ML20126K051

Forwards Instrumentation & Control Sys Branch Request for Addl Info Re OL Application Review.Response Should Be Provided in Amend to FSAR within 120 Days.Background Info Encl

ML20126K051
Person / Time
Site:	South Texas
Issue date:	04/30/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Oprea G HOUSTON LIGHTING & POWER CO.
References
NUDOCS 8105110135
Download: ML20126K051 (29)

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1. R 3 0 1981 i

Docket Nos.: 50-498/499 r

Mr. George W. Oprea, Jr.

Executive Vice President Houston Lighting and Power Company P. O. Box 1700 Houston, Texas 77001

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - SOUTH TEXAS PROJECT i

The Instrumentation and Control Systems Branch has identified four concerns that must be addressed prior to completion of its review of your operating license applications. The specific concerns are listed in Enclosure 1 to this letter. contains background information that will be useful in the development of your response.

Your responses should be in the form of an Amendment to your FSAR. You are requested to provide your response within 120 days of receipt of this letter.

If you can not meet that date please advise us what date you can meet.

If you have any question, contact D. Sells (301) 492-7792, the Project Manager.

Sincerely, W

Wh*^

Robert L. Tedesco, Assistant Director for Licensing Division of Licensing r

cc: See next page.

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Io, UNITED STATES I '* 3,.  %

NUCLEAR REGULATORY COMMISSION I

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APR 3 0 1981 i

Docket Nos. : 50-498/499 r

Mr. George W. Oprea, Jr.

I Executive Vice President Houston Lighting and Power Company P. O. Box 1700 Houston, Texas 77001 l

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - SOUTH TEXAS PROJECT The Instrumentation and Control Systems Branch has identified four concerns l

that must be addressed prior to completion of its review of your operating license applications. The specific concerns are listed in Enclosure 1 to this letter. contains background infonnation that will be useful in the development of your response.

l Your responses should be in the form of an Amendment to your FSAR. You are requested to provide your response within 120 days of receipt of this f

letter.

If you can not meet that date please advise us what date you can meet.

If you have any question, contact D. Sells (301) 492-7792, the Project Manager.

Sincerely,

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Robert L. Tedesco, Assistant Director I

for Licensing i

Division of Licensing l

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Mr. C. Robertson Lowenstein, Newman, Axelrad & Toll Houston Lighting and Power Company 1025 Connecticut Avenue, N.W.

P. O. Box 1700 Washington, O.C.

2003C Houston, Texas 77001 Melbert Schwarz, Jr., Esq.

Baker & Botts One Shell Pla:a Houston, Texas 77002 Mr. J. R. Geurts Brown & Root, Inc.

j P. O. Box 3 Houston, Texas 77001 Mrs. Peggy Buchorn Executive Director Citi: ens for Equitable Utilities, Inc.

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ENCLOSURE 1 I

t 032.0 REQUEST FOR ADDITIONAL INFORMATION 032.42 Loss of Non-Class IE Instrumentation and Control Power System Bus During Power Operation (IE Bulletin 79-27)

If reactor controls and vital instruments derive power from common electrical distribution systems, the failure of such electrical distribution systems may result in an event requiring operator action concurrent with failure of important instrumentation upon which these i

operator actions should be based. This concern was addressed in IE Bulletin 79-27 (enclosed). On November 30, 1979. IE Bulletin 79-27 was sent to operating license (OL) holders, the near term OL applicants (North Anna 2, Diablo Canyon, McGuire, Salem 2 Sequoyah, and Zimmer), and other holders of cons truction permits (CP), including South Texas Project. Of these recipients, the CP holders were not given explicit direction for making a submittal as part of the licensing review. However, they were informed that the issue would be addressed later.

You are requested to address these issues by taking IE Bulletin 79-27 Actions 1 thru 3 under " Actions to be Taken by Licensees".

Within the response time called for in the attached transmittal letter, complete the review and evaluation required by Actions 1 thru 3 and provide a written response describing your reviews and actions.

032.43 Engineered Safety Features (ESF) Reset Co,trols (IE Bulletin 80-06)

If safety equipment does not remain in its emergency mode upon reset of an engineered safeguards actuation signal, system modification, design change or other corrective action should be planned to assure that protective action of the affected equipment is not compromised once the associated actuation signal is reset. This issue was addressed in IE Bulletin 80-06 (enclosed).

For facilities with operating licenses as of March 13, 1980, IE Bulletin 80-06 required that reviews be conducted by the licensees to determine which, if any, safety functions might be unavailable after reset, and what changes could be implemented to correct the problem.

For facilities with a construction permit including OL applicants.

Bulletin 80-06 was issued for information only.

The NRC staff has determined that all CP holders, as a part of the OL review process are to be requested to address this issue.

Accordingly, you are requested to take the actions called for in Bulletin 80-06 Actions 1 thru 4 under " Actions to be Taken by Licensees". Within the response time called for in the attached transmittal letter, complete the review verifications and descriptions of corrective actions taken or planned as stated in Action 1 thru 3 and submit the report called for in Action Item 4.

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032.44 Qualification nf Control Systems (IE Information Notice 79-22)

I Operating reactor licensees were informed by IE Information Notice 79-22, issued September 19, 1979, that certain non-safety grade or i

control equipment, if subjected to the adverse environment of a high energy line break, could impact the safety analyses and the adequacy of the protection functions performed by the safety grade equipment. Enclosed is a copy of IE Information Notice 79-22, and reprinted copies of an August 20, 1979 Westinghouse letter and a September 10, 1979 Public Service Electric and Gas Company letter which address this matter., Operating Reactor licensees conducted reviews to determine whether such problems could exist at operating facilities.

We are concerned that a similar potential may exist at light water facilities now under construction. You are, therefore, reques ted to perform a review to determine what, if any, design changes or operator actions would be necessary to assure that high energy line breaks will not cause control system failures to complicate the event beyond your FSAR analysis. Provide the results of your l

reviews including all identified problems and the manner in which l

you have resolved them to NRR.

The specific " scenarios" discussed in the above referenced Westinghouse letter are to be considered as examples of the kinds of interactions which might occur. Your review should include those scenarios, where applicable, but should not necessarily be limited to them.

Applicants with other LWR designs should consider analogous interactions as relevant to tneir designs.

032.45 Control System Failures The analyses reported in Chapter 15 of the FSAR are intended to demonstrate the adequacy of safety systems in mitigating anticipated operational occurrences and accidents.

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Based on the conservative assumptions made in defining these design-basis events and the detailed review of the analyses by the staff, it is likely that they adequately bound the consequences of single control system failures.

To provide assurance that the design basis event analyses adequately bound other more fundamental credible failures you are requested to provide the following information:

(1)

Identify those control systems whose failure or malfunction could seriously impact plant safety.

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(2)

Indicate which, if any, of the control systems identified in (1) receive power from common power sources, The power sources considered should include all power sources whose failure or malfunction could lead to failure or malfunction of more i

than one control system and should extend to the effects of i

cascading power losses due to the failure of higher level i

distribution panels and load centers.

i (3)

Indicate which, if any, of the control systems identified in (1) receive input signals from common sensors. The sensors l

considered should include, but should not necessarily be i

limited to, common hydraulic headers or impulse' lines' feeding i

pressure, temperature, level or other signals to two or j

more control systens.

f (4) Provide justification that any simultaneous malfunctions of the control systems unidentified in (2) and (3) resulting l

from failures or malfunctions of the applicable common I

power source or sensor are bounded by the analyses in Chapter 15 and would not require action or response beyond the i

capability of operators or safety systems.

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ENCLOSURE 2 l

BACXGROUND INFORMATION 4

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UN: ten STILES siiNi ho :

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NgCr. EAR REGULATORY CO N155 ION Accession Nc.:

CTT!;E OF IHi~E~ TION ANO E'N CEC 3%NI 7910iOIII

'JA;H;N TON, D.C.

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November 30, 1979 1

l IE Eulletin No. 75-E7 L:55 CF NON-CLA55-1-E INSTRUMENTATION AND CONTROL POWER SYSTEM BUS

L'RINO OTEF.ATION 1

Ccscription of Circumstances:

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.n hoverber 10,157? an ever.: c:curr:d :t e Oc: ?e Fe.er itation, Unit I,

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nat resulted in loss of power to a ncn-class-1-E 120 Vac single phase power panel that supplied power to the Integrated Control System (ICS) and the N:n-Nuclear Instrunentation (NNI) System. This loss of power resulted in control system calfunctions and significant loss of inf;rmation to the

ntr:1 re:: : pert :r.

~:e:ifically, et 3:15 p.e., with Unit 3 at 100 per:ent po.er, the main conder. sate

s os tripped, a;oarently as a result of a technician performing maintenance on t'e Bet-ell level control system. This led to reduced feed-cter fles to the stea generators, which resultec in a reactor trip cue t

high coolant system At 3:17:15 p.m., the e essure and simultar.eous turbine trip at 3:16:57 p.m.

r.ca-class-1-E inverter po-er supply feeding all power to the integrated control system (which provides proper coordination of the reactor, steam generator feet.ater control, and turbine) and to one NNI chann'el tripped and failed to astomatically transfer its loaos from the DC power source to the regulated AC

er source. The inverter tripped due to blown fuses.

Loss of power to the

WI rendered control room indicators and recorders for the reactor coolant system (ex:ept for one wide-range RCS pressure recorder) and most of the secondary plant tjsta:s ineperaole, causing loss of indication for systems used for decay heat Upon loss removal anc water addition to the reactor vessel and steam gener'eters.

of pe-er, all valves controlled by the ICS assumed their respective failure The loss of power existed for approximately three minutes, until an.

ositions.

erator could reach the equipment room and manually swite.} the inverter to the regulated AC source.

73e above event was discussed in IE Information Notice No. 79-29, issued Nov e=cer 15, 1979.

huREG 0500 " Investigation into the March 28, 1979 TMI Accident" also discusses j

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TMI LER 78-021-03L whereby the RCS depressurized and Safety Injection occured on loss of a vital bus due to inverter failure.

A::iens to Be Taken by Licensees For all powtr reactor facilities with an coerating license and for those nearing 1

comoletion of construction (North Anna 2, Diablo Canyon, McGuire, Salem 2, Sequoyah, and Zimmer):

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he.er x 30. M.57i IE Eu'ht.in %. 75-27 Face 2 of 3 F.eview the class-1-E and non-class 1-E buses supplying po er to safety and 1.

non-safety related instrumentation and control systems which could affect the ability to achieve a cold shutdnen condition using existing procedm s For each bus:

or procedures deveiopd undar item 2 below.

identify and review the alare and/or indication provided in the' control l

a) re:m to alert the operator to the loss of power to the bus.

identify the instrument and control system loads connected to-the bus j

b) and evaluate the effect's of loss of power to these 1 cads inc1cding the ability to achieve a cold shutdsn condition.

describe any preposed de:ign modifications resulting from those ikvic.ts c) and evaluations, and your proposed schedule for icplementing those l

t modifications.

l Prepare emergency procedures or review existing ones that will be used by 1d 2.

c:ntr:1 re:m cperators, incTuding procedures rcquired to achieve a c:

shutd:.n condition, upon loss of pcmer to ez.ch class 1-E and non-class 1-E tus su;; plying power to safety ano non-safety related ins rument and i

control systems. The emergencf procedures should include:

t.e diagnostics / alarms / indicators /symptem resulting from the review a) 5 anc evaluation conducted per item 1 above.

j the use of alternate indication and/or control circuits which may be b) powered from other non-class 1-E or clas's 1-E instrumentation and j

control buses.

c) methods for restoring power to the bus.

Describe any proposed design modification or administrative controls to be icplemented resulting from these procedures, and your proposed schedule,for j

implementing the changes.

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Re-review IE Circular No. 79-02, Failure of 120 \\'olt Vital AC Power Suppiias, l

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cated January 11, 1979, to include both class 1-E and non-class 1-E safety Based on a review of operating experiunce related power supply inverters.

i and your re-review of IE Circular No. 79-02, describe any proposed design modifications or administrative controls to be implemented as a result of trie re ' review.

i Within 90 days of the date of this Bulletin, complete the review and 4.

evaluation required by this Bulletin and provide a written response l

describing your reviews and actions taken in response to each item.

f Repor*s should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and 20555.

j Enforce:nent, Division of Reactor Operations Inspection, Washington, D.C.

i If you desire additional informatiert regarding this matter, please contact the r

j IE Regional Office.

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9, :'979

'- e u-2:. huiieein he. 7s-27 Ta;e 2 of 3 f

A.; proved by GA0 5150225 (Ts0072); cleara..ce expires 7/31/B0. A;;reval w as given under a blanket clearance specifically for identified generi: pr:blems.

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E...k a; c IE Eu11etin No. 7E-27 Nv eLaer-30, 1579 RECENTLY ISSUED IE BULLETINS.

- Bulletin Subject Date Irsued Issued To No.

79-26 Boron Loss From BwR 11/20/79 All 9UR pewer rez: tor facilities with an Control 91ades OL 79-25 Failures of Westinghouse 11/2/79 All po-er raactor facilities with an BFD Relays In Safety-Related OL or CP Systems 79-17 Pipe Cracks In Stagnant 10/29/79 All P'n'R's with an CL and for ir.for:ation (Rev. 1)

Ecrated Water System At to other pc.'er rea: tors j

PVR Pinnts

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79-24 Frozen Lines 9/27/79 All power reactor j

facilities which have-either Ols or cps a.d are in the late stage l

of construction j

l 75-22 Potential Failure of 9/12/79 All Power Reactor Facilities with an Emergency Diesel Generater Field Operating License or a construction per=it Excite.- Transformer

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79-14 Seismic Analysep For 9/7/79 All Power Reactor -

Facilities with an (Sup-lement 2) As-Built Safety-Related OL or a CP Piping Systems 79-22 Possible Leakage of Tubes 9/5/79 To Each Licensee of Tritium Gas in Time-sibo Receives Tubes of Tritium Gas i

pieces for Luminosity Used in Timepieces for Lc=inosity.

79-13 Cracking in Feedwater 8/30/79 All Designated (Rev. 1)

System Piping Applicants for OLs 79-02 Pipe Support Base Plata 8/20/79 All power Reactor Facilities with an (Rev. 1)

Designs Using Concrete OL or a CP (Supplement 1) Expansion Anchor Bolts 79-14' Seismic Analyses For 8/15/79 All P:wer Reactor Facilities with (Supplement)

As-Built Safety-Related an OL or a CP Piping' Systems 9

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.va-ch 13,1 ;n IE Bulletin No. So-C5 1

CNGINEERED. 5AFETi FEATURE (EST) Rt0LT CON 1T.DL5 ji Description of Circumstances:

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Cn Neverber 7,1979, Virginia Eiectric and Power Ccrpany (VEPCO) reported that

_i follosing initiatien of Safety Injection (SI) at North Anna Power Station l

Unit 1, the use cf the II Occ:t,';shbu; tons alone resuitad in cert:in v:ntila-tion da:pers changing pcsitica frcm thair safety or emergency code to their

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Turther investigatica by VEPCO and the architect-engineer resulted j

normal node.-

l in discovery of circuitry which similarly affected components actuated by a' j

Contain:ent Depressuri:ation Actuation (CDA, activated on Hi-Hi Centainment P ras f ure). The ci'rcuits in questien are listed below:

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Probica Ccmponent/ System Outside/Inside Recirculation Spray Pump motors will not start after l

actuation if CDA Reset is depressed

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Pump !btors prior to starting timer running l

out (approx. 3 minutes) t

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Pressurized Control Room Dampers will open on SI Reset

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I Ventilation Isolation Dampers i

Safeguards Area Filter Dampers Dampers reposition to bypass filters when CDA Reset is depressed

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Containment Recirculation Cooler Fans will rest' art when CDA Reset o

is depressed l

Fans Service Water Supply and Discharge If service water is being used as l

the cooling medium prior to CDA Valves to Containment actuation, valves will reopen upon depressing CDA reset l

Service Water Radiation Monitoring Pumps will not start after actuation if CDA reset is d,epressed Sample Pumps prior to motor starting timers running out

'!:ain condcaser Air Efector Exhaust After receiving a high radiation

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Isolation Valves to the Containment monitor alarm on the air ejector i

exhaust, SI actuation would shut these valves and depressing SI Re>et wouid reopen them

i R.wice of :::c...ry :or ventiiation

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VJTTO o:.d *. di:ccvcry cf cimilcr '- i;ac in F SF-actuated co. poneo;s at 535 been found that caricin equipc.:.t 3dr.y Uait I a.r.d C cver Vc11cy; where i:

would re turn to its normal mode folicwing the reset of an ESF 3:;. 61;. t!..5, protective actiens of the affected systems could be compromised cnce the associated actuation signal is reset.

Thece two plants had Stone p.nd Webster Engineering Corporation for the architect-cngineer as did the Narth Anna Units.

I The Stene and Vebster Engineering Corporation and VEPCO are preparing design changes to preclude safety-related equipment from moving out of its emergency mode upon reset of an Engineered Safety Features Actuatica Signal (ESFAS).

been fcund acceptable by the NRC, in that, upe7 This correctise action he:

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reset of E5FAS, all af fect3d equip:. cat rcmcias in its erergency mcda.

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reviews of selected areas of ESFAS reset action on Ph'R The NRC has performed facilities and, in some cases, this review was limited to examination of logic, i

It has been determined that logic diagrams may not diagrars and prncedures.

adequately reflect' at-built conditions; thereforc, th.e requestad review of drasings nust te dene at the schematic /elem:ntary diagram level.

There have bcen several ccmmunicatio'ns to licensees from the NRC on ESF reset For example, scme of these ccmmunications have been in the form of actions.

Generic letters issued in November,1978 and October,1979 on contcinment Inspection and Enforcement venting and purging during normal operation.

Bulletins Nos. 79-05, 05A, 05B, 06A, 068 and 08 that addressed the events at THI-2 and NUREG-0578, THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations.

However, each of these communications has aodressed only a limited area of the ESF's.

We are requesting that the reviews undertaken for this Bulletin address all of the ESF's.

Actions To Be Taken By Licensees:

For all PWR and Skt facilities with operating licenses:

Review the drawings for all sys,tems serving safety-related functions at 1.

the schematic level to determine whether or not upon the reset of an ESF actuation signal, all associated safety-related equipment remains in its emer,gency mode.

Verify th'e actual installed instrumentation and controls at the facility are consistent with the schematics reviewed in Item 1 above by conducting 2.

a test to demonstrate that all equipment remains in its emergency mode upon removal of the actuating signal and/or manual resetting of the various isolating or actuation signals.

Provide a schedule for the performance of the testing in your response to this Bulletin.

If any safety-related equipment does not remain in its emergency mode upon 3.

reset of an ESF signal at your facility, describe proposed system rodification, design change, or other corrective action planned to resolve the prob 1cm.

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c0 days, the re c ui t s e t,. :.;r r.vi::.. a. J ir.:l' y

~.;rert in writing within s

3 ab~ve, acii :s E

a T i:.t of all deviccs ur.ich respund as discu:::d ir. ih l

t:1.:.n or plc.r..ed to a:!ure Ader:n-te equi.ccat ccr. trol, and a schedule for This information is requested under it.pi:.cntation of corrective action.Accordingly, you are requested to j

the provisicns of 10 CTC 50.54(f).

provide within the time period specified above, written stett:2nts of-R p rts shall!

the above infer: ticri, sign:d under oath or affire.ation.

j be submitted to the Director of the appropriata !.'RC Rc;ional Of fice and a copy shali be, f:.,. arded to the t RC Of fice.of Inspection and Enforcement, 20555.

i Division of Renctar Cperations Inspection, k'ashington, D.C.

this Suitetin is l

Fcr all' pc.er reactor facilities with a construction parmit, l

fr.r inferration cr.ly cnd no 1:ritten response is rcquired.

Appreval was l

Approved.by GAO, E:20225 (R0072); clearance expires 7-31 I

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3 ENOLC':UF.' 1 l

t Uu' ED STATES i

NUCLEAR kECUU.TDRY COMKISSION DFF::E C: I h'.* F T,* TION AM: EUFORCEMENT t'ASNIN;TCN, D.C.

20EE.;

l September 14, 1979 IE Inf ermati:n N:tice N:. 79-::

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' QUA.!?!!ATION CF !ONTROt if 57EM3 Public Service Electric and Gas Company notified the N?.C cf a p:tential unrevie.cf This notification was based on a safety cuestion at their Salem Unit I facility.

centinuin; revie. by Westingnouse of the environmental qualificatiens of equi;nent l

Based en the present st:tus that they su;;1y for nuett:r : tea-su; ply systems.

cf this effert, Westinghouse has infermed their customers that the performan:e of non-safety g*ade e;uipment sucjectec te an adverse envir:nner.: could impa:t i

tne prete:tive f unctions performe: by saf ety grace equipmint.

These non-safetf i

grace systems in:1uce:

Steam generater power operated relief valve control system i

• Pressuri er power operated re, lief valve control syste Main fee sater control system Automatic roc control system These systems could potentially malfunction due to a high energy line break inside or outside of containment.

NRC is also concerned that the adverse environment c:uld also give erroneous information to the plant operators.

Westingneuse states

• hat the consequences of such an event could possibly be more limiting than results presented in Safety Analysis Reports, however, Vestinghouse also states that the severity of the results can be limited by operator actions together with operating characterisitics of the safety Fur,ther, Westinghouse has recommended to their customers that they systems.

review tneir systems to catermine whetner any unreviewed safety questions exist.

l This Informatien N tice is provided as an early notification of a possibly It is expected that recipients will review the information No specific action or resp:nse significant matter.

for pessibie applicability to their facilities. evaluations so indica *e,further licensee

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is repuested at

• his time.

If NR a:tions may be requested or recuired.

If you have questiens regarding this matter, j

please contact the Director of the appropriate NRC Regional Office, written response to this Inf ormation Nctice is required.

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Er-2776 Pittsourgh, Pennsylvania 1Ei30 l

August 30, 1979 l

PSE-79-21 Mr. F. P. Libri::i, General Manager Eic tric Pro:v:ti:n Public Service Ele::ric and Gas Ccmpany 80 Park Place Newark, New Jersey 07101 l

Dear Mr. Libri:

i:

Public Service Electric and Gas Co.

Salem Unit Nc.1 00ALIFICATION OF CONTROL SYSTEMS l

f As part of a continuing review of the environmental qualifications of Westinghouse supplied RSSS equipment, Westinghouse has also found it l

necessary to consider the interaction with non-safety grade systems.

This investigation has been conducted to determine if the performance 3

of non-safety grade systems which may not be protected from an adverse environment could impset the protective functions performed by RSSS r

The NSSS control and protection systems were safety grace equipment.

included in this review to assess the adequacy of the present environ-i i

mental qualification requirements.

As a result of this review, several systems were identified which, if i

subjected to an adverse environment, could potentially lead to control system operation which may impact protective functions. These systems are:

l Steam generator power operated relief valve control system I

Pressuri:er power operated relief valve control system Main feedwater control system i

1 Automatic rod control system e

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t E:ch ;# the !bove ren;ioned sys: cms could c::entially malfur.ction if l i r. ~ b r "..'. i n~ i d:

ieca::ec ey acverse er.vironm'.nts cue te i t.igh Enc EJ or outsice c:ntainment.

In ea:h case, a lic1 cc set of creaks, couple' i

with possible consecuential control malfunction in an adverse dire:: ion, i

of the above events Aculd yield results which are re limiting than those In all cases, however, the preseated in the plant Safety Analysis Reports.

severity of the results can be limited by operater actions toge:ner with

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p:r::1:; ch:ra:::rt::ic; of the sately systems.

We believe these systems identified do not constitute a substantial safety However, Westinghouse recgunends you review them to determine if hazard.

any unreviewed safety questions or significant deficiencies exist in your 4

pl an:( s).

To assist you in understanding these concerns, Westinghouse will hold a i

seminar in Pittsburgn on Thursday, Septemoer 6 at Westinghouse R10 Center, The seminar will Suilding 701, with all cur operating plant customers.

eccress the poter. ial impact of these concerns for various plan designs and various licensing bases.

Please contact your WNSD Regional Service office to confirm your attancance at the seminar. We IK11 provide additional details concerning the agenda and other meeting arrangements as they become available.

Very truly yours, ORIGINAL SIGNED BY F. Noon, Manager Eastern Regional & WNI Support SR4/CC13&l4 cc:

H. J. Midura H. J. Heller R. D. Rippe T. N. Taylor R. A. Uderitz C. F. Barclay W p_

l EN!LOSURE 3 j

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RErR!NI T'.!::.!; CE~J/IC.. L::...

1.2 61.5 C.

Salem Nuclecr Gencrctir.g 5:ction P. O. Box 56 Hancocks Bridge, New Jersey 0803E

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P S ept embe* 10, 1979 i

Pr. Boyce H. Grier*

l Director of USNRC O ice ef Inssertien and Enfcrcement Rc-ten !

631 Park Avenue King of Prussia, Pennsylvania 19406 i

Dear Sir:

REF0ETAELE,0CCURRENCE 79-53/01F SALEM NO 1 UNIT LER o

This lecter will serve to confirm our telephone re; ort tc Mr. Gary Senneicer of the Regional NRC office on Friday, Septemoer 6,1979, advising of a potential reportable occurrence in accordance with Technical Specification 6.g.1.8.

We have been notified by our Engineering Department that a Westing-house conducted review of the environmental qualifications of Westinghouse supplied K555 ecuipment has identified that conditions associated with high energy line breaks inside or outside containment anc their impact on non-safety control systems may constitute an unreviewed safety question. The control systems concerned are steam generator power operated relief valve control, pressurizer power operated relief valve control, main feedwater centrol and automatic red control systems.

A detailed report will be submitted in the time period specified by the Technical Specifications.

Very truly yours, 0}-

Original Signed By H. J. Midura 7[j t(

A Manager - Salem Generating Station A'4::j ds CC: General Manager - Electric Production Manager - Quality Assurance

EN;LOSURE 2

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MEET!fy HIGHLIGHTS l

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i II.E Infon.4tiv.. i..,.iu 7511, Gud SLptc. cr l', lg7E, t.: 3 issur.t it.f: ming the nuclear indu:try of a potential unreviewed safety question at Sale., Unit 1 f

I of Public Service Eledric and Gas Ccmpany, based on a Westinghouse review of the environ +ntal cuciific: tion of ccui;:m:.t.

C:rt:in nen.::f:ty e :d: ceuf;:m:nt, if subje:ted to an adverse environment such as results from a high-energy line

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I break inside or outside of containment, ccJld impact the safety analyses and the prota:tive fun:tions perfomed by stftty gradc c:uipm:nt.

l f

Meetings were arranged with all fcur light water reactor vencers ac::rding to the 1

following schedule:

Westinghouse - Tuesday, September 1S Combustion Engineering - Wednesday, September lg Babcock and Wilcox - Thursday a.m., September 20 General Electric - Thursday, p m.,

September 20 During the Westinghouse meeting, they identified, for all high. energy line breaks.and possible locations, the control systems that could be affected as a result of the adverse environment and whose consequential failure could invalida1;e the

-l accident analyses presented in Westinghouse plants' SARs. Recomendations were also presented for resolving the adverse interactions identified.

Westinghouse's investigation identified seven e.ccidents and seven control systems that could possibly interact and presented them in a matrix fem as shown in Enclosure l'.

As can be seen the potential interactions that could degrade the accident analyses are in the:

a.

Auto.atic Rod Control System

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Pressurizer P02.V Ceatrol - Syrte-I c.

hcin reecw:ter L:ntroi systLa j

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Steam Generator PORY Centrol System Westinghouse presented their recomiended short-tem and long-tem solutions, l

presented as Enclesure 2.

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Westinghouse stated that the possible matrix interactions may increase as more j

detailed analyses are perfomed but the interactions will remain for all of their t

plants and the interactions may be eliminatec only if concitiens are suen tnat plan

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specific designs mitigate the interactions because of:

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system layout t

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type of equipment used t

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qualification status of equipment utilized r

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design basis events considered for license applications

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prior comitments made by utility to the NRC.

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1 Westinghouse stated that their investigations were carried further than FSAR analysis and they would need to evaluate consequential failurey on a realistic basis; this

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evaluation may eliminate some problems. Westinghouse also stated that their 3

investigations are lower probability subsets of FSAR analyses whicy in themselves i

are sets of low probability.

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e W:: tin;h:u:e end the utility rcpresente.tivcs cil dcutt that the can conclusive 1,*

oetemine the qualification status of all of the involved equipment in 20 days.

l Both Westinghouse and the utility representatives stated that they will respond to the 20-day letter by addressing the four control syste.?.s identified in a manner suggested by the Westinghouse recommendations unless the NRC staff provides

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dire:ti:ns to the c:ntI'ary and further establishes guidelines stating their

' p:.siti:r. c. tha pr::Lic.. blong with their reco=.endaticac.

The NF.C staff stated that they are sympathetic to the requests by the nuclear industry regarding position and direction but this can be fomulated only at the conclusion of the scheduled meetings with all four light water reactor vendors. At that l

time the staff will present their results, magnitude and dire: tion to industry for resolution of the problem.

l At this time, it is not evident which utilities are faced with what environmental interacticn problem. The effects of implementing all of the Westinghouse reco=:. ended short-term " fixes" may be contradicted by other sequences. There are three parts to the problem dealing with the basis of short-t rm operation:

1.

qualify equipment to the appropriate environment; this would take longer

. than 20 days and would, more likely, for most utilities, be a long-tem j

i partial solution.

2.

short-term " fixes" should be in place pending long-term solutions such as the above.

It must be noted that in this situation, some components that are relied upon to operate might possibly be wiped out by consequential failures under certain conditions and accident sequences if the postulated 2

adverse environment is estab13shed.

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the " worst case" plant should be selected and a bounding analysis performed to detemine the time frame available for qualification of equipnent.

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