ML20196H629

Director Status Report on Generic Activities Action Plans Generic Communication & Compliance Activities April 1999

ML20196H629
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Issue date:	04/30/1999
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NUDOCS 9907060414
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DIRECTOR'S STATUS REPORT on GENERIC ACTIVITIES I

Action Plans Generic Communication and Compliance Activities APRIL 1999 /

p ps-Office of Nuclear Reactor Regulation 9907060414 PDR ORG 99o43o .

NRRA 1

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INTRODUCTION l The purpose of this report is to provide information about generic activities, including generic l 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.

Attachment 1, "NRR Action Plans," includes generic or potentially generic issues of sufficient complexity or scope that require substantial NRC staff resources. The issues covered by action plans include concerns 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.

Attachment 2, " 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.

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umi niuUTION for NRR Dircctor's Quart:rly Status 3: port

/C ntralFils PDR RGEB R/F FJMiraglia, EDO i SJCollins, NRR RZimm:rmin, NRR WFKana, NRR BWSh:ron, NRR J 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, SER Manager

& Chief Nuclear Offi r Bechtel Power Corpo Nuclear Energy I stute 9801 Washington Ivd.

1776 i Stree Gaithersburg, M land 20878-5356 Suite 4 Washi ton, D.C. 20006-3708 Mr. R. P. LaRhette institute of Nucle Power Operations 700 Galleria P way Atlanta, G gia 30339-5979 Mr. Lee Watkins Assistant Manager For High Level Waste U.S. DOE P.O. Box A I Aiken, South arolina 29892 Mr. R. W. Barber Safety and Quali ssurance, DOE 270 Corporat enter (E-853) {

20300 Cent Blvd.

Germantown, MD 20874 Mr. S. Scott j Office of Nuclear Safety, DOE Century 21 Building (E-H72) 9gil(

19901 Germantown Roajr Germantown, MD 20874-1290 Mr. Bob Borsum 1700 Rockville Pike,jii0ite 525 Rockville, MD 208ji2 Ms. Norena G. Robinson, Licensing Technician ,

Nebraska Public Po r District - .. , ,g '

Cooper Nuclear St on \ ~

P.O. Box 98 ,

Brownsville, 68321 i DOCUMENTNAME: DIST. PUB OFFCE ROEB DRIP DRP ' REXB.DRP ROEB. DRIP RGEB:DRP D DRPA --"

j RGEB.DQP j NAME S$ureensybeh EWen0 WBurton h EMcKenne ((/[1 ketulowEr , CCar /

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4 TABLE OF CONTENTS l

BOILING WATER REACTOR INTERNALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 PRAiMPLEMENTATION ACTION PLAN 1.2 (c) Inservice Inspection Action Plan..............................................................6 STE AM G E N E R ATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 MEDIUM-VOLTAGE & LOW-VOLTAGE BREAKER FAILURES . . . . . . . . . . . . . . 15 ENVIRONMENTAL SRP REVISION ACTION PLAN . . . . . . . . . . . . . . . . . . . . . . . . 17 EMERGENCY ACTION LEVEL GUIDANCE DEVELOPMENT . . . . . . . . . . . . . . . . 19 PRA IMPLEMENTATION PL AN 1.2(d) Graded Quality Assurance Action Plan . 22 ACCIDENT MANAGEMENT IMPLEMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . 30 CORE PERFORMANCE ACTION PLAN Final Update . . . . . . . . . . . . . . . . . . . . . . 34 HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN . . . 36 WOLF CREEK DRAINDOWN EVENT: ACTION PLAN . . . . . . . . . . . . . . . . . . . . . 39 NEW SOURCE TERM FOR OPERATING REACTORS . . . . . . . . . . . . . . . . . . . . . 41 1

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ATTACHMENT 1 NRR ACTION PLANS l

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BOILING WATER REACTOR INTERNALS TAC Nos. M91898, M93925, M93926, M94959, M94975, M95369, Last Update: 3/26/99 M96219, M96539, M97373, M97802, M97803, M97815, M98266, Lead NRR Division: DE M98708, M98880, M99638, M99870, M99894, M99897, M99898, Supporting Division: DSSA M99895, M99897, MA1102, MA1104, MA1138, MA1226, MA1926, GSI: Not Available

, MA1927, MA2326, MA2328, MA3395, MA3683, MA4203, MA4464, MA4465, MA4467, MA4468 MILESTONES DATE (T/C)

PART 1: REVIEW OF GENERIC INSPECTION AND EVALUATION CRITERIA

1. Issue summary NUREG 1544 03/96C o Update NUREG-1544 3O/FY99T j

2. Review BWRVIP Re-inspection and Evaluation Criteria I o Reactor Pressure Vessel and internals Examination Guidelines (BW RVI P-03) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 06/08/98Cl o BWRVIP-03, Section 6A, Standards for Visual inspection of Core Spray Piping, Spargers, and Associated Components . . . . . . . . . . . . . . . . . . . 06/08/98Cl !

o BWR Vessel Shell Weld inspection Recommendations (BWRVIP-05) . . 07/28/98CA '

o BWR Axial Shell Weld inspection Recommendations . . . . . . . . . . . . . . . 12/31/99T j

'o Guidelines for Reinspection of BWR Core Shrouds (BWRVIP-07) . . . . . 04/27/98CA I

3. - Review of generic repair technology, criteria, and guidance TBD

4. Review generic mitigation guidelines and criteria TBD

5. Review of generic NDE technologies developed for examinations of BWR internal components and attachments TBD i

6. Other Intemals reviews (safety assessments, evaluations, mitigation .

measures, inspections, and repairs)  !

o Safety Assessment of BWR Reactor internals (BWRVIP-06) . . . . . . . . 09/15/98CA o Bounding Assessment of BWR/2-6 Reactor Pressure Vesselintegrity issues (BWRVIP-08 & BWRVIP-46) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 03/27/98CA o Evaluation of Crack Growth in BWR Stainless Steel RPV Internals (BW RVI P- 14) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 06/08/98Cl o Intemal Core Spray Piping and Sparger Replacement Design Criteria (BW RVI P- 16) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11/16/98Cl o Roll / Expansion of Control Rod Drive and in-Core Instrument Penetrations in BWR Vessels (BWRVIP-17) . . . . . . . . . . . . . . . . . . . . . . 03/13/98CD o BWR Core Spray internals inspection and Flaw Evaluation Guidelines (BWRVIP-18) . . . . . . . . . . . . . . . .............................. 06/0e/98CA o BWRVIP-18, Appendix C, BWR Core Spray intemals Demonstration of Compliance With TechnicalInformation Requirements of License Renewal Rule (10 CFR 54.21) . . . . . . . . . . . . . . . . . . . . . . ...... TBD 1

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o Internal Core Spray Piping and Sparger Repair Design Criteria (BW RVI P- 19) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11/16/98CA o Core Plate Inspection and Flaw Evaluation Guideline (BWRVlP 25) . . . l 04/30/99T l o Top Guide Inspection and Flaw Evaluation Guideline (BWRVIP-26) . . . 04/30/99T l o Standby Liquid Control System / Core Plate AP inspection and Flaw l Evaluation Guidelines (BWRVIP-27) . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T l o Assessment of BWR Jet Pump Riser Elbow t<> Thermal Sleeve Weld i Cracking (BW RVIP-28) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T o Technical Basis for Part Circumferential Weld Overlay Repair of Vessel Internal Core Spray Piping (BWRVIP-34) . . . . . . . . . . . . . . . . . . . . . . . TBD o Shroud Support inspection and Flaw Evaluation Guidelines (BW RVI P-38) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T o BWR Jet Pump Assembly inspection and Flaw Evaluation Guidelines

( BW R VI P-41 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 06/30/99T o BWR LPCI Coupling inspection and Flaw Evaluation Guidelines (BW RVI P-4 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . 04/30/99T o Update of Bounding Assessment of BWR/2-6 Reactor Pressure Vessel ,

Integrity issues (BWRVIP-46) . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . 03/27/98CA I o BWR Lower Plenum inspection and Flaw Evaluation Guidelines (BW RVI P-47) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T ,

o Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines l (BW RVI P-48) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 03/21/99C o Instrument Penetration inspection and Flaw Evaluation Guidelines (BWRVIP-49 Fop Guide / Core Plate Repair Design Criteria (BWRVIP-50) 08/04/98 CA o Jet Pump Repair Design Criteria (BWRVIP-51) . . . . . . . . . . . . . . . . . . . . 12/30/99T o Shroud Support and Vessel Repair Design Criteria (BWRVIP-52) . . . . . 12/30/99T l o Standby Liquid Control Line Repair Design Criteria (BWRVIP-53) . . . . . 12/30/99T 12/30/99T o Lower Plenum Repair Design Criteria (BWRVIP-55) . . . . . . . . . . . . . . . . 12/30/99T o LPCI Coupling Repair Design Criteria (BWRVIP-56) . . . . . . . . . . . . . . . . 12/30/99T o instrument Penetrations Repair Design Criteria (BWRVIP-57) . . . . . . . . 12/30/99T o CRD Internal Access Weld Repair (BWRVIP-58) . . . . . . . . . . . . . . . . . . 12/30/99T o Evaluation of Crack Growth in BWR Nickel-Base Austenic Alloys in RPV Inte mals (BWRVIP-59) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12/30/99T o BWR Vessel and Internals induction Heating Stress Improvement Effectiveness on Crack Growth in Operating Plants (BWRVIP-60) . . . . . 12/30/99T o Technical Basis for inspection Relief for BWR Intemal Components with Hydrogen injection (BWRVIP-62) . . . . . . . . . . .. ...... .... .. 12/30/99T 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 internals. This includes plant specific reviews and the assessment of the generic crihria that have been proposed by the BWR Owners Group and the BWRVIP technical subcommittees address IGSCC in core shrouds and other BWR internals.

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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 intemals cracking. In July 1994, the NRC issued Generic Letter 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 (Boiling Water Reactor Vessels and Intemals Project - BWRVIP) was formed to focus on resolution of reactor vessel and intemals degradation. This group was instrumental in facilitating licensee responses to NRC's Generic Letter. The NRC evaluated the review group's reports, submitted in 1994 and early 1995, and all plant responses.

All of the plants evaluated have been able to demonstrate continued safe operation untilinspection 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.

~ Proposed Actions: The staff will continue to assess the scopes that have yet to be submitted by licensees conceming inspections or re-inspections of their core shrouds. The staff will also continue to assess core shroud reinspection results and any appropriate core shroud repair designs on a case-by-case basis. The staff will issue separate safety evaluations regarding the acceptability of core shroud reinspection results and core shroud repair designs. The staff has been interacting with the BWRVIP and individual 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 internals.

The BWRVIP has submitted 29 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.

Oriainatino Document: Generic Letter 94-03, issued July 25,1994, which requested BWR licensees to inspect their core shrouds by the next outage and to justify continued safe operation until inspections can be completed.

I Reaulatory Assessment: In July 1994, the NRC issued Generic Letter 94-03 which required licensees to inspect their shrouds and provide an analysis justifying continued operation until inspections could be performed. The staff has concluded in all cases that licensees have provided sufficient evidence to  ;

support continued operation of their BWR units to the refueling outages in which shroud inspections or  !

repairs have been scheduled. In addition, in October 1995, industry's special review group submitted a safety assessment of postulated cracking in all BWR reactor internals and attachments to assure continuing safe operation.

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Current Status Almost all BWRs completed inspections or repairs of core shroud' s during refueling outages in the fall of 1995. Various repair methods have been used to provide altemate load carrying capability, including preemptive repairs, installation of a series of clamps and use of a series of tie-rod L

assemblies. The NRC has reviewed and approved all shroud modification proposals that have been submitted b'y BWR licensees. Review by NRC continues on individual plant reinspection results and plant-specific assessments. .

! In' October 1995, industry's spbcial review group issued a report '(BWRVIP-06) which the NRC staff's preliminary review indicated was not comprehensive. The NRC staff requested additional information 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 1996f 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 significance of and recommende6?.spections for top guide and core plate ring cracking. Technical review of the " Reactor Pressure Vessel a~ nd intemals Examination Guidelines (BWRVIP-03)" is complete and the staff's SE has been issued.

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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 inform 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 for licensees 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 review of BWRVIP-14 and -18 is complete and the staff's SEs have been issued. The staff's i review of BWRVIP-16 and -19 on intamal core spray piping inspection and flaw evaluation and repair j design criteria, respectively, is complete.

By letter dated December 20,1996, the BWRVIP submitted Appendix C to BWRVIP-18. This appendix addresses the use of BWRVIP generic intamal core spray inspection guidelines for compliance with requirements of the license renewal rule (10 CFR Part 54). The staff is reviewing this appendix in  !

conjunction with its review of BWRVIP-18 guidelines.  !

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!. The BWRVIP submitted BWRVIP-28 to address the safety implications of recent cracking found in BWR l~ ' 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 l

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 l to ensure that the vertical core shroud welds, and the core shroud repair, is adequately inspected.

By letters dated April 25 and May 30,1997, the BWRVIP provided a reaffirmation of the BWR member licensees to the BWRVIP, and committed, on behalf of 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 Contactsi . Keith Wichman, EMCB,415-2757 Kerri Kavanagh, SRXB,415-3743.

Kamal Manoly, EMEB,415-2765 NRR Lead PM: C. E. Carpenter, EMCB,415-2169

References:

1 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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i PRAiMPLEMENTATION ACTION PLAN 1.2(c)

Inservice Inspection Action Plan l

TAC Nos. M95125, M97153, Last Update: 3/26/99 M99389, M99756, MA0125, Lead NRR Division: DE MA0867, MA0868 Support Division: DSSA, EMCB 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 l

7. Office Director /OGC concurrence 07/08/970

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 j

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 j

15. SECY from EDO to Commissioners (SECY-97-190) 08/20/97C l

16. Publish draft for : ^1ic comments 10/15/97C

17. Public comment t < . lod for draft RG/SRP ends 01/13/980 l

18. Public Workshop 11/20/97C

19. ComplC.e draft for ACRS/CRGR review / comments 04/98C l

20. Complete draft for Inter-Office concurrence 05/98C l

21. Issue RG/SRP for trial use by the staff 06/98C WOG TOPICAL REPORT MILESTONES DATE (T/C) i

1. Technical Meeting 9/22/98C 2c WOG Commitment Letter 9/30/98C

3. Issue FSER 12/15/98C 6

m 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 4/15/99T

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

3. Issue FSER ANO-2 12/29/98C l

' Clarification of some aspects of the program are stillin progress.

INSPECTION PROCEDURES MILESTONES DATE (T/C)

1. Issue Draft Inspection Procedure Number 73753 6/98C

2. Issue Final Inspection Procedure Number 73753 6/98C

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 acceptability of the pilot programs with respect to their RI ISI applications and prepare plant specific safety evaluation reports (SER). The action plan describes the process for the review of RI-ISI 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.

Historical Backaround: On August 16,1995, the U.S. Nuclear Regulatory Commission (NRC) published a policy statement (60FR42622) on the use of probabilistic risk assessment (PRA) methods in nuclear regulatory activities. In the statement, the Commission stated its belief that the use of PRA technology ,

in NRC regulatory activities should be increased to the extent supported by the state-of-the-art in PRA I methods and data and in a manner that complements the NRC's deterministic approach. in a l 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 l 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 des eloped 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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ASME is working on three Code Cases for alternate examination requirements to ASME Section XI, Division 1 for piping welds. Code Case N-577 is based on the WOG methodology and Code Cases c 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 ISI. 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 irom 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 l parallel effort is being 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 plant for the WOG methodology is Surry 1 and pilot plants for the EPRI methodology are Vermont i Yankee and ANO-2.

l The acceptability of the RI-ISI pilot plant programs has been documented in SERs for each of the pilot i plant licensees and issued to the pilot plant licensees to allow use of the RI-ISI methodology. j ASME is working on three Code Cases for alternate examination requirements to ASME Section XI, Division 1 for piping welds. Codb Uase N-560, for the alternate examination requirements for Class 1, l 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 intends to review the WOG methodology as well as the Code Case N 577 and the consistency of the Surry 1 pilot program for RI-ISI to both of these. The Code Case N-577 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 methodology is that the scope of the Code Case is limited to ASME Class 1,2, and 3 systems while the EPRI methodology may encompass 8

+ ,

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 L 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 altamate examination 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 will review 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. ,

The staff utilized the acceptable alternative provision of 10 CFR 50,55a (a)(3)(i) to approve the pilot plants' applications.~ The staff is working closely with ASME to expedite changes involving ISI.

" Lono-Ranae Plan ' . , ,

This action plan will b monitored up to and including the comple:lon 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 seek relief from NRC to the current 50.55a requirements. A minimal review cycle is expected for the

.' approval of RI-ISI submittais during this time frame.

it is anticipated that subsequent to the issuance of safety evaluation reports (SER) for the pilot plants and the topical reports, the industry will revise the ASME Code Cases to incorporate lessons leamed

. from pilot plants and topical report reviews. iThe 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 RI-ISI methodology has been endorsed in RG 1.147 (and endorsed in L10 CFR 50.55a, as necessary), the staff anticipates that the licensee will develop an Al-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.

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.

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_ Oriainatina Documentsi Inl tbvember 30,1995, me orandum to J.' M. Taylor, the NRC EDO,

. Chairman Jackson directed that the staff prepare an action plan together with a timetable for developing RGs and SRPs applicable to use of PRAs to be completed in'two years in his response of January 3, 1996, the EDO presented a plan that established milestones for the development of regulatory guidance documents for 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.

f Reaulatory Assessment: The operational readiness and functional integrity of certain safety-related 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.

Licensees are currently interested in optimizing inspection by applying resources in more safety- 1

-- significant areas. They are also interested in maintaining system availability and reducing overall l

maintenance costs in ways that do not have an adverse effect on safety.

- On a parallei path, ASME is developing Code' Cases for altemate examination requirements to the I current ASME Section XI selection and inspection requirements. These Code Cases utilize procedures that are based on the relative risk significance of piping locations within individual systems.

The NRC is using probabilistic methods, as an adjunct to deterministic techniques, to help define the scope, type, and frequency of 151. 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 Al-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-98-139) for information, The RG and SRP were issued in the Federal Register in October 1998. j The staff completed its review of the Westinghouse Owners Group (WOG) methodology documented in i 1

WCAP-14572, Rev 1, and issued its safety evaluation report (SER) on December 15,1998. The staff

- comply'ed its review of the Vermont Yankee (RI ISI) pilot program and issued its safety evaluation report ]

(SER) on November 9,1998. The staff completed its review of the Surry Unit 1 (RI-ISI) pilot program l' and issued its safety evaluation report (SER) on December 16,1998. The staff completed its review of the ANO-2 (RI-lSI) 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 q 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 EPRI's responses to NRC's Request for Additional Information (RAl) related to the approach described in EPRI topical report, TP.-106706, Risk-  :

l: Informed inservice inspection Evaluation Procedure. Based on the discussion, EPRI plans to revise the i topical report by April 15,1999, to incorporate lessons leamed from the pilot applications (Vermont Yankee and ANO-2) of the methodglogy; methodology enhancements which have evolved since the 10  ;

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l June 1996 report was issued; and to provide further clarification and guidance as necessary, based on l NRC RAls. The staff's plan is to issue the draft SER by mid-June,1999, to support ACRS meetings, and complete the final SER by October 31,1999.

The NRC issued information Notice 93 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 j program for piping.

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The staff has also been actively participating in ASME Code activit%s related to RI-ISI.

NRR Contacts: S. 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 Staternent," August 16,1995.

2. Memorandum from Shirley Ann Jackson, Chairman, NRC to James M. Taylor, Executive Director i for Operations, " Follow-up Requests in Probabilistic Risk Assessment and Digital Instrumentation and Control," November 30,1995. l

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-106'706," 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.

I1

g 1 STEAM GENERATORS

' TAC Nos. M88885, M99432, MA4265 Last Update: 3/26/99 Lead Division: 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 I 4, Regulatory Analysis : 5/97(C) -

5. Proposed GL Pkg -

10/97(C) j 6.~ ACRS Endorsement ~ -

4~ 9 9/97(C) f 7.- CRGR Concurrence > '

On hold m

8. EDO .i

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 GL1 On hold

' Orig.' Publish Final Rule 12/95 Brief C;;;iiction: The NRC originally p' lanned to develop a rule pertaining to steam generator tube integrity. The proposed rule was to implement a more flexible regulatory framework for steam generator surveillance and maintenance activities that allows a degradation specific management approach. The results of the regulatory analysis suggested that the more optimal regulatory approach was to utilize a generic letter. The NRC staff suggested, and the Commission subsequently approved, a = vision to the regulatory approach to utilize a generic letter, in SECY-98-248, the staff 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\ecommendation. If sufficient progress is made with NEl in resolving technical and regulatory implementation issues, then the GL effort may be permanently halted.

< Mhis revision reflects the status of the proposed GL as "ori hold". The staff estimates that by mid-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.

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Reaulatory Assessment: The current regulatory framework providas reasonable assurance that .

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. operating PWRs are safe. The current regulatory framework that implements governing requirements l: ,  ; through the plant technical specifications can be improved. The staff is currently working with NEl to find r

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.~ GUDG-1074/TSs - September 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 L

- NEl submits NEl 97-06 " Steam Generator Program Guidelirns"- December 16,1997

. - CRGR package concurred on by NRR and sent to CRGR April 14,1998 '

- - Met with CRGR on June 12,1998, for inf armation briefing on package

- Met with CRGR on July 21,1998, for detailed 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 Isa,ued 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 . i September 1998 Hopenfeld memorandum to the Commission, for public comment  !

- The Commission, in SRM dated December 21,1998, agreed to above recommendations 1

- Held technical and management meetings with industry on 10/7/98,10/28-29/98,11/12/98,11/18/98, i 2/10/99, and 2/24/99 to resolve technical and regulatory implementation issues regarding NEl 97-06. i 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 j

, - Intemal guidance to SG inspectors was issued on 1/25/99 indicating that DG-1074 should not be used j for inspection guidance as directed by the Commission's SRM of 12/21/98.

4 Briefed ACRS Materials S/C on 3/24/99 regarding the status of current regulatory approach.

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l NRR Techr.ical Contacts: Ted Sullivan, EMCB,415-3266 Tim Reed, EMCB,415-1462 '

RES

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MEDIUM-VOLTAGE & LOW-VOLTAGE BREAKER FAILURES Last Update: 3/26/99 Lead Division: DE MILESTONES DATE (T/C)

1. Develop and issue a Ti and conduct inspection

a. Issue Temporary Instruction to inspect two sites for Regions I, ll, Ill, and IV (EElB) 12/31/97C l

b. Issue revision to Temporary Instruction (EElB) 03/09/98C

c. Conduct Tlinspections (REGIONS, EElB,IOMB) 09/30/98C

i. Region 1 (8/3 - 8/7 & 9/21 - 9/2S) li. Region ll (3/16 - 3/20 & 5/4 - 5/8) lii. Region lll (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, lOMB) 08/28/98C

4. Inspect at least three major circuit breaker vendors' facilities and at least three of 09/30/99 the third-party repair / refurbishment outfits (IOMB, EElB)

a. GE facilities

b. Westinghouse facilities

c. ABB facilities

d. PDS ERAM

e. NLI

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 submittats (NRR/REXB) Ongoing

6. Participate in Industry meedngs (EElB, IOMB, REXB)

a. GE Users Group meetings 2/15-2/19/99

b. Westinghouse Users Group meetings 8/23-8/27/99

c. ABB Users Group meeting 6/14-6/18/99

7. a. Summarize inspection findings and evaluate if any additional regulatory actions is required (EElB, REXB, lOMB) TBD04/30/99

b. Complete the required action (EElB, REXB, lOMB) 06/30/99

==

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 instructions, procedures, and drawings. Also, the use of inadequately manufactured and dedicated parts 15

appears to be responsible for some recent breaker failures. Over the years the NRC issued information notices (ins) d(. scribing the breaker failures and expected that licensees would review their maintenance l

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-rpecific 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 (greater than 1.0E 4) reported in the ASP reports (NUREG-4674). Based on these risk insights, generic regulatory actions would not be warranted. However, reviews performed by NRC contractors indicate that breaker malfunctions 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 i 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 j 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 a temporary instruction (TI) 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 mediurn-voltage and low-voltage nielsi-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 (IOMB and EElB). NRR completed the breaker inspections in October 1998 and is currently evaluating whether 1 additional regulatory actions are needed. I 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 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 failurel and review of 10 CFR Part 21 submittals; (6) participating in industry meetings; and (7) summarizing Tl inspection findings and issuing generic communication if required. .

1 Current Status: Tl was issued on December 31,1997. A revision to the Tl was issued on March 9,1998. l 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 i indicated above. The staff is currently evaluating whether additional regulatory actions are needed for j addressing circuit breaker maintenance programs, and is in the process of preparing an Information i Notice summarizing the inspection findings. I NRR Technical

Contact:

A. Pal,415-2760 16 J

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~ ENVIRONMENTAL SRP REVISION ACTION PLAN TAC No. MA0837. Last Update: 04/13/99 Y 1GSI: Not Available Lead NRR Division: DRIP

, MILESTONES DATE (T/C)

1. Reflect Potential impacts and integrated impacts in Options for Resolution

a. .ldentification of potentialimpacts

b. Identification of integrated impacts ._ .

03/96C

c. Proposed options for resolution and develop initial draft of 06/96C

. revised ESRP L d. Staff / contractor meeting 'to resolve format and content of 10/96C revised ESRP :

11/960

2. Prepare Final Draft of ESRP Sections for Public Comment

a. Draft updated ESRP for staff review ^

01/97C

b. ACRS and/or CRGR review, if necessary 06/970

c. Publish (electronic) for public comment 09/97C 3J Disposition Public Comments 02/98C

4. Publish Final NUREG-1555 05/99T

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5. Maintenance of program data Ongoing Brief Descriotion: 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.

Reaulatory 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. 1 Current Status: The PNNL/NRC_ staff workshop on the restructured and revised ESRP was held during  !

November 13-14,1996. Now that the Part 51 rule for license renewal is final, particular emphasis is 1

. being placed on assuring that license renewal 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 i

- 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 diske"e) 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 17 L

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l :1 T Committee determined that it no longer needed a subsequent SAMDA/SAMA briefing on the ESRP or 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 concern raised by the public was the consolidation of guidance for the technology area across disparate licensing frameworks (i.e., Parts 50,52, and 54); the staff restructured the document to segregate guidance into a Part 50/52 ("greenfield"-type review) and that for Part 54 (renewal of a license for an existing facility). This segregation took the form of a supplement to the ESRP and was completed in draft form on July 3,1998. During December 1-3,1998, the final PNNUNRC 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 final 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 i Alternatives 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. As of the date of this report, the letter issue has not been resolved.

NRR Technical

Contact:

B. Zalcman, RGEB,415-3467 -

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EMERGENCY ACTION LEVEL GUIDANCE DEVELOPMENT '

TAC No.: MA3695 Revision to NESP-007 Last Update: 03/31/99 ,

M98020 Shutdown EAL Guidance Lead NRR Division: DIPM .

I REVISION TO NESP-007 (NEW DOCUMENT NEl-97-03) l MILESTONES DATE (T/C)

1. Meet with NEl to discuss NEl-97-03 10/19/98C

2. NEl to provide revised NEl-97-03 with shutdown EAL guidance 11/2/98C removed for NRC comment-

3. NRC provide comments on NEl-97-03 to NEl 12/3/98C

_4. NEl submit NEl-97-03 for NRC endorsement 1/11/99C l

5. Draft Guide developed (' Revision to Regulatory Guide 1.101) endorsing 5/99T NEl-97-03 for interim use and comment

6. CRGR/ACRS meeting on draft guide 6/99T

7. Draft Guide issued for public comment 7/99T

8. Public comments addressed (any needed revision to NEl-97-03 10/99T completed)

9. CRGR/ACRS meeting on final guide' 6/00T

10. Regulatory Guide issued' 8/00T

• 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) i i

1. Meet with NEl to resolve staff concerns on WEl'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 4/07/99T review

3. NRC provides comments to NEl on NEl-99-01 5/07/99T

4. . Meet with NEl to discuss comments 5/12/99T

5. Comments resolved and final draft of NEl-99-01 submitted for 7/15/99T endorsement

6. ' Draft guide developed endorsing NEl-99-01 developed in form of a draft 10/99T guide for CRGR/ACRS review.

7. Determination made on whether to issue a Generic Letter on plant- 10/99T specific implementation of shutdown EALs

8. CRGR/ACRS meetino on draft culde and aeneric letter 12/99T 19

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9. Draft Guide issued for public comment 1/00T I 10. ' Public comments addressed (NEl-99 01 revised as needed) 4/00T l

11. CRGR/ACRS meeting on final guide and generic letter 6/00T

12. Regulatory Guide and generic letter issued - 8/00T

Description:

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, 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 l EAL guidance and licensee's EAL schemes were deficient: (1) loss of power EALs were ambiguous and I (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.

Proposed 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.

Oriainatina Documents: Vogtle llT EDO Staff Action item 4a [7]  !

NUREG-1449 l

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. Regulatory 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 c

- 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 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 1on 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.

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 expects to provide a draft of NEl 99-01 to the NRC for its review by April 7,1999.

References:

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. . .

i

'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 fmm A. Nelson to J. Roe, September 16,1997.

7. Memorandum from J. Taylor to T. Murley, June 21,1990.

8. Letter from B. Zalcman 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.

l 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 S. Magruder, DRIP,415-3139

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PRA IMPLEMENTATION PLAN 1.2(d) ,

Graded Quality Assurance Action Plan

{

. TAC Nos. M91429, M92447, Last Update: 4/9/99 M92448, M92449, M88650, M91431, Lead NRR Division: DIPM M91432, M91433, M91434, M91435, Support Division: DSSA 3 M91436. and M91437, M92420 and M94163 GSI: Not Available j i

MILESTONES - DATE (T/C) 1

1. Issued SECY-95-059 03/95C

2. Begin interactions with volunteer licensees 05/95C Palo Verde letter dated 4/6/95 l Grand Gulf meeting 5/4/95

- South Texas meetings on 4/19/95 and 5/8/95

3. NRC Steering Group meetings to guide working level staff activities As Needed

- 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 GA controls NRC letter conveying trip reports issued on 12/4/95 Meeting on 4/11/96 to discuss the staff evaluation guide 1 Letter from licensee on 4/24/96 providing comroents on staff . 1 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 I

GOA closeout letter transmitted to licensee on 7/2/98

5. Staff interactions with South Texas Ongoing

- Meeting on 7/17/95 on project status through Site meeting on 10/3-4/95 on risk ranking and QA controls 3/98C Meeting on 12/7-8/95 to discuss risk ranking and QA controls

- South Texas Submittal of 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 auidance 22

q.

u, l Meeting on 6/19/96 to discuss staff comments on the QA plan submittal for graded QA, review questions transmitted -

to STP on 8/16/96 Site visit on August 21-22 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 Additional information (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, QA 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 RAl response STP submittal of draft QA Plan on 5/21/97 STP submittal of GOA related procedures, responses to RAI,

.5 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.

1 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 c - 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 Letter to GGNS dated 5/29/96 regarding implementation of OAP 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 QAP commitments for low-safety significant items, meeting summary issued on 12/17/96 Site visit on 11/21/96 to review procurement activities, trip report was issued on 11/6/97 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 i

9. Issue letter to 3 volunteer plants outlining program objectives and review 1/96C expectations. Distributed staff evaluation guide to licensees.

10. Evaluation Guide issued for use by staff in evaluating volunteer plants Meeting held with volunteer plants to receive feedback on 1/96C staff evaluation guide on 4/11/96.

- Industry comments on staff evaluation guide provided by 4/96C 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 SRM conditionally approves issuance of GOA RG 6/29/98C GOA final RG issued 8/98C 24

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12. ACRS Briefings -

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Expert Panel and deterministic considerations 2/27-28/96C ,

l.

Graded QA 4/11/96C I L

PRA implementation Plan and pilot projects 7/18/96C Risk Informed Pilots - 8/7/960 Graded QA Regulatory Guide 11/22/96C Graded QA Regulatory Guide 2/21/97C ACRS Concems 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 3/11/970

= Graded QA Regulatory Guide 2/27/98C Graded QA Inspection Procedure 12/8/98C

- 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 procedere 6/99T

15. Conduct NRC Staff Training ^ 7/99T i

Description:

Prepare staff evaluation guidance and regulatory guidance for industry implementation for the grading of quality assurance (QA) practices commensurate with the safety significance of the plant

. equipment. The development of this guidance will be based on staff reviews of regulatory requirements, proposed changes to existing 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 QA implementation practices that have evolved have often not been graded. In the development of implementation guidance for the maintenance rule, a methodology to determine the risk significance of plant equipment was proposed by the industry (NUMARC 93-01). During a public mee'ing on December 16,1993, the staff suggested that the industry cauld build on the experience gained from the maintenance ruis to develop implementation methodologies for graded QA. The staff had numerous interactions with the Nuclear Energy Institute (NEl) during ca'endar year 1994 as the graded QA concepts were discussed and the initial industry guidelines were devel3 ped 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 effortr,.

Proposed Actions: The goal of the action plan is to utilize the lessons learned from the 3 volun:eer licensees to modify staff-developed draft guidance to formulate regulatory guidance on acceptable i methods for implementing graded QA. The staff will develop a regulatory guide based in part on input  !

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g i from Brookhaven Natidnal LEboratory, arid will also prepare a baseline and reactive inspection procedure

(IP) for graded QA. An inter-office team has been established to prepare the regulat6 guidance

documents and test their implementation during the evaluation of volunteer plant activities.

Orldinatina Document: Letter from b. Sniezek, NRC to J. Colvin (NUMARC) dated January 6,1994, describing the estaNdhment of NRC steering group for the graded QA initiative.

Regulatory 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 leamed 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 for inspections of.the risk significance determination

aspects of graded QA programs.~ '1 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 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 OA 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 mede, and accepted by the Chairman, to focus staff efforts on revising the general regulatory guide and standard review plan first.

The proposal to sequentially complete the application specific guidance documents, including GOA, 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 GQA 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 GOA RG.- Prior to issuance of the RG the staff will have to review, and revise accordingly, the RG i 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.

j 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 A

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k By letter dated November 13,1998, the ACRS stated that the proposed IP was found adequate to perform an evaluation of licensee graced 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 req'uisite PRA expertise to implement the IP in a uniform manner.

2. The IP 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.

l

3. l The inclusion of high-safcty c!gnificant 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 consirier the development of plant-specific Temporary Instructions to verify graded QA 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 concems.

Conversion of the IP to a Tl that would be specific to South Texas Project was consbered, 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 staff plans to revise the draft IP to address generic OA issues and submit it for reconsideration by CRGR, The target date to issue the IP is June 18,1999.

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 J 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, i 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 issd a letter to the Chairman which recommended that the GOA RG (RG 1.176) be issued for use. The ACHS expressed a concern that RG 1.176 does not take 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 OA controls on PRA parameters, and for the staff to prepare a plan for improvements to RG 1.176 after gaining j experience with its application and to brief the committee within the next 2 years. I I

27

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- South' Texas submitted their QA program revision for their graded QA effort on March 28,1996. The change has been reviewed by the staff (lOMB, SPSB, RES, RIV, and NRC contractors). A meeting was held with STP on June 19 to discuss the staff's comments and concems. 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 OA controls. The staff also observed both a Working Group and Expert Panel meeting et which time licensee safety significance evaluations for 2 systems (Radiation Monitoring and Essential Service Water) were discussed. Staff review of the updated QA program submittal was completed and a

- second RAl 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 RAI. Staff (from both lOMB and SPSB) performed a site evaluatio'n during the week of May 5-8 to review aspects associated with: PRA quality, OA controls for the PRA, corrective action and performance monitoring feedback processes, OA controls for low safety significant items,' detailed information presented to -

address issues raised in the RAl,~and the audit scheduling process. Further dialogue has occurred

~

between the staff and STP 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, i 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 discum the experience with implementing GOA. STP Indicated that 6 systems had been evaluated and

. that a majority (89%) of the 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 j 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 facility. 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- l 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 performed 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 rev~iew plan By letter dated June 8, NEl indicated that their 96-02 guide will be revised.

Further, NEl requeste'i 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 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 initiativei NEl stated their concerns regarded the questions raised by the staff in the area of QA controls for items determined to be low safety significant and in the area of safety i 28 y

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' isignificance 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 -

L flexibility to manage their QA programs. The vo.lunteer 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. Snlezek (NRC) to J. Colvin (NEl) dated 1/6/94.

2). Regulatory Guide 1.160. .

~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. L .

. . , .. l

8) NEl 96-02, " Guideline for implementing a Graded Approach to Quality."

-4, 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-informeci 9egulation 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.

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ACCIDENT MANAGEMENT IMPLEMENTATION TAC # M91966 Overall Last Update: 4/7/99 M91641 - BWROG SAMG Review Lead NRR Division: DSSA Mit ESTONES DATE (T/C)

1. BWROG Severe Accident Management Guidance (SAMG) documents Complete review of SAMG documents 7/98C Resolve remaining technical concerns 6/99T

2. Review severe accident training materials and BWROG prioritization 6/95C {'

methodologies

3. Develop guidance for NM 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. Commission paper regarding inspection and oversight of voluntary 8/99T programs

5. Complete A/M audits TBD

6. Hold public workshop TBD

7. Report to Commission on audit findings and recommendations for TBD achieving closure Descriotion: This action plan is intended to guide staff efforts to assess the quality of utility j implementation of accident management (A/M), and the manner in which insights from the IPE program have been incorporated into the licensees NM program. Specific review areas will include: development l 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 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 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 l

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 NM 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 NM 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 commitment with the understanding that the staff would inspect utility implementation [7].

Proposed Actions: Specific actions included in the NM action plan are: (1) complete the review of BWROG SAMG documents, (2) conduct NM demonstration visits to observe how the elements of the formal industry position are being implemented, (3) complete NM audit guidance using the information and perspectives obtained through the demonstration visits, (4) conduct NM audits, and (5) hold a public 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, f 30 1

F l- , 1 Oriainatina Document: 'SECY-88147, Integration Plan for Closure of Severe Accident issues, May 25,1988.

l Reaulatory Assessment: Accident management programs are being implemented by licensees as part of l an initiative to further reduce severe accident risk below its current, and acceptable, level. Consequently, l this is a non-urgent regulatory action and continued facility operation is justified.

l Current Status-Severe Accident Management Guidelines

- Severe accident management guideline documents have been submitted by each of the PWR owners 9  : groups, and reviewed by the staff [8]. The BWROG submitted Rev. O of the Emergency Procedure and i

Severe Accident Guidelines (EP/ SAG) and associated technical basis documents to NRC for information on August 29,1996 [9].EThe staff and Oak Ridge National Laboratory have completed a high level review of the EP/ SAG documents. Areas where additionalinformation and discussion with the BWROG are

' corJered 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 responsei 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 concerns and expects to provide a written response to the NRC by May_1999.' l l

' Utility implementation Licensee commitments and target dates for completing W 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 a 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 22,1997, letter [16]. NRC provided comments on this guidance in a January 28,1998, letter to NEl (17]. In 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 concern 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 NEI's and NRC's expectations, and that industry should continue to proceed with implementation.

. Implementation has now been completed at almost all plants. Two utilities have rescheduled their completion dates to the end of 1999. For about 20 other sites, licensees have not provided letters confirming that implementation has been completed. Staff will follow up with those sites regarding their implementation status.

I NRC Evaluation - l 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, j i

l 31 j

(3) performing pilot inspections ct cbout fivs pl:nts using the TI, (4) developing an inspection procedure (IP) 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.

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 in orporate insights from the A/M demonstration visits, staff positions contained in NRC letters to NEl, anc ieedback received on the draft Tl. 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. ,

The staff committed to carefully consider the industry proposal, and provide a response to NEl. l The staff is planning to prepare a Commission paper addressing the manner in which voluntary programs, such as A/M and shutdown risk, should be included in the risk-informed inspection and assessment process. The question regarding A/M audits and oversight will be addressed in this paper. I

References:

1. Memorandum from F. Rowsome to W. Minners, "A New Generic Safety issue: Accident Management," April 16,1985

2. SECY-88-147, Integration Plan for Closure of Severe Accident issues

3. SECY-95-079, implementation Plan for Probabilistic Risk Assessment

4. SECY 89-012, Staff Plans for 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 from W. Russell to W. Rasin, January 9,1995

' 8. Letter from W. Russell to W. Rasin, February 16,1994

9. Letter from K. Donovan to Document Control Desk, August 29,1996

10. Letter from D. Matthews to K. Donovan, April 2,1997

11. Letter from K. Donovan to Document Control Desk, May 10,1997

12. Letter from T. Rausch to Document Control Desk, January 9,1998

13. Letter from T. Essig to T. Rausch, July 20,1998

14. Letter from A. Thadani 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

l l' .

l' 17. , Letter from G. Holahara 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 f

20. ' - Letter from S. Collins to R.' Beedle, August 10,1998 f NRR Technical Contegt: R. Palla, DSSA,415-1095 g ' NRR Lead PM: . Ramin Assa, DLPM,415-1391 l

l

?

?

5 4

4 v

l i

I 1

1 t

'P l.

I i

i t

33 I

e . .- __

CORE PERFORMANCE ACTION PLAN '

1 Final Update TAC Nos. M91257 - DSSA Status: COMPLETE M91602 DIPM Lead NRR Division: DSSA GSI: Ll 179 Supporting Division: DIPM Descriotion: The action plan covered the assessment of the impact of reload core design activities on fuel and control rod reliability affecting plant safety. The activities were accomplished through performance-based inspections of fuel vendors, evaluation of licensees' reload analyses, supplemented with independent evaluation of core performance information, and coordinated with regional training and interaction.

Historical Backaround: The action plan addressed a review of fuel fabrication, reload core design, and reload analysis issues that were discussed during 1994,1996, and 1997 briefings given to the Executive Director for Operations (EDO), and at the June 19,1997 Chairman briefing. The briefings presented by the Reactor Systems Branch (SRXB), Division of Systems Safety and Analysis (DSSA), covered generic fuel and core performance issues and related evaluations of fuel failures. The former Special Inspection Branch (PSIB), Division of Inspection and Support Programs (DISP), supported the briefings. As a result of these briefings, the Office of Nuclear Reactor Regulation (NRR) was directed to focus on licensee activities and the licensee / vendor interfaces.

Oriainatina Document: Memorandum from Gary M. Holahan and R. Lee Spessard to Ashok C. Thadani, dated October 7,1994, ' Action Plan to Monitor, Review, and improve Fuel and Core Components Operating Performance" and the enhanced focus on licensee reload design participation resulting from briefing feedback.

Reaulatory Assessment: Core design is a fundamental component of plant safety because maintaining fuel integrity is the first principal safety barrier (i.e., fuel cladding, reactor coolant system boundary, or the containment) against serious radioactive releases. Likewise, the safety analyses must be property performed in order to verify, in conjunction with startup tests and normal plant parameter monitoring, that the core reload design is adequate and to provide assurance that the reactor can safely be operated.

Evaluation of activities that affect the quality of fuel and core components are important to ensure that safety and quality are not degraded and that the core performs as designed.

Resolution: An action plan was developed to: (1) evaluate fuel vendors' quality assurance (OA) performance through performance-based inspections that evaluate the reload core design, safety analysis, licensing process, fuel assembly mechanical design, and fuel fabrication activities; (2) evaluate the OA performance of licensees that perform core reload analysis functions; (3) identify, document, and categorize core performance problems and root cause evaluations that will be further evaluated during these inspections and provide input to licensee evaluations as well as support regional enforcement actions, as appropriate; (4) train and coordinate regional support staff participating in these activities; and (5) eva!uate the results of these activities for use in formulating generic communications, revisions of regulatory guidance and guidance for regionalinspectors, and other appropriate regulatory actions. A Core Pe.iormance Workshop was conducted on October 24-25,1996, to inform the industry of generic issues, and solicit feedback. The status of core performance inspection evaluations and emerging issues has been covered at the yearly Regulatory Information Conferences.

The action plan included ten vendor inspections, including all five domestic reload fuel vendors and their major fuel cladding suppliers, that were performed by multi-disciplined inspection teams led by lOMB with contracted technical assistance, as required. These inspections are documented in the individual vendor inspection reports. The action plan also included selected licensee inspections in each region, with the regional team leaders, to assess licensee performance in reload core analysis vendor oversight and participation. Licensee inspections were normally issue-driven.

34

{

The data acquired thrsugh licensee / vendor inspections has been integrated with information supplied by the regions and other sources and is evaluated for generic core performance indicators and industry conformance to current regulatory requirements, as part of our normal activity. One end product of the l on-going assessment is continuing interaction and guidance for resident inspectors and regional staff.

The ongoing activities to capture and address early warning of emerging core performance issues will

' also continue to evaluate vendor / licensee interface issues and potential generic problems, with our continuing assessment of core performance under the DSSA operating plan.

~ The data acquired from the completed vendor and licensee inspections have been evaluated for generic impact and identification of emerging issues. Results from recent NUPlO Nuclear Fuel Committee joint vendor surveillances and audits have been reviewed to supplement our vendor evaluations. We are attending the NUPIC Nuclear Fuel Committee vendor-specific Affinity Group annual meetings, and evaluating options _for sharing vendor oversight results. DSSA is evaluating the action plan results to better integrate and prioritize its on-going activities, consistent with the DSSA operating plan. Options

. and recommendations are being prepared for management review to capture the lessons teamed and to provide ongoing interaction and guidance for regional staff. This was documented in a memorandum from Jared S. Wermiel to Gary M. Holahan, "Closecut of Core Performance Action Plan," dated February 16,1999.

The action plan activity to review the adequacy'of vendor lead testing programs for new fuel designs has been transferred to the Agency Program Plan for High Burnup Fuel, issued on July 6,1998.

NRR Technical Contacts: E. Kendrick, DSSA/SRXB,415-2891 G. Cwalina, DIPM,415-2983 l

i J

35  ;

' HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN

. (Previously Part of the Dry Cask Storage Action Plan)

TAC Nos. M93821: Action Plan: Last Update: 04/09/99 M91955: DSC generic review Lead NRR Division: DSSA M95546: Generic review of NRCB 96-02' 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 documentsi

- < Develop summary of guidance.' ' 2/96C

2. Determine significant lkeavy 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. . 2/96C 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.' e

- lasue bulletin on load movement during operations. 4/96C

6. Draft staff guidance and disseminate to appropriate raanagement (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 TI). 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. J lssue final inspection procedures. TBD

11. Issue final guidance. TBD*

Note: .* Indicates that the activity is contingent on the results of NRR's/RES's review of NRR's proposed GSI on the potential risk and consequences of heavy load drops in nuclear power plants.

Description:

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)..

36

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 regulatory progrn.m for 1:.e 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 concems regarding '

what is required for the movement of heavy loads while the pl9nt 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.

Prooosed 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.

Orlainatina Document: Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, r July 28,1995, " Dry Cask Storage Action Plan."-

BeIlulatory. 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,'several licensees 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 responses 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 failures were abandoned in early 1998 due to budget shortfalls. The staff will propose to RES (ETC 04/99), 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. The heavy loads issue'(USl 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 mid-1999.

The staff visited Calvert Cliffs in 1997 for the purpose of obtaining an understanding how t'n e 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 37

y. -

Refefangen:

} i ,

[ Memorandum from Robert M. Bernero and William T. Russell to James M. Taylor, March 15,1995, 1

( " Realignment of Reactor Decommissioning Program."  !

i l- Memorandum f rom Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995, " Dry i . Cask Storage Action Plan."  ;

1 4

)

)

38

i WOLF CREEK DRAINDOWN EVENT: ACTION PLAN l

TAC Nos.: M66278 Last Update: 03/31/99 Lead NRR Division: DSSA MILESTONES DATE (T/C)

1) Draft Generic Letter (GL) 11/95(C) l

2) Issue Supplement to IN 95-03 03/96(C)

3) Complete Draft Tl/ Issue to the Regions for Comments 12/98(C)

4) CRGR Concurrence of the GL for 1st time 09/96(C)

CRGR Concurrence of the GL for 2nd time (after reconciling Public Comments) 01/98(C)

5) ACRS Briefing 11/97(C)

6) GLissued 05/98(C) )

7) Receive Regional Comments on Tl 1/99(C) _

8) Complete Evaluation of the Responses to the GL 03/99(C)

9) Issue Tl 04/99(T) l

==

Description:==

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 tranF, ferring 9200 gallons of hot RCS water to the RWST in 66 seconds.

The Wolf Creek event represents a LOCA with the potential to consequentially fail all the ECCS pumps and bypass the containment. Another important feature of this event is the short time available for corrective action. Based upon calculations by the licensee and the staff, it is estimated that if the draindown had not been isolated within 3-5 minutes, net positive suction head would have been lost for 4

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) isshe 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 pror;de information on draindown vulnerabilities and the measures they implemented to diminish the p robability of a draindown. 1 Oriainatina DocumeDj: AEOD/S95-01," Reactor Coolant System P.owdown at Wolf Creek on September 17,1994".

39

Reaulatorv 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.

Current Status,: Information Notice IN 95-03, and its Supplement was issued. CRGR concurred the

i. proposed GL in 9/96 for the first time; but as directed by an SRM, the GL was published in the Federal Register in 2/97 for public comments. ACRS was briefed on 11/6/97. 2nd CRGR concurrence was obtained in 1/98 after reconciling the public comments. The GL was issued on 5/28/98 after reconciling all the comments. Staff issued the draft Tl to the Regions for comments on 12/98. Comments from all the Regions have been received. The staff does not anticipate any significant problems in incorporating the comments from the Regions, and is currently modifying the draft Tl in order to reconcile the comments from the Regions. s NRR Technical

Contact:

M. M. Razzaque, SRXB,415-2882 NRR Lead PM: Kristine Thomas, NRR,415-1362

References:

1) AEOD/S95-01," Reactor Coolant System 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.

1 i  :

4 40

i NEW SOURCE TERM FOR OPERATING REACTORS TAC No. M89586 Last ')pdate: 04/08/99 GSI No.155.1 Lead NRR Division: DSSA Sup>orting Division: DE MILESTONES DATE (T/C)

1. NEl Letter 07/94C

2. Commission Memo 09/94C l

3. NEl Response 09/94C

4. NEl/NRC Meeting 10/94C

5. Publication of NUREG-1465 02/95C

6. NEl/NRC Meetings 10/94C,06/95C,10/95C, 01/960,02/96C,05/96C, 08/96C,10/96C,04/97C

7. Submittal of Generic Framework Document (from NEl) 11/95C

8. First Pilot Plant Submittal 12/95C

9. Issue Memo to Commission, Updating Status 08/96C

10. Present Commission Paper in E-Team Briefing 09/96C

11. Brief CRGR on Commission Paper 10/96C

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 P!Iot 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) 07/99T

23. Finish Third Pilot Plant Review (Indian Point Unit 2) 07/99T

24. Finish Fourth and Fifth Pilot Plant Review TBD 41 1

Description:

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 I this new kn~.redge to operating reactors could result in cost savings without sacrificing real safety margin. In a# tion, safety enhancements may also be achieved.

Historical 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 TID 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 l 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. i 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 app!ication review will be used in develop'ag the associated regulatory guide and standard review plan that will be part of the final rulemaking for the alternative source term.

priainatina 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.

Reaulatorv 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 submitted 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 purs'iing 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 i help shape rebaselining, in June 1997, RES circulated an early draft of the proposed RG that would j consider updated source term insights (NUREG-1465) (the RC would be analogous to RGs 1.3 and 1.4 l 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.

42 m

The st:ff briefed the NRR Executiva Tcm on SECY-96-242 in September 1996, the CRGR in l

October 1996, and the ACRS full committee in November 1996. A limited number of pilot plants j submittals and exemptions are expected - four submittals have been received so far (Browns Ferry, l Perry, Oyster Creek, and Indian Point-2). An application is also expected from Grand Gulf. In addition, l 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 1 EDO issued the letter response to NEl. The staff has initiated the rebaselbing effort. The staff briefed l 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 l 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 j 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 I the lead for rebaselining and NRR is now the lead for rulemaking and regulatory guidance). The i 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 rebaselining, in addition, NRR transferred its technical assistance resources with SNL, ORNL, and PNNL that were designated for rebaselining to RES. These changes will 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-phased plan and preliminary findings from Phase 1, Phase 11, and the DBA portion (with the updated I assumptions) of Phase lit, respectively, for the rebaselining effort. On April 1 and 2,1998, RES and NRR staff t:riefed the ACRS and DONRR, respectively, on the progress of the rebaselining effort.,initialinsights 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 entry in Attachment 2.

l The staff cornpleted 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 July 1999. 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-1465, " Accident Source Term for Light Water Nuclear Power Plants," February 1995. 1 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".

43

Y December 22,1995, pilot plant submittal, letter to Document Control Desk from 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 50, 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,"Rulernaking Plan for implementation of Revised Source Term at Operating Reactors," SECY-98-158.

June 30,1998, memorandum to t'he Commission from NRC staff,"Results of the Revised (NUREG-1465)

Source Term Rebaselining for Operating Reactors," SECY-98-154.

Summaries of public meetings:

e 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; 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; e dated July 28,1997, for public meeting with PWR (Surry) held on June 18,1997.

44

1 y-. Y l

I ATTACHMENT 2 l l

GENERIC COMMUNICATION AND COMPLIANCE ACTIVITIES l

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DIRECTOR'S STATUS REPORT on GENERIC ACTIVITIES Action Plans Generic Communication and Compliance Activities APRIL 1999 Office of Nuclear Reactor Regulation St 9c 7 o ca .1/ y

i INTRODUCTION The purpose of this report is to provide information about generic activities, including genenc 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 genenc compliance activities. Generic communications and compliance activities (GCCAs) are potential generic issues that are safety significant, require technical resolution, and possibly require genenc communication or action.

I Attachment 1, 'NRR Action Plans," includes genene or potentially generic issues of sufficient complexity or scope that require substantial NRC staff resources. The issues covered by action plans include concems

! identified through review of operating experience (e.g., Boiling Water Reactor intemals 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.

Attachment 2, " 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.

9 O

1 TABLE OF CONTENTS BOILING WATER REACTOR INTERNALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 PRA IMPLEMENTATION ACTION PLAN 1.2 c Inservice ins Pla n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . (. .). . ..............

. . . . . . . . .pec tion Action

..6 STEAM G EN E R ATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 MEDIUM-VOLTAGE & LOW-VOLTAGE BREAKER FAILURES . . . . . . . . . . . . . . 15 ENVIRONMENTAL SRP REVISION ACTION PLAN . . . . . . . . . . . . . . . . . . . . . . . . 17 l EMERGENCY ACTION LEVEL GUIDANCE DEVELOPMENT . . . . . . . . . . . . . . . . 19 PRA IMPLEMENTATION PLAN 1.2(d) Graded Quality Assurance Action Plan . 22 ACCIDENT MANAGEMENT IMPLEMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . 30 CORE PERFORMANCE ACTION PLAN Final Update . . . . . . . . . . . . . . . . . . . . . . 34 HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN . . . 36 WOLF CREEK DRAINDOWN EVENT: ACTION PLAN . . . . . . . . . . . . . . . . . . . . . 39 NEW SOURCE TERM FOR OPERATING REACTORS . . . . . . . . . . . . . . . . . . . . . 41

ATTACHMENT 1 l

NRR ACTION PLANS I 'd's . . . _ _ _ _ . _ _ . _ _ _ . . . _

BOILING WATER REACTOR INTERNALS TAC Nos. M91898, M93925, M93926, M94959, M94975, M95369, Last Update: 3/26/99 M96219, M96539, M97373, M97802, M97803, M97815, M98266, Lead NRR Division: DE M98708, M98880, M99638, M99870, M99894, M99897, M99898, Supporting Division: DSSA M99895, M99897, MA1102, MA1104, MA1138, MA1226, MA1926, GSI: Not Available MA1927, MA2326, MA2328, MA3395, MA3683, MA4203, MA4464, MA4465, MA4467, MA4468 MILESTONES DATE (T/C) l PART I: REVIEW OF GENERIC INSPECTION AND EVALUATION CRITERIA

1. Issue summary NUREG-1544 03/96C o Update NUREG-1544 3O/FY99T l 2. Review BWRVIP Re-inspection and Evaluation Criteria o Reactor Pressure Vessel and Internals Examination Guidelines (BWRVIP-03) . . . . . . . . . . . . . . ....... ....... ... ... . .. 06/08/98Cl o BWRVIP-03, Section 6A, Standards for Visualinspection of Core Spray Piping, Spargers, and Associated Components . . . . . . . . . . . . . . . . . . 06/08/98Cl o BWR Vessel Shell Weld inspection Recommendations (BWRVIP-05) . 07/28/98CA o BWR Axial Shell Weld Inspection Recommendations . . . . . . . . . . . . 12/31/99T o Guidelines for Reinspection of BWR Core Shrouds (BWRVIP-07) . . . . 04/27/98CA

3. Review of generic repair technology, criteria, and guidance TBD

4. Review generic mitigation guidelines and criteria TBD

5. Review of generic NDE technologies developed for examinations of BWR internal components and attachments TBD

6. Other intemals reviews (safety assessments, evaluations, mitigation measures, inspections, and repairs) o Safety Assessment of BWR Reactor Internals (BWRVIP 06) . . . . . . . . 09/15/98CA o Bounding Assessment of BWR/2-6 Reactor Pressure VesselIntegrity Issues (BWRVIP-08 & BWRVIP-46) . . . . . . . . . . .. .. .... . 03/27/98CA o Evaluation of Crack Growth in BWR Stainless Steel RPV Intemals (BWRVIP-14) . . . . . . . . . . . . . .. ... . .. ... . ..... 06/08/98Cl o Internal Core Spray Piping and Sparger Replacement Design Criteria (BWRVIP-16) . . . . . . . . . . . . . . .. ..... . .. ..... . ..... 11/16/98Cl o Roll / Expansion of Control Rod Drive and in-Core Instrument Penetrations in BWR Vessels (BWRVIP-17) ... .. . ... .... 03/13/98CD o BWR Core Spray Internals inspection and Flaw Evaluation Guidelines (BWRVIP-18) . . . ... . . . . . . . .. .. .. . .. ...... 06/08/98CA 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 1

-a

< Intemal Core Spray Piping and Sparger Repair Design Criteria (BW RVI P- 19) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

11/16/98CA o Core Plate Inspection and Flaw Evaluation Guideline (BWRVIP-25) . . . 04/30/99T o Top Guide Inspection and Flaw Evaluation Guideline (BWRVIP-26) . . . 04/30/99T o Standby Liquid Control System / Core Plate AP inspection and Flaw Evaluation Guidelines (BWRVID-27) . . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T o Assessment of BWR Jet Pump Riser Elbow to Thermal Sleeve Weld Cracking (BWRVIP-28) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T o Technical Basis for Part Circumferential Weld Overlay Repair of Vessel Internal Core Spray Piping (BWRVIP-34) . . . . . . . . . . . . . . . . . . . . . . TBD o Shroud Support inspection and Flaw Evaluation Guidelines (BW R VI P 38) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T o BWR Jet Pump Assembly inspection and Flaw Evaluation Guidelines

( BW RVI P-41 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 06/30/99T o BWR LPCI Coupling Inspection and Flaw Evaluation Guidelines (B W R VI P4 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T o Update of Bounding Assessment of BWR/2-6 Reactor Pressure Vesse!

Integrity issues (BWRVIP-46) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 03/27/98CA o BWR Lower Plenum inspection and Flaw Evaluation Guidelines (BW RVI P-4 7) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 04/30/99T o VesselID Attachment Weld inspection and Flaw Evaluation Guidelines

( BW RVI P-4 8) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 03/21/99C o Instrument Penetration inspection and Flaw Evaluation Guidelines (BWRVIP-49 Fop Guide / Core Plate Repair Design Criteria (BWRVIP-50) 08/04/98 CA o Jet Pump Repair Design Criteria (BWRVIP-51) . . . . . . . . . . . . . . . . . . . . 12/30/99T o Shroud Support and Vessel Repair Design Criteria (BWRVIP 52) . . . . . 12/30/99T o Standby Liquid Control Line Repair Design Criteria (BWRVIP-53) . . . . . 12/30/99T 12/30/99T o Lower Plenum Repair Design Criteria (BWRVIP-55) . . . . . . . . . . . . . . . . 12/30/99T o LPCl Coupling Repair Design Criteria (BWRVIP-56) . . . . . . . . . . . . . . . . 12/30/99T o Instrument Penetrations Repair Design Criteria (BWRVIP-57) . . . . . . . 12/30/99T o CRD Internal Access Weld Repair (BWRVIP-58) . . . . . . . . . . . . . . . . . . 12/30/99T o Evaluation of Crack Growth in BWR Nickel-Base Austenic Alloys in RPV Internals (BW RVIP-59) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12/30/99T o BWR Vessel and intemals Induction Heating Stress improvement Effectiveness on Crack Growth in Operating Plants (BWRVIP-60) . . . . . 12/30/99T o Technical Basis for Inspection Relief for BWR Intemal Components with Hydrogen injection (BWRVIP-62) . . . . . . . . . ............. .. 12/30/99T 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 internals. 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 IGSCC in core shrouds and other BWR intemals.

2

i-Historical Backaround: Significant crackir.1 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 I

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 94-03 which requires licensees to inspect their shrouds and provide an analysis justifying continued operation untilinspections can be completed.

1 A special industry review group (Boiling Water Reactor Vessels and Intemals 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 Generic Letter. The NRC evaluated the review group's

. reports, submitted in 1994 and early 1905, and all plant responses.

All of the plants uvaluated have been able to demonstrate continued safe operation untilinspection 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 r; actor. 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 specialindustry r; view 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.

Procosed Actions: The staff will continue to assess the scopes that have yet to be submitted by licensees concerning inspections or re-inspections of their core shrouds. The staff will also continu to assess core shroud reinspection results and any appropriate core shroud repair designs on a case-by-case basis. The staff will issue separate safety evaluations regarding the acceptability of core shroud r; inspection results and core shroud repair designs. The staff has been interacting with the BWRVIP and individual 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.

The BWRVIP has submitted 29 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 intemals.

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 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 until inspections 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 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 internals and attachments to assure continuing safe operation.

3

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 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 significance of and recommended inspections for top guide and core plate ring cracking. Technical review of the " Reactor Pressure Vessel and internals 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 (IN) 97 63, " Status of NRC Staff's Review of BWRVIP-05," was issued August 7,1997, to inform 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 for licensees 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 BWRVlP-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 review of BWRVIP-14 and -18 is complete and the staff's SEs have been issued. The staff's review of BWRVIP-16 and -19 on intemal core spray piping inspection and flaw evaluation and repair design criteria, respectively, is complete.

By letter dated December 20,1996, the BWRVIP submitted Appendix C to BWRVIP-18. This appendix addresses the use of BWRVIP generic intemal core spray inspection guidelines for comp'iance with requirements of the license renewal rule (10 CFR Part 54). The staff is reviewing this appendix in conjunction with its review of BWRVIP-18 guidelines.

4

e.

. 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 9717, " 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 BWRVlP-07 to ensure that the vertical core shroud welds, and the core shroud repair, is adequately inspected.-

By letters dated April 25 and May 30,1997, the BWRVIP provided a reaffirmation of the BWR member licensees to the BWRVIP, and committed, on behalf of 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: Keith Wichman, EMCB,415-2757.

Kerri Kavanagh, SRXB,415-3743 Kamal Manoly, EMEB,415-2765 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.

5

PRA IMPLEMENTATION ACTION PLAN 1.2 (c)

Inservice inspection Action Plan TAC Nos. M95125, M97153, Last Update: 3/26/99 M99389, M99756, MA0125, Lead NRR Division: DE MA0867, MA0868 Support Division: DSSA, EMCB 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 /comrnents 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 07/08/97C

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 tris' se by the staff 06/98C WOG TOPICAL REPORT MILESTONES DATE (T/C)

1. Technical Meeting 9/22/98C 2c WOG Commitment Letter 9/30/98C

3. Issue FSER 12/15/98C 1 I

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EPRI TOPICAL REPORT MILESTONES DATE (T/C)

1. Issue RAls in 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 4/15/99T

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

3. Issue FSER ANO-2 12/29/98C

' Clarification of some aspects of the program are stillin progress.

l INSPECTION PROCEDURES MILESTONES l DATE (T/C)

) 1. Iri Draft inspection Procedure Number 73753 6/98C

2. It e Final Inspection Procedure Number 73753 6/98C

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 acceptability of the pilot programs with respect to their RI-ISI applications and prepare plant specific safety evaluation reports (SER). The action plan describes the process for the review of RI-ISI 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.

Historical Backaround: On August 16,1995, the U.S. Nuclear Regulatory Commission (NRC) published a policy statement (60FR42622) on the use of probabilistic risk assessment (PRA) methods in nuclear regulatory activities, in the statement, the Commission stated its belief that the 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 Rsgulation/ 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).

7

ASME is working on three Code Cases for alternate examination requirements to ASME Section XI, Division 1 for piping welds. Code Case N-577 is based on the WOG methodology and Code Cases N-578 is based on the EPRI methodology. Code Case N-560 is based on the EPRI methodology but is being revised to encompass both methodologies.

ProDosed 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 ISI. 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 is being 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 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 attemate 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 alternate examination requirements for Risk-Based Selection Rules for Class 1,2, and 3 piping welds, is based on the EPRI methodology. The NRC staff intends to work with the industry and ASME to ensure integration and consistency of the various approaches being proposed.

The major difference between Code Case N-577 and the WOG methodology submitted to the staff (Reference 4,6) is that the scope of the Code Case is limited to ASME Class 1,2, and 3 systems while the WOG methodology may encompass all the safety significant systems in the plant. In 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 intends to review the WOG methodology as well as the Code Case N-577 and the consistency of the Surry 1 pilot program for RI-lSI 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 methodology is that the scope of the Code Case is limited to ASME Class 1,2, and 3 systems while the EPRI methodology may encompass 8

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 RMSI 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 alternate examination 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 will review the EPRI methodology as well as Code Case N-560 and the consistency of the Vermont Yankee RI-ISI pilot i

program to both of these.

The staff utilized the acceptable alternative provision of 10 CFR 50.55a (a)(3)(i)to approve the pilot plants' applications. The staff is working closely with ASME to expedite changes involving ISI.

Lono-Ranoe 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.

Fm N RI-ISI programs submittals subsequent to the approval of the pilot plant programs and topical fu; . .s, but prior to the endorsement of ASME Code Cases, it is expected that the licensees will utilize the approved WOG or EPRI Topical Report as guidance for developing RI-ISI programs but will need to seek relief from NRC to the current 50.55a requirements. A minimal review cycle is expected for the approval of Al-ISI submittals during this time frame.

It is anticipated that subsequent to the issuance of safety evaluation reports (SER) for the pilot plants and the topical reports, the industry will revise the ASME Code Cases to incorporate lessons learned 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 161 programs without the need for specific NRC review and approval. For the RMSI programs developed after the RI-ISI methodology has been endorsed in RG 1.147 (and endorsed in 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 normalISIinspection 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 Al-ISI program at the next available opportunity using the staff approved topical reports on WOG or EPRI methodology.

During the two-year extension period, the licensees would continue to implement their current ISI program. In order to disseminate the information to the licensees, the staff issued information Notice 98-44.

9

m Oriainatina Documents: In a November 30; 1995, memorandum to J. M. Taylor, the NRC EDO, Chairman Jackson directed that the staff prepare an action plan together with a timetable for developing RGs and SRPs applicable to use of PRAs to be completed in two years. In his response of January 3, l 1996, the EDO presented a plan that established milestones for the development of regulatory guidance I

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.

Reaulatory Assessment: The operational readiness and functionalintegrity of certain safety-related piping and associated structural elements (e.g., pressure retaining welds) are vital to the safe operation of nuclear power plants. 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.

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 parallel path, ASME is developing Code Cases for altemate 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 locations within individual systems.

The NRC is using probabilistic methods, as an adjunct to deterministic techniques, to help define the scope, type, and frequency of ISI. The development of RI-ISI programs has the potential to optimize the i 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-98-139) for information. The RG and SRP were issued in the Federal Registerin 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) on November 9,1998. The staff completed its review of the Surry Unit 1 (RI-ISI) pilot program l l and issued its safety evaluation report (SER) on December 16,1998. The staff completed its review of l 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 EPRI's responses to NRC's Request for Additional Information (RAI) related to the approach described in EPRI topical report, TR-106706, Risk-Informed Inservice Inspection Evaluation Procedure. Based on the discussion, EPRI plans to revise the topical report by April 15,1999, to incorporate lessons leamed from the pilot applications (Vermor:t l Yankee and ANO-2) of the methodology; methodology enhancements which have evolved since the 10

June 1996 report was issued; and to provide further clarific ation and guidance as necessary, based on >

NRC RAls. The staff's plan is to issue the draft SER by mid-June,1999, to support ACRS meetings, and complete the final SER by Octcber 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: S. 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 Shirley 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, *1mprovements 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.  ;

i

6. WCAP-14572. Revision 1, Westinghouse Owners Group Application of Risk-Informed Methods to I Piping inservice inspection Topical Report," October 1997.

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STEAM GENERATORS TAC Nos. M88885, M99432, MA4265 Last Update: 3/26/99 Lead Division: DE (#394)

MILESTONE DATE (T/C)

1. Commission /EDO Approve.1 02/94(C)

2. Receive NEl Document 02/96(C)

3. Review NEl Document Revisions Continuous Process

4. Regulatory Analysis 5/97(C)

5. Propcsed GL Pkg 10/97(C)

6. ACRS Endorsement 9/97(C)

7. CRGR Concurrence On hold W

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

Description:

The NRC originally planned to develop a rule pertaining to steam generator tube integrity. The [Foposed rule was to implement a more flexible regulatory framework for steam generator surveillance and maintenance activities that allows a degradation specific management approach. The results of the regulatory analysis suggested that the more optimal regulatory approach was to utilize a generic letter. The NRC staff suggested, and the Commission subsequently approved, ,

a revision to the regulatory approach to utilize a generic letter. In SECY-98-248, the staff recommended to the Commission that the proposed GL be put on hold for 3 rnonths 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.

Mhis revision reflects the status of the proposed GL as "on hold". The staff estimates that by mid-1999, dependent on the timing of industry submittals, that the staff should be able to reach a determination of i whether the proposed GL should be reactivated. At that time (if the GL is reactivated), the staff will provide revised schedular estimates.

12 o

l Reoulatory Assessment: The current regulatory framework provides reasonable assurance that operating Pi 7s are safe. The current regulatory framework that implements governing requirements through the s a nt technical specifications can be improved. The staff is currently working with NEl to find G 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 l

- 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 NEVindustry senior mgmt re. GL status - July 22,1997 '

- Briefed ACRS re. GL/DG-1074/DPO issues -- August 26,27, September 3,1997 l

-Information Brief for CRGR re. GL and backfit -- September 9,1997 '

- Met w/NEl re. GL/DG-1074/TSs - September 11,1997

- ACRS endorsement to issue GL and DG-1074 for public comment -- September 15,1997

- 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 information briefing on package

- Met with CRGR on July 21,1998, for detailed 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

- Internal 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.

13

NRR Technical Contacts: Ted Sullivan, EMCB,415-3266 Tim Reed, EMCB,415-1462 RES

Contact:

N/A l

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1 MEDIUM-VOLTAGE & LOW-VOLTAGE BREAKER FAILURES Last Update: 3/26/99 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) 09/30/980

1. Region I (8/3 - 8/7 & 9/21 - 9/25) li. Region 11 (3/16 - 3/20 & 5/4 - 5/8) iii. Region 111 (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, lOMB) 08/28/980

4. Inspect at least three major circuit breaker vendors' facilities and at least three of 09/30/99 the third-party repair / refurbishment outfits (lOMB, EElB)  :

a. GE facili9es

b. Westinghouse facilities

c. ABB facilities

d. PDS ERAM e.NLI

f. Wyle Labs

5. Monitor circuit breaker f ailures 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, IOMB, REXB)

a. GE Users Group meetings 2/15-2/19/99

b. Westinghouse Users Group meetings 8/23-8/27/99

c. ABB Users Group meeting 6/14-6/18/99

7. a. Summarize inspection findings and evaluate if any additional regulatory actions is required (EElB, REXB, lOMB) TBD04/30/99

b. Complete the required action (EElB, REXB, IOMB) 06/30/99

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 instructions, procedures, and drawings. Also, the use of inadequately manufactured and dedicated parts 15

r l

l appears to be responsible for some recent breaker failures. Over the years the NRC issued information j notices (ins) describing the breaker failures and expected that licensees would review their maintE nance 4 programs and correct the deficiencies described in the ins. Desp:te 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 (greater than 1.0E-4) reported in the ASP reports (NUREG-4674). Based on these risk indghts, generic regulatory actions would not be warranted. However, reviews performed by NRC contractors indicate that breaker malfunctions 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 problemmith medium-voltage and low-voltage .

circuit breakers, the staff prepared a generic letter to address this 3 .e. On May 13,1997, the staff requested that the Committee to Review Generic Requirements (C iGR) 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 a temporary instruction (TI) 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, ll, 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 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 issuing 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 comple'ed in October 1998 as indicated above. The staff is currently evaluating whether additional regulatory actions are needed for l addressing circuit breaker m'aintenance programs, and is in the process of preparing an Information I Notice summarizing the inspection findings.

NRR Technical

Contact:

A. Pal,415-2760 l

I 16 I

t 1

1

I 1

ENVIRONMENTAL SRP R5 VISION ACTION PLAN 4 l

TAC No. MA0837 Last Update: 04/13/99 GSI: Not Available Lead NRR Division: DRIP MILESTONES DATE (T/C)

1. Reflect Potential Impacts and Integrated impacts in Options for Resciution

h. Identification of p atentialimpacts

b. Identification of ir tegrated impacts 03/96C

c. Proposed optims for resolution and develop initial draft of 06/96C revised ESGP

d. Staff /coraractor m seting to resolve format and content of 10/96C revised ESRP 11/96C

2. Prepare Final Draft of ESRP Sections for Public Comment

a. Draft updated ESRP for staff review 01/97C

b. ACRS and/or CRGR review,if necessary 06/97C

c. Publish (electronic) for public comment 09/9'7C

3. Disposition Public Comments 02/98C

4. Publish Final NUREG-1555 05/99T

5. Maintenance of program data Ongoing Brief

Description:

The Environmental Standard Review Plan (ESRP) Revision Action Plan deals with the r vision 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.

Reaulatory 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 r:quirements 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 Pad 51 for license renewal r: views.

Current Status: The PNNL/NRC staff workshop on the restructured and evised ESRP was held during November 13-14,1996. Now that the Part 51 rule for license renewal is final, particular emphasis is bring placed on assuring that license renewal 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 forits 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 raay request a briefing during the public comment period. The June draft was provided to CRGR for information; the CRGR declined to consider it. Technica! editor, legal (OG,C), and technical (laad technical branches) comments were receked 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 aistributed for comment.

ACRS discussed the NUREG at its May 1,1998, meeting; in subsequent interactions with ACRS staff, the 17

Committee determined that it no longer needed a subsequent SAMDA/SAMA briefing on the ESRP or 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 a orkshop 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) and that for Part 54 (renewal of a license for an existing facility). This segregation took the form of a supplement to the ESRP and was completed in draft form on July 3,1998. During December 1-3,1998, the final PNNUNRC 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 fina! 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 Attematives 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. As of the date of this report, the latter issue has not been resolved.

NRR Technical

Contact:

B. Zaleman, RGEB,415-3467 1

l 18 i

EMERGENCY ACTION LEVEL GUIDANCE DEVELOPMENT TAC No.: MA3695 Revision to NESP-007 Last Update: 03/31/99 M98020 Shutdown EAL Guidance Lead NRR Division: DIPM REVISION TO NESP-007 (NEW DOCUMENT NEl-97-03)

MILESTONES DATE (T/C)

1. Meet with NEl to discuss NEl-97-03 10/19/98C

2. NEl to provide revised NEl-97-03 with shutdown EAL guidance 11/2/98C removed for NRC comment

3. NRC provide comrr. ants on NEl-97-03 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 5/99T NEl-97-03 for interim use and comment

6. CRGR/ACRS meeting on draft guide 6/99T

7. Draft Guide issued for public comment 7/99T

8. Public comments addressed (any needed revision to NEl-97-03 10/99T completed) 9' CRGR/ACRS meeting on final guide

• 6/00T

10. Regu'atory Guide issued

• 8/00T I

• NEl intends to combine NEl-97-03 with NEl-99-01 into a single EAL guidance document l I

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 4/07/99T i review ,

3. NRC provides comments to NEl on NEl-99-01 5/07/99T

4. Meet with NEl to discuss comments 5/12/99T

5. Comments resolved and final draft of NEl-99-01 submitted for 7/15/99T  :

endorsement I

6. Draft guide developed endorsing NEl-99-01 developed in form of a draft 10/99T guide for CRGR/ACRS review.

7. Determination made on whether to issue a Generic Letter on plant- 10/99T specific implementation of shutoown EALs

8. CRGR/ACRS meetino on draft auide and aeneric letter 12/99T

, 19 l

9. Draft Guide issued for public comment 1/00T

10. Public comments addressed (NEl-99-01 revised as needed) 4/00T

11. CRGR/ACRS meeting on final guide and generic letter 6/00T

12. Regulatory Guide and generic letter issued 8/00T

Description:

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 i long-term fuel storage.

Historical Backoround: 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, 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 1 not available [2]. The NRC's evaluation of shutdown and low power operation in NUREG-1449 also I 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 q

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.

Orioinatino Documents: Vogt!e llT EDO Staff Action item 4a [7]

NUREG-1449 20

Reaulatory 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 needed to improve licensees' capability (Ch regard to timeliness and accuracy) for assessing and classifying the significance of events that o. : urin 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.

The industry provided the final draft of NEl 97-03 Revision 3 for NRC review and approvalin a letter dated January 11,1999. The industry expects to provide a draft of NEl 99-01 to the NRC for its review by April 7,1999.

References:

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-Luop 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. nit, Rev. 3 "6mergency 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. Zalcman to A. Nelson, March 13,1998.

9. Memoranoom from S. Magruder to T. Essig, June 26,1998.

10. Memorandum from S. Magruder to T. Ess'g, 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: S. Magruder, DRIP,415-3139 21 7

7 PRA IMPLEMENTATION PLAN 1.2(d)-

Graded Quality Assurance Action Plan

, TAC Nos. M91429, M92447 Last Updata: 4/9/99 M92448, M92449, M88650, M91431, Lead NRR Division: DIPM M91432, M91433, M91434, M91435, Support Division: DSSA M91436, and M91437, M92420 and M94163 GSI: Not Available MILESTONES DATE (T/C) I 1.' lasued SECY-95-059 03/95C

2. ' Begin interactions with volunteer licensees 05/950 Palo Verde letter dated 4/6/95 Grand Gulf meeting 5/4/95 South Texas meetings on 4f19/95 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 cn 12/4/95 -

Meeting on 4/11/96 to di-scuss the staff evaluation guide Letter from licensee on 4/24/96 providing comments on staff L evaluation guidance Site visit on 6/5-6/96 to observe expert panel and review revised procurement QA controls, trip report sent to licensee on 8/6/96 -

Letter from licensee on 9/12/96 transmitting responses to procurement issues raised in earlier staff trip reports Letter from licensee dated 11/13/96 responding to PRA issues raised in 12/4/95 trip report ~

Overview of GQA initiative provided by PVNGS at 2/27/97  ;

meeting with staff - l GOA close_out letter transmitted to licensee on 7/2/98

5. Staff interactions with South Texas . Ongoing i Meeting on 7/17/95 on project status through l Site meeting on 10/3-4/95 on risk ranking and QA controls 3/98C {

, Meeting on 12/7-8/95 to discuss risk ranking and QA controls  !

South Texas Submittal of QA Plan for implementation of I graded QA, dated 3/28/96 is currently under staff review i 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 ouidance i

22 l

Meeting on 6/19/96 to discuss staff comments on the QA plan submittal for graded QA, 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 Additional Information (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 performence 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 RAl response STP submittal of draft QA Plan on 5/21/97 STP submittal of GQA 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 GQA implementing procedures and associated change control on 7/31/97 GTP submittal on 8/4/97 responding to PRA RAl and provided procedures related to shutdown operations Negative consent SECY paper (97 229, dated October 6, 1997) and Safety Eva!aation 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

1 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 I 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 Letter to GGNS dated 5/29/96 regarding implementation of OAP commitments l

)

Staff review comments on GGNS safety significance l 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 1 of QAP commitments for low-safety significant items, i 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 Woup Review

9. Issue letter to 3 volunteer plants outlining program objectives and review 1/96C expectations. Distributed staff evaluation guide to licensees.

10. Evaluation Guide issued for use by staff in evaluating volunteer plants Meeting held with volunteer plants to receive feedback on 1/96C staff evaluation guide on 4/11/96.

Industry comments on staff evaluation guide provided by 4/96C 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 j Public workshop held on draft RG 8/12/97C Publish final RG in SECY-98-067 4/2/98C SRM conditionally approves issuance of GOA RG 6/29/98C GOA final RG issued 8/98C i

24

12. ACRS Briefings Expert Panel and deterministic considerations 2/27-28/96C Graded QA 4/11/96C PRA Implementation Plan 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 concems 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

14. Issue draft Staff Inspection Guidance (Daseline + 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 concems with IP Issue final inspection procedure 6/99T

15. Conduct NRC Staff Training 7/99T Descriotion: 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 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 Backoround: 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 implementation 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 th3 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 w:re discussed and the initialindustry 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.

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 m;thods for implementing graded OA. The staff will develop a regulatory guide based in part on input 25

from Brookhaven National Laboratory, and will also prepare a baseline and reactive inspection procedure (IP) for graded OA. An inter-office team has been established to prepare the regulatory guidance documents and test their implementation during the evaluation of volunteer plant activities.

Oriainatino 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 learned 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 for inspections of the risk significance determination aspects of graded QA programs.

The safety benefits to be gained from a graded OA 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 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 beer 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 OA 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 comrnnt period on the risk-informed guidance documents has expired.

At this time,42 sets of comments have been received. A decision has been made, and accepted by the Chairman, to focus staff efforts on revising the general regulatory guide and standard review plan first.

The proposal to sequentially complete the application specific guidance documents, including GOA, 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 necesnry. 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 GOA 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

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 rueeting 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 Pquisite PRA expertise to implement the IP in a uniform manner.

2. The IP does not provide objective or deterministic standards for inspectors to make decisions on the adequacy of licensee programs. This could result in backtitting 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.

1 The CRGR recommended that the staff consider the development of plant-specific Temporary

(' Instructions to verify graded QA 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 concems.

Conversion of the IP to a Tl that would be specific to South Texas Project was considered, however, the l 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 staff plans to revise the draft IP to address generic OA issues and submit it for reconsideration by CRGR. The target date to issue the IP is June 18,1999.

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 stafi 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 OA 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 full advantage of PRA information. However, the ACRS acknowledged the inherent difficulty given the lack of a model to assess quar,titatively 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 OA controls on PRA parameters, and for the staff to prepare a plan for improvements to RG 1.176 after gaining enperience with its application and to brief the committee within the next 2 years.

27

South Texas submitted their QA program revision for their graded QA effort on March 28,1996. The change has been reviewed by the staff (lOMB, 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. Staff 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 IOMB and SPSB) performed a site evaluation during the week of May 5-8 to review aspects l

associated with: PRA quality, OA centrols 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 the 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 facility. 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 discontinuaniof certain technical provisions. STP stated that for low and non-risk significant, safety- '

related equipn.erMhat exemptions are warranted to remove those components from the scope of seismic and environmelital 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 performed 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 f 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 OA program implementation. NEl has proposed that the amended 96-02 guidelines will be submitted to the staff for endorsement by a regulatory guide. A subsequent letter was received from NEl on July 16 that provided an updated version of NEl 96-02 based on comments they 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 concern with the graded QA initiative. NEl stated their concerns regarded the questions raised by the staff in the area of QA controls for items determined to be low safety significant and in the area of safety 28

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) Regulatory Guide 1.160.

3) NUMARC 93-01, ".. dustry 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 f rom 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

ACCIDENT MANAGEMENT IMPLEMENTATION TAC #: M91966 - Overall Last Update: 4/7/99 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 6/99T

2. Review severe accident training materials and BWROG prioritization 6/95C methodologies

3. Develop guidance for A/M audits initial draft (for intemal use) 11/95C industry, sponsored A/M demonstrations 3/98C Revised draft (to NEl and public) 8/98C Final TBD

4. Commission paper regarding inspection and oversight of voluntary 8/99T programs

5. Complete A/M audits TBD

6. Hold public workshop TBD

7. Report to Commission on audit findings and recommendations for TBD achieving closure Descriotion: This action plan is intended to guide staff efforts to assess the quality of utility implementation of accident management (A/M), and the manner in which insights from the IPE program have been incorporated into the licensees A/M program. Specific review areas will include: development and implementation of plant-specific severe accident management guidelines (SAMG), integration of SAMG with emergency operating procedures and emergency plans, and incorporation of severe accident information into training programs.

Historical Backoround: The issue of A/M and the potential reduction in risk that could result from 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 activities in late 1989. Completion of A/M is a major remaining element of the Integration Plar' 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 l 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 commitment with the understanding that the staff would inspect utility implementation [7).

Proposed Actions: Specific actions included in the A/M action plan are: (1) complete the review of BWROG SAMG documents, (2) conduct A/M demonstration visits to observe how the elements of the

, formal industry position are being implemented, (3) complete A/M audit guidance using the information l and perspectives obtained through the demonstration visits, (4) conduct A/M audits, and (5) hold a public workshop to discuss audit findings. Following the workshop, the staff will report to the Commission on audit findings and recommendations for remaining actions to achieve closure.

30

Oriainatina Document: SECY-88-147, Integration Plan for Closure of Severe Accident issues, May 25,1968.

Reautatorv 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 additionalinformation 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 NP.0/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 concerns. The BWROG has evaluated the staff concerns and expects to provide a written response to the NRC by May 1999.

Utilityimplementation Licensee commitments and target dates for complet!ng A/M implementation were submitted to NRC in 1995 as part of the industry initiative. Irnplementation 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 22,1997, letter [16]. NRC provided comments on this guidance in a January 28,1998, letter to NEl [17]. In 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 concern 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.

Implementation has now been completed at almost all plants. Two utilities have rescheduled their completion dates to the end of 1999. For about 20 other sites, licensees have not provided letters confirming that implementation has been completed. Staff will follow up with those sites 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 irmluded: (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, 31

(3) performing pilot inspections at about five plants using the TI, (4) developing an inspection procedure (IP) 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 regionalinitiative.

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 Amold (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 Tl. 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.

The staff committed to carefully consider the industry proposal, and provide a response to NEl.

The staff is planning to prepare a Commission paper addressing the manner in which voluntary programs, such as A/M and shutdown risk, should be included in the risk-informed inspection and assessment process. The question regarding A/M audits and oversight will be addressed in this paper.

References:

1. Memorandum from F. Rowsome to W. Minners, "A New Generic Safety issue: Accident Management," April 16,1985

2. SECY-88-147, Integration Plan for Closure of Severe Accident issues

3. SECY-95-079, implementation Plan for Probabilistic Risk Assessment

4. SECY-89-012, Staff Plans for 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 from W. Russell to W. Rasin, January 9,1995

8. Letter from W. Russell to W. Rasin, February 16,1994

9. Letter from K. Donovan to Document Control Desk, August 29,1996

10. Letter from D. Matthews to K. Donovan, April 2,1997

11. Letter from K. Donovan to Document Control Desk, May 10,1997

12. Letter from T. Rausch to Document Control Desk, January 9,1998

13. Letter from T. Essig to T. Rausch, July 20,1998

14. Letter from A. Thadani 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, DSSA,415-1095 NRR Lead PM: Ramin Assa, DLPM,415-1391 33

CORE PERFORMANCE ACTION PLAN Final Update TAC Nos. M91257 - DSSA Status: COMPLETE M91602 - DIPM Lead NRR Division: DSSA GSI: LI-179 Supporting Division: DIPM Descriotion: The action plan covered the assessment of the impact of reload core design activities on fuel and control rod reliability affecting plant safety. The activities were accomplished through performance-based inspections of fuel vendors, evaluation of licensees' reload analyses, supplemented with independent evaluation of core performance information, and coordinated with regional training and interaction.

Histoncal Backaround: The action plan addressed a review of fuel fabrication, reload core design, and reload analysis issues that were discussed during 1994,1996, and 1997 briefings given to the Executive Director for Operations (EDO), and at the June 19,1997 Chairman briefing. The briefings presented by the Reactet Systems Branch (SRXB), Division of Systems Safety and Analysis (DSSA), covered generic fuel and core performance issues and related evaluations of fuel failures. The former Special inspection Branch (PSIB), Division of Inspection and Support Programs (DISP), supported the briefings. As a result of these briefings, the Office of Nuclear Reactor Regulation (NRR) was directed to focus on licensee activities and the licensee / vendor interfaces.

Oriainatina Document: Memorandum from Gary M. Holahan and R. Lee Spessard to Ashok C. Thadani, dated October 7,1994, " Action Plan to Monitor, Review, and improve Fuel and Core Components Operating Performance" and the enhanced focus on licensee reload design participation resulting from briefing feedback.

Reaulatory Assessment: Core design is a fundamental component of plant safety because maintaining fuel integrity is the first principal safety barrier (i.e., fuel cladding, reactor coolant system boundary, or the containment) against serious radioactive releases. Likewise, the safety analyses must be properly performed in order to verify, in conjunction with startup tests and normal plant parameter monitoring, that the core reload design is adequate and to provide assurance that the reactor can safely be operated.

Evaluation of activities that affect the quality of fuel and core components are important to ensure that safety and quality are not degraded and that the core performs as designed.

Resolution: An action plan was developed to: (1) evaluate fuel vendors' quality assurance (QA) performance through performance-based inspections that evaluate the reload core design, safety analysis, licensing process, fuel assembly mechanical design, and fuel fabrication activities; (2) evaluate the QA performance of licensees that perform core reload analysis functions; (3) identify, document, and categorize core performance problems and root cause evaluations that will be further evaluated duririg these inspections and provide input to licensee evaluations as well as support regional enforcement actions, as appropriate; (4) train and coordinate regional support staff participating in these activities; and (5) evaluate the results of these activities for use in formulating generic communications, revisions of regulatory guidance and guidance for regionalinspectors, and other appropriate regulatory actions. A Core Performance Workshop was conducted on October 24-25,1996, to inform the industry of generic issues, and solicit feedback. The status of core performance inspection evaluations and emerging issues has been covered at the yearly Regulatory information Conferences.

The action plan included ten vendor inspections, including all five domestic reload fuel vendors and their major fuel cladding suppliers, that were performed by multi-disciplined inspection teams led by lOMB with contracted technical assistance, as required. These inspections are documented in the individual vendor inspection reports. The action plan also included selected licensee inspections in each region, with the i regional team leaders, to assess licensee performance in reload core analysis vendor oversight and l participation. Licensee inspections were normally issue-driven.

34

The data acquired through licenseekendor inspections has been integrated with information supplied by the regions and other sources and is evaluated for generic core performance indicators and industry conformance to current regulatory requirements, as part of our normal activity. One end product of the on-go;ng assessment is continuing interaction and guidance for resident inspectors and regional staff.

The ongoing activities to capture and address early warning of emerging core performance issues will also continue to evaluate vendor / licensee interface issues and potential generic problems, with our continuing assessment of core performance under the DSSA operating plan.

The data acquired from the completed vendor and licensee inspections have been evaluated for generic impact and identification of emerging issues. Results from recent NUPIC Nuclear Fuel Committee joint vendor surveillances and audits have been reviewed to supplement our vendor evaluations. We are attending the NUPlc Nuclear Fuel Committee vendor specific Affinity Group annual meetings, and evaluating options for sharing vendor oversight results. DSSA is evaluating the action plan results to I

better integrate and prioritize its on-going activities, consistent with the DSSA operating plan. Options and recommendations are being prepared for management review to capture the lessons learned and to provide ongoing interaction and guidance for regional staff. This was documented in a memorandum from Jared S. Wermiel to Gary M. Holahan," Closeout nf Core Performance Action Plan," dated l February 16,1999.

The action plan activity to review the adequacy of vendor lead testing programs for new fuel designs has been transferred to the Agency Program Plan for High Burnup Fuel, issued on July 6,1998.

NRR Technica' Contacts: E. Kendrick, DSSA/SRXB,415-2891 G. Cwalina, DIPM,415-2983 35

HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN (Previously Part of the Dry Cask Storage Action Plan)

TAC Nos. M93821: Action Plan Last Update: 04/09/99 M91955: DSC generic review Lead NRR Division: DSSA M95546: Generic review of NRCB 96-02 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. 2/96C Single-Failure-Proof Crane (reliability). TBD*

Spent fuel cask drop accident prior to cucuring the lid. 2/96C 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 TI). 12/99

8. Obtain feedback (meeting, FRN, or other means) conceming the statt 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 finalinspection procedures. TBD

11. Issue final guidance. TBD*

Note: ' Indicates that the activity is contingent on the results of NRR's/RES's review of NRR's proposed GSI on the potential risk and consequences of heavy load drops in nuclear power plants.

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).

36

Historical Backaround: R:: cent 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 l 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 fra mework and inform the staff; (2) review the general issues and identify specific problems to be addressed; p) develop corrective actions to resolve the problems; and (4) implement the corrective actions. Specific cc. rective 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.

Oriainatino Documenj: Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995, " Dry Cask Storage Action Plan."

Reaulatory 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 transler casks for transfer to ISFSis) over safe shutdown equipment, and over spent fuel during plant operation. Questions have been raised regardi1g 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 responses 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 failures were abandoned in early 1998 due to budget shortfalls. The staff will propose to RES (ETC 04/99), 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. The heavy loads issue (USl 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 mid-1999.

The staff visited Calver1 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 37

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."

l l-38

WOLF CREEK DRAINDOWN EVENT: ACTION PLAN TAC Nos.: M66278 Last Update: 03/31/99  ;

Lead NRR Division: DSSA l MILESTONES DATE (T/C) ,

1

1) Draft Generic Letter (GL) 11/95(C) l

2) issue Supplement to IN 95-03 03/96(C) i

3) Complete Draft Tl/ Issue to the Regions for Comments 12/98(C)

4) CRGR Concurrence of the GL for 1st time 09/96(C)

CRGR Concurrence of the GL for 2nd time (after reconciling Public Comments) 01/98(C)

5) ACRS Briefing 11/97(C)

6) GLissued 05/98(C)

7) Receive Regional Comments on Tl 1/99(C)

8) Complete Evaluation of the Responses to the GL 03/99(C)

9) Issue Tl 04/99(T)

Description:

The objective of this action plan is to collect and evaluate information from the licensees I r:garding 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 v;lve, 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 adecuately control two incompatible act;vities resulted in transferring 9200 gallons of hot RCS water to '.he RWST in 66 seconds.

The Wolf Creek event represents a LOCA with the potential to consequentially fail all the ECCS pumps and bypass the containment. Another important feature of this event is the short time available for corrective action. Based upon calculations by the licensee and the staff, it is estimated that if the draindown had not been isolated within 3-5 minutes, net positive suction head would have been lost for Ell 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.

Oriainatina Document: AEOD/G95-01, " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994".

39  ;

1 l

l l Reoulatorv Assessment: Th3 staff perform d an evaluation of th3 probibility for evsnt initiation and of i

the conditional core damage probability. The value of this probability for co,e damage along with licensee awareness for this scenario makes the risk for continued PWR oneration acceptably small.

I Current Status: Information 100tice IN 95-03, and its Supplement was issued. CRGR concurred the proposed GL in 9/96 for the first time; but as directed by an SRM, the GL was published in the Federal Register in 2/97 for public comments. ACRS was briefed on 11/6/97. 2nd CRGR concurrence was obtained in 1/98 after reconciling the public comments. The GL was issued on 5/28/98 after reconciling

~

611 the comments. Staff issued the draft Tl to the Regions for comments on 12/98. Comments from all the Regions have been received. The staff does not anticipate any significant problems in incorporating the comments from the Regions, and is currently modifying the dra't Tl in order to reconcile the comments from the Regions.

NRR Technical

Contact:

M. M. Razzaque, SRXB,415-2882 NRR Lead PM: Kristine Thomas, NRR,415-1362

References:

1) AEOD/S95-01," Reactor Coolant System Blowdown at Wolf Creek on September 17,1994."

2) IN 95-03, issued January 18,1995.

3) Suoplement 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.

l 1

l l 40 l

l l

r NEW SOURCE TERM FOR OPERATING REACTORS TAC No. M89586 Last Update: 04/08/99 GSI No.155.1 Lead NRR Division: DSSA 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. Publication 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

10. Present Commission Paper in E-Team Briefing 09/96C l

11. Brief CRGR on Commission Paper 10/96C

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 l

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) 07/99T

23. Finish Third Pilot Plant Review (Indian Point Unit 2) 07/99T

24. Finish Fourth and Fifth Pilot Plant Review TBD 41

f

Description:

Mors thin a decide of r: search his led to en enhancid und:rstanding of tha 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 j margin. In addition, safety enhancements may also be achieved.

Historical Backaro.u.Lnd: 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 TID-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 I 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 dote 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 inetitute (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.

ProDosed 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 rulemaking for the attemative source term.

Oriainatino 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.

Reculatory Assessmen_t: There will be no mandatory backfit of the new source term for operating  ;

reactors. The design-oasis accident analyses for current reactors based on the TID-14844 source term 1 are still valid. Therefore, non-urgent regulatory action and continued facility operation are justified. i Current Status: NEl submitted its generic framework document in November 1995 for NRC review and approval. TVA submitted 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 diseJss the staff's plan to issue exemptions while pursuing rulemaking, and on April 2,1997, to provide a s.atus report on the staff's actions regarding rebaselining and rulemaking subsequent to the Commis : ion's SRM. The pilot plant applications for 1 Browns Ferry, Peny, 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 l term insights outlined in NUREG-1465; NRR believed that the rulemaking process can be initiated prior to  !

the completion of rebaselining, 42

r 1 I

Tha st;ff briefed the NRR Executiva Term 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, Perry, Oyster Creek, and Indian Point-2). An application is also expected from Grand Gulf. In addition, the staff and Virginia Power met on November 26,1996, March 25 and June 18,1997, to discuss the i

rebaselining of Surry; the staff and Entergy met on August 29,1996, and March 27.,1997, to discuss the r:baselining of Grand Gulf. In a February 12,1997, SRM, the Commission approved the Option 2 cpproach 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 totivities involved with rebaselining; as a consequence of that briefing, the user need memorandum r2garding 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

' November 1998, NRR and RES discussions and shifts in lead technical responsibility resulted in an improvement in the schedule. At a Commissioners' Technical Assistants briefing on October 9,1997, the Task Force Lehder outlined a new schedule that would result in the completion of rebaselining and the - ,

rulemaking pian in June 1998; this was accomplished by reversing the lead responsibilities (RES is now i the lead for rebaselining and NRR is now the lead for rulemaking and regulatory guidance). The~ l 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 rebaselining. In addition, NRR transferred its technical assistance resources with SNL, ORNL, and PNNL that were designated for rebaselining to RES. These changes will be reflected in the next revision to the NRR Operating Plan. On ,

November 13,1997, January 7,1998, February 24,1998, and March 30,1998, RES presented its four-i phased plan and preliminary findings from Phase I, Phase ll, and the DBA portion (with the updated 1 assumptions) of Phase 111, respectively, for the rebaselining effort. On April 1 and 2,1998, RES and NRR staff briefed the ACRS and DONRR, respectively, on the progress of the rebaselining effort, initial insights from the assessments completed, and the essential elements of the Rulemaking Plan. The results of the r:baselining 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 entry in Attachment 2.

The staff completed its review of the first pilot plant application in February 1999 and issued a safety cvaluation and its license amendment for Perry plant in March 1999. The staff is currently reviewing Grand Gulf and Indian Point Unit 2 pilot pla,1t applications and the staff expects to complete its review by July 1999. Browns Ferry and Oyster Creek requested to have their pilot plant anplication reviews "on hold."

NRR Technical

Contact:

J. Lee, SPSB,4151080

References:

NUREG-1465, " Accident Source Term for Light Water Nuclear Power Plants," February 1995.

July 27,1994, letter to A. Marion, NEl, from D. Crutchfield, NRC, " Application of New Source Term to Operating Reactors".

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".

43

Drosmber 22,1995, pilot plant submittal, litt:r to Documsnt Control D:sk from Tenn sste 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 50, 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 12,1955; 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; e dated April 24,1997, for public meeting with BWR (Grand Gulf) held on March 27,1997; i e dated May 8,1997, for public meeting with NEl held on April 2,1997; e dated July 28,1997, for public meeting with PWR (Surry) held on June 18,1997.

i l

44

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