ML19208A238

SER for Facility,Suppl 1

ML19208A238
Person / Time
Site:	Green County
Issue date:	09/30/1978
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0283, NUREG-0283-S01, NUREG-283, NUREG-283-S1, NUDOCS 7909130264
Download: ML19208A238 (45)

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Regulatory Comm s or jrelated to construction of office of Nuclear i

Reactor Regulation

-Greene County Nuclear Power Plant oocket mo. so-s49 l Power Authority of the oa september 1978

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State of New York l Supplement No,1 l

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Printed Copy $5.25; Microfiche $3.00 The price of this document for requesters outside of the North American Continent can be obtained from the National Technical Information Service.

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NUREG-0283 Supplement No. 1 September 5, 1978 SUPPLEMENT NO. 1 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF POWER AUTHORITY OF THE STATE OF NEW YORK GREENE COUNTY NUCLEAR POWER PLANT DOCKET N0. 50-549 739003

TABLE OF CONTENTS fagg

1.0 INTRODUCTION

AND GENERAL DISCUSSION.

1-1 1.1 Introduction.

1-1 1.7 Requirements for Future Technical Information.

1-2

1. 8 Outstanding Items.

1-2 1.10 Significant Changes to the Application Found Acceptable by the Staff.

1-6 2.0 SITE CHARACTERISTICS.

2-1 2.1 Geography and Demography.

2-1 2.1.2 Exclusion Area Control.

2-1 2.2 Nearby Industrial, Transportation and Military Facilities.

2-1 2.4 Hydrology.

2-2 2 4.5 Design Basis Groundwater Level.

2-2 3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS.

3-1 3.2 Classification of Structures, Components, and Systems.

3-1 3.2.1 Seismic Classification.

3-1

3. 5 Missile Protection.

3-1 3.5.1 Missile Selection and Protection Criteria.

3-1 3.5.2 Barrier Design Procedures.

3-2 3.6 Protection Against Dynamic Effects Asscciated with the Postulated Rupture of Piping.

3-4 3.6.2 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping Outside Containment.

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3. 7 Seismic Design.

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EBLEOFCONTENTS(Continued)

PAGE 3.7.2 Seismic System and Subsystem Analysis.

3-5 3.11 Environmental Cesign of Mechanical and Electric Equipment.

3-6 4.0 REACTOR.

4-1 4-1 4.2 Mechanical Design.

4.2.1 Fuel Design.

4-1 4.4 Thermal and Hydraulic Design.

4-1 4.4.3 Thermal and Hydraulic Analysis.

4-1 5.0 REACTOR COOLANT SYSTEM.

5-1 5.4 Component and Subsystem Design.

5-1 5.4.5 Decay Heat Removal System.

5-1 6.0 ENGINEERED SAFETY FEATURES.

6-1 6.2 Containment Systems.

6-1 6.2.1 Containment Functional Design.

6-1 6.2.2 Containment Subcompartment Design.

6-1 6.2.5 Containment Isolation System.

6-3 6.2.6 Combustible Gas Control System.

6-3 6.3 Emergency Core Cooling System.

6-4 6.3.3 Discussion of Loss-of-Coolant Accident Analyses.

6-4 7.0 INSTRUMENTATION AND CONTROLS.

7-1

7. 6 All Other Systems Required for Safety.

7-1 7.6.1 Decay Heat Removal System.

7-1 73 W 6

TABLE OF CONTENTS (Continued)

PAGE 8.0 ELECTRIC POWER SYSTEMS.

8-1 8.2 Of fsite Power Systems.

8-1 8.3 Onsite Power Systems.

8-2 8.3.1 Alternating Current Power Systems.

8-2

9. 0 AUXILIARY SYSTEMS.

9-1 9.1 Fuel Storage and Handling.

9-1 9.1. 2 Spent Fuel Storage.

9-1 9.1.3 Fuel Pool Cooling and Purification System.

9-1 9.2 Water Systems.

9-1 9.2.1 Reactor Plant Service Water System.

9-1 9.2.2 Reactor Plant Component Cooling Water System.

9-2 9.2.3 Ultimate Heat Sink.

9-3 9.4 Air Conditioning, Hea'ing, Cooling and Ventilation Systems.

9-3 9.4.1 Control Building Air Conditioning ar': Ventilation Systems.

9-3 9.4.3 Fuel Building Ventilation System.

9-4 10.0 STE. 1 AND POWER CCNVERSION SYSTEM.

10-1 10.4 Other Features of Steam and Power Conversion System.

- l 10.4.5 Steam Generator Heatup System.

10-1 13.0 CONDUCT OF OPERATIONS.

13-1 13.1 Organizational Structure of Applicant.

13-1 13.6 Industrial Security.

13-1 733006 iii

TABLE OF CONTENTS (Continued)

E 15-1 15.0 A"CIDENT ANALYSIS.

15-1 15.3 Moderate Frequency Events.

15-1 15.3.1 Increase in Heat Removal by the Secondarj System.

15-1 15.4 Infrequent Transients and Accidents.

15.4.4 Feedwater System Breaks Inside and Outside of 13-1 Containment.

15.4.5 Spectrum of Steam Pipir g failures Inside and Outside of 15-3 Containment.

15-4 15.5 Radiological Consequences of Accidents.

15-4 15.5.1 Loss-of-Coolant Accident.

15.5.4 Main Steamline Break Outside of Containment.

15-7 18-1 18.0 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.

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21.0 CONCLUSION

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APPENDICES PAGE APPENDIX A - CONTINUATION OF CHRONOLOGY C' RADIOLOGICAL REVIEW.

A-1 APPENDIX B - BIBLIOGRAPHY (Continued).

B-1 APrENDIX C - CHANGES AND ERRATA TO THE SAFETY EVALUATION REPORT ISSUED S~PTEMBER 1977.

C-1 APPENDIX 0 - ADVISORY COMMITTEE ON REACTOR SAFEGUARDS REPORT ON GREENE COUNTY NUCLEAR POWER PLANT.

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1.0 INTRGDUCTION AND GENERAL DISCUSSION 1.1 Introduction The Nuclear Regulatory Commission's Safety Evaluation Report for the Greene County Nuclear Power Plant (NUREG-0283) was issued on September 1, 1977. In that report, we concluded that upon favorable resolution of the outstanding matters set forth in Section 1.8 of the Safety Evaluation Report, we would be able to reach the conclu-sions required in accordance with the provisions of Section 50.35(a) of 10 CFR Part 50.

Section 1.8 of the Safety Evaluation Report identified 17 outstanding items for which additional informatior was required from the applicant, or for which the staff had not completed its review of recent Preliminary Safety Analysis Report supplements.

This Supplement No. I to the Safety Evaluation Report summarizes the results of the

+echnical evaluation of the proposed Greene County Nuclear Power Plant performed by the Commission's staff since the Safety Evaluation Report was issued.

Since the Safety Evaluation Report was issued:

(1) The Advisory Committee on Reactor Safeguards has completed its review of the Greene Conty Nuclear Power Plant application. Its report is included as Appendix 0 The Committee's comments are discussed in Section 18.0 of this supplement.

(2) The applicant has submitted Preliminary Safety Analysis Report Sup,lement Numbers 22 through 28 responding to the outstanding items identified in Section 1.8 of the Safety Evaluation Report and +o matters raised by the siaff since issuance of that report. These Preliminary Safety Analysis Report supplements also describe significant changes to the application found accept-able by the staff as listed in Section 1.10 of this supplement.

(3) We have resolved 13 of the 17 outstanding items reported in Section 1.8 of the Safety Evaluation Report, and we have identified seven additional outstanding items in Section 1.8 of this supplement. We have resolved two of these seven additional outstanding items. Therefore, a total of nine outstanding items remain unresolved at this time. We will report on the resolution of these nine outstanding items in a future supplement to the Safety Evaluation Report.

Our review of the financial qualifications of the applicant is continuing and will be reported in a future supplement to the Safety Evaluation Report.

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Appendix A to this supplement is a continuing chronology of the principal actions related to the radiological review of the Greene County Nuclear Power Plant application.

Appendix B incorporates additions to the bibliography appended to the Safety Evaluation Report. Appendix C is a licting of errata to the Safety Evaluation Report. The changes effected by Appendix C do not alter the staff conclusions presented in the earlier Safety Evaluation Report or the significant information upon which these conciusions are based.

Each of the following sections in this supplement is numbered the same as the corresponding section of the Safety Evaluation Report. This supplement is an addition to and is not in lieu of the discussion in the Safety Evaluation Report.

1.7 Requirements for Future Technical Information We have identified the need to review certain information before the applicant begins the installation of certain equipment or construction of certain structures in the event of a favorable decision on issuance of a construction permit. Since the evaluation of this information may affect subsequent design and construction actions, we conclude that our review of these matters should be made during con-struction rather than later during the operating stage of review. The applicant has committed to provide the following information listed below on a schedule consistent with the design and construction of the plant. This will provide an opportunity for our review of this information and the associated corrective actions by the applicant should these actions oecome necessary.

(1) A report describing an acceptable environmental qualification program and its schedule for the balance-of plant mechanical and electrical equipment (Section 3.11).

1.8 Outstanding Items The Safety Evaluation Report identified 17 outstar. ding items for w"ich our review was not complete. These items are listed below and are discussed fu.ther in the sections of this report as indicated. The applicant and the staff have resolved 13 of the outstanding items as described below. At this time, four items identified in the Safety Evaluation Report still remain outstanding; they are numbers (2),

(9), (10), and (14). For each of the items below the applicant has submitted additional information. Therefore, the status of their review and our requirements have changed. Accordingly, we have restated the outstanding items to reflect the present status while retaining their earlier identification numbers.

(1) We required additional infermation as to the applicant's acquisition authority unde

• the laws of the State of New York. The applicant submitted additional information which resolved this item (Section 2.1.2).

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(2) We required additional information on analyses of c ound shock as a conse-quence of the accidental detonation of an explosive-laden truck on Route 9W (relocated) or nearby explosive magazine. The applicant submitted a partial response and more recently additional information which the staff i, reviewing.

This matter remains an outstanding item (Section 2.2).

(3) We required that the ultimate heat sink cooling towers must be designed to protect against vertical tornado missiles. The applicant's commitment to do this resolved this item (Sections 3.5.1 and 9.2.3).

(4) We recuired increased wall thickness for concrete tornado missile barriers.

The applicant adopted an alternate staff position on wall and rocf thicknesses which resolved this item (Section 3.5.2).

(5) We required a commitment that the f acility design criteria will limit concrete barrier design to ductil1 U factors lower than 10.

The applicant's commitment to do this resolved this item (Section 3.5.2).

(6) We required a commitment that the applicant provide a technical report within one year af ter issut ice cf the construction permit describit an acceptable s

environmental qual " eat.on program and its schedule for the balance-of plant mechanical and electrical equipment. The applicant's commitment to do this resolved this item (Section 3.11).

(7) We required additional information from the applicant to demonstrate that decay heat removal system isolation from the reactor coolant system can be accomplished in the event of a postulated pipe failure in the decay heat removal system outside containment, during shutdown cooling, assuming a single active component failure, in accordance with Standard Review Plan, Section 3.6.1.

The, applicant submitted the additional information which resolved this item (Sections 5.4.5 and 7.6.1).

(8) We required a commitment that the applicant will provide the results of addi-tional nodalization studies of dynamic asymmetric pressure loads within the reactor cavity, steam generator subcompartments and pressurizer subcompartment to provide conservati"* loads for the design of these component supports. The applicant's commitmen, to do this resolved this item (Section 6.2.2).

(9) The applicant proposes to purge the containment during hot shutdown. We required that the applicant restrict containment purging ta cold shutdown and refueling, or provide a system design and analysis that complies with the reconmendations of Branch Technical Position CSB 6-4, "Containrent Purging During Normal Plant Operations." The applicant submitted additional informa-tion which the s.aff is reviewing. This matter remains an outstanding item (Section 6.2.5).

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(10) We required the applicant to confirm that the combustible gas control system will be designed to the larger metal-water reacti, given in Regulatory Guide 1.7, " Control of Combustible Gas Concentration in Containment Following d loss-of-Coolant Accident. " The rpplicant submitted a reanalysis to show the proposed design capable of acceptably controlling the combustible gas concen-trations following a loss-of-coolant accident assuming the larger metal-water reaction. We then requested additional information on the corrosion rate for metals in the post-accident environment. The applicant submitted additional infor.dation which the staff is reviewing. This matter remains an outstanding item (Section 6.2.6).

(11) At the time tre Safety Evaluation Report was issued we had not completed our review of the emergency core cooling system analysis in conformance sith Appendix K to 10 CFR Part 50 far the specific design parametcrs used in the Greene County Nuclear Power Plant. This review has now been completed. We find the applicant's analysis to be acceptable, and this item is resolved (Section 6.3.3).

(12) We required additional information on the grid stability and frequency decay rate to demonstrate that grid transients will not result in fuel damage, or we requir ed the applicant to provide safety grade breakers and associated instra-mentation and controls for the reactor coolant pumps. The applicant submitted a commitment to show by analyses that grid transients will rat result in fuel damage, or to provide the required safety grade instrumentation and controls for the reactor coolant pumps. The grid transient analyses will be provided in the Final Safety Analysis Report. The applicant's commitment resolved this item (Section 8.2).

(13) We required a cesign modification to eliminate the manual transfer switching 4.ircuit for the third make up/hich pressure injection system pump. The appli-cant subr..:.ed a description of the design change which resolved this item (Section 8.3.1).

(14) We required the component cooling water system be designed to preclude unaccept-able damage to the seals on the reactor coolant pumps. The applicant submitted additional information to s A that a loss of component cooling water to the pump seals will not result in a treach of the reactor coolant pressure boundarv.

The applicant further committed to perform a test to demonstrate that capability.

We requested a commitment from the applicant that in the event the proposed test fails to demonstrate that a loss of component cooling water to the reactor coolant pump seals could not result in a breach of the reactor coolant pressure boundary, this piping will be provided with the grade classification and instrumentation to satisfy the stiff position as stated in the Safety Evalua-tion Report. This matter remains an cutstanding item (Section 9.2.2).

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(15) We required additional informa' ion to confirm that the bounding transient for steam releases from the secondary system is the inadvertent opening of the turbine bypass valves. The applicant submitted additioral information on the design o.' the control system for the main steam dump valves and turbine bypass valves to demonstrate that the analyzed transient is the bounding event. The submittal resolved this item (Section 15.3.1).

(16) We required additional information on the system performance for a postulated feedwater sys em break inside and outside containment. The applicant submitted additic,1al information which resolved this item (Section 15.4.4).

(17) We required additional informatior on the system performance for a postulated steam system pipirg failure inside and outside containment. The applicant submitted additional information which resolved this m (Section 15.4.5).

Additional outstanding items, raised by the staff since the Safety Evaluation Report was issued, are listed below with the sections of this supplement where they are discussed. At this time we have reviewed and found acceptable item numbers (18) and (21).

(18) We required a reanalysis of the peak pressures to be used for structural design of the walls of the reactor cavity, steam generator subcompartments, and pressurizer subcompartment using as input the Lass and energy rates given in Preliminary Safety Analysis Report Supplement No. 20, response 222.5.

The applicant submitted the reanalysis which resolved this item (Section 6.2.2).

(19) We required that the main steam and feedwater valve house be designed to withstand the environmental ef fects of a break in a main steam or feedwater line without causing a loss of safe shutdown capability. The applicant sub-mitted a commitment to perform analyses and design the valve house to ensure structural integrity. Our review of this matter is incomplete. This matter remains an outstanding item ('.,ection 3.6.2).

(20) We require additional information on the new minimum flow distribution factor andanevaluationofitsimpactynthethermalandhydraulicanalysis. The applicant rubmitted additional information which the staff is reviewing. This matter remains an outstanding item (Section 4.4.3).

(21) On its own initiative the applicant revised the criteria for selecting pipe break locations for high and moderate energy pipe breaks outside containment.

This information was presented in Supp1raent No. 24 to the Preliminary Safety Analysis Report. We recuired that these )ipe break locations be selected consistent with Branch Technical Position MEB 3-1,

• Postulated Break and Leakage Locations in Fluid System Piping Outside Containment." In Supplement No. 26 the applicant again revised these criteria, this time making them 739013 1-5

consistent with the Branch Technical Position MEB 3-1 and thereby resolving this item (Section 3.6.2).

(22) The applicant advised us that the double-ended main steam line break inside containment appeared not to be tt.e break size that results in the highest containment temperature. We r,equired the applicant to provide an taalysic of a spectrum of steam line break sizes to identify the highest containment pres-sures and temperatures. The applicant sLbmitted a respon,e which the staff is reviewing. This matter remains an outstanding item (Section 6.2.1).

(23) We require that the plant design have the capability for transferring heat from the reactor to the environment during the transition from normal reactor operating conditions to cold shutdown, using only safety grade systems, and assuming (a) only offsite or only onsite power is available, and (b) the most limiting single failure has occurred. The applicant submitted a response which the staff is reviewing. This matter remains an outstanding it.em (Section 5.4.5).

(24) W have determined that a portion of the fuel building is not designated a seis:ic Catean y I structure. We will require conformance of this structure and its associated ventilation system with the Comr.ission's rules and regula-tions. This matter remains an outstandina item (Sections 3.2.1 and 9.4.3).

In summary, nine items remain outstanding before we can complete our review of the Greene County Nuclear Power Plant. We will require retolution of all of these items prior to a decision on issuance of a construction permit. Our review and conclusions on these items will be described in a future supplement to the Safety Evaluation Report.

I 10 Significant Changes to the Application Found Acceptable by the Staf f The following significant changes to the application for the Greene County Nuclear Power Plant were made on the applicant's initiative since the Safety Evaluation Report was issued. The staff evaluations and acceptance of these changes are described in the noted sections of this supplekent.

(1) The spectrum of design basis missiles to be used foi the design of the tornado missile barriers nas beea changed froc the Revision 0 to the Revision 1 spectrum as proposed tur Section 3.5.1.4 of NUREG-75/087, " Standard Review Plan for the Review of Safety Analysis Reports f or Nuclear Power Plants" (Section 3.5.1).

(2) Ti.e ceismic analysis method for the soil to structure interaction has been changed from a lumped mass spring approach to a fixed base approach for seismic Category I structures founded on undisturbed rock (Section 3.7.2).

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(3) The spent fuel storage system has been changed to increase the number of spent fuel assemblies which May be stored on the site during the life of the plant (Sections 9.1.2 and 9.1.3).

(4) The reactor plant service water system has been modified to add a half capacity, nonsafety grade pump to each of the two trains for use during normal plant operation (Section 9.2.1).

(5) The design of the ultimate heat sink cocling towers has been revised to utilize a single water storage basin under the mechan cal draft cooling to ;rs (Section t.2.3).

(6) The applicant's technical organization for design of the facility has been revised to abolish the position of Manager-Thermal Power Generation; the Principai Nuclear Engineer new reports to the Assistant Chief Engineer -

Projects (Section 13.1).

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2.0 SITE CHARACTERtSTICS 2.1 Geography and Demography 2.1.2 Exclusion Area Control In the Safety Evaluation Report we reviewed the applicant's right to acquire full ownership of all site property by eminent domain proceedings under the laws of the State of New York. We concluded that the applicant's responses had narrowed the concern to a nuestion of the legal interpretation of the above acquisition authority.

We requested a legal memorandum by the applicant analyzing New York State law and prior court decisions on this question. The applicant responded in a letter dated August 5, 1977, setting forth the legal basis for the taking of land in fee or otherwise. In a letter dated January 6,1978, the applicant demonstrated that it had employed these methods of acquisition and that their lega ity has been upheld t

by the courts of the State of New York.

Based on our review of the information provided by the applicant, we conclude that the applicant has the legal means to acquire by eminent domain or other proceedings the full ownership of all site property needed to establish the exclusion area described in the application. We find the above two letters an acceptable response to our request and consider this matter, outstanding item (1) in Section 1.8, to be resolved.

2.2 Nearby Industrial, iransportation and Military Facilities In the Safety Evaluation Report, n reported on the applicant's analyses of the ground shock which might be caused b) six possible accidental explosions of explo-sive material transported and stored for quarrying operatiens in the vicinity of the site. These analyses were provided in Supplement No. 13 to the Preliminary Safety Analysis Report and revised in Supplement No. 18.

Upon review of these analyses, we requested additional information on the relationship between the ground motions and the plant design response spectrum for earthquake motion.

In Supplement No. 21 to the Preliminary Safety Analysis Report, the applicant provided additional information describing the resulting response spectrum at the plant site for the worst case accident, the postulated detonation of an explosive laden truck on relocated State Route 9W at about 1600 feet from the control build-ing. The resulting response spectrum was described as being lower than the response spectrum for the cperating basis earthquake. In addition, the response spectrum for the worst case routine quarry blast was described as being lower than the the response spectrum for the operat ng basis earthquake. The applicant later 733016 2-1

provided copies of a report, " Evaluation of Ground Motions Induced by Postulated Explosions in the Vicinity of Greene County Nuclear Power Plant f in support of the results for accidental explosions given a Supplement No. 21 to the P eliminary Safety Analysis Report.

In December 1977, the applicant provided copies of blast monitoring records furnished by the Alpha Cement Company. Upon further study of these records, the appli ant advised us in a letter dated January 30, 1978, that the' accuracy of the data pre-sented in the Alpha Cement Company records was subject to question and that the applicant was initiating a blast monitoring program to obtain site specific seismo-graphic data. Seismometers were located on a number of rock outcrops at the plant site and in the general area to record ground motion resulting from blasts at adjacr

• quarries. These site specific blast monitoring records were submitted by a letter dated July 25, 1978, and we are reviewing this information. We will complete che review of this matter, outstanding item (2) in Section d, and report our conclusion _in a future supplement to the S.aety Evaluation Report.

2. 4 Hydrolog 2.4.5 Design Basis Groundwater Levet In Supplement No. 23 to the Preliminary Safet.inalysis Report, the applicant updated the Preliminary Safety Analysis Report to include additional piezometric data collected at the site. These data are from 26 piezometers and have been collected through May 1977. The hydrologic information contained in Supplement No. 23 to the Freliminary Safety Analysis Report does not alter any related conci -

sions or information contained in the Safety Evaluation Report.

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3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMP 0NENis 3.2 Classification of Structures, Components, and Systems 3.2.1 Seismic Classification We have determined that a portion of the fuel building which is used for the decontamination and shipment of the spent fuel cask and for the receipt of new fuel is not designated a seismic Category I structure. We will require conformance of the fuel building and its associated ventilation system with the Commission's rules and regulations. botil the applicant either demonstrates that the fuel building and ventilation system satisfies the requirements of 10 CFR Part 100, or makes acceptable design changes such as meeting the requirements of Regulatory Guide 1.29, we consider this matter, item (24) in Section 1.8, to be an outstanding item. We will report on the tesolution of this matter in a future supplement to the Safety Evaluation Report.

3.5 Missile Protection 3.5.1 Missile Selection and Protection Criteria (3) Tornado Missiles triterion 2 of the General Design Criteria requires that structures, systems and cc,,moents important to safety be designed to withstand the effects of natural phenomena such as tornadoes without loss of capability to perform their safety functions. In consideration of this requirement, we notea in the Safety Evaluation Feport that we have specified a spectrum of selected design basis missiles which we conclude are acceptable for the design of tornado missile barriers in NUREG-075/087, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 3.5.1.4.

We further noted that Section 3.5.1.4 of NUREG-075/087 was being revised to include the addition of a second spectrum of selected design basis missiles, referred to

- the Revision 1 spectrum, which we conclude is also acceptable for the design of tornado missile barriers. Our position on this Revision 1 spectrum is set forth in our letter to the applicant dated November 11, 1976.

At the time the Safety Evaluation was issued, the Preliminary Safety Analysis Report showed that the facility walls and roofs of safety-related structures would be designed for the initial (Revision 0) spectrum of missiles, which we found to be acceptable. In Supplement No. 25 to the Preliminary Safety Analysis Report, the applicant changed the spectrum of design basis missiles to be used for the design of tornado missile barriers. The applicant now states that the 733018 3-1

facility walls and roofs of safety-related structures will be designed to the spectrum of design basis missiles given in the staff posiiion 31(.6 (Revised) as set forth in our letter to the applicant dated November 11, 1976. We conclude this change in design criteria by the applicant conforms tc the staff position identified above and the Revision 1 spectrum of design basis tornado missiles as proposed for Revision 1 to Section 3.5.1.4 of NUREG-75/087, " Stand-ard Review Plan for the Review of Safety Analysis Reports for Nuclear Fower Plants." Therefore, we find the Supplement 25 spectrum of aesign basis tornado missiles to be acceptable.

In the Safety Evaluation Report we identified as outstanding item (3) a require-ment that the ultimate heat sink cooling towers must be designed to protect against vertical tornado missiles. At the time our repoit was issued, the applicant stated that the cooling tower structures for the ultimate heat sink would be designed to withstand tornadoes and horizontal tornado missiles, but that no attempt would be made to design the cooling towers for vertica' tornado missiles. We found this design was not in conformance with Regulatory Guide 1.117, " Tornado Design Classification," or Branch Technical Position AAB 3-2, " Tornado Design Classif W nn."

These two documents state, "The physical separation of redundant or alternative structures or components required for the safe shutdown of the pl e t is generally not considered an acceptable method for protecting against tornado efrects, including tornado-cenerated missiles." Therefore, we informed the applicant that the proposea design was unacceptable and the cooling tower structures must be designed to withstand vertical missiles.

In Supplement 22 to the Preliminary Sa'fety Analysis Report the applicant committed to design the cooling towers to withstand the effects of tornadoes, and horizontal and vertical tornado missiles. Tower internals will be fully protected by the structure from tornado missiles. With this commitment, we conc'lude that the design of the ultimate heat sink cooling towers will be consistent with the above quoted criteria for protecting against torna o effects, including tornado generated missiles. Therefore, we find the tornado missile protection design for the ultimate heat sink cooling towers to be acceptable and consider this matter, outstanding item (3) in Section 1.8, to be resolved.

3.5.2 Barrier Design Procedures In the Safety Evaluation Report, we identified as an outstanding item a requirement that the wall thickness be increased for concrete tornado missile barriers to be consistent with the staff position 130.22 given in our letter to the applicant dated September 29, 1976. We also advised the applicant that the concrete strength niven in our position 130.22 refers to the 28-day cure strength of the concrete.

The use of the 90-day cur-.trength in the design of the barriers would be an 733019 3-2

acceptable methu/ of conforming to the above staff position only if pozzolan material will be used in tre concrete mix.

The applicant has not proposed the use of this type of concrete. This matter is listed as outstanding item (4) in Section 1.8 of this supplement.

In Sucplement No. 25 to the Preliminary Safety Analysis Report, the applitant adopted the Revision 1 spectrum of design Lasis missiles to be used for the ae. g.,

of tornado missile barriers as described in Section 3.5.1 (3) above. In the same Supplement No. 25 the applicant adopted the associated staff position 130.22 (Revised) given in our letter to the appl' ant dated November ll, 1976. The applicant states that the external reinforced concrete missile barriers will be designed to include wall and roof thickness criteria in accordance with thc attachment to staff position 130.22 (Revised), " Attachment 130-1, Wall and Roof Ihickness Requirements to Resist the Effects of Tornado Hissile Impact." D Supplement 26, the applicant further stated that the concrete strengths listed in "Attacheent 130-1" are considered to be the 28-day cure strength of the concrete. We conclude these commitments are consistent with the staff positions identified above and as set forth in the Safety Evaluation Report. On this basis, we find the criteria for wall and roof thickness requirements to resist the effects of tornado missile impact to be acceptable, and we consider this matter, outstanding item (4) in Section 1.8, to be resolved.

At the time the Safety E v aluation R3 port was issued, the applicant,2roposed to utilize concrete ductility factors larger than 10 in the design analyses for reinforced-concrete missile barriers. The applicant contended that the furthcoming Stone and Webster report, "Missle Barrier Interaction" would justify its position that the criteria set forth in staf f pnsition 130.23 given in our letter to the applicant dated September 29, 1976, were unnecessarily conservative. In the Safety Evaluation Report, we found the applicant's position on the use of ductility factors larger than 10 for tne design of concrete missile barriers to be unacceptable.

In Supplement No. 22 of the Freliminary Safety Analysis Report the applicant revised its position to adopt the staff criteria as set forth in our position 130.23 refer-enced above. The applicant will limit missile barriers to ductility factors not greater than 10 and will design members subjected to simultaneous bending and compression loads according to the ductility ratio criteria as follows:

(1) For beam.olumn members where the compressive load is equal to or less than one-thi! of that which would produce balanced conditions (i.e., P, the axial b

load re alting in a balanced condition; 0.lf 'A, the specified compressive c g streng+ of concrete multiplied by the gross area of the section; whichever is smaller) *he allowable ductility is 10.

(2) For beam-column members where the design is controlled t,y compression, the allowable ductility is 1.3.

7390'40 3-3

(3) For members which are between the cases of items (1) and (2), above, the ductiltty ratio should te taken as decreasing linearly from 10 to 1.3.

We conclude that the applicant's commitment as given in Supplement No. 22 to the Preliminary Safety Analysis Report is cnnsistent with the criteria given in the above staff position. On this basis, we find this commitment to be acceptable and consider this matter, outstanding item (5) in Section 1.8, to be resolved.

The Stone and Webster report entitled "5WECO-7703 (EMIR-801), " Missile Barrier In'.eraction," was submitted for our review as a topical report in September 1977.

The applicant states that if discussions between the staff and Stone and Webster on this report lead to staff acceptance of ductility factors greater than 10 for barrier design, the applicant may elect to design missile barriers for this application based cn the results of these discussions. Upon completion of our review and acceptance of this report, the applicant may request approval to utilize the design criteria found acceptable in this topical report.

3.6 Protection Against Dynamic Effect., Associated with the Postu M ed Rupture of Piping 3.6.2 Prctection Against Dynamic Effects Associated with the Postulated Rupture of Piping Outside Containment in the Safety Evaluation Report, we reported that the criteria to be followed in design of piping systems and associated nmponents and structures are consistent with the criteria contained in Branch Technical Positions APCSB 3-1, " Protection Against Postulated Piping Failures in Fiuid Systems Outside Containment," and MEB 3-1, " Postulated Break and Leakage locations in Fluid System Piping Outside Containment." After issuance of the Safety Evaluation Report, the applicant modi-fied the design basis pipe break criteria in Supplement No. 24 to the Preliminary Safety Analysis Report. We reviewed thesc changes in the Preliminary Safety Analysis Report and concluded the pipe break criteria were not consistent in certain specifics with Branch Technical Position MEB 3-l.

We informed the applicant that the revised pipe break criteria were not completely acceptable in our letter of March 16, 1978, and identified the changes required to resolve this issue. The applicant responded to our position on this matter by making the identified changes in aupplement No. 26 to the Preliminary Safety Analysis Report. With these changes, we conclude the pipe break criteria as presently given are consistent with Branch Technical Position 3-1.

On this basis, we find them to be acceptable and consider this matter, outstanding item (21), to be resolved.

By a letter dated Novemoer 8, 1977, we advised the applicant that we are preparing a revis.on to Sections 3.6.1 and 3.6.2 of NUREG 75/087," Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." The revision will establish a position that the valve room for the main steam line valves and feedwater line valves be designed to withstand the environmental effects of a main steam or 739021 3-4

feedwater line break equivalent to the flow area of a single ended pipe rupture.

Venting is an acceptable means of limiting environmental conditions within the valve room.

We furtM r advised the applicant that in the interest of avoiding potentially costly plant modifications later, we are establishing the following position at this time for the Greene County Nuclear f ewer Plant.

We require that this isolation valve compartment for the u*eene County Nuclear Power P! ant be aesigned to withstand the environment <l effects of a main steam or feedwater line break equivalent to the flow area of a single ended pipe rupttre.

In conjunction with the above, we require that a subcompartment pressure analysis be provided to confirm that the valve compartment or room, as designed, will with-stand the envi onmental effects of a break in this area.

In addition, we require that any equipment required tar safe shutdown which will be located ir, t' valve room, including the main ste n isclation valves and operators, be qualified to be capable of operating in the environment resulting from the postulated pipe rupture.

We requested additional information on this matter. In Supplement No. 25 to the Preliminary Safety Analysis Report, the applicant responded that the plant will have two completely separated main steam and feedwater valve houses, and tnat these valve houses will be structurally designed to withstand the pressure incre.se that wculd resuit from a main steam or feedwater line break equivalent to the flow area of a single-ended pipe rupture. Vents to the atmosphere will be provided.

'he applicant will provide in the Final Sr.fety Analysis Report, a subcompartment pressure analysis and description of all desigi changes associated with ensuring that the structural integrity of these valve houses will be maintained. Our review of thii matter is incomplete. We will corcplete the review of this matter, outstandi7g item (19) in Section 1.8, and report our conc lusion in a future supplement to the Safew Evaluation Report.

Seismic Design Seismic System and Subsystem Analysis In the Safety Evaluation Report, we repc.rted that tae applicant will use the lumped mass spring approach to evaluate the soil structure interaction for seismic Category I buildings. In Supplement No. 25 to the Preliminary Safety Analysis Report. the applicant revised the method of analysis for the containment-annulus building.

ihe reinforced concrete f oundation mat f(r this building will be supported completely on bedrock, and the applicant has ad(ptEd the fixed base approach to evaluate the soil-structure interaction for this t Jilding. Ihe use of the fixed base approacn for a struiure supported completely on bedrock is consistent with our criteria as given in Section 3. 7.2, NUREG-75/087, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Fower Plants," and we find this change in the design analysis'to be acceptable.

739022 3-5

Nonmental Design of Mechanical and Electrical Equipment 3.11 r

in the Safety Evaleation Report, we identified as an outstanding item a requirement that the applicant commit to provioe a technical report for the balance-of plant Class 1E electrical, instrumentatic., and control equipment within one year af ter issuance of the construction permit. This report should identify: (1) how each piece of Class lE equipment has been, or will be, qualified; (2) the acceptance criteria; (3) test procedures; and (4) test results, if available or a schedule for submittal of test results. The applicant document-M in Supplement No. 22 to the Preliminary Safety Analysis Report a commitment to provide a technical report addressing the above for the balance-of plant component environmental qualification program approximately one year after receipt of the construction permit. In addition, the applicant will meet periodically with the staff to provide status updates and additional information as it becomes available.

We conclude that the applicant has described the development program require 1 to resolve our concern about the operability of the balance-of plant mechanical and electrical equipment in an accident environment, and committed to a schedule which shows that tne adequacy of this equipment will be resolved prior to the date for a decision on issuing an operating license. We have determined that the required additional information is of the type that, in accordance with the provisions of Section 50.35 of 10 CFR Part 50, can be lef t for later consideration. Therefore, we find the environmental design of the balance-of plant eqJipment to be acceptable, and consider this matter, outstanding item (6) in Section 1.8 to be resolved for the construction permit stage of review.

3-6

4.0 REACTOR 4.2 Mechanical Desico 4.2.1 Fuel Design

4. 2.1. 3 Mechanical Performance On page 4-9 of the Safety Evaluation Report, we stated that we had requested a topical report on the cycling and fatigue analyses planned by Babcock and Wilcox.

This topical report was to have been submitted for review at least one year prior to the date of submittal of a Final Safety Analysis Report on a plant usiig Mark C fuel. We have subsequently been notified, however, that this information is con-tained in the Final Safety Analysis Report for the Bellefonte Nuclear Plant (Docket Nos. 50-438/439). Therefore, we have withdrawn our request for the topical report, and are currently reviewing the Bellefonte information on cycling and fatigue. The results of our evaluation of the Bellefonte information will be applicable to the Greene County Nuclear Powei Plant. This earlier review will provide sufficient time, before fuel fabrication, to incorporate any design changes that may be needed as a result of our review. Since this subject will not be resolved by a generic topical repnrt, this matter should be addressed in the Final Safety Analysis Report.

4.4 Thermal and Hydraulic Design 4.4.3 Thermal and Hydraulic Analysis The Safety Evaluation Report described our review of the thernal and hydraulic analyses performed bj Babcock and Wilcox in Section 4.4.3.

We noted the 0.99 minimum flow distribut ion factor utilized by Babcock and Wilcox and stated "that the minimum flow factor of 0.99 is acceptable at the preliminary design review stage."

In a letter dated December 20, 1977 (J. H. Taylor to S. A. Varga, NRC),

Babcock and Wilcox advised us that the flow distributian factor presented as 0.99 in Topical Report BAW-10025P, " Reactor Vessel radel Flow Tests for 205-FA Core,"

should be reduced to 0.965.

Using the parallel channel flow stability analyses described in the Safety Evaluation Repo;t, this flow reduction could result in a penalty in the departure f rom nucleate be ing ratio.

In a letter dated February 2,1978, we informed the applicant of our concerns in this matter and identified this as a new review item. We required additional information on the new minimum flow distribution factor and in evaluation of its impact on the thermal and hydraulic analysis.

In the meantime, Babcock and Wilcox submitted additional informat. q on the impact and significance of the change in this minimum flow distribution fa.'.or on the 739024 4-1

BSAR-205 Stanoard Reference Plant (Docket No. 3TN 50-561). By a letter dated May 29, 1978, we requested additional information on the Greene County Nuclear Power Plant. The applicant submitted additional information on this matter in Supplement 28 to the Preliminary Safety Analysis Report. We will complete the review and report on the resolution of this matter, outstanding item (20) in Section 1.8, in a future supplement to the Safety Evaluation Report.

733025 4-2

5.0 REACTOR COOLANT SYSTEM 5.4 Component and subsystem Design 5.4.5 Decay Heat Removal System In the Safety Evaluation Report we concluded that tl.e p;cvisions for isolating the decay heat removal system from the high pressure reactor coolant system were not in conformance with staff criteria on the mitigation of decay heat removal system piping fa:1ures, as cescribed in Section 3.6.1 of NUs " 75/087, " Standard Review Plan for the Review of Safety Analysis Reports for Nur. car Power Plants." h r that reason, we found the design to be unacceptable and identified this as an outstand-ing

.em.

The design reviewed in the Safety Evaluation Report consisted of two parallel suction lines each with two motor-operated valves located inside containment. The power supply arrangement to these isol3 tion valves was such that loss of power to one decay beat removal train negated the ability to close either of the two isola-tion valves in that train.

The applicant has changed the design of the decay heat removal system to replace one of t:1e two electric motor-operated isolation valves in each decay heat suction line with an electrohydraulic / stored energy valve. With the new design, we find that all functional requirements have been satisfied. Assuming a loss of a vital bus, the decay heat removal system will be able to be actuated and 't can be isolated in case of a postulated failure in the decay heat removal suction ?ine. With this modification, we find the Jecay heat removal system acceptable because it conforms to the staff criteria described in Section 3.6.1 of NUREG 75/087, and it meets the requirements of Criterion 34 of the General Design Critecia. We consider this matter, outstanding item (7) in Section 1.8, to be resclved as further described in Section 7.6.1 of this supplement.

In the Safety Evaluation Report, we reviewed the decay heat removal system and concluded that provisions for prolonged operation (20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />) in the hot shutdown condition using only safety grade equipment was acceptable. However, the staff has been considr-ing, on a generic basis, whether the capability should be provided for transferring heat from the reactor to the environment during the transition from normal reactor operating conditions to cold shutdawn, using only safety grade systems, and assuming (1) only offsite or only onsite power is available and (2) the most limiting single failure has occurred.

733026 5-1

A final decision o..

this matter has now been reached by the staff. This decision establishes that spe"ific technical requirements shall be incorporated in overall plant designs for which constructioq permit or Prelir.;inary Design Approval applica-tions were docketed before January 1,1978, and for which operating license issuance is expected af ter January 1,197?

It is includes, among others, the Greene County Nuclear Power Plant. In a letter dated May 29, 1978, we advised the applicant of this change in our reouirement for sa"ety grade equipment to achieve cold shutdown conditions, and requested a commitment to comply with certain staff cold shutdown criteria. The applicar t submitted adc itional information on his matter in Supole-ment 28 to the Preliminary Safety Analysis Neport. We will complete our review and report our conclusion on this matter, outstanding item (23) in Section 1.8, in a future supplement of the Safety Evaluation Report.

730027 5-2

6.0 ENGINEERED SAJETY FEATURES 6..

Containment Systems 6.2.1 Containment Functional Desion.

Subsequent to the issuance of the Safety Evaluation Report, the applicant informed the staff that the analysis for the double-ended main steam line break submitted in the P eliminary Safety Analysis Report may not be the break size that results in the highest containment atmospheric temperature. Therefore, we required the appli-cant to re-do the analysis for a spectrum of break sizes to identify the break that results in the highest containment temperature. The applicant submitted a response which the staff is reviewing. This requirement is identified as outstanding item (22) in Section 1.8 of this supplement. We will report on the resolution of this matter in a future supplemer.t to the Safety Evaluation Report.

6.2.2 Containment Subecmpartment Design in the Safety Evaluation Report we required a commitment from the applicant to provide the results of additional nodalization studies of dynamic asymmetric pres-sure loads within the r) actor cavity, steam generator ar.J pressurizer compartments to provide conservative loads for tha design of these component supports. Further-more, we requested a schedule commitment for the submittal of this information as a post-construction permit item.

In Supplement No. 22 to the Preliminary Safety Analysis Report, the applicant responded that the results of additional nodaliza-tion studies of dynamic asyksetric pressure loads within these compartments will be provided, and that the methodology tn be used will be resolved with the NRC staff.

In addition, there is an ongoing generic task within the NRC to further investigate the analytical methods and assumptions used for such analyses. The objective of the task is to develcp guidelines for both the assumptions and analytical models to be used.

We will require the applicant to comply wth any analytical or design requirements resulting from the resolution of the NRC's Gen ic Task A-2, Asymmetric LOCA Loads. The completion of the task is presently scheduled for February 1979. In combination with the NRC generic program, we find the applicant's commitment for submittal of the additional nodalization studies to be acceptable.

We consider this matter, outstanding item (8) in Section 1.8, to be satisfactor ly i

resolved for the construction permit stage of review.

In the Safety Evaluation Report > reported that we had requested noding sensitivity studies of the applicant. We had requested similar studies on the Sabcock and Wilcox BSAR-205 Stanaard Reference Design (Docket No. STN-50-561).

The BSAR-205 standard plant is similar to the Greene County Nuclear Power Plant but is designed frr a slightly higher power level. Babcock and Wilcox performed noding studies H

NM

which showed that a conservative solution is obtained for the calculated mass and entrgy release rates. We accepted those noding studies in our review of the BSAR-205 standard plant. The applicant submitted the mass and energy release data calculated for '.he 85AR-205 standard plant in Supplem nt No. 20 to the Preliminary Safety Ardlysis Report. The Safety Evaluation Report stated that we have determined that the BSAR-205 nass and energy release rates are acceptable for use on the Greene County Nuclear Power Plant and this resolves our requirement for additional noding sensitivity analyses to test calculstional methods.

At the time the Safety Evaluation Report was issued, our review of Supplement No. 20 to the Preliminary Safety Analysis Report was incomplete. Upon continued review of the information included in supplement No. 20, we questioned whether the mass and energy release rates sucmitted for the noding sensitivity analyses could impact the design pressure for tne subcompartment walls. The BSAR-205 analyses included a revision to the method of analysis during the subcooled period of blow-down whi3h showed an increase in the rates of mass and energy icleased in the early part of blowdown, and gave the mass and energy flow rates for smaller increments o, time than previous analyses.

Our letter to the applicant dated N)vember 8, 1977, advised that we required a reanalysis of the peak pressJres to be used for structural design of the walls of the reactor cavity, steam generator succompartments, and pressuri2er subcompartment using as inout the mass and energy rates given in Supplement No. 20 to the Prelimi-nary Safety Analysis Report.

The applicant revised the mass and energy release rate data for calculating the peak pressures to be used for the structural design of the subcompartment walls to adopt the values calculated for the 85AR-205 design. This revision in the mass and energy release data also reflects a re/ision in the method af analysis. The CRAFT 2 code is used tc calculate these mass and energy release rates. Break flow is maximizad by use of the Burnell criticil flow correlation when the primary coolant is subcooled and by use of the Moody critical flow correlation when the primary coolant is saturated. We have determined that this model is conservative by comoar-ison to experimental data such as that described in TREE-NUREG-1006, "A Study of Critical Flow Predication for Semi-Scale MOD-1 Loss-of-Coolant Accident Experiments."

We conclude that the revised method of analysis will produce tcnservative mass and energy release rates for the subcompartment analyses, and we have determined that the BSAR-205 mass and energy release rates are acceptable for use on the Greene County Nuclear Power Plant.

The applicant also redid the analyses for the reactor cavity, the steam generator subcompartments and the pressurizer subcompartment based on the revised mass and energy release rate data. Based upon our evaluation of the applicant's reanalysis and our confirmatory calculations, we conclude that the results of the applicant's 6-2 7 MOM

analyses submitted in Supplement Nos. 23 and 25 to the Preliminary Safety Analysis Report are acceptable for use in the structurai design of the reactor cavity, the steam generator subcompartments and th pressurizer subcompartment. We consider this matter, outstanding item (18) in Section 1.8 to be resolved.

6.2.5 Containment Isolation Systey The applicant has stated that it plans to operate the containment purge system during the reactor operating mode of hat shutdown, and has committed to isolate the system during the other reactor operating modes requiring containment integrity.

We informed the applicant of our position that purge system performance should satisfy Branch Technical Position CSB 6-4, " Containment Purging During Normal Plant Operation," if the system is to te usea during the hot shutdown mode. The informa-tion provided by the applicant in respanse to our position was incomplete and, therefore, we requested additional information on the use of the containment purge systam during the hot shutdown mode of reactor operation. The applicant submitted additional information on this matter in Supplement Nos. 21 and 26 to the Preliminary Safety Analysis Report. At this time, our review of this additional information is not complete. We will complete our resiew of this matter, outstanding item (9) in Section 1.8, and report on its resolutien in a future supplement to the Safety Evaluation Report.

5.2.6 Combustible Gas Control System The Safety Evaluation Report deferred i conclusion on the acceptability of the hydrogen generation analysis until the applicant's evaluation of the emergency core cooling :,, stem aerformance il acccrdance with Appendix K to 10 CFR Part 50 was approved. Siace then, the emergency care cooling system evaluation has been approvef and the applicant has rev: sed the hjdragen generation analysis. The ne. h drogen y

analysis, which was submitted in $uoplement No. 21 to the Preliminary Safety Analysis Report, is based upon a rive percent zirconium meta! -ater reaction. This five percent reaction is conservative with respect to the reCCmmendations of Regulatory Guide 1. 7, " Control of Combustible Gas Concentraticns in Containment f ollowing a Loss-of-Coolant Accident," and we find it to be acceptable.

Since the Safety Evaluation Report was issued, the staff developed a question concerning the corrosion rates of mate *ials used to determine the rate of hydrcgen generation within the containnent aad the experimental data upon which these design values were based. The applicant responded in Supplement No. 26 to the Preliminary Safety Analysis Report. At this time, our revie of this additional informatiun is not complete and this matter, outstanding item (10) in Section 1.8, remains unre-solved as restated. We will comple:e our review and repcrt on the resolution of this matter in a future supplement to the Safety Evaluation Report.

7D3030 e-,

6.3 Emergency Core Cooling System 6.3.3 Discussion of loss-of-Coolar.t Accident Analysis The applicant submitted analyses for the Greene County Nuclear Power Plant to demonstrate acceptable consequences for a postulated loss-of-coolant accident and to assure emergency core cooling system performance adequacy.

The applicant has cited Babcock and Wilco < generic studies presented in Topical Report BAW-10102, "ECCS Evaluation of Babcock and Wilcox 205 FA NSS," Revision 2 and 3, to shcw that the most limiting break for the Greene County Nuclear Power Plant is an 8.55-square foot double-ended guillotine rupture of a reactor ccolant pum discharge pipe, having a discharge coefficient of 1.0.

The studies considered large and small breaks and were performed with models described in Topical Report BAW-10104, " Babcock and Wilcox's Emergency Core Cooling System Evaluation Model,"

Revisions 1 and 3, which conform to Appendix K of 10 CFR Part 50, and have been approved by the staff.

The applicant submitted a loss-of-coolant accident analysis for the Greene County Nuclear Power Plant of the most limiting break identified above and calculated with the model described in 8AW-10104, Revision 3, which coaforms to Appendix K of 10 CFR Part 50.

Containment parameters used are the same as those found acceptable in Section 6.3.

of the Safety Evaluation Report. Results from this analysis are 2059 degrees Fahrenheit peak cladding temperature, 4.2 percent maximum local metal-water oxidation, and corewide oxidation less than cne percent. These reported results are within the limits specified in Section 50.46 of 10 CFR Part 50 for peak cladding temperature, local oxidation, and corewide oxidation of 2200 degrees Fahrenheit. 17 percent, and one percent, respectively.

The analysis also indicates a cladding temperature turnaround, with maximum fuel channel blockage of 59.1 percent. Therefore, for these conditions the core will remain amenable to cooling, and the emerpncy core cooling system design for long-term cooling will ensure that Continued cooling will be provided to the core, satisfying the requirements of Section 50.46(b) of 10 CFR Part 50. We find the results of these analyses acceptable because they conform to the acceptance criteria of Section 50.46 of 10 CFR Part 50.

We consider this matter, outstanding item (11) in Sect 8, to be resolved.

730031 6-4

7. 0 INSTRUMENTAf!ON AND CCNTROLS

7. 6 All Other Systems Required for Safety 7.6.1 Decay Heat Removal System In the Safety Evaluation Report, we reported that there was insufficient informa-tion in the Preliminary Safety Analysis Report to determine the power assignment to the valve motor operators for the two main decay be,t removal system suction lines from the reactor ccolant system. We were onable to establish that the decay heat removal system could be isolated from the high pressure reactor coolant system in ccmbination with a single failure in the power sources for these valves.

We con-cluded the design did not meet our criteria for mitigating consequences of pipe break s within this sy stem as descrit;ed in Section 5.4.5.

In Supplement Nos. 21 and 26 to the Preliminary Safety Analysis Report, the appli-cant modified the design. The modified design will permit isolation of the decay heat removal system as.,uming the most limiting single failure. Each decay heat removal system suction line from the reactor coolant system will be provided with two isolation valves in series inside the containment. The upstream v31ve will be electric motor operated and the downstream valve will be electro-hydraulic to open, stored energy tc close. The electro-nydraulic/ stored energy valves require alter-nating current electrical energy to open and will automatically close upon loss of Each of the redundant suction flow paths will be supplied from a separate power.

essential bus to ensure that a single failure within the essential power supply system cannot prevent proper operation of the decay heat removal system.

The applicant did not identify the, specific electric-hydraulic / stored energy v31ve design to be used for this application.

Therefore, duri g the operating license n

stage of review we will require design and functional information on the selected valve design to demonstrate its operational reliability to perform its safety function. To meet this objective, the applicant will submit information on the qualification program for accident environmental conditicos as described in Section 3.l! of this supplement.

We conclude that the applicant has described the development program required to resolve our concern about operability of this valve design in an accident environ-ment, and cownitted to a schedule which shows that adequacy of the design will be resolved prior to the date for a decision on issuing the operating license. We have determined that the required additional information is of the type that, in 739032 7-1

accordance with the provisions cf Section 50.35 of 10 CFR Part 50, can t;e lef t foi later consideration. Therefore, we find the modified design of the decay heat removal systsm tu be acceptable, and consider this matter, outstanding item (7) in Section 1.8, to be resolved.

733033 7-2

8.0 ELECTRIC POWFR SYSTEM 8.2 Offsite Power Systems In the Safety Evaluation Report, we identified a concern that adverse offsite grid frequency conditions might result in reactor coolant pump deceleration more rapid than the rate used in the applicant's loss of reactor coolant flow analysis. The applicant stated in the Preliminary Safety Anai, sis Report that the grid frequency limiting condition is 56 Hertz and for f reqce.Cies above this value there are no safety implications for aly frequency decay rate.

The applicant also stated that for frequency decay rates of 2.3 Hertz per second or less, the power to flow trip will protect the core f rom experiencing departure f ro,a nucleate boiling. The applicant further stated that the atove limiting conditions would never be reached in the New York Power Pool system.

We concluded that there was insufficient information in the Preliminary Safety Analysis Report to support the above statements, and we requested additional informa-tion on the grid underfrequenc) and frequency decay rate as they relate to the reactor coolant pump deceleration.

The applicant provided the following commitment on the Greene County Nuclear Power Plant.

(1) The Topical Report BAW-10121

",'leactor Protection System Limits and Set Points,"

was submitted by Babcock and Wilcox on February 27, 1978. To the extent that this topical repoct requires revisiuns to existing Preliminary Safety Analysis Hepart analyses, these revisions will be provided.

(2) The applicant will provide on the docket at the Final Safety Analysis Report stige the results of an analysis of grid stability to demonstrate that the worst credible case underfrequency and maximum frequency decay rate for the applicant's grid will be within the correspording criteri.i cescribed in the Preliminary Safety Analysis Re:, ort, modified as necessary by the ' ove topical report.

(3) In the event that the applicant's grid underfrequency and frequency decay rate limits exceed those corresponding values established for the reactor protection system, underfrequency devices will be provided to trip the reactor coolant pump breakers to disconnect the pumps from the distribution system. The trip breakers and the instrumentation and controls will be designed and qualified in accordance with the requirements of IEEE 279-1971 and IEEE STD 308-1974.

8-1 733034

If credit is taken for such a reactor coolant pump breaker trip. it is the staff's position that the reactor colant purp breakers and associated instrumentation and cont ' be located in a seismic Category I structure.

We conclude that these design criteria are acceptable for the construction permit stage of review, and c a sider this matter, outstanding item (12) in Section 1.8, to be resolved.

8.3 Onsite Power Systems 8.3.1 Alternatir-Currsnt Power Systems The makeup /high pressure injection system will contain three pumps, one of which is an installed spare. The three pumps will be connect?d to the 4160 volt essential buses 1A and 18.

In order to provide power to the spare pump f rom either of the two essential buses, the applicant proposed a manual bus transfer in which a single circuit breaker mechanism would be physically moved to the housing associated with the other bus.

During our review, we requested the applican' to provide an analysis to verify that a single failure of this switching circuit would not result in the two engineered safety feature buses being paralleled. The applicant stated it was his intention eliminate the r,eed for the manual transfer device and upon development of alter-nate system arrangements, the Preliminary Safety Analysis Report would be revised to reflect the elimination of the manual transfer devices. The applicant in Supplement No. 22 to the Preliminary Safety Analysis Report eliminated the need for the manual transfer switch by modifying the system design. The spare pump will be permanently installed on the 4160 volt essential bus lA.

We conclude that the proposed design of the onsite alternating current p^wer system hfies the applicable criteria stated in Section 8.1 of the Safety Evaluation si Report. Therefore, we find the above modification to be acceptable, and consider this matter, outstanding item (13) in Section 1.8, to be resolved.

733035 6-2 1

e

9.0 AUXILIARY SYSTEMS 9.1 Fuel Storage and Handling 9.1.2 Spent Fuel Storage in Supplement No. 24 to the Preliminary Safety Analysis Report, the applicant proposed a design change that increases the sto-age capacity of the spent fuel pool from 1.87 cores (384 fuel assemblies) to 4.68 cores (960 fuel assemblies). The increase in storage capacity is achieved by modifying the design of the seismic Category I rack from a modular, multiple-cell structure of 24 spaces in a 4 x 6 array to one with 54 spaces in a 6 x 9 array. The new rack design will have 14-inch centerline spacing. An analysis performed by the applicant and independently verified by us, indicates that the spacing will be sufficient to maintain an effec-tive multiplication facter of equal to or less than 0.95, assuming that unexposed fuel assemblies containing the hignest probabie enrichment (3.5 percent) are loaded in a spent fuel rack and the pool becomes inad,_rtently filled with unborated Based upon the above, we find the change in rack design for the spent fuel water.

storage to be acceptable.

9.1. 3 Feel pool Cooling and purification Systems On changing the number of spent fuel assemblies to be stored in the spent fuel storage pool, the applicant reanalyzed the capability of the fuel pool cooling and purification syttems to maintain acceptable temperature limits within the pool.

The applicant determined that the total heat load resulting from the increased stora9e caDacity did not change the design pool temperatures.

We independertly reviewed the heat load calculations and the capability of the fuel pool cooling and purification systems to maintain temperatures in the spent fuel pool within acceptable limits. Based upon our review, we conclude that the design for the fuel pool cooling and purification systems to accommodate the spent fuel pool storage capacity of 960 fuel assemblies is acceptable.

9.2 Water Syste,5 9.2.1 Reactor plant Service Water System In the Safety Evaluation Report we described the reacto plant service water system as consisting of two trains. Each train will contain two 100 percent capacity, safety grade pumps, associated piping and instrumentation and will be powereri f.om an independent engineered safety features bus. Only one full capacity p;mp is required to safely shutdown the facility or to provide cooling water during a postulated loss-of coolant accident.

rvo y.3 -

$bs) job 9-1

In Supplement No. 25 tc the Preliminary Safety Analysis Report, the applicant modified the design of the reactor plant service water system. An additional half capacity, nonsafety grade pump has been added in each train. With this new arrange-ment, during normal operation, only the nonsafety grade half capacity pump in each train will be operating. The modification does not reduce the safety grade equip-ment available for a ststion shutdown or for mitigating the consequences of a postulated loss-of-coolant accident. We have reviewed this modification and find it to be acceptable since it does not redace the capability of the reactor plant service water system to perform its safety-related functions.

9.2.2 Reactor Plant Component Cooling Water System In the Safety Evaluation Report we reviewed the reactor plant component cooling s ?ter system design and concluded that the portion of the system which supplies cooling water to the reactor coolant pump seals was not acceptable.

Previously the applicant had modified the design of the component cooling water system to the reactor coolant pump bearings to prevent fuel damage from occurring.

However, the applicant did not modify the design to include prevention of a poten-tial breach of the pressure boundary.

In Suppl?msnt No. 18 to the Preliminary Safety Analysis Report, the applicant, in respon,a to our concern thi.t loss of component coolant to the reactor coolant pumps could result in shaft seizure, modified the design to provide safety grade instru-mentation to both the upper and lower motor bearing coolers and pump oil coolers, However, the applicant did not agree to providing safety grade instrumentation to the pump seals. In lieu of providing safety grade instrumentation the applicant, in Supplement No. 25 to the Preliminary Safety Analysis Report, provided an analysis to demonstrate that the component coolant water system is only the backup system.

That is, the seal injection system is the primary source of coolant for the pump seals. We reviewed the analysis and determined that without pump test data we could not verify the analysis. We notified the aM licant of our conclusions concern-ing the analysis.

In further respc7se, the applicant stated in Supplement 27 of the Preliminary Safety Analysis Report that he would perform tests to demonstrate that a complete loss of component coolant to the seals of the reactor coolant pumps for longer than 30 minutes would not result in breach" of reactor pressure boundary. The applicant cited pumps similar in design to the Greene County Nuclear Power Plant coolant pumps that have been successfully tested for time periods longer than 30 minutes without cooling the seals. Successful testing of the reactor ccolant pumps seals is acceptable to us; however, the applicant diu not commit to providing an accept-able alternative if the tests were unsuccessful. This system will remain an out-standing item until the applicant commits to either complying fully with our position 739037 91

or provio <

ec;eptable alternati e for our review and approval. Therefore, s

outstand item (14) in Section. 8, remains unresolved. We will report on the resolution of this matter in a future supplement to the Safety Evaluation Report.

9.2.3 Ultimate Heat Sink In the Safety Evaluation Report we noted that the coolir,' tower structures for the ultimate heat sink were not designed to provide protectio. against vertical tornado missiles, and we found that to be unacceptable. In Supplement No. 22 to the Pre-liminary Safety Analysis Report, the applicant modified the design of the ultimate heat sink to include protection against vertical missiles. We reviewed this modifi-cation in Section 3.5.1 of this supplement and found the tornado missile protection design for the ultimate heat sink cv..ing towers to be acceptacle, and we consider this matter, outstanding item (3) in Section 1.8, to be resolved.

In Supplement No. 25 to the Preliminary Safety Analysis Report, the applicant also changed the design of the ultimate heat sink by locating the redundant cooling towers over a single water storage basin. The cooling tower structure and all components will be designed to withstand the safe shutdown earthquake. The towers will be able to withstand the effects of tornadoes as well as horizontal and vertical tornado missiles, and the tower internals will be fully protected by the structure from tornado missiles. The cooling tower storage basin will be almost entirely below grade, missile protected, designed to seismic Category I requirements, and capable of withstanding each of the most severe phenomeni postulated. The emergency 30-day supply of water will be located entirely below grade. Based on our review, we conclude that this design modification complies with the criteria as stated in Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Pcwer Pla...s."

Therefore, we find this design modification to be acceptable.

9.4 Air Conditionino. Heatirg, Cooling and Ventilation Systems 9.4.1 Control Building Air Conditioning and Ventilation Systems In the Safety Evaluation Report we stated that the control room emecgency pressuri-zation system will include two air supply trains with separate remote outside air intakes. We described these intakes as physically separated a distance of 1400 feet and located 180 degrees apart on opposite sides of the control building. The applicant relocated the remote air intakes for the control room area ventilation system. The intakes are now physically separated a distance of 1650 feet and located

.) proximately 180 degrees apart on opposite sides of the control building.

We conclude that this slight relocation of the remote air intakes will have no impact on the design objectives for the system and find it to be acceptable.

733038 9-3

9.4.3 Fuel Building ventilation System In the Safety Evaluation Report we reviewed the fuel building ventilation system and concluded that the design criteria and bases for this system were acceptable.

We have determined that a portion of the fuel building, as described in Section 3.2.1 of th.s supplement, is not classified a seismic Category I structure. We will require conformance of the fuel building including its ventilation system with the Commission's rules and regulations. Until the applicant either demonstrates that the fuel building and ventilation system satisfies the requirements of 10 CFR Part 100, or makes acceptable design changes such as meeting the ret.,uirements of Regulatory Guide 1.29, we consider this matter, item (24) in Section '.8, to be an outstanding item. We will report on the resolution of this matter in a future supplement to the Safety Evaluatien Report.

7330:39 9-4

10.0 STEAM AND POWER CONVERSION SYSTEM 10.4 Other Features of Steam and Power Conversion System 10.4.5 Steam Generator Heatup System The steam generater heatup system will be desicned to ameliorate two thermal condi-tions during plant stortuo from cold shutdown. First, the system will provide for concurrent warmup of both the shell and tubes tt minimize compressive loading of the steam generator tubes otherwise caused by temperature differences between the hatter tubes and the cooler shell. Tube warmup will occur as the primary coolant is circ" lated by the reactor coelant pumps. Shell warmup will be accomplished by drawing steam from the steam generator through the turbine bypass lines or unloaded turbine to the main condenser. Second, the system will provide for warmup of the water contained in the main feedwater piping to the steam generator. The system will draw the shell-side water through four shell-side drains and pump it into the feedwater line immediately downstream of the outboard isolation valve. This small purrp will circulate and cause the warming of the water in the main feedwater lines between the isolation valve and the steam generator.

Secondary uses of t'is system will include chemical control during cold wet layup, and steam generator solids removal from the shell-side. The system will be shutdown and isolated prior to power operation. In all cases the system will be functional only when the reactor is held in the shutdown condition by the control rods and boron concentration in the reactor coolant. The boron concentration in the reactor coolant will be maintained at the cold shutdown concentrations. Thus, a break in the steam generator heatup system piping will not result in reactor criticality by cooling the reactor coolant.

Isolation of the steam generator heatup system suction line will be provided by the normally closed valves on each of the four drain lines and the single pump suction valve located outside containment. Isolation of the system return lire will be provided by the isolation valve outside containment in the return line to the feedwater line. The steam generator heatup system will be neither a part of the reactor coolant pressure toundary nor connected directly to the containment atmo-sphere. We have determined that the containment isolation provisions for the lines penetrating containment conform to Criterion 57 of the Gereral Design Criteria.

Based on our review we find the design of the steam generator heatup system to be acceptable for the construction permit tage of review.

739040 16-1

13.0 CONDUCT Of OPERATICNS e

13.1 0_rqanizational Structure of Applicant In Supplement No. 24 to the Pre';minary Safety Analysis Report, the applicant amended its organizational structure for technical support and design of the facil-ity by deleting the position of Manager - Thermal Power Generation and reassigning the responsibilities of this position. We have reviewed this change and find that the applicant has an acceptable orginization to implement its responsibilities for the design of the Greene County Nuclear Power Plant. The change does not alter our previous conclusion that the Power Authority of the State of New York is technically qualified to design and construct the Greene County Nuclear Power Plant.

13.6 Industrial Security In its report to the Commission on the Greene County Nuclear Power Plant applica-tion, the Advisory Committee on Reactor Safeguards expressed a concern about the substantial qL l ties of explosives used near the site and recommended that this be given special consideration in the development of security measures for the plant. On December 20, 1977, representatives of the applicant met with the staff to discuss the plant security provisions responsive to the above concern. The applicant is studying a combination of physical barriers and administrative pro-cedures to prevent the approach of explosive loaded vehicles to the plant. Based upon these discussions and plant prntection features studied by the staff on a generic basis, we conclude that adequate protection for the hypothetical threat can be provided.

Final details on this matter will be provided during the operating license staga of review. In the interim we find that our earlier conclusions are not altered. That is, we conclude that a satisfactory planning base has been described by the appli-cant upon which a complete security program can be developed to demonstrate c' pliance with the new regulations and to provide an acceptable level of physical protection to this site at the appropriate time. We will continue to work and provide guidance to the applicant to this end.

739041 13 1

15.0 ACCIDENT ANALYSIS 15.3 Moderate Frequency Events 15.3.1 Increase in Heat Removal by the Secondary System In the Safety Evaluation Report we noted that the applicant identified the simul-taneous inadvertei' opening of the power turbine bypass valves as the bourding event for steam releases from the secondary system. This occurrence will withdraw 15 percent additional steam from the steam generators. The applican stated that the analysis has not been done for this f acility, but that the analy: :s has been performed for the 3800 megawatt ESAR-205 Standard Reference Plant, and that the re.u!ts will be less severe for the Greene County Nuclear Power Plant. The BSAR-205 analysis shows fhat the departure from nucleate boiling ratio remains above 1.32 and the reactor coolant system pressure is reduced by the transient. However, we were unable to accept the applicant's assumption for the BSAR-205 analysis that the largest increate in steam flow for a moderate f requency event will be limited to 15 percent and we required additional information to confirm that inadvertent actuation of the atmospheric dump valves, or decay heat release valves in combina-tion cannot occur.

In Supplement No. 25 to the Preliminary Safety Analysis Repor t, the applicant submitted additional information on the design of the control system for the main steam cump valves and turbine bypass valves. Af ter a review of the preliminary design of the contrul system, we conclude that multiple failures, multiple operator actions, or multiple set point calibration errors will be required to cause more than a 15 percent increase in ste n flow.

On this basis, we find the BSAR-205 analysis of the simultaneous opening of the four condenser dump valves acceptable as the bounding event. As the analysis was done on a similar 3300 MWt plant, we consider it conservative for the Greene Cconty Nuclear Power Plant. This analysis is acceptable because it shows that this transient does not result in either a violation of the departure from nuc'eate boiling limit or reactor coolant system pressure limit for moderate frequency events. We consider this matter outstanding item (15) in Section 1.8, to be resolved.

15.4 Infrequent Transients and Accidents 15.4.4 Feedwater System Breaks Inside and Outside of Containment The applicant provided analyses and discussions of feedwater s em breaks that it considered cc m erv3tive and bounding for these en nt:,

T3* C In u of feedwater was cited as bounding for a feedwater break outside of containment. An analysis M

15-1

uf a feedwater line break inside containment was presented. In neither case, during the thirty-second transient period analyzed, were fuel or reactor co lant system accident limits exceeded.

Although similar analyses have been founo acceptable for earlier nuclear steaa supply systems, designed by the Babcock and Wilcox Company, we requested additional information to complete our understanding of these esents. In particular, we requested studies c.? analyses of both short-term and long-term periods following the postulated pipe failure. The impact of the following parameters and events on the plaot response to a spactrum of feedwater system breaks inside and outside containment were to be investigated:

(1) Loss of offsite power;

(?) Single failure; (3) Operator actions; (4) Seismically initiated events; (5) Ereak size and location; and, (6) Systems required or assumed to cperate.

The BSAR-205 The applicant provided only limitea responses to our requests.

analysis of the feedwater line break was referenced as bounding in terms af pri-mary system pressure. The referenced analysis showed that primary system pressure remained below 110 percent of system oesign pressere ar.J reactor power level remained below 112 percent of rateo thermal power. The analysis showed that this feedwater system break does not result in either a violation of the departure from nucleate boiling limit of reactor coolant system pressure criteria for moderate frequency events.

On the basis that the additional information provided by the applicant still shows that the consequences of feedwater system breaks are acceptable with margins relative to acceptable limits for fuel damage, reactor coolant system pressure limit, and release of radioactive materials from infrequent incidents and accident.,

we find the facility design

'.o be acceptable for the construction permit stage of review. We consider this matter, outstanding item (16) in Section 1,8, to be resolved.

We will require that the applicant provide analyses in the final Safety Analysis Report that address the concerns noted above in order to confirm acceptable con-sequences of the feedwater system break as impacted by the above parameters and events. We will review the applicant's analysis in the Final Safety Analysis Report for confirmation of acceptable performance based on the above requirements.

In the event that the analyses fa;l to confirm acceptable performance, modifica-tions will be considered at that time.

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On the basis that the additional information provided by tL applicant still shows that the consequences of the steam line break are acceptable with margins a isting relative to acceptable limits for fuel damage, reactor coolant system pressure limit, and release of radioactive materials from infrequent incidents and accidents, we find the facility design to be acceptable for the construction permit stage of review.

We consider this matter, outstanding item (17) in Section 1.8, to be resolved.

We require that the applicant provide analyses in the Final Safety Analysis Repoit that address the sta'f cone'_rns nated above in order to confirm acceptable conse-quences of the steam system piping failures as impacted by the above parameters and events. We will review the applicant's analyses in the final Safety Analysis Report for confirmation of acceptable performance based on the above requirements.

In the event the analyses fail to confirm acceptable performance, modifications will be considered at that time.

15.5 Radiological Consequences of Accidents 15.5.1 loss-of-Coolant Accident The radiological ccnsequences of the deposition of radionuclides into the water sc ply reservoirs near the site after a postulated loss-of-coolant accident were reviewed. We performed an independent analysis to determine the maximum doses which might be received by individuals through this pathway. For the Greene County Nuclear Power Plant, three reservoirs have been identified as of concern with regard to deposition of radioactivity following a postulated loss-of-coolant accident. Table 15.1 identifies the reservoirs, their respective distances from the plant, surface areas and volumes. We calculated the expected iodine-131 concentration in the three reservuirs at the end of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the postu-lated accident and these values are also given in Table 15.1.

Because of its distance and surface-to volume ratio, the Saugerties Reservoir was found to be the most limiting.

To estimate the exposure to an individual assumed to ingest the reservoir water, the following assumptions were made:

(1) By the end of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, appropriate steps have been taken to interdict public use of reservoir water.

(2) The maximum exposed individual is an infant whose assumed intake of reservoir water is 0.9 liters during the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(3) for the maximum individual (infant) the thyroid dose conversion factor is 13.9 rem /nillicurie ingested (Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effiuents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I").

739045

Tt3LE 15.1 RESERVOIR CONTAMINATION FROM A POSTULATED LOS5-OF-COOLANT ACCIDENT Concentration Distance and Direction Surface Area Volume Iodine-131 after from Greene County Plant (square (cubic 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (curies Reservoir (kilometers) meters) meters) per cubic meter)

Saugerties 9.7 WSW 20,000 17,400 3.8 x 10-6 Potoc 21 N 235,000 1,000,000 0.54 x 10-6 Ashokan (at its 24 SW 32,000,000 480,000,000 0.077 x 10-6 closest point) h 15-5

(4) The iodine-131 dose is the only dose calculated because of its longer half-life and higher dose conversion factor. The other iodines would contribute a relatively small added dose. Radioisotopes other than the iodines contribute an insignificant added dose.

(5) No credit is given for radionuclide removal during transport through the water distribution system, cod no credit is given for radionuclide removal by water purification processes.

Based upon the above assumptions, the estimated dose to the infant from ingestion of water from the Saugerties Reservoir is calculated to be 0.05 rem thyroid. The dose conversion factor for an adult is approximately 14 percent of that for the infant, and the resulting adult thyroid dose would be 0.007 rem for the same intake.

The applicant also identified a spring-fed quarry 1.5 miles to the southwest of the site as a source of water for a limited number of users. We performed an analysis in which we calculated the concentration of iodine-131 in the quarry

-6 water at 4.3 x 10 curies per cubic meter. We conclude, based upon the above assumptions, that an individual drinking water drawn from this source following a postulated design basis loss-of-coolant accident might receive a dose comparable to that given for the Saugerties Reservoir.

Wet deposition (washout of materals f rom the atmosphere due to rainf all) results in greater amounts of material being removed from the atmosphere. If it is assumed to be raining at the time of the postulated accident, the amount of iodine-131 deposited in the reservoir would be significantly less because of rapid removal of material by the sc3 verging action of rainfall in the intervening distance between the plant and the reservoir. The quantity deposited in the reservoir would be greater only if it is assumed to be raining in the area of the reservoir but not in the intervening distance. We used that assumntion and also assumed that it rained continuously for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> over the Saugerties Reservoir.

Under these conservative assumptions. the estimated thyroid dose to the infant from ingestion of the rescrvoir water is calculated to be 2.8 rem.

Based upon the conservatisms used in the above analyses and the relatively low dases calculated, we conclude that the postulated loss-of-coolant accident at the Greene County Nuclear Power Plant is not likely to result in significant exposures of the public because of deposition in the Saugerties, Protoc, or Ashokan Reservoirs.

733047 15-6

15.5.4 Main Steamline Break Outside Containment The assumptions used to analyze the consequences of the steamline break accident are given in Table 15.2, and are consistent with those given in Regulatory Guide 1.5, " Main Steamline Break Accident" A guillotine break of a main steam line outside the containment is assumed to occur, and the entire content of the steam generator connected to the broken line is assumed to be emptied to the atmosphere, together with all the radioactivity it contained. Reactor trip cccurs at the time of the accident but the primary system pressure remains at or near normal operating pressure. The pressure differential across the steam generator tubes maintain the leakage of one gallon per minute, and no additional tube failure is assumed to occur. All radioactivity contained in the primary coolant leaking to the shell side is available for release to the atmosphere.

We calculated the thyroid doses to be 68 rem (0-2 hours) at the exclusion radius, and 10 rem (0-8 hours) at the low population zone boundary. The results show that the potential radiological consequences are within the guidelines of 10 CFR Part 100 and therefore are acceptable.

TABLE i5.2 STEAMLINE BREAK ACCIDENT ASSLNNS AND TkFJT TIMETERS Power level 3600 megawatts thermal Failed fuel fraction 0.7 percent (due to accident)

Quantity of iodine released 10 percent from failed fuel elements Iodine activity level in secondary coo 1 ant 0.1 micro curies per gram Steam generator primary-to-secondary leak rate 1 gallon per minute Atmospheric disperson factors 4

0-2 hours at 472 meters 6.8 x 10 seconds per cubic meter 0-8 hours at 4023 meters 1.02 x 10 seconds per cubic meter 15-7

18.0 PEPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS At its 210th meeting o. October 6-8, 1977, the Advisory Committee on Reactor Safe-guards completed its review of the application by the Power Authority of the State of New York for a permit to construct the Greene County Nuclear Power Plant. A copy of the committees report on the Greene County Nuclear Power Plant, dateJ October 12, 1977, contains certain comments and recommendations, is included as Appendix 0 to this report. The actions we have taken or plan to take in response to the Committee's comments and recommendations are described in the following paragraphs.

(1) The Committee noted that our Safety Evaluation Report identified a number of outstanding safety items, and the Committee recommendeu that these matters should be resolved in a manner satisf actory to the staf f.

Since the October 6, 1977 meeting with the Committee, we have revitwed additional information and commitments submitted by the applicant concerning the outstanding issues identified by the staff. Our evaluations, resolutions, and status of unresolved outstanding items are reported in appropriate sections of this supplement to our Safety Evaluation Report.

(2) The Committee identified those generic issues relating to large water reactors which were discussed in the Committee's report to the Commission on February 24, 1977 (Report No. 5), and which the Committee considered relevant to the Greene County Nuclear Power Plant. The Committee noted that the issues should be dealt with by the staff and applicant as solutions are found.

We have transmitted the Committee's recommendations to the applicant for its consideration in proceeding with the design of the Greene County Nuclear Power Plapt. Appendix C to the Safety Evaluation Report discusses the disposition and status of the generic matters raised by the Advisory Committee on Reactor Safeguards.

(3) The Advisory Committee on Reactor Safeguatds expressed a concern about the substantial quantities of explosives used near the site, and recommendec this should be given special consideration in the develooment of security measures.

Our review to this matter is oiscussed in Section 13.6 to this supplement.

W90@

18-1

21.0 CONCLUSION

S In Section 21.0 of the Safety Evaluation Report, we stated that we would be able to make certain conclusions upon favorable resolution of the outstanding items set forth in Section 1.8 of the Safety Evaluation Report. We have described the favor-able resolution of many of these iteis in this supplement, and we have identified new outstanding items which are listed in Section 1.8 of this supplement.

We will require acceptable resolutions for all of the remaining outstanding items listed in Section 1.8 of this supplement prior to reaching a favorable conclusion for the Greene County Nuclear Power Plant.

4 730050 21-1

APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW Of POWER AUTHORITY OF THE STATE OF NEW YORK'S APPLICATION COUNTY NUCITER POWER PLANT DOCKET NO. 50-549 September 1, 19/7 Safety Evaluation Report issued.

September 12, 1977 Summary of meeting with PASNY held July 19, 1977.

September 21, 1977 Advisory Committee on Reactor Safeguards subcommittee meeting with staff and applicant.

September 29, 1977 Meeting with PA5NY to review outstanding technical items identified in Safety Evaluation Report.

September 30, 1977 Letter from PASNY transmitting information to resolve open items in Safety Evaluation Report.

October 6, 1977 Advisory Committee on Reactor Safeguards meeting with staff and applicant.

October 6, 1977 Summary of meeting with PASNY held September 29, 1977.

October 7, 1977 Letter from PASNY transmitting Amendment No. 23, consisting of revision to general and financial information.

October 12, 1977 Letter from Advisory Committee on Reactor Safeguards.

October 13, 1977 Meeting with PASNY to discuss staff requirements for information concerning the analyses for long-term system performance following a postulated feedwater system or steam system break.

October 13, 1977 Summary of meeting with ACRS subcommittee held 5eptember 21, 1977.

October 25, 1977 Letter to PASNY regarding physical security assessment models.

October 31, 1977 Summary of meeting with ACRS held October 6, 1977.

November 2, 1977 Letter from PASNY transmitting Amendment No. 24, consisting of revised and additional information, k)f3k) 5.),

A-1

November 8, 19//

Letter to PANSY identifying two additional items since the Safety Evaluation Report was issued.

November 16, 1977 Summary of meeting with PASNY held October 13, 1977.

November 22, 1977 Letter to PASNY concerning long-term plant performance for fe.dwater or steam system pipe breaks.

November 23, 1977 Letter to PA5NY regarding withholding of proprietary information.

December 1, 1977 Letter from PASNY transmitting Amendment No. 25, consisting of revised and additional information.

December 6, 1977 Letter from PASNY transmittirQ copies of blast monitoring records.

December 20; 1977 Meeting with PASNY to discuss plant security provisions for external explosions.

December 27, 19/7 Summary of meeting with PASNY held December 20, 1977.

December 30, 1977 Letter from PASNY transmitting Amendment No. 26, consisting of revised and additional information.

January 4, 1978 Meeting with PAS.d to discuss proposed change to use a fixed base seismic analysis.

January 6, 1978 Letter from PAN 5Y regarding acquisition of land within exclusion area.

January 9, 1978 Letter from PASNY ansmitting comments on the Safety Evaluation Repcrt.

January 19, 1978 Summary of meeting with PA5NY held January 4,1978.

Janua y 30, 1978 Letter from PASNY addressing outstanding safety review open items.

January 31, 1978 Letter to PASNY concerning comments by Corps of Engineers regarding foundation conditions.

February 2, 1978 Letter to PASNY transmitting request for additional information on fire protection program.

February 2, 1978 Letter from PASNY transmitting Amendment No. 27, CCnsisting of revised and additional information.

A-2 bO[32

February 2,1978 Letter to PASNY identifying additional outstanding item since Safety Evaluation Report was issued.

March 3, 1978 Letter f rom PASNY transmitting reference report entitled, " Evaluation of Ground Motions Induced by Postulated Explosions in the vicinity of Greene County Nuclear Power Plant, Revision 1."

Marts 7, 1978 Letter from PASNY transmitting information on steam line break.

March 9, i978 Letter from PASNY concerning backfill density testing recommended by NRC and Corps of Engineers.

March 16, 1978 Letter to PASNY transmitting staff positions on design basis pipe break criteria.

Apri! 5, 1978 Megting with PASNY to discuss outstanding items.

Apri' 7, 1978 Letter to PASNY requesting additional information concerning contain-ment system design.

April 19, 1978 Letter to PASNY requesting additional information concerning auxiliary system design.

April 19, 1978 Letter from PASNY transmitting information concerning feedwater line break.

April 19, 1978 Letter to PASNY transmitting safeguard handbooks (Oifici31 Use Only).

May 5, 1978 Letter to PASNY transmitting for comment, Draft 2 to NUREG-0219,

" Nuclear Security Personnel for Power Plants, Review Plan and Accept-ance Criteria for a Security Training Program" (Official Use Only).

May 11, 1978 Letter from PASNY transmitting respanses to request for additional information on fire protection program.

May 26, 1978 Letter from PASNY transmitting Amendment No. 28, consisting of revised aid additional information.

May 29, 1978 Summary of meeting with PASNY held April 5, 1978.

May 29, 1978 Letter to PASNY concerning staff position and request for additional information on reactor systems design and reactor analysis.

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June 8, 1978 Letter from PASNY authorizing Mr. P. J. Early to sign correspondence on behalf of applicant.

June 12, 1978 Letter to PASNY transmitting two security reports:

"The Barrier Technology Handbook," and "A Systematic Approach to the Conceptual Design of Physical Protection Systems for Nuclear Facilities."

June 27, 1978 Letter from PASNY transmitting Amendment No. 29, consisting of revised and additional information.

July 14, 1978 Letter to PASNY transmitting Revision 2 of the " Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors,"

NUREG-0103, Rev. 2.

July 18, 1978 Letter to PASNY transmitting " Barrier Penetration Database,"

NUREG/CR-0181.

July 25, 1978 Letter from PASNY providing additional blast monitoring program data.

luly 25, 1978 Letter from PASNY transmitting additional details concerning basis for evacuation plans.

August 1, 1978 Letter f ron, PASNf transmitting Amendment No. 30, consisting of revised and additional information.

August 2, 1978 Letter to PASNY transmitting " Nuclear Security Personnel for Power Plants, Content and Review Procedures for a Security Training and Qualification Program," NUREG-0219.

August 3, 1978 Letter to PASNY concerning staff position on loss of component cooling water to reactor coolant purp seals.

August 11, 1978 Letter to PASNY regarding the standard format for meteorc' gical data on magnetic tape.

August 15, 1978 Letter to PASNY announcing a conference, " Pressurized Water Reactor Steam Generator Workshop," to be held on September 7 and 8, 1978.

August 16, 1978 Letter from PASNY documenting transmittal of meteorological data tapes and requesting their return.

t.

s A-4

APPlhDIX B 4

BIBLIOGRAPHY (Continuedj REACTOR 27.

Lowe, R.

J., et al., "ECCS Evaluation of Babcock and Wilcox's 205 CA NSS,"

BAW-10102, Revision 2, Babcock and Wilcox Company, December 1975.

28.

Taylor, J. H. (Babcock and Wilcox), Letter to S. A Varga (NRC), "Large Break Evaluation of Babcock and Wi.cox 205 FA NSS Usi.ig the August 1977 ECCS Evaluation Model (BAW-10104, Revision 3)," Supplement to BAW-10102, Supplement to BAW-19102, September 30, 1977.

29.

Dunn, B.

M., et al., "Babcuck and Wilcox's ECCS Evaluation Model," BAW-10104A, Revision 1, Babcock and Wilcox Company, March 1976.

30.

Dunn, B. M., et al., " Babcock and Wilcox's ECCS Evaluation Model Revision 3,"

BAW-10104, Revision 3, Babcock and Wilcox Company, August 1977.

31.

Varga. 5. A.

(NRC), letter to K. Suhrke (Babcock and Wilcox), " Babcock and Wilcox ECCS Evaluation Model," Februa y 18, 1977.

32.

Ross, D. F. (NRC), Memorandum to D. Vassallo (NRC), " Babcock ard Wilcox Reflood Model Changes," February 8, 1978.

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B-1

APPENDIX C CHANGES AND ERRATA TO THE SAFETY EVALUATION REPMTT5TdED SFPTEMBER 1977 Chanq)s and errata effected by this appendix to the report do not alter the staff conclusions presented in the earlier Safety Evaluation Report or the significant information upon which those calculations are based.

Table of Contents Page ti, line 27 Change "Agains" to "Against" Page is, line 18 Change " Malfunction" to " Malfunctions" Page ix, line 25 Change " Seisure" to " Seizure" Chapter i Page 1-5, line 20 Change " facilities" to " facility" Page 1-5, line 22 Cha7ge "f acilities" to "f acility" Page 1-6, line 6 Change " Appendix B" to " A. endix A" t

Page 1-8, line 3 Change "tehcnical" to " technical" Page 1-9, line 9 Change "dasymmetrit" to " asymmetric" Page 1-10, line 10 Change "Accpetance" to " Acceptance" Page 1-12, line 6 Change "matal-water to " metal-water" W390a,,o C-1

Chapter (

Page 2-4, line 24 Change "authvity" to " authority" Page 2-7, line 1 Change "Marquett" to "Marquette" Page 2-7, line_24 Change "l700" to "1600" Page 2-9, lines 3 and 4 Delete the first sentence in this paragraph ard substitute the following: "There are no operating air facilities within five miles of the site."

Page 2-9, lines 16 through 21 Delete the east two sentences in this paragraph and substitute the following:

" Route J68 is the closest jet route serving aircraf t at altitudes above 18,000 feet.

In-flight accidents in jet routes leading to crashes from high altitudes are extremely rare, and were such to occur, the possible locations of impact would be diffusely distributed over very wide areas. Plant sites reviewed in the past which had equivalent aircraft traf fic in equal or closer proximity were, af ter careful examination, found to present no undue risk to the safe operation of those plants.

Based upon this experience, in the staff's judgment, no undue risk is present from aircraft hazard at the plant site now under consideration."

Page 2-13, line 10 De'ete "488 meters to the south" and ada as a third sentence to this paragr oh:

"This relative concentration was calculated for the south sector of the exclusion area using a site boundary distance of 488 meters in accordance with the model."

Page 2-15, line 4 Change "25 year hurricant surge at the Battery;" to "40 miles per hour overland wind; Page 2-19, line 5 Delete "following" Page 2-19, line 6 Following the word " levels" insert "given in Table 2.4" Page 2-22, line 36 ange " structure found" to " structural trend found"

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?380!;7

Page 2-24, line 33 Change " degreed" to " degree" Page 2-25, line 14 Change " observe" to " observed" Page 2-27, line 17 Change "than" to "then" Page 2-29, line 35 Change " insite" to "in situ" Page 2-29, lines 37 and 38 Change " building, containment structure, and the two service water cooling towe:i will" to " building and containment structure wil'"

Page 2-30, line 7 Following "mean sea level" insert "the two service water cooling towers to be founded at mean sea level,"

Page 2-30, line 38 Change "0.07" to "0.10" Chapter 3 Page 3-5, lines 18 and 19 Change the first sentence of this paragraph to read:

" Exterior walls and slabs of seismic Category I structures will be waterproofed below plant grade to protect systems and components located below the ground level.

The solid waste and decontamination building will similarly be waterproofed below plant grade.

Page 3-8, lines 1 and 2 Change "Section 3.5.1.4 (Revision 1) was revised with" to "a revision (Revision 1) to Section 3.5.1.4 was prepared to include" Page 3-14, line 6 Change "the American" to "the applicable sections of the American" Chapter 4 Page 4-3, line 30 Change "heen" to "been"

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Page 4-3, line 37 Change "Babacock" to " Babcock" Page 4-6, lines 1 and 2 Change the comma to a period and delete the phrase "and a destructive examination will be conducted after three cycles of operation."

Page 4-9, lines 1 through 4 Change this sentence to read "We have requested a tcpical report on the results and interpretation of test data."

Page 4-9, line 4 Change "will" to "should" Page 4-9, line 7 Change "fatique" to "f atigue" Page 4-22, line 28 Following " Maximum fuel centerline temperature" insert "at the beginning of core life" Page 4-23, line 24 Change " low" to " bow" Chapter 5 Page 5-5 line 22 Change "3800" to "3780" Page 5-5, ines 31 and 32 Change "during long-term cooling" to read "at all times the reactor coolant system pressure 'oundary is filled and closed."

Chapter 6 Page 6-2, lines 32 and 33 Change "has calculated" to " submitted the results of a calculation for a typical Babcock and Wilcox 205 FA class plant showing" Page 6-9, line 10 Change "contingvous plant areas." to " areas contiguous to the containment structure."

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Page 6-12, lines 18 and 19 Delete "the decay heat removal system return line,"

Page 6-14, line 14 Change "one percent of the fuel cladding," to "the amount of hydrogen that would evolve from a core wide average depth of reaction into the original cladding of 0.00023 inch,"

Page 6-18, line 4 Delete "and low pressure" Chapter 7 Page 7-2, line 28 Change "imcomplete" to " incomplete" Chapter 8 Page 8-3, line 7 Change "alternatig" to " alternating" Page 8-3, line 18 change "contains" to "contain" Chapter 9 Page 9-10, line 3 Change " electrical tunnel ventilation" to " electrical tunnel ventilation, kitchen, and toilet exhaust system and purge air supply and exhaust" Page 9-10, line 24 Change " intake" to "doct valves" Page 9-12, line 11 Change "acutate" to " actuate" 7330GO Chaptery Page 11-7, Table 11.2 "D

(a) Add a footnote:

Quality Group and seismic design in accordance with BTP ETSB 11-1 (Revision 1).

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(b) Delete the reference to footnote "a" from the subheading Process Gas Subsystem (c) Add a reference to footnote "a" to the component; Degasifier (d) Add a reference to footnote "b" to the following components:

Charcoal Sed Absorbers Process Gas Compressors Process Gas Receiver Tank Page 11-8, line 36 Change " seismic Category 1" to " designed to the seismic criteria contained in Branch technical Position ET58 No. 11-1 (Revision 1)."

Chapter 13 Page 13-2, line 11 Change "Qaulity" to " Quality" Page 13-3, lines 22 and 23 Delete ", either the Radiation Protection and Radiochemistry Supervisor, or" Chapter 15

.P_ age _,15-17, line 29 Change "0.5" to "0.05" Page 15-17, line 30 Change "0.5" to "0.05" Page 15-17, line 32 Change "0-hours" to "0-2 hours" Page 15-17, line 32 Change "4.8 x 10 " to "4.8 x 10'4" 4

Page 15-17, line 33 5

- 5,,

Change "2.5 x 10 " to "2.5 x 10 Page 15-19, line 21 Change "2 hrs" to "0-2 hours"

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Appendix A Insert in respective calendrical positions:

November 5, 1975 Summary of meeting with PASNY held October 24, 1975.

March 15, 1976 Summary of meeting with PASNY held February 10, 1976.

March 25, 1970 Summary of site visit to review meteorology program on March 18, 1976.

June 23, 1976 Summary of meeting with PASNY held May 6, 1976.

July 3, 1976 Summary of meeting with PASNY held May 14, 1976.

August 27, 1976 Summary of meeting with PANSY held August 5, 1976.

December 21, 1976 Summary of meeting with PASNY held November 1, 1976.

December 21, 1976 Summary of meeting with PA5NY held November 9, 1976.

March 8, 1977 Summary of meeting with PASNY held February 3, 1977.

April 20, 1977 Summary of meeting with PASNY held April 6, 1977.

Appendix D Page 0-1, Line 24 Change "Satefy" to " Safety" Page 0-1, line 30 Change " TABLE C-1." to " TABLE D-1."

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APPENDIX D p

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UNITED STATES e,

NUCLEAR REGULATORY COMMISSION g

l ADVISORY COMMITTEE ON REACTOR SAFEGUAROs j

msamoTom o. c. 2assa October 12, 1977 Honorable Josepn M. Hendrie Chair:ran U. S. Nuclear Regulatory Comission Washington, D. C.

20555 SUBJEC: REPCRT CN GREENE CXNIY NUC:. EAR PCEER PUN 2

Dear Dr. BeMrie:

During its 210th Meeting, October 6-8, 1977, the Adviscry Corm:ittee ca Reactor Safeguards ccepleted its review of the application of the Pwer Authority of the State of New Ycrk (Applicant) for a permit to construct the Greene County Nuchar Power Plant. A Shm.ittee meeting was held in Catskill, New York on Septemoer 21, 1977 and the plant site was visited by members cf the Sebecrmittee the same day. The Cocrnittee had the bene-fit of discussions with representatives and censultants of the Applicant, Bacccck and Wilecx Cotpany, Stone and Webster Engineering Ccrporation, aM the Nuclear Regulatcry Cccraission CGC) Staff. The Carnittee also had the tenefit of the docu::ents listed.

The Greene County Plant will utilize a 2600 E(t) Bahcock & Wilecx pres-surized water reacter, enclosed in a steel-lined reinforced cencrete con-tainment.

• hr. basic design of the Nuclear Stea:n System (NSS) is similar to designs for the Washington Public Pwer Supply Systen Nuclear Projects, RG 1 aM 4, the Bellefcnte Nuclear Power Plant, Units 1 aM 2 aM tae Peccle Springs Nuclear Plant, Units 1 and 2 repcrted on in Cxmittee let-ters of Jure 11,1975, July 16,1974 and February 11, 1976, resrectively.

The NSS design is also similar to the 3800 MW(t) Babcock-205 Standard NSS design on which the Cemittee repcrted in its letter of August la, 1977. The balance-of plant design is similar to the Stone and Wecster staMard balance-of plant design for Westingnouse reactors en which the Cor:mittee previously repcrted in its letter of August 18, 1976.

The preposed Greene County Plant will be located en a 190 acrc site on the west bank cf the Hudson River apprcximately 35 :niles south of Alhany, New York and 13 miles north-ncrtheast of Kingston, New Ycrk (the nearest pcpulation center,1970 population 25,500). The :nini.m:n exclusion dis-tance is 1500 feet frcen the center of mntainment and the radius of the low ;cpulatien zene is 2 1/2 miles.

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Honorable Joseph M. Herdrie The Applicant and the Staff have agreed on a horizontal ground accelera-tion cf 0.2g for the safe shutdown earthgaake and 0.lg for the operating basis earu g m e.

The Comittee considers these values acceptable for tais plant.

The Staff has identified a nu::ter of safety itec:s which will require reso-lution before issuar.ce of a construction permit. These matters should be resolved in a :rarner satisfactory to the Staff. The Ccrm:iteee believes these ite::s can be resolved prior to the issuance of a constraction per:::it.

The Ccruittee has concerns about the substantial quantities of explosives uM near the site, and believes this should be given special consideraticn in the developnent of security :neasures.

With regard to generic problems cited in the Co:mtittee's report " Status of Generic Itsus Relating to Lignt water Reactors: Report No. 5,* dated February 24, 1977, itec:s censidered relevant to the Greene County Plant II-2, 3, 4, 5 (loose parts ::enitor resolved), 6, 7,10; IIA-3, 4, are:

5, 6, 7; IIB-1, 2; IIC-1, 2, 3, 4, 5, 6; IID-2. 'itese problen:s should be dealt with by the Staff and the Applicant as solutions are found.

The Advisory Cnmittee on Reactor Safeguards telieves that if due con-sideration is given to the foregoing, the Greene County Nuclear Power Plant can te constructed with reasonable assurance that it can be oper-7.ted without urdue risx to the health and safety of the public.

Sincerely, M. Sender Chair:ran References 1.

Greene County Nuclear Power Pl. int Preli=inary Safety Analysis Re-port, Volu:nes 1 through 12 and Supplements 1 through 19.

2.

Safety Evaluation Report related to construction of Greene County Nuclear Power Plant, NUFm-0283, Septeter 1977.

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UNITED STAttS

._py NUCLE AR REGULATOMY COMMtSSION posraos Ano rats paso WASHINGTON, D. C. 20S55

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