceptable. j

. ,, 8.7 Circulating Water System The once-throuah condenaer cooling water systea withdraws

.r water from Lake Ontario through a single aector-shaped aubmerged intake atru*cture located about 900 feet offshore. The atructure and tunnel which connect* it to the pumphouse are designed to aeiamic Class I requirements. Electrically heated rack bars on the intake structure prevent ice blockage.

Three circulating water pumps take their auction froa two

• • parate bays in the pumphouse. The* deaian flow rate is 825 cfs, of which 785 cfs are for the aain condenaer and the reaainder for aervice water require.. nta.

The diachara* i* returned to the lake via the diacharae tunnel wih a aix-headed diffuaer located aoae 1100 feet offahore. The

,1 ayatea deaian incorporate* provLaiona for reveraina the flow in the tunnel* auch that the diacharae tunnel becoae* the intake, and vice versa, if the need ahould ever ari***

,),

9-1 9.0 STRUCTURES AND EQUIPMENT 9 .1 Structural Classification The app_licant has classified the plant structures and equipment into two categories dependent upon their relationship to safety.

Structures (e.g., primary containment vessel, reactor building and plant stack) and equipment (e.g., reactor pressure vessel and inter nals, primary coolant system and the emergency core cooling system) whose failure could cause significant release of radioactivity or which are vital to a safe shutdown of the facility and the removal of decay heat have been classified as seismic Class I. Seismic Class II structures and equipment are defined as those which are necessary for station operation but are not essential to a safe shutdown. We have reviewed the applicant's classification of structures and equipment and we conclude that they have been appropriately classified.

9.2 Structural Deaian Criteria The varioua design criteria which were applied to c1 ... I structure* are diacuaaed below.

9 .,2.1 Wind and Tornado Criteria

'!be deaian wind velocity for the *eiaai.c c1... I atructur**

ia 90 ..,h at 30 feet above around b .. ed on a recurrence interval of 100 year*.

Coaponent* which dire ctly affe ct the aafe ahutdawn of the plant (Cl ..

• I) are houaed within atructur** that provide

9-2 protection from the effects of a tornado. These structures include the reactor building below the refueling floor level, the electrical bay portion of the turbine building, the control room, diesel gener ator building and the intake structure. The main 385-foot stack is not specifically designed to resist a tornado but is located more

,i than 400-feet away from any equipment necessary for the safe shut down of the plant. The design tornado for these structures has a 300 mph rotational velocity at the periphery and a translational velocity of 60 mph. 'ftle simultaneous atmospheric pressure drop is 3 psi for a duration of 3 seconds.

ASCE P.aper No. 3269 was utilized to determine the loads resulting from these wind and tornado effects. The load factor associated with the winds is 1.25. Por the tornado loads the structures were designed for a load factor of 1.0. These load factors and allowable stressea are consistent with those used for previously _approved plants. The methods of converting wind and tornado velocities into forces on the structures are in accordance with the state-of-the-art.

9.2.2 Flood Deaian Criteria The facility waa designed to resi*t flood water* to an elevation of 255 feet, an elevation which wa* arrived at conaiderina the flood

.. well .. the effect* of the wind on Laite Ontario, aa deacribed in Section 2. The buoyant fore** c reated by noraal ground water and 1

durina floodina were both conaidered in the de aian of aeiaaic c1 ... I l atructur** with load factors of 1.0. The flood atructural d**ian criteria are acceptable.

I

• '" I

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9-3 9.2.3 Mi.ssile Protection Criteria

'!be tornado generated missiles include a spectrum of possible ite1111 that could.be dislodged during tornadic winds to become mis siles. The applicant's worst case missile, a 40 foot, 12 inch diam eter utility pole at 150 mph, is resisted locally in reinforced concrete structures by allowina the concrete to crush in the area of impact. Structures supporting or housing critical equipment were reviewed to ensure that deflections would not result in loss of equip ment function. The missile protection criteria are acceptable on the basis that they have been used on previous similar plants and repre sent the present state of knowledge in providing a means of damage 9.2.4 Seiaaic Deaian Criteria The seismic deaian for Class I structures is based on dynamlc analyses uaina acceleration response spectrum curves which are noraaliaed to a around eotion of 0.08 g for the OBE and 0.15 g for the DBE.

Modal reaponse spectrum aulti-dearee-of-freedom and noraa1 aocle-ti.. hiatory .. thod* were uaed for the* analyaia of all aeiaalc c1... I *tructure*, *yate.. , and coaponenta. Governina reaponae par ...ter* have been cOlllbined by the square root of the a\911 of the aquar .. to obtain.the aoda1 ..xiaumr when the aodal reapon*e spectrUIII

.. thod w.. uaed. The ab*olute aUlll of re*pon*e* i* uaed for clo*ely apaced frequenciea. Ploor *pectra input* uaed for deeian and teat L

---

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9-4 verification of structures, systems and omponents were generated by semi-empirical methods and confirmedby the normal mode-time history method. The vertical response spectrum acting at the base of the structures was assumed to be 2/3 the horizontal response spectrum and both were considered to act simulatneously.

In addition, a vertical seismic system design analysis was employ ed to account for significant vertical seismic amplifications in the higher elevations of the structures and for the mounted systems and coaponents.

The applicant applied an empirical factor of 1.3 to the peak resonant r*aponae used in the original response spectrum approach for design of equipment. The applicant now advises that he will use a factor of 1.5 for the moat highly stressed components to account for potentially more than one dynamic mode significantly partici pat1na in the peak resonant frequency range. We expect to obtain the re*ults of this analysis prior to issuance of an operating license.

so that we may verify that the resulting stresses remain within the allowable limits described in Section 9.3.

Our aeialllic consultant, Nathan M. Newmark Consulting Engineering Servi ce*, of Urbana Illinoi*, was requested to review and evaluate the aeiaaic desipi criteria eaployed by the applicant with reference to atructurea, ayste.. and coaponenta. The report of our consultant i* included a* Appendix D to this report.

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9-5 We and our consultants conclude that the seismic design criteria and seismic system dynamic methods and procedures proposed and docu mented by the applicant provide an acceptable basis for the seismic design.

9.3 Evaluation of Class I Structures and Equipment The review and evaluation of the seism:1.c Class I structures included the structural foundations, the reactor building, th control room, main stack, diesel generator building, radwas.te building and screenwell-p1.a0phouse building. These structures are founded on competent bedrock.

The structural design stress levels for Class I structures were presented in Table 12.4-3 of the PS.All. For normal loadin119 colllbined with either wind or OBE loads, they are within the AISC code allowable values for structural steel and ACI 318-63 code allowable value* for reinforced concrete. Por loadina colllbination*

that include either DBE, tornado or probable aaxiaua floodina loacla

• atr***** are lia:lted to 901 of yield atrenath for atructural and reinforcina ateel and 751 of Che ulti..te atrenath of concrete.

'ftle analy*** were bued on *l* tic analytical procedure* and th*

d.. ip w* executed uaina the workina atr*** cleaip -thod.

Analya.. and/or aupportina teat* were conducted and uaed in the deaip of **1*1111.c Cl** I -chanical equipaant auch

• fan*, PUIIIIP dri-va* and val,,. operator* to d...,natrate their ability to rea:l.at

... iallic vibratory loadina conditiona. Bquipaant operation durina the

9-6 tests was considered, where appropriate. The applicant will submit for our review, prior to licensing, a summary of the test methods and

' il results of typical classes of equipment to demonstrate that the test

. l

'.. l

,* i

' i

., l levels exceeded the levels predicted by the dynamic system and

,I

I

1 subsystem analyses.

i

'11le basic seismic instrumentation program provided by the applicant corresponds to the recommendations of Safety Guide 12 with respect to 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 he containment structure. The applicant will provide suppo rting instrumentation such as peak deflection recorders for selected locations on Class I structures, systeDB and coaponents in order to provide data for the verification of the seisad.c *responses determined analytically for such Class I iteDB.

As a result of our review and evaluation of the criteria and procedures related to the deaian and construction of the aeiaaic Cl-* I atructurea and equip-nt we conclude that th-* structure*

and equip..nt are acceptable aubject to our review of the additional docuaentation to be furniahed by the applicant, - deacribed herein.

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. .. l 10-1 10.0 ACCIDENT ANALYSIS 10.1 General We have evaluated the applicant's analyses of various anticipated operating transients. The events that characterize abnormal operating transients have been described in Setion 14 of the FSAR and include such events as process system control malfunctions, inadvertent control rod withdrawal, turbine . trip, and variations in operating parameters.

On the basis of this evaluati.on we conclude that the design of the facility, including the protection and control syste118, is such that the occurrence of such transients would not result in damage either to the fuel or to the primary coolant boundary. Consequently, the*

occurrence of these abnormal transients would not lead to a sianificant release of fission products to the environs.

We alao have evaluated a broad spectrum of accidents that mial'at reault froa .. 1operation or failures of equipment. We have selected four hiahly unlikely, but etandard, design baais accidents that envelop the apectrum of type*, cauaea and locations and involve the varioua enaineered aafety ayate ... The calculated potential conaequenc.. of the deaip b-i* accidents exceed thoae of all other accident* conaidered. We performed conaervative analyaea of th-*

d*ian buia accident* to ..*e** the adequacy of the enaineered aafety feature* to control and a:lniaise the poaaible **cape of fiaaion product* froa the facility. The d**ian ba*i* accident*

r J

10-2 analyzed were: (1) loss-of-coolant, (2) refue ling, (3) control-rod-drop, and (4) steam-line-break.

Our evaluation of these postulated accidents yields calculated resulting doses which are well within the 10 CFR Part 100 guideline values. The results of our analyses, and the models and assumptions used in each case, are described in the following sections.

10.2 Loss-of-Coolant Accident In calculating the potential consequences of the postulated loss-of-cooiant accident, to provide a conservative assessment, we have assumed that in spite of emergency core cooling system operation, large amounts of fission products would be released from the reactor fuel. The fractions of the total core fission product inventory we assumed to be released from the core are those aiven in AEC Safety Guide 3, i.e., 100% of the noble gases, and 50% of the halogens. In additon, 50% of the halogens released froa the core are aaauaed to plate out onto internal surfaces of the containment building or onto internal coaponents. Of the reaainina 251 of haloaen* aasuaed available for leakage, we used Safety Guide 3 aaauaption* of 87%, 8%, and 5% for the elemental, oraanic and particulate foraa, reapectively. The primary contain

-nt *- -*uaed to leak at a conatant rate of 0.5 percent*. of it*

voluae per day at accident condition* for the 30 day accident duration without conaideration of the lllitiaatina effects of decre-ina pr.. *ure durina the poat-accident interval.

10-3 We have assumed a 90% halogen removal efficiency for the elemental and particulate forms of iodine, and 70% for the organic forms of iodine in the HEPA filters and charcoal adsorbers of the standby *gas treataent system (SGTS) in the secondary containment building. In our analysis, we adopted the conservative assumption that leakage from the drywell goes directly to the *standby gas treatment system without mixina in the reactor building and then passes through the SG'l'S to the environment via the 117 meter stack.

The calculated two hour doses at the nearest site boundary (97S aeters) are 36 rem and 2 re* to the thyroid and whole body, respectively. The stack heiaht was 111Ddified by terrain heipt at this location and fUllliaation conditions ere assuaed to exist durina the first four hours of the post-accident release period.

In calculating the 30 day duration dose at the 3.4 aile low population zone radiua an effective stack heiaht of SO meter* was uaed to account for the riae in terrain. The calculated thyroid *J11 t*

1 and whole body doaea are 60 rea and 2 rea, reapectively, for theae conditiona. In all ca*** the diffuaion factor* uaed*in our calcula tion* are b.. ed on onaite aeteoroloaical .... ureaent* which are diacuaaed in Section 2.S.

10.3 ..fuelina Accident In evaluatina the poatulated refuelina accident, we* .. auaed that durina fual handlina operation*, a fuel bundle fall* with aufficient force to daaaae (perforate the clacldina) 111 fuel roda 24 houra after

10-4 reactor shutdow n. We also assume that 10% of the noble gases and 10% of the halogens from the damaged rods are released to the spent fuel storage pool water. Ninety-nine percent of the halogens thus released from the perforated fuel rods are assumed to remain in the pool water. Halogens released from the pool water are assumed to be 25% organic and 75% elemental. The airborne fission products within the building are assumed to pass through the standby gas treatment system (with a charcoal adsorber iodine removal efficiency of 90% for elemental forms and 70% for organic forms) and be dis charged from the 117 meter stack over a two-hour period. 'nle meteorological conditions assumed are those for the 0-2 hour period following the loss-of-coolant accident.

The calculated two-hour radiation exposures at the nearest site boundary are 2 rem to the thyroid and leas than 1 rem whole body.

10.4 Control Bod Drop Accident The postulated control rod drop accident aasuaea that a bottom entr.y control rod has been fully inserted and becomes stuck in this poeition unbeknown* to the operator. The drive is then aaauaed to becoae uncoupled and withdrawn from the rod. The rod subsequently falls from the core, inserting an aaount of reactivity correapondina to its worth.

In evaluatina the radiological conaequence* of thi* accident.

we have ..de aaauaptiona b.. ed upon the applicant'* analytical aodel aa preaented in the PSAR. Aa diacuaaed in the aubaequent paraaraph**

the analyaia techniques for thi* particular accident have been

10-5 revised by General Electric and are under review by the AEC staff.

As the result of these analyses, we will likely require modifications, in addition to the safety features presently provided, to mitigate the potential consequences .

Hot standby, zero power is the worst operating condition at which this accident could occur because it results in a high energy release and a direct re-lease path for unfiltered fission products could exist through the mechanical vacuum pump at the condenser.

In the applicant's analysis the most reactive control rod, with reactivity worth of 2.5% k, is assumed to drop out of the core during atartup which occurs 30 mi.nutea after ahutdown from full p ower operation, causing 330 fuel rods to exceed a calculated energy input of 170 cal/gm. These rods are assumed to perforate, releasing 100% of the contained nob1e gases and 50% of the contained halogena to the reactor coo1ant aystem. Of the halogens *o released, 90% are aesumed to be retained in the primary aystem and half of the remainder 1* reaoved by p1ate-out. Thus, all of the noble gases and 25% of the haloaeaa in th* affected rods are assumed to be available for release.

Detection of a hiah radiation *1811*1 in the main *teaaline* autoaatically clo* .. the main-ateallllin* isolation valve*, *huts down the aechanical vacuua puap and clo*es the isolation valve down*tream of the puap.

The actiity entrained in the conden*er is aaau.. d to be released at around level froa the turbine buildina by leakage froa the condenaer

10-6 at the rate of 0.5% of the condenser volume per day. -The duration of the release is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and dilution in the turbine building wake is assumed.

The calculated two-hour doses at the nearest site boundary (exclusion area) are 1 rem thyroid and less than 1 rem whole body.

The 24-hour doses at the low population zone are calcu1ated as less than 1 rem to the thyroid and whole body. The meteorological conditions assumed in a11 cases are those discussed in Section 2.5.

The Atomic Energy Commission has for some time utilized Brookhaven National Laboratory (BNL) as its consultant as part of the regulatory assistance program. Por the past few months, personnel of BNL have been performing independent calculations of boiling water reactor control rod worths and potential consequences of a desian basis control rod drop accident. On the basis of the work performed to date at BNL,* it was determined that the model used by General Electric to evaluate the design baaJ.s control rod drop accident required revisions. Specifically, the assumed rate of negative reactivity.insertion from control rod scram was not suitably con*ervative since it used insertion characteristics now considered to be not readily attainable in large boil_ing water* reactors. In addition, the actual reactivity insertion rates are not lnear as

• - -*uaed.

The General Electric Coapany h- now revised the analysis of the effect:* of a control rod drop accident and has aublllitted two topical

• BIIL 16717-Dl021, "llod Drop and Scraa in Boilina Water lleactora,"

dated April 1972.

10-7 reports* to the regulatory staff. The analysis presented in the main report applies to those reactor plants using control curtains in the core for initial reactivity control. The supplement presents a similar analysis for plants such as FitzPatrick which use gadolinia poison in the fuel. The regulatory staff, with the assistance of BNL, is current1y evaluating the adequacy of the revised model and the resultant consequences of this postulated accident. Included in the revised analyses are, among other features, a chanae in the method for modeling the rate of neaative reactivity insertion froa a contro1 rod scraa.

We have concluded that modifications are required to auapaent the Rod Worth Minimizer, as described in Section 3.4, to ..ke the occur rence probability of this postulated accident nealiaibly saall. 'ftlis is required through the ranae of reactor operation where the conse quences of the accident would be unacceptable by cauaina a peak enthalpy deposition in the fuel areater than the lower threshold value (about 280 calories/am) at which rapid fuel diaperaal and damaaina pressure pulses to the reactor might occur.

The applicant baa propoaed a Rod Sequence Control Syatea (llSCS) aa a backup to the llod Worth Minia:lzer. Thia ia the *- ayatea that baa been proposed for the Browns Pery Plant and which is presently

• NE1ii-10S27, 11Rod Drop Accident Analyai.a for Larae Boilina Water Reactors," dated March 1972, and NBIJO 10527, Supple-nt 1, "llod Drop Analysis for Larae BW1la," dated July 1972.

10-8

,1nder review by the staff. Its acceptabi1ity will be determined when our evaluation is comple*ted.

Although the dose consequences wil1 be approximately doubled using the revised accident model, the calculated doses at the site boundary and LPZ will still be we11 below 10 CFR Part 100 guidelines.

10.5 Main Steamline Break Accident The break of a main steamline outside of the drywell represents a potential escape route for reactor coolant from the vessel to the atmosphere without passage through the standby gas treatment system.

This escape route wou1d exist only for the few seconds required

,c

• {

for the iso1ation control instrumentation to sense the break and close the main steam1ine isolation valves.

The occurrence of a main steamline break outside the containment f would be sensed by either high steam f1ow or increased temperature in the steamline tunnel area. The steamline iso1ation valves would start to close within 0.5 seconds after the steamline break*is sensed.

The app1icant has provided analyses to show that fuel rod c1adding l perforations would not occur as a result 0£ a steamline break if the isolation valve c1osure times, including instrument delay, are less than 10.5 seconds. To provide additiona1 margin to asau that cladding perforations will not occur during the transient before the valves are closed and to reduce the amount of radioactivity releaaed, the Technical Specifications will require a valve closure ti- of not t.

greater than 5 seconds.

• *4. , * - . * ,. * * ,_ .

10-9 The meteorological considerations and other conditions as sumed for this accident are those given in AEC Safety Guide 5, "Assumptions Used For Evaluating The Potentia1 Radiological Conse quences of a Steam Line Break Accident For Boiling Water Reactors."

The consequence of this accident will be limited by virtue of the Technica1 Specifications which will prohibit a reactor coolant iodine fission product specific activity in excess of 20 m.icrocuries per cubic centimeter. In our analysis, the re1eased mass of liqtdd primary coolant (120,000 lbs i.n 10.5 seconds) is assumed to have this activity even though the normal operating level will be about an order of magnitude below this. The Technical Specifications will also require an isolation valve closure time considerably less than 10.5 seconds. The calculated 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exposure dose at the nearest site boundary is 33 rem thyroid. The whole body dose is estimated to be less than 1 rem.

10.6 Conclusion On the basis of our evaluation, the calculated radiological doses that might result from any of the potential design basis accidents are well within the guidelines given in 10 CPR Part 100.

11-1 11.0 CONDUCT OF OPERATIONS 11.1 Station Organization and Staff Qualifications The James A. FitzPatrick Nuclear Power Plant (JAFNNP) organization will consist of approximately 120 full-time employees of NMPC. Person

• ' nel wil1 be shared with Nine Mile Point Nuclear Unit 1 except for a Plant Superintendent, Maintenance Foreman, Assistant Instrument and Contro1 Supervisor, Assistant Radiochemistry and Radiation Protection Supeaor and operating shift personnel. The plant is under the ultiaate supervision of the Genera1 Superintendent who reports to the Vice President of Operations of the Niagara Mohawk Power Corporation.

The site sta£f consists of a maintenance group under the supervi*ion of a Maintenance Supervisor, a results group under the supeaion of a llesults Supervisor and a Plant Superintendent for each plant (Nine Mile Point and FitzPatrick) reportina to the 0.Jiera1 Superintendent. The Plant Superintendent is in charge of the functiona1 operation of his respective unit, and is responsible for the **f* and efficient operation 0£ hi* unit. The nor1111l shift coapla..nt for the 3Al'NPP will be 5 aen and there will be 4 such coapla..nt** Thi* conai*t* of a Plant Shift Supervisor (SRO), a Shift Operatina Poreaan (RO), two Nuclear Auxiliary Operators, one of whoa 1* a liceneed *actor Operator, and an Auxiliary Operator.

The *hift *taffina **ti*fi.. the auideline* of the ABC'* propoeed t i ,,ttfiW'atttWtt&i:it @if Y"i "" --** t Hhfl '. * , 71 tt'e-*wrtnMttf1fI5 r' l'ffi'SMH"trtrt-W..T/t1 *-:ti vwrrwt ttttt'f *. titt@Odf ro&d i* *1W:"H*¢nl-*itnilel

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11-2 Operati onal Safety Guide No. 11, dated July 1969. The minimum crew size will be stated in the Technical Specifications. The applicant will have a Senior Reactor Operator at the plant at all times from fuel loading to the end of plant life.

The Maintenance Supervisor is responsible for the mechanical and electrical maintenance, excluding instrumentation, f or both plants.

The Results Supervisor is responsible for the proper execution of technical work in both plants. He has \Dlder his direction a Reactor Analyst Supervisor, a Radiation Chemistry and Radiation Protection Supervisor, and an Instrument and Controls Supervisor. The qualifica tions of the key supervisory and staff personnel with regard to educational background, experience and technical specialties have been reviewed and are in accordance with those defined in ANSI N 18.1-1971 "Selection and Training of Nuclear Power Plant Personnel". The Technical Specifications will require a confirmed collllllitment to that standard or the equivalent. The training program has been conducted by NMPC, GE and others. The majority of key supervisory personnel have had operating experience in comparable positions at the Nine Mile Station. Members of the technical 1roupa have received trai.ning specifically oriented to their assianed responsibilities.

Technical support for the plant staff will be provided by the Niagara Mohawk Power Corporation and the Power Authority of the State of New York.

11-3 We have concluded that the organizational structure, the training, and qualifications of the staff for the James A. FitzPatrick Nuclear Power Plant are adequate to provide an acceptabl e operating staff and technical support for the safe operation of the facility.

11.2 Test and Startup Program Test procedures will be reviewed and approved by a Joint Test Group composed of representatives of NMPC, GE, PASNY, and Stone and Webster. The Plant Superintendent is chairman of this Joint Test Group. Test procedures will also be approved by the Site Operations Review Conmittee. Test results will be approved by the Joint Test Group and the Site Operations Review Comnittee. Copies- of approved tests will then be forwarded to the Safety Review and Audit Board which will assure that the test results are in conformance with the scope and acceptance criter_ia for each test.

Technical support for initial startup and testing will be supplied from. the Nine Mile Point Nuclear Station and from General Electric. The General Electric Coapany will assign a site aanaaer and a supervisor on each shift. In addition, GE will aasian technical personnel as required.

We have reviewed the applicant's preoperational and startup testina proaram and conclude that it is aenerally in accordance with the ABC publication*, "Guide for the Plannina of Preoperational I

• • 1*i.. iri' J..

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11-4 Testi.ng Prograns" and "Guide for the Planning of Initi.al Startup."

The program will provide an adequate basis to confirm the safe operation of the plant and is therefore acceptable.

11.3 Plant Procedures Plant operations are to b performed in accordance with written and approved operating and emergency procedures. Areas covered i.nclude normal startup. operation and shutdown. abnormal conditions and emergencies, refueli.ng 9 intenance, surveillance and testing, and radiation control. All procedures and changes thereto will be reviewed by the Site Operations Review Coamd.ttee prior to

'l i.11pleaentation. 1

• f We conclude that the provisions for preparation, review, approval, and uae of written procedues are satisfactory.

11.4 Safety llevi.ew and Audit The safety review and -audit f"lction for the James A.* FitzPatrick Nuclear Power Plant will be conducted by a Site Operati.ons Review 1 ttee and the Safety a.view and Audit Board. These two bodies ar* a1reacly in existence and functioning for the Ni.ne Mile Point Nuclear Station. The Si.te Operations lleview Colali.ttee ia advisory to the General Superintendent and will review all propoaed tests.

chan..

• in plant operating procedure* and deaian aodifi.cationa. The Saety llavt.ew and.Audit Board ia adviaory to the Vice Preaident Qf Operation* and Vice Pre*ident-Chief Engineer of NMPC and provide*

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• 11-5 corporate manapaent with a review and audit capability to verify that organizational checs and ba1ances are functioning to assure continued safe operation and desi.p adequacy of the plant. The Principa1 Nuc1ear Engineer of PASNY i.s a ae111ber of the Safety Beview and Audit Board. The quorum. aeeting frequency, responai.bi.1ities

• and authorities of both coaaitteea wi.11 be de1ineated in the Techni*ca1 Speci.fications. Minutes of coaa:l.ttee aeetinp w i.11 be docuaented. We. conc1ude that the revi.ew and audit structure proposed by the appli.cant is acceptab1e.

11.s Eaeraency P1anni.na The app1i.cant has eatabli.shed an organization for copi.ng with eaeraencies. Zt inc1udea written agreeaents 9 1iaison and communica tion with appropriate local, State and Pedera1 agencies which have reaponaibilit:ies for coping with emergencies. The applicant has defined categories of incidents inc1udng criteria for the notifica tion and participation of off-site support: groups and for determining when protective -.. urea ahould be conaidered. Arrangeaent:a have been ...s. by the applicant: to provf.de for aedical aupport in the ent of a radio1oaica1 or other eaergency. Provisions for periodic t:rainina for both p1ant: personnel and off-site eaeraency oraaniza tiona have b ..n 1.iac1uded 1.n the app11.cant:'* *-raency plan.

We have rev1.a,ed the *-raency p1an and conclude that 1. t: -eta the cr1.t:eria of Append1.x B to 10 Cl'll Part 50 and that adequate

11-6 arrangements have been made to cope with the possible consequences of accidents at the site. Regulatory Operations is reviewing the detailed procedures for implementation of the plan. This includes verification and e valuation of all arrangements which have been made with individuals, groups, facilities and government bodies.

11.6 Industrial Security

'nle applicant has submitted an industrial security plan which describes his provisions for the protection of the plant from industrial sabotage. These provisions include control of access to the facility, personnel selection and review procedures, and means for the monitoring of vital equipment. We have reviewed the indus trial security plan and conclude that it satisfies the intent of Safety Guide No. 17 and provides an appropriate program for protection aaainst industrial sabotage.

+. .. .', ' -

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12-1 12.0 QUALITY ASSURANCE The app1icant has desc ribed the Quality Assurance (QA) program for the FitzPatrick P1ant in Appendix D of the FSAR. Supp1ements 6 and 7 to the FSAR provide additional information in response to staff questions. This program has been modified to satisfy the criteria established in Appendix B of 10 CFR Part SO, which was issued after construction was begun.

PASNY maintains the u1timate responsibility for implementing and execut.ing the Quality Assurance Program during all phases of construction and operation. During the construction and testing phases, Stone & Webster was delegated the major responsibility for carrying out the day-to-day QA activities as presc ribed by PASNY.

Niaaara Mohawk Power Corporation, as operators of the p1ant, wi11 have the major responsibility for carrying out qua1ity contro1 activities durina plant operation.

PASNY's lteaident Engineer, - a member of the Site Operations Review Colllllittee, wil1 audit procedura1 and operationa1 activities and aaintain poeeeaeion of all re1evant records.

Various inapectiona conducted durina conatruction revea1ed certain inadequacies in *the QA proaraa and areaa of noncoap1iance with ABC reau1ationa and the l'SAR. Th-* have all b..n reaol,,.d.

We conclude, therefore, that the Quality Aaaurance Proaraa, when properly iaple..nted by the applicant and verified by the Directorate of ltasuJ.atory Operationa, will provi.de reaaonable aaaurance of adequate and acceptable quality.

. . *- *_*;, .: .-, . . - . * .: . . . C * )*.,; :., ) '-. * * *

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13-1 13.0 TECHNICAL SPECIFICATIONS The Technica l Specifications in the license define certa in features. characteristics. and conditions governing operation of the facility that cannot be changed without prior approva l of the AEC. We have reviewed the proposed Technical Specifications for the plant in deta il and have held meetings with the applicants*

to discuss their contents. Modifications to the propo sed Technical Specifications have been suggested by the Staff to more clearly describe the allowed conditions for plant operation.

The fina lly approved Technica l Specifications for the plant 111.11 be included a a part of the operating license. They will a ddress safety lilllits and liaiting safety system settings, lillliting conditions for operation. surveillance requirements. design features. an d adw:lnistrative controle. On the basis of our review. we conclude that normal plant .operation within the lilllits of the Technica l Specificationa will not result in potentia l offsite exposures in exc..

• of 10 Cft Part 20 lilld.ta. Purtheraore. the lim:1.tiog conditions of operation and aurveillance requirell8nt* will aaaure that necessary enat.neered aafety feature* for continued operation or safe shutdown will be available in the event of ..1functiona within the plant.

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14-1 14.0 CONFOBMANCB WZ'l11 GENERAL DESIGN CRITERIA Based on our evaluation 0£ the design and design criteria for the Jaaea A. PitzPatrick Nuclear Power Plant, we conclude that there is reasonable assurance that the intent 0£ the General Desian Criteria £or Nuclear Power Plants, published in the Federal Register on May 21, 1971 as Appendix A to 10 CPR. Part 50 9 will be met.

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lS-1 15.0 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The report of the ACRS on this project will be placed 1.n the Comission' s Public Document Boom and will be published in a supple ment to this report. It can be discussed by the regulatory staff at the public hearing in the event such a public hearing is required.

• • -- *,., ** .:..,'c ' *

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• 16-1 16.0 COMMON DEFENSE AND SECURITY The app1icants state that the activities to be conducted wi11 be within the jurisdiction of the United States and that all of their directors and principal officers are United States citizens. PASNY is a public benefit corporation constituting a pQ1itical sub-division of the State of New York. NMPC is incorporated under the la,s of the State of New York and renders electric service to the public.in a portion of that state. We find nothing in the application to suggest that either PASNY or NMPC is owned, controlled, or dominated by an alien, a _foreip corporation or a foreign government. The activities .to be conducted do not involve any restricted data, but the applicants have agreed to safeguard any such data which llli(lht become involved in accordance with the requirements of 10 CPR Pa rt SO.

The applicants will obtain fuel as it is needed' froa sources of s'1J)ply available for civilian purposes, so that no diversion of specia1 nuclear material froa a:llitary purposes is involved. l'or these reasons, and in the absence of any information to the contrary, we have found that the activities to be performed will not be inillllcal to the co on defense and security.

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17-1 1

17.0 PINANCIAL QUALIFICATIONS We have reviewed the financial information presented in the application. The funds necessary to meet operating costs of the facility will come from operating revenues of PASNY, as more fully set forth in its application. Information contained in the application indicates that such revenues will be ample to cover the estimated cost of operating this reactor as well as the safe decommissioning of the unit if such should become necessary.

We conclude that the applicants are financially qualified to*

enaaae in the activities which would be authorized by the operating license. Our detailed evaluation of the applicants' financial q-ualifications is pre*ented in the attached Appendix E.

17.1 Financial Protection and Indemnity Requirements P ursuant to the financial protection and indeamification provisions of the Atomic Energy Act of 1954, as amended (Section 170 and related sections), the Co dssion has issued reaulations in 10 en. Part 140. 'lbese reaulations set for*th the Coaa:lasion's require11ents with regard to proof of financial protection by,*and inde11111ification of, liceneee* for facilitie* such as power reactors licenaed under 10 CPll Part 50. \

Preoperacional scoraae of Nuclear Puel taaion'* regulations in Part 140 require that* each hlder of a conacruction permit under 10 en. Part SO, who 1* aleo to be the holder of a licenae under 10 Clll *Pare 70 auchorisina the ownerahip

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' l

17-2 and poaseasion for atorage only of special nuclear material at the reactor conatruction site for future use as fuel in the reactor (after iasuance of an operating licenae under-10 CFll Part 50), shall, J

. j

• . J durina the interim storage period prior to licensed operation, have

.l and maintain financial protection in the amount of $1,000,000 and

. I execuce an inde11111ity aareement with the Coamdssion. The applicants

f have provided proof of financial protection in the form of a Nuclear i Ener8Y Liability Insurance Aaaociation Policy, have paid the required I

' fee and have executed Inde11111ity Aareement B-63 with the Collllllisaion

• Special Nuclear Material Licenae No. SNM-1311 has been issued to the applicants and fuel 1* presently beina atored at the aite.

17.3 Operatina Licenae Under the Co iasion'* reaulationa, 10 CPR Part 140, a licenae auchoriaina the operation of a reactor ..Y not be isaued until proof of financial protection in the aaount required for auch operation baa been furni*h*d, and an indeanity aare... nt, or ...ndaent thereto, coveriaa auch operation ha* been executed. The amount of financial proteccion which 111Uat be ..intained for reactor* which have a rated ca*acicy in exc*** of 10 5 kW i* $15 1111.llion, which ia the aaximua allOUDt available froa the coabined private aourcea of the two nuclear liability inauranc* poola.

Accordinaly, no lic*n** authoriaina operation of th* 3 ...

• A.

Wit*Pacrick Nuclear Power Plant will be iaauad until proof of financial

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17-3 protection ht the requisite amount has been received and the requisite inde11nity aareeaent or amendment executed.

We expect that. in accordance with the usual procedure. the nuclear liability insurance pools will provide. in advance of anticipated issuance of the operating liceuse docdllent 9 evidence in writin& 9 on behalf of the applicants. that the present coveraae has been appro priately amended and that the policy lilllits have been increased tQ an amount that meets the requirements of the COIIIDiasion's reaulations for reactor operation.

17.4 Conclusion on the basis of the above conaiderationa. we conclude that the presently applicable requirements of 10 CPR Part 140 have been satisfied and that. prior to issuance of any operatina license. the applicant* will be required to coaply with the provisions of 10 CPR Part 140 applicable to operatina licenses. inc1udina thoae ae to proof of financial protection in the requisite aaount and to execution of an appropriate inde1111ity aareeaent or aaendlllent thereto with the COlllllieeion.

18-1 18.0. OONCLUSIONS .

1

• Based on our evaluation of the application as set forth above, and assuming favorable resolution of outstanding unresolved matters as discussed in Sections 4.2, 4.4, 5.2.4, 6.4, 7.6.2, 7.7.1, 7.7.2, 9.2.4, and 10.4, we have concluded that:

1. The application for an operating license filed by the Power Authority of the State of New York and the Niagara Mohawk Power Corporation, as amended, complies with the requirements of the Atomic Energy Act of 1954, as amended (Act), and the Comadssion's regulations set forth in 10 CFR Chapter l; and

2.

• Construction of the facility has been substantia11y completed, and there is reasonable assurance that it will be completed, in conformity with Construction Permit No . CPPR-71 and the application as amended, the provisions of the Atomic Energy Act of 1954, as aaended, and the rules and regulations of the Coadssion; and

3. The facility wil1 operate in conformity with the app1ication as aaended, the provisions of the Act, and the ru1es and regulations of the Coaaission; and

4. There is reasonable aasurance (i) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public. and (ii) that such activities I

I will be conducted in coapliance with the regul*ations of the Co d*sion set forth in 10 Cl'll Chapter 1; and I

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18-2

5. The applicants are technically and financially qualified to engage in the activities authorized by an operating license in accordance with the regulations of the Commission set forth in 10 CFR Chapter 1; and

6. The issuance of an operating license will not be inimical to the co11111on defense and security or to the health and safety of the public.

Prior to final consideration of the matter of issuance of an operating license to the applicants for the James A. FitzPatrick Nuclear Power Plant, the Collllli.ssion's Directorate of Regulatory Operations will prepare a supplement to this Safety Evaluation which will deal with those matters relating to the status of construction completion and conformance of that construction to the construction permit, the application, the Act, and the rules and regulations of the CoDlllission. Such completeness of construction as is required for safe operation at the authorized power level must be verified by the Co11111ission's Directorate of Regulatory Operations prior to issuance of a license. Further, before an operating license is issued, the applicants will be required to satisfy the applicable provisions of 10 CPR Part 140.

.. .-, . .....,-) ...... --- --.-""' "........_.....

A-1 APPENDDC A CHRONOLOGY REGULATORY REVIEW OF THE POWER AUl'RORITY OF THE STATE OF NEW YORK*

JAMES A. FITZPATRICK NUCLEAR POWER PLANT May 21, 1971 PASNY submi.ts Environmental Report. 1 '

June 4, 1971 PASNY submi.ts amended and substituted license application and volumes 1-5 of the Final Safety Analysis Report.

June 24, 1971 Meeting with PASNY to discuss revised pipe whip protection criteria.

June 25, 1971 AEC letter requests additional information not included in FSAR.

September 14, 1971 Meeting with PASNY to discuss reliability and acceptance testing of onsite power system *.

September 17, 1971 PASNY submits Amendment No. 1 together with Supplement No. 1 to FSAR containing responses to the AEC letter of 6/25/71, updated information, and proposed Technica Specifications.

October 6, 1971 AEC letter requests definition of detailed est program to justify the overall adequacy and reliability of the onsite emergency power system.

October 12, 1971 ABC letter requests information relative to the requirements of Safety Guide No. 7 .

,l

• \ on control of combustible gases following

'j I a LOCA.

I

• Power Authority of the State of New York and Niagara Mohawk Power Corporation are jont applicants. Reference to PASIIY herein implies both applicants.

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A-2 October 19, 1971 PASNY submits show cause information.

October 22, 1971 Meeting with PASNY for further discus sions on pipe whip protection and onsite power system test program.

October 28*, 1971 PASNY submits additional show cause information providing further clarification.

Ncvember 8, 1971 PASNY submits Amendment No. 2 together with Supplement No. 2 to the FSAR containing revised information.

November 19. 1971 PASNY submits further show cause information on transmission lines.

November 22 9 1971 PASNY submits Supplement No. 1 to Environmental Report pursuant o 10 CFR SO Appendix D.

November 23 9 1971 AEC publishes Show Cause Determination and Discussion and Findings not to suspend construction of the FitzPatrick Plant pending completion of the HEPA review.

November 29 9 1971 AEC letter requests additional information on PSAR ite1DS November 30 and December 1 9 1971 Meeting with PASNY to review several aspects of the PSAR and discuss pre liminary comments on the proposed Technical Specifications.

December 2. 1971 U.S. Departaent of Interior Fish &

Wildlife Service submits co11111ents on PSAR which are foxvarded to PASNY.

December 13, 1971 PASRY submits Alllendment No. 3 together wf.th Supplement No. 3 to PSAR containing reviaed and updated infoanaion.

Dec..ber 21 and 22. 1971 Meeting with PASRY for prelilllinary diecuaeions on control. inatrwaentation.

and-electrical ayat... aapecte.

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A-3 January 12, 1972 ABC letter requests additional infor mation on FSAR items. ,!

February 1, 1972 Meeting with PASRY to discuss hydro logical analysis models used in PSAR.

February 7, 1972 ABC letter requests additional information and detail on emergency plans.

February 9, 1972 PASNY submits Amendment Ro. 4 together with Supplement Ro. 4 to PSAll contaf.Ding revised and supplementary inforaation and a partial response to the ABC letter of 11/29/71.

February 10 & 15, 1972 Meeting with PASRY to discuss responaea

• t.

.'! l to open items froa prior aeetinga and to discuss operator training.

February 14, 1972* PASNY subaits response to U.S.

Department of Interior Pish

• Wi1d1ife Service c0111111ents and reco11111endationa.

February 18, 1972 ABC letter requests additional informa tion on confonaance to requir...nta of 10 CPR 50.55(a) Codee and Standarda.

March 9, 1972 lleeting with PASRY for futher diacuaaiona on onaite power ayat.. teat_ proar ...

March. 10, 1972 PASNY aublllita Aaend..nt No. 5 together vith Suppl...nt Ro. 5 to PSAll containina reviaed and auppl...ntary infonaation and partial reaponaea to ABC letter* of 11/29/71 and 1/12/72.

March 10, 1972 PASIIY aubllita reaponae to ABC letter of 10/6/71 on the onaite ... rgency pCNer ayat.. teat progr ...

March 13, 1972 .-c letter requeat* evaluation of the effect* of blowdown fore** on torua atructur***

March 14. 1972 ABC 1etter requeat* info..-tion needed to calculate effluent aource ter111a.

(,* '* *. ,

A-4 March 28. 1972 Meeting with PASNY to discuss emergency plan and industrial security requirements and.responses to questions on Conduct of Operations.

March 31. 1972 PASNY submf.ts response to AEC letter of 3/14/72 on sourc* term data.

Apr11 6. 1972 ABC letter requests additional information on pipe whip analysis and drywall impact prevention.

Aprf.1 10. 1972 PASNY subad.ts Alllenc:laent No. 6 together with Supplement No. 6 to FSAR containing revised and supplementary information and partial responses to AEC letters of 11/29/71 and 1/12/72.

April 17, 1972 PASNY submits response to AEC letter of 2/18/72 on coap1iance to 10 CPR 50.55(a).

Aprf.1 2s. 1972 PASNY submits supplementary information to their 1etter of 4/17/72.

Aprf.1*2S. 1972 PASNY subad.ts response to ABC 1etter of 3/13/72 on effects of blowdown forces on torus structure.

Aprf.1 27 9 1972 lleeting with New York State Departaent of Environaenta1 Conservation concerning p1ant and site information requirements.

Aa,i::1.1 2a. 1912 PASNY submits response to ABC 1etter of 10/12/71 df.scuasing p1ana for contro1 of the containment environaent.

Nay 3. 1972 ABC 1etter request detailed supporting data on propo*ed containment atao*phere control concept.

11ay.12. 1972 PASRY aubaita letter updating the con*tructf.on status of the tranalld.**ion 11.ne.

Nay 15. 16, 617. 1972 llaetina vf.th P.ASIIY to review control.

f.n*tr1mantatf.on and electrical *Y*tw

-dradnp.

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A-5 May 18, 1972 PASNY submi.ts Amendment No. 7 together with Supplement No. 7 to FSAR containing

revised and supplementary information and partial responses to ABC letters of :i 11/29/71 and 1/12/72.

May 22 9 1972 PASNY submi.ts letter providing additional infoanation on conformance to 10 CPR 50.55(a).

Nay 26, 1972 PASNY sublllits response to ABC letter of 4/ 6/72 describing additional-pipe whip protection.

June 5 9 1972 ABC letter requests additional informa tion for.preparation of Draft Environ mental Statement.

June 6 9 1972 llaeting with PASNY to discuss question responses and need for additional infonaatian in site safety related areas and to discuss applicant's apparent nonconforaance with Codes and Standarda requirements.

June 9 9 1972 PASRY aublld.ta Suppleaent No. 2 to Environmental Report containing a Coat and Benefit Ana1yais.

June 13 9 14 9 6 15 9 1972 Meeting with PASNY to continue contro1.

instruaentation . and electrical ayat81D8 drawing review.

June 22, 1972 PASRY aubllits additional clarification of torus structural analyaia.

June 27 6 *21, 1972 Meeting with PASNY to diacuss additional nforaation needed for enviromaenta1 revt.ew and to viait the site.

June 28 9 1972 PASIIY auba:lta Allend-nt No. 8 together Id.th Supp1eaent No. 8 to l'SAll contdnina revf.aed and aupp1eaentary information and ***inder of reapon*** to ABC letters of 11/29/71 and 1/12/72.

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1 A-6 June 30, 1972 Meeting with PASNY's attorneys to revi.ew schedu1e and adlestones.

- July 7, 1972 AEC letter requests additiona1 information and clarification on PSAR site safety re1ated itelllB.

Ju1y 18, 1972 Site vi.sit to review safety and env1.ronaental deta11a.

August 7, 1972 ;._ ABC letter requests updated financial inforaation for preparation of safety eva1uation.

Auauat 7, 1972 PASNY aublllf.ts Aaendlllent Ro. 9 toaether with Suppleaent Ro. 9 to the PSAll containing revf.sed and suppleaentary inforaation and reeponeee to te ABC letter of 7/7/72 * .

August 11, 1972 ABC 1etter requests additional information and clarification of PSAll it... in several areas.

Auauat 11, 1972 PASRY euba:lta Supp1eaent Ro. 3 to Bnvf.roamental lleport containina reaponaee to ABC letter of 6/5/72.

Aupat 15, 1972 PASNY auba:lta Aaendaent Ro. 10 toaether with Suppleaent Ro. 10 to the l'SAll containina reaainder of reaponaea to the ABC letter of 7/7/72.

Auauat 24

• 25, 1972 Neetina with PASIIY to caapl*t* control inatruaentation, and electrical ayat...

dravina review.

Auauat 28, 1972 MC iaau** Special Nuclear Material*

licena* to PASRY.

SapteFlber 6, 1972 PASlff aublllt* Acsnd-nt llo. 12 to the application containina the financi*l

lnforaation Eequeated in the ABC letter of 8/7/72 *

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A-7 September 22, 1972 PASNY submits Amendment No. 11 together with Supplement No. 11 to the FSAR containing revisions and updated information as well as the Emergency and Industrial Security Plans.

September 25, 1972 PASNY submits Amendment No. 13 together with Supplement No. 12 to the FSAR containing responses to the ABC letter of 8/11/72.

September 29, 1972 AEC letter requests that PASNY prepare and submit system operating pressure and temperature limitations based on the 1972 Sumner Addenda, Appendix G, to Section III of the ASHE Boiler and Pressure Vessel Code.

October 3 1 1972 Notice of Hearing and Consideration of Issuance of Facility Operating License published in Federal Register.

October 4 1 1972 PASNY sublllits some supporting data on proposed containment atmosphere control concept in partial response to AEC letter of 5/3/72 .

October 10, 1972 AEC letter requests that PASNY aodify their Industri.al Security Plan to eliminate deficiencies n certain areas.

October 11, 1972 AEC letter requests analyses together with augmented testing and surveillance progra1D8 to measure and evaluate bypass leakage through the vacuum breaker valves.

October 16, 1972 PASNY submits Aaendaent No. 14 together

• with Supplement No. 13 to the PSAll containing 1111.scellaneous revi*iona to text and previoualy answered queationa.

October 16, 1972 PASNY sublllits reaponae to AEC letter of 9/29/72 indicating pl.anned conformance with 1972 S1-*r Addenda to ASNE Boiler and Pr***ure Veaael Code.

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• . i A-8 October 25, 1972 PASNY submits Revision 1 to the Industrial Security Plan. PASNY also submits Amend ment No. 15 together with Supplement No.

14 to the FSAR containing miscellaneous revisions to the FSAR text and to the Emergency Plan.

October 30, 31 and Site visit to inspect implementation of November 1, 1972 control, instrumentation and electrical

' i systems.

October 31, 1972 PASNY submits Amendment No. 16 together with Supplement No. 15 to the FSAR con taining revised pages and miscellaneous additional information.

l "'

October 31, 1972 AEC letter requests description of system proposed to augment the Rod Worth Minimizer in connection with the postulated Control

, ., Rod Drop Accident.

November 3, 1972 PASNY submits partial response to AEC letter of 10/ 11/72 pertaining to surveil lance of primary containment leakage paths.

November 3, 1972 PASNY submits additional infor mation on containment atmosphere dilution system.

November 10, 1972 PASNY submits Amendment No. 17 together with Supplement No. 16 to the FSAR con taining miscellaneous additional information.

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9&01 LITTLE fl'ALL.S ROAD. N.W.

WA8HINGTON. D.c. aoo1*

CEREN-DE 11 October 1972 Docket No. 50-333 Mr. J. F. Stoly. Chief Boiling Water Reactors Branch 2 Directorate of Licensing

u. s. Atomic Eneray Coanission Washinaton. D. c. 20545

Dear Mr. Stoly:

Reference is made to your letter of 30 September 1971 regarding Docket No. 50-333. the Port Authority of the State of New York. James A.

Fitzpatrick Nuclear Power Plant. FSAR. Volumes 1-9 and Amendments 1 thrOll&h 13 thereto.

In accordance with our arrangements. an engineer of the CERC staff has reviewed pertinent information in the report leading to the establishment of the maxilDUJll and minimwn desian water levels at the plant site. It is his opinion with which I concur. that the Probable Maximum flood levels of El. 254.0 feet (USLS 1935. datum) in the screenwell and El. 261.0 feet (USLS. 1935. datum) at the shoreline. and a Probable Minimum low water level of El. 236.S feet (USLS. 1935. datum) discussed by the applicant.

are compatible with the conservative desian requirements associated with the passaae of a Probable Maxi11U111 aeteorological event.

Analysis of the factors involved with overland and screenwell floodina determined that surface access into the plant appears to be at or above El. 272.0 feet (USLS. 1935. datua). and that safety related equipment in the screenwell is above El. 255.0 feet (USLS. 1935. datum). (or one-foot above Probable MaxilllUJII flood level). It would thus appear that sufficient

£reeboard exists to preclude any potential problems from either overland or sreenwell flooding from interferina with the safe operation of the plant.

Comparison of the miniaua desian lake level at El. 236.5 feet (USLS. 1935.

datum) with top elevations of the lake intake structure indicates that adequate water supply should be available to the screenwell from Lake Ontario throu1hout Naxiaua Probable lake level fluctuations.

CP: Mr. L. G. Hulaan/ABC DON s. McCOY DABN-CWZ-R/Major 811ith Lieutenant Director

.. r,i* *.i

C-1 APPENDIX C UNITED STATES DEPARTMENT OF THE INTERIOR GEOLOGICAL SURVEY WASHINGTON, D.C. 20242

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Dear Mr. i"1t.1.ntzi:

Tra.nsm-i-tted herewith in resonse to a reg_uest, by .rvir. Ro3er S. Boyd.,

c.:.tec1 October ll, l97l, i..s a :r*eview of' the geologi.ca.l aspects of" a minor f'e.ul"i.:, disclosed du.r:!..ng excavD.tj_on o:C' t.r..e int.a.lee and d:Lscha:rce tunneJ.s i"or

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c-2 Power Authority of the State of New York James A. FitzPatrick Nuclear Power Plant AEC Docket No. 50-333 Geo1ogy Field Review The FitzPatrick Nuclear Power Plant site in New York was visited on December 16, 1970, to inspect and discuss a sma11 fault encountered in the Intake and Discharge Tunnels. Dr. D. U.

Deare,

consulting engineer-ing geologist, and Messrs. Asa George of PASNY, and W. F. Swiger, J. Wainrib, C. E. Goodman, J. P. Allen, and others of Stone and Webster Engineering Corporation accompanied me on the inspection.

The fau1t was encountered in the tunnels at locations about 500 feet offshore and approximately 50 feet below the lake; first in the Intake Tunnel, on October 16, at Station 106+32, and then in the Discharge Tunnel, on October 19, at Station 6+20. The tunnels are being excavated in the nearly f1at-1ying Oswego Sandstone of Ordivician age. This formation consists of thin-bedded quartzitic sandstone and some interbedded sha1e.

The formation breaks well and stands well without support.

The fault is a sma11 normal fault that crosses the tunnels at nearly right angles, trending approximately N. 78 ° w. and dipping about 60° to 0

64 south. The fault consists of a thin clayey gouge, 1/2 to 3 inches thick, and a bordering fractured zone that ranges in width from a few inches in the sandstone beds to as much as 24 inches in the shale beds.

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C- 3 Slickensided surfaces are rare, but where observed they indicate normal dip-slip movement. Laboratory analyses of the gouge material indicate that it is essentially unaltered, ground-up shale.

Measurable displacement along the fault is about 16 to 18 inches vertically (the hanging-wall has moved downward with respect to the foot wall). The amount of lateral movement, if any, is not known, but the evidence from the slickensides suggests that most of the movement was dip-slip. Both the fault and bordering fractured zone are tight.

An examination of the projected trace of the fault in bedrock ex posed in the barge slip area and in a trench excavated down to rock,.

about 300 feet east of the barge slip, showed only a system of joints with approximately the same attitude as the tunnel fault; no displacement of the bedrock or of the overlying till or soil could be detected. Conse quently, although the age of the last movement along the fault could not be determined, the available evidence suggest s that the fault is geo logically old--certainly older than the last glacial episode in the area and probably much older. It is concluded, therefore, that the fault encountered in the tunnels is a minor geologic. feature that does not significantly affect the safety of the tunnels.

The proposed treatment of the fault zone .is essentially the standard treatment provided for such features encountered in tunnels. The principal purpose of the treatment is to provide for drainage of the zone and to improve the character of the material within the faulted zone.

January 20, 1972 2

D-1 l

APPENDIX D NATHAN M. NEWMARK CONSULTING ENGINEER:NG SERVICES 1114 CIVIL ENGINEERING BUILDING URBANA. ILLINOIS 81801 3 November 1972 FINAL REPORT

• /

ON STRUCTURAL ADEQUACY l '

OF THE JAMES A. FITZPATRICK NUCLEAR POWER FLANT PCMER AUTHORITY OF THE STATE OF NEW YORK AEC Docket No. S0-333 by W. J. Ha 1 I and N . M . Newmark After our review of the FSAR ., Including Amendments 1 through !i and 8 through 12 ., It Is believed that the desin of the James A. FftzPatrfck Nuclear Pc:Mer Plant can be considered satisfactory for safe shutdONn for a Design Basis Earthquake of 0.1Sg maximum transient horlzonta l ground acceleration and capable otherwise of withstanding the effects of an Operating Basis Earthquake of half this Intensity. Our review was based.

on consideration ., among other things ., of the foundations ., earthquake hazard,

• spectra 1 damping, dynamic analysis of structures 1 piping and equipment 1 stress 1 lmlts ., burled piping ., critical Instrumentation and electrlcal systems.

It Is our conclusion that the design Incorporates a satfsfactory range of margins of safety for the hazards considered.

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E-1 APPENDIX E FINANCIAL QUALIFICATIONS The Conmission's regulations which relate to the financial data and information required to establish financial qualifications for an applicant for an operating license are 10 CFR S0.33(f) and 10 CFR SO, Appendix C.

The basic application of Power Authority of the State of New York (PASNY) and Niagara Mohawk Power Corporation , Amendment No. 15 thereto, the amended and substituted application, Amendment No. 12 thereto, and the accompanying certified annual financial statements of the applicants provide the financial information required by the Cmmnission's regulations. This information includes the estimated annual costs of operating the James A. PitzPatrick Nuclear Power Plant for the firs five years of operation plus the estimated I I

il cost of permanently shutting down the facility and maintaining it in a safe *l r,

shutdown condition. I Our evaluation of the financial data submitted by the applicants, suaaarized below, provides reasonable assurance that the applicants poeeeee or can obtain the necessary funds to meet the requ1.reaents of 10 CPR S0.33(f) to operate the James A. PitzPatrick Nuclear Power Plant and, if neceaaary, permanently shut down the facility and aaintain it in a safe shutdown condition.

PASNY is a public benefit corporation constituting a political aub divieion of the State of Rew York. The Authority conaiata of *five truatee*

appointed b y the Governor of Nev York. PASRY f.a a vholeaale power aupplier and sell* it:* power to about 50 cuat:oaera, includina three ut:f.lity coapani**

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and 42 nunicipally and cooperatively owned electric systems in New York.

The FitzPatrick Plant will be owned by the Authority and operated and maintained by Niagara Mohawk under contract with the Authority. Costs incurred by Niagara Mohawk in operating and maintaining the plant will be reimbursed by the Authority* under its contractual arrangements with Niagara Mohawk. Revenues to cover operating expenses will be derived from the sale and transmission of power produced by PASNY's generating plants. The estimated costs of operating the James A. PitzPatrick Nuclear Power Plant for the initial five year period (1973-1977) are presently estimated by the applicants to be (in mllliona of dollars) $4.2; $27.6; $30.1; $31.9; and $34.6, in that order. These coats include amounts for operation and maintenance, fuel, insurance, overhead and interest on investment. These costs do not include mnortization or depreciation as the Bond Reao1ution convering bonds issued to finance construction of the plant does not provide for principa1 retirements prior to 1981, and the Resolution excludes depreciation as an operating expense. The Authority is exempt from taxation.

In addition to operating coats, the applicants estimate the cost of peraanent1y shutting down the faci1ity at the conclusion of its uaefu1 1ife (baaed on 1972 dol1ars) wf.11 be $7 m:l.11ion. It is estimated that an annual coat of $150,000 (in 1972 dollars) wf.11 be incurred to maintain the facility in a *afe *hutdown condition.

The Authority wil1 provide funds to cover coats to *hut down and

-intain the *hut down facility in a safe condition. Prf.or to 1985 the Authority would finance theae co*t* through iaauance of notes. After 1984

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E-3 revenues from its St. Lawrence and Nagara Projects and reserve funds established from such revenues would be available to cover these costs.

The information contained in PASNY's calendar year 1971 financial repor t indicates that operating revenues for 1971 totaled $100.4 million; operating expenses were $13.S million. The interest on long-teEm debt was earned 3.2 times; and the net income for the year was $63.9 million.

all of which was used in the business. As of December 31. 1971. the Authority's assets totaled $1,658.0 million, most of which was invested in gross utility plant ($1.371.0 million). Pinancial ratios ccmputed from the 1971 statements indicate an adequate financial condition, e.g.*

long-term debt to total capitalization - .69; and to gross utility plant -

.80; the operating ratio - .13; and the rate of return on total investment

- 5.6%. The record of PASNY's operations over the past 2 years reflects that operating revenues have been relatively constant froa $102.6 ad.11:lon in 1969 to $10 0.4 million in 1971. Net income liktndse has been constant at $63.9 ad.11:lon. Grose investment in plant :lncreaaed $1.123.4 111111:lon to $1.371.0 alllion. wh:lle the nuaber of ti-* :lntereat earned increaaed froa 2.9 to 3.2. Moody's Inveatora Service rate* PASRY'a bonds as A-l (strona-upper -d:lua arade) and Ena-A (under conatruct:lon).

The :lnfonaat:lon conta:lned :ln N:laaara llohavk Paver Corporat:lon'*

calendar year 1971 f:lnancial report :lnd:lcates that operat:lng revenues for 1971 totaled $571.4 mill:lon; operat:lng expenses were $ 477.0111:lll:lon,.

of wh:lch $52.8 alllion repreaented deprec:lation. The :lntereat on long-term debt vaa earned 2.3 t:l-*; and the net :lncoae for the year vaa $56.8 a:lll:lon.

E-4 of which $44.6 mf.llf.on was distr:lbuted as divf.dends to stockholders and the remainder of $12.2 m:lll:lon was retaf.ned for use f.n the buaf.ness. As of December 31, 1971, the Company's assests totaled $1,774.2 llon. aost of which was invested in net util:lty plant ($1,660.8 ad.11:lon); retained earn:lngs aaounted to $192.2 mf.11ion. Financial ratios coaputed fro* the 1971 statements indicate an adequate financ:la1 c ondition, e.g., 1ong-term debt to total capitalization -

• SS; and to net utility p1ant -

• S3.; net p1an to capitalization - 1.04; the operating ratio - .83; and the rates of return on coaaon - 9.1%, on stockho1ders' investment - 8.01, and on tota1 investment - 5.6%. The record of Niagara'* operations over the past 2 years reflects that operat:lng revenue* increased from $4S3.S 1111111.on in 1969 to $571.4 mi11ion in 1971. Ret 1.ncoae increased froa $51.3 1111111.on to

$56.8 a:l.11ion and net investment in plant increased froa $1,453.9 111111ion to $1,660.8 1111111.on, while the nuaber of tfllles interest earned declined from 2.S to 2.3. lloody'* *Inveatora Service rates the Corporation'*

first a>r taqe bonds as A (upper -diua grade). Their current Dun and Bradstreet credit rating ia 5Al. Niaaara appears fully qualified financially to operate the faci1ity for PASNY.

Our financia1 analyai* of each oraani.*ation reflect:ina the ratios and other pertinent financia1 data i* au arisecl in t:he att:ached t:altlea.

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E-S POWER AUTHORITY OP THE STATE OP NEW YOIIK DOCltET NO. S0-333 nNANCZAL ANALYSIS.

(dollar* in ailliona)

Calendar Year Ended Deceaber 31 1971 1970 1969 Long-tera debt $1.092.7 $ 939.S $ 878.9 Utility plant (groaa) 1.37 1.0 1.193.9 1.123.4 llatio - debt to fixed plant .80 .79

• 78 Net incoae before intereat 92.S 94.4 98.1 Liabilitiea and Capitalization 1 9 658.0 1.497.4 1.298.1 Rate of earnina* on total inveataent S.S8% 6.30% 7.56%

Net inco.. before intereat 92.5 94.4 98.1 Intereat 28.6 31.S 34.2 No. of ti..* intereat earned 3.23 3.00 2.87 Operatina expenae* 13.5 11.8 10.,

Total operatina revenue* 100.4 99.5 102.6 Operatina ratio .13 .12 .10 Net incoaa 63.9 62.9 Total revenue* 106.0 106.2 101.*.6 Net incoae ratio .60 .59 .59 Utility plant 1.371.0 1.193.9 . 871.t.

Utility operatina revenue 100.4 99.5 l.02.6 Ratio of plaat inveat..nt to ravenuea 13.65 12.00 8.57 11z1 lfZll * .* * * *.*

Capitalisatioqt trrvnt I of of tes,1 Tote) MIOUIIS Lona-tena debt t1.ot2.1 69.21 939.5 69.41 Bonda retired 379.3 24.0 302.5 22.3 llaveauea allocated Total 1,.ffl:i 6.1 100.0I 11.iHB Dl:Ji lloody'a Bond llatina* A-1 and<****>*

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...--------,..,loo,," *....__ ....._.. -....,..- .... ,.- ...-,.. *or*** ......,,.,e-,._,, __ ,.,.,..** .,,.-.*****"'""'-'"'-" *** *,, 4* *., ..:,,.".--"'* "".., ,..,....,* .,-* ,,,..._..._,_,,.._..,. *'" *..... -...,....., _ . .,.-..,.,. _-_,_,.,_, *.,........, _ ...,*'--.+.L..'""'""..,.____ :.... .._... _.c,,...,. *...,_* .,....,.-,*..,...,.-..... .. ,-w-.-.....-_.. - .*..,. .. *'* -4 ,...,.,..., ....., .... _........,_......_._ * *,.._..,_..__ . ..---.**.....i E-6 MIAGAIIA ll>IIAWIC POWBa COaPOIIATl<B DOCKET RO. 50-333 PDIAIICIAL ANALYSIS (dollars in 11d.lliona)

Calendar Year Ended Decellber 31 1971 1970 1969 Lona-tera debt $ 882.8 $ 818.3 $ 820.2 Utility plant (net) 1.660.8 1.534.9 1.453.9 Ratio - debt to fixed plant .53 .53 .56 Utility plant (net) 1.660.8 1.534.9 1.453.9 Capitalization 1.596.9 1.485.3 1.439.4 Ratio of net plant to capitalisation 1.04 1.03 1.01 Stockholder*' equity 714.1 667.0 619.2 Total .. aeta 1.774.2 1.644.5 1.554.6 Proprietary ratio .40 .41 .40 Barnina* available to coaaon equity 49.3 43.8 43.7 Ca on equity 543.0 495.9 448.1 aate of earnina* on c011111on equity 9.11 8.81 9.81 Ret incoaa 56.8 51.4 51.3 Stockholder*' equity 714.1 667.0 619 *.2 aate of earnina* on atockholder*' equity 8.01 7.71 8.31 llet incoae before intereat 99.9 91.1 85.6 L:labilitiea and capital 1.774.2 1.644.5 1.554.6 II.ate of earnina* on total inveat..nt 5.61 5.51 5.51

• • t incaae before intereat 99.9 91.1 85.6 Xntereat on lona-t*ra debt 42.8 39.5 34.1

• o. of tiae* lona-tera intereat earned 2.3 2.3 2.5

• • t bcoae 56.8 51.4 51.3 Total 577.0 530.9 463.7 llet 9.81 9.71 11.11 Total utility operatiaa expeaae* 477.0 439.8 377.2 Total utility operattaa reYe11Uea 571.4 522.2 453.5 Operattaa ratio .83 .84 .13 Utility .1..t <aroa*> 2.11,., 2.015.4 1.1,2.1 Utility ...ratiaa re,,...... 571.4 522.2 453.5 r,

llatio of pl*t iave*t.. at to revenue* 3.12 4.17 3.86 11z I 1 Tosal

• 8.1111.

• TaHI 0- .1aD11111 Laaa-t** ...t N2.I 55.31 111.3 55.11 hef*rr** atock 171.1 10.7 171.1 11.5 Ct 111 *toek * *urplua ,,,.o 34.0 495.9 33.4 Total 11,211,1 191,91 11,*11,i 1,ag,91

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