ML20086U410
| ML20086U410 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/29/1984 |
| From: | Gucwa L GEORGIA POWER CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20086U411 | List: |
| References | |
| GL-83-28, NED-84-054, NED-84-54, NUDOCS 8403070251 | |
| Download: ML20086U410 (24) | |
Text
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N, N N ,. N 5 2y j f [,;~J NED-84-054 February 29, 1984 -
Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Comission Washington, DC 20555 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57. NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 RESIONSE 'IO GENERIC LE' ITER 83-28, SALEM RICUIREMENTS Gentlenen:
Our status report dated Noveder 7, 1983 on the Salen Generic Rajuirenents provided infomation on current confomance with the positions of the lettet and also provided plans and schedules for canpliance where these were known. Attachnent 1 to this letter is an update on Plant Hatch conformance and further infomation related to plans and schedules for any needed improvenents.
Attachment 2 is a detailed description of a Vendor Riuipnent Technical Information Program developed by the INPO Nuclear Utility Task Action Comittee (NUTAC) on Generic Letter 83-28, Section 2.2.2. The report is provided as a supplanent to our response to Section 2.1 and 2.2.2.
Please contact this office if you have any questions or car:nents.
Yours very truly, fH &a
- L.T. Gucwa PLS/mw Attachnents ,
xc: J.T. Beckham, Jr. f H.C. Nix, Jr.
J.P. O'Reilly (NRC - Region II) (9 Site Resident Inspector - Plant Hatch ,(&
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o ATTACINENT 1 PIANT EDWIN I. HA'IGI UNITS 1 AND 2 RESPONSE 'IO GENERIC IFI'IER 83-28 SALEM GENERIC REQUIRDENTS FEBRUARY 29, 1984 _
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NCTIT:
A stmnary of requirenents of Generic Letter 83-28, dated July 8,1983, are stated in bold letters. Paragraph ntsnber designations in the responses correspond to number designations of positions of the Generic Letter. Plant Hatch status of compliance is provided along with plans and schedules for any needed improvenents. A schedule stmnary is provided at the end of the attachnent.
1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)
POSI'" ION LICENSEES AND APPLICANTS SHALL DESCRIBE 'INEIR PROGRAM ECR ENSURING 'IHAT UNSCHEDULED REACIOR SHUI 1XMNS ARE ANALYZED AND 'IHAT A DETEPMINATION IS MADE THAT THE PIANT CAN BE RESTARTED SAFELY.
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Status:
- Our response duted Novt. ter 7,1983 described Plant Hatch practices for post-trip reviews and restart authorization.
Plans and schedules for ccznpliancer As reported in the Novenber 7 response, current Plant Hatch procedures and practices meet the requirenents 4 the generic letter. Improvenents based on INPO and BWR Owners Group reccmnendations will be incorporated, where appropriate, but no further action is required for the purposes of this response.
1.2 POFJ 'IRIP REVIEW - DATA AND INEORMATION CAPABILITY POSITION LICENSEES AND APPLICANTS SHALL HAVE OR HAVE PIANNED A CAPABILITY 'IO RECDRD, RECALL AND DISPLAY DATA AND INFORMATICN 'IO PEINIT DIAGNOSING 'INE CAUSES OF UNSCHEDUED REACIOR SHUIDOWNS PRIOR 'IO RESTART AND FOR ASCERTAINING THE PROPER EUNCTIONING OF SAFETY-RELA'IED ECUIEMENT.
ADBCUATE DATA AND INFOINATION SHALL BE PROVIDED 'IO CORRECILY DIAGNOSE THE CAUSE OF UNSCHEDULED REAC10R SHUID00ES AND 'INE PROPER EUNCTIONING OF SAFETY-REIA'IED EQUIPMENT DURING 'INESE EVENTS USING SYSTEMATIC SAFEIY ASSESENENT PROCEDURES (ACTION 1.1) . 'INE DATA AND INFORMATION SHALL BE DISPIAYED IN A EOPM 'IHAT PEBMITS EASE OF ASSIMIIATION AND ANALYSIS BY PERSONS TRAINED IN 'INE USE OF S'._iTEMATIC SAEErY ASSES! MENT PROCEDURES.
FEB 2 91984
Status:
.Our Novenber 7 response provided a description of information sources used for post-trip review, and stated that an evaluation of the adequacy of existing instruentation would be provided with this response.
Se evaluation consisted of _a review of data required to execute procedure HNP-426, "Scran/ Transient Reporting", and an evaluation of operating experience in 1982 and 1983. 'Ihe evaluation considered both
. detemination of the root cause of the transient and the performance of safety related equipnent. 'Ihe purpose of procedure HNP-426 is to collect and doceent all pertinent information concerning a scran or transient, and to analyze plant response at all levels to determine if corrective action is needed prior to returning the unit to service.
A review of scran reports fran calender years 1982 and 1983 shows there were a total of 28 unscheduled reactor shutdowns during the two year period. In all cases, the signal initiating the trip was identified, and the analysis of the transient identified the initiating event. .
For evaluating the performance of safety related equipnent, a section of HNP-426 rsluires a review of the operation of equipnent including all safety i systems. Se control room instrunentation is adequate for determining the
, start and stop times and for checking the maximun and minimun values of process parameters which initiate the systens. Actual performance of the
.. equipnent to the level .of run-up time, flows and pressures is not determined during post trip review unless there is a malfunction which requires greater scrutiny. Performance of safety- related sluipnent is monitored during f
periodic surveillance required by the Technical Specifications.
Based .on favorable operating experience with existing systems, we
! conclude thet no additional enhancement is required for the purpose of
! performing post-trip analysis. Adequacy of control rom instrumentation for a wide range of needs will be evaluated during a detailed control roan design review -(DCRDR) required by NUREG 0737 Supplement 1. A report on the
. Hatch DCRDR is scheduled for NRC sutaittal in June,1986.
Plans and schedules for compliance:
( Existing systens, as described in the Novenber 7 response, provide adequate information for a canprehensive and thorough post-trip review. No
- additional enhancenents are required to meet the positions of Generic Letter 83-28..
L L ! H B 2 3 1984
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I 2.1 EQUIINDTI CIASSIFICATION AND VENDOR INTERFACE (REACIOR TRIP SYSTD4 ~
CCMPONENIS)
POSITION LICENSEES AND APPLICANTS SHALL CONFIIM THAT ALL CCMPONDTTS hBOSE MJNCTIONING IS REQUIRED TO TRIP THE REACIOR ARE IDENTIFIED AS SAEETY-REIATED ON DOCLNENTS, PROCEDURES, AND INEOINATION HANDLING SYS7TMS ' USED IN THE PLANP TO CONTROL SAFETY-REIATED ACTIVITIES, INCIUDING MAINITNANCE, NORK ORDERS, AND PARTS REPIACEMENT. IN ADDITION, FOR THESE C04PONDFIS, LICENSEES AND APPLICANTS SHALL ESTABLISH, IMPIDENT AND MAINTAIN A CONTINUING PROGRAM TO ENSURE 7 EAT VENDOR INECENATION IS 03(PIEPE, CURRENP AND OJN1 POLLED THROUGHOUT THE LIFE OF 7EE PIANT, AND APPROPRIATELY REFERENCED OR IN00RPORATED IN PIANT INSTRUCTIONS AND PROCENJRES.
Status:
he Novenber 7 response stated that an evaluation of existing cmpliance and plans for any needed improvenents would be provided with this response. The wording of . Generic Letter 83-28 has created a minor sanantic problen which will be clarified based on a position d,eveloped by the BWR Owners Group Salen Requirenents Carnittee.
NUREG-1000 notes that the Reactor Trip System is the systen that initiates a scran including sensors, power supplies, etc. Plant Hatch does not have a specific systen naned the Reactor Trip Systen; rather the trip function is performed by several systens in the plant. In particular the reactor trip function (RTF) is performed by:
(a) Control Rod Drive
~ (CRD, Systen Cll), Reacter Protection Systen (RPS, Systen C71), Neutron Monitoring (Systen C51), Nuclear Boiler (Systen B21) Process Radiation Monitoring (Systen Dll) and Turbine
&- Systen (Systen N32) are systens that provide sensor inputs to the Reactor Protection Systen.
(b) The Reactor Protection Systen which contains logic, power supplies, etc.
(c) The Control Rod Drive Systen which contains the scram pilot valves that emplete initiation of a scran.
Since creation of a Reactor Trip Systen (R7S) would cause confusion with the existing plant systens, GPC decided to respond to Letter 83-28 Iten 2.1 on a systen level basis by covering the systens that perform the RTF as part of. this Iten 2.1 response and then cover the renaining safety related systens as part of the Iten 2.2 response. The Iten 2.1 review consequently includes all RTS cmponents as well as any other safety relate 3 cmponents in theso systens. The specific cmponents that form the RTS were not separately identified.
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'Ibe systens that provide the reactor trip function will be reviewed for correct docunentation of quality requirenents in the Plant Hatch Muipent Iocation Index (ELI). Approximately 600 relays in the reactor trip function systens will be added to the ELI to provide additional assurance that quality requirenents are properly used and docunented.
For vendor interface, Georgia Power Company has an equipent information program with General Electric Canpany (GE) , the NSSS vendor. %is progran consists of two major categories: 1) information regarding
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safety-related systens and components; and, 2) technical information intended to enhance safety and non-safety related equignent reliability and thereby improve plant performance. %ese programs include:
- 1. 10CFR21 Reporting -- We General Electric Company has established a reporting systen to handle safety concerns that conplies with the requirenents of 10CER21.
- 2. - Urgent Connunications - In addition to the 10CFR21 reports, a proc 2 dure for handling urgent connunications to BWR owner / operators -
has been established for use in providing fast notification of safety concerns. nese canunications are usually in the form of a short letter which provides a _ brief explanation and advice or precautionary - measures to be observed to avoid potential operational hazards. Due to their urgent nature, these i
connunications are processed to operating plants by the most effective method (i.e. . telex, telecopy, cable, special mail handling, etc.).
-3. . Service Information Letters (SILs) -
%ese docinents are usually brief, providing reemnendstions for equipnent modification, plant design improvements 'or changes to procedures to improve plant performance.
- 4. Service Advice Letters - ' %ese documents are issued by GE Products t-Departments other than the San Jose based Nuclear Energy Products Departments and are used to provide notification of product problems and/or service information on a broad range of GE constner and industrial products. Service Advice Letters that are recognized by the issuing product department as applying to devices l-
" used in nuclear plants are specially identified and are flagged for distribution to all_ nuclear plants.
- 5. 'Iurbine Information Letters (TILs) -
%ese documents are issued by GE's Large Steam 'Ibrbine Generator Department to provide descriptions of- product problens/ improvements and to recanend modifications that will mitigate problems or improve product performance.
FEB 2 91984
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- 6. Application Infomation Docments -
%ese docments are " white papers" that describe potential operating problens and provide i
design change or operating recemnendations to mitigate or avoid then. . 'Ibese doments are primarily aimed at requisition plants, but are also forwarded to operating plants when they have any applicablility to those plants.
For SIIs 'and TILs, Georgia Power Capany has a program to review
(. supplied information to assure that it is emplete, current and controlled and appropriately referenced or incorporated into plant instructions and procedures. A system of positive feedback to GE is in place. Additional infomation on vendor interface is contained in the response to Section 2.2.2.
Plans and schedules for empliance:
' B;uipnent Incation Index review and incorporation of RPS relays into the ELI will be completed by January 1,1986. The administrative program and procedures now in place for vendor infomation will be reviewed to
' confim - that they implenent the guidelines of the Vendor B;uipnent
- Technical Infomation Program described in attachnent 2. His review is scheduled to be cmpleted and appropriate changes made prior to January 1, 1985.
i .
2.2 EQUIPMENT CIASSIFICATION AND VENDOR INIERFACE (PROGRAMS EUR ALL SAETIY-REIATED CINPONENTS) .
j POSITION
, LICENSEES AND APPLICANIS SHALL SUIMIT, FOR STAFF REVIEW, A DESCRIPTION j OF .THEIR PIOGRAMS FOR SAFETY-REIATED EQUIPMENP CIASSIFICATION AND VEPEOR INIERFACE AS DESCRIBED BEION:
- l. FOR EQUIPMENP CLASSIFICATION, ' LICENSEES AND APPLICANTS SHALL DESGIBE 'HIEIR PROGRAM POR ENSURING 'IflAT ALL CINPONENTS OF SAFETY-REIA'IED SYS'HMS NECESSARY POR ACCCMPLISHING RECUIRED SAFETY FUNCTIONS ARE IDENTIFIED AS SAFETY-REIATED ON DOCLNENIS,
! PROCEDURES, AND INECENATION HANDLING SYTEMS USED IN THE PIANT 'IO CONISOL SAFETY-REIA'IED ACTIVITIES INCIDDING MAINNNANCE, WORK ORDERS AND REPIAGMENT PARTS. 'ITIIS DESCRIPTION SHALL INCLUDE:
- l. 'IffE CRIIERIA POR IDENTIFYING CINPCEENTS AS SAFETY-RELA'IED WITHIN SYS'HMS CURRENTLY CIASSIFIED AS SAFETY-REIATED. 'I1fIS SHALL NOT BE INTERPRETED 'IO REQUIRE CHANGES IN SAFETY CIASSIFICATION AT 'ITE SYSTEMS LEVEL.
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- FEB - 2 9 1984
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- 2. A DESCRIPTION OF 7EE INFORMATION HANDLING SYSTB4 USED TO IDENTIFY SAFETY-REIATED CMPONENIS (E.G., C MPUTERIZED EQUIEMENT LIST) AND THE MEIEODS USED FOR ITS DEVEICIMENT AND VALIDATION.
3.
A DESCRIPTION OF THE PROCESS BY WHICH STATION PERSONNEL USE
'IHIS INFOINATION HANDLING SYS77M 'IO DETERMINE THAT AN ACTIVITY IS SAFEIY-REIATED AND WHAT PROCEDURES FOR MAINIENANCE, SURVEILIANCE, PARTS REPIACEMENT AND OIEER ACTIVITIES DEFINED I IN THE INTRODUCTION TO 10 CFR 50, APPENDIX B, APPLY TO !
SAFETY-REIATED OCMPONENTS.
- 4. A DESCRIPTION OF THE MANAGEMENT CONIROIS UTILIZED TO VERIFY [
THAT THE PROCEDURES FOR PREPARATION, VALIDATION AND ROUTDE "
UTILIZATION OF 'INE INFORMATION HANDLING SYSTEM HAVE BEEN FOLIONED.
- 5. A DEMONSTRATION THAT APPROPRIATE DESIGN VERIFICATION AND QUALIFICATION TESTING IS SPECIFIED EOR PROCUREMENT OF SAFETY-REIATED OO4PONENIS. THE SPECIFICATIONS SHALL INCLUDE QUALIFICATION TESTING FOR EXPECTED SAFETY SERVICE CONDITIONS AND PROVIDE SUPPORT EOR TEE LICENSEES' RECEIPT OF TESTING DOCIBENTATION TO SUPPORT THE LIMITS OF LIFE RECOMENDED BY 'IEE SUPPLIER.
- 6. LICENSEES AND APPLICANTS NEED ONLY TO SUEMIT EOR STAFF REVIEN THE EQUIINENT CLASSIFICATION PROGRAM FOR SAFETY-REIATED OCMPONENTS. ALTHOUGH NOT REQUIRED TO BE SUEMITTED EUR STAFF REVIEN, TOUR EQUIPMENT CIASSIFICATION PROGRAM SHOUID AISO INCLUDE THE BROADER CLASS OF SIEUCIURES, SYSTEMS, AND CI24PONENTS IMIORTANT 'IO SAFETY REQUIRED BY GDC-1 (DEFINED IN 10 CFR PART 50, APPENDIX A, " GENERAL DESIGN CRITERIA, INISODUCTION").
Status:.
- 1. Classifying cmponents as safety-related requires an engineering determinatior. (usually by the Architect Engineer) based on the following:
A safety-related structure, system, or cmponent is one that is required for safe shutdown of the reactor, post design basis accident (DBA) shutdown core cooling, or that contribute l to the prevention of or mitigation of the consequences of
! accidents which could cause undue risk to health and safety of the public.
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1 FEB 2 91s84
o 2. The muipnent Iocation Index (ELI) is used to doctnent and control quality rs;uirenents for plant systens and cmponents. %e Unit 1 ELI is maintained by Southern Cmpany Services; the Unit 2 ELI is maintained by Bechtel. Both cmpanies have procedures for validation and use of the ELI.
- 3. The ELI is used by the Plant Hatch Quality Control department to identify quality requirenents for Maintenance Rquests (MRs) and Purchase Rs;uisitions. Procedures for controlling activities described in 10Cm50, Appendix B and apply to safety related cmponents are the Plant Hatch administrativo procedures . The adninistrative procedures contain 157 procedures related to plant activities including controls for design change, maintenance, procedure use ri control, procurenent, quality assurance and quality control,
- 4. %e Plant Hatch GPC corporate Quality Assurance Departments conduct periodic audits covering all aspects of plant activities including requirenents for procedural cmpliance. Use of the ELI at Plant Hatch and architect engineering cmpanies is controlled by procedures which are subject to audits by the Plant Hatch and GPC corporate QA organizations.
- 5. Procurenent of safety related cmponents is controlled by procedures for material and services procurenent and ratuisition review for quality requirenents. Procedures require that appropriate design verification and qualification testing are specified for procurenent of safety related cm.ponents. This is accmplished by the QC department by reference to information in the ELI, procurenent standards, GE and Architect Engineer equipnent lists, purchase specifications, and applicable codes and standards. Requisitions for which this information is not available are sent to the corporate headquarters for determination of quality requirenents.
G. Georgia Power has a quality assurance progran which includes equipnent not classified as " safety related" but which we have determined is necescary for safe or reliable operacion. However, we do not recognize a separate category of equipnent called "important to safety". We endorse the position taken by the AIF Canittee on Safety Classification that the ' terms " safety related" and "important to safety" as currently used in the regulations are synonymous and interchangeable, and apply to squipnent described in 10 CFR 50, Appendix A. Implenentation of a third formal category of equipnent would represent imposition of a previously unimposed requirenent. Should the NRC wish to impose such a requirenent, it should be done through the regulatory process so that the additional burden on the industry and the NRC staff is properly evaluated. We recognize that there is confusion within 'the industry and the NRC related to the use of the terms, but we do not feel that there is any safety problen resulting frm the lack of consensus on the issue.
FEB g g ggy
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f Plans and schedules for cmpliance:
- 1. Criteria for identifying cmponents as safety-related meet applicable requirenents. No enhancenent is required.
- 2. Information handling. for maintenance, purchasing, inventory !
control, and surveillance requirments will be improved by implenentation of a planned. c a puterized data base managenent system. %e systen was under developnene prior to the issuance of
, Generic T.etter 83-28, and the scope of the systen goes .beyond our i
interpretation of the minimtzn requirenents of the Generic Letter.
However, the system specification will address rquirenents resulting from the Salen events. %e system is scheduled for operation prior to January 1, 1986.
- 3. Plant Hatch procedures for use of the ELI meet the positions of the Generic Letter. Procedures developed for users of the data base managenent cmputer systen will include existing re uirements and any ne .equirenents resulting from the Salem events.
- 4. %e Plant Hatch and GPC corporate @ staffs are responsible for ,
t verifying that procedures for controlling quality requirements are !
being followed. Future audits will include verification'of correct performance of activities related to positions of the Salen letter. While @ is an on going activity, an audit which includes consideration of the Salen rquirenents will be empleted by March 1, 1985.
I
- 5. Hatch procedures for procurement of safety related cmponents meet the positions of the Generic Letter. Procurenent procedures are currently being revised as a result of a GPC initiative taken prior to the issuance of Generic Ietter 83-28. %e procedure revisions are expected to improve our staff's efficiency and to provide additional assurance that quality requirenents are met. Revisions are expected to be incorporated in the Hatch procedures prior to Deceber, 1984. Procedures for monitoring limits of life for equipnent are expected to be implenented by April 1, 1984. %e l procedures were developed as part of a program to meet the l requirenents of I&E Bulletin 79-01B. !
- 6. Georgia Power Capary recognizes that the definition and use of the term "important to safety" is an unresolved issue and will continue to monitor industry and NRC positions. No change in our current definitions are planned.
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FEB 2 91984
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e 2.2.2 FOR VENDOR INTERFAG, LICENSEES AND APPLICANTS SHALL ESTABLISH, IMPLEMENT AND MAINTAIN A CONTINUING PROGPJM 'IO ENSURE 'IHAT VENDOR ,
INEOPMATION FOR SAFEIT-REIATED OCNPONENIS IS (XMPLEIE, CURRENT AND CONIROLIED THROUGiKX7T THE LIFE OF 'INEIR PIANTS, AND APPROPRIATELY REFERENCED OR INCDRPORA'11!D IN PIANT INSTRUCTIONS AND PROCEDURES.
Status:
GPC participated in the INPO Nuclear Utility Task Action .. Ccanittee
- (NUIRC) on Generic,Intter 83-28, Section 2.2.2. His carnittee has developed and approved an industry-wide Vendor Etluipnent Technical Infomation Progran (VETIP), which is described in detail in Attachment 2 of this sutmittal. His progran prmotes interaction among the major organizations involved in the generation of canercial nuclear power.
As illustrated in Figure 1 of attactznent 2, individual utilities exchange information related to safety related systens and components with vendors, the NRC, INPO and other utilities. %is exchange of infomation takes place via written notification (ie. Licensee Event Reports, NRC I&E Bulletins and Information Notices, industry newsletters, etc.) as well as industry meetings and day to day verbal cmsnunications. We purpose of these information exchanges is to share equipnent technical information to improve the safety and reliability of t " nuclear generatirxJ stations. The primary purpose of the VETIP program ,
i
-is to ensure that current information will be made available to those personnel responsible for developing and maintaining plant ' instructions and procedures. %ese information programs and systens currently exist ,
and are capable' of identifying to the industry precursors that could i lead to a Salen-type event. %e VETIP progran is industry controlled L and does not rely on vandor action to provide information to utilities, i' Instead, the VETIP provides information developed through industry
- experience' and analyzed in Significant Event Reports (SERs) and ,
i
'Significant Operating Experience Reports (SOERs). Vendors - becme involved by review of SERs and SCERs before they are circulated to the utilities.
Plans and schedules for empliance:
l 2e administrative program and procedures now in place at Plant Hatch will be reviewed to confirm that they implenent the reccanended guidelines presented in the NUTAC report. His review is scheduled to be cmpleted and any appropriate changes are scheduled to be made prior to January 1, 1985. Ebll implenentation of the Nuclest Plant Reliability Data Systen '(NPRDS) at Plant Hatch is scheduled for Decenber
-1984. ,
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3.1 POST MAINTENANCE TESTING (RFACIOR TRIP SYSTEM OCMPONENIS) 1 POSITION THE EOLIONING ACTIONS ARE APPLICABIE TO POST-MAINIENANCE TESTING:
1.. . LICENSEES AND APPLICANTS SHALL SUINIT THE RESULTS OF TH8IR REVIEW OF TEST AND MAINTENANCE PR03DURES AND TECHNICAL SPECIFICATIONS TO ASSURE THAT POST-MAINIENANCE OPERABILITY TESTING OF SAEETY-REIATED CI:MIONENTS IN THE REACIOR TRIP SYSTEM IS REQUIRED TO BE CI)NDUCITD AND THAT TEE TESTING DEMONSTRATES THAT THE EQUIINENT IS CAPABLE OF j
PEREOR4ING ITS SAFEIT EVNCTIONS BEFORE BEING RETURNED TO SERVICE.
- 2. LICENSEES Ato APPLICANTS SHALL SUEMIT TEE RESULTS OF THEIR CHECK OF VENDOR AND FNGINEERING REOCMENDATIONS 70 ENSURE THAT ANY APPROPRIATE TEST GUIDANCE IS INCLUDED IN TEE TEST AND MAINIENANCE ~
PROCEDURES OR THE TECHNICAL SPECIFICATIONS, WHERE REQUIRED.
- 3. LICENSEES AIO APPLICANTS SHALL IDENTIFY, IF APPLICABLE, ANY POST-NAINTENANCE 7EST. REQUIREMENIS ~IN EXISTING TECHNICAC SPECIFICATIONS WHICH CAN BE DE40NSTRATED TO DEGRADE RA7HER THAN ENHANCE SAFETY. APPROPRIATE CHANGES TO THESE TEST REQUIREMENIS, WITH SUPPORTING JUSTIFICATION, SHALL BE SUENITTED EOR STAFF APPROVAL.
Status l 1. We have initiated a review of maintenance procedures to assure that L post-maintenance operability testing is adquate. and required for j all components needed to trip the reactor. This review cannot be
- l. adequately performed until after implementation of the Analog Trip Tranamitter Systen (ATTS) is canplete after the next outage on each unit. The response to Section 4 of the Generic Ietter contains a discussion of current Tech Spec requirements for testing of components required for a reactor trip.
- 2. 7te review described in (1) above will include a check of available vendor and engineering reconnendations to verify that they are included or referenced in appropriate procedures.
- 3. Item 3.1.3 has been referred to the Technical Specification Consnittee of the BNR Owners group to pursue on a generic basis for l BWRs. GPC will participate on this cannittee and pursue any l appropriate improvements to the Hatch Technical Specifications.
l FEB 2 9 584
Plans and Schedules for Cmpliance:
1 & 2. Due to changes resulting frm ATIS installation, a report on the results of a procedure review cannot be empleted in the near term. A report describing.the results of our review of maintenance procedures for systes needed to trip the reactor will be subnitted in August,1986.
- 3. 'Ihe BWROG Technical Specification Improvm ent Comnittee is expected to be a long term activity which will eventually address all aspects of plant Technical Specifications. ne progran dealing with this section of the Salem Requirements is expected to be emplete within two years.
3.2 POST-MAINIENANCE TESTING (ALL OTHER SAFETY-REIATED C3HOETIS)
POSITIO.N_
'IEE EOLIOWING ACTIONS ARE APPLICABLE 'IO POST-MAINIDIANCE TESTING)
- 1. LICENSEES AND APPLICANTS SHALL SUINIT A PEPORT DOCINENTING 'IEE EXTENDING OF 'IEST AND MAINTENANCE PROCEDURES AND TECHNICAL SPECIFICATIONS REVIEW 'IO ASSURE 'IHAT POST-MAINIENANCE ODERABILITY
'IESTING OF ALL SAFETY-REIATED EQUIINENT IS REQUIRED 'IO BE CONDUCIED AND THAT THE TESTING DEMONSTRATES THAT 'IEE EQUIINENT IS CAPABE OF PEREOINING ITS SAFETY FUNCTIONS BEFORE BEING RETJRNED 'IO SERVICE.
- 2. LICENSEES AND APPLICARTS SHALL SUIMIT 'IEE RESULTS OF THEIR CHECK OF VENDOR AND ENGINEERING RECONENDATIONS 'IO ENSURE 'INAT ANY l
l APPROPRIATE 'IEST GUIDANCE IS INCIDDED IN 'INE TEST AND MAINTENANCE PfQ.x.ujRES OR 1EE 'IECHNICAL SPECIFICATIONS WHERE REQUIRED.
- 3. LICENSEIS PND APPLICARTS SHALL IDENTIFY, IF APPLICABLE, ANY POST-MAINIENANCE 'IEST REQUIREMENIS IN EXISTING 'IECHNICAL SPECIFICATIONS WHICH ARE PEDCEIVED 'IO DEGRADE RATHER 'IHAN ENHANCE CAFETY. APPROPRIA'IE CHANGES 'IO THESE TEST REQUIREMENTS, WITH SUP. SORTING JUSTIFICATION, SHALL BE SUIMITTED EUR STAFF APPROVAL.
Status:
1 & 2. Response to position 3.2 is essentially the sane as the response provided in Section 3.1. Where practical, a higher
- i. priority will be given to review of procedures related to i
caponents required to trip the reactor
- 3. GPC has identified testing requirments- for the diesel generators which were degrading their reliability. Proposed Tech Spec changes are in the final stage of the review process and are scheduled for NRC subnittal in March,1984. @e BWR Owners Group has initiated a Tech Spec Dnprovenent progran. GPC is a participant in the effort.
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Plans and Schedules for Canpliance:
1 & 2.- A report describing the results of our review of maintenance procedures for all safety related equipnent will be sutaitted along with the Section 3.1 response in August, 1986.
- 3. - Diesel Generator Technical Specification improvenents are scheduled for NRC subnittal in March, 1984. GPC will continue to participate in the BWROG generic effort and implenent any appropriate revisions the the Plant Hatch Technical Specifications.
4.1 RFACIOR 'IRIP SYS'1TM RFLIABILITY (VENDOR REIA'IED MCDIFICATIONS) i 4.2' REACIOR 'IRIP SYSTEM RELIABILITY (2RIEVENTIVE MAINIENANCE AND SURVEILIANCE PROGRAM POR REACIOR TRIP BREAKERS) -
4.3 REACIOR 'IRIP SYSTEM RELIABILITY (AUT04ATIC ACTUATION OF SHUNT 'IRIP ATTACINENT FOR WESTINGHOUSE AND B & W PIANIS) 4.4 REACIOR TRIP SYSTIM RELIABILITY (IMPROVEMENIS IN MAINTENANCE AND 'IEST PROCEDURES FOR B & W PIANIS)
Status:
Positions 4.1 through 4.4 do not . apply to a BWR. Accordingly, no response to these sections is required for Plant Hatch.
4.5 REACIOR 'IRIP SYSTEM RELIABILITY (SYSTTM EUNCTIONAL 'IESTING)
. POSITION ON-LINE FUNCTIONAL TESTING OF THE REACIOR 1 RIP SYSTEM, INC[UDING IlOEPEREENT TESTING OF THE DIVERSE TRIP FEATURES, SHAIL BE PERPOINED ON ALL PIANTS.
. Status:
Our response of Novenber 7 stated that a report would be provided with our February 29 sutaittal. The following discussion of reactor trip systen functional testing provides the information requested in position 4.5.
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REAC10R TRIP SYSTEM MJNCTIONAL TESTING Generic Ietter 83-28 requests that licensees of operating reactors review their plants' operation in areas related to reactor trip systen reliability.
Specifically, " action 4.5" of the letter requests that operating reactors' reactor trip systs be reviewed to ensure that proper on-line functional testing ic perfomed. Unless otherwise noted, the following discussion applies to both Unit 1 and Unit 2.
'Ihe Plant Hatch Reactor Protection Systen (RPS) design emplies with all applicable regulatory rquirenents for the reactor protection systen.
A review of the Plant Hatch RPS on-line functional testing and testing intervals was performed and found to be consistent with achieving a high scram system reliability. The following is a cromary of the on-line functional testing and testing intervals performed on the RPS.
On-lino channel functional testing of multiple and diverse reactor trip sensors is performed monthly. Average Power Range Monitor and Intermediate Range Monitor Reactor trip signals are channel functionally tested prior to reactor start-up and weekly thereafter. The multiple and diverse Scram Discharge Voltrae (SDV) high water level trips are functionally tested monthly. Each scram contactor which actuates the scram pilot solenoid valves is tested during each trip sensor channel test. @e simple operation of the scram contactors minimizes concerns of wear, end frequent testing assures that any failures are detected early. %e Scram Pilot Solenoid Valves which are actuated by the serm contactors are tested during rod insertion time testing on a testing frquency shc m in the table on page 18 of this report. Redundant Electrical Protection Assenblies (EPAs) protect the Scram Pilot Solenoid Valves fran low voltage chattering and the associated potential consequence of accelerated wear.
Surveillance testing requirenents related to the Scram Pilot Solenoid valves assure that the probability of undetected failures of these independently acting solenoid valves is small.- In simnary, the current Reactor Protection System on-line surveillance testing requirenents, in conjunction with multiple and diverse scram sensors, assure that the probability of failure of enough control rods to prevent reactor shutdown is negligible.
14 -
FEB 2 9 G84
The following evaluation further discusses the functional testing and reliability of the Plant Hatch Reactor Protection Systs.
REACIOR PROIECTION SYSTEM (RPS) TESTING AND RELIABILITY Introduction RPS and CRD Systs test intervals have been developed in the Plant Hatch Technical Specifications to provide early identification of cmponent failures during stand-by operation and to ensure that any indication of systmatic probles will be identified and corrective action initiated on a timely basis. By identifying cm ponent failures and any systmatic probles early in reactor operation, corrective actions can be taken to ensure that systec achieve and maintain high scre syste reliability.
'Ihe purpose of this evaluation is to review the current Plant Hatch on-line functional testing for the RPS and CRD systes rs{uired by Technical Specifications.to verify that testing intervals are consistent with achieving high scra system reliability. It is concluded frm this evaluation that the current Plant Hatch on-line functional testing intervals for the RPS and CRD system are consistent with achieving a high scram syst a reliability.
RPS INSTREMENTATION SURVEILIANCE RDDUIREMENIS Channel Ebnctional Test Channel funtional tests are performed monthly for the following sensor trips:
o Reactor Vessel Dme Pressure-High o Reactor Vessel Water Level-Iow o Main Stem Line Isolation Valve-Closure o Main Steam Line Radiation-High o Drywell Pressure-High o Turbine Control Valve Fast Closure, Control Oil Pressure-Iow o 'Ibrbine Stop Valve-Closure o Scran Diccharge Volme High Water Level Channel functional tests are performed prior to startup, and weekly thereafter for Main Stem Line Radiation-High, Average Power Range Monitors and Intermediate Range Monitors.
FEB 2 9 8
In reference 1 and 2, it is shown that each of the above plant variables used to initiate a protective function is backed up by a cmpletely different plant variable. In fact, it can be seen frm Table 1 that for the most frequent transients, scre is initiated by three diverse sensors in all but one case (regulator failure-primary pressure increase which is initiated by two diverse sensors). Ihis indicates that adequate redundancy exists in the design to provide protection against multiple independent sensor failures. Also, diversity mong sensor types reduces the potential for cmmon cause failures, failures due to hwan error, and increases in failure rate due to wearout. A pictorial representation of the RPS logic configuration with the frquency of channel functional tests is provided in Figure 1.
Each sensor channel functional test includes full actuation of the associated logic, the two output serm contactors in each channel, and the individual CRD scram air pilot valve solenoids for the associated logic division (solenoids frm both logic Division A and B required for scram initiation) . Each individual channel output contactor is tested during each sensor channel functional test.
%e most credible failures within the RPS logic will de-energize a set of scram solenoids which causes a half scrm, i.e., one of the two scram solenoids required for scram initiation is de-energized at sme or all hydraulic control units.
%ese failures would be " SAFE" failures that would increase the probability of plant shutdown.
%e less credible logic failures which prevent a channel frm de-energizing . will be detected during channel functional tests in cmpliance with Technical Specification requirments. The frequency of tests described above ensures that an increase in failure rate due to a wearout condition or a canon cause failure potential could be detected early and corrective action taken before the failure condition becmes systmatic.
Backup Scrm Ingic In addition to the primary RPS logic, a backup scram logic is provided in the RPS. %e backup scram function is accmplished by two air operated solenoid valves which isolate the main air supply and vent the air supply header which connnects to the individual Hydraulic Control L
Units. The backup scrm valves are redundant valves with redundant trip
! signals frm both RPS logic A and B. %e logic is diverse fra the primary RPS since the backup scrm valve solenoids are energized and D.C. powered to trip versus the primary scraa pilot valves which are de-energized to trip and A.C. powered. Although one-half of the backup scrm logic is actuated for each valve during channel functional tests, the only time the backup scram solenoids are actuated is when a cmplete scram signal is initiated.
FEB 29 8
On-line testing of the backup scram function is not appropriate for the following reasons:
- 1. The backup scram function was incorporated as an additional improvment in response to an already extreely remote event and is not required by applicable regulatory requirments.
- 2. .%e backup scram function has been designed to be highly reliable by use of redundant valves and actuating logic,
- 3. Testing during operation would require either a plant scram or plant modifications and test procedures that have a potential for human error or failures which could result in unnessary shutdowns.
- 4. We primary scra pilot valve solenoids which are normally energized and tested frequently are diverse to the backup scre solenoids which are normally de-energized and not cycled frequently. Due. to the lower operating frequency of the backup scrm valves, the potential for a canon cause or hwan error affecting both the primary and backup RPS is reduced.
Other Channel Functional Tests Other channel functional tests include quarterly testing of the manual scram trip and test of the reactor made switch in the shutdown position every refueling. % e first involves on-line testing and the latter mode switch test can only be conducted during reactor shutdown. %e manual scrm trip can be tested on-line without creating a scram. The testing frequency for this trip is considered adequate based on the automatic trips and alternate means of manually scramming the reactor.
We monthly testing of the SUV Water Level-High trip is considered adequate based on the current designed redundancy and diversity incorporated into the system. %ere are two diverse and redundant sets of level sensors which scram the reactor in the unlikely event of high water level in the SDV during power operation. %ese trips are designed to allow sufficient scram water discharge volme given the scram trip point is reached.
Scram Insertion Tests
%e following tests of the scram insertion times are required by the Plant Hatch Technical Specifications:
g3 2 919M I
Ntznber of Control Rods Frtpuency ;
- 1) 100% Each Refueling (Unit 1) after core alterations or reactor shutdowns of greater than 120 days (Unit 2) .
- 2) Specific Rods After maintenance or modifications affecting scram insertion times (Unit 2 only)
- 3) 10% of Control Rods 16 weeks (Unit 1) on a Rotating Basis 120 Days (Unit 2)
Reference 2 concluded that reactor shutdown can be achieved if at least 50% if the control rods in checkerboard pattern and 69% in a rands pattern are inserted in the core. The probability of independent failure of enough rods to prevent shutdown is negligible. The most likely type of failure would be sme comnon cause mechanism that if undetected over a long period of time could cause an unsafe condition.
The surveillance requirments given above adquately ensure that a failure mechanism affecting several individual drives (considered to be very remote) would not go undetected. One of the major features that ensures that several drives do not fail at one time due to wearout or a ccanon cause is' the staggered maintenance and overhaul of selected degraded CRDs or Hydraulic Control Units (HCDs) at refueling outages.
This ensures a mix of drives by age, cmponent lot, maintenance time and servicing personnel, and testing.
The scram insertion time tests include, in addition to drive timing and insertion capability, a test of operability of the HCU scre insert and discharge valves including associated scram air pilot valves. As stated in the previous paragraph, the required frequency of testing given in the Technical Specification ensures that a systematic failure mechanism in the HCUs would be detected early and corrective action taken before the condition becmes a critical failure preventing scram.
References
- 1. NEDO-1-189, "An Analysis of Ebnctional Comon-Mode Failures in GE BWR Protection and Control Instrmentation," L. G. Frederick, eta al., July 1970.
- 2. "BWR Scra System Reliability Analysis," W. P. Sullivan, et al.,
Septenber 30,1976 (Transmitted in letter frm E. A. Hughes (GE) to D. F. Poss (NRC) , " General Electric Cmpany AIWS Reliability Report," September 30, 1976).
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p %= mary Schedule for Salen Generic Requirements
, e o Generic Letter 83-28 Position Action Schedule Date io G
T. 1.1 Post Trip Review Procedure revision Complete 9/83 1.2 Data and Information Capability Report on existing capability Complete 11/83 for Post Trip Review Evaluation of existing capability Complete 2/84 Analog Trip System 12/84 Safety Parameter Display System 6/86 2.1 Equipment Classification and ELI review and enhancement 12/85 VeniorInterface for RTS Components Report on vendor interface program Complete 2/84 Database management computer 1/86 NPRDS implementation 12/84 4
] 2.2 Equipment Classification for 1 Safety Related Components i
- 1. Classification criteria Provide description Complete 2/84
- 2. Information system " " Complete 2/84
- 3. Procedures for quality " " Complete 2/84 i requirements '
- 4. Management controls " " Complete 2/84 i QA audit 2/85 i
- 5. Procurement requirements Provide description Complete 2/84 Procedures for monitoring " " Complete 2/84 equipment life l
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, Summary Schedule for Salen Generic Requirements (Continued) w Generic Letter 83-28 Pbsition Action Schedule Date -
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,. g 2.2.2 Vendor Interface for Safety Report on General Electric vendor interface Complete 2/84 Related Components Implementation of hPRDS and INPO VETIP . 1/85 1
infomation program 3 Post Maintenance Testing for .
RTS Components
- 1. Procedure adequacy Procedure review 6/86
- 2. Vendor infomation Check of vendor infomation 6/86-
- 3. Tech Specs Report on BWROG activity Complete 2/84.
3.2 Post Maintenance Testing for '
Safety Related Components, .
- 1. Procedure adequacy Procedure review 6/86
- 2. Vendor infomation Check of vendor infomation 6/86 ,
- 3. Tech Specs Report on BWROG activity Complete 2/84 4.1 through 4.4, Scram Breaker PWR related modifications - no action required N/A Modifications for Plant Hatch I
) 4.5 RTS Reliability, System Report of current testing requirements and Complete 2/84 4
Functional Testing evaluatiot of adequacy PLS 2/28/84 t