ML20003H627
| ML20003H627 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/24/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Abel J COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 8105060529 | |
| Download: ML20003H627 (4) | |
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g Dis APR 2 41981 g oc et Fi 9 Lord Kdg DEisenhut bec: TERA BJYounglood NRC/PDR ABournia L/PDR Docket Nos.: 50-173 MRushbrook NSIC an@ALd RTedesco TIC RVollmer ACRS (16)
TMurley Mr. J. S. Abel Dross Director of Nuclear Licensing RHartfield, MPA Commonwealth Edison Company VNoonan Post Office Box 767 OELD Chicago, Illinois 60690 OIE (3) l
Dear Mr. Abel:
Requ st for Additional Information - La Salle County Station.
Subject:
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Units 1 nd 2 Our Instrumentation and Control Systems Branch has identified additional concerns that must be resolved prior to the first refueling of La Salle Unit 1.
The specific concerfis are delineated in the enclosure. We request that you amend your Final Safety Analysis Report to reflect your responses as soon as possible.
l If you desire any discusswa or clarification of the infomation requested, please contact A. Bournia, Licensing Project Manager, (301) 492-7200.
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Sincerely,
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Robert L. Tehsco Robert L. Tedesco, Assistant Director '
for Licensing l
Division of Licensing g,\\,NdJ$f f 1'
Enclosures:
As stated eg
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cc: See next page
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Mr. J. S. Abel Director of Nuclear Licensing Corrmonwealth Edison Company Post Office Box' 767 Chicago, I;111nois ' 60590 cc: Philip P. Steptoe, Esq.
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One First National Plaza Chicago, Illinois 60603 Dean Hansell, Esquire Assistant Attorney General 188 West Randolph Street
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' Suite 2315 Chicago, I.llinois 60601
.Mr. Roger Walker, Resident Inspector -
U. S. Nuclear Regulatory Commission Post Office Box 224 Marseilles, Illinois 61364 ao D es e M*
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ENCLOSURE Request for Additional Information 3
LaSalle County Station Units Nos.1 & 2 Docket Nos. 50-373 and 50-374 030.0 Instrumentation & Contrcl Systems Branch 031.288 Qualification of Control Systems (IE Information Notice 79-22)
J Operating reactor licensees were informed by IE Information Notice
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79-22, issued September 19, 1979, that certain non-rafety 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 j
equipment. Attachment 1 is a copy cf IE Information Notice 79-22, and reprinted copies of an August 30, 1979 Westinghouse letter and a September 10, 1979 Public Service Electric ard Gas Company letter which address this matter. Operating Reactor licensees I
l 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, requested to perform a review to determine what, if any, design changes or l
operator actions would be necessary to assure that high energy line breaks will not cause control system failures to complicate the i
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event beyond your FSAR analysis. Provide the results of your:
reviews 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 interactions which might occur. Your review should include those scenarios, where applciable, but should not necessarily be limited to them.
Applicants with other LWR designs sho31d consider analogous interactions as relevant to their designs.
031.289 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. Both Congress and ACRS have expressed such concern. Conunissioner Ahearne has responded to Congress l
regarding this issue (Attachment 2) and part of his response referred l
to control systems reviews to be performed in connection with OL licensing.
Based on the conservative assumptions made in defining these Chapter 15 l
design-basis events and the detailed review of the analyses by the staff, it is likely that they adequately bound the consequences of single l
control system failures.
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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 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 malfuction 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.
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(3)
Indicate which, if any.of the control systems identified in (1) receive input sfgnals from common sensors. The sensors considered should include, but should not necessarily be limited to, common hydraulic headers or impulse lines feeding I
pressure, temperature, level or ot!.er signals to two or I
more control systems.
(4)
Provide justification that any simultaneous malfunctions of the control systems indentified in (2) and (3) resulting 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 capability of operators or safety systems.
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ATTACHMENT 1 i
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UN ED STATES NUCLEAR kfGULATORY CO WISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 September 14, 1979 IE Infom.ation Notice No. 79-22
' QUALIFICATION OF CONTROL SYSTEMS Public Service Electric and Gas Company notified the NRC of a potential unrevir-ei This notification was based on a saf ety cuestion at their Salem Unit 1 facility.
continuing revie. by Westingnouse of the environmental qualifications of equipment Based on the present status that they supply for nuclear steam supply systems.
of this effort, Westinghcuse has informed their customers that the performance of non-safety grade equipment sub.iected to an adverse environment could impact These non-safety the protective functions performed by saf ety grade equipment.
grade systems incluce:
Steam generator power operated relief valve control system 1
i Pressurizer power operated relief valve control syste:
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Main feer.<ater control system i
l Automatic rod control system l
These systems could potentially malfunction due to a high energy line break i
inside or outside of containment.
NRC is also concerned that the adverse
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environment could also give erroneous information to the plant operators.
Westinghouse 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 Further, Westinghouse has recoc, ended to their customers that they review their systems to determine whether any unreviewed safety questions exist.
systems.
This Information Notice is provided as an early notification of a p No specific action or response significant matter.
for possible applicability to their facilities. evaluations so indicate, further licensee is requested at this time.
If NR If you have questiens regarding this r.atte-actions may be requested or required.
please contact the Director of the appropriate NRC Regional Office.
No written response to this Information Nctice is required.
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REPRINT Westinghouse Electric Corporation Water Reactor Division Nuclear Service Division Box 2728 Pittsburgh, Pennsylvania 15230 August 30, 1979 e
PSE-79-21
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Mr. F. P. Librizzi, General Manager Electric oroduction Public Service Electric and Gas Company 80 Park Plbr.e Newark, New Jersey 07101
Dear Mr. Librizzi:
Public Service Electric and Gas Co.
Salem Unit No.1 00ALIFICATION OF CONTROL SYSTEMS As part of a continuing review of the environmenta' 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
.of non-safety grade systems which may not be protected from an adverse environment could impact the protective functions performed by RSSS The NSSS control and protection systems were safety grade equipment.
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 which, if subjected to an adverse environment, could potentially lead to cont"ol These systems system operation which may impact protective functions.
are:
Steam generator power operated relief valve control system Pressurizer power operated relief valve control system
Main feedwater control system Automatic rod control system a
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Page 2 PSE-79-21 Each of the above mentioned systems could potentially malfunction if impacted by adverse enviromnents due to a high energy line break inside In each case, a limited set of breaks, coupled or outside containment.
with possible consequential control malfunction in an adverse dire: tion, of the above events pould yield results which are more limiting than thoseIn all c presented in the plant Safety Analysis Reports. severity of the results operating characteristics of the safety systems.
We believe these systems identified do not constitute a substantial safety However, Westinghouse recommends you review them to determine if hazard.
any unreviewed safety questions or significant deficiencies exist in your pl ant ( s).
To assist you in understanding these concerns, Westinghouse will hold a seminar in Pittsburgh on Thursday, September 6 at Westinghouse RA0 Center, The seminar will Building 701, with all~ our operating plant customers. address the pote and various licensing bases.
Please contact your WNSD Regional Service office to ecnfirm your attendance We will provide additional details concerning the agenda at the seminar.
and other meeting arrangements as they become available.
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Very truly yours, p
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. Tayl or R. A. Uderitz C. F. Barclay W
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REPRINT FUGLIC SERVICE ELECTRIC AND GAS'COMDANY e
Salen Nuclear Generating Station P. O. Box 56 Hancocks Bridge, New Jersey 08038 Septer:ber 10, 1979 Mr. Bcyce H. Grier' Director of USNRC Office of Inspection and Enforcement Region 1 631 Park Avenue King of Prussia, Pennsylvania 19405
Dear Sir:
REPORTABLE OCCURRENCE 79-58/OlP SALEM NO. 1 UNIT LER This letter will serve to confim our telephone report to Mr. Gary Schneider of the Regional NRC office on Friday, Seotember 6,1979, advising of a potential reportable occurrence in accordance with Technical Specification 6.9.1.8.
We have been notified by' cur 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 generator power operated relief valve control, pressurizer power operated relief valve control, main feedwater control and aetomatic rod control systems.
A detailed report will be submitted in the time period specified by the Technical Specifications.
Very truly yours, Original Signed By H. J. Midura Manager - Salem Generating Station AT4: ids 7
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November 17, 1980 l
Tne Honorable Morris K. Udall, Chairman Subco.mittee on Energy and the Environment Cemittet on Interior and Insular Affairs United States House of Representatives Washington, D. C.
20515
Dear Mr. Chairman:
This responds to your letter of June 11, 1980 in which you asked the Comission again to consider seriously control system failures in nuclear power plants.
Since such failures may have severe consequences, the NRC staff has begun to better define their safety significance. summarizes these actions.
Recent ope *ating experience, such as the Crystal River event of last February, and continuing evaluation of the control system failure issue, has led the staff to propose to the Commission that the issue has sufficient safety sig-nificance to justify its being designated as an " Unresolved Safety Issue" (USI) and reported to the Congress under Section 210 of the Energy Reorgani-That proposal is currently under consideration zation Act (See Enclosure 2).
by the Comission.
Classification as an USI would assure priority f,or resources needed for timely and effective resolution of this issue.
At the present time, the Comission is relying on the consensus engineering judgment of senior staff that the risk associated with control system failures is not sufficient to require imediate corrective actions such as power derating. This judgment is not based on any special analyses or calculations beyond those normally performed in the course of staff review of postulated transients and accidents. We recognize (as you noted in Mr. Denton's October 22, 1979 memorandum) that the analyses do not take into account all events that can be postulated. The program outlined in Enclosure 2 is intended to provide a better basis for judging the adequacy of plant protection featu es and operator actions to mitigate control system failures.
With respect to the differences you noted between the wording of our May.14 response and that of the previous staff statement enclosed with that response, the wording in the May 14 letter does convey a greater sense of certainty about the adequacy of analyses performed to evaluate the interaction between high energy lines and control systems than does the December 19, 1979 memorandum.
We regret any misunderstar. ding this may have caused.
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MEM3RANDUM FOR:
FROM:
Harold Denton, Director Office of Nuclear Reactor Regulation
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THRU:
William J. Dircks Acting Executive Director for. Operations SU5]ECT:
ACRS AND AEOD COMMENT 5 CONCERNING NEW UNRESOLVED SAFETY ISSUES In esp:nse to your memorandum of August 19, 1980, we have reviewed the ACRS and AEOD coments on SECY S0-325 and have the following coments and recomendations.
We recommend that the two issues that AEOD recommended be considered as
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Unresolved Safety Issues be added to the litf of items recuiring further study before deciding on designation as USI's. This further study will We recomend that the other two be performed over the next few months.
concerns raised by AEOD be incorporated into existing or proposed Unresolved Safe.ty Issues es described'in Enclosure 1.
i The ACRS :omments" include a recomendation to add to the list of Unresolved l
Safety Issues ihree issues that the staff had initially screened out.
Upon reconsidera, tion we now rec: mend that one of these (." Control System Reliability"1 warrants designation as an Unresolved safety Issue; Enclosure l
3 provides a. description of the issue that we propose to add to the In di.scussing the issue of Control System Special Report to Congress.
Reliability, the ACRS also noted the related issue of the reliability of nonsafety system informa. tion displayed for use of the reactor We recomend tha.t this ACRS concern be added to the list of cDerator.
items requiring further study to evaluate their.'imoact on overall risk This fu'r'ther study will be before deciding on designation as a USI.
performed over the next few months.
We do not agree that the two other issues recomended by ACRS (D.C.
Power Reliability and the Single Failure Criterion) warrant designation as Unresolved Safety Issues for the reasons described in Enclosure 2.
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