ML20003F958
| ML20003F958 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 04/16/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Draper E GULF STATES UTILITIES CO. |
| References | |
| IEB-79-22, IEB-79-27, IEB-80-06, IEB-80-6, IEIN-79-22, NUDOCS 8104270161 | |
| Download: ML20003F958 (7) | |
Text
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UNITED STATES 8%
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Docket Nos. : 50-458/459 Dr. E. Linn Draper, Jr.
Vice President - Technology Gulf States Utilities Company P. O. Box 2951 Beaumont, Texas 77704
Dear Dr. Draper:
SUBJECT:
RE00EST FOR ADDITIONAL INF0PyATION - RIVER BEND NUCLEAR PLANT, UNITS 1 AND 2 The Instrumentation and Control Systems Branch hos identifkd four concerns The that will be addressed in its review of operating license applications.
specific concerns are delineated in the enclosure. We request that you Should provide your responses within 45 days of tne date of this letter.
you have any questions, please contact us.
Sincerely,
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Robert L. Tedesco, Assistant Director for Licensing Division of Licensing 4
Enclosure:
As stated cc: See next page.
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Dr. E. Linn Draper, Jr.
Vice President - Tecnnology Gulf States Utilities Co:pany P. O. Box 2951 Beaumont, Texas 77704 cc: Troy B. Conner, Jr., Esq.
Conner, Moore & Corber 1747 Pennsylvania Avenue, N. W.
Washington, D. C.
20006 Mr. J. E. Becker, Safety & Licensing Gulf States Utilities Cc=pany P. O. Box 2951 Beau =cnt, Texas 77704 Stanley Plett=an, Esq.
.Orgain, Bell & Tucker Beaumont Savings Butiding Beauront, Texas 77701 Karin P. Sheldon, Esq.
Sheldon, Harmon & Weiss 1725 I Street, N. W.
Washington, D. C.
20006 William J. Guste, Jr., Esq.
Attorney General State of Louisiana P. O. Box 44005 State Capitol Baton Rouge, Louisiana 70804 Richard M. Troy, Jr., Esq.
Assistant Attorney General in Charge J
State of Louisiana Department of Justice 234 Loyola Avenue New Orleans, Louisiana 70112 e
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ENCLOSURE Instrumentation & Control Systems Branch Loss of Non-Class IE Instrumentation and Control Power System Bus During Power Operation (IE Bulletin 79-27)
If reactor controls and vita 1' 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 importand instrumentation upon which these operator actions should be based. This concerns was addressed in IE Bulletin 79-27. On November 30, 1979, IE Bulletin'79-27 was sent to operating license (OL) holders, the near term OL applicants (North Anna 2, DiU : Canyon, McGuire, Salem 2, Sequoyah, and Zimmer),
and other holders of construction permits (CP), including River Bend.
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 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 transmitted as a part of your FSAR (if the FSAR has already been printed it may be incorporated in an FSAR amendment and submitted to the NRC Office of Nuclear Reactor Regulations as a licensing submittal)
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Engineered Safety Features (ESF) Reset Controls (IE Bulletin 80-06) l 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 c7 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 tnat reviews be conducted by the licensees to determine which, if any, safety functions might be unavailabe after reset, and what changes could be implemented to correct the problem.
For facilities with a construction permit including OL applicantsBulletin 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 descriptionf a
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of corrective actions taken or planned as stated in Action 1 thru 3 and submit ~ the report called for in Actions Item 4.
The report should be submitted to the NRC Office of Nuc1 car Regulation as
..a licensing submittal.in the FSAR for River Bend (if the FSAR has not already been printed, it may be incorporated in the orginal FSAR),
Qualification of Control Systems (IE Infomation Notice 79-22)
Operating reactor licensees were informed by IE Infomation 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.
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 detemine 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 operator actions would be necessary to assure that high energy line breaks will not cause control system failutres to complicate the event beyond your FSAR analysis. provide the results of your revies 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 should consider analogous interactions as relevant to their designs.
Control System Failures The analyses reported in Chapter 15 of the FSAR are intended to l
democatrate the adequacy of sofety systems in mitigating anticipated
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l 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 J
that they adequately bound the consequences of single control system failures.
To provide assurance that the design basis event analyses adegntely bound other more fundamental credible failures you are requestc1 to provide the following infonnation:
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e (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.
(3)
Indicate which, if any, of the control systems identified in (1) receive input sugnals 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.
(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 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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3 BACKGROUfl0 INFORMATION l
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UNITED STATES SSINS No.:
5320 NUCLEAR REGULATORY CCPMISSION Accession No.:
OFFICE OF INSPECTION AND ENFORCE'ENT 7910250499 WASHINGTON, D.C.
20555 November 30,, 1979 IE Bulletin No. 79-27 LOSS OF hCN-CLASS-1-E INSTRUMENTATION AND CONTROL POWER SYSTEM BUS CURING OPERATION Cescription of Circu: stances:
Cn Novecber 10, 1979, an event occurred at the Oconee Power Station, Unit 3, inat resulted in loss of power to a non-class-1-E 120 Vac single phase power panel that supplied power to the Integrated Control System (ICS) and the Ncn-Nuclear Instrumentation (NNI) System. This loss of power resulted in control systec calfunctions and significant loss of information to the control rocm cperator.
5:ecifically, at 3:15 p.m., with Unit 3 at 100 percent power, the, main condensate I
- cos tripped, apoarently as a resuit"of a technician performing maintenanca on t7e hot-ell level control system. This led to redu
- ed feedwater fics to the stear generators, which resulted in a reactor trip due to high coolant system
- essure and simultaneous turbine trip at 3
- 16:57 p.m.
At 3:17:15 p.m., the r.:n-class-1-E inverter po-er supply feeding all power to the integrated control sy s te-: (which provides proper coordination of the reactor, steam generator fee $ater control, and turbine) and to one NNI channel tripped and failed to aatoratically transfer its loads from the DC power source to the regulated AC
- wer source.
The inverter tripped due to blown fuses.
Lcss of power to the MI rendered cen. col room indicators and racerders for the reactor coolant system (exce?t for One wide-range RCS pressure recorder) and most of the secondary plant i
sjstans incpera:le, causing loss of indication for systems used for decay heat removal and water adoition to the reactor vessel and steam generators. Upon loss t
of se=ar, all valves controlled by the ICS assumed their respective failure
- ositions.
The loss of pcwer existed for approximately three minutes, until an.
- erator could reach the equipment room and manually switch the inverter to the regulated AC source.
Tne above event was discussed in IE Information Notice No. 79-29, issued Nove:ter 15, 1979.
i WREG 0500 " Investigation into the March 28, 1979 TMI Accident" also discusses l
i TMI LER 78-021-03L whereby the RCS depressurized and Safety Injection occured 4
on loss of a vital bus due to inverter failure.
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ENGINEEREO 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-tion dacpers changing position from their safety or emergency mode to their normal node.
Further investigaticn by VEPCO and the architect-engineer resulted in discovery of circuitry which similarly affected ccaponents actuated by a Contain ent Depressuri:ation Actuation (CDA, activated on Hi-Hi Containment i
Press ure). The ci'rcuits in question are listed below:
l Compcnent/ System Prcbiea Outside/Inside Recirculation Spray Pump motors will not start after Pump Motors actuation if CDA Reset is depressed prior to starting timer running out (approx. 3 minutes)
Pressurized Control Room Campers will open on SI Reset Ventilation Isolation Da pers Safeguards Area Filter Dampers Campers reposition to bypass filters when CDA Reset is depressed
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Containment Recirculation Cooler Fans will rest' art when CDA R'eset Fans is depressed Service Water Supply a'nd 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 Pumps prior to motor starting timers running out i
Main Condenser Air Ejector Exhaust After receiving a high radiation Isolation Valves to the Containment monitor alarm on the air ejector g
exhaust, SI actuation would shut
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these valves and depressing SI Reset would reopen them l
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- i 3 R.: view of circr:try for ventilation da.; rs, estors, and valvas raport'.d by VEPCO result.'d in discovery of similar designs in ESF-actuated components at would return to its normal mode following the reset of an ES protective actions of the affected systems could be compromised once theThe associated actuation signal is reset.
Engineering Corporation for the architect-engineer as did the North Anna Units.
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The Stone and Vebster Engineering Corporation and VEPCO are preparing design changes to preclude safety-related equip er.t from coving out of its erargtney mode epon reset of an Engineered Safety Features Actuatien Signal (ESFAS).
This corrective action has been found acceptable by tha NRC, in that, upo1 reset of ESFAS, all affected equipment remains in its energency code.
reviews of selected areas of ESFAS reset action on PWR The NRC has perforcedfacilities and, in some cases, this review was limited to exa It has been determined that logic diagrams cay not diagrans and proce,dures. adequately reflect as-built conditions; therefore, the drawings must be done at the schematic / ele sntary diagram level.
8 There have been several cceeunicati6ns to licensees from the NRC For example, some of these communicaticns have been in the form of Generic letters issued in Nove:-ber,1978 and Octcber,1979 on containment actions.
Inspection and Enforcement venting and purging during normal operatien.79-05, 05A, 05B, OSA, 053 and TMI-2 and NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Bulletins Nos.
Fowever, each of these communications has Short-Term Recommendations.
We are requestirg that the acdressed only a linited area of the ESF's.
reviews undertaken for this Bulletin address all of the ESF's.
Actions To Be Taken By Licensees:
For all PWR and SWR facilities with operating licenses:
Rev'iew the drawings for all systems serving sa-fety-related functions at the schematic level to determine whether or not upon the reset of an ESF 1.
l actuation sigr.al, all associated safety-related equipment remains in its cmergency mode.
Verify the actual installed instrumentation and contro i
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a test to demonstrate that all equipment remains in its emergency mode upon removal of the actuating signal and/or manual resetting of theProvid various isolating or actuation signals.
performance of the testing in your response to this Sulletin.
If any safety related equipment does not remain in its emergency code up reset of an ESF signal at your facility, describe proposed system 3.
d to rodification, design change, or other corrective action planne l
resolve the problem.
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q:rt in writing within 90 days, the results of ycur r2 view and inc1?-::
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a list of all devices which respond as discussed in ite: 3 above, acti :s titen or planned to assure adequate equip:ent control, and a schedule f r This infer:atica is requested under inple:entation of corrective action.
the provisiens of 10 CFR 50.54(f).
Accordingly, you are requ sted to provide within the tima period specified above, written state:ents of the above infermatien, signed under oath or af firmation.
R ports chall be submitted to the Director of the appropriate 1RC Regional Office and a copy shall be fc:-arded to the tiRC Of fice of Inspection and Enforcement, e
20555.
Division of Reactor Operations Inspection, Washington, D.C.
For all pc-er reactor facilities with a constructica parait, this Sulletin is f;r information only and no written response is required.
clearanceexpires7-51-30
- pproval was Appr:ved by G*0, B:50225 (20072);
given t.nder a blanke.t clearance specifically for identificd generic prc51ers..
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ENCLOSURE 4-Uh" ED STATES NUCLEAR kEGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT 1
WASHINGTON, D.C.
20555 September 14, 1979 IE Inferr.ation Notice No. 79-22 0
QUALIFICATION OF CONTROL SYSTEMS Public Service Electric and Gas Company notified the NRC of a potential unrevirwed safety question at their Salem Unit 1 facility.
This notification was based on a continuing revie= by Westinghouse of the environmental qualifications of equipsent that they supply for nuclear steam supply systems.
Based on the present status of this effort, Westinghouse has informed their customers that the performance of non-safety grade equipment subjected to an adverse environment could impact tne protective functions performed by saf ety grade equipment.
These non-safety grade systems include:
5 team generater power operated relief valve control system Pressurizer power operated relief valve control system Main feersater control system i
Automatic red control system 1
These systems could potentially malfunction due to a high energy line break l
inside or ou* side of containment.
NRC is also concerned that the adverse environment could also give erroneous information to the plant operators.
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l Wastinghouse states that the consequences of such an event could possibly be more limiting than results presented in Safety Analysis Reports, however, West.inghouse aise 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 cetermine whether any unreviewed safety questions exist.
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This Infermation Notice is provided as an early notification of a possibly i
significant matter.
It is expected tnat recipients will review the information l
for possible a;plicability to their facilities.
No specific action or response is requested at this time.
If HRC evaluations so indicate, further 1,icensee a:tions may be requested or required.
If you have questiens regarding this matter, pieuse contact the Director of the aopropriate NRC Regional Office.
No written response to this Information Notice is required.
hloe 79p822p12_1
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4 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 Cc.
Salem Unit No.1 OUALIFICATION OF CONTROL SYSTEMS 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 of non-safety grade systems which may not De protected from an adverse environment could impact the protective furetions performed by NSSS The NSSS control and protection systems were safety grade equipment.
included in this review to astess the adequacy of the present environ-I mental qualification requirements.
As a result of this review, several systems were identified which, if subjected to an adverse environment, could potentially lead to control 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 doe 79_46Wob w-
Page 2 PSE-79-21 9
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 severity of the results can be limited by operator actions together with operating characteristics of the safety systems.
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).
To assist you in understanding these concerns, Westinghouse will hold a seminar in Pittsburgh on Thursday, September 6 at Westinghouse RA0 Center, Building 701, with all cur operating plant customers. The seminar will address the ootential impact of these concerns for various plant designs and various licensing bases.
Please contact your WNSD Regional Service office to confirm your attendance at the seminar. We 9ill provide additional details concerning the agenda and other neeting arrangements as they become available.
Very truly yours, ORIGINAL SIGNED BY F. Noon, Manager Eastern Regional 3 WN! Support SR4/CCl3314 cc:
H. J. Midura H. J. Heller R. D. Ripce T. N. Taylor R. A. Uderitz C. F. Barcl ay W
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REPRINT PUBLIC SERVICE ELECTRIC 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 !
631 Park Avenue King of Prussia, Pennsylvania 19406
Dear Sir:
REPORTABLE OCCURRENCE 79-58/0l?
SALEM NO. I UNIT LER This letter will serve to confim our telephone report to Mr. Gary Schneider of the Regional NRC office on Friday, September 6,1979, advising of a potential recortable occurrence in accordance with Technical Specification 6.9.1.8.
We have been notified by our Engineering Department that a Westing-house conduc*ed review cf the environmental qualifications of Westinghouse supplied NSSS ecuipment has identified that concitions 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 coeratad relief valve control, main feedwater control and automatic red control systems.
A detailed report will be submitted in the time period specified by the Technical Specifications.
Very truly yours, Originel Signec By H. J. Midura J*w= er. Salem Generating Station 5
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