ML20126F005
| ML20126F005 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1435, NUREG-1435-S02, NUREG-1435-S2, NUDOCS 9212300075 | |
| Download: ML20126F005 (174) | |
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NUREG-1435 Supplement 2
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Sta:us of Safety Issues at Licensed Power Plants TMI Action Plan Requirements Unresolved Safety Issues Generic Safety Issues Other Multiplant Action Issues U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation
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l AVAILABILITY NOTICE i
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 Pubhc Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555 2.
The Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082 l
3.
The National Technical information Service, Springfield, VA 22161
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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 l
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NUREG-1435 Supplement 2 Status of Safety Issues at Licensed Power Plants TMI Action Plan Requirements Unresolved Safety Issues Generic Safety Issues Other Multiplant Action Issues Manuscript Completed: November 1992 Date Published: December 1992 Omce of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 pa a.w,
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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, a program was established whereby an annual NUREG report would be published 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 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 i
unresolved safety issues (USis). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSis). The first annual NUREG report, Supplement 1, which combined these volumes into a single report and provided updated information as of September 30,1991, was published in December 1991. This second annual NUREG report, Supplement 2, provides updated information as of September 30,1992. In addition, this supplement also provides the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSis.
The data contained in this supplement is a product of the NRC's safety issues management system (SIMS) data base, 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, 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. An additional purpose of this NUREG. report is i
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.
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a CONTENTS ABSTRACT ill EXECUTIVE
SUMMARY
.........................vil ABBREVIATIONS....
xi 1
INTRODUCTION..........
.....................1 1.1 Background...
I 1.2 Process and Accountability...........
............. 2 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 Stat us by l ssue.............,....................................... 17 2.5 Conclusions
.....................31 3
UNRESOLVED SAFETY lSSUES
..................33 3.1 Implementation Status..
....... 33 3.2 Verification Status.
. 37 3.3 = Status by Plant.
...................................41 3.4 Status by lssue.
. 45 3.5 Conclusions
.............49 4
GENERIC SAFETY lSSUES..............
..........51 4.1 Implementation Status.
............ 51 4.2 Verification Status....
....... 57 4.3 Status by Plant....
61 4.4 Status by lssue...
. 65 4.5 Conclusions.
.................69 5
OTHER MULTIPLANT ACTION ISSUES.
.... 71 5.1 Implementation Status...
71 5.2 Verification Status
...... 92 5.3 Status by Plant 95 5.4 Status by issuo.
..................99 5.5 Conclusions.....
............... 113 APPENDICES A
Usting of Unimplemented TMI Items by issue.
........ A.1 B
Listing of Unimplemented USl items by issuo.......
B-1 C
Listing of Unimplemented GSI Items by Issue..
. C-1 D
Listing of Other Unimplemented MPA ltems by issue....
.. D-1 FIGURES 2.1 TMl Action Plan Requirements Implementation Status at Licensed Plants....
.8 2.2 Projected Schedules for Remaining TMI Items..........
...................9 3.1 Unresolved Safety issues - Implementation Status at Licensed Plants
. 34
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Figures continued 3.2 Summary of Three Unimplemented USIs...........
,....,............. 36 4.1 Generic Safety issues Implementation Status at Licensed Plants 53 4.2 Summary of Seven Unimplemented GSis......
.... 55 5.1 Other MPA lssues - Implementation Status at Ucensed Plants,
.............. 74 5.2 Summary of Six Unimplemented MPAs.
. 75 TABLES 2.1 Summary of the Remaining TMl items by Area.............................. 10 2.2 Summary of the Remaining TMl items by Plant
....... 12 2.3 Status of TMI Action Plan - Summary by Plant
. 14 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 USl items Requiring Verification....
... 39 3.3 Catus of USl Plants - Summary by Plant....
.........................42 3.4 Status of USl Plants Summary by item.....
.. 46 4.1 GSI Numbers and Corresponding SIMS Item Numbers.....
52 4.2 Summary of Unimplemented GSI Items by Plant.
. 54 4.3 lemporary Instructions for Resolved GSis....
, 58 4.4 Summary of GSI Items Requiring Verification..
... 59 4.5 Status of GSI Plants Summary by Plant
.. 62 4.0 Status of GSI Plants Summary by item.
. 66 5.1 SIMS lssue Numbers and Corresponding MPA Number
.... 72 5.2 Summary of Unimplemented MPA Items by Plant...,
.......... 73 5.3 Temporary Instructions for Resolved MPAs........................... 93 5.4 Summary of Other MPA Items Requiring Verification..
... 94 5.5 Status of Other MPA Plants - Summary by Plant.....
.. 96 5.6 Status of Other MPA Plants Summary by item 100
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EXECUTIVE
SUMMARY
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This U.S. Nuclear Regulatory Com' mission (NRC) NUREG report covers the j
implementation and verification status of the Three Mile Island (TMI) Action Plan requirements, unresolved safety issues (USis), generic safety issues (GSis), and other-4 multiplant action (MPA) issues not related to TMI Action Plan requirements, USis, or GSis at 110 licensed nuclear power plants. The implementation and verification status i
l are current as of September 30,1992.
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Backoround h
The implementation and verification status of TMI Action Plan requirements, USis, and GSis was initio'ly reported in Volumes 1, 2, and 3 of NUREG 1435, published in 1991.
The first annual NUREG report, Supplement 1, consolidated and updated the information provided in the earlier three volumes l it was published in December 1991' and provided updated information as of September 30,1991. This second NUREG report, Supplement 2, provides updated information as of September 30,1992, and also provides the status of licensee implementation and NRC verification of other t
multiplant action (MPA) _ issues not related _to TMI Action Plan requirements, USis, or GSis. The data contained herein is a product of the NRC's Safety _lssues Management System (SIMS), which is maintained by the project management staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. The NHC has given significant attention to the quality review of TMI, USI, GSI, and other MPA implementation and verification data in SIMS.
Three Mile Island Action Plan Reauirements Supplement 1 of this NUREG series reported data on TMI Action Plan requirements for 111 plants, including Yankee Rowe which is now permanently shut down. For the purpose of this report, Yankee Rowe has been excluded from the data on the _
Implementation and verification status of TMI Action Plan requirements as well as USIs, GSis and other MPAs.
Imolementation Status.- More than 99 percent of the TMI Action Plan items have been implemented at 110 licensed plants. Of the 13,408 applicable items,13,322 have been completed or closed and only 86 remain open from an implementation standpoint.
About 47 percent of the remaining 86 open items are projected to be implemented by
- the end of calendar year 1993. As noted in Supplement 1,'some slippages have occurred in projected implementation dates. Delays in the restart of Browns Ferry Units 1 and 3 account for 34 of the unimplemented items. All schedules for implementation of TMI Action Plan items remain within the timeframe previously-reported to the Commission and to Congress.
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From an issue perspective, three major areas account for about 68 percent of the 86 remaining items: detailed control room design review items (39); plant safety parameter display system items (10), and accident monitoring (10).
From a plant perspective, the TMI Action Plan requirements have been fully implemented or closed at 62 of the 110 licensed plants. Three plants account for approximately 44 percent of the 86 remaining items: Browns Ferry Units 1,2 and 3 (37 items). The rest of the plants have no more than two items each remaining to implement.
Verification Status. Seventy-eight Tis have been issued to provide guidance for the field verification of licensee implementation. Of the 6,055 TMI items requiring verification,5,995 (99 percent) have been completed.
Unresolved Safety Issues (USIs) j imolementation Status. Approximately 88 percent of the USI items have been implemented at licensed plants. Of the 1,803 applicable items,1,576 have been completed and 227 remain van from an implementation standpoint. On average, each plant has approximate) 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 87 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.
I Verification Status. Four 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, l
Qualification of Class 1E Safety-Related Equipment; and A-26, Reactor Vessel Pressure Transient Protection. Licensee implementation of A-44, Station Blackout, will require NRC field verification.
The requirements to perform field verifications have resulted in a total of 425 items to be verified at the 110 p! ants. As of September 30,1992, the NRC field verification had i
been completed on 291 (68 percent) of the 425 required items.
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Generic Safety Issues (GSis)
Imolementation Status. Approximately 90 percent of the applicable items associated with GSis have been implemented at licensed plants. Of the 2,647 applicable items, 2,389 have been completed and 258 remain open from an implementation standpoint.
On average, each plant has fewer than 3 items to implement, and no plant has more than 7 items to implement. Fourteen plants have implemented all applicable items related to GSis. Seven GSts account for 95 percent of the items for which implementation is incomplete. These GSis are specifically addressed in Section 4.1 of this report.
Verification Status. Seven Tis have been issued to provide guidance for the field verification of licensee implementation. Of the 1,080 GSI items requiring verification, 1,039 (97 percent) have been completed.
Other Multiolant Actions (MPAs)
Imolementation Status, Approximately 84 percent of the applicable items associated with MPAs have been implemented at licensed plants. Of the 7,175 applicable items, 6,197 have been completed and 978 remain open from an implementation standpoint.
On average, each plant has fewer than 9 remaining items to implement, and no plant has more than 14 itemt, to implement except Browns Ferry Units 1 and 3. Each of these units has 22 items. No plant has implemented all applicable items related to other MPAs. Six MPAs account for 62 percent of the items for which implementation is incomplete. These six MPAs as well as those with more than three open items are specifically addressed in Section 5.1 of this report.
Verification Status. Thirteen Tis have been issued to provide guidance for the field verification of licensee implementation. Of the 720 MPA items requiring verification, 587 (80 percent) have been completed.
Conclusions After a detailed review of the implementation and verification status of TMI Action Plan requirements, USIs, GSis, and other MPAs, the NRC staff has come to the following conclusions:
The NRC closure process for TMI Action F,an issues, USIs, GSis, and other MPAs is adequate for protecting the public health and safety.
Licensees continue to make progress in implementing actions that are voluntary or
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imposed or requested by the staff.- The framework exists to verify future implementation of delayed items.
The NRC continues to make progress in verifying the implementation actions reported complete for those issues that require verification.
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The schedule slippages related to implementing TMI Action Plan items do not pose 1
a threat to the public health and safety. The NRC staff will conti.1ue 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, 1
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r ABBREVIATIONS ACRS Advisory Committee on Reactor Safeguards ATWS anticipated transient without scram -
BL bulletin B&O bulletins and orders -
B&W Babcock and Wilcox l
BWR boiling water reactor-l BWROG Boiling. Water Reactors Owners Group 4
CE Combustion Engineering i.
. containment performance improvement CRGR Committee for the Review of Generic requirements DBA design-basis accident
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ECCS emergency core cooling system 1-GIP generic implementation procedure j
GL-generic letter l
GSER generic safety evaluation report i
GSI generic safety issue HPCI high-pressure coolant injection i
IGSCC intergranular stress corrosion cracking j
IN information notice (NRC) i IPE individual plant examination I
IPEEE individual plant examination of external events j
IST inservice testing l
LCO limiting conditions for operation l
MOV motor-operated valve MPA multiplant action j(
NRC U.S. Nuclear Regulatory Commission i
NRR Office of Nuclear Reactor Regulation (NRC) l NUMARC Nuclear Management and Resource Council ODCM Offsite Dose Calculation Manual
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- process control program--
PORV power-operated relief valves j
PWR pressurized-water reactor PZR pressurizer l
RCIC reactor core isolation cooling RCS reactor coolant system
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' RES Office of Nuclear Regulatory Research (NRC)'-
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- radioactive effluent technical' specifications i
RG regulatory guide I
- RWCU reactor water cleanup L
SBL supplement bulletin L
SIMS safety issues management system 1
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SOER significant operating experience report SPDS safety parameter display system SOUG Seismic Qualification Utility Group SRM staff requirements memorandum SSER supplementary safety evaluation report Tl temporary instruction TMI Three Mile Island TS technical specifications TU Texas Utilities (Electric)
USl unresolved safety issue VlB vital instrument bus i
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i 1 INTRODUCTION This second annual NUREG report, Supplement 2, updates the implementation and verification status of the Three Mile Island (TMI) Action Plan requirements, unresolved safety issues (USIs), and generic safety issues (GSis), and provides the status of other multiplant actions (MPA). The NRC previously published three volumes of this NUREG series. Volume 1, published in March 1991, discussed the status of TMI Action Plan requirements. Volume 2, published in May 1991, identified the implementation and verification status of actions associated with US!s. 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. This second annual NUREG report, Supplement 2, provides updated information as of September 30,1992. In addition, this supplement also provides the status of licensee implementation and NRC verification of MPA issues not related to TMI Action Plan requirements, USIs, or GSis. 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 110 licensed plants in the United States and makes this information available to interested parties, including the public. Supplement 1 of this NUREG series reported data on TMI Action Plan requirements for 111 plants, including Yankee Rowe which is now permanently shut down. For the purposes of this report, Yankee Rowe has been excluded from 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 110 licensed plants, there are 13,408 applicable items for TMI Action Plan issues,1,803 for USIs,2,647 for GSis, and 7,175 for other MPAs. A total of 25,033 applicable items are addressed in this report. The information presented in this volume is current as of September 30,1992.
1.1 Backoround 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 this supplement. Actions to be taken by licensees in response to these generic issues apply to more than one plant.
The NRC evaluates the status of each licensee's implementation in conjunction with its evaluation of other NRC requirements, unique plant considerations, and interim 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 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 safety.
The data contained in this NUREG report are a product of the NRC's Safety issues Management System (SIMS), which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation (NRR) and by NRC regional personnel. The NRC has given significant attention to the quality review of TMI, USI, GSI, and other MPA implementation and verification data in SIMS.
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:
Identifving Relevant issues - Generic concerns are typically identified by the NRC 4
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, licensees, vendors, the Advisory Committee on Reactor Safeguards (ACRS), and the public. The NRC staff identified the TMI requirements by compiling and evaluating information from all available sources concerning the accident at TMI.
Prioritizing issues - Once identified, an issue is evaluated by the staff for its potentialimportance 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.
Resolving issues - The staff evaluates corrective actions that might be taken to 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.
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Imoosino Reouirements (USIs and TMI Action Plan Reouirements)- Each affected licensee or applicant is required to prepare a schedule for implementing the resolution consistent with a rule, policy statement, regulatory guide, generic letter, bulletia, 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.
Reouestina Action (GSis)-The NRC staff evaluates the importance of an issue i
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 is resolved, each action to be implemented is assigned a multiplant action (MPA) number for tracking purposes.
Imolementino Actions - Licensees of affected plants take corrective actions to satisfy commitments made in response to the imposed requirements (TMI Action Plan requirements and USts) or the staff's 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 TMl Action Plan requirements or USis 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.
Verifvino 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. 4 L_
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SIMS is designed to track issues from their identification throu0h implementation of associated actions and field verification. The NRC Project Manager periodically obtains data periaining to the licensee's implementation dates from meetings, site j
visits, and discussions with resident inspectors or the licensee. Recent NRC initiativos i
to improve the quality and the accountability of data include requirements that (1) any conclusion that a corrective action has been implemented be accompanied by a referenco document from the licenseo staff providing the basis for closure of the issuo i
at the particular plant, and (2) the inspection report number and the date of the i
inspection be ontored into SIMS if verification is required.
1.3 DpfioltiODS l
A number of terms are used to describe generlo issues and their status. These terms are important not only because they categorize issues and their origin, but becauso j
their use implies both applicability and degree of completeness. For the purposes of this report, the following definitions apply:
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. Generic Safety 1swe (GSI) - A safety concern that affects the design, construction, i
or operation of all, several, or a class of nuclear power plants and may have the potential for safety improvements at such plants.
J EUR!amented item - An Itt is implemented when a licenseo has completed the j
activities necessary to sanc, the requirements (or assumptions) made in the staff's technical resolutbn in accordance with commitments concerning the generic issue.
Item - The application of a 1MI Action Plan requirement, USl, GSI or other MPA
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issue to a specific plant.
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.MPA A multiplant action item originates from industry experience, new l
regulations / requirements, or from resolution of generic issues resulting in the issuance of a generic communication requiring action by the licenseos. TMl 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 j
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, USl, GSI or other MPA) to all 110 plants.
Total Aoplicable Plant items - The total number of applicable items determined by reviewing the applicability of each issue at each of the 110 licensed plants.
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Unimolemented item - An item is unimplemented when a plant has not completed the activities necessary to satisfy the actions requetted or required by the staff following the resolution of a particular generic issue.
Unresolved Safety Issue fUSJ)- A matter affecting a number of nuclear power
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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.
Verification Comoteted - A licensee's actions to implement a technical resolution
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for a generic issue have been inspected and verified by the NRC in accordance with the guidance provided by the applicable ternporary instruction for that issue, 4
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2 THREE MILE ISLAND ACTION PLAN REQUIREMENTS This section describes the overall status of implementation and verification of TMI Action Plan items at the 110 currently licensed plants. Supplernent 1 of NUREG 1435 reported the status of 111 licensed plants, including data on Yankee Rowe, which is now permanently shut down. For the purposes of this report, Yankee Rowe is not included in the data on the implementation and verification status of TMI Action Plan requirements.
2.1 Imolementation Stplus More than 99 percent of the TMI Action Plan items have been implemented or closed at licensed plants. Of the 13,408 items,13,322 have been completed and only 86 have not yet been implemented. Figure 2.1 presents the overall status of implementing the TMI Action Plan requirements. The 13,322 items completed at the 110 licensed plants have been disposed of as follows:
A total of 12,926 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 been effectively addressed by other items or through other regulatory means. The superseded items are discussed in detail in Volume 1 of NUREG 1435.
The following observations are made about the remaining 86 unimplemented iterns:
Approximately 47 percent of these items are projected to be implemented by the end of calendar year 1993, as shown in Figure 2.2. Licensees continue to make progress toward implementation of the remaining items.
As noted in previous status reports, some slippages have occurred in projected implementation dates. Delays in the restart of Browns Ferry Units 1 and 3 (34 items), along with the rescheduling of refueling outages at other plants account for a large number of the slippages in the implementation of remaining TMI items.
Browns Ferry Unit 1 has 15 TMl items that will not be implemented until the end of calendar year 1996. All schedules for implementing the remaining TMI Acticn Plan items are within the timeframe previously reported to the Commission and to Congress..
From an issue perspective, three major areas account for about 68 percent of the 86 unimplemented items, as shown in Table 2.1: detailed control room design review items (39), safety parameter display system items (10), and accident monitoring (10).
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_ _ - _.... _.. _ _ _... _. _ _. _ _ _. _ _ = _ _ _.. _ _ _ _. _.
I C
TMI ACTION PLAN REQUIREMENTS IMPLEMENTAT!ON STATUS AT UCENSED PLANTS i
i f
I I
l 25000-300 I
OVERVIEW
.- 2M v
19092 250 --
m g
200 -.
180 g
i y 15000,.
tM 13322 155 3g 6
ld
$E150--
t j
118 D
1 icooo.,
g-105 i
2 100 -.
,86
{
a 2
5000-50 _.-
f a
f I
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I f
f I
I O
4 4
4 4
i e
i 1
6 6
I e
i TOTAL
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.LE' June 1989 9009912f31/99 3/30fB0 7G1/90 Sf30sc 12/31/90 2/2691 EGorgt ecoret 1131/B2 3G1/92 &~ t/92 9/3WB2 DATEOFSTATUS REPORT
- These totals do not include items for Rancho Seco, Shoreham, Ft. St. Vrain and Ywkoe Rowe plants. These plante are
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permanentfy or indefinitely shut domm. The total number of l
ficensed plants conssdered in this report is110.
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i PROJECTED SCHEDULES FOR REMAINING TMI ITEMS ITEMS NOT IMPLEMENTED AT END OF CALENDAR YEAR 100 i
86 JUNE 1969 I
l SCHEDULES f
R CURRENT l
SCHEDULES f
o 75 --
68 I
N I
5 57 g
I I
f f
50 -.
45 O
b 3'
l 5'
i 25 _.
19 18 I
16 15 14 12 0_
l l
... l
-l l
l 09/92 12/92 12/93 12/94 12/95 12/96 4
I l
09/92 12/92 12/93 12/94 12/95 12/96 l
JU~ 41989 SCHEDULES
- 57 31 19 15 14 12
)
)i 09/30/92 SCHEDULES
- 86 68 45 18 16 0
l 1
l I
- Based on dates of items listed as not implemented I
Figure 2.2
SUMMARY
OF THE REMAINING TMI ITEMS BY AREA j
AREA OPEN 4
o Plant SPDS Console 10 4
i 4
I o Post-Accident Sampling 5
o Relief & Safety Valve Test Report 2
l o Containment isolation Dependability 4
t o Accident Monitoring 10
- o..
o Instrumentation for Detection of inadequate Co _: C.v: Pug 5
l i
o B&O Task Force issues 8
r t
I l-o Technical Specifications - NUREG-0737 2
o Detailed Control Room Design Review 39 l
s TOTAL as i
Table 2.1
From a plant perspective, the TMI Action Plan has been fully implemented or closed at 62 of the 110 licensed plants. Table 2.2 summarizes the 86 unimplamented items by plant. Three plants account for approximately 44 percent of the remaining 86 items: Browns Ferry 1,2, and 3 (37 items). Four plants have 2 items and 40 plants have 1 item remaining to implement.
Appendix A lists the unimplemented TMI Items by issue and gives projected implementation dates.
2.2 Verification Status For generic items, such as ;he 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 Tl, 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 6,551 items at the 110 licensed plants. Upon initial inspection of certain items and further review by the regional offices,496 items covered by the Tis were found to be inapplicable from a serification standpoint, leaving a net total of 6,055 items requiring _
verification. The mtjority 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 30,1992,5,995 items (99 percent) had been verified. Only 60 items remain to be verified, including some items not yet implemented by licensees.
I i
I
SUMMARY
OF THE REMAINING TMI ITEMS BY PLANT I
PLANT OPEN-PLANT OPEN PLANT OPEN Arkansas 1
'1 Haddam Neck 2
Prairie Island 1 1
i Arkansas 2.
1 Hatch 1 1
Prairie Island 2 1
(
l Big Rock Point 1
Hatch 2 1
Quad Cities 1 1
Beaver Valley 1 1
Millstone - 1 1
Quad Cities 2 1
Browns Ferry 1-15 Millstone 2 1
Salem 1 1
l j
Browns Ferry 2 3
Nine Mile Pt 1 2
Salem 2 1
i j'
Browns Ferry 3 19 North Anna 1 1
San Onofre 2 1
l j-Brunswick 1.
1 North Anna 2 1
San Onofre 3 1
Brunswick 2.
1 Oconee 1 1
Sequoyah 1 1
l Calvert Cliffs 1 1
Oconee 2 1
Sequoyah 2 1
j Calvert Cliffs 2 1
Oconee 3 1
Surry 1 1
Crystal River 3 1
Palo Verde 1 1
Surry. 2 1
1
- Diablo Canyon 1 1
Palo Verde 2 1
Zion 1 2
Y Diablo Canyon 2 1
Palo Verde 3 1
Zion 2 2
Dresden 3 1
Pilgrim 1 1
i Fermi.2 1
Point Beach:1 1
Ft Cafhoun-i 1
Point Beech 2 1
4 L
i~
l I.
i i
i I
i i
?
i Table 2.2 f
f
+,.
2.3 Status by Plant 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 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 has been fully implemented at 62 of the 110 licensed plants. Browns Ferry Units 1,2, and 3 (37 items) account for approximately 44 percent of the 86 remaining open items.
From a verification standpoint, all required inspections have been completed at 75 of the 110 licensed plants. Browns Ferry Units 1 and 3, have 10 items each that will require verification following implementation by the licensee. Thirty three plants have 2 or less items remaining to be verified. _
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTIOu FLA4T - SLN"AUlf ET PLMT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED
-=---- -
ARMANSAS 1 122 121 (99 }
1 62 61 60 (98 ARMANSAS 2 112 111 (99 )
1 59 58 56 (96 BEAVER VALLEY 1 117 116 (99 l 1
62 61 61
< 100 BEAVER VALLEY 2 126 126 (200l 0
62 57 57 l 100 i
BIG ROCK POINT 1 104 103 (99 y 1
56 50 50 1 100 i
BRAIDWOOD 1 126 126 (100) 0 62 56 56 i 100 t
BRAIDWOOD 2 126 126 (100) 0 62 56 56 l 100 l
BROWMS FERRY 1 110 95 186 )
15 ST 51 41 1 to
'l i
BROWMS FERRY 2 110 107 (97 )
3 57 51 50 l 98 BROWNS FERRY 3 110 91 (82 1 19 57 51 41 i 30 i
BRUNSWICM 1 110 109 (99 )
1 57 51 51 1 100 1
BRUNSWICM 2 110 109 (99 )
1 57 51 51 l 100 i
^
BYRCN 1 126 126 1.100) 0 62 58 SE (100 i
i BYRON 2 126 126 1 1001 0
62 58 58 I100 i
CALLAWAY 1 123 123 l 100 1
0 60 60 CO l 100 CALVERT CLIFFS 1 113 112 I 99 l
1 59 53 53 1 100 CALVERT CLIFFS 2 113 112 1 99,,
1 59 53 53 i 100 i
CATAWBA 1 129 126 (200) 0 62 60 59 1 98 CATAWEA 2 126 126 l100) 0 62 62 61 1 98 CLINTON 1 120 120 1 100) 0 56 56 56 1 100 COMANCHE PEAM 1 119 119 1 100) 0 55 50 50 l 100)
C00M 1 117 117 1 1001 0
62 55 55 1 100 1
C00M 2 117 117 1 100) 0 62 55 55 1 100 COOPER STATION 110 110 1 100) 0 57 31 51 1 100 Ce 4L RIVER 3 122 121 l 99 y 1
62 56 54 1 96 DA - BESSE 1 121 121
- 100) 0 61 55 55 1 100 t
DImAV CANYON 1 126 125 (99 l
1 61 55 55 l 100 i
DEASLO CANYON 2 126 125 (99 l
1 61 57 57 l 100 DRESDEN 2 110 110 1 100 1
0 58 52 51 (98 i
DRESDEN 3 110 109 l 99 i
1 58 52 51 198 DUANE ARNOLD 110 110 1 100) 0 57 54 54 (100 I
FARLEY 1 118 118 1 100) 0 62 56 56 (100 t
FARLEY 2 128 128
! 100) 0 62 56 56 (100 1
FERMI 2 120 119 1 99 )
1 SS 50 50 (200 t
FIT 2 PATRICK 110 110 1 100) 0 57 56 55 (98 i
FORT CALHOUM i 113 112 1 99 )
1 59 55 54 98 l
GINNA 116 116 1100) 0 61 56 58 100)
GRAND GULF 1 120 120 (100) 0 56 50 50 100 t
HADDAM NECM 117 115 (98 )
2 62 55 53 (96 HARRIS 1 125 125 (100f 0
61 60 60 1100 MA TCH 1 110 109 (99 l
1 57 57 57 (100?
HATCH 2 110 109 (99 l
1 57 57 56 (98 )
HOPE CREEK 1 120 120 (100 1
0 56 50 50 (100 INDIAN POINT 2 118 118 (100l 0
62 59 59 (100 Table 2.3
m..
.m SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAR - SlPMARY BT PLMT l
IMPLEMENTATION VERIFICATION
........-- - = - - - -___...__...............
ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PE F CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED INDIAN POINT 3 117 117 (100) 0 62 59 59 1100)
I MEWAUNEE 117 117 (100) 0 62 58 57 (98 )
LASALLE 1 120 120 (2001 0
56 52 52 (100)
LASAttt 2 120 120 (1001 0
56 52 52 (100)
LIMERICK 1 120 120 (100) 0 56 52 52 (1001 i
LIMERICK 2 120 1I0 (1001 0
56 50 50 (100)
MAINE YANKEE 113 113 (2001 0
59 56 56 (100) 1001 0
62 62 61 (98 )
MCGUIRE 1 127 127
{100) 1 0
62 62 61 (98 )
MCGUIRE 2 127 127 1
MILLSTONE 1 109 108 (99 l 1
56 46 45 (9T I MILLSTONE 2 113 112 (99 I 1
59 54 54 1100) i MILLSTONE 3 127 127 (1001 0
62 58 58
{100)
MONTICELLO 110 110 (1005 0
57 51 51 (100) g MINE MILE POINT 1 107 105 (98 )
2 56 54 54 (100)
NINE MILE POINT 2 119 119 (2001 0
55 52 52 (2001 NORTH ANNA 1 118 117 (99 1 1
62 60 60 (100)
NORTH ANNA 2 128 127 (99 l 1
62 60 60 (100i
[
OCONEE 1 122 121 (99 )
1 62 57 56 (98 i
OCONEE 2 122 121 (99 )
1 62 58 57 (98 i
DCONEE 3 122 121 1 99 )
1 62 58 57 (98 l
100) 0 56 47 47 (100l OYSTER CREEM 1 107 107 PALISADES 113 113 1100) 0 59 51 51 (100) l PALO VERDE 1 120 119 (99 l 1
59 52 52 (100)
PALO VERDE 2 120 119 (99 )
1 59 53 53 (1003 PALO VERDE 3 120 119 (99 )
1 59 54 54 (1001 PEACH BOTTDM 2 110 110 (100)
O 57 51 51 (1001 PEACH BOTTOM 3 110 110 (100) 0 57
-51 51 (100)
O 56 55 55 (200)
[
( 100 l*
PERRY 1 120 120 (99 1
57 48 48 (100)
PILGRIM 1 110 109 POINT BEACH 1 117
.116 (99 l 1
62 56 56 (100) j POINT BEACH 2 117 116 (99 3 1
62 56 56 (100) j PRAIRIE 15 LAND 1 117 116 (99 1 1
62 52 52 (teo)
PRAIRIE ISLAND 2 117 116 (99 1 1
62 53 53 (1001 OUAD CITIES 1 110 109 (99 9 1
57 51 50 (98 l
QUAD CITIES 2 110 109 (99 3 1
57 51 50 198 (100[3 -
0 56 50 49 (97 l
(100 RIVER BEND 1 119 119 0
62 56 54 (96 l
ROBINSON 2 117 117 SALEM 1 116 115 (99 l 1
61 54 52 (96 i
SALEM 2 127 126 (99 y 1
62
$7 55 (98 )
SAN ONOFRE 1 116 116 (1001 0
61 58 56 (96 )
SAN ONOFRE 2 122 121 (99 )
1 59 53 53 (1GP) 5AN ONOFRE 3 122 121 199 )
1 59 55 55 (2001 4
SEA 8 ROOM 1 127 127 (100) 0 62 57 57 (100)
SEQUOYAH 1 127 126 (99 1 1
62 56 55 (98 )
4 Table 2.3
w_.
.. - - - - - =. ~
_. - - _ - _. - ___. _~
i i
SAFETY ISSUE MANAGEMENT SYSTEM L
l STATUS OF TMI ACTICN PLANT - SLNPMT BT PLAMT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED SEQUOYAH 2 127 126 (99 )
1 62 58 58 (100r SOUTH TEMAS 1 126 128 11003 0
62 56 56 (100 i
SOUTH TEXAS 2 126 126 l 100) 0 62 56 56 (100 ST LUCIE 1 113 113 1 100) 0 59 55 55 100 t
1 St LUCIE 2 122 122 l 100) 0 59 54 54 1 100 t
SUMMER 1 127 127 F100) 0 62 61 61 i 100 t
SURRY 1 117 116 1 99 )
1 62 56 56 1 100 i
SURRY 2 117 116 (99 )
1 62 56 56 (100 i
SUSOUEHANNA 1 120 120 (100) 0 5e 56 56 (100 i
SUSCUEHANNA 2 120 120 (100) 0 56 56 56 (100 I
THREE MILE ISLAND 1 128 128 (100) 0 62 55 55 (100 1
TROJAN 118 118 (1003 0
62 58 58 (100 1
.A TURKEY POINT 3 117 117 (1003 0
62 60 59 (98 I
)
00 TURMEY POINT 4 117 117 1 100) 0 62 60 59 (98 i
l VERMONT YANKEE 1 110 110 1 1001 0
57 33 53 (100 V0GTLE 1 124 124 1 100) 0 60 54 53 (98 V0GTLE 2 124 124 (100) 0 60 56 55 198 WASHINGTON NUCLEAR 2 120 120 (200) 0 56 55 55 (100l WATERFORD 3 121 121 (Ico) 0 59 55 55 1100)
WOLF CREEK 1 126 126 (1003 0
60 54 54 (200)
ZION 1 117 115 (98 )
2 62 60 59 (98 )
ZION 2 117 115 (98 1 2
62 60 59 (98 )
TOTALS / AVERAGES 1301f 1292f 99 86 6551 6055 5995 99
-I
- Excludes 396 superseded items closed at the 110 licensed plants i
i' Table 2.3
A 2.4 Status by Issue i
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 i
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.
j Of the 172 TMI Action Plan issues,144 have been fully implemented and 4 have been superseded. Three categories of TMl Action Plan issues account for about 68 percent of the TMI requirements to be implemented: detailed control room design review, with 39 plants yet to complete implementation; the safety parameter display system, with 10 plants remaining to complete implementation; and accident monitoring, with 10 plants remaining to complete implementation. The next largest contributor is bulletins and orders (B&O) task force issues, still open at 8 plants.
a
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN SUP9GRY BY ITEM IMPLEMENTATION W RIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAIN!ssG REQUIRED CCVERED REQUIRED Cos*LETED COMPLETED
)
87.4.1 73 T3 (2005 0
NO REACTCR COOLANT PUMP TRIP GL.s3-02' 110 110 (2001 0
NO MUREG-0737 TEC M ICAL SPECIFICATIONS CL-83-36 110 108 (98 J 2
NO MUREG-0737 TECHNICAL SPECIFICATIONS I.A.1.1.1"-
110 110 (100) 0 YES 110 lie 110 (190)
SHIFT TECHNICAL ADVISOR - 04 DUTY a
CD I.A.1.1.2 lie.
110 (100) 0 too SHIFT TECNeeICAL ADVIsoft -- TECM SPECS I.A.1.1.3 110 110 (100) 0 YES 110 110 110 (1903 SHIFT TECHNICAL ADVISOR - TRAINED PER LL CAT 8 1.A,1.1.4 110 110 f100f 6
NO SHIFT TECHeeICAL ADVISOR - DESCRIBE LoseG. TERM PROGRAM l
1.A.1.2-110 110 (1003 0
YES 110 110 110 (1903 SHIFT SUPERVISOR RESPONSIBILITIES I.A.1.3.1 110 110 (1C0) 0 YES 110 110 110 (2003 SHIFT MANNING - LIMIT OVERTIPE I.A.I.3.2 110 110 (1001 0
YES 110 110 110 (100)
SHIFT M48eNING - MIN SHIFT CREW I. A _ 2.1 '.1 110 110 (2001 0
88 0 IP94EDIATE UPORADING OF RO & SWO TRAINING AteD OUAL. - SRO EXPER.
I.A.2,1.2 110 110 (1001 0
NO It**EDIATE U> GRADING OF RO & SRO TRAINIIeG AND QUAL.
SRO*S BE RO'S AYR I.A.2.1.3-110-110 (2001 0
NO IMMEDIATE UPORADING OF RO & SRO TRAINING AND QUAL. - 3 MO. TRAINING I A 2.1. 4 :
110
. 110 (1001 0
'YES 110 110 t10 (100) 199EDI ATE UPORADING OF RO & SRO TRAINIIOG AND QUAL. - MODIFY TRAINIIeG I.A.2.1.5-110 119 (1001 0
NO
'It**EDIATE UPGRADIIeG OF RO & SRO TRAINING AseD QUAL.. F ACILITY CERTIF.
Table 2.4
______--.__.___,_m,_.
9 k'
l t
l i
1 t
i SAFETY ISSUE MANAGEMENT SYSTEM l
I f
STATUS OF TMI ACTION PLAN - StRe#RY BY ITEM i
i IMPLEMENTATIDW VERIFICATION
[
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT f
ITEM APPLICABLE COMPLETED COMPLETED REM 4INING REQUIRED COVERED REQUIRED COMPLETED COMPLETED
==
t I.A.2.3 110 110 (1001 0
WO I
ADMINISTRATION OF TRAINING PROGRAMS
}
I I.A.3.1.1 110 110 (1001 0
NO l
REVISE SCOPE & CRITERIA FOR LICENSING EXAMS - INCREASE SCOPE 4
I. A. 3.1. 2 '
110 110 (1001 0
NO REVISE SCOPE & CRITERIA FOR LICENSING EXAMS. INCREASE PASSING GRADE j
t I.A.3.1.3.A
' T9 79 (1003 0
NO REVISE SCOPE & CRIT. FOR LIC. EXAMS - SIMULATOR PLANTS WITH SIMULATORS j
i I.A.3.1.3.8 32 32 (1001 0
NO g)
REVISE SCOPE & CRIT. FOR LIC. EXAMS. SIMULATOR - OTHER PLANTS i
l I.8.1.2 49 49 '
(1001 0
NO FVALUATION OF ORGANIZATION & MANAGEMENT 3
1.4.1.1 110 110 II I YES 110 110 110 (190)
SHORT-TERM AOCIDENT & PROCEDURES ItEVIEW - $8 LOCA I.C.1.2.A 119 110 11001 0
YES IIS 93 93 f100)
SHORT. TERM ACCIO. & PROCEDURES REV. - INADEQ. CORE COOL. REANAL. GUIDE 3
I.C.1.2.8 110 110 (2003 0
YES 119 110 110 (100)
SHORT-TERM ACCID.~& PROCEDURES REV. - INADEO. CORE COOL. REVISE PROCED I.C.1.3.A 119 110 (100) 0 YES 119 93 93 (100)
'SHORT-TERM ACCID. & PROCEDURES REV - TRANSIENTS & ACCDTS. REAMaL GUIDE i
I.C.1.3.8 119 110' (100) 0 YES 119 119 110 (100)
SHORT. TERM ACCID. & PROCED8MtES REV. - TRANSIENTS & ACCDRS. REVISE PROC r
I.C.2 119 110 (100) 0 YES 119 119 119 (199)
SHIFT & RELIEF TURNOVER PROCEDURES
'bl I.C.3 110 110 1100) 9 YES 119 119 110 t100I i
SHIFT-SUPERVISOR RESPONSISILITY i
I.C.4 119 110 11003 0
YES 119 119 10 t199) l CONTROL-ROOM ACCESS t
1 4
I,C.5 119 110 11001 0
YES 110 110 110 1100)
[
}
FEEDBACK OF OPERATING EXPERIENCE t
i Table 2.4 i
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$AFE7Y I5 SUE MANAGEMENT SYSTEM I'
(
STATUS OF Tt*I ACTION PLAN - Supe 4ARY ST ITEM 1
j i
IMPLEMENTATION VERIFICATION 4
...............--- ------ = - -. -------.......
PLANTS PLANTS.
PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED I
4 II.8.2 3 110 110 (1003 0
YES 119
.110 110 (100)
PLANT SHIELDING - PLANT MODIFICATIONS (LL CAT 81 II 8 3.1 109 109 (1003 0
.YES 109 103 103 (100)
POSTACCIDENT SAMPLING - INTERIM SYSTEM 1
11.8.3.2-110 109 (99 )
1 NO POSTACCIDENT SAMPLING - CORRECTIVE ACTIONS
.11.8.3.3 109 107 (98 3 2
NO i
i s
. POSTACCIDENT. SAMPLING ~ PROCEDURES t
i e
]
- FC 11.8.3.4 110 108 (98 1 2
YES 119 110 107 (ST 1 y
POSTACCIDENT SAMPLING. PLANT MODIFICATIONS (LL CAT 8) r 1
II.8 4 1 110
. 110 (1001 0
NC
{'
TPAINING FOR MITIGATING CORE DAMAGE - DEVELOP TRAINING PROGRAM t'
7 II.8.4.2.A 110 110 (1001 0
YES 119 110 110 (100)
TRAINING FOR MITIGATING CORE DAMAGE - INITIAL 11.8.4.2.8
'110 110 (1001 0
YES 119 119 110 (1001 t
F TRAINING FOR MITIGATING CORE DAMAGE - COMPLETE II.D.1.1-110 110 (1001 0
WO RELIEF & SAFETY VALVE TEST REQUIREMENTS - SUEMIT PROGRAM II.D.1.2.A 119 110 (100)
O NO RELIEF & SAFETY. VALVE TEST REQUIREMENTS -. COMPLETE TESTING l
II_D.1.2.8 119 109 (99 )
'1
' No RELIEF & SAFETY VALVE TEST REQUIREMENTS - PLANT SPECIFIC REPORT i
RELIEF & SAFETY VALVE TEST REQUIREMENTS - BLOCK-VALVE TESTING
' No II.D.1.3 TO SS 195 )
1 i
i I
II.D.3 1 119
- -110 (1901 9
YES 119 119 119 (100)
VALVE PO'-1TIDst IWICATION - INSTALL DIRECT IWICATIONS OF VALVE POS.
II.D'3.2 11e 110 (100) e NO VALVE POSITICN INDICATION - TECH SPECS II.E.1.1.1 73 T3 (100) 0 NO
]
AFS EVALUATION-AWALYSIS i
i i
Table 2.4 i
'l i
l
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l t
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SAFETY ISSUE MANAGEMENT SYSTEM I
l f
STATUS OF TMI ACTION PLAN - St#844RY SY ITEM t
l.
IMPLEMENTATION VERIFICAT104
}
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = = = = - -
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT 1
1 ITEM APPLICAeLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED C W rTED COMPLETED II.E.1.1.2-72 72 (2001 0
YES 72 72 (190) l AFS EVALUATION-SHORT TERM MODS.
i II.E.1.1.3 73 73 (1001 0
YES 3
72-73 (190) 4 AFS -LONG TERM MODS.
II.E.1.2.1 A 87 87 (1001 0
YES 67 67 87 titel i
AFS INITIATION & FLOW-CONTRDL GRADE t
II.E.1.2.1.8 73 73 (Ice) 0 YES 73 73 73 (109)
AFS INITIATION & FLOW - SAFETY ORADE r
'U.
II.E.1.2.2.A 88 SS (1003' O
YES St SS 53 (199) i AFS INITIATION & FLOW - FLOW I W UCTIDW CONTROL GRADE l
II.E.1.2.2.8 73 73 (ISO) 0 NO AFS INITITATION & FLOW - LL CAT A TECM SPECS.
l 4
1.
II.E.1.2.2.C 73-73 4109) 0 YES 73 73 73 titel AFS INITITATIDW & FLOW - SAFETY ORADE i
i II.E.3.1.1 73 73 (1993 0
YES 73 73 73 (2001 y
EMERGENCY POWER FOR PRESSURIZER NEATERS - UPORADE POWER SUPPLY
[
4 II.E 3.1.2 73 73 (1001 0
40
[
E.
EMERGENCY POWER FOR PRESSURIZER NEATERS - TECH SPECS
- i i
II.E.4.1.1 109 109 (2003 0
WO i
i DEDICATED HYDROGEN PENETRATIDMS - DESIGN e
t II.E.4.1.2 199 109 (100) 0 YES 199 194 192 (98 3 DEDICATED HYDROGEN PENETRATIONS - REVIEW & REVISE M2 C04 Trot PROC II.E.4.1.3 199 109 (1001 0
YES 199 Its 194 (98 I DEDICATED MYDROGEN PENETRATION - INSTALL l
.1 II E.4.2.1-4 110 108 (St 1 2
YES 119 119 109 (99 I
}
CONTAIMENT ISOLATION DEPEWABILITY - IMP. DIVERSE ISOLATION j
II.E.4.2.5.A 110 110 (1001 0
NO i
f CONTAINMENT ISOLAT. DEPENDA81LITY - CNTMT PRESS. SETFT. SPECIFY PRESS.
i II.E.8 2.5.s 110 110 (1001 0
YES 11e let 10s ts9 )
[
j COMTAIsogENT ISOLATIDW DEPENDA6ILITY - CNTMT PRESSURE SETFT. MODS.
[
i i
Table 2.4 -
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SAFETY ISSUE MAMAGEMENT SYSTEM i
STATUS OF TMI ACTION PLAN -
SUMMARY
SV ITEM i
IMPLEMENTATION VERIFICATION
==-=---
..........=
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICA8LE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED 1
4 II,E.4.2.6 110 109 (99 1
YES 110 109 108 199 )
4 CONTAINMENT ISOLATION DEPENDABILITY - CNTMT '~JRGE VALVE II.E~4.2.7 110 110 (1001 0
YES 110 10F 106 199 3 i
CONTAINMENT ISOLATION DEPENDARILITY - RADIATION SIGNAL 04 PURGE VALVES COM d NMkMT ISCLATION DEPENDABILITY.-
H SPECS i
II.F.1.1 110 108 898 1 2
NO ACCIDENT-MONITORING - PROCEDURES d)
II.F.1.2.A 110 108 198 3 2
YES 119 119 110 (190)
[
j 00 ACCIDENT-MONITORING - NOBLE GAS MONITOR II.F.1.2.8 110 108
[9s ]
2 YES 119 119 110 (109)
ACCIDENT-MONITORING - IDDINE/ PARTICULATE SAMPLING II.F.1.2.C 110 108 198 1 2
YES 110 110 105 (98 )
[
[
ACCIDENT-MONITORING - CONTAINMENT HIGM-RANGE MONITOR II.F.1.2.D 110 109 (99 l 1
.YES 11 9 110 108 (St 1 f
ACCIDENT-MONITORING - CONTAINMENT PRESSURE
}
I II.T 1.2.E 110 109 (99 1 1
YES 119 110 110 (109)
ACCIDENT-MONITORINO - CONTAINMENT WATER LEVEL
(
II.F.1.2.F 119 110 (1001 0
YES 119 109 108 (99 I f
ACCIDENT-MONITORING - CONTAINMENT HYDROGEN II.F.2.2 T3 73 (1001 0
YES T3 T3 70 (95 I l
i INSTRUMENTATION FOR DETECT. OF INADEQUATE CORE COOLING - SU6C00L METER L
4 II.F.2.3 110 110 (2001 0
NO l
INSTRUMENTATION FOR DETECT. OF INADEQUATE CORE COOLING - DESC. OTHER
{
II.F.2.4 109 104 195 )
5 YES 109 1DS 94 (88 I l
INSTRMNTATM FOR DETECT. 17 INADEO CORE CLNG INSTLL h *L INSTRMNTATM i
II.G.I.1 73 73 f1001 0
YES T3 73 72 (98 )
POWER SUFP. FOR PRESSURIZER RELIEF.5 LOCK VALVES & LfVEL IMO.- UPGRADE i
11.0.1.2 73 T3 (1001 0
WO POWER SUPP. FOR PRESSURIZER RELIEF. BLOCK VALVES & LEVLE IMD.- TECM SP.
Table 2.4
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!j-IMPLEMENTATIDW VERIFICATION
_--- = - - -.......... - - - ---
{
PLANTS PLANTS PER ' CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED t
II.M.1.21 1
1 11001 0
NO IE BULLETINS - AUTO SO ANTICIPATORY REACTOR TRIP II.M 1.22 14 14 (1001 0
NO 4
IE BULLETINS - AUX. HEAT REM SYSTM, PROC.
l:
II.M.I.23 14 14 (1001 0
NO IE BULLETINS'- RV LEVE
L. PROCEDURE
S II.M 1.5 49 49 (1001 0
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?
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l M
II.M.2.10 7
7 (1001 0
YES 7
7 7
(2001 Y-ORDERS ON B&W PLANTS - SAFETY-ORADE TRIP l.
'II.M.2.11
?
7 (1001 0
YES 7
7 7
(100)
ORDERS ON B&W PLANTS - OPERATOR TRAINING i -
II M.2.13 71-71 1100]
O NO ORDERS ON B&W FLANTS - THERMAL MECMANICAL REPORT (CE & W PLANTS ALS0)
II.M.2.14 7
7 (1001 0
NO 4
ORDERS 04 B&W PLANTS - LIFT FREQUENCY OF PORV*S & SV'S II.M.2.15 7
7 (100) 0 NO CaDERS OM 84W PLANTS - EFFECTS OF SLUG FLOW -
II.M.2.16 7
. 7 (1001 0
- NO ORDERS ON 84W PLANTS. RCP SEAR DAM 4GE II.M.2.17 73 73 (1001 0
NO ORDERS CW 8&W PLANTS - VOIDING IN RCS (CE & W PLANTS ALSol 4
II M.2.19 7
7 (100)
-0 NO SENOHMARM ANALYSIS OF SEQUENTIAL AFW FLOW TO ONCETHROUGH STM OENERATOR II M.2.2 7
7 (100) 0 NO i
ORDERS ON B&W PLANTS - PROCEDURES TO CONTROL AFW IND OF ICS 3
II.M.2.20 7
7 (2003 0
i NO ORDERS OM B&W PLANTS - SYSTEM RESPONSE TO $8 LOCA II M.2.s-7 7
(100) 0 YES 7
e 6
(2003 4
ORDERS ON B&W PLANTS - UPORADE AFW SYSTEM i
1 Table 2.4 2
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=-
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PLANTS PLANTS PE R CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICA8LE COMPLETED COMPLETED REMAINING REQUIRED COVERED 3EQUIRED COMPLETED COMPLETED II.K.3.18.8 36 38 (1001 0
mo B&O TASM FORCE - ADS ACTUATION PROPOSED MODIFICATIDMS II.K.3.18.C 36 34 (94 1 2
YES 38 35 33 (94 3 840 TASK FORCE. ADS ACTUATIDW MODIFICATIONS II.K.3.19 3
3
!!003 0
YES 3
3 3
(1003 B&O TASK FORCE - INTERLOCM RECIRCULATORY PUMP MODIFICATIDWS II.K.3.2 71 71 (1001 0
NO S&O TASK FORCE - REPORT 04 PORY FAILURES y
II.K.3.20 1
1 (2001 0
YES 1
1 I
(100)
B&O TASK FORCE - LOSS OF SVC WATER AT BRP II.K.3.21.A' 37 37 (1001 0
N0 i
l B&O TASK FORCE - RESTART OF CSS & LPCI LOGIC DESIGN II.K.3.21.8 37 37 (100) 0 YES 37 35 35 (1001 B&O TASK FORCE - RESTART OF CSS & LPCI LtNIC DESIGN MODIFICATIONS II.K.3.22.A 32 32 (2001 0
NO B&O TASK FORCE - RCIC SUCTION VERIFICATI04 PROCEDURES II.K.3.22.8 32 32 (1001 0
wo B&O TASK FORCE - RCIC SUCTIDW MODIFICATIONS II.K.3.24 34 34 (100]
O YES 34 34 34 (190) 840 TASK FORCE SPACE Co0 LING FOR MPCI/RCI LOSS OF AC POWER II.K.3.25.A 193 103 (100 9
NO 8&O TASK FORCE - POWER ON PUMP SEALS PROPOSED MODIFICAT 045 II.K.3.25.s 102 102 (2001 0
YES Ier
-se s7 iss 3 B&O TASK FORCE - POWER ON PUMP SEALS MODIFICATIDWS II.K.3.27 37 35 194 3 2
YES 37 37 37 (100) 8&O TASK FORCE - COBSEDW REFERENCE LEVEL FOR BWR$
II.K.3.25 37 35 (34 3 2
YES 37 3g 3a gge 3 8&O TASK FORCE - QUALIFICATI0W OF ADS ACCUMULATORS II.K.3.29 8
8 (100) 0 NO B&O TASK FORCE - PERFOR944NCE OF ISOLATIC3 COISENSERS Table 2.4
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._._.-__u
- _. ~ _. ~ _ _ -.. -.-_
k l
t 4
E 4
j r
r t
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF TMI ACTION PLAN - StpmARY BY ITS t
IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED MPA-F06A 110 110 (1001 0
TES 110 50 50 (1001 j
III.A.1.2 OPERATIONAL SUPPORT CENTER MPA-F065 110 110 (2001 0
YES 110 35 25 (100]
III.A.1.2 EMERGENCY CPERATIONS FACILITY i
i MPA-F071 110 71 (6A 1 39 N0 i
I.D.1 2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-Si to t
to.
t l
1
\\
I r
i Table 2.4 r.
2.5 Conclusio0S After a detailed review of the implementation and verification status of the TMl Action Plan requirements at alllicensed plants, the NRC staff has concluded the following:
Progress has been rnade in the implementation of TMl Action Plan requirements at all licensed 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.
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, 31-l
3 UNRESOLVED SAFETY ISSUES This section presents the overall status of implementation and verification of the requirements imposed following the resolution of USIs.
3.1 Imolementation Status Ucensees achieve implementation of USl 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 USl items related to the 110 licensed plants considered in this report.
Approximately 88 percent of the USl items have been implemented at licensed plants.
Of the 1,803 applicable items,1,576 have been completed and only 227 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, USls. Of the 110 licensed plants,11 have fully implemented all applicable USls. Table 3.1 lists the number of unimplemented USl items by plant, Appendix B lists the unimplemented USI items by issue and projected implementation dates.
USIs A-44, A 46, and A 47 account for 87 percent of the 227 unimplemented items.
Figure 3.2 summarizes the implementation status of these issues. These three USis are in varying stages of NRC review and licensee implementation, as described below.
A 44 Station Blackout The station blackout rule was issued in July 1988. The licensees' responses addressing the rule were due within 270 days. 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 essentially all of the safety evaluation reviews of licensee responses.
Most of these responses are for plants that have proposed major hardware modifications. The remaining plants are expected to implement minor hardware and procedure modifications. About 40 percent of the plants have already implemented procedure modifications, and it is expected that a large majority of licensees will complete implementation of the station blackout rule by the end of 1994.
33-
i l
l l
Unresolved Safety issues l
Implementation Status at Licensed Plants 2000 -
~
1500 --
l E
d 1803 5
1576 i
h j
500 --
0 1
i Applicable irr.plemented Unimplemented Figure 3.1
f Summary of' Unimolemented USI Items by Plant itoms itoms ito m s Remaining PLANT Remaining PLANT Remaining PLANT _,
' Arkansas 1 3
Grand Gulf 1 2
Point Beach 1 3
Arkansas 2 3
Haddam Neck 3
Point Beach 2 3
Beaver Valley 1 2
Harris 1 1
Prairle Island 1 2
Beaver Valley 2
-1 Hatch 1 3
Prairie Island 2 2
Big Rock Pt 1 2
Hatch 2 3
Quad Cities 1 4'
Browns Ferry 1 6:
Hope Creek 1 1
Quad Cities 2 4
Browns Ferry 2 "
3 Indias' Pt 2 2
River Bend 1 2
Browns Ferry 3 6
Indian Pt 3 2
Robinson 2 2
"Nnswick 1 '
2 Kewaunee 3
Salem 1 -
2 Brunswick 2 2
LaSalle 1 3
- Salem 2 2
Callaway 1 1
LaSalle. 2 3
San Onoire 2 2
Calvert Cliffs 1
'3 McGuire 1 2
San Onofre 3 2
Calvert Cliffs 2
'3' McGuire 2 2-Seabrook 1 1
Catawba ' 1
'2' Millstone 1 3
Sequoyah 1 1
6 Catawba 2 2
Millstone 2 3-Sequoyah 2.
1
??
Clinton 1 3
Millstone 3 1
St. Lucie 1 3
Comanche Peak 1-1
' Monticello 2
St. Lucie 2 2
Cook 1 3
Nine Mile Pt 1 3
Surry '1 2
Cook 2..
3 Nine Mile Pt 2 2
.Surry 2 2
Cooper Station
'3 North Anna 1' 2
-Susquehanna 1 1
Crystal River 3' 3
North Anna 2 2
Susquehanna 2.
1 Davls-Besse 1 2
Oconee 1-3 Three Mlle Island 1 1
Diablo Canyon 1 1
Oconee 2 3
. Trojan 2
Diablo Canyon 2 1.
Oconee 3.
3 iurkey Pt 3 2
Dresden 2 4
Oyster Creek 1 3
Turkey Pt 4 2
Dresden 3 '
4 Pallsades 4
Vermont Yankee 1 1
. Duane ' Arnold 1
Palo Verde 1 1
Vogtle 1 1
Farley 1 2
Palo Verde 2 1
Vogtle 2
'.1 Farley 2 1
Palo Verde 3 '
2 Washington Nuclear 2 1
Fermi 2 1
Peach Bottom 2 3
Waterford 3 L'
.. Fitzpatrick '
3 Poech Bottom 3 3
Wolf Creek 1 1
Ft Calhoun 1 3
Perry 1 -
2 Zlon 1 4
Ginna ;
2 Pilgrim 1 4
Zion 2
.3 Table 3.1
Summary of Three Unimplamsnted USIs l
l 120 --
(110)
(110)
~~
(93) i i
80 --
(68) 160 --
l(62) co E
P 3
)
4' 40 --
20'- -
o-1 I
i
'A-44
' A-46 Seismic A-47 Safety Station Qualification of implications of Blackout Equipment Control.
g9g in Operating Plants Systems i
M Applicable Plants U mplemented Plants l
I E Unimplemented Plants Figure 3.2 l
3 i
A-46 _ Seismic Qualification of Eouloment in Ooeratina Plants On May 22,1992, the NRC staff completed its review of the Generic Implementation Procedure (GIP) developed by the Seismic Oualification Utility Group (SOUG) based on a comprehensive experience data base. The GIP will be used by licensee's for their walkdown inspections to verify the seismic adequacy of equipment in those operating plants that are subject to USI A 46 review. The 4
Generic Safety Evaluation Report (GSER) on GIP, Revision O, was issued in July,1988. The Supplemental Safety Evaluation Report, No. 2 (SSER 2) was issued May 22,1992. This was developed by the SOUG based on a comprehensive e::perience data base.
Each affected licensee was required to submit its schedule for implementing the resolution of USl A-46 within 120 days after that supplement was issued, that is by September 19,1992. The plant-specific implementation schedule shall be such that the affected plant should complete its implementation within 3 years after the issuance of this supplement or for some plants, until one of the following conditions is met:
the receipt of staff approval of the in-structure response spectra to be used to resolve USl A-46 60 days following the licensee's September 1992 submittal of procedures and 4
criteria in generating those in-structure response spectra, if a prior staff evaluation of the submittal has not been issued A-47 Safety lmolications of Control Systems The primary focus of the resolution of this USI is to provide a mechanism to trip the main feedwater pumps when a high water level occurs in the reactor vessel or steam generators. In 1990, the staff reviewed licensee responses to GL 89-19, which requested information on these modifications. Westinghouse pressurized water reactors (PWRs) have, to a large degree, already implemented the modifications. Boiling water reactors (BWRs) and Combustion Engineering (CE)
PWRs provided responses to the generic letter concluding that the modifications for the generic letter are not cost beneficial. NRR is preparing a CRGR review i
package that BWRs and CE PWRs should not be required to provide the system modifications. Review of seven Babcock & Wilcox (B&W) plants is continuing. The B&W Owners Group has not proposed any relaxation since they do not yet have a position on this issue.
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 USl resolution. The NRC performs these l
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.
l Four 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 Reactor Vessel Pressure Transient Protection Tl 2500/020 A-9 Anticipated Transient Without Scram Tl 2515/76 A 24 Oualification of Class 1E Safety Related Equipment Tl 2515/85 A-7 Mark I Long-Term Program, NUREG-0661, Supplement 1 Additionally, verification of licensee implementation of the requirements of 10 CFR 50.63, which resolve USI A-44, Station Blackout, will be necessary. A draft Tl has been prepared and issued to the NRC regional offices for comment. All plants must be verified for compliance with station blackout requirements.
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 425 items requiring NRC verification, 291 items (68 percent) have been completed.
l l
4 s
i Summary of USI Items Requiring Verification i
Plants Plants Plants I
LLSJ Covered Reauired Verified 24 24 A-7 Mark i Long-Term Program 24 109 91 A-9 Anticipated Transient Without 110
. Scram b
A-24 Qualification of Class IE' Safety-110 109
[
110 Related Equipment
.72 67 A-26 Reactor Vessel Pressure Transient 73 Protection 110.
O A-44 Station Blackout 110 NOTE:
Covered Piarts are those for which Usts are applicable.
{
t Plants Floquired are those plants requiring field verification.
Plants covered but for which field verification is not necessary have implemented i
the resolution in a manner not requiring plant hardware changes.
l r
Table 3.2 4
=2_-
+
3.3
.Etstus by Plant 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 USIs that require the NRC to verify implementation actions, the table shows the number of items covered by a Ti at each p; ant, the number of items requiring verification, and the number and percentage of items completed.
Eleven plants have completed all applicable USIs. Seven plants have four items remaining to be implernented and two plants have six items remaining to be implemented. The remaining 90 plants have three or less items remaining to be implemented.
Five USIs require inspection to verify that implementing actions have been completed.
Of the 110 plants,106 have completed at least 50 percent of the applicable USIs requiring verification. For the remaining four plants, NRC verification is complete for one of the four USis that are applicable at those plants.
Appendix B lists the unimplemented USI items by issue and gives the projected implementation date, where applicable.
._ l
t i
f SAFETY ISSUE MANAGEMENT SYSTEM i
STATUS OF USIs -
SUMMARY
BY PLANT.
r i
s IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS-ITEMS PER CENT j
UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED ARKANSAS 1 16 13 (81 )
3 4
4 3
i75 I
)
ARKANSAS 2 -
16 13 (81 )
3 4
4 3
i 75 l
75 BEAVER VALLEY 1 16 14 (87 )
2 4
4 3
BEAVER VALLEY 2 15 14 193 )
1 4
3 2
1 66 r
BIG ROCK POINT.1 14 12 1 85 1 2
3 2
1 1 50
. I
.BRAIDWOOD 1 15 15
- 100) 0 4
4 3
(75 l
1 1001 0
4 4
3
- 75 t
BRAIDWOOD 2 15 15 i
66 )
6 4
4 2
1 50 t
- i BROWNS FERRY 1 18 12 i
BROWNS FERRY 2 18 15 1 83 )
3 4
4 3
! 75 l
BROWNS FERRY 3 18 12 (66 3 6
4 4
2
.I 50 t
BRUNSWICK 1 18 16 i 88 8 2
4 4
3 (75 1
88 )
2 4
4 3
- 75 I
BRUNSWICK 2 18 16 75 i
. Jb BYRON 1 16 16 1 100 1
0 4
4 3
DJ' BYRON 2 15 15 l 100 0
4 4
3 l 75 c
CALLAWAY'1 18 15 l 93 1
1 4
4 3
(75 t
CALVERT CLIFFS 1 16 13 (81 l
3 4
4 3~
(75
'l l
i CALVERT CLIFFS 2 18
-13 1 81 l
3 4
4 3
75 t
CATAWBA 1 16 14 i ST I
2 4
4 3
75'
[
' CATAWBA 2 16 14 87.I 2
4 4
3 75 i
j CLINTON 1 15 12 1 79 l
3 3
3 2
(66 l
COMANCHE FIAK 1 15 14 (93 l
l' 4
4 3
(75 t
COOK 1 17 14 1 82 t
3 4
4 3
175 4
COOK 2 17 14 1 82 3
4 4
3 1 75 COOPER STATION 18
.15
'l 83 3
4 4
3 1 75 l
3 4
4 3
1 75 CRYSTAL RIVER 3
-16 13 81 i
' DAVIS-BESSE 1 le 14 87 1
2 4
4 3
i 75 75, t
DIABLO CANYON 1 18 15 93 1
1 4
4-3 1
DRESDEN 2 18 14
- 77 4
4 4
.3 1 75
. i 75 i
DIABLO CANYON 2 15 14
- 93 1
4 4
3 DRESDEN 3 18 14 77 )
4 4
4 3
1 75 i
i DUANE ARNOLD 18 17 1 94 1
4 4
3 i 75 l
i FARLEY 1.
18 14 1 87~
I 2
'4 4
3
.! 75 l
l 1
4 4
3 1 75-t 18
'15 l 93 FARLEY Z '
18 15 1 93 i
1 4
4 3
1 75 5
FERMI 2 FITZPATRICM 18 15 83 1
3 4
4 3
175 l
FORT CALHOUN 1 18
- 13 81 3
-4 4
3 (75 I
GINNA 16'
'14 87 2
4 4
3 I 75 l
GRAND GULF 1 16
'14 1 87 2
3.
~ 3 2
1 56 I
i HADDAM NECK 18
'15 I 83 1 3
'4 4
3 1 75 t
HARRIS 1 18
.15
! 93 1
4-4 3
1 75 i
HATCH 1 18 15 1 83 3
4 4
3 1 75 t
HATCH 2 18 15 1 83 1
3 4.
4 3
l 75 i
i HOPE CREEK 1 17 16 (94 i
'I 4
4 3-l 75-1 INDIAN POINT 2 16 14 187 3 2'
4 4
3-(75 i
Table.3.3 i
i
(
t SAFETY ISSUE MANAGEKtM SYSTEM STATUS OF USIs -
SUMMARY
BT PLANT r
IMPLEMENTATION vtRIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PE R CENT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REOUIRED COMPLETED COMPLETED
.......... UNIT IMOIAN POINT 3 16 14 L87 )
2 4
4 3
175 KEWAUNEE 16 13 1 81 1
3 4
4 3
(75 t
LASALLE 1 17 14 1 82 1
3 3
3 2
(66 I
I LASALLE 2 16 13 i 81 1
3 3
3 2
(66 LIMERICK 1 16 16 1 100 1
0 3
3 2
(66 i
LIMERICK 2
.16 16 1 100 1
0 3
3 2
(66 i
75 1
0 4
4 3
MAINE YANKEE
.16 16 1 100 1
MCGUIRE 1 18 16 1 88 2
4 4
3 i 75 1
MCGUIRE 2 18 16 I88' 2
4 4
3' i 75 I
i
' MILLSTONE 1
'19 16
.1 84 1
3 4
4 3
1 75 MILLSTONE 2 16 13 1 81 3
4 4
3 (75 MILLSTONE 3 16 15 I 93 )
1.
4 4
3 (75 JA MONTICELLO 18 16 i 88 )
2 4
4 3
175 i
Y NINE MILE POINT 1 18 15
! 83 3 3
4 4
3
{75 i
87 )
2 3
3 2
(66 i
NINE MILE POINT 2 le 14 l
NORTH ANNA 1 16 14
- 87 )
2 4
4 2
(50 NORTH ANNA 2 17 15
- 88 )
2 4
4 2
(50 OCONEE 1 16 13 1 81 )
3 4
4 1
(25 i
DCONEE 2 16 13 I 81 1 3
4 4
.1 (25 OCONEE 3
.16 13 I 81 1 3
4 4
1 (25 OYSTER CREEK 1
.18 15 1 83 3
4 4
3 i75 l
PALISADES 16 12 1 75 4
4 4
3 1 75 I
PALO VERDE 1 15 14-1 93 1
1 4
4 2
50 I
PALO VERDE 2 15 14 (93 )
1 4
4 2
1 50 I
PALO VERDE 3 15 13 1 86 )
2 4
4 2
i 50 t
PEACH BOTTOM 2.
18 15 i 83 3
4 4
3 l 75 t
75 PEACH BOTTOM 3 13 15 i 83 3
4 4
3 PERRY 1 15 13
- 86 2
3 3
2-(66 i
77 4
4 4
3 175 l
-PILGRIM 1 18 14 81 3
4 4
3 1 75 i
POINT BEACH 1 16 13 i
1 75 I
I POINT BEACH 2 16 13
. 81 3
4 4
3' 87 l
2 4
4 3
I 75 PRAIRIE ISLANO 1 16 to k87 l
2 4
4 3
! 75 PRAIRIE ISLAND 2
.16 14 QUAD CITIES 1 18 14 (77 t
4 4
4 3
1 75 QUAD CITIES 2 18 14 177 4
4 4
3 1 75 RIVER BEND 1
'15 13 186 2
3 3
2' 1 66 t
ROBINSON 2 16 14 F87 -
2 4
4 3
1 75 SALEM 1 16 14 i 87, 2
4 4
3 i 75 t
SALEM 2 17 15 1 88 2
4 4
3 i 75 a SAN ONOFRE 1 16 16 i 100 0
4 4
3 1 75 y SAN ONOFRE 2 16 14 I 87 )
2 4
4 3
175 )
SAN ONOFRE 3 16 14 I 87 )
2 4
4 3
(75 )
i SEABROOK 1 15 14
-(93 )
1 4
4 3
(75 )
SEQUOYAH 1 18 17 (94 )
1 4
4 2
(50 )
9 Table 3.3 4
i
~
_m l
SAFETY ISSUE MANAGEMENT SYSTEM l
i STATUS OF USIs -
SUMMARY
BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED SEQUOYAH 2 18 17 (94 1 1
4 4
2 l'50 l
SOUTH TEXAS 1 15 15 (1001 0
4 4
3 1 75 i
SOUTH TEXAS 2 15 15 (100 1
0 4
4 3
1 75 l
t 3
4 4
2 1 50 i
ST LUCIE 1 16 13 181 ST LUCIE 2 16 14 1 37 i
2 4
4-2 1 50 l
SUt#4E R 1 16 16 1 100 l
0 4
4 3
i 75 SURRY 1 16 14 I 87 1
2 4
4 2
1 50 SURRY 2 16 14 1 87 )
2 4
4 2
(50 t
SUSQUEHANNA 1 17 16 (94 )
1 3
3 2
(68 i
SUSQUEHANNA 2 16 15 l93 i
1 3
3 2
(66 t
THREE MILE ISLAND 1 16 15 l 93 1
4 4
3 (75 I
TROJAN 16 14
- 87 2
4 4
3 175 i
Sh TURKEY POINT 3 16 14
' 87 l
2 4
4 2
1 50 l
- 87 l
2 4
4 2
1 50 t
JE TURKEY P0lNT 4 16 14 94 1 1
4 4
3 i 75 i
VERMONT YANKEE 1 18 17 V0GTLE 1 15 14 (93 l
1 4
4 2
1 50 l
V0GTLE 2 16 15 l93 i
1 4
4 1
(25 f
WASHINGTON NUCLEAR 2 16 15 i 93 1
1 3
3 2
(66 t
WATERFORD 3 15' 13 i 86 l
2 4
4 3
(75 WOLF CREEK 1 16 15 (93 1 1
4 4
3 (75 i
ZION 1 16 12 -
(75 1 4
4 4
3 (75 I
ZION 2 16 13 (81 )
3
/.
4 3
(75 )
TOTALS / AVERAGES 1803 1576 88 227 427 425 291 68 Table 3.3
J 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.
Of the 27 USls,17 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 87 percent of the unimplemented items:
A-44, Station Blackout, with 93 plants remaining to complete implementation; A-46, Seismic Qualification of Equipment in Operating Plants, with 62 plants remaining to complete implementation; and A 47, Safety implication of Control Systems, with 43 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. Twelve plants have not implemented corrective actions for A 9, Anticipated Transient Without Scram, and ten plants have not implemented corrective actions for USl A-48, Hydrogen Control Measures and Effects of Hydrogen Burns. The remaining USIs have from one to two plants remaining to complete I.
implementation.
NRC inspection to verify licensee implementation is required for five USIs and is complete for USI A-7, Mark I long-term program. Station blackout accounts for 110 of the 134 outstanding verifications.
__.m_
l 1'
l SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF USIs -
SUMMARY
BT ITEM
?
i IMPLEMENTATION VERIFICATION i
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT i
ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED i
A-1 110 110 (100) 0 NO WATER HAf#9ER
[
A-2 73 73 (100)-
0 NO ASYP99ETRIC BLOWDOWN LOADS ON REACTOR PRIMARY COOLANT SYSTEMS t
A-3 4 AND 5 73 73 (100) 0 NO STEdM GENERATOR TUSE INTEGRITY A-6 23 23 (100' O
NO MARK I SHORT-TERM PROGRAM
.A A-7 24 22 (91 1 2
YES 24 24 24 (100)
MARK I LONG-TERM PROGRAM A-8 8
8 (100) 0 NO MARK II CONTAINMENT POOL DYNAMIC LOADS LONG-TERM PROGRAM A-9 110 98 (89 )
12 YES 110 109 91 (83 )
ATWS h
A-10 38 3e (100) 0 NO BWR FEEDWATER WOZZLE CRACKING A-11 110 109 (99 )
1 NO REACTOR VESSEL MATERIALS TOUOHNESS i
'A-17
'110 110 (1CO) 0 NO SYSTEM INTERACTIONS'IN NUCLEAR POWER PLANTS
'A-24 110 108
)
2 YES 110 110 109=
199 )
QUALIFICATION OF CLASS IE SAFETY-RELATED EQUIPMENT (98 A-20 73 70 (100) 0 YES 73 72 87 (93 )
REACTOR VESSEL PRESSURE TRANSIENT PROTECTION
'A-31 50 49 (97 )
1 NO RHR SHUTDOWN REQUIREMENTS r
A-38 (C010) 110 110 (100) 0 NO CONTROL OF HEAVY. LOADS OVER SPENT FUEL POOL (PHASE ONE)
A-36 (C015) 78 78 (100) 0 NO CONTROL OF HEAVY LOADS - PHASE II (FOLLOWUP OF MPA8 C-10)
Table 3.4 -
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF USis -
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 4 39 36 36 (100) 0 NO DETERMINATION OF SRV POOL DYNAMIC LOADS & TEMP. LIMITS FOR BWR CNTMNTS A-40 3
3 (100) 0 NO SEISMIC DESIGN CRITERIA A-42 38 38 (100) 0 NO PIPE CRACKS IN 8 OILING WATER. REACTORS A-43 110-110 (100) 0 NO CONTAINMENT EMERGENCY SUMP PREFORMANCE b7 STATION 8LACMOUT.
110 17 (15 )
93 YES 110 110 0
(0 1 A-44
-A-45 110 110 (100)
O NO
-SHUTDOWN DECAY HEAT REMOVAL REQUIREMENTS A-46 88 8
(8 )
82 NO SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS A-47 110 87 (60 1 43 NO SAFETY IMPLICATIDW OF CONTROL SYSTEMS A-48
~ 47 37 (78 l 10 NO HfDROGEN CONTROL MEASURES AND EFFECTS OF HYDROGEM BURNS A 73 72 (98 1 1
NO PRESSURIZED THERMAL SHOCK.
Table 3.4
3.5 Conclusions After a detailed review of the implementation and verification status of the resolution of the 27 USts, the NRC staff has concluded the following:
The NRC closure process for USIs ensures continued adequate protection of the public health and safety.
All 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 implementing requirements imposed following the NRC's resolution of USis, and the framework 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 with regulatory requirements.
4 GENERIC SAFETY ISSUES This section presents the overall status of implementation and verification of GSis applicable at the 110 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.
4.1 Imolementation Status Ucensees 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 110 licensed plants considered in this report.
Approximately 90 percent of the GSI items have been implemented at licensed plants.
Of the 2,647 items,2,389 have been comp!sted and 258 remain open from an implementation standpoint. On average, each plant has less than three items to i
implement, and no plant has more than seven remaining items. Figure 4.1 presents the overall status of, and progress on, GSis. Of the 110 licensed plants,14 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.
Seven GSIs have not been implemented at a significant number of plants for which they are applicable; these account for approximately 95 percent of the unimplemented items. Figure 4.2 summarizes the implementation status of these issues. A brief description of each issue follows.
GSI 43 Reliability of Air Systems This issue arose from concerns related to the TMl accident and air-operated l
equipment failures at other plants. In August 1988, the staff issued Generic Letter 8814 to specify the performance of a design and operations verification of n
instrument air systems and descriptions of licensees' programs for maintaining proper instrument air quality. The staff gave licensees 6 months in which to i
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 cornplete 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.
- t
GSI Numbers and Corresponding SIMS Item Numbers SIMS MPA Item No.
GSINo.
- _No, SIMS Title 40 40 B-065 Safety Concems Associated With Pipe Breaks in BWR Scram System 41 41 B-050 BWR Scram Discharge Volume Systems GL-88-14 43 B-107 Instrument Air Supply System Problems Affecting Safety-Related Equip.
GL-89-13 51 L-913 Service Water System Problems Affecting Safety-Related Equipment 67.3.3 67.3.3 A-017 Improved / xident Monitoring 70 70 B-114 PORV and t31ock Vafve Reliability 75 (8076) 75, item 1.1 B-076 Item 1.1 - Post-Trip Review; Program Description & Procedures 75 (8085).
75, item 1.2 B-085.
Item 1.2 - Salem ATWS 1.2 Data Capabihty 75 (8077) 75, !!em 2.1 0-077 I;em 2.1 - Equipment Classification & Vendor interf ace - RTS Component 75 (8086) 75, item 2.2.1 B-086 flem 2.2.1 - Salem ATWS 2.2 S-R Components GL-90-03 75, item 2.2.2 L-003 Item 2.2.2 - Re'.axation of Staff Pos in Gen Letter 83-28, item 2.2 Part 2 75 (8078) 75, items 3.1.1 & 3.1.2 B-078 Items 3.1.1 & 3.1.2 - Post-Maintenance Test Procedures & Vendor Recomm.
75 (B079) 75, item 3.1.3 B-073 Item 3.1.3 - Post-Maintenance Testing - Changes to Tech Specs - RTS Component 75 (8087) 75, Items 3.2.1 & 3.2.2 B-087 Items 3.2.1 & 3.2.2 - Salem ATWS 3.2.1 & 3.2.2 S-R Components 75 (8088) 75, item 3.2.3 B-088 Item 3.2.3 - Salem ATWS 3.2.3 T.S. S-R Components) 75 (8080) 75, item 4.1 B-080 item 4.1 - Reactor Trip System Reliability - Vendor Related Mods n
IP 75 (B081) 75, items 4.2.1 & 4.2.2 B-081 Items 4.2.1 & 4.2.2 - Preventative Maint Prog for Reactor Trip Breakers 75 (8082) 75, item 4.3 B-082 Item 4.3 - Automatic Actuation of Shunt Trip Attach. for West & B&W 75 (8090) 75, item 4.3 B-090 item 4.3 - Salem ATWS 4.3 W and B&W T.S.
75 (8091) 75, item 4.4 B-091 Item 4.4 - Salem ATWS 4.4 B&W Test Procedures 75 (B092) 75, item 4.5.1 B-092 Item 4.5.1 - Salem ATWS 4.5.1 Diverse Trip Features 75 (8093) 75, items 4.5.2 & 4.5.3 B-093 Items 4.5.2 & 4.5.3 - Salem ATWS 4.5.2 & 4.5.3 Test Altematives 86 86 B-084 Long Range Plan Dealing With Stress Corrosion Cracking in BWR Piping GL-88-03 93 B-098 Resolution of GSI 93," Steam Binding of Auxiliary Feedwater Pumps" 94 94 B-115 Additionc! Low-Temp Oveyressure Protection for LWRs GL-88-17 99 L-817 Loss of Decay Heat Removal 124 124 S-001 Auxiliary Feedwater System Reliability GL-80-099 A-13 B-107 Technical Specification Revision for Snubber Survei!Iance GL-84-13 A-13 B-022 Technical Specification for Snubbers A-16 A-16 D-012 Steam Effects on BWR Core Spray Distribution M PA-8023 A-35 B-023 Degraded Grid Voltage B-10 B-10 S-008 Behavior of BWR Mark 111 Cortainments B-36 B-36 none Dev Design. Test & Maint Criteria for Atmo Cleanup Sys Air Filter & Adsorption Units GL-80-014 B-63 B-045 LWR Primary Coolant System Pressure isolation Valves Table 4.1
l i
l l
Generic Safety issues I
implementation Status at Licensed Plants I
i sooo i
l 2500 --
2000 --
l 1500..
l d
2647 2389 9
l 1000 i
j soo..
o i
1 1
l l
Applicable implemented Unimplemented I
Figure 4.1 i
l l
I l
l h
l-i
< a
.-.i
_=.
Summary of Unimolemented GSI Items by Plant items items items PLANT Remaining PLANT Remaining PLANT Remaining Arkansas 1 1
Haddam Neck 4
Prairie Island 2 2
Arkansas 2 3
Hatch 1 1
Quad Cities 1 3
Beaver Valley 1 5
Hatch 2 1
Quad Cities 2 4
Beaver Valley 2 3
Hope Creek 1 2
River Bend 1 2
l Braldwood 1 3
Indian Pt 2 4
Robinson 2 2
Braldwood 2 4
Indian Pt 3 3
Salem 1 2
Browns Ferry 1 6
Kewaunee 3
Salem 2 2
Browns Ferry 2 3
LaSalle 1 2
San Onofre 2 2
Browns Ferry 3 7
LaSalle 2 2
San Onofre 3 2
Brunswick 1 1
Maine Yankee 2
Seabrook 1 2
Brunswick 2 1
McGuire 1 4
South Texas 1 4
Byron 1 2
McGuire 2 4
South Texas 2 4
Byron 2 2
Millstone 1 3
St. Lucle.1 3
Callaway 1 2
Millstone 2 3
St. Lucie 2 3
Calvert Cliffs 1 5
Millstone 3 4
Summer 1 3
Calvert Cllits 2 5
Nine Mlle Pt 1 2
Surry 1 2
Catawba 1 4
Nine Mile Pt 2 1
Surry 2 2
Catawba 2 4
North Anna 1 3
Susquehanna 1 1
Cilnton 1 2
North Anna 2 3
Susquehanna 2 1
Cook 1 4
Oconee 1 3
Trojan 1
. Cook 2 3
Oconee 2 3
Turkey Pt 3 3
Cooper Station T
Oconee 3 2
Turkey Pt 4 3
Crystal River 3 S
Oyster Creek 1 2
Vogtle 1 2
Diablo Canyon 1 3
Palisades 4
Vogtle 2 2
Diablo Canyon 2 3
Palo Verde 1 1
Waterford 3 1
Dresden 2 2
Palo Verde 2 2
Wolf Creek 1 3
Dresden 3 3
Palo Verde 3 2
Zion 1 6
Farley 1 2
Peach Bottom 2 1
Zlon 2 6
Farley 2 2
Peach Pottom 3 1
Fermi 2 1
Perry 1 2
Fitzpatrick 2
Pilgrim 1 2
Ft Calhoun 1 4
Point Beach 1 2
Glnna 4
Point Beach 2 2
Grand Gulf 1 1
Pralrle Island 1 2
Table 4.2
Summary of Seven Unimpl2m:nted GSis i
V 120 (110)
(110)
(110)
(110)
L' 100 L90) 80 __
(73)
IO I 7)
(66) p (g) 60 (61).
40 41) 20 (20)-
(16) 0__
g l
l l
i GL-88-14 GL-89-13 67.3.3 70 GL-90-03 94 GL-88-17 Legend f
SIMS Numbers E Total j
l 0 inen=nied E uni',5:;r.ented-Figure 4.2 I
I
i l
i GSI 51 Procosed Reouirements for imorovina the Reliability of Ooen-Cvele 4
I Service Water Systems
~
~
~
i l
This issue was developed as a result of uncertainties regarding the compliance of 3
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 l
approximately two thirds of all plants. The staff considers the status of this GSI acceptable, j
GSI 67.3.3 Imoroved Accident Monitorino i
L This issue addresses conformance with RG 1.97, " Instrumentation for i
Ught 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 l
to request schedules and details of the licensees': plans.to implement the l
provisions of RG 1.97, Revision 2. - Based on responses to this generic letter, the staff issued confirmatory orders in 1985l Because the industry has taken exception-l to and appealed some_of the provisions of RG 1.97, Revision 2, implementation is j
- incomplete at many plants. Most BWRs have not installed the Category i neutron j
flux monitoring system. Many PWRs have not installed Category 2 qualified j-instrumentation to monitor containment sump temperature. The staff is j-reevaluating these positions for both BWRs and PWRs.
l l
GSI 75 fltem 2.2.2) Vendor Interface for Safetv Related Comoonents 1
This issue addresses one of several anticipated transient without scram (ATWS) items; namely establishing vendor interface programs for safety related j
components. In March 1990, the staff issued GL 90-03 as revised guidance j
addressing GSI 75, Salem ATWS Item 2.2.21 GL-90-03 supersedes.the original l
guidance on this item issued in GL 83 28.1 Industry responses to GL 90-03 were 4
due by the end of September 1990. Allindustry responses have been received.
As a result of GL 90-03, many licensees have revised their vendor interface L
program and all programs are expected to be revised by February 1993.
i L
GSI 99 Reactor Coolant System / Residual Heat Removal Suction Line Valve i
Interlock on PWRs-j.
This issue concerns inadvertent closing of residual heat removal (RHR) suction
' valves when the RHR system is in use, especially during mid-loop operation,
-resulting in possible loss of decay heat removal capability. GL 88-17, " Loss of Decay Heat Removal,". issued in October _1988, superseded and closed out an j-_
earlier generic letter of more limited scope and was intended to improve industry's understanding of the significance of such events. As recommended in GL 88-17,-
-56
a several expeditious actions have been completed at all plants. In addition, GL 88-17 requested plans and schedules for implementation of a number of longer term program enhancements to be completed by specified subsequent refueling outages. Essentially all of those plants that still have this issue open are scheduled to complete implementation by the end of 1992 in accordance with the generic j
letter. The safety improvements provided by GL 88-17 reduced the importance of GSI 99 to the degree that the staff suggested in GL 88-17, but did not require, licensees to consider removing the auto closure interlocks.
GSI70 PORV and Block Valve Reliability. and GSI94 Additional Low Temoerature Overoressure Protection for Light-Watei Reactors GL 90-06 was issued to licensees on June 25,1990, to be implemented by the end j
i of the first refueling outage. Some licensees may need to resubmit their responses for conformance with the staff positions. GI 70 is applicable to all Westinghouse and B&W-designed plants and CE-designed plants with power-operated relief valves (PORVs). Issue 94 is applicable to all Westinghouse-designed and CE-designed plants whether or not they have PORVs and block valves.
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 prionties, to verify proper implementation of the requirements.
Verification is not considered complete until the required inspection is conducted in accordance with the Tl 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 fe safety verification inspection are completed before fullimplementation of all aspects of the issue's resolution by the licensee. Of the 1080 items requiring NRC verification,1039 (97 percent) have been completed.
Seven Tis provide guidance for the field ve,ification of licensee implementation of GSis.
Tl designations and the corresponding GSIs are provided in Table 4.3. Table 4.4 summarizes the items remaining to be verified.
Temporary Instructions for Resolved GSis 1
SIMS ltem SIMS Title IJ 41 BWR SCRAM DISCHARGE VOLUME SYSTEM 2515/090 67.3.3 IMPROVED ACCIDENT MONITORING 2515/087 75 (8077)
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 (B079)
ITEMS 3.1.3 - POST MAINTENANCE TESTING -
2515/064 CHANGES TO TECH SPECS - RTS COMPONENT 75 (8080)
ITEM 3.1 - REACTOR TRIP SYSTEM RELIABILITY -
2515/091 VENDOR RELATED MODS 75 (8081)
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 i
i 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 ATWS 4.5.1 - DIVERSE TRIP FEATURES 2515/064 i
l 86 LONG RANGE PLAN DEALING WITH STRESS-2515/089 l
CORROSION CRACKING IN BWR PIPING l
GL-88-17 LOSS OF DECAY HEAT REMOVAL 2515/101 2515/103 Table 4.3 l
l 58-l
Summary cf GSI Itema Requiring Verificiticn Plants Plants Plants SIMS Item Covered Reauired Verified 41 BWR SCRAM DISCHARGE VOLUME SYSTEMS 37 37 36 67.3.3 IMPROVED ACCIDENT MONITORING 110 109 94 75 (B077)
ITEM 2.1 - EQUIPMENT CLASSIFICATION & -
110 97 94 VENDOR INTERFACE -RTS COMPONENT 75 (8078)
ITEMS 3.1.1 & 3.12 -POST MAINTENANCE 110-98 95 TEST PROCEDURES & VENDOR RECOMM.
75 (B079)
ITEM 3.1.3 - POST MAINTENANCE TESTING -
110 98 95 CHANGES TO TECH SPECS - RTS COMPONENT-75(8080)
ITEM 4.1 - REACTOR TRIP SYSTEM 73 73 72 RELIABILITY -VENDOR RELATED MODS
[
75(8081)
ITEMS 42.1 & 422 -PREVENTATIVE MAINT 73 67 67 PROG FOR REACTOR TRIP BREAKERS 75 (8086)
SALEM ATWS 22 S-R COMPONENTS 110 98 95 75(8087)
SALEM ATWS 3.2.1 & 322 S-P COMPONENTS 110 98 97 75 (B088)
SALEM ATWS 32.3 T.S. S-R COMPONENTS 110 98 95 75 (B092)
SALEM ATWS 4.5.1 DIVERSE TR!P FEATURES 110 98 97 86 LONG RANGE PLAN DEALING WITH STRESS
-36 36 34 CORROSION CRACKING IN BWR P1 PING GL-88-17 LOSS OF DECAY HEAT REMOVAL 73 73 68 NOTE: Plants Covered are those for which GS!s are %W I
Plants Required are those piants requirmg field venficaton.
[
Plants covered but for which field verdicaton is not necessary have 'rnplemented the resolution in a manner not requiring plant hardware cha ges.
Table 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 110 plants,14 have completely implemented all GSI items. Fifty plants have completed implementation actions for all except 1 or 2 GSis; and 38 plants have 3 or 4 items to implement. The remaining eight plants have seven or less items to implement. All of the items requiring verification by inspection (in accordance with a TI) have been completed at 85 plants. Another 22 plants have completed at least 89 percent of the applicable items requiring verification.
6- !
l SAFETY ISSUE MANAGEMENT SYSTEM iTATUS OF GSIs -
SUMMARY
BY PLANT IMPLEMEN7ATION VERIFICATION ITEMS ITEMS "ER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED
(;0MPLETED REMAININO COVERED REQUIRED COMPLETED COMPLETED ARKANSAS 1 27 26 (96 )
1 11 3
3 1:100l ARMANSAS 2 25 22 (87 )
3 11 3
3 L100 f
BEAVER VALLEY 1 26 21 (80 t
5 11 11 11 1 100 i
EEAVER VALLEY 2 27 24 (88 t
3 11 3
3 1 100 i
BIG ROCK POINT 1 21 21 (100 l
0 10 9
9 100 i
88 3
11 11 11
- 100 BRAIDWOOD 1 27 24 85 4
11 11 11 L100 i
BRAIDWOOD 2 27 23 l
71 1
0 10 10 9
' 89 BRDWNS FERRY 1 21 15 BROWNS FERRY 2 21 18
' 85 l
J 10 10 10 1 100 1
BROWNS FERRY 3 21 14
.66 )
7 10 10 9
1 89 i
BRUNSWICK 1 19 18 i 94 1 1
10 10 10 1 100 I
BRUNSWICK 2 19 18 1 94 )
1 10 10 10 1 100 1
CD BYRON 1 27 25 1 92 )
2 11 11 11 1 100) h3 BYRON 2 27 25 1 92 )
2 11 11 11 l 100 t
CALLAWAY 1 26 24 1 92 l 2
11 11 11 100 t
5 11 11 11
- 100 1
CALVERT CLIFFS 1 23 18 l 78.
5 11 11 11 1 100 i
CALVERT CLIFFS 2 23 18
- 78 t i 85 l 4
11 11 11 1 100s CATAWBA 1 27 23 CATAWBA 2 27 23
! Ph )
4 11 11 11 1 100 CLINTON 1 21 19
'90 1
2 10 10 10 1 100 t
COMANCHE PEAK 1 26 26 (100 1
0 11 11 11 4100 1
COOK 1 25 21 1 83 1
4 11 11 11 1100 1
COOK 2 25 22 1 87 l
3 11 11 11 1 100 1
1 10 3
3 1 100 1
COOPER STATION 21 20 1 95 CRYSTAL RIVER 3 27 22 1 81 5
11 11 11 1 100 l
DAVIS-BESSE 1 26 26 1 100 0
11 11 10 1 90 11 11 11 (100 DIABLO CANYON 1 26 2"
- 88.
~
11 11 11 (100 t
t 88 )
DIABLO CANYON 2 26
- 3 DRESDEN 2 20 18
' 89 )
i 10 10 9
(89 I
3 to 10 9
(89 DRESDEN 3 20 17 I 84 DUANE ARNOLD 21 21 1 100 1
0 10 10 10 (100 i
FARLEY 1 26 24 92 t
2 11 11 11 (100 l
FARLEY 2 26 24 1 92 l
2 11 11 11 (100J FERMI 2 21 20 1 95 1
10 10 9
(89 i
FITZPATRICK 21 19 i 90 2
10 10 10 (100 1
FORT CALHOUN 1 25 21 l 83 4
11 8
8 (100 i
GINNA 26 22 l 84 1
4 11 11 10 (90 l
GRAND GULF 1 21 20 1 95 i
1 10 10 9
89 4
11 11 11 100 84 100)1 HADDAM NECK 26 22 l
0 11 11 11 100 I
HARRIS 1 27 27 i
HATCH 1 21 20 1 95 )
1 10 10 10 (100i HATCH 2 21 20 1 95 )
1 10 10 10 (100)
HOPE CREEN 1 22.
20 1 90 )
2 10 3
3 (100)
INDIAN POINT 2 25 21 (83 1 4
11 11 11 (100)
Tab?
4.5
m 5
i-c.
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF GSIs - SUPa?.RY BY PLANT t
{
IMPLEMENTATION VERIFICATION l.
iI ITEMS ITEMS PER CENT
-ITEMS ITEMS
-ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED-COMPLETED
. COMPLETED t
INDIAN POINT 3 26
' 23 '
! 88 1 3
11 11.
11 100 1
t I
KEWAUNEE
.25 22 1 87 )
3 11 11 11 100 1
LASALLE 1 22 20
. 90 l 2
10 10 -
9 89 :l 90')
2 10-10 9
89 l
LASALLE 2 22 20
' LIMERICK 1 20 20 1 1001 0
10 3
3 (100 t
LIMERICK 2 20 20
. 100) 0 10 3
3 (100 l
I 100 I
j.
MAINE YANKEE
.24 22 1 91 1
2 11 11 11 MCGUIRE 1 26 22 1 84 i
a 11 11L 11
- 100 I
l PCCUIRE 2 26 22 1 84 1
4 11 11 11 100 I
f 1001 F
MILLSTONE 1 21 18 i 85 1
3 10 10 10 i
100 i
l
, MILLSTONE 2 24 21 1 87 l
3 11' 11 11 1
100 l
3 3
MILLSTONE 3 27 23 1 85 1
4 11
. 10 10 i
100t MONTICELLO 20 20 (100 l
0 10 l
NINE MILE POINT 1 22-20 1 90 1
2 10 10 10 4 1001 i
NINE MILE POINT 2-22 21
.1 95 l
1 10
- J 3
(100)
NORTH ANNA 1 26 23 1 88 )
3 11 11 11 1100) i NORTH ANNA 2 26 23 1 88 i 3
11
'11 11 1 1001 i
90 OCONEE 1
~ 28 23 l 88 3
11 11 10 i
OCONEE 2 26 23 1 88 3
11 11 10 90 l~
OCONEE.3 26 24 1 92 2
11 11 10 90 l
OYSTER CREEK 1 22 20 1 90
.2 10
- O 10 1 100 i
PALISADES 24 20 83 4
11 11 10 LSO i
PALO VERDE 1 23 22 1 95 1
11 11 11 (100 i
PALO VERDE 2 23 21 (91 2
11 11 11 1100 i
PALO VERDE 3 23 21 1 91 2
11.
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2 9
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2 11 11 11
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2 11 11 11 100 18 t
l i
SAN ONOFRE 1 26-26
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2 11 11 6
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2 11 11 11.
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SEQUOYAH 1-27 27 l 100 1
0.
11 11 10 (90 )
Table 4.5
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SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF GSIs - SLM4ARY BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT t
UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED j
SEQUOYAH 2 27 27 (100) 0 11 11 10 (90 l
SOUTH TEXAS 1 27 23 (85 1
4 11 11 11 (100 i
SOUTH TEXAS 2 27 23 1 85 4
11 5
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11 11 11 (100 t
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3 11 11 11 (100 i
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2 11 11 11 (100 i
SURRY 2 26 24 1 92 2
11 11 11
! !00 i
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SUSQUEHANNA 2 20 19 l 94 1
10 10 10 i 100l THREE MILE ISLAND 1 25 25 (100 0
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- l 96
[
1 11 11 11 (100)
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3 11 11 10 190 4
TURKEY POINT 4 25 22 i "7 1
3 11 11 10 l 90,
i 0
10 10 10 1 100 VERMONT YANMEE 1 21 21 (100 V0GTLE 1 27 25 192 2
11 11 11 1 100 I
92 2
11 11 11 1 100 V0GTLE 2 27 25 i
WASHINGTON NUCLEAR 2 21 21 i 100 0
10 10 10 1 100 WATERFORD 3 23 22 1 95 1
1 11 3
3 1 100:
WOLF CREEK 1 27 24 1 88 3
11 11 11 1 100 ZION 1 26 20 1 76 )
5-11 11 11 1 100 ZION 2 26 20 (76 )
6 11 11 11 (1001 TOTALS / AVERAGES 2647 2389 90 258 1172 1080 1039 97 Table 4.5
1 i
4.4 Status by issue.
4 Table 4.S summarizes the status of implementation and verification of each GSI and j
sub-Issue. For each issue, the table shows the number of applicable plants, the f
number and percentage of plants that have completed implementation, and the number.of plants remaining to complete implementation. For those issues requiring verification of corrective actions, the table shows the number of plants covered by a TI, the number of plants requiring verification, and the number and percentage of plants that have completed verification.
Of the 34 GSis and sub-issues,18 have been fully implemented. Five issues remain to be implemented at only one plant each and four more issues remain to be implemented at two or three plants each. The seven icsues discussed in Section 4.1 of this report account for 244 ( 95 percent) of the 258 items remaining to be implemented.
3 iP J
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t i
SAFETY ISSUE MANAGEMENT SYSTEM l
STATUS OF GSIs -
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 40 33' 31 93 )
2 NO SAFETY CONCERNS ASSOCIATED WITH PIPE BREAKS IN BWR CRAM SYSTEM 41 37 36 (97 l' 1
YES 37 37 36 (97 l-BWR SCRAM DISCHARGE VOLUME SYSTEMS 67.3.3 110
' 61 (55 )
49 YES 110 109 94 (86 )
i L
IMPROVED ACCIDENT MONITORING' 70 67 11 (16 )
56 NO PORY AND BLOCK VALVE RELIABILITY
'75 (8076) 110-110-(100) 0 NO ITEM'i.1 - POST TRIP REVIEW: PROGRAM DESCRIPTION & PROCEDURES 75 (8077) 110
'110 (100) 0 YES 110' 97 94 (96 )
t ITEM 2 1 - EQUIPMENT CLASSIFICATION & VENDCR INTERFACE - RTS COMPONENT v
75 (8078) 110 110 (1001 0
YES 110 98 95 (96 )
ITEMS 3.1.1 & 3.1,2 '-POST MAINTENANCE TEST PROCEDURES & VENDOR REcope4.
75 (8079) 110-110 (1001 0
YES 110 98 95 (96 1, ITEM 3.1;3 -. POST MAINTENANCE TESTING - CHANGES TO TECH SPECS - RTS CO 75 (B080) 73-73 (1001 0
YES 73 73-72 (98 )
ITEM 4.1 - REACTOR TRIP SYSTEM RELIABILITY - VENDOR RELATED MODS 75'(8081) 73 73
-(1001 0
YES 73 67 67 (100)
-ITEMS 4.2.1 & 4.2.2 -PREVENTATIVE MAINT PROC FOR REACTOR TRIP BREAKERS-l t
75 18082) 58 58 -
(1001 0
No i
ITEM 4.3 - AUTOMATIC ACTUATION OF SHUNT TRIP ATTACH. FOR WEST & B&W 75 (8085).
110 107 (97 )
3 NO SALEM ATWS 1.2 DATA CAPABILITY i
75 (B086) 110' 110 (100) 0 YES 110 98 95 (96 )
t
' SALEM ATWS 2.2 S-R Co*1PONENTS 75 (B087) 110 110 (100) 0 YES 110 98 97 (98 )
SALEM ATWS 3.2.1 & 3.2.2 S-R COMPONENTS
.. [
75 (8088)
- 110 110 (1001 0
YES 110 98 95 (96 )
SALEM ATWS 3.2.3 T.S.'S-R COMPONENTS
'f Table 4.6 -
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4.5 Conclusions After detailed 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.
i 2
Licensees 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 seven largely unimplemented GSis is generally acceptable because of the relatively recent issuance of staif positions on three of the GSis, and projected implementation schedules for the remaining four. The NRC staff is preparing a generic resolution to modify and/or clarify certain provisions of -
RG 1.97, Revision 2, relating to the implementation of GSI 67.3.3, Improved Accident Monitoring.
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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 110 licensed plants. Because cach MPA may be tracked under different designations, Table 5.1 cross references the MPA number and the SIMS j
number used in the tables and appendices of this report.
5.1 Imolementation Status i
~
Ucensees achieve implementation of MPA items either by incorporating corrections Into the plant design before licansing or by making the modifications necessary to meet the requested, required or voluntary actions at licensed plants. The information i
presented here includes all MPA items related to the 110 licensed plants considered in i
this report.
i Approximately 84 percent of the MPA items have been implemented at licensed plants.
Of the 7,175 applicable items,6,197 have been completed and 978 remain open from an implementation standpoint. On average, each plant has less than 9 remaining items to implement. No plant has more than 14 remaining items except Browns Ferry Units 1 and 3. Each unit has 22 items. Figure 5.1 presents the overall status of, and progress on, MPAs. Of the 110 licensed plants, none have fully implemented all applicable MPAs. Table 5.2 lists the number of unimplemented MPA items by plant 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 90-01, BL 92 01, GL-88 20, GL 8910, GL-92-01, and MPA-B118 account for 62 percent of the 978 unimplemented items. Figure 5.2 summarizes the implementation status of these issues. A brief description of these six MPAs and those MPAs with more than three open items follows:
9 BL-88-02 Steam Generator Tube Rupture (X802)
This issue is an outgrowth of the steam generator tube rupture event which occurred at North Anna 1 on July 15,1987. As a result of this event, the staff concluded that the necessary general conditions leading to a rapidly propagating fatigue failure were the denting at the upper support plate, a fluid-elastic stability ratio approaching that for the tube that ruptured at North Anna, and the absence of effective antivibration bar support.
The NRC issued Bulletin (BL) 88-02 on February 5,1988, to all holders of operating licenses or construction permits for Westinghouse-designed nuclear power reactors with steam generators having carbon steel support plates. The bulletin required,
SIMS issue Numbers and Corresponding MPA Numbers SIMS MPA Item No, Ng, SIMS Title BL-88-02 X802 Steam Generator Tube Rupture (BL 88-02)
BL-88-04 X804 St Pump Failure (BL 88-04)
BL-88-08 X808 Thermal Stress in Piping BL-88-10 X810 Circuit Breaker Material Problems (BL 88-10)
BL-88-11 X811 Thermal Stratificatxm in PZR Surge Line (BL 88-11)
BL-89-32 X902 Stess Corrosen Cracking of Anctor Darling Check Valve Botting (BL 89-02)
BL-90-01 X001 Loss of FZ' Oil in Rosemount Transmitters (BL 90-01)
BL-92-01 X201 Themmi Lagging 330 (BL 92-01)
GL-84-09 A019 Recombiner Capability BWR Mark i 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 Problems in BWR Piping GL-88-11 A023 R.G.1.99 Rev 2 (pressunzed Thermal Shocx Rule) (GL 88-11) 4 GL-88-12 D022 Removal of Fire Protection Tech Specs (GL 88-12) 4 r,o GL-88-16 D021 Removal of Cycle-Specific Parameter Limits (GL 88-16)
GL-88-20 B111 Indandual Plant Evaluations (GL 88-20)
GL-89-01 0025 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 Valve Testing and Surveillance (GL 89-10)
GL-89-16 8112 Installation of Hardened Wetwell Vent (GL 89-16)
GL-9049 D028 Elimination of 325 Requirement in Tech Spec 4.0.2 (GL 90-09)
GL-91-01 D029 Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens GL-91-06 L106 Adequacy of Safety-related DC Power Supplies (GI A-30)
GL-91-08 D030 Removal of Component Lists from Tech Specs (GL 91-08)
GL-91-11 L111 Vital Instruments Buses and Tie Breakers (Gi-48,49)
GL-92-01 B120 Reactor VesselStructuralIntegnty GL-92-04 B121 BWR Water LevelInstrumentation MPA-B11G B116 Consder Results of Spunsvimi Motor-operated Tests (GL 89-10. Supp 3)
MPA-B117 B117 Failure of Westinghouse SG Tube Mai.on.2 Plugs (BL 90-01, Supp 2)
MPA-B118 B118 IPE Extemal Events (GL 88-20. Supp 4)
Table 5.1 i
Summary of Unimolemented MPA Items by Plant items items items PLANT Remaining PLANT Remaining PLANT Remaining Arkansas 1 7
Haddam Neck 7
Prairie Island 2 9
Arkansas 2 7
Harris 1 9
Quad Cities 1 11 Beaver Va!!ey 1 6
Hatch 1 12 Quad Cities 2 11 Beaver Valley 2 6
Hatch 2 12 River Bend 1 6
Big Rock Point 1 10 Hope Creek 1 7
Robinson 2 7
Braidwood 1 9
Indian Pt 2 7
Salem 1 12 i
Braidwood 2 9
Indian Pt 3 7
Salem 2 12 i
Browns Ferry 1 22 Kewaunee 6
San Onofre 1 2
Browns Ferry 2 13 LaSalle 1 11 San Onofre 2 8
Browns Ferry 3 22 LaSalle 2 11 San Onofre 3 9
Brunswick 1 11 Limerick 1 7
Srabrook 1 6
Brunswick 2 11 Limerick 2 7
Sequoyah 1 7
Byron 1 8
Maine Yankee 8
Sequoyah 2 7
Byron 2 8
McGuire 1 8
South Texas 1 8
Callaway 1 11 McGuire 2 7
South Texas 2 7
Calvert Cliffs 1 7
Mi!Istone 1 10 St. Lucie 1 7
Calvert Cliffs 2 7
Millstone 2 10 St. Lucie 2 7
4 Catawba 1 7
Millstone 3 6
Summer 1 4
9 Catawba 2 7
Monticello 9
Surry 1 7
Clinton 1 13 Nine Mile Pt 1 12 Surry 2 7
Comanche Peak 1 7
Nine Mile Pt 2 9
Susquehanna 1 9
Cook 1 12 Nonh Anna 1 8
Susquehanna 2 9
Cook 2 1O North Anna 2 8
Three Mile Island 1 7
Cooper Station 13 Oconee 1 9
Trojan 7
Crystal River 3 9
Oconee 2 9
Turkey Pt 3 8
Davis-Besse 1 8
Oconee 3 9
Turkey Pt 4 8
Diablo Canyon 1 8
Oyster Creek 1 12 Vermont Yankee 1 10 Diablo Canyon 2 9
Palisades 9
Vogtle 1 9
Dresden 2 11 Palo Verde 1 11 Vogtle 2 9
Dresden 3 10 Palo Verde 2 11 Washington Nuclear 2 8
Duane Arnold 9
Palo Verde 3 11 Waterford 3 7
Farley 1 7
Peach Bottom 2 10 Wolf Creek 1 7
Farley 2 7
Peach Bottom 3 10 Zion 1 10 l
Fermi 2 8
Perry 1 9
Zion 2 7
Fitzpatrick 9
Pilgrim 1 8
Ft Calhoun 1 6
Point Beach 1 8
i Ginna 14 Point Beach 2 8
l Grand Gulf 1 8
Pra!rle Island 2 11 Table 5.2
_. _ -. _ _ _ _ ~
I l
Other MPA issues j
Implementation Status at Licensed Plants l
l 8000 7000 l
6000 5000 4000 h3000 2000 1000 Applicable Implemented Unlmplemented Figure 5.1 74-l
l Summary of Six Unimplemented MPAs j
i 120 110 110 110 no no no 109 109 107 104 100 --
90 l
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BL-90-01 BL-92-01 GL-88-20 GL-89-10 GL-92-01 MPA-B118 1-SIMS Item Number E Applicable Phuts f
)
l-o Imphnnented Plants a Unimplemented Plants Figure 5.2
{
i
licensees to submit within 45 days a report detailing the status of their compliance with the actions specified in the bulletin.
The initial action was to examine the most recent inspection data for evidence of tube denting at the uppermost support plate. If inspection records were not adequate, an inspection of a sufficient number of tubes was required at the next refueling outege, if there was evidence of denting, implementation of an enhanced primary to secondary leak rate monitoring program was required pending NRC stati review and approval. Enhanced leak rate monitoring was also to be implemented if future inspections were required to confirm that no denting existed.
The monitoring to be described in the 45 day report was to assess and implement j
as necessary, the need for long term corrective actions to minimize the probability i
of a rapidly propagating fatigue failure such as occurred at North Anna 1. Plants with no evidence of denting at the uppermost support plate were required to commit to the bulletin denting actions if future inspections provided evidence of denting.
BL 88-04 SI Pumo Failure (X804)
This issue arose when a licensee requested that Westinghouse determine if parallel pump operation while on miniflow is acceptable. The review determined that the potentia! exists for the stronger pump to dead head the weaker pump during parallel pump operating conditions while on miniflow only. In addition, it was determined that even without pump interaction the recirculation flow available was not adequate to ensure continuous operation of even a single pump on miniflow.
To resolve this issue, the NRC staff issued BL 88-04, ' Potential Safety Related Pump Loss," on May 5,1988. The purpose of the bulletin was to (1) alert all holders of operating licenses and construction permits to the above two concerns, (2) request addressees to determine if they have similar problems at their facilities, (3) take appropriate actions, (4) provide a schedule for their short and long-term corrective actions, and (5) report completion of theli activities.
All operating reactor licensees have responded to BL 88-04. Approximately one third of the licensees identified the need for procedure or design changes.
Implementation of these changes is completed for most plants. However, the staff has noted substantial variations in the level of information provided by licensees.
The pump dead heading concerns have been adequately addressed for the most part, but additional efforts and better understanding is needed to resolve minimum flow adequacy concerns. The minimum flow issue has been identified to the NRC office of Nuclear Regulatory Research (RES) as a potential generic issue and is currently in the prioritization process, to assess the safety significance of the issue and the need for additional NRC guidance. Additional work, if any, will be under a separate MPA..
BL 88-08 Thermal Strestin Piplag ConnectettQ 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 result 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 monitoring 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 was 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 contained insufficient information.
BL-88-10 Bonconforming Molded - Case Circuit Breakers (X810)
This issue arose when information was received from Pacific Gas and Electric Company that Square D circuit breakers, purchased for non safety related applications, were apparently rebuilt instead of being new as the purchase order specified.
On June 3,1988, U.S. marshalls seized equipment and records related to Square D products from five firms in the Los Angeles area that were refurbishing and marketing the circuit breakers as new equipment. A review of the seized records identified utilities and other government agencies that had received material from the five firms. On July 8,1988, the NRC issued Information Notice (IN) 88-46 alerting the industry to the problem On July 21,1988, the NRC issued Supplement 1 to IN 88-46 and on December 20,1988, Supplement 2 to IN 88-46.
The NRC issued BL 8810, "Non-conforming Molded-Case Circuit Breakers," on November 22,1988. The bulletin requested licensees to determine the traceability of a sample of molded-case circuit breakers. If traceability to the original manufacturer could not be established, the sample was to be expanded. Also, any non-traceable breakers installed in the plant had to be replaced. The NRC issued _...
o
supplement 1 to BL 88-10 on August 3,1989, to clarify the requests of the original bulletin. Initial responses were required by April 1,1989. All corrective actions required as a result of the reviews conducted by licensees were requested to be completed after two refueling outages, commencing after March 1,1989.
BL 8811
. Pressurizer Surge Line Thermal Stratification (X811)
This issue arose from staff concerns over the unexpected movement of the pressurizer surge line observed by the licensee at the Trojan plant. The extent of the movement was such that the piping actually contacted several pipe whip restraints. The licensee's investigation determined that the problem was a result of thermal stratification.
The NRC issued IN 88-80, " Unexpected Piping Movement Attributed to Thermal 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,"
on December 20,1988, describing a series of short-and long term actions to address the problem.
Licensees of operating PWRs were required to (1) perform a visual inspection of the surge line at the first 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 regulatory commitments for the licensed life of the plant. If the analysis did not show 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. The issue is closed for some Westinghouse plants, including plants that required modifications to meet the applicable Code requirements. The issue remains open for the remainder of the Westinghouse plants pendin0 verification of the Westinghouse Owners Group generic detailed analysis. The staff issued a safety evaluation on the Westinghouse analysis in May of 1991. The staff is still l
reviewing open items in the analyses of the CE and B&W Owners Groupt BL-89-02 Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel (X902)
This issue arose as a result of finding Anchor Darling check valves with cracked and broken internal bolting at the Diablo Canyon and D.C. Cook nuclear plants.
The failures were the result of stress corrosion cracking of bolis that (1) are constructed of type 410 or other high-hardness types of stainless steel, (2) are internal to the valve and (3) are preloaded.
The purpose of BL 89-02 was to address this problem and request certain actions.
Specifically, BL 89-02 requested licensees to identify, disassemble, and inspect l
L - -
safety related swing check valves containing internal botting made of high-hardness stainless steel, if the botting was determined to be cracked or broken, or was of sufficiently high hardness that it was susceptible to stress corrosion cracking, licensees were requested to replace the botting with lower hardness material.
BL-90-01
. Loss of Fill-Oil in Transmitters Manufactured by Rosemount (X001)
This issue arose when the staff found that the reported failures of transmitters at Northeast Utilities's Millstone 3 in 1987 were the result of a gradual loss of fill-oil from the transmitter's sealed sensing model.
The staff issued IN 89-42, " Failure of Rosemount Models 1153 and 1154 Transmitters," on April 21,1989, to alert the industry of the loss of oil fill problem.
Rosemount, the vendor for the transmitters, investigated the causes of these 1
failures and also confirmed the potential for leaking fill oil, particularly in Model 1153, Series B & D and Model 1154 (suspect lots). The staff further issued BL 90-01, " Loss of Fill Oil in Transmitters Manufactured by Rosemount," on March 9,1990, which requested licensees to promptly identify and take appropriate corrective actions for Model 1153, Series B, Model 1153, Series D, and Model 1154 transmitters that may be or have the potential for leaking fill oil.
During the summer and fall of 1990, the Nuclear Management and Resource Council (NUMARC) surveyed the industry to gather data on all Models 1153 and 1154 transmitters, as well as safety related Models 1151 and 1152 at all commercial nuclear facilities. The purpose of the survey was to address the staff concerns related to installed transmitters and the appropriateness of the Rosemount methodology. The staff has reviewed licensee responses, data from licensee event reports, technical information from Rosemount, site visits, and NUMARC Report 91-02, " Summary Report of Numarc Activities to Address Oil Loss in Rosemount Transmitters, " and held numerous meetings with the industry. Based on the review and the meeting, the staff found a relationship, as had been previously identified by Rosemount and NUMARC, between operating pressure and time-in service in identifying where the transmitters would most likely fail.
The staff concludes that actions requested by BL 90-01 are insufficient in that they do not achieve the desired high functional reliability and proposes to issue a supplement bulletin (SBL) with requested actions that will supersede the actions requested in the original bulletin. A draft SBL was published for public comment in the Federal Register on April 27,1992, and submitted to the NRC Committee to Review Generic Requirements (CRGR) on August 19,1992. The staff is currently incorporating CRGR comments into the staff proposed SBL, and when approved, the SBL will be issued. Upon issuance of this SBL, the staff plans to close out the original bulletin (BL 90-01). The staff closecut effort is expected to begin in late December 1992. 1
I BL-92-01 Failure of Thermo Lao 330 Fire Barrier S.yMem (X201)
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 j
endurance tests performed by the Gulf States Utilities Company on Thermo-Lag i
330 fire barrier systems installed on wide aluminum cable trays and the associated 4
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 330 fire barrier material. Recognizing the concerns stated in ins 91-47 and 9179 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-Lag 330 fire barrier system to perform its specified function as a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 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 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 which use this material for protection and separation of safe shutdown capability, (2) implement appropriate compensatory measures for an inoperable barrier, and (3) verify that the requested actions have been taken and describe the measures being taken to assure operability.
The responses to BL 92-01 Indicate that 83 operating plants have Thermo Lag fire barrier material installed and appropriate compensatory measures have been implemented. The staff will review responses to Supplement 1 when they are received. An action plan has beea developed to address the concerns identified by the special review team. Resolution is scheduled for the end of 1993.
GL-84 09 Recombiner Caoability Reouirements of 10 CFR 50.44 [Q){3){Q LA01%
The issue arose as a result of the TMI-2 accident. 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. As a result,-the staff revised 10 CFR 50.44, " Standards for Combustible Das Control Systems," effective __
January 4,1982 (48 FR 58484) to address this safety concern. For plants with Mark I and M k 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 l 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 d3 sign 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.
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 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 iri containment other than that resulting from radiolysis of the reactor coolant.
GL 87-09 Sections 3.0 and 4.0 of the Standard Technical Specifications on the ADolicability of Limitina Conditions for Ooeration and Surveillance ReauirementtM024)
The NRC issued GL 87-09 on May 4,1987 to all power reactor licensees and applicants to provide guidance for three specific problems that have been encountered with the general requirements on the applicability of limiting conditions for operation (LCO) and surveillance requirements in Section 3.0 and 4.0 of the Standard Technical Specifications. The first problem involves unnecessary restrictions on mode changes by Specification 3.0.4 and inconsistent application of exceptions. The specification was changed to define the conditions under which the requirements apply. The second problem involves unnecessary shutdowns required by Specification 4.0.3 when surveillance intervals are inadvertently exceeded. The action requirement applicability was clarified to specify an acceptable time limit for completing a missed surveillance in certain instances and to clarify when a missed surveillance constitutes a violation of the operability requirements of an LCO. The third problem involves possible conflicts between Specification 4.0.3 and 4.0.4. Staff guidance clarified these conflicts.
Implementation of the guidance contained in GL 87-09 is voluntary. -.m,
+
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i GL-88-01 NRC Position on IGSCC in BWR Austenitic Stainless Steel Pioing 16091) i 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). BWR licensees and construction permit holders 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 included a commitment to revise technical specifications (TS) to be consistent with the NRC staff positions in GL 88-01.
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-8811 NRC Position on Radiation Embrittlement of Reactor Vessel Materials
.(602.3)
Revision 2 to RG 1.99, " Radiation Embrittlement of Reactor Vessel Materials,"
became effective in May,1988. GL 88-11 was issued on July 12,1988,and 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 estirnate of vessel beltline embrittlement (except those for pressurized thermal shock).
GL 88-11 requested that all licensees of operating reactors use the methods described in Rev. 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 then required to submit the results of their analyses and an implementation plan for proposed actions within 180 days of the issuance of the generic letter. Responses have been received from all licensees.
GL-88-12 Removal of Fire Protection Reouirements from Technical
_ Specifications (QQ22)
The NRC issued GL 86-10, " Implementation of Fire Protection Requirements," on April 28,1986, which provided information and the staff's interpretation concerning the implementation of the fire protection requirements contained in Appendix R to 10 CFR Part 50. GL 86-10 allowed licensees to remove the descriptions of and requirements for fire protection systems from the plant TS and implement the fire protection program by a license condition. To remove the fire protection requirements from the TS, licensees must request an amendment to their license.
This generic letter was issued for information only. No licensee response was required.
1.
GL 8812, ' Removal of Fire Protection Requirements from Technical Specifications,'
provided specific guidance for the removal of references to the fire protection systems from the TS and for the addition of a license condition for implementation of the fire protection program. This is a voluntary action based on guidanes contained in GLs 8610 and 8812, which permit individual licensees to remove fire protection references from their plant TS.
GL 8816 Removal of Cycle Specific Parameter Limits From Technical
. Specification (D021)
The NRC issued GL 8816,
- Removal of Cycle Specific Parameter Umits," on October 4,1988. It provided guidance for the preparation of a license amendment request to modify TS that have cycle specific parameter limits, implementation of GL 8816 is a voluntary action on the part of licensees. An acceptable alternative to specifying the values of cycle specific parameter limits in TS was developed on the basis of the review and approval of a lead plant proposal for this change to the TS for the Oconee units. The implementation of this alternative will result in a resource savings for the licensees and the NRC by eliminating the majority of license amendment requests on changes in values of cycle specific parameters in TS.
This alternative consists of the following three separate actions required to modify the plant's TS: (1) the addition of the definition of a named formal report that includes the values of cycle-specific parameter limits that have been established using an NRC approved methodology and that is consistent with all applicable limits of the safety anal sis, (2) the addition of an administrative reporting
/
requirement to submit the formal report on cycle-specific parameter lirnits to the Commission for information, and (3) the modificatk,n of individual TS to note that cycle specific parameters shall be maintained within the limits provided in the defined formal report.
GL-88 20 Individual Plant Examination for Severe Accident Vulnerability 10 CFR 50.54(F) (B111)
The NRC issued GL 88-20, " Individual Plant Examination for Severo Accident Vulnerability," on November 23,1988. The purpose of this generic letter was to:
(1) request holders of operating licenses and construction permits to perform an individual plant examination of their plant specific internal event severe accidents and (2) report the results of their analysis. No response to the generic letter was required until 60 days after receipt of the forthcoming final guidance documents.
The NRC issued Supplement 1 to GL 88-20, " Initiation of the Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54," on August 29, 1989. Supplement 1 required all holders of operating licenses and construction permits to submit an individual plant examination (IPE) report approximately 3 years from the date Supplement 1 was issued. The purpose of the IPE was to identify 83-
plant specific severe accident vulnerabilities using probabilistic risk analysis methodology.
The staff established a two step review process to review the IPE reports submitted to the NRC in response to GL 88 20, Supplement 1. The e J performs a relatively short Step 1 review of each IPE submitted: (1) to determine whether the licensee's IPE process meets the intent of GL 88 20 and (2) to store important IPE insights and findings in a data base for future use. The staff performs a more detailed Step 2 review on only selected IPE submittals. The IPEs selected for a Step 2 review are normally those (1) for plants with Stop 1 review findings that appear inconsistent with the staff's probabilistic risk assessment (PRA) experiences or expectations, suggesting weaknesses in the applied methodology or the plant's operational characteristics or (2) for plants with unique characteristics that are not well understood.
The IFE effort is more complex than estimated. Licensees have delayed submittal of several IPEs for 2 to 18 months. The staff has issued two evaluation reports documenting the results of the review for Seabrook and Millstone 3 IPEs The following three plants heve been selected to date for Step 2 IPE reviews: Turkey Point 3, Turkey Point 4, and Fitzpatrick.
GL-89-01 Jmplementation of Prooram Controls for RETS in Administration Control section (D025)
The NRC issued GL 89-01 on January 31,1989 to all power reactor licensees and applicants 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 of procedural details of current RETS to the Offsite Dose Calculation Manual (ODCM) or process control program (PCP). The NRC staff's intent in recommending these changes to the TS is to fulfill the goal of the Commission Policy Statement for Technical Specification improvements. 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. This GL provided guidance to licensees for voluntary TS revisions.
GL 89-04 lST Reviews and Schedules (A025)
This issue arose from staff observation that certain generic weaknesses were being observed in licensee Inservice Testing (IST) programs. The weaknesses noted stemmed from a lack of understanding of code testing requirements, technical specification requirements, and acceptable alternatives to code requirements.
The NRC issued GL 89-04, " Guidance on Developing Acceptable Inservice Testing Programs," on April 3,1989. The generic letter noted that certain plants need not 84-
respond occause the staff had recently, or would in the near future, review the plant's IST program and would address the positions of GL 89-04 in the SER for that program. 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 were also required to submit a schedule for equipment and program modifications required as a resuit of the review. GL 89-04 granted approval for licenses to change their 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 lett6t. The staff has completed individual reviews for most of these plants.
GL-89-06 Task Action Plan item I.D.2 SPDS (F072)
On October 31,1980, the NRC issued NUREG-0737, which provided guidance for implementing TMI 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.
On April 12,1989, the NRC staff issued GL 89-06, which required all licensees (except for 15 specific plants) to provide information to the staff 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-8910 Egfelv Related Motor-Ocerated Valve Testing and Surveillance (B110)
The purpose of GL 8910 is to (1) alert all holders of operating licenses and construction permits to problems concerning the operability of safety-related motor-operated valves, (2) request addressees to establish programs to demonstrate the operability of these valves and to ensure continued operability over the life of the plant, (3) provide a commitment to establish such a program and complete the demonstration of operability within the timeframe specified in GL 8910, and -..
(4) report completion of the demonstration phase of their programs. The subject matter of this generic letter is related to that of DL 85-03, " Motor-Operated Valve Common Mode Failures During Plant Transient Due to Improper Switch Setting,"
and its supplements.
Supplement 1 to GL 8910, *Results of the Public Workshops," was issued in June 1990. The purpose of this supplement was to forward the opening remarks by the NRC representatives and the responses provided by the NRC staff to all significant questions at the public workshops held to discuss the original generic letter.
Supplement 2 to GL 89-10, " Availability of Program Descriptions," was issued in August 1990. The purpose of this supplement was to announce the staff's decision to delay commencement of onsite inspections for 4 months to allow licensees time to incorporate the Information provided by Supplement 1 into their motor-operated-valve (MOV) programs.
Supplement 3 to GL 8910, " Consideration of the results of NRC Sponsored Tests of Motor Operated Valves," was issued in October 1990. (Unlike the GL ltself and the earlier supplements, this supplement is tracked under MPA B 116.) The purpose of this supplement was to request the addressees to (1) assess the applicability to GL 8910 of the data from the NRC sponsored MOV tests, (2) determine the as is capability of the high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), and reactor water cleanups (RWCU) MOVs described therein, (3) identify any deficiencies in those MOVS, (4) where applicable, evaluate the MOVS used for containment isolation !n the line to the isolation condensers, (5) perform a plant specific safety assessment to verify that the generic safety assessment performed by the staff and the BWR owners group were applicable, (6) consider the implementation cf short-term corrective actions, and (7) report their findings.
Supplement 4, " Consideration of Valve Mispositioning in Bolling Water Reactors,"
was issued in March 1992. The purpose of this supplement was to rescind the recommendation that licensees for BWRs consider inadvertent mispositioning of valves when determining the valves to b6 included in the program described in GL 8910.
GL-8916 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 vont capability, (2) improved reactor pressure vessel depressurization system reliability, (3) an alternative water supply to tiie reactor vessel and drywell sprays, and (4) updated emergency procedures and training.
On September 1,1989, the staff issued GL 89-16 informing the BWR Mark I plant licensees that the recommended plant improvements, with the exception of hardened wetwell vent installation, should be evaluated by the licensees as a part of the IPE program. The staff requested that the licensees make the required i
modifications to the hardened vent capability voluntarily under the provisions of 10 CFR 50.59 of the commission's rules. The staff further stated that for those plants who's licensees elected not to make modifications voluntarily, the staff would initiate a plant specific backfit analysis. Where the backfit analysis supports the impositions of a requirement to provide a hardened wetwell vent capability, the staff would issue orders for such modifications.
The licensees coordinated their response to GL 8916 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.
GL 90-09 Alternative Reouirements for Snubber Visual Insoection intervals and Corrective Actions (D028)
NRC issued GL 90-09, ' Alternative Requirements for Saubber Visual Inspection Intervals anct Corrective Actions," on December 11,1990. The purpose of GL 90-09 was to provide guidance for alternative roothods for determining visual inspection intervals for snubbers. The staff sought to maintain the same confidence level in the new Inspection interval criteria while recognizing that licensees were spending significant resources and experiencing unnecessary radiological exposure under the existing criteria.
GL 90-09 encouraged licensees to voluntarily request changes to the TS implementing the staff guidance on inspection intervals and corrective actions. The alternative inspection interval is based on the number of unacceptable snubbers found during the previous inspection in proportion to the size of snubber populations in various categories.
GL-91-01 Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens (D029)
The purpose of GL 91-01 was (1) to provide all holders of operating licenses and construction permits with the staff's recommended technical specification wording with respect to reactor vessel material specimens, and (2) to request all addressees to voluntarily propose revised TS that are consistent with these staff positions for their faci!!'ies.
GL 91-06 Adeguacy of Safetv Related D.C. Power Sucolies (L106)
The purpose of GL 91-06 was (1) to inform all holders of operating licenses that the staff had completed the evaluation of Generic issue A 30, " Adequacy of Safety-Related DC Power Supplies,
- as part of the resolution of Generic issue 128,
" Electrical Power Reliability," and (2) to request addressees to provide written responses to a number of questions pertaining to multiple and common cause i
failures of their DC electrical systems. The information was req' ired to be u
submitted to NRC, signed under oath and affirmation, within 180 days.
The specific areas of concern of GI A 30 are the multiple and common causes failures of the safety related de power supply systems. The staff believes that certain maintenance., surveillance, and monitoring provisions are appropriate for safety related do systems and that most plants have already implemented a major portion of these provisions because of the number of actions taken by the staff and industry. During the development of plans to resolve Gl A 30, it was observed by the staff that several previously issued notices, bulletins, and letters included l
recommendations similar to those that have been identified to resolve Gl A 30.
{
More specifically, it has been determined that the provisions of BL B5 74, BL 79-27, the Institute of Nuclear Power Operations Significant Operating Experience Report -
(SOER) 83-6, and separate actions taken to resolve GI 49, included the elements that would resolve GI A-30. The response to the GL 91-06 questionnaire was necessary to provide the staff with information to determine whether any further regulatory action was required.
All licensee certified responses to the GL 91-06 questionnaire have been received by NRC. Review of licensee responses is being performed by RES, Division of Safety issues Resolution (DSIR).
1 GL 9108 Removal of comoonent Lists from Technical Soecifications (QQ30)
GL 91-08 was issued on May 6,1991 to all power reactor licensees and applicants to provide guidance for licensee amendment requests to remove component lists from TS. Guidance was also provided for the Incorporation of component lists into plant procedures. The removal of the component lists from TS permits administrative control of changes to these lists without processing licensee 1
amendment requests. Any request by licensees to implement the guidance contained in GL 91-08 is voluntary.
GL 91 11 Vital Instrument Buses and Tie Breakers (GI-48 & 49) (L111)
The staff evaluated the concerns raised in Gls 48, "LCOs for Class 1E Vital Instrument Buses," and 49, " Interlocks and LCOs for Class 1E Tie Breakers." The staff concluded that the concerns identified could be resolved by the verification or implementation of appropriate administrative controls in plant procedures for Class 1E buses and tie breakers.
. 88-
j GI-48 was initiated upon discovering that some operating nuclear power plants do not have any administrative controls governing operational restriction for their Class 1E 120 V ac vital instrument buses (VIBs) and associated inverters. Without such restrictions, the normal or alternato power sources for one or more VIDs could be out of service indefinitely. This condition could prevent certain safety systems from meeting the plant design basis.
The issue described in GI 49 arose because of an ir'cident that occurred at Point Beach 2 in 1980. A tie breaker between safeguard wses at the plant was discovered left closed after a plant shutdown. The concern was that independence i
between redundant safeguards buses can be lost if manually actuated tie breakers are inadvertently left closed.
GL 9111 required all holders of operating licenses to certify that plant procedures included time limitations and surveillance requirements for vital instrument buses, inverters or other onsite power sources to the vitalinstrument buses, and tie breakers that can connect redundant Class 1E buses between units at the same site, if plant procedures did not include time limitations and surveillance requirements as requested, a documented evaluation was needed to justify why such provisions were not needed.
GL-92-01 Reactor Vessel Structural Integrity (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 j
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 50.60,10 CFR 50.61, Appendices G and H to 10 CFR Part 50, and are fulfilling the requirements of GL 8811, "NRC Position On Radiation Embrittlement of Reactor Vessel Materials and its impact On Plant Operations." The information will also be used to provide the information requested in the Staff Requirements Memorandum (SRM) M910711 A, dated July 19,1991. The SRM directed the staff to provide the Commission with a list of any other plants (besides Yankee Rowe) that have not, but should have, requested exemptions from the requirements of Appendiv H to 10 CFR Part 50.
I -
i j
GL 92-04 Enactor Vessel Water Level Instrumentation in BWRs (B121) l The staff issued IN 92 54,
- Level Instrumentation inaccuracles Caused by Rapid l
Depressurization," on July 24,1992. IN 92 54 was issued to alert BWR licensees to the possibility of noncondensible gases becoming dissolved in the reference leg of l
BWR water level instrumentation, which could lead to a false high-level Indication i
after a rapid depressurization event. On July 29,1992, the staff held a public meeting with the Regulatory Response Group of the BWROG to discuss th3 effect j
of inaccuraf.,les in the reactor vessel level instrumentation system in BWRs.
The staff issued GL 92-04,
- Resolution on the Issues Related to Reactor Vessel i
Water Level instrumentation in BWRs Pursuant to 10 CFR 50.54(F) (GL 92-04)," on August 19,1992. The purpose of this generic letter was to alert all holders of operating licenses for BWRs to errors related to instrumentation accuracy in BWR water level instrumentation and the results of the staff's review of the BWROG's generic analysis of these errors and to request addresses to (1) determine the impact of these errors on automatic safety systems response, operator short-and l
lorig term actions and emergency operating procedure at their facilities, (2) take i
short-and long term corrective actions, and (3) submit a report that includes the i
results of their determinations, a discussion of their short. and long-term actions, j
and the schedule for completion of their long-term programs.
MPA B116 Results of NRC Testino of MOVs (GL-8910. Suco 3) (B116) i 1
On June 5,1990, the staff issued IN 90-40, *Results of NRC Sponsored Testing of Motor Operated Valves (MOVs)," which described the results of tests on MOVs.
t These tests are part of an ongoing research effort and_were conducted on gate valves typically used to provide containment isolation in the steam supply lines of j
the HPCI, RCIC systems, and the supply line for the RWCU system of BWRs.--The test 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.
l As a result of the findings, the staff issued Supplement 3 to GL 8910 on
- October 25,1990, which described required actions for licensees of BWRs.-
n Ucensees were required to provide (1) criteria reflecting operating experience and the latest test data that were applied in determining whether the deficiencies exist in j
the subject MOVs (2) a list of the MOVs found to have deficiencies, and (3) a j
schedule for the necessary corrective action.
P 4
_ _. _ _ _.. _ _ ~. _, - _ _ _.
MPA B117 Failure of WLstinghouse SG Tube Mechanical Plugs (BL 89-01.
Suco 2) (B117)
BL 89-01 requested that licensees determine whether certain mechanical plugs supplied by Westinghouse were installed in their steam generators (SG). If so, affected licensees were requested to implement an action plan (including plug repairs and replacements) to ensure that the mechanical plugs continue to provide adequate assurance for reactor coolant pressure boundary integrity. On the basis of additional Information, Supplement 1 to BL 89-01 was issued January 14,1990, to alert licensees of the potential for primary water stress corrosion (PWSC) cracking in SG tube plugs not addressed in BL 89-01.
Supplement 2 to BL 89-01 was issued June 28,1991, and requested that actions =
similar to those requested in BL 89 01 be extended to include all Westinghouse mechanical plugs fabricated from thermally treated Inconel 600. These actions were necessary to ensure that the mechanical plugs will continue to provide adequate assurance of reactor coolant pressure boundary integrity under normal operating, transient, and postulated accident conditions.
MPA B118 lPE External Events (GL-88 20. Suco 4) (B118) l l
The staff issued Supplement 4 to GL 88-20 on June 28,1991 to initiate the IPE process for external events. Five categories of external events were specified and licensees were required to submit to a schedule and methodology by December 26,1991. Licensees were requested to submit the results of the individual plant examination of external events (IPEEE) within 3 years of the issuance date of Supplement 4, or 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 to GL 88 20.
Alllicensee responses to Supplement 4 have been received and reviewed independently and jointly by NRR and RES. A common difficulty with a large number of the responses was the linkage of IPEEE to implementation of the USl A-46 resolution required by GL 87-02 to verify seismic adequacy of mechanical and electrical equipment. Supplement 4 to GL 88 20 encouraged licensees to combine the walkdown that would be required for the seismic portion of the IPEEE with the walkdown required by GL 87-02.
Supplement 1 to GL 87-02 was issued on May 22,1992, approving the seismic qualification utility group generic implementational procedure for USl A 46 implementation and starting the clock for both A-46 and the IPEEE. Following i
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 l
licensee's response would be considered acceptable. The NRC staff reported the l
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 1
o _ _. _
the IPEEE results must be submitted to the staff by June 1995 or within 3 years after issuance of tlie staff's evaluttion approving the A-46 GIP, whichever occurs first. Therefore, since the evaluation was issued on May 22,1992 with GL 87-02, Supplement 1, the latest acceptable date for IPEEE sublaittal becomes May 22,1995.
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 eight singled out by the Eastern United Stabs Seismic Hazards Program as needing further NRC staff evaluation.
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 MPS 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.
Tis provide guidance for the field verification of licensee implementation of other MPAs.
The NRC issued thirteen Tis for 13 individual MPA issues, which cover a total of 812 items at the 110 licensed plants. Upon initialincpection of certain items and further review by the regional offices,92 items covered by the Tis were found to be inapplicable from a verification standpoint, leaving a total of 720 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 30,1992,587 items (80 Percent) had been verified. Tl designations and the corresponding MPAs are summarized in Table 5.3.
Table 5.4 summarizes the items remaining to be verified.
92-
Temporary Instructions for Resolved MPAs SIMS ltem Mi%
SIMS Title ILNutohnt BL 7915 B031 Deep Draft Pump Deficiencies 2500/01 BL-80-11 B059 Masonry Wall Design 2515/037 BL-88-04 X804 S1 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 Quality Assurance Requirements Regarding Diesel 2515/093 Generator Fuel Oil GL 8121 B066 Natural Circulation Cooldown 2515/086 GL-83-08 D021 Modification of Vacuum Breakers on 2515/096 Mark l Containments GL-89-07 L907 Power Reactor Safeguards Contingency Planning for 2515/102 Surface Vehicle Bombs GL-89-10 B110 Safety Related Motor Operated Valve Testing and 2515/109 Surveillance 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
.~.
-..-.-.~._- _ _
a Summary of Other MPA Items Requiring Verification Plants Plants Plants I
SIIAS Item Covered Reouired Verified BL-79-15 DEEP DRAFT PUMP DEFICIENCIES 110 104 104 i
BL-63-11 MASONARY WALL. DESIGN 65 65 63 I
BL-88-04 St PUMP FAILURE (BULLETIN 88-04) (OLD MPA 107 37 37 B103) t BL-88-07 POWER OSCILLATIONS IN BIOILING WATER 37 37 36 REACTORS (BWRS) l
(
GL-80-002 OUAtJTY ASSURANCE REQUIRMENTS 41 38 37 REGARDING DIESEL GENERATOR FUELOIL GL-81-21 NATURAL CIRCULATION COOLDOWN 73 72 62 l
i i.
GL-83 MODIFICATION OF VACUUM BREAKERS ON 23 22 21 i
MARK I CONTANMENTS.
T GL-89-07 POWER REACTOR SAFEGUARDS 110 110 110 COTINGENCY PLANNING FOR SURFACE i
VEHICLE BOMBS l
GL-89-10 SAFETY-RE!.ATED MOTOR-OPERATED 110 110 1
VALVE TESTING AND SURVEILI.ANCE i
t MPA-B003 PWR MODERATOR DILUTION 39 36 34 i
MPA-B011 FLOOD OF EQUIPMENTINPORTANTTO 10 4
1 SAFETY
(
MPA-8041 FIRE PROTECTION - FINALTECH SPECS 66 64 62 (NOLUDING SER SUPPLEMENTS) j l
1
}.
. MPA-C002 BWR-RECIRC. PUMP TRIP (ATWS) 21 21 19 i
)
1.
t j'
NOTE: Plants Covered are those for which MPAs are apphcable Plants Required are those plants requinng field verscation-i 1'
Plants covered but for whch field venicatson is not necessary have implernented
[
the resotution in a manner fx1 requinng plant hardware changes.
}
j Table 5.4
5.3 Status by Plant Table 5.5 summarizes information on the status of implementation and verification of MPAs at all licensed phnts. 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 covsred 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 110 plants, none have completely implemented all MPA items. On average, each plant has less than riine remaining items to implement. No plant has more than 14 remaining items, with the exception of Browns Ferry 1 and 3, which have 22 each, i
m r
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S) - SLWMARY BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED ARKANSAS 1 99 92 (92 1
7 9
8 7
- 87 1 ARKANSAS 2 83 76 l91 1
7 9
7 6
- 85 i
BEAVER VALLEY 1 90 84 1 93 6
9 8
7 37 BEAVER VAtLEY 2 39 33 1 84 1
6 6
5 4
1 79 BIG ROCK POINT 1 67 57 I 85 )
10 8
7 6
1 85 BRAIDWOOD 1 39 30 1 76 I 9
6 5
4 i 79 i
BRAIDWOOD 2 36 27 1 75 )
9 5
4 3
1 75 l
BRDWNS FERRY 1 81 59 1 72 )
22 10 9
5 (55 t
BROWNS FERRY 2 76 63 I 82 t
13 9
9 8
(88 i
BROWNS FERRY 3 78 56 1 71 1
22 9
3 6
(75
'l BRUNSWICK 1 78 67 l 15 11 8
8 6
(75 I
,Q BRUNSWICK 2 78 67 l 85 l
11 9
8 7
I87 I
i vJ BYRON 1 39 31 i 79 1 8
5 4
3 1 75 l
8 5
4 3
1 75 BYRON 2 37
.29 78
'CALLAWAY 1 41 30 73 11 5
4 3
1 75
.I 4
l CALVERT CLIFFS 1 89 82
- 92 7
9 8
7 87 CALVERT CLIFFS 2 84 77 1 91 1
7 9
8 7
87 CATAWBA 1 40 33 1 82 1
7-5 4
2
- 50 CATAWBA 2 39 32 (82 1
7 5
4 2
1 50 CLINTON 1 36 23 (63 1 13 6
5 4
79 i
COMANCHE PEAK 1 74 67 (90 1 7
9 5
4
- 79 l
COOK 1 87 75 1 36 i
12 8
7 6
' 85 I
10 8
7 6
I 85 COOK 2 86 76 I 88 87 COOPER STATION 77 64 I 83 i
13 9
8 7
CRYSTAL RIVER 3 38 79 1 89 l
9 8
8 7
l 87 DAVIS-BESSE 1 85 77 i,90 1
8 9
7 5
1 71 DIABLO CANYON 1 48 40 i 83 I
8 6
6 5
1 83 l
DIABLO CANYON 2 13 34 1 79 i
9 5
5 4
1 79 I
DRESDEN 2 79 68 1 86 l
11 10 10 9
1 89 DRESDEN 3 78 68 1 87 l
10 9
9 8
' 88 DUANE ARNOLD 82 73 i 39 l
9 10 9
8 i 88 FARLEY 1 86 79
' 91 1
7 8
8 7
l 87 i
FARLEY 2 52 45 E86 1
7 7
7 6
1 35 I
FERMI 2 39 31
- 79 1
3 6
5 4
1 79 I
FITZPATRICK 79 70
' 88 1
9 10 9
8 1 88 i 94 1
6 10 8
7 i 37 t
GINNA
~101 95 FORT CALHOUN 1 89 75
' 84 1
14 9
9 8
! 88 I
GRAND GULF 1 39 31 1 79 1
8 5
4 4
1 100 I
HADDAM NECK 88 81 i 92 l'
7 8
8 7
l 87 t
HARRIS 1 38 29
' 76 I
9 5
4 2
1 50 I
84 1
12 9
8 7
1 87 i
HATCH 1 78 66 i
83 12 10 9
8 I 88 HATCH 2 73 61 HOPE CREEK 1 37 30 (81 1
7 5
4 3
1 75 'l INDIAN POINT 2 90 83 (92 )
7 9
8 7
- 87 Table 5.5
SAFETY' ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S) - SLHMARY BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS ITEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUI;tED COMPLETED COMPLETED INDIAN POINT 3 37 80 (91 )
7 8
7 6
185 KEWAUNEE 90 84 (93 i
6 9
8 6
(7E LASALLE 1 36 25 l 69 11 5
4 3
(75 LASALLE 2 36 25 69 11 5
4 3
(75 80 7
5 4
3 (75 i
LIMERICK 1 36 29 7
5 4
3 (75 LIMERICK 2 34 27 1 79 i
MAINE YANKEE 93 85 1 91 1
8 10 9
8 (88 8
5 4
3 (75 t
MCGUIRE 1 49 41 1 83 MCGUIRE 2 43 36 (83 l
7 5
4 3
(75 t
MILLSTONE 1 73 63 1 86 l
10 9
8 7
(87 i
MILLSTONE 2 90 80 i 88 )
10 9
8 7
(87 t
y MILLSTONE 3 37 31 1 83 l
6 5
5 4
(79 i
N MONTICELLO 78 69 88 1
9 9
8 7
(87 t
NINE MILE POINT 1 81 69
' 85 12 10 10 9
i89 NINE MILE POINT 2 33 24 72 1
9 5
4 3
i 75 i
' NORTH ANNA 1 69 61 1 88 l
8 8
8 7
l 87 OCONEE 1 '
52 44 1 84 )
8 6
5 4
i 79 NORTH ANNA 2 90 81 89 l
9 9
8 7
1 87 OCONEE 2 90 81 1 89 l
9 9
8 7
f87 OCONEE 3
-89 80 i 89 i
9 9
8 7
g87 i
OYSTER CREEK 1 75 63
! 83 i
12 10 10 9
(89 PALISADES 92 83
' 90 1
9 10 8
6 (75 i
PALO VERDE 1 37 26
, 70 1
11 5
5 4
(79 I
PALO VERDE 2 35 24 68 l
11 5
5 4
(79 i
PALO VERDE 3 36 25 1 69 11 5
5 4
(79 t
PEACH BOTTOM 2 77 67 1 87 10 10 9
8 (88 I
86 10 10 F
S 188 PEACH BOTTOM 3 76 66 i
PERRY 1 34 25 i 73 9
5 4
3 1 75 PILGRIM 1
-82 74
' 90 8
9 8
7 I 87 r
POINT BEACH 1 88 80 (90 )
8 9
9 7
1 77 POINT BEACH 2 88 80 (90 )
8 8
8 6
1 75 PRAIRIE ISLAND 1 90 79 187 )
11 8
7 6
1 85
' PRAIRIE ISLAND 2 90 81 1 89 )
9 8
7 6
I 85
.t 11 9
8 7
(87 QUAD CITIES 1 81 70' I 86
QUAD CITIES 2 80 69 i 86 11 9
8 7
i87.'i RIVER BEND 1 33 27
' 81 6
5 4
3 1 75 I
ROBINSON 2 38 79
' 91 7
9 8
7 1 87 i
SALEM 1 88 76 86 12 8
8 7
1 87 i
SALEM 2 -
53 41 1 77 l
12 7
7 6
1 85 SAN ONOFRE 1 85 83
! 97 )
2 10 9
6 (66 l
78 8
5 5
4 179 l
SAN ONOFRE 2 38 30
' SAN ONOFRE 3 39 30 1 76 l-9 5
5 4
1 79 i-SEABROOK 1 33 27 1 81 1 6
5 5
4 1 79 l
SEQUOYAH 1 48 41 1 85 1 7
5 5
4 1 79 i
Table 5.5
. SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S) - SUM ARY BY PLANT IMPLEMENTATION VERIFICATION ITEMS ITEMS PER CENT ITEMS ITEMS
?TEMS ITEMS PER CENT UNIT APPLICABLE COMPLETED COMPLETED REMAINING COVERED REQUIRED COMPLETED COMPLETED SEQUOYAH 2 40 33 (82 )
7 5
5 4
(79 i
SOUTH TEXAS 1 36 28 177 )
8 5
3 2
! ss SOUTH TEXAS 2 33 26 78 )
7 5
3 2
1 66 92 7
9 9
7
' 77 ST LUCIE 1 91 84 ST LUCIE 2 37 30 i 81 7
5 5
3
. 59 SUMMER 1 41 37 l 90 4
4 4
2 l 50 SURRY 1 93 86 1 92 l
7 10 9
7 i 77 92 7
10 9
7 i 77 i
$URRY 2 96 89 l
34 25 i 73 )
9 5
4 3
. 75 1
SUSQUEHANNA 1
^
34 25 1 73 )
9 5
4 3
1 75 t
SUSQUEHANNA 2 THREE MILE ISLAND 1 97 90 1 92 )
7 9
8 7
1 87 1
' TROJAN 88 81 L92 l
7 9
9 6
1 66 i
h TURKEY POINT 3 95 87 (91
.f 8
9 7
6 I 85 (o
TURKEY POINT 4 96 88 (91 8
9 7
6 (85 VERMONT YANKEE 1 78 68 (87 l
10 10 8
7 (87 V0GTLE 1 39 30 176 )
9 5
5 3
159 VDGTLE 2 31 22 1 70 )
9 5
5 3
1 59 WASHINGTON NUCLEAR 2 40 32 1 79 )
8 5
5 4
1 79 l
WATERFORD 3 36 29 i 80 1
7 5
3 2
1 66 i
7 5
3 2
1 66 t
WOLF CREEK 1 38 31 1 81 ZION 1 94 84 1 89 10 8
6 5
1 83 t
ZION 2 93 88 1 92 )
7 3
6 5
1 83 TOTALS / AVERAGES 7175 6197 84 978 812 720 587 80 Table 5.5
I 4
5.4 Status by Issue 1
l 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 irnplementation, and the r, umber 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 l
number and percentage of plants that have completed verification.
i Of the current 166 MPA issues,119 have been fully implemented,~ twenty six issues l
remain to be implemented at less than 8 plants and 13 issues remain to be implemented at 9 to 23 plants. The remaining oight MPA issues are to be j
implemented at 29 or more plants.
F h
(
l 1
1_
e 99
_.~.,. _-_
_----_..m__~
.m m...__. -......_ _ _
._-.__~
L 1
S A F-E T Y ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S). StM4ARY BY ITEM IMPLEMENTATION VERIFICATION l
ITEM APPLICABLE COMPLETED COMPLETED REMAINING' REQUIRED COVERED REQUIRED COMPLETED COMPLETED 47.
.2 2
(1001 0
NO LOSS OF OFF. SITE POWER 75 IB089) 69 69 (1001 0
NO
' SALEM ATWS 4.2.3 & 4.2.4 LIFE COMPONENTS 8059 (E004) 15 15 (2001 0
No i
'BWR SINGLE LOOP OPERATION B059 (E005) 9 9
(100) 0 No W N-1 LOOP OPERATION E
BL-79-06 60 60 (100) 0 No Q
REVIEW OF OPERATIONAL.EPRORS AND SYSTEM MI$ ALIGNMENTS. IDENTIFIED DURIN
~
BL -7 9-06 A
'5 5
(100) 0 No REVIEW OF OPERATIONAL ERRORS AND SYSTEM MISALIGNMENTS IDENTIFIED DURIN t
BL 79-08
'5 5'
(100) 0 NO EVENTS RELEVANT TO BOILING WATER REACTORS-IDENTIFIED DURING THREE MILE
~BL-79-13 50-50 (1001 0
NO CRACKING IN FEEDWATER SYSTEM PIPING BL 79-15 110 110 (100) 0 YES 110 104 104 (100)
. DEEP DRAFT PUMP DEFICIENCIES S b NON-CLASS-1-E INS UMENTATION A CONTROL S BUS DURING OP BL-80-04 46 46 (100) 0 NO ANALYSIS OF A PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDIT i
BL.80-06
- 63 63 (1001 0
No ENGINEERED SAFETY FEATURE (ESF) RESET CONTROLS BL-80-07 33 33 (1001 0
No BWR JET PUMP ASSEMBLY FAILURE
'BL-80-11
'85 65 (1001 0.
YES 95 85 (96 )
' MASONRY WALL MSIGN BL-80-18 45 45 (1001 0
NO MAINTENANCE OF ADEQUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING PUMPS F Table 5.6
4'
+
$AFETY ISSUE M A'M A G E M E N 1 SYSTEM STATUS OF OTHER MPA(S) - SUPMARY 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-87-01 110 110 (100%
0 NO THINNING OF PIPE WALLS IN NUCLEAR POWER PLANTS BL-88-01 110 110 (1001 0
NO DEFECTS IN WESTINGHOUSE CIRCUIT DREAMERS BL-88-02 35 29 (82 )
6 N0 STEAM GEERATOR TUBE RUPTURE (BULLETIN 88-02) (OLD MPA 80991 6L-88-03
'110 108 (98 )
2 NO GE HFA RELAYS (BULLETIN 88-03) 1 BL-88-04 107 98 (91-)
9 YES 107 37 37 (2001 O
SI PUMP FAILURE (BULLETIN 88-04) (CLD MPA B103)
BL-88-05
'110 110-(2001 0
NO NONFORMING MATERIALS SUPPLIED BY PIPING SUPPLIES. INC. AT FOLSOM BL-88-07 37 35 (94 1 2
YES 37 37 36 (97 I-POWR OSCILLATIONS IN BOILING WATER REACTORS (BWRS)
BL-88-08 110 91 (82 1 19 NO THERMAL STRESSES IN PIPING CONNECTED TO REACTOR COOLANT SYSTEMS BL-88-09 50 50-(1001 0
NO THIMBLE TUBE THINNING IN WESTINGHOUSE REAXCTORS BL-88 110 103 (93 )
7 NO NONCOCORMING' MOLDED-CASE CIRCUIT BREAKERS BL-88-11 73 44-(60 1 29-NO PRESSURIZER SURGE LINE THERMAL STRATIFICATION BL-89-01 74-
~73 (98 1 1
NO FAILURE OF WESTINGHOUSE STEAM GENERATOR TUBE MECHANICAL PLUGS-BL-89-02 110 104 (94 )
6.
NC STRESS CORROSION CRACKING OF HIGH-HARDNESS TYPE 410 STAINLESS STEEL BL-89-03 73 73 (2001 0
NO POTENTIAL LOSS OF REQUIRED SHUTDOWN MARGIN DURING REFUELING OPERATIONS BL-90-01 110
-20 (18 )
90 NO LOSS OF FILL-0IL IN TRANSMITTERS MANUFACTURED BY ROSEMOUNT Table 5.6 4
s m.-
m-_-..
-m l
SAFETY ISSUE MANAGEMENT' SYSTEM STATUS OF OTHER MPA(S) -
SUMMARY
BY ITEM IMPLEMENTATION VERIFICATION l
PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT
......' ITEM
. APPLICABLE, COMPLETED COMPLETED REMAINING.
REQUIRED. COVERED - REQUIRED CGMPLETED COMPLETED 1
t BL-90-02 37 37 (100) 0 NO LOSS OF THERMAL MARGIN CAUSED BY CHANNEL BOX BOW' BL-92-01 110
'24 (21 )
86 NS i
FAILURE OF THERMO-LAG 330 FIRE BARRIER SYSTEM GL-79-25 19 19 (100) 0-MO INFORMATION REQUIRED TO REVIEW CORPORATE CAPABILITIES GL-79-32 61
'61 (100) 0 NO TMI-2 LESSONS LEARNED TASK FORCE REPORT - NUREG-0578 ea
- GL-79-33 63 63 (100)
.O NO C].
- TRANSMITTING NUREG-0567-SECURITY TRAINING AND QUALIFICATIONS PLAN GL-79-36 65 65 (1001 0
NO ADEQUACY'CF STATION ELECTRIC DISTRIBUTION SYSTEMS c
- GL-79-43'
'I7' 17 (100) 0 NO
= REACTORS CAVITV' SEAL-RING GENERIC ISSUE (PWR) i
- GL-79-46 65 65
-(100) 0.
NO l
CONTAINMENT PURGING AND VENTING DURING NORMAL OPERATION - GUIDELINES GL-79-58 43 43 (100]
0 s NO ECCS CALCULATIONS ON FUEL CLADDING GL-80-002 41 41' (1001 0
YES 41 38 37 (97 )
CUALITY ASSURANCE REQUIREMENTS REGARDING DIESEL GENERATOR FUEL OIL
- GL-80-024 68 68 (100) 0.
NO 4 '
'NRC NUCLEAR DATA LINM (NDL)
GL-80-030 63 63 (1001 0
NO CLARIFICATION OF THE TERM "0PERABLE" AS IT APPLIES TO SINGLE FAILURE GL-80-061 20 20 (1001 0
NO TMI-2 LESSONS LEARNED GL-81-01~
54?
54 (100)
^0 NO QUALIFICATION OF INSPECTION,. EXAMINATIONS, AND TESTING AND AUDIT PERSO GL-81-04 66 66-(1001 0
NO EMERGENCY PROCEDURE AMD TRAINING FOR STATION BLACKOUT EVENTS Table 5.6 -
4 k
m m_
i t
SAFETY ISSUE MANAGEMENT ' SYSTEM I
STATUS OF OTHER MPA(S) - SUMMERY BY ITEM IMPLEMENTATICN :
VERIF CATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED GL-81-14 :
<44 41 (93 )
3 NO SEISMIC QUALIFICATION OF AUXILIARY FEEDWATER SYSTEM GL-81-21 73
-73 (1001 0
YES 73 69 62 (89 )
NATURAL CIRCULATION C00LDOWN GL-83 23 22 (95 1 1
YES 23 25 21 (91 1 MODIFICATION OF VACUUM BREAKERS ON MARK I CONTAINMENTS GL-83-43 80 80 (100)
'O NO 1
REPORTING REQUIREMENTS CF-10 CFR PART 50, SECTIONS 50.72 AND 50.73. AN GL-84-09 21 14
)
7 NO RECOMBINER CAPABILITY REQUIREMENTS OF 10 CFR 50.44({86 a
)(3)(II)
.g C
GL-84-15 86 86-(1001 0
NO PROPOSED STAFF ACTIONS T0-IMPROVE AND MAINTAIN DIESEL GENERATOR RELIAS GL-87-05 123 23 (100) 0 NO REQUEST FOR ADDITIONAL INFORMATIDN-ASSESSMENT OF LICENSEE MEASURES TO GL-87-09 65 -
51.
(78 )
14 NO SECTIONS 3.0 AND 4'0 0F THE STANDARD TECHNICAL SPECIFICATIDMS (STS) ON GL-87-12 71 71 (2001 0
NO LOSS. 0F RESIDUAL. HEAT REMOVAL (RNR) WHILE IN THE REACTOR COOLANT SYSTE GL-88-01 37 22 f59 )
15 NO NRC POSITION ON IGSCC IN BWR AUSTENITIC STAINLESS STEEL PIPING GL-88-12 52 31-(59 )
21
.NO t
REMOVAL OF FIRE PROTECTION REQUIREMENTS FROM. TECHNICAL SPECIFICATIONS ii GL-88-05
'73 73
-(1001 0
N0' i
BORIC ACID CORROSION OF CARBON STEEL REACTOR PRESSURE BOUNDARY COMPONE GL-88-06 103 101 J98 1 2
' NO i
REMOVAL OF ORGANIZATIONAL CHARTS FROM TECHNICAL SPEwIFIMTION ADMINIST GL-88-11
'110 103 (93 )
7' NC NRC POSITION ON RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS AN GL-88-16 82 75 (91 )
7
'NO REMOVAL OF CYCLE-SPECIFIC PARAMETER LIMITS FROM TECHNICAL SPECIFICAT-Table 5.6
.. h!
l 1
t SAFETY ISSUE MANAGEMENT SY$ TEM I
I STATUS OF OTHER MPAts) - SUmARY BY ITEM r
IMPLEPENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED. COMPLETED COMPLETED i
'f GL-86-20 110 13 (2 )
107 NG
. INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT VULNER. 10CFR50.54(F)
[
GL-89-01 59 38 (64 1 21 NO IMPL OF PROG CONTROLS FOR RAD EFFLUENT TECH S*ECS IN ADM CCNT SECTION GL-89-04 43 26 (60 1 17 NO GUIDANCE ON DEVELOPING ACCEPTABLE INSERVICE TESTING PROGRAMS GL-89-06 98 88 (92 1 7
'NO
- i TASK ACTION PLAN ITEM I.D 2 - SAFETY PARAMETER DISPLAY SYSTEM j
2 GL-89-07 110 110 (1001 0
YES 110 J10 110 (100)
POWER REACTOR SAFEGUARDS COTINGENCY PLANNING FOR SURFACE VEHICLE BOMBS GL-89-08 110 108 (98 l 2
NO EROSION / CORROSION-IleUCED PIPE WALL THINNING GL-89-10 110 0
(0 1 110 YE!
110 110 1
(0 )
SAFETY-RELATED MOTOR-OPERATED VALVE TESTING AND SUPVEILLANCE 75 72 (95 )
3 NO GL-89-14
.. IN TECH SPECS - REMOVAL OF 3.25 LIMIT ON EXT SURV LINE-ITEM IMPROV.
GL-89-16 24 1
(4 )
23' NO INSTALLATION OF.A HARDENED WETWELL VENT (GL 89-161 GL-90-02:
10 7
(69 )
3 NO ALTERNATE REQ FOR FUEL ASSEMBLIES IN THE DESIGN FEAT SECT OF TECH SPEC GL-90-09 64 59 (92 1 5
NO ALT REO FOR SNUBBER VISUAL I'1SPECTION INERVALS & CORRECTIVE ACTIONS GL-91-01 22-18
($1 1 4
NO REMOVAL OF THE SCHEDULE FOR THE WITHDRAWAL OF REACT VESSEL MAT SPEC.
GL-91-04 7
4 (57 )
3 NO
' CHANGES IN TECH SPEC SURV. INTERils TO ACCOMODATE A 24-M0 FUEL CYCLE GL-91-06' 110 87 (79 )
23 NO ADEQUACY OF SAFETY-RELATED DC POWER SUPPLIES (GL 91-06) (GI A-30)
GL-91-08 22
-10
[45 i 12 NO REMOVAL OF COMPONENT LISTS FROM TECHNICAL SPECIFICATIONS i
Table 5.6 1
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPAIS) -
SUMMARY
BY ITEM VERIFICATION IMPLEMENTATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMA*NING REDUIRED COVERED REQUIRED COMPLETLD COMPLETED
.....~...
MOD OF SURV INTERVAL FOR ELEC PROTO ASSEM IN POWER SUPPL GL.91-11 110 94 (85 1 16 NO VITAL INSTRUMENT BUSES & TIE BREAKERS (GI 48/49.GL 91-11)
GL-91-13 13 10 (76 1 3
NO ESSENTIAL SERVICE WATER SYSTEM FAILURES 1GL-91-13) (GI-130)
GL-92-01 110 6
(5 )
104 NO REACTOR VESSEL STRUCTURAL INTEGRITY ea C3 GL-92-04 37 0
(0 )
37 NO gn REACTOR VESSEL WATER LEVEL INSTRUMENTATION IN BWRS MPA-A024 97 97 (100) 0 NO MISCELLANEOUS AMEN 0MENTS AND SEARCH REDUIREMENTS MPA-A001 65 64 (98 1 1
NO 10 CFR 50.55 A(0) - ISI MPA-A002 65 65 (100) 0 NO APPENDIX I ALARA MPA-A003 61 61 (1001 0
NO SECURITY REVIEWS-MODIFIED AMENDMENT PLANS MPA-A004 49 47 (95 1 2
NO APPENDIX J - CONTAINMENT LEAM TESTING MPA-A005 19 19 (100) 0 NO GE MARM I CONTAINMENT TECH SPECS-SHORT TERri MPA-A006 11 11 (100) 0 NO RESPIRATORY PROTECTION SYSTEM MPA-A007 10 10 (1001 0
NO APPENDIX 0 - FRACTURE TOUGHNESS MPA-A008 9
9 (100)
O NO ECCS EVALUATION-GENERIC PER 50.46 COMPLIANCE HPA-A009 62 62 (100) 0 NO PRESSURE VESSEL BELTLINE MATERIAL SURVEILLANCE Table 5.6
........ _ - -.. ~.
-. -. -... - - - ~ ~. _.... ~... -... -. ~......
- ~. - -
Y f
i
?
S A'F'E T Y ISSUE MANAGEMENT SYSTEM i
t STATUS OF-0THER MPA(S) - SUPNARY BY ITEM r
IMPLEMENTATION VERIFICATION
' PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS.
PLANTS PER CENT
'i
. ITEM.
APPLICABLE.
COMPLE TED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED COMPLETED 6
' CONTINGENCY PLANNING MPA-A010 63 63 (100) 0 NO 3
k MPA-A012 61 61 (100) 0 NO VITAL AREA ANALYSIS MPA-A014 51 51 (100) 0 NO 10CFR.50.55 A(G).' INSERVICE TESTING
-MPA-Blis 37 26 (70 1 11 No RESULTS OF N'C TESTING OF MOVS (OL 89 10, SUPP3)
.,1 i
R h
MPA-B117-
. ' 73 57 (78 -)
16 NO FAILURE OF WESTINGHOUSE SG TUBE MECHANICAL PLUGS (BL 90-01, SUPP2) s l'
MPA-B119 109.
-0 (0 l 109 NO IPE EXTERNAL EVENTS (OL'88-20, SUPP 4)
MPA-B001' 26 26 (100) 0 NO DIESEL GENERATOR LOCKOUT
?
MPA-8002'.
.56 56 (1001 0
NO FIRE PROTECTION
'PWR MODERATOR DILUTION. 39 39 (100) 0 YES 39 36 34 (94 )
MPA-8003 MPA-8006 22 22
-(100) 0 NO.
BWR RELIEF VALVE l
' 25 '-
'25 (1001 0
NO MPA-B007 STEAM GENERATOR FEEDWATER FLOW INSTABILITY' MPA-8008 14 14 (1001 0
NO PWR HPSI-LPSI FLOW RESISTANCE
'MPA-8009 17 17 (100]
O NO CHARGING SYSTEMS PIPE VIBRATIONS MPA-8010'.
3..
'3'
.(1001 0
NO BURNABLE POISON ROD FAILURE'- B&W
- 10 10' (1001 0
YES-10 4
1 (25')
MPA-8011 FLOOD OF EQUIPMENT IMPORTANT TO SAFETY-Table 5.6
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l t
i s
l SAFETY ISSUE
-MANAOEMENT SY$ TEM 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 COMPLETED MPA-8035 5
5 (1001 0
NO ORIFICE ROD ASSEMBLY INTEGRITY - B&W MPA-B036' 7-7 (1001 0
NO
-RESISTANCE TEMPERATURE DETECTOR (RTD) RESPONSE - CE MPA.B037 10 10
-(200) 0 NO STEAM GENERATOR. TUBE DENTING AND SUPPORT PLATE MODIFICATIONS - CE MPA-8038~
2 2
(1001 0
NO TENDON SURVEILLANCE'- BECHTEL CONTAINMENTS' C3 MPA-0039-18 38 (1001 0
NO 9
PWR PRESSURE TEMPERATURE LIMIT TECH SPECS MPA-8040 1
1 (100) 0 NO PIPE SUPPORT BASE PLATES i
MPA-8041 66 64 (96 )
2 YES 68 61 62 (96 )
FIRE PROTECTION - FINAL TECH SPECS (INCLUDES SER SUPPLEMENTS) l F
MPA-B046L 54 54 (100) 0 NO.
i ANALYSIS OF TURBINE DISC CRACKS MPA-6049
'10 10 (100)'
O NO PWR CONTROL ROD MISALIGNMENT
.[
.MPA-8052.
28 28 (100)
.O NO
' REVIEW OF SAFETY ASPECT OF INADVERTENT SAFETY ACTIONS DURING SUR. TEST j
l MPA-8055' 5
5 (200) 0 NO B&O REPORT ON BWRS MPA-8356 8
8 (1001 0
NO
- CONTROL RODS F AILURE TO INSERT. BWR
.i
- MPA-8037 31 -
31 (1001 0
NO DHR CAPA8ILITY MPA-8064 7
7 (100)
O.
NO ACC INDUCED FLUX ERRORS (B&W)
MPA-8067
,8 8
(100) 0 NO.
THERMAL SHOCK Table ' 5.6 5
4 s
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. MF MP MT MP MP MB MQ MF MC MF MB MP MR MB ME e '" h
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4 i
SAFETY ISSUE MANAGEMENT SYSTEM ST TUS OF OTHER MPAIS) - StM4ARY BY ITEM IMPLEMENTATION VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED C0hPLETED REMAINING REQUIRED COVERED REQUIRED COMPLET!D COMPLETED
~
f!PA-0003
. 15 15 (1001 0
No l
PRESSURIZER HEATUP RATE ERROR MPA-D005 6
6 (100) 0 NO
. PLANT UPI MODEL' PROBLEM MPA-D006 8
8.
(1001 0
NO PEAKING MODEL' CHANGE FOR CE REACTOR CORE MPA-D007' 3
3 (100) 0 NO
[
_A A
MPA-0008 3
3 (2001 0
NO 9
DEFICIENCY.IN CHEM ADDITION TO CONTAINMENT SPRAYS MPA-D009 1
1 (1001 0
NO GE ECCS INPUT ERRORS MPA-Doll el 61 (1001 0
NO FISSION GAS RELEASE-MPA-D013.
8 8
(100) 0 NO.
B&W SMALL BREAK ERROR r
MPA-0014 10 10 (100) 0 NO REACTOR VESSEL WELD - WIRE DEFICIENCY MPA-Dol $
83 63;
-(100) 0 NO
. HIGH ENERGY LINE BREAK & CONSEQUENTIAL SYSTEM FAILURE MPA-D018 7'
.7.
(100) 0 NO NUREG 0630 CLADDING MODELS (8&W PLANTS)
MPA-E001 31 31
.(1001 0'
NO
[
SPENT FUEL POOL EXPANSIONS e
L MPA-E002 7
7 (100) 0-NO FUEL CASK DROP MPA-E003 -
30 30 (1001 0
. NO CORE RELOADS REQUIRING PRIOR NRC APPRD'.?AL
'MPA-E006 7
7 (1001 0
NO CEA POSITION INDICATION FAILURES - CE P
Table 5.6
SAFETY ISSUE MANAGEMENT SYSTEM STATUS OF OTHER MPA(S) - SUP94ARY BY ITEM IMPLEMENTATIDM VERIFICATION PLANTS PLANTS PER CENT PLANTS PLANTS PLANTS PLANTS PER CENT ITEM APPLICABLE COMPLETED COMPLETED REMAINING REQUIRED COVERED REQUIRED COMPLETED C NPLETED MPA-E007 5
5 (100) 0 NO REACTOR PROTECTION SYSTEM LOGIC - CE
. s
_s a.
Table 5.6
5.5 Conclusions After a detailed review of the implementation and verification status of the resolution of the 166 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 P;er implementation actions and schedules proposed by licensees to ensure that they are completed in accordance with regulatory requirements.
l l
l i
-113-
Appendix A
-LISTING OF UNIMPLEMENTED TMI ITEMS BYISSUE l-1
APPENDIX A This appendix provides a detailed list, by issue, of the 86 TMI Action Plan items not implemented, along with the projected target date for completing the item. Status and projected implementation dates are presented as of September 30,1992.
A-1
TMI Issues (Listing of Open Items)
IMPL DATE ISSUE MPA PLANT TAC TITLE 11/92-1.
GL-83-36 8083 ZION 1 M54580 NUREG-0737 TECHNICAL SPECIFICATIONS 2.
GL-83-36 B083 ZION 2 M54581 NUREG-0737 TECHNICAL SPECIFICATIONS 11/92 3.
I.D.2.2 F075 BROWNS FERRY 1 M74602 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - INSTALLED 07/96 4.
I.D.2.2 F075 BROWNS FERRY 2 M74607 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - INSTALLED 04/93 5.
I.D.2.2 F075 BROWNS FERRY 3 M74612 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - INSTALLED 01/94 6.
I.D.2.3
.F009 BEAVER VALLEY 1 M51221 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 04/93 7.
I.D.2.3 F009-BROWNS FERRY 1.
M51223 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 07/96 8.
I.D.2.3 F009 BROWNS FERRY 2 M51224 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 06/93 9.
I.D.2.3 F009 BROWNS FERRY 3 M51225 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 01/94 10.
I.D.2.3 F009 PALO VERDE 1
-M56654 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IIFLEMENTED 11/93
.11.
I.D.2.3 F009 PALO VERDE 2 M62796 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 04/93 12.
J.D.2.3 F009 PALD VERDE 3 M64581 PLANT-SAFETY PARAMETER DISPLAY CONSOLE - FULLY IMPLEMENTED 11/93 13.
II.B.3.2 F076 BROWNS FERRY 3 M74613 POSTACCIDENT SAMPLING - CORRECTIVE ACTIONS 01/94 14.
11.8.3.3 F077 BROWNS FERRY 1 M74603 POSTACCIDENT SAMPLING - PROCEDURES 07/96 15.
11.8.3.3 F077 BROWNS FERRY 3 M74614 POSTACCIDENT SAMPLING - PROCEDURES 01/94 16.
II.B.3.4
'F012 BROWNS FERRY 1 M44423 POSTACCIDENT SAMPLING - PLANT MODIFICATIONS (LL CAT B) 07/96
.p 17.
II.B.3.4-F012. BROWNS FERRY 3 M44425 POSTACCIDENT SAMPLING - PLANT MODIFICAT10h5 (LL CAT B) 01/94 di 18.
II.D.1.2.B F014 CRYSTAL RIVER 3 M44571 RELIEF & SAFETY VALVE TEST REQUIREMENTS - PLANT SPECIFIC REPORT 10/92 19.
II.D.1.3 F084 FOR: CALHOUN 1 M75832 RELIEF & SAFETY VALVE TEST REQUIREMENTS - BLOCK-VALVE TESTING 05/94 20.
II.E.4.2.1-4 F078 BROWNS FERRY 1 M74604 CONTAIMENT ISOLATION DEPENDABILITY - IMP. DIVERSE ISOLATION 07/% '
21.
II.E.4.2.1-4
- F078 BROWNS FERRY 3-M74615' CONTA! MENT ISOLATION DEPENDABILITY - IMP. DIVERSE ISOLATION 01/94 22.
II.E.4.2.6 F079 SROWNS FERRY 3 M74616 CONTAINMENT ISOLATION DEPENDABILITY - CNTMT PURGE VALVES 01/94 23.
II.E.4.2.8 F080 NINE MILE POINT 1 M76986 CONTAINMENT ISOLATION DEPENDABILITY - TECH SPECS 10/92 07/96 I
24.
II.F.1.1
.F081 BROWNS FERRY 1 M74605 '. ACCIDENT-MONITORING - PROCEDURES 25.
II.F.1.1 F081. BROWNS FERRY 3 M74617 ACCIDENT-MONITORING - PROCEDUPES 01/94 26.
II.F.1.2.A F020 LBROWNS FERRY 1
'M44903-ACCIDENT-MONITORING - NOBLE GAS MONITOR 07/96 01/94 27.
II.F.1.2.A F020 BROWNS FERRY 3 M44905'.. ACCIDENT-MONITORING - NOBLE GAS MONITOR 28.
II.F.1.2.8 F021 BROWNS FERRY 1
'M44974 ACCIDENT-MONITORING - IODINE / PARTICULATE SAMPLING 07/96 29.
II.F.1.2.8 F021-' BROWNS FERRY 3' M44976' ACCIDENT-MONITORING - IODINE / PARTICULATE SAMPLING 01/94 30.
II.F.1.2.C
'F022 BROWNS FERRY 1 M45045 ACCIDENT-MONITORING - CONTAINMENT HIGN-RANGE MONITOR 07/96-31.
II.F.1.2.C F022 BROWNS FERRY 3 M45047 ACCIDENT-MONITORING - CONTAINMENT HIGH-RANGE MONITOR 01/94
- 32.
!!.F.1.2.D' F023. BROWNS FERRY 3 M475841 ACCIDENT-MONITORING - CONTAINMENT PRESSURE 01/94 33.
II.F.1.2.E F024 BROWNS FERRY 3 M47655 ACCIDENT-MONITORING - CONTAINMENT WATER LEVEL-01/96 34.
II.F.2.4 F026 BROWNS FERRY 1 M45116, INSTRMNTATN FOR DETECT. OF INADEO CORE CLNG INSTLL AD0'L INSTRMNTATM 07/96-35.
II.F.2.4 F026 : BROWNS FERRY 3-M45118 INSTPMNTATM FOR DETECT. OF INADEO CORE CLNG INSTLL AD0'L INSTRMNTATN01/94-
_. -- _. ~ ~ -
r J
36.
II.F.2.4 7326 DRESDEN 3 M45130 INSTRMNTATM FOR DETECT. OF IMADEQ CORE CLNG INSTLL AD0'L INSTRMNTATM 11/93 37.
'II.F.2.4 F026.cuAD CITIES 1
.M45164 INSTRMNTATM FOR DETECT. OF INADEQ CJRE CLNG INSTLL AD0'L INSTRMNTATA 11/92 38.
II.F.2.4 F026 QUAD CITIES 2 M45165 INSTRMNTATN FOR DETECT. OF INADEO CORE CLNG INSTLL AD0'L INSTRMNTATM 05/93 39.
II.K.3.13.8 F043 - BROWNS FERRY 1 M45532 B&O TASK TORCE - HPCI & RCIC INITIATION LEVELS MODIFICATION 07/96 40.
II.K.3.13.B F043 BROWNS FERRY 3 M45534 B&O TASK TORCE - NDCI & RCIC INITIATION LEVELS MODIFICATION 01/94 41.
II.K.3.18.C F048 BROWNS FERRY 1
'M45680 'B&O TASK FORCE - ADS ACTUATION MODIFICATIONS 07/96 42.
II.K.3.18.C F048 BROWNS FERRY 3 M45682 B&O TASK FORCE - ADS ACTUATION MODIFICATIONS 01/94 43.
II.K.3.27 F054 ' BROWNS FERRY 1 M45776 B&O TASK FORCE - Come0N REFERENCE LEVEL FOR BWRS 07/96 44.
I I.K.3.27 F054 BROWNS FERRY 3 M45778 B&O TASK FORCE - Copeq05 REFERENCE LEVEL FOR BWRS 01/94 45.
II.K.3.28 F055.'. BROWNS FERRY 1 M48260 B&O TASK FORCE - QUALIFICATION OF ADS ACCUMULATORS 07/96 46.
II.K.3.28 FOSS BROWNS FERRY 3 M48262 B&O TASK FORCE.- QUALIFICATION OF ADS ACCUMULATORS 01/94 47.
III.D.3.4.3
'F070 HADDAM NECK
' M46 '1 - CONTROL ROOM HABITABILITY - IMPLEMENT MODIFICATIONS 12/93 48.
MPA-F071' F071 ARKANSAS 1 M561ud 1.D.1.2 DETAILED CONTROL ROOM REV!EW (FOLLOWUP TO F-8) 12/92 49.
MPA-F071 F071 ARKANSAS 2-M56101 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FDLLOWUP TO F-8) 12/92 50.
MPA-F071 F071 BIG ROCK POINT' M56103 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/92 51.
MPA-F071 7071 BROWNS FERRY 1 M56104 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 07/96 2
52.
MPA-F071 F071 BROWNS FERRY 2 M56105 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 04/93 53.
MPA-F071 F071 'SROWNS FERRY 3 M56106 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP 70 F -8) 01/94 54.
MPA-F071 F071. ' BRUNSWICK 1' M5o107 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FDLLOWUP TO F-8) 12/92 55.
MPA-F071 F071 BRUNSWICK 2 M56108 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWP TO F-8) 06/93 56.
MPA-F071
-F071 CALVERT CLIFFS 1 M56110 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWP TO F-8)
J5/93 L
57.
MPA-F071 F071 CALVERT CLIFFS 2 M56111 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 05/93 58.
MPA-F071 F071 DIABLO CANYON 1 M56117 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FDLLOWUP TO F-8) 05/94 59.
MPA-F071 F071 DIABLO CANYON 2 M68040
'I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 10/94 60.
MPA-F071 F071 FERMI 2 M74599 1.D.1.2 DETAILED CONTROL' ROOM REVIEW (FOLLOWUP TO F-8) 10/92 61.
MPA-F071
.F071 HADDAM NECK M56128 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 03/96-62.~
MPA-F071 F071 HATCH 1 M56129 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/92 63.
MPA-F071 F071 HATCH 2 M56130 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/92 64.
MPA-F071
- F071 ' MILLSTONE 1 M56138 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWLp To F-8) 12/95 I
. 65.
MPA-F071 F071 MILLSTONE 2, M56139" I.0.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 11/92 66.
MPA-F071 F071 NINE MILE MINT 1 M56141 I.D.1.2 DETAILED CONTROL. ROOM REVIEW (FOLLOWUP TO F-8) 03/95 67.
MPA-F071-
'F071 WORTH ANNA 1 M56142 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/94 68.
.MPA-F071 F071 NORTH ANNA 2 M56143.
I.D.1.2 DETAILED CONTROL ROOM REVIEW (FDLLOWUP TO F-8) 12/94-
" 69.
MPA-F071-F071 OCONEE 1 M56144
'I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 05/93 70.'
MPA-F071
.. F071 OCONCE 2 M56145:
I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 10/93
. - = - -. -. - _ -. -
... ~....
~.
i i
r t
'f 71.
MPA-F071
.F071 OCONEE 3 M56146 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP 70 F-8) 11/92 72.
MPA-F071 F071 PILGRIM 1 M59329 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 03/93 73.
MPA-F071 F071 Po!NT BEACH 1 M56152 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/93 74.
MPA-F071 F071 POINT BEACH 2 M56153 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/93
' 75.
MPA-F071 F071 PRAIRIE ISLAND 1 M56154 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 04/94 76.
MPA-F071 F071 PRAIRIE ISLAND 2 M56155 I.D.1.2 DETAILED CONTROL ROOM REV!EU (FOLLOWUP TO F-8) 11/93
- 77. 'MPA-F071 F071 SALEM 1 M56160 -
I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 10/92
' 78.'
MPA-F071 F071 -SALEM 2 -
M56161 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 10/92.
I
. 79.
MPA-F071 F071-SAN ONOFRE 2 M56163
!.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 03/93 80.
MPA-F071 F071 SAN ONOFRE 3 M56164 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 10/93' 81.
MPA-F071
.F071 SEOUOTAM 1 M56165
.I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/93 82.
MPA-F071 F071 SEQUOTAN 2 M56166 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/93 83.
MPA-F071 F071 SURRY 1 M56170 1.D.1.2 DETAILED CONTROL ROOM REVIEW (FDLLOWUP TO F-8) 12/94 i
84.
MPA-F071 F071 SURRY 2 M56171 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 12/94 85.
MPA-F071 F071 ZION 1 M56182 I.D.1.2 DETAILED CONTROL ROOM REVIEW (FOLLOWUP TO F-8) 10/92 86.
MPA-F071 F071 ZION 2 M56183 I.D.1.2 DETAILED LONTROL ROOM REVIEW (FDLLOWUP TO F-8) 12/92 Oi h
1 t.
W 4
1
Appendix B LISTING OF UNIMPLEMENTED USl ITEMS BY ISSUE
APPENDIX B This appendix provides a detailed list, by issue, of the 227 USI items not implemented, along with the projected date for completing the item. Status and projected implementation dates are presented as of September 30,1992.
l i
l B-1 l
Unresolved Safety Issues (Listing of Open Items)
IMPL ISSUE MPA PLANT TAC TITLE DATE 1.
A-11 S007 PALISADES A77021 REACTOR YESSEL MATERIALS TOUGNNESS 10/92 2.
A-24 8060 BROWNS FERRY 1 M42481 QUALIFICATION OF CLASS IE SAFETY-RELATED EQUIPE NT 07/96 3.
A-24 8060 SROWNS FERRY 3 M42483 QUALIFICATION OF CLASS IE SAFETY-RELATED EQUIPMENT 01/94 4.
A-31 S004 HADDAM NECK M77025 RNR SHUTDOWN REQUIREMENTS 04/93 5.
A-44 A022 ARKANSAS 1 M68508 STATION BLACKOUT 12/94 6.
A-44 A022 ARKANSAS 2 M68509 STATION BLACKWT 12/94 7.
A-44 A022 BEAVER VALLEY 1 M68510 STATION BLACKOUT 06/93 8.
A-44 A022 BEAVER VALLEY 2 M68511 STATION BLACKOUT 06/93 9
A-44 A022 BIG ROCK POINT 1 M68514 STATION BLACKOUT 10/94 10.
A-44 A022 BROWNS FERRY 1 M68517 staff 0N BLACK 0UT 01/94 11.
A-44 A022 BROWNS FERRY 2 M68518 STATION BLACKOUT 04/93 12.
A-44 A022 BROWNS FERRT 3 M68519 STATION BLACKOUT 01/94 13.
A-44 A022 BtDNSWICK 1 M68520 STATION BLACKOUT 12/92 14.
A-44 A022 BRUNSW1CK 2 M68521 STATION BLACK 00T 12/92 15.
A-44 A022 CALLAWAY 1 M68524 STATION BLACK 0UT 12/92 16.
A-44 A022 CALVERT CLIFFS 1 M68525 STATION BLACK 0UT 05/96 17.
A-44 A022 CALVERT CLIFFS 2 M68526 STATION BLACKOUT 05/96 gy 18.
A-44 A022 CATAWBA 1 M68527 STATION BLACKOUT 06/94 19.
A-44 A022 CATAW9A 2 M68528 STAT 104 BLACKOUT 06/94 20.
A-44 A022 CLINTON 1 M68529 STATION BLACKOUT 06/93 21.
A-44 A022 COMLNCHE PEAK 1 M68530 STAT!0h BLACKOUT 07/94 22.
A-44 A022 COOK 1 M68532 STATION BLACKOUT 11/93 23.
A-44 A022 COOK 2 M68533 STAtl0N BLACKOUT 11/93 i
24 A-44 A022 COOPER STATIO4 M68534 STATION BLACKDJT 06/94 25.
A-44 A022 CRTSTAL RIVER 3 M68535 STATION BLACKOUT 06/94 26.
A-44 A022 DAVIS-BESSE 1 H68536 STATION BLACKOUT 05/93 27.
A-44 A022 DIABLO CANYON 1 M68537 STATION BLACKOUT 03/93 28.
A A022 DIABLO CANTON 2 M68538 STATION BLACK 0UT 03/93 29.
A-44 A022 DRESDEN 2 M68539 STATION BLACK 0UT 12/95 30.
A-44 A022 DRESDEN 3 M68540 STATION BLACKOUT 12/95 31.
A-44 A022 FARLEY 1 M68543 STATION BLACKOUT 04/94 32.
A-44, A022 FARLEY 2 M68544 STATION BLACK 0UT 10/93 33.
A-44 4022 FIT 2 PATRICK M68546 ' STATION BLACKOUT 06/94 34.
A-44 A022 FORT CALHWN 1 M68547 STATION BLACKOUT 11/93 35.
A-44 A022 GINNA M68548 STATION BLACKOUT 09/94
I 36.
A-44 A022 GRAND GULF 1 M68549 STATION BLACKauf 12/92 37.
A-44 A022. MADDAM NECK M68551 STATION BLACKOUT 03/93 38.
A-44 A022 HARRIS 1 M68552 STATION BLACKOUT 06/94 39.
A-44
.A022 MATCH 1 M68553 STAT 104 BLACK 0UT 05/93 40.
A-44 A022 MATCH 2 M68554 STATION BLACKOUT 05/93 41.
A-44 A022 HOPE CREEK 1 M68555 STATION BLACKOUT 06/94 42.
A-44 A022 INDIAN POINT 2 M68556 STATION BLACK 0UT IT/93 43.
A-44 A022 IN0!AN POINT 3 M68557 STATION SLACK 007 06/94 44.
A-44 A022 KEWAUNEE M68558 STATION BLACKOUT 05/93 45.
A-44 A022 LASALLE 1 M68559 STATION BLACKOUT 07/93 46.
A-44 A022 LASALLE 2 M68560 STAftau BLACKOUT 07/93 47.
A-44 A022 MCGUIRE 1 M68564 STATION BLACKOUT 06/94 48.
A-44 A022 MCCUIRE 2.
M68565 STATION BLACKOUT 06/94 49.
A-44 A022 MILLSTONE 1 M68566 STATION BLACKOUT 02/94 50.
A-44 A022.- MILLSTONE 2 M68567 STATION BLACKOUT 06/93 51.
A-44 A022 MILLSTONE 3 M68568 STAft04 8LACKOUT 06/93 m
52.
A-44 A022 MONTICELLO.
M68569 STAT!ON BLACKOUT 12/94 4
53.
A-44 A022 NINE MILE POINT 1-M68570 STATION BLACKOUT 11/92 54 A-44 A022 NINE MILE POINT 2 M68571 STATION SLACK 00T.
02/93 55.
A-44 A022 NORTH ANNA 1 M68572 STAT 10N BLACKOUT 12/94 56.
A-44 A022 NORTH ANNA 2 M68573 STATION BLACK 0UT 12/94 57.
A-44 A022 OCONEE 1 M68574 - STATION BLACKOUT 12/92 58.
A-44 A022 OCONEE 2 M68575 STATION BLACKOUT.
12/92 59.
A-44 A022 OCONEE 3 M68576 STATION SLACK 0UT 12/92 60.
A-44 A022 OYSTER CREEK 1 M68577 STATIDW BLACKOUT 03/93 61.
A-44, A022' PALISADES M68573 STATION BLACKOUT 02/93 1
62.
A-44 A022 PALO VERDE 1 M68579 STATION BLACKOUT 11/93 63.
A-44
'A022 -PALO VERDE 2 W8580 STATION BLACKOUT, 11/94 64.
A-44' A022 PALO VERDE 3 M68581 STATION BLACK 007 04/94 65.
A-44 A022. PEACH BOTTOM 2 M68582 STATION BLACKOUT 10/94 66.
A-44 A022 PEACH BOTTOM 3 M68583 STATION 8 TACK 001 10/94 67 A-44 A022 PERRY 1 M68584 STATION BLACKOUT 04/94 68.
A-44.
A022 PILGRIM 1 M68585. STATION BLACKOUT 06/93-69.
A-44 A022 Po1NT SEACH 1 M68586 STAfl0N BLACKDJT 10/92 70.
A-44 A022 POINT BEACH 2 M68587 STATION BLACKOUT 10/92
~
^ ^ ~
~
71.
A-44 A022 PRAIRIE ISLAND 1 M68588 STATION BLACKUUT OS/W 72.
A-44 A022 PRAIRIE IstAND 2 M68589 STATION 8 TACK 1UT 05/93 73.
A-44 A022 QUAD CITIES 1 M68590 STATION BLACKOUT 12/M 74.
A-44 A022 QUAD CITIES 2
'M68591 STATION BLACK 0UT 12/95 i
75.
A-44 A022 RIVER 8END 1 M68593 STATION BLACKOUT 03/94 76.
A-44 A022 RostNSON 2 M68595 STAft0N BLACKOUT W/93 77.
A-44 A022 SALEM 1 M68596 STATION BLACKOUT 01/94 78.
A-44 A022 SALEM 2 M68597 STATION BLACKOUT 01/94 1
79.
A-44 A022 SAN ONOFRE 2 M68599 STAT 10N BLACK 0UT 09/94 80.
A-44 A022 SAN ONOFRE 3 M68600 STATION BLACKOUT 09/94 81.
A-44 A022 SEA 8 ROOK 1 M68601 STATION BLACK 0UT 03/93 82.
A-44' A022 SEQUOYAH 1 M68603 STATION BLACKOUT 06/94 83.
A-44 A022 SEQUOTAN 2 M68604 STAfl0N BLACKOUT 06/94 84.
A-44 A022 ST LUCIE 1 Mo8608 STATION 8LACKOUT 09/93 85.
A-44 A022 ST LUCIE 2 M68609 STAil0N 8LACKO'T 09/93 J
86.
A-44 A022 SURRT 1 -
M68611 STATION BLACKOUT 01/96 87.
A-44 A022 SURRT 2 M68612 STATION BLACKOUT 05/96 gg 23.
A-44
.A022 SUSQUEHANNA 1 M68613 STATION BLACKOUT 06/94 89.
A-44 A022 SUSQUEHANNA 2
.M68614 STATION BLACK 0UT 06/94
. 90, A-44 A022 TROJAN-M68617 STATION BLACKOUT 06/93 91.
A-44 A022 YOGTLE 1 M68621 STATION SLACK 0UT 02/94 92.
A-44 A022 V0GTLE 2 M73447 STATION BLACKOUT 02/94 93.
A-44 A022 WASHINGTON NUCLEAR 2-M68626 STATION BLACKOUT 06/94 94 A-44 A022 WATERf0RD 3 M68623 STATION BLACKOUT
' 06/M 95.
A-44 A022 WOLF CREEK 1 M68628 STATION BLACK 0UT 12/92 A-44 A022 ZleM 1 M68630 STATIO1 BLACKOUT 04/93 97.
A-44 A022 ZION 2 M68631 STATION BLACKalT i2/92 98.
A-46 8105-ARKANSAS 1 M69426 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS fro /95 l
99.
A-46 8105 ARKANSAS 2 M69427 SEISMIC QUALIFICATION OF EQUIPMENT IN OPEltATING PLANTS 06/95 100.
A-46 8105 BEAVER VALLET 1 M69428 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS
/
101.
A-46 8105 BIG ROCK POINT 1 M69429 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATlWG PLANTS
/
102.
A-46 8105 BROWNS FERRY 1 M69430 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/96 103.
A-46 8105 BROWNS FERRY 2' M69431 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS
'07/96 104.
A 46 8105-BROWNS FERRT 3 M69432 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 07/96 105.
A-46 8105 BRUNSWICK 1 M69433 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 12/92
_m k
i l
106.
A-46 8105 BruusutCE 2 M69434 SEISMIC QUALIFICATION OF EQU!PMENT IN OPERATING PLAtTS 12/92
[
107.
A-46 8105 CALVERT CLIFFS 1 M69435 SEISMIC QUALIFICATION OF EQtnPMENT 14 OPERAT1mG PLANTS 06/96 108.
A-46 8105 CALVERT CLIFFS 2 M69436 SEISMIC QUALIFICATION OF EQUIPMENT fu OPERAT!aG PLANTS 06/96 109 A-46 8105 COOK 1 M69437 SEISMIC QUALIFICATI0u OF EQtJIPuf47 IN OPERATING PLANTS 12/92 110.
A-46 8105 C00C 2 M69438 SEISMIC QUALIFICAT104 0F EOJ.PMENT tu & ERAT!wG PLANTS 12/92 111.
A-46 8105 COOPER STATION M60439 SEISMIC QUALIFICATION OF EQUIPMENT 14 OPERATING PLAsTS 12/93 112.
A-46 8105 CRTSTAL river 3 M69440 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 04/95 i
113.
A-46 8105 Davis-8 ESSE 1 M69441 SEISMIC QUALIFICATION OF E13JIPMENT 19e OPERATimG PLANTS 09/95 114 A-46 8105 DRESDEu 2 M69442 SEISMIC QUALIFICATION OF EQJIPMNT In OPERATING FtANTS
/
115.
A-46 8105 DRESDEN 3 M69443 SEISMIC QUALIFICAfl0E OF EQUIPMEET 14 TERATING PLANTS
/
116.
A-46 8105 DUANE Atw0LD M69444 SEISMIC QUALIF1 CATION OF EQUIPMENT 14 OPERAfluG PLANTS 11/95
}
117.
A-46 8105 FARLEY 1 M69445 SE1SMIC QUALIFICATI04 OF EQtJ!P9ENT IN OPERAT!4G PLAETS 05/95 i
118.
A-46 8105 FIT 2PATRICE M69446 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERAfluG PLANTS 05/95 I
119.
A-46 8105 FORT CALMOUN 1 M69447 SEISMIC QUALIFICAT10N OF EJUIPMEuf tu OPERATING PLANTS 09/95 120.
A-46 8105 GINEA M69449 SEISMIC QUALIFICATION OF EQtJIPMENT I4 CPERAf twG PLANTS 05/95 121.
A-46 8105 kADDAM NECK M69450 SEISMIC QUALIFICATION OF EQUIPMEsf tu UPERAfinG PLANTS C3/93 122.
A-46 8105 aATCM 1 M69451 SEISMIC QUALIFICATI04 OF EQUIPMENT IN OPERAf tWG PLANTS 06/95 g
6 123.
A-46 8105 HATCM 2 M69452 SEISMIC QUAttFICATION OF EQUIPMENT Im OPERAT!WG PLANTS 06/95 124.
A-46 8105 INDIAN PCINT 2 M69453 SEISMIC QUALIF;CAft0N OF EQUIPMENT I4 OPERATING PLANTS 11/95 125.
A-46 8105 INDIAN Po!NT 3 M69454 SEISMIC OtJALIFICATION OF EQUIPMENT IN CPERATING PLANTS 12/94
[
126.
A-46
'8105 KEvaumEE M6% 55 SEISMIC QUALIFICATION Or EQUIPMEET IN CPERATIWG PLANTS 05/95
[
127.
A-46 8105 MILLSTONE 1 M69458 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS 12/95 128.
A-46 8105 MILLSTONE 2 M69459 SEISMIC QUALIFICATION OF EQUIPMENT la T ERATING PLANTS 05/v5 129.
A-46 8105 MONT1 CELLO M69460 SEISMIC QUALIFICAftDN OF EQt!!PM 4T In OPERATING PLANTS
/
130.
A-46 8105 NINE MILE Po!NT 1 M69461 SEISMIC QUALIFICATIOf OF EQUIPMEtt tu crERAftwG PtApTS 05/95 131.
A-46 8105 NORTM AnwA 1 M69462 SEISMIC QUALIFICATION OF EQtJIPMENT 15 OPERATING PLANTS 12/94 132.
A-46 8105 WORTN A4mA 2 M69463 SEISMIC QUALIFICATION OF EQUIPMENT IN OPERAT!WG PLANTS 06/95 133.
A-46 8105 OCONEE 1 M69464 SEISMIC QUALIFICATION OF EntJIPMFart 14 OPERATIuG PLAwTS 06/95
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134 A-46 8105 OCONEE 2 M69465 SEtSMIC QuALIFICATI04 cF EQUtrMENT su OPERAf tes PtAuTS 06/95 135.
A-46 8105 OCOMEE 3 M69466 SEISMIC QUALIFICATION OF EQUIPMENT I4 OPERATIuG PLANTS 06/95 136.
A-46 8105 OYSTER CREEK 1 M69467 SEISMIC QUALIFICATION OF EQUIP *'ENT Im OPERATING PLANTS C3/95 137.
A-46 8105 PAllSADES M69468 SEISMIC QUALIFICATI0E 7 EQUIPMENT 15 OPERATING PLAsTS
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A-46 8105 PEACH ECTTOM 2 M69469 SEISMIC QUALIFICATION OF EQLJIPMENT is CPERAfimG PLAsts 11/95 139.
A-46 8105 PEACB BOTTOM 3 M69470 SEISMIC QUALIF! CATION OF EQUIPMENT In OPERAtluG PLANTS 11/v5 140.
A-46 8105 PILG8tM 1 M69471 SEISMIC QUALIFICATION OF EQUIPMENT 15 OPERATim: PLAETS 12/95
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b APPENDIX C This appendix provides a detailed list, by issue, of the 258 GSI items not implemented, along with the projected date for completing the item. Status and projected implementation dates are presented as of September 30,1992.
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C C C Y A A A U U E E A A ON A A I I R R R R R A A A A L R R I C C D F M M N OOOOP P P Q Q B B C C C I P P Z Z A B B B FC C CC C D 7 777 7 777 77 777 7 777 777777 7333 3 3333 333 0 0 0 0 0 0 0 00 0 0 00 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 88 88 888 88 899999999999 8 8 8 8 8 8 8 8 8 8 8 8 8 8 L L L L L L L L L L 1 L L L L L L L L L L 4 4 4 4 4 4 4 44 4 4 4 4 4 77 77 7 77777 3333 3333333 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 888 8 88 8 8 8 8 8 88 8 8 8 8 8 88 8 8 8 89 999 9 999 999 8 8 88 88 8 88 8 888 88888888 88 8 8888 88 88888 L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L L G C G C G G GG G G G G G GGG C G GCC G C C C G G G G G CG CG G 67 89 01 234 5 678901 2 3 4 5 67 8 9 0 1 234 5 67 890 7 77 7 88 8 888 8 88 8 99 9 9 9 99 99 9 0 00 0 0 0 00 0 0 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2222 2222222
() d
)
i 5'
211.
GL-89-13 L913 FITZPATRICK M74003 SERVICE WTER SYSTEM PROBLERS AFFECTING SAFETY RELATED EeulPENT 10/92 212.
GL-89-13 L913 FORT CALMOUN 1 M74004 SERY1CE WTER SYSTEM PROBLEMS AFFECTIN3 SAFETY RELATFD EQUIPENT 11/92 i
213.
GL-89-13 L913 GINNA M74007-SERVICE WTER SYSTEM PR08LEMS AFFECTING SAFETY RELATED EcutPENT 12/92 214.
GL-89-13 L913. MOPE CREEK 1 M74014 SERVICE WTER SYSTEM PROBLEMS MFECTING SAFETY RELATED EeulPMNT 12/92 215.
CL-89-13 L913 MAINE YANKEE M74022 SERVICE W TER SYSTEM PROSLEMS AFFECTING SAFETY RELATED EQUIPMENT 12/93 216.
CL-89-13 L913 MCGUIRE 1 M74023 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 08/94 217.
GL-89-13 L913 McGb!RE 2 M74024 SERYtCE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED E001PMENT 08/94 218..
GL-89-13 L913 MILLSTONE 1 M74025 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED ESUIPMENT 07/93 l
219.
GL-89-13 L913-MILLSTONE 3 M74027 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 01/93 220.
CL-89-13 L913 NINE MILE POINT 1 M74029 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 03/93
- 5 221 GL-89-13 L913 NORTN ANNA 1 M74031 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 12/93 222.
GL-89-13 L913 NORTN ANNA 2 M74032 SERVICE W TER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 12/93 4,
223.
CL-89-13 L913 OtIINEE 1 M74033 SERVICE W TER SYSTEM PROSLEMS AFFECTING SAFETY RELATED EaUIPMENT 10/92 224.
GL-89-13 L913 OCONEE 2 M74034 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED Ecu1PMENT 10/92 225.
.L913 PALISADES M74037 SERVICE WTER SYSTEM PR05LEMS AFFECTING SAFETY RELATED EQUIPMENT 05/93 226.
GL-89-13 L913 PERRY 1 M74043 SERYlCE WATER SYSTEM PROELENS AFFECTING SAFETY RELATED EGutPMENT 12/93 227.
GL-89-13 L913-QUAD CITIES 2 M74050 SERVICE WTER SYSTEM PR00LEMS AFFECTING SAFETY RELATED EQUIPMENT 11/92 gg 228.
GL-89-13 L913-RIVER BEND 1 M74052 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 10/92 229.
GL-89-13 L913 SOUTH TEXAS 1 M74064 SERVICE WATER SYSTEN PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 10/92 l
230.'
GL-89-13 L913 SOUTM TEXAS 2 M74065 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPENT 10/92 231.
GL-89-13 L913 - SURRY 1 M74069 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 04/93 4
l 232.
GL-89-13 L913 SURRY Z M74070 SERVICE WATER SFSTEM PROBLEMS AFFECTING SAFETY RELATED Eeu!PMENT 04/93 233.
GL-89-13 L913 $USQUENANNA 1 M74071 SERvtCE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 10/92 1
234.
GL-89-13 L913. SUSouENANNA 2 M74072 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQU!PMENT 10/92 235.
GL-89-13 L913 TURKEY POINT 3 M74076 - SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EeUIPMENT 12/92
'913 TURKEY POINT 4 M74077 SERVICE WATER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 06/93 236.
GL-89-13 L
237.
GL-89-13 L913 WATERFORD 3 M74081 SERVICE WTER SYSTEM PROSLEMS AFFECTING SAFETY RELATED EQUIPMENT 04/93 238.
GL-89-13 L913 WOLF CREEK 1 M74088 SERVICE WTER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQUIPMENT 06/93 239.
GL-89-13 L913 ZION 1 M74090 TERYtCE WTER SYSTEM PtoeLENS AFFECTING SAFETY RELATED EQUIPMENT 12/93 240.
GL-89-13 L913 ZION 2 M74091 SERVICE WATER SYSTEM PROBLEMS AFFECTING SAFETY RELATED EQU1PMNT 12/93 241.
CL-90-03 LOO 3 BRAloWOOD 1 M76208 RELAXATION OF STAFF POS IN GEN LETTER 83-28, ITEM Z.2 PART 2 12/92 242.
GL-90-03 LOO 3 3RAtouo0D 2 M76209 RELAMATION OF STAFF POS IN GEN LETTER 83-28, ITEM 2.2 PART 2 12/92 243.
GL-90-03 LOO 3 3RUNSWICK 1 M76213 RELAXATION OF STAFF POS IN GEN LETTER 83-28, ITEM 2.2 PART 2 12/92 244 GL-90-03~
LD03 ' BRUNSWICK 2 M76214 RELAMATION OF STAFF POS IN GEN LETTER 83-28, ITEM 2.2 PART 2 12/92 4
245.
GL-90-03 LOO 3 eTRON 1 M76215 RELAXATION OF STAFF POS 16 GEN LETTER 83-28, ITEM 2.2 PART 2 12/94 4
i i
y i
246.
GL-90 LOU 3' SYaou 2 M76216 RELAXAf tDN OF STAFF POS IN GEk LETTEa 83-28, ITEM 2.2 PMT 2 12/94 247.
GL-90-03 LOO 3 D#ESDEN 2 M76232 RELAxAf t0N OF STAFF POS la GEM LETTER 83-28, ITEM 2.2 Peti 2 12/92 248.
GL-90-03 LOO 3-D*ESDEu 3 M76233 RELARATIOE OF STAFF POS IN Em LETTER 83-28, ITEM 2.2 Puf 2 12/92 249.
GL-90-03 LOQ3 luotAn P0tsi 2 M76247 RELARATION OF STAFF POS le GEW LETTER 83-28, ITEM 2.2 Puf 2 12/92 250.
GL-90-03 LOO 3 LASALLE 1 M76250 aELAXATION OF STAFF POS le GEu LET!En 83-28, ITEM 2.2 Pmi 2 12/92 251.
GL-90-03 LOU 3 LASALLE 2 M76251 mELARAf ton or STAFF POS tu GEu LETTER 83-28, ITEM 2.2 PMT 2 12/92 252.
GL-90-03 LOO 3 ' PILGRIM 1 M76276 RELARA1104 0F STAFF POS Im E4 LETTER 83-25, ITER 2.2 PmT 2 01/93 253.
GL-90-03 L003 ouA0 CITIES 1 M76281. RELARATION OF STAFF POS te Em LETTER 83-28, ITEM 2.2 PMT 2 12/92 254.
GL-90-03 LOQ3 cuAD CITIES 2 M76282 RELARATION OF STAFF POS IN gem LETTER 83-28, ITEW 2.2 PMT 2 T2/92 255.
GL-90-03 LOQ3 ' 210s 1 M76317 RELAKAff05 08 STAFF POS 15 Et LETitt 83-28, ITEM 2.2 Put 2 12/92 256.
GL-90-03 LOO 3 Ztou 2 -
- 76318 ' RELAXATION OF STAFF Pos tu GE4 LETTER 83-28, ITEM 2.2 PMT 2 T2/92 257 MPA-8023 8023 CRYSTAL sivEt 3 M10017 DEGRADED Ger0 VOLTAGE 12/93 258.
MPA-8023 3023' MtLLStosE 1 M60207-DEGRADED GRID VOLTAE 06/93 Oa
. o.
t f
1
Appendix D LISTING OF OTHER UN!MPLEMENTED MPA ITEMS BYISSUE
APPENDIX D This appendix provides a detailed list, by issue, of the 978 MPA items not implemented, along with the projected date for completing the item. Status and projected implementation dates are presented as of September 30,1992.
l l
l D1 l
~'
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l Other Mutti-Plant Actlens (Listirg of Opm items) twL ISSUE.
1.
SL-88-02 x802 DIAsLO CANTou 1 M67304 STEAM EuERATOR TUBE RUPTURE (888-02) (CLD WA 9099)
/
2.
SL-88-02 x802 OtASLO CAuYou 2 M67305 STEM GEuERATOR TUBE rJPTURE (B88-02) (OLD wt 8099)
/
3.
SL-88-02 x802 GlauA M67308 STEAM GENERATOR TUBE RUPTURE (B88-02) (OLD esPA 3099) 01/93 4.
BL-88-02 x802 nA00AM utCE M67309 STEM GEutRATOR TUDE RUPTURE (888-02) (OLD WA 9099)
/
5.
St-88-02 x802 McCutRE 1 M67313 STEAM CEuERATOR TUBE RUPTURE (B88-02) (OLD MPA 3099)
/
6.
BL-88-02 x802 ZION 1 M67332 STEAM GENERATOR TUBE RUPTURE (888-02) (OLD MPA 3099) 10/92 7.
SL-88-03 x803. stalms FERRY 1 M73852 CE WFA RELAYS (888-03) 07/96 8.
BL-88-03 x803 BRohass FERRY 3 M73854 GE elFA RELAYS (888-03) 01/94 9.
8t-88-04 x804 SROWuS FERRY 1 M69888 St PJeP FAILURE (888-04) (OLD MPA 5103) 07/96 10.
BL-88-04 x804 BROWNS FERRY 3 M69890 $1 SUeP FAILURE (888-04) (OLD MPA 8103) 01/94 11.
SL-88-04 x804 KEWAUMEE M69928 St PtMP FalLUPE (888-04) (OLD MPA 8103) 05/93 12.
SL-88-04 x804 NINE MILE posui 1 M69940 St PtpeP FAILURE (s88-04) (OLD uPA 5103) 03/95 13.
St-88-04 x804 OccufE 1 M69944 St PtreP FAILURE (888-04) (OLD MPA 3103) 12/92 14 BL-88-04 x804 OccuEE 2 M69945 St PueP FAILURE (888-04) (CLD MPA 3103) 12/92 C
15.
BL-88-04 x804 OccuEE 3 se69946 St PUMP FAILURE (888-04) (OLD MPA 3103) 12/92
+
$3 16.
8L-88-04 x804 TROJAu M69984 St PUMP FAILURE (B88-04) (CLD MPA 5103) 08/93 17; BL-88-04 x804-2 ION 1 M69995 St PUMP FAILURE (988-04) (CLD MPA 5103)-
10/92 18.
BL-88-07 x807 SRtkus FERRY 1 M72805 POWER OSCTLLATIONS tu SWR *$ (888-07) 07/96 5
19.
BL-88-07 x807. SROWNS FLRRY 3 M72769 POWER OSCILLATIONS In 9WR'S (888-97) 01/94 20.
BL-88 08 x808 BRAIDWOOD 1.
se69602 inERMAL STRESS 14 PIPluG CONNECTED TO RCS (888-08)(CLD MPA B107) 04/94 21.
BL-88-08 x808 BRAIDWOOD 2 se69603 THERMAL STRESS tu P! Plug ComuECTED TO RCS (988-08)(OLD MPA 5107) 04/93 22.
SL-88-08 x808 STR04 1 M69609 TMERenAL STRESS la P! Plug ComuECTED TO RCS (888-08)(OLD wA 3107) 04/93 23.
BL-88-08 x808 BTRom 2 M69610 THERMAL STRESS fu PIPluG ComuECTED TO RCS (888-08)(CLD sePA 3107) 10/93 24 st-88-08 x808 CLluTou 1 se6%16 THERMAL STRESS lu PIPluG CONNECTED TO RCS (988-08)(CLD MPA 3107) 12/92 25.
BL-88-06 x808 COOK 1
. M69618 TMEMAL STRESS In PIPluG ComuECTED TO RCS (888-08)(OLD MPA 5107) 12/93 26.
BL-88-08 x808-COOK 2 M69619 TWERetal STRESS la PIP!aG ComuECTED TO RCS (308-08)(OLD MPA 3107) 12/93 5I 27.
BL-88-08 ISOS utuE setLE Potsi 1
' se69655 inERMAL STRESS tu *IPlus ComuECTED TO RCS (sSS-08)(OLD MPA 3107) 03/95 1
28.
8t-88 08 x808 ss0RTu AmeA 1 se69657 inERMAL STRESS lu PIPluG C3suECTED TO RCS (388-08)(OLD sePA 3107)
/
4 29.
8t-88-08 x808 WORTW AsusA 2 se69658 TWEPMAL STRESS la PIPluG ComuECTED TO RCS (988-08)(OLD sePA 3107)
/
30.
St-88-08 2808 OConEE 1 M69659 TMEteeAL STRESS 14 PIPluG COuuECTED TO RCS (388-08)(OLD MPA 5107) 03/93 31, 8L-88-08 x808 CCouEE 2 se69660 TNEtenAL STRESS la PIPluG ComuECTED TO RCS (SM-08)(OLD sePA 3107) 03/93 32.
BL-88-08 x808 OCOMEE 3 M69661 THERMAL STRESS tu PIPING ComufCTED TO RCS (908-08)(OLD sePA 3107) 03/93 33.
8L-88-08 x808 PALO VERDE 1 M69664 THEteeAL STRESS fu P! Plug ComuECTED TO RCS (908-08)(CLC MPA 5107)
/
34.
BL-88-08 x808 PALD VERDE 2 M69665 TWEResAL STRESS tu P! Plug ComutCTED TO RCS (888-08)(OLD MPA 3107)
/
35.
8L-88-08 x808 PALO VERDE 3 se69666 THERenAL STRESS In P! Plug ComuECTED TO RCS (388-08)(OLD MPA 3107)
/
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A 106.
OL-90-01 N001 FAntET 1 M76554 toss OF FILL Ott tu WOSEnouuf Tamuss!TTERS (s90-01) 12/92 107.
BL-90-01 x001 FAttET 2 M76555 LOSS OF FILL O!L tu #0SEMOUni TRAuSRtTTERS (990 01) 12/92 108.
St-90-01 2001 FITZPATatCK M76557 LOSS OF FILL Ctt tu ROSEuOUmf TRAuSMITTERS (890-01) 12/92 109 SL-90-01 N001 GaAn0 CULF 1 M76560 LOSS OF FILL Ott tu aoStuount TamusstTTERS (390-01) 03/93 110.
SL-90-01 x001- #AePts 1 M76562 LOSS OF FILL CIL tu postuousi TanusmrTTEa5 (390-01) 12/92
'111.
BL-90-01 X001, uATCu 1 476563 LOSS OF FILL 01L Im ROSE 40Unf TRAuSMITTERS (890-01)
/
112.
BL-CD-01 N001 NATCu 2 M76564 LOSS OF FILL OIL Ju 90SEm0Uuf TRANSp!TTERS (890-01)
/
113.
BL-90-01 x001. n0PE CREEE 1 M76565 LOSS OF FILL CIL IM #0SEMOUuT TemusptTTERS (8f0-01) 01/93 114 BL-90-01 N001 INDIAu 70!uf 2 M76566 1055 0F FILL Ott tu #0SEm0UuT TRAuSmITTERS (3(0-01) 12/92 115.
SL-90-01 N001 suDIAu Potui 3 M76567 LOSS OF FItt Ott tu ROSEMOLust TRAusntTTEas (00-01) 12/92 116.
SL-90-01 x001 LASALLE 1 M76569 LOSS OF FILL ott tu ROSEmaunt itAuSurTTERS (SM-01) 12/92 117.
St-90 01 N001 LASALLE 2 M76570 Loss OF FILL ott la ROSEpoUNT TRAuSatTTEa5 (890-01) 12/92 118.
BL-90-01 N001 LIMERICK 1 M76571 LOSS OF FILL CIL Im ROSEmmut TRAusmITTERS (390-01) 12/92 119.
BL-90-01 K001 LluERICK 2 m76572. LOSS OF FILL OIL tu #0SEn0 Uni TaAusmITTERS (390-01) 12/92 C
120.
SL-90-01 N001 MAINE TAutEE M76573 LOSS OF FILL CIL Im ROSEn0 Uni TamuSp!fitas (390-01) 12/93 d) 12Y.
BL-90-01 2001 MCtsteE 1 M76574 LOSS OF FILL Ctt tu ROSEn0Umf TRAuSMITTEt3 (890-01) 12/92 122.
BL-90-01 x001 mCGutet 2 M76575 LOSS OF FILL O!L Im ROSEn0Uut TRAusm!TTEas (590-01) 12/92 123.
st-90-01 x001 MitLSiout 2 m76577 LOSS OF FILL Ott 14 aOStuoumf TaAuSmtTTEa5 (s90-01) 12/92 124.
'8L-90-01 "N001 MILtSTONE 3 M76578 LOSS OF FILL O!L Im ROSEMOUut TRANSMITTEt3 (890-01) 12/92 125.
st-90-c*
x001 MouTICELLO M76579 LOSS OF FILL CIL tu ROSEMl3MT TRAuSMITTERS (890-01)
/
126.
SL-90-01 x001 usuE MILE P01mf 1 M76580 LOSS OF FILL Oil tu ROSE *WUmf TaAusm!TTEt3 (390-01) 12/92 127.
BL-90-01 "N001 stuE p LE P014T 2 M76581 LOSS OF FILL CIL tu ROSEMOUuT TWAuSRITTERS (390-01) 12/92 128. ' SL-90-01
'N001 sw im AmuA 1 M76582 LOSS OF FILL Ott tu ROSE 40tmi inauSm:1 tea 1 (390-01) 12/92 129.
SL-90-01 K001 NORTu AmeA 2 M76553 - LOSS OF FILL OtL tu ROSEMOUnt TRAuSRITTERS (390-01) 12/92 130.'
SL-90-01 x001 OccuEE 1 M76584 LOSS OF FILL Ott tu ROSEu0uut TRANSMITTERS (990-01) 12/92 131.
BL-90-01 XOOT OCouEE 2 M76585 LOSS OF FILL Ott tu 90SE40UmT T8AuSMITTERS (990-01) 12/92 132.
BL-90-01 N001 000mEE 3 M76586 LOSS OF FILL CIL tu SCSEMOUut itAusmITTERS (390-01) 12/92 133.
x001 OTSTER CREEE 1 M76587 LOSS OF FILL OIL In 90SENOUni TatuSRITTERS (890-01) 12/92 134.
St-90-01 N001 PALO VERDE 1 M76584 LOSS OF FILL Ott tu WOSE40UnT TRauSMITTERS (340-013 12/92 135.
SL-90-Of N001 PALO vtaDE 2 M76590 Loss CF FILL OfL In 80SEn0 Uni TRANSm!TTERS (390-01) 12/92 136.
.M001 PALO DERDE 3 M76591 Loss OF fitt O!L is nOSFwiunt TamusntTTEas (s90-01) 12/92 137.
et-90-01 N001 PEACE 80TTOR 2 M76592 LOSS OF FILL OIL te WOSEnouuf TRAuSMITTERS (390-01) 12/92 138.
BL-90-01^
2001 peace BOTTCue 3 m76593 Loss CF F LL CIL tu eOSEm0Umf TeAusmtTTERS (390-01) 12/92 139.
St-93-01 K001 PEpei 1 N76594 LOSS OF FILL Ott tu ROSEMOUni inAusuf fitas (390-01) 01/93 140.
8L-90-01 N001 FILGain -
M76595 LOSS OF FILL CIL tu BOSF40Umf TRAusMiTTEts (340-01) 12/93
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st-92-01 E201 SIG ROCK P0 tai 1 M83845 TERMO-LAG (80LLEttu 92-01)
/
177.
BL-92-Oi M201 saAIDWOOD 1 M83846 inEf*IO-LAG (ButtEf te 92-01)
/
178.
St-92-01 K201 ssAIDuctxt 2 M83547 inEasuo-TAG (But1Ef ts 72-01)
/
179 BL-92-01 x201 sa0Wm3 FEttY 1 M83848 TMEDMO-LAG (90LLEf ts 92-01)
/
180.
st-92-01 r201 se0WNS FERef 2 e483849 TMERMO-LAG (SULLETis 92-01)
/
181.
st-92-01 x201 BROWes FERRY 3 M83850 fuERMO-LAG (9ULLETIs 92-01)
/
182.
St-92-01 x201 seuuSWICK 1 M83851 inER*C-LAG (9ULLETim 92-01) 11/92 183.
BL-92-01 x201 SeUNSWICK 2 -
M83852 TuteMo-LAG (0:1LEf ta 92-01) 11/92 186 St-92-01 X201 STRON 1 MS3853 THERMO-LAG (SULLEf ts 92-01)
/
185.
SL-92-01 x201 STRON 2 M83854 TMERMO-LAG (SULLEf ta 92-01)
/
186.
SL-92-01 X201 CALLAWAY 1 M83855 inEWWLAG (SULLETsu 92 01) 11/02 187.
st-92-01 N201 CLINTou 1 M83860 inEMO-TAG (SULLEf f e 92-01)
/
188.
sL-92-01 K201' CCMANCME PEAK 1 M83861 TMEWMO-LAG (SULLEf f e 92-01) 70/92 189 BL-92-01 x201 CCCK 1 M83863 TMER'C-LAG (BULLEf ts 92-01) 12/93 h
190.
sL-92-01 K201 Coot 2 M83864 TpfRMO-LAG (BULLET!s 92-01) 12/93 CO 191.
st-92-01 x201 COOPER STAitou Nf3865 inERMD-LAG (BULLEf ts 92-01)
/
192.
St-92-01 N201 CRTSTAL river 3 M83566 TNEROLAG (BULLETis 92-01)
/
193.
BL-92-01 E201 DAVIS-BESSE 1 M83867 TNERMO-LAG (BULLEfta 92-01) 12/93 196 8L-92-01 X201 01A310 CAmTou 1 M83868 TWERMD-LAG (EULLETIE 92-01) 01/93 195.
st-92-01 K201 DIAst0 CANTCM 2 M83869 f pERMO-LAG (BULLETIN 92-01) 01/93 196.
st-92-01 x201 DUAhE Atm0LD M83872 TwERMO-LAG (BULLEf ta 92-01)
/
197 BL-92-01 X201 FERMt 2 M83875 THER'C-LAG (BULLEf ta 92-01)
/
198.
SL-92-01 x201 GINEA M83878 THER*ED-LAG (BULLET!W 92-01)
/
199.
sL-92-01 K201 GR MD CULF 1 M83879 inERMO-LAG (ButLEf te 92-01) 12/93 200.
st-92-01 K201 pA00AM mEcr 8483880 THERMO-LAG (90LLETim 92-01)
/
201.
st-92-01 N201 matris 1 M83851 TMER40-LAG (BULLEf ts 92-01) 12/92 202.
St-92-01 M201 mATCn 1 M83882 TmERMO-LAG (BULLETIE 92-01)
/
203.
sL-92-01 X201 MATCn 2 M83883 TPEmeo-LAG tButLETtu 92-01)
/
20r..
SL-92-01 K201 tuotAs Potsi 2 M83885 TMERMO-LAG (BULLETis 92-01)
/
205.
BL-92-01 K201 LASALLE 1 M83888 7MERMO-LAG (BULLEttu 92-01)
/
206.
BL-92-01 K201 LASALLE 2 M83829 "EPMo-TAG (9ULLEf ts 92-01)
/
207 sL-92-01 X201 LIMERICK 1 M83890 THEPMD-LAG (9ULLETIs 92-01)
/
208.
St-92-01 1201 LIMERICK 2 M83891 inEaMO-LAG (9ULLEfts 92-01)
/
209.
sL-92-01 x201 MCtattrE 1 M83893 TNEEMD-LAG (SULLEf ts 92-01)
/
210.
BL-92-01 x201 MCtarteE 2 M83896 TuEWMD-LAG (SULLEfts 92-01)
/
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.211.. SL-92-01 X201' #1LLST0mE 1 m83895 fuEauo-TAG (eutLEfte 92-01) 12/92
'212.
SL-92-01 X201. MittsfouE 2 483896 TufRsO LAG (SULLET1e 92-01)
/
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213.
SL-92-Of R201 atttsfomE 3 43897 TMEam0-LAG (SULLEffe 92-01)
/
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214.
SL-92 x201 acusTICELLO M83808 inEWuO-LAG (SULLEfte 92-01) 03/93
- 215.
SL-92-01 x201 stuE MILE PoteT 1 uS3899 ' TREEuc-LAG (
1774 92-01) 12/92 21)
Oe/93 216. -SL-92-01 x201 miaE MILE POluf 2 M83900 TWEsuo-LAG (tv-217.
SL-92-Ot-x201 monin AmeA 1 uS3901 inEeuO-LAG ' gL i t
/
218.
St-92-01 X201-morts Amen 2 483902 inEamo-th A
/
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/
219.'
SL-92-01 x201' CTSTER CREEE 1 M83906 - TMERuG-LAr. o
+
'31)
/
220.
SL-92-01 M201 PALisa0E5 863907 Tutono-LAG (t 221.
SL-92-01 x201 Petc wade 1 M83908 TpEauO-TAG (K1lM
. !?
/
2224 SL-92-01 x201 PALO W NDE 2 M83909 inEsmo-LAG (gutLEti 41 )
/
223.-
BL-92-01 E201-PALO VERDE 3 M83910 TMEauo-LAG (OULLE14 w J1)
/
226 SL-92-01' x201 ' PEACW SOTTtM 2 -
483911 TuE9a0-LAG (SJLLETl= 92-01)
/
9 225.
BL-92-01 x201. - PEACM SOTTOR 3 483912 Tuttuo-LAG (SULLEfts.,2 31)
/
EO 226.
BL-92-01.
x201-PERRY 1 -
MS3913 TMEauO-LAG (SULLEtta 92-01)
/
-227. ' BL-92-01
-x201 PRAta!E Islamo 1
' m3917 TwauO-LAG (sULLEf ta 92-01)
/
222.
BL-92-01 x201 Pentatt Islaan 2 M83918 Tutan0-TAG (autLE!!s 92-01)
/
22 0 SL-92-01 x201-ouse CITIES 1 963919 TutauO-LAG (autLEfts 92-01)
/
230.
St-92-01' x201 eUno CITits 2 M83920. ' TWE840-LAG (SULLEf ts 92-01)
/
231 SL-92-01
.x201 San OuDF9E t M83927 TuEauO-LAG (SULLETle 92-01) 05/93 232.
SL-92-01 x201 SAs ONOFWE 2 M83928 TWE940-LAG (BULLET 1s 92-01) '
/
'233.
st-92-01 x201 saa omoFaE 3 563929 THERuo-LAG (EULLETTe 92-01)
/
236.
SL-92-01 x201 sEmancot t 863930 Turmao-LAa (autLEfta 92-01) 12/92 235. 'SL-92-01
-x201 steUOTAa 1 863931' TWEWup-LAG (SULLETtu 92-01)
/
236.
SL-92-01 K201 'SEeUOTA# 2 M83932 TWEase-LAG (SULLEf ts 92-01)
/
. 237 BL-92-01
.. N201 SouTu TEMAS T' e63933 TWEa40-LAG (SULLEf fe 92-01)
/
238.'
SL-92-01
' X201 50UTa TEMAS 2 M83936 TMEEuo-LAG (EULLETIe 92-01)
/
m3923,. TgEsse-LAG (SULLEf te 92-01)
/
239.
BL-92-01 K201 ST LUCIE 1 240..BL-92-01 M201 ST LUCIE 2 n83926 - Tutano-teG (sulttitu 92-01),
/
241.
BL-92-01 X201 StsumEn 1 4 3935 : TaEauo-LAG (sultEf te 92-91)
/
242.
SL-92-01 x201 susti 1 M83936 TM ano-LAG (sulLEffe 92-01)
/
263. 'BL-92-01 x201 sUmet 2 863937 THEamo-LAG (SULLEf te 92-01)
/
244.
SL-92-01 X201 SusouEummeA 1 M83938. THEeuo-LAG (SULLETis 92-01) 06/93 265.
SL-92-01 x201 SuseuEgamma 2 M83939 inE8MO-LAG (EULLETim 92-01) 06/93
__._...___._._,_m____.
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246.
SL-92-01-.
K201 inmEE MILE ISLAe 1 suR39mo THE8ee-LAG (EULLEf fe 92-01) 12/93 j~
247.
sL-92-01 x201' Ta0 Jam n83961 - TEnno-tas (autLEf te 92-01) 02/96
[
268.
SL-92-01 E201-TuuET Potut 3 M83962 TE nne-lag (autLETsu 92-01)
/
4-269.
BL-92-01
-.x201 TUetET POluf 4 M83963 inEmme-Lac (auttEffe 92-01)
/
(
250.
SL-92-01 x201 VEautpri YnutEE 1 M83964 THEauc-LAG (gulLETru 92-01)
/
l -'-
251.
SL-92-01
.x201 v0GTLE 1.
m83965 THEamo-LAG (OULLETtu 92-01)
/
[
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252.
SL-92-01 K201 v0GTLE 2 M83966 THEenO-LAG (gulLETis 92-01)
/
i 253.
SL-92-01 R201 WasatuGTou muCLEAR 2 M83950 inEauO-LAG (autLET!u 92-01)
/
)
254.
SL-92-01 E201-WATERFORD 3 M83947 TuEeuO-LAG (90LLETim 92-01)
/
I 255.
BL-92-01 N201' WOLF CREEE 1 M83951 TuE1Me-Las (autLEflu 92-01)
/
256.
BL-92-01 K201 2 ION 1 se83952 Tutsee-LAG (SULLEffe 92-01)
/
j 257.
St-92-01 K201 2tou 2.
se3953 THE8MO-LAG (SULLETtu 92-01)
/
}
258.
'C014. OccutE 1 M43643 ' aurittARY FEEcautTEa SEtSatC ouattFICAfstas (GL81-14) 01/93 259.. GL-81-14 C014 OCouEE.2 se63644 AuxtttAtv FEEDWATER set 5ptC OUALIFICATION (GL81-14) 01/93 l
260.
GL-81-14 C014 OtouEE 3 M&3645. AUKILIART FEEDWATER set 9trC OunLIFICATIO6 (GL81-14) 01/93 I
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261.
GL 83-08 0020' se0MuS FERWT 1 M57146 STARE I Ot?WELL VACUUN OstaKERS (GL83-08) '
07/96 i
262.
GL-84-09 A019 - COOPER STaftom M55322 attpsetuEt CAPASILITT DEGU1REMEuTS OF TO CFR 50.44 (GL86-09)
- 70?92 e
l 263.
GL-84-09 A019 DRESDEN 2 sG6579 REC 04tutt CarastLITT REQUIREuEuTS OF 10 CF# 50.44 (GL86-09)
C9/93 U
264.
GL-84-09 A019 DeESDEu 3 M56580 RECanstatt CAPAstLTTY RE3utateEEuTS OF to CFR 50.44 (GL84-09) 09/93 I
l 265.
GL-84-09 A019 MILLST0mE 1 M65067 REtapu3 sta CAPASILITY WEGuteEMuTS OF 10 CFR 50.44 (GLM-09) 06/93 l'
' 266. ' 'CL-84-09' A019 'CTSTER CREEE 1 se62980. RECDuetuER CAPASILITT RE001sE8EuTS OF 10 CFR 50.44 (GL86-09) 03/93 267.
GL-84-09 A019 Oump CITIES 1 M55148 af ttpstata CAPASILITY REGutBENEuTS OF TO CFt 50.44 (GLE-09) 09/93 j
269. ~ GL-84-09' A019 CUmD CfTIES 2 M55149 DECoustuER CApastLITY REGUt#EMEuTS OF 10 CFt 50.44 (GL84-09) 09/93 j
268.
GL-87-09 '
0024 D CALLaun? 1 8482221. MODE CNAGES ase (CD*S - TECu SPECS 3.0 auD 4.0 (GL 87-09) 12/92 I
270.
GL-87-09 '
'0024 COOK 1 M79836 MODE CMAGES AND LCO*5. TECW SPECS 3.0 AuD 4.0 (GL 87-09) 11/92
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2 71.
GL-87-09 0024 C00E 2 '
M79837 MieE CanGES anD LCD*3 - TECW SPECS 3.0 a.JD 4.0 (GL 87-09) t0/92 -
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~ 272.
GL-87-09 0024 maarts 1 M19410 ' suoDE Cnemt ase LCD*$ i TECM SPECS 3.0 aus 4.0 (GL 87-09) 12/92 2 73.
CL-87-09 D024 LASALLE 1 M75789 setBE CnaGES nuD LCD'S - TEtn SPECS 3.0 ase 4.0 (GL 87-09) 11/92 i
2 74 GL-87-tr#
D024 ~ LA$aLLE 2 M75790 seDE DAGES ase LCD*S
- TECB SPECS 3.0 auD 4.0 (GL 87-09) 11/92 275. : CL-87-09 0024 Patt$aDES M72752. MODE Cn4GES mee LCD'S - TECN SPECS 3.0 aup 4.0 (GL 87-09)
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. 276.. GL 09 0024 ' PALO WESDE 1 8e1840 moeE CunGES anD LCo*$ - 7ttu SPECS 3.0 amp 4.0 (GL 87-09)
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GL-88-20 8111 PitGRIM 1 M74451 luDIVIDUAL PLANT EVALUATIOrs (GL 88-20) 08/95 407.
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GL 88-20 8111 WASNINGTON NUCLEnn 2 M74489 INDIVIDUAL PLANT EVALUATIONS (GL 88-20) 06/93 l
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701 GL-92-01 8120 BROWNS FERRY 2 M83439 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/73 702.
GL-92-01 8120 BROWNS FERRY 3 M83440 REAC10R VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 703.
GL-92-01 8120 BRUNSWICK 1 M83441 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 704 GL-92-01 8120 BRUNSWICK 2 M83442 Rf A. TOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 705.
GL-92-01 8120 STROM 1 M83443 REACTOR VESSEL STRUCTLRAf. INTEGRITT (GL 92-01) 12/93 706.
GL-92-01 8120 BTRON 2 M83444 REACTOR VESSEL STRUCTUht INTEGRITY (GL 92-01) 12/93 707.
GL-92-01 8120 CALLAWAT 1 MS3445 RE ACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01)
/
708.
GL-92-01 8120 CALVERT CLIFFS 1 M83446 REACTOR VESSEL STRUCTURAL INTEGRITY (CL 92-01) 72/93 709.
GL-92-01 8120 CALVERT CLIFFS 2 M83447 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 710.
GL-92-01 8120 CATAWBA 1 M83448 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01)
T2/93 711.
GL-92-01 8120 CATAWBA 2 M83449 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 712.
GL 92-01 8120 CLINTON 1 M83450 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 72/92 733.
GL-92 01 8120 COMANCHE PEAK 1 MS3451 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 714.
CL-92-01 8120 C00K 1 MS3453 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 715.
GL-92-01 8120 COOK 2 MS3454 REACTOR VESSEL STRUCTURAL INTEGRlTY (GL 92-01) 12/93 gg (a)
)#6.
GL 92-01 8120 COOPER STATION M83455 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 717.
GL-92-01 8120 CRYSTAL RIVER 3 M83731 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01)
/
718.
GL-92-01 8120 DAVIS-BESSE 1 M83732 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/94 719.
f;L-92-01 8120 OIA8LO CANTON 1 MS3456 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 02/93 720.
Gt-92-01 8120 DIA8t0 CANTON 2 M87457 REACTOR VESSt *. STRUCTURAL INTEGRITY (GL 92-01) 02/93 721.
GL-92 01 8120 DRESDEN 2 M83458 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 722.
CL-92-01 8120 DRESDEN 3 M83459 REACTOR VESSEt STRUCTURAL INTEGRITY (GL 92-01) 12/93 723.
CL-92-01 8120 DUANE ARNOLD M83460 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 724. ~ GL 92-01 8120 FARLEY 1 M83461 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-07) 12/93 725.
GL 92-01 8120 FARLEY 2 M83462 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 726.
GL-92-01 8120 FITZPATRICK MS3464 REACTOR WESSEL STRUCTURAL INTEGRITY (GL 92-01)
T2/93 727.
GL-92-01 8120 FORT CALHOUN 1 M83465 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 01/94 728.
GL-92-01 8120 'GINNA M83733 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01)
T2/93 729.
GL 92-01 8120 GRAWO GULF 1 MS3466 REACTOR VESSEt STRUCit*AL INTEGRITY (GL 92 01) 12/94 730.
GL-92-01 8120 HADDAM WECK MS3467 REACTOR VES!EL STRUCTURAL INTEGRITY (GL 92-01) 12/93 731.
GL-92-01 8120 HARRIS 1 M83468 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 732.
GL-92-01 8t20 MATCH 1 M83469 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 733.
CL-92-01 8120 HATCM 2 M83470 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 734.
CL-92-01 8120 HOPE CREEK 1 M83471 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 735.
GL-92-01 8120 INDIAN POINT 2 M83472 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93
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CL-92-01 8120 INDIAN POINT 3 M83473 REACTOR VESSEL STRUCTURAL INTEGRITY (CL 92-01) 12/92 73 7.
GL-92-01 8120 REWAUNEE MS3474 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92 01) 12/93 738.
GL-92-01 8120 LASALLE 1 M83475 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 739 Gt-92-01 8120 LAMLLE 2 MS3476 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 740.
GL-92-01 8120 LINERICK 1 M834TT REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 741.
GL-92-01 8120 LIMERICK 2 :
M83478 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01).
12/93 742.
CL-92-01 8120 MAINE TAsKEE M83479 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/92.
E 743.'
GL-92-01 8120 MCGUIRE 1 M83480 REACTOR VES?tt STRUCTURAL INTEGRtTY (GL 92-01) 12/93
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GL-92-01 8120 MCGulRE 2 M83481 REACTOR VESSEL STRUCTURAL INTEGetTT (GL 92-01) 12/93-745.
'8120 MILLSTONE 1 MB3482 REACY(R VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 746.
GL-92-01 8120 MILLSTONE 2 M83483 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 l
747.
GL-92-01 8120 MILLSTONE 3-M83484 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 7
748.
CL-92-01 8120 MONTICELLO M83485 REACTOR VESSEL STRUCTL1tAL INTEGRITY (GL 92-01) 12/93 749 GL-92-01 8120 WINE MILE PolNT 1 M83486 REACTOR VESSEL STRUCTURAL 8WTEGRITY (GL 92-01) -
12/93 9
750.
CL-92 01 8120- NfhE MILE P0 TWT 2 M83487 REACTOR VESSEL STRUCTURAL INTEGRITT (GL 92-01) 12/93 FO 751.
GL-92 8120 NORTN ANNA 1 M83488 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01)
,12/93 752.
GL-92-01 8120 NORTW AN64 2 M83489 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 753.
.CL-92-01 8120 OCONEE 1 M83734 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) --
12/93 754.
CL-92-01 8120 OCOMEE 2 M83735 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 755.
GL-92 01 8120 OCOMEE 3.
M83736 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 756.
GL-92-01 8120 CTSTER CREEK 1 M83490 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 07/94 757 CL-92-01 8120' PALISADES M83491. REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01)
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GL-92-01 8120 PALO VERDE 1 M83492 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 759.
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GL 92-01 8120 PEACM SOTTCM 3 M83496 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 763.
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GL-92-01 8120 ' PRAIRIE ISLAND 1
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CL-92-01 8120 PRAIRIE ISLAND 2 M83500 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 768.
CL-92-01 8120 - GUA0 CITIES 1 MB3501 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 769.
GL-92-01 8120 QUA0 CITIES 2 863502 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 770.
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Gt-92-01 8120 - R00!NSON 2 M83504 REACTOR VESSEL STRUCTURAL INTEGRlif (GL 92-01) 12/93 772.
GL-92-01 8120 SALEM 1-M83507 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 773.
-8120 SALEM 2 M83508 REACIOR VESSEL STRUCTURAL INTEGRiiY (GL 92-01) 12/93 774.
GL-92 8120 SAN ONOFRE 2 MB3510 REACTOR VESSEL STRUCTURAL laiTEGRITY (GL 92-01) 12/93 775.
CL-92-01 8120 SAN ONOFRE 3 M83511 ' REACTOR VESSEL STRUCTURAL' INTEGRITY (GL 92-01) 12/93 776.
GL-92-01 8120 SEASROOK 1 M83512 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 01/94 777.
GL-92-01 0120' SEQUOTAN 1 M83513 REACTGt VESSEL STRUCTURAL INTEGR!iY (GL 92-01) 12/93 773.
GL-92-01 8120- SEQUOTAR 2 M83514 REACTOR VESSEL STRUCTURAL 11TEGRITY (GL 92-01) 12/93 779.
GL-92-01 8120 SOUTH TEMAS 1 M83515 REACTOR WESSEL STRUCTURAL INTEGRiff (GL 92-01) 12/93 780.
GL-92-01 8120 SOUTH TEXAS 2 M83516 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 781.
GL-92-01, 8120 Si LUCIE 1 MS3505-REACTOR VESSEL STRUCTURAL luiEGRITY (GL 92-01) 12/92 782.
GL-92-01 8120 ST LUCIE 2 M83506 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/92 783.
GL-92-01 8120 SURRY 1' M83739 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 784.
GL-92-01 8120 SURRY 2 M83740 REACTOR VESSEL STRUCTURAL INTEGRiff (GL 92-01) 12/93 9
785. 'GL-92-01 8120 SUSOUENANNA 1 M83518 REACTOR WESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 M
786.
'8120 SUSQUEHANNA 2 M83519 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93' 787.
GL-92-01 8120 TNREE MILE istase 1-M83741 - REACTOR VESSEL STRUCTURAL' INTEGRITY (GL 92-01) 12/93 788 ' GL-92-01 8120 - TROJAN M83520 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 789.
.8120 TURKEf POINT 3 M83742 REACTOR VESSEL ST4UCTURAL INTEGRiff (GL 92-01) 07/93 790.
GL-92-01 8120 > TURKEY POINT 4 M83743 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 07/93 791.
GL-92-01 8120 ~ WERMONT TaawEE 1 M83521 REACTOR VESSEL STRUCTURAL INTEGRITY (CL 92-01)
/
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GL-92-01 l8120 V0GTLE 1 M83522 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93
.793 GL-92 8120 V0GTLE 2 M83523 REACTOR VESSEL STRUCTURAL INTEGRITY (WL 92-01) 12/93 794.
GL-92-01 8120 WASRINGTON NUCLEAR 2 M83527 REACTOR VESSEL ffRUCTURAL INTEGRiff (GL 92-01) 12/93e 795.
GL-92-01 8120 WATERFORD 3 M83524 REACTOR VESSEL STRUCTURAL INTEGRIff (GL 92-01) 12/93 796.
GL-92-01 8120' WOLF CREE 1C 1 M83528 ~ REACTOR WESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/92 4
797.
GL-92-01 8120 2tDN 1 M83744 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 798.
GL-92-01 8120 ZION 2 M83745 REACTOR VESSEL STRUCTURAL INTEGRITY (GL 92-01) 12/93 799.
GL-92-04 8121 BIG ROCK P03NT 1 M84268 BWR LEVEL INSTRUMENTATION (GL-92-04)
/
800.
GL-92-04 8121 BROWNS FERRY 1 M84269 BWR LEVEL INSTRtRENTATION (GL-92-04)
/
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GL-92-04 8121 840WNS FERRY 2 M84270 ' BWR LEVEL INSTRUMENTATION (Gt-92-04) ~
/
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GL-92-04 8121 BROWNS FERRY 3 M84271 'BWR LEVEL INSTRUMENTAll0N (GL-92-04)
/
803.
GL-92-04 8121 BRUNSWICK 1 M64272 8WR LEVEL INSTRUMENTATION (CL-92-04)
/
804. ' GL-92-04 0121 BRUNSWirr 2 M84273 BWR LEVEt INSTRtmENTATION (GL-92-04)
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GL-92-04 8121 CLINTON 1 nTA274 - aWR LEVEL INSTRUMENTATIDW (GL-92-04)
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/
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GL-92-04 8121 DRESDEN 2 M84276. SWR LEVEL INSTRUMENTATION (CL-92-06)
/
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GL-92-04 812T DRESDEN 3 M842TT BWR LEVEL INSTRUMENTATION (GL-92-04)
/
809.
GL-92-06 5121 DUANE ARNOLD M64278 BWR LEVEL INSTRUMENTATION (GL-92-04)
/
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/
811.
GL-92-04 8121 FITZPATRICK M84280 BWR LEVEL INSTRUMENTATION (GL-92-04)
/
812.
'8121 CRAND GULF 1 M64281 BWR LEVEL INSTRWENTATION (Gt-92-06) 12/93 813.
GL-92-06 8121 MATCH 1
.M64282 8WR LEVEL INSTRUMENTATION (GL-92-06)
/
814.
GL-92-06 e121 MATCN 2 M84283 SWR LEVEL INSTRUMENTATION (GL-92-04)
/
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.B121 n0PE CREEK.1 M64286 8WR LEVEL INSTRWENTATION (GL-92-04)
/
816.
CL-92-04.
8121 LASALLE 1 M64285 SWR LEVEL INSTRUMENTATION (GL-92-04)
/
817.
CL-92-04
'8121 LASALLE 2 M64286 BWR LEVEL INSTRUMENTATION (CL-92-04)
/
818.
GL-92-04 B121 LIMERICK 1 M64287 BWR LEVEL INSTRUMENTATION (GL-92-04)
/
819.
Gt-92-04 s121 LIMERICK 2 M84288 BWR LEVEL INSTRUMENTATION (GL-92-06)
/
O 820.
GL-92-04 B121 MILLSTONE 1 M84289 BWR LEVEL INSTRtmENTATION (GL-92-06) 12/92 e
h 821.
GL-92 06 8121. MONTICELLO M64290 BWR LEVEL INSTRUMENTATION (GL-92-04)
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GL 92-04 8121 NINE MILE POINT 1 M84291 SWR LEVEL INSTRLMENTATION (CL-92-04) 03/93 823.
. 8121 NINE MILE POINT 2 M64292 8WR LEVEn. INSTRUMENTATION (GL-92-04) 10/93 l
826.
GL-92-04 B121 OTSTER CREEK 1 M64293 BWR LEVEL INSTRUMENTATION (CL-92 04)
/
825.
CL-92-04 8121 ~PEACM SOTTOM 2 M64294 BWR LEVEL INSTRUMENTATION (GL-92-04)
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GL-92-04 8121 PEACM BOTTOM 3 M84295 8WR LEVEL INSTRtMENTATION (GL-92-04)
/
827.
GL-92-04 8121 PERRY 1 M64296 8WR LEVEL INSirJMENTATION (GL-92-06)
/
828.
CL-02-04 5121 PILGRIM 1 M84297 BJR LEVEL INSTRUMNT ATION (GL-92-04)
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829.
3121 QUAD CITIES 1 M64298 SWR LEVEL INSTRUMENTATION (GL-92-04)
/
830.
GL-92-04 s121 QUAD CITIES 2 M64299 BWR LEVEL INSTRUMENTATION (GL-92-04)
/
831.
GL-92-06 B121 RIVER 8END 1 M84300.. nWR LEVEL INSTRUM NTAfl0N (GL-92-06)
/
832.
GL-92-04 8121 SUSQUENANNA 1 M84301 ' BWR LEVEL l#STRUMENTATION (CL-92-04) 04/93 833.
GL-92-04 B121 'SUSQUENANNA 2 M84302 OWR LEVEL l#STRUMENTATION (GL-92-04) 04/93 834 GL-92-06 B121. VERMONT TANKEE 1 M64303. SWR LEVEL INSTRUMENTATION (GL-92-04)
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835.
GL-92 8121.- WASHINGTON NUCLEAR 2 M64304-BWR LEVEL INSTRLMENTATION (GL-92-04)
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MPA-A001 A001 CDMANCHE PEAK 1 M77939 10 Cf t 50.55 A(G) - ISI 10/92 837.
MPA A006 '
' A006 ' BROWNS FERRY 1 M08715 APPENDIX J - CONTAlteEENT LEAK TESTING 07/96 838.
npA-A004 A004 BROWNS FERAT 3 M08717 APPENDIN J - CONTAINMENT LEAK TESTING OT/94 839.
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8032 3NADDAM NECK -
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8041 BROWNS FERRT 3 M48136 FIRE PROTECTION - FIPAL TECM SPECS (INCLLDES SER SUPPLEMENTS) 01/94 f
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MPA-8116 8116 SIG RC"K POINT 1 M77764 SUPP 3, NRC SPONSORED TESTS OF MOTOR-OPERATED VALVES (GL89-10) -
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MPA-8116' 8116 COOPER STATION MT7771 SUPP 3, NRC SPONSORED TESTS OF MOTCR-OPERATED VALVES (GL89-10) 10/9?
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MPA-8116 8116 FERMI 2 M77775 SUPP 3, NRC SPONSORED TESTS OF MOTOR-OPERATED VALVES (GL89-10) 10/92 846.
MPA-8116 8116 MATCM 1 M77778 SUPP 3, NRC SPONSORED TESTS OF MOTOR-OPERATED VALVES (GL89-10) 06/93
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MPA-8116 8116 NATCM 2 MTT779 SUPP 3, NRC SPONSORED TESTS OF MOTOR-OPERATED VALVES (GL89-10) 12/92 848.
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MPA-8116 5116 M!LLSTONE 1 M77785 SUPP 3, NRC SPONSORED TESTS OF MOTOR-OPERATED VALVES (GL89-10) 12/92 i
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MPA-8116 8116 OfSTER CREEK 1 M77789 SUPP 3, NRC SPONSORED TESTS OF MOTOR-OPERATED VALVES (GLB9-10) -
06/94 852.
MPA-8116 8116 VERMONT TANKEE 1 M77800 SUPP 3, WRC SPONSORED TESTS OF MOTOR-OPERATED VALVES (GL89-10) 06/94 I
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MPA-8117 8117 ARKANSAS 1 M81588 SUPP 2 - FAILURE OF WESTINGMOUSE SG TU8E MECMANICAL PLUGS -
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MPA-8117 8117 ARKANSAS 2 M81589 ' SUPP 2 - FAILURE OF WESTINGMOUSE SG TU8E MECMANICAL PtuGS 10/92 g
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MPA 8117 8117 CALLAWAT 1 M61598 SUPP 2 - FAILURE OF WESTINGNOUSE SC TUBE MECMANICAL PLUGS
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MPA-8117' 8117 COOK 1 M81607 SUPP 2 - FAILURE OF WE$ilNGMOUSE SG TUBE MECMANICAL PLUGS 12/94 857.:
MPA-8117 8117' CRTSTAL RIVER 3 M81609 SUPP 2
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MPA-8117 8117-GINh4 M81622. SUPP 2 - FAILURE OF WESilNGNOUSE SG TUBE MECMANICAL PtuGS 12/98
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MPA+8117 8117 MILLSTONE 2 M81635 $UPP 2 - FAILURE OF WESTINGHOUSE SG TU8E MECMANICAL PLUGS
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MPA-8117 8117 MILLSTONE 3 M81636 $UPP 2 - FAILURE OF WESTINGNOUSE SG TU8E MECMANICAL PLUGS 12/96 861.
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MPA-8117 8117 SEA 8R00K 1 M61658 SUPP 2'- FAILURE OF WESf!NGMOUSE SG TUBE MECMANICAL PLUGS 12/95 865.
MPA-8117 8117 V0GTLE 1 M81673 SUPP 2 -. FAILURE OF WCSilNGMOUSE SG TUBE MECMANICAL PLUGS 05/93 866., MPA-8117 8117 V0GYLE 2 M81674 SUPP 2 - FAILUPE OF WESilNGNOUSE SG TU8E MECMANICAL PLUGS 08/93 l
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8118 ARKANSAS 1 M83588 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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MPA-8118 8118 SEAVER VALLET 2 -
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MPA-8118 8118 OIG ROLE POINY'1 M83592 IPE EXTERNAL EVENTS (GL88-20, SUPP 4).
05/95 874.
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'8118 8RAIDWOOD 1 M83593 IPE EXTERNAL EVENTS (CL88-20, SUPP 4)
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MPA-8118 8118. 8EAIDWOOD 2 M83594 IPE EXTERNAL EVENTS (GL88-20, SUPP A)
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MPA-8118 8118 BROWNS FERRY 1 M83595 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 07/96 877.
MPA 8118 8118 BROWNS FEter 2 M83596 IPE EXTERNAL EVENTS (CL88-20, SUPP 4) 07/96 8 78.
MPA-8118 8118 BROWNS FERRY 3 M83597 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 07/96 879.
MPA 8118 8118 BRUNSWICK 1 M83598 IPE EX1ERNAL EVENTS (GL88-20, SUPP 4)
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881.
MPA-8118 8118 SYRON 1 M83600 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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882.
MPA-8118 8118 STRON 2 M83601 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
/
883.
MPA-8118 8118 CALLAWAY 1 M83602 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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854.
MPA-8118 -
8118 CALVERT CLIFFS 1 M83603 IPE EXTERuAL EVENTS (GL88-20, SUPP 4) 12/95 i
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MPA-8118 8118 CALVERT CLIFFS 2 M83604 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 12/95 886.
MPA 8118 8118 CATAWBA 1 M8360$ IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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MPA-8118 8118 CATAWEA 2 M83606 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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888.
MPA 8118 8118 CLINTON 1 M83607 IPE EXTERNAL EVENTS (GLS8-20, SUPP 4)
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889.
MPA-8118 8118 COMANCHE PEAK 1 M83608 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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MPA 8118 8118 COOPER STAf f 0N M83611 IPE EXTERNAL EVENTS (Gt88-20, SUPP 4)
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MPA 8118 8118 CRYSTAL RIVER 3 M83612 IPE EXTERNAL EVENTS (GL88 20, SUPP 4)
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MPA-8118 8118-DAVIS 8 ESSE 1 M83613 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 09/9$
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MPA-8118 8118 DIABLO CANTON 1 M83614 IPE EXTERhAt EVENTS (GL88-20, SUPP 43 01/96 876.
MPA 8118 8118 DIABLO CANTON 2 M8361$ IPE EXfERNAL EVENTS (Gt88-20, SUPP 4) 01/95 897.
MPA-8118 8118 DRESDEN 2 M83616 IPE EXTERNAL EVENTS (GL88-20,1 SUPP 4)
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MPA-8118 8118 DRESDEN 3 M83617 IPE EXTERNAL EVENTS (GL88-23, SUPP 4)
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MPA-8118 8118 DUANE ARNOLD M83618 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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MPA-8118 8118 FARLET 1 M83619 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 06/95 901 MPA-8118 8118 FARLE' 2 M83620 IPE EXTERNAL EVENTS (GL88 20, SUPP 4) 06/95 902. 'MPA-8118 8118 FERMI 2 M83621 IPE EXTERNAL EVENTS TGL88-20, SUPP 4) 06/95 903.
MPA-8118 8118 FITZPATRICK M83622 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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MPA-8118 8118 GINNA M83624 IPE EXTERNAL EVENTS (GL88-20, SUPP 4)
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MPA-8118 8118 GRAND WLF 1 M83625 lPE EXTERNAL EVENTS (GL88-20, SUPP 4) 12/95 907.
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MPA-8118 8118 ftARRIS 1 M83627 IPE EXTERNAL EVENTS (GL88-20, SLFP 4) 06/95 909.
MPA-8118 8118 NATCN 1 MS3628 IPE EXTEtNAL EVENTS (GL88-20, SUPP 4)
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MPA-8118 8118, NATCH 2 M83629 IPE EXTERNAL EVENTS (GL88-20, SUPP 4) 06/95
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NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION
- 1. REPORr NUMDER h
3v ar Num-C 1102, 320s,3202 BIBUOGRAPHIC DATA SHEET t* *. 81 say-)
(see instructions on ine r.versei NUREG-1435 Supplement 2
- 2. mtE ANo suomtE
- 3. oATe REPORT emiSsEo Status of Safety issues at Licensed Power Plants MONTH l
YEAR i
TMl Action Plan Requirements December 1992 Unresolved Safety Issues
- 6. AulHOH(S)
- 6. TYPE OF AEPORT Annual
- 7. PERICO COVERED (inclusive Dates) 10/1/91 - 9/30/92
- 8. PERFOHMING ORGANIZATION - NAME AND ADORESS (if NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and ma6 ting address; if contractor, provide name and malling address.)
Program Management, Policy Development and Analysis Staff Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
- 9. SPONSORING,r4GANt2ATION - NAME AND ADORESS (It NRC, type "Same as above* ; if contractor, provide NRC Div6sion. Office or Region, U.S. Nuclear Regulatory Commission, and maliing address;)
Same as above
- 10. SUPPt.EMENTARY NOTES
- 11. ABSTRACT (200 words or less)
As part of ongoing U.S. Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG report will be published on the status of licensee im-piementation and NRC verification of safety issues in major NRC requirement areas. This information was compiled and re-ported in three NUREG volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (FMI) 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). Supplement 1, published in December 1991 combined these volumes into a single report and provides updated information as of September 30, 1991. This annual NUREG report pmvides updated information on FMI, US1, and GS! issues and includes status of'all Other Multiplant Actions (MPAs).
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 Staffin the Office of Nuclear Reactor Regulation and by NRC regional person-
- net, nis report is to provide a comprehensive description of the implementation and verification status of TMI Action Plan Require-ments, 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 1 ublic. An additional purpose of this NUREG report is to serve as a follow-on to NU REG-0933, "A Prioritization of Generic St.fety 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.
13 AVAILABlUTY STATEMENT
- 12. KEY WOROS/DESCRIPTORS (Ust words or phrases that will assist researchers in locating the report.)
Unlimited
- 14. SECURITY CLASSIFICATION Status of Safety issues at Licensed Power Plants (Th" rase)
TMI Action Plan Requirements Unclassified j
Unresolved Safety issues (inia Report)
Generic Safety issues Unclassified Other Multiplant Action Issues is. NuMeeR Or eAoES
- 16. PRICE NRC FORA /l 335 (2-8g)
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Federal Recycling Program
UNITED STATES seEcet rounm CLOSS RATE POSTAGE CND FEES C*oD NUCLEAR REGULATORY COMMISSION
- USNac WASHINGTOM, D.C. 20555-0001 PERMIT 9J0. G-67 CFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 IANICDIC01C91 1
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