ML20216F016
| ML20216F016 | |
| Person / Time | |
|---|---|
| Issue date: | 07/31/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| References | |
| NUDOCS 9909210016 | |
| Download: ML20216F016 (64) | |
Text
DIRECTOR'S STATUS REPORT on GENERIC ACTIVITIES Action Plans Generic Communication and Compliance Activities
/
C, JULY 1999
,D P
Office of Nuclear Reactor Regulation o-lH'Y gpag16 =as an o - c
r.
l lNTRODUCTION The purpose of this report is to provide information about generic activities, including generic communications, under the cognizance of the Office of Nuclear Reactor Regulation. This report, which focuses on compliance activities, complements NUREG-0933, "A Prioritization of Generic Safety issues."
This report includes two attachments: 1) action plans and 2) generic communications under development and other generic compliance activities. Generic communications and compliance activities (GCCAs) are potential generic issues that are safety significant, require technical resolution, and possibly require generic communication or action., "NRR Action Plans," includes generic or potentially generic issues of sufficient complexity J
or scope that require substantial NRC staff resources. The issues covered by action plans include
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concems identified through review of operating experience (e.g., Boiling Water Reactor Internals Cracking and Wolf Creek Draindown event), and issues related to regulatory flexibility and improvements (e g., New Source Term and Probabilistic Risk Assessment (PRA) Implementation Plan). For each action plan, the report includes a description of the issue, key milestones, discussion of its regulatory significance, current status, and names of cognizant staff., " Generic Communications and Compliance Activities," consists of three status reports:
- 1) Open GCCAs,2) GCCAs added since the previous report, and 3) GCCAs closed since the previous report. The generic communications listed in the attachment includes bulletins, generic letters, and information notices. Compliance activities listed in the attachment do not rise to the level of complexity that require an action plan, and a generic communication is not currently scheduled. For each GCCA, there is a short description of the issue, scheduled completion date, and name of cognizant staff.
ATTACHMENT 1 NRR ACTION PLANS a
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TABLE OF CONTENTS BOILING WATER REACTOR INTERNALS................................
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PRAIMPLEMENTATION ACTION PLAN 1.2(c)Inserviceinspection Action Plan..............................................................6 STEAM G EN ER ATO RS..............................................
12 MEDIUM-VOLTAGE & LOW-VOLTAGE BREAKER FAILURES..............
15 ENVIRONMENTAL SRP REVISION ACTION PLAN........................
17 EMERGENCY ACTION LEVEL GUIDANCE DEVELOPMENT................
19 PRA IMPLEMENTATION PLAN 1.2(d) Graded Quality Assurance Action Plan.
22 ACCIDENT MANAGEMENT IMPLEMENTATION..........................
30 HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN (Fin al U pdate).....................................................
34 WOLF CREEK DRAINDOWN EVENT: ACTION PLAN (Final Update).........
37 ECCS SUCTION BLOCKAGE (Initial Update)............................
39 NEW SOURCE TERM FOR OPERATING REACTORS.....................
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s BOILING WATER REACTOR INTERNALS TAC Nos. M94975, M95369, M96219, M96539, M97802, M97803, Last Update: 7/14/99 M97815, M98266, M98880, M99638, M99870, M99894, M99897, Lead NRR Division: DE M99895, MA1102, MA1104, MA1138, MA1926, MA1927, MA2326, Supporting Division: DSSA MA2328, MA3395, MA4203, MA4464, MA4465, MA4467, GSI: Not Available MA4468, MA5057, MA5058, MA6015 MILESTONES DATE (T/C)
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i PART1: REVIEW OF GENERIC INSPECTION AND EVALUATION I
CRITERIA l
- 1. Issue summary NUREG 1544 03/96C '
o Update NUREG-1544 3Q/FY99T 1
- 2. Review BWRVIP Re-inspection and Evaluation Criteria j
o Reactor Pressure Vessel and Intemals Examination Guidelines j
(BW RVIP-03)................................................
06/08/98 CI' o BWRVIP-03, Section 6A, Standards for Visual Inspection of Core Spray Piping. Spargers, and Associated Components....................
06/08/98 Cl o BWR Vessel Shell Weld inspection Recommendations (BWRVIP-05)..
. o BWR Axial Shell Weld inspection Recommendations...............
07/28/98 CA o Guidelines for Reinspection of BWR Core Shrouds (BWRVIP-07).....
12/31/99 T 04/27/98 CA
- 3. Review of generic repair technology, criteria, and guidance TBD 4.z Review generic mitigation guidelines and criteria TBD
- 5. Review of generic NDE technologies developed for examinations of BWR intemal components and attachments TBD 6.: Other Intemals reviews (safety assessments, evaluations, mitigation
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measures, inspections, and repairs) o Safety Assessment of BWR Reactor internals (BWRVIP-06).........
09/15/98 CA o Bounding Assessment of BWR/2-6 Reactor Pressure Vesselintegrity issues (BWRVIP-08 & BWRVIP-46).............................
03/27/98 CA o Evaluation of Crack Growth in BWR Stainless Steel RPV Intemals (BW RVI P-14)...............................................
06/08/98 Cl o intemal Core Spray Piping and Sparger Replacement Design Criteria
- (BW RVIP-16)............,.................................
11/16/98 Cl 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 Intemals inspection and Flaw Evaluation Guidelines (BWRVI P-18).... 4.........................................
06/08/98 Cl o BWRVIP 18, Appendix C, BWR Core Spray Internals Demonstration of Compliance With Technical Information Requirements of License Renewal Rule (10 CFR 54.21)..................................
TBD I
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DATE (T/C) o intamal Core Spray Piping and Sparger Repair Design Criteria (BW RVIP-19)...............................................
11/16/98 Cl o Core Plate Inspection and Flaw Evaluation Guideline (BWRVIP-25)...
04/28/99 Cl o Top Guide Inspection and Flaw Evaluation Guideline (BWRVIP-26)...
05/18/99 Cl o Standby Liquid Control System / Core Plate AP inspection and Flaw Evaluation Guidelines (BWRVIP-27)............................
04/27/99 CF o Assessment of BWR Jet Pump Riser Elbow to Thermal Sleeve Weld Cracking (BWRVIP-28).......................................
09/30/99 T o Technical Basis for Part Circumferential Weld Overlay Repair of Vessel intemal Core Spray Piping (BWRVIP-34).........................
TBD o Shroud Support inspection and Flaw Evaluation Guidelines (BW RVI P-38)...............................................
07/30/99 T o BWR Jet Pump Assembly inspection and Flaw Evaluation Guidelines (BW RVI P-41 )................................................
10/30/99 T o BWR LPCI Coupling Inspection and Flaw Evaluation Guidelines (BW RVI P-42)...............................................
06/14/99 Cl o Update of Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity issues (BWRVIP-46)..................................
03/27/98 CA o BWR Lower Plenum inspection and Flaw Evaluation Guidelines (BW RVI P-47)...............................................
04/30/99T o VesselID Attachment Weld Inspection and Flaw Evaluation Guidelines (BW RVI P-48)...............................................
03/21/99C o ' instrument Penetration Inspection and Flaw Evaluation Guidelines (BW RVIP-4 9)..............................................
08/04/98 CA o Top Guide / Core Plate Repair Design Criteria (BWRVIP-50).........
12/30/99 T o Jet Pump Repair Design Criteria (BWRVIP-51)....................
12/30/99 T o Shroud Support and Vessel Repair Desi n Criteria (BWRVIP-52).....
12/30/99 T 9
o Standby Liquid Control Line Repair Design Criteria (BWRVIP-53).....
12/30/99 T o Lower Plenum Repair Design Criteria (BWRVIP-55)................
12/30/99 T o LPCI Coupling Repair Design Criteria (BWRVIP-56)................
12/30/99 T
~ o instrument Penetrations Repair Design Criteria (BWRVIP-57)........
12/30/99 T o CRD intemal Access Weld Repair (BWRVIP-58)..................
12/30/99 T o Evaluation of Crack Growth in BWR Nickel-Base Austenitic Alloys in RPV Intemals (BWRVIP-59)..'.................................
12/30/99 T o BWR Vessel and intemals Induction Heating Stress improvement.
Effectiveness on Crack Growth in Operating Plants (BWRVIP-60).....
07/08/99CF o Technical Basis for inspection Relief for BWR Intemal Components with Hydrogen injection (BWRVIP-62).......................'........
12/30/99 T o Shroud Vertical Weld inspection and Evaluation Guidelines (BW RVI P-63)........... ~....................................
12/30/99 T
'Cl= complete interim (i.e., draft SER)
Descriotion: Many components inside boiling water reactor (BWR) vessels (i.e., internals) are made of materials such as stainless steel and various alloys that are susceptible to corrosion and cracking. This degradation can be accelerated by stresses from temperature and pressure changes, chemical interactions, irradiation, and other corrosive environments. This action plan is intended to encompass the evaluation and resolution of issues associated with intergranular stress corrosion cracking (IGSCC) in BWR intemals. This includes plant specific reviews and the assessment of the generic criteria that have been proposed by the BWR Owners Group and the BWRVIP technical subcommittees to address lGSCC in core shrouds and other BWR intemals.
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Historical Backaround: Significant cracking of the core shroud was first observed at Brunswick, Unit 1 nuclear power plant in September 1993. The NRC notified licensees of Brunswick's discovery of significant circumferential cracking of the core shroud welds. In 1994, core shroud cracking continued to be the most significant of reported internals cracking. In July 1994, the NRC issued Generic Letter (GL) 94-03 which requires licensees to inspect their shrouds and provide an analysis justifying continued operation until inspections can be completed.
A special industry review group (Bolling Water Reactor Vessels and Internals Project - BWRVIP) was formed to focus on resolution of reactor vessel and internals degradation. This group was instrumental in facilitating licensee responses to NRC's GL 94-03. The NRC evaluated the review group's reports, submitted in 1994 and early 1995, and all plant specific responses.
All of the plants evaluated were able to demonstrate continued safe operation until inspection or repair on the basis 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 or inspections to their core shrouds.
In late 1994, extensive cracking was discovered in the top guide and core plate rings of a foreign reactor. The design is similar to General Electric (GE) reactors in the U.S., however, there have been no observations of such cracking in U.S. plants. GE concluded that it was reasonable to expect that the ring 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 assessment was acceptable and that top guide ring and core plate ring cracking is not a short term safety issue.
J Proposed Actions: The staff will continue to assess the scopes that have yet to be submitted by J
licensees conceming inspections or re-inspections of their core shrouds. The staff will also continue to i
assess core shroud reinspection results and any appropriate core shroud repair designs on a case-by-j case basis. The staff willissue separate safety evaluations regarding the acceptability of core shroud reinspection results and core shroud repair designs. The staff has been interacting with the BWRVlP and individuel 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 interacting with the industry on a generic basis in order to encourage them to continue their proactive efforts to resolve IGSCC of BWR intemals as a voluntary industry initiative. The BWRVIP has submitted 53 generic documents, supporting plant-specific submittals, for staff review. The staff is ensuring that the generic reviews are incorporating recent operating experience on all BWR internals.
Oriainatina 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 4
can be completed.
Reaulatorv 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 untilinspections could be performed. The staff has concluded in all cases that licensees have provided sufficient evidence to support continued operation of their BWR units to the refueling outages in which shroud inspections or J
repairs have been scheduled. in addition, in October 1995, industry's special review group submitted a safety assessment of postulated cracking in all BWR reactor intemals and attachments to assure continuing safe operation.
Current Status: Almost all BWRs completed inspections or repairs of core shrouds during refueling outages in the fall of 1995. Various repair methods have been used to provide alternate load carrying capability, 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 3
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submitted by BWR licensees. ' Review by NRC continues on individual plant reinspection results and plant-specific assessments.
In October 1995, industry's special review group issued a report (BWRVIP-06) which the NRC staff's preliminary review indicated was not comprehensive. The NRC staff requested additionalinformation which the BWRVIP provided in letters dated December 20,1996, and June 16,1997. The staff has completed its review of this submittal. The industry group submitted a report on reinspection of repaired and non-repaired core shrouds (BWRVIP-07) in February 1996. The staff completed its review and issued an SER with several open items. The staff met with the industry to resolve these open items, and completed its final SER. The NRC is also reviewing information submitted by GE on the safety significanoe of and recommended inspections for top guide and core plate ring cracking. Technical review of the " Reactor Pressure Vessel and intemals Examination Guidelines (BWRVIP-03)' is complete and the staff's 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 inspection 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 welds. Information Notice (lN) 97-63, " Status of NRC Staff's Review of BWRVIP-05," was issued August 7,1997, to infomi the industry of both the status of the staff's review and that the staff would consider technically-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 staff's independent assessment of the BWRVIP-05 report was transmitted by letter dated August 14, 1997, to the BWRVIP, along with a request for additional information and information that needed to be addressed forlicensees requesting 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 evaluation of the BWRVIP-05 report. IN 97-63, Supplement 1, was issued May 7,1998, to inform the industry that the staff would continue to consider technically-Justified scheduler relief requests from performing augmented inspections of the RPV shell circumferential welds for 40-months or two operating cycles, whichever was longer. A proposed GL informing the industry of the staff's SE was published August 7,1998 (63 FR 42460). No public comments were received, and the staff issued the final GL (GL 98-05) November 10,1998.
The staff's initial review of the BWRVIP-03,14, -18, -25, -26, -47, and -48 is complete and the staff's SEs have been issued. The BWRVIP has responded to the open items in the staff's SEs, and the staff is presently completing its final review of the these reports. The staff's initial review of BWRVIP-16,19, and -42 is completo. The staff is awaiting the BWRVIP's response to the open items in the staff's SEs in order to complete the final review.
The BWRVIP has submitted Appendices to the inspection and Flaw Evaluation Guidelines. These appendices address the use of BWRVIP generic inspection guidelines for compliance with requirements of the license renewal rule (10 CFR Part 54). The staff is reviewing these appendices in conjunction with its review of the BWRVIP guidelines, and has issued the first of thirteen license renewal SEs on BWR internals, with the remaining to be completed by December 1999.
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 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 Welds 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 staff that it plans to revise BWRVIP-07 4
to ensure that the vertical core shroud welds, and the core shroud repair, is adequately inspected. This report was provided for staff review as BWf3 VIP-63, dated July 1,1999.
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 their 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 applicable BWRVIP products. The staff is requesting that the BWRVIP have each BWR licensee confirm to the NRC staff that each BWR licensee is aware of the NRC staff's understanding of these commitments, and recognizes their individual commitment to the BWRVIP program and reports.
NRR Technical Contacts:
Robert Hermann, EMCB,415-2768 Amy Cubbage, SRXB,415-2875 Jai Rajan, EMEB,415-2788 NRR Lead PM:
C. E. Carpenter, EMCB,415-2169
References:
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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PRA IMPLEMENTATION ACTION PLAN 1.2 (c) r Inservice Inspection Action Plan TAC Nos. M95125, M971.53 Last Update: 6/30/99 M99389, M99756, MA012b Lead NRR Division: DE MA0867, MA0868 Support Division: DSSA RG/SRP MILESTONES DATE (T/C) 1.
Draft for RI-ISI team review / comments 04/05/96C 2.
First draft for Branch Chiefs review / comments 08/14/96C 3.
Revised draft for Branch Chiefs review / comments 01/24/97C 4.
Revised draft for Branch Chiefs review / comments 04/08/97C 5.
Draft for Division Director review / comments 04/29/97C 6.
Draft for Office Director /OGC review / comments 05/16/97C 7.
Office Director /OGC concurrence 07/08/97,C 8.
Draft for CRGR review / comments 07/08/97C 9.
Draft for ACRS review / comments 06/03/97C 10- Initial presentation to ACRS full Committee 06/11/97C
- 11. Initial presentation to CRGR 06/11/97C
- 12. Meeting with ACRS Subcommittee 07/08/97C
- 13. Meeting with ACRS full Committee 07/09/97C
- 14. Meeting with CRGR 07/17/97C
- 15. SECY from EDO to Commissioners (SECY 97-190) 08/20/97C
- 16. Publish draft for public comments 10/15/97C
- 17. Public comment period for draft RG/SRP ends 01/13/98C
- 18. Public Workshop 11/20/97C
- 19. Complete draft for ACRS/CRGR review / comments 04/98C
- 20. Complete draft for Inter-Office concurrence 05/98C
- 21. Issue RG/SRP for trial use by the staff 06/98C WOG TOPICAL REPORT MILESTONES DATE (T/C) 1.
Technical Meeting 9/22/98C 2.
WOG Commitment Letter 9/30/98C 3.
Issue FSER 12/15/98C t
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EPRI TOPICAL REPORT MILESTONES DATE (T/C) 1.
Issue RAls to EPRI 6/12/97C 2.
EPRI Response to RAls 11/13/98C 3.
Open items Technical Meeting 3/2/99C 4.
Receive Revised Report from EPRI 4/15/99C 5.
Issue FSER 10/31/99T PILOT PLANT REVIEW MILESTONES DATE (T/C) 1.
Issue FSER Vermont Yankee 11/9/98C*
2.
Issue FSER Surry 2/16/98C j
3.
Issue FSER ANO-2 12/29/98C
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- Clarification of some aspects of the program are stillin progress.
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INSPECTION PROCEDURES MILESTONES DATE (T/C) 1.
Issue Draft inspection Procedure Number 73753 6/98C 2.
Issue Final inspection Procedure Number 73753 6/98C
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Description:==
Develop risk informed inservice inspection (RI-ISI) application-specific Regulatory Guide (RG), corresponding Standard Review Plan (SRP) sections and related inspection 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 i
acceptability of the pilot programs with respect to their RI-ISI applications and prepare plant specific j
safety evaluation reports (SER). The action plan describes the process for the review of RI-ISI i
submittals subsequent to the approval of the pilots by referencing the topical reports, the addition of a description for the future reviews and approvals of the ASME Code Cases. This action plan will be monitored up to and including the completion of RI-ISI RG and SRP, pilot plant reviews, topical report reviews, and development of inspection procedures. All of these items except the review of the EPRI Topical Report and clarification of some aspects of the Vermont Yankee program are completed.
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Historical Bacl<around: 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 activities. In the statement, the Commission stated its belief that tue use of PRA technology in NRC regulatory activities should be increased to the extent supported by the state-of-the-art in PRA methods and data and 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 RGs and SRPs associated with the use of PRA in specific applications. A Nuclear Reactor Regulation / Nuclear Regulatory Research (NRR/RES) joint task group was 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).
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L ASME is ~ working on three Code Cases for altamate 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 revised to encompass both methodologies.
Proposed Actions The NRC has encouraged licensees to submit pilot plant applications organized under one umbrella sponsoring 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 ISl. The NRC is reviewing the industry submittals with focus on the licensees characterizing the proposed change including the identification of the particular piping systems and welds 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 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 documentation of the analyses and the request for NRC review and approval.
Additionally, using the results from the review of the above-mentioned pilot plant applications, from the PRA insights obtained from the risk-ranking of piping elements, and in cooperation with the RES staff, a parallel effort was carried out to develop: (a) an RI-ISI application-specific RG and (b) the corresponding SRP chapters and associated 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 1
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 has been documented in SERs for each of the pilot plant licensees and issued to the pilot plant licensees to allow use of the RI-ISI methodology.
ASME is working on three Code Cases for altemate examination requirements to ASME Section XI, Division 1 for piping welds. Code Case N-560, for the attemate 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 Selection Rules for Class 1,2, and 3 piping welds, is based on the WOG methodology.' Code Case N-578, for the attemate 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 addition, the
. Code Case is an abbreviated version and does not have all the details presented in the WOG topical report (Reference 4,6). The staff reviews the WOG methodology as well as the Code Case N-577 and the consistency of the Surry 1 pilot program for Rl-ISI to both of these. The Code Case N-577 will be reviewed and, if found acceptable, will be endorsed by RG 1.147.with any necessary additions or deletions.
The major difference between Code Case N-578 and the EPRI me' hodology is that the scope of the i
t Code Case is limited to ASME Class 1,2, and 3 systems while the EPRI methodology may encompass 8
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all 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.
Code Case N-560 for the attemate examina$ ion requirements for Class 1, Category B-J piping welds is being revised to encompass both WOG and EPRI methodologies. This Code Case has limited applicability in that it is applicable only to ASME Class 1 piping systems. The staff reviews the EPRI
, methodology as well as Code Case N 560 and the consistency of the Vermont Yankee RI-ISI pilot program to both of these.
q The staff utilized the acceptable altemative provision of 10 CFR 50.55a (a)(3)(/) to approve the pilot plants' applications. The staff is working closely with ASME to expedite changes involving ISI.
Lona-Ranae Plan -
This action plan will be monitored up to and including the completion of RI-ISI RG and SRP, pilot plant reviews, topical report reviews, and development of inspection procedures. All of these items except the review of the EPRI Topical Report and clarification of some aspects of the Vermont Yankee program are completed.
For the RI-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
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seek relief from NRC to the current 50.55a requirements. A minimal review cycle is expected for the
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. approval of RI-ISI submittals during this time frame.
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lt is anticipated that subsequent to the issuance of safety evaluation reports (SER) for the pilot plants i
and the topical reports, the industry will revise the ASME Code Cases to incorporate lessons leamed from pilot plants and topical report reviews. The ASME Code Cases will be endorsed by RG 1.147 with exceptions and/or additions, if necessary, consistent 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 Al-ISI methodology has been endorsed in RG 1.147 (and endorsed in
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l 10 CFR 50.55a, as necessary), the staff anticipates that the licensee will develop an RI-ISI program using the approved ASME Code Case. No NRC approval will be required, and the staff will oversee the acceptable implementation as part of the normal ISI inspection program.
For the non-pilot plant licensees that intend to implement RI-ISI starting with their next ten year interval, the staff will consider granting a relief from the current deterministic requirements of ISI of piping, of up
. to two years. These licensee 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, i
During the two-year extension period, the licensees would continue to implement their current ISI program. In order to disseminate the information to the licensees, the staff issued information Notice 98-44.
Onpinating 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 I
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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 utilizing PRA in reactor related activities including ISI. This action plan is in conformance with the agency-wide implementation plan for PRA and any future changes will be consistent with the overall plan.
k Reculatory Assessment: The operational readiness and functional integrity of certain safety-related 1
piping and associated structural elements (e.g., pressure retaining welds) are vital to the safe operation of nuclear power plants. ISI is 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 endorsed through the rulemaking process. The current ASME Code ISI requirements and practices have only an implicit consideration of risk-informed information, such as failure probability and consequence of failure.
4 Licensees are currently interested in optimizing inspection 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 parassl path, ASME is developing Code Cases for alternate examination requirements to the current ASME Section XI selection and inspection requirements. These Code Cases utilize procedures that are based on the relative risk significance of piping locctions within individual systems.
The NRC is using probabilistic methods, as an adjunct to deterministic techniques, to help define the scope, type, and frequency of ISt. The development of RI-ISI programs has the potential to optimize the use of NRC and industry resources and continue to assure adequate protection of public health and safety.
Acceptability of the RI-ISI pilot programs is documented in safety evaluations. To provide the permanent approach to RI-ISI, the staff intends to utilize the experience gained through the pilot applications in the proposed rulemaking process to modify 10 CFR 50.55a to explicitly endorse RI-ISI methodology.
Current Status: The staff completed final drafts for trial use of risk-informed inservice inspection (RI-ISI) of piping regulatory guide (RG) (RG-1.178) and standard review plan (SRP) Section 3.9.8 which were submitted to the Commissioners (SECY 96-139) for information. The RG and SRP were issued in the Federal Register in October 1998.
The staff completed its review of the Westinghouse Owners Group (WOG) methodology documented in WCAP-14572, Rev.1, and issued its safety evaluation report (SER) on December 15,1998. The staff completed its review of the Vermont Yankee (RI-ISI) pilot program and issued its safety evaluation report (SER) ori November 9,1998. The staff completed its review of the Surry Unit 1 (RI-ISI) pilot program and issued its safety evaluation report (SER) on December 16,1998. The staff completed its review of the ANO-2 (RI-ISI) pilot program and issued its safety evaluation report (SER) on December 29,1998. It should be noted that subsequent to issuance of the SER on the Vermont Yankee RI-ISI program, some issues arose regarding clarification of how augmented inspection programs for stress corrosion cracking are treated in the program. The staff is pursuing this clarification with the licensee.
On March 2 and 3,1999, the staff met with EPRI to discuss EPRrs responses to NRC's Request for Additional information (RAl) related to the approach described in EPRI topical report, TR-106706, Risk-Informed inservice Inspection Evaluation Procedure. Based on the discussion, EPRI revised the topical report and submitted to the staff on April 15,1999, to incorporate lessons learned from the pilot applications (Vermont Yankee and ANO-2) of the methodology; methodology enhancements which have 10
1 evolved since the June 1996 report was issued; and to provide further clarification and guidance as necessary, based on NRC RAls. The staff has completed the draft SER to support ACRS meetings, and plans to ccmplete the final SER by October 31,1999.
The NRC issued Information Notice 98-44 to inform addressees that for licensees that intend to implement a RI-ISI program for piping and do not have a pilot plant application currently under staff review, the staff will consider authorizing a delay of up to two years in the implementation of the next ten-year ISI program for piping only in order for the licensee to develop and obtain approval for the RI-ISI program for piping.
The staff has also been actively participating in ASME Code activities related to RI-ISI.
NRR Contacts:
,3. Ali,415-2776 S. Dinsmore,415-8482
References:
1.
Federal Register, Vol. 60, No.158, "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement," August 16,1995.
2.
Memorandum from Shiriey Ann Jackson, Chairman, NRC to James M. Taylor, Executive Director for Operations, " Follow-up Requests in Probabilistic Risk Assessment and Digital Instrumentation 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 Utilization of Probabilistic Risk Assessment and Digital Instrumentation and Control Technology," January 3, 1996.
4.
WCAP 14572, " Westinghouse Owners Group Application of Risk-Based Methods to Piping inservice inspection Topical Report," March 1996.
5.
EPRI TR 106706," Risk-Informed Inservice inspection Evaluation Procedure," June 1996.
6.
WCAP-14572, Revision 1," Westinghouse Owners Group Application of Risk-Informed Methods to Piping inservice inspection Topical Report," October 1997.
11 w
' STEAM GENERATORS TAC Nos. M88885, M99432, MA4265 Last Update: 6/30/99 Lead Civision: DE (#394)
MILESTONE DATE (T/C) 1.
Commission /EDO Approval 02/94(C) 2.
Receive NEl Document 02/96(C) 3.
Review NEl Document Revisions Continuous Process 4.
Regulatory Analysis 5/97(C) 5.
Proposed GL Pkg 10/97(C) 6.
ACRS Endorsement 9/97(C) 7.
CRGR Concurrence On hold 08 8.
EDO On hold 9.
Publish Proposed GL On hold Orig. Publish Proposed Rule 03/95(C)
- 10. Public Comment (120 day comment period)
On hold
- 11. Revise GL Pkg On hold
- 12. ACRS Comments On hold
- 13. CRGR Concurrence On hold
- 14. EDO Concurrence On hold
- 15. Commission Approval On hold
- 16. Publish Final GL On hold Orig. Publish Final Rule 12/95 Brief Descriotion: The NRC originally planned to develop a rule pertaining to steam generator tube integrity. The proposed rule was to implement a more flexible regulatory framework for steam i
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. In SECY 98-248, the staff "This revision reflects the status of the proposed GL as "on hold". The staff estimates that by September 1999, dependent on the timing of industry submittals, that the staff should be able to reach a determination of whether the proposed Gl. should be reactivated. At that time (if the GL is reactivated),
the staff will provide revised schedular estimates.
12 J
l recommended to the Commission that the proposed GL be put on hold for 3 months while the staff works with NEl on their NEl 97-06 initiative. In the staff requirements memorandum dated December 21,1998, the Commission did not object to the staff's recommendation. If sufficient progress is made with NEl in resolving technical and regulatory implementation issues, then the GL effort may be permanently halted.
Reaulatory Assessment: The current regulatory framework provides reasonable assurance that operating PWRs are safe. The current regulatory framework that implements governing requirements through the plant technical specifications can be improved. The staff is currently working with NEl to find a performance-based, risk-informed solution to the current problems, that utilizes industry guidance wherever possible.
Current Status:
- Briefed ACRS on ANPRM - August 1994
- SG rule ANPRM - September 1994
- SECY-95-131 - May 1995 - justifies continuation of rulemaking
- Briefed Commission on SG rule -- June 1995
- Briefed Commission on SG rule status -- February 1996
- Memo to Commission re. revised schedule - May 1996 Briefed Chairman on status - July 1996
- Information Brief for CRGR - October 1996
- ACRS Brief on SG rule -- November 5,6,1996
- Briefed Chairman on SG rule status -- December 1996
- Briefed ACRS re. risk-informed approach for SG rule - January 1997
- Briefed ACRS re. risk assessment and regulatory analysis results -- March 4,5, and
. April 3,1997
- COMSECY-97-013 suggests revising 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. GL/DG-1074/DPO issues - August 26,27, September 3,1997
- Information Brief for CRGR re. GL and backfit -- September 9,1997
- Met w/NEl re. GL/DG 1074/TSs - Spptember 11,1997
- ACRS endorsement to issue GL and DG 1074 for public comment -- September 15,1997
- Briefed ACRS re. DPO issues - October 2,1997
- 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 hformation briefing on package
- Met with CRGR on July 21,1998, for oatailed review of proposed GL package
- Memo from Collins to Callan dated September 11,1998, suggests putting proposed GL on hold for 3 months to work with NEl on NEl 97-06
- Staff issued Commission paper SECY-98-248 (October 28,1998) recommending a 3 month hold on issuance of proposed GL. SECY 98-248 also recommended issuance of (1) DG-1074, (2) the DPO consideration document, and (3) the September 1998 Hopenfeld memorandum to the Commission, for public comment
- The Commission, in SRM dated December 21,1998, agreed to above recommendations Held technical and management meetings with industry on 10/7/98,10/28-29/98,11/12/98,11/18/98, 2/10/99, and 2/24/99 to resolve technical and regulatory implementation issues regarding NEl 97-06.
- Draft regulatory guide DG-1074 was issued for public comment (appeared in federal register on 1/20/99) with the DPO consideration document, and the Hopenfeld memorandum to the Commission 13
- Intemal guidance to SG inspectors was issued on 1/25/99 indicating that DG-1074 should not be used for inspection guidance as directed by the Commission's SRM of 12/21/98.
- Briefed ACRS Materials S/C on 3/24/99 regarding the status of current regulatory approach.
- May 6,1999 technical meeting with NEl/ industry: Discussed industry proposed TS & Technical Requirements Manual (TRM).
- May 18,1999 senior management meeting with NEl/ Industry: Discussed status
- June 24,1999 technical meeting with NEl/ industry: Discussed open technical issues; NEl presented a revised TS & TRM.
NRR Technical Contacts:
Ted Sullivan, EMCB,415-3266 Kristine Thomas, EMCB, 415-1362 RES
Contact:
N/A i
)
14
l 1
MEDIUM-VOLTAGE & LOW-VOLTAGE BREAKER FAILURES Last Update: 6/30/99 Lead Division: DE MILESTONES DATE (T/C) 1.
Develop and issue a Tl and conduct inspection
- a. Issue Temporary Instruction to inspect two sites for Regions I,11, Ill, and IV (EElB) 12/31/97C
- b. Issue revision to Temporary Instruction (EElB) 03/09/98C
- c. Conduct Tl inspections (REGIONS, EElB, lOMB) 09/30/98C
- 1. Region I (8/3 - 8/7 & 9/21 - 9/25) ii. Region ll (3/16 - 3/20 & 5/4 - 5/8) j iii. Region ill (5/18 - 5/22 & 6/8 - 6/12) iv, Region IV (7/6 - 7/10 & 8/31 - 9/4)
- d. Issue finalinspection report 10/16/98C 2.
Prepare white paper describing status of staff and industry actions to address circuit breaker problems (NRR/(EElB, IOMB, REXB) and RES 03/25/98C 3.
Develop and issue an information notice describing recent (Pre Tl Inspections) findings regarding maintenance practices, review of ins and industry
)
experience (EElB, REXB, IOMB) 08/28/98C 4.
Inspect at least three major circuit breaker vendors' facilities and at least three 09/30/99 of the third-party repair / refurbishment outfits (lOMB, EElB)
- a. GE facilities
- b. Westinghouse facilities 1/22/99C
- c. ABB facilitics
- d. PDS ERAM 2/17/99C
- e. NLI 6/10/98C
- f. Wyle Labs 5.
Monitor circuit breaker failures via review of LERs and identify potential safety problems as a result of breaker failures (RES) and review 10 CFR Part 21 submittals (NRR/REXB)
Ongoing 6.
Participate in Industry meetings (EElB, lOMB, REXB)
- a. GE Users Group meetings 2/15-2/19/99C
- b. Westinghouse Users Group meetings 8/23-8/27/99
- c. ABB Users Group meeting 6/14-6/18/99C 7.
- a. Summarize inspection findings and evaluate if any additional regulatory actions is required (EElB, REXB, IQMB) 4/22/99 I
- b. Complete the required action (EElB, REXB, lOMB) 4/29/99C
==
Description:==
The action plan is intended to address medium-voltage and low-voltage power circuit breaker reliability issues.
Historical Backaround: Over the past several years, the NRC evaluated a number of events at nuclear power plants that involved the failure of circuit breakers. The major causes of breaker failures were inadequate lubrication, improper repair and refurbishment, and lack of adequate maintenance l
instructions, procedures, and drawings. Also, the use ofinadequately manufactured and dedicated parts appears to be responsible for some recent breaker failures. Over the years the NRC issued information 15 J
notices (ins) describing the breaker failures and expected that licensees would review their maintenance programs and correct the deficiencies described in the ins. Despite these notices, a number of events similar to those addressed in the ins have recently occurred, thus indicating continuing problems with these breakers.
The staff identified a few accident sequence precursor (ASP) events at the plant-specific level, involving medium-voltage circuit breaker failures in which conditional core damage probabilities (CCDPs) were in the low E-5 range. The magnitude of CCDPs for these events is in the low risk significance category as compared to other events (grecter than 1.0E-4) reported in the ASP reports (NUREG-4674). Based on these risk insights, generic reguletory actions would not be warranted. However, reviews performed by NRC contractors indicate that breaker ma!! unctions are significant contributors to pump unavailability and reliability. The majority of these breaker failures were discovered when pumps failed to start on demand either in service or during surveillance testing.
The staff has issued eight ins (four on medium-voltage circuit breaker problems and four on low-voltage circuit breaker problems) since 1996. In view of recurring problems with medium-voltage and low-voltage circuit breakers, the staff prepared a generic letter to address this issue. On May 13,1997, the staff requested that the Committee to Review Generic Requirements (CRGR) review and endorse the proposed generic letter entitled Problems With Medium-Voltage Circuit Breakers." At the CRGR meeting on June 12,1997, the committee determined that the generic letter was not the appropriate vehicle for
{
correcting the breaker problem and instead recommended that the staff issue e temporary instruction (TI) j and conduct targeted inspections to determine the extent of the generic problem with breakers and to
{
ensure licensee compliance with NRC regulations, especially the provisions of 10 CFR Part 50,
{
Appendix B, and the maintenance rule, as appropriate. The staff prepared a Tl covering both medium-voltage and low-voltage metal-clad circuit breakers, which was issued on December 31,1997. This Tl was performed at two sites in Regions I,11, Ill, and IV with oversight and support from NRR (lOMB and EElB). NRR completed the breaker inspections in October 1998 and is currently evaluating whether additional regulatory actions are needed.
Proposed Actions: Specific actions included in the action plan are: (1) issuing a temporary instruction and conducting targeted inspections to ensure licensee compliance with NRC regulations, especially the provisions of 10 CFR Part 50, Appendix B, and the maintenance rule, as appropriate; (2) issuing a white paper describing status of staff and industry actions to address circuit breaker problems; (3) issuing an Information notice describing recent (Pre Tl inspection) findings regarding maintenance per vendor j
manuals, review of ins and industry experience; (4) inspecting the major circuit breaker vendors' facilities and some of the third party repair / refurbishment outfits; (5) monitoring the breaker failures via review of LERs and identify potential safety problems as a result of breaker failures and review of 10 CFR Part 21 submittals; (6) participating in industry meetings; and (7) summarizing Tl inspection findings and iscuing generic communication if required.
Current Status: Tl was issued on December 31,1997. A revision to the Tl was issued on March 9,1998.
The white paper describing status and industry actions to address circuit breaker problems was issued on March 25,1998. Also, the plant specific inspections per Tl 2515/137 were completed in October 1998 as indicated above. On April 22,1999, the staff evaluated the inspection findings and concluded that no additional regulatory action is required for addressing circuit breaker maintenance programs. The staff issued an Information Notice (IN 99-13) to inform the industry of circuit breaker inspection findings and insights on April 29,1999. Item 4 (Inspect at least three major circuit breaker vendor's facilities and at least three of the third-party repair / refurbishment outfits) of the Task Action Plan is scheduled to be completed by September 30,1999, items 5 (Review of LER and 10 CFR Part 21 submittals) and 6 (Participate in Industry meetings) are ongoing. A memorandum closing the Action Plan is in the concurrence process.
NRR Technical Ccatact:
A. Pal,415-2760 16
ENVIRONMENTAL SRP REVISION ACTION PLAN TAC No. MA0837 Last Update: 07/19/99 GSI: Not Available Lead NRR Division: DRIP MILESTONES DATE (T/C) 1.
Reflect Potential impacts and Integrated Impacts in Options for Resolution a.
Identification of potentialimpacts 03/96C b.
Identification of integrated impacts 06/96C c.
Proposed options for resolution and develop initial draft of revised ESRP 10/96C d.
Staff / contractor meeting to resolve format and content of revised ESRP 11/99C 2.
Prepare Final Draft of ESRP Sections for Public Comment
- a. Draft updated ESRP for staff review 01/97C
- c. Publish (electronic) for public comment 09/97C 3.
Disposition Public Comments 02/980 4.
Publish Final NUREG-1555 10/99T 5.
Maintenance of program data Ongoing Brief
Description:
The Environmental Standard Review Plan (ESRP) Revision Action 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 environmental Justice) since originally published in 1979, and to support the review of license renewal applications.
Reoulatory Assessment: NRR has established the ESRP Update Program for use in the life cycle review of environmental protection issues for nuclear power plants, especially license renewal applications, but also operating reactors, and future reactor site approval applications. The ESRP will reflect current NRC requirements and guidance, consider other statutory and regulatory requirements (e.g., the National Environmental Policy Act, Presidential 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: A PNNL/MCC sMff workshop on the restructured and revised ESRP was held during November 13-14,1996. Now that the. Part E1 rule for license renewal is final, particular emphasis is being placed on assuring that license reewat needs are being addressed in a schedule consistent with the RES regulatory guide and pilot plant application. 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 staff indicated 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 notice 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.
ACRS discussed the NUREG at its May 1,1998, meeting; in subsequent interactions with ACRS staff, the Committee determined that it no longer needed a subsequent SAMDA/SAMA briefing on the ESRP or 17 l
I i
any environmental document prepared by the staff for license renewal unless staff practice changes.
During the week of February 9,1998, the staff developed the comment binning and disposition plan; subsequently, a PNNUNRC staff workshop was held during February 24-25,1998, to disposition technical comments and make decisions regarding the organizational 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) arvi that for Part 54 (renewal of a license for an existing facility). This segregation took the form of a supple. nt to the ESRP and was completed in draft form on July 3,1998. During December 1-3,1998, the final I. INUNRC staff workshop was held to consider how and whether comments raised on the companion RG for license renewal should be dispositioned for the ESRP. The fir.al draft of the ESRP for NRC concurrence was provided by the contractor in February 1999.. When the staff finalizes its positions on Severe Accident Mitigation Altematives and scope of the transmission lines appropriate for license renewal in conjunction with its reviews for Calvert Cliffs and Oconee, the staff will assure that the updated ESRP reflects those positions. The ESRP will now be published in a time frame consistent with Regulatory Guide 4.2, j
Supplement 1, on the format and content of the environmental report for license renewal.
NRR Technical
Contact:
B. Zalcman, RGEB,415-3467 i
l 1
l l
1 18 l
l l
EMERGENCY ACTION LEVEL GUIDANCE DEVELOPMENT TAC No.: MA3695 Revision to NESP-007 Last Update: 06/30/99 M98020 Shutdown EAL Guidance Lead NRR Division: DIPM REVISION TO NESP407 (NEW DOCUMENT NEl-97-03)
MILESTONES DATE (T/C) 1.
Meet with NEl to discuss NEL97-03 10/19/98C 2.
NEl to provide revised NEl-97-03 with shutdown EAL guidance removed 11/2/98C for NRC comment 3.
NRC provide comments on NEh97-03 to NEl 12/3/98C 4.
NEl submit NEl-97-03 for NRC endorsement 1/11/99C 5.
Draft Guide developed (Revision to Regulatory Guide 1.101) endorsing 8/99T NEl-97-03 for interim use and comment 6.
CRGR meeting on draft guide 9/99T j
7.
Draft Guide issued for public comment 10/99T j
8.
Public comments addressed (any needed revision to NEl-97-03 1/00T completed) i 9.
CRGR/ACRS meeting on final guide' 5/00T
- 10. Regulatory Guide lesued' 7/00T
- NEl intends to combine NEl-97-03 with NEl-99-01 into a single EAL guidance document EAL GUIDANCE FOR COLD SHUTDOWN, REFUELING AND LONG TERM FUEL STORAGE (" SHUTDOWN EAL GUIDANCE" NEl-99-01)
, MILESTONES DATE (T/C) 1.
Meet with NEl to resolve staff concerns on NEl's guidance (proposed in 1/28/99C NEl-97-03) for EALs applicable in the shutdown mode of operation 2.
NEl to provide new shutdown EAL guidance (NEl-99-01) for NRC review 4/07/99C 3.
NRC provides comments to NEl on NEl-99-01 5/11/99C 4.
Meet with NEl to discuss comments 5/13/99C 5.
Comments resolved and final draft of NEl-99-01 submitted for 8/99T endorsement 6.
Draft guide developed endorsing NEl-99-01 developed in form of a draft 9/99T guide for CRGR/ACRS review.
7.
Determination made on whether to issue a Generic Letter on plant-9/99T specific implementation of shutdown EALs 8.
CRGR/ACRS meetino on draft auide and aeneric letter 10/99T 19
9.
Draft Guide issued for public comment 12/99T
- 10. Public comments addressed (NEl-99-01 revised as needed) 3/00T
- 11. CRGR/ACRS meeting on final guide and generic letter 5/00T
- 12. Regulatory Guide and generic letter issued 7/00T Descriotion: This action plan is intended to guide staff efforts to review (and endorse, if appropriate) industry-developed emergency action level (EAL) guidance. This action plan consists of two elements:
(1) review of the NEl revision to NUMARC/NESP-007 existing guidance for EALs, and (2) review of new guidance under development for EALs for the shutdown and refueling modes of reactor operation and for long-term fuel storage.
Historical Backaround: 10 CFR 50.47(b)(4) and Appendix E to 10 CFR Part 50 require licensees to develop EALs for activating emergency response actions. NUREG-0654/ FEMA-REP 1, issued in 1980, i
provides example initiating conditions for development of EALs [1].
The NRC's evaluation of the 1990 Vogtle Loss Vital AC Power event identified two areas where NRC's EAL guidance and licensee's EAL schemes were deficient: (1) loss of power EALs were ambiguous and (2) EAL guidance for classifying events that could occur in the shutdown mode of plant operations was not available [2]. The NRC's evaluation of shutdown and low power operation in NUREG-1449 also identified a need for guidance for EALs applicable in the shutdown mode of operation [3].
In 1992, the industry issued EAL guidance in NUMARC/NESP-007, Revision 2 [4]. This guidance is more detailed than the guidance provided in NUREG-0654 (e.g., it includes example EALs and bases for the EALs in addition to example initiating conditions) and is based upon 10 years of industry experience in developing EAL schemes. In 1993, the NRC endorsed the industry guidance as an acceptable alternative to the NUREG-0654 guidance in Regulatory Guide 1.101, Revision 3 [5]. The industry guidance addressed the concerns regarding ambiguities in the loss of power EALs and, to a limited degree, addressed concerns with EAL guidance for events initiated in the shutdown mode of operation.
However,it was recognized that further guidance for EALs applicable in the shutdown mode was needed.
In September 1997, the Nuclear Energy Institute (NEI) submitted a proposed revision to NUMARC/NESP-007 (issued as NEl 97-03)[6]. This revision provided additional guidance for EALs applicable in the shutdown and refueling modes of plant operation and incorporated a number of improvements and clarifications to the existing EAL guidance in NUMARC/NESP 007. The need for these changes was identified during the development and review of site-specific EAL schemes based on the NUMARC/NESP-007 guidance.
ProDosed Actions: Endorse industry-developed EAL guidance in revisions to Regulatory Guide 1.101.
Determine whether development of a Generic Letter which requests licensees to incorporate EAL guidance for classifying events initiated in the shutdown and refueling modes of plant operation is warranted. Issue generic letter if it is determined to be warranted.
Oriainatino Documents:
Vogtle llT EDO Staff Action item 4a [7]
NUREG-1449 Reoulatory Assessment: EALs are used to classify events in order to initiate emergency response efforts.
Multiple indicators are used in EAL schemes to determine the significance of events. Licensees' current EAL schemes include EALs that can be used to classify events initiated in the shutdown and refueling modes of operation (e.g., radiation monitor-based EALs and judgement EALs). However, guidance is 20
needed to improve licensees' capability (with regard to timeliness and accuracy) for assessing and classifying the significance of events that occur in the shutdown mode of plant operation.
Current Status: NRC provided comments to NEl on the proposed EAL guidance in a letter dated March 13,1998 [8]. In meetings held in March and June 1998 [9,10] the proposed EALs were discussed
/
and the industry provided proposed modifications to the shutdown EALs. The NRC provided comments on the proposed modifications in a letter dated August 3,1998 [11].
In a letter dated August 13,1998, NEl proposed an adjustment to the approach for the development of industry EAL guidance [12]. A two-phase approach was proposed. The first phase would focus on incorporating clarifications to the existing guidance. The second phase of the project will produce a new document to be numbered NEl 99-01, " Methodology for Development of Emergency Action Levels for Cold Shutdown, Refueling, and Long Term Fuel Storage," and will provide EAL guidance for cold shutdown and refueling conditions, defueled plants, and dry-cask fuel storage users.
4 The industry provided the final draft of NEl 97-03 Revision 3 for NRC review and approval in a letter dated January 11,1999. The industry provided a draft of NEl 99-01 to the NRC for its review by April 7, 1999. NRC provided comments on NEl 99-01 to NEl in a letter dated May 11,1999. In a memorandum dated June 3,1999, ACRS informed the EDO that it decided not to review NEl 97-03 but would like a briefing on NEl 99-01.
Referenegg:
1.
NUREG-0654/ FEMA-REP-1," Criteria for the Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Revision 1, November 1980.
2.
NUREG-1410," Loss of Vital AC Power and the Residual Heat Removal System During Mid-Loop Operations at Vogtle Unit 1 on March 20,1990," June 1990.
3.
NUREG-1449, " Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States," September 1993.
4.
NUMARC/NESP-007, Revision 2," Methodology for Development of Emergency Action Levels,"
January 1992.
5.
Regulatory Guide 1.101, Rev. 3, " Emergency Planning and Preparedness for Nuclear Power Reactors," August 1992.
6.
Letter from A. Nelson to J. Roe, September 16,1997.
7.
Memorandum from J. Taylor to T. Murley, June 21,1990.
8.
Letter from B. Zaleman to A. Nelson, March 13,1998.
9.
Memorandum from S. Magruder to T. Essig. June 26,1998.
10.
Memorandum from S. Magruder to T. Essig, June 26,1998.
11.
Letter from C. Miller to A. Nelson, August 3,1998.
12.
Letter from A. Nelson to C. Miller, August 13,1998.
NRR Technical Contacts:
J. O'Brien, DIPM,415-2919 R. Sullivan, DIPM,415-1123 L. Lois, DSSA,415-2897 Lead PM:
J. Birmingham, DRIP,415-2829 21
r; PRA IMPLEMENTATION PLAN 1.2(d)
Graded Quality Assurance Action Plan TAC Nos. M91429, M92447 Last Update: 7/19/99 M92448, M92449, M88650, M91431 Lead NRR Division: DIPM M91432, M91433, M91434, M91435 Support Division: DSSA M91436, M91437, M92420 and M94163 GSI: Not Availab!e MILESTONES DATE (T/C) 1.
Issued SECY 95-059 03/95C 2.
Begin interactions with volunteer licensees 05/95C Palo Verde letter dated 4/6/95 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 Meetings on: 8/25/95,10/10/95,10/25/95 4.
Staff interactions 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 3/98C 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 GA 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 GOA initiative provided by PVNGS at 2/27/97 meeting with staff GOA closeout letter transmitted to licensee on 7/2198 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 OA 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 22 1
1
I Meeting on 6/19/96 to discuss staff comments on the OA plan submittal for graded OA, review questions transmitted to STP on 8/16/96 Site visit on August 2122 to observe working group and expert panel meetings, and to discuss staff review items, trip report in preparation Management meeting on 10/15/96 to discuss PRA initiatives and staff activities Letter from licensee dated 10/30/96 responding to PRA questions Revised QA plan submitted on 1/21/97 Overview of STP initiative provided at 2/27/97 meeting with the staff Staff Request for AdditionalInformation (RAl) 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, OA controls for the PRA, corrective action and performance monitoring feedback processes, audit scheduling, and responses to the RAI concerns. Trip report issued on 7/10/97.
STP submittal on 5/8/97 for preliminary RAI response STP submittitl of draft QA Plan on 5/21/97 STP submittal of GOA related procedures, responses to RAl, and follow-on OA Plan on 5/22/97 STP submittal of revised QA Plan on 6/10/97 Staff RAIissued on 6/13/97 STP submittal on June 26,1997, response to staff RAI STP submittal of revised QA Plan on 7/16/97 STP transmittal of additionalinformation regarding GOA 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 STP meeting with staff on 9/15/98 to discuss GOA implementation and issues associated with technical requirements imposed on low risk significant, but safety-related equipment 23
STP letter of 10/14/98 proposes to be a pilot to use GOA risk ranking results to support discontinuance of technical provisions (seismic and equipment qualification, ASME requirements) for low and non-risk significant safety-related equipment 6.
Staff interactions with Grand Gulf Ongoing Site meeting on 7/11-14/95 to observe expert panel through Meeting at hdqt. on 10/24/95 on OA controls 3/98C 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 1.etter to GGNS dated 5/29/96 regarding implementation of QAP commitments 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 OAP 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 GOA closeout letter transmitted to licensee on 1/7/98 7.
Revision 3 of Draft Evaluation Guide for Volunteer Plants issued for staff 07/95C comment 8.
Revision 4 of Draft Evaluation Guide for Volunteer Plants issued for Steering 10/95C Group Review 9.
Issue letter to 3 volunteer plants outlining program objectives and review I
expectations. Distributed staff evaluation guide to licensees.
1/96C 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.
4/960
)
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 8/1/96C Draft RG for ACRS/CRGR review 11/22/96C Draft RG for public comment 6/25/97C Draft RG public comment period ends 9/23/97C Public workshop held on draft RG 8/12/97C Publish final RG in SECY 98-067 4/2/98C l
SRM conditionally approves issuance of GOA RG 6/29/98C GOA final RG issued 8/980 24
12.
ACRS Briefings Expert Panel and deterministic considerations 2/27 28/960 Graded QA 4/11/96C PRA implementation P'an and pilot projects 7/18/96C Risk Informed Pilots 8/7/96C Graded QA Regulatory Guide 11/22/96C Graded QA Regulatory Guide 2/21/97C ACRS Concerns on GOA Regulatory Guide 3/6/97C ACRS memo to Commission expressing concerns with GOA approach 3/17/97C Public Comments on GOA Regulatory Guide 10/21/97C Application RG/SRP discussions with Subcommittee 2/19/98C Application RG/SRP discussions with Full Committee 3/3/98C 13.
CRGR Briefings Graded QA Regulatory Guide 11/26/96C Graded QA Regulatory Guide 3/11/97C Graded QA Regulatory Guide 2/27/98C Graded QA Inspection Procedure 12/8/98C Graded QA Inspection Procedure 6/22/99C Graded QA Inspection Procedure 8/4/99T 14.
Issue draft Staff Inspection Guidance (Baseline + Reactive IP) for comment 9/29/98C ACRS Full Committee Meeting 11/6/98C Letter from ACRS endorsing IP dated 12/13/98 CRGR meeting on GOA IP 12/8/98C Memo from CRGR dated 12/17/98 expressing concerns with IP Issue finalinspection procedure 8/99T 15.
Conduct NRC Staff Training 9/99T
==
Description:==
Prepare staff evaluation guidance and regulatory guidance for industry implementation for the grading of quality assurance (OA) practices commensurate with the safety significance of the plant j
equipment. The development of this guidance will be based on staff reviews of regulatory requirements,
)
proposed changes to existing practices, staff development of a draft regulatory guide with input from a national laboratory, and assessment of the actual programs developed by the three volunteer utilities implementing graded quality assurance programs.
Historical Backaround: The NRC's regulations (10 CFR Part 50, Appendices A & B) require OA programs that are commensurate (or consistent) with the importance to safety of the functions to be performed.
However, the OA implementation practices that have evolved have often not been graded. in the development of implemen~tation guidance for the maintenance rule, a methodology to determine the risk significance of plant equipment was proposed by the industry (NUMARC 93-01). During a public meeting on December 16,1993, the staff suggested that the industry could build on the experience gained from the maintenance rule to develop implementation methodologies for graded QA. The staff had numerous interactions with the Nuclear Energy Institute (NEI) during calendar year 1994 as the graded QA concepts were discussed and the initial industry guidelines were developed and commented on. In early 1995, three licensees (Grand Gulf, South Texas, and Palo Verde) volunteered to work with the staff. The staff has reviewed the licensee developmental graded QA efforts.
25 f
Proposed Actions: The goal of the action plan is to utilize the lessons leamed from the 3 volunteer licensees to modify staff-developed 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 baseline and reactive inspection procedure (IP) for graded QA. An inter-office team has been established to prepare the regulatory guidance documents and test their implementation during the evaluation of volunteer plant activities.
Oriainatina Document: Letter from J. Sniezek, NRC to J. Colvin (NUMARC) dated January 6,1994, describing the establishment of NRC steering group for the graded QA initiative.
Reaulatory Assessment: Existing regulations provide the necessary flexibility for the development and implementation of graded quality assurance programs. The staff willissue a NUREG report regarding the lessons leamed from the volunteer plant implementations. Additional regulatory guidance will be issued to either disseminate staff guidance or endorse an industry approach. Planned guidance for the staff will involve an evaluation guide for application to the volunteer plants, the lessons learned report, training sessions and public workshops, and inspection guidance in the form of a baseline and a reactive IP. The staff is evaluating the appropriate mechanism fet inspections of the risk significance determination aspects of graded QA programs.
The safety benefits to be gained from a graded QA program could be significant since both NRC reviews and inspections and the industry's quality controls resources would be focused on the more safety significant plant equipment and activities. Secondarily, cost savings to the industry could be realized by j
avoiding the dilution of resources expended on less safety significant issues. The time frame to complete this action plan is directly related to the overall PRA implementation plan schedules.
Current Status: A draft evaluation guide for NRC staff use has been prepared for application 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 staff's proposed interaction framework has been transmitted in a letter to the three volunteer licensees. The letter sought licensee comments. Draft 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 8,1997) transmitted the draft regulatory guides, including the GOA guide, to the Commission. Commission approval was obtained on June 5,1997, to issue the documents for a 90 day public comment period. Senior management briefings were provided to the Director, NRR (on April 22,1997) and to the Deputy, EDO (on April 24,1997). The public 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 j
Chairman, to focus staff efforts on revising the general regulatory guide and standard review plan first.
The proposal to sequentially complete the application specific guidance documents, including GQA, was also accepted. SECY 97-229 forwarded the staff's evaluation of the STP GOA program with a recommendation that it be approved. The Commission did not object to the issuance of the SER. The staff presented the revised GOA RG to the ACRS (Subcommittee and Full Committee) and the CRGR, comments received during those reviews were addressed as necessary. On April 2,1998, SECY-98-067 was issued which transmitted the GOA RG, along with the other application specific guidance documents, to the Commission. By SRM dated June 29,1998, the Commission conditionally approved the issuance of the GQA RG. Prior to issuance of the RG the staff will have to review, and revise accordingly, 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. The GOA RG was issued in August 1998.
Work has been initiated on developing a GOA inspection procedure (IP). The draft IP was issued for comment on September 29,1998. The IP was transmitted to the regions, ACRS, CRGR, OGC, SRAs, RES, and OE. The staff presented the proposed IP to the ACRS Full Committee on November 6,1998.
26
r By letter dated November 13,1998, the ACRS stated that the proposed IP was found adequate to perform an evaluation of licensee graded QA programs. A meeting was held with CRGR to discuss the IP on December 8,1998. By memorandum dated December 17,1998, the CRGR expressed concerns about the IP and recommended that it not be issued, and the following concerns were expressed:
1.
The IP is not performance based and is structured more as a Standard Review Plan. Most NRC inspectors do not possess the requisite PRA expertise to implement the IP in a uniform manner.
2.
The l'P does not provide objective or deterministic standards for inspectors to make decisions on the adequacy of licensee programs. This could result in backfitting situations where individual inspector views are inappropriately imposed on licensees.
3.
The inclusion of high-safety significant but non-safety-related components in the OA program would be considered unauthorized backfitting if pursued by an inspector.
The CRGR recommended that the staff consider the development of plant-specific Temporary Instructions to verify graded OA implementation.
The staff has considered the CRGR comments and while there is some disagreement with the perceived flaws in the IP, has concluded that the proposed IP should be redrafted to resolve the CRGR concerns.
Conversion of the IP to a Tl that would be specific to South Texas Project was considered, however, the issues arising from implementation of the STP graded QA program are being addressed principally through the licensing review program rather than the inspection program. Therefore the statt plans to revise the draft IP to address generic OA issues and submit it for reconsideration by CRGR. The target date to issue the IPis June 18,1999.
The staff met with CRGR on June 22,1999, to discuss the revised IP. The CRGR acknowledged that the staff had addressed its original concems. However, the Committee raised several new cnneerns and recommendations. The staff is scheduled to meet with CRGR again on August 4,1999, to discuss the staff's response to the Committee's recommendations.
A meeting was held with the three volunteer licensees on April 11,1996, to receive their feedback on the staff developed evaluation guide. The licensees expressed concerns about the level of detail contained in the guide, particularly that related to PRA and commercial grade item dedication. The licensees contend that exiting industry guidance (PSA Application 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 additional regulatory guidance for the graded QA application.
NEl contends that existing industry guidance is sufficient. STP and PVNGS letters providing comments on the evaluation guide were dated April 17,1996, and April 24,1996, respectively. 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 GOA 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 public comment. On March 12,1998, the ACRS issued a letter to the Chairman which recommended that the GOA RG (RG 1.176) be issued for use. The ACRS expressed a concern that RG 1.176 does not take 1
27
]
L
r-full advantage of PRA information. However, the ACRS acknowledged the inherent difficulty given the lack of a model to assess quantitatively the impact of modified QA controls upon the PRA model. The ACRS further recommended that RES consider a research project to assess the impact of QA controls 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 change has been reviewed by the staff (HOMB, SPSB, RES, RIV, and NRC contractors). A meeting was held with STP on June 19 to discuss the staff's 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 significant items. The staff visited the site on August 2122 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 also observed both a Working Group and Expert Panel meeting at which time licensee safety significance evaluations for 2 systems (Radiation Monitoring and Essential Service Water) were discussed. Siaff review of the updated QA program submittal was completed and a second RAI was issued on April 14,1997, for both PRA and QA controls aspects. A meeting was held on April 21,1997, during which the licensee provided some responses to the issues raised in the RAl. Staff (from both HOMB and SPSB) performed a site evaluation during the week of May 5-8 to review aspects associated with: PRA quality, QA controls for the PRA, corrective action and performance monitoring feedback processes, QA controls for low safety significant items, detailed information presented to address issues raised in the RAI, and the audit scheduling process. Further dialogue has occurred between the staff and STP during the review of the subsequent STP submittals and following issuance of staff RAls. SECY paper 97-229 was issued on October 6,1997, to inform the Commission of the staff's 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 staff's safety evaluation was transmitted to the licensee. The licensee provided their interpretation on 1/26/98 of selected aspects of the staff's SER. By letter dated February 19,1998, the staff agreed with the licensee's interpretations. On September 15,1998, the staff met with STP to discuss the experience with implementing GOA. STP indicated that 6 systems had been evaluated and that a majority (89%) of 'he equipment had been found to be low or non-risk significant. STP stated that they had not derived the expected benefit from GOA due to other technical provisions (such as the ASME Code and seismic qualification) that are required for safety-related equipment. STP further informed the staff that they desired to identify a mechanism that could provide broad regulatory relief in these areas for low safety significant equipment. The staff acknowledged STP's concerns and indicated that these issues are related to the initiative to evaluate Part 50 with respect to making it more risk-informed. The staff agreed to meet again in the October time frame to continue the discussions with STP. In addition on September 15,1998, STP provided a presentation to all interested NRC staff on their overall strategy to implementing risk-informed approaches at their f acility. By letter dated October 14,1998, STP expressed their desire to be a pilot plant to utilize the risk ranking results from the graded QA effort, to justify the discontinuance of certain technical provisions. STP stated that for low and non-risk significant, safety-related equipment that exemptions are warranted to remove those components from the scope of seismic and environmental qualification programs, in addition, ASME code requirements should be removed through relief requests. The STP proposal has been integrated into the Part 50 risk-informed SECY paper.
Also, NEl submitted 96-02, " Guideline for implementing a Graded Approach to Quality" dated March 21, 1996. The staff has perforrned a cursory review of the document and concluded that it does not reflect the progress 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 objective means to assess the adequacy of QA program implementation. NEl has 28
n s
l l
l 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 received from the volunteer plants and industry sources. The staff has reviewed the modified document. On October 10,1996, NEl submitted a letter expressing their concem with the graded QA initiative. NEl stated their concems 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 meeting 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 flexibility to manage their QA programs. The volunteer plant staff stated their firm desire to obtain copies of the draft GOA Regulatory Guide in a timely manner, following 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 GOA efforts.
NRR Contacts:
T. Quay,415-1017 D. Dorman,415-1425 RES
Contact:
H. Woods,415-6622
References:
1)
Letter from J. Sniezek (NRC) to J. Colvin (NEI) dated 1/6/94.
2)
3)
NUMARC 93-01, " Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants."
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.
6)
Letter from C. Thomas (NRC) to W. Stewart (APSCo) dated 12/4/95.
7)
Memorandum from S. Black to W. Beckner and W. Bateman dated 1/24/96, Draft Staff Evaluation Guidance.
8)
NEl 96-02, " Guideline for implementing a Graded Approach to Quality."
9)
Draft Regulatory Guide-1064, "An Approach for Plant-Specific, Risk-informed Decision Making:
Graded Quality Assurance," dated March 24,1997, 10)
SECY-97 229," Graded Quality Assurance /Probabilistic 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 11/6/97.
12)
Letter from J. Donohew to J. Hagan (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.
14)
Regulatory Guide 1.176,"An Approach for Plant-Specific, Risk informed Decisionmaking:
Graded Quality Assurance," August 1998.
29 i
w ACCIDENT MANAGEMENT IMPLEMENTATION t
i TAC #: M91966 - Overall Last Update: 7/15/99 M91641 - BWROG SAMG Review Lead NRR Division: DSSA MILESTONES DATE (T/C) 1.
BWROG Severe Accident Management Guidance (SAMG) documents Complete review of SAMG documents 7/98C Resolve remaining technical concerns 9/99T 2.
Review severe accident training materials and BWROG prioritization -
6/95C methodologies 3.
Develop guidance for A/M audits initial draft (for internal use) 11/95C Industry-sponsored A/M demonstrations 3/98C Revised draft (to NEl and public) 8/98C Final TBD 4.
Develop approach for A/M oversight as part of guidelines for use of 5/00T industry initiatives i
i 5.
Complete A/M audits TBD 6.
Hold public workshop TBD 7.
Report to Commission on audit findings and recommendations for TBD achieving closure
==
Description:==
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 willinclude: 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.
Historical Backaround: The issue of A/M and the potential reduction in risk that could result from i
developing procedures and training operators to manage accidents 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 i
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 continued to work with industry to define the scope and content of utility A/M programs and these efforts culminated in industry-developed A/M guidance for utility implementation. Industry committed to implement an accident management program at each NPP [6). NRC accepted the industry j
commitment with the understanding that the staff would inspect utility implementation [7).
J ProDosed 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 1
workshop to discuss audit findings. Following the workshop, the staff will report to the Commission on audit findings and recommendations for remaining actions to achieve closure.
30 J
T Oriainating Document: SECY-88-147, Integration Plan for Closure of Severe Accident Issues, May 25, 1988.
Reaulatory Assessment Accident management 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 action and continued facility operation is justified.
Current Status-Severe Accident Management Guidelines Severe accident 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 high level review of the EP/ SAG documents. Areas where additional information and discussion with the BWROG are considered necessary were identified in an April 2,1997, letter to the owners group [10). A BWROG submittal describing a time line for actions that operators would take according to the EP/ SAG was received in May 1997 [11). The BWROG response to the April 2,1997, staff letter together with Rev.1 of the EP/ SAG was received in January 1998 [12). The staff has completed its review of the BWROG response. Remaining concems with the EP/ SAG were provided to the BWROG in a July 20,1998, letter
[13). During an NRC/BWROG management meeting on August 5,1998, the BWROG proposed and the NRC agreed that technical discussions on remaining issues be deferred until early 1999 to permit BWR licensees to complete A/M implementation before redirecting their resources towards addressing the remaining technical concems. The BWROG has evaluated the staff concems and expects to provide a written response to the NRC. This response has not yet been received.
Utilityimplementation Licensee commitments and target dates for' completing A/M implementation were submitted to NRC in 1995 as part of the industry initiative implementation was scheduled to be completed at all sites by the end of 1998. Several areas of the industry initiative needing clarification were brought to NEl's attention by licensees during A/M implementation. In response, NEl developed supplemental guidance to address these areas and provided this guidance to industry and to NRC in a July 22,1C7, letter [16). NRC provided comments on this guidance in a January 28,1998, letter to NEl [17]. ~.n an April 3,1998, letter
[18), NEl expressed concem that NRC appears to be reversing previously understood positions and
. escalating expectations. The staff positions on licensed operator training and evaluation, use of a systematic approach to training, and application of 10 CFR 50.59 were of greatest concem to industry, in a June 25,1998, response [19], NRC provided clarification regarding the staff positions and the approach to reaching closure. The staff indicated that they do not see major differences in NEl's and NRC's expectations, and that industry should continue to proceed with implementation.
j Implementation has now been completed at almost all plants. Two utilities have rescheduled their completion dates beyond 1998. All but 8 utilities have provided letters confirming that implementation has been completed. Staff is following up with those utilities regarding their implementation status.
NRC Evaluation The staff outlined plans to evaluate licensee A/M implementation in separate communications with NEl and the Commission in 1995-1996 [14,15]. Major steps included: (1) conducting information gathering visits at two to four sites to observe how the elements of the formal industry position are being implemented, (2) completing a temporary instruction (TI) using insights obtained through the site visits, (3) performing pilot inspections at about five plants using the TI, (4) developing an inspection procedure (lP) for use at remaining plants based on findings from the pilot inspections and feedback from industry, (5) evaluating implementation at remaining plants using the IP, and (6) in the longer term, evaluating A/M maintenance on a for-cause basis as a regional initiative.
31
In January 1997, the staff agreed to participate in a limited number of industry-organized A/M demonstration visits in lieu of the information gathering visits, and to reassess the need for inspections at the remaining plants after the A/M demonstrations. The A/M 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).
In June 1998, upon further consideration of the voluntary nature of this program, the staff concluded that the A/M evaluations should be performed as audits rather than inspections [19]. The objectives of the audits would be to assess how licensees have evaluated and implemented enhancements to A/M capabilities in accordance with formal industry position, and to establish a basis for a decision regarding the need for future inspections or any other regulatory action.
A draft Tl for use in planned pilot inspections was completed in February 1996, and discussed with industry, ACRS, and NRC Regional office staff in separate meetings in early 1996. The Tl was subsequently recast as audit guidance, and updated to incorporate insights from the A/M demonstration visits, staff positions contained in NRC letters to NEl, and feedback received on the draft TI. The audit guidance was provided to NEl in an August 10,1998, letter, and placed in the Public Document Room
[20). During an October 1998 meeting with NEl regarding the audit guidance, NEl proposed that NRC cancel plans for the A/M audits and workshop on the basis that the four completed demonstrations provide a sufficient understanding for NRC to decide on the acceptability of industry improvements regarding A/M. NRC agreed to reconsider the necessity of the audits and workshop, and to explore ways to achieve the goals and objectives of the audits within the context of the risk-informed reactor oversight process.
The staff prepared a draft Commission paper describing an approach for monitoring voluntary programs, including A/M, as part of the risk-informed inspection and assessment process. However, on May 27, 1999, the Commission issued a Staff Requirements Memorandum (SRM) on SECY 99-063,"The Use by industry of Voluntary Initiatives in the Regulatory Process," directing the staff to develop an approach and guidelines for using industry initiatives in the regulatory process, and to ensure that the guidance developed accounts for the inspection and enforcement of voluntary initiatives that are implemented in lieu of regulatory requirements. The issue of A/M oversight will be subsumed by this activity, which is being led by the Division of Engineering (DE). DSSA has suspended efforts on the draft paper regarding monitoring voluntary programs, and will support DE's activities to develop guidelines for use of industry initiatives, including planned stakeholder meetings. The guidelines are scheduled to be provided to the Commission for review in May 2000.
References:
1.
Memorandum from F. Rowsome to W. Minners, "A New Generic Safety issue: Accident Management," April 16,1985 2.
SECY-88147, 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 A/M 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 f rom 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 T. Essig to T. Rausch, July 20,1998 i
14.
Letter from A. Thadant to T. Tipton, August 3,1995 15.
SECY 96-088, Status of the Integration Plan for Closure of Severe Accident Issues and the Status of Severe Accident Research 16.
Letter from D. Modeen to G. Holahan, July 22,1997 32
17.
Letter from G. Holahan to D. Modeen, January 28,1998
' 18.
- Letter from R. Beedle to S. Collins, April 3,1998 19.
Letter from S. Collins to R. Beedle, June 25,1998 20.
Letter from S. Collins to R. Beedle, August 10,1998 NRR Technical
Contact:
R. Palla, SCSB,415-1095 NRR Lead PM-Ramin Assa DRPW,415-1391 i
l i
i 33
HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN (Previously Part of the Dry Cask Storage Action Plan)
Final Update TAC Nos. M93821: Action Plan Last Update: 07/13/99 M91955: DSC generic review Lead NRR Division: DSSA M95546: Generic review of NRCB 96-02 Status: Com)lete ACTION DATE (T/C)
- 1. Review, summarize and issue existing NRC guidance on heavy load control.
Review NUREG-0554, NUREG-0612, GL 80-113, GL 81-07, 2/96C GL 85-11, and other supporting documents.
Develop summary of guidance.
2/96C
- 2. Determine significant heavy load issues that need to be addressed and develop resolution method.
Generic letter 85-11 and NUREG-0612.
2/960 Single-Failure-Proof Crane (reliability).
TBD*
Spent fuel cask drop accident prior to securing the lid.
12/96C Risk significance of multiple failures within safe load path.
TBD*
- 3. Review licensee implementation of heavy load control, including applicable (ongoing) correspondence from a sample of licensees and site visits.
- 4. Review NRC audit / inspection procedures, practices, inspection reports, 5/96C enforcement actions, and experience.
- 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 appropriate management (SPLB, TBD*
Region I, NRR) and obtain/ resolve any comments. (Propose form of guidance).
(Contingent on resolution of item 2 above)
- 7. Issue the draft inspection procedure (s) (issue draft inspection procedure).
12/99
- 8. Obtain feedback (meeting, FRN, or other means) conceming the staff position TBD*
from industry representatives and resolve any discrepancies with the industry position.
- 9. Develop final version of guidance and obtain management concurrence.
TBD*
- 10. Issue final inspection procedures.
- 11. Issue final guidance.
TBD*
Note:
The Task Action Plan is closed based on RES's acceptance of a proposed Generic Safety issue (GSI).
The assigned GSI number is 186 and the prioritization evaluation is expected to be completed by November 1999. (Refer to RES letter dated May 27,1999). The development of an inspection Procedure for heavy load handling operations will be pursued by SPLB independent of the GSI activities.
34
Descriotion: The Heavy Load Control (HLC) and Crane issues task action plan will identify, 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 description of the scope of the actions under the action plan).
Historical Backaround: Recent increases in licensees' activities involving the transfer of spent fuel to independent spent fuel storage installations (ISFSis) have given rise to a number of concerns 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 existing 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 existing 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, and was recently revised to better reflect the scope and magnitude of the task.
Proposed Actions: Actions included in the plan are: (1) understand the current regulatory framework and inform the staff; (2) review the general issues and identify specific problems to be addressed; (3) develop corrective actions to resolve the problems; and (4) implement the corrective actions. Specific corrective actions may include the issuance of guidance to licensees alerting them to the potential problems and requesting that corrective measures be taken to preclude accidents.
Oriainatina Document: Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995, " Dry Cask Storage Action Plan."
Reaulatorv Assessment: Several licensees have either developed 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. Questions 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, severallicensees have undertaken efforts to assess their plans, capabilities, 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.
Current Status: Review of the respcnses to Bulletin 96-02 was closed, on a generic basis, in April 1998.
The staff committed to evaluate licensees' heavy load handling programs on a plant specific basis.
Projects continue to issue licensee specific closeout letters. The staff continues to interact with licensees on a plant specific basis.
Staff efforts to work with RES to evaluate risks of crane f ailures were abandoned in early 1998 due to budget shortfalls. The staff proposed to RES, a Generic Safety Issue (GSI) on the potential risks and consequences of heavy load drops (probability of crane failure) during the movement of heavy loads (completed 04/19/1999). Based on RES's acceptance of the proposed GSI, the TAP is closed and activities involving the development of an inspection procedure for heavy load handling operations will be pursued by SPLB independent of the GSI. The heavy loads issue (USI A-36) was previously reported to Congress as resolved based on the implementation of NUREG-0612. However, the proposed GSI on the same issue is based on the results of the staff's review of licensee responses to NRC Bulletin 96-02. The staff found that there is a substantially greater potential for severe consequences to result from a load drop than previously envisioned. The staff will coordinate with RES on determining whether the issue is a valid GSI. NRR will work with RES to prioritize the proposed GSI by November 1999.
35
i The staff visited Cai/ert Cliffs in 1997 for the purpose of obtaining an understanding how the various elements of the licensees' programs are being implemented, information and perspectives gained through such 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 Carl F. Lyon,415-2296 Joseph E. Carrasco, RGN-l/DRS, (610) 337-5306
References:
Memorandum from Robert M. Bemero and William T. Russell to James M. Taylor, March 15,1995,
" Realignment of Reactor Decommissioning Program."
Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995, " Dry Cask Storage Action Plan."
36
i WOLF CREEK DRAINDOWN EVENT: ACTION PLAN Final Update TAC No.: M66278 Statue: Complete Lead NRR Division: DSSA MILESTONES DATE (T/C) i
- 1) Draft Generic Letter (GL) 11/95(C)
- 2) Issue Supplement to IN 95-03 [Ref. 3]
03/96(C)
- 3) CRGR Concurrence of the GL for 1" time 09/96(C)
CRGR Concurrence of the GL for 2"8 time (after reconciling public omments) 01/98(C)
- 4) ACRS Briefing 11/97(C)
- 5) GL lssued [Ref. 4]
05/98(C) 1
- 6) Complete Draft Tl/ Issue to the Regions for Comments 12/98(C) l l
- 7) Receive Regional Comments on Tl 01/99(C)
I
- 8) Issue Ti [Ref. 5]
06/99(C) 1 Descnotion: The 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 Backaround: On September 17,1994, the Wolf Creek plant experienced loss of reactor coolant system (RCS) inventory, while transitioning to a refueling shutdown. The event occurred when operators cycled a valve in the train A side of the RHR system cross-connect 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 (RWST) 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.
4 The Wolf Creek event represents a LOCA with the potential tn 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, net 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, and (3) issue a Temporary Instruction (TI).
Oriainatina Document: AEOD/S95-01, " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994".
37
Reaulatory Assessment: The staff performed an evaluation of the probability for event initiation and of the conditional core damage probability. The value of this probability for core damage along with licensee awareness for this scenario makes the risk for continued PWR operation acceptably small.
NRR Technical
Contact:
M. M. Razzaque, SRXB,415-2882 NRR Lead PM:
Kristine Thomas,415-1362
References:
- 1) AEOD/S95-01, " Reactor Coolant System Blowdown 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 Emergency Mitigation Functions while in a Shutdown Condition," issued May 28,1998.
- Draindown During Shutdown and Common-mode Failure," issued June 18,1999.
I 4
)
38
ECCS SUCTION BLOCKAGE l
(NEW)
TAC Nos. MA0626, MA0704, MA1719, Last Update: Initial Update and MA0698 Lead NRR Division: DSSA Supporting Divisions: DE, DRCH, and DRA (RES)
GSI: 191 MILESTONES DATE (T/C)
PART 1:
BWR ECCS SUCTION STRAINER CLOGGING ISSUE l
1.
NRCB 96-03," Potential Plugging of Emergency Core Cooling Suction l
Strainers by Debris in Boiling-Water Reactors" l
o Complete review of licensee responses 8/99 T o Complete audits of 4-6 plants 9/99 T l
o Complete hydrodynamic load review 11/99 T
)
o Evaluate impact of coatings research on BWR resolution 1/00 T PART 11:
NPSH EVALUATIONS 1.
GL 97-04," Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps" Complete review of licensee respons6s 10/99 T o
o Complete revision of RG 1.i/RG 1.82 12/00 T PART lil:
CONTAINMENT COATINGS 1.
GL 98-04," Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Materialin Containment" o Complete review of licensee responses 12/99 T o Complete revision of RG 1.54 12/99 T o Publish summary of GL responses 6/00 T 2.
NRC-sponsored research program on the potential for coatings to fail during an accident Phase I analytical evaluation / coating degradation model 12/98 C o
Phase il test program to validate model and test key parameters 10/99 T o
o Evaluate need for rugulatory action based on research results 1/00 T PART IV:
GSI 191," ASSESSMENT OF DEBRIS ACCUMULATION ON PRESSURIZED WATER REACTOR SUMP PERFORMANCE" 1.
NRC-sponsored research program on the potential for loss of ECCS NPSH during a LOCA due to clogging by debris o Preliminary (qualitative) risk assessment (NRR) 3/99 C o Complete collection of plant data to support research program 6/99 C o Integrate industry activities into this Action Plan 9/99 T l
o Complete research program on PWR sump blockage (including final risk 12/01 T l
assessment) l o Evaluate need for regulatory action based on research program results 3/02 T l
(NRR)
==
Description:==
This action plan has been prepared to comprehensively address the adequacy of ECCS suction design, and to ensure adequate ECCS pump NPSH during a LOCA. Specifically, the concern is whether debris could clog ECCS suction strainers or sump screens during an accident and prevent the 39
ECCS from performing its safety function. The plan will be risk informed. For PWRs, a detailed risk assessment will be conducted when sufficient information is gathered to perform an assessment of the potential for clogging the ECCS sump screens. A preliminary risk assessment has been performed by the staff and the recults are discussed below under the Regulatory Assessment. For BWRs, a risk assessment was performed as part of the development of NRC Bulletin (NRCB) 96-03, " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors" dated May 6,1996. This risk assessment formed part of the basis for issuing NRCB 96-03.
This plan has four parts. First, for boiling-water reactors (BWRs), this issue has been addressed by licensee responses to NRCB 96-03. The staff is currently confirming the adequacy of the licensee solutions implemented in response to the bulletin; therefore, the staff's confirmatory effort is included in this action plan for completeness. Second, the adequacy of licensee (both PWR and BWR) net positive suction head (NPSH) calculations is being evaluated through NRR review of licensee responses to Generic Letter (GL) 97-04, " Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps," dated October 7,1997. The third part of the plan consists of two efforts by the staff. The first effort assesses the adequacy of the implementation ar'd maintenance of current licensee coating programs through NRR review of licensee responses to GL 98-04," Potential for Degradation of the Eme,gency Core Cooling System and the Containment Spray System after a Loss-of-coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment," dated July 14,1998. The second effort is a research program to assess the potential for coatings to become debris, including the timing of any failures that might occur, and the cause and the characteristics of the debris. These two efforts combined will provide NRR the necessary technical bases on which to assess the potential threat to the ECCS by coating debris and the adequacy of current coating licensing bases (both PWR and BWR). The results of these two programs will also feed into the fourth part of the action plan: an evaluation of the potential for clogging of PWR ECCS recirculation sumps during a LOCA. As with the coating research discussed above, this part of the plan is being conducted by the Office of Nuclear Regulatory Research (RES). RES is evaluating the potential for pressurized-water reactor (PWR) sumps to become clogged during an accident based on new information teamed during the development of NRCB 96-03 for the BWRs.
Historical Backaround: During licensing of most domestic power plants, consideration of the potential for loss of adequate net positive suction head (NPSH) due to blockage of the ECCS suction by debris generated during a loss-of-coolant accident (LOCA) was inadequately addressed by both the NRC and licensees. The staff first addressed ECCS clogging issues in detail during its review of Unresolved Safety Issue (USI) A-43, " Containment Emergency Sump Performance." The NRC staff's concerns related to the potential loss of post-LOCA recirculation capability due to insulation debns were discussed in Generic Letter (GL) 85-22, " Potential for Loss of Post-LOCA Recirculation Capability due to insulation Debris Blockage," dated December 3,1985. This generic letter documented the NRC's resolution of USl A-43.
The staff concluded at that time that no new requirements would be imposed on licensees; however, the staff did recommend that Regulatory Guide 1.82, Revision 1, " Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," be used as guidance for the conduct of 10 CFR 50.59 reviews dealing with change out and/or modification of thermal insulation installed on I
primary coolant system piping and components. NUREG-0897, Revision 1,' Containment Emergency j
Sump Performance" (October 1985), contained technical findings related to USI A-43, and was the principal reference for developing the revised regulatory guide.
l Since the resolution of USl A-43, new information has arisen which challenged the adequacy of the NRC's conclusion that no new requirements were needed to prevent clogging of ECCS strainers in boiling-water reactors (BWRs). On July 28,1992, an event occurred at Barseback Unit 2, a Swedish boiling-water reactor (BWR), which involved the plugging of two containment vessel spray system I
(CVSS) suction strainers. The strainers were plugged by mineral wool insulation that had been dislodged by steam from a pilot-operated relief valve that spuriously opened while the reactor was at 435 psig. Two j
of the three strainers on the suction side of the CVSS pumps that were in service became partially
]
l 6
I 40
plugged with mineral wool. Following an indication of high differential pressure acrose both suction strainers 70 minutes into the event, the operators shut down the CVSS pumps and backflushed the stralners. The Barseback event demonstrated that the potential exists for a pipe break to generate insulation debris and transport a sufficient amount of the debris to the suppression pool to clog the ECCS strainers.
Similarly, on January 16 and April 14,1993, two events involving the clogging of ECCS strainers occurred at the Perry Nuclear Power Plant, a domestic BWR. In the first Perry event, the suction strainers for the residual heat removal (RHR) pumps became clogged by debris in the suppression pool. The second Perry event involved the deposition of filter fibers on these strainers. The debris consisted of glass fibers from temporary drywell cooling unit filters that had been inadvertently dropped into the suppression pool, and corrosion products that had been filtered from the pool by the glass fibers which accumulated on the surfaces of the strainers. The Perry events demonstrated the deleterious effects on strainer pressure drop caused by the filtering of suppression pool particulates (corrosion products or " sludge") by fibrous materials adhering to the ECCS strainer surfaces. This sludge is typically present in varying quantities in domestic BWRs, since it is generated during normal operation. The amount of sludge present in the pool depends on the frequency of pool cleaning /desludging conducted by the licensee. The effect of particulste filtering on head loss had been previously unrecognized and therefore its effect on PWRs had not been previously considered.
On September 11,1995, Limerick Unit 1 was being operated at 100-percent power when control room personnel observed alarms and other indications that one safety relief valve (SRV) was open. Attempts by the reactor operators to close the valve were unsuccessful, and a manual reactor scram was initiated.
Prior to the opening of the SRV, the licensee had been running the "A" loop of suppression pool cooling to remove heat being released into the pool by leaking SRVs. Shortly after the manual scram, and with the SRV still open, the "B" loop of suppression pool cooling was started. The reactor operators continued their attempts to close the SRV and reduce the cooldown rate of the reactor vessel. Approximately 30 minutes later, operators observed fluctuating motor current and flow on the "A" loop of suppression pool cooling. Cavitation was believed to be the cause, and the loop was secured. After it was checked, the "A" pump was successfully restarted and no further problems were observed. After the cooldown following the blowdown event, the licensee sent a diver into the Unit 1 suppression pool to inspect the condition of the strainers and the general cleanliness of the pool. The diver found that both suction strainers in the "A" loop of suppression pool cooling were almost entirely covered with a thin " mat" of material, consisting mostly of fibers and sludge. The "B" loop suction strainers had a similar covering, but less of it. Analysis showed that the sludge primarily consisted of iron oxides and the fibers were polymeric in nature. The source of the fibers was not positively identified, but the licensee determined that the fibers did not originate within the suppression pool, and contained no trace of either fiberglass or asbestos. This event at Limerick demonstrated the importance of foreign material exclusion (FME) practices to ensure adequate suppression pool and containment cleanliness. In addition, it re-emphasized that materials other than fibrous insulation could clog strainers.
NRCB 96-03," Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors," was issued on May 6,1996, requesting BWR licensees to implement appropriate procedural measures and plant modifications to minimize the potential for clogging of ECCS suction strainers by debris generated during a LOCA. Regulatory Guide 1.82, Revision 2, (RG 1.82), " Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," was issued in May 1996 to provide non-prescriptive guidance on performing plant-specific analyses to evaluate the ability of the ECCS to provide long-term cooling consistent with the requirements of 10CFR50.46. On November 20,1996, the Boiling Water Reactor Owners Group (BWROG) submitted NEDO-32686, " Utility Resolution Guidance for ECCS Suction Strainer Blockage" (also known as the URG) to the staff for review. The purpose of the URG is to give BWR licensees detailed guidance for complying with the requested actions of NRCB 96-03. The staff approved the URG in a safety evaluation report (SER) dated August 20,1998.
In response to NRCB 96-03, all affected BWR licensees have installed, or will install during their upcoming refueling outage, new large-capacity passive strainers. As noted above, the staff is presently confirming the adequacy of licensee resolutions implemented in response to NRCB 96-03.
41
RES has begun an evaluation of the potential for PWRs to lose NPSH due to clogging of ECCS sump screens by debris during an accident because of new information learned during the development of NRCB 96-03. As noted above, the effect of filtering of particulates on head loss across the sump screen had previously been unrecognized. In addition, it was also learned that more debris could be generated than was previously assumed, and that the debris would be significantly smaller than was previously expected. With more and finer debris, the potential for clogging of the ECCS sump screen becomes greater leading to the need for the staff to evaluate the potential for clogging of PWR sumps. RES's evaluation will include a risk assessment.
Recent events at a number of plants have raised concems regarding potential for coatings to form debris i
during an accident which could clog an ECCS suction. Several cases have occurred where qualified j
coatings have defaminated during normal operating conditions. Typically, the root cause has been attributed to inadequate surface preparation. This led the staff to raise questions regarding the adequacy of licensee coating programs. The staff issued GL 98-04 to obtain necessary information from licensees to evaluate how they implement and maintain their coating programs. In addition, Rylatory Guide (RG) 1.54 has been revised and issued for comment as DG-1076 with the objective to u?
guidance for the selection, qualification, application, and maintenance of protective coatings in nuclear power plants to be
{
consistent with currently employed ASTM Standards. The endorsement of industry consensus standards is responsive to OMB Circular A 119 and the NRC's Strategic Plan. RES has also begun a research program aimed at providing sufficient technicalinformation regarding the failure of coatings to allow the staff to evaluate the potential for clogging of ECCS suctions by coating debris (or for coatings to contribute to ECCS suction clogging). The program will evaluate the failure modes of coatings, the likely causes, the characteristics (e.g., size, shape) of the debris, and the timing of when coatings would likely fail during an accident. This information will be used to evaluate the ability of the coating debris to transport to the ECCS suction screens or strainers during an accident and the ultimate effect on head loss. The conclusions from the coatings portion of this action plan will be utilized in both RES's assessment of PWR sump clogging and in the staff's confirmatory evaluation of BWR solutions to the strainer clogging issue.
1 Proposed Actions: This action plan is divided into four parallel efforts. The first effort is for the staff to complete its review of the resolution of NRCB 96-03. Most licensees installed their new strainers under 10 CFR 50.59, concludi g that installing the new strainer modification did not constitute an unreviewed safety question. Since the staff did not receive detailed responses from these licensees describing their resolutions, the staff will audit 4-6 plants to determine if any significant issues exist. If inspection effort is needed to evaluate licensee resolutions, a Temporary Instruction will be prepared. Otherwise, the issue will be closed based on the audit findings and the findings of the staff's review of coatings related issues (discussed below).
The second effort is for the staff to complete its review of GL 97-04 responses. This will ensure that there are acceptable methods utilized throughout the industry for evaluating NPSH margin. This is important to the ECCS clogging issue because the calculation of adequate NPSH is the ultimate success criteria for determining ability of the ECCS to provide the required flow needed to meet the criteria of 10 CFR 50.46.
The third effort involves the evaluation of coatings as a potential debris source. Concerns raised in this area are due to recent events where qualified coatings have failed during normal operation at a number of sites. The failure of qualified coatings during normal operation has led to two specific staff concems.
The first concem is whether the qualification of coatings is adequate to ensure that coatings do not pose a potential threat to the ECCS. Accordingly, the staff has begun a research effort led by RES to evaluate the potential for coatings to become debris during an accident N onsequently, become a threat to the ECCS performing its safety function. The second concern rs
. to the adequacy of licensee programs to apply and maintain coatings consistent with their licensing basss. This concern will be addressed by NRR staff through review of license responses to GL 98-04. The staff will evaluate licensee responses to GL 98-04 to determine if licensee coating programs (application and maintenance of protective coatings in containment) are adequate to meet their current licensing bases. This issue is applicable to BWRs and PWRs.
42
r-The fourth cffort involves an evaluation of PWR sumps based on new information learned during the development of the staff's resolution for NRCB 96-03. RES has begun a program to evaluate PWR sump designs and their susceptibility to blockage by debris. This evaluation will include a detailed risk assessment. Risk insights will be used to support any conclusions drawn relative to the need for licensees to address the potential for ECCS suction clogging. Support for the research program is also needed from the industry to provide RES with the necessary plant data so that RES can bound the problem to be evaluated. The Nuclear Energy Institute (NEI) has conducted a survey of PWR licensees to provide the information needed by RES. Additionally, the staff is still seeking at least two additional reference plants to be used in the research study. Specifically, the staff needs a reference Westinghouse plant with a large dry containment and a reference Babcock and Wilcox plant. The staff will also coordinate its work with industry to eliminate duplication of effort and to ensure effective utilization of resources.
Oriainatina Document: Not Applicable.
Reaulatorv Assessment: Title 10, Section 50.46 of the Code of FederalRegulations (10 CFR 50.46) requires that licensees design their ECCS systems to meet five criteria, one of which is to provide long-term cooling capability following a successful system initiation for a sufficient duration so that the core temperature is maintained at an acceptably low value and decay heat is removed for the extended period of time required by the long-lived radioactivity remaining in the core. The ECCS is designed to meet this criterion, assuming the worst single failure.
However, for BWRs, experience gained from operating events and detailed analyses (including a detailed risk assessment) demonstrated that excessive buildup of debris from thermal insulation, corrosion products, and other particulates on ECCS pump strainers could occur during a LOCA. This created the potential for a common-cause failure of the ECCS, which could prevent the ECCS from providing long-term cooling following a LOCA. This led to the issuance of NRCB 96-03, and the subsequent installation of new larger strainers by BWR licensees.
The staff believes that there is sufficient new information and concerns raised relative to the potential for debris clogging in PWRs that part of this action plan has been prepared to address PWR sump blockage concems. However, it is not clear whether a significant threat to PWR ECCS operation exists. The staff believes that continued operation of PWRs is justified because of PWR design features which would tend to prevent blockage of the ECCS sumps during a LOCA. These features would tend to be effective for insulation and coating debris. For instance, the containments in PWRs tend to be very compartmentalized making the transport of debris to the sump screens difficult, in addition, PWRs typically do not need to switchover to recirculation from the sump during a LOCA until 20-30 minutes after the accident initiation allowing time for much of the debris to settle in other places within the containment.
Coating debris, in particular, would have pier,ty of time to settle. Clearly, the results of the staff's research program are needed before a final conclusion regarding the potential to clog the ECCS sump can be reached. In addition to these design considerations, the staff considers continued operation of PWRs to be justified because the probability of the initiating event (i.e., large break LOCA) is extremely low. More probable (although still low probability) LOCAs (small, intermediate) will require less ECCS flow, take more time to use up the water inventory in the refueling water storage tank (RWST), and in some cases I
may not even require the use of recirculation from the ECCS sump because the flow through the break would be small enough that the operator will have sufficient time to safely shut the plant down. In l
addition, all PWRs have received approval by the staff for leak-before-break (LBB) credit on their largest l
RCS primary coolant piping. While LBB is not acceptable for demonstrating compliance with 10 CFR 50.46, it does demonstrate that LBB-qualified piping is of sufficient toughness that it will most likely leak (even under safe shutdown earthquake conditions) rather than rupture. This, in turn, would allow operators adequate opportunity to shut the plant down safely (although debris generation and j
transport for an LBB size through-wall flow will still be investigated). Additionally, the staff notes that there are sources of margin in PWR designs which may not be credited in the licensing basis for each plant. For instance, NPSH analyses for most PWRs do not credit containment overpressure (which would likely be present during a LOCA). Any containment pressure greater than assumed in the NPSH analysis provides additional margin for ECCS operability during an accident. Another example of margin 43
would be that it has been shown, in many cases, that ECCS pumps would be able to continue operating for some period of time under cavitation conditions. Some licensees have vendor data demonstrating this. Design margins such as these examples may prevent complete loss of ECCS recirculation flow or increase the time available for operator action (e.g., refilling the RWST) prior to loss of flow.
GL 97-04 is a review of NPSH calculations. No specific generic concerns have been identified in the review of licensee responses. Any issues identified with an individual licensee's response will be handled on a case by case basis.
The Probabilistic Safety Assessment Branch of NRR recently completed a preliminary assessment of the risk associated with the potential clogging of the ECCS sump in PWRe during a LOCA. In a memo from Richard J. Barrett to John N. Hannon dated March 26,1999, it was concluded that "(d)ue to the J
unavailability of probabilistic models for debris-induced loss of ECCS NPSH and the plant-specific nature
(
of the sump screen clogging issue, the scope of this risk assessment was limited to assessing the frequency of accident sequences requiring ECCS recirculation to prevent core damage for an average PWR plant. Because the probability and timing of sump screen clogging depends on LOCA size and location, among other parameters, an effort was made to present the results, for each LOCA category, I
separately.
The following major conclusions were reached by performing this preliminary risk assessment.
Results presented in this analysis strongly justify research to re-evaluate the potential for clogging o
of PWR sump screens by taking into account new information, thus enabling more realistic evaluation and management of associated risks.
Continued operation of PWRs is justified because, based on available current information, there o
is no evidence that the risk associated with the sump clogging issue is high enough to compromise public health and safety."
These conclusions clearly support this action plan as outlined herein.
Current Status: The review of NRCB 96-03 responses is nearly complete. The staff has completed two
)
audits and is currently preparing audit reports for Dresden and Limerick. The responses to GL 98-04 are 4
currently under review, and RES has begun Phase 2 of its coatings research program. RES's PWR sump study has begun, and NEl is compiling the results of its industry survey for submittal to the staff.
NRR review of GL 97-04 responses is about 95% complete. The remaining issue for those reviews still open is the use of some containment overpressure to demonstrate adequate net positive suction head (NPSH) of the ECCS and containment spray pumps. Since the probability of the initiating event is low, 1
and the conditional probability of a major loss of containment integrity (which would be necessary to lose containment overpressure) is also low, the staff finds that the short time remaining to complete the GL 97-04 reviews is acceptable, j
Generic Letter (GL) 98-04 is scheduled to be complete this calendar year (1999). Available evidence from limited industry tests of the transport of coating debris indicates that coating debris may not transport very well under conditions approximating those of containment sump flow. In fact, very small amounts of debris actually reached the screens in these tests. This consideration, in addition to the low probability of the initiating event and the difficulty of transporting the debris to the sump given the circuitous geometry of a containment flow path, leads to the conclusion that the target date for completing the reviews of GL 98-04 is acceptable.
44
Rob Elliott, SPLB,415-1397 NRR Technical Contacts:
Jim Davis, EMCB,41S-2713 Rich Lobel, SPLB,415-2865 Jack Kudrick, SPSB,415-2871 Tony D'Angelo, SPLB,415-2857 Kerri Kavanagh, SRXB,415-3743 Larry Campbell, HOMB,415-2976 Nicholas Saltos, SPSB,415-1072 RES Technical Contacts:
Michael Marshall, ERAB,415-5895 Al Serkiz, ERAB,415-6563
References:
- Net Positive Suction Head for Emergency Core Cooling ar.d Containment Heat Removal System Pumps"(Safety Guide 1), dated November 1970.
Regulatory Guide 1.54," Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants" (Draft DG-1076, Proposed Revision 1, publisned March 1999),
dated June 1973.
NRC Bulletin 93-02," Debris Plugging of Emergency Core cooling Suction Strainers," dated May 11,1993.
NRC Bulletin 93-02, Supplement 1," Debris Plugging of Emergency Core Cooling Suction Strainers,"
dated February 18,1994 NUREG/CR-6224," Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris" dated October 1995.
NRC Bulletin 95-02, " Unexpected Clogging of Residual Heat Removal (RHR) Pump Strainer While Operating in Suppression Pool Cooling Mode," dated October 17,1995.
NRC Bulletin 96-03," Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors" dated May 6,1996.
Regulatory Guide 1.82, Revision 2," Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," dated May 1996.
l GL 97-04," Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps," dated October 7,1997.
GL 98-04," Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-coolant Accident Because of Construction and Protective Coating Deficiencies I
and Foreign Material in Containment," dated July 14,1998.
Memorandum from Richard J. Barrett to John N. Hannon," Preliminary Risk Assessment of PWR Sump Screen Blockage issue," dated March 26,1999.
I 45 i
NEW SOURCE TERM FOR OPERATING REACTORS TAC No, M89586 Last Update: 07/13/99 GSI No.155.1 Lead NRR Division: DSSA CTL 1.1 Supporting Division: DE MILESTONES DATE (T/C) 1.
NEl Letter 07/94C 2.
Commission Memo 09/94C 3.
NEl Response 09/94C 4.
NEl/NRC Meeting 10/94C 5.
Publicati_on of NUREG-1465 02/95C 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 I
10.
Present Commission Paper in E-Team Briefing 09/96C 11.
Brief CRGR on Commission Paper 10/960 12.
Send Commission Paper to EDO/ Commission 11/96C 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 06/98C 21.
Finish First Pilot Plant Review (Perry) 02/99C 22.
Finish Second Pilot Plant Review (Grand Gulf) 09/99T 23.
Finish Third Pilot Plant Review (Indian Point Unit 2) 12/99T 24.
Finish Fourth and Fifth Pilot Plant Review TBD 46
[
y Descnotion: 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 could result in cost savings without sacrificing real safety margin. In addition, safety enhancements may also be achieved.
Histoncal Backaround ' in 1962, the U.S. Atomic Energy Commission published TID-14844, " Calculation of Distance Factors for Power and Test Reactors." Since then licensees and the NRC have used the accident source term presented in TlD-14844 in the evaluation of the dose consequences of design basis
- accidents (DBA).
~ After examining years of additional 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 evaluations, 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 description of release timing and composition. Since the NUREG source term results in lower calculated DBA dose 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 dose evaluations in support of plant, technical specification, and procedure modifications.
NRC and NEl met several times to discuss the industry's plans to use the new source term. To make efficient use of NRC's review resources, NRC encouraged the industry to approach the issue on a generic basis. The Nuclear Energy Institute (NEI) unveiled its plans for the use of the new source term at operating plants at the Regulatory Information Conference in May 1995. NEl, Polestar (EPRI's consultant), and pilot plant (Grand Gulf, Beaver Valley, Browns Ferry, Perry, and Indian Point) representatives met with NRC staff in June and October 1995 to discuss more detailed plans.
Proposed Actions The staff will continue working with industry and complete its review of the pilot plant applications on an expedited basis. The knowledge gained from the pilot plant application review will be used in developing the associated regulatory guide and standard review plan that will be part of the final rulemakMg for the altemative source term.
Oriainatina Document: EPRI Technical Report TR-105909, " Generic Framework Document for Application of Revised Accident Source Term to Operating Plants," transmitted by letter dated November 15,1995.
Reaulatory Assessment: There will be no mandatory backfit of the new source term for operating reactors. The design-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: NEl submitted its generic framework document in November 1995 for NRC review and approval. TVA subm!tted part of its pilot plant application for Browns Ferry in December 1995. The staff met with NEl on January 23,1996, to discuss 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 staff's plan to issue exemptions while pursuing rulemaking, and on April 2,1997, to provide a status report on the staff's actions regarding rebaselining and rulemaking subsequent to the Commission's SRM. The pilot plant applications for Browns Ferry, Perry, Indian Point, and Oyster Creek have been circulated to the task force members to help shape rebaselining. In June 1997. 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 TID-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 initiated prior to the completion of rebaselining.
47
The staff briefed the NRR Executive Team on SECY-96-242 in September 1996, the CRGR in October 1996, and the ACRS full committee in November 1996. A limited number of pilot plants
=submittals and exemptions are expected - four submittals have been received so far (Browns Ferry, Per y, Oyster Creek, and Indian Point-2). An application is also expected from Grrnd Gulf. In addition, the staff 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 Commission approved the Option 2 approach of SECY-96-242 and a modification 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 activities involved with rebaselining; 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 Commission inquiries regarding the deferral of the completion of rebaselining until 4
November 1998, NRR and RES discussions and shifts in lead technical responsibility resulted in an i
improvement in the schedule. At a Commissioners' Technical Assistants briefing on October 9,1997, the Task Force Leader outlined a new schedule that would result in the completion of rebaselining and the rulemaking plan in June 1998; this was accomplished by reversing the lead responsibilities (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.
)
NRR is working closely with RES to transfer technical insights gained on robaselining. In addition, NRR j
transferred its technical assistance resources with SNL, ORNL, and PNNL that were designated for
{
rebaselining to RES. These changes w!lI be reflected in the next revision to the NRR Operating Plan. On i
November 13,1997, January 7,1998, February 24,1998, and March 30,1998, RES presented its four-f phased plan and preliminary findings from Phase 1, Phase !!, and the DBA portion (with the updated assumptions) of Phase Ill, respectively, for the rebaselining effort. On April 1 and 2,1998, RES and NRR staff briefed the ACRS and DONRR, respectively, on the progress ot 'he rebaselining effort, initial insights i
from the assessments completed, and the essential elements of the Rulemaking Plan. The results of the rebaselining effort were reported in SECY-98-154 dated June 30,1998. The Rulemaking Plan was provided in SECY-98-158 dated June 30,1998. SRM on SECY-98-158 issued 9/4/98. See rulemaking entryin Attachment 2.
The staff completed its review of the first pilot plant application in February 1999 and issued a safety evaluation and its license amendment for Perry plant in March 1999. The staff is currently reviewing Grand Gulf and Indian Point Unit 2 pilot plant applications and the staff expects to complete its review by September 1999 for Grand Gulf and December 1999 for Indian Point Unit 2. Browns Ferry and Oyster Creek requested to have their pilot plant application reviews "on hold."
NRR Technical
Contact:
J. Lee, SPSB,415-1080
References:
NUREG-45, " 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".
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 Commission from NRC staff,"Use of NUREG-1465 Source Term at Operating Reactors".
48
I December 22,1995, pilot plant submittal, letter to Document Control Desk f rom Tennessee Valley Authority, " Brown's Ferry Nuclear Plant (BFN) - Units 1,2, and 3 - Technical Specifications (TS) No. 356 and Cost Beneficial Licensing Action (CBLA) 08 - Increase in Allowable Main Steam isolation Valve (MSIV) Leakage Rate and Request for Exemption from 10 CFR 00, Appendix J... and 10 CFR 100, Appendix A...".
August 9,1996, memorandum to the Commission 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 Commission from NRC staff,"Results of the Revised (NUREG-1465)
Source Term Rebaselining for Operating Reactors," SECY-98-154.
Summaries of public meetings:
dated November 10,1994, for public meeting with NEl held on October 6,1994;.
e dated July 26,1995, for public meeting 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; e
dated February 27,1996, for public meeting with Browns Ferry held on February 7,1996; e
dated September 27,1996, for public meeting with Grand Gulf held on August 29,1996; e
dated October 11,1996, for public meeting with NEl held on October 2,1996; e
dated January 24,1997, for public meeting with Surry held on November 26,1996;
{
e dated April 24,1997, for public meeting with PWR (Surry) held on March 25,1997, i
e dated April 24,1997, for public meeting with BWR (Grand Gulf) held on March 27,1997; e
dated May 8,1997, for public meeting with NEl held on April 2,1997; I
e dated July 28,1997, for public meeting with PWR (Surry) held on June 18,1997.
e I
49
ATTACHMENT 2 i
GENERIC COMMUNICATION AND COMPLIANCE ACTIVITIES
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DISTRIBUTION f:r NRR Director's Quirt:rly St:tus R: port
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./PDR RGEB R/F FJMireglin, EDO SJCollins, NRR RZimm:rm:n, NRR WFKina, NRR BWSharon, NRR DBMatthews, NRR CACarpenter, NRR FMAkstulewicz, NRR EMMcKenna, NRR WFBurton, NRR EYWang, NRR BJSweeney, NRR BABoger, NRR JAZwolinski, NRR GMHolahan, NRR JRStrosnider, NRR JESilber, NRR RCEmrit, RES Regional Administrators 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 i 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.
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Germantown, MD 20874 Mr. S. Scott Office of Nuclear Safety, DOE
' Century 21 Building (E-H72) 19901 Germantown Road Germantown, MD 20874-1290 Mr. Bob Borsum cf(f S4 1700 Rockville Pike, Suite 525 Rockville, MD 20852 Ms. Norena G. Robinson, Licensing Technician Nebraska Public Power District Cooper Nuclear Station P.O. Box 98
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