ML20236V815
| ML20236V815 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 12/01/1987 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20235V135 | List: |
| References | |
| FOIA-87-644 GL-83-28, NUDOCS 8712070113 | |
| Download: ML20236V815 (10) | |
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SAFETY EVALUATION REPORT GENERIC LETTER 83-28, ITEM 2.1 (PART 1) EQUIPMENT CLASSIFICATION (RTS COMPONENTS) i HOPE CREEK NUCLEAR GENERATING STATION I DOCKET NO. 50-354 INTRODUCTION AND SUPNARY On February 25,1983, both of the scram circuit breakers at Unit 1 of the Salem l Nuclear Power Plant failed to open upon an automatic reactor trip signal from l the reactor protection system. The incident was terminated manually by the. operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior to this incident, on February 22 - 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was In generated based on steam generator low-low level during plant start-up. this case, the reactor was tripped manually by the operator almost coin-cidentally with the automatic trip. I i Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the staff to investigate and report on the generic l implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem I unit incidents are reported in NUREG-1000. " Generic Implications of the ATWS Events at the Salem Huclear Power Plant." As a result of this investigation, the Comission (NRC) requested (by Generic Letter 83-28 dated July 8,1983I) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the l analyses of these two ATWS events. 8712070113 871201 PDR FOIA SDR GI B7-644 PDR,
4 l i This report is an evaluation of the response submitted by Public Service Electric and Gas Company, the applicant for the Hope Creek Nuclear Gbner- J ating Station, for Item 2.1 (Part 1) of Generic Letter 83-28. The actual fl doctinents reviewed as part of this evaluation are listed in the references ] at the end of the report. ) l Item 2.1 (Part 1) requires the applicant to confinn that all Reactor Trip f System components are identified, classified and treated as safety-related [ l as indicated in the following statement: ) i Licensees and applicants shall confirm that all components whose functioning is required to trip the reactor are identified as- I saiety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, in-cluding maintenance, work orders, and parts replacement. I i i EVALUATION _ The applicant for The Hope Creek Nuclear Generating Station, responded to the 2 requirements of Item 2.1 (Part 1) with submittals dated March 30, 1984 , 4 December 17, 1984 and May 21, 1985 . In these submittals the applicant des-cribed their plan to develop a Master Equipment List (MEL) that will meet these , requirements, reviewed progress made, and in the last submittal stated that all components that are required to perform the reactor trip function were reviewed i The to verify that these components are classified as safety-related equipment. applicant further stated that while the MEL had not been completed for all components of the reactor trip systes, this effort would be completed by September 30, 1985.
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CONCLUSION Based on our review of these responses, we find the licensee's statements - b. confirm that a p*ogram exists for identifying, classifying and treating J 4 components that are required for perfonnance.of the reacter trip function as,, ' safety related. This program meets the' requirements of Item'2.1 (Part 1) ofs,. the Generic Letter 83-28, and is therefore acceptable. I REFERENCES , ,
- 1. NRC Letter D. G. Eisenhut to all Licensees of Operating Paactors. -
~ #
Applicants for Operating License, and Holders of ConstrucMon Permits.
" Required A tions Based on Generic Implications of Salem ATWS Events D l
(Generic Letter 83-28)," July 8,1983.
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- 2. Letter, R. L. Mitt 1, Public Service Electric and Sas Co., to A'. Schwencer, ff l
NRC, March 30, 1984 R
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- 3. Letter, R. L. Mitt 1, Public Service Electric and Gas Co., to A.. Schwttcer[o t NRC, Decerter 17,1984. !
'l 4 Letter, R. L. Mitt 1, Public Service Electric' and Gas Co., to W. Butler, i l
NRC, May 21, 1985. q q i i r d' H c j
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l 41FULTON Robject: MINUTES CF DE'CRMAL MEETING WIS PRC 70 D28 CUSS ASS.
't9CR SPEC SAMS, BETHESQh, MD, APRIL 1) -1987 L i .)
Enclosures:
- 1) List of Atten6tw - 1 1
- 2) Presentation Material i ' ['
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This letter sunnarises the discussions < held with the IEC during' an inforenal meeting in BethesdQ ND on April 1,1987. The. purpose,of.the meeting was to discuss tiijypes of changes that should be incorW into th(Technical gecifientionitywhen isr$muunting plant modifications to ocorasudate the EE5 rule Il0CPR50.62). . ( ' 1 7 ,. y :, . t , j lb ' l; .f:ONh
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.i The initiki divmaafan fracused cm the Ted esse changes 'associabei with - C'-
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; implementors alternstrered inrArtion -(ARI). 'Psar plasts[with/ARI al% ]
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; danges wonM be reMre$. For plants whi&lamTently de act have. ARI i comed in Tech speas, t}A information. to be included abould be'similar '!
$ to Um Ntandard ted: Specs for Ans Recirculation Pump Tdp (RPf), l sooounting for the differences n between ARI and = RET es apprtr;riate. - f Also, while not neowmary to be covered in Sech specs, tho'Isc wants to i c .
have covered in .the ATWB- dar= =ntatite autenittals ' reptired by .i 10CFR50.62(c)(6) hay administrate'Jely, ARI valve operability will be h pettiodically demsrstrated. > ( 4 ir f j.> .g l/- 6 9
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The discussion on Standby Liquid Control System'(St4S) modifications was covered in two pazts, two-pump operation and enriched boron.' For both approaches, the NRC concurred that changes to the SI4 system made . to meet the MWS rule most not invalidate the original design bases. For the_ two-punp. operation, the following Tech Spec changes should be considered: -{
- 1) Revise SIC punp discharge relief valve setpoint pressure if required.
- 2. '
Revise co. centration requirements if necessary 'to account for minimum two-punp, flow rate. Further, it.was decided that current periodic testing requirements. are' adequate to assure system capability. . However, the 10C does want.to see evidence as a part of the MWS documentation submittal to demonstrate the adequacy of two-punp operation (e.g. two-punp performance test). For the enriched boron option, .it was decided 'that1 current Volume, temperature, ocncentration, and flow rate Tech 8pec surveillance' re-quiruments should be retained. For senitoring of . the . 210 enrichment lowl, two" separate approaches 'were discussed, depending upon whether the enriched sodium pentaborate was formulated at the chemical vendor's facility, or at the BWR plant site. For 'either approach, interim procedures should be' developed to assure that the presently mixed non-enriched sodium pentaborate is properly disposed of. For plants utilising sodium pentaborate formulated l at. the chemical vendor, an enrichment test should be done as c part of the inecning veterial evaluation (administrative pah). '! hen, at the beginning. of each cycle, the enrichment of the boron material 'in the SI4 tank should be analyzed to assure proper shutdown capability. ( _ . _ - . - - - - "
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For plants which plan to utilize enriched boric acid and mix batches of . ~ sodium pentaborate on site, an enrichment test should be performed each ~ time new chemicals are added to the tank.-In addition to the once-per-cycle test for pre-mixed plants 'and the test each ti:se chemicals are - mixed at l other plants, the following would need to be includedMin enriched boron technical specifications: {
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- 1) !
A 30-day tim period is allowed for receiving the test results back i which determine the enrichment of the sodium pentaborate. .
- 2) ;
Once the enrichment is determined, an evaluation should be; j made to assure that both the original design requirement and i the new A M requirement are met, 3)' If the original design requirmnent is not mt (inndequate. i 10 total amount of B to achieve cold shutdown) proceed to the i j existing '!*cb Spec operability requirement. 4) If the ATWS rule is not mt, a period of 7 days 'is allowed to - bring the Baron enrichment into cu pliance. '
- 5) i If at the.end of the 7-day period, coupliance cannot be assured, !
then within 7' days, the NRC mat be notified, and plans to j I bring the material into conpliance with tlw A2WS rule met be J submitted to the NRC. t i For boron flow testing procedures which include provisions for saving and recycling 'the enriched material trapped in .the SLC linea during testing, enrictment checks are not ru,uired when this materi'l a is recycled to the SIC tank. l Please feel free to contact GI Samstad, (408) 925-6229, or me if- you have any questions or coninents regarding these meeting minutes. ards, , ickens, Chairmn BWROG A'IMS Conpliance Alternatives Comittee (612) 337-2037 cc: RP Janecek, BHROG Vice Chairman Meeting Attendeec e i
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- M JRE ONE MEETING ATTENDEES:
M.'W. Hodges,.NRC l T. E. Collins, NRC H. C. Li, HRC J. L. Hauck, NRC T. A. Pickens, Chairman, A'NS.Ctspliance Altematives Comittee B. Teletski, CP&L. R. T. Earle, GE G. I. Samstad, GE l 4 6 f
n ,ce s w a : x L c u s:4 ' g J PROPOSED GENERIC ATWS TECH SPECE f a MODIFICATIONS OVERALL o CHAfiGES TO SLC SYSTEM TO' MEET ATWS RULE MUST NOT. d INVALIDATE THE ORIGINAL DESIGN BASIS
- TWO-PUMP OPERATION o u REVISE SLC PUMP DISCHARGE RELIEF VALVE SETPOINT PRE1 IF REQUIRED-o REVISE CONCENTRATION REQUIREMENTS IF NECESSARY T0 ACCOUNT FOR MINIMUM TWO-PUMP FLOW RATE o SYSTEM CAPABillTY ASSURANCE l
- CURRENT PERIODIC TESTING REQUIREMENTS ADEQUATE.
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... . ;c PROPOSED GENERIC TECH SPEC '
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o RETAIN CURRENT VOLUME, TEMPERATURE, . CONCENT *' FLOW RATE TECH. SPEC. SURVEILLANCE REQU o. CHECK BORON-10 ISOTOPE ENRICHMENT
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PENTABORATE IS FORMULATED SODIUM PENTABORATE FORMULATED 'AT VENDO RECEIPT INSPECTION.ONLY'- NO TECH.LSPEC VERIFICATION REQUIREMENTS SODIUM PENTABORATE FORMULATED AT h . S SPEC, VERIFICATION-o EXIS. TING QUALITY ASSURANCE PROCEDURE " ADMINISTRATIVELY CONTROL THE QUALITY:0o AVAILABLE TO SERVICE THE TANK o l NO NEED FOR PERIODIC ENRICHMENT SURV STABLE ISOTOPE-o q OPERABILITY REQUIREMENTS IMPOSED ONLY.l DESIGNREQUIREMENTS-(NOTATWS)
.ESTABLISHMINIMUMBORON-10WEIGHTNECES >
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SUMMARY
- q THE PROPOSED CHANGE TOLTHE.SPECS. SYSTEM TECH COUPLED WITH THE Q.A PROCEDURES AND CONTROLS.:lS CO -;
ADEQUATE FOR PROVIDING FULL' ASSLRANC ) DOWN CAPABILITY IS BEING. MAINTAINED R o l_____ _ _ _ _ _ _ . . . _ _ _ _ _ _ . . _ = . __
P. ! o.. , t EGG-NTA--7188 Revised Draft
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CONFORMANCE TO GENERIC LETTER 83-28 '. ITEM 2.1 (PART 1) EQUIPMENT CLASSIFICATION (RTS COMPONENTS) SELECTED GENERAL ELECTRIC BOILING' WATER REACTOR PLANTS HOPE CREEK' PEACH BOTTOM 2 AND 3 PERRY 1 AND'2 PILGRIM 1 . R. HAROLOSEN Published September 1986 EG&G Idaho, Inc. Idaho Falls Idaho 83415 9 Prepared for.the U.S. Nuclear Regulatory Comission Washington, D. C. 20555 Under DOE Contract No. DE-AC07-161001570-FIN Nos. 06001 and 06002 t/ ' DOCKO$000354 l CF i
r' ' ABSTRACT-This EG&G Idaho,-Inc. report provides a review'of the submittals from selected operating and applicant Boilir.g Water Reactor (BWR) plants for conformance to Generic Letter 83-28 Item 2.1 (Part 1). The following plants are included in this review. Plant Name Docket Number TAC Number Hope Creek 50-354 OL , Peach Bottom 2 50-277 52865 Peach Bottom 3 50-278 52866 Perry 1 50-440 61705 . Perry 2 50-441 OL ' Ptigrim 1 50-293 52867 FOREWORD { This report is supplied as part of the program for evaluating licensee / applicant conformance to Generic Letter 83-28, ' Required Actions Based on Generic Implications of Sales ATWS Events." This work is being conducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Regulation, Division of PWR Licer. sing-A, by the' EG&G Idaho, Inc. i The U. S. Nuclear Regulatory Commission funded this work under the ! i authorization B&R 20-19-10-11-3 and 20-19-40-41-3, FIN Nos. 06001 and 06002. i I l O [
CONTENTS. ASSTRACT .............................................................. 11 FOREWORD.............................................................. 11; INTRODUCTION AND
SUMMARY
. 1. 1
. 2. PLANT. RESPONSE. EVALUATIONS ....................................... 3 2.1 Hope _ Creek'....................................., .......... 3' 2.2 Conclusion ................................................. 3' 2.3 Peach Bottom 2 and 3'....................................... 5-2.4 Conclusion ................................................. 5.
2.5 Perry 1 and.2 ..............................................- 6' 2.6 Conclusion .................................................. 6 2.7 Pilgrim 1 .................................................. 7 2.8 Conclusion ................................................. 7 l 1
- 3. GENERIC REFERENCES .....'.......................................... 8 1
l l l e 9
- 1. INTRODUCTION AND
SUMMARY
On February 25, 1983, both of the scram circuit breakers at unit 1 of. the Salem Nuchar Power Plant failed to open upon an automatic reactor trip
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signal from the reactor protection system. This inci6ent was terminated manually by the operator about 30 seconds after the' initiation of the automatic trip signal. The failure of the circuit breakers'was determined to be related to the sticking of the undervoltage trip-attachment. Prior to this incident, on February 22.1983,: an automatic trip signal was generated at Unit 1 of the Salem Nuclear Power Plant based on steam In this case, the reactor generator low-low level during plant startup.- was tripped manually by the operator almost coincidentally with the . , automatic trip. , Following these incidents, on February 28,1983. ' the NRC Executive Director for Operations (E00), directed the staff to investigate and report on the generic impitcations of the occurrences at Unit 1 of .the $41em Nuclear Power Plant. The results of the staf f's . inquiry into the generic
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implications of the Salen Unit 1 incidents are reported in NWEb1000
" Generic Implications of the ATWS Events at the Sales Nuclear Power plant. " As a result of this investigation, the Commission (BRC) requested (by Generic Letter 83-28, dated July. 8,1983 ) all licensees 'of operating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two ATWS events.
This report is an evaluation of the responses submitted from a. selected' group of Boiling idater Reactors (BWRs) for' Item 2.1 (Part 1) of Generic Letter 83-28. The results of the review of four individual plant responses are
' combined and reported on in this document to enhance review efficiency.
The specific plants reviewed in this report were selected based on the-1
I convenience of review. The actual' documents which were reviewed for each' j evaluation are listed at the end of each plant evaluation. . The generic- ] documents referenced in this report are listed at the.end of the report. Part 1 of Item 2.1'of Generic Letter 83-28 requires the licensee or ') applicant to confirm that all reactor trip system components are Identified, classified, and treated as safety-related, as indicat'ed in the following statement: 1 Licensees and applicants shall confirm that all components whose functioning is required to trip the. reactor are identified as safety-related on documents, procedures, and information handling-systems used in-the plant to control safety-related activities,. ~ including maintenance, work orders, and parts replacement. l l i 1
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- 2. ' PLANT RESPONSE EVALUATIONS 2.1 Hope Creek 50-354 (0L)
The applicant for Hope Creek (Public Service Electric and-Gas
- Company) provided responses to the requirements of Item 2.1'(Part 1) of Generic Letter 83-28 in submittals dated March 30,1984, 8ecember 17. 1984 ,
and May 21, 1985. In the first submittal the applicant sescribed their ) plan to develop a Master Equipment List (MEL) which would identify.the . components required to initiate reactor trip and designate these components- . as safety-related. The MEL imposes quality. assurance requirements for the saf6ty-related components and is the controlling document for ., j safety-related activities. The applicant stated intentisms- to be in . compliance with Item 2.1 (Part 1) prior to September 1984. The second submittal reviewed progress to December 17, 1984 and j outlined a revised program which would meet ' the requirements of Item 2.1 (Part 1) prior to March 1985.. The applicant confirmed is their May 21, I 1985 submittal that review of the reactor trip system had been completed and that reactor trip system components were verified to be classified safety-related on appropriate design documents, however, the MEL had not been completed for all components of the reactor trip system. The applicant stated that this effort would be completed by September 30,.1985. 2.2 Conclusion Based on a review of the applicant's submittals, we find that' the applicant's responses confirm that' components required to trip the reactor have been designated safety-related and that the MEL is used to control all j activities relating to safety-related components. We, theref ore, find that . l the applicant's responses meet the requirements of Iten 2.1 (Part 1) of .! Generic Letter 83-28, and are acceptable. 3
REFERENCES-
- 1. Letter, R.L. Mitt 1, Public Service Electric and Gas Co., to A.
Schwencer, NRC, March 30, 1984.
- 2. Letter, R.L. Mitti. .Public' Service Electric and. Gas Co., to A.
. Schwencer, NRC, <0ecember 17, 1984.
- 3. Letter, R.L. Mitt 1, Public Service Electric and Gas Co., to W. Butler, NRC, May 21.-1985.
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1 2.3 Peach Bottom 2 50-277 TAC NO. 52865 Peach Bottom 3 50-278 TAC NO. 52866 The licensee for Peach Bottom 2 and 3 (Philadelphia Electric Co.). provided responses to the requirements of Item 2.1 (Part 1) of. Generic Letter 83-28 in submittals dated November 4,1983, April 23,1984 and May 29,1985. The responses state that all systems that contribute to the. reactor trip f unction have been identified as saf ety-related in the current' "Q" list and that all components of safety-related systems are safety-related unless specifically excluded by safety evaluation. The 'Q" list is used to , I identify the applicable codes, standards and procedures' to be used for . activities relating to the safety-related components. 1 Each ites or service to be procured is r'eviewed to determine if it is safety-related. The review is performed by a congnirant member of the plan staf f or the Engineering and .Research Department. 2.4 Conclusion Item 2.1 (Part 1) requires licensees to confirm that all components whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance, work orders, and paris replacement. Based on the licensee's submittal we find that the list of components required to trip the reactor is incomplete. We also find that the licensee's program does not identify safety-related components on relevant plant documents. The response, therefore, does not meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 and is unacceptable.
,1
{ i REFERENCES
- 1. Letter S.L. Daltrof f, Philadelphia Electric Co.,' to D.G. Eisenhut, i NRC, November 4, 1983.- q Letter, S.L. Daltrof f, ' Philadelphia Electric Co., to D.G. 'Eisesut,. i
. 2.
NRC, April 23, 1984.
- 3. Letter, S.L. Daltrof f, Philadelphia Electric Co. - to .1.F. Stolz, NRC' May 29 1985. _
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2.5 Perry 1 50-440'(OL1 and Perry 2 50-441 (Oll-The applicant for Perry -1 and 2 (Cleveland Electric Illuminating Co.)- provided responses to the requirements of: Item 2.1 (Part 1) of Generic: Letter 83-28 in submittals dated April 6,1984 and August 28.1985.IThe' applicant reported in the first submittal that the "Q'-list for the plants.
' was undergoing review to verify.the correct classification of safety-related components. The "Q"-list is . to be used toL determine classification for maintenance, work orders and procurement. activities.
The second submittal reported that the "Q*-list evaluation had been-completed and that all numbered components from the 5 systems that . contribute to the reactor' trip function had been reviewed and classified as . safety-related or nonsaf ety-related. . The "Q"-list is the safety-related ' subset of the Perry Equipment Master Files System (PEMS) used to determine the classification for work orders, maintenance and parts procurement. t 2.6 Conclusion Based on the review of the applicant's submittals, we find that the applicant has verified that the components necessary to perform reactor trip are classified as safety-related and that' this classification program imposes saf ety-related procedures on work orders, maintenance, and. procurement activities. We, therefore, find that the applicant's response meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 and are acceptable. REFERENCES
- 1. Letter, M.R. Edelman, Cleveland Electric Illuminating Co. to' D.G.
Eisenhut, NRC, April 6, 1984.
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2 Letter, M.R. Edelman, Cleveland Eleetric Illuminating Co., to 8.J. Youngblood, August 28, 1985 7
J 2.7 Pilgrim 1. 50-253. TAC No. 52867
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1 { The itcensee for Pilgria 1 (Boston Edison Co.) provided responses to j the requirements of Ites 2.1 (Part 1) of Generic Letter 83-28 in submittals dated November 7, 1983 and June 28, 1985. In the submittals the lic'ensee confirmed that the components required to function for reactor trip are
- identified in the plant 'Q'-list ar.d are controlled at a quality level )
which reflects the safety-related functions. Documents (Purchase Orders, f f Maintenance Requests) used to control activities associated with the 1 j
"Q" listed equipment are identified as "Q" which designates the use of l
safety-related procedures. 2.8 Conclusion . Based on the review of the licensee's submittals, we find that the licensee has verified that the components necessary to perform reactor trip are classified as safety-related and that the classification program imposes safety-related procedures on maintenance and procurement activities relating to the components. We, therefore, find that the licensee's response meet the requirements of Items 2.1 (Part 1) of Generic . Letter 83-28 and are acceptable. REFERENCES .
- 1. Letter, W.D. Harrington, Boston Edison Co., to D.B. Vassallo, NRC, >
November 7, 1983.
?
- 2. Letter, W.D. Harrington, Boston Edison Co., to D.8. Vassallo. NRC, June 28, 1985.
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- 1 I
- 3. GENERIC ' REFERENCES .
- 1. Generic Implications of ATWS' Events at the Sales Nuclear Power ' Plant.
NUREG-1000, Volume 1. April 1983; . Volume 2. July 1983.
- 2. NRC Letter, D.G. Eisenhut. to all ' Licensees of 0perating Reactors, ,
Applicants for. Operating License and Molders of Construction Permits.c
'-
- Required Actions Based on Generic Impitcations of Salem ATWS Events (Generic Letter 83-28)," July 8, 1983. .
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, , UNITED STATES
.,*- .- 4 NUCLEAR REGULATORY COMM!sslON
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\, , JANUARY 8 197g j r r-Docket No. 50-293 g7f s%w d.!<jyL .
00cuMENT C0Hia Wh 1 4-Mr. J. E. Howard SM s 5 57? .l Boston Edison Company jj -- > 800 Boylston Street -ENS Boston, Massachusetts 02199 j
Dear Mr. Howard:
i d Attached for your information is a copy of NUREG 0460,' Volume 3 _ which details our current view related to ATWS. :In.this supplement a variety of options are considered regarding ATWS. We intend to select one of the ATWS options in the near future and to pursue it. to adoption. However, it is important .to note that all of the options- under serious consideration by the NP,C staff (options #2, 3, and 4 in' Volume' 3 of. NUREG 0460) regarding resolution' of the ATWS issue for BWRs require installation of an RPT. While you have committed to install a RPT on your facility, Pilgrim Nuclear Power Station,-you 'have not yetL begun to take steps toward such installation, on the grounds that you were awaiting firmer requirements by NRC. The NRC staff now has a firm position that RPT is required for your facility. Therefore, we see no bases for any further delay in implementing an RPT for your - facility. The RPT designs discussed in this letter are compatible with ATWS requirements. To expedite your installation of an approved RPT,'the-staff .is j providing a modified description (Appendix A, attached) of design i requirements which provide some additional- flexibility over those previously provided (May,1978), but which the staff has found acceptable for RPT systems to be installed in the: near future. - i j For all operating plants, the Monticello RPT design described in NEDO j 25016 and summarized in Appendix B has been accepted by the staff as meeting the Appendix A criteria. ' Sections of NEDO 25016'.related to ) ARI should be ignored as that system is not addressed by this letter. Some operating plants have already installed the "BWR/4" or " Hatch" C RPT, and the staff also accepts that design as meeting the Appendix A ' J criteria provided the changes specified in Appendix B, or equivalent changes, are incorporated. a
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,g .. .t Mr. J. E. Howard Both the Monticello design and the modified "BWR/4"~ or " Hatch" design a utilize generator field breakers which have been modified so that they .
are providef with two trip coils. One coil for. each breaker is actuated only by reactor pressure and ~ water level sensors in'RPT division A, and ~ the other coil 'is actuated by pressure and level sensors in RPT division B,- thereby providing redundancy of power supplies available to the overall system and increasing trip reliability. Either the Monticello or modified "BWR/4" or " Hatch" design, would be an acceptable RPT design provided diverse final trip relays' of .a different type are used,- or obtained from a different manufacturer than the primary scram relays used in the RPS.
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The staff has not reviewed the specific design of the time delay circuitry recently proposed for the.Monticello RPT design for low-level initiated , pump trips. We agree that time delays on the order'of 10. seconds ~ are desirable to avoid making the consequences of a postulated 1.0CA more severe, an:f we agree that.such delays of around 10 seconds have insigniff-( cant effect on ATWS consequences (for low-level initiated ATWS . pump trips l only). Therefore, we find incorporation of such circuitry on either RPT design discussed above to be acceptable, provided: l 1. The time delay is realized only for low-level initiated pump trips; I and, 1
- 2. The circuitry is incorporated in such a way that it does not signifi-cantly affect the overall reliability of the RPT; that is,- that no single failure in the timing circuit (s) can cause failure of the pump trip to occur. This could be accomplished, for example, by use of a-separate, independent timing (delay) circuit with each low-level I
sensor, or equivalent. Implementation as soon as possible of an RPT in accordance with the attached design criteria will provide an increased level of safety over the lifetime of the plant and should be installed as promptly as is reasonable. 1 ! 1
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y Mr. J. E. Howard - The staff has given careful consideration to the concern expressed by some licensees that RPT design requirements may change in the- future.. We have concluded that the design criteria outlined in this letter (Appendix A) are, for operating plants, equivalent to those enclose"f with the May,1978 letters to all BWR licensees, and we intend to effect no changes to those criteria. in the future. We believe that RPT design, procurement, and installation can be accomplished within a two year period-without requiring additional outage time beyond refueling outages. We have given consideration to steps that can be taken at present, in order to reduce the risk from ATWS events during the interim period, before recirculation pump trip circuitry and any other necessary plant modifications are completed. We have detennined that many of- the following steps are practicable and appropriate for your facility for this interim period. We therefore, request that you inform us within 90 days that you have done the following:
- 1. Developed emergency procedures to enable operators to recognize an ATWS event, including consideration of scram indicators, rod position indicators, flux monitors, vessel level and pressure indicators, relief valve and isolation valve indicators, and containment temperature, pressure, and radiation indicators.
- 2. Train operators to take actions in the event of an ATWS including consideration of manually tripping the recirculation pumps and scramming the reactor by using the manual scram buttons, changing individual rod scram switches to the scram position,- stripping the feeder breakers on the reactor protection system power distribution buses, opening the scram discharge volume drain valve, prompt actuation of the standby liquid control system, and prompt placement of the RHR in the pool cooling mode.to L reduce the severity of the containment conditions.
Early operator action as described above would provide significant. protection from those ATWS events which occur at low power levels where the rise in the vessel pressure and the containment temperature is limited to acceptable values by manual recirculation pump trip'and actuation of the existing stanaby liquid control system. If the i operator were to promptly (in a few seconds) trip the recirculation pumps to assure that the short term rise in vessel ~ pressure is not - excessive, protection will also be provided for those ATWS events where the common mode failure occurs in either the electrical portion of the scram system or in some portions of the drive system. i
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? Mr. J. E. Howard Within 90 days inform us of your schedule for implementation of your commitment to install an RPT system for your plant. Such system should conform to the acceptable systems described in this letter and your schedule should be consistent with the staff's overall objective of
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assuring that an acceptable RPT system is installed at your facility within two years. Sincerely, W Harold R. Denton, Director Office of Nuclear' Reactor Regulation
Enclosures:
- 1. NUREG 0460, Volume 3
- 2. Appendices A and B cc w/ enclosure No. 2:
see next page I ( 1 l. I l 1 i i
[ Boston Edisan Company CC Mr. Paul J. McGuire Pilgrim Station Acting Manager Boston Edison Company. RFD #1,. Rocky Hill Road Plymouth, Massachusetts 02360 Anthony Z. Roisman
. Natural Resources Defense Council 91715th Street, N. W. ~
Washington, D. C. 20005 Henry Hermann, Esquire '. Massachusetts Wildlife Federation 151 Tremont Street Boston, Massachusetts 02111 Plymouth Public Library North Street Plymouth, Massachusetts 02360 ( 4 1
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1 i APPECIX'A CRITERIA FOR HIGH PRESSURE-LOW LEVEL INITIATED . d RECIRCULATION PUMP TRIP (RPT) TO BE INSTALLED IN OPERATING BWRs BEFORE NOVEMBER 1, 1979* A. General Functional Requirement The RPT system shalizautomatically initiate the appropriate action whenever the c'onditions monitored by the system reach a ' preset level. B. Independence and Integrity The RPT system and components shall be independent and separate from components and/or systems that initiate anticipated transient (s) being analyzed and diverse from the normal scram system to minimize the probability 'of disabling the operation of the mitigating system. Diversity.can be achieved by incorporating as many of- the following methods as is practicable:
- 1. Use o' RPT final trip relays from different manufacturers (required).
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- 2. Use of energized versus de-energized trip status..
- 3. Use of AC versus DC power sources.
It shall be demonstrated that the function of the RPT system and components will not be disabled as a consequence of events being.- analyzed. Diversity of the RPT pressure and level sensing devices '(including . relays used in such sensing devices) from similar'or identical- devices used on the RPS is not required, since failure of those devices.on both i the RPT.and the RPS is not likely to cause an ATWS due to the' presence l of other diverse trips on the RPS (high flux,. valve position, etc.).
*The HRC staff has reviewed the Mo'nticello RPT design and- thel" Hatch" RPT i design, and finds that they' meet these criteria (provided the changes; specified in the cover letter are made to the'" Hatch" design). Plant' specific reviews will be conducted only-as necessary to ascertain that the plant design is the same 'as, or equivalent to, one of the' approved :
designs.
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C. Equipment Qualification The RPT system equipment and components shall be tested to verify that the system will provide, on a continuing basis, its' functional . capability under conditions relevant to postulated ATWS events, in ' cluding extremes of conditions (as applicable) relating to environ-ment, which are expected to occur in the lifetime of a plant. D. Periodic Surveillance and Preventative Maintenance Testing and Calibration-1 Periodic surveillance and preventative maintenance tests and calibra-tion requirements shall be identified to provide continuing assurance that the RPT system, including sensors 'and actuated equipment, . is' capable of functioning as' designed and that system accuracy and per formance have not deteriorated with time and usage. These requirements
- shall be particularly directed toward the. detection of those failures or degradation of accuracy and perfomance which would.not. otherwise -
be likely to be detected during the course of nomal operations. Integrated system testing shall also be perfomed to verify overall system performance. ( E. Quality Assurance A quality assurance program in conformance with the requirements of 10 CFR 50 Appendix B shall be applied to the RPT system design and equipment. F. Administrative Controls Administrative controls shall be established to control the access to all set point adjustments, calibration and test points. i G. Information Readout The RPT system shall be designed to provide the operator with accurate, complete and timely information regarding its status. For those -1 functions, including operations, test or maintenance, and calibration,. which require direct operator interaction, human engineering factors such as information displays (e.g., display formats, layout and con-trols) and functional controls (e.g... methods, location and identifi-cation) shall be included in the design. h
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- g. Maintainability
( The design-shall include measures which enhance. maintainability to reduce mean-time-to-repair and to assure the continued availability and reliability of the systen for the life.of the: plant. :The system ' < ; design shall' include features which facilitate the. recognition ~, loca-- ., tion, replacement, repair and/or adjustments of malfunctioning equipment ; and components or modules.
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p ,,JW Appendix B ' a[ Acceptable RPT Designs' 3 lW Monticello RPT Design j f
.a The Monticello design simultaneously trips both_ MG sets!"A" aNdNB" m, j
generator. field brukers upon receipt of either reactor-high' pressur;e, or low-low water.luel control . logic input signals; The-logic to; a y, i each breaker is two-out-of-two (pressure) or two-out-of-two -(level), i% a' (2/2 or 2/2), i.e., contacts "A" and "C" or contacts ."B" and "D" must, 3 close to trip the breaker. The Monticello'. design employs diversity,3 testability, separation and redundancy. y
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Modified BWR/4 or Hatch RPT Design , 9 x .. The modified "BWR/4" or " Hatch" design results -in the independent . D,' ?i ' (separate) trip of each of the two recirculation pumps' upon receipt
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o of either one reactor high pressure. signal or one. low-low water ' level , 9: .' Oe signal. The logic to each MG set "A" and "B"Lgenerator field-)toaker Wlq c , is one-out-of-two (level) or one-out-of-two (pressure)1(1/2' or' 1/2); .
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The modified "BWR/4" or. " Hatch" . design employs diversity, testability,~ separation, and redundancy. The modification to the existing " Hatch" design which makes it acceptable ' is accomplished as follows:
- 1) Add a second trip coil to'each recirculation loop's.M-G set; generator field breaker, as per the identical modification ~made to- ,1 Monticello. ,J,
- 2) Connect one of the pressure sensors- and one of the low level sensors in RPT train A to the old (existing) trip' coil in the recirculation. Y loop A M-G set . generator field breaker. Connect one of the pressure sensors and one of the low level sensors in RPT train B to theLww.
n trip coil in the recirculation. loop A' M-G ' set ' generator field breaker..
- 3) Connect the other pressure sensor and'the other low level sensor in -
RPT train A to the new trip coil in the recirculation loop B M-G' set generator field breaker. Connect the. other pressure sensor and the other low level sensor in RPT train B .to the old (existing) trip ~ C coil in the recirculation loop B M-G set generator field breaker. p
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-['i;.,,, UNITED ST ATES 5
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Dear Mr. Andognini:
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,' e The Comission today has issued the'- enclosed Confirmatory Order. for Pilgrim i ;
. Nuclear Power / Station: This '0rder confirms your' comitment, as stated i'n
-y6u'r0 13yer dd;.;d April 10, 1979; to install 'aL recirculation pump trip by-May 31, U80,'6/ in cny case, prior to' operatio'n in.' calendar. year 1981.'
A copy of this. Order is bei$g filed.with the Office ef th'e. Federal' Register-o d. O for l}/ biication. m.,5 ;, yI incerely, i p \
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' (*/ Harold' R. - Denton , ihector < ,j
)s,$ ? . a Office of' Nuclear Reactor Regulation. - 4
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Enclosure:
,y Confint.atory Ordre ,5 v .
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CONFIPs,ATORY' ORDER e ey 1
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Lost 6n Edis'en Company (the licensee) is the h'olderLof. Facility Operating i t" ' License No. DPR-35 which ' authorizes the Licensee' to io;9 Mate?th$ilgrim iluclear1 -'
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t d ' d j I.t b Power Ve' Statipn at peaer levels. not in exce)ssf0f1998megawQtsty,.,=.ermal.: 'y (ram power), dL . (;p7 Eq The facility is a boiling wetebreactor loca.ted .c e at4+helicensee w ,. , . sf * "
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.y n Over the past, eleven years the. subject ~ gf anticipated tr' ansientswithout o^p
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c a scram (ATWS) events and the manner ir/wbich thefUshciuld'be ,4 .s consi.dered;in'th j t t ,
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design of nuclear power plants has tepdiscussed extensive!,f,. l'd. ..'b eween th'e ' -l
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In April 1978, the Staff (published at report on~,0 " Anticipated Transients:
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i; my 1) Withoub Scram for Light-Wateo1 Reactord, NURE4 046G,"YoWuss .1 and 2? whidh. yM r. >4 e.
1 5 o Q-summarized technical considerhiit>fs v y ;- relateddo . y ATWSYnd nadi lrec Vr ;
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k D i Following additional investigations by the Staff and by the ACRS, the S af f issued Volume 3 of NUREG-0460, in December 1978. I Although final determination of all the design changes to nuclear pcmer plants which may be necessary to respdnd to ATWS events has not yet been
.i reached, the Staff has concluded that the addition of a Recirculation Pump Trip !
(PPT) in boiling water reactors (BWRs) would significantly limit the immediate consequences of an ATWS event. Therefore, letters da'.ed January 9,1979 were sent to the BWR licensees who did not have installed RPTs. These letters:
- described the reasons for requiring an RPT at this time,
- described two alternative ways to p* ovide an acceptable RPT, and
( - requested that licensees provide an RPT implementation schedule which would provide for installation within two years. III. Ee:ause ar P.PT provides considerable additional assur:.nce that a BWR can safely respond to an ATWS event, I have determined that installation of an RPT by DP licensees should be completed as soon as practicable and in no event J J later P.an December 31, 1980. In response to the letter of January 9,1979, and acditional discussions with the Staff, the licensee . committed, by letter dated April 10, 1979, to installation of an RPT before reactor operation during calendar year 1981. I have determined that this cocinitment should be ; for,melized by Order requiring that RPT installation be completed no later iner Decenber 31, 1980.
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IV. Accordingly, pursuant. to the Atomic Energy Act' of 1954, as amended, andsthi . Comission's regulations . in' 10 CFR Parts. 2 'and.50,. IT (IS HEREBY' ORDERED THAT: The licensee shall,. by Deceinber '31,1980, .. install 'a recirculation-pump trip or in the' alternative, ' place and maintairi'its facility- , n ' in a cold shutdown or' refueling mode ' of.: operation. V. Any person who has an interest 'affected by this: Order may request a . hearing within twenty (20) days of the date of, th'e10rder. Any request for ' ., a hearing will not stay the effectiveness 'of thisl0rder. AnyL request for < a hearing shall oe addressed to the' Direct.or, Officei.of Nuclear Reactor-Regulation, U. S. Nuclear Regulatory Conmission, ' Washington, D. C. . 20555.- , n If a hearing is requested by a person whose interst may1be affected by th.is' N Order, the Conrnission will issue.an Order designating the tire 'and place ' of any such hearing. ' 1 In light of the Licensee's expressed willingness to install'an RPT Tif a hearing is held concerning this Order the issue to be considered at the hearin'g; ) shall be: whether RPT installation should'be: implemented as prescribed in
-i this Order. '
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. Operation of the facility on terms consistent' wi.thl thih.Or'deriisinotj
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'F R'.THE NUCt. EAR REGULATORY' COMMISSION:
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, y o lh Darre 1 G. Eisenhut.c Acting ; Directors.
Division of;0perating Reactors- j. , 4 Office of Nuclear, Reactor.; Regulation < e Effective date: FEBRUARY 2 1 1330 - 9= " .
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'%'..... February 19, 1981 Docket No. 50-293 i
Mr. ' A. Victor Morisi, Manager Nuclear Operations Support Department Boston Edison Company-M/C Nuclear 800 Boylston Street-l Boston, Massachusetts 02199 .l l
Dear Mr. Morisi:
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SUBJECT:
ATWS RPT/ARI DESIGN' l Reference (a) BEco 80-57, April 7,1980 .l We have completed our review on the subject and forward the enclosed safety ! evaluation for.your infonnation. The Technical Specifications of the ' enclosure were effected by Amendment No. 42. to your license. However; the q evaluation also documents the telecon between our staff and BEco (Williams / J Fulton, Deacon,1/21/81) regarding installation / testing of the power supplies and inverters. These items should be installed during the first prolonged j outage (seven days) after April 1,1981' and, in. any case, not later than your j scheduled October refueling outage, f No response to this letter is necessary unless the above installation schedule is changed. 'Thank you for your cooperation in this matter. 3 1 i Sincerely, ' A /) l ( R 1 Thornas A.- ppolito, Chief Operating Reactors Branch #2 Division of Licensing
Enclosure:
Safety Evaluation cc w/ enc 1: I See next page , h f,Ja,
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Mr. A. Victor Morisi
- Boston Edison Company Cc:
Mr. Richard D. Machon Pilgrim Station Manager Boston Edison Company RFD #1, Rocky Hill Road Plymouth, Massachusetts 02360 Henry Herrmann, Esquire Massachusetts Wildlife Federation 151 Tremont Street Boston Massachusetts 02111 Plymouth Public Library North Street Plymouth, Massachusetts 02360 Resident Inspector c/c U. S. NRC P. O. Box 867 Plynouth, Massachusetts 02360 j i i l i i i t l < 1 . .__: . . _ _ _
, b ENCLOSURE .
SAFETY EVALUATION PILGRIM UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS FOR-
' ATWS RPT/ARI Introduction Boston Edison Company (BEC), .by letter dated Apr11l 7,1980,: submit'ted proposed Technical Specification changes-in response to the January 8,1979,.
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NRC letter on anticipated transients without scram (ATWS). The BECo. submittal addressed the staff requirements in volume 3 of NUREG 0460 for a recirculation pump trip system as a means of substantially reducing maximum reactor vessel pressure in thi unlikely event of a failure to scram. In addition the licensee proposed an alternate rod insertion system to provide a deverse means for initiation of control rod insertion. This safety evaluation report addresses the proposed Technical Specifi- ! cation changes and the differences between the Pilgrim Unit.~1 RPT/ARIL and the staff accepted Monticello RPT/ARIL design'. ' A safety evaluation for.the ATWS Recirculation Pump Trip and. Alternate. Rod Insertion systems ATWS RPT/ARI was submitted on the; docket for the Monticello Nuclear Generating Plan (Docket No. 263, License No. DPR-22). This evaluation was reviewed by the NRC staff and a. favorable Safety Evaluation Report was issued on February 23, 1977 for the.RPT: function. Since the RPT/ARI systems proposed for the. Pilgrim Nuclear Power Station are essentially identica1' to that described-in'the.- . Monticello evaluation only the minor differences. unique = to the technicalc j' specifications will be considered. l Proposed Technical Specification Changes l Page 44a Add a new LCO as follows: .j "G. Recirculation Pump Trip / Alternate Rod Insertion Initiation i Whenever the reactor is in the RUN mode, the limiting ; conditions for operation for the instrumentation listed in Table 3.2.G shall be met." Add a new Surveillance Requirement as follows: 1 j "G. Recirculation Pump Trip / Alternate Rod: Insertion i Surveillance for instrumentation which initiates Recirculation Pump Trip and Alternate Rod. Insertion shall be as specified in Table 4.2-G." ; 1 l l l l f & o* Q h),,, $
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2- l l I Page 59a - Insert a new page as shown in Exhibit BI This page includes Table 3.2.G and notes pertaining to the table. Page 66a - Insert an additional item in table 4.2.G as shown in Exhibit - B which provides RPT/ARI instrumentation surveillance require-ments. Page 67 - Modify notes to include Table 4.2.G. ! Page 73 - Insert the paragraphs shown in Exhibit B which state the bases i for the recirculation pump trip and alternate rod insertion systems and the limiting conditions for operation. Addition- f ally, three references are added. Page 77 - Insert the paragraph shown in Exhibit B_ which states the bases for the surveillance frequency of the recirculation pump trip and alternate rod insertion instrumentation. The proposed limiting condition for operation requires the recirculation pump trip system to be operable when the reactor is in the RUN mode and i the alternate rod insertion system to be operable in all modes except i REFUELING. Since the capacity of the safety / relief valves is far in l excess of the steam generation rate achievable in any other mode, there j is no potential for vessel overpressurization in modes other than RUN. ( Restricting the LC0 to the RUN mode for the RPT function is therefore { appropriate. The proposed operability requirements are similar to those of like systems. These requirements were assumed in the design and reliability analysis of the trip system. The proposed surveillance requirements incorporate the fact that analog transmitters are used in ATWS RPT/ARI . systems. These devices are a new, improved line of BWR instrumentation. The calibration frequency is therefore proposed to be once per operating cycle which is consistent with both the equipment capabilities and the requirements for similar equipment used by other reactor vendors. The calibration frequency for the trip units is proposed to be quarterly, the same as other similar protective instrumentation. Likewise, the test frequency is specified as monthly like that of other protective instrumentation. A sensor check is proposed once per day; this is considered'to be an appropriate frequency, commensurate with the design applications and the fact that the recirculation pump trip / alternate rod insertion systems are backups to existing protective instrumentation. SEEREF.(A)
Differences Between Pilgrim and Monticello RPT/ARI The Monticello RPT design includes a' " Manual Initiation" push button on - the operator control console. The proposed RPT/ARI design removes this - push button but does provide manual control of the ARI function from the operator control console. Manual initiation.of RPT at the console is unnecessarily redundant due to the variety of means already available to the operator for manually tripping the recirculation pumps or otherwise reducing recirculation flow. The addition of.the ARI function results in additional crowding of. the - operator _ control console. In order to reduce this crowding the manual reset push buttons have been eliminated and automatic reset logic sub-stituted. Although the Monticello RPT ' design included a seal-in logic . l with manual reset it is unnecessarily redundant. Once a trip signal. actuates the field breakers, it can be removed without affecting the state / of the field breakees. The field breakers must be manually reset. Therefore, the automatic reset feature only reduces the manual actions required to reset the pumps for operation and does not affect the trip function. The ARI automatic reset logic includes' a seal-in logic for a 30 second interval to assure sufficient time to blow down the pilot air header and insure complete rod insertion. The automatic reset cannot , function, however, if the trip signal is still- present. . In this case an additional 30 seconds of delay will occur before reset and this* sequence will continue until the trip signal is. removed. I l The high pressure setpoint for ATWS. RPT/ARI (1160 psig) as proposed. is ) higher than the specification for the Monticello .RPT (1150 psig). ' The ' licensee's analysis indicates.that the raised setpoint will result in a - 25 ps1 increase in the peak pressure during an ATWS event. .The peak pressure would still be more than 150 psi below the ATds criteria of 1500 psig. The results of this analysis appear-reasonable based on comparisons 1 with the calculations for the Monticello plant (NEDO 25016). . J With the current plant configuration, initiation of the ATWS RPT function is predicted during certain pressurization transients if the ATWS RPT 1 setpoint is not raised from 1150 psig' to the proposed value of 1160.psig, 1 Since the initiation of ATWS RPT causes an increase in severity of the
~
transients this is an undesirable condition. Raising the setpoint decouples the more frequent pressurization transients from ATWS RPT effects. With the proposed setpoint, the only events which will initiate ATWS RPT are the turbine / generator trip with bypass failure and the ASME overpressure protection event (MSIV closure with trip scram failure). The turbine / generator trip events will not result in exceeding the vessel . 1 pressure limit despite the increased severity due to ATWS RPT fnitiation and the limiting event will remain the MSIY closure with trip scram failure..
The design of each ATWS cabinet includes 'two qualified 24 VDC. power supplies. One power supply is sourced by offsite 115 VAC which. transfers to onsite generator power in the event of loss of.offsite power. . The second power supply is sourced by the station batteries (125 VDC) throughL an inverter (125 VDC to 115 VAC). The uninterruptible power supplied by the inverter allows .the system logic to remain ' energized during the event of loss of offsite power until the transfer to onsite generators occurs. The present systems.will require manual action to initiate a pump trip and ARI during a loss of offsite power event. The presently installed' ATWS RPT/ARI systems include one non-qualified power supply per ATWS cabinet sourced by offsite 115 VAC. The qualified DC power supplies and DC to AC inverter- specified for the ATWS RPT/ARI systems are onsite1 and will be installed as the final system during the first seven-day outage after April 1,1981 and not later than October 1981. Conclusions Based on our review of the licensee's proposed Technical Specification changes, we conclude that they are in conformance with the requirements of volume 3 of NUREG-0460. The differences between the Pilgrim and Monticello RPT/ARI systems are also acceptable. However, we recommend that the licensee be required to install the qualified DC power supplies and DC to AC inverter specified for the' ATWS RPT/ARI systems. .The present systems require manual action to initial a recirculation pump trip and ARI during a loss of offsite power event. The present system is acceptable as an interim measure,only until the next scheduled refueling (October 1981). a l l l i
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( '- ' o{ UNITED STATES g 3 n-t NUCLE AR REGULATORY COMMISSION . WASHINGTON, D. C. 20555 k.**"*f December 2,.:1983. y ,, para l#- - I 0#D $ j;p pia'S
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W City. ^ TO ALL POWER REACTOR LICENSEES AND APPLICANTS.FOR 0PERATING LICENSES'
' Gentlemen:
Subject:
NRC Staff Recommendations Regarding Operator' Action for Reactor Trip and ATWS (Generic Letter. 83-32)~ - The NRC staff has developed'the enclosed " Staff Position on.0perator Actions for Reactor Trip and ATWS." The need for the.' development of-a position became apparent during discussions with several. utilities.'as a result of reactor trip' failures. The attached position is being forwarded to you for information as it is being used within the' staff for guidance. It does not constitute. a requirement. '
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arre i enhut, D ector Division of Licensing . Office of Nuclear Reactor Regulation
Enclosure:
As stated i O I
'831202030 4 2.-%TU 2-y? ,
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STAFF RECOMMENDATIONS ON OPERATOR ACTIONS'FOR-REACTOR TRIP ANDJATWS (OCTOBER 1983) Operators have responsibilities and take some actions for all freactor trips... In the-case of a~ failure of the automatic trip system, operator action.is the : only backup to initiate roc:frisertion to shut down.the reactor. Operator action to immediately back up all automatic trips with,a manual: trip based-solely' on receipt of " positive ' indication" 'of ar. " automatic trip demand" without evaluating the automatic; trip system's success is believed-by the;NRC: staff to be conservative and,'therefo're,J the preferred method. (This; operator action is not currently required by :the NRC. Al though not: as < conservative..- another method is to have operators manuallyitrip the reactorLwhen the' successful completion of an automatic: trip cannot"be .both immediately and a ' positively confirmed. With' either method, .if. successful completioniof j reactor trip cannot be immediately and.' positively confirmed after. manual trip. " of the reactor, appropriate backup measures must be' prescribed as part' of the; ATWS emergency instructions. ,, 3 Facility procedures should identHv'the instruments that provide the' .
^
" positive indication" of an " automatic ' trip demand." ;In -addition. .~if the choice is made to verify failure of the reactor trip system' prior.cto -
inserting a manual trip, the:; procedures should'identifyf the instruments' that provide the "insnediate and positive confirmation" of the success'of the, automatic' reactor trip. The instruments. selected must _ provide timely.
. indications with adequate reliability' to . minimize the number of' unnecessary manual' scrams and yet assure the operator is' alerted to a5 system failure when-operator action is necessary.
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Generic Letter 83-29: Section 2.2.2;; EA
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l J /' Fln ^33-28. the wake 'of the 'Salme Evest*, NRC issued NUREG 1000 and -Genettit.4etter$G Among thei requirements 2ef. Generic Letter 83-28 was -Item 212,Which 9 instructed licensees to develop and asintain contact with vendors of; safety' related equipment. . Such contact would entail a " feedback" system to ensure ,,: receipt of information, and' en internal . review system to dispositten-the. ' inferination such that appropriate changes to practices and procedures would be' . ; made. in a timely manner. '
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. Beston Edison (BEto) believed that a generic approach by' the various nuclear . ]
. utilities, as suggested by the NRC in Generic Letter 83-29, would be the mest i ef fective way to address this ites; therefore, SECo actively: participated',in - l
' the Nuclear Utility Task Action Committee ;(NUTAC) on. Ites 2.2.2 which has -
formulated a Vendor Equipment Technical Information Program (VE11P) .- BLCe J herein submits a description of the VE11P as Attachment A to this ^ 1etter, and:. endorses, in conjunction with internal progress- described in. Attachments 8 and ' R C. all' but Section 3.2 as BEco's progras for addressing Item 2.2.2. At this time we cannot endorse Section 3.2 because; the ' enhancements
- talked of in i this .section have not been developed to a point where we can ef fectively- i assess their impact on safety or resources. - Naturally. BECo will review these-items as they become more defined and implement those .' assessed 1o4 be-beneficia'.
BLCo endorses the VEllP because. as various alternatives were explored, it became apparent workable program.that. only- a centralized approach would provide an ef f ective, Further. this progree serves to enhance both the interf ace between other vendors. utilities and the N555 vendor. and the *as-needec* interf ace with
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- i e 4;. * , : p; Page-l Provideo as Attachmer: B to t*is letter is Nuc lear Operations Procedure (h0')
8401, ' Operating Expe rienc e seview Program *, which describes an existing BL;o-prog as, developed in accorda9ce -ith the requirements of NUREG 0737 and the guidance of INPD, whi h we belie e implement s . the ' YE11P guidance . and L the requirement of Generic Letter 83-28 for the timely review and dispositioning of. vendor information. Provided as Attachment C is NOP 94 A4, " vendor Manual Control'.- This NOP establishes a system f or the everall control of vendor manuals associated with the installation, operation and usintenance of Pilgrim Nuclear Power Station. [ Sased om our review we believe the . vender. interface concerns"of Generic Letter l' h6 S3-79 mee satisfied by the esisttag.SEco
. tde _ areprograms and the VETIP ete ? described is . S
$jejQ 1: Pchments to thissataittal.. prepared, of 'courseg change " g t experience indicates ways to sette them more offactive, Should.yes - % y[L des. Ng" .further information on this_ issue, please contact us. - %2 l*
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Very truly yours,
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Attachments: A) Vendor Equipment ' Technical inf orination Program
- 8) NDP S401: Operating Experience Review Program '
C) NOP E t A4 : Vendor Manual Control PMK/kx cc: Mr. Darrell G. Eisenhut, tirector Division of Licensing . Of fice of Nuclear acactor Regulations U.S. Nuclear Regulatory Cassirsion Washington, D. C. 20555
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I s 4 \ l Nuclear Utility Task Action Committee r ON GENERIC LETTER 83-25 SECTION 2 2 2. Vendor i Equipment s ., 4 t-
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Program March 1984
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4.,; s. Developed By ' Waclear Utility Task Action Comittee l for Generte Letter 83-28, Section 2.2.2 i l INPC 81-010 i (frJTAC) ] Ma rch 1984 l 1 l
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***s pu*alication *as been ;"o3., rec by the %stAC on ',*aeric Letter s} 29,
- ton 2.2.2., w'tn t"e ss;;0*! c' tre Institute of %sclear Power *A*atic,es N O). The officers of Inis 5.!A: were Chairman Ec.a rc P. Griffia;;s-3 vi:e C*a rman Walter E. Andre =s.
Jttlities that participated in this NOTAC- incluce the following: Alabama Power Company Net raska Public Power District . '1 A erican Electric Po. e Service Corporation New Yort Power Authority . artzoea Public Service Cogany Niagara Mohaak Power Corporation , Northeast Uttlities arkansas Power & Light Company . f attimore Gas and Electric Company MW Northern States Power Ctm;:a.iy ; Boston Edison Company 't;pt(baana Public Power District [ ' Carolina Power & Light Company ^ ",
~~ Pacific Gas and Electric Cogaay -
Cincinnati Gas & Electric Company U Pennsylvania Power & Lignt Conpef - > The Cleveland Electric illuminating Company Philadelphia Electric Company, ~ 3 Commonwealth Edison Company .Portlaed General Electric Company. ~ 1 Consolidated Edison Company of New Yo rk , Inc. Public Service Company of Colora:o j Coasumers Power Company Pablic Service Company of1 ]nciaaa, : : Tne Detroit Edison Company Public Service Company of New wa ;s - Dake Power Company Public service Electric and Gas :: ::*. j Dscuesne Light Company Rochester Gas and Electric Corpora! :' I Florida Power Corporation Sacramento Municipal Utility Dist":' Florida Power & Light Company South Carolina Electric & Gas Co ;a SPU Nuclear Corporation Southern California Edtsoa Coga j i Georgia Powe* Company Tennessee Valley Authority ' Gulf States Utilities Company Temas Utilities Generattag Co9any 1 Mcaston Lighting & Power Company The Toledo Edison Compaay l Illinois Power Company Union Eiectric Company - Iowa Electric Light and Power Company Vermont Yankee Nuclear Power Cor :'at *
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tansas Gas and Electric rnnpany Virginta Electric ar.d Power Con;aaf Leag Island Ligntiag C~';any Washington Pubile Power Seply Syste-Losistana Power & Light Company Wisconsin Electric Powee Company ma ine Yankee Atomic Power Company Wisconstn Public Service Corporatica l't ssissippi Power & Light Company Yankee Atomic Electric Company ICTICE: This document was prepared by a nuclear utility task action cose-ittee-(%.< TAC) with the staf f support of the institute of Nuclear Power Operations (!NC. l heither this NJTAC, !NPO, members and participants of INPO, other pers:as cont:- l uting to or assisting in the preparation of the docume-t, nor any'persea actiag :* ! t>enalf of these parties (a) makes any warranty or representation,.tnpressed or 1* plied, with respect to the accuracy.. completeness, c usefulness of 1"e lef:"ad" tion contained in this document, or that the use of .an,, information, a;paratut', method or process disclosed in.this document may not irf ringe on privately own.'Or rights, or (b) assumes any llaD111ttes with respect to the use of any inforn3tica, j a;paratus, method, or process disclosed in this docume-t, i
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( A[CL * . ! h*G ' I 1 i inis re et was prepared by the Nuclear Utility Tass An t on Cov . tee ( UT A' ) . on Geme-sc lette r 63-?8 " Required Act ions Based on Gene ic Im;11 cations of Salem a*.5 [ vents
- Section 2.2.2. It describes tee Veador [qsipment Tech- )
rocal la'Demat ton Program (VETIP) developed by the NUTA: in response to the conceres on vendor information and interf ace addressed in Section 2.2.2 of the I generic letter. VETIP is a program that enhances information exchange and 1 evaluation among utilities constructing. or operating nuclear power plants and , provides for more ef fective vendor interface. I l
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The NUTAC was comprised of representatives of 56 utilities that are mehrs~cfM[ m m. the Institute of Nuclear Power Operations (INPO). Staff support for the?NUTAC "J r was provised by INPO. . Tnis report unanimously presents the final conclusions'% +. <
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-of the trJTAC and is provided to assist individual utilities in seveloping specific programs to meet the intent of the generic letter.
Generic Letter 83-28 was developed following investigations by the NRC on the Salem events. As a result of these investigations, the hRC determined that better control and utilization of information regarding safety-related compo-nents might have helped to prevent these events. The NUTAC identified a program to better ensure that plant personnel have timely access to such information. The NUTAC ef forts were guided by the recognition that individual utilities have the greatest experience with and are most cognizant of the application of safety-related equipment. Vendor involvement witn such equipment is generally greatest during construction and initial operation of the plant. Vendors are not familiar with the surveillance or maintenance histories, nor with the . application of the equipment or its environment. Tht 5 type of information is nost readily available at the plant level within indi vidual utilities. _ Based on this recognition, the NUTAC investigated the mechanisms rently available to facilitate information exchange among utilities. The NJTAC i identified four activtties that currently address information a o' ut safety-i I
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re ~ s.ee :: peaerts. *aese c e : 'ae A 'Iit i<*e-::* or- / ** . - . ,- - iate-cnange, and t *e SE E-:s a-: ' N.'S pec ; rams maae;ed t, It as the assessment of tae %.' C inat tnese existing act i.ities, -* : - . inte; rated and implemented, wd : provide a f ramework for an overa : .- to casure ef fective cownicatiem of sa f ety-relat ed inf ore'.at i on am: ; a ' ' ) utilities. Accordingly, the pregram developed to accompitsh tnis goal .i* : I 1 uses the esisting ef forts as elements of a more comprehensive progra .. l
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The 1/ETIP combines these existing programs, incorporating ennance eats, . - ,
. : coordinated program within each wtility. A key element of the . VET!D is tne development. by each litflity.of an active internal program.to contri@te la'- -'
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. mation to the NPRD5 'and SEE-IN Programs 'and to use the results of these ' 1
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-l The effectiveness of the VETIP will be determined by the level of utility :sa. !
ticipation in these programs. To ir plement the VET!P, eace utility s*oA:: .I assess the type of information estrently being provided to NPRDS anc SEE.:'. j and expand the scope of reporting if appropriate. Additionally, eaca att' . Should evaluate current administrative controls for reporting infor attoa e-: for disseminating the results of the NPRDS and SEE-IN Programs to t*e :laat level. These administrative controls may require modification to ens.e ertaat effective coordination is established. Concurrent with these ef forts er a*:f-1 ments will be made to both NPRDS and SEE-IN by INPO within its prese-t tast$- i tutional ob.Jectives. The VETIP has been developed to ensure that nuclear utilities have peoret I access to and ef fective handling of safety-related equipment technical taf:'- mation. In addition, VETIP is responsive to the intent of Generic Letter -J 83-28. Section 2.2.2. Furthte details are provided in the body of tais rtport.
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.- J e: ,4 y 22 and 25,19 3, ds ag si e : .:s c' t *e d e ie- L" t . J :- 1 4*. M '.
- ca:ter t rip bream ers (wes t ingno.se sc re! Ot-$0: f a' led to open om er a,to-I ratic t rip signa l . As a conseaseace, the N clear Regalatory Con-i s s i on (hP :
ferme: an investigating task force to ceterrine the f actual inf ormation perti-neat to the management and administrative controls tnet should have ensu r ed p cper operation of the trip breamers. The findings and conclusions of the task force are documented in NUREG-0977 "NRC Fact finding Task Force Report oe the ATW5 Events at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25,1983." A second task force determined the extent to which tnese investigative findings were generic in nature. The NRC subsequently issued WUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant" and Generic Letter 83-28,
- Required Actions'8ased on Generic Inclinations of Salem ATES Events." ~N
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l- ,. 2 3 .~ On September 1,1983, a group of utility representatives net at the of fices of the Institute of Nuclear Po er Operations (!NPO) to discuss the establishment of an ac hoc utility group to address issues relative to the NRC Generic Letter 83-28, Section 2.2.2. The representatives decided that such a group could provide direction that would be of generic benefit to the utilities and consequently formed the NJClear Utility Task Action Committee (NUTAC) on Generic Letter 83-28, Section 2.2.2. The specific charter for the NUTAC ( Appendix A) was adopted, and the target date for completion of activities was established as February 1984 iii 4m **J r e n' S l - gp ,, d , .'
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2.2 Definitions...................................................... .4 . 1 1 3. VENDOR E091PSEhi TECHNICAL INFORMA!!ON PROGRAM (VETIP) DESCRIPTION.... 7 3.1 Esisting Programs................................................ 8 3.1.1 Nuc lear Plant Reliabi li ty Dat a Syst em (NPRD5). ... .... . . . 9 3.1.2 Signir *it Event Evaluation and Information Network . . (5EE-IN)...............................................11
, ,,'- /# .3.1.3 I n t e ra ct i on wi t h Vendors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ; 14 ?-
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,q 3.1.4 ' Re gul at o ry Report i ng Requi rement s . . . . . . . . . . . . . '. . . . .'. . . 2 16 e / " ,
i j t 3.2 Rec ommended Enhancement s 1to Es i st i ng Programs . . . . . . .'. . . . . . . . i. . . . 17 <
' 3.2.1 Enhancements.to WPRD5...................'....~...~....'... . 17
- 3.2.2 E nha n c eme n t s t o 5E E - l h . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 9- 3.3 S umm a ry E x a mp l e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 j 4
I MPL E ME N T A T I ON OF VE T ! P . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 1 4.1 Respons i bi li t i es for Impl ement at i on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 4.1.1 La il i ty Impl ement a t i on Responsi bi l i t 1es . . . . . . . . . . . . . . . . 21 ) 4.1.2 lhPO Impl ement a t i on Respons ibi l i t i es . . . . . . . . . . . . . . . . . . . 24 i 4.2 Sc hedu l e f or imp l emen t a t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 ! 4.2.1 E x i s t i n g P ro g ra ms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 ) 4.2.2 Enha ncement s to E xi st i ng Programs . . . . . . . . . . . . . . . . . . . . . . 25 i rigsres i F i gu re 1 . VE T I P Bl oc k Di a gram. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 Figure 2 - Operating Experience Review Process and Related Activities................................................ 27 i APPENDIX A: SPECIFlt CHARTER FOR NUCLEAR UTILITY TASK ACTION COMMITTEE ON GENERIC LETTER 83 28, SECTION 2.2.2 APPENDIX 8: LIST OF REFERENCES APPENDIX C: - ' SEE-IN FUNCTIONS APPEND!r D: GENERIC LETTER 83 28. SECTION 2.2.2 v l ,, ' , '-
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m; re e tre saf ety a : rella 1*ty :' n sc lea pp.er generet ", st e: . s : ,. ensveing inat the ut'lities a*e pec. :e: wita sig-ificant an: t i me 'y te;*- f
*1 cal Information roe:erning reliability of saf ety-related co poae-t s . **
l a typical nuclear station. h.,-creds of vendors su; ply the thoLsarcs ef q compoaents that perform safety-related functions. The variations in vintage and design of plants ensure that althoup comon applicat10's of I specific components may exist, there are an equal or greater n.eDer of u n icue applications. To attain t*e objective in a cost-ef fective ame
, efficient manner, this NUTAC has developed the program outlined in tais cocument. This_ positive program has been found to be the most realtstic approach to attain the objective. ,
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m I The veneor Equipment Technical Information Progree (VETIP)isescribec .. in , tnis decament establishes a more formal interaction among the major organ-tratiens involved with corrnercial nuclear power generation. The goal of tne Interaction is to improve the quellty and availability of easipvat tecarical Inf ormation for use Cy the utilities. The major components of EMe VE p are set infore.ation treesfer system and a centralized evaluation of in:Jstry experiences. I inis document provides the unanimous kJTAC position on the g idelines *:e sa e*fective technicel information program. The determination of ea:"
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Sacivi sal utility to support s'id utilize these guidelines is the ke,. : t*e ef festiveness of t*is pro;*am for the industry .as a w% Die. Tne pe:- ge-m coes not reasire the use of nor prescribe sta*dard administ rati.e processres, but it allows the use of plant-specific process es r o ; eta;*e with the utility's internal or;aairat'on and needs. Howeve*, t*e re:r - mendations in this cocument pr:vice tne basis f or a uni f o*m i nd s t ry response to NRC asestions and reg.,1revnts relative to a tecent:a1 1 ' - l mat ion program. This program will be oeneficial to the utilities aae. a: l the same time, /it will be respeasi ve to Sect ion ?.?.2 of the hkC Geace it Letter 83-?8. i
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C'perati onal Dat a ATW$ Anticipated Transtent Witnost Scts* l CFR Code o' Federal Regstations ; EPRI Electric Powe* Research Institute i ETI Equipmeret Technical Information IEB, IEN Inspection anc Enforcement Bulletins ancu ;
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*fMd!*;f3 Licensee Event Report.' issued by a utility j l p3R Monthly Operating Report l NDRD5 huclear Plant Reliability' Dat a Syste- 'I haC Nu: lear Regslatory Commission h5AC haclear Safety Analysis Center I h555 haclear Stes'n Sapply System kJTAC hsclear Utility Task Action Coarittee !
O&MR Operations and Maintenance Remiade- ; 1 PRA Probabilistic Risk' Assessment OA Quality Assurae:e SEE 1% Significant Event Evaluation anc I femat toa Network SER Significant Event Report
$3ER Significant Operating Emperience Re:o-t l
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. , : t yne cr i a e. clear powe* geaa at'or.. A5 11 .itrate: 'r Fig.are 1, a utity enchar ges 5 4'ety-re' at ed e;u i pment i n f ore a t i c e. =$th vee ors, ho ,- ;
IV ?, aad otner uti ttles via reports, bulletini, e:tices, newsletters, anc meetings. The pse pose of these information eac**dnges is to share eqs *pme*t tecenical information to improve the sa'ety aad re11et111ty o' nuclear power generating St at ions. The WTAC concisdec that the lack of information .15 not a problem, but that the various Information Systems- j availaele are not integrated properly. The purpose of VET!P is to en5u ea that current information and data will be available to those personnel .
.. 1 % in.], m respon5itle for developing and maintaining plant ' instructions,and pro . j hcedures. Theseinformationsy'Stee15andprogramsC.urrentlyeAistcendare;[.f,'
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$ @, ce Salem type event. JETIP 15~p..- A c i.n . . .. ~. ?an indestry-controlled and mainly hardware +4(p'J :~ - - '
"- f ortented prograe that does not rely on vendor action, otner tban~ thel 5551-sup;'ier, to provide inforn.ation to utilities. lestead VET!p provides ief or at ioa developec by indu5try experience thro.,gm SERS and 50ERS to the '
ven:or for c oment be 'ere i
- is circulated to the ' utilities concerned, i
j Tne ca.;ority o' infor-ation provioed by vendors is commercial in nature.
%is us., ally is p*ovided voluntarily by the vencoe, but does little to improve tne sa'ety or reliability of emi$ ting easipaent . j A weedor-oriented program to provide information !*at mosld traprove the safety aes rellatility of ent Sting r:wipment relies ca t*e vendor having an saternal pr:g*aa to develop the information. Such programs typically are act in existence. Following oesi gn and qua li f icat t oe testing, vead:rs normally do not coatiese extec51.e testing or eag+aeertrg progra s i -
anticipation of e: sip wn: prc:le 5. Sut5eauent f a 1.res discovercs esring operation 5 regs i re Several Ste?S t o C ompl et e t he 1 - for . t i on feem ack loop. For end*;le, =*en a proble* occars am a 10;u' vea m represea. tative provide 5 4 Solution, he wou'C have to prov':e that i n.f orma t i on t o the vendor hea2;ua *t ers. Then, toe headavarters aoslJ reed a tracking prog *am to idertify a trend a'd Sub5egsently a prof ra- to provide the inf ormat ion t o the indust ry, in addition, the ver:o cften 15 rot ia the
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' face h'$tO*y of Ine egutpment. ?*e .,tility, not the maas'actJPe",8*!.'
**e coalconent 's actual application and enviro'ime"t. The ut1Ilty is !"e
;-' mary source of information on the f ailure, and the ut 11ty has t*e I
geedtest need for the solution. As such, the utility is the ceatra' reganizer in any approach to the solution, whether or not the maas 'a t , , pets involved.. The utility is in the position to know of. t*e f ailure ,, a*alysis'and"its. solution et.the earliest possible time. Tne utility ca.- . a 'vg
.aen disseriinate the in'ormation to other utilities. with an Indica'. ion j
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its significance and urgency 1, 7,o' . c. -. y fy sharing the operating history, prod 1 ems, and solutions within t*e - j Nclear industry, independent of any no-mal vendor contacts, tne t't'e' sees will be tafoemed in a much more timely and uniform way. In nis ,
=sy, the distribution of information is controlle3 entirely by tae n :'es' s _
The programs that comprise the VETIP currently are ia wi t lity industry. esistence. The recor'vnended enhanceNets contained within t*u s resce se s ,ggested ways to improve the current use and application of these entst-
'ag programs.
3.1 Esisting Programs - The existing systems and programs included in the VETIP are the huclear Plant Reliability Data. System (hPRDS) and the Significant Evert Evaluation and Information hetwork (SEE-IN), both manages by thP0. Also, the VET!P includes esisting prograr s that the ut111t'es ' now conduct with vendors and other sources of ETI, particularly tne NSSS vendor i teraction progra rs and the NRC report tng programs trat disseminate significant failure information. _Ut ility-vendor inter. ' action is further enhanced by the thPO supplier participant prectices. Through participation in this program, NSSS vendors and A/E firms are working toward greater participation in the hPRDS and 5EE-!N Progra*5.
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,; s,,1c3r :aat N:;?$ is aa t e: At ry wi ce sy st e- *a' t ;a! by IN;s f 7 mo ^ ' t ; * ' * ;
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- - per d re:* of s < ' e : t *:' sy s t a-i a1 co pe eats at a : :6*
p?.er plants. ' %Q -embe" ut i l i t i e s ha ve a ga ead t o 'pa r.t i c i . pate in the pr w ar. United States plaa ts in comercial. f operation (except fce sia 4 typical, early .vintige units) i supply basic engince-ing informatio- and subsequent f ailure - data on the selected systems and components (typically s.im 'to seven thousand components. from some 30 systems per unit). The value-of NPRDS lies in the rehy availability of th'is data base to operation and engineering groups:for a broad: range of 9 applications. The criteria used to determine the scope of.
'n NPRDS reports are as follows: .
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g d) 1 systems and c.omponents that provide functions. necessary fo
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initiate a significant plant transient ~ Uniform scoping and reporting criteria are set forth in the' Nuclear Plant Reliability Data Syste (NPRDS) Reportable - j
-d System and Compoeent Scope Manual (thPO 83-020) 'and in the Reporting Drocedures Manual for the nuclear Plant Reliability-Data system (ikPO 64-011).
To support the benefits that can be obtainea from hpRDS usage, utilities submit three kinds of information to-the NpRDS data base: engineering / test information, failure reports, and operating history. The engineering / test record on a component contains information necessary to identify the component and its application, such as manufacturer, model number, operating environment, site, horsepower, and test frequencies. The , information is submitted when the coroonent is placed in ,f service and is stored in the data base. .lf that component ] e f ails to perforw as intended, a report is submitted containing l 1 a description of the f ailure mode anc- cause, the f ailure's- i effect on plant operations, corrective actions taken,- and other 9
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Dd51s, plaat is in cif f ere-t modes of ope,a;ica, imig 3 . < ;; , . , ; , 7, ,g J used In conjun ction with tne engineering and f ailur e re;:rts to generate f ailu re stat istics for systeats- and coapone*!s. I The' data is retrievable from a computer, and the engive-tog and failure information can be combined in various ways. A scare,i of the failure records can identify problems espeet- ) enced with components in other plants and the corrective .
, actions taken. There are several hundred searches of the data 'I
-.
- base in a typical month. Following are;some example uses~of' h-(f' '-
the data base: i - y .* v - 1 ,, y . .w j Q, , u; f h ijh 1
?A- Utility and Plant Staffs 7 -
~ a 3 o ' accessing comprehensive equipment history files to sspport- '
3 I maintenance planning and repair o avoidance of forced or prolonged outages by identifying other plants with similar or identical equipment that may have spares for a possible loan o determination of spare parts stocking, based on indsstry mean time between failures o comparison of component failure rates 'at a given plant witn the industry average failure rates l Design Groups o identification of common failure modes and causes j o selection of vendors based on component applicati.,i av . i performance o identification of compo"ent wearout and aging patteres-o studies of component performance as a function of operating ; i characteristics, such as test frequer y and operatir; ,
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environment o input to plant availability improveme t programsf J
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'a ?!-tree ta- ses (reliab;lity or'PRA stuc'esl r
NEC3 data is s ea oble to users through va r10.,s periodi c reports an: te ,oa oa-line access of the data f ror a co%ter terminal. 3.1. 2 Significant Evrt ~ valuation and Information Network (SEE-!%'
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- Since Ine emely cuy.cf. nuclear power plant operations, utili-
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-41l. ~ ' Xties and manuf ach,*ees4 ave attecipted ,to share what' has' been ;i
" 1
. 7iearned from p5.an-Nrating experience. - Es nuclear tecN- +
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.sr' j jg g( f nology 'becomes e complea and.more demandir!g,;the need.'.f.0,r q sharing operattag 4cerience continues to grow aAd becoss'
more 170* tant . % safety benefits 'of avoiding problems alreacy enco;ntece sac resolved rore tha'i justifies tne costs ! and estra e' foe- % red fer utilities to *eep ea:" otne" 4 in'o mes, t%e %:*ee Safety Acalysis Center jN5A;). wi? . tae - l suoport o' tts st% ecvisory group, began developing a> - program tc share 'evation lear",t3 fec ana1yt ng nuc1es* . j plant empericace5. sacrtly af ter its formatica '9 late 19 4 . tae Inst state c' %,:ese Power Operations (!h:0) Joiae: N5 t.: ia the deve!op-e t ce implementation of the program. Tne progra- nas oee r_ tac "Significant Evert Evalsation and information %et-ce, (SEE-lh). In 1991, the rnanagement of tae l SEE.!N Pro;*ar Secsa the sole respons10111ty o' INPO. Ot:a c t i ve The obje:tive o' SL R is to ensure teat the cua lat:ve I learning process f c operating and ma ntenante esperience 35 ' ef fecpive and t*at te lessons learned are reporte;! and'cor. : rective action trac n a timely manner to improve plant safety, re11acil-ty sne availabtitty.. This objective is met I l i
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1 and app'?;-tate ces ;ners and rans'a ! vers. l l J Scope The funct i:na' approach to SEE-lN is an eigat-ste: p e: s; .1 outlined 'n Appendia C. While INP0 has the prograi a a;* function, no single organization is responsible for per': -- ; all of these functions; rather, the responsibility. 's s: -s * ! 3 among key participants .in the network. Thepr[inchte::e-.- [ y. . i ' T$ations involved in the initial screening of plant event :t:L .c . b > 1
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.are the~ A111t'ies and INPO.: ~ EacEn.:..a,muclear. . (11ty )es an
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pl[nt(s). INP0 has ta broader charter tottereen 411 nscte3- 'W;a;. e ,, plant events. The sources of input to the ' screening pr::ess . .f include NPRDS. NUCLEAR NETWORK, NRC-mandated reports,'!!!s, C IENs, etc. The provision to contrel the data norrally is governed by agreements between INPO and the supplyta; orge - zation (e.g., utilities, NRC, NSSS vendors, international l1 participants, etc.). When a significant event or tread ass been identified from the screening process, a Significe-:- 'q Event Report (SER) is prepared by INPO and transriitted te tae utilities and other participants on NUCLEAR NETWORK. Ta's event then undergoes an action analysis by .!NPO. Tne pe::se l of the action analysis is to investigate the event or trea: sa more detail and to develop and evaluate practical reeecies.
?or events requiring utility action, the results of the action ( ;
analysis are conynunicated to the utilities, norrully'in t*e f form of a Significant Operating Experience' Report (50!#1. la~ these instances, recommendations are made to resolve tre ua :e'- lying problems. The implementation of applicable recoar'e-re: u \ remedial actions is the responsibility.'of the individ$al .til- 1 1 ity. Implementation may include changes in plant Voceos es, equipment design, and/or operator training programs. Tne t.o final steps in the SEE-!N process are (1) feedback and IVJ
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I for eveats we*ca, t***.,g* t e sc"eerin9 process, are ce* e . . r-seed not s i ge l ficaa! P.,t ha ve va 7 sac Ie oDef at io*5 o "d ~~a ' * ! f - nance information, aa Opera! tcas & "d'at ena*ce Re*1adee
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is prepared and processer ie the same way as SEDs. The SEE-!N Program provides copies. of draf t SERs, 0&was, a-: ,
' 50ERs ,to t'ie af fected vendors for review. Vendor.co eneat.5 are -
w;Sf,, n Lconsidered in preparation.of-final 'SEE-IN reports'. 0cce.. - fh7 ,,4 Affgha112ed, thelreports are sefit to the utilities. kb - Q.-n ' .w ,y
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<. ,1udes' a cross-referenc'e capabilitrto:
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N identify SERs, O&*ts. 50ERs. LERs, 'etc., whicn report como . l ) nent problems t*iat coule cause a' significant event. Tn15 l cress-re'erence factiitates utility reniew cf the cor poae*!'s 1
- l prior history before using tnat compont*t in a sa'et -re'a'e' application.
Procra* Ope'atton Plant operating expe*ie9ce cata is reviewer f ro se<erkI-perspectives teclud'ng desig , component and system rea' . maace, plant procedures, har.am ' actors, persernel t rainteg, maintenance and testing p a:t *ces . ara manage eat systeas !;- i identify significant eveat s a-d t reads. Icemal Review Sources A for% ) review is como ,cted Cn AE ~ tof ormation not1:es, 1 l bulletins, AE03 repo*ts, ever t.related generic letteri, etc.. A f ormal review also 's c0'0,:!e0 ca indust ry-p repa red . i n',0 mation (including those cent red d NR Ci .* s u c h a s' L E R s , ene- t - f . operating reports' NRC eveat , elate 3 eeports, NSSS technical bulletins, NPRDS date, kJCLE A2 NEtw]RK operating emperieace
-13
-, -q 9 l et q . p+ ence _;-ets, c~ r, . ente,et u ,3 t 3 ' :; tier c.' . ea:, ren s, sa*ety ce'ect cer*:s anc trees
.1 ideatte: as signi'icant 1- tee *NP hW5 a-: 1.E : cata-bases. **e f :rmal revie , -i ncludes a dua l , i- epen.;ent sc ree%
1
'ing pr::ess. 'me review status is docseentec a-a tra:ked by J computer. ,
t ) i Other sources of operating experience information are used.byl } I the SEE !h Program on an ad hoc. basis as reference or supple-mental material but do not receive a formal review.- The' sources include such-items as NRCe NOREG docuarats EPRI and s .
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. n g'h.kSAC reports, and other -inde'stry# reports or esta.concerne ,.
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.Jhk Utility Centact (SEE-IN)
In addition to the formal 'and reference' information sources, another vital information source is direct contact with power > plant technical personnel on an ad hoc basis. Each utility designates a SEE-!N contact to respond to questions.from INPD on plant events. The majority of such communications was - l handled over the telephone or via NUCt. EAR NErdO4(. Files are i maintained by INPO on nuclear utilities and centain names and 1 telephone numbers of designated contacts,; telecopier numbers, status of nuclear units (ije., operating, under construction i or planned), and NS$$ vendor (s). 3.1.3 Interaction With vendors . 4 In the interest of operating the plant safely ar4 ef ficiently, the utility-vendor contact is essential. To accomplish this goal, utilities already interact with various vendors. !
. . i l The contractual obligations for furnishing eg.ipment a~ d n sof tware (manuals, drawings. etc.) are fslfilled upon accep-tance at the plant site. Interaction between utilities. and
~
vendors, due to deficiencies, may be brought about by. the 4 e n 14 re
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aa icitial ;u ;*ase. Im acDt'oa, m'a of tae.' ate'e't* ' with the vendors during pleet Infe is teittatec in "espo*se t*, significant failures, to failur e trends'espertenced at tee plan', to spare parts procurement, or to subsecuent -psechase orders of new equipment.
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The interaction with the NS$5 vendor, whL typically supplies a-large portion of the safety-related plant. equipment. penerally;
.i i d Tpree are esist ' g m
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w th the other venc - ~, uw ors.M zn - 4, %,,. . w 7 s.,q . .~; information
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of laterest to their C11ent utilltles. 7 Thete%6.:
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, ,i.'%y; A .; include the following: p?y 4
w; t - L;.s. r qQ. . yy e 9.; w0 h Ngular meetings. W5$$ representatives outline eccent.g developments and maintenance / design recosumendations. Any . , special . concerns of the' utility can' be ' addressed in fo11Cne. up correspondence with the NSSS supplier *$ service depart-ment. o Bulletins or advisories from the MS5$ supplier's service department alert client utilities to special protlees experienced by similar plants. Typically included in this correspondence are a description of the problem and the corrective actions taken to resolvt it. Rec ontw adat ions - for preventive actions or for particular cautions to be considered by the utility usually are included. o Owners groups provide an additional forum f or the enchaage of information that may be of generic interest ' to meaiDer - utilities. For example, problems in the design' or oper. ation of a system or component. may be shared.with tne gNo ' e
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and potential resolutions identified. The owners groups'
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ef forts of ten are directed at seeking improvements or anti-cipating problems rather than be ng only rea:tive in natsre.
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3.t.a se g.,' at : , a port e i mg sees i rene.t g . Otwe esist lag toweces of tafo**4tte are t v se ,* **s.?*et resv i t f rom t he tutC 's report t a, r,,s t r eme.c s , . tarse em.:s tac twee 10 Cra 21 reports.10 Cr# 50.55(el retort s, t ice see i
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archttect-engtnotr$. constructors, vengers. and anaa w'actsrees - assecf etsd netth nuclear factittles., 10 CFR 50.55(e) reoutres that the holeer of a coastr c"w permit net tfy the IWC of eech deficiency found in testpa and construction, which, if vacorrectea, could af fect the safe . operatten of the nuclear power plant adverself . 10 Cra 50.73 requires the holder of an operet tag licease for
~
a nuclear power plant to submit a Licensee'fecet Report uf E) for events aescribed in 50.13(a)(2). These Leas are incor. porated into the IsIPO Ltt esta base, which provides informa-tion to teentify and Isolate precursor eveats sac ideatifs emerging trends or patterns of pot eat t al saf ety sige'ft:aate.
;The astC Of fice of Inspection' and Enforce
- ent (![) issues
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,/ various documents, including bwllet tet er i informat ion
/ notices, to inform licensees and construt' ton perett holSers of $19ntf 4 cant concerns that may result f on the leaC evalue-j ti or re, ort s , a s re . i re. ., i o C r, z i . n . so. s s ( e ) . a <
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- n: e*: t w *N *, ,*e.; 'a est' tete t 9 reestt*tity or tasse eerwuneMat':-s. It '>,w feas t e t e. ea t ac ' ese-t et t oa p rogram s*N i f De se re l DeaP f .
3,7.1 tansaceer*ts to are05 e The present deffettien of Camponent in Wd@S (estracted from
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- IIlf 40M900) 56 88ft appifcatie to electrical c.ampenents. .
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i The deftattige h, r ~. ,..ld he .9eyteegd to Arttribe sectestget
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a. 0 75w prese*t faIivre reportiag pildence asees taprovement .ta tsie'fetlowing stees: T
.. Eu v deme 15 meeded - t e peevi de het t er. t af ermet i en f or 4*alpring the rels. of piece pertt et a facter 'ia castlag Covt>vit f4Sitere6. _
.. The g,igage shog1g te rg,1'6ed to indicate 18tet ut t19t tel showI8 Supp1y infeeneifom sette 1modegeete vender iafprus.
tio* 15 tOentted as e Cowle' er ceatritwt tag facter' tm a fativ re. The pside*Ce showid provide users' of the date base tae abtitty to retrieve react ly thole 'f ailuret t a vo t e' ag inadequate ve*ser inf ormat ten - (taegle s ey . worc s etlag.,Codtag). Prese.t . f ailur e report s are of f ee saetcafie provistag detet!S of the f ailwee analysis coad.icted by utilities. Tot 9,, * ::46Ce ' 560.lc em;*a s tle the import ance cf pro'e t ci ng - more c omplete resul t s of f ai lure ' analylls whe+ ene . t l' Conc.*!e3. Althow?M'Otta11ed failure analyses Gre.act. l 17 9 a 4
a . atys conducted for every ~ f ailure, when -they 'are co.. . dected they should be provided in NPRDS failure reports.
-In this way, the SEE-!N Program' and. other^ utilities .3a derive more benefit from the work 1 oficach ut111ty.
. ~
t .t ?..o. Utilities should develop interna 1 Leethods to.~ ensurepan m _A 1
;.mi.thef e,: ,, et PRD$ N reports are clear and complete end that'th, ...al; myM#aar, . -
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f:7p,yrogeam guidancebts ?i'011omned appropriat'ely. . Q'h ; gW lbi
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to?- % ; provide a complete failure description.within' the time frames for. reporting to NPRDS. ifft f lities sho'uld 'still' submit preliminary failure reports within the established time frame. titilities should revise;these reports when t9e necessary information is 'available. However, the present systeft does not provide methods for utilities to indicate. J that reports will ' be revise: later. NPRDS- should be r.odi-fled to permit each utility to readily identify'which of . their reports still requires follow.up . information. . Ut t lf-ties should report a failure event promptly' and include' an-initial analysis. Detailed and complete information shoulo be.provided in a timely manner once final analysis-has bee-- completed. l o The present scope of NPRDS reporting 'may not-meet all the needs of individual utilities for monitoring the rellak bility of their own safety-related components.. Each utility 1 that decides that additional systems and components 5hould ce i added to their basic scope of NPRDS systems and components should request that INP0 accept these systems. INPD will consider tuese requests, identify the additional' resource requirements needed to handle these requests. and notify I utilities when it is able to-accept additiona'1 ~information.: _ _ . _ _ _ . _ _ - _ _ . - _ - - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - - - - ^ ^ - - ^ - ' ^ ^ ~ ^ ^ ^ ' ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ~ ^ ^ ^ ^ ^ ~ ^ ^ ^ ^ ^ - ^ ^ ^ ' ^ ~ ~ ~' i
l 3.2.2 Enhancements to SEE-IN 0 Reports should be generated .for potertial: failures caused - by f aulty:or missing vendor-supplied information ' or other ET1. . The VETIP recognizes' that the utility will. uncover errors in ETI (e.g.. during review of the. information,. writing.cf instructions, testing, etc.) before anyone else. . It is. reroepended that ET!' faults be reported over NUCLEAR NETWORK for review by INP0 under the' SEE-IN ' Program,
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h. 4 ,~ em - .. , , ,. ,, , ,. yo The SEE-IN Program should.be broadened by]NP0 tol.1a% prove g 6 SU the ability to trend NPRDSi dat'a.CPresent methods of iridb 7 % D- E.~% are largelyl qualitative and subjective. in natur fhe
, s. a .' -
9 depend largely on the ability of analysts'to recognize the need to look for degrading or: unacceptable system and com-ponent reliability. INPO should develop methods to use NPRDS in a more quantitative fashion to detect trend prob- .. Iems. inis enhancement is presently under' development.by INPD. 3.3 Sumary Eranple One problem that led to the Salem event was' that the 'information con-tained in the NSSS vendor technical bu11etin'(issued-in'1974) was not processed appropriately and therefore not incorporated into plant procedures. If the systems that comprise. the' YETIP were functional
' ': ')
in the early 1970s, this oversight probably would not- have occurred or would have been rectified. Westinghouse had prepared the techni-cal bulletin based on a precursor event that occurred at another nuclear unit. This type of precursor avent would have required that an LER be written and submitted to the: NRC; INPO also 'would have-reviewed the Westinghouse technical' buitetin and the LER.- The cur-rent criteria for significance screent 9 used by 1%PO personnel. j identify this type event as a significe9t single failure. It is--
\
I highly likely that ar .,r.R would have been generated by . !NPO and. disseminated to utilities via NUCLEAR hETWORK. Utilities. wouid have reviewed the SER through their operating experience report review . proyrsas.
'I .
m l In addition, utilities would have had an ongoing program with their - NSSS vendors to obtain ET!. Utilities .would have had. systems..in place:to track and process this information. . Therefore, there are two pathways that would have ensured this type of information was received and evaluated by,the utility:
~o' MPRDS/SEE-IN.(SERs 50EAs). ,
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io ".NSSS vendor technical' bulletins
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lw 4 IwhEMENTATION Of VETIP 4.1 Responsibi11 ties for implementation 4.1.1. Utility implementation' Responsibilities-4.1.1.1 Esisting Programs s em- N555 Vendor Contact
.cn ._... a? ,. -, m MW, ;Each uti11ty should Stave a program ia place .wi20til15 . . . . - d '?
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bf. j TJ / program consists of 4 technical. bulletin: system' and.".h. *
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ettwiththe,NS$5,M .3[h & {
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'h ' s ~+R' Each utility should indicate or reaffirm its active 1 participate on. in. tWNPRDS and SEE-IN Programs. .The utility should : supply the',necessary besic.'.information anti should report failures.and problems 'on AItimely basis. ' Adequate-internal, controls should be in ' place to ensure that this activity is timely. consistent. .
and controlled and should include incorporation of'
~
future revisions to these programs.- Other Veedoes Each utility should continue to ~ seek ; assistance and Eil from other . safety-relatec ecaipmentivendors when .
~
the utility's evaluation of an equipment or ET! , problem concludes that such-direct interaction is-necessary or would be.. twneficial . These' problems and-those of. lesser significance will. continue to be ~
/ reported by means of the NPRD5 Lead /or the SEE-lh Programs.
./
I l: Internal Mandling of. Equ$pment Technical s Informati'Jn - 1' The utility should process .'incoa*ng ET! so .the objec-tives noted below are achievec. h -_ __i____ _-____.____mm_____2_---_ A-_ - - - _ - - _
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o Administ rat ive procedures should' provide.contr:*. i of incoming ET1 whether it arrives 'directly f ro-- the vendor or from other industr.i or: regulatory. ) a sources . (i.e., ~ NUCLE AR NETWORK, NPRDS, SEE-14,E: l bulletins, etc.), so 'it receives the appropriate j engineering / technical review, evaluation, and distribution for the following-
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t t 99d apcompt. esernings to key personne1'. .. g' gh, 1 L.; ,.. . e.+ a .m r~.s.; w p:s. s; m. , :+ r - -n .:.
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~:perchasing; records, and training programs k ;
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future procedure review and: revision cycle's -
-- notification on MdCLEAR'sETWORK of significa-t ETI The incorporation of such safety-related infor-- ]H mation (or change's) remains within the' scope o' .j the utility's review and approval requirements. '!
J o The administrative ' program should require that maintenance or operating procedures cite appro-priate ET! in the reference section of the proce- . dure. ; i o' Within the performance se: tion of the procedure, . .; appropriate ET!. should be incorporated and approved th the engineert 99, technical. and. 'l quality; review of the saf ety-related procedure. {
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_ . _ . _ . . . . _.m.. _. __.._
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o Intemal Handling of ' Vendor Services The vendor, contractor, or technical represen-tative who will perfore safety;related services. -1 should bela 'QA approved / qualified supplier of such nuclear safety-related services. Furthermore, the1 g services should be specified in the. proc'urement' .)
, ,7 -l documentation 'so that a combination.'of procedural o - l
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, ' NWA weerfor service may(te performed usik '~%,g' g.utilij y gp . , ,
XX .prochs. If 'sc, the procurement documentation j should specify that the service is' performed using -!_ uttitty procedures that have been approved after a . technical and quality review cycle typical for other utility service, maintenance,L repair, er f operating procedures.. As an alternative, the : i i serv 1 e may be performee using ven6er or \
.}
contractor. procedures. 'In this case, the 1 docume~tation should.specify that the service is ; performed using vendor, contractor, or technical-representative procedures' that have been reviewed and approved in accordance with the. utility ; procurement program,L O4 program, and administrative review program.' .This is to ensure ~ that' their. documents are processed and approved in a man 6er equivalent. to the utility procedures conceming siellar activities. ' , 1 le addition to specifying the' procedures that will ' be used, the QA/Qc program to be used should also
~
be specified. The utfif ty 0A/QC program may be: -l wSed. In this case the procurement documentation should specify that the activity. will' be performed under the cognit#nce of the utility 04/QC t program. Altematelyi the vendor or; contractor na ny nenner a*v he ,, us . -in ehte cae. th. i
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. doc 9e-tat 'on should specify that trc act ,*:, i will be pe* formed under the Cog *12ence' of :"e1 vendor, co6 tractor, or technical. represen:ah,,.
QA/QC program tha't has been reviewed separgIe1,;., and . approved in accordance 'with the utili4f,A - program. In addation, during the performa*ce.3f 2 r
. ff the offect1veness of, their.perfoyace 7
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2m r ' compliance mith' tts approved program my santa:! erg,j:c.,5f. -
~h k surveillance, inspection,andaucith?f? M.N;.; / b[- . .,.. y y.mMn + ' '
y ~W 4 x , # 4.1.1.2 Enhanced Progrees _ _ o %DRDS Each utility should incorporate the. enaancer.:r to the %P435 recomeaded in Section 3.' 2. . >3 could mro),e revisicas -to esisting ad'inf st a: s progrees or procedures.- It also coald rey *e revised trat*tng'or other _ actions ~ heeded to eas. + a meaning *wl and ef'ective ihpleavntat1on of **e NPRD3.progra enha7 cements. o SEE-lh
. Ecch utility should incorp; rate' the enhanceme*t s -
to the SEE-lh' program recorrended in 'Section - 3.2. 'As in t9e NPR35 program, thisL cowls ta C v . 'e. revisions to esisting admin 3strative program " procedures or to tra tning,o ;'other activities s': the data reported to the SE!.jk Program is L co - plete.and deta11ed enough 'to- Support the syste ,
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enhancements being undertakt9 'by INPO.; ' i '
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4.1.2 1NPO Implementation Responsibilities' 'I o Esisting Progrees b
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t The NUTAC determined that present: NPRDS/SEE-14. Programs.- properly used, currently provide an adeq0 ate framework f3r ' the effortive e .
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e' o' [nhanced Programs INPO should implement the enhancements of; the hPRD5 an: SEE-!N Programs (noted in Se: tion 3.2) to augment this , VETIP.. 4.2 Schedule for laplementation
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. 4.2.1 Emisting Programs '
, . g w:$' (MejdA=prQ ;g(titt11 ties that find that their esisting internal program ana
' e7.Z ONE 2, Mpresiedures do not" support those evtlined .in.5ections>3.1 and p'~D.. .s. , *.; y ,
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&~.' - > Q4.1.1.A above should make the. aecessary timely revisions as -
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> :-part of the established review and updating ~cycIe forf such i documentation. A specific schedule should be established by.
the individual utility with a target date for. full implementa-tion by January 1,1985. 4.2.2 Enhancements to Existing Programs 4.2.2.1 lhPO snould work with the NPOS user's group with tw goal of establishing schedules by July 1,1984, for l implementation of the enhancements of the hPRDS ] program. l 4.2.2.2 Utilities should incorporate tne enhancements to tne . NPRDS and SEE .IN progrees, . recommended'in Section 3.2 and 4.1.1.8 above into their internal program anc procedures on a timely basts. 4.2.2.3 Schedules should be established that are consisteat- i with an overall goal ta inclement the recomeacec l
/ enhancements to both programs by January.1,1956. - 'I i
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5, h j i Non-NSSS I venoors ; i i I i IES TEN LER 10 CF R 21 - , i 10 CFR 50 55e Ge'wrsc Letwrs Q Notification ,
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;' 04 GENERIC LETTER 83-28, . '
SECTION 2.2.2 s 4 i 1 1 i 4 j
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$PECIFIC CHARTER FOR NUCLEAR UTILITY Tass:. ACTION C0mlTTEE ON GENERIC LETTER 83-28 SECTION 2.2.2 i
'I This nuclear' otility Task Action Committee (NUTAC) .has been established by .a , g group of utility representatives who have recognized a need for nuclear indes try guidance on Generic Letter 83-28. 5ection 2.2.2. . The establishment of, .~ .
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this NUTAC *as been in accordance with the general charter governing the'_; ;G& j organization and operation of a NUTAC, as approved by the, Institute of Nuclear j Po er Operations (INPO) Board of Directors. This NUTAC is committed to ccm. ] pliance with this specific charter, its bylaws, and the general ~ charter. This. ; charter his been reviewed and approvec by the. chairman of the Analysis and I Engineering Dtvision Inoustry Review Group and the president of thPD, and the president of 1NPO authorizer staff support for this. NUTAC. :l 1 l This committee has adopted the following objective to ensure fulfillment:of ] the goal of aMieving industry consenses and guidance on Generic Letter 83 28, ( Section 2.2.2: ] o development of guidance for use by utilities in response to Generic Letter 83-28, Section 2.2.2 l To ensure that this objective results in products'that are of generic benef tt to the utilities, voting memoership on tnis committee is limited to permanent- ; employees of U.S. nuclear utilities. The chairman and vice chairman of this comr ittee will be elected by the NUTAC f rom a list of candidates approved by the chairman of the sponsoring IRG. To further. ensu e that 'this NUTAC ' pro.. j vides products that are of generic benefit to utili, ties, the NUTAC chairman will maintain close liaison with the sponsoring IN#0 Industry Review Group.
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6 A:s't tonelly, tais %.' AC sneu f d es tabi t s" liaison wita ci e recognizee. i : 5 , try groups, such as A!F. ' A%5. eel. EMI. and N555 owners groups and will l maintain communication on this ine stry init iat i ve. with tae 4RC, as apprope t. ate. 1
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3 .. r, Approved 11 as n Af fj% SI Chairman -WTAC . y Date Chairman, IE Date ' ' ' . '
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, 4 'o LIST OF REFERENCES #' ~I- .
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e n yr ', a .ft. },$ l 3 t a r- - j, s s re: A: pans Based on Generic implications of Setem ATW$ (vents *. e f (Geh oc tettir 8M8). Washington. D.CW ' L'.5. Nuclear' Regulatory . \
Coc>hsio6.'p'AM 1983-s
- 2. NC Fak t-f tedthg 7a 't N.jrFr Report on the ATWS Events at Sale,m Octear
/ finerating Station (nit 7. on February 22 and 25.1953 [huREG-09/7). .
+
Wasnington. D.C.: U.5,' hucleaf Re991 story Commission, March 1983. t x <
;u
} _ J. Generic Implication of <fd*f$ Eventsi et tg,5p]eg Aclear Power Plant . (buREG-1000). Washingmi,7.C.: - U.S. Es im; Asgu16 tory Comission,' < , :%. 9 2 April 1963. - ' mj i tgA 1 ,;, x s4 4 0.QB!/ ' O ' W;
< [? kg" $f'_
' 4~4T 'Significe : Eve 4r
- t ' j J hPD 83. ffT.a,Atlant a , Ga . :
baluatien and Informatica Itety (SEE-feO g . j u
; Institt:teW of *MtA nucThiU)wer pfFretionI# p" r.,,
~ g
, ', #v.februsty
% -8961, @ l , - '#
- 5.
.. iw - , , j- 4
%uclear /plaat Re?J4111ty Tht a' System (NPC31 ReportdM? System and C6mpo.
cent 5c5E%nual (ihPD 53-D20). Atlanta, fe,: Institute of, hoc 1ga t 4 - c/p \ . powerpg6pFidrT,*i1N3.. (- 6.
;].{ f. 8 Reporttec #roceduresAtlanta,Manual Ga.: for Octear Plaii?. Reliability Ma Systems ' N 4
; G%EO B4DIT)-
1984 Institute of.%cicar Power operations,- y 7 10 Ct2 21 Coe of Federal Requ'titions: i
' t' Defects anc honcompi1ance."' W4NiQton, D.C.:
Title - 10 Part 21,
- Report ing ofl # '!
- 07. ' :e. 1 U 5. Government Printing ', / % -i V( (
13 CFR 50. Code of Federa k Requ1ations: Title ~10 Pa t 50. ~ Domestic Liceesta; oT Procucpon amt u'silization Fac ilit ies.' Washington . D.C.:
- .S. Govemet Prlw ingn!f s ce.
('"
, . .d g
'j 9 ( .qgo
' StaasaYa f;3-50). Ceiteria for Safety Systems for %:leae Gener46ng Stations (IEEl he. Torr, k.f.:
E eg'neers , 1980 J. stitute of Electrical and jf,y;tr6nics i o )
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- 1. Provide basic report of plant event (utilities).
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- 2. Screen events for significance and transmit l Significant Event Reports (SERs) via NUCLEAR NETWORK (utilities and INPO with. vendor input solicited l when specific product is.~ identified).
' e. ,v M
.+< .- ' . ;Provide backup data .on contributing factors and .; probablef couses Jed'. con-o.x ; y ... . .. ,
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vggri"m- l-r -b,,er ' Perform'w. ar.tton anal'ysis on significant events to. eval'uate possible options *t'.
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~ %;for short-term remedies and feasible long-term solutions ~ that might' be . '
implemented (utilities, INPO, and ' vendors).
- 5. Disseminate information, along with an alert of potential implication, to the utilities (INPO). 's i
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- 6. Evaluate the info mation and implewnt remedies as appropriate (utili-ties). :
i 7 Provide feedback on implementation actions (ut tlities and INPO)..
- 6. Evaluate periodically the ef festiveness of the process, including steps . l l-7 above (INDO).
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,..- GENERIC LETTER 83-28' Nt;hg7;; -
- nyj'L SEC710h 2.2.24 ,
(Generic Letter 83-28, Section 2.2.2 is er:losed verbatim) . 1 I i i t i
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r UNIT r o s t AYt s P. JC L C A R A ( G U L A T L E v C OWi' SIO'.
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6::EME A4D HDLDERS 0F CONSTRUCTION PERMITS g, Gemtlemen: N b: 5%;ECT: ItEQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SA'.EM hyQq . . ATW5 [ VENTS (Generic Letter $3 28)' yp;c. . .c u- m - ;, a pw w. mThe consission has recently, reviewed intermediate. term actions to be taken by
~
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J N1(censees and applicants as a rescit of the $alen' anticipated transient without. Y screr l ATW5) events. Dese actions have been developed .by the staff based on : te*:mation contained in NWREG-1000 '
- Generic' Implications of AW5 twents at
'",m tw salen Nuclear Power Plant." These actions' address issues' related to reactor
;ri: systea reliability and general management capability.
.e a:tions covered by this letter fall into the. following four areas:
1 Post. Trip Review - This action addresses the program, procedures and sata collection capability to assure' that the causes for unscheduled reactor shutdowns, as well as the response of ' safety-related equipment, see fully understood prior to plant restart. 2, !;ai;r=ent Classification and Vendor Interface . This action addresses the
;*ogra s for assuring that all compo'wnts necessary for accomplishing
*e:;uired safety.related functions are properly identified -in documents,
;ro:eL*es, and infomation handling systems that are used to control safety.relatec plant activities. In additioe, this action addresses the establistrent and maintenance of a program to ensure that vendor infomation for safety-related components is comolete.
2, r ost Nintenaate Testing - This action addresses post-maintenance operability testing of saf ety-related components. 4 Geactor Trip System Reliablity _ Improvements - This action is aimed at assoring that vendo*. recommended' reactor trip breaker modifications and essociated reactor protection system chtiges dee completed in PWRs,_ that a comprehensive program of preventive' maintenance and surveillance testing is implemented for the reactor trip breakers in PWRs,' that the shunt trip' 4*tachment attivates automatically in all PWRs tha. use circuit breakers in their reactor trip system, and to ensure that on-Itne' functional testing cf the reactor trip systm is performed on all LWRs.
. e e= -
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n j 1 1 t e e ;':s., e to this lettee : coa s' kg !* e s e 4: 'c- t '" *. r s a.e e ' . ; :c 4-k s .
% 1: '1ed t hat . all act'cas, ence!K o.r (attice.'.2. 4.1, c.3. arc.4.E , !
re:c e 5:f t-are (preceh es, tratris ,'etc.) cea ;es 4-:':r :- " tat e s 4 d :c a- : a'fect ecutreat cnanges or req ire rea:ter $~ ; tee. to co .p' ete, j H l
- ti:n 1.2 ay result in s: e changes t o the se;,,ence of events reccrder or en'st * ; plant computers., tot will not res it in a pl nt sn teo a n to irrpt e ent, g Acticas 4.1, 4.3 and 4.5.2, if applica:le, would regst re the clart to be .. .
fj$*:stoonn . in order lte implement. ag @@d b.
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, jf' :;,s,,.T XfThe reactor-trip;&ystem is fundamental to reactor safety for 'all %:)sar pomer -p- @pleet des *gns. ' All transient- and accicunt- analyses ' are predicated lo49tt ' _ "Q'i successful cperation to assure acceptable consequences.
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,[n The*eferet the' actions -
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. .Itsted te'ow,-watch relate directly to the reactor trip-systea, are of the c "Mg*est priority and should be integrated into existing ;1 ant. Scr.eeules-fi st. .%ggg@$
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. r" a 1.1 Dost.* rip Review (Program Description and Procedure). I I
2.1 Eauipment Classification and vendor Interface (&ea: c
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I Systen Cor ponents) 3.1 DostMaintenaece Testing (Eeactor Trip System' o p;ae-is t.1 Reactor Trip Syste : Rellatility (Vendor-Related Mcci f t:ations)- 4.2.1 and 4.2.2 Reactor Trip System Reliability '(Preventive.
- maintenance and Sarve111ance Program. for Reactor Trip ! e eners) .)
1 a.3 Deactor Trip Syste* Reliability. ( Autenatic Actuation of 5*unt-trip Attachtment for westinghouse and B&4 plants)
Dst of tae reeaining inte**ediate. term actions concern all other safety.
related systees. Trese syster:'. while .not sharing the sa-e relati ve import aate - , to safety as the reactor trip System, are essential in teltigating the conse- R csences :,f transients and accidents. Therefore, these act.t oas SC C be- l 1-te; at ed into existing plant schedules over the _ longer. team 09 4 medlun. j
.pe1cet ty ca st s. Some of :ne actioa.s discussed in the enclos sre e. tit. dest te served t %ners' Groap participation, and this is encourage d to the eatea: -]
**act1Ca'.
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Accordingly, pursuant to 10 CFR 50.5a(f), operating reactor licersees and appitcants for an operating license (this letter is for infe matton only-for 'those utilities that have not applied for an operating iicense) are reasested to furnish, under oath and af firmation, no later t'an 120 days f rom the date of this letter, the status of current conforwance r th the posit tens 'e containec nerein,' and plans and schedules for any needed improverwnts for : conforman:e with the positions. The schedule for the implement'atton of these improvtce*ts is to be negotiated with the Project Manager.
p . 1 . ls [ e:cesees regst red and applicants inf orma may ti on. Such requestmust a request an eateasion set .forth aofproposed time for schwste
'subrnitta's c'
-c justi fication for the. delay. Such a request shall be directed to the
.*e: tor, Division of Licensing, NRR. ' Any such request must' beisubmitted
-: later than 60 days from the date of this letter. If a licensee orlappl u. int .
I c:es not intend to implement any of the enclosed items, the response-shoulo t 5: indicate and a safety basis should be provided for each item not. intended- -
- og implemented. Value. impact analysis can be usec' to support such responses c- to argue in favor of alternative positions that licensees might propose..
& s- .
A ;
$!/ 3,rr r 4tperating o Reactors,'the schedules for implementation of dhese actions shalif d,, I d.# - se seveloped consistent with,the staf f's post of;1ntegrating new requirements.E Wss-ce-sidering the. maique-status of each plant 'and the relative safety. importeweir#11 of the improvements,Scambined with all other existing plant programs. Therefore,1y%
- -semecules for implementation of these actions-will .be negotiated between the a f W
~ ;V: Project Manager 'and licensees. QbW;
- g. '- ",
, y v .
* for plants undergoing operating license review at this time, pl'antispecific ~i*
scwdales for the implementation of these requirements shall be developed a manner similar to that being used for operating' reactors,'taking into ' c: sideration the degree of completion of the power plant. For'constructipi
- eet holders not under OL review and for construction pemit applicants, tre requirements of this letter shall be implemented prior to the issuance D c' en operating license. j h
is reque t for information was- approvec by the Of fice of ' Management and ' .i S.cpet under clearance number 3150-0011 e ich expires April 30, 1985.
- re.ents on burden and duplication may be directed to the Of fice of -
N age ent and Bsdget, Reports Manage.nent Rocyn 3208, hew Executive Of fice
!.*:cing, e shingion. D. C. 20503.
Sincerely, G >
$Y LL.
Da rrell G. 'Eisennut . Direc tor. 01rision'of. Licensing .j E9c'esure: , Se .sred Actions Based on Generic belications of Salem ATWS Events i
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1; . .. 1 EN:.Osst' ' GEN:s:0 AC'!Oh5' E ASEO Oh GENER: C IMPLICATIONS OF SA.ER A!WS EVENTS 1.1 POST. TRIP. REVIEW t RDGRAM DESCRIPTION AND PROCEDURE) { Position 1,1censees and applicants shall describe their' program for ensuring. . j that unscheduled reactor shutdowns are analyzed and that a determination 1
" is made that the plant can be restarted safely. . ' A report describing.the - !
MJ . ' , program for review and analysis,of such' unscheduled reactor, shutdowns 7/yp . j
, should include, as a minimum:, , 4 u NS q 1 M b,"
Pew Eu ^" ~ #,?
$ffV- 1. The' criteria for determining the' acceptability of'tyrstart. f: "
' l J* W? . .
, ~%^ *- '
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- 2. ~ The responsibilities and authorities of. personnel who will;~ @, # m.3M_3
$i %' '
perform the review and' analysis of these events. '4 3- ,
- 3. The ne:essary qualifications and training for the responsible
-l personnel. J
- 4. The sources of plant information necessary to conduct the review and analysis. The sources of information should include the
.q measures and equipment thet provide the necessary detail. and- !<
type of information to reconstruct the event accurately; and in i suf ficient detail for proper understanding. ~(See Action ).2)
- 5. The mettods and criteria for comparing tne event information with j kno.n or expected plant behavior (e.g., that safety.related equip.
rert operates as required by the Technical Specifications or other performance specifications related to the safety function).
- 6. Tne criteria for determining the need for independent assessment of an event (e.g., a case in which the cause of the event i cannot be positt vely identified, a competent group such as the Plar.t 0;erations Revie Corrrnittee, will be consulted prior to, authorizing restart) and guidelines or, the preservation cf _ physic'al evicence (Doth hardware and software) to support independent -
analysis of the' event. .I 7 Items 1 inrough 6 above are cons 1 cered to be, the basis _ for the j establishment of a systematic method to ass,eis unstneduled reactor j shutco ns. The tjstematic safety assessmertf or.:ecures compiled ; fror- the above items, which are to De used ir conducting the ; evaluation, should be in the report. >
/
1 A;clicability . s
/ ;
This position applies to all licensees and ft applicants. i
a 1.
- L Type of Review For licensees, a post implementation review of the program'and. procedures w111 be conducted or. the- staf f will perform a pre-implementation review if desired by the licensee. kRR will perform.the review andl1ssue Safety Evaluations..
2- .
-f;1tcensingschedule.For h! OL applicants, the hRR review will.' be perforised g g, , con
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;gg S) Dec
,Mistion Required .. _I < - . ~ ' + '
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- " Licensees and applicants shall submit a repoit describing their program-
.sedressing all the items in the posities. 'g .
, j,. : ~ jg,Q + :N :'y Techncial Specification Changes Required me changes to Technical Specifications' are required. -
References ,
' Section 2.2 of NUREG-1000-negulatory . Guide 1.33 AmSI M18.7 1976/ AMS-3.2 Ites 1.C.5 of huREG-0660 10 CFR 50 - 50;72-
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,,- FOST TEM REVIEm . Dt.* A Ahl t hf 0RE'10h CtJ AB!ll!'
l P0sition
,1censees and applicants sh'all have' orl have planned a Capability to redord, recall and display data and information.to persrtt diagnosing 19e causes of unschecwled reactor shutdowns prict to restart:and for es:ertalr.ir.g the proper functioning of safety-related equipment. .
Adequate data and information shall be provided to correctly diagnose the cause of unscheduled reactor' shutdowns and the proper functioning . .. a I
'cf safety.related equipment during these ewestts using systematic safety - QMh q s 'assesseent procedures (Action '1.1).. ;Ine data lane information shall .be 5 i
.stsplayed in 4 form that permits ease 'of e' assimilation end.' analysis by . @7 j
. persons trained in the use of systematic safety assessment i 7 procedures. . .
l
. , .4 i%,' .;
- 1 L report 'shall be prepared which describes .
. : l % .l k n .
'A stifiesandfju,%
the adequacyLWlW dj ' of equipment for diagnosing an unscheduled reactor shutdown. The report ;'. , small describe as a einimum < I. Capattlity for assessing sequence of events (on.cff indications)
- 1. Brief description of equipment (e.g., plant computer,-
eedicated computer, strip chart)
- 2. Parameters monitored
- 3. Time discrimination between events
- 4. Format for displaying data and infomation-
- 5. Capability for retection of data'and information
- 6. Power source (s) (e.g. , Class IE, non. Class 'IE, non.
Interruptible)
- 2. Ca;at'lity for assessing the time history of: analog variaoles neeced '
i to detemine tne cause of unscheduled reactor shutdowns,' and the f unctior.tng of saf ety.related equipment.
- 1. Beief description of equipment (e.g. , pleet computer, :
occicated computer, strip' charts)
- 2. Parameters monitored. sampling rate, and basis for sele: ting parameters and sam;1ing rate .l
- 3. Duratice of time history (rrinutes before trip and -
etnutes af ter trif
l 4 Format f or cisplaying data inclucing scale -(reacodility) of time histories
- 5. Capability for eetention of data, information, and physical evidence (both naramare and sof t.are)
- 6. Power source (s) (e.g., Class '!E. non. Class IE, non. .
interruptabit) e .'- y cc . .. , u ., a .j
.j g
, M ne :.V. ..w . a _ s: - *
.- ungtp . . ., 4,;ty'"wa,,j e.~ _. . .
"f Y4 D ', 3.@C0ther date and information' Sik provided' topdWilj assess Unscheduled reactor
- M %t;='ij@ $ 6
". shutdowns.
pd ?M - 1 7
.~,~ 4.~j$chedule capabil.1;ty,4fofSany4
~
h%f M , anned changes to esistirag jPgg. data and Informatio
. , . . x
,Q
- olicability . g-; 4 g .
n- ,, This position applies to all licensees and OL applicants. "7
~
Ty;e of Review Data and information capability will De revie.ed by %R to ceteretr.e -
. tether adequate data anc information util be availacle to support the systesatic safety assessne-t of.unschecated reactor snutoowns. hER
.all perform the reviews and issue a Safety Evaluation.
For licensees, a post-implementation review of the program and precedures will be concucted by NRR or the staf f mill perf orm a pre-implementation review if desired by the Itcensee. For OL applicants, the haR review will be performed consistent with the licensing schedule. Occam entatier Required Licensees and applicants shall submit a report cescribing their cata anc inf crmation capability f or arschedoled reactor snutco.ns.-
'echnical 5: edification Chaages Required To be determined based on evaluation of required documentation.
Feferences-
. 4 Section 2.2 of NUREG 1000.
i 0 _ . - _ - - . _ . -m __
l .g spMihi CL A551FIC A110N AC vth:0k 1%'EEF A;E; (RLaCTOR TRIP $Y51EM CMON[hi$) Pes 1 tion Licensees and applicants shall confirm that all component s 'whose ~ f unction. ing is required to trip the reactor are identified as safety-related on-documents, procedures, and information handling systems used fin 'the plart-to control safety.related activities, including maintenance . work orders, and -
, parts replacement. In addition, for these components.' licensees and applicants; T u.. - ishall establish, implement and maintain a Contiaving progree to ensure that .li,05Y -vender .tnforsation is complete, current and centrelhed throughout the Itfe et &
D7" the plant', ano appropriately referenced er incorporated? tfa plant;1astructeems~g 7 Ond procedures. : Vendors of these components shosid be contacted ~and an fater.E
~
face established. mere . vendors can not be identified, have pone out of 7
$~ P
* 'c twsiness,~or will not supply the inferination,< the.1tcensee or appiteent a J shall assure that sufficient attention ts ~ paid to equipment maintenance, "
y$$.; #' replacement, and repair, to compensate:for the lack of vendor backup, to
" assure reactor trip system reliability. Tne vendor interfaca program'snell include periodic communication with vendors to assure that'sel applicable i-f ormation has been received. The program shovid'use a system of oositive fee::ack with.vencors for mailings containing technical information. LTnis:
coule be accomplished by licensee acknowledgement for receipt' of, technical' O. mailings. The program shall also define.the interface and' division of ~
~
responsibilities among the Itcensees and the nuclearrand nonnuclear divisions-of their vendors that provide service on reactor trip system components to assure that requisite control'of and' applicable instructions 'for' mainteunce work are provided. Applicability inis action applies to all licensees and OL applicants'. Type of Review For licensees, a post-implementation review will be conducted. .NRW will perform these licensing reviews and issue a safety Evaluation. For OL applicants, the NRR review will be performed consistent with the licensing sc:.:cule. Documentation Required Licensees and applicants should submit a staterrsnt confirming that they have reviewed the Reactor Trip System components and conform.to the position regarding equipment' classification. 1+ addition, a summary , j report describing the vendor interface program shall be submitted for. ,i staff review. Vendor lists of technical information, and the' techncial / information itself, shcIl be available for inspection at each reactor- l. site, t
r-i 1 Tecnnical Specificat3e (nanges heq tree no changes to Technical 5pect f tcations 'are reewtr'eo. Reference Sectten-3.1.1 of apets-1000.
+,If Section 2.3 2 of In4L5-1000. . l..n 1 -
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pceegges and a;; tcoats tra}* s. W t, for sta*f r e w t e. , a . de s c r i p t i on ' i I of their prog *ses f or sa f ety .re!sted* eq.W t c la s si f ic a t ion and
,easo r int e r f a e a s ee s c r ibec be t o :
1, for egetPatet classifIcattom, i1censees and apsltcent S Shall; sescrese - , tneir pro vree f or enswris.g that all components of safety.related- ~
.. N, /~; systems necessary for accomeltshing required safety. fernctlent ere , ..
hD I hteemt1f ted AS. Safety.related en documents, proceewrong eed informetsas: h. p a gti j$endisag systosej eted 14 $8te plent te Control Safety.lPeleted.-et$MLles' h ) Nd y gy$ tit ' - - 1
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'descript1 ste % %ecyWee:NQ aeelget18ers uy sad, rep gg @leCement M ,,pGgy 7
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- v Uj.i i peritnin systees lCereently 'Clesstfled. 46 &&fety.eeleted.u ' e 3 "6; . C fhts shall not te interpreted to retutre Changes in ' %% h_,.~ ~ *
' Safety ClasTTTicatio's et take systems level. ' M* , . 6jy -
v ty . A description of the reformation mancitng systee sses to
~
~ 2,
. seer.tify safety.related concorent s (e.g. , computert red - a 5
equipaent Inst) and the meteocs usec for its' development
,w and asitest1en, ,
.i
- 3. A Description of the p'ocess by weich stat toa personnel use this taf ormat ion rianct ing systet to deterwine that an activity is safety.ref ate: and erhat procedures for main.
tenance , surveillance.. ;4rt s replacement and other 'actie t ttes oefleed in the $ntrod ctice to 10 Cf R 50, A;penot a b, apply to saf ety. relate: c om;onents. 1
- 4. A Description of the adaapement cont rols utilt ted to vert fy- 1 that the procedures for gre;aration, valication anc rowitne ut i l lia t i on of t he I n f orma t i on ha nci t n g sy s t em ha ve bee n folio.ec.
$. A demonstration that a;;'3;rt at e design verification and q.alification testing 15 s;e:ifica f er procur ewnt of safety -
i relatec c or porient s. The specifications shall inclooe quali. fication testig for espe te safety sere 'ce c onditions anc pro. ice 56pport f or the licensees' receipt of testias cocumen.. -i tation to support the limits of In f e re:prvnenced by the supplier,
*Ia'ety.related structures, systems .anc :omponents are those that are reitec
.;rf. to remain f u ct n s'o tal during and fotioning design' basis events to ensure: ;
i;; the-integrity of the reactor coclant Dow n ca ry. (2)' the capactitty to .sh ! co<n the reactor and maintain it in a sa'e shutdown concition, and (3) tne ' 'Lj
- a;4bility to prevent or nitigate the C 'se:Wences cf accicents that'cc ld !
result in potential of fsite caposu es aceparable to the- 9Liceltnes of 10 CFE - . Part 100. i
^
- ~. 1 Type of Rev1e= For licensees, a post-imp ementation l review of the program'and procedures will be conducted or the staf f will perform a pre implementation.' review ] if desired by the licensee. 'hRR will' perform the review and issue safety Evaluations. For OL applicants, tne hRR review will be performed consistent with the licensingschedule.M. : ,:c . ;Q .f q Q,.y';;,;. ].; , s. > g . , 1 DocumentationRequired:}.}g;*. e 7-
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. ,, V c*K Licensees and applicants'c.-.shall submit a re'portidescribing .:their ..' program? "'.~/
i addressing all the itensiin the position. ' W" ~ N,h,*+ . um:n. p:..M. ;cx
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%4 Techncial,$ specification Changes Required - ,;' +9 r Jho changes to Technical Specifications are required.s /
R2ferences Section 2.2 of NUREG 1000 .- Regulatory Guide 1.33 ~ AkS! N18.7 1976/ANS-3.2 l Item I.C.5 of NUREG-0660 ~ 10 CFR 50 - 50.72 .
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; r;$i.*R!P RE.;En . DATA AC lhf 0RM.Ai;0h CAS AB;Ll!Y F05 i t ' on t scersees and applicants shall have 'or have' planned a capability to'. j recore, recall anc display data and inf orir.ation to peririt ciag sosing
~
ine causes of unstneculed reactor shutdo.ns prior to restart and' for. ) as:ertaining the proper functioning of safety.related equipment. q J
~
Adeqt.Jte data and information shall be provided to correctly diagnose. " i the cause of unscheduled reactor shutdowns and the proper functioning' ;,1
,E
+
' of safety.related equipment during these events using systematic safety - "
assessment procedures (Action 1.1). .The data and information shall be , - Y'
,^ .
cisplayed in a form that permits ease of assimilation and analysis by ' %g[y persons trained in the use of systematic safety assessment procedures. .f. .
- s. ' . .,, . . . . , . . . . . % s _ gW A rehrt shall be prepared which describes and justifies the adequacy 5c 5 i 3 '
of egoipment for diagnosing ~ an unscheduled reactor shutdown. The report c [" shall describe as a rinimum: .
)
I
- 1. C4; ability for assessing sequence of events-(on-off indications) 1 Brief description of equipment (e.g., plant computer,.
dedicated computer, strip chart) 2 Parameters monitored
- 3. Time discrimination between events
- 4. Fomat f or displaying data and information
- 5. Capability for retention of data and information i
- 6. power source (s) (e.g., Class IE, non-Class. IE, non-interruptable) i
- 2. Ca: ability for assessing the time history of analog variables needed )
to determine the cause of unscheduled reactor shutdowns, and the functioning of Safety-related equipment.
- 1. 3rtef description of equipment (e.g. . plant computer, dedicated computer, strip charts)
/
- 2. Parameters monitored, see.pling rate, and Das 's for selecting parameters and sampling rate
- 3. Duration of time history (minates bef ore 1 1p and :
minutes af ter trip'
\
1 9
r 4 t 3:pMINT CL A55 :F IC A'10h AO vih:O ]%'E M A;t. PI A; TOR ~ Thlr 5151tn - C M ONINTS) Pc'sition Licensees and applicants shall confirm that all components whose; function. Ing is required to t*1p the reactor are identified as safety.relatec orc documents, procedures, and infomation' handling systems used in the. plant to control . safety-related activities, including maintenance, work . orders, and parts replacement. ]* addition, for. these components, licensees .and ' applicants ' shall establish, implement and maintain a continuing program to ensure that vendor inforination is complete, current .end' controlled throughout the life of.. ';; the plant, ano appropriately referenced er tacorporated in plant instructions and procedures. Vendors of these components should 'be centacted and~ an inter.
- face established, mere vendors can not be identified, have pone. OutLof W f' business. or will not supply the information..the licensee erospplicant. G.T
'shall assure that suf ficient attention is'~ pate to equipment maintenance', '
N replacement, and repair, to compensate for the 1ack of vendor, backup, to assure reactor trip system reliability. The vendor interface program small: include periodic communication with vendors, to assure that al' applicable. trformation has been received. 'The program should'use a system.of positiveL feenach with vendors for mailings containing technical information. :ints' , could be accomplished by licensee acknowledgement ' for: receipt of tect.nical' mailings. The program shall:atso define.the interface and division of' responsibilities among the licensees and the nuclear and nonnuclear divisions of their vendors that provide service on reactor trip system components :to assure that requisite control of and applicable instructions for maintenance work are provided. Applicability Tnis action applies to all licensees and OL applicants. Type of Neview For licensee:,, a post-implementation revi+w will be conducted. NRW will perform these licensing reviews and issue's Safety' Evaluatt.on. For OL applicants, the NRR review will be performed consistent with theL licensing schaule. Documentation Required Licensees and applicants should submit a statement confirming that 'they have reviewed the Reactor Trip System components.'and conform to the position regarding equipment classification. In addition, aL nummary . l report describing the vendor interface program shall be submitted for , staff review. Vendor lists of technical information, and the>techncial; l inf ormation itself, shc11 be available' for inspection at each reactor-site. l
1
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?eW 'c ai ~b e:*: stic- ;ts'1es 6e;.~ rec.
w, changes t ol fece.':# F 5pe: 1f ?cat'<oes'are,rea.,P ec. Reference- ..- Sectton 2.3.1-.of huREG-1000. - o1
' Section' 2.3.2 of MURE3 1000. :
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- p.e e t c l a s s i f s c a t i o^ , I t r e n s ets : an t a;;t c o at 'n ' sha l l ec s c r i te their prog ram f or ensurlAp that 41) compore**ts cf $4fety.relatt0 siste-s necesse*y f or accomplishing reeveres safety functions are tsee.tif ind ics saf ety.esisted ori sucpeats, procesares,. and inferNt 984
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$n.r.y. l eh, Mhe Criteese for 1deeUfyths touwene9l&Let . Safety-fel~ette1 G , nW 1< N;4 Q.$ 4mithir. 6pstees- gverently Classif. fed 45 Safety. '
'#[v b %' @j relates
- p r < . Tass shall not be saterpeetes to retutee changes trif ^, , ~%g ;W gf '
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r$sfety clasIITication at the syste<ita level,- ~ , l' '
, .y 2, a description c t %e inf efw.at ten Aancl u g sy stee used to Sce in fy safety.related comweets (e.g. , computert red es. ireat I n st ) a re the met e o:s usec f s+ i t s se mel opaient -
sat va t icat toe.
- 3. A sescript ton of the process by etch stetten personnel. -
.se this u f eraation hancling syste* te deteamine that en act seits is saf ety-relatec and e. rut procecores for main.
tenance, surveillance, part s reg,'ecement and ether actie1 ties oefinec in toe tetroc.ct io*. to 10 CFA $*, A;;endia b, apply to saf ety.re: Atec com;onent s.
- a. A ces: r t pt iori cf the edesgewnt co9t roi s 5t ilitec to veri fy t*at the ; c<eesres for preparat son, valisation anc routine -
witltrettoe of the inf ormation hencling system have been fcllo.ec.
- t. A ceeonstratioe that a;;*opriate cestga verification anc q.d?1ficatnor testtag 11 5;,e:1fIec f or ; eu; w etoent of saf ety.
re i e tte c cr;osent s. The spe: 1fications snell.incluot quali. fica! ton testieg f or empe;ted Saf ety se's ice C oncit 10es anc pro tcc 5 ; port f or t r.e liter. sees
- re;e
- pt of testif'i, Occumen-tat son to sup;, ort' the limits of- 31 f e re; o.eence ' by the.
Sw;pl l' t.
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*!s'ety.reieted strsctures, sy stems , and component s are tf 0se, that afe rei t ec -
w;;,*. to ret.a t n . f unc t iondi curing anc following design. basis events to-ensure ', i., the integrity of 'the reactor Coolant bouncary, '(2) Ihe capability to sh t. Os.t the rextor anc maintain it in a' safe shwt00 n con ('It ton, and (3) the I ca: ability to prevent or ntigate the consedea.ces cf acc10ents that' could' resdt in potential of f site caposures coraparacle to the guidelines of 10. CrR , H
~part'100.
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. 6. . ? c e%ees d'c a?;I ' ca
- t5 ne ed o311 to s M t 've , y f r e v i ew j 16e eg.'rert classi'itatica prog *a- f: .sa'ei..rd ated- l
. c o* pone nt s. ' Al t houg * ' not . requir e Llo be ssDw'!tec fcr . l st a
- f review, your - eqst ynent cl a s s : f ic at 'on program shoul df j also include the. broader class of struc t s es.asystems,' r and components important to safety required by GDC-1;(defined '
d in 10 CFA Part 50 . Appendia A.' *Geaeral Design Criteria,
' Introduction").
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- 2. For veneer interface licensees'and:applicasts shall. establish.
implepeat and natatatn a centinsirig program to ensure that veneer .' information for- safety-related components. it complete, corrent; . y f j and controlled throughout the life.ef their. plants. and appropriately , 4 H
> referenced or incorporated in plant instructions: end precedures.s t .;: , 2 Ve3ters of = safety.related equipment'should tie contacted 'and an' interf ace esta 11shes. Where vendors cannot be ident fied. Anave : pone ~ out of -
bustyss, or wil1 not supply information, t*e licensee or appitcant ' shall assare that sufftctent attention.15-p:16 to equipment malatenance. replace =4et , and re;a t r.. to compensate 'for the' laca.-of vendor. beckup, to .i assse e rei s attitty canaeasurate with .its sa'ety function (GDC-1). The l
; rag e sui t be cl osely ' couplee with 'act tee 2.2.1 "above (equi pment . s qua'
- f scat t om). T*e progr see 'sha!1 inc1wce pe -1 odic comunu m i.catioW wIth were:rs to assare' that all applicable infor ation has been received. The .
preges= she,,lJ use 's system of positive feectack with venoors for mattings contatning te:hnical information. This' cou! $,be accomp111hed by. Itcensee ackno.1e:gment' for receipt of technical mait ings.. -It shal); al so define the inte-f ace sec divi st or. of ' responsibility es ano69 the licensee aad the % clear sad r0*wclear divisions of thetc.veneces that provide service or. sdfety related equipment to assu r e that requisite control, of amo at:11ca:1e' instructions f or ' maintenance =crt oe ' s a f ets . rel a ted : equipment a-e provided. Ap:' t c anil i ty TM s action a;si t es to all I t censees a63 OL - app 1 t cants. _Ty e e ' p e, t e w Fo* I t ce 9s ee s , a pes t. ' *;' eme - t a t i oe. r ev i ew will t,e coafucted. 4RE =111~ Pe rf om t he ret t ee and is sse .a Sa f ety (,alwat t on. j For OL ap;1 t cant s , t he nR R r e. t ee wi l l be je a f or e n c oe s t s ter.t wi t h the licensing schedule. Doceeetation Reevired . 4 ,
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. Ltcersees and apellcants sho.I t Smit a re; ort that. describes 'the .
$ eq * ><*t cla ssi ficat t os. and vendor' inter f ace pr;grees outlinec the
' l PCs it ion st0ve.
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h: cndr yes to the Techrical 5;ecificatior.s' are react red. Eeferences Section 2.3.1 of NLkEG 1000. Section 2.3.2 of heEEG-1000. x . V 4-', e e ,
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J 3,1 :C$i.M :%'!%A%:( TEST:s~a ';[ 4 TOP * ; ^ ;' i ' $ ' E w U"; .NE C $) . cesittom The foilcaing actioas are ap::31 cat" e to post-ma19tenen*e testing:
- 1. Liceesees and applicants shall sube.?! the results of their review :j of. test ' and na16tenance prc,cedu+es 'amd Technical' Speci fication.. to j assu*e that post. maintenance operability testing of safety-relatedL .j components in the reactor trip systee.is -requireo to be conhted j and ttat the testing 6 demonstrates that the equipne.1t'i' capable ofj '
perfomir; its safety fasectionslbefore being returne1
- service.
2. Ticensees and .apst icants'shall:. subatt the JAN residk . . ~
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%.,, Nendor and engineering recommendations to entdeithat 'any appropriate- q ftest ' guidance is' includeejiri the' test and maintenance procedures.or.- 1
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- 3. Liceesees and applicants shall = sdentify, if applicable,May- post.
ma tateca9te test re:;uirements : tm existing - Tecn,v c. .
!pedifications-which can be de tonstrated.to degrade' rather t; an to..ance safety. -
A:;ee:riate changes to these test re::wirements, with. supporting j ustificatior., 59411 De s bmitte::!fc" staf f ap;* oval. -(hote.that .) action 4.5 ciscusses on-Inne system fe.ctienal: testing.) Acp11cability This action applies to all licensees and OL applicants. .; Type of Review For licensees, a p~ost-impleme-tatioe review will be coeducted for actions 3.1.1 and 3.1.2 above. The Regions will perform these licensing reviews an:- issue Safety Evaluations. Preposed Technical Spect fication ' changes resultirt - l f rom action -3.1.3 above will receive's. pre-implementation' review by hRR. ; I For OL applicarts, the review will be perfomed consistent with the licens1ng schedule. l l Occ ument a t i on' Requir ed .j Licensees and applicants shou' d sube tt 4' state-ent confirming that" actions 3.1.1 a nd 3.1.2 of t he a bov e po s i t i ce ha v e ' be en ' i m p ' eme nt ed . Technical 5perificat ton Chaege1 Requ red - l' I Changes to Technical Speci f ica t ions .. as a result' of ' act ion. 3.1.3. ' are to 1 be detemi eed by the lit.ensee or app licant and 'submi tted, for - staf f approval ~. as.necessary. Re f er eng Sect ton 2. 3.4 of. irJREG-1000. 1 _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ '1
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y . .' *:5'.6:siENAN:t TEST!h' ; A;L Oir4E; ' SA5 E l v fiE L A
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;ce following actions are applicabie to _ post. maintenance testing:
- 2. Licensees and applicants shall submit a report ' documenting' thel
~
extencing of test and maintenance procedures and Technical ^ .. Specifications review to assure that post-maintenance operability testing.of all safety-related equipment is required to be conducted ' and that the testing demonstrates that' the equipment is capable of performing its safety functions before being returned to. Service. d? ; .- .. . .. 4> i
. 2.c l Licensees and applicantsishall: submit the results of' their check; ih4of vendor end engineering recommendations' to ensure that'any &.3;. jf
~
gg J L -[M Harpropriate test guidance is included in the test and asiatenance31 g 4 Q, 1 l%gkh*pgpproced. w ires or the TechnicaliSpecifications;where: p'.@ f. # ~'T required 44A
-i.iuresees and applicants shall identify, if applicable, any postMjF :
maintenance test- requirement 5 im' esisting Technica1J Spectficatiens; ' ' which are perceived to degrade rather .than enhance safety.1 Appropriate H changes to these test requirerests, with supporting justification, shall be submitte1 for. staf f aporoval. Applicability This action applies to all' licensees and OL applicants.- Type of Review For licensees, a post-implementattom review will . be conducted forLactions J 3.2.1 and 3.2.2 above. The Regions will perform these licensing' reviews; and issue Safete Evaluations. Proposed Technical . 5 specification changes resulting f rom action '3.2.3 above .will receive a pre.implementatten rev.iew . by NRR. . For OL applicants, the review will be performed consistent with the i licensing schedule. ] Doc eentation Required Licensees and applicants should submit a statement confirr.ing that actions 3.2.1 and '3.2.2 of the above position have been implemented. Technical Spect fication Changes Reqsired Changes to Technical Specifications. as a result of action 3.2.3..are to be determined by. the. licensee or applicant for 51af f ' approval . as neces sa ry. i heference Sect ion 2.3.4 of Ni$EG-1000. a 1
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- E . !" : I *. ' ( i E A D:E: E A t? N0;r:D !chs)
Fesitdon All vendor-reccmnded . reactor trip. breaker modifications shall: be reviewed to vert fy that either: (1) each modification _has, in f act, been implemented; or (2) a ~ written evaluation of the technical reasons. for not implementing a modification exists, i For example, the modific'ations _ recommended by Westinghouse'in' uCD-Elec-t4L
. for the D6-50 breakers 'and a March -31,1983.. letter for. the D5-416 breakers ,
shall .he, implemented or a justification for not' implementing sanall he mede J _.svailable. Modifications not previously made shall be incorporated 'or aj.;
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g, #W.','., written; evaluation shall. De provided. 3, m ;. . .
-l N;'_} Applicability;, pp{ , T,? e f j [ D " 7 [ Y M. , ~. ;ij[ M !
t y ,a ,, >i..; , .. .y . ~ . . ;. y,, This acti6n applies to'all PWR. licensees and OL applicants. Tyoe of Review < For licensees, a post.iglementation review will be conducted, The Regions will perform these licensing reviews L and-issue. Safety Evaluations. ' j For OL applicants, the NRR, review will' be performed consistent with the licensing schedule.. Documentation Required Licensees and an;;11 cants should subeit 's statement confirming' that this action has been implemented. .q Technical Specifications Required . No chsnges to Technical Specifications are required. Reference Section 3 of WUREG-1000. l 1
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e i 4,; s[ A;;0a *R;D $'5 TEM ret I AE!61TY -(: Et NTAfivE MA!h*iNah;E ! A%3 SAVE :a Ah;E PROGR AM F OR RE A;'M 91r BRE AKERS) j i Fosition
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Licensees and applicant $ shall descr:e their preve'.tative maintenan'ce a9e surveillance program to ensure retable reactor trip breaker _ operation.. Tne program shall include the fo11owts:
- 1. A planned prog' ram of periodic imettenance, including lubrication.-
housekeeping, and other itens renamended by. the equipment: supplier. o ,. .h .. , ,. ; _ 7. Tr'ending of parameters'affecting meration.and seas'ered during - 3.t "i, ; o i esting to forecast degradattom e operability. . 4;
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-y%d d.1 Life testing of the breakers (tec'ading the trip attachments)..en! '.C 14 vf .i;g4 %
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4 periodic typlacement. s' breakers .c components: consistent with. demonstrated Itfe cycles. Reliability ints action applies to. all NR Itcenses and 01. applicants.- ) bpe of keview Act tons 4.2.1 and 4.2.2 will receive + post. implementation review by
&% A pre-implementation review will m performed by-NRR for actions-j 4.2.J and 4.2.4 (the circuit breaker "Ife testing program and the Corn-weet testing / replacement requirement based vpon the It fe. testing- ')
res.,1t s). . A safety Evaluation will' tw issued. . F:e OL applicants, NRR will perform tw reviews for actions 4.2.1 and 4.?.? on a schedule consistent 'with tw licensing schedule.: NRR will' De*f or's a pre. implementation review . fr actions 4.2.3 and 4.2.4 (the-circuit breaker It fe testing 'progems ac the. component testing / replace'- emt re:;ut rements based upon the li'e 'esting results). Safety
~.aiwatiors will be issued.
i
~.<ccentat ion Required Mcensees and applicants s Id s. butt descriptions of their 'prograres .to :
e s' u re compliance with this actioe.
/ _: e Technical specification Ch'anges Requtg 1
% changes to Technical pecif scatto,ns are required.
~
.Ee'ereng, I 4
Section 3 of Nutts.1090..
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4.3 RE ACTOR TRIP SYSTEw RE.11B:. T' > AU* ? ~!C ACTU ATION Or L 5Nht - TR!P ATTAC*Elii FOR 6CS*14-NSC. AC E &W PL A ATS)' . Position Westinghouse :and B&W reactors shall be modified by providing automatic ' reactor trip systes actuatica of the breaker, shunt trip attachments.- The shunt. trip attachment shall he considered safety related (Class !E). Applicability < j"
. ;- :. i .
E ..;;x j?..((This 7 4 applicants. actic'n applies to all tiestinghouseTand - '. 8&W licensees .,apd kaOL.3'.y;@m % m
%a.;es.Se;9W; % .* D
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Ma Q(? Type of Review
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L. , g< p @7 $ j .: . , 4 . gp - w For licensees, s' pre-implementatten review shall be performed iW
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for'tti
. design modifications by NRR. ' A Safety Evaluation will be. issued.
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^
For OL applicants, the NRR review will' be' performed consistent with~ the licensing schedule. ., Technical Specification changes, if required, will be' reviewed prior to implementation. Documentation Required ) i Licensees and applicants should sabeit a report describing' the -j modifica tions.
~
Technical Specification Charges Required. Licensees are to subett any needed Technical . Specification ' change requests prior to declaring the modified system operable. Reference Section 3 of WUREG-1000. i I f
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- : : :: C: i& P 1'!*!" FE;1 h h :T* i w m .Ewthit Ih M*:U t haf E a: TEST'F: : Er.:E5 FOR E&f PL uis) 1 c e sition-Licensees and applicants with' B&W reactors shall apply: safety-related e.aintenance and test procedures to the diverse reactor trip feature provided by interrupting power to. control rods *hrough the. silicon controlled rectifiers.
This action shall not be (hterpreted to require hardware Changes er accitional erwirofunental Dr seismic qualification of. these components.:
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Applicability , 'ff " " M ' .. ;.[s ' 35
. oc & . .- .w h:;w g.@ wp D This action applies' to'lani licensees and OL[ applicants only. . s. >:ri.
y g#.,.( : 4 ;.4 3 gf -
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n.i 4p[ .. eO. y Type of Revie7; @a @,jg O e,c,E,; MVh,,.
.,K! .sy%$,4.e f '- ,.,
n .e i. Ter licensees, a post-implewntation review /will'be' conducted. The-bgions will conduct the 11 censing' review and issue a Safety Evaluation. -q For OL spplica9ts, the review will be performed consistent with the licensing schedule. Occumentation Recuired - 1 Licensees and applicants should submit a statement confiming that-this action has been implemented. Technical Specification Changes Required Include the silicon controlled rectifers in the-' appropriate surveillance l and test sections of the Technical Specifications. l Eeference Section 3 of WUREG-1000.
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9 i l a.? C[A:*0R TR]P SY$i[M RE. N IL'if (SYSTEw FUN ~ TIONAL TEST N ) . Position- <f -]
.l Cn.1ine functions) testing. of the reactor. trip system, . including independent testing of the diverse. trip features, shall De performed on all plants. I
- 1. The diverse trip features .to' be tested include the breaker -
undervoltage and' shunt trip features on Westinghouse. 88W.(see3. Action 4.3 abora)..and CE plants; the circuitry used for power p) intermption with the st11 con controlled rectifiers 'on 84W plants , '7 (see Action 4.4 above); and the scram pilot valee and backup scram' . 1, " /A, T + , q. p qyf ivalves (including;a11' initiating 5:fgcuitry) on GE plants.Wy
- n. n ..J 3 r .y . .. .- .
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& S% . Plants hall justifynot carrently not making modifications designed to'permitJsuch to persittesting. periodic:J
- 4. on-Alternatives to on-line testing proposed by 15censees will be' '-:1 !'
X? 'q considered where special circumstances entst sad wherie the objective of high reliability can be met in another way.. . a 1 :u
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- 3. Existing intervals ' for. on-line. functional testing reovired by.
Technical Specifications shall be reviewed to determi- at ce 'o the intervals are consistent with achieving high reach trip, 6 system availability when accountic; for considerations'such ; as: f" l
- 1. uncertainties in camponent. failure rates .e ~
- 2. uncertainty in common mode feilure rates- A'
- 3. reduced reiundancy during testing- ; j 4 ' ope-ator errors during testing , ,f
- 5. component
- wear-est* caused by the. testing g:
Licensees currently not performing periodic on-line testing shall- I determine appropriate test intervals as ' described above. ' Changes to- -] esisting required intervals for on-line testing as well as the- > interva s to be determined by licensees currently not performing.. M on-line testing shall be justified by information on' the: sensitivity . 4 3M/ of reactor trip system availability to parameters such as the test O intervals, component failure rates, and common mode' failure rates. y,j i Applicability .. j
- 1 This action applies to all licensees and OL applicants. p 'Lj g
Type of Review
,J For licensees, a post-implementation review 'will be conducted for action 4.5.1. The Regions will perform'these 1.icensing reviews and issue! )4 Safety Evaluations. . Actions 4.5.2 and 4.5.3 wil1~ require a pre implemen a - l tation- review by WRR. Results will be issued in a Safety Evaluation.- 1 l
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