ML20063H746
| ML20063H746 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1435, NUREG-1435-S03, NUREG-1435-S3, NUDOCS 9402220173 | |
| Download: ML20063H746 (165) | |
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NUREG-1435 Supplement 3 Status of Safety Issues at Licensed Power Plants TMI Action Plan Requirements Unresolved Safety Issues Generic Safety Issues Other Multiplant Action Issues l
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U.S. Nuclear Regulatory Commission l
Office of Nuclear Reactor Regulation l
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5 9402220173 931231
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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
1.
The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2.
The Superintendent of Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, DC 20402-9328 3.
The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu-ments and correspondence.
The following documents in the NUREG series are available for purchase from the GPO Sales Program; formal NRC staff and contractor reports, NRC-sponsored conference proceedings, international agreement reports. grant publications, and NRC booklets and brochures. Also available are regulatory guides, NRC regulations in the Code of Federal Regulations, and Nu-clear Regulatory Commission issuances.
Documents available from the National Technical Information Service include NUREG-s $ ries reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literature items. such as books. journal articles, and transactions. Federal Register notices, Federal and State legislation, and congressional reports can usually be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations, and non-NRC con-ference proceedings are available for purchase f rom the organization sponsoring the publica-tion cited.
Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Administration, Distribution and Mail Services Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, for use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American Na-tional Standards Institute,1430 Broaoway, New York, NY 10018.
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i-4 NUREG-1435 Supplement 3
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Status of Safety Issues at Licensed Power Plants 1
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TMI Action Plan Requirements i
i Unresolved Safety Issues Generic Safety Issues i
Other Multiplant Action Issues l
j Manuscript Completed: November 1993 Date Published: December 1993 l
4 OITice of Nuclear Reactor Regulation l
U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
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ABSTRACT As part of ongoing U.S. Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSis). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30,1991, was published in December 1991.
The second annual supplement, which provided updated information as of September 30,1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMl Action Plan requirements, USIs, or GSis. This third annual NUREG report, Supplement 3, presents updated information as of September 30,1993.
The data contained in this supplement is produced using the NRC's safety issues management system (SIMS) database, which is maintained by the project management staff in the Office of Nuclear Reactor Regulation and by the staff in NRC's regions.
This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSis, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, "A Prioritization of Generic Safety Issues," which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees.
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CONTENTS ABSTRACT....................................................
iii EXECUTIVE SU M MARY..........................................
vil ABBREVIATIONS
...............................................xi 1
INTRO D U CTIO N............................................ 1 1.1 Backg round........................................... 1 1.2 Process and Accountability................................ 2 j
1.3 Definitions............................................. 4 2
THREE MILE ISLAND ACTION PLAN REQUIREMENTS............... 7 2.1 Implementation Status.................................... 7 2.2 Verification Status...................................... 11 2.3 Status by Plant........................................ 13 2.4 Status by issue........................................ 17 2.5 Conclusions.......................................... 31 3
UNRESOLVED SAFETY ISSUES............................... 33 3.1 Implementation Status................................... 33 3.2 Verification Status................
.....................38.
3.3 Status by Plant........................................ 41 3.4 Status by issue.........
..............................45 3.5 Conclusions.........
................................49 4
GENERIC SAFETY lSSU ES.................................. 51 4.1 Implementation Status................................... 51 4.2 Verification Status...................................... 58 4.3 Status by Plant........................................ 63 4.4 Status by issue............. -........................... 67 4.5 Conclusions.......................................... 71 5
OTHER MULTIPLANT ACTION ISSUES.......................... 73 i
5.1 Implementation Status................................... 73 5.2 Verification Status...................................... 90 5.3 Status by' Plant........................................ 93 5.4 Status by Issue........................................ 97 5.5 Conclusions.........................................
111 4
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1 APPENDICES A
Usting of Unimplemented TMI Items by Issue.................. A-1 B
Usting of Unimplemented USl items by issue.................. B-1 C
Usting of Unimplemented GSI Items by issue.................. C-1 D
Usting of Other Unimplemented MPA Items by issue............. D-1 i
FIGURES 2.1 TMI Action Plan Requirements -
Implementation Status at Ucensed Plants...................... 8 2.2 Projected Schedules for Remaining TMI Items................... 9 3.1 Unrescived Safety issues - Implementation Status at Ucensed Plants. 34 3.2 Summary of Three Unimplemented USIs...................... 36 '
4.1 Generic Safety issues - Implementation Status at Ucensed Plants
... 54 4.2 Summary of Five Unimplemented GSis....................... 56 4
5.1 Other MPA issues - Implementation Status at Ucensed Plants....... 76 5.2 Summary of Eleven Unimplemented MPAs.................... 77 TABLES 2.1 Summary of the Remaining TMI Items by Area..................
10 2.2 Summary of the Remaining TMI Items by Plant.................
12 7
2.3 Status of TMl Action Plan - Summary by Plant..,...............
14 4
2.4 Status of TMI Action Plan - Summary. by item..................
18 3.1 Summary of Unimplemented USl items by Plant
................35 3.2 Summary of USI Items Requiring Verification................... 40 4
3.3 Status of USis - Summary by Plant.......................... 42 3.4 Status of USIs - Summary by item.......................... 46 4.1 GSI Numbers and Corresponding SIMS Item Numbers............_. 53 4.2 Summary of Unimplemented GSI Items by Plant
................55 4.3 Temporary Instructions for Resolved GSis..................... 60 4.4 Summary of GSI Items Requiring Verification................... 61 4.5 Status of GSis - Summary by Plant.......................... 64 l
4.6 Status of GSis - Summary by item.......................... 68 5.1 SIMS Issue Numbers and Corresponding MPA Number
.......... 74 5.2 Summary of Unimplemented MPA Items by Plant................ 75 5.3 Temporary Instructions for Resolved MPAs.................... 91 5.4 Summary of Other MPA Items Requiring Verification.............. 92 5.5 Status of Other MPAs - Summary by Plant
....................94 5.6 Status of Other MPAs - Summary by item..................... 98
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EXECUTIVE
SUMMARY
I This U.S. Nuclear Regulatory Commission (NRC) report covers the implementation 2
and verification status of the Three Mile Island (TMI) Action Plan requirements, unresolved safety issues (USls), generic safety issues (GSis), and other multiplant action (MPA) issues not related to TMI Action Plan requirements, USls, or GSis at 109 licensed nuclear power plants. The implementation and verification status are current as of September 30,1993.
Backaround The implementation and verification status of TMl Action Plan requirements, USis, and GSis was initially reported in Volumes 1,2, and 3 of NUREG-1435, published in 1991.
The first annual supplement consolidated and updated the information given in the earlier three volumes; it was published in December 1991 and provided updated information as of September 30,1991. The second supplement was published in December 1992 and gave updated information as of September 30,1992.
Supplement 2 also gave the status of licensee implementation and NRC verification of other muttiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSis. This third annual report, Supplement 3, gives updated information as of September 30,1993. The data contained herein is a product of the NRC's Safety d
Issues Management System (SIMS), which is maintained by the project management staff in the Office of Nuclear Reactor Regulation and by the staff in NRC's regions.
The NRC has given significant attention to the quality review of TMI, USl, GSI, and other MPA implementation and verification data in SIMS.
Supplement 2 reported data on 110 plants. It included San Onofre 1 and Trojan, which are now permanently shut down. It did not include data for Comanche Peak 2, which is now fully operational. Supplement 3 reports data for 109 plants, including data for Comanche Peak 2, but excluding data for San Onofre 1 and Trojan.
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Three Mile Island Action Plan Reauirements 1
l Imolementation Status. More than 99 percent of the TMI Action Plan items have been implemented at 109 licensed plants. Of the 12,898 applicable items,12,837 have been completed or closed and only 61 remain open from an implementation standpoint.
About 38 percent of the remaining 61 open items are projected to be implemented by the end of calendar year 1994. As noted in previous supplements, some slippages have occurred in projected implementation dates. Delays in the restart of Browns Ferry Units 1 and 3 account for 34 of the 61 unimplemented items. All schedules for implementation of TMI Action Plan items with the exception of Browns Ferry 1, remain within the timeframe previously reported to the Commission and to Congress.
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From an issue perspective, four major areas account for about 77 percent of the 61 remaining items: detailed control room design review items (21), accident monitoring (10), plant safety parameter display system items (8), and B&O Task Force issues (8).
From a plant perspective, the TMI Action Plan requirements have been fully implemented or closed at 82 of the 109 licensed plants. Two plants, Browns Ferry Units 1 and 3, account for approximately 56 percent, or 34 of the 61 remaining items.
Of the remaining 25 plants, each has 1 remaining item to implement, with the exception of Browns Ferry 2 and Haddam Neck, which have 2 each.
Verification Status. Tis have been issued for 78 individual TMI requirements to provide guidance for the field verification of licensee implementation. Of the 5,992 TMI items requiring veri 5 cation,5,939 (99 percent) have been completed.
Unresolved Safety issues (USis)
Imolementation Status. Approximately 90 percent of the'USlitems have been implemented at licensed plants. Of the 1,787 applicable items,1,610 have been completed and 177 remain open from an implementation standpoint. On average, each plant has approximately 2 remaining items to implement, and no plant has more than 6 items to implement.
Three USIs (A-44, Station Blackout; A-46, Seismic Qualification of Equipment in Operating Plants; and A-47, Safety implications of Control Systems) account for 90 percent of the unimplemented items. These three USIs are in varying stages of NRC review and licensee implementation, as further described in Section 3.1 of this report.
Verification Status. Five Tis have been issued to provide guidance for the field verification of licensee implementation. These Tl designations correspond to USis A-7, Mark i Long-Term Program; A-9, Anticipated Transients Without Scram; A-24, Qualification of Class 1E Safety-Related Equipment; A-26, Reactor Vessel Pressure l
Transient Protection; and A-44, Station Blackout.
The requirements to perform field verifications have resulted in a total of 423 items to be verified at the 109 plants. As of September 30,1993, the NRC field verification had been completed on 292 (69 percent) of the required items.
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Generic Safety issues (GSis)
Imolementation Status. Approximately 94 percent of the applicable items associated with GSis have been implemented at licensed plants. Of the 2,621 applicable items, 2,463 have been completed and 158 remain open from an implementation standpoint.
On average, each plant has fewer than 2 items to implement, and no plant has more than 7 items to implement. Thirty-nine plants have implemented all applicable items i
related to GSis. Five GSIs account for 90 percent of the items for which implementation is incomplete. These GSis are specifically addressed in Section 4.1 of this report.
Verification Status. Eight Tis have been issued to provide guidance for the field verification of licensee implementation. Of the 1,176 GSI items requiring verification, 1,045 (89 percent) have been completed.
1 Other Multiolant Actions (MPAs) j imolementation Status. Approximately 87 percent of the applicable items associated with MPAs have been implemented at licensed plants. Of the 7,517 applicable items, 6,576 have been completed and 941 remain open from an implementation standpoint.
On average, each plant has approximately 9 remaining items to implement, and no 4
plant has more than 12 items to implement except Browns Ferry Units 1 and 3. Each l
of these units has 20 items. No plant has implemented all applicable items related to other MPAs. Eleven MPAs account for 85 percent of the items for which i
implementation is incomplete. These 11 MPAs as well as those with more than 3 open items are specifically addressed in Section 5.1 of this report.
Verification Status. Fifteen Tis have been issued to provide guidance for the field verification of licensee implementation. Of the 776 MPA items requiring verification, 589 (76 percent) have been completed.
Conclusions After a detailed review of the implementation and verification status of TMl Action Plan requirements, USis, GSis, and other MPAs, the NRC staff has drawn the following conclusions:
The NRC closure process for TMI Action Plan issues, USIs, GSis, and other MPAs 2
is adequate for protecting the public health and safety.
Licensees continue to make progress in implementing actions that are voluntary or that are imposed or requested by the staff. The framework exists to verify that open items are implemented in the future.
I The NRC continues to make progress in verifying the implementation actions that licensees reported complete.
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The schedule slipages related to implementing TMI Action Plan items do not pose.
a threat to the pu alic health and safety. The NRC staff will continue to maintain -
close oversight of the implementation actions and schedules proposed by the -
licensees to ensure that remaining TMI requirements are completed in accordance.
with regulatory requirements and within acceptable time frames.
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i ABBREVIATIONS ACRS Advisory Committee on Reactor Safeguards ATWS anticipated transient without scram BL bulletin B&O bulletins and orders B&W Babcock and Wdcox l
BWR boiling-water reactor BWROG Boiling Water Reactors Owners Group CE Combustion Engineering CPI containment performence mprovement CRGR Committee for the Review of Generic requirements J
DBA design-basis accident ECCS emergency core cooling system GIP generic implementation procedure GL generic letter
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GSER generic safety evaluation report GSI generic safety issue HPCI high-pressure coolant injection i
IGSCC intergranular stress corrosion cracking IN information notice (NRC)
IPE individual plant examination IPEEE individual plant examination of external events IST inservice testing LCO limiting conditions for operation MOV motor-operated valve MPA multiplant action NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation (NRC)
NUMARC Nuclear Management and Resource Council ODCM Offsite Dose Calculation Manual PCP process control program PORV power-operated relief valves PWR pressurized-water reactor PZR pressurizer 4
RCIC reactor core isolation cooling RCS reactor coolant system RES Office of Nuclear Regulatory Research (NRC)
RETS radioactive effluent technical specifications RG regulatory guide RHR residual heat removal RWCU reactor water cleanup SBL supplement bulletin SIMS safety issues management system
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l SOER significant operating experience report SPDS safety parameter display system SOUG Seismic Qualification Utility Group l
SRM staff requirements memorandum SSER supplementary safety evaluation report Tl temporary instruction TMI lhree Mile Island l
l TS technical specifications l
USI unresolved safety issue VIB vital instrument bus i
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1 INTRODUCTION i
This third annual report, Supplement 3, updates the implementation and verification i
status of the Three Mile Island (TMI) Action Plan requirements, unrcsolved safety issues (USIs), generic safety issues (GSis), and other multiplant actions (MPAs). The NRC previously published three volumes of this NUREG series, bolume 1, published in March 1991, discussed the status of TMI Action Plan requioments. Volume 2, published in May 1991, identified the implementation and verification status of actions associated with USIs. Volume 3, published in June 1991, detailed the status of GSI actions. The first annual NUREG report, Supplement 1, combined these volumes into a single report and provided updated information as of September 30,1991.
Supplement 1 was published in December 1991. The second annual NUREG report, Supplement 2, provided updated information as of September 30,1992, in addition, Supplement 2 also provided the status of licensee implementation arid NRC verification of MPA issues not related to TMI Action Plan requirements, USis, or GSis. This third annual NUREG report, Supplement 3, provides updated information as of September 30,1993 for all TMI, USI, GSI, and MPA issues. Subsequent volumes will continue to be published annually to document the progress of implementation and verification of these items.
This report describes the implementation and verification status at the 109 licensed plants in the United States and makes this information available to interested parties, including the public. Supplement 2 of this NUREG series reported data on 110 plants, including San Onofre 1 and Trojan which are now permanently shut down but not including Comanche Peak 2, which is now fully operational. For the purpose of this report, San Onofre 1 and Trojan have been excluded from the data and Comanche Peak 2 has been included in the data on the implementation and verification status of TMI Action Plan requirements as well as USIs, GSIs and other MPAs.
included herein is information on the implementation and verification status of the TMI Action Plan requirements, USIs, GSis, and other MPAs. For the 109 licensed plants, there are 12,893 applicable items for TMl Action Plan issues,1,787 for USis,2,621 for GSis, and 7,517 for other MPAs. A total of 24,823 applicable items are addressed in this report. The information presented in this volume is current as of i
Septembu 30,1993.
t 1.1 Backoround j
TMI Action Plan requirements, USis, GSis, and other MPAs are all types of generic issues that originated from increased technical understanding of the safety of nuclear power plants. This increase in understanding occurred over time and resulted from operating events, research, testing, and experience. The specific origins of these issues and the development of requirements in each area, with the exception of other MPA's, were discussed in Volumes 1 through 3 of this NUREG series. The origin for other MPAs is discussed in section 1.3 of this supplement. Actions to be taken by licensees in response to these generic issues apply to more than one plant.
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l The NRC evaluates the status of each licensee's implementation in conjunction with its evaluation of other NRC requirements, unique plant considerations, and iriterim measures. Similarly, the NRC authorizes a licensee to restart or begin operation of its plant only after carefully reviewing the plant's compliance with NRC requirements and 1
evaluating the licensee's demonstrated capabilities to safely operate the plant. The NRC has allowed operation of a new plant, or continued operation of a licensed plant,-
when the licensee has not fully implemented all items discussed in this report only after ensuring that sufficient compensatory measures have been taken or after determining that plant operation presented no undue risk to the public health and i
- safety, i
i The data contained in this NUREG report are a product of the NRC's Safety issues l
Management System (SIMS), which is maintained by the Project Management Staff in
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the Office of Nuclear Reactor Regulation (NRR) and by NRC regional personnel. The NRC has given significant attention to the quality review of TMl, U31, GSI, and other MPA implementation and verification data in SIMS.
i 1.2 Process and Accountability in 1989, the Commission adopted a six-step program for closure of generic safety issues. Although TMI requirements were treated separately, the process to achieve and verify closure of these issues is similar to that used for USIs, GSis and MPAs.
The overall NRC program consists of the following steps:
i' Identifyino Relevant issues - Generic concerns are typically identified by the NRC i
staff as a result of perceived problems at one or more operating nuclear power plants, or as a result of revised analyses pertaining to matters previously considered resolved. Issues may also be identified by others, for example, j
licensees, vendors, the Advisory Committee on Reactor Safeguards (ACRS), and the public. The NRC staff identified the TMI requirements by compiling and j
evaluating information from all available sources concerning the accident at TMI.
Prioritizino issues - Once identified, an issue is evaluated by the staff for its potential importance to nuclear safety. The staff classifies the issue and establishes a priority for resolution based on this evaluation and on other factors, such as value-impact analysis and risk assessment. The primary purpose of prioritizing issues is to assist in the timely and efficient allocation of resources to those safety issues that have a high potential for reducing risk. Four priority categories are used: high, medium, low, and drop. A high priority ranking means that a concentrated effort is appropriate to achieve the earliest resolution practical.
Resolvino issues - The staff evaluates corrective actions that might be taken to 4
satisfactorily resolve a safety issue, in addition to using experience, tests, and experiments, the staff may use the results of analyses, probabilistic risk assessments, or other calculations in this evaluation. The staff uses the results of such work to propose the action or actions it would consider an acceptable basis
for closing the issue. The evaluation may require NRC to change requirements or guidance.
imoosino Reauirements (USLs and TMI Action Plan Reauirements)- Each affected
'icensee or applicant is requ red to prepare a schedule for implementing the resolution consistent with a rule, policy statement, regulatory guide, generic letter, bulletin, or licensing guidance developed during the resolution stage. The NRC staff evaluates the importance of the issue and determines whether it is to be imposed only on plants licensed after resolution of the issue, or if the required corrective actions should be backfit to existing plants.
Recuestina Action (GSis)-The NRC staff evaluates the importance of an issue and determines the types or classes of plants to which the resolution applies.
The staff also determines whether corrective action is appropriate for existing plants or only for plants licensed after resolution of the issue. These corrective actions may be imposed on licensees in the form of a rule, policy statement, regulatory guide, generic letter, bulletin, or licensing guidance. Each affected licensee or applicant is required to prepare a schedule for implementing the resolution. Once an issue o resolved, each action to be implemented is assigned a multiplant actian (MPA) number for tracking purposes.
Imolementina Actions - Uc6nsees of affected plants take corrective actions to satisfy commitments made in response to the imposed requirements (TMI Action Plan requirements and USIs) or the staffs request (GSis and other MPA issues).
These actions may include modifications or additions to equipment, structures, procedures, technical specifications, operating instructions, and so forth.
The role of the NRC Project Manager in implementing the resolution of a particular issue depends on the safety significance of the issue and the manner in which the issue is to be addressed. Significant TMI Action Plan requirements or USls may require backfits to existing plants. Backfitting is imposed by rule or order unless the licensee volunteers to comply, in which case a confirmatory order may be issued. In any case, a deadline is set or negotiated for completing action at the particular nuclear plant. The Project Manager monitors licensee progress toward closure and ensures that the work is completed by the negotiated date. The Project Manager ensures that the status of the item is properly documented for each plant.
Verifvina Imolementation - NRR staff members, NRC regional personnel and NRC resident inspectors ensure that licensees meet commitments made to the NRC for those issues requiring verification. Temporary instructions (Tis) have been issued to guide inspectors in verifying licensee implementation of corrective actions for certain generic issues that require plant hardware changes and subsequent verification by the NRC. Other issues may require engineering analysis to demonstrate continued safety of the plant, but require no specific plant configuration changes. For these issues, the NRC headquarters staff reviews and - -
ensures the acceptability of each analysis, and no verification at the plant site is required.
SIMS is designed to associated actions a, track issues from their identification through implementation of nd field verification. The NRR Project Manager periodically obtains data pertaining to the licensee's implementation dates from meetings, site visits, and discussions with resident inspectors or the licensee. Recent NRC initiatives to improve the quality and the accountability of data include requirements that (1) any conclasion that a corrective action has been implemented be accompanied by a reference document from the licensee staff providing the basis for closure of the issue at the particular plant, and (2) the inspection report number and the date of the inspection be entered into SIMS if verification is required.
1.3
_D_qfinitions A number of terms are used to describe generic issues and their status. These terms are important not only because they categorize issues and their origin, but because
- heir use implies both applicability and degree of completeness. For the purposes of this report, the following definitions apply:
Generic Safety issue (GSI)- A safety cone,ern that affects the design, construction, or operation of all, several, or a class of riuclear power plants and may have the potential for safety improvements at sucii plants.
Imolemented item - An item is implemented when a licensee has completed the activities necessary to satisfy the requirements (or assumptions) made in the staff's technical resolution in accordance with commitments concerning the generic issue.
Item - The application of a TMI Action Plan requirement, USI, GSI or other MPA issue to a specific plant.
MPA - A multiplant action item originates from industry experience, new regulations / requirements, or from resolution of genenc issues resulting in the issuance of a generic communication requiring action by the licensees. TMI items, USis and GSis are all MPAs; however, there are also other MPAs that do not fit into one of these categories. These other MPAs may be either required or voluntary.
.TMI Action Plan item - An issue applicable to one or more licensed plants as derived from NUREG-0737, Supplement 1, thereto.
Total Plant items - The theoretical maximum number of potential items resulting from applying each issue (TMI, USI, GSI or other MPA) to all 109 plants.
. Total ADolicable Plant items - The total number of applicable items determined by reviewing the applicability of each issue at each of the 109 licensed plants. l
Unimolemented item - An item is unimplere..ted when a p' ant has not completed the activities necessary to satisfy the actions requested or required by the staff following the resolution o.' a particular generic issue.
Unresolved Safety issue (USI)- A matter affecting a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants affected as identified in NUREG-0510 and subsequent annual reports to Congress.
Vecification Comoleted - A licensee's actions to implement a technical resolution for a generic issue have, been inspected and verified by the NRC in accordance with the guidance provided t.y the applicable temporary instruction for that issue.
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i 2
THREE MILE ISLAND ACTION PLAN REQUIREMENTS This section describes the overall status of implementation and verification of TMI Action Plan items at the 109 currently licensed plants. Supp% ment 2 of this NUREG series reported data on TMl Action Plan requirements for 11C plants, including San Onofre 1 and Trojan which are now permanently shut dowr but not hcluding i
Comanche Peak 2, which is now fully operational. For the purpose of this report, San i
Onofre 1 and Trojan have been excluded from the data and Comanche Peak 2 has been included in the data on the implementation and verification status of TMI Action Plan requirements.
2.1 Imolementation Status 1
More than 99 percent of the TMl Ac+ ion Plan items have been implemented or closed at licensed plants. Of the 12,898 items,12,837 have been completed and only 61 have not yet been implemented. Figure 2.1 presents the overall status of implementing the TMI Action Plsn requirements. The 12,837 items completed at the 109 licensed plants have been disposed of as follows:
A total of 12,441 have been implemented or closed by either incorporating fixes into the plant design before licensing or by implementing the necessary requirements at operating plants.
A total of 396 items have been superseded and the associated requirements have i
been effectively addressed by other items or through other regulatory means. The superseded items are discussed in detail in Volame 1 of NUREG-1435.
I The following obsen/ations are made about the remaining 61 unimplemented items:
i Approximately 38 percent of these items are projected to be implemented oy the end of calendar year 1994, as shown in Figure 2.2. Ucensees continue to make progress toward implementation of the remaining items.
4 As noted in previous status reports, some slippages have occurred in projected implementation dates. Delays in the restart of Browns Ferry Units i and 3 i
(34 items), along with the rescheduling of refueling outages at other plants account for a large number of the sl ppages in the implementation of remaining TMI items.
Browns Ferry Unit 1 has 15 TMI items that will not be implemented until 1997. All schedules for implementing the remaining TMl Action Plan items are within the timeframe previously reported to the Commission and to Congress.
From an issue perspective, four major areas account for about 77 percent of the 61 unimplemented items, as shown in Table 2.1: detailed control room design review items (21), accident monitoring (10 ), safety parameter display system items (8), and B&O Taqk Force issues (8).
TMI ACTION PLAN REQUIREMENTS 1
IMPLEMENTATION STATUS AT LICENSED PLANTS 300 OVERVIEW
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- These totolo do not include items for R. St. Vrain, Rende Seco, Sen Onofre 1, Shorehem, Trolen and Yanhoe Rowe plants. These pients are penneneney or :.1.C, ehut down. The total nurnber of licensed plante conaldered in this report is 100.
Figure 2.1 Y
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PROJECTED SCHEDULES FOR REMAINING TMI ITEMS ITEMS NOT IMPLEMENTED AT END OF CALENDAR YEAR 100 I
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09/93 12/93 12/94 12/95 12/96 12/97 09/93 1253 1254 1255 12S 6 1257 JUNE 1989 SCHEDULES
- 31 19 15 14 12 0
09/3043 SCHEDULES
- 61 46 38 16 15 0
- Based on dates of Roms listed as not implomonted Figure 2.2
SUMMARY
OF THE REMAINING TMi ITEMS BY AREA AREA OPEN o Plant SPDS Console 8
o Post-Accident Sampling 5
o Relief & Safety Valve Test Report 1
o Containment isolation Dependability 3
o Accident Monitoring 10 a
o o Instrumentation for Detection of inadequate Core. Cooling 4
. o B&O Task Force issues 8
o Control. Room ' Habitability 1
o Detailed Control Room Design Review 21
' TOTAL 61 Table 2.1 '
i
... ~.
From a plant perspective, the TMl Action Plan has been fully implemented or closed at 82 of the 109 licensed plants. Table 2.2 summarizes the 61 unimplemented items by plant. Two plants account for approximately 56 percent of the remaining 61 items: Browns Ferry 1 and 3 (34 items). Browns Ferry 2 and Ha:! dam Neck have 2 items and the remaining plante have 1 item each to irnpitimmt.
Appendi< A lists the pnimplemented TMI items by issue and gives projected implementation dates.
2.2 Verification Status For generic items, such as the TMI requirements, the Office of Nuclear Reactor Regulation issues temporary instructions (Tis), when appropriate, to specify which requirements are to be verified by the NRC after licensees have implemented the corrective actions specified in the resolution. The NRC performs these inspections, consistent with other inspection priorities, to verify proper implementation of the requirements. Verification is not considered complete until the NRC conducts the required inspection in accordance with the TI, and issues an inspection report documenting that the licensee'has adequately satisfied the requirements. On occasion, there may be issues for which the verification requirements according to the Tl are completed before the licensee has fully implemented all aspects of the issue.
Tis have been issued for 78 individual TMI requirements, which cover a total of 6487 items at the 109 licensed plants. Upon initial inspection of certain items and further review by the regional offices,495 items covered by the Tis were found to be inapplicable from a verification standpoint, leaving a net total of 5992 items requinng verification. The majority of items found not applicable are cases in which initial inspections did not reveal any significant findings and for which further inspection effort cannot be justified.
As of September 30,1993,5939 items (99 percent) had been verified. Only 53 items t
remain to be verified, including some items not yet implemented by licensees.
l l
SUMMARY
OF THE REMAINING TMI ITEMS BY PLANT PLANT OPEN PLANT OPEN PLANT OPEN Big Rock Point 1
Millstone 1 1
Prairie Island 1 1
Browns Ferry 1 15 Nine Mile Pt 1 1
Prairie Island 2 1
Browns Ferry 2 2
North Anna 1 1
Quad Cities 2 1
Browns Ferry 3 19 North Anna 2 1
San Onofre 2 1
Diablo Canyon 1 1
Palo Verde 1 1
San Onofre 3 1
Diablo Canyon 2 1
Palo Verde 3 1
Sequoyah 1 1
Dresden 3 1
Pilgrim 1 1
Sequoyal' 2 1
Ft Calhoun 1
Point Beach 1 1
Surry 1 1
Haddam Neck 2
Point Beach 2 1
Surry 2 1
F Table. 2.2.
i 4
j 2.3 Status by Plant e
Table 2.3 presents summary information on the status of TMI Action Plan items (except superseded items) at all licensed plants. For implementation, the table shows the number of applicable items, the number of items completed, the percentage i
completed, and the number of items remaining. For verification, the table shows the number of items covered by Tis at each plant, the number requiring verification, the number completed, and the percentage completed. Appendix A lists the
}
unimplemented items by issue and gives projected implementation dates.
From an implementation standpoint, the TMI Action Plan har been fully implemented at 82 of the 109 licensed plants. Browns Ferry Units 1 and 3 (34 items) account for approximately 56 percent of the 61 remaining open items.
1 From a verification standpoint, all required incoections have been completed at 78 of j
the 109 licensed plants. Browns Ferry Units 1 and 3 have 10 items each that will j
require verification following imp!smentation by the licensee. Twenty-nine plants have 1
2 or less items remaining to be verified.
i d
i 4
4 -
l SAFETY ISSUE MANAGEMENT SYSTEM
\\
STATUS OF TMt ACTION NAN -
SUMMARY
BY PLANT IMP 1EMENTATION VERIFICATION ITEMS I ttml PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETE.*
COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED ARKAWSAS 1 122 at?
(1001 0
62 61 61 (100)
ARMANSAS 2 112 112 (100) 0 59 58 58 (100)
BEAVER VALLEY 1 117 117 (2001 0
62 61 61 (100)
BEAVER VALLEY 2 126 126 11001 0
62 57 57 (100)
BIG ROCK POINT 1 104 103 199 1 1
56 50 50 (100)
BRAIDWOOD 1 126 126 11001 0
62 56 56 (100)
BRAIDWOOD 2 126 126 (1001 0
62 56 56 (100)
BROWNS FERRY 1 110 95 186 )
15 57 51 41 (80 )
BROWNS FERRY 2 110 108 198 1 2
57 51 50 (98 )
BROWNS FERRY 3 110 91 (82 )
19 57 51 41 (80 )
BRUNSWICK 1 110 110 11005 0
57 51 51 1100)
SRUNSWICK 2 110 110 (1001 0
57 51 51 (100)
OYRON 1 126 126 (1001 0
62 58 58 (100)
BYRON 2 126 126 (1001 0
62 58 58 11001 CALLAWAY 1 123 123 (2005 0
60 60 60 (100) i CALVERT CLIFFS 1 113 113 (1001 0
59 53 53 (100) l CALVERT CLIFFS 2 113
!!3 (1001 0
59 53 53 (2001 I
CATAWBA 1 126 126 (2005 0
62 60 59 198 i CATAWBA 2 126 126 (1001 0
62 62 61 (98 i A
CLINTON 1 120 120 (2005 0
56 56 56 (1001
.h COMANCHE PEAK 1 119 119 (1001 0
55 50 50 (100)
COMANCHE PEAK 2 120 120 t1001 0
59 53 53 (100) l COOK 1 11T 117 11001 0
62 55 55 (1001 C00M 2 117 117 (100) 0 62 55 55 (100)
COOPER STATION 110 110 (2005 0
57 51 51 (2001 CRYSTAL RIVER 3 122 122 (1001 0
62 56 54 196 )
DAV15-8 ESSE 1 121 121 (1001 0
61 55 55 (1001 DIABLO CANYON 1 126 125 199 1
1 61 55 55 (100)
DIABLO CANYON 2 126 125 199 l
1 61 57 57 (1001 DRESDEN 2 110 110 (1001 0
58 52 51 (98 l DRESDEN 3 110 109 199 l 1
58 52 51 (98 i DUANE ARNOLD 110 110 (1001 0-57 54 54 (2001 FARLEY 1 118 118 (1001 0
62 56 56 (100)
FARLEY 2 128 128 (1001 0
62 56 56 (100)
FERMI 2 120 120 (1001 0
56 50 50 (100)
FITZPATRICK 110 110 (2001 0
57 56 55 (98 i FORT CALHOUN 1 113 112 199 1 1
59 55 54 (98 )
GINNA 116 116 (1001 0
61 56 56 (100)
GRAND GULF 1 120 120 (1001 0
56 50-50 (100)
HADDAM MECK 117 115 (98 1 2
62 55 54 (98 i HARRIS 1 125 125 (1001 0
61 60 60 (100!
HATCH 1 110 110 (1001 0
57 57 57 (103)
HATCH 2 110 110 (2001 0
ST' 57 56 (92 I HOPE CREEK 1 120 120 11001 0
-56 50 50
!!30b Table 2.3
.--..m.
.____._m..
.-...m=._.4 m..
m... _.. _ ~
<__._.~.-.-.-u.
._m SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TM1 ACTION PLAN.
SUMMARY
BY PLANT IMPLEMENTATION VERIFICATI00t ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICA3LE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLE TED INDIAN POINT 2 118 118 (1001 0
62 59 59 (100)
INDIAN POINT 3 117 117 (1001 0
62 59-59 (200)
MEWAUMEE
~~
117 117 (2001 0
62 58 57 (98 I LASALLE 1 120 120 11001 0
56 52 52 (100)
LASALLE 2 120 120 (1001 0
56 52 52 (100)
LIME R ICK 1 120 120 (1001 0
56 52 52 (100)
LIMERICK 2 120 120 (1001 0
56 50 50 (2001 MAINE YANKEE.
113 113 (1001 0
59 56 56 (100)
MCGUIRE 1 127 127 (1001 0
62 62 61 (98 i MCGUIRE 2 127 127 (1001 0
62 62 61 (98 )
MILLSTONE-1 109 108 (99 1 1
56 46 45 (97 1 MILLSTONE 2 113 113 (1001 0
59 54 54 (200)
MILLSTONE 3 127 127 (1001 0
62 58 58 (100)
MONTICELLO 110 110 (1006 0
57 51 51 (2001 NINE MILE POINT 1 107 106 (99 3 1
56 54 54 (1001 NINE MILE POINT 2 119 119 (1005 0
55 52 52.
(2001 NORTH ANNA 1 118 117 (99 1 1
62 60 60 (100)
NORTH ANNA 2 128 127 (99 1 1
62 60 60 (100)
OCONEE 1-122 122 1100t 0
62 57 56 (98 )
j, OCONEE 2 122 122 (1001 0
62 58 ST-(98 l g
OCONEE 3 122 122 (1001 0
62 58 57 (98 )
Ov5TER CREEK 1 107 107 (1008-0, 56 47 47 (1001 PALISADES 113 113 (100) 0 59 51 51 (100)
PALO VERDE 1 120 119 (99 l 1.
59 52 52 (100)
PALO VERDE 2 120 120 (100)
O 59 53 53 (100)
PALO VERDE-3 120 119 (99 1 1
59 54 54 (100)
PEACH 80TTOM 2 110 110 (2005 0
57 51 51 (100)
PEACH BOTTOM 3 110 110 11001 0
57 51 51 (100)
PERRY I 120
'120 (1001 0-56 55 55 (100)
PILGRIM 1 110.
109 (99 l 1
57 48 48 (100)
POINT BEACH 1 117 116-199 l 1
62 56 56 (100)
. POINT BEACH 2 117 116 (99 1 1
62-
- 56-56.
(100)
PRAIRIE ISLAND 1
-117 116 (99 l 1
62 52 52 (100)
PRAIRIE ISLAND 2
.117 116 (99~l 1
62 QUAD CITIES 1 110 110 (1001 0
57
. 53 53 (1001
'51 50-(98 l QUAD CITIES 2 110 109 (99 l 1
57 51 50 (98 I RIVER BEND 1 119 119 (1001 0
56 50 50 (1001 ROBINSON 2 117
!!F (100) 0 62 56 54 (96 l' SALEM i 116 116-(2003 0
61 54-52 (96 SALEM 2 127 127 (1001 0
62-
- 57 55
!96 i SAN ONOFRE 2 122 121 (99 l 1
59 53 53 (1001 SAN ONOFRE 3 122 121 (99 l 1
59 55 55 (100)
SEABROOK 1 127 127 (lool 0
62 57-57-(1001 SEQUOVAH 1 127 126 (99 I I
62 56 55 (98 I Table 2.3 3
. =. -
. - a
t l
l l
SAFETY ISSUE MANAGPsENT SYSTEM STATUS OF TMt ACTKP' & -
SUMMARY
BY PLANT l
IMPLEF mTATION VERIFICATION l
ITEMS ITF*S PER CENT ITEMS ITEMS ITEMS ITEMS PER CEIIT i) NIT APPLICABLE COW LETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED
.SEQUOYAH 2 127 126 (99 1 1
62 58 58 (100)
SOUTH TEXAS 1 126 126 (1003 0
62 56 56 (100)
SOUTH TEXAS 2 12F 126 (100) 0 62 56 56 (100)
ST LUCIE 1 143 113 1100)
O 59 55 55 (100)
ST LUCIE 2 122 122 11001 0
59 54 54 (100)
SUD94ER 1 127 127 (1001 0
62 61 61 (100]
SURRY 1 117 116 (99 1 1
62 56 56 (100)
SURRY 2 117 116 199 l 1
62 56 56 (1001 SUSQUEHANNA i 120 120 (1001 0
56 56 56 (1003 SUSQUEHANNA 2 120 120 11005 0
56 56 56 (100)
THREE MILE ISLAND 1 128 128 11001 0
62
.55 55 (100)
TURKEY. POINT 3 117 117 (1001 0
62 60 59 (98 )
TURKEY PF;NT 4 117 117 (1001 0
62 60-59 (98 )
VERMON' YANKEE 1 110 110 (2005 0
57 53 53 (1001 V0GTL 1 124 124 fl005 0
60 54 53 198 )
VPGILE 2 124 124 (1001 0
60 56 55 (98 l WASHINGTON NUCLEAR 2 120 120-11001 0
56 55 55 (1001 WATERFORD 3 121 121 (2001-0 59 55 55 (100*
WOLF CREEK 1 126 126 l1001 0
60 54 54 (100l
_A ZION 1 117 117 (2003 0
'62 80 59 198 )
g ZION'2 117 117 (100) 0 62 60-59 (98 ).
TOTALS'/ AVERAGES 12898
- 12837*
100 61 6487
-.5992
$939 99
- Excludes 396 superseded items at the 109' licensed' plants-Table 2.3 a
d 2.4 Status by Issue Table 2.4 summarizes information on each TMI issue. For implementation, the table shows the number of applicable plants, the number of plants completed, the percentage completed, and the number of plants remaining. For verification, the table shows whether the issue requires verification, the number of plants covered by the TI, the number of plants requiring verification, the number of plants completed, and the percentage completed.
Of the 172 TMI Action Plan issues,147 have been fully implemented and 4 have been superseded. Four categories of TMI Action Plan issues account for about 77 percent of the TMI requirements to be implemented: detailed control room design review, with 21 plants yet to complete implementation; accident monitoring, with 10 plants remaining to complete implementation; safety parameter display system, with 8 plants remaining to complete implementation; and B&O Task Force issues, with 8 plants still open. The next largest contributor is post accident sampling issues, still open at 5 3
plants.
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j SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN - SUPNARY BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAIP!NG REQUIRED COVERED REQUIRED-COMPLETED COMPLETED I.A.2.3 109 109 (100) 0 NO ADMINISTRATION OF TRAINING PROGRAMS I.A 3.1.1 109 109 (100)
O NO-REVISE SCOPE & CRITERIA FOR LICENSING EXAMS - INCREASE SCOPE
'I.A.3.1.2 109 109 (100) 0 NO REVISE SCOPE & CRITERIA FOR LICENSING EXAMS - INCREASE PASSING GRADE I.A.3.1.3.A 80 80 (1001 0
NO REVISE SCOPE & CRIT. FOR LIC. EXAMS - SIMULATOR PLANTS WITH SIMULATORS e
I.A.3 1.3.8 30 30 (100]
O NO REVISE SCOPE & CRIT. FOR-LIC. EXAMS - SIMULATOR - OTHER PLANTS I.B 1.2 SO SO (1001 0:
NO EVALUATION OF ORGANIZATION & MANAGEMENT I.C.1.1 109 109
'(1001 0
YES 109 109 109 (100)
SHORT-TERM ACCIDENT & PROCEDURES REVIEW - SB'LOCA I.C.1.2.A 109 109 (100) 0 YES 109 92 92 (100)
SHORT-TERM ACCID. & PROCEDURES REV. - INADEO. CORE COOL.'REANAL. GUIDE I C.1.2.8 109'.
109 (1001 0'
YES 109 109' 109 (100)
SHORT-TERM ACCID. & PROCEDURES REV.'- INADEQ. CORE COOL. REVISE PkOCED I.C.1.3.A 109.
109-(100) 0-YES 109 92-92 (100)
SHORT-TERM ACCID. & PROCEDURES REV - TRANSIENTS & ACCDTS. REANAL GUIDE I.C.1.3.8
'109 109 (100) 0 YES 109 109 109 (100)
SHORT-TERM ACCID. 4 PROCEDURES REY.. TRANSIENTS & ACCDRS. REVISE PROC' I.C.2 109 109 (100)
-0
-YES-109 109 109 (100l-SHIFT & RELIEF TURNOVER PROCEDURES I.C.3 109 109
-(1001 0
YES -
109 109 109 (100)
SHIFT-SUPE
- VISOR RESPONSIBILITY I C.4 109 109 1100)
-O' YES 109 109
.109 (100)
CONTROL-ROW ACCESS I.C.S'-
109 109
-(1001 0
.YES 109 109 109 (100)
FEEDBACK OF OPERATING EXPERIENCE.
Table 2.4 -
i
..____....._..m
?
SAFETY ISSUE MANAGEMENT SYSTEM
{
STATUS OF TMI ACTION PLAN - SUPNARY BY ITEM-IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED.
I C.6 109 109 (100l 0
YES 109 109 109 (100)
VERIFY CORRECT PERFORMANCE OF OPERATING ACTIVITIES I.C.7.1 49 49 (100) 0 NO NSSS VENDOR REV. OF PROC - LOW POWER TEST PROGRAM I.C.7.2 49 49 (100) 0 NO NSSS VENDOR REV. OF PROC - POWER ASCENSION & EMER. PROCS I.C 8 47 47 (100)
O NO -
PILOT MON OF SELECTED EMERGENCY PROC FOR NTOLS
.h I.D.2.1 109 109 (100) 0 NO PLANT-SAFETY PARAMETER DISPLAY CONSOLE - DESCRIPTION i
i I.D.2.2 109 106 (97 1 3
YES 109 49 49 (100) l PLANT. SAFETY PARAMETER DISPLAY CONSOLE - INSTALLED
.I.D.2.3-109 -
104 195 )
5 YES 109 39 39 (100)
PLANT-SAFETY _ PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED I. G 1.' 1
' 49 49 (100) 0
-NO TRAINING DURING LOW-POWER TESTING - PROPOSE *ESTS-I Gfl.2 49 49 (100)
C:
NO TRAINING DURING LOW-POWER TESTING - SUBMIT ANAL. & PROCS, I.G.1.3
- 49
'49 11001' 0
NO -
TRAINING DURIneG LOW. POWER TESTING - TRAINING & RESULTS-II.B.I.1 109 109 (1001-
- 0-NO REACTOR-COOLANT SYSTEM VENTS - DESIGN VENTS 11.8.1.2 109' 109 (1001 0 ~
YES-109i 108-
'108 (100)
REACTOR-COOLANT SYSTEM VENTS
- INSTALL VENTS (LL CAT 8)
II'8.1.3 109 109-(100) 0 YES 109 108-108 (100)-
REACTOR-COOLANT SYSTEM VENTS'- PROCEDURES II.B.2.1 109 109_.
(100)
.0 NO PLANT. SHIELDING - REVIEW DESIGNS 11.8 2.2' 109-109 (1001'-
0 NO PLANT. SHIELDING - CORRECTIVE ACTIONS TO ASSURE ACCESS
. Table 2.4-
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN. StMHARY BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED-COMPLETED COMPLETED II.B 2 3 109 109 l1001 0
YES 109 109 109 (100)
PLAN
- SHIELDING - PLANT MODIFICATIONS (LL CAT B)
II.B.3.1 108 108 (100) 6 YES 108 102 102 (100)
POSTACCIDENT SAMPLING - INTERIM SYSTEM
'II.B 3.2 109 108 (99 )
1 NO POSTACCIDENT SAMPLING - CORRECTIVE ACTIONS II.B.3.3 108 106 (98 )
2 NO POSTACCIDENT SAMPLING - PROCEDURES eh II.B.3.4 109 107 (98 )
2 YES 109 109 106 (97 )
POSTACCIDENT SAMPLING - PLANT MODIFICATIONS (LL CAT B) e 11.B.4.1 109-109 (100) 0 NO TRAINING FOR MITIGATING CORE DAMAGE - DEVELOP TRAINING PROGRAM II.B.4.2.A
- 109 109 (100) 0 YES 109-109 109 (100)
TRAINING FOR MITIGATING CORE DAMAGE - INITIAL II.B.A.2.B
~109 109 (1003' 0-YES 1091 109 109 (100)
TRAINING FOR MITIGATING CORE DAMAGE'. COMPLETE II.D.1 1 109 109 (1001 0
NO -
RELIEF & SAFETY VALVE TEST REQUIREMENTS - SUBMIT PROGRAM II.D.1.2.A
'109 109 (100)
'O.
No
' RELIEF & SAFETY VALVE TEST REQUIREMENTS - COMPLETE TESTING II.D.1.2.8 109-109 (100)
O' NO RELIEF & SAFETY VALVE TEST REQUIREMENTS - PLANT SPECIFIC. REPORT II.D.I.3-69 - -
' 68 -
-(98 1 1
N0' RELIEF & SAFETY VALVE. TEST REQUIREMENTS - BLOCK-VALVE TESTING II.D 3.1
-109 109 l100) 0-YES 109' 109 109 (100)
VALVE POSITION INDICATION - INSTALL DIRECT INDICATIONS OF VALVE'POS II.D.3.2s 109 109' (100) 0 NO VALVE POSITION INDICATION - TECH SPECS II.E.1.1.1 72 72 (100)
O.
NO AFS EVALUATION-ANALYSIS
-Table 2.4
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN -
SUMMARY
BY ITEM
~
IMPLEMENTATI0N VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED II.E.1.1.2 71 71 (100) 0 YES 71 71 71 (1003 AFS EVALUATION. SHORT TERM HODS.
II.E.1.1.3 72 72 (100) 0 YES 72 72 72 (100)
AFS -LONG TERM NODS.
II.E.1.2.1.A 66 66 (100) 0 YES 66 66 66 (100)
AFS INITIATION & FLOW-CONTROL GRADE II.E.1.2.1.8 72 72 (100) 0 YES 72 72 72 (1001 AFS INITIATION & FLDW - SAFETY GRADE h$
II.E.1.2.2.A 67 67 (100) 0 YES 67 67 67 (100)
AFS INITIATION & FLOW - FLOW INDUCTION CONTROL GRADE e
II.E 1.2 2 8 72 72 (1001 0
NO AFS INITITATION & FLOW - LL CAT A TECH SPECS.
II.E.1.2.2.C 72 72 (1001 0
YES 72 72 72 (100)
AFS INITITATION & FLOW - SAFETY GRADE II.E.3.1.1 -
72 72-(100) 0 vES 72 72 72 (1001 l
EMERGENCY POWER FOR PRESSURIZER HEATERS - UPGRADE POWER SUPPLY II.E.3.1.2 72 72 (100) 0 NO EMERGENCY POWER FOR PRESSURIZER HEATERS - TECH SPECS II.E.4.1.1 107 107 (100) 0 NO DEDICATED HYDROGEN PENETRATIONS - DESIGN II.E.4.1.2 107 107 (1001 0
YES 107 102 100 (98 1 DEDICATED HYDROGEN PENETRATIONS - REVIEW & REVISE H2 CONTROL PROC II.E.4.1 3 107 107 (1001 0
YES 107 104 102 (98 )
DEDICATED HYDROGEN PENETRATION - INSTALL II.E.4.2.1-4 109 107 (98 )
2 YES 109 109 108 (99 )
CON T AIME N T ISOLATION DEPENDABILITY - IMP. DIVERSE ISOLATION II E.4.2.5.A 109 109 (1001 0
NO CONTAINMENT ISOLAT. DEPENDABILITY - CNTMT PRESS. SETPT SPECIFY PRESS.
II.E.4.2.5.8 109 109 (100) 0 YES 109 108 107 (99 )
CONTAINMENT ISOLATION DEPENDABILITY - CNTMT PRESSURE SETPT. MODS.
Table 2.4 I
L
m m m SAFETY TSSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN -
SUMMARY
BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED I I '. E. 4. 2. 6 109 108 199 )
1 YES 109 108 108 (100)
CONTAINMENT ISOLATION DEPENDABILITY - CNTMT PURGE VALVES II.E.4.2.7 109 109 (100) 6 YES 109 106 105 (99 )
CONTAINMENT ISOLATION DEPENDABILITY - RADIATION SIGNAL ON PURGE VALVES II.E.4.2.8 109 109 (100)
D NO CONTAINMENT ISOLATION DEPENDABILITY - TECH SPECS II.F.1.1 109 107 (98 )
2 NO ACCIDENT-MONITORING.- PROCEDURES
,h II F.1.2.A 109 107 (98 )
2 YES 109 109 109 (100)
ACCIDENT-MONITCRING - NOBLE GAS MONITOR
=
II.F.1.2.8 109 107 (98 )
2 YES 109 109 109 (100)
ACCIDENT-MONITORING - IODINE / PARTICULATE SAMPLING II.F.1.2.C 109 107 (98 1 2
YES 109 109 107 (98 )
ACCIDENT-MONITORING - CONTAINMENT HIGH-RANGE MONITOR II.F.1.2.0 109 -
108 (99 )
1 YES 109 109 107 (98 )
ACCIDENT-MONITORING - CONTAINMENT PRESSURE II.F.1.2.E 109 108 (99 )
1 YES 109 109 109 (100)
ACCIDENT-MONITORING - CONTAINMENT WATER LEVEL II.F.1.2.F 109 109 (100) 0 YES 109 108 107 (99 )
ACCIDENT-MONITORING - CONTAINMENT HYDROGEN II.F.2.2 72 72 (100)
O YES 72 72 69 (95 )
INSTRUMENTATION FCR DETECT. OF INADEQUATE CORE COOLING - SUBC00L METER II.F.2.3 109 109 (100) 0 NO INSTRUMENTATION FOR DETECT. OF INADEQUATE CORE COOLING - DESC. OTHER II.F.2.4 108 104 (96 )
4 YES 108 105 94 (89 )
INSTRMNTATM FOR DETECT. OF INADEQ CORE CLNG INSTLL ADD *L INSTRMNTATN II.G.1 1 72 72 (100) 0 YES 72 72 72 (100)
POWER SUPP. FOR PRESSURIZER RELIEF, BLOCK VALVES & LEVEL IND.- UPGRADE 72 72 (100) 0 NO II.G.I.2
. PRESSURIZER RELIEF. BLOCK VALVES & LEVLE IND.- TECH SP.'
P0b'ER SUPP. FOR Table 2.4-i
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SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN - SIM4ARY BY ITEM IMPLEMENTATION VERIFICATION
--....................e........-..................
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLE TED REMAINING REQUIRED COVERED' REQUIRED COMPLETED CCMPLETED II.K.1.21 1
1 (100) 0 NO IE BULLETINS - AUTO SG ANTICIPATORY REACTOR TRIP II.K.1.22 14 14 (100)
O NO IE BULLETINS - AUX. HEAT REM SYSTM, PROC.
II.K.1.23 14 14 (100)'
.D NO IE BULLETINS - RV LEVEL, PROCEDURES II.K.I.5 50 50 (100]
O NO IE BULLETINS - REVIEW ESF VALVES e
h.
II.K.2.10 7
7
( *.001 0
YES 7
7 7
(100)
ORDERS ON B&W PLANTS - SAFETY-GRADE TRIP II.K.2.11 7
7 (1001 0
.YES 7-7 7
(100)
ORDERS ON B&W PLANTS - OPERATOR TRAINING II.K 2.13 70 70 (1001 0
NO
-ORDERS ON B&W PLANTS - THERMAL MECHANICAL REPORT (CE & W PLANTS ALS0)
II.K.2.14' 7
7
- (100) 0 NO -
ORDEP'a ON B&W PLANTS - LIFT FREQUENCY OF PORV's & SV*S
.II.K.2.15 7
7 (1001 0
NO ORDERS ON B&W PLANTS - EFFECTS OF SLUG FLOW II.K.2.16 7
7 (100) 0 NO ORDERS ON B&W PLANTS - RCP SEAL DAMAGE II.K 2.17 72 72' (100) 0 NO ORDERS ON B&W PLANTS - VOIDING IN RCS (CE & W PLANTS ALS0)
-II.K.2.19 7 -
7 (100)
-0 NO BENCHMARK ANALYSIS OF SEQUENTIAL AFW FLOW TO'ONCETHROUGH STM GENERATOR-0 NO-(100)ICS II.K.2.2
-7 7
ORDERS ON B&W PLANTS'- PROCEDURES TO CONTROL AFW IND OF II.K.2.20 7
. 7 (1001 0
NO ORDERS ON B&W PLANTS'- SYSTEM RESPONSE TO SB LOCA II.K.2.8 7
7 (100) 0.'
YES 7
6 6-(100)
ORDERS ON B&W PLANTS - UPGRADE AFW SYSTEM Table 2.4 i-
=
m.
. = _ --.. _. - _ __
m.._.
i i
l i
SAFET Y ISSUE MANAGEMENT SY STEM STATUS OF TMI ACTION PLAN -
SUMMARY
BY ITEM I
IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED II.K.2.9 7
7 (1001 0
YES 7
7 7
(100)
ORDERS ON B&W PLANTS - FEMA DN ICS II.K.3.1.A 72 72 (1001 O
NO B&O TASK FORCE - AUTOMATIC PORV ISOLATION DESIGN II.K.3.1.B 72 72 (100) 0 YES 72 62 62 (100]
FINAL RECOMMENDATIONS.B&O TASK FORCE - AUTO PORV 150 TEST / INSTALL II.K.3.10
'45 45 (1001 0-YES 45 41 41 (100)
B&O TASK FORCE - PROPOSED ANTICIPATORY TRIP MODIFICATIONS e
h.@
II.K.3.11-9 9
(100l'
-0 NO cr B&O TASK FORCE - JUSTIFY USE OF CERTAIN PORV-II.K.3.12.A 50' 50 (1001 0
NO B&O TASK FORCE - ANTICIPATORY TRIP ON TURBINE TRIP PROPOSED MODS
+
.II.K.3.12.5; 50 50 (1001 0
YES 50J 48 48 (100]
B&O TASK FORCE -. ANTICIPATORY TRIP DN TURBINE TRIP INSTALL MODS.
_II.K.3.13.A 33 33 (100) -
0 NO B&O TASK FORCE - HPCI & RCIC SYSTEM INITIATION LEVELS ANALYSIS II.K.3.13.8 33 31-(93 1 2-YES' 33 133 33 (100)
B&O TASK FORCE - HPCI & RCIC INITIATION LEVELS MODIFICATION II.K;3.14 5
6
. (100) 0 YES
~6 5~
5 (100)
B&O TASK. FORCE - ISO CONDENSER ISOLATION ON HIGH RAD II.K.3.15 33 33 (1001 0
YEst 33 33 33-(100]
B&O TASK FORCE - MODIFY MDCI & RCIC BRE DETECTION CIRCUITRY II.K.3.16.A 37 37 (100) 0 NO B&O TASK FORCE.- CHALLENGE & FAILURE.0F RELIEF VALVES STUDY.
II.K.3.16.8'
.. 37 37-(1001' 0
YES 37 37 37 (100)
B&O TASK FORCE -. CHALLENGE & FAILURE OF RELIEF.-VALVES MODIFICATIONS II.K.3.17' 95 -
-95 (100)
'O
- NO.
B&O TASK FORCE - ECC SYSTEM OUTAGES
.II.K.3.18.A 36 36 (100)-
O' NO:
B&O TASK FORCE - ADS ACTUATION STUDY Table 2.4 i
y e
l l
SAFETY ISSUE MANAGEMENY SYSTEM STATUS OF TMI ACTION PLAN - SUP94ARY BY ITEM IMPLEMENTATION VERIFICATION l,
_________~____...___.. ____..__________.__________
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED II.K.3.18 8 36 36 (100]
O NO B&O T ASK FORCE - ADS ACTUATION PROPOSED MODIFICATIONS II.K.3.18.C 36 34 (94 1 2
YES 36 35 33 (94 i B&O TASK FORCE - ADS ACTUATION MODIFICATIONS II.K.3.19 3
3 (1001 0
YES 3
3 3
(100)
B&O TASK FORCE - INTERLOCK RECIRCULATORY PUMP MODIFICATIONS II.K.3.2 70 70 (1001 0
NO B&O TASK FORCE - REPORT ON PORV FAILURES
-q II.K.3.20 I
1 (1001 0
YES-1 1
1 (1001 h3 B&O TASK FORCE - LOSS OF SVC WATER AT BRP II.K.3.21.A 37 37 (1001 0
NO B&O TASK FORCE - RESTART OF CSS & LPCI LOGIC DESIGN II.K.3.21.8 37 37 (1001 0
YES 37 35 35 (1001 B&O TASK FORCE - RESTART OF CSS & LPCI LOGIC DESIGN MODIFICATIONS II.K.3.22.A-32-32 (1001 0
NO B&O TASK FORCE - RCIC SUCTION VERIFICATION PROCEDURES II.K.3.22.8 32 32 (1001 0
NO B&O TASK FORCE - RCIC SUCTION MODIFICATIONS 6 TA2 FORCE - SPACE CO ING FOR HPCI CI LOSS OF OWER II.K.3.25.A 102 102 11001 0
NO B&3 TASK TORCE - POWER ON PUMP SEALS PROPOSED M001FICATIONS II.K.3.05.B 101 101 (1001 0
YES 101 97 97 (1001-B&O. TASK FORCE - POWER ON PUMP SEALS MODIFICATIONS II.K.3.27 37' 35 (94 1 2
YES 3T 37-37 (1001 B&O TASK FORCE - CopO90N REFERENCE LEVEL FOR BWRS II.K.3.28 37' 35 (94 )
2
.YES 37 36 34 (94 1 B&O TASK FORCE - QUALIFICATION OF ADS ACCUMULATORS
'II.K 3.29 6
6 (1001 0.
NO B&O TASK FORCE PERFORMANCE OF ISOLATION COBOENSERS Table 2.4
i I
l -
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN -
SUMMARY
BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICA8LE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED II.K.3.3 109 109 (200) 0 NO B&O TASK FORCE - REPORTING SV & RV FAILURES AND CHALLENGES II.K.3.30.A 109 109 (100) 0 NO B&O TASK FORCE - SCHEDULE FOR OUTLINE OF SB LOCA MODEL 1
l II.K.3.30.5 109-109 (100) 0 NO
(
B&O TASK FORCE - SB LOCA MODEL. JUSTIFICATION II.K.3.30.C 109 109 (1001 0
NO 840 TASK FORCE - SB LOCA METHODS NEW ANALYSES h3 00 II.K.3.31 109 109 (1001 0
NO B&O TASK FORCE - COMPLIANCE WITH CFR 50.46 II.K.3.44 37 37 (1001 0
NO B&O TASK FORCE - EVALUATE TRANSIENT WITH SINGLE FAILURE II.K.3.45 36 36 (100) 6 NO B&O TASK FORCE - ANALYSES TO SUPPORT -
II.K.3.46 37 37 (100) 0 NO RESPONSE TO LIST OF CONCERNS FROM ACRS CONSULTANT
.II.K;3.5.A 72-72 (100)
O NO B&O TASK FORCE - AUTO TRIS OF RCP*$ PROPOSED MODIFICATIONS II.K.3.5.B 72 72 (100) 0 YES 72 66 66.
(100)
B&O TASK FORCE - AUTO' TRIP OF RCP*S MODIFICATIONS _
II.K.3.57 23 23 (100) 0-YES 23 23-20 tOS )
IDENTIFY WATER SOURCES PRIOR'TO MANUAL ACTIVATION OF ADS-II K.3.7 7
7 (100) 6-NO B&O TASK FORCE - EVALUATION OF PORV OPENING PROBABILITIES II.K.3.9 50 50' (100)'
0 YES
' 50 50 50 (100)
B&O TASK FORCE - PID CONTROLLER MODIFICATION MPA-F008 109 109
'(100) 0 NO.
I.D.1.1 DETAILED CONTROL ROOM DESION REVIEW PROGRAM PLAN
- MPA-F063 109 109 (1u3) 0_
YES 109 34 34 (100)
.III.A.I.2 TECHNICAL SUPPORT CENTER-l Table'2.4 l-l t
i F
F SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN - SUP94ARY BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED MPA-F064 109 109 (1001 0
YES 109 48 48 (100).
III.A.1.2 OPERATIONAL SUPPORT CENTER MPA-FOss 109-109 (100) 6 YES 109 33 33 (100)
III.A.1.2 EMERGENCY OPERATIONS FACILITY MPA-F071 109 88 (80 1 21 NO I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8)
.E.
-Table ~2.4
=..
... =.
=
2.5 Conclusions After a detailed review of the implementation and verification status of the TMI Action Plan requirements at all licensed plants, the NRC staff has concluded the following:
Progress has been made in the implementation of TMI Action Plan requirements at alllicensed plants.
Ucensees continue to make progress toward implementing the remaining requirements. The schedules currently proposed by licensees for completing the remaining items are acceptable and are within the timeframes given to the Commission and to Congress, with the exception of Browns Ferry 1.
The NRC closure process for TMI Action Plan items ensures continued adequate protection of the public health and safety.
The NRC staff will maintain close watch over the implementation actions and schedules proposed by licensees to ensure that the TMI requirements that remain to be implemented are completed in accordance with regulatory requirements.
3 UNRESOLVED SAFETY ISSUES This section presents the overall status of implementation and verification of the l
requirements imposed following the resolution of USis.
3.1 Imolementation Status Ucensees achieve implementation of USI items either by incorporating corrections into.
the plant design before licensing or by making the modifications necessary to meet requirements at licensed plants. The information presented here includes all OSI items related to the 109 licensed plants considered in this report.
Approximately 90 percent of the USl items have been implemented at licensed plants.
Of the 1,787 applicable items,1,610 have been completed and only 177 remain open from an implementation standpoint. On average, each plant has approximately 2 '
remaining items to implement. No plant has more than 6 remaining items. Figure 3.1 presents the overall status of, and progress on, USIs. Of the 109 licensed plants,18' have fully implemented all applicable USis. Table 3.1 lists the number of unimplemented USI items by plant. Appendix B lists the unimplemented USl items by issue and projected implementation dates.
USIs A-44, A-48,' and A-47 account for 90 percent of the 177 unimplemented items.
Figure 3.2 suinmarizes the implementation status of these issues. These three USis are in varying stages cf NRC review and licensee implementation, as described below.
A-44 Station Blackout (A022)
The station blackout rule was issued in July 1988. According to the rule, licensees are required to implement their proposed modifications (hardware and procedural)
{
within 2 years of NRC notification approving the licensee's approach. - The staff has completed all of the safety evaluation reviews of licensee responses. About half of the plants have proposed major hardware modifications, while the remaining plants-are expected to implement minor hardware and procedure modifications. About 40 -
percent of the plants have already implemented procedure modifications, and the staff expects that a large majority of licensees will complete implementation of the station blackout rule by the end of 1994.
Ti 2515/120 was issued on September 24,1993, and will be performed at 8 sites, involving all 5 regions, at which time an evaluation will be performed by NRR to determine if additional sites need to be inspected.
A-46 Seismic Qualification of Eouloment in Ooeratino Plants (B105)
The Generic Implementation Procedure, Revision 2 (GIP-2), was developed by the Seismic Qualification Utility Group (SOUG) for implementation of USl A-46. On i
i Unresolved Safety issues l
Implementation Status at Licensed Plants l
2000 --
l 1500 --
l
}
1000 --
161 500 --
0 1
1 Applicable implemented Unimplemented Figure 3.1 i,
Summary of Unimpl@mented USI Items by Plant items items items PLANT Remaining PLANT Remaining PLANT Remaining Arkansas 1 2
Hatch 1 2
Point Beech 1 2
Arkansas 2 2
Hatch 2 2
Point Beach 2 2
Beaver. Valley 1 1
Hope Creek 1 1
' Prairie stand 1 1
l Big Rock Pt 1 2
Indian Pt 2 2
Prairie taland 2 1
Browns Ferry 1 6
Indian Pt 3 2
Quad Cities 1 4
Browns Ferry 2 2
Kewaunee 2
Quad Cities 2 4
Browns Ferry 3 6
LaSalle 1 1
River Bond 1 2
. Brunswick 1 2
LaSalle 2 2
Robinson 2 2
Brunswick 2 2
McGuire 1 1
Salem 1 2
Calvert Cliffs 1 3
McGuiro 2 1
Salem 2 2
Calvert Cliffs 2 3
Millstone 1 3
San Onofra 2 2
Catawba 1 1
Millstone 2 3
San Onofre 3 2
Catawba 2 1
Millstone 3 1
Sequoyah 1 1
Clinton ' 1 '
1 Monticello 2
Sequoyah 2 1
6 Comanche Peak 1 1
Nine Mile ' Pt 1 2
St. Lucio 1 2
?
Cook 1 3
Nine Mile-Pt 2 1
St. Lucie 2 1
Cook 2 3
North Anna 1 2
Surry 1 2
Cooper Station 2
North Anna 2 2
- Surry 2 2
Crystal River 3 3
Oconee-1 2
Susquehanna.1 1
Davis-Besse 1 1
Oconee 2-2 Susquehanna 2 1
Dresden 2.
2 Oconee 3 2
Three Mlle Island 1 1
Dresden 3.
3 Oyster. Creek 1 2 ~
Turkey Pt 3 -
2 Duane Arnold
-1 Palisades 3
Turkey Pt 4 2
Farley - 1 2
- Palo Verde.1 1
Vermont Yankee 1 1-1 Farley 2 '
_1 Palo Verde.2 ~
1
_Vogtle 1 -
1 Fitzpatrick -
3 Palo Verde 3 1
. Vogtle 2 1
Ft Calhoun 1 -
3 Peach Bottom 2 3
Washington Nuclear 2 1
Ginna.
2
. Peach Bottom 3 3
Waterford 3
.2 Grand Gulf 1 1
Perry '1 2
Zion.1 2
Haddam Neck 3
Pilgrim.1 2-Zion 2 -
2 Harris 1 1
Table 3.1.
a s
A
Summary of Three Unimplemented USIs 150 --
i (109)
(109) l cwq l
100 -
W2M3 I
RG 1
bN @)
(es)-
- )
50 I
I ""*""'
'I-0 I-1 I
A-44
_ A-46 Seismic _
A-47 Safety Station Qualification of implications of Blackout Legend Equipment Control in Operating Plants Systems I
E Applicable Plants -
9 Implemented Plants E Unimplemented Plants Figure 3.2
May 22,1992, the NRC staff issued its Supplemental Safety Evaluation Report (SSER 2) identifying the conditions under which the GIP-2 resolution is acceptable.
Each licensee was required to submit its schedule for implementing the resolution by September 19,1992. Most licensees have committed to use the GlP-2 as supplemented and clarified by SSER 2 and will provide their seismic evaluations for staff review in 1995. Florida Power Corporation (for Crystal River) and Florida Power end Ught (for St. Lucie 1 and Turkey Point) are implementing plant-specific resolutions. In addition, by letter dated August 28,1991, from Dr. T.E. Murley, the licensee for Maine Yankee was informed it rieed not respond to A-46.
A-47 Safety Imolications of Control Systems (8113)
The primary focus of the resolution of this USl is to provide a mechanism to trip the main feedwater pumps when a high water level occurs in the reactor vessel or i
steam generators. In 1990, the staff reviewed the licensees' responses to Generic Letter 89-19 and determined that:
The Westinghouse pressurized-water reactors (PWRs) have completely implemented the GL reccmmendations in their designs.
- The boiling water reactors (BWRs) (except Oyster Creek and Big Rock Point) and the Combustion Engineering (CE) (except Palo Verde) PWRs concluded that the modifications recommended in the GL are not cost beneficial. The staff has agreed with the BWR Owners Group justification that no further rnodifications to the existing reactor vessel overfill protection system are.
necessary. Letters to individual BWR licensees requesting their commitment to 4
the BWROG resolution are being prepared. The staff is continuing its review of the CE Owners Group justification as it relates to assumptions on steam generator tube rupture probability.
Review of the Babcock & Wilcox (B&W) plants is continuing on a plant-specific -
basis because the B&W Owners Group has not taken a position on this issue.
4 Other UC issues with more than 3 open items are discussed below.
J A-9 Anticioated Transient Without Scram (ATWS) (A020)
Most operating reactors have installed systems to comply with the ATWS rule.
Some of the plants listed as having unimplemented items have systems that are installed and operable but that may require modification or inclusion of a design aspect to fully comply with the A1WS rule. An example of a remaining item that requires resolution before full implementation at the remaining 6 operating plants is General Electric trip unit delivery. Implementation is projected to be completed at all licensed plants by April 1994..
A-48 Hverooen Burns (S003)
The final hydrogen rule for BWR Mark 111 and PWR ice condenser containment types was published on January 25,1985. In September 1989, NUREG-1370,
Resolution of Unresolved Safety issue A-48, Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment," was published. NUREG-1370 concluded that no additional rule making, requirements, or guidance are necessary. NUREG-1370 stated that there are certain staff actions in the final stages, including completion of the generic SER on the Hydrogen Combine Owners Group (HCOG) repon for BWR Mark 111 containments, and analyses to demonstrate equipment survivability in ice condenser containments. The staff evaluations for these final stages have been completed.
On May 26,1993, the staff issued its safety evaluation regarding the Catawba and McGuire equipment survivability analyses. The staff concluded that essential-equipment in ice condenser containments would survive a wide spectrum of accident sequences involving hydrogen generation.
On June 26,1993, the staff issued a supplement to its August 6,1990 evaluation
" Acceptance for Referencing of Ucensing Topical Report Titled ' Generic Hydrogen Control Information for BWR-6 Mark 111 Containments - HGN-112-NP."' This supplement resolves HCOG concerns with the evaluation of August 6,1990, and enables licensees to finanAc and document their plant-specific analyses, in addition to completing the above reviews relating to Mark lli and ice condenser containments, the staff has resolved the recombiner issue of GL 84-09 for plants having Mark I containments but lacking post-accident hydrogen recombiners.
Ucensees for these facidties have upgraded or committed to upgrade existing nitrogen inerting systems to provide reliable post-accident nitrogen containment atmosphere dilution capability. Safety evaluations have been issued for all Mark 1 plants.
3.2 Verification Status For generic items such as USIs, NRR issues Tis, when appropriate, to specify which requirements are to be verified by the NRC after licensees have implemented the corrective actions specified in the USI resolution. The NRC performs these inspections, consistent with other inspection priorities, to verify proper implementation of the requirements. Verification is not considered complete until the required inspection is conducted in accordance with the TI, and an inspection report has been issued documenting that requirements have been adequately satisfied by the licensee.
On occasion, there may be issues for which the requirements'specified in the Tl for safety verification inspection are completed before total implementation of all aspects of the issue's resolution by the licensee.
Five Tis have been issued to provide guidance for the field verification of licensee implementation. The Tl designations and the corresponding USis are listed below.
Tl 2500/019 A-26 Reactot Vessel Pressure Transient Protection-Tl 2500/020 A-9 Anticipated Transient Without Scram Ti 2515/76 A-24 Qualification of Class 1E Safety-Related Equipment Ti 2515/85 A-7 Mark I Long-Term Program, NUREG-OSS1, Supplement 1 Tl 2525/120 A-44 Station Blackout Temporary instruction (TI) 2515/120, Station Blackout, was issued on September 24, 1993, and will be performed at 8 sites, involving all 5 regions, at which time an _
evalua'.lon will be performed by NRR to determine if additional sites need to be inspected.
Table 3.2 illustrates the items remaining to be verified for these five USIs Table 3.3 includes a summary of the verification status for each plant. Of the 423 items rr, quiring l
NRC verification,292 items (69 percent) have been completed.
l t
Summary of USI Items Requiring Verification Plants Plants Plants UE_1 Covered Required Verified A-7 Mark I Long-Term Program 24 24 24 A-9 Anticipated Transient Without 109 109 94 g
- A-24 Qualification'of Class 1E Safety-109-
~ 109 108 i
Related Equipment
-(
A-26 Reactor Vessel Pressure Transient 7 21
- 72 --
66-Protection
~ A-44.. Station Blackout -
109 109 0
NOTE:
Covered Plants are those for which uses are W-e?. -
Plants Requwod are those plants requenng field verscath.n.
Plants covered but for which field venfication is not necessary have implemented the resoluton in a rnanner not requmng plant hardware changos.
. Table.' 3.2 l
l
...# J m,.
1,
,,-- m m
. m..
3.3 Status by Plant 1
Table 3.3 summarizes information on the status of implementation and verification of USIs at all licensed plants. For each plant, the table shows the total number of applicable items, the number and percentage of items implemented, and the number of items remaining to be implemented. For those USts tha require the NRC to verify implementation actions, the table shows the number of items covered by a Tl at each plant, the number of items requiring verification, and the number and percentage of
. cms completed.
Eighteen plants have comp!eted all applicable USIs. Two plants have 6 items remaining to be implemented and 2 plants have 4 items remaining to be implemented.
The remaining 87 plaats have three or less items remaining to be implemented.
Five USIs require inspection to verify that implementing actions have been completed.
]
Of the 109 plants,105 have completed at least 50 percent of the applicable USis requiring verification. For the remaining 4 plants, NRC verification is complete for 1 of the 4 USIs that are applicable at those plants.
Appendix B lists the unimplemented USI items by issue and gives the projected 1
implementation date, where applicable.
SAFETY (SSUE MANAGEMENT SYSTEM STATUS OF USts -
SUMMARY
BY PIANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS I TEMS ITE9tS ITEMS PE R CENT UNIT APPLICABLE COMPLETED COMPLETED REMAIN!NG CDvtRED REQUIRED COMPLETED COMPLETED ARKANSA$.1 16 14 (87 1 2
4 4
3 (75 l
ARKANSAS 2 15 14 187 )
2 4
4 3
(75 BEAVER VALLEY 1 16 15 (93 l 1
4 4
3 175 l
BEAVER VALLEY 2 15 15 11001 0
4 4
3 l75 BIG ROCK POINT 1 14 12 185 1 2
3 3
2 (88 t
BRAIDWOOD 1 15 15 t1003 0
4 4
3 175 BRAIDWOOD 2 15 15 t1001 0
4.
4 3
1 75 l
BROWMS FERRY 1 18 12
-(66 l 6
4 4
2 i 50 I
l 2
4 4
3
- 75 l
EROWNS FERRY 2 18 16 (88 BROWNS FERRY 3 18-12 (66 l
6 4
4 2-($9 i
BRUNSWICK 1 18 16 (88 1
2 4
4 3
i75 i BRUNSWICK 2 18 16 188 )
2 4.
4 3
1 75 )
l BYRON 1 18 16 (1001 0
4 4
3 1 75 )
l BYRON 2 15
.15 (2001 0
4 4
3 i 75 l
'l CALLAWAY 1 16 16 11001 0
4 4
3 (75 3 i
CALVERT CLIFFS 1 18 13 181 l
3 4
4 3
(75 3.
CALVERT CLIFFS 2 18 13 181 l
3 4
4' 3
(75 CATAWBA 1 18 15 (93 1
4 4
3 (75 E
CATAWBA 2 16 15 (93 3 1
4 4
3
- 75 i
f CLINTON 1 15 14 (93 1
3 3
2
' 66 t
g a
COMANCHE PEAM 1 15 14 03 1
4 4.
1
' 75 i 50 i COMANCHE PEAM 2 18 16 Ila.
0 4
4 COOK.1 17 14 (82 1 3
4 4
3 1 75 l C00M 2 17 14 182 1 3
4 4
3 l 75 )
COOPER STATIO88 18 16 188 1 2
'4
~4 3
I.75 CRYSTAL RIVER 3 18 13 (81 l' 3
4 4
3 (75:
1 DAVIS-8 ESSE 1 16 15 (93 1 1
4 4
3 (75 DIABLO CANYDM 1 16 16-(1001
'O 4
4 3
(75 1
DIABLO CANYON 2' 15 15 (1001 0
4 4
3
.(75 DRESDEN 2 18 16
.(88 1 2
4 4
3 175 DRESDEN 3 18 15 (83 )
3 4
4.
3 (as DUANE ARMOLD 18 17 194 3 1-4 4
3 (75 I
FARLEY 1 16 14 187 1 2
4 4
3 (75 i
FERMI 2 '
18 15 (93-)
1 4
4 3
175 I FARLEY 2 16 16
.(1001 0
4 4
3 175 I FITZPATRICK 18 15 883 3 3
4 4'
3 (75
' FORT CALHOUN 1 16 13 181.I 3
4
-4
'3 (75 i
GINNA 16 14 187 8 2
4.
4-3 175 i
GRAND GULF 1 Ib 15 193 1 1
3 3
2 (88 l 1.
HADDAM NECK 18 15' 183: I
-3 4
4
.3 175 HARRIS 1 16 15
-893 1 1
4 4
3 f 75 75 HATCH 1 1R 16 (28 8 2
4
'e 3-HATCH 2 18 16 188 l'
2 4
4-3 L75 l
HOPE CREEh 1 17.
16
-494 3.
1 4
4 3
175.)
- Table 3.3
___ __.=____
_=
~
e
.m SAFETY ISSUE MANAGEMENT SYSTEM STATLfS OF (JSis - StNWARY BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PE R CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED INDIAN POINT 2 16 14 187 1 2
4 4
3 (75 i
INDIAN POINT 3 16 14 (87 5 2
4 4
3 (75 REWAUNEE 16 14 187 1 2
4 4
3 (75 LASALLE 1 17 16 194 l 1
3 3
2 (86 t
LASALLE 2 18 14
($F I 2
3 3
2 ISS 1
LIME RICK 1 16 16 11001 0
3 3
2 (se )
LIMERICK 2 16 16 (1001 0
3 3
2 las l
MAINE YANKEE is 16 11003 0
4 4
3 gy5 i
MCGUIRE 1 18 17 194 l
1 4
4 3
175 MCGUIRE 2 18 17 (94 i
'l 4
4 3
75 1
MILLSTONE 1 19 16 (84 1
3 4
4 3
75 MILLSTONE 2 18 13 181 )
3 4
4 3
i 75 MILLSTONE 3 18 15 193 1 1
4 4
3 175 )
MONTICELLO 18 16 (88 l 2
4 4
3 175 )
- NINE MILE POINT 1 18 16 188 l 2
4 4
3 175 l NINE MILE POINT 2 18 15 193 1 1
3 3
2 (se 3 NORTH ANNA 1 16 14 (87 l 2
4 4
2
{50 )
NORTH ANNA 2-17 15 tes 1 2
4 4
2 (50 1 e
OCONEE 1 18 14 18F )
2 4
4 1
i25 )
A OCONCE 2 18 14 (8F )
2 4
4 L
l'25 )
Ed OCONEE 3 18 14 (87 l 2
4 4
1 1 25 )
OYSTER CREEK 1 18 16 (88 1 2
4 4
3 1 15
)
PALISADES 16 13 (81 1 3
4 4
3
' F5 I
PALO VERDE 1 15 14 (93 3 1
4 4
3 75 i
PALO VERDE 2 15 14 193 1 1
4 4
3
- 75 t
PALO VERDE 3 15 14 (93 l 1
4 4
3
- 75 i
PEACH BOTTOM 2; 18 15 (83 1 3
4 4
3 (75 PEACH BOTTOM 3 18 15 183 )
3 4
=
3 175 i
3 3
2 165
' PILGRIM 1 18 16 (88 2
4 4
3 175 POINT BEACH 1 16 14 (SF 2
4 4
3 (75 POINT BEACH 2 18 14 187 l 2
4 4
3
-(F5 PRAIRIE ISLA@ 1 18 15 193 J 1
4
-4 3.'
175 PRAIRIE ISLAND 2 18 15 193 1 1
4 4
3 (75 i
OUAD CITIES 1 18 14 177 1 4
4 4
3 175 )
OUAD CITIES 2 18 14 177 3 4-4 4
3 P75 i
RIVER BEND 1 15 13 (86 3 2
3 3
2
' 68 i
ROSINSON 2 18 14 (S F -l -
2' 4
4 3
- 75 i
SALEM 1 16
.' 14
!** i 2
4 4
3 i 75 i
SALEM 2 I7-15 188 l 2
4 4
3 1 75 i
SAN ONOFRE 2 18 14 lSF )
2 4
4 3
1' 75 SAsi GNUFRE 3 is 14 (87 8 2
4 4
3 t15 i
. SEA 8 ROOM i 15 15 t106l 0-4 4
3 iF5 t
SEQUOYAH 1 18 17 (94 3 1
4 4
2 150 l i
Table 3.3 m
m m
m
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF Usts -
SUMMARY
BY PLANT IMPLEMENTATION VERIFICAT1001 r
ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE -COMPLETED COMPLETED REMAINING COVERED RTOUIRED COMPLETED COMPLETED SEDUOYAH 2 18 17 194 1 1
4 4
2 (59 south TEzAS 1 15 15 11001 0
4 4
3 (75
- ~
SOUTH TEKAS 2 15 15 (2001 0
4 4
3 (75 ST LUCIE 1 16 14 187 3 2
4 4
2 (50 ST LUCIE 2 16 15 193 1 1
4 4
2 (50 l
SUD99E R 1 16 16 (100)
O 4
4 3
(F5 SURRY 1 16 14 (8F 9 2
4 4
2 (50 SURRY 2 16 14-
.tSF l 2
4 4
2 (50 SUSOUEHANNA 1 1F 16 194 1 1
3 3
2 (66 SUSQUEHANNA 2 16 15 193 1 1
3 3
2 (66 i THREE MILE ISLAMD!1 16 15 (93 1 1
4 4
3 (F5 I TURKEY POINT 3 16 14 187 3 2
4 4
2 (50 )
TURKEY POINT 4 18 14 (87 3 2
4 4
2 (50 y VERMONT YANREE 1 18 17 198 1 1
4 4
3 (75 )
VOGTLE 1 15 14 193 1-1 4
4 2
(50 I VOGTLE 2 16 15 193 8 1
4 4
1 (25 t
WASHINGTON NUCLEAR 2 16 15 (93 5 1
3 3
2 (66 WATERFORD 3 15 13 tas i 2
4 4
3 (75 WOLF CREEK 1 16 16 (1001 0
4 4-
.3 (75 t
= $5 ZION 1 16 14
'tSF !
2 4
4 3-175.)
p ZION 2 16 14 (87 3.
2 4
4 3
(F5 i TOTALS / AVERAGES 1787 1610 90 177 423 423 292 89 Table 3.3
3.4 Status by Issue Table 3.4 presents summary information on the status of implementation and verification of each USI. For each issue, the table shows the number of applicable plants, the number and percentage of plants that have completed implementation, and the number of plants remaining to complete implementation. For those issues requiring NRC verification of corrective actions, the table shows the number of plants covered by the issue, the number of plants at which verification is required, and the number and percentage of plants that have completed verification.
i Of the 27 USis,19 have been fully implemented. (USis A-3, A-4, and A-5 relate to steam generator tube integrity for the three major PWR vendors and are considered separate issues.) Three USIs account for 90 percent of the unimplemented items:
A-44, Station Blackout, with 60 plants remaining to complete implementation; A-46, Seismic Qualification of Equipment in Operating Plants, with 61 plants remaining to complete implementation; and A-47, Safety implication of Control Systems, with 39 plants remaining to complete implementation. These three, largely unimplemented, USis are in varying stages of NRC review and licensee implementation, as discussed in Section 3.1 of this report. Six plants have not implemented corrective actions for A-9, Anticipated Transient Without Scram, and 6 plants have not implemented corrective actions for USl A-48, Hydrogen Control Measures and Effects of Hydrogen Burns. The remaining USIs have 1 or 2 plants remaining to complete implementation.
1 NRC inspection to verify licensee implementation is required for five USIs and is l
complete for USl A-7, Mark I long-term program. Station blackout accounts for 109 of the 292 outstanding verifications.
i l
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.___m._
I SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF USis -
SUMMARY
BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLArlS '
PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLE TE D REMAINING REQUIRED COVERED REQl';#ED COMPLETED COMPLETED
-.-..a.--
A.39 36 36
- (1001 0
No DETERMINATION OF SRV POOL DYNAMIC LOADS & TEMP. LIMITS FOR SWR CNTMNTS A-40 3-3 (100) 6 NO SEISMIC DESION CRITERIA A.42 33-38 (1001 6
NO PIPE CRACKS IN BOILING tmTER REACTORS A-43 109 109 (2001 0
NO CONTAINENT EMERGENCY SUMP PREFORMRNCE A
%d A-44 109 49 144 )
80 YES 109 109 0
le t STATION BLACKOUT A-4S 158 109 (2001 0
NO SHUTDOWN DECAY HEAT RCdOVAL REQUIREENTS A-48 98 S
(7-l 81 NO SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS A.47 109 70 (64 1 39 NO SAFETY IMPLICATION OF CONTROL SYSTEMS A.48 47 41 (37 I 8
NO HYDR 00EN CONTROL MASURES AND EFFECTS OF HYDROGEN BURNS A-49 72 LT2 (100).
.O NO PRESSURIZED THERMAL SHOCK
~
Table 3.4; a-.
---Jan.'
m' w
v s
r mr d
A 4'*^*'W 2
-w-. - -
.2 1
x a
l l
3.5 Conclusions l
l After a detailed review of the implementation and verification status of the resolution of j
the 27 USIs, the NRC staff has concluded the following:
The NRC closure process for USIs ensures continued adequate protection of the j
public health and safety.
A!! USis have been resolved by the NRC, and progress has been made in implementing and verifying required changes at plants.
Ucensees are making adequate progress toward ir.alementing recluirements imposed following the NRC's resolution of USIs, and the framewom exists to oversee future implementation of delayed items.
Although the resolution of USis involves complex technical issues and analyses, it appears that all required implementation items can be completed in accordance j
with regulatory requirements.
j l
i i
i j
4 GENERIC SAFETY ISSUES This section presents the overall status of implementation and verification of GSis applicable at the 109 licensed plants. Because each GSI may be tracked under different designations, Table 4.1 cross-references the GSI and sub-issue number and the SIMS numbers used in the tables and appendices of this report.
i 4.1 Imolementation Status Licensees achieve implementation of GSI items either by incorporating corrections into the plant design before licensing or by making the modifications necessary to meet the requested actions at licensed plants. The information presented here includes all GSI items related to the 109 licensed plants considered in this report.
Approximately 94 percent of the GSI items have been implemented at licensed plants.
Of the 2,621 items,2,463 have been completed and 158 remain open from an implementation standpoint. On average, each plant has less than 2 items to implement, and no plant has more than 7 remaining items. Figure 4.1 presents the 4
overall status of, and progress on, GSis. Of the 109 licensed plants,39 have implemented all applicable GSIs. Table 4.2 lists the number of unimplemented items by unit. Appendix C lists the unimplemented GSI items by issue and projected implementation dates.
Five GSis have not been implemented at a number of plants for which they are applicable; these account for approximately 90 percent of the unimplemented items.
Figure 4.2 summarizes the implementation status of these issues. A brief description of each issue follows, t
GSI 43 Reliability of Air Systems (8107) in August 1988, the staff issued GL 88-14 to specify the performance of a design and operations verification of instrument air systems and descriptions of licensees' programs for maintaining proper instrument air quality. The staff gave licensees 6 months in which to confirm that these actions had been accomplished or to commit to perform them during a subsequent outage. The licensees for operational plants that still have this issue open are scheduled to complete implementation by the end of 1993. Most of the plants that still have this issue open have completed 80 to 90 percent of the significant recommended actions and are awaiting a suitable outage opportunity to complete the final actions. The staff believes that the planned completion schedules do not pose any significant safety risk.
h GSI Numbers and Corresponding SIMS Item Numbers SIMS MPA ltem No.
GSI No.
_N_o.,
SIMS Title 40 40 B065 Safety Concems Associated With Pipe Breaks in BWR Scram System 41 41 B050 BWR Scram Discharge Volume Systems GL-88-14 43 B107 Instrument Air Supply System Problems Affecting Safety-Related Equip.
GL-89-13 51 L913 Service Water System Problems Affecting Safety-Related Equipment 67.3.3 67.3.3 A017 Improved Accident Monitoring 70 70 B114 PORV and Block Valve Reliability 75 (B076) 75, item 1.1 8076 Item 1.1 - Post-Trip Review; Program Description & Procedures 75 (8085).
75, item 1.2 B085.
Item 12 - Salem ATWS 12 Data Capability 75 (8077) 75, item 2.1 B077 Item 2.1 - Equipment Classification & Vendor interface - RTS Component 75 (8086) 75, item 2.2.1 B086 Item 2.2.1 - Salem ATWS 22 S-R Components GL-90-03 75, item 2.22 LOO 3 Item 2.22 - Relaxation of Staff Pos in Gen Letter 83-28. Item 22 Part 2 75 (8078) 75, items 3.1.1 & 3.1.2 B078 Items 3.1.1 & 3.1.2 - Post-Maintenance Test Procedures & Vendor Recomm.
75 (B079) 75, item 3.1.3 B079 Item 3.1.3 - Post-Maintenance Testing - Changes to Tech Specs - RTS Comoonent 75 (8087) 75, items 3.2.1 & 322 B087 Items 3.2.1 & 3.22 - Salem ATWS 32.1 & 322 S-R Components 75 (8088) 75, item 32.3 B088 Item 32.3 - Salem ATWS 32.3 T.S. S-R Components) in 75 (8080) 75, item 4.1 B080 item 4.1 - Reactor Trip System Reliability - Vendor Related Mods P
75 (8081) 75, items 4.2.1 & 4.22 B08t items 42.1 & 4.22 - Preventative Maint Prog for Reactor Trip Breakers 75 (B082) 75, item 4.3 8082 Item 4.3 - Automatic Actuation of Shunt Trip Attach. fer West & B&W 75 (B090) 75, item 4.3 8090 item 4.3 - Salem ATWS 4.3 W and B&W T.S.
75 (8091) 75, item 9.4 8091 Item 4.4 - Salem ATWS 4.4 B&W Test Procedures 75 (B092) 75, item 4.5.1 B092 Item 4.5.1 - Salem ATWS 4.5.1 Diverse Trip Features 75 (8093) 75, items 4.5.2 & 4.5.3 B093 Items 4.52 & 4.5.3 - Salem ATWS 4.52 & 4.5.3 Test Altematives 86 86 B084 Long Range Plan Dealing With Stress Corrosen Cracking in BWR Piping GL-88-03 93 B098 Resolution of GSI 93," Steam Binding of Auxi!iary Feedwater Pumps" 94 94 B115 Additional Low-Temp Overpressure Protection for LWRs GL-88-17 99 L817 Loss of Decay Heat Removal 124 124 S001 Auxiliary Feedwater System Reliability GL-80-099 A-13 B107 Technical Specification Revision for Snubber Surveillance GL-84-13 A-13 B022 Technical Specification for Snubbers A-16 A-16 D012.
Steam Effects on BWR Core Spray Distribution MPA-B023 A-35 B023 Degraded Grid Voltage B-10 B-10 S008 Behavior of BWR Mark til Containments Dev Design, Test & Maint Criteria for Atmo Cleanup Sys Air Filter & Adsorption Units B-36 B-36 none GL-80-014 B-63 B045 LWR Primary Coolant System Pressure Isolation Valves Table 4.1
Generic Safety issues implementation Status at Licensed Plants i
a00.-
j 2000 --
)
2 l
1500 --
l p
2621 2463 H
)
1000 500 --
{
0 I
I l
Applicable implemented Unimplemented i
I Figure 4.1 l !
l
(
Summary of Unimolemented GSI Items by Plant Items items items PLANT Remaining PLANT Remaining PLANT Remaining Arkansas 2 1
Haddam Neck 4
Palo Verde 2 1
Beaver Valley 1 3
Hatch 1 1
Perry 1 2
i Beaver Valley 2 2
Hatch 2 1
Point Beach 1 2
Braidwood -1 1
Indian Pt 2 3
Point Beach 2 2
Braidwood 2 2
indian Pt 3 2
Quad Cities 1 2
Browns Ferry 1 6
Kewaunee 3
Quad Cities 2 3
i Browns Ferry 2 2
LaSalle 1 1
Robinson 2 2
Browns Ferry 3 7
LaSalle 2 1
Salem 1 2
Calvert Cliffs 1 4
Maine Yankee 2
Salem 2 2
Calvert Cliffs 2 5
McGuire 1 4
San Onofre 2 -
1 Catawba 1 2
McGuiro 2 4
San Onofre 3 1
Catawba 2 2
Millstone 1 3
South Texas 1 -
3 Cook 1 3
Millstone 2 3
South Texas 2 3
Cook 2 3
Millstone 3 3
St. Lucie 1 1
Cooper Station 1
Nine Mile Pt 1 1
St. Lucie 2 1
Crystal River 3 5
- Nine Mile Pt 2
_1 Summer 1 3
Dresden 2 1
North Anna 1 2
Surry 1 2
+
4 Dresden 3 2
North Anna 2 2-
- Surry 2 2
.m Farley 1 1
.Oconee 1 1
Turkey Pt 3 2
Farley 2 1
Oconee 2-1 Turkey Pt 4 2
Fermi 2 1
Oconee 3 1
- Wolf Creek 1 1
Ft Calhoun 1 3
Oyster Creek 1 2
' Zion 1 4
Ginna 4
Palisades
.1 Zion 2 4
Grand Gulf 1 1
i Table 4.2 J
Summary of Five Unimplemented GSis 120 -
(109)
(109)
(109) 18__
100 -
3881 80 -
(74) 60 -
n I**I 40 -
(36) 4
(
?>
g
( )
20 -
(12) 0-l GL-88-14 GL-89-13 67.3.3 70 94 (8107)
(L913)
(A017)
(B114)
(B115) 8 ""*
Legend
@ Total U Implemented E unimpi.mented Figure 4.2
GSI 51 Prooosed Reauirements for Imorovina the Reliability of Open-Cvcle Service Water Systems (L913) j This issue was developed as a result of uncertainties regarding the compliance of service water systems with the regulations. In July 1989, the staff issued GL 89-13 requesting licensees to take certain actions and establish programs to ensure continued compliance of their service water systems with the applicable regulations. The staff asked licensees to submit implementation plans and schedules by early 1990. The actions and programs have been implemented at approximately 80 percent of all plants. Temporary instruction T12515/118 was issued on December 29,1992, to assess the licensees' planned or completed actions in response to GL 89-13. The staff considers the status of this GSI acceptabic.
GSI 67.3.3 Imoroved Accident Monitorina (A017)
This issue addresses conformance with RG 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident." The staff issued GL 82-33 in December 1982 to request licensees to submit schedules and details of their plans to implement the provisions of RG 1.97, Revision 2. The licensee responses to this generic letter prompted the staff to issue confirmatory orders in 1985. Because the industry has taken exception to and appealec'some of the provisions of RG 1.97, Revision 2, implementation is incomplete at many plants.
The issue of Category 1 neutron flux monitoring system at BWRs has been resolved, although some PWR licensees have yet to install Category 2 qualified instrumentation to monitor containment sump water temperature. The staff is seeking closure on the PWR issue on a generic basis. Some other plant-specific issues still remain, and supplemental safety evaluations are being prepared to close those issues.
GSI 70 PORV and Block Valve Reliability (B114). and GSI 94 Additional Low-Temoerature Overoressure Protection for Liaht-Water Reactors (B115)
The staff determined that the low-temperature overpressure protection (LTOP) system unavailability is the dominant contributor to risk from low-temperature overpressure transients. The staff further concluded that a substantive improvement in availability, when the potential for an overpressure event is the j
highest, and especially during water-solid operations, can be achieved through improved administrative restrictions on the LTOP system.
The staff considered the conditions under which a low-temperature overpressure transient is most likely to occur. While LTOP is required for all shutdown models, the most vulnerable period was found to be MODE 5 (cold shutdown) with the,
i
reactor coolant temperature less than or equal to 200 F. The basis of the detailed evaluation of operating reactor experiences performed in support of GI 94. LTOP transients that have challenged the system have occurred with reactor coolant temperatures in the range of 80 F to 190 F. In addition, a review of the STS for containment integrity indicates that there are no specific requirements imposed during MODE 5, when the reactor coolant temperature is below 200 F. Industry responses to GL 87-12, " Loss of RHR While RCS Partially Filled," dated July 9, 1987, also indicate that containment integrity during MODE 5 is often relaxed to allow for testing, maintenance, and the repair of equipment.
In all instances when pressure / temperature limits in the TS have been exceeded, one LTOP channel was removed from service for maintenance-related activities.
During these events the redundant LTOP channel failed to mitigate the overpressure transient as a result of a system / component failure that had not been detected.
The current 7-day AOT for a single channel is considered to be too long under certain conditions. The staff concluded that the AOT for a single channel should be reduced to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when operating in MODE 5 or 6 when the potential for an overpressure transient is highest. The operating reactor experiences indicate that these events occur during planned heatup (restart of an idle reactor coolant pump) or as a result of maintenance and testing errors while in MODE 5. The reduced AOT for a single channel in MODES 5 and 6 will help to emphasize the importance of the LTOP system in mitigating overpressure transients and provide additional assurance that plant operation is consistent with the design basis transient analyses.
On the basis of the forgoing concerns, sdded assurance of LTOP availability is to be provided by revising the current technical : pacification for overpressure protection to reduce the AOT for a single channel from 7 days to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the plant is operating in MODES 5 and 6. The guidance provided also is applicable to plants that rely on both PORVs and RHR SRVs or that rely on RHR SRVs only.
4.2 Verification Status For generic items such as GSIs, NRR issues Tis for those items that need to be verified in the field by the NRC staff after licensees have implemented the actions specified in the GSI resolution. The NRC performs these inspections, consistent with other inspection priorities, to verify proper implementation of the requirements.
Verification is not considered complete until the required inspection is conducted in accordance with the Ti and an inspection report has been issued documenting that the requirements have been adequately satisfied by the licensee. On occasion, there may be issues for which the requirements specified in a Tl for safety verification inspection are completed before full implementation of all aspects of the issue's resolution by the licenser Of the 1,176 items requiring NRC verification,1,045 (89 percent) have beer, ampleted.
58-
Eight Tis provide guidance for the field verification of licensee implementation of GSIs.-
Ti designations and the corresponding GSIs are provided in Table 4.3. Table 4.4 summarizes the items remaining to be verified.
j i
1 1
4 1
i i
e 1
f I
i j
1 d
4 k
- 4
Temporary Instructions for Resolved GSis
_SIMS item SIMS Title
.Tl 41 BWR SCRAM DISCHARGE VOLUME SYSTEM 2515/090 67.3.3 IMPROVED ACCIDENT MONITORING 2515/087 75 (B077)
ITEM 2.1 - EQUIPMENT CLASSIFICATION &
2515/064 VENDOR INTERFACE - RTS COMPONENT 75 (B078)
ITEMS 3.1.1 & 3.1.2 - POST MAINTENANCE TEST 2515/064 PROCEDURES & VENDOR RECOMM.
75 (8079)
ITEMS 3.1.3 - POST MAINTENANCE TESTING -
2515/064 CHANGES TO TECH SPECS - RTS COMPONENT 75 (B080)
ITEM 3.1 - REACTOR TRIP SYSTEM RELIABILITY -
2515/091 VENDOR RELATED MODS 75 (B081)
ITEMS 4.2.1 & 4.2.2 - PREVENTIVE 2515/064 MAINTENANCE PROGRAM FOR REACTOR TRIP BREAKERS SALEM ATWS 3.2.1 & 3.2.2 S-R COMPONENTS 75 (B086)
SALEM ATWS 2.2 S-R COMPONENTS 2515/064 75 (8087)
SALEM ATWS 3.2.1 & 3.2.2 S-R COMPONENTS 2515/064 75 (B088)
SALEM ATWS 3.2.3 T.S. S R COMPONENTS 2515/064 75 (B092)
SALEM A1WS 4.5.1 - DIVERSE TRIP FEATURES 2515/064 86 LONG RANGE PLAN DEALING WITH STRESS 2515/089 CORROSION CRACKING IN BWR PIPING GL-88-17 LOSS OF DECAY HEAT REMOVAL 2515/101 2515/103 GL-89-13 SERVICE WATER SYSTEM PROBLEMS 2515/118 AFFECTING SAFETY-RELATED EQUIPMENT Table 4.3 60-
._._.m l
L Summ ry cf GSI Itams Requiring Vsrificatien Plants flants Plants SIMS Rein Covered Required Verified 41 BWR SCRAM DISCHARGE VOLUME SYSTEMS 37 37
'36 67.3.3 IMPROVED ACCIDENT MON 110 RING 109 108 94 75 (B080)
ITEM 4.1 - REACTOR TRIP SYSTEM 72 72 71 RELIABluTY-VENDOR RELATED MODS 86 LONG RANGE PLAN DEALING WITH STRESS 36 36 4
CORROSION CRACKINGIN BWR PIPING M
GL-88-17 LOSS OF DECAY HEAT REMOVAL 72 -
72 66 Y
. SERVICE WATER SYSTEM PROBLEMS 109-
_107 0
GL-89-13 AFFECTING SAFETY RELATED EQUIPMENT NOTE: Plants Covered are those for which GSis are apphcable.
Plants Required are those plants requenng field venheation. :
Plants covered but for which field venfication is not necessary have implemented the reaahann in a menner not requinng plant hardware changes.
r t.
Table 4.4 4
4.3 Status by Plant Table 4.5 summarizes the status of implementation and verification of GSis at all licensed plants. For each plant, the table shows the total number of applicable items, the number and percentage of items implemented, and the number of items remaining to be implemented. For those GSis that require NRC to verify implementation of corrective actions, the table shows the number of items covered by the Tis at each plant, the number of items requiring verification, and the number and percentage of items completed. Appendix C lists the unimplemented GSI items by issue and gives projected implementation dates.
Of the 109 plants,39 have completely implemented all GSI items. Forty-seven plants have completed implementation actions for all except 1 or 2 GSis; 19 plants have 3 or 4 items to implement; and the remaining 4 plants have 5 to 7 items to implement.
Of the 109 plants, one plant has completed all of the items requiring verification by inspection (in accordance with a TI); 107 plants have completed all but 1 or 2 items requiring verification; and one plant has completed all but 3 items.
II SAFETY ISSUE MANAGEMENT SYSTEM j
STATUS OF GSts - SUnmARY BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PE R CENT ITEMS ITEMS
-ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLE TED COMPLE TE D REMAINING COVE RED. REQUIRED. COMPLETED
-COMPLETED
---~~----
ARMANSAS 1 27 27 (1001 0
12 4
3 (75 l ARRANSAS 2 25 24 (95 l 1
12-4 3
( 75 ")
BEAVER VALLEY !
. 26 23 (88 1 3
'12 12 11 (91 l Bt AVER VALLE Y 2 :
27 25 (92 )
2-12' 4
3 (75 )
8IG ROCK P011ET.1 21 21
{100) 0 11 to 9
(89 )
BR AID 40D - 1 27 26 (96 l 1
12 12 11 (91 i BRAirHOOD 2 27 25 (92 l 2
12 12-11 (91 i BROWNS FERRY 1.
21 15 (71 )
6 11 11-9 (81 (90 -l BROWNS FERRY 2 21 19 (90 )
2 11 11 10 l.
BROWNS FERRY 3 21 14 (66 4 7
11 11 9
(81 )
8RUNSWICM 1-19-19 (100)'
O 11 11 10-(90 )
BRUNSWICK 2 19 19 (2001 0
11 11 10 (90 l BYRON 1 27 27 (100) 0 12 12 -
11 (91 )
BfRON 2 27.
27-(1001 0
12 12 11
'(91 l CALLAWAY 1
~ 26 26
{100).
0 12.
12 111 (91 l CALVERT CLIFFS 1 23 19 (82 )
4 12 12 11 (91 )
CALVERT CLIFFS 2 23 18 (78-)
5 I2-12 11 (91 )
CATAW8A 1 27 25 (92 3
?'
12 12 '
11
-(91 l CATAW8A 2 27
'25 (92 l 2
12
- 12 11 (91 I f
CLINTON 1-21 21 (1001 0
11 11
-10' (90 )
COMANCHE PEAK 1 26 26~
(100) 0 12' 11 11 (100)
COMANCHE PEAK 2 26 26
-(1001 0
12 11
.8 (72 )
COOK 1-25 22' (87 1 3
12 12 11 (91 )
CDOM 2 25-22.
(87 8 3
12.
12 11 (91 l
. COOPER STATION 21 20 (55 l 1
11 4
3 (75 3 CRYSTAL RIVER.27 22 (81 1 5
12 '11 (91 l-DAV.S-FESSE 1 DIA3 0 CANYCN 1.
.26 26 (100) 0 12 12 10-(83 i 26 26 (100l' O
12.
.12 <
11 (91 1 DIAELO CANYON 2 26 26 (100) 0
- 12 11 (91.)
.DRESDEN 2-20
- 19
-(94 1:
1 11 11 3
(81 )
DREsDEN 3
'20
. 18.
{s9 1 2
11 11-s (81 3-DUANE ARNOLD
~21.
21
.(1003 0
11 11.
10_
(90 I FARLEY 1 26
~25 (96: 1 L'
12 12.
11' (91 )
FARLEY 2 26 25 (96 8 1.
12-12
.11 -
191 l
' FERMI 2 21 20
-(95 1
-1 11, 11
. 10 (90 i 9
.(81 )
FITEPATRICK-21 --
i21 (100)-
0
.11 11.
FORT CALHOUIl 1" 25
-22 (87 3 3
12 9
3 (88 i GINNA 26 22' (84 3 4-
'12' GRAND GULF 1.
21:
20 (95 3 1
11
~ 12 11-(91 3-gg g
.gg! g l:
HAODAM NECK 26 22
' ( 8 4 -. I 4
12 12
,11-(91 )
j HARRIS 1 27 27 (200)
.O.
-12 12 11
. ( 91 I i.
HATCH 1 21 20 -
(95 J '.
1
-11 11 10 (90 1-
. HATCH 2 21 20 (95 l 1
11-
~11 10 (90 -l
. HOPE CREEK 1 '
.22
-22 (1001-O
}11-4 3
.(75 )
I l'
i L
l-l' Table 4.5-I l
l
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF GSts -
SUMMARY
BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS IIEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED INDIAN POINT 2 25 22 (82 1 3
12 12 11 (91 1 INDIAN POINT 3 26 24 (92 1 2
12 12 11 (91 1 MF.WAUNE E 25 22 (87 1 3
12 12 11 (91 l LASALLE 1 22 21 (95 1 1
11 11 9
(81 )
LASALLE 2 22 21 (95 1 1
11 11 9
(81 l LIMERICK 1 20 20 l1001 0
11 4
3 (75 i
LIMERICK 2 20 20 (100) 0 11 4
3 (75 1
MAINE YANKEE-24 22 (91 1 2
12 12 11 (91 MCGUIRE 1 26 22 (84 1 4
12 12 11 (91 MCGulRE 2 26 22 (84 )
4 12 12 11 (91 MILLSTONE 1 21 18 (85 1 3
11 10 (90 i
MILLSTONE 2 24 22 (91 l 2
12 12 11 (91 1 MILLSTONE 3 27 24 (88 1 3
12 4
3 (75 1 MONTICELLO 20 20 (1001 0
11 11.
10 (90 i NINE MILE POINT 1 22 21 (95 l 1
11 11 10 (90 I NINE MILE POINT 2 22 21 (95 1 1
11 4
3 (75 )
NORTH ANNA 1 26 24.
(92 3 2
12 12 11 (91 )
NORTH ANNA 2 26 24 (92 1 2
12 12 11 (91 )
OCONEE 1 26 '
25 (96 l 1
12 12 10 (83 1 1
OCONEE 2 26 25 (96 1 1
12 12 10 (83 )
W OCOMEE 3 26
-25 (96 3 I
12 12 10 (83 )
Y OYSTER CREEK 1 22 20 (90 1 2
11 11 10 (90 )
PALISADES 24 22 (91 1 2
12 12 10 (83 i PALO VERDE 1 23 23-(1001 0
12-
'12 11 (91 i PALO VERDE 2 23 22 (95 1 1
12 12 11 (91 )
PALO VERDE 3 23 23 (100) 0 12 12 11 (91 1 PEACH BOTTOM 2 21 21 (2001 0
11 11 10 (90 i PE ACH BOTTDM 3 21 21 (100) 0 11 11 10 (90 3.
PERRY 1 22 20 (90 1 2
11 11 10 (90 i PILGRIM 1 21 21 (1001 0
11 11-10 (90 l POINT BEACH.1' 25 23 (91 1-2 12 12
-10 (83 i POINT BEACH ?
25 23 (St i 2
12 12 10 (83 i PRAIRIE ISLAND 1 26 26 (1001 0
12 12 11 (91 1 PRAIRIE ISLAND 2 26 26 (1003 0
12 12 11 (91 I QUAD CITIES 1 21 19 (90 l 2
11 11 9
- (El )
QUAD CITIES 2 21 18-(85 1 3
11 11 9
(81 )
- ROBINSON 2 26 24 (92 l-2 12 12 11
. (6E I RIVER BEND 1 20 20 (1001 0
to 3
2 (91 1 SA LEM 1-26 24 (92 1 2
12-12 11 (91 i SALEM 2 26 24 (92 1 2
12 12 11 (91 l SAN ONOFRE 2 24 23 (95 l-1-
12 12 11 (91 l SAN ONOFRE 3.
24 23 (95 1 1
12 12 11 (91 l SEABROOK l' 27 27 (100) 0 12 12-11 (91 i SEQUOYAH 1 27 27 (1001 0
!?
12 10 (83 I Table 4.5'
.m
.... m m
. m
-+-
.,---e e-m
.ui
.....m
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF GSis -
SUMMARY
BY PLANT IMPLEMENTATION VERIFICATION ITEMS TTEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLE TED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED SEQUOYAH 2 27 27 (1001 0
12 12 10 183 l SOUTH TEXAS 1 27 24 (88 1 3
12 12 il 191 )
SOUTH TEXAS 2 27 24 188 5 3
12 6
5 183 )
ST LUCIE 1 24 2
195 1 1
12 12 11 (91 ST LUCIE 2 25 2
(95 1 1
12 12 11 191 SUD99ER 1 26 23 (88 )
3 12 12 11 I91 SURRY !
26 24 192 )
2 12 12 11 (91 )
SURRY 2 26 24 (L2 1 2
12 12 11 191.
SUSouEHANNA 1 20 20 (2005 0
11 11 10 (Se SUSQUEHANNA 2 20 20.
I1001 0
11 11 10 t90 i THREE MILE ISLAND 1 25 25 t1001 0
12 12 11 (91 1 TURKEY POINT 3 25 23 191 l 2
12 12 to (3-p TURKEY POINT 4 25 23 (91 1 2
12 12 10 183 )
VERMONT YANKEE 1 21 21 11001 0'
11 11-10 190 )
V0GTLE 1 27 27 (1005 0
12 12 11 (91 )
VDGTLE 2 27 27 (1001 0
12 12 11 191 )
WASHINGT001 NUCLEAR 2 21 21 (1001 0
11 11 10 (90 )
WATERFORD 3 23 23 (2003 0
12 4
3 (75 )
WOLF CREEK 1 27 26' (96 l 1
!?
12 11 191 )
ZION 1 26 22 (84'I 4
12 12 11 191 )
ZION 2 26 22 (84 )
4 12 12 11 191 P TOTALS / AVERAGES 2621
.'463 94 158 1270 1176 1045-88
be-
1 4.5
_Qpnclusions After detalicd review of the implementation and verification status of the resolution of GSis and sub-issues, the NRC staff has concluded the following:
The NRC closure process for GSis is adequate to protect the public health and safety.
Ucensees are making significant progress toward implementing GSI-related actions
]
requested by the staff, and the framework exists to oversee future implementation of delayed items.
Significant progress has been made in verifying the completion of implementation actions associated with those GSis that have been resolved.
The overall status of the 5 largely unimplemented GSis is generally acceptable because of the relatively recent issuance of staff positions on three of the GSis, I
and projected implementation schedules for the remaining 2, 4
1 2
k i
a
(
1 2
5 OTHER MULTIPLANT ACTIONS This section presents the overall status of implementation and verification of other MPAs not related to TMI Action Plan requirements, USis, or GSis. The MPAs are applicable to the 109 licensed plants. Because each MPA may be tracked under different designations, Table 5.1 cross-references the MPA number and the SIMS number used in the tables and appendices of this report.
j 5.1 Imolementation Status l
l Ucensees achieve implementation of MPA items either by incorporating corrections i
into the plant design before licensing or by making the modifications necessary to i
meet the requested, required or voluntary actions at licensed plants. The information presented here includes all MPA items related to the 109 licensed plants considered in this report.
1 l
Approximately 87 percent of the MPA items have been implemented at licensed plants.
Of the 7,517 applicable items,6,578 have been completed and 941 remain open from l
an implementation standpoint. On average, each plant has less than 9 remaining items to implement. No plant has more than 12 remaining items except Browns Ferry Units 1 and 3. Each unit has 20 items. Figure 5.1 presents the overall status of, and progress on, MPAs. Of the 109 licensed plants, none have fully implemented all applicable MPAs. Table 5.2 lists the number of unimplemented MPA items by plant i
and projected implementation dates. Appendix D lists the unimplemented MPA items by issue and projected implementation dates. MPAs are of a dynamic nature. New I
MPAs can and will be added as the situation dictates during coming years.
MPAs BL-88-11, BL-92-01, BL-93-02, BL-93-03, GL-88-20, GL-89-10, GL-92-01, GL 08, GL-93-04, MPA-B118, and MPA-B122 account for 85 percent of the 941 unimplemented items. Figure 5.2 summarizes the implementation status of these issues. A brief description of these 11 MPAs follows:
BL-88-11 Pressurizer Surae Line Thermal Stratification (X811) j The NRC issued IN 88-80, " Unexpected Piping Movement Attributed to Thermal i
Stratification," on October 7,1988, to alert licensees of PWRs of the phenomenon.
The NRC further issued BL 88-11, " Pressurizer Surge Line Thermal Stratification,"
i on December 20,1988, describing a series of short-and long-term actions to cddress the problem.
Ucensees of operating PWRs were required to (1) perform a visual inspection of the surge line at the tast available cold shutdown (of greater than 7 days duration) after issuance of the bulletin and (2) perform an analysis that demonstrated that the surge line met applicable design codes and other FSAR and reguintory commitments for the licensed life of the plant. If the analysis did not show
. 2
SIMS issue Numbers and Corresponding MPA Numbers l
SIMS MPA l
Item No.
No, S!MS Title BL-88-08 X808 Thermal Stress in Piping BL-88-11 X811 Thermal Stratification in PZR Surge Line (BL 88-11)
BL-92-01 X201 Thermal Lagging 330 (BL 92-01)
BL-93-02 X302 Debris Plugging of Emergency Core Cooling Suction Strainers BL-93-03 X303 Reactor Vessel Water level Instrumentation in BWRs GL-84-09 A019 Recombiner Capability BWR Mark l GL-87-09 D024 Mode Changes & LCO's - Tech Specs 3.0.4 and 4.0.4 (GL 87-09)
GL-88-01 B097 IGSCC Problemsin BWR Piping GL-88-11 A023 R.G.1.99 Rev 2 (pressurized Thermal Shock Rule) (GL 88-11)
GL-88-20 B111 Individual Plant Evaluations (GL 88-20)
GL-89-01 D025 Relocate RETS to Admin Section of Tech Specs GL-89-04 A025 IST Reviews and Schedules (GL 89-04)
GL-89-06 F072 Safety Parameter Display System - Response to GL 89-02 GL-89-10 B110-Motor Operated Vahre Testing and Surveillance (GL 89-10)
GL-89-16 B112 Installation of Hardened Wetwell Vent (GL 89-16)
GL-91-08 D030 Removal of Component Lists fror Tech Specs (GL 91-08) l GL-91-11 L111 Vital Instruments Buses and Tie Breakers (G-48.49)
GL-92-01 B120 Reactor Vessel Structural Integrity GL-92-04 B121 BWR Water LevelInstrumentation GL-92-08 L208 Thermo-Lag 330-1 Fire Barriers GL-93-04 L304 Rod Control System Failure ard Withdrawalof Rod Cetrol Cluster Assemblies MPA-B116 B116 Consider Results of Soonsored Motor-operated Tests (GL 89-10, Supp 3)
MPA-B117 8117 Failure of WestinghNse SG Tube Mechanical Plugs (BL 90-01, Supp 2)
MPA-B118 B118 IPE Extemal Eve:4s (GL 88-20 Sulp 4)
MPA-B122 B122 Loss of Fill-Oilin Transmitters Manufactured by Rosemount (BL-90-01)
L i
l Table 5.1 1
I
_m _ _ _.. _
i Summary of Unimolemented MPA Items by Plant
!?
Iteras
_ Items items PLANT Remaining PLANT Remaining PLANT Remaining ~
f Arkansas 1 8
Ginna 12 Pilgrim 1 -
7 Arkansas 2 8
Grand Gulf 1 8
Point Beach 1 9
Beaver Valley 1 10 Haddam Neck 6
Point Beach 2 -
9 Beaver Valley 2 10 Harris 1 9
Prairie Island 1 9
Big Rock Point 1 6
Hatch 1 12 Prairie Island 2 9
Braidwood 1 11 Hatch 2 12 Quad Cities 1 9
Braidwood 2 12 Hope Creek 1 7
-Quad Cities 2 9
Browns Ferry 1 20 Indian Pt 2 8
River Bond _1 7
Browns Ferry 2 10 Indian Pt 3 6
Robinson 2 8
Browns Ferry 3 20 Kewaunee 8
Salem.1 10 Brunswick 1 9
LaSalle 1 11 Salem 2 9
s Brunswick 2 9
LaSalle 2 12 San Onofre 2 10
'lL Byron 1 11 Limerick 1 8
. San Onofre 3 11 i
Byran 2 11 Limerick 2 9
Seabrooit 1-5 Callaway.1 10 Maine Yankee 7-Sequoyah 1 7
Calvert Cliffs _1 5
McGuire 1 9
Sequoyah 2 8
Calvert Cliffs 2 6
McGuire 2 9
South Texas 1 8
Catawba 1 7
Millstone 1 8-South Texas 2 7
O Catautba 2 7
Millstone 2 6
St. Lucio 1 7
?
Clinton 1 8
Millstone' 3 7-
.St. Lucie 2 7
Comanche Peak-1
-8 Monticello 6'
Summer 1 7
Comanche Peak 2
. 3:
Nine' Mile Pt 1 10 Surry.1 8
Cook 1 11
' Nine Mile Pt 2 8-Surry 2:
8 Cook 2 19 North Anna 1-9 Susquehanna 1
-11 Cooper Station:
~7 North' Anna 2 9-
>Susquehannai2 11 l
.- Crystal River 3.
8 Oconoa 1
.7 Three Mlle Island 1 8
Davis-Besse 1 8
Ocora; 2 7
Turkey Pt 3 7
Diablo Canyon.1 6
Oconee 3 7.
Turkey Pt 4 12 6
Diablo Canyon 2 6
Oyster Creek 1 8
' Vermont Yankee 1
.Dresden 2 8
Palisades
.11
_ Vogtle 2 L 9
9 Vogtle 1-9 Dresden 3 8
. Palo Verde 1 Duane Arnold 7.-
Palo Verde 2
' 10 :.
Washington Nuclear 2 8
Farley 1
.8 Palo ~ Verde ~ 3'
.1 1 -
Waterford13 _
8 Farley 2 -
8 Peach Bottom 2 10 ~
Wolf. Creek 1 8
Fermi 2 9
Peach Bottom 3 11
' Zion 1 8
Fitzpatrick ~
- 5 Perry.1 11 Zion 2 8
Ft Calhoun 1
-5 l
Table 5.2 '
I j
Other MPA issues implementation Status at Licensed Plants l
i 8000 7000 l
l 6000
$5000 Y
3 4000 7517 h3000 1
6576 l
2000 l
0
[
t Applicable implemated Unimplemented Figure 5.1 i
l Summary of Eleven Unimplemented MPAs i
120--
109 109 108 109 109 109 109 109 108 108 107 100--
93 80--
79 64 2
60--
i 5
,)
40--
34 33 20--
1 0
l l
l i
l l
BL-88-11 BL-92-01 BL-93-02 BL-93-03 GL-88-20 GL-89-10 GL-92-01 GL-92-08 GL-93-04 MPA-B118 MPA-B122 (X811)
(X201)
(X302)
(B121)
(B111)
(B110)
(B120)
(L208)
(L304)
(B118)
(8122)
SIMS Item Number i
I Legend E Applicable Plants
[I mplemented Plants i
E Unimplemented Plants -
Figure 5.2
e 4
compliance with the applicable codes, licensees were required to obtain plant specific data on thermal stratification, striping and line deflection.
Most PWR licensees coordinated their efforts through their respective owners groups. For Westinghouse plants, the issue is closed for all except six units.
These six units require minor modifications that will ensure surge line stresses remain acceptable for the design life of the plant.
The staff resolved open items regarding the CE Owners Group analysis and issued a Safety Evaluation in June 1993. The staff also resolved open items regarding the B&W Owners Group analysis and issued a Safety Evaluation in September 1993.
CE and B&W licensees must confirm the applicability of the respective owners group analysis to their plants.
BL-92-01 Failure of Thermo-Lao 330 Fire Barrier System (X201)
On June 24,1992, the NRC issued BL 92-01, " Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage."
TU Electric and the NRC sponsored additional testing of Thermo-Lag 330 materials.
As a result of failures in these tests, BL 92-01, Supplement 1, was issued on August 28,1992, to extend the scope of the original bulletin by requesting licensees to (1) identify the areas of the plant that use this material for protection and separation of safe shutdown capability, (2) implemer. spropriate compensatory measures for an inoperable barrier, and (3) verify that the requested actions have been taken and describe the measures being taken to ensure operability.
The responses to BL 92-01 indicate that 83 operating plants have Thermo-Lag fire barrier materialinstalled and appropriate compensatory measures have been implemented. The staff is reviewing responses to Supplement 1 as they are received. An action plan has been developed to address the concerns identified by the special review team.
BL-93-02 Debris Pluaaina of Emeraency Core Coolina Suction Strainers (X302)
This issue arose when a recent experience demonstrated that the capability of the emergency core cooling system (ECCS) to perform its intended function is less than expected. While Perry 1 was shut down in January 1993, the licensee discovered that two ECCS strainers were clogged with particulates and deformed by hydraulic forces. After taking corrective action, and while testing the strainers, the licensee discovered that a strainer was again clogged. Fiberglass from air l
filters in the drywell and iron corrosion products were clogging the strainers. The strainers clogged with fibrous material act as filters; progressively they filter out finer material and develop larger pressure drops than previously anticipated.
Clogging of the strainers can lead to loss of net positive suction head (NPSH),
cavitation of the ECCS pumps, and loss of the ECCS Failure of the ECCS to perform its intended function can cause failure of fuel cladding and the cordainment and a release of radiounuclides to the environment.
On May 11,1993, the NRC issued NRC Bulletin 93-02, " Debris Plugging of Emergency Core Cooling Suction Strainers." The bulletin discussed several instances in which ECCS suction was blocked because fibrous material clogged the strainer. All operating reactor licensees were requested to identify fibrous air 3
fitters or other temporary sources of fibrous material, not designed to withstand a i
i LOCA, that are installed or stored in the primary containment, take prompt action to remove any such material, and take any immediate compensatory measure that may be required to ensure the functional capability of the ECCS. Ucensees were required to provide a written response stating if the actions requested have been or will be performed, the locations and quantity of any identified material, and any immediate compensatory measures taken. Reports on the completion of the requested actions and justification for any deviations from the requested actions also were required.
The responses to BL 93-02 indicate that approximately 75 percent of the licensees do not need, or had already performed, any necessary corrective actions. For the remainder of the licensees whose responses did not provide sufficient information or who took exception to the actions requested in the bulletin, the staff will review further. The issue should be resolved for all facilities within 1 year of the date of the bulletin.
j 1
BL-93-03/GL-92-04 Reactor Vessel Water Level Instrumentation in BWRs (B121/X303)
The staff issued GL 92-04, " Resolution on the issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)" on August 19,1992, to alert licensees of BWRs to errors related to instrumentation accuracy in water level instrumentation and to the results of the staff's review of the j
BWROG's generic analysis of these errors. The staff also requested addressees to 4
(1) determine the impact of these errors on automatic safety system response, operator short-and long-term actions, and emergency operating procedures at i
their facilities; (2) take short-and long-term corrective actions; and (3) submit a report that includes the results of their determinations, a discussion of their short-4 4
and long-term actions, and the schedule for completion of their long-term programs.
All addressees responded by September 28,1992. Most licensees requested deferral of the long-term corrective actions to allow BWROG to complete testing and ana!ysis of the BWR water level instrumentation. The staff accepted delays in the implementation of long-term corrective actions pending BWROG development of plant and/or procedure modifications by July 1993. -
I Following an event at the WNP-2 plant in January 1993, additional analyses by the l
l BWROG revealed additional safety concerns related to RPV water level instrumentation at low pressures following normal depressurizations. This led the staff to issue NRC Bulletin 93-03 on May 28,1993, requesting additional actions by the addressees. These actions were to (1) provide additional procedures and training to address the new concerns and (2) implement, at the first cold shutdown after July 30,1993, hardware modifications to ensure high functional reliability of the RPV water level instrumentation for long-term operation. Addressees have provided their responses, and the NRC is reassessing implementation schedules based on preventing undue hardship or adverse consequences.
Temporary Instruction 2515/119 was issued on March 31,1993, to verify licensee implementation of operator guidance and training to ensure required operator actions concerning reactor water level following rapid depressurization transients.
GL-88-20 Individual Platl Examination for Severe Accident Vulnerability 10 CFR 50.54(Fi '8111)
The NRC issued GL 88-20, " Individual Plant Examination for Severe Accident Vulnerability," on November 23,1988 o request addressees to perform an individual plant examination (IPE) of theli plant-specific internal event severe accidents and report the results of their analysis.
The NRC issued Supplement 1 to GL 88-20. " Initiation of the Individual Plant Examination for Severe Accident Vulnerabilit.es - 10 CFR 50.54," on August 29, 1989 requiring licensees to submit an IPE to identify plant-specific severe accident vulnerabilities using probabilistic risk analysis methodology.
The IPE effort is more complex than estimated. Ucensees have delayed submittal of several IPEs from 2 to 18 months. The staff has issued seven evaluation reports documenting the results of the Step 1 review for Seabrook, Turkey Point 3/4, Oconeo 12/3, Surry 1/2, Beaver Valley 2, Diablo Canyon 1/2, and Millstone 3 IPEs. The following three plants have been selected to date for Step 2 IPE reviews: Turkey Point 3 and 4 and Fitzpatrick.
With the exception of Browns Ferry 1 and 3 (which are shutdown), all IPE reports are scheduled for submittal by June 1994, the scheduled date for Braidwood 1 and
- 2. Staff closure is estimated to occur by the end of 1995.
j GL-89-10 Safety Related Motor-Ocerated Valve Testino and Surveillance (8110)
NRC staff issued GL 89-10 to inform lice:1 sees of problems concerning the operability of safety-related motor-operated valves, and request addressees to (1) establish programs to demonstrate the operability of these valves and to ensure continued operability over the life of the plant, (2) provide a commitment to establish such a program and complete the demonstration of operability within the I.
l timeframe specified in GL 89-10, and (3) report completion of the demonstration phase of their programs. The subject matter of this generic letter is related to that of BL 85-03, " Motor-Operated Valve Common Mode Failures During Plant Transient Due to improper Switch Setting," and its supplements.
Supplements 1 through 4 of GL 89-10 were addressed in Supplement _2 of NUREG 1435. Since then, Supplement 5, " inaccuracy of Motor-Operated Valve Diagnostic Equipment," was issued on June 28,1993, to request licensees to reexamine their MOV programs in light of new information on MOV diagnostic equipment inaccuracies and to identify measures taken or planned to account for uncertainties in valve thrust. Ucensees are also to determine the schedule necessary to satisfy this supplement.
l GL-92-01 Reactor Vessel Structural Intearity (B120)
NRC issued GL 92-01, " Reactor Vessel Structural Integrity," on February 28,1992.
Revision 1 was issued on March 6,1992. The background section concerning NRC's assessment of embrittlement in the Yankee Rowe reactor vessel was updated by Revision 1 to better reflect the licensee's extensive technical efforts regarding reactor vessel integrity. The information was requested within 120 days from issuance of GL 92-01, Revision 1. All licensees have responded.
GL 92-01 is part of the staff's continuing program to evaluate reactor vessel integrity. The information provided will be issued to confirm that all licensees are complying with the requirements of 10 CFR 50.60 and 50.61 and Appendices G and H to 10 CFR Part 50 and are fulfilling the requirements of GL 88-11. "NRC Position On Radiation Embrittlement of Reactor Vessel Materials and its impact On Plant Operations."
A status paper (SECY-93-048) dated February 25,1993, provided the results of the staff's initial screening of GL 92-01. Many requests for additional information have been issued to resolve discrepancies or inconsistencies in the licensees responses.
All issues related to the methodology used to determine compliance with Appendix.
G to 10 CFR Part 50 will be resolved'or the review and approval of equivalent 4
margin analysis, as requested by licensees, will be performed. The staff also is reviewing the current values of reactor vessel material brittle-to-ductile transition temperature because of the information provided in response to the generic letter.
An expanded reactor vessel materials data base is being developed on the basis of the detailed information provided in responses to the generic letter.
GL-92-08 Thermo-Laa 330-1 Fire Barriers (L208) -
t On August 6,1991, the NRC issued IN 91-47, " Failure of Thermo-Lag Fire Barrier Material to Pass Fire Endurance Test," which contained information on the fire endurance test per,'ormed by the Gulf State Utilities Company on Thermo-Lag 330 fire barrier system installed on wide aluminum cable trays and the associated 1
failures. On December 6,1991, the NRC issued IN 91-79, " Deficiencies in The Procedures for installing Thermo-Lag Fire Barrier Materials," which contained information on deficiencies in procedures that the vendor (Thermal Science, Inc.)
supplied for installing Thermo-Lag Fire Barrier Material. Recognizing the concerns stated in ins 91-47 and 91-79 regarding the Thermo-Lag 330 fire barrier system, Texas Utilities (TU) Electric instituted a full-scale fire endurance testing program to qualify its Thermo-Lag 330 electrical raceway fire barrier systems for its Comanche Peak Steam Electric Station. The results of these tests have raised questions regarding the ability of the Thermo tag 330 fire barrier system to perform its specified function as a 1-hour fire barrier.
On June 23,1992, the NRC issued IN 92-46, "Thermo-Lag Fire Barrier Material Special Review Team Final Report Findings, Current Fire Endurance Testing, and l
Ampacity Calculation Errors," in which it discussed the safety implications of these questions. On June 24,1992, the NRC issued BL 92-01, " Failure of Thermo-Lag 330 Fire Barrier Systgm To Maintain Cabling in Wide Cable Trays and Small Conduits Free From" Fire Damage." TU Electric and the NRC sponsored additional ~
testing of Thermo-Lag 330 materials. As a result of failures in these tests, BL 92-01, Supplement 1, was issued on August 28,1992, to extend the scope of the original bulletin.
Following this ac' tion, on December 17,1992, the NRC issued GL 92-08, which -
l required the licensees to confirm (1) that the Thermo-Lag 330-1 barrier systems have been qualified by representative fire endurance tests, (2) that the ampacity derating factors have been derived by valid tests, and (3) that these qualified barriers have been installed with appropriate procedures and quality controls to ensure that they comply with the NRC's requirements.
in response to GL 92-08, most licensees have indicated that they await the results of the NUMARC tests of Thermo-Lag 330 fire barrier material. The NUMARC tests l
are scheduled to be completed in January 1994, at which time, licensees will develop their plan of action for the resolution of the Thermo-Lag 330 issue.
GL-93-04 Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies.10 CFR 50.54(f) (L304)
The NRC issued GL 93-04 on June 21,1993, (1) to notify addressees about a I
single failure vulnerability within the Westinghouse solid-state rod control system that could cause an inadvertent withdrawal of control rods in a sequence resulting in a power distribution not considered in the design basis analyses and (2) to require that all action addressees provide the NRC with information describing their plant specific findings related to this issue and actions taken. The GL was addressed to all holders of operating licenses or construction permits for Westinghouse-designed nuclear power reactors except Haddam Neck.
l l
The GL requested licensees to assess within 45 days if their licensing basis is still satisfied with regard to single failure in the rod control system in light of the Salem event. If the licensing basis is not satisfied, the NRC requested licensees to provide an assessment of the impact and describe any compensatory short term actions taken within 45 days and provide a plan and schedule for long-term resolution within 90 days. On July 26,1993, the NRC granted relief to the schedules to extend the licensing basis assessment portion of the 45-day response 4
to the 90-day response, in response to a request from the Westinghouse Owners a
Group.
All licensees have provided their 45-day response and these are under NRC staff review. The 90-day responses were due by September 19,1993 and the staff is reviewing these responses.
i MPA-B118 iPE Extemal Events (GL-88-20. Suco 4) (B118)
The staff issued Supplement 4 to GL 88-20 on June 28,1991 to initiate the IPE process for external events. Five categories of extemal events were specified and licensees were required to submit to a schedule and methodology by December 26,1991. Ucensees were requested to submit the results of the individual plant examination of external events (IPEEE) within 3 years of the issuance date of Suppiement t er no later than June 28,1994. A' copy of NUREG-1407, " Procedural and Submittal Guidance for the IPEEE for Severe Accident Vulnerabilities," was sent to each licensee with Supplement 4. All licensee responses to Supplement 4 with respect to schedules md methodology, have been received and reviewed independently and jointiy by NRR and RES.
Supplement 1 to GL 87-02 was issued on May 22,1992, approving the seismic qualification utility group generic ;mplementational procedure for USI A-46 implementation and starting the clock for both A-46 and the IPEEE. Following review of the licensees' responses to Supplement 4 to GL 88-20 (IPEEE), the NRC staff met to develop guidelines that would be used in determining whether a licensce's response would be considered acceptable. The NRC staff reported the results to the commission (SECY-92-130), and estimated a delay of approximately 1 year may be warranted. The guidelines reported to the Commission were that the IPEEE resutts must be submitted to the staff by June 1995 or within 3 years after issuance of the staff's evaluation approving the A-46 GlP, whichever occurs first. Therefore, since the evaluation was issued on May 22,1992 with GL 87-02, Supplement 1, licensees were advised that the latest acceptable date for IPEEE submittal would be June 1995. A second round of licensee responses indicated that by the best effort by the industry will have the IPEEEs for 72 plants submitted by the target date of June 1995 but for the remaining 38 plants, the submittal dates will range from September 1995 to July 1997.
There are a small number of plants that have unique problems requiring a more customized response (1) because the licensee proposed alternative methods or
failed to provide any method at all for its IPEEE or (2) because the licensee's plant was one of the alght singled out by the Eastern United States Seismic Hazards Program as needing further NRC staff evaluation.
MPA-B122. Loss of Fill-Oilin Transmitters Manufactured by Rosemount (BL-90-01) (B122)
On April 21,1989, the staff issued Information Notice 89-42, " Failure of Rosemount Models 1153 and 1154 Transmitters," to alert the industry of the loss of oil-fill problem. On March 9,1990, the staff issued Bulletin 90-01, " Loss of Fill-Oil in Transmitters Manufactured by Rosemount," to request the licensees to promptly identify and to take appropriate corrective action for Model 1153, Series B, Model 1153, Series D, and Model 1154 transmitters that may have the potential for leaking fill-cil. From mid 1990 through 1992, the staff reviewed information from the (1) licensee responses to Bulletin 90-01, (2) data from related licensee event reports, (3) visits to the sites, (4) NUMARC Report 91-02, " Summary Report of NUMAC Activities to Address Oil Loss in Rosemount Transmitters," and (5) meetings with the industry. The staff found a relationship between operating pressure and time-in-service that can be trended for use in identifying transmitters that are most likely to fail. The staff has concluded that (1) the requested actions in Bulletin 90-01 were insufficient in that they did not provide the desired high functional reliability and (2) a supplemental bulletin would be needed for ensuring appropriate licensee corrective action to the loss of fill-oil problem.
Subsequently, on December 22,1992, the staff issued Bulletin 90-01, Supplement 1, " Loss of Fill-Oil in Transmitters Manufactured by Rosemount," to request new action from the licensees. Specifically, licensees were to provide information on specified models of the Rosemount transmitters manufactured before July 11, 1989, that are in use or may be used in the future. The information shall detail the use of the devices in either a safety-related system or a system governed by the NRC's ATWS (anticipated transient without scram) requirements where normal operating pressure is greater than 500 pounds per square inch. Requested corrective action includes the replacement of the suspect transmitter or the use of an enhanced surveillance monitoring program until the transmitter reaches the time-in-service pressure criterion recommended by the vendor.
Responses to the supplemental bulletin have been received from all applicable licensees and currently are under the staff review.
A brief description of other multiplant actions with more than three open items follows:
BL-88-08 Thermal Stress in Pioina Connected to RCS (X808)
Following a circumferential crack in an unisolable section of emergency core cooling piping at Farley 2, the NRC issued BL 88-08, " Thermal Stresses in Piping Connected to Reactor Coolant Systems," dated June 22,1988. The Bulletin requested all licensees and applicants to take the following three actions: (1) review their reactor coolant systems (RCSs) to identify any connected unisolable piping that could be subjected to temperature distributions that could resuit in unacceptable thermal stresses, (2) examine unisolable piping sections for existing flaws, and (3) implement a program to provide continuing assurance that unisolable sections will not be subject to stresses that could cause fatigue failure.
In summary, BL 88-08 was closed for those BWRs and PWRs whose responses to action item 3 above were consistent with the stated modification or monitodng alternatives. However, some plants replied that assurance for certain lines would be provided by inspection alone, when conducted as part of their inservice inspection program. The licensee responses for these plants were unacceptable without further justification, because inservice inspection was not identified by BL 88-08 as an acceptable alternative. The basis for this position is that the fundamental precept of the actions of BL 88-08 is to prevent the initiation of cracks in piping. Inservice inspection is not a technique that prevents the initiation of cracks. Rather, inservice inspection identifies cracks after they appear, and then a safety significance determination is made and corrective action is proposed. The staff is reviewing the supplemental responses of licensees whose initial submittals i
contained insufficient information.
4 GL-84-09 Recombiner Caoability Reauirements of 10 CFR 50.44 (C)(3)(ll)
)
(A019)
As a result of the TMI-2 accident, it became clear that the amount of hydrogen produced from the metal-water reaction was far in excess of that previously considered by the NRC staff during the licensing process. As a result, the staff revised 10 CFR 50.44, " Standards for Combustible Gas Control Systems," effective l
January 4,1982 (48 FR 58484) to address this safety concern. For plants with Mark I and Mark 11 type containments, the staff determined that containment inerting (with nitrogen) and recombiner capability were sufficient measures to accommodate hydrogen from a 75-percent metal-water reaction without resulting in a burnable mixture. Certain licensees with Mark I containment took exception to the staff's position of providing recombiner capability because they believed the assumptions in NEDO 22155 were questionable. Therefore, using the models in NEDO-22155, they calculated that a typical Mark I design equipped with containment inerting was sufficient to preclude a burnable mixture resulting from a 75 percent metal-water reaction for the 30 days following an accident, both within the design-basis-accident (DBA) envelop and slightly beyond. The NRC staff concluded that, on balance, costs outweighed the benefits to address this limited situation. To reflect this position, the NRC issued GL 84-09, dated May 8,1984.
- 1
GL 84-09 allowed licensees with Mark I type containments that rely on purge /repressurization systems as a means of hydrogen control, an option in lieu of installing recombiner capability if they met the following conditions: (1) the plant has technical specifications (limiting conditions for operation) requiring that the j
containment is less than 4-percent oxygen while inerted, (2) the plant has only nitrogen or recycled containment atmosphere for use in all pneumatic control systems within containment, and (3) there are no significant sources of oxygen in containment other than that resulting from radiolysis of the reactor coolant.
GL-87-09 Sections 3.0 and 4.0 of the Standard Technical Soecifications on the Acolicability of Umitino Condit ons for Ooeration and Surveillance Reouirements (D024)
The NRC issued GL 87-09 on May 4,1987, to provide guidance for three specific problems that had been encountered with the general requirements on the applicability of limiting conditions for operation (LCOs) and surveillance 2
requirements in Section 3.0 and 4.0 of the Standard Technical Specifications. The problems involve (1) unnecessary restrictions on mode changes and inconsistent application of exceptions, (2) unnecessary shutdowns when surveillance intervals are inadvertently exceeded, and (3) possible conflicts between Specification 4.0.3 and 4.0.4. Staff guidance addressed these problems. Implementation of the guidance contained in GL 87-09 is voluntary.
GL-88-01 NRC Position on IGSCC in BWR Austenitic Stainless Steel Pioina (8097)
The NRC issued GL 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," on January 25,1988, to seek information from BWR licensees and construction permit holders regarding implementation of new staff positions regarding intergranular stress corrosion cracking (IGSCC). Addressees were asked to respond within 120 days of receipt of GL 88-01. The response was to indicate whether the utility intended to follow the staff positions included in the letter or propose alternative measures. An acceptable response from licensees also i
included a commitment to revise technical specifications (TS) to be consistent with the NRC staff positions in GL 88-01.
1 GL 88-01, Supplement 1, was issued on February 4,1992. The supplement provided clarification, guidance, and acceptable alternative staff positions to the positions delineated in GL 88-01. The supplement did not require a response.
GL-88-11 NRC Position on Radiation Embrittlement of Reactor Vessel Materials (A023)
Revision 2 to RG 1.99, " Radiation Embrittlement of Reactor Vestel Materials,"
became effective in May 1988. GL 88-11 was issued on July 12,1988, and 1
indicated that RG 1.99 (Rev. 2) would be used by the staff for evaluating all submittals regarding pressure temperature limits and for all analyses that require an estimate of vessel beltline embrittlement (except those for pressurized thermal shock).
i GL 88-11 requested that all licensees of operating reactors use the methods described in Revision 2 to RG 1.99 to predict the effect of neutron radiation on reactor vessel materials as required by Appendix G to 10 CFR Part 50, unless they could justify the use of different methods. The licensees were required to submit the results of their analyses and an implementation plan for proposed actions.
Acceptable responses have been reaived from all licensees and all technical i
i reviews have been completed.
GL-89-01 Imolementation of Prooram Controls for RETS in Administration Control Section (D025)
The NRC issued GL 89-01 on January 31,1989 to provide guidance for the implementation of programmatic controls for radiological effluent technical specifications (RETS) in the administrative controls section of TS and the relocation 4
of procedural details of current RETS to the offsite dose calculation manual (ODCM) or process control program (PCP). It is not the staff's intent to reduce the level of radiological effluent control. Rather, this proposed TS change will provide programmatic controls for RETS consistent with regulatory requirements and allow relocation of the procedural details of current RETS to the ODCM or PCP.
Implementation of the guidance contained in GL 88-11 is voluntary.
GL-89-04 IST Reviews and Schedules (A025)
The staff noted that certain generic weaknesses were being observed in licensee inservice testing (IST) programs. The NRC issued GL 89-04, " Guidance on Developing Acceptable Inservice Testing Programs," on April 3,1989. With the exception of certain plants noted in the generic letter, licensees of the remaining plants were required to (1) review their most recently submitted IST program and procedures against the positions of GL 89-04 and (2) confirm in writing within 6 months their conformance with the staff positions. These licenses also were required to submit a schedule for equipment and program modifications required as a result of the review. GL 89-04 granted approval for licensees to change their i
IST program without specific prior approval for changes that conformed to the staff positions.
In response to GL 89-04, many facilities revised their programs to reflect the relief granted by the staff positions in GL 89-04. In many cases, the facilities submitted revised programs, including additional relief requests that were outside the scope of the generic letter. The staff has completed individual evaluations for such plants.
These facilities must confirm correction of any program anomalies identified in the staff's safety evaluation.
Temporary Instruction 2515/114 was issued on January 15,1992, to provide uniform guidance for inspecting the activities of nuclear power plant licensees regarding inservice testing of pumps and valves.
l GL-89-06 Task Action Plan item 1.D.2 - SPDS (F072)
The NRC issued NUREG-0737 on October 31,1980 to provide guidance for implementing TMl action items. On December 17,1982, GL 82-33 transmitted Supplement 1 to NUREG-0737 to clarify the TMI action items related to emergency response capability, including item 1.D.2, safety parameter display system (SPDS).
The staff evaluated licensee / applicant implementation of the SPDS requirements at 57 units and found that a large percentage of designs did not fulfill the requirements identified in Supplement 1 to NUREG-0737.
The NRC staff issued GL 89-06 on April 12,1989 to provide information to licensees regarding the implementation status of the SPDS at their facilities.
NUREG-1342 was enclosed with GL 89-06 to aid in implementing the SPDS requirements. Licensees were required to furnish one of the following: (1) certification that the SPDS fully meets the requirements of NUREG-0737, Supplement 1, taking into account the information provided in NUREG-1342; (2) certification that the SPDS will be modified to fully meet the requirements of NUREG-0737, Supplement 1, taking into account the information provided in NUREG-1342; or (3) if a certification cannot be provided, the licensee must provide a discussion of the reasons for that finding and a discussion of the compensatory action the licensee intends to take or has taken.
GL-89-16 Installation of Hardened Wetwell Vent (B112)
The Mark I containment performance improvement (CPI) program identified a number of plant modifications that substantially enhance a plant's capability to both prevent and mitigate the consequences of severe accidents. The improvements that were recommended to the Commission included (1) improved hardened wetwell vent capabirity, (2) improved reactor pressure vessel depressurization system reliability, (3) an alternative water supply to the reactor vessel and drywell sprays, and (4) updated emergency procedures and training.
In a staff requirements memorandum (SRM) of July 11,1989, the Commission directed the staff to (1) proceed with a generic implementation of installation of hardened wetwell vents at all Mark I containment plants; (2) forward the remaining CPI improvement requirements to the licensees of the Mark I containment plants for incorporation into their individual plant examination (IPE) programs; and (3) expedite the staff actions to implement the station blackout rule at the Mark I containment plants. The staff issued GL 89-16 to address generic implementation of hardened wetwell vent installation.
The licensees responded to GL 89-16 through the Boiling Water Reactor Owners Group (BWROG). The staff has completed the evaluation of the licensees' actions implementing the hardened vent capability at all 24 Mark I plants and has either approved the modification schedules or accepted the existing wetwell venting capability. The' staff is currently preparing a Tl for verification of hardened vent installation.
GL-91-11 Vital Instrument Buses and Tie Breakers (GI-48 & 49) (Liii) l GL 91-11 required all licensees to certify that plant procedures included time limitations and surveillance requirements for vital instrument buses, inverters or other onsite power sources to the vital instrument buses, and tie breakers that can connect redundant Class 1E buses between units at the same site. ff plant procedures did not include time limitations and surveillance requirements as reques'.ed, a documented evaluation was needed to justify why such provisions were not needed.
MPA-B116 Results of NRC Testina of MOVs (GL-89-10. Suco 3) (B116) l On June 5,1990, the staff issued IN 90-40, "Results of NRC-Sponsored Testing of 1
i Motor-Operated Valves (MOVs)." The tests revealed that the valves required more thrust for opening and closing under various differential pressure and flow conditions than would have been predicted from standard industry calculations using typical friction factors. Therefore, the staff issued Supplement 3 to GL 89-10 on October 25,1990, o aich described required actions for licensees of BWRs.
Licensees were required to provide (1) criteria reflecting operating experience and the latest test data that were applied in determining whether the deficiencies exist in the subject MOVs, (2) a list of the MOVs found to have deficiencies, and (3) a i
schedule for the necessary corrective action.
MPA-B117 Failure of Westinchouse SG Tube Mechanical Pluas (8117) i Bulletin 89-01, Supplement 2, requested that actions similar to those requested in NRC Bulletin 89-01, " Failure of Westinghouse Steam Generator Tube Mechanical Plugs," be extended to include all Westinghouse mechanical plugs fabricated from thermally treated inconel 600.
Bulletin 89 01, requested that licensees determine whether certain mechanical i
plugs supplied by Westinghouse were installed in the steam generator (SG) and, if so, that an action plan (including plug repair and/or replacement) be implemented to ensure that the plugs would continue to provide adequate assurance of reactor coolant pressure boundary (RCPB). The request applied to only four plugs fabricated from inconel heats and referred to as group 1 heats. These plugs were highly susceptible to primary water stress, corrosion cracking (PWSCC).
After issuance of Bulletin 89-01, Westinghouse compiled a complete listing of all inconel 600 plug lifetime categorized by plant, date of installation, and heat number. All plugs, not included in group 1 heats were specified as group 2 heats.
During the summer and autumn of 1990, two plants experienced PWSCC affecting group 2 heats. These events were described in Bulletin 89-01, Supplement 1.
Subsequently, Westinghouse revised its algorithm for estimating plug lifetimes PWSCC rather than temperature relationship, on the basis of operating experience trends.
Cumulated field experience prompted issuance of Bulletin 89-01, Supplement 2 on June 28,1991, which requested that actions similar to those requested in BL 89-01, be extended to include all Westinghouse mechanical plugs group 2 heats fabricated from thermally treated inconel 600. These actions were measured to ensure that the mechanical plugs would continue to provide adequate assurance of RCPB integrity under normal operating, transient, and postulated conditions.
5.2 Verification Status For generic items such as MPAs, NRR issues Tis for those items that need to be verified in the field by the NRC staff after licensees have implemented the corrective actions specified in the MPA resolution. The NRC performs these inspections, consistent with other inspection priorities, to verify proper implementation of the j
requirements. Verification is not considered complete until the required inspection is conducted in accordance with the TI, and an inspection report has been issued -
documenting that requirements have been adequately satisfied by the licensee. On occasion, there may be issues for which the requirements specified in the Tl for safety verification inspection are completed before totalimplementation of all aspects of the issue's resolution by the licensee.
Tis provide guidance for the field verification of licensee implementation of other MPAs.
The NRC issued 15 Tis for 15 individual MPA issues, which cover a total of 879 items at the 109 licensed plants. Upon initialinspection of certain items and further review by the regional offices,103 items covered by the Tis were found to be inapplicable from a verification standpoint, leaving a total of 776 items requiring verification. The majority of items found not applicable are cases in which initial inspection did not reveal any significant findings and for which further inspection effort cannot be justified.
As of September 30,1993,589 items (76 percent) had been verified. Tl designations and the corresponding MPAs are summarized in Table 5.3. Table 5.4 summarizes the l
items remaining to be verified.
90
)
Temporary Instructions for Resolved MPAs 1
i SIMS Item MPA SIMS Title Tl Number i
BL-79-15 B031 Deep Draft Pump Deficiencies 2500/001 BL-80-11 B059 Masonry Wall Design 2515/037 BL-88-04 X804 SI Pump Failure (Bulletin 88-04) (Old MPA B103) 2515/105 BL-88-07 X807 Power Oscillations in Boiling Water Reactors (BWRs) 2515/099 GL-80-002 A015 Ouality Assurance Requirements Regarding Diesel 2515/093 Generato' MI Oil GL-81-21 B066 Natural Clrculation Cooldown 2515/086
]
GL-83-08 D021 Modification of Vacuum Breakers on 2515/096 Mark l Containments GL-89-04 A025 Guidance on Accepting inservice Testing Programs.
2515/114 i
GL-89-07 L907 Power Reactor Safeguards Contingency Planning for : 2515/102 1
Surface Vehicle Bombs GL-89-10 B110 Safety Related Motor Operated Valve Testing and 2515/109 Surveillance GL-92-04 X303 Reactor Vessel Water Level Instrumentation in BWRs 2515/119 MPA-B003 B003 PWR Moderator Dilution 2515/094-MPA-B011 B011 Flood of Equipment important to Safety 2515/088 MPA-B041 B041 Fire Protection - Final Technical. Specification 2515/062 (including SER Supplements)
MPA-C002 C002 BWR Recirculation Pump Trip (ATWS) 2515/095 Table 5.3 _.
i.
Summ:ry of Oth;r MPA Itama Requiring Vcrific tion i.
Plants Plants Plants SIMS Item Covered fl.stnuired Verified BL-88-07 POWER OSCILLATIONS IN BIOfLING WATER.
37 37 36 REACTORS (BWRS) i l
GL-81-21 '
NATURAL CIRCULATION COOLDOWN 72 ~
67 60 l
GL-83-08 MODIFICATION OF VACUUM BREAKERS ON.
23-23 21 MARK I CONTAINMENTS GL-89-04 GUIDANCE ON ACCEPTABLE INSERVICE
~ 40 -
38 3'
TESTING PROGRAMS GL-92-04 REACTOR VESSEL WATER LEVEL 37.
37 0
l-INSTRUMENTATION IN BWRs GL49-10 SAFE 1Y-RELATED MOTOR-OPERATED 109 109 9-lo.
VALVE TESTING AND SURVEILLANCE
+
Y MPA-B011 FLOOD OF EQUIPMENT INPORTANT TO
'9
.3 1
SAFETY MPA-B041 FIRE PROTECTION - FINALTECH SPECS 65 62 61
. (INCLUDING SER SUPPLEMENTS).-
,MPA-C002 BWR-RECIRC.PUhr TRIP (ATWS)
' 21' 21-
.19 '
- NOTE Plants Coveed we those for wNch MPAs are appbcable
- Plants Required are those plants requinng field venhcahon
. Plants covered but for which Reid venNcaten is not necessary have implemented the resoluten in a rnenner not requenng plant hardware changes.
t Table 5.4 t
l'
..u.
- m..
.__.._,.;_-..,,m._,
e..
5.3 Status by Plant Table 5.5 summarizes information on the status of implementation and verification of I
MPAs at all licensed plants. For each plant, the table shows the total number of applicable items, the number and percentage of items implemented, and the number of items remaining to be implemented. For those MPAs that require the NRC to verify implementation actions, the table shows the number of items covered by a Tl at each plant, the number of items requiring verification, and the number and percentage of items completed. Appendix D lists the unimplemented MPA items by plant and gives projected implementation dates.
Of the 109 plants, none have completely implemented all MPA items. On average, each plant less than 9 remaining items to implement. No plant has more than 12 remaining items, with the exception of Browns Ferry 1 and 3, which have 20 each.
i r
4
)
I 3
i l
i a
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPAs -
SUMMARY
BY PLANT IMPLEMENTATION VERIFICATI0sl ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVE R E D REQUIRED COMPLETED COMPLETED ARMANSAS 1 102 94 192 )
8 10 8
7 (87 )
ARKANSAS 2 86 78 190 )
8 9
7 6
i85 )
BEAVER VALLEY 1 96 86 (89 l 10 9
8 7
l 87 )
BEAVER VALLEY 2 45 35 (77 )
10 7
6 4
1 66 )
BIG RDCM POINT 1 70 64 (91 )
6 10 9
6 (66 )
BRAIDWOOD 1 43 32 (74 )
11 6
5 4
(79 )
BRAIDWOOD 2 41 29 (70 l 12 5
4 3
(75 )
BROWNS FERRY 1
($
65 (76 )
20 11 10 5
(50 B20WNS FERRY 2 80 70 (87 )
10 10 10 8
(79 BROWNS FERRY 3 82 62 (75 1 20 10 9
5 (66 BRUNSWICK 1 82 73 (89 )
9 9
9 6
(66 l'
BRUNSWICK 2 82 73 (89 l 9
10 9
7 (77 i
BYRON 1 43 32 (74 l 11 5
4 3
(75 i
BYROM 2 41 30 (73 )
11 5
4 3
(75 i
CALLAWAY 1 48 36 (78 )
10 5
4 3
(75 CALVERT CLIFFS 1 92 87 (94 )
5 9
8 7
(37 CALVERT CLIFFS 2 88 82 (93 )
6 9
8 7
(87 I
CATAWBA 1 44 37 (84 )
7 5
4 2
(50 CATAWBA 2 43 36 (83 )
7 5
4 2
(50
,m CLINTON 1 41 33 (80 )
8 7
6 4
(66 Ji COMANCHE PEAK 1 78 70 (89 )
8 9
5 4
(79 COMANCHE PEAK 2 73 70
-(95 )
3 9
3 2
(86 COOK 1 91 80 (87 )
11 8
7 8
(85 i
COOK 2 90 81 (89 )
9 8
7 a
(85 )
C00'ER STATION 82' 75 (91 )
?-
11 9-7 (77 )
CRYSTAL RIVER 3 91 83 (91 )
8' 9
9 7
(77 1
DAVIS-BESSE 1 88 30 (90 )
8 9
6 5
(83 1
DIABLO CANYO4 1 52
- 46
-l88 1
6 6
6 6
(100 1
DIABLO CANYDN 2 47 41 (87 l
6 5
5 5
(100 DRESDEN 2 83 75
.(90 I
.8 12 12 9
(75 I
DRESDEN 3 82 74 (90 1 8
11 11
-3 (72 )
DUANE ARNOLD 87 80 (91 )
7 12 11 8
(72 )
FARLEY 1 90 82 191 )
8 8
8 7'
(87'l FERMI 2 '
56 48
-(85 )
8.
7 7
6' (85 FARLEY 2 42 33
-(78 3 9
7 6
4 (66 FITZPATRICK 83 78:
(93 1 5-12
' 11
'O (72 FORT CALHOUN 1'
'105 100
-(95 )
5 10 S-7 (87 )
' G19H0A 93 81 (87 )
12 10 to 8
(79.)
GRAND GULF 1 45 37 I82:)
8 7
6 4
(66 )-
-HADDAM NECK 92 86 (93 )
6 9
9 7
(77 l
-HARRIS 1 42 33 (78 1 9
5 4
2 (50 l
HATCH 1 82 70
( 85 ')
12 10 9
7 (77 l
HATCH 2 77 65 (84 l-12 11 10 8
(79 )
NOPE CREEK 1 41 34 (82 1 7
6 5
3 (59 )
Table 5.5
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHR MPAs - StWMARY BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT I TEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED INDIAN POINT 2 95 87 (91 )
8 10 9
7 (77 I INDIAN P0lMT 3 91 85 (93 1 6
9 8
8 (75 )
KEWAUNCE 95 87 (91 )
8 9
8 7
(87 )
LASALLE 1 40 29 (FJ B 11 7
6 4
(66 - )
LASALLE 2 40 28 (69 l 12 7
6 4
(66 LIMERICK 1 40 32 i79 )
8 6
5 3
(59 I
L IME R ICK 2 38 29 1 76 I 9
6 5
3 (59 MAINE YANKEE 96 59 i 92 1 T
11 10 8
(79 HCGUIRE 1 53 i 83 )
9 5
4 3
(75 MCGUIRE 2 47 l 80 1 9
5 4
3 (75 MILLSTONE 1 76 I 89 I 8
11 10 7
.(89
)-
MILLSTONE 2 93 a
1 93 3 6
9 8'
7
'q87
).
MILLSTONE 3 42 3 '.
1 83 )
7 5
5 4
(79 )-
MONTICELLO 82 76 (92 ) 11 10 7
(69 l-NINE MILE POINT 1 85 75 (88 I 10 11
-11 9
(81 )
NINE MILE POINT 2 38 30 (78 )
.8 6
5 3
(59 )
NORTH ANNA 1 73 64 (87 l 9
9 9
7 (77 )
NORTH ANNA 2 56 47
-(83 1 9
7 6
4 (66 )
OCONEE 1 93 86
-(92 I 7
10 9
7-(77 )
go OCONEE 2 93 86 192 )
7 10 9
7 (77 )
Ln OCONEE 3 92 85 192 3 7
to 9
7 177-)
OYSTER CREEK 1 79 71 (89')
8 12 12 9
(75 )
PALISADES 95 86 (90')
9 11 8
6 (75 i
~
PALO VERDE 1 40
. 29 172 )
-11 5
5 5
(1003 j
PALO VERDE 2 38 28 (73'l 10 5
5
.5 (100)
-PALO VERDE 3 39.
25 (71 1 11'
-5
'5
_5
-(100)
PEACH BOTTDM 2 81 71 187 )
10 11 10 8
(79 )
PE ACH BOTTOM 3 80 69 (86 1 11 11 10 8
(79 )
PERRY 1 38 27 (71'l 11 7
6-4 (66 )
PILGRIM-1
- 86 79 (91 )
7 11 10 7
(69 )
POINT BEACH 1 92 83 190 )
9 10' 9
7 (77 )
POINT BEACH 2 92 83 (90_l 9
9' 8.:
6 (75 l PRAIRIE ISLAND-1 94 85 190 )
9
'8 7
6 (85 l-OUAD CITIES 1 86 77 (89 )
- 9 8.
7.
6 (85 )
PRAIRIE ISLAND 2 94 85 190 1 9
11 10 7
(69 )
OUAD CITIES 2 85 76 (89 )
9' 11 10
'7 (69 )
RIVER BEND 1 57
.30 (81 1.
7 6
5 3
(59 )
ROBINSON 2 90 82
( 91 ' )
8 9-8 7
(87 I SALEM 1 92 82
-(89 l 10 9
~9 7
( 77. )
9-8_
8 6
(75 1--
SALEM 2 57 48 (84 I
._10 5
5 5
(1001 SAN ONOFRE 2 41 31 (75 1' SAN ONOFRE 3 42 31
-(73 1 11 5
5-5
-(1001-SEABROOK 1 37
.32
-(86')
-5 5
5-
'4 (79 i SEQUOYAH I 53.
46-
- (86 1 7
5
'5 4
(79 )
Table 5.5L y
k
+
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPAs -
SUMMARY
BY PLANT IMPLEMENTAYION VERIFICAT1088 ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVE RED REQUIRED COMPLETED COMPLETED
- StouGYAH 2 45 37 (s2 i e
5 5
4 (79 )
SOUTH TEMAS 1 41.
33 (so-l 8
5 3-2 166 )
SOUTH TEXAS 2 38 31 181 )
7 5
3
'2 166 )
ST LUC 1E 1 95 35 192 )
7 10.
!O.
7 (69 )
ST LUCIE 2 41 34
- (32 )
7 5
5 3
(59 )
SUPMER 1 45 38 184 )
7 4
4
?
150 )
SuRRY I 97 89 191 )
3 10 9
7 (77 )
SURRY 2 100 92 (91 )
8 10 9
7 (77 t
[
- $USQUEHAN004 1 40 29 172 )
11 7
6 3
150 I
SUSQUEHANNA 2 40 29 (72 )
11 7
6 3
1 50 l
_ THREE MILE ISLAle 1 100 92
. (91 l 8_
9 8
7 37 TURMEV POINT 3 99 92 (92 )
7 10 8
8 1
75 f
TURMEY POINT 4 100 94 (93 )
6 10 8
6 (75 )
i VERMONT YANKEE 1 82 70 (85 3
-12 12' 10 7
(69 )
V0GTLE 1 44 35 (79 )
9 5
5 3
(59 )
[
VOGTLE 2
-36 27 175 )
9 5
5
.3 (59 )
WASHINGTON NUCLEAR 2 44 36
- 131 )
a l,
- WATE RF ORD 3 -
~39 31 179 )
8 ~
6 6
5 I33 )
5 3
2 166 t
WOLF CREEK 1 43 35 181 )
8 5
3 2
166 l
m' ZION 1
.99 91 191 )
8 8
6 5.
(83
. (T,)
ZION 2 99 91 191 )-
5 8
6 5
(83 i
TOTALS-/ AVERAGES 7517 6576 941 373 yyg 3,, -
75 l.
l:
~
1'
?
p-r I ~
Table!5.5-4 t
g'T M* e W t + ir*
v*w---
w '++v*--r e
V-v v
w=t-v
-v~c
- i'W-'"%T
+'Y**
w'
--+w="YmwT*-'
+ + ' - --i-- - ----- -
4 i
5.4 Status by issue j
Table 5.6 presents summary information on the status of implementation and verification of each MPA. For each issue, the table shows the number of applicable plants, the number and percentage of plants that have completed implementation, and the number of plants remaining to complete implementation. For those issues i
requiring NRC verification of corrective actions, the table shows the number of plants covered by the issue, the number of plants at which verification is required, and t,6 j
number and percentage of plants that have completed verification.
Of the current 171 MPA issues,127 have been fully implemented, twenty-f /e issues remain to be implemented at 5 or less plants and 8 issues remain to be implemer.ted at 6 to 15 plants. The remaining 11 MPA issues are to be implemented at 26 or more plants.
SAFETY ISSUE MANAGEMEN T SY$ TEM STATUS OF OTHER MPAIS) -
SUMMARY
BY ITEM IMPitMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED LOSS OF OFF-SITE PDWER 75 (8089) 68 68 (100) 0 NO SALEM ATWS 4.2.3 & 4.2.4 LIFE COMPONENTS 8059 (E004) 15 15 (100) 0 NO BWR SINGLE L6C' OPERATION
(
B059 (E005) 8 8
(100) 0 NO W N-1 LOOP OPERATION 0.3 BL-79-06 58 58 (100) 0 NO REVIEW OF OPERATIONAL ERRORS AND SYSTEM MISALIGNMENTS IDENTIFIED DURIN BL-79-06A 5
5 (100) 0 NO l
REVIEW OF OPERATIONAL ERRORS AND SYSTEM MISALIGNMENTS IDENTIFIED DURIN BL-79-08 5
5 (100) 0 NO EVENTS RELEVANT 10 BOILING WATER REACTORS IDENTIFIED DURING THREE MILE BL-79-13 48 48 (100) 0 NO CRACKING IN FEEDWATER SYSTEM PIPING BL-79-15 109 109 (100) 0 YES 109 102 102 (100)
DEEP DRAFT PUMP DEFICIENCIES OSS b NON-CLASS-1-E INS RUMENTATION AND CONTROL S SM BUS DURING P BL-80-04 45 45 (100) 0 NO ANALYSIS OF A PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDIT BL-80-06 82 82 (100) 0 NO I
ENGINEERED SAFETY FEATURE (ESF) RESET CONTROLS BL-80-07 33 33 (100) 0 NO BWR JET PUMP ASSEMBLY FAILURE BL-80-11 84 84 (100) 0 YES 64 63 63 (100J
. MASONRY WALL DESIGN BL-80-18 43 43 (100) 0 NO MAINTENANCE OF ADEQUATE MINIMUM FLOW THRU CENTRIFUGAL CHAEGING PUMPS F Table 5.6
SAFETY ISSUE MANAGEMEN T SYSTEM STATUS OF OTHER MPA(S) -
SUMMARY
BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED
- tMAINING REQUIRED COVERED REQUIRED COMPLF.TED COMPLETED BL-87-01 109 109 (100) 0 NO THINNING OF PIPE WALLS IN NUCLEAR POWER PLANTS BL-88-01 109 109 1100) 6 NO DEFECTS IN WESTINGHOUSE CIRCUIT BREAKERS BL-88-02 33 32 196 1 i
NO STEAM GEERATOR TUBE RUPTURE (BULLETIN 88-02) (OLD MPA B099)
BL-88-03 109 107 (98 )
2 NO GE HFA RELAYS (BULLETIN 88-03)
BL-88-04 106 103 (97 )
3 YES 106 35 35 (100)
SI PUMP FAILURE (BULLETIN 88-04) (OLD MPA B103) i BL-88-05 109 109 (100) 0 NO NONFORMING MATERIALS SUPPLIED BY PIPING SUPPLIES. INC, AT FOLSOM BL-88-07 37 35 194 1 2
YES 37 37 36 (97 )
POWR OSCILLATIONS IN BOILING WATER REACTORS (BWRS)
BL-88-08 109 S5 (87 )
14 NO THERMAL STRESSES IN PIPING CONNECTED TO REACTOR COOLANT SYSTEMS BL-88-09 50 50 (100)
O NO THIMBLE TUBE THINNING IN WESTINGHOUSE REAXCTORS BL-88-10 109 107 (98 )
2 NO NONCONFORMING MOLDED-CASE CIJCUIT BREAKERS BL-88-11 71 45 (63 1 26 NO PRESSURIZER SURGE LINE THERMAL STRATIFICATION BL-89-01 73 72 (98 )
1 NO FAILURE OF WESTINGHOUSE STEAM GENERATOR TUBE MECHANICAL PLUGS BL-89 109 106 (97 )
3
.NO STRESS CORROSION CRACKING OF HIGH-HARDNESS TYPE 410 STAINLESS STEEL BL-89-03~
72 72 (100) 0 N0' POTENTIAL LOSS OF REQUIRED SHUTDOWN MARGIN DURING REFUELING OPERATIONS BL-90-01 109 109 (1001 0
NO LOSS OF FILL-CIL IN TRANSMITTERS MANUFACTURED BY ROSEMOUNT Table 5.6
3 SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S).
SUMMARY
BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS'
-PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED BL-90-02 37 37 (1001 0
NO LOSS OF THERMAL N%RGIN CAUSED BY CHANNEL BOX BDW BL-92-01 109 57 (52 1_
52 NO
- FAILURE OF THERMO-LAG 330 FIRF BARRIER SYSTEM BL-93 02 109 J4
'(58 1 45-NO
>L IMPL OF PROG CONTROLS FOR RAD EFFLCENT TECH SPECS IN ADM CONT SECTION.
BL-93-03 34 1
12- )
33 NO j,
-RESOLUTION OF ISSUES REL TO REACTOR VESSEL WATER LEVEL INST IN BWR'S h.3 GL-79-25 19 19 (100)
.0 NO 1
INFORMATION REQUIRED TO REVIEW CORPORATE CAPABILITIES-GL-79-32 59 -
59 (100).
O NO TMI-2 LESSONS LEARNED TASK FORCE REPORT - NUREG-0578 GL-79-33
- 62
~ 62~~
(1001-0.
NO
. TRANSMITTING.NUREG-0567. SECURITY TRAINING AND QUALIFICATIONS PLAN.
GL-79-36 64 64 (1001 0
NO ADEQUACY OF-STATION ELECTRIC DISTRIBUTION SYSTEMS l
GL-79-43 16
" 16 (1001 0-NO
~
REACTORS CAVITY SEAL RING GENERIC ISSUE (PWR)
GL-79 -
- 64 64
'(1001-0-
NO CONTAINMENT PURGING AND VENTING DURING NORMAL OPERATION GUIDELINES :
-. 42 42
-ECCS CALCULATIONS 0N FUEL CLADDING ^
~
(1001 0-NO'
.GL-80-002 40' 40
-(100)-
0
-YES 40 37.
37
-(100)
QUALITY ASSURANCE -REQUIREMENTS REGARDING DIESEL GENERATOR FUEL DIL '
b
-GL-80-024-66- -(1001 0
'NO NRC NUCLEAR DATA LINK (NDL)
GL-80-030 62 62 (I00)
O' NO CLARIFICATION OF-THE1 TERM *0PERABLE" AS IT APPLIES TO SINGLE FAILUREL
~GL-20-061-20 20 L(100) 0 ~-
NO-TMI-2 LESSONS LEARNED' t
x Table 5.6 h
+,, -
.,-., --, I w# Emzn, i r
yv v Ewe
.c..
m.._ m.
..m._.-..__m_m_.-~.~._,m.__m~~.-
_..-m--.-_....
-_...~m mm___ _.._... __... _. -. _..
I i
6 i
?
SAFETY ISSUE MANAGEMEN T SYSTEM s
STATUS-OF OTHER MPAIS) - StM4ARY BY ITEM IMPLEA NTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT TTEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED ' COVERED REQUIRED COMPLETED COMPLETED GL-81-01 54 54 (1001 0
NO QUALIFICATION OF. INSPECTION, EXAMINATIONS, AND TESTING AND AUDIT'PERSO GL-81-04 65 65 11001 6:
NO EMERGENCY PROCEDURE AND TRAINING ~FOR STATION BLACKOUT EVENTS
.GL-81-14 42 42 (100) 6 NO-SEISMIC QUALIFICATION OF AUXILIARY FEEDWATER SYSTEMS GL-81-21 72 72 (1001 0-YES 72 67 60
('89 )
NATURAL CIRCULATION COOLDOWN g
a 23 22-
- (95 ')
1 YES 23.
23
-21 191 )
o
..GL-83-08 MODIFICATION OF VACUUM BREAKERS ON MARK I CONTAIIe4ENTS a
GL-83-43 79 79-(100)
-0
'NO REPORTING REQUIREMENTS OF 10 CFR PART 50;' SECTIONS 50.72 AND 50.73, AN GL-84-09 21 15 '
- (71 1 6
NO RECOMBINER CAPABILITY REQUIREM NTS OF.10 CFR 50 44(C)(3)(II].
GL-84-15 85 85
- (100)-.
O NO.
PROPOSED STAFF ACTIONS TO IMPROVE AND MAINTAIN DIESEL GENERATOR RELIA 8-GL-87-05 23 (100)- -
0 NO REQUEST FOR^ ADDITIONAL INFORMATION-ASSESSMENT OF LICENSEE MEASURES TO GL-87-09
. :64 60 (93 )
4' NO SECTIONS 3.0 AND 4.0 0F-THE STANDARD TECHNICAL SPECIFICATIONS (STS) ON GL-87-12 70 -
70/.
1(100),
0.
NO LOSS OF RESIDUAL HEAT REMOVAL (RHR) WHILE IN THE REACTOR COOLANT SYSTE-GL-88-01 37' 28
.'(75 )'
9'
^ NO MRC POSITION.ON IGSCC IM BWR AUSTENITIC STAINLESS STEEL PIPING GL-88-12 52-49
'(94 )
3-NO REMOVAL.0F FIRE PROTECTION REQUIREMENTS FROM TECHNICAL SPECIFICATIONS -
~
GL-88-05 72 72 -
' (200)~
0 NO
'{
BORIC ACID CORROSION OF. CARBON STEEL REACTOR PRESSURE BOUNDARY COMPONE-GL-88-06
-101' 101-
-(1001'
'O NO REMOVAL OF ORGANIZATIONAL CHARTS FROM TECHNICAL SPECIFICATION ADMINIST TaMc 5.6
i l
i SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S)'- SUpetARY BY ITEM IMPLEE,NTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED GL-88-11 109
-104 (95 )
5-NO NRC POSITION 04 RADIATION E98RITTLEE NT OF REACTOR VESSEL MATERIALS AN GL-88-16 84 84 '
(100) 0 NO REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS FROM TECHNICAL SPECIFICAT-I GL-88-20 108 7
~(6
)
-101 NO INDIVIDUAL PLANT. EXAMINATION FOR SEVERE ACCIDENT VULNER. 10CFR50.54(F)
GL.89-01 SS 51 (77 )-
-15 NO IMPL OF PROG CONTROLS FOR RAD EFFLUENT TECH SPECS IN AOM CONT SECTION O
GL-8g-04
'40 30 -
~(75 )
10 ~
-YES 40 38 3
(7 l'
Y GUIDANCE ON DEVELOPING ACCEPTABLE-' INSERVICE TESTING PROGRAMS GL-89-06
-95 90-
- ~(94-)
'S
'NO TASK ACTION PLAN ITEM I.D.2 - SAFETY PARAN TER DISPLAY-SYSTEM L
l-
-.GL-89-07 109
. 109
.'l100).
0 YES 109 108 108 1100)
POWER REACTOR SAFE 00ARDS COTINGENCY PLANNING FOR SURFACE VEHICLE BOMBS-
-109
-107
.(98:)
' 2' NO
- EROSION / CORROSION-IleUCED PIPE WALL THINNING :
- 109 1.
(0 )
108-
-.Y E S
'109 109 9
(8 )
f-
-SAFETY-RELATED MOTOR-OPERATED VALVE TESTING AND SURVEILLANCE.
l-GL-89-14
-82
-80
-(97 )
2-NO.
i f
LINE-ITEM IMPROV. IN TECH SPECS - REMOVAL OF.3.25 LIMIT.ON EXT SURV.
l-l-
GL.89.
24 20-
- 183- )
4 NO '
INSTALLATION OF A HARDENED WETWELL VENT-(GL.89-15):
i 1
I GL.90-02 12 12-(100) 0-.
NO.
l ALTERNATE REQ FOR FUEL ASSEMBLIES IN THE-DESIGN FEAT SECT OF TECH SPEC.'
GL-90.~71-70
. - (98 )
1.
NO ALT. REQ FOR $NU88ER VISUAL INSPECTION INERVALS & CORRECTIVE ACTIONS -
GL-91-01 22
- 21~
(95 l'..
1
- NO -
REMOVAL OF-THE SCHEDULE FOR THE WIT 2RAWAL- 0F REACT : VESSEL MAT SPEC.
~GL.91-04
'7
-4 (57 )
3 NO :
CHANGES..IN TECH SPEC SURV. INTERVALS TO ACCOMODATE A 24-M0 FUEL CYCLE Table 5.6
..r..
e,
.m
=
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPAIS) - SUM 4ARY BY ITEM IMPLEMENTATION VERIFICATION
..........................-_....................:1.
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETr0 COMPLETED GL-91-06 103 108 l100) 0 NO ADEf,UACY OF SAFETY-RELATED DC POWER SUPPLIES (GL 91-06) (GI A-30)
GL-91-08 24
- 22 (91 1 2
NO REMOVAL OF COMPONENT LISTS FROM TECHNICAL SPECIFICATIONS GL-91-09 5
5 (100) 6 NO MOD OF SURV INTERVAL FOR ELEC PROTO ASSEM IN POWER SUPPL GL-91-11 109 102 (93 )
7 NO VITAL INSTRUMENT BUSES & TIE BREAKERS (GI 48/49,GL 91-11]
GL-91-13 14 12 (85 )
2 NO g
ESSENTIAL SERVICE WATER SYSTEM FAILURES (GL-91-13) (GI-130)
GL-92-01 109 0
(0 1 109 NO REACTOR VESSEL STRUCTURAL INTEGRITY GL-92-04 37 27 (72 )
10 YES 37 37 0
(0 )
REACTOR VESSEL WATER LEVEL INSTRUMENTATION IN BWRS GL-92-08 109 30
~ (27 l-79 NO THERMO-LAG 330-1 FIRE BARRIERS GL-93-c4 49 0
(0 1 49 NO ROD CONTROL SYS FAIL & WITHDRAWAL OF ROD CONT CLUSTER ASSEMBLIES MPA.A024 96 96 (100) 0 NO MISCELLANEOUS AMENDMENTS AND SEARCH REQUIREMENTS MPA-A001 63 63 1100) 0 NO 10 CFR 50.55 A(G) - ISI MPA-A002 64 64 (100) 0 NO APPENDIX I - ALARA MPA-A003 61 61 (1001 0
NO SECURITY REVIEWS-MODIFIED AMENDMENT PLANS MPA-A004 49 47 195 )
2 NO APPENDIX J - CONTAINMENT LEAK TESTING MP'-A005 19 19 (100) 0 NO A
GE MARK I CONTAINMENT TECH SPECS-SHORT TERM Table 5.6
i' i
(
t SAFETY ISSUE MANAGEMENT SYSTEM 3
STATUS OF OTHER MPA(7) -
SUMMARY
BY ITEM IMPLEMENTATION -
__. VERIFICATION t
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT' ITEM APPLICABLE COMPLETED COMPLETED REMAINING.
REQUIRED COVERED REQUIRED COMPLETED COMPLETED l
MPA-A006 11' 11 (100) 0 NO RESPIRATORY PROTECTION SYSTEM MPA-A007 11 11 (1001 0
NO r
APPENDIX G - FRACTURE TOUGHNESS MPA-A008-10 10 (100) 0 NO ECCS EVALUATION-GENERIC PER 50.46 COMPLIANCE
.MPA-A009 60 60 (100) 0 NO PRESSURE VESSEL BELTLINE MATERIAL SURVEILLANCE g
MPA-A010 62 62 (100) 0 NO 3
CONTINGENCY PLANNING.-
MPA-A012 59 59 (100)'
O NO VITAL AREA ANALYSIS f
MPA-A014-50 50 (100) 0 NO 10CFR 50.55 A(G). INSERVICE TESTING MPA-B116 39 31 (79 ).
8 NO-RESULTS CF-NRC TESTING OF MOVS (GL 39-10 SUPP3)
MPA-B117 72 67
- (93 l-5 MO E
FAILURE.0F WESTINGHOUSE SG TUBE MECHANICAL PLUGS (BL 90-01 SUPP2)
MPA-Bilt 108
.1
- ( 0 -: )
107 NO
.IPE EXTERNAL EVEN1S (GL 88-20..SU9P 4)
MPA-8122 -
109 16
. (14 l-
. 93 -
NO
. LOSS OF FILL-0It' IN TRANSMITTERS MANUFACTURED BY ROSEMOUNT -
'MPA-8001 27 27
( 1C01 -
.0 NO DIESEL GENERATOR LOCMOUT-
'MPA-8002 55 55
[(10015 0~
NCI FIRE PROTECTION MPA-B003~
38 38
'(100)
?O
- YES-38.
.34
-34 (100)
PWR MODERATOR DILUTION MPA-B006..
22 22 (100)
'0.
- NO
' BWR RELIEF VALVE -
Table 5.6.
l:
. ~
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i e et w et E et ( Ms aC O (w (p (4 et W :(w (E 43 (I '(M
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w
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S] - SUP91ARY BY ITEM IMPLEMENTATION VERIFICATION i
i PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED MPA-8027 22 22 (1001 0
NO REVIEW RESPONSES TO IE BULLETIN 78-03 (OFFGAS EXPLOSIONS)
MPA-8029 5.
5 (100) 0 NO BWR FEEDWATER PUMP TRIP' MPA-8030 5
5 (100) 0 NO STEAM GENERATOR REPLACEMENT PROGRAM MPA-B032 23' 22 (95'l' 1
NO i
BLOCKED SI SIGNAL DURING COOLDOWN
.^
.MPA-8034 1
1 (100) 0 NO 8,
BWR JET PUMP INTEGRITY ASSURANCE
~
MPA-8035 5
5 (100) 0 NO ORIFICE ROD ASSEMBLY. INTEGRITY - B&W i
MPA-8036 7
7 (100)
'O NO
9 1100) 0 NO STEAM GENERATOR TUBE DENTING AND SUPPORT PLATE MODIFICATIONS - CE.
MPA-8038 2
2 (100)-
0' NO
. TENDON SURVEILLANCE - BECHTEL. CONTAINMENTS j
MFA-8039 36 36 (1001-
'O NO PWR PRESSURE'. TEMPERATURE LIMIT: TECH SPECS 4
-MPA-8040 1
1
.(1001 0'
NO t
PIPE SUPPORT BASE PLATES
'MPA-8041 85 63 (96 l' 2'
'YES 65 62-61 (98 1 FIRE PROTECTION - FINAL TECH SPECS (INCLUDES SER SUPPLEMENTS)
MPA-B046 52 52 (1001 0 '-
NO ANALYSIS OF TURBINE DISC CRACKS 3
MPA-8049 11
' 11
~(1001 0
NO PWR CONTROL ROD MISALIGNMENT MPA-B052 27
.27 (1001 -
0 NO '
+
REVIEW 0F SAFETY ASPECT OF INADVERTENT SAFETY ACTIONS DURING SUR. TEST
' Table 5.6 t
r i
m
~
a w
4
+
_...p
=
a m
m
m m.__.
___.m__.--
b i
L l
l I
f SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S). SUPNARY BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED CrsMPLETED MPA-B055 5
5 (100)
O NO-
'B&O REPORT ON BWRS MPA-8056 6
6
'(100) 6 NO CONTROL RODS FAILURE TO INSERT _ BWR MPA-8057 31 31 (100) b NO DHR CAPABILITY MPA-8064
. 7 7
(100) 0 NO ACC INDUCED FLUX ERRORS (B&W) 6 MPA-B067 8
8 (100) 0 NO N
THERMAL SHOCK e
-MPA-B070 44
'44 (100) 0 NO FATIGUE TRANSIENT LIMIT TS MPA-8073 7
7 (1001 0.
NO-PLANS FOR PREVENTING EXCEEDING PTS SCREENING CRITERION MPA-8074 4
4 (100) 0 NO THERMAL SHIELD FOLLOW UP ANALYSIS MPA-C006 34 34 (1001 0
NO.
PUMP SUPPORT-LAMELLAR TEARING MPA-C001 35 35 (100)
-0 NO PWR SECONDARY WATER CHEMISTRY MONITORING REQUIREMENTS MPA-C002 21 21 (100) 0 YES
.21' 21 19 (90 l' BWR-RECIRC. PUMP TRIP'(ATWS)
MPA-C003 15 15 (2001-0 NO
. QUALIFICATIONS OF RADIATION PROTECTION MANAGER MPA-C004 16
.16 (100) 0
. NO -
FILTER TECH SPECS MPA-C005 3
8 (100) 0 NO CONVERSION TD STANDARD TECH SPECS MPA-C007-
~ 36 36
'(1001 0
NO FUEL HANDLING ACCIDENT INSIDE CONTAINMENT
~ Table 5.6
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S) -
SUMMARY
BY ITEM IMPLEMENTATION VERIFICATIDN PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED C0%fLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED MPA-C005 9
9 (1001 0
NO BWR POST LOCA H2 CONTROL MPA-C009 7
7 (100) 0 NO 1%#t AUX FW PUMPS
.1 MPA.C011 22 21 (95 )
1 NO RPS PCWER SUPPLY MPA-C012 6
6 (1001 0
NO j,
BORON SOLUBILITY DURING LONG TERM COOLING FOLLDWING LOCA h.h MPA-0002-21 21 (100) 0 NO ECCS ZIRC' CLAD MODEL ERROR-COMPLIANCE WITH 10 CFR-46
~ 100]-
0 NO MPA-D003 16
'16
(
PRESSURIZER HEATUP RATE ERROR MPA-D005 6
6 (1001 0
NO PLANT UPI MODEL PROBLEM l
MPA-D006 6
6 (1001 0
.NO PEAKING MODEL CHANGE'FOR CE REACTOR CORE MPA-000T 3
.3 (2001 0
MPA-0008 3
3 (100)-
0
.: NO l
DEFICIENCY IN CHEM ADDITION TO CONTAINMENT SPRAYS MPA-D009 1
1 (100)-
0 NO
.GE ECCS INPUT ERROES MPA-Doll' 59 59 (100) 0 NO FISSION GAS RELEASE
.B&W SMALL~ BREAK ERROR.
6
~ 6'
'[1001 0-NO MPA-0013 MPA-D014' 10 10 (100) 0 NO REACTOR VESSEL WELD - WIRE DEFICIENCY MPA-D015 61
'61' (100) 0 NO HIGH ENERGY LINE. BREAK & CONSEQUENTIAL SYSTEM FAILURE
' Table 5.6
SAFETY ISSUE MANAGEMEN T SY STEM STATUS OF OTHER MPA(S) - SUMNMRY BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED MPA-DOIS 7
7 1I00) 0 NO NUREG 0630 CLADDING MODELS (B&W PLANTS)
MPA-E001 30 30 1I00) 6 NO SPENT FUEL POOL EXPANSIONS MPA-E002 7
7 (I00) 6 NO FUEL CASK DROP MPA-E003 30 30 (I00) 0 NO CORE RELOADS REQUIRING PRIOR NRC APPROVAL Q
MPA-E008 7
7 (I00) 0 NO y
CEA POSITION INDICATION FAILURES - CE MPA-E007 5
5 (IOO) 0 NO REACTOR PROTECTION SYSTEM LOGIC - CE Table 5.6
5.5 Conclusions j
After a detailed review of the implementation and verification status of the resolution of the 171 MPAs, the NRC staff has concluded the following:
The NRC closure for MPAs is adequate to protect the public health and safety.
Licensees are making progress toward implementing MPA-related actions requested by the staff, and the framework exists to oversee future implementation actions associated with those MPAs that have been resolved.
Progress is being made in verifying the completion of implementation actions associated with those MPAs that have been resolved.
The NRC staff will maintain close watch over implementation actions and schedules proposed by licensees to ensure that they are completed in accordance with regulatory requirements.
i
-l
-111-
i Appendix A-LISTING OF UNIMPLEMENTED TMI,lTEMS BYISSUE l
i v
tr w-m i
e4.
l i
k 1
j APPENDIX A i
j This appendix provides a detailed list, by issue, of the 61 TMI Action Plan items not j
implemented, along with the projected target date for completing the item. Status and projected implementation dates are presented as of September 30,1993.
4 1
l i
I i
i j
i f
i 4
1 1
I i
l 1
4
)
i i
I A-1
CI Ircuez (Listing of Open Items)
IMPL ISSUE MPA PLANT TAC TITLE DATE 1.
I.D.2.2 F075 BROWNS FERRY 1 M74602 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - INSTALLED 07/97 2.
I.D.2.2 F075 BROWNS FERRY 2 M74607 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - INSTALLED 10/93 3,
I.D.2.2 F075 BROWNS FERRY 3 M74612 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - INSTALLED 06/95 4.
I.D.2.3 F009 BROWNS FERRY 1 M51223 PLANT-SAFETY PARAMETER OISPLAY CONSOLE - FULLY IMPLEMENTED 07/97 5.
1.D.2.3 F009 BROWNS FERRY 2 M51224 PLANT-SAFETY FARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 10/93 6.
I.D.2.3 F009 BROWNS FERRY 3 M51225 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 06/95 7.
I.D.2.3 F009 PALO VERDE 1 M56654 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 11/93 S.
I.D.2.3 F009 PALO VERDE 3 M64581 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 11/93 9.
II.B.3.2 F076 BROUNS FERRY 3 M74613 POSTACCIDENT SAMPLING - CORRECTIVE ACTIONS 06/95 10.
II.B.3.3 F077 BROWNS FERRY 1 M74603 POSTACCIDENT SAMPLING - PROCEDURES 07/97 11.
II.B.3.3 F077 BROWNS FERRY 3 M74614 POSTACCIDENT SAMPLING - PROCEDURES 06/95 12.
II.B.3.4 F012 BROWNS FERRY 1 M44423 POSTACCIDENT SAMPLING - PLANT MODIFICATIONS (LL CAT B) 07/97 13.
II.B.3.4 F012 BROWNS FERRY 3 M44425 POSTACCIDENT SAMPLING - PLANT MODIFICATIONS (LL CAT B) 06/95 y
14.
II.D.1.3 F084 FORT CALHOUN 1 M75832 RELIEF & SAFETY VALVE TEST REQUIREMENTS - BLOCK-VALVE TESTING 06/94
$3 15.
II.E.4.2.1-4 F078 BROWNS FERRY 1 M74604 CONTAIMENT ISOLATION DEPEISABILITY - IMP. DIVERSE ISOLATION 07/97 16.
II.E.4.2.1-4 F078 BROWNS FERRY 3 M74615 CONTAIMENT ISOLATION DEPENDABILITY - IMP. DIVERSE ISOLATION 06/95 17.
II.E.4.2.6 F079 BROWNS FERRY 3 M74616 CONTAINMENT ISOLATION DEPENDABILITY - CNTMT PURGE VALVES 06/95 18.
II.F.1.1 F081 BROWNS FERRY 1 M74605 ACCIDENT-MONITORING - PROCEDURES 07/97 19.
II.F.1.1 F081 BROWNS FERRY 3 M74617 ACCIDENT-MONITORING - PROCEDURES 06/95 20.
II.F.1.2.A F020 BROWNS FERRY 1 M44903 ACCIDENT-MONITORING - NOBLE M S MONITOR 07/97 21.
II.F.1.2.A F020 BROWNS FERRY 3 M44905 ACCIDENT-MONITORING - NOBLE GAS MONITOR 06/95 22.
II.F.1.2.8 F021 BROWNS FERRY 1 M44974 ACCIDENT-MONITORING - IODINE / PARTICULATE SAMPLING 07/97 23.
II.F.1.2.8 F021 BROWNS FERRY 3 M44976 ACCIDENT-MONITORING - 10 DINE /PARTIQJLATE SAMPLING 06/95 24.
II.F.1.2.C F022 BROWNS FERRY 1 M45045 ACCIDENT-MON!TORING - CONTAINMENT HIGM-RANGE MONITDt 07/97 25.
II.F.1.2.C F022 BROWNS FERRY 3 M45047 ACCIDENT-MONITORING - CONTAINMENT NIGH-RANGE MONITOR 06/95 26.
II.F.1.2.0 F023 BROWNS FERRY 3 M47584 ACCIDENT-MONITORING - CONTAINMENT PRESSURE 06/95 27.
II.F.1.2.E F024 BROWNS FERRY 3 M47655 ACCIDENT-MONITORING - CONTAINMENT WATER LEVEL 06/95 28.
II.F.2.4 F026 BROWNS FERRY 1 M45116 INSTReetTATM FOR DETECT. OF IMADEQ CORE CLNG INSTLL ADD'l INSTRMNTATM 07/97 29.
II.F.2.4 F026 BROWNS FERRY 3 M45118 INSTRMNTATM FOR DETECT. OF INADEQ CORE CLNG INSTLL AD0'L INSTRMNTATM 06/95 30.
II.F.2.4 F026 DRESDEN 3 M45130 INSTRMNTATM FOR DETECT. OF INADEQ CORE CLNG INSTLL AD0'L INSTRMNTATM 11/93 31.
II.F.2.4 F026 QUAD CITIES 2 M45165 INSTRMNTATM FOR DETECT. OF INADEQ CCRE CLNG INSTLL ADD'L INSTRMNTATM 11/93 32.
II.K.3.13.8 F043 BROWNS FERRY 1 M45532 B&O TASK FORCE - HPCI & RCIC INITIATION LEVELS MODIFICATION 07/97 33.
II.K.3.13.8 F043 BROWNS FERRY 3 M45534 B&O TASK FORCE - HPCI & RCIC INITIATION LEVELS MODIFICATION 06/95 34.
II.K.3.18.C F048 BROWNS FERRY 1 M45680 B&O TASK FORCE - ADS ACTU4 TION Mm)IFICATIONS 07/97 35.
II.K.3.18.C F048 BROWNS FERRY 3 M45652 B&O TASK FORCE - ADS ACTUATION MG)IFICATIONS 06/95 s
._-a
,~~u....
--- ~.~ ~ ~. - ---
- -. -. - -. ~....-.. - -.
36.
II.K.3.27 F054 BROWS FERRY 1 9145776 380 TASK FORCE - COBOION REFERENCE LEVEL FOR BWtt 07/97 37.
II.K.3.27 F054 SROWS FERRY 3 M45778 - 380 TASK FORCE - COINION REFERENCE LEVEL FOR OMt3 06/95 38.
II.K.3.28 FQ55 BROWS FERRY 1 M48260 880 TASK FORCE - GUALIFICATION OF ADS ACCUMULATORS 07/97 39.
II.K.3.28 F055 stoiAss FERRY 3 1848262 580 TASK FORCE - GUALIFICATION OF ADS ACCISRALATORS 06/95 40.
III.D.3.4.3 F070 NADDAM NECK M46450 CONTROL ROOM MARITABILITY
- IMPLE9ENT MODIFICATIONS
-12/93 41.
MPA-F071 F071 BIG ROCK POINT 1 M56103 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLL0lAJP TO F-8) 12/93 42.
MPA-F071 F071 sRotAss FERRY 1 M56104 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLotAAP TO F-8) 07/97 43.
MPA-F071 F071 SRotAfS FERRY 3 M56106 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 06/95 44.
MPA-F071 F071 DIASLO CANYON 1 M56117 f.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 05/94 45.
MPA-F071 F071 DIASLO CAllY0Il 2 M68040. I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 10/94 46.
MPA-F071 F071 IIADDAM NECK M56128 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 03/96 47 1MPA-F071 F071 MILLSTONE 1 M56138 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLotAAP TO F-5) 12/95 48.
MPA-F071 F071 NINE MILE POINT 1 M56141 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP 70 F-8) 03/95 y 49 MPA-F071
,F071 IIORTN ANNA 1
-M56142 I.D.1.2 del %ILED CONTROL 1t0134 REVIEW (FOLL0lAAP TO F-8) -
12/94 A 50.
MPA-r071 ro71 NORTN ANNA 2
-M56143 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/94
- 51.. MPA-F071 F071-PILeftIn 1 M59329 I.D.1.2 DETAILED COIITROL ROOM REVIEW (FOLLOWUP 70 F-8) ~
03/95 52.
MPA-F071 F071 PetNT BEACM 1.
M56152.-
I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/95 53.
MPA-F071
-- F071 POINT BEACM 2 M56153 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/93 54 MPA-F071 F071 PRAIRIE IStase 1 M56154 f.D.1.2 DETAILED CONTROL ROOM REVIEW (FollotAJP TO F-8) 04/94 55.
MPA-F071 F071. PRAIRIE ISLAND 2 M56155 I.D.1.2 DETAILED CONTitel ROOM ItEVIEW (FOLLOWUP TO F-8) 11/93 56.
MPA-F071
.F071 SAN ONOFRE 2:
M56163-.
I.D.1.2 DETAILED CONTROL ROOM REVIEW (FDLLOWUP TO F-8)'
.10/93 57.
MPA-F071 -
F071 SAN ONOFRE 3 M56164 f.D.1.2 DETAILED COWilt0L ROOM REVIEW (FOLLOWUP YO F-8) 10/95 58.
MP4-F071 F071 SEGUOTAM 1 M56165-I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP 70 F-8)-
12/93 59.
MPA-F071 '
'F071 SEGUOTAN 2 M56166' I.D.1.2 DETAILED CONTROL R0084 REVIEW (FOLLOWUP TO F-8)
~12/93 i
60.
MPA-F071 F071 -SURRT.1
-M56170 f.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/94 61.
MPA-FOTI F071-SURRf 2l M56171. I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) -
12/94.
a n., ~.. _.
=.
i
-1 Appendix B LISTING OF UNIMPLEMENTED USl ITEMS BYISSUE Y
t
)
i i.
1
APPENDIX B This appendix provides a detailed list, by issue, of the 177 USI items not implemented, along with the projected date for completing the item. Status and projected implementation dates are presented as of September 30,1993.
1 i
i
(
B-1 4
~_.-_u
... - ~. - ~.. _ -
-n
~- - -
.... ~.
-.-._,_. -.~ - ~
.. _. - _.~ ~
Unresolved Safety Issues (Listing of Open Items)
IMPL ISSUE MPA PLANT TAC TITLE DATE 1.
A-24 8060 BROWNS FERRY 1 M42481 QUALIFICATION OF CLASS 1E SAFETY-RELATED EQUIPMENT 07/97 2.
A-24 8060 BROWNS FERRY 3 M42483 QUALIFICATION OF CLASS 1E SAFl!TY-RELATED EQUIPMENT 06/95 3.
A-31 S004 HADDAM WECK M77025 RHR SHUTDOWN REQUIREMENTS 10/94 4.
A-44 A022 ARKANSAS 1 M68508 STATION BLACKOUT 12/94 5.
A-44 A022 ARKANSAS 2 M68509 STATION BLACKOUT 12/94 6.
A-44 A022 BIG ROCK POINT 1 M68514 STATION BLACKOUT 10/94 7.
A-44 A022 BROWNS FERRY 1 M68517 STATION BLACKOUT 06/95 8.
A-44 A022 BROWNS FERRY 2 M68518 STATION BLACKOUT 06/95 9.
A-44 A022 BROWNS FERRY 3 M68519 STATION BLACKOUT 06/95 10.
A-44 A022 BRUNSWICK 1 M68520 STATION BLACKOUT 10/93 11.
A-64 A022 BRUNSv!CK 2 M68521 STATION SLACK 0UT 10/93 12.
A-44 A022 CALVERT CLIFFS 1 M68525 STATION SLACK 0VT 05/M 13.
A-44 A022 CALVERT CLIFFS 2 M68526 STAT 10N BLACKOUT 05/96 g) 14.
A-44 A022 CATAWBA 1 M68527 STATION BLACKOUT 06/94 U 15.
A-44 A022 CATAWBA 2 M68528-STATION 8LACKOUT 06/94 16.
A-44 A022 COMANCHE PEAK 1 M68530 STATION BLACKOUT 07/94 17.
A-44 A022.
COOK 1 M68532 STAT 10N BLACKOUT 11/93 18.
A-44 A022 COOK 2 M68533 STATION BLACKOUT
- 11/93 19.
A-44
.A022 CRTSTAL RIVER 3 M68535 STATION BLACK 0'UT 06/94 20.
A-44 A022 FARLEY 1 M68543. STATION BLACKOUT 04/94 21.
A-44 A022 FARLEY 2 M68544 STAfl0N BLACKOUT 10/93 22.
A-44 A022 FIT 2 PATRICK M68546. STATION SLACK 0UT 06/94 23.
A-44 A022 FORT CALHOUN 1 M68547 STAft0N BLACKOUT 11/93 24.
A-44 A022 GINMA M68548. STATION BLACKOUT 09/94 25.
A-44 A022 HADOAM NECK M68551 STAT 10N BLACKOUT 10/93 26.
A A022 HARRIS 1 M68552 STATION BLACKOUT 06/94 27.
A-44 A022' HOPE CREEK 1 M68555 STATION BLACKOUT 06/94 28.
A-44 A022 INDIAN POINT 2 M68556 STATION BLACKOUT 12/93 29.
A-44 A022 INDIAN POINT 3 M68557 STATION BLACKOUT
- 06/94 30.
A-44 A022 KElmuMEE M68558' STATION BLACKOUT 11/93 31.
A-44 A022 McGu!RE 1 M68564 STATION SLACK 0UT' 06/94 32.
A-44 A022 MCGUIRE 2 1168565-. STATION BLACKOUT 06/94 33.
A-44 A022 MILLSTONE 1
- H8566 STAfl0N BLACKOUT 02/94' 34.
A-44 A022 MILLSTONE 2 M61ri'?. STATION BLACEDUT 04/94 35.
A-44 A022 MILLSTONE 3 M68568 STATION BLACKOUT
- 10/93 r
2
--r.
2
l l
l l
l 36.
A-44 A022 MONTICELLO M68569 STATION BLACK 0UT 12/94 37.
A-44 A022 NORTM ANNA 1 M68572 STATION BLACKOUT 12/94 38.
A-44 A022 NORTM ANNA 2 M68573 STATION BLACKOUT 12/94 39.
A-44 A022 PALISADES M68578 STATION BLACKOUT 10/93 40.
A-44 A022 PALO VERDE 1 M68579 STATION BLACKOUT 11/93 l
41.
A-44 A022 PALO VERDE 2 M68580 STATION BLACKOUT 11/94 t
42.
A-44 A022 PALO VERDE 3 M68581 STATION BLACKOUT 04/94 43.
A-44 A022 PEACH BOTTOM 2 M68582 STATION BLACK 0UT 10/94 44.
A-44 A022 PEACH BOTTOM 3 M68583 STATION BLACKOUT 10/94 l
45.
A-44
'A022 PERRY 1 M68584 STATION BLACKOUT 07/94 46.
A-44 A022 QUAD CITIES 1 M68590 STATION BLACK 0UT 12/95 47.
A-44 A022 QUAD CITIES 2 M68591 STATION BLACK 0UT 12/95 48.
A-44 A022 RIVER SEND 1 M68593 STATION BLACKOUT 03/94 (I) 49.
A-44
-A022 R08tNSON 2 M68595 STATION BLACKOUT 11/93 E 50.
A-44 A022 SALEM 1 M685% ' STATION 8tACKOUT 01/94 51.
A-44 A022 SALEM 2 M68597 STATION BLACKOUT 01/94 52.
A-44 A022 SAN ONOFRE 2 M68599 STATION BLACKOUT 09/94 53.
A-44 A022 SAN ONOFRE 3 M68600 STATION BLACKOUT 09/94
' 54.
A-44 A022 SEQUOTAM 1 M68603 STATION BLACKOUT 06/94 55.
A-44 A022 SEQUOTAN 2 M68604 STATION BLACKOUT 06/94 56.
A-44 A022 SURRY 1 M68611 STATION BLACKOUT--
.01/96 57.
A-44 A022 SURRY 2 M68612 STATION BLACKOUT 05/96 58.
A-44 A022 SUSQUENANNA 1 M68613 STATION BLACKOUT 06/94 59.
A-44 A022 SUSQUENANNA 2 M68614 STATION BLACKOUT 06/94 l
60.
A-44~
A022 V0GTLE 1 M68621 STATION SLACK 0UT 02/94 61.'
A-44 A022 V0GTLE 2 M73447 STATION BLACK 0UT 02/94 62.
A-44 A022 WASHINGTON NUCLEAR 2 M68626 STATION BLACKOUT 06/94 63.
A-44 A022 WATERFORD 3 M68623 STATION BLACKOUT 06/94 64 A-46 B105 ARKANSAS 1 M69426 SEISMIC GUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/95 65.
A-46 3105 ARKANSAS 2 M69427 SEISMIC GUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/95 66.
A-46 3105 BEAVER VALLET 1 M69428 SE!SMIC GUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/95
. 67.
A-46 5105 BIG ROCK POINT 1 M69429 SEISMIC GUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/95 l.
68.
A 5105 3ROWNS FERRY 1 M69430 SEISNIC GUALIFICATION OF EeUIPDENT IN OPERATING PLANTS 07/97 69.
A-46 5105 BROWNS FERRY 2 M69431 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/96
. 70.~
A-46 8105 BROWNS FERRY 3 M69432 SEISMIC OUALIFICATION'0F EeUIPMENT IN OPERATING PLANTS 07/96
~+
-...--...a-.m.w
. - - ~ - - - - _ _ ~ - -
71.
A-M 3105 3RUNSUICK 1 M69433 SEISMIC QUALIFICATION OF (YJI M NT IN OPERATING PLANTS 06/95 72.
A-M siO5 sRUNSUICK 2 M69434 SEISMIC O W FICATION OF E W AENT IN CPERATING PLANTS 10/95 73.
A;4 5103 CALVERT CLIFFS 1 M69435 SEthMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/96 74.
A-M 5105 CALVERT CLIFFS 2 M69436 SEISMIC ouALIFICATION OF EQUIPMENT IN OPERATING PLANTS -
07/96 75.
A-M B105 COOK 1 M69437 SEISMIC GUALIFICATION OF EQUl> MENT IN C*ERATING PLANTS 07/95 76.
A-46 8105 COOK 2 M69438 SEISMIC QUALIFICP. TION D' EQUIPMENT IN OPERATING PLANTS 07/95 77.
A-M B105 COOPER STATION M69439 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/95 78.
A-46 B105 CRTSTAL RIVER 3 M69440 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATIha PLANTS 12/95 79.
A-46 5105 DAVIS-BESSE 1 M69441 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 09/95 80.
A-4 5105 DRESDEN 2 M69442 SEISMIC QUALIFICATION OF EQUIPMENT lu OPERATING PLANTS 11/95 81.
A-M ~
B105 DRESDEN 3 M69443 SEISMIC cuALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 82.
A-46 8105 Du'NE ARNOLD M69444 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95
- 83.
A-46 BIOS FARLET 1 M69445 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 05/95 g 84.
A-46 5105 FIT 2 PATRICK M694M SEISMIC QUALIFICATI0st OF EQUIPMENT IN OPERATING PLANTS 05/95 U3 85.
A-4 3105. FORT CALMOUM 1 M69447 SEISMIC cuALIFICATION OF EQUIPMENT IN OPERATING PLANTS 09/95 86.
A-4 3105-GINNA M69449 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 05/95 87.
A-4 5105 NA00AM NECK-M69450 SEISMIC GuALIFICATION OF EQUIPMENT IN OPEkATING PLANTS 11/93 88.
A-46 5105 ' NATCM 1 M69451 SEISMIC GudlFICATION OF EQUIPENT IN OPERWNG PLANTS 08/95 89.
A-M 5105 NATCN 2 M69452 SEISMIC cuALIFICATION OF EQUIPMENT IN OPE,4?'IG PLANTS 08/95' 90.
A-M B105 INDIAll POINT 2 M69453' SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 91.
A-M 3105 INDIAN POINT 3
- M69454 SEISMIC OUALIFICATION OF EQUIPE NT IN OPERATING PLANTS.
11/95 92.
A-46 5105- ~KEWAUNEE M69455 SEISMIC ouALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 93.
A-M BIOS. MILLSTONE 1 M69458-SEISMIC cuALIFICATION OF EQUIPMENT IN OPERATING PLANTS 08/96-94.
A-M -
5105.. MILLSTONE 2 M69459. SEISMIC GUALIFICATION OF EQUIPMENT.IN OPERATING PLANTS.
05/95 95.
A-46 5105' MONTIu'lJ M69460 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 96 A-M 5105 ' NINE f.iLE POINT 1-M69461. SEISMIC QUALIFICATION OF EeulPMENT.IN CPERATING PLANTS 05/95 97.
A-M B105 NORTN ANNA 1 M6942. SEISMIC GUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 12/94 2
98.
A-4 5105..NORTN ANNA 2' M69 43 SEISMIC cuALIFICAtl0N OF EetlIPMENT IN OPERATING PLANTS 06/95 99.
A-4 3105. OCONEE 1 M69464 SEISMIC GUALIFICATION OF EQUIPE NT IN OPERATING PLANTS 06/95
' 100.- ' A-4
. 3105 - OCONEE 2 :
M69465 SEISMIC auALIFICATION OF EeUIPE NT IN f9ERATING PLANTS 06/95 101.
A-4 5105. OCGIEE 3 M6946 SEISMIC euRLIFICATION OF EeuiPE NT IN OPERATING PLANTS '
06/95 102. ' A-M
' 3105. OTSTER CREEK 1 M69467 SEISMIC OUALIFICATION OF EOUIPENT AN OPERATING PLANTS :
09/95 103.z. A-M -'
3105'- PALISADES M69468 SEISMIC euALIFICATION OF EeulP E NT IN OPERATING PLANTS
. 07/95 104.
A-M 3105 - PEACN 90T!0M 2' M69469 SEISMIC cuALIFICATION OF EeUIPMENT IN OPERATING PLANTS 11/95 105.
A-46 5105 PEACM BOTTON 3 M69470. SEISMIC auALIFICATION OF EeUIPE NT.IN OPERAfl#G PLANTS
- 11/95 2
_.._._._,_m._m.___m m.
mm mm en Y-m m
x=..
l 106.
A-46 B105 PILGRIM 1 M69471 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 107 A-46 B105 POINT BEACM 1 M69472 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 08/95 108.
A-46 8105 POINT BEACH 2 M69473 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 08/95 109.
A-46 8105 PRAIRIE ISLAND 1 M69474 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 110.
A-46 BIOS PRAIRIE ISLAND 2 M69475 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 111.
A-46 8105 QUAD CITIES 1 M69476 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 112.
A-46 8105 QUAD CITIES 2 M69477 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 113.
A-46 8105 ROBINSON 2 M69478 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATIkG PLANTS 12/94 114.
A-46 B105 SALEM 1 M69479 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 115.
A-46 8105 SALEM 2 M69480 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 116.
A-46 9105 ST LUCIE 1 M69483 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 12/93 117.
A-46 8105 SURRY 1 M69484 3EISMIC QUALIFICATION OF EQUIPMENT IM OPERATING PLANTS 11/95 118.
A-46 8105 SURRY 2 M69485 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 Q 119.
A-46 BIOS THREE MILE ISLAND 1 M69486 SEttatIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 12/95
) 120.
A-46 8105 TURKEY POINT 3 M68303 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 12/93 121.
A-46 BIOS TURKEY POINT 4 M68304 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 12/93 122.
A-46 8105 VERMn N YANKEE 1 M6%90 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/95 123.
A-46 B105 ZION 1 M69492 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11/95 124 A-46 B105 ZION 2 M69493 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 11ra" 125.
A-47 B113 BROWNS FERRY 1 M74915 SAFETY IMPLICATION OF CONTROL SYSTEMS 47/97 126.
A-47 B113 BROWNS FERRY 3 M74917 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/95 l
127.
A-47 B113 CALVERT CLIFFS 1 M74923. SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 128.
A-47 8113 CALVERT CLIFFS 2 M74924 SAFETY IMPLICATION OF CONTROL SYSTFMS 06/94 1
129.
A-47 B113 COOPER STATION M74932 SAFETY IMPLICATION OF CONTROL SYSTEMS 12/93 130.
A-47 B113 CRYSTAL RIVER 3 M74933 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 131.
A-47 8113 DRESDEN 2 M74937 SAFETY IMPLICATION OF CONTROL SYSTEMS 01/94 C2.
A-47 8113 DRESDEN 3 M74938 SAFETY IMPLICATION OF CONTROL SYSTEMS 01/94 I
133.
A-47 B113 FITZPATRICK M74943 SAFETY I w LICATION OF CONTROL SYSTEMS 12/93 i
134 A-47 B113 FORT CALHOUN 1 M74944 SAFETY IMPLICATION OF CONTROL SYSTEMS 12/93 135.
A-47 B113 NATCM 1 M74949 SAFETY IWLICATION OF CONTROL SYSTEMS 06/94 l
136.
A-47 B113 NATCH 2 M74950 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 137.
A-4 7 8113 LASALLE 1 M74955 SAFETY IMPLICATION OF CONTROL SYSTEMS
/
138.
A-47 B113 LASALLE 2 M74956 SAFETY IMPLICATION OF CONTROL SYSTEMS
/
139.~ A-47 B113 MILLSTONE 1 M74962 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 140.
A-47 Bil3 MILLSTONE 2 M74963 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 i
141.
A-47 8113 NINE MILE POINT 1 M74966 SAFETY IMPLICATION OF CONTROL SYSTEMS 12/93 142.
A-47 8113 NINE MILE POINT 2 M74% 7 SAFETY IMPLICAliON OF CONTROL SYSTEMS 12/93 143.
A-47 8113 OCONEE 1.
M74970 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 144.
A-47 8113 OCONEE 2 M74971 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/93 145.
A-47 B113 OCONEE 3 M74972 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 146.
A-47 8113 OYSTER CREEK 1 M74973 SAFETY IMPLICATION OF CONTROL SYSTEMS 12/94 147.
A-47 8113 PALISADES M74974 SAFETY IMPL'r.ATION OF CONTROL SYSTEMS 10/93 148.
A-47 8113 PEACH OOTTOM 2 M74978 SAFETY IMPLIL'. TION OF CONTROL SYSTEMS 04/94 149.
A-47 8113 PEACM BOTTOM 3 M74979 SAFETY INPLICATION OF CONTROL SYSTEMS 04/94 150.
A-47 8113 PILGRIM 1 M74981 SAFETY IMPLICATION OF CONTROL SYSTEMS 12/93 151.
A-47 8113 POINT SEACH 1 M74982 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 152.
A-47 8113 PotNT BEACH 2 M74983 SAFETY IMPLICATION OF CONTROL SYSTEMS 06/94 153.
A-47 8113 QUAD CITIES 1 M74986 SAFETY IMPLICATION OF CONTROL SYSTEMS
/
154.
A-47 8113 QUAD CITIES 2 M74987 SAFETY IMPLICATION OF CONTROL SYSTEMS
/
g 4 155.
A-47 8113 SAN ONOFRE 2 M74994 SAFETY IMPLICATION OF CONTROL SYSTEMS 05/94 156.
A-47 8113 SAM ONOFRE 3 M74995 SAFETY IMPLICATION OF CONTROL SYSTEMS 05/94 157.
A-47 8113 ST LUCIE 1 M75002 SAFETY IMPLICATION OF CONTROL SYSTEMS 07/94 158.
A-47 8113 ST LUCIE 2 M75003 SAFETY IMPLICATION OF CONTROL SYSTEMS 07/94 159.
A-47 8113 TURKEY PotNT 3 M75011 SAFETY INPLICATION OF CONTROL SYSTEMS 04/94' 160.
A-47 8113 TURKEY PotNT 4 M75012 SAFETY IMPLICATION OF CONTROL SYSTEMS 04/94 161.
A-47 8113 WATERFORD 3 M75016 SAFETY IMPLICATION OF CONTROL SYSTEMS 01/94 162.
A-47 8113 ZION 1 M75022 SAFETY IMPLICATION OF C0hTROL SYSTEMS 12/93 163.
A-47 8113 ziaN 2 M75023 SAFETY IMPLICATION OF CONTROL SYSTEMS 12/93 164.
A-48
$003 CLINTON 1-M81719 NYDROGEN CONTROL MEASURES AND EFFECTS OF HYDROGEN SURNS 03/94 165.
A-48 S003 COOK 1 M49520 NYDROGEN CONTROL MEASURES AND EFFECTS OF NYDROGEN BURNS.
12/93 166.
A-48 5003 C00K 2 M49519 HYDROGEN CONTROL MEASURES AND EFFECTS OF HYDROGEN BURNS 12/93 167.
A-48
$003 GRANO GULF 1 M77024 HYDROGEN CONTROL MEASURES AND EFFECTS OF NYDROGEN SURNS 12/93 168.
A-48 5003 PERRY 1 M66121 HYDROGEN CONTROL MEASURES AND EFFECTS OF NYDROGEN OURNS 12/93 169.
A-48 5003 RIVER BEND 1 M$9714 HYDROGEN CONTROL MEASURES AND EFFECTS OS HYDROGEN 8tutNS 02/94 170.
A-7 0001 BROWNS FERRY 1 M07929 MARE I LONG-TERM Pe0 GRAM C7/97 171.
A-7 0001 BROWNS FERRY 3 M07931 MARK I LONG-TiaM PROGRA;:
06/95 07/97 172.
A-9 A020 BROWNS FERRY 1 M59072 ATUS 06/95 173.
A-9 A020 BRobMS FERRY 3 M59074 ATUS-03/94 174.
A-9 A020 DRESDEN 3 M59090 ATWS 11/93 175.
Ae9 A020 LASALLE 2 M59108 ATUS
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Appendix C.
.I 4
1 LISTING OF UNIMPLEMENTED GSI ITEMS 1
BYISSUE 1
1 I
1 1
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I APPENDIX C This appendix provides a detailed list, by issue, of the 158 GSI items not impleme.3ted, along with the projected date for completing the item. Status and projected implementation dates are presented as of September 30,1993.
1 e
4 I
C-1
Generic Safety Istues (Listing of Open Itew O IMPL ISSUE MPA PLANT TAC TITLE DATE 1.
124 5001 CRYSTAL RIVER 3 M67802 AUXILIARY (EEDWATER SYSTEM RELIABIL TY 10/93 2.
40 8065 BROWNS FERRY 1 M43727 SAFETY CONCERNS ASSOCIATED WITH PIPE BREAKS IN BWR SCRAM SYSTEM 07/97 3.
40 B%5 BROWNS FERRY 3 M43736 SAFETY CONCERNS ASSOCIATED WITH PIPE BREARS IN BWR SCRAM SYSTEM 06/95 4
4' B058 BROWNS FERRY 3 M51014_ BWR SCRAM DISCHARGE YOLtME SYSTEMS 06/95 5.
67.3.3 A017 BEAVER VALLEY 1 M51071 IMPROVED ACCIDENT MONITORING 02/94 6.
67.3.3 A0i? BROWNS FERRY 1 M51073 IMPROVED ACCIDENT MONITORING 07/97 7.
67.3.3 A017 BROWNS FERRY 3 M51075 IMPROVED ACCIDENT MONITORING 06/95 8.
67.3.3 A017 CALVERT CLIFFS 1 M51078 IMPROVED ACCIDENT MONITORING 06/94 9.
67.3.3 A017 CALVERT CLIFFS 2 M51079 IMPROVED ACCIDENT MONITORING 06/94 10.
67.3.3 A017 COOK 1 M51080 IMPROVED ACCIDENT MONITORING 10/93 11.
67.3.3 A017 C00K 2 M51081 IMPROVED ACCIDENT MONITORING 10/93 12.
67.3.3 A017 COOPER tTATION M51082 IMPROVED ACCIDENT MONITORING 11/94 13.
67.3.3 A017 FERMI 2 M59620 MPROVED ACCIDENT MONITORING 12/94 O 14.
67.3.3 A017 FFti CALHOUN 1 M51091 IMPROVED ACCIDENT MONITORING 11/93 b 15.
67.3.3 A017 GINNA M51093 IMPROVED ACCIDENT MONITORING 12/93 16.
67.3.3 A017 GRAND GULF 1 M51094 IMPROVED ACCIDENT MONITORING 11/93 17.
67.3.3 A017 HADDAM NECK M51095 IMPROVED ACCIDENT IONITORING 12/94 18.
67.3.3 A017 MATCH 1 M51096 IMPROVED ACCIDENT MONITORING 12/94 19.
67.3.3 A017 HATCH 2 M51097 IMPROVED ACCIDENT MONITORING 12/94 20.
67.3.3 A017 :NDIAN POINT 2 M51098 IMPROVED ACCIDENT MONITORING 12/93 21.
67.3.3 A017 KEWAUNEE M51100 IMPROVED ACCIDENT MONITORING 10/93 22.
67.3.3 A017 LASALLE 1 M51102 IMPROVED ACCIDENT MONITORING
/
23.
67.3.3 A017 LASALLE 2 M56407 IMPROVED ACCIDENT MON!TORING
/
24.
67.3.3 A017 MILLSTONE 1 M51106 IMPROVED ACCIDENT MONITORING 11/93 25.
67.3.3 A017 MILLSTONE 3 M65327 IMPROVED ACCIDENT MONITORING 12/93 26.
67.3.3 A017 NINE MILE-POINT 2 M79172 IMPROVED ACCIDENT MONITORING 10/93 27, 67.3.3 A017 OYSTER CREEK 1 M51115 IMPROVED ACCIDENT MONITORING 12/94 28.
67.3.3 A017 PERRY 1 M79010 IWROVED ACCIDENT MONITORING
/
29.
67.3.3 A017 QUAD CITIES 1 M51124 IMPROVED ACCIDENT MONITORING 05/94 30.
67.3.3 A017 ouAD CITIES 2 M51125 IMPROVED ACCIDEET MONITORING 05/94 31.
67.3.3 A017 SAN ONOFRE 2 M51131 IMPROVED ACCIDENT MONITORING 11/93 32.
67.3.3 A017 !%N ONOFRE 3 M51132 IMPROVED ACCIDENT MONITORING 11/93 33.
67.3.3 A017 tOJTH TEXAS 1
'M63480 IMPROVED ACCIDENT MONITORING 02/94 34.
67.3.3 A017. SOUTN TEXAS 2 M77956 IMPROVED ACCIDENT MONITORING 02/94 35.
67.3.3 A017 ST LUCIE 1 M51135 -IMPROVED ACCIDENT MONITORING 07/96
l l~
l 4
36.
67.3.3 A017 ST LUCIE 2 M51136 IMPROVED ACCIDENT MONITORING 09/95 37.
67.3.3 A017 $ Ulster 1 M51137 IMPROVED ACCIDENT MuellTORING 01/94 38.
67.3.3 A017 ZION 1 M51367 IMPROVED ACCIDENT MONITORING 10/93 39.
67.3.3 A017 ZION 2 M51368 IMPROVED ACCIDENT-MONITORING 10/93 40.
70 8114 BEAVER VALLEY 1 M77328 PORY AND BLOCK VALVE RELIABILITY 03/94 41.
70 8114 DEAVER VALLEY 2 M77329 PORY AND BLOCK VALVE RELIABILITY 03/94 42.
70 8114 BRAIDWOOD 1 M77332 PORY AND BLOCK VALVE RELIABILITY
/
43.
70 8114 BRAIDWOOD.2 M77333 PORY AND BLOCK VALVE RELIABILITY
/
44 70 B114 CALVERT CLIFFS e.
M77337 PORV AND BLOCK VALVE RELIA 8ILITY 12/93 45.
70 8114' CALVERT CLIFFf 2 M77338 PORY A W BLOCK VALVE RELIABILITY 12/93 46.
70 8114 COOK 1 M77343 PORY AW BLOCK VALVE RELIABILITY 11/93 47 70 8114 C00K 2 M77344 PORY AND BLOCK VALVE RELIABILITY 11/93 48.
70 8114 CRYSTAL RIVER 3 M77345 PORY AND BLOCK VALVE RELIABILITY.
12/93 O 49.
70 8114. FORT CALHOUN 1 M77351 PORY AND BLOCK VALVE RELIABILITY 12/93 b 50.
70 -
8114 GINNA M77352 PORY AND BLOCK VALVE RELIABILITY 01/94 51.
70 5114 MADDAM NECK M77353 PORY AND BLOCK VALVE RELIA 8ILITY 10/93 52.
70 B114-INDIAN POINT 2 M77355 PORV A W BLOCK VALVE RELIABILITY 01/94
-53.
70 8114 INDIAN POINT 3 M77356 PORY A W SLOCK VALVE RELIA 8ILITY 12/93 54.
70 8114 KEWAUNEE M77357 20RV AND BLOCK VALVE RELIABILITY 12/93 55.
TO B114 MCGUIRE 1 M77359 PORV AND SLOCK VALVE RELIABILITY-01/94 56.
70 5114 MCGUIRE 2 M77360 ' PORY AND BLOCK VALVE PELIABILITY-01/94 57.
70 8114. MILLSTONE 2 '
~ M77361 PORY AND BLOCK VALVE RELIABILITY 01/94 58.
70 8114' MILLSTONE 3 M77362 PORY AND BLOCK VALVE RELIABILITY 11/93
- 59. - 70
.B114 NORTN ANetA 1 M77363 PORY AND BLOCK VALVE RELIABILITY 12/93 60.
.70 5114 If0RTN AMesA 2 M77364 PORY AND BLOCK VALVE RELIABILITY 12/93.
61.
70 c 3114 : OCONEE 1 M77365 PORY AW SLOCK VALVE RELIABILITY 01/94 62.
70 5114 OCONEE 2-MT7366 PORY AND BLOCK VALVE RELIABILITY-01/94 63.
70 8114' OCONEE 3-M77367 PORY AND SLOCK VALVE RELIABILITY 01/94-64.
70 B114 PALISADES M77368 PORY AND RLOCK VALVE RELIASILITY 1*/93 65.
73 B114 POINT BEACM 1 M77369 PORY AND SLOCK VALVE RELIABILITY G4/94 66.
70-8114 POINT SEACM 2 M77370 - PORY AND BLOCK VALVE RELIABILITY' 04/94 67.
70 s114 R0sINSON 2 M77373 ' PORY AND BLOCK VALVE RELIABILITY 10/93 68.
70.
3114 SALEM 1 M77374 PORY AND SLOCK VALVE RELIABILITY-05/94
- 69., 70 '
B114 SALEM 2 M77375 PORY AND SLOCK VALVE RELIABILITY' 05/94 70.
70 5114 ' SouTN TEXAS 1 M77380.PORY AND BLOCK VALVE RELIABILITY.-
10/93
I f
71.
70 5114-SOUTH TEXAS 2 M77381 PORV AND BLOCK VALVE RELIABILITY 10/93 72.
70 site SUIeER 1 M77384 PORY AND BLOCK VALVE RELIABILITY 01/96 73.
70 8114 - SURRY 1 M77385 PORV AND BLOCK VALVE RELIABILITY 12/93 74.
70 5114 SURRY 2 M77386 PORY AND BLOCK VALVE RELIABILITY 12/93 75.
70 3114 TURKEY POINT 3 M77389 PORY AND BLOCK VALVE RELIABILITY 12/93 76.
70 8114.. TURKEY POINT 4 M77390 PORY AND BLOCK VALVE RELIABILITY ~
- 12/93 77.
70 5114 ZION 1 M77397 PORY AND BLOCK VALVE RELIABILITY 12/93 78.
70 a114. - ZION 2' M77398 PORY AND BLOCK VALVE RELIASILITY
.12/93 79.
75 (3085) 3085 mRolars FERRY 1 -
M53571 SALEN ATUS 1,2 DATA CAPASILITY 07/97 80.
75 (8085) 3085 BRohmes FERRY 2 M53572 SALEM ATWS 1.2 DATA CAPA8ILITY -
10/93 81.
75 (8085)-
t085 mRolaIS FERRY 3
- M53573-SALEM ATWS 1.2 DATA CAPA81LITY 06/95
' 82.'
75 (8090) 3090; CRYSTAL RIVER 3 M55351 SALEM ATWS 4.3 v Als stW T.S.
10/93
- 83.
75 (3091):
3091 CRYSTAL RIVER 3 M53953 SALEM ATWS 4.4 B&W TEST PROCEDURES '
12/93 O / 84.
75 (8093) 3093 sRolaIS FERRY 1 MS3964 SALEM ATWS 4.5.2 & 4.5.3 TEST ALTERNATIVES 07/97 01 85. - 75 (8093) 8093 eRotat$ FERRY 3.
M53966 SALEM ATWS 4.5.2 & 4.5.3 TEST ALTERNATIVES ~
06/95 i
86.
96 5115-ARKANSAS 2 M77399 ADDITIONAL LOW-TEMP. 0VERPitESSURE PROTECTICII FOR LWS '
01/96 87.
94 5115 BEAVER VALLEY 1'.
M77400 A00lTIONAL LOW-TEMP. OWERPRESSURE PROTECTICII FOR LWits -
03/94 88.'
94 5115 OEAVER WALLEY 2 M77401 ; : ADDITIONAL LOW-TEMP. OVERPRESSURE PROTECTICII FOR LWRS :
' 03/96.
- 89.. ' 94 -
3115 - CALVERT CLIFFS 1~
M77407 I ADDITIONAL LOW-TEIW4 OYERPRESSURE PROTECTION FOR LWRS
- 12/93
+
8115-CALVERT CLIFFS 2
- M77408 ADDITIONAL LOW-TEMP. OVERPRES$URE PROTECTION FOR LWits 12/93
.90.
94
.91.
94 5115-COOK 1 M77413 ~ ADDITIONAL LOW-TEMP. OVERMtESSURE PROTECTION FOR LWRS -
11/93
-l 92..
96 3115" C00K'2'
'M77414. ADDITIONAL LOW-TEMP. OWERPRESSURE PROTECTICII F0lt LURS 11/93 93.
94 B115 FARLEY 1'
' M77419 ? ~ ADDITIONAL LOW-TEMP. OWERPRESSURE PROTECTICII FOR LWRS -
12/93-96.
96 8115 FARLEY 2 M77420- ADDITICIIAL LOW-TEMP. OWEllPRESSURE PROTECTICII FOR LWR $
'12/93
- 95. : 96 3115 FORT CALMOUN 1-
-MT7421-ADDITIONAL LOW-TEMP. OWERMtESSURE PROTECTION FOR LWR $ '
12/93 94 8115 GIINIA M77422 ~ ' ADDITIONAL Lol6 TEMP. OWEltPRESSURE PROTECTION FOIt LWRS-01/96 97.
96 5115 IIADOAM NECK '
M77423 ADDITIONAL LOW-TE M. OYERPitES$URE PROTECTION FOR LWRS
- 10/93'
- 98. 1 94 5115' I mIAll POINT 2
- M77425 : AmelTICIIAL LOW-TEMP. OWERMIESSURE PROTECTICII FOR LWRS.
01/96 s
99.
96 -
3115..ImIAN POINT 3 M77426 AeoITIONAL LOW-TEMP. OVERMIESSURE PROTECTICII FOR LWits-112/931 100.
96 5115' REWUNEE MT7427 - ADDITIONAL LOW-TEMP. OVERPREStultE PROTECTION FOR LWIts ?
- 12/93 101..
96 3115. MAINE YANREE -
477428-. ACOITICIIAL LOW-TEMP. OWEllMIESSURE PROTECTION FOR LWRS 01/94 102.. 94 3115' MCWIRE 1-MT7429 ' A001TIONAL LOW-TEMP. OWERMIESSURE Mt0TECTION Felt LWits J 01/96 103.
96 3115 MCEUIRE 2-M77430 ADDITIONAL LOW-TEMP. OWERPREssuRE PROTECTIGN FOR LWRS.
- 01/96
-106.
94
-3115 MILLSTONE 2 MT7431 ADDITIONAL. LOW-TEW. OWEltPRESSURE PROTECTICII Folt LWIts -
'10/93-
-105.:.'96.'
3115 -' L IIORTu ANNA 1 M77433 ADDITICIIAL' LOW-TEMP. OWEllMIESSURE PROTECTION FOR LWRS
'12/93
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/ / / / / / / / / / /
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141.
GL-89-13 L913 CALVERT CLIFFS 1 M73978 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPENT 12/94 142.
GL-89-13 L913 CALVERT CLIFFS 2 M73979 SERVICE M TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 12/94 143.
GL-89-13 L913 CATAWBA 1 M73980 SERVIE ETER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 10/93 144.
GL-89-13 L913 CATAWBA 2 M73981 SERVICE WTER STSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT
'10/93 145.
GL-89-13 L913 DRESDEN 3 M73996 SERVICE ETER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 10/93 146.
GL-89-13 L913 GINNA M74007 SERVI E ETER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 12/93 147.
GL-89-13 L913 MAINE YANKEE M74022 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 12/93 148. - GL-89-13 L913 MCGUIRE 1 M74023 SERVICE W TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 08/94 149.
GL-89-13 L913 MCGUIRE 2 M74024 SERVICE M TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 08/94
.150.
GL-89 L913 MILLSTONE 1 M74025 SERVICE W TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 02/94 151.
GL-89-13 L913 MILLSTONE 3 M74027 SERVICE W TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EGUIPE NT 10/93 152.
GL-89-13 L913 PERRY 1 M74043 SERVICE W TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 12/93 153.
GL-89-13 L913 euAD CITIES 2 M74050 SERVICE M TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EGUIPMENT' 10/93 154 GL-89-13 L913, WOLF CREEK 1 M74088 SERVICE W TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 10/93 g
y 155.
L913 ZION 1 M74090 SERVICE W TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 10/93 156.
GL-89-13 ^
L9131 ZION 2 M74091-SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 10/93 157.
MPA-0023 B023 CRYSTAL RIVER 3 M10017. DEGRASED GRID VOLTAGE -
12/93 158.
MPA-5023
-3023 MILLSTONE 1 M60207 DEGRADED GRID VOLTAGE 10/93
i Appendix D I
LISTING OF OTHER UNIMPLEMENTED MPA ITEMS BYISSUE I
4 i
i 4
i.
i 4
APPENDIX D i
This appendix provides a detailed list, by issue, of the 941 MPA items not implemented, along with the projected date for completing the item. Status and projected implementation dates are presented as of September 30,1993.
1 4
i 4
i l
1 4
1 l
1 i
1
)
l i
?
l i
(
D-1 i,
. ~,
Other Multi-Plant Actions (Listin8 of Open Itans)
IWL ISSUE MPA PLANT TAC TITLE DATE 1.
BL-88-02 X802 COOK 1 M67301 STEAM GENERATOR TUBE RUPTURE (888-02) (OLD MPA 8099) 06/94 2.
BL-88-03 X803 BROWNS FERLv 1 M73852 CE NFA RELAYS (888-03) 07/97 3.
8L-88-03 X803 SROWNS FERRY 3 M73854 GE HFA RELAYS (888-03) 06/95 4.
8L-88-04 X804 BROWNS FERRY 1 M69883 $! PLMP FAILURE (888-04) (OLD WA 8103) 07/97 5.
BL-88-04 X804 BROWNS FERkY 3 M69890 SI PUW FAILURE (888-04) (OLD MPA 8103) 06/95 6.
BL-88-04 X804 NINE MILE POINT 1 M69940 St PUMP FAILURE (888-04) (OLD MPA 8103) 03/95 7.
BL-88-07 X807 SROWNS FERftY 1 M72805 POWER OSCILLATIONS IN 8 tnt's (888-07) 07/97 8.
8L-88-07 X807 BROWNS FERRY 3 M72769 POWER OSCILLATIONS IN BWR'S (888-07) 06/95 9.
8L-88-08 X808 BRAIDWO(x) 1 M69602 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA 8107) 04/94 10.
BL-88-08 X808 SRAIDWOOD 2 M69603 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD W A 8107) 10/93 11.
BL 88-08 X808 BYRON 1 M69609 THERMAL STRESS IN PIPING CIntNECTED TO RCS (888-08)(OLD MPA 8107) 10/93 12.
BL-88-08 X808 BYRON 2 M6%10 THERMAL STRESS IN PIP!NG CONNECTED TO RCS (888-08)(OLD MPA B107) 10/93 13.
8L-88-08 X808 COOK 1 M6%18 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA $107) 12/93 O 14 BL-88-08 X808 COOK 2 M69619 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA B107) 12/93 b 15.
BL-88-08 X808 KEWAUNEE M69643 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA 8107)
/
16.
BL-88-08 X808 NINE MILE IMINT 1 M6%55 THERMAL STRESS IN PIPING CONNECTED TO RCS (8 4 08)(OLD W A 8107) 03/95 17.
8L-88-08 X808 PALO VERDE 1 M69664 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA 8107)
/
18.- BL-88-08 X808 PALO VERDE 2 M69665 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA 8107)
/
19.
BL-88-08 X808 PALO VERDE 3 M69666 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA 3107)
/
20.
BL-88-08 X808 SALEM 1 M69680 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA 8107) 12/93 21.
8L-88-08 X808 SALEM 2 M69681 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA 8107) 12/93 22.
BL-88-08 X808 SOUTH TEXAS 1 M69689 THERMAL STRESS IN PIPING CONNECTED TO RCS (888-08)(OLD MPA 8107) 12/93 23.
BL-88-10 X810 POINT BEACH 1 M71338 NONCONFORMING MOLDED-CASE CIRCUIT BREAKERS (888-10) 12/93 24.
BL-88-10 X810 POINT BEAC:: 2 M71339 NONCONFORMING MOLDED-CASE CIRCUIT BREAKERS (888-10) 12/93 25.
BL-88-11 X811 ARKANSAS 2 M72109 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 10/93 26.
BL-88-11 X811 BRAIDWOOD 2 M72115 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 10/93 27.
8L-88-11 X811 BYRON 1 M72116 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 10/93 28.
BL-88-11 X811 STRON 2 M72117 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 10/93 29.
BL-88-11 X811 CALVERT CL:7FS 1 M72119 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 02/94 30.
BL-88-11 X811 CALVERT CL NFS 2 M72120 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 02/94 31.
BL-88-11 X811 CRYSTAL RitER 3 M72127 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 10/93
' 32.
BL-88-11 X811 DAVIS-SESSr 1 M72128 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 12/94 33.
BL-88-11 X811 FORT CALHOGI 1 h D 34 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 12/93 34.
8L-88-11 X811 MAINE YANKFE M72141 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 12/93 35.
BL-88-11 X811 OCONEE 1 M72148 THERMAL STRATIFICATI0ld IN PZR SURGE LINE (888-11) 12/93
36.
BL-88-11 X811 OCONEE 2 M72149 THERMAL STRATIFICATION IN PZR SURGE LINE (B88-11) 12/93 37.
BL-88-11 X811 OCONEE 3 M72150 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 12/93 38.
BL-88-11 X811 PALISADES, M72151 THERMAL STRATIFICATION IN PZR SURiaE L!!E (B88-11) 11/93
~
39.
BL-88-11 X811 PALO VERDE 1 M72152 THERMAL STRATIFICATION IN PZR SURGE LINE (888-11) 12/93 40.
BL-88-11 X811 PALO VERDE 2 M72153 THERMAL STRATIFICATION IN P2R SURGE LINE (888-11) 12/93 41.
BL-88-11 X811 PALO VERDE 3 M72154 THERMAL STRATIFICATION 1N PZR SURGE LINE (888-11) 12/93 42.
8L-88-11 X811 POINT BEACF 1 M72155 THERMAL STRATIFICATIGN IN PZR SURGE LINE (B88-11) 10/93 43.
BL-88-11 X811 POINT BEACH 2 M72156 THERMAL STRATIFICMION IN PZR SURGE LINE (888-11) 10/93 44.
8L-88-11 X811 ROBINSON 2 M72160 THERMAL STRATUICATION IN PZR SURGE LINE (888-11) 01/94 45.
BL-88-11 X811 SALEM 1 M72161 THERMAL STPATIFICATION IN PZR SURGE LINE (B88-11) 12/93 46.
BL-88-11 X811 SAN ONOFRE 2 M72164 TRERMAL STRATIFICATION IN PZR SURGE LINE (B88-11) 11/93 47.
BL-88-11 X811 SAM ONOFRE 3 M72163 THEVAAL STRATIFICATION iN PZR SURGE LINE (888-11) 11/93 48.
BL-88-11 X811 SEQUOYAN 2 M72167 NERMAL STRATIFICATION IN PZR SURGE LINE (B88-11) 04/94 O 49.
BL-88-11 X811 THREE MILE !SLAND 1 M72174 THERMAL STRATIFICATION IN PZR SURGE LINE (B88-11) 12/93 h 50.
BL-88-11 X811 WATERFORD 3 M72180 THERMAL STRATIFICATION IN PZR SURGE LINE (888 11) 12/93 51.
BL-89-01 X901 COOK 1 M73164 FAILURE OF WEST STEAM GENERATOR TUBE MECHANICAL PLUGS (889-01) 12/94 52.
sL-89-02 X902 PRAIRIE ISLAND 1 M74298 STRESS CORR CRACKING OF ANCHOR DARLING CK VALVE BOLTING (E89-02) 05/94 53.
BL-89-02 X902 PRAIRIE ISLAND 2 M74299 STRESS CORR CRACKING OF ANCHOR DARLING CK VALVE BOLT!NG (889-02) 11/93 54.
BL-89-02 5902 TURKEY P0llti 3 M74325 STRESS CORR CRACKING OF ANCHOR DARLING CK VALVE BOLTING (s89-02) 06/94 55.
BL-92-01 X201 ARKANSAS 1 M83839 THERMO-LAG (BULLETIN 92-01)
/
56.
BL-92-01 X201 ARKANSAS 2 MS3840 THERMO-LAG (BULLETIN 92-01) 06/95 57.
BL-92-01 X201 BEAVER VALLEY 1 M83841 THERMO-LAG (BULLETIN 92-01) 12/99 58.
BL-92-01 X201 BEAVER VAtlCY 2 M83842 THERMO-LAG (BULLETIN 92-01) 12/99 59.
BL-92-01 X201 BRAIDWOOD 1 M33846 THERMO-LAG (BULLETIN 92-01)
/
60.
8L-92 01 X201 BRAIDWOOD 2 M83847 THERMO-LAG (BULLETIN 92-01)
/-
61.
BL-92-01 X201 BROWNS FERRY 1 M83848 THERMO-LAG (8ULLETIN 92-01) 07/97 62.
BL-92-01
.X201 BROWNS FERRY 2 M83849 THERMO-LAG (BULLETIN 92-01) 12/93 63.
BL-92-01 X201 BROWNS FERa? 3
'M83850 THERMO-LAG (BULLETIN 92-01) 03/9'.
64 BL-92-01 X201 BYRON 1 M83853 THERMO-LAG (BULLETIN 92-01) f 65.
sL-92-01 X201 BrRON 2 M83854 THERMO-LAG (8ULLETlu 92-01)
/
66.
SL-92-01 X201 CLINTON 1 M83860 THERMO-LAG (BULLETIN 92-01) 06/94 67 BL-92-01 X201 COMAEiiE SHK 1 M83861 THERMO-LAG (BULLETIN 92-01) 12/93
- 68. 'BL-92-01 X201 COOK 1 M83863 THERMO-LAG (8ULLETIN 92-01) 12/93 69.
BL-92-01 X201 C00K 2 M83864 THERMO-LAG (80LLETIN 92-01) 12/93 70.
BL-92-01 X201 DAVIS-8 ESSE 1 M83867 THO M0-LAG (BULLETIN 92-01)'
12/94
71.
BL-92-01 x201 DUANE ARNOLD M83872 THERMO-LAG (8ULLETIN 92-01) 12/93 72.
BL-92-01 X201 FERMI 2 M83875 THERMO-LAG (BULLETIN 92-01) 12/93 73.
BL-92-01 X201 GRAND GULF 1 M83879 THERMO-LAG (BULLETIN 92-01)~
12/94 74.
BL-92-01 X201 HARRIS 1 M83881 THERMO-LAG (BULLETIN 92-01) 12/93 75.
BL-92-01 X201 MATCH 1 M83832 THERMO-LAG (BULLETIN 92-01) 12/95 76.
BL-92-01 x201 HATCH 2 M83883 THERMO-LAG (BULLETIN 92-01) 12/94 77.
BL-92-01 X201 INDIAN POINT 2 M83885 THERMO-LAG (BULLETIN 92-01) 12/93 78.
BL-92-01 X201 LASALLE 1 M83888 THERMO-LAG (8ULLETIN 92-01)
/
79.
BL-92-01 X201 LASALLE 2 M83889 THERMO-LAG (BULLETIN 92-01)
/
80.
BL-92-01 x201 MCGUIRE 1 M83893 THERM 0-LAG (BULLETIN 92-01)
/
81.
BL-92-01 X201 MCGUIRE 2 M83894 THERMO-LAG (8ULLETIN 92-01)
/
82.
BL-92-01 x201 MILLSTONE 1 M83895 THERMO-LAG (8ULLETIN 92-01) 06/94 83.
BL-92-01 X201 NINE MILE Po!NT 1 M83899 THERMO-LAG (BULLETIN 92-01) 12/93 0 84.
BL-92-01 x201 NINE MILE POINT 2 M83900 THERMO-LAG (BULLETIN 92-01) 12/93 On 85.
BL-92-01 X201 NORTH ANNA 1 M83901 THERMO-LAG (BULLETIN 92-01) 12/93 86.
8L-92-01 X201 NORTH ANNA 2 M83902 THERMO-LAG (8ULLETIN 92-01) 12/93 87.
BL-92-01 X201' -PALISADES' M83907 THERMO-LAG (SULLETIN 92-01)
/
88.
BL-92-01 x201 PALO VERDE 1 M83908 THERMO-LAG (8ULLETIN 92-01)
/
89.
BL-92-01 x201 PALO VERDE 2 M83909 -THERMO-LAG (BULLITIN 92-01)
/
90.
BL-92-01 X201 PALC VE N ',
M83910 THERMO-LAG (BULLETIN 92-01)
/
91.
BL-92-01
.x201 PEACM SOTTOM 2 M83911 THERMO-LAG (BULLETIN 92-01) 12/94 92.
BL-92-01 x201 PEACH 90TTOM 3 M83912 THERMO-LAG (BULLETIN 92-01) 12/94 93.
8L-92-01 x201 PRAIRIE ISLANO 1 M83917 THERM 0-LAG (BULLETIN 92-01)
/
94.
BL-92-01 x201 PRAIRIE ISLAse 2 M83918 THERMO-LAG (BULLETIN 92-01)
/
95.
BL-92-01 x201 SAN ONOFRE 2 M83928 THERMO-LAG (90LLETIN 92-01)
/
96.
BL-92-01 x201 SAN ONOFRE 3 M83929 THERM 0-LAG (SULLETIN 92-01)
/
97.
BL-92-01 X201 SURRT 1 M83936 THERMO-LAG (BULLETIN 92-01) 12/93
. 98.
E-92-01
.X201' SURRY 2 M83937 THERMD-LAG (SULLETIN 92-01) 12/93 99.
BL-92-01 X201 SuseUENANNA 1 M83938 THERM 0-LAG (suLLETIN 92-01) 12/93 100.
E-92-01 x201 suseUENAMMA 2 M83939 THERMO-LAG (BULLETIN 92-01) 12/93 101.
K-92-01 x201 THREE MILE ISLAND 1' M83940.THERMO-LAG (00LLETIN 92-01) 12/93 102.
K-92 x201 VERMONT YANKEE 1 M83944-JNERMD-LAG (EULLETIN 92-01) 12/93 103.
E-92-01 X201 VOCTLE 1 M83945 THERIO-LAG (BULLETIN 92-01) 12/93 104.
K-92-01 X201 V0GTLE 2 M83966 THERMO-LAG (BULLETIN 92-01) 12/93 105.
E-92-01 x201 WATERFORD 3 M83967 THERMD-LAG (BULLETIN 92-01)
/
- ~
106.
BL-92-01 X201 WOLF CREEK 1 M83951 THERMO-LAG (BULLETIN 92-01)
/
107.
BL-93-02 1302 ARKANSAS 1 A86526 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINER $
/
108.
BL-93-02 X302 BEAVER VALLEY 1
- t36528 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCT40N STRAINERS 12/93 109.
BL-93-02 X302 BEAVER VALLEY 2 f486529 DEBRIS PLUGGINC OF EMERGENCY CORE COOLING SUCTION STRAINERS 12/93 110.
BL-93-02 X302 BIG ROCK PotNT 1 FS6532 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
it).
BL-93-02 x302 BRAIDWOCD 1 M6533 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINER $
/
112.
BL-93-02 x302 BRAIDWOOD 2 PC6534 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINER $
/
113.
BL-93-02 x302 BROWNS. FERRY 1 P46535 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 07/97 114.
BL-93-02 x302 BROWNS FERRY 3 M6537 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 06/95 115.
BL-93-02 x302 BYRON 1 M%540 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
116.
BL-93-02 F302 BYRON 2 MO6541 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
117.
BL-93-02 x302 CALLAWAY 1 Mt6542 DEBRIS PLUGGING OF EMERENCY CORE COOLING SUCTION STRAINERS
/
118.
BL-93-02 X302 CATAWBA 1 Mf4545 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 12/93 O 119.
BL-93-02 X302 CATAWBA 2 ME5546 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRA!NERS 12/93 b 120.
BL-93-02 x302 COOK 1 M66550 DEBRIS PLUGGING OF EMERENCY CORE COOLING SUCTION STRAINERS 12/93 121.
BL-93-02 x302 COOK 2 MM551 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 12/93 122.
BL-93-02 x302 CRYSTAL RIVER 3 M84553 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
123.
BL-93-02 x302 FARLEY 1 M85360 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
124.
BL-93-02 x302 FARLEY 2 M86561-DEBRIS PLUGGING CF EMERGENCY CORE COOLING SUCTION STRAINER $
/
125.
BL-93-02 x302 GINNA M85565 DEBRIS PLUGGING OF EMERENCY CORE COOLING SUCTION STRAINERS
/
126.
BL-93-02 x302 NOPE CREEK 1 M86571 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 12/93 127.
BL-93-02 x302 LASALLE 1 MS 575 DEBRIS PLUGGING OF EMERENCY CORE COOLING SUCTION STRAINERS
/
128.
BL-93-02 X302 LASALLE 2 M85576 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION ITidINFr.1
/
129.
BL-93-02 x302 LIMERICK 1 MM577 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTI0tt S?RAINERS 12/93 130.
BL-93-02 x302 LIMERICK 2 MMSM DEBRIS PLUGGING OF EMCRGENCY CORE COOLING SUCTION STRAINERS 12/93 131.
BL-93-02.
x302 MCGUIRE 1 M66580 DEBRIS PLUGGING OF EMERENCY CORE C00 LIV O lau STRAINERS 12/93 132.
BL-93-02 x302 MCGUIRE 2 MMSE1 DEBRIS PLUGGING CF EMERGENCY CORE C00L13 9 410N STRAINERS 12/93 133.
BL-93-02 x302 PALISADES M865% DEBRIS PLUGGtWG OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
134.
BL-93-02 X302 PALO VERDE 1 MS6595 CEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
135.
BL-93-02 x302 PALO VERDE 2 M365% DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
136.
BL-93-02 x302 PALO YERDE 3 M86597 DEBRIS PLUGGING OF EMERENCY CORE COOLING EUCTION STRAINERS
/
137.
BL-93-02 x302 PEACH BOTTOM 2 MS659E-DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 10/93 138.
BL-13-02 X302 PEACH BOTTOM 3 pa6599 DEBRIS PLUGGING Gr EMERGENCY CORE COOLING SUCTION STRAINERS 10/93 139.
BL-93-02 x302 PERRY 1 R4600 DEBRIS PLUGGING OF EMERENCY CORE COOLING SUCTION STRAINERS
/
140.
BL-93-02 x302 PILGRIM 1 t46601 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTICK STRAINER $
12/93
ss 141.
SL-93-02 X302 POINT BEACM 1 M86602 DEBRIS PLUGGING OF EMERENCY CORE C00 LING SUCTION STRAINERS
/
142.
BL-93-02 X302 POINT BEACH 2 M86603 OEBRIS PLUGGING OF EMERENCY CORE COOLING SUCTION STRAINERS
/
143.
BL-e3-02 X302 ROSINSON 2 M8M09 tEBRIS PLUGGING OF EMERENCY CORE COOLING SUCTION STRAINERS 11/93 144.
BL-93-02 X302 SALEM 1 M86610 CEsRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 12/93 145.
BL-93-02 X302 SALEM 2 M86611 DIBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 12/93 146.
BL-93-02 X302 SAN ONOFRE 2 M86612 DI8RIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
147.
BL 93-02 X302 SAN ONOFRE 3 M86613 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
148.
BL-93-02 X302 SUSQUENANNA 1 M86624 DIBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINER $
05/94 149.
BL-93-02 X302 SUSQUENANNA 2 M86625 DEBRIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 05/94 150.
8L-93-02 X302 YERMONT YANKEE 1 M86629 DE3RIS PLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS 12/93 151.
BL-93-02 X302 WATERFORD 3 M8M32 DEERIS FLUGGING OF EMERGENCY CORE COOLING SUCTION STRAINERS
/
152.
BL-93-03 X303 BROWNS FERRY 1 M86882 VELSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 07/97 153.
BL-93-03
'X303 BROWNS FERRY 2 M86883 VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 12/94 O 154 BL-93-03 X303 BROWNS FERRY 3 M86884 VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 06/95 N 155.
BL-93-03 X303 BRUNSWICK 1 M86885 VESSEL WATER LEVEL INSTRUMENTATION IN 8WR'S: RES. OF ISSUES
/
156.
BL-93-03 X303 BRUNSWICK 2 M86886 VES!EL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES
/
157.
BL-93-03 X303 CLINTON 1 M86887 VESSIL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 11/93 i
158.
BL-93-03 X303 COOPER STATION M86888 VESSCL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES
/
159.
BL-93-03 X303 DRESDEN 2 M86889 VFSSEL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 09/94 160.
BL-93-03 X303 DRESDEN 3 M86890 VCSSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 03/94 161.
8L-93-03 X303 DUANE ARNOLD M86891 VESSE1 WATER LEVEL INSTRUMENTATION IN SWRaS: RES. OF ISSUES 10/93 162.
BL-93-03 X303 FERMI 2 M86892 - VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 04/94 163.
BL-93-03 X303 GRAND GULF 1 M86893 VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 12/93 164.
BL-93-03 X303 NATCN 1 M86894 VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 12/95
.165.. BL-95-03 X303 NATCM 2 M86895 VESSEL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 12/95 166.
8L-93-03 X303 h0PE CREEK 1 M86896 VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF. ISSUES 03/94 167.
BL-93-03 X303 LASALLE 1 M86897 VESSEL WATER LEVEL INSTRtmENTATION IN SWR'S: RES. OF ISSUES
/
168.
SL-93-03 X303-LASALLE 2 M86898 VESSEL WATER LEVEL INSTRtmENTATION IN SWR'S: RES. OF ISSUES
/
169.
BL-93-03 X303 LIMERICK 1 M86899 VESSEL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 12/93 170.
BL-93-03 X303 LIERICK 2 M86900 VESSEL bATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 12/93 171. - SL-93-03 X303 MONTICELLO M86901-VESSEL b4TER LEVEL INSTRUMENTATION IN tlat*S: RES. OF ISSUES 11/94 172.
BL-93-03 '
X303 NINE MILE POINT 1-M86902 VESSEL WATER LEVEL INSTRUENTATION IN Blat'S: RES. OF ISSUES 03/95 1 73.
BL-93-03 X303 'NINE MILE Po!NT 2' M86903 VESSEL WhTER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 12/93 1 74.
BL-93-03 X303 OYSTER CREEK 1 M86904 WESSEL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 11/94 175.
sL-93-03 X303 PEACM 90TTOM 2 M86905 VESSEL WATER LEVEL INSTRLMENTATION IN SWR'S: RES. OF ISSUES 12/93
l i.
r 176.
BL-93-03 X303' PEACM BOTTOM 3 M86906 VESSEL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 12/93 177.
BL-93-03 X303 PERRY 1 M86907 VESSEL WATEF LEVEL INSTRUMENTATION IN SWR 8S: RES. OF ISSUES 10/93 178.
BL-93-03 X303 QUAD CITIES 1 M86909 VESSEL WATER LEVEL INS'TRUMENTATION IN BWR'S RES. OF ISSUES
/
179.
BL-93-03 X303 QUAD CITIES 2 M86910 VESSEL WATEft LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES
/
180.
BL-93-03 X303 RIVER BEND 1 M86911 VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF' ISSUES
/
181.
BL-93-03 X303 SUSQUEHANNA 1 M86912 VESSEL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 12/93 182.
BL-93-03 X303 SUSQUENANNA 2 M86913 VESSEL WATER LEVEL INSTRUMENTATION IN SWR'S: RES. OF ISSUES 05/94 183.
BL-93-03 X303 VERMONT YAKEE 1 M86914 VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES 11/93 184.
BL-93-03 X303 WASHINGTON NUCLEAR 2 M86915 VESSEL WATER LEVEL INSTRUMENTATION IN BWR'S: RES. OF ISSUES
/
185.
GL-83-08 0020 BROWNS FERRY 1 M57144 MARK I DRYWELL VACUUM BREAKERS (GL83-08) 07/97 186.
GL-84-09 A019 DRESDEN 2 M56579 RECOMBINER CAPABILITY REQUIREMENTS OF 10 CFR 'J.44 (GL84-09) 12/97 187..
GL-84-09 A019 DRESDEN 3 M56580 RECOMBINER CAPABILITY REQUIREMENTS OF 10 CFr 50.44 (GL84-09) 12/97 188.
GL-84-09 A019 MILLSTONE 1 M65067 RECOMBINER CAPABILITY REQUIREMENTS OF 10 r.R 50.44 (GL84-09) 02/94
. g 189.
'GL-84-09 A019 OYSTER CREEC 1 M62980.RECOMBINER CAPABILITY REQUIREMENTS OF 10.FR 50.44 (GL84-09) 11/93
- (n 190.
GL-84-09 A019 0UA0 CITIES 1 M55148 RECOMBINER CAPABILITY REQUIREMENTS OF 10 ;FR 50.44 (GL84-09) 09/95
,191.
GL-84-09 A019 cuaD CITIES 2 M55149 RECOMBINER CAPABILITY REQUIREMENTS OF iu CFR 50.44 (GL84-09) 12/94-192.
GL-87-09 D024 BRUNSWICK 1 M64910 MODE CHAGES AND LCO'S -.TECK SPECS 3.0 AND 4.0 (GL 87-09)
/
193.
GL-87-09 D024 BRUNSV!CK 2 M64911 MODE CHAGES AND LCO'S - TECH SPECS 3.0 AND 4.0 (GL 87-09)
/
194.
CL-87-09 D024 - LASALLE 1
- M75789 - MODE CHACES AND LCO'S - TECM SPECS 3.0 AND 4.0 (GL 87-09) 10/93
~
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GL-87-09 0024 LASALLE 2 M75790. MODE CHAGES AND LCO'S ~ TECM SPECS 3.0 AND 4.0 (GL 87-09) 10/93 196 GL-88-01 5097-DRESDEN 2 M69132 IGSSCC PROBLEMS IN SWR PIPING 11/94 197.
GL-88-01 8097 DRESDEN 3-M69133 IGSSCC PR08LEMS IN SWR PIPING 11/94 198.
GL-88-01 8097 PERRY 1 M69152 IGSSCC PROBLEMS IN BWR PIPING 12/93~
199.
.CL-88-01 8097 QUAD CITIES 1 M69154-IGSSCC PROSLEMS IN SWR PIPING 11/95-2001 GL-88-01 8097 QUAD CITIES 2-
'M69155 IGSSCC PROBLEMS IN BWR PIPING-11/9?.
201.
GL-88-01 9097 SUseUENANNA 1 M69158: IGSSCC PROBLEMS IN SWR PIPING 11/93
.202.
GL-88-01 8097 SUSQUENANNA 2 M69159 ICSSCC PROBLEMS IN BWR PIPING
-11/93 203.
GL-88-01 B097. VERMONT YANKEE 1
'M69160 IGSSCC PROBLEMS IN SWR PIPING 1 12/93 204.
GL-88-01 3097 - WASHINGTON NUCLEAR 2 M69161 ICSSCC PROBLEMS IN BWR PIPING.'
04/94 205. -GL-88-11 A023: SEAVER VALLEY 2
. M71463 R.G.1.99 REV 2 (PRESSURIZED THERMAL SNOCK RULE) (GL 88-11) -
/
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GL-88 A023-sROWits FERRY 3
.M71469 R.G. 1.99 REV 2 (PRESSURIZED THERMAL SNOCK RULE) (GL 88-11) 06/95 208.
GL-88-11 A023 SAN ONorRE 2 M71546-R.G. 1.99 REV 2 (P9ESSURIZED THERMAL SNOCK RULE) (CL 88-11).
/
209.
GL-88-11 A023 SAN ONOFRE 3.
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/
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GL-88-12 10022 f. eRUNSWICK 1 -
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/.
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GL-88-12 0022 BRUNSWICK 2 M79417 REMOVAL OF FIRE PROTECTION TECM SPECS (GL 88-12) 12/93 212.
GL-88-12 D022 FORT CALMOUN 1 M87825 REMOVAL OF FIRE PROTECTION TECM SPECS (GL 88-12) 05/94 213.
GL-88-20 sitt ARKANSAS 1 M74376 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
214.
GL-88-20 8111 ARKANSAS 2 M74377 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 03/94 215.
GL-88-20 8111 BEAVER VALLEY 1 M74378 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 04/95 216.
GL-88-20 8111 81G RO M POINT.1 M74381 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
217.
GL-88-20 sitt sRAIDWOOD 1 M74382 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
218.
GL-88-20 s111 BRAIDWOOD 2 M74383 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
219.
GL-88-20 B111 BROWNS FERRY 1 M74384 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
220.
GL-88-20 8111 BROWNS FERRY 2 M74385 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 11/93 221.
CL-88-20 8111 BROWNS FERRY 3 M74386 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
222.
GL-88-20 sitt BRUNSWICK 1 M74387 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 02/94 223.
GL-88-20 8111 BRUNSWICK 2 M74388 INDIVIDUAL PLANT EVALUAT!0NS (GL 88-20) 02/94 224.
GL-88-20 B111 8YRON 1 M74389 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)-
04/94 g
g 225.
GL-88-20 sit 1 BYRON 2 M74390 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 04/94 226.
GL-88-20 8111 CALLAunY 1 M74391 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 227.
GL-88-20 8111 CALVERT CLIFFS 1 M74392 INDIVIDUAL PLANT EVALUATIONS (CL 88-20)
/
228.
GL-88-20 8111 CALVERT CLIFFS 2 M74393 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
229.
GL-88-20 Bill CATAWBA 1 M74394 ' INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 230.
GL-88-20 8111 CATAWBA 2 M74395 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 231.
GL-88-20 8111 CLINTON 1 M74396 INDIVfDUAL PLANT EVALUATIONS (GL 88-20) 12/93 232.
GL-88-20 B111 COMANCHE PEAK 1 M74397 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 09/94 233.
GL-88-20_
B111 C00K-1 M74398 INDIVIDUAL PLANT. EVALUATIONS (GL 88-20) 10/93 234. - GL-88 8111 COOK 2 M74399 INDIVIDUAL PLANT EVALUATl0NS (CL 88-20) 10/93-235.
GL-88-20 Bill COOPER STATION M74400 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93
-236.
Gt-88-20 8111 CRYSTAL RIVER 3 M74401 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/94 237.
GL-88-20 8111 DAVIS-BESSE 1 M74402 INDIVIDUAL PLANT EVALUAT10NS (GL 88-203' 12/94 238.
CL-88-20 8111. DRESDEN 2 M74405 INDIVIDUAL PLANT. EVALUATIONS (GL 88-20)
/
239.
GL-88-20 8111 DRESDEN 3 M74406 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
240.
GL-88-20 8111 DUANE ARNOLD M74407 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
241.
GL-88-20 8111 FARLEY 1
'M74408 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
242.
GL-88-20 Bill FARLEY Z M74409 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
./
l 243.
GL-88-20 8111 FERMI 2 M74410 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 06/94 l
244.
GL-88-20 Bill FIT 2 PATRICK M74411 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 10/93 245.
_B111 FORT CALMOUN 1 M74412 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
l 246.
GL-38-20 8111 GINNA M74414 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 01/95 247.
GL-88-20 B111 GRAND GULF 1 M74415 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 248.
GL-88-20 B111 HADDAM NECK M74417 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 06/94 249.
GL-88-20 8111 HARRIS 1 M74418 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 08/94 250.
GL-88-20 8111 MATCN 1 M74419 INDIVIDUAL PLANT EVALUAfl0NS (GL 88-20) 12/M 251.
GL-88-20 stil HATCH 2 M74420 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/96 252.
GL-88-20 8111 MOPE CREEK 1 M74421 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/94 253.
GL-88-20 8111 INDIAN POINT 2 M74422 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 01/94 254 GL-88-20 8111 INDIAN PolNT 3 M74423 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
255.
GL-88-20 B111 KEVAUNEE M74424 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
256.
GL-88-20 8111 LASALLE 1 M74425 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
257.
GL-88-20 Bill LA141E 2 M74426 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
258.
GL-88-20 B111 LIOKt 1 M74427 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 10/93 9 259.
GL-88-20 8111 LIMERItx 2 M74428 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 10/93
$260.
GL-88-20 B111 MAINE YANKEE M74429 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 08/94 261.
GL-88-20 B111 MCGUIRE 1 M74430 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 09/94 262.
GL-88-20 Bill MCGUIRE 2 M74431 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 09/94 263.
GL-88-20 B111 MILLSTCHE 1 M74432 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 10/93 264.
GL-88-20 B111 MILLSTONE 2 M74433 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
265.
GL-88-20 8111 MONTICELLO M74435 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93' 266.
GL-88-20 8111 NINE MILE POINT 1 M74436 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 07/94 267.
GL-88-20 B111 NINE MILE POINT 2 M74437 1N0lVIDUAL PLANT EVALUATIONS (GL 88-20) 11/93 268.
CL-88-20 8111 NORTH ANNA 1 M74438 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 269.
GL-88-20 8111 NORTN ANNA 2 M74439 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 2 70.
GL-88-20 B111 OCONEE 1 M74440 INDIVIDUAL PLANT EVALUATIONS (GL 68-20) 06/94 271.
GL-88 20 8111 OCONEE 2 M74441 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 06/94 272.
GL-88-20 8111 OCONEE 3 M74442 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 06/94 2 73.
GL-88-20 8111 OYSTER CREEK 1 M74443 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 274.
GL-88-20 B111 PALISADES M74444 INDIVIDUAL PLAMT EVALUATIONS (GL 88-20)
/
275.
GL-88-20 B111 PALO VERDE 1 M74445 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 10/93 276.
GL-88-20 8111 PALO VERDE 2 M74446 INDIVIDUAL PLANT EVALUATIONS (GL 58-20) 10/93 277.
GL-88-20 8111 PALO VERDE 3 M74447 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 10/93 278.
GL-88-20 B111 PEACN BOTTOM 2 M74445 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 279.
GL-88-20 Bill PEACN BOTTOM 3 M74449 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 280.
GL-88-20 8111 PERRY 1 M74450 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
281.
GL-88-20 8111 PILGRIM 1 M74451 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 08/95 282.
GL-88-20 8111 POINT BEACH 1 M74452 INDIVIDUAL PLANT EVALUATIONS (CL 88-20)
/
283.
GL-88-20 8111 POINT SEACM 2 M74453 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
284.
GL-88-20 8111 PRAIRIE ISLAND 1 M74454 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
285.
GL-88-20 8111 PRAIRIE ISLAND 2 M74455 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
286.
GL-88-20 8111 QUAD CITIES 1 M74456 INDIVIDUAL PLANT EVALUATIONS (CL 88-20)
/
287.
GL-88-20 8111 QUAD CITIES 2 M74457 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
288.
GL-88-20 8111 RIVER BEND 1 M74459 INDIVIDUAL PLANT EVALUATIONS (CL 88-20)
/
289.
GL-88-20 8111 ROBINSON 2 M74460 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 290.
GL-88-20 8111 SALEM 1 M74461 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 03/94 291.
GL-88-20 8111 SALEM 2 M74462 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 03/94 292.
Gt-88-20 8111 SAN ONOFRE 2 M74464 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
293.
GL-88-20 8111 SAN ONOFRE 3 M74465 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
9 294.
GL-88-20 8111 SEQUOTAN 1 M74468 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 08/94 d 295.
CL-88-20 8111 SEQUOTAN 2 M74469 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 08/94 296.
Stil SOUTN TEXAS 1 M74471 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 08/94 297.-
GL-88-20 8111 SOUTH TEMAS 2 M74472 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 08/94 298.
GL-88-20 8111' ST LUCIE 1 M74473 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
299.
'8111 ST LUCIE 2 M74474 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
300.
GL-88-20 8111 SUNNER 1 M74475 INDIVIDUAL PLANT EVALUATIONS (GL 88 20)
/
301.
GL-88-20 8111 SURRY 1 M74476. INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 302.
GL-88-20 8111-SURRY 2 M74477 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 303.
CL-88-20 8111 SUSJJENANNA 1 M74478 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 304.
GL-88-20 8111 SUSQUENANNA 2 -
M74479 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93 305.
- 8111 THREE MILE ISLAND 1 M74480 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/94 306.
GL-88-20 8111 VERMONT TANKEE 1 M74484 IWIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
307.
GL-88-20 8111 V0GTLE 1 M74485 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/93' 308.
GL-88-20 8111 V0GTLE 2 M74486. INDIVIDUAL PLANT EVALUATIONS (GL 88-20)-
12/93 309.
GL-88-20 8111 MASNINGTON NUCLEAR 2 M74489 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 06/94 310.
GL-88-20 8111 WATERFORD 3 M74487 INDIVIDUAL PLANT EVALUATIONS (GL 88-20)
/
311.
GL-88-20 8111 WOLF CREEK 1 M74490 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 12/94 312.
GL-88-20 8111 ZION 1 M74492 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 06/94 313.
GL-88-20 8111. ZION 2 M74493 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 06/94 314.
,0025 REA 'ER VALLET.1 M86770 RELOCATE RETS TO ADMIN. SECTION OF TECW SPECS (GL89-01) '
12/93 315.
GL-89-01 0025-REAVER VALLET 2 M86771 RELOCATE RETS TO ADMIN. SECTION OF TECM SPECS (GL89-01) 12/93
l.
i 316.
GL-89-01 D025 BROWitS FERRY 1 M83108 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GL89-01) 12/93 317.
GL-89-01 D025 BROWNS FERRY 2 M83109 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GL89-01) 12/93 318.
GL-89-01 0025 BROWNS FERRY 3 M83110 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GL89-01) 12/93 319.
CL--89-G1 D025 BRUNSWICK 1 M67946 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GL89-01) 12/93 320.
GL-89-01 0025 BRUNSWICK 2 M67947 RELOCATE RETS TO ADM!N. SECTION OF TECH SPECS (GL89-01) 12/93 321.
GL-89-01 0025 FARLEY 1 M84009 RELOCATE RETS TO ADMIN. SECTION OF TECH SPCCS (CL89-01) 01/94 322.
CL-89-01 D025 FARLEY 2 M84010 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GL89-01) 01/94 323.
GL-89-01 DC25 HARRIS 1 M84127 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GL89-01) 12/93 324 GL-89-01 D025 HATCH 1 M84635 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GLB9-01) 12/93 325.
GL-89-01 D025 MATCH 2 M84636 RELOCATE RETS TO ADMIN. SECTION OF TECW SPECS (CL89-01) 12/93 326.
GL-89-01 D025 KEWAUNEE M86417 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GL89-01) 11/93 327.
CL-89-01 D025 ST LUCIE 1 M85766 RELOCATE RETS TO ADMIN. SECTION OF TECH SPECS (GL89-01) 10/93 328.
GL-89-01 0025 ST LUCIE 2 M85767 RELOCATE RETS TO ADMIN. SECTION OF TECM SPECS (GL89-01) 10/93 9329.
GL-89-04 A025 ARKANSAS 1 M74756 IST REVIEWS AND SCHEDULES (GL 89-04) 11/93 M 330.
CL-89-04 A025 GINNA M74767 IST REVIEWS AND SCHEDULES (GL 89-04) 12/93 331.
GL-89-04 A025 NORTN ANNA 1 M74777 IST REVIEWS AND SCHEDULES (GL'89-04) 10/94 332.
GL-89 A025 WORTN ANNA 2 M74778 IST REVIEWS AND SCHEITJLES (GL 89-04) 10/94 333.
GL-89-04 A025 OCONEE 1 M74779 IST REVIEWS AND SCHEDULES (GL 89-04) 07/94 334.
GL-89-04 A025 OCONEE 2 M74780 IST REVIEWS AND SCNEDULES (GL'89-04) 07/94 335.
'A025 OCONEE 3 M74781 IST REVIEWS AND SCHEDULES (GL 89-04)-
07/94
'336.
GL-89-04 A025 Petty 1 M74784 IST REVIEWS AND SCHEDULES (GL 89-04) 04/94 337.
GL-89-04 A025-SALEM 1 M74790 IST REVIEWS AND SCHEDULES (GL 89-04)-
11/93 338.
GL-89 A025 SALEM 2 M74791 IST REVIEWS AND SCHEDULES (GL 89-04) 11/93 339.
GL-89-06 F072 3ROWNS FERRY 1 M73634 I.D.2 SAFETY PARAMETER DISPLAY' SYSTEM (GL89-06) 07/97 340.
GL-89-06 F072 BROWNS FERRY 2 M73635 1.D.2 SAFETY PARAMETER DISPLAY SYSTEM (GL89-06) 10/93 341.
GL-89-06 F072 BROWNS FERRY 3 M73636' l.D.2 SAFETY PARAMETER DISPLAY SYSTEM (GL89-06) 06/95 342.
GL-89-06 F072 'PALO VERDE 1
.M73686 1.D.2 SAFETY PARAMETER DISPLAY SYSTEM (GL89-06) 11/93 343.
CL-89-06
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04/94 344.
GL-89-06 L908 BR0htet FERRY 1 M73457 EROS 10N/CORROSt0W INDUCED PIPE WALL TNINNING (GL89-08)(OLD 3108) 07/97 345.
GL-89-08 L908 sROWNS FERRY 3 M73459 - EROSION / CORROSION INDUCED PIPE WALL TMINNING (GL89-08)(OLD B108) 06/95 346.
GL-89-10 5110 ARKANSAS 1 M75626 -MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 347.
-GL-89-10.
3110 ARKANSAS 2 M75627 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 348.
GL-89-10 8110 BEAVER VALLEY 1 M75628 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL'89-10) 06/94 349.
GL-89-10 8110 SEAVER VALLEY 2
~ M75629 MOTOR OPERATED VALVED. TESTING Als SURVEILLANCE (GL 89-10) 06/94 350.
GL-89-10 8110 BIG ROCK PotNT 1 M75632 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94
I l
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GL-89-10 8110 BRAIDWOOD 1 M75633 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 i
352.
GL-89-10 8110 BRAIDW0tB 2 M75634 MOTOR OPERATED VALVED TESitMG AND SURVEILLANCE (GL 89-10) 06/94 353.
GL-89-10 8110 BROWNS FERRY 1 M75635 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 07/97 354.
GL-89-10 8110 ~8ROWNS FERRY 2 M75636 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 05/95 355.
GL-89-10 8110 BROWNS FERRY 3 M75637 MOTOR OPERATED YALVED TESTING AhD SURVEILLANCE (GL 89-10) 06/95 356.
GL-89-10 8110 BRUNSUICK 1 M75638 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 357.
GL-89-10 8110 BRUNSWICK 2 M75639 MOTOR OPERATED YALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 358.
GL-89-10 8110 SYRON 1 M75640 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 359.
GL-89-10 8110 BYRON 2 M75641 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 360.
GL-89-10 E110 CALLAWAY 1 M75642 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 361.
CL-89-10 8110 CALVERT CLIFFS 1 M75643 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 07/94 362.
GL-89-10 8110 CALVERT CLIFFS 2 M75644 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 07/94 363.
GL-89-10 8110 CATAWBA 1 M75645 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 12/98 9 364.
GL-89-10 8110 CATAWBA 2 M75646 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 12/98 j
y 365.
GL-89-10 8110 CLINTON 1 M75647 MOTOR OPERATED YALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 l
366.
GL-89-10 8110 COMANCHE PEAK 1 M75648 MOTOR OPERATED VALVED TESTING AsiD SURVEILLANCE (GL 89-10) 06/94 367.
GL-89-10 8110 COMANCHE PEAK 2 M75649 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 368.
CL-89-10 8110 C00K 1 M75650 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 369 GL-89-10 8110 COOK 2 M75651 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 06/94 370.
GL-89-10 8110 CRYSTAL RIVER 3 M75653 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 06/94 371.
GL-89-10 8110 DAVIS-8 ESSE 1 M75654 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 11/94 372.
GL-89-10 8110 DIABLo CANYON 1 M75655 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 373.
GL-89-10 8110 DIABLO CANYON 2 M75056 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 374.
GL-89-10 8110 DRESDEN 2 475657 MOTOR OPERATED VALVED TESTING AND SURVE!LLANCE (GL 89-10) 06/94 375.
GL-89-10 8110 DRESDEN 3 M75658 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 376.
GL-89-10 8110 DUANE ARNOLD M75659 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 377.
GL-89-10 8110 FARLEY 1 M75660 HOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 378.
GL-89-10 8110 FARLEY 2 M75661. MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 379.
GL-89-10 8110. FERMI 2 M75662 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 380.
GL-89-10 8110 FITZPATRICK M75663 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 381.
GL-89-10 8110 FORT CALHOUN 1 M75664 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 06/94 382.
GL-89-10 8110' GINNA M75665 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 06/94 383.
GL-89-10 8110 GRAND GULF 1 M75666 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 384 GL-89-10 8110 NADOAM NECK M75667 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 02/96 385.
GL-89-10 8110 HARRIS 1 M75668 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 I
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421.
GL-89-10 5110 PRAIRIE ISLAND 1 M75704 MOTOR OPERATED YA'.VED TESTING AND SURVEILLANCE (GL 89-10) 06/94 422.
GL-89-10 8110 PRAIRIE ISLAe 2 M75705 MOTOR OPERATED VACtED TESTING AND SURVEILLANCE (GL 89-10) 06/94 423.
GL-89-10 B110 QUAD CITIES 1 M75706 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 424.
GL-89-10 8110 QUAD CIflES 2 M75707 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 425.
CL-89-10 8110 RIVER BEND 1 M75708 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 426.
CL-89-10 8110 ROBINSON 2 M75709 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 427.
GL-89-10 8110 SALEM 1 M75710 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 05/95 428.
GL-89-10 8110 SALEM 2 M75711 MOTOR OPERATED YALVED TESTING AND SURVEILLANCE (GL 89-10) 11/94 1
429.
GL-89-10 8110 SAN O MFRE 2 M75713 MOTOR OPERATED VALVYD TESTING AND SURVEILLANCE (GL 89-10) 06/94 430.
GL-89-10 8110 SAN ONOFRE 3 M75714 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 431.
GL-89-10 8110 SEABRON 1 M75715 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 432.
GL-89-10 8110 SEQUOYAN 1 M75716 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 03/95 433.
CL-89-10 8110 SEQUDYAH 2 M75717 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 03/95 O 434.
GL-89-10 8110 SOUTH TEXAS 1 M75719 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94
)
g435.
CL-89-10 8110 SOUTN TEXAS 2 M75720 MOTOR OPERATED VALVED TESTING AND SURYEILLANCE (GL 89-10) 06/94 436.
GL-89-10 8110 ST LUCIE 1 M75721 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 06/94 437.
CL-89-10 8110 ST LUCIE 2 M75722 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 438.
GL-89-10 8110 SLMMER 1 M75723 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 j
439.
GL-89-10 8110 SURRY 1 M75724. MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 j
440.
GL-89-10 B110 SURRY 2 M75725 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94
)
441.
- B110 SUSQUEHANNA 1 M75726 MOTOR OPERATED YALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 l
442.
GL-89-10 B110 SUSOUEHANNA 2 M75727 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 443.
GL-89-10 8110 THREE MILE ISLAND 1 M75728 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 444 GL-89-10 8110 TURKEY POINT 3 M75730 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89 10) 06/94 445.
GL-89-10 8110 TURKEY POINT 4 M75731 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 446.
CL-89-10 8110 VERMONT YANKEE 1 M75732 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 447.
CL-89-10 8110 V0GTLE 1 M75733 MOTOR OPERATED YALVED TESTING AND SURVEILLANCE (GL 89-10) 06/95 448.
GL-89-10 B110 V0GTLE 2 M75734 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89 10) 06/95 449.
GL-89-10 B110 WASNINGTON NUCLEAR 2 M75738 MOTOR OPERATED YALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 450.
GL-89-10 8110 WATERFORD 3 M75735 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 06/94 451.
GL-89-10 5110 WOLF CatEK 1 M75739 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (GL 89-10) 12/94 452.
GL-89-10 B110 ZION 1 M75741 MOTOR OPERATED VALVED TESTING AND SURVEILLANCE (CL 89-10) 06/94 453.
GL-89-10 5110 ZION 2 M75742 MOTOR OPERATED VALVED TESTING AND SL4VEILLANCE (GL 89-10) 06/94
'454.
GL-89-14' 0026 QUAD CITIES 1 M84319 ELIMINATION OF 3.25 REQUIREMENT IN TECH SPEC 4.0.2 (CL89-14)
/
455.
GL-89-14 D026 QUAD CITIES 2 M84320 ELIMINATION OF 3.25 REQUIREMENT IN TECN SPEC 4.0.2 (GL89-14)
/
l 1
j
456.
GL-89-16 3112 sROWNS FERRY 1 M74858 INSTALLATION OF HARDENED WETE LL VENT (GL 89-16) 07/97 457.
GL-89-16 5112 3ROWNS FERRY 3 M74860 INSTALLATION OF HARDENED WETWELL VENT (GL 89-16) 06/95 458. ~ GL-89-16 5112 MILLSTONE 1 M74872 INSTALLATION OF HARDENED UETWELL VENT (GL 89-16) 02/94 459.
3112 PEACM BOTTOM 3 M74877 INSTALLATION OF MARDENED WETWELL VENT (GL 89-16) 12/93 460.
CL-90-09 0028 GINNA M83570 Visual inspection FregJency for Sm&bers (GL-90-09) 12/93 461.
GL-91-01 D029 SAN ONOFRE 3 M84517 Removal of W/D Schedute for RV Meteriet rgecfmens (GL-91-01)
/
462.
GL-91-04 D031 - PEACM 90TTOM 2 M83704 TS SURVEILLANCE INTERVAL REeutREMENTS FOR 24 NO CYCLE (CL91-04) 10/93 463.
-0031 PEACM 90TTOM 3 M83705 TS SURVEILLANCE INTERVAL REQUIREMENTS FOR 24 Mo CYCLE (GL91-04) 10/93
-464.
CL-91-04
.0031,P!LGRIM 1-
'M83787 TS SURVEILLANCE INTERVAL REQUIREMENTS FOR 24 M0 CYCLE (GL91-04) 03/94 465.
CL-91-08 0030- CALVERT CLIFFS 2 M87559 Removat of Component Lists from Tech Spec :
02/94 466.
GL-91-08 0030 GINNA M77349 Resoval of Component Lists from Tech spec 01/94 467.
GL-91-11 L1111 CALLAWAY.1 M82391 VITAL INSTRUMENT BUSES & TIE BREAKERS (GI 48, GI 49)
/
468.
GL-91-11 L111 GINNA M82414 VITAL' INSTRUMENT SUSES & TIE BREAKERS (GI 48, GI 49) 12/93 M 469.
GL-91-11 L111' MATCH l'
'M82418. VITAL INSTRUMENT SUSES & TIE SREAKERS (GI 48, GI 49) 12/94
.$470.
GL-91 L111 MATCN 2 M82419 VITAL INSTRUMENT BUSES & TIE BREAKERS (GI 48,.GI 49) -
12/94
'471.
GL-91-11 L111 KEWRUNEE M82423 VITAL INSTRUE NT SUSES & TIE GREAKERS (Gl 48, GI 49) 11/93 472.
-L111 MCGUIRE 1 M82429 VITAL INSTRUMENT SUSES & TIE BREAKERS (GI 48, GI 49)
-12/93 473.
GL-91-11 L111 MCGUIRE 2..
M82430 VITAL INSTRLSE!NT BUSES & TIE tutEAKERS (GI 48, GI 49) 12/93 474 GL-91-13
'3119-~BRAIDWOOD 1 M61167 ESSENTIAL SERVICE WATER SYSTEM FAILURES (GSI~130) 10/93 475.
GL-91-13 5119. BRAIDWOOD 2 M81168 - ESSENTIAL SERVICE WATER SYSTEM FAILURES (GSI 130) 10/93 476.
GL-92-01 3120i ARKANSAS 1 M83730 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) -
03/94 ~
477 GL-92-01 3120' ARKANSAS 2.
M83430 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01)-
12/93 478.
GL-92-01 8120- SEAVER VALLEY 1 MB3431 ' REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 479.
GL-92-01 3120; BEAVER VALLEY 2 M83432: REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01)7
'12/93-
' 480.'
'GL-92 -8120 BIG ROCK POINT 1-M83435- ' REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01).
12/93-481.-
GL-92 8120 '3RAIDWOOD 1.
M83436 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93
-482.
3120.ORAIDWOOD 2 MB3437 REACTOR W~,SEL STRUCTURAL INTEGRITY (GL 92-01) -
-12/93 483.
GL-92 5120 BRONNS FERRY 1
' MB3438 ' REACTOR VESSEL. STRUCTURE INTEGRITY (GL 92-01) 12/93 484.
GL-92-01 3120 sROWNS FERRY 2 MB3439 REACTOR VESSEL STRUCTURAL INTE M ITY (GL 92-01) 12/93 485..GL-92-01 8120 SRonals FERRY 3 -
M83440 REACTOR VESSEL STRUCTURAL INTEGRITY ( R 92-Ot) 12/93 486.
GL-92-01 3120' ONUNSUICK 1 MB3441 REACTOR VESSEL STSUCTUR E INTEGRITY (GL 92-01)-
12/93.
487.c GL-92-01.
a120 mRUNSulcK 2 MB3442 REACTOR VESSEL STRUCTUR E INTEORITY (GL 92-01) 12/93 48B..
GL-92-01 B120 :sYRON 1
- M83443 ' REACTOR VESSEL STRUCTim E INTEGRITY (GL 92-01) 12/93 489.'
GL-92-01 3120' BTRON 2 matm REACTOR WESSEL STRUCTURE INTEGNITY (GL 92-01) 12/93
-490..
GL-92-01:
3120 CALLAMAY 1
'MB3445 REACTOR DESSEL STRUCTUR E INTEGRITY (GL 92-01) 12/93
,--w-__,
491.
GL-92-01 8120 CALVERT CLIFFS 1 M83446 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 492.
GL-92-01 8120 CALVERT CLIFFS 2 M83447 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 493.
GL-92-01 8120 CATAWBA 1 M83448 REACTOR VESSEL STRUCTURAL INTEGRITY (CL 92-01) 12/93 494.
GL-92-01 8120 CATAUBA 2 M83449 REACTOR YESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 495.
GL-92-01 8120 CLINTON 1 M83450 REACTOR v1ESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 496.
GL-92-01 8120 COMANCHE PEAK 1 M83451 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 497.
GL-92-01 8120 COMANCHE PEAK 2 M83452 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 498.
GL-92-01 8120 C00K 1 M83453 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 499.
GL-92-01 8120 COOK 2 M83454 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 500.
GL-92-01 8120 COOPER STATION M83455 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 501.
GL-92-01 8120 CRYSTAL RIVER 3 M83731 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 502.
GL-92-01 9120 DAVIS-sESSE 1 M83732 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/94 503.
GL-92-01 8120 DIABLO CANYON 1 M83456 REACTOR VESSEL STRUCTU!tAL INTEGRITY (GL 92-01) 01/94 O 504.
GL-92-01 8120 DIABLO CANYON 2 M83457 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 01/94 Q^505.
GL-92-01 8120 ORESDEN 2 M83458 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 506.
GL-92-01 8120 DRESDEN 3 M83459 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93
{
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GL-92-01 8120 DUANE ARNOLD M83460 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 508.
GL-92-01 8120 FARLEY 1 M83461 REACTOR VESSEL iTRUCTURAL INTEGRITY (GL 92-01) 12/93 509.
GL-92-01 8120 FARLEY 2 M83462 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 510.
GL-92-01 8120 FERMI 2 M83463 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 511.
GL-92-01 8120 FITZPATRICK M83464 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 512.
GL-92-01 B120 FORT CALMOUN 1 M83465 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 01/94 513.
GL-92-01 C120 GINNA M83733 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 01/94 514 GL-92-01 8120 GRAND CULF 1 M83466 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 01/94 515.
GL-92-01 s120 NADOAM NECK M83467 REICTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 j
516.
GL-92-01 3120 NARRIS 1 M83468 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 517.
GL-92-01 s12C NATCH 1 M83469 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 518.
GL-92-01 8120 NATCN 2 M83470 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/94 519.
GL-92-01 8120 NOPE CREEK 1 M83471 REACTOR VESSEL STRUCTURAL INTEGR!1Y (GL 92-01) 12/93 520.
GL-92-01 8120 INDIAN POINT 2 M83472 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 I
521.
GL-92-01 8120. INDIAN POINT 3 M83473. REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 522.
GL-92-01 s120 KEWAUNEE M83474 REACTOR VESSEL STRUCTURAL IkTEGRITY (GL 92-01) 12/93 523.
GL-92-01 8120 LASALLE 1 M83475 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 524.
GL-92-01 8120 LASALLE 2 M83476 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 525.
GL-92-01 3120 LIMERICK 1 M83477 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 i
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E-92-01 8120 MCailRE 1 M83480 REACict VESSEL STRUCTUR E INTEGRITY (E 92-01) 12/95 529.
E-92-01 R120 MCeUIRE 2 M83481 REACTOR VESSEL STRUCTURE INTEGtITY (E 92-01) 12/93 t
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GL-92-01 8120 -MILLSTONE 2 M83483 ItEACTOR VESSEL STRUCTURR INTEGRITY (EL 92-01)-
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GL-92-0; 3120 NFSE MILE P0fMT 2 M83487 - RLacTOR VESSEL STRUCTURE INTEGRIR (E 92-01) 12/95 536.
GL-92-01 5120 Nrsti:t ANNA 1 M834PS. REACTOR VESSEL S'VUCTURAL INTEERITY (GL 92-01) 12/93 l'
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GL-92 :8120 ' OTSTER CNEEK 1 M83490 - REACTOR WSSEL STRUCTURE INTEGRITY (et'92-01) -
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GL-92-01 3120' PILeRIM 1 MB3498 - REACTOR WESSEL STRUCTUR E.INTEERITY (el 92-01)
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fi-92-01 3120 SEmeR00K 1 MB3512 REACTOR VESSEL STRUCTURAL INTEGRITY ( R 92-01) 01/96 563.
R-92-01 5120 SEGUOVAN 1 M83513 REACitNt VESSEL STRUCTURE INTEGRITY (GL 92-01) 12/94 564.
E-92 01 3120 SEGUDYAN 2
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E-92-01 5120 SOUTM TEXAS 1
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R-92-01 3120 SOUTN TENAS 2 43516 REACTOR W SSEL STRUCTURE INTEERITY (GL 92-01) 12/93 567..
GL-92-01 3120 ST LUCIE 1 M83505 REACT 3t WSSEL STRUCTURAL INTEGRITY (GL 92-01) 01/96-563. ' R-92-01 3120 ST LUCIE 2 M83506 REACTOR WSSEL STRUCTURAL INTEGRITY (GL 92-01) 01/94 569. - R-92-01 3120 SupeqER 1' MB3517 REACTOR W SSEL STRUCTURE INTEGRITY (et 92-01) 12/93 570.
GL-92-01 3120 SURRY 1 MB3739 REACTOR VESSEL STRUCTUltAL INTEGPTTY (GL 92-01).
12/93 571.
GL-92-01 5120 SURRY 2 MB3740 REACTOR VESSEL STRUCTURAL l#TEtallTY (GL 92-01) 12/93 5 72..GL-92-01
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GL-92-01 5120 - SUSGUENANNA 2 M83519 REACTOR VESSEL STRUCTURE INTEGRITY (GL 92-01) 12/93 9574.
GL-92 8120 THREE MILE ISLAND 1 MB3741 REACTOR DESSEL STRUCTURE INTEGRITY (GL 92-01) 12/93
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01/96 580.
E-92-01 8120 MSWINGTON NUCLEAR 2 'MB3527 REACTOR WSSEL STRUCTUltR INTEGRITY (GL 92-01)
- 12/93
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GL-92-01' B120 MTERFORD 3 MB3524 REACTOR WSSE1 STRUCTURE INTE45tITY (GL 92-01)-
12/93
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12/93 584. GL-92-01 B120 IION 2 IEE3745 REACTOR VESSEL STRUCTURAL INTEERITY (GL 92-01).
12/95' i
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'3121 CODDER STATION M86275 Bist LEVEL INSTRISWITATION (eL-92-06) 12/93 586.
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EL-92-04
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GL-92-08 L208 SEAVER VALLET 1 M85516 THEINED-LAG (GENERIC LETTER 92-08) 12/95 597.
GL-92-08 L208 SEAVER VALLET 2 M85517 TNERMD-LAG (GENERIC LETTER 92-08)'
12/93 598.
GL-92-08 L208 SRAIDWOOD 1 M85521 THERMO-LAG (GENERIC LETTER 92-08)
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GL-92-08 L208 BRAIDWOOD 2-M85522 TMERMO-LAG (GENERIC LETTER 92 08)
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GL-92-08 L208 SR0lRIS FERRY
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GL-92-08 L208 Stonals FERRY 2 M85524 THERMD-LAG (GENERIC LETTER 92-08) 11/95-602.
GL-92-08 L208 BROWNS FERRY 3 M85525 THERMO-LAG (GENERIC LETTER 92-08)'
11/93 603.
GL-92 'L208 SRUNSWICK 1 M85526 -TMERMD-LAG (GENERIC LETTER 92-08)'
03/95-604.
GL-92-08 L208 BRUNSWICK 2 1485527 THEIWeD-LAG (GENERIC LETTER 92-08) 03/95 605.
GL-92-08 L208 BYRON 1
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GL-92-08 L208 CALLAWAY 1 MB5530 THERM 0-LAG (GENERIC LETTER 92-08)
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GL-92-08 L208 CLINTON 1 MB5535 THElueD-LAG (GENERIC LETTEQ 92-08) ~
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GL-92-08 L208 - Cook 2
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GL-92-08 L208 -FERMI 2 1485550 1 'THEINID-LAG (GENERIC LETTER 92-08) -
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GL-92-08 L208. GRAND GULF 1 M85554 ' TIIEINED-LAG (GENERIC LETTER 92-06) '
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GL-92 L208 NARRIS 1~
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GL-92-08 L208 NATCN 1 te85557. ' THEINID-LAG (GENERIC LETTER 92-08) 12/95
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GL-92-08 LL208 'IIATCM 2
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GL-92-08 L208, 1181411 POINT 2 M85560 THElWID-LAG (GENERIC LETTER 92-08)
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GL-92-08 L208 MILLSTONE 2 M85571 THERMO-LAG (GENERIC LETTER 92-03) 06/94 632.
GL-92-08 L208 MILLSTONE 3 M85572 THERMO-LAG (GENERIC LETTER 92-08) 12/93 653.
GL ~82-08 L208 NINE MILE PolNT 1 M85574 THERMO-LAG (GENERIC LETTER 92-08) 12/93 634.
GL-92-03 L208 N'NE MILE POINT 2 M85575 THERMO-LAG (GENERIC LETTER 92-08) 12/93 635.
GL-92-08 L208 NORTH ANNA 1 M85576 THERMO-LAG (GENERIC LETTER 92-08) 12/93 636.
GL-92-08 L208 NORTN ANNA 2 M85577 THERMO-LAG (GENERIC LETTER 92-08) 12/93 637.
GL-92-08 L208 OYSTER CREEK 1 M85581 THERMO-LAG (GENERIC LETTER 92-08) 04/94 638.
GL-92-08 L208 PALISADES M85582
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639.
GL-92-08 L208 PALO VERDE 1 M85583 THERMO-LAG (GENERIC LETTER 92-08)
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640.
GL-92-08 L208 PALO VERDE 2 M85584 THERMO-LAG (GENERIC LETTER 92-08)
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641.
CL-92-08 L208 PALO VERDE 3 M85585 TsrfRMO-LAG (GENERIC LETTER 92-08)
/
642.
GL-92-08 L208 PEACM BOTTOM 2 M85586 THERMO-LAG (CENERIC LETTER 92-08) 12/94 643.
GL-92 08 L208 PEACH BOTTOM 3 h55 W THERMO-LAG (GENERIC LETTER 92-05) 12/94 9 644 GL-92-08 L208 PERRY 1 M85588 THERMO-LAG (GENET'C LETTER 92-08)
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GL-92-08 L208 Po!NT SEACM 1 M85590 THERM 0-LAG (GENERIC LETTER 92-08)
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GL-92-08 L208 PotNT BEACM 2 M85591 THERM 0-LAG (GENERIC LETTER 92-08)
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l 647.
GL-92-08 L208 PRAIRIE ISLAND 1 M85592 THERMO-LAG (GENERIC LETTER 92-08)
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648.
GL-92 06 L208 PRAIRIE ISLAND 2 M85593 THERMO-LAG (GENERIC LETTER 92-04)
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GL-92-08 L208 RIVER BEND 1 M855%
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650.
GL-92-08 L208 SAN ONOFRE 2 M85601 THERMO-LAG (GENERIC LETTER 92-08)
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651.
GL-92-08 L208 SAN ONOFRE 3 M83602 THERMO-LAG (GENERIC LETTER 92-08)
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GL-92-08 L208 SEQUOYAN 1 M85604 THERMO-LAG (GENERIC LETTER 92-08) 01/94 653.
GL-92-08 L208 SEQUOYAM 2 Mb5605 THERM 0-LAG (GENERIC LETTER 92-08) 01/94 654.
GL-92-08 L208 SOUTN TEXAS 1 M85606 THERMO-LAG (GENERIC LETTER 92-08) 06/94 655.
GL-92-08 L208 SOUTH TEXAS 2 M85607 THERMO-LAG (GENERIC LETTER 92-08) 06/94 656.
GL-92-08 L206 ST LUCIE 1 M85608 TERMO-LAG (GENERIC LETTER 92-08) 12/93 657.
GL-92-08 L208 9 LUCIE 2 M85609 THERMO-LAG (GENERIC LETTER 92-08) 12/93 658.
GL-92-08 L208 SupMER 1 M85610 THERMO-LAG (GENERIC LETTER 92-08) 10/93 659.
GL-92-08 L206 SURRY 1 M85611 THERMO-LAG (GENERIC LETTER 92-08) 06/94 660.
GL-92-08 L206 SURR 2 M85612 THERMO-LAG (GENERIC LETTER 92 08) 06/94 661.
GL-92-08 L208 $UscuERANNA 1 M85613 THERMO-LAG (GENERIC LETTER 92-08) 12/95 662.
GL-92-08 L208 SUSQUEMANNA 2 M85614 THERMO-LAG (GENERIC LETTER 92-08) 12/94 663.
GL-92-06 L208 THREE MILE ISLAND 1 M85615 THERMO-LAG (GENERIC LETTER 92-08) 12/94 666 GL-92-08 L208 TURKEY POINT 3 M85616 inERMO-LAG (GENERIC LETTER 92-06) 12/93 665.
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GL-92-08 L208 V0GTLE 1 M85619 TIIERMD-LAG (GENERIC LETTER 92-08) 12/96 668.
GL-92-06 L208 V0GTLE 2 M55620 TNERMD-LAG (GENERIC LETTER 92-08) 12/96 669.
GL-92-06 L208 WASNINGTON NUCLEAR 2 M85624 THERMD-LAG (GENERIC LETTER 92-06)
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GL-92-08 L206 WTERFORD 3 M85621 THERMD-LAG (GENERIC LETTER 92-06)
/
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GL-97-06 L208 WOLF CREEK 1 M85625 TMERMD-LAG (GENE 2tC LETTER 92-06)
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GL-V2-08 L208 ZION 1 M85626 THERMD-LAG (GENERIC LETTER 92-06)
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GL-92-06 L208 ZION 2 M85627 THERMO-LAG (GENERIC LETTER 92-06)
/
674.
GL-93-04 L304 SEAVER VALLEY 1 M86831 ROD CCatTROL SYSTEM FAILURE & WIT WRA WL OF RCCA'S (GL 93-06) 12/93 675.
GL-93-04 L304 SEAVER VALLEY 2 M86832 RCD CONTROL SYSTEM FAILURE & WITMDRAWL ff RCCA*S (GL 93-04) 12/93 676.
GL-93-04 L304 8RAIDWOOD 1 M86833 RCD CONTROL SYSTEM FAILURE & WITMDetAEL OF RCCA'S (GL 93-04)
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GL-93-04 L304 8RAIDWOOD 2 M86834 RCD CONTROL STSTEM FAILURE & WIT WRA mt OF RCCA's (GL 93-06) 678.
GL-93-04 L304 SYRON 1 M86835 RCD CONTROL SYSTEM FAILURE & WITluNtAWAL OF RCCA'S (GL 93-04)
/
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GL-93-04 L304 8YRON 2 M86836 ROD CONTROL SYSTEM FAILURE & WITieRAW L OF RCCA'S (GL 93-06)
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GL-93-04 L304 CALLAWY 1 M86837 ROD CONTROL SYSTEM FAILURE & WITNDRAWAL OF RCCA'S (GL 93-04)
/
681.
GL-93-04 L304 CATAWBA 1 M86838 RCD CONTROL SYSTEM FAILURE & WITMDRA E L OF RCCA'S ( K 93-06)
/
682.
GL-93-04 L306 CATAus4 2 M86839 RCD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA's (GL 93-04)
/
683.
GL-93-04 L304 COMANCNE PEAK 1 M86840 RCD CONTROL SYSTEM FAILURE & WITMDRA E L OF RCCA'S (GL 93-04)
/
684.-
GL-93-04 L304 COMANCNE PEAK 2 M86841 RCD CONTROL SYSTEM FAILURE & WITieRAMAL OF RCCA's (GL 93-04) 12/93 685.
GL-93-04 L304 COOK 1 M86842 RCD CONTROL SYSTEM FAILURE & WITMDRA W L OF PCCA*5 (GL 93-04) 03/94 686.
GL-93-04 L304 COOK 2' M86843 RCD CONTROL SYSTEM FA**URE & WITilDRAMAL OF RCCAng ggt 93 06) 03/96 687.
GL-93-04 L304 DIA8t0 CANYON 1 M86844 ROD CONTROL SYSTEM FA D 58 & WITNDRAW L OF RCCA'S (GL 93-06) 03/96 688.
GL-93-04 L304 DIA8to CANYON 2 M86865 RCD CONTROL SYP8M FAILustE & WITNDRAMAL OF RCCA'S (GL 93-06) 03/96 689.
GL-93-04 L304 FARLEY.1 M86846 RCD CONTROL JYSTEM FAILURE & WITMDRA W L OF RCCA'S (GL 93-06)
/
690.
GL-93-04 L304 FARLEY 2.
F447 ' RCD CONTROL SYSTEM FAILURE & WITNDRAMAL OF RCCA'S (GL 93-06)
/
691.~
GL-93-06 L304 -GINNA M4A68
- 0D CONTROL SYSTEM FAILURE & WITIEMtAMAL OF RCCA'S (GL 93-06)
/
692.
GL-93-04 L304 NARRIS 1
. M86849 RCD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA'S (GL 93-04) 03/94 693.
GL-93-04 L306 INDIAll POINT 2 M86850 RCD CONTROL SYSTEM FAILURE & WITWRAM6L OF RCCA'S (GL 93-04)
/
'696.
'L304 IWIAll POINT 3 M86851 :. ROD CONTROL SYSTEM FAILURE & WITIEMtAMAL OF ltCCA'S (GL 93-06) 03/96 695..GL-93-04 L306 KEWRUNEE M86852 RCD CONTROL SYSTEM FAILURF & WITMDRAMAL OF RCCA'S (GL 93-06).
/
696.
GL-93-04 L306 MCGUIRE 1 M86853 RGD CONTROL SYSTEM FAILIAE & WITINHtAMAL OF RCCA'S (GL 93-04) 03/96 697..
GL-93-04 L304 -MCGUIRE 2 M86854 RCD CONTROL CYSTEM FAllultf a WITNDitAMAL OF RCCA*$ (GL 93-06) 03/96'-
698.
'GL-93-06 L306 MILLSTONE 3 M86855 RCD CONTROL SYSTEM FAILURE & WITNDIIAMAL OF RCCA'S (GL 93-04) 03/96
[
699.
GL-93-04 L304 lecitTN ANNA 1 M86856. RCD CONTROL SYSTEM FAILUltE & IwITWRAMAL OF RCCA'S (GL 93-04) -
12/93 l
.700.
GL-93-04 L306 NORTN ANNA 2 M86857 RCD CONTROL SYST8". rC'*E E WITNDRAWL r4 RCCA'S (GL 93-04) 12/93 l
l l
l
.,m_
.-.m 1
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t 701.
E-93-04 L3W POINT BEACM 1 M86858 RCD CONTROL SYSTEM FAILURE & WITIORAWK OF RCCA'S (R 93-06) 03/96 702.
GL-93-06 L304 POINT BEACM 2 M86859 RCD CONTROL SYSTEM FAILURE & WITNDRAMAL OF RCCA'S (R 93-06) 03/96 703. _GL-93-04 L304 PRAIRIE ISLAND 1 M86860 ROD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA'S (GL 93-06) 03/96 704.
GL-93-04 L3% PRAIRIE ISLAND 2 M86861 RCD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA'S (GL 93-06) 03/94 705.
GL-93-04 LLO4 RosINSON 2 Mann o ROD CONTROL SYSTEM FAILURE & WITMDRAWAL OF RCCA'S ( R 93-06) 12/93
'706.
GL-93-04 L306 SALEM 1 MM'l R00 CONTROL SYSTEM FAILURE & WITNDRAWAL OF RCCA'S (GL 93-06) 03/94 707.
GL-93-04 L304 SALEM 2 M86864 ROD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA'S (GL 93-06) 03/94 708.
GL-93-06 L304 SEABROOK 1 M86865 R00 CONTROL SYSTEM FAILURE & WITNDRAMAL OF RCCA'S (GL 93-06) 03/96 1
I 709.
.L306 SEQUOYAN 1 MRAAAA RCD CONTROL SYSTEM FAILURE & WITMDRAnabL OF RCCA'S (GL 93-06) 03/96 710.
GL-93-04 L306 SEGUOYAN 2 M86867 RCD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA'S (GL 93-06) 03/96 711.
GL-93-06 L306 SOUTN TEXAS 1 M86868 ROD CONTROL SYSTEM FAILURE & WITMDRAWAL OF RCCA'S (GL 03-06) 03/96 712.
L306 SOUTIE TEXAS 2 M86869 RCD CONTROL SYSTEM FAILURE & WITMDRAWAL OF RCCA'S (GL 93-06) 03/96 713.
GL-93-04 L304 SUMMER 1 M86870 RCD CONTROL SYSTEM FAILURE & WITMDRAWAL OF RCCA'S (GL 93-06)
/
9 714 GL-93-04 L306 SURRY 1-M86871 RCD CONTROL SYSTEM FAILURE & WITNDRAMAL OF RCCA*S (GL 93-06)
/-
[
$715.
GL-93-04 L306 SURRY 2 M86872 RCD CONTROL SYSTEf1 FAILURE & WITNDRAWAL OF RCCA'S (GL 93-06)
_/
716.
GL-93-06 L304 TURKEY POINT 3 M86873 Rap CONTROL SYSTEM FAILLNtE & WITINHtAWAL OF RCCA'S (GL 93-06) 03/94 717.
GL-93-06' L306 TURKEY POINT 4 M86876 RCD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA'S (GL 93-06) 03/96 718.
GL-93-06 L304 V0GTLE 1 M86875 RCD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA'S (GL 93-06) 03/96 t
719.
L306 V0GTLE 2 M86876 RCD CONTROL SYSTEM FAILURE & WITMDRAWAL OF RCCA*S (GL 93-06) 03/96
- 720.
~GL-93-06 L304 WOLF CREEK 1 M86878 RCD CONTROL SYSTEM FAILURE & WITINMtAMAL OF RCCA's (GL 93-06) -
03/94
'721. ~ GL-93-04 L306 Il0N 1 MnAnan RCD CONTROL SYSTEM FAILLNtE & WITMDRAMAL OF RCtA's (GL 93-06) -
03/96 r
722.
-GL-93-06
.L304 ZION 2 M86881 RCD CONTROL SYSTEM FAILURE & WITMDRAMAL OF RCCA*S (GL 93-06) 03/96 723.
MPA-A004 A006 3ROWNS FERRY 1 M08715. APPENDIX J - CONTAINNENT LEAK TESTING 07/97 726.
24-A006 A004-BROWNS FERRY 3 -
M08717 APPENDIX J - CONTAlINENT LEAK TESTING.
06/95 725.
MA-5032 9032 NASDAM NECK M49625 - SLOCKED SI SIGIIAL DURING C00LDonal
/
726.
WA-0061 3061 BRotAIS FERRY 1 1868136 FIRE PROTECTION - FINAL TECM SPECS (INCLimES sER SUPPLEE NTS) 07/97
-727.
MPA-0061 3041 eROWNS FERRY 3
>M68136 FIRE PROTECTICII --FIIIAL' TECII SPECS (INCLISES SER SUPPLEENTS) 06/95
=728.
MPA-3116 B116 FERMI 2 MTTT75 --.. suPP 3, INic sPOWs0 RED TESTS OF IIDTOR-CPERATB WALVES (GL89-10) 12/95' 729.
IN'A-0116 3116 NATCN 1 -
M777T8 suPP 3, INtc SPOWs0 RED TESTS OF IIDTOR-CPERATED VALVES (GLSD-10) 12/95~
730.. MPA-3116 3116 INLTCM 2 317T779. SUPP 3, turc SPOWs0 RED TESTS OF IIDTOR-OPEIIATED VALVES (GLS9-10) 12/96 731.z IFA-8116
-3116 LASALLE 2 M77F82 SUPP 3, NRC sPouenasa TESTS OF IIDTOIt-CPERATED VALVES (GL89-10) 10/93 732.-
MPA-5116 3114 OfSTER CREEK 1 M77789 SUPP 3, NRC spmanasa TESTS OF IIDTOR-OPERATED VALVES (GL89-10) 06/96 733.
IW'A-8116 '
_B116 VERMOIIT YAINEE i MTF800 SUPP 3, NRC sPONenusa TESTS OF IIDTOIt-OPERATED VALVES (GL89-10) 06/94 '
MR5210 sUPP 3, INIC SPOWs0 RED TESTS OF IIDTOR-OPERATED VALVES (GL89-10) 99/95 734.-
5F4-8116 3116 40GTLE 1._
9185211 SUPP 3, NRC SPOWs0 RED TESTS OF MOTOR-CPERATED VALVES (GLS9-10) 09/95 735.
Irn-3116 -
5116~.40GTLE 2 1
736.
MPA-8117 8117 CALLAWAY 1 M81598 SUPP 2 - FAILURE OF WESTINGHOUSE SG TUBE MECHANICAL PLUGS
/
737.
MPA-8117 8117 GINNA M81622 SUPP 2 - FAILURE OF WESTINGHOUSE SG TU8E MECHANICAL PLUGS 12/98 738.
MPA-B117 8117 MILLSTONE 3 M81636 $UPP 2 - FAILURE OF WESTINCHOUSE SG TUBE MECHANICAL PLUGS 12/94 739.
MPA-8117 8117 ~ 2 ION 1 M81680 SUPP 2 - FAILURE OF WESTINGHOUSE SG TUBE MECHANICAL PLUGS
/
740.
MPA-B117 B117 2 ION 2 M81681 SUPP 2 - FAILURE OF WESTINGHOUSE SG TUBE MECHANICAL PLUGS
/
741.
MPA-B118 8118 ARKANSAS 1 M83588 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
742.
MPA-8118 8118 ARKANSAS 2 M83589 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
743.
MPA-B118 0118 BEAVER VALLEY 1 M83590 IPE EXTERNAL EVENTS (GL88-20, SUPP A) 06/97 744.
MPA-B118 B118 BEAVER VALLEY 2 M83591 IPE EXTERNAL EVENTS (CL88-20. SUPP 4 07/99 745.
MPA-8118 B118 BIG ROCK POINT 1 M83592 IPE EXTERNAL EVENTS (GL88-20, SUPP 4 05/95 746.
MPA-B118 B118 BRAIDWOOD 1 M83593 IPE EXTERNAL EVENTS (CL88-20, SUPP 4)
/
747.
MPA-8118 8118 BRAIDWOOD 2 M83594 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
74 8.
MPA-8118 8118 BROWNS FERRY 1 M83595 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 07/97 9 749.
MPA-8118 B118 BROWNS FERRY 2 M835% IPE EXTERNAL EVENT 3 (GL88-20, SUPP 4) 07/%
$ 750.
MPA-B118 B118 BROWNS FERRY 3 M83597 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 07/M 751.
MPA-8118 8118 BRUNSWICK 1 M83598 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 12/95 752.
MPA-B118 8118 BRUNSUICK 2 M83599 IPE EXTERNAL EVENTS (CL88-20, SUPP 4) 12/95 753.
MPA-B118 B118 BTRON 1 M83600 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
754.
MPA-8118 8118 BTRON 2 M83601 IPE EXTERNAL EVENTS (CL88-20, SUPP 4)
/
I 755.
MPA-8118 B118 CALLAWAY 1 M83602 IPE EXTERNAL EVENTS (CL88-20, SUPP 4)
/
756.
MPA-B118 8118 CALVERT CLIFFS 1 M83603 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
757.
MPA 8118 8118 CALVERT CLIFFS 2 M83604 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) -
/
758.
MPA-B118 B118 CATAWBA 1 M83605 IPE EXTERNAL EVENTS (CL88-20, SUPP 4)
/
759.
MPA-8118 B118 CATAWBA 2 M83606 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
760.
MPA-B118 8118 CLINTON 1 MS3607-IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
761.
MPA-8118 8118 COMANCHE PEAK 1 M83608 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
762.
MPA-8118 8118 COCK 1 M83609 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 12/93 763.
MPA-8118 B118 COOK 2 M83610 IPE EXTERNAL EVENTS (GL88-20, SuPo 4) 12/93 764 MPA-8118 B118 COCPER STATION M83611 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
.12/95 765.
MPA-8118 8118 CRYSTAL RIVER 3 M83612 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
766.
MPA-8118 8118 DAVIS-BESSE 1 M83613 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 09/95 767.
MPA-8118 8118 DIABLO CANYON 1 M83614 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 01/96 768.
MPA-8118 8118 OIABLO CANTON 2 M83615 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 01/96 769 MPA-8118 8118 DRESCEN 2 M83616 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
77D.
MPA-8118 8118 DRESDEN 3 M83617 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
- - _ ~._ _._ _-.-._. _ _.
l 4
771.
WA-3118 B118 OURNE ARNOLD '
4 M83618 IPE EXTERNAL EWNTS (GL88-20, SUPP 4)
/
772. -MPA-3118 B118 FARLET 1 MB3619 IPE EXtreWAL EVENTS (E 88-20, suPP 4) 06/95 M83620 IPE EXTERWi EVENTS (E 88-20, SUPP 4) 06/95 773. -MPA-3116 8118 FARLET 2 774..MPA-3118
-3118 FERMI 2 W M83621 IPE EXTERNAL EWNTS (E88-20, SUPP 4) 06/95 775.
MPA-B118 3118 FIT 2 PATRICK M83622 IPE EXTERN E EVENTS (GL88-20, SUPP 4)
/
7 776.
wA-3118.
- 8118 GINNA M83624 IPE EXTERNAL E W NTS (GL88-20, SUPP 4) 05/95
- 777.
MPA-3118:
8118 GRAND EULF 1 MB3625 IPE EXTERNAL EVENTS (GL88-20, suPP 4).
'12/95 778.
MPA-B118 B118 IM00AM NECK MB3626 IPE EXTERNAL EVENTS (E88-20, SUPP.4)
/
-; 779. ' MPA-3118 B118' NARRIS 1 M83627 - IPE EXTERNAL EVENTF (E 88-20, SUPP 4) 06/96 780. -MPA-B118 5118 NATCM 1 MB3628 IPE EXTERNAL EVENTS (E88-20,' SUPP 4) 12/95 781.
MPA-3118 8118 NATCM 2 MB3629 IPE EXTERNAL EVENTS (E 88-20, SUPP 4)
-06/95 782. ' wA 3118
'5118 NOPE CREEK 1 MB3630 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 02/96
'783.
MPA-3118 3118 INDIAN POINT 2 M83631 IPE EXTERNAL EVENTS ( R 88-20, SUPP 4)-
/
9 784. : MPA-3118' 5118 ' INDIAN POINT 3 M83632 IPE EXTERNAL EVENTS (GL88-20, SUPP 4).
/
@785. -MPA-3118
'3118 EEWhuMEE M83633 IPE EXTERNAL EVENTS (GL88-20, suPP 4);
/
786.
MPA-3118 -
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/
787.'
MPA-3118 8118 LASALLE 2 M83635 IPE EXTERNAL EVENTS (E88-20, SUPP 4).
/
7 788.
MPA-B118 5118 LIERICK 1 M83636.IPE EXTERNAL EVENTS (E 88-20, SUPP 4)
-/
789. -.MPA-B118.
8118L LIMERICK 2-M83637 'IPE EXTERNAL EVENTS (GL88-20, SUPP 4)'
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790. -MPA-3118
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/
791. <MPA-5118
.3118 ' MCEtIRE 1 MB3639 IPE EXTERN E EVENTS ( E 88-20, SUPP 4)
/
792.
MPA-3118 5118 MCsUIRE 2 MB3640 ~IPE EXTERNAL EVENTS (E 88-20, SUPP 4)
/-
793.-
MPA-8118 5118 - MILLSTONE 1 M83641 IPE EXTERN E EVENTS (GL88-20, SUPP 4)
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l 795.
MPA-3118 B118 MILLSTONE 3 M83643 -IPE EXTERNAL EVENTS ( E 88-20, SUPP 4).
12/93 796.
wA-3118' 5118 - MONTICELLO MB3644 IPE EXTERNAL EVENTS'(GL88-20, SUPP 4):.
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/
'799. -MPA-3118 3118 NORTN ANNA 1 M83647 IPE EXTERNE EVENTS (GL88-20, SUPP 4)
.12/95 800.
WA-B118 B118 NORTN ANNA *:-
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12/95 801. ---WA-3118f 3118-OCONEE 1 iM83649 IPE EXTERNAL EVENTS ( E88-20, SUPP 4).
12/95
- 802.
WA-8118
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12/95 803.
WA-3118 0118 OCONEE 3 M83651' IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 12/95 804. - MPA-3118
. 8118 : UTSTER CREEK 1 M83652..IPE EXTERNAL FVENTS (E 88-20, SUPP 4)
- /
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MPA-8118 3118 PRO WNDE 1 MSM54 IPE EXTERNK EVENTS ( E88-20, SUPP 4) 1h94 807.
MPA-B118 3118 PALO VERDE 2 MSM55 IPE EXTEGNAL EM NTS ( E88-20, SUPP 4) 12/94 808.
MPA-B118 3118 PALO W ADE 3 M8M56 IPE EXTERNAL EVENTS (E 88-20, SUPP 4) 12/94 809.
MPA-M118
-3118 PEACM BOTTOM 2 M83657 IPE EXTERN R EVENTS (Gt.88-20, SUPP 4)
/
810.
WA-B118 5118 PEACM BOTTOM 3 MB3658 IPE EXTERNAL EVENTS (E 88-20, SUPP 4)
/
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MPA-B118 5118 PERRY 1
~
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WA-3118 5118 PILGRIM 1 M836(4 IPE EXTERNE EVEN13 (E 88-20, SUPP 4) 12/99 813.
WA-3118 5118 POINT BEACN 1 MB3661 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/'
814.. WA-3118 B118 POINT BEACN 2 MB3662 IPE EXTERNAL EVENTS ( E 88-20,' SUPP 4)
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815..8tPA-B118 3118 PRAIRIE ISLAIS 1 MA'Laat IPE EXTERNAL EVENTS (E88-20, SUPP 4)
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816.
WA-B118 3118 PRAIRIE ISLAte 2 MB3664 IPE EXTERNAL EVENTS ( E 85-20, SUPP 4)
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MPA-B118 5118. eUAD CITIES 1 M83665 :IPE EXTERNAL EWNTS (E88-20, SUPP 4)
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MFA-B118 5118 ROBINSON 2.-
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/
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MPA-3118 B118. SEEM 1.
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/
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MPA-3118 B118 SAN ONOFRE 2 1 MB36711 IPE EXTERN R EVENTS (GLOB-20, SUPP 4) ~
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MPA-B118-u3118 '-SAN ONOFRE 3
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'12/93' 826.
IPA-3118 5118 'SEeUOTAN 1 M83674 'IPE EXTERNAL EVENTS'(GL83-20, SUPP 4)
~06/95.
827.
MPA-3118 5118' SEeU0 FAN 2 -
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-06/95:
.,828.
MPA-B118 3118 'SOUTN TEXAS 1 MB3676 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 01/96 829.
MPA-3118' B118 - SOUTH TEXAS 2 -
-NB3677 'IPE EXTERNAL EVENTS (E 88-20, SUPP 4)-
' 01/96 830.
frA-3118-3118' -ST LUCIE 1 NB3678 : IPE EXTERNAL EVENTS (E88-20, SUPP 4)
- /
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IPA-3118 3118 ST LUCIE 2
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- 12/95 839.
IFA-3118 -
3118 TURKEY POINT 4 Immat.IPE EXTERNAL EVENTS (GL M -20, SUPP 4) 12/95
~ 840.- ;MPA-3118 5118 VERMONT YAIN(EE 1 MB3689 IPE EXTERN E EVENTS (GL8B-20, SUPP 4)
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MPA-B118 8118 V0GTLE 1 M83690 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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MPA-5118 5118 V0GTLE 2 M83691 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
843.
WA 8118 B118 nNLSNINGTON NUCLEAR 2 M83695 IPE EXTERNAL EVEN'S (GL88-20, SUPP 4)
/
844.
MPA-B118 8118 WATERFORD 3 M83692 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
845.
MPA-5118 5118 WOLF CREEK 1 M83696 IPE EXTERNAL EVENTS (GL88-20, SU,PP 4)
/
846.
MPA-8118 8118 ZION 1 M83697 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
847.
MPA-8118 8118 ZION 2 M83698 IPE EXTERNAL EVENTS (GL88 20, SUPP 4)
/
848.
MPA-B122 5122 ARKANSAS 1 M85352 ~ RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 849.
MPA-B122 5122. ARKANSAS 2 M85353 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 850.
MPA-8122 8122 BEAVER VALLEY 1 M85354 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 851.
MPA-8122 8122 BEAVER VALLEY 2 M85355 RESPONSE TO B-90-01, LOSS OF FILL CIL IN ROSEMOUNT TRANS (X001) 12/93 852.
MPA-5122 5122 BIG ROCK POINT 1 M85358 RESPONSE TO 8-90-01, LOSS OF FILL Ott IN ROSEMOUNT TRANS (X001) 12/94 853.
MPA-B122 B122 BRAIDWOOD 1 M85359 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 0 854 MPA-s122 8122 sRAIDWOOD 2 M85360 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 y855.
MPA-8122 8122 BROWMS FERRY 1 M85361 CESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT 1RANS (X001) 12/93 856.
MPA 3122 B122 mROWNS FERRY 2 M85362 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 857.
MPA-5122 5122 mROWNS FERRY 3 M85363 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93
'858.
MPA-3122 5122 'SYRON 1 M85366 RESPONSE TO 8-90-01,' LOSS OF FILL OIL IN ROSEt m NT TRANS (X001) 12/93 859.
MPA-5122 8122 BYR9 2 M85367 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 860.
MPA-8122 8122 CALLAMAY 1
.M85368 - RESPONSE TO S-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001)
- 12/94 861.
MPA-8122 5122 CATAWBA 1
.M85371 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93
'862.
MPA-8122 B122 CATAWEA 2 M85372 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 863.
MPA-B122 5122 CLINTON 1 M85373 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001)
?2/93 864.
MPA-B122 8122 COMANCHE PEAK 1 M85374 RESPONSE TO.8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12, %
865. 2MPA 8fik 8122 COOPER STATION M85378 RESPONSE TO 8-90-01,--LOSS OF FILL CIL IN ROSEMOUNT TRANS (X001) 12/M 866."'$FA-3122 B122 CRTSTAL RIVER 3 M85379 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/91 867.
MPA-8122-8122 DAVIS-BESSE 1 M85380. RESPONSE TO 8-90-01, LOSS OF FILL DIL IN ROSEMOUNT TRANS (X001) 12/93 868.
MPA-B122 8122 DIABLO CANYON 1 M85381 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) i2/93 869. -MPA-8122 8122 DIASLO CANYON 2 M85382 RESPONSE TO B-90-01, LOSS OF FILL Oll IN ROSEMOUNT TRANS (X001) 12/93 870.
MPA-8122 -
8122 DRESOEM 2 M85383 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001)'
12/93 871.
W A-8122 5122 DRESDEN 3 M85384 RESPONSE TO B-90-01, LOSS OF FliL OIL IN ROSEMOUNT TRANS (X001)
. 12/93 872.
MPA-8122 8122 FARLEY 1 M85386 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 873.. MPA-3122 8122 FARLEY 2 M85387 RESPONSE TO 8-90-01, LOSS OF FILL CIL IN ROSEMOUNT TRANS (X001) 12/93 874.
MPA-8122 -
B122-FERW2 2 M85388 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 03/94 875.
MPA-3122 8122 FITZPATRICK M85389. RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93
876.
MPA-B122 8122 GINNA M85391 RESPONSE T0 s-90-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 12/93 877.
MPA-8122 8122 GRAND GULF 1 M85392 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 01/94 878.
MPA-B122 B122 HARRIS 1 M85394 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 879.
MPA-B122 8122 MATCH 1 M85395 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/95 880.
MPA-B122 B122 HATCH 2 M853% RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/95 881.
MPA B122 B122 HOPE CREEK 1 M85397 RESPON4E TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 882.
MPA-8122 8122 INDIAN POINT 2 M85398 RESPONSE TO B-90-01, LOSS OF FILL CIL IN ROSDK)UNT TRANS (X001) 12/93 883 MPA-B122 B122 IN0!AN POINT 3 M85399 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 884.
MPA 8122 8122 LASALLE 1 M85401 RESPONSE TO B-90-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 12/94 885.
MPA-B122 8122 LASALLE 2 M85402 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 886.
MPA-B122 B122 LIMERICX 1 M85403 RESPONSE TO 8-93-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 01/95 887.
MPA-8122 B122 LIMERICK 2 M85404 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 01/95 888.
MPA-B122 B122 MA!NE YANKEE M85405 RESPONSE To B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 O 889.
MPA-B122 B122 MCGUIRE 1 M85406 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94
@890.
MPA-B122 8122 MCWIRE 2 M85407 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 891.
MPA B122 B122 MILLSTONE 2 M85409 RESPONSE TO B-90-01, LOSS OF FILL Dil IN ROSEMOUNT TRANS (X001) 12/94 892.
MPA-B122 B122 MILLSTONE 3 M85410 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 01/95 893.
MPA-B122 8122 MONTICELLO M85411 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 894.
MPA-8122 8122 NINE MlLE POINT 1 M85412 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 895.
MPA-B122 B122 NINE MILE POINT 2 M85413 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 896.
MPA-8122 8122 MORTN ANNA 1 M85414 RESPONSE TO B-90-01, LOSS OF FILL CIL IN ROSEMOUNT TRANS (X001) 12/93 897.
MPA-8122 8122 NORTH ANNA 2 M85415 RESPONSE TO B-90-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 12/93 898.
MPA-8122 B122 OCONEE 1 M85416 RESPONSE TO 8-90-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 12/93 899.
MPA-B122 B122 OCONEE 2 M85417 RESPONSE TO 8-90-01, LOSS OF FILL CIL IN ROSEMOUNT TRANS (X001) 12/93 900.
MPA-9122 B122 OCONEE 3 M85418 RESPONSE TO B-90-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 12/93 901.
MPA-8122 5122 PALISADES M85420 RESPONSE TO B-90-01, LOSS OF FILL Oll IN ROSEMOUNT TRANS (X001) 12/94 902.
MPA-8122 B122 PALO VERDE 1 M85421 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 903.
MPA 8122 B122 PALO VERDE 2 M85422 RESPONSE TO B-90-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 12/93 904.
MPA-8122 8122 PALO VERDE 3 M85423 RESPONSE TO B-90-01, LOSS CF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 905.
MPA-8122 8122 PEACH BOTTON 2 M85424 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 906.
MPA-B122 8122 PEACM BOTTOM 3 M85425 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 907.
MPA-8122 8122 PERRY.1 M85426 RESPONSE TO B-90-01, LOSS OF FILL Oil IN ROSEMOUNT TRANS (X001) 12/94 908.
MPA-B122 5122 PILGRIM 1 M85427 RESPONSE To B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 909.
MP,A-B122
-B122 PRAIRIE ISLAM 0 1 M85430 RESPONSE TO 8-90-01, LOSS OF FILL Cll IN ROSEMOUNT TRANS (X001) 12/93 910.
MPA-8122 5122 PRAIRIE ISLAND 2 M85431 RESPONSE TO 8-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93
~.
911.
MPA-B122 B122 QUAD CITIES 1 M85432 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMCUNT TRANS (X001) 12/94 912.
MPA-8122 8122 QUAD CITIES 2 M85433 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 913.
MPA-B122 8122 RIVER BEND 1 M85434 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 914.
MPA-B122 8122 ROBINSON 2 M85435 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 915.
MPA-8122 B122 SALEM 1 M85436 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 916.
MPA-B122 B122 SALEM 2 M85437 RESPCWSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 917.
MPA-B122 8122 SAN ONOFRE 2 M85439 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 918.
MPA-B122 B122 SAN ONOFRE 3 M85440 RESPONSE TO B 90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 919.
MPA-B122 B122 SEABROOK 1 M85441 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 920.
MPA-B122 8122 SEQUOYAN 1 M85442 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 921.
MPA-8122 8122 SEQUOTAN 2 M85443 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 922.
MPA-8122 B122 SOUTH TEXAS 1 M85444 RESPONSE TO B-90-01, LOSS OF FILL oil IN ROSEMOUNT TRANS (X001) 12/93 923.
MPA-8122 B122 SOUTH TEXAS 2 M85445 RESPONSE TO B-90-01, LOSS OF FILL CIL IN ROSEMOUNT TRANS (X001) 12/93 O 924 MPA-B122 8122 ST LUCIE 1 M85446 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 925.
MPA-B122 B122 ST LUCIE 2 M85447 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 926.
MPA-B122 B122 SUMER 1 M85448 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 927.
MPA-B122 8122 SURRY 1 M85449 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 928.
MPA-B122 8122 SURRY 2 M85450 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 929.
MPA-B122 8122 SUSQUEHANNA 1 M85451 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001).
12/95 930.
MPA-B122 8122 SUSQUEMAhWA 2 M85452 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/95 931.
MPA-B122 8122 THREE MILE ISLAND 1 M85453 RESPONSE TO B-90-01, LOSS OF FILL DIL IN ROSEMOUNT TRANS (X001) 12/93 932.
MPA-B122 8122 TURKEY POINT 3 M85454 RESPONSE TO B-90-01, LOSS OF FILL CIL IN ROSEMOUNT TRANS (X001) 12/94 933.
MPA-8122 8122 TURKEY PotNT 4 M85455 RESPONSE TO B-90-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 12/94 934.
MPA-B122 B122 VERMONT YANKEE 1 M85456 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 935.
MPA-8122 8122 V0GTLE 1 M85457 RESPONSE TO B-90-01, LOSS OF FILL O!L IN ROSEMOUNT TRANS (X001) 12/94 936.
MPA-B122 B122 V0GTLE 2 M85458 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/94 937.
MPA-B122 B122 UASN!NGTON NUCLEAR 2 M85462 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001)
- 2/94 938.
MPA-B122 8122 WOLF CREEK 1 MS5463 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 939.
MPA-B122 B122 2 ION 1 M85464 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNT TRANS (X001) 12/93 940.
MPA-B122 B122 2 ION 2 M85465 RESPONSE TO B-90-01, LOSS OF FILL OIL IN ROSEMOUNY TRANS (X001) 12/93 941.
MPA-Coll C011 BROWMS FERRY 3 M08931 RPS POWER SUPPLY 06/95
~
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v I
NRC DORM 336 U.S. NUCLEAR REGULATORY COMMISSION
- 1. REPORT NUMBER M 1102, ev Ond Num-saci,sao2 BIBUOGRAPHIC DATA SHEET b"'- N '"YJ 1
(See instruction. on the revw )
NUREG-1435 SUPP ement 3.
l
- a. Tm AND SueTm Status of Safety Issues at Licensed Power Plants
- a. oATE REPORT.PUBU5HED TMI Action Plan Requirements
- b December ' 1993 Unr6 solved Safety Issues Generic Safety Issues
Other Multiplant Action Isuses
- 5. AUTHOR (s)
- 6. TYPE OF REPORT JAnnual
- 7. PERIOD COVERED (incluelve Dates)
. 10/1/92 - 9/30/93 I
i
- 8. PERFORMING ORGANIZATION - NAME AND ADORE 88 (if NRC, provide Olvision, Office or Region, U.S. Nuclear Regulatory Commission, and malling addrees; if contractor, provide name and malling addrses.) z Program Management, Policy Development and Analysis Staff Office of Nuclear Reactor Regulation U.S. Nuclear Re ;ulatory Commission s
Washington, DC 20555-0001
' 8'?"if&#Jait*^ Cars.0""an^.T#!"a'J's.'." >"" '" "" " *""' " ""'"'" "'""' ""
Same as above
- 10. SUPPLEMENTARY NOTE 8
- 11. ABSTRACT (200 words or lese)
As part of ongoing U.S. Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program was established whereby an annual NUREG report would be published on the status of licensee imple.
mentation and NRC verification of safety issues m major NRC requirement areas. This information was compiled and reported in three NUREG volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Re-quirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). Supplement 1, pubbshed in De-cember 1991 combined these volumes into a single report and provided updated information as of September 30,1991. Supple-1 ment 2, published in December 1992, provided updated information on TMI, USI, and GSIissues and meluded status of all other
.l Multipla tt Actions (MPAs). This annual NUREG report provides updated information on TMI, USI, and GSI and other MPAs as of Septeraber 30,1993. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel.This report is to provide a comprehensive description of the implementation and verification status of TMI Action Plan Requirements, USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action j
or actions by licensees. This report makes the information available to other interested parties, including the public. An addi-tional purpose of this NUREG report is to serve as a follow-on to NUREG-0933, "A Prioritization of Generic Safety Issues,"
which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees.
- 12. KEY WoROS/DESCRIPTORS (List words or phrases that will asslet researchers in locating the report.)
- 13. AVALA81UTY 8TATEMENT Unlimited
- 14. 8ECURrrY CLA881FICATION Status of Safety Issues at Licensed Power Plants
- (Thi* Pas')
TMI Action Plan Requirements Unclassified Unresolved Safety Issues
. (This meport)
Generic Safety Issues Unclassified Other Multiplant Action Issues is. NUMBER OF ProES
- 16. PRICE -
NRC FORM 335 (2-89)
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- ou recycled paper-Federal Recycling Program
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