ML20195K223

Director Status Rept on Generic Activities,Action Plans, Generic Communication & Compliance Activities

ML20195K223
Person / Time
Issue date:	10/31/1998
From:	NRC (Affiliation Not Assigned)
To:
References
NUDOCS 9811250209
Download: ML20195K223 (58)

Text

- . . . - - _ - - - . _ - - _ - - - - . - _ _---.

1 DIRECTOR'S STATUS REPORT l

on GENERIC ACTIVITIES Action Plans l Generic Communication and Compliance Activities 0

OCTOBER 1998 q d Office of Nuclear Reactor Regulation 7 g12 09 981031

@r_

MISC pg g z n ,rg 4.:1,l ,

a l

INTRODUCTION The purpose of this report is to provide information about generic activities, including generic communications, under the cognizance of the Office of Nuclear Reactor Regulation. This report, which focuses on compliance activities, complements NUREG-0933. "A Prioritization of Generic Safety issues."

This report includes two attachments: 1) action plans and 2) generic communications under development and other generic compliance activities. Generic communications and compliance activities (GCCAs) are potential generic issues that are safety significant, require technical resolution, and possibly require generic communication or action.

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 plasm 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 cf cognizant staff.

l 1

i 1

ATTACHMENT 1 4

0 3

1 i

4 i

i 4

NRR ACTION PLANS i

l 1

1 1

i TABLE OF CONTENTS l BOILING WATER REACTOR INTERNALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1

G RI D R E LI AB I LITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 PRA IMPLEMENTATION ACTION PLAN 1.2 (c) Inservice Inspection Action Plan . 8 STE AM G EN E R ATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 NEW SOURCE TERM FOR OPERATING REACTORS . . . . . . . . . . . . . . . . . . . . 16 ENVIRONMENTAL SRP REVISION ACTION PLAN . . . . . . . . .............. 20 PRA IMPLEMENTATION PLAN 1.2(d) Graded Quality Assurance Action Plan . 22 EXTENDED POWER UPRATE ACTION PLAN . . . . . . . . . . . . . . . . .......... 29 ACCIDENT MANAGEMENT IMPLEMENTATION . . . . . . . . . . . .............. 31 CO R E P E R FO R M AN C E ACTION P LAN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN . . . 37 HIGH BURNUP FUEL ACTION PLAN Final Update . . . . . . . . . . . . . . . . . . . . . . . 40 WOLF CREEK DRAINDOWN EVENT: ACTI O N P L AN . . . . . . . . . . . . . . . . . . . . 42

l I6 BOILING WATER REACTOR INTERNALS TAC Nos. M91898, M93925, M93926, M94959, M94975, M95369, Last Update: 09/22/98 M96219, M96539, M97373, M97802, M97803, M97815, M98266, Lead NRR Division: DE i M98708, M98880, M99638, M99870, M99894, M99897, M99898, Supporting Division: DSSA M99895, M99897, MA1102, MA1104, MA1138, MA1226, MA1926, GSI: Not Available l MA1927, MA2326, MA2328, MA3395 l MILESTONES DATE (T/C) -

i PART 1: REVIEW OF GENERIC INSPECTION AND EVALUATION CRITERIA -

l 1. Issue summary NUREG-1544 03/96 C l

o Update NUREG-1544 30/FY99 T

2. Review BWRVIP Re-inspection and Evaluation Criteria 3

o Reactor Pressure Vessel and intemals Examination Guidelines '

(BWRVIP-03) . ..... ......... ... ... ...... .... 06/08/98 C o BWRVIP-03, Section 6A, Standards for Visualinspection of Core  ;

Spray Piping. Spargers, and Associated Components . ..... 06/08/98 C i o BWR Vessel Shell Weld Inspection Recommendations (BWRVIP-05)N ....... .. .... .. .. ... ... .. 07/28/98 C o Guidelines for Reinspection of BWR Core Shrouds (BWRVIP-07) 04/27/98 C

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

! 4. Review generic mitigation guidelines and criteria TBD l 5. Review of generic NDE technologies developed for examinations of BWR TBD i intemal components and attachments M The Commission, in SRM M9705128 dated May 30,1997, requested that the staffs SER (a) should address the I ' BWRVIP proposal to examine 100 percent of the axial welds which would include examinations of some circumferential weld lengths near the intersections of the weld types to determine if this proposal could provide an appropriate level of sampling of the circumferential welds, (b) should provide a comprehensive evaluation of the probabilistic analysis l contained in the BWRVIP proposed alternative in determining the acceptability of a proposed technical alternative and/or l

In pursuing changes to the rule. (c) should consider a tiered approach in gathering additional baseline information and/or implementing the rule and, (d) should receive appropriate review, including review by ACRS. The industry was not timely

'in completing its RAI responses, and this resulted in extending the previously established schedule. The NRC staff completed it s safety evaluation of the BWRVIP-05 report and presented the completed SE to CRGR and ACRS for their review and concurrence, which was given by letter from the ACRS dated July 21,1998, and by memorandum from the i CRGR dated September 15.1998.

I 1

L

6. Other Intemals reviews (saf ety assessments, evaluations, mitigation measures, inspections and repairs) l o Safety Assessment of BWR Reactor Intemals (BWRVIP-06) . 09/15/98 C o Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity issues (BWRVIP-08 & BWRVIP-46) . ........, .. 03/27/98 CA

, o Evaluation of Crack Growth in BWR Stainless Steel RPV Intemals (BWRVIP 14) . . . ... . ........ .......... . . 06/08/98 C o Intemal Core Spray Piping and Sparger Replacement Design i Criteria (BWRVIP 16, . ...... . ... . . . .... .. 10/30/98 T o Roll / Expansion of Control Rod Drive and In-Core Instrument l Penetrations in BWR Vessels (BWRVIP-17) .. . .. .. 03/13/98 CD l o BWR Core Spray intemals Inspection and Flaw Evaluation l

Guidelines (BWRVIP-18) . . . .. . .... . . . 06/08/98 C o BWRVIP 18. Appendix C. BWR Core Spray Intemals Demonstration of Compliance With Technicallnformation Requirements of License Renewal Rule (10 CFR 54.21) . 12/98 T o Intemal Core Spray Piping and Sparger Repair Design Criteria (BWRVIP-19) .... . . .... . . ..... .. . .. . 10/30/98 T o Core Plate Inspection and Flaw Evaluation Guideline (BWRVIP-25) . ... .. .. ..... . . . . . . .. . 10/30/98 T o Top Guide Inspection and Flav Evaluation Guideline (BWRVIP-26) .. .. . . . . ... . 10/30/98 T o Standby Liquid Control System / Core Plate AP Inspection and Flaw Evaluation Guidelines (BWRVIP-27) .... , 10/30/98 T o Assessment of BWR Jet Pump Riser Elbow to Thermal Sleeve Weld Cracking (BWRVIP-28) . . . . . . . . . . . . . .. .. . . 10/30/98 T o Technical Basis for Part Circumferential Weld Overlay Repair of Vessel Intemal Core Spray Piping (BWRVIP-34) ..., .. ... TBD o Shroud Support inspection and Flaw Evaluation Guidelines (BWRVIP-38) . . . . . .. ...... ..... .. . .. ........ 03/30/99 T o BWR Jet Pump Assembly inspection and Flaw Evaluation Guidelines (BWRVIP 41) . . . .. . .. . ....... ... 10/30/98 T o BWR LPCI Coupling Inspection and Flaw Evaluation Guideliaes (BWRVIP-42) . . . ... .. ..... .. . . .. . . 12/98 T o Update of Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity issues (BWRVIP-46) . ..... . . . 03/27/98 CA o BWR Lower Plenum Inspection and Flaw Evaluation Guidelines (BWRVIP-47) . . . . . .. . . .. .... . ... .. 12/98 T o VesselID Attachment Wold Inspection and Flaw Evaluation Guidelines (BWRVIP-48) ... .. .. . .. . 03/30/99 T o Instrument Penetration Inspection and Flaw Evaluation Guidelines (BWRVIP-49) . .. .. .... ... . . .. 03/30/99 T o Top Guide / Core Plate Repair Design Criteria (BWRVIP-50) . 03/30/99 T o Jet Pump Repair Design Criteria (BWRVIP-51) 03/30/99 T o Shroud Support and Vessel Repair Design Criteria (BWRVIP-52) 03/30/99 T l o Standby Liquid Control Line Repair Design Criteria (BWRVIP-53) 03/30/99 T Descriotion: Many components inside boiling water reactor (BWR) vessels (i.e.. intemals) are made of l 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 l 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 2

e intamals. This includes plant specific reviews and the assessment of the generic criteria that have been

! proposed by the BWR Owners Group and the BWRVIP technical subcommittees to address lGSCC in core

! shrouds and other BWR intemals.

Historical Backaround: Significant cracking of the core shroud was first observed at Brunswick, Unit i 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 proVde an analysis justifying continued operation untilinspections 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 instrumentalin 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 until inspection or repair on the basis of: 1) no 360' through-wall cracking observed to date,2) low frequency of pipe breaks, and 3) short period of operation (2-6 months) before all of the highly susceptible plants complete repairs of or inspections to their core shrouds.

In late 1994, extensive cracking was discovered in the top guide and core plate rings of a foreign rr. 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 tha'. the ring cracking could occur in GE BWR. with operating time greater than 13 years, in the special industry review group's report, that was issued in January 1995, ring cracking was avaluated. The NRC concluded that the BWRVIP's assessment was acceptable and that top guide ring and coro olate ring cracking is not a short term safety issue.

Prooosed Actions: The staff will continue to assess 'he 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 willissue 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 individuallicensees.

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 chrouds by the next outage and to justify continued safe operation until inspections can be completed.

Reaulatory Assessment: In July 1994, the NRC issued Generic Letter 9443 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 i 3

O 4

1 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 intemals and attachments to assure continuing safe operation.

Current Status: Almost all BWRs completed inspections or repairs of core shrouds during refueling outages in the fall of 1995. Various repair methods have been used to provide attemate 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 indicates 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 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 submittod 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 intemals Examination Guidelines (BWRVIP-03)* is complete and the staff's SE has i been issued. I By letter daled 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 I we'ds. 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 I independent assessment of the BWRVIP-05 report was transmitted by letter dated August 14,1997, to the l BWRVIP, along with a request for additional information and information that neoded 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 plans to issue the final GL in the near term.

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 continuing.

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

4

l d

The BWRVIP submitted BWRVIP-28 to address the safety implications of recent cracking found in BWR jet I pump riser elbows. The staff is reviewing the BWRVIP-28 report. The staff issued NRC hformation Report  !

IN 97-02, " Cracks Found in Jet Pump Riser Assembly Elbows at Boiling Water Reactors," on February 6, l 1997, '

information Notice 97-17, " Cracking of Vertical Welds in the Core Shroud and Degraded Repair," was issued April 4,1997, to inform the industry of vertical weld cracks and a degraded core shroud repairs found at Nine Mile Point, Unit 1. The BWRVIP has informed the staff that it plans to revise BWRVIP-07 to l 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 send sictions, including ,

implementing the BWRVIP topical reports at each BWR as appropriate considering individuai p'cnt .

i 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.

@R 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 LeMer 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors,"

July 25,199 t.

Action Plan dated April 1995.

5

i GRID RELIABILITY TAC No. M98444 Last Update: 9/25/98 Lead NRR Division: DE Supporting Division: DSSA MILESTONES DATE (T/C)

1. Assess and evaluate the risk significance of potential grid instability due to deregulation.

a. Survey past and expected electric gM performance (EELB) 9/98 (C)

b. Assess projected risk from grid-centered loss of offsite power events 1/99(T)

c. Inform Commission (RES) 3/99(T)

2. Monitor industry deregulation and implement mechanisms to institutionalize ongoing staff level contacts. (EELB)

a. Conduct meetings with NERC, Regional Reliability Councils and ONGOING FERC/ DOE

b. Develop reliability assesstnent tool 7/98 (C)

c. Inform Commission (RES) 3/99 (T)

3. Issue generic communication (EELB)

a. Draft Generic Communication 11/97 (C) b.: Office Concurrences (if necessary) 2/98 (C)

c. ACRS Review (if necessary) N/A

d. CRGR Review (if necessary) N/A

e. EDO Concurrence (if necessary) N/A

f. Commission Approval (if necessary) N/A

g. Issue Generic Communication 2/98 (C)

4. Evaluate based on Task 1 the need for regulatory actions. Evaluate method (s) to identify grid-centered event precursors. Evaluate the impact of deregulation on SBO risk reduction goals.

a. Review AEOD study for implications regarding grid-center events 2/98 (C)

(EELB)

b. Complete feasibility study on methods to identify grid-centered events N/A (RES)

c. Assess the implications of grid-centered events to SBO risk redixtion 1/99 (T) goals (RES)

d. Determine what additional regulatory actions are necessary (RES) 2/99 (T)

e. Inform Commission (RES) 3/99 (T)

Description The action plan is intended to address the Commission's concems regarding the impact of utility deregulation on the reliability of the electric grid to supply offsite power to nuclear power plants for safe operation.

Historical Background in recent years, two relatively new factors are emerging: non-utility generation and deregulation. It is anticipated that, in the not too distant future, power suppliers, whether utilities or independent power producers, will actively compete for sales to customers who may be located anywhere on the power grid. Regional grid control would be the responsibility of centralized independent System Operators (ISOs). The responsibilities and authority of an ISO have yet to be defined, but it is expected 6

l that they will be charged with maintaining grid reliability to facilitate the marketing of power, it is also l

I uncertain how, or even whether, the current method of maintaining relinbility through voluntary compliance with guidelines established by consensus associations will transition to the new utility structure. These uncertainties raise questions with respect to the continued supply of reliable offsite power to nuclear power plants.

Prooosed Actions: Specific actions included in the action plan are: (1) issuing generic communications to l reemphasize the need for licensees to maintain their design basis with respect to the stability and reliability l of offsite power and to maintain a process for ensuring that they continue to meet their design basis for the remainder of their license; (2) monitoring industry deregulation developments and its impact on the reliability of offsite power to nuclear power plants; (3) assessing and evaluating the risk signifcance of potential grid instability due to deregulation; and (4) reassessing the risks and effectiveness of SBO issue resolution efforts due to grid-centered loss of offsite power event initiators.

Oriainatina Document: AEOD C97-01, Grid Performance Factors, dated March 20,1997; Staff Requirements Memorandum dated May 27,1997.

Regulatory Assessment Based on the rapid changes in utility deregulation, the Commission requested that the staff give greater urgency to ensuring that related health and safety issues within NRC's jurisdiction are addressed, particularly in reviewing the terms of the licensing basis and validating grid reliability assumptions, Given that there is no evidence at this time that the reliability of the grid is degraded, continued operation is justified.

Current Status Contract has been authorized for Oak Ridge National Laboratories (ORNL) to begin work and data collection trips with staff are in progress. Twelve out of twelve visits completed as of 5/29/98.

ORNL has submitted the final report to support Milestone 1.a. In accordance with SECY-97-220 (DSI 22)

DSSA transferred the responsibility to complete subject risk studies (Milestones 1.c and 4.c) to RES on September 10,1998. Given the actions completed to date and the transfer of responsibility of future tasks to RES the staff will take steps to close the subject Action Plan by 10/98.

NRR Technical

Contact:

Ronaldo Jenkins,415 2985 NRR Lead PM Chester Poslusny,415-1402 References-(1) SECY-97-248,"Infomiation on Staff Actions to Address Electric Grid Reliability issues," October 23, 1997.

(2) Memorandum from S. J. Collins to H. J. Thompson, "Rebaselining Chairman Tracking List,"

March 11,1998.

i 7

k j

l PRA IMPLEMENTATION ACTION PLAN 1.2 (c)

Inservice inspection Action Plan

. TAC Nos. M95125, Last Update: 9/23/98

M97153, M99389 Lead NRR Division: DE

< M99756, MA0125, Support Division: DSSA, EMCB l MA0867, MA0868

} RG/SRP MILESTONES DATE (T/C)

! 1. Draft for RI-ISI team review / comments 04/05/96 C

2. First draft for Branch Chiefs review / comments 08/14/96 C

3. Revised draft for Branch Chiefs review / comments 01/24/97 C l 4. Revised draft for Branch Chiefs review / comments 04/08/97 C i 5. Draft for Division Director review / comments 04/29/97 C j 6. Draft for Office Director /OGC review / comments 05/16/97 C j 7. Office Director /OGC concurrence 07/08/97 C l 8. Draft for CRGR review / comments 07/08/97 C

9. Draft for ACRS review / comments 06/03/97 C

{ 10. Initial presentation to ACRS full Committee 06/11/97 C

11. Initial presentation to CRGR 06/11/97 C l 12. Meeting with ACRS Subcoir vi;11ee 07/08/97 C

! 13. Meeting with ACRS full Commdtee 07/09/97 C

14. Meeting with CRGR 07/17/97 C j 15. SECY from EDO to Commissioners (SECY 97-190) 08/20/97 C

16. Publish draft for public comments 10/15/97 C

{ 17. Public comment period for draft RG/SRP ends 01/13/98 C

18. Public Workshop 11/20/97 C j 19. Complete draft for ACRS/CRGR review / comments 04/98 C

20. Complete draft for Inter-Office concurrence 05/98 C l 21. Issue RG/SRP for trial use by the staff 06/98 C

22. lasue final RG/SRP 06/99 T l WOG TOPICAL REPORT MILESTONES DATE (T/C)

J

1. TechnicalMeeting 9/22/98 C

2. WOG Commitment Letter 10/15/98 T l 3. Issue FSER 1/99 T l

4 l

i

)

i i

8 1

3 yw et- -- - *-s.,-

__ . - - - _ ~. - _ -- - _. - . . - - . . - -

d EPRI TOPICAL REPORT MILESTONES

• DATE (T/C)

1. Issue RAls to EPRI 6/12/97 C  ;

2. EPRI Response to RAls 10/31/98 T

3. Open items Technical Meeting TBD

4. Receive Revised Report from EPRI TBD l S. Issue FSER TBD EPRI Topical Report review milestones cannot be established due to lack of any input from EPRI.

PILOT PLANT REVIEW MILESTONES " DATE (T/C)

1. Issue FSER Vermont Yankee 11/30/98 T

2. Issue FSER Surry 12/31/98 T l

3. Issue FSER ANO-2 12/31/98 T

4. Issue FSER ANO-1 7/31/99 T Subject to change based on licensees

• actual submittal dates and responses to Staff RAls.

INSPECTION PROCEDURES MILESTONES DATE (T/C)

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

2. Issue FinalInspection Procedure Number 73753 6/98 C Descriotion: 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). 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. The action plan describes the process for the review of Rl-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.

Historical Backaround: On August 16,1995, the U.S. Nuclear Regulatory Commission (NRC) published a policy statement (60 FR 42622) 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 Regulation / Nuclear Regulatory Research (NRR/RES) joint task group has been established to accomplish the above delineated specific tasks in the RI-ISI area as directed by Chairman Jackson.

The nuclear industry has submitted two methodologies for implementing RI-ISI. One methodology has been jointly developed by ASME Research and Westinghouse Owners Group (WOG)(Reference 4,6) and the other methodology is being sponsored by Electric Power Research Institute (EPRI) (Reference 5).

9

l l

ASME is working on three Code Cases for altemate examination requirements to ASME Section XI, Division i 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.

Prooosed Actions: The NRC has encouraged licensees to stmit pilot plant applications organized under one umbrella sponsoring organization, e.g., Nuclear Enngy Institute (NEI), for demonstrating risk-informed methodologies to be used for piping segmerN 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-IS). 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-lSI pilot plant programs will be documented in SERs for each of the pilot plant licensees and forwarded to the Commission. Upon Commission approval, the staff willissue SERs authorizing attemative inspection strategies for tne pilot plant licensees to allow use of the RI-ISI methodology.

ASME is working on three Code Cases for altemate examination requirements to ASME Section XI, Division 1 for piping welds. Code Case N-560, for the attenWe 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 attemate 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 pilot plant RI ISI program review will be docu nented in the staff SER.

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 all safety significant systems in the plant. Also, the Code Case is an abbreviated version and does not 10

have all the details presented in the EPRI topical report (Reference 5). The staff will review the EPRI methodology as well as Cods Case N-578 and the consistency of the ANO-2 RI-ISI pilot program to both of these. Code Case N-578 will be reviewed and, if found acceptable, will be endorsed by RG 1.147 with any necessary additions or deletions. The pilot plant RI-ISI program review will be documented in the staff SER.

Code Case N-560 for the attemate 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.-

l l The staff will use the acceptable altemstive provision of 10 CFR 50.55a (a)(3)(/)to approve the pilot plants' applications. The staff is working closely with ASME to expedite changes involving ISI.

Long-Range Plan i- 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. The staff plans to perform in-L depth reviews of pilot plant submittals and WOG and EPRI Topical Reports during the FYs 1998 and 1999 l in order to ensure that the RI-IS! programs are consistent with the staff RG and SRP This process may entail revisions in indust:y documents as well as in the staff RG and SRP.

i For the RI-ISI programs submittals subsequent to the approval of the pilot plant programs and topical l 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 Rl-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 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 loamed from pilot plants and topical report reviews. The ASME Code Cases will be endorsed by RG 1.147 with exceptions andor 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, the staff anticipates that the licensee will develop an RI-ISI program using the approved ASME Code Case. No NRC approval will be required, and the staff will oversee the acceptable implementation as part of the normal ISI inspection program.

l. For the non-pilot plant licensees that intend to implement RI-ISI starting with their next ten year interval, the staff is considering granting a relief from the current deterministic requirements of ISI of piping, of up i

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 IS!

program, in order to disseminate the information to the licensees, the staff is considering issuing an information notice.

Originating Documents In a November 30,1995, memorandum to J. M. Taylor, the NRC EDO, i Chairman Jackson directed that the staff prepare an action plan together with a timetable for developing i 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 4

11 I

l h

I , , . - - - ,.-,. _ , _ , - - . _ -,,.-...m . . ,,-

___ _ _ _ _ . _ _ . _ . _ __ ..__._____._______m.. . . _ _ __

4 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 f uture changes will be consistent with the overallplan.

Bagulptory 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 expetience 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 attemate examination requirements to the current 4

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 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 expected to be documented in safety evaluations. The staff recommendatic, whether to authorize an attemative inspection program pursuant to 10 CFR 50.55a (a)(3)(/) will be presented to the Commission prior to its implementation. 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 Rl-ISI methodology.

Current Status Since the formation of the Al-ISI team, several meetings have been held wie NEl and industry / utility representatives. In these meetings, the NRC staff and industry have discussed their respective plans for the Al-ISI programs. NEl has submitted WOG technical report (Reference 4), and EPRI technical report (Reference 5). The staff has also been actively participating in ASME Code activities related to Rl-ISI. NEl has submitted a revised WOG technical report (Reference 6) that addresses the staff's comments and requests for additional information (RAI). The staff has also submitted its comments and RAls on the EPRI report to the industry.

The staff completed final drafts for trial use of RI-ISI RG (RG 1.178) and SRP Section 3.9.8 which were submitted to the Commissioners (SECY 98-139) for information. The RG and SRP are expected to be issued during October 1998.

The staff has received pilot plant submittals from Vermont Yankee, Surry, and ANO-2. The staff review for these pilot applications is currently under progress.

NRR Contacts S. Ali(415-2776)

S. Dinsmore (415-8482)

RES Contact J. Guttmann (415-7732)

12

l l

l l

References:

1. Federal Register, Vol. 60, No.158, "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement," August 16,1995. 1 l 2. Memorandum from Shirley Ann Jackson, Chairman, NRC to James M. Taylor, Executive Director for l Operations, " Follow-up Requests in Probabilistic Risk Assessment and DigitalInstrumentation and Control," November 30,1995.

3. Memorandum from James M. Taylor, Executive Director for Operations, NRC to Shirley Ann Jackson, i Chairman, "Irnprovements Associated with Managing the Utilization of Probabilistic Risk Assessment l and Digital Instrumentation and Control Technology," January 3,1996.

'4. WCAP-14572," Westinghouse Owners Group Application of Risk-Based Methods to Piping inservice  !

l Inspection Topical Report," March 1996. l

5. EPRI TR-106706, " Risk-informed inservice Inspection Evaluation Procedure," June 1996.

6. WCAP-14572, Revision 1, " Westinghouse Owners Group Application of Risk-Informed Methods to ,

l Piping inservice Inspection Topical Report," October 1997. l 4

I I

l l

l 13 l

l l

1 STEAM GENERATORS TAC' Nos. 88885 and 99432 Last Update: 9/30/98 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

4. Regulatory Analysis 5/97 (C)

5. Psoposed GL Pkg 10/97 (C)

6. ACRS Endorsement 9/97 (C)

7. CRGR Concurrence 1/99 (T)<23

8. EDO 1/99 (T)

9. Publish Proposed GL 3/98 (T)

Orig. Publish Proposed Rule 03/95 (C)

10. Public Comment (120 day comment period) 3/98-5/98

11. Revise GL Pkg 3rd qtr 99(T)

12. ACRS Comments 3rd qtr 99(T)

13. CRGR Concurrence 4th qtr 99 (T)

14. EDO Concurrence 4th qtr 99 (T)

15. Commission Approval 4th qtr 99 (T)

16. Publish FinalGL 4th qtr 99 (T)

Orig. Publish Final Rule 12/95 Brief Description The NRC originally planned to develop a rule pertaining to steam generator tube integrity. The proposed rule was to implement a more flexible regulatory framework for steam 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 a letter to Callan dated September 11,1998, NRR management suggested that the proposed GL be put on hold for 3 months while the staff works with NEl on their NEl 97-06 initiative. The staff is currently processing a Commission paper to gain Commission agreement with this approach, if sufficient progress is made with NEl in resolving technical and regulatory implementation issues, then the GL effort may be permanently halted.

Regulatorv Assessment' The current regulatory framework provides reasonable assurance that operating PWRs are safe. However, the current regulatory framework has numerous shortcomings. To resolve l

l M All revised dates reflect a 3 month " hold" period and assume that the Cornrnission agrees with NRR l= management with the revised regulatory approach, 14 f

4 i-l .'

f~ ' these shortcomings the staff is revising the regulatory framework to utilize a risk-informed and performance-based approach that will ensure compliance with current regulations (i.e., GDC, Appendix B,

ASME code, Part 100).

i Current Status:

- Briefed ACRS on ANPRM - August 1994

- SG rule ANPRM - September 1994

- SECY-95-131 - May 1995 - justifies continuation of rulemaking 4

- Briefed Commission on SG rule -- June 1995 i Briefed Commission on SG rule status - February 1996

Memo to Commission re, revised schedule - May .1996 l - Briefed Chainnan on status'- July 1996

- Information Brief for CRGR -- October 1996

- ACRS Brief on SG rule - November 5,6,1996

- Briefed Chainnan 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 4 - 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. GUDG-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 andorsement to issue GL and DG-1074 for public comment - September 15,1997 i - Briefed ACRS re DPO issues -- October 2,1997

- ACRS endorsement to issues DPO document for public comment - October 10,1997

~

l - 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, fer information briefing on package l - 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 currently processing Commission paper for 3 month hold on issuance of proposed GL NRR Technical Contacts- - Tod Sullivan, EMCB,415 3266 Tim Reed, EMCB,415-1462 RES Contact. N/A 15

. .. - ~ . - . . .. - - - _ . . . . - - _ - - ._- - ._.

9 NEW SOURCE TERM FOR OPERATING REACTORS TAC No. M89586 Last Update: 09/22/98 GSI No.155.1 Lead NRR Division: DRPM CTL 1.1 Scoorting Division: DSSA & DE MILESTONES DATE (T/C)

1. NEl Letter 07/94C

2. Commission Memo 09/94C

3. NEl Response 09/940

4. NEl/NRC Meeting 10/940

5. Publication of NUREG 1465 02/95C

6. NEl/NRC Meetings 10/940,06/95C,10/95C, 01/96C,02/96C,05/96C, 08/960,10/96C,04/970

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

11. Brief CRGR on Commission Paper 10/96C

12. Send Commission Paper to EDO/ Commission 11/960

13. Brief ACRS on Commission Paper 11/960

14. Response to NEl Framework Document 02/97C

15. Begin Pilot Plant Reviews 02/97C

_16. Begin Rebaselining 02/97C

17. Brief E Team on Status of Rebaselining 07/97C

18. Issue User Need for Rulemaking 08/97C

19. Finish Rebaselining 06/98C

20. Finish Rulemaking Plan 06/980

21. Finish Pilot Plant Reviews 01/99T Descriotkgi: More than a decade of research has led to an enhanced understanding of the timing, magnitude and chemical form of fission product releases following nuclear accidents. The results of this work has been summarized in NUREG 1465 and in a number of related research reports. Application of this new knowledge to operating reactors could result in cost savings without sacrificing real safety rnargin, in addition, safety enhancements may also be achieved.

16

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 ur>ed 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 Februaiy 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 tho TID source term; however, it includes a more realistic description of release timing and composition. Since the NUREG source term results in lower calculated DBA dose consequences, NRC decided not to requim current plants to revise thcJr DBA an9% us@he new source term. However, many licensees want to use the new source toe n to periorm DBA do',e 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 l efficient use of NRC's review resources, NRC encouraged the industry to approach the issue on a generic l basis. The Nuclear Energy Institute (NEI) unveiled its plans for the use of the new source term at l operating plants at the Regulatory Information Conference in May 1995. NEl, Polestar (EPRI's l consupant), and pilot plant (Grand Gulf, Beaver Valley, Browns Ferry, Perry, and Indian Point) l representatives met with NRC staff in June and October 1995 to discuss more detailed plans. l Proposed Actions: The staff has reviewed the framework document has prepared a Commission paper and decision letter that describes c generic implementation approach. The staff presented the l Commission paper and decision letter to the NRR Executive Team in September 1996, briefed CRGR in October 1996, and briefed the ACRS full committee in November 1996. The staff sent the Commission paper and decision letter to the Ccanassion in November 1996 (SECY-96-242). As described in the Commission paper, the current plan is to rebaseline two NUREG-1150 plants; one a PWR and one a BWR. The staff is reassessing tbs availability of key resources needed tu t .mplete rebaselinirq on an expedited schedule and issued a memorandum to the Commissioners informing them of the ruus of the project, briefed the Commissioners' Technical Assistants and accelerated the completion schedule with a shift in responsibilities between NRR and RES. The staff will also review each pilot plant application and prepare an exemption package addressing the use of each feature of the NUREG-1465 source term while pursuing rulemaking. The plan for issuing each remaining generic exemption is to brief the CRGR, issue for public comment, and then issue the exemption.

Orlainatina 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 Assessmoal: 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 pursuing rulemaking, and on April 2,1997, to provide a status report on the staff's actions regarding rebaselining and mismaking subsequent to the Commission's SRM. The pilot plant applications for Browns Ferry, Peny, Indian Point, and Oyster Creek have been circulated to the task force members to 17

help shape rebaselining. In June 1997, RES circulated an early dratt 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 TlD-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.

The staff briefed the NRR Executive Team on SECY-96-242 in September 1996, the CRGR in October 1996, and the ACRS full committee in November 1996. A limited number of pilot plants submittals and exemptions are expected - four submittals have been received so f ar (Browns Ferry, Perry, Oyster Creek, and Indian Point-2). An application is arso expected from Grand Gulf. In addition, the stati and Virginia Power met on November 26,1996, March 25 and .iune 18,1997, to discuss the rebaselining of Surry; the staff and Entergy met on August 29,1996, and March 27,1997, to discuss the rebaselining of Grand Gulf, in a February 12,1997, SRM, the Commission approved the Option 2 approach of SECY-96-242 and a modification to the letter response to NEl. On February 26,1997, the EDO issued the letter response to NEl. The staff has bitiated the rebaselining effort. The staff briefed the NRR Executive Team on the status of the project on July 8,1997, with an emphasis on the scope of activities involved with rebaselining; as a consequence of that briefing, the user need memorandum regarding' rulemaking was issued on August 1,1997, and the staff status report to the Commissioners was issued on September 9,1G97, indicating that the completion of rabaselining will be deferred.

In response to Commission inquiries regarding the deferral of the completion of rebaselining until November 1998, NRR and RES discussions and shius in lead technical responsibility resulted in an 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 accomp.ishc0 by reversing the lead responsibilities (RES is now the lead for rebaselining and NRR is now the had fc: rulemaking and regulatory guidance). The schedule for the completion of the pilot plant miews also improved by approximately 5 months as well.

NRR is working closely with RES to transfer te;hnical 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-phased plan and preliminary findings from Phase I, Phase ll, and the DBA portion (with the updated assumptions) of Phase Ill, respectively, for the rebaselining effort. On April 1 and 2,1998, RES and NRR staff briefed the ACRS and DONRR, respectively, on the progress of the rebaselining effort, initialinsights from the assessments completed, and the essentcl 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-98158 dated June 30.1999. SRM on SECY-98-158 issued 9/4/98.

NRR Technical Conts.stti: R. Emch, PERB,415-1068 S. LaVie, PERB,415-1081 NRR Lead PM: B. Zalcman, PGEB,415-3467

References:

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

July 27,1994, letter to A. Marion, r4El, from D. Crutchfield, NRC, " Application of New Source Term to Operating Reactors".

l I

18 i

,_.m , . __ _ _ . . - . _ . _ __ __ - _ _. .

September 6,1994, memorandum to the Comrnission from NRC staff, "Use of NUREG-1465 Source Term at Operating Reactors".

July 21,1995, memorandum to the Comrnission from NRC staff, "Use of NUREG-1465 Source Te.w at Operating Reactors".

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 S.%rce Term at Operating Reactori."

February 12,1997, Staff Requirements Memorandum to SECY-9u?42.

February P6,1997, 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 frota 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 ci public meetings:

e dated November iO,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 (Sutry) held on March 25,1997; e dated April 24,1997, for public meeting with BWR (Grand Gulf) hald 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 (Suny) held on June 18,1997.

I 19 r

, . ~- - . . - ~ . .- -- -- - . - . - - - - . - - - - . - - - _ . --

' er 1

ENVIRONMENTAL SRP REVISION ACTION PLAN l TAC No. MA0837 Last Update: 09/30/98 GSI: Not Available Lead NRR Division: DRPM

. MILESTONES DATE (T/C)

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

a. IdentNication of potentialimpacts 03/96C

b. IdentNication of integrated impacts 06/96C

c. Proposed options for resolution and develop initial draft of revised ESRP 10/96C j d. Staff / contractor meeting to resolve format and content of 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 CROTl review, N necessary 06/97C i

c. Publish (electronic) for public comment 09/97C l 1

3. - Disposition Public Comments 02/98C i 4. Publish Final NUREG-1555 02/99T l-

5. Maintenance of program data Ongoing I ' ada[ Descrhtion 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 j environmental protection issues (e.g., SAMDA and environmental justice) since originally pub:ished in i 1979, and to support the review of license renewal applications.

2 Regulatory Assessment NRR has established the ESRP Update Program for use in the INe cycle review j of environmental protection issues for nuclear power plants, especially license renewal applications, but

also operating reactors, and fvture reactor site approval applications. The ESRP will reflect current NRC -;

['

requirements and guidance, r.onsider other statutory and regulatory requirements (e.g., the National i Environmental Policy Act, Presidential Executive Orders), and incorporate the generic environmental i impact work and plant-specifb requirements developed during amending of Part 51 for license renewal reviews.

Current Status: The PNNUNRC staff workshop on the restructured and revised ESRP was held during November 13-14,1996. Now that the Part 51 rule for license renewal is final, particular emphasis is being

placed on assuring that license renewal needs are being addressed in a schedule consistent with the RES regulatory guide and pilot plant application. The results of the November workshop were provided by i PNNL in January 1997; followup discussions were held with the contractor through August 1997. The j'

June 1997 draft of the ESRP was forwarded to ACRS for its consideration. In light of the current ACRS schedule, ACRS staff indicated hat the ACRS will have no objection to publishing the draft ESRP; the ACRS may request a briefing during the public comment period. The June draft was provided to CRGR

. for information; the CRGR declined to consider it. Technical editor, legal (OGC), and technical (lead technical branches) comments were received on the July draft in early August and were included in the final draft. The FR notice of availability of Draft NUREG-1555 was published on October 3,1997; the electronic version (CD and diskette) is available in the PDR and wiH be made available to the public at no cost. Approximately 300 CDS and 500 hard copies of the Draft NUREG were distributed for comment.

ACRS discussed the NUREG at its May 1,1998, full corranittee meeting; the Committee indicated that the staff should retum to further discuss the SAMDA guidance. During the week of February 9,1998, the staff developed the comment binning and disposition plan; subsequently, a PNNUNRC staff workshop was 20

i. I I

l o I

held du bg February 24-25,1998, to disposition technical comments and rnake decisions regarding the organc tional structure of the ESRP. A primary concem raised by the public was the consolidation of git.'.*n.e for the technology area across disparate licensing frameworks (i.e., Parts 50,52, and 54); the r,bff r6 structured the document to segregate guidance into a Part 50/52 ("greenfield"-type review) and l that for Part 54 (renewal of a license for an existing facility). This segregation took the form of a l

supplement to the ESRP and was completed in draft form on July 3,1998.  !

)

NRR Technical

Contact:

B. Zaleman, PGEB,415-3467 l

1 l

21

l*

l PRA IMPLEMENTATION PLAN 1.2(d)

Graded Quality Assurance Action Plan TAC Nos. M91429, M91431, Last Update: 9/30/98 l M92447, M92448, M92449, Lead NRR Division: DRCH 1 M88650, M91431, M91432, Support Division: DSSA M91433, M91434, M91435, GSI: Not Available i M91436, and M91437 MILESTONES DATE (T/C) j 1. Issued SECY-95-059 03/95C

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

- 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: 8/25/05,10/10/95,10/25/95

4. Staff interactions with Palo Verde Ongoing Site visit on 5/23/95 on ranking and QA controls through NRC letter dated 7/24/95 on proposed QA controls 3/98C Site visit on 8/29-30/95 on risk ranking Site visit on 9/6-7/95 on procurement QA controls NRC letter conveying trip reports issued on 12/4/95 Meeting on 4/11/96 to discuss the staff evaluation guide Letter from licensee on 4/24/96 providing comrnents on staff evaluation guidance Site visit on 6/5-6/96 to observe expert panel and review revised procurement QA controls, trip report sent to licensee on 8/6/96 Letter from licensee on 9/12/96 transmitting responses to procurement issues raised in earlier staff trip reports Letter from licensee dated 11/13/96 responding to PRA issues raised in 12/4/95 trip report Overview of GQA initiative provided by PVNGS at 2/27/97 meeting with staff 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 OA controls 3/98C Meeting on 12/7-8/95 to discuss risk ranking and QA controls

- South Texas Submittal of QA Plan for implementation of graded QA, dated 3/28M6 is currently under staff review Meetings on 4/11/96 and 4/25/96 to discuss the staff evaluation guide and f uture interaction milestones and schedules Letter from llcensee on 4/17/96 providing comments on staff evaluation ouidance i

22 I

l '

l Meeting on 6/19/96 to discuss staff comments on the QA plan submittal for l graded QA, review questions transmitted to STP on 8/16/96 l -

Site visit on August 21-22 to observe working group and expert panel meetings, and to discuss staff review items, trip report in preparation  ;

l Management meeting on 10/15/96 to discuss PRA initiatives and staff >

l activities ,

Letter from licensee dated 10/30/96 respondeng to PRA questions

- Revised QA plan submitted on 1/21/97  :

Overview of STP initiative plovided 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 5/54 to evaluate: PRA quality, graded QA controls, QA contrcis for the PRA, corrective action and performance monitoring feedback processes, audit scheduling, and responses to the RAI concoms. Trip l report issued on 7/10/97.

STP submittal on 5/8/97 for preliminary RAl response STP submittalof draft QA Plan on 5/21/97 STP submittal of GQA related procedures, responses to RAl, and follow-on QA Plan on 5/22/97 n - 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 additional information regarding GQA implementing _

procedures and associated change control on 7/31/97 STP submittal on 8/4/97 responding to PRA RAI and provided procedures related to shutdown operations '

Negative consent SECY paper (97-229, dated October 6,1997) and Safety Evaluation has been issued that documents the staff's review of the QA -

program change.  ;

Commission did not object to issuance of STP SER as documented in 10/30/97 SRM Staff SER transmitted to licensee on 11/6/97 STP comments and interpretations submitted on SER on 1/26/98 Staff accepted STP interpretations of SER content on 2/19/98 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 '

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 QA 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 rankhg 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 evaluatbn guide Letter to GGNS dated 5/29/96 regarding implementation of QAP commitments i

i 1

23 L

c

l' ,

i Staff review comments on GGNS safety significance detennination 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 i 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 11/8/97 GQA closeout letter transmitted to licensee on 1/7/98 1

7. Revision 3 of Draft Evaluation Guide for Volunteer Plants issued for staff 07/95C comment  !

8. Revision 4 of Draft Evaluation Guide for Volunteer Plants issued for Steering 10/95C Group Review

9. Issue letter to 3 volunteer plants outlining program objectives and review ,

expectations. Distributed staff evaluation guide to licensees. 1/96C l

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

Industry comments on staff evaluation guide provided by letter dated 5/24/96 The staff reviewed the industry comments with respect to the need to revise, and finalize, the evaluation guide.

11. Regulatory Guide development milestones per PRA Action Plan Draft RG for Branch / division review and comment 7/31/96C i Draft RG for inter-office review and concurrence 8/1/96C l Draft RG for ACRS/CRGR review 11/22/96C  ;

- Draft RG for public comment 6/25/97C l Draft RG public comment period ends 9/23/97C i

- Public workshop held on draft RG 8/12/97C Publish final RG in SECY-98467 - 4/21980 SRM conditionally approves issuance of GQA RG 6/29/98C GOA final RG issued 8/98C

12. ACRS Briefings

- Expert Panel and deterministic considerations 2/27-28/960 Graded OA 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 Concems on GOA Regulatory Guide 3/6/97C ACRS memo to Commission expressing concems with GQA approach 3/17/97C Public Comments on GQA Regulatory Guide 10/21/97C Application RG/SRn 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 l

Graded QA Regulatory Guide 3/11/97C

- Graded QA Reaulatory Guide 2/27/980 l

l I

24

l

\

14. Issue draft Staff Inspection Guidance (Baseline + Reactive IP) for comment 9/29/98C CRGR meeting on GOA IP 12/8/98T

- Issue final inspection procedure 12/98T

15. Conduct NRC Staff Training 2/99T Descriotion: Prepare staff evaluation guidance and regulatory guidance for industry implementation for the grading of quality assurance (QA) practices comrnensurate 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 nationallaboratory, 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 QA 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 meeting on December 16,1993, the staff suggested that the industry could build on the experience gained from the maintenance rule to develop implementation methodologies for graded QA. The staff had numerous interactions with the Nuclear Energy Institute (NEI) during calendar year 1994 as the graded QA concepts were discussed and the initial industry guidelines were developed and commented on. In early 1995, three licensees (Grand Gulf, South Texas, and Palo Verde) volunteered to work with the staff. The staff has reviewed the licensee developmental graded QA efforts.

Prooosed Actions: The goal of the action plan is to utilize the lessons leamed from the 3 volunteer licensees to modify staff-developed draft guidance to formulate regulatory guidance on acceptable methods for implementing graded QA. The staff will develop a regulatory guide based in part on input from Brookhaven National Laboratory, and will also prepare a baseline and reactive inspection procedure (IP) for graded QA. An inter-office team has been established to prepare the regulatory guidance documents and test their implementation during the evaluation of volunteer plant activities.

Oriainatina Document: Letter f rom J. Sniezek, NRC to J. Colvin (NUMARC) dated January 6,1994, describing the establishment of NRC steering group for the graded QA initiative.

Reaulatorv 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 teamed 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.

The safety benefits to be gained from a graded QA program could be significant since both NRC reviews l 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.

25 i

l 0 l l .

Current Statur A draft evaluation guide for NRC staff use has been prepared for application to the volunteer plants implementing graded quality assurance programs. The statt 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 prepsred and circulated for review by both the ACRS and CRGR. SECY-97-077 (dated April 8,1997) transmitted the draft regulatory guides, including the GOA guide, to the Commission. Commission approval was obtained on June 5,1997, to Issue the documents for a 90 day public comment period. Senior management briefings were provided to the Director, NRR (on April 22,1997) and to the Deputy, EDO (on April 24,1997). The public comment period on the risk-informed guidance documents has expired. At this time,42 sets of comments have been received. A decision has been made, and accepted by the 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 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 GQA 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. A meeting has been scheduled with CRGR to discuss the IP on December 8,1998.

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 concems 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 Apphcation Guide and EPRI-5652) are sufficient for those topics. The staff received written comments from NEl on the evaluation guide by letter dated May 24,1996. The N51 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 concems with the staff focus on simply proposing to reduce quality controls for low safety significant items.

However, in recognhion of industry interest in the guido, 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 GQA RG (RG 1.176) be issued for use. The ACRS expressed a concem 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 l PRA parameters, and for the staff to prepare a plan for improvements to RG 1.176 after gaining experience with its application and to brief the committee within the next 2 years.

26 l

l

e South Texas submitted their QA program revision for their graded QA effort on March 28,1996. The change has been reviewed by the staff (HOMB, SPSB, RES, RIV, and NRC contractors). A meeting was held with STP on June 19 to discuss the staff's comments and 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 j 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 at which time licensee safety significance evaluations for 2 systems (Radiation Monitoring and Essential i

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 tf CIAl. Staff (from both HOMB and SPSB) performed a site evaluation during the week of May 5-8 to rev. nects associated with: PRA quality, OA controls for the PRA, corrective action and performance mon ang feedback processes, OA 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 concems 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 allinterested NRC staff on their overall strategy to implementing risk-informed approaches at their facnity.

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 review plan. By letter dated June 8, NEl indicated that their 96-02 guide will be revised.

Further NEl requested a meeting with the staff (in the August time frame) to discuss the changes and to discuss more objective means to assess the adequacy of QA program implementation. NEl has proposed that the amended 96-02 guidelines will be submitted to the staff for endorsement by a regulatory guide. A subsequent letter was received from NEl on July 16 that provided an updated version of NEl 96-02 based on comments they received from the volunteer plants and industry sources. The staff has reviewed the modified document. On October 10,1996, NEl submitted a letter expressing their concem with the graded QA initiative. NEl stated their concems regarded the questions raised by the staff in the area of OA controls for items determined to be low safety significant and in the area of safety significance determination. A meeting with NEl and staff from the volunteer plants (STP and PVNGS) was held on February 27,1997. NEl stated that 50.54(a) needs to be revised to offer licensees greater flexibility to manage their QA programs. The volunteer plant staff stated their firm desire to obtain copies of the draft GOA Regulatory Guide in a timely manner, following Commission approval, these were released for comment on June 25,1997. NEl additionally outlined a conceptual approach to integrate a performance monitoring methodology into the GOA efforts.

l 27

1

. l NRR Contacts: S. Black,415-1017 R. Gramm, 415-1010 RES Contact H. Woods,415 6622 References l

1) Letter from J. Snlenk (NRC) to J. Colvin (NEI) dated 1/6/94. I

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 l 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: l Graded Quality Assurance," dated March 24,1997,

10) SECY-97-229," Graded Quality Assurance /Probabilistic Risk Assessment Implementation Plan for the .

South Texas Project Electric Generating Station," dated October 6,1997, and SRM dated 10/30/97.

11) Letter from T. Alexion to W. Cottle (HL&P) dated 11/6/97.

12) Letter from J. Donohew to J. Hagan (Entergy) dated 1/7/98.

13) SECY-98-067," Final Application-Specific Regulatory Guides and Standard Review Plans for Risk-Informed Regulation of Power Plants," and SRM dated 6/29/98.

14) Regulatory Guide 1.176,"An Approach for Plant-Specific, Risk-Informed Decisionmaking: Graded Quality Assurance," August 1998.

l 28

l l

l l

EXTENDED POWER UPRATE ACTION PLAN  !

TAC No. M91571 Last Update: 09/30/98 Lead NRR Division: DRPW GSI: RI-182 Supporting Division: DSSA MILESTONES DATE (T/C)

1. Receive GE Topical ELTR1 (Generic Review Methodology). 3/95 C

2. Issue Staff Position Paper on ELTR1 l Meeting with GE/NSP. 4/95 C Identity differences between LTR1 and ELTR1. 8/95 C

- Issue RAls as appropriate. 9/95 C Incorporate information on foreign exoerlence obtained from SRXB. 10/95 C Develop power uprate database for all U.S. plants. 10/95 C

+ lssue Staff Position Paper. 2/96 C

3. Receive GE Topical ELTR2 (Generic Bounding Analyses).

GE plans to submit ELTR2 in two parts: the first part in March 96 3/96 C and the second part in July 1996. 7/96 C

4. Issue Staff SE on GE ELTR2.

Meeting with GE/ industry. 2/96 C

- Issue RAl. 3/97 C GE response to the RAl. 7/97 C Receive revised ATWS analysis from GE. 5/98 C

- Input to the draft SE from technical branches. 5/98 C

- Issue draft SE to the ACRS. 5/98 C

5. Receive Lead Plant Application (Monticello). 7/96 C

6. Issue Staff SE for Lead Plant.

Meeting with Monticello. 10/96 C

- RAI input from tech branches. 1/97 C Issue RAl. 4/97 C

- NSP response to the RAl. 9/97 C Receive revised application from Monticello. 12/97 C

- Issue additional RAls as appropriate. 2/98 C

- Input to the SE from tech branches. 5/98 C Issue draft SE to the ACRS. 5/98 C ACRS Presentation 6/98 & 7/98 C Issue SE. 9/98 C

7. Support the ongoing staff effort in developing a Standard Review TBD l Procedure for power uprates. Incorporate lessons leamed from Lead Plant activity.

29 f

i l

l

E Descriotion' This action plan describes the stratogy for completing both the generic and plant-specific reviews for extended power uprate submittals for boiling water reactors (BWRs). General Electric Company (GE) submitted a licensing topical report (ELTR1), which outlines the methodology for implementation of an extended power uprate program. ELTR1 encompasses power uprates of up to 120 percent of the originallicensed thermal power. Individual plant submittals for uprates willlikely contain requests for an optimum power level specific for that plant which is something less than the full 120 percent.

Each technical branch will review the applicable portions of both the ELTR2 (GE topical report containing generic analyses) and the lead plant application (Monticello), and will provide input into the staff's safety l

evaluation reports. The experience gained from these reviews will be incorporated into the ongoing staff effort in developing a standard review procedure for power uprates.

Historical Background The generic BWR power uprate program was created to provide a consistent

means for individual licensees to recover additional generating capacity beyond their current licensed l limit. In 1990, GE submitted licensing topical reports to initiate this program by proposing to increase the  !

rated thermal power levels of the BWR/4, BWR/5, and BWR/6 product lines by approximately 5 percent. i Since 1990, the staff has reviewed and approved at least 10 such power uprate requests under this i generic BWR power uprate program. As a follow-on to this program, GE submitted ELTR1 in March 1995 to propose " extended" power uprates of up to 120 percent of the originallicensed thermal power.

Prooosed Actions: Specific actions included in the generic actbn plan are: (1) review ELTR1 and issue a staff position paper, (2) review ELTR2 and issue a safety evaltation report, (3) review the lead plant ,

application and issue a safety evaluation report, and (4) develop a standard review procedure based on 1 l ELTR1, ELTR2, and the lead plant review.

Orignatina Document GE Licensing Topical Report (NEDC-32424), " Generic Guidelines for General l

Electric Boiling Water Reactor Extended Power Uprate," dated February 1995.

l j Reaulatory Assessment: Not applicable. (A safety assessment is not needed for this action plan because l l a justification for continued operation of a plant is not required.) This program is an inductry initiative that is strictly voluntary.

Current Status' The final SERs for both the ELTR2 and the lead plant application (Monticello) were l issued in September 1998. The fina! SER for the Hatch's 8% uprate is currently going through j management concurrences.

J l

With the issuance of the final SE for the lead plant, the staff has met its original objectives for this Task Action Plan except the action item of developing a Standard Review Procedure for power uprates, which is being tracked by separate tracking system (Maine Yankee Lessons Leamed issues).

NRR Lead PM T. J. Kim, DRPW,415-1392 i

i l

4 30

_ .~. _ _ . _ _ _ . _ _ _ ._ _ ___ , _ ___ _ _ _ _ _ _ . _ _ . _

l ACCIDENT MANAGEMENT IMPLEMENTATION  !

TAC #: M91966 - Overall Last Update: 10/1/98 ,

M91641 BWROG SAMG Review Lead NRR Division: DSSA I l

MILESTONES DATE (T/C)

1. BWROG Severe Accident Management Guidance (SAMG) documents l

Complete review of SAMG documents 7/98C Resolve remaining technical concems 6/99T L 2. Review severe accident training materials and BWROG 6/95C l prioritization methodologies I i

l 3. Develop guidance for NM audits j initialdraft (for intamal use) 11/95C

industry-sponsored NM demonstrations 3/980 j Revised draft (to NEl and public) 8/98C Final 11/98T

4. Complete A/M audits 3/99T

5. Hold public workshop 5/99T

6. Report to Commission on audit findings and 7/99T recommendations for achieving closure l

l Description This action plan is intended to guide staff efforts to assess the quality of utility implementation of accident management (A/M), and the manner in which insights from the IPE program have been incorporated into the licensees NM program. Specific review areas will include: development and implementation of plant-specific severe accident rnanagement guidelines (SAMG), integration of SAMG l with emergency operating procedures and emergency plans, and incorporation of severe accident

! information into training programs.

Hatorical Background The issue of NM 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 NM is a major remaining element of the Integration Plan for Closure of Severe Accident lasues [2]. The development of generic and plant-specific risk insights to support staff evaluations of utility NM programs is also identified in the implementation Plan for Probabilistic Risk i Assessment (3). NRC's goals and objectives regarding NM were established at the inception of this program (4]. Generic NM strategies were issued in 1990 for utility consideration in the IPE process (5).

l The staff has continued to work with industry to define the scope and content of utility NM programs and

! these efforts have culminated in industry-developed NM gukiance for utility implementation. Industry has

l. committed to implement an accident management program at each NPP [6]. NRC has accepted the

!' industry commitment and developed tentative plans for staff evaluation of utility implementation [7).

Proposed Actions Specific actions included in the A/M action plan are: (1) complete the review of l BWROG SAMG documents, (2) conduct A/M 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 Origmating Document SECY-88-147, Integration Plan for Closure of Severe Accident issues, May 25,

! 1988.

l l

31

--- ~. - - - _ - . - - - - ---- - - . - -. - - - _

i ,

Ruaulatorv 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 Statgg: 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 l Procedure and Severs Accident Guidelines (EP/ SAG) and associated technical basis documents to NRC i

for information on August 29,1996 [9]. The staff and Oak Ridge National Laboratory have completed a high level review of the EP/ SAG documents. Areas where additional information and discussion with the BWROG are considored 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 was received in January 1998 [12]. The staff has completed its review of the BWROG response. Remaining concems with the EP/ SAG were provided to the BWROG in a July 20,1998, letter [13].

During an NRC/BWROG management meeting on August 5,1998, the BWROG proposed and the NRC agreed that technical discussions on remaining issues be deferred until early 1999. This will permit BWR j licensees to complete A/M implementation, before redirecting their resources towards addressing the remaining technical concems. The milestone chart has been modified to show that the review of BWROG guidance documents is complete. A new milestone has been added to track the resolution of the remaining technical concems. The target date for resolving remaining concems is June 1999, based on the assuraption that technical discussions with the BWROG are initiated in January 1909.

Licensee target dates for completing A/M implementation have been submitted to NRC. Implementation has been completed at approximately one-fourth of the sites. Implementation at the balance of sites, including most of the BWR sites, will be completed by the end of 1998.

The staff outlined plans to evaluate licensee 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, l (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. NRR staff also attended an NEl-sponsored workshop on accident management implementation in March 1997. The workshop provided an opportunity to better understand plant-specific implementation approaches and issues, and the major elements of implementation.

The 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). No additional A/M demonstrations are planned. The A/M demonstration visits provided insights into the licensee's A/M implementation and evaluation process, and areas where changes to the guidance for evaluating A/M implementation may be needed. However, the demonstrations did not permit sufficient time and flexibility for the staff to evaluate the licensee's supporting analyses and resolve several staff concems identified during the visits. The staff needs to better understand the implementation process to determine the effectiveness of the voluntary industry initiative, the significance of the issues identified in the demonstration visits, and the need for inspections at remaining plants. Accordingly, the staff intends to proceed with further evaluations of A/M implementation.

In previous communications with industry and the Commission (e.g., SECY-98-131 and SECY-97-132) the staff had characterized the planned A/M evaluations as inspections against licensee commaments, but upon further consideration of the voluntary nature of this program, has concluded that these 32

l i

l evaluations should be performed as audits rather than inspections. The objectives of the audits will be to assess how licensees have evaluated and implemented enhancements to 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. Remaining actions in the revised approach are: (1) complete the A/M audit guidance, (2) conduct audits at 4-5 plants, (3) hold a public meeting / workshop to discuss l audit findings and staff / industry views on program completion, and (4) report to the Commission on audit l findings and recommendations for remaining actions to achieve closure.

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 concem to industry. In a June 25, 1998, response [19). NRC provided clarification regarding the staff positions and the approach to reaching closure. The staff indicated that they do not see major differences in NEl's and NRC's expectations, and that industry should continue to proceed with implementation.

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 has been recast as audit guidance, and updated to incorporate insights from tne 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].

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. Thadanl 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

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. Boedle, August 10,1998 NRR Technical

Contact:

R. Palla, SCSB,4151095 NRR Lead PM: Ramin Assa, DRPW,415-1391 33

l CORE PERFORMANCE ACTION PLAN TAC Nos. M91257 - DSSA Last Update: 09/25/98 i M91602 DRCH Lead NRR Division: DSSA l GSI: Ll-179 Supporting Division: DRCH MILESTONES DATE (T/P/C) I Task 1 - Inspection of Nuclear Fuel Vendors (DRCH) complete Siemens Power Corporation [PWR AIT followup) 06/94C l ABB/ Combustion Engineering [PWR reloads] 11/94C Teledyne Wah Chang (1WC) 12/94C Sandvik Specialty Metals (SSM) 12/94C Westinghouse CNFD 07/95C General Electric NEP 10/95C Framatome/Cogema Fuels (B&W Fuels) 09/96C GE (SLMCPR & low density pellets) 09/96C SPC (comprehensive re-Inspection of open items and new issues) 04/97C GE (RWE/RBM issues and followup)'[ licensee enforcement pending] 07/97C l

Task 2 - Inspection of Licensee Reload Analyses (DSSA) complete RI - 2 licensees: PSE8G (Hope Creek) 08/980 PP&L (Susquehanna) 10/97C Ril - 1 licensees:

SNOC (Hach) 12/97C Rlli - 1 licensee: Comed Detroit Edison (Fermi) 08/97C RIV -2 licensees: WPPSS (WNP2) 06/97C WNP2 re-inspection for enforcernent issues 02/98C Union Electric (Callaway) 11/97C Task 3 Core Performance Data Gathering / Evaluation (DSSA) complete Regions - Moming Reports & Event Notification ongoing Other Data Acquisition and Collation ongoing PNNL - Core Performance Evaluation Analysis (CY95) 12/97C Task 4 - Participation of Regions in Action Plan (DSSA) complete Identification of Vendorissues Feedback from Licensee Inspections Counterparts Meetings (Rl-RIV)

Task 5 - Evaluate Inspection Guidance (DSSA/DRCH) ongoing Evaluate Results of Licensee inspections incorporate Feedback from Region Inspectors Draft Guidance for Resident and Region Inspectors 09/98T Draft inspection Criteria and Action Plan Closeout 09/98T 34 L

0 Task 6 - Evaluate Licenses / Vendor 8.ead Test Programs for identification of 12/97P Core Performance Problems (DSSA) [ transferred to agency program on high bumup fuel]

Task 7 - Workshop on Core Performance issues (TAC No. M95674) complete identify issues Conduct workshop 07/96C Followup on Comments and Questions (RIC session) 10/96C 04/97C Desenption The action plan is intended to assess the impact of reload core design activities on plant safety through inspections of fuel vendors, evaluation of licensees' reload analyses, and independent evaluation of core performance information, with regional training and interaction.

Historical Background The action plan addresses the review of fuel fabrication, core design, and reload analysis issues that were discussed during 1994,1996, and 1997 briefings given to the Executive Director for Operations, and the 1997 Chairman briefing. The briefings presented by the Reactor Systems Branch (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 expand the action plan to monitor and improve core performance in operating reactors to include focus on licensee activities and the licensee /vendorinterfaces.

Procosed Actions Specific actions included in the action plan are: (1) evaluate fuel vendors' performance through performance-based inspections that evaluate the reload core design, safety analysis, licensing process, fuel assembly mechanical design, and fuel fabrication activities; (2) evaluate the 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 SALP evaluations as well as regional enforcement actions, as appropriate; (4) train and coordinate regional support staff participating in these activities; and (5) evaluate the results of these activities for use in formulating generic communications, revisions of regulatory guidance and guidance for regional inspectors, and other appropriate regulatory actions. In addition, as a result of recent generic concems, including the failure of control rods to fully insert, the action plan was  ;

expanded to review the adequacy of vendor lead testing programs for new fuel designs (Task 6); and a i workshop was conducted on core performance issues (Task 7) in the fall of 1996. The status of core performance inspection evaluations and emerging issues was covered at the last Regulatory Information Conference.

DSSA - The action plan identifies that licensee inspections in each region shall be performed, in coordination with the regional inspectors, to assess licensee performance in reload core analysis oversight and participation. Licensee inspections will normally be issue-driven. The data acquired through licensee / vendor inspections will be integrated with information supplied by the regions and other sources and will be evaluated for generic core performance indicators and industry conformance to current regulatory requirements. The end product of the initial assessment willinclude guidance for resident inspectors and regional staff. The ongoing activities to capture and address earfy waming of emerging issues will continue into FY98, and the action plan will be closed out by a memorandum response to the originating document, with summary attachments on vendor / licensee interface issues and potential generic problems, outlining continuing assessment of core performance under the DSSA operating plan.

DRCH - The action plan currently identifies 10 completed vendor inspections that shall be performed by multi-disciplined inspection tearns led by HOMB with contracted technical assistance, as required. These inspections are currently scheduled to be completed in 1998. The planned inspections of ABB/CE and Westinghouse Siemens are on hold, pending re-evaluation of the vendor inspection function within NRR.

35

i

. 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 f rorn briefing feedback.

Regulatorv 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 l containment) against serious radioactive releases. Likewise, the safety analyses must be properly l

l performed in order to verify, in conjunction with startup tests and normal plant parameter monitoring, that the core reload design is adequate and 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. l l

Current Status:

DSSA The data acquired from the completed vendor and licensee inspections have been evaluated for l generic impact and identification of emerging issues. Results from the NUPlc Nuclear Fuel Committee l

joint vendor surveillances and audits are being reviewed to supplement our vendor evaluations DSSA is .  ;

evaluating the action plan results to better integrate and prioritize its activities, consistent with the DSSA 1 operating plan. Options and recommendations for management review are being prepared for a closeout plan to capture the lessons leamed and to provide guidance for regional staff.

DRCH DRCH is re-evaluating options for the inspection program, and only the Siemens re-inspection for l DFl and non-conformance issues is still being considered.

l l

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

l l

4 k

1 4

i 36

_ ._ _ _ ._ . _ . _ .._._._ _ _ . _ _. _._.. _ _.~. .. _ . _ . _ _ _

HEAVY LOAD CONTROL (HLC) AND CRANE ISSUES TASK ACTION PLAN 1

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

TAC Nos. M93821: Action Plan Last Update: 9/24/98 i

M91955

DSC generic review Lead NRR Division: DSSA l M95546: Generic review of NRCB 96-02 ACTION DATE (T/C) l .1. Review, summarize and issue existing NRC guidance on heavy load control.

- Review NUREG-0554, NUREG-0612, GL 80-113, GL 81-07, GL 8511, and 2/96C' l other supporting documents.

l. - Develop summary of guidance. 2/96C

2. Determine significant heavy load issues that need to be addressed and develop resolution method.

! - Generic letter 85-11 and NUREG-0612. 2/960 4 i - Single Failure-Proof Crane (reliability). TBD* )

j- - Spent fuel cask drop accident prior to securing the lid. 2/96C l Risk significance of multiple failures within safe load path. TBD* I l- i j 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, 5/960 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. If an inspection procedure (or procedures) is planned, issue the inspection 12/98 procedure (s) in draft.

8. Obtain feedback (meeting, FRN, or other means) concoming the staff position TBD*

from industry representatives and resolve any discrepancies with the industry position.

9. Develop final version of guidance and cbtal:, r,anagement 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 the risk of crane failure during the movement of heavy loads.

37

s 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).

Historical Background 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 concems 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 concoms regarding what is required for the movement of heavy loads while the plant is operating. Because of anticipated future increases in industry efforts in this area, the staff needs to fully understand the existing problems and to undertake efforts to reduce such problems in the future. This plan was identified as a near-term issue under the dry cask storage action plan, and was recently revised to better reflect the scope and magnitude of the task.

, Proposed Actions Actions included in the plan are: (1) understand the current regulatory framework and i

j 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 l

actions may include the issuance of guidance to licensees alerting them to the potential problems and l requesting that corrective measures be taken to preclude accidents.

Originatina Document Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, i July 28,1995, " Dry Cask Storage Action Plan." '

Regulatory Assessment Severallicensees have either developed or are implementing plans to move I heavy loads in various areas of nuclear power plants (i.e., offloading spent f uel via dry storage and/or j

, transfer casks for transfer to ISFSis) over safe shutdown equipment, and over spent fuel during plant l l operation. Questions have been raised regarding the adequacy of NRC's guidance and the licensees'

l. 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 l the concems. As a result of the bulletin, several licenseas have undertaken efforts to assess their plans, l

capabilities, and licensing basis for heavy loads. The rction plan is intended to improve NRC guidance, address site specific issues, and aid licensees in their tuture plans to move heavy loads.

Current Status Review of the responses to Bulletin 96-02 v.miosed in April 1998. 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 due to budget shortfalls.

The staff will propose to RES a generic safety issue on evaluating the probability of crane failure during the movement of heavy loads. The heavy loads issue (USl A-36) was previously reported to Congress as closed via NUREG-0933, dated August 1987.

The staff visited Calvert Cliffs in 1997 and plans to visit 2 sites in 1998 for the purpose of obtaining an understanding how the various elements of the licensees' programs are being implemented. Information and perspectives gained through these visits, as well as input from the Regions, could be used to help determine and develop further guidance.

NRR Contacts Brian E. Thomas, DSSA,4151210 Kevin A. Connaughton, DRPW,415-3018 Joseph E. Carrasco, RGN-l/DRS, (610) 337-5306 38 I .

_ _._ ._ . - . . . . . . _. . . ._ . . - _ __ . ~ . . . .

6

'o References-Memorandum from Robert M. Bemero and William T. Russell to James M. Taylor, March 15,1995,

" Realignment of Reactor Decommissioning Program."

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

i 1

1 l

I l

l 1

h l-l' l

l l

39

o HIGH BURNUP FUEL ACTION PLAN Final Update l

TAC No. M91256 Status: COMPLETE-Lead NRR Division: DSSA GSl: 170 Supporting Office: RES l Desenption The action plan covered the assessment of fuel performance for high bumup fuel and evaluation of the adequacy of SRP licensing acceptance criteria.

Histoncal Background Experimental data on performance of high bumup (>50 GWD/MTU) under reactivity insertion conditions became available in rnid-1993. The unexpectedly low energy deposition (30 CAUGM) to initiation of fuel failure in the first test rod (with bumup of 62 GWD/MTU) led to a re-evaluation of the licensing basis assumptions in the SRP. As a result, the Office of Nuclear Reactor Regulation (NRR) was requested to prepare an action plan, in coordination with the Office of Nuclear Regulatory Research (RES).

Originating Documents Commission Memorandum from James M. Taylor (EDO), " Reactivity Transients ,

and High Bumup Fuel," dated September 13,1994, including IN 94-64, ' Reactivity insertion Transient l and Accident Limits for High Bumup Fuel,' dated August 31,1994. Commission Memorandum from James M. Taylor, " Reactivity Transients and Fuel Damage Criteria for High Bumup Fuelf dated November 9,1994, including an NRR safety assessrnent and the joint NRR/RES action plan.

l Regulatorv Assessment. It was determined that there was no immediate safety issue, because of the low

to medium bumup in most currently operating cor9s. Since the fuel failure threshold declines with i

increasing burnup, it was determined that the licensing basis design acceptance criteria might need to be redefined as a function of bumup. The end products of this action plan were 1) to determine the need for regulatory action, and 2) to establish and define the need for further action on extended bumup cycles and high bumup fuelissues.

Resolution After a preliminary safety assessment was performed, an action plan was developed. The

, first actions were to issue a user need letter to RES and the issuance of contracts to assess all aspects of l the high bumup fuel issue. Concurrently, meetings were scheduled with the non-domestic experimenters

( and regulatory authorities to discuss the experimental data and to assess potential consequences and j regulatory actions. In Information Notice (lN) 94-64), dated August 31,1994, the NRC staff described l experimental data suggesting that high bumup fuel could be more prone to failure during design-basis j transients and reactivity insertion accidents than previously thought. These data, on the relationship between fuel failure enthalpy and bumup for pressurized water reactor fuel rods tested in foreign experimental facilities, indicated that failure initiation enthalpy thresholds may be lower than considered in L

the evaluation of currently approved fuelbumup limits. On October 26,1994, the NRC staff presented additional technical information at the 22nd Water Reactor Safety information Meeting (WRSM).

Meetings with industry were held to discuss their planned actions and to solicit cooperation. In November 1994, the staff issued letters to the fuel vendors transmitting the WRSM papers and requesting industry action. The industry performed an initial generic safety assessment of the preliminary data and concluded that there was no significant impact on the public health and safety. Supplement 1 to IN 94-64 was issued on April 6,1995, with the purpose of providing addressees with additional information on high bumup fuel '

performance data acquired after the original notice and to discuss NRC and industry actions.

A letter from NEl summarizing the industry position was received in April 1996, and the EPRI report supporting this position was sent by NEl on September 20,1996. This report documented an evaluation

, - of the foreign research programs and concluded that the experimental failure data are not directly l applicable to the U.S. light-water reactors and that reactivity accidents are not a limiting constraint in high

, bumup fuel behavior. The industry's position was that revision of the existing fuel failure criteria for

reactivity accidents is not necessary, and additional expenditure of industry resources on this issue was 2

not warranted.

40

e e A Commission paper on the status of the high bumup issue and planned actions was prepared by NRR, i

reviewed by RES, and issued November 25,1996. This paper described the previously described events as well as the Core Performance Workshop held on October 24-25,1996, and the then<urrent and future research on high-bumup fuel.

By letter dated August 12,1997, the NRC staff responded to the NEl and industry report on high bumup fuel stating that the staff agrees with the industry "that reactivity accidents are of low safety significance and are not an immediate safety concem in fuel behavior for current cores restricted to a 62GWD/MTU lead rod average bumup limit. However, we do not agree with the industry position that revision of existing i fuel failure criteria used to evaluate reactivity accidents is not necessary and additional expenditure of I

industry resources on this issue is not warranted. The industry position appeared to be based on rejection of experimental results (non-prototypical) in favor of interpretations based on analytical results described in EPRI TR-106387". Furthermore, the staff stated that it had been following the foreign research programs and performing its own analysis and evaluation of those data. The staff acknowledged that the foreign data were not completely prototypical, but the staff believes that the foreign data provide a clear l Indication, reasonably justified by fundamental reasons, that f uel failure enthalpy limits decline with increasing bumup and make the existing regulatory criteria non-conservative. Thus the existing criteria may be inappropriate to evaluate the acceptability of high bumup fuel transient performance. A meeting with the industry was held on November 18-20,1997, to discuss interim criteria and exchange information.

The NRC staff (NRR, RES, and NMSS) met with the ACRS Reactor Fuels, Onsite Fuel Storage, and Decommissioning Subcommittee on April 23 and 24,1996, to discuss all aspects of high bumup fuel and l the Agency High Bumup Program Plan. The program plan outlines the following licensing and research l strategy. Testing and analyses will be continued to provide confirmation of acceptable fuel behavior for current fuel designs up to the present bumup limit of 62GWD/MTU (lead rod average). To obtain higher bumup limits, additional data and analyses of a similar nature will have to be provided by the industry prior to receiving NRC approvals. This licensing and ressarch strategy for bumup extensions involves a shift in responsibility to the industry for work that the NRC traditionally did to establish and justify regulatory criteria. In addition, fuel behavior would have to be addressed during normal operation, transients, and postulated accidents, and at a minimum the high bumup issues identified in the plan would have to be covered. A program for monitoring fuel performance should also be included. On June 4,1998, the ACRS was briefed on the plan. The Agency Program Plan for High Bumup Fuel was issued on July 6, 1998, in a letter from L. Callan to the Commission. Future work on the subject of high bumup fuelis l outlined in the Agency Program Plan. With the issuance of the Agency Program Plan for High Bumup l Fuel, the tasks of the High Bumup Action Plan are completed and the Action Plan is closed.

NRR Technical Contacts M. Chatterton, NRR/DSSA/SRXB,415-2889 Shih-Liang Wu, NRR/DSSA/SRXB,415-3284 Edward Kendrick, NRR/DSSA/SRXB,415-2891

RES

Contact:

Ralph Meyer, RES/ DST /RPSB,415-6789 i

i i

41 L

1

(

a a DIRECTOR's QUARTERLY STATUS REPORT October 1998 Generic Communication and Compliance Activities Added Since July 01,1998 TAC Type Contact Lead Tech Branch TR Comp LA Comp Title Reason Added MA2309 IN TKmhy Plant Systems Branch -/-/- N: Design Deficiency in Emergency The EAP authorued devekynent d N as its Diesel Generatws that Prevent its 7/1498 meeting.

Img Term Operauon MA1314 W 1Xoshy Mechamcal -/-/- N:Ibential Inadequacies in t!r The EAP authaued devekyment d IN at its Engineering Branch Installation d Oneck Valves Made 7/1488 meeting.

by Anderson Greenwood and Bag Warner MA3273 N TACarene Reacta Systems -/-/- 11/1088 T N: Potentiallhelems wnh a Re PECB Management approved devekpment d N Branch Commercially Available Computer 8/l1N8.

Code MA3377 LT %TBurton Civil Engineering and i1/3068 LT: Ten-Year Inservice Ingeason he PECB management agproved LT followw d Gewciences Brandi Program Update fm Plants hat this issue on 8/20N8.

Intend to inclement Risk-Inf<rmed Insavice Inspectam MA3433 IN TAGreene Civil Engineering and -/-/- 12/3088 T N: ASME Code,Section XI Required The FAP autixvimed devehpnent d N as its Geonciences Brandi Inservice Inspection Exanunzions 8/25B8 meeting.

MA3581 IN TAGreene Civd Enginecrmg and i1/3038 2/25B9 T IN: Concrete Containment he PECR management auttarized devekpnent d IN Geosciences Brandi Degradation and Carusion d 9/156 8.

Prestressed Tendons and Steel Containnent Liner MA%49 N JRTappert Maierials and 03/12 S 9 12/3168 T IN: Vibration Induced Failure of he EAP authorized devekyment d IN at its Chemical Engineenug Socket Welds due toIxtdown Orifice 9/22M8 meeting.

Branch Degradation MA3653 N ENFiekts Reacta Systems -/-/- N: Design Control Deficiencies The EAP autherued devekpirat of IN as its Branch Resulted in Reduced ECCS Pump NTSII 9/22S8 mecung.

NOTES: Total Number of Records = 8

" /--/ " for a "TR Comp" date means that at least one reviewer is unscheduled.

"11/11/11" for a "TR Comp" date means that at least one reviewer is constant load scheduled.

Page1of 1 08-Oct-98

s DIRECTOR's QUARTERLY STATUS REPORT October 1998 Generic Communication and Compliance Activities Closed Since July 01,1998 TAC Type Contact Lead Tech Branch TR Comp LA Comp Title Reason Closed M94565 LT DLSkeen Reada Syssems 08A)3S8 P 8/3S8 C Skm Saam Solenoid Pdot Valves The SRXB has deternused the bcensee respuse Branch Caused by Viton Dugdragms to the B%1 TOG recomnwndaions is adequate, bcensees have either replaced the suspect SSPV diaphragms a have an enhanad nxunwing program.

M95444 LT TAGreene Materials and 07/24S8 P 7/2488 C lead Tedinical Review -Inducaxa 1his TAC is closed per Tom Greemen request Owsrucal Engineering IIem Stress L,.m.am fw (7/2468). He infamation from industry to Branch Stainless Steel Piping write an informahan will acs be availaNe in the near future.

M96354 LT TAGreene Mechamcal 10/0168 P 10/188 C Contmement Rec'rculauon Spray and Ccurpleted un&r TIA (TAC MA0273)

Engineering Branch Qtwach Spray Piping Outside Design Basis M96912 LT WTBurton Plana Systems Branch 09ANS8 P 9BS8C LT: Potential Generic Concern with The EAP authorized devekipmem d IN (MA1617) on regard to Fire Protection Aauahon 4/2868. This TACis closed.

System M97146 GL JWShapaker Containment Systems 07/1488 P 7/14/98 C GL: Degradatmo d ECC GL 98-04 issued 7/14S8.

and Severe Accident Recirculsion Following a LOCA ese Branch to Fcreign Matenal in the Containment M97331 BL JWShapaker ReactaS 3stems 08A)668 P 8M/98 C BL: Inadequate Procedural Guidance The prToned BL is cancelled per SRXB (lackson)

Branch dtring S/D and Site Specific e-mail to JWS on 8/388. This TACis ckned.

Vulnerabihties due to Gas Accumulation M97920 GL JWShapaker Civil Engineering and OE/1968 P 8/1988 C GL: Seismic Capaluhty of The TAC is canoeRed per GBagchii request Geosciences Brandi 7bermaLLag Panels 8/1988.

M97981 GL WFBurton Ovil Engmeering and 07/2098 P 7/2098 C N: Scalement Monitoring and N 98-26 issued 7/24/98.

Gemciences Branda Inspection d Plant Strucures Alfected by Degradation d Perous Canacte Subfoundaions M98751 GL JWShapaker Emergency 07/10/98 P 7/1008 C GL: Oarification d NURECER 5055, The grrposed GL was canceled by PERB on 6/26/98.

Preparedness and "Atmosidwric Diffusum fu Control Rmbrion Prosectma Room flabitability Assessment" Branch M99304 LT TAGreene Ovil Engineering and 08/1388 P 8/13/98 C ll%IEC Part 21 Cornputer Cok issue lhe TAC is ckmed per GBagchils request 7/28S8.

Geaciences Branch 1his issue will be comtuned wah a ptpwed IN under TAC MA3273.

Page1 4 3 08-Oct-98

. a Generic Communication and Compliarme Activities Closed Since July 01,199f TAC Type Contact Lead Tech Branch TR Conap LA Comep Title Reason Closed MA0202 N CDPetrone Plant Systems Brand 05/1988 P 5/1968 C N: latergretmion ci dEPA & SPLB to address issue in Rev to RG 1.54 & .140; Giacoal Filter Testue ficquency N not needed.

Folkwing Expanse to Manic ega MA0311 N RABenedict Operata Ucensing 07/3188 P 7/3TS8 C N:Ik-A tdSystems2 N 98-28 issued 7/31S8..

Eranch Sample Plan for Operator Facensing h ;. .,w.

MA0376 N TKoshy Events Assessmentand 07/13SS P 7/1368 C N: Issues identified During Recent N 98-22 issued 6/17S8.

GCs, and Special ' NRC Design Inspectons Inspedian Branch MA0300 N CVIlodge Plant Systems Branch 06/26S8 P 6/26/98 C N: Stem Binding in Turtune N 98-24 issued 6/2668.

Governor Valves in RCIC and Auuhary Feedwater Systems MAM75 N CDPetrne a Civil Enginemng and 07/22S8 P 7/22S8 C N: BWR Ovethead CYane Design Basis ECGB decided that the prposed N is not needed.

Geosciences Brandi Calatations DoNot Ad%s 7/2268.

Occurrena of Seism: Event When The Crane is in Un MA1947 N CDPetrone Containment Systems 07/2288 P 7/2288 C N:75 Surveills ace T se Cancelled due to hmited resowees and pimity and Severe Accident Requirements f.r Containment confhcts.

Branch isolation Valve s MA1616 N JRTarpert Materials and 06/23S8 C 6/2388 C N: Crosby T ebet Valw Setpoint N 98-23 issised 6/23S8.

Chenucal Engineering Drift Probb.ns Branch MA1623 LT JRTarpert Electncal 07/1588 P 7/15/98 C LT: Rmew of AEOD Study on 'the TAC was closed by PECB task manager i Engineering Branch Wesreghouse RPS Unavailability Tappent on 7/15/98.

MA1690 LN TMKhan Plant Systems Branch 07K)888 P 7/888 C .: Less of Inventwy from N 98-25 issued 7/SS8.

Safety-Related, Closed-lap Cooling Water Systems MA1691 N TAGreene Reacta Systems 07/24S8 P 7/2J % C N: Channel Bowing in GEli Fuel The GE BWR Owners Gitap have been involved ie Branch Assembly this issue. All plants that are susceptibte to this pctential poblem have been irl.urd.

1herefwe, thee is no need for any additional generic comminication.

MA2034 N ElBenner Muerials and 07/24S8 P 7/24S8 C E: Steam Generator Tube End N 98-27 issued 7/24/98.

Cheniscal Engineering M ag Branch MA2098 N CDPetrone Medianscal 07/22S8 P 7/22S8 C Sy. to N 96-48, Mota-Operated LN 9M8, Sup 1, issued 7/24/98.

Engineering Branch Valve Performana Issues Page 2 of 3 08-oct-98

e o- .;.

Generic Communication and Compliance Activities Closed Since July 01,1998 TAC Type Contact Lead Tech Branch TR Conny LA Conny Title Reason Closed M 3099. N TAGreene Inser.uneatation and 07/24/98 P 7/24/98 C N:Inndvertest Aar.atson of "this issue only affects I.iawick, llope Creek, Controis Brandi Staaeyliquid Cemetel System and Perry.

MA2' 4 N CDPetrone Plant Systems Brandi OV17/98 P 8/17/98 C P!: Fkmding of ECCS Rooms at WhP-2 N 98-31 issued 8/l8/98.

CM by Fire Waer System Valve Rugnure MACM N TAGreene Decencal 08/28S8 P 1'28/98 C N: Configurauon Control Errors N 9E-34 issued 8/28/98.

Engineering Branch MA3272 LT , JRTagpert Maserials and 09/30/98 P 9/30/98 C Gemene letdown Orifice Degradatxe ThisTACwasreplacedIr-MA3649:N Vibratica Chesmcal Engineering Reviews Induced Faikse d Soc'tes Welds ese toletdown Branch Orifice DegradatuxL MA3378 N ENTields Decincal 09/798 P 9/211/98 C N: "- .

or Poorly N 98-36 issued 9/18/98.

Engineenag Branch Controlled. Non-Safety-Related Maintenana Activities Unneassarily ChaBenged Safety Systems ,

MA3379 N CDPetrone Operata Liccasing 09/30/98 P 9/30f98 C IN: licensing Exam Ehgibihty N 98-37 issued 10/1/98. [

Branch

. Expenema Requirements fa SRO Appbcants l

L I

i NOTES: Total Nunnber of Records = 28 i

" /--/ " for a "TR Comep" date mecans that at least one reviewer is mascheduled.

"11/11/11" for a "TR Counp" date meesas that at least one reviewer is constaat load scheduled.

Page 3 of 3 08-Oct-98

i

?

WOLF CREEK DRAINDOWN EVENT: ACTION PLAN TAC No.: M66278 Last Update: 9/25/98 Lead NRR Division: DSSA MILESTONES DATE (T/C)

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

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

3) Complete Draft Tl/ lasue to the Regions for Comments 10/98(T)

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

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

Comments)

5) ACRS Briefing 11/97(C) l l

6) GL lasued 05/98(C)

7) Receive Regional Comments on Tl 12/98(T)

8) Complete Evaluation of the Responses to the GL 02/99(T)

9) lasue Tl 03/99(T) l l- 10) Complete inspections (As necessary) 07/99(T)

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 addresseas comply l with NRC regulatory requirements and conform with current licensing bases for their facilities.

l Historical Background 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 sye*.em, train B, to the refueling water l storage tank (RWST) for reborating train B. The failure of the reactor operating staff to adequately control l two incompatible activities resulted in transferring 9200 gallons of hot RCS water to the RWST in 66

-seconds.

The Wolf Creek event represents a LOCA with the potential to consequentially fail all the ECCS pumps and bypass the containment. Another important feature of this event is the short time available for corrective action. Based upon calculations by the licensee and the staff, it is estimated that if the draindown had not been isolated v%in 3-5 minutes, not positive suction head would have been lost for all ECCS pumps, and cors uncovery would follow in about 25-30 minutes. This event represents a PWR vulnerability which was not previously recognized.

Pmoosed Actions. Specific actions of this generic action plan are: (1) issue IN 95-03 (issued January 16.

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.

Originating Document: AEOD/S95-01, " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994'.

42 L

l t

a Romitatory Aeeeeement: 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 have been issued. CRGR concurred the proposed GL in 9/96; but as directed by an SRM, the GL was published in the Federal Register in 2/97 for public comments. ACRS was briefed on 11/6/97. 2nd CRGR concurrence was obtained in 1/98. The GL was issued on 5/28/98. Staff is preparing draft Tl for issuance to the Regions for comments.

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 Rea:: tor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions while in a Shutdown Condition," issued May 28,1998.

l l

l L

l I

i-I i

i 43 l l

[

gm..,.... . . . . . . - . . . ,

e ATTACHMENT 2 GENERIC COMMUNICATION AND COMPLIANCE ACTIVITIES

. 3 DIRECTOR's QUARTERLY STATUS REPORT October 1998 Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comep LA Conip Title Description Division of Engineenng Civil Engineering and Geosciences tiranch MA3377 LT %TBurton ll/XV98 LT: Ten-Year Inservim Inspection Considershon d extentions on the subnuttal d 10Lyear ISI Program Update for Plants That programs fw piping to allow time to deveky risk-i-fanxd latend to implement Risk-Informed ISI programs fa pipng.

Insenice Inspection MA3433 IN TAGreene - /-t- 12/XV98 T N: ASMECode,SectionXI Required Remind liansees that they need to subnut request fa rehef Inservim Inspection Examinations to the NRC when inservia inspxtnan examinations required by Section XI d the ASME Boiler and Pressure vessel Code are impractical MA3581 N TAGreene i1/yW98 2/25S9 T N: Concrete Containment Alert liansees regarding the prrper monitaing d Degralation aal Carrasion d prestressed concrete containments.

Prestressed Tendons and Steel Contaicment liner ECGB has 3 GCCA(s)

Electrical Engineering Branch MA0734 N DLSkeen -/-/- 12/XV98 T IN: low & Medium Vokage Circuit Prmides examples of breaker failures due to inadequate Breaker Fadures maintenance a refutbishment. i EELR has 1 GCCA(s) -

Materials and Chemnical Engineering Branch M95279 GL JWShapaker  !!/XV98 12/3168 T GL: Modification d the Extending to gerating reador liansees, on soluntary Requirements fm Post-Amident basis, relaxations in PASS program requirements.

Sampling System M99226 GL JWShapaker -/-/- 12/3168 T GL: Augmented Inspxtnan Issue fa Prgoses augmented inspection d smaH diameta, Class 1 Small Diameter, Class 1 Pipeng in pping in PWR high-pessure-injection systems toovercome PWR liigh-Pressure-Injedian System ASME code oversigtt M99432 GL JWShapaker 12/3169 12/3168 T CL: Steam Generata Tube Integnty Informs licensees that actions beyond curtent TS requirements may be necessary to ensure steam generascr tube integnty.

Page 1 of 5 08-Oct-98

a >

Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comep LA Comep Title Description M99n58 GL JWShapaker

12/3168 T GI; RCS Onennstry Effects on Flaw During the review d a flaw evahaahan fcr a anck in the Growth Estimates core shroud it was observed that the facdnty had kss restrrctiw TS s,.m G fa primary water chermstry control:, then were rehed upon by the hcensee in estabbshing a anck growth rate.

MA1689 GL JWShag=ker 08/28/00 11/3068T GL:NRCStaff ReviewdBWRVIP-05 MA3649 N JRTagpert 03/12M9 12/3168T N: Vibration t=And Fadure of Alerts hcensees torecent operational czperienz Socket Wekts due toletdown Orifice concerning failure of socket welds on leadinn knes.

Degradauce Specifically, the Infwmahor. Ndice will address crusion d orifices and subsequent vibration induced fatigue failures d wekts in letdown lines.

EMCB has 6 GCCA(s)

Mechanical Engineering Branch M96354 LT TAGreene 100188 P 10/188 C Containment Recirculation Spray and Minstone 3 desermined that the containment recuculataan Quench Spray Piping Outsmie Design sgray and quench spray piping and suppats cruald be Basis suljected to higher accident temperatures than those previcansly assumed in the design basis.

MA2097 N CDPetrone 06/13n9 6ft3S9 T N: leaks in the Emergency Diesel Alerts hansees so leaks in the skid-rmamted tube oil Generator Imbe Od Piping pigung of emergency diesel generatas.

MA1514 IN TKosby -/-./- N: IbcatialIna&quacaes in the Discusses maintenance errors that could make the check Incan euw of check Valves Made valves inoperable.

by Anderson Greenwad and Bag Warner EMEB has 3 GCCA(s)

DE has a total of 13 GCCA(s)

Division of Reactor Controis and Hannan Facton Qnality Assarance and Maintenance Branch M98441 GL JWShapaker i1/1388 11/30M8T GL:Quahty Assurana of Deoronic la view d te<+=aL=w =1 advancements, changes in NRC Rectrds regulahans, a request was made to update the gindana provided in GL 88-18.

MA1618 LT %TBurton IOr3088 LT: Guidernes for Safety Considers providrag guidance on the classification and Nha= d Struaures, evahantion of struaures, systems and astrols.

Systems, and Components in Nuclear Power Plants HQMB has 2 GCCA(s)

Page 2 of 5 08-Oct-98

- _ _ _ _ - _ - _ _ . . _ = .

n 2 Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Cosep LA Conny Title Description DRCH has a total of 2 GCCA(s)

Division of Reactor Prograan Managesment Events Assessament and GCs, and Special Inspection Branch MA2124 GL JWShapaker -/-I- 12/30f98 T AL: ADAMS DectronseSubnuttal Infcrms licensees to the OCIO gumiana a the eledronic l Capability subcuttal d documents per 50.4 (c).

PECB has 1 GCCA(s)

Generic Issues and Environniental Projects Branch M99811 GL RABenedid --/-/-- GL: Interim Guidlines fa Updating Reiterates need fa licensees to ipdate theirs FSARs, in FSAR keeping with 10 CIR 50.71(c).

PGER has 1 GCCA(s)

Safeguards Branch MAOSl4 N RABenedics -/-I- N: Annual Summary of Fitness for Provides a general summary and analysis d the data Duty Program Performance Repcuts subnutted by liansees in their fitness fa duty program for Calendar Year 1996 performance reports for cylender year 1996.

PSGB has 1 GCCA(s)

DRPM has a total of 3 GCCA(s)

Division of Systeins Safety and Analysis Containement Systemis and Severe Aaident Branch M98125 N TJCarter 03AMA12 4/1/99 T N: BWR Containment Bypass Flow A plant mafigtration during rowinc <peration could During Purging potentially result in containment b3pass folkwring an acident M99813 GL JWShapaker 11/06/98 12/31/98 T GL:(.Way Requirements for Va!ves required to be open for heat removal purposes may Dual Function Valves also have a containment isolatace function.

MA0349 GL JWShapaker 01/31/99 12/31/98 T GL: Request for information Based on staff actions that were undertaken to address the Related to ECCS Rma- matter of containment detris and potential fw ckyging of Capability suction strainers in B%1 facilities, there is a need to similarly pursue this matter on PWRs.

MA0626 LT JWShapaker 06/12/00 12/15/98 T LT: PWR Sump Blockage Issue To develop and implement a plan to esh the potential for clogging d ECCS sump sacens by LOCA debris in PWRs.

Page 3 of 5 08-Oct-M

Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Coenp Title Description SCSB has 4 GCCA(s)

Plant Systems Branch M96913 GL JWShapaka 0480N9 1200N8 T GL: Post-Fire Safe Shutdown Circiut To alat heensees to re:ent noncompliances and associated Analyses civil penakies regarding has*een lack of demonstrable p'ttection froan a control rtxun hot slurt condition.

M97978 GL JWShapaker 12/1IN8 2126/99 T GL- Latrratory Testing d Informs addressees about NRC staff views on diarmal testing Nucl.:ar-Grade Activated Charmal praaices and diers model technical specifications for voluntary adtytion by the addressees in parparation for future testag obhgations. L i

M99066 IN DBenner -/-/- 600N8 L IN: Misundastanding d the Deveig N to infwm bcensees d several instances of Ultimate IIcat Sink Licensing Basis errtrs in licenseen understanding d l'kimate fleat Sink licensing basis.

M992n3 GL JWShapaker 12 BIN 8 128068 T GL 96-06, Sup 2- Assurance d Notifies liansees ahmt safety-significant issues that 14tupment Operabihty and cwid affect antainment integrity and equipment operabihty Containment Integnty dirir.; dunsg design-basis axident andations.

Design-Basis Accident Conditions M99331 N DDenner 12 SIN 8 60068 L IN: Pcetulated Imss cfIkedwater as Alats addressees to concans related to fkxxiing as a a Resuk d a Pipe Break in the resi:lt of a non-design basis pipe tweak in the cirulating Circulating Wasa System water system MA1361 IN WFBurton --/-/- 1000N3 T IN: Inadequate Analysis of Reqd Alerts licensees to potentiad prttlems assa with pmt-fre and Assa Dectncal CKTs Rsit. in safe S/D CKT analysis. Ihese problems could resuk in the Potentialloss of Post-Fire S/D fire-induced CKT failures w*iida cxmikt prevent the operatim Capabihty cr lea ho maloperse e of equipment necessary to adzieve post-r --sfe S/D.

MA1617 IN WFBurton 12OINR 10G188 T N: Failure of Automatic Sgrinkler Discusses recent fathnes d gre-action sprinkler valves at Valves Farley.

MA1940 N 'IXoshy 10G1/98 N: Changes on OSilA Requirernents in Alerts liansees to recent rate change from OSilA regarding Fire Brigade respiratory grotection MA1941 N TACreene -/-/- N: IF.300 Spent Fuel Shyping Cs ' Alerts licensees to a ptential prthlem concermag the IF-300 spent furt sgrppmg cask. ,

MA23tt9 IN TKoshy -/-/- IN: Design Defaciency in Dnerency Discusses the design deficiency that relies on ccuitrol as Diesel Generators that Preven' ats to keep the engine running even during emergency mode.  ;

long Term Operation <

SPLB has 10 6CCA(s) t 4

Page 4 of 5 08-Oct-98  ;

I

Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Cosep LA Cosmp Title Description Reactor Systenes Branch M99332 GL JWShapaker -./-t- 12/31/98 T GL:Guidana on tN Reguletsy Pnmdes a compilarkm d the current NRC staff guidance on Requzements fw Chticality regulatory requirements fw cntacanty analysis d new and Analysis dITel Staage at LWR spent fuel storage at LWR plants.

Power Plants M99e59 N EFGoodwin 12/31/98 9/30/98 L N 97-15. Siv I, Reporting d Informs licensees d .@a :s d 10 CFR 50.46 concerning Errors and rh .res in LB LOCA Evin repmting d aR errors asmusted with code d recad. not Models of Fuel Vendors and just errws assonated with large-break LOCA analysis.

Comphance with 50.46(a)(3)

MA0519 LT EKsoodwin -/-I- LT: Licensed Thermal Power Pnmdes licensees and staff with explicit guidance on i regulmay lea.r A d licensed core thermal power and I guidance on enforarnent.

MA0779 N WFBurton *

-/-I- 11/21/98 T N: Gas Binding of St Pumps Informs licensees about gas accun=ds-in high head safety t

injxtion piping at Beaver VaBey.

MA1538 N RABenedict -/-I- N: Ice Condenser Operability Desaibes experiences at DC Cook plant regarding maintenance Assurance Problems and surveitana dice condenser MA3273 N TAGreene .-/-/- 11/10/98 T N: Potentsal Problems with a infam liansees to potential prnblems with a we_ My Commerciagy AvailableComputer available computer code.

Code MA3653 N ENFiekis -/-/- N: Design Control Defmencies Expenenes at three plants regarding the impact on STSIl d Resuhed in Reduad EOCS Pump NPSit - " - - -- level cahbration pucedures.

SRXB has 7 GCCA(s) i DSSA has a total of 21 GCCA(s) t r

l NOTES: There are a total of 39 GCCA(s) l

" /--/ " for a "TR Comip" date uncans that at least one reviewer is unscheduled.

"11/11/11" for a "TR Cosop" date nicans that at least one reviewer is constant load scheduled.

Page 5 of 5 08-Oct-98 i

- . . . . = - _ _ - . _ - .

DISTRIBUTION for NRR Dir:ctor's Quart:rly Status Rtport C:ntr:I File PDR- PGEB R/F TOMartin, EDO

! a SJCollins, NRR FJMiraglia, NRR BABoger, NRR BWSheron, NRR JWRoe, NRR DBMatthews, NRR THEssig, NRR FMAkstulewicz, NRR EMMcKenna, NRR PCWen, NRR EYWang, NRR BJSweeney, NRR GMHolahan, NRR FPGillespie, NRR RLSpessard, NRR GClainas, NRR RCEmrit, RES Regional Administrators Mr. Ralph Beedle, Senior Vice President Nancy G. Chapman, SERCH Manager

& Chief Nuclear Officer Bechtel Power Corporation Nuclear Energy Institute 9801 Washingtonian Blvd.

1776 i Street NW Gaithersburg, Maryland 20878-5356 Suite 400 Washington, D.C. 20006-3708 Mr. R. P. LaRhette

! Institute of Nuclear Power Operations 700 Galleria Parkway Atlanta, Georgia 30339-5979 Mr. Lee Watkins Assistant Manager For High Level Waste U.S. DOE l P.O. Box A Aiken, South Carolina 29892 Mr. R. W. Barber Safety and Quality Assurance, DOE 270 Corporate Center (E 853) I l 20300 Century Blvd.

Germantown, MD 20874

Mr. S. Scott Office of Nuclear Safety, DOE Century 21 Building (E-H72) 19901 Germantown Road Germantown, MD 20874-1290 Mr. Bob Borsum 1700 Rockville Pike, Suite 525 Rockville, MD 20852 Ms. Norena G. Robinson, Licensing Technician Nebraska Public Power District Cooper Nuclear Station P.O. Box 98 Brownsville, NE 68321

r. .-~. . cm* 'NMZl.'O*m v . cm m .~-.e.w. v c- -m .~ wow.. w . e n omce poss oseu . E eoceaneu p eoes oneu lF eoceppu 4 eagggs poes.oneu f .ggReu M_

=ue n-,uns tw., fin ew- w J e=(4 A yW/, re . 6-I onTe - some swJAe J

s& sb  % mli'l% eps &m ,

/ _ =+ OenCIAL RECORD COPY '/

6h <