ML19207B578

SER Re Operation of Facilities.Suppl 2

ML19207B578
Person / Time
Site:	Crane, Farley
Issue date:	10/31/1976
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0117, NUREG-0117-S02, NUREG-0117-S2, NUREG-117, NUREG-117-S2, NUDOCS 7909040054
Download: ML19207B578 (32)

Text

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Y EUPPL. NO. 2 TO NUREG-75/034 3DBDED GDHD elD0DriT Regulatory Comm s on related to operation of Office of Nuclear Joseph M. Farley Riuclear Plant Units 1 and 2 Docket No. 50-348 "U~ 84 Alabama Power Company OCTOBER 1976 Supplement No. 2

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NUPEG-Oll7 Sutplement 2 to NUREG-75/034 SL'FFLEMENT NO. 2 TC THE SAFETY EVALUATION PEPCRT BY TFE C;FICE OF NUCLE AR PE ACTOR PEGULATIO'4

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NUCLEAR PEGULATORY COMMISSION IN THE MATTER OF ALACAMA F0 HEP CCMP A.'d JOSEPH M, F A' LEY 'UCLE AR PLANT U'4ITS 1 T'iD 2 00CVEi N05. 50-348 AND 50-364 Available f rom.

National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy 4.CQ Microfiche

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TABLE OF CONTENT *-

PAGE 1.0 INTPODUCTION At.D GENERAL DFSCRIPTICN CF FLANT.

I 1.1 Introduction.

I 1.

Surrary of Outstanding Peview Itars and Issues.

1 4.0 PEACTOR.

3 4.2 Fuel Pechanical Design.

3 6.0 ENGINEERED cETY FEATURES.

4 E.3 Emerger.cy Core Cooling Systen.

4 6.3.2 Systen Design.

4 6.3.3 Tests and Insoections.

7 6.4 Er.ergency Shutdown Cooling Capability.

9 7.0 INSTRUMENTATION AND CONTROLS.

10 7.3 Engineered Safety Features Actuatien and Control.

10 7.8 Cable Separation and Identification Criteria.

11 7.8.1 Bypassed and Inoperable Status Indication.

11 7.8.2 Erergency Power Board.

12 7.8.3 Color Code Identification Criteria.

12 9.0 AUXILIARY SYSTEMS.

14 9.2 Fual Storage and Handling.

14 9.2.4 Fuel Pandling Systen.

14 9.3 Cooling Wa ter Systens.

15 9.3.1 Auxiliary Feedwater System.

lb 10.0 STEAP AND POWER CONVERSIC'. SYSTEM.

18 10.5 Pain Feedwa ter Syster..

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TABLEOFCONTENTS(Continuedl PAGE 15.0 ACCIDENT ANALYSIS.

20 15.2 Thennal and Hydraulic Analyses.

20 15.2.2 Accidents.

20 16.0 TECHNICAL SPECIFICATIONS.

23 APPENDICFS APPENDIX A SUPFLEMENT TO THE CHPONOLOGY OF RADIOLOGICAL REVIEW.

A-1 APPENDIX B SUPPLEMEkT TO THE BIBLIOGRAPHY.

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1.0 INTRODUCTION

AND GENERAL CESCPIPTION OF THE PLfNT 1.1

'atroduction The Safety Evaluation Report for the Joseph M. Farley Nuclear Plant, Units 1 and 2 (Farley plant), cated May 2,1975, concluded that upon favurable completion and resolu-tion of the outstanding matters described therein the plant could t)e operated without endangering the health and safety of the public.

Supplement No.1 to the Safety Evaluation Peport described our evaluation of additional technical and financial infon"ation provided in Amendments 48, 49, 50, and 51 to the application, and described our plans for the implementation of the recomendations made by the Advisory Comittee on Reactor Safeguards.

This supplemert. Supplement No. 2 to the Safety Evaluation Report, describes our evaluation of additional technical information provided in Amendments 52 through 59 to complete and resolve outstanding matters. The sections ' this supplement are nu t bered and titled so as to correspond to the sections of the Safety Evaluation Peport that have been affected by our additional evaluations. Appendix A is a continuation of the chronology of principal everts that have occurred during the safet:/ review. Appendix B lists additional documents used in the supplemental review.

1.7 Sumary of Outstanding Review Items and Issues In Section 1.7 of Supplement No.1, we identified eleven items tha. were outstanding because additional information was required from the applicant or b?cause the staff had not completed its review of recently suNi*ted amendments. Sinc ? preparation of Supplement No. I was completed, eight of the eleven outstanding items have been completed, and progress has been made towards completion of the other thrte items. An additional matter regarding tests of the emergency core cooling s, stem has been raised by the staff for resolution in the Farley plant application.

Our evaluation of the eight resolved items and the resolution of the additional issue is provided in the text of this report. The current status of the three outstanding review items and the references that provide our evaluaticn of each item are tabulated below. Two of the three remaining outstanding items a' e expected to be completed satisfactorily prior to loading fuel into Uni? 1. now scheduled for February 15, 1977. The scheduled date for Unit 2 fuel loading is December 15, 1978. The satis-factcry completion of the other item, anticipated transients without scram, is expected to be completed prior to the first fuel reload of Unit I and the initial fuel loading of Unit 2.

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OUTSTANDING ISSUES Safety Evaluation Peport Feference item Status Section 3.9.1 f.nalysis of LOCA Staf f's June 24, 1976 letter lcads cn reactor requested additional infonna tion for vessel support our review of the reactor vessel structures support system. Applicant's report ef review is expected in the f al' of 1976.

Section 5.4 Anticipated Staff's July 2, 1976 let:(i requested transients without that recorrendations in our " Status scram Report on Anticipated Transients Without Scram fce Westinghouse Reactere be considered and tha t design changes resulting therefrom be provided by March 30, 1977.

Section 7.7 Seismic and West 19ghous( has p.ovided acceptable environmental scope for test ar.d analysis programs, quali fica tion scheduled for completion in 1976.

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4.0 PE ACTOR 4.2 Fuel Mechanical Design Irradiation - indated densification of uranium dioxide fuel pellets causes decreases in pellet radius ar.d length which can result in corresponding axial gaps between pellets. Cladding time-to-collapse calculations predict the tir.

' r unsupported cladding to beccme dimensicnally unstable and to flatten into the <ial gap between pellets.

Our Safety Evaluation Report concluded that the methods used by Westinghouse to calculate the cladding time-to-Collapse were acceptable and stated that the predicted me-to-collapse fcr the Farley plant initial core loading would be required prior to i

a c cision on issuance of an operating license. Final Safety Analysis Peport (FSAR)

A e'dn., nt 52 provides the minimum predicted clad flattening time for the Farley fuel as 41,00- effective full pcwer hours, which is greater than the expected residence time of fu 1 assemblies in the core.

We have m ae an independent calculation of the fuel cladding tine-to-collapse for the Farley f uel assembly and operating conditions. Results confirm the applicant's con-clusion that the predicted cladding time-to-collapse is greater than the expected residence tire of fuel assenblies in tne core.

Basea on our resiew, we nave concluded that clad flattening will not occur during the service life of tne fuel.

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6.0 ENGINEERED SAFETY FEATURES 6.3 Emergency Core Cooling System (ECCS)

One of the outstanding issues in Supplement No. I to our Safety Evaluation Report was an analysis of the postulated loss-of-coolant accident to demostrate confermance of the emergency core cooling system with Appendix K to 10 CFR Part 50.

In a letter dated June 30, 1975, we provided guidance for the preparation of this analysis and re-quested certain infornation regarding system design. The applicant his provided this analysis and design information in a letter dated September 15, 1975 and in FSAR ATendment hos 39, 46, 4 7, 51 a nd 53.

Our evaluation of the performance analysis is presented in Section 15.2 of this report. Our evaluation of the design information is provided in Section 6.3.2 below.

6.3.2

System Design

6.3.2.1 Failure Mtdes and Effects Analyses Appendix K to 10 CFR Part 50 of the Comnission's regulations requires that the com-bination of erergency core cooling systen subsystems to be assumed operative sh311 be those available af ter the rest limiting single failure of ECCS equipment has occurred.

The worst single failure which woulu minimize the emergency core coolant available to cool the core and provide m.aximum containment cooling was identified by Westinghouse as the loss of a low pressure energency core cooling system pump. The staff concluded in its October 15, 1974 Status Report

• that the application of the single failure criterion was tc confirred during subsequent plant reviews.

In Section 6.3.2 of the Safety Evaltation Report, the staff concluded that the single failure of certain manually-controlled motor-operated energency core cooling system valves could negate or unacceptably degrade a safety function by inadvertent valve novement. The staff stated it would incorporate appropriate administrative cantrols into the plant technical specifications to assure that power would te locked out of the motors of these valves during power operation to preclude inadvertent nation. In its June 30, 1975 letter, the staff requested the applicant to ide'tify these valves.

On pages RSB 3-2 and 3-3, of FSAR Amendnent Nos. 53 and 54, the applicant has identified the following valves and plant conditions for which power will be locked out at the rotor control centers.

• See Bibliography for Accident Analyses in Appendix B to the Safety Evaluation Report.

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(1) Valve n ebers 8884, 8886 and 8889 (discharge from charging and residual heat removal pumps into the reactor coclant hot legs) shcwn on FSAR Figure f.os. 6.3-1 and 6.3-2 will be closed and have power locked out during normal operation. The power will be restored prior to the switchover from cold leg recirculation.

(2) Valve numbers 8706A and 8706B (crossover from the residual heat removal pumps discharge to the charging pump suction) shown on FSAR Figure 5.5-6 will be closed and have power to the valve operators during normal operation The pw er will be locked out dJring normal Coolduwn.

(3) Valve numbers 8808A, 8808B and 880SC (discharge from the accunulators into the reactor system cold legs) shcwn on FSAR Figure 6.3-2 will be open and have power locked out during noro.al operation. The power will be restored during normal cooldown.

(4) Val,e numbers 8132A and 8132B (charging pump discharge header isolation valves) shown on FSAR Figure 9.3-4 will be open and have power locked out if charging pump 1A is declared inoperable.

We have reviewed the applicant's selection of prcposed valves requiring control and plant conditions for which power will be locked out at the notor control centers and find them to be acceptable on the bases that (l) the power will be available to the valves at times when they are required for norql and short term emergency functions, and (P) there also will be access to the locked motor control centers and ample time to restore power (greater than one Fcur) when they are requirec for long tern emergency functions.

6.3.2.2 Boric Acid Concantration During Long Term coolinq In Amendment hos. 50 and 51 to the FSAR the applici.nt has described on page 6.3-10a the procedures he proposes to implement during post-loss-of-coolant accident long term cooling in order to prevent excessive concentration of boric acid in the reactor vessel. The procedures are based on an April 1,1975 study performed by Westinghouse.*

According to these procedures, for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter a postulated loss-of-coolant accident, boric acid solution will be injected into tne cold legs of the reactor coolant system. After that time, the energency core cooling system will be realigned and the solution will be injected into the hot legs of the reactor coolant system.

The staff has reviewed the applicant's proposed procedures and requires that the time for realignment of the emergency core cooling system injection to the reactor coolant syste-hot legs be changed fram 24 to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> af ter a postulated loss-of-coolant accident. This shorter time interval will assure that for cold leg breaks, the concentration of boric acid will not exceed 23.5 weight percent which is ft,r weight

• See Biblicjraphy in Appendix B to this report.

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percent below the solubility limits at 212 F.

The staf f tas concludad that the 4 weight percent safety mar 91n is required to account for uncertainties in prediction of the concentration of boric acid in the reactor vessel. TSe staff also required that the procedures be changed so th1t both the hat lens and the cold legs rather than only hot legs will be used af ter the initial 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> cold leg injection period. Either of the following procedures is acceptable:

(1) sinultaneous hot and cold leg injection, or (2) alternate hot and cold leg injection every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

In Ar endrent 59 to the FSAR, the applicant has agr eed to modify its emergency operatino procedures to meet the staff's requirments. We conclude that the modified emergency operating procedures for the postulated loss-of-coolant accident will have an acccpt-able time interval and acceptable redlignrent of energenCy Core Cooling systen to prevent excessive concentration of boric acid in the reactor vessel.

6.3.2.3 Sutrerged Valves In its letter of June 30, 1975, to the applicant, the staff requested a review of the arrangenent of piping systms within the containrent to deternine whether any valve motors may becore sutrerged following a postulated loss-of-coolant accident when coolant is collected cn the containnent floor for recirculation.

In its letter of Septenter 15, 1975, the applicant identified the following valves that could be sutrerged following a postulated LOCA; accumulater divharge valves (0803A, SE0BB ar.d 8805C) and residual heat removal ty. im suction isolation vcives (8701A, 8701B, 8702A and 8702B). The accumulator discharge valves will be open and have power locked out of the motor operator during nurral power operation. The residual heat removal suction valves will be closed during power oper ation. ?.either type of valve is required to operate folicwing a postulated loss-of-coolant accident.

The applicant concludes that sutrergence of valve motors is not expected to result in adverse consequences.

We have reviewed the arrangerent and functions of these valves. When the valve motors are flooded, there is no power at the valve motors on the accumulator valves and faults in the control circuits to the residual heat reroval valve motors caused by flooding inside containrent would not cause residual heat rmoval valves to open. We agree with the applicant's conclusions that flooding of valve mators near the contain-rent floor will not cause adverse consequences following a postulated loss-of-coolant accident.

6.3.2.4 Automatic Backup to K4nual Switchover In our Safety Evaluation Report we required an autonatic backup to the m.m.

over of the emergency core coolirg system from the injection mode of operation to the o-

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recircula tion tode of operation. We have discussed two alternative neans for rreeting this requirtrent in reetings with the applicant on Noverter 13, 1975, April 1, 1976, and April 14, 1976.

One alternitive proposed by the applicant in Arcrdment 51 consisted of the a.3 t oira t i c tripping of residual heat re" oval purps by a Kw water level signal from the refuelirg water storage tarJ r to assure the maximum possible time for manual saitchover of pump suction to the containrent su p lines and manual restarting of the pur ps.

The staff cencluded tnat a' tomti' tripping of the pumps was unacceptable as an automatic backup becaus_ it would not assure a continuous sJpply of w3ter to the Core in the event the manual swittnoier procedure was not started.

The other alternative proposed by the applicant in fcerdent 55 consisted of the autonatic opening of co. tairrent sur p line isolation valves by a level signal from the refutline water storage tark The level set point will te the lowest calculated lewl for which the automatic back up switchover can te safely made while continuously supplying + ter to the cuction of the "sidual heat ren wal pun s.

Upon annuncietion of a low level alam, the cperator will perforn a r ar,

witchover in accordance with i

the energency operating procedures established for the uriginal design of the system.

$hould the operator fail to perform the manual switchover, the pumps would te auto-matically aligred to take suction from the spilled water in the containment before all of the stored water is used. The containment sump isolation valve automatic actuation circuits will be designed so that a single failure would rot unacceptably degrade safety functions including isolation of the containrent surp lines during plant operation arJ safety injecticn, adiition of sodium hjdroxide to containment spray water for removal of iodine from the containment atmosphere, and provision of adequate rargin for vortex-free operation during recirculation.

As indicated in Section fs.3.3 of this report, additional information confirming vertex-f <ee operation will be provided for our review. Our evaluation of instrumen-taticn and control circuits for opening sep valves is provided in Section 7.3 of this report.

We conclude that the design bases for an autoratic bac!up to the Lanual switchover of the energency core ccoling systen frcn the injection rode of operation to the re-circulation mode of operation are acceptable for the f arley plant u.3.3 Tests and Insjections Subsequent to the issuance of tre Safety Evaluation peport cn May 2, 1975, we fou, that the applicant proposed preoperatianai *0sts of the erergency core cooling sfster in the recircul' ion rode did not meet the recorr endations as stated in Paragraph C.l.b(2) of.egulatory Guide 1.79, "Preoperational Testing of En ergency Core Ccoling Systems for Pressurized Water Peactors. ' We require that preoperational testing of the recirculation rode include taking suction through the containment surp

. lines to verify adequate vortex control and acceptable pressure drop through the inlet screen, suction lines, and valves.

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The containment surp lines for the Farley plant tJke suction from the containment floor, rather than f rom a surp below floor level. There ere two sump lines for tht emergency ccre cooling system and two sump lines for the containment spray systen. As shown in FSAR Figure 6.2-18, the vertical sump line inlet is about one foot above the containrent floor and a rectangular box-shaped structure with screening and grating on the vertical sides covers the inlet. As indicated on FSAR pages 14.1-R and 3A-1.79.2, the apslicant had not planned to take water through the sura lir es to test vortex control.

In a meeting with the applicant an November 13, 1975, we advised the applicant of two acceptable alternative means for meeting the recomendations of Pegulatory Guide 1.79.

(1) Provide an evaluation of parameters that are significant to the deternination of adequate vertex control at the s e p inlet and a comparison of significant para-meters for the Farley plant with those for a similar facility in which tests have demonstrated acceptable performance.

(2) Test the containment sump line inlet and screen arrangerent.

Cur letter to the applicant dated January 12, 1976 requested that the FSAP be a ended to describe an acceptable means for reeting the objectives of Fegulatory Suide 1.79.

In April 1976, we visited the plant to see the physical arrangement of the sur p inlets in Farley Unit I and had further discussions ;ith the applicant regarding tests of the ECCS in the recirculation node. The applicant plans to test pressure drop in the sump lines by connecting two surp inlets, supplying water from a storage tank, and measuring pressure drops in the line for several flos rates. Analyses to den cnstrate adequate vortex control will be tased on test results from a mcdel of the Farley plant vortex breaker, screcning, and surp line inlet georetry. A description of these analyses and tests was provided to the staff in a June 29, 1976 letter from the applicant.

The applicant expects to have test results by the end cf 1976 and a final test report in March 1977. The scheduled Farley Unit I fuel loading date is February 15,1977 and the start of power escalation tests is about one month later. An operating license may be needed prior to our review and acceptance of the test report. If so, we will condition the license to limit operation at or below five percent of rated power until an acceptable test report has been provided to the staff.

We conclude that the applicant has described an acceptable means for demonstrating performance of the erercency core cooling system in the recirculation mode. However, we will evaluate confirratory test results prior to authorizing operation of Farley Unit I above five percent of rated power.

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6.4 frergency Shutdown CnolTn M anability In Supplenent No. I to ot r Saf ety Evaluation Peport, we identified the need for addi-tional information regard ng the backup air supply for the r,ain steam relief valves.

The backup air supply could be utilized should a high energy line break in the main steam valve room incapacit. te the normal air supply to the valves and mak e the room inecessible for manual va'se operation.

Amendments 52 and 53 to the FSAR state that an assured backup air supply system will be provid;d fer the main steam relief salves to allow controlled atmospheric venting of the steam generators, in conjunction with auxiliary feedwater acdition, during the initial cooldo n stage, until the primary coolant temperature and pressure permits residual heat renoval system initiation. The system will include redundant air compressors, pressure regulator and piping, and will te designed to seismic Category I requirerents. The compressors will be located in an accessible area isolated from the rain steam valve roon In Supplement No. I to our Safety Evaluaticn Peport we also identified the need for additional infomation regarding the environmental qualification of solenoid valves that control the air supply to the main steam isolation valves. FSAR Section 5.5.5.1, Amendment 53, states that the manufacturer of these solenoid valves used standards and high temperature co"'ponents that are the same as those it used to manuf acture similar pilot valves for use inside containrent. The containment pilot valves have pass >d tests in a steam environment of 20*F and 90 pounds per square inch for cne hour followed by steam at 290'F and 56 pounds por square inch for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. These envircn-rental conditions envelope the conditions that could occur in the main steam valve room for hig! energy line breaks in that room.

Based on our review, we conclude that the applicant has adequate'y designed and pro-tec*ed areas and systems req; ired fcr safe plant shutdown following a postulated high ynergy pipe failure, including the design of a backup air supply for the main stean relief valves and the design of solenoid va:ses that control air supply to the main stean isolation valves.

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7.0 INST M'f N T ATIM 21 (m.T%Li 1.3 f nl neered 5 f oty_ F ea tures Ac tua tion and Con trol i

t Ine applicant has revised the design of the ergir:eered 5 3 fety f ea tures acttation circuits in two areas: (1) the initiation of rain sten line isolation, and (2) the automatic opening of containrent surp valves.

Curing its redesign of the auxiliary feedw3ter discharge piping, the applicant modi-fitd the logic for initiation of main steam line isolation to assure rcre rapid isolation of the two intact steam gererators in tre event of a rain feedwater line upture.*

The arplicant described the revised logic in a reeting on February 4, 1976 ano n Arenaaent 54 to the FS M.

For the cid logic systen, steam line isolation i

would have teen initiated for high steam line flow coincident with either lcw stes line pressure or Icu-low reactcr coolant te"perature. For a postulated feedwater line breu, steam flow may not te suf ficiently high to generate a signal until the af fected steam generator is emptied. Since each rain steam line has two forward-flow stop-check valves (corpared to one forward-ficw stcp-check valve and a reverse-ficw check valve in most other WestingFouse nuclear stead sJpply systeUs), a rain sted9 line isolation signal rust be generated to isolate the ste3m g:nerators, for tne modified logic systen, main steam line isolation will be initiated for eithrr low stcan line pressure in two of three loops cr high steam line flow coincident with low-low reactor coolant te perature. We have reviewed the new logic diagiars provided in FS AR Figure 7. 3-1 and conclude that this logic tysten will perform the design safety functions and is acceptable.

As a back up to manual alignment of the ecergency core cooling syster for the re-c' cilation mode of operation, the applicant has revised the design to include a;to-matic gening of containment sump line isolation valves by a low-Icw level signal f ron the refueling v.ater storage tank.**

In Vend ent 5f and in neetings on April 14, 1976 and April 15, 1976, the applicant described the additional instru'xnt logic that will be used. In ;cer "ent SS the applicant provi nd drawing; of the instruments and controls. Power to actuate the valves will be supplied from four erergency buses (two alternating current buses and two direct current bnes) in a two-out-of-fo;r logic. Four separate signal c6bles, one for each contair. rent isolation valve, will

• Sae Section 9.3.1 oT tnis report for our evaluation and acceptance of tnis rcJified auxiliary feedwa ter systen design.

• • See 5ection 6.3.2 of this report for evaluation and acceptance of this backup systen.

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be run from separate level senscrs in the refueling water storage tank.

The level set point will te tFe level at which water can be safely and continuously supplied to the residual heat renoval pu ps.

The residua? heat removal pu ps will rct be auto-ratically tripped; however, they n'ay be rianually tripped in the ranual switchover to the recirculaticn node of operation. The valve 3ctuaticn instrurents and ccntrols will reet the single failure triterion and the other criteria identified in FSAR Secticn 7.3 for engireered safety features actuatico systems. We conclude that the proposed instru ents and controls for automatic opening of the contain+ent sump isolation valves fol? ) wing a postulated loss-of-coolant accident are acceptable.

We conclude that the redified engineered safety features actuation system desians reet our criteria and are acceptable.

7.B Cable Separation and Identification Criteria Supplement No.1 to cur Safety Evaluation Report concluded that component and cable separation criteria for safety significant systems were acceptable subject to accept-able revisions to tne FSAR in two areas: (1) bypassed and inoperable status indication of systens, and (2) assessrent of the fla e retardance of the car;lo installed in the emergency power beard. These two areas were revised by FSAR k enanent 53.

Our evaluatico is provided in Sections 7.P.1 and 7.8.2 of this report.

Supplement No. I to our Safety Evaluation Peport errcneously concluded that rarking of conduit at each end and eich sid_ cf wall penetration; was acceptable. In a reet-iiig on Noverter 13, 1975, we advised the applicant that conduit should also be marked et 20-foot interval < alcrg the length. Color code identification criteria. vere accordingly revised by k,ertnent 53.

Our evaluaticn is provided in Section 7.8.3 of this report.

7.8.1 Bg assed and Irocerable Status Indication FSAR page 3A-1.47-1 states that a nanually-operated light display board to be installed in the main control roc cn a single panel will indicate those engineered safety fea-tures that are bypassed by deliberate operator action. Celiberate operator action neans either the granting of permission to perform any activity that would or ccu!d affect a safety significant system or the perforrance of such an activity. The indicating lights will be arranged to clearly show all portions of redundant systens that are bypassed or rade inoperable by any activity such as the performance of periodic tests or naintenance on one channel of instrumentation, one train of electri-Cal power supply, or che Component of a systen. Adninistrative procedures outlined in the technical specificat.ons will require that the operator's pernission be cbtained prior to initiating any activity that would or could affect a safety signifi-cant systen.

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The bypassed and incperable status indication described in FSAR knendment 53 will not aJtoratically indicate bypassed or inoper3ble portions of safety significant Syste~s as reconnended in Pegulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Fower Plant Safety Systems.

However, the indication systen will have corronent lights interconrected so that ranually actuatirg the indicating light for are inoperable corscnent or portion of a systen will actuate indicating lights for all other components or portions of safety significant systens that are rendered inoperab'e.

We conclude that tne proposed bypassed and incperable status indication systen will adequately supplecent administrative precedures and is acceptable.

7.8.2 Emergoncy Power Board _

FSaR page 8.3-30a describes the fire protection for one redundant train of control cables in steel conduit routed through sections of the erergency power board that contain control cables and switches for the other redundant train. The four-inch conduits will be shielded fron potential fire sources by a metal shroud around the conduit. The shroud will be insulated on the top and rear surfacts by two-inch thick Cera Forn rated to withstand continuous exposure to a fire temperature o' 2100*F.

Calculations indicate that for a fire in the board, it would require rore than three hours for cables in the conduit to reach 194*F ar.d more than 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> for cables in the conduit to reach 244*F.

In a meeting cn November 13, 1975, the applicant provided additional information re-garding a potertial fire in a section of the emergency power board. The coplicant stated that the c.cnduit would have suf ficient insulation that a fire which completely burned all combustible material in that section (cab e insulation, plastic wireways and other plastic assemblies) would not make the cabita inside the conduit inoaerable.

Anendrents 54 and E6 have provided sufficient information to support this statement.

We conclude that f ar redundant safety significant cables in the Farley e ergency power board, the proposed insulated shroud around the steel conduit is an acceptable means for protecting cne cable in the conduit from a pcstulated fire that is assumed to completely burn all combustible material within that section of the cower board.

l.8.3 Color Code Identification Crite_ria FSAR Amendment 53 revised the color Code identification criteria to require that con-duit and four-inch channels be marked with the appropriate color at 20-foot intervals along their lengths, where visible, ard at each siGe of wall and floor penetrations.

In addition, ccnduit and four-inch channels are markeo with the appropriate color and the last six digits of the raceway number at each end.

Cable trays are narked with the appropriate color and the last six digits of the raceway number at each end and at 20-foot intervals along their length.

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)nclu - that the criteria in FSAR Section 7.3.1.5 as awnded through trendment 53 We d

are acceptable because they allow a deterrnination by visual inspection that the cri-t eria f cr separation of redJndarit electrical trains have t:een satisfactorily irrplement-ed in the plant.

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9.0 aux! LIAR ( SYSTEMS 9.2 Fuel Storage anj Hardlinq 9.2.4 Fuel Handlin1 ystem S

Our "ay 2,1975 Safety Evaluatic' Peport concluded that the fual ter.dlirl syste design as described in the Firil Safety Analysis Peport, as a~ ended thror;h ren tent afi, is acceptable provided the spent fLel cask handling crane is shawn to reet tne staff's sir.gle failure requirc ents Our Octcter 1975 5;pple wnt No. I to the Safets Evaluation ?eport cercluded that the revised crane design and safety features, as described in c re*ent 47 to the FSAR, dia not reet tre stif f's requirenats in the Auxiliary and Pcwer Ccnversion Syste-s Branch Technical Fosition M CSB 9-1, "Overreid Handling Systens for nuclear Pcwer Plants.

FSAR Amend ents 50, 52 and 53 provide additienal inf er ation regarding the spent fuel cask crane design criteria, perfernance specifications, safety criteria, inspection, and testing.

WP have evaluated this revised descripticq of tre cr3ne by cq-paring it with our requirerents in 5 ranch Technical FCsition APCSB 9-l.

The s; ent f uel cask cr3re is an outdoor overhead travelir.g unequal leg gantry trane that is shared t, Units 1 and 2.

Ine ain toist is rated at 125 tons. The crane is designed to seis-ic Category I require!ents 3nd will withstard, without less of lo3d carrying functiers, tre forces resulting fron a safe shutdoan earthquD e w*en the crare is harfling the full rated load. Tne cask handling crane gantry is also designed to prevent overturnir:q or horizontal rove ent in tre event of a tornado. The crane has redund3ncy of brakes, geir trains ree,-ing systen, and Icad attachina points. Crane control components and systems are designed to fail safe. Although the reeving system does not r'eet our guidelines for wire rope strength safety nargin and fleet angles, the applicant has provided acceptable alterna r,e procedures to inspect the rope and to replace it for specific conditions that would indicate the occurrence of a significant decrease in the rope strength safety nargin. Cold proof tests will be run at 125 ]f the rated load capacity to demonstrate that rated load can be handled at te peratures equal to or above the test te perature. The technical specifications will include requiraments for wire rope inspections art cold proof tests in accord 3nce with the precedures described in FSAR Section 9.l.4.4 The design of devices for lif ting the spent fuel cask has not been provided. A condition in the operating license will require the submission for our approval of a report describing the design of lif ting devices for spent fuel casks prior to use of the crane to handle the casks.

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Wt conclude that the srent fuel cask crane design, inservice inspection program, ard proof test progran equal or aceed the staf f's regairments in JCSB 9-1, ' O,crr ead Handling Systers for Naclear Power Plants,' and are acceptable.

9.3 Ccolirq P ter Snters 9.3.1 Aur ili 3r1 eec%1ter S v stm F

The auxiliary feedwater systen is designed to supply water to the ste r generators f or removal of sensible beat and decay beat fron the reactor coolant syste, when the nontal f eedwa ter systc.. is not a vailable. The system will be utilized daring nornal startup and shutdewn, in the event of anticipated operaticnal occurrercCS such as lor of offsite po-er, and also in the event of accia nts.

The system is desigred to s

seismic Category I requirerents and is protected from tornado missiles.

Tne auxiliary fee %3ter system contains two motor driven pu ps and one turbine driven pap.

Stean 54 ply to the turbine is taken from one of two rain steam lires upstcean of the rain ster isolation valves. The notor driven pur.ps are cornected to separate er.ergency power bases. The motor driven peps will start automatically in the event of a safety injecticn signal, loss of of fsite power, low-low water lesel in any cre steam generator, or tripping of both steam generator feed pur ps.

The turbine dr;~,

pur p will start astomatically in the event of undervoltage on two reactor coolant rumo bases or low-low water level in to out of three steam generators. Auxiliary f tedwater flow will be adjusted by re ote operatej fb cnntrol valves. Each rotor driven pump h35 a norlinal cpacity of 350 gallrns per minute while the turbine driven vp h3s a nominal capacity of 700 gallons per rinute. For normal plan' ccoldown, one motor driven pump feeding 350 gallors per minate to all three steam generators is adequate. For safe cooldown in the event of a main feedwater lire break, 150 gallcns per minute supplied to the two intact steam gererator3 is adequate provided this rate of fceding starts at one ninute following the break and is increased to 350 g311ons per minate within 30 minutes following the break.

Our previous evaluations have resulted in design change, to the auxiliary feedaatar system The May 2,1975 Safety Evaluation Peport concluded that tre auxiliary f eed-water system described in the FSAP through Amendment 47 was acceptable provided it was denonstrated that an adequate quantity of feedwater could be suppl'ed to the two intact iteam generators in the event of a main feedwater line rupture and coincident fail *are of any one of the notor-operated valves in the pump discharge lines. With the design then proposed, a 'purious closure of ore valve could have result?d in feedwater being supplied to only one of the two intact steam generators In the June 1975 Advisory Comittee on Peactor Safeguard neeting and suosequently in A ene.nt 50 to the FSAR, the applicant stated that there would be ample time for an operator to go to the purp room and manually open any spurinusly-closed valve that prevented water flow to an intact steam generator. The October 1975 Supplement No. 1 to the Safety Evaluation Report concluded that the consequences of an operator's failure to open an inadvertently closed valve had not bern analyzed ard requested either system 15 eU$*

design modifications or analyses to show that water automatically supplied to steam gererators would be adequate until cperators co;1d align systers to assure 1cng-tern cooling. In a february 4,1976 2eting and in subsequent FSAR amendments, the arpli-cant described a revised design of the auxiliary feemter system that will autcmatically provide at least 150 gallons per minute of teedatar to the two intact steam generators from one minutt to 30 minutes fo'scwing a feedwater line break coircident with the nost adverse single active failure of a corpcnent.

In FSAR Section 15.4.2.2.2., the applicant has provided an analysis of the performance of this revised system following a main feedaater pipe bre 3k.

The applicant has concluded that during the 30 minute time interval following the pipe break, the reactor coolant pressure would not exceed the pressurizer safety valve setroint nor would the w1ter lesel in the reactor coolant systen fall below the top of the core.

At 30 minutos f ollowing the break, the auxiliary feed *ater purp discharge piping system will be manually aligred, if necessary, to suLply at least 350 gallons rer ninate to the two intact steam generatcrs which is adequate to remove core residual teat at that tinc.

In FSAK Table 6.5-2, the applicant his provided the results of i failure mode and eff ects analysis which indicate that sufficient flow to effect cold shutdown will be provided to at least two ster generators with various cortinations of a single active failure and a high energy pipe break.

We have evaluated the revised design of the aJMiliary feedwatcr system as described and analyzed in the F$AR as verded thror;5 Arend ent 56 by cor paring it with our requirements in Auxiliary and Power Conversion Systers Branch Techn' cal Fositico APCSB 10-1. " Design Guidelires for Auxiliarj Feedwater Systen Pump Pive ard f eaer Supply Diversity for Pressurized Water Feactors."* The systen includes provision of flow limiting orifices M purp discharge lines, two notor-operated valves in e*ch pump discharge line that are manually-eperated from the control room, and discharge lines fron the turbine driven pump to each main feedaater line. During power opera-tion, the motor-operated stop check valves in auxiliary feedwater lines (FSAR Figure 5.5-1, valve nurbers 3350A, 3350B, 3350C) will have Ocwer renoved by opening the switch a t the rotor control center. ** For a large break in the secondary coolant system, the two mctor driven purps and the turbine driven pump are started automatically and deliver at least 150 gallcns per minute to two stea-' generators within one minute following the pipe break. The auxiliary feedwater purp discharge piping is manually aligned to deliver at least 350 gallons per mirute to tv.o stear generators within 30 minutes following the pipe break. We will require a preoperational test to verify the design of the flow-limiting orifices and of the auxiliary feedwater pump performar.ce for normal shutdown and for sirulated accident ccMitions.

Si e bibliography' In' Appendix B to this report.

The safety function of these valves is to isolate the containment ir the event of excessive leakage from the system during long-tem coaling following an accident; however, they are required to be open to supply feedwater to the stean generators.

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Based on our review, we conclude that the a;xiliary feedwater system design is in conformance with our technical position APCSB 10-1 regarding diversity of power sources, system flexibility and redundancy including the corbination single active failure and high energy line break and is, therefore, acceptable.

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10.0 STEAM AND POWER CONVERSION SYSTEM 10.5 W in Feedwater Systen in its June 12, 1375 report cn the Farley plant, tre Advisory Coroittee en Reactor Safeguards noted that a potentially da aging water haarer phenorenon has been observed in feedwater inlet piping of some pressurized water reactor steam generators and that cnrrective reasures were planned for the Farley plant. The Advisory Comittee on Peactor Safeguard recorrended that corrective treasures implerented on the Farley plant be experimentally verified.

In FSAR Arendrent No. 53, the applicant has stated that nodifications to be made to the Farley feedaatei system anc' stean generators include installation of J-tubes en the steam generator feed ring, elimination of the feed ring botton holes and reduc-tion of the horizontal length of tha feed line adjacent to the steam generator. The applicant further states that, because of the similarity between the Trojan and Farley stear generators and since the Trojan tests did not produce any water hammer for the full range of steam generato, operatirg pressures and auxiliary, feedsater addition rates (410 gallons p9r minute per steam generator), neither ?r,alyses ner an, additional in-plant tests are planned to be perforned at this time. However, tho applicant will consider the recoTendations resulting from completed NRC and indus:

studies of the.ou3es and et tects of water

• icwe in feedwater pipinc,.

We are currently evaluating the feedwater pipe water hamer pr olen en a generic basis, tithough the Trojan plant tests that were perforred en a Westinghouse steam generator with "J-tubes" en the feed ring re;ulted i., no significant water harrer for the maxirum rate cf steam generator water level rise available frora the auxiliary feedwater systen (440 gallons per minute per steam generator), the tests were perforced

.sino ctor coolant pump heat to cbtain operating coolant pressure and temperature.

Tes she Calvert Cliffs plant which has a Ccrbustion Engineering steam cenerator witnout "J-tubes,' that were perforced using nuclear heat to obtain operating coolant terperature and pressure resulted in limiting the maxinun rate of steam generator water level rise to 1.2 inches per minute (equivalent to 168 gallons per minute per steam generator) when the water level has fallen below the feed ring to prevent recurrence of significant flow instability (water ha mer). Apparently sten generator design, feedwater piping arrangement, water level rise rate and primary coolant heat source are significant parameters in the water ha~rer phencmeron. We hase concluded that there is insufficient test data for the rcnge of significant parameters that occur at different plants to assure accurate pradiction of the onset of significant flow instability for ;ne plant based on test data for another plant. We therefcre reaffim our requirenent stated in Supplment No.1 to the Safety Evaluation Peport t~

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that tests be run daring initial power escalation of Farley Unit 1 to demonstrate that significant flow instability (water harrer) will not occur for the modified pipir.g and feed ring design at auxiliary feedwater flow rates up to the maximum available rate. We will ir.clude satisfactory cor pletion of these tests as a condition in the operating license.

Subject to our acceptance of the auxiliary feedwater tests, we conclude that the modified desiga of the main feedwater system piping and feed ring is acceptable.

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15.0 ACCIDENT ANALYSIS 15.2 Thermal and Hydraulic Analysis 15.2.2 Accidents 15.2.2.1 Introduction Section 15.2.2 of the Safety Evaluation Report issued on May 2,1975 concluded that the results of thermal and hydraulic analyses of design basis accide ts are accept-able subject to an acceptable loss-of-coolant accident loss-of-coola accident analysis with an approved Westinghouse emergency core cooling Syste > u orgency core cooling syten evaluation model in confornance with fppendix K to 10 CFR Part 50.

In a letter dated June 30, 1975, we provided guidance for the preparation of loss-cf-coolant accident analyses regarding the spectrum of breaks, potential boron precipita-tion, single failure analyses of emergency core cooling system, flooding of safety related valve motors inside containment, contain ent pressure calculation, and the reflood evaluation model.

In its letter dated September 15, 1975 and in FSAR A~endment has. 39, 46, 47, 51 and 53, the applicant has provided 'oss-of-coolant accident analyses to demonstrate conforrance with Appendix K to 10 CFR Part 50 and to respond to our letter of June 30, 1975. Our evaluation and conclusions regarding loss-of-coolant accident analyses are provided in Sections 15.2.2.2, 15.2.2.3, and 15.2.2.4 of this report. Our evaluation and conclusions regarding design, erergency operating proce6res and technical specifi-cations for the e-ergency core cooling system are provided in Section 6.3 of this report.

15.2.2.2 Loss-of-Coolant Accident Analyses The applicant submitted loss-of-coolant accident analyses in FSAR Section 15.4.1 that considered a spectrun of reactor coolant systen pipe ruptures. A three-break spectrum, specific for the Farley plant, was submitted and an applicable generic plant sensi-tivity study was referenced in conforrity with the break spectrum requirements in Section 50.46(a) of 10 CFR Part 50. The analyses submitted were perfor"ed with cn acceptable evaluation redel which is in conformance with Appendix K to 10 CFR Part 50.

Our approval of the Westinghouse model was provided in two letters to Westinghouse dated May 33, 1975.*

The analyses identified the worst break size as the double-ended cold leg guillotine break having a coefficient of discharge of 0.4.

The calculated peak cladding

• See Bibliography in Appendix B to shis report.

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temperature of 2133 F is within the acceptable limit of 2200 F specified in Section 50.46(b) of 10 CFR Part 50.

The calculated raximum local retal/ water reaction of 5.7 percent and total core wide metal / water reaction of less than 0.3 percent are well below the allowable limits of 17 percent and I percent, respectiviely. The analyses are based on fuel assemblies having a 17 x 17 array of fuei rods, a total power peaking factor of 2.32 and reactor operation at 102 percent o' the rated core power level of 2774 megawatts therral. These results are valid for all first cycle fuel.

Since analyses were presented only for all three reactor coolant loops in operation, the technical specifications will also include a condition that power operation will not be pernitted with one loop out of service.

Westinghouse inforced the staff in a letter dated October 28, 1975* that the observed fuel rod bowing is larger than anticipated in operating pressucized water reactors that use fuel designed by Westinghouse. In its Interin Safety Evaluation Peport on Westinghouse Fuel Rod Bowing, April 1976, the staff has evaluated the effects of this increased rod bowing on load power peaking and on peaking factor limits cited in the technical specifications. For fuel assemblies having a 17 x 17 array of fuel rods, such as the Farley fuel assemblies, the staf f concluded that the peaking factor limit included in the present technical specifications is adequate for fuel burnups to 24,000 regawatt-days per metric ton of uranium, which corresponds to the end of the

.cond fuel cyr e.

During the third fuel cycle it is likely that the decrease in core flux p -0; will more than offset the increased local peaks due to rod bowing.

The technical specificaticns will include a requirement that the adequacy of the power peaking limit be confirred prior to the start of the third fuel cycle.

15.2.2.3 Containment Pressure Evcluation The containnent pressure used in m a loss-of-coolant accident analyses for the Farley plant were calculated using the Westingnouse emergency core cooling systen evaluation model. The sta'f reviewed Westinghouse's nodel and published a status report on October 15, 1974 which was amended fiovember 13, 1974. We concluded that Westinghouse's containment presstre redel was acceptable for emorger.cy core cooling systen evaluation.

We required, howevtr, that justification of the plant-dependent input parameters used in the analysis be s ibnitted for our review of each plant.

Information regarding t'e containment pressure calculation was submitted in FSAR Section 15.4.1.1.3 and Table 15.4-3.

The contair. ment net-free volume, the passive heat sinks, and operation of the containment heat renoval systens were evaluated in a conservative manner for energency core cooling system analyses. For emergency core cooling system analyses, it is conservative to minimize containment pressure, since this will increase the resistance to steam flow in the reactor coolant loops and reduce the reficod rate in the core. This evaluation was based on measurements within the containment and as-built drawings to which additional nargin was added. The

• See Bibliography in Appendix B to this report.

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containment heat removal systens were assumed to operate at their maximun capacities and the spray water and service water temperatures were assumed to be at the lowest values exrecteJ during plant operation, We h3ve ccncluded that the plant-dependent information used for the emergency core cooling system containr:ent pressare analysis for the Farley plant is conservative and therefore the calculated contairrent pressure is in accordance with Appendix A to 10 CFR Part 50.

15.2.2.4 Conclusions Based on our review, we conclude that:

1.

The loss-of-coolant accident analyses that were performed by the applicant are wholly in conferr,ar.ce with the regairements of Appendix r to 10 CFP Part 50.

2.

Ibe emergency core cooling systen cooling performance conforms to the peak cladding temperature and maximum oxidation and hydrogen generation criteria in Section 46 of 10 CFR Part 50, provided fuel rod bowing is accoanted for in the peaking f actor limit 3.

Emergency core cooling system cooling performance will be adequate despite any postu'ated f ailure of a single component, provided all three reactor coolant system loops are operating and power is luthed cJt of Certain valves idtitifled in sect:en 6.3 of this recort.

4.

Adequate syster:s exist to provide long term cooling to the reactor vessel, provided the ererger.cy operating procedures are rodified as required in Section f>. 3 of thi s report.

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16.0 TECHNICAL SPECIFICATIONS In the Safety Evaluation Report and Supplement No. I we identified siqnificant specifications resuiting from our review that are not included in the f 5AR but that will be included in the technical specifications to be issued with the operating license. Subsequently, we have concluded that specifications which will be renoved from the license after satisfactory completior: during iritial plant operetion snould be incorporated as license conditions rather than technical speci fications (see e.g.,

Items (2) ar.m '4) below). The following list identifiec addi tior,al speci fica tiens and special reports that will be required by either license conditions or technical specifications as a result of the review described in this report.

(1) Specifications for administrative ccatrol of safety significant valves that could, if improperly positioned by a single failure, unacceptably degrade the capability of an engineered safety feature to perforn its safety function (5ecticos 6.3.2 and 9.3.1 of this report).

(2) A special report providing the results of tests to verify acceptable perfonrance of the erergency core cooling system in the recirculation mode of operatico (Section 6.3.3 of this report).

(3; Specifications for the spent fuel cask crane (a) to define periodic inspections and criteria for replacement of the wire rope, (b) to define air temperature as a limiting condition for crane operation, based on cold proof tests, and (c) to preclude its use to handle spent fuel casks until the design of devices for lif ting the cask have been evaluated by the staff (Section 9.l.4 of this report).

(4) A special report providing the results of preoperational tests to verify design flow rates of the auxiliary feedwater systen for simulated accident conditior s.

(Section 9.3.1 of this report).

(5) Specifications of numbers of operating reactor coolant purps and pe k fuel rod linear heat rating, including an allowance for fuei rod bowing, as limiting conditions for plant operation (Section 15.2.2.2 of this report).

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APPENDIX A SUPPLEMENT TO THE CHRONOLOGY OF THE PADI0 LOGICAL SAFETY REVIEW September 15, 1975 Letter from applicant providing infonration regarding ECCS analyses requested in our June 30, 1975 letter.

September 15, 1975 Letter from applicant transmitting initial draft of Farley Technical Specifications.

October 3,1975 Letter to applicant transmitting staff's Supplement No. I to the Farley Safety Evaluation Report (RSER-1).

October 17, 1975 Amendment No. 52 from applicant providing revised FSAR pages in partial response to outstanding issues identified in staff's SSER-1.

October 20, 1975 Letter from applicant advising that Westinghouse steam generator tube implants will be included in Farley Uni t 1.

October 22, 1975 Meeting with applicant to discuss assymretric loads on reactor vessel supports resulting from a LOCA.

October 28, 1975 Letter from applicant transmitting Pevision 4 to the draf t of Farley Technical Specifications.

November 7, 1975 Meeting with applicant to discuss preparation of the Technical Specifications for the Farley plant.

November 13, 1975 Meeting with applicant to discuss outstanding issues in Supplement No.1 to the staff's Safety Evaluation Report.

November 26, 1975 Letter i. forming applicant of a potential safety quet-tion regarding design of reactor vessel support systems to withstand LOCA.

December 19, 1975 Letter from applicant responding to our November 26, 1975 letter and advising that evaluation would be completed in the Fall of 1976.

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January 5, 1976 Arenanent No. 53 from applicant revising FSAP to include additional information about the crane design and respond to cur Aoril 30, 1975 letter.

January le, 1976 Letter to applicant requesting tests of the recircula-tien mode of operation of the emergency ccre cooling system.

February 4, 1976 Meeting with applicant to hear and discuss redesign of the auxiliary feedwater system.

February 20, 1976 nrendment No. 54 from applicant revising FSAR to include information abe ' the revised design of the auxiliary feedwater syster February 23, 1976 Letter from applicant transmitting revisions to the security plan.

February 23, 1976 Letter to applicant requesting additional information concerning radioactive waste trea tnent equi pment per-formance compared to requirenents in Append L I to 10 CFR 50.

March 19,1976 Letter from applicant adsising schedule for providing infornation requested in our February 23, 1976 letter.

March 30, 1976 Letter to applicant advising of staff meeting with applicants in our Pegio-II office of I&E April 1, 1976 Mee*"3s with applicant at the Farley plant to view April 2,1976 equipment and discuss outstanding issues regarding automatic backup for ECCS canual operations and ECCS recirculation tests.

April la, 1976 l'eetings with applicant to discuss outstanding issues April 15, 1976 regarding automatic backup to ECCS ranual operations ar.d redesign of the auxiliary feedaater systen.

, April 15,1976 Meeting with applicant regarding steam generator tube implants.

April 20, 1976 Meeting with applicant to discuss conments on the draft Farley Technical Specification 3.

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May 4, 1976 Letter from applicant agreeing to provide results of inspection and tests of tube implants in the steam generator.

May 10, 1976 Letter to applicant providing draft Technical Specifica-tions for i~plerentation of requirements in Appendix I to 10 CFR 50.

May 14, 197:

Amendment No. 55 from applicant providing revised FSAR pages in par tial response to outstanding issues identified in staff's review.

June 24, 1976 Letter to =pplicant requestir) additional i nf orna ti c o regarding reactor pressure vessel supports.

Jure 29, 1976 Letter from applicant providinq outline of model tests sirulating containment surp inlets.

. r,e 30,1976 Letter frca applicant providing revisions to the Security Flan for the Joseph M.

Farley Nuclear Plant regarding patro frequency and equal eroloynent reovisions.

July 2, 1976 Letter to applicant requesting the identification and schedule for design changes pursuint to our recomrenda-tivns regarding anticipated transients without scram.

July 2, 1976 retrdment No. 56 from applicant providing reviscd FSAR pages in partial response to outstanding issues identified in staff's review.

July 29, 1976 Letter from applicant providing schedule for response to our June 24, 1976.etter.

August 4,1976 Application for withhciding proprietary inforr:ation from public disclosure. Ssbject: Steam line break release to the containrent.

August 6,1976 irendment '.o. 57 from applicant providing revised FSAR pages to correct material used for h*gh energy pipe restraints and correct errors.

August 20, 1976 reendment No. 58 from applicard providing instrunenta-tion ard control drawings for automatic backup to ECLS manual switchover, e

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September 27, 1976 feendment No. 59 from applicant providing additional infomation on ECCS manual ;witchover proce+;re.

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APPENDIX B SUPPLEMENT TO THE BIBLIOGRAPHY Engineered Safety Features Regulatory Guide 1.79, Preo?erational testing of energency core cooling systens for pressurized water reactors, USNRC, September 1975.

April 1, 1975 letter from C. L. Caso (Westinghouse) to T. ti. Novak (USNRC) regarding boron concentration increase in reactor due to boiling during long-term cooling following a loss-of-coolant accident.

Auxiliary Systems Regulatory Standard Review Plan, Auxiliary and Power Corversion Systems Branch Technical Posi-tion APCSB 9-1, " Overhead Handling Systems for Nuclear Power Plants," attached to Section 9.1.4,

" Fuel Handling Systems,' USNRC, November 24, 1975.

Accident Analyses May 30, 1975 letter from D. B. Vassallo (USNRC) to C. Eicheldinger (Westinghouse) transmitting staff's evaluation of the Westine ouse energcncy core cooling system evaluation model.

May 30, 1975 letter from D. B. Vassallo (USNRC) to C. Eicheldinger (Westinghouse) transmitting staff's evaiuation of WCAP-8471, "The Westinghouse ECCS Evaluation Model: Supplecental Information" October 28, 1975 letter from C. Eicheldinger (Westinghouse) to D. B. Vassallo providing an analysis of fuel rod bowing data.

" Interim Safety Evaluation Report on Westir.ghouse Fuel Rod Bowing", Division of Systems Safcty, Office of Nuclear Peactor Regulation, U.S. Nuclear Regulatory Comnission, April 1976.

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