ML19345G781
| ML19345G781 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 04/16/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Bryan J, Koester G KANSAS GAS & ELECTRIC CO., UNION ELECTRIC CO. |
| References | |
| NUDOCS 8104220342 | |
| Download: ML19345G781 (5) | |
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UNITED STATES a,E NUCLEAR REGULATORY COMMISSION a
WASHINGTON, D. C. 20555 APR 161981 Docket Nos.: STN 50-482 and STN 50-483 Mr. John K. Bryan Mr. Glenn Koester Vice President Vice President - Nuclear Union Electric Company Kansas Gas and Electric Company 1901 Gratiot Street 201 North Market Street Post Office Box 149 Post Office Box 208 St. Louis, Missouri 63166 Wichita, Kansas 67201
Dear Gentlemen:
Subject:
SNUPPS FSAR - Request for Additional Infomation As a result of our review of your application for operating licenses we find that we need additional infomation regarding the SNUPPS FSAR. The specific information required is as a result of the Instrumentation and Control Systems Branch's review and is listed in the Enclosure.
To maintain our licensing review schedule for the SNUPPS FSAR, we will need responses to the enclosed request by May 15, 1981.
If you cannot meet this date, please inform us within seven days after receipt of this letter of the date you plan to submit your responses so that we may review our schedule for any necessary changes.
Please contact Mr. Dromerick, SNUPPS Licensing Project Manager, if you desire any discussion or clarification of the enclosed report.
Sincerely,
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Robert L. Tedesco, Assistant Director for Licensirig Division of Licensing m
Enclosure:
f As stated h
cc: See next page
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8104220D W A
Mr. J. K. Bryan Mr. Glenn L. Koester Vice President - Nuclear Vice President - Nuclear Union Electric Corpany Kansas Gas and Electric Cnmpany P. O. Box 149 201 North Market Street St. Louis, Missouri 63166 P. O. Box 209 Wichita, Kansas 67201 cc: Gerald Charnoff, Esq.
Shaw, Pittman, Potts, Dr. Vern Starks Trowbridge & Madden 1800 M Street, N. W.
Route 1, Box 863 Ketchikan, Alaska 99901 Washington, D. C.
20036 Mr. William Hansen Kansas City Power & Light Compan" U. S. Nuclear Regulatory Commission ATTN: Mr. D. T. McPhee Resident Inspectors Office Vice President - Productioa RR #1
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1330 Baltimore Avenue Steedman, Missouri 65077 Kansas City, Missouri 64101 Ms. Treva Hearn, Assistant General Counsel Mr. Nichclas A. Petrick Missouri Public Service Commission Executive Director, SNUPPS P. O. Box 360 5 Choke Cherry Road Jefferson City, Missouri 65102 Rockville, Maryland 20850 Jay Silbera, Esquire Mr. J. E. Birk Shaw, Pittman, Potts & Trowbridge Assistant to the General Counsel 1800 M Street, N. W.
Union Electric Company Washington, D. C.
20036 St. Louis, Missouri 63166 Mr. D. F. Schnell Kansans for Sensible Energy Manager - Nuclear Engineering P. O. Box 3192 Union Electric Company Wichita, Kansas 67201 P. O. Box 149 St. Louis, Missouri 63166 Francis Blaufuse -
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Westphalia, Kansas 66093 Ms. Mary Ellen Salava Route 1, Box 56 Mr. Tom Vandel Burlington, Kansas 66839 Resident Inspector / Wolf Creek NPS c/o USNRC Mr. L. F. Drbl P. O. Box 1407 Missouri - Kansas Section Emporia, Kansas 66801 American Nuclear Society
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15114 Navaho Mr. Michael C. Keener Olathe, Kansas 66062 Wolf Creek Project Director State Corporation Commission Ms. Wanda Christy State of Kansas 515 N. 1st. Street Fourth Floor, State Office Building Burlington, Kansas 66839 Topeka, Kansas 66612 Floyd Mathews, Esq.
Birch,.Horton, Bittner & Monroe 1140 Connecticut Avenue, N. W.
Washington, D. C.
20036
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Page 1 of 10 ENCLOSURE 420.0 Instrumentation & Control Systems Branch 420.1 Loss of Non-Class IE Instrumentation and Control Power System bus During Power Operation (IE Bulletin /9-2/)
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. On November 30, 1979, IE Bulletin 79-27 was sent to operating license (OL) holders, the near term OL applicants (North Anna 2 Diablo Canyon, McGuire, Salem 2, Sequoyah, and Zi=mer),
and other holders of construction permits (CP), including.Callaway 1 and Wolf Creek. Of these recipients, the CP holders were not given explicit direction for making a submittal as part of the licensing review. However, they were informed that the issue would be addressed later.
You are requested to address these issues by taking IE Bulletin 79-27 Actions 1 thru 3 under " Actions to be Taken by Licensees".
Within the response time called for in the attached transmittal letter, complete tae 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 and submitted to the NRC Office of Nuclear Reactor Regulations as a licensing submittal.
420.2 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 (enclosed).
For facilities with operating licenses as of March 13, 1980, IE bulletin 80-06 required that reviews be conducted by the licensees to determine which, if any, safety functions might be 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 description
Page 2 of E 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 Regulation as a licensing sutaittal in the form of an FSAR amencment.
420.3 Oualification of Control Systems (IE Information Notice 79-22)
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 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 30, 1979 Westinghouse letter and a September 10, 1979 Public Service Electric and Gas Company letter which address this matter. Operating Reactor licensees conducted reviews to determine whether such problems could exist at operating facilities.
We are concerned tnat 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 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 interactions which might~~ occur. ~ Your review should~ 1nclude thase scenarios,-
=
where applicable, but should not necessarily bc.crited to them.
Applicants with other LWR designs should consider analogous interactions as relevant to their designs.
420.4 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.
Based on the conservative assumptions made in defining these design-basis events and the detailed review of the analyses by the staff, if is likely that they adequctely bound the consequences of single control system failures.
To provide assurance that the design basis event analyses adequately bound other more fundamental credible failures you are requested to provide the following information:
4
Page 3 Of 10 (1)
Identify those control systems whose failure or ma' function 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 hicher level distribution panels and load centers.
(3)
Indicate which, if any, of the control systens 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 reauire acticn or response beyond the capability of operators or safety systems.
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av::b u3-ch 13,1950 IE Sullstin "o. 53-C5 E.'.'2I.!EERED SAFETY FEATURE (ESF) RF. SET CCNTF.0LS Description of Circumstances:
Ser 7, '979, Virginia E5c'.Ht and Pcwer Ccepany (VE?CO) repcried t.'ut
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following initiatien of Safety Infection (SI) at Ncrth Anna Pc-er Statica
ait 1, the use of the SI Reset,Sshbuttens alone resulted in certain ventila-
.tica da:pers changing positi=n frc= thair safety or emergency made to th31r Furthar investigatica by VEPCO and the architect-enginear rasultad
.;;rmal node.
in disccvery of circuitry which siailarly affected ccaponents actuated by a Ccntain ent Depressurization Actuatica (CDA, activated on Hi-Hi Contaic: ant Pressura).
The ci'rcuits in questica are listed belew:
Pr:blea Ccaponent/ System Cutside/Inside Recirculatien Spray Pump motors will not start after actuatica if CDA Reset is d2 pressed Purp ": tors prior to starting timer running out (approx. 3 minutes)
Dampers will open on SI Reset Pressurized Control Room Ventilatien Isolation Da:pers.
Da=pers reposition to bypass Safeguarls Area Fiiter Da=pers filters when COA Reset is depressed i
f Fans will rest' art when CDA Reset Containment Recirculation Cooler i
is depressed Fans If service water is being used as Service Vater Supply and Discharge the cooling medium prior to CDA Valves to Contain=ent actuation, valves will reopen upon depressing CDA reset l
Service Water Radiation Monitoring Pumps will not start after actuation if CDA reset is depressed Sacple Purps prior to motor starting timers running out After receiving a high radiation Main endenser Air Efecto7 Exhaust monitor alarm on the air ejector Isolation Valves to the Containment exhaust, SI actuation would shut these valves and depressing SI Reset would reopen them
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P00R ORIGINAL Page Ei of f0
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R. view of.:ir...:.ry for ventilati:n D. ;;rs,
- t:rs, and vil e ts r2, ;rt. ! '.y 1t:d in discovery of simiiar designs in ESF-actuated componceas a:.
VEPCO :- w:
Surry Cait I a.d Ec:ver Valley; where it has been fcund that cartain e.i,- :nt would return to its normal mode fo11cwing the reset of an ESF 2i;r.31; th.:,
protective actions of the af fected systems cauld be compromis4d once the asscciat.d actuation signal is reset.
These two plants had St:ne md Vebste.-
Engineering Corporation for the architect-engiaaer as did tha.tet.1 Anna
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Units.
The St:ne and Vebster Engineering Corporation and VEPCO are preparing c'esign changes to preclude s.ifety related equipment from c:ving out of it:
amaracacy mode upon reset of an Engineered Safety Features Actuatica Signal (ESFAS).
This corrective action has been found acceptable by the NRC, in that, upe7 reset of ESFAS, all affected equipment remains in its e erg:ncy m.:de.
reviews of selected areas of ESFAS reset aci. ion en F'-M The N.o.C has performed facilities and, in some cases, this review was limited to examination of 1:;ic.
It has been determined that logic diagrams may not diagre.s and procedures.
adequaiely reflect' as-built conditions; therefore, the requestad review of drawings must ta done at the schematic /elemantary diagram level.
There have been several cor.municatic'ns to licensees from the NRC on ES For example, some of these ccmmunications have been in the form of actions.
Generic Letters issued in November,1978 and October,1979 on cont:inment Inspection and Enforcement venting and purging during normal operation.79-05, 05A, 05B, 06A, 068 and Bulletins Nos.
TMI-2 and NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations.
However, each of these communications has acdressed 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 FWR and SWR facilities with operating licenses:
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 1.
actuation signal, all associated safety-related equipment remains in its l
crer,gency mode.
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Verify th'e actual installed instrumentation and controls a 2.
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 various isolating or actuation signals.
Provide a schedule for the performance of the testing in your response to this Bulletin.
If any safety-related equipment does not remain in its emergency code upon reset of an ESF signal at your facility, describe proposed system 3.
codification, design ' change, or other corrective action planned to i
resolve the problem.
,a e o. of Iv, r g m i2 ' odira:I.2,
- rt in riting wit;.in 0 days, the results -f 2,
2 3 itava, ac 2
i;;t of all d:. ices which r2 p r.d as discussed in ;'.:.?
. s a..n 2:sure adequate eculp:cn. c:n.ro,t, y ~ a > ', " ". -..,,.
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/,ccordingly, y:u..:
- _s r:visi.:as of 10 CFR 50.5;(f).
- c5da ithin the tic.2 paried specified ab ve, writ'en st-t7
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b= su mitted ta the Direct =r of the appropriate ::RC,,.eq1cnas u,s w
r a c:;:v shall ha fo -arded to the N.RC Of fice of Inspec.len and E crear:n.,
4.0. 5.
e-erations Inspection, sashington, D.C.
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__F:r all pc-ar reactor facilities with a c:nstruction pa-.it, this 2 ll? tin is f..r inf rcation only and.s written response is rcquirad.
E-reved bY GM. S:20225 (20072); clearance expires 7-}i-C').
- p;r-val es gf.en :.nder a bi ar.ke.t clearance speci fica 3,.y,.or igen.... < e-og.. - ---u3 ns-er ue.-.
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Page 7 of 10 05" ED STATES NUCLEAR kEGULATORY CCM..ISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 Septemmer 14, 1979 IE Information N:tice No. 79-22 QUALIFICATION OF CONTROL SYSTEMS Puniic Service ETectric and Gas C:mpany nctified the NRC of a potential unreviewed This notification was based on a safety cuestien at their Salem Unit 1 f acility.
continuing revie= cy Westingnouse of the environmental qualificatiens of ecuipment Basec on the present status inat tney su::iy fc-nuclear steam su:cly systems.
cf nis eff:rt, Westinghcuse has informed tneir custacers that the performance of n:n-saf ety grace equipment sucjectec to an acverse envir:nment could impact These non-safety tne protective functions performee by saf ety grace equipmant.
grade systems incluce:
Steam generator power operated relief valve con:rol system Pressurizer power operated relief valve control system Main feeesater control system Automatic rod control system These systems could potentially malfunction due to a high energy lina break l
inside or outside of containment.
NRC is also concerned that the adverse environment c:uld als,o 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, Westingneuse nas recommended to their customers that they review their systems to cetermine whetner any unreviewed safety quash,1,ons exist.
I systems.
This Information Netice is provided as an early notification of a possibly It is expected that recipients will review the information significant matter.
No specific action or resp:nse for pessible applicability to their facilities.If NRO evaluations so indicate, further lice is requested at this time.
If you have questicas regarding this r.atter, a:tions may be recuested or required.
please contact the Director of the acpreoriate NRC Regional Office.
No written response t= this Information Notice is required.
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P00R ORIGINAL Page 8 of 10 REDR!NT Westinghouse Electric Corporation Water Reactor Division Nuclear Service Division Box 2723 Pittsourgh, Pennsylvania 15230 August 30, 1979 PSE-79-21 Mr. F. P. Litri::i, General Manager Electric Production Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101
Dear Mr. Litri:
Puolic Service Electric anc Gas Co.
Sal em Unit No.1 OUAL:FICAT:0N OF CONTRCL 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 witn non-safety grade systems.
This investigation has been conducted to determine if the performance of non-safety grade systems wnien may not be protected from an adverse environment could impact the protective functions performed by NSSS safety grace equipment. The NSSS control and protection systems were included in :nis review to assess the adecuacy 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 control system operation wnich may impact protect'4e functions. These systems are:
Steam generator power operated relief valve control systen Pressuri:er power operatec relief vaive control system Main feedwater control system Automatic rod control system (twwL
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Page 9 of 10 Page 2 PSE-79-21 Each of the a: eve mentioned systems could potentially malfunc: ion if imoactec ey acverse environments due to a hign energy line break inside er outsioe containment.
In eacn case, a limited se: Of creaxs, couple:
with cessicle consequential control malfunction in an adverse direc:1:n, of :ne a:cve events s:ule yiele results which are more limiting than those In all casas, Mcwever, :ne presented in the plant Safety Analys1s Reports.
severity of the results can ce limited by operator actions toge:ner with 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 ycur pl an:( s).
Tc assist you in understancing nese concerns, Westinghouse will hcid a seminar in Pittsburgn on Thursday, Septencer 6 at Westingnouse R&O Center, The seminar will Builc1ng 701, with all our operating plant customers.
accress :ne potential impact of these concerns for various plant designs and various licensing bases.
Please contact your WNSD Regional Service office to confirm your attendance We will provide additional details concerning the agenda at the seminar.
and other meeting arrangements as they become available.
Very truly yours, ORIGINAL SIGNED BY F. Noon, Manager Eastern Regional & WNI Support SR4/CC13bl4 cc:
H. J. Midura H. J. Heller R. D. Rippe T. N. Tayl or R. A. Uderit:
C. F. Sarcl ay W
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Page 10 of 10 REDoINT PUBLIC SERV!CE El aina ~~ u~ CCMPANY t
Salem Nuclear Generating Station P. O. Box 56 Hancocks Bridge, New Jersey CEC 35 Septemoer 10, 1979 Mr. Boyce H. Grier*
Director cf USNRC Office of Inspection ano Enforcement Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406
Dear Sir:
REPGRTABLE CCCURRENCE 79-53/0l?
SALEM NO. I UNIT LER This letter will serve to confirm our tele:none re: ort to Mr. Gary f
Senneioer of the Regional NRC office on Friday, September 6,1979, aovising of a potential reportable occurrence in accordance with Tecnnical Specification 6.9.1.8.
i We have been notifieo by our Engineering Department that a Westing-house conducted review of the environmental qualifications of Westingnouse sucolied NSSS eouipment 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 pover operated relief valve control, pressurizer power operated relief valve control, main feedwater control and automatic rod control systems.
A detailed report will be submitted in the time period specified by l
the Technical Specifications.
I Very truly yours, Original Signed By H. J. Midura Manager - Salem Generating Station l
AWX:jds CC: General Manager - Electric Production Manager - Quality Assurance ONLC-$W]I's0020
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