ML20126K134

Forwards Instrumentation & Control Sys Branch Round 2 Requests for Addl Info to Complete Review of OL Application. Util Must Respond to Concerns Re IE Bulletins & Identify Control Sys Posing Threat to Facility Safety

ML20126K134
Person / Time
Site:	Waterford
Issue date:	04/29/1981
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Aswell D LOUISIANA POWER & LIGHT CO.
References
NUDOCS 8105110515
Download: ML20126K134 (21)

Text

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+  %, UNITED STATES y .g NUCLEAR REGULATORY COMMISSION g ;E WASHINGTON, D. C. 20655 -1

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APR 2 9 1981 1

Docket No.: 50-382 ,

1 Mr. D. L. Aswell Vice President, Power Production Louisiana Power & Light Company 1 142 Delaronde Street New Orleans, Louisiana 70174 1

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Dear Mr. Aswell:

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SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - WATERFORD 3 We have determined that certa.in additional information is required in order to permit us to complete our review of your application for an operating license for Waterford Steam Electric Station, Unit 3. The enclosed round two requests for additional information were prepared by the Instrumentatio.n

, and.. Control Branch.and are numbered 030,44 threegb 030:47w Please advise us of the date you expect to provide responses to the enclosed reques t. If you require any clarification, please contact the staff's assigned project manager. -

Sincerely, 8b'.t. A M Robert L. Tedesco, Assistant Director for Licensing Division of Licensing

Enclosure:

Request for Additional Information ,

cc w/ encl: See next page.

. 1 18105310516 A ]

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4 Mr . . D. L. Aswell Vice President, Power Production Louisiana Power & Light Company ,

142 Delaronde Street l' N{w Orleans, Louisiana 70174 .

cc:. W. Malcolm Stevenson, Esq.

Monroe & Lemann 1424 Whitney Building  ;

New Orleans, Louisiana 70130 1 Mr. E. Blake  ;

Shaw, Pittman, Potts and Trowbridge  !

I 1800 M Street, N. W.

Washington, D. C. 20036 Mr. D. B. Lester .

Production Engineer  :

Louisiana Power & Light Company 142 Delaronde Street New Orleans, Louisiana 70174 -

Lyman L. Jones, Jr. , Esq.

Gillespie & Jones "'" " " '

P. O. Box 9216 ' - " -

Metairie,.Lquigiana 70005 Luke Fontana, Esq.

Gillespie & Jones 824 Esplanade Avenue New Orleans, Louisiana 70116 Stephen M. Irving, Esq.

One American Place, Suite 1601 Baton Rouge, Louisiana 70825 Resident Inspector /Waterford NPS P. O. Box 822 K111ona, Louisiana 70066 e

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I 030f44 ALL OPERATING LICENSE APPLICANTS d

. 'SUBJECTI CONCERNS RELATED TO I&E BULLETIN 79-27 Operating reactors and some near term OL applicants have previously P received I&E Bulletin 79-27 which is enclosed. The concerns If you havewhich not prompted the Bulletin apply to all OL applicants. -

already responded to the concerns of BulletinFirst, 79-27,theyou arefor time now re-re- i quested to do so, but with two exceptions. ~

sponse will be determined on a case by case basis so that the 90 day limit in Item 4 is not applicable. Secondly, your reply should be made in the same way as other responses to requests for additional informa-tion by NRR.

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- UNITED STATES SSINS No.: 6820

NUCLEAR REGULATORY COMMISSION Accession No.:

o- 0FFICE OF INSPECTION AND ENFORCEMENT 7910250499 WASHINGTON, D.C. 20555 November 30,1979 IE Bulletin No. 79 LOSS OF NON-CLASS-1-E INSTRUMENTATION AND CONTROL POWER SYSTEM BUS ,

C'JRING OPERATION .

I Description of Circumstances. ,

i On Novenber 10, 1979, an event occurred at the Oconee Power Station, Unit 3 tnet resulted in loss of power to a non-class-1-E 120 Vac single phase power panel t. hat supplied power to the Integrated Control System (ICS) and the ..

Non-Nuclear Instrumentation (NNI) System. This loss of power resulted in control system malfunctions and significant loss of information to the.  !

control roca operator. i Soecifically, at 3:16 p.m., with Unit 3 at 100 percent power, the main condensate '

pacos tripped, apoarently as a result of a technician performing maintenance on

7e hotwell level control system. This led to reduced feedwater flow to the

"* stear generators, which resulted' in a reactor trip due to high coolant.sr! tem 3.:

p essure and simultaneous turbine trip at 3:16:57 p.m. At 3:17:15 p.m. , the r.on-class-1-E inverter power 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 channel tripped and failed to:

aatomatically transfer its loads from the DC power source to the regulated AC

-er source. The inverter tripped due to blown fuses. Loss of power to the iWI rendered control room indicators and recorders for the reactor coolant system (ex:ept for ene wide-range RCS pressure recorder) and most of the secondary plant sjste::s inoperasle, causing loss of indication for systems used for decay heat removal anc water addition to the reactor vessel and steam generators. Upon loss of power, all valves controlled by the ICS assumed their respective failure

. positions. The loss of power existed for approximately three minutes, until an .

erator 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 Novecber 15, 1979.

SUREG 0600 " Investigation into the March 28, 1979 TMI Accident" also discusses e i

TMI LER 78-021-03L whereby the RCS depressurized and Safety Infection occured i

on loss of a vital bus due to inverter failure.

A tions to Be Taken by Licensees for all power reactor f acilities with an operating license and for those nearing completion of construction (North Anna 2, Diablo Canyon, McGuire, Salem 2, 5equoyah, and 'Zirner):

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f f i I .I IE Bulletin:Ho; 79-27 November 30, 1579  ;

Page 2 of 3 l

~1. Review the class-1-E and non-class 1-E buses supplying power to safety and  !

i non-safety related instrumentationL and control systems which could affect '

the ability to achieve a cold shutdo.tn condition using existing proceduras or procedures developed under item 2 below. For each bus:

a); 'ident[fy and review the alarm and/or indication provided in th'e control f

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room to alert the operator to the loss'of power to the bus.

b) identify the instrument and control system loads connected to-4.he bus  !

and evaluate the effects of loss of power to these loads including  ;

the ability to achieve a cold shutdown condition. ,

c) describe any proposed design modifications resulting from these ieviews and evaluations, and your proposed schedule for implementing those -  !

modifications. ',

2. Prepare emergency procedures or review existing ones that will be used by -l control room operator.s, including procedures required to achieve a cold .

shutdown condition, upon loss of power to each class 1-E and non-class i 1-E bus supplying power to safety and non-safety,#eleted instrument and l control systems. The emergencf procedures should include:  ;

the diagnostics / alarms / indicators / symptom resulting from the review e W a) j

- and evaluation conducted per item 1 above.

I b) the use of arternate indication and/or control cl'rcuits which may be

• i powered from other non-class 1-E or class 1-E instrumentation and control buses. l c) methods for restoring power to the bus. '

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Describe any proposed design modification or administrative controls to be  !

implemented resulting from these procedures, and your proposed schadule for implementing the changes. l>

3. Re-review IE Circular No. 79-02, Failure of 120 \'olt Vital AC Power Supplias, j dated January 11, 1979, to include both class 1-E and non-class 1-E safety- ,

related power supply inverters. Based on a review of operating experience  !

j 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 .

tfie re ' review. i

4. Within 90 days of the date of this Bulletin, complete the review and i evaluation required by this Bulletin and provide a written response j{i describing your reviews and actions taken in response to each item.

l Reports should be submitte2 to the Director of the appropriate NRC Regicnal Office and a copy should be fon.'arded to the NRC Office of Inspection and  ;

Enforcewnt, Division of Reactor Operations Inspection, Washington, D.C. 20555.

j If you desire additional infonnation regarding this matter, please contact the

.- IE Regional Office. l t

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1 l'ovember 30, 1979  ;

' IE Bulletin No. 79-27' Page 3 of-3  ;

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Approved by GAO B180225 (R0072); cleara..ce expires 7/31/80. Approval was given -

under a blanket clearance specifically for identified generic problems. r i

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. IE Bulletin No. 79-27 Enclosure Novebmer- 30, 1979 d r

• - . RECENTLY ISSUED

.IE BULLETINS Bulletin Subject Dr.te Issued Issued To  ;

No, t

79-26 Boron Loss From BWR 11/20/79 All SWR power reactor 1 Control Blades facilities with an  ;

OL  ;

79-25 Failures of Westinghouse 1U2/79 All power reactor .

i BFD Relays In Safety-Related facilities with an Systems OL or CP .

79-17 Pipe Cracks In Stagnant 10/29/79 All PWR's with an '

(Rev. 1) Borated Wa.er System At , OL and for infor..ation PWR Plants -: -

to other power' reactors ,

79-24

" ~ " '

Frozen Lines 9/27/79 All power reactor-facilities which have g .., o a u, either OLs or cps ud - ,'..',.,

are in the latei tsge"

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of construction

• I, 79-23 Potential Failure of 9/12/79 All Power Reactor Emergency Diesel Facilities with an Generator Field Operating License or a construction permit Exciter Transformer 79-14 Seismic Analyses For 9/7/79 All Power Reactor .

(Supplement 2) As-Built Safety-Related Facilities with an i Piping Systems OL or a CP ,

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

To Each Licensee of Tritium Gas in Time- s;bo Receives Tubes pieces for Luminosity of Tritium Gas

'. Used in Timepieces for Lu:.inosity i I

Cracking in Feedwater 8/30/79 All Designated 79-13 System Piping Applicants for OLs - l (Rev. 1)

Pipe Support Base Plate 8/20/7S All power Reactor 79-02 Designs Using Concrete Facilities with an i (Rev. 1) OL or a CP (Supplement 1) Expansion Anchor Bolts  ;

Seismic Analyses For 8/15/79 All Pcwer Reactor 79-14' Facilities with (Supplement) As-Built Safety-Related ,

Piping Systems an OL or a CP ,

i 030.45 ALL OPERATING LICENSE APPLICANTS

SUBJECT:

CONCERNS RELATED TO I&E BULLETIN 80-06

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Operating reactors and some near term OL applicants have previously  :

received I&E Bulletin 80-06 which is enclosed. The concerns which prompted the Bulletin apply to all OL applicants. If you have not already ,

responded to the concerns of Bulletin 80-06, you are now requested to do ,

so, but with two exceptions. First, the time for response will be deter-  !

mined on a case by case basis rather than within 90 days. Secondly, your  ;

reply should be made in the same way as other responses to requests .

for additional information by NRR. .

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Of flCE GT ::J. i.Ci!ON AND E:ERCEW.NT i.M':EST ON , D. C. 20555 ,

J March 13,1980 .

IE Bulletin No. 80-D6 l, '

ENGINEERED SAFETY FEATURE (ESF) RESET CONTROLS l;

Description of Circumstances:

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On November 7,1979, Virginia ETe~ctric and Power Company (VEPCO) reported that  !.

following initiation of Safety Injection (SI) at North Anna Power Station i Unit 1, the use of the SI Reset pushbuttons alone resulted in certain ventila-tion dacpers changing position from their safety or emergency mode to their l normal node. Further investigation by VEPCO and the architect-engineer resulted in discovery of circuitry which similarly affected components actuated by a .' d Contain .ent Depressurization Actuation (CDA, activated on Hi-Hi Containment "

Pressure). The ci'rcuits in question are listed below:

Component / System Problem Outside/Inside Recirculation Spray Pump motors will not start after m > .m v.s ' ^ attbation if. CDA Ree~.t :is tdepre'ssed  ;

• " Pump Motors

- - prior to starting timer running  ;

- out (approx. 3 minutes) ,

Dampers will open on SI Reset  ;

Pressuiized Control Room

• Ventilation Isolation Dam; ers 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 Fans is depressed .

p 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 ..

y upon depressing CDA reset y

Pumps will not start after Service Water Radiation Monitoring actuation if CDA reset is depressed .

Saaple Pumps prior to motor starting timers  !

running out ,

U MainCondcaserAirEjeciorExhaust After receiving a high radiation monitor alarm on the air ejector n

Isolation Valves to the Containment exhaust, SI actuation would shut 1 these valves and depressing SI Reset i vould reopen them  ;

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: of 1 Riview of circuitry for ventilation dnf ers, motors, and valves ' reported by i i

VEPCO result:d in discovery of similar designs in ESF-actuated components at _  ;

8 i turry tinit 1 ar.d Beaver Valley;.where it has been found that certain equipent would return to its normal mode following the reset of an ESF signal; . thus,. !l protective actions of the affected systems could be compromised once the ,.l associated actuation signal is reset. These two plants had Stone and Webster j i

Engineering Corporation for the architect-engineer as did the North Anna j Units. I I

i i The Stone and Webster Engineering Corporation and VEPCO are preparing design '

1 changes to preclude safety-related equipment from moving out of its emwgtncy mode upon reset of an Engineered Safety Features Actuation Signal (ESFAS).

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

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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 proce,dures. It has been determined that logic diagrams may not l adequately reflect as-built c.ondi.,tions; therefore, the' requested review of ,

drawings must be done at the schematic / elementary diagr;am level. l l

There have been several communicatio*ns to licensees from the NRC on ESF reset actions. .,for ep; amp)p,:csoevof these.ccmr.unf atiam.,have been in the form of '

l Generic Letters issued Tn November,1978 and October,1979 on containment ,

l venting and purging during normal operation. Inspection and Enforcement i l

Bulletins Nos. 79-05, 05A, 05B, 06A, 058 and 08 that addressed the events at I l

THI-2-and NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and  :

lr l Short-Term Recommendations. However, each of these communications has -

addressed only a limited area of the ESF's. We are requesting that-the - i 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: ,

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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 (

i actuation signal, all associated safety-related equipment remains in its '

cmergency mode.

2. Verify the actual installed instrumentation and controls at the facility i are consistent with the schematics reviewed in Item 1 above by conducting i 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 ,. 1 performance of the testihg in your response to this Bulletin. .

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3. If any safety-related equipment does not remain in its emergency mode upon reset of an ESF signal at your facility, de>cribe proposed system .

codification, design change, or other corrective action planned to )

resolve the problem. ,

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4. h1ert in writing within 90 days, the results of your review and intim.'e  ;

a list of all devices which respond as discussed in item 3 above, acti as j taken or planned to assure adequate equipment control, and a schedule for -l This information is requested under '

ir.plementation of corrective action. Accordingly, you are requested to '

the provisions of 10 CFR 50.54(f). .

provide within the time period specified above, written statements of l the above information, signed under oath or affire.ation. P.aports shall l be submitted to the Director of the appropriate NRC Regional Office and  ;

l a copy shall be forwarded to the.NRC Office.of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C. 20555.

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For all pc er reactor facilities with a construction parmit, this Bulletin is l far information only and no written response is required. .

Approval was '. l Approved by GAO, B180225 (R0072); clearance expires 7-31-80.  !

given under a blanke.t clearance specifically for identificd generic problems. l 1

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030'.46 ALL OPERATIMG LICEMSE APPLICAPTS f 4 i

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SUBJECT:

HIGH EMERGY LINE BREAKS AND CONSEQUENTIAL COHTROL SYSTEM FAILURES ,

i Operating reactor licensees were informed by IE Information 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 3 impact the ' safety analyses and the adequacy of the protection functions performed. ,

by the safety grade eauipment. 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. ,

'The HRC is concerned that a similar potential may exist at light water ficilities  ;

now under construction. You are, therefore, requested to perform a review to  :

determine what, if any, design changes or operator , actions would be necessary to l assure that high energy.line breaks will not cause control system failures to  !

complicate the event beyond the FSAR analysis. Provide the results of your review including all identified problems and the manner in which you have resolved them to NRR. j

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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 applicaDie, but snould not  :

necessarily be limited to them. BWR applicants should consider analogous inter- '

actions as relevant to the BWR designs.

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' ENCLOSURE 1 J

UF ED STATES NUCLEAR kEGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 September 14. 1979 IE Information Notice No. 79-22 l

QUALIFICATION OF CONTROL SYSTEMS Public Service Electric and Gas Company notified the NRC of a potential unreviewed .,

safety question at their Salem Unit 1 facility. This notification was based on'a "li continuing review by Westinghouse of the environmental qualifications of equipment Based on the present status l that they supply for nuclear steam supply systems. l of this effort, Westinghoupe has informed their customers that the performance of non-safety grade equipment subjected to an adverse environment could impact the protective functions performed by safety grade equipment. These non-safety l j

grade systems include: -

Steam generator power operated relief valve control system Pressurizer power operated relief valve contro), system Main feedwater control system ,

Automatic rod control system These systems could potentially malfunction due to a high energy line break inside or outsid6 of containmer.t. NRC is also concerned that the. adverse '

environment could also give erroneous information to the plant operators.

Wastinghouse states that the consequences of such an event could.possibly be ._; .

more limiting than results presented in Safety Analysis Reports,.however, ~.

Westinghouse also states that the severity of the results can be limited ~ '

by operator actions together' with operating characterisitics of _the safety systems. Fur,ther. Westinghouse has recommended to their customers .that'they review their systems to determine 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 No will review the information specific action or response for possible applicability to their facilities.

is requested at this time. If NR" evaluations so indicate, further licensee a:tions may be requested or required. If you have questiens regarding this matter, please contact the DirecteF'of the apprepriate NRC Regional Office.

No written response to this Information Notice is required.

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l-ENCLOSURE 2 i

REPRINT '

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Westinghouse Electric Corporation-  ;

Water Reactor Division i Nuclear Service Division Box 2728 Pittsburgh, Pennsylvania 15230 l

• August 30, 1979 i

. PSE-79-21  !

Mr. F. P. 'Librizzi, General Manager  :

Electric Production  !

Public Service Electric and Gas Company. - - -

l 80 Park Place *' -

Newark, New Jersey 07101 .

Dear Mr. Librizzi:

Public Service Electric and Gas Co. l Salem Unit No.1 I QUALIFICATION OF CONTROL SYSTEMS  !

As part of a continuing review of the envirorinental qUa'lNi$a$ibns' of i

Westinghouse supplied NSSS equipment, Westinghouse has also found it . l necessary to consider the interaction with non-safety grade systems. -

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' This investigation has.been conducted to determine if the perfomance  : ,. j of non-safety grade systems which may not be protected from.an adverse - ; -

l environment could impact the protective functions perfomed by NSSS  !

safety grade equipment. The NSSS control and protection systems we,re  !

included in this review to assess the adequacy of the present environ- . . j mental qualification requirements. .

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

subjected to an adverse environment, could potentially lead to control - -

system operation which may impact protective functions. These. systems -

l are:

- Steam generator power operated relief valve control system 1 l I

- Pressurizer power operated relief valve control system i

- Main feedwater control system l t

- Automatic rod control system ,

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3 Each of the above mentioned systems could potentially malfunction if  ;

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

'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 l operating characteristics of the safety systems. i J

- 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 pl ant (s) .

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' . j To assist you in underst$nding these concerns, Westinghouse will hold a l seminar in Pittsburgh on Thursday, September 6 at Westinghouse R&O Center, i Building 701, with all our operating plant customers. The seminar will l address the potential impact of these .' concerns for various plant designs  !

and various licensing bases.

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Please contact your WNSD Regional Service office to conf.irm your attendance at the seminar. Wc Eill provide additional details concerning the agenda l and other meeting arrangements as they become available. _.

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Very truly yours, j ORIGINAL SIGNED BY l F. Noon, Manager i Eastern Regional & WNI Support l SR4/CCl3&l4 l cc: H. J. Midura j H. J. Heller  ;

R. D. Rippe j T. N. Taylor ,

R. A. Uderitz C. F. Barclay W l v

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ENCLOSURE 3 t

REPRINT j c PUBLIC SERVICE ELECTRIC AND GAS' COMPANY Salem Nuclear Generating Station .  :

P. O. Box 56 Hancocks Bridge, New Jersey 08038 l September 10, 1979 l 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 N-58701P SALEM NO. 1 UNIT LER This. letter .wi1MrrMe to confirm our telephone report to Mr. Gary '

Schneider of the Regional NRC offici oh' Friday, September 6,1979l' 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 conditipns ,

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 generator power operated relief valve control, pressurizer power -

operated relief valve control, main feedwater control and automatic -- ,

rod control systems.

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

the Technical Specifications.

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

CC: General Manager - Electric Production 1 Manager - Quality Assurance l

l ENCLOSURE 2 o -

MEETINgHIGHLIGHTS t

a 4 POTENTIAL UNREVIEWED SAFETY QUESTION ON INTERACTION BETWEEN NON-SAFETY GRADE SYSTEMS AND 5AFETY GRADE SYSTEMS ,

I&E Infomation Notice 79-22, dated September 14, 1979, was issued infoming j the nuclear industry of a potential unreviewed safety question at Salem,- Unit 1  ;

of Public Service Electric and Gas Company, based on a Westinghouse review of the environmental qualification of equipment. Certain non-safety grade equipment, if subjected to an adverse environment such as results from a high-energy line - ---:

break inside or outside of containment, could 1,mpact the safety analyses- and the  ;

t protective functions perfonged by safety grade equipment.

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were arranged

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with all. four light water reactor vendors according to the following schedule: }

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Wesiinghouse - T'uesday, September 18 .

1 Combustion Engineering - Wednesday, September 19 .-- -

Babcock and Wilcox - Thursday a.m., September 20 -

General Electric - Thursday, p..m., September 20 i

During the Westinghouse meeting, they identified, for all high-energy -li.ne. -1 -

f i breaks.and possible locations, the control systems that could. be affected as a-. result . {

of the adverse environment and whose consequential failure could invalidate the  ;

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

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i Westinghouse's investigation identified seven accidents and seven control systems l that could possibly interact and presented them in a matrix fom as shown in Enclosure T. As can be seen the potential interactions that could degrade the I

accident analyses are in the:  ;

a. Automatic Rod Control System ,.

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b. Pressurizer PORY Control System l

'c . Main Feedwater Control System Steam Generator PORY Control System d.

Westinghouse presented their reconnended short-term and long-term solutions, -

presented as EnclosureI 2.

Westinghouse stated that the possible matrix interactions may increase as more-detailed analyses are performed but the interactions will remain for all of ~their-plants and the interactions.jmay be eliminated only if' conditions are such that plant specific designs mitigate the interactions because of:

,..r.,,.,, a. system layout , ,,,, '

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b. type of equipment used

c. qualification status of equipment utilized I

d. design basis events considered for license applications

e. prior connitments made by utility to the NRC. ~

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

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

investigations are lower probability subsets of FSAR analyses whicy in themselves are sets of low probability.

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, 3 Westinghouse and the utility representatives all doubt that they can conclusively

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detemine the qualification status of all of the involved equipment in 20 days.  :

Both Westinghouse and the utility representatives stated that they will respond to the 20-day letter by addressing the four control systems identified in a manner suggested by the Westinghouse recomendations unless the NRC staff provides ,

directions to the contI'ary and further establishes guidelines stating their position on the problem along with their recomen'dations. ,

The NRC staff stated that they are sympathetic to the requests by the nuclear industry i i

regarding position and direction but this can be fomulated only at the conclusion ~

t of the scheduled meetings with all four light water reactor vendors. At that j time the staff will pra:ent their results, magnitude di.FGireM6i/ tb ' industry for !

resolution of the problem. , j r

At this time, it is not evident which utilities are faced with what environmental  ;

interaction problem. The effects of implementing all of the Westinghouse recomended j short-tem " fixes" may be contradicted by other sequences. There are three parts {

to the problem dealing with the basis of short-term operation ~:  ;

1. qualify equipment to the appropriaw environment; this would take Tong ~er

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

partial solution.  ;

2. short-term " fixes" should be in place pending long-tem solutions such as  !

the above. It must be noted that in this situation, some components that i I

are relied upon to iperate might possibly be wiped out by consequential  ;

failures under certain conditions and accident sequences if the postulated  !

adverse env'ironment is established. .

3. the " worst case" plant should be selected and a bounding analysis performed to

\ determine the time frame available for qualification of equipment.

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i 030.47 ALL OPERATIHG LICEt!SE APPLICANTS l

4

SUBJECT:

, C0l! TROL SYSTEM FAILURES i i

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.

Based on the conservative assumptions made in defining these " design bases" 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. l 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. .

(2) Indicate which, if any of the control systems idntified 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 4 c r. ' , .e malfunction of more than one control system and should extend to the effects ,

of cascading power losses due to the failure of higher level distribution panels and load centers.

(3) Incicate which, if any, of the control systems identified in (1) receive input signals from common sensors. The sensors considered should include, ,

but should not necessarily be limited to, common hydraulic headers or impulse lines feeding pressure, temperature, level or other signals to.two or more control systems.

l (4) Provide justification that any simultaneous malfunctions of the control systems identified in (2) and (3) resulting from failures or malfunctions of the ,

applicable common power source or sensor are bounded by the analyses-in Chapter. !

15 and would not require action or response beyond the capability of operators or safety systems.

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&Nb g DISTRIBUTION: APR 2 9 198 '

- Dodet F;;e 50-382 TLPDR%

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