ML20126J723

Requests Addl Info Re Four Potential Problem Areas in Design & Analysis of Instrumentation & Control Sys.Fsar Should Be Amended to Reflect Response to Encl Questions & Any Other Outstanding Items by 810801

ML20126J723
Person / Time
Site:	Byron, Braidwood
Issue date:	04/30/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Abel J COMMONWEALTH EDISON CO.
References
NUDOCS 8105050009
Download: ML20126J723 (13)

Text

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UNITED STATES

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y,( .g de E NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ,

APR 3 01981 ,

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' Docket Nos.: 50-454/455 ,

.and' 50-456/457

-l Commonwealth Edison Company - f  %

'% 04 yggg~ j ATTN: Mr. J. S. Abel **uE D S*

Director lof Nuclear Licensing  % '

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P. O. Box 767 Chicago, Illinois 60690 ' ,b g Q

Subject:

Request' for.Information Regarding Four.. Potential Prob 1cm Areas in the Design and Analysis of the Instrumentation and Control Systems at Byron /Braidwood '

h The staff ha's identified four potential areas in the instrumentation and contro1' systems' that must be addressed in your OL license application. The

'first two are related to events that have occurred at operating plants, the- .

third,- generic questions regarding environmental qualification of Westinghouse .

NSSS equipment, and the fourth, seeks to ensure that all credible design basis

. events which can lead.to failure of power or control systems are considered,in [

Chapter 15.of the FSAR. These four items of concern are listed as questions l 30.040 through 30.043 in the Enclosure. l t

S We request that you amend your FSAR to reflect your response to these questions and any other outstanding items by August 1,1981. f If you need further clarification on these items, contact the Licensing Project Manager for Byron, Larry White (301/492-8285) or the Licensing Project Manager for Braidwood, Kenneth Kiper (301/492-7318).

Sincerely ,/ l

, i Rob rt . Tede co, Assista t Director I or icensing l Division of Licensing 4

Enclosure:

.As stated-cc: See next page i

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i 8105050004- '

A THIS DOCUMENT CONTAINS I

I P00R QUALITY PAGES I

> Mr. J . S.. Abel Director'of Nuclear Licensing' l l

Commonwealth Edison Company Post Office. Box 767 .' ,

q Chicago, Illinois. 60590 l CCs . Mr. Edward'R. Crass  !

Mr. William Kortier . Nuclear Safeguards and Licensing Divisior Atomic Power Distribution - .

Sargent & Lundy Engineers Westinghouse Electric. Corporation

• 55 East Monroe Street .

P. O. Box 355 Chicago, Illinois 60603 l; Pittsburgh,; Pennsylvania 15230 .,

Nuclear Regulatory Commission, Region 111 j Paul M.. Murphy, Esq. .

Office of Inspection and Enforcement Isham, Lincoln & Beale 799 Roosevelt Road One First National Plaza Glen Ellyn, Illinois 60137 l 42nd Floor t Chicago, Illinois 605.03 -

Myron Cherry, Esq. ~

Mrs. Phillip'B. Johnson Cherry, Flynn and Kanter ,

1 IEM Plaza, Suite 4501 i 1907 Stratford Lane Chicago, Illinois 60611 r- Rockford, Illinois 61107 s e >

8 Ms. Julianne Mahler Center for Governmental Studies  !

Northern Illinois University DeKalb, Illinois 60115  :

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C. Allen Bock, Esq.

• P. O.-Box 342 Urbanan, Illinois 51820

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t Thomas J. Gordon, Esq.

[' Waaler, Evans & Gordon

^L 2503 S. fleil  !

! Champaign, Illinois 61820 ,

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' Ms. Bridget Little Rorem l o Appleseed Coordinator ,

117 North Linden Street "j-Essex, Illinois 60935 -

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.t Kenneth F. Levin Esq.

.i Beatty, Levin,. Holland, .,

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Basofin & Sartany  ;

3 11 South LaSalle Street p, ' l n Suite 2200 3 Chicago, Illinois 60603 I

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ENCLOSURE BYRON /BR'XIDWOOD STATIONS Instrumentation & Control Systems Branch

-30.040 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 operator actions should be based. This concern was addressed in IE Bulletin 79-27 (Attachment 1). On November 30, 1979, IE Bulletin 79-27 was sent to operating license (0L) holders, the near term OL applicants (North Anna 2 Diablo Canyon, McGui.re, Salem 2, Sequoyah, and Zimmer), and other holders of construction permits (CP), including Byron /Braidwoo' d . 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 this issue 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. This report should be in the form of an amendment to your FSAR. .

30.041 Engineered Safety Features (ESF) Reset Controls (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 (Attachment 2). 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 description

of corrective actions taken or planned as stated in Action 1 thru 3 and submit the report called for in Action Item 4. The report should be submitted to the NRC Office of Nuclear Reactor Regulation as a licensing submittal in the form of an FSAR amendment.

30.042 Qualification of Control Systems (IE Information Notice 79-22)

Operating reactor licensees were informed by. IE Jnformation Notice ,

79-22, issued September 19, 1979, that certain non-safety grade or 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. Attachment 3 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 l . letter which address this matter.. Operating reactor licensees I conducted reviews to determine whether such problems could exist ,

l at operating ' facilities.

We are concerned that a similar potential may exist at light water facilities now under construction. You are, therefore, requested to perform a review to determine what, if any, design changes or '

operatcr actions would be necessary to assure that high energy line breaks will not cause control system failures to complicate the i event beyond your FSAR analysis. Provide the results of your review including all identified problems and the manner in which 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 :nteractions which might occur. Your review should include those scenarios, where applicable, but should not necessarily be limited to them.

You should consider analogous interactions relevant to the designs ,

of the Byron and Braidwood plants.

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

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 fundamental credible failures you are requested to provide the following information:

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- (1) .

!dentify lthose control systems whose failure or malfunction - -

could. seriously impact plant . safety.  ;

.i L(2) - Indicate which, if any, of;the control systems identified in .

1l (1) receive power from common power sources. . The power sources i considered should include. all. power sources whose failure I or malfunction could lead to failure or malfuction of.more  ;

than one control system and should extend to 'the effects of  !

- cascading power lossec iue to the failure of higher. level  !

distribution panels and load centers. . j i

(3) Indicate' which if any, of the control systems identified  :

in '(1) receive input signals from. common sensors. The sensors .!

considered shouldL include, but should not necessarily be j l' - limited to, common hydraulic' headers or impulse lines feeding.  !

-pressu:'e,~ temperature, level or other ' signals to two or l jE. more control systems.  :

i .!

(4) Provide justification that any simultaneous malfunctions of j the control systems indentified in (2) and (3) resulting-  !

- from failures or malfunctions of the applicable common  !

. power source'er sensor are bounded by the analyses in Chapter l 15 and would not require action or response beyond the i capability of operators. or . safety . systems.

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ATTACHMENT l' s

. UNITED STATES SSINS No.: 6820 NUCLEAR REGULATORY COMMISSION Accession No.:

OFFICE OF INSPECTION AND ENFORCEMENT 7910250499 WASHINGTON, D.C. 20555 ,

November {0,,1979

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IE Bulletin No. 79-27 LOSS OF NON-CLA55-1-E. INSTRUMENTATION AND CONTROL POWER SYSTEM BUS .

CURING OPERATION Description of Circumst.ances: i

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On Novecber 10, 1979, ah event ~ occurred at the Oconee Power Station, Unit 3, -

tnat resulted in loss of power to a non-class-1-E 120 Vac single phase power panel that supplied powcr to the Integrated Control System (ICS) and the ,

Non-Nuclear Instrumentation (NNI) System. This loss of power resulted .

j in control system calfunctions and significant loss of information to the  ;

control roca operator.  ;

' 'cifically, at 3:16 p.m. , with Unit 3 at 100 percent power, the main condensate '

os tripped, apoarently as a result of a technician performing maintenance on hotwell level control system. This led to reduced feedwater flow to the  ;

.ac generators, which resulted in a reactor trip due toAt high coolant system 3:17:15 p.m. , the essure and simultaneous turbine trip at 3:16:57 p.m'.

t.on-class-1-E inverter power supply feeding all power to the integrated control ,

system (which pr.ovides proper coordination of the reactor, steam generator, fee:Nater control, and turbine) and to one NNI channel ~ tripped and' failed to  ;

automatically transfer its loads from the DC power source to the regulated AC  !

p:wer source. The inverter tripped due to blown fuses. Loss of power to the MI rendered control room indicators and recorders for the reactor coolant system (except for one wide-range RCS pressure recorder) and rnost of the secondary plant j s/stens ineperable, causing loss, of indication for systems used for decay heat removal and water addition to the reactor vessel and steam generators. Upon loss l

of pcver, all valves controlled by the ICS assumed their respective failure positions. Tne loss of power existed for approximately three minutes, until an .

cperator could reach the equipment room and manually switch the inverter to the regulated AC source. . .,

The above event was discussed in IE Information Notice No. 79-29, issued f

Noven:ber 16,1979. ,

i hUREG 0600 " Investigation into the March 28, 1979 TMI Accident" also discusses  ;

l TMI LER 78-021-03L whereby the RCS depressurized and Safety Injection occured  ; )

on loss of a vital bus due to inverter failure.

• A:tions to Be Taken by Licensees For a.11 power reactor f acilities with an operating license and for those nearing l 1

completion of construction (North Anna 2, Diablo Canyon, McGuire, 2alem 2, Sequoyah, and Zic:mer):

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. November 30,1979 ,

' lIE' Bu11'et n TNo. 79-27 Page '2 of 3 '

Review the_ class-1-E. and non-class :1-E buses supplying. power. to : safety and

, 1.-  ?

non-safety'relatedlinstrumentation and control systems which could affect  :

the ability to achieve a cold shutdmin condition using existing procedures - -

or procedures developed under item 2 below. :For each bus: .

l a) identify and review the alarm and/or indication provided in the control j room to alert the. operator:to the loss of power to the bus.. 1

.b) identify .the : instrument and control system loads connected to-the bus ,j and evaluate.the effects of loss.of power to these loads including ,

the' ability to achieve a cold shutdown condition. s describe any proposed design modifications resulting from these t evie is [

c) 4 and evaluations,: and y'our proposed schedule for implementing those 3 modifications.

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2. Prepare. emergency procedures or review existing oiles that will be used by ,

control room-operators, including procedures required To achieve a cold :I shutdown condition, upon loss of power to each class.1-E and non-class j 1-E bus supplying power to safety and non-safety related instrument and control systems. The emergencf procedures should include: i a) the diagnostics /alarins/ indicators / symptom resulting from the review -

- and evaluation conducted per item 1 above. .

b)- the use of arternate indication and/or control circuits which may be '

powered from other non-class 1-E or class 1-E instrumentation and ~

. control buses. - i  !

- c') methods.for restoring power to the bus. ,

' Describe any proposed design modification or administrative controls to be implemented resulting from these procedures, and your proposed schedule for -i implementing ihe changes. l

3. Re review IE Circular No. 79-02, Failure of 120 \'olt V, ital AC Power Supplias, ,

dated January 11, 1979, to include both class 1-E and nop-class 1-E safety related power supply inverters. Based on a review of operating expericce r and your re-review of IE Circular No 79-02, describe any proposed design 4 modifications or administrative controls to be implemented as a rcsult of '

. the re ' review. ,

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

evaluation required by this Bulletin and provide'a written response -l

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describing your reviews and actions taken'in response to each item. r it:

Reports should be submitted to the Director of the appropriate NRC Regional- .j a .

Office and a copy should be , forwarded to the NRC Office of Inspection and 20555.

l

' Enforcement, Division of- Reactor Operations Inspection, Washington, D.C. >

If you desire additional information regarding tiiis matter, please contact the l IE Regional Office. l

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P Enclosure IE Bulletin No. 79-27 i

.Novebmer 30, 1979  !

RECENTLY. ISSUED '

IE BULLETINS ,

l Issued To Bulletin Subject Date Issued ,

i No.

Boron loss From BWR 11/20/79 All SWR power reactor  :

79-26 Control Blades facilities with an l 1

OL 79-25 Failures of Westinghouse

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11/2/79 All power reactor J BFD Relays In Safety-Related facilities with an Systems OL or CP i 79-17 Pipe Cracks In Stagnant 10/29/79 All PWR's with an ,

Borated Wa er System At OL and for information  ;

'(Rev. 1) to other power reactors  !

PWR Plants 79-24 Frozen Lines 9/27/79 All power reactor- 4

- facilities which have .

l' either Ols or C?s and are in the late stage of construction

.n 79-21 Potential Failure of 9/12/79 ' All Power Reactor Emergency Diesel Facilities with an ,

Generator Field Operating License or  !

Exciter Transformer a construction permit 75-14 Seismic Analyses For 9/7/79 All Power Reactor (Supplement 2) As-Built- Safety-Related . Facilities with an  ;

Piping Systems OL or a CP ,

79-22 Possible Leakage of Tubes 9/5/79 ,

To Each Licensee  !

i I of Tritium Gas in Time- s,to Receives Tubes pieces for Luminosity of Tritium Gas

. Used in Timepieces for Luminosity Cracking in Feedwater 8/30/79 All Designated 79-13 Applicants for OLs (Rev. 1) System Piping Pipe Support Base Plate 8/20/79 All power Reactor 79-02 Facilities with an (Rev. 1) Designs Using Concrete (Supplement 1) Expansion Anchor Bolts OL or a CP Seismic Analyses For B/15/79 All Po er Reactor 79-14' Facilities with

.(Sepplement) As-Built Safety-Related Piping Systems an OL or a CP e

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ATTACHt1ENT 2 . cct:icn f.'o.:

4: .-1;LO559 -

viiic0 diATES N JC L'_'a ..,:&./o CT.Y C' '::: 2 : 0N OFFICE OT ::..UT.Ci!ON /.ND ENFCRCEmHT 1..&'.1N3 TON, D.C. 20555 March 13,1980 -

IE Bulletin .No. 80-06 ENGINEERED SAFETY FEATURE (ESF) RESET CONTROLS Description of Circumstances: ,

On Novecher 7,1979, Virginia Electric and Power' Company (VEPCO) reported that following initiation of Safety Injection (SI) at North Anna Power Station Unit 1, the use of the SI Reset pushbuttons alone resulted in certain ventila- i tion dacpers ch'anging positidn from their safety or emergency mode to their normal node. Further investigatica by VEPCO and the architect-engineer resulted in discovery of circuitry which similarly affected components actuated by a ,'

Contain:ent Depressurization Actuation (COA, activated on Hi-Hi Containment P ressure). The ci'rcuits in question are listed below:

Component / System Problem ,

! Outside/Inside Recirculatio.n Spray Pump motors will not start after actuation if CDA Reset is depressed l Puap Motors

- - prior to starting timer running I

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- out (approx. 3 minutes)

Pressurized' Control Robm ,

Dampers vil1 open on.SI Reset Ventilation Isolation Dampers i,

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

)

Containment Recirculation Cooler Fans will rest' art when CDA R'eset Fans is depressed .

Service Water Supply and Discharge If service water is being 'used as Valves to Containment the cooling medium prior to CDA actuation, valves will reopen .,-

upon depressing CDA reset Service Water Radiation Monitoring Pumps will not start after actuation if CDA reset is depressed Sacple Pu..:ps prior to motor starting timers running out Main Condenser Air Ejector Exhaust After receiving a high radiation monitor alarm on the air ejector Isolation Valves to the Containment exhaust, SI actuation would shut thess valves and depressing SI Reset would reopen them I

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2 ::f 3 Review of ciredtry for ventilation dr.;srs, cotors, and valves r sported by VEPCO resulted in discovery of similar designs in ESF-actuated components at i I -

Surry Unit 1 and Beaver Valley; where it has been found that certain equipr. ant  !

would return to its normal mode following the reset of an ESF signal; . thus, '

protective actions of the affected systems could be compromised once the ,

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associated actuation signal is reset. These two plants had Stone and Webster -

Engineering Corporation for the architect-engineer as did the North Anna Units. i The Stone and Vebster Engineering Corporation and VEPCO are preparing design changes to preclude safety-related equipment from moving out of its emergtacy mode upon reset of an Engineered Safety Features Actuation Signal (ESFAS).

This corrective action has been found acceptable by the.NRC, in that, upoy reset of ESFAS, all affected equipment remains in its emergency co'fe.

The NRC has performed reviews of selected areas of ESFAS reset action on PWR

' facilities and, in some cases, this review was limited to examination of logic.

diagrams and procedures. It has been determined that logic diagrams may not , ,

l adequately reflect' as-built conditions; therefore, the requested review of drawings must be done at the schematic / elementary diagram level.

I There have been several communicatidns to licensees from the NRC on ESF reset actions. For example, some of these cccmunications have been in the form of  :

' Generic Letters issued in November,1978 and October,1979 on containment ,

venting and purging during normal operation. Inspection and Enforcement i

i Bulletins Nos. 79.05, 05A, 05B, 06A, 06B and 08 that addressed the events at  !,

TMI-2 and NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and'  ;

Short-Term Recommendations. However, each of these communications has addressed 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 BWR facilities with operating licenses: ,

1. Review the drawings for all sys,tems serving safety-related functions at the schematic level to determine whether or not upon the reset of an ESF t actuation signal, all associated safety-related equipment remains in its ,.- l coergency mode. .

2. Verify the actual installed instrumentation and controls at the facility are consistent with the schematics reviewed in Item I above by conducting a test to demonstrate that all equipment remains in its emergency mode upon removal of the actuating signal and/or manual resetting of the i various isolating or actuation signals. Provide a schedule for the performance of the testing in your response to this Bulletin.

3. If any safety related equipment does not remain in its emergency mode upon reset of an ESF signal at your facility, describe proposed system rodification, design change, or other corrective action planned to resolve the problem. .

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'erch 13, 6 0

?*'.en N >. ::-05 *

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4. hport in writ'ing within C0' days, the results of your review and incluG a list of all devices which respond.as discussed in item 3 above, acti ns j l

. t?. ken or planned to assure adequate equipment control, and a schedule for ir.plementation'of corrective action. This information is iequested under the ~ provisions 'of.10 CFR 50.54(f). . Accordingly, you are requested to i provide within the time period specified above, written statements of -

the above information, signed under oath or affirr.ation. Raports shall be submitted to the Director of the appropriate NRC Regional Office and  ;

.a copy shall be forwarded to the NRC Office.of Inspection and 20555. Enforcenent, [

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Division of Reactor Operations Inspection, Washington, D.C.  ;

l For- all poser reactor facilities with a construction permit, this Bulletin is  !

f Or inforration only and no written response is required. $ i t'

Approval was .

Approved by GAO, B180225 (R0072); clearance expires 7-31-80.

given under a blanke.t clearance specifically for identificd generic problems. l, l.

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- ATTACHMENT 3 .i

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i UF ED STATES . l

. NUCLEAR kEGULATORY COMISSION -

OFFICE OF INSPECTION'AND ENFORCEMENT WASHINGTON, D.C. ;20555 l

September 14, 1979

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-t IE Information Notice No. 79-22 ,

3

QUALIFICATION OF CONTROL SYSTEMS f

Public Service Electric and Gas Company notified the NRC of a potential unreviewed I

safety question at their Salem Unit I facility. This notification was based on a l continuing review by Westinghouse of the environmental Basedqualifications on the present of equiprient statuF i that they supply for nuclear steam supply systems.  !

of this effort, Westinghouse has informed their customers that the performance j of non-safety grade equipment subjected to an adverse environment could impact  :

the protective functions performed by safety grade equipment. These non-safet/ 7 grade. systems include. ,

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Steam generater power operated relief. valve control system .

" Pressurizer

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power operated relief valve control. system

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• i Main feeesater control rystem Automatic red control system j

I 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 could also give erroneous information to the plant operators. t Wastinghouse states that the consequences of such an event could possibly be l more limiting than results presented in Safety Analysis Reports, however, -  :

Westinghouse also states that the severity of the results can be limited i by operator actions together with operating characterisitics of the safety I systems. Fur,ther, Westinghouse has recommended to their customers that they  ;

review their systems to cetermine whether any unreviewed safety questions exist. '

This Information Notice is provided as an early notification of a possibly  ;

significant matter. It is expected that recipients will review the information No specific action or response  :

, for possible applicability is requested at this time. to their ficilities.If NRC evaluations so indicate, further lice f a:tions may_be requested or required. If you have questiens regarding this matter, l please contact the Director of the appropriate NRC Regionel Office.  :

No written response to this Information Nctice is required.

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i REPRINT Westinghouse Electric Corporation- -

Water Reactor Division

• Nuclear Service Division >

- Box 2728' -

Pittsburgh, Pennsylvania 15230

• August 30, 1979

. PSE-79-21 '

Mr. F. P. Librizzi, General Manager  ;

Electric Production  ;

Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101

Dear Mr. Librizzi:

Public Service Electric and Gas Co. ,

f Salem Unit No. I OUALIFICATION OF CONTROL SYSTEFG As part of a continuing review of the environmental qualifications of Westinghouse supplied NSSS equipment, Westinghouse has also found it necessary to consider the' interaction with non-safety grade systems.  !

This investigation has been conducted to determine if the performance l

of non-safety grade systems which may not be protected from an adverse ,

environment could impact the protective functions performed by NSSS safety grade equipment. The NSSS control and protection systems were [

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

mental qualification requirements.

As.a.. result of this review, several systems were identified whic.h, if i subjected to an adverse environment, could'potdntially lead to control  !

system operation which may impact protective functions. These systems [

are: - l

- Steam generator power operated relief valve control system l Pressurizer power operated relief valve control system

- Main feedwater control system {

- Automatic rod control system 7 I

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Page 2 PSE-79-21

't Each of the above mentioned systems could potentially malfunction if , ,

impacted by adverse environments due to a high energy line break inside or outside containment. In each case, a limited set of breaks, coupled -

with possible consequential control malfunction in an adverse direction, '

of the above events pould yield results which are more limiting than those presented in the plant Safety Analysis Reports. In all cases, however, the r, severity of the results can be limited by operator actions together with .

operating characteristics of the safety systems. '

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

any unreviewed safety questions or significant deficiencies exist in your plant (s).  !

To assist you in understanding these concerns, Westinghouse will hold a l seminar in Pittsburgh on Thursday, September 6 at Westinghouse R&O Center, ,

Building 701, with all our operating plant customers. The seminar will  ;

address the potential impact of these concerns for various plant designs l

and various licensing bases.

Please contact your WNSD Regional Service office to confirm your attendance at the seminar. We will provide additional details concerning the'ag~enda and other meeting arrangements as they become available. l i

Very truly yours,  ;

ORIGINAL SIGNED BY ,

F. Noon, Manager -

Eastern Regional & WNI Support j

,SR4/CCl3&l4 cc: H. J. Midura H. J. Heller .

R. D. Rippe T. N. Tayl or R. A. Uderitz ,.

C. F. Barclay W i

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REPRINT c PUBLIC SERVICE ELEC.TRIC AND GAS COMPANY Salem Nuclear Generating Station ,

P. O. Box 56 ',

. . . . . Hancocks Bridge, New Jersey 08038- .

September 10, 1979  !

Mr. Boyce H. Grier* '

. Director of USNRC Office of Inspection and Enforcement  !

Region I -

631 Park Avenue ,

King of Prussia, Pennsylvania 19406  ;

Dear Sir:

REPORTABLE OCCURRENCE 79-58/OlP SALEM NO. 1 UNIT LER l, This letter will serve to confirm our telephone report to Mr. Gary  ;

l. Schneider of the Regional NRC office on Friday, September 6,1979,  :

advising of a potential reportable occurrence in accordance with ,

Technical Specification 6.9.1.8. ,

We have been' notified by our Engineering Department that a Westing-  !

house conducted review of the environmental qualifications of .

Westinghouse supplied NSSS equipment has identified that conditions associated with high energy line breaks inside or outside containment -

and their impact on non-safety control systems may constitute an unreviewed safety question. The control systems concerned are steam  !

i generato.r power operated relief valve control, pressurizer power

• operated relief valve control, msin feedwater control and automatic rod control- systems. -

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

- Very truly yours, Original Signed By l H. J. Midura Manager - Salem Generating Station l AWK:jds  !

CC: General Manager - Electric Production Manager - Quality Assurance l

l

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