ML20205S374
| ML20205S374 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1988 |
| From: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-1272, NUREG-1272-V02-N01, NUREG-1272-V2-N1, NUDOCS 8811110038 | |
| Download: ML20205S374 (216) | |
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- j NUREG-1272 )
Vol. 2, No.1 Repod to the ; U.S. Nuclear Regulatory Commission on Analysis and Evaluation of i Operational Data 1987 Power Reactors .
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i i U.S. Nuclear Regulatory ! Commission , Office for Analysis and Evaluation of Operational Data ! I pa esco {
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NOTICE Availab.:'ty of Reference Materials Cited in NRC Publications Most documents cited in NRC ; ublications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Docurrents, U.S Government Printing Of fice, Post Office Box 37082, Washington, DC 20013 7082
- 3. The National Technical information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Docu ment Room include NRC corresoondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspect i on and investigation notices: Licensee Event Reports; vendor reports and correspor dence; Commission papers; and applicant and licensee documen's and correspondence. The following documents in the NUREG series are available for pu :hase from the GPO Sales Program: formal NRC staff and contractor reports, NRC sponsored conference procexlings, and NRC booklets and brochures. Also available are Regulatory Guides, N RC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issvances. Documents available from the National Technical Information Service include NUREG series reports and tect.nical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries. Documents such ss theses, dissertations, foreign reports and translations, and non NRC conference proceedings are available for purchase from the organization sponsoring the publication cited. Sin;pe copies of NRC draf t ..,) orts are available free to the extent of supply, upon written request to the Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory Commissior . Washington, DC 20555. Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Asenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if tney are American National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10018. l l [
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!o s - UNITED STATES j'*,j e
NUCLEAR REGULATORY COMMISSION Washington, D.C. 205$5 )
%.,*...*/ October 1988 I
ERR ATA SHEET Report Number: NUREG-1272 Repert
Title:
Report.to the U.S. Nuclear Regulatory Comission on Analysis and Evaluation of Operational Data - 1986 Prepared by: Office for Analysis and Evaluation of Operational Data ! 18C Date Published: May 1987 Instructions: Picase change NUREG-1272 to NUREG-1272, Vol. 1, as follows on the cover, title page and Fom 335. NUREG-1272 Vol. 1 I i
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i i 1 Distslon of Freedom of Information and Publications Feevices Office of Administration and Rr,ources Managenneet ; i
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1 NUREG-1272 AEOD/S804 Vol. 2, No.1 1 Reaort to the U.S. Nuclear Regulatory Commission on Analysis and Evaluation of Operational Data 1987 Power Reactors ate $uIsh $ctobr 1N Offico for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Cornmission
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l l l ABSTRACT I This annual report of the U.S. Nuclear Regulatory Commission's Office for Anal-ysis and Evaluation of Operational Data (AE00) is devoted to the activities ' performed during 1987. The report is published in two volumes. NUREG-1272, Vol. 2, No. 1, covers Power Reactors and presents an overview of the operating experience of the nuclear power industry, with comments regarding the tre:1ds of some key performance measures. The report also includes the principal fi1 dings and issues identified in AE00 studies over the past year, and summarizes Infor-nation from Licensee Event Reports, the NRC's Operations Center, and Diagnostic l Evaluations. NUREG-1272, Vol. 2, No. 2, covers Nonreactors and presents a re- . view of the nonreactor events and micadministration reports that were reported ! in 1987 and a brief synopsis of AE00 studies pubitsbed in 1987. Each volume contains a list of the AE00 Reports issued for 1980-1987. 1 1 i I l i i f
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EXECUTIVE SUMARY l AE00 reviews and evaluates operating experience of U.S. and foreign commercial nuclear power reactors to identify safety significant events and concerns, i their root causes, the significance of the trends and patterns displayed by ' the events, the corrective actions taken to address the cont.arns, and the generic applicability of these events and concerns. From 1984 to 1987 there was an appreciable increase in the number of licensed commercial power reactors; 28 commercial nuclear power plants with a total capacity of more than 30,800 Mwe were licensed. In 1987, five new units were licensed for low power and eight units were licensed for full power operation. [ By the end of 1987, there were 109 licensed commercial nuclear power plants in ' the U.S. AE00 compared the 1987 industry average data on reactor trips, safety system ! actuations and failures and other operating data with those of previous years ' to identify trends of industry performances. From 1984 to 1987, some key per- , formance measures improved. The number of significant events at operating l plants dropped sharply from an average of about 2.4 events per plant in 1985 , to 1.6 in 1986 and 0.8 in 1987. The average number of unplanned automatic l reactor trips per year decreased from 5.2 to 3.2 per reactor. The average ! number of demands of the erergency core cooling systems and of the emergency diesel generators decreased from 2.8 to 1.7. The frequency of plant shutdowns ' required by plant Technical Specifications remained low. The number of risk significant events, as measured by the AEOD Accident Sequence Precursor Method-ology, decreased (improved from 17 in 1984 to 6 in 1986). These trends suggest 1 that overall nuclear plant operational safety is improving. ' Emphasis by the regulatory community and the plant operators, including the Institute of Nuclear Power Operations, on the reduction of unplanned challenges to safety systems have encouraged various industry improvement programs. It appears that these initiatives are having a beneficial effect. For example, } significant reactor trip reductions have been achieved at mature plants with , Westinghouse reactor designs. The improvements were due to reductions in the j frequency of feedwater system and electrical system initiated rear. tor trips as : the result of a reduction in the frequency of equipment failures. , t However, many plant specific problems continue to exist. AE00 stuoies (dis- l cussed in Section 3) have identified several areas, including air systems, i decay heat removal, motor-operated valves, and human performance where safety ' and plant performance could be substantially improved. Significant events that reveal problems associated with plant design, modification, and operation con-tinue to occur and plants requiring additional regulatory attention continue to i be identified. Industry forced outage rates and equipment forced outages are r not improving. In addition, failure trends need to be monitored and appropri-ate maintenance programs implemented, t In 1987, AE00 performed a study of the first two years of operation of commer-cial nuclear power reactors, "Operational Experience at Newly Licensed Plants," l NUREG 1275, Vol. 1 (Section 3.1 of this report). The study focused on the i startup experience of 22 plants for which operating licenses were issued be- i tween 1983 and 1986 and analyzed data drawn from operational reports and site l l l l V
i visits to 11 of the plants. The analyses showed that, without early corrective action, the root causes of events that occurred during startup would likely persist during early commercial operation. The study concluded that it was possible to achieve significant improvements in learning curves and early plant performance and made suggestions concerning plant management, personnel train-ing, plant equipment, and some NRC practices. Approximately 3000 licensee event reports (LERs) covering reportable events and conditions were reviewed in 1987. While the overall industry trends were improving, there were several plant specific events that were significant and that revealed deficiencies in operations, maintenance, and design. Examples include a steam generator tube rupture at North Anna, extended loss of offsite power at Palisades and Pilgrim, and some noteworthy cases of personnel error. Thirteen Augmented Team Inspections were conducted in 1987 (Section 4). Two Diagnostic Evaluation Teams were established in 1987 and performed evaluations at the Dresden and McGuire facilities (Section 5). Personnel errors continued to contribute a large fraction of the reported events. Although most of these errors were not safety significant by themselves, they can cause inadvertent reactor trips, actuation of engineered safety features, and in some cases challenges to safety systems. An example of a personnel error was the failure to detect the loss of 20 percent of the primary system coolant inventory during a three-day period at the North Anna Station. This event was the only ECCS actuation in response to a loss of primary coolant in 1987. (See AE0D Report T707, Section 3.20, for a description of this event.) Several AE00 technical review studies performed in 1987 dealt with human factor considerations. For example, "Hispositioning of ' Reverse Acting' Valve Controllers Causing Safety System Inoperability," T713 (Section 3.24), dealt with mispositioning of a controller at Susquehanna 2 due in part to a lack of labeling and inadequate procedures; "Occurrence of Events Involving Wrong Unit / Wrong Train / Wrong Component," T705 (Section 3.18), focused on weaknesses in licensee programs and practices; and "Unplanned Criticality Events at U.S. Power Reactors Similar to that at Oskarshamn Unit 3 on 07/30/87," T712 (Section
. 3.23), examined unplanned criticality events at U.S. plants while a principal f
safety system was inoperable. AE00 found that human factors involving deficient procedures, deviation from procedures, a lack of adequate procedures, or inadequate training continued to be the cause of problems. "Loss of Decay Heat Removal Function at Pressurized j Water Reactors with Partially Orained Reactor Coolan+ Systems," 5702 (Section l 3.3), recommended corrective measures in planning, coordination, procedures, and personnel training during shutdown. "Discharge of Primary Coolant Outside of Containment of PWRs While on RHR Cooling," E704 (Section 3.7), recomended increased attention to persorel and procedural deficiencies. "Inadequate Mechanical Bloc ing of Valves," E706 (Section 3.9), suggested, among other l things, that licensees should ensure clear and verifiable procedures for restoring blocked valves to service (Information Notice 87-38). l i Events involving previously identified and new potential common mode failure l continue to be studied. "Air Systems Problems at U.S. Light Water Reactors," NUREG-1275, Vol. 2 (Section 3.2), recommended additional industry and vi
l l regulatory actions co ensure that air systems are maintained and operated at functionally ade.auate levels. "Auxiliary Feedwater Pump Trips Caused by Low Suction Pressute," E709 (Section 3.12), resulted in issuance of Information Notice 87-53 and review of the adequacy of the safety design basis for the AFW system in the ongoing GI-124 program for assuring AFW system reliability and availability. "Inadequate Net Positive Suction Head in Low Pressure Safety Systems in PWRs," E710 (Section 3.13), resulted in issuance of Information Notice 87-63 advising licensees to establish strong root cause programs for taking the proper corrective action. 1 Equipment failures, actual and incipient, continue to be reported. As a result, programmatic initiatives were established by many licensees in areas such as motor-operated valve (MOV) operability, equipment qualification veri-fication, and fire protection upgrades. AE00 reports associated with these problems included "MOV Failure Oue to Hydraulic Lockup from Excessive Grease in Spring Pack," E702 (Section 3.5). This report recommended industry effort to identify conditions that result in hydraulic lockup, to develop a solution for motor operators currently in use, and to disseminate the appropriate corrective ] action. NRC staff is working with industry to resolve this problem. Design, fabrication and installation errors are the causes for approximately half of the LERs. Many of these deficiencies were identified by an adverse event that challenged the particular component or system, or by a special in-I vestigation or design review. "Design and Construction Problems at Operating Nuclear Plants," E707 (Section 3.10), suggested that licensees be alerted of the need to review the adequacy of the plant's QA and QC programs used in plant i modifications. "Problems with High Pressure Safety Injection Systems in 1 Westinghouse PWRs," T708 (Section 3.21), suggested that the HPSI system , unavailability receive further consideration, f Additional studies completed by AE00 in 1987 addressed diverse issues. For d example: "Feedwater Regulating and Bypass Valves - NPROS Prototype Study," l P702 (Section 3.25), found that variations in MFW control and bypass valve i failure rates appeared tv be due to differences in operational philosophies and maintenance practices among plants. "RWCU System Automatic Isolation and Safety Considerations," E705 (Section 3.8), documented ways for plant operators to reduce the number of RWCU system isolations caused by operational and design problems in the leakage detection system. "Depressurization of Reactor Coolant Systems in PWRs," E708 (Section 3.11), highlighted areas of safety analyses and plant operation that needed improvement and suggested correction of deficien-cies in emergency procedures. "Compression Fitting Failures," T701 (Section 3.14), concluded that the failures were adequately iddressed by NRC and indus-try. "Leaking Pulsation Dampener Leads to Loss of Charging System," T702 (Section 3.15), determined that the only safety-related system that could be affected by a similar failure of the pulsation dampeners was at San Onofre 2 and 3 and informed the licensee. "Potential for Loss of Emergency Feedwater Pump Due to Pump Runout During Certain Transients," T703 (Section 3.16), found ! that the concern regarding the loss of EFW pump runout for other plants was adequately addressed in IE Bulletin 80-004 and related industry actions.
"Pressurizer Code Safety Valve Reliability," T704 (Section 3.17), initiated a detailed study by AEOD to determine the extent of the problem and assess the adequacy of present efforts toward increasing safety vaive reliability. "Recent Events Involving Turbine Runbacks at PWRs," T706 (Section 3.19), found I
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limited safety significance associated with events involving turbine runback and suggested no further review. "Loss of Offsite Power Due to Unneeded Actua-tion of Startup Transformer Protective Differential Relay," E703 (Section 3.6), suggested that this event be considered in ongoing licensing reviews and that the lessons learned be incorporated into the next revision of the Standard Review Plan. ) viii
r _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ TABLE OF CONTENTS
.P_a21 EXECUTIVE SUMARY ...................................................... v
1.0 INTRODUCTION
...................................................... 1 2.0 NUCLEAR POWER PLANT OPERATIONAL EXPERIENCE ........................ 3 2.1 Industry Operational Performance ............................. 3 2.1.1 Significant Events .................................... 3 2.1.2 Reactor Trips ...............................'.......... 5 2.1.3 Actuations of Engineered Safety Features .............. 8 2.1.4 Safety System Failure Events .......................... 12 2.1.5 Forced Outage Rates and Equipment Forced Outages ...... 15 2.1.6 Collective Radiation Exposure ......................... 15 2.1.7 Technical Specifications Related Events ............... 21 2.1.8 Accident Sequence Precursor Program ................... 21 2.1.9 Foreign Events .............................m. ......... 23 2.2 Analysis of the 1987 Licensee Event Reporting ................ 23 2.3 Abnormal Occurrence (AO) Reporting ........................... 29 2.4 Data from Operations Center in 1987 .......................... 30 3.0 RESVLTS OF AE00 STUDIES ........................................... 35 3.1 Operatirnal Experiences at Newly Licensed Nuclear Power Plants, NUREG-1275 (Vol. 1) ............................ 35 3.2 Air Systems Problems at U.S. Light Water Reactors, NUREG-1275 (Vol. 2) .......................................... 38 3.3 Loss of Decay Heat Removal Functicn at Pressurized Water Reactors with Partially Drained Reactor Coolant Systems, S702 ......................................................... 39 3.4 Potential Containment Airlock Window Failure Due to Radiation AE00/E701 ......................................... 40
- 3. 5 MOV Failure Oue to Hydraulic Lockup from Fxcessive Grease in Spring Pack, AE00/E702 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 3.6 Loss of Offsite Power Due to Unneeded Actuation of Startup Transformer Protective Differential Relay, AE00/Ev~ . 42 3.7 Discharge of Primary Coolant Outside of Containment at PWRs While on RHR Cooling, AE00/E704 ...................... 42 3.8 RWCU System Automatic Isolation and Safety Considerations, AE00/E705 .................................................... 43 3.9 Inadequate Mechanical Blocking of Valves. AE00/E706 .......... 44 3.10 Design and Construction Problems at Operating Nuclear Plants, AE00/E707 ............ ............................... 45 3.11 Depressurization of Reactor Coolant Systems at PWRs, AE00/E708 .................................................... 45 3.12 Auxiliary Feedwater Pump Trips Caused by low Suction Pressure AE00/E709 .......................................... 46 ix
l TA8LE OF CONTENTS (CONTINUED) i P, age ; s 3.13 Inadequate NPSH in Low Pressure Safety Systems in PWRs, i' AE00/E710 .................................................... 47 3.14 Compression Fitting Failures, AE00/T701 ...................... 47 l j 3.15 Leaking Pulsation Dampeners Lead to Loss of Charging ; System AE00/T702 ............................................ 48 ! 3.16 Potentia 1forlossofEmergencyFeedwaterDuetoPump Runout During Certain Transients, AE00/T703 .................. 48 l 3.17 Pressurizer Code Safety Valve Reliability, AE00/T704 ......... 48 : 3.18 Occurrence of Events Involving the Wrong Unit / Wrong Train / Wrong Component, AE00/T705 ................................... 49 ! 3.19 Recent Events Involving Turbine Runbacks at PWRs, AE00/T706 .. 49 ; 3.20 Undetected Loss of Reactor Water, AE00/T707 .................. 50 r 3.21ProblemsWithHighPressureSafetyInjectionSystems I in Westinghouse M s, AE00/T708 .............................. 50 l 3.22 Heating, Ventilation and Air Conditioning System l Problems, AE00/T710 .......................................... 50 t
, 3.23 Unplanned Criticality Events at U.S. Power Reactors l 4
Similar to That at Oskarshamn Unit 3 on July 30, 1987, ' i AEOD/T712 .................................................... 51 3.24 Mispositioning of "Reverse Acting" Valve Controllers - Causing Safety System Ino)erability, AE00/T713 ............... 51 ! 3,25 Feedwater Regulating and lypass Valves, NPRDS Prototype L J Study AE00/P701 .............................................. 52 4 3.26 Ongoing Studies .............................................. 54 l i 4.0 INCIDENT INVESTIGATION PROGRAM .................................... 55 l
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5.0 DIAGNOSTIC EVALUATION PROGRAM ..................................... 56 [ .I f APPENDICES l I Appendix A - Data on Plant Operational Experience ................. A 't Appendix 8 - Summary of 1987 Abnormal Occurrences ................. B' 1 Appendix C - Listing of AE00 Reports 1987 ......................... C-1 i Appendix D - Listings of AEOD Reports 1980-1986 ................... D-1 i Appendix E - Status of AEOD Recommendations .............. ........ E-1 ( i l x l
TABLE OF CONTENTS (CONTINUED) List of Figures
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Figure 1 - Industry Trends in Operating Experience /Significant Events ... 4 1 Figure 2 - Industry Trends in Operating Experience / Automatic Reactor , Trips......................................................... 6 ; Figure 3 - Industry Trends in Operating Experience / Automatic Reactor l Trips Below 15% Power......................................... 7 , Figure 4 - Industry versus Mature Plant unplanned Reactor Trip Rate ..... 9 : Figure 5 - Industry Trends in Operating Experience /ECCS and Emergency Diesel Actuations ............................................ 11 ; Figure 6 - Rates of ESF Actuations ...................................... 13 i Figure 7 - Industry Trends in Operating Experience / Safety System ' Failures ..................................................... 14 ( Figure 8 - Industry Trends in Operating Experience / Forced Outage Rate (%)...................................................... 16 Figure 9 - Industry Trends in Operating Experience /All Plants l 1984 to 1987 ................................................. 18 Figure 10 - Industry Trends in Operating Experience / Excluding Plants ! Shutdown 1984 to 1987 ........................................ 19 Figure 11 - Industry Trends in Operating Experience / Radiation Exposure.... 20 Figure 12 - Technical Specification-Related Events ....................... 22 Figure 13 - Accident Sequence Precursor Program Results ................ . 24 Figure 14 - LERs Submitted in 1987 ....................................... 28 Figure 15 - U.S. Nuclear Power Plants Abnormal Occurrences vs. Year ...... 31 Figure 16 - U.S. Nuclear Power Plants Abnormal Occurrences / Plant vs. Year ......................................................... 32 List of Tables Table 1 - U.S. Wuclear Power Plants Not Operating During Port.c N of Period 1984-1987 ............................................. 17 Table 2 - 1986 Significant Precursor Events ............................ 25 Table 3 - 1987 Significant Precursor Events ............................ 26 Table 4 - LERs Submitted by Year ....................................... 27 Table 5 - LER Reporting by 50.73 Requirement ........................... 29 Tele 6 - Average Number of LERs by Major NSSS Vendors in 1987 ......... 29 Table 7 - Classification of Events Under Emergency Plans 1984-1987 ..... 33 xi
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1.0 INTRODUCTION
Since its formation in 1979, the Office for Analysis and Evaluation of Opera-tinnal Data (AE00) provides, as one of its primary roles, a strong, independent et,pability for the analysis of operational data. This role was strengthened j and expanded in 1987 in accordance with the Commission's emphasis on opera-i tional safety matters. The Office continues to serve as the focal point for l the independent assessment of operational events, and manages the review,
- ! analysis, and evaluation of reactor plant safety performance. Since May 1987,
] it is also responsible for the NRC's Incident Response Program, Diagnostic : l Evaluation Program, Technical Training Center, and the Incident Investigation l' Program (IIP). Additionally, AE00 provides support for the Committee to
- Review Generic Requirements (CRGR). ,
The Office consists of two divisions: the Division of Operational Assessment which includes the Incident Response Branch, the Diagnostic Evaluation and Incident Investigation Branch, and the Technical Training Center; and the Division of Safety Programs which includes the Reactor Operations Analysis Branch, the Trends and Patterns Analysis Branch, and the Nonreactor Assessment Staff. AE00 reports directly to the Executive Director for Operations (E00).
- AE00's activities involve the review and evaluation of operating experience in f order to identify
- significant events and the associated safety concerns and root causes; the trends and patterns displayed by these events; the adequacy of 7 corrective actions taken to address the concerns; and generic applicability of these events and concerns. In performing these activities, AEOD's specific 1 functions include:
t t l Analysis of operational safety data associated with all NRC-licensed
- activities and identification of safety issues which require NRC staff i actions. >
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! Development and implementation of the agency program on reactor 1 performance indicators for use by Regional and Headquarters management.
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- Development of the NRC program for diagnostic evaluations of licensee ;
J performance and direction of the diagnostic evaluation teams. q
- Development of policy, program requirements, and procedures for NRC :
incident insestigations of significant operational events, j l i
- Identification of needed operational data to support safety analysis f
- activities, and development of agency-wide operational data reporting, and !
. retrieval methods and systems. ; s I
' Oevelopment of a coordinated system for feedback of operational safety l
information to NRC offices, licensees, and other organizations, as . appropriate, and preparation of the Abnormal Occurrence Report to Congress. l l l
- Development in consultation with other NRC offices of the NRC policy for .
response to incidents and emergencies, and assessment of the NRC response l i capabilities and performance. ; ) I ! i :
1 l Development of an agency-wide technical qualification program for a broad rance of technical positions within the NRC staff, and providing for technical training needed by NRC personnel through operation of the NRC's Technical Training Center at Chattanooga, Tennessee. Continuous manning of the NRC Optrations Center to screen reactor and nonreactor events and other information reported to the Operations Center to assure the proper NRC reaction to reported events. Acting as a focal point for coordination of generic operational safety information and data systems with industry, foreign governments, and other agencies involved with the collection, analysis, and feedback of operational data. The 1987 AE00 Annual Report is composed of two parts: Power Reactors and Nonreactors. This part on Power Reactors presents an overview of the opera-tional experience of the nuclear power industry, with comments regarding the trends of some key performance measures. The report also includes the principal findings and issues identified in AE00 studies over the past year and summarizes information from licensee event reports, the NRC's Operations Center and Diagnostic Evaluations. The report includes appendices as follows: AppendIxAcontainsdatatosupportthesectionontheoperationalexperience; Appendix B lists and summarizes 1987 Abnormal Occurrences; Appendix C lists AE00 reports issued in 1987; and Appendix D lists AE00 reports issued from 1980 through 1986. Appendix E presents the status of outstanding recommendations included in AE00 studies. l 2
- 2. 0 NUCLEAR POWER PLANT OPERATIONAL EXPERIENCE This section is intended to provide a broad overview of operating experience and characteristics of the nuclear industry and individual facilities. The data are derived in large measure from a review of Licensee Event Reports (LERs) by NRC Office for Analysis and Evaluation of Operational Data (AE00).
On January 1, 1984, NRC implemented a new licensee event report rule to establish a uniform threshold for event reporting. The threshold included consideration of infrequent safety significant events as well as the more frequent events of lesser significance which are more amenable to statistical analysis and trending. Since that time, the events that met the threshold have provided a consistent basis to assess the performance trends of the industry as a whole and those of individual licensees. These events are reviewed individually, on a plant or unit specific basis, and collectively by AE00 as part of its engineering assessment, performance indicator, and trends and patterns review responsibilities. Aggregated data for the industry provide the basis for the industry trends presented in section 2.1. Data aggregated by component or system provide the basis for the studies and evaluations summarized in section 3. 2.1. Industry Operational Performance AE00 monitors operating experience to discern trends and patterns of events and their causes. Data are routinely aggregated for (a) industry as a whole, (b) t vendor, (c) system, (d) age of plant, and (e) a plant specific basis Plant i specific data are provided in quarterly Performance Indicator reports. The seven performance indicators currently monitored by NRC for all operating
; plants are: automatic reactor trips while critical, safety system actuations, significant events, safety system failures, forced outage rate, equipment forced outages per 1000 critical hours, and collective radiation exposure. The definitions of the seven performance indicators used in the program are in Appendix A-1. Industry average data are discussed below in terms of these indicators. Observations on trends in technical spe:ification violations and accident precursors are also presented below to provide a more complete picture.
From 1984 to 1987 there was an appreciable increase in the number of licensed commercial power reactors in the U.S.; 28 commercial nuclear power plants with a total capacity of more than 30,800 Hwe were licensed. In 1987, five new units were licensed for low power and eight units were licensed for full power i operation. By the end of 1987, there were 109 licensed commercial nuclear power plants in the U.S. 2.1.1 Significant Events Significant events are associated with degradation in safety equipment, unexpected plant response to transients, degradation of primary coolant t
- boundary, reactor trips with complications, and operations outside the plant Technical Specifications limits. Degradation of fuel integrity or unplanned releases of radicactivity could also be classified as significant, should they occur. As shown in Figure 1, the number of significant events at operating 3
Figure 1 Industry Trends in Operating Experience All Plants 1984 to 1987 Significant Events 4 3-W 4 2.3s o 2-E i.s E b i 4 1-0.40 o
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1984 1985 1986 1987 Year 4 i
1 l plants dropped sharply from an average of about 2.4 per plant in 1985 to 1.6 in 1986, and to 0.8 in 1987. A list of the 1987 significant events can be found in Table 1 of Appendix A-2. 2.1.2 Reactor Trips ' Any challenge to the reactor trip system reflects less than optimum operating , practices. Some reactor tripe, are the result of relatively minor incidents; others could be severe accident precursors. Reactnr trips represent rapid , changes in plant status, increasing the likelihood of operator error or , equipment malfunction. In 1987 there were a total of 430 unplanned reactor trips
- at 98 U.S. LWRs. l The corresponding figure for 1986 was 469 reactor trips at 93 LWRs. Automatic ,
reactor trips, while the reactor is critical, show a declining trend from 460 per year (5.0 per reactor) in 1985 to 386 (4.0 per reactor) in 1986, and to 341 < (3.2 per reactor) in 1987. The annual trend in automatic reactor trips for , operating U.S. plants is illustrated in Figure 2. Additional insight can be , gained by subdividing the reactor trips and examining the trends in (a) man- , ually actuated, (b) low reactor I was complicated by failures and/ power(d) or errors level, (c) situations mature reactors, in andwhich recovery (e) new i reactors. Trends are also examined by reactor type, cause, and initiating - system. ! A total of 69 (16%) of the 430 unplanntd reactor trips in 1987 were manual, t representing an increase from 51 (11%) in 1986. This increase was mostly the ' result of contributions from one new plart that recorded 11 manual trips during its startup testing program from January, when it achieved initial criticality, to December 1987. There was a significant decline in the number of automatic reactor trips at ! low power (less than 15% power). This decline can be attributed to improve- f ments in feedwater control, preventive maintenance programs, training programs, : and a reduction in the number of personnel errors. The average number of ! automatic reactor trips per reactor at low power between 1984 and 1987 is presented in Figure 3. Unplanned reactor trips in which the recovery is complicated by additional I equipment failures or personnel errors are a matter of concern because of the l higher level of stress and demands placed upon the operating personnel and ; mitigating systems. Associated failures, defined as "component failures or ' personnel errors that did not contribute directly to the cause of the reactor , trip, but are associated with post reactor trip recovery (i.e., normally the ; failure was discovered or occurred when the component was actuated to mitigate the consequences of the reactor trip)," were experienced in 17 percent of all reactor trips above 15 percent power in 1987, a slight decrease from earlier ! years. I i
*A reactor trip is defined as an actuation of the reactor protection system, l whether automatic or manual, that results in control rod motion. Plants were f included in these statistics if they (1) were issued a full power operating licen=e and (2) accumulated critical hours for some portion of 1987. l l
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Figure 2 Industry Trends in Operating Experlence All Plants 1984 to 1987 Automatic Reactor Trips 8-8 E 6-s.04 o 4-2 1984 1985 1986 1987 Year 6 _. O
l l Figure 3 l Industry Trends in Operating Experienco i All Plants 1984 to 1987 , l l
- Automatic Reactor Trips Below 15% Power i l 1.0 l i t I
l o ' 'l 1 3e 0. 8 - ! l ! 4e ! r ,
- $ 0.6- (
i tie i t g 0.49 j o ! j 0.4- ; E
- S o.as
' 0.22 m
, e
- $ 0.2--
j g 0.15 0 t i i i 1984 1985 1986 1987 Year
- Normalized reactor trips per 1000 critical hours show the same declining rate as total trips. The 1987 rate of 0.7 reactor trips per 1000 critical hours represents an approximate 14 percent reduction from the 1986 rate of 0.8 i
, reactor trips per 1000 critical hours. The reactor trip rates for individual i j reactors in 1987 are provided in Table 2 in Appendix A-2. The initiating systems are shown in Table 3 in Appendix A-2. Table 4 in Appendix A-2 lists the initiating systems for each NSSS vendor. The average reactor trip rate for mature reactors (plants with more than two years of operation) has dropped from 0.8 teactor trips per 1000 critical hour: 1 in 1984 to 0.5 in 1987 as shown in Figure 4. In large part, this reduction 's ! i attributed to improvements in the feedwater system of Westinghouse reactors. !
, Because of the large number of Westinghouse reactors in operation, the !
l improvements in the feedwater system are reflected in the statistics for the l entire industry.
- AE00 issued a study on startup experience at newly licensed plants in 1987, 1 NUREG-1275, Vol. 1 (Section 3.1), and has also evaluated reactor trips in
- general. Newly licensed reactors tend to trip at a much higher rate than do j mature reactors. The average reactor trip rate in 1987 was 1.8 reactor trips
- per 1000 critical hours for plants operating during the first two years after receiving an operating license. This reactor trip rate average was 3.5 times the average for mature reactors. Balance of Plant (BOP) (the main turbine and main feedwater) systems dominate as the initiator of unplanned reactor trips at new plants. Equipment failures are responsible for about half of the unplanned f trips at new plants, while human errors are responsible for about 27 percent.
Equipment failures were responsible for approximately 60 percent of all l unplanned trips at mature reactors compared with approximately 30 percent due ! to error and deficiencies in proceJures. Tables 5 and 6 in Appendix A-2 list the reactor trip rates by causes and Initiating systems during the initial 24 months of operation for 1984 through i 1987. Table 7 in Appendix A-2 provides the mature reactor trip cause trend for ! the period 1984 through 1987. Table 8 in Appendix A-2 provides the same I information for each NSSS vendor. Analysis of the unplanned reactor trip rates l for individual NSSS vendors consistently show the main feedwater, reactor i protection system, electrical distribution, and main turbine as the systems l primarily involved. Unplanned reactor trip rates for the four NSS$ vendors are 4 shown in Appendix A-2. Table 9 (Westinghouse). Table 10 (General Electric), j Table 11 (Combustion Engineering), and Table 12 (Babcock and Wilcox). I 2.1.3 Actuations of Engineered Safety Features l An evaluation of operational events involving either necessary or inadvertent
- actuations of Engineered Safety Features (ESFs) provide insight into plant l performance. The number of inadvertent ESF actuations, together with the root cause(s), can provide a measure of performance of plant personnel in maintain-ing, calibrating, and testing vital plant equipment. A needed ESF actuation may be viewed as an integral test, the results of which can be compared with the performance parameters stated in plant safety analyses. Such a comparison may be used to confirm system performance of the ESF involved, or to discover 8
FIGURE 4 ! INDUSTRY VERSUS MATURE PLANT UNPLANNED REACTOR TRIP RATE 1984 - 1987
, 1.2 E
I g
~
LEGEND l $'~ M mous m o h MATURE l ; - - g .s-l !" , n_
,_ .3 .
8 ; a . l @ . : . E o_ I I. - _ 1b 1985 lb 1967 l
hidden deficiencies or undetected failures. Excessive inadvertent challenges can degrade the systems, and thereby decreate their reliability. The number of ESF actuations, together with the observed failures, may be used to quanti-tatively estimate system reliability of the ESFs involved. These estimates are useful in the quantitative discussion of reactor accident risks for operat-ing plants. Inadvertent emergency core cooling system (ECCS) actuations are challenges that can have detrimental effects such as the potential for over-pressurization or overcooling the reactor coolent system. ESF actuation data for the last three years have been found to be consistent with the general downward trend that has been observed for reactor trips. Insight can be gained by trending ESF actuations in terms of safety function, systems involved, reactor type, and whether the actuation involved reactor trips. ECCS and emergency diesel loads are among the most significant safety rystem actuations and, therefore, are trended in the Performance Indicator program. From January 1, 1985 to January 1, 1988, average unplanned actuations on a per plant basis declined as follows: ECCS by 50 percent, to one event per year; emergency power p ) by 29 percent, to about ons event per year. Other less significant W actuations, such as those involving heating, ventilation and air condit? fcg (HVAC) systems declined by 36 percent, to approximately five act.uations' per year. A large part of these improvements can be attributed to overcoming early operational problems by new plants. ESF actuations that were actual and reeded provide a perspective on the real frequency of operating conditions or events. Because of the low frequency of these events, however, no trends were identified. Additionally, because most ESF actuations do not involve the reactor protection system (reactor trip), the analysis reflicts only reported inadvertent actuations not involving the reactor protection system. The analysis, for the three-year period from 1985 to 1987 indicates that: one of 215 ECCS actuations involved a response to a loss of reactor coolant; six percent (21 of 365) of the EP actuations occurred in response to a loss of offsite power; one percent (21 of 1,883) of the HVAC actuations involveu actual atmospheric radioactivity o' hazardous material (all were minor and about half involved minor leakage of ' eine at one reactor); approximately 10 percent of the 555 reactor water ci m iup (RWCU) system actuations involved external leakage. None of the actual and needed ESF actuations resulted in any significant oublic health and safety consequences. Equipment problems accounted for about one-third of the unplanned ESF actuations, personnel error caused about 30 percent, and proccoural problems caused less than 10 percent. The remaining actuations were due to aatural phenomena, miscellaneous, or unknown ceases. The North Anna steam generator tube rupture event resulted in the only ECCS actuation in response to a loss of primary coolant in 1987. In the same year, there were 108 demands of one or moro emergency diesels due to low voltage on an emergency bus. The average of the sum of ECCS (actual and inadvertent injections) and these EP actuations (actual demands due to low voltage on emer-gency buses) has maintained a downward trend from an average of 2.8 per plant in 1984 to 1.7 per plant in 1987, as shown in Figure 5, 10 i
l i. Figure 5 Industry Trends in Operating Experience All Plants 1984 to 1987 l i ECCS and Emergency Diesel Actuations 4- , 6 l 13 ! l 3- ! 2.76 j 2.6 N U 4 2.04 l g 2-g i.a E ! 5 b 1-e : 2 0 - - - - - - - l 1984 1985 1986 1987 I Year i , 11
From January 1985 through December 1987, over 4000 ESF actuations were reported. These are summarized in Table 13 of Appendix A-2. From 1965 to 1987, the average annual rate of non-RPS ESF actuations* improved from 14 to 10 per reac-tor. The improvement in the annual rate per plant varied by reactor type as follows: W plants from 9.3 to 8.4, GE plants from 22 to 15, CE plants from 16 to 8.1 and B&W plants from 3.5 to 2.0. Plant design differences, such as those in the ESF initiation logic and actuation signals, influence the rate of ESF events among the reactor vendors and architect engineers. The ESF actuations were dominated by heating, ventilation and air conditioning 4 (HVAC) system and reactor water cleanup (RWCU) isolation, which generally do not have high safety significance. ESF actuations show a decreasing trend because of the reduction in HVAC actuations at all types of reactors, and RWCU system isolations at BWRs (Figure 6). The trend data for each type of reactor are shown in Figure b 1 thruugh A-4 in Appendix A-2. The ESF database identifies approximately 80 safety systems. Lowever, three PWR systems--ECCS, emergency power (EP) and HVAC systems-- accounted for approximately 50 percent of the reported PWR ESF actuations. Similarly, four systems--ECCS, EP, HVAC, and RWCV--accounted for approximately 60 percent of the BWR ESF actuations. A detailed summary of the ESF actuations ' by event sequence in terms of the safety function, i.e., whether a safety sys-tem was actuated, a normally operating system was isolated, or both occurred, for each type of reactor is provided in Table 14 of Appendix A-2. Table 14 also shows whether single or multiple systems were involved in the actuation. AE00 study E705 (Section 3.8) discusses RWCU system automatic isolation and safety considerationc. t Overall, the trends and patterns of ESF actuations show a decline and leveling off in the rate of challenge of these safety systems during the last year or so. The improvement during this period wat largely influenced by the learning curve of new plants. The current rate of occurrence of unneeded actuations is still high for all the systerr.s reported (about 8 non-RPS ESF actuations per plant per year). Particular attention is needed during maintenance and tesMng i activities on equipment associated with the HVAC and RWCU systems. ; 2.1.4 Safety System Failure Events Safety system failures (SSF) are evants in which there has been - or could be - a failure of a system to perform its safety function, regardless of whether the system was required to function at the time. AE00 performed several safety systen, failure evaluations, including: "Auxiliary Feedwater Pump Trip Caused by Low Suction Pressure" E709 (Section 3.12); "Inadequate NPSH in Low Pressure Safety Systems in PWRs" E710 (Section 3.13); and "Potential for loss of Emer-gency Feedwater Pump Due to Pump Runout During Certain Transients" T703 (Sec-tion 3.16). Figure 7 shows trends in SSFs. The industry trend in SSFs indicates a drop from 3.6 per plant in 1984 to 2.4 in 1985, and 2.3 in 1986. The number of failures in 1987 is s11ghtly lower than in previous years. Plant specific data for SSFs are shown in Table 15 of Appendix A-2. i "Non-RPS Efr actuations are those that do not involve actuation of the reactor protection system (RPS). l a i 12 l _ _ - - - . _ - . - - _ . - _ _
FIGURE 6 RATES OF ESF ACTUATIONS JANUARY 1985 - DECEMBER 1987 o ALL REACTORS 2 10 m E un 9- LEGEND n. F _ ALL ESF Z _.a 8- - - NON-RPS n. E - S ---- VAUD NON-RPS ua 7_ NON-RPS g a. u) \ ---- NON-RPS HVAC" z 6- A s 9 F N
% - - - - - - - NON-RPS ECCS*
4 5- - f - - - NON-RPS EP" O
< 4-. HVAC % = . - - - - NON-RPS RWCU*
- u. ~
O .-
- E m
3- RWCU
- N ' ' g.
- TOTAL
* * - RPS - Reactor Protection System E 2-V ,. . .
U z us VAUD,f #"e
~~'N' %g , / ~~~~
HVAC - Heating VentBation, and o 1-s* ECCS - Emergency Core Cootmg ( EP - - - . % E ECCS* * * * * * * ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ * * * - U. ~. .'.7 # " us o y l l l l 1 [ EP - Emergency Power System i 4 85-6 85-1f! 86-6 86-12 87-6 87-12 RWCU - Reactor Water Cleanup System PERIOD BY 6 MONTH INTERVALS
1 1 1 Figure 7 Industry Trenda in Operating Experience All Plants 1984 to 1987
)
4 Safety System Failures , 8 i l. y : 6- : ( I E I d' ! il
, 4-l j 3.56 E i s t 2 l I
2.44 ,,33 l 2- 1. 4 i i i 0- .
, j 1986 1987
,. 1984 1985 Year e l 14 l I
2.1.5 Forced Outage Rates and Equipment Forced Outages Plant forced outages as defined in Appendix A-1, are directly related to overall plant unavailability. Such forced outages are often caused by balance of plant problems. This parameter is important because equipment fa' lures are indicative of the overall maintenance of the piant, particularly the efficacy of preventive and predictive maintenance programs. Hence a monitoring of forced outages provides a perspective on overall plant performance which complements the narrower scope of safety system actuations and failures. The data for forced outage rates and equipment forced outages per 1000 critical ' hours, shown in Figure 8, are obtained from the monthly operating reports for each plant submitted in accordance with the plant Technical Specifications. When this information is used to calculate an industry average, forced outage rates increased from 11 to approximately 17 percent between 1984 and 1987. However, several plants listed in Table 1 were not operating during this period. Industry trends fc.r several categories of operating experience for all plants are shown in Figure 9. Figure 10 shows the same categories exclusive of the shutdoen plants. Although the number of shutdown plants during any quarter is small, the forced outage rate for these plants is considered to be 100 percent and significantly skews the averages. If the shutdown plants shown in Table 1 are exciuded, the industry average forced outage rate is a comparatively flat 11 percent between 1984 and 1987, as shown in Chart 5 of Figure 10. The equipment forced outages similarly show a flat trend. These flat trends are significant and indicate a need for improved preventive t.nd predictive maintenance programs in order to improve equipment dependability, especially of balance of plant. Balance of plant equipment failures dominate the forced outage rate and hence affect plant availability. 2.1.6 Collective Radiation Exposure Radiation exposure of pi nt personnel is a measure of management effectiveness. Low totc1 exposures to personnel usually result from effe:tive and conscien-tious efforts to assure: well maintained and wc11 designed equipment; experi-enced and well trained maintenance personnel; good plant water chemistry; effective decontamination and cleanup practices; good ALARA and maintenance planning programs; and an alert health physics staff ' The collective radiation exposure trend shows a decline during the last four years, indicating that better chemistry control and ALARA programs have been effective in reducing the total exposure to personnel at the plants. The overall industry performance is shown in Figure 11. Average man-rem exposure has declined from 714 man-rem in 1984 to 574 in 1985, 431 in 1986, and 425 man-rem in 1987. The collective radiation exposura for each plant for 1984 through 1987 is listed in Table 16 of Appendix A-2. Individual plants show uneven rates of increase and decrease. Steam generator inspection and maintenance in PWRs, and primary system piping repairs and replacement in BWRs were major factors in higher exposures at some plants. Better water chemistry control in the secondary system of PWRs should reduce steem generator tube ' 4 corrosion. The application of hydrogen water chemistry in BWRs should mitigate intergranular stress corrosion cracking in the primary system. These actions should further reduce personnel exposures in the future. 15
_m _ . __ i Figure 8 i Industry Trends in Operating Experience All Plants 1984 to 1987 l 4 Forced Outage Rate (%) ; 40 - ! 35 _ t i
$ I 80 -
3 ! 1 { g 25- , o ) O 3 20 - P., 17.9 i 2 16.4 i
~
13,4 11 1 < .- l 1984 1985 1986 1987 Year I l l l 16 [ t
TABLE 1 U.S. Nuclear Power Plants not Operating-During Portions of Period 1984-1987 Plant Date of Shutdown Restart Date Browns Ferry 1 03/19/85 Browns Ferry 2 09/03/85 Browns Ferry 3 03/09/85 Davis Bess 6 06/09/85 12/22/86 Ft. St. Vrain 11/08/85 04/01/87 Palisades 05/19/86 04/03/87 Peach Botton 2 03/31/87 Peach Bottom 3 03/31/87 Pilgrim 04/11/86 Rancho Seco 12/26/85 Sequoyah 1 12/20/85 Sequoyah 2 08/21/85 Shoreham and Seabrook are not included. 4 i I, h 4 17
Figura 9 Industry Trends in Operating Experience - All Plants 1984 to 1987
- 1. Automatic Trips While Critical 2. Safety Sysiem Actuations
. 2 e
H g-
!u 3-s.24 4 2.s o s.04 ,
- o 4,
3.9s 3.24 fE 2-
, se j
. z .
f2- h1"
- A i -
_ _ _ p- - , - , - , - g _1984_1985 1986 1987 1984 1985 1986 1987 Year Year
- 4. Safety System Failures
- 3. Sl9sificant Eva .t. g n $
5 5
& 3- E *-
2.38 2- j 4- , 1.s y o 24* 2.32
& 1- E 2-NA 0 -~
\ 0- 1 . . . - s ~ . ~. - ~ l 1984 1985 1986 1987 1984 1985 1986 1987 Year Year j 5. Forced Outage Rate (%) 6. Equipment Forced Outages / g 1000 Critical Hours 4_ I j 35-E 30- e h 25- 0 5 o 20- *
] g3,4 18.4 &'-
2 15- 2 2 11 4 I o.s7 1.os o.79 g 10- g. o.se
) 5-4 0- - - - -
0- - - - - 1984 1945 1986 1987 1984 1985 1986 1987 i Yeef Year 18
Figura 10 Industry Trends in Operating Experience Excluding Plants Shutdown 1984 to 1987 1 Automatic Trips While Critical 2. Safety System Actuations i o_ E E N 4- 3- 3,75 ;
) 6.24 s.22
$ i 2.02 E 4- o 2-2- l' y- -
0- . . . 1984 1985 1986 1987 1984 1985 1986 1987 Year Year
- 3. Significant Events 4. Safety System Fallures g
a 2 c 2 33 y 6- l 0 2.3s i 2- ,,,, 4- .u Y 2.3 2.21 NA 0 . s _ . . _ O s s . . 1984 105 1986 1987 1984 1985 1986 1987 : j Year Year l 1 1
- 5. Forced Outage Rate (%)
- 6. Equipment Forced Outages /
g 4 1000 Critical Hours e l 5 %0 - i a:
$ $ 3~
{ 60 - o O u. i
] 2-g 40 - g 11 11 8 '
11 10.8 o.se 0.s0 O.Ss g. l E 20-I 0- -- - - - o- - - - 1984 1985 1986 1987 1984 1985 1986 1987 Year Year 19 l l
3 Figure 11 ;,
^
Industry Trends in Operating Experience All Plants 1984 to 1987 ljl h y Radiation Exposure (Man-Rem / Year) E f, I 1000 - 1. I i 8% - j g 714 ij e e ; i D i E , j y 600 - 570 I
$ 491 'I s
e 425 400 - I 200 - j 0- { 1984 1985 1986 1987 ; Year i l i
?u !
~ - - _ - - _ - ___ - , . - - . - . . ,
2.1.7 Technical Specifications Related Events i A study of LERs concerning Technical Specifications (TS) for 1984, 1985 and 1986 indicates that the distribution of TS-related reports was dominated by TS violations (93%). A summary of completed shutdowns required by TS is shown in Table 17 of Appendix A-2. The average reported violation rate per plant rose [. from 7.4 violations in 1984 to 10.7 violations in 1985, and remained relatively constant (10.2 violations per plant) in 1986. Preliminary results for 1987 indicate a slight increase in the average number of detected TS violations per plant to 10.8. This increasing trend is distributed among three sources: (1) relatively high violation rates experienced for short periods by a small group of mature plants with repetitious events, (2) a slight increase in the average violation rate for the remaining mature plants, and (3) a clear contri-bution from newly licensed plants. Violations of the limiting condition for operation (LCO) were more numerous than violations of surveillance requirements and were dominated by four areas: radiation monitoring, fire detection, containment isolation control, and passive fire suppression (barriers). (See Table 18 in Appendix A-2.) The in-creased attention to implementation of fire protection requirements (Appendix R to 10 CFR 50) may have increased the number of violations. Problems in these ; four areas reflect deficiencies in administrative control over widely different safety features which may not be receiving the same 'ttention as more prominent line systems such as ECCS and auxiliary feedwater. h.a overall assessment of the reported TS violations is difficult, however, and the Technical Specifica-tions contest varies crnsiderably among plants. Thus, reporting is not con-sistent from all plants. TS-related reports reflect plant conditions that vary widely in safety potential. To reflect this variance, a qualitative assessment of the events in terms of low, medium and high ranking was made. The distribution of TS-related events according to ever.t type und relat've safety potential is shown in Figura 12. 2.1. 8 Accident Sequence Precursor Program The accident sequence precursor (ASP) program systematically evaluates LERs to identify and assess the more significant reactor operational (i.e., precursor) events. The measure of precursor significance is the estimated likelihood of inadequate core cooling given the occurrence of an event. It is one means of evaluating in a structured manner what might happen if one or more additional events or failures had occurred. A precursor event is any one of the following which occurs (or could be postulated to occur) at power or when needed to function:
- 1. One or more plant safety systems fails or are found failed.
- 2. The redundancy of two or more safety systems are found degraded.
l 3. An initiator occurs, such as: a) LOOP - Less of Offsite Power l l l
?1
FIGURE 12 TECHNICAL SPECIFICATION-RELATED EVENTS Distribution According to Event Type and Relative Safety Potential 1200 1100-c 1000-LEGEND 900- - h $,' 9
& LOW e 800-o 700-
$)
/
f
,5 MED m
g o 600 - g 9
) /.
'i; HIGH i
z 5"- 4 .= = 4m- 8 _.; Si, -- 300- -- -
-~
200- h 5 100-0 ; g i T Y 84 85 86 84 85 86 84 85 86 VIOLATIONS SHUTDOWN COMPLIANCES
i l b) LOCA - Loss of Cooling Accident c) SLB - Steam Line Break o) Any transient or plant upset that proceeds in an offnormal manner (e.g., plant equipment was unavailable or degraded or the event was not previously analyzed in safety evaluations) A significant precursor is a precursor having an estimated core damage conditional probability of IE-4 or higher. (Stated differently, a significant precursor is one where the core damage likelihood given the occurrence of an event is estimated to be one chance in ten thousand or greater.) The number of . highly significant precursor events determined by AE00's Accident Sequence Precursor Program for the period 1984 through 1987 exhibits a downward trend. Figure 13 shows a decrease from 17 significant precursors in 1984 to 6 in 1986. The ASP models used to estimate event significance are being changed for the 1987 evaluations to better reflect present knowledge regarding likelihood of some operator actiors given the occurrence of a station blackout event. Because of these model changes, the 1987 precursor events (type and quantity) are not directly comparable to those of prior years. The ASP program has tentatively identified 10 events in 1987 that are significant. The 1986 and 1987 signifi-cant precursor events are summarized in Tables 2 and 3 respectively. 2.1.9 Foreign Events AE0D routinely reviews reports of foreign reactar operating experience from the Incident Reporting Systems of the Organization for Economic Cooperation and Development's Nuclear Energy Agency and the International Atomic Er,ergy Agoney, as well as information obtained through bilateral exchange programs with several countries. Two issues, of importance to domestic reactors were high-lighted by these information sources in 1987: thimble tube thinning events in Westinghouse reactors, and shaft cracks in reactor coolant pumps designed by the German KSB Pump Company. The pumps are used only at the Palo Verde Plant in the U.S. 2.2 Analysis of the 1987 Licensee Event Reporting The Licensee Event Report (LER) is one of the most widely analyzed documents in the nuclear industry. It provides a means of identifying safety problems and concerns which may not be otherwise recognized or understood as significant. Table 4 shows a comparison between the number of LERs submitted in 1987 from plants licensed to operate and previous years. The LER rule (10 CFR 50.73) be- ; came effective on January 1, 1984 and revised the reportability requirements. Implementation of the revised rule resulted 11 an increase in the information provided in events of greater interest and an overall decrease in the number of reportable events. The safety significance of the LERs is assessed by engineering review of each LER and trends and patterns analysis of the content of the LERs. Variations in't.ER counts from plant to plant can result from a host of factors, only one of which is an actual difference in safety performance. Thus, the number of LERs is not a parameter of particular safety significance. 23
l Figure 13 Accident Sequence Precursor Program Results 20 - 18-17 16-14-0 12-10 10* 10-8- I g .... . .. .. 1984 1985 1986 1987 Year CD - Core Demete
- The methodology for analysle was revised in 1947 24
r j
,, I Table 2 1986 Significant Precursor Events LER Plant Date Description 247/86-017 Indian Pt. 2 05/28/86 Open turbine bypass valves (to condenser) and one high pressure injection train failure 247/86-035 Indian Pt. 2 10/20/86 Reactor trip, loss of faedwater, and two auxiliary feedwater train failures to autostart 250/86-039 Turkey Pt. 3 12/27/86 Reactor trip and stuck open pressurizer power operated relief valve 261/86-005 Robinson 2 01/28/86 Electrical bus fails with loss of offsite power and diesel generator unavailable 413/86-031 Catawba 1 06/13/86 Small loss of coolant accident forces plant trip 414/86-028 Catawba 2 06/27/86 Steam generator power operated relief valves open during test, and reactor trip with other failures l
[ I 25
Table 3 1987 Significant Precursor Events LER Plant Date Description 255/87-024 Palisades 07/14/87 Seven and one-half hour loss-of-offsite power event 285/87-025 Ft. Calhoun 1987-88 Undetected failures from water intrusion into instrument air, failing emergency diesels (LERs 87-025, 87-033, and 88-010 considered and evaluated together) 293/87-014-1 Pilgrim 11/12/87 LOOP and one DG removed from service for inspection 317/87-012-1 Calvert 07/2387 Loss of offsite power and one auxiliary Cliffs 1 feedwater train failed initial start attempts 324/87-001-2 Brunswick 2 01/05/87 Reactor trip with high pressure coolant injection and reactor core iso?ation cooling systems unavailable 331/87-009 Duane Arnold 05/27/87 Emergency power system unavailability due to incorrectly set overcurrent relays 338/87-014 North Anna 1 07/15/87 Steam generator tube rupture. Site alert. 440/87-009 Perry 1 02/27/87 Two diesel generators inoperable 455/87-019 Byron 2 10/02/87 Trip with loss of offsite power 482/87-015-1 Wolf Creek 12/xx/87 Unavailability of DGs within a 9-day period 26
i i Table 4 LERs Submitted by Year Year LERs Units LERs Per Unit 1981 4016 75 53 1982 4400 81 54 1933 4839 8A 57 1984 2435 92 (1) 26 1985 2997 97 (2) 31 1986 2818 104 (3) 27 1987 2810 109 (4) 26 (1) Palo Verde not included - licensed 12/31/84. (2) Dresden 1, Humboldt Bay, and Three Mile Island 2 are not included in the 1985 data. (3) Dresden 1, Humboldt Bay, Three Mile Island 2, Seabrook 1 and Braidwood 1 are not included in the 1986 data. (4) Dresden 1, Humboldt Bay, Three Mile Island 2, Lacrosse and Seabrook 1 are not included in the 1987 data. Analysis of the 2810 LERs submitted in 1987 indicates a broad range in the number of LERs submitted by each licensee (from 4 LERs to 78 LERs in this 1 year period). The mean number of LERs per unit is 26, while the median is
- 21. The distribution of LERs per unit is shown in Figure 14.
Analysis of the content of the LERs indicates that the largest number of reports (46 percent) was associated with ESF actuations and reactor trips. Multiple ESFs and reactor trips can be reported in one LER. There were more than twice as many ESF actuations (other than reactor trips) as reactor trips. A reactor trip was reported in about 19 percent of the 1987 LERs. The second most frequently reported type of event (38 percent) was a condition prohibited by TS or a shutdown required by the TS. Examples of reportable items in this category include a missed surveillance test or the completion of a plant shut-down because of the unavailability of required safety components. The third most frequent LER catagory concerned events that did or could have resulted in a loss of safety function at the system level. This condition was reported in 8 percent of the LERs. The percent of LERs in terms of the individual LER 50.73 report requirements is shown in Table 5. In addition to reports received pursuant to 50.73 require-ments, about 9% of the tctal LERs were reported pursuant to other requirements, such as Parts 21, 50.36, etc. 27
FIGURE 14 LERs SUBMII I t-D IN 1987 2. 18-r 16 - g 14 - - . z -
- l -
j ! {12- : u. g O 10 - '
, g : .:
m : '; :,
! 8- I
- ;}
z , - 6* f ,; I - F: ; .
] 'j ::y
- , L X
7 z;
.y ,
, . ,' l ,
s 'to ii lis io-20 2i-2s 2s'so adss as'40 41-4s 4dso sdss ss'so cdss ss'707 7 ! s 7b ds NUMBER OF LERs
Table 5 LER Reporting by 50.73 Requirement Percent of Reference Requirement 50.73 Reports
- 50.73(a)(2)(iv) RPS/ESF Actuation 46 50.73(a)(2)(1) TS Shutdown or TS Violation 38 50.73(a)(2)(v) Real or Potential Loss of a Safety System 8 50.73(a)(2)(ii) Unanalyzed Conditions 5 '
50.73(a)(2)(vii) Failures in Multiple Systems 3 50.73(a)(2)(x) Internal Threat 1 50.73(a)(2)(iii) External Threat <1 ! 50.73(a)(2)(viii)(A) Airborne Radioactive Releases 0 ! 50.73(a)(2)(viii)(B) Liquid Effluent Radioactive Releases 0
- Percentages sum to greater than 100 percent because an LER can be reported under more than one reporting requirement.
When the LERs are assessed in terms of classifications such as NSSS vendor or architect-engineer, a wide variation typically results. Specific details are provided in Table 6. Table 6 Average Number of LERs by Major NSSS Vendors in 1987 Number of Average Number of LERs NSSS Vendor Units Per Plant - B&W 8 17 CE 15 19 W 47 26 GE 37 30 It is important that the LERs contain complete information about the events , being reported. To assess the quality of the contents of LERs, AE00 entered d into a contract with the Idaho National Engineering Laboratory in 1985. There ;
< has been a general improvement in LER quality since 1985. Although plant ,
differences still exist in the quality of LER reports, AE00 terminated the ' evaluation contract at the end of FY 1987 because of budgetary constraints and d in ear'.y 1988 began to ovaluate LER quality in-house. 2.3 Abnormal Occurrence (AO) Reporting ; AE00 prepares the quarterly A0 Report to Congret,s (as well as the associated Federal Register Notices) and, af ter staff crordination, forwards them to the 29 1
-- -- y ,,, --, , ,,- ,- -e v , ec - - - -
EDO, and in turn, to the Commission for review and approval. Five quarterly A0 Reports to Congress were issued during CY 1987 (second, third, and fourth quarters 1986, and first and second quarters 1987). The 5 reports described 34 A0s at NRC licensees and 6 A0s at Agreement State licensees. The reports also described 27 Other Events of Interest. Of the 34 A0s at NRC licensees, 10 occurred at nuclear power plants and 24 at other licensees (e.g., fuel cycle facilities, industrial radiographers, medical institutions). Thirteen of the 34 events reported in 1987 occurred during 1987, 19 during 1986, 1 during 1985, and 1 during 1984. [ Note: The third and fourth quarter CY 1987 A0 reports described 7 A0s at NRC licensees (2 at nuclear power plants and 5 at other licensees), 2 A0s at Agreement State licensees and 7 Other Events of Interest]. , The number of A0s reported for nuclear power plants in the Reports to Congress for each calendar year since 1977 is shown in Figure 15. The history of A0s on a per plant basis is shown in Figure 16. An increasing trend in the number of A0s was noted starting in 1982, with a sharp drop noted for 1987. Although there was an increasing number of nuclear power plants over this period, a review of Figure 16 shows that the number of A0s per plant remained relatively level for the years 1983 through 1986. It is not clear why the sharp drop during 1987 occurred or whether this reduction will continue. 1987 was the first year since 1931 in which no A0s occurred at plants licensed less than 2 years at the time of the A0. This was a change from previous years; of the 49 A0s reported from 1981-1986, 29 percent (14 events) occurred at plants licensed less than 2 years at the time of the A0. A summary of 1987 A0s is provided in Appendix B. This sunmary includes power reactor, nonreactor, and medical misadministration A0s. 2.4 Data from Operations Center in 1987 The NRC's Operations Center is the NRC's center for direct communications, through dedicated telephone lines, with licensed nuclear power plants and certain fuel cycle facilities. The NRC Operations Center is staffed 24 hours a day by an NRC Headquarters Operations Officer (H00). The H00 is a Reactor Systems Specialist trained in the receipt, evaluation and response to events which are required to be reported to the NRC Operations Center. The NRC Operations Center received about 3500 notifications in 1987 from licensees, primarily nuclear power plants, concerning events falling within NRC's piompt reporting requirements (10 CFR 50.72). A small subset of these notifications involved events that were classified by licensees into one of the emergency classes specified in licensees' emergency plans. The remainder of the reportable events did not meet the emergency classification levels. The number of events classified under the emergency plans for the past four years is shown in Table 7 below. Unusual Events represent conditions that are no immediate threat to public health, and Alerts indicate actual or i potential substantial degradation of plant safety. For comparison, Site Area and General Emergencies (if they occurred) would indicate major failures of , systems required for public safety or the potential for a major offsite release, j 30
FIGURE 15 U.S. NUCLEAR POWER PLANTS ABNORMAL OCCURRENCES VS. YEAR 2a u) 18 - O Z 16 - E
] 14 -
8 22 -
, q _
10 - - 9 : j k 8-
~
- 8 S-g;
- :
g: _ i ll L l l i d L I 77 I 78 79 I i 80
~~
81 l l 82 CALENDAR YEAR L ~ 1 83 I 84 i 85
~
i 86 1
~~
87 l
FIGURE 16 ! U.S. NUCLEAR POWER PLANTS l ABNORMAL OCCURRENCES /P'dNT VS. YEAR l teseNo
.18- - 180 AOs/ PLANT
.16- - 160 # PLANT g.- .., g
=
g .,2- - ,20 l g .,- : -1.g 8 .m : -=8
~
b .0s- ' i - s0 $
! s 1
l l j 0 l 77 I ? 78 79 l 1 80 1 81 i i 82 83 84 i i 85 {i 86 1 87 0 CALENDAR YEAR
l l Table 7 Classification of Events Under Emergency Plans 1984-1987 1 1984 1985 1986 1987 Notification of Unusual Event 224 312 209 231 Alert 8 11 9 9 Site Area Emergency 0 0 0 0 General Emergency 0 0 0 0 The alerts reported in 1987 were as follows:
- 1. Braidwood - 01/21/87 Cause: Lost Level in Component Cooling Water (CCW) System Duration: 15 minutes
- 2. Big Rock Point - 03/31/87 Cause: Fire in Station Power Room Duration: 3 minutes
- 3. Perry - 03/16/87 Cause: Loss of Control Room Annunciators Duration: 34 minutes
- 4. Big Rock Point - 06/23/87 Cause: Suspected Fire Inside Containment Duration: 1 hour 15 minutes
- 5. North Anna - 07/15/87 Crise: Steam Generator Tube Rupture Duration: 6 hours 37 minutes
- 6. Calvert Cliffs - 07/23/87 Cause: Loss of Offsite Power Duration: I hour 40 minutes
- 7. Fort St. Vrain - 10/0/387 Cause: Fire in Turbine Building Duration: 7 hours 45 minutes
- 8. Palisades - 10/15/87 Cause: Loss of Shutdown Cooling Duration: 1 minute
- 9. Crystal River 3 - 10/16/87 i Cause: Loss of Offsite Power Duration: I hour 8 minutes an H00 in response to these event notifications ranged from a Actions taken by/ log book entry to establishing emergency conference calls simple computer between the H00 the licensee, and senior NRC regional and HQ staff members.
ForverysignifIcantevents,suchconferencecallswouldresultinactivation ! 33
t of the agency's Incident Response Plan. While none of the events ifsted above resulted in activation of the NRC Incident Response Plan, the Unusual Event at Palisades on 7/14/87 and the North Anna and Calvert Cliffs Alerts (7/15/87 and
, 7/23/87, respectively), resulted in additional NRC staff members being called i
into the headquarters and regional incident response centers to monitor them because of their potential for degrading into more serious events. , i l I 1 P l l { i L i k i d i ! t d I .i ( l ) 34 , [ L
3.0 RESULTS OF AEOD STUDIES I In 1987, AE00 issued a number of case studies, special studies, engineering evaluations, and technical reviews, in addition to routine reports. The results of these studies are summarized in tho following sections. AEOD case studies involve substantive in-depth analyses on significant safety issues identifiad through the review of operating experience. They document I the bases for A't0D recommendations for regulatory or industry actions. A case i study report goes through a rigorous peer review process prior to its publica- l tion to ensure technical adequacy. Special studies document accelerated investigations, and contain suggestions ; or recommendations for regulatory act Uns on an expedited basis. Engineering evaluations document evaluations of significant operating events and contain suggestions for remedial actions where appropriate. Technical reviews document lesser work products, and typically conclude that corrective actions implemented or planned are adequate. I 3.1 Operational Experiences at Newly Licensed Nuclear Power Plants, , NUREG-1275 (Vol. 1) In July 1987, AE00 issued a study of the operational experience of commercial power reactors during their first two years of operation. The goals of the study were to: (1) characterize the trends in operational events experienced at new licensed plants; (2) identify correlations between plant attributes and performance; and (3) provide feedback to facilitate improvement. The study focused on the startup experience of 22 plants for which operating licensees were issued between January 1983 and June 1986 and analyzed data drawn from operational reports and site visits to 11 of the plants. The report documenting the study, Operating Experience Feedback Report - New Plants," NUREG-1275 Vol. 1 (July 1987), was subjected to the formal AE00 peer review process that included the staffs of the plants that were analyzed, industry organizations, and all major NRC staff offices. , The study concluded that it was possible to achieve significant improvements in early plant performance, learning curves, and early commercial operation. The study disputed the assumption that all new plants experience high event frequencies during their first two years of operation. A number of suggested improvement. were developed in areas such as plant management, personnel train-ing and plant equipment, and in some NRC practices. The analyses showed that, without early effective corrective action, the root causes of a high evet frequency will likely persist during early commercial operation. At this stige, the relatively high challenge frequency coupled I with the potential of unietected systems problems might present a significant challenge to a relative 1) inexperienced operating crew. It was determined that a root cause corrective action program, in response to events, is a necessary factor in achieving good performance. I 35 t l
Among the lessons learned from this study were: A. Management Related (1) Establish an operating plant mentality prior to initial criticality. For example, ensure that plant operations personnel have the responsi-bility for operating all equipment as early as possible in the con-struction completion process. Take early, complete control of the transition from construction to operation. (2) Conduct a deliberate, evenly paced, thorough, and well planned preoperational and startup test program. For example, conduct thorough reviews and dry runs for planned testing and allow time for additional testing during either the preoperational or startup testing program. Emphasize planning to red 'ce the frequency of unplanned scrams and unnecessary ESF ac..*',1ons. A detailed review of operational experience of similar plants should be a principal guide to the areas needing additional attention. (3) Use the finalized Technical Specifications (TSs) to generate and validate surveillance testing procedures as early as possible against the as-built plant. (4) Improve administrative control of surveillance. For example, evaluate the location and nature of work activities during operation in terms of adverse effects on plant operation and take appropriate administrative actions. (5) Give high visibility to the sources (i.e., organizational element) of unplanned scrams (and other unplanned events) caused by human error and establish performance goals. (6) Ensure that operating experience feedback programs: (a) combine internal events with relevant events from similar plants, (b) are communicated directly to the appropriate first level supervisors and working level staff at the plant on a periodic basis, including prior to startup, and (c) address preventive measures. (7) Establish as a major goal an increased commitment to training in performing surveillance testing, calibration, and troubleshooting activities well prior to operations. I&C training initiatives, such as repeated practice for those surveillance testing activities that could cause a transient and which should be conducted on actual in-plant equipment on live systems prior to operations, should be emphasized. (6) Emphasize training for routine operations involving power level changes and the associated communications among shift personnel (i.e. , feed flow and turbine evolutions) that have historically caused trips. (9) Establish extensive, detailed training for all segments of the onsite plant staff, including I&C techniciaits, maintenance mcchanics, security staff, operations, and management. This 36
training should emphasize: (a) the applicability of the various TSs to the changing plant modes of operation and associated schedules, (b) the relationship of the TSs to the plant procedures, (c) the NRC requirements for reportability of violations, and (d) the basis for the TSs and discussion of LCO requirements. l l B. Equipment Related I (1) Focus on the balance-of plant prior to operation and early in life l appears to provide a high return regarding the reductiun of unplanned scrams and ESF actuations. Within this area, attention could be given to:
- Conducting aiditional reviews of feedwater and turbine control and bypass systems to identify sensitivities and plant-specific characteristics that could contribute to transients or the ability of the system to cope with expected transients.
Conducting a systematic review of equipment protective logics and setpoints on components such as pumps (suction trip, time delay, vibration trip) or power supplies to identify areas where a time delay or additional channels for coincidence could reduce the potential for unnecessary transients or spurious actuations. (2) Install test jacks and bypass switches at appropriate points in actuation circuitry. (3) Implement on a priority basis vendor or licensee trip reduction measures. Licensee trip reduction programs should focus on safety-related equipment as well as on GOP equipment. (4) Pay attention to the design and installation of equipment located in the vicinity of radiation monitors and associated cabling to ensure that adequate grounding of equipment, cable shielding, etc., are provided to prevent the occurrence of electromagnetic interference which can trigger this extremely sensitive instrumentation. (5) Thoroughly test new or unique plant features, such as new RPS systems, electrical systems, etc., prior to fuel load to reduce unanticipated failures or unexpected erratic behavior. ! (6) Give preference for proven designs and standardization of design in l plant feedwater and turbine systems for future designs or major plant modification. Conduct further analyses of any first of a kind, one of a kind, and state-of-the-art features, since they have generated a large number of problems during plant startups. (7) Incorporate scram prevention measures 3uch as:
- Develop a color coding scheme for single point scram components whose misoperation could cause a scram (for example, pressure sensing lines).
37
Install cages or covers ov . . :es or racks that could provide trip signals. I Since October 1986, both the l.C staff and the industry have given increased attention to newly licensed plants. The new plant study reinforced the need for these efforts. Following publication of NUREG-1275 Vol. 1, the staff met with the Commission (August 4, 1987) to discuss new plant performance. As a follow-up, the staff transmitted at the Commission's request a copy of NUREG-1275 Vol. 1 to new licensed ruclear power plants, to plants still under construction, and to plants undergoing a prolonged shutdown, and requested i them to review the applicable improvement lessons. Some power reactor i licensees iave already recognized the need for preventive programs and i developed programs that, when implemented fully, could result in improved
- performance in many respects.
AE00 also conducted a review of the operational experience of the 12 plants licensed since June 1986 which were not included in NUREG-1275 Vol. 1 This study documented in "Technical Review Report: Recent N u Plant Operational Experience," AE0D/T709, 1987, indicated that these plants were performing in a manner similar to those studied in NUREG-1275 Vol. 1. AE00 discussed the findings of these studies with senior utility and plant man-agement at an INP0 workshop on new plant performance improvement in December 1987. AE0D also supplied the lessons learned during the study to the following plant reviews: Vogtle 1 Three site meetings with licensee senior management concerning plant performance during startup and power 1 ascension. 4 Beaver Valley 2 Site meeting with licensee on readiness to operate the plant prior to issuance of the initial operating license. South Texas 1 Special pre-startup inspection. 1 Paio Verde 3 Site meeting wi*h licensee. j Clinton Special review of the initial plant operating experience. l 3.2 Air Systems Problems at U.S. Ligh, w - . wtors, NUREG-1275 (Vol. 2) J
- This study provides a comprehensive review and evaluation of the potential t
safety implications associated with air systems problems at U.S. LWRs. The study analyzes operating data, focusir.g upon degraded air systems, and the vulnerability of safety-related equipment to common mode failures associated witi, air systems. Several recommendations are prosented to reduce risk, enhance safety, and improve plant performance. i- Air systems are not safety grade systems at most oper .ng plants. As a result, plant accident analyses assume that safety-related equipment dependent upon air systems will either "fail safe" upon loss of air or perform their intended function with the assistance of backup accumulators. The report highlights 29 failures of safety-related systems that resulted from degraded 38
or malfunctioning air systems. Some of the sy;. ems which were significantly i degraded or failed were decay heat removal, auxiliary feedwater, BWR scram, l main steam isolation, salt water cooling, emergency diesel generator, contain-ment isolation, and the fuel pool seal system. These are viewed as important precursor events. (he report identifies the following as specific deficiencies: (1) the air j quality capability of the instrument air system filters and dryers dcas not always match the design requirements of the equipment using the air, resulting in mismatched equipment; (2) maintenance of instrument air systems is not always, performed in accordance with the manufacturers' recommendation; (3) air quality is not usually monitored periodically; (4) plant personnel frequently do not nderstand the potential consequences of degraded air systems; (6) operators are not well trained to respond to losses of the instrument air, and emergency operating procadures are frequently inadequate; (6) at many pichts the response of key equipment to a loss of instrument air has not been v rified to be consistent with the FSAR; (7) safety-related backup u pmulatou do not necessarily undergo surveillance testing or monitoring to confirm their readiness; and (8) the size and the seismic capability of safety-rela %d bhckup accumulators at several plants are inadequate. The study concludes that additional industry and regulatory actions are nuessasj to assure that air systems are maintt ned and operated at levels which will enable plant equipment to function 6. designed and not ce subject to unanalyzed falure modes. To date, such failures have not occurred in connection with a limiting transient or accident. The recommendatioM resulting from the study include: (1) ensure that air system quality meets the requirements specified by the manufacturers of the plant's air-operated equipment; (R) ensure adequate operator response by formulating and implementing anticipated transient and system recovery procedures for loss-of-air events; (3) ensure, by improsed training, that plant operations and maintenance personnel are sensitized to the importance of air systems and the vulnerability of safety-related equipment served by the air systems to common mode failures; (4) confirm the adequacy and reliability of safety-related backup accumulators; and (5) verify equipment response to gradual losses of air to ensure that such losses do not result in events which fall outside licensing analyses. Subsequsnt to issuance of the case study, additional safety signific3nt events occurred which were related to air systems problems. As a result, the case study wos updated and publish;d as NUREG-1275, Voluae 2 in December 1987. The NUREG was distributed to industry via Information Notice 87-28, Supplement 1 December 1987. The need for further regulatory action remains under review. 3.3 .. _ecay H o, Remo.al Function at Pressurized Water Reactors with C S./ Ora M - .~. tor Coolant Systems, AE00 Special Study Repoi't 5702 On Ap- + . ' dual he't removal (RHR) pumps at Diablo Canion 2 were ' a ng/ cavitation <. Nile the reactor coolant system was parth.i t 'sult, the plant lost its ability to remove decay hea*. ror '
.o .ing this period, the reactor coolant system (RCS) heated up uc > ' -
ng was present in the RCS. 39
i Based on a review of this event, AEOD issued a special report, S702, which ; noted that Diablo Canyon plant't, loss of the decay heat removal function (CHR) was one of 37 reported events at U.S. pressurized water reactors (PWRs) over a 10-year period. Ali. hough there was no impact on public health and safety from these events, there is a potential for a more serious event. The report notes aspects of partially drained -(mid-loop) operation which contribute to risk, especially operation with loss of containment integrity, e.g., an open eouipment hatch. It also presents results of recent probebilistic risk assessmentt which address risks associated with PWRs whilo shutdown. The report identifies a composite set of NRC and industry recommended remedial actions, including those in the 1985 AE00 case study report C503 dealing with decay heat removal problems for PWR operation. The report also presents a cost benefit analysis for the implementation of remedial actions. The report recommends the following actions for licensees:
- 1. Maintain containment integrity to the maximum extent practicable during periods of highest DHR risk (i.e., early stages of shutdown and drain-down operations).
- 2. Improve planning, coordination, procedures, and personnel training during shutdown to ensure the availability of DHR.
- 3. Develop a reliable method for measuring and monitoring reactor vessel level during shutdown modes of operation and develop technical specification requirements for operability of vessel level instrumentation.
- 4. Perform a task analysis of DHR operation.
- 5. Consider removal or disabling of valve autoclosure interlocks in order to minimize loss-of-DHR events.
- 6. Modify plant Technir:a1 Specifications to ensure that the DHR system is available during Mode 4 and the early stages of Mode 5.
- 7. Analyze the hydraulics associated with drained-down operations.
Subsequent to issuance of this report, NRR issued a Generic Letter (GL 87-12) requesting PWR licensees to describe their operations during shutdown condi-tions. NRR has reviewed the licensees' submittals and is presently consider-ing issuance of a generic com unication requiring licensees to implement corrective actions similar to those described in S702 and C503. 3.4 Potential Containment Airlock Window Failure Due to Radiation, Engineering Evaluation Report AE00/E701 This study evaluated Licensee Event Report 50-315/80-005 at the D.C. Cook plant. A vendor report noted that samples of glass used in airlock windows shattered in a beta radiation test. The D.C. Cook plant has two airlocks in 40
each unit. Because of the uncertainty associated with the integrity of the windows under accident conditions, the licensee installed a 3/8-inch thick cover plate over the inner tirlock windows. The investigation noted that airlock windows are used at other plants and that failure of the windows could compromise containment integrity following design basis accidents or severe core damage accidents. It also found some test results that indicated a degraded window strength under gamma radiation. NRR review of this issue concluded that the test scenario was not representative of the containment environment following a design basis accident and regional review did not indicate this to te a generic concern; consequently no further NRC action was considered necessary. 3.5 MOV Failure Due to Hydraulic Lockup from Excesa <e Grease in Spring Pack, Engineering Evaluation Report AE00/E702 This study was initiated as the result of an event at Vermont Yankee that involved motor burnout of a motor-operated valve (MOV). The purpose of the study was to investigate the phenomenon referred to as "hydraulic lockup" together with its effect on motor-operated valve ope ation. The evidence con-firms that hydraulic lockup, which is believed due to the presence of grease in the spring pack araa that inhioits deflection of the Belville spring, can occur with L1mitorque SMB motor operators. The safety concern is that hy-draulic lockup resy represent common mode failure mechanism for ssfety-related MOVs. Further, this condition may result in a failure that would not be de-tected by plant operators and, in fact, the M0V could be inoperable for the next demand af ter an apparently successful closm e. Examples of such undetected failures are: motor burnout, component dauaga due to overloading, or inability to open due to over-tightening during closing. Although hydraulic lockup appears to ba associated with the use of EXXON HEBULA EP-0 grease, one of the two environmentally qualified greases available, it I seems that an adequate understanding of the combination of parameters or con-ditions that cause the phenor::enon has not existed. For example: (1) motor operators manufactured prior to 1975 can be fitted with a modification kit to provide a grease relief path, although this modification may not be fully effective in preventing hydraulic lockup; (2) motor operators manufactured subscquent to 1975 have a design change to provide an internal grease relief . path to prevent lockup ar.d this change may also not be adequate; and (3) mis- ' information continues to exist throughout the int'ustry and there is a lack of awareness of the occurrence of hydraulic lockup in motor opt.rators. Further, current emphasis on the use of environmentally qualified gre ue might expose licensees to a greater risk of occurrence of hydraulic lockup. The study recommendad that NRC issue an information notice to alert licensees about the complex situation. It also recommended immediate industry effort to: (1) identify conditions, secuences, or procedures that result in hydraulic lockup; (2) develop solutions for all motor operators currently in use (moaification kits, design change, etc.); and (3) disseminate the corrective action to all users. 41 i
a The recommended industry effort on hydraulic lockup is similar to an irdustry ' wide program on MOVs that was initis.ed by NUMARC in response to the AE00 report on motor-operated valves (MOV). (See AE0D Case Study C603 "A Review . 4 of Motor-Operated Valves Performance, December 1986.) !
- 3. 6 Loss of Offsite Power Due to Unneeded Actuation of Startup Transformer Protective Differential Relay, Engineering Evaluation Report AE00/E703 This report concerned an event at H.B. Robinson Unit 2 that was initiated by the loss of one of the unit's emergency buses, which led to a reactor trip.
The loss of offsite power occurred during a transfer of auxiliary loads from a unit auxiliary transformer to a startup transformer. Investigations performed subsequent to the Robinson event concluded that loss of offsite power to the emergency buses had occurred because of direct current (DC) saturation of a current transformer (CT) attendant to a startup transformer protective differential relay. The associated safety concern is that power to the emergency buses may be lost at a tin when needed to operate Safety-related > electrical equipment. At Robinson, the main problem was corrected by connecting a second set of cts and rewiring selected circuits. Although the review did not identify any similar reported event, suscepti-4 bl.ity to DC saturation of cts, associated with power transformer protective differential relay circuits, could potentially exist at other operating nu-
; clear plants. Based on the results of the ongoing study, Information Notice 86-87 "Loss of Offsite Power Upon an Automatic Bus Transfer," was issued in Octeber 1986 to alert other plant licensees to the potential problem. The i.ifsrmation notice suggested that review of susceptibility of transformers to
- DC saturation could reduce unwanted transformer isolations at other plants.
The report suggested that NRR consider this event in ongoing licensing reviews
. and incorporate the lessons learned into the next revision of the Standard ;
J Review Plan. t 1 3.7 Discharge of Primary Coolant Outside of Containment at PWRs While on RHR Cooling, Engineering Evaluation Report AEOD/E704 This study was prompted by er.rlier AEOD studies on loss of residual heat removal (RHR) cooling and the inadvertent draining of PWRs during shutdown enoling. The focus of this effort was on conditions involving valve mis-alignments that consequently resulted in uncontrolled loss of priegry coolant outside of containment. Seven operating events which occurred at different PWRs in the last nine years were identified and evaluated in the study. All of these operating events involved discharging 2000 to 5000 gallons of primary water outsido ef containment. If the events were to proceed in an uncon-trolled fashion to severe core aamage, and containment bypass was not manually isolated, the offsite doses could be significant. No automatic mitigation would occ;r because the ECCS actuation is blocked at about 19W psi when a plant is cooled down. Thus, the response is totally dependent on proper operttor action and appropriate procedures to handle this type of an incident. An evaluation indicated that the root causes of these events fell into two categories: personnel errors and procedure deficiencies. Personnel errors included operation of the wrong valve due to poor labeling or ope'ator 3 C
cognitive error. Lapses in operator concentration could be corrected by good communications feedback from coworkers who are monitoring relevant instru-mentation. Procedures deficiencies were related to omissions or lack of specificity in sequential valve operations when conducting tests on the RHR , system. NRR issued Generic Letter GL 87-12 on July 9,1987 on the broader issue of i "Loss of Decay Heat Removal While the Reactor Coolant System is Partially i Filled," and is consider'ng issuing a followup generic letter. These
- communications are expected to cover the concerns expressed in this report.
3.8 RWCU System Automatic Isolation and Safety Considerations, Engineering Evaluation Report AE0D/E705 i This study u dressed the causes for the frequent automatic isolation of the [ l BWR reactor water cleanup tystem and the safety significance of the events. ; Unplanned automatic isolations of the reactor water cleanup (RWCU) system - accounted for 20 percent of all boiling water reactor (BWR) LERs since January 1 1985. During the summer of 1986, two events occurred at BWRs that resulted in leakage of substantial quantities of reactor coolant from the RWCU system to ' areas in the reactor building. The first occurred at Millstone 1 when a small pipe to a relief valve on a heat exchanger failed due to vibration and thermal cycling, resulting in the discharge of 2200 gallons of coolant. The second leak was at Dresden 3, where 140,000 gallons of reactor coolant leaked from a failed valve packing over a 30 hotr period. In both cases the leak was expeditiously identified, although not immediately isolated at Cresden 3. There were no significant radiological consequences. Automatic isolation of the RWCU system i did not occur under loss of inventory conditions because these plants do not ; have such a design feature. ! As a consequence of the two events, this study was conducted to determine the safety significance of RWCU system failures and determine whether all RWCU systems should have the capability to be automatically isolated. The study i
- found that all but eight of the oldest boiling water reactors had some [
j capability for automatic isolation of the system based on signals from the leak detection system. After evaluating approximately 300 instances of RWCU j isolation, the study concluded that only about 10 percent of all automatic I isolations were associated with any leakage of coolant external to the ' system. Futtner, except for the events at Millstone 1 and Dresden 2, no 4 leakage greater than 20 gallons was identified, and there was no evidence of other safety consequences. Thus, the report concluded there was no basis to i require automatic isolation of the RVCU system on all plants. ' The report documented ways for plant operators to reduce the number of RWCU i system isolations caused by operational and design problems in the leakage i i detection system. INP0 distributed the report to BWR operatore. The BWR ' f Owners Group prepared a comprehensive lessons learned and operating experience t manual for the RWCU system (see Walter W. McNeil, "Reactor Water Cleanup Sys- l l tems, A Comprehensive Summary of Design, Corrective Actions and Improvements," ( Detroit Edison Company, September 1987). ! ! I i l i 43 i I
3.9 Inadequate Mechanical Blocking of Valves, Engineering Evaluation Report AEOD/E706 An investigation of domestic LWR operating experience involving inadequate or inadvertent bioching of valves by mechanical methods was init'.ated by an event that occurred at a foreign Westinghouse two-loop PWR. The event resulted in a sustained, uncontrolled blowdown of high energy steam from an unanticipated opening of an upstream isolation valve that was not properly blocked in the "closed" position. The root cause of the foreign reactor event was identified to be inadequate procedural controls for assuring that the valve was incapable of subsequent automatic movement. This study investigated 19 events of mechanical blocking of automatic valves at domestic reactors during a 5 year period from October 1981 to February 1986. Nine of the events involved the application of mechanical blocking devices to safety relief valves. The remaining 10 events in this study involved the misapplication of a mechanical blocking device to one motor-operated and nine air-oparated valves. The study found that:
- 1. Six of the valves were inadvertently blocked from automatic motion and were incapable of responding to a remote command.
- 2. There was one instance of a mechanical blocking device failing, and it was incapable of preventing undesired valve motion.
- 3. Two automatic valves were not mechanically blocked in the "safe" position and each valve subsequently .ycled to a position that led to degradation of safty-related equipment at the plant.
I The study found that the cisapplication of mechanical blocking devices to automatic valves was caused by human error deficiencies (i.e., personnel : errors or inadequate procedures). The events were not repetitive at any ! individual plant so it would appear that the corrective actions taken at these i specific plants were effective. Although this study found that the mi.applicaticn of mechanical blocking devices was an infrequent and unrepetitive occurrence, a high proportion of the events could have and, in some cases, actually did result in significant i compromises ir, safety. Accordingly, it was suggested that an information notict be issued to describe several of these events and the underlying cause and actions that could be taken to minimize the possibility of this type of , problem. Subsequently, Information Notice 87-38, "Inadequate or Inadvertent Blocking of Valve Movement" was issued. The information noti c included the following suggestions fo licensees: verify that procedures for piping modifications l assure that a single isolation valve is aligned in the safe position; identify possible causes of unintended valve actuation duri maintenance, such as te" ;rary 105s of power or surveillance tests that 'd cause an unblocked valve to cycle to an unsafe position, particularly .an the valve tycling would result in an open system; and ensure that procedures for restoring blocked valves to service are clear and include ve.'ification. 44
3.10 Design and Construction Problems at Operating Nuclear Plants, Engineering Evaluation Report AE0D/E707 This study was prompted by design and construct "n deficiencies associated with safety-related systems at operating nuclear piants. The study evaluated 55 licensee event reports (LERs) from 34 plants. Twenty-one of the plants began cosaercial operation in or prio; to 1970; the remaining 13 plants had been in operation for less than two years. The deficiencies were not detected by the plants' QA or QC programs during plant construction or system mndification c d had existed since the initial startups. The deficiencies remained latent -- in the sense that the plants' testing programs, such as post-constructional, preoperational, startup or ; surveillance testing, failed to identify them -- and were discovered during : reviews or insoections conducted for reasons other than the requirements of ) l the plants' QA or QC programs, 1 e The deficiencies fell into six categories: (1) piping stress exceeding code ; limits; (2) incorrect hardware or improper installation of hardware * - offiresaalsforelectricalcablepenetrations;(4)electricalwi'dng(3) . lack errors; i
! (5) errors in electrical, instrumentation and control circuits; and (6) !
electrical and control panels not seismically supported. The deficiencies l were attributed to inadequate document controls, unreviewed safety and design i conditions, or inadequate review during modifications. The deficiencies could i have posed potential hazards to safe plant operation had they remained t uncorrected. The study suggested that an information notice be issued to describe the [ operational experiences identified in the report and alert licensees of , operating plants of the need to review the adequacy of the plants' QA and QC i l programs current'y used in plant modifications. The study also suggested that ; j the findingr. be used as reference material for future inspections and reviews of plant modifications. Subsequently, NRR used the findings in revising several manual chapters and inspection procedures. 3.11 Depressurization of Reactor Coolant Systems at PWRs, Engineering l Evaluation Report AE00/E708 i This study was prompted by an inadvertent reactor trip at Salem Unit 2 or August 26, 1986, which resulted in a loss of normal pressurizer spray, loss of auxiliary spray, and loss of one of the pressurizer power operated relief
; valves (PORVs). Repeated PORV actuations caused by continued operation of . ! high pressure safety injection occurred. Quick impromptu operator action f rectored normal pressurizer spray, secured high pressure safety injection, and !
returned the plar.t to a stable condition. Extrapolation of this sequence of : events to a steam generator tube rupture accident or a natural circulation ! , cooldown (as during an extended loss of offsite power) highlighted areas of ! safety analyses, plant operation, and emergency procedures that could be f improved. , 1 . Operational experience shows that one or more pi'tces of depressurization equipment may be out of se?vice for extended times because technical l l
- i
, 45 : i
specifications do not contain limiting conditions of operation for these systems. Similarly, the Salem event illustrated electrical dependencies that may reduce the availability of the depressurization function. Generally, these problems stem from a lack of specificity in the plant safety analyses acceptance criteria for systems used in mitigating steam generator tube rupture accidents. Further, inconsistencies in the order in which the depressurization systems are used were noted in the emergency procedures for Westinghouse plants. Although emergency procedures were intended to be symptom oriented, special procedures exist for steam generator tube rupture events, and the preferred order of equipment use is different. This adds an element of potential uncertainty to the operator's response and is especially important consideri'ig potantial improvisation in control room operations as an event evolves. The study also noted that efforts to reduce reactor coolant system pressure may be hampered by void formation in the reactor vessel head. The likelihood of severe core damage from steam generator tube rupture accidents was previously estimated to be small. Actions to facilitete the operator's recovery from this event may nevertheless be prudent because of the potential for bypassing containment during this accident. As a result of this study, the following suggestions were made: (1) issuance of an information notice, (2) revision of the standard review plan for steam generator tube rupture to reflect staff's position on systems needed to mitigate the accident, (3) incorporation of limiting conditions for operation for depressurization equipment needed to mitigate a steam generator tube rupture accident in the plant Technical Specifications, (4) reexamination of the use of PORVs before the auxiliary spray system in certain emergency pro-cedures, (5) reexamination of the impact of the uppt> plenum temperature on l depressuritation, and (6) consideration of the problems noted above in advanced reactor designs. As a result of the study, Information Notice 87-60, "Depressurization of Reactor Coolant Systems in Pressurized Water Reactor," was issued in December 1987. 3.12 Auxiliary Feedwater Pump Trips Caused by Low Suction Pressure, Engineering Evaluation Report AEOD/E709 This report was initiated by four events involving auxiliary feedwater (AFd) pump trips on low suction pressure. The low suction pressure trip presen'.s a potential common cause failure of the AFW pumps that could result in a tetal or partial loss of the AFW system. The study found that the AFW trips were
'aused by pressure oscillations or fluctuations in the suction lines whech s.reated momentary pressure drops, as sensed by the suction pressure monitors, despite sufficient available suction head. The pressure oscillatims or fluctuations were a result of hydraulic hammering, excessive suction flow, or pump speed oscillation.
The operational conditions that cause these low suction pressure transients have not been considered in AFW system design. Although the four events involve only the AFW system. similar pressure transients can occur in other safety-related systems. T.< report cited several corrective actions that had 46
l i proven effective by the affected licensees including extending the time i delay associated with the low suction pressure trip function, removing the low ! pressure trip function, or replacing the automatic trip function with an r
- alarm / operator-action combination. The report led to issuance of Information l
- j Notice 87-53, "Auxiliary Feedwater Pump Trips Resulting from Low Suction .
Pressure," and to a review of the adequacy of the safety design basis for the i j AFW system in the ongoing GI-124 program for assuring AFW system reliability q and availability. l t ^ 3.13 Inadequkte NPSH in Low Pressure Safety Systems in PWRs, Engineering ! Evaluation Report AEOD/E710 ; This evaluation of inadequate net positive suction head (NPSH) was prompted by '
; four similar LERs within a short period of time. In three of the four cases, 4
the deficiency went undetected for considerable time because normal plant ' )! ooeration did not illuminate the problem. Inadequate NPSH was attributed to high flow rates generally caused by lower hydraulic resistance downstream of i i the pumps than was originally considered in the design calculations. Higher flows lead to larger pressure losses in the suction piping and therefore a ' , decrease in the available NPSH at the inlet to the pumps. Inadequate NPSH i i causes cavitation which pits and erodes the impeller blades and ultimately results in pump failure. Since a number of systems such as containment spray
! and residual heat removal involve functions which mitigate loss of coolant i accidents, failure of the associattJ pumps due to inadequate NPSH would j constitute a significant reduction in safety margin. ,
.I For example, residual heat removal (RHR) pumps may provide several functions I following a LOCA in Westinghouse plants. Normally, they serve as the low head j i' safety injection and recirculation pumps taking water from the refueling water , i storage tank or the containment sump. In many plants, the high pressure f safety injection pumps are fed from the discharge of the RHR pumps during the t l recirculation phase following a LOCA. In a few plants, the RHR pumps also provide containment spray during recirculation. Thus, it is possible that one RHR pump (assuming a failure of the second RHR pump) could simultaneously
! discharge into four cold legs, two to four HPSI pumps, and to two trains of i
] containment spray. This arrangement significantly reduces the effective ( i hydraulic resistance on the discharge side of the RHR pumps which potentially ; j increases the pump flow beyond design conditions. This situation could then ! i lead to inadequate NPSH and premature failure of the involved pumps. The '
- licensees corrected the deficiencies by modifying operating procedures to pre- !
! clude inappropriate lineups or by increasing the hydraulic resistance in the !
i discharge piping. ! j Staff issued Information Notice No. 87-63, "Inadequate Net Positive Suction [ { Head in Low Pressure Safety Systems," citing the corrective actions taken by the four plants that experienced the low NPSH problem. { f 1 3,14 Compression Fitting Failures, Technical Review Report AEOD/T701 ! I i j This study was initiated as a result of the screening of a Catawba 1 Licensee i ) Event Report (LER 85-59) involving a leaking compression fitting on a primary j system flow transmitter. The review found that 16 compression fitting i 47 ' i
failures we,e reported to have occurred at 11 nuclear power plants within the past 6 years. In about half of these events, the licensees either did not know or did not specify the root causes. For the 9 events in which the root causes were reported, 6 events were attributed to human errors (installation or maintenance errors), 2 were attributed to vibration, and 1 was attributed to thermal stress and cycling. Corrective actions included replacement of compression fittings with welded connectors, improved procedures for the installation and maintenance of compression fittings, and counseling of plant personnel. AE0D's review of industry and NRC actions indicated that t h failures ha9e been adequately addressed and that NRC's follow-up actions on this subject are adequate. As a result of the technical review, AEOD made no additional recommendations. 3.15 Leaking Pulsation Dampeners Lead to Loss of Charging Sys+em, Technical Review Report AE00/T702 This report describes three events in which there was a loss of charging pump suction at Palo Verde due to the failure of the gas-filled pulsation dampener or the suction stabilizer on a positive displacement charging pump. In two I events, a failed pulsation dampener led to all three positive displacement pumps becoming gas bound through their common suction header. Although the charging system at Palo Verde is not safety-related, a study was initiated because of the possibility of similar failures affecting safety-related charging pumps at other plants. The study found that the only safety-related system that could be affected by a similar fLilure of the pui,ation dampeners is at San Unofre 2 and 3. Since similar failures of the suction stabilizers have bec, investigated and documented by previous reports, no additional suggestions or recommendations were made. 2 3.16 Potential for loss of Emergency Feedwater Pump Oue to Pump Runout During Certain Transients Technical Review Report AEOD/T703 This report was initiated by an LER submitted for the Oconee Station which identified a lack of pump runout protection for the emergency feedwater (EFW) pumps. The problem had potential implications for all PWRs. If uncorrected, , the lack of EFW pump protection could lead to a degradation or loss of EFW i function during a steamline break accident. The licensee attributed this
- problem to a weakness in its design engineering program. To correct the problem, the licensee has, in the thort-term, instituted procedures and training for operator actions to prevent pump runout. The licensee is eval-uating long-term modifications whici would eliminate the need to rely on i operator action. The study found that the concern regarding the loss of EFW pump runout for other plants was adequately addressed in IE Bulletin 80-004 and related industry actions. No further action was recommended.
3.17 Pressurizer Code Safety Valve Reliability, Technical Review Report AE00/T704 This study was initiated by an LER from Diablo Canyon 1 documanting that test-ing of the unit's pressurizer code safety valves found their lift setpoints to be aeove their technical specification limits. The valves were subsequently reset correctly. A search of operational data indicated that since January 48
1983 a total of 34 pressurizer code safety valves at 17 plants had setpoint drift, leakage, misadjusted ring settings, or maintenance and installation l problems. These problemi could lead to overpressurization of the reactor coolant system, degradation of the reactor coolant system pressure boundary, or inadvertent trips of the reactor. These problems appear to be generic to all safety relief valves. The report suggested that AEOD initiate a detailed study to determine the , extent of the problem and assess the adequacy of present efforts toward increasingsafetyvalvereliability. Subsequently, AE00 initiated a study entitled Safety and Safety / Relief Valve Reliability." 3.18 Occurrence of Events Involving the Wrong Unit / Wrong Train / Wrong Component, Technical Review Report AEOD/T705 This report updates NUREG-1192, published in April 1986 to document the results of an NRC stu'.y of weaknesses in licensees' programs and practices that contributed to s. vents involving wrong unit / wrong train / wrong components (WV/WT/WC). The ea111er study found that to reduce the potential for these events, it would bs necessary for licensees to upgrade programs in such areas as labeling and it'entification, procedures, and training. However, this type of problem continues to exist. For example, in Octobe.' 1987, three events occurred stue to licensee activities on a wrong component: one at Byron which resulted \n a loss of offsite power; one at Palisades, resulting in a loss of decay heat removal; and one at Quad Cities, destroying a 4.16 KV electrical breaker ar.a resulting in a reactor trip. While the data for 1987 are still considered to be preliminary, the data indicated that: (1) events involving the wrong unit, wrong train, or wron component continue to occur at about the same rate as in recent years, (2)g the averall rate of occurrence at rew plaats is about twice that for older plants, and (3) progress by licensaes in reducing the rate of occurrence of these ; events needs to be evaluated on a plant specific basis. The review found that year-to year changes in the rate of occurrence for the industry can be .; misleading because improvements at some sites can mask the lack of improvement I at other sites. ; 3.19 Recent Events Invc,1ving Turbine Runbacks at PWRs, Technical Review Report , AE0D/T706 This study was initiated following an event at Haddam Neck on April 21, 1987 in which a spurious signal from the nuclear instrumentation caused a turbine load runback from 90 percent power to 85 percent power. The purpose of the study was to investigate the effe:tiveness of the turbine runbac( feature in preventing unecessary reactor scrass and to evaluate the significance of the transients caused by turbine runbacks. The study evaluated 48 events involving turbine runbacks reported at domestic pressurized water reactors from January 1, 1985 to January 31, 1987. fhe stuoy found that the turbine runback feature did not cause a significant transient in any of the events and appears to be generally effective in preventing unnecessary reactor scrams. 49
The study concluded thers was limited safety sigt ficance associated with events involving turbine runbacks and suggested no further review at this time. 3.20 Undetected Loss of Reactor Water, Technical Review Report AE00/T707 This review was prompted by an event at North Anna on June 17-21, 1987 in which the undetected loss of 20 percent of reactor coolant system inventory occurred over three days during shutdown. A leak developed in one of the reactor coolant pump motors during startup. The licensee elected to complete the motor repair without draining the primary system. The repair required disconnecting the motor from the pump which res'J1ted in a 1 gpm leak with 15 psig across the pump seal. The primary system was depressurized and pressurizer level was established at 20 percent. Reactor coolant system makeup needs were met by passive flow from the volume control tank whMh was maintained at 30 psi and refilled by the boric acid blending systein, The plant stayed in this mode for approximately three days with the operstors relying on the pressurizer level to indicate pressure vessel inventory. Over the 3-day period, the plant slowly depressurized and the dissolved gases in the reactor water came out of solution and collected in the upper head. At least three shifts of operators failed to note the ongoing loss of approximately 20,000 gallons of primary coolant inventory. After the Threa Mile Island accident, multiple modifications in equipment, procedures, and training were made to address the insufficiency of the pres-surizer level as an indication of reactor vessel inventory. The focus was on transients and accidents at power rather than all plant modes. Based on the AEOD report, Irformation Notice 87-46, "Undetected Loss of Reactor Coolant," was issued in September 1987 to focus licensee attention on inappro-priate reliance on the pressurizer level as an indication of reactor inventory. 3.21 Problems with High Pressure Safety Injection Systems in Westinghouse PWRs, Technical Review Report AEOD/T708 This study provides an overview of operating and design problems with high pressure safety injection (HPSI) systems in Westinghouse plants. Three out of approximately 500 licensee event reports reflected a total HPSI system failure; the remaining reports discussed degraded HPSI system situations. The study found that previous staff or licensee actions had addressed most of the problems that lead to loss of HPSI. These included problems with valves, loss of pump cooling, boron solidification, debris in the pumps, and deficiencies in the HPSI design. Although no specific action was suggested as a result of this study, the data indicated a relatively high contribution to system unavailability due to equipment out of service, coupled with another random component failure. 3.22 Heating Ventilation, and Air Conditioning System Problems, Technical Review Report AE00/T710 This study was initiated as a follow-up to a memorandum from the ACRS to the EDO on the generic extent of heating, ventilation, and air conditioning (HVAC) 50
system problems. A search of the data base of operational experience involv- i ing HVAC system problems identified 183 licensee event reports (LERs) from 1981 to mid-1987. These LERs were reviewed for the root cause of the prob-i less, the safety-related systems affected, and the actions taken. The review ' identified three root causes: (1) design deficiency (2) equipment or com- ; I ponent failure, and (3) personnel or procedural error. l j A review of actions taken by the NRC showed that many of the HVAC problems related to design deficiency have been the subject of follow-up and feedback ections. The review of events caused by personnel and procedure problems l j showed that the root cause was generally corrected at the plant where the !
; problet accurred, but that lessons learr.ed were not effectively shared with ;
other I.aclear plants. The study found equipment and component failures are
~
4 i the predominant root cause of the problem of HVAC systems and recommended that ! this area receive further attention in a future study. Based on this effort, ! j AE00 will undertake a coeprehensive study of HVAC problems. ;
- 3.23 Unplanned Criticality Events at U.S. Power Reactors Similar to That at !
l Oskarshamn Unit 3 on July 30, 1987, Technical Review Rebort AE00/T712 3 ; J On July 30, 1987, Oskarshamn Unit 3, a BWR in Sweden, experienced an unplanned ; criticality during routine shutdown margin testing. AE00 LER data bases at j i ORNL (SCSS) and at INEL were screened to see if similar events had occurred at : U.S. commercial power reactors. Several events were found with characteristics ! l of this event at U.S. commercial nuclear power plants within the past few l l years. An event at Peach Bottom 3 on March 17, 1986, appeared to be most like l j that of Oskarshamn 3 in that an unplanned flux anoma'ty occurred as a result of 1 multiple operators being inattentive while a principal safety system was 4 inoperable. Fermi 2, San Onofre 3, Summer, and Vo t1e 1 also experienced flux ; l anomaliesasaresultofmultipleoperatorsbeingknattentiveduringpla.it 1 ops.7 tion, In contrast to the Oskarshamn event, in none of these U.S. reactor j I events were the principal reactor control systems inoperable. Thus, serious ! l consequences from these events were prevented. I
- Prompted by this study, Information Notice 88-21 "Iradvertent Criticality
- Events at Oskarshamn and at U.S. Nuclear Power Plants" was issued on May 9 -
l 1988. The Information Notice included additional event samples and noted that j these events highligl.t the importance of maintaining an operable fast-acting
- scram capability whenever any coupled control rods are withdrawn from a rear, tor a
core. The notice also stated that these events highlight the importance of I following procedures during control rod manipulations and staying alert to the ! relevant instrumentation even when the reactor is not expu ted to become ) critical. The notice encouraged the licensees to review their procedures and
- training programs to ensure that there were no ambiguities on these points. !
I. l 3.24 Mispositioning of "Reverse Acting" Valve Controllers Causing Safety l System Inoperability. Technical Review Report AE00/T713 l 1 > 1 This review was prompted by an event at Susquehanna 2 on October 24, 1986, in [ ! which a water regulating valve failed to open during the performance of a ' l surveillance test of the "B" emergency switchgear room cooler. An operator { i then repositioned the valve's controller from the "100 percent" position to : ? I i 6 l 51 ; i, I.--..-,_~_.-__________,______,__ _,__-,_.m.
, _ _ . , _ _ , . , _ _ , _ ~_.__.,__,m._ m. _ _ .
the "O percent" position and the valve opened. This design is referred to as a "reverse acting" controller. Similar results occurred on the "A" emergency switchgear room cooler when it was tested. The licensee postulated that the equipment in the emergency switchgear and emergency load center rooms might have failed due to loss of room cooling, resulting in a loss of power to the lowpressurecoolantinjectionandcorespraysystemsduringanaccident i because both valve controllers would be inaccessible. To prevent recurrence, a procedural change was implemented requiring that the controller be placed at the "O percent" (open) position instead of the "100 percent" (closed) position and that labels be installed at the controllers to indicate open and closed l positions. The study concluded that the corrective action implemented at Susquehanna 2 in revising the procedure and labeling the controllers should prevent further
! mispositioning of the valves. The root causes of the event were: (1) design error in utilizing a controller as an actuating device for a valve which re-l quires only an open and closed position to perform its intended safety function; !
(2) lack of human factors consideration in failing to label the open and closed !' controller functions; and (3) inadequate procedure which directed that the [ valve controllers be placed in the "100 percent" (closed) position. An , s investigation of approximately 400 LERs involving mispositioned valves from 1984to1987didnotidentifyanyothersafetyconcernsrelatedtothe No further mispositioning of valves due to ' reverse acting" controllers. i follow-up actions are contemplated, j 1 ; 3.25 Feedwater Regulating and Bypass Valves, NPRDS Prototype Study AE00/P701 ( i I l AEOD began a pilot program to analyze NPRDS data in 1985. The purpose was to 1 identify lessons from operating experienc, and recommend appropriate measures I to resolve safety concerns. As part of f 11s program, Idaho National Engi- t < neering Laboratory, under NRC contract, :nalyzed NPRDS failure data for main , feedwater (MFW) flow control valves and ' low control bypass valves in U.S. ! commercial PWRs during 1984 and most of 1985. Major components within the t MFW system were the first ones selected for a.ialysis because they were identi- , fied as a significant cause of unplanned reactor trips which result in demands i 1 on safety systems. Upgrades of the MFF system and supporting systems, such as l ! control air and oil systems, could reduce reactor scrams, and thereby lessen I unnecessary demands on safety systems. ) The AEOD analysis indicated that characteristics and practices for the par.icu-lar plant in which a component was installed had greater influence on the per-j formance of the component than the component itself. The main variation in i MFW control and bypass valve failure rates appeared to be due to differences in operational philosophies and maintenance practices among plants. The various types of generic problems and preventive actions are summarized on the following
- page.
The analysis contains proprietary information and has been issued as an ! internal report. It provided supporting information to the RES Nuclear Plant i j Aging Research (NPAR) Program and to NRR's trial program for inspecting a l generic balance of plant (B0P) system. l 52
Findings of the Flow Control Valve Studies { Problem Cause Actions to prevent problems ! Valve System or Use flexible stainless steel instrument ! operator valve-induced air lines. . failure vibration ' Use vibration-resistant connectors and l fasteners (especially for the soleno d ! valves). Valve Oil, moisture Upgrade the instrument air system with I operator and/or rust, improved blowdown valves and dryers. > or foreign l particles in Monitor instrument air quality and ; the instrueent establish maintenance schedules allowing j air system prompt corrective action. ! I i' Valve Outdoor weather Use waterproof solenoids. operator conditions f failure . t Valve and Poor Use detailed maintenance procedures that i j valve maintenance assure the completion of proper maintenance ' operator procedures and oajustments before system startup. ; failures Provide adequate training and support of ! the maintenance personnel. [ Consult with valve manufacturers to i establish efficient routins maintenance I schedules. ! Have valve manufacturers refurbish the ! i valve trim instead of doing this in-house. }
' i Cover disassembled valves during maintenance. [
Valve Packing leaks Use new packing materials with low i released shrinkage and designs that caintain t t leakage constant pressure on the packing t (spring-loaded, for example). . l Bonnet /flenge In maintenance, carefully inspect the ; flange before reassembly, , Valve Improperly Use improvad, valve-specific maintenance l ! contained adjusted valve procedures. 6
- leakage operators [
i i Damaged valve Consult valve manufacturers for advice on ; tria (plug and improved valve tria designs and materials ,
- cage or seats) for actual plant conditions such as higher !
pressure drops. ! 53 i
.. - - - - _ ~..-.,. _ _ , _ - . - ... .-_ . - -- . . - _ . _ - - - . _ - _ - - _ _
3.26 Ongoing Studies The following is a listing of currently active AE0D studies involving reactor operational events. ; Case Studies Service Water System Performance at LWRs Engineerina Studies l
- Potential LOCA Due to Energized Uncovered Pressurizer Heaters l
Water Damage to Safety-Related Equipment and Electrical Conduits ;
- Design and Other Deficiencies in Contioi Room Emergency Ventilation Systems Improper Application of Grease, Solvents, and Sealants 4
- l Maintenance Problems Causing Extended Unavailability of Equipment ;
'! Large Containment Leakage Events
- Reliability of Non-Safety Related Breakers in ATWS - RPT Pruoence of Operation with Equipment Out of Service in Shutdown Modes !
1 Reactor Scrams Caused by a loss of Feedwater and Complicated by a Failure of the HPCI or RCIC System l i Safety and Safety Relief Valve Reliability l l Pump Cavitation rc Low Flow Conditions
- Tn nds cnd Patterns Studies
- Operational Experience Feedback - Technical Specifications, 1984-1986 '
Operatio6a1 Experience Feedback - Unplanned ESF Actuations, 1984-1986 r Progress in Scram Reduction, 1984-1987 l l 2
- Insights from Accident Sequence Precursors, 1984-1986 :
Trends and Patterns of Main Feedwater Pump Failures J Trends and Patterns of MSIV Failures : i I 1 l
)
i L t r I 54 [
i a l 4.0 INCIDENT INVESTIGATION PROGRAM I The Incident Investigation Program (IIP) is intended to assure that the NRC's i investigation of significant events is timely, thorough, well coordinated, and l formally administered. The scope of the IIP includes the investigation of
- significant operational events involving reactors and nonreactor activities
- licensed by the NRC. The IIP generates two investigatory responses based on !
i the safety significance of the operational events. Both are provided by an l , NRC team which is assembled to determine the circumstances and causes of an operational event. For an event of potentially major significance an Incident , l Investigation Team (IIT) is established by the EDO. AninvestigatIonofless ! j significant operational events is conducted by Augmented Inspection Teams (AITs). During 1987, no event was judged to have a high enough level of safety significance to warrant an IIT investigation. l t Thirteen AITs were conducted at the following plants during 1987. 4 Event Report , i Date Plant AIT Report Date Descrintjon i l 03/11/87 Turkey Point 4 50-251/87016 05/15/87 Boric acid buildup on RV i head. l l: i Loss of decay heat removal. 04/11/87 Diablo Canyon 2 50-323/87018 06/18/87 [ i 06/17/87 Perry 1 50-440/87014 07/30/87 MSIV solenoids not powered I 1 independently. , i i 1 07/02/87 McGuire 1 & 2 50-369/87022 08/31/87 Failure of Westinghouse ! l 07/28/87 50-370/87022 reactor trip breaker. ! j 07/14/87 Palisades 50-255/87019 10/08/87 Loss of offsite power. l 07/15/87 North Anna 1 50-338/87024 08/28/87 Steam generator tube rupture. . 50-339/87024 l 50-313/87-29 i il !
- 08/03/87 Arkansas 1 50-368/87-29 12/03/87 Containment high temperatures, j J !
l 08/07/87 Oresden 3 50-237/87029 10/16/87 Feedwater oscillations result : i 50-249/87028 in reactor trip. l 4 09/06/87 Davis Besse 50-346/87025 10/01/87 Scram with multiple equipment I failures. [ 09/11/87 Oyster Creek 30-219/87029 09/28/87 Violation of safety limit, j } 10/03/87 Ft. St. Vrain 50-267/87-26 10/29/87 Loss of coolin0 due to fire. f i ) 10/29/87 Perry 1 50-440/87-24 01/22/88 Multiple MSIV Failures. { 1 11/13/87 ; P 11/12/87 Pilgrim 50-293/87-53 12/14/87 Loss of offsite power with EDG failure. ( 55
- \
1 5.0 DIAGNOSTIC EVALUATION PROGRAM The Diagnostic Evaluation Program (DEP) was established in 1987 in order to provide an assessment of licensee performance at selected reactor facilities. The DEP evaluates in a safety-related framework the root causes of performance l problems with an adverse effect on plant safety, such as the degree of involve-ment of licensee management and staff in ensuring safe plant operations, the effectiveness of their actions and the adequacy of plant operations, mainte-nance activities, engineering proce3ses, and support programs. When a diagnostic evaluation is approved by the EDO for a specific reactor facility a Diagnostic Evaluation Team (DET) is established. The DET consists ofexperIencedtechnicalscaffmembersfromheadquartersoffices,regionaland l resident inspectors, and contractors, if appropriate. The evaluation process 1
- involves observations of plant activities, in-depth technical reviews, employee interviews, equipment walkdowns, and programmatic reviews in a number of functional areas important to plant safety. Functiunal areas evaluated may
- include conduct of operations maintenance, surveillance testing, corrective j actions sa protection,feguards,plantmodificationsanddesignchanges,radiationquality assuranc
! During 1987, two DETs were established to evaluate performance at the Dresden l and McGuire facilities. A summary of the major technical findings from the evaluations conducted at Dresden Units 2 and 3, and McGuire Units 1 and 2 are L discussed below. l Dresden Diagnostic i Eased upon the team assessment, it was concluded that the fundamental or i root causes of Dresuen's frequently low and fluctuating performance : 4 history were due tu (a) a lack of a strong and in-depth corporate ! I attention in the va t; (b) an attitude and approach that had not been directed at achir.v ng or maintaining a high standard of safety perform- , i ance; and (c) pr.st insrovement initiatives that had been largely a l l reaction to findings sy INP0 and the NRC, and had not been developed in a i J specific and complete way to L'ercome Dresden deficiencies. In addition, ! j major weaknesses in maintenance, inservice testing, communications and ! training were contributing causes. ! Some of the difficulties were believed to be due to past and continuing 4 rescurce limitations. For example, it was found that Dresden's management i
- and staff were overextended and additional management and staff resources were required for long-term improvement. Ihis aspect was believed to q explain the existence of mixed attitudes of station personnel toward j management and the management initiatives for improvement. The team
- found that rany operating and support personnel did not understand the l l improvement initiatives and were skeptical of management's commitment to ;
j them. ! The success of past improvement programs at Dresden has been mixed. Some
- programs had met with a considerable degree of short-term success. For (
j example, there had been a dramatic reduction in contaminated areas, ; j 56 [ 4 i
reductions in personnel contamination, reduction in personnel errors, and a significant improvement in cleanliness inside the plant. These initiatives, combined with a painting program, had greatly improved plant appearance. Personnel morale, althuugh generally low overall, had been improved by these initiatives. The results of other initiatives such as new management trending tools (weekly and monthly reports), an upgraded modification pr'agram, a maintenance assessment, a safety system functional inspection of an emergency diesel generator, and extensive management changes were not yet to the point where a judgmcnt on effectiveness could be made. The team found that for the most part the Dresden staff had a positive attitude towards safety. Despite past and present improver.ient programs, the team found a number of 1 majorweaknesses. These included maintenance (particularly of motor-operated valves), inservice testing, organizational communication, and , operator training. Because of a history of poor maintenance and testing practices, the team concluded that wear, aging, and an accumulation of equipment deficiencies could cause system and component unreliability. Further, the team found that communication was poor across the organiza-
- tion, and that the operator requalification program remained unsatis-i factory. As discussed earlier, these weaknesses were significant contributors to Dresden's poor performance history.
An immediate safety concern associated with exesssive operator overtime was also identified by the team. The team found that a Unit 2 NSO (licensed operator), who as responsible for unit startup, had worked six i double shifts (each shift is 8 hours) in a 7-day period. This overtime,
, which was unconMolleo by management, had been a recurring problem since late 1986. The licensee took prompt corrective action on this concern.
Mr.Guire Diagnostic 1 Although it was determined that the performance at McGuire was improving, the team concluded that the improvement efforts could be slowed by several factors. Foremost among these are the limited utilization of Design Engineering in the evaluhtien of plant operating problems and programs. The team found that althougn Duke's Design Engineering Department was a large ar.d capable resource, it was not being utilized in the day-to-day support of the operating plants due to attitudes within both the Design Engineering and the Nuclear Production Departments which tended to limit ; Design Engineering involvement. Other factors of concern included: the ' 3 4 near-term limitation on the contributions of QA for enhancing plant safety I performance; and some instances of inadequate performance of Construction and Maintenance Department personnel working at McGuire due to inadequate - training. The team was also concerned about the potential for reduced l ! corporate oversight, direction, and leadership for the operating nuclear i I stations due to the competing demands with Duke's growing outside business interests. The operations, maintenance, and testing functional areas were found to have a number of noteworthy programmatic strengths, and some progrates were judged to be above the industry average in overall quality. For example, , a 12-hour operating shift contributed to good morale among the operators j 57 1 1
=
I
, i a and good communication and cooperation between operations and support ; ) groups. Additionally, the preventive maintenance program was found to be '
comprehensive and the completion of surveillance tests was ensured by an
~
j integrated scheduling group at the station. Notwithstanding the above strengths, a num! er of programmatic weaknesses, I
- technical problems, and concerns were identified in each of the functional i areas which were uncharacteristic of Duke's commitment to quality in all L
{ activities. In maintenance, for example, weak root cause determinations, combined with the lack of a formal integrated failure trending program , resulted in recurring common-cause bearing damage for five of the six : j)j McGuire auxiliary feedwater pumps. Significant deficiencies were found in the Inservice Testing Program for safety related check valves and some i air-operated valves. The Inservice Testing Program deficiencies resulted ! in check valve failures in the auxiliary feedwater system and the . team t supply system to the turbine-driven auxiliary feedwater pump not being l detected in a timely manner. The team found that poor technical reviews, resulting from weak involvement by Design Engineering in the development 1 of the initial Inservice Testing Program and, subsequently, in the i, {a development of a comprehensive action plan to address check valve failures ! and problems were a significant underlying cause for the identified test- i
- ing deficiencies. Lack of adequate management review and weaknesses in '
the technical capabilities of the QA surveillance group were also found r
; to be important underlying causes for administrative limits regarding i 1 reactor coolant system and pressurizer cooldown rates being exceeded on a !
i recurring basis, i The diagnostic evaluations conducted at Dresden and McGuire provided senior j NRC management with an in-depth and independent assessment of the licensees' '
- abilities to ensure safe plant operations. .
i ! Each licensee responded positively to the DET findings. The Dresden licensee i ! made a commitment to provide the necessary resources to strengthen and expand . l the improvement programs. The McGuire licensee made a commitment to move l j quickly to resolve the deficiencies identified by the Team. ) l l l l 6 f 58 ? 6
- ~ - - , -. ,- - ,,, , ,,-. - _ - _ .n.- - , - r---------,,------- - , . , - - . - - - , . - - - . - - . . - - - - - ---
l l l l APPENDIX A DATA ON PLANT OPERATIONAL EXPERIENCE A-1 PERFORMANCE INDICATOR PROGRAM A-2 OTHER PERFORMANCE DATA l
i F APPENDl'. A-1 PcRFORMANCE !!.01CATOR PROGRAM r 1 h n 4 A-3 : i l
l PERFORMANCE INDICATOR PROGRAM The Performance Indicator (PI) Program for operating nuclear power plants has been established to provide an objective and timely view of operational per-formance and enhance NRC's ability to a re promptly recognize changes in the safety performance of operating plants. However, it is only a tool, and is ' used in conjunction with other tools such as the results of routine and special inspections and the Systematic Assessment of Licensee Performance (SALP) for providing input to senior NRC management decisions regarding the need to adjust plant-specific regulatory programs. The PI program is a single and coordinated overall NRC program under tne direction of AE00. There are seven PIs currently monitored in the program for all operating plants. The definitions of the PIs are provided below. Automatic Scrams While Critical (Scrams): This is identical to the indir.ator, . unplanned automatic scrams while critical, used by the Institute of Nuclear Power Operations (INPO). In addition, the number of automatic scrams from above 15% power per 1000 critical hours and the number of automatic scrams while critical below 15% power are monitored. - t Safety System Actuations (SSA): This is identical to the indicator, unplanned i !afety system actuations, used by INPO and includes actuations of ECCS (actual and inadvertent) and e: orgency ac power system (actual). ]
,Sionificant Events (SE): These events are identified by the detailed scre ening of operating experience by HRR and AE00, and include degradation of important safety equipment, unexpected plant response to a transient or a
- major transient, discovery of a major condition not considered in the plant safety analysis, or degradation of fuel integrity, primary coolant pressure t boundary, or important associated structures.
J I Safety System failures (SSF): This includes any event or conditions that J alone could prevent the fulfillment of the safety function of structures or j systems. Twenty-four systems or subsystems are monitored for this indicator. Forced Outage Rate (FOR): This indicator's definition is identical to the one used by INPO and the NRC Grey Book (NUREG-0020), and is the number of forced j outage hours divided by the sum of forced outage hours and service hours. ( d Equipment Forced Outages Per 1000 Critical Hours (EFO): This is the inverse of the mean time between forced outages caused by equipment failures. The ' mean time is equal to the number of hours the reactor is critical in a period divided by the number of forced outages caused by equipment failures in that period. Collective Radiation Exposure: This is identical to the one used by INPO. It ; is the total dose at the station. The station total is divided by the number of units at the site contributing to exposure to obtain unit values. ; I The PI data are extracted from licensee event reports submitted in accordance !'
! with 10 CFR 50.73, immediate notifications to the NRC Operations Center in i ! A-5 1 ;
i
- , -- --w -
- , - - - , , - - , - - - - - , - - - . 4
i accordance with 10 CFR 50.72, monthly operating ceports in accordance with j plant Technical Specifications, and radiation exposure reports in accordance with 10 CFR 20.407. Quarterly data for all PIs except collective radiation exposure are presented in the form of charts and tables for each of the operating plants. The collective radiation exposure data on a quarterly basis are in the process of being collected. In the meantime the exposure data are provided on an ennual basis. Since February 1987, the PI reports are being provided to the senior NRC management and Commission on a quarterly basis. The senior management reviews these reports in conjunction with other information el a periodic basis to identify the plants that may require increased NRC attent on. I A-6
l t l l i l l i
\
i 1 i L A APPENDIX A-2 OTHER PEP.FORMANCF DATA A-7
TABLE 1 LISTING OF SIGNIFICANT EVENTS FOR 1987 PLANT DATE TYPE RG DESCRIPTION OF EVENT Arkansas 1 08/14/87 B&W 4 Average temperature in contair. ment approximately 140'F during normal operation Arkansas 2 04/24/87 CE 4 Unisolable leak on pressurizer due to boric acid corrosion Beaver Valley 2 11/17/87 W 1 Personnel induced turbine trip with failure of arto-transfer resulted in scram with 1,s,. of offsite power (17 sec) Braidwood 1 01/29/87 W 3 Oiluted phosphoric acid solution drained into control room due to obstruction in floor drainage system Browns Ferry 2 11/02/87 GE 2 Self-sustaining, self propagating cable tray fire inside containment. Fire lasted 33 minutes Brunswick 1 07/01/87 GE 2 Srv J and L failed to function on demand due to cured loctite RC 620 on plunger Brunswick 2 01/05/87 GE 2 Turbine trip and resultant reactor scram due to failure of generator automatic voltage regulator. Loss of HPCI injection capability due to stuckclosedinjectionvalve, PartiallossofRCICinjectioncapa-bility due to stuck open bypass valve Byron 1 04/08/87 W 3 All component cooling water was lost to Unit 1 05/09/87 W 3 Main steam line supports damaged outside containment Byron 2 10/02/87 W 3 Loss of all offsite power to Unit 2 for 10 hours due to operator error Cr11away 08/15/87 W 3 Degraded essential service water s stem. Condition has existed since 5 84 Calvert Cliffs 1 03/11/87 CE 1 Failure of RCS quench tank rupture disk, Pressurizer code safety valves leakage and setpoint drift A-9
TABLE 1 (Cont'd) PLANT DATE TYPE RG DESCRIPTION OF EVENT Calvert Cliffs 1 04/01/87 CE 1 Plant shutdown due to EQ problems identified riuring audit l j 04/23/87 CE .1 Commercial grade materials installed in safety grade piping systems 07/23/87 CE 1 Both 500 KV feeders to the plant were lost for several hours Calvert Cliffs 2 03/24/87 CE 1 Loss of shutdown cooling and LPSI due to pipe crack in LPSI relief value pipe stub 04/23/87 CE 1 Commercial grade materials installed in safety grade piping systems 05/07/87 CE 1 Crack discovered in LPSI piping, same location as recently repaired crack. Both trains of S/0 cooling were lost 07/23/87 CE 1 Both 500 KV feeders to the plant were
; lost for several hours Catawba 2 03/23/87 W 2 Loadrejectiontestperformedwith unidentifi o RCS leak requiring shutdown. Condition of primary coolant boundary not adequately known Cook 1 08/26/87 W 3 Potential for contairuent over-pressure during a large break LOCA Cook 2 08/26/87 W 3 Potential for containment over-pressure during a large break LOCA -
Crystal River 3 10/16/87 B&W 2 Offsite power lost for over one hour when contract personnel inadvertently shorted startup transformer Davis Besse 03/13/87 B&W 3 Reactor trip on accidental interrup-t'.on of FW to ore S/G. Main steam safety valve fat., to reseat 09/06/87 B&W 3 Scram complicated by atin steam safety valve failure to reseat and loss of three RCPs A-10
TABLE 1 (Cont'd) PLANT DATE TYPE RG DESCRIPTION OF E'!ENT Diablo Canyon 1 01/05/87 W 5 Reactor scram when EHC leak resulted in turbine trip and reactor trip. i 4 Scram complicated by loss of vital , bus of manual transfer prior to l reactor trip a i Diablo Canyon 2 04/11/87 W 5 RCS thermal transient due to loss of RHR while both RCS and contain-l ment boundaries breached - i i t 1 Dresden 3 08/07/87 GE 3 Scram complicated by two steam leaks ( , due to oscillations / vibrations in l main feedwater system [ Farley 2 11/22/87 W 2 2200 gal spill from ruptured pres- l 4 surizer relief tank rupture disk ! due to operator error while testing ! RHR j 12/09/87 W 2 Unisolable leak in RCS (cold leg [ safety injection) [ 1 Fermi 2 06/26/87 GE 3 Unplanned mode change from mode 4 ) to mode 3. Operating staff did not c notice RCS leakage over several hours, j 09/08/87 GE 3 Single failure of Division 1 DC -
- control power resulted in loss of !
- all LPCI i i !
- Fort Calhoun 1 03/21/87 CE 4 Station blackout caused by maintenance j activities underway on both onsite and offsite AC electrical systems 07/06/87 CE 4 During fire surveillance maintenance 1
personnel introduced water into j instrument air lines i Fort St. Vrain 10/03/87 GGA 4 Loss of both helium primary coolant i loops subsequent to fire caused by hydraulic fluid on steam lines Hope Creek 1 10/10/87 GE 1 Scram with safety relief valve stuck open during surveillance testing j Indian Point 2 01/28/87 W 1 Station Class 1E battery found below ' ] temp. required to deliver required f i capacity. Cold weather induced ! j problem ; i A-11
TABLE 1 (Cont'd) PLANT DATE TYPE RG DESCRIPTION OF EVENT Indian Point 2 02/03/87 W 1 Two motor driven AFW pumps fail 18 month surveillance test due to leakage of valve on recirculation line of CST i 02/10/87 W 1 Inadvertent manual scram caused by operator error. After the scram one tie breaker failed to transfer non-safeguards bus to offsite power. Then operator inadvertently opened the second tie breaker Indian Point 3 05/23/07 W 1 Exposure of maintenance foreman due to radioactive particle found on his clothing McGuire 1 08/16/87 W 2 Sct'am with safety injection, 40 gpm primary systein leakage and loss of condenser vacuum McGuire 2 01/20/87 W 2 Scram c7mplicated by main feedwater pump seal failure, pump shaft damage and waterhammer 07/02/87 W 2 Reactor trip breaker failed to open due to mechanism binding and cracked link weld. Breaker erroneously indicated open in the control room Millstone 3 0608/87 W Unplanned mode change from Mode 5 to Mode 4 Monticello 06/07/87 GE 3 Intermittent non-class 1E drywell fan cooler short circuit and trip of class 1E bus feeder- ground fault protection not coordinated i North Anna 1 06/17/87 W 2 LossofRCSinventorg"(Modes 4&5) while repairing RCP A motor 07/15/87 W 2 Steam generator tube rupture scram, SIanddeclarationofsitealert. Region II response team upgraded to AIT i 4 North Anna 2 10/26/87 W 2 High pressure charging pump aligned to reactor coolant system at cold shutdown A-12
TABLE 1 (Cont'd) PLANT DATE TYPE RG DESCRIPTION OF EVENT Oconee 1 04/01/87 B&W 2 Reactor building and decay heat coolers degraaed due to fouling ) Oconee 2 04/01/87 B&W 2 Peactor building and decay heat coolers degraded due to fouling Oconee 3 03/31/87 B&W 2 Reactor building and decay heat coolers degraded due to fouling i 04/11/87 B&W 2 Valves to HPI injecticn pumps found isolated during heatup Oyster Creek 04/24/87 GE 1 Shift supervisor approved opening drywell to torus vacuum breakers i during power operation j i 09/11/87 GE 1 Closure of recirculation discharge valve violated safety limit c i Palisades 07/14/87 CE 3 Loss of offsite power caused by failure of start-up transformer ;
] bushings Palo Verde 1 07/04/87 CE 5 Excessive seal leakage from LPSI [
pump with black foreign substance in lower oil resivoirs of both LPSI j pumps i Peach Botton 2 03/31/87 GE 1 NRC order suspending operations due h to operator inattentiveness to duties , J ; Peach Bottom 3 03/31/87 GE 1 NRC order suspending operations due i to operator inattentiveness to duties Perry 1 03/02/87 GE 3 Operator error induced reactor trip i complicated by RCIC malfunction j' 06/17/87 GE 3 Loss of single RPS bus caused MSIV closure and subsequent reactor trip. i' 4 FSAR description states that MSIV . closure requires trip of both RPS l trains l l 10/29/87 GE 3 MSIV closure tise on 3 of 8 valves l 3 outside of technical specification ' limit for operability , 11/29/87 GE 3 One MSIV failed to close during l
- testing--inoperable solenoid ,
l f A-13 l1 (
TABLE 1 (Cont'd) PLANT DATE TYPE RG CESCRIPTION OF EVENT Pilgrim 11/12/87 GE 1 Snow storm caused total loss of offsite power for 21 hrs Point Beach 2 08/16/87 W 3 Both main steam isolation valves inoperable due to personnel error Rancho Seco 02/04/87 B&W 5 Cable routing errors discovered during cable routing inspection 06/17/87 B&W 5 Lube oil coolers for containment spray pumps found clogged with silt River Bend 1 09/20/87 GE 4 Inadvertent partial drainage of upper fuel pool Salem 1 12/09/87 W 1 Procedure and testing inadequacies in the reactor protection and control systems Salem 2 08/12/87 W 1 Leakage of seal weld on instrument-ation penetration led to pitting of reactor vessel upper head 12/09/87 W 1 Procedure and testing inadequacies in the reactor protection and control systems San Onofre 2 08/31/87 CE 5 Unisolabic leak in reactor coolant system during cold shutdown San Onofre 3 06/21/67 CE 5 Reactor trip with excessive cooldown - and SI injection Shearon Harris 1 10/09/87 W 2 Small LOCA to containment during testing due to improperly installed downstream block valves Surry 1 05/16/87 W 2 Disk separated from stem on hot leg isolation valve causing scram from 100% power 05/23/87 W 2 Boric acid accumulation at seam of reactor vessel head flange studs and mirror insulation. 06/23/87 W 2 Reactor coolant leakage of 47 GPH past loop isolation valve stem packing A-14
f 1 TABLE 1 (Cont'd) l PLANT DATE TYPE RG DESCRIPTION OF EVENT Susquehanna 1 09/23/87 GE 1 Steam line plug blew into reactor vessel Trojan 04/17/87 W 5 Radioactive particles found on technician conducting rad survey of refueling equipment Turkey Point 3 05/28/87 W 2 Loss of boric acid flow path due to tschanical seal failure. N2 intrusion into pump building. 05/19/87 W 2 Orifices in containment spray were never installed. Could result in inadequate NPSH to containment spray pumps. Turkey Point 4 03/13/87 W 2 Boric acid corrosion of equipment on and around the reactor vessel upper head due to small leak from instru-ment port seal 05/19/87 W 2 Orifices in containment spray were never installed. Could result in inadequate NPSH to containment spray pumps. 05/28/87 W 2 Loss of boric acid flow path due to mechanical seal failure. N2 intrusion into pump building. Vermont Yankee 01/29/87 GE 1 Revised minimum flow values of RHR pumps specified by manufacturer may require substantial modifications of miniflow logic and/or miniflow lines Vogtle 04/01/87 W 2 Recent reactor trips considered collectively. Wash. Nuclear 2 03/22/87 GE 5 Manual scram complicattd by water in steam lines (overfill of RPV). Wolf Creek 01/08/87 W 4 Safety injection and subsequent scram with complication due to personnel error. Scram complicated by PORV position indicator failure. Because position indicator erroneously showed PORV open, operator closed block valve. A-15
TABLE 1 (Cont'd) PLANT DATE TYPE RG DESCRIPTION OF EVENT Zion 1 04/30/87 W 3 Failure to follow procedure resulted in opening of MSIV's and inadvertent safety injection. A-16
TABLE 2 REACTOR TRIP RATES 1987 REACTOR AUTO- TRIPS PER PLANT MANUAL MATIC CRITICAL 1000 CRITICAL TRIPS TRIPS HRS HRS ARKANSAS 1 0 2 7856 0.25 ARKANS\S 2 0 4 7715 0.52 ~ BEAVER VALLEY 1 0 4 7340 0.54 BEAVER VALLEY 2 3 14 2319 7.33 BIG ROCC POINT 0 4 6216 0.64 BRAIDWOOD 1 5 5 3429 2.92 BRUNSWICT 1 0 2 5789 0.35 BRUNSWICH 2 0 2 8380 0.24 BYRON 1 0 4 6210 0.64 BYRON 2 2 9 6813 1,61 CALLAWAY 0 2 6228 0.32 CALVERT CLIFFS 1 2 4 6616 0.91 CALVERT CLIFFS 2 2 4 5958 1.01 CATAWBA 1 3 4 6076 1.15 CATAWBA 2 4 5 7213 1.23 CLINTON 1 2 7 5350 1.68 COOK 1 0 3 6001 0.50 COOK 2 0 6 6283 0.95 COOPER STATION 1 5 8424 0.71 CRYSTAL RIVER 3 0 2 5334 0.37 DAVIS-BESSE 0 5 7426 0.67 DIABLO CANYON 1 0 5 8476 0.59 DIABLO CANYON 2 1 4 6059 0.83 DRESDEN 2 0 4 5764 0.69 Dr.ESDEN 3 2 4 7209 0.83 DUANE ARNOLD 0 0 5668 0 FARLEY 1 1 4 8307 0.60 FARLEY 2 0 1 6538 0.15 FERMI 2 0 7 5148 1.36 FITZPATRICK 0 6 6161 0.97 FORT CALHOUN O O 660S 0 GINNA 0 0 8015 0 GRAND GULF 0 2 7203 0.28 HADOAM NECK 0 1 4729 0.21 HATCH 1 0 4 7192 0.56 HATCH 2 0 4 8520 0.47 HOPE CREEK 2 5 7570 0.92 INDIAN POINT 2 0 2 6347 0.32 INDIAN POINT 3 0 6 5496 1.09 KEWAUNEE O 3 7861 0.38 LASALLE 1 0 6 5609 1.07 LASALLE 2 0 1 4781 0.21 LIMERICK 0 2 6151 0.33 MAINE YANKEE 1 1 5724 0.35 MCGUIRE 1 0 4 6836 0.59 MCGUIRE 2 0 5 7047 0.71 A-17
r l l l TABLE 2 (Cont'd) l REACTOR AUTO- TRIPS PER PLANT MANUAL MATIC CRITICAL 1000 CRITICAL TRIPS TRIPS HRS HRS HILLSTONE 1 0 4 6971 0.57 MILLSTONE 2 0 5 8242 0.61 MILLSTONE 3 0 10 6351 1.57 MONTICELLO O 4 7174 0.56 NINE MILE PT. 1 1 2 8171 0.37 NINE MILE PT. 2 0 6 2703 2.22 NORTH ANNA 1 1 3 4585 0.87 NORTH ANNA 2 0 1 6842 0.15 OCONEE 1 1 0 6914 0.14 OCONEE 2 0 3 8605 0.35 OCONEE 3 0 0 6142 0 OYSTER CREEK 0 3 5620 0.53 PALISADES 5 1 4227 1.42 PALO VERDE 1 0 3 4591 0.65 PALO VERDE 2 1 3 6985 0.57 PALO VERDE 3 0 1 946 1.06
'EACH BOTTOM 2 0 0 1730 0 PEACH BOTTOM 3 0 2 1823 1.10 PERRY 6 6 4504 2.66 POINT BEACH 1 0 2 7389 0.27 POINT BEACH 2 0 1 7583 0.13 PRAIRIE ISLAND 1 0 2 7288 0.27 PRAIRIE ISLAND 2 0 0 8760 0 QUAD CITIES 1 0 1 6252 0.16 QUAD CITIES 2 0 5 6041 0.72 RIVER BEND 2 3 5995 0.83 ROBINSON 2 0 7 6354 1.10 SALEM 1 0 2 6413 0.31 SALEM 2 0 4 6423 0.62 SAN ONOFRE 1 0 1 7383 0.14 SAN ONOFRE 2 2 1 6193 0.48 SAN ONOFRE 3 0 2 7135 0.28 SHEARON HARRIS 11 10 6241 3.36 ST. LUCIE 1 0 7 6972 1.00 ST. LUCIE 2 0 5 7382 0.68 SUMMER 0 5 6222 0.80 SURRY 1 1 2 6178 0.49 SURRY 2 0 2 6659 0.30 SUSQUEHANNA 1 1 1 64~5 0.31 SUSQUEHANNA 2 0 1 8484 0.12 THREE MILE ISL 1 0 3 6435 0.47 TROJAN 1 2 4731 0.63 TURKEY POINT 3 2 4 1910 3.14 TURKEY POINT 4 0 1 4503 0.22 VERMONT TANKEE O 3 7375 0.41 A-18
TABLE 2 (Cont'd) REACTOR AUTO- TRIPS PER PLANT MANUAL MATIC CRITICAL 1000 CRITICAL TRIPS T7IPS HRS HRS V0GTLE 1 2 25 5386 5.01 WASH. NUCLEAR 2 1 5 6199 0.97 WATERFORD 3 0 7 7224 0.97 WOLF CREEK 0 9 6153 1.46 YANKEE-ROWE O 2 7248 0.28 ZION 1 0 1 6877 0.15 ZION 2 0 0 5570 0 69 361 A-19
TA8LE 3 INITIATING SYSTEMS FOR MATURE OPERATION Reactor Trips per 1000 Critica) Hours Systems 1984 1985 1986 1987 Feedwater. 0.24 0.18 0.17 0.14 RPS 0.13 0.13 0.10 0.07 Turbine 0.10 0.09 0.10 0.09 Electrical 0.10 0.12 0.09 0.05 Main Generator 0.04 0.05 0.02 0.05 l Mein Steam 0.06 0.04 0.04 0.03 I Control Rod Drive 0.04 0.06 0.04 0.02 l Condensate 0.04 0.04 0.05 0.02 ' RCS 0.05 0.04 0.03 0.02 , Others 0.03 0.03 0.04 0.02 ; I l l I i I i i t s a t i i f l l I t A-20 i
7 I TABLE 4 INITIATING SYSTEMS BY NSSS VENDOR Initiatina Systems - Mature Westinghouse _ Reactors Reactor Trips per 1000 Critical Hours Systems 1984 1985 1986 1987 Feedwater 0.34 0.18 0.21 0.10 RPS 0.16 0.14 0.13 0.09 Electrical 0.16 0.16 0.10 0.05 Turbine 0.11 0.05 0.12 0.10 Main Generator 0.08 0.05 0.02 0.04 Control Rod Drive 0.02 0.06 0.05 0.04 Other 0.02 0.04 0.05 <0.01 RCS 0.06 0.03 <0.01 0.03 Main Steam 0.03 0.02 0.03 0.03 Condensate 0.04 0.04 0.02 0.01 Initiatina Systems - Mature General Electric Reactors Reactor Trips per 1000 Critical Hours Systems 1984 1985 1986 1987 Feedwater 0.11 0.10 0.12 0.13 Turbine 0. 7.3 0.13 0.09 0.08 RPS 0.12 0.17 0.07 0.07 Main Steam 0.13 0.07 0.06 0.05 Electrical 0.05 0.10 0.07 0.05 Condensate 0.07 0.07 0.09 0.04 Other 0.05 0.03 0.05 0.03 Main Generator <0.01 0.05 0.02 0.06 RCS 0.04 0.03 0.03 0.01 Control Rod Drive 0.02 0.04 0.03 0.01 A-21
TABLE 4 (Cont'd) Initiating Systems - Mature Combustion Engineerina Reactors Reactor Trips per 1000 Critical Hours i Systens 1984 1985 1986 1987 Feedwater 0.23 0.23 0.18 0.22 Turbine 0.07 0.09 0.08 0.12 Electrical 0.09 0.05 0.11 0.05 Control Rod Drive 0.13 0.15 0.03 0.04 RPS 0.13 0.05 0.08 0.06
) RCS 0.09 0.08 0.08 0.04 ;
Main Generator 0 0.08 0.05 0.07 l Main Steas 0.05 0.06 0.04 0.02 Condensate 0 0.02 0.07 0.04 i l Other 0.05 0.02 0.01 0.04 i l 1 l Initiatina Systems - Mature Babcock and Wilcox Reactor ]
+
Reactor Trips per 1000 Critical Hours l 1985 1986 1987 Systems 1984 l Feedwater 0.19 0.42 0.14 0.14 ! Turbine 0.04 0. 17 0.08 0.07 0.04 i
! RPS 0.08 0.01 0.05 '
Electrical 0.04 0.12 0.05 0.04 Condensate 0.02 0 0.08 0.02 0.02 0.05 0.02 0 RCS Main Generator 0.02 0.02 0.02 0.02 Control Rod Drive 0 0.02 0 0.02 0.02 0.02 0 0 Main Steam Other 0 0 0 0.02 t i e f h t , t t I ! 1 i i A-22
TABLE 5 REACTOR TRIPS CAUSES - EARLY OPERATION
- Reactor Trips per 1000 Critical Hours l Causes 1984 1985 1986 1987 Hardware 1.61 1.62 0.95 0.92 i Human Errors 0.78 0.57 0.25 0.53 l Procedures 0.31 0.13 0.16 0.24 Unknown 0.09 0.10 0.08 0.14 SG Level 0.07 0.04 0.18 0.07 "Reactor operating from OL date to 24 months.
TABLE 6 INITIATING SYSTEMS - EARLY OPERATION
- Reactor Trips per 1000 Critical Hours Systems 1984 1985 1986 1987 Feedwater 0.73 0.77 0.54 0.77 Turbine 0.57 0.33 0.16 0.30 Condensate 0.31 0.10 0.13 0.19 Electrical 0.19 0.33 0.11 0.05 Control Rod Drive 0.12 0.23 0.13 0.17 RPS 0.33 0.38 0.25 0.24 Main Generator 0.14 0.10 0.11 0.10 RCS 0.17 0.10 0.03 0.06 Main Steam 0.24 0.04 0.08 0.09 Other 0.07 0.07 0.09 0.05 "Reactor operating from OL date to 24 months.
A-23
i l l l i 1 TABLE 7 REACTOR TRIP CAUSES - MATURE REACTORS Reactor Trips per 1000 Critical Hours 19g4 1985 1986 1967 l Hardware 0.55 0.49 0.42 0.33 i Human Error 0.20 0.22 0.17 0.12 Procedures 0.02 0.04 0.04 0.03 Unknown 0.03 0.01 0.02 0.04 SG Level 0.04 0.02 0.02 0.01 I l 1 A-24 l
I TA8LE 8 REACTOR TRIP CAUSES - 14ATURE REACTORS BY NSSS VENDOR Reactor Trio Causes - Nature Westinghouse Reactors Reactor Trips per 1000 Critical Hours Causes 1984 1985 1986 1987 Hardware 0.67 0.45 0.45 0.31 Human Error 0.25 0.24 0.21 0.13 SG Level 0.07 0.03 0.02 <0.001 Procedures <0.01 0.03 0.05 0.02 Unknown 0.03 0.01 0.01 0.03 Reactor Trip Causes - Nature General Electric Reactors Reactor Trips per 1000 Critical Hours Causes 1984 1985 1986 1987 Hardware 0.46 0.51 0.38 0.37 Human Error 0.19 0.22 0.19 0.10 Procedures 0.05 0.06 0.05 0.03 Unknown 0.02 0.01 0.01 0.04 A-25
l TABLE 8 (Cont'd) l Reactor Trips causes_ - Mature Combustion Engineerina Reactors. Reactor Trips per 1000 Critical Hours Causes 1984 1985 1986 1987 Hardware 0.55 0.46 0.45 0.34 Human Error 0.13 0.26 0.13 0.18 ( Procedures 0.05 0.06 0.03 0.07 l i Unknown 0.05 0.02 0.05 0.07 i l 0.07 0.02 0.05 0.02 SG Level l l l l Reactor Trip Causes - Nature Babcock and Wilcox Reactors [ l f Reactor Tript per 1000 Critical Hours ; 1984 1985 1986 1987 l i Causes l 0.72 0.30 0.27, l Hardware 0.31 Human Error 0.10 0.10 0.08 0.08 j Procedures 0 0.05 0.02 0 : i 0.02 0.02 0 0.04 Viknown 0 0.02 0.05 0 SG Level l { w t ( I I f i l t A-26 l I
TABLE 9 Westinghouse Sumstry At Westinghouse reactors for the first three years of the analysis, the main feedwater system dominaitd as the primary initiating system. However, in 1987 the feedwater system in'tiated reactor trip rate approached the reactor trip rate of the RPS (spurious operation) and main turbine system. The table below lists the reactor trip rates for the principal initiating systems at Westing-house reactors. Principal Initiating System at Mature Westinghouse Reactors Reactor Trips per 1000 Critical Hours Systems 1987 Feedwater 0.10 RPS 0.09 Main Turbine 0.10 Electrical 0.05 The reduction in 1987 main feedwater system reactor trip rates when compared with previous years can be attributed to a reduction in hardware induced failures. An analysis of feedwater system initiated reactor trips revealed that although reactor trip reductions occur in all power ranges, the most dramatic reductions occurred at low poner due primarily to a reduction in hardware failure initiated trips. TABLE 10 General Electric Summary T** statistics that follow provide the reactor trip rates for the three principal initiating system in 1987. Principal Initiating Systems at Mature General Electric Reactors Reactor Trips per 1000 Critical Hours Systems 1987 Feedwater 0.13 RPS 0.07 Main Turbine 0.08 The reactor trip rate due to the main feedwater system shows an upward trend in each year since 1984. Spurious trips initiated within the boundaries of the RPS have shown the most significant reduction. The reductions in the RPS initiated reactor trip rate at GE reactors in 1986 and again in 1987 were the result of a reduction in the human error contribution. A-27
TABLE 11 Combustion Engineering Summary The reactor trip rate at mature Combustion Engineering (CE) reactors have shown no significant change over the entire evaluation period. The reactor trip rate at nature CE reactors rose in 1987 when compared with the 1986 trip rate. The upward trend in 1987 is due to an increase in the feedwater system and main turbine systems reac6er trip rates. Statistics for nature CE reactors in 1987 I are as follows: ' Principal Initiating Systems at Mature Combustion Engineering Reactors Reactor Trips per 1000 Critical Hours J Systems 1987 Feedwater 0.22 Turbine 0.12 Electrical 0.05 RPS 0.06 Problems with the main feedwater pump were the primary reasons for main feedwater system initiated reactor trips at CE reactors. TABLE 12 Babcock and Wilcox Summary Unplanned reactor trips at B&W reactors are primarily initiated by the main feedwater system. In 1987 the feedwater system reactor trip rate was 0.14 trips per 1000 critical hours. The 1987 feedwater initiated reactor trip rate repre-sents an increase when compared to the 1986 rate. The reactor trip rate for other systems decreased in 1987. A-28
l TABl.E 13 SlM4ARY OF NUMBERS OF ESF ACTUATIONS l JANUARY 1985 THP90GH DECEMBEP 1987 ; OTHER W PLANTS GE PLANTS CE PLANTS 8&W PLANTS PLANTS
- TOTAL Number of Plants 39-48 32-37 13-15 8 2 92 109 Category of Event All E5F 1231 2311 494 101 98 4235 '
I Valid ESF 412 568 128 58 28 1194 Unneeded 819 1743 366 43 70 3041
; ECCS 136 164 43 25 17 385 En. Power 198 163 54 29 22 466 t HVAC 527 1230 369 18 25 2169 !
RWCU - 689 - - - 689 i i ESF Actuations , w/o RP5 Actuation ~ 6 Total 1045 1982 450 77 78 3632 Valid 300 409 % 37 15 857 ; Unneeded 745 1575 354 40 63 2777 ECCS 81 81 32 10 11 215 Valid 13 7 8 2 - 30 i t i Unneeded 68 74 24 8 11 185 i Emergency Power 156 136 35 25 13 365 i i Valid 51 40 19 12 3 125 t
] Unneeded 105 96 16 13 10 M0 (
HVAC 484 1018 345 15 21 1883 [
; Valid 84 2 04 65 4 3 360 L ; tanneeded 400 814 280 11 18 1523 j RWCU -
555 - - - 555 ! Valid - 153 - - - 153 l
! 402 - -
402 Unneeded - - j ESF Actuations f i With Equipment g Ralfunctions 35 37 19 6 1 98 <
"Lacrosse and Ft. St. Vrain l
f i l A-29 !
TABLE 14 ESF ACTUATIONS FUNCTIONAL DISTRIBUTION Plant Isolation Events Actuation Events h Total Sincie Multiple Total Single Multiple Combination . All 64% 81% 19% 60% 83% 17% 19% W 49% 84% 16% 70% 91% 9% 20% 1 GE 70% 78% 23 55% 80% 20% 26% ! CE 60% 88% IN 52% 78% 25 15 i B&W 35 75 28% 84% 80% 20% 17% i 1 - 1 i i i i l I t 1 (
- t f
~
a i i I I E ! j i h l I 4 L i I !, t 1 i A-30 a l
RGURE A-1 RATES OF ESF ACTUATONS JANUARY 1985 - DECEneER 1987 g weSmc+OUSE-TYPE REACT @S e 7 m E LEGE30
>- 6- ALL ESF 2
< Au.
g - - Nca-HeS g 5-a -nes, ^g ---- VAUD NON-RPS N
$ f --.- NON-RPS HVAC O 4- /
y P \ f - - - - - - NON-RPS ECCS m < g ~ 3 F- / --- NON-RPS EM.PWR. O
< 3- \' /
- u. ,b o wAc /
r e 2-g / g , g ,-- ~~ ~, ~~ , z vAuo f ' ' 3_ s, f N PM ' *
-. .-._y ."
g E ccs
. . .'."".". . . . - - - - - .*..~p.'. . .....
W o i
- p. i i i
- 85-6 85-12 86-6 86-12 87-6 87-12 PEROD BY S R40 NTH INTERVALS
i FIGURE A-2 RATES OF ESF ACTUATIONS JANUARY 1985 - DECEMBER 1967 GEERAL ELECTRIC-TYPE REACTORS g K 13.5 - g E LEGEND E ( 12.0~ s' - E io.5-g
's m ---- VAUD NON-RPS
=
z s.0-N
, N -- - NON-RPS HVAC p % '4 4 7 ,3 _ N, ------ u +Res eccS 1 B o c\ --- NON-ReS Eu.ewR.
l 4 6.0 - % ,O ' l g * " g - - - - NON-RPS RWCU E 4.5 - m . - 2 2 3.0 - nwcu - .- - b-z ,,
>g%
g O 1.5 -
< VAuo * ,""p"," ~
%,ge**
.:ee -
M -~~ ~
,$ Eccs ....---M-= -r-- -
n I I I I I I 4 85-6 85-12 86-6 86-12 87-6 87-12 POWOD BY $ asONTH WTERVALS
FIGURE A-3 RATES OF ESF ACTUAT)ONS JANUARY 1985 - DECEMBER 1987 g CNBUSTKE NWPE REACTORS e 10 m E 9-LEGa m M m
< s- now-nes ,ag Au_ESF d - - NON-ReS e 7_ \
r c -7 s g ---- VAUD NON-RPS Z 6_ s \ ?" 9 ---- NON-RPS H"/AC w t- \ " < 5-s . . . . . . NON-RPS ECCS 3 \ 6- t O
< 4- (% %,, "" - - - NON-RPS EM.PWR.
s s . g 3- , , g y.--s s 2_ - ,\ D # Z p % g VAUD" \ I '~ \ e"# 's O ECCS
.*'w\..........%.,.%
g ERA. PWR M* . W 0 ----Sa
> l i I I I i < 85-6 85-12 86-6 86-12 87-6 87-12 PERIOD BY 6 MONTH INTERVALS I
f9GURE A-4 RATES OF ESF ACTUA10NS JANUARY 1985 - DECEMBEK '987 BABCOCK Af0 WLCOX-T(PE REACTORS x E N
,_ S-z E ESF g - - non-ses i
E 5-1 g ---- VAUD NoN-RPS
= A _..._ nou-aps wAc P 4- /\
x .. .... maps Eccs
< / \
E / \ --- Non-nes z.v.pwn. o 3-
< / \
g a/ .
\
m 2- /s e s m wow-nes,
- a /
/ s s
\
i 3 3 2 vauo,/ s \ f i_ f g t_- 8
- m. m w - - s 's /
g mac ~ -m I -, o ECCS . 7, , , , j , p, - - Q - - - e
,p'%* .,
1 m I I y I I I I
< 85-6 85-12 86-6 86-12 87-6 87-12 POWOO BY 5 200 NTH INTERVALS 1
4
-er m w,-v - - ---- , ,---w---w, w-r-- ---
--------m9- , -yw--v---y--- - - - - -
p _ - . , _. __ _ _ _ _ __ _
TA8(E 15 SAFETY SYSTEM FAILURES UNIT 86-1 86-2 86-3 86 4 87-1 87-2 87 3 87-4 1 0 0 0 0 0 0 0 ARKANSAS 1 1 1 0 0 0 0 0 0 ARKANSAS 2 0 0 0 0 1 0 0 0 BEAVER VALLEY 1 0 2 0 BEAVER VALLEY 2 0 0 0 0 1 0 0 0 BIG ROCK POINT BRAIDWOOD 1 2 3 0 1 1 0 0 0 1 1 1 1 l BROWNS FERRY 1 0 BROWNS FERRY 2 0 0 0 0 2 0 3 1 i BROWNS FERRY 3 0 0 0 0 2 1 0 0 ?, 1 3 2 2 BRUNSWICK 1 2 0 0 0 6 2 0 . BRUNSWICK 2 0 0 0 0 0 5 1 0 BYRON 1 0 1 2 0 C BYRON 2 0 1 0 0 0 1 3 2 , CALLAWAY 1 0 0 0 3 0 0 0 CALYZRT CLIFFS 1 0 0 0 0 0 1 0 0 CALVERT CLIFFS 2 1 0 1 0 2 1 0 1 CATAWBA 1 0 0 0 0 1 1 0 1 CATAWBA 2 1 0 0 1 3 3 . CLINTON 1 1 1 0 0 0 0 2 0 COOK 1 I 1 1 0 0 0 1 0 0 COOK 2 2 1 0 1 1 0 1 0 COOPER STATION 1 0 0 0 1 0 2 1 CRYSTAL RIVER 3 1 0 2 2 0 0 0 1 DAVIS-BESSE
' 0 0 1 0 0 1 0 1 DIABLO CANYON 1 1 0 0 0 1 3 0 1 ,
DIABLO CANYON 2 ' DRESDEN 2 0 0 0 0 v 2 1 2 0 0 1 0 2 0 7 0 [ DRESDEN 3 2 2 1 3 1 2 1 1 DUANE ARNOLO 0 0 0 0 1 0 0 1 FARLEY 1 1 0 0 0 0 0 0 0 i FARLEY 2 0 1 0 1 3 4 4 1 FERMI 2 2 2 2 2 0 0 3 0 FITZPATRICK i 0 0 0 0 1 3 0 0 FORT CALHOUN FORT ST, VRAIN 0 0 1 1 2 0 0 0 0 0 0 0 3 0 0 0 GINNA 0 3 0 1 1 0 0 0 t GRAND GULF 2 2 1 0 0 0 1 0 HADDAM NECK 0 1 1 1 0 3 2 1 HATCH 1 0 0 3 2 0 2 3 2 HATCH 2 . l 1 1 3 1 2 2 0 HOPE CREEK l 0 0 0 0 3 3 0 0 INDIAN POINT 2 0 0 0 0 0 1 0 0 : INDIAN POINT 3 1 0 0 0 1 0 0 0 , KEWAUNEE r I l l A-55 i t i
4 TABLE 15 SAFETY SYSTEM FAILURES (Cont'd) UNIT 86-1 86-2 86-3 86-4 87-1 87-2 87-3 87-4 LASAllE 1 0 0 1 1 2 0 1 2 LASALLE 2 1 1 1 0 0 0 1 5 LIMERICK 0 1 0 0 0 3 1 3 MAINE YANKEE 0 0 0 0 0 1 0 0 MCGUIRE 1 2 1 1 2 2 0 2 4 MCGUIRE-2 2 0 1 1 2 0 0 3 MILLSTONE 1 0 0 0 1 0 2 3 0 MILLSTONE 2 0 3 0 2 0 0 0 0 MILLSTONE 3 4 0 0 0 0 1 0 1 MONTICELLO 2 1 1 0 0 2 0 1 NINE MILE PT. 1 0 1 1 0 2 0 1 0 NINE MILE PT. 2 2 1 3 5 7 NORTH ANNA 1 0 0 0 1 0 3 0 0 NORTH ANNA 2 0 0 0 1 1 1 1 0 OCONEE 1 0 0 0 1 2 0 0 1 OCONEE 2 0 0 0 1 2 0 0 0 OCONEE 3 0 0 0 1 2 2 0 0 OYSTER CREEK 1 1 3 3 3 2 1 3 l PALISADES G 0 3 0 4 1 0 2 PALO VERDE 1 0 2 0 0 1 0 0 0 PALO VERDE 2 0 2 0 0 1 0 0 0 PALO VERDE 3 0 1 0 PEACH BOTTOM 2 0 0 1 0 0 2 5 1 PEACH BOTTOM 3 1 0 0 0 0 1 2 0 PERRY 1 0 2 0 5 9 6 7 6 l PILGRIM 0 2 2 0 2 1 1 1 POINT BEACH 1 0 0 0 0 0 0 1 0 POINT BEACH 2 0 0 0 0 0 0 1 0 PRAIRIE ISLAND 1 0 0 0 0 1 1 0 1 PRAIRIE ISLAND 2 0 0 0 0 0 0 0 1 : QUAD CITIES 1 0 1 1 3 2 1 2 2 : QUAD CITIES 2 2 3 0 1 4 0 1 2 , RANCHO SEC0 2 2 1 6 4 3 3 2 ' RIVER BEND 3 0 1 2 0 2 2 1 l ROBINSON 2 1 0 0 0 0 6 4 5 . SALEM 1 0 2 1 0 0 2 2 1 SALEM 2 0 2 1 0 0 1 2 3 SAN ONOFRE 1 0 0 0 0 0 0 0 0 SAN ON0FRE 2 1 0 0 0 0 0 0 0 SAN ONOFRE 3 0 0 1 0 0 0 0 0 SEABROOK 0 0 0 0 1 0 0 0 SEQUOYAH 1 0 0 0 1 3 1 7 4 SEQUOYAH 2 0 0 0 1 2 1 5 4 SHEARON HARRIS 1 0 2 1 1 1 r A-36 '
. , , -. ,,-- -v -- --
w-
TABLE 15 SAFETY SYSTEM FAILURES (Cont'd) UNII 86-1 86-2 86-3 86-4 87-1 87-2 87-3 87-4 SHOREHAM 0 0 0 0 0 0 1 0 SOUTH TEXAS'1 0 3 l ST. LUCIE 1 0 0 0 0 0 3 0 1 l ST. LUCIE 2 1 0 1 0 0 0 0 0 SUlWER 0 0 0 0 0 1 2 1 SURRY 1 0 1 0 0 2 2 0 1 SURRY 2 0 0 0 1 0 0 0 0 0 1 1 0 2 3 0 0 t SUSQUEHANNA 1 1 0 1 1 2 2 0 0 SUSQUEHANNA 2 THREE MILE ISL 1 1 0 0 0 1 0 0 0 TROJAN 2 2 0 0 1 4 0 1 TURKEY POINT 3 3 3 1 2 1 3 1 1 TURKEY POINT 4 2 4 1 2 1 5 1 2 VERMONT YANKEE 4 2 0 0 0 1 1 2 V0GTLE 1 1 4 5 1 WASH. NUCLEAR 2 0 2 0 0 0 0 0 0 WATERFORD 3 0 0 0 0 0 0 0 0 WOLF CREEK 0 1 0 0 1 2 2 1 YANKEE-ROWE O 2 0 0 1 1 0 0 ZION 1 1 0 2 1 1 1 0 0 ZION 2 1 0 2 2 0 0 0 0 l i l l A-37 1
TABLE 16 COLLECTIVE RADIATION EXPOSURE MAN-REM UNIT 1984 1985 1986 1987 ARKANSAS 1 403 155 571 191 ARKANSAS 2 403 155 571 191 BEAVER VALLEY 1 504 60 627 210 BEAVER VALLEY 2 BIG ROCK POINT 155 291 84 222 BRAIDWOOD 1 BRAIDWOOD 2 BROWNS FERRY 1 647 386 525 394 BROWNS FERRY 2 647 386 525 394 BROWNS FERRY 3 647 386 525 394 BRUNSWICK 1 1630 1402 955 710 BRUNSWICK 2 1630 1402 955 710 BYRON 1 104 769 BYRON 2 CALLAWAY 70 225 393 CALVERT CLIFFS 1 240 347 174 206 CALVERT CLIFFS 2 240 347 174 206 CATAWBA 1 143 225 CATAWBA 2 143 225 CLINTON 1 COOK 1 381 473 373 333 COOK 2 381 473 373 333 l COOPER STATION 799 1333 320 103 I CRYSTAL RIVER 3 49 689 472 524 l OAVIS-BESSE 177 71 124 47 DIABLO CANYON 1 304 168 DIABLO CANYON 2 168 DRESDEN 2 591 843 1398 623 DRESDEN 3 591 843 1398 623 DUANE ARNOLD 189 1112 187 667 FARLEY 1 451 1276 429 299
- FARLEY 2 451 1276 429 299 I FERMI 2 l FITZPATRICK 971 1845 411 940 1
FORT CALHOUN 563 632 74 388 FORT ST. VRAI.': GINNA 394 426 357 344 GRAND GULF 436 420 HADDAM NECK 1216 101 1567 750 HATCH 1 1109 445 749 408 HATCH 2 1109 445 749 408 [ HOPE CREEK 117 A-38
1 1 TABLE 16 COLLECTIVE RADIATION EXPOSURE MAN-REM (Cont'd) UNIT 1984 1985 1986 1987 INDIAN POINT 2 2644 192 1250 1217 INDIAN POINT 3 230 570 202 500 KEWAUNEE 139 176 169 226 LASALLE 1 252 343 475 697 LASALLE 2 252 343 475 697 LIMERICK 175 MAINE YANKEE 884 700 100 722 MCGUIRE 1 507 386 508 522 MCGUIRE 2 386 508 522 MILLSTONE 1 836 608 150 684 MILLSTONE 2 120 1581 918 253 MILLSTONE 3 253 MONTICELLO 2462 327 596 568 NINE MILE PT, 1 890 265 1275 141 NINE MILE PT. 2 NORTH ANNA 1 973 420 361 761 NORTH ANNA 2 973 420 361 761 OCONEE 1 369 435 475 381 OCONEE 2 369 435 475 381 OCONEE 3 369 435 475 381 OYSTER CREEK 2054 748 2436 522 PALISADES 573 507 672 456 PALO VERDE 1 335 PALO VERDE 2 335 PALO VERDE 3 PEACH BOTTOM 2 1225 1739 540 1098 PEACH BOTTOM 3 1225 1739 540 1098 , PERRY 1 PILGRIM 4082 893 1579 POINT BEACH 1 395 241 201 277 POINT BEACH 2 395 241 201 277 PRAIRIE ISLAND 1 74 208 128 68 PRAIRIE ISLAND 2 74 208 128 68 790 519 496 388 QUA0 CITIES 1 790 519 496 388 QUAD CITIES 2 RANCHO SECO 222 756 300 RIVER 8END 378 ROBINSON 2 2880 311 539 499 SALEM 1 341 102 300 300 SALEM 2 341 102 300 300 SAN ONOFRE 1 513 189 412 232 SAN ONOFRE 2 473 267 412 232 A-39
TABLE 16 COLLECTIVE RADIATION EXPOSURE MAN-REM (Cont'd)' UNIT 1984 1985 1986 1987 SAN ONOFRE 3 267 412 232 SEABROOK SEQUOYAH 1 559 536 210 SEQUOYAH 2 559 536 210 SHEARON HARRIS 1 SHOREHAM SOUTH TEXAS 1 ST. LUCIE 1 632 672 246 47S ST. LUCIE 2 632 672 246 413
, SUMER 295 379 23 560 SURRY 1 1124 908 1178 ~356 SURRY 2 1124 908 1178 356 SUSQUEHANNA 1 308 1106 414 311 SUSQUEHANNA 2 414 311 THREE MILE IS. 1 344 429 613 149 TROJAN 433 363 381 363 <
TURKEY POINT 3 628 627 473 686
, TURKEY POINT 4 628 627 473 686 i VERMONT YANKEE 603 1051 1188 303 V0GTLE 1 l WASH. NUCLEAR 2 136 222 406 i WATERFORD 3 223 156
- WOLF CREEK 142 134 YANKEE-ROWE 348 211 45 217 ZION 1 393 583 249 346 ZION 2 393 583 249 346 I
4 1 i i m 4 A-40 i
1ABLE 17 COMPLETED SHUT 00WNS REQUIRED BY TECHNICAL SPECIFICATIONS 1984 1985 1986 Number of Shutdowns 57 73 65 j Shutdowns Per 10,000 Critical Hours 0.09 0.13 0.11 Number of Plants Affected 41 41 44 Total Technical Specification Shutdown Hours 11,205 7,674 10,558 , Average Hours Shut Down 220 105 165 A-41
i ' d 1 y k TABLE 18 StM4ARY OF SYSTEM INVOLVEMENT IN LCO VIOLATIONS 1 i'. Contribution To Total Number of . System LCO Violations L] L 1984 1985 1986 Radiation Monitoring 61/442 54/714 86/800 8 h Fire Detection 47/442 135/714 129/800 [ Containment Isolation Control 45/442 51/714 57/800 [j Fire Protection (Passive) 43/442 44/714 38/800 0( y l Reactor Coolant (PWR) 21/442 * * : l l ! Control Rod Drive 29/714 1 Main / Reheat Steam *
- 27/800 1
- This system was not among the top five contributors to system involvement in LCO violations for the year shown.
t A-42
I l APPENDIX B SUM 4ARY OF 1987 ABNORMAL OCCURRENCES
Abnormal Occurrences CY 1987 Report No. A0 Criterion Title of A0 NUREG-0090 or Example Event Description A0 # NRC Order Suspends Vol. 10, No. 1 A-10 On March 31, 1987, the NRC issued an Order 87-1 Suspending Power Operation and Order to Power Operations and A-11 Show Cause (Effective Immediately) to of Peach Botton Philadelphia Electric Company. The Order facility Due to directed the licensee to place Peach Bottom Inattentiveness Unit 3, operating at about 100% power at the of the Control time, in cold shutdown (Unit 2 was already in Room Staff cold shutdown for refueling) and maintain both Units in cold shutdown pending further Order. The Order was based on the fact that at times during various shifts one or more of the Peach Bottom operations control room staff (including licensed operators, senior licensed operators, co J, and shift supervisors) had for at least five months periodically slept or had been other-wise inattentive t0 licensed duties. In addition, plant management either knew of or condoned this inattent!veness, or should have known of these facts, and either took no action or inadequate action to correct this situation. Prior NRC inspections had identified other instances of inattention to duty or failure to adhere to procedures on the part of licensed operators in the control room at Peach Bottom. 87-2 Diagnostic Vol. 10, No. 1 G In a January 6, 1987 letter, Allegheny Valley Medical Hospital, Natrona Heights, Pa, notified NRC Region 1 that on November 21, 1986, patient Misadministration received an intravenous dose of 100 millicuries of technetium-99m rather than the prescribed dose of 20 millicuries. Estimated doses to various organs of the patient were: stomach
Abnomal Occurrences CY 1987 (Continued) Report No. A0 Criterion A0 # . Title of A0 NUREG-0090 or Example Event Description 87-2 (Continued) wall, 25 rads; thyroid,13 rads; intestinal wall, 6-7 rads; and bladder wall, 5 rads. These doses are about five times those which would have been expected had the prescribed j doses been administered. No significant health effects are expected by the licensee. 87-3 Diagnostic Vol. 10, No. 1 G On January 21, 1987, NRC Region IV was notified Medical by St. Anthony Hospital, Oklahoma City, Okla. Misadministration that on January 12, 1987, a 15 year old female was administered 400 microcuries of I-131 rather than the prescribed dose of 400 oo microcuries of I-123, resulting in a thyroid f, dose of about 1490 rads. This may result in a small increased risk of reduction in thyroid function, and a small increased risk of latent thyroid cancer. 87-4 Diagnostic Vol. 10, No. 1 i Medical G In a le:'er dated March 2, 1987, the NRC Misadministration received written notification that on February 19, 1987 a patient referred to the Nuclear Medicine Department of the University of Massachusetts Medical Center in Worchester, Mass. , received a 5.5 millicurie dose of iodine-131 rather than the prescribed 5.0 microcuries. Based upon the 24 hour uptake and the measured effective half-life, the licensee estimated that the radiation dose to the patient's thyroid was 730 rads and the total body dose was 1.7 rads. The effect on the thyroid, if any, would be of no importance because prior to the event, the patient was scheduled for a thyroidectomy to be performed in March.
Abnormal Occurrences CY 1987 (Continued) Report No. A0 Criterion NUREG-0090 or Example Event Description A0 # Title of A0 Vol. 10, No. 1 A-11 On March 17, 1987, the NRC issued a Notice of 87-5 Significant Violation and Proposed Im m sition of Civil Breakdown in Management Penalty in the amount of $10,000 (later reduced i Oversight and to $7,500) to Radiation Sterilizers, Inc. of l Control of Menlo Park, Cal. The violations were found at Radiation Safety the licensee's irradictior facilities in Program at Two Schaumburg, Ill. and Westerville, Ohio. Some of the violations related to unsafe practices of a Licensee's which could have resulted in serious overexpo-Irradiator sures of licensee personnel. The base civil C
.acilities penalty for the violations would be $5,000.
However, this was escalated because of: the licensee's p for knowledge of the problems; the licensee's failure to take prompt and effective
, corrective measures for previously identified J, violations; and the duration of some of the violations (some had existed for several months).
Diagnostic Vol. 10, No. 1 G On April 27,1%7, NRC Region IV was notified 87-6 by Veterans Administration Medical Center, Medical Boise, Idaho, that on April 1, 1987, Misadministration 400 microcuries of I-131 was administered to an adult male for a total body scan; on April 6, 1987, it was discovered that a bone scan using technetium-99m was the desired study. The licensee calculated that the patient received a whole body and thyroid dose of about 0.47 and 400 rads, respectively. The physician user evaluated the exposure and concluded that the irradiation posed a small, but still significant, risk of reduction in thyroid function.
Abnormal Occurrences CY 1987 (Continued) Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Event Description 87-7 Significant Break- Vol. 10, No. 1 A-11 On ?.pril 1, 1987, the NRC issued a Demand for down in Management ) Ovetsight and Con- Infora3 tion and Notice of Violation and Pro-posed Imposition of Civil Penalties to Grede ] trol of Radiation Foundries, Inc., Milwaukee, Wis. This action j Safety Program at was taken after an October 1986 inspection i an Industrial Radiog- showed a significant breakdown in the raphy Licensee licensee's oversight and control of its radia-
! tion safety program. The inspection showed not only that the Radiation Safety Officer was not familiar with NRC requirements for the training, but also that an unqualified / untrained l
~ radiographer made 43 radiographic exposures a m on August 6, 7, and 8, 1986, which was in viola-a tion of NRC requirements and contrary to the conditions of Grede's license. In addition, i the individual made the exposures with the knowledge of an authorized radiographer, who in turn entered the information into a log i and signed off on it as though he had made i the exposures himself. 87-8 Significant Break- Vol. 10, No. 1 A-11 On April 10, 1987, the NRC issued an Order down of Management Temporarily Suspending License (Effective Controls for Radio- Immediately) and Order to Show Cause why the i graphic Operations ! license should not be reycked to A-1 Inspec- ) tion, Incorporated, of Evanston, Wyo. The Order was based on NRC inspections which identified two instances where the licensee permitted unauthorized individuals to conduct radiography. In one instance, the licensee stated to an NRC inspector that he had not employed such individuals to conduct radiag-raphy while later he admitted to an investi-gator thrt he had. These actions inoicated a l
Abnormal Occurrences CY 1987 (Continued) Report No. A0 Criterion AD # Title of A0 NUREG-0090 or Example Event Description 87-8 (Continued) disregard for requirements and lack of reason-able assurance that the licensee could be trusted in the future. AS87-1 Breakdown in Vol. 10, No. 1 A-11 On February 17, 1987, the Arizona Radiation Management and Regulatory Agency issued an order to U.S. Test-Procedural Controls ing Company, Unitech Services Group, San at an Industrial leandro, Cal., to cease all radiographic Radiography Licensee operations within the state of Arizona. Ti.e (Agreement State order was issued based ca the findings of an Licensee) inspection performed on February 6 and 7, 1987, to investigate the circumstances associated with two employees (a radiographer and an assistant radiographer) of the licensee co receiving radiation exposures in excess of .' regulatory limits while performing radiographic operations at the Navajo Generating Station, Page, Ariz. The licensee had not properly trained the radiographers. AS87-2 Breakdown in Vol. 10, No. 1 A-11 On February 27, 1987, an Emergency Order Management and suspending all radiographic operations was Precedural Controls issuet ?y an inspector for the California at an Industrial Departaent of Industrial Relations to Radiography Continental Testing and Inspection (CTI), Licensee (Agree- Signal Hill, Cal. During a routina compliance ment State Licensee) inspection of CTI's licensed radiographic operations, it was determined that individuals acting as radiographers may have lacked the required training and experience, since sub-stantiating records were not available for inspection.
l i Abnormal Occurrences CY 1987 (Continued) ! Report No. A0 Criterion
- AD # Title of A0 NUREG-0090 or Example Event Description i
j 87-9 Diagnostic Medical Vol. 10, No. 2 G On January 21, 1987, a 66 year-old female at
- Misadministration Halifax-South Boston Community Hospital, South
- Boston, Va., received 782 microcuries of I-131 j
instead of a 100-microcurie dose usually given for a thyroid scan. No adverse effects to the patient are expected from the reported mis-i 4 administration. The dose to the whole body was estimated as 0.37 rem and a thyroid tissue dose of 625 rem. I 87-10 Therapeutic Vol. 10, No. 2 G From April 20-22, 1987, a patient treated on Medical Misadmin- the cobalt-60 teletherapy unit at St. Peter's istration Medical Center, New Brunswick, N.J., received
- , a radiotherapy administration of 600 rads to a the lumbar spine area, which was not the prescribed treatment site. The patient's l referring physician and radiotherapist con-cluded that the dose would have no detrimental i clinical effect due to the patient's current disease state (i.e., breast cancer with I
metastasis to the bone).
- 4 87-11 Diagnostic Medical Vol. 10, No. 2 G On June 3, 1987, NRC received written notifica-Misadministration tion that on May 20, 1987, a patient at the National Institutes of Health, Bethesda, tid. ,
received 120 millicuries of tecnnetium-99m 1 pertechnetate rather than the prescribed radio-pharmaceutical, 10 millicuries of gallium-67 citrate. The patient experienced no adverse effect from this misadministration but received the following unwarranted approximate organ doses (rads): bladder wall,10.2; stomach ! wall, 6.1; upper large intestinal wall, 14.4; lower large intestinal wall, 13.2; thyroid, , j 15.6; and red marrow, 2.0. ' J l
l Abnormal Occurrences CY 1987 (Continued) Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Event Description 87-12 NRC Order Issued Vol. 10, No. 2 A-11 On June 15, 1987, an Order Modifying License, to Remove a Hos- Effective Immediately, was issued to Milford , l pital's Radiation Memorial Hospital, Milford, Del. The action j Safety Officer was based on (1) the falsification of daily constancy checks of the dose calibrator by the licensee's two technologists, and (2) the falsification of records of Radiation Safety Committee meetings by the Radiation Safety Officer for about 15 years. The consequence I f of these occurrences was a reduction in the level of safety associated with the use of l licensed material by this licensee. No ( specific hazard was identified. Significant Break- Vol. 10, No. 2 A-11 On June 17, 1987, the NRC issued an Order
$ 87-13 down in Management Modifying License (Effective Immediately) to and Procedural Con- United States Testing Company, Inc., Unitech trols at an Indus- Services Group (USTU), San Leandro, Cal.,
trial Radiography which required the licensee to temporarily Lice...ee cease all operations until certain specified corrective actions were taken. At the time, USTU was licensed by the NRC and several
~
Agreement States tc perform industrial radiog-raphy. In-depth special safety inspections identified numerous radiation safety viola-tions, including (1) allowing individuals to perform radiography after failing one or more certification examinations, (2) allowing individuals to perfonn radiography before all training and examinations were completed, and (3) allowing individuals with expired certifi-cations to perform radiography.
I i Abnormal Occurrences CY 1987 (Continued) 4 ! Report No. A0 Criterion Title of A0 AD # NUREG-0090 or Example Event Description AS87-3 Radiographer Over- Vol. 10, No. 2 A-1 On December 9,1986, an industrial radiographer exposures (Agree- and a radiographer's assistant, employed by ment State Licensee) Northwest X-ray, Idaho Falls, Idaho, received ovarexposures while performing radiography in i a multi-level hot cell at the Chemical Process-ing Plant at the Idaho National Engineering Laboratory (INEL) near Idaho Falls. ! The assistant received a dacumented exposure j of 3.4 rem whole body, and INEL-estimated ex-j posures of 6 rem to the lens of the eye, 5 rem j to the left hand, and 20 rem to the right 2 hand. The radiographer received documented } ? exposure of 7.8 rem whole body and INEL- ] g estimated exposures of 50 ren to the lens
- of the left eye, 70 rem to the lens of the right eye, and entrance doses of .?000 and 1700 rem to the left and right rands, respec-tively. Both individuals were examined by i
INEL's Medical Director. No s'agns of injury l were found. The assistant was released and 1 the radiographer will be followed medically j for several months. I ! 87-14 Significant Degra- Vol. 10, No. 3 B-5 and On April 24, 1987, while the reactor was being i dation of Plant A-ll shut down, personnel errors resulted in a Safety at Oyster condition which could have resulted in Creek containment failure had a loss of coolant l accident (LOCA) occurred. With reactor power at about 23%, the licensee began to purge tSe i containment nitrogen atmosphere so that entry i l could be made into the drywell. In order to accelerate the deinerting process, the group L--_v _ _ , - . . _=y,, , _,-- -_-...,,,,-,_,._-,---c- . , - - - - - - = , - - , . - . - _ - - - - - , - _ - . , - _ _ , _ _ . . _ _ _ _ _ _ _ _ _ _ _ _
Abnormal Occurrences CY 1987 (Continued) Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Event Description 87-14 (Continued) shift supervisor authorized the blocking open of the torus-t7 drywell vacuum breaker valves. However, this rendered the containment vulner-able to steam bypass of the suppression chamber, potentially resulting in containment over-pressurization for small, intermediate, and large LOCAs. Furthermore, blocking open of the suppression chamber-drywell vacuum breakers resulted in the plant being in an unanalyzed condition. 87-15 Steam. Generator Vol. 10, No. 3 8-2 At approximately 6:35 a..a. on July 15, 1987, S' Tube Rupture at North Anna Unit I was manually tripped from C North Anna Ui.it 1 100 percent power due to indications of a steam generator tube rupture. Inspection showed that a tube had failed over 360 degrees i of its circumference, and the fractured ends j were displaced in the axial direction approxi- j nately one-half inch. The cause was found to be due to fatigue. It is estimated that a total of 1.59 X 10 1 curies was released, which consisted primarily of radiogases. There was no detectable increase in normal background levels of radioactivity at the site boundary in the affected sector (s). The release was less than 1% of Technical Specification limits.
Abnormal Occurrences CY 1987 (Continued)
- Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Event Description i
87-15 (Continued) The primary-to-secondary leak in this event
- was estimated to be between 550 to 637 gallons per minute (gpe). The North Anna Updated Final Safety Analysis Report estimated that a double-i ended rupture of a sir.gle tube at full power
; would result in a flow rate of 710 gpe. The ; highest flow rate in the 1982 5/G tube rupture i
at Ginna was estimated to be 760 gpe. l This event was investigated by an NRC Augmented l Inspection Team. l' 87-16 Therapeutic Vol. 10, No. 3 G On August 24, 1987, the NRC was notified that 9' Medical Misad- a 75 year-old patient at Parkview Memorial R; ministration Hospital, Fort Wayne, Ind., received two therapeutic radiation exposures to the wrong i part of the body. The patient was scheduled to receive radiation therapy exposure of ) 250 rads per exposure per day to the right hip, for a total of 3000 rads. However, two i exposures of 2a0 rads each was administered before it was discovered that the left hip was being treated instead of the right hip. The patient was examined by a physician and no i medical side effects were noted as a result of the misadministration. l 4
l Abnormal Occurrences CY 1987 (Continued) Report No. A0 Criterion A0 i Title of A0 NUREG-0090 or Example Event Description 87-17 Failure to Report Vol. 10, No. 3 A-11 On August 24, 1987, the NRC issued an Order to Diagnostic Medical Show Cause Why the License Should Not Be Misadministrations Modified to the Edward Hines, Jr. , Veterans Administration Hospital, Hines, Ill., l l directing that a hospital staff member be l removed from NRC-licensed activities and l l that the hospital take certain steps to improve its control over its nuclear medicine program. An NRC investigation determined that the Assistant Chief Physician of the Hospital's Nuclear Medicine Service: failed to ensure 5' that two diagnostic misadministrations of ll radioactive pharmaceuticals were reported to the NRC a< required; made a false statement to a Veterans Administration Investigatory Board and to NRC investigators; destroyed evidence; and attempted to impede the NRC investigation by influencing the testimony of a witness. 87-18 Suspension of a Vol. 10, No. 3 A-11 On September 8, 1987, the NRC issued an Well Logging inimediately effective order to Log-Tec of Company's License Cleveland, Okla. , that suspended the NRC license, ordered all byproduct material be placed in locked storage, and ordered the licensee to show cause why the license should not be revoked. During August 1987, an inspection showed several apparent violations associated with use and
_ - - - .=_. _ _ _ - -__ - - _ . . - 1 i Abnormal Occurrences CY 1987 (Continued) j
- Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Event Description 87-18 (Continued) possession of sealed radioactive sources.
4 When these violations we m discussed with the I company's sole proprietor, the NRC inspector was told that the sources had not been used since about June cf 1986. Later, when confronted with evidence to the contrary, he not only admitted the violations, but also that he had used the sources af ter June of 1986. 3 Subsequently, in respucse to the licensee's l request, the NRC is terminating the license. ! 87-19 Suspension of an Vol. 10, No. 3 A-11 On September 21, 1987, tha NRC issued an j , Industrial Radiog- Order Suspending License (effective immediately)
,, raphy Company's to Finlay Testing Laboratories, Inc., Aiea,
- License Hawaii. The Order required the licensee to suspend all activities authorized by the i
license and to place all byproduct material in the licensee's possession in locked storage. During inspections and investigations conducted I in September 1987, it was determined that i contrary to NRC and DOT regulations, licensee employees had placed a radiographic exposure } device containing radioactive material in lug- ] gage which was loaded and transported on 1 commercial passenger and military cargo / passenger ) aircraft. It was further noted that: licensee i personnel failed to prepare and use required j shipping papers and labels for these shipments; j and licensee representatives (including the 4 Radiation Safety Officer) had failed to maintain ] required records of licensed activities.
Abnormal Occurrences CY 1987 (Continued) Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Event Description AS87-4 Hospital Contami- Vol. 10, No. 3 A-ll and G On the morning of June 2, 1987, an 87 year old nation Incident patient at Buffalo General Hospital, Buffalo, New York, was administered a 200-millicurie therapy dose of iodina-131 in the hope of relieving esophageal compression caused by metastatic thyroid carcinoma. The patient had had a total thyroidectomy in April 1987, and had a gastrostomy tube and a foley catheter in place. On the evening of June 3,1987, approxi-mately 34 hours after receiving the dose, the patient had a cardiopulmonary arrest and i expired. During an attempt at resuscitation I in the patient's room by sixteen staff members, oo which included insertion of a pacemaker, con-
. taminated blood and urine were spilled and no l surveys of the clothing of those present were done.
Even though the contamination was extensive, subsequent thyroid bioassay showed no uptakes by involved staff, and the highest personnel monitoring badge reading was 30 millirem for one of the nurses. AS87-5 Therapeutic Vol. 10, No. 3 G On August 5,1987, the New York State Depart-Medical Misodmin- ment of Health, Bureau of Environmental istration Radiation Protection (State Agency) was notified of a series of therapeutic medical misadministrations to patients at Northern Westchester Medical Center, Westchester County, New York. l l l 1 l
u _. _a a __a%-.ae.__ _Am +.__L & -A- ',___E -e _c-m.. _- - _.b---.-en _- ,_ ,_* - 4 a - _ - m. M.-__ __ ___w a. _ _ _ .u a.._t 4 1. m 4.___ _ _.__ _ _._n _ -. + - - l Abnormal Occurrences CY 1987 (Continued) Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Event Description 1 I AS87-5 (Continued) It was found that a dosimetrist, employed by a
! consulting group for the hospital, had made .'
numerous serious errors in calculating cob &l.t teletherapy treatment times for patients. There were 22 cases in which the therapy doses , delivered to patients differed from the pre- , scribed doses by more than 10% (this included overtreatments as well as undertreatments). The largest error found was an adr .7istered j dose that was about 2.4 times the pres ribed j dose. Some patients receiving overtreatments
- had exhibited physical symptoms apparently due j to the exposures.
T i 5 87-20 Suspension of Vol. 10 No. 4 G On October 30, 1987, the WRC issued an Order
- License of an Suspending License (Effective Immediately) and
{ 011 and Gas Well Order to Show Cause why the license should not ] Tracer Company be revoked to Tracer Profiles, Inc. , of Oklahoma i City, Okla. I During March 1987, several violations of NRC ] requirements were found. Contacts with the company to satisfactorily resolve the viola-tions were not successful. Subsequently, it was fou.w! that the company vacated its offices
- and moved to a new and unknown location with-l out notifying the NRC. Therefore, the NRC is considering action to revoke the license.
i I 1 l .___ _ .._ - - - - -. _ - _ _ -_ .-_.~ , -, _. ,._,,_.- - . - _ _ _ _ _ ,_ __ -__._._ _ - _ _ ._ _.
i
. i
)
l l l l l APPENDIX C LISTING of AEOD REPORTS 1987 I 9 J
b APPENDIX C ! AE00 PEPORTS ISSUED IN 1987 l Case Studies ! Defe" Sub.iect No. Author 03/87 Air Systems Problems at U.S. Li0ht Water C701 H. Ornstein j Reactors j Reactor Engineering Evaluations Date Subject No. Author 3 01/22 Potential Containment Airlock Window E701 S. Israel l I Failure Due to Radiation 4 ' 03/19 MOV Failure Due to Hydraulic Lockup from E702 E. Brown ' j Excessive Grease in Spring Pack 03/20 Loss of Offsite Power Due to Unneeded E703 F. Ashe -
. Actuation of Startup Transformer Protection i Differential Relay , ? 03/26 Discharge of Primary Coolant Outside of E704 S. Israel
- Containment at Pb.s While on RP.R Cooling v
l 03/31 RWCU System Automatic Isolat.in & Safety E705 N. Thomasson ! Considerations l l l 03/31 Inadequate Mechanical Blocking of Valves E706 T. Cintula l
! 03/31 Design and Construction Problems at E707 C. Hsu i 1 Operating Nuclear Plants i 1
t 08/13 Depressurization of Reactor Coolant Systems E708 S. Israel j at PWRs 08/24 Auxiliary Feedwater Pump Trips Caused by E709 C. Hsu , Low Suction Pressure l 10/L/ Inadequate NPSH in Low Pressure Safety E710 S. Israel ; i Systems in PWRs ! L k ) ,
, i 6 ; C-3 l - _ - , - - _ , _ . _ . . _ _ _ . . - . _ . _ _ _ _ _ - . - - , _ _ _ _ _ , _ _ _ _ , _ _ _ _ _ - . --___.._..,,-,_,m-- _ _ _ .
l l { Nonreactor Engineering Evaluations Date subiact No. Author 01/14 Diagnostic Mi'iadministrations Involving N701 S. Pettijohn the Administration of Millicurie Amounts of Iodine-131 i 03/20 Medical Misadninistrations Re]orted to NRC N702 S. Pettijohn i for the Period January 1986 tirough December 1986 03/27 Report on the 1986 Nonreactor Events N703 K. Black l Program Support Reports Date Subject No. Author 09/14 Trends and Patterns Program Ranort - P701 G.L. Plumlee i Operational Experience Feedback on Main Feedwater Flow Control and Main Feedwater Flow Bypass Valves and Valve Operators 07/87 Operational Experiences at Newly Licensed NUREG R. Dennig l Nuclear Power Plants 1275 I Vol. 1 Technical Review Reports l Date Subject No. Author 1 1 01/29 Compression Fitting Failures T701 H. Ornstein 03/10 Leaking Pulsation Dampener Leads to Loss T702 T. Cintula of Chargiag System 03/26 Potential for loss of Emergency Feedwater T703 H. Wegner Due to Pump Runout During Certain Transients 03/27 Pressurizer' Code Safety Valve Reliability T704 M. Wegner 05/20 Occurrence of Events Involving Wrong Unit / T705 E. Trager l Wrong Train / Wrong Component - Update I through 1986 l C-4 l l
Technical Review Reports (Continued) Date Subject M. Author 06/15 Recent Events Involving Turbine Runbacks T706 E. Leeds at PWRs 08/05 Undetected Loss of Reactor Water T707 S. Israel 08/05 ProblemswithHighPressureSafetyInjectionT708 5. Israel Systes.s in Westinghouse PWRs 10/22 Recent New Plant Operational Experience T709 T. Wolf l ! 11/13 Heating, Ventilating and Air Conditioning T710 H. Chiramal System Problems 11/13 Review of Data on Teletherapy Misadmini- T711 S. Pettijohn strations Reported to the State of New York that were the Subject of PN0-1-87-74A 11/19 Unplanned Criticality Events at U.S. Power T712 T. Wolf Reactors Similar to that at Oskarshamn Unit 3 on 07/30/87 12/03 Hispositioning of "Reverse Acting" Valve T713 J. P. Stewart Controllers 12/01 Distribution of Information Notices and T714 S. Pettijohn Other "Mass Mailing" Information to Licensees that have Users at locations Remote from the Headquarters locations Special Study Reports Date Subject No. Author 05/87 AEOD Annual Report for 1986 5701 S. Massaro 05/18 loss of Decay Heat Removal Function at 5702 H. Ornstein Pressurized Water Reactors with Partially Drained Reactor Coolant Systems 10/22 Radiation Overexposure Events Involving 5703 S. Pettijohn Industrial Field Radiography 12/87 Operating Experience Feedback Report - HUREC H. Ornstein Air Systems Problems Comercial Power 1275 Reactors Vol. 2 C-5
APPENDIX D LISTINGS OF AEOD REPORTS 1980 - 1986 1 h 6 i l t i l l l
i i TABLE D-1 REPORTS ISSUED IN CY 1986 ; Case and Special Studies Date Sd3ect No. Author l S. Pettijohn i 8/86 Rupture of an Iodine-125 Brachytherapy C601 l Source at the University of Cincinnati , Midical Center l j 8/86 Operational Experience Involving Turbine C602 C. Hsu ! Overspeed Trips 12/86 A Review of Motor-Operated Valve C503 E. Brown ! Performance ! 12/86 Effects of Ambient Temperature on C604 M. Chiramal ! Electronic Components in Safety-Related : Instrumentation and Control Systems ( 12/86 Operational Experience Involving Losses C605 F. Ashe i of Electrical Inverters i i 1/86 Trends and Patterns Program Plan - P601 R. Dennig i FY86-FY88
! 8/86 Trends and Patterns Report of Unplanned P602 L. Bell l i Reactor Trips at U.S. Light Water j ! Reactors in 1985 ;
8/86 Trends and Patterns Report of Engineered P603 H. Harper l f Safety Feature Actuations at Commercial l U.S. Nuclear Power Plants !
; r I. 8/86 Trends and Patterns Report of the P604 T. Wolf I
l Operational Experience of Newly Licensed j U.S. Nuclear Power Reactors ! ' i 4/86 AE00 Annual Report for 1985 5601 J. Heltemes f
; 5/86 An C<erview of Nuclear Power Plant 5602 J. Crooks e i
Operating Experience Feedback rograms i I
! 6/86 Adequacy of the Scope of IE B.11etin 5603 E. Leeds !
j 86-01 l
< l l i; l
4 I I 0-3 f 1 ;
TABLE D-1 (Continued) Reactor Engineerina Evaluations and Technical Reviews Date Subject No. Author 1/9/86 Deficient Operator Actions Following E601 F.. Leeds Dual Function Valve Faili.res 1/15/86 Unexpected Criticaily Due to Incorrect E602 k. Freemen Calculation and Failure to Follow Procedures E/19/86 Delayed Access to Safety-Related Areas E603 T. Cintula Durit,C Plant Operation 3/14/86 Spurious System Isolations Caused by the E604 F. Leeds Panalarm Model 86 Thermocouple Monitor 4/28/86 Lightning Events at Nuclear Power Pla-ts E605 M. Chiramal 5/27/86 Loss of Safety Injection Capability at E606 R. Tripathi Indian Point Unit 2 5/27/86 Core Damage Precursor Event at Trojan E514 D. Zukor , (Revision 1) 7/3/86 Degradation or loss of Charging Systems E607 F. Ashe with Swing Putop Designs 7/14/86 Reexamination of Water Hammer E608 E. Leeds l Occurrences ! 8/8/86 Inadvertent Drai'ing of Reactor Vessel E609 P. Lam During Shutdown e aling Operation 8/14/8G Less of Low Pressure Coolant Injection E610 E. Leeds Loop Selection logic at Millstone Unit 1 10/16/86 Deficiencies in Seismic Anchorage for E611 N. Thomasson Electrical and Control Panels 12/17/86 Emergency Diesel Generator Component E612 C. Hsu Failures Due to Vibration 12/13/86 Localized Rod Cluster Control Assembly C613 E. Brown Wear at PWR Plants 1/27/86 Pressure Sensitive Temperature Switch T601 T. Cintula Results in Spurious Actuation of Fire Suppression System 4/29/86 Emergency Diesel Generator Cooling T602 E. Leeds Water System Design Deficiencies at Maine Yankee and Hadda, Neck 0-4
TABLE D-1 (Continued) Reactor Engineerina Evaluations and Technical Reviews (Continued) , Date Subject No. Author 4/30/86 Inadvertent Pump Suction Transfer and T603 R. Tripathi Potential Auxiliary Feedwater Pump Cavitation at Davis-Besse 5/7/86 Events Resulting from Deficiencies in T604 E. Trager Labeling and Identification Systems l 6/17/86 Failure of Main Steam Safety Valves T605 R. Freeman to Properly Reseat 8/7/86 Inadvertent Recirculation Actuation T606 T. Cintula Signals at Combustion Engineering Plants 9/19/86 Occurrence cf Events Involving Wrong T607 E. Trager Units / Wrong Train / Wrong Component - Update through June 1986 11/17/86 Hydrogen Fire and Failure of Detection T608 M. Chiramal System 12/16/86 Foreign Material and Debris in Safety- f609 E. Leeds Related Fluid Systems 12/19/86 ADS /RCIC System Interaction Events at T610 E. Leeds River Bend Unit 1 12/19/86 Denied Access Due to Negative Room T611 T. Cintula Pressura 12/31/86 Degradation of Safety Systems Due to T612 R. Tripathi l Component Misalignment and/or Mispositioned Control / Selector Switches D-5
TABLE 0-1 (Continued) Nonreactor Engineering Evaluations Date Subject No. Author 6/17/86 Report on 1985 Nonreactor Events and N601 K. Bldek Five-Year Assessment for 1981-1985 6/25/86 Medical Misadministrations Reported for N602 S. Pettijohn 1985 and Five-Year Assessment of 1981-1985 Reports Incident Investiaation Program Reports Date Subject Designation 1/86 Loss of Power and Water Hammer Event NUREG-1190 at San Onofre, Unit 1, on November 21, 1985 2/86 Loss of Integrated Control Systcm NUREG-1195 Power and Overcooling Transient at Rancho Seco on December 26, 1985 8/86 Incident Investigation Manual -- 12/86 Incident Investigation Manual - -- Revsion 1 I l D-6 _ _ - ____________--_______- _-____-________A
l k TABLE 0-2 REPORTS ISSUED IN CY 1985 Case and Special Studies Date Subject No. Author 6/85 Safety Implications Associated with C501 H. Ornstein In-Plant Pressurized Gas Storage and Distribution Systems in Nuclear Power i Plants ; 9/85 Overpressurization of Emergency Core C502 P. Las Cooling Systems in Boiling Water , Reactors , 1 12/85 Decay Heat Removal Problems at U.S. C503 H. Ornstein ! 1 Pressurized Water Reactors ! i 12/85 Loss of Safety System Function Events C504 E. Trager j t
- 12/85 Therapy Misadministrations Reported to C505 S. Pettijohn !
$ HRC Pursuant to 10 CFR 35.42 - a i,
, 6/85 Trends and Patterns Analysis of 1981 P502 B. Brady !
; through 1983 LER Data (NUREG/CR4129) {
7/85 Feedwater Transient Incidents in P501 R. Dennig f Westinghouse PWRs j 8/85 Engineered Safety Feature Actuations at P503 T. Wolf ! Commercial U.S. Nuclear Power Reactors t l January 1 through June 30, 1984 j
.i 8/85 Trends and Patterns Report of Unplanned P504 L. Bell
- 1 Reactor Trips at U.S. Light Water ,
; Reactors in 1984 !
l
- 3/85 Review of Operational Experience From S501 D. Zukor I Ncn-Power Roactors ,
~ i 4/85 AE00 Semiannual Report for July-Decereber 5502 J. Heltemes i , 1984 ; i 9/85 Evaluation of Recent Valve Operator Motor $503 E. Brown ; 4 Burnout Events . I i i
- i i r
} f i i i 1 0-7 !
TABLE D-2 (Continued) l Reactor Engineerina Evaluations and Technical Reviews , Date Subject No. Author i 1/17/85 'k. tor Operated Valve Failures Due to E501 M. Chiramal ; Hammering Problem 1/25/85 Failure of Residual Heat Removal E502 C. Hsu , Suppression Pool Cooling Valve to l l Operate j i 3/4/85 Partial Failures of Control Rod Systems E503 M. Chiramal
! to Scram i 1
a 3/29/85 Loss or Actuation of Various Safety- E504 F. Ashe ; i Related Equipmer,t Due to Removal of 1 Fuses or Opening of Circuit Breakers ; I 3/29/85 Service Water System Air Release Valve E505 S. Salah !
; Failures '
1 ! 5/13/85 Valve Stem Susceptibility to inter- E506 C. Hsu ; granular Stress Corrosion Cracking ; -] Due to Improper Heat Treatment ; i I l 5/17/85 Electrical Interaction Between Units E507 M. Chiramal ! 3 During Loss of Offsite Power Event of ! August 21, 1984 at McGuire Units 1 i and 2 l 5/24/85 Nuclear Plant Operating Experience E508 S. Rubin ! Involving 3afety System Disturbances Caused By Bumped Electro-Mechanical 4 Components 1 ! 7/25/85 Salem Unit 2 Depressurization Event E509 R. Freeman j 7/30/85 Disabling of a Shared Diesel Generator E510 F. Ashe . Set Due to Electrical Power Supply
- Arrangement for Support Auxiliaries 8/9/85 Closure of Emergency Core Cooling System Eb11 E. Leeds Minimum Flow Valves 9/4/85 Failure of Safety-Related Pumps Due to E512 R. Freeman
! Debris ] 9/16/85 High Pressure Core Spray System Relief E513 S. Salah j Valve Failures 1 10/8/85 Core Damage Precursor Event at Trojan E514 D. Zukor D-8
TABLE D-2 (Continued) Reactor Engineering Evaluations and Technical Reviews (Continued) Date Subject No. Author 12/11/85 Inadvertent Actuation of Safety System E515 M. Chiramal i Due to Cross-Talk j 1/22/85 Failure of Automatic Protection for T501 R. Freeman l Boron Dilution Event at Callaway i Unit 1 l 3/18/85 Comparative Analysis of Recent Feedline T502 E. Leeds [
- Water Hammer Events at Maine Yankee, '
Calvert Cliffs, Palisades, and Salem i , l 5/2/85 Pressurizer Level Instrumentation of T503 M. Chiramal i Combustion Engineering Reactor Units 4 I 5/17/85 Loss of Instrument Air and Subsequent T504 R. Freeman ! Pressure Transient at Callaway Unit I r 7/17/65 Beaver Valley Component Cooling Water T505 C. Hsu i
, Pump Damage l 7/25/85 Primary System Release Due to Pres- T506 T. Cintula ( ) surizer Degas Relief Valve Lifting j
- 8/13/85 Standby Liquid Control System Pressure T507 E. Brown i l Relief Valves Lift at a Pressure Lower
! than Reactor Coolant Pressure f I l ! 8/14/85 Browns Ferry Nuclear Plant High T508 E. Leeds ! ! Pressure Coolant Injection System ,
J Performance Assessment ; 9/29/85 Inadequate Surveillance Testing Pro- T509 F. Ashe l cedures for Degraded Voltage and 1 Undervoltage Relays Associated with ; a 4160-Volt Emergency Buses !
! ?
9/4/58 Xenon Induced Power Oscillations at T510 R. Freeman i ). Catawba l I 9/16/85 Technicians Perform Work on Wrong T511 E. Trager ] Control Rod Drive Mechanism 10/24/85 Incorrect Plugging of Steam Generator T512 R. Freeman Tubes 11/7/85 Flooding of Safety-Related Valves in T513 D. Zukor ; Pits l l i i D-9
TABLE 0-2 (Continued) Reactor Engineering Evaluations and Technical Reviews (Continued) Date Subject No. Author 11/25/85 Potential Loss of Component Cooling TS14 D. Zukor Water Due to Ha1 adjustment of Relief Valves 12/5/85 Residual Heat Removal Service Water T515 S. Salah Booster Pump Air Binding at Brunswick Unit 1 12/11/85 High Pressure Coolant Injection Over- T516 E. Trager speed Trip (oss Events and Subsequent Damage Oue t1 Water Hammer Nonreactor Engineering Evaluavions Date Subject No. Author 5/20/85 Summary of the Nonreactor Event Report N501 K. Black Data Base for the Period January-June 1984 6/28/85 Summary of the Nonreactor Event Report N502 K. Black Data Base for the Period July-December 1984 7/25/85 Report on Medical Misadministrations for N503 S. Pettijohn January 1984-December 1984 Incident Investigation Program Reports Date Subject Designation t 7/85 Loss of Main and Auxiliary Feedwater NUREG-1154 Event at the ' Davis-Besse Plant on June 9, 1985 D-10
TABLE 0-3 REPORTS ISSUED IN CY 1984 l Case and Special Studies Date subject No. Author , W. Lanning ! 3/84 Low Temperature Overprerdure Events at C401 Turkey Point Unit 4 ! 6/64 Operating Experience Related to Motsturr C402 M. El-Zeftawy i Intrusion in Electrical Equipment at Commercial Power Reactors i 5/84 Hatch Unit 2 Plant Systems Interacti6n C403 S. Rubin i Event on August 25, 1982 : i 7/84 Steam Binding of Auxiliary Feedwater C404 W. Lanning j Pumps ) 9/84 Breaching of the Encapsulation of Sealed C405 S. Pettijohn Well Logging Sources ; 1 i 2/94 Operating History Overview for Diesel P401 R. Dennig i Generators in Nuclear Service M. Chiramal ;
! 3/84 AE00 Trends and Patterns Program Plan P402 R. Dennig !
' I t 5/84 AE00 Trends and Patterns Evaluation P403 F. Hebdon , i Report, "Preliminary Assessment of LER i Reporting Under 10 CFR 50.73" } 3/64 LER Data on Personnel Errors P404 F. Hebdon i 11/84 Oraft Trends and Patterns Analysis of P405 M. Harper ! Feedwater Transients at Westinghouse 1 ) PWRs l
; 11/84 Trends and Patterns Analysis of Reactor P406 L. Bell t Scrams (Pilot Study) I
] 1/84 Human Error in Events Involving Wrong 5401 E. Trager l Unit or Wrong Train l i I
- 5. Rubin l 7/84 Pressure Locking of Flexible-Disk Wedge- S402
) Type Gate Valves a 6/84 Annual Report of U.S. NRC Participation 5403 J. Crooks ! in the Nuclear Energy Agency ?.ncident } j Reporting System Ouring 1983 f l !
- 0-11 !
s
TABLE D-3 (Continued) Case and Special Studies (Continued) Date Subject No. Author 6/84 Aaalysis of Foreign IRS Reports 5404 D. Zukor Submitted During CY 1983 9/84 Semiannual Report on AE00 Activities $405 J. Heltemes 10/84 Application of Risk Perspectives: A S406 P. Las Procedures Guide ; Reactor Enaineerina Evaluations and Technical Reviews Date Subject No. Author 1/4/84 Temporary Loss of All AC Power Due to E401 M. Chiramal Relay Failure in Diesel Generator Load Shedding Circuitry at Fort St. Vrain ! 1/'.0/84 Water Hawer in Boiling Water Reactor E402 S. Rubin High Prt:ssure Coolant Injection Systems , 1/17/84 Deficiency in Automatic Switch Company E403 F. Ashe ! (ASCO) Spare Parts Kits for Scram Pilot i Solenoid Valves i 2/28/88 Failures in the Upper Head Injection E404 D. Zukor System 3/22/84 Common Mode Failure of HPCI Steve Flow E405 M. El-Zeftawy Isolation Capability at Browns ?erry 3/22/84 Mechanical Snubber Failure E406 C. Hsu 3/26/84 Initiation and Indication Circuitry for E407 F. Ashe HighPressureCoolantInjectionSystems 3/27/84 Load Reduction Transient at the Salen E323 N. Trehan Unit 2 on January 14, 1982 (Revision 1) 4/13/84 Reversed Differential Pressure E408 S. Rubin Instrument Sensing 1.ines 5/16/84 Operating Experience Involving Air in E409 S. Salah Instrument Sensing Lines i l l 0-12
TABLE 0-3 (Continued) Reactor Engineerina Evaluations and Technical Reviews (Continued) Date subject No. Author l l Operational Experiences Involving 5/21/84 E410 F. Ashe Statedby Gas Treatment Systems Which Illustrate Potential Common Cause Failure or Degradation Mechanisms 5/22/84 Failure of Anti-Cavitation Device in E411 C. Hsu l Residual Heat Removal Service Water ! Heat Exchanger Outlet Valve { 5/25/84 Adverse System Interaction with Domestic E412 T. Cintula 1 Water Systems t ) W. Lanning I . 5/25/84 Natural Circulation in Pressurized Water E413 i j Reactors lP 9 l 5/31/84 Stuck Open Isolation Check Valve on the E414 P. Lam ( Residual Heat Removal System at Hatch t
- Unit 2
)
6/6/84 Overcooling Transient E415 E. Imbro - i l 6/11/84 Erosion in Nuclear Power Plantt E416 E. Brown f 7/2/84 Loosening of Flange Bolts on Residual E417 C. Hsu , Heat Removal Heat Exchanger Leading to ; Primary to Secondary Side Leakage 1 7/24/84 Feedwater Transients During Startup at E418 D. Zukor i 2 Westinghouse Plants 7/84 Fail"res of Fischer-Porter Transmitters E419 M. Chiramal l ! Used in Safety-Related Systems j i 8/23/84 Operational Experiences Involving E420 M. Chiramal Shorted Lamp Sockets of Indication Lights i 1 l
! 8/27/84 Loss of Pressurizer Heaters During E421 T. Cintula f l Precore Hot Functional Testing j ! 8/27/84 High Pressure Coolant Injection System E422 T. Wolf l l Performance at Hatch Units 1 and 2 ; ! l
) 9/20/84 Failure of Large Hydraulic Snubbers to E423 E. Brown !
] Lock Up !
I r ! 10/1/84 Failure of Anchor Bolt on Diesel E424 C. Hsu ! Generator Day Tank at Davis Besse Unit 0-13 { i l
TABLE 0-3 (Continued) Reactor Engineering Evaluations and, Technical Reviews (Continued) Date Subject No. Author 10/11/84 High Pressure Coolant Injection System E425 M. Chiramal Lockout at Vermont Yankee 10/24/84 Single Failure Vulnerability of Power E426 E. Imbro Operated Relief Valve Actuation Circuitry , for Low Temperature Overpressure Protection 11/6/84 Licensee Event Reports that Address E427 F. Ashe Situations Which Potentially Could Rc= ult in Overloading Electrical Equip.nent in the Emergency Power System or Prevent Operation of the Onsite Power System Sequencer 3/2/84 Failures of Containment Air Monitors T401 D. Zukor at Farley Units 1 and 2 3/22/84 Chemical Contamination of Primary and T402 M. El-leftawy Secondary Systems in Light Water Reactors 3/23/84 Setpoint Drift of Barton Model 288 T403 H. Chiramal ' Switches 4/13/84 Cable Fire and Loss of Control Power to T404 M. Chiramal Engineered Safeguards Valves 4/25/84 Cold Weather Events 1983-1984 T405 T. Cintula 4/25/84 Improper Spare Parts Procurement Event T406 T. Wolf at Grand Gulf Unit 4/30/84 Failure of 4 kV Circuit Breaker to Trip T407 M. Chiramal 5/7/84 Diesel Generator Inoperability Due to T408 M. Chiramal Overheating of Ventilation Cowling 5/1/84 Multiple Failure of Bell and Howell Dual T409 F. Ashe Potentiometer Modules Which Occurred at the Fort Calhoun Nuclear Station 5/1/84 Injection Valve for the High Pressure T410 E. Brown Coolant Injection System Failure to Open During a Surveillance Test 6/18/84 Contamination of the Nitrogen System at T411 H. El-Zeftawy Sacramento Municipal Utility District 0-14
TABLE D-3 (Continued) Reactor Engineering Evaluations and Technical Reviews (ContinueO Date subject No. Author l 6/18/84 Failure of an Access Door Between the T412 T. Wolf ; Drywell and the Wetwell 6/28/84 Failure of Fire Damper in Safeguards T413 W. Lanning Ventilation System 7/12/84 Station Operating Restrictions for Lost T414 F. Ashe or Out of Service Power Transformers Through which Electrical Power is Supplied to the Emergency Buses 7/17/84 Destruction of Charging Pump T415 W. Lanning 8/1/84 Loss of Enineered Safety Feature T416 D. Zukor Auxiliary Feedwater Pump Capability at TrojanonJanuary 22, 1983 8/2/84 Excessive Cooldown Rate Event at T417 S. Salah LaSalle Unit 1 8/6/84 Events Involving Fires or Other Related T418 F. Ashe Abnormalities in Motor Control Centers i
, with Aluminum Bus Bars 8/20/84 Contamination of Snubber Bleed Screw and T419 C. Hsu Lockup Poppet Valve ,
8/23/84 Failure of an Isolation Valve of the T420 P. Lam Reactor Core Isolation Cooling System to Open Against Operating Reactor , Pressure i 8/23/84 Design Deficiency in Standby Gas T421 M. thiramal ! Treatment System j l j 8/29/84 Inoperability of Safety Injection Pump T422 0. Zukor at Salem Unit 1 on October 17, 1983 10/25/84 Inoperability of Helium Circulator Over- T423 E. Imbro ! speed Trip Channels Due to Impedence l Variations in Speed Sensing Cables l Exposed to Steam Leak l l 11/20/84 Fire Water Main Leakage Into 4 kV T424 T. Cintula l Switchgear Room at San Onofre Unit 1 I l l D-15
l TABLE 0-3 (Continued) Nonreactor Engineerina Evaluations Date Subject No. Author 5/8/84 Report on Medical Misadministrations N2040 S. Pettijohn for January 1983 through June 1983
- 6/11/84 Nonreactor Event Report Database for N401 K. Black l the Period July-December 1983 l
1 1 6/26/04 Events Involving Undetected Unavail- N402 E. Trager ability of the Turbine-Driven Auxiliary Feedwater Train 7/3/84 Report on Mt 1 al Misadministrations N403 S. Pettijohn for July 196,' "scecher 1983 t l l l 0-16
l TABLE D-4 REPORTS ISSUED IN CY 1983 Case and !pecial Studies , Date Subject No. Author 4/83 Failures of Class 1E Safety-Related C301 M. Chiramal ; Switchgear Circuit Breakers to Close j on Demand 9/83 Potentially Damaging Failure Modes of NUREG/ M. Chiramal L j High and Medium voltage Electrical CR-3122 , Equipment ! 7/83 Report on the Implications of the P301 J. Crooks [ Anticipated Transient Without Scram ! Events at the Salem Nuclear Power : Plant on the NRC Program for Collection ; I and Analysis of Operational Experience , t l Reactor Engineerina Evaluations i Technical Reviews ! l Subject Author
)
Date No. f f I 1/19/83 Fuel Degradation at Westinghouse Plants E301 D. Zukor ; I
; 1/31/83 Potential Loss of Service Water Flow E302 E. Imbro l J Resulting from a Loss of Instrument Air '
2/16/83 Valve Flooding Event at Surry E303 D. Zukor f I 3/11/83 Investigation of Backflow Protection in E304 T. Cintula Common Equipment and Floor Orain Systems 1 1 to Prevent Flooding of Vital Equipment in f
! Safety-Related Compartments [
l i 4/13/83 Inoperable Motor Operated Valve E305 E. Brown Assemblies Due to Premature Degradation F. Ashe ;
- of Motors and/or Improper Limit Switch / i TorqueSwitchAdjustment 4/14/83 Cooldown During Loss of Control Room E306 D. Zukor t Test at McGuire Unit 1 l I
l 4/14/83 Update to AEOD/E301 (Fuel Degradation E301 D. Zukor l at Westinghouse Plants)(Revision 1) l I r 0-17 i - (
- - ~ - . . - - - - - . - - _ - _ - _ . - ._-_.-.. - . - - . - ____ - - _. ._- - - - . ..
4 t TABLE 0-4 (Continued) Reactor Engineerina Evaluations and Technical Reviews (Continued) Date Subject No. Author 4/18/83 Degradation of Safety-Related Batteriet E307 F. Ashe Due to Cracking of Battery Cell Cases and/or Other Possible Aging-Relating Hechanisms 4/19/83 Cracks and Leaks in Small Diameter E308 E. Brown i Piping 4/21/83 The Potential for Water Hammer During E309 S. Rubin the Restart of Residual Heat Removal Pumps at BWR Nuclear Power Plants 4/25/83 Loss of Shutdown Cooling and Subsequent E310 T. Cintula Boron Dilution at San Onofre Unit 2 4/25/83 Loss of Salt Water Flow to the Service E311 T. Cintula Water Heat Exchangers for 23 Minutes at Calvert Cliffs Unit 2 5/10/83 Operability of Target Rock Safety Relief E312 J. Pellet Valves in the Safety Mode with Pilot Valve Leakage 6/15/83 Potential Contamination of the Spent E313 E. Brown Fuel Pool and Primary Reactor System 6/28/83 Loss of All Three Charging Pumps Due E314 T. Cintula to Empty Common Reference Leg in the Liquid Level Transducers for the Volume Control Tank at St. Lucie 1 7/5/83 Hisuse of Valve Resulting in Vibration E315 E. Brown ar.d Damage to the Valve Assembly and Pipe Supports 7/11/83 Frozen Ice Condenser Intermediate Deck E316 0. Zukor Doors 8/1/83 LossofHighPressureInjectionSystem E317 N. Trehan 8/15/83 Biofouling at Salem Units 1 and 2 E318 E. Imbro 9/8/83 Loss of Drywell-Torus Pressure E319 S. Rubin Differential During Residual Heat Removal Pump Flow Testing at Cooper 0-18
TABLE 0-4 (Continued) Reactor Engineering Evaluations and Technical Reviews (Continued) Date Sub.icct No. Author 9/8/83 Power Operated Reitef Valve (PORV) E320 E. Imbro ActuationResultinginSafetyInjection Actuation at Calvert Cliffs 9/12/83 Three Similar Events of a Loss of Shut- E321 T. Cintula down Cooling Flow at Combustion
- Engineering Plants 9/1C/88 Damage to Vacuum Breaker Valves as a E322 C. Hsu Result of Relief Valve Lifting at Peach Botton Unit 2 l 9/19/83 Load Reduction Transient at Salem Unit 2 E323 N. Trehan on January 14, 1982 l
9/21/83 Review nf Events Involving Failures of E324 M. Chiramal Power Supply in Instrumentation and Contro' Systems 11/21/83 Vapor Binding of Aaxiliary Feedwater E325 W. Lanning Pumps at Robinson Unit 2 11/28/83 Steam Voiding in Oconee Unit 3 on E326 H. Ornstein June 13, 1975: A Precursor Event to the TMI-2 Accident 11/28/83 Gaseous Releases From Waste Gas Disposal E327 N. Trehan System 1/19/83 Diesel Generator Load Sequencer Design T301 M. Chiramal Deficiency - LER 82-025/0IT 2/9/83 Postulated Loss of Auxiliary Feedwater T302 E. Imbro System Resulting from a Turbine Driven , Auxiliary Feedwater Pump Steam Supply Line Rupture 3/2/83 Seat Degradat. ion in Henry Pratt T303 E. Brown Butterfly Valves 3/23/83 Cause of Containment Isolation T304 S. Salah Valve F042A to Close at Brunswick Unit 1 3/28/83 Flow Blockage in Essential Raw Cooling T305 E. Imbro Water System Due to Asiatic Clam Intrusion at Sequoyah Unit 1 D-19
! h l TA8LE D-4 (Continued) Reactor Engineerina Evaluations and Tevnical Reviews (Continued) t Date Sub.iect No. Author 4/23/83 Scram Discharge Volume Level Switch T306 J. Pellet Failure at Hatch Unit 2 4/19/83 Condensate Domineralizer Resin Migration T:(7 J. Pel'et ( through the Plant Vent and the Standby Gas Treatment System at Pilgrim Unit 1 4/20/83 Unietectable Failure in Westins5ouse T308 M. Chiramal Solid State Protection System 4/25/83 Air in Reactor Water Cleanup System T309 S. Salah Instrument Sensing Lines at Brunswick Unit ? 4/25/83 Blocking of Automatic Safety Injection T310 M. Chiramal Signals 5/5/63 Rod Control Urgent Failure on June 25, T311 N. Trthan 1982 at Surry Unit 2 5/9/83 Failure of 5 kV Cable Terminations Ta12 M. Chiramal 5/11/83 Capped Containment Pressure Sensing T313 S. Rubin Lines 5/24/83 Improper Size of Inlet Piping to T314 E. Imbro Primary Safety Valves 5/24/83 Events Involving Losses of or Perturba- T315 F. Ashe tions in a Single 120 Volt AC Vital Power Supply Inverter and Attendant Distribution Bus which Resulted in Inadvertent Actuations of Safety Systems 5/31/83 Thermal Nonrepeatability Problem with T316 M. Chiramal Barton Model 763 and 764 Electronic Transmitters 6/13/83 Probler.s with Dietel-Driven Containment T317 D. Zukor Spray Pimp at Zion Unit 2 on December 16, 1982 0-20
l i f TASLE D-4 (Continuid) f
,hactorEncin,eerinoEvaluationsandTechnicalReviews(Continued)
] Date Suht No. Author I 6/13/88 failure of Recirculation Spray Service T318 0, Zukor , { Water Motor Operated Valves [ 6/13/83 Design Deficiency in Control Circuits of T319 M Chirama) i j Feedwater Isolation Valves and Boron l Injection Tank Recirculation Valves 1
- 6/14/83 Inadvertent Safety Injections Attributed T320 F. Ashe j to Personnel Error at Summer !
6/15/83 Check Valve Installed Backwards in T321 0, Zukor j ] Instrument Air Line to the Pcwer ! 1 Operated Relief Valve at Surry Unit 1 ! t 6/15/83 Gow ts in Main Coolant System Piping T322 0. Zukor ! 1 at diablo Canyon on April 19, 1983 I i i I 6/17/83 Turbine Trip Bypass Delay at Grand Gulf T323 S. Salah [ I Unit 1 l 7/27/83 Events Involving Two or More Simultan- T324 F. Ashe ! l eously Dropped Rod Control Cluster ( Assemblies : f
- 8/1/83 Leakage in Static-0-Ring Pressure T325 M. Chiramal ;
l Switches j 8/2/83 Safety / Relief Valve Corrosion at a T326 E. Brown r Foreign Reactor I 8/11/83 Auxiliary Feedwater Header Problems at T327 H. Ornstein , Babcock & Wilcox Plants ! 8/12/83 Two of Three Emergency Core Cooling T328 0. Zukor ! System Accumulators Ir. operable at } Surry Unit 1 l P 8/24/83 Leak in Reactor Water Cleanup System "B" T329 C. Hsu ! Regenerative Heat Exchanger Relief Line ! l 8/29/83 Steam Generator Tube Rupture at Oconee T330 H. El-Zeftawy j Unit 2 i l 8/29/83 Review of Events at Operating Nuclear T331 M. Chiramal Plants Involving Plant Computers 0-21 !
1ABLE 0-4 (Continued) Reactor Enaineerina Evaluations and Technical Reviews (Continued) t Date Subject M. Author 10/7/83 Reactor Vessel Orainage at Grand Gulf T332 S. Salah Unit 1 10/31/83 Degradation of Saltwater Cooling System T333 H. Ornstein
.t San Onofre Unit 1 Caused by a Loss of Instrument Air 11/15/83 Reactor Vessel Orainage at Grand Gulf T334 S. Salah Unit 1 11/15/83 Simultaneous Oafety Injection Actuation T335 E. Imbro Signal and Recirculation Actuation Signal at San Onofre Unit 3 11/17/83 Design Deficiency Resulting in Isolation T336 M. Chiramal of Both loops of the Emergency Condenser System Ot Nine Mile Point Unit 1 11/21/83 Water Hammee in the Main F cdwater T337 E. Imbro System Resulting in a Feedeater Line Crack at Maine Yankee 11/28/S3 Water Leak Through Containment Spray T338 0. Zukor Block Valves at San Onofre 1 11/29/83 Redundant Emergency Core Cooling System T339 T. Cintula Pump Room Air Coolers Out of Service far 22 Hours at Calvert Cliffs Unit 1 12/2/83 Evaluation of a Control Rod Hismanipul- T340 T. Wolf ation Event at Hatch Uni t 2 12/19/83 Corrosion of Carbon Steel Fipe in T341 E. Brown Service Water Headers Nonreactor Engineerina Evaluations and Technical Reviews Date Subject @. Author 1/11/83 Nonreactor Event Report Database for N109A E. Trager the Period January-June 1982 0-22
\ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _
I TABLE 0-4 (Continued) f I Nonreactor Engineerina Evaluations and Technical Reviews (Continued) Date Sub.iect g. Author 3/18/83 1125/1131 Effluent Release by Material N301 S. Pettijohn Licensees ' 6/10/83 Mound Laboratory Fabricated PuSe Sources N302 K. Black 6/10/83 Americium Contamination Resulting from N303 K. Black i Rupture of Well-Logging Sources 6/14/83 Nonreactor Event Report Database for the N2098 K. Black Period July-December 1982 < 7/14/03 Americium-241 Sources N304 7/14/83 Report on Mec"cal Misadministrations N204C S. 'ettijohn : { for January sal December 1982 l 1 I , 8/4/83 H n Factors Contributions to Accident N305 E. Trager [ l 5 ace Precursor Events l i 12/1/83 Potentially Leaking Americium-241 N306 5. Pettijohn Sources Manufactured by Amersham ) Corporation i ! j 12/28/83 honreactor Event Report Database for N307 K. Black ' ) the Period January-June 1983 l 3/10/83 Internal Exposure to Am-241 NT301 K. Black [ 4/5/83 Kay-Ray, Inc. Reports of Suspected NT302 S. Pettijohn i Leaking Sealed Sources Manufactured ) by General Radioisotope Products l 8/24/83 Possession of Unauthorized Sealed NT303 5. Pettijohn 7 l Source / Exposure Device Combinations ! j by Mid-Con Inspection Services, Inc. ;
- i i 11/4/83 Human Factors Involvement in Events at NT304 K. Black [
1 Oconee Units 1, 2 and 3 l 1 I < l l l l .i
- D-23 Y
_ t
TABLE D-5 REPORTED ISSUED IN CY 1982 Case & Special Studies Date Subject & Author 1/82 Safety Concern Associated with Reactor C201 M. Chiramal Vessel Level Instrumentation in Boiling Water Reactors 2/82 Report on Service Water System Flow C202 E. Imbro Blockages by Bivalve Mollusks at Arkansas Nuclear One and Brunswick 5/82 Survey of Valve Operator-Related Events C203 E. Brown Occurring During 1978, 1979, and 1980 7/82 San Onofre Unit 1 Loss of Salt Water C204 H. Ornstein Cooling Event on March 10, 1980 8/82 Abnormal Transient Operating Guidelines C205 J. Pellet as Applied to the April 1031 Overfill Event at Arkansas Nuclear One, Unit 1 10/82 Inadvertent Loss of Reactor Coolant C206 W. Lanning Ev2M s at the Sequoyah Nuclear Plant, Units 1 and 2 Reactor Engineering Evaluations Date Subject No. Author 1/12/82 Methodology for Vital Area Determination E201 W. Lanning 1/13/82 Loss of High Pressure Injection Lube Oil E202 J. Pellet Cooling at Rancho Seco 1/21/82 Inadvertent Isolation of Containment Fan E203 W. Lanning Units at Salen Unit 1 1/28/82 Effects of Fire Protection System Actua- E204 M. Chirasal tion on Safety-Related Equipment 2/16/82 Potential Consequences of Heavy Load Drop E205 M. El-Zeftawy Accidents in LWRs 2/22/82 Load Reduction Transient on January 14 E206 N. Trehan 1982 at Sales Unit ? D-24 t _ --- -
TABLE 0-5 (Continued) Reactor Engineering Evaluations (Continued) Date Sub. lect No. Author 2/22/82 LER 50-336/81-26: Investigation of the E207 E. Imbro Relative Frequency of Valve Overtravel Anomalies that Could Result in a Potential Centrifugd Pump Runout Exceeding Net Positiva Suctw.' Head 2/22/82 An Observet! Difference i.' Lif t Setpoint E208 E. Imbro for Steam Ggnerator and Pttssurizer Safety Valves 2/23/82 Generator Rotar Retaining Rint, as a E209 M. Chiramal Potential Miss'le (Incident at carseback Unit 1 on April 13,1979) 2/23/82 Inadequate Switchjear Cooling at Betver E210 W. Lanning Valley Unit 1 2/7.4/82 Repetitive Failures cf Emergency Feed- E211 E. Imbro water Flow valves at Arkansas Unit 2 Because of Valve Operator Hydraulic Problems 2/24/82 Spurious Trip of the Generator Lockout E212 F. Ashe Relay Associated with a Diesel Generator Unit 2/24/82 Trip of Two Inservice Auxiliary Feedwater E213 D. Zukor Pumps from Low Suction at Zion Unit 2 on December 11, 1981 3/1/82 Duane Arnold Loss of River Water System E214 T. Wolf Lcop 3/18/82 Engine cing Evaluation of the Salt E215 E. Imbro Service Water System Flow Blockage at the Pilgrim Nuclear Power Station by Blue Mussels 3/28/82 4 Recently Evaluated Preoperational Test E216 H. Ornstein Precursor of the THI-2 Accident 3/31/82 Scram Pilot Solenoid Valve Failures Due E217 M. Chiramal to Low Voltage - Grand Gulf Unit 1 0-25
TABLE D-5 (Continued) keactor Engineerina Evaluations (Continued) i Date Subject No. Author T1/82 Potential for Air Binding or Degraded E218 5. Rubin F.rformance of BWR Residual Heat Removal i I System Pumps During the Recirculation Phase of a Loss of-Coolant Accident 4/1/82 Proposed Circular: Contamination of Air E219 H. Ornstein ; l Serving Safety-Related Equipment I 4/6/82 Water in the Fuel Oil Tank at Surry Power E220 N. Trehan Station Unit 2
! 4/22/82 Indian Point Unit 2 Flooding Event Ekil W. Lanning l r
i 5/10/82 Loss of Reserve Station Service E222 N. Trehan ; Transformer "8" on January 18, 1982 at l l surry Unit 2 l 5/11/82 Inadvertent loss of Coolant Events at E223 W. Lanning I i Sequoyah Units 1 and 2 (
) [
5/21/82 Generic Concerns Associated with the E224 W. Lanning Ginna Steam Generator Tube Rupture Event i 4 i 6/1/82 Degradation of BWR Scram Pilot Solenoid E225 M. Chiramal Valves Due to Abnormal Power Supply Voltage . 6/18/82 Inoperability of Instrumentation Due to E226 M. Chirama) i Extreme Cold Weather 6/24/82 Failure of Engineered Safety Features E227 F. Ashe Manual Inititation Pushbutton Switches 6/25/82 Repetitive Overspeed Trips of the Steam E228 E. Imbro l Oriven Emergency Feedwater Pump on Initial l Start at Arkansas Nuclear One, Unit 2 l Potential for Flooding in Control Room at E229 E. 1mbro ) 6/29/82
- San Onofre Units 2 and 3 7/7/82 Water in the Fuel Oil Tank at Surry Power E230 N. Trehan l
Station, Unit 2 - Additional Information 7/19/82 Millstone Unit 2 Loss of Shutdown Cooling E231 M. Chiramal Due to Trip of Low Pressure Safety InjectionPump i ! D-26 i
i TABLE D-5 (Continued) Reactor Engineerina Evaluations (Continued) Date Sub.iect M. Author ! 7/19/82 Potential Deficiency in the Sigma E232 F. Ashe { Lusigraph Indicators Model Number 9270 I 7/28/82 Carbon Dioxide Systems Used for Fire E233 M. Chiramal f Protection in or Adjacent to Critical i Areas [ 8/11/82 Failure in a Section of 4 kV Bus Cable E234 F. Ashe l Henufactured by Okonite j ! 8/11/82 Wiring Error in Handswi'.ch for Solenoid E235 S. Rubin ! Control Valves Associated with High Pressure ! CoolantInjectionsystemSteamCandensing l Mode Pressure Contial Vahe at Duane Arnold t t 8/25/82 Brunswick Steam Electric Plant Unit 2 E236 T. Wolf f Loss of Residut Heat Remova! Service Water on Januar,< 16, 1982 8/25/82 Power Operated Relief Valve Failure at E237 E. Brown Robinson 8/25/82 Water in the Lube Oil in Safety Injection E238 N. Treha' { Pump IA-A at Sequoyah - LER 81-076 6 9/24/82 Main Steam Isolation Valve Closures and E239 T. Cintula Pressurizer Safety Valve Actuations at ( St. Lucie Unit 1 on December 19, 1981 i 9/29/82 Preliminary Account of Events Associated E240 S. Rubin i with a Reactor Trip at Hatch Unit 2 on j August 25, 1982 ; 10/1/82 Emergency Diesel Generator System E241 M. Chiramal ! Problems at Fitzpatrick 10/21/82 Fuel Assembly Degradation While in the E242 E. Brown Spent Fuel Storage Pool 10/21/82 Plant Trip Followed b T. Cintula Caused by loss of "A"y a Safety Cooling Injection E243 Tower Pump at Palisades on February 4, 1982 0-27
)
4 1 1 ! TA8Li 0-5 (Continued) Reactor Engineerina Evaluations (Continued) Date Subject No. Author l 10/21/82 Loss of Residual Heat Removal System E244 T. Wolf a Event at Pilgrim Nuclear Power Station , on December 21, 1981 l 10/21/82 Failure of Westinghouse Type SC 1 No. E245 f. Ashe i 1876-072 Relays 10/21/82 Events Involving Loss of Electrical E246 f. Ashe In't.ters Including Attendant Inverters 1 to Vital Instrument Buses ! 10/26/82 Engineering Evaluation of Turbine / Reactor E247 J. Pellet
. Trip at Rancho Seco on August 7, 1981 j i 11/2/82 Engineering Evaluation Report on McGuire E248 D. Zukor l
Overpressurization Event of August 27, 1981 I 11/4/82 Engineering Evaluation Memorandum. E249 H. Ornstein 1 i Licensee Reporting of the Turbine / Reactor j Trip at Rancho Seco on August 7, 1981 , I i 11/8/82 Quad Cities Unit 2 Loss of Auxiliary E250 M. Chiramal l Electrical Power Event on June 22, 1982 l
- I 11/9/82 Salem Unit 2 Loss of Vital Bus No. 2A E251 M. Chiramal
{ { { 11/17/82 Potential Control Logic Problem Resulting E253 F. Ashe ! ' in Inoperable Auto-Start of Dieset j I Generator Units Under the Conditions of loss-of Coolant Accident and toss of ! j Station l i 11/17/82 Review of Prairie Island Unit 1 E254 M. Chiramal ! ! LER 82 015-0!T on Diesel Generator
! Operability l 11/17/82 Failure of the Vent Line on the Common E255 T. Cintula ! ; Discharge of the Two Motor-Driven Auxiliary }
I Feedwater Pump; at San Onofre Unit 2 from t ] an Improper Valve Lineup ( l 11/24/82 Loss of Shutdown Cooling and Subsequent E256 T. Cintula Boron Dilution at San Onofre Unit 2 l kt 4 D-28
- r
l TABLE D-5 (Continueo) , l Reactor Engineerina Evaluations (Continued) Date Subject No. Author j 12/2/82 Insufficient Net Positive Suction Head E257 D. Zukor [ for Charging Pump Service Water Pun.ps at . Surry Nuclear Power Station i Nonreactor Engineerina Evaluations , Ste Subject @. Author . l j 2/1/82 Report on Medical Misadministrations for N201 S. Pettijohn l I the Period November 10, 1980 - September 30, I
- 1981 [
1/21/82 Buildu of Uranium-Bearing Sludge in N202 K. Black I WastegetentionTanks
- 2/18/82 Lost Plutonium - 238 Sourse N203 K. Black 3/5/82 Report on Medical Misadministrations for N204 5. Pettijot j CY 1981 f
l 4/27/82 Preliminary AE00 Review of lodine-125 N205 E. Trager { q Sealed Source Leakage Incidents ! i 5/6/82 Eberline Instrument Corporation - Part 21 N206 K. Black i 1 Report i 5/25/82 AE00 Review of Iodine-125 Sealed Source N207 E Trager [ j Leakage Incidents l
; r 1 8/2/82 Potentially Leaking Plutonium-Beryllium N208 S. Pettijohn i j Neutron Sources i f ! 8/30/82 A Summary of the Nonreactor Ever,t Report N209 K. Black l Database for 1981 j l
11/15/82 Leaking Hoses on Self Contained Breathing N210 K. Black Apparatus (SCBA) Manufactured by MSA ] J i 1 l ) 0-29 i _. _ - _ _ _
____7__-___-________-___________________________________. TABLE 0-6 REPORTS ISSUED IN CY 1981 Case Studies Date Subject No. Author
- 3/81 Report on the St. Lea.ie Unit 1 Natural C101 E. Intro Circulation Cooldown on June 11, 1980 l
3/81 Robinson Reactor :loolant System Leak C102 W. Lanning on January 29, 1981 3/81 AE00 Safety Concerns Associated with C103 S. Rubin l Pipe Breaks in the BWR Scram System l l 4/81 Millstone Unit 2 Loss of 125 V DC Bus C104 M. Chiramal l Event on January 2, 1981 ; I 12/81 Report on the Calvert Cliffs Unit 1 C105 E. Imbro ! Loss of Service Water on May 20, 1980 I i Reactor Engineerina Evaluatio 3 l Date Subject No. Author ! 1/19/81 Cegradation of Internal Appurtenances E101 E. Brown f' in LWR Piping 1/30/81 Sequoyah Unit 1 Loss of Annunciation E102 M. Chiramal f 3/2/81 Engineering Evaluation of Feedwater E104 5. Sands [ Transient and System Pipe Break at - Turkey Point 3 i 3/31/81 Water Hammer During Restart of Residual E105 J. Huang r Heat Remeval Pueps [ 3/31/81 Water Hammer in the Steam condensing E106 J. Huang Mode of the Residual Heat Removal I Systea Operation } 1 I r 1 [ 0-30 l
I TABLE 0-6 (Continued) l l Reactor Encineering Evaluations (Continued) Date Sub.iect No. Author, 4/17/81 Peach Bottna Unit 3 Occurrence on E107 F. Ashe f February 25, IC81 j 4/21/81 Hatch Units 1 and 2 - Alternate Offsite E108 M. Chiramal ! Source Interlock with Emergency Diesel Generators i 4/24/81 Potential Creman Mode Failure of Diesel E109 M. Chiramal i
- Generators !
l ! 4/29/81 Requirements of the Preferred or E110 F. Ashe I ! Offsite Power System i l 5/22/81 Evaluation of High Pressure Safety Elli E. Imbro i InjectionPumpOperabilityWithout i Service Water i { 6/15/81 Inoperability of Instrumentation Due E112 M. Chirama) i
- to Extreme Cold Weather l l 6/24/81 Deliberate Pump Trip at Browns Ferry E113 W. Lanning !
; Unit 2 on April 6, 1981 ! ) i l 6/24/81 Control System Failures that could E114 F. Ashe [
Cause or Exacerbate Nuclear Power i
,I Plant Accidents '
i I 7/8/81 Additional Information on Events at E115 H. Ornstein TMI 2 During Prooperational Testing
] (9/5 12, 1977) i i
7/14/81 Failure of B Phase Main Transformer and E116 M. Chiramal I Subsequent Fire in the Transformer Area - North Anna Unit 2 7/16/81 Events at TMI-2 During Pr" .. m tional E117 H. Ornstein i j Testing ) 7/20/81 Setpoint Drift Occurrences r .he E118 F. Ashe q Barton Model 288 Instrument l 7/22/81 Loss of Residual Heat Removal Capability E119 E. Imbro l at Brunswick Units 1 and 2 l' i 8/6/81 Ignition of Gaseous Waste Decay Tank at E120 H. Ornstein I San Onofre Unit 1 - July 17,1981 j i l ! l 0 31
1 l TABLE 0-6 (Continued) Reactor Engineering Evaluations (Continued) Date Subject No. Author 8/28/81 Crystal River 3 Engineered Safeguards E121 M. Chiramal Relay Failures 9/4/81 AE00 Concern Regarding Inadvertent E122 H. Ornstein Opening of Atomospheric Dump Valves on B&W Plants During Loss of Integrated Control System Nonnuclear/ Instrumentation 9/15/81 Immediate Action Memo: Common Cause E'.23 H. Ornstein Failure Potential at Rancho Seco - Desiccant Contaminatt of Air Lines 9/24/81 Review of Information on Purge Valves E124 E. Brown 10/15/81 Engineering Evaluation Report on E125 G. Lanik Shutdown Cooling System Hoat Exchanger Failures at Oyster Creek, August 1081 i 10/16/81 Event Sequences Not Considered in the E126 F. Ashe Design of Emergency Eus Control Logic 10/28/81 Pressure Boundary Degradation Due to E127 W. Lanning Pump Seal Failure at Arkansas Nuclear One 11/10/81 Inoperable Teledyne Solenoid Valves E128 F. Ashe 12/7/81 Brunswick Unit 2 Diesel Generator E129 M. Chiramal Jacket Water Temperature Control Valve and Manual Bypass Valve 12/7/81 Davis Besse LER 79-062 on Auxiliary E130 M. Chiramal Feedwater System Pressure Switches 12/10/81 High Circulating Current Associated E131 F. Ashe with Inverter Output Due to Lack of Circuit Tuning 12/23/81 Abnormal Wear Encountered on Aloyco E132 T. Cintula Swing Check Valves Installed in the Low Pressure Safety Inje: tion System at Calisades 4/15/81 Inadequacies ir. Permdic Testing of E133 H. Chiramal Combustion Engineering PWR Reactor Protection System 0-32
l' l l TABLE 0-6 (Continued) NonreactorEngineerinaEvaluations Date Subject No. Author 3/16/81 Interim Report on Brown Bovert Betatron N101 E. Trager Calibration Check Source 3/26/81 Irradiator Incident at an Agreement N102 K. Black State Licensee's Facility (Becton-Dickinson, Broken Bow, Nebraska) 4/13/81 Interim Report on the October 1980 Fire N103 E. Trager at the Sweetwater Uranium Hill l l 4/30/81 Interim Report on the January 2, 1981 N104 E. Trager l Fire at the Atlas Uranium Mill 5/18/81 Interim Report on Tailings Impoundment N105 E. Trager Liner Failure at the Sweetwater Uranium ! Hill ; 8/12/81 Review of Reports of Leaking Radioactive N106 E. Trager Sources 12/10/81 Engineering Evaluation of Fire N107 E. Trager Protection at Nonreactor Facilities 'f 12/16/81 Notes on AE00 Review of Emissions from N108 E. Trager Tritium Manufacturing and Distribution Licensees , I l 1 l l I D-33 i
I TABLE 0-7 REPORTS ISSUED IN CY 1980 Case Studies Date subject No. Author 7/80 Report on the Browns Ferry Unit 3 ' ^-01 S. Rubin Partial Failure to Scram Event on June 28, 1980 9/80 Report on the Interim Equipment and C002 G. Lanik Procedures at Browns Ferry Unit 3 to Detect Water in the Scram Discharge Volume 10/80 Report on Loss of Offsite Power Event C003 W. Lanning at Arkansas Nuclear One, Units 1 and 2 11/80 AE00 Actions Concerning the Crystal C004 H. Ornstein River Unit 3 Loss of Nonnuclear Instrumentation and Integrated Control System Power on February 26, 1980 12/80 AEOD Observations and Recommendations C005 E. Imbro Concerning the Problem of Steaa Generator Overfill and Combined Primary and Secondary Side Blowdown Reactor Engineerina Evaluations
,Date Subject No. Author 3/20/80 Crystal River Nuclear Power Plant Decay E001 H. Ornstein Heat Closed Cycle Cooling Water Pumps /
OCP-IA and DCP-IB 5/23/80 BWR Jet Pump Integrity E002 S. Rubin 6/19/80 Comparison of Reactor Coolant Pump E003 E. Brown Events Contained in LERs, NPRDS, RECON, and Plant Records 0-34
TABLE D-7 (Continued) Reactor Engineering Evaluations (Continued) Date Subject No. , Author 7/10/80 Data Summaries of Licensee Event Reports E004 H. Ornstein of Pumps at U.S. Comercial Nuclear Power Plants, January 1, 1972 to April 30, 1978 7/15/80 Operational Restrictions for Class 1E E005 M. Chiramal 210 V AC Vital Instrument Buses 8/4/80 Loss of Residual Heat Removal at Beaver E006 W. Lanning l Valley, LER 80-n31 8/18/80. Potential for Unacceptabla Interaction E007 S. Rubin Between the Control Rod Drive System and Nonessential Control Air System at Browns Ferry 8/20/80 Operational Restrictions During Surveil- E008 H. Chiramal lance Testing of Emergency Diesel Generators 8/22/80 Failures of Containment Isolation Valves E009 W. Lanning at Zion
- 8/27/88 Tie Breaker Between Redundant Class 1E E010 M. Chiramal Buses-Point Beach Units 1 and 2 I
8/29/80 Concerns Relating to the Integrity of a E011 E. Imbro Polymer Coating for Surfaces Inside Containment
- 9/12/80 Salem Unit 1-Solenoid Valve of E012 M. Chiramal Containment Fan Coil Unit Service Water Flow (,ontrol Valve 9/12/80 Excessive Main Feedwater Transient E013 J. Creswell 10/8/80 Transient at Crystal River Unit 3- E014 H. Ornstein September 30, 1980 J
^ 0-35 I
TABl.E D-7 (Continued) Reactor Engineerina Evaluations (Continued) Date Subject No. Author 10/20/80 January 3,1977 Quad Cities Unit 1 E015 G. Lanik Loss of Air Event and Its Effects on Scram Capability 10/21/80 Flow Blockage in Essential Equipment at E016 E. Imbro ANO Caused by Corbicula sp. (Asiatic Class) 10/29/80 Engineering Evaluation of Steam E017 W. Lanning Generator Overfill 12/18/80 Potential Failure of BWR Backup Scram E018 M. Chiramal (Mode Switch in Shutdown) Capability 12/19/80 Davis Besse Unit 1-Emergency Core E019 M. Chiramal Cooling System Actuation During Hot Shutdown on December 5, 1980 12/24/80 Internal Appurtenances in LWRs E020 E. Brown D-36 i
i APPENDIX E STATUS OF AEOD RECOMMENDATIONS
STATUS OF RECOMMENDATIONS At the beginning of 1987, 80 AE00 recommendations were outstanding as listed in the "AE00 Recommendation Tracking System." As of December 31, 1987, the status of AE00 recommendations is as follows: Status of AEOD Recommendation Added since isst report 7
- Resolved, deleted or combined 24 Currently outstanding 63 Included in USI, GI, & TMI 37 Addressed to staff, INPO, etc. 26 RES prioritization of current issues High 37 Medium 8 Low 2 Not prioritized 16
' Functioned areas addressed by current issues Procedures, training, etc. 34 Hardware modification 16 Equipment testing 6 Other 7 AE00's tracking system ensures that all formal AE00 recommendations are tracked until resolution is achieved. At this time, there are no issues involving AE00 recommendations which would warrant EDO attention. The majority of current issues have been prioritized as high and many are include <l as NRC USIs and GIs as shown above. Important recommendations related to decuy heat removal, MOV problems, and air system problems have been acted upon by the responsible program offices.
In addition to the formal recommendations which are tracked and included in this section, additional actions are routinely implemented by NRC program offices on AEOD suggestions contained in engineering evaluations and special reports. These AE00 suggestions are not formally tracked or closed out by AE00. E-3
AEOD RECOMENDATION TRACKING SYSTEM, RECOMMEN0ATION SOURCE: Case Study AEOD/C101 Responsible AEOD Engineer: T. Cintula TITLE OR
SUBJECT:
"St. Lucie Natural Circulation Cooldown" RECOMMENDATION 1 Provide a supply of cooling water to reactor coolant pum not be disabled by a single failure. (Recommendation 4e)p seals that will RESPONSIBLE OFFICE /0lV/BR CONTACT PRIORITY RES/DE/EIB J. Jackson High STATUS This recommendation is included in Generic Issue 65, "Probability of Core Melt Due to Component Cooling Water System Failures." Generic Issue 65 Seal Fcilures.p' Generic Issue 23 will address the reliability of RCP sealhas cooling systems. No milestones were scheduled or completed in 1987 for Generic Issue 23. The next scheduled step is preparation of a proposed :
resolution package for CRGR review. i T d l i i C-101-1 i E-4
l l AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C105 Responsible AE00 Engineer: T. Cintula TITLE OR
SUBJECT:
"Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980" RECOMMENDATION 1 Installation of dual atmospheric dump valve capability for each steam generator on two-loop PWRs. (Study recommendation 8(b)3)
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/0RPS/RPSIB H. Woods High STATUS This recommendation originally was to have been addressed by a revision to SRP 15.6.3 as part of Generic Issue 67.5.1, "Reassessment of Radio-logical Consequences Following a Postulated Steam Generator Tube Rupture." This recommendation is currently addressed in USI A-45, "Shutdown Decay Heat Removal Requirements." During 1987, comments were received on the regulatory draft "Shutdown DHR Analysis" and NPR was briefed on the status of A-45. As a result of this meeting, A-45 may be rescoped to include plant specific decay heat removal probabilistic risk assessments I and A-45 may be included in the individual plant examinations being done as a part of the Severe Accident Program. REC 0f01EhDATION 2 Review of stear ger.erator tube rupture (SGTR) analyses for plants licensed prior to the SRP. (Study recommendation 8(b)2) RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE K. Shaukat Medium STATUS This issue was inadvertently unassigned during 1986. The recommendation nowmaybeaddressedaspartofGI-67,"ReevaluationofSGTRDesignBasis Event. However, ongoing SGTR programs including GI-67 are being integrated into a single generic issue (GI-135). This recommendation may be included in the integrated program. GI-135 recommends to drop the "Reevaluation of SGTR Design Basis Event" because LOFT and steam generator vendor test facilities have demonstrated ability to control SGTR events. C-105-1 E-5
AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AE00/C105 (continued) Responsible AE00 Engineer: T. Cintula TITLE OR
SUBJECT:
"Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980" RECOM ENDATION 5 IST of check valves in the instrument air system used to isolate safety-related portions of the system. (Study recommendation 8(e)2)
RESPONSIBLE OFFICE /DIV/8R CONTACT PRIORITY RES/ ORA /ARGIS W. Milstead Low / drop STATUS This issue is to be addressed as part of Generic Issue 43, "Contamination of Instrument Air Lines" which was prioritized by NRR as low / drop. AE00 did not agree with the prioritization. NRR agreed to reevaluate the prioriti-zation when a comprehensive AE00 case study on air system problems was issued. The AE00 case study was issued in April 1987. Since then, GI-43 has been rescoped to include all causes of inttrument air system unavail-ability. Reptioritization of new GI-43 is currently underway and a priority of high is tentatively contemplated. RECOMMENDATION 6 Accessibility of ADVs for local manual operations for RCS cooldown following a steam generator tube rupture. RESPONSIBLE OFFICE /DIV/8R CONTAE PRIORITY NRR/OSR0/RSIB A. Marchese High STATUS This recommendation is currently included in USI A-45, "Shutdown Decay Heat Removal Requirements." A site walkdown survey of nine plants substantiated the AE00 concern of ADV accessibility. At some plants, the ADVs were readily accessible, while ADVs at other plants were difficult to open manually and may cause personnel radiation exposure. During 1987, comments were received on the regulatory draft "Shutdown OHR Analysis" and NRR was briefed en the status of A-45. As a result of this meeting, A-45 may be rescoped to include plant specific decay heat removal probabilistic risk assessments and A-45 may be included in the individual plant examinations being done as part of the Severe Accident Program. C-105-2 E-6
AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Memorandum dated January 20, 1982 from C. Michelson to Harold R. Denton Responsible AE00 Engineer: M. Chiramal TITLE OR
SUBJECT:
"Safety Concerns Associated with Reactor Vesse:1 Level Instrumentation in BWRs" R_ECOMENDATION 1 Safety-related low-low reactor vessel level start of HPCI and RCIC systems should not be prevented or delayed by nonsafety-related high level trip.
! RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/EIB 0. Thatcher High ) STATUS Assigned as Generic Issue 101. Draft report issued. RECOMENDATION 2 Protective functions of ilarrow range level instreuentation must be assured 1 in spite of adverse control system protection syszem interaction. ) j RESPONSIBLE OFFICE /0IV/BR CONTACT PRIORITY t j RES/EIB A. Szukiewicz High i STATUS Ongoing USI-A47 and Generic Issue 101. GI-101 draft report is issued. USI A-47 CRGR review completed. Report being issued for public comment.
REFERENCE:
Memo dated March 19, 1982 from H. R. Denton to C. Michelson l J i i C-201-1 E-7
AE00 REcomENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AEOD/C202 Responsible AE00 Engineer: M intula TITLE OR
SUBJECT:
"Flow Blockage by Bivalve Mollusks at Arkansas Nuclear One and E'unswick" REC 0mENDATION 1 c n=hility to measure cooling water flow should be provided for all saf;'.y-related equipment.
RE'J ONSIBLE Okic2/DIV/BR CONTACT PRIORITY RES/DE/EIB J. Jackson Medium l STATf/S This recommendation has been included in Generic Issue 51, "Reliability of Open Cycle Service Water Systems." During 1987, a contractor study (FIN-B-2977), which includes this recommendation, has been completed. RECOMENDATION 2 Develop and implement biofouling control strategies. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE/EIB J. Jackson Medium STATUS (See status of Recommendation 1 above.) REC 0mENDATION_3 Periodic inspection of service water system piping. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE/EIB J. Jackson Medium STATUS (See status of Recommendation 1 above.) C-202-1 E-8
AE00 RECOMENDATION TRACKING SYSTEM REC 0mENDATION SOURCE: Case Study AEOD/C202 (continued) Responsible AE00 Engineer: T. Cintula TITLE OR
SUBJECT:
"Flow Blockage by Bivalve Mollusks at Arkansas Nuclear Ora and Brunswick" RECOMENDATION 4 Periodic verification of overall heat transfer coefficient on multiple pass heat exchangers.
I RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY ' r RES/DE/EIB J. Jackson Medium STATUS (See status of Recommendation 1 above.) RECOMENDATION 5 Periodic verification of cooling water flow to all safety-related equipment should be specified in technical specifications. P RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE/EIB J. Jackson Medium 1 STATUS { (See status of Recommendation 1 above.) C-202-2 E-9 i
AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C203 and Memorandum dated May 28, 1982 Responsible AE0D Engineer: E. J. Brown TITLE OR
SUBJECT:
"Survey of Valve Operator - Related Events Occurring During 1978, 1979, and 1980" (See also Recommendation 1 on page S-503-1)
RECOMMENDATION 1 Existing guidance to bypass thermal ovr.rload protective devices associated with safety-related valve motor operators should be reassessed. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE/EIB 0. Rothberg Medium STATUS Delete. Incorporate into C603. RECOMMENDATION 2 Improved methods and procedures for the setting of torque switches should be developed and evaluated relative to valve operability and functierial quali-fication. 1 RESPONSIBLE OFFICE /0IV/BR CONTACT PRIORITY RES/DE/EIB 0. Rothberg Medium STATUS Delete. Incorporate into C603. RECOMMENDATION 3 Signature tracing techniques should be developed and tried on selected motor-operated valves as part of the periodic inservice testing program. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE/EIB 0. Rothberg Medium NRR/00EA/0GCB R. Kiessel STATUS Delete. Incorporate into C603. C-203-1 E-10
AE00 RECOMEN0ATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AEOD/C204 "San Onofre Unit 1 Loss of Salt Water Cooling Event on March 10, 1980," dated July 1982 Responsible AE00 Engineer: H. Ornstein TITLE OR
SUBJECT:
"Single Failure Vulnerability of San Onofre l's Salt Water Cooling System" REcomENDATION 1 Ongoing efforts of the SEP focus on single failure vulnerability and con-sequences for the salt water cooling system and other equivalent service and I cooling water systems.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/PRAB E. Chan High STATUS Resolved. NRR has reviewed SEP plants for such vulnerabilities. Modifica-tions have been made at SEP plants as appropriate. San Onofre has made several modifications and has performed a reliability analysis of the modified salt water cooling system to confirm the adequacy of the system modifications. The analysis was submitted to NRR in April 1987. C-204-1 E-11
AE00 RECOMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Memorandum, C. Michelson to Chairman Ahearne, "New Unresolved Srfety Issues" dated August 4, 1980; Memo, C. Michelson to H. Denton, "Resolution of Issue Concerning Steamline Break with Small LOCA," dated June 23, 1982; Case Study AE00/C205, "ATOG as Applied to the April 1981 Overfill Event at ANO-1" Responsible AE0D Engineer: H. Ornstein TITLE OR
SUBJECT:
"SafetyImglicationsofSteamGeneratorTransientsand Accidents RECOMENDATION 1 Combined primary / secondary side blowdown should be a USI for B&W plants.
RESPONSIBLF 0FFICE/0IV/8R CONTACT PRIORITY NRR/SRXB R. Jones High RES/RPSB R. Lee STATUS B&W licensees /EPRI/NRC are jointly funding a test facility to obtain integral systems test data to resolve the uncertainties associated with B&W plant response to SBLOCA and other transients and accidents. Quick look reports have been completed. Final test reports are being prepared. Based upon the preliminary results of USI's A-3, 4, and 5, the priority of this issue is being reconsidered. Expenditure of additional resources to resolve this issue will depend upon the outcomo of the reprioritizing. MCOMMENDATION2 TAP-A47 should focus on equipment niodifications or additions to preclude SG overfill as a credible event. RESPONSIBLE OFFICE /DIVLBR CONTACT PRIORITY RES A. Szukiewicz High STATUS Included in ongoing Generic Issue A-47 and B&W Owners group reassessment. Most B&W plants have committed to install safety grade systems to prevent SG overfill protection. CRGR has approved a package, which includes a proposed generic letter for resolution of the issue. The generic letter is currently being reviewed by several NRC affices. C-205-1 E-12
% v . AE00 RECOMMENDATION TRACKING SYSTEM RECOMMEN0ATION SOURCE: Case Study AE00/C301 , Responsible AE00 Engineer: M. Chiratal TITLE OR
SUBJECT:
"Failure of Class IE Safety-Related Switchgear Circuit Breakers to Close to Oemand" ,
RECOMMENDATION 1 l Provide for monitoring the status of the closing circuit of Class IE circuit breakers and for appropriately selected breakers such as diesel generator output breakers. Make the status indication available to the control room operator. RESPONSIBLE OFFICE /0IV/BR CONTACT PRIORITY RES/EIB R. Baer Drop STATUS t NRR is reprioritizing Generic Issue 55 based on fieference 3. GI-55 < reptioritized as Low. AE00 agreed to drop. _ RECOMMENDATION 2 In the short-term, licensees of operating reactors should establish regular ! local surveillance of Class 1E switchgear circuit breakers to monitor the ! readiness status of the spring-charging motor of each unit. I RESPONSIBLE . OFFICE /DIV/BR CONTACT PRIORITY
< t NRR/OSR0/SPEB K. Kniel N/A ,
STATUS (See status of Recommendation 1.) i RECOMMENDATION 3 In addition to the above, measures that tend to preclude dirty or corroded , contacts, poor electrical connections, blown control circuit fuses, and ' improper return of breakers to operable status should be incorporated into the maintenance procedures and used in actual maintenance practice. C-301-1 , 4 E-13 , i 4
n . - - . . _ _ - _ . AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION
^
SOURCE: Case Study AEOD/C301 (continued) Respo.isible AE00 Engineer: M. Chiramal TITLE OR
SUBJECT:
"Failure of Class IE Safety-Related Switchgear Circuit Breakers to Close to Demand RECOMENDATION 3 (continued)
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/OSR0/SPEB K. Kniel N/A STATUS Delete. (See status of Recommendation 1.) RECOMENDATION 4 Shift operating personnel should receive periodic training in the logic and operation of circuit breakers equipped with anti pumping controls. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR K. Kniel N/A V. Thomas S_TATUS Resolved. IE IN 83-50 issued and followed up. '
REFERENCES:
- 1) Memo to C. J. Heltemes, Jr. from H. R. Denton, June 17, 1983, i "AE00 April 1983 Report on Failure of Class 1E Safety-Related '
Switchgear Circuit Breakers to Close on Demand"
- 2) Memo to D. G. Eisenhut from R. L. Spessard, June 1, 1984, 1
"Unmonitored Failures of Class 1E Safety-Related Switchgear Circuit Breakers to Close on Demand"
- 3) Memo to R. M. Bernero from H. R. Denton, March 27, 1985,
, "Scheduled for Resolving and Completing Generic Issue No. 55
-FailureofClass1ESafety'-RelatedSwitchgearCircuit !
Breakers to Close on Demand t
- 4) Hemo to H. R. Denton from C. J. Heltemes, Jr. , April 12, 1985, "Generic Issue No. 55 - Failure of Class 1E Safety-Related l Switchgear Circuit Breakers to Close on Demand" ,
- 5) Memo to C. J. Heltemes, Jr. from H. R. Denton, May 9,1985, 4 "AE00 Concerns Regarding Generic Issue No. 55" C301-2 l l E-14 ,
i
l AE00 RECOMMENDA110N TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE00/0401 and Memorandum from C. J. Heltemes, Jr. to , H. Denton, dated March 16, 1984 i Responsible AE00 Engineer: S. Salah TITLE OR
SUBJECT:
"Low Temperature Overpressure Events at Turkey Point Unit 4" RECOM4ENDATION 1 Correct the LTOP technical specifications for the five areas identified in the report.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY l RES/DSPS/RPSIB E. Throm High l STATUS ! This study has been transferred to RES from NRR as of April 12, 1987, with the reorganization of NRC. Previously, NRR had identified this activity as Generic Issue 94. Also, NRR/OSR0 had revised the prioritization to "High" based on our memorandum dated June 3, 1985. Oak Ridge National Laboratory performed an LER search and provided the results to NRR (letter from G. Mays i to E. Throm dated September 2, 1986). NRR had also signed a contract with PNL to perform some of the work related to this generic issue. Presently PNL work is nearing completion and RES has prepared a CRGR package which is currently in the review process as of September, 1988. )I i C-401-1
- E-15
-. _ - _ - - . . . . . _ _ - _ _ - . - . - - _ . . _ . _ . _ - . _ _ _ - , - - - . . ~
AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C402 Responsible AE00 Engineer: H. Ornstein TITLE OR
SUBJECT:
"Operating Experience Related to Moisture Intrusion in Electrical Equipment at Commercial Power Reactors"
- RECOMMENDATION 1 IE should revise the inspection program to ensure licensee adherence to NRC requirements.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY
- NRR/ILRB M. Johnson N/A STATUS Resolved. This recommendation was addressed by including them in five i separate maintenance procedure-related IE inspection modules.
1 L 4 t t. l C-402-1 ( l
"-16 j
AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AEOD/C403 Responsible AE00 Engineer: T. Cintula TITLE OR
SUBJECT:
"Edwin I. Hatch Unit 2 Plant Systems Interaction Event on August 25, 1982" RECOMENDATION 1 Evaluate the common mode failure potential of safety systems due to the harsh environment of breaks outside containment being back channelled through floor drain systems.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY
- RES/DE/EIB D. Thatcher High STATUS This recomendation was originally to be evaluated as part of Generic Issue In 77, 1986"Backflow Protection NRR consolidated GI-77ininto Comon A-17, Equip' Systems Interactions."ent The staff and Floor Dr prepared a proposed resolution packags for A-17 which predominantly involves comon-mode flooding of nuclear plant equipment. The package was transmitted to all NRC offices and ACRS for coment and is currently being revised to address the coments received in preparation for CRGR review.
The status of this recomendation is the saide as last year and no additional milestones were scheduled or achieved for Generic Issue A-17. RECOMENDATION 2 Supplemental arrangements should be provided to assure timely isolation of the affected floor drain system if the results of the above evaluation result in unacceptable comon-mode safety system failures. RESPONSIBLE OFFICE /DIV/BR CONTAC1 PRIORITY RES/DE/EIB D. Thatcher High STATUS (See status of Recomendation 1.) C-403-1 E-17
AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Memorandum from C. J. Heltemes, Jr. to H. Denton, dated July 23, 1984 Responsible AE00 Engineer: S. Israel TITLE OR
SUBJECT:
"Steam Binding of Auxiliary Feedwater Pumps" RECOMMENDATION 1 PWR licensees should establish a method to regularly monitor the AFW system to minimize the potential for steam binding.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/DSR0/RSIB A. Spano High l IE/DEPER/EGCB V. Hodge 1 STATUS Resolved. This recommendation has been implemented by a generic letter on February 17, 1988 which embraces the continued enforcement of IE Bulletin 85-01. C-404-1 E-18
AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AEOD/C501 (NUREG/CR-3551) Responsible AE00 Engineer: H. Ornstein TITLE OR
SUBJECT:
"Safety Implications Associated with In-Plant Pressurized Gas Storage and Distribution Systems in Nuclear Power Plants" RECOMENDATION 2 Require protection to prevent hydrogen explosions or fires in areas containing or impacting operation of safety-related equipment.
l RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/RPSIB C. Graves Medium STATUS Generic Issue No. 106 "Piping and the Use of Highly Combustible Gases in VitalAreas"wasorigInallyprioritizedbyPNLforthestaffandgivena low priority. The priority of this issue was reevaluated. Reprioritization was completed in December 1987, resulting in a medium priority. A task action plan is currently being formulated for resolution of this issue. d i a L C-501-1 [ t E-19 -
. __ ,__ ___ _ _ _ _ , _ _ , . _ _ . ~ . - _ _ _ - . _ _ _ . . .
AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AE00/C502 Responsible AE00 Engineer: P. Lam TITLE OR
SUBJECT:
"Overpressurization of Emergency Core Cooling Systems in Boiling Water Reactors" RECOMENDATION 1 , Disable the nonsafety-related air operator associated with the testable isolationcheckvalveontheinjectionlineintheemergencycorecooling systems.
RESPONSIBLE OFFICE /DIV/Bt CONTACT PRIORITY RES/EIB 0. Rothburg High I STATUS Assigned as Generic Issue 105 "Interfacing Systems LOCA at BWRs." BNL has completedanevaluationandrIskanalysisoftheissuecoveringbothBWMs and PWRs. The draft report is now being reviewed by EIB. . RECOMENDATION 2 I Perform leakage testing of the testable isolation check valve prior to plant startup after each refueling outage or following maintenance, repair or replacement of the valve. < l l RESPONSIBLE ) 0FFICE/DIV/BR CONTACT PRIORI,TJ j RES/EIB 0. Rothburg High STATUS t i Assigned as Generic Issue 105. BNL has completed an evaluation and risk ! 4 analysis of the issue covering both BWRs and PWRs. The dr:ft re,nort is now [ being reviewed by EIB. RECOMENDATION 3 Reduce human errors in maintenance and surveillance testing activities.
- i RESPONSIBLE ,
OFFICE /DIV/BR CONTACT PetIORITY r RES/EIB 0. Rothburg High ; STATUS (See status of Recommendation 1 above.) s C-502-1 l 4 ] E-20
AE00 REC 0 win 0ATION TRACKING SYSTEM - RECOM4ENDATION SOURCE: Case Study AEOD/C502 (continued) Responsible AE00 Engineer: P. Lam TITLE OR
SUBJECT:
"0verpressurization of Emergency Core Cooling Systems in Boiling Water Reactors" RECOW4ENDATION 4 Study reducing the frequency of surveillance testing of the isolation barriers of the emergency core cooling systems during power operation, i RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE/EIB 0. Rothburg High STATUS l
(See status of Recommendation 1.) l h J l i I i 1 i C-502-2 4 j E-21 1
.--. - , . . . - - . = _ . .
AEOD RECOMENDATION TRACKING SYSTEM RECOMEN0ATION SOURCE: Case Study 4 0D/C503 Responsible AEOD Engineer: H. Ornstein TITLE OR
SUBJECT:
"Decay Heat Removal Problems at U.S. PWRs" RECOMENDATION 1 NRR assess the need for NRC requirements to improve planning, coordination, procedures, and personnel training during shutdown to ensure the avail-ability of the DHR system.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/RPSIB A. Spano High ; STATUS l Generic Issee 99, "RCS/RHR Suction Line Valve Interlock on PWRs" has been expanded to include all of the recommendations of C503. RES has contracted with BNL to oerform a PRA on the effects of implementing thir recommendation 4 j at the Zion plant. Methods of extrapolating the work for Zion to other 4 plants are also to be studied. The study was published as NUREG/CR5015 in May, 1938. l In May 1987, AE00 issued a supplemental report, S702, addressing loss of DHR during mid-loop operation (see writeup on S702). GI-99 is also j addressing this issue. Information Notice issued May 1987 (IN 87-23). Generic letter 87-12 sent to PWR owners July 1987. NRR prepared a generic
- letter requiring corrective action which was approved by the CRGR on August 24, 1988. The generic letter will be issued in the Fall, 1988.
RECOMMENDATION 2 j NRR require PWR licensees to have a reliable method of measuring and
- monitoring reactor vessel level during shutdown modes of operation and corresponding technical specification requirements for operability.
RESPONSIBLE OFFICE /DIV/BR CON 1ACT PRIORITY NRR/0$R0/RSIB A. Spano High STATUS (See status of Recommendation 1.) C 503-1 E-22
t i i AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION i SOURCE: Case Study AE00/C503 (continued) { Responsible AE00 Engineer: H. Ornstein TITLE OR
SUBJECT:
"Decay Heat Removal Problems at U.S. PWRs" i RECOMENDATION 3 f NRR require licensees to improve the man-machine interfaces related to DHR i cperation.
RESPONSIBLE , OFFICE /DIV/BR CONTACT PRIORITY 3 NRR/DSR0/RSIB A. Spano High , STATUS 1 (See status of Recommendation 1.) RECOMEN0ATION 4 NRR should consider DHR suction bypass lines as alternatives to redundant drop lines (if A-45 concludes that single drop line configuratior.s are unacceptable). RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY HRR/OSR0/RSIB A. Spano High STATUS (See status of Recommendation 1.) RECOMENDATION 5 NRR consider removal of autoclosure interlocks to minimize loss-of-OHR events. RESPONSIBLE OFFICE /DIV/BR CONTACT @0RITY NRR/DSR0/RSIB A. Spano High STATUS (See status of Recommendation 1.) C-503-2 E-23
AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE00/C503 (continued) Responsible AE00 Engineer: H. Ornstein TITLE OR
SUBJECT:
"Decay Heat Removal Problems at U.S. PWRs" RECOMMEN0ATION 6 NRR should address the issue of DHR system redundancy to ensure that the OHR system is available during Mode 4 and the early stages of ".>de 5.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/DSR0/RSIB A. Spano High STATUS (See status of Recommendation 1.) C-503-3 E-24
AE00 RECOMMENUATION TRACKING SYSTEM ! REC 0lWENDATION SOURCE: Case Study AEOD/C504 Responsible AE00 Engineer: E. Trager [ i TITLE OR SUCJECT: "Loss of Safety System Function Events" REC 06NEN0ATION 2 , NRR review the Maintenance and Surveillance Frogram Plan, the Human Factors Program Plan, and the INPO training accreditation program to ensure the adequacy of training programs for all types of nuclear power plant personnel. ! RESPONSIBLE OFFICE.(DIV/BR CONTACT PRIORITY NRR/DCEA/0GCB C. Berlinger lo STATUS t Resolved. A NRR memorandum dated August 18, 1987, states that NRR continues t to work on this issue in the Division of Licensee Performance and Quality '
- Evaluation. Specifically, NRR staff actions to ensure the adequacy of train- !
, ing for all types of nuclear power plant personnel are described as follows: Maints. nance and Surveillance Program Plan. 1 The Performance Evaluation Branch has been activel'/ working to develap the I Maintenance and Surveillance Program Plan. NRC staff efforts ace now r directed toward assessing the effectiveness of industry efforts in this ! area using performance indicators and SALP reports by observing INPO plant ! evaluationsandbyreviewingmaintenanceprogramsInconjunctionwithNRC team inspections and assessments. ! INPO Teaining Accreditation Plan: The Human Factors Assessment. Granch is responsible for monitoring the effectiveness of this INP0-managed progrma through a variety of methods ! including observations of Accreditation Team visits, observations at the l National Nuclear Accreditation Board, review of plant specific self-evalua- ' tion reports and INPO Team Reports, review of Training Inspections, and SALP post-accreditation reviews of HRC examination results. ! Human Factors Program Plan: ! 1
) Responsibility for the resolution of human factors generic issues has been j transferred to RES. The Program Plan for Htman Factors Research was j developed with input from NRR and otner NRC offices and considers recommen-dations from the National Academy of Sciences.
l Based on these and other NRC activities in this area, NRR actions on this l reconnendation were found to be acceptable. i C' t-1 :
- c6 f
AE00 RECOMMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AEAD/0505 Responsible AE00 Engineer: S. Pettijohn TITLE OR
SUBJECT:
"Therapy Hisadministrations Reported to the NRC Pursuant to 10 CFR 35.42" RECOMMENDATION 2 The Office of Nuclear Material Safety and Safeguards should consider the following actions in regard to establishing quality assurance requirements for radiotherapy facilities licensed by NRC:
-- Contact appropriate professional organizations to encourage and support the initiation of a volur.tary, industry-directed physical quality assurance program for radiotherapy facilities. We believe that the commitment of the professional organizations in this regard should be assessed by the NRC and a conclusion reached as to the effectiveness of the voluntary program within two years.
-- If substantial progress toward completion of the voluntary program, including a final completion date, has not been demo strated at the end of two years, we recommend that NHSS initiate the necessary studies to deturmine whether a rulemaking is justified to require that radiotherapy facilities licensed by HRC have quality assurance programs to insure the accuracy of patient doses. The program should include such things i as: independent verification of patient dose calculations and indepen-dent verification of the activity of brachytherapy sources before the sources are implanted.
-- The voluntary quality assurance program should contain a+,least the elements outlined above.
RESPONSIBLE OFFICE /DIy/BR CONTACT PRIORITY NMSS R. Cunninght. N/A t STATUS Resolved. This recommendation has been addressed in current rulemaking on quality assurance for radiotherapy facilities, "Base Quality Assurance in Radiation Therapy." C-505-1 E-26
AE00 RECOMENDATION TRACKING SYSTEM RECOW".NDATION SOURCE: Case Study AEOD/C505 (continued) Responsible AE0D Engineer: S. Pettijohn TITLE OR SUBJECT 1 "Therapy Misadministrations Reported to the NRC Pursuant to 10 CFR 35.42" RECOMENDATION 3 10 CFR Part 35.21 should be amended to include the calibration of beam modifiers such as wedge filters, shaping filters, trays, etc. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NMSS/FCMC R. Cunningham N/A STATUS Resolved. This recommendation has been addressed satisfactorily in the current rulemaking on quality assurance for radiotherapy facilities, "Basic Quality Assurance in Radiation Therapy." RECOMENDATION 4 In addition, to the extent that the NRC implements Re:ommendation 3, the action should be made an item of compatibility for Agreement States. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY , GPA/SLITP D. Nussbaumer N/A STATUS Action will be taken st.bsequent to Recommendation 3. I C-505-2 E-27
AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C601 Responsible AE00 Engineer: S. Pettijohn TITLE OR
SUBJECT:
"Rupture of an Iodine-125 Brachytherapy Source at the University of Cincinnat', Medical Center" RECOMMENDATION 2 RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NMSS/FC/FCML K. Smith N/A STATUS 2 Resolved. 3M Company modified package inserts for Iodine-125 seeds to provide clearer information to licensees.
RECOMMENDATION 3 RESPONSIBLE OFFICE /DIV/BK CONTACT PRIORITY NHSS/FC/FCML K. Smith N/A [ STATUS NMSS is evaluating the prerogatives of licenseas and the context of the i recommendation to determine to what extent the recommendation might be , implemented. [ t [ i i C-601-1 l l E-28 1
AE00 RECOMENDATION 1 RACKING SYSTEM RECOMEN0ATION SOURCE: Case Study AEOD/C602 Responsible AE00 Engineer: C. Hsu TITLE OR
SUBJECT:
"Operational Experience Involving Turbine Overspeed Trips" RECOMENDATION 1 Alleviate the effect of slow response of the governor valve during the turbine startup transient by implementing a steam bypass modification for turbines equipped with a Woodward Model EG governor.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/PRAB S. Diab High STATUS
- Included in ongoing Generic Issues 122 and 124.
RECOMENDATION 2 Avoid entrapped oil in the saeed setting cylinder by establishing administrative controls for aleeding off the entrapped oil, installing a controllable dump valve, or providing indication in the control room for a spinning turbine. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/PRAB S. Diab High STATUS i Included in ongoing Generic Issues 122 and 124.
; RECOMENDATIGN 3 Ensure the adequacy of the existing vendor-supplied calibration procedures for governor speed setting.
RESFONSIBLE g 0FFICE/DIV/BR CONTACT PRIORITY NRR/PRAB S. Diab High i STATUS Included in ongoing Generic Issues 122 and 124. l C-602-1 i E-29 i i
. _ ~ _
AE00 RECOMENDATION TRACKING SYSTEM J RECOMENDATION SOURCE: ___ Case Study AE00/C602 (continued) Responsible AE00 Engineer: C. Hsu TITLE OR
SUBJECT:
"Operational Experience Involving Turbine Overspeed Trips" RECOMENDATION 4 Prevent water induction into the turbine by providing adequate provisions for condensate removal from the steam supply line to the AFW turbine.
RESPONSIBLE
, OFFICE /DIV/BR CONTACT PRIORITY NPR/PRAB S. Diab High i
I STATUS l 1 Included in ongoing Generic Issues 122 and 124. l l l RECOMENDATION 5 i Minimize trip and reset problems by assessing the adequacy of the existing ; procedural instruction, upgrading the training program, and providing local [ indication as well as control room indication for trip and reset . conditions. [ RESPONSIBLE . OFFICE /DIV/BR CONTACT PRIORITY { l NRR/PRA8 S. Diab High STATUS : Included in ongoing Generic Issues 122 and 124. [ l i l Memo dated October 22, 1986 f rom H. R. Denton to C. J. Heltemes, Jr. I
REFERENCE:
I l ' f i I i C-602-2 f t E-30 i
AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AE00/C603 ResponWh AEOD Engineer: E. Brown TITLE h j uBJECT: "A Review of Motor-Operated Valve Performance" Rp ci1fp 2LQ} Expv .'ma.g tyvie, mt the recomendations in AE00/C203 (May 1982) and AEOD. a ($epueaber 1985). RESR i V . OFFIC.# ,g '.ONTACT PRIORITY NUMARC T. Tipton N/A STATUS Cerrently under review. AE00 Case Study C603 transmitted to NUMARC by Reference 1 requesting that they take actions necessary to implement the recomendations in the case study. In Reference 2 NUMARC agreed to initiate a program to address the concerns and recomendations. NUMARC met with NRC staff on September 1, 1987 to discuss HOV program initiated by 1n70 and EPRI. A follow-up meeting was held on March 31, 1988, and periodic reviews are planned. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE/EIB 0. Rothberg Medium RES/DE/EME S. Aggarwal Medium STATUS EIB is preparing a draft report on thermal overload devices. This report will be used as input to a proposed change in Regulatory Guide 1.106. RECOMMENDATION 2 Requirt licensees to establish procedures and diagnostic capability to determine root causes of MOV failure to operate. RESPCNSIBLE OFFICE /DIV/8R CONTACT PRIORITY NUMARC T. Tipton N/A STATUS Currently under review. (Same as Recomendation 1 above.) C-603-1 E-31
AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AEOD/0603 (continued) Responsible AE00 Engineer: E. Brown, TITLE OR
SUBJECT:
"A Review of Motor Operated Valve Performance" RECOMENDATION 3 Require licensees to develop a strong training program to ensure appropriate information and instructions are disseminated to operating and maintenance personnel.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NUMARC T. Tipton N/A STATUS i Currently under review. (Same as Recommendation 1 above.) } RECOMENDATION 4 I j The scope of IE Bulletin 85-03 should be extended to cover all ! i safety-related MOV tssemblies required to be tested in accordance with . 10 CFR 50.55a(g). l RESPONSIBLE l i 0FFICE/DIV/BR CONTACT PRIORITY t ! L NUMARC T. Tipton N/A f STATUS { Currently under review. (Same as Recommendation 1 above.) RESPONSIBLE l Il OFFICE /DIV/BR CONTACT PRIORITY f
- RES/DE/EIB 0. Rothberg N/A ,
. t STATUS RES has concurrent review for this issue as part of Generic Issue II.E.6. A draft Value Impact Statemer.t (see reference) has been issued for NRC ; staff review. i i
REFERENCE:
Memo, O. Rothberg to Distribution, Brcokhaven National Laboratory l j (BNL) Draft Report. "Value-Impact Analysis for Extension of NRC i Bulletin 85-03 to cover all Safety-Related Motor Operated Valves," , February 17, 1988. ( l
- C-603-2 E-32 !
AE00 RECOMENDATION TRACKING SYSTEM RECOMENDATION SOURCE: Case Study AE00/C604 Responsible AE00 Engineer: M. Chiramal TITLE OR
SUBJECT:
"Effects of Ambient Temperature on Electronic Components in Safety-Related I&C Systems"
,RECOMENDATION 1 Procedures for (a) loss of HVAC systems supplying instrumentation and control system equipment rooms and areas, and (b) loss of forced cooling to instrument cabinets, should be provided.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/0EAB R. Scholl N/A STATUS Action de' erred until actions addressing Recommendation 4 are empleted. RECOMENDATION 2 Training of control room and plant operators in using the procedi:res should be provided. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/0EAB R. Scholl N/A STATUS Action deferred on Recommendation 4 action. REcomENDATION 3 Supplemental cooling equipment should be readily available and identified for use in the event (of a loss of control room cooling). RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/0EAB R. Scholl N/A STATUS Action deferred on Recommendation 4 action. C-604-1 E-33
A.E00 RECOM4ENDATION TRACKING SYSTEM RECOM4EN0ATION SOURCE: Case Study AE00/C604 (continued) Responsible AE00 Engineer: M. Chiramal TITLE OR
SUBJECT:
"Effects of Ambient Temperature on Electronic Components in Safety-Related I&C Systems" RECOMENDATION 4 Periodically measure or continuously monitor the environmental conditions inside the instrument cabinets that contain heat sensitive solid-state components.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/0EAB R. Scholl N/A STATUS A temporary instruction for inspection of all licensed Westinghouse plants using a solid state arotection system is planned to determine what actions had been taken and wiat actions will be taken to resolve concerns raised by IE IN 85-89. RECOMENDATION 5 The room ambient temperature limit specified in the plant technical specifications for operability of the control room cooling and ventilation systems should reflect the actual measured temperatures in the safety-related instrumentation and control system cabinets located in the control room area. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/0EAB R. Scholl N/A STATUS Deferred on outcome of Recommendation 4 action. C-604-2 E-34
AEOD RECOW4ENDATION TRACKING SYSTEM REC 0W4EN0ATION SOURCE: Case Study AE00/C604 (continued) Responsible AE00 Engineer: M. Chiramal TITLE OR
SUBJECT:
"Effects of Ambient Temperature on Electronic Components in Safety-Related I&C Systems" RECOW9ENDATION 6 In the ongoing plant specific evaluations associated with the resolution of USI A-44, Station Blackout," the following considerations rega, ding the effects of high ambient temperature on solid-state electronic components should be included:
(a) The design adequacy should be evaluated for instrumentation and contini system equipment needed to function during and recovering from a station blackout, as well as other eouipment whose malfunction would impact operability of such equipment, ,' (b) Plant specific equipment qualification data should be required unless the equipment qualification data can be verified by actual f measurements of as-built and as-installed conditions. j RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/EIB A. Serkiz
- 1 j STATUS l Included in USI A-44.
REFERENCE:
Memo dated May 21, 1987 from T. E. Murley to E. L. Jordan 1 l i l i
! C-604-3 l
E-35 1
AE00 RECOMENDATION TRACKING SYSTEM
. RECOMENDATION SOURCE: Case Study AE00/C605 Responsible AE00 Engineer: M. Chiramal TITLE OR
SUBJECT:
"Operational Experience Involving Losses of Electrical Inverters" RECOMENDATION 1 The Office of Nuclear Reactor Regulation (NRR) should issue an information notice which addresses events involving inverter losses.
RESPONSIBLE 1 0FFICE/DIV/BR CONTACT PRIORITY NRR V. Thomas Medium STATUS Resolved. IE IN 87-24 issued. . RECOMENDATION 2 ) The circuitry which monitors the position of the circuit breakers for t l reactor coolant pump motors should be reassessed by NRR. RESPONSIBLE i 0FFICE/DIV/BR CONTACT PRIORITY NRR/0EAB R. Scholl N/A i i STATUS L i 4 Actions being coordinated with C604 Recommendation 4 actions. RECOMENDATION 3 ! Technical specifications which specifically address inverters and/or atten-dant buses for comparable plant designs should be reviewed to ensure that action statements addressing plant operating restrictions are consistant. RESPONSIBLE ; i 0FFICE/DIV/BR CONTACT PRIORITY I NRR/0EAB R. Scholl N/A 1 l STATUS (Coordinating actions with C604 Recommendation 4.) l
REFERENCE:
Memo dated May 21, 1987 from T. E. Murley to E. L Jordan j i C-605-1 l l E-36 l
l AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE00/C701 (March 1987)/NUREG-1275 Vol. 2 f. december 1987) Responsible AE0D Engineer: H. Ornstein TITLE OR
SUBJECT:
"Air Systems Problems at U.S. Light Water Reactors" RECOMMENDATION 1 Licensees should ensure that air system quality is consistent with equipment specifications and is periodically monitored.
RECOMMENDATION 2 Anticipated transient and system recovery procedures and related training for loss of air systems should be reviewed for adequacy and revised as necessary. I RECOMMENDATION 3 i Plant staff should be trained regarding the importance of air systems. RECOMMENDATION 4 The adequacy of safety grade backup accumulators for safety-related equipment should be verified. RECOMMENDATION 5 All operating plants should be required to perform gradual loss of instrument air system pressure tests. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/ DEST /SPLB J. Warmiel High RES/DRPS G. Burdick STATUS - ALL 5 RECOMMENDATIONS Two Information Notices, 87-28 and 87-P8 Supplement 1, have been issued. The issue has been prioritized as High NRR is preparing a generic communication requiring licensees to c. ry out Recommendations 1-4. Recommendation 5 will be addressed subsequently. RES has been tasked with formulating a set of operating practices / guidelines for plant air systems. C-701-1 E-37
AE00 RECOMENDATION TRACKING SYSTEM i REC 0 MEN 0ATION SOURCE: Memo: C. Michelson to R. Mattson, "NRC Action Plan Developed as a Result of TMI-2 Accident - Draft 3, Task II.E.3 Decay Heat Removal," /.pril 24, 1980 Responsible AE00 Engineer: H. Ornstein TITLE OR
SUBJECT:
"Reliability of DHR Systems" RECOMENDATION 1 Reliability of DHR systems should be reviewed and where necessary upgraded on an expedited basis.
I. RESPONSIBLE
! 0FFICE/DIV/BR CONTACT PRIORITY RES/EIB A. Spano High STATUS l Deleted. Subsumed in C503 and S702.
j + ! I i : i ; ) i ) : ) ! ' [
\
i l ! i
- 1 E-001A-1 l l E-38 [
I ,
k AE00 REC 044EN0ATION TRACKING SYSTEM RECOW9EN0ATION __ SOURCE: 1. Memorandum dated July 15, 1980 from C. Michelson to H. R. Denton
- 2. Memorandum dated August 27, 1980 from C. Michelson to
, H. R. Denton : Responsible AE00 Engineer: M. Chiramal TITLE OR
SUBJECT:
"Operaticnal Restrictions for Class IE 120 V/ic Vital -
Instrument Buses" "Tie Breaker Between Redundant Class IE Buses - Point Beach , Nuclear Plant Units 1 & 2" !
; RECOW4ENDATION 1 Interconnection between redundant safety-related electrical load groups ,
should comply with requirements of Regulatory Guide 1.6. l i RESPONSIBLE
, OFFICE /DIV/BR CONTACT PRIORITY t RES/EIB D. Theitcher High !
STATUS l Currently being pursued under Generic Issue 128. Draft report issued for l comment, < i ,
REFERENCES:
- 1) Memo dated September 29, 1980 from H. Denton to C. Michelson :
- 2) Memo dated October 25, 1983 from H. Denton to D. Eisenhut
- 3) Memo dated October 16, 1980 from H. Denton to C. Michelson i
l l E-005-1 E-39
( , AE00 REC 0W4ENDATION TRACKING SYSTEM 3 REC 0W4EN0ATION 1 SOURCE: Memorandum to H. R. Denton from C. Michelsor, dated January 19, 1981 ; ) j Responsible AE00 Engineer: E. Brown TITLE OR
SUBJECT:
"Degradation of Internal Appurtenances in LWRs" l RECOW4ENDATION 1 RESPONSIBLE 0FFICE/DIV/BR CONTACT PRIORITY i RES H. Vander Molen Low STATUS l
J Dropped. Low safety significance, l
REFERENCE:
Memo, K. V. Seyfrit to W. Minners February 15, 1985, "Evaluation i ofGenericIssueNo.35,DegradatIonofInternalAppurtenancein LWRs" l ! I t l i 1 ) i ) j E-101-1 E-40
i i AE00 REC 0mENDATION TRACKING SYSTEM l REC 0 MEN 0ATION
- SOURCE: Memorandum from C. Michelson to V. Stello and H. Denton, "Immediate Action Memo: Common Cause Failure Potential at Rancho Seco - Dessicant Contamination of Air Lines," l September 15, 1981 Responsible AE00 Engineer: H. Ornstein TITLE OR
SUBJECT:
"Plant Air Systems" l l
i RECOMENDATION 1 Obtain licensee's experience and assessment of this problem and determine j course of corrective action if required. , RESPONSIBLE !
- 0FFICE/DIV/BR CONTACT PRIORITY ;
I : RES W. Milstead Low / Drop ! l
- STATUS j l Delete. Incorporated into C701.
) l
)
REFERENCE:
Memo from C. J. Heltemes to H. R. Denton, "Contamination of [ Instrument Air Lines," December 14, 1983 , 1 l [
! l \
l l , i J
] ! h i ;
I [ E-123-1 i r
] E-41 i i l 1 ~
l
AE00 RECOMENDATION TRACKING SYSTEM RECOMEN0ATION SOURCE: Memorandum from C. Michelson to R. Vollmer and R. Mattson dated February 24, 1982 Responsible AE00 Engineer: M. Chiramal TITLE OR
SUBJECT:
"Spurious Trip of the Generator Lockout Relay Associated with a Diesel Generator Unit" RECOMENDATION 1 Should explicitly verify that seismic qualification of all protective devices used in the control and protection circuitry of DG units has been performed with these devices in their energized, de-energized, tripped and non-tripped states.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/EIB T. Chang High STATUS Issue has been incorporated into USI A-46. USI A-46 is being implemented.
REFERENCE:
Memorandum from H. R. Denton to C. Michelson dated May 11, 1982 E-212-1 E-42
,1 A_E00 RECOMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Engineering Evaluation AE00/E215 Responsible AE00 Enginett: T. Cintula TITLE OR
SUBJECT:
"Salt Water System Flow Blockage at Pilgrim NPS by Blue Mussels" RECOMMENDATION 1 Internal inspection of RBCCW HX supply headers RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE J. Jackson Medium STATUS Delete. Incorporated into C202.
RECOMMENDATION 2 Periodic measurement of overall heat transfer coefficient on RBCCW HXs at Pilgrim RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE J. Jackson Medium STATUS Delete. Incorporated into C202. RECOMMENDATION 3 Periodic measurement of Salt Water System flow to RBCCW HXs RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE J. Jackson Medium STATUS Delete. Incorporated into C202. E-215-1 E-43
1 4 AE00 RECOMENDATION TRACKING SYSTEM RECOMEN0ATION SOURCE: Engineering Evaluation AE00/E304 Responsible AE00 Engineer: T. Cintula TITLE OR
SUBJECT:
"Backflow Protection in Common Equipment and Floor Drain Systems" l!
RECOMENDATION 1 Provide backflow protection for drain systems in older operating plants. RESPONSIBLE
- 0FFICE/DIV/BR CONTACT PRIORITY RES/DE/EIB D. Thatcher High STATUS This recommendation was initially prioritized as high and subsequently became Generic Issue 77, "Backflow Protection in Common Equipment and Floor j Drain Systems." In 1986 NRR consolidated GI-77 into A-17 "Systems 1 Interaction." The staff prepared a proposed resolution package for A-17
- which principally involves common mode flooding of nuclear plant vital 1 equipment spaces. The package was transmitted to all NRC offices and ACRS i
for comment and is currently being revised to address the comments received ,
- in preparation for CRGR review. The status of this recommendation is
- unchanged from last year and no additional milestones were scheduled or ,
- achieved for Generic Issue A-17 in the past year. i t
i I i i 3 4 t i i k r i ! i i i E-304-1 E-44
AE00 REC 0fe4ENDATION TRACKING SYSTEM REC 0mENDATION SOURCE: Special Study Report - C. J. Heltemes, Jr. to H. R. Denton dated January 13, 1984 and follow-up memorandum dated August 8, 1984 Responsible AE00 Engineer: E. Trager TITLE OR
SUBJECT:
"Human Error in Events Involving Wrong Unit or Wrong 1 rain" RECOMENDATION 1 Consider the need for further clarification or guidance on what constitutes an acceptable independent verification program.
RESPONSIBLE OFFICE /DIV/BR CCYlACT PRIORITY NRR/0HFT/MTB G. Cwalina High STATUS The reference stated that "The NRC should provide clarifying guidance regarding independent verification." NRR plans to request that RES prepare regulatory guidance on activities such as independent verification programs to minimize the potential for their occurrence.
REFERENCE:
"An Investigation of the Contributions to Wrong Unit or Wrong Train Events." NUREG-1192. April 1986.
RECOMEN0ATION 2 NRR review wrong unit / wrong train events and develop appropriate guidance to minimize such events. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY NRR/DLPQE/HFAB R. Eckenrode High STATUS NRR staff is preparing a memo describing actions that have and will be taken to resolve Generic Issue 102 "Huuan Error in Events Involving Wrong Unit or Wrong Train" to resolve this item in 1988. 5-401-1 E-45
AE00 REcomENOATION TRACKING SYSTEM RECOMMENDATION SOURCE: Special Study Report AE00/S503 and Memorandum, C. J. Heltenes to H. Denton dated September 19, 1985. (See also recommendations related to AE00/C203 on pages C203-1 and C203-2.) Responsible AE00 Engineer: E. Brown TITLE OR
SUBJECT:
"Evaluation of Recent Valve Operator Motor Burnout Events" RECOMMENDATION 1 In view of the more than 200 motor burnout events, the NRR plan to address motor burnout should be expedited.
RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY RES/DE/EIB 0. Rothberg Medium STATUS Delete. Incorporate in C603. , I f I L l i S-503-1 i E-46
- _ _ _ _ . , _ _ _ - . _ . . . _ _ , _ _ _ _ _ _ _ . _ _ . _ . _ . _ _ _ _ _ _ , _ _ . _ _ _ _ __ __l
AE00 RECOMENDATION TRACKING SYSTEM REcoletENDATION SOURCE: Special Study Report AE00/S602 Responsible AE00 Engineer: J. L. Crooks TITLE OR SUSJECT: "An Overview of Nuclear Power Plant Operating Experience Feedback Programs" REC 0 MEN 0ATION 1 NRR should consider issuing an Information Notice (IN) to licensees to communicate the conclusions from this AE00 study and to suggest that licensees review their operating experience (0E) activit?as for effectiveness and for the intent and scope of conformance with the existing NRC requirements. The IN should note that licensees and the industry have lead responsibility for ensuring that the OE review activities are effective and reemphasize that the NRC conside's the OE activities important to the safety of operations. The findings from the three IIT reports can bt 'ited in this regard. Specifically, the ! IN should suggest that licensees, consistent with existing NRC requirements: { (a) Emphasize root cause analysis and implementation of permanent
. corrective actions to prevent recurrence of safety-related 1 in-house events; j (b) Be diligent in the screening and assessment of industry feedback i
for applicability to their plant and in implementing corrective actions when appropriate; (c) Disseminate information to non-supervisory plant operators, trades
- personnel, technicians and other support staff both directly and I through training and retraining to keep them aware of anomalous i events and their actual or potential consequences, and; (d) Implement self-monitoring of the effectiveness of their OE review j
activities through management reviews. I RESPONSIBLE 0FFICE/DIV/RR CONTACT PRIORITY i NRR/00EA/0GCB V. Hodge Medium l. l STATUS In July 1986, an IN was drafted that addressed $602 findings and conclusions, but the IN was not issued. AE00 issued a summary of $602 in i NUREG-1272. Report to the U.S. Nuclear Regulatory Commission on Analysis and Evaluation of Operational Data - 1986. l ] S-602-1 E-47
l l l i l , t AE00 RECOMENDATION TRACKING SYSTEM : RECOMENDATION i SOURCE: Special Study Report AE00/5602 (continued) ! l Responsible AE00 Engineer: J. L. Crooks : i l TITLE OR
SUBJECT:
"An Overview of Nuclear Power Plant Operating Experience l
Feedback Programs" REC 0mENDATION 1 (continued) l l RESPONSIBLE ; i 0FFICE/DIV/BR CONTACT PRIORITY ! NRR/DOEA/0GCB V. Hodge Medium STATUS (continued) A contract effort was initiated in mid-1986 to determine options for ; improving generic communications with licensees. The contract study was i completed in late 1987 and a report issued in January 1988 - Evaluation and i Proposed Improvements to Effectiveness of U.S. Nuclear Regulatory Commission l Generic Communications (NUREG/CR-4991). The recommendations for action are ; currently under staff review. REcomENDATION 2 ; AEOD should have further discussions with the industry on additional initiatives, such as on means to increase the effectiveness of OE review activities, and on assessing the effectiveness of the OE review activities. RESPONSIBLE OFFICE /DIV/BR CONTACT PRIORITY AE00/DSP/TPAB J. Crooks Continuing activities STATUS Resolved. Het with INP0 in February and June 1987 to discuss their initiatives based on their findings in this area. One major initiative focused on the utility's implementation of INP0's recommendations for corrective actions from SOERs. This area is to receive additional attention in plant evaluations. Other improvements were made to i program guidance (e.g., good practice documents) and the evaluation techniques for assessing the effectiveness of utility OE programs. TPAB also included discussions on licensee's OE fesdback programs during plant visits related to "new plant" and "plant restart after extended shutdown" activities. Future plans are to visit additional licensee facilities, meet further with industry organizations, and evaluate the current status of industry OE feedback activities. 5-602-2 j E-48 i 1
AE00 RECO MENDATION TRACKING SYSTEM REC 0 MEN 0ATION SOURCE: Special Study Report AE00/5602 (continued) Responsible AE00 Engineer: J. L. Crooks TITLE OR
SUBJECT:
"An Overview of Nuclear Power Plant Operating Experience Feedback Programt,"
RECOMEN0ATION 3 The NRC should continue to monitor operatin experience and industry performance in OE activities for some time two years or so). If after this period of time sufficient improvement n effectiveness has not been observed, then further regulatory action would be a)propriate. In this regard, AE00 believes the following actions should se initiated at this time: (a) NRR, in con; unction with AE00, should reappraise how the operating experience feedback activities might be improved through further guidance. The guidance could address the shortcomings noted in this study. The elements of an acceptable program could be defined in sufficient detail to permit determinaticas of program adequacy. Further, the guidance could address that both in-house and industry-wide experience are to be thoroughly assessed; that the applicable lessons are to be fed back to the various non-supervisory operating staff through direct feedback information systems and training programs; and that corrective actions are to be taken through procedure changes, training, and plant modifications. The guidance developed 5%old not, however, be overly prescriptive, and should be consistent with present requirements. (b) Based upon the above guidance, NRR should davelop a draft inspection module for reviewing the effectiveness of licensee practices for implementing operating experience feedback activities, including verification by NRC on a plant-by plant basis (at least annually). (c) NRR should consider revising current OE fevdback reviews to routinely include a sampling of industry feedback docus, ants. RESPONSIBLE OFFICE /0!V/BR CONTACT PRIORITY AEOD/DSP/TPAB J. Crooks Various NRR/ILRB F. Febdon STATUS All groups are monitoring the 05 feedback programs as part of their normal functions. 5-602-3 E-49
i l i AE00 RECOMEN0ATION TRACKING SYSTEM RECOMENDATION SOURCE: Special Study Repc t AE00/$602 (continued) l Responsible AE00 Engineer: J. L. Crooks TITLE OR
SUBJECT:
"An Overview of Nuclear Power Plant Operating Experience Feedback Programs" RECOMENDATION 3 (continued)
RESPONSIBLE OFFICE /0!V/BR CONTACT PRIORITY AE00/OSP/TPAB J. Crooks Various NRR/ILRB F. Hebdon STATUS (continued)
- NRR and AE00 are assessing the recommendation of the contract study on proposed improvements to generic communications, prior to developing additional guidance.
- A sampling of industry feedback documents currently exists in the .
inspection procedures, j RECOMENDATION 4 I Regarding OE feedback, as soon as practical NRR and AE00 should consider j consolidating existing systems (e.g. , generic letters that deal with oper- o ational events, information notices that discuss operational events, and [ Power Reactor Events) for transmitting operational experience into a single i "NRC Notice" system. This "notice" would constitute the sole system for ! the feedback of operating experience to licensees for appropriate action j and closecut. This system could specify the initiating office and the purpose (i.e. , OE feedback), provide an indication of the significance and of the priority of attention expected an standard distribution list. InadditIon,dcouldbesenttoasinglethe system couldi information available for licensees to fully understand the concerns for [ assessment purposes and to conduct training when appropriate. In light of ! several licensee suggestions, consideration should be given to transmitting i these "notices" via an electronic mail system. RESPONSIBLE OFFICE /0!V/BR CONTACT PRIORITY AE00/DSP/TPAB J. Crooks various NRR/00EA/0GCB J. Ramsey NRR/00EA/0GCB V. Hodge b S-602-4 E-50 : i t (
i n ; I AE00 REC 0l#4ENDATION y ACKING SYSTEM i RECOW4ENDATION ! SOURCE: Special Study Report AE00/S602 (continued) l Responsible AE00 Engineer: J. L. Crooks f TITLE OR
SUBJECT:
~ "An Overview of Nuclear Power Plant Operating Experience Feedback Programs" REcole4ENDATION 4 (continued) !
STATUS Further actions are pending staff review of recommendations from the { contract effort for options on improvements to generic NRC communications. The contsactor recommended that the current system not be changed. Staff, t however, is continuing to review the issue of consolidation. l NRR is pursuing development of an electronic communication (mail) system. !
,t RECOW4ENDATION 5 3 In conjunction with the implementation of Recommendation 4 above, AE00 !
j should: (1) terminate the publication of Power Reactor Events, and (2) i { terminate the publication of the monthly Licensee Event Report (LER) { i Compilation, unless there are conflicting statutory or legal requirements i i or sufficient subscriptions to offset the cost. ! I i RESPONSIBLE I j 0FFICE/DIV/8R CONTACT PRIORITY } l; AE00/OSP/TPA8 J. Crooks low I STATUS l 4 r j Resolved. Both publications will continue. AE00 plans to revise PRE to I
! better meet current reeds. I l '
6 I r
! [
i t i . C-602-5 I 1 1 E-51 i _-- _ . _ _. _ . _ __ _~
f AE00 REC 044EN0ATION TRACKING SYSTEM_ AECOM4tN0ATION SOURCE: Special Study Report AE00/5702 Responsible AE00 Engineer: H. Ornstein TITLE OR SU6 JECT: "Loss of Decay Heat Rewoval Function at Pressurized Water Reactors with Partially Drained Reactor Coolant Systems" RECOW4ENDAT!0N 1 Licensees should maintain containment integrity to the maximum extent practicable during periods of highest DHR risk (i.e., early stages of shutdown and drain-down operations). RECOM4ENDATION 2 , Licensees should analyze the hydraulics associated with drained-down operations. (See C503 tor other related recommendations which also appear in this report.) RESPONSIPLE OFFICE /DIV/BR CONTACT PRIORITY RES/RPSIB A. Spano High , NRR/SRXB W. Lyons STATUS Information notice issued May 1987 (IN 87-23). Generic Letter 87-12 issued in July 1987. Responses have been reviewed. Presently MRR is preparing a generic communication requiring licensees to take corrective actions. As part of GI- % , BNL has written a draft report quantifying risk associated with this issue (and issues highlighted in C503). Draft results support 5702 and C503 recommendations. S-702-1 E-52
go n, u s =veu.a .covum, co .u.o., i. n,o.n uuna ,w., m., m m . , o
!'%"/df Bl8UOGRAPHIC DATA SHEET NUREG-1272
- u. n. vet.o io= , ia 2.,, Vol. 2, No. 1 2 Tif tt .NO .W 7 s?ti 3 Lt.V $L.4.
Ri or, 'o the U.S. Nuclear Regulatory Comission on j c,e and Evaluation of Operational Data - 1987 ,,,,,..,,,,,,,n,
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[ September 1988 i October 1988 T' .. _ .~.. o o. s . . . , ,o . .. . o .. .s , o . c o. . a ,,.. i. c , ...oacia. .o..v.,,. .... Office for Analysis and Eveluation of Operational Data . . . o. o. . i . . . U.S. Nuclear Regulatory Cc<nmission Washi.gton, DC 20555
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...........o,n CY 1987 This annual report of the U.S. Nuclear Regulatory Comission's 01'fice for Analysis and Evaluation of Operaticnal Data (AE00) is devoted to the activities performed during 1987. The report is published in two volumes. NUREG-1272, Vol. 2, No. 1, covers Power Reactors and presents an overview of the operating experience of the nuclear power industry, with coments regarding the trends of some key performance measures. The report also includes the principal findings and isstes identified in AE00 studies over the past year and sumarizes information from Licensee Event Reports, the NRC's Operations Center, and Diagnostic Evaluations. NUREG-1272, Vol. 2, No. 2, covers Nonreactors and presents a review of the nor, reactor events and misadministration reports that were reported in 1987 and a brief synopsis of U.00 studies published in 1987 Each volume contains a list of the AE00 Reports issued for 1980-1987.
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