ML20211G705

Directors Status Rept on Generic Activities - Actions Plans & Generic Communication & Compliance Activities, Oct 1997

ML20211G705
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Issue date:	10/31/1997
From:	NRC (Affiliation Not Assigned)
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NUDOCS 9710030210
Download: ML20211G705 (88)

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t DIRECTOR'S STATUS REPORT on GENERIC ACTIVITIES l

Action Plans Generic Communication and Compliance Activities October 1997 nr>w

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

This report includes two attachments: 1) action plans and 2) generic communications under development and other generic compliance activities. Generic communications and compliance activities (GCCAs) are potential generic issues that are safety significant, require technical resolution, and possibly require generic communication or action., *NRR Action Plans," includes generic or potentially generic issues of sufficient complexity or scope that require substantial NRC staff resources. The issues covered by action plans include concerns identified through review of operating experience (e.g. Boiling Water Reactor internals Cracking and Thermolag), and issues related to regulatory flexibility and improvements (e.g. New Source Term and Probabilistic Risk Assessment (PRA) Implementation Plan). For each action plan, the report includes a description of the issue, key milestones, discussion of its regulatory significance, current status, and names of cognizant staff., " Generic Communications and Compliance Activities," consists of three monthly status reports.1) open GCCAs, 2) GCCAs added since the previous report, and 3) GCCAs closed since the previous report. The generic communications listed in the attachment includes bulletins,

_ generic letters, and information notices. Compliance activities listed in the attachment do not rise to the level of complexity that require an action plan, and a generic communication is not currently scheduled. For each GCCA, there is a short description of the issue, scheduled completion date, and name of cognizant staff.

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DISTRIBUTION for NRR Director's Status Report Central File PDR PGEB R/F FMiragliaJRoe T0 Martin, RZimmerman Dmatthews Scollins FAkstulewicz EmcKenna Pwen OEDO EWang Bsweeney Bsheron Gholahan FG111espie WTravers LSpessad Glairas Regional Administrators Mr. William Rasin, Vice President Technical / Regulatory Division Nuclear Energy Institute 1776 " Eye" Street i

Washington D.C.

20006 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 PO Box: A Aiken South Carolina 29892 i

Mr. R. W. Barber Safety and Quality Assurance. DOE 270 Corporate Center (E-853) 20300 Century Blvd.

cj Germantown, MD 20874 7po3 f, p y,Y Y (., - \\ A Mr. S. Scott Office of Nuclear Safety. DOE q

7pg Century 21 Building (E-H72) yF,g-%

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 bhlll)' h{ [))0ll[j[]

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l NRR ACTION PLANS

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e TABLE OF CONTENTS DE BOILING WATER REACTOR INTERNALS CRACKING 1

PRA IMPLEMENTATION 1.2(C) ISI ACTION PLAN.................

5 STEAM GENERATOR ACTION PLAN 10 DRCH UPDATE OF SRP CHAPTER 7 TO INCORPORATE DIGITAL INSTRUMENTATION AND CONTROLS (l&C) GUIDANCE........

12 GRADED QUALITY ASSURANCE ACTION PLAN.................

14 DRPM NEW SOURCE TERM FOR OPERATING REACTORS...............

20 ENVIRONMENTAL SRP REVISION ACTION PLAN,,..............

24 10 CFR 50.59 ACTION PLAN DEVELOPMENT,.................

26 INDUSTRY DEREGULATION AND UTILITY RESTRUCTURING,.......

29 DRPW GENERAL ELECTRIC EXTENDED POWER UPRATE ACTION PLAN.....

32 DRY CASK STORAGE ACTION PLAN 34 DSSA ACCIDENT MANAGEMENT IMPLEMENTATION 37 CORE PERFORMANCE ACTION PLAN........................

40 ENVIRONMENTAL QUALIFICATION TASK ACTION PLAN..........

43 FIRE PROTECTION TASK ACTION PLAN......................

45 HEAVY LOAD CONTROL ACTION PLAN,.....................

47 HIGH BURNUP FUEL ACTION PLAN 50 WOLF CREEK DRAINDOWN EVENT ACTION PLAN,..............

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BOILING WATER REACTOR INTERNALS TAC Nos. M91898, M93925, M93926, Last Update: 08/28/97 M93627, M94959, M94975, M95369, Lead NRR Division: DE M96219, M96539, M97802, M97803, Supporting Division: DSSA M97815, M98266, M98708, M98880 GSI: Not Available MILESTONES DATE (T/C)

PART 1: REVIEW OF GENERIC INSPECTION AND EVALUATION CRITERIA

1. Issue summary NUREG 1544 03/96 C o Update NUREG 1544 12/97 T

2. Review BWRVIP Re-inspection and Evaluation Criteria o Reactor Pressure Vessel and internals Examination Guidelines (BWRVIP-03) 09/97 T BWRVIP-03, Section 6A, Stanc'$rds for Visual Inspection of Core 09/97 T o

Spray Piping, Spargers, and Associsted Components BWR Vessel Shell Wold Inspection Recommendations (BWRVIP-05)'

12/97 T o

o Guidelines for Reinspection of BWR Core Shrouds (BWRVIP-07) 06/97 T

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

4. Review generic mitigation guidelines and criteria TBD

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

The Commission, in SRM M9705128 dated May 30,1997, requested that the staff's SER (a) should address the BWRVIP proposal to examine 100 percent of the axial welds which would include examinations of some circumferential weld lengtns 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 contained in the BWRVIP proposed attemative in determining the acceptability of a proposed technical alternative and/or 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.

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6. Other Internals reviews (safety assessments, evaluations, mitigation measures, inspections and repairs) o Safety Assessment of SWR Reactor Internals (BWRVIP-06) 09/97 T o Evaluation of Crack Growth its BWR Stainless Steel RPV Internals 12/97 T (BWRVIP 14)

Internal Core Spray Piping and Sparger Replacement Design Criteria 12/97 T o

(BWRVIP 16) o Roll / Expansion of Control Rod Drive and in-Core Instrument 12/97 T Penetrations in EWR Vessels (BWRVIP-17) o BWR Core Spray Internals inspection and Flaw Evaluation Guidelines 12/97 T (BWRVIP-18) o BWRVIP-18, Appendix C, BWR Core Spray Internals Demonstration of 12/97 T Compliance With TechnicalInformation Requirements of License Renewal Rule (10 CFR 54.21) o Internal Core Spray Piping and Sparger Repair Design Criteri (BWRVIP.

12/97 T 19)

Core Plate inspection and Flaw Evaluation Guideline (BWRVIP-25) 12/97 T o

o Top Guide inspection and Flaw Evaluation Guideline (BWRVIP-26 12/97 T o Standby Liquid Control System / Core Plate AP inspection and Flaw 04/98 T Evaluation Guidelines (BWRVIP-27) o Assessment of BWR Jet Pump Riser Elbow to Thermal Sleeve Weld 12/97 T Cracking (BWRVIP-28)

Technical Basis for Part Circumferential Weld Overlay Repair of Vessel 05/98 T o

l Internal Core Spray Piping (BWRVIP-34) l Descriotion: Many components inside boiling water reactor (BWR) vessels (i.e., internals) are made of materials such as stainless steel and various alloys that are susceptible to corrosion and cracking. This degradation can be accelerated by stresses from temperature and pressure changes, chemical interactions, irradiation, and other corrosive environments. This action plan is intended to encompass the evaluation and resolution of issues associated with intergranular stress corrosion cracking (IGSCC) in BWR internals. This includes plant specific reviews and the assessment of the generic criteria that have been proposed by the BWR Owners Group and the BWRVIP technical subcommittees to address IGSCC in core shrouds and other BWR internals.

Historical Backaround: Significant cracking of the core shroud was first observed at Brunswick, Unit 1 nuclear power plant in September 1993. The NRC notified licensees of Brunswick's discovery of significant circumferential cracking of the core shroud welds. In 1994, coro shroud cracking continued to be the most significant of reported internals cracking. In July 1994, the N.RC issued Generic Letter 94-03 which requires licensees to inspect their shrouds and provide an analysis justifying continued operation until inspections can be completed.

A special industry review group (Bolling Water Reactor Vessels and Internals Project--BWRVIP) was formed to focus on resolution of reactor vessel and internals 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 operatbn untilinspection or repair on the basis of: 1) no 360' through-wall cracking observed to date, 2) low frequency of pipe breaks, and 3) short period of operation (2-6 months) before all of the highly susceptible plants complete repairs of or inspections to their core shrouds.

In late 1994, extensive cracking was discovered in the top guide and core plate rings of a foreign reactor. The design is similar to General Electric (GE) reactors in the U.S., however, there have 2

0 been no observations of such cracking in U.S. plants. GE concluded that it was reasonable to expect that the ring cracking could occur in GE BWRs with operating time greater than 13 years. In the specialindustry review group's report, that was issued in January 1995, ring cracking was evaluated. The NRC concluded that the BWRVIP's assessment was acceptable and that top guide ring and core plate ring cracking is not a short term safety issue.

Prooosed Actions: The staff will continue to assess the scopes that have yet to be f;ubmitted by l

licensees concerning 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 regarcing 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 internals. The BWRVIP has submitted 15 generic documents, supporting plant-specific submittels, for staff review. The staff is ensuring that the generic reviews are incorporating recent operating experience on all BWR internals.

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

Recu!storv Assessment: In July 1994, the NRC issued Generic Letter 94 03 which required licensees to inspect their shrouds and provide an analysis justifying continued operatloa until inspections could be performed. The staff has concluded in all cases that licensees have provided sufficient evidence to support continued operation of their BWR units to the refueling outages in which shroud inspections or repairs have been scheduled. In addition, in October 1995, industry's special review group submitted a safety assesament of postulated crricking in all BWR reactor internals and attachments to assure continuing safe operation.

Current Status: Almost all BWRs completed inspections or repairs of core shrouds during refueling outages in the fall of 1995. Various repair methods have been used to provide alternate load carrying capability, including preemptive repairs, installation of a series of clamps and use of a series of tie-rod assemblies. The NRC has reviewed and approved all shroud modification proposals that have been 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 has sent a request for additionalinformation. The BWRVIP provided its response to the RAls in a letter dated December 20,1996, The staff met with the BWRVIP to discuss its expanded basis for prioritization as part of W entinuing review of BWRVIP-06. The staff requested additionalinformation from the BWRVIP ring this meeting in April 1997 in order to complete the review of BWRVIP-06; the BWRVIP has e

submitted the requested information and the staff is completing the review. In addition, the industry group subm!tted a report on reinspection of repaired and non-repaired core shrouds (BWRVIP-07) in February 1996. The staff is currently reviewing both this report and the supplemental information provided in the BWRVIP's response to the NRC staff's request for additionalinformation. The NRC is also reviewing information submitted by GE on the safety significance of and recommended inspections for top guide and core plate ring cracking Review of the " Reactor Pressure Vessel and Internals Examination Guidelines (BWRVIP-03)" is continuing with RAls sent in February 1997. By letter dated September 20,1996, the BWRV P 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 3

i 100% of circumferencial welds and 0% of longitudinal welds. The staff is developing its safety evaluation on this issue. The NRC staff will complete its evaluation of the BWRVlP-05 report by December 1997. Informaion Notice (IN) 97-63, " Status of NRC Staff's Review of BWRVIP-05,"

was issued August 7,1997, to inform the industry of botn the status of the staf f's review and that the staff would consider technically-Justified schedular relief requests from performing augmented inspections of the RPV shell circumferential welds for 40-months or two operating cycles, whichever was loager. The staff's independent assessment of the BWRVIP-05 report was transmitted by letter dated August 14,1997, to the BWRVIP, along with a request for additional information and information that needed to be addressed for licensees requesting schedular relief.

The staff is presently reviewing one such relief request. The statf briefed the ACRS subcommittee on Augus,t 26,1997, and will brief the full committe on September 4,1997.

The staff's review of BWRVIP 14 is continuing, and RAls were issued on December 9,1996. The staff is awaiting a response from the BWRVIP. The staff's review of BWRVIP 18 anc 19 on internal core spray piping inspection and repair design criteria is continuing. RAls on inese two documents were issued on January 16,1997.

By letter dated December 20,1996, the BWRVIP submitted, " Appendix C to BWRVIP 18. This appendix addressoa the use of BWRVIP generic internal 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.

The BWRVIP submitted BWRVIP-28 to address the safety implications of recent cracking found in BWR jet pump riser elbows. The staff is reviewing the BWRVIP-28 report and is developing RAls.

The staff issued NRC Information Report IN 97-02, " Cracks Found in Jet Pump Riser Assembly Elbows at Boliing Water Reactors," on Fcbruary 6,1997 and is developing a generic letter on the same subject.

Information Notice 9717, " Cracking of Vertical Welds in the Core Shroud and Degraded Repair,"

was issued April 4,1997, to inform the industry of vertical weld cracks and a degraded core shroud repairs found at Nine Mile Point, Unit 1. The BWRVIP has informed the staff that it plans to revise BWRVIP-07 to ensure that the vertical core shroud welds, and the core shroud repair, is adequately inspected.

By letters dated April 25 and May 30,1997, the BWRVIP provided a reaffirmation of the BWR member licensees to the BWRVIP, and committed, on behalf of your member licensees, to several actions, including implementing the BWRVIP topical reports at each BWR as appropriate considering individual plant schedules, configurations and needs, and providing timely notification to the NRC staff if a plant does not implement tho 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.

Technical Contacts:

Keith Wichman, EMCB, 415 2757 Merrilee Banic, EMCB, 415-2771 Kerri Kavanagh, SRXB, 415 3743 Frank Grubelich, EMEB 415 2784 NRR Lead PM:

C. E. Carpenter, EMCB, 415-2169

References:

Generic Letter 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," July 25,1994 Action Plan dated April 1995 4

PRA IMPLEMENTATION ACTION PLAN 1.2 (c)

Inservice inspection Action Plan TAC Nos. M93710, M95125, Last Update: 08/30/97 M95126, M97153 Lead NRR Division: DE Support Division: DSSA RG/SRP MILESTONES DATE (T/C) 1.

Draf t for RI-ISI team review / comments 04/05/96 C 2,

1st draft for Branch Chiefs review / comments 08/14/96 C 3.

Revised draft for Branch Chiefs review / comments 01/24/97 C 4.

Revised draft for Branch Chiefs review / comments 04/08/97 C 5.

Draft for Division Director review / comments 04/29/97 C 6.

Draft for Office Director /OGC review / comments 05/16/97 C 7.

Office Director /OGC concurrence 07/08/97 C 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 12.

Meeting with ACRS Subcommittee 07/08/97 C 13, Meeting with ACRS full Committeo 07/09/97 C 14.

Meeting with CRGR 07/17/97 C 15.

SECY from EDO to Commissioners (SECY 97190) 08/20/97 C 16.

Publish draft for public comments 09/97 T 17.

Public comment period for draft RG/SRP ends 12/97 T 18.

Complete final draft for ACRS/CRGR review / comments 02/98 T 19.

Complete final draft for Inter-Office concurrence 03/98 T 20.

Publish final RG/SRP 04/98 T 5

PILOT PLANTS MILESTONES DAT E (T/C) 21.

Meeting with NEl and Industry to discuss RI ISI plan 02/10/90 C 22.

Meeting with NEl and Industry to discuss revised Rl ISI 02/28/90 C plan 23.

Meeting with NEl and industry to discuss technical details 03/0E/96 C

~

of WOG and EPRI methodologies 24.

Meeting with NEl and industry to discuss technical details 09/20/90 C of WOG and EPRI methodologies 25.

Meeting with ASME and Industry to discuss ASME Code 03/10/97 C Case N 500 26.

Meeting with NEl to discuss coordination of industry 04/16/97 C submittels and ASME Cods Cases c

l 27.

Meeting with NEl and EPRI to discuss technical details of 04/29/97 C EPRI methodology l

l 28.

Receive complete submittals of pilot plants (Surry 1, ANO.

09/97' T 2 Fitzpatrick) 29.

Issue SERs on pilot plants 12/98' T i

Subject *o change based on licensees' actual submittat dates.

Descriotion: Develop risk informed inseivice 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 ilSI) application and related American Society of Mechanical Engineers (ASME) Code Cases: review acceptability of th3 three pilot programs with respect to their RI ISI applications and prepare plant specific safety evaluation reports (SER).

11istorical Backaround: On August 16,1995, the U.S. Nuclear Regulatory Commist.lon (NRC) published a policy statement (60FR42622) on the use of probabilistic risk assessramt (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 Di'ector 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 defineated specific tasks in the RI ISI area in two years as directed by Chairman Jackson.

The nuclear industry has submitted two methodologies for implementing RIISI. One methodology has been jointly developed by ASME Research and Westinghouse Owners Group (WOG) (Reference

4) and the other methodology is being sponsored by Electric Power Rosesrch Institute (EPRI)

(Reference 5).

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ASME is working on three Code Cases for alternate examination requirements to ASME Section XI.

Division 1 for piping welds. Code Case N 577 is based on the WOG methodology and Code Cases N 578 is based on the EPRI methodology. Code Case N 500 is based on the EPRI methodology but is being revised to encompass both methodologies.

Pronosed Actions: The NRC has encouraged licensees to submit pilot plant applications organized under one umbrella sponsoring organization, e.g., Nuclear Energy Institute (NEI), for demonstrating risk informed methodologies to be used for piping segment and piping structural element selection in systems scheduled for ISI. The N3C will review the industry submittels with focus on the licensees chcracterizing the proposed change including the identification of the particular piping systems and welds that are affected by the change, engineering evaluations perfor*ned, 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 IMensing basis of the plant, develovnent 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 frorn the risk ranking of piping elements, and in cooperation with the RES staff, a parallel of fort will be carried out to develop: (a) an RI ISI application specific RG and (b) the corresponding SRP chapters and associated inspection proceduto documents.

The nuclear industry, under the umbrella of NEl, has submitted two methodologies for the implementation of the RI-ISI. One methodology has been jointi;fsveloped by ASME Research and WOG (Reference 4) 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 ANO 2 and Fitzpatrick.

The acceptability of the three RI ISI pilot plant programs will be documented in SERs for each of the three pilot plant licensees and forwarded to the Commission. Upon Comm:ssion approval, the staff willissue SERs authorizing alternative inspes Jor strategies br the pilot plant licensees to allow use of the RIISI methodology.

ASME is working on three Code Cases for alternate examination requirements to ASME Section XI.

Division 1 for piping welds. Code Case N-560, for the alternate ecamination requirements for Class 1, Category B J piping welds, is based on the EPRt methodology and has been approved by the ASME Section XI Subcommittee. This Code Cese is bems levised to encompass both WOG and EPRI methodologies. Code Case N 577, for the alternate examination requirements for Risk Based Selection Rules for Class 1,2, and 3 piping welds, is based on the WOG methooology and is being reviewed by the ASME Section XI Subcommittee. Code Case N 578, for the alternate examination requirements for Risk Based Selection Rules for Class 1,2, and 3 piping welds, is based on the EPRI methodology and is also being reviewed by the ASME Section XI Subcommittee. The NRC staff intends to work with the industry and ASME to ensure integration and consistency of the various approaches being proposed.

ASME Code Case N 577, based on the WOG methodology, is expected to be approved by the ASME Section Xi Subcommittee during 1997. The major difference between Code Case N 577 and the WOG methodology submitted to the staff (Reference 4) is that the scope of the Code Case is limited to ASME Class 1,2, and 3 systems while the WOG methodology encompasses 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). 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 incorporated into RG 1.147 with any necessary additions or deletions. The pilot plant Rl ISI program review will be documented in the staff SER.

7

ASME Code Case N 578, based on the EPRI methodology,is also expected to be approved by the ASME Section Xl Subcommittee during 1997. 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 encompasses all safety significant systems in the plant. Also, the Code Case is an abbreviated version and does not have all the details presented in the EPRI topical report (Reference 5). The staff will review the EPRI methodology as well as Code Case N-578 and the consistency of the ANO 2 and Rtzpatrick RIISI pilot programs to both of these Code Case N 578 will be reviewed and,if found acceptable, will be incorporated into RG 1.147 with any necessary additions or deletions. The pilot plant RI ISl programs' review will be documented in the staff SER.

Code Case N 5t0 for the alternate examination requirements for Class 1, Category B J piping welds is being revised to e.ncompass 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 use tha acceptable alternative provision of 10 CFR 50.55a (al(3//// to approve the pilot plants' applications. The staff will work closely with ASME to expedito changes involving ISI.

Orialnatina Documents: In a November 30,1995, memorandum to J. M. Taylor, the NRC EDO, Chairman Jackson directed that the staff prepare an action plan together with a timetable for developing RGs and SRPs applicable to use of PRAs to be completed in two years. In his response of January 3,1990, the EDO presented a plan that established milestones for the development of regulatory guidance documents for utilizing PRA in reactor related activities including ISt. This i

action plan is in conformance with the agency wide implementation plan for PRA and any future changes will be consistent with the overall plan.

Reaulatorv Assessment! The operational readiness and functionat integrity of certain safety related piping and associated structural elements (e.g., pressure retalning welds) are vital to the safe operation of nuclear power plants. ISIis one of the mechanisms used by the licensees to ensure piping integrity. The type and frequency of ISI are based on past experience and collective best judgment of the NRC and industry in a consensus Code 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 costo in ways that do not have an adverse effect on safety.

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

The NRC lh using probabilistic methods, as an adjunct to deterministic, techniques to help define the scope, type, and frequency of ISI. The development of Rl ISI programs has the potential to optimize the use of NRC and industry resources and continue to assure adequate protection of i

public health and safety.

Acceptability of the RI ISI pilot programs is expected to be documented in safety evaluations. The staff recommendation whether to authorize an alternative inspection program pursuant to 10 CFR 50.55a (al(3//// will be presented to the Commission prior to its implementation. To provide the permanent opproach to Rl ISI, the staf f intends to utilize the experience gained through the pilot applications in the proposed rulemaking process to modify 10 CFR 50.55a to explicitly endorse RI-ISI methodology, 8

Current Status; Since the formation of the RI ISI team, several meetings have been held with NEl and industry / utility representatives, in these meetings, the NRC staff and industry have discussed their respective plans for the RI ISI programs. NEl has submitted WOG technical report (Reference 4), and EPRI technical report (Reference 5). The statf has also been actively participating in ASME Code activities related to RI-lSt.

The staff has completed draf ts of RIISl RG and SRP which have been submitted to the Commissioners (SECY 97190) to request approval to publish the RG and SRP for public comments.

The staff made its initial presentation of the Rl ISI program to the ACRS on June 11,1997.

Detailed presentations by the staff and industry to the ACRS were made on July 8,1997 during its i

443rd meeting. The ACRS has recommended (i.etter dated July 14,1997) that these documents be issued for public comments subject to incorporation of changes in response to ACRS comments.

The staff also met with the Committee to Review Generic Requirements (CRGR) on July 17,1997 during its Meeting No. 308. CRGR has issued its letter usted August 14,1997, in which it stated that CRGR has no objection to the proposed RG and SRP being issued for public comments, subject to the incorporation of its comments and recommendations. A workshop to solicit and discuss public comments is also scheduled for October 30 and 31,1997.

The pilot plant submittels scheduled for 1996 (ANO 2, Fitzpatrick, and Surry) have not been received by the staff. The staff has submitted its request for additionalinformation (RAll on the EPRI methodology to the industry.

NRR

Contact:

S. All(415 2776), S. Dinsmore (415 8482)

RES

Contact:

J. Guttmann (415 7732)

Referencent 1.

Federal Register, Vol,60, No.158, "Use of Probabilistic Risk Assessment Methuds in Nuclear Regulatory Activities; Final Policy Statement," August 16,1995.

2.

Memorandum from Shirley Ann Jackson, Chairman, NRC to James M. Taylor, Executive Director for Operations, ' Follow up Requests in Probabilistic Risk Assessment and Digital Instrumentation and Control," November 30,1995, 3.

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

4.

WCAP 14572, " Westinghouse Owners Group Appt' cation of Risk Based Methods to Piping inservice inspection Topical Report," March 1996.

5.

EPRI 1R 106706,* Risk Informed inservice inspection Evaluation Procedure," June 1996, l

9

STEAM GENERATORS Last Update: 8/31/97 TAC No. 88885 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)

I S. Proposed GL Pkg 9/97 (T)

6. ACRS Endorsement 10/97 (T)

I

7. CRGR Concurrence 11/97 (T) 8.EDO 12/97(T)

9. Concurrence 10. Publish Proposed GL 12/97(T)

Orig. Publish Proposed Rule 03/95 (C) i

10. Public Comment 12/97 (T)

11. Revise GL Pkg 1st qtr 98 (T)

12. ACRS Comments 1st qtr 98 (T)

13. CRGR Concurrence 2nd qtr 98 (T)

14. EDO Concurrence 2nd qtt 98 (T)

15. Commission Approval 6/98 (T)

16. Publish Final GL 7/98 (T)

Or/g. Publish Finel Rule 12/95 10

Brief Descrintl0D: The NRC originally planned to develop a rule pertainin0 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 rep'n approach was to utilire a generic letter. The NRC staff suggested, and the Commission subsequently approved, a revision to the regulatory approach to utilite a generic letter.

flgoulatorv Assessment: The current regulatory framework provides reasonable assurance that operating PWRs are safe. However, the current regulatory framework has numerous shortcomings.

To resolve these shortcomings the staff is revising the regulatory framework to utilize a risk-informed and performance based approach that will ensure compilance with current regulations I

(i.e., GDC, Appendix B, ASME code, Part 100).

l Current Status: The ACRS was briefed on the ANPR on Augus' 3 4,1994. The staff revised the ANPR based on Commission comments in the SRM dated August 23,1994, which indicated that the EDO should add the issues raised in the recent DPO on this subject. The advanced notice of proposed rulemaking was published in the FederalRep/ ster on September 19,1994. The comment period expired on December 5,1994. Two comments were received. The staff has reviewed the comments. A Commission paper (SECY 95131) to justify continuance of rulemaking in light of the public comments was issued. Th9 staff briefed the Commission concerning steam generator problems and the justification for continuance of rulemaking on June 1,1995. The staff briefed the i

Commission on February 27,1996 and apprised the Chairman that the SG rulemaking scheduled required revision. The staff developed a revised schedule for draf t rule issuance and provided the schedule to the Commission via a memorandum signed by the EDO on May 20,1996. The staff briefed the Chairman on July 18,1996 concerning the current status and schedule for the draft SG rule. An information briefing by the staff was held on October 29,1996, for the CRGR. The proposed rule was presented to ACRS on November 5,6,1996. The ACRS provided comments to the staff in a letter dated November 20,1996.

The staff briefed the Chairman on the current 7

j status and the revised schedule on December 10,1996. The staff responded to the November 20, 1996, letter from ACRS in a letter dated January 2,1997. Subsequently the staff met with the ACRS on January 9,1997.

The staff has completed a draf t risk assessment and draf t regulatory analysis and met with ACRS on March 4,5, and April 3,1997, to discuss the two efforts. The results of these two efforts caused the staff to conclude that the regulatory approach needed to revised. The staff subsequently drafted, and sent to the Cornmission COMSECY 97 013 (May 23,1997) which discussed the basis for revisin0 the regulatory approach to utilize a generic letter. The Commission approved the revised approach in the SRM dated June 30,1997.

NRR Technical Contsets:

Ted Sullivan, EMCB, 415 3266 Tim Reed, EMCB, 415 1462 RES

Contact:

N/A 11

i UPDATE OF SRP CHAPTER 7 TO INCORPORATE DIGITAL INSTRUMENTATION AND CONTROLS (l&C) GUIDANCE

• Final Update

• TAC Nos. M86387, M86392, M86423, Last Update: 08/26/97 i

M86769, M86997, and M87680 Lead NRR Division: DRCH a

MILESTONES DATE (T/C) 1.

Develop Update of SRP Chapter 7 10/95C 2.

ACRS Subcommittee Briefings 3/96C, 5/90C,10/96C 4

3.

Incorporate new Regulatory Guides (provided by 8/96C RES)in SRP Chapter 7 Update 4.

Draf t SRP to Chairman 9/19/96C' 1

5.

Publish Draft SRP Chapter 7 for Public Comment 12/03/96C 6.

Incorporate Public Comments and National 6/97C i

Academy of Sciences study recommendations 7.

Final ACRS/CRGR Review of SRP Chapter 7 6/97C I

1 8.

Final SRP to Chairman 7/17/97C 1

9.

Publish Final SRP Chapter 7 7/31/97C l

Descrintion: This task action plan is used to track and manage the final phase of codifying the digitall&C regelatory approach and criteria by updating the existing Standard Review Plan (SRP)

Chapter 7.

4 Historical Backaround: By a staff requirements memorandum (SRM) dated November 30,1995, from the Chairman, Shirley Ann Jackson, to the Executive Director of Operations, James M. Taylor, the Chairman requested that the staff develop an action plan in the area of dighalInstrumentation and controls. The action plan is for the expeditious development of a Standard Review Plan (SRP) to ensure that safety margins are addressed and that NRC regulatory requirements are available and ready for use when reviewing licensee proposed installation of digitalinstrumentation and control systems in nuclear power plants. The staff has an ongoing effort for updating Chapter 7 of the SRP that deals with instrumentation and control systems to accomplish the requested action and this task action plan was initiated to track and manage the final phase of that effort in response to the SRM.

i Etangled Actions: Specific actions included in this task action plan are: (1) to develop the update of SRP Chapter 7, (2) to periodically brief the ACRS as sections of the SRP update are completed, (3) to incorporate new regulatory guides on digital l&C that will be provided by the Office of Nuclear Regulatory Research (RES), (4) to incorporate results from the National Academy of Sciences (NAS) study of digital I&C at nuclear plants, (5) to publish the draft SRP Chapter 7 for public comments, (6) to incorporate the public comments, (7) to have final ACRS and CRGR review of the SRP Chapter 7 update, and (8) to publish the final revised SRP Chapter 7.

12

Orlainatino Document: The memorandum from the EDO to Chairman Jackson dated January 3, 1996, *lmprovements Associated with Managing the Utilization of Probabilistic Risk assessment (PRA) and Digital Instrumentation and Control Technology "

1Reaulatorv Assessment: The approach and criteria that form the current regulatory framework for review and acceptance of digitall&C systems in nuclear power plants is being codified in the update to SRP Chapter 7. This framework has been communicated to the industry and public in safety evaluations for digital modifications to operating plants and design certification of the advanced reactor designs, and in Generic Letter 95 02, "Use of NUMARC/EPRI Report TR.102348, Guideline on Licensing Digital Upgrades,' in Determining the Acceptability of Performing Analog.to.

Digital Replacements Under 10 CFR 50.59 dated

• dated April 20,1995. This action plan tracks and manages the codification of the existing framework by updating SRP Chapter 7. Consequently, this is not an urgent regulatory action, and continued plant operation is justified.

l Current Status: The staff and its contractor, Lawrence Livermore National Laboratories (LLNL),

l have revised the seven existing sections of SRP Chapter 7 and developed two new sections and I

several new branch technical positions (BTPs) to incorporate criteria and guidance related to digital l&C systems, in parallel, the Office of Nuclear Regulatory Research (RES) has developed several regulatory guides that endorse national standards related to digitall&C.

The updated draf t SRP Chapter 7 was issued for public comment and the notice of availability was published in the Federal Register on December 3,1990, it was also posted on the NRC Homepage on the World Wide Web in December 1996.

The public comment period closed on January 31,1997 and all public comments received in February 1997 are being addressed in the revision of SRP Chapter 7. The National Research Council / National Academy of Sciences' (NAS) final report on Digital Instrumentation and Control Systems in Nuclear Power Plants, Safety and Reliability issues was received by the staff in late January 1997. The recommendations in the report were reviewed andappropriate changes incorporated in the revision to SRP Chapter 7. The revised SRP Chapter 7 was presented to ACRS on June 12,1997 and to CRGR on June 23,1997. By letter dated June 23,1997, the ACRS endorsed the final SRP Chapter 7 update and associated regulatory guides. In the minutes of CRGR Meeting Number 307 dated August 14,1997, CRGR endorsed the document for issuance. The notice of availability of the updated SRP Chapter 7, Revision 4 was published in the Federal Register on July 31,1997.

Contacts:

Matthew Chiromal DRCH,415 2845 Joe Joyce DRCH,415 2842 13

PRA IMPLEMENTATION ACTION PLAN 1.2(d)

Graded Quality Assurance Action Plan TAC Nos. M91429, M91431, M92420, Last Updato: 09/05/97 M92450, M92451, M92447, M92448, Lead NRR Division: DRCH M92449, M88650, M91431, M91432, Support Division: DSSA M91433, M91434, M91435, M91436, GSI: Not Available M91437 MILESTONES DATE (T/C)

1. lssued 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 Grcup meetings to guide working level statf activities

- As Needed Meetings on: 8/25/95,10/10/95,10/25/95

4. Staff interactions with Palo Verde Ongoing Site visit on 5/23/95 on ranking and QA controls through NRC letter dated 7/24/95 on proposed QA controls Site visit on 8/29 30/95 on risk ranking Site visit on 9/6 7/95 on procurement GA controls 3/98 NRC letter conveying trip reports issued on 12/4/95 Meeting on 4/11/96 to discuss the staff evaluation guide Letter from licensee on 4/24/96 providing comments on staff evaluation guidance

- Site visit on 6/5 6/96 to observe expert panel and review revised procurement QA controls, trip report sont to licensee on 8/6/96 Letter from licensee on 9/12/96 transmitting responses to procurement issues raised in earlier statf 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 14

5. Staf f interactions with South Texas Ongoing i

Meeting on 7/17/95 on project status through Site meeting on 10/3 4/95 on risk ranking and QA controls Meeting on 12/7 8/95 to discuss risk ranking and QA controls South Texas Submittal of QA Plan for implementation of graded j

QA, dated 3/28/96 is currently under staff review W38 l

Meetings on 4/11/96 and 4/25/96 to discuss the staff evaluation i

guide and future interaction milestones and schedules Letter from licensee on 4/17/96 providing comments on staf f evaluation guidance

]

Meeting on 6/19/96 to discuss staff comments on the 0.A plan i

submittal for graded QA, review questions transmitted to STP on 8/16/96 Site visit on August 2122 to observe working group and expert j

panel meetings, and to discuss staff review items, trip report in j

preparation Management meeting on 10/15/96 to discuss PRA Initiatives and j

staff activities Letter from licensee dated 10/30/96 responding to PRA questions Revised QA plan submitted on 1/21/97 j

Overview of STP Initiative provided at 2/27/97 meeting with the staff f

Staff Request for AdditionalInformation (RAll issued on 4/14/97 for both PRA and QA controls 1

j Meeting on 4/21/97 to discuss STP responses to RAl j

Site visit on 5/5 8 to evaluate: PRA quality, graded QA controls, QA j

controls for the PRA, corrective action and performance rnor.iisiing feedback processes, audit scheduling, and responses to the RAI concerns. Trip report issued on 7/10/97.

STP submittal on 5/8/97 for preliminary RAI response STP submittal of draft QA Plan on 5/21/97 9

STP submittal of GOA related procedures, responses to RAI, and j

follow on QA Plan on 5/22/97 j

STP submittal of revised QA Plan on 7/10/97

- Staff RAI losued on 6/13/97 STP submittal on June 26,1997 response to staff RAI STP submittal of revised QA Plan on 7/16/97 j

STP transmittal of additionalinformation regarding GOA l

implementing procedures and associated change control on 7/31/97 i

STP submittal on 8/4/97 responding to PRA RAI and provided '

procedures related to shutdown operations Negative consent SECY paper and Safety Evaluation has been prepared that documents the staff's review of the QA program change.

15

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 Meeting at RIV on 11/10/95 on graded QA effort Site meeting on 11/17/95 to observe expert panel 3/98 GGNS system and component ranking criteria under staff evaluation, the comments are scheduled to be provided to GGNS by the end of June Meeting on 4/11/96 to discuss the staff evaluation guide

- Letter to GGNS dated 5/29/96 regarding implementation of QAP commitments Staff review comments on GGNS safety significance determination process transmitted to licensee on July 15 Meeting on August 27 to discuss staff comments on safety significance process and to discuss GGNS implementation of QAP commitments for low safety significant items, meeting summary issued on 12/17/96 Site visit on 11/21/96 to review procurement activities, trip report has been prepared

7. Revision 3 of Draft Evaluation Guide for Volunteer Plants issued for st6ff 07/95C comment

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

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

1/96C i

10. Evaluation Guide lasued for use by staff in evaluat!ng volunteer plants 1/96C Meeting held with volunteer plants to receive feedback on staff evaluation guide on 4/11/98.

4/96C I

- Industry comments on staff evaluation guidc provided by letter dated 5/24/96 The staff reviewed the industry comments with respect to the l

need to revise, and finalize, the evaluation guide.

11. Regulatory Guido development milestones per PRA Action Plan Draft RG for Branch / division review and comment 7/31/96C Draft RG for inter office review and concurrence 8/1/96C Draft RG for ACRS/CRGR review 11/22/96C Draf t RG for public comment 6/25/97C

- Draft RG public comment period ends 9/23/97T Public workshop held on draft RG 8/12/97C Publish final RG 12/31/97T

12. ACRS Briefings Expert Panel and deterministic consideration:

2/27 28/96C graded QA 4/11/96C

- PRA implementation Plan and pilot projects 7/18/96C

- Risk Informed Pilots 8/7/96C Graded QA Regulatory Guide 11/22/96C Graded QA Regulatory Guide 2/21/97C

- ACRS Concerns on GOA Regulatory Guide 3/6/97C ACRS memo to Commission expressing concerns with GOA 3/17/97C epproach 16

13. CRGR Briefings

. Graded QA Regulatory Guide 11/26/96C Graded QA Regulatory Guide 3/11/97C

14. Issue Lessons Learned NUREG report regarding Graded QA Programs at 2/98T volunteer plants

15. Public Workshop on Graded QA 4/98T

16. Issue draf t Staff Inspection Guidance (Baseline + Reactive IP) for 12/97T comment

. Issue finalinspection procedure 2/98T

17. Conduct NRC Staff Training 3/98T

18. Issue SECY Update (close-out of action plan) 4/98T Descriotion: Prepare staff evaluation guidance and regulatory guidance for industry implementation l

for the grading of quality assurance (QA) practices commensurate with the safety significance of the plant equipment. The development of this guidance will be based on statf reviews of regulatory requirements, proposed changes to existing practices, staff development of a draf t 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 60, 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 practicos that have evolved have often not been graded. in the development of implementation guidance for the maintenance rule, a methodology j

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 initialindustry guidelines were developed and commented on. In early 1995, three licensees (Grand Gulf, South Texas, and Palo Verde) volunteered to work with thr; staff. The staff has reviewed the licensee developmental graded QA efforts.

Prooosed Actions: The goal of the action plan is to utilire the lessons learned from the 3 volunteer licensees to modify staff-developed draft guidance to formulate regulatory guidance on acceptable methods for imp lementing 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 activ; ties.

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

Reaulatory Assessment: Existing regulations provide the netassary flexibility for the development and implementation of graded quality assurance programs. The staff willissue a NUREG report regarding the lossons learned from the volunteer plant implementetions. Additional regulatory guidance will be issued to either disseminate staff guidance or endorse an industry approach.

Planned guidance for the staff willinvolve an evaluation guide for application to the volunteer plants, the lessons learned report, training sessions and public workshops, and inspection guidance in the form of a baseline and a reactive IP. The staff is evaluating the appropriate mechanism for inspections of the risk significance determination aspects of graded QA programs.

17

The safety benefits to be gained from a graded QA program could be significant since both NRC reviews and inspections and the industry's quality controls resources would be focused on the rnore 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 onectiy related to the overall PRA implementation plan schedules.

Current status: A draft evaluation guide for NRC staff use has been prepared for application to the volunteer plants implementing graded quality assurance programs. The staff will utilize the guide for the review of the volunteer plant graded QA programs. The guide and the staff's proposed interaction framework has been transmitted in a letter to the three volunteer licensees. The letter sought licensee comments. Draft regulatory guides for both risk ranking and grading of QA controls have been prepared and circulated for review by both the ACRS and CRGR. SECY 97 077 (dated April 8,1997) transmitted the draf t regulatory guldes, 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).

i A meeting was held with the three volunteer licensees on April 11,1996 to reesive their feedback on the staff developed evaluation guide. The licensees expressed concerns about the level of detail contained in the guide, particularly that related to PRA and commerchi grade item dedication. The licensees contend that exiting industry guidance (PSA Application Guide and EPRI 5652) are sufficient for those toples. The staff received written comments from NEl on the evaluation guide by letter riated May 24,1996. The NEl letter questions the need for additional regulatory guldance 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 M the staf. exp6ctations for the conduct of expert panel activities. The ACRS was further L.cfed on the development of the GQA Regulatory Guide on November 22,1996 and F6bruary 21,1997, and March 6,1997. The ACP.S lasued a letter to the Chairman on March 17,1997 regarding their review of the risk informed guichnce documents. The ACRS expressed some concerns with the staff focus on simply proposing to reduce quality controls for low safety significant items. However, in recognition of industry interest in the guide, the ACRS recommended that it be issued for public comment.

South Texas submitted their QA program revision for their graded QA effort on March 28,1996.

The change has been reviewed by the staff (HOMB, SPSB, RES, RIV, and NRC contractors). A meeting was held with STP on June 19 to discuss the staff's comments and concerns. STP indicated their willingness to re-examine the content of the QA plan with respect to the proposed QA controls for the low safety significant items. The staff visited the site on August 2122 to receive information from STP in response to earlier staff questions about the STP approach towards determining safety significance categorization and adjustment of QA controls. The staff also observed both a Working Group and Expert Panel meeting at which time licensee safety significance evaluations for 2 systems (Radiation Monitoring and Essential Service Water) were discussed. Staff review of the updated QA program submittal was completed and a second RAI was issued on April 14,1997 for both PRA and QA controls aspects. A meeting was held on April 21,1997 during which the licensee provided some responses to the issues raised in the RAl. Staf f (from both HOMB and SPSB) performed a site evaluation during the week of May ~ 8 to review aspects associated with: PRA quality, QA controls for the PRA, corrective action t, $d performance monitoring feedback processes, QA controls for low safety significant items, detailed information 18

0 presented to address issues raised in the RAl, and the audit scheduling process. Further dialogue has occurred between the staff and STP during the revie v of the subsequent STP submittals and following issuance of staff RAls. The staff has prepared a Safety Evaluation fur the STP OA program change (which is in management concurrence review), and a companion negative consent SECY paper to inform the Commission of the staff's review conclusions, i

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 las 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 Jene 8, NElindicated that their 96-02 guide will be revised. Further NEl requested a meeting with the statf (in the August time frame) to discuss the changes and to discuss more objective means to assess the j

adequacy of OA program implementation. NEl has proposed that the amended 96 02 guidelines

- will be submitted to the staff for endorsement by a regulatory guide. A subsequent letter was l

received from NEl on July 16 that provided an updated version of NEl 96 02 based on comments they received from the volunteer plants and industry sources. The staff has reviewed the modified document. On October 10,1996 NEl submitted a letter expressing their concern with the graded QA initlative. NEl stated their concerns regarded the questions talsed by the staff in the area of QA controls for items determined to be low safety significant and in the area of setety significance determination. A meeting with NEl and staff from the volunteer plants (STP and PVNGS) was held on February 27,1997. NEl ttated 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 diaft GQA 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 etforts.

NBR

Contact:

S. Black 415 1017, R. Gramm 4151010 RES

Contact:

J. Guttman 415 7732

References:

1)

Letter from J. Sniezek (NRC) to J. Colvin (NEl) dated 1/6/94 2)

Regulatory Guide 1.160 3)

NUMARC 93-01, " Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" 4)

SECY 95-059, " Development of Graded Quality Assurance Methodology",3/10/95 5)

Letter from B. Hollan (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 Staf f Evaluation Guidance 8;

NEl 96-02, " Guideline for implementing a Graded Approach to Quality" 9)

Dtatt Regulatory Guide 1064, "An Approach for Plant Specific, R!sk informed Decision Making: Graded Quality Assurance", dated March 24,1997 19

NEW SOURCE TERM FOR OPERATING REACTORS TAC No. M89586 Last Updato: 09/09/97 GSI No.155.1 Lead NRR Division: DRPM CTL l.J Supporting Division: DSSA & DE MILESTONES DATE (T/C) 1.

NEl Letter 07/94C 2.

Cornmission Memo 09/94C 3.

NEl Response 09/94C 4.

NEl/NRC Meeting 10/94C 5.

Publication of NUREG 1465 02/95C 6.

NEl/NRC Meetings

/

10/94 C, 06/95C,10/95C, 01/90C, 02/96C, 05/96C, 08/96C,10/96C,04/97C 7.

Submittal of Generic Framework Document (from NEI) 11/95C 8.

First Pilot Plant b1bmittal 12/95C 9.

Issue Memo to Commission, Updating Status 08/96C 10.

Present Commission Paper In E Team Briefing 09/96C 11.

Brief CRGR on Commission Paper 10/96C 12.

Send Commission Paper to EDO/ Commission 11/96C 13.

Brief ACRS on Commission Paper 11/90C l

14.

Response to NEl Framework Document 02/97C l

15.

Begin Pilot Plant Reviews 02/97C 16.

Begin Rebaselining 02/97C 17.

Brief E Team on Status of Rebaselining 07/97C 18.

Issue User Need for Rulemaking 08/97C 19.

Finish Rebaselining 11/98T 20.

Finish Pilot Plant Reviews 06/99T Descrintion: 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 accidento. 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 sacriticing real safety margin. In addition, safety enhancements may also be achieved.

20

Historien! Backaround: In 1962, the U. S. Atomic Energy Commission published TID 14844,

" Calculation of Distance Factors for Power and Test Reactors." Since then licensees and the NRC have used the accident source term presented in TID 14844 in the evaluation of the dose l

consequences of design basis accidents (DBA).

After examining years of additional research and operating reactor experience, NRC published NUREG 1465, " Accident Source Terms for Light Water Nuclear Power Plants," in February 1995.

The NUREG describes the accident source term as a series of five release phases. The first three phases (coolant, gap, and early in vessel) are applicable to DBA evaluations, and all five phases are applicable to severe accident evaluations. The DBA source term from the NUREG is comparable to the TfD source term; however, it includes a more realistic description of release timing and composition. Since the NUREG source term results in lower calculated DBA dose consequences, NRC decided not to require current plants to revice their DBA analyses using the new source term.

However, many license 0s want to use the new source term to perform DBA dose evaluations in support of plant, technical specification, and procedure modifications.

NRC and NEl met several times to discuss the industry's plans to use the new source term. To make efficient use of NRC's review resources, NRC encouraged the industry to ap,aoach the issue on a generic basis. The Nuclear Energy Institute (NEI) unvolled its plans for the use of the new source term at operating plants at the Regulatory Information Conference in May 1995. NEl, Polestar (EPRl's consultant), and pilct plant (Grand Gulf, Beaver Valley, Browns Ferry, Perry, and Indian Point) representatives met with NRC stalf in June and October 1995 to discuss more detailed plans.

Eronosed Actions: The staff has reviewed the framework document has prepared a Commission paper and decision letter that describes a generic implementation approach. The staff presented the Commission papet 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 Commission in November 1996 (SECY 96 242).

As described in the Commission paper, the current plan is to rebaseline 2 NUllEG-1150 plants; one a PWR and one a BWR. The staff is reassessing the availability of key resources needed to complete rebaselining on en expedited schedule ad has prepared a memorandum to the Commissioners informing them of the status of the project. The staff will also review each pilot plant application and prepare en 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 Assessment: There will be no mandatory backfit of the new source term for operating reactors. The design basis accident analyses for current reactors based on the tid 14844 source term ere still valid. Therefore, non-urgent regulatory action and continued facility operation are justified.

Current Stain: This issue is item I.J on the Chairman's Tracking List. 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 submittels.

The staff met again with NEl and the industry on October 2,1996, to discuss the staff's plan to issue exemptions while pursuing rulemaking, and on April 2,1997, to provide a status report on the staff's actions regarding rebaselining and rulemaking subsequent to the Commission's SRM. The pilot plant applications for 21

Browns Ferry, Perry, Indian Point, and Oyster Creek have been circulated to the task force members to help shape rebaselining, in June 1997, RES circulated an early draf t of the proposed RG that would consider updated source term insights (NUREG 1465) (the RG would be analogous to RG) 1.3 and 1.4 that use the TID 14844 source term). On August 1,1997, D:NRR issued a request to D:RES to initiate a rulemaking that would establish a regulatory framework for operating reactors to use the updated source term insights outlined in NUREG 1465; NRR believed that the rulemaking process can be initiated prior to the completion of rebaselining.

The staff briofed 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 submittels and exemptions are expected four submittels have been received so far (Browns Ferry, Perry, Oyster Creek and Indian Point 2). An application is also expected from Grand Gulf. In addition, the staff and Virginia Power met on November 26,1996, March 25 and June 18,1997, to discuss the rebaselining of Surry; the staff and Entergy mot on August 29,1996, and March 27, 1997, to discuss the rebaselining of Grend Gulf. In a February 12,1997, SRM, the Commission approved the Option 2 approach of SECY 96 242 and a modification to the letter response to NEl.

On February 26,1997, the EDO issued the letter response to NEl. The staff has initiated the rebaselining effort. The staff briefed the NRR Executive Team on the status of the project on July 8,1997, with an emphasis on the scope of activities involved with rebaselining; as a consequence of that briefing, the user need memorandum regarding rulemaking was issued on August 1,1997, i

i and the staff status report to the Commi-31oners was signed on August 29,1997, indicating that the cornpletion of rebaselining will be de'arred.

The next update should include the implementation plan and schedule for rulemaking and the development of regulatory guidance.

NRR Technical Contacts:

R.Emch,PERB, 415 1068 NRR Lead PM:

B. Zaleman, PGEB, 415 3407

References:

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

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

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

July 21,1995, memorandum to the Commission from NRC staff, "Use of NUREG 1465 Source Term at Operating Reactors".

December 22,1995, pilot plant submittal, letter to Document Control Desk from Tennessee Valley Authority, " Brown's Ferry Nuclear Plant (BFN) Units 1,2, and 3 Technical Specifications (TS)

No. 356 and Cost Beneficial Licensing Action (CBLA) 08 Increase in Allowable Main Steam isolation Valve (MSIV) Leakage Rate and Request for Exemption from 10 CFR 50, Appendix J... and 10 CFR 100, Appendix A...".

August 9,1996, memorandum to the Commission from NRC staff, "Use of NUREG 1465 Source Term at Operating Reactors".

November 25,1996, SECY-96 242, "Use of the NUREG 1465 Source Term at Operating Reactors."

22

February 12,1997, Staff Requirements Memorandum to SECY 96 242.

February 26,1997, letter to T. Tipton, NEl, from J. Callan, NRC, responding to the NEl Framework Document.

August 1,1997, memorandum from D:NRR to D:RES to request that rulemaking be initiated for operating reactor use of updated source term insights.

August 29,1997, memorandum to the Commission from NRC staff, "Use of NUREG 1465 Source Term at Operating Reactors."

Summaries of public meetings:

e dated November 10,1994 for public meeting with NEl held on October 6,1994; dated July 26,1995 for public meeting with NEl held on June 1,1995; e

l e

dated November 17,1995 for public meeting with NEl held on October 12,1995; I

e dated February 1,1996 for public meeting with NEl held on Jr. wary 23,1996; dated February 27,1996 for public meeting with Browns Fei.y held on February 7,1996; y

e l

dated September 27,1996 for public meeting with Grand Gulf held on August 29,1996; dated October 11,1996 for public meeting with NEl held on October 2,1996; dated January 24,1997 for public meeting with Surry held on November 26,1996; e

e dated April 24,1997 for public meeting with PWR (Surry) held on March 25,1997; e

dated April 24,1997 for public meeting with BWR (Grand Gulf) held on March 27,1997; dated May 8,1997 for public meeting with NEl held on April 2,1997; e

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

e 23

ENVIRONMENTAL SRP REVISION ACTION PLAN TAC No, M80177 Last Update: 09/09/97 l

GSI: Not Available Lead NRR Division: DRPM i

MILESTONES DATE IT/C) 1 1

1.

Reflect Potential Impacts and integrated Impacts in Options for Resolution a.

Identificatlon of potentialimpacts 03/96C b.

Identification of Integrated impacts 06/96C l

c.

Proposed options for resolution and develop initial 10/96C draft of revised ESRP d.

Staff / contractor meeting to resolve format and j

content of revised ESRP 11/96C 2.

Prepare Final Draft of ESRP Sections for Public Comment

a. Draf t updated ESRP for staff review 01/97C i

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

c. Publish (electronic) for public comment 09/97C 4

i l

3.

Disposition Public Comments 02/98T l

4.

Publish Final NUREG 1555 08/98T 5.

Maintenance of program data Ongoing Bdat Descriotion: The Environmental Standard Review Plan (ESRP) Revision Action Plan deals with the revision to NUREG 0555 to reflect changes in the statutory and regulatory arena, to i

incorporate emerging environmental protection issues (e.g., SAMDA and environmental justice) since originally published in 1979, and to support the review of license renewal applications. The ESRP will take the form of the SRP (including acceptance criteria) and follows the same update

'l criteria outlined under the SRP-UDP project (with the exception of maintaining the MDB at this time). The objective of the tasks outlined in the action plan is to complete the identification of potential Impacts by April 1996 (completed in March 1996), the integrated impacts by June 1996 (completed), and the options for resolution beginning in August 1996 with levelizing across ologies occurring earlier at the options stage rather than later at the draft stage, initialinteractions on j

options stage indicate that, at a minimum, the existing ESRP sections will need restructuring to conform to NUREG-0800 format; contractor is combining resolution options and format restructuring to accelerate schedule. After submittal of the draft by February 1997 for staff and CRGR review, if necessary, the sections will be published for public comment in September 1997.

Disposition of public comments and staff review of the update (NUREG 1555) leads to a publication date of August 1998.

4

' Raoulatorv Assessment: N3R has established the ESRP Update Program for use in the life cycle review of environmental protection issues for nuclear power plants, especially license renewal applications, but also operating reactors, and future reactor site approval applications. The ESRP will reflect current NRC requirements and guidance, consloer other statutory and regulatory requirements (e.g., the National Environmental Policy Act, Presidential Executive Orders), and incorporate the generic environmentalimpact work and plant specific requirements developed 1

j during amending of Part 51 for license renewal reviews.

24 v

.rm nn~

~

+r

-vcwe

--,..w

Current Status: The PNNL/NRC staff workshop on the restructured and revised ESRP was held during November 13 14,1990. Now that the Part 51 rule for license renewalis 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 PNNL in January 1997; followup discussions were held with the contractor through August 1997. The June 1997 draft of the ESRP was forwarded to ACRS for its consideration in light of the current ACRS schedule, ACRS staff indicated that the ACRS will have no objection to publishing the draft ESRP; the ACRS may request a briefin0 during the public comment period. The June dmit 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 draf t in early August and were included in the final draft. The final draft was completed on August 20,1997, and is at the printer; it will carry an August 1997 date with a Federal Register notice of avmilability in mid September. The Federal Register notice of availability is being prepared.

NRR Technleal

Contact:

B. Zalcman, PGEB, 415 3407 I

l C

25

10 CFR 50.59 ACTION PLAN TAC No. M94269 Lac. Update: 09/05/97 id NRR Division: DRPM

.pporting Divisions: all MILESTONES DATE (T/C) i 1.

Action plan approval / copy to Commission (04/10/96)(C) i 2.

Identify work group members 05/24/96(C) 1 Brief C/NRR on issues N/A l

4.

Conduct workshor 06/18/96(C) 5.

Brief D/NRR on proposed positions 07/24/96(C) 6.

Draf t position papers 08/29/96(C) 7.

Obtain regional comments 09/30/96(C) 8.

Policy issues and position paper to Commission with Lessons (02/12/97(C)

Learned Report 9.

Issue document for public comments 05/07/97(C) 10.

Obtain comments 0707/97(C) 11.

Recommendations and rulemaking plan issued to NRC (08/97)(C) management 12.

Commission Paper (09/07/97)(T) 13.

Follow up Actions TBD Descriotion: This action plan defines measures to improve licensee implementation and NRC staff oversight of the 10 CFR 50.59 process.

Historical Backaround: 10 CFR 50.59 was promulgated in 1962 to describe the circumstances under which licensees may make changes to their facility (or to make changes to procedures, or to conc uct tests and experiments) without prior NRC approval when the change does not involve the Technical Specifications or an unreviewed safety question. Licensees are required to submit periodically information related to changes made pursuant to 50.59. The NRC has programs for monitoring licensee processes for implementing 50.59. In a memorandum dated October 27,1995, Chairman Jackson raised a number of questions concerning 50.59 implementation and NRC oversight, and proposed a systematic reconsideration and reevaluation of the process.

The statf developed an action plan to identify actions to be undertaken to improve both the licensee's implementation and the NRC staff's oversight of the 50.59.

- Donosed Actions: In accordance with the action plan, the staff's approach to development of regulatory guidance would proceed in phases. Over the latt several months, the staff has daveloped specific positions (guidance) in particular areas related to 50.59 Implementation and has coasidered the feasibility of implementing such guidance w; thin the existing regulatory framework.

Pubic comments on the position paper (s) will be obtained. The ACRS was asked requested to provide its comments on these positions. At the end of the first phase, the staff will take stock of its progress and make recommendations on issuing guidance, undertaking rulemaking or other 26

l actions. Actions, milestones and schedules for further phases of this ef fort will be developed af ter

}

the results of the first phase are assessed.

Orioinatino Document: April 15,1996 memorandum from the EDO to Chairman Jackson,

Subject:

Action Plan for improvements to 10 CFR 50.59 Implementation and Oversight.

Reoulatorv Assessment: The action plan was developed to identify actions to improve implementation of the 50.59 process. A number of improvements have been implemented, such as directing inspectors conducting all routine inspections to specifically address FSAR compliance, and reviewing spent fuel pool / core offload procedures and practices at all facilities. As stated in the December 15,1995, memorandum, 'The staff concludes that there is currently no indication that implementation of 10 CFR 50.59, as it is carried out today, has led to decreased safety, based on inspection experience. While improvements can be made to achieve a higher degree of uniformity of review, the current process as it is being implemented provides reasonable assurance that plant safety has not been decreased." The above conclusion is confirmed by the additional analysis of inspection experience presented in the staff review document. Therefore, non urgent regulatory action and continued facility operation are justified.

C.unent Status: A revision to the action plan was issued on August 20,1996, which revised the scheduN milestones such that the Commission would have the opportunity to consider the policy issues ass)ciated with 50.59 along with other policy issues from the Millstone lessons learned review.

A Cr. mission 9 aper, SECY 97-035, was sent to the Commission on February 12,1997, that forwart.1 the re suits of the staff's review to the Commission in the paper, the staff identifies areas whose in olorm,1tation would benefit from clarification. The staff proposed to issue regulatory guidance to provide these clarifications, and the paper requested Corr. mission approval to publish the staff paper for public comment. A Commission briefing was conducted on March 10,1997. In a Staff Requirements Memorandum dated April 25,1997, the Commission approved the staff recommendation for a 60-day comment period on the staff's proposed guidance. The Federal Realster notice of availability for comment of draft NUREG-1000 was published on May 7,1997.

The Commission also directed the staff to provide a paper by September 8,1997, that would provide staff recommendations including consideration of the public comments and Commission guidance on SECY 97-036 (Millstone Lessons Learned Part 2 report), and a rulemaking plan for a risk informed approach for 50.59 determinations. This Commission paper is in final concurrence.

(This Commission paper (SECY 97-205) has been issued on Sep 10,1997.)

In an SRM dated May 20,1997, the Commission provided guidance concerning SECY 97-036. In the SRM, the Commission requested a paper in September 1997 that discusses experience with short term actions, any license or design basis issues that remain, and any recommendations for further irnprovements to the regulatory process. The response to these items is included in the above-mentioned Commission paper, in addition, the staff briefed the Chairman on these topics on July 31,1997 and on September 2,1997. Briefings for Commission Technical Assistants were conducted on July 24,1997 and on August 26,1997.

The staff briefed the ACRS on April 2,1997, on SECY 97-035. In a letter dated April 8,1997, the ACRS recommended that the staff positions not be issued for public comment but instead that the NRC and industry continue efforts to revise industry guidance (draft NEl 96-07). The staff met with NEl on April 28,1997, June 11,1997 and July 24,1997, to discuss possible revisions to NEl 96-07.

NRR Technical

Contact:

E. McKenna, PGEB, 415 2189 27

Balalantant October 27,1995 memorandum from Chairman Jackson to EDO November 30,1995 memorandum from Chairman Jackson to EDO December 15,1995 memorandum from EDO to Chairman Jackson i

December 28,1995 memorandum from EDO to Chairman Jackson April 15,1990 memorandum from EDO to Chairman Jackson August 20,1996 memorandum from EDO to Commission t'ebruary 12,1997, SECY 97 035, Proposed Re9ulatory Guidance Related to implementation of 10 CFR 50.59 (Changes, Tests, or Experiments)

April 25,1997, Commission SRM on SECY 97-035.

May 20,1997, Commission SRM on SECY 97 030, 28

INDUSTRY DEREGULATION AND UTILITY RESTRUCTURING ACTION PLAN TAC Nos. M78003 Last Update: 8/31/97 GSl: Not Available Lead NRR Division: DRPM MILESTONES DATE (T/P/C)

Task 1 Develop NRC Policy Statement and SRP 09/97T Draft Policy Statement 05/96C Office Concurrences 06/96C EDO Concurrence 06/96C Commission Paper 07/96C l

Draft SRP 07/96C Publish Draft P.)lir / Statement 09/96C l

Of fice Concurrences on SRP 09/96C EDO Concurrence on SRP 09/96C Commission Paper on SRP 09/96C Publish Draf t SRP 1/97C Public Comment Policy Statement 2/97C Public Comment SRP 03/97C Final Policy Statement 06/97C Office Concurrences 06/97C ACRS 06/97C CRGR N/A EDO Concurrence 06/97C Commission Approval 07/97C Publish Final Policy Statement 08/97C Final SRPs 09/97T Publish Final SRPs 09/97T Task 2 lasue Administrative Letter to Licenseea on Financial Reporting 06/96C Requirements Draf t Administrative Letter 05/96C Office Concurrences 05/96C Commission Information Paper 06/96C lasue Admin Ltr to Licensees w/WTR Letter to CEOs 06/96C Task 3 Develop Non Rulemaking Option for Periodic Reporting 09/97T Requirements as Necessary Determine Necessity for Action 09/96C Draft Option 01/97C Office Concurrence 01/97C EDO Concurrence 09/97T Publish Draft 09/97T 29

Task 4. Update prior NUREG documents on owners and antitrust 02/97C license conditions issue Task Order Contract 05/96C Draft NUREG Updated 09/96C Publish NUREGs 12/96C Task 5 Institutionalize Staff Level Contact with NARUC,SEC,FERC.

ONGOING Develop MOUs as necessary.

Letter to agencies 06/96C Staff level meetings 11/96C Draft MOUs to Commission (as required)

TBD Sign MOUs TBD Task 6 - Develop and implement rulemaking to clarify 10 CFR 50.80 if necessary l

Commission determination of need (Policy paper in preparation) 10/97T Proposed ANPR or rulemaking package TBD l

Office Concurrences TBD 1

ACRS Comments TBD CRGR Concurrence TBD EDO Concurrence -

TBD Commission Approval TBD Publish ANPR or Proposed rule TBD Public Comment TBD Revise Rulemaking Package TBD l

Office Concurrences TBD ACRS Comments TBD CRGR Concurrence TBD EDO Concurrence TBD Commission Approval TBD Publish Final Rule TBD Task 7 Assict Office of Research (RES) on Decommissioning Funding ONGOING Assurance Rule.

Milestones for this task provided by RdS under rulemaking action, ' Decommissioning Costs and Funding Evaluations

• Descriotion; The action plan is intended to address the Commission's concerns regarding the impact of utility deregulation and resulting reorganizations and restructuring on licensee's financial qualifications and their ultimate ability to safely operate and decommission their f acilities.

Historical Backaround in recent years, several restructurings and reorganizations have occurred with the electric utility industry.1 in addition, State public utility commissions (PUCs) have increased pressure for improvements in economic performance of electric utilities they regulate in order to reduce the rates paid by wholesale and retail consumers. The accelerated pace of this restructuring may affect the ability of power reactor licensees to pay for safe plant operations and decommissioning. Specifically, the restructuring may affect the factual underpinnings of the NRC's previous conclusion that power reactor licensees can reliably accumulate adequate funds for operations and decommissioning over the operating lives of their facilities.

30-

Pronosed Actionst Specific actions included in the action plan are: 1) issuing a policy statement delineating NRC's expectations with respect to future financial and anti trust reviews and developing a standard review plan regarding NRC's current financial review requirements: 2) issuing an administrative letter to alllicensees delineating their current responsibilities with respect to getting prior NRC approval for changes that may affect their previous financial qualification determinations or ownership; 3) formulating non-rulemaking periodic reporting requirements,4) updating NUREG documents containing financial information: 5) establishing staf f level contacts with the Securities and Exchange Commission (SEC), the Federal Energy Regulatory Commission (FERC), and the National Association of Utility Regulatory Commissions (NARUC); 6) implementing rulemaking if necessary; and 7) assisting the Office of RES in their decommissioning funding assurance rulemaking.

Current Status: PGEB has developed a final policy statement, administrative letter, and has conducted meetings with FERC and SEC. Staff level contacts with NARUC have been identified and implemented. The administrative letter was issued with a letter to the CEOs of alllicensees on June 21,1996. A Commission Information Paper informed the Commission of our intentions for sending the Admin letter and CEO letter. The final policy statement was published in the Federal Register on August 19,1997.

NRR Technical Contacts:

R. Wood, PGEB, 415 1255 M. Davis, PGEB, 415 1016 1

l 31

EXTENDdD POWER UPRATE ACTION PLAN TAC No. M91571 Last Update:

09/02/97 Lead NRR Division: DRP'N GSI: RI 182 Supporting Division: [gSA MILESTONES DATE (T/C) 1:

Receive GE Topical ELTR1 (Generic Review Methodology).

3/95 C 2:

lasue Staff Position Paper on ELTR1 Meeting with GE/NSP.

4/95 C Identify (Mforences between LTR1 and ELTR1, 8/95 C lasue RAls as appropriate.

9/95 C Incorporate information on foreign experience obtained 10/95 C from SRXB.

Develop power uprate database for all U.S. plants.

10/95 C lasue 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 90 3/96 C and tho second part in July 1990.

7/96 C 4:

Issue Staff SE on GE ELTR2.

Meeting with GE/ Industry.

2/96 C Issue RAls as appropriate.

3/97 C Input to the SE from technical branches.

11/97 T Issue SE.

12/97 T 5:

Receive Lead Plant Application (Monticello).

7/96 C 6:

Issue Staff SE for Lead Plant.

Meeting with Monticello.

10/96 C RAls input from tech branches.

1/97 C Issue RAls as appropriate.

4/97 C Issue additional RAls as appropriate.

11/97 T input to the SE from toch branches.

3/98 T ACRS Presentation 4/98 T Issue Secy Information Paper 5/98 T Issue SE.

6/98 T 7:

Support the ongoing staff effort in developing a Standard Review TBD Procedure for power uprates, incorporate lessons learned from Lead Plant activity.

Descriotion: This action plan describes the strategy 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 submittels for uprates willlikely contalo r

• sts for an optimum power level specific for that plant which is something less than the full

,ercent.

32

e Each technloal branch will review the applicable portions of both the ELTR2 (GE topical report containing generic analyses) and the lead plant application, and will provide input into the staf f's safety 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.

Historleal Backoround: The generic BWR power uprate program was created to provide a consistent means for Individuallicensees to recover additional generating capacity beyond their current licensed limit. In 1990, GE submitted licer' sin 0 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. Since 1990, the staff has reviewed and approved at loost 10 such power uprate requests under this generic BWR power uprate program. As a follow on to this program, GE submitted ELTR1 in March 1995 to propose " extended" power uprates of up to 120 perennt of the originallicensed thermu power.

l Pronosed Actions: Specific actions included in the generic action plan are: (1) review ELTR1 and l

Issue a staff position paper, (2) review ELTR2 and issue a safety evaluation report, (3) review the lead plant application and issue a safety evaluation report, and (4) develop a standard review procedure based on ELTR1, ELTR2, and the lead plant review.

Orlainatina Document: GE Licensing Topleal Report (NEDC 32424), " Generic Guidelines for General Electric Bolling Water Reactor Extended Power Uprate," dated February 1995.

Reaulatorv Assessmient: Not applicable. (A safety assessment is not needed for this action plan because a justification for continued operation of a plant is not required.) This program is on industry initiative that is strictly voluntary.

Current Status:

The GE's response to the staff's RAI was received in July 1997. The Monticello's response to the staff's RAI is expected in early September,1997. In August 1997, Hatch submitted an application for an extended power uprate (8%). A meeting with GE/Monticello/ Hatch to discuss the status of the BWR power uprate program is scheduled for September 10,1997.

NRR Lead PM: T. J. Kim, DRPW, 415 1392 33 I

1

DRY CASK STORAGE ACTION PLAN TAC Nos.: M93927 (load / unload proc)

Last Update: 09/04/97 M94107 Unspection act;vities)

Lead NRR Division: DRPW M94108 (action plan)

GSI: Not Available M96608 (ISFSI support pad) nm.

MILESTONES DATE (T/C)

Im*"

7 l

1. Heavy Loads / Cranes l

develop working group plan 11/95C prure & issue Bulletin 96-02 4/96C Issue Heavy Loads Action Plan 6/97C

- complete Heavy Loads Action Plan 8/98T a.(i) Movement of Casks Prior to Securing Ud

-Issue RAI for Bulletin 96-02 responses 12/96C

- review site specific responses 9/97T

- identify and resolve generic itsue 12/97T

2. Ccsk Loading / Unloading Procedures contact Nuclear Energy Institute about industry efforts 8/95C resolve high priority issues 9/95C form working group 10/95C

- complete working group determination on further issues 4/96C

-issue information notice 7/97C

3. Inspection Guidance

- issue revised ISFSI inspection procedures 2/96C issue MC 2690 2/97C

- issue 10 CFR 72.212 and 72.48 inspection procedures 2/98T

- Ravise MC 2515 Inspection Procedures for ISFSI 12/97T support activities

4. Enforcement Guidance

- establish working group 5/97C

- Issue revised onforcement guidance pertaining 12/98T to ISFSI violations

5. VSC-24 Weld issue

- issue inspection report 4/97C

-issue Confirmatory Action Letters 5/97C

- review responses (a) loading of future casks 6/98T (b) loaded casks 6/98T

- address othr4 cask designs / generic issue 6/98T

6. VECTRA QA Performance

- issue Demand for Information (DFI) 1/97C

- review responces to DFI 8/97C perform verification inspection 12/97T

7. Emergency Planning

}

- issue EP review / inspection guidance 12/97T 34 L

8. ISFSI Support Pad Designs issue design / inspection guidance for ISFSI support pads 12/97T Descriotion: The Plan was originally developed in June 1995 to identify and resolve major issues and problems in the arca of dry cask storage of spent reactor fuelin independent spent fuel storage installations (ISFSis), Specific issues encompassed by the plan include broad technica; areas such as heavy load control and cask loading / unloading, cask nr vendor specific concerns such as the VSC-24 seal weld integrity issue and the performance of VECTRA in the area of quality assurance, and enhancement of inspection and enforcement guidance, 1

Historical Backaround: The number of U.S. nuclear power plant licensees having or considering ISFSis, is increasing significantly. Licensees have encountered a number of problems durlag the fabrication, installation, and licensing of the existing ISFSls and there has been an inconsistent level of performance by involved licensees and cask fabricators with respect to the use of dry cask storage of spent reactor fuel, Because of the anticipated increased industry effort in this area, the staff needed to fully understand the problems that occurred and take appropriate measures to reduce such problems in the future. Therefore, NMSS and NRR developed a plan to resolve major issues and problems The plan was revised in June 1997 to remove some completed items as described in the referenced January 1997 memorandum and, thereby, improve the readability and usability of the document.

Pronosed Actions: Actions included in the plan are: (1) review each general issue and identify the specific problems to be addressed; (2) develop corrective actions for each problem; and (3) implement the corrective actions.

O ioinatino Document: Memorandum from Carl J. Paperiello and William T. Russell to James M.

Taylor, July 28,1995, " Dry Cask Storage Action Plan."

Reaulatorv Assessment: The plan addresses dry storage of fuel that is several years old, Technical issues have been addressed on a site-specific basis for existing facilities. The action plan will improve guidance, enhance communications with industry and the public, and aid future applicants, i

Current Status: As described in the memorandum dated January 30,1997, from C. Paperiello and F. Miraglia to H. Thompson, the following issues included in the original action plan have been completed or closed following a determination that staff action was not required: cask trunnions, hydrostatic testing, pad seismic requirements, cask weeping, safeguards concerns,10 CFR Part 72 reporting requirements, vendor inspections, and communientions, As discussed in the memorandum dated August 25,1997, the current version of the action plan was revised to include only the open items described in the January 1997 memorandum and several new items requiring significant staff attention by both NRR and NMSS, Although retained as an item in this action plan, issues related to the control of heavy loads have also been incorporated into a separate action plan for planning and monitoring of associated staff activities. The item related to unloading procedures for dry storage casks was reopened to reflect the planned issuance of an information notice on this topic that had been deferred from 1996. Information Notice 97-51, " Problems Experienced with Loading and Unloading Spent Fuel Storage and Transportation Casks," was issued on July 11, 1997. Draft revisions of existing inspection procedures and a new procedure related to 10 CFR 72.212 evaluations were issued by NMSS in June 1997. The target schedule for issuance of these procedures and a new inspection procedure for 10 CFR 72.48 evaluations is February 1998. The action plan retains an item for the review of inspection procedures within the MC 2515 program and subsequent development of proposed revisions for ISFSI support activities. A task force has been formed to propose changes to the NRC Enforcement Policy to ensure that ISFSI activities are appropriately addressed. Two items associated with concerns on specific casks / vendors were added to the action plan because of the impact of these probtams on the transfer of spent fuel to 35

ISFSis. The issues involve possible challenges to the integrity of seal welds on VSC 24 casks and concerns related to quality assurance and oversight by VECTRA, vendor for the NUHOMS storage casks. An item was added to the action plan to resolve questions on review responsibilities and staff expectations for licensee emergency plans for ISFSis, in addition, this item will address (for ISFSis) a general question that has arisen regarding the sequired level of NRC review and approval of licensee-developed emergency actien !evels. An item was added regarding NRC expectations for ISFSI support pads in respor.se to questions identified during a regionalinspection. Because site-specific questions related to the inspection have been resolved, the action plan item addresses proposed revisions to inspection procedures and preparation of other review or inspection guidance.

Contacts:

William Reckley, DRPW/NRR, 415-1314 Patricia Eng, SFPO/NMSS, 415-8577

References:

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

" Realignment of Reactor Decommissioning Program."

l l

Memuandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995, l

" Dry Cisk Storage Action Plan,"

Memoroadum from Carl J. Paperiello and William T. Russell to James M. Taylor, January 25,1996,

" Update u the Dry Cask Storage Action Plan."

Memorandum from Carl J. Paperiello and Frank J. Miraglia to Hugh L. Thompson, January 30,1997, " Dry Cask Storage Action Plan Update."

Memorandum from Carl J. Paperiello and Samuel J. Collins to Hugh L. Thompson, August 25, 1997, " Update on the Dry Cask Storage Action Plan."

36

ACCIDENT MANAGEMENT IMPLEMENTATION TAC #:

M91966 - Overall Last Update: 09/03/97 M91641 - BWROG SAMG Review Lead NRR Division: DSSA MILESTONES DATE (T/C) 1.

Review BWROG Severe Accident Management 11/97T Guidance (SAMG) documents 2.

Rev:ew severe accident training materials and BWROG 06/95C prioritization methodologies 3.

Develop Tl for pilot inspections initial draft (for internal use) 11/95C Industry-sponsored A/M demonstrations 12/97T Revised draft (to NEl and public) 02/98T Final Tl 04/98T 4.

Complete pilot inspections and fo' low-up 07/98T 5.

Revise inspection procedures (IP) and hold public workshop 09/98T Draft IP 10/98T i

Public meeting / workshop 12/98T l

Final IP l

6.

Review remaining plants TBD Descriotion: This 6etion plan is intended to guide staff efforts to assess the quality of utility implementation of accident management (A/M), and the manner in which insights from the IPE program have been incorporated into the licensees A/M program. Specific review areas will include: development and implementation of plant specific severe accident management guidelines (SAMG), integration of SAMG with emergency operating procedures and emergency plans, and incorporation of severe accident information into training programs.

Historical Backaround: The issue of A/M and the potential reduction in risk which could result from developing procedures and training operators to manage accidents beyond the design basis was first identified in 1985 (1). A/M was evaluated as Generic issue 116 and subsumed by A/M-related research activities in late 1989. Completion of A/M is a major remaining element of the Integration Plan for Closure of Severe Accident Issues 12]. The development of generic and plant-specific risk insights to support staff inspections utility A/M programs is also identified in the implementation Plan for Probabilistic Risk Assessment [3]. NRC's goals and objectives regarding A/M were established at the inception of this program (4]. Generic A/M strategies were issued in 1990 for utility consideration in the IPE process [5]. The staff has continued to work with industry to define the scope and content of utility A/M programs and these efforts have culminated in industry-developed A/M guidance for utility implementation. Industry has committed to implement an accident management program at each NPP (6]. NRC has accepted the industry commitment and developed tentative plans for staff inspection of utility implementation 17].

Pronosed Action:i: Specific actions included in the A/M action plan are: (1) complete the review of BWROG SAMG documents, (2) conduct site visits to observe how the elements of the formal industry position are being implemented, (3) complete the draf t Temporary Instruction (TI) using the information and perspectives obtained through the site visits, (4) complete pilot inspections and 37

follow-up, and (5) develop an inspection procedure for use at remaining plants and hold a public workshop. Based on feedback from the workshop, the staff will finalize the inspection procedure, and the approach and schedule for evaluating A/M implementation for the remaining plants.

Oriainatino Document: SECY 88-147, Integration Plan for Closure of Severe Accident issues, May 25,1988.

Reautatory Assessme0.1: Accident management programs are being implemented by licensees as part of an initiative to further reduce severe accident risk below its current, and acceptable, level.

Consequently, this is a non-urgent regulatory action and continued facility operation is justified.

Current Status: Severe accident management guideline documents have been submitted by each of the PWR owners groups, and reviewed by the staff [8). The BWROG submitted Rev. O of the Emergency Procedure and Severe Accident Guidelines (EP/ SAG) and associated technical basis documents to NRC for information on August 29, 1996[9]. The staff and Oak Ridge National Laboratory have completed a high level review of the EP/ SAG documents. Areas where additional information and discussion with the BWROG is considered necessary were identified in an April 2, 1997 letter to the owners group [10). A BWROG submittal describing a timeline for operator actions that would oe taken in accordance with the EP/ SAG for a limited number of sequences was received in May 1997(11). However, the BWROG has not yet responded to the April 2,1997 staff letter. A meeting to discuss specific questions / concerns regarding the BWROG products will be delayed until the submittalis received and the BWROG is prepared to address staff concerns.

Licensee target dates for completing A/M implementation have bean submitted to NRC Licensees for 7 sites have revised (delayed) their original completion dates citing higher priority issues and resource limitations, implementation will be completed at approximately 9 sites within the next two months, and an additional 18 sites by late-1997. Implementation at the balance of sites (43),

includin0 the majority of the BWR sites, will be completed within the latter half of 1998.

A draft Tl for use in the pilot inspections nas been completed. Comments on the draft Ti have been received from the NRC Region offices. The staff met with industry on February 22,1996 and ACRS on March 1,1996 to discuss plans for inspecting utility implementation of the formal industry position on severe accident mar.agement and major elements of the draft Tl These plans included staff visits to approximately 2 to 4 sites for the purpose of obtaining an early understanding of how the various elements of the formalindustry position are being implemented.

The information and perspectives obtained through these visits as well as comments from th.

Region offices would be used to update the draft Tl. The draft Tl would be revised and made available to NEl and the public af ter the information-gathering visits.

A meeting with NEl to discuss the scope and schedules of the information gathering visits was held on December 19,1996. At that time, NEl proposed to take the lead in organizing " demonstrations" of completed A/M implementation at four to six plants. These demonstrations would be in lieu of the information gathering visits and follow on pilot inspections envisioned by the staff, and would occur in the June / July 1997 timeframe. NEl also informed the staff of an industry sponsored workshop concerning severe accident management implementation planned for March 11-13, 1997, and proposed that NRC staff attend in order to better understand implementation approach and status. In a follow-up meeting with NEl on January 24,1997, the staff indicated that attendance at the A/M workshop together with participation in the A/M demonstrations should serve the role of the information gathering visits, but that the staff is not in a position at this time to alter the plans outlined in SECY-96-088 concerning the need for pilot inspections and the nature of the inspections that would be performed at the balance of plants in the longer term. This aspect of the program will be reassessed and refocussed after the A/M demonstrations.

38 I

NRR staff attended the NEl sponsored workshop on accident management implementation on March 11 13,1997. The workshop provided an opportunity to better understand plant specific implementation approaches and issues, and the major elements of implementation.

Two A/M implementation demonstration visits have been completed to date -- the first at Comanche Peak on May 29 30,1997, and the second at North Anna on July 24 25,1997. The visits provided insights into: (1) the licensee's implementation / evaluation process, (2) how well the draft Tl on A/M functions to provide an inspection process, and where changes to the guidance may be needed, and (3) whether the industry-sponsored demonstrations may justify modification of the planned inspection process. Staff is currently awaiting confirmation from NEl regarding the l

schedule and locations of additional A/M demonstrations. Such demonstrations would preferably involve CE, B&W, or GE plants since implementation approaches for these NSSS designs may differ considerably due to the differences in implementation guidance and materials provided by each owners group. These demonstrations are not expected to occur untillate 1997, approximately 6 months later than originally planned. Milestone dates have been updated based on the assumption that the A/M demonstrations will be completed by the end of 1997.

References:

1.

Memorandum from F. Rowsome to W. Minners, "A New Generic Safety issue:

Accident Management," April 16,1985 2.

SECY-88147, Integration Plan for Closure of Severe Accident issues l

l 3.

SECY-95-079, Implementation Plan for Probabilistic Risk Assessment i

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 NRR Technical

Contact:

R. Palla, SCSB, 4151095 NRR Lead PM:

Ramin Assa, DRPW, 415 1391 39

CORE PERFORMANCE ACTION PLAN

-t TAC Nos. M91257 - DSSA Last Update: 08/19/97' M91602 - DISP Lead NRR Division: DSSA GSl: LI 179 Supporting Division: DISP MILESTONES DATE (T/P/C)

Task 1 -

Inspection of Nuclear Fuel Vendors (DISP) ongoing' Sieraens Power Corporation [PWR AIT followup) 06/94C ABB/ Combustion Engineering (PWR reloads) 11/94C To:edyne-Wah Chang (TWC) 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 04/97C issues)'

GE (new issues and followup)*

ongoing l

ABB/CE IBWR) (WNP-2 transition core)*

10/97T Task 2 -

Inspection of Licensee Reload Analyses (DSSA) ongoing' RI - 3 licensees: PSE&G 10/97T PP&L 09/97T tbd 06/98T Ril 2 licensees: CP&L 11/97T TVA 02/98T Rill - 3 licensees: Comed 12/97T Detroit Edison 08/97P(ongoing) tbd 04/98T RIV - 2 licensees: WNP2 06/97P(ongoing)

Entergy 03/98T Task 3 -

Core Performance Data Gathering / Evaluation (DSSA) 12/97T.

Regions - Morning Reports & Event Notification ongoing' Other Data Acquisition and Collation ongoing PNNL - Core Performance Evaluation Analysis (CY95) 12/97T Task 4 -

Participation of Regions in Action Plan (DSSA) ongoing Identification of Vendor issues Feedback from Licensee Inspections Counterparts Meetings (Rl RIV)

Task 5 -

Evaluate Inspection Guidance (DSSA/ DISP) 12/97T Evaluate Results of Licensee Inspections incorporate Feedback from Region Inspectors Draft Guidance for Resident and Region Inspectors issue Inspection Criterlynd Action Plan Update 40 i

Task 6 -

Evcluate Licensee / Vendor Lead Test Programs for Identification 12/97T*

of t', ore Performance Problems (DSSA/ DISP)

Task 7 -

Workshop on Core Performance issues (TAC No. M95674)

Identify issues 07/96C Conduct workshop 10/96C Followup on Comments and Questions (RIC session) 04/97C

• lasue Driven Descriotion: 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 Backoround: The action plan addresses the review of fuel f abrication, core design, and reload analysis issues that were discussed during 1994 and 1996 briefings given to the Executive Director for Operations. The briefings presented by the Reactor Systems Branch (SRXB). Division of Systema Safety and Analysis (DSSA), covered generic fuel and core performance issues and related evaluations of fuel failures. The Special Inspection Branch (PSIB), Division of Inspection and Support Programs (DISP), supported the briefings. As a result of 'hese briefings, the Office of Nuclear Reactor Regulation (NRR) was requested to expand the action plan to monitor and improve core performance in operating reactors to include focus on licensee activities and the licensee / vendor interf aces.

Pronosed 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) evcluate 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 regionalinspectors, and other appropriate regulatory actions. In addition, as a result of recent generic concerns, including the failure of control rods to fully insert, the action plan is being expanded to review the adequacy of vendor lead testing programs for new fuel designs (Task 6); and to conduct a workshop on core performance issues (Task 7) in the f all of 1996. The status of core performance inspection evaluations and emerging issues was covered at the recent 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 will include guidance for resident inspectors and regional staff. The ongoing activities to capture and address early warning of emerging issues will continue into FY97, and the action plan will reflect the planned inspection of 10 licensee / plants,5 vendor LTA program inspections, and four anticipated event-reactive inspections.

DISP - The action plan currently identifies 8 completed and two planned vendor inspections that shall be performed by multi-disciplined inspection teams led by the Special Inspection Branch (PSIB) with contracted technical assistance. These it'spections are currently scheduled to be completed in 1997. In addition, DISP will support the FY97 vendor LTA and licensee inspections, as required.

41

1 Orioinatino 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 participation.

Reculatorv Assessment: Core design is a fundamental component of plant safety because m9intaining fuelintegrity is the first principal safety barrier (i.e., fuel cladding, reactor coolant sys*em boundary, or the containment) against serious radioactive releases. Likewise, the safety analyses must be properly performed in order to verify, in conjunction with startup tests and normal plant parameter monitoring, that the core reload design is adequate and 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.

Current Status:

DSSA -The data acquired from the ongoing vendor inspections are being evaluated for generic impact and identification of emerging issues. The issue-driven inspections at GE and Siemens, were supported by SRXB/DSSA staff and contract specialists in reload design. Interaction with the regions is ongoing to participate in region-led licensee inspections. SRXB has participated in two Region I and one Region 11 inspector counterparts meetings. DSSA is re-evaluating the action plan to better integrate and prioritize its activities, consistent with the available FY97 TA funding.

Options and recommendations for management review are being prepared to support new emphasis on licensee inspection.

' DISP - The remaining issue driven inspections include ABB Combustion Engineering's supply of a l

l BWR transition core reload for WNP-2 (unscheduled), and a comprehensive (4 team weeks) follow-up inspection of Siemens Power Corporation issues, which began 2/10/97, and ended on 4/4/97.

NRR Technical contacts:

E. Kendrick, SRXB, 415 2891 S. Matthews, PSIB, 415-3191

• ime spent on-site at vendor inspections (Task 1) is allocated to appropriate fuel vendor docket #

t 42

ENVIRONMENTAL QUALIFICATION TASK ACTION PLAN TAC No. M85648 Last Update: 09/02/97GSI: 168 Lead NRR Division: DSSA MILESTONES DATE (T/C) 1.

Inform Commission 05/93C 2.

Meet With Industry Ongoing 3.

Programmatic Review 10/97T 4.

Risk Assessment 10/97T 5.

Data Collection and Analysis 4/96C 6.

Review and Evaluation of the Status 10/97T 7.

Technical Issues 10/98T 8.

Options for Resolution TBD 9.

Implementation TBD l

Descrintion: This action plan will evaluate environmental qualification (EO) issues, including operating experience, testing methodology, and adequacy of current rule and guidance for operating reactors, it will resolve EQ issues for aging operating reactors and license renewal.

Historical Backaround: A review of environmental qualification requirements for license renewal and f ailures of qualified cables during research tests led to the development of the EQ Task Action Plan (TAP), which was issued in July 1993. The EQ TAP was developed to address: (1) staff concerns regarding the differences in EQ requirements for older and newer plants: (2) concerns raised by some research tests which indicate that qualification of some electric cables may have been non-conservative; and (3) concerns that programmatic problems identified in the staff Fire Protection Reassessment Report might also exist in the NRC EQ Program.

Prooosed Actions: The EQ TAP includes meetings with industry, a program review of EQ, data collection and analysis, a risk assessment, and research on aging and condition monitoring. Annual Commission papers are written to update the status of the EQ TAP. The staff will develop options for resolving EQ concerns, which may include issuing a generic letter, changing the rule, or documenting the acceptability of the current EQ rule and standards. The basis for the appropriate regulatory action will be oocumented.

Oriainatino oocumer June 28,1993, memorandum from Samuel J. Chilk to James M. Taylor (SECY 93-049); May 27,1993, letter to the Commission from J. Taylor on Environmental Qualification of Electric Equipment.

Reaulatorv Assessment: Depending on the application, failure of these cables during or following design-basis events could affect the performance of safety functions in nuclear power plants.

There is no immediate safety issue because of the degree of conservatism already included in the EQ qualification test margins.

43

e Current Stahul: The draf t reports on the programmatic review and risk issues regarding EQ are currently under management review (Milestones 3 and 4).

BNL is continuing with the cable testing program, which includes investigating condition monitoring methodologies (Milestone 7). The cable test program includes thermal aging, radiation aging and exposure of cable samples to LOCA environments.

j The first set of cable testing (XLPE) has been completed. Preliminary draft reports from the testing contractor are currently being reviewed by BNL. BNL is scheduled to provide the staff a report on the first set of results in October 1997. This report willinclude an analysis of the test anomalies exhibited in the first test series. The second set of test cables (EPR) is scheduled to be aged in the i

first quarter of FY 98 and LOCA tested in the second quarter of FY98. A third set of cables is scheduled to be aged in the second quarter of FY98 and LOCA tested in the third quarter of FY98.

Overall results from the test program are expected in fiscal years 1998 and 1999.

Contacts:

NRR Technical Cor laet:

G. Hubbard, SPLB, 415-2870 RES

Contact:

S. Aggarwal, EMEB, 415-5849 l

NRR Lead PM:

L. Olshan, DRPE, 415-3018

References:

Letter to the Commission from J. Taylor on Environmental Qualification of Electric Equipment dated May 27,1993 (Accession No. 9308180153).

Staff requirements memorandum (SECY 93-049) dated June 28,1993 (Accession No. 9409010107).

Task Action Plan for Environmental Qualification and updates, July 1,1933, April 8,1994, November 16,1994, June 27,1995, August 22,1996, and November 15,1996.

RES Program Plan for Environmental Qualification, July 7,1994 (Accession No. 9407250066) 44

FIRE PROTECTION TASK ACTION PLAN TAC Nos.

M86652, M82809, M84592, Last Update: 09/02/97 M85142, and M89509 Lead NRR Division: DSSA GSI: LI 181 MILESTONES DATE (T/C) 1.

Annual Commission status report Last: 05 '30/97C Next: 05/30/98 f 2.

Recommendations for 04/98T action (Part 1) 3.

Mecommendations for 10/96C yture study (Part II) 4.

Confirmation issues 10/96C

,,8 art Ill) 5.

Other issues (Part IV) 08/95C Descriotion: The Fire Protection Task Action Plan (FP-TAP) is used to track and manage l

implementation of the recommendations made in the " Report on the Reassessment of the NRC Fire l

Protection Program," of February 27,1993.

l Historical Backaround: In February 1993, the Office of Nuclear Reactor Regulation (NRR) completed a reassessment of the reactor fire protection review and inspection programs in response to programmatic concerns raised during the review of Thermo-Lag fire barriers. The results of the reassessment were documented in the " Report on the Reassessment of the NRC Fire Protection Program," of February 27,1993. The staff prepared the FP TAP to implement the recommendations made as a result of the reassessment report.

(

Pronosed Actions: The FP-TAP tracks the implementation of a wide range of technical and programmatic fire protection issues, it includes recommendations for action (Part I),

recommendations for further study (Part 11), confirmation issues lPart lil), and lessons learned (Part IV). The staff is implementing the recommendations, in priority order, as resources allow.

The staff focus is now on irtplementing its plan for future direction of the NRC fire protection pro 0 ram with emphasis on the fire protection functional intpection (FPFI) program.

Oriainatina Document: " Report on the Reassessment of the NRC Fire Protection Program,"

February 27,1993.

Beautatorv Assessment: Each operating reactor has an NRC-approved fire protection plan that, if properly implemented and maintained, satisfies 10 CFR 50.48, " Fire protection," and General Design Criterion 3, " Fire protection," Therefore, each plant has an adequate level of fire safety and the individual action plan items are receiving appropriato priority.

Current Status: The staff issued a semiannual report to the Commission on the status of the FP-TAP on May 30,1997. The next status report is due to the Commission on May 30,1998.

The staff completed additional smalbscale fire tests of fire barrier materials other than Thermo-Lag at NIST. The test results were provided by NIST in its Report of Test FR 4008, " Pilot-Scale Firo-Endurance Tests of Fire-Barrier Panels ard Panel / Blanket Combinations," dated August 20,1996.

The staff's review of the Report of Test FR 4008 and fire barrier materials other than Thermo Lag is 45

ongoing. Due to competing higher priority tasks, the staff changed the completion schedule for this task from September 1997 to April 1998 The Plant Systems Branch (SPLB) continued to work with Probabilistic Risk Assessment (PRA)

Branch staff and Brookhaven National Laboratory (BNL), its technical assistance contractor, to evaluate the risk associated with the post-fire safe shutdown methodology that imposes a self-induced station blackout. The stafI had planned to apply the PRA model for assessing the risk significance of the self-induced station blackout methodology to two plant specific cases. Due to limited staff resources, the staff now plans to address the insights gained from the BNL evaluation during the performance of the FPFI program.

The staff is working on an issue recommended for further study regarding fire barrier reliability, under Generic Safety Issue (GSI) 149, " Adequacy of Fire Barriers." The staff and BNL have performed scoping analyses, using f ault trees and event trees, to assess the effectiveness of a degraded fire barrier in mitigating the consequences of a fully developed fire in a plant area that is important to post fire safe shutdown. The staff and BNL discussed the preliminary results of these two studies and future plans with the Advisory Committee on Reactor Safeguards (ACRS) on February 29,1996. By letter of March 15,1996, the ACRS submitted its comments to the Commission. The staff responded to the ACRS by letter of April 25,1996. The staff assessed the re immendations made by the ACRS and redefined the scope of this work. The staff is evaluating this task in the context of fire protectior, program defense-in-depth principles with a focus on fire initiation, its uncertainties, limitations, and data assumptions. The insights gained will be f ac.ored into the framework of future rulemaking.

The completion of Part ll and Part lll of the FP-TAP is documented in a memorandum of l

October 31,1996, from J. Taylor to the Commission.

in an SRM dated February 7,1997 the Commission agreed with the Fire Protection Functional inspection (FPFI) program described in SECY-96 267. The staff issued the first draf t of the FPFI procedures and conducted the first pilot inspection at River Bend in June 1997. The pilot inspection at Clinton, scheduled for August 1997, was cancelled due to an emergent assessment by a Special Evaluation Team (SET). The three remaining pilot inspections are scheduled to be performed at Susquehanna in October 1997, at St. Lucie in March 1998, and at Prairie Island in May 1998).

The staff will provide the Commission with a post-pilot inspection program report describing inspection results and discussing strategies which would expand the benefits of the pilot inspections to alllicensees (e.g. licensee self assessments with followup NRC reviews). Post-pilot inspection program activities willinclude a public workshop to discuss inspection results and request comments.

The development of a staff fire protection training program will remain on hold until the FPFI program is implemented.

Contact:

D. Oudinot, DSSA, 301-415-3731

References:

" Report on the Reassessment of the NRC Fire Protection Program," of February 27,1993.

• SECY 95-034, " Status of Recommendations Resulting From the Reassessment of the NRC Fire Protection Program," February 13,1995.

• Memorandum of October 31,1996, from J. M. Taylor, EDO, to the Commission, " Semiannual Report on the Status of the Thermo-Lag Action Plan and Fire Protection Task Action Plan."

• SECY-96-267, " Fire Protection Functional Inspection Program," Decemoer 24,1996.

46

7 4

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

TAC Nos. M93821: Action Plan Last Update: 09/04/97 M91955: DSC generic review Lead NRR Division: DSSA M95546: Generic review of NRCB 96-02 ACTION DATE (T/C) 1.

Review, summarize and issue existing NRC guidance on heavy load control.

Review NUREG 0554, NUREG-0612, GL 80-113, GL 8107, 2/96C GL 8511, and other supporting documents.

- Develop summary of guidance.

2/96C 2.

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

- Generic letter 8511 and NUREG 0612, 2/96C Single Fallure-Proof Crane (reliability).

TBD Spent fuel cask drop accident prior to securing the lid.

2/96C Risk significance of multiplc failures within safe load path.

TBD

- Other TBD 3.

Review licensee implementation of heavy load control, including applicable correspondence from a sample of licensees and site visits.

01/98T 4.

Review NRC audit / inspection procedures, practices, inspection reports, enforcement actions, and experience. Document the findings and determine whether additional inspection procedures are needed.

5/96C Conduct review.

02/98T

- Document Findings 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.

- Issuo bulletin on load movement during operations.

4/96C

- Issue guidance on load movement during shutdown (if needed).

TBD

- Issue guidance on safe load path determination and associated TBD risks (if needed).

6.

Draft staff guidance and disseminate to appropriate management 10/97T (SPLB, Region I, NRR) and obtain/ resolve any comments. (propose form of guidance).

7.

If an inspection procedure (or procedures) is planned, issue the 11/97T inspection procedure (s) in draft.

8.

Obtain feedback (meeting, FRN, or other means) concerning the staff 2/98T position from industry representatives and resolve any discrepancies with the industry position.

I 47

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

ACTION DATE (T/C) 9.

Develop final version of guidance and obtain management 5/98T concurrence.

10.

Issue finalinspection procedures.

12/97T 11.

Issue final guidance.

8/98T Descriotion: The Heavy Load Control (HLCi 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 powei plants. (See the Enclosure for a detailed description of the scope of the actions under the action plan),

lii:ligrical Backoround: Recent increases in licensees' activities involving the transfer of spent fuel to independent spent fuel storage installations (ISFSis) have given rise to a number of concerns with NRC's regulatory program ior the control and handling of heavy loads, and with the licensees' programs for complying with the requirements in NRC's existing guidance. For example, there are concerns regarding what is required for the.novement of heavy loads while the plant is operating.

Decause 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 probbms in the future.

This plan was identified as a near-term iscue under the dry cask storage action plan, and was recently revised to better reflect the scope and magnitude of the task.

Prooosed Actions: Actions included in the plan are: (1) understand the current reguietory framework and inform the staff; (2) review the general issues and identify specific problems to be addressed; (3) develop corrective actions to resolve the problems; and (4) implement the corrective actions. Specific corrective actions may include the issuance of guidance to licensees alerting them to the potential problems and requesting that corrective measures be taken to preclude accidents.

Oriainatino Document: Memorandum from Carl J.Paperiello and William T. Russell to James M.

Taylor, July 28,1995, " Dry Cask Storage Action Plan."

Reaulatorv Assessment: Severallicensees have either developed or are implementing plans to move heavy loads in v9rious areas of nuclear power plants (i.e. offloading spent fuel via dry storage and/or transfer casks for transfer to ISFSis) over safe shutdown equipment, and over spent fuel during plant operation. Questions have been raised regarding the adequacy of NRC's guidance and the licensees' methods of precluding an accident, and in the event of an accident, mitigating the severity of the consequences. An urgent NRC bulletin (NRCB) 96-02, " Movement of Heavy Loads Over Spent Fuel, Over Fuelin the Reactor Core, or Over Safety Related Equipment," has been issued to alert licensees to the concerns. As a result of the bulletin, several licensees have j

undertaken efforts to assess their plans, capabilities, and licensing basis for heavy loads. The action plan is intended to improve NRC guidance, address site specific issues, and aid licensees in their future plans to move heavy loads.

Current Status: All the licensees have responded to NRC Bulletin 96-02. General information about the responses is being assessed, and RAls will be developed and issued as appropriate. As a follow-up to the bullotin, the staff issued an RAI regarding the potent 61 accident scenario involving a tipped over spent fuel cask prior to the cask tid being secured and during transfer of the cask between the spent fuel cask pit area and the spent fuel pool. Responses to the RAI have been received from all licensees with near term plans to move cusks. The staff has reviewed responses from ANO, Oconee, Palisades, Harris, and Surry. Responses from Robinson and WNP-2 are currently being reviewed. The RAI was sent only to licensees without singlo-failure-proof cranes t

48

who plan to move casks within the next two years.- As other licensees approach their future -

schedules for cask movements, they too will have to address the issues raised by the RAl. The staff continues to interact with licensees un a plant specific basis.

The staff is currently working with RES and SFPO to get RES to perform a heavy loads PRA as part of its efforts to develop a dry cask systems PRA. The staff will visit about 2 to 4 sites in 1997 for the purpose of obtaining an understanding how the various elements of the licensees' programs are being implemented. Information and perspectives gained through these visits, as well as input from the Regions, will be used to help determine and develop further guidance.

NRR Contacts: Brian E. Thomas, DSSA, 415 1210 Phillip M. Ray, ADPR, 415 2972 Joesep5 E. Carrasco, RGN-l/DRS, (010) 337 5306 Rafsrances:

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

" Realignment of Reactor Decommissioning Program."

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

49 i

HIGH BURNUP FUEL ACTION PLAN TAC NO.

M91256 Last update: 8/19/97 Lead NRR Division: DSSA GSI: 170 Supporting office: RES MILESTONES DATE (T/C) 1.

Issue user need letter to RES 10/93C 2.

Contracts issued by RES 03/94C 3.

Schedule and coordinate meetings with foreign experimenters and 09/95C regulatory authorities 4.

Issue Information Notice (IN 94 64) Announcing new RIA data 08/94C 5.

Present high burnup data at water reactor safety meeting 10/94C 6.

Schedule / coordinate industry meetings to discuss actions 10/94C 7.

Determine need for further generic communications 11/94C 8.

Issue letter to vendors 11/94C 9.

Issue IN 94 64, Suppl.1 Providing Data and Vendor Letter 03/95C 10.

RES Update NUREG-0933 on Generic issue

• and Plan of Action 03/95C' 01/96C 11.

Review industry (NEI) Response 09/95C 12.

Assess effects on design basis accidents of reduced failure 09/95C threshold for high burnup fuel 13.

Committee on the safety of nuclear insteilcions specialists 09/95C l

meeting on the transient behavior of high burnup fuel l

14.

CNRA (OECD) Committee on nuclear regulatory activities and 11/95C CSNI annual meetings.

15.

Issue ltr to NEl assessing industry actions (vendor /EPRI response 8/97C to IN) 16.

Water reactor safety information meetings (high burnup session) 10/95C core performance issues workshop 10/96C 17.

RES briefs ACRS and completes response to NRR user need letters 04/96C 9/97T 18.

Complete review of available fuel transient data relevant to design 4/97C basis event e

19.

Develop interim acceptance enteria (e.g., Based on cladding oxide) 4/97C 20.

Meeting with NEl and industry on interim criter;a 9/97T 21.

Complete agency program plan on high burnup fuel 10/97T 50

4 MILESTONES DATE (T/C) 22.

Initiate revision to SRP 12/97T 23.

Establish schedule for LOCA resolution and final assessment 12/97T Jtermine need for further regulatory action

'hES HAS PRIORITIZED AS GENERIC ISSUE #170 NUREG 0933.

Descriotion: The action plan covers assessment of fuel performance for high burnup fuel and evaluation of the adequacy of SRP licensing acceptance criteria.

Historical Backaround: Recent experimental data on performance of high burnup (> 50 GWD/MTU) under reactivity insertion conditioris became available in mid 1993. The unexpectedly low energy deposition (30 CAL /GM) to initiation of fuel failure in the first test rod (at 62 GWD/MTU) led to a re-evaluation of the licensing basis essumptions in the SRP. As a result, the office of nuclear reactor regulation (NRR) was requested to prepare an action plar, in coordination with the Office of Nuclear Regulatory Research (RES).

Prooosed actions: After a preliminary safety assessment was performed, an action plan was i

developed, to include a user need letter to RES and the issuance of contracts to assess all aspects of the high burnup fuelissue. Concurrently, meetings would be scheduled with the non domestic experimenters and regulatory authorities to discuss the experimental data and to assess potential consequences and regulatory actions. Meetings with industry would be scheduled to discuss their planned actions and to solicit cooperation with the safety evaluations. Based on a complete review of all available fuel transient data, relevant to design basis events, NRR/RES would define acceptance criteria, establish a schedule for final assessment, and state need for further regulatory action.

Oriainatino Documents: Commission Memorandum from James M, Taylor (EDO), " Reactivity Transients and High Burnup Fuel," dated September 13,1994, including IN 94-64, ' Reactivity insertion Transient and Accident Limits for High Burnup Fuel,' dated August 31,1994. Commission Memorandum from James M. Taylor, " Reactivity Transients and Fuel Damage Criteria for High Burnup Fuel," doted November 9,1994, including an NRR safety assessment and the joint NRR/RES action plan.

Reaulatorv Assessment: There is no immediate safety issue, because of the low to medium burnup in currently operating cores. Since the fuel failure threshold declines with increasing burnup, the licensing basis design acceptance criteria may need to be redefined as a function of burnup. The end nroduct of the plan will determine the need for regulatory action and will establish and define the need for further action on extended burnup cycles and high burnup fuelissues.

Current Status: An ACRS Subcommittee Meeting on the status of RES contractor programs was held in 4/96. An NElletter summarizing the industry position was received in April, and the EPRI report supporting this position was sent by NEl on 9/20/96.

A commission paper on the status of the high burnup issue and planned actions was prepared by NRR, has been reviewed by RES, and was issued on November 25,1996. A Commission briefing was completed on March 25,1997.

A letter with an enclosure was issued on 8/12/97 to respond to the NEl and industry report on high

'ournup fuel.

NRR Technical Contacts:

Laurence Phillips, NRR/DSSA/SRXB, 415-3232 Shih-Liang Wu, NRR/DSSA/SRXB, 415-3284 Edward Kendrick, NRR/ OSSA /SRXB, 415-2891 RES

Contact:

Ralph Meyer, RES/ DST /RPSB, 415-6789 51 1

WOLF CREEK DRAINDOWN EVENT: ACTION PLAN TAC Nos. M92635 Last Update: 9/3/97 Lead NRR Division:DSSA MILESTONES DATE (T/C)

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

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

3) Complete Draft Tl/ Issue to the Regions for Comments 12/97(T)

4) CRGR Concurred GL for 1ST Time / 2ND Time (after Public Comments) 9/96(C) /11/97(T)

GL issued 12/97(T)

5) Receive Regional Comments on Tl 2/98(T)

6) Complete Evaluation of the Responses to the Generic Letter 05/98(T) 7)lssuo Tl 05/98(T)

8) Complete Inspections (As necessary)

OP/98(T)

Descriotion: The objective of this action plan is to collect and evaluate information from the I?censees regarding plant system configurations and vulnerabilities to draindown events. A 10 CFR 50.54(f) letter will be used to gather the information, and the licensees are expected to take corrective actions, as appropriate, in accordance with the requirements of App B to 10CFR50 to ensure compliance with the intent of GDCs 34 and 35.

Historical Backaround: On September 17,1994, the Wolf Creek plant expenenced loss of reactor coolant system (RCS) inventory, while transitioning to a rclueling shutdown. The event occurred when operato;s cycled a valve in the train A side of the RHR system cross-connect line following maintenance on the valve, while at the same time establishing a flow path from the RHR system, train B, to the refueling water storage tank for reborat.ag train B. The failure of the reactor operating staf f to adequately control two incompatible activities resulted in transferring 9200 gallons of hot RCS water to the RWST in 66 seconds.

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

This event represents a PWR vulnerability which was not previously recognized.

Prooosed Actions: Specific actions of this generic action plan are: (1) issue IN 95-03 (issued January 18,1995) and supolement to IN 95-03 (issued March 25,1996),(2) Request a!t PWR licensees, via an information gathering (10 Cl R 50.54(f)) Generic Letter (GL), to provide information on draindown vulnerabilities aM the measures they implemented to diminish the probability of a draindown. The staff considers the proposed action as a conditional compliance backfit issue.

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

52

Raoulatorv Assemament: The staff performed an evaluation of the probability for event initiation and of the conditional core damage probability. Tha value of this probability for core damage along with licensee awareness for this scenario makes the risk for continued PWR operation acceptably -

small, i

Current Status: Information Notice IN 95-03, and its Supplement have been issued. CRGR concurred the proposed GL in 9/96. Directed by an SRM, the GL was published in the Federal Register in 2/97 for public comments.

NRR Technical

Contact:

M. M. Razzaque, SRXB, 415 2882 NRR Lead PM:

J. C. Stone, DRPW, 415 3063

References:

• AEOD/S95 01, " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994"

• IN 95-03, issued January 18,1995.

• Supplement to IN 95-03, issued March 25,1996.

.j I

l 53

bium p

GENERIC COMMUNICATION AND COMPLIANCE ACTIVITIES l

i

c.

Page No.

1

09/11/97 PUBLIC' SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT

. Open Generic. Communication and Compliance Activities Sorted by Lead Technical Division and Branch.

TAC-Type Contact

-TR Comp LA Comp Title-Description-

" LTD = Associate Director for Projects

• LTB =

M99000 IN WFBurton

--/--/--

12/31/97 T.-INi Dispositioning of Alerts licensees to )roblems the NRC Technical Specifications staff has noted in t1e dispositioning-

. Discovered Not to be of-technical specifications found to Sufficient to Assure Plant specify parameter values or required Safety actions that even if complied with, would not assure safety..

• LTD = Division of Engineering

• LTB = Civil Engineering and Geosciences Branch M94293 GL JWShapaker 06/01/98 9/19/97 T GL: NRC Preliminary Findings Develop a GL to advise licensees that Related To The Use Of Reduced the use of reduced seismic criteria for' Seismic Criteria For Temporary temporary conditions may involve Conditions.

unreviewed. safety questions and staff review may be neaded.

M95688 LT TAGreene 09/30/97.

4/30/97 L Study of The Adequacy of After completion of contract JCN Enveloped Response Spectrum J-2354, an-IN might be issued to Method caution operating plant licensees that under certain conditions ERS analysys method may not provide adequate estimates of seismic response of piping -

systems.

l l

Page No.

2

'09/11/97

'PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT 0 pen Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC-Type Contact TR Comp LA Comp Title Description M97920 GL JWShapaker 10/31/97 9/20 97 T GL: Seismic Capability of Informs addressees about reduced;-

Thermal-Lag Panels seismic capability of Thermo-Lag panels in high temperature areas of plants.

and need'for corrective actions.

M97981 GL' JWShapaker

09/30/97 9/26/97 T GL

Monitoring of Containment Informs addressees of need to' review.

Structure Settlement due to subfoundation designs and.' as Degradation of Porous Concrete appropriate. describe plans for Sub-foundations.

foundation settlement monitoring.

M99304 LT TAGreene

-09/30/98 10/30/98 T HOLTEC Part 21 Computer Code To review information submitted by Issue Holtec International concerning ANSYS compute code.

• LTB = Electrical Engineering Branch M95215 LT DLSkeen

--/--/--

9/1/97:L Charging / Discharging of-Study and interact with the industry Safety-Related AT&T Round Cell group on the AT&T round cell battery Batteries degradation problems.

M96616 GL JWShapaker 11/11/11 9/26/97 T GL: Medium-Voltage Circuit GL to address continued breaker Breaker Failures problems because of' refurbishment

. practice.s. licensee maintenance, and inadequate review of industry operating experience.

Page No.

3 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR *S BIMONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Disision and Branch TAC Type Contact TR Comp LA Comp Title Description M97147 LT DLSkeen

--/--/--

8/31/97 L LT: Failure of Westinghouse Evaluate failure of breakers due to Type DS-206 Circuit Breakers degraded lubricant.

M97328 IN DLSkeen

--/--/--

8/31/97 L-IN 95-22.Sup 1. Hardened or Supplement to IN to discuss additional ContaminatM Lubricants Cause area of operating mechanism where Metal-Clad Circuit Breaker hardened lubricant can cause breaker Failures failure.

M97397 IN JRTappert 11/11/11 10/31/97 T IN: Potential Deficiency of Notifies licensees cbout information Electric Cable Connections obtained from aging and LOCA testing of electrical cable connections as contained in the Sandia National Laboratory draft report NUREG/CR-6412.

M98643 IN DLSkeen 11/11/11 8/31/97 L IN: Reversed Current Miswired current sensors in ABB K-line Transformer Leads Resulted in breakers may result in premature Loss of Multiple Safety tripping on overcurrent.

Functions M99300 IN CDPetrone

--/--/--

10/29/97 T IN: Reactor Trip Breaker Alerts licensees that three licensees Maintenance and Surveillance have recently identified potential Testing of Auxiliary Contacts problems regarding reactor trip breaker maintenence and/or surveillance testing.

M99470 GL JWShapaker

--/--/--

1/30/98 T GL: Electrical Grid Requests licensees to confirm that the Reliability Trends design basis for offsite power sources will be maintained for each nuclear power plant.

Page No.

4 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Comp Title Description o LTB = Materials and Chemical Engineering Branch M95279 GL JWShapaker 09/30/97 11/21/97 T GL: Modification of the Extending to operating reactor Requirements for Post-Accident licensees, on voluntary basis.

Sampling System relaxations in PASS program requirements.

M95290 GL JWShapaker

--/--/--

11/14/97 T GL: Degradation of Steam Identification of steam generator Generator Internals internals degradation mechanisms based on foreign reactor operating experience.

M95373 GL JWShapaker 12/31/96 11/21/97 T GL: Implementation of App.

Discusses the need for lecensees to VIII of Sec XI of The 1995 adopt the Appendix VIII to improve the Edition of The ASME Boiler And quality and confidence level of Pressure Vessel Code inservice inspections.

P95444 LT TAGreene 12/31/97 3/15/98 T Lead Technical Review -

Cracking has been found in several Induction Heat Stress utilities' austentic stainless steel Improvement for Stainless piping which had been subjected to IHSI Steel Piping in the 1980's. Staff concerns include that IHSI may not have been properly applied.

M96401 GL JWShapaker

--/--/--

11/21/97 T GL: Steam Generator Tube Informs licensees of the importance of Inspection Techniques performing steam generator tube inservice inspections using cualified techniques and requests that 1icensees

)

implement described actions.

Page No.

5' 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPnRT Open Generic Communication and Compliance Activities Sorted by Lead. Technical Division and Branch TAC

' Type Contact TR Comp '

LA Comp Title Description M97743 LT EJBenner

--/--/--

12/31/97 T -LT: Weld Toughness of Moment Evaluate need for further generic Connection action related to weld failures during<

Northridge earthquake.

M98888 IN WFBurton

--/--/--

IN: Procedure Problems with Inadequate controls on valve-Leak Sealing.

leak-sealing process. leads to sealant-affecting other equipment.

M99226 GL JWShapaker

--/--/--

12/30/97 T GL: Augmented Inspection Issue Proposes augmented inspection of small for Small Diameter. Class 1-diameter. Class 1 piping in PWR Piping in PWR high-pressure-injection systems to High-Pressure-Injection System overcome ASME code oversight.

M99340 IN RABenedict

--/--/--

10/29/97 T IN: Fire Hazard in Use of Leak Not-normally-flammable mineral. oil in -

Sealant raw leak sealant may spontaneously ignite if it drips onto fibrous insulation that covers hot equipment.

M99432 GL JWShapaker 12/31/97 12/30/97 T GL: Steam Generator Tube Informs licensees that actions beyond' Integrity current TS requirements may be necessary to ensure steam generator tube integrity.

M99434. IN EJBenner

--/--/--

10/31/97.T IN: Recent SG Inspection Informs licensees the findings from the Experience

. examination of steam generator tubes.

and secondary side components in PWRs.

i

Page No.

6 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY' STATUS REPORT

-Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC.

Type Contact 1R Comp LA Comp Title

. Description c LTB = Mechanical Engineering Branch M96354 LT TAGreene

--/--/--

12/31/97 T Containment Recirculation Millstone 3 determined that the.

Spray'and Quench Spray Piping

. containment recirculation' spray and Outside Design Basis cuench spray piping and supports could:

~e subjected to higher accident temperatures than those previously assumed in the design basis.

M9661d LT TKoshy 11/14/97-6/26/97 L LPSI Pump Mission Time When the RCS pressure' remains higher than.LPSI injection. head the pumps may.

be required to run for long durations.

with minimum flow.

It appears that there is no demonstrated evidence to ensure LPSI puinp capability for the require mission time.

-M97327 LT CDPetrone

--/--/--

9/30/97 T LT: Target Rock Two-Stage SRV Consider. Issuing an information notice Setpoint Drift when BWR owners group comes to a conclusion regarding the cause of the larget Rock two-stage SRV setpoint drift.

M98808 IN TAGreene

--/--/--

5/30/98 T IN: Issues Associated with Alerts. licensees to potential generic CHECWORKS MODEL Predictions problems _related to the' occurrence and prediction of flow-accelerated corrosion in extraction steam systems.

Page No.

~7

<09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMDNTHLY STATUS REPORT

'Open' Generic Comunication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact' TR Comp LA Comp Title Description M99024 IN TAGreene

--/--/--

12/30/97 T IN: Use of Nonconservative Alerts licensees.to potential problems; Acceptance Criteria in associated with safety-related pump:

Inservice Pump Tests surveillance inservice' testing.

M99352 GL JWShapaker

--/--/--

12/15/97 T_ GL: MSLB Leakage Calculations Addresses calculational.inconsistencyf Performed in Support of SG between MSLB. leakage assessment Tube Voltage-Based Repair performed in support of'SG tube repair Criteria & Site Allowable LR criteria and site-allowable leak? rate Limits limits.

• • LTD = Division of Inspection and Support. Programs

• LTB = Inspection Program Branch M98889 IN TAGreene'

--/--/--

12/30/97 T IN: Part 9990~ Technical Informs licensees of NRC position Guidance on TS Interpretation concerning TS interpretation and the steps to take if they would like NRC formal review.

• LTB = Special Inspections Branch M97801 IN DLSkeen

--/--/--

8/31/97 L IN: Setpoint Drift in ITT Sulfur-induced corrosion may cause -

Barton Model 753 Gage Pressure excessive setpoint drift in Model 753 Transmitters transmitters.

-_=

Page No.

8 j

09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Open Generic Comunication and Compliance Activities Sorted by Lead Technical Division and Branch 4

TAC Type Contact TR Camp LA Comp Title Description oo LTD - Division of Reactor Controlc and Human Factors o LTB = Human Factors Branch M99022 IN EJBenner

--/--/--

9/22/97 T IN: Use of Operator Actions in Alerts licensees to a recent instance Place of Automatic Actions for where a licensee inappropriately used Design Basis Accident operator actions in place of automatic i

Mitigation actions for design basis accident mitigation without the regiaired prior NRC review and approval.

o LTB = Quality Assurance and Maintenance Branch i-M98441 GL JWShapaker 10/10/97 9/26/97 T GL: Quality Assurance of In view of technological advancements.

Electronic Records changes in NRC regulations. a request was made to update the guidance i

provided in GL 88-18.

• o LTD - Division of Reactor Program Management o LTB = Emergency Preparedness and Radiation Protection Branch M98237 IN TAGreene

--/--/--

2/25/98 T IN: Remova of FTS Lines froin Ale is licensees that NRC is removing Service from service some direct access telephone lines located at their facilities.

d w--

Page No.

9 09/11/97 f1JBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Comp Title Description M98751 GL JWShapaker

--/--/--

10/31/97 T GL: Clarificatior, of NUREG/CR Clarifies that methods discussed in 5055. " Atmospheric Diffusion contractor-prepared NUREGs (NUREG/CRs) for Control Room Habitability are not necessarily approved or Assessment" endorsed by the NP,C o LTB = Generic Issues and Environmental Projects Branch M99271 GL JWShapaker

--/--/--

9/30/97 T GL 91-18. Rev 1. Info to Informs licensees of change in NRC Licensees Re two NRC Insp inspection manual guidance for the Manual Sectns on Resltn of resolution.* degraded and Degraded and Nonconf. Conds nonconfoming conditions at facilities.

and on Operab.

oo LTD = Division of Systems Safety and Analysis o LTB - Analytical Support Group M96947 LT TAGreene 06/21/99 12/31/97 T LT : Possible Computer Code Identical computer models launched from Platform Dependency different personal computer platforms can result in different calculations.

o LTB = Containment Systems and Severe Accident Branch M96537 GL JWShapaker 12/30/97 9/30/97 T GL: Assurance of Sufficient Notifies licensees about a NPSH for ECCS and Containment safety-significant issue that could Heat Recoval System Pumps affect the ability for long-term core cooling and containment beat removal under accident conditions and which has generic implications.

Page No.

10 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC

' Type Contact TR Camp LA Camp Title Description

. M97146 GL JWShapaker 12/31/97 11/14/97 T GL: Degradation of ECC Notifies addressees about the potential Recirculation Following a LOCA safety impact of foreign material in oue to Foreign Material in the sur.ps and suppression pools, which Containment could render. safety-related equipment inoperable.

M97297 LT EJBenner

--/--/--

11/30/97 T LT: Errors in Containment Code Identify generic actions necessary as a Analysis result of potential errors in Oconee'sBulletin 80-04 respor.se.

M98125 LT TJCarter 09/30/97 LT: BWR Containment Bypass A plant configuration during routine Flow During Purging operation could potentially result in containment bypass following an accident o LTB = Plant Systems Branch M80296 LT TAGreene

--/--/--

9/30/97 T Generic Communications -

Development of staff NUREG or other Assessment of Turbine Failure publication to document turbine at Vandellos 1 building fire issues for U.S. plants in light of Vandellos fire.

M91323 LT CVHodge 09/11/98 7/11/97 L Reactor Water Cleanup (RWCU)

Review of the effects of an unisolated Study in Response to ACRS RWCli break at several BWR s.

Result of Concern ACRS concerns during the review of the ABWR

l Page No.

11 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATb5 REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Comp LA Comp Title Description M93335 LT WFBurton 03/13/98 8/31/98 T Main Control Room Envelope Use improved methodology to verify the Unfiltered Inleakage effects of potential inleakage rates on compliance with radiation and toxic gas exposure limits inside the main control room.

M96912 LT WFBurton 03/31/97 12/31/97 T. LT: Potential Generic Concern Farley - Failure of numerous pre-action with regard to Fire Protection sprinklers in fire protection systems 4

Actuation System providing fire protection service to safety-related system components.

M96913 BL JWShapaker 09/30/98 9/30/97 T BL: Potential for Loss of To alert licensees to recent 4

Remote Shutdown Capability noncompliances and associated civil 4

during a Control Room Fire penalties regarding licensee's lack of demonstrable protection from a control room hot short condition.

i M97299 GL JWShapaker

--/--/--

10/31/97 T GL: Spent Fuel Pool Compliance Requests licensees to describe their Activities spent fuel pool offload practices.

l temperature limits and bases. and decay heat removal redundancy and include the information in the FSAR.

M97978 GL JWShapaker 03/13/98 12/24/97 T GL: Laboratory Testing of Informs addressees about NRC staff Nuclear-Grade Activated views on charcoal testing practices and Charcoal offers model technical specifications for voluntary adoption by the addressees in preparation for future l

testing obligations.

i o

Page No.

12 09/11/97-PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Open Generic Comunication and Compliance Activi.-ies Sorted by Lead Technical Division and Branch TAC Type Contact TR Camp LA Camp Title Description M98066 IN EJBenner

--/--/--

10/11/97 T IN: Misunderstanding of the Develop IN to inform licensees of Ultimate Heat Sink Licensing several instances of errors in Basis licensee's understanding of Ultimate Heat Sink licensing basis.

M98970 IN ENFields

--/--/--

IN: Failure of Omega Series Alerts licensees to the potential Sprinlker Heads problem with Omega sprinkler heads manufactured by Central Sprinkler Company.

M99203 GL JWShapaker 09/30/98 10/31/97 T GL %-06. Sup 1: Assurance of Notifies licensees about Equipment Operability and safety-significant issues that could Containment Integrity during affect containment integrity and Design-Basis Accident equipment operability during Conditions design-basis accident conditions.

M99331 IN WFBurton 12/31/97 12/31/97 T IN: Postulated Loss of Alerts addressees to concerns related Feedwater as a Result of a to flooding as a result of a non-design Pipe Break in the Circulating basis pipe break in the cirulating Water System water system M99408 IN WFBurton

--/--/--

12/31/97 T IN: Control Room Evacuation Alerts addressees to a recent event at due to Inadvertent Halon Haddam Neck involving Halon actuation Actuation as a result of a camera flash.

M99409 IN TJCarter

--/--/--

IN: Potential Proolems with Alerts licensees to potential problems Fire Barrier Penetration Seals in installed fira barrier penetration seals that may have gone.r.etected as a result of inadequate surveillance inspection procedures and inadequate

acceptance criteria.

09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Camp LA Ccmp Title Description o LTB = Reactor Systems Branch M92635 GL JWShapaker 05/21/01 1/16/98 T GL: Reactor Coolant Inventory Loss of ECCS function due to steam Loss and Potential Loss of voiding in RWST line to suction of ECCS Emergency Mitigation Functions pumps due to loss of RCS inventory.in While Shutdown Mode 4 (Wolf Creek).

M94565 LT DLSkeen

--/--/--

7/31/97 L Slow Scram Solenoid Pilot Scram solenoid pilot valves with viton Valves Caused by Viton diaphragms showing degraded scram times Diaphragms within 6-8 months. Currently tracking licensee response to RRG recommendations.

M96192 IN WFBurton

--/--/--

12/31/97 T IN: ECCS Throttle Valves May High differential pressure across ECCS Degrade Due To Cavitation throttle valves during LOCA could cause Induced Erosion During LOCA pump runout flow and subsequent ECCS pump damage M96615 LT TKoshy 12/31/97 6/25/97 L Boron Precipitation in B&W Design bases concern on active means of Reactors preventing boron precipitation following a LOCA.

M9/150 LT TJCarter

--/--/--

6/30/97 L LT: Evaluate Postulated A potential scenario not adequ&tely Concern During Cool 90wn of adcressed by E0Ps was discovered during Reactor Following a Reactor an inspection at Cooper.

Shutdown after ATWS Event v

Page No.

14 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT

. Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Camp LA Comp Title Description M97331 BL JWShapaker 12/03/97 10/31/97 T BL: Inadecuate Procedural Requests PWR licensees to take action Guidance curing S/D and Site to assure that there is adequate Specific Vulnerabilities due procedural guidance during shutdown to Gas Accumulation operation and that gas accumulation vulnerabilities are identified. and actions are taken to limit or preclude adverse system performance.

M97396 BL JWShapaker 10/30/97 9/30/97 T BL %-01. Sup 1. Control Rod-Informs addressees of issues Insertion Problems concerning incomplete control rod insertion due to distortion of thimble tubes.

M98812 LT TKushy 05/31/99 LT: Generic Issues on the Studies generic issues on the Interruption of Flow to the interruption of flow to the core for Core for Aligning Sump aligning sump flow from the injection Recirculation to recirculation mode.

.M99046 IN WFBurton

--/--/--

11/30/97 T IN: Recent loss of Inventory Alerts addressees to recent loss of Events inventory events at Sequoyah and Zion.

M99332 GL JWShapaker 10/21/97 12/31/97 T GL: Guidance on the Regulatory Provides a compilation of the current Requirements for Criticality NRC staff guidance on regulatory Analysis of Fuel Storage at requirements for criticality analysis LWR Power Plants of new and spent fuel storage at LWR plants.

o Page No.-

15 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact TR Camp LA Comp Title Description M99433 IN WFBurton

--/--/--

12/31/97 T IN: Effects of Crud Buildup Alerts addressees to axial offset and Boron Deposition on Power anomaly problems at Callaway.

Distribution and Shutdown Margin

Page No.

1 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Comunication and Compliance Activities Added Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Camp LA Camp Title Reason Added M98643 IN DLSkeen Electrical 11/11/11 8/31/97 L IN: Reversed Current The EAP authorized development of Engineering Transformer Leads IN at its 5/6/97 meeting.

Branch Resulted in Loss of Multiple Safety Functions M98751 GL JWShapaker EmergcN /

--/--/--

10/31/97 T GL: Clarification of The EAP authorized development of Prepared.ess NUREG/CR 5055.

GL at its 5/20/97 meeting.

and Radiation

" Atmospheric Protection Diffusion for Branch Control Room Habitability Assessment" M98808 IN TAGreene Mechanical

--/--/--

5/30/98 T IN: Issues The EAP authorized development of Engineering Associated with IN at its 5/27/97 meeting.

Branch CHECWORKS MODEL Predicticas M98812 LT TKoshy Reactor Systems 05/31/99 LT: Generic Issues The EAP authorized LT follow-up Branch on the Interruption of this issue at its 5/27/97 of Flow to the Core meetimg.

for Aligning Sump Recirculation M98888 IN WFBurton Materials and

--/--/--

IN: Procedure The EAP authorized development of Chemical Problems with Leak IN at its 6/3/97 meeting.

Engineering Sealing Branch

Page No.

2

'09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT.

Generic Communication and Compliance Activities Added Since the Last Public Report (April 1997)

TAC-Type Contact.

Tech Branch TR Camp LA Comp Title Reason Added M98889 IN TAGreene Inspection

--/--/--

12/30/97 T IN: Part 9990 The EAP authorized development of Program Branch Technical Guidance IN at its 6/3/97 meeting.

on TS Interpretation M98970 IN ENFields Plant Systems

--/--/--

IN: Failure of Omega Thhe EAP authorized developtrent Branch Series Sprinlker of IN at its 6/10/97 meeting.

Heads M99000 IN WFBurton

--/--/--

12/31/97 T IN: Dispositioning Thhe EAP authorized development of Technical of IN at its 6/10/97 meeting.

Specifications Discovered Not to be Sufficient to Assure Plant Safety M99022 IN EJBenner Human Factors

--/--/--

9/22/97 T IN: Use of Operator The EAP authorized development of Branch Actions in Place of IN at its 6/17/97 meeting.

Automatic Actions for Design Basis Accident Mitigation M99024 IN TAGreene Mechanical

--/--/--

12/30/97 T IN: Use of The EAP authorized developr vit of Engineering Nonconservative IN at its 6/17/97 meeting.

Branch Acceptance Criteria in Inservice Pump Tests

Page No.

3 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Communication and Compliance Activities Added Since the Last Public Report (April 1997)

TAC Ty p Contact Tech Branch TR Camp LA Camp Title Reason Added

'M99046 IN WFBurton Reactor Systems --/--/--

11/30/97 T IN: RecentLdssof The EAP authorized development of Branch Inventory Events IN at its 6/17/97 meeting.

M99203 GL, JWShapaker Plant Systems 09/30/98 10/31/97 T GL 96-06. Sup 1:

The EAP authorized development of Branch Assurance of GL at its 7/8/97 meeting.

Equipment Operability and Containment Integrity during Design-Basis Accident Conditions M99226 GL JWShapaker Materials and

--/--/--

12/30/97 T GL: Augmented The EAP authorized development of Chemical Inspection Issue for GL at its 7/8/97 meeting.

Ergineering Small Diameter.

Bi.:nch Ciass 1 Piping in PWR High-Pressure-Inject ion System M99271 GL JWShapaker Generic Issues --/--/--

9/30/97 T GL 91-18. Rev 1.

The EAP authorized development of and Info to Licensees Re GL at its 7/23/97 meeting.

Environmental two NRC Insp Manual Projects Branch Sectns on Resltn of Degraded and Nonconf. Conds and on Operab.

Page No.

4 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Com;nication and Compliance Activities Added Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Camp LA Comp Title Reason Added M99300 IN CDPetrone Electrical

--/--/--

10/29/97 T IN: Reactor Trip The EAP authorized development of Engineering Breaker. Maintenance IN at its 7/29/97 meeting.

Branch and Surveillance Testing of h;xiliary Contacts M99304 LT TAGreene Civil 09/30/98 10/30/98 T HOLTEC Part 21 The acting branch chief (P1D)

Engineering and Computer Code Issue authorized issuing this TAC.

Ce sciences Branch M99331 IN WFBurton Plant Systems 12/31/97 12/31/97 T IN: Postulated Loss The EAP authorized development of Branch of Feedwater as a IN at its 8/5/97 meeting.

Result of a Pipe B.eak in the Circulating Water System

. eactor Systems 10/21/97 12/31/97 T GL: Guidance on the The EAP authcrized development of M99332 GL JWShapaker R

Branch Regulatory GL at its 8/5/97 meeting.

Requirements for Criticality Analysis of Fuel Storage at LWR Power Plants M99340 IN RABenedict Materials and

--/--/--

10/29/97 T IN: Fire Hazard in The EAP authorized developtrent of Chemical Use of Leak Sealant IN at its 7/29/97 meeting.

Engineering Branch

O h

Page No.

5 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Communication and Comoliance Activities Added Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Comp LA Comp Title Reason Added M99352 GL JWShapaker Mechanical

--/--/--

12/15/97 T GL: MSLB Leakage The EAP authorized development of Engineering Calculations GL at its 8/5/97 meeting.

Branch Perfonned in Support of SG Tube Voltage-Based Repair Criteria & Site Allowable LR Limits M99408 IN WFBurton Plant Systems

--/--/--

12/31/97 T IN: Control R om The EAP authorized development of c

Branch Evacuation due to IN at its 8/12/97 meeting.

Inadvertent Halon Actuation M99409 IN TJCarter Plant Systems

--/--/--

IN: Potential The EAP authorized development of Branch Problems with Fire IN at its 8/12/97 meeting.

Barrier Penetration Seals M99432 GL JWShapaker Materials and 12/31/97 12/30/97 T GL: Steam Generator The EAP authorized development of Chemical Tube Integrity GL at its 8/19/97 meeting.

Engineering Branch M99433 IN WFBurton Reactor Systems --/--/--

12/31/97 T IN: Effects of Crud The EAP authorized developrrent of Branch Buildup and Boron IN at its 8/19/97 meeting..

Deposition on Power Distribution and Shutdown Margin

+

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Page No.

6 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Comunication and Compliance Activities Added Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Camp LA Comp Title Reason Added i

M99434 IN EJBenner Materials and

--/--/--

10/31/97 T IN: Recent SG The EAP authorized development of Chemical Inspection IN at its 8/19/97 meeting.

i Engineering Experience Branch 7

M99470 GL JWShapaker.

Electrical

--/--/--

1/30/98 T GL: Electrical Grid The EAP authorized development of i

Engineering Reliability Trends GL at its 8/26/97 meeting.

Branch i

l I

t I

t I

L

I Page No.

1 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Communication and Compliar,ce Activities Closed Since the Last Public."eport (April 1997)

TAC Type Contact Tech Branch TR Comp LA Conp Title Reason Closed M91544 GL JWShapaker Events 05/16/97 P 5/16/97 C GL: Defining Info in GL 97-02 issued 5/15/97.

Assessment and Monthly Operating Generic Report Required by Communications Tech Specs Branch M95278 GL JWShapaker Reactor Systems 09/05/97 P 9/5/97 C GL: Use of The proposed GL was canceled.

Branch Thermal-Hydraulic The EAP agreed with this Codes for Licensing recommendation at its 8/26/97 Applications meeting.

-M95871 IN TAGreene Plant Systems 06/10/97 P 6/10/97 C IN: Emergency IN 95-36. Sup 1 issued 6/10/97.

Branch Lighting Issues M96073 IN EJBenner Mechanical 07/11/97 P 7/11/97 C IN: Problems IN 97-51 issued 7/11/97.

Engineering Experienced with Branch loading and Unloading Spent Nuclear Fuel Storage and Transporation Casks M96714 IN TKoshy Mechanical 07/17/97 P 7/17/97 C IN 91-50. Sup 1.

IN 91-50. Sup 1. issued 7/17/97.

Engineering Water Hammer Events Branch Since 1991

o

-Page No.

2 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Ccamenication and Compliance Activities Closed Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Camp LA Comp Title Reason Closed M96961 IN CDPetrone Reactor Systems 05/14/97 P 5/14/97 C IN: Extended IN 87-10. Sup 1. issued 5/15/97.

Branch Operation in Suppress. ion Pool Cooling Moca M97151 IN TAGreene Plant Systems 07/09/97 P 7/9/97 C IN: Inadequate or IN 97-48 issued 7/9/97.

Branch Inappropriate Fire Protectica Compensatory Measures M97329 IN EJBenner Materials and 05/19/97 P 5/19/97 C IN: Degradation in IN 97-26 issued 5/19/97.

Chemical Small-Radius U-Bend Engineering Regions of Steam Branch Generator Tubes M97667 IN JRTappert Mechanical 06/27/97 P 6/27/97 C IN: Undersized Oil IN 97-41 issued 6/27/97.

Engineering Heat Exchangers Branch M97799 LT ENFields Analytical 09/10/97 P 9/10/97 C LT: Loop Seal The SASG*s report "Small Break Support Group Clearing LOCA Loop Seal Sensitivity ~

Investigation -

indicated that for the case Westinghouse.

analyzed. prolonged core uncovery did not pose problem.

.Page No.

3 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Communication and Compliance Activities Closed Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Comp LA Comp Title Reason Closed M97800 LT ENFields Analytical 09/10/97 P 9/10/97 C LT: Loop Seal The SASG's report "Small Break Support Group Clearing LOCA Loop Seal Sensitivity

• Investigation - CE indicated that for the case analyzed. prolonged core uncovery did not pose problem.

M98029 IN CDPetrone Emergency-06/18/97 P 6/18/97 C IN: Unplanned Worker IN 97-36 issued 6/20/97.

Preparedness Intakes of and Radiation Transuranics and Protection External Exposure Branch due to Inadequate Control of Work M98030 IN CVHodge Events 07/01/97 P 7/1/97 C IN: Inadequate IN 97-39 issued 6/25/97.

Assessment and Safety Evaluation at Generic Licensed Independent Communications Spent Fuel Storage Branch Installations M98064 IN JRTappert Reactor Systems 06/26/97 P 6/26/97 C IN: Nitrogen IN 97-40 issued 6/26/97.

Branth Intrusion into ECCS Piping M98065 IN ENFields Plant Systems 07/17/97 P 7/17/97 C IN: Inadvertent loss IN 97-52 issued 7/17/97.

Branch of ECCS Motor Cooling Capability

Page No.

4 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Comunication and Compliance Activities Closed Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Comp LA Camp Title Reason Closed' M98126 IN TAGreene Electrical 07/18/97-P 7/18/97 C IN: Circuit Breakers IN 97-53 issued 7/18/97.

Engineering Left Racked Out in Branch Non-seismically Oualified Position M98182 IN EJBenner Materials and 07/10/97 P 7/10/97 C IN: B&W Once-Through IN 97-49 issued 7/10/97.

Chemical Steam Generator Tube Engineering Inspection Findings Branch M98183 IN CVHodge Non-Power 07/01/97 P 7/1/97 C IN: Potential IN 97-44 issued 7/1/97.

Reactors and Undetectable Failure Decommissioning in Linear Neutron Project Flux Monitor at Directorate Non-Power Reactor Facilities M98233 IN EJBenner Mechanical 06/03/97 P 6/3/97 C IN: Reactor Coolant IN 97-31 issued 6/3/97.

Engineering Pump Degradation Branch Experience in Foreign Plants M98234 IN TJCarter Electrical 07/02/97 P 7/2/97 C IN: Environmental IN 97-45 issued 7/2/97.

Engineering Qualification Branch Deficiency for Cables and Containment Penetration Pigtail

Page No.

5 09/11/97-PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Communication and Compliance Activities Closed Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Comp LA Camp Title Reason Closed M98235 IN DLSkeen Special 06/13/97 P 6/13/97 C IN: Defective IN 97-32 issued 6/10/97.

Inspections Critical Component Br6nch in Limitorque Actuator M98238 IN JRTappert Technical 07/01/97 P 7/1/97 C IN: License IN 97-43 issued 7/1/97.

Specifications Condition Compliance Branch M98239 IN TKoshy Instrumentation 05/09/97 P 5/9/97 C IN: Dynamic Range IN 97-25 issued 5/9/97.

and Controls Uncertainties of Branch Reactor Vessel Level Instrumentation System M98323 IN CVHodge Instrumentation 05/30/97 P 5/30/97 C Elimination of IN 97-28 issued 5/30/97.

and Controls Instrument Response Branch Time Testing Under The Requirement of 10 CFR 50.59 M98379 IN TAGreene Civil 05/30/97 P 5/30/97 C Implementation of IN 97-29 issued 5/30/97.

Engineering and Containment Geosciences Inspection Rule Branch

..-........................--.-:--..,u___

1

o Page No.

6 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT

. Generic Comunication and Compliance Activities Closed Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Comp LA Camp Title Reason Closed M98442 IN TJCarter Emergency 09/03/97 P 9/3/97 C IN: Loss of Control IN 97-68 issued 9/3/97.

Preparedness of Diver in a Spent and Radiation Fuel Storage Pool Protection Branch M98443 IN EJBenner Electrical 07/02/97 P 7/2/97 C IN 96-44. Sup 1.

IN 96-44. Sup 1 issued 7/2/97.

Engineering Failure of RTB from Branch Cracking of Phenolic Material in Secondary Contact Assembly M98644 IN TKoshy Non-Power 07/18/97 P 7/18/97 C IN: Expiration of IN 97-54 issued 7/18/97.

Reactors and Non-pow r Reactor Decomissioning Operator Licenses Project Directorate M98691 IN TJCarter Containment 05/24/97 C 5/16/97 C IN: Effect of IN 97-27 issued 5/16/97 Systems and Incorrect Strainer Severe Accident Pressure Drop on Branch Available NPSH M98692 IN CDPetrone Reactor Systems 06/11/97 P 6/11/97 C IN: Unanticipated IN 97-33 issued 6/11/97.

Branch Effect of Ventilation System on Tank Level Indications and ESFAS Setpoint L

Page No.

7 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Comunication and Compliance Activities Closed Since the Last Public Report (April 1997)

TAC Type Contact Tech Branch TR Camp LA Camp Title Reason Closed M99700 IN TKoshy Emergency 06/20/97 P 8/20/97 C IN: Oversight on IN 97-66 issued 8/20/97.

Preparedness Respirator Qual. &

and Radiation Corrective Lenses Protection fo.^ Operator Branch M98748 IN ENFields Instrumentation 06/24/97 P 6/24/97 C IN: Level IN 97-38 issued 6/24/97.

and Controls Instrumentation Branch System Initiates Comon Mode Failure of HPI Pumps M98749 IN EJBenner Mechanical 07/09/97 P 7/9/97 C IN: Unisolable IN 97-46 issued 7/9/97.

Engineering Pressure Boundary Branch Leak M98750 IN CDPetrone Operator 08/19/97 P 8/19/97 C IN: Failure to IN 97-67 issued 8/21/97.

Licensing Satisfy Requirements Branch for Significant Manipulations of the Controls for Power Reactor Operator Licensing M98807 IN DLSkeen Probabilistic 06/24/97 P 6/24/97 C IN: Main Transformer IN 97-37 issued 6/20/97.

Safety Fault with Ensuing Assessment Oil Spill into Branch Turbine Building

Page No.

8 09/11/97 PUBLIC SEPTEMBER 1997 DIRECTOR'S BIMONTHLY STATUS REPORT Generic Communication and Compliance Activities Closed Since the Last Public Report (April 1997)

TAC Type Contact

. Tech Branch TR Comp LA Comp Title Reason Closed M98811 IN TKoshy Reactor Systems 08/01/97 P 8/1/97 C IN: Incorrect USO IN 97-60 issued 8/1/97.

[

Branch Determination Related to ECCS Swapover from the Injection Mode to the Recirculation Mode M98849 IN TJCarter Emergency 06/12/97 P 6/12/97 C IN: Terminology in IN 97-34 issued 6/12/97.

Preparedness Emergency Plan and Radiation Revisions Protection Branch M98890 IN TKoshy Reactor Systems 08/06/97 P 8/6/97 C IN: Unrecognized IN 97-62 issued 8/6/97 Branch Reactivity Addition during Plant Shutdown M99023 EJBenner Mechanical 07/09/97 P 7/9/97 C IN: Crack in High This TAC was inadvertently taken Engineering Pressure Injection out Branch Piping it is the same as M98749 M99155 IN-TJCarter Plant Systems 08/01/97 P 8/1/97 C IN: Fire Endurance IN 97-59 issued 8/1/97.

Branch Test Results of Versawrap Fire Barriers

.