ML20151T429
| ML20151T429 | |
| Person / Time | |
|---|---|
| Issue date: | 07/31/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| References | |
| NUDOCS 9809090412 | |
| Download: ML20151T429 (73) | |
Text
_ _. _ _. _. _
DISTRIBUTION for NRR Director's Quartarly Status Report
' Central File
.'JWRoe, NRR DBMatthews, NRR THEssig, NRR FMAkstulewicz, NRR EMMcKenna, NRR PCWen, NRR EYWang, NRR BJSweeney, NRR
. GMHolahan, NRR FPGillespie, NRR WDTravers, NRR RLSpessard, NRR GClainas,NRR RCEmrit, RES Regional Administrators Mr. Ralph Beedle, Senior Vice President Nancy G. Chapman, SERCH Manager
& Chief Nuclear Officer Bechtel Power Corporation Nuclear Energy Institute 9801 Washingtonian Blvd.
1776 l Street NW _ Gaithersburg, Maryland 20878-5356 Suite 400
- Washington, D.C. 20006-3708 Mr. R. P. LaRhette Institute of Nuclear Power Operations 700 Galleria Parkway Atlanta, Georgia 30339-5979 Mr. Lee Watkins Assistant Manager For High Level Waste U.S. DOE P.O. Box A Aiken, South Carolina 29892 Mr. R. W. Barber Safety and Quality Assurance, DOE 270 Corporate Center (E-853) 20300 Century Blvd.
Germantown, MD 20874 Mr. S. Scott Office of Nuclear Safety, DOE h ;!
Century 21 Building (E-H72) 19901 Germantown Road
. Germantown, MD 20874-1290:
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Mr. Bob Borsum 1700 Rockville Pike, Suite 525 gLS{J{'ij b ig-,.m
,f i o~?;pH Rockville, MD 20852 ty Ms. Norena G. Robinson, Licensing Technician N Nebraska Public Power District Cooper Nuclear Station 0 kk b
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9909090412 990731 i " "' Y#
i DIRECTOR'S STATUS REPORT on GENERIC ACTIVITIES Action Plans Generic Communication and Compliance Activities JULY 1998 Office of Nuclear Reactor Regulation
INTRODUCTION The purpose of this report is to provide informaton about generic activibes, including generic communications, under the cognizance of the Office of Nuclear Reactor Regulation. This report, which focuses on compliance activities, complements NUREG-Og33, A Prioritization of Generic Safety lesues."
This report includes two attachments 1) acton plans and 2) generic communications under development and other generic compliance activibes Generic communicatons and compiiance actMbes (GCCAs) are potential generic issues that are safety significant, require technical resolution, and possibly require generic communication or action.
Attachment 1 "NRR Action Plans," includes generic or poten6 ally generic issues of sufficient complexity or scope that require substantial NRC staff resources. The lesues covered by schon plans include concems identlSed through review of operaung experience (e.g., Boihng Water Reactor internals Cracking and Wolf Creek Draindown overd), and issues related to regulatory flexibility and improvements (e.g., New Source Term and Probabihetic Risk Assessment (PRA) Implementabon Plan). For each action plan, the report includes a -t 42-:-7 of the issue, key milestones, discussion of its regulatory significance, current status, and names of cognizant staff.
Attachment 2," Generic Communications and Compliance Achvibes," consists of three status reports.
- 1) Open GCCAs, Q GCCAs added since the previous report, and 3) GCCAs closed since the previous report. The genonc communicatons listed in the attachment includes bulletins, generic letters, and information notices Compliance activibes listed in the attachment do not rise to the level of complexity that require an action plan, and a generic communication is not curren#y scheduled. For each GCCA, there is a short descripton of the issue, scheduled complebon date, and name of cognizant staff.
ATTACHMENT 1 NRR ACTION PLANS
TABLE OF CONTENTS BOILING WATER REACTOR INTERNALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 GRID RELIABILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 PRA IMPLEMENTATION ACTION PLAN 1.2 (c) Inservice inspection Action Pian..............................................................8 NEW SOURCE TERM FOR OPERATING REACTORS . . . . . . . . . . . . . . . . . . . . . 15 ENVIRONMENTAL SRP REVISION ACTION PLAN . . . . . . . . . . . . . . . . . . . . . . . . 19 l STEAM G EN ERATORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 PRA IMPLEMENTATION PLAN 1.2(d) Graded Quality Assurance Action Plan . 23 INDUSTRY DEREGULATION AND UTILITY RESTRUCTURING ACTION PLAN Final U pdate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 EXTENDED POWER UPRATE ACTION PLAN . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 DRY CASK STORAGE ACTION PLAN Final Update . . . . . . . . . . . . . . . . . . . . . . . 35 ACCIDENT MANAGEMENT IMPLEMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . 39 CORE PERFORMANCE ACTION PLAN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 FIRE PROTECTION TASK ACTION PLAN Final Update . . . . . . . . . . . . . . . . . . . 45 HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN . . . 47 HIGH B U RN U P FU EL ACTION PLAN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 WOLF CREEK DRAINDOWN EVENT: ACTION PLAN . . . . . . . . . . . . . . . . . . . . . 52
BOILING WATER REACTOR INTERNALS TAC Nos. M01898, M93925, M93926, M94959, M94975, M95369, Last Update: 07/07/98 M96219, M96539, M97373, M97802, M97803, M97815, M98206, Lead NRR DMeion: DE M98708, M98880, M99638, M99870, M99894, M9G897, M99898, Supportmg Division. DSSA M99895, M99897, MA1102, MA1104, MA1138, MA1226, MA1926, GSI: Not Available MA1927 MILESTONES DATE (T/C)
. PART1: REVIEW OF GENERIC INSPECTION AND EVALUATION CRITERIA
- 1. lesue summary NUREG-1544 03/96 C o - Update NUREG-1544 3Q/FY99 T
- 2. Review BWRVIP Re-inspection and Evaluation Criteria o Reactor Pressure Vessel and intomais Examination Guedelmes (BWRVIP- 06/08/98 C 03)........................................................... 06/08/98 C o BWRVIP-03, Section 6A, Standards for Visual inspection of Core Spray 09/98 T Piping, Spergers, and Associoned Components . . . . . . . . . . . . . . . . . . . . . . . 04/27/98 C o BWR Vessel Shell Wold inspection Recommendabons (BWRVIP-05)N . . . 1 o Guidelines for Rainspection of BWR Core Shrouds (BWRVIP-07) . . . . . . . .
- 3. Review of generic repair technology, critorie and guidance TBD
- 4. Review generic mitigallon guidelmes and criteria TBD l
- 5. Review of generic NDE technologies developed for examinations of BWR TBD intomal components and attachmer,ts i
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NThe Commacon. in SRM M970512B dated May 30,1997, requested that the staffs SER (a) should address the BWRVIP proposal to examme 100 percent of the axial welds which would include examine 6ons M some circumferential weld lengths near the intersectons of the weld types to determine if this proposal could provide an approprints level of sampling of the circumferental welds. (b) should provide a comprehenelve evaluation of the probabilistic analysis ,
contained in the BWRVIP proposed ellematve in determming the acceptability of a proposed technical apwnshve and/or j in pursuing changes to the rule, (c) should consider a tiered approach in gathering additional baseline iryocnabon and/or i implemening the rule end. (d) should receive appropriate review. including review by ACRS. The indue!ry was not timely
'n completing its RAl responses, and this has resulted in extending the previously established schedule. The NRC staff has completed les safety evalushon of the BWRVIP-05 report and the completed SE wil be provided to CRGR ond ACRS for their review and concurrence 1
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- 8. Other intomals reviews (safety assessments, evaluations, miliga6on measures,
!- inspections and repairs) o Safety Aeoessment of BWR Reactor intemals (BWRVIP-08) . . . . . . . . . . . . 7/30/98 T o Bounding Assessmerd of BWR/2-6 Reactor Pressure Vessel Integrity lesues (BWRVlP-08 & BWRVIP-48) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 03/27/98 CA o Evalusuon of Crack Growth in BWR Stainless Steel RPV Intemals l (BWRVIP-14) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 08/08/98 C l o intomal Core Spray Piping and Sparger Replacement Design Criteria j (BWRVIP-18) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8/30/98 T o Roll / Expansion of Control Rod Drive and in-Core Instrument Penetrations in BWR Vessels (BWRVIP-17) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 03/13/98 CD
- o BWR Core Spray infomais inspec6on and Flaw Evaluation Guidelines l' (BWRVIP-18) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . 08/08/98 C !
j o BWRVlP-18, Appendix C, BWR Core Spray intemals Demonstration of l Compliarme With Technical Information Requirements of Ucense Renewal j Rule (1 C W 54.21 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12/98 T
- x, l 8/30/98 T l o imemal Core Spray Piping and Sparger Repair Design Criteria 8/30/98 T (BWRVIP-1 9) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8/30/98 T l
o Core Plate inspecton and Flaw Evaluation Guideline (BWRVIP-25) . . . . . .
! o Top Guide inspection and Flow Evaluation Guideline (BWRVIP-28) . . . . . . 9/30/98 T i o Standby I.iquid Control System / Core Plate AP inspechon and Flaw I. Evaluation Guidelines (BWRVIP-27) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10/30/98 T o Assessmerd of BWR Jet Pump Riser Elbow to Thermal Sleeve Weld Crackin0 (BWRVIP-28) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . TBD o Technical Basis for Part Circumferential Weld Overlay Repair of Vessel intamal Core Spray Piping (BWRVIP-34) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 03/30/99 T o Shroud Support inspechon and Flaw Evalua6on Guidelines (BWRVIP-38) .
o BWR Jet Pump Assembly inspechon and Flaw Evaluation Guidelines 09/98 T (BWRVIP-41 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
( o BWR LPCI Coupling inspechon and Flaw Evaluation Guidelines (BWRVIP- 12/98 T l 42)...........................................................
i o Update of Bounding Assessment of BWR/2-8 Reactor Pressure Vessel 03/27/98 CA Integrity lesues (BWRVIP-48) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
o BWR Lower Plenum inspec6on and Flaw Evaluation Guidelines (BWRVIP- 12/98 T 47)........................................................... ;
o VesselID Attachment Wold inspechon and Flaw Evaluation Guidelines 03/30/99 T !
(BWRVIP-48) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
o instrument Penetration inspection and Flaw Evaluation Guidelines 03/30/99 T j (BWRVIP-49) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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Description:
Many components incide boiling water reactor (BWR) vessels (i.e., intemals) are made of materials such as stainless steel and various alloys that are susceptible to corrosion and cracking. This degradabon can be accelerated by stresses from temperature and pressure changes, chemical interac6ons,irradia5on, and other corrosive environments. This action plan is intended to encompass the evalua6on and resolution of issues associated with intergranular stress corrosion cracking (IGSCC) in BWR intomals. This includes plant specific reviews and the assessment of the generic criteria that have been proposed by the BWR Owners Gram and the BWRVIP technical subcommittees to address IGSCC in core ;
shrouds and other BWR internals. !
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i Historical Backaround SigniAcant cracking of the core shroud was Srst observed at Brunswick, Unit i nuclear power plant in September 1993. The NRC no66ed hcensees of Brunswick's decovery of significant circumferential weeks of the core shroud welds in 19N, core shroud cracking continued to be the most signlAcant of reported intemals cracking in July 19M, the NRC issued Generic Letter 94-03 which requires licensees to inspect their shrouds and provide en analysis jusefying congnued operation until inspections ,
can be w.,
A special industry review group (BoiNng Water Reactor Vessels and Intemals Project - BWRVIP) was ;
formed to focus on resolution of reactor vessel and intemals degradation. This group was instrumental in fedlitating licensee responses to NRC's Generic Letter. The NRC evaluated the review group's reports, i
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submined in 19M and early 1995, and all plant responses.
All of the plants evaluated have been able to demonstrate con 6nued safe operation untilinspechon or repair on the beels of: 1) no 360' through-wall cracking observed to date,2) low frequency of pipe breaks, ,
and 3) short period of operation (2-6 months) before all of the highly susceptible plants complete repairs of ;
orinspectons to their core shrouds.
In late 19M, extensive asakg was decovered in the top guide and core plate rings of a foreign reactor, t The design is similar to General Electric (GE) reactors in the U.S., however, there have been no ;
observatons of such creciong in U.S. plants. GE concluded that it was reasonable to expect that the ring i cracking could occur in GE BWRs with operating time greater than 13 years. In the special industry review ,
group's report, that was issued in January 1995, ring cracking was evaluated. The NRC concluded that the BWRVIP's messesment was acceptable and that top guide ring and core plate ring cracking is not a short term safetyissue. )
Proposed Actions The staff wHl contnue to assess the scopes that have yet to be submitted by licensees concoming inspections or re-inspechons of their core shrouds. The staff will also continue to assess core shroud reinspection results and any appropriate core shroud repair designs on a case-by-case basis. The staff will losue separate safety evaluations regarding the acceptability of core shroud reinspection results and core shroud repair designs. The staff has been interacting with the BWRVIP and indudual licensees.
In an effort to lower the number of industry and staff resources that will be needed in the future, it is important for the staff to continue interactng with the industry on a generic basis in order to encourage them to continue their proactive efforts to resolve IGSCC of BWR internals. The BWRVIP has submitted 29 generic documents, suppor9ng plant-spece6c submittels, for staff review. The staff is ensuring that the generic reviews are incorporating recent operating experience on all BWR internals.
Orininanna Document Generic Letter 94-03, issued July 25,1994, which requested BWR licensees to inspect their core shrouds by the next outage and to justify continued safe operation until inspections can be completed Regulatory Assessment in July 1994, the NRC issued Generic Letter 94-03 which required licensees to inspect their shrouds and provide an analysis justifying continued operation until inspections could be performed. The staff has concluded in all cases that licensees have provided sufficient evidence to support contnued operation of their BWR units to the refueling outages in which shroud inspections or repairs have been scheduled. In addition, in October 1995, industry's special review group submitted a safety assessment of postulated crocking in all BWR reactor internals and attachments to assure continuing safe operation.
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Current Status
- Almost all BWRs completed inspec6ons or repairs of core shrouds during refueling outages in the fall of 1995. Various repair methods have been used to provide altemate load carrying catsahany, including preemptive repairs, installation of a series of clamps and use of a series of tie-rod assemblies. The NRC has reviewed and approved all shroud modification proposals that have been submitted by BWR liconeses Review by NRC con 6nues on individual plant reinspecton results and plant-specific assessments in October 1995, industry's special review group issued a report (BWRVIP-06) which the NRC staffs preliminary review indicates was not comprehensive. The NRC staff requested additional information whicn the BWRVIP provided in letters dated December 20,1996, and June 16,1997. The staff has completed its review of this submlBal, and the SE is presendy in final concurrence. The industry group submitted a report on reinspection of repaired and non-repaired core shrouds (BWRVIP-07) in February 1996. The staff wi-E':2 j les review and lesued an SER with several open items. The staff met with the industry to resolve these open items, and wT-g't j les final SER. The NRC is also reviewing information submitted by GE on the safety significance of and recommended inspec6orn for top guide and core plate ring crecidng.
Technical review of the " Reactor Pressure Vessel and Intemals Examination Guidelines (BWRVIP-03)" is complete and the staffs SE has been issued.
By letter dated September 20,1996, the BWRVIP informed the staff of its intention to Petition for Rulemaking to change the augmented inspechon requirements contained in 10 CFR 50.55a(g)(6)(ii)(A), in accordance with the recommendations of BWRVIP-05, which would change the inspection requirements from " Essentially 100%" of all RPV shell welds to 100% of circumferential welds and 0% of longitudinal wolds, information Notco (lN) 9743," Status of NRC Staffs Review of BWRVIP-05," was issued August 7, 1997, to inform the industry of both the status of the staffs review and that the staff would consider twh'-- ", justified scheduler relief requests from performing augmented inspections of the RPV shell circumferential welds for 40-months or two operating cycles, whichever was longer. The staffs independent assosoment of the BWRVIP-05 repost was transmitted by letter dated August 14,1997, to the BWRVIP, along with a request for arishnalinformation and information that needed to be addressed for licensees requesung scheduler relief. The staff has granted such relief requests. The staff briefed the ACRS subcommittee on August 26,1997, and briefed the full committee on September.4,1997. The NRC staff has completed its evaluaton of the BWRVIP-05 report and the completed SE will be provided to CRGR and ACRS for their review and concurrence. IN 97-63, Supplement 1, was issued May 7,1998, to inform the industry that the staff would continue to consider technically-jushfied scheduler relief requests from performing augmented inspechons of the RPV shell circumferential welds for 40-months or two operating cycles, whichever was longer.
The staffs review of BWRVIP-14 and -18 is complete and the staffs SEs have been issued. The staffs review of BWRVIP-19 on internal core spray piping repair design criteria is continuing.
By letter dated December 20,1996, the BWRVIP submitted, " Appendix C to BWRVIP-18. This appendix addresses the use of BWRVIP generic intemal core spray inspection guidelines for compliance with requirements of the license renewal rule (10 CFR Part 54). The staff is reviewing this appendix in conjunction with its review of BWRVIP-18 guidelines.
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I The BWRVIP submitted BWRVIP-28 to address the safety implications of recent cracking found in BWR Jet pump riser elbows. The staff is reviewing the BWRVIP-28 report. The staff issued NRC Information Report i IN 97-02," Cracks Found in Jet Pump Riser Assembly Elbows at Boiling Water Reactors," on February 6, 1997.
Information Notice 97-17," Cracking of Vertical Wolds in the Core Shroud and Degraded Repair," was issued April 4,1997, to inform the industry of vertical weld cracks and a degraded core shroud repairs found at Nine Mile Point, Unit 1. The BWRVIP has informed the staf' that it plans to revise BWRVIP-07 to ensure that the vertical core shroud welds, and the core shroud repair, is adequately inspected.
By letters dated April 25 and May 30,1997, the BWRVIP provided a reaffirmation of the BWR member licensees to the BWRVIP, and committed, on behalf of your member licensees, to several actions, including implementing the BWRVIP topical reports at each BWR as appropriate considering individual plant schedules, configurations and needs, and providing timely notification to the NRC staff if a plant does not ,
implement the applMable BWRVIP products. The staffis requesting that the BWRVIP have each BWR 1 licensee confirm to the NRC staff that each BWR licensee is aware of the NRC staffs understanding of these commitments, and recognizes their individual commitment to the BWRVIP program and reports.
NRR Technical Contacts: Keith Wichman, EMCB,415-2757 Merriiee Banic, EMCB,415-2771 Kerri Kavanagh, SRXB,415-3743 Frank Grube!ich, EMEB,415-2784 NRR Lead PM C. E. Carpenter, EMCB,415-2169 R6derences:
Generic Letter 94-03,"Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors,"
July 25,1994.
Action Plan dated April 1995.
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GRID RELIABILITY TAC No. M98444 Last Update: 6/30/98 Lead NRR Dmson: DE Supporting DMsson: DSSA MILESTONES DATE (T/C)
- 1. Assess and evaluate the risk significance of potential grid instability due to deregulation.
- a. Survey past and expected electric grid performance (EELB) 10/98 (T)
- b. Assess projected risk from grid-centered loss of offsite power 1/99(T) events (SPSB) ,
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- c. Inform Commission (SPSB / EELB) 3/99(T)
- 2. Monitor industry deregulation and implement mechanisms to institutionalize on9oing staff level contacts. (EELB)
- b. Develop reliability assessment tool 7/98 (T)
- c. Inform Commission (EELB) 3/99(T)
- 3. Issue generic communication (EELB)
- a. Draft Generic Communication 11/97 (C)
- b. Office Concurrences (if necessary) 2/98 (C)
- c. ACRS Review (if necessary) NA
- d. CRGR Review (if necessary) NA
- e. EDO Concurrence Of necessary) NA
- f. Comrnession Approval (if necessary) NA
- g. Issue Generic Communication 2/98 (C)
- 4. Evaluate based on Task i the need for regulatory actions. Evaluate method (s) to identify grid-centered event precursors. Evaluate the impact ,
of deregulation on SBO risk reduction goals. j
- a. Review AEOD study for implications regarding grid-center events 2/98 (C) j (EELB) !
- b. Complete feasibility study on methods to identify grid-centered events N/A (EELB)
- c. Assess the implications of grid-centereci events to SBO risk reduction 1/99 (T) goals (SPSB)
- d. Determine what additional regula' ay actions are necessary (NRR) 2/99 (T)
- e. Inform Commission (SPSB / EELB) 3/99 (T)
Descriobon The action plan is intended to address the Commission's concerns regarding the impact of utility deregulation on the reliability of the electric grid to supply offsite power to nuclear power plants for safe operation.
Historical Backarour d: In recent years, two relatively new factors are emerging: non-utility generation and deregulation. It is anticipated that, in the not too distant future, power suppliers, whether utilities or independent power producers, will actively compete for sales to customers who may be located anywhere on the power grid. Regional grid control would be the responsibility of centralized independent System 6
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Operators (ISOs). The responsitulibes and authority of an ISO have yet to be defined, but it is expected that they will be charged with maintaining grid reliabikty to fackste the marketing of power. It is also t uncertain how, or even whether, the current method of maintaining reliability through voluntary compliance with guidelines established by consensus associations will transition to the new utility -
structure. These uncerm raise ques 6ons with respect to the continued supply of reliable offsite power to nuclear power plants. j Procomed Achons Specille schons included in the action plan are: (1) issuing generic communications to -
reemphasize the need for hcensees to maintain their design basis with respect to the stability and reliability of offsite power and to mentam a process for ensuring that they continue to meet their design basis for ,
the remainder of their license; (2) monitoring industry dersgulation developments and its impact on the !
reliability of offsite power to nuclear power plants; (3) assessing and evaluating the risk significance of i potential grid instability due to deregulation; and (4) re======g the risks and effechveness of SBO issue j resolution efforts due to grid-centered loss of offsite power event initiators. '!
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Orininatma Document AEOD C97-01, Grid Performance Factors, dated March 20,1997; Staff Requirements Memorandum dated May 27,1997.
Reaulatory Assessment Based on the rapid changes in utility deregulation, the Commission requned j that the staff give greater urgency to ensuring that reisted health and safety issues within NRC's jurisdicbon are addressed particularly in reviewing the terms of the licensing basis and validating grid reliability assumphons Given that there is no evidence at this time that the reliability of the grid is degraded, continued operation is justified Current Staf2: Contract has been authorized for Oak Ridge National Laboratories (ORNL) to begin work and data collecton trips with staff are in progress.12 out of 12 visits completed as of 5/29/98. ORNL is working on the final report to support Milestone 1.a.
NRR Technical Contact Ronaldo Jenkins,415-2985 NRR Lead PM Chester Posiusny,415-1402 References (1) SECY-97-246, *information on Staff Actions to Address Electric Grid Reliability issues," October 23, 1997.
(2) Memorandum from S. J. Collins to H. J. Thompson, "Rebaselining Chairman Tracking List,"
March 11,1998.
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' PRA IMPLEMENTATION ACTION PLAN 1.2 (c) inservice inspection Action Plan ;
l TAC Nos. M95125, M97153, M99389 Last Update: July 29,1998 Complete Revision l M99756, MA0125, Lead NRR DMeion* DE l MA0067, MA0868 Support DMelon DSSA, EMCB i RG/SRP MILESTONES DATE (T/C)
- 1. Draft for Rl-ISl team review / comments 04/05/96 C
- 2. First draft for Branch CNefs rewow/ comments 08/14/96 C
- 3. Revised draR for Branch Chiefs rewow/ comments 01/24/97 C
- 4. Revised draR for Branch Chiefs review / comments 04/08/97 C i
- 5. Draft for DMeion Director review / comments 04/29/97 C
- 6. Draft for Office Director /OGC rewow/ comments 05/16/97 C :
- 7. Office Director /OGC concurrence 07/08/97 C )
- 8. Draft for CRGR review / comments 07/08/97 C
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- 9. ' Draft for ACRS review / comments 06/03/97 C i
- 10. Inllial presentation to ACRS full Committee 06/11/97 C
- 11. Initial presentation to CRGR 06/11/97 C i
- 12. Meeting with ACRS Subcommittee 07/08/97 C
- 13. Meeting with ACRS full Committee 07/09/97 C
, 14. Meeting with CRGR 07/17/97 C j
- 15. SECY from EDO to Commissioners (SECY 97-190) 08/20/97 C l
- 16. Publish draft for public comments 10/1547 C
- 17. Public comment period for draft RG/SRP ands 01/13/98 C
- 18. Public Workshop 11/20/97 C
- 19. Complete draft for ACRS/CRGR review / comments 04/98 C
- 20. Complete draft for inter-Office concurror.ce 05/98 C !
- 21. Issue RG/SRP for trial use 08/98 T
- 21. Issue final RG/SRP 06/99 T WOG TOPICAL REPORT MILESTONES
- DATE (T/C)
- 1. Issue Second RAI(if necessary) 7/31/98 T
- 2. Receive RAI Responses 9/30/98 T
- DATE (T/C)
- 1. Receive Revised Report 7/06/98 T
- 2. Issue Second RAI(if necessary) 9/30/98 T
- 3. Receive RAI Responses 1/31/99 T
- 4. Issue FSER 3/31/99 T 8 l l
a I
I PILOT PLANT REVIEW MILESTONE 0
- DATE (T/C)
- 2. leaue FSER Surry 12/31/98 T
- 3. lesuo FSER ANO-2 12/31/98 T .
- i Subject to change based on licensees' actual submittel dates and responses to Staff RAls.
l INSPECTION PROCEDURES MILESTONES DATE (T/C)
- 1. lesue Draft inspection Procedure Number 73753 6/98 C ,
- 2. lesuo Finalinspechon Procedure Number 73753 TBD Description
- Develop risk informed inservice inspechon (RI-ISI) application-specific Regulatory Guide (RG), correspondmg Sta idard Review Plan (SRP) sections and related inspec6on procedures; determine acceptability of industry's proposed risk-informed methodologies for inservice inspection (ISI) application and related American Society of Mechanical Engineers (ASME) Code Cases; review acceptability of the pilot programs with respect to their RI-ISI applications and prepare plant specific safety evaluation reports (SER). This achon plan ~will be monitored up to and including the completion of RI-ISI RG and SRP, pilot '
plant reviews, topical report reviews, and development ofinspechon procedures. The action plan has been completely revised to clarify the process for the review of RI-ISI submittals subsequent to the approwel of the pilots by referencing the topical repotts, the addition of a descripton for the future reviews and approvals of the ASME Code Cases,i.e., the long-range plan, and miscellaneous editorial changes.
Historical Backaround On August 16,1995, the U.S. Nuclear Regulatory Commission (NRC) published a policy statement (60FR42622) on the use of probabilistic risk assessment (PRA) methods in nuclear regulatory actMties. In the statement, the Commission stated its belief that the use of PRA technology in i
NRC regulatory actMeios should be increased to the extent supported by the state-of-the-art in PRA methods and data sad in a manner that complements the NRC's deterministic approach. In a
- November 30,1995, memorandum to J. M. Taylor, the NRC Executive Director for Operations (EDO),
Chairman Jackson directed that the staff prepare an action plan, together with a timetable for developing i RGs and SRPs awsociated with the use of PRA in specific applications. A Nuclear Reactor Regulaton/ Nuclear Regulatory Research (NRR/RES) joint task group has been established to accomplish the above delineated specific tasks in the RI-ISI area as directed by Chairman Jackson.
The nuclear industry has submitted two methodologies for implementing RI-ISI. One methodology has been jointly developed by ASME Research and Westinghouse Owners Group (WOG) (Reference 4,6) and the other methodology is being sponsored by Electric Power Research Institute (EPRI) (Reference 5).
ASME is working on three Code Cases for alternate examination requirements to ASME Section XI, Division 1 for piping welds. Code Case N-577 is based on the WOG methodology and Code Cases N-578 is based on the EPRI methodology. Code Case N-560 is based on the EPRI methodology but is
- being revleed to ancompass both methodologies Proposed Achons The NRC has encouraged licensees to submit pilot plant applications organized under one ambrella sponsonne organization, e.g., Nuclear Energy institute (NEI), for demonstrating risk-informed methodologies to be used for piping segment and piping structural element selection in systems scheduled for ISI. The NRC is reviewmg the industry submittals with focus on the licensees characterizing the proposed change including the identification of the particular piping systems and weldc that are affected by the change, engineering evaluations performed, PRA analysis to provide risk insight to support the selection of welds to inspect, traditional engineering analysis to verify that the proposed changes do 9
m __ ____ -
i not compromise the existing regulations and the licensing basis of the plant, development of implementation and monitoring programs to assure that the reliability of piping can be maintained; and 4
documentation of the analyses and the request for NRC review and approval. Additionally, using the l results from the review of the above-mentioned pilot plant applications, from the PRA insights obtained j from the risk-ranking of piping elements, and in cooperabon with the RES staff, a parallel effort is being
, carried out to develop (a) en RI-ISI apphcation-specific RG and (b) the corresponding SRP chapters and j ==~4= led inspection procedure documents.
The nuclear industry, under the umbrella of NEl, has submitted two methodologies for the implementation ,
of the RI-ISI. One methodology has been jointly developed by ASME Research and WOG (Reference 4, !
- 6) and the other methodology is being sponsored by EPRI (Reference 5). The pilot plant for the WOG !
methodology is Surry 1 and pilot plants for the EPRI methodology are Vermont Yankee and ANO-2.
{
The acceptability of the RI-ISI pilot plant programs will be documented in SERs for each of the pilot plant licensees and forwarded to the Commission Upon Commesson approval, the staff will issue SERs l authorizing allemative inspection strategies for the pilot plant licensees to allow use of the Rl-ISI methodology.
i ASME is working on three Code Cases for alternate examination requirements to ASME Section XI, Dueion 1 for piping welds. Code Case N-560, for the altemate examination requirements for Class 1, '
Category B-J piping welds,is based on the EPRI methodology. This Code Case is being revised to encompass both WOG and EPRI methodologies. Code Case N-577, for the alternate examination requirements for Risk-Based Selecton Rules for Class 1,2, and 3 piping welds,is based on the WOG j methodology. Code Case N-578, for the alternate examination requirements for Risk-Based Selection Rules for Class 1,2, and 3 piping welds, is based on the EPRI methodology. The NRC staff intends to work with the industry and ASME to ensure integration and consistency of the various approaches being ,
proposed.
- The major difference between Code Case N-577 and the WOG methodology submitted to the staff ,
(Reference 4,6) is that the scope of the Code Case is limited to ASME Class 1,2, and 3 systems while ;
the WOG methodology may encompass all the safety significant systems in the plant. In addebon, the i Code Case is an abbreviated version and does not have all the details presented in the WOG topical report (Reference 4,6). The staff intends to review the WOG methodology as well as the Code Case N-577 and the consistency of the Surry 1 pilot program for RI-ISI to both of these. The Code Case N-577 l will be reviewed and,if found acceptable, will be endorsed by RG 1.147 with any necessary addshons or :
deletions The pilot plant RI-ISI program review will be documented in the staff SER. l The major difference between Code Case N-578 and the EPRI methodology is that the scope of the l Code Case is hmited to ASME Class 1,2, and 3 systems while the EPRI methodology may encompass all i safety significant systems in the plant. Also, the Code Case is an abbreviated version and does not have all the details presented in the EPRI topical report (Reference 5). The staff will review the EPRI methodology as well as Code Case N-578 and the consistency of the ANO-2 RI-ISI pilot program to both of these. Code Case N-578 will be reviewed and,if found acceptable, will be endorsed by RG 1.147 with any necessary additions or deletions. The pilot plant RI-ISI program review will be documented in the i staff SER.
Code Case N-560 for the attemate examination requirements for Class 1, Category B-J piping welds is l being revised to encompass both WOG and EPRI methodologies. This Code Case has limited i apphcability in that it is apphcable only to ASME Class 1 piping systems. The staff will review the EPRI l methodology as well as Code Case N-560 and the consistency of the Vermont Yankee RI-ISI pilot program to both of these. ;
1 i
10 l I
)
The staff will use the acceptable altemative provimon of 10 CFR 50.55e (a)(3)(i) to approve the pilot plants' applications The staff will work closely with ASME to expedite changes involving ISI.
Lono-Ranae Plari This acuon plan will be monitored up to and including the completion of Rl-ISI RG and SRP, pBot plant reviews, topical report reviews, and development of inspechon procedures. The staff plans to perform in-depth reviews of pilot plant submittals and WOG and EPRI Topical Reports during the FYs 1998 and 1999 in order to ensure that the RI-lSi programs are consistent with the staff RG and SRP. This process may entail revisions in industry documents as well as in the staff RG anr1 SRP.
For the Rl-ISI programs submittals subsequent to the approval of the pilot plant programs and topical reports, but prior to the endorsement of ASME Code Cases, it is expected that the licensees will utilize the approved WOG or EPRI Topical Report as guidance for developing RI-iSI programs but will need to seek relief from NRC to the current 50.55a requirements. A six month review cycle is expected for the approval of RI-ISI submittels during this time frame.
It is anticipated that subsequent to the issuance of safety evaluation reports (SER) for the pilot plants and the topical reports, the industry will revise the ASME Code Cases tc incorporate lessons learned from pilot plants and topical report reviews The ASME Code Cases will 64 endorsed by RG 1.147 with exceptions and/or additions,if necessary, consstent with past practice. Subsequently, the Code Cases are expected to be incorporated into the ASME Code. In the long term, the staff will proceed with rulemaking to approve the ASME Code with caveats, if necessary, so that other licensees can voluntarily adopt risk-informed ISI programs without the need for specific NRC review and approval. For the RI-ISI programs
- developed after the RI-ISI methodology has been endorsed in RG 1.147 but prior to approval via endorsement of ASME Code in 50.55a, the staff will approve the RI-ISI programs as relief requests. ,
Subsequent to endorsement in 10 CFR 50.55a, the staff anticipates that the licensee will develop an l RI-ISI program using the approved ASME Code method. No NRC approval will be required, and the staff will oversee the acceptable implementation as part of the normal ISI inspecbon program.
For the non-pilot plant hcensees that intend to implement RI-ISI starting with their next ten year interval, the staff is considering granting a relief from the current determirustic requirements of ISI of piping, of up to two years. These hcensee would then be able to develop and obtain approval for their RI-ISI program at the next available opportunity using the staff approved topical reports on WOG or EPRI methodology.
During the two-year extension period, the licensees would continue to implement their current ISI program. In order to dieseminate the information to the licensees, the staff is considering issuing an information notice.
Orimnatina Documents
- In a November 30,1995, memorandum to J. M. Taylor, the NRC EDO, Chairman Jackson directed that the staff prepare an action plan together with a timetable for developing RGs and SRPs applicable to use of PRAs to be completed in two years. In his response of January 3, 1996, the EDO presented a plan that established milestones for the development of regulatory guidance documents for utikring PRA in reactor-related actmbes including ISI. This action plan is in conformance with the agency-wide implementabon plan for PRA and any future changes will be consistent with the .
overall plan.
Reaulatory Assessment The operational readiness and functionalintegrity of certain safety-related piping and associated structural elements (e.g., pressure retaining welds) are vital to the safe operation of nuclear power plants. ISIis one of the mechanisms used by the licensees to ensure piping integrity. The type and frequency of ISI are based on past experience and collective best judgment of the NRC and industry in a consensus Code endors+1 through the rulemaking process. The current ASME Code ISI requirements and pracbees have only an implicit consideration of risk-informed information, such as failure probability and consequence of failure.
11
Liconeses are currently interested in optimizing inspechon by applying resources in more safety-significant areas. They are also interested in maintaining system availability and reducing overall maintenance costs in ways that do not have an adverse effect on safety.
On a parallel path, ASME is Cz':-$g Code Cases for altemate examination requirements to the current ASME Section XI selecton and inspec6on requirements. These Code Cases utilize procedures that are based on the relaHvo risk significance of piping locations withm indudual systems.
The NRC is using probabilielic methods, as an adjunct to determiniebe, techniques to heip define the i scope, type, and frequency of ISI. The development of RI-ISI programs has the potential to opbmize the !
use of NRC and industry resources and conhnue to assure adequate protection of public health and
. safety.
Acceptability of the Rl-ISI pilot programs is expected to be documented in safety evaluations. The staff recommendation whether to authorize an altama#ve inspechon program pursuant to 10 CFR 50.55e (a)(3)(I) will be presented to the Commission prior to its implementation. To provide the permanent approach to RMSI, the staff intends to utilize the experience gained through the pilot apphcations in the proposed rulemaking process te modify 10 CFR 50.55e to explicitly endorse RMSI methodology.
Current Status' Since the formation of the RI-ISI team, several meetings have been held with NEl and industry / utility representatives. In these meetings, the NRC staff and industry have discussed their respeceve plans for the RI-ISI programs. NEl has submstted WOG technical report (Reference 4), and EPRI technical report (Reference 5). The staff has also been actively participating in ASME Code schwilies related to RMSI. NEl has submitted a revised WOG technical report (Reference 6) that addresses the staffs commons and requests for addibonal information (RA!). The staff has also submitted its comments and RAls on the EPRI report to the industry.
The staff cosiC:I drafts of RMSI RG and SRP which were submitted to tne Commissioners (SECY-97-190) to request approval to publish the RG and SRP for public comments. The staff and the industry made detailed presentations to the ACRS on July 8,1997, during its 443rd meebng. The ACRS recommended (Letter dated July 14,1997) that these documents be issued for public comments subject to incorporation of changes in response to ACRS comments. The staff also met with the Committee to Review Generic Requirements (CRGR) on July 17,1997, during its Meeting No. 308. CRGR issued its letter dated August 14,1997,in which it stated that CRGR has no objection to tim proposed RG and SRP being issued for public comments, subject to the incorporation of its comments and recommendations.
On August 20,1997, the staff forwarded the draft guidance documents to the Commission and requested its approval for issuing the documents for comment by the public. Commasion approval was received in an October 1,1997, SRM. The RG and SRP were issued for public comments by a Federal Reg'e ter Notice (FRN) on October 15,1997, which stated that public comment period will expire on January 13, j 1998. i To facilitate solicitation of public comments, the staff held a workshop on November 20 and 21,1997, to explain the draft documents and answer questions. The workshop was well attended by industry representabwes. The public meeting and the written comments offered a number of constructive comments, some criticisms, and some suggestions for changing the guidance. Overall, the comments indicated strong support for pursuing risk-informad inservice inspection (RMSI) but in a manner which would neceseltate some modifications to the draft guidance. The more significant comments regarding the RI-ISI regulatory guide and standard review plan included:
. the flexibiNty to pursue partial plant or system level applicabon should be supported;
. the role of qualitative risk impact assessment is not adequately discussed in the guide and supporting i appendices are extremely detailed and provide prescriptive methodolog'es instead of a framework;
. the EPRI approach should be incorporated as a way to classify safety significant pipe segments; 12
. there should be a strong recommenda6on to validate pipe failure potential by benchmarking against service exportance;
- ASME Code Cases N-560, N-577, and N-578 should be endorsed as acceptable alternatives per 10CFR50.55a(s)(3)(I).
In response to public comments and discussions with ACRS, the following changes to the RI-ISI guide and SRP have been made.
. The staff modi 8ed DG-1063 and its accompanying SRP and plans to issue these documents for trial -
use (as RG-1.178 and SRP Section 3.9.8). The staffis proposing to issue the documents for trial use for the following reasons:
- a. Cr:':-;-r,u.t of the documents did not have the benefit ofinsights gained from pilot plant applications. Schedules for the receipt of licensee submittals have slipped such that these reviews are now expected to be completed by the end of December 1998.
- b. The industry submitted topical reports that apply two different methods for incorporating risk insights into their RI-ISI programs. The staff issued requests for additional information (RAis) and is reviewing the reports in parallel with the pilot submittals and thus are also expected to be completed by the end of December 1998.
- c. The ASME passed three code cases (N-560, N-577, and N-578) that implement RI-ISI programs.
The topical reports identi8ed in (b.) above provide the technical details required to implement these code cases. The staff is reviewing these code cases in parallel with the pilot plant submittals.
. The RG and SRP have been revised to indicates that both partial and full scope applications of RI-ISI are acceptable.
. In modifying the RG, the staff has concluded that the appendices should be removed from the RG.
~ Regulatory guidance will be provided to the industry through the staffs safety evaluations of the topical reports and pilot plant sutettals. Upon completion of the topical report and pilot reviews, a decision will be made on the need for and content of any append 6ces for the final RG. In the interim, the staff will document the technical content of the appendices, including elements of a qualitative approach, in a draft NUREG report (s) and make it available for information.
. Guulance regarding the scope, level of detail, and quality of the PRA needed for RI-ISI programs was modilled to provide more flexibility in meeting the guidance.
. Guidance regarding the performance monitoring program for high safety significant piping (HSS) and low safety significant piping (LSS) remained unchanged, except for the second barrier to fission product release (e.g., primary system piping). The RG was modified to state that if the categorization of the primary system (Class-1 piping) identified the reactor coolant system or a subset, such as the cold and hot leg piping, to be LSS, then some level of inspection will continue to be required for defensehdepth considerations.
- Documentation requested for an ISI program submittal was reduced.
The staff has received pilot plant submittals from Vermont Yankee, Surry, and ANO-2. The staff review for these pilot applications is currently under progress.
13
- - - - - .- r _ _
i l
NRR Contact S. Ali(415-2776), S. Densmore (415-8482)
RES Contact. J. Guttmann (415-7732) 1 References-
- 1. Federal Register, Vol. 60, No.158, "Use of Probatulistic Risk Assessment Methods in Nuclear [
Regulatory AclM6es; Final Policy Statement," August 16,1995.
- 2. Memorandum from Shirley Ann Jackson, Chairman, NRC to James M. Taylor, Executive Director for Operatons," Follow-up Requests in Probabilmtic Risk Assessment and Digital lastrumentation and Control," November 30,1995.
- 3. Memorandum from James M. Taylor, Executive Director for Operations, NRC to Shirley Ann Jackson, .
Chairman," improvements Associated with Managing the U611zation of Probabilistic Risk Assessment and Digital Instrumentaton and Control Technology," January 3,1996.
I
- 4. WCAP-14572,"Wesenghouse Owners Group Application of Risk-Based Methods to Piping inservice J inspec6on Topical Report," March 1996. l l
- 5. EPRI TR-106706," Risk-informed inservice inspection Evaluation Procedure," June 1996.
- 6. WCAP-14572, Revision 1,"Wesbnghouse Owners Group Applicatiori of Risk-Informed Methods to Piping inservice inspecbon Topical Report," October 1997.
l4
I NEW SOURCE TERM FOR OPERATING REACTORS TAC No. M89586 Last Update: 07/10/98 GSI No.155.1 Lead NRR Division: DRPM CTL 1.1 Supporting Division: DSSA & DE MILESTONES DATE (T/C)
- 1. NElletter 07/94C !
- 2. Commission Memo 09/94C
- 3. NEl Response 09/94C
- 4. NEl/NRC Meeting 10/94C
- 5. Publication of NUREG-1465 02/950
- 6. NEl/NRC Meetings 10/94C,06/95C,10/95C, 01/96C,02/96C,05/96C, 08/96C,10/96C,04/97C
- 7. Submittal of Generic Framework Document (from NEI) 11/95C
- 8. First Pilot Plant Submittal 12/95C
- 9. Issue Memo to Commission, Updating Status 08/96C
- 10. Present Commission Paper in E-Team Briefing 09/96C I
- 11. Brief CRGR on Commission Paper 10/960 j 12. Send Commission Paper to EDO/ Commission 11/960 l
- 13. Brief ACRS on Commission Paper 11/96C
- 14. Response to NEl Framework Document 02/97C
- 15. Begin Pilot Plant Reviews 02/97C
- 16. Begin Rebaselining 02/97C
- 17. Brief E-Team on Status of Rebaselining 07/97C
- 18. Issue User Need for Rulemaking 08/97C
- 19. Finish Rebaselining 06/98C
- 20. Finish Rulemaking Plan 06/980
- 21. Finish Pilot Plant Reviews 01/99T Descriotion: More than a decade of research has led to an enhanced understanding of the timing, magnitude and chemical form of fission product releases following nuclear accidents. The results of this work has been summarized in NUREG-1465 and in a number of related research reports. Application of this new knowledge to operating reactors couki result in cost savings without sacrificing real safety margin. In addition, safety enhancements may also be achieved.
15
t i
Historical Backaround In 1962, the U.S. Atomic Energy Commiselon published TID-14844,"Calculadon i
~ of Distance Factors for Power and Test Reactors " Since then licensees and the NRC have used the l accident source term presented in TID-14844 in the evalustion of the dose consequences of design beeis accidents (DBA).
AAer examining years of addulonel research and operating reactor experience, NRC published NUREG-1465, " Accident Source Terms for Light-Water Nuclear Power Plants," in February 1995. The
~
. NUREG describes the accident source term as a series of five release phases. The first three phases !
(coolant, gap, and early in-vessel) are applicable to DBA evalua8ons, and all five phases are applicable to !
severe accident evaluations The DBA source term from the NUREG is comparable to the TID source term; however,it includes a more realistic cx@ i of release tming and compositon. Since the NUREG source term results in lower calculated DBA does consequences, NRC decided not to require current plants to revise their DBA analyses using the new source term. However, many licensees want to use the new source term to perform DBA does evaluations in support of plant, technical specificahon, and procedure nwese mans, NRC and NEl met several times to discuss the industry's plans to use the new source term. To make ofilcient use of NRC's review resources, NRC encoura0ed the industry to approach the lesue on a generic beels. The Nuclear Energy inattute (NEI) unveiled its plans for the use of the new source term at
, operating plants at the Reguistory information Conference in May 1995. NEl, Polestar (EPRrs consultant), and pilot plant (Grand Gulf, Beaver Valley, Browns Ferry, Perry, and Indian Point)
. representallves met with NRC staff in June and October 1995 to discuss more detailed plans.
Proposed Achons' The staff has reviewed the framework document has prepared a Commiselon paper ,
and decision letter that describes a generic implementm6on approach. The staff presented the Commission paper and decision letter to the NRR Executive Team in September 1996, briefed CRGR in October 1996, and briefed the ACRS full committee in November 1996. The staff sent the Comrmonion paper and decision letter to the Commiselon in November 1996 (SECY-96-242). As described in the Commiselon paper, the current plan is to rebaseline two NUREG-1150 plants; one a PWR and one a BWR.~ The staffis reassessin0 the availability of key resources needed to complete rebaselining on an evt-mad schedule and issued a memorandum to the Comminaioners informing them of the status of the i project, briefed the Commiseloners' Technical Anmetants and accelerated the completion schedule with a shift in responsibilities between NRR and RES. The staff will also review each pilot plant apphcation and prepare an exempton package addreemng the use of each feature of the NUREG-1465 source term while pursuing rulemaidng.' The plan for issuing each remaining generic exemption is to brief the CRGR, issue for public comment, and then issue the exemption.
Onainatum Document: EPRI Technical Report TR-105909," Generic Framework Document for Appbcadon of Revised Accident Source Term to operating Plants," transmitted by letter dated November 15,1995. .
Reaulatory Assessment- There will be no mandatory back1It of the new source term for operating reactors The desi0n-basis accident analyses for current reactors based on the TID-14844 source term are still valid. Therefore, non-urgent regulatory action and continued facility operation are justified.
Current Status This issue is item 1.1 on the Chairman's Tracking List. NEl submitted its generic framework document in November 1995 for NRC review and approval. TVA submitted part ofits pilot plant apphcation for Browns Ferry in December 1995. The staff met with NEl on January 23,1996, to discues the generic framework document and separate meetings were held on February 7, May 30, and August 29,1996, to discuss the pilot plant submittals. The staff met again with NEl and the industry on October 2,1996, to discuss the staffs plan to issue exemptions while pursuing rulemaking, and on April 2, 1997, to provide a status report on the staffs actions regarding rebaselining and rulemaking subsequent to the Commiselon's SRM. The pilot plant applications for Browns Ferry, Perry, Indian Point, and 16
_ _ _ . _ _ . _ _ . _. _ _ _ _ _ .. _ ._ ~ _ . _ _ . . _ . . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _
Oyster Creek have been drculated to the task force members to help shape rebaselining. In June 1997, i RES circulated an early draft of the proposed RG that would consider updated source term insights (NUREG-1465) (the RG would be analogous to RGs 1.3 and 1.4 that use the TfD-14844 source term).
On August 1,1997, D:NRR issued a request to D:RES to initiate a rulemaking that would establish a regulatory framework for operating reactors to use the updated source term insights outlined in
- NUREG-1465; NRR believed that the rulemaking process can be inibated prior to the completion of re' x:"-As.
The staff briefed the NRR Executive Team on SECY-96-242 in September 1996, the CRGR in i October 1996, and its ACRS full committee in November 1996. A limited number of pilot plants !
sebmittels and exemptions are expected - four submsttals have been received so far (Browns Ferry, Perry, Oyster Creek, and indian Point-2). An application is also expected from Grand Gulf. In addition, the staff j and Virginia Power met on November 26,1996, March 25 and June 18,1997, to discuss the rebaselining of Surry; the staff and Entergy met on August 29,1996, and March 27,1997, to discuss the rebaselining of Grand Gulf. In a February 12,1997, SRM, the Comminaion approved the Option 2 approach of SECY-96-242 and a modlAcetion to the letter response to NEl. On February 26,1997, the EDO issued the letter response to NEl. The staff has initiated the rebaselining effort. The staff briefed the NRR Executive Team on the status of the project on July 8,1997, with an emphasis on the scope of actMbes involved vdth r4 M,;,g; as a consequence of that briefing, the user need memorandum regarding rulemaking was issued on August 1,1997, and the staff status report to the Commissioners was issued on September 9,1997, indicating that the completion of rebaselining will be deferred.
In response to Commession inquiries regarding the deferral of the completion of rebaselining until November 1998 NRR and RES discussions and shifts in lead technical responsetxlety resulted in an improvement in the schedule. At a Commessioners' Technical Assistants briefing on October 9,1997, the ,
Task Force Leader outimed a new schedule that would result in the completion of rebaselining and the I rulemaking plan in June 1998; this was accomplished by reversing the lead responsitzleties (RES is now
( the lead for rebaselining and NRR is now the lead for rulemaking and regulatory guidance). The schedule for the completion of the pilot plant reviews also improved by approximately 5 months as well.
1 NRR is working closely with RES to transfer technical insights gained on rebaselining. In addition, NRR I transferred its technical anoistance resources with SNL, ORNL, and PNNL that were designated for i rebaselining to RES. These changes will be reflected in the next revision to the NRR Operating Plan. On l November 13,1997, January 7,1998, February 24,1998, and March 30,1998, RES presented its four-phased plan and preliminary findmgs from Phase I, Phase ll, and the DBA portion (with the updated assumptions) of Phase lit, respectively, for the rebaselining effort. On April 1 and 2,1998, RES and NRR staff briefed the ACRS and DONRR, respectively, on the progress of the rebaselining effort , initial insights from the assessments completed, and the essential elements of the Rulemaking Plan. The results of the r# - "-#4 effort were reported in SECY-98-154 dated June 30,1998. The Rulemaking Plan was j i provided in SECY-98-158 dated June 30,1998.
NRR Technical Contacts- R. Emch, PERB,415-1068 S. LaVie, PERB,415-1081 NRR Lead PM B. Zalcman, PGEB,415-3467 References-NUREG-1465," Accident Source Term for Light Water Nuclear Power Plants," February 1995.
July 27,1994, letter to A. Marion, NEl, from D. Crutchfield, NRC," Application of New Source Term to Operating Reactors".
17 I
-, ,. - , y ., , , - . - - . - ,
September 6,1994, memorandum to the Commission from NRC staff,"Use of NUREG-1465 Source Term at Operating Reactors".
July 21,1995, memorandum to the Commiseen from NRC staff,"Use of NUREG-1465 Source Term at Operating Reactors".
December 22,1995, pilot plant submiltal, letter to Document Control Desk from Tennessee Valley Authority, " Brown's Ferry Nuclear Plant (BFN) - Units 1,2, and 3 - Technical Specificatons (TS) No. 356 and Cost Beneficial Licensing Action (CBLA) 08 - Increase in Allowable Main Steam isola 6on Valve (MSIV) Leaka0e Rate and Request for Exemp6on from 10 CFR 50, Appendix J... and 10 CFR 100, Appendix A...".
AuGuet 9,1996, memorandum to the Commiselon from NRC staff,"Use of NUREG-1465 Source Term at Operating Reactors".
November 25,1996, SECY-96-242,"Use of the NUREG-1465 Source Term at Operating Reactors."
February 12,1997, Staff Requirements Memorandum to SECY-96-242.
February 26,1997, letter to T. Tipton, NEl, from J. Callan, NRC, responding to the NEl Framework Document.
August 1,1997, memorandum from D:NRR to D:RES to request that rulemaking be initiated for operating reactor use of updated source term insights.
September 9,1997, memorandum to the Commission from NRC staff,"Use of NUREG-1465 Source Term at Operating Reactors."
June 30,1998, memorandum to the Commission from NRC staff, "Rulemaking Plan for implementation of Revised Source Term at Operating Reactors,." SECY-98-158.
June 30,1998, memorandum to the Commiseson from NRC staff,"Results of the Revised (NUREG-1465)
Source Term Rebaselining for Operating Reactors," SECY-98-154.
Summaries of public meetmgs e dated November 10,1994, for public meeting with NEl held on October 6,1994; e dated July 26,1995, for public mee6ng with NEl held on June 1,1995; e dated November 17,1995, for public meeting with NEl held on October 12,1995; e dated February 1,1996, for public meeting with NEl held on January 23,1996; 1 e dated February 27,1996, for public meeting with Browns Ferry held on February 7,1996; e dated September 27,1996, for public mee6ng with Grand Gulf held on August 29,1996; e dated October 11,1996, for public meetmg with NEl held on October 2,1996; e dated January 24,1997, for public meetng with Surry held on November 26,1996; ;
e dated April 24,1997, for public meelmg with PWR (Surry) held on March 25,1997; i l
e dated April 24,1997, for public meetmg with BWR (Grand Gulf) held on March 27,1997; e dated May 8,1997, for public meetmg with NEl held on April 2,1997; e dated July 28,1997, for public meetmg with PWR (Surry) held on June 18,1997.
I 18
ENVIRONMENTAL SRP REVISION ACTION PLAN TAC No.MA0837 Last Update: 07/24/98 GSI: Not Available Lead NRR Division: DRPM MILESTONES DATE (T/C)
- 1. Reflect Polential impacts and integrated impacts in Op6ons for Resolution
- a. Iden6ficebon of potendalimpacts 03/96C
- b. idenUlication of integrated impacts 06/96C
- c. Proposed options for resolution and develop initial draft 10/96C of revised ESRP
- d. Staff / contractor meebng to resolve format and content of revised ESRP 11/96C
- 2. Prepare Final Draft of ESRP Sechons for Public Comment
- a. Draft updated ESRP for staff review
- c. Publish (electronic) for public comment 06/97C 09/97C
- 3. W-::"=, Public Comments 02/980
- 4. Publish Final NUREG-1555 02/99T
- 5. Maintenance of program data Ongoing Brief Description The Environmental Standard Review Plan (ESRP) Revision Achon Plan deals with the revision to NUREG-0555 to reflect changes in the statutory and regulatory arena, to incorporate emerging environmental protection issues (e.g., SAMDA and envbonmental jus 6ce) since originally published in 1979, and to support the review of license renewal appbca6ons.
Renulatory Assessment: NRR has estabhshed the ESRP Update Program for use in the life cycle review of environmental protechon issues for nuclear power plants, especially bconse renewal applications, but also operating reactors, and future reactor site approval applicabons. The ESRP will reflect current NRC requirements and guidance, consider other statutory and regulatory requirements (e.g., the National Environmental Policy Act, Presidental Executive Orders), and incorporate the generic environmental impact work and plant-specific requirements developed during amending of Part 51 for license renewal reviews.
Current Status The PNNL/NRC staff workshop on the restructured and revised ESRP was held during November 13-14,1996. Now that the Part 51 rule for license renewal is final, particular emphasis is being placed on assuring that license renewal needs are being addressed in a schedule consistent with the RES regulatory guide and pilot plant appbce6on The results of the November workshop were provided by PNNL in January 1997; followup discussions were held with the contractor through August 1997. The June 1997 draft of the ESRP was forwarded to ACRS for its consideration. In light of the current ACRS schedule, ACRS staffindicated that the ACRS will have no objection to publishing the draft ESRP; the ACRS may request a briefing during the public comment period. The June draft was provided to CRGR for information; the CRGR declined to consider it. Technical editor, legal (OGC), and technical (lead technical branches) comments were received on the July draft in early August and were included in the final draft. The FR nobce of availability of Draft NUREG-1555 was published on October 3,1997; the electronic version (CD and diskette) is available in the PDR and will be made available to the public at no cost. Approximately 300 CDs and 500 hardcopies of the Draft NUREG were distributed for comment.
19
ACRS discussed the NUREG at ils May 1,1998, full committee meeting ; the Committee indicated that the staff shouki retum to further discuss the SAMOA guidance. During the week of February 9,1998, the staff developed the comment binning and depossbon plan; subsequently, a PNNUNRC staff workshop was held during February 24-25,1996,to depossbon technical comments and make decisions regarding the orgardza6onal structure of the ESRP A primary concem raised by the public was the consolidation of guidance for the technology area across disparate licensing frameworks (i.e., Parts 50, 52, and 54); the staff restructured the document to segregate guidance into a Part 50/52 ("greenfield"-type review) and that for Pari S4 (renewal of a license for an existing facility). This segregation took the form of a >
supplement f a the ESRP and was completed in draft form on July 3,1996.
NRR TechrkalCo@gt: l B. Zaleman, PGEB,415-3467 L
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-TAC Nos. 88885 and 99432 Last Update: 8/30/98 !
MILESTONE DATE (T/C)
- 1. - Comminaion/EDO Approval 02/94 (C)
- 2. Receive NEl Document 02/96 (C)
- 3. Review NEl Document Revisiors ' Continuous Process L
- 4. Regulatory Analysis 5/97 (C)
- 5. Proposed GL Pk0 10/97 (C)
- 6. ACRS Endorsement 9/97 (C)
- 7. CRGR Concurrence 8/98(T)
- 8. EDO 8/98 (T)
- 9. Publish Proposed GL 10/98(T)
Orig. Put2sh Proposed Rule 03/95 (C)
- 10. Public Comment (120 day comment period) 10/98 to 1/99(T)
- 11. Revise GL Pke 1st qtr 99(T)
- 12. ACRS Comments 2nd qtr 99(T)
- 13. CRGR Concurrence 3rd qtr 99(T) i
- 14. EDO Concurrence 3rd qtr 99(T) i
- 15. Commission Approval 3rd qtr 99(T)
- 16. Publish Final GL 3rd qtr 99(T)
Orig. Publish Final Rule 12/95 Brief Description
- The NRC orI0inally planned to develop a rule pertaining to steam generator tube integrity. The proposed rule was to implement a mo's flexible regulatory framework for steam generator surveillance and maintenance activities that allows a degradation specific management approach. The results of the regulatory analysis suggested that the more optimal regulatory approach was to utilize a generic letter. The NRC staff suggested, and the Commission subsequently approved, a revision to the regulatory approach to utilize a generic letter.
Regulatory Assessment The current regulatory framework provides reasonable assurance that operating PWRs are safe. However, the current regulatory framework has numerous shortcomings. To resolve these shortcomings the staffis reveng the regulatory framework to utilize a risk-informed and performance-based approach that will ensure compliance with current regulations (i.e., GDC, Appendix B, ASME code, Part 100).
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I Current Status-
- Briefed ACRS on ANPRM- August 1994 ;
- SG rule ANPRM- September 1994 l
- SECY-95-131 - May 1995 -Justifies continuation of rulemaking j
- Briefed Commission on SG rule -June 1995
- Briefed Commiselon on SG rule status - February 1996 i
- Memo to Commission re: revised schedule - May 1996
- Briefed Chairman on status - July 1996
- Informadon Brief for CP,GR - October 1996
- ACRS Brief on SG rule - November 5,6,1996 )
- Briefed Chairman on SG rule status - December 1996 l
- Briefed ACRS re risk 4nformed approach for SG rule - January 1997 l
- Briefed ACRS re. risk assessment and regulatory analysis results - March 4,5, and April 3,1997
- COMSECY-97-013 su90ests revismg approach to a GL - May 1997
- Briefed Commissioner Assistants re. revised approach - June 5,1997
- SRM of June 30,1997 agrees with revised regulatory approach
- Briefed ACRS re revised approach - June 12,1997
- Met with NEl/ industry senior mgmt re. GL status - July 22,1997
- Briefed ACRS re. GlJDG-1074/DPO issues - August 26,27, Septeriber 3,1997
- Information Brief for CRGR re. GL and backfit - September 9,1997 i
- Met w/NEl re. GL/DG-1074/TSs - September 11,1997 !
- ACRS endorsement to issue GL and DG-1074 for public comment - September 15,1997 l
- Briefed ACRS re. DPO issues - October 2,1997 1
- ACRS endorsement to issues DPO document for public comment - October 10,1997
- GL package into concurrence - October 21,1997
- NEl submits NEl 97-06 ' Steam Generator Program Guidelines"- December 16,1997
- CRGR package concurred on by NRR and sent to CRGR April 14,1998
- Met with CRGR on June 12,1998, for information briefing on package NRR Technical Contacts- Ted Sullivan, EMCB,415-3266 Tim Reed, EMCB,415-1462 RES
Contact:
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PRA IMPLEMENTATION PLAN 1.2(d)
Graded Quality Assurance Action Plan TAC Nos. M91429, M91431, Last Update: 7/9/98 M92447, M92448, M92449, Lead NRR Division: DRCH M88650, M01431, M01432, SupportDmsson: DSSA M91433, M91434, M91435, GSI: Not Available M91436, and M91437 MILESTONES DATE (T/C)
- 1. Issued SECY-95-059 03/950
- 2. Begin interactons with volunteer licensees 05/95C
- Palo Verde letter dated 4/6/95 l
- Grand Gulf meeting 5/4/95
- South Texas meetings on 4/19/95 and 5/8/95
- 3. NRC Steering Group meetings to guide working level staff activities As Needed
- Moetir;os on: 8/25/95,10/10/95,10/25/95
- 4. Staffinteracbons with Palo Verde Ongoing
- Site visit on 5/23/95 on ranking and QA controls through
- NRC letter dated 7/24/95 on proposed QA controls 3/980
- Site visit on 8/29-30/95 on risk ranking
- Site visit on 9/6-7/95 on procurement QA controls
- NRC letter conveying trip reports issued on 12/4/95
- Meeting on 4/11/96 to discuss the staff evaluation guide
- Letter from licensee on 4/24/96 providing comments on staff evaluation guidance
- Site visit on 6/5-6/96 to observe expert panel and review revised procurement QA controls, trip report sent to licensee on 8/6/96
- Letter from licensee on 9/12/96 transmitting responses to procurement issues raised in earlier staff trip reports
- Letter from licensee dated 11/13/96 responding to PRA issues raised in 12/4/95 trip report
- Overview of GQA initiative provided by PVNGS at 2/27/97 meeting with staff
- GQA closeout letter transmitted to licensee on 7/2/98
- 5. Staff interactions with South Texas Ongoing
- Meeting on 7/17/95 on project status through
- Site meeting on 10/3-4/95 on risk ranking and QA controls 3/98C
- Meeting on 12/7-8/95 to discuss risk ranking and QA controls
- South Texas Submittal of QA Plan for implementation of graded QA, dated 3/28/96 is currently under staff review
- Meetings on 4/11/96 and 4/25/96 to discuss the staff evaluation guide and future interaction milestones and schedules
- Letter from licensee on 4/17/96 providing comments on staff evaluation cuidance 23
. _ _ _ _ ._ _ . _ . . . _ .. ~__.__.- _ _ _ . _ . __._ . - . _ . _._ _. . _ . _ . _ _ .
- Mee6ng on 6/10/96 to discuss staff comments on the QA plan sutmttal for graded QA, review questons transmitted to STP on 8/16/96
- Site vielt on August 21-22 to observe worlung group and expert panel meebngs, and to discuss staff review items, trip report in preparation
- t=+;+T, sat mee6ng on 10/15/96 to discuss PRA initiatives and staff actMties
- Letter from licensee dated 10/30/96 responding to PRA ques 6ons
- Revised QA plan submitted on 1/21/97
- Overview of STP india #ve provided at 2/27/97 moeung with the staff
- Staff Request for Addibonal Information (RAI) issued on 4/14/97 for both PRA and QA controls
- Meeting on 4/21/97 to discuss STP responses to RAI
- Site visit on 5/5-8 to evaluate: PRA quality, graded QA controls, QA controls for the PRA, corrective action and performance monitoring feedback processes, audit scheduling, and responses to the RAI concerns.
Trip reportissued on 7/10/97.
- STP submittel on 5/8/97 fu orciirninary RAl response 3 - STP submittal of draft QA Plan on 5/21/97
- STP submittal of GQA related procedures, responses to RAI, and follow-on QA Plan on 5/22/97
- STP submittal of revised QA Plan on 6/10/97
- Staff RAIir, sued on 6/13/97
- STP submittal on June 26,1997, response to stafi RAI
- STP submittal of revised QA Plan on 7/16/97
- STP transmittal of addihonalinformation regarding GQA implementing procedures and associated change control on 7/31/97
- STP submittal on 8/4/97 responding to PRA RAI and provided procedures related to shutdown operations
- Negative consent SECY paper (97-229, dated October 6,1997) and Safety Evaluation has been issued that documents the staff's review of the QA program change.
- Commission did not object to issuance of STP SER as documented in 10/30/97 SRM
- Staff SER transmitted to licensee on 11/6/97
- STP comments and interpretations submitted on SER on 1/26/98
- Staff accepted STP interpretations of SER content on 2/19/98
- 6. Staffintersebons with Grand Gulf Ongoing
- Site meeting on 7/11-14/95 to observe expert panel through
- Meeting at hdqt. on 10/24/95 on QA controls 3/960
- Meeting at RIV on 11/16/95 on graded QA effort
- Site meeting on 11/17/95 to observe expert panel
- GGNS system and component ranking criteria under staff evaluation, the comments are scheduled to be provided to GGNS by the end of June
- Meeting on 4/11/96 to discuss the staff evaluation guide
- Letter to GGNS dated 5/29/96 regarding implementation of QAP commitments 24
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Staff review comments on GGNS safety significance determination process transmitted to licensee on July 15
- Meeting on August 27 to discuss staff comments on safety significance process and to discuss GGNS implementation of QAP commitments for low-safety significant items, meeting summary issued on 12/17/96
,~ - Site visit on 11/21/96 to review procurement activities, trip report was issued
' on 11/6/97.-
- GQA closoout letter transmitted to licensee on 1/7/98 4
- 7. Revision 3 of Draft Evaluation Guide for Volunteer Plants issued for staff 07/95C comment l
- 8. Revision 4 of Draft Evaluation Guide for Volunteer Plants issued for Steering 10/95C Group Review i
- 9. lssue letter to 3 volunteer plants outlining program objectives and review expectations. Distributed staff evaluation guide to licensees. 1/960
- 10. Evaluation Guide issued for use by staff in evaluating volunteer plants 1/96C ;
- Meeting held with volunteer plants to receive feedback on staff evaluation guide on 4/11/96. .
4/96C
- Industry comments on staff evaluation guide provided by letter dated 5/24/96
- The staff reviewed the industry comments with respect to the need to .
revise, and finalize, the evaluation guide.
- 11. Regulatory Guide development milestones per PRA Action Plan
- Draft RG for Branch / division review and comment 7/31/96C
- Draft RG for inter-office review and concurrence 8/1/960
- Draft RG for ACRS/CRGR review 11/22/96C
. Draft RG for public comment . 6/25/97C
- Draft RG public comment period ends 9/23/97 C
- Public workshop held on draft RG 8/12/97C
- Publish final RG in SECY-96-067 ' 4/2/98 C
- SRM conditionally approves issuance of GQA RG 6/29/980
- 12. ACRS Briefings .
- Expert Panel and deterministic considerations 2/27-28/96C l
- graded QA .. . .
. 4/11/96C
' -PRA implementation Plan and pilot projects 7/18/96C
- Risk Informed Pilots 8/7/96C - -l
- Graded OA Regulatory Guide 11/22/96C )
- Graded QA Regulatory Guide 2/21/97C
- ACRS Concems on GQA Regulatory Guide. . .
3/6/97C
- ACRS memo to Commission expressing concems with GOA approach 3/17/97C
- Public Comments on GQA Regulatory Guide 10/21/97C
' -Application RG/SRP discussions with Subcommittee 2/19/980
- Application RG/SRP discussions with Full Committee 3/3/98C
- 13. CRGR Briefings
- Graded QA Regulatory Guide 11/26/960
- Graded QA Regulatory Guide 3/11/97C
- 14. leeue draft 8taff Inspecton Guidance (Baseline + Reactive IP) for comment 9/98 T
-leeue Analinspecdon procedure 12/98 T
- 15. Conduct NRC Staff Training 2/99T Descriction: Prepare staff evalua6on guidance and regulatory guidance for industry implementation for the gradmg of quality assurance (QA) practices commensurate with the safety significance of the plant equipment The development of this guidance will be based on staff reviews of regulatory requirements, proposed changes to existing prachces, staff development of a draft regulatory guide with input from a natonal laboratory, and assessment of the actual programs developed by the three volunteer utilities implemonung graded quality amurance programs.
Historical Backaround The NRC's regulebons (10 CFR Part 50, Apperxhces A & B) require QA programs that are commensurate (or consistent) with the importance to safety of the functions to be performed.
However, the QA implementabon practices that have evolved have often not been graded. In the development of implementation guidance for the maintenance rule, a methodology to determine the risk significance of plant equipmerd was proposed by the industry (NUMARC 93-01). During a public meetmg on December 16,1993, the staff suggedsd met the industry could build on the experience gained from the maintenance rule to develop hpte A on methodologies for graded QA. The staff had numerous interactions with the Nuclear Ens iy in3Giutt (NEl) during calendar year 1994 as the graded QA concepts were discussed and the initialindew, falines were developed and commented on. In cady 1995, three licensees (Grand Gulf, South Texas, and Palo Verde) volunteered to work with the staff. The staff has reviewed the heensee developmental graded QA efforts.
Procomed Actions The goal of the schon plan is to utilize the lessons leamed from the 3 volunteer licensees to modify staffti-g+1 draft guidance to formulate regulatory guidance on acceptable methods for implementing graded QA. The staff will develop a regulatory guide based in part on input from Brookhaven National Laboratory, and will also prepare a beseline and reactive inspection procedure (IP) for graded QA. An inter-office team has been established to prepare the regulatory guidance documents and test their implementabon during the evaluation of volunteer plant activities Orimnalma Document: Letter from J. Sniezek, NRC to J. Colvin (NUMARC) dated January 6,1994, CNGg the establishment of NRC steering group for the graded QA initiative.
Reaulatory Assessment: Existmg regulations provide the necessary flexibility for the development and implomontston of graded qusHty assurance programs. The staff will issue a NUREG report regarding the lessons loamed from the volunteer plar;t implementations. Addstional regulatory guidance will be issued to either dieseminale staff guidance or endorse an industry approach. Planned guidance for the staff will involve an evaluation guide for applicebon to the volunteer plants, the lessons learned report, training seselons and public workshops, and inspechon guidance in the form of a baseline and a reactive IP. The staffis evaluating the appropriate mechanism for inspechons of the risk significance determination aspects of graded QA programs.
The safety beneAls to be gained from a graded QA program could be significant since both NRC reviews and inspections and the industry's quality controis resources would be focused on the more safety significant plant equipmerd and activities. Secondarily, cost savings to the industry could be realized by avoidmg the dilution of resources expended on less safety significant issues. The time frame to complete this action plan is direcoy related to the overall PRA implementation plan schedules.
26
Current Status A draft evaluation guide for NRC staff use has been prepared for applica6on to the volunteer plants implementing graded quality assurance programs. The staff will utilize the guide for the review of the volunteer plant graded QA programs. The guide and the staffs proposed interschon framework has been transmilled in a letter to the three volunteer licensees. The letter sou0ht hcensee comments DraR regulatory guides for both risk ranking and grading of QA controls have been prepared and circulated for review by both the ACRS and CRGR. SECY-97-077 (dated April 8,1967) transmitted the draft regulatory guides,inchading the GQA guide, to the Comminaion. Commission approval was obtained on June 5,1997, to issue the documents for a 90 day pubhc comment period. Senior mana0ement brioengs were provided to the Director, NRR (on April 22,1997) and to the Deputy, EDO (on April 24,1997). The pubhc comment period on the risk-informed guidance documents has expired. At this time,42 sets of comments have been received A decision has been made, and accepted by the Chairman, to focus staff efforts on revising the general regulatory guide and standard review plan first.
The proposal to sequentially complete the applicaton specific guidance documents, including GQA, was also accepted. SECY-97-229 forwarded the staffs evaluation of the STP GQA program with a recommendation that it be approved. The Comminaion did not object to the issuance of the SER. The staff presented the revised GQA RG to the ACRS (Subcommittee and Full Committee) and the CRGR, comments received durin0 those reviews were addressed as necessary. On April 2,1998, SECY-98-067 was issued which transmitted the GQA RG, along with the other application specific guidance documents, to the Commission By SRM dated June 29,1998, the Commiselon conditionally approved the issuance rd the GQA RG. Prior to issuance of the RG the staff will have to review, and revise acco,Gigly, the RG with respect to prior Commission guidance and direction contained in SRMs associated with the general risk-informed guidance and the policy issues associated with risk-informed regulation. Work has been initiated on dr;i4 5#a GQA inspection procedure.
A meetng was held with the three volunteer licensees on April 11,1996, to receive their feedback on the j staff dr;if+1 evaluation guide. The licensees expressed concerns about the level of detail contained in the guide, particularly that related to PRA and commercial grade item dedicabon. The licensees contend 1 that enlbng industry guidasece (PSA Apphcaton Guide and EPRI-5652) are sufficient for those topics. The staff received written comments from NEl on the evaluation guide by letter dated May 24,1996. The NEl letter questions the need for addibonal re0ulatory guidance for the graded QA apphcation. NEl contends that existing industry guidance is sufIlcient STP and PVNGS letters providing comments on the evaluation guide were dated April 17,1996, and April 24,1996, respec9vely. The staff considered suggested changes to the evaluation guide in response to the industry comments.
A presentation on graded QA was made to the full ACRS on April 11th. During the ACRS meeting some questions arose with respect to the staff expectations for the conduct of expert panel activities The ACRS was further briefed on the development of the GQA Regulatory Guide on November 22,1996, and February 21,1997, and March 6,1997. The ACRS issued a letter to the Chairman on March 17,1997, regarding their review of the risk informed guidance documents. The ACRS expressed some concerns with the staff focus on simply proposing to reduce quality controls for low safety significant items.
However, in recognition of industry interest in the guide, the ACRS recommended that it be issued for pubhc comment. On March 12,1998, the ACRS issued a letter to the Chairman which recommended that the GQA RG (RG 1.176) be issued for use. The ACRS expressed a concern that RG 1.176 does not take full advantage of PRA information However the ACRS scknowledged the inherent difficulty given the lack of a model to aseoas quantitatively the impact of modsfied QA controls upon the PRA model. The ACRS further recommended that RES consider a research project to assess the impact of QA controis on PRA parameters, and for the staff to prepare a plan for improvements to RG 1.176 after gaining experience with its application and to brief the committee within the next 2 years.
South Texas submitted their QA program revision for their graded QA effort on March 28,1996. The chan0e has been reviewed by the staff (HQMB, SPSB, RES, RIV, and NRC contractors). A meeting was held with STP on June 19 to discuss the staffs comments and concerns. STP indicated their willingness to re-examine the content of the QA plan with respect to the proposed QA controls for the low safety 27
__ _ _ _ _ _ _ _ _ _ _ __ _ . _ _ - _ _ _ _ _ . ~ . . _ _ _ _ _ _ _
significant items The staff visited the site on August 21-22 to receive information from STP in response to earlier staff questions about the STP approach towards determining safety significance categorization and adjustment of QA controls. The staff miso observed both a Working Group and Expert Panel meeting at
' which time licensee safety si 0ni6cance evaluations for 2 systems (Radiation Monitoring and Essential Service Water) were discussed. Stoff towow of the updated QA program submittal was completed and a second RAl was losued on April 14,1997, for both PRA and QA controls aspects. A meetmg was held on April 21,1997, during which the licensee provided some responses to the issues raised in the RAl. Staff (from both HQMB and SPSB) performed a site evaluation during the week of May 5-8 to review aspects had with: PRA quality, QA controls for the PRA, corrective schon and performance monitoring feedback processes, QA controls for low safety significant items, detailed information presented to address issues raised in the RAl, and the audit scheduling process. Further dialogue has occurred between the staff and STP durin0 the renew of the subsequent STP submittals and followmg lesoance of staff RAls. SECY paper 97-229 was losued on October 6,1997, to inform the Commission of the staffs review conclusions, and the recommendation to accept the STP program. The Commission did not object to the issuance of the SER as documented in their SRM of October 30,1997. On November 6, 1997, the staffs safety evaluation was transmitted to the licensee. The licensee provided their interpretation on 1/26/98 of selected aspects of the staffs SER. By letter dated February 19,1998, the staff agreed with the licensee's interpretations.
Also, NEl submitted 96-02,"Guidelme for implementing a Graded Approach to Quality" dated March 21, 1996. The staff has performed a cursory review of the document and concluded that it does not reflect the progrees and level of detail that has been achieved through the volunteer plant effort. The staff informed NEl by letter dated May 2,1996, that the guide is not adequate (as a stand alone document) to implement graded QA but that it will be considered as the staff develops the graded QA regulatory guide and standard review plan. By letter dated June 8, NEl indicated that their 96-02 guide will be revised.
Further NEl requested a meeting with the staff (in the August time frame) to discuss the changes and to discuss more objectwo means to assess the adequacy of QA program implementation. NEl has proposed that the amended 96-02 guidelines will be submitted to the staff for endorsement by a regulatory guide. A subsequent letter was received from NEl on July 16 that provided an updated version of NEl 96-02 based on comments they receeved from the volunteer plants and industry sources. The staff has reviewed the modified document. On October 10,1996, NEl submitted a letter expressing their d
concem with the graded QA iniliative. NEl stated their concerns regarded the questions raised by the staff in the area of QA controls for items determined to be low safety significant and in the area of safety significance determination. A meetmg with NEl and staff from the volunteer plants (STP and PVNGS) was held on February 27,1997. NEl stated that 50.54(a) needs to be revised to offer licensees greater flexibihty to manage their QA programs. The volunteer p%nt staff stated their firm desire to obtain copies of the draft GQA Regulatory Guide in a timely manner, foliosing Commission approval, these were released for comment on June 25,1997. NEl additionally outlined a conceptual approach to integrate a performance monitoring methodology into the GQA efforts NEtR Contacts S. Black,415-1017 R. Gramm,415-1010 I
RES Contact H. Woods,415-6622 References-
, 1) Letter from J. Sniezek (NRC) to J. Colvin (NEl) dated 12/94.
- 3) NUMARC 93-01," Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear i Power Plants."
l 4) SECY-95-059," Development of Graded Quality Assurance Methodology," dated 3/10/95.
- 5) - Letter from B. Holian (NRC) to W. Stewart (APSCo) dated 7/24/95.
i 6) Letter from C. Thomas (NRC) to W. Stewart (APSCo) dated 12/4/95.
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- -4 , . -. .__. . - , . 4 - ,, , , - , , ,-. ,,, , .. ,,. -
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- 7) Memorandum from S. Black to W. Beckner and W. Bateman dated 1/24/96, Draft Staff Evaluebon Guidance.
- 8) NEl 96-02," Guideline for Irnplemen#ng a Graded Approach to Quality."
- 9) Draft Regulatory Guide-1064,%1 Approach for Plant-Specific, Risk-informed Decision Making:
Graded Quality Assurance," dated March 24,1997. :
- 10) SECY-97-229, " Graded Quality Assurance /Probabiliste Risk Assessment implementation Plan !
for the South Texas Project Electric Generating Station," dated October 6,1997, and SRM dated
]
10/30/97.
- 11) Letter from T. Alexion to W. Cottle (HL&P) dated 11 Al/97.
- 12) Letter from J. Donohew to J. Ha9an (Entergy) dated 1/7/98.
- 13) SECY-98-067,
- Final Application-Specific Regulatory Guides and Standard Review Plans for Risk-informed Regulation of Power Plants," and SRM dated 6/29/98. ,
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INDUSTRY DEREGULATION AND UTILITY RESTRUCTURING ACTION PLAN Final Update l TAC Nos. M78003 Last Update: 6/30/98 )
GSI: Net Available Lead NRR Dmaion DRPM ,
MILESTONES ! DATE (T/P/C)
Task 1 - Develop NRC Policy Statement and SRP 12/97C Dran Policy Statement 05/96C Office Concurrences 06/96C EDO Concurrence 06/96C Commission Paper 07/96C Draft SRP 07/960 Publish Draft Policy Statement 09/96C Office Concurrences on SRP 09/96C EDO Concurrence on SRP 09/96C Commission Paper on SRP 09/960 Publish Draft SRP 1/97C Public Comment Policy Statement 2/97C Public Comment SRP 03/97C Final Policy Statement OS/97C Office Concurrences 06/97C ACRS 06/97C CRGR N/A EDO C:,acurrence 06/97C Commission Approval 07/97C Publish Final Poky Statement 08/97C Final SRP on Antitrust to Commission 10/97C Publish Final Antitrust SRP 12/97C Task 2 - Issue Administrative Letter to Licensees on Financial Reporting 06/96C Requirements Draft Administrative Letter 05/96C Office Concurrences 05/960 Commission Information Paper 06/96C issue Admin Lir to Licensees w/WTR Letter to CEOs 06/96C Task 3 - Develop Non-Rulemaking Option for Periodic Reporting 09/97C Requirements as Necessary Determine Necesssty for Action 09/96C Draft Option 01/97C Office Concurrence 01/97C EDO Concurrence 09/97C Publish Draft 09/97C Task 4 - Update prior NUREG documents on owners and antitrust license 02/97C conditions issue Task Order Contract 05/96C Draft NUREG Updated 09/96C Publish NUREGs 12/960 30
l Task 5 - Inedtutionalize Staff Level Contact with NARUC,SEC,FERC. ONGOING Develop MOUs as necessary.
Letter to agencies 06/96C Stafflevel meetings 11/96C Draft MOUs to Comnwesion (as required) TBD Sign MOUs TBD l Task 6 - Develop and implement rulemaking to clarify 10 CFR 50.80 if necessary Commission determination of need (Policy paper in preparation) 10/97C Policy options paper 8/97C Office Concurrences 9/97C ACRS Comments NA CRGR Concurrence N.A.
EDO Concurrence 10/97C Commission Approval 10/97C Putssh ANPR or Proposed rule 1/98C (
Public Cornment TBD :
Revise Rulemaking Package TBD OfRce Concurrences TBD 1 ACRS Comments TBD CRGR Concurrence TBD EDO Concurrence TBD Commission Approval TBD Puthsh Final Rule TBD TBD !
Task 7 -Complete on Decommissioning Funding Assurance Rule. ONGOING ,
Milestones for th6s task provided under rulemaking action, ,
"C+:=, . ' ' -7. .s Costs and Funding Evaluations" Descriollon The action plan is intended to address the Commission's concems regarding the impact of I'
. utility deregulation and resulting reorganizations and restructuring on licensee's Anancial qualifications and their ultimate ability to safely operate and decommesson their facellbes Historical Backaround in recent years, several restructurings and reorganizations have occurred with the electnc utility industry. In addition, State public utility commissions (PUCs) have increased pressure for improvements in economic performance of eixtric utilities they regulate in order to reduce the rates paid by wholesale and retail consumers.- The accelerated pace of this restructuring may affect the ability of power reactor licensees to pay for safe plant operations and decommissioning. Specifically, the restructuring may affect the factual underpinnings of the NRC's previous conclusion that power reactor licensees can retably accumulate adequate funds for operations and decommesioning over the operating lives of their fadlities.
Proposed Actions Specinc actions included in the action plan are: 1) issuing a policy statement delineating NRC's expectations with respect to future financial and anti-trust reviews and developing a standard review plan regarding NRC's current financial review requirements; 2) issuing an administrative letter to all licensees delineating their current responsibilities with respect to getting prior NRC approval for changes that may affect their previous financial qualification determinations or ownership; 3) formulating 31
l non-rulemaking periodic reporting requirements,4) updating NUREG documents containing financial information; 5) establisNng staff level contacts with the Securities and Exchange Commiseen (SEC), the Federal Enerpy Regulatory Commiselon (FERC), and the Natk>nal Association of Utility Regulatory j Commincions (NARUC); 6) implementing rulemaking if necessary on financial qualifications, including l 50.80 transfers; and 7) completing decommesioning funding assurance rulemaking. ,
Current Status' t Task 1: The final policy statement on industry deregulation & restructuring was published in the Federal Register on August 19,1997. The final SRP on antitrust was forwarded to the Commission on 10/3/97 (SECY-97-227). In that paper, finalization of the financial assurance SRP past the draft phase was tied to the ongoing decommissioning funding assurance rule (TAC # M93806).
Completion of this item b tow part of the implementation plan for this rulemaking.
Task 3: OfRce of Research has leeued a Reg Guide for periodic financial reporting requirements. 1 Task 6: The staff has prepared, and the EDO has sont to the Commesion, a policy options paper on financial qualllications requirements in light of restructuring (SECY-97-253; October 24,1997).
Task 7: The staff submitted a final rule on financial assurance for decommissioning in light of restructuring to the Commesion on June 30,1998.
NRR Technical Conlada- R. Wood, PGEB,415-1255 M. Davis, PGEB,415-1016 4
4 32
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EXTENDED POWER UPRATE ACTION PLAN TAC No. M91571 Last Update: 07/10/98 Lead NRR Division: DRPW l GSI: Al-182 Supporting Division: DSSA l MILESTONES DATE (T/C)
- 1. Receive GE Teoical ELTR1 (Generic Review Methodology). 3/95 C
- 2. Issue Staff Position Paperon ELTR1
- Meeting with GE/NSP. 4/95 C
- Identify differences between LTR1 and ELTRI. . 8/95 C
- Issue RAls as appropriate. 9/95 C
- Incorporate information on foreign experience obtained from SRXB. 10/95 C Develop power uprate database for all U.S. plants.
- Issue Staff Position Paper. 10/95 C 2/96 C
- 3. Receive GE Topical ELTR2 (Generic Bounding Analyses).
GE plans to submit ELTR2 in two parts: the first part in March 96 3/96 C and the second part in July 1996. 7/96 C
Meeting with GE/ industry. 2/96 C Issue RAl. 3/97 C
- GE response to the RAl. 7/97 C Receive revised ATWS analysis from GE. 5/98 C
- Input to the draft SE from technical branches. 5/98 C Issue draft SE to the ACRS. 5/98 C
- 5. Receive Lead Plant Application (Monticello). 7/96 C
- Meeting with Monticello. 10/96 C
- RAIinput from tech branches. 1/97 C
- Issue RAl. 4/97 C
- NSP response to the RAl. 9/97 C
- Receive revised application from Monticello. 12/97 C
- Issue additional RAls as appropriate. 2/98 C
- Input to the SE from tech branches. 5/98 C Issue draft SE to the ACRS. 5/98 C ACRS Presentation 6/98 & 7/98 C
- Issue SE. 8/98 T
- 7. Support the ongoing staff effort in developing a Standard Review TBD Procedure for power uprates. Incorporate lessons leamed from Lead Plant activity.
33
Descriollon This adion plan describes the strategy for completmg both the generic and plant-specific reviews for extended power uprate submittals for boiling water reactors (BWRs). General Electric Company (GE) submitted a licensing topical report (ELTR1), which outhnes the methodology for
'wi'-T " " -i of an extended power uprate program. ELTR1 encompasses power uprates of up to 120 percent of the original licensed thermal power. Individual plant submittals for uprates will likely contain requests for en op6 mum power level specific for that plant which is something less than the full 120 percent Each technical branch will review the applicable portons of both the ELTR2 (GE top' cal report containmg generic analyses) and the lead plant applicaton (Monbcello), and will provide input lito the staffs safety ,
evaluation reports. The experience gained from these reviews will be incorporated into the ongoing staff l effort in C;ig'-,g a standard review procedure for power uprates.
Historical Backaround. The generic BWR power uprate program was created to provide a consistent means for individual liconeses to recover addi6onal generating capacity beyond their current licensed limit in 1990, GE submitted licenang topical reports to initiate this program by proposing to increase the rated thermal power levels of the BWR/4, BWR/5, and BWRM product lines by approximately 5 percent.
Since 1990, the staff has reviewed and approved at least 10 such power uprate requests under this generic BWR power uprate program. As a follow-on to this program GE submitted ELTR1 in March 1995 to propose " extended" power uprates of up to 120 percent of the original licensed thermal power.
Proposed Actionr Specific actions included in the generic action plan are: (1) review ELTR1 and issue a staff position paper, (2) review ELTR2 and issue a safety evaluation report, (3) review the lead plant apphcation and issue a safety evaluation report, and (4) develop a standard review procedure based on ELTR1, ELTR2, and the lead plant review Orlainatina Document GE Licensing Topical Report (NEDC-32424)," Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," dated February 1995.
Reaulatory Assessment Not applicable. (A safety assessment is not needed for this action plan because a justificaton for continued operation of a plant is not required.) This program is an industry initiative that is stnctly voluntary.
l Current Status-Draft SERs for both the ELTR2 and the lead plant apphcabon were issued to the ACRS in May 1998. The ACRS met on Jurne 2,3, and July 8,1998, to review the staffs draft SERs. Final SERs are expected to be issued in August 1998. The Southem Nuclear Company submitted its appliathn for a 8% power uprate for Hatch in August 1997 following the Monticello's lead. Completion of the sta'rs review for the Hatch submittal is currently scheduled for September 1998.
NRR Lead PM T. J. Kim, DRPW,415-1392
. 1 s
4 34 i
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DRY CASK STORAGE ACTION PLAN Final Update TAC Nos.: M93927 (load / unload proc) Last Update: 6/18/98 M94107 (inspechon actMbes) Lead NRR Dmsion: DRPW M94108 (ac6on plan)
M96608 (ISFSIsupport pad)
GSI: Not Available MILESTONES DATE (T**/C)
- 1. Heavy Loads / Cranes
-develop working grcup plan 11/95C
- prepare & issue Bulletin 96-02 4/96C
-Issue Heavy Loads Acton Plan 6/97C
-complete Heavy Loads Acton Plan 6/98C*
a.(l) Movement of Casks Prior to Securing Lid
-issue RAI for Bulletin 96-02 responses 12/96C
- review site specific responses 12/97C
-Identify and resolve generic issue 4/98C
- 2. Cask Loading / Unloading Procedures
- contact Nuclear Energy Institute about industry efforts 8/95C !
l
- resolve high priontyissues 9/95C
-form working group 10/95C
- complete worldng group determination on further issues 4/96C
-issue information notice 7/97C
- 3. Inspecton Guidance
-issue revised ISFSI inspechon procedures 2/96C
-issue MC 2690 2/97C
-issue 10 CFR 72.212 and 72.48 inspechon procedures 6/98C"
- Revise MC 2515 Inspection Procedures for ISFSI 6/98C" support activebes
- 4. Enforcement Guidance
- establish working group 5/97C
-issue revised enforcement guidance pertaining 6/98C" to ISFSIviolations
- 5. VSC-24 Weld issue
-issue inspechon report 4/97C
-Issue Confirmatory Action Letters 5/97C
- review responses (a) loading of future casks 6/98C" (b) loaded casks 6/98C"
- address other cask designs / generic issue 6/98C"
- Closed. The Heavy Loads Action Plan is being tracked by the action plan program.
- Closed. These issues are being tracked by the SFPO Oparating Plan.
35
- 6. VECTRA QA Performance
-issue Demand forinformation (DFl) 1/97C
-review responses to DFl 8/97C f
- cerform vertlication inspection 6/98C"
- 7. Emergency Planning
-issue EP reviewAnspection guidance 6/98C"
- 8. ISFSI Suppart Pad Desi0ns l
- leeue desinnAnspection guidance for ISFSI support pods 6/98C"
- Closed. The Hoovy Loads Ar, tion Plan is being tracked by the action plan program.
" Closed. These losua= tra heing tracked by the SFPO Operating Plan.
Descriotion The Plan was originally c'ri-;+j in June 1995 to identify and resolve major issues and problems in the area of by cask stora0e of spent reactor fuel in independent spent fuel storage installations (ISFSis). Specific issues encomposeed by the plan include broad technical areas such as heavy load control and cask loadin0 /unloedmg, cask or vendor specific concerns such as the VSC-24 seal weld integrity issue and the performance of VECTRA in the area of quality assurance, and enhancement of inspection and enforcement guidance.
Historical Backaround The number of U.S. nuclear power plant licensees having or considering ISFSis is increasing sipruficantly. Licensees have encountered a number of problems during the fabrication, installation, and licensing of the existing ISFSis and there has been an inconsistent level of performance by involved licensees and cask fabricators with respect to the use of dry cask storage of spent reactor fuel. Because of the anticipated increased industry effort in this area, the staff needed to fully understand i
the problems that occurred and take appropriate measures to reduce such problems in the future.
Therefore, NMSS and NRR developed a plan to resolve major issues and problems. The plan was revised in June 1997 to remove some completed items as described in the referenced January 1997 memorandum and, thereby, improve the readability and usability of the document.
Onainatma Document Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995," Dry Cask Storage Action Plan."
1 Reaulatory Assessment The plan addresses dry storage of fuel that is several years old. Technical l Issues have been addressed on a site-specific basis for exishng facilities. The action plan willimprove j guidance, enhance communications with industry and the public, and aid future applicants.
Resolution A thorough review of the resolution paths for the remaining items in this action plan was conducted prior to the February 17,1998, NMSS/NRR Office Director's interface meeting between C. Paperiello and S. Collins. The results of the review provided information that presented a method by which to close the remaining open items in the action plan.
Item 1, Heavy Loads / Cranes, contains one previously open issue, the completion of the Heavy Loads Action Plan. The Heavy Loads Action Plan is already being tracked by the action plan program and therefore should not be tracked twice by the Dry Cask Action Plan. This item within this action plan is closed, but the Heavy Loads Action Plan will continue to be tracked as an individual action plan.
36
- - - - . . - - - - . - - ~ . - - - - - - . . . - _ . . - . . - - . . ~ - . . -
l' I
4 f item 2, Cask Loading / Unloading Procedures - this item was previously closed.
- llem 3, inspechon Guidance, contains two previously open lesues' the completion of 10 CFR 72.212 and 72.48 inspection procedures, and revision of IMC 2515 Inspecton Procedures for ISFSI support schvibes Comments on a draft procedure to inspect 72.212 souvides are under revow by SFPO as resources permit. Procedures on 72.48 activibes will be developed to support issuance of the final rule revising 50.59 and 72.48. Schedules for both of these schwilies are being hacked in the SFPO Operating Plan , ,
NMSS does not require input from NRR to complete these issues I Regarding the issue of revising IMC 2515, input from regional offices and field inspectors on the need for additional inspection guidance will be evaluated as part of SFPO's rouW,c bspection program oversight per IMC 2690. SFPO would then determine if changes to IMC 2515 inspechon procedures should be requested from NRR. Any such lesues idenulled by SFPO would be tracked in the SFPO Operating Plan.
If changes are required to IMC 2515, NMSS will formally request NRR to initiate the changes. Thin item is closed, item 4, Enforcement Guidance, contains one previously open issue for issuing enforcement guidance on ISFSI violations. The proposed enforcement guidance has been issued for comment. SFPO will resolve the comments and forward the guidance to OE for submission to the Commission. Schedules are being tracked in the SFPO Operaung Plan. NMSS does not require further input from NRR to resolve this issue.
Thisitem is closed item 5, VSC-24 Wold lesus, contains three previously open issues: reviewing responses for currently loaded casks, reviewing responses for the loadmg of future casks, and addressing the potential for
. generic implications The issues regarding the loadmg of future cask and currently loaded casks are being tracked in the SFPO Operating Plan. NRR has no further input on these issues The issue regarding other cask designs and generic lesues has been evaluated by NMSS and no immediate acbons are required. The issue has been referred to the Office of Research for long-term assessment of the corrective achons associated with generic implications of cask designs currently in use. This issue is tracked by the Generic System "m g+ ment and Control System Information. This item is closed.
Item 6, VECTRA QA Performance, contains one previously open issue regarding verification inspechons inspections will be scheduled when fabrication of new casks is underway. Inspections are tracked in the SFPO Operaung Plan. NRR has no input to the resolution of this issue. This item is closed.
Item 7 Emergency Planning, contains one previously open issue regarding review and irepechoa guidance. Staff guidance for reviewing emergency plans is provided in Appendix C of NUREG-1567,
" Standard Review Plan for Spent Fuel Dry Storage Facilibes " Guidance for conductirvs emergency plan inspectons can be obtamed from inspection Procedure 88050, Emergency Preparov. ness, which is for fuel cycle facillbes (Manual Chapter 2600). Emergency preparedness requirement:s for ISFSis and fuel cycle facili6es are similar; therefore IP 88050 is adequate. Emergency preparedness reviews and inspections are tracked in the SFPO Operating Plan. NMSS will make formal request of NRR for asestance on emergency planning as needed. This item is closed, item 8, ISFSI Support Pad Designs, contains one previously open issue regarding design and inspechon guidance for ISFSI support pods. A rulemaking plan has been approved that includes requirements for ISFSI support pods. The rulemaking schvibes and inspecton guidance are being tracked in the SFPO Operating Plan. NMSS will make formal request of NRR for assistance on this issue as necessary.
Thisitem is closed 1 Contacts Phillip Ray, DRPW/NRR,415-2972 Lawrence Kokajko, SFPO/NMSS,415-1309 37
. . _ __ . _ _ . _. . . _ . _ ._ _ __ __ _ _ _ . . ~ _
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References-Memorandum from Robert M. Bemero and William T. Russell to James M. Taylor, March 15,1995, "Reali0nment of Reactor Decom,T'- 's.irg Program."
Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995," Dry Cask Stora0e Action Plan."
Memorandum from Cart J. Paperiello and William T. Russell to James M. Taylor, January 25,1996,
" Update to the Dry Cask Storage Action Plan."
Memorandum from Carl J. Paperiello and Frank J. Miraglia to Hugh L. Thompson, January 30,1997," Dry Cask Storage Action Plan Update."
Memorandum from Carl J. Paperiello and Samuel J. Collins to Hugh L. Thompson, August 25,1997,
" Update on the Dry Cask Storage Action Plan."
38 ,
1
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ACCIDENT MANAGEMENT IMPLEMENTATION TAC #: M91986 - Overall . . Last Update: 7/1/98 M91641 -BWROG SAMG Review Lead NRR Division: DSSA MILESTONES DATE (T/C)
- 2. Review severe accident training materials and BWROG 6/95C prioritization rnethodologies
- 3. Develop guidance for A/M audits initial draft (forinternal use) 11/95C Industry-sponsored A/M demonstrations 3/98C Revised draft (to NEl and public) 7/98T Final 9/98T
- 4. Complete A/M audits 12/98T
- 5. Hold public workshop 2/99T
- 6. Report to Commission on audit findings and recommendations for 4/99T achieving closure Descrioton This action plan is intended to guide staff efforts to assess the quality of utility implementation
~ of accident management (A/M), and the manner in which insights from the IPE program have been incorporated into the licensees A/M program. Specific review areas will include: development and implementation of plant-specific severe accident management guidelines (SAMG), integration of SAMG with emergency operating procedures and emergency plans, and incorporation of severe accident information into training programs.
Histoncal Background The issue of A/M and the potential reduction in risk that could result from -
developing procedures and training operators to manage accklents beyond the design basis was first identified in 1985 [1]. A/M was evaluated as Generic issue 116 and subsumed by A/M-related research activities in late 1989. Completion of A/M is a major remaining element of the Integration Plan for Closure of Severe Accident issues [2]. The development of generic and plant-specific risk insights to support staff evaluations of utility A/M programs is also identified in the Implementation Plan for Probabilistic Risk Assessment [3]. NRC's goals and objectives regarding A/M were established at the inception of this program [4]. Generic A/M strategies were issued in 1990 for utility consideration in the IPE process [5].
The staff has contmuod to work with industry to define the scope and content of utility A/M programs and those efforts have culminated in industry-developed A/M guidance for utility implementation. Industry has committed to implement an accident management program at each NPP [6]. NRC has accepted the
- industry cammitment and developed tentative plans for staff evaluation of utility implementation [7].
Proposed Actions Specific actions included in the A/M action plan are: (1) complete the review of BWROG SAMG documents, (2) conduct A/M demonstration visits to observe how the elements of the formal industry position are being implemented, (3) complete A/M audit guidance using the information and perspectives obtained through the demonstration visits, (4) conduct A/M audits, and (5) hold a public workshop to discuss audit findings. Following the workshop, the staff will report to tha Commission on audit findings and recommendations for remaining actions to achieve closure.
Origmating Document SECY-88147, Integration Plan for Closure of Severe Accident issues, May 25, 1988.
39
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Raoulstory Assessment Accident mana0ement programs are being implemented by licensees as part of an initiative to further reduce severe accident risk below its current, and acceptable, level. Consequently, this is a non-urgent regulatory schon and continued facility operation is justified Current Status Severe accedent management guideline documents have been submitted by each of the PWR owners groups, and reviewed by the staff [8]. The BWROG submitted Rev. O of the Emergency Procedure and Severe Accident Guidelines (EP/ SAG) and associated technical basis documents to NRC ,
for information on August 29,1996 [9]. The staff and Oak Ridge National Laboratory have completed a hl0h level review of the EP/ SAG documents. Areas where addibonal information and discussion with the BWROG are considered necessary were idenblied in an April 2,1997, letter to the owners group [10]. A BWROG submittal describing a time line for acdons that operators would take acco,d g to the EP/ SAG i was received in May 1997 [11]. The BWROG response to the April 2,1997, staff letter was received in January 1998 [12]. The staff has completed its review of the BWROG response A letter to the BWROG providing staff views on the EP/ SAG 9nd expectations regarding plant-specific implementation of the generic guidelines is under review and is expected to be issued in July 1998. ;
Licensee target dates for completing AM implementation have been submitted to NRC. Implementation has been completed at approximately one-fourth of the sites. Implementation at the balance of sites, including most of the BWR sites, will be completed by the end of 1998.
The staff outlined plans to evaluate licensee AM implementation in separate communications with NEl and the Commession in 1995-1998 [13,14]. Major steps included: (1) conducting information gathering visits at two to four sites to observe how the elements of the formal industry posibon are being implemented, (2) completing a temporary instruc6on (TI) using insights obtained through the site visits, (3) performing pilot inspechons at about five plants using the TI, (4) developing an inspecbon procedure (IP) for use at remaining plants based on findings from the pilot inspechons and feedback from industry, (5) ovaluabng implementation at remaining plants using the IP, and (6) in the longer term, evaluating AM maintenance on a for-cause basis as a regional initiative.
In January 1997, the staff agreed to participate in a limited number ofindustry-organized AM demonstration visits in lieu of the information gathering visits, and to reassess the need for inspectiore at the remaining plants after the AM demonstrations. NRR staff also attended an NEl-sponsored workshop on acciderd mana0ement implementation in March 1997. The workshop provided an opportunity to better understand plant-specific implementation approaches and issues, and the major elements of implementation.
The AM implementation demonstration visits were completed in March 1998. A total of four sites were visited - Comanche Peak (5/97), North Anna (7/97), Duane Arnold (2/98), and Calvert Cliffs (3/98). No addibonal AM demonstrations are planned. The AM demonstration visits provided insights into the licenswe's AM implementation and evaluation process, and areas where changes to the guidance for evaluating AM implementation may be needed. However, the demonstrations did not permit sufficient time and flexibility for the staff to evaluate the licensee's supporting analyses and resolve several staff concems identified during the visits. The staff needs to better understand the implementation process to determine the effecbveness of the voluntary industry initiative, the significance of the issues identified in i i
the demonstration visits, and the need for inspections at remaining plants. Accordingly, the staff intends to proceed with further evaluations of AM implementation.
In previous communications with industry and the Commission (e.g., SECY-98-131 and SECY-97-132) '
I the staff had characterized the planned AM evaluations as inspections against licensee commitments, but upon further consideration of the voluntary nature of this program, has concluded that these evaluations should be performed as audits rather than inspections. The objectives of the audits will be to ;
assess how licensees have evaluated and implemented enhancements to AM capabilities in accordance with formal industry posibon, and to establish a basis for a decision regarding the need for future inspecbons or any other regulatory action. Remaining actions in the revised approach are: (1) complete the AM audit guidance, (2) conduct audits at 4-5 plants, (3) hold a public meeting / workshop to discuss 40
audit SndinOs and staffAndustry views on program completion, and (4) report to the Commesion on audit Andings and recommendations for remaining actions to achieve closure. The milestone chart has been modited to reflect the modiRed approach.
Several areas of the industry initiative needmg clarification were brought to NEl's attention by licensees during AM implementation. In response, NEl developed supplemental guidance to address these areas and provided this guidance to industry and to NRC in a July 22,1997, letter [15]. NRC provided comments on this guidance in a January 28,1998, letter to NEl [16]. In an April 3,1998, letter [17], NEl expressed concern that NRC appears to be reversing previously understood positions and escalating ,
^^ '
e:- + _ %-r,s. The staff positons on licensed operator training and evaluation, use of a systematic approach to training, and application of 10CFR50.59 were of greatest concem to industry. In a June 25, 1998, response [18), NRC provided clart6 cation regarding the staff positions and the approach to reaching closure. The staff indecated that they do not see major differences in NEl's and NRC's expectations, and that industry should continue to proceed with implementation. A draft Tl for use in planned pilot l Inspections was completed in February 1996, and discussed with industry, ACRS, and NRC Regional j ofRce staff in separate meetings in early 1996. The Tl has been recast as aucait guidance, and updated to incorporate insi0hts from the AM demonstration visits, staff poseons contained in NRC letters to NEl, and fsedback received on the draft Tl. The audit guidance is currently under review and is expected to be released by the end of July 1998. References-
- 1. Memorandum from F. Rowsome to W. Minners, "A New Generic Safety issue: Accident Mana0ement," April 16,1985
- 2. SECY-88-147, integration Plan for Closure of Severe Accident issues
- 3. SECY-95-079, implementation Plan for Probabilistic Risk Assessment
- 4. SECY-89-012, Staff Plans for AM Regulatory and Research Programs
- 5. Generic Letter 88-20, Supplement 2, April 4,1990
- 6. Letter from W. Rasin to W. Russell, November 21,1994
- 7. Letter from W. Russell to W. Rasin, January 9,1995
- 8. Letter from W. Russell to W. Rasin, February 16,1994
- 9. Letter from K. Donovan to Document Control Desk, August 29,1996
- 10. Letter from D. Matthews to K. Donovan, April 2,1997
- 11. Letter from K. Donovan to Document Control Desk, May 10,1997
- 12. Letter from T. Rausch to Document Control Desk, January 9,1998
- 13. Letter from A. Thadani to T. Tipton, August 3,1995
- 14. SECY-96-088, Status of the integration Plan for Closure of Severe Accident issues and the Status of Severe Accident Research
- 15. Letter from D. Modeon to G. Holahan, July 22,1997 i
- 16. Letter from G. Holahan to D. Modeen, January 28,1998 l i
- 17. Letter from R. Boedle to S. Collins, April 3,1998
- 18. Letter from S. Collins to R. Beedle, June 25,1998 NRR Technical Contact ~ R. Palla, SCSB,415-1095 NRR Lead PM Ramin Assa, DRPW,415-1391 41
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CORE PERFORMANCE ACTION PLAN TAC Nos. M01257 - DSSA Last Update: 06/30/98 M01602 - DRCH Lead NRR Dwision* DSSA GSI: Li-179 Supporting Division DRCH MILESTONES DATE (TP/C)
Task 1 - Inspection of Nuclear Fuel Vendors (DRCH) ongoing' Siemens Power Corporation [PWR AIT followup] 06/94C A88 Combustion Engineering [PWR reloads] 11/94C Toledyne-Wah Chang (TWC) 12/94C Sandvik Specialty Metals (SSM) 12/94C Westinghouse CNFD 07/95C General Electric NEP. 10/95C Frematome/Cogema Fuels (B&W Fuels) 09/96C GE (SLMCPR & low density pellets)" 09/96C SPC (comprehenswe re-inspechon of open items and new issues)" 04/97C GE (RWE/RBM issues and followup)* [ licensee enforcement pending] 07/97C ABB/CE [BWR] (WNP-2 transition core)* on-hold Westinghouse (re-inspection & recent issues) on-hold Siemens (re-inspection on corrective actions) 01/99T Task 2 - Inspection of Licensee Reload Analyses (DSSA) ongoing
- RI - 2 licensees: PSE&G (Hope Creek / Salem) 07/98T PP&L (Susquehanna) (enforcement pending) 10/97P Rif - 2 licensees CP&L(HB Robinson) on-hold SNOC (Hatch)[ enforcement pending] 12/97P Rlli-2 licensees: Comed on-hold Detroit Edison (Formi) [ enforcement pending] 08/97P RIV-2 licensees WPPSS(WNP2) 06/97C WNP2 re-inspechon for enforcement issues 02/98P Union Electnc (Callaway) 11/97C Task 3 - Core Performance Data Gathering / Evaluation (DSSA) ongoing Regions - Moming Reports & Event Nctification ongoing
- Other- Data Acquisition and Collation ongoing PNNL - Core Performance Evaluation Analysis (CY95) 12/97P Task 4 - Participation of Regions in Action Plan (DSSA) ongoing idonellcation of Vendor lesues Feedback from Licensee inspections Counterparts Meetings (RI-RIV)
Task 5 - Evaluate inspecton Guidance (DSSA/DRCH) ongoing Evaluate Results of Licensee inspections incorporate Feedback from Region inspectors Draft Guidance for Resident and Region Inspectors 09/98T Draft inspecton Criteria and Action Plan Update 09/987 42
l Task 6 - Evaluate Licensee / Vendor Lead Test Programs for identification of 12/97P' Core Performance Problems (DSSA)
! Task 7 - Workshop on Core Perfonnance issues (TAC No. M95674)
Identityissues 07/96C Conduct workshop 10/96C Followup on Comments and Questions (RIC session) 04/97C l I
- lasuE DRIVEN '
Description
- The action plan is intended to assess the impact of reload core design activities on plant .
safety through inspections of fuel vendors, evaluation of licensees
- reload analyses, and independent evaluation of core performance information, with regional training and interaction.
Historical Background he action plan addresses the review of fuel fabrication, core design, and reload analysis issues that were discussed during 1994,1996, and 1997 briefings given to the Executive Director {
for Operations, and the 1997 Chairman briefing. The briefings presented by the Reactor Systems Branch 1 (SRXB), Division of Systems Safety and Analysis (DSSA), covered generic fuel and core performance !
~ issues and related evaluations of fuel failures. The former Special Inspection Branch (PSIB), Division of l Inspection and Support Programs (DISP), supported the briefings. As a result of these briefings, the l Office of Nuclear Reactor Regulation (NRR) was directed to expand the action plan to monitor and improve core performance in operating reactors to include focus on licensee activities and the licensee /vendorinterfaces.
Proposed Actions Specific actions included in the action plan are: (1) evaluate fuel vendors'
- performance through performance-based inspections that evaluate the reload core design, safety analysis, licensing process, fuel assembly mechanical design, and fuel fabrication activities; (2) evaluate t the performance of licensees that perform core reload analysis functions; (3) identify, document, and categorize cer: pctformance problems and root cause evaluations that will be further evaluated during these inspections and provide input to SALP evaluations as well as regional enforcement actions, as appropriate; (4) train and coordinate regional support staff participating in these activities; and (5) evaluate the results of these activities for use in formulating generic communications, revisions of regulatory guidance and guidance for regional inspectors, and other appropriate regulatory actions. In addition, as a result of recent generic concoms, including the failure of control rods to fully insert, the action plan was expanded to review the adequacy of vendor lead testing programs for new fuel designs (Task 6); and a workshop was conducted on core performance issues (Task 7) in the fall of 1996. The status of core performance inspection evaluations and emerging issues was covered at the last Regulatory information Conference.
DSSA - The action plan identifies that licensee inspections in each region shall be performed, in coordination with the regional inspectors, to assess licensee performance in reload core analysis oversight and participation. Licensee inspections will normally be issue-driven. The data acquired through licenses / vendor inspections will be integrated with information supplied by the regions and other sources and will be evaluated for generic core performance indicators and industry conformance to current regulatory requirements. The end product of the initial assessment willinclude guidance for resident inspectors and regional staff. The ongoing activities to capture and address early waming of
_ emerging issues will continue into FY98, and the action plan reflects the planned inspection of 5 licensee / plants, and four anticipated event-reactive inspections.
DRCH - The action plan currently identifies 10 completed and 1 planned vendor inspections that shall be performed by multi-disciplined inspection teams led by HOMB with contracted technical assistance, as required. These inspections are currently scheduled to be conpleted in 1998c The planned inspections 4 of ABB/CE and Westinghouse are on hold, pending re-evaluation of the vendor inspection function within
[
~
NRR.
43 4
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i Onoinalina Document Memorandum from Gary M. Holahan and R. Lee Spessard to Ashok C. Thadani, dated October 7,1994, " Action Plan to Monitor, Revww, and improve Fuel and Core Components
< Operating Performance" and the enhanced focus on licensee reload design partopation resulting from i brienne seedback.
Reaulatory Anasesment Coro design is a fundamentgl component of plant safety because maintaining fuel integrity is the first principal safety berrier (i.e., fuel cladding, reactor coolant system boundary, or the >
containment) against serious radioacthe releases. Likewise, the safety analyses must be property j performed in order to verify, in conjunction with stastup tests and normal plant parameter monitoring, that :
the core reload design is adequate and provide assurance that the reactor can safely be operated.
Evaluadon of actMeles that affect the quality of fuel and core components are impoetant to ensure that i safety and quality are not degraded and that the core performs as designed.
Current Status-DSSA - The data acquired from the ongoing vendor inspecbons are being evaluated for generic impact j and identification of emerging issues The recent issue-driven inspections at GE and Siemens, were supported by SRXB/DSSA staff and contract speculists in reload design. Interaction with the regions is ongoing to participate in region-led licensee inspechons SRXB has participated in two Region I and one Region ll inspector counterparts meetings. DSSA is re-evaluating the action plan results to date to better s
integrate and prioritize its actMbes, consistent with the DSSA operating plan. Options and 4
recommendabons for management review are being prepared for a closeout plan to capture the lessons loomed and to provide guidance for regional staff.
DRCH - The remaining issue-driven vendor inspectons include assessment of ABB Combusbon l Engmeering's supply of a BWR tranelbon core reload for WNP-2 and re-inspections of Siemens on 4
correctve actions and Weebnghouse on recent claddmg issues DRCH is re-evaluating options for the I inspection program, and only the Siemens re-inspection for DFl and n in-conformance issues is still .
scheduled. !
NRR Technical Contacts E. Kendrick, SRXB,415-2891 G. Cwalina, HQMB,415-2983 4
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FIRE PROTECTION TASK ACTION PLAN
- i Final Update l TAC Nos. M86652, M82809, M84592, Final Update: 07/06/98
- M85142, and M89509 Lead NRR Division: DSSA l GSI: LI-181 l l
MILESTONES DATE (T/C)
- 1. Annual Commission status report Final: 06/09/980
- 2. Recommendations for action (Part I) 06/09/98C
- 3. Recommendations for future study (Part II) 10/96C
- 4. - Confirmation issues (Part lil) 10/960
- 5. Other issues (Part IV) 08/95C Descripton The Fire Protection Task Action Plan (FP-TAP) was used to track and manage implementation of the recommendations made in the " Report on the Reassessment of the NRC Fire Protection Program," of February 27,1993.
Hatorical Backaround in February 1993, the Office of Nuclear Reactor Regulation (NRR) completed a reassessment of the reactor fire protection review and inspection programs in response to programmatic concems raised during the review of Thermo-Lag fire barriers. The results of the reassessment were documented in the
" Report on the Reassessment of the NRC Fire Protection Program," of February 27,1993. The staff prepared the FP-TAP to implement the recommendations made as a result of the reassessment report. <
Actions The FP TAP tracked the implementation of a wide range of technical and programmatic fire protection issues. It included recommendations for action (Part I), recommendations for further study ;
(Part II), confirmation issues (Part lil), and lessons loamed (Part IV). !
l Originatina Document " Report on the Reassessment of the NRC Fire Protection Program," l February 27,1993. l Regulatory Assessment Each' operating reactor has an NRC-approved fire protection plan that, il property implemented and maintained, satisfies 10 CFR 50.48, " Fire protection," and General Design Criterion 3, " Fire protection." Therefore, each plant has an adequate level of fire safety and the individual action plan items are receiving appropriate priority. 1 Current Status: The FP-TAP is complete. The staff issued the final report on the FP-TAP on l June 9,1998, in a memorandum from L. Joseph Callan, EDO, to the Commission entitled " Annual Status Report on the Fire Protection Task Action Plan and Plant-specific Thermo-Lag Corrective Action Programs."
The staff has completed its small-scale fire barrier scoping tests at the National Institute of Standards and Technology (reference " Report of Test FR 4008'). The staff continues to review fire barrier issues on a case-by-case basis as needed The staff issued information Notice (lN) 97-59, " Fire Endurance Test Results of Versawrap Fire Barriers," August 1,1997, and IN 95-52, Supplement 1, " Fire Endurance Test Results For Electrical Raceway Fire Barrier Systems Constructed From 3M Company Interam Fire Barrier Materials," March 17,1998, in response to a task interface agreement issued by Region 11, the staff is evaluating the fire-resistive capabilities of Knowool fire barriers. The staff will complete this review and will 45
. . . . ~ . . - . .. -. -. .~--.--- . _ _ _ - - - . - - - . -
i e
review Are border issues on a case-by-case basis as warranted.
The staff, with the technical assistance of Brookhaven Nabonal Laboratory (BNL), has developed a prahahia= air risk acessement (PRA) model for assessing the risk associated with post-fire safe-shutdown !
methodologies that impose a self4nduced station blackout. As reported in the status report of October 31,1996, the staff presented the draft station blackout study to the Advisory Committee on l Reactor Safeguards (ACRS) on February 29,1996. Sece submission of that status report, the staff has ;
considered the recommendations made by the ACRS and has submitted a revised scope of work to BNL -
addreeoing the ACRS recommendations. BNL has revised the subject report to address the ACRS !
comments The staff intends to apply the revised PRA model as part of its FPFI program to assess the
- risk significance of entering a self-induced stabon blackout. Therefore, the FPFI addresses this recommendation.
t in a Staff Requirements Memorandum of (SRM) February 7,1997, the Commiselon agreed with the fire :
protection functional inspection (FPFI) program described in SECY-96-267. The program described I included a pilot program followed by a permanent program. The pilot program consists of four pilot inspections (one per region) and an industry workshop l
i The staff conducted the first pilot inspection at River Bend in June 1997, the second pilot inspection at Susquehanna during November 1997, and the third pilot inspechon at St. Lucie during April 1998. The final pilot inspection, which will be conducted at Prairie Island, has been delayed until summer 1998 because of the need to divert the staffs hmited fire protecton inspection resources to a fire protecton/ post-fire safe-shutdown inspecton at Quad Cities.
The SRM of February 7,1997, requested that at the completon of the pilot program the staff provide a j report to the Commiselon which r8imeu==== the FPFI program and plans for the program. The report was scheduled to be completed during June 1998 (WITS 9700021). However, because of the program delays describert above, the FPFI pilot program described in SECY-96-267 will not be completed within the originally proposed schedule. Therefore, the staff will not issue its final report to the Commission during July 1998. However, the staff will provide the Commisalon with an interim pilot inspection program raport in June 1998 describing inspection results to date and the plans to complete the pilot program. Post-pilot inspection program activibes will include a public workshop in the Fall of 1998 to discuss inspecton results and request public and industry feedback. The staff is tracking its remaining FPFI acevities under WITS 9700021 and TAC M90284. q The final' Annual Status Report on the Fire Protechon Task Action Plan and Plant-specific Thermo-Lag Corrective Action Programs," dated June 6,1998, completes Part i of the FP-TAP. The completion of Part i 11 and Part til of the FP-TAP is documented in a memorandum of October 31,1996, from J. M. Taylor, EDO, to the Commission," Semiannual Report on the Status of the Thermo-Lag Action Plan and Fire Protection Task Action Plan." l Contact- D, Oudinot, DSSA,301-415-3731 ;
References- !
" Report on the Reassessment of the NRC Fire Protection Program," of February 27,1993.
I SECY-95-034," Status of Recommendations Resulting From the Reassessment of the NRC Fire i Protection Program," February 13,1995.
Memorandum of October 31,1996, from J. M. Taylor, EDO, to the Commission, " Semiannual Report on i the Status of the Thermo-Lag Acton Plan and Fire Protection Task Action Plan."
SECY-96-267, " Fire Protection Functional Inspecbon Program," December 24,1996.
Memorandum of June 9,1998, from L. Joseph Callan, EDO, to the Commission, ' Annual Status report on the Fire Protecton Task Acton plan and Plant-specific Thermo-Lag Corrective Action Programs.'
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HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES I TASK ACTION PLAN !
(Previously Part of the Dry Caok Storage Action Plan) .
TAC Nos. M93821: Action Plan Last Update: 7/7/98 _
i M91955: DSC generic review Lead NRR Division: DSSA i M95546: Generic review of NRCB 96-02 ACTION DATE (T/C) l
- 1. Review, summarize and issue existing NRC guidance on heavy load control.
- Review NUREG-0554, NUREG-0612, GL 80-113, GL 81-07, GL 8511, and 2/96C i other supporting documents.
- Develop summary of guidance 2/960 i
- 2. Determine signliicant heavy load issues that need to be addressed and develop resolution method. !
- Generic letter 85-11 and NUREG-0612. 2/96C
- Single-Failuro-Proof Crane (reliability). TBD
- Spent fuel cask drop accident prior to securing the lid. . 2/960 l
- Risk significance of multiple failures within safe load path. TBD i
- 3. Review licensee implementation of heavy load control, including applicable (ongoing) '
correspondence from a sample of licensees and site visits.
4.L Review NRC audit / inspection procedures, practices, inspection reports, enforcement actions, and experience.
- Conduct review. 5/960
- 5. Document the staff's position on heavy loads issues. Determine a proposed method of disseminating this information to the staff and industry as
appropriate and issue.
Issue bulletin on load movement during operations. 4/96C
- 6. Draft staff guidance and disseminate to appropnate management (SPLB, (TBD)*
Region I, NRR) and obtain/ resolve any comments. (propose form of guidance).
(contingent on resolution of item 2 above)
- 7. If an inspection procedure (or procedures) is planned, issue the inspection 12/98 procedure (s)in draft.
from industry representatives and resolve any discrepancies with the industry position. *
- 9. - Develop final version of guidance and obtain management concurrence. (TBD)*
10.- Issue finalinspection procedures. (TBD)
- 11. Issue final guidance TBD Note: --
Indicates that the activity is contingent on the results of NPR's/RES's review of the risk
- of crane failure during the movement of heavy loads.
47
Descriobon The Heavy Load Control (HLC) and Crane issues task action plan willidentify, evaluate and resolve major issues and problems regarding the handling of heavy loads (i.e., spent fuel assemblies, spent fuel dry storage casks, reactor cavity biological shield blocks, reactor vessel head, etc.) within nuclear power plants. (See the Enclosure for a detailed descriphon of the scope of the actions under the action plan).
Historical Backaround Recent increases in licensees' actwlbes involving the transfer of spent fuel to independent spent fuel storage installations (ISFSis) have given rise to a number of concarns with NRC's regulatory program for the control and handling of heavy loads, and with the licensees' programs for complying with the requirements in NRC's exishng guidance. For example, there are concerns regarding what is required for the movement of heavy loads while the plant is operating. Because of anticipated future increases in industry efforts in this area, the staff needs to fully understand the exisbng problems and to undertake efforts to reduce such problems in the future. This plan was identified as a near-term issue under the dry cask storage action plan, end was recently revised to better reflect the scope and magnitude of the task.
Proposed Achons Achons included in the plan are: (1) understand the current regulatory framework and inform the staff; (2) review the general issues and iden#fy specific problems to be addressed; (3) develop correc#ve schons to resolve the problems; and (4) implement the correctwo actions. Specific corrective actions may include the issuance of guidance to licensees aler1ing them to the potential problems and requeshng that correctwo measures be taken to preclude accidents.
Onainatina Document Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995, " Dry Cask Storage Achon Plan."
Renulatory Assessment Several licensees have either develope,d or are implementing plans to move heavy loads in various areas of nuclear power plants (i.e. offloading spent fuel via dry storage and/or transfer casks for transfer to ISFSis) over safe shutdown equipment, and over spent fuel during plant operation. Queshons have been raised regarding the adequacy of NRC's guidance and the licensees
- methods of precluding an accident, and in the event of an accident, mitigating the severity of the consequences. An urgent NRC Bulletin (NRCB) 96-02," Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment," has been issued to alert licensees to the concems. As a result of the bulletin, several licensees have undertaken efforts to assess their plans, capabilldes, and licensing basis for heavy loads. The action plan is intended to improve NRC guidance, address site specific issues, and aid licensees in their future plans to move heavy loads.
Currer,t Status Review of the responses to Bulletin 96-02 was closed in April 1998. Projects continue to issue hcensee specific closeout letters. The staff continues to interact w6th licensees on a plant specific basis.
Staff efforts to work with RES to evaluate risks of crane failures were abandoned due to budget shortfalls.
The staffis examining opbons for any further effosis, considering budget limits. The staff may undertake limited efforts to understanti current crane failure data, assess the risks involved and make a decision on the future direction towards establishing a
, regulatory posjtion The staff visited Calvert Cliffs in 1997 and plans to visit 2 sites in 1998 for the purpose of obtaining an understanding how the various elements of the licensees' programs are peing implemented. Information and perspectwas gained through these visits, as well as input from the Regions, could be used to help determine and develop further guidance.
NRR Contacts Brian E. Thomas, DSSA,415-1210 Phillip M. Ray, ADPR,415-2972 4 Joseph E. Carrasco, RGN-l/DRS, (610) 337-5306 4
48
References.
Memorandum from Robert M. Bernero and William T. Russell to James M. Taylor, March 15,1995, "Reali0nment of Reactor Decommissioning Program."
Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995, " Dry Cask Stora0e Action Plan."
l 49
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HIGH BURNUP FUEL ACTION PLAN f
TAC NO. M91256 Last Update: 7/2/98 l Lead NRR Division: DSSA '
GSI: 170 ,
Supporting Office: RES MILESTONES DATE (T/C)
- 1. Issue user need letter to RES 10/930
- 2. Contracts issued by RES l 03/94C j
- 3. Schedule and coordinate meetings with foreign experimenters and regulatory 09/950 authorities -
- 4. Issue Information Notice (IN 94 64) Announcing new RIA data 08/94C
- 5. Present high bumup data at water reactor safety meeting 10/940
- 6. Schedule / coordinate industry meetings to discuss actions 10/94C
- 7. Determine nood for further generic communications 11/940 l
- 8. Issue letter to vendors 11/94C '
- 9. Issue IN 94-64, Suppl.1, Providing Data and Vendor Letter 03/95C
- and Plan of Action 03/95C' 01/960
- 11. Review industry (NEI) Response - 09/95C
- 12. Assess effects on design basis accidents of reduced failure threshold for high 09/95C bumup fuel
- 13. Committee on the safety of nuclear installations specialists meeting on the 09/95C transient behavior of high bumup fuel
- 15. Issue ltr to NEl assessing industry actions (vendor /EPRI response to IN) 8/97C
- 16. Water reactor safety information meetings (high bumup session) 10/95C core performance issues workshop 10/960 )
17.- Complete review of available fuel transient data relevant to design basis event 4/97C
- 18. Develop interim acceptance criteria (e.g., Based on cladding oxide) 4/970
- 19. Meeting with NEl and industry on interim criteria 11/97C
- 20. Complete agency program plan on high bumup fuel; consider possible 7/98C revision to SRP
- 21. RES briefs ACRS and completes response to NRR user need letters 04/960 6/98 P 11/98T 22.' Establish schedule for LOCA resolution and final assessment 12/98T Determine need for further regulatory action i RES HAS PRK)RITIZED AS GENERIC ISSUE #170 NUREG-0933.
j 50
1 Descriotion: The action plan covers assessment of fuel performance for high bumup fuel and evaluation -
j of the adequacy of SRP licensing acceptance criteria.
j Historical Backaround Recent experimental data on performance of high bumup (>50 GWDMTU) urxler reactMty inserton conditions became available in mid-1993. The unexpectedly low energy deposition ,
(30 CAUGM) to initiation of fuel failure in the first test rod (at 62 GWDMTU) led to a re-evaluation of the licensing beels assumptions in the SRP. As a result, the office of nuclear reactor regulation (NRR) was requested to prepare an action plan,in coordination with the Office of Nuclear Regulatory Research l (RES),
t
- Proposed Actions After a preliminary safety assessment was performed, an action plan was developed, to include a user need letter to RES and the issuance of contracts to assess all aspects of the high -
bumup fuel issue. Concurrently, meetings would be scheduled with the non-domestic experimenters and !
regulatory authorities to discuss the experimental data and to assess potential consequences and regulatory actions. Meetngs with industry would be scheduled to discuss their planned actons and to solicit cooperation with the safety evaluations. Based on a complete review of all available fuel transient i . date, relevant to design basis events, NRR/RES would define acceptance criteria, establish a schedule for
! final assosoment, and state need for further regulatory action.
- Oriainanna Documents Commiselon Memorandum from James M. Taylor (EDO), "ReactMty Transients i and High Bumup Fuel," dated September 13,1994, including IN 94-64, ' Reactivity insertion Transient and i Accident Limits for High Bumup Fuel,' dated August 31,1994. Commission Memorandum from
, James M. Taylor, "ReactMty Transients and Fuel Damage Criteria for High Bumup Fuel," dated
- November 9,1994, including an NRR safety assessment and the joint NRR/RES action plan.
j Reaulatory Assessment There is no immediate safety issue, because of the low to medium burnup in j currently operating cores. Since the fuel failure threshold declines with increasire burnup, the licensing 4 basis design acceptance criterie may need to be redefined as a functon of bumup. The end product of the plan will determine the need for regulatory action and will establish and define the need for further action on extended bumup cycles and high bumup fuelissues Current Status An ACRS Subcommittee Meetmg on the status of RES contractor programs was held in 4/96. An NEl letter summarizing the industry posibon was received in April, and the EPRI report i supportng this positon was sent by NEl on 9/20/96. A comrmesion paper on the status of the high burnup i issue and planned actions was propered by NRR, has been reviewed by RES, and was issued on November 25,1996. A Commission briefing was completed on March 25,1997.
i i A letter with an enclosure was issued on 8/12/97 to respond to the NEl and industry report on high burnup 8
fuel. A meeting with the industry was held on November 18-20,1997, to discuss the interim criteria and
, exchange information, thus completing the Milestone 19. An agency program plan for high burnup fuel
} for Milestone 20 was completed in 7/98. RES briefing ACRS was completed on 6/4/98.
l NRR Technical Contacts M. Chatterton, NRR/DSSA/SRXB,415-2889
< Shih-Liang Wu, NRR/DSSA/SRXB,415-3284 Edward Kendrick, NRR/DSSA/SRXB,415-2891
! RES Contact. Ralph Meyer, RES/ DST /RPSB,415-6789 l
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- - - , - . . = , - - , - - - - , .~ r . . -,, ,.u. ,-. ,_ w .._. - - - - - , . . - . ,
WOLF CREEK DRAINDOWN EVENT: ACTION PLAN TAC Nos.:' M66278 - Last Update: 6/30/98 Lead NRR Division: DSSA MILESTONES DATE (T/C) _
- 1) Draft Generic Letter (GL) 11/95(C)
- 2) lasue Supplement to IN 95-03 03/96(C)
- 3) Complete Draft Tl/ lasue to the Regions for Conrnents 07/98(T)
- 4) CRGR Concurrence of the GL for 1st time 09/96(C)
' CRGR Concurrence of the GL for 2nd time (after reconciling Public Comments)' 01/98(C)
- 5) ACRS Briefing 11/97(C)
- 6) GLlasued 05/98(C)
- 7) Receive Regional Comments on Tl 09/98(T)
- 8) Complete Evaluation of the Responses to the GL 01/99(T)'
- 9) lasue Tl 01/99(T)
- 10) Complete Inspections (As necessary) - 4/99(T)
==
Description:==
Ths objective of this action plan is to collect and evaluate information from the licensees -
regarding plant system configurations and vulnerabilities to draindown events. A 10 CFR 50.54(f) letter will be issued to gather the information which will enable NRC staff to verify whether addressees comply with NRC regulatory requirements and conform with current licensing bases for their facilities.
Historical Background . On September 17,1994, the Wolf Creek plant experienced loss of reactor coolant system (RCS) inventory, while transitioning to c refueling shutdown. The event occurred when operators cycled a valve in the train A side of the RHR system cross <onnect line following maintenance on the valve, while at the same time establishing a flow path from the RHR system, train B, to the refueling water storage tank for reborating train B. The failure of the reactor operating staff to adequately control two incompatible activities resulted in transferring 9200 gallons of hot RCS water to the RWST in 66 seconds.
The Wolf Creek event represents a LOCA with the potential to consequentially fail all the ECCS pumps and bypass the containment. Another important feature of this event is the short time available for corrective action. Based upon calculations by the licensee and the staff, it is estimated that if the draindown had not been isolated within 3-5 minutes, not positive suction head would have been lost for all ECCS pumps, and core uncovery would follow in about 25-30 minutes. This event represents a PWR vulnerability which was not previously recognized.
Proposed Actions. Specific actions of this generic action plan are: (1) issue IN 95-03 (issued January 18, 1995) and supplement to IN 95-03 (issued March 25,1996), (2) Request all PWR licensees, via an information gathering (10 CFR 50.54(f)) Generic Letter (GL), to provide information on draindown ,
vulnerabilities and the measures they implemented to diminish the probability of a draindown. !
Originatina Document. AEOD/S9541, " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994".
52 l
I Reaulatory Aaesamment The staff performed an evaluation of the probability for eventinitiation and of the ,
condisonal core dema0e probatility. The value of this probability for core damage along with licensee l awareness for this acenario makes the risk for continued PWR operation acceptably small. l Curter 18tatus information Notice IN 95-03, and its Supplement have been issued. CRGR concurred the proposed GL in 9/96; but as directed by an SRM, the GL was published in the Federal Register in 2/97 for public comments. ACRS was briefed on 11 A5/97. 2nd CRGR concurrence was obtained in 1/98. The GL was issued on 5/28/98. Staff is properir'g draft Tl for issuance to the Regions for comments. j NRR Technical Contact M. M. Razzaque, SRXB,415-2882 NRR Lead PM Kristine Thomas, NRR,415-1362
References:
- 1) AEOD/S95-01," Reactor Coolant System Blow.down at Wolf Creek on September 17,1994."
- 2) IN 95-03, issued January 18,1995.
- 3) Supplement to IN 95-03, issued March 25,1996.
- 4) Generic Letter 98-02," Loss of Reactor Coolant inventory and Associated Potential for Loss of i Emergency Mitigation Functions while in a Shutdown Condition," issued May 28,1998. )
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ATTACHMENT 2 GENERIC COMMUNICATION AND COMPLIANCE ACTIVITIES
DIRECTOR's QUARTERLY STATUS REPORT July 1998 Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Cosep LA Cesap Title Description Associate Dieteter for Projects Technical Specifications Branch M990ee OL JW5hapaker -./- /- 7/13/98 L OI; C: , ,ofTeduncal Alerts licensees to problems the NRC staffhas noted in the dnpoestxwung of Specificahons Discovered Not to be tedaucal spectScaixma found to specify parameter values or required actxes Sufficiers to Assure Plart safety th( even ifcomplied with, would not assure safety.
_T3B has 1 GCCA(s)
ADPR has a total of 1 GCCA(s)
Division of Engineering ,
Civil Engineering and Geosciences Branch M97920 OL JW5hapaker 08/15,98 10/30.98 T GI2 Seismic Capability ofTherrnal-l.ag Informs aderssees about :=A==t seismic capability of% panels Panels in high 6, J areas ofplares, and need for corrective actions.
M97981 OL WFBurton 08/31/98 6/30,98 L IN: Settiensere Mansonng and Inspection of Informa addressees ofneed to review subfoundaban designs and, as Pfare Structures Afkded by Degradation of appropriate, describe plans for foundation settlement monitoring Perous Concrete Subfoundations M99394 LT TAOrtene 09/30/98 10/30,98 T IlOLTEC Part 21 Computer CodeIssue To review infonnehon subnuned by llohec heernational concermng ANSYS compute code.
MA0675 IN CDPetrone 08/31/98 10/31/98 T IN: BWR Overhead Crane Design Basis Alerts licermees to the fact that the overhead anne and reactor building l Calculations Do Not Address Occurrence of ,a A may not be analyzed for a safe abutdown earthquake when the Seurruc Event When The Crane Is in Use erane is in operation and is act in the parked pondon.
Mall 42 IN TAGreene 09/30/98 11/30/98 T IN: Tendon surveillance Analysis To inform licensees in the proper use ofobserved data points for performing regression analyse to ensure that the irsegnty of the tendon system is maintained throughout the plare's life.
i ECGB has 5 GCCA(s)
Electrical Engineering Branch MAe734 IN DIEkeen -/-/- 12/30/98 T IN: 1.sw & Medium Vohage Circuit Breaker Provides exarryles ofbreaker failures dae to inadequate rnmintenance or Failures .J L.2.. ;.
'Page 1 of 7 21-Jal-98
Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Comp Title Description MA1623 LT JRTappert 4-/- 6/15SS L LT: Review of AEOD Study on Review and prmide comments on draft AEOD study concerrung W4. RPSUnevailability Westinghouse RPS unavailability.
MA2276 IN TAOreene -./-./- IN: EDO Configurshon Control Errors Inform licensees about recent inspedian fmdings related to configurabon control ofsafety4 elated equipmert.
EELB has 3 GCCA(s)
Materials and Chemical Engineering Branch M95279 GL JWShepaLar 08/3168 1000/98 T 01; Modification of the Requirements for Extending to operating reactor licemees, on voluntary basis, relaxations in
' Pest-Accident Sampling System PASS program requirements.
M95444 LT TAOreene 12/31/99 3/1539 T lead Teclencal Review -Induction flest Cracking has been found in several utilities
- austermic stainless steel piping Stress L, -. ; for Stainicas Steel Piping which had been subjected to 11151 in the 1980's . Staffconcerns include that I1151 may not have been properly applied.
M99226 GL JWShepaker 0900,98 8/28.98 T GL: Augmented Inspectmn issue for Small Pr ynnes augmented espechan ofsenali diameter, Class I piping in PWR Diameter, Class I Pipeg in PWR Iligh- high-pressurunjection systerra to overcome ASME code oversighL L. '-J.c System M99432 GL JWShepaker 12GIS9 900.98 T GL: Steam Generator Tube Integrity Informs licermees that actions beyond current TS requirements may be inacessary to ensure steam generator tube integrity.
M99958 GL JWShapaker /-/- 10/30,98 T CL: RCS Chemistry Effects on Flaw Growth During the review of a flaw evaluatian for a crack in the core shroud it was Eshmates observed that the facility had less restrictive TS requirements for pnmary water chermstry controls then were relied upon by the licensee in establishing a crack growth rate.
MAI689 GL JWShepaker I1/11/11 I1/30/98 T OL: NRC Staff Review of BWRVIP-05 MA2634 IN EJBenner 07/31198 9/15/98 T IN: Oracking of SG Tube Ends Informs licermees the fmdings from the - --. --- ' --+ of steam generator tubes in PWR EMCB has 7 GCCA(s)
Mechanical Engineering Branch M96354 LT TAOreene -/- /- 900/98 T Cornamment Ra.a-m . Spray and Millstone 3 determmed that the m: 2 .. ; recirculation spray and quench Quench Spray Pipeg Outside Design Basis arvey piping and supports could be subjected to higher accident t .. ,..a -
than those previously assumed in the design basis.
MA1798 LT CDPetrone 06/1399 5/13.99 T LT: Air Operated Valves AEOD is doing a review of problens with air operated valves in the industry and will prepare an IN to summarize the results of the review.
Page 2 of 7 21-Jul-98
Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Comp Title Description MA2997 IN CDPetrone 4-/- IN: leaks in the Emergency Diesel Alerts licerwees to leaks in the skid <nounted tube oil pipmg ofemergency Generator Lebe Oil Piping diesel generators MA1998 IN CDPetrone 12/30,98 Sup.toIN 9M8,hiotor-Operated Valve Aleits licensees to recerd guidance frorn IJnutarque Corporation for the Performance issues predictiori of torque output from its motor actuators used to open and close motorwed valves. .
EMEB has 4 GCCA(s)
DE has a totalof 19 GCCA(s)
Division of Reactor Controls and Human Factors Instrumentation and Controls Branch MA2o99 IN TAOrtene --/- /- IN: LW.ent Actuation of Stan&y Uquid Informs licensees of the root cause for an automatic actuation of the saaney Control System liquid control system that ocrurred at an nuclear power plard.
HICB has 1 GCCA(s)
Operator Licensing Branch MA0311 IN RAHenedict iI/tI/lI IN: Developing NRC Exammation Sample 1;armee prepared exams not teduncally accurate, are of flawed A,s rian and are too easy.
150LB has 1 GCCA(s)
Quality Assurance and Maintenance Branch M98441 GL JWShapaker 07/17,98 9/18.98 T OL:Quatriy Assurance of Electroruc Records in view oftechnological 2.-~.-.c._, changes in NRC regulations, a request was made to update the guidance pnmded in GL 88-18.
MA1618 LT %TDurton --/- /- LT: Ouidelines for Safety Classification of Structures, Systems, and Components in Nuclear Power Plants HQMB has 2 GCCA(s)
DRCH has a total of 4 GCCA(s)
Page 3 of 7 21-Jul-98
Open Generic Ccamunication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Comp Title Description Division of Reactor Program Management Emergency Preparedness and Radiation Protection Branch M9970s IN TAOreene -/-I- 9/30,9s T IN: Dese Calculatione fw an Array ofCask Dmcusses fuuling wish respect to nonconservatsve done calculations for an on a Pad arrey ofcasks on a pad PERB has 1 GCCA(s)
Events Assessment and GCs, and Special Inspection Branch MA2124 OL JW5hapaker -/-/- 12/30,98 T. OL: ADAMS Electroruc Subnuttal i Carabil*y PECB has 1 GCCA(s)
Generic Isenes and Envirvoneestal Projects Branch M99811 OL RABenedict -/-/- OL: lederirn Guidlines for Updating FSAR Reitersecs need for licensees to updmee theirs FSARs, in keeping with 10 CFR
$0.7I(e).
PGEB has 1 GCCA(s)
Non-Power Reactors and Decommissioning Project Directorate MA14M IN TAOreene -/-/- 11/30.98 T IN: Release of Radmactive Materials front To inform licensees orthe poternial probleens resuking from free release of ,
the Radiologically Controlled Area at radioactive masenais froen the radiologically controlled ores at nuclear power Nuclear Power Plants plants.
PDND has 1 GCCA(s)
Safegnards Branch MAe514 IN RABenedn:t -M- IN: Arumal Sununary of Fitness fw Duty Pnmdes a general summary and analysis of the data submitted by licensees Progreen Performance Reports fw Calender in their fitness for duty program performance reports fw cylender year 1996.
Year 1996 PSGB has 1 GCCA(s)
DRPM has a total of 5 GCCA(s)
Page 4 of 7 21-Jul-98
Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Comp Title Description Division of Systems Safety and Analysis Containment Systemis and Sewre Accident Branch M97146 GL JWShepsker 11/11/11 7/31/98 T GL: Degradation of ECC Recirculation Notifies adeeserce about the potential safety irnpact of foreign maserial in Followmg a LOCA due to Foreign Material sumps and suppression pools, which eld render safety-related equipmere in the Containment inoperable.
M98125 IN TJCarter 0394M2 7/19/98 L IN: BWR Contamment Bypam flow During A plant configuration during ruutine operaten could potentially result in
- Purgmg crunninmere bypam following an acndent M99813 OL TJCarter 09/25/98 9/l/98 T - OL: Operability R.quiremmts for Dual Valves required to be open for heat renovat purposes may also have a Function Valws containmere isointion function.
MA0349 GL JWShepaker 12/31/98 12/30!98 T GL: Request for Informaten Related to Based on staff actions that were undertaken to address the matter of ECCS Recirculation Capability containmere debris and potereial for cloggmg of suction stramers in BWR facilities,there is a need to similarly pursue this matter on PWRs.
MA9626 LT JWShepaker 06/12/00 12/15/98 T LT: PWR Sump Blockageissue To develop and implement a plan to evaluate the potereial for clogging of ECCS sump screem by IDCA debris in PWRs.
MAID 47 IN CDPetrone --/-/- 10/31/98 T IN: T3 Surveillance Test Requirements for Small valves used for testing, ventmg, and draining ofcontairunent isolation Contammere Isolation Valves valves are also belongingto antainmere isolation valves. These valves are not being property surveilled per T3.
SCSB has 6 GCCA(s)
Flant Systemis Branch M96912 LT WFBurton - /-/- 8/31/98 T LT: Petereial Genmc Concern with regard Farley- Failure ofnumerous pre acten synnklers in fire protection systems to I~are Protection Actuaten System providing fire protection servics to safety related systemm., _ J .
M96913 OL JWShepaker 04/30/99 10/30/98 T GL: Post-Fire Safe Shutdown Circuit To alert licensees to recent -,.g. _ and associated civil penalties Analyses regardmg licenace's lack ofdemonstrable protecten from a control room het short conditiert M97978 OL JWShepaker 08/30/98 900/98 T OL: Laborusory Testing of Ndear-Orade Informs a&hessees abcut NRC staff views on charecel testing pracuces and Actrvased Charcoal offers model technical specifications for voluntary adoptwn by the addressees in perparation for future testing obligations.
M99066 IN EJBenner -/-/- 6!30/98 L IN: Misunderstandag of the Ultimate liest Develop IN to inform licensees ofeeveral instances oferrors in hcenseei Sirdt Iacensmg Desis understandmg ofUltanate ifcat Sink liansmg basis.
Page 5 of 7 21-Jul-98
Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Comp Title Description M99283 GL JWShepaker 09G0S8 12/3038 T GL 96-06. Sup 2: Assurance of Equipment Notifies licensees 6out safety-significam issues that could affect contairunem Operability and Contamment Integrity etegnty and equipmed operabilrty during design-basis accident conditions.
during Design-Desis Accident Conditions M99331 IN EJDermer 083IS8 600/98 L IN: Postulated Imas of Feedwater as a Resuk Alerts addressees to concerns related to floodmg as a renuit of a norsiesign of a Pipe Dreak in the Circulating Water basis pipe break in the cirulating water system .
System MA1361 IN WFDurton 44- 600S8 L IN: nadequate Analysis of Reqd and Asso. Alerts licensees to potential problems asso. with post-fre safe S/D CKT Electncal CKTs Rsit. in the Potential Ims of analysis.These problems could result in faranduced CKT failures which Post-Fire S/D Capability could prevent the operation ce lead to maloperation ofequipment necessary to achieve post-fre safe S/D.
MA1617 IN %TBurton 120168 IN: Failure of Automatic Sprinkler Valves MA1940 IN TKoehy 100158 IN: Changes on OSilA Requirements in Fire Alerts licensees to recers rule change from OSilA regarding respiratory Brigade protectiert MA1941 IN TAOreene -M- IN: IF-300 Sperd Fuel Shippmg Cask Alerts licemees to a potential probierc conarmng the IF-300 spent furl 88Wmg cask-MA2294 IN CDPetrone 06/26/00 IN:Iloodmg of ECCS Rooms at WNP-2 An IN is being prepared to describe the flooding ofseveral ECCS rooms Caused by Fire 5% System Valve Rupture (RIIRC and IECS) at WNP2 which resulted when a fue main riser in a reactor buildmg stairwell ruptursd due to a water hammer SPLB ' is 11 GCCA(s)
Reactor Systems Branch M94565 LT DLSkeen 03G1/98 6G0/98 L Slow Saam Solenoid Pilot Valves Caused Scram solenoid pilm valves with viton diaphragms showing degraded scram by Viton Diaphragms times within 6-8 montin Currently tracking licermee response to RRO M97331 DL JWShepaker 12/3168 8/28/98 T DL: hudequate Procedural Guidance during Requests PWR licensees to take action to assure that there is adequate S/D and site Specific Vulnerabilities due to proculural guidance during shutdown operation and that gas accumulation Gas Accumulation vulnerabilities are identified, and actions are takers to limit or preclude adverse system performance.
M99332 GL JWShepaker 1200/98 10G098 T GL: Guidance on the Regulatory Provides a m.p.'S . of the current NRC staff guidance on regulatory Requirements for Criticality Analysis of requirements for criticality analysis ofnew and spers fuel storage at LWR Fuel Storage at LWR Power P!aras plants.
Page 6 of 7 21-Jul-98
Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Cosep LA Comp Titic Description M99eS9 IN EFOoodwm 12/31/98 8/28/98 T IN 97-15, Sup 1, Reportmg of Ermrs and Idorms licensees ofrequiremcens of 10 CFR 50.4 concermng repcsteg d Changes inIEIDCA Evin Models of Fuel all errors asanciated with cxule ofrecord, not just enors associated with large-Vendors and t'- - with 50.M(a)(3) break IDCA analynn.
MA0076 IN EFOoodwun 12/31/98 7/22/98 T IN: Westmghouse Fuel Gap Reapemng Alerts licensees about recent notificahon to the NRC by Weahnshouse related to W J -- . plants fuel rod irmemal pressure causing 10 CFR s 50.4(b)(2)- ' - concern (17 % local claddmg oxidahon)
MA0519 LT EFOoodwin -/-/- LT:lacensed1hennal Power Provides liunsees and staffwith explicit guidance en regulatory meerpretshonoflicensed core thermal power and guidance on enforcernent.
MA9779 IN WFBurton /-/- 8/21/98 T
- IN: Oas Bmdag of 51 Purnpa Informs licermees about gas acx:umulation in high head safety injection piping at Beaver vency.
MA1538 IN RABenedict - /-/- IN: Ice Condenser Operability Assurance Describes expenences at DC Cook plant regarding mamicnance and Problems surveillance ofice condenser MA1691 IN TAOreene -M- IN: Channel Bowmg in OEI I Fuel Assembly Informs licenseen of a recent proble n that resuhed in degraded corarol red performance due to fuel channel bow.
SRXB has 9 GCCA(s)
DSSA has a totalof 26 GCCA(s)
NOTES: There are a total of 55 GCCA(s)
-I-/ " for a "TR Comp" date mecans that at least one reviewer is unscheduled.
"11/11/11" for a "TR Cosep" date mecans that at least one reviewer is constant load scheduled.
4 Page 7 of 7 21-Jul-98
DIRECTOR's QUARTERLY STATUS REPORT July 1998 Generic Communication and Compliance Activities Added Since April 23,1998 TAC Type Contact Ixad Tech Branch TR Comp LA Comp Title Reason Added MA1617 IN WFDurton Plant Systems Branch 12SISS IN: Failure of Automahc Sprinkler Valves The EAP authorued 1 4-._.; ofIN st its 4/2898 meetus s MA1618 LT WFDurton Quality Aasm and 4-I- LT: Guidelines for Safety Classification of The EAP autherned kmg-term followup on this issue.
Maintenana Branch Structures. Systema, and Corngments in Nuclear Power Plares MA1623 LT JRTappert Electncal Ergpneerms -/-/- 6/15.98 L LT: Review of AEOD Study on The PECB management autherned kmg-term followup this Dranch Wesunghouse RFS Unavailability issue.
MA1689 GL JWShawker Materials and Chermcal IIlllllI I100/98 T GL: NRC Staff Review ofDWRVIP45 The EAP authorued 1. ;,,. .; ofGL at its 5/5.98 Engmeenng Branch meetag MA1691 IN TAGreene Reactor Speems Branch - /-/- IN: Channel Dowmg in gel 1 Fuel Assembly The EAP autherned i. 6,_; ofIN at its 5/5S8 meetirs MA1700 LT CDPetrone Mechanr:al Engmeenns 06/13/99 5/13.99 T LT: Air Operated Valves The PECD .., .-.; autherned LT follow-up this issue Branch 5/5,94.
MA1940 IN TKoshy Ptare Systems Branch 10/31/98 IN: Dianges on OSIIA Requirements in Fe' re The EAP nethorned i..:,,.-.; ofIN at its 5/2698 Brigade meetag MA194I IN TAGreene Plant Systems Branch -/-/- IN:IF-300 Spent Fuel Stuppmg Cask The EAP authorned dewlopmers ofIN at its 5/2698 meetirs MA2034 IN EJDenner Materials and Chenucal 07G168 9/15S8 T IN: Cradung ofSG Tube Ends The EAP authorned 1.n ,_; ofIN at its 6998 meetmg.
Engmeermg Branch MA2097 IN CDretrone Mechanical Engmeenng - /-/- IN: leaks in the Emergency Diesel The EAP authorized 1.J,.._.; ofIN at ia 6/1698 Branch Generator Imbe Oil Prpmg meeting.
MA2098 IN CDPetrone Mecharucal Engmeenng 1200/98 Sap. to IN 96-48, Motor-Operated Valve The EAP authorned developr:=nt ofIN at its 5/1698 Dranch Performance Issues meetng MA2099 IN TAGreene 1.m.m.-.; i- and - /-/- IN: Inadvertern Actuation of Stanchy liquid The EAP authorned i . 'm_.; ofIN at its 6/1698 Controls Dranch Control System meeting. ,
MA2124 GL JWShepaker Events Assessment and --/-/- 12/30/98 T GL: ADAMS Electroruc Submittal The TAC was approved by PECD .- e.,-.; 6/1198.
GCs, and Special Capability lasrection Branch Page 1 of 2 21-Jul-98
Generic Communication and Compliance Activities Added Since April 23,1998 TAC Type Contact Lead Tech Branch TR Comp LA Camp Title Reason Added MA2294 IN CD8etrone Plant Systems Branch 06/26/00 IN: Flooding of ECCS Rooms at WNP-2 The EAP authorued development of!N at its 600,98 Caused by Fire Waser Syneem Valve Rupture meeting.
MA2276 IN TAOreene Electncal Engmeenns -M- IN: EDO Connguration Coreol Ermrs The EAP authorued 1.J,,_./ orIN at its 7/1498 ,
Branch nm '
NOTES: Total Number of Reconis = 15
" /-4 " for a "TR Comp" date means that at least one reviewer is unscheduled.
J "11/11/11" for a "TR Comp" date means that at least one reviewer it constant load scheduled.
Page 2 of 2 21-Jul-98
DIRECTOR's MONTHLY STATUS REPORT July 1998 Generic Communication and Compliance Activities Closed i Since April 23,1998 TAC Type Contact lead Tech Branch TR Comep LA Coasp Title Reason Closed M92635 GL JWShepaker Reactor Symems Branch 05/29S8 C 5/2868 C GL: Ims of Reactor Coolam invasory and GL 98-02 inssed 5/2E98.
Associated Potesenal for less of Emergency Mitigation Fundaans W1 mile Shusdown M96614 LT TKoshy Mechamcal Engmeermg O&V368P 6/368 C LPSI Pump Mission Tiene "Ilnis LT item is ciceed h the Crystal River test Branch reeffirmed the confidence.
M97397 IN JRTappert Electrical Engmeenns 06/0498 P 6/498 C IN: Poesntial Deficiency of Eleanc Cable IN 98-21 issued 6/498.
Branch Connections M98751 GL JWShapaker Emergency Preparedness 07/1068 P 7/1068 C GL:ClariGcation ofNUREGCR 5055, The prW GL was canceled by PERB en 6/2698, and Ra<lin'iaa Procedian "Almosphenc Ddfusion for Control Roose Branch IIalutabilny Assessment
- MA8138 GL JWShepaker Instrumerdation and 05/12S8 P 5/12/98 C GL: Year 2000 Readmess ofComputer GL 98-01 issued 5/1168.
Centrols Branch Sysseen at Nuclear Power Plants MA0376 IN TKoshy Events Assessmera and 07/13/98 P 7/13/98 C IN: Issues Idemified During Recent NRC IN 98-22 issued 6/1768.
GCs, and Special DesigriInspections Impaction Branch MA0380 IN CViiodge Plant Symems Branch 06/26S8 P 6/2698 C IN: Stem Bandmgin Turbine Governor IN 98-24 issued 6/2698.
- Valves in RCIC and Auxiliary Feedweser Systems MA0608 IN TKoshy Emergency Nr.,L 0693,98 P 6/368 C IN: Problems with Emergency Preparabiess IN 98-20 issued 6/3/98.
, and Radiation Protectson Reaperatory Protectiion Propams Branch MA1339 IN RABenedict Safeguards Branch 05/0768 P 5/768 C IN: ANSIR Pro 8 ram IN 98-17 issued Sn/98.
MA1539 IN DLSkeen Matenals and Cherrucal 05/07/98 P 5n,98 C IN 9743, Sup 1, Status ofNRC Staff IN 98-63, Sup 1,inaued 5/7/98.
Engmeenng Branch Review of BWRVIP-05 MA1616 IN JRTappert Materials and Chenucal 06/23/98 C 6/23/98 C IN: Crosby Relief Valve Seepost DriR IN 98-23 issued 6/2368.
Engmeenrig Branch Problems MA1688 IN DI.Skeen Quality Assurance and 06/03S8 P 6/3/98 C IN: Spring Return Bindmgin GEType IN 98-19 issued 6/3/98.
Massenance Branch SBM Control Switches Page 1 of 2 21-Jul-98 -
Generic Communication and Compliance Activities Closed Since April 23,1998 TAC Type Contact lead Tech Branch TR Comp LA Comp Title Reason Closed MA1690 IN TMKhan Plant Syncms Branch 07/08.98 P 7/8/98 C IN: Ims ofInventory San Safety-Related. IN 98-25 innued 7/1/98.
Closed-tmp Cooling Water Systems NOTES: Total Number of Records = 13
" I-/ " for a "TR Comp" date means that at least one miewer is unscheduled.
"11/11/11" for a "TR Comp" date means that at least one miewer is constaat load sch Page 2 of 2 21-Jul-98
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