ML20134G611
| ML20134G611 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| References | |
| NUDOCS 9611130298 | |
| Download: ML20134G611 (88) | |
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DIRECTOR'S STATUS REPORT on GENERIC ACTIVITIES Action Plans Generic Communication and Compliance Activities OCTOBER 1996 Office of Nuclear Reactor Regulation
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1 INTRODUCTION The purpose of this report is to provide information about generic activities, including generic communications, under the cognizance of the Office of Nuclear Reactor Regulation. This report, which focuses on compliance activities, complements NUREG-0933, "A Prioritization of Generic Safety issues."
This report includes two attachments: Il action plans and 21 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 scrficient complexity or scope that require substantial NRC staff resources. The issues covere.f by action plans include concerns identified through review of operating experience (e.g. Boilirvj Water Reactor internals Cracking and Thermolag), and issues related to regulatory flexibilhy and improvements (e.g. New Source Term and Probabilistic Risk Assessment (PRA) lmplementation 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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NRR ACTION FLANS I
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TABLE OF CONTENTS DE BOILING WATER REACTOR INTERNALS CRACKING 1
REACTOR PRESSURE VESSEL FRACTURE TOUGHNESS...........
4 MOTOR OPERATED VALVES ACTION PLAN...................
8 DRCH UPDATE OF SRP Ci! APTER 7 TO INCORPORATE DIGITAL INSTRUMENTATION AND CONTROLS (l&C) GUIDANCE
....... 11 GRADED QUALITY ASSURANCE ACTION PLAN 13 DRPM NEW SOURCE TERM FOR OPERATING REACTORS 18 ENDANGERED SPECIES ACTION PLAN....................... 21 EFFECT OF HURRICANE ANDREW ON TURKEY POINT............ 23 ENVIRONMENTAL SRP REVISION ACTION PLAN................ 25 10 CFR 50.59 ACTION PLAN DEVELOPMENT.................. 27 INDUSTRY DEREGULATION AND UTILITY RESTRUCTURING........ 30 DRPW GENERAL ELECTRIC EXTENDED POWER UPRATE ACTION PLAN..... 33 DRY CASK STORAGE ACTION PLAN 35 DSSA ACCIDENT MANAGEMENT IMPLEMENTATION................. 38 FIRE PROTECTION TASK ACTION PLAN...................... 40 THERMO-LAG ACTION PLAN............................
42 PRA IMPLEMENTATION ACTION PLAN 44 ENVIRONMENTAL QUALIFICATION TASK ACTION PLAN.......... 48 GENERIC SPENT FUEL STORAGE POOL PART A: OPERATING FACILITIES....................... 50 GENERIC SPENT FUEL STORAGE POOL PART B: PERMANENTLY SHUTDOWN FACILITIES 53 CORE PERFORMANCE ACTION PLAN........................ 55 HIGH BURNUP FUEL ACTION Pl.AN......................... 58 RRG TOPIC AREA 55: CYCLE SPECIFIC PARAMETER LIMITS IN TECH SPECS....................................
60 WOLF CREEK DRAINDOWN EVENT ACTION PLAN 62 1
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BOILING WATER REACTOR INTERNALS TAC Nos. M91898, M93925, M93926, M94959, Last Update: 09/24/96 M94975, M95369, M96219, M96539 Lead NRR Division: DE GSl: Not Available Supporting Division: DSSA MILESTONES DATE (T/C)
PART 1: REVIEW OF GENERIC INSPECTION AND EVALUATION CRITERIA
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03/96 C
- 1. Issue summary NUREG 1544
- 2. Review BWRVIP Re-inspection and Evaluation Criteria 10/96 T o Reactor Pressure Vessel and internals Examination Guidelines (BWRVIP-03) 12/96 T o BWRVIP-03, Section 6A, Standards for Visual Inspection of Core Spray Piping, Spargers, and Associated Components 08/96 T o BWR Vessel Shell Wold inspection Recommendations (BWRVlP-05) 08/96 T o Guidelines for Reinspection of BWR Core Shrouds (BWRVIP-07)
- 3. Review of generic repair technology, criteria and guidance TBD
- 4. Review generic mitigation guidelines and criteria TBD
- 5. Review of generic NDE technologies developed for examinations of BWR internal components and attachments,
- 6. Other internals reviews (safety assessments, evalustians, mitigation measures, inspections and repairs) 10/96 T o Safety Assessment of BWR Reactor Internes (BWRVIP-06) 12/96 T o Evaluation of Crack Growth in BWR Stainle\\s Steel RPV intomals (BWRVIP-14) 03/97 T o BWR Core Spray Internals inspection and Fisa Evaluation Guidelines (BWRVIP-18) 04/97 T o internal Core Spray Piping and Sparger Repair Design Criteria (BWRVIP-19) pescnot#on: Many components inside boiling water reactor (BWR) vessels (i.e., internals) are ma of materials such as stainless steel and various alloys that are susceptible to corrosion and j
cracking. This degradation can be accelerated by stresses from temperature and pressure changes, chemical irG::ctions, 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 lGSCC in core shrouds and other BWR internals Significant cracking of the core shroud was first observed at Brunswick, l
Historical Racharound:
Unit I nuclear power plant in September 1993. The NRC notified licensees of Brunswick's l
discovery of significant circumferential cracking of the core shroud welds. In 1994, core shroud cracking continued to be the most significant of reported internals cracking. In July 1994, the N 1
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issued Genenc Letter 94-03 which requires licensees to inspect their shrouds and provide an analysis justifying continued operation until inspections can be completed.
A special industry review group (Boiling Water Reactor Vessels and internals Project-BWRVIP) was formed to focus on resolution of reactor vessel and intemals degradation. This group was instrumental in facilitating licensee responses to NRC's Generic Letter. The NRC evaluated the review group's reports, submitted in 1994 and earfy 1995, and all plant responses.
All of the plants evaluated have been able to demonstrate continued safe operation untilinspection or repair on the basis of: 1) no 360' through-wall cracking observed to date, 2) low frequency of pipe breaks, and 3) short penod 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 creciung was discovered in the top guide and core plate rings of a foreign reactor. The design is simstar to General Electnc (GE) reactors in the U.S., however, there have been no observations of such cracking in U.S. plants. GE concluded that it was reasonable to expect that the ring cracking could occur in GE BWRs with operating time greater than 13 years.
In the special industry review group's report, that was issued in January 1995, ring cracking was evaluated. The NRC concluded that the BWRVIP's assessment was acceptable and that top guide ring and core plate ring cracking is not a short term safety issue.
Pronosed Actions: The staff will continue to assess the scopes that have yet to be submetted by licensees concoming inspections or re-inspections of their core shrouds. The staff will also continue to assess core shroud inspection results and any appropriate core shroud repair designs on l
a case-by-case basis. The staff willissue separate safety evaluations regarding the acceptability of core shroud inspection results and core shroud repair designs. The staff has been interacting with the BWRVIP and individual licensees. In an effort to lower the number of industry and staff j
resources that will be, needed in the future, it is important for the staff to continue interacting with j
the industry on a generic basis in order to encourage them to continue their proactive efforts to resolve IGSCC of BWR intomals. The BWRVIP has subemtted nine generic documents, supporting j
plant-specific submittals, for staff review. The staff is ensuring that the generic reviews are incorporating recent operating experience on all BWR intomals.
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Oriainatsno Docurrent: Generic Letter 94-03, issued July 25,1994, which requested BWR I
licensees to inspect their core shrouds by the next outage and to justify continued safe operation i
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until inspections can be completed.
l Raautatorv Assessment: In July 1994, the NRC issued Generic Letter 94-03 which required l
licensees to inspect their shrouds and provide an analysis justifying continued operation until inspections could be performed. The staff has concluded in all cases that licensees have provided sufficient evidence to support continued operation of their BWR units to the refueling outages in which shroud inspections or repairs have been scheduled. In addition, in October 1995, industry's 1
special review group submitted a safety assessment of postulated cracking in all BWR reactor I
intomats and attachments to assure continuing safe operation.
Almost all BWRs completed inspections or repairs of core shrouds during refueling Current Status:
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 l
series of tie-rod assemblees. The NRC has reviewed and approved all shroud modification proposals i
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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 additional information. In addition, the industry group submitted a report on reinspection of repaired and non-repaired core shrouds (BWRVIP-07) in February 1996. The staff is currently reviewing this report and has sent a request for additional information. The NRC is also reviewing information submitted by GE on the safety significance of and recommended inspections for top guide snd core plate ring cracking. Review of the " Reactor Pressure Vessel and internals Examination Guidelines (BWRVIP-03)" has slipped to allow for review by NRC contractor.
NRR Technical Contacts:
David Torso, EMCB, 415-3317 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:
i Generic Letter 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," July 25,1994 j
Action Plan dated April 1995 1
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l REACTOR PRESSURE VESSEL FRACTURE TOUGHNESS ACTION PLAN TAC Nos.
M92310, M92313, M93329, Last Update: 09/30/96 M93330, M93331 Lead NRR Division: DE GSI: Not Available MILESTONES DATE (T/C)
- 1. ISSUE SUPPLEMENT TO GL 92-01 5/95 (C)
- 2. COORDINATION WITH RESEARCH 7/97 (T)
- 3. NRC/lNDUSTRY WORKSHOP ON RPV ISSUES 7/95 (C)
- 5. NUREG 1511 RPV STATUS PPPORT SUPPLEMENT 1 10/96 (T)
- 6. REVIEW OF GL 92-01 SU' PLEMENT 1,2ND ROUND 12/96 (T) r
- 7. NUREG 1511 RPV STATUS REPORT SUPPLEMENT 2 6/97 (T)
- 8. ISSUE OF RVID REVISION 1 6/96 (C)
- 9. ISSUE OF RVID REVISION 2 6/97 (T)
- 10. OBSERVE INDUSTRY ANNEALING DEMONSTRATION 12/96 (T) 11a. REVIEW PALISADES HAZ FRACTURE PROGRAM TEST 12/96 (T)
RESULTS 11b. REVIEW PALISADES HAZ CHARPY IMPACT TEST RESULTS 12/97 (T) 11c.NRC RESEARCH HAZ SIMULATION TEST PROGRAM 12/96 (T) 11d.NRC RESEARCH HAZ WELD TEST PROGRAM 12/97 (T) lle.NUREG 1511 RPV STATUS REPORT SUPPLEMENT 3 7/98 (T)
- 12. PREPARE RECOMMENDATIONS TO REVISE PROCESS FOR 7/98 (T)
ESTIMATING EMBRITTLEMENT
- 13. INDUSTRY BEGINS MAINTENANCE OF RVID 12/99 (T)
Descriotion: Appendix G to 10 CFR 50 and 10 CFR 50.61 establish requirements to prevent fracture of the reactor pressure vessel (RPV). These rules require licensees to project the amount of embrittlement of RPV materials. As a result of the review of responses to Generic Letter (GL) 92-01, the review of Palisades PTS issue, and recent inspections conducted at Combustion Engineering, several issues related to RPV evaluations have been identified. These issues can be summarized as follows:
(1) It appears that licensees may not have been aware of or considered all relevant information and data in previous assessments of their RPVs, (2) The variability in copper and nickel chemical composition may be independent of weld heat number and is greater than previously recognized by the staff, 4
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(3) The Palisades reactor vessel will be the first commercial nuclear vessel annealed in the U.S. to improve its fracture toughness.
Histoncal Backaround In March 1992, the NRC issued Generic Letter (GL) 92-01, Revision 1,
" Reactor Vessel Structural inteenty,10 CFR 50.54(f)." As a result of the information provided by the licensees in response to GL 92-01, Revision 1, the staff issued NUREG-1511, " Reactor Pressure Vessel Status Report," and the Reactor Vessel integrity Database (RVID). NUREG 1511 provides a summary of the critical issues and regulatory requirements involved in RPV structural integrity and the status of each RPV with respect to the regulatory requirements. The RV'D contains all the data that.was submitted by licensees to demonstrate comphance with ine regulatory requirements. Since Iscensees provide data during the life of the plant to c'amonstrate their comphance with regulatory requirements, NUREG-1511 and the RVID will require penodic upgrading.
In April 1995, the staff completed its evaluation of the Palisades plant comphance with the l
pressurized thermal shock (PTS) rule,10 CFR 50.61. The staff concluded that the Palisades RPV could be operated in comphance with the requirements of the PTS rule through the plant's 14th refuehng outage, which was scheduled for late 1999. To extend the life of the Palisados RPV beyond 1999, the licensee for Palisades has begun to plan for annealing of the Palisades RPV. The staff will review the licensee's annealing plan prior to its implementation. The Palisades anneal is scheduled for the 1998 refuehng outage Prior to this anneal the industry will be performing l
demonstration annealt, at the Marble Hill and Midland-2 sites.
As a result of information received during the Palisades PTS review, a meeting with Combustion Engeneering and two inspections at the Combustion Engineering offices in Windsor, Connecticut, the staff determined that licensees may not have been aware of or considered all relevant information and data in previous RPV assessments. Based on the above finding, the staff concluded that the most effective way to resolve this issue was through a supplement to GL 92-01 requiring the licensees to collect all data relevant to their RPVs, and if there are data that they had not previously considered, to perform a reassessment of their RPV.
l As a result of the data supplied in response to GL 92-01 and the Palisades PTS review, the Office of Nuclear Reactor Regulation roguested in a letter dated August 11,1995 that the Office of Nuclear Regulatory Research evaluate whether changes to the PTS rule or Regulatory Guide 1.99 are necessary.
Charpy impact and hardness tests by AEA Technology on simulated course grain heat affected rone (HAZ) material indicates that irradiated and annealed HAZ RPV material could be susceptible to temper embrittlement. In these tests, annealing did not result in a recovery of embrittlement (decrease in transition temperature), but resulted in an increase in transition temperature. This effect must be investigated for the Palisades RPV prior to its anneal.
l Specific actions included in the generic action plans are: (1) issue Supplement 1 Bonosed Actions:
ta GL 92-01, (2) coordination with RES on RPV integrity issues, (3) hold an NRC/ industry workshop r.c RPV issues, (4) review first and second round of responses to GL 92-01 Supplement 1, (5) issue sepplement 1 to NUREG-1511 in 1996 and issue supplement 2 to NUREG-1511 in 1997, (6) issue revision 1 of tta P.Vib in 1996 and issue revision 2 of the RVID in 1997, (7) observe industry antwaliry, demonstrations, (8) review and evaluate the Palisades annealing plan, and (9) review the l
Palisades anneal.
l The investigation of the effect of irradiation and annealing on fracture resistance of HAZ material will consist of five phases. Consumers Power will examine the fracture surface of irradiated and I
annealed HAZ material from the Palisades surveillance weld to determine whether its HAZ materi
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susceptible to temper embnttlement, the licensee will need to conduct Charpy impact tests of irradiated and annealed HAZ material. This phase of the program would need to be completed by i
12/97 (prior to the thermal anneal in 1998). These tests are not presently included in the j
licensee's program. A schedule for this program is being discussed with the licensee. This phase j
of the program is not necessary if the Palisades HAZ surveillance weld material is not susceptible to temper embattlement.
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The NRC confirmaton research on the effect of irradiation and annealing on the fracture resistance of HAZ material will casist of two phases. The initial phase will consist of tests on simulated HAZ material to determee whether the materials are susceptible to temper embnttlement and to j
determine whether annealmg results in a recuvery of embrittlement. The second phase will test i
HAZ material from submerged arc welds removed from a RPV.
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The test results from the Consumer Power and NRC programs on the effect of irradiation and i
annealing on fracture resistance of HAZ material will be published in the Safety Assessment of the Palisades annealing program and in a supplement to NUREG 1511.
I After reviewing the licensee and owners groups responses to GL 92-01, Supplement 1 and programs to define the best-estimate chemistry, the staff will provide recommendations to revise i
the process for estimating embrittlement and update the RVID. After differences between the RVID and the industry database are resolved, the staff / industry will develop procedures to permit i
industry representatives to maintain the database with NRC oversight.
l Oriainatina Document Memorandum from Jack R. Strosnider to Ashok C. Thadani, NRR, August 9,1995.
l Reaulatorv Assessment: This plan would allow for resolution of the issues discussed above in i
about two years. The staff anticipates that it will take the industry and the NRC this long to collect l
and assess all the relevant date. The staff assessed the impact of increased variability in chemistry l
on the RTm value of PWR reactor vessels in a memorandum from J.R. Strosnider to A.C. Thadani dated May 5,1995. The staff's assessment indicates that there is no immediate cause for concem and that there is adequate time to perform a more rigorous assessment of the issue. Based on the staff's genenc assessment of the impact of increased variability, the staff has concluded that this is an acceptable schedule.
t Current Status GL 92-01, Supplement 1 has been issued. NRC/ industry workshop has been completed. A request for research on RPV integrity issues has been issued. The Reactor Vessel Integrity Database (RVID) has been issued (NRC Administrative Letter 95-03) to all licensees and to r
all individuals requesting a copy. The staff has completed the review of licensees' initial responses to Supplement 1 to GL 92-01. The licensee for Kewaunee in a letter from Clark R. Steinhar.it dated August 21,1995 provided the only notable response. They provided three methods of 7
analysis of their surveillance data that indicate the Kewaunee reactor vessel will be below the PTS screening criteria at the expiration of its license. The licensee for Ginna in a letter from dated October 11,1995 has also submitted a revised PTS evaluation. The Kewaunee PTS evaluation is l
being reviewed by the staff. The staff has completed the review of the Ginna PTS evaluation, which is documented in a March 22,1996 letter to the licensee. Based on the currently available chemistry and surveillance data, the Ginna reactor vessel is projected to be below the PTS screening criteria at the expiration of its license.
The industry is currently conducting the Marble Hill annealing demonstration prciect and will conduct another annealing demonstration at Midland later this year. The demonstration at the Marble Hill site used a gas-fired heating method and was completed in July 1996. Preliminary evaluation of tne Marble Hill anneal indicates that temperatures, stresses and displacements were maintained within the expected ranges and that there were no measurable permanent dimensional i
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changes to the RPV nor damage to key components from the anneal. A detailed report on the Marble Hill anneal is expected by October 1996. The second demonstration at the Midland site will employ an electric resistance heating approach and is tentatively scheduled for November 1996.
lasued Reactor Pressure Vessel Integrity Database (RVID) Revision 1 on World Wide Web (WWW) on June 26,1996. Issued Administrative Letter that describes changes to RVID and informs licensees that RVID is on the WWW.
This action plan will be closed out. All tasks will be scheduled as part of individual plant reviews or as multi-plant action items.
NRR Technical
Contact:
Berry J. Elliot, EMCB, 415-2709 NRR Lead PM:
Daniel G. Mcdonald, PD1-1, 415-1408 Marsha K. Gamberoni, PD3-1,415-3024 References-Memorandum to Ashok C. Thadani from Jack R. Strosnider, " Plan for Addressing Generic Reactor Pressure Vesselissues," August 9,1995.
NUREG-1511, " Reactor Vessel Status Report," December,1994.
Generic Letter 92-01, Revision 1, (and Supplement 1) March 6,1992 and May 19,1995.
Memorandum to Ashok C. Thadani from Jack R. Strosnider, " Assessment of impact of increased Variability in Chemistry of the RTm Value of PWR Reactor Vessels," May 5,1995.
NRC Administrative Letter 95-03, August 4,1995
" Temper Embrittlement, Irradiation induced Phosphorus Segregation and Implications for Post irradiation Annealing of Reactor Pressure Vessels," R.J. McElroy et. al.
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MOTOR-OPERATED VALVES ACTION PLAN TAC Nos.
M80330, M82J72, Last Upciate: 9/30/96 M75089, M88898 Lead NRR Division: DE GSI: II.E.6.1 -
MILESTONES DATE (T/C) r Regulatory improvements:
1/96-9/96 (T)
(1) Staff is wortung with ASME to improve the inservice testing requirements in the ASME Code and (2) Staff is working with OM to develop guidelines for penodic verification of MOV design-basis capability to replace stroke-time testing.
New Generic Letter on MOV Periodic Verification:
Staff preparing generic letter to provide recommendations on i
the periodic verification of MOV design-basis capability.
Issue for public commeat 2/96 (C)
Final issuance 9/96 (C )
MOV inspection Module: the staff will prepare an inspection 10/96 (T) module for inspecting MOV programs over the long-term and provide appropriate training for inspectors.
Review of EPRI MOV Performance Prediction Program: NRR and RES are currently reviewing a topical report submitted by NEl on the EPRI MOV Performance Prediction Program.
I SER 2/96 (C)
SER SUPPLEMENT 10/96 (T)
I Descnotion: Appendices A and B to 10 CFR Part 50 and 10 CFR 50.55(a) require nuclear power plant licensees to establish programs to ensure that structures, systems, and components important to the safe operation of the plant are designed, installed, tested, operated, and maintained in a manner that provides assurance of their ability to perform their safety functions.
GL 8910 and its supplements, asked licensees to help ensure the capability of MOVs in safety-related systems by reviewing MOV design bases, verifying MOV switch settings initially and periodically, testing MOVs under design-basis conditions where practicable, improving evaluations of MOV failures and necessary corrective action, and looking for trends in MOV problems. EMEB has programmatic oversight responsibility of regional inspection activities conducted to verify that licensee MOV programs are being implemented. EMEB provides support to the regions, either by staff or contractor expertise, for the conduct of inspections in this area and closure of licensee actions pursuant to GL 89-10.
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Historical Backaround: In 1985, the Davis-Besse nuclear power plant experienced a total loss of i
feedwater when, following a loss of main feedwater, safety-related MOVs in the auxiliary foodwater system could not be reopened after their inadvertent closure. As a result of this and other information, the NRC staff issued Bulletin 85-03 (November 15,1985) requesting that licensees verify the design-basis capability of safety-related MOVs used in high pressure systems.
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The information from the implementation of Bulletin 85-03, additional operating events, and NRC-sponsored research indicated the need to expand the scope cf Bulletin 85-03 to all safety-related systems.
i in Generic Letter (GL) 8910 (June 28,1989) and its supplements, the NRC staff asked licensees l
to help ensure the capabelity of MOVs in safety-related systems by reviewing MOV design bases, verifying MOV switch settmos initially and penodically, testing MOVs under design-basis conditions where practicable, improving evaluations of MOV failures and implementing necessary corrective action, and loolung for trends in MOV problems. The NRC staff requested that licensees complete the verification of he design-basis capability of MOVs included in the scope of GL 8910 withm three refusimg outages or five years from the date of issuance of the generic letter, whichever was later. The NRC starf has issued seven supplements to GL 8910 that provide additional guidance and information on GL 8910 program scope, design-basis reviews, switch settings, testing, periodic verification, trendme, and schedule extensions.
In June 1990, the NRC staff issued NUREG-1352, " Action Plans for Motor-Operated Valves and Check Valves," describing actions to organize the activities aimed at resolving the concoms about i
the performance of MOVs and check valves. These actions included evaluating the current i
regulatory requirements and guidance for MOVs, preparing guidance for and coordinating NRC inspections, completing NRC MOV research programs and implementing the research results, and providing the nuclear industry with information on MOVs.
Pronosed Actions: Specific activities included in the generic action plan to improve MOV performance are:
(1) Regulatory improvements - The staff is working with ASME to improve the inservice testing requirements in the ASME Code and the staff is working with OM to develop guidelines for periodic verification of MOV design-basis capability to replace stroke-time testing. Recently, ASME issued Code Case OMN-1, " Alternative Rules for Preservice and inservice Testing of Certain Electric Motor Operated Valve Assemblies in LWR Power Plants OM - Code - 1995 Edition; Subsection ISTC."
The staff references the code case in recently issued Generic Letter 96-05.
(2) EPRI MOV Performance Prediction Program - On March 15,1996, the staff issued the Safety Evaluation on the topical report on EPRI MOV Performance Prediction Program. The staff is reviewing the hand-calculation models for two unique gate valve designs.
(3) MOV Periodic Verification Generic Letter - The staff prepared a generic letter to provide recommendations on the periodic verification of MOV design-basis capability. On September 18, 1996, the staff issued GL 96-05, "Pariodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves."
(4) MOV inspection Module - The staff plans to prepare an inspection module for inspecting MOV programs over the long-term and provide appropriate training for inspectcrs.
l Oriainatino Document: NRC Bulletin 85-03 issued November 15,1985.
Reaulatory Assessment While it is important for the licensee to take steps to ensure that MOVs l
will operate reliably under design-basis conditions, the probability of any individual MOV failure is l
small and safety systems are robust enough to provide reasonable assurance of public health and
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safety.
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Current Status:
Coordination with industry and support to NRC regional staff, efforts on codes and standards, and MOV research and analysis are ongoing activities.
On September 18,1996, i
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the staff issued GL 96-05, " Periodic Verification of Design Basis Capability of Safety-Related Motor-Operated Valves."
On March 15,1996, the staff issued a non-proprietary Safety Evaluation on the EPRI MOV Performance Prediction Program. The staff is reviewing the remaining EPRI models for two unique gate valve designs and plans to issue a supplement to the SE addressing these two models later in 1996. The staff has been alerting licensees, NEl and EPRI to the staff's findings from the EPRI program review, and has been communicating staff views with industry regarding periodic verification. On August 21,1996, the staff issued information Notice 96-48 to alert licensees to lessons learned from the EPRI MOV program. In addition, the staff has been factoring the overall findings from the EPRI program into staff activities.
With the issuance of Generic Letter 96-05, EMEB is working to complete the supplement to the SE on the EPRI MOV Topical Report before closing the MOV Action Plan and will complete the remaining tasks as part of the implementation phase of the generic letter.
NRR Technical
Contact:
Thomas G. Scarbrough, EMEB, 415-2794 NRR Lead PM:
A:len G. Hansen, Df1PW, 415-1390 References-Bulletin 85-03, November 15,1985 Generic Letter 89-10, June 28,1989, and 7 supplements NUREG-1352, " Action Plans for Motor-Operated Valves and Check Valves," June,1990 Generic Letter 96-05, September 18,1996.
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UPDATE OF SRP CHAPTER 7 TO INCORPORATE DIGITAL INSTRUMENTATION AND CONTROLS (l&C) GUIDANCE TAC Nos.
M86387, M86392, M86423, Last Update: 9/30/96 M86769, M86997, and M87680 Lead NRR Division: DRCH GSI: Not Available MILESTONES DATE (T/C)
- 1. Develop Update of SRP Chapter 7 10/96T
- 2. ACRS Subcomtmttes Briefings 3/96C, 5/96C, h
10/96T
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- 4. Incorporate results from National Academy of Sciences 02/97T study l
- 5. Draft SRP to Chairman 9/19/96C
- 6. Publish Draft SRP Chapter 7 for Public Comment 12/96T
- 7. Incorporate Public Comments 3/97T l
- 8. Final ACRS/CRGR Review of SRP Chapter 7 4/97T
- 9. Final SRP to Chairman 3/31/97T 10.
Publish Final SRP Chapter 7 5/97T Descrintion: This task action plan is used to track and manage the final phase of codifying the l
digital 1&C regulatory approach and criteria by updating the existing Standard Review Plan (SRP) i i
Chapter 7.
l tiistor6 cal 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 digital instrumentation 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 reaty for use when reviewing licensee proposed installation of digital instrumentation 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.
Proposed 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 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 revisw of the l
SRP Chapter 7 update, and (8) to publish the final revised SRP Chapter 7.
i 11
Oriainatino Document: The memorandum from the EDO to Chairman Jackson dated January 3, 1996, *1mprovements Associated with Managing the Utilization of Probabilistic Risk assessment (PRA) and Digital Instrumentation and Control Technology."
Reculatory Assessment The approach and criteria that form the current regulatory framework for review and acceptance of digital l&C systems in nuclear power plants is being codified in the update to SRP Chapter 7. This framework has boon communicated to the industry and public in safety evaluations for digital modifications to operating plants and design :artification of the advanced reactor designs, and in Generic Letter 95-02, "Use of NUMARCMPRI 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 26,1995. 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.
Current Status: The staff and its contractor, Lawrence Livermore National Laboratories (LLNL), are currently revising the seven existing sections of SRP Chapter 7 and daveloping two new sectionr.
and several new branch technical positions (BTPs) to incorporate criteria and guidance related to digital l&C systerr.s. In parallel, the Office of Nuclear Regulatory Research (RES) is developing several regulatory guides that endorse national standards related to digital l&C.
By the letter dated June 6,1996, the ACRS stated their agreement with the staff approach to the update of SRP Chapter 7, and their plan to continue to interact with the staff on the remaining changes to SRP Chapter 7. By memorandum dated September 16,1996, NRR requested CRGR review of the complete draft SRP Chapter 7.
The complete SRP Chapter 7 update is to be presented to the ACRS in October 1996. Following the review of the complete draft SRP Chapter 7 by CRGR and ACRS, the updated draft SRP Chapter 7 will be published in the Federal Register to provide an opportunity for public comment.
Contacts:
Matthew Chiramal, DRCH, 415-2845 Joe Joyce, DRCH, 415-2842 12 J
PRA IMPLEMENTATION ACTION PLAN 1.2(d)
Graded Quality Assurance Action Plan TAC Nos. M91429, M91431, M92420, Last Update: 9/30/96 M92450, M92451, M92447, M92448, Lead NRR Division: DRCH M92449, M88650, M91431, M91432, Support Division: DSSA M91433, M91434, M91435, M91436, M91437 GSI: Not Available MILESTONES DATE (T/C) 03/95C
- 1. Issued SECY 95-059 05/95C
- 2. Begin interactions with volunteer licensees
- Palo Verde letter dated 4/6/95
- Grand Gulf meeting 5/4/95
- South Texas meetings on 4/19/95 and 5/8/95 As Needed
- 3. NRC Steering Group meetings to guide working level staff activities Meetings on: 8/25/95, 10/10/95,10/25/95
- 4. Staff interactions with Palo Verde Ongoing through Site visit on 5/23/95 on ranking and QA controls
- NRC letter dated 7/24/95 on proposed QA controls 6/97 1
- Site visit on 8/29-30/95 on risk ranking Site visit on 9/6-7/95 on procurement GA controls
- NRC letter conveying trip reports issued on 12/4/95
- Meeting on 4/11/96 to discuss th6 staff evafwtion 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 OA controls, trip report sent to licensee on 8/6/96
- Letter from licensee on 9/12/96 transmitting responses to issues raised in earlier staff trip reports Ongoing
- 5. Staff interactions with South Texas through
- Meeting on 7/17/95 on project status
- Site meeting on 10/3-4/95 on risk ranking and QA controls 6/97
- Meeting on 12/7-8/95 to discuss risk ranking and QA controls
- South Texas Submittal of QA Plan for implementation of graded QA, dated 3/28/96 is currently under staff review
- Meetings on 4/11/96 and 4/25/96 to discuss the staff evaluation guide and future interaction milestones and schedules
- Letter from licensee on 4/17/96 providing comments on staff evaluation guidance
- Meeting on 6/19/96 to discuss staff comments on the OA plan submittal for graded QA, review questions transmitted to STP on 8/16/96
- Site visit on August 21-22 to observe working group and expert panel meetings, and to discuss staff review items, trip report in preparation
- SECY paper to be prepared prior to staff approval of OA program change only if policy positions are taken in the staff approval 13
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- 6. Staff interactions with Grand Gulf Ongomo
- Site meeting on 7/11-14/95 to observe expert panel through
- Meeting at hdot. on 10/24/95 on OA controls 6/97 l
- Meeting at RIV on 11/16/95 on graded QA effort
- Site meetmg on 11/17/95 to observe expert panel 1
- GGNS system and component ranking criteria under staff evaluation, the comments are scheduled to be provided to GGNS by the end of June
- Meetmo on 4/11/96 to discuss the staff evaluation guide r
- Letter to GGNS dated 5/29/96 regarding implementation of OAP i
commitments
- Staff review comments on GGNS safety asonificance determination l
process transmitted to licensee on July 15 i
- Meeting on August 27 to discuss staff comments on safety significance process and to discuss GGNS implementation of OAP l
commitments for low-safety significant items, meeting summary in l
preparation
- Tentative plans to visit site in t 1/96 to review procurement activities
- 7. Revision 3 of Draft Evaluation Guide for Volunteer Plants issued for staff 07/95C comment l
- 8. Revision 4 of Draft Evaluation Guide for Volunteer Plants issued for 10/95C Steenne Group Review l
- 9. lasue letter to 3 volunteer plants outlining program objectives and j
review expectations. Distribute staff evaluation guide to licensees.
1/96C j
- 10. Evaluation Guide issued for use by staff in evaluating volunteer plants 1/96C i
- Meeting held with volunteer plants to receive feedback on staff evaluation guide on 4/11/96.
4/96C
- Industry comments on staff evaluation guide provided by letter dated 5/24/96
- The staff will review the industry comments with respect to the need to revise, and finalize, the evaluation guide by the end of October.
Meetmg of GOA steenne group will be scheduled to discuss finalization of staff evaluation guide for volunteer implementation phase 11.
Regulatory Guide and SRP development milestones per PPA Action Plan
- Draft SRP and RG for Branch / division review and comment 7/31/96C for RG
)
- Draft SRP and RG for inter-office review and concurrence 9/30/96 (SRP) T
- Draft SRP and RG for ACRS/CRGR review 8/1/96 (RG) C
- Draft SRP and RG for public comment 11/1/96T
- Draft SRP and RG public comment period ends 12/31/96T
- Final draft SRP and RG for ACRS/CRGR review 3/3/97T
- Final draft SRP and RG for inter-office concurrence 9/1/97T
- Publish final SRP and RG 12/1/97T 12/31/97T
- 12. ACRS Briefings
- Expert Panel and deterministic considerations 2/27-28/96C
- graded QA 4/11/96C
- PRA implementation Plan and pilot projects 7/18/96C
- Risk informed Pilots 8/7/96C i
14
__.- - - - -. -. - -. _-_ ~.
- 15. Public Workshop on Graded QA 2/98T
- 16. Issue Staff Inspection Guidance (Reactive IP) 12/97T i
- 17. Conduct NRC Staff Training 1/98T
- 18. Issue SECY Update (close-out of action plan) 4/98T Descnotion* Prepare staff evaluation guidance and regulatory guidance for industry implementation I
for the grading of qualty assurance (QA) practices commensurate with the safety significance of
[
the plant equipment. The development of this guidance will be based on staff reviews of regulatory requirements, proposed changes to existing practices, staff development of a draft regulatory guide with input from a national laboratory, and assessment of the actual programs developed by the three volunteer utilities implementing graded quality assurance programs.
Histoncal Backoround: The NRC's regulations (10 CFR Part 50, Appendices A & B) require QA programs that are commensurate (or consistent) with the importance to safety of the functions to be performed. However, the QA implementation practices that have evolved have often not been l
graded. In the development of implementation guidance for the maintenance rule, a methodology to determene the risk significance of plant equipment was proposed by the industry (NUMARC 93-01). During a publec meeting on December 16,1993 the staff suggested that the indusuy could build on the experience gained from the maintenance rule to develop implementation methodologies for graded QA. The staff had numerous interactions with the Nuclear Energy Institute (NEI) during calendar year 1994 as the graded QA concepts were discussed and the initial industry guidelines were developed and commented on. In early 1995, three licensees (Grand Gulf, South Texas, and Palo Verde) volunteered to work with the staff. The staff has reviewed the licensee developmental graded QA efforts.
Pronosed Actions: The goal of the action plan is to utilize the lessons learned from the 3 volunteer-licensees to modify staff-developed draft guidance to formulate regulatory guidance on acceptable methods for implementing graded QA. The staff will develop a regulatory guide based in part on input from Brookhaven National Laboratory, a standard review plan revision for Chapter 17, and a reactive inspection procedure (IP) for graded QA. An inter-office team has been established to prepare the regulatory guidance documents and test their implementation during the evaluation of volunteer plant activities.
Onoinatina Document-Letter from J. Sniezek, NRC to J. Colvin (NUMARC) dated January 6, 1994, describing the establishment of NRC steering group for the graded QA initiative.
i Raoulatory Assessment: Existing regulations provide the necessary flexibility for the development 1
and implementation of graded quality assurance programs. The staff will issue a NUREG report regarding the lessons learned from the volunteer plant implementations, Additional regulatory guidance will be issued to either disseminate staff guidance or endorse an industry approach.
Planned guidance for the staff will involve an evaluation guide for app;ication to the volunteer plants, the lessons learned report, training sessions and public workshops, Standard Review Plan revision, and inspection guidance in the form of a reactive IP. The staff is evaluating the appropriate mechanism for inspections of the risk significance determination aspects of graded QA programs.
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The safety benefits to be gained from a graded QA program could be significant since both NRC
)
reviews and inspectsons and the industry's quality controls resources would be focused on the more safety significant plant equipment and activities. Secondarily, cost savings to the industry 1
)
could be realized by avoiding the dilution of resources expended on less safety significant issues.
The time frame to complete this action plan is directiy 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 qualsty assurance programs. The staff will utilize the guide i
i for the review of the volunteer plant graded QA programs. The guide and the staff's proposed interaction framework has been transtnetted in a letter to the three volunteer licensees. The letter j
j aseks licensee comments. Outlines of a draft regulatory guide and SRP for both risk ranking and 1
grading of OA controls have been prepared and circulated for review for the inter-office team. A i
meeting was held with the three volunteer licensees on April 11,1996 to receive their feedback on the staff developed evaluation guide. The licensees expressed concems about the level of detail y
contained in the puede, particularly that related to PRA and commercial grade item dedication. The i
licensees contend that exiting industry guidance (PSA Application Guide and EPRl-5652) are sufficient for those topics. The staff received written comments from NEl on the evaluation guide by letter dated May 24,1996. The NEl letter questions the need for additional regulatory guidance for the graded QA application. NEl contends that existing industry guidance is sufficient. STP and PVNGS letters provideng comments on the evaluation guede were dated April 17,1996 and April 24,1996 respectively. The staff will compelo suggested changes to the evaluation guide in response to the industry comments and a meeting will be held to brief the graded QA steering group on the proposed changes. In addition, a presentation on graded QA was made to the full ACRS on April 11th. During the ACRS meeting some questions arose with respect to the staff expectations for the conduct of expert panel activities.
i South Texas submitted their QA program revision for their graded QA effort on March 28,1996.
i The change has been reviewed by the staff (HOMB, SPSB, RES, RIV, and NRC contractors). A j
meeting was held with STP on June 19 to discuss the staff's comments and concoms. STP indecated their willingness to re-examine the content of the OA 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 segnificance categorization and adjustment of 0.A controls. The staff also observed both a Wortung 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 will resume when STP submits a revision to the OA program to address staff concoms.
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 i
effort. The staff informed NEl by letter dated May 2,1996 that the guide is not adequate (as a stand alone document) to implement graded QA but that it will be considered as the staff develops the graded QA regulatory guide and standard review plan. By letter dated June 8, NElindicated that their 96-02 guide will be revised. Further NEl requested a meeting with the staff (in the August time frame) to discuss the changes and to discuss more objective means to assess the adequacy of QA program implementation. NEl has proposed that the amended 96-02 guidelines will be submitted to the staff for endorsement by a regulatory guide. A subsequent letter was
((
received from NEl on July 16 that provided an updated version of NEl 96-02 based on comments they received from the volunteer plants and industry sources. The staff will review the modified I
document and then brief the steering group on the results.
I 16 i
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Contact:
S. Black 415-1017, l
R. Gramm 4151010 l
Contact:
R. Woods 415-6622
References:
- 1) Letter from J. Sniezek (NRC) to J. Colvin (NEI) dated 1/6/94
- 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. Holian (NRC) to W. Stewart (APSCo) dated 7/24/95
- 6) Letter from C. Thomas (NRC) to W. Stewart (APSCo) dated 12/4/95 l
- 7) Memorandum from S. Black to W. Beckner and W. Bateman dated 1/24/96, Draft Staff l
Evaluation Guidance
- 8) NEl 96-02, " Guideline for implementing a Graded Approach to Quality' i
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NEW SOURCE TERM FOR OPERATING REAL, TORS TAC No. M89586 Last Update: 09/30/96 GSI No.155.1 Lead NRR Division: DRPM Supporting Division: DSSA & DE MILESTONES DATE (T/C)
- 1. NEl Letter 7/94C
- 2. Commission Memo 9/94C
- 3. NEl Response 9/94C
- 4. NEl/NRC Meeting 10/94C
- 5. Publication of NUREG 1465 2/95C
- 6. NEl/NRC Meetings 10/94C, 6/95C,10/95C, 1/96C
- 7. Submittal of Generic Framework Document (from NEl) 11/95C
- 8. First Pilot Plant Submittal 12/95C
- 9. Issue Memo to Commission, Updating Status 8/96C i
10.
Present Commission Paper in E-Team Briefing 9/96C 11.
Brief CRGR on Commission Paper 10/96T 12.
Send Commesseon Paper to EDO/ Commission 10/96T 13.
Brief ACRS on Commission Paper 11/96T 14.
Response to NEl Framework Document 11/96T 15.
Bogen Pilot Plant Reviews 11/96T
- a. lasue RAls on Pilot Plants 1/97T
- b. Response to RAls from Pilot Plants 4/97T
- c. Provide Pilot Plant SERs to Projects 6/97T
- d. Prepare Exemption Package for 7/97T Pilot Plants Descnotion. More than a decade of research has led to an enhanced understanding of the timing, magnitude and chemecal form of fission product releases following nuclear accidents. The results of this work has been summarized in NUREG 1465 and in a number of related research reports.
Application of this new knowledge to operating reactors could result in cost savings without sacrificing real safety margin. In addition, safety enhancements may also be achieved.
Histoncal Backaround: In 1962, the U. S. Atomic Energy Commission published TID-14844,
" Calculation of Distance Factors for Power and Test Reactors." Since then licensees and the NRC have used the accident source term presented in TID-14844 in the evaluation of the dose consequences of design basis accidents (DBA).
After examining years of additional research and operating reactor experience, NRC published NUREG 1465, " Accident Source Terms for Light Water Nuclear Power Plants," in February 1995.
18
i The NUREG descrioes 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 TlO source term; however, it includes a more realistic description of release timing and composition. Smce the NUREG source term results in lower calculated DBA dose consequences, NRC decided not to require current plants to revise their DBA analyses using the new source term.
However, many licensees want to use the new source term to perform DBA dose evaluations in support of plant, technical specification, and prnedure 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 oncouraged the industry to approach the issue on a generic basis. The Nuclear Energy Institute (NEl) unveiled its plans for the use of the new source term at oparating plants at the Regulatory Information Conference in May 1995. NEl, Polestar (EPRI's consultant), and pilot plant (Grand Gulf, Millstone, Beaver Valley, Browns Ferry, Perry, and Indian Point) representatives met with NRC staff on June 1 and October 12,1995, to discuss more detailed plans.
Pronosed Actens: The staff has reviewed the framework document and is preparing a draft Commession paper and decision letter that describes a generic implementation approach. The staff presented the Commission paper and decision letter to the NRR Executive Team in September, and plans to brief CRGR and then send the Commission paper and decision letter to the Commission and ACRS by November 1996 (SRM M960612). As described in the Commission paper, the plan is to review each pilot plant application and prepare an exemption package addressing the use of each feature of the NUREG-1465 source term while pursuing rulemaking. The current schedule is to prepare the first exemption after the decision letter is issued to industry and the staff completes a technical review of the first pilot plant submittal. The plan for issuing each remaining exemption is to brief the CRGR, issue for public comment, and then issue the exemption.
Orininatina Document-EPRI Technical Report TR 105909, " Generic Framework Document for Applecation of Revised Accident Source Term to Operating Plants," transmitted by letter dated November 15,1995.
Raoulatory Assessment: There will be no mandatory backfit of the new source term for operating reactors. The design-basis accident analyses for current reactors based on the TID 14844 source term are still valid. Therefore, non-urgent regulatory action and continued facility operation are justified.
Current S14 hit: 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 and Northeast Utilities System submitted its pilot plant application for Millstone in April 1996. 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 Browns Ferry, Perry, and Grand Gulf pilot plant submittals, respectively. The staff has completed its review of the framework document and is preparing a Commission paper describing how it intends to conduct its generic review of pilot plant submittals. The NRR Executive Team was briefed on the Commission paper in September, and briefings of CRGR and ACRS are being scheduled. A limited number of pilot plants submittals and exemptions are expected, but only two have been received so far (Browns Ferry and Millstone). Remaining pilot plant submittals are expected before the end of 1996. On a related issue, as a result of the June 12,1996, Commission briefing on the Reactor Site Criteria rulemaking 110 CFR 100 (Seismic and Non-seismic Provisions)L the staff was directed to seek guidance from the Commission regarding the application of the new source term to l
operating reactors. Consequently, the staff has prepared a memorandum to the Commission, i
which was issued on August 9,1996, that summarizes the current status and future actions.
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... -. - _.- __ - - -...- ~ ~
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NRR Technical Contacts:
R. Emch, PERS, 415-1068 A. Huffert, PERS, 415 1081
)
J. H. Wilson, PGEB, 415-1108
References:
NUREG 1465, " Accident Source Term for Light Water Nuclear Power Plants," February,1995.
i July 27,1994, letter to A. Marion, NEl, from D. Crutchfield, NRC, " Application of New Source l
Term to Operating Reactors".
j t
September 6,1994, letter to the Commission from NRC staff, "Use of NUREG-1465 Source Term i
4 j
at Operating Reactors".
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July 21,1995, letter to the Commission from NRC staff, "Use of NUREG 1465 Source Term at Operating Reactors".
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December 22,1995, pilot plant submittal, letter to Document Control Desk from Tennessee Valley Authonty, " 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...".
Summaries of pubhc meetings:
e dated November 10,1994 for public meeting with NEi held on October 6,1994; e dated July 26,1994 for public meeting with NEl held on June 1,1995; e dated November 17,1995 for public meeting with NEl held on October 12,1995.
dated February 1,1996 for public meeting with NEl held on January 23,1996.
e b
20
ENDANGERED SPECIES ACTION PLAN TAC No. M88282 Last Update: 09/30/96 GSI: El-184 Lead NRR Division: DRPM l
MILESTONE DATE f
- 1. Development of action plan.
06/95C
- 2. Develop list of currently listed protected species in the vicinity of each 11/95C nuclear power plant site l
- 3. Identify individual licensee programs and activities being condur:ted to 05/96C l
further the conservation of protected species.
- 4. Determene priority for sites warranting follow-up actions.
10/96T l
- 5. Recommend site-specific follow-up actions to Projects.
11/96T
- 6. Development and implementation of process for maintaining status and 03/97T compliance with the ESA at each site.
Descnotson Develop a list of currently listed protected species in the vicinity of each nuclear l
power plant site, identify individual licensee programs and activities being conducted to further the conservation of protected species, and conduct informal or formal consultation with either the National Marine Fisheries Service or the Fish and Wildlife Service, as warranted for any specific site.
Historical Backaround In 1973, Congress passed the Endangered Species Act for the protection of endangered or threatened species. In responding to a Commission memorandum of July 30,1991, i
concerning efforts of the Commission, applicants, and licensees for protection of endangered species in the vicinity of nuclear power facilities, it was identified that the NRC may not have l
completed all the necessary activities required by the Endangered Species Act for some of the facilities that have identified endangered species. This action plan will determine the additional actions, if any, that need to be taken at individual sites so that the NRC can meet its obligations under the act.
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Proposed Actions: Conduct evaluations of plant-specific lists of endangered species and existing licensee commitments to further the conservation of the protected species and determine if informal or formal consultation with either the National Marine Fisheries Service or the Fish and i
Wildlife Service is warranted.
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Oriainatina Document Commission Memorandum of July 30,1991.
Reaulatorv Assessment Continued facility operation is appropriate because this action plan does not involve a health and safety issue.
l Current Status: A list of currently listed protected species in the vicinity of each nuclear power i
plant site was developed by a contractor and delivered to NRC on 4/25/96. This report is now under review. This action plan will be evaluated with respect to separating generic and plant-l specific aspects as part of Milestone 4. The contractor will provide a final report in October that l
prioritizes sites and makes recommendations for follow-up actions.
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1
Contacts:
NRR Technical Contacts:
Mike Masnik, PDND, 415-1191 Jim Wilson, PGEB, 415 1108 NRR Lead PM:
Jim Wilson, PGEB, 415-1108 References-Commission Memorandum of July 30,1991.
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EFFECT OF HURRICANE ANDREW ON TURKEY POINT TAC No. M86716/17 Last Update: 9/30/96 GSI: Ll-178 Lead NRR Division: DRPM Mll.ESTONES DA TE (T/C)
- 1. Evaluate the Adequacy of 1.icensee Offsite Communications for 11/96T Natural Disasters Within the Plant Design Basis.
Collect information on licensee communication capabilities and 6/96C vulnerabilities via region inspection.
Analyze inspection findings and report on results.
8/96C l
Established schedule for issuance of Information Notice.
10/96T f
- 2. Evaluate the Adequacy of NRC quidance for Reviewing t.icensee 5/96C Preparation and Response to Natural Disasters and Industry Proplanned Support.
Descrintion: This action plan was developed to address the actions necessary to resolve the issues identified in the " Report on the Effect of Hurricane Andrew on the Turkey Point Nuclear Generating Station from August 20-30,1992." Two of the issues are still being considered. They are:
- 1) Whether there is a need for generic guidance to licensees to ensure that their offsite communication circuits can reliably survive or recover from the impact of a severe natural event such as a hurricane. These circuits are required to provide reliable notification to offsite authorities of emergency conditions at the licensee's power reactor facility.
l
- 2) Whether there is a need for generic guidarwe to inspectors to review licensees' preparation for and response to natural disasters, including industry proplanned support.
Histoncal Backnround On August 24,1992, Category 4 Hurricane Andrew hit south Florida and caused extensive onsite and offsite damage at Turkey Point. An NRC/ industry team was organized to review the damage that the hurricane caused the nuclear units and the utility actions to prepare for the storm and recover from it, and to compile lessons that might benefit other nuclear reactor facilities. Results of the team review are presented in the report, " Report on the Effect of Hurricane Andrew on the Turkey Point Nuclear Generating Station From August 20-30,1992,*
issued in March 1993. This report was distributed to all power reactor licensees by the Institute of Nuclear Power Operations on June 10,1993.
The EDO requested a review of the NRC/ industry report to determine the actions necessary for resolving the issues identified in the report. An action plan was established on July 22,1993, to perform this function. Annual written status reports are provided until allitems are closed. The October,1995 report contained two open items, listed above.
A temporary instruction (Tl 2515/131), issued 1/18/96, incorporating Regional comments, was written to provide Regional inspectors guidance for collecting information on offsite notification circuits. The Tl was performed at seventeen plants from February 1,1996 to June 30,1996. The I
results from these seventeen plants, along with data gathered on several plants prior to the TI issuance, was analyzed.
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Prooosed Actions: For item 1) a6ve, an Information Notice is proposed for distribution to all 4
reactor licensees discussing th' rasults of the Temporary Instruction used to collect information about these circuits. A draf*, of che IN will be shared with the Federal Emergency Management Agency (FEMA) since these communication circuits are used by offsite emergency management agencies, and FEMA is responsible for oversight of the radiological emercency preparedness of these agencies.
Raoulatory Assessmang: Justification for non-urgent regulatory action: A qualitative ar'ety assessment of the technical issues being addressed for item 1) demonstrates thn tre significance of the issue is at a level that will allow both continued facility operation and traatment of the issue as a son-urgent regulatory action.
Current Status: For item 1) the results of the Tl were documented in a memorar'dum to the Director, NRR.
For item 2), the action to provide guidance for inspectors has been incorporated into the PRA implernentation Plan. On that basis, milestone 2 is considered closed.
Also, interim guidance was recently issued to Regional offices and NRR on coordination with FEMA following impact of natural disasters on power reactors and the areas surrounding them. This guidance ensures that offsite preparedness is re-affirmed before authorizing the restart of any power reactor that shuts down in anticipation or as a result of a natural disaster. This effort is related to this action plan because it originated in lessons learned from the Hurricane Andrew disaster.
NRR Technical Contacts:
W. Maier, PERB, 415-2926 NRR Lead PM:
R. Croteau, DRPE, 415-1475 24 t
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ENVIRONMENTAL SRP REVISION ACTION PLAN TAC No. M80177 Last Update: 10/02/96 GSI: Not Available Lead NRR Division: DRPM MILESTONES DATE (T/C)
- 1. Reflect Potential impacts and integrated impacts in Options for Resolution
- a. Identification of potentialimpacts 03/96C
- b. Identification of integrated impacts 06/96C
- c. Proposed options for resolution and develop initial draft of 10/96T revised ESRP
- d. Staff / contractor meeting to resolve format and content of 12/96T revised ESRP
- 2. Prepare Final Draft of ESRP Sections for Public Comment
- a. Draft updated ESRP for staff review 02/97T
- c. Pubhsh (electronic) for public comment 08/97T
- 3. Disposition Public Comments 01/98T
- 4. Publish Final NUREG-1555 08/98T
- 5. Maintenance of program data Ongoing Bnef Desenntion The Environmental Standard Review Plan (ESRP) Revision Action Plan deals with f
the revision to NUREG-0555 to reflect changes in the statutory and regulatory arena, to incorporate emerging environmental protection issues (e.g., SAMDA and environmental justice) since originally published in 1979, and to support the review of license renewal applications. The l
ESRP will take the form of the SRP (including acceptance criteria) and follows the rame update criteria outhned under the SRP-UDP project (with the exception of maintaining the MD8 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. Initial interactions on 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 August 1997.
Disposition of public comments and staff review of the update (NUREG-1555) leads to a publication date of August 1998.
Reaulatory Assessment: NRR has established the ESRP Update Program for use in the life cycle review of environmental protection issues for nuclear power plants, especially license renewal applications, but also operating reactors, and future reactor site approval applications. The ESRP will reflect current NRC requirements and guidance, consider other statutory and regulatory requirements (e.g., the National Environmental Policy Act, Presidential Executive Orders), and incorporate the generic environmental impact work and plant-specific requirements developed during amending of Part 51 for license renewal reviews.
i 25 l
Current Status: Two contracts are currently in place to support the ESRP Program, JCN J-2028 with Pacific Northwest National Laboratory (PNL) for overall coordination and most of the ESRP sections and JCN J-2039 with Lawrence Livermore National Laboratory (LLNL) for the seismology and geology sections. The work approach and detailed procedures rely heavily on the framework established for 6 SRP-UDP. The project team was established in 1994; resources were diverted twice to work off higher priority activities (i.e, the Watts Bar Environmental Statement Update and the RADTRAD project). Potential impacts were completed for all existing ESRP sections, integrated impacts were completed for most existing ESRP sections and work for new sections (i.e., Environmental Justice, SAMDAs) is underway; Proposed Options for Resolution phase is underway with sectums to be delivered beginning in August 1996 leading to a working session on the options for each section of the document scheduled for November 1996.
NRR Techrucal
Contact:
B. Zalcman, PGEB, 415-3467 I
t I
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i 1
10 CFR 50.59 ACTION PLAN 4
3 TAC No. M94269 Last Update: 9/30/96 i
Supporting Divisions: all 1
MILESTONES DATE (T/C) j l
4 1
- 1. Action plan approval / copy to Commission 04/15/96(Cl f
- 2. Identify work group members 05/24/96(C) 4 l
- 3. Brief D/NRR on issues N/A k
- 4. Conduct workshop 06/18/96(C) i
- 5. Brief D/NRR on proposed positions 07/24/96(C) 2 f
- 6. Draft position papers 08/29/96(C)
- 7. Obtain regional comments 09/30/96(C)
- 8. Policy issues and position paper to EDO with Lessons 10/96(T) i Learned Report l
- 9. Issue document for public comments 12/96(T) 9 10.
Obtain comments, inclujing ACRG 02/97(T) 11.
Revised positions and secommendations issued to 04/97(T) l NRC management 06/97(T) 4
- 12. Commission Paper 4
- 13. Followon Actions TBD l
Descriotion This action plan defines measures to improve licensee implementation and NRC staff i
oversight of the 10 CFR 50.59 process.
i l
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 j
conduct tests and experiments) without prior NRC approval when the change does not involve the j
1 Technical Specifications or an unreviewed safety question. Licensees are required to submit l
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 concoming 50.59 implementation and NRC oversight, and proposed a systematic reconsideration and reevaluation of the process.
i The December 15,1995, memorandum from the EDO responded to the specific questions and k
stated that within 120 days from the date of the memorandum, the staff would review previously
{
issued guidance on implementation of the 50.59 process to define areas where the guidance needs j
to be emended and to develop 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. The staff has completed j
its review of existing guidance and has identified certain issues for further examination, which this
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action plan addresses.
j!
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The staff has made the results of its review of guidance, the action plan, and its interim inspection guidance publecly avaelable.
Planned Actions-The staff's approach to development of regulatory guio, ice would proceed in phases. Over the next several months, the staff will attempt to provide spw. e posi ions (guidance) to accomplish t
the objectives listed below, and will evaluate the fearibility of implementing such guidance within the existing regulatory framework. At the end of thr2 first phase, estimated to take about eight months, the staff would take stock of its progress a 1d make recommendations on issuing guidance, undertaking rulemaking or other actions.
i 1
Specifically, the objectives of this effort are to develop guidance that would-define the elements of safety evaluation review or screening processes within the context of o
various licensee design or change control processes, to provide greater assurance that effects i
on safety of changes, whether to equipment, procod res, or methods of system operation, are appropriately evaluated.
1 define more specifically the scope of applicability of 50.59 (that is, to identify those changes, o
tests, or experiments) that need to be evaluated to determine if NRC approval is needed). This would include a more coniprehensive description of change, and guidance for broader consideration of "as described."
establish the process for resolving nonconforming conditions such that differences from the o
FSAR are reconciled (from both safety and regulatory viewpoints) in a tirra frame commensurate with their safety significance. This will also consider wt an such conditions should be evaluated under 50.59 as temporary modifications because resolution of the nonconforming condition has been delayed.
o improve USO determinations in the following respects:
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address the extent to which short and long term compensating actions may be considered as part of change under 50.59 so that it can be determined that the probability has not increased or margins of safety as defined in the basis for any technical specification has not been reduced. Also address when consideration of compensating actions should be reviewed as part of the basis for approving a proposed license amendment.
clarify the extent to which PRA techniques may be usefulin evaluating the effects on safety of a change, and in addressing the " probability may be increased" criterion for unreviewed safety questions.
clarify what is meant by " margin of safety" in relation to numerical parameters, analysis methods, calculated results of safety analyses, and licensing limits such that changes that might affect the basis for staff's safety conclusions with respect to Technical Specifications are more consistently identified.
Finally, as part of the development of possible guidance, consider whether additional definitions are needed, such as for malfunction of equipment important to safety, 28 k
3 Public comments on the position paper (s) will be obtained. The ACRS will be requested to provide
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its comments on these positions. Actions, milestones and schedules for further phases of this j
effort will be developed after the results of the first phase are assessed.
i in the area of staff oversight, the staff conducted, on June 18,1996, a roundtable discussion with j
regional staff, resident inspectors and NRR staff who have participated in 50.59 inspection efforts to share experiences and to discuss such topics as the mix of programmatic and implementation l
reviews, sampimg and team composition. Appropriate changes to inspection procedures will be l
made.
j Other related efforts are being tracked under other programs.
)
i Onoinatina Document:
j April 15,1996 memorandum from the EDO to Chairman Jackson,
Subject:
Action Plan for l
Improvements to 10 CFR 50.59 Implementation and Oversight.
4
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Ranulatorv Assessment The action plan was developed to identify actions to improve J
implementation of the 50.59 process. A number of improvements have been implemented in the t
last few months, such as directing inspectors conducting all routine inspections to specifically j
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 i
is currently no indication that implementation of 10 CFR 50.59, as it is carried out today, has led to 3
decreased safety, based on inspection experience. While improvements can be made to achieve a higher degree of uniformsty of review, the cur.ent process as it is being implemented provides reasonable assurance that plant safety has not been decreased." The above conclusion is j
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.
I Current Status-l A revision to the action plan was issued on August 20,1996, v4kh revised the scheduled milestones such that the Commission will have the opportunity w consider the policy issues associated with 50.59 along with other policy issues from tt-Wistone lessons learned review.
j The draft position paper, including policy issues, was issued for intomal review and comment on August 29,1996. A number of comments have been received and are being reviewed for j
incorporation into the paper. Schedule changes were made to conform with planned dates for i
submitting Part 2 of the Millstone Lessons-learned report to the EDO by the end of October,1996.
Schedule for submittal to the Commission has not yet been established.
}
NRR Technical Contar.t:
E. McKenna, PGEB, 415-2189 f
References:
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October 27,1995 memorandum from Chairman Jackson to EDO 1
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November 30,1995 memorandum from Chairman Jackson to EDO i
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December 15,1995 memorandum from EDO to Chairman Jackson December 28,1995 memorandum from EDO to Chairman Jackson
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April 15,1996, memorandum from EDO to Chairman Jackson l
August 20,1996, memorandum from EDO to Commission i
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i INDUSTRY DEREGULATION AND UTILITY RESTRUCTURING ACTION PLAN TAC Nos. M78003 Last Update: 09/30/96 Lead NRR Division: DRPM GSI: Not Available Supporting Division:
MILESTONES DATE (T/P/C)
Task 1 - Develop NRC Policy Statement and SRP 06/97T Draft Policy Statement 05/96C Office Concurrences 06/96C EDO Concurrence 06/96C Commission Paper 07/96C Draft SRP 07/96C Publish Draft Policy Statement 09/96C Office Concurrences on SRP 09/96C EDO Concurrence on SRP 09/96C Commission Paper on SRP 09/96C -
Publish Draft SRP 10/96T Pubiec Comment Policy Statement 10/96T Public Comment SRP 11/96T Final SRP/ Policy Statement 12/96T Office Concurrences 01/97T ACRS 02/97T CRGR 02/97T EDO Concurrence 03/97T Commission Approval 05/97T Publish Final SRP and Policy Statement 06/97T Task 2 - lasue Administrative Letter to Licensees on Financial 06/96C Reporting Requirements Draft 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 Reposting 2/97T Requirements as Necessa'y Determine Necessity fo Action 09/96C Draft Option 12/96T Office Concurrence 12/96T CRGR Review 01/97T EDO Concurrence 02/97T Publish Draft 02/97T 30
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Task 4 - Update prior NUREG documents on owners and financial 02/97T license conditions issue Task Order Cnntract 05/96C Draft NUREG Updated 09/96C Publish NUREGs 10/96T Public Comment 01/97T Revise and Publish Final 02/97T Task 5 - Institutionalize Staff Level Contact with NARUC,SEC,FERC.
10/96T Develop MOUs as necessary.
Letter to agencies 06/96C Staff level meeting.
10/96T Draft MOUs to Commission (as required)
TBD Sign MOUs TBD Task 6 - Develop and implement rulemaking to clarify 10 CFR 50.80 TBD if necessary Commission determination of need TBD Proposed ANPR or rulemaking package TBD Office Concurrences TBD i
I ACRS Comments TBD CRGR Concurrence TBD EDO Concurrence TBD l
Commission Approval TBD Publish ANPR or Proposed rule TBD Public Comment TBD Revise Ruiomaking Package TBD Office Con:urrences TBD i
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CRGR C'>ncurrence TBD EDO Concurrence TBD Cornmission Approval TBD Publish Final Rule TBD l
Task 7 - Assist Office of Research (RES) on Decommissioning ONGOING Funding Assurance Rule.
Milestones for this task provided by RES under rulemaking action,
" Decommissioning Costs and Funding Evaluations" l
Descriotion: The action plan is intended to address the Commission's concerns regarding the impact of utility deregulation and resulting reosganizations and restructuring on licensee's financial qualifications and their ultimate ability to safely operate and decommission their facilities.
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Historical Backaround: In recent years, several restructurings and reorganizations have occurred with the electnc utilety industry. 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 j
decommissoorung. 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.
Pronomad Actans: 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 all licensees delineating their current responsibilities with respect to getting prior NRC approval for changes that may affect their previous financial qualification determenations or ownership; 3) formulating non-rulemalung periodic reporting requirements,
- 4) updating NUREG documents containing financial information 5) establish 6ng staff level contacts with the Secudties and Exchange Commission (SEC), the Federal Energy Regulatory Commission (FERC), and the National Association of Utility Regulatory Commissions (NARUC); 6) implementing rulemaldng if necessary; and 7) assisting the Office of RES in their decomenissioning funding assurance rulemaking.
Current Status: PGEF has developed a draft policy statement, administrative letter, and liaison letters to FERC and." iC. Staff level contacts with NARUC have been identified. The administrative lettei was issued with a letter to the CEOs of alllicensees on June 21. A Commission Information Paper informed the Commission of our intentions for sending the Admin letter and CEO letter. A Commission Paper forwarding the draft policy statement was submitted on July 2,1996, as SECY 96-148. The Commission approved publication of the draft policy statement by SRM dated August 16,1996. The draft policy statement was sent to the Federal Register for publication on September 13,1966.
NRR Techn6 cal Contacts:
R. Wood, PGEB, 415-1255 M. Davis, PGEB, 415-1016 32
EXTENDED POWER UPRATE ACTION PLAN Tac No. M91571 Last Update: 10/06/96 Lead NRR Division: DRPW GSI: RI-182 Supporting Division: DSSA MILESTONES DATE (T/C) 1:
Receive GE Topical ELTR1 (Generic Review Methodology).
-3/95 C 2:
lasue Staff Position Paper on ELTR1 Meeting with GE/NSP.
4/95 C Identify differences between LTR1 and ELTRI.
8/95 C Issue RAls as appropriate.
9/95 C Incorporate information on foreign experience obtained from SRX8.
10/95 C Develop power uprate database for all U.S. plants.
lasue Staff Position Paper.
10/95 C 2/96 C 3:
Receive GE Topical ELTR2 (Generic Bounding Analyses).
GE plans to submit ELTR2 in two parts: the first part in March 96 3/96 C and the second part in July 1996.
7/96 C 4:
Meeting with GE/ industry.
2/96 C Issue RAls as appropriate.
10/96T
- Input to the SE from technical branches.
2/97 T lasue SE.
4/97 T 5:
Receive Lead Plant Application (Monticello).
7/96 C 6:
lasus Staff SE for Lead Plant.
Meeting with Monticello.
10/96 T RAls input from toch branches.
11/96 T Issue RAls as appropriate.
11/96 T Input to the SE from tech branches.
6/97 T ACRS Presentation 6/97 T Issue Secy Information Paper 6/97 T Is',ue SE.
6/97 T 7:
Develop a Standvd Review Procedure. Incorporate lessons learned 6/97 T from Lead Plant activity.
Descrintion: 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 (ELTRI), which outlines the methodology for implementation of an extended power uprate program. ELTR1 encompasses power uprates of up to 120 percent of the original licensed thermal power. Individual plant submittals for uprates willlikely contain requests for an optimum power level specific for that plant which is something less than the full 120 percent.
The technical branches will review the applicable portions of the ELTR2, GE topical report I
containing generic analyses and the lead plant application, and provide input into both safety evaluation reports. Review criteria from the reviews performed on ELTR1, generic analyses, and 33
3 i
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the lead plant submsttal will be developed and asssmbled into a review procedure for individual PMs to use for subsequent plant-specific reviews. If an area in an individual plant submittal is outside the bounds of the previously established criteria, the applicable technical branch will perform a l
review of that specific area and provide input into the safety evaluation.
I Historical Backaround: The generic BWR power uprate program was created to provide a consistent means for individual licensees to recover additional generating capacity beyond their i
current licensed limit. In 1990, GE submetted licensing topical reports to initiate this program by j
proposing to increase the rated thermal power levels of the BWR/4, BWR/5, and BWR/6 product i
lines by approximately 5 percent. Since 1990, the staff has reviewed and approved at least 9 such power uprate requests under this generic BWR power uprate program. As a follow-on to this i
i program, GE submstted ELTRI in March 1995 to propose " extended" power uprates of up to 120 l
j percent of the original licensed thermal power.
4 l
Pronosed Actions Specific actions included in the generic action plan are: (1) review ELTR1 and issue a staff position paper, (2) review ELTR2 and issue a safety evaluation report, (3) review the lead plant application and issue a safety evaluation report, and (4) develop a standard review
)
procedure based on ELTR1, ELTR2, and the lead plant review.
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Oriainatina Document GE Licensing Topical Report (NEDC-32424), " Generic Guidelines for General i
Electric Boeleng Water Reactor Extended Power Uprate," dated February 1995.
Reaulatory Assessment-Not appiscable. (A safety assessment is not noeded for this action plan i
because a justification for continued operation of a plant is not required.) This program is an industry initiative that is strictly voluntary.
Current Status-The lead plant application from Monticello was received on August 1,1990. Supplement 1 to the GE's ELTR2 vras received on July 17,1996. The staff 5 eld a kick-off meeting <m August 8,1996 to initiate the review of the lead plant application.
NRR Lead PM: T. J. Kim, DRPW, 415-1392 t
34
DRY CASK STORAGE ACTION PLAN TAC Nos.
M93821 (issue 2.a)
Last Update: 10/08/96 i
M93927 (issue 3.b)
Lead NRR Division: DRPW M94107 (issue 4.c.)
M94108 GSI: Not Available MILESTONES DATE (T/C)
- 1. Develop action plan 07/95C
- 2. Near-term technical issues
- a. Heavy Loads / Cranes
- develop working group plan 11/95C
- complete actions 6/97T
- b. Cask Trunnions'
- develop staff position 09/95C
- modify standards / guidance No changes required (C)
- c. Hydrostatic Testing' 12/95C
- d. Seismic Requirements for Pads 06/95C
- issue Information Notice
- 3. Long-term technical issues
- a. Cask weeping'
- meet with NEl 08/95C
- determine NRC actions to resolve As Necessary
- b. Cask loading / unloading procedures
- contact NEl about industry efforts 08/95C
- resolve high priority issues 09/95C
- form working group 10/95C
- complete working group determination on further issues 04/96C
- c. Off I oading after fuel pool is decommissioned'
- develop guidance and modifications to inspection As required in procedures response to submittals
- d. Failed Fuel Storage'
- review proposed solutions Reviewing first submittal, ECD 03/97T
- e. Safeguards Concerns'
- complete analysis of designs 12/95C
' NMSS has the lead for this issue.
35
MILESTONES DATE (T/C)
- 4. Procedural issues
- a. Change processes
-issue SRP and 50.59 guidance 03/96C
- training for staff 05/96C
- b. Reporting Requirements'
- develop posetion, communicate to licensees 09/95C
- c. Inspection of site activities
-issue revised procedures 02/96C
- develop resource estimates and inspection schedule 02/96C
- d. Vendor inspections'
- issue revised procedures 02/96C
- develop resource estimates and inspection schedule 10/95C
- e. Cask and SAR differences'
- contact vendors 09/95C
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- 5. Communcations
- a. Interface meetings Ongoing
- b. Staff training' 10/95C
- c. Industry workshop
- 07/95C Descrintion: The Plan was developed to identify and resolve maior issues and problems in the area of dry cask storage of spent reactor fuel in independent spent fuel storage installations (ISFSis).
'i Specific issues encompassed by the plan include haavy load control, procedures for cask loading and unloadeng, failed fuel storage, change processes, inspection activities, and communications l
(internal and external). Issues have been divided into the following categories: near-term technical, long-term technical, communications, and process issues.
Historcal Backaround Since 1986, several U.S. nuclear power plant licensees have installed independent spent fuel storage installations (ISFSis), that is, licensee-owned dry cask storage i
facilities. Other licensees are also planning such installations. In recent years, licensees have encountered a number of problems during the fabrication, installation and licensing of some of these ISFSis and there has been an inconsistent level of performance by involved licensees and cask fabricators with respect to the use of dry cask storage of spent reactor fuel. Because of the anticipated increased industry effort in this area, the staff needed to fully understand the problems that occurred and take appropriate measures to reduce such problems in the future. Therefore, l
NMSS and NRR reviewed the lessons learned from past experience with ISFSis, both our l
experience and the experience of other headquarters and regional offices, and developed a plan to resolve ma#or issues and probierrr.
j Proposed Actions: Actions included in the plan are: (1) review each general issue and identify the I
specific problems to be addressed, (2) develop corrective actions for each problem, and l
(3) implement the correctivo actions.
l Oriainatina Document: Memorandum from Carl J. Paperiello and William T. Russell to James M.
Taylor, July 28,1995, " Dry Cask Storage Action Plan".
i 8 An additional workshop has been tentatively scheduled for May 1996.
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Raoulatory Assessment: The plan addresses dry storage of fuel that is several years old. Technical i
issues have been addressed on a site specific basis for existing facilities. The action plan will j
improve guidance, enhance communications with industry and the public, and aid future applicants.
Current Status: The following action plan issues have been completed: cask trunnions, cask j
weeping, hydrostatic testing, safeguards concoms, Part 72 reporting requirements, inspection of site activities, and vendor inspections. The inspection procedures for dry cask activities (site and vendor) were issued in February,1996. These procedures included resource estimates for inspection activities. The staff has incorporated additional guidance on seismic issues into j
inspection Procedure (IP) 80851 and additional guidance concoming consideration of failed fuel in 1
1 unloadmg procedures into IP 60854. The wortung group has completed its review of the issues
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1 associated with cask loadmg and unloadeg. The working group is implementing its recommendatsons, includeg the preparation of an information notice concoming loading and unloadmg issues. Recent information (e.g., hydrogen ignition at Point Beach) may require additions to the action plan. The schedule for the remaining technical issue (heavy load control) has been extended to allow resolution of issues related to NRC Bulletin 96-02, issued April 11,1996. In j
addition, lasues related to Oyster Creek licensing actions and 50.59 evaluations may provide j
insights into the final resolution of the control of heavy loads. If possible, the issue of potential i
cask drop events prior to securing the lids will be resolved as part of closure of Bulletin 96-02. The variety of issues related to heavy loads and impact on staff resources may justify a separate action I
plan. Croation of a heavy loads action plan may also support closure of this ISFSI action plan since most other activities have been incorporated into the NMSS operating plan. A determination of I
whether to adopt will follow an as of yet unscheduled meeting between the directors of NRR and NMSS. The staff are reviewing and resolving public comments received on the draft dry cask i
storage SRP. Lessons learned from the hydrogen ignition event at Point Beach are being incorporated into the SRP as well as inspection procedures. All of the communications issues are ongoing efforts with no specific criteria for closure. However, there have been significant j
improvements in these areas. The Regions, NMSS, and NRR hold regular interface calls to discuss j
dry cask issues, training has been given to the affected staff, and NRC has established open communications with the newly-formed Nuclear Energy Institute Dry Cask Storage issue Task l
Force. Based on these improvements, the staff will review these issues for closure in the coming I
months. The ECD for the review of the first submittal for the storage of damaged fuel has been changed to 03/97 due to untimely responses from the applicant. NMSS/SFPO sponsored a public
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workshop on dry cask storage on May 17,1996. Well over 300 people attended from the plants, j
states, industry groups, vendors, and the staff. The staff briefed the Commission on the status of the dry cask storage program on May 30,1996. In response to a hydrogen ignition event at Point Beach on May 28, the staff issued Information Notice 96-34 on May 31,1996 NRC Bulletin 96-04 1
I on July 5,1996. The bulletin requires all storage and transportation cask vendors and users to l
submit a report to the NRC that documents evaluations of cask material compatibility with all environments that the casks are expected to encounter. Staff review of the bulletin responses is j
ongoing.
Contact:
Contact:
William Reckley, DRPW, 415-1314 NMSS
Contact:
Patricia Eng, SFPO, 415-8577 i
References-1 l
Memorandum from Robert M. Bernero and William T. Russell to James M. Taylor, March 15,1995, 2
" Realignment of Reactor Decommissioning Program *
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Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, July 28,1995,
" Dry Cask Storage Action Plan" j
Memorandum from Carl J. Paperiello and William T. Russell to James M. Taylor, January 25,1996,
" Update to the Dry Cask Storage Action Plan
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ACCIDENT MANAGEMENT IMPLEMENTATION l
TAC #: M91966 - Overall Last Update: 9/26/96 i
M91641 - BWROG SAMG Review Lead NRR Division: DSSA MILESTONES DATE (T/C)
- 2. Review severe accident trasning materials and BWROG 06/95C prioritization methodologies
- 3. Develop Tl for pilot inspections initial draft (for internal use) 11/95C Site visits of "in-progress" activities 11/96T Revised draft (to NEl and public) 12/96T Final Tl 03/97T
- 4. Complete pilot inspections and follow-up 12/97T
- 5. Revise inspection procedures (IP) and hold public workshop Draft IP 03/98T Public meeting / workshop 05/98T Final IP 07/98T
- 6. Review remaining plants TBD DescnntiGD: This action plan is intended to guide staff efforts to assess the quality of utility implementation of accident management (A/M), and the manner in which insights from the IPE program have been incorporated into the licensees A/M program. Specific review areas will include: development and implementation of plant-specific severe accident management guidelines (SAMG), integration of SAMG with emergency operating procedures and emergency plans, and incorporation of severe accident information into training programs.
Histor6 cal Backaround The issue of A/M and the potential reduction in risk which could result from developeng procedures and training operators to manage acciderats 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. Compietion of A/M is a major remaining element of the integration Plan for Closure of Severe Accident issues (2). The development of generic and plant-specific risk insights to support staff 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 (7).
Proposed Actions: Specific actions included in the A/M action plan are: (1) complete the review of BWROG SAMG documents, (2) conduct site visits in late 1996 and early 1997 to observe how the elements of the formal industry position are being implemented, (3) complete the draft Temporary Instruction (TI) using the information and perspectives obtained 38
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l through the site visits, (4) complete pilot inspections and foSow-up, and (5) develop an inspection
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procedure for use at remaining plants and hold a public workshop. Based on feedback from the
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j workshop, the staff will finalize the inspection procedure, and the opproach and schedule for
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j evaluating A/M implementation for the remaining plants.
Orinmat na Document SECY 88-147, Integration Plan for Closure of Severe Accident lasues, i
May 25,1988.
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Ranulatorv Assessment Accident management programs are bemg implemented by licensees as i
j part of an irwtsetive 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.
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Current Status Severe accident management guideline documents have been submitted by each j
of the PWR owners groups, and reviewed by the staff. The BWROG submitted a severe accident j
management overview document on February 3,1995, and a draft emergency procedure and severe accident guidelines (EPG/ SAG) document on April 6,1995. The BWROG response to staff comments on the overview document was received on March 6,1996. The BWROG submitted a revised draft version of the EPG/ SAG document and an associated draft technical basis document j
to NRC for information on May 10,1996. The revised EPG/ SAG document was developed by a
]
different contractor than used to develop the April draft, thereby circumventing an issue between GE and their ongmal contractor concoming ownership of the earlier document. A final version of the EPG/ SAG document (Rev. 0) and technical basis report was submitted by the BWROG on August 29,1996. In light of staff commitments on other activities and the need to expedite the 2
j review of the BWROG guidelines, the staff has contracted with Oak Ridge National Laboratory to perform a high level review of the EPG/ SAG documents. The target date for concleting the review e
of the BWROG material has been rescheduled for December 1996. A meeting to diswss specific j
questions / concerns regarding the BWROG products is expected in October 1996.
)
Licensee target dates for completing A/M implementation have been submitted to NRC, and a draft j
Tl for use in the pilot inspections has been completed. Comments on the draft Tl have been received from the NRC Region offices. The staff met with industry on February 22,1996 and 1
ACRS on March 1,1996 to discuss plans for inspecting utility implementation of the formal j
industry position on severe accident management and major elements of the draft Tl. The staff will j
visit approximately 2 to 4 sites in late 1996 and early 1997 for the purpose of obtaining an early j
understanding of how the various elements of the formal industry position are being implemented.
j A meeting with NEl to discuss the scope and schedules of the information gathering visits is anticipated in the October 1996 timeframe. The information and perspectives obtained through these visits as well as comments from the Region offices will be used to update the draft Tl. The
}
draft Tl will be made available to NEl and the public after the information-gathering visits.
~
References:
1.
Memorandum from F. Rowsome to W. Minners, "A New Generic Safety issue: Accident Management," April 16,1985 2.
SECY-88-147, Integration Plan for Closure of Severe Accident issees 3.
SECY 95-079, implementation Plan for Probabilistic Risk Assessment 4.
SECY-89-012, Staff Plans for A/M Regulatory and Research Programs 5.
Generic Letter 88-20, Supplement 2, April 4,1990 6.
Letter from W. Rasin to W. Russell, November 21,1994 7.
Letter from W. Russell to W. Rasin, January 9,1995 NRR Technical
Contact:
R. Palla, SCSB, 415-1095 NRR Lead PM:
Ramin Assa, DRPW, 415-1391 39
_ _ _ _ _. _ _ _ _ _ _ _. _. _ _ _ _. _ _ _ ~ _.
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FIRE PROTECTION TASK ACTION PLAN l
TAC Nos.
M86652, M82809, M84592, Last Update: 09/26/96 M85142, and M89509 Lead NRR Division: DSSA GSI: Li-181 f
MILESTONES DATE (T/C)
- 1. Semiannual Commission status reports Last: 04/03/96C Next: 10/96T
- 2. Recommendations for 01/97T 1
action (Part 1)
- 3. Recommendations for 05/97T future study (Part II)
- 4. Confirmation issues 05/97T (Part lil)
- 5. Other issues (Part IV) 08/95C Descriotion: The Fire *rotection Task Action Plan (FP-TAP) is used to track and manage
' nplementation of the recommendations made in the " Report on the Reassessment of the NRC Fire w
Protection Program," of February 27,1993.
Histor6 cal BmLwwgi: le February 1993, the Office of Nuclear Reactor Regulation (NRR) completed a rwsessment of the reactor fire protection review and inspection programs in response to prastammatic 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.
Proposed Actions The FP-TAP tracks the implementation of a wide range of technical and programmatic fire protection issues, it includes recommendations for action (Part 1),
recommendations for further study (Part II), confirmation issues (Part lli), and lessons learned (Part IV). The staff is implementing the recommendations, in priority order, as resources allow.
The staff focus is now on implementing its plan for future direction of the NRC fire protection program with emphasis on the fire protection functional inspection (FPFI) program and centralizing the management, by NRR, of the FPFI program and all other reactor fire protection work. The principal objective of these efforts is to ensure that the NRC has a strong, broad-based and coherent fire protection program which is commensurate with the safety significance of the subject.
Oriainatina Document: " Report on the Reassessment of the NRC Fire Protection Program,"
February 27,1993.
Ranulatory 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 appropriate priority.
Current Status: The staff issued a semiannual report to the Commission on the status of the FP TAP on April 3,1996. The next status report is due October 1996.
40
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 f
l l
evaluate the risk associated with the post fire safe-shutdown methodology that imposes a j
self-induced station blackout. The staff plans to apply the PRA model for assessing the risk segnificance of the self-induced station blackout methodology to two plant-specific cases during FY 97. The staff is working on an issue recommended for further study regarding fire barrier reisabelety, under Genenc Safety lases (GSI) 149, " Adequacy of Fire Barriers." The staff and BNL have performed scoping analyses, using fault trees and event trees, to assess the effectiveness of a degraded fire barrier in metigating 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 prelimmary 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. Technical assistance l
fundmg to complete these two projects has been budgeted for FY 97.
l Scientech and BNL are providing technical assistance for developing the FPFI procedures.
Scientech and BNL submetted the first draft of their work product on September 6 and September 16 respectively, j
Several tasks are on hold until an expected increase of fire protection resources is implemented.
l The tasks that need to be rescheduled include (1) a fire protection training program, (2) two recommendations for further study, shutdown operability requirements, and (3) several remaining confirmation issues.
Contact-D. Oudinot, DSSA, 301-415-3731 References-
" Report on the Reassessment of the NRC Fire Protection Program," of February 27,1993.
l SECY-95-034, " Status of Recommendations Resulting From the Reassessment of the NRC Fire i
Protection Program," February 13,1995.
l Memorandum of April 3,1996, 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."
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4 e
THERMO-LAG ACTION PLAN i
l
- *
- FINAL REPORT * *
- TAC Nos.
M82809, M90203, M90284 Last Update: 09/26/96 i
GSI: Not Available Lead NRR Division: DSSA MILESTONES DATE (T/C) 1 i
- 1. Semi-annual Commission status reporss Last: 04/03/96C Next: 10/96T
- 2. Resolve technical issues (Part I) 09/96C 1
j
- 3. Testing (Part II) 04/95C
- 4. Assess NRC fire prot. program (Part IV) 02/93C Descnotion: Evaluation and resolution of generic Thermo-Lag fire barrier issues regarding toxicity, construction and installation, fire endurance, ampacity dorating, combustibility, seismic capabilities, 4
and uniformsty of materials. Includes special review team findings, public concoms, coordinating I
with Nuclear Energy Institute (NEI) and licensees, conducting fire endurance and ampacsty dorating j
tests, and assessing NRC reactor fire protection program. The staff has issued 16 genonc 4
communecations regarding Thermo-Lag fire bemers.
{
Histoncal Backaround: In June 1991, the Office of Nuclear Reactor Regulation (NRR) established a l
special team to review the safety significance and generic applicability of technical issues regarding the use of Thermo-Lag fire barriers. In April 1992, the special review team issued its final report, which identified concoms about fire endurance, combustibility, and ampacity dorating.
j Subsequently, the NRR staff prepared an action plan to address the issues associated with i
Thermo-Lag and the NRC fire protection program. The scope of the action plan includes coordination with industry and testing by the staff.
Pronosed Actions: Specific actions include (1) the resolution of concerns and generic issues raised by the special review team and (2) resolution of plant-specific issues that emerge from the generic issues, in June 1994, the Commission approved a staff recommendation to resolve Thermo-Lag concerns by requiring compliance with existing NRC requirements and to permit plant-specific exemptions, where justified.
Orininatsna Document-Final Report of the Special Review Team for the Review of Thermo-Lag Fire Barrier Performance, April 1992.
Reaulatory Assessment: In response to Bulletin 92-01 and its supplement, licensees with Thermo-Lag fire barriers established NRC-approved measures, such as fire watches, to compensate for possibly inoperable fire barriers. The combination of compensatory measures and the defense-in-depth fire protection features provides an adequate level of fire protection until licensees implement permanent corrective actions, i
Current Status: The staff issued the semiannual report to the Commission on the status of the Thermo-Lag Action Plan on April 3,1996. The next report, due during October 1996, will be the final report on the status of the Thermo-Lag Action Plan.
42
. - - - - - _ - = =._
_ _ - - ~ ~. - - -. _... -. -.
Work on the final generic issue-mechanical properties tort program-has been completed. NIST submitted its final report in July 1996. The staff evaluated the test results and is considering issuing an information notice to inform industry of the test results, in June 1996, NIST submitted its final report regarding the feasibility of developing fire curves for rating fire barriers on the basis of representative nuclear power plant fire hazards rather than the fire curves specified in existing fire test standards.The study was published as NUREG 1547,
" Methodology for Developing and Implementing Alternative Temperature-Time Curves for Testing the Fire Resistance of Barners for Nuclear Power Plant Applications," August 1996. Staff action on the feassbelety study is complete.
The staff is reviewing plant-specific corrective actions as a multiplant action (MPA) under Generic Letter 92-08 (MPA-L208), and the review of related plant-specific issues, such as exemption requests, as licensing actions. The MPA and plant specific licensing actions are tracked by the Workload information and Scheduling Program (WISP) and are not a part of the Thermo-Lag Action Plan. The plant specific activities also appear in the Chairman's Tracking List as item II.N.1.
i Contacts-D. Oudinot, SPLB, 301-415-3731 L. Tran, DRPW, 301-415 1361 i
i 1
43
PRA IMPLEMENTATION ACTION PLAN TAC Nos.
M90370, M90371, M90227, Last Update: 9/30/96 M90977, M91787-M91802 Lead NRR Division: DSSA GSI: Not Available MILESTONES DATE(T/C)
- 1. ACRS Meeting 07/94C 08/96C 11/96T
- 2. Commission Briefing 08/94C 04/95C 04/96C 10/96T
- 3. Publish PRA Policy Statement for 60-day comment period 12/94C
- 4. ACRS Subcommittee Meeting 09/94C 07/96C
- 5. Conduct Public Workshop on PRA implementation Plan 12/94C
- 6. Publish final PRA policy statement 08/95C
- 7. Detailed implementation NA 1.1(a)
Develop draft Standard Review Plans for risk-informed 11/96T regulation for ACRS review 1.1(b)
Publish draft Standard Review Plans for Public comment 12/96T 1.1(c)
Final draft Standard Review plans for ACRS review 9/97T
'.1(d)
Publish final Standard Review Plans 12/97T 1.2 Pilot Applications to Specific Regulatory Initiatives:
(a) MOVs (a) 2/96C (b) IST (b) 3/97T (c) ISI (c) 6/97T (d) Graded QA (d) 6/97T (e) Maintenance Rule (e) 09/95C (f) Technical Specifications (f) 12/96T (g) Other applications to be identified later 1.3(a)
Develop inspection Guidance to Use IPEs and Plant-12/96T Specific PRAs 1.3(b)
Develop training course for inspectors 12/96T 1.3(c)
Support regionalinspection activities Ongoing 1.4 Operator Ucensing - Revise Examiner's Handbook to Reflect 12/96T Revised Knowledge & Abilities Based on Risk Insights 1.5 Event Assessment -
(a) Conduct event assessment of reactor events (a) Ongoin (b) Assess desirability of risk assessment on non-power reactors (b) TBD
.~
i MILESTONES DATE(T/C) 1.6 Review Adequacy of Licensee Analysis in IPEs/IPEEEs 6/97T 1.7 Apply Guidance to Assess Effectiveness of SBO and ATWS 09/97T Rules 1.8(a)
Staff review of PRAs for design certification applications Ongoing 1.8(b)
Develop SRP for Review of PRAs for Evolutionary Reactor 12/99T Designs j
1.8(c)
Develop Guidance for Use of Risk in Simplification of 12/96T Emergency Planning Requirements i
1.9 Accident Management - Develop Risk insights to Review and TBD i
Inspect Industry Accident Management Programs i
1.10 Evaluate IPE insights to determine followup activities TBD
}
Descriotion: This action plan is intended to describe the process for the staff to use PRA method j
l and technology in the agency's effort toward risk-informed regulatory approach. The plan encompasses methods development, pilot applications, and staff training. The plan will be used to t
ensure timely and integrated agency-wide effort that is consistent with the PRA Policy Statement.
i Historical Backaround: The NRC has been making use of PRA technology to varying degrees in its regulatory activities since WASH-1400. Prior to 1991, this had been an ad hoc application, depending on the availability of expertise in various technical groups. Since 1991, there have been
)
a number of high-level studies within NRC that have focused on the status of PRA use and its role i
in the regulatory process. Collectively, the findings and recommendations from these studies support the view that there is a need for increased emphasis on PRA technology applications. For the full value of our investment in risk assessment me,thodology to be achieved, it is important that consistent high-level agency guidance be provided or. the appropriate use of PRA. To this end, in November 1993, the Office Directors of NRR, AEOD, NMSS, and RES proposed to take the initiative in providing guidance on coordination and expe';tations for PRA efforts. Specifically, they proposed to develop an integrated plan for the staff's ris'. assessment and risk management practices. In August 1994, the staff submitted SECY-94-219, " Proposed Agency-Wide Implementation Plan For Probabilistic Risk Assessment," for the Commission's irJormation. On March 30,1995, The staff submitted SECY-95-079, " Status Update of the Agency-Wide implementation Plan for PRA," and briefed the Commission on the subject on April 5,1995. On May 18,1995, the staff forwarded SECY-95-126, " Final Policy Statement on the Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities," for Commission vote. On June 8,1995, the staff briefed the ACRS on the PRA policy statement. The final PRA policy statement was published in the FederalRegister on August 16,1995.
1 Prooosed actions: The PRA implementation Plan includes activities for NRR, RES, AEOD, and NMSS staff to increase the use of PRA methods in all regulatory matters. NRR focuses on the PRA applications in reactor regulations, the development of standard review plans, the pilot programs to use PRA technology in specific regulatory initiatives, events assessment, and working with regions on risk-informed inspections. RES focuses on the IPE/IPEEE reviews, PRA method and quality, and the development of PRA regulatory guides for the industry. AEOD focuses on risk-informed trends
]
and patterns analysis, reliability data for PRA applications, and staff training. NMSS focuses on using PRA in high and low level waste issues. The detailed actions are described in the PRA implementation ?lan.
45
. -.. - - _. - -.. - - - - = -. - -
1 Onninatina Document-Memorandum dated November 2,1993, T. Murley et al. to J. Taylor,
" Agency Directions For Current and Future Uses of Prob 6bilistic Risk Assessment".
1 l
Raoulatory Assessment: This action plan is meant to improve the regulatory process by developing
}
state-of-the-art PRA tools that will expand the use of PRA technologies in making regulatory decisions. The plan is not intended to correct safety problems at licensed facilities. Therefore, I
continued facility operation is justified.
l j
Current Status: On February 27 and 28,1996, the staff met with the ACRS PRA subcommittee to i
j discuss technecol issues related to risk-informed regulation. This was followed by a meeting with a
the ACRS full commettee on March 8,1996. On March 26,1996, the EDO forwarded a l
memorandum to the Commission updating the progress and status of the PRA implementation Plan.
j On April 4,1996, the staff briefed the Commission on the status of and progress made regarding l
j the activities in the PRA implementation Plan. On May 15,1996, the Commission issued an SRM j
directing the staff to prepare a policy paper to address the four emerging policy issues raised in the i
j March 26,1996 PRA implementation Plan update. The staff was also asked to update the j
Commission on the use of the Safety Goal subsidiary objectives and to clarify how it intends to j
address uncertainties in risk-informed and perfcfmance-based regulation. On June 20,1996, the EDO forwarded the quarterly status update of the PRA implementation Plan to the Commission.
p The staff met with the ACRS PRA subcommittee on July 18,1996, to discuss resolution for the four policy issues. The subcommittee meeting continued on August 7 to discuss the risk informed l
pilot applications. An ACRS full committee meeting was conducted on August 8 to discuss the options and staff proposed recommendations for resolution of the four policy issues. On August 15,1996, the ACRS forwarded a letter to Chairman Jackson expressing its agreement with the j
staff proposed recommendation for resolution of the key policy issues raised in the March 26 update of the PRA implementation Plan.
The staff has prepared an " Issues List" encompassing the key technical, policy and process issues related to risk-informed regulation. The staff continues to work toward resolution of these issues and these issues will be addressed in the RGs and SRPs. This " Issues List" and the staff response to the SRM dated May 15,1996, were attached to the most recent PRA implementation Plan i
updateforwarded to the EDO's office on September 27.
NRR Techmcal Contacts: Tony Hsia, SPSB, 415-1075 References-SECY 94-219, " Proposed Agency-Wide implementation Plan for Probabilistic Risk Assessment" SECY-95-079, " Status Update of The Agency-Wide implementation Plan for Probabilistic Risk Assessment" i
SECY 95-126, " Final Policy Statement on The Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities
- SECY-95-280, " Framework For Applying Probabilistic Risk Analysis in Reactor Regulation" Memorandum from James M. Taylor to Chairman Jackson, "lMPROVEMENTS ASSOCIATED WITH MANAGING THE UTILIZATION OF PROBABILISTIC RISK ASSESSMENT (PRA) AND DIGITAL INSTRUMENTATION AND CONTROL TECHNOLOGY," January 3,1996.
Memorandum from James M. Taylor to the Commission, " Status Update of the Agency-Wide implementation Plan for Probabilistic Risk Assessment (PRA) (From March 30,1995 to February i
29,1996)," March 26,1996.
46 i
Staff Requirements - Briefing on PRA implementation Plan,10:00 a.m., Thursday, April 4,1996, Commissioners' Conference Room, One White Flint North, Rockville, Maryland (Open to Public j
Attendance), May 15,1996.
Memorandum from James M. Taylor to the Commission, " Status Update of the Agency-Wule implementation Plan for Probabilistic Risk Assessment (PRA) (From March 1,1996 to May 31, 1996)," June 20,1996.
Letter from T. S. Cross, ACRS Chairman to Chairman Jackson, NRC, " Risk-informed, performance-based regulation and related matters" dated August 15,1996.
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d 47
ENVIRONMENTAL QUALIFICATION TASK ACTION PLAN l
l TAC No.
M85648 Last Update: 09/26/96 GSI: 168 Lead NRR Division: DSSA l
f i
MILESTONEL DATE (T/C) f 1.
Inform Commission 05/93C r
2.
Meet Wnh Industry Ongoma 3.
Programmatic Review TBD 4.
Risk Assessment TBD I
5.
Data Collection and Analysis 4/96C 6.
Review and Evaluation of the Status TBD l
7.
Technical lasues 10/98T 8.
Options for Resolution TBD t
i 9.
. implementation TBD j
i Descriotion: This action plan will evaluate environmental qualification (EO) issues, including i
operating experience, testing methodology, and adequacy of current rule and guidance for i
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 i
and failures of qualified cables during research tests led to the development of the EO Task Action Plan (TAP), which was issued in July 1993. The EO TAP was developed to address: (1) staff l
concerns regarding the differences in EQ requirements for older and newer plants; (2) concems i
I raised by some res6 arch tests which indicate that qualification of some electric cables may have been non-conservs tive; and (3) concems that programmatic problems identified in the staff Fire Protection Reassessment Report might also exist in the NRC EO Program.
Procosed Actions The EQ TAP includes meetings with industry, a program review of EO, 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 EO TAP. The staff will develop options for resriving EO 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 regulat Jry action will be documented.
Cridnatina Document: June 28,1993, memorandum from Samuel J. Chiik to James M. Taylor (SECY 93-049); May 27,1993, letter to the Commission from J. Taylor on Environmental Qualification of Electric Equipment.
Reaulatory 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 EO qualification test margins.
Current Status: Members of RES, NRR, and BNL met at BNL to discuss the comments from the pubhc meeting held August 6-7,1996. During the public meeting, several members of the public indicated that certain unresolved issued included in the RES research program have been resolved 48
l and that reports are available. RES is currently in the process of obtaining the reports and 1
evaluating them for their impact on the EO cable test program.
The draft reports on the programmatic "eview and risk issues regarding EO are currently under management review (Milestones 3 and 4).
BNL is continuing with the cable testirag program, which includes investigating condition monitoring i
methodologies (Milestone 7). The cable test program began in May 1996 and is expected to continue for two years. In addition to aging cable samples, several condition monitoring methodologies are being investigated. Results from the test program are expected in fiscal years I
1998 and 1999.
The staff is preparing an evaluation to resolve whether differences that currently exist in EO requirements between older and newer plants are safety significant. This is one of three fundamental task action plan issues. The evaluation is scheduled to be completed by November 8, 1996. The evaluation will be forwarded to the Commission in the next EO-TAP update, which is also expected to be completed by November 8,1996.
SPLB will continue as the lead branch for the EO Task Action Plan, and will continue to be
[
responsible for coordinating this monthly update. EELB has responsibility for all other EQ issues.
1 Contacts: NRR Technical
Contact:
G. Hubbard, SPLB, 415-2870 RES
Contact:
S. Aggarwal, EMEB, 415-5849 NRR Lead PM:
L Olshan, DRPE, 415 3018 Referenegg:
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 i
(Accession No. 9409010107).
Task Action Plan for Environmental Qualification and updates, Task Action Plan for Environmental l
Qualification and updates, July 1,1993, April 8,1994, November 16,1994, and June 27,1995 (Accession Nos. 9308120145, 9404260206, 950110431, 9507110203, respectively).
RES Program Plan for Environmental Qualification, July 7,1994 (Accession No. 9407250066),
i i
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1 i
l GENERIC SPENT FUEL STORAGE POOL 3
PART A: OFERATING FACILITIES i
4 l
TAC No.
M88094 Status: Cornplete GSI: Gl-173.A Lead NRR Division: DSSA Descriotion: The action paan encompassed Spent Fuel Pool (SFP) issues identified through a 1994 3
special inspection at Dresden 1, the staff's review of loss of SFP coolmg concerns at Susquehanna i
i Steam Electric Station (SSES), and other SFP concerns identified as part of this plan. Specific j
review areas identified through implementation of this action plan include plant desson features and
]
administrative controls that affect the probatwisty of spent fuel poci boiling, adverse environmental effects on essential equipment due to bosieng, significant loss of spent fuel pool coolant inventory,
)
j adverse radiologecal condetions, unplanned spent fuel pool reactivity changes, undetected spent fuel pool events, and adverse effects of control system actuations.
j Histoncal Backaround In November 1992, two engineers, who formerly worked under contract for
{
the Pennsylvania Power and Light Company (PP&L), filed a report contending that the design of the l
Susquehanna station failed to meet regulatory requirements with respect to sustained 1:ss of the j
cooling function to the SFP that mechanistically results from a loss-of-coolant acciGnt (LOCA) or a loss of offsite power (LOOP). The licensee (PP&L) and the engineers each made a series of addstional submettals to the NRC and participated in public meetings with the NRC staff to describe their respective positions on a number of technecal and licensing issues. In order to inform the i
nuclear power industry of the issues, the agency issued information Notice (lN) 93 83 on October 7,1993. The staff evaluated these issues as they related to Susquehanna using a probabilistic l
safety assessment, a deterministic engineering assessment, and a licensing basis analysis. The staff issued their final safety evaluation report on June 19,1995. This closed the Susquehanna action plan (TAC No. M85337).
A genenc action plan was developed and adopted on October 13,1994, with two parts. Part A
)
(TAC No. M88094) encompasses the staff's review of generic issues relating to the SFP at operating reactor facilities. Part B (TAC Nos. M40004, M90441, and M93805) includes applicable issues from the Part A review and concems from the Dresden 1 special inspection particular to permanently shutdown facilities with stored, irradiated fuel to establish evaluation criteria for spent fuel pools at permanently shutdown facilities. Part B was included after the special inspection at Dresden 1 determmed that problems in implementing the facility's decommissioning plan combined with certain SFP design features created the potential for a substantial loss of SFP water inventory.
Dresden 1, which is permanently shutdown, experienced containment flooding due to freeze damage to the service water system on January 25,1994, and the licensee for Dresden 1 reported a similar threat to SFP integrity. This licensee report resulted in the special inspection.
The principal concoms included in Part A of the generic action plan involve the potential for a sustained loss of SFP cooling capability, which was identified through the report filed with the N1C relating to Susquehanna, and the potential for a substantial loss of SFP coolant inventory, which was given renewed emphasis following the Dresden 1 special inspection. Postulated adverse conditions that may develop following a LOCA or a sustained loss of power to SFP cooling system components could prevent restoration of SFP decay heat removal. The heat and water vapor added to the building atmosphere by subsequent SFP boiling could cause failure of accident mitigation or other safety equipment and an associated increase in the consequences of the initiating event. Incomplete administrative controls combined with certain design features, particularly at the oldest facilities, may create the potential for a substantial loss of SFP coolant inventory and the associated consequences, which include high local radiation levels due to loss of shielding, unmonitored release of radiologically contaminated coolant, and inadequate cooling of stored fuel.
50
Oriainatino Documents: (1) Letter from D.A. Lochbaum and D.C. Prevatte to T. Martin, NRC, November 27,1992, "Susquehanna Steam Electric Station Docket No. 50-387, License No. NPF-14,10 CFR 21 Report of Substantial Safety Hazard;" (2) Inspection Report No. 50-010/94001.
Reaulatory Assessment: The postu!ated events do not pose an undue risk to the public based on the availability of common design features that help protect stored irradiated fuel, protect estential reactor safety systems, and prevent development of adverse radiological conditions. These design features include the provision of diverse means of cooling, the strong structural design of the spent fuel pool, the absence of drainage paths from the pool, the anti-syphon protection on piping within the scent fuel pool, the availability of multiple sources of make-up water, spent fuel pool instrumentation with control room annunciation, the maintenance of a substantial shutdown reactivity margin in the pool, radiation shielding provided by coolant inventory, and spent fuel pool water purificdtion systems. Additionally, the relatively slow evolution of these events in the spent fuel pool resulting from the initial large coolmg water inventory creates significant opportunity for operator recovery prior to experiencing, adverse conditions or consequences. Therefore, continued facility operation is justified.
l Resolution: The staff identified concoms for evaluation and reviewed existing guidance documents.
On the basis of the identified concoms, the staff developed plans for on-site safety assessments of spent fuel storage. The on-site assessments were conducted at Brunswick, Monticello, Comanche Peak, and Ginna. The assessment teams cordded that the potential for a sustained loss of spent fuel pool cooling or a significant loss of sperr. tuel pool coolant inventory at the sites visited was remote on the basis of certain design features and operational controls. The teams found that other identified concems within the scope of the action plan review were much less significant than a sustained loss of spent fuel pool cooling or a significant loss of spent fuel pool coolant inventory in terms of risk at the plants visited. The staff completed individual assessment reports documenting the findings from visits to Brunswick, Monticello, Comanche Peak, and Ginna.
The staff then performed an FSAR-based review to identify facilities whose design was not well represented by any of the facilities reviewed through the on-site assessments. On the basis of this FSAR review of 16 sites in addition to the sites visited, the staff determined that the s.gnificant spent fuel pool issues are best resolved through a site-specific evaluation because of the small number of facilities affected by each particular concern and because of site-specific variations in design and operation of the spent fuel pool and associated systems. To accomplish this task, the staff expanded the FSAR-based review to encompass development of a data-base specifying the current licensing basis for the SFP cooling system, selected design basis parameters, and current operating procedures relevant to SFP cooling for all facilities. The staff initiated this expanded review on January 16,1996. Proiect managers conducted the data collection function, which was performed under TAC M94480. In order to develop a consistent licensing basis determination, the technical staff devoted substantial resources to a plant by-plant licensing basis review, which was forwarded to the respective project managers prior to on-site visits.
The staff briefed Chairman Jackson regarding SFP issues on February 1,1996, and on April 4, 1996. Following these briefings, the staff committed to provide results of the plant specific licensing basis review effort to the Chairman. Additionally, the staff committed to prepare a course of action for resolution of significant issues and identify plant-specific and generic areas for regulatory analysis. The licensing basis review effort was completed on May 21,1996.
The staff completed its evaluation of significant spent fuel pool issues and its identification of plant specific and generic areas for regulatory analysis. These findings were transmitted to the Commission in a report dated July 26,1996. The staff identified 38 plants at 22 sites having spent fuel pool design features from seven categories that the staff found to warrant regulatory analysis for potential safety enhancement backfits. Also, plants with SFP related design features in three additional categories were identified for further review. Generic areas identified for regulatory 51
analysis include applecation of the shutdown operations rule to spent fuel pool refueling activities I
and revision to NRC design review guidance.
During separate public meetings, the staff presented its findings to the Commission on August 1, 1996, and to the Advisory Commntee on Reactor Safeguards on August 9,1996. In response to comments received from the Commission in a staff requirements memorandum dated August 26, 1996, the staff committed to complete regulatory analyses associated with plant-specific design features by May 1997, implement plant specific backfits by October 1997, and complete revisions to Regulatory Guide 1.13, and Standard Review Plan sections 9.1.2 and 9.1.3 by October 1998.
The staff is currently conducting the regulatory analyses supporting application of the shutdown operations rule to the spent fuel pool under the action plan for that rulemaldng activity. Revisions to Regulatory Guide 1.13, and Standard Review Plan sections 9.1.2 and 9.1.3 will be managed I
under separate TACs for each document. Similarly, plant specific regulatory analyses will be managed under individual TACs for each affected plant.
Contacts-S. Jones, 415-2833 J. Shea, 415-1428 References
- Letter from Lochbaum and Prevatte, November 1992 Task Action Plan for Spent Fuel Storage Pool Safety, October 13,1994 (publicly available, Accession No. 9410190155)
SER for Susquehanna, June 19,1995 (publicly available, Accession No. 9507070008)
Information Notice 95-54', December 1,1995 (SFP cooling design basis at Millstone 1 and Cooper)
Information Notice 93-83 (and Supplement 1), October 7,1993 and August 24,1995.
l Information Notice 94-38, May 27,1994 (Dresden 1 Special inspection Results) i inspection Report No. 50-010/94001, April 14,1994 (Dresden 1 Special Inspection)
Report to the Commission, " Report on Survey of Refueling Practices," from J. M. Taylor, May 21, 1996 (publicly available)
Report to the Commission, " Resolution of Spent Fuel Storage Pool Action Plan issues," from J. M.
Taylor, July 26,1996 (publicly available) 52
GENERIC SPENT FUEL STORAGE POOL PART B: PERMANENTLY SHUTDOWN FACILITIES TAC Nos.
M90441 & M93805 Status: Complete GSl: Gl-173.B Lead NRR Division: DSSA Desenotion This Part B effort will use the results of Part A activities to establish evaluation criteria for spent fuel pools (SFPs) at permanently shutdown plants to support rulemaking and other generic activities initiated by the Decommissioning and Non-Power Reactor Project Directorate (PDND).
Histoncal Backaround. A generic action plan was developed and adopted on October 13,1994, with two parts. Part A (TAC No. M88094) encompasses the staff's review of generic issues relating to the SFPs at operating reactor facilities. Part B (TAC Nos. M40004, M90441, and M93805) includes applicable issues from the Part A review and concerns from the Dresden 1 special inspection particular to permanently shutdown facilities with stored, irradiated fuel to establish evaluation criteria for SFPs at permanently shutdown facilities. Part B was included after the special inspection at Dresden 1 determined that problems in implementing the facility's decommissioning plan combined with certain SFP design features created the potential for a substantial loss of SFP water inventory. Dresden 1, which is permanently shutdown, experienced containment flooding due to freeze damage to the service water system on January 25,1994, and the licenses for Dresden 1 reported a similar threat to SFP integrity. This licensee report resulted in the special inspection.
The staff issued NRC Bulletin 94-01, " Potential Fuel Pool Draindown Caused by inadequate Maintenance Practices at Dresden Unit 1," on April 14,1994. This bulletin requested all holders of licenses for nuclear power reactors that are permanently shut down with spent fuelin the spent fuel pool to take actions to ensure the quality of the SFP coolant, the ability to maintain an adequate coolant inventory for cooling and shielding, and the necessary support systems are not degraded, in order to evaluate the management controls and SFP activities at permanently shutdown reactors, the NRC staff initiated a series of special team inspections at permanently shutdown fa::ilities with stored, irradiated fuel in the SFP. These inspections were completed at all of the subject facilities by the first quarter of 1995.
Oriainatina Documents: Inspection Report No. 50-010/94001 for Dresden Unit 1.
Reaufztorv Assessment: The postulated events involving a loss of cooling do not pose undue risk to the public, because of the low residual decay heat in the spent fuel at permanently shutdown reactors and the associated long period of time available for recovery. Concerns involving maintenance of the coolant quality and ability to control coolant inventory have been addressed through the special int,pection activities. Therefore, continued facility operation is justified.
Resolutiom The staff determined that all significant identified concerns from Part A applicable to permanently shutdown facilities were encompassed by the special inspection activities. The special inspections found no significant deficiencies other than at Dresden 1. In response to the Dresden 1 Special Inspection findings, PDND will issue decommissioning guidance consistent with the newly revised decommissioning regulations. The Division of Systems Safety and Analysis will provide technical support for these activities. Staff resources will be tracked through related TACs assigned to PDND.
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Contacts' NRR Technical
Contact:
S. Jones,SPLB, 415-2833 NRR Lead PM:
R. Dudley, PDND, 415-1116
References:
Task Action Plan for Spent Fuel Storage Pool Safety, October 13,1994 (publicly available, Accession No. 9410190155)
Information Notice 94-38, May 27,1994 (Dresden 1 Special Inspection Results)
NRC Bulletin 94-01, April 14,1994.
l Inspection Report No. 50-010/94001, April 14,1994 (Dresden 1 Special inspection) t I
54 i
____ __ _ _ _ _._ _ _ _. _ _.q t
l CORE PERFORMANCE ACTION PLAN i
t TAC Nos.
M91257 - DSSA Last Update: 09/26/96 l
M91602 - DISP Lead NRR Division: DSSA l
GSI: Ll-179 Supporting Division: DISP i
i MILESTONES DATE (T/P/C) i j
Task 1 -
Inspection of Nuclear Fuel Vendors (DISP) 03/97T
}
Seemens Power Corporation (PWR AIT followup) 06/94C l
AB8/ Combustion Engineering (PWR reloads) 11/94C I
Tolodyne-Wah Chang (TWC) 12/94C Sandvik Specialty Metals (SSM) 12/94C 1
Westinghouse CNFD 07/95C j
General Electric NEP 10/95C j
Framatome/Cogoma Fuels (B&W Fuels) 09/96C i
GE (SLMCPR & low density pellets)*
09/96C A88/CE (BWR) (WNP-2 transition core)*
12/96T l
SPC (comprehensive re-inspection of open items)*
02/97T Task 2 -
Inspection of Licensee Reload Analyses (DSSA) ongoing *
)
Ri - GPU [TMI-1);
12/95C l
l Ril - Duke (Oconeel:
03/95C j
Rlli - Comed (Zion);
10/94C RIV - APS (Palo Verdel (original pilot audit) 04/93C Task 3 -
Core Performance Data Gathering / Evaluation (DSSA) 12/96T i
j Regions - Morning Reports & Event Notification ongoing i
Other - Data Acquisition and Oct'ation ongoing
{
PNL
- Core Performance Evat a ion Analysis (CY95) 12/96T 1
Task 4 -
Participation of Regions in Action Plan (DSSA) ongoing i
l Identification of Vendor issues i
Feedback from Licensee inspections i
Counterparts Meetings (RI-RIV) 1 1
Task 5 -
Evaluate inspection Guiwnce (DSSA/ DISP) 12/96T 1
Evaluate Results of Vendor / Licensee inspections incorporate Feedback from Regions Draft Guidance for Residents i
Draft inspection Criteria and Plan Outline i
Task 6 -
Evaluate Lead Test Programs for identification of Core 12/96T Performance Problems (DSSA/ DISP) i Task 7 -
Workshop on Core Performance issues (TAC No, M95674)
Identify issues 07/96C Conduct workshop 10/96T
=1..u. 0,iven 55
Descrioten: The action plan is intended to assess the impact of reload core design activities on plant safety through inspections of fuel vendors, evaluation of licensee's reload analyses, independent evaluation of core performance information, with regional training and interaction.
Historical Backoround: The action plan addresses the review of fuel fabrication, core design, and reload analysis issues that were discussed duririg the March 29,1994, briefing given to James M.
Taylor, Executive Director for Operations. The briefing presented by the Reactor Systems Branch (SRXB), Division of Systems Safety and Analysis (DSSA), covered generic fuel and core performance issues and related evaluations of fuel failures. Representatives of the Vendor inspection Branch (VlB), Division of Reactor inspection and Licensee Performance (DRit),
participated in the briefing. As a result of this briefing, the Office of Nuclear Reactor Regulation (NRR) was requested to prepare an action plan for a proactive approach to monitor and improve core performance in operating reactors.
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 desien, and fuel fabrication activities; (2) evaluate the performance of licensees that perform core reload analysis functions; (3) identify, document, and categorize core performance problems and root cause evaluat'ons that will be further evaluated during these inspections and provide input to SALP evaluatio.'s as well as regional enforcement actens, as appropriate; (4) train and coordinate regional support staff participating in these activities; and (5) ov.sluate the results of these activities for use in formula:ing genonc communications, revisions of regulatory guidance and guidance for regional inspectors, and other appropriate regulatory actsons. 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 fall of 1996, as stated at the recent Regulatory information Conference.
DSSA - The action plan identifies that licensee inspections in each region shall be performed, in coordination with the regeonal 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 t
conformance to current regulatory requirements. The end product of the initial assessment will include guidance for resident inspectors and regional staff. These activities are scheduled to be completed in FY96. The ongoing activities to capture and address early warning of emerging issues will continue into FY97, and the action plan will be updated to reflect the planned inspection of 10 licensee / plants, 5 vendor LTA program inspections, and four anticipated event-reactive ic.spections.
DISP -The action plan currently identifies 8 completed and two planned vendor inspections that shall be performed by multi-discipiened inspection teams lead by the Special inspection Branch (PSIB) with contracted technical assistance. These inspections are currently scheduled to be completed in 1997. In addition, DISP will support the FY97 vendor LTA and licensee inspections, as required.
Onoinatina 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 revision, in progress.
56 1
.. _ _ _ _. _ _ _ ~. -. _... _
Raoulatorv Assessment: Coro design is a fundamental component of plant safety because maintaining fuel integrity is the first principal safety barrier (i.e., fuel cladding, reactor coolant system boundary, or the containment) against serious radioactive releases. Likewise, the safety analyses must be properfy performed in order to venfy, 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 being acquired from the ongoing vendor inspections are being evaluated for
~
generic impact and identification of emerging issues. The issue-driven inspection at GE, conducted in May 1996, was supported by SRXB/DSSA staff and contract specialists in reload design. The first part of the Framatome inspection was supported, with the final phase to be conducted during actual fuel fabrication campaigns. Interaction with the regions is ongoing to participate in region-led licensee inspections. However, due to diversion of resources to review lead test assembly programs, it is uniekely that licensee inspections in each region can be supported this year. SRXB participated in the Region I inspector counterparts meetings in December 1995 and May 1996.
DSSA is re-evaluating the action plan to better integrate and prioritize its activities, consistent with the available TA funding. Options and recommendations for management review are being prepared.
DISP - The inspection of Framatome Cogoma Fuels (formerly Babcock and Wilcox Fuel Company),
j located in Lynchburg, Virginia, began in March 1996; however, FCF production scheduling delays will result in delaying the manufacturing end of the inspection. An assessment of GE SLMCPR orrors and the low density pellet issue was conducted in May. The remaining planned issue-driven inspections include ABB Combustion Engineering's supply of a BWR transition core reload for WNP-2, and a comprehensive follow-up inspection of Siemens Power Corporation issues.
NRR Technical Contacts:
E. Kendrick, SRXB, 415-2891 S. Matthews, PSIB, 415 3191
- ime opent on-site et vendor inspections (Teek U is allocated to appropriate fuel vendor docket #
t 57
HIGH BURNUP FUEL ACTION PLAN TAC No M91256 Last Update: 10/2/96 Lead NRR Division: DSSA l
I GSI:170 Supporting Office: RES MILESTONES DATE (T/C) i 1.
Issue User Need Letter to RES 10/93C i
2.
Contracts lasued by RES 03/94C i
3.
Schedule and Coordinate Meetings with Foreign Experimenters and 09/95C Regulatory Authorities 4.
Issue Information Notice (IN 94-84) Announcing New RfA Data 08/94C l
5.
Present High Bumup 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 lasue* and Plan of Action 03/95C*
01/96C t
11.
Review industry (NEI) Response 09/95C j
12.
Assess Effects on Design Basis Accidents of Reduced Failure Threshold 09/95C for High Burnup Fuel j
13.
Committee on the Safety of Nuclear installations Snecialists Meetina on 09/95C the Transient Behavior of Heah Bumun Fuel 14.
CNRA (OECD) Committee on Nuclear Regulatory Activities and CSNI 11/95C annual meetings.
15.
Issue Letter to NEl Assessing Industry Actions (Vendor /EPRI response to 11/96T IN) 16.
Water Reactor Safety information Meetings (High Burnup session) 10/95C i
Core Performance lasues Workshop 10/96T 17.
RES Briefs ACRS and Completes Response to NRR User Need Letters 04/96C 02/97T 18.
Complete Review of Available Fuel Transient Data Relevant to Design 12/96T Basis Event 19.
Develop Interim Acceptance Criteria (e.g., based on cladding oxide) 12/96T 20.
lasue GL to Define Interim Critoria and Request post-LOCA Evaluation 02/97T 21.
Establish Schedule for LOCA Resolution and Final Assessment 010/96T Determme Need for Further Regulatory Action
- RES hee priontired as Generic leave #17o NUREG4933.
58 i
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Descriotion: The action plan covers assessment of fuel performance for high bumup fuel and j
evaluation of the adequacy of SRP licensing acceptance criteria.
Historical Backaround Recent experimental data on performance of high bumup (> 50 GWd/MTU)
I under reactivity insertion conditions became available in mid-1993. The unov,.ctedly 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 assumptions in the SRP. As a result, the Office of Noclear Reactor Regulation (NRR) was requested to prepare an action plan, in coordination with the O'tfice of Nuclear Regulatory Research (RES).
Procosed Actions: After a preliminary safety assessment was performed, an action van was developed, to include a user need letter to RES and the issuance of contracts to assess su aspects of the high burnup fuel issue. Concurrently, meetings would be scheduled with the non-domestic experimenters and regulatory authorities to discuss the experimental data and to assess potential consequences and regulatory actions. 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 transeent data, relevant to design basis events, NRR/RES would define I
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 Bumup Fuel," dated September 13,1994, including IN 94-64, ' Reactivity insertion Transsent and Accident Limits for High Bumup Fuel,' dated August 31,1994.
Commission Memorandum from James M. Taylor,
- Reactivity Transients and Fuel Damege Criteria for High Burnup Fuel," dated November 9,1994, including an NRR safety assess"nont and the joint NRR/RES action plan.
Reaulatory Assessment There is no immediate safety issue, because of the low to medium bumup in currently operating cores. Since the fuel failure threshold declines with increasing bumup, the licensing basis design acceptance criteria may need to be redefined as a function of burnup. The end product of the plan will determine the need for regulatory action and will establish and define the need for further action on exter:ded burnup cycles and high bumup fuel issues.
Current Status: An ACRS subcommittee meeting on the status of RES contractor programs was held in 4/96. An NEl letter summarizing the industry position was received in April, and the EPRI report supporting this position was sent by NEl on 9/20/96. These documents will be reviewed to prepare the NRR response. A Commission paper on the status of the High Burnup issue and planned actions has been prepared by NRR, has been reviewed by RES, and is in concurrence.
NRn Technical Contacts: Laurence Phillips, NRR/DSSA/SRXB, 415-3232 Shih-Liang Wu, NRR/DSSA/SRXB, 415-3284 Edward Kendrick, NRR/DSSA/SRXB, 415 2891 BES Contact-Ralph Meyer, RES/ DST /RPSB, 415-6789 59
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RRG TOPIC AREA 55: CYCLE SPECIFIC PARAMETER LIMITS IN TECH SPECS AND GENERIC LETTER 88-16 REVISION TAC Nos. M89033 and M85023 Last Update: 9/24/96 j
MILESTONES DATE (T/C) 1.
Complete draft guidance for GL 88-16 revision 8/94C 2.
Office concurrences on GL (NRR/OGC/RES/OC) n/a 3.
Contractor report received on reload report content 6/94C 4.
Complete draft guedance on contents of reload package (Reg. Guide) 9/94C and GL 83-11 revision 5.
Office concurrences on GL 83-11 revision 9/95C 6.
CRGR concurrence on GL 83-11 revision 10/95C 7.
EDO concurrence on GL 83-11 revision n/a 8.
Pubksh proposed GL 83-11 revision for public comment 10/25/95C 9.
Receive public comments on GL 83-11 revision 12/11/95C 10.
Office concurrence on GL 8311 revision TBD 11.
CRGR concurrence on GL 83-11 revision TBD 12.
EDO concurrence on GL 8311 revision TBD 13.
Publish GL 8311 revision TBD Brief Descriotson* This item recommended actions to reduce schedule and resource requirements for the NRC's review of reactor core reloads and the reload analysis methodology.
l Historical Backoround: The objective of this task is to respond to the Regulatory Review Group (RRG) Item #55. The RRG recommendations were to provide quicker review of core reload codes and to revise current Tech Specs to permit changes in accordance with approved core topical reports to take advantage of improved analyses without a license amendment by revising Generic Letter 8GL) 88-16 (Core Operating Limits Report (COLR) Guidance. The task was subsequently revised to address the first recommendation only by preparing a supplement to GL 8311 (Licensee Qualification for Performing Safety Analyses).
Pronosed Actions Prepare a supplement to GL 8311 which presents criteria intended for licensees who wish to perform their own licensing analyses using previously approved methods. By complying with these criteria, the licensee would eliminate the need to submit a topical report qualifying its use of a previously approved methodology.
l Orioinatina Document: Regulatory Review Group Topic Area item #55, Cycle Specific Parameter Limits in Tech Specs and Generic Letter 88-16 Revision.
Ranulatory Assessment This regulatory action has no safety impact on operating plants: it is intended to reduce resources required for methodology reviews.
60 l
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l Current Status: The proposed supplement to GL 8311 was published for comment in the Federal Register on October 25,1995. The comment period expired December 11,1995. A final package has been developed. However, because of recent issues regarding improper application of l
approved methods by licensees, as well as increased complexities in core reload analyses due to mixed core designs, the reduction in staff oversight is not justified. Therefore, issuance of the supplement to GL 83-11 is being cancelled and RRG ltem #55 will be closed out.
NRR Technical
Contact:
Larry Kopp, SRXB, 415-2879 NRR Lead PM:
Steve Bloom, DRPW, 415-1313 References Generic Letter 83-11 (February 8,1983) and Federe/ Register Notice 60 FR 54712 (October 25,1995).
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WOLF CREEK DRAINDOWN EVENT: ACTION PLAN TAC Nos.: M91621, M92635, M93568 Last Update: 09/26/96 Lead NRR Division:DSSA MILESTONES DATE (T/C) 1.
Draft Generic Letter 11/95(C) 2..
Issue Supplement to IN 95-03 03/96(C) 3.
Complete Draft Tl/ lasue to the Regions for Comments 10/96(T) i 4.
Generic Letter to be Concurred by CRGR/ Letter issued 10/96(T) 5.
Receive Regional Comments on Tl 12/96(T) 6.
Complete Evaluation of the Responses to the Generic Letter 03/97(T) 7.
Issue Tl 03/97(T) 8.
Complete inspections (As necessary) 06/97(T)
Descnotion: The objective of this action plan is to collect and evaluate information from the licensees regarding plant system configurations and vulnerabilities to draindown events. A 10 CFR 50.54(f) letter will be used to gather the information.
L Historical Backaround: On September 17,1994, the Wolf Creek plant experienced loss of reactor coolant system (RCS) inventory, while transitioning to a refueling shutdown. The event occurred when operators cycled a valve in the train A side of the RHR system cross-connect line following maintenance on the valve, while at the same time establishing a flow path from the RHR system, train B, to the refueling water storage tank for reborating train B. The failure of the reactor operating staff to adequately control two incompatible activities resulted in transferring 9200 gallons of hot RCS water to the lWYST in 66 seconds.
The Wolf Creek event represents a LOCA with the potential to consequentially fail all the ECCS pumps and bypass the containment. Another important feature of this event is the short time available for corrective action. Based upon calculations by the licensee and the staff, it is estimated that if the draindown had not been isolated within 3-5 minutes, not positive suction head would have been lost for all ECCS pumps, and core uncovery would follow in about 25-30 minutes.
This event represents a PWR vulnerability which was not previously recognized.
Proposed Actions: Specific actions of this generic action plan are: (1) issue IN 95-03 issued January 12,1995: and supplement to IN 95-03 issued March 25,1996, (2) Request all PWR licensees, via an information gathering (10 CFR 50.54(f)) Generic Letter (GL), to provide i 'ormation on draindown vulnerabilities and the measures they implemented to diminish the probability of a draindown.
Onainatino Document: AEOD/S95-01, " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994".
62
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Renulatory Assessment: The staff performed an evaluation of the probability for event initiation and of the conditional core damage probability. The value of this probability for core damage along with licensee awareness for this scenario makes tha risk for continued PWR operation acceptably j
small.
Current Status: Information Notice IN 95-03 has been issued. Information Notice Supplement has also been issued.
l 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.
- Action Plan dated October 20,1995 e
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GENERIC COMMUNICATION AND COMPLIANCE ACTIVITIES a
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I 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic. Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact Date Title Description
- LTD = Associate Director for Projects
- LTB - Technical Specifications Branch M91404 GL JWShapaker 1/15/97 T GL: Administrative Controls Section Line item improvement, guidance on revising the admin controls section of T.S.
M92544 GL JWShapaker 3/12/97 T GL: Design Features Technical Guidance to revise the design features Specifications section of T.S. (line item improvement)
- LTD - Division of Engineering
- LTB - Civil Engineering and Geosciences Branch M92553 LT RABenedict 12/15/96 T Investigate Impact of Failure of Certain steel framing members failed in SMRFs (During Northridge EQ) to NPP earthquake. Determine if same construction Steel Structures used in other plants.
M94293 GL JWShapaker 1/15/97 T GL: NRC Preliminary Findings Develop a GL to advise licensees that the use Related To The Use Of Reduced of reduced seismic criteria for temporary Seismic Criteria For Temporary conditions may involve unreviewed safety l
Conditions.
questions and staff review may be needed.
M94861 lii RABenedict 11/30/96 T IN: Liner Plate Corrosion in Corroded liner might be weakened against Concrete Containment post-accident leakage.
M95688 LT TAGreene 9/30/97 T Study of The Adequacy of Enveloped After completion of contract JCN J-2354, an Response Spectrum Method IN might be issued to caution operating plant licensees that under certain conditions ERS analysys method may not provide adequate estimates of seismic response of piping systems.
Page No.
2 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact DATE Title Description
- LTB - Electrical Engineering Branch M94841 IN ENFields 10/30/96 T IN: Loss of Offsite Power and Develop IN to discuss loss of offsite power Reactor Trip with One of Two EDGs and reactor trip with one of two EDGs Unavailable at Catawba Unit 2 unavailable at Catawba Unit 2.
r M95215 LT DLSkeen 8/1/97 T Charging / Discharging of Study and interact with the industry group on I
Safety-Related AT&T Round Cell the AT&T round cell battery degradation i
Batteries problems.
M96076 LT EJBenner 10/31/96 T Cracking of Phenolics in Reactor Evaluate need for additional GC beyond an IM Trip Breakers on cracking of phenolics in reactor trip breakers M96611 IN JRTappert 11/30/96 T IN: Improper Groundin'g Results in Alerts to licensees to potential problems Fire at Palo Verde concerning component grounding problems which i
could result in simultaneous fires.
i M96616 GL DLSkeen 1/29/97 T GL: Medium-Voltage Circuit Breaker GL to address continued breaker problems
[
Failures because of refurbishment practices, licensee i
maintenance, and inadequate review of industry operating experience.
l M96055 LT CVHodge 1/31/97 T GE Magne-Blast Breaker Failure Risk insight for common-mode failures
- LTB = Materials and Chemical Engineering Branch M95279 GL JWShapaker 12/31/96 T GL: Modification of The Extending to operating reactor licensees, on Requirements for Post-Accident voluntary basis, relaxations in PASS program Sampling System requirements.
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10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT i
Open Generic Communication and Compliance Activities I
Sorted by Lead Technical Division and Branch
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l TAC Type Contact Date Title Description i
M95280 GL JWShapaker 11/15/96 T GL: Primary Water Stress Corrosion Identification of need to implement program Cracking of Control Rod Drive for assuring timely inspection of PWR vessel Mechanism And Other Vessel Head head penetrations.
l Penetrations M95290 GL JWShapaker 12/20/96 T GL: Degradation of Steam Generator Identification of steam generator internals r
Internals degradation mechanisms based on foreign i
reactor operating experience.
M95373 GL JWShapaker 11/29/96 T GL: Implementation of App. VIII of Discusses the need for lecensees to adopt the i
Sec XI of The 1995 Edition of The Appendix VIII to improve the quality and ASME Boiler And Pressure Vessel confidence level of inservice inspections.
r Code M95444 LT TAGreene 6/15/97 T Lead Technical Review - Induction Cracking has been found in several utilities' Heat Stress Improvement for austentic stainless steel piping which had Stainless Steel Piping been subjected to IHSI in the 1980's. Staff concerns include that IHSI may not have been properly applied, f
M96074 GL NKHunemuller 10/30/96 T AL: Notice of Technical Guidance Informs licensees of technical guidance Issued in USNRC Inspection Manual included in the NRC Inspection Manual on ASME Regarding 10 CFR 50.55a/ASME Code Sec III and XI, as it relates to code issues interpretations, use of engineering judgement i
and flaw evaluation.
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4 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact Date Title Description M96401 GL JWShapaker.
1/17/97 T GL: Steam Generator Tube Inspection Informs licensees of the importance of Techniques performing steam generator tube inservice i
inspections using qualified techniques and requests that licensees implement described l
actions.
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- LTB - Mechanical Engineering Branch M93841 LT JMSebrosky 10/31/96 T Implications of Target Rock 2-Stage Evaluate safety inplications of leakage en i
SRV Pilot Leakage valve operability and adequacy of leak detection.
M95443 IN WFBurton 11/30/96 T IN: Thermal Fatigue Cracking In Alerts licensees an event of weld crack on l
Residual Heat Removal Lines safety injection line at Sequoyah.
M96073 IN EJBenner 1/31/97 T IN: Concerns with Dry Cask Loading Alerts licensees to several identified I
and Unloading Procedures problems with procedures for the loading and unloading of spent fuel storage casks.
i M96354 LT NKHunemuller Containment Recirculation Spray and Millstone 3 determined that the containment Quench Spray Piping Outside Design recirculation spray and quench spray piping Basis and supports could be subjected to higher i
accident temperatures than those previously assumed in the design basis.
1 M96614 LT TKoshy 12/3/96 T LPSI Pump Mission Time When the RCS pressure remains higher than l
i 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 pump capability for the require mission time.
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5 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact Date Title Description
- LTD - Division of Reactor Controls and Human Factors
- LTB - Instrumentation and Controls Branch M96322 IN CDPetrone 10/31/96 T IN: Problems Associated with Develop IN to alert licensees to recent Testing or Tuning of Digital reactor transients, reactor trips, and Control Systems While at Power engineered safety feature actuations caused by testing, tuning, or resetting of digital control systems while at power.
- LTB - Operator Licensing Branch M94840 GL JWShapaker 12/31/96 T GL: Changes in The Operator Notify licensee of NRC's decision to change Licensing Program and Issuance of the operator licensing process by giving Rev. 8 of NUREG-1021 licensees the opportunity to directly participate in the preparation of draft written examinations and operating tests for NRC use.
- LTD - Division of Reactor Program Management
- LTB = Emergency Preparedness and Radiation Protection Branch l
M91620 GL JWShapaker 3/12/97 T GL: Revision to Augmentation Ensuring adem! ate staffing for emergencies.
Staffing Levels For Nuclear Power Plant Emergencies l
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6 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact Date Title Description l
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- LTB = Events Assessment and Generic Communications Branch M91544 GL JWShapaker 11/15/96 T GL: Defining Info in Monthly Reducing reporting requirements to the Operating Report Required by Tech minimum needed by the staff (part of RRG).
Specs M95686 IN RABenedict 10/30/96 T IN: Main Steam Safety Valve Develop IN to alert licensees of the Failure to Reseat Caused by an experience from ANO-1 event in which a steam Improperly Installed Release Nut generator dryout occurred due to improper maintenance of main steam safety valve.
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- LTB = Safeguards Branch M96535 IN DLSkeen 10/31/96 T IN: Licensee Response to Indication Information Notice to address the recent of Tampering, Vandalism, or security enents at St Lucie and Beaver Malicious Mischief Valley.
- LTD - Division of Systems Safety and Analysis
- LTB = Containment Systems and Severe Accident Branch M96400 IN CVHodge 1/31/97 T IN: Revised Calculation of High Recalculation to support power uprate at Energy Line Break from Reactor Monticello significantly increased mass Water Cleanup Piping release rate for HELB. RWCU break outside containment exceeds MSLB under certain conditions.
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10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch j
TAC Type Contact Date Title Description i
M96536 IN EJBenner 10/9/96 T IN: Inadequate NPSH of ECCS and Discusses recent discoveries by licensees i
Containment Heat Removal Pumps that the available NPSH requirements for ECCS under Design Basis Accident and containment heat removal pumps may not be Conditions adequate under all postulated design basis senarios.
M96537 GL JWShapaker 2/22/97 T GL: Assurance of Sufficient NPSH Notifies licensees about a safety-significant for ECCS and Containment Heat issue that could affect the ability for Removal System Pumps long-term core cooling and containment heat removal under accident conditions and which has generic implications.
- LTB = Plant Systems Branch M80296 LT TAGreene 9/30/97 T Generic Communications - Assessment Development of staff NUREG or other of Turbine Failure at Vandellos 1 publication to document turbine building fire issues for U.S. plants in light of Vandellos fire.
M91323 LT NKHunemuller 12/27/96 T Reactor Water Cleanup (RWCU) Study Review of the effects of an unisolated RWCU i
in Response to ACRS Concern break at several BWR's. Result of ACRS concerns during the review of the ABWR f
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8 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch i
TAC Type Contact Date Title Description M93335 LT WFBurton 4/30/97 T Nain 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.
s M95871 IN NKHunemuller 12/2/96 T IN: Emergency Lighting Issues Develop IN to alert licensees to potential problems regarding emergency lighting for Plant areas needed for operation of post-fire safe shutdown equipment and in the access and egress routes.
M96260 IN NKHunemuller 11/30/96 T IN: Potential for Exceeding Develop IN to alert licensees of the f
Compartment Pressures due to potential for ineffective blowout panels or Inadequate Blowout Panels or Vent vent areas associated with HELBs outside Areas containment. The inoperability may be due to Permanent or temporary interferences or improper modifications.
1 I
M96437 LT JRTappert 11/30/% T Vulnerability of EDGs to Fuel Follows up on Calvert Cliffs root cause Oil / Lubricating 011 Incompatibility analysis that identified fuel all/ lubricating I
oil incompatabilities.
If validated as potentially generic issue, an IN may be issued.
M96502 LT DLSkeen 12/30/96 T Potential for Air Regulator Detemine if industry response to IN 88-24 f
Failures to Overpressurized and GL 91-15 is adequate, in light of 9/6/96 Safety-Related SOVs Nillstone 3 discovery of susceptible SOVs.
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Paga N3.
9 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by lead Technical Division and Branch TAC Type Contact Date Title Description
- LTB - Reactor Systems Branch M80326 LT SSKoenick 1/31/97 T Accumulation of Volume Control Tank Not a new issue, there have been several i
Cover Gass in ECCS Piping Connected generic communications already issued. SRX8
+
to the Charging System.
would like to close this out by meno.
j M91599 GL JWShapaker 6/30/97 T GL 83-11 Supp: Licensee PART OF A TASK ACTION PLAN -- Provides Qualification For Performing Safety alternative means of licnesee qualification Analyses in Support of Licensing for perfoming sanalyses using generically Actions approved methods.
M92635 GL JWShapaker 10/31/96 T GL: Reactor Coolant Inventory loss Loss of ECCS function due to steam voiding in and Potential loss of Emergency RWST line to suction of ECCS pumps due to Mitigation Functions While Shutdo loss of RCS inventory in Mode 4 (Wolf Creek).
M94565 LT DLSkeen 11/1/96 T Slow Scram Solenoid Pilot Valves Scram solenoid pilot valves with viton Caused by Viton Diaphrages diaphrages showing degraded scram times within 6-8 months. Currently tracking licensee response to RRG recommendations.
M95278 GL JWShapaker 12/31/96 T GL: Use of Thermal-Hydraulic Codes Discusses the fact that a computer code has for Licensing Applications been developed and assessed primarily with NRC funds does not per se mean that it is acceptable as a licensing code.
l M96191 IN RABenedict 11/15/96 T IN: Plant Specific E0Ps Contain Alerts licensees the problem with ICS/FWCS l
Inadequate Technical Info to interactions. The proposed IN was prompted
[
Accomplish Timely and Effectively by the ANO-1 5/19/96 event.
Feeding of OTSG
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10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Open Generic Communication and Compliance Activities Sorted by Lead Technical Division and Branch TAC Type Contact Date Title Description M96192 IN WFBurton 10/31/96 T IN: ECCS Throttle Valves May High differential pressure across ECCS Degrade Due To Cavitation Induced throttle valves during LOCA could cause pump Erosion During LOCA runout flow and subsequent ECCS pump damage 1
M96355 LT SSKoenick 11/20/96 T Concerns Regarding Siemens Large Changes to large break ECCS model may have Break LOCA ECCS Evaluation Model resulted in significant changes in calculated 1
peak clad temperatures for some plants.
l M96361 IN SSKoenick 11/13/96 T IN: Boron Dilution and Other Excess boron dilution at Byron due to r
l Activities Affecting Reactivity non-representative sampling techniques. Will l
incorporate WNP event.
l M96615 LT TKoshy 12/3/96 T Boron Precipitation in B&W Reactors Design bases concern on active means of I
preventing boron precipitation following a i
LOCA.
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I 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Generic Communication and Compliance Activities Added Since the Last Public Report (July 1996)
TAC Type Contact Tech Branch Date Title Reason Added M95871 IN NKHunemuller Plant Systems 12/2/96 T IN: Emergency Lighting The EAP authorized development of IN at Branch Issues its 7/2/96 meeting.
M96055 LT CVHodge Electrical 1/31/97 T GE Magne-Blast Breaker 7/23/96: The EAP authorized a long-tern Engineering Failure follow up for this issue.
Branch M96073 IN EJBenner Mechanical 1/31/97 T IN: Concerns with Dry The EAP authorized development of IN Engineering Cask Loading and at its 7/16/96 meeting.
Branch Unloading Procedures M96074 GL NKHunemuller Materials and 10/30/96 T AL: Notice of Technical This GC has been changed to an AL. per Chemical Guidance Issued in USNRC e-mail from JWS to PXW,10/2/96*
Engineering Inspection Manual Branch Regarding 10 CFR 50.55a/ASME Code issues M96076 LT EJBenner Electrical 10/31/96 T Cracking of Phenolics in The EAP authorized a long-term follow Engineering Reactor Trip Breakers up for this issue (beyond an IN) at its Branch 7/16/96 meeting..
M96191 IN RABenedict Reactor Systems 11/15/96 T IN: Plant Specific E0Ps The EAP authorized development of IN at Branch Contain Inadequate its 7/23/96 meeting.
Technical Info to Accomplish Timely and i
Effectively Feeding of OTSG 4
Page No.
2 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Generic Communication and Compliance Activities Added i
Since the Last Public Report (July 1996)
?
r TAC Type Contact Tech Branch Date Title Reason Added M96192 IN WFBurton Reactor Systems 10/31/96 T IN: ECCS Throttle Valves The EAP authorized development of IN at Branch May Degrade Due To its 7/23/96 meeting.
Cavitation Induced Erosion During LOCA M96196 RABenedict 7/30/96 C IN: Main Steam Safety This TAC was inadvertently opened. This r
Valve Failure To Reseat TAC is the same as M95686.
Caused By An Improperly
+
Installed Release Nut M96260 IN NKHunemuller Plant Systems 11/30/96 T IN: Potential for The EAP authorized development of IN at Branch Exceeding Compartment its 8/6/96 meeting.
Pressures due to Inadequate Blowout Panels I
or Vent Areas M96322 IN CDPetrone Instrumentation 10/31/96 T IN: Problems Associated The EAP authorized development of IN at and Controls with Testing or Tuning of its 8/14/96 meeting.
Branch Digital Control Systems While at Power t
M96354 LT NKHunemuller Mechanical Containment Recirculation The EAP authorized long-term followup Engineering Spray and Quench Spray for this issue at its 8/20/96 meeting.
Branch Piping Outside Design Basis l
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3 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Generic Communication and Compliance Activities Added Since the Last Public Report (July 1996)
TAC Type Contact Tech Branch Date Title Reason Added M96355 LT SSKoenick Reactor Systems 11/20/96 T Concerns Regarding The EAP authorized long-term followup Branch Siemens large Break LOCA for this issue at its 8/20/96 meeting.
ECCS Evaluation Model M96361 IN SSKoenick Reactor Systems 11/13/96 T IN: Boron Dilution and The EAP authorized development of IN at Branch Other Activities its 7/9/96 meeting.
Affecting Reactivity M96400 IN CVHodge Containment 1/31/97 T IN: Revised Calculation The EAP authorized development of IN at Systems and of High Energy Line Break its 8/27/96 meeting.
Severe Accident from Reactor Water Branch Cleanup Piping M96401 GL JWShapaker Materials and 1/17/97 T GL: Steam Generator Tube The EAP authorized development of GL at Chemical Inspection Techniques its 8/27/96 meeting.
Engineering Branch M96437 LT JRTappert Plant Systems 11/30/96 T Vulnerability of EDGs to The EAP authorized long-term follow up Branch Fuel Oil / Lubricating 011 of this item.
If validated as Incompatibility potentially generic issue, an IN will be issued.
M96502 LT DLSkeen Plant Systems 12/30/96 T Potential for Air The EAP authorized LT follow-up of this Branch Regulator Failures to issue at its 9/10/96 meeting..
Overpressurized Safety-Related SOVs l
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l 10/15/96 t
PUBLIC OCTOBER 19% DIRECTOR'S MONTHLY STATUS REPORT Generic Communication and Compliance Activities Added
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Since the Last Public Report (July 1996)
TAC Type Contact Tech Branch Date Title Reason Added f
f i
M96535 IN DLSkeen Safeguards 10/31/96 T IN: Licensee Response to The EAP authorized development of IN at
(
Branch Indication of Tampering, its 9/17/96 meeting Vandalism, or Malicious Mischief F
l M96536 IN EJBenner Containment 10/9/% T IN: Inadequate NPSH of The EAP authorized development of IN at
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l Systems and ECCS and Containment Heat its 9/17/96 meeting Severe Accident Removal Pumps under Branch Design Basis Accident Conditions M96537 GL JWShapaker Containment 2/22/97 T GL: Assurance of The EAP authorized development of GL at Systems and Sufficient NPSH for ECCS its 9/17/96 meeting.
Severe Accident and Containment Heat Branch Removal System Pumps M96611 IN JRTappert Electrical 11/30/96 T IN: Improper Grounding The EAP authorized development of IN at i
Engineering Results in Fire at Palo its 9/24/96 meeting.
Branch
. Verde M96614 LT TKoshy Mechanical 12/3/96 T LPSI Pump Mission Time The EAP authorized LT follow-up of this I
Engineering issue at its 9/24/96 meeting.
Branch M96615 LT TKoshy Reactor Systems 12/3/% T Baron Precipitation in The EAP authorized LT follow-up of this Branch B&W Reactors issue at its 9/24/96 meeting.
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5 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Generic Communication and Compliance Activities Added Since the Last Public Report (July 19%)
TAC Type Contact Tech Branch Date Title Reason Added M96616 GL DiSkeen Electrical 1/29/97 T GL: Medium-Voltage Authorized development of GL by RLD*
Engineering Circuit Breaker Failures acting for AEC, on 9/26/96.
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10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Generic Comunication and Compliance Activities Closed I
Since the Last Public Report (July 1996)
TAC Type Contact Tech Branch Date Title Reason Closed M93336 GL JWShapaker Operator 8/20/96 C GL: Exemption For TAC closed per memo from SARichards to Licensing Applicants For the Senior AEChaffee 8/14/96.
Branch Reactor Operator License Limitad to Fuel Handling (LSRO) 993706 GL JWShapaker Mechanical 9/18/96 C GL: Periodic Verification GL 96-05 issued 9/18/96 Engineering of Design-Basis Branch Capability of t
Safety-Related Motor-Operated Valves M93707 GL JWShapaker Civil 9/3/96 C GL: Plant Shutdown TAC closed per ECGB's request (RRothman t
Engineering and Criteria Following an e-mail to JWS 8/29/96)
Geosciences Earthquake Branch M93979 IN JRTappert Special 9/16/96 C IN 92-68, Supp:
IN 92-68, Sup I, issued 9/16/96" Inspections Potentially Substandard 1
t Branch Slip-On, Welding Neck, and Blind Flanges 1
M94794 IN ENFields Special 7/25/96 C IN: Deficiencies in IN 96-40 issued 7/25/96.
Inspections Material Dedication and l
Branch Procurement Practices and Vendor Audits m
Page No.
2 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Generic Comunication and Compliance Activities Closed Since the Last Public Report (July 1996)
TAC Type Contact Tech Branch Date Title Reason Closed M95074 IN DLSkeen Special 9/4/96 C IN: Problems with IN 96-50 issued 9/4/96 Inspections Westinghouse DHP Circuit Branch Breaker Levering-In Device i
l M95251 IN EJBenner Materials and 7/10/96 C IN96-09, Sup 1, Damage in IN 96-09, Sup 1, issued 7/10/96.
Chemical Foreign Steam Generator Engineering Internals Branch M95356 IN TAGreene Civil 9/20/96 C IN: Possible ANSYS Code Based on the 8/27/96 memo from GBigchi Engineering and Platform Dependency to AEChaffee, this TAC is closed.
It Geosciences is not necessary to issue an IN Branch regarding the Holtec's erroneous analysis results using the ANSYS
<omputer code.
M95374 IN SSKoenick Reactor Systems 7/26/96 C IN: Overpower due to 7.N 96-41 issued 7/26/96.
Branch large Reduction in Feedwater Temperature i
M95684 IN TAGreene Mechanical 8/20/96 C IN
'4 96-48 issued 8/21/96.
l Engineering Motor-Operated Valve l
Branch Performance Issues l
l M95685 IN TJCarter Instrumentation 8/5/96 C IN: Multiple Unexpected IN 96-42 issued 8/5/96.
l and Controls -
Opening of Safety Relief l
Branch Valves 1
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3 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Generic Communication and Compliance Activities Closed Since the Last Public Report (July 1996)
TAC Type Contact Tech Branch Date Title Reason Closed M95720 IN DLSkeen Electrical 8/2/96 C IN: Failures of GE IN 96-43 issued 8/2/96.
Engineering Magne-Blast Circuit l
Branch Breakers l
M96075 IN EJBenner Electrical 8/5/96 C IN: Cracking of Phenolics IN 96-44 issued 8/5/96.
l Engineering in Recator Trip Breakers Branch j
M96194 IN DLSkeen Electrical 8/13/96 C IN: Zinc Plating in GE IN 96 #5 issued 8/12/96.
Engineering Magne Blast Breakers Branch M96195 IN JRTappert Plant Systems 8/12/96 C IN: Post-Accident Failure IN 96-45 issued 8/12/96.
Branch of Containment Coolers M96259 IN TJCarter Containment 8/20/96 C IN: Thermally Induced IN 96-49 issued 8/20/96.
Systems and Overpressurization Severe Accident Branch M96276 GL JWShapaker Plant Systems 9/27/96 C GL: Assurance of GL 96-06 issued 9/30/96 Branch Equipment Operability and Containment Integrity during Design Basis Accident
Page No.
4 10/15/96 PUBLIC OCTOBER 1996 DIRECTOR'S MONTHLY STATUS REPORT Generic Communication and Compliance Activities Closed Since the Last Public Report (July 1996)
TAC Type Contact Tech Branch Date Title Reason Closed M96323 JWShapaker Materials and 9/6/96 C GL: TS Revisions TAC closed per request from EMCB (Treed Chemical Supporting Implementation e-mail to JWShapaker 8/21/96).
I Engineering of 10 CFR 50.xx " Steam Guidance on the development of TS to Branch Generator Tube Integrity comply with the "SG rule" will be for Operating NPPs" included in the FR notice, therefore, there is no need for a GL.
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October 31c 1996 DISTRIBUTION for NRR Director's Status Report Central Filei Public Document Room i
PGEB R/F TTMartin DMatthews FAkstulewicz 1
EMcKenna PWen EWang Mr. William Rasin, Vice President Technical / Regulatory Division j
Nuclear Energy Institute 1776 " Eye" Street Washington, D.C.
20006 Mr. R. P. LaRhette Institute of Nuclear Power Operations 700 Galleria Parkway i
Atlanta, Georgia 30339-5979 Mr. Lee Watkins Assistant Manager For High Level Waste U.S. DOE PO Box: A Aiken, South Carolina 29892 Mr. R. W. Barber Safety and Ouality Assurance, DOE 270 Corporate Center (E-853) 20300 Century Blvd.
Germantown, MD 20874 Mr. S. Scott Office of Nuclear Safety, DOE 3
Century 21 Building (E-H72) f.
I fU I
19901 Germantown Road Germantown, MD 20874-1290 Mr. Bob Borsum 1700 Rockville Pike, Suite 525 C+N H j
Rockville, MD 20852 Y Wy S( 6e4/'CJl Ms. Norena G. Robinson a
ricofy Licensing Technician Nebraska Public Power District y
Cooper Nuclear Station P. O. Box 98 1200G3 Brownsville, NE 68321 y
(
OFFICE PG :D PM PGEB:DRPM P
RPM iG B/bRPM
- D PM D
NAME EV9ang PWen ((W EMcKenna F%sstulewicz
'Dheatthe9 ts' TTMArtiQ DATE 10Af/96 10hq/96 10/,p//96 1N/96 10/'2 @ 6 10/1\\/9 6 OFFICIAL RECORD COPY
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