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| NUREG-1272 Report to the U S. Nuclear Regulatory Commission on Analysis and Evaluation of Operational Data 1986 U.S. Nuclear Regulatory Commission Office for Analysis and Evaluation of Operational Data f a arce, y ,k itl N.h'9 ...
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| plS6'n8AES"'"g[n 1272 R
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| NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
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| : 1. The NRC Public Document Room,1717 H Street, N.W. ,
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| Washington, DC 20555
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| : 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013 7082
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| : 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.-
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| Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda: NRC Office of inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers;and applicant and licensee documents and correspondence.
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| The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC+ponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.
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| Documents available from the National Technical information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
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| Documents available from public and special technical libraries include all open literature items, such as books, joumal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.
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| Documents such as theses, dissertations, foreign reports and translations, and non NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.
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| Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
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| Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, 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 they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.
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| i NUREG-1272 l AEOD/S701 '
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| ; Re aort to the
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| 'U.S. Nuclear Regulatory Commission
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| ;on Analysis and Evaluation of iOperational Data -
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| 1986 M:nuscript Completed: April 1987 Dita Published: May 1987 !
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| Offica for Analysis and Evaluation of Operational Data l U.S. Nuclear Regulatory Commission
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| ' WIshington, DC 20666 l
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| o a a'*% i W..... ) .
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| , _ - _ _ _ _ . _ . ~ _ _ _ _ _ . _ - _ _ . _ _ . _ _ _ , _ _ . _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ . . _ , _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ .
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| ABSTRACT i
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| This annual report of the U.S. Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AE00) is devoted to the activities performed during calendar year 1986. Comments and observations are provided on operating experience at nuclear power plants and other NRC
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| ,' licensees, including results from selected AE00 studies; summaries of abnormal Cccurrences involving U.S. nuclear plants; reviews of licensee event reports and their quality, reactor scram experience from 1984 to 1986, engineered ;
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| safety features actuations, and the trends and patterns analysis program; and assessments of nonreactor and medical misadministration events. In addition, the report provides the year-end status of all recommendations included in '
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| :i AE0D studies, and listings of all AE00 reports issued from 1980 through 1986.
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| 't 111
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| CONTENTS
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| :.' Page
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| : 1. INTR 00VCTION................................................ 1
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| : 2. ORGANIZATION AND STAFFING................................... 3
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| : 3. COMMENTS AND OBSERVATIONS ON NUCLEAR F0llER PLANT OPERATING EXPERIENCE...................................... 4 3.1 Overview of Operating Experience....................... 4 3.2 Results from Selected AE00 Studies..................... 11 3.3 Abnormal Occurrences Involving U.S.
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| 4 Nuclear Power Plants................................... 18 3.4 Analysis of the 1986 Licensee Event Reporting.......... 21 3.5 Assessment of Licensee Event Report Quality............ 23 3.6 Reactor Scram Experience in 1984, 1985, and 1986....... 26 3.7 Engineered Safety Features (ESF) Actuations............ 39 3.8 NPRDS Trends and Patterns Analysis Program............. 45
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| : 4. COMMENTS AND OBSERVATIONS ON 1986 OPERATING EXPERIENCE AT OTHER LICENSEES........................................ 50 4.1 Nonreactor Events...................................... 50 4.2 Medical Misadministration Events....................... 61
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| : 5.
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| ==SUMMARY==
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| OF AE00 ACTI VI TI ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 5.1 Reactor Operations Analysis Branch (R0AB). . .... . .... ... 67 5.2 Program Technology Branch (PTB)........................ 83 5.3 Honreactor Assessment Staff (NAS) . . . . . . . . . . . . . . . . . . . . . . 97 5.4 IncidentInvestigationStaff(IIS)..................... 103
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| : 6. STATUS OF AE00 RECOMMENDATIONS.............................. 111 APPENDIX A - Summary of 1986 Abnomal Occurrences. . . . . . . . . . . . . .. . A-1 APPENDIX B - Listings of AE00 Reports, 1980 - 1986............... B-1 l
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| v
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| - , - - - _ . - _ . - ,- 3 - +- - - - w - - w - , -m, -
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| CONTENTS t
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| List of Figures Page Figure 1 - Unplanned Reactor Scrams............................. 12 Figure 2 - U.S. Nuclear Power Plants, Abnormal Occurrences VS. Year........................................... 20 Figure 3 - U.S. Nuclear Power Plants, Abnormal Occurrences / Plant Vs. Year........................................... 20 Figure 4 - LERs Submit ted in .198 6. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 Figure 5 - Distribution of Overall Average LER Scores as of December 1985...................................... 25 Figure 6 - Distribution of Overall Average LER Scores as of December 1986...................................... 25 Figure 7 - 1985 and 1986 Reactor Scram Rates vs. Critical Hours by P1 ant..................................... 29 Figure 8 - Initiating System Summary (Reactor Scrams)........... 35 Figure 9 - Primary Balance of Plant Systems (Reactor Scrams).... 36 Figure 10 - Cause Summary (Reactor Scrams)....................... 37 Figure 11 - Systems Impacted Most Frequently (Reactor Scrams).... 38 Figure 12 - Unit Distribution of 1984/1985/1986 ESF Actuations... 41 List of T_ ables Table 1 - Significant Events Meeting Selection Criteria, January throu gh June 1986. . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Table 2 - Even t Charac te r i s t ic s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Table 3 - LERs Submitted by Year................................ 21 Table 4 - LER Reporting by Report Requ irement. . . . . . . . . . . . . . . . . . . 23 Table 5 - Average Number of LERs by Major NSSS Vendors in 1986.. 23 Table 6 - Average Reactor Scram Frequency Data.................. 27 Table 7 - Plants with High Critical Hours Rate Less than Two Tr i p s pe r 1000 H ou r s , 1986. . . . . . . . . . . . . . . . . . . . . . . . . . 30 Table 8 - Licensed Reactors with Zero Critical Hours,1986...... 31 Table 9 - Reactor Trip Rates, 1986.............................. 32 Table 10 - Measured Parameters and Associated System functions for Valid ESF Actuations Occurring in 1985.......... 43 Table 11 - ESF Actuations for 1984, 1985, and 1986............... 46 Table 12 - Findings of the Flow Control Valve Studies............ 48 Table 13 - Types of Licensees that Submitted Reports During 1986. 51 Table 14 - Categorization of Nonreactor Event Reports Occurring During 1986......................................... 52 Tab le 15 - Program Support Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97 Table 16 - Events involving the Wrong Unit. Train, or System Considering Event Type and Plant Operating ExperienC0.......................................... 102
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| , Table 17 - Augmented Inspection Team (AIT) Responses............. 1 04 i
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| : vi
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| OFFICE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA ANNUAL REPORT 1986
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| : 1. INTRODUCTION NRC's Office for Analysis and Evaluation of Operational Data (AE00) was estab-
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| ! lished in 1979, as one of the Comission's earliest major steps toward improv-ing the use of licensee operating experience to identify and resolve problems with potential safety-related implications. The Office, which reports directly to the Executive Director for Operations, is dedicated to the collection, assessment, and feedback of operational data.
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| AE00's focus and role in the regulatory program is to provide a strong capabil-ity for the analysis of operating experience, independent of the routine regulatory activities associated with licensing, inspection, or enforcement and to feed back the lessons learned to the NRC, the teuclear industry and the public.
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| AE0D accomplishes its mission through analysis and evaluation of operational safety data associated with all NRC-licensed activities. These include the operations of commercial power reactor licensees and radioactive material and
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| , fuel cycle licensees. The Office also coordinates the overall NRC operational i
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| data program, and serves as the focal point for interaction with outside and fcreign organizations performing similar work. The Office's objectives and its major tasks and activities are highlighted below.
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| Objectives Collect, screen, analyze, and feed back operating experience to appropri-ate NRC offices, the nuclear community (domestic and international), and the public for all NRC-licensed activities.
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| Coordinate the overall NRC operational data program.
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| Administer the NRC's Incident Investigation Program.
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| Specific Tasks and Activities Screen U.S. and foreign operational events for significance.
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| Systematically and independently analyze operational events and issue technical reports on the results.
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| Seek trends and patterns in operating experience which may indicate l potential safety problems.
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| l
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| Develop the procedures for and support the establishment of Incident l Investigation Teams. j
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| * Develop and track AEOD recommendations for action by other NRC Offices for resolution of safety issues.
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| Develop and coordinate operational data retrieval systems, including foreign data.
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| * Prepare and coordinate Abnormal Occurrence reports to Congress and the i public.
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| ! Prepare Power Reactor Events reports and other feedback documents.
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| Provide documentation of U.S. events for reporting to the Nuclear Energy Agency's Incident Reporting System.
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| Serve as principal point of contact with ACRS, INPO, and NSAC on opera-tional data activities.
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| AEOD maintains an awareness of the studies undertaken by other organizations such as the NRC Offices of Nuclear Reactor Regulation (NRR) and Inspection and Enforcement (IE), and the Institute of Nuclear Power Operations (INPO).
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| Normally, AE00 will not overlap or duplicate the study efforts of other organ-izations unless a particular need or special circumstance exists. Thus, AE0D does not review in-depth all events or operating problems because many problems identified through operating experience are being studied extensively and comprehensively by other organizations. Thus, to be aware of all identified safety problems and lessons of experience, one must integrate the output of many NRC offices and outside programs.
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| The recommendations contained in AE00 studies discussed in the following sections are not final NRC positions. They are internal recommendations for action by appropriate NRC program offices (e.g., NRR, IE) or Regional Offices.
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| The program office or Regional Office is responsible for reviewing and, where appropriate, implementing AE00 recommendations. A written response to each recommendation is required and a formal action tracking system has been
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| ; established.
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| . 2. ORGANIZATION AND STAFFING ;
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| At the close of CY 1986, major changes in the scope of the Office's objectives !
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| ; and activities were announced as part of the pending NRC reorganization !
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| ! (April 1987). Following this reorganization, the staff will more than double, ;
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| j and additional responsibilities will include developing and managing the NRC !
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| , program for emergency and incident response, developing and providing technical '
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| ! training for NRC staff, developing and managing the NRC program for reactor I perfonnance indicators, and managing and conducting the support functions for the Committee to Review Generic Requirements. AE00's key duties of analyzing !
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| and evaluating operational safety data associated with NRC-licensed activities, j independently identifying issues that re i j findings to the appropriate recipients other (quire NRC action, and feeding offices, licensees, the back those
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| : industry, Congress,etc.)arethusenhanced. (For details, see NUREG-0325, i j Rev.10, issued February 1987.) l During 1986, a number of changes were made in AE00 management. Vacancies in i thepositionsofChiefoftheReactorOperationsAnalysisBranch(ROAB), Chief l of the Program Technology Branch (PTB), and Chief of Reactor Systems Section 1 in ROAB were filled. Also during 1986, AEOD recruited for four Reactor Systems '
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| Engineers for vacancies in ROAB, and one Plant Systems Engineer and one Reactor ,
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| Systems Engineer were added to PTB. One of these individuals is a Senior
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| {
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| Reactor Operator-licensed engineer and two are former resident inspectors. !
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| 1 AE00'sstaffinglevelforFY87(excludingthependingreorganization)is
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| ) essentially at full strength (43 FTE), and thus remains at approximately the j same level as in FY 86. The AE00 organizational structure and staffing as of December 31, 1986 are shown below.
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| I wrist pas same as me sommette w speentiema enes See letteest geleefem e5.4.
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| ee.e e. mese me.* J. meses t one.e e, enanmen one.# e, um l f $ flin - $
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| : e. 6. Aaltet mest B. Gusema j e. e I:r ,
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| 4
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| : 3. COMMENTS AND OBSERVATIONS ON NUCLEAR POWER PLANT OPERATING EXPERIENCE AE00 activities include studies to identify and evaluate potentially signif t-cant events and safety concerns involving U.S. comercial power reactors, based on events reported to the NRC by nuclear power plant licensees. During 1986, these events were reviewed individually as well as collectively by AE00 as part of its engineering assessment and trends and patterns activities. Studies were conducted that provide a broad overview of the operating experience and characteristics of the nuclear industry as well as individual facilities.
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| This section provides some observations and perspectives on selected aspects of operating experience analyzed in 1986, and provides overviews of Licensee Event Reports as well as the significant events forwarded to the Comission for consideration as Abnormal Occurrences.
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| 3,1 Overview of Operating Experience In the review of operating experience, certain common characteristics in a number of areas and a number of potentially generic issues were noted which may warrant increased NRC and industry attention. These are discussed below.
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| 3.1.1 Review of Selected Significant Events AEOD has a study in progress of significant events that occurred at nuclear power plants in the period from January 1985 to July 1986 The objectives of this study are to determine what comon characteristics the events shared, and to assess the additional lessons to be learned from the collection of events.
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| Events are included in the study if they met one or more of the following criteria:
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| The event was studied by an Incident Investigation Team (!!T) or an AugmentedInspectionTeam(AIT).
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| t
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| * The event was detemined to be an abnormal occurrence.
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| i The event had a calculated conditional probability of core damage of 1 x 10(-4) or greater (see Section 5.2.6.1).
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| A total of 21 events met one or more of the selection criteria. Sixteen of i the 21 events were operating events, four were events involving latent defects, and one event involved premature criticality during startup. Table ! lists the i
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| events considered in this study.
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| The 21 events in the study included a range of situations. Many events represent scrams or trips that were accompanied by equipment failures.
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| None of the 21 events resulted in significant plant damage, or caused a significant increase in the amount of radioactive material released from the plant. Furthemore, with a few exceptions, the significant events included in the review did not generally represent severe challenges to plant equipment or the operating staff.
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| Table 1 Significant Events Meeting Selection Criteria '
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| January 1985 through June 1986 I !
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| I i
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| ; REACTOR LER NO. DATE IIT AIT A0 CDP Operating Events
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| - DAVIS-BESSE 34685013 6/9/85 X X X i
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| SAN ONOFRE 1 20685017 11/21/85 X X X RANCHO SECO 1 31285025 12/26/85 X X D. C. COOK 2 31685035 10/29/85 X X*
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| PILGRIM 1 29386009 4/12/86 X j PALISADES 25586018 5/19/86 X
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| ! LASALLE 2 37486011 6/1/86 X X j CATAWBA 2 41486028 6/27/86 X X TURKEY POINT 3 25085021 7/22/85 X
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| ) HATCH 1 32185018 5/15/85 X r i TROJAN 34485009 7/20/85 X
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| { DAVIS-BESSE 34685002 1/15/85 X i OYSTER CREEK 21985012 6/12/85 X i HATCH 1 32185010 1/16/85 X ,
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| { GRAND GULF 1 41685050 12/31/85 X :
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| i HATCH 2 36685030 11/5/85 X 4
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| Events Involving Latent Defects ,
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| i CALVERT CLIFFS 1 N/A 7/85 X X* i l MAINE YANKEE 30985009 8/7/85 X 1
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| BRUNSWICK 2 32585008 9/27/85 X [
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| P!LGRIM N/A 5/19/86 X j
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| Startup Event !
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| SUMMER 39585003 2/28/85 X
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| ]
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| i
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| ! !!T = Event was subject of incident investigation j A!T = Event was subject of augmented inspection A0 = Event was determined to be an abnormal occurrence
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| ! CDP = Conditional core damage probability estimated to be greater than 1 x 10(-4)
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| ' Update of A0 prior to 1985. ,
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| i l
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| i J
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| j !
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| i 5
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| L __ - . . - - - . - . . _ - - . - .- - - - - - - - _ -
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| l The characteristics of the events can be divided into standard subsets:
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| ! hardware (e.g., design, common mode failures) and human factor deficiencies, i Table 2 lists the types of deficiencies identified with each of the events.
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| Table 2 Event Characteristics Deficiencies 1 4
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| Hardware Human Factors ;
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| 1 Comon System Significant Event Date Design Mode Inter- Generic j Problems Failures action Problems Procedure Training l
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| 1 Events involving Transients DAVIS-BESSE 6/9/85 X X X X X i
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| SAN ONOFRE 1 11/21/85 X X X X X X i RANCHO SECO 1 12/26/85 X X X X X X i D. C. COOK 2 10/29/85 X X X t
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| t PILGRIM 4/12/86 X X X X X
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| ; PALISADES 5/19/86 X X X X v
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| ; LASALLE 2 6/1/86 X X X X i CATAWBA 2 6/27/86 X X X X X
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| : TURKEY POINT 3 7/22/85 X X X X X X l HATCH I 5/15/85 X X X X
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| ! X TROJAh 7/20/85 X X X X
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| ] DAVIS-BESSE 1/15/85 X X X
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| 0YSTER CREEK 6/12/85 X
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| ! X HATCH 1 1/16/85 X GRAND GULF 1 12/31/85 X X X
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| { HATCH 2 11/5/85 X X 1
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| i Events involving Latent Defects I j
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| CALVERT Cliffs 1 7/85 X X X X l MAINE YANKEE 8/7/85 X X BRUNSWICK 2 9/27/85 X X X X
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| ; PILGRIM 5/19/86 X X I Startup Event ,
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| } SUMMER 2/28/85
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| ' X X I
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| f i +
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| $ i i
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| j 6 ;
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| i
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| I i ,
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| 3.1.1.1 Hardware f i It was found that 20 of the 21 events involved one or more hardware problems.
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| i In general, these could be divided into four categories, although the i categories are not independent: design problems, common mode failures, 1 system interaction problems, and generic problems. Some examples of these
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| ! included:
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| 1 1 Design Problems i -- Design did not meet requirements (e.g., a residual heat removal i i
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| system was vulnerable to single failures; an air system did not -
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| i produce the quality of air required for the associated components !
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| j toperformasintended). '
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| i j
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| Inadequate instrumentation and control misled or confused operators i
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| ! (e.g., a lack of indication of the status of steam generator blow- l
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| { down on control panels).
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| l
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| : -- Failure to understand and account for component behavior !
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| ! under actual operating conditions (e.g., incorrect specification of l
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| ) initial torque switch bypass settings; check valves specified for l use in flow conditions under which they will be damaged, such as j high turbulence for extended periods of time). i 1 Comon Mode Failures 1
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| -- Failure of a number of motor operated valves to open for the same reason (e.g.,inadequatetorqueswitchbypassswitchsettings).
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| -- Potential failure of multiple trains as the result of procedure / t design problems (e.g., residual heat removal system was susceptible !
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| to single failures multiple pressure sensing lines were unavailable ,
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| i due to an inadequate procedure). !
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| i i l --
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| Failures ascribed to degradation of identical equipment (e.o., main !
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| i stean isolation valves reactor trip breaker Static "0" Ring switches), t i !
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| * System Interactions Problems i
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| f {
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| j --
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| Interaction between nonsafety air system and safety systems, j l
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| Interaction between nonsafety integrated control system and i
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| ! safety systems. t i !
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| j -- Interaction between systems caused by electrical noise.
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| 1 l Generic Problems
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| ) --
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| In many of the events, the deficiency or failure could be generic j in nature (i.e., identical components used in other nuclear power l j plants might be subject to the same type of failure). i i
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| t 1
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| i
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| $ 7 i
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| 1
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| It can be seen (Table 2) that many of the events involved several different types of hardware problems. Design problems were seen in 19 of the 21 events; potential or actual common mode failures in 16 of 21; generic problems in 12; and system interactions in nine of the events. In some of the events, the hardware problem occurred as the result of plant modifications; in others, the event stemed from the original plant design. Essentially all of the operating $
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| events (13 out of 16) occurred at plants that had been operating for 7 or more years. All of the IIT events, and three of the five AIT events occurred at ,
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| plants that had been operating for 10 years or more. This finding points out that not all hardware problems have been uncovered and corrected early in plant life.
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| 3.1.1.2 Human Factors
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| ; The dominant human factors problems found in the 21 events were:
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| Procedures - Procedural deficiencies, use of ad hoc procedures, deviations from procedures, or lack of procedures were contributing factors to many l of the 21 events. Some examples include:
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| Failure of a procedure to give clear direction on when to take a i specific action.
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| Inservice testing procedures that were inadequate to detect a damaged l or degraded valve.
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| Training - Training was identified as a problem in a number of the 21
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| , events, especially in those events that were the subject of an IIT or an I
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| AIT investigation. Some examples of training deficiencies included:
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| failure of operator training to focus on normal operations and operational events; training for complex calibration or maintenance activities; and operator training on a simulator that did not sufficiently represent the actual plant.
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| As noted above, none of these 21 events resulted in significant plant damage, and few of the events presented severe challenges to plant equipment or opera-1 ting staff. However, these events (especially the operating events) illustrate events or sequences of events that, under different circumstances, could result in challenges to plant equipment or operators. Generally, the operatin were initiated because of sone random component failure or human error,g events i
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| frequently in a non-safety system. In most cases, the reactor tripped or scramedcorrectly(automaticallyormanually)followingtheinitial difficulty. However, often in the case of significant operating events, there were additional system difficulties that occurred following the trip or scram.
| |
| These difficulties resulted from a variety of reasons, but rarely from random equipment or component failure. In several cases, the systen difficulties arose from a latent defect such as a potential common cause failure of redundant equipment or an unexpected system interaction or a failure of personnel to understand the systen design and component behavior. In some events, human error, arising because of procedural deficiencies or training deficiencies, contributed to the extent of post-trip problems, or to the duration of a problem.
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| \
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| 8
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| 1 ,
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| The preliminary findings from the review of these significant events are that most events involved more than one contributor; i.e., hardware and human factors problems. The many design problems and common mode / generic failures in the events are of concern, particularly since some of the equipment deficiencies experienced in these events were not identified by previous design reviews and testing. The preliminary review of these 21 events showed that l 7
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| nost events involved a reactor trip or scram complicated by safety system l i
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| losses after the event. In some cases the safety system loss was attributable 1 to a lack of understanding of how system operating conditions could affect the I operation of a component. To reduce the occurrence of the type of significant ovents reviewed in this report, efforts should be directed to reducing scrams or trips. In addition, efforts to reduce or eliminate safety system failures l
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| should include, as an initial step, a process designed to develop an under-1 standing of the root cause of the safety system failure, and to correcting the
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| ! root cause.
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| ; 3.1.2 Common Mode Failure Mechanisms Ccntinue to be Identified through In-Depth Studies of Operational Events i Since 1980, AEOD has evaluattd operational data with a primary goal of finding i
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| and feeding back information on safety significant situations which, if left uncorrected, could result in potentially serious operating events. AE00 has found many such situations, a number of which involved construction deficien-cies, system interactions, and common mode failure mechanisms that eluded plant
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| ; safety analysis reviews. Upon discovery of such deficiencies and situations, AEOD has alerted other NRC program offices and industry groups of their exis- ,
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| t tence, potential significance, and the need for corrective action. l i
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| The potential for common mode failure represents a latent problem that may pose i a significant challenge to plant safety following a major transient or accident.
| |
| Such failures normally should be identified through detailed design reviews and related test procrans, yet a number of potential common mode failures continue i to be identified af ter plants begin commercial operation. The potential common
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| ! mode failures found and the AE00 reports
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| * which documented the findings to alert the NRC program offices and industry include:
| |
| * BWR scram system failures (case studies C001, C002, C103, and C403; engineering evaluations E007. E015, and E018)
| |
| J j
| |
| Airsystemfailures[casestudyC204andadraftcasestudyC701(see :
| |
| Section 3.2.3); engineering evaluation E123,1 l
| |
| Instrumentation and control system and electrical equipment failures (case studies C402 and C604) 1 Service water system failures (case study C202; engineering evaluations E016andE215) ,
| |
| Motor operator valve failures (case studies C203 and C603; special i study 5501) i
| |
| * Auxiliary feedwater system failures (case studies C404 and C602; engineering evaluation E325) l l M istings of AE00 reports issued from 1980 to the present are provided in Appendix B.
| |
| 9 .
| |
| i
| |
| | |
| I
| |
| )
| |
| i Diesel generator failures (engineering evaluation E110; technical review T602) l Gate valve failures (special study S402)
| |
| Agingofbatteries(engineeringevaluationE307)
| |
| ; Check valve failures affecting emergency core cooling systems (case study C502)
| |
| Failure of intermediate head emergency core cooling system pumps (special study S603)
| |
| Erosion of water and steam lines (engineering evaluation E416)
| |
| I These potential common mode failures exemplify the fact that it is necessary to carefully review operating experience and to identify and understand the exact cause of failure. Otherwise the corrective action is not likely to be l effective and the probability of recurrence may not be reduced. The discovery of root causes, however, is often difficult and can involve a wide range of 3 in-depth studies of operating events. The identification of common mode
| |
| ; failures, such as those above, continues to be important, since as discussed i in Section 3.1.1, serious operating events typically involve a combination of
| |
| ; hardware and human performance problems.
| |
| 3.1.3 Repeated Failures May Indicate Incorrect Identification of Root Cause, i or ineffective Corrective Actions AE00 case studies in 1986 addressing failures of inverters, motor operated valves, and electronic components in instrumentation and control systems identified recurring failures of this equipment at operating nuclear plants.
| |
| ; k'hile the reasons for these recurring failures cannot be determined with <
| |
| j certainty, such failures seem to reflect incorrect or incomplete root cause determinations, or a lack of effective corrective actions.
| |
| In the AE00 case study on " Operational Experience involving Losses of Electri-
| |
| ; cal Inverters," discussed in Section 3.2.5, it is concluded that the number of i inverter losses per reactor-year shows little or no improvement in each I calendar year in the 3 years (1982,1983 and 1984) included in the study.
| |
| 1 The overal lack of inprovement has occurred despite numerous NRC and industry i fe?dback communications on this problem and the recommendations for corrective l a r.: ions . The review indicates that a major contributing factor for inverter i
| |
| failures is an incompatibility between the actual plant service conditions and
| |
| } tN design service conditions (i.e., actual plant service conditions are more l severe than thise assumed in the design of inverters). The report identifies !
| |
| ; three potential failure mechanisms for inverters: (1) high ambient temperature and/or humidity within inverter enclosures; (2) electrical interconnections and physical arrangements of components which form the inverter circuitry; and (3) voltage spikes and perturbations at inverter inputs and outputs in con-junction with relatively rapid response time of solid state devices. However, effective corrective action depends on proper identification of the precise
| |
| ; underlying failure mechanism, and this appears to be lacking.
| |
| l i 10 l 1
| |
| | |
| The AEOD case study on "A Review of Motor Operated Valve Performance,"
| |
| discussed in Section 3.2.4, represented an extension and update of previous NRC studies and reports of motor-operated valve (MOV) operating experience. This study indicates that recent MOV events (from 1981 to mid-1986) involve failures that are similar to those observed earlier, and there is no apparent improve-ment in the rate of failure. Thus, the report concludes that the previous recommendations are still valid. Most of the valve inoperability was associ-ated with items such as torque switch / limit switch settings, adjustments, or failures; motor burnout; improper sizing or use of thermal overload devices; premature degradation related to inadequate protective devices; damage due to misuse (valve throttling, and hamering of valve operator); mechanical problems (loosened part or improper assembly); or the bypass circuit around the torque
| |
| ; switch not being installed or being improperly set. In many events, however, the true root causes of failures to operate were not adequately determined.
| |
| The most important conclusion from this study concerning valve assembly operability and performance / reliability is that current methods and procedures at many operating plants require improvement to assure that M0V assemblies will operate when needed (e.g., under credible accident conditions). The study also concludes that assurance of valve operability appears to be strongly dependent upon the capability to correctly diagnose, assess, and evaluate valve assembly operational failures so that root causes of failure, including erroneous switch setpoints, are correctly determined and proper changes are implemented.
| |
| The AE0D case study on the " Effects of Ambient Temperature on Electronic Components in Safety-Related Instrumentation and Control Systems," discussed in Section 3.2.6, documents the review and evaluation of four events involving failures of solid state electronic components due to overheating. One of the findings of the study was that identifying elevated room ambient temperatures or instrument cabinet temperatures as the root cause for the failure of electronic components generally has not been easy or immediate. Licensees, over an extended period, experienced several failures and many corrective actions before finally identifying overheating of components as the underlying reason for many of the failures they had experienced. It is also noted in the report that although several other operating plants have experienced failures of electronic components similar to those reviewed in the study, high ambient or internal cabinet temperature was not always pursued as a possible root cause of the failures. The report notes that overheating of electronic components raises two concerns that are generic to all operating nuclear plants that utilize heat sensitive electronic components: (1)decreasedreliabilityof electronic equipment due to increased failure rate of components, and (2) the potential for common cause failure of safety-related instrumentation channels if there is extended loss of normal cooling air flow to the cabinets in which the instruments are located.
| |
| 3.2 Results from Selected AE0D Studies In 1986 a number of AE0D studies were completed or progressed to the point where conclusions could be formed. Several of the significant reports are highlighted in this section. Discussed below are the AE0D: (1) review of new plant operating experience; (2) study on the effectiveness of licensees' and I
| |
| the NRC's operational experience feedback programs; (3) case study on air system problems and their safety implications; (4) case study on motor-operated ;
| |
| valve performance; (5) case study on operational experience involving electri- l cal inverters; (6) case study on the effects of elevated room temperature on ;
| |
| electronic components; and (7) case study on operational experience involving )
| |
| turbine overspeed trips.
| |
| l 11
| |
| | |
| .~s- -w u- . , ~ - ~ = - - - - - . - -
| |
| 3.2.1 Review of Operational Experiences at Recently Licensed (New) Plants l Among the concerns that arose from the TMI-2 accident, which occurred 13 months after licensing, was the adequacy of attention paid to operational safety by the regulators and plant operators during the initial phases of the plant's operation. The pace of ifcensing increased in the mid-1980s and, in this
| |
| ! regard, both the 1985 and 1986 Policy and Planning Guidance of the Consnission
| |
| ] set goals to "... continue to closely monitor the first 2 years of operation of new plants coming on line." This led AE00 to conduct a study of the experi-ence at new plants. Although the initial study, published in August 1986,
| |
| , identified significant differences in the experience of new plants, an analysis of the causes underlying the observed variation in event rates or the high frequencies was not perfonned. As a result, this study was expanded and continued to further identify the causes for these differences and the opera-tional problems experienced by new plants. Three plants licensed in 1986 were added to the scope of the August report, and the plant event data (through June 1986) were examined. In order to take advantage of AEOD's Trends and i Patterns Program for LER data, the event types covered in the analyses of this
| |
| ] latter report are reactor scrams (a subset of reactor protection system i actuations), engineered safety feature (ESF) actuations, violations of tech-nical specifications and losses of system safety function.
| |
| { >
| |
| { This study includes 22 plants that received initial operating licenses after January 1,1983, and reviews the reportable operational events which occurred i
| |
| within the first 24 months after OL issuance, up to a cutoff of June 30, 1986.
| |
| ; The analysis takes explicit account of the startup period and the plant status in the data for each plant. For example, Figure I shows scram counts displayed
| |
| ! by months since OL. The months of initial criticalit licensing (FP) and start of commercial operation (CO)y are (IC), full power indicated, along with !
| |
| { the reactor critical hours for each month and' the cause category for each j scram.
| |
| }
| |
| i l Figure 1. Unplanned Reactor Scrams is none i
| |
| EMta_ <
| |
| han : erm unna.E" eso 1
| |
| E se i i -
| |
| )
| |
| i %
| |
| h-a ase e a 4 lih..m .il i i e
| |
| e to sa 34 se se as as as j Months since OL lesuance i
| |
| : 12 i
| |
| i____________________.___.__.___.._.___ . _ . _ _ _
| |
| | |
| A formal statistical analysis using average unplanned event rates during both the startup progran and the first 180 days after completion of the startup program showed no statistically significant correlation of these event rates with NSSS vendor, type of AE (the utility itself or an outside firm), the length of the startup program, date of OL, or prior nuclear plant experience.
| |
| The analysis did indicate a significant correlation between the average rate during the startup program and the average rate during the first 180 days of commercial operation for scrams, ESF actuations and losses of system safety function. For violations of technical specifications, a significant correlation existed between the commercial rate and whether or not the plant was a second unit at a site (i.e., lower violation rates prevailed for subsequent units at a site).
| |
| The study of causes through a simple categorization process such as shown in Figure 1 too often masks the true nature of the events' root causes. Indeed, the information in LERs, even though subjected to a thorough review by licensees, often does not reflect underlying factors such as the overall management of operations or schedule pressures. Therefore, more detailed cause analysis was performed using startup reports, site discussions with licensees, discussions with NSSS vendors and other available documentation of early post-licensing operational experience. Site visits were conducted for the following plants: Diablo Canyon 1 and 2, Callaway, Byron 1, Wolf Creek, St. Lucie 2, Palo Verde 1 and 2 Washington Nuclear 2 (WNP-2), Susequehanna 2, and Hope Creek.
| |
| The preliminary findings and conclusions resulting from the continued AE00 study of new plants include the following:
| |
| Some plants achieve commercial operation in a smooth manner. They have relatively few unanticipated events, similar to Japanese plants, which appear to be relatively event-free in their early years. Other domestic plants have difficult early years, and some may experience over 25 unan-ticipated scrams, over 100 unplanned actuations of plant safety features, and over 50 violations of their technical specifications during their first 2 years of operation.
| |
| In general, newly licensed reactors experience a higher frequency of unplanned operational events during their early years of operation when compared to later years. This observation has historically been trans-lated into segregating the statistics for this period from general industry figures, and has given rise to the views that (1) startup experience is less important than post-commercial operation, and (2) the high frequency of unplanned operational events is an inevitable and hence acceptable part of early operation. However, this study found that frequent difficulties are not necessarily inherent in the startup process.
| |
| During this study, AE0D found that, individually, power reactor licensees already have recognized the need for action to achieve good early opera-tion. They have developed programs in the areas of training and opera-tions that cover the spectrum of actions which, if implemented 1 effectively, would result in improvement.
| |
| l This study confirmed that it is possible to achieve significant reductions 1 i
| |
| in the frequency of reportable events for new plants. The data analysis 13 1
| |
| ____ _ _ , . __ _-,.r - -.---<---,-----,----i-w- - - - - - - - -
| |
| --w- - ,^w - -- - -w- -
| |
| c- -- - - _
| |
| | |
| of the operating experience provided a pointer to focus discussions with licensees and identified a number of lessons that could be used to achieve improvements.
| |
| Both the NRC and industry have recently taken additional steps to focus on newly licensed plant operation, including the NRC senior management focus on this topic in October 1986 and the recently proposed INP0 program for improving performance during early operation. The findings of this study reinforce the need for and benefit of these efforts.
| |
| For a new plant, an aggressive root cause determination and corrective i action program for responding to failures is a key ingredient to improve- !
| |
| ment. The analyses of operational data for scrams, engineered safety feature (ESF) actuations and loss of system safety function clearly demonstrate the need to correct the root causes for reportable events early in life. Without correction, these root causes will likely exist during the early commercial operation of the facility. At this time of life, the relatively high challenge frequency coupled with the potential of undetected systems problems may present a significant challenge to a new operating crew. Therefore, the analyses indicate that increased attention to operations is appropriate during the period of early commer-cial operation. The improvement measures identified in the report repre-sent an effective action list to consider in formulating a program to prevent high reportable event frequencies.
| |
| Nuclear safety measures consist of prevention and mitigation items. This study identified measures to reduce the frequency of reportable events, which are prevention initiatives. Mitigation measures such as operator training for response to plant transients are also necessary, since the two sets of measures do not overlap. However, it is necessary to carefully assess event reduction measures such as revisions of setpoints to avoid a net decrease in plant safety.
| |
| 3.2.2 Operational Experience Feedback Programs Subsequent to the TMI-2 accident, various studies highlighted a need for increased attention to the feedback of operating experience (0E) by the NRC, the industry, and each licensee in order to use the lessons of experience to prevent serious nuclear incidents from occurring. One action taken b was to require that each nuclear power plant licensee have a formal (yi.e.,
| |
| the NRC controlled by procedures) program to incorporate the lessons learned from operating experience into plant hardware and procedures, and to assure that these lessons are communicated to plant operators and other personnel.
| |
| In fall 1984, AE00 initiated a survey to assess the usefulness of operational data feedback documents and to determine the characteristics of licensees' operating feedback programs. AE0D staff made on-site visits to seven licensees, met with INPO to discuss their programs, observed an INP0 plant evaluation in the operational experience area, and in May 1986 issued AE0D report S602, "An Overview of Nuclear Power Plant Operating Experience Feedback Programs."
| |
| In general, the study found that, over the last 5 years, many significant and worthwhile initiatives have been implemented to understand the lessons of 14
| |
| | |
| cxperience. As a result, there has been increased attention paid to OE feed-back. However, many licensee programs have not yet achieved effective per-formance levels in all areas. Most plants are making moderate, not extensive, use of their in-house operating experience, and in general are making less use of the large body of knowledge associated with events and concerns that originate elsewhere in the industry. Thus, many licensee OE review activities still do not provide high confidence that all of the important lessons of operating experience are being thoroughly explored and used. To increase the effectiveness of OE reviews in order to correct past and potential operational problems (i.e., assuring safety of operations), licensees and the industry, or the NRC need to take further actions.
| |
| Based on this study, AE0D recommended that IE consider infonning licensees of the conclusions drawn from this study and suggesting that each licensee review their OE activities for effectiveness and conformance with the intent and scope of the existing NRC requirements. No major regulatory actions were recommended because AE00 was aware that industry initiatives in this area were still ongoing. AEOD recommended that the effectiveness of OE activities continue to be monitored by staff through operating experience reviews and observations of OE review activities. If, after sufficient time (2 years or so), the effectiveness of OE review activities has not improved, further regulatory actions should be considered. In this regard, AE00 recommended as a prudent measure that NRC initiate the development of better guidance for licensees' OE activities, and an inspection module based on that guidance. In addition, AE0D recommended that the staff consolidate some of the existing systems for com-municating operating experience into a single "NRC Notice" system. This system would thus be the sole NRC system for the feedback of operating experience to licensees. Staff responses and the status of actions related to those recom-mendations are contained in Section 6.0.
| |
| 3.2.3 Air Systems Problems at U.S. LWRs A preliminary AE00 case study report on " Air Systems Problems at U.S. Light Water Reactors" was issued for peer review on December 11, 1986. The study
| |
| ; analyzes and evaluates the operational experience related to, and the safety implications associated with, failures and degradations of air systems at U.S.
| |
| LWRs.
| |
| The report presents aspects of air system degradations and plant responses to air system losses which are not addressed in previous studies. It also highlights more than two dozen events in which, contrary to licensing assumptions, a safety-related system failed due to an air system degradation or I
| |
| failure. Operating events involving the loss or degradation of air systems l
| |
| were judged to be safety significant because they may lead, under different circumstances, to potentially serious events and conditions which have not been analyzed in a plant's Final Safety Analysis Report (FSAR). Some of the systems which were significantly degraded or failed were decay heat removal, auxiliary feedwater, BWR scram, main steam isolation, salt water cooling, emergency diesel generator, containment isolation, and the fuel pool seal system.
| |
| The root causes of most of those failures were traceable to design and/or management deficiencies. The design and operating problems found appear to reflect a lack of sufficient regulatory requirements and review, and the view by many applicants and licensees that air systems are not highly important to 15
| |
| | |
| assuring plant safety. The report addresses specific deficiencies which were found in the following areas: (1) mismatched equipment - the air quality capability of the instrument air system filters and dryers does not always match the design requirements of the equipment using the air; (2) maintenance of instrument air systems is not always performed in accordance with manufacturer's recommendations; (3) air quality is not usually monitored I periodically; (4) plant personnel frequently do not understand the potential I consequences of degraded air systems; (5) operators are not well trained to respond to losses of instrument air, and the emergency operating procedures for such events are frequently inadequate; (6) at many plants the response of key equipment to a loss of instrument air has not been verified to be consistent with the FSAR; (7) safety-related backup accumulators do not necessarily undergo surveillance testing or monitoring to confirm their readiness; and (8) the size and the seismic capability of safety-related backup accumulators at several plants have been found to be inadequate.
| |
| This final case study was formally issued in March 1987.
| |
| 3.2.4 A Review of Motor-Operated Valve Performance AE0D issued case study C603 in response to Action Item 6(1) of the actions directed by the ED0 to respond to the NRC staff investigation of the June 9, 1985 event at Davis-Besse. The report brings together previous NRC studies and other related documents (i.e., AEOD reports and IE bulletins, circulars, and information notices), reviews approximately 1200 valve operator events (found in the SCSS and NPRDS data bases; see Sections 5.2.2 and 5.2.7) from 1981 to mid-1985, and incorporates new data from an NRC valve testing program that utilized signature tracing techniques on valve assemblies in operating plants.
| |
| The report concludes that current methods and procedures at many operating plants require improvement to assure that valves will operate when needed.
| |
| Furthermore, the deficiencies would generally not be detected by existing plant procedures intended to assure operability, such as surveillance testing (plant technical specifications and ASME Code, Section XI inservice testing) or plant operator observations. Thus, assurance of valve operability appears to be strongly dependent upon the diagnostic capability to assess and evaluate fail-ures to operate so that root causes of failure--including switch setpoint--dre correctly determined and proper changes implemented. Further, the issue of valve assembly performance and reliability is a complex sub.iect that involves several technical disciplines.
| |
| AE00 currently is working with industry to develop and implement improved guidance, procedures, and/or equipment to address all aspects of safety-related motor-operated valve assembly operability. The overall goal for this industry effort is expected to involve the development of uniform guidance, procedures, and/or equipment which licensee management could adopt with confidence, and which field personnel could implement consistently and effectively.
| |
| 3.2.5 Operational Experience Involving Losses of Electrical Inverters AE0D case study C605 was initiated primarily because of the observed lack of overall improvement in the operational performance of safety- and nonsafety-related electrical inverters. Operating experience has shown that the loss of 16
| |
| | |
| an inverter is a frequent occurrence, with a total of 152 events involving )
| |
| inverter losses reported between 1981 and 1985.
| |
| l l Occurrences of events involving inverter losses raises two principal concerns:
| |
| 1 (1) increased frequency of plant transients, and (2) inoperable or improper functioning of both safety-related and other important plant equipment. Review and analysis of related operational data indicate that inverter loss events result from component failures and/or personnel actions. Further review indi-l cates that a major contributing factor and/or cause for the occurrence of component failure events is an incompatibility between the actual inplant i service conditions for inverters and their design service conditions. Approxi-mately 50% of the events attributed to component failures were due in part or whole to high temperature and/or humidity within inverter enclosures and/or voltage spikes and perturbations at inputs and outputs of inverters. Con-tributing factors for events involving personnel actions included inadequate maintenance and testing procedures and deficient practices. In addition, inadequate plannino, training, and verification for related maintenance and testing activities were contributing factors.
| |
| Overall, no significant decrease in the frequency of inverter losses has been detected on an industry-wide basis. Although the plants involved in these '.
| |
| events have recovered without any serious consequences occurring, the frequency '
| |
| and potential consequences of these events continue to be of concern.
| |
| 3.2.6 Effects of Ambient Temperature on Electronic Components in Safety-Related Instrumentation and Control Systems In reviewing operating experience from 1982 to 1986, AEOD case study C604 identified four events at four separate nuclear plants involving failures of solid state electronic components in safety-related instrumentation and control (18C) systems due specifically to overheating.
| |
| Overheating1)(ofdecreased concerns: electronicreliability components in safety-related of electronic 18Cdue equipment systems raises two to increased failure rate of printed circuit cards and other heat sensitive electronic components, and (2) the potential for common cause failure of redundant safety-related instrument channels due to extended loss of normal cooling air flow to the cabinets in which the instruments are located. These concerns are generic to all operating nuclear plants that use solid state electronic conponents. Failures of instrument system components due to overheating have caused malfunctions of control systems, inoperability, and spurious actuation of protection and engineered safety features actuation systems (ESFAS) channels, inadvertent actuation and failure of an ESFAS train, and erroneous indications and alarms in the control room.
| |
| The review of the four events found that, in general, licensees did not readily identify elevated room ambient temperature or cabinet internal temperature as the root cause for the failure of the electronic components. The plant licensees, over an extended period, experienced several failures and many corrective actions before finally identifying overheating of components as the underlying reason for many of the failures they had experienced. A limited review of operating experience at other plants that have experienced electronic component failures of a similar type found that, in general, elevated room ambient or cabinet internal temperature was not identified as a possible root 17
| |
| | |
| i j' cause of the failures. The study also found that plant technical specifications regarding area ventilation cooling systems and instrumentation systems have not adequately considered actual temperatures in the instrument cabinets.
| |
| ! In addition, a review of the staff's proposed resolution of USI A-44 regarding
| |
| ; design. adequacy and capability of instrumentation and control system equipment i needed to function in environmental conditions associated with a station black-
| |
| ! out found that plant specific evaluations are needed with regard to the actual temperature and condition of heat sensitive components inside the instrument cabinets. l 1
| |
| 3.2.7 Operational Experience Involving Turbine Overspeed Trips <
| |
| AEOD case study C602 was performed in response to the action item 8(f) of the action directed by the EDO to respond to the NRC staff investigation of the
| |
| , June 9, 1985 event at Davis-Besse. In this event, after main feedwater was
| |
| ; lost, both auxiliary feedwater (AFW) turbines of the plant coincidently tripped on overspeed, and resulted in a total loss of feedwater flow to the steam generators. The study evaluated past operating experience involving overspeed trips of PWR AFW turbine-driven pumps. The overspeed trips of AFW turbines
| |
| , were found to be. somewhat wide-spread and one of the major causes for-loss of '
| |
| operability or availability of AFW systems. A total loss of the AFW system:
| |
| when needed will significantly increase the probability of core damage.
| |
| In addition, the study was extended to include a review of turbine overspeed ,
| |
| trip operating experience on BWR high pressure coolant injection (HPCI) and.
| |
| reactor core isolation cooling (RCIC) turbine-driven pumps. Almost all of the turbines which are used in AFW, RCIC, and HPCI systems at operating plants are
| |
| , manufactured by Terry Company and are equipped with Woodward governors. The service conditions for these three systems are quite similar and, as such, review of HPCI and RCIC events aided in identifying causes for overspeed trips a on AFW turbines. A total of 128 events involving overspeed trips of steam i turbines associated with AFW, HPCI, and RCIC systems were reviewed. These l events occurred between January 1972 and September 1985.
| |
| l Review of these events indicates that the dominant attributed causes of AFW i
| |
| turbine overspeed trips are speed control problems associated with governors, l and trip and reset problems associated with trip valves and overspeed trip mechanisms. These problems are primarily the result of inadequate performance by plant personnel, inadequate procedures, and insufficient system design i considerations. The governor speed control problems are: .(1) slow response of~
| |
| ! the Woodward Model PG-PL governor, (2) entrapped oil in the speed setting cylinder of the Woodward Model PG-PL governor, (3) incorrect governor. setting,
| |
| , and (4) water induction into the turbine. The trip and reset problems stem j from the complexity of reset operations and a lack of trip position indication.
| |
| 3.3 Abnormal Occurrences Involving U.S. Nuclear Power Plants
| |
| }
| |
| Each calendar quarter, AE0D prepares and coordinates a Report to-Congress on
| |
| ! Abnormal Occurrences _(A0s). A0s may be individual incidents, recurring events, generic concerns, or a series of incidents which the Commission determines are significant from the standpoint of public health or safety. The criteria for selecting A0s have not changed since publication in February 1977. Thus, the number of A0s per year can be viewed as a rough, yet reasonably constant index
| |
| * f 1
| |
| . 18
| |
| | |
| to the performance of the nuclear power industry with regard to potentially strious or significant occurrences.
| |
| The number of A0s reported for nuclear power plants in the Reports to Congress for each calendar year since 1977 is shown in Figure 2. The number of A0s for 1986 is a preliminary estimate, since the last two quarterly reports have not yst received final staff concurrence and/or Commission approval. The history of A0s on a per plant basis is shown in Figure 3. An increasing trend in the number of A0s is noted starting in 1982, with a slight drop noted for 1985.
| |
| , One reason for this increase is related to the increasing number of nuclear
| |
| ; power plants. A review of Figure 3 shows that the number of A0s per plant has l remained relatively level for the past 4 years.
| |
| A significant fraction of A0s has been associated with plants licensed less than 2 years. For example, of the 49 A0s reported from 1981 through 1986, about 29% (14 events) occurred at plants licensed less than 2 years at the time of the A0.
| |
| A summary of 1986 abnormal occurrences, including those still under con-sideration, is provided in Appendix A. This summary includes power reactor, nonreactor, and medical misadministration abnormal occurrences. The two latter categories are discussed in Sections 4.1.1 and 4.2.4.
| |
| On November 18, 1986, AE0D coordinated an A0 Reporting Meeting at headquarters (Bethesda), attended by representatives of most Headquarters and Regional Offices. These representatives were those generally involved in the A0 reporting program (e.g., the Office A0 Coordinators). The purposes of the meeting were to (1) assure consistency of reporting, (2) improve timeliness and quality of reporting, (3) reduce personnel resources, and (4) resolve questions concerning the program. The need for such a meeting was indicated by (1) the large turnover in A0 Coordinators during the life of the A0 program so that some coordinators may not be ' completely familiar with the policies and procedures, (2) staff questions regarding the threshold for reporting certain events, (3) uneven quality (content) of event writeups submitted to AEOD, and (4) the generally extensive time expended in identifying A0s and processing them.
| |
| The meeting was coordinated by members of AE0D who have been with the A0 program since its inception in 1975; these individuals were the principal NRC staff coordinators for the development of the A0 criteria, NRC A0 Policy Statement, and the staff guidelines and procedure for the A0 reporting process.
| |
| From the questions and comments received, it was evident that the meeting was of mutual benefit to all of the participants.
| |
| Conclusions In summary, the number of significant items of this magnitude remains relative-ly constant.
| |
| I 19
| |
| | |
| U. S. NUCLEAk7dWER PLANTS ABNORMAL OCCURRENCES VS. YEAR
| |
| .o so-s... .
| |
| no.
| |
| s..
| |
| 4 nlml l[d 77 78 79 80 81 52 83 84 lE 85 86 U. S. NUCLEA YbWER PLANTS ABNORMAL OCCURRENCES / PLANT VS. YEAR
| |
| *** "" LEGEND
| |
| .t s - too N AOs/ PLANT
| |
| ,,,, .,o # PLANTS
| |
| .5 4 - **o 3 .... .._,
| |
| g .... ...,
| |
| : g. .
| |
| l l[ !!
| |
| 77 7s 79 so a s2 84 as as 20
| |
| | |
| 3.4 Analysis of the 1986 Licensee Event Reporting On January 1, 1984, the LER rule (10 CFR 50.73) became effective. As a result, the NRC and the industry have a description by the licensee, in a reasonably complete and detailed manner, of: all actuations of Engineered Safety Features (ESFs), including scrams; all losses of safety function at a system level; all significant systems interactions; all technical specification shutdowns /
| |
| violations; and all significant internal and external threats to plant safety.
| |
| The number of LERs submitted by year is shown in Table 3.
| |
| Table 3 LERs Submitted By Year Year LERs Units LERs Per Unit 1981 4016 75 53 1982 4400 81 54 1983 4839 84 57 1984 2435 92 (1) 26 1985 2997 97 (2) 31 1986 2818 104 (3) 27 (1) Palo Verde not included - licensed 12/31/84.
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| (2) Dresden 1, Humboldt Bay, and Three Mile Island 2 are not included in the 1985 data.
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| ' (3) Dresden 1 Humboldt Bay, Three Mile Island 2, Seabrook 1 and Braidwood I are not included in the 1986 data.
| |
| The following discussion provides an overview of the LERs submitted in 1986. As noted on a number of occasions, we attach no safety significance to the raw number of LERs per se. The safety significance is assessed through engineering review of each LER, and through trends and patterns analysis of the content of the LERs. Variations in LER counts from plant to plant can result from a host of factors, only one of which is an actual difference in safety performance.
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| Thus, the reporting pattern is examined with attention to both high and low reporters,. in order to gauge the reporting process itself.
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| Analysis of the 2818 LERs submitted in 1986 indicates a broad range in the number of LERs submitted by each licensee (from three LERs to 93 LERs in this 1-year period). The mean number of LERs per unit is 27, while the median is
| |
| : 24. The distribution of LERs per unit is shown in Figure 4.
| |
| In terms of what was reported, the largest number of reports (44%) was asso-ciated with scrams and ESF actuations. The number is roughly split between ESF actuations (other than scrams), and scrams. A scram was reported in about 20%
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| of the 1986 LERs, compared to only 1% of the 1981-1983 LERs (note: scrams were not specifically reportable prior to 1984).
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| 1 21
| |
| | |
| The second most frequently reported type of event (31%) was a condition pro-hibited by technical specifications (TSs) or a shutdown required by the TS.
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| This category covers conditions ranging from a missed surveillance test to the completion of a plant shutdown because of the unavailability of required safety components.
| |
| Figure 4 LERs SUBMITTED IN 1986 m
| |
| m.
| |
| m-Ib is- -
| |
| I ..t
| |
| ~
| |
| . i
| |
| :-e lllllll1.11 si-i i- i-m i-4e t-ee NUMBER OF LERs i-ee ri-m t-es i-se .
| |
| i The third most frequent LER category concerned events that did or could have resulted in a loss of a safety function at the system level. This condition l was reported in 7% of the LERs. The percent of LERs by the individual LER report requirement is shown in Table 4.
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| 22 l
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| | |
| Table 4 LER Reporting by Report Requirement Reference Requirement Percent
| |
| * 50.73(a)(2)(iv) RPS/ESF Actuation 44 50.73(a)(2)(1) TS Shutdown or TS Violation 31 50.73(a)(2)(v) Real or Potential Loss of a Safety System 7 50.73(a)(2)(ii) Unanalyzed Conditions 4 50.73(a)(2)(vii) Failures in Multiple Systems 3 50.73(a)(2)(iii) External Threat 1 Other associated reporting requirements 11 (e.g.,Part 21, 50.36, 73.71, voluntary)
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| * Percents sum :o greater than 100% because an LER can be reported under more than one reporting requirement.
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| When the LERs are assessed in terms of other classifications, such as NSSS vendor or architect-engineer, typically a wide variation results. Specific d2 tails are provided in Table 5.
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| Table 5 Average Number of LERs by Major NSSS Vendors in 1986 Number of Average Number of LERs NSSS Vendor Units Per Plant 88W 8 18 CE 14 22 W 43 25 GE 37 34 The analysis of 1986 LERs is continuing. A number of major studies are in progress or planned in order to better characterize and understand the nature and significance of these operational events. The preliminary results of some of these studies (e.g., scrams and ESF actuations) are reported in Sections 3.6 and 3.7.
| |
| 3.5 Assessment of Licensee Event Report Quality Th2 Licensee Event Report (LER) is one of the most important and widely analyzed documents in the nuclear industry. It is the principal means of identifying and analyzing safety problems and concerns which may not be recognized or properly understood as to potential significance. In order to 23
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| | |
| evaluate the overall quality of the contents of the LERs, a representative sample of a licensee's LERs is evaluated against the reporting requirements contained in 10 CFR 50.73. Each sample consists of:
| |
| LERS with event dates during the assessment period.
| |
| From the LERs submitted during the period, a maximum of 15 (selected randomly) are evaluated. If the licensee has submitted fewer than 15 LERs for the period, all of the LERs for the period are assessed.
| |
| Beginning in early 1986, for a multi-unit site where the LERs for the site are prepared by a single organizational entity, a single sample that meets the criteria described above is generally used for the entire site evaluation.
| |
| Prior to 1986, a separate report was prepared for each unit at a multi-unit site.
| |
| The assessment of LER quality is performed by an AE0D contractor, Idaho National Engineering Laboratory (INEL) at Idaho Falls, Idaho. The evaluations performed by the contractor consist of a detailed review of each selected LER to evaluate the content of its text, abstract, and coded fields. The evalua-tion process for each LER is divided into two parts. The first part of the evaluation consists of commenting on the content of each LER. The second part consists of determining a score (0-10 points) for the text, abstract, and coded fields of each LER.
| |
| The LER-specific comments serve to point out what the analysts considered to be the specific deficiencies or observations concerning the information pertaining to the event, and also provide a basis for evaluating the general deficiencies for the overall sample of LERs that was reviewed.
| |
| A separate report is prepared for each licensee which includes a detailed dis-cussion of the strenaths and weaknesses of each LER. AEOD asks each Region to forward these reports to the licensee to assist in the preparation of future LERs. In addition, a brief summary statement of the results of the assessment is presented for each licensee. Graphic presentations of the distribution of the overall grades for the licensees assessed as of the end ~of December for-1985 and 1986 are shown in Figures 5 and 6, respectively.
| |
| It is noted that there has been a favorable shift in the number of units /
| |
| stations receiving higher overall average scores during 1986. A significant portion of this increase is attributed to licensees which have been evaluated more than once. For example, the plant scoring 6.6 in 1985 scored above the industry average when reevaluated during 1986. (The plant scoring 6.6 in 1986 had not been evaluated previously.) Almost without exception during 1986, the ovefallaveragescoreshowedimprovementforeachlicenseewhichhadbeenpre-viously evaluated. Neverthe-less, there is still considerable room for improve-meet for the majority of licensees to provide LER reports of consist-ently high quality. Future con-sideration will be given to varying the frequency of review based on the results of the evaluations.
| |
| During the December 1986 Region II Resident Inspectors Meeting, a presentation was given of the methods used to evaluate the quality of LERs, and of the l results obtained for each licensee. Similar presentations are planned for l other Regions.
| |
| 24
| |
| | |
| Figure 5. DlStribution Of OVerall average LER sc' ores as of December 1985 11 10 - (Evaluations were performed E g. on an individual nuclear .
| |
| ,o unit basis) 5 8- -
| |
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| |
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| |
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| |
| 1 9.5 9.0 8.5 8.0 7.5 7.0 6.5 6.0 Overall average scores Figure 6 Distribution of overall average LER scores as of December 1986 11 .... .... ....
| |
| 10 - 7 -
| |
| f (Evaluations were based on E g. / a multi-unit basis, if o # .
| |
| / LERs were prepared by the 8- / sare licensee group) -
| |
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| 9.5 9.0 8.5 8.0 7.5 7.0 6.5 6.0 Overall average scores ;
| |
| l l
| |
| 25 .
| |
| | |
| Conclusion There was general improvement in the quality of LERs during the year. However, there still is a rather wide divergence in the quality of LERs submitted by the licensees. In addition, there is still considerable room for improvement for many licensees.
| |
| 3.6 Reactor Scram Experience In 1984, 1985, and 1986 This section analyzes unplanned reactor trips (i.e., scrams) which occurred at U.S. light water power reactors in 1984, 1985, and 1986. Data on reactor scrams were extracted from LERs submitted by licensees in conformance with 10 CFR 50.73. The results from completed studies of 1984 and 1985 scram data (AE0D/P504 and AE0D/P602, respectively) and preliminary results for a similar study for 1986 experience, are summarized below.
| |
| There are generally three phases to a scenario or sequence of events involving a reactor scram. First, there is the generation of some off-normal plant state which results in operation of the reactor protection system (RPS) or the need for a manual scram. Second, there is the operation of the RPS and control rod drive system. Third, there is the plant and operator response to the scram.
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| Each phase has safety significance. For example, the NRC has concluded that a reduction in the frequency of challenges to plant safety systems should be a prime goal of each licensee, and that large Anticipated Transient Without Scram risk reductions can be achieved by reducing the frequency of transients which call for the RPS to operate.
| |
| 3.6.1 1986 Scram Experience A reactor scram is defined as an actuation of the RPS, whether automatic or manual, that results in control rod motion. Plants were included in these statistics if they: (1) held a full power operating license, and (2) accumu-lated critical hours for some portion of the calendar year in question.
| |
| Reactor years were calculated for portions of the calendar year where necessary, based on the date of initial criticality.
| |
| Based on preliminary data assessment, in 1986 there was a total of 469 unplanned scrams at 93 U.S. LWRs which were licensed to operate at above 5%
| |
| power and which had accumulated some critical hours. The corresponding figures for 1985 were 552 scrams at 92 LWRs.
| |
| Of the 469 scrams in 1986, a total of 52 (11%) were manual. The number of total scra'ns is down in 1986 as compared to 1985; however, the industry scram rate, while reduced, remained at nearly one scram per thousand hours of criti-cal operation.
| |
| In 1986, the industry average rate (number per plant per year) for automatic trips showed its first significant reduction. The comparative statistics for 1983 through 1986 are as follows:
| |
| 26
| |
| | |
| t r
| |
| Scram Type Average Number Per Plant Per Year l 1983* 1984 1985 1986 i
| |
| Manual 0.9 0.7 0.6 0.6 Automatic 5.6 5.2 5.4 4.5
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| . Total U D D U l *From the NRC (NRR) study, "1983 Reactor Trip Statistics," dated August 1984.
| |
| l j 3.6.2 Reactor Scram Frequency l Reactor scram rates for unplanned reactor scrams occurring in 1984, 1985, and 1986 are displayed for the ma.ior NSSS vendors in Table 6. Overall, there was a significant reduction in the overall PWR aggregate rate. There was a decrease i in the total number of reactor scrams from 1985 to 1986 (reduction of 83 i scrams), while the total critical hours and number of plants remained almost i constant.
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| Totte 6
| |
| ] Average Reecter Seren Frequency Date 1984 1906 1906 eYPer Per Per asector 1000 Critical asector 1000 Critical Reector 1000 Critteel j Teer Moors Year Moors Year Moors a
| |
| W 7.1 1.22 6.8 1.04 5.5 0.04 CE 5.9 0.86 7.5 1.24 6.2 0.91 3/W 3.0 0.44 5.0 0.88 2.4 0.46 l Total 6.3 1.04 6.7 1.06 5.3 0.82 OWR GE 5.5 1.12 5.3 0.M 4.4 0.00 I
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| i '
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| ; The decrease in the Westinghouse (W) average scram rate from 1.04 to 0.84 i
| |
| scrams per 1000 critical hours is reflective of a general decrease of the
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| ! individual ~ plant scram rates for plants operating in both years, as well as i relatively low scram rates for units which achieved initial criticality in i 1986. The two Westinghouse units achieving initial criticality in 1986 had a j trip rate below two trips per 1000 critical hours. Overall there were 37
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| ; Westinghouse-designed plants with data for 1985 and 1986; 19 showed a decrease 1 in the number of scrams, 15 showed an increase, and three registered no change.
| |
| i The major cause for the reduction at CE-designed plants was the major reduction in the number of scrams at Waterford 3, a net reduction of 21 scrams. Seven other CE units also experienced a reduction in the overall scram rate.
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| 1 27
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| !__,.. _ . _ _ . . _ _ _ . _ _ . _ _ _ . . . _ _ - . _ . , , _ _ _ _ _ _ _ _ _ _ _ , , _ . . _ ~ _ _ _ . _ _ . _ _ _ - - _ _ - _ . _ _ _ . _ _ _ _ . . _ _ _
| |
| | |
| The Babcock & Wilcox (B&W) average is based on the smallest number of plants (eight in 1985 and seven in 1986), and hence is most sensitive to individual plant behavior. In 1986, four of the seven plants experienced a reduction in their overall trip rate.
| |
| Finally, the General Electric (GE) BWR average shows a significant decrease from 0.94 to 0.80 scrams per 1000 critical hours. The decrease reflects a drop in the scram rate at 15 of the 28 GE-designed plants that operated in both 1985 and 1986.
| |
| In 1984,1985, and 1986, the majority of reactor scrams occurred with the reactor power above 15% (i.e., 68% in 1984, 74% in 1985, 76% in 1986). The 1986 data is consistent with the 1985 results, which we believe reflect the short time spent at lower power. Over the past 3 years, 31%, 38%, and 38% of the total scrams in 1984,1985, and 1986, respectively, occurred while the plant was at 95% power or above. Because the overwhelming majority of scrams occur at power levels greater than 15%, and because of the inherent greater decay heat removal requirements at higher power levels, our analysis focused on the power level greater than 15% power.
| |
| In 1984, a scram frequency of 2.0 scrams (above 15% power per 1000 critical hours) was selected as a breakpoint for examining relatively high scram rates.
| |
| Figure 7 provides a plot of each plant's reactor scram rate for scrams above 15% power. The axes are (X) scrams per 1000 critical hours and (Y) the number of critical hours. In addition, horizontal and vertical reference lines bound plants into quadrants. The horizontal reference line represents a plant availability of approximately 70%. Plants above the 70% availability line and to the left of the vertical two scrams per 1000 critical hours reference line have high availability and low trip rates. Plants bound in other quadrants represent those plants with high availability /high scram rate, low availability / low scram rate, and low availability /high scram rate. Table 7 lists plants with high availability and low scram rates. Table 8 is provided to list those plants that are currently licensed but had no critical hours in 1986. In 1986, there was a modest increase in the average number of critical hours; i.e., 5878 critical hours in 1985 to 6016 critical hours in 1986.
| |
| Table 9 is provided to show plant by plant scrar rates for 1986. The criteria for relatively high trip experience was met in 1984 by ten plants, in 1985 by eight, and in 1986 four plants met this criteria. The maximum rate in 1986 was 3.0 scrams per 1000 critical hours versus the 5.7 scrams per 1000 critical hours in 1984 and 4.8 in 1985. Only one of the four plants that exceeded the cutoff achieved initial criticality in 1986 (i.e., achieved criticality 4/18/86).
| |
| In conclusion, the scran frequency is noticeably improved (lower) for 1986 as compared to 1985. The number of plants exhibiting relatively high scran rates (based on their scram rate above 15% power exceeding 2.0 scrams per 1000 critical hours) has decreased significantly since 1984 3.6.3 Initiating Systems (Above 15% Power)
| |
| As in previous years, each scram has been examined to determine the system containing the root cause of the scram. That system is defined as the
| |
| " initiating system." The scran that occurred may have been the result of 28
| |
| | |
| Figure 7 ,
| |
| 1986 REACTOR SCRAM RATES VS CRITICAL HOURS BY PLANT POWER GREATER THAN 155
| |
| +::$ %
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| i $ i 5 5 5 5 5 5 5 E $ 5 PLANT TRIP RATE PER 1000.0 CRITICAL HLS 1985 REACTOR SCRAM RATES VS CRITICAL HOURS BY PLANT i
| |
| POWER GREATER THAN 15X i
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| O e i I 2 2 3 3 4 4 5 5 0 E E i $ i $ E $ E E E E E PLANT TRIP RATE PER tees.e CRITICAL HRS 4
| |
| 29
| |
| | |
| . Table 7 PLMi$ uliH NIGH CRITICAL HOWS RATE LEl$ THM 2 IR!PS PER 1000 HRS YEAR 1996 j NAE CRITMS RATE ARKANSAE2 6370.0 0.63
| |
| ;i MAVERVALLEYl 6243.I 0.32 l Bil ROCK POINT 8387.3 0.12 IRUNSN!CK 1 B317.7 0.60 BYRON 1 7820.9 0.51 i CALLWAY l 7307.6 0.32 CALVERTCLIFFl! 6906.2 0.50 j CALVERTCLIFFS2 1442.0 0.59 COOK 1 e5M.4 0.53 COOPER 6570.1 0.30 ,
| |
| l!ABLOCANTON2 7296.3 1.37 IRE 5KN 2 7110.1 0.42
| |
| ;: BUANEARNOLI 7348.2 0.00 FARLD 1 7276.4 0.55
| |
| ; FARLEY2 7549.7 0.40 Fli! PATRICK 0075.5 0.25 FT. CALHOUN 1 8485.2 0.24 ElmA U16.3 0.52 l HATCH 2 6451.9 0.77 i INDIM POINT 3 6501.6 1.22 i KENAtalEE 7584.3 0.26 4
| |
| LASALLE 2 6614.0 0.45 LinERICK1 7146.1 0.14 NAINE YAMEE M91.1 0.D 1 MILLSiont ! 8276.5 0.24 RILLS 10hE 2 6599.6 0.61 MILLSichE 3 6201.1 0.11
| |
| ) NONilCELLO 6984.9 0.29 NORTH MNA 1 7560.0 0.H NORTH MNA 2 7301.3 0.41 OCONEE 2 7253.7 0.M i OCONEE3 M35.4 0.26
| |
| ; PEACH ICTIOR 2 7272.8 0.M
| |
| ; POINT MACH I 7905.4 0.25 POINT M ACH 2 7262.7 0.14
| |
| ; PRAIRIE ISLA W I M90.4 0.13
| |
| } PRAIRIE 15 LAW 2 7972.1 0.30 1 R00!NSON2 7118.9 0.98 4
| |
| SALER1 7097.2 1.13 lAN ONOF E 2 6400.0 1.00 SAN ONOF E 3 7422.3 0.54 ST. LK!E I M24.0 0.36 l St.LKit2 7326.7 0.41 SUMER1 M53.2 0.M SURRYl 6233.2 0.H TRI-l 6268.6 0.40 TROJM 7M4.1 0.57 ,
| |
| TWRED PolNT 3 M00.4 0.06 j NATERFOR03 7011.6 0.06 '
| |
| i WP-2 6391.5 0.70
| |
| ! NILF CRE R i 6523.7 0.N j TAKE RONE 8343.5 0.36 j
| |
| Ilm 2 D83.5 0.51 ;
| |
| ; 30 i
| |
| | |
| t l
| |
| Table 8 LICDISED REACTORS NITH ZERO CRITICAL IEMil YEM 19N
| |
| -_..~ VDIDOR=MK0CK WILCOI -- - -
| |
| 005 llME 1 RMCHO SECO 2 THI-2e
| |
| .--__. VENDOR 4ElilllSHOUSE -- ---
| |
| ! 0S mE i
| |
| 3 SEGUOYM 1 4 SEDUOYM 2 5 HARRIS !
| |
| < 6 MMN 2 7 MAIDWC00 1 8 SEAM 00r
| |
| -l
| |
| __._. VENDOR 4ENERAL ELECTRIC ---- -
| |
| OBS IIME 9 BRESODI le 10 IUl90LDT MYe
| |
| !! IR0 lull FERRY l 12 BROWS FE MY 2 13 3ROINil FEMY 3
| |
| ; 14 NINE N!LE POINT 2 15 CLINTON 1 i
| |
| * Plants are permanently shut down.
| |
| i l
| |
| 31
| |
| | |
| Table 9 REACTORTRIPRATES YEAR 1986 NAME NAIR)E AUTO LESS TNAll GREATER CRlilCAL TRIP MTE PER KM TIE MATIC OREIUAL THAN HOURS 10 M NOURS KTEENTRIPS 151PONER 151PONER PONERli15 POERli151 ARKMSAS 1 0 4 2 2 5536.7 0.36 2761.4 ARKANSAS 2 0 5 1 4 6370.0 0.63 1592.5 KAVER VALLEY l 1 3 2 2 6243.I 0.32 3121.9 316ROCKPOIN7 0 1 0 1 8307.3 0.12 8387.3 BRUNSN!CK 1 1 & 2 5 1317.7 0.H IM3.5 BRUNSNICK2 0 2 0 2 4232.5 0.47 2116.3
| |
| , BYRON 1 2 4 2 4 7820.9 0.51 1955.2 CELANY l 1 6 1 6 7307.6 0.02 1217.9 CEVERT CLIFFS 1 0 4 0 4 69M.2 0.50 1726.6 CEVERTCLITTS2 3 3 1 5 I442.0 0.59 1600.4 CATMBA i 0 9 2 7 5425.2 1.29 US.0 CAIMBA 2 2 5 2 5 2678.4 1.87 535.7 COOK 1 0 5 1 4 7536.4 0.53 III4.1 COOK 2 0 3 1 2 5560.5 0.36 2700.3 COOPER 0 2 0 2 6570.1 0.30 3205.0 CRYST E RIVER 3 0 1 0 1 3691.1 0.27 3691.1 OAVIS DESSE 1 0 1 1 0 178.0 0.M .
| |
| DIA K O CANYON I e 3 1 2 5967.4 0.34 2983.7 514K 0 CANYON 2 1 10 1 10 7296.3 1.37 729.6 BRESDEN2 2 4 3 3 7110.1 0.42 2370.0 DRESDEN 3 1 5 0 6 27H.4 2.17 41.1 DUAE AMOLD 1 1 2 0 7348.2 0.M .
| |
| FMLEY l 0 4 0 4 7216.4 0.55 1819.1 FMLEY 2 0 4 1 3 7549.7 0.40 2516.6 FERMI 2 2 2 3 1 143.0 0.71 lH3.0 FIT! PATRICK 0 2 0 2 8075.8 0.25 4037.9 FT.-CALHOUN I i 1 0 2 8405.2 0.24 4242.6
| |
| $1NM i 3 0 4 7716.3 0.52 1929.1 GRAND SLA.F 1 1 5 2 4 5624.6 0.71 IH6.1 MDDAM ECX 3 3 2 4 5H0.9 0.71 1265.2
| |
| ; NATCH1 0 4 0 4 5521.2 0.72 1300.3 MATCH 2 0 7 2 5 6451.9 0.D 1290.4 NOPECREEKI 2 9 I 3 2M9.5 1.12 II9.9 INDIANPolNT2 2 8 0 10 5101.9 1.96 510.2 IN9!M POINT 3 0 9 1 I 6501.6 1.22 822.7 KEMAUNEE 0 3 1 2 7584.3 0.26 3792.1 LACROSSE 2 16 le I 44M.8 1.82 550.1 LASELE ! 0 2 1 1 2395.7 0.42 2395.7 LASALLE2 0 5 2 3 M14.0 0.45 2204.7 LIMERICKI 6 1 0 1 71 4 .1 0.14 714.1 kBU!RE 1 e 3 0 3 5022.2 0.H 1674.1 kSUIRE2 1 5 1 5 5770.4 0.07 1154.1 RAIE TANKEE 2 4 4 6 7791.1 0.77 1290.5 MILLS 10E I I 2 1 2 1276.5 0.24 4130.3 RILLSTOE 2 0 4 0 4 6599.6 0.61 1649.9 RlLLSTDE 3 0 14 9 5 6201.1 0.01 1240.2 MONTICELLO 6 2 0 2 6904.9 0.29 3492.4 NIE MILE POINT l e 2 2 0 5023.5 0.00 .
| |
| NORTHAMAI 2 5 2 5 7560.0 0.H 1512.0 NORTHAWA2 0 4 1 3 7301.3 0.41 2433.I OCONEEI 0 2 0 2 5948.7 0.34 2974.4 BCONEE2 0 5 0 5 7253.7 0.69 1450.7 32
| |
| | |
| l Table 9 (cont'd)
| |
| E KTOR TRIP RATES YDR lth l NAME RMUAL AUTO LESS TMN IREATER CRiitCAL TRIP RATE PER EMTIE MATIC ORE 00AL TMN MURS IMO MIRS ETES TRIPS 151 PO ER 151POER PON G IT 15 POER IT 151 OCOEE 3 1 1 0 2 7835.4 0.26 3917.7 OYSTERCREEE 6 3 2 1 23M.! 0.42 2389.1 PALISADES 6 2 0 2 1490.5 1.34 745.3 PALO WRK I 4 13 0 13 5499.0 2.34 422.9 PALO VERE 2 0 10 2 3 3803.1 2.10 475.4 PDCH DOTTOM 2 1 4 4 5 7272.I 0.69 1454.6 P DCH 30iTOM 3 3 6 3 6 5919.3 1.01 tN.5 PGRY l ! 4 5 0 1429.6 0.00 .
| |
| P!LSRIR1 0 4 3 1 1715.5 0.58 1715.5 PolNT KACH 1 0 2 0 2 7905.4 0.25 3952.7 POINT M ACH 2 0 2 1 1 7262.7 0.14 7262.7 PRAIRIE ISLMS 1 0 2 1 1 7891.4 0.13 7890.4 PRAIRIE ISLMD 2 0 3 0 3 7972.1 0.38 2657.4 0U40 CITIES 1 0 4 2 2 6151.3 0.33 3075.6 00A0 CITIES 2 0 2 0 2 5728.0 0.35 2N4.0 RIVER KND 1 0 18 2 16 5305.7 3.02 331.6 R0l!NSON2 0 11 4 7 7111.9 0.90 1017.0 SALEM i e t i I 7097.2 1.13 887.2 SALEM 2 0 9 3 6 5629.3 1.07 138.2 SM ONOFRE 1 0 2 1 1 2975.1 0.34 2975.1 SM ONOFRE 2 1 7 1 7 6480.0 1.00 925.7 S M ONOFRE 3 1 5 2 4 7422.3 0.54 1855.6 SHOREHAM 1 0 0 0 0 1299.5 0.00 .
| |
| ST.LUCIE1 1 3 1 3 0424.0 9.36 2000.0 ST.LUCIE2 1 4 2 3 7326.7 0.41 2442.2 SUMER 1 0 6 1 5 8453.2 0.59 1690.6
| |
| $URRY ! 3 2 1 4 6233.2 0.64 1558.3 SUR4Y2 1 3 0 4 6171.1 0.65 1542.I SUSOUEHAWAi 0 0 0 0 6196.3 0.00 .
| |
| SUS 90EHMil4 2 1 1 1 1 59 4 .6 0.17 59 4 .6 TMI-! 0 4 1 3 6268.6 0.40 2009.5 TROJ411 0 4 0 4 7N4.1 0.57 17M.9 TURKEYPOINT3 1 5 0 6 6901.4 0.M !!64.7 TURKEY POINT 4 1 2 1 2 3040.1 0.M 1524.1 VDM0NTYANEE 4 2 2 0 43M.6 0.00 .
| |
| IIATERFORD 3 0 7 1 6 7011.6 9.M 1168.6 l INF-2 4 7 2 5 6391.5 0.79 1270.3 WOLF CREEE 1 0 7 2 5 6523.7 0.77 13M.7 YANKEE ROIE 6 4 1 3 8343.5 0.36 2781.2 Il0E 1 0 2 0 2 5411.0 0.36 2745.5 Il0N2 0 4 4 4 7783.5 0.51 1945.9
| |
| . m. =.
| |
| 52 417 123 344 33
| |
| | |
| hardware failure; inappropriate system operation, maintenance, or calibration; or testing of the initiating system. The results of these categorizations are shown in Figure 8. The data for the major balance-of-plant (B0P) systems (i.e., feedwater, turbine, condensate, main generator, and main steam) are provided in Figure 9. Fifty-seven percent of all reactor scrams at power levels greater than 15% were due to B0P systems in 1986. Nearly half of the B0P trips were attributed to the main feedwater system. The scram frequency (in terms of scrams per thousand critical hours) attributed to various initi-ating systems for 1986 is consistent with and follows the same general trend
| |
| ' seen in 1984 and 1985. AEOD's ongoing analysis of the 1986 scram data will provide additional details about root causes, component failures, and personnel interactions as they relate to initiating systems.
| |
| 3.6.4 Causes of Scrams (Above 15% Power)
| |
| The LER description of each scram was reviewed to determine the general classi-fication of cause or causes, and the results are shown in Figure 10. As has been the trend in past years, hardware failures dominated in 1986. Personnel related problems (i.e., human error, manual steam generator control problems and procedural deficiencies) in 1986 accounted for 29% of all reactor scrams above 15% power, which is similar to the experience in 1984 and 1985.
| |
| Unlicensed personnel (technicians, contractors, nonlicensed operating staff, and other utility staff personnel) were responsible for 16% of all scrams above 15% power; this represents a small increase from the 14% found in 1985. In 1985 technicians were involved in approximately one of every 12 scrams; in 1986, one in every nine scrams.
| |
| The four activities (maintenance, testing, calibration and troubleshooting) require interaction of personnel with plant systems. During 1986, maintenance, testing, and calibration were responsible for 36% of all scrams that occurred.
| |
| This compares with 30% of all' scrams when these activities were ongoing in 1984 and 1985. In 1986, when troubleshooting is added, the figure is raised by 4%.
| |
| It is noteworthy that in 1986, for all scrams above 15% power attributed to human error, 79% were caused by human errors during maintenance, testing, cali-bration, and troubleshooting activities. Unlicensed personnel (technicians) were responsible for 63% of the human errors during these activities.
| |
| 4 Hardware failures were encountered in 25% of the cases during these evolutions.
| |
| Figure 11 is provided to display the systems most frequently impacted during maintenance, testing, troubleshooting and calibrations. Testing only was respor.sible for 22% of all scrams in 1986. Testing associated with the RPS and turbir.e were responsible for the majority of these scrams (12% of all scrams).
| |
| A detailed analysis of these scrams will be provided in the AE0D 1986 scram report.
| |
| 3.6.5 Scrams With Associated Failures Unplanned scrams where the recovery was complicated by additional equipment failures or personnel error can be of concern because of the higher level of stress and demands placed upon the operating personnel and mitigating systems.
| |
| We define " associated failures" as component failure or personnel error that did not contribute directly to the cause of the scram, but are associated with post scram recovery (e.g., normally the failure was discovered or occurred when the component was actuated to mitigate the consequences of the scram).
| |
| 34
| |
| | |
| .a..---+_e ..a , m m -e _ _. - --
| |
| * _u --._m _ - _ . .-h e. - --- - - .
| |
| bb t
| |
| < E
| |
| ; }g ~
| |
| d a- f
| |
| ! "I "
| |
| 3"e E@
| |
| l 5 bs 4
| |
| l 9 eg
| |
| ~
| |
| ..I.l.il.lli.is l
| |
| lh o
| |
| g 8 n l
| |
| !s" a
| |
| ss:3 l8 3 i
| |
| W31SAS ON11VillNI rigure 8
| |
| | |
| i PRIMARY BALANCE OF PLANT SYSTEMS POWER GREATER THAN 15%
| |
| YEAR 1984 vss/s////r//r////r/r/ssr/r-A is.
| |
| E 39ss l <----------s l
| |
| TURBINE
| |
| '2 w '8 2 V//////A M MAIN GENERATOR t
| |
| o_
| |
| O
| |
| \ D i V////////4 l CONDENSATE I
| |
| I uAiN sTrAu g ,fe .52 E i NO. SCRAMS /1000 CRITICAL HOURS
| |
| | |
| CAUSE
| |
| | |
| ==SUMMARY==
| |
| | |
| POWER GREATER THAN 15%
| |
| YEAR 6 1984 i
| |
| DWME - g j 1985
| |
| ' Vffff/#ff/A 1986 HUMAN ERROR -
| |
| .'2 72 3 PROCEDURE -
| |
| - to O u)
| |
| .i D UNKNOWN -
| |
| r O
| |
| OTHER -
| |
| ENVIRONMENTAL -
| |
| E SG LEVEL CML -
| |
| g
| |
| .125 .250 .375 .500 O
| |
| NO. SCRAMS /1000 CRITICAL HOURS I
| |
| | |
| SYS"RS IPAC"D EOS:' ?EQXJ BY TESTINGMAINTENANCETROUBLESHOOTINGCAl.IBRATIONS POWER GREATER THAN 15%
| |
| YEAR 1986 SYSTEM FRED TURBINE 26 RPS 25
| |
| . ELECTRICAL 17 RCS 16 FEEDVATER 16 CTRI. ROD DR I1 CONDENSATE l 8 MAIN STEAM 6 OTHER 14 1
| |
| .......... .. r................ .............................,.
| |
| 0 5 10 15 20 25 30 l moen OTHER INCLUDES REMAINING SYSTEMS INITIATING SCRAMS i
| |
| Figure 11 i
| |
| f i
| |
| l i
| |
| 38
| |
| | |
| I In 1986 associated failures were experienced in 18% of scrams. The comparative statistics for the period 1984 through 1986 are shown below:
| |
| Scrams with Percentage of all Associated Scrams With Year Failures Associated Failures 1984 77 20 j 1985 109 23 i
| |
| 1986 66 18 3.6.6 Quantitative Safety Significance Measures In 1985, a concurrent program was in place to quantitatively measure the safety significance of an event sequence (transient) involving a reactor scram. The specialized probabilistic event tree technique called " Accident Sequence Precursor (ASP)" was used to evaluate the 1985 reactor scram sequences as documented in LERs. (See Section 5.2.6 for a description of the ASP program.)
| |
| Our use of this technique was exploratory in 1985, and results are not yet available for scrams occurring in 1986.
| |
| 3.7 Engineered Safety Features (ESF) Actuations ,
| |
| 3 Safety systems should work reliably and properly when challenged, and should not be challenged frequently or unnecessarily. In order to gain an understand-ing regarding the need and frequency of the challenges to safety systems, the Comission required, as part of the revised LER system, that actuation of an engineered safety feature (ESF) be reported to the NRC as an LER. This report-ing requirement became effective on January 1,1984. Prior to this date, the actuations of these systems were not directly. reportable.
| |
| ESF systems are designed to control and mitigate specific occurrences that might challenge the integrity of the reactor and/or adversely affect plant ptrsonnel or the general populace. Generally, these include systems designed to control reactor core reactivity, isolate and cool containment, supply emergency cooling to the reactor fuel, remove residual core heat, assure habitability of the control room under all conditions, control radioactivity releases to the environment, and provide a source of emergency power.
| |
| As part of the AE0D trends and patterns analysis program, ESF actuations are the subject of periodic detailed review. A study of ESF actuations for the first half. of 1984 was published as AE0D/P503 in August 1985. The analysis of the second half of 1984 was completed and was issued in August 1986 as AEOD/ !
| |
| P503. Data for 1985 were compiled, and analysis results will be issued in 1987. A summary of the findings for 1985, and a discussion of the trends from 1985 to 1986 (based on preliminary data), are presented below.
| |
| 3.7.1 Frequency of ESF Actuations It was found that 1582 ESF actuations were reported in 1985 involving 91 of the 97 reactors that were eligible for reporting per 10 CFR 50.73. Of the 97 units eligible, six units (6%) experienced no actuations during the year, 31 units (32%) had between one and five actuations, 18 units (19%) had between six and 39 l
| |
| | |
| 1 1
| |
| 1 ten actuations, eight units (8%) had between 11 and 15 actuations, six units
| |
| ! (6%) had between 16 and 20 actuations, and 28 units (29%) experienced more than i .20 ESF actuations, with the maximum number of actuations at any one unit being
| |
| , 86. Figure 12 shows the unit distribution of total ESF actuations for 1984, i
| |
| 1985, and 1986. Only about 9% of all reported ESF actuations involved an emergency core cooling system (ECCS), and none of these occurrences were
| |
| ; necessary to control an actual loss-of-coolant accident (LOCA). Over 80% of the actuations that occurred in ESF systems were associated with either an isolation function or a ventilation function.
| |
| I 3.7.2 Valid ESF Actuations l
| |
| j In 407 of the 1582 cases studied (26%), the measured parameter reached the i intended setpoint for ESF actuation; however, in only 15 of these 407 cases I were the actuations considered needed in terms.of providing a required ESF ,
| |
| i system response for protection from an actual design basis event. These 15 -
| |
| " design basis" ESF actuations represented less than 1% of the 1582 cases.. They l
| |
| occurred at 12 different units, with no unit having more than three events.
| |
| They included eight cases involving loss of offsite power resulting from a loss of a bus, four events involving a control building heating, ventilating, and air conditioning system, and five high radiation events.
| |
| i In the remaining 392 valid cases, the ESF actuations resulted from conservative
| |
| : alarm setpoints, equipment failures, or personnel errors. These ESF actuations l were considered to be valid (i.e., the measured parameter reached the intended actuation setpoint), but did not represent a response to a design basis event.
| |
| I Rather, they were actuations resulting from non-design basis conditions, such j as radioactive trash being moved near a radiation monitor. These valid but l non-design basis actuations were primarily associated with water level (reactor 4
| |
| water level in BWRs and steam generator level in PWRs), loss of power, flow
| |
| ! (about 90% were reactor water cleanup flow), or toxic gas monitors. Most of
| |
| ! them were associated with isolation of the reactor water cleanup system or the containment, or the ESF mode of actuation of the control room ventilation. The distribution of the ESF system functions involved and the measured parameters j
| |
| (signals) for valid ESF actuations are shown in Table 10.
| |
| ! 3.7.3 ECCS Actuations i
| |
| In 144 of the ESF actuations at 62 units, an ECCS was actuated. .Only 96
| |
| ! resulted in actual injection of fluid. These 96 actual injections occurred in
| |
| } 41 different units. Three units had more than four actual injections. The 96
| |
| ; actual injections were more than double the 43 actual injections in 1984. Of l these 96 actual injection events, none were needed to control an actual LOCA.
| |
| I 3.7.4 False ESF Actuations ;
| |
| ) A total of 1175 false actuations, or 74% of the 1582 ESF actuations, occurred
| |
| : at 87 units. These false ESF actuations principally affected systems whose functions were associated with either isolation or ventilation. False ESF t
| |
| l
| |
| , 40 i
| |
| f l
| |
| _ , _ . . , . _ _ . __.,___.,__..____,_-_-._,____..____,,--.m.___
| |
| m.,__--,m- - ., _ . _ _ _ _ . . - - -_ . _ _ _ . _ _ _ . . . , _ , _ _
| |
| | |
| UNIT DISTRIBUTION OF 1984/1985/1986 ESF ACTUATIONS ARKANSAS 1 ' YEAR ARKANSAS 2 -*'
| |
| BEAVER VALLEY 1 W 1984 BIG ROCK POINT ,"
| |
| BRAIDWOOD 1985 BROWNS FERRY 1 Ehiw.
| |
| BROWNS FERRY 2 -mas = 1986 BROWNS FERRY 3 -E1 BRUNSWICK 1 -~
| |
| BRUNSWICK 2 -BRL.-
| |
| w BYRON 1 EEliF'"-
| |
| s BYRON 2 5 CALLAWAY -mm--,
| |
| ' CALVERT CLIFFS 1 -n b
| |
| z CALVERT CLIFFS 2 -r U CATAWBA 1 -"-------
| |
| CATAWBA 2 -
| |
| CLINTON -
| |
| COOK 1 L COOK 2 - --
| |
| COOPER w CRYSTAL RIVER -W DAVIS-BESSE -s=
| |
| DIABLO CANYON 1 -EEE" DIABLO CANYON 2 -umr DRESDEN 2 -us-DRESDEN 3 -ama.
| |
| O 25 50 75 100 125 150 175 NUMBER OF ESF ACTUATIONS UNIT DISTRIBUTION OF 1984/1985/1986 ESF ACTUATIONS YEAR DUANE ARNOLD -,
| |
| FARLEY 1 -w .*M*
| |
| FARLEY 2 -r 1984 FERMI 2 - - -
| |
| FITZ PATRIC K -le 1985 FORT ST.VRAIN -ma FT. CALHOUN -" 1986 GINNA -=
| |
| GRAND GULF e HADDAM NECK -m-
| |
| $ H TCH -""------
| |
| 5 HATCH 2 "-
| |
| HOPE CREEK -
| |
| t INDIAN POINT 2 -L.
| |
| Z U IN DIAN POINT 3 4 KEWAUNEE -EP LACROSS E Rh - --
| |
| LASALLE 1 EEEEEE"'"-
| |
| LME Ck ~- ----------
| |
| MAINE YANKEE 4" MILLSTONE 1 -e MILLSTONE 2 -La.
| |
| MILLSTONE 3 - l N c"G UIR E W -I~
| |
| l O 25 50 75 100 125 150 175 NUMBER OF ESF ACTUATIONS j l Figure 12 41 l l
| |
| l
| |
| | |
| UNIT DISTRIBUTION OF 1984/1985/1986 ESF ACTUATIONS McGUlRE 2 -ha YEAR NINE MILE PT 1 a------ f"
| |
| ,f NINE MILE PT 2 -w -
| |
| 1984 NORTH ANNA 1 % l NORTH ANNA 2 1 1985 i OCONEE 1 l OCONEE 2 -r 1986 ;
| |
| OCONEE 3 - - l OYSTER CREEK L ;
| |
| W PALISADES -ICL.
| |
| PALO VERDE 1 ------
| |
| PALO VERDE 2 m- -
| |
| @ PEACH BOTTOM 2 -h z
| |
| rPEACH BOTTOM 3 -h.ua E PERRY D PILGRIM -Km.
| |
| POINT BEACH 1 P POINT BEACH 2 -C PRAIRIE ISLAND 1 ->
| |
| PRAIRIE ISLAND 2 -w OUAD CITIES 1 -r.ee=
| |
| OUAD CITIES 2 -l>=
| |
| RA g ;,y ______
| |
| ROBINSON %
| |
| SALEM 1 "
| |
| O 25 50 75 100 125 150 175 NUMBER OF ESF ACTUATIONS UNIT DISTRIBUTION OF 1984/1985/1986 ESF ACTUATIONS SALEM 2 -L YEAR SAN ONOFRE 1 4 = - - - - - - - g-# 1984 SAN ONOFRE 2 SAN ONOFRE 3 ii.ii --
| |
| SEABROOK , 1985 SEQUOYAH 1 -
| |
| SEQUOYAH 2 4. ""'" "
| |
| 1986 SHOREHAM ---
| |
| ST. LUCIE 1 1 W
| |
| ST. LUCIE 2 #
| |
| SUMMER F
| |
| @ SURRY 1 -16 .
| |
| Z SURRY 2 -6
| |
| - SUSOUEHANNA 1 dCElta=
| |
| 5 SUSOUEHANNA 2 -iKP'--
| |
| R TMI- 1 4 TROJAN -P TURKEY POINT 3 -b=
| |
| TURKEY POINT 4 -Ita VERMONT YANKEE -D WATER FOR D W WNP-2 =''i--
| |
| WOLF CREEK -----
| |
| YANKEE ROWE 1"*
| |
| ZION 1 ima ZION 2 4.e=
| |
| 0 25 50 75 100 125 150 175 NUMBER OF ESF ACTUATIONS Figure 12 (cont'd) 42
| |
| | |
| Table 10 I
| |
| Measured Parameters and Associated System ,
| |
| Functions for Valid ESF Actuations !
| |
| Occurring in 1985 General System Functi0n EASSED TLUID EATIES AfD ISOLAfim MMNT NOT TOTAL PARAETER VERTILAflW POWER ETtmINS (Signal) (wAC) (ntsc) 7 1 3 1 1 13 i
| |
| nasuaL LafVOLT 21 18 18 38 1 94 Laf $$ PRES $ tat 1 0 3 0 0 4 Lei RX VES$tL LEVEL 15 4 28 2 0 47 Laf RX PatSSWE 4 1 2 0 0 7 IIIs STEAft LIE ftf1M 1 0 1 e e t C et RA0!Afi m 8 $3 25 0 1 90 FIRE O 9 s 0 0 0 m m TORIC 4AS 0 3 4 0 0 3 Et RX WATER CLEARF SYSTDI FLal e e at e e It Ittel ROOM DIFERDITIAL TDFDAM S 0 14 0 0 14 met PRIMARY CONTAlmstT PRESM e 1 0 0 1 2 j Laf CONTAllBOIT VActAgt 3 8 e e 9 11 CIpelRATIM 95 fe es as 8 150 j IN NOIA e 0 0 0 1 1
| |
| ' G S e 0 e 0 IIIM STEAft FLW J Let $6 LEVEL 24 1 1 0 1 27 met RESIDunL EAT RDOVAL SYSTDI FLOli 0 0 1 0 0 1
| |
| {
| |
| 31 TRIP 2 0 1 0 0 3
| |
| ! met RX Ptflstat 5 0 t t t 9 Itt e 54 LEVEL 5 0 4 0 0 11 RIPPtB HA!E Ftta FIBF 10 0 0 0 t " 24 STWR 11 2 12 2 4 33 l met RYWELL Ptt$sWE G e e 3 0 0 Asf L0lf LEVEL 1 0 9 0 9_ 1 Laf STEAft LIE PESM 5 1 7 0 1 14
| |
| < L0lf CONODISOR VACIA38 0 0 3 0 e t l Laf SAMPLE PL W G G G G e e l Cet TDFDIAM 1 9 3 e e 3 i TOTAL ACTUATlWB 196 113 306 E IS IBF ,
| |
| 1 I
| |
| i I
| |
| 43 0 . .
| |
| | |
| 1
| |
| ['
| |
| I actuations were caused mostly by spurious actuations, equipment failures, or problems related to personnel. The main parameters (signals) involved with
| |
| : these false actuations were radiation and loss of power.
| |
| ! 3.7.5 ESF Failures In only ten of the 1582 ESF actuations (less than 1%) did an ESF system (the entire system) fail to actuate properly. Only one unit had more than one such
| |
| ; event.
| |
| In 78 additional cases the ESF actuations performed properly, but one or more -
| |
| failures were associated with the actuation. There was no identifiable trend i
| |
| or pattern for these failures, and redundant systems were available to perform required safety functions. Only three units had more than four such failures.
| |
| l 3.7.6 Conclusions Based Upon 1985 Data ;
| |
| I Analyses of the 1985 data support the following general conclusions: !
| |
| Fifteen units were identified as experiencing repeated unresolved ESF 3' actuations which could ultimately challenge continued equipment operability and proper personnel response.
| |
| i' The remaining events necessitating ESF actuations, including ECCS actua-j tions, have not been individually significant. This also holds true for failures and problems associated with the ESF actuations. In addition, it was readily apparent that the majority of the ESF actuations were unnecessary, and that the rate of these actuations could be greatly decreased by: (1) reducing the number of equipment failures during normal
| |
| ! operation, (2) reducing the number of personnel errors during maintenance i
| |
| and testing, and (3) revising actuation setpoints to more appropriate protective levels. ESF functions associated with isolation or ventilation I should receive first priority.
| |
| 3.7.7 Use of ESF Actuations as Performance Indicators The wide variety of ESF systems and the differences in the types of ESF actua-tions (including variations in imediate safety significance) make comparisons i
| |
| among units very difficult. As a result, only selected ESF actuations are
| |
| , being used as perfomance indicators [i.e., unplanned ECCS actuations that result from reaching an ECCS actuation setpoint or from a spurious or inad-vertent ECCS signal, and the number of emergency ac power system actuations that result from the loss (deenergization) of a safeguard bus.]
| |
| 3.7.8 Comparison of 1984, 1985, and 1986 Data Data for 1986 is currently being analyzed in detail, and complete results are not available at this time. However, some preliminary trends can be described
| |
| ' using the preliminary 1986 data. The average number of ESF actuations per reporting plant per unit of time increased from 1984 to 1985, but appears to be 1 decreasing from 1985 to 1986.
| |
| 4 44 4
| |
| e.- .. .w- v.... ..-, .,i, , , . , , . ,.3.-, -,._ .. , , . . . ,, ,w.s, _%,.
| |
| | |
| 1 The number of units reporting at least one ESF actuation increased from 1984 to 1985, and again from 1985 to 1986. The maximum number of actuations at a single unit decreased from 1984 to 1985, but increased from 1985 to 1986.
| |
| Safety injections more than doubled from 1984 to 1985, but decreased dramatic-ally in 1986. While valid actuations remained fairly level, false actuations increased greatly from 1984 to 1985, but decreased in 1986. .
| |
| Assuming the 1986 preliminary data-is confirmed, it appears that the rate of ESF actuations is decreasing for the first time since such events became reportable. Comparisons of ESF actuations for 1984, 1985, and 1986 are summarized in Table 11.
| |
| )
| |
| 3.8 NPRDS Trends and Patterns Analysis Program I
| |
| In 1986, AE00 began the full scale implementation of its NPRDS Trends and i
| |
| ; Patterns Analysis Program, a part of the AEOD Trends and Patterns Program.
| |
| This program analyzes component failure data from the Nuclear Plant Reliability
| |
| '- DataSystem(NPRDS),adatabasevoluntarilysupportedbytheU.S. nuclear power plant (NPP) industry and maintained by the Institute of Nuclear Power ;
| |
| Operations (INP0). The data base includes records that provide nameplate information on the design characteristics of the components as well as records
| |
| ! d: scribing their failures.
| |
| Major components within the main feedwater (MFW) system of pressurized water reactors (PWRs) were the first components selected for analysis. This system was selected because loss of MFW is a part of a majority of the dominant accident sequences in probabilistic risk assessments of PWRs. Furthermore, MFW transients are the major source of unplanned reactor scrams.
| |
| The goal of the trend and pattern analysis of.MFW component failure data is to identify component attributes.that are associated with relatively high inci-dences of failure. Manufacturer, valve type, and pump rotational speed are examples of component attributes under study; in addition, the pattern of failures among NPP units is studied. 7 The initial step in the trend and pattern analysis is a review of_ data with the utilities to verify or clarify any data that seem atypical or inconsistent.
| |
| Although differences exist among PWR plant MFW systems, the common function of the specific components under study provides a relatively homogeneous population.
| |
| i The second step is to select, from the attributes listed above, a set of component attributes (i.e..' statistical variables) to analyze for their sta-tistical relation to failure.
| |
| A series of statistical methods from the' field of survival analysis is then applied to the data to identify trends and patterns in failures with regard to the variables. The methods use the component lifetimes (i.e., times between failures) and the component attributes. These methods identify possible l
| |
| i 45
| |
| 'T m-w.,-. ,e v-m-e u '---=m>------,-.--v= ~~~a-n s~- =rn, w ,e v-~n -vm- , e ~mw-ew-, e, -e-~s-ww--w---wn m- -w+-,rn
| |
| | |
| Table 11 ESF Actuations for 1984, 1985, and 1986 Category 1984 1985 1986*
| |
| Total ESF Actuations 1102 1582 1422 Units Eligible to Report 91 97 106 Units Reporting ESF Actuations 81 91 101 UnitsReporting(%):
| |
| No. of ESF Actuations 0 11% 6% 5%
| |
| 1-5 49% 32% 34%
| |
| 6-10 16% 19% 16%
| |
| 11-15 3% 8% 13%
| |
| 16-20 3% 6% 8%
| |
| > 20 16% 29% 25%
| |
| Maximum ESF Actuations at Single Unit 152 86 104 ECCS Actuations (%) 8.4% 9.1% 6%
| |
| No. of Actual Injections 43 96 55 No. of injections for LOCA 0 0 N/A Isolation / Ventilation Actuations (%) 70% 80% N/A Valid ESF Actuations 400 (36%) 407(26%) 418(29%)
| |
| Design Basis 34(3%) 15(1%) 33(2%)
| |
| Non-Design Basis 366(33%) 392 (25%) 385(27%)
| |
| False ESF Actuations 702 (64%) 1175 (74%) 1004 (71%)
| |
| * Preliminary Data N/A: Not available at this time.
| |
| 46
| |
| | |
| factors that influence the failure time distributions. For numerical attrib-utes such as valve inlet size, a method tests for trends. All of these sta-tistical methods are tailored to use truncated lifetimes (shortened by the data cutoff date or the study start date) as well as actual times between failures.
| |
| In addition, the failures are studied in calendar time to detect shifts in the rate of component problems and identify specific attributes with rates j significantly above an average baseline failure rate.
| |
| l The final step is an engineering follow-up of the statistical results. The factors identified by the statistical analysis are starting points for this
| |
| , investigation. Failures for components with attributes flagged in the sta-l tistical analysis are reviewed, considering reported event details as well as component attributes. Root causes for these failures are sought through
| |
| @xtensive discussions with appropriate plant personnel, NRC inspectors, and component manufacturers. By identifying the underlying causes and the prac-tices that prevent recurrence, the benefits of experience can be shared to upgrade the MFW system performance for the entire PWR nuclear power plant industry.
| |
| To date, AE0D has studied MFW flow control valves, bypass flow control valves, and turbine-driven pumps. During 1986, three proprietary reports were prepared. The first two, " Failure Time Analysis of Feedwater Flow Control Valve and Valve Operator Components Within the Nuclear Plant Reliability Data System" and " Failure Time Analysis of Feedwater Flow Control Bypass Valve and Valve Operator Components Within the Nuclear Plant Reliability Data System" described the statistical analysis of these components. The third report,
| |
| " Resolution of Feedwater Flow Control Valve and Feedwater Bypass Valve Issues Identified in NPRDS Trends and Patterns Analysis," discussed the follow-up engineering investigation and focused on resolving the issues identified in the statistical analysis and on identifying the practices and conditions that have been responsible for some of the problems. It gave reconnendations for practices that help remedy these problems and minimize their recurrence.
| |
| The primary finding of the three studies issued in 1986 is that differences among units and stations have a greater influence on the performance of these components than any of the other component attributes studied. The reports on the flow control valves and the bypass flow control valves identified ten and six units, respectively, as outliers because of their frequent problems with these valves. All the units except one identified as having relatively high rates for bypass valve failures also were identified as having high reporting fer full-power main feedwater flow control valve problems.
| |
| Results of the follow-up engineering investigation for the flow control valves and the bypass (startup) flow control valves are similar; therefore, they are l discussed together. Table 12 presents the findings. Most of the recommenda-tions involve improved maintenance and maintenance procedures, the upgrading i of the instrument air system, the use of improved valve packing, and the use of components resistant to operating environments.
| |
| 47
| |
| | |
| 1 Table 12 l
| |
| 1 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 fasteners (especially for the solenoid valves).
| |
| Valve 011. moisture Upgrade the instrument air system with
| |
| . operator and/or rust, improved blowdown valves and dryers, or foreign particles in Monitor instrument air quality and the instrument establish maintenance schedules allowing j air system prompt corrective action.
| |
| Valve Outdoor weather Use waterproof solenoids.
| |
| operator conditions failure
| |
| ; Valve and Poor Use detailed maintenance procedures that valve maintenance assure the completion of proper maintenance operator procedures and adjustments before system startup.
| |
| failures Provide adequate training and support of the maintenance personnel.
| |
| , Consult with <alve manufacturers to establish efficient routine maintenance schedules.
| |
| Have valve manufacturers refurbish the j valve trim instead of doing this in-house.
| |
| ! Cover disassembled valves during maintenance.
| |
| Valve Packing leaks Use new packing materials with low released shrinkage and designs that maintain leakage constant pressure on the packing (spring-loaded,forexample).
| |
| Bonnet / flange In maintenance, carefully inspect the I flange before reassembly.
| |
| ! Valve Improperly Use improved, valve-specific maintenance contained adjusted valve procedures.
| |
| j leakage operators Damaged valve Use proper maintenance.
| |
| trim (plugand Consult valve manufacturers for advice cage or seats) on improved valve tria designs and materials for actual plant conditions such as higher pressure drops.
| |
| 48
| |
| : .. . . - _ _ . _ - . _ . - _ _ _. . , _ . . _ _ _ . . _ - - - _ _ , _ . _ _ - _ _ , _ . , _ . . _ . _ ~ . _ - _, _ , - . _ _ . _ _
| |
| | |
| a .- _.:_.a..,,. r ---- - - -
| |
| 1 In conclusion, the engineering evaluation of the statistical results from the trend and pattern analysis shows that proper maintenance and the use of appro-priate materials dominate in avoiding problems with MFW components. Although the MFW system is not a safety system, upgrades of that system and supporting systems such as the control air and oil systems to make the MFW system more reliable will reduce reactor scrams and unnecessary demands on safety systems..
| |
| l l
| |
| 4 t
| |
| i I
| |
| i 4
| |
| i 49 i
| |
| - . . . - -. , ,_. . -. - . . .. - - . . . . . , _ , , . -._m. - __.
| |
| : 4. COMMENTS AND OBSERVATIONS ON 1986 OPERATING EXPERIENCE AT OTHER LICENSEES During 1986, a number of events involving NRC and Agreement State licensed nonreactor facilities and activities were reported to the NRC. This section provides an overview and summary of reported events involving these nonreactor facilities, and medical misadministrations occurring in 1986.
| |
| 4.1 Nonreactor Eveng The AE00 Nonreactor Event Report (NRER) data base contains infomation on licensed nuclear materials and fuel cycle operational events and on personnel radiation exposure events. The NRER data base management system provides for input, storage, retrieval, and computer-assisted analyses of operational event data, and may be used to identify trends in operational safety events which may signal a need for remedial actions by the NRC and/or licensees.
| |
| 4.1.1 Occurrences in 1986 )
| |
| The NRER data base includes 202 records of events that were entered into the data base during 1986. Infonnation on these events was contained in reports submitted by nonreactor licensees to the Regional Offices or in other docu-ments, primarily inspection reports. The data base does not include infonna-tion from certain fuel cycle licensee reports, such as those related to routine effluent releases, nor does it include information from reports of medical misadministrations (a separate data base on this subject is maintained - see Section 4.2). Table 13 provides infonnation on the types of licensees for which information was entered into the data base.
| |
| An NRER data base item may be associated with more than one category of event.
| |
| For example, a report from a radiography licensee concerning a personnel radiation exposure would be counted in the total number of radiation exposure events as well as in the total number of events involving radiography. The 202 nonreactor licensee reports were cataloged as 317 entries in ten different areas (see Table 13 for details); in addition, reports received for which no program code was available, were delegated to the "Other" category. Note that, because some reports are associated with more than one event category, the total number of events exceeds the total number of reports.
| |
| Certain categories in Table 14 are primary categories; that is, they contain events from all types of licensees. These primary categories are: exposures; lost, abandoned, or stolen material; leaking sources; release of material; and consumer products. With few exceptions, most of these events are assigned to only one of the categories.
| |
| The secondary categories in Table 14 are designed to capture events by the type of licensee involved in the event. Many of the events assigned to these categories are also assigned to primary event categories. Secondary categories generally serve as a measure of the frequency with which certain types of licensees make reports to the NRC.
| |
| 50
| |
| | |
| I l
| |
| ! Table 13 Types of Licensees That Submitted Reports During 1986 l Number of License Type _ Reports Received
| |
| * Academic 2 Medical 41 Commercial / Industrial Measuring Systems 54 Well Logging (22)
| |
| Other Measuring Systems (32)
| |
| Manufacturing and Distribution (Excluding Medical) 12 12 Industrial Single Radiography Location (In Plant) (3)
| |
| Multiple Locations (Field) (9)
| |
| Irradiator 3 R&D 12 Source Materials ** 8 Mills (1 UF-6 Facilities 3 Other 4 Special Nuclear Material (Including Plutonium) 10 Agreement State 17 Other*** 31 Total 202
| |
| * Medical misadministration reports are not included, c* Routine environmental effluent release reports, e.g., reports required ,
| |
| by 40.65 and 70.59 were not included in the totals for source and special nuclear materials licensees.
| |
| *** Number includes reports received for which no program code was available.
| |
| I 51
| |
| | |
| Table 14 Categorization of Nonreactor Event Reports Occurring During 1986
| |
| }
| |
| i Category
| |
| * Number of Reports Associated i Primary Categories:
| |
| ; Personnel Radiation Exposures 37 Lost, Abandoned, and Stolen Material 68 Leaking Sources 21 Release of Material 22 Consumer Products 4 Secondary Categories:
| |
| Fuel Cycle (e.g., Mills, UF-6 Facilities 13 Special Nuclear Material)
| |
| Industrial Radiography 19 l Manufacturing and Distribution 34 l
| |
| (IncludingMedical) 1 Commercial / Industrial Measuring Systems 31 (Excluding Well Logging)
| |
| Other** 68 i
| |
| Total 317
| |
| *An NRER database item may be associated with more than one category of event, for example, a report from a radiography licensee concerning a
| |
| ; personnel radiation exposure would be counted in the total number of
| |
| , radiation exposure events as well as in the total number of events involving radiography.
| |
| i
| |
| **0ther includes categories such as medical, transportation, miscellaneous,
| |
| ; etc.
| |
| i d
| |
| }
| |
| d 52
| |
| | |
| 4.1.2 Radiation Exposure Events The NRER data base contains information from 37 reports of events that were entered during 1986 in which there was the potential for or an actual radia-tion overexposure. Of these 37 events, 13 involved actual radiation overexpo-sures; a fourteenth event (Department of Air Force, Wright-Patterson Air Force Base) may have resulted in an overexposure. The types of licensees associated t:ith the actual overexposures reported during 1986 were as follows:
| |
| Number of Total Number of Licensee Type Overexposure Events Individuals Exposed Medical / Academic 2 2 Radiography 8 10 Commercial / Industrial 1 1 Fuel Cycle 1 2 Broad Scope 1 1 possible General Licensee 1 1 Total 14 16 + 1 possible Medical / Academic Licensee - The two reports of overexposures at medical or academic licensees involved licensee personnel and included: two whole body exposures (3.4 rem / quarter and 21 rem). The reports came from different licensees.
| |
| Radiography Licensee - All of the radiography overexposures were received by licensee personnel. Five of the eight reports were whole body expo-sures, and generally less than 5 rem / quarter. In a sixth event, occurring at a DOE site, film badges of two individuals showed whole body exposures of less than 5 rem. Extremity exposures for this event were unavailable at the time this report was prepared.
| |
| Two events occurred in prior years in Agreement States. In the first, an overexposure that was received in 1984 was not reported to the State until late in 1985. The radiographer may have received a hand exposure of 29,000 rem and a whole body exposure of 47 rem. In a second overexposure event, a radiographer received an estimated hand dose of 2000 rad and a whole body dose of about 6 rad. The overexposures occurred at different licensees.
| |
| Comercial/ Industrial Licensee - The commercial / industrial overexposure received by a licensee employee involved an extremity overexposure (42 rad).
| |
| Fuel Cycle Licensee - The reported fuel cycle overexposure resulted in calculated internal exposures of two individuals that exceeded 40 MFC hours.
| |
| 53
| |
| | |
| Broad Scope Licensee - An individual involved in decontamination activi-ties at Wright-Patterson Air Force Base inhaled radioactive material (Am-241) that may have resulted in an overexposure.
| |
| General Licensee - A member of the public could have received a radiation exposure of from 0.6 to 1.7 rem (buttocks) and from 69-200 rem to the leg from a gauge. Effective control over the gauge had been lost.
| |
| In general, all of the 1986 overexposures were minor. There were two substan-tial overexposures to radiographers that occurred in prior years in Agreement States. Thesetwoeventswerereportedasabnormaloccurrences(seeAppendixA for brief sumaries of abnormal occurrences).
| |
| 4.1.3 Lost, Abandoned, and Stolen Material Licensees are required to report the loss or theft of ifcensed material that has occurred in such quantities and under such circumstances that it appears to the licensee that a substantial hazard may result to persons in unrestricted areas (10CFR20.402(a)(1)).
| |
| Sixty-eight events occurred during 1986 that involved lost, abandoned, or stolen licensed material. These events consist of 47 reports of lost or stolen material, plus 21 reports of abandoned, irretrievable well-logging sources.
| |
| One of the 68 events resulted in a radiation overexposure. In this event, an Abnormal Occurrence, effective control over an industrial gauge was lost and a member of the public was overexposed.
| |
| Of the 47 reports of lost or stolen sources, 21 were found, two pacemakers were lost (one probably was buried and the other was sent to a land fill). The location of the remaining sources is not known. Generally, the sources whose whereabouts are unknown were small. There were however three reports of lost (or misplaced) sources containing 1 Ci or more:
| |
| Cs-137 -
| |
| Kansas University Medical Center reported the loss of 5 Ci Cs-137 source.
| |
| Am-241 - Dresser Industries reported that a 4.8 Ci Am-241 source could not be located. It is believed that the source is still in the licensee's possession.
| |
| Kr-85 - Sweetheart Products Group reported the loss of a 1 Ci Kr-85 source on a comon carrier.
| |
| 4.1.4 Leaking or Contaminated Sources Certain licensees are required to leak test sources and to report leakirig sources under 10 CFR 34.25; others are required to leak test sources and to report leaking sources as a license condition. In both cases, a removable contamination exceeding the most comon test limit for removable contamination (0.005 microcuries) is considered evidence of leakage, and must be reported to the NRC.
| |
| Twenty-one events of leaking or contaminated sources occurred during 1986.
| |
| None of the events resulted in a radiation overexposure. The isotopic sources 54
| |
| | |
| fcund to be leaking or contaminated contained americium, cesium, cobalt, j tritium, iodine, nickel, and technetium.
| |
| l Most events were reports of small, individual sources found to be leaking or ccntaminated. One source leakage event was attributed to damage to sources during cleaning; another event was attributed to packaging; and a third event ccncerned damage to stolen radiopharmaceutical packages. One sealed source was discovered to have removable contamination after a steam leak in the facility.
| |
| A preliminary overview of the reports shows that the only generic problem displayed in the events, in which sources were damaged by cleaning, has been resolved by the licensee.
| |
| 4.1.5 Release of Materials Twenty-two events occurred in 1986 involving the release of radioactive materi-als. Several of these had significant consequences:
| |
| In an event at Sequoyah Fuels in January 1986, a cylinder filled with from 14 to 15 tons of hot liquid uranium hexafluoride (UF-6) ruptured, releas-
| |
| : ing a large fraction of the contents as UF-6 gas. The UF-6 hydrolyzed, forming uranyl fluoride UF-2F-2, and hydrogen fluoride (HF, hydrofluoric acid). The cloud of UF-2F-2 and HF was dispersed by a 25 mph wind. A highway adjacent to the plant was closed for a period of time. (See Section4.1.7.)
| |
| In an event at Wright-Patterson Air Force Base, a building used to house low level radioactive waste became contaminated when an unlabeled waste drum was opened on two occasions to inspect its contents in an attempt to determine the origin of the waste. After the first occasion, which resulted in contamination of the building, the Air Force Radiological Assistance Team was activated and arrived, together with a private con-tractor, to assist in decontamination and repackaging of the waste. The same drum was again opened, and resulted in further release of radioactiv-ity. One individual may have received an uptake (inhalation) of airborne americium-241 that exceeded the NRC limit of 3-8 nanocuries. The uptake is still being evaluated. Costs to date for decontamination, repackaging, and disposal are about $S00,000. The fiscal decision whether to decontam-11 ate the building further or to dismantle and dispose of it is still open.
| |
| In an event in North Dakota, a tractor-trailer truck carrying yellowcake had an accident with a train. The accident resulted in contamination by yellowcake. The State Department of Health supervised the cleanup.
| |
| Several reports of krypton-85 releases were were received in 1986, of which none had serious consequences. There were six reports of losses of krypton-85 from fine leak test devices. The devices were manufactured by two different firms. The leaks were attributed to various causes: piping (plumbing) leaks; failure of a ten-year-old logic printed circuit board; installation of a new electrical component that was apparently inadequate-ly sized; leaking valves; and inadequate maintenance. In two events, an 0-ring seal on the device contributed to or caused the leak.
| |
| i 55
| |
| | |
| 4.1.6 Consumer Products An additional category, " consumer products," was defined for the data base in 1985. These reports describe events in which radioactive material was found in, or had a reasonable probability for being introduced into, nonlicensed i consumer products. Four reports of this nature were received in 1986: l
| |
| * A technician at Oyster Creek possessed several one dollar bills contami-nated with I-131. He had received the money from a bank the day before the event.
| |
| * Several (97) contaminated or potentially contaminated snubbers were shipped from Farley Nuclear Station by Alabama Power and Light from Alabama to California. Ten snubbers were not located.
| |
| * A 51orida licensee, Florida Steel, found a small piece of radioactive material, a Sr/Y beta source, in an incoming shipment of scrap.
| |
| * About 700 gallons of masking tape adhesive were contaminated with radioac-tive material from a static elimination device. The licensee, an Agree-ment State licensee, stated that the adhesive would not be used in products for commercial distribution without authorization.
| |
| 4.1.7 Fuel Cycle Facility Event Reports There were 13 fuel cycle events entered into the nonreactor data base in 1986.
| |
| On January 4,1986, a large cylinder filled with molten uranium hexafluoride ruptured at the Sequoyah Fuels Corporation Facility in Gore, OK. The event was perhaps the most significant fuel cycle event in many years, and as such, resulted in an increased sensitivity to events involving UF-6. A total of five of the 13 fuel cycle events concerned UF-6-related events; seven other events occurred at fuel fabrication facilities, and one event occurred at a uranium mill.
| |
| UF-6-Related Events:
| |
| UF-6 Release at Sequoyah Fuels Facility, Gore OK - On Saturday, January 4,1986, a large cylinder of molten uranium hexafluoride ruptured while it was being heated in a steam chest. The force of the explosion damaged the steam chest enclosure. The escaping UF-6 rapidly reacted with moisture in the air to form uranyl fluoride and hydrofluoric acid. The resulting vapor cloud of these materials was carried south by southeast by a wind gusting to 25 mph. The cloud enveloped the process building, and the acidic vapor caused the death of an operator who was working approxi-l mately 70 feet from the cylinder. The va)or was drawn into the plant ventilation system. Approximately 40 workers in the building evacuated to an upwind location on site, some passing through the cloud. Plant person-nel manned water hoses with fog nozzles in an attempt to suppress further airborne release of material.
| |
| Notification of NRC and civil authorities is estimated to have begun within 10 minutes. The injured workers were transported by fellow workers to nearby hospitals for treatment. The general population dcwnwind was contacted and advised to evacuate and report to hospitals for examination.
| |
| 56
| |
| | |
| l Various local. State, and Federal officials were notified by the licensee from the corporate office in Oklahoma City. Resider.ts downwind of the site were contacted by licensee employees and advised to proceed to a l local hospital. The general public was notified by local radio.
| |
| Within an hour and a half, radiological surveys b'egan both on and off the site. State, Federal, and company officials arrived at the site during the ensuing hours, and recovery operations began.
| |
| The NRC initiated an Augmented Investigation Team (AIT) inspection (the first such AIT for a fuel cycle event). The AIT report, " Rupture of Model 48Y UF-6 Cylinder and Release of Uranium Hexafluoride" (NUREG-1179),
| |
| concluded that the cylinder was not defective but failed because of stress caused by hydraulic pressure that resulted from the expansion of the UF-6 in the cylinder when it was heated.
| |
| There were two major causes of the event:
| |
| I --
| |
| The physical equipment and facilities used for filling and weighing UF-6 cylinders were inappropriate for safe use with 14-ton cylinders.
| |
| The training of workers in operating procedures and ensuring the implementation of these procedures were not carried out effectively.
| |
| In addition to forming an AIT, the NRC undertook three other major actions to appraise the significance of the event at Sequoyah Fuels:
| |
| An Ad Hoc Interagency Public Health Assessment Task Force was estab-lished. The Task Force published a report, " Assessment of the Public Health Impact from the Accidental Release of UF-6 at the Sequoyah fuels Corporation Facility at Gore, Oklahoma" (NUREG-1189), document-ing their assessments.
| |
| A Lessons-Learned Task Force (LLTF) was established. Their findings were furnished in the report, " Release of UF-6 from a Ruptured Moded 48Y Cylinder at Sequoyah Fuels Corpuration Facility: Lessons-Learned Report"(NUREG-1198).
| |
| A Study Group was established to review the regulation of nuclear
| |
| ^i materials safety by the NRC. Their report, " Materials Safety Regula-tion Review Study Group Report," was published for public comment in the Federal Register on December 17, 1986.
| |
| Recomendations in the LLTF and the Study Group reports, as well as the public comments on the Study Group report are currently under review within the NRC.
| |
| UF-6 Cylinder Deformation at Allied Chemical Company Metropolis, IL -
| |
| Following the Sequoyah event, the NRC Region !!! Office questioned Allied about cylinder overfills at its facility in Metropolis, IL. The NRC had been called by a newspaper reporter asking about a 1984 overfill at Allied. Allied, on questioning by HRC, stated that an overfilled cylinder had been heated in December 1984 with resulting damage to the cylinder (cracked stiffening rings; deformed wall). There were no releases of UF-6 57
| |
| | |
| during the event. NRC held that the event was reportable under Chapter 10 CFR Part 20. (Allied informed the NRC of an additional UF-6 cylinder overfilling event of March 23, 1986. The excess UF-6 was removed from the cylinderwithoutheating.)
| |
| Leaking Valve on UF-6 Cylinder - Allied Chemical Company filed a report of a leaking valve on a UF-6 cylinder undergoing an air pressure test.
| |
| Transportation Accident on August 7, 1986 - A UF-6 cylinder being shipped from Babcock and Wilcox Company (Lynchburg, VA) to DOE broke free of the transport truck and fell on the s?reet. Although there was superficial damage to the cylinder, the vent valve cover and plugs were intact and all seals were in place.
| |
| UF-6 Pigtail Failure at Nuclear Fuel Services - On September 14, 1986, a pigtail between a heated UF-6 cylinder (in a glove box) and pro:ess equipment failed, releasing UF-6 to the glove box and the ventilation system. The calculated dose commitment to a child would be 0.02 mrem whole body exposure (at 250 meters). The pigtail was of a design that was not nonnally used.
| |
| Other Fuel Cycle Events:
| |
| Milling - Plateau Resources Ltd. calculated that exposures of two employ-ees working in the yellowcake packaging area exceeded 40 MPC hours.
| |
| Uranium Fuel Fabrication -
| |
| -- At the Babcock and Wilcox Navy Nuclear Fuels Division facility, there was an accumulation of uranium in ventilation ducts that exceeded 75%
| |
| of the critical mass value of 0.63 kg U-235. The accumulation of uranium was not adequately controlled.
| |
| -- During a routine survey of an unrestricted area at the Babcock and Wilcox Research Center, it was discovered to be contaminated. Signs were posted, and a fence was extended to encompass the restricted area.
| |
| -- A finishing unit at Nuclear Fuel Services (NFS) became pressurized, resulting in the release cf radioactivity into the processing area.
| |
| No overexposures resulted from the release.
| |
| -- NFS personnel were washing con'tamination off their hands in a lunch-room sink.
| |
| -- There was an elevated uranium concentration in sludge at a waste treatment plant in Erwin, TN. It was determined that the uranium l
| |
| comes largely from releases at an NFS onsite laundry facility. NFS was to take action to reduce the releases.
| |
| -- There was a hydrogen fire at NFS occurred on November 5, 1986.
| |
| Reserve tubes of hydrogen located outside of the process buildings.
| |
| 58
| |
| | |
| i The licensee declared an unusual event and implemented its Radiolog- l ical Contingency Plan. Most plant areas were temporarily evacuated. ;
| |
| No radioactive materials were involved.
| |
| l I --
| |
| General Electric Company reported that defective installation of high efficiency particulate air filters occurred in 1985. Installation procedures were deficient.
| |
| 4.1.8 Industrial Radiography Nineteen 1986 events involved radiography. One event occurred at a fixed radiography site and 18 occurred at remote (field) radiography sites or during transportation.
| |
| S;venteen of the reports concerned overexposure or potential overexposure events, and have been discussed previously. The number of radiography events does not differ substantially from the number of events reported during prior years.
| |
| 4.1.9 Manufacturing and Distribution Thirty-four 1986 events were identified from the program code of the licensee as being associated with manufacturing and distribution. These licensees have no unique reporting requirements for events involving health and safety, unless the requirements are incorporated into a license condition or an order.
| |
| N:ne of the events was significant.
| |
| 4.1.10 Gauges / Measuring Systems H]lders of specific licenses to possess gauges are required to report failures of or damage to shielding; on/off mechanisms, or indicators of the gauge; or detection of removable contamination on the gauge. In addition, these licensees must make reports required pursuant to 10 CFR Part 20 (lost, or stolen materials, releases of material, etc.). Gauge licensees that submitted reports of events occurring in 1986 were identified by the program codes of the licensee. None of the events was by itself significant.
| |
| 4.1.11 Abnormal Occurrences In the Report to Congress on Abnomal Occurrences for the first half of 1986, fcur events at NRC licensees and five events at Agieement State licensees were determined to be abnormal occurrences (see Appendix A for listing). All of the events in the Agreement States occurred in 1985 or earlier.
| |
| The abnormal occurrences at NRC licensees involved:
| |
| The rupture of a UF-6 cylinder and subsequent release of material at Sequoyah Fuels Corporation, Sequoyah, OK, facility.
| |
| Two overexposures, one to a member of the public, and the other to a licensee employee.
| |
| Breakdcwn of management controls at an irradiator facility.
| |
| 59
| |
| | |
| With the exception of the occurrence at Sequoyah Fuels, the events were discov-ered as the result of routine inspections. The overexposures fell into the moderate range; i.e., they were less than 25 rem whole body.
| |
| l The five abnormal occurrences at Agreement State licensees involved:
| |
| I Three exposures of industrial radiographers and an assistant radiographer l (oneoftheoverex l
| |
| Report AE0D/N601)posures was discussed in the 1985 Nonreactor Event Contamination of a scrap steel facility (discussed in the 1985 Nonreactor EventReport);and Release of Kr-85 from a Trio-Tech " Tracer-Flo" system, resulting from a failure in the machines logic board. No overexposures were caused by the release.
| |
| The two radiography events in Agreement States that were not discussed in earlier NRER reports resulted in: a hand exposure of about 29,000 rem and a whole body exposure of 47 rem; and a hand exposure of about 2000 rad and a l whole body dose of 6 rad.
| |
| Three nonreactor events were reconnended to the Commission as abnormal occur-rences during the second half of 1986. The events, all of which occurred at NRC licensees, include:
| |
| i The event at Wright-Patterson Air Force Base that resulted in contamina-tion of a building when a waste drum was opened on two separate occasions.
| |
| Suspension of a license of a radiography and teletherapy servicing firm.
| |
| An order removing an officer of a radiography licensee from any assignment l or position related to NRC-licensed activities.
| |
| l The event at Wright-Patterson was reported to the NRC. The other events, not incorporated into the Nonreactor Event file, resulted from investigation of allegations (suspension of license) and observations by an NRC inspector during I
| |
| a routine inspection (removal of official).
| |
| 4.1.12 Conclusions i As in prior years, most 1986 nonreactor events concerned reports of modest I overexposures, lost or abandoned sources, or leaking sources. For these types I of events, the 1986 data do not differ substantially from the same types of r l events reported in prior years, i
| |
| The UF-6 cylinder rupture in January 1986 at Sequoyah fuels was the most i significant event reported by fuel cycle or material licensees for many years.
| |
| That event resulted in a substantial effort by NRC to review its licensing and l inspection activities in the nonreactor area, and to address the lessons taught i by the event.
| |
| 1 i
| |
| Another item of interest occurred, the contamination of a building at Wright-Patterson Air Force Base. The opening and dumping of an unlabeled waste 60
| |
| | |
| drum to identify its origin resulted in contamination of a building. Recovery costs of at least $500,000 are estimated. ;
| |
| j 4.2 Medical Misadministration Events l
| |
| i The NRC regulates certain aspects of the uses of reactor-produced radioisotopes in nuclear medicine and therapeutic radiology. Certain diagnostic and therapy aisadministrations must be regorted to the NRC in accordance with 10 CFR 35.41 ,
| |
| through 10 CFR 35.43. Diagnostic misadininistration, as used in NRC regula-
| |
| ; tions, refers to the misadministration of radioisotopes in nuclear medicine ,
| |
| l studies such as brain scans and bone scans. Therapy misadministration, as used I in NRC regulations, refers to the misadministration of radiation from cobalt-60 l
| |
| teletherapy or radioisotopes in radiation therapy. -
| |
| The significance of any event stems from the potential impact of the event on
| |
| ! public health and safety. One dimension of event risk is the frequency of the i event; a second is the magnitude of the potential impact of the event. AE00
| |
| ; has used the data collected on misadministrations for 6 years (1981-1986) to estimate error rates for certain of these misadministration events. i Regarding the frequency of events over the 6-year period, there were 25 therapy
| |
| =l misadministration reports that involved teletherapy machines. In these 25 1 events, a total of 76 patients were overtreated or undertreated. Using patient l statistics from the " Patterns of Care" study of the American College of Radiol-i ogy, the error rate per patient is estimated to be about 0.015%. For diagnos-tic misadministrations, there were about 2400 reported to NRC over this 6-year
| |
| ! period. A recent study by the Technologist Section of the Society of Nuclear l Medicine estinated that about 10 million diagnostic procedures are performed annually in the United States. In 1986, NRC regulated only 22 of the 50 i statest it is therefore estimated that about 4 million procedures are performed
| |
| , annually by NRC licensees. The diagnostic error rate per procedure is estimat-i ed to be about 0.01%.
| |
| ; Regarding the magnitude of the potential or actual impact of the event, therapy i misadministrations are associated with procedures in which large doses of .
| |
| radiation are administered to patients to achieve a therapeutic effect.
| |
| Diagnostic misadministrations are associated with procedures designed to perinit i a diagnosis to be made with little exposure to the patient. An exception is a .
| |
| diagnostic procedure known as an iodine-131 whole body scan which is discussed
| |
| . in detail below.
| |
| l Therapy misadministrations have larger potential impacts on the health of the
| |
| : patient than diagnostic misadministrations. Diagnostic misadministrations that
| |
| ! result in the erroneous administration of an iodine-131 whole body scan can
| |
| ! result in thyroid doses that are near the therapy range. Since both ,
| |
| $ teletherapy misadministrations and diagnostic misadministrations have about the l
| |
| ! same estimated error rate, the therapy misadministrations and some iodine-131 '
| |
| j misadministrations as a class appear to be individually and collectively more significant than diagnostic misadministrations. AEOD, therefore, reviews in i d2 tail therapy misadministration reports and diagnostic misadministration i reports that involve the administration of therapy amounts of radioisotopes (o.g.,1-5 millicurie of fodine-131 administered for a whole body iodine scan).
| |
| Most diagnostic misadministration reports are reviewed from a collective or statistical viewpoint.
| |
| I r
| |
| 61 l
| |
| | |
| I 4.2.1 Occurrences Reported for 1986
| |
| ! I I For this period, 369 of the approximately 2600 NRC licensees authorized to
| |
| , perform nuclear medicine studies or radiation therapy reported one or more i f
| |
| misadministrations, a total of 446 reports involving 498 patients (eight were ,
| |
| i therapypatients). Of the 446 reports of misadministrations, 438 (98%) report-
| |
| ! ed diagnostic misadministrations, and eight (1%) reported therapy
| |
| ; misadministrations.
| |
| 4.2.2 Therapy Misadministrations
| |
| ; Six therapy misadministrations were reported in 1986. Three of the misadminis- !
| |
| i trations involved teletherapy, one involved brachytherapy and two involved l
| |
| ; radiopharmaceutical therapy. The type and probable cause of the misadmints-
| |
| ; trations is shown below: ;
| |
| : Therapy Misadministrations Reported to NRC for 1986 i
| |
| ; Type of Procedure /Cause of Misadministration 1
| |
| l Teletherapy j Error in dose calculations 1
| |
| ) Wrong patient was selected for treatment 1 l Radiation administered from the wrong source 1 1 (cobalt-60 instead of linear accelerator)
| |
| ) Source head rotation switch left in rotation mode 1 l Brachytherapy i Error in dose calculations 1
| |
| ; Wrong activity sources loaded into source appifcation 1 1
| |
| Radiopharmaceutical Therapy
| |
| ?{ Misunderstanding of verbal order Failure to properly screen a patient for pregnancy 1
| |
| 1*
| |
| While one of the teletherapy misadministrations, the two brachytherapy mis-adninistrations, and one of the radiopharmaceutical therapy misadministrations were ascribed to causes previously identified in the AEOD report on therapy _
| |
| misadministrations (AEOD/C505, " Therapy Misadministrations Reported to NRC
| |
| ! Pursuant to 10 CFR 35.42"), the causes of three of the misadministrations were i
| |
| new (i.e., not previously identified as causing a misadministration).
| |
| The four new causes were: (1) the wrong patient was selected for treatment )
| |
| (2) a teletherapy treatment order usin a linear accelerator was misunderstood as an order for a cobalt-60 machinel (g) 3 the source head rotation switch on a cobalt-60 machine was inadvertently left in the rotation mode positions and (4) a patient was not properly screened for pregnancy. Although these causes are l
| |
| j *This event was reported by a licensee as a possible misadministration, but
| |
| ; does not seem to fit the criteria of the regulations. It appears that the I d
| |
| regulations did not address this type of event. It has been included because it seens to AE00 to be as significant as the other therapy misadministrations. l I 62 1
| |
| | |
| l l
| |
| l different from previously identified causes, the general conclusion drawn from i the analysis of previous misadministration reports (i.e., that the occurrence of therapy misadministrations can be reduced by improvements in licensee quality assurance procedures) appears to apply equally to preventing misadmin-1strations from the four newly identified causes.
| |
| 4.2.3 Diagnostic Misadministrations This section discusses the 1986 experience in tems of the number of diaonostic misadministrations by the types and causes of the misadministrations. Of the
| |
| , 438 total reports 337 involved the administration of the wrong radiopharma-
| |
| ; c utical to a patient and 80 involved the administration of a radiophama-
| |
| ; c utical to the wrong patient (95% of the reported misadministrations were of these two types). The remaining diagnostic misadministrations involved eight
| |
| : reports of the wrong route of administration, and 13 reports in which the i diagnostic dose of a radiopharmaceutical differed from the prescribed dose by i greater than 50%. The number of misadministrations reported for 1986 was about 13% higher than the average of the number of reports received for the last 2
| |
| ] y ars. However, because of the expected variations in the rate of occurrence of diagnostic misadministrations, we are not able to draw any conclusions from the data for the higher reporting rate. The types and causes of the diagnostic misadministrations were about the same as reported in 1985.
| |
| 1
| |
| ! Effectively, all of the causes for these types of misadministrations involved i human error. With regard to administration of the wrong radiopharmaceutical, j the data show that 34 of the 337 events (10%) resulted from receipt of misla-i beled doses from a radiophamacy. In the remainder of the events, 30 (9%)
| |
| : resulted from misinterpretation of the physician's order, and errors in the i preparation of delivery of doses accounted for 175 (52%). Another 98 reports I
| |
| (about 29%), had other causes or contained inadequate information from which to assign a cause.
| |
| The domirant causes for the wrong patient events were: the wrong patient's name on the requisition, 17 events (21%); the wrong patient was delivered to I
| |
| the nuclear medicine department,18 events (23%); a failure to correlate the i patient's identification with the study,11 events (14%); and the patient answeringtothewrongname,13 events (16%). Relatively simple quality i
| |
| assurance procedures (checking the patient's identification against the study J and the patient's medical history; asking the patient to state his name) might i reduce the frequency of these events.
| |
| j The remaining two types of diagnostic misadministrations, excess dose and wrong l
| |
| route of administration, had diverse causes.
| |
| i 4.2.3.1 lodine Misadministrations
| |
| ! Although most of the diagnostic misadministrations involved the administration i cf the wrong technetium-99m compound to a patient or the administration of the i technetium-99m compound to the wrong patient, five diagnostic misadmints-i trations involved therapy doses of iodine-131 to patients. These misadmints-l trations typically involved events in which the technologist perfomed a whole 63
| |
| | |
| body iodine scan on the patient, where the referring physician ordered a thyroid uptake or scan. The dose for whole body iodine scans is typically 1 to 5 millicuries; the typical dose for thyroid uptake study or scan is 30 microcuries of iodine-131, or 5 to 10 millicuries of technetium-99m. The
| |
| " iodine-131 whole body scan" is the only diagnostic study where this large l amount of iodine-131 is used. j The radiation dose to a patient is significantly higher if he is administered therapy doses of iodine-131 instead of the prescribed diagnostic dose of a radiopharmaceutical. For exanple, a patient prescribed a 5-millicurie dose of technetium-99m for a " thyroid scan" would receive 0.7 rad to the thyroid, whereas a 1-millicurie dose of iodine-131 can produce a dose of 4000-9000 rad to the thyroid.
| |
| During 1986, AE00 undertook an engineering evaluation of diagnostic misadmints-trations that involved administering therapy amounts of iodine-131 to patients.
| |
| Although the study had not been issued as final at the end of CY 1986, pre-liminary findings included the following:
| |
| (1) The direct causes of ten of the 14 reported iodine misadministrations (71%) were ascribed to either the physician's order being misinterpreted by or miscommunicated to the technologist (seven cases), or the techno-logist not knowing the correct dosage to administer for thyroid scan procedures that involved scanning the chest area (three cases).
| |
| (2) Causal factors associated with the occurrences of the misadministrations appeared to include:
| |
| Use of verbal orders for nuclear medicine studies.
| |
| Use of similar terms by referring physicians and licensees to refer to different procedures.
| |
| * Lack of technologist training.
| |
| * Lack of procedures.
| |
| Failure of technologist to follow procedure.
| |
| (3) Theunderlyingcauseof11of14(79%)ofthemisadministrationsappears to have been a lack of licensee control over the administration of milli-curie amounts of fodine-131 to patients. These 11 misadministrations could likely have been prevented, despite the errors that led to the misadministrations, if the prescription for the fodine-131 dosage had been verified for each patient before the iodine-131 was administered to the patient.
| |
| 4.2.4 Abnormal Occurrences Four misadministration events were included in the two Abnormal Occurrence Reports to Congress that were published for the first half of 1986; and five misadministration events have been reconnended to the Connission as abnormal occurrences for the last half of 1986 (one of these latter events occurred in 1984 and one occurred in an Agreement State). The events deemed to be suffi-ciently serious to be classified as abnonnal occurrences include five therapy misadministrations (two teletherapy, two brachytherapy, and one radiopharma-ceutical therapy) and four diagnostic misadministrations. The diagnostic 64
| |
| | |
| cisadministrations each involved the administration of a therapeutic amount of iodine to a patient.
| |
| 4.2.5 Proposed and Pending Regulatory Changes Involving Misadministrations Tha Comission's purpose in requiring the submittal of misadministration r ports to the NRC is to verify that the causes are properly identified and that licensees implement appropriate corrective actions to prevent recurrence.
| |
| If potential generic problems are identified, the Commission notifies other licensees of the generic problem or concerns and assesses the need for addi-tional actions (e.g., changes in regulations to reduce the occurrence of similar and perhaps more serious events).
| |
| In this regard, as a result of the findings in AE00 case study AE00/C505,
| |
| " Therapy Misadministrations Reported to the NRC Pursuant to 10 CFR 35.42." and th2 findings from the investigation (by Regional Offices) of several recent cisadministrations, the Commission directed NMSS to initiate rulemaking to require radiotherapy facilities to have quality assurance programs to, among other things, ensure the accuracy of patient doses. An ANPRM for this rulemaking is expected to be published in early 1986.
| |
| In addition, in late 1986 AEOD informed NMSS that AEOD was preparing to issue an engineering evaluation report that could be related to the ANPRM and pro-posed rule on quality assurance for radiotherapy facilities being developed by NMSS. The study would document (see Section 4.2.3.1) a review of diagnostic misadministrations that involved the administration of millicurie amounts of iodine-131 to patients.
| |
| 4.2.6 Conclusions Eight therapy misadministrations were reported in 1986, a rate not too differ-ent from prior years. Both the teletherapy and the brachytherapy misadminis-trations that occurred in 1986 might have been prevented by quality assurance procedures directed to verifying dose calculations, type of treatment, and patient identification.
| |
| Essentially all of the diagnostic misadministrations for 1986 involved either the administration of the wrong radiopharmaceutical or the administration of a radiopharmaceutical to the wrong patient. The type and cause of diagnostic misadministrations are about the same as reported for 1985. The causes report-i ed by licensees are generally the same as have been reported in the past; that is, simple errors associated with (1) preparation of radiopharmaceuticals, (2) processing nuclear medicine requisitions, and (3) patient identification. An additional cause for misadministrations involving the administration of milli-curie amounts of iodine to patients was the failure of licensees to exercise adequate control over the administration of millicurie amounts of iodine-131 to patients.
| |
| l l
| |
| l l
| |
| 65 ,
| |
| l
| |
| : 5.
| |
| | |
| ==SUMMARY==
| |
| OF AE0D ACTIVITIES An overall summary of AE0D reports issued in 1986 and in prior years is shown below:
| |
| Summary of AE0D Reports issued January 1,1986 through December 31, 1986 Case Studies 5 issued Special Studies 3 issued Trend and Pattern Studies 4 issued Reactor Engineering Evaluations 13 fr. sued I
| |
| Reactor Technical Reviews 12 issued Nonreactor Engineering Evaluations 2 issued IIP Reports:
| |
| - IIT Investigation Manual 1 issued
| |
| - IIT Investigation Manual Revision 1 issued
| |
| - IIT Investigation Report 2 issued Summary by Year i 1980 1981 1982 1983 1984 1985 1986 Case Studies 5 5 6 2 5 5 5 Engineering Evaluations and Technical Reviews 20 33 57 68 51 31 26 Special Study and Trends and Patterns Reports 0 0 0 1 12 7 7 l Nonreactor Engineering Evaluations and Technical l Reviews 0 8 10 14 4 3 2 IIP Reports -- -- -- -- -- 1 4 Appendix B provides listings, by year, of all AE00 reports issued from 1980 through December 1986, including date published, title, and author.
| |
| i i
| |
| 66
| |
| | |
| 5.1 Reactor Operations Analysis Branch (ROAB)
| |
| The Reactor Operations Analysis Branch (ROAB) performs the major technical AEOD studies on individual events and potential generic concerns. ROAB provides strong engineering and systems capabilities for the review of operational events involving U.S. and foreign commercial LWR plants. The operating exper-1ence of Fort St. Vrain, the only U.S. commercial gas-cooled reactor, is reviewed for ROAB under contract by Oak Ridge National Laboratory. ROAB is responsible for screening LERs and other pertinent operating information; identifying events involving particular safety significance; conducting in-depth engineering evaluations and case studies as warranted; and formulating appropriate recomendations for action by other NRC offices. The full scope of ROAB responsibilities is discussed below.
| |
| REACTOR OPERATIONS ANALYSIS BRANCH STUART D. RUBIN, CHIEF Responsibilities and Work Products 5.1.1 Systematically screen LERs and other pertinent domestic and foreign event reports and determine their significance 5.1.2 Analyze and evaluate selected individual events and potentially generic safety concerns associated with operational experience 5.1.3 Request followup NRC actions including:
| |
| -- Recommendations to address potentially significant safety concerns
| |
| -- Suggestions for feedback of important lessons learned 5.1.4 Document independent technical assessments including:
| |
| -- Case study reports (including a peer review process)
| |
| -- Special study reports
| |
| -- Engineering evaluation reports l
| |
| -- Technical review reports
| |
| -- Memoranda 5.1.5 Implement the Memorandum of Agreement with INP0 5.1.6 Provide operational experience perspective and input to related agency activities 67
| |
| | |
| l During this reporting period, a number of significant accomplishments and studies were completed. These are sumarized below for each of the {
| |
| i individual activities identified above. i 5.1.1 Systematically Screen LERs and Other Pertinent Event Reports and '
| |
| Detemine Their Significance ROAB screens each LER and other pertinent event reports to identify those events or situations that warrant additional analysis and evaluation. The ROAB screening process is controlled by means of a written office procedure.
| |
| From this process, ROAB determines whether: (1) further engineering review of the event by AE0D or others is warranted; (2) the event meets the criteria established for abnomal occurrence reporting; (3) the event meets the criteria established for reporting to the Nuclear Energy Agency (NEA) and the International Atomic Energy Agency (IAEA) Incident Reporting Systems; and/or (4) the event should be described in the bimonthly AE0D publication, Power
| |
| _ReactorEvents(seeSection5.2.4).
| |
| The fundamental objectives of the ROAB screenino process are to identify and isolate precursor events and other situations where the margin of safety has been significantly degraded, and to identify from the operational experience generic situations or concerns which may have potential safety significance.
| |
| A precursor is considered an event that could have been potentially serious if plant conditions, personnel actions, or equipment failure had been slightly different. Screening accomplishments during this reporting period include:
| |
| A total of about 3000 LERs were screened by ROAB. Of these, approximately 150 LERs were judged to be significant, warranting additional NRC attention.
| |
| Of these significant LERs, about 7% were related to generic issues or to matters being actively pursued by NRR, IR, or RES. The remainder (about 40 events) were designated by ROAB for further analysis and evaluation by AE00.
| |
| In addition to U.S. experience, ROAB routinely reviews foreign operating experience reports. During this reporting period, 61 NEA/ IRS reports and 31 IAEA reports were screened. Additionally, approximately 100 foreign reports obtained through bilateral agreements were reviewed.
| |
| Other U.S. operational information reviewed by ROAB during this reporting period included approximately 2500 regional inspection reports, 350 Part 21 re,norts, 4200 licensee 50.72 report sumaries, 470 preliminary notification reports, and 50 INPO documents.
| |
| 5.1.2 Analyze and Evaluate Selected Individual Events and Potentially Generic Safety Concerns During this reporting period, ROAB completed one preliminary and four final case studies, one special report, 13 engineering evaluations, and 12 technical review reports. Representative studies are discussed below, a complete listing of completed studies is provided in Section 5.1.4. (Section 5.1.3 provides examples of follow-up NRC actions recomended by ROAB in the case studies discussedbelowandinselectedengineeringevaluations.)
| |
| The AE00 preliminary case study on " Air Systems Problems at U.S. Light Water Peactors" was issued for peer review in December 1986. The report 68
| |
| | |
| i l
| |
| l provides a detailed, comprehensive and systematic analysis and evaluation of the potential safety implications associated with air system problems reported by domestic LWR licensees. The analysis of the operational data focuses on degraded air systems and the attendant vulnerability of safety-I related equipment to common mode failures caused by air system degrada-tions. The report analyzes the data for trends and patterns of the underlying causes and consequences, and evaluates the safety importance of the problem in terms of both risk assessments, and cost / benefit studies.
| |
| The major finding was that the root cause of most air systems problems is traceable to either design or management deficiencies. The deficiencies appear to reflect a lack of sufficient regulatory requirements and review and the view by many applicants and licensees that air systems are not highly important to plant safety. The study presents several preliminary recommendations to reduce reactor accident risk, enhance safety and improve plant performance, including: (1) ensuring that air systems quality meets the requirements plants' air-operated equipment; sp(ecified by the manufacturers of the2) ensuring fonnulating and implementing anticipated transient and system recovery procedures for loss-of-air events; (3) improving training programs to ensure that plant ope.ations and maintenance personnel are sufficiently sensitized to the importance of air systems and the potential vulnera-bility of safety-related equipment served by the air systems to conunon mode failures; (4) confirming the adequacy and reliability of safety-related backup air accumulators; and (5) verifying predicted equipment i
| |
| response to gradual losses of air by testing to ensure that such losses do not result in events which fall outside FSAR analysis assumptions.
| |
| Following peer review, the final case study was issued in March 1987.
| |
| Case study AE00/C605, " Operational Experience Involving Losses of Electri-cal Inverters," was issued in December 1986. The study analyzed opera-tional experiences involving losses of electrical inverters. Analysis of the data indicates that inverter loss events may be attributed to two causes: component failures and personnel actions. A major contributing factor and/or cause for inverter loss events was found to be incompati-bility between actual plant service conditions and design service con-ditions for inverters. Contributing factors and/or causes for inverter losses were due to incorrect personnel actions which were traced to inadequacies in planning for maintenance and testing activities, and deficiencies in personnel training. The case study presented several recommendations to improve the operational performance reliability of inverters and related areas, including: issuance of an IE information notice focusing on the identified causes for occurrences of inverter l loss events attributed to component failures and personnel actions; review of the technical specifications action statements addressing inverters and/or attendant bases for comparable plant designs to ensure consistency; and reassessment of a portion of the electrical circuitry associated with the Westinghouse Solid State Protection System. The report was forwarded to IE and NRR for review and appropriate followup regulatory or industry actions.
| |
| Case study AE00/C604, " Effects of Ambient Temperature on Electronic Components in Safety-Related Instrumentation and Control Systems," was issued in December 1986. The report documents the analysis and evaluation 69 I
| |
| | |
| of four events involving the failures of solid state electronic components in safety-related instrument and control systems due to overheating caused by inadequate cabinet cooling. These events occurred at the McGuire 1 Davis-Besse, Summer 1, and Palo Verde I nuclear plants. Overheating of electronic components was found to raise two concerns: (1) decreased reliability of electronic components due to increased component failure rate, and (2) potential common mode failure of redundant safety-related instruments due to extended loss of normal cooling air flow to the in-strument cabinets. Based on the study findings and conclusions, rec-ommendations to address the significant safety concerns are presented in the report. The study recommendations include the need to: (1) establish emergency procedures and operator training to cope with a loss of cooling 1 to instrument cabinets; (2) monitor actual local temperature conditions in i
| |
| the instrument cabinets; (3) provide technical specification requirements for operability of the control room cooling and ventilation systems which reflect the actual local temperature in the instrument cabinets; and (4) consider this issue in the plant-specific evaluation of the station l blackout issue (USI A-44). The report was forwarded to NRR for review and appropriate followup action.
| |
| Case study AE00/C602, " Operational Experience Involving Turbine Overspeed Trips," was issued in August 1986. The study reviewed and evaluated past operating experience involving overspeed trips of PWR auxiliary feedwater (AFW) turbine-driven pumps. The study found that the causes of AFW turbine overspeed trips are dominated by speed control problems associated
| |
| # with governors, and trip and reset problems associated with trip valves and overspeed trip mechanisms. These problems result primarily from inadequate performance by plant personnel, inadequate procedures, and insufficient system design considerations. The governor speed control problems are caused by: (1) slow response of the Woodward Model PG-PL governor, (2) entrapped oil in the speed setting cylinder of the Woodward Model PG-PL governor, (3) incorrect governor setting, and (4) water induction into the turbine. The trip and reset problems were found to stem i
| |
| from the complexity of reset operations and a lack of trip position indication. Recommendations for preventing or reducing the overspeed trip problems were also presented in the study. The report recommended issu-ance of an IE information notice to feed back the lessons learned in the study to all LWR licensees, and consideration of specific additional recommendations in the ongoing NRR programs for improving AFW system reliability (particularly in response to action items directed by the i
| |
| EDO in connection with the NRC staff investigations of the June 9,1985 event at Davis-Besse). Following issuance of the case study report IE 1
| |
| issued Information Notice 86-14, Supplement 1, "Overspeed Trips of AFW, HPCI and RCIC Turbines," on December 17, 1986. Additionally, NRR agreed to consider the study recommendations in the resolution of Generic Issues 122 and 124.
| |
| The final AE00 case study on "A Review of Motor-Operated Valve Perfor-mance"wasissuedinDecember1986(AEOD/C603). The study reviewed nuclear plant operating experience involving motor-operated valves in all safety-related systems from 1978 through mid-1985. The study found that: (1) valve assembly failure to operate continues to be a safety concern with little improvement in performance or reliability over the study period, and (2) current programs and procedures at many plants need 70
| |
| | |
| l improvement to assure that motor-operated valve assemblies will operate when needed (e.g., under credible accident conditions). The study report
| |
| ! presented several recommendations, including: reiteration of the rec-ommendations presented in previous AEOD reports on the subject of motor-operated valves (C203 and S503); development of procedures and diagnostic capability for determining the root cause of valve assembly failures; i development of training programs for plant personnel; and expansion of the scope of IE Bulletin 85-03 to all safety-related motor-operated valve assemblies. Following issuance of the case study report, it was trans-mitted to the Nuclear Utilities Management and Resources Comittee (NUMARC) with the request that they take the lead in developing and implementing an industry program to address the findings, conclusions, and recomendations presented in the study.
| |
| In addition to the above case studies, AE0D issued a number of other studies which identified potentially significant safety concerns. Some representative oxamples are as follows:
| |
| Engineering evaluation AE0D/E606, " Loss of Safety Injection Capability at Indian Point Unit 2," was issued in May 1986. The report evaluated a December 28, 1984 event in which all three safety injection pumps were inoperable. The malfunction of the safety injection pumps was caused by boric acid crystallization and possibly gas binding. Although the issue of boric acid crystallization is not viewed as a broad generic problem, the report identified additional plants that have a similar design which I
| |
| could also be prone to similar failures. Subse licensee removed the boron injection tank (BIT)quent which wastothe thesource event,ofthe boron crystallization. Other similar plants have also removed the BIT and/or obtained approval from NRR to operate with reduced boric acid i levels to prevent similar malfunctions. The report's findings were discussed in IE Information Notice 86-63, " Loss of Safety Injection Capability."
| |
| Engineering evaluation AE0D/E609, " Inadvertent Draining of Reactor Vessel During Shutdown Cooling Operation," was issued in August 1986. The report analyzed and evaluated 11 operational eveats which occurred at nine boiling water reactors over a period of 4 years. The operational events were primarily caused by human errors associated with the operation of the residual heat removal (RHR) system in the shutdown cooling mode. The cause of these human errors can be traced to deficient procedures, im-proper or inadvertent actions, lack of knowledge or training, man / machine interface problems, cognitive errors, or maintenance errors. The need for manual operation of the RHR shutdown cooling valves, the elevational differences and interconnections between the RHR subsystems, and the absence of comprehensive valve interlocks also contributed to the occur-rence of these operational events. In light of the frequency and sig-nificance of these events, the report suggested several relatively low cost measures which could be implemented to prevent their recurrence. The study suggestions are consistent with the program elements stated in I the NRC Human Factors Program Plan in the training, procedures, man- !
| |
| machine interface, and human performance areas. The study was transmitted J to NRR for review, and suggested followup action within their ongoing i human factors program. I i i 71
| |
| | |
| Technical review AE00/T605, " Failure of Main Steam Safety Valves (MSSVs) to Properly Reseat," was issued in June 1986. The review of an event that occurred at Oconee with other operating experience indicates that the potential for MSSVs failing to reseat and initiating an overcooling transient is higher for Babcock & Wilcox designed plants than other PWRs.
| |
| However, the likelihood of such an overcooling event challenging the structural integrity of the pressure vessel was found to be small. The study concluded that the industry evaluation of ways to reduce the number of challenges to MSSVs and improve the reliability of the valves was an adequate response to these concerns.
| |
| Engineering evaluation AE0D/E602, " Unexpected Criticality Due to Incorrect Calculation and Failure to Follow Procedures," was issued in January 1986.
| |
| The evaluation addressed an event at Virgil C. Summer Unit I that occurred on February 28, 1985. A reactor trip occurred on a high positive flux rate trip during a reactor startup. The event was attributed to a number of causes. First, the licensed operator conducting the startup failed to adhere to applicable procedures in that criticality was not anticipated during control rod bank withdrawal, and an awareness of plant conditions was not maintained at all times. The second cause which contributed to the event was a lack of adequate guidance in the procedures used to calculate the estimated critical rod position and reference critical data.
| |
| Finally, the last cause identified which could have contributed to the event was procedural inadequacy in the licensee's administration of the plant's on-the-job training program. Although the event was bounded by accident analyses and adequate core protection was maintained, a review of related operating experience indicates that licensee operator training programs may need to be reviewed in order to minimize the potential for premature criticality events.
| |
| Engineering evaluation AE0D/E603, " Delayed Access to Safety-Related Areas During Plant Operation," was issued in February 1986. The study reviewed several events of delayed access to vital areas at nuclear power plants caused by security, radiological protection or administrative provisions in order to determine their underlying causes and impact on plant opera-tional safety. None of the events occurred during an actual plant emergency and therefore, no imediate safety concerns were involved. Most of the events were also of sufficiently short duration that had they occurred during an actual emergency, timely operator actions likely could have been taken if needed inside the affected cornpartnent to avoid a significant degradation of plant safety margins. However, an event at Limerick involved delayed local operator actions during a remote shutdown demonstration test because plant procedures did not include provisions for having a set of equipment or compartment keys available t.t the remote shutdown panel. This procedural deficiency has been addressed by improve-ments implemented in the planning and procedures for remote shutdown operations. The engineering evaluation report suggested issuance of an IE information notice to remind licensees of the need for adequate key availability during remote shutdown operations. Following issuance of the report, IE Information Notice 86-55, " Delayed Access to Safety-Related Areas and Equipment During Plant Emergency," was issued.
| |
| Technical review report AEOD/T606, " Inadvertent Recirculation Actuation Signals at Combustion Engineering Plants," was issued in August 1986. The 72
| |
| | |
| study was prompted by the different system responses to an inadvertent recirculation actuation signal (RAS) at Arkansas Nuclear One Unit 2 and Calvert Cliffs Unit 2. The study found that at CE plants without anti-draindown check valves in the emergency core cooling system pump suction lines, a significant inventory of borated water could transfer from the reactor water tank (RWT) to the containment sump on an inadvertent RAS.
| |
| l Antidraindown check valves prevented a similar loss of RWT inventory at
| |
| ! the other plant design. Because plant responses were so different, the safety aspects of a postulated safety injection actuation signal subse-quent to an inadvertent RAS were analyzed for each plant design. The analysis showed that, for either plant design, adequate suction flow would be provided to the ECC pumps in event of coincident RAS and LOCA signal.
| |
| Engineering evaluation AE0D/E604, " Spurious System Isolations Caused by the Panalarm Model 86 Thermocouple Monitor," was issued in March 1986 The report evaluated a number of high pressure coolant injection, reactor core isolation cooling, and reactor water cleanup system isolations caused by spurious trips of a Panalarm Model 86 thermocouple monitor. The study found that the elevated sensitivity of the Panalarm thermocouple monitor makes the instrument highly susceptible to spurious trips caused by momentary electrical disturbances. The spurious system isolations caused by the instrument were found to have potentially adverse impacts on system reliability, isolation valve operability, and the distractions presented to the plant operating personnel. The study found that a modification to the leak detection system trip circuitry at Duane Arnold appears to have proven successful in significantly reducing spurious system isolations caused by the thermocouple monitor. The study suggested that other licensees be informed of the spurious system isolations caused by the Model 86 thermocouple monitor and the modification implemented at Duane Arnold which appeared to an effective corrective action. In accordance with the study suggestion, IE issued Information Notice 86-69, " Spurious System Isolations caused by the Panalarm Model 86 Thermocouple Monitor,"
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| on August 18, 1986.
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| Engineering evaluation .AE0D/E610, " Loss of LPCI Loop Selection Logic at Millstone-1," was issued in August 1986. The report evaluated an event at Millstone-1 in which three Model 288 Barton differential pressure (D/P) switches in the low pressure coolant injection (LPCI) loop selection logic were found in the tripped condition. Under certain pump flow conditions, improper operation of these switches could result in the LPCI system i injecting into the broken recirculation loop following a postulated design l basis loss-of-coolant accident. The study found that the seven BWRs still equipped with LPCI loop selection logic utilize this type of switch. The switch is also used in a number of safety-related applications at various other LWRs. The study concluded that Model 288 Barton D/P switches are ,
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| susceptible to setpoint drift and this susceptibility appears to increase as the switches age. The study also concluded that other BWRs could be
| |
| . vulnerable to this type of LPCI system degradation due to the suscepti-bility of the switches .to setpoint drift. Suggestions were made to issue an IE information notice discussing the Millstone-1 event and the suscept-ibility of the Barton switch to setpoint drift. The study also suggested that IE evaluate the adequacy of the manufacturer's surveillance and maintenance recommendations.
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| 73
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| Engineering evaluation AE00/E611, " Seismic Anchors for Electrical and Control Panels," was issued in October 1986. The report evaluated recent LERs involving inadequacies in the seismic anchorage of safety-related electrical and control system equipment at older operating plants. The report details several instances of design and installation deficiencies found in the seismic anchorage of electrical and control panels at the Dresden and Quad Cities nuclear power plants. The report noted that these deficiencies represented a failure to properly implement seismic design requirements in contrast to inadequacies in the seismic requirements themselves, which was the principal concern of USI A-46. Even so, the study provides considerable support for the proposed resolution of USI A-46, " Seismic Qualification of Equipment in Operating Nuclear Power Plants." The principal suggestion made in the report was to inform those plants not included in the resolution of USI A-46 of the types of design and installation errors identified in the study. Following issuance of the report, a generic letter on USI A-46 was issued which referenced the events reviewed and the concerns raised in the engineering evaluation.
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| Engineering evaluation AE0D/E607, " Degradation or Loss of Charging Systems with Swing Pump Designs," was issued in July 1986. The report identified situations involving functional degradation or loss of charging systems using swing pump designs. The degradations or system losses were due to design deficiencies in interlock circuitry or inadequate maintenance procedures. In view of these deficiencies the study concluded that these or similar problems could exist at other plants. As a result of this study IE Information Notice 86-79, " Degradation or Loss of Charging Systems at PWR Nuclear Power Plants Using Swing-Pump Designs," was issued in July 1986.
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| Engineering evaluation AE0D/E605, " Lightning Events at Nuclear Power Plants," was issued in April 1986. The report W s based on the review of 62 event reports involving lightning strikes at 9.S. auclear plants in the period from 1981 to 1985. The events occurred at 30 plant sites and involved 32 reactor units. A comparison of the number of lightning events, the geographic location of the affected units and the annual lightning strike density at the location, found a direct correlation between the annual lightning strike density and the number of events reported. The reported systems affected were: (1) the offsite power system, (2) the safety-related instrumentation and control system, (3) the meteorological and weather systems, (4) the radiation, gas and effluent flow monitoring systems, and (5) the air intake tunnel halon system. The report concluded that although lightning strikes have adversely affected the operation of some nuclear plants, in most cases there has been no significant degradation of safety systems and minimal equipment damage.
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| In particular cases where damage has been extensive or where failures caused by lightning strikes have been repetitive, the licensees have taken corrective actions to reduce the consequences of future strikes. Since IE Information Notice 85-86 had been issued on the subiect of lightning strikes, the report suggested that no further actions be taken.
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| ! 74 i
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| | |
| Technical review AE0D/T608, " Hydrogen Fire and Failure of Detection System," was issued in November 1986. The report documents the review of an event at a foreign reactor which involved a leak in the unit's electric generator hydrogen system followed by an explosion and a fire. Due to a !
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| design deficiency in the fire detection system that allowed too many l detectors to be installed in a detection loop, the fire detection system did not respond correctly during the event--instead of annunciating a fire, it indicated an "out of order" condition. The fire was detected promptly in spite of the detection system problem, and the plant was safely shut down. The review was initiated to assess whether such a fire detection system problem could exist at U.S. plants, since such a design deficiency could potentially prevent the detection system from performing its design function when needed. Based on a review of recent operational events in domestic plants involving fire protection and detection systems, such a design deficiency was not identified in U.S. nuclear plants.
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| Although a review of detailed design drawings of fire detection systems installed in specific U.S. plants was not performed, the report concluded that a large fire, that renders the fire detection system incapable of functioning properly, remaining undetected is a low possibility.
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| Technical review report AE0D/T612, " Degradation of Safety Systems Due to Component Misalignment and/or Mispositioned Control / Selector Switches,"
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| was issued in December 1986. The report addressed in detail three events at D.C. Cook and one event at Kewaunee where the ability of a safety system to automatically perform its designed function was compromised due to human error. Additionally, 55 similar incidents which occurred at U.S.
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| operating plants between 1981 and 1985 were examined. Nearly 91% of these 55 events occurred at PWRs and the rest at BWRs. The systems most affect-ed in these events were the safety injection system, the residual heat removal system and the containment spray system. The events at D.C. Cook and Kewaunee and other similar occurrences reviewed illustrate that human error can result in the degradation or complete loss of a safety system.
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| These events emphasize the need for appropriate procedural controls, ade-quate comunications between plant personnel, and independent verification of system lineup after maintenance or surveillance activities.
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| Engineering evaluation AE0D/E612 " Emergency Diesel Generator Component Failures Due to Vibration," was issued in December 1986. The report evaluated resulted ineight events that inoperability involved of an cracking of small onsite emergency dieselbore piping (thatEDG).
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| generator The cracked lines were found in the EDG lube oil, fuel oil and cooling water systems. The piping cracks, which were caused by cyclic fatigue that resulted from engine-induced vibration, were not detected by either the inservice inspection or the preoperational testing program for the piping. The failures were discovered after the cracks propagated com-pletely through the tube wall and flufd was observed leaking from the pipes. The study found that fatigue failures induced by steady-state operation, such as plant equipment or engine-induced vibration, are not normally analyzed in the original. piping design. A more complete design review is typically done only after related problems are identified during plant operation. Often the problem is not detected until a leak occurs.
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| Such leaks could result in the sudden disabling of the EDGs when needed 75
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| and could adversely affect a safe plant shutdown following of a loss of offsite power. The report suggested that IE issue an information notice to address the potential safety concern of fatigue cracks in those small ,
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| bore lines associated with EDGs.
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| Engineering evaluation AE0D/E613, " Localized Rod Cluster Control Assembly (RCCA) Wear at PWR Plants," was issued in December 1986. The report found that RCCA wear should be anticipated in Westinghouse designed PWRs. The degradation was related to: (1) wear during rod motion associated with startup, shutdown, reactor trip, and load following; (2) flow induced vibration leading to fretting wear between the rodlet and the guide card when RCCAs are at fixed positions for long periods of time; and (3) rodlet cracking (intergranular stress corrosion cracking) that appears related to absorber / clad interaction. The concerns raised by this study were for-warded to IE with a request for an IE information notice and to NRR with a request to review whether periodic inspections by licensees were needed.
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| Engineering evaluation AE00/E601, " Deficient Operator Actions Following Dual Functions Valve Failures," was issued in September 1986. The study evaluated events involving dual function valve failures at Dresden 3, Brunswick 1 and 2, and Peach Bottom 3. The study found that LWRs gener-ally are equipped with a number of valves which perform both an emergency core cooling (or safety-related) function and a containment isolation function. However, operating experience shows that the proper and con-servative operator action for a failure of one of these valves has not always been taken by the operating staff in a manner which is fully consistent with plant technical specifications. In each of the events reviewed, the operator's actions preserved one function of the valve but disabled the valve's alternate function. Furthermore, the operator's actions were taken without declaring the adversely affected function of the valve to be inoperable as required by the plant technical specifi-cations. At Dresden 3, operability of an ECCS loop was preserved, but containment isolation capability was disabled. At Brunswick and Peach Bottom 3, containment integrity was maintained, but a safeguards train was disabled. The study also found that the affected valves were not included in the plant technical specification table of containment isolation valves. As a result, IE Information Notice 86-38, " Deficient Operator Actions Following Dual Function Valve Failures," was issued, and the Technical Specification Improvement Program adopted a suggestion to highlight these dual function valves in all plant technical specifications.
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| Engineering evaluation AE0D/E604, " Spurious System Isolations Caused by the Panalarm Model 86 Thermocouple Monitor," was issued in March 1986.
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| The report evaluated a number of high pressure coolant injection, reactor core isolation cooling, and reactor water cleanup system isolations caused by a spurious trip of a Panalam Model 86 thermocouple monitor. The study found that the elevated sensitivity of the Panalarm thermocouple monitor makes the instrument highly susceptible to spurious trips caused by mcmentary electrical disturbances. The spurious system isolations caused by the instrument are undesirable because of the potentially adverse impacts on system reliability, isolation valve operability, and the distractions presented to the plant operating personnel. The study found 76
| |
| | |
| that a modification to the leak detection system trip circuitry at Duane Arnold appears to have proven successful in significantly reducing spur-ious system isolations caused by the themocouple monitor. , The study recommended that other licensees be informed of the spurious system isolations caused by the Model 86 thermocouple monitor and describes the modification implemented at Duane Arnold as a possible corrective action.
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| IE issued Information Notice 86-69, " Spurious System Isolations caused by the Panalarm flodel 86 Thermocouple Monitor," on August 18, 1986.
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| In addition, the following special study report was performed and issued by ROAB during this reporting period:
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| Special study AE0D/S603, " Adequacy of the Scope of Bulletin 86-01," was issued in June 1986. This study evaluated a potential inadequacy in the scope of IE Bulletin 86-01, " Minimum Flow Logic Problems That Could Disable RHR Pumps," which had been issued in May 1986. Bulletin 86-01 was addressed to only BWR facilities. The AE0D study included information which strongly indicated that a problem similar to the BWR residual heat removal (RHR) minimum flow control logic single failure vulnerability identified at the Pilgrim station (the topic of the IE Bulletin) might also exist at a number of Westinghouse-designed PWRs. Documents from Wisconsin Electric and Carolina Power and Light described almost identical and analogous design deficiencies at Point Beach and H. B. Robinson. The deficiency at both plants involved a single failure that could cause the loss of the common minimum flow recirculation path for all of the safety injection (SI) pumps. The deficiency at Point Beach had also been the subject of a Part 21 report. Due to the potential safety significance which could be associated with a failure of the SI system at Westinghouse PWRs and the timeliness of IE Bulletin 86-01, the memorandum recomended that IE give imediate consideration to issuing either a supplemental or a separate IE bulletin to all Westinghouse PWR facilities concerning the SI recirculation path design deficiencies identified at Point Beach and H. B. Robinson. Subsequently, on October 8, 1986, IE issued Compliance Bulletin 86-03, " Potential Failure of Multiple ECCS Pumps due to Single Failure of Air-0perated Valve in Minimum Flow Recirculation Line."
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| 5.1.3 Request for Followup NRC Actions The reports issued by ROAB during this reporting period included a number of formal recommendations for NRC program office followup actions related to potentially significant safety concerns or suggestions for feedback of l important lessons learned from plant operating experiences. In addition, during this reporting period, a number of actions were initiated or completed by one or more NRC program offices as a result of recommendations or suggestions previously made by ROAB. Selected examples of recommendations or t suggestions made by ROAB during this period, or actions taken by others as a l result of earlier requests, are presented below (see Section 6 for a complete status report on all AE0D recommendations that were outstanding during 1986).
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| Engineering evaluation AE0D/E602 was forwarded to NRR, Region II, and Region IV for information and review. As a result of this review, NRR performed a special evaluation of the operator licensing examinations conducted at Surry, and subsequently upgraded future examinations to place more emphasis on reactor start procedures.
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| 77
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| Engineering evaluation AE0D/E606 suggested that an IE information notice be issued to make the licensees of potentially affected plants more aware of the safety concerns which were raised by the Indian Point Unit-2 loss of all safety injection pumps due to boric acid crystallization. It was suggested that licensees of potentially affected plants be made aware of the safety concerns of boric acid crystallization, and emphasize the need to follow strict administrative controls to prevent similar occurrences until their boron injection tanks are removed. In response, IE Informa-tion Notice 86-63, " Loss of Safety Injection Capability," was issued.
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| Engineering evaluation AE0D/E609 suggested that several effective and relatively low cost measures, developed by the study to reduce the like- l lihood of recurrence of inadvertent draining of the reactor vessel, be !
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| considered by NRR as part of their ongoing systematic assessment of human factors concerns. As a result, NRR agreed to give consideration to these measures as part of their ongoing evaluations in the human factors area.
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| Engineering evaluation AE0D/E603 suggested that an IE information notice be issued on an event at Limerick where plant procedures did not include provisions for having a set of equipment or compartment keys available at the remote shutdown panel. As a result, IE Information Notice 86-55,
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| " Delayed Access to Safety-Related Areas and Equipment During Plant Emergency," was issued on this subject.
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| Engineering evaluation AE0D/E601 suggested that an IE information notice be issued on several events involving the failure of dual function valves.
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| In each of the events reviewed, the operator's actions preserved one function of the valve, but disabled the valve's alternate function. As a result, IE Information Notice 86-38, " Deficient Operator Actions Following Dual Valve Failures," was issued and the technical specification improve-ment program included the suggestion that valves which perfonn both an ECCS and containment isolation function be highlighted in all plant technical specifications.
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| Engineering evaluation AE0D/E604 suggested that an IE information notice be issued on operating events involving high pressure coolant injection, reactor core isolation cooling, and reactor water cleanup system isola-tions caused by spurious trips of Panalarm Model 86 thermocouple monitors.
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| As a result, IE Information Notice 86-69, " Spurious System Isolation Caused by the Panalarm Model 86 thermocouple Monitor," was issued.
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| Engineering evaluation AE00/E610 suggested that an information notice be issued that describes the loss of LPCI selection logic which occurred at Millstone-1 and the susceptibility of the Barton Model 288 differential pressure (D/P) switches to setpoint drift. It was also suggested that the IE Vendor Inspection Branch evaluate the adequacy of the manufacturer's recommended surveillance and maintenance with regard to the intended usage of the Model 288 D/P switch in the LPCI loop selection logic. The Vendor Inspection Branch is investigating these concerns and IE will issue an information notice pending the results of its investigation. NRR plans to include information on the Barton Model 288 as part of its generic letter on differential switches.
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| 78
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| l Engineering evaluation AE0D/E611 suggested: (1) issuance of an updated IE information notice to characterize the continuing deficiencies in seismic anchorage of critical safety equipment; (2) reconsideration of the limi-tation of the application of the resolution of USI A-46 to 70 older plants since the deficiencies identified in the LERs and in recent NRC inspec-tions are caused by construction / quality assurance deficiencies and not
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| ! design; and (3) preliminary walk-through reinspections to identify gross l deficiencies, as indicated by the LERs, before the detailed reinspections I and engineering analyses are conducted. In response to these suggestions, l a generic letter for plants not included in the USI A-46 program was revised to address the study findings and. conclusions.
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| Special study AE0D/S603 alerted IE to design deficiencies identified at several Westinghouse-designed PWRs involving the potential loss of all safety injection pumps due to a single failure in the minimum flow recirc-ulation path. As a result, IE issued IE Compliance Bulletin No. 86-03,
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| " Potential Failure of Multiple ECCS Pumps due to Single Failure of Air-Operated Valve in Minimum Flow Recirculation Line."
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| Case study AE0D/C602 contained several recommended actions aimed at preventing or reducing the unwanted turbine overspeed trips identified in past operational events. Specifically, the report recommended ttat:
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| (1) an IE information notice be issued to address the study findings; (2) measures be taken to alleviate the effect of slow response of the governor valve; (3) administrative controls be implemented or an addi-tioral device be added to prevent hydraulic fluid from entrapping in the speed setting cylinder; (4) additional review take place for the existing vendor supplied calibration proceduret for governor speed setting to assure adequacy; (5) adequate provisions be provided for condensate removal from the steam sunply line to prevent water induction into the turbine during startup; (G) the adequacy of the existing procedural instruction be assessed; (7) training programs be upgraded; and (8) posi-tion indication to minimize trip and reset problems be provided. In response to these recommendations, IE Information Notice 86-14, Supplement 1, "Overspeed Trips of AFW, HPCI and RCIC Turbines," was issued. Addi-tionally, NRR agreed to consider the study recommendations in the reso-lutions of Generic Issues 122 and 124.
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| Engineering evaluation AE0D/E607 suggested that an infonnation notice be issued for the events, causes, and possible corrective actions associated with functional degradation or loss of charging systems using swing pump design. In response to this request, IE Information Notice 86-79, "Degra-dation of Loss of Charging Systems at PWR Nuclear Power Plants Using Swing-Pump Designs," was issued.
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| Case study AE0D/C603 contained a number of. recomendations associated with the need for high priority licensee action in several areas related to motor operated valve performance. The ED0 subsequently sent a request to NUMARC to assume the lead in a concerted industry effort to address the AE0D recomendations.
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| 79
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| 5.1.4 Document Independent Technical Assessments During 1986, the following studies were completed by ROAB personnel:
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| Reactor Case Study Reports - 1986 No. Subject Date Issued Author C602 Operational Experience Involving Turbine 8/20/86 C. Hsu
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| . Overspeed Trips C603 A Review of Motor-0perated Valve 12/5/86 E. Brown Performance C604 Effects of Ambient Temperature on 12/30/86 M. Chiramal Electronic Components in Safety-Related Instrumentation and Control Systems 1
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| C605 Operational Experience Involving Losses 12/31/86 F. Ashe of Electrical Inverters Special Study Reports - 1986 No. Subject Date Issued Author S603 Adequacy of the Scope of IE Bulletin 6/17/86 E. Leeds 86-01 Reactor Engineering Evaluation Reports - 1986 No. Subject Date Issued Author E601 Deficient Operator Actions Following Dual 1/9/86 E. Leeds Function Valve Failures E602 Unexpected Criticality Due to Incorrect 1/15/86 R. Freeman Calculation and Failure to Follow Procedures E603 Delayed Access to Safety-Related Areas 1/19/86 T. Cintula During Plant Operation E604 Spurious System Isolations Caused by the 3/14/86 E. Leeds Panalarm Model 86 Thermocouple Monitor 80
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| No. Subject Date Issued Author E605 Lightning Events at Nuclear Power Plants 4/28/86 M. Chiramal E606 Loss of Safety Injection Capability at 5/27/86 R. Tripathi Indian Point Unit 2 1 E514 Core Damage Precursor Event at Trojan 5/27/86 D. Zukor l (Revision 1)
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| E607 Degradation of Loss of Charging Systems 7/3/86 F. Ashe with Swing Pump Designs E608 Re-examination of Water Hamer Occurrences 7/14/86 E. Leeds E609 Inadvertent Draining of Reactor Vessel 8/8/86 P. Lam During Shutdown Cooling Operation E610 Loss of LPCI Loop Selection Logic at 8/14/86 E. Leeds Millstone-1 E611 Deficiencies in Seismic Anchorage for 10/16/86 N. Thomasson Electrical and Control Panels E612 Emergency Diesel Generator Component 12/17/86 C. Hsu Failures Due to Vibration E613 Localized Rod Cluster Control Assembly 12/23/86 E. Brown (RCCA) Wear at PWR Plants Reactor Technical Review Reports - 1986 No. Subject Date Issued Author T601 Pressure Sensitive Temperature Switch 2/27/86 T. Cintula Results in Spurious Actuation of Fire Suppression System T602 Emergency Diesel Generator Cooling Water 4/29/86 E. Leeds system Design Deficiencies at Maine Yankee and Haddam Neck i
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| T603 Inadvertent Pump Suction Transfer and 5/6/86 R. Tripathi Potential Auxiliary Feedwater Pump Cavitation at Davis-Besse T604 Events Resulting from Deficiencies in 5/7/86 E. Trager Labeling and Identification Systems T605 Failure of Main Steam Safety Valves 6/17/86 R. Freeman (MSSVs) to Properly Reseat 81
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| No. Subject Date Issued Author T606 Inadvertent Recirculation Acutation 8/7/86 T. Cintula Signals at Combustion Engineering Plants 1 T607 Occurrence of Events Involving Wrong 9/19/86 E. Trager Units / Wrong Train / Wrong Component - Update Through June 1986 i T608 Hydrogen Fire and Failure of Detection 11/12/86 M. Chiramal System T609 Foreign Material and Debris in Safety- 12/16/86 E. Leeds
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| ; Related Fluid Systems T610 ADS /RCIC System Interaction Events at 12/19/86 E. Leeds River Bend-1 l T611 Denied Access due to Negative Room 12/19/86 T. Cintula l
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| Pressure T612 Degradation of Safety Systems Due to- 12/31/86 R. Tripathi Component Misalignment and/or Mispositioned
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| , Control / Selector Switches j 5.1.5 Implement the Memorandum of Agreement with INP0 i AE00 and INP0 share results of completed studies related to review and assess-ment of operational data. Additionally, listings are periodically exchanged i for studies planned or in progress. Periodic informal meetings are held to
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| ! discuss concerns of mutual interest. During 1986, two such meetings were held l (May 29 and November 19,1986). In addition to these meetings, considerable i
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| informal discussion occurs between the ROAB staff and their INPO counterparts.
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| 5.1.6 Provide Operational Experience Perspectives, Input, and Support to Related Agency Activities 1
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| ROAB is often requested to provide comments, perspectives and support on a number of technical issues and concerns and to serve on various task forces or teams. Such assignments are authorized when it is believed that AE00 can be responsive and the request offers a desirable method of communicating the lessons learned from operating experience and related AE00 studies. Examples of this type of activity during this reporting period, included:
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| Participating in NRC's review of the B&W Owner's Group Reassessment Program 1
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| l Providing input to NRR on the resolution of Generic Issue 99 "RCS/RHR Suction Line Interlocks" l Participating in the third regular bilateral NRC/MITI meeting on nuclear safety 82 i
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| * Presenting a paper at the International ANS/ ENS Topical Meeting on Thermal Reactor Safety (February 1986)
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| Submitting a paper for publication at the ANS Topical Meeting on Nuclear Power Plant Maintenance (March 1986)
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| Assisting the IIT team on the Rancho Seco investigation Participating in the Augmented Inspection Team at the Pilgrim Nuclear Power Station
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| * Participating in the IAEA consultants' meeting on " IRS Generic Safety Issues Identification," and preparing a technical report on the subject Providing input to Commissioner's site visit briefings Providing input to NRR on the resolution of Generic Issue 51, " Proposed Requirements for Improving the Reliability of Open Cycle Service Water Systems" Participating in a team inspection of San Onofre Units 2 and 3 in connection with the ability of these units to safely respond to events that have occurred at other facilities.
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| 5.2 Program Technology Branch (PTB)
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| PTB is responsible for a broad range of program, technical, and administrative activities within AEOD, as shown below.
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| PROGRAM TECHNOLOGY BRANCH MARK H. WILLIAMS, CHIEF Responsibilities and Work Products 5.2.1 Conduct a comprehensive and systematic trends and patterns program 5.2.2 Conduct Nuclear Plant Reliability Data System (NPRDS) evaluation program and coordinate NRC guidance and usage of the NPRD System 5.2.3 Fomulate Abnomal Occurrence (AO) guidance and prepare A0 reports 5.2.4 Issue operational data feedback reports, including bimonthly Power Reactor Events reports and monthly LER Compilation reports 5.2.5 Prepare reports on U.S. events to the Nuclear Energy Agency's Incident Reporting System (NEA-IRS) and provide support to the International Atomic Energy Agency's Incident Reporting System (IAEA-IRS) 5.2.6 Develop techniques to apply PRA perspectives to the screening and analysis of operational events 83
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| l-5.2.7 Develop and manage data bases, including:
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| SequenceCodingandSearchSystem(SCSS)
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| Foreign Event File i 5.2.8 Access INP0 data files, including the NPRD System l 5.2.9 Formulate and operate internal management information systems,
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| } including:
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| LER screening results (WAMS)
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| Resource accountability (TACS) 3 IncidentReportingSystem(IRS) i i During this reporting period, a number of significant milestones and actions d
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| were accomplished. These are sumarized below for each of the individual l
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| : j. activities identified above. l 5.2.1 Trends and Patterns Program Plan, FY 86-88 4
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| i i The AE0D Trends and Patterns Programs Plan for FY 86-88 was developed in i 1985 and distributed to NRC Headquarters and Regional Offices for comment-in January 1986. This plan describes the AE00 program for the periodic
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| ! analysis of sets of operational event data reported by commercial power i reactor licensees in the Licensee Event Reports. (LERs).and to INP0's Nuclear '
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| 4 Plant Reliability Data System (NPRDS). " Trends and patterns" is used to describe a program for analyzing incidents of low individual significance
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| ; but which may be signficant because of their frequency or distribution. Trends.
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| and patterns analysis usually assesses operational data with limited prior formulation of a concern, whereas an engineering review usually begins with
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| { the formulation of a specific concern, i
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| i The AE00 Trends and Patterns Program consists of three separate program l
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| ; elements:
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| (1) Anal.ysis of Event Data. Four indepth reports, each treating a different category of operational events covered by the requirements-of 10 CFR 50.73 i
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| on reactor trips ESF actuations, loss of system safety function, and
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| '! technical specifications violations and shutdowns, .and a fifth report l
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| ! reviewing the performance of recently licensed plants, will be issued i covering the events from the most recent calendar year and comparing the latest results with that from earlier years. Further, each report I will contain findings, conclusions, and recommendations for correcting l l problems found to be most significant. Some preliminary results of the !
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| ; 1986 studies are discussed in Section 3. l (2) Analysis of Component Data. AE00 has developed and pilot tested a program for 1he systematic analysis of NPRDS data. The methods of analysis were tested, and the full-scale implementation of this program began in January 1986.
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| 4 1
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| 1 I
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| 84
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| (3) Performance Indicator Data. At present six indicators are being tracked by NRC's Performance Indicator Program. The basic input data for five of these indicators are developed by AE0D using contract assistance from INEL.
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| l Results of analysis of event data:
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| The results and conclusions of the analysis of events documented in LERS were presented in the following reports issued in 1986:
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| AE00 issued P603, " Trend and Patterns Analysis of Engineered Safety feature Actuations at Commercial U.S. Nuclear Power Plants," in August 1986. This study covers the July-December 1984 time period. The analysis of 1985 data will be issued in the spring of 1987. The preliminary results from the 1986 data, now being compiled, are discussed in Section 3.7.
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| AEOD issued report P602, " Trend and Patterns Report of Unplanned Reactor Trips at U.S. Light Water Reactors in 1985," in August 1986. The 1986 report is scheduled for publication in June 1987.
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| AEOD initiated trends and patterns analyses of losses of system safety function and violations of shutdowns required by technical specifications as reported in 1984 and 1985. The reports on these topics for 1985 will be issued in spring 1987.
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| In August 1986, AE0D issued P604, "AE0D Evaluation of New Plant Experience," an analysis of operational data gathered in the first 2 years of licensed operation for new plants. A more in-depth study of root causes of events at new plants will be issued later in 1987.
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| Methods and results of analysis of component failure data:
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| In 1985, AE0D pilot tested a program to systematically analyze NPRDS data. The NPRD System is an industry-wide, voluntary reporting system for tracking the performance of selected systems and components at nuclear power plants. Since almost all U.S. plants in commercial operation supply detailed design data, operating characteristics, and performance data, NPRDS provides an extensive data source for analysis of operating experience. AE0D worked with INP0 in d:veloping a list of critical components, called " key components," which are considered to have the greatest impact on safety system availability and the occurrence of plant transients. The NPRDS Trends and Patterns Analysis Program focuses on these key components.
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| The statistical analysis of the key components utilizes four different methods using the time to failure, engineering information and failure counts in NPRDS:
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| (1) Trend Analysis - Identifies the trend in the failure rate of key components over time and to determine if the rate is increasing or decreasing, and the significance of the trend.
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| 85
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| (2) The Page Procedure - Detects a shift in the failure rate of a particular key component above a preselected target or threshold failure rate.
| |
| (3) Survival Analysis - Determines the factors (pedigree information) which appear to influence the failure pattern of key components and the failure rate of the component over the life of the component.
| |
| (4) Reliability Modeling - Fits well-known reliability models to the failure patterns of key components; that is, to model the failure rate of the component over time, thus providing insight into failure causes and identifying outliers (components whose failure pattern suggests more detailedengineeringanalysis).
| |
| The results of these statistical analysis for each key component is then used by AE0D engineers as a starting point for an engineering evaluation to develop findings, conclusions, and recommendations for the selected key components.
| |
| The full scale implementation of this program began in January 1986 with the analysis of failures of PWR main feedwater system components. The preliminary results of this analysis are discussed in Section 3.9. Three proprietary reports were prepared by INEL under contract to AEOD in 1986.
| |
| 5.2.2 Conduct the NPRDS Evaluation Program and Coordinate NRC Guidance for NPRDS As directed by the Commission, an NPRDS evaluation program continues to be implemented. Semiannual evaluation re
| |
| , July 1986 (SECY-86-216) and February (ports on NPRDS progress were forwarded in SECY-87-54).
| |
| These reports highlighted the continued commitment of the industry to the improvement of the NPRDS. INP0's progress in the conversion of the NPRDS to an IBM based system will significantly improve the NPRDS capability and greatly enhance its usefulness. The reports also commended INP0 for improvements in the number of plants participating, in the number of reports being submitted by the plants, and in the timeliness and quality of the data during the period covered by these reports, fourth quarter 1985 through third quarter 1986.
| |
| In particular, the following specific observations resulted from the staff's activities involving NPRDS:
| |
| (1) All but one of the eligible plants submitted at least one failure report during the four quarters covered by these reports.
| |
| (2) Over 4,000 failure reports were submitted by the plants in each quarter.
| |
| (3) The timeliness of NPRD failure reporting for the quarters evaluated has improved compared to the same period in the preceding year. The reporting timeliness has stabilized over the years to the point where within 9 months of the end of the quarter approximately 75% of the NPRDS-reportable failures described in a sample of LERs were in the NPRDS data base; half of the failure reports are in NPRDS within about 6 months of the failure.
| |
| 86
| |
| | |
| (4) To assess the quality of the narrative information provided in NPRDS failure reports, a sample of reports was reviewed to determine if the text described the failure in sufficient detail such that system users could understand the failure events. For the four quarters evaluated, over 85%
| |
| of the samples were rated as at least "probably adequate" (i.e., a knowledgeable person could understand the characteristics of the failure.)
| |
| Although these improvements are very notable, the data base completeness and quality continue to warrant attention. At present, its usefulness is limited by the resources required to verify and upgrade the data after it is taken from
| |
| ! NPRDS. Particular observations in this regard are:
| |
| The staff continues to assess whether its current program for measuring the level of NPRDS reporting, using the component failures reporting in LERs as a yardstick, accurately reflects the overall level of reporting.
| |
| In an effort to improve the NPRDS, INP0 has promoted the reportina of failures documented in LERs. Therefore, the reporting level of component failures from LERs that are also in NPRDS may be higher, perhaps significantly, than the overall reporting level. During the past evaluation period, AE0D tested two methods of measuring the overall completeness of NPRDS failure reporting. Although no conclusions can be drawn from these two checks as to the overall completeness of the data base, they do indicate that diversified measuring methods should be pursued.
| |
| 3 From an NRC user perspective, some difficulties have been found in using NPRDS data, such as inconsistencies in the engineering reports, missing engineering records, and errors in the data. User's comments are becoming an increasingly important part of the NPRDS evaluation / improvement process. We have been and will continue to provide INP0 with this user feedback.
| |
| Recent case studies that attempted to use NPRDS data found that narrative descriptions in the NPRDS failure reports were often too brief to adequately determine failure causes. For this reason our recent evaluations have also measured the percentage of failure reports that provided good root cause information. Few reports met this test.
| |
| 8:cause of our concerns with the possible preferential reporting of LER documented failures and the problems in meeting NPRDS user needs, our future evaluations by the staff will be directed at determining failure reporting completeness and NPRDS appropriateness for the various end uses, including regulatory activities, and at stimulating needed improvements.
| |
| 5.2.3 Prepare Abnormal Occurrence (A0) Reports AE0D prepares the quarterly A0 Reports to Congress (as well as the associ-ated Federal Register Notices) and, after ' staff coordination, forwards them to the Commission. These reports, issued as the NUREG-0090 series, serve as a feedback of significant event information to Congress, government agencies, licensees, and the public. The reports are widely distributed (PDR, LPDRs, U.S. nuclear plants, Agreement States, other government agencies, and various foreign nations) and are available individually or on a subscription basis through the GP0 sales program.
| |
| 87
| |
| | |
| Three quarterly A0 Reports to Congress were issued during calendar year 1986 (third and fourth quarter CY 85, and the first quarter CY 86). The three reports described 19 A0s at NRC licensees and four A0s at Agreement State licensees. The reports also described 11 additional events as other reports of interest. Of the 19 A0s at NRC licensees, seven occurred at nuclear power plants and 12 occurred at other licensees (e.g., fuel cycle facilities, industrial radiographers, medical institutions). The details of the 19 A0s at NRC licensees were also published in the Federal Register.
| |
| Five of the 19 events reported in 1986 occurred during CY 86,13 occurred during CY 85, and one occurred late in CY 84.
| |
| The second quarter CY 86 A0 report was approved by the Comission in January 1986 and issued in February 1986. The report contained seven A0s at NRC licensees (two A0s occurred at the nuclear power plants and five occurred at other NRC licensees), and two A0s at Agreement State licensees.
| |
| The report also described seven additional events as other reports of interest.
| |
| The third quarter CY 86 A0 report was sent to the Comission for approval by SECY-87-73 dated March 18, 1987. The proposed report contains five A0s at NRC licensees (four of the A0s occurred at nuclear power plants),
| |
| and one A0 at an Agreement State licensee. The report also describes four additional events as other reports of interest.
| |
| The fourth quarter CY 86 A0 report is presently under development by the staff.
| |
| It is expected that it will propose nine A0s at NRC licensees (three of the nine A0s are associated with nuclear power plants) and no A0s at the Agreement State licensees. The report is also expected to describe four additional events as other reports of interest.
| |
| A sumary of the 1986 A0s are provided in Appendix A. These include those approved by the Comission (as of late February 1986) and those still in the approval chain.
| |
| 5.2.4 Issue Operational Data Feedback Reports, Including Bimonthly Power Reactor Events and Monthly LER Compilation Reports In 1986, AE0D continued with the routine publication of operating experience feedback. Two routine publications issued are discussed in this section.
| |
| Power Reactor Events is a bimonthly newsletter intended to feed back operating experience information and lessons learned to personnel at comercial nuclear power plants. These personnel include licensing engineers, plant managers, and training coordinators, as well as reactor operators and support personnel.
| |
| Other recipients total approximately 1000 and include NRC employees; employees of DOE and other Federal agencies; applicable State agencies; various domestic and foreign industry groups; and interested individuals.
| |
| Each issue contains the following information:
| |
| "Sumaries of Events" provides detailed write-ups of events that may be significant because of their safety implications and/or because of operator and licensee action taken during and after the events.
| |
| 88
| |
| | |
| -* " Excerpts of Selected Licensee Event Reports," added in June 1984, i
| |
| provides direct excerpts from LERs. In general, the information describes conditions or events that are unusual or complex, or that demonstrate a problem or condition that may not be obvious.
| |
| " Abstracts of Other NRC Operating Experience Documents" provides
| |
| ! abstracts and/or titles from pertinent NRC documents reflecting l operational experience such as Abnormal Occurrence reports, IE
| |
| ! Bulletins and Infonnation Notices, AEOD case studies and engineer-ing evaluations, NRR generic letters, and NRR operating reactor j event memoranda.
| |
| L l In CY 86, six issues of Power Reactor Events were published. They included 52 event summaries and 66 LER excerpts. Consideration is being given to terminating this publication, since the study of licensee programs mentioned in Section 3.1.1 indicated the document, while read and used by some licensees as source material for training programs, was not resulting in widespread use by licensees to initiate voluntary actions in response to the lessons communicated. The decision to terminate is pend-ing the results of three inputs--completion of a contract initiated by IE to improve generic communications with licensees, a survey of users planned for spring 1987, and possible impacts from the upcoming major reorganization of NRC functions.
| |
| ! Twelve issues of the monthly LER Compilation were published using data from the SCSS/ RECON databases. This compilation provides an abstract of each event sorted by facility name and chronologically by event date. Each event I is also cross-referenced by system, component, component vendor, and general i keyword indexes.
| |
| 5.2.5 Prepare Reports on U.S. Events to the Nuclear Energy Agency's Incident Reporting System (NEA-IRS), Provide Support to the International Atomic Energy Agency's Incident Reporting System (IAEA-IRS), and Support International Data Usage Programs i The U.S. continued to participate during the reporting period with 12 other [
| |
| countries (Belgium, Canada, Finland, France, Federal Republic of Germany, .
| |
| Italy, Japan, Netherlands, Spain, Sweden, Switzerland, and United Kingdom) '
| |
| in a centralized Incident Reporting System (IRS) for exchanging information' on operational experience. This system is operated by the Nuclear Energy '
| |
| i Ag:ncy (NEA) under the direction of the International Committee on the Safety of Nuclear Installations (CSNI).
| |
| AE0D serves as the principal U.S. technical representative to Principal Work- !
| |
| , ing Group (PWG) No. 1, " Operating Experience and Human Factors." During this
| |
| ; period. PWG-1 met in September 1986 with representatives from IAEA countries i to discuss significant operating events and also met in September 1986 to evaluate the significance of operating events and data in NEA countries. A
| |
| , total of four technical presentations were made by the AE0D representative in i these meetings. In addition, AEOD provided suggestions to both organizations regarding ways to obtain better consistency in the quality and the quantity of reporting.
| |
| J i
| |
| 2 89
| |
| , , - . -. -- ,~e-- -
| |
| 4 -, <c--- - - - - ~, , . - - . - . - . - - , - - , . -n- - - . - - , . , . . - .
| |
| | |
| i AE0D screens U.S. operating experience to select and prepare reports on those events meeting pre-established reporting criteria. During this report period, AEOD prepared and submitted to NEA a total of 65 IRS reports on U.S. operational events. In addition, AE0D prepared and submitted to NEA supplemental informa-tion on several U.S. events that were of particular interest to other IRS participants.
| |
| In 1985, the U.S. formalized its comitment to support the IAEA-IRS. The U.S.
| |
| is fulfilling its comitment through the NEA-IRS with NEA coordinating with IAEA on the exchange of incident reports. AE00, as mentioned above, also provided support to IAEA in 1986 on significant operating events and in the program direction area.
| |
| AE0D also provided considerable support to the NEA's International Symposium on Reducing Reactor Scram Frequency, held in Japan in April 1986. The objectives of the symposium were to: (1) compile consistent statistical data regarding scram frequencies in NEA member countries, (2) exchange insights on the reasons for the differences between countries, and (3) look for measures by which plant performance can be improved. AE0D served on the program comittee, coordinated the U.S. participation, and presented two papers and chaired three sessions at the meeting. One of the papers was a summary overview of the responses at the member countries to an NEA questionnaire on their scram experience. The other was the Trends and Pattern Analysis of Unplanned Reactor Trips at the U.S.
| |
| Light Water Reactors in 1984 and 1985 (AE0D/P504).
| |
| In addition, subsequent to the Chernobyl accident, the Subcomittee on Energy, Nuclear Proliferation, and Government Process of the Comittee on Governmental Affairs, U.S. Senate, requested that the NRC participate in a hearing on international nuclear safety concerns. AE0D provided technical support to the Office of International Programs, attended the hearing on May 8,1986, and pro-vided responses to Congressional questions for the record subsequent to the hearing (S. HRG 99-669).
| |
| 5.2.6 Develop Techniques to Apply PRA Perspectives to the Screening and Analysis of Operational Events An AE00 goal is to use the results and methods of probabilistic risk assess-ment (PRA) in reactor operational event assessment. Because of the large number of LERs received, routine use of PRA for prioritizing LERs may not be !
| |
| practical. Yet such techniques may offer promise to help assess and "cali- i brate" the significance of selected operating events and component failures. i In 1985, AE0D assumed responsibility for the Accident Sequence Precursor (ASP)
| |
| Program from the Office of Nuclear Regulatory Research. The obje.ctive of this program is to systematically evaluate operational data from U.S. nuclear power plants. The evaluation is intended to assess potentially serious operational incidents from a wide perspective. The information from this program is intended to serve as one element in NRC's assessment of nuclear plant opera-tional safety.
| |
| 90
| |
| | |
| The ASP program owes its genesis to the Risk Assessment Review Group, which concluded that " unidentified event sequences significant to risk might contribute...a small increment...[to the overall risk]" (NUREG/CR-0400).
| |
| The report continues, "It is important, in our view, that potentially signif-icant (accident) sequences, and precursors, as they occur, be subjected to the kind of analysis contained in WASH-1400." Evaluetions done for the 1969 te 1981 period [NUREG/CR-2497 (1969-79) andNUREG/CR-3591(1980-81)] were the l first efforts in this type of analysis.
| |
| I The ASP program involves a systematic review of operational events that have occurred at light-water power reactors to identify and categorize precursors to potential severe core damage accidents. Such precursors could be infre-quent initiating events or equipment failures that, when coupled with one or more postulated events, could result in a plant condition in which core cooling may not be adequate.
| |
| 5.2.6.1 ASP Evaluation of Reactor Operational Data A nuclear plant is designed with safety systems for mitigating accidents 4
| |
| or off-normal initiating events that may occur during plant operation.
| |
| These systems have a finite probability of failing or being in a failed state when required to operate. The ASP program uses LERs and other plant data, estimated system unavailabilities, the expected average frequency of initiating events [ loss of feedwater, loss of offsite power, loss-of-coolant accidents (LOCAs)], and event details to evalute the potential impact of the following two situations:
| |
| (1) Safety System Unavailability. Given a failure of a safety system or partial failures in two or more systems, the expected initiating event occurrence rates are used to determine the number of initiating events that may challenge the failed and backup systems during the period asso-ciated with the failure. The expected challenges are multiplied by system failure probabilities, using event trees to evaluate the likeli-hood of the occurrence of the overall event sequence.
| |
| (2) Initiating Event Occurrences. The probability exists that the standby safety systems may fail when called on to mitigate expected transients or transient-initiating events. The likelihood of potential severe core damage for precursors that included initiating events is calculated based on expected response of the safety systems. Failed or degraded systems existing at the time of the initiating event are accounted for in the calculations.
| |
| Initiating event frequency and system failure probability estimates are used, in conjunction with precursor event trees, to estimate a conditional prob-ability of potential severe core damage associated with each precursor. This probability is an estimate of the chance of potential severe core damage (in-adequate core cooling), given that the precursor event occurred in the manner it did, and can be considered a measure of the residual protection against potential severe core damage available during the event. The conditional probabilities associated with each precursor are used to evaluate precursors as to significance and to identify dominant sequences among all postulated sequences to potential severe core damage for the more significant events.
| |
| 91
| |
| | |
| I 1
| |
| 5.2.6.2 'Recent AE0D Publications e 'In 1986, AE00 published two reports: " Precursors To Potential Severe Core i Damage Accidents: 1985 - A Status Report" (NUREG/CR-4674), and " Precursors to i Potential Severe Core Damage Accidents: 1984 - A Status Report" (Draft).
| |
| j (Following the transfer of ASP responsibilities to AE00, the 1985 report was- l j perfomed before the 1984 report in order to develop an understanding of the i
| |
| ; mostcurrentdata.)
| |
| J
| |
| ! Studies of this nature are subject to certain inherent limitations. Often
| |
| ! there is limited data, and the results may be affected by many of the decisions I inherent in the process as well~ as in the methodology itself. However, a detemined effort is made to address these problems. Although uncertainties j exist in the numeric probability estimates associated with each event addressed in the report, the identification of the more serious events from a core damage standpoint is considered reasonably valid. The significant findings of these -
| |
| two studies are highlighted here.
| |
| l Accident sequences of interest in these studies are those that, if completed, would have'resulted in inadequate core cooling in the short term (typically up to 20 to 30 minutes) and then would have potentially resulted in severe core j damage. Accident sequence precursors are events that are important elements in such accident sequences. Such precursors could be infrequent initiating l events or equipment failures that, when coupled with one or more postulated
| |
| ! events, could result in a plant condition leading'to severe core damage.
| |
| l Precursors were selected and evaluated using a similar screening process and significance quantification methodology as was used in NUREG/CR-3591 for the l 1980 to 1981 period. This methodology permits a reasonable quantification of
| |
| ; the significance of an event without the laborious detail associated with evaluation using event trees and fault trees down to:the component level, while including observed human and system interactions.
| |
| The 1985 Study i
| |
| i Approximately 3000 LERS concerning events that occurred during 1985 were screened fcr accident sequence precursors. Of these, a total of more than 1400 LERs were selected for detailed review if they included a reactor trip or more serious initiator, included two or more component failures or un-i availabilities, or described an event that proceeded differently than expected.
| |
| All LERs selected for detailed review were subjected to an indepth evaluation.
| |
| As a result of this detailed review, 63 events were selected as precursors.
| |
| ; As a result, certain preliminary conclusions can be drawn.
| |
| The number of auxiliary feedwater (AFW), reactor core isolation cooling (RCIC),
| |
| j and isolation condenser failures following an initiating event observed in .
| |
| 4 1985 was greater than expected. In the case of AFW, three recoverable failures on demand following an initiating event were seen. No AFW system failures during testing were reported. Such failures would be expected more frequently than those following initiating events, because testing occurs more-frequently. -
| |
| ; if testing adequately addressed potential failure modes.
| |
| i s
| |
| ]
| |
| 92 1
| |
| | |
| Although the number of precursors per reactor ear identified in 1985 increased over the number identified in earlier reports primarily reflecting the revised LER requirements and changed program emphasis), the number of more significant events (i.e., with higher conditional probabilities of potential severe core damage) appears about the same as in the 1980 to 1981 time period. This observation is based on analyses that used system failure probabilities con-sistent with those used in the 1980 to 1981 evaluations. The use of higher failure probabilities for AFW, RCIC, and isolation condensers, which may be indicated by the number of failures seen in these systems in 1985, has not been factored into the precursor conditional probability estimates.
| |
| The dominant severe core-damage accident sequences associated with the more important precursors in past probabilistic riskfor BWRs were assessments generally) consistent with those predictedand (PRAs program for BWRs. For PWRs, the number of small-break-LOCA-related events pre-viously predicted (and seen) was not observed. This resulted in a relatively greater emphasis for sequences associated with loss of all secondary cooling and failure of feed and bleed for PWRs.
| |
| The 1984 Study Approximately 2400 LERs concerning events that occurred during 1984 were screened for accident sequence precursors. Of these, a total of more than 900 LERs were selected for detailed review if they included a reactor trip or more serious initiators, included two or more component failures or unavail-abilities, or described an event that proceeded differently than expected.
| |
| Forty-eight operational events reported in LERs during 1984 are considered to be precursors to potential severe core damage.
| |
| The number of precursors per reactor year identified in 1984 increased over the number identified in 1980-81 to the same level previously noted for 1985 events (about 0.6/ reactor year). Part of this increase is believed to reflect the changed program emphasis and revised LER reporting requirements (NUREG-1022). In addition, the frequency of events with conditional prob-abilities of core damage greater than 1 x 10(-4) was twice that observed in 1980-81. However, many of these events were only slightly above 1 x 10(-4),
| |
| and only one event greater than 1 x 10(-3) was observed, compared to six in the 1980-81 period.
| |
| The dominant severe core damage accident sequences associated with the more important in precursors past probabilistic forassessments risk BWRs. were (PRAsgenerally) and consistent previously with thoseinpredicted observed this
| |
| , program for BWRs. For PWRs, the number of small-break LOCA-related events previously predicted (and seen) was not observed. As was the case with the 1985 data, this resulted in a relatively greater emphasis in PWRs for sequences associated with loss of all secondary cooling and failure of feed and bleed.
| |
| 5.2.6.3 Future Plans The ASP evaluations of LERs occurring through 1986 will use the " potential for i core damage" as the primary measure of LER significance. LERs reported in 1987 i
| |
| and later will be evaluated and documented in ASP NUREG reports using the Commission
| |
| ! Safety Goals as the measure of event significance, to the extent practical. This means that the program goal will be to measure the operational event significance
| |
| : 93
| |
| | |
| to consider the likelihood and extent of radiological release through contain-ment to the environment. In order to perform operational data evaluations using the Commission Safety Goals, new methods beyond those currently used for estimating " core damage likelihood" are needed. Models and program results from the Office of Research's " Reactor Risk Reference Report" (NUREG-1150) are planned to be used to augment ASP assessment of LERs using the Comission Safety Goal measures of significance.
| |
| 5.2.7 Develop and Manage Data Bases AE0D utilizes a contract with the Nuclear Operations Analysis Center (NOAC) at Oak Ridge National Laboratory to assist in the data storage, retr' eval, and evaluation efforts, as discussed below:
| |
| (1) SequenceCodingandSearchSystem(SCSS)
| |
| The SCSS was developed with two objectives: (1) encode all of the rele-vant technical information provided by the licensee in the LER, and (2) encode the information with sufficient " tags" so that the individual pieces can be precisely retrieved. The SCSS data base currently contains data from 1980 to the present.
| |
| During this period, approximately 3300 current LERs were added to the data base. Also, due to the backfit program, approximately 3100 LERs from 1980 were added. This increases the number of LERs on the data base to almost 24,280.
| |
| Amendments 4 and 5 to Revision 1 of the controlled copies of the SCSS Coder's Manual were issued in January and August 1986, respectively.
| |
| Revision 1 of the Ouality Assurance Program for SCSS was issued in April 1986.
| |
| An amendment to the SCSS Programer's Manual reflecting additions and changes to the software and procedures which were made during FY 86 was issued in September 1986.
| |
| In January 1986, an IBM disk dedicated to support SCSS computer activities was installed to accomodate the 20,000+ LER records and other supplemental files.
| |
| In March 1986, the coding process was modified to begin coding LERs directly on-line interactively versus coding on paper forms and sub-mitting these forms for data entry via keypunch operations. The QA review and scan operations were performed on-line as well.- The computer QA checks are executed at each of the three processing stages, thus providing immediate feedback at each stage of the overall process.
| |
| By March 1986, the majority of LER abstracts were being entered via an optical character reader (0CR) for subsequent matching with the SCSS matrix.
| |
| 94
| |
| | |
| t :
| |
| Plant specific data was added to the SCSS data base.. This infomation included design electrical rating, criticality date, L commercial operation date, the State in which the plant is located, .
| |
| and the number of coolant loops. !
| |
| Work was completed in September 1986 to backfit the SCSS data base to include LER' data from 1980.
| |
| I There were 11 requests for on-line access to the SCSS during 1986 from employees of the NRC and the DOE national laboratories. _ This
| |
| ; made a total of 53 individuals who have direct access to the data
| |
| . base.
| |
| l AE0D and its contractor responded to about 215 requests for data j output from SCSS from individuals without direct access.
| |
| (2) Foreign Event File (FEF) j Beginning in 1980, NOAC. developed the Foreign Event File (FEF) data base, j- and in late 1985 p_ laced it (as a protected file) on the DOE computerized information system (RECON) to increase its availability to authorized users. On December 31, 1986, the RECON system was teminated and the FEF was transferred to DOE's Integrated Technical Infomation System (ITIS). ;
| |
| The FEF file is similar to the abstract file of LERs that has also been I maintained on RECON by the Nuclear Safety Information Center (NSIC), which
| |
| ;. is part of NOAC. The scope of the Foreign Event File is limited to events 4 at foreign light water reactors (LWRs) with an output-greater than 200 MWe.
| |
| i During the period. 903 events were added to the FEF data base and events identified as potentially safety significant and possibly
| |
| , applicable to U.S. reactors were forwarded for further review to AE0D, NRR, RES, IE, and INP0. FEF currently contains information on
| |
| ; about 7000 foreign events (1120 with full abstracts).
| |
| During 1986, the principal users of the system were NOAC and the NRC.
| |
| The file was accessed 205 times, i
| |
| In 1986. AEOD requested that NOAC review foreign reporting and suggest methods for obtaining consistency in quality and quantity in both the -
| |
| ! NEA-IRS and IAEA-IRS. AEOD then communicated its recomendations to the NEA and IAEA for consideration.
| |
| j (3) RECON LER file i An LER abstract and search capability is also maintained on the DOE RECON
| |
| : system, a computerized information retrieval system designed to provide users with remote terminal access to bibliographic data bases. RECON
| |
| ' access is available on-line to DOE offices, to institutions holding DOE j contracts, to other Federal agencies with energy-related or energy intensive missions, and to State agencies with State-wide responsibilities for energy programs or infomation. Printouts from the NOAC data base are i also available to other organizations on a cost-recovery basis by request.
| |
| From 1963 to 1983, the LER data for this data base was manually i
| |
| ___ -++..-..._--.--,,,-,------.,e.-,. s, m-. . --,,.y,w -,.#,. r e--e,> - - -w-,- --%, , .w.., -m-, -ve-. .-r-+* - + - . r-- - v3 e r, -
| |
| | |
| i i
| |
| l abstracted, and keywords were. manually assigned. Once the SCSS became operational, the NSIC data was generated by the computer from the SCSS
| |
| ! data base. Therefore, with minimal resources, the NSIC LER file' continued i to be available on-line to the government agencies and contractors.
| |
| However, in 1986, AEOD was informed that RECON would be terminated as of 12/31/86. Therefore, RM and AEOD have worked to set up the NSIC data base on a commercial data base network or another DOE system. Negotiations are in progress.
| |
| I~ Over 4400 LERs were added to the RECON database which now contains approximately 50,700 LERs received since 1963. i 1
| |
| l Of the more than 50 data bases included.in the RECON system, the LER
| |
| , data base is typically in the top six with respect to the frequency
| |
| !. of use. For example, for the last.3 months of 1986, the number of :
| |
| ] accessions to this data base ranged from 36 to 68 per month.
| |
| AEOD and its contractor responded to about 40 requests for data output from the RECON LER file.
| |
| , (4) Access INP0 Data Files i i Included in the Memorandum of Agreement with INPO is a provision to share data bases and results of analyses. As a result, AE00 maintains the -
| |
| l capability to conduct searches of selected INPO data files such as NPRDS.
| |
| f During this period, AE00, along with RM and the EDO, worked with INPO to i resolve problems concerning the need for MPRDS output. These problems i
| |
| resulted from the need to provide access for additional NRC users, to j train users on the on-line NPRDS data base management system SEEK, and to perform NPRDS searches for the NRC. Thus the number of individuals i outside of AE00 who were authorized to have access remained at seven.
| |
| } Late in 1986, contract negotiations were begun in response to INP0's j request for cost-recovery of the NRC's use of the NPRDS data. It is 1
| |
| anticipated that the contract will be finalized early in 1987.
| |
| ; AEOD responded to about 45 requests for data output from NPRDS.
| |
| 1 l (5) Fomulate and Operate Internal Management Infomation Systems
| |
| ! AE00 maintains several internal data bases for such purposes as docu-l mentation of screening results and resource accountability. An additional i small data base.is maintained on reports of U.S. events to NEA/ IRS.
| |
| 1 I
| |
| Routine reports of LER screening results (WAMS) and manpower utilization -
| |
| (TACS) are prepared periodically and used by AE00 managers.
| |
| j (6) Other Activities j PTB prepared or coordinated office procedures and prepared the office's i input to the NRC Annual Report to Congress. Program support reports l completed by PTB are listed in Table 15.
| |
| 1 j
| |
| l 96 i.--_---_._,. -
| |
| | |
| . -. . = . . . _ - . ..
| |
| Table 15 Progran Support Reports Date Subject. No.
| |
| i 1/86 Trends and Patterns Program Plan - FY86-FY 88 P601 5/86 An Overview of Nuclear Power Plant Operating S602 Experience Feedback Programs 8/86 Trends and Patterns Report of Unplanned Reactor P602 4
| |
| Trips at U.S. Light Water Reactors in 1985 8/86 Trends and Patterns Analysis of Engineered Safety P603 Feature Actuations at Commercial U.S. Nuclear Power Plants (July 1984-December 1984) 8/86 Trends and Patterns Analysis of the Operational P604
| |
| < Experience of Newly Licensed U.S. Nuclear Power Reactors 5.3 Nonreactor Assessment Staff (NAS)
| |
| NAS screens and analyzes the operating experience associated with the activi-ties and facilities licensed by the Office of Nuclear Material Safety and Safeguards (NMSS) and by Agreement States (i.e., nonreactor operating experi-ence). In addition, NAS conducts studies from a human factors perspective on both reactor and nonreactor operating events.
| |
| The major activities of NAS include: reviewing events reported to the five Regional Offices by licensees; reviewing inspection reports; entering coded information on events identified from the review into computer data files; ovaluating the events as a whole; and performing detailed reviews of specific events and concerns. The full scope of NAS responsibilities is shown below:
| |
| NONREACTOR ASSESSMENT STAFF KATHLEEN BLACK, CHIEF Responsibilities and Work Products 5.3.1 Screen individual events associated with NMSS and Agreement
| |
| , State licensed activities and facilities and determine significance 5.3.2 Analyze and evaluate individual nonreactor and medical misadministration events and related potentially generic safety concerns 5.3.3 Analyze reactor and nonreactor events from a human factors perspective 97
| |
| | |
| - . _-.. . . - . ~. - -. . . .-- .- - - - - . . . _ - - --
| |
| 5.3.4 Document independent technical assessments in:
| |
| Case studies Engineering Evaluations Technical Reviews Memoranda
| |
| [ 5.3.5 Develop, maintain, and provide updating data to these computerized data files:
| |
| j
| |
| * Nonreactor data file Pedical misadministration data file 4
| |
| . The milestones and progress on these NAS responsibilities are summarized i
| |
| individually below.
| |
| 5.3.1 Screen Individual Events Associated with NMSS and Agreement State l
| |
| Licensed Activities and Facilities and Determine Significance j
| |
| l About 6000 reports per year are received from the approximately 8300 NRC
| |
| ; licensees. Of these reports, approximately' 400 involved an operational event, such as an overexposure, spill,' or medical misadministration. In addition, the I 8000 Agreement State licensees are required to submit reports to the cognizant j Agreement State which, in turn, provides reports on certain of the reported events to the NRC. The Agreement State reports are also reviewed principally from the standpoint of generic problem identification and from their contribu-l tion to the review of events reported by NRC licensees.
| |
| I As a result of these NAS screening activities, the following individual ~ events
| |
| : associated with NMSS and Agreement State licensed activities and facilities I were identified: (1) events that are determined to be A0s; (2) events that are
| |
| ! themselves not of high individual significance but are events with potential
| |
| ; generic significance; and (3) events that appear to have an immediate sig-
| |
| ! nificance, either specifically or generically, i
| |
| l Of the many hundreds of event reports from NRC and Agreement State
| |
| ! licensees, 16 were determined by the Commission to meet the A0 reporting i criteria. Another eight events occurring in 1986 were recommended by AE00 l as meeting the criteria, but were not yet acted on by the Commission. Of
| |
| : the 24 events, 14 occurred in 1986 and ten occurred in 1984 or 1985; four l were diagnostic misadministrations, five were therapeutic misadministra-
| |
| ! tions, and four were occupational overexposures (three of which were j overexposures of radiographers or assistant radiographers). 'There were 11 additional A0s: exposure of a member of the public; the rupture of a j
| |
| ; UF-6 cylinder at Sequoyah Fuels; loss of management controls at an irra-diator licensee; contamination of a scrap steel facility; two willful
| |
| !~ failures to report misadministrations; a release of krypton-85; distri-bution of molybdenum-contaminated technetium; contamination of a building
| |
| , at Wright-Patterson Air Force; suspension of the, licensee for a radi- i
| |
| ! ography and teletherapy servicing fim; and removal of an officer of a l radiography firm from any assignment of position related to NRC licensed l activities. In one A0, a member of the public was exposed.
| |
| l
| |
| {
| |
| ; 98 i
| |
| b
| |
| ,rr,,,,- ---,-.yv-.- ,-,-w,. ,,e-- -,-y,--,,-w- -,, ~,- .,, ,,,-r,-.v- ,_ ,--, y,-~~---y,,- ,
| |
| | |
| l Each of the four diagnostic misadministrations involved the misadministra- l tion of iodine-131.
| |
| The five therapy misadministrations all resulted from human factors.
| |
| ! AE0D published a case study (AE0D/C505) on therapy misadministrations that had occurred prior to 1985. A major conclusion of this case study--
| |
| that independent verification of dose calculations and patient chart reviews could reduce the number of therapy misadministrations--is further substantiated by these A0s.
| |
| 5.3.2 Analyze and Evaluate Individual Nonreactor and Medical Misadministration Events and Related Potentially Generic Safety Concerns During the reporting period, case study AE0D/C601, " Rupture of an Iodine-125 -
| |
| Brachytherapy Source at the University of Cincinnati Medical Center," was completed and issued August 12, 1986. The study documents the anal rupture of an iodine-125 seed (nominal activity of 40used millicuries) in yses o brachytherapy treatment of brain tumors at the University of Cincinnati in SIptember 1984.
| |
| I The primary conclusion of the study was that the the risk of an iodine-125 seed rupture is relatively high when the seeds are reused for several' patients. The i risk of a seed rupture is associated with:
| |
| )
| |
| * The susceptibility of the seeds to damage from typical tools used for removing the seeds (razor blade, scissors, etc.); and 4
| |
| * The discolored or stained condition of the catheters after use in therapy, i making viewing of the seeds difficult.
| |
| The consequence of the seed rupture at the University of Cincinnati, involving
| |
| : patient and other personnel uptakes and facility contamination, could have been
| |
| ) mitigated by adequate radiation surveys of the work area and the tools used to remove the seeds from the catheter, or by performing a leak test of the seeds.
| |
| Additionally, personnel uptakes other than the patient and the facility contam-ination could have likely been prevented if the seed removal operation ha'd been j performed under a fume hood.
| |
| l It appears that the consequence (personnel uptakes, and personnel and facility
| |
| ; contamination) of a similar event could also be mitigated by employing radia-i tion safety procedures designed to promptly detect if a seed is ruptured and to j prevent personnel uptakes and personnel and facility contamination. Such procedures would include: performing the removal / reloading operation in a fume hood; performing wipe surveys of tools and the area used for the removal-and i reloading of the seeds; or leak testing the seeds following the removal /
| |
| {
| |
| reloading operation.
| |
| The report reconnended that the Office of Inspection and Enforcement send an; l information notice to the affected licensees describing the event and the l l action taken by the licensee and the source manufacturer to prevent the !
| |
| 1 j recurrence of similar events; specific instructions and safety precautions for l j reusing the seeds be communicated; and NMSS should explore the option of 1 addressing the reuse of the high activity iodine-125 seeds during the license I
| |
| , issue, renewal, or amendment process.
| |
| 99 [
| |
| I l
| |
| | |
| I 5.3.3 Analyze Reactor and Nonreactor Events from a Human Factors Perspective In December-1985, AEOD issued case study report AE0D/C504, " Loss of Safety
| |
| ' System Function Events." The study identified 133 events involving a loss
| |
| .of safety system function (LSSF) in the '1981 to June 1984 period, 87 of which (65%) were the result of human factors contributions. The results i of the study November 1986 were studyaby principal the CSNIimpact Principal and wereGroup Working published as p(art of a
| |
| #1 0perating Experience and Human Factors) report entitled " Loss of Safety System Functions." '
| |
| Special study AE0D/S401, " Human Error in Events Involving Wrong Unit or Wrong . Train," was published in January 1984. Because of the implications of wrong unit / wrong train events, NAS has continued to track and analyze similar subsequent events including events involving the wrong component.
| |
| Because wrong unit / wrong train / wrong component (WU/WT/WC) events continued l to occur, AEOD and the Office of Nuclear Reactor Regulation (NRR) under-i t
| |
| took a cooperative field survey of the problem during 1985. A series of site visits was undertaken to obtain first-hand knowledge of plant layout, t equipment identification, and personnel practices. By the end of 1985, NRR and AEOD staff had completed site visits and reviews at ten multi-unit sites to obtain information on the factors that appeared to have contrib- :
| |
| uted to these types of events. NRR published the results of the study in "An Investigation of the Contributions to Wrong Unit or Wrong Train Events," NUREG-1192, dated April 1986.
| |
| ' Industry has recognized the' potential significance of this type of event, and NUMARC has stated that it will study this issue. At the suggestion of NRR, AE00 will continue to monitor the occurrence of this type of event on
| |
| ' a normalized basis in order to determine the progress made by industry.
| |
| Reports on results of the monitoring will be published semiannually.
| |
| Technical review AE00/T607, dated September 19, 1986, includes the annual i
| |
| total number of WU/WT/WC events over the 1981 to June 1986 time period.
| |
| l The report concluded that: (1) there does not appear to be a reduction in the number of events involving the wrong unit / wrong train / wrong component,
| |
| , and (2) plants with little operating experience (less than 2 years from initial criticality) seem to experience a disproportionately high number i of these events. Preliminary data for the.second half of 1986 confirm 1
| |
| this (see Table 16). However, the apparent high rate of occurrence in more recent years may reflect the improved investigation and reporting i
| |
| ' of the detafis of causes of events. rather than an increase in event occurrence.
| |
| Of the significant operating reactor events recently investigated by the NRC, a large fraction involved procedures to some degree. As a result, AE00 began a study of 1984 and 1985 events that involved procedures and 1 that were significant from a safety standpoint. The initital study 4
| |
| focused on 1984 and 1985 LERs that were assigned an AEOD significarce level of "1" or "2". Of the 179 such reports from 1984 and the 110 re-ports from 1985, totals of 54 and 47 reports, respectively, were found to 1
| |
| 100 f
| |
| | |
| .!i involve procedures.* The 54 events in 1984 and the 47 events in 1985 that involved procedures were then exam'ned ~to detemine the characteristics of I
| |
| L the events.
| |
| The initial review of the characteristics of the events indicated that problems associated with procedures were contributing factors in a high
| |
| ! percentage oil recent significant events, and that some licensees are more
| |
| -prone to these probless than others. However, the data-also indicated that some plants may tend to use procedure revisions asia general 4
| |
| J i'-
| |
| '- corrective action following an event. This may be due to a lack of understandino as to'the fundamental cause and contributing factor (perhaps due to lack of thorough event investigation); (2) a focus on equipment problems rather the personnel performance problems due to deficiencies in 4 personnel qualifications, training, or environment; and/or (3) a desire to auickly "close out" an LER, when the complete action to prevent recurrence is far-reaching. .
| |
| To achieve a better understanding of the contribution of procedures to the occurrence of significant events, a series of site visits and discussions with resident inspectors and licensees' staff was undertaken beginning in-late 1986. The site visits should be completed and the special study report issued in early 1987.
| |
| Some prelirainary conclusions of the study are that problems associated with procedures were contributing factors in a high percentage of recent (1984-1985) significant events; and licensees frequently reported that-procedures were incomplete, possibly indicating inadequate programs for 3
| |
| assuring procedure quality (e.g., procedure improvement programs).
| |
| 5.3.4 Document Independent Technical Assessments Technical Reviews:-
| |
| * AE0D/T604, " Events Resulting from Deficiencies in Labeling and Identifi-cation System," dated May 6, 1986, described the characteristics of 38 recent events that resulted from deficiencies in nuclear power plant labeling and identification (L&ID) systems; that is, administrative-sys-tems for ensuring the correct identification of units, trains, and compo- ,
| |
| nents. The report provides additional ~ evidence that deficiencies in L&ID systems frequently contribute to human error and that this continues .to be a problem in the nuclear power industry.
| |
| *For the purpose of the review, " events resulting from defective procedures" were those events that were at least partially the result of a lack of, deviation from, or deficiencies in, operating, maintenance or administrative
| |
| ! control procedures. An event was deemed to be the result of a ~ defective procedure if (1) the LER stated that a procedural deficiency or deviation from a procedure was a contributing factor or that a procedure change (s) had
| |
| : or would be made to prevent recurrence, or (2) in LERs in which " procedures"
| |
| ! were not explicitly mentioned, a judgement was made as to whether procedures were a contributing cause.
| |
| - 101
| |
| _. .- _. _ _ _ l
| |
| | |
| Table 16 Events Involving the Wrong Unit, Train, or System Considering Event Type and Plant Operating Experience
| |
| * 1981 1982 1983 1984 1985 1986 Total Wrong Unit Events at Plants with:
| |
| >2 years experience 2 3 4 3 P 5 25 12 years experience 1 0 1 3 1 3 9 Wrong Train Events at Plants with:
| |
| >2 years experience 6 8 12 18 14 25 83 12 years experence 1 1 1 4 8 5 20 Wrong Component Events at Plants with:
| |
| >2 years experience N/A** N/A** 4 10 19 33 66
| |
| $2 years experience N/A** N/A** 0 7 22 18 47 Total Events at Plants with:
| |
| >2 years experience 8 11 20 31 41 63 174 12 years experience 2 1 2 14 31 26 76 Total 10 12 22 45 72 89 250 Number of Multi-Unit Sites 20 20 23 25 26 28 Total Number of OperatingReactor Plants ** 69 74 77 83 92 99 Number of plants with
| |
| >2 years experience 63 66 69 74 77 83 Number of plants with 52 years experience 6 8 8 9 15 16
| |
| * Experience is defined as the amount of time from initial criticality.
| |
| **N/A = Not Available
| |
| *** Totals do not include Lacrosse and Fort St. Vrain.
| |
| Total does include TMI-1 for 1985 and 1986, 102
| |
| * AE0D/T607, " Occurrence of Events Involving Wrong Unit / Wrong Train / Wrong Component - Update through June 1986," dated September 19, 1986.
| |
| l Othsr Reports:
| |
| * " Wrong Unit / Wrong Train Events, 1981-1985," dated February 13, 1986.
| |
| * " Trip Report for McGuire and Oconee Site Visits Regarding Wrong Unit / Wrong Train Events," dated March 6, 1986.
| |
| * " Trip Report for Turkey Point Site Visit Regarding Wrong Unit / Wrong Train L Events," dated March 6, 1986.
| |
| l 5.3.5 Develop, Maintain, and Provide Updating Data to the Nonreactor and Medical Misadministration Data' Files l Computerized data files continued to be updated by NAS on (1) ncnreactor
| |
| ; events, and (2) medical misadministrations.
| |
| i From the events screened by NAS, approximately 200 nonreactor events and
| |
| ! 400 misadministrations were coded by NAS and entered into the nonreactor data file during 1986.
| |
| 5.4 Incident Investigation Staff (IIS)
| |
| The overall goal of the Incident Investigation Program (IIP) is to promote the i public health and safety by thoroughly understanding the nature and causes of significant events and thereby reduce the frequency and consequences of such incidents. The objective of the IIP is to ensure that operational events are investigated in a systematic and technically sound manner to gather information partaining to the probable causes of the events, including any NRC contribu-tions or lapses, and to provide appropriate feedback regarding the lessons of i experience to the NRC, industry, and public. By focusing on probable causes of operating events and identification of associated corrective actions, the i
| |
| results of the IIP process should improve nuclear safety by ensuring a com-J plate technical and regulatory understanding of significant events.
| |
| Tha IIP has two investigatory responses based on the safety significance of tha operational events. Both involve responses by an NRC team to detemine
| |
| ; th3 circumstances and causes of an operational event. For the most signifi-I cant operational events, an Incident Investigation Team (IIT) is dispatched l
| |
| by the Executive Director for Operations (EDO) to investigate the event. The response to less significant operational events are designated Augmented In-
| |
| ); spection Teams (AITs) which are usually initiated and directed by the Regional i Office. Although no IITs were activated during CY 1986, nine AITs were con-t ducted, as shown in Table 17.
| |
| ] These AIT and IIT responses ensure that significant operational events are investigated in a manner thac is timely, objective, systematic, and technically scund; that factual information pertaining to the event is documented; that i probable cause(s) are ascertained; and that a complete technical and regulatory
| |
| ] understanding of the event is achieved.
| |
| +
| |
| 103 i
| |
| | |
| Table 17 Augmented Inspection Team (AIT) Responses i Date of Event Facility Event / Comments 1/4/86 Sequoyah Fuels Facility Rupture of Model 48Y UF-6 Cylinder and Release of Uranium Hexafluoride 1/31/86 Perry Earthquake .5.0 on Richter Scale 4/12/86 Pilgrim Recurring Residual Heat Removal (RHR) Isolation and Valve Leakage 5/19/86 Palisades Multiple Equipment Failures, Maintenance Backlog 6/1/86 LaSalle Unit 2 Feedwater Transient with Failure of Static-0-Ring Switches 6/27/86 Catawba Unit 2 Unit 2 Depressurization During Remote Shutdown Test 9/11/86 Hope Creek Multiple Failures During 9/19/86 Loss of Offsite Power Tests Faulty Bailey Solid State Logic Modules 12/3/86 Hatch Loss of Water from Spent Fuel Pools to Site Due to Failed Inflatable Seals 12/9/86 Surry Unit 2 Main Feed Pump Suction Pipe Rupture i
| |
| } 1 04
| |
| | |
| . ,._m - . ~ .. ._ m.. -
| |
| l i
| |
| L l
| |
| l In January 1985, the Incident Investigation Staff (IIS) was established as an organizational element within AE00. The IIS has the responsibility to develop policy, program requirements, and procedures for IIT investigations of significant operating events. An IIS member will accompany each IIT to provide administrative support, liaison, and technical guidance to ensure that the IIT l activities are consistent with established procedures and coordinated with NRC cffices and other organizations. The IIS maintains and integrates plans, pr:ctdures, team rosters, and training, and coordinates staff activities to
| |
| ; achieve the IIP objectives as noted below:
| |
| l INCIDENT INVESTIGATION STAFF WAYNE D. LANNING, CHIEF Responsibilities and Work Products
| |
| , 5.4.1 Develop formal guidance for and issue the NRC Incident Investigation Program (Manual Chapter 0513) 5.4.2 Prepare and maintain personnel rosters of candidate IIT leaders and members 5.4.3 Develop and provide training for IIT candidates 5.4.4 Develop formal protocol for the treatment of guaranteed i equipment and information during an IIT i 5.4.5 Provide and maintain administrative support for the IITs j Th2 following describes the activities associated with the IIP that were completed during CY 1986.
| |
| 1 5.4.1 Issuance of Manual Chapter NRC-0513. "NRC Incident Investigation Program" An NRC Manual Chapter was prepared that defines the duties, responsibilities, 4 authorities, and schedule for event investigation under the IIP. This Manual
| |
| ! Chapter contains guidance regarding the significant operational events to be investigated by an IIT and by an AIT. The Manual Chapter was issued for com-ment on March 5, 1986, concurred by all offices, forwarded to the ED0 for approval on July 31, 1986, and issued final on August 8, 1986.
| |
| ! 5.4.2 Establishment of Personnel Rosters Parsonnel rosters were developed to enable a timely identification of NRC
| |
| : staff having specific expertise to be considered as IIT leaders and members for evsnts at reactor sites. Candidates for the roster were reconsnended by their
| |
| , office director and approved by the EDO. The rosters are updated and revised cvsry 6 months.
| |
| l l The initial roster was completed in February 1986 and approved by the EDO in April 1986. In August, the rosters were revised to update each member's expertise based on the background information provided. The rosters and i
| |
| 105
| |
| | |
| background information have been entered into a database file using an IBM personal computer for ready access and convenience for updating and revising.
| |
| 5.4.3. Development of an IIT Training Program The purpose of the IIT training program is to provide IIT candidates with a comprehensive range of guidelines and methodologies to ensure that NRC investigations of significant events are timely, thorough, coordinated, and femally administered. The training program was developed and organized by the IIS after discussions with representatives from the National Transportation Safety Board, the Federal Aviation Administration, Ontario Hydro, the Depart-ment of Energy, and the National Aeronautics and Space Administration. The i experience and insights from previous IIT team leaders and members were incor-porated into the training course.
| |
| The course consists of an intensive 2-week curriculum that begins with an everview of the NRC Incident Investigation Program, including perspectives from previous IITs, and covers the following areas:
| |
| Incident Investigation Program Investigation Perspectives Investigation Guidelines Investigation Analytical Techniques The two-week training course covering event investigation methodologies and the IIT procedures was given to 25 representatives from Headquarters and Regional Offices on July 7-18, 1986. This class composition was selected on the basis of establishing five IITs. As part of the training course, a training manual was prepared for IIT class participants containing background material and draft procedures.
| |
| 5.4.4 Development of Standard Confimation of Action Letter and Order A generic Confirmation of Action Letter (CAL) and Order were developed that cstablish the protocol for the treatment of quarantined eouipment and informa-tion during an IIT investigation. The licensee's responsibility with respect to the quarantined equipment is clearly defined. The generic CAL and Order also address parallel investigations by confinning that the licensee will ensure that any investigation to be conducted by the licensee or third party will not interfere with the IIT investigation. These documents have been included in the Incident Investigation Manual.
| |
| 5.4.5 Development and Implementation of IIT Administrative Requirements The IIS developed procedures to guide the administrative functions necessary to dispatch, support, and close out IITs. The completion of these administrative procedures was a major activity during CY 1986. The administrative procedures address the following subject:
| |
| 5.4.5.1 Imprest Fund An imprest fund was established in AE00 to enable IIT members to obtain a i
| |
| travel advance if an IIT is activated during off-duty hours for IIS and IIT 106
| |
| | |
| travel during non-duty hours. The IIS has obtained a safe and cash box, and has implemented appropriate administrative controls for safeguarding and dispersing this fund.
| |
| 5.4.5.2 Travel Arrangements
| |
| * Ability to process travel authorizations and obtain travel advances on short notice regardless of day or time. Procedures address the use of
| |
| ! transportation requests and accessing the imprest fund for emergency l travel.
| |
| * Ability to identify the proximity of the major airports and available lodging near sites. The IIS maintains a Site Data Notebook containing information enabling IITs to travel to the site, with confinned hotel reservations (e.g., proximity of major airports, available lodging near sites, reactor site addresses, and directions to site).
| |
| * Ability to process travel claims in a timely manner. Procedures have been established to ensure timely and correct submission of travel vouchers for IIT members.
| |
| 5.4.5.3 Stenographers The IIS is responsible for ensuring that stenographers are'available at the site in a timely manner to transcribe IIT interviews. The Atomic Safety and Licensing Board Panel (ASLBP) manages the NRC contract for stenographers and assists the IIS in obtaining this service. Procedures have been established addressing the coordination with ASLBP to obtain stenographers for IITs, including provisions for off-duty hours and holidays.
| |
| 5.4.5.4 Equipment Th2 IIS has the following equipment available for use by IIT members, along with controls for assigning and returning equipment: 35m camera with flash and lens attachments, carrying case, and film supply; mini-cassette recorders with tapes and batteries; mini-transcribers; key cards; kastle cards; a portable computer; and a "go-bag" containing this equipment, procedures, and supplies for the IIT when in the field.
| |
| 5.4.5.5 Onsite Support Arrangements Procedures have been established to ensure that proper arrangements are made for the IIT while on site, including obtaining access to the site and making a.rrangements for providing the IIT with the physical facilities necessary to accomplish the onsite investigation efficiently and with a minimum of disrup-ticn to all involved. Procedures also address the proper recordkeeping of time and attendance of IIT members while onsite.
| |
| 5.4.5.6 Records and Documentation Control Pr:cedures have been established to ensure that all infonnation will be handled in a systematic manner to minimize the probability of lost information, to cnsure that safeguards and proprietary information is properly controlled, and to ensure that information collected during the investigation is readily 107
| |
| | |
| . . - - =. _ . - -- . _ . _ - _ . - - - ... _ . . -
| |
| i retrievable. Specific guidance is provided for the handling of sensitive information such as transcripts and allegations of wrongdoing.
| |
| 5.4.6 Development of the Incident Investigation Manual 4
| |
| i The Incident Investigation. Manual.was prepared to provide guidelines for the conduct of investigative activities by the IITs. The guidelines were developed and coordinated by the IIS aad reflect the experience gained from previous IIT investigations and other pertinent investigations. The Incident Investigation Manual addresses the specific points and concerns .
| |
| } identified in the Commission Paper (SECY-85-208) and includes the following: '
| |
| i 5.4.6.1 Activating an IIT I
| |
| This provides guidance to NRC management for activating an IIT response and addresses the following areas:
| |
| Selection and Scope of Events for IIT Response IIT Membership IIT Activation Process Participation by Industry Organizations AugmentedInspectionTeam(AIT) Response j -
| |
| Upgrading or Downgrading an IIT
| |
| + 5.4.6.2
| |
| . Conduct of Investigation s This provides guidance to the IIT leader for conducting an investigation including the following topics:
| |
| Scope of the Investigation Team Leader Responsibilities Role of the Region l
| |
| Initial Actions by the Team Leader Entrance Meeting with the Licensee Plant Tour of Equipment and Systems Interviewing Personnel l
| |
| Sequence of Events j -
| |
| Development of the Quarantine Equipment List (QEL) 4 i -
| |
| Responding to Press Inquiries IIT Coordination Meetings Identifying Additional Expertise and Outside Assistance Industry Participation in the Investigation Paralle1 Investigations
| |
| - Status Reports i -
| |
| IIT Recordkeeping Activities Collection of Information Referral of Investigation Information to NRC Offices Confidentiality Subpoena Power and Power to Administer Oath and Affirmation IIT Investigation Sequence Return Site Visit
| |
| [
| |
| Report Preparation ar' Presentation I
| |
| t j 10s
| |
| . . . . - . - - . ,-~~,n- --w..-,,n-r---,n.. , - , - ,~ - , - . . , . . . . . , - , - - - . . . . , - , . - , , - , - , , , . . - , . . - - - - , , - - -
| |
| | |
| . . - ~. __ _ . _ . . .. _. _ - . _ . _ _ . _ . _ _ - . __ _ _
| |
| i i
| |
| 5.4.6.3 Interviews i
| |
| ! This provides guidance to IITs to ensure interviews are conducted in a uniform, I systematic, and complete manner and addresses the following areas:
| |
| 4
| |
| - Guidance for Conducting Interviews
| |
| : - Handling of Transcripts and Access by Other Parties
| |
| < - Attendance of Third Parties at Interviews
| |
| : 5.4.6.4 Treatment of Ouarantined Equipment
| |
| { This provides guidance for handling quarantined equipment and developing related
| |
| ! troubleshooting action plans including.the following areas:
| |
| 4
| |
| ! - Licensee's Responsibilities
| |
| ! - QuarantinedEquipmentList(QEL) Guidelines 1 - Guidance for Developina Troubleshooting Action Plans
| |
| > 5.4.6.5 Report Preparation This provides q'uidance for the preparation, release, and distribution of the IIT report inc uding the following areas:
| |
| ! - Writing and Publishing Guidelines
| |
| - Report Writing Guidelines
| |
| - Graphic Guidelines
| |
| ; - Publication Forms
| |
| - Distribution of the Advance Copy
| |
| - Distribution of the Published NUREG ;
| |
| - Schedule t
| |
| 5.4.6.6 Background Information i This section contains background information for the IIP including:
| |
| - NRC Manual Chapter 0513 Incident Investigation Program .
| |
| ) - IE Procedure for Augmented Inspection Team Response to Operational Events i - SECY-85-208 entitled "NRC Incident Investigation Program" i ?
| |
| : The manual was issued for review and connent on June 16, 1986, discussed during ;
| |
| ) the training program, transmitted to the Owners' Groups, NSAC, and INP0 for review on August 4, 1986, and issued for trial use on September 12, 1986.
| |
| i 5.4.7 Regional Workshops 1 During the last months of 1986, the IIS made preparations to hold a one-day I workshop on the NRC's Incident Investigation Program in each of the regional offices. The purpose of the workshop is to acquaint utilities with the IIP i and provide an understanding of the process by which NRC will investigate potential significant incidents. It also provides an opportunity for the utilities to be better prepared should an incident at one of their facilities trigger the establishment of an IIT or an AIT. The workshop provides a forum for questions to be raised and answered, and for industry connents to be incorporated into the next revision of the Incident Investigation Manual.
| |
| 109
| |
| | |
| i i 5.4.8 Industry Participation 1
| |
| The objective of this activity is to have representatives from outside organt-zations, such as INPO, NSAC, and utility Owners' Groups, participate fully as
| |
| ! team members in future IITs. Their participation would add additional perspec-
| |
| ; tive to the investigation and expert knowledge of industry practices in numerous l
| |
| ' areas. Guidance concerning industry representatives on IITs was developed which addresses the qualifications, selection, responsibilities, and role of industry participants on IIT investigations. This guidance was forwarded and j subsequently discussed in meetings with INPO, NSAC, and all utility Owners' Groups to elicit their comments concerning industry involvement in the IIT process. A number of issues raised during these discussions, including questions about an industry representative's liability, about NRC supervision
| |
| ! of non-government members of an IIT, about the need for a statement of conff-
| |
| : dentiality from non-NRC IIT members, and about the application of the Federal i
| |
| Advisory Connittee Act are being addressed. Resolution of these issues in
| |
| { coordination with the Office of General Counsel is expected in 1987.
| |
| j i
| |
| i i
| |
| 4 v
| |
| )
| |
| i i
| |
| I a
| |
| 110
| |
| | |
| 1 l
| |
| l
| |
| ! 6. STATUS OF AE00 RECDMMENDATIONS This section sunnarizes the year-end status of all AE00 recanmendations which are either new or still outstanding from the last report. The status of a
| |
| : total of 100 recommendations are provided in this section. These have been categorized (as of December 31,1986) as follows:
| |
| I Status of AE00 Recommendations Added Since Last Report 28 Resolved Since Last Report 11 Total Not Resolved 79 Currently Under Review 17
| |
| - Proceeding Satisfactorily 59 Not Proceeding Satisfactorily 3 4
| |
| For tracking purposes, " Resolved Since Last Report" means that formal NRC
| |
| ; action has been taken requesting licensee action. " Currently Under Review" means that the recommendation is being assessed by another NRC office and no i position has yet been taken by the responsible program office. " Proceeding 4
| |
| Satisfactorily" means that the responsible program office has agreed with the recommendation and has initiated appropriate action. "Not Proceeding Satisfactorily" means that the recommendation was not accepted by the program office and additional discussion between the program office and AE0D is being or will be held. At this time, there are no issues involving AE00 recommendations which would warrant EDO attention. AE00's recommendation tracking system ensures that all formal AE0D recommendations are tracked until r: solution is achieved.
| |
| In addition to formal recommendations which are tracked and included in this section, additional actions are routinely implemented by NRC program offices
| |
| , on AE00 suggestions contained in engineering evaluations and special reports.
| |
| These AE00 suggestions are not forwelly tracked or closed out, and no formal
| |
| ! response is required. Section 5.1 highlights a number of examples of where action was initiated based upon AE0D suggestions.
| |
| l l
| |
| : 111
| |
| | |
| a AE00 RECOPMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AEDD/CIO1 Fesponsible AEOD Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "St. Lucie Natural Circulation Cooldown" REcoletENDATION 1 Provide a supply of cooling water to reactor coolant pump seals that will not be disabled by a single failure.
| |
| (Recommendation 4e)
| |
| RESPONSIBLE CRGR 4 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/EIB J. Jackson High 6/87 Proceeding satisfactorily.
| |
| This recommendation is included
| |
| "' in Generic Issue 65, "Probabi-lity of Core Melt due to Component Cooling Water System Failures." Generic Issue 65 comprises task 2 of Generic Issue 23 " Reactor Coolant Pump
| |
| ' Seal Failures" and will address the reliability of RCP seal cooling systems. During 1986 the technical studies and research needed for GI-23 were completed. The next step is for the staff to prepare a proposed resolution package for CRGR review.
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM I
| |
| RECOMENDATION SOURCE: Case Study AE0D/C105 RGsponsible AEOD Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980" RECOMENDATION 1 2 Installation of dual atmospheric dump valve capability for each steam generator on two-loop PWRs.
| |
| (Study recommendation 8(b)3)
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS
| |
| ; NRR/DSR0/RSIB A. Marchese High 4/87 P stcd ing satisfactorily. This 4 recommendation originally was to 4
| |
| have been addressed by a revision C to SRP 15.6.3 as part of Generic
| |
| " Issue 67.5.1, " Reassessment of Radiological Consequences Follow-ing a Postulated Steam Generator Tube Rupture." This recommenda-tion may be addressed in USI A-45,
| |
| " Shutdown Decay Heat Removal Requirements" depending on the final resolution of this issue.
| |
| In the past year a draft regula-tory analysis and backfit report that includes this reconnenda-tion as part of A-45 was prepared. This report is being reviewed by NRR prior to CRGR review.
| |
| C-105-1
| |
| | |
| f i .
| |
| i
| |
| ! AE00 REcopetENDATION TRACKING SYSTEM l
| |
| l i RECSOIENDATION :i l SOURCE: Case Study AE00/C105 (continued) !
| |
| ) ,
| |
| $ Responsible AE00 Engineer: T. Cintula i j TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980" i' ,
| |
| REC 0f08ENDATION 2
| |
| ! Review of steam generator tube rupture (SGTR) analyses for plants licensed prior to the SRP. (Study
| |
| ; recommendation 8(b)2)
| |
| I RESPONSIBLE CRGR 1 0FFICE/DIV/8R CONTACT PRIORITY REVIEW STATUS NRR/DSR0/EIB K. Shaukat Medium Not scheduled Not proceeding satisfactorily.
| |
| As a result of contacts made I
| |
| with cognizant NRR staff after '
| |
| ; O*
| |
| January 1,1987 it was learned-that this issue was inadvertently
| |
| ! left unassigned by the NRP staff i during 1986. It was to have
| |
| . been included with USI A-47,
| |
| " Safety Implications of Control j
| |
| : Systems." Shortly after contact I was made the recomunendation was 7 assigned a contact and it may now be addressed as part of GI-67, " Reevaluation of SGTR ..
| |
| Design Basis Event." However, j ongoing SGTR programs including GI-67 are being integrated into l a single generic issue (GI-135).
| |
| This reccounendation may be in-- -
| |
| ! cluded in the integrated program.
| |
| i
| |
| ^
| |
| ' C-105-2
| |
| | |
| ! AE00 REcoIWENDATION TRACKING SYSTEM e j RECOBSENDATION SOURCE: Case Study AE0D/C105 (continued) 4 i Responsible AE00 Engineer: T. Cintula ,
| |
| j TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980" REC 0erENDATION 3 i Revise SRP 9.2.2 to clarify isolation of nonsafety-related portions of service water system.
| |
| (Study recommendation 8(a)6)
| |
| ., RESPONSIBLE CRGR j 0FFICE/DIV/8R CONTACT PRIORITY REVIEW STATUS i NRR/DSRO/SPEB L. Riani N/A N/A Resolved. Standard Review
| |
| - Sections 9.2.1 and 9.2.2 were
| |
| , revised to address this l C recommendation. The revised 4
| |
| * SRPs were issued in June'1986.
| |
| i 4
| |
| 9 i
| |
| 1 e
| |
| 4 I
| |
| i .C-105-3 i
| |
| | |
| 1 AE00 REcopetENDATION TRACKING SYSTEM RECONKENDATION SOURCE: Case Study AEOD/C105 (continued)
| |
| Responsible AEOD Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980" RECOMMENDATION 5 IST of check valves in the instrument air system used to isolate safety-related portions of the system.
| |
| (Study reconmendation 8(a)2)
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS 2
| |
| NRR/DSR0/SPEB W. Milstead Low / drop N/A Not proceeding satisfactorily.
| |
| 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 this prioritization.
| |
| l NRR agreed to reevaluate the prioritization when a comprehensive AEOD case study on air system problems was issued. In December 1986 AEOD issued for peer review a preliminary case study on this sub-ject. The final report is expected 4
| |
| to be issued in early 1987 at which time we would expect NRR to reassess the priority of this recommendation.
| |
| C-105-4
| |
| | |
| AE00 RECC9mENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Memorandum to Harold Denton from C. J. Heltemes, dated May 2, 1983 R2sponsible AEOD Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| Response to NRR Comments on AE0D Report, "Calvert Cliffs Unit 1 Loss of Service Water on May 20, 1980"
| |
| ]
| |
| 4 RECOMMENDATION 1 Accessibility of ADVs for local manual operations for RCS cooldown following a steam generator tube rupture.
| |
| RESPONSIBLE CRGP OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS
| |
| ; NPR/DSR0/PSIB A. Marchese High 7/87 Proceeding satisfactorily.
| |
| ! This recommendation is currently i included in USI A-45. A
| |
| :: site walkdown survey of nine
| |
| -a 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.
| |
| In 1986 a draft regulatory and backfit report that addresses this recommendation has been prepared. This report is currently undergoing internal NRR staff review prior to being forwarded to CRGR.
| |
| C-105-5
| |
| | |
| k AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Memorandum dated January 20, 1982 from C. Michelson to Harold R. Denton R:sponsible AEOD Engineer: M. Chiramal/F. Ashe TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Safety Concerns Associated with Peactor Vessel Level Instrumentation in BWRs" I i
| |
| RECOMMENDATION 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 CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB J. Joyce High Not scheduled Proceeding satisfactorily.
| |
| Assigned as Generic Issue 101.
| |
| RECOMMENDATION 2 Q
| |
| Protective functions of narrow range level instrumentation must be assured in spite of adverse control system protection system interaction.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/EIB A. Szukiewicz High Not scheduled Proceeding satisfactorily.
| |
| On-going USI-A47 and Generic Issue 101.
| |
| | |
| ==REFERENCE:==
| |
| Memo dated March 19, 1982 from H. R. Denton to C. Michelson.
| |
| -C-201-1
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C202 R;sponsible AE0D Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Flow Blockage by Bivalve Mollusks at Arkansas Nuclear One and Brunswick" RECOMMENDATION 1 Capability to measure cooling water flow should be provided for all safety-related equipment.
| |
| RESPONSIBLE CRGR
| |
| : OFFICE /DIV/BR CONTACT PRIORITY REVIEW
| |
| * STATUS NRR/DSR0/EIB C. Hickey Medium 1/88 Proceeding satisfactorily.
| |
| This recommendation has been included in Generic Issue 51,
| |
| > - " Reliability of Open Cycle C$ Service Water Systems." During 1986 a contractor study which includes this recommendation has been approved (FIN-B-2977),
| |
| funded and initiated. The final contractor report is scheduled for completion by 10/87.
| |
| RECOMMENDATION 2 Develop and implement biofouling control strategies.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW
| |
| * STATUS NRR/DSR0/EIB C. Hickey Medium 1/88* Proceeding satisfactorily.
| |
| (See status of Recommendation I above.)
| |
| *May not be required. If necessary, will be scheduled following completion of Generic Issue 51, " Proposed Requirements for Improving the Reliability of Open Cycle Service Water Systems." Generic Issue 51 is incorporating Task V of Pro.iect FIN B-2977 by RES.
| |
| C-202-1
| |
| | |
| AEOD RECOMPFNDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C202 (continued)
| |
| RGsponsible AEOD Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Flow Blockage by Bivalve Mollusks at Arkansas Nuclear One and Brunswick" RECOMMENDATION 3 Periodic inspection of service water system piping.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW
| |
| * STATUS KRR/DSR0/EIB C. Hickey Fedium 1/88 (See status of Recommendation 1 above.)
| |
| ,. RECOMMENDATION 4 Periodic verification of overall heat transfer coefficier.t on multiple pass heat exchangers.
| |
| RESPONSIBLE CRGP OFFICE /DIV/BR CONTACT PRIORITY REVIEW
| |
| * STATUS NRR/DSR0/EIB C. Hickey Medium 1/88 (See status of Recommendation I above.)
| |
| *May not be required. If necessary, will be scheduled following completion of Generic Issue 51, " Proposed Requirements for Improving the Reliability of Open Cycle Service Water Systems." Generic Issue 51 is incorporating Task V of Project FIN B-2977 by RES.
| |
| C-202-2
| |
| | |
| AE00 RECOMMEFDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C202 (continued)
| |
| R2sponsible AEOD Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Flow Blockage by Bivalve Mollusks at Arkansas Nuclear One and Brunswick" RECOMMENDATION 5 Periodic verification of cooling water flow to all safety-related equipment should be specified in technical specifications.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW
| |
| * STATUS NRR/DSR0/EIB C. Hickey Medium 1/88 (See status of Recommendation 1 above.)
| |
| C
| |
| *May not be required. If necessary, will be scheduled following completion of Generic Issue 51, " Proposed Requirements for Improving the Reliability of Open Cycle Service Water Systems." Generic Issue 51 is incorporating Task V of Project FIN B-2977 by RES.
| |
| C-202-3
| |
| | |
| AE0D RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AEOD/C203 and Pemorandum dated May 28, 1982.
| |
| RGsponsible AEOD 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 overload protective devices associated with safety-related valve motor operators should be reassessed.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW
| |
| * STATUS NRR/DSR0/EIB 0. Rothberg Medium Not scheduled Proceeding satisfactorily.
| |
| E3 This issue is being addressed in Generic Issue II.E.6, "In-Situ Testing of Valves," which is being evaluated under contract with BNL. This issue will also be further evaluated by NRR as part of their review of the final AEOD Case Study on MOV performance, which was issued in December 1986.
| |
| | |
| 1 AE00 RECOMMENDATION TRACKING SYSTEM i REC 0f9fENDATION SOURCE: Case Study AE00/C203 and Memorandum dated May 28, 1982 (continued).
| |
| Responsible AE00 Engineer: E. J. Brown TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Survey of Valve Operator - Related Events Occurring During 1978, 1979, and 1980" (See also Reconeendation 1 on page S-503-1)
| |
| REC 0f94ENDATION 2 1
| |
| Improved methods and procedures for the setting of torque switches should be developed and evaluated relative to valve operability and functional qualification.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS l' NRR/DSR0/EIB 0. Rothberg Medium Not scheduled Proceeding satisfactorily.
| |
| Based on a letter from the g Director, NRR dated 2/23/83, i
| |
| an existing RES program was j expanded to address this sub-j ject. This has been included 1 in RES contract B3050 ." Valve Performance Testing."
| |
| 1 l
| |
| +
| |
| 1 C-203-2
| |
| | |
| ~ - .- .
| |
| l AEOD RECOMPENDATION TRACKING SYSTEM REC 0frENDATION
| |
| ! SOURCE: Case Study AE00/C203 and Memorandum dated May 28, 1982 (continued).
| |
| Responsible AE00 Engineer: E. J. Brown TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Survey of Valve Operator - Pelated Events Occurring During 1978, 1979, and 1980" (See also Recomunendation 1 on page S-503-1)
| |
| RECOMMENDATION 3 Signature tracing techniques should be developed and tried on selected motor-operated valves as part of the periodic inservice testing program.
| |
| RESPONSIBLE CPGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/EIB 0. Rothberg Medium Not scheduled Proceeding satisfactorily.
| |
| U IE/DEPER/EGCB R. Kiessel
| |
| * This was included as part of a proposed draft plan (not yet approved) for Generic Issue II.E.6, "In-Situ Testing of Valves." It will be covered by research programs. A user request memorandum, dated 5/14/84, from the Director, NPR to the Director, RES addressed this item. Valve testing was completed and a final report was issued in January 1986, as NUREG/CR-4380. Signature tracing is being used by many licensees in implementing IE Bulletin 85-03.
| |
| l l C-203-3
| |
| | |
| AE00 RECOMMENDATION TRACKIMG SYSTEM l
| |
| 0 REC 00mENDATION
| |
| , SOURCE: Case Study AE00/C204 " San Onofre Unit I Loss of Salt Water Cooling Event on March 10, 1980, l dated July 1982 Responsible AEOD Engineer: H. Ornstein l TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Single Failure Vulnerability of San Onofre l's Salt Water Cooling System" RECOMMENDATION 1
| |
| , Ongoing efforts of the SEP focus on single failure vulnerability and consequences fnr the salt water cooling system and other equivalent service and cooling water systems.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS 1
| |
| NRR/DPLA/PDI E. McKenna High N/A Proceeding satisfactorily. NRR
| |
| ! ;; has reviewed SEP plants for such ta I vulnerabilities. Modifications
| |
| : have been made at SEP plants as i
| |
| appropriate. San Onofre has made several modifications and is performing a reliability analysis
| |
| ^
| |
| of the modified salt water cool-ing system to confirm the adequacy i of the system modifications. The analysis was scheduled for comple-l tion in August 1986, however, it is not yet completed and no new i completion date has been received.
| |
| 1 I
| |
| T C-204-1
| |
| | |
| AFDD RECOMMENDATION TPACFING SYSTEM RECOMMENDATION SOURCE: Memorandum, C. Michelson to Chairman Ahearne, "New Unresolved Safety 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 AEOD/C205, "AT0G as Applied to the April 1981 Overfill Event at ANO-1" Pesponsible AEOD Engineer: H. Ornstein TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Safety Implications of Steam Generator Transients and Accidents" RECOMMENDATION 1 Combined primary / secondary side blowdown should be a USI for B&W plants.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DPLB/RSB R. Jones High N/A Proceeding satisfactorily.
| |
| EI RES B. Beckner
| |
| '" R8W licensees /EPRI/MPC are
| |
| .iointly 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.
| |
| -(t-fliR-J1 _ . _ _ _ _ _ _ - - - - - - _ - - - - -
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM i
| |
| i RECOMMENDATION SOURCE: Memorandum, C. Michelson to Chairman Ahearne, "New Unresolved Safety 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 AEOD/C205, "ATOG as Applied to the April 1981 Overfill Event at ANO-1" l
| |
| l R:sponsible AEOD Engineer: H. Ornstein TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Safety Implications of Steam Generator Transients and Accidents" (continued)
| |
| RECOMMENDATION 2 TAP-A47 should focus on equipment modifications or additions to preclude SG overfill as a credible event.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/EIB A. Szukiewicz High 2/87 Proceeding satisfactorily.
| |
| > - Included in on-going Generic 0$ Issue A-47 and the B&W Owners Group reassessment. Most B&W plants have committed to install safety-grade systems to prevent I
| |
| SG overfill protection. A CRGR package, which includes a proposed generic letter for resolution of the issue, is currently being reviewed by several NRC offices.
| |
| C-205-2
| |
| - - ----------_______m_ _ _ _ _ _ .
| |
| | |
| e AEOD RECOMPENDATION TRACKING SYSTiiM RECOMMENDATION SOURCE: Case Study AEOD/C301 Responsible AEOD Engineer: M. Chiramal TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Failure of Class 1 E Safety-Related Switchgear Circuit Breakers to Close to Demand" REC 0fmENDATION 1 Provide for monitoring the status of the closing circuit of Class 1 E Circuit Breakers and for appropriately selected breakers such as diesel generator output breakers, make the status indication available to the control room operator.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/SPEB K. Kniel N/A N/A Proceeding satisfactorily. NRR
| |
| _ is re-prioritizing Generic Issue 55 g based on Reference 3.
| |
| REC 000tENDATION 2 In the short-term, licensees of operating reactors should establish regular local surveillance of Class 1 E switchgear circuit breakers to monitor the readiness status of the spring-charging motor of each unit.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/SPEB K. Kniel N/A N/A Proceeding satisfactorily.
| |
| (See status of Recommendation 1 above.)
| |
| i i
| |
| C-301-1
| |
| | |
| 1 l AE00 REC 0mENDATION TRACKING SYSTEM
| |
| : RECOMENDATION
| |
| . SOURCE: Case Study AE00/C301 (continued) t R:sponsible AEOD Engineer: M. Chiramal i TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Failure of Class 1 E Safety-Related Switchgear Circuit Breakers.to Close to Demand" RECOMENDATION 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.
| |
| RESPONSIBLE CRGR 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS. .
| |
| NRR/DSR0/SPEB K. Kniel N/A Proceeding satisfacto'rily.
| |
| L 0 (See status of Pecommendation 1). .
| |
| ;. e REC 0mENDATION 4 Shift operating personnel should receive periodic training'in the logic and operation of circuit breakers equipped-with anti-pumping controls.
| |
| i ' RESPONSIBLE- CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS- ,
| |
| l' NRR/DSR0/SPEB K. Kniel Proceeding satisfactorily. !
| |
| IE V. Thomas N/A (See status of Recommendation 1).
| |
| IE issued Information Notice'83-50 L advising licensees of the problems.
| |
| 1 This concern is being reconsidered in NRR re-prioritization of the ,
| |
| issue.
| |
| ! l I
| |
| C-301-2
| |
| ,- , . . . , - -. , 4 - - . - . - . _ _ _ _ _ _ _ _ _ . _ . . _ _ _
| |
| | |
| AEOD RECOPMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AEOD/C301 (continued)
| |
| Responsible AEOD Engineer: M. Chiramal TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Failure of Class -1 E Safety-Related Switchgear Circuit Breakers to Close to Demand" a
| |
| | |
| ==REFERENCES:==
| |
| : 1) Memo to C. H. Heltemes, Jr., from H. R. Denton, June 17, 1983 "AE00 April 1983 Report on Failure of Class 1 E Safety-Related Switchgear Circuit Breakers to Close on Demand
| |
| : 2) Memo to D. G. Eisenhut from R. L. Spessard, June 1,.1984, "Unmonitored Failures of Class 1 E
| |
| ' 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
| |
| ; g Issue No.: 55 - Failure of Class 1 E Safety-Related Switchgear Circuit Breakers to Close on Demand"
| |
| : 4) Memo to H. R. Denton from C. J. Heltemes, Jr., April 12,1985 " Generic Issue No. 55 - Failure of -
| |
| Class 1 E Safety-Related Switchgear Circuit Breakers to Close on Demand" 4
| |
| : 5) - Memo to C. J. Heltemes, Jr. from H. R. Denton, May 9,1985 "AEOD Concerns Regarding Generic Issue No. 55" t
| |
| t s
| |
| C-301-3
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE Case Study AE0D/C401 and Memorandum form C. J. Heltemes, Jr. to H. Denton, dated March 16, 1984 Responsible AE00 Engineer: S. Salah TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Low Temperature Overpressure Events at Turkey Point Unit 4" RECOMMENDATION 1 Correct the LTOP technical specifications for the 5 areas identified in the report.
| |
| RESPONSIBLE CRGR '
| |
| CONTACT PRIORITY REVIEW STATUS OFFICE /DIV/BR Ed Throm High N/A Proceeding satisfactorily. NRR NRR/DSR0/RSIB has identified this activity as Generic Issue 94. NRR/DSR0 has revised the prioritization to C "High" based on our memorandum
| |
| " dated 6/3/85. Oak Ridge National Laboratory performed an LER search and provided the results to NRR (Letter from Mays to E.
| |
| Throm dated Sept. 2, 1986). NRR has also' signed a contract with.
| |
| PNL to perform some of the work related to this generic issue.
| |
| C-401-1 1
| |
| | |
| AE0D RECOP9tENDATION TRACKING SYSTEM REC 0fetENDATION SOURCE: Case Study AE0D/C402 Responsible AEOD Engineer: H. Ornstein
| |
| . TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Operating Experience Related to Moisture Intrusion in Electrical Equipment at Commercial Power Reactors" REC 0pe!ENDATION 1 IE should revise the inspection program to ensure licensee adherence to NRC requirements.
| |
| * RESPONSIBLE CRGR
| |
| : OFFICE /DIV/8R. CONTACT PRIORITY REVIEW STATUS IE/DQASIP/0RPB M. Johnson None Proceeding satisfactorily.
| |
| 1 This-recommendation is to be M addressed.in five separate maintenance procedure-related IE inspection modules.
| |
| P.eaional consents ' have been incorporated in these modules, and are in the concurrence chain for issuance.
| |
| 1 l
| |
| I C-402-1
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM RECOMENDATION
| |
| ( SOURCE: Case Study AE0D/C403 l
| |
| f Responsible AE00 Engineer: P. Baranowsky TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "Edwin I. Hatch Unit 2 Plant Systems Interaction Event on August 25, 1982" RECOMMENDATION 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 CPGR CONTACT PRIORITY REVIEW STATUS OFFICE /DIV/BR NRR/DSR0/EIB D. Thatcher High 3/87 Proceeding satisfactorily.
| |
| ' This recommendation was E3 1 "' originally to be evaluated.
| |
| as part of Generic Issue 77,
| |
| " Backflow Protection in Common Equipment and Floor Drain Systems " In 1986 NRR consolidated GI-77 into A-17,
| |
| " Systems Interactions." The staff prepared a proposed resolution package for A-17 which predominantly involves common mode flooding of nuclear
| |
| . plant equipment. The package
| |
| ' wasttransmitted to all NRC'
| |
| . offices and ACRS for comment and is currently being revised to address the comments received in preparation for CRGR review..
| |
| C-403-1
| |
| | |
| AE0D RECOMMENDATION TRACKING SYSTEM RECOPMENDATION SOURCE: Case Study AEOD/C403 (continued)
| |
| Responsible AE00 Engineer: P. Baranowsky TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "Edwin I. Hatch Unit 2 Plant Systems Interaction Event on August 25, 1982" RECOMMENDATION 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 common-mode safety system failures.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/EIB D. Thatcher High 3/87 (See status of Recommendation 1).
| |
| C-403-2
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEP RECOMMENDATION
| |
| ; SOURCE: Memorandum from C. J. ' Heltemes, Jr. to H. Denton, dated July 23, 1984- t i
| |
| Responsible'AEOD 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 CRGR CONTACT PRIORITY REVIEW STATUS i OFFICE /DIV/BR NRR/DSR0/RSIB A. Spano High TBD Proceeding satisfactorily.
| |
| : Identified 'as Generic Issue 93. . 'i IE/DEPER/EGCB V. Hodge The Task Action Plan for GI 93 .
| |
| $k was issued on 1/22/85. The present schedule for resolution-is mid-1987. IE. Bulletin 85-01 has been issued. All PWR plants-
| |
| -are monitoring pipe temperature ,
| |
| every eight hours in the. interim.
| |
| i C-404-1 i
| |
| .I
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AEOD/C405 Responsible AE00 Engineer: S. Pettijohn TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Breaching of the Encapsulation of Sealed Well Logging Sources" RECOMMENDATION 1 Part 39 should require that well logging licensees have specific emergency procedures for handling specific source rupture incidents.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS RES/RAMBR M. L. Ernst N/A Resolved. This recommendation was included in the final re-t; vised version 10 CFR 39.
| |
| 3 0%
| |
| RECOMMENDATION 2 Part 39 should require special physical characteristics for well logging sources.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS RES/RAMBR M. L. Ernst N/A Resolved. This recommendation was included in the final re-vised version of 10 CFR 39.
| |
| C-405-1
| |
| | |
| _ m. __ _ . _ .__ ___ ._. _ __ _ .- - -__ - - - _ _ _ - _ - _ _ _
| |
| AEOD RECOMPENDATION TPACKING SYSTEM RECOMENDATION SOURCE: Case Study AE0D/C405 (continued)
| |
| , Rssponsible AE0D Engineer: S. Pettijohn TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Breaching of the Encapsulation of Sealed Well Logging Sources" RECOMMENDATION 3 Part 39 should preclude licensees from removing or attempting to remove sources from source holders-without specific authorization in the license.
| |
| RESPONSIBLE CRGR
| |
| , OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS RES/RAMBR M. L. Ernst N/A Resolved. This recommendation was included in the final re--
| |
| vised version of 10 CFR 39.
| |
| w RECOMMENDATION 5 r
| |
| The interface between the well owner, well logging company, and companies specializing in recovery operations should be defined in order to establish the regulatory responsibility and authority over recovery and well logging operations. ,
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS RES/RAMBR M. L. Ernst N/A Resolved. This recommendation was included in the final re-vised version of 10 CFR 39.
| |
| C-405-2 ,
| |
| m .---< + . _ _ _ _ _
| |
| | |
| i AEOD RECOMMENDATION TPACKING SYSTEM ;
| |
| RECOMMENDATION '
| |
| SOURCE: Case Study AEOD/C50I (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" RECOMMENDATION 2 i
| |
| Require related equipment. protection to prevent hydrogen explosions or fires in areas containing or impacting operation of safety
| |
| ~
| |
| RESPONSIBLE CRGR
| |
| ,, OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/SPEB W. Milstead Low N/A Proceeding satisfactorily. Generic Issue No. 106, " Piping and the Use 3
| |
| of Highly Combustible Gases in Vital Areas" iwas prioritized by PNL'for the staff and given a low priority... ;
| |
| i Due to increased hydrogen.use pro-posed for BWRs to. reduce oxygen in the. recirculation system for IGSCC .
| |
| control, the priority of this issue will:be re-evaluated by NRR.
| |
| i 4
| |
| b C-501-1
| |
| | |
| l AE00 RECOMMENDATION TRACKING SYSTEM-a l RECOMENDATION SOURCE: Case Study AEOD/C502 Responsible AE00 Engineer: P. Lam i
| |
| TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "0verpressurization of Emergency Core Cooling Systems in Boiling Water Reactors" REC 0 mENDATON 1 Disable the nonsafety-related air operator associated with the testable isolation check valve on the injection line in the emergency core cooling systems.
| |
| 4 RESPONSIBLE CRGR CONTACT PRIORITY REVIEW STATUS 0FFICE/DIV/BR
| |
| ! H. Woods High 1988 Proceeding satisfactorily. Assigned.
| |
| - NRR/DSR0/RSIB as. Generic Issue 105 " Interfacing M Systems LOCA at BWRs." BNL has been contracted to conduct an a
| |
| evaluation and risk analysis of the issue with the draft report due by-May 1987. -
| |
| RECOMPENDATION 2 l 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.
| |
| RESPONSIBLE CRGR CONTACT PRIORITY REVIEW STATUS OFFICE /DIV/BR H. Woods 'High 1988 Proceeding satisfactorily. Assigned NRR/DSR0/RSIB as Generic Issue 105. BNL has been contracted to conduct an evaluation and risk analysis of the issue with the draft report due by !
| |
| May 1987.
| |
| 1
| |
| | |
| l AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C502 (continued)
| |
| Responsible AE0D Engineer: P. Lam TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "0verpressurization of Emergency Core Cooling Systems in Boiling Water Peactors" RECOMMENDATION 3 Reduce human errors in maintenance and surveillance testing activities.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS
| |
| % NRR/DSR0/RSIB H. Woods High
| |
| '' 1988 Proceeding satisfactorily. Assigned as Generic Issue 105. " Interfacing Systems LOCA at BWRs." BNL has been contracted ;
| |
| to conduct an evaluation and risk analysis of the issue with the draft report due by May 1987.
| |
| RECOMMENDATION 4 Study reducing the frequency of surveillance testino of the isolation barriers of the emergency core cooling systems during power operation.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS i NRR/DSR0/RSIB H. Woods High 1988 Proceeding satisfactorily. (See status of Recommendation 3 above.)
| |
| C-502-2
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM l RECOMPENDATION SOURCE: Case Study AE00/C503 R2sponsible AEOD Engineer: H. Ornstein i
| |
| TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Decay Heat Removal Problems at U.S. PWRs" RECOMMENDATION 1 NRR assess the need for NRC requirements to improve planning, coordination, procedures, and personnel training during shutdown to ensure the availability of the DHR system.
| |
| RESPONSIBLE CRGR CONTACT PRIORITY REVIEW STATUS-OFFICE /DIV/BR g; NRR/DSR0/RSIB A. Spano High N/A Proceeding satisfactorily. Generic
| |
| - Issue 99. "RCS/RHR Suction Line Valve Interlock on PWRs" has been expanded to include this recommenda-tion. NRR has contracted with BNL-to perform a PPA on the effects of implementing this recommendation at the Zion plant. Methods of extrapola-ting the work for Zion to other plants are also to be studied.. The study is to be completed by the spring of 1987.
| |
| C-503-1
| |
| | |
| AEOD RECOPMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE00/C503 (continued)
| |
| Responsible AEOD Engineer: H. Ornstein TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Decay Heat Removal Problems at U.S. PWRs" RECOMMENDATION 2 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 CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB A. Spano High N/A Proceeding satisfactorily. .(See status of Recommendation 1.)
| |
| RECOMMENDATION 3 NRR require. licensees to improve the man-machine interfaces related to DHR operation.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB A. Spano High N/A Proceeding satisfactorily.
| |
| status of Recommendation 1.)
| |
| C-503-2
| |
| | |
| AEOD RECOPPENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C503 (continued)
| |
| R2sponsible AEOD Engineer: H. Ornstein TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Decay Heat Removal Problems at U.S. PWRs" RECOMMENDATION 4 NRR should consider DHR suction bypass lines as alternatives to redundant drop lines (if A-45 concludes that single drop line configurations are unaccaptable).
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB A. Spano High N/A Proceeding satisfactorily. (See.
| |
| status of Recommendation 1.)
| |
| }}
| |
| RECOMMENDATION 5 NRR consider removal of autoclosure interlocks to minimize loss-of-DHR events.
| |
| RESPONSIBLE CRGR OFFICE /DIV/RSIB CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB A. Spano High F/A Proceeding satisfactorily. (See status of Recommendation 1.).
| |
| RECOMMENDATION 6 NRR should address the issue of DHR system redundancy to ensure that the DHR system is available during Mode 4, and the early stages of Mode 5.
| |
| RESPONSIBLE CRGP STATUS OFFICE /DIV/BR CONTACT PRIORITY- REVIEW Proceeding satisfactorily. (See NRR/DSR0/RSIB A. Spano High N/A status of Recommendation 1.)
| |
| C-S03-3
| |
| | |
| 1 AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION j SOURCE: Case Study AEOD/C504 1
| |
| Responsible AEOD Engineer: E. Trager TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Loss of Safety System Function Events" RECOMMENDATION 1 IE should issue an information notice to feed back the results of the case study to industry.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS h$ IE/DEPER/EGCB R. Baer N/A N/A Completed. IE determined and document by memorandum dated February 1987 that an.Information Notice (IN) would not be an appro-priate vehicle for. feeding back to industry the results of the case study. This was because the wide variety of systems and problems covered in the case study would not permit reconnending corrective action (s) and because IE has issued ins and Bulletins on specific problems that have occurred in systems covered in the case study.
| |
| During 1986 and 1987, two IE Bulletins, 18 ins, and one IN C-504-1
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C504 (continued)
| |
| R:sponsible AEOD Engineer: E. Trager TITLE OP
| |
| | |
| ==SUBJECT:==
| |
| " Loss of Safety System Function Events" RECOMMENDATION 1 IE should issue an information notice to feed back the.results of the case study to industry.
| |
| PESPONSIBLE CRGP CONTACT PRIORITY REVIEW STATUS OFFICE /DIV/BR
| |
| !! IE/DEPER/EGCB R. Baer (continued) supplement have been issued to address different facets of this subject. AEOD also distributed the case study report to INPO, NSAC,.and the Owners' Groups, and to the licensees- that had experienced the events. In addition, the Performance Indicator program has included safety system failures as one of the six indicators for quarterly review. Finally, the results of the study were published in a November 1986 study by the CSNI Principal Working Group #1 (Operatin Factors) report g Experience entitledand
| |
| " Loss Human of Safety System Functions."
| |
| C-504-2
| |
| _-__ - _ _______-_i
| |
| | |
| AEOD PECOMMENDATION TPACKING SYSTEM RECOMMENDATION S0llPCE: Case Study AE0D/C504 (continued)
| |
| Responsible AEOD Engineer: E. Trager TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Loss of Safety System Function Events" RECOMMENDATION 2 NRR review the Maintenance and Surveillance Program Plan, the Human Factors Program Plan, and the INP0 training accreditation program to ensure the adequacy of training programs for all types of NPP personnel.
| |
| RESPONSIBLE CPGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS j{ NRR/DHFT/HFIB D. Jones low N/A Not proceeding satisfactorily. '[
| |
| Understaffing and other higher priority HFIB projects have delayed l work on this item.
| |
| RECOMMENDATION 3 AEOD determine whether or not to perform further evaluations of losses of ECCS injection systems events, containment spray isolation events, and the Salem loss of component cooling water event.
| |
| .! RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS AE0D E.-Trager N/A N/A Pesolved. The San Onofre 1 and Beaver Valley 1 Loss of ECCS injection systems events were evaluated in NUREG/CR-3591,
| |
| " Precursors to Potential Severe Core Damage Accidents: -1980-1981, C-504-3
| |
| | |
| i AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C504 (continued)
| |
| R;sponsible AEOD Engineer: E. Trager TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Loss of Safety System Function Events" RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS i
| |
| i' (continued)
| |
| AE0D E. Trager N/A N/A A Status Report," where they were assigned i estimated conditional core damage probabili-ties of IE-4 and 2E-6, respectively. A similar analysis using the Accident Sequence J
| |
| ~ Precursor (ASP) evaluation techniques was per-O formed on the remaining four events and they were found to have a low to moderate potential for core damage. The four estimated conditional core damage probabilities ranged-from 1E-6 to 3E-5, and were smaller than the earlier San Onofre 1 event. The monitoring of containment spray isolation events since issuance of AEOD/C504 has not indicated the need for a. specific' study on this topic.
| |
| The loss of compontat cooling water event at j- Salem has been included in an ongoing AEOD study entitled " Failures of Service Water System at Westinghouse Plants from 1981 to
| |
| '1985."
| |
| C504-4
| |
| | |
| _ . _ = - - - .
| |
| i.
| |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C505 R:sponsible AE0D Engineer: S. Pettijohn TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Therapy Misadministrations Reported to the NRC Pursuant to 10 CFP 35.42"-
| |
| RECOMMENDATION 1 4
| |
| The Office of Nuclear Material Safety and Safeguards should-communicate the information contained in this report to the affected licensees.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NMSS R. Cunningham N/A N/A Resolved. . Copies of the case study i- $$
| |
| report were transmitted to affected licensees in December 1986 by a letter summarizing the report findings.
| |
| RECOMMENDATION 2 The Office of Nuclear Material Safety and Safeguards should consider the following actions in regards to establishing quality assurance requirements for radiotherapy facilities licensed by NRC:
| |
| Contact appropriate professional organizations to encourage and support the initiation of a voluntary, industry-directed physical quality assurance program for radiotherapy facilities. We believe that the commitment of the profession 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 demonstrated at the end of two years, we recommend that NMSS initiate the necessary studies to determine whether a rulemaking is justified to require that -radiotherapy facilities licensed by NRC have quality assurance programs to insure the accuracy of patient doses.- The program should include such things as: independent C-505--I
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOPO4ENDATION SOURCE: Case Study AE0D/C505 (continued)
| |
| R:sponsible AEOD Engineer: S. Pettijohn TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Therapy Misadministrations Reported to the NPC Pursuant to 10 CFR 35.42" verification of patient does calculations and independent verification of the activity of brachytherapy sources before the sources are implanted.
| |
| The voluntary quality assurance program should contain at least the elements outlined above.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS b NMSS/FCMC R. Cunningham N/A N/A Proceeding satisfactorily. NMSS has undertaken rulemaking in regard to requiring quality assurance programs for radiotherapy facilities.
| |
| An ANPRM is scheduled to be submitted to the Commission in early 1987. It appears that this rulemaking will satisfy our concerns as stated in Recommendation 2 in regard to quality assurance for radiotherapy facilities. We will continue to monitor the progress and content of the rulemaking.
| |
| C-505-2
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C505 (continued)
| |
| Responsible AE00 Engineer: S. Pettijohn TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Therapy Misadministrations Reported to the NRC Pursuant to 10 CFR 35.42" i
| |
| RECOMMENDATION 3 10 CFR Part 35.21 should be amended to include the calibration cf beam modifiers.such as wedge filters, shaping filters, trays etc.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW. STATUS G
| |
| c) NMSS/FCMC R. Cunningham N/A N/A Proceeding satisfactorily. An ANPRM is being submitted to the Commission in early 1987. This ANPRM includes Recommendation 3.
| |
| We will continue to monitor the pronress and content of the proposed rulemaking.
| |
| RECOMMENDATION 4 1
| |
| In addition, to the extent that the NRC implements recommendation 3, the action should be made an item of compatibility for Agreement States.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS-SP W. Kerr N/A N/A Proceeding satisfactorily. (See status of Recommendation 3.)
| |
| C-505-2
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C601 Responsible AEOD Engineer: S. Petti, john TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Rupture of an Iodine-125 Brachytherapy Source at the University of Cincinnati Medical Center" RECOMMENDATION 1 IE should publish an information notice describing the event and actions taken by the licensee and manufacturer to prevent recurrence.
| |
| - RESPONSIBLE CRGR g 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS IE/DI/SMP8 H. Karagiannis N/A N/A Resolved. IE IN-86-84 was issued on September 30, 1986.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NMSS/FC/FCML K. Smith N/A N/A Proceeding satisfactorily, hHSS is working with Region III to get 3M to provide improved guidance to licensees.
| |
| C-601-1
| |
| _ _ _ _ _ _ - - _ . _ _ _ _ _ _ _ _ _ _ -_m--_
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C601 (continued)
| |
| R;sponsible AEOD Engineer: S. Pettijohn TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Rupture of an Iodine-125 Brachytherapy Source at the University of Cincinnati Medical Center" RECOMMENDATION 3 t
| |
| RESPONSIBLE CRGR 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS NMSS/FC/FCML X. Smith N/A N/A Proceeding satisfactorily.
| |
| t; NMSS is evaluating the k) prerogatives of broad licensees and the context of the recommendation to determine to what extent the recommendation might be implemented.
| |
| C-601-2
| |
| | |
| i AEOD RECOMMENDATION TRACKING SYSTEM I
| |
| RECOMMENDATION SOURCE: Case Study AEOD/C602 Responsible'AE00 Engineer: C. Hsu TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Operational Experience Involving Turbine Overspeed Trips" RECOMMENDATION 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 CRGR
| |
| {j OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB S. Diab High N/A Proceeding satisfactorily.
| |
| Included in on-going Generic Issues 122 and 124.
| |
| RECOMMENDATION 2 Avoid entrapped oil in the speed setting cylinder by establishing administrative controls for bleeding off the entrapped oil, installing a controllable dump valve, or providing indication in the control. room fnr a spinning turbine.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB S. Diab High N/A Proceeding satisfactorily..
| |
| 4 Included in on-going Generic Issues 122 and 124.
| |
| i C-602-1
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM i
| |
| RECOMMENDATION SOURCE: Case Study AE00/0602 (continued)
| |
| Responsible AEOD Engineer: C. Hsu ,
| |
| i ;
| |
| 1 TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Operational Experience Involving Turbine Overspeed Trips" RECOMMENDATION 3 l Ensure the adequacy of the existing vendor-supplied calibration-procedures for governor speed setting.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW. STATUS G
| |
| 'NRR/DSR0/RSIB S. Diab High N/A Proceeding satisfactorily.
| |
| Included in on-going Generic Issues 122 and 124.
| |
| RECOMMENDATION 4 Prevent water induction into the turbine by providing adequate provisions for condensate removal from the steam supply line to the AFW turbine. .
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB S. Diab High N/A Proceeding satisfactorily.
| |
| Included in on-going Generic Issues 122 and 124.
| |
| C-602-2
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION i
| |
| SOURCE: Case Study AE0D/C602 (continued)
| |
| R;sponsible AE0D Engineer: C. Hsu TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Operational Experience Involving Turbine Overspeed Trips"
| |
| ; RECOMMENDATION 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.
| |
| 1 $k RESPONSIBLE CRGR 4
| |
| 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/RSIB S. Diab High N/A Proceeding satisfactorily.
| |
| Included in on-going Generic Issues 122 and 124.
| |
| RECOMMENDATION 6 Issue an information notice to alert licensees of operating reactors of the findings that led to the above recommendations.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS IE/DEPER/EGCB' J. Henderson High N/A Resolved. Information Notice IE IN-86-14, Supplement I was issued on December 17, 1986.
| |
| 1
| |
| | |
| ==REFERENCE:==
| |
| Memo dated October 22, 1986 from H. R. Denton to C. J. Heltemes, Jr. .
| |
| C-602-3
| |
| | |
| AE0D REC 0FFENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C603 4
| |
| Responsible AE0D Engineer: E. Brown TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "A Review of Motor-0perated Valve Performance" RECOMMENDATION 1 Expeditiously implement the recommendations in AE00/C203 (May 1982) and AECD/S503 (September 1985).
| |
| RESPONSIBLE CRGR
| |
| ,, OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NUMARC Not assigned N/A N/A Currently under review. AEOD Case Study C603 was transmitted to NUMARC by Reference 1
| |
| -requesting that they take actions necessary to implement the recom-mendations in the case study.
| |
| In Reference 2 NUMARC agreed to i
| |
| initiate a program to address the concerns and recommendations.
| |
| 1
| |
| | |
| ==REFERENCE:==
| |
| : 1. Letter from V. Stello, NRC to Warren Owen, RUMARC dated December 10; 1986
| |
| : 2. Letter dated January 7,1987 to V. Stello from W. H. Owen C-603-1
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION .-
| |
| SOURCE: Case Study AE0D/C603 (continued)
| |
| R:sponsible AEOD Engineer: E. Brown TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "A Review of Motor-Operated Valve Performance" RECOMMENDATION 2 Require licensees to establish procedures and diagnostic capability to determine root causes of MOV failure to operate.
| |
| - RESPONSIBLE CRGR E3 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS NUMARC .Not assigned N/A N/A Currently under review. (Same as recommendation 1 above.)
| |
| RECOMMENDATION 3 Require licensees to develop a strong training program to ensure appropriate information and instructions are disseminated to operating and maintenance personnel.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NUMARC Not assigned N/A N/A Currently under review. (Same as recommendation 1 above.)
| |
| C-603-2
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AE0D/C603 Responsible AEOD Engineer: E. Brown
| |
| - TITLE OP
| |
| | |
| ==SUBJECT:==
| |
| "A Review of Motor-Operated Valve Performance" RECOMMENDATION 4 The scope of IE Bulletin 85-03 should be extended to cover all safety-related MOV assemblies required to be tested in accordance with 10 CFR 50.55a(g).
| |
| RESPONSIBLE CRGR E3 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS NUMARC Not assigned N/A N/A Currently under review. (Same as recommendation 1 above.)
| |
| ! 1 C-603-3
| |
| | |
| i AE00 RECOMMENDATION TRACKING SYSTEM i
| |
| ;. RECOMENDATION i
| |
| SOURCE: Case Study AE0D/C604 R2sponsible AE00 Engineer: M. Chiramal t
| |
| .J .
| |
| 2 TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Effects of Ambient Temperature on Electronic Components in Safety-Related 18C Systems" i
| |
| RECOMMENDATION 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 CRGR PRIORITY REVIEW STATUS 1 g .0FFICE/DIV/BR CONTACT NRR Not assigned N/A N/A Currently under review.
| |
| RECOMMENDATION 2
| |
| . Training of control room and plant operators in using the procedures should be provided.
| |
| i RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR Not assigned N/A N/A Currently under review.
| |
| 'C-604-1 d
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Case Study AEOD/C604 (continued)
| |
| R2sponsible AE00 Engineer: M. Chiramal TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Effects of Ambient Temperature on Electronic Components in Safety-Related I&C Systems" RECOMMENDATION 3 Supplemental cooling) control room cooling . equipment should be readily available and identified for use in the event (of a loss RESPONSIBLE CRGR g 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS NRR Not assigned N/A N/A Currently under review.
| |
| RECOMMENDATION 4 Periodically measure or continuously. monitor the environmental conditions inside the instrument cabinets that contain heat sensitive solid-state-components.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR Not assigned N/A N/A Currently under review.
| |
| j C-604-2
| |
| | |
| i ,
| |
| i AEOD REC 000tENDATION TRACKING SYSTEM .
| |
| i l REC 0f0tENDATION i SOURCE: Case Study AE0D/C604 (continued) l R9sponsible AE00 Engineer: M. Chiramal TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Effects of Ambient Temperature on Electronic Components in Safety-Related ISC Systems" t
| |
| RECODEENDATION 5
| |
| ]'
| |
| i 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 i safety-related instrumentation and control system cabinets located in the control room area.
| |
| g RESPONSIBLE CRGR
| |
| ~
| |
| OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS i
| |
| i i NRR Not assigned N/A N/A Currently under review.
| |
| I
| |
| : REC 00MENDATION 6 I
| |
| I In the on-going plant-specific evaluations associated with .the resolution of USI A-44, Station Blackout, the l following considerations regarding the effects of high ambient temperature on solid-state electronic
| |
| ; components should be included:
| |
| i (a) The. design adequacy should be evaluated for instrumentation and control system equipment needed to function during and recovering from a station blackout, as well as other equipment whose malfunction would. impact operability of such equipment.
| |
| li (b) Plant-specific equipment qualification data should be required unless the equipment qualification data ,
| |
| 4 can be verified by actual measurements of as-built and as-installed conditions. 1 3
| |
| .C-604-3 1
| |
| 1
| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
| {
| |
| i}!;!ii !4 j. I ' i:
| |
| | |
| AE0D RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE:_ Case Study AEOD/C605 (continued)
| |
| R2sponsible AEOD Engineer: F. Ashe TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Operational Experience Involving Losses of Electrical Inverters
| |
| , RECOMENDATION 3 Technical specifications which specifically address inverters and/or attendant buses for comparable plant designs should be reviewed to ensure that action statements addressing plant operating restrictions are consistent.
| |
| $$ RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR Not assigned N/A N/A Currently under review.
| |
| C-605-2
| |
| | |
| AE00 REC 0f#tENDATION TRACKING SYSTEM i
| |
| I RECOMENDATION SOURCE: Memo: C. Michelson to R. Mattson, "NRC Action Plan Developed as a Result of TMI-2 Accident -
| |
| Draft 3, Task III.E.3 Decay Heat Removal," April 24, 1980.
| |
| RGsponsible AE00 Engineer: H. Ornstein TITLE OR SU8 JECT: " Reliability of DHR Systems" RECOMENDATION I Reliability of DHR systems should be reviewed and where necessary upgraded on an expedited basis.
| |
| CRGR RESPONSIBLE STATUS CONTACT PRIORITY REVIEW
| |
| ; ~ OFFICE /DIV/BR S High 3/87 Proceeding satisfactorily.
| |
| NRR/DSR0/RSIB A. Marchese Included in TAP A-45. NRR is presently drafting its
| |
| ; position on this issue based l upon several contractor studies.
| |
| e l
| |
| E-001A-1
| |
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| r e
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| |
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| |
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| |
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| |
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| |
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| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATIOff 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," September 15, 1981 R:sponsible AEOD Engineer: H. Ornstein TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Plant Air Systems" RECOMMENDATION 1 Obtain licensees' experience and assessment of this problem and determine course of corrective action if required.
| |
| RESPONSIBLE CRGR PRIORITY REVIEW STATUS OFFICE /DIV/BR CONTACT i
| |
| E?
| |
| "' W. Milstead Low / Drop N/A Not proceeding satisfactorily.
| |
| NRR/DSR0/SPEB NRR originally prioritized this issue (Generic Issue 43) and recommended that it be dropped.
| |
| AE00 did not agree with the prioritization, and sent a memo (Reference) requesting NRR to hold final disposition of this issue in abeyance until a comprehensive AE00 report was written on this subject. In December 1986, AE00 issued for peer review a preliminary case study on this subject. The final report is expected to be E-123-1
| |
| | |
| AE0D RECOMMENDATION TRACKING SYSTEM RECOMMENDATION 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," September 15, 1981 (continued)
| |
| Responsible AE00 Engineer: H. Ornstein TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Plant Air Systems" RECOMMENDATION 1 (continued)
| |
| Obtain licensees' experience and assessment of this problem and determine course of corrective action if required.
| |
| RESPONSIBLE CRGR 01 0FFICE/DIV/BR CONTACT PRIORITY c' REVIEW STATUS NRR/DSR0/SPEB W. Milstead Low / Drop (continued)
| |
| N/A issued in early 1987 at which time NRR will reassess the priority of this recommendation.
| |
| I REFRENCE:
| |
| Memo from C. J. Heltemes, Jr., to H. R. Denton, " Contamination of Instrument Air Lines,"
| |
| December 14, 1983 E-123-2
| |
| | |
| i AEOD RECOMENDATION TRACKING SYSTEM
| |
| ! RECOMMENDATION SOURCE:
| |
| Memorandum from C. Michelson to R. Vollmer and R. Mattson dated February 24, 1982 l
| |
| Responsible AE0D Engineer: M. Chiramal 1
| |
| ! TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Spurious Trip of the Generator Lockout Relay Associated with a Diesel Generator Unit" I RECOMMENDATION 1 Should explicitly verify that seismic qualification of all protective devices used in the control and protection circuitry of DG units has been perforn'ed with these devices in their energized, de-energized, tripped and non-tripped states.
| |
| ~
| |
| CRGR RESPONSIBLE REVIEW STATUS
| |
| ~ CONTACT PRIORITY OFFICE /DIV/BR Proceeding satisfactorily.
| |
| T. Chang High Not scheduled 4
| |
| NRR/DSR0/EIB Issue has been incorporated into USI A-46.
| |
| 3 4
| |
| | |
| ==REFERENCE:==
| |
| Memorandum from H. R. Denton to C. Michelson dated May 11, 1982 1
| |
| I E-212-1
| |
| .i
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM 4
| |
| RECOMMENDATION SOURCE: Engineering Evaluation AEOD/E215 t Responsible AE00 Engineer: T. Cintula TITLE CP
| |
| | |
| ==SUBJECT:==
| |
| " Salt Water System Flow Blockage at Pilgrim NPS by Blue Mussels" RECOMMENDATION 1 Internal inspection of RBCCW HX supply headers RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW
| |
| * STATUS NRR/DSR0/EIB C. Hickey Medium 1/88 Proceeding satisfactorily.
| |
| Ri (See status of recommendation 1 on page C-202-1.)
| |
| RECOMMENDATION 2 Periodic measurement of. overall heat transfer coefficient on RBCCW HXs at Pilgrim RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW
| |
| * STATUS NRR/DSR0/EIB C. Hickey Medium 1/88 Proceeding satisfactorily.
| |
| (See status of recommendation 1 on page C-202-1.)
| |
| *May not be required. If necessary, will be scheduled following completion of Generic Issue 51, " Proposed Requirement for Improving the Reliability of Open Cycle Service Water Systems."
| |
| E-2IS-I
| |
| | |
| ' AE00 REC 0petENDATION TRACKING SYSTEM RECOPMENDATION SOURCE: Engineering Evaluation AE00/E215 (continued)
| |
| Responsible AEOD Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Salt Water System Flow Blockage at Pilgrim NPS by Blue Mussels" REC 000tENDATION 3 Periodic measurement of Salt Water System flow to RBCCW HXs CRGR RESPONSIBLE STATUS CONTACT PRIORITY REVIEW
| |
| * OFFICE /DIV/BR C. Hickey Medium 1/88 Proceeding satisfactorily.
| |
| NRR/DSR0/EIB (See status of recommendation 1 on page C-202-1.)
| |
| 0
| |
| *May not be required. If necessary, will be scheduled following completion of Generic Issue SI, " Proposed Requirement for Improving the Reliability of Open Cycle Service Water Systems."
| |
| E-215-2
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Engineering Evaluation AEOD/E304 Responsible AEOD Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Backflow Protection in Common Equipment and Floor Drain Systems" REC 099tENDATION 1 Provide backflow protection for drain systems in older operating plants.
| |
| : s. RESPONSIBLE CRGR 3d 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/EIB D. Thatcher High Early 1987 Proceeding satisfactorily.
| |
| This recommendation was ini-tially prioritized as high and was to be included in Generic Issue 77, " Backflow
| |
| - Protection in Common Equipment and Floor Drain Systems." In 1986 NRR consolidated GI-77 into A-17. " Systems Inter-action." The staff prepared a proposed resolution package for
| |
| , A-17 which principally involves common mode flooding of nuclear plant vital equipment spaces.
| |
| E-304-1
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM RECOMMENCATION SOURCE: Engineering Evaluation AEOD/E304 (continued)
| |
| Responsible AE00 Engineer: T. Cintula TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Backflow Protection in Common Equipment and Floor Drain Systems" RECOMMENDATION 1 (continued)
| |
| Provide backflow protection for drain systems in older operating plants.
| |
| RESPONSIBLE CRGR CONTACT PRIORITY REVIEW STATUS OFFICE /DIV/BR
| |
| - (continued) di D. Thatcher High Early 1987 The package was transmitted to NRR/DSR0 all NRC offices and ACPS for comment and is currently being revised to address the comments received in preparation for CRGR review.
| |
| t 4
| |
| E-304-2
| |
| | |
| i
| |
| , AEOD RECOMPEFDATION TRACKING SYSTEM i
| |
| RECOMMENDATION SOURCE: Special Study Report - C. J. Heltemes, Jr. to H. R. Denton dated January 13, 1984 and follow-up
| |
| ; memorandum dated August 8,1984 4
| |
| Responsible AEOD Engineer: E. Trager TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Human Error in Events Involving Wrong Unit or Wrong Train" i
| |
| ; RECOMMENDATION 1 Consider the need for further clarification or guidance on what constitutes an acceptable independent j
| |
| verification program.
| |
| n.
| |
| SI RESPONSIBLE CPGR j OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS I
| |
| NRR/DHFT/MTB G. Cwalina High N/A Proceeding satisfactorily.
| |
| (See status of recommendation 2 below.) The reference stated that "The NRC should provide clarifying guidance regarding' independent j
| |
| ' verification." NRR plans to request that RES prepare regulatory guidance on.
| |
| activities such as independent verification programs that i
| |
| help to minimize the potential for the occurrence of these J
| |
| S-401-1
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM
| |
| ,' REC 0094ENDATION SOURCE: Special Study Report - C. J. Heltemes, Jr. to H. R. Denton dated January 13, 1984 and follow-up memorandum dated August 8, 1984 (continued)
| |
| Responsible AEOD Engineer: E. Trager
| |
| (
| |
| TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Human Error in Events Involving Wrong Unit or Wrong Train" 4
| |
| i RECOMMENDATION 1 (continued) 4 RESPONSIBLE CRGR CONTACT PRIORITY REVIEW STATUS OFFICE /DIV/BR j (continued)
| |
| G. Cwalina High N/A types of events. Action on O NRR/DHFT this recomunendation will be t
| |
| considered complete when the NRR request has been forwarded to RES. AE00 will continue to l
| |
| ' monitor the rate of occurrence of these types of events.
| |
| I
| |
| | |
| ==REFERENCE:==
| |
| "An Investigation of the Contributions to Wrong Unit or Wrong Train Events," NUREG-1192, April.1986.
| |
| 4 S-401-2
| |
| | |
| ._ - -____._ ._.__ _ _ . _ _ . _ . . _ _ _ _ _.. _ - . _ ~ _ _ . . . . _ _ __ __ . _ .
| |
| , AEOD REC 0WENDATION TPACKING SYSTEM RECOMENDATION i SOURCE:
| |
| j Special Study Report - C. J. Heltemes, Jr. to H. R. Denton dated January 13, 1984 and follow-up memorandum dated August 8, 1984 (continued)
| |
| I Responsible AEOD Engineer: E. Trager TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Human Error in Events Involving Wrong Unit or Wrong Train" REcom ENDATION 2 NRR review wrong unit / wrong train events and develop appropriate guidance to minimize such events.
| |
| RESPONSIBLE CRGR l C 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS
| |
| ; m NRR/DHFT/MTB G. Cwalina High N/A Proceeding satisfactorily.
| |
| 1 NRR and AE00 staff completed a ,
| |
| s study of the factors that appeared to have contributed to these types of events. In April 1986 the results of the study were published in '
| |
| NUREG-1192 and were forwarded to INPO and MUMARC for i'
| |
| consideration by industry.
| |
| [In a meeting on August 15, s
| |
| 1986 with DNFT, NLSIARC described industry efforts-i (e.g., procedure and labeling improvements) to S-401-3
| |
| (
| |
| I
| |
| | |
| 1
| |
| +
| |
| i AE00 REttWWEENDATION TRACKING SYSTEM i
| |
| 2 RECWWENDATION .
| |
| I SOURCE: Special Study Report - C. J. Heltames. Jr. to H. R. Denton dated January 13, 1984 and follow-up memorandum dated August 8.1984 (continued) l1
| |
| [
| |
| Responsible AE00 Engineer: E. Trager
| |
| ) TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Human Error in Events Involving Wrong Unit or Wrong Train" l
| |
| T
| |
| ! REC (NSENDATION 2 (continued) i RESPONSIBLE CRGR CONTACT PRIORITY REVIEW STATUS q OFFICE /DIV/BR (continued) reduce the occurrence of these O MRR/DHFT/MTB G. Cwalina High N/A '
| |
| .l
| |
| * types of events.] The results i
| |
| of the study were also consider-ed while developing site visit protocols for use in Phase II of the MSPP during which
| |
| .: industry actions to resolve j the noted problems will be monitored. MRR has not been-l satisfied with the progress being made by industry and r
| |
| plans to request that RES i
| |
| provide regulatory guidance in
| |
| ' this area. Action on this
| |
| -recommendation will be con-
| |
| $ sidered complete when the NRR request has been issued to '
| |
| RES. AE00 will continue to j track the rates of occurrence I of these types of events.
| |
| ! S-401-4
| |
| | |
| AEOD RECOP9tFNDATION TRACKING SYSTEM RECOMENDATION SOURCE: Memorandum - C. J. Heltemes, Jr., to R. DeYoung, May II,1984 Responsible AEOD Engineer: S. Rubin l
| |
| TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Pressure Locking of Flexible Disk Wedge Type Gate Valves" REco mENDATION 1 Give isumediate consideration to issuing a bulletin on the subject of pressure locking of flexible disk medge-type gate valves.
| |
| RESPONSIBLE CRGR g 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS IE/DEPER R. Singh High N/A Resolved. IE initially pre-pared a draft bulletin as part of a CRGR review package.
| |
| Shortly thereafter, it was learned that INPO had begun work on an 50ER on this subject.
| |
| AE00 and IE agreed to allow IMPO to address the concerns, and activity by IE on the bulletin was suspended with AE00 agreement. On 12/I4/84, INP0 issued SOER 84-7, " Pressure Locking and Thermal Binding of Gate Valves." This document S-402-1
| |
| | |
| _ - - - _ - - . = - - - -. -_- . . . _ . . _ . - - . . . . -
| |
| i AE00 RECOMIENDATI0r. TpACKING SYSTEM l
| |
| REC 0fetENDATION SOURCE: Memorandum - C. J. Heltenes, Jr., to R. DeYoung, May 11, 1984 Responsible AE00 Engineer: S. Rubin TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Pressure Locking of Flexible Disk Wedge-Type Gate Valves" RECONtENDATION 1 (continued)
| |
| RESPONSIBLE CRGR CONTACT PRIORITY REVIEW STATUS OFFICE /DIV/BR (continued)
| |
| IE/DEPER R. Singh High N/A was found by AE00 and IE to be -
| |
| - an acceptable substitute for E the bulletin originally proposed.
| |
| IE and AE00 have monitored the LERs, and 50.72 reports for adequacy of corrective actions taken in response to the IMPO SOER. A review of the 1986 LER and RPRDS data related to this issue indicates that the 50ER has been effective in addressing this issue and '
| |
| therefore no further NRC actions are required.
| |
| S-402-2 i
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECCMMENDATION SOURCE: Special Study Report AEOD/S503 and memorandum, C. J. Heltemes, Jr. to H. Denton dated September 19, 1985. (See also recommendations related to AEOD/C203 on pages C203-1 and C203-2). l l
| |
| Responsible AEOD Engineer: E. Brown l TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| " Evaluation of Recent Valve Operator Fotor Burnout Events" REC 0fetENDATION 1 In view of the more than 200 motor burnout events, the NRR plan to address motor burnout should be expedited.
| |
| - RESPONSIBLE CRGR E OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS NRR/DSR0/EIB 0. Rothberg Medium N/A Proceeding satisfactorily.
| |
| This issue will be addressed in Generic Issue II.E.6, "In-Situ Testing of Valves,"
| |
| which is being evaluated under contract with BNL. As of mid-April 1986, the contract with BNL had been funded, approved and initiated. This issue will also be further evaluated by NRR during their review of the final AEOD Case Study on MOV performance which was issued in December 1986.
| |
| S-503-1
| |
| | |
| l' AE00 RECOMENDATION TRACFING SYSTEM 1
| |
| i
| |
| }
| |
| REC 0 mENDATION
| |
| ; SOURCE: Sps:ial Study Report AE00/5602 i
| |
| i Responsible AE00 Engineer: J. L. Crooks l
| |
| 1
| |
| ' TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "An Overview of Nuclear Power Plant Operating Experience Feedback Programs" REC 0mENDATION 1
| |
| ) (1) IE should consider issuing an Infor1 nation Notice (IN) to licensees to communicate the conclusions from
| |
| ; this AEOD study and to suggest t845t licensees review their operating experience (0E) activities 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 5 activities are effective and reemphasize that the PPC considers the OE activities important to the safety
| |
| " of operations. The findings from the three IIT reports can be cited in this regard. Specifically, the i 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 in-house events, (b) Be diligent in the screening and assessment of industry feedback 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
| |
| )
| |
| j support staff both directly and 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 activities through management i reviews.-
| |
| i -
| |
| S-602-1 - ~
| |
| | |
| AE00 RECOMMENDATION TRACKING SYSTEM 1
| |
| RECOMMENDATION SOURCE: Special Study Report AE00/S602 (continued)
| |
| R:sponsible AEOD Engineer: J. L. Crooks TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "An Overview of Nuclear Power Plant Operating Experience Feedback Programs" RECOMMENDATION 1 (continued) a RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS IE/EPER/EGCB V. Hodges On hold N/A In July 1986, an IN was
| |
| ' drafted that addressed S602 Es
| |
| '' findings and conclusions, and an IE contract effort was being initiated to determine options for improving generic communications with licensees The IE contract was let with -
| |
| completion due late in FYI987.
| |
| The IN was not issued.
| |
| ~ The IE contract includes interviews with NRC Headquarters and Regional Staff, licensees, INPO and i
| |
| others.
| |
| Further actions are pending
| |
| ; completion of the contract.
| |
| S-602-2
| |
| | |
| AE00 RECOMMENDATION TPACKING SYSTEM RECOMMENDATION SOURCE: Special Study Report AEOD/S602 (continued)
| |
| R;sponsible AE0D Engineer: J. L. Crooks TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "An Overview of Nuclear Power Plant Operating Experience Feedback Programs" RECOMMENDATION 2 AEOD should have further discussions with the industry on additional initiatives, such as on means to increase the effectives of OE review activities, and on assessing the effectiveness of the OE review activities.
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS AEOD/PTB J. Crooks Continuing N/A Met with INPO in February 1987 Activities to discuss their initiatives in the area. One initiative was focused on the implementa-tion of prinrity corrective actions from SOERs. Other 4 improvements were made to program guidance and the evaluation techniques for assessing utility OE programs.
| |
| PTB also included discussions on licensee's OE programs during plant visits related to the "new plant" study. In addition, discussions were S-602-3
| |
| | |
| .i
| |
| ; AEOD RECOMMENDATION TRACKING SYSTEM ;
| |
| RECOMMENDATION SOURCE: Special Study Report AE0D/S602 (continued) t Responsible AEOD Engineer: J. L. Crooks
| |
| . TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "An Overview of Nuclear Power Plant Operating Experience Feedback ' Programs" RECOP91ENDATION 2 (continued)
| |
| RESPONSIBLE CRGR 0FFICE/DIV/BR CONTACT PRIORITY REVIEW STATUS c AEOD/PTB J. Crooks Continuing N/A held with licensees who had !
| |
| read the report and wanted l E more information. Future *
| |
| ' plans are to visit additional licensee facilities and to ,
| |
| meet further with INPO.
| |
| RECOMMENDATION 3 The NRC should continue to monitor operating experience and industry performance in OE activities for some '
| |
| time (2 years or so). If, after this period of time, sufficient improvement in effectiveness has not been observed, then further regulatory action would be appropriate. In this regard, AE00 believes the following
| |
| -actions should be initiated at this time:
| |
| (a) NRR, in conjunction with AE00 and IE, 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 determinations 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 S-602-4
| |
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| _ . _~ . . __ ._ . _ _ _ . _ . _ . - _ _ . _ . _ .. .
| |
| j AE00 RECOMMENDATION TRACKING SYSTEM-1 RECOMMENDATION SOURCE: Special Study Report AE0D/S602 (continued)
| |
| Responsible ~AE00 Engineer: J. L. Crooks TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "An Overview of Nuclear Power Plant Operating Experience Feedback Programs" RECOMMENDATION 3 (continued)
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS AE00/PTB J. Crooks Various N/A communications, prior to
| |
| ,. NRR/0 RAS F. Brenneman developing ~ additional 8 IE/DI/0RPB P. McKee guigance.
| |
| IE has modified several" pro-ceduresEto better address i the effectiveness of OE feedback.
| |
| RECOMMENDATION 4 Regarding OE feedback, as soon as practical, NRR, IE, and AE00 should consider consolidating existing systems (e.g., Generic Letters that deal with operational events Information. Notices'that' discuss operational events, Power Reactor Events) for transmitting operational experience into a single "NRC Notice" system. This
| |
| " notice" would constitute the sole. system for the feedback of operating experience to licensees for appropriate action and closeout. 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, and could. be sent to a single standard distribution list. In addition, the system could make. sufficient 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 these " notices" via an electronic mail: system.
| |
| S-602-6
| |
| | |
| AEOD RECOMMENDATION TRACKING SYSTEM RECOMMENDATION SOURCE: Special Study Report AE0D/S602 (continued)
| |
| Responsible AEOD Engineer: J. L. Crooks TITLE OR
| |
| | |
| ==SUBJECT:==
| |
| "An Overview of Nuclear Power Plant Operating Experience Feedback Programs" RECOMMENDATION 4 (continued)
| |
| RESPONSIBLE CRGR OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS AEOD/PTB J. Crooks On hold N/A further actions are pending NRR/0 RAS F. Brenneman the results of the IE contract Es IE/EGCB V. Hodges for options on improvements to
| |
| * generic communications.
| |
| RECOMMENDATION 5 In conjunction with the implementation of recommendation-(4) above, AEOD should: (1) terminate the publication of Power Reactor Events, and (2) terminate the publication of.the monthly Licensee Event Report (LER) Compila-tion, unless there are conflicting statutory or legal requirements or sufficient subscriptions to offset the cost.
| |
| RESPONSIBLE CRGP OFFICE /DIV/BR CONTACT PRIORITY REVIEW STATUS AE00/PTB J. Crooks low N/A Both publications are continuing. Termination is at present contingent on the consolidation efforts in (4) above, which is pending completion of the IE contract.
| |
| S-602-7 I
| |
| | |
| Af0D RECOMMENDAfl0N fpACKIN8 EViffM
| |
| !! 8pesial 8tudy Report At00/8608 (Meme dated dune 6, 1986 frns F. Hebden to R Vollmer)
| |
| Reipensible AROD Ingineeri L Lasds fifMJHL1Hdiffi "Adequesy of the isspe of It Bulletin 86=01" 11105UDAU0lL1 h!s!! hushes!g!sd'PWT!afetff!le!5!s!pul" ' ' '
| |
| g ob!!!bhBR CONTACT REDRHY llVith UB It/DEPER/tAl H iciley N/A 10/96 r8 Roselved,fnOil'ishfne4=1188, gtiiivedi i I
| |
| 4 vill'$i'!<!nei '""
| |
| rei el =le ifMe'"
| |
| gern!d v.ive ''
| |
| lfn!.gg4 nes end uen,
| |
| "!.ss;i,11:l'"'"""
| |
| i=60l=1
| |
| | |
| AMPE M E A SEEMIttir SF 1!41RE ABIBEM. MEN DInstindlihms h 4ppened! Ily illle ammiissiimm and lause stIn inn ime Staff IBlfWliamKammiissilem Appramm11 Olmiin@
| |
| | |
| Abnormal Occurrences CY 1986 i
| |
| Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Comments 86-1 Loss of Power and Vol. 9, No. 1 G-2 On November 21, 1985, San Onofre Nuclear Water Hammer Event Generating Station (SONGS) Unit 1 experienced a' partial loss of inplant ac electrical power while the plant was operating at 60 percent power.
| |
| Following a manual reactor trip, the plant lost all inplant ac power _ for 4 minutes and experienced a severe water hammer'in the feedwater system which caused a leak, damaged plant equipnent, and challenged the integrity of the plant's heat sink. The most significant aspect of the event y- involved the failure of five safety-related check n>
| |
| valves in the feedwater system, without detection, which jeopardized the integrity of safety systems.
| |
| The event involved a number of equipment malfunctions, operator errors, and procedural deficiencies.
| |
| The' incident was investigated by an NRC Incident Investigation Team.
| |
| 86-2 Loss of Integrated Vol. 9, No. 1 G-3 On December 26, 1985, Rancho Seco Nuclear !
| |
| Control System Generating Station experienced a loss of de power Power and Over- within the integrated control system.(ICS) while cooling Transient .the plant was operating at 76 percent power.
| |
| Following.the loss of.ICS de power, the reactor tripped on high reactor coolant system (RCS) pressure followed by a rapid overcooling transient and automatic initiation of the safety features actuation system on low RCS pressure. The overcooling transient continued until ICS de power was restored 26 minutes after its loss. The significance of the event is that a nonsafety-
| |
| | |
| Abnormal OccurrIncts CY 1986 ,_
| |
| Report No. A0 Criterion . .
| |
| or Example Comments A0 # Title of A0 NUREG-0090 related system failure initiated a plant transient which could have been more severe under other postulated scenarios.
| |
| The incident was investigated by an NRC Incident Investigation Tean.
| |
| 86-3 Rupture of a Vol. 9, No. 1 G On January 4,1986, a cylinder filled with uranium Uranium Hexa- and hexafluoride (UF-6) ruptured while it was being
| |
| ~
| |
| fluoride Cylinder A-11 heated in a steam chest at the,Sequoyah Fuels l'"' and Release of Corporation's Sequoyah Facility near Gore, Gases Oaklahoma. One worker died from pulmonary edema caused by inhalation of hydrofluoric acid, a
| |
| ' reaction product of UF-6 and airborne moisture.
| |
| Much of the facility complex and some offsite areas to the south were contaminated with hydrofluoric acid, and a second reaction product, uranyl fluoride. The interval of release was
| |
| ' approximately 40 minutes.
| |
| The incident was investigated by an NRC Augmented Investigation Team.
| |
| 86-4 Therapeutic Med- Vol. 9, No. 1 G On February 7, 1986, a patient at Washington ical Misadminis- Hospital Center, Washington, D.C. received a tration cobalt-60 teletherapy treatment of 150 rads to the abdomen, which was intended for another patient.
| |
| 86-5 Overexposure to a Vol. 9, No. 1 A-2 On February 19, 1986, while checking a licensee Member of the which had apparently ceased operations, an NRC Public from an Region III inspector determined that an industrial Industrial Gauge gauge, containing a sealed source of cobalt-60,
| |
| | |
| Aine.i Deen,rense, er i a
| |
| ! A0 # Title of A0 NNN0090 e tha he Coments was in an unrestriated area of the former fastery
| |
| , lub detemined tha siteiiwe.sequentinstcet,teliisreesivedes, isa m.o siv tat a reivii is ndiaiien a.ri of he J ihe imge,e, dii,eJfi IIaSt T!'a R!ilin'e fe!RNJT iiheuMi ! nit"letR3'"f. t'''""'' ''" '''"ijnle M, M6 Breakdown of Vol. 9 No. 1 A 11 On March a 1986, the NRC 15 i ed an Order luspending P Management Controls Lisense Effsetive imediit ) la Radialien
| |
| * at an Irradiator Teshnolog(y, interparated R of Reshawap N Jersoy, Facility The Order wel based en NR nepostions wh sh n if:ed a number of insterees of hypall:ny gefety inter e k these ins tealed a l' systemsi's 11eensee management sentre'gnifisant system, breakdown n the 84 7 Tritium Over- Vol. 9. No. 1 A 11 During ing! gen on
| |
| " Marsh lli 80atPg'eril
| |
| '"U'Tidi!Jn''
| |
| r ila ne, a tie '
| |
| sien asuiiig'*leni!,yad re e
| |
| is sureofirtn.ser(eeareher e
| |
| /peuting'lais ivumpegn e viva eni is a rete:ved an ug'NirT,8fT!se.d 4
| |
| 2!UI of ,4 ,re nei re l fr
| |
| .lfli!.:iessiive .14.e'iti"t ra A
| |
| re i.ntified,
| |
| .dit!'une"Jd !4, l A584 1 RadiationIn,furp Vol. 9, No. 1 A=1 On Apri' foi 1984 en indinidual emplo of an Industria inspeelo on servise,s in Mid'and, fenas,yed byanIF resolved
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| | |
| Abnemal Occurrences CY 1984 l
| |
| ; A0 # Title of A0 N!IG0!ss h $ p5e " Coments Radiographer radi li b n asi e n uensee of reaas
| |
| ' " ^ '"'"" ' ' " " I eNposurethatresultedinlamin gerfaming niit"lini !!!'eiTeb rese knewtheradiographe! ved a red atten burn. The burn Wan reported by Pem'an Lndustrial X Ray present emp' eyer of the indiv' dual, to the Agency, on Nov eber 8 1985 When the radfegrapher had an apparent recurrence of the Wound.
| |
| $ Theradiogra!chis'efthandandabe!t!?resher 29,000 rems map have reso Whole body exposure fra the enpened soures.
| |
| Asse-t contamination of Vol. 9 No. 1 0 On May 24, 1985, it was diseevered that some a sera 5 eel facilities of fames Bleel Company of Ontartei Facili licensee Californi i ereW sentaminated Wilh redigast ye of Cali o nia, an material later detemined la be aesfWm), peerently, t
| |
| AgreementState) the sesfu Was part of(delsmined a rodisaative devise to be er abeus sours les sensuri r s)te ned in a scrap stes shipment to the planti he rad's=
| |
| activity was released When the scrap meta was melted un a furness.
| |
| AS86-3 Radiation Injury Vol. 9. No. 1 A1 of an Industrial OnAugustti,diatio,nindryfromansneeled1985 received a ra an industr Radiographer (licenseeofCali- seures 2000 redsof)his and aleftwhohond e b dyestimated to be evereupesure ybout l ash' mated fornia, an A At tas time of the :ne denti mentState) gree- tobeabout6(reds).the employes mployed by feethe=f
| |
| | |
| Abnormal Occurrences CY 1986 I
| |
| Report No. A0 Criterion
| |
| . A0 # Title of A0 NUREG-0090 or Example Coments was performing radiography at the company's field site in the Kern River oil field in Bakersfield, i California. He was using a 46-curie iridium-192 source contained in a radiographic projector.
| |
| l AS86-4 Radiation Injury Vol. 9, No. 1 A-1 On November 9, 1985 an individual employed as of an Industrial an assistant radiographer by Basin Industrial Assistant Radio- X-Ray in Odessa, Texas, received a radiation '
| |
| grapher (licensee of burn of his left hand and an estimated 129 rems
| |
| , Texas, an Agreement whole body exposure. The licensee failed to a State) notify the Texas Bureau of Radiation Control (Agency) of the incident. Another licensee informed the Agency on November 26, 1985 that an
| |
| ; incident had occurred involving Basin Industrial X-Ray.
| |
| The estimated exposure to the employee's hand from the exposed source may have been as high as 30,000 rems, or even considerably higher.-
| |
| 86-8 Out of Sequence Vol. 9, No. 2 A-11 On March 18, 1986, during a startup of Peach Bottom 4
| |
| Control Rod With- Unit 3, personnel errors by four licensed operators drawal resulted in a control rod being withdrawn out-of-sequence without being detected by these operators.
| |
| The next operating shift detected the error and manually scramed the unit.
| |
| 86-9 Boiling Water Vol. 9, No. 2 G-3, On May 19, 1986, the Boston Edison Company (BECO) i' Reactor Emergency A-10, notified the NRC that a significant design deficiency Core Cooling System and in the residual heat removal (RHR) system minimum
| |
| . 2 _ - _ _ _ _ _ _ _ _ _
| |
| | |
| l l
| |
| Abnormal Occurrtnces CY 1986 Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Coments Design Deficiency A-12 flow protection logic at the Pilgrim Nuclear Power Station (PNPS) had been discovered. Later, it was found that some other GE-designed BWRs (Dresden 2, 3 and Quad Cities 1, 2) also contained the same design deficiency.
| |
| A similar deficiency was also discovered in some PWRs. This was discussed in an update to A0 86-9 in NUREG-0090, Vol. 9, No. 3.
| |
| s.
| |
| 3 O 86-10 Willful Failure to Vol 9, No. 2 G On May 8,1985, a ' patient at Mercy Hospital, Report a Diagnostic and Wilkes-Barre, Pennsylvania, received an injection Medical Misadminis- A-11 of a radiophamaceutical (a diagnostic dose tration of technetium-99m) intended for another patient.
| |
| The misadministration was willfully not reported to the NRC as required by 10 CFR 635.43.
| |
| 86-11 Therapeutic Medical Vol. 9, No. 2 G On April 9, 1986, at Maryview Hospital, Portsmouth, Misadministration Virginia, a patient received a therapy dose in a chemical fom other than that intended. This resulted in an unintended dose of several hundred rad to the patient's bone marrow. This could result in an increased chance of the patient contracting leukemia.
| |
| 86-12 Willful Failure to Vol. 9, No. 2 G On April 22, 1986, the NRC Office of Inspection and.
| |
| Report Diagnostic and Enforcement issued an Order, effective imediately.
| |
| Medical Misadminis- 11 removing a physician from the position of Radiation trations Safety Officer (RS0) and Authorized User at Bloomington Hospital, Bloomington, Indiana. The physician had willfully not-reported five
| |
| | |
| no i ..urr.n.. ey an 1 m, ini. ., a 18tli!JI6 !!idli!'" e nu
| |
| '#n.!ill!P.J!.idM'in'All';
| |
| I in ,. ar , n!i.!!f#'Ha Mi..
| |
| a.a g g;;;;;;;gi ui, e, n, i a g;gg ;;,g6g,gg;g,g, "
| |
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| |
| ! g 911.*ri"4.n.vir,la.
| |
| fHi'i.Ili"Wn g;;g,ai l"
| |
| ..u'sr.flfn;.itP...R.!!"i;'t..
| |
| iyr.. aye.
| |
| au p,gg!:;,;;tgiaiba
| |
| * Eg'!g!!'ut*,!!j;irigiaM I!'
| |
| :'1'd iinV'jP"t"?lj'(!#11.7'"NuWI l 'EI ' 'i.'.
| |
| I'is."il nyr ..uil ;r.#..fi in ;.,.irin,it, fi! L; i.t'.n' AM6=5 gn V.i 9, N., R 0= #
| |
| !!p.gr.g.B"f.L"!
| |
| "" E '
| |
| "g.@gp. i ' ! g . g g f .
| |
| .n Ainl 01 11.016 '" I'". !IPriW31'WO'n:Pt4.It!'nt!!il..
| |
| r a . J " ' . . f " ! .i;
| |
| | |
| - - . - - - - - - .-- - - _ - . - _ _ _ . - - - .~ - . - - - . - ---- - ----- -
| |
| i Abnsmal Occurrences CY 1986 Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example' Consents ;
| |
| ..t i AS86-6 Contaminated Radio- Vol. 9, No. 2 G On May 9, 1985, a breakthrough'of molybdenum-99 i phamaceutical Used (a radioactive contaminant) occurred in a molyb-l in Diagnostic Administra- denum-99/ technetium-99m generator at Scripps tions (licensee'of Memorial Hospital, Encinitas, California. The California, an breakthrough went unrecognized and the AgreementState) contaminated' technetium-99m radiopharmaceutical was administered to four patients scheduled for diagnostic medical tests. Therefore, these.
| |
| patients received l exposures higher than necessary.
| |
| (NUREG-0090, Vol. 9, No. 3 was sent to the Commission for approval'on March 18, 1987, by SECY-87-73. It contains the following five proposed A0s and one proposed A0 for NRC and Agreement-State licensees, respectively.)
| |
| 86-15 Differential .Vol. 9, No. 3 G-2 On June 1, 1986, LaSalle Unit 2 experienced a feed-4 Pressure Switch and water transient that resulted in low water level '
| |
| I- Problem in Safety A-12_ in the reactor vessel. The level reached a point Systems at LaSalle where an-automatic reactor scram would be expected; 3
| |
| Facility however,.no'such scram occurred. Subsequent
| |
| != investigation found that.the problem was caused primarily by inadequate calibration of mechanical' 34 differential. pressure switches supplied by SOR, ;
| |
| * Incorporated (formerly Static "0" Ring Pressure Switch Company). Similar switches have been installed in safety systems at many' nuclear power i plants.
| |
| 86-16 Abnormal Cooldown Vol. 9, No. 3 B-4 On June 27, 1986,.while Duke Power Compan'y was-and Depressuriza- conducting a startup test at Catawba Unit 2 from-tion Transient at remotely located control panels, the reactor :
| |
| Catawba Unit 2 ' experienced an unexpected depressurization and
| |
| | |
| Abnorwel Occurrences CY 1986 Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Comments 1
| |
| cooldown. There were no actual consequences to public health or safety. However, if the decay heat load of the reactor core had been greater and'if the use of the remote shutdown panels had actually been required during a plant emergency, a more severe transient could have occurred.
| |
| > 1 86-17 Significant Safe- Vol. 9, No. 3 A-8 On July 7, 1986, NRC Region IV issued enforcement guards Deficien- letters containing Severity Level II violations 2 cies at Wolf Creek to the. licensees of two nuclear power stations for g; and Fort St. Vrain serious deficiencies in plant physical barriers.
| |
| In the most serious example, it was determined at the Wolf. Creek site that multiple uncontrolled access paths existed from the Owner Controlled Area- (OCA) into the Protected Area (PA) and in two instances into Vital Areas (VAs). At the Fort St. Vrain site, NRC inspectors identified paths from the OCA~to the PA and VA. In this situation, each access had a barrier installed, but each was evaluated to be inadequate and not capable of preventing an intruder from defeating it easily..
| |
| ! 86-18 Significant De- Vol. 9. No.'3 On August 7, 1986, NRC Region IV issued an enforce -
| |
| ficiencies in ment letter to the licensee containing a Severity Access Controls at Level II; violation regarding serious deficiencies River Bend Station in controlling the access of personnel to vital areas. Conditions existed whereby an intruder could have obtained unauthorized and undetected access into vital areas from either the protected
| |
| | |
| Abnormal Occurrences CY 1986 Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Comments area or other vital areas.
| |
| 86-19 Therapeutic Medical Vol. 9, No. 3 G During late August 1984, a high-activity Misadministration iodine-125 radiation source, which had been implemented for treatment of a brain tumor in a patient of University of Cincinnati Medical Center, Cincinnati, Ohio, had leaked, causing an unintended radiation exposure of 2087 rad to the patient's thyroid. This dose could reduce the
| |
| }* thyroid's function.
| |
| C Leakage occurred because one of the high activity seeds has been inadvertently cut into by a technician in the brachytherapy source storage room (BSR). The BSR was considerably contaminated and about 60 people received minor thyroid updates from airborne migration of the iodine-125.
| |
| The event was not previously reported as an abnorwel occurrence because at the time of the incident it was not classified as a medical 4 misadministration as defined in 10 CFR 935.45.
| |
| However, a reevaluation of the event by the NRC Staff during the latter part of 1986 concluded that the event should have properly been classified as a medical misadministration, and reportable as an abnormal occurrence.
| |
| | |
| i Abnormal Occurrences CY 1986 Report No. A0 Criterion A0 # Title of A0' NUREG-0090 or Example Comments i >
| |
| AS86-7 Therapeutic Medical Vol. 9, No. 3 G On September 5,1986, the Iowa : Radiological Health
| |
| : Misadministration Section, Bureau of Environmental Health. was '
| |
| (licensee of Iowa, notified.of a therapeutic medical .misadministra-an Agreement State) tion received by a patient at the University of l
| |
| i 7 Iowa Hospitals and Clinics, Iowa City. Iowa.
| |
| i The. patient's bronchial tenor was 'being treated by an iridium-192. source placed.in his bronchus tu be. While. sedated and asleep, the patient .
| |
| ! 3, apparently: pulled the tube containing -the source i- J. out of the bronchus.where.it came to rest on his
| |
| ; ^3 chest. The patient received an. estimated 1500 rad-to the chest in an area 3.4 cm long and.2 mm
| |
| , wide. a 1 '
| |
| (NUREG-0090, Vol. 9, No. 4 is under development by the NRC staff. The following nine events at NRC licensees are being considered by the staff for submittal to the Commission for approval.)- -
| |
| t 86-20 Loss of Low Vol. 9 No. 4 G-3 On October 1, 1986, while Unit'2 was-in a. refueling Pressure Service outage the Unit 2 load shed test was twice performed.
| |
| Water Systems at During both tests, the low pressure service. water- ,
| |
| Oconee. Unit 2 . system pump suction was. lost. Investigation showed that due to a design deficiency. the - l condenser circulating water system (which performs various safety-related-functions) was degraded.. A.similar design deficiency existed on Units ILand 3 which were operating at the time. ,
| |
| These units ~were taken to cold shutdown until the
| |
| ; problem was corrected. '
| |
| i
| |
| | |
| Abnorwel Occurrences CY 1986 1
| |
| i j
| |
| Report No. A0 Criterion or Example Comments ,
| |
| : A0 # Title of A0 NUREG-0090 Vol. 9, No. 4 A-10 On October 23, 1986, the licensee discovered that 86-21 Degraded Safety many valves in safety systems were degraded at
| |
| +
| |
| ! Systems Due to Catawba. On October 28, the licensee found a-Incorrect Switch licensee using improper torque switch settings i Settings on Rotork on the valve's Rotork motor operators. The Motor Operators at problem was caused by the licensee using improper '
| |
| Catawba and McGuire torque switch settings on the valve's Rotork Nuclear Stations motor operators. This could result.in the valves-not performing as designed (e.g.,-the activator motors switching off before the associated . valves 2, completed their travel).
| |
| $ ik' -
| |
| Vol. 9, No. 4 A-10 On December 9, 1986, with both Units 1 and 2 86-22 Secondary System at 100% power, Unit 2 tripped due to a low-Pipe Break Resulting in Death of Four low level in the "C" steam generator, followed Persons at Surry by a. rupture of an 18-inch suction line to the ,
| |
| "A": train main feedwater pump. The reactor Unit 2 was taken to a. cold shutdown condition with no release of radioactivity. 'However, eight-
| |
| , individuals in the vicinity of the pipe rupture-were injured due to the. release of steam and water. Four of.the individuals subsequently died.
| |
| I This incident was. investigated by an NRC Augmented Investigation Team.
| |
| Vol . 9, No. 4 G -On September 18 and October 6, 1986, a drum 86-23 Release of Ameri- containing radioactive waste was opened ~to
| |
| .cium-241 Inside a and-Waste Storage A-11 inspect its contents at Wright-Patterson Air Force Building at Base, near Dayton, Ohio. Both openings resulted L Wright-Patterson in a significant release of radioactive americium-Air Force Base
| |
| | |
| Abnorwel Occurrences CY 1986 i
| |
| Report No. A0 Criterion A0 # Title of A0 NUREG-0090 or Example Comments 241 inside the waste storage building, which significantly contaminated the building. .The costs to date as a result of this incident is approximately $500,000. Further costs will be incurred for either further decontamination of the '
| |
| building, or to dismantle and dispose of the building as radioactive waste.
| |
| 86-24 Therapeutic Medi- Vol. 9 No. 4 G On October 6-8, 1986, a patient at the Cleveland cal Misadministra- Clinic Foundation, Cleveland, Ohio, received a T* tion 30 series of therapeutic radiation exposures which
| |
| ; resulted in a radiation dose of approximately 2000 rad (head to waist), instead of an intended dose of 1200 rad.
| |
| 86-25 Suspension of Vol. 9, No. 4 A-11
| |
| ' On October 10, 1986, the NRC issued an Order License for suspending certain NRC-licensed service activities Servicing Tele- of Advanced Medical Systems, Inc., Geneva, Ohio.
| |
| ; therapy and Radiography This action was taken after the NRC determined Units that the firm had been using untrained and.unquali-fied employees to service cobalt-60 teletherapy units.
| |
| 86-26 Diagnostic Medical Vol. 9, No,'4 G On October 21, 1986, a patient at St. Luke's .
| |
| Misadministration Hospital, Racine, Wisconsin, received a whole body iodine-131 diagnostic scan while the intended procedure was to be a thyroid scan. The whole body scan involved 1.53 millicuries of iodine-131, about 30 times the normal dosage for a thyroid -
| |
| 4 scan. The patient may experience reduced thyroid function.
| |
| 9
| |
| | |
| Abnormal Occurr:nces CY-1986 Report No. A0 Criterion or Example Comments A0 # Title of A0 NUREG-0090 Diagnostic Medical Vol. 9, No. 4 G On November 18, 1986, a patient at Toledo 86-27 Hospital, Toledo, Ohio, received a mis-Misadministration administration of a radiopharmaceutical when the.
| |
| wrong radioactive material was administered. The patient's physician prescribed a bone scan, which normally involves about 20 millicuries of-techne-tium-99m MDP. Instead, the patient received about 20 millicuries of iodine-131. As a result, the patient's thyroid received a dose of several thousand rad which is expected to significantly reduce the thyroid's function.
| |
| }
| |
| "' G-3 On December 30, 1986, an Order was issued to Met-86-28 Immediately Effective Vol. 9, No. 4 Chem Testing. Laboratories of Utah, Inc., located in Order Modifying License Salt Lake City, that removed a senior management and Order to Show Cause employee from any assignment or position influenc-Issued to an ing or involving the performance or supervision of Industrial Radiography Company any NRC-licensed activities. This action was taken after the individual admitted he had typed a letter and forged on it the signature of a radio- ,
| |
| grapher for the purpose of explaining away an over-exposure indicated.on the radiographer's film badge, t
| |
| | |
| V y ww .,--_ _ _ ____
| |
| l APPENDIX B LISTINGS OF AEOD REPORTS 1980 - 1986 J
| |
| | |
| TABLE B-1 REPORTS ISSUED IN CY 1986 Case and Special Studies Date Subject No. Author 8/86 Rupture of an Iodine-125 Brachytherapy C601 S. Pettijohn Source at the University of Cincinnati Medical Center 8/86 Operational Experience Involving Turbine C602 C. Hsu 1
| |
| Overspeed Trips 12/86 A Review of Motor-Operated Valve C603 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 of Electrical Inverters 1/86 Trends and Patterns Program Plan - P601 R. Dennig FY86-FY88 8/86 Trends and Patterns Report of Unplanned P602 L. Bell Reactor Trips at U.S. Light Water Reactors in 1985 8/86 Trends and Patterns Report of Engineered P603 M. Harper Safety Feature Actuations at Connercial U.S. Nuclear Power Plants 8/86 Trends and Patterns Analysis of the P604 T. Wolf Operational Experience of Newly Licensed U.S. Nuclear Power Reactors 4/86 AE0D Annual Report for 1985 S601 J. Heltemes 5/86 An Overview of Nuclear Power Plant S602 J. Crooks Operating Experience Feedback Programs 6/86 Adequacy of the Scope of IE Bulletin S603 E. Leeds 86-01 l
| |
| l l
| |
| B-2
| |
| | |
| 'R2 actor Engineering Evaluations and Technical Reviews Date Subject No. Author Deficient Operator Actions Following Dual E601 E. Leeds 1/9/86 l Function Valve Failures Unexpected Criticality Due to Incorrect E602 R. Freeman 1/15/86 Calculation and Failure to follow Procedures Delayed Access to Safety-Related Areas E603 T. Cintula l 2/19/86 l
| |
| During Plant Operation Spurious System Isolations Caused by the E604 E. Leeds 3/14/86 Panalarm Model 86 Thermocouple Monitor Lightning Events at Nuclear Power Plants E605 M. Chiramal
| |
| ) 4/28/86 i
| |
| Loss of Safety Injection Capability at E606 R. Tripathi 5/27/86 Indian Point Unit 2 4 5/27/86 Core Damage Precursor Event at Trojan E514 D. Zukor
| |
| ! (Revision 1)
| |
| Degradation or loss of Charging Systems E607 F. Ashe i 7/3/86 with Swing Pump Designs Reexamination of Water Hamer Occurrences E608 E. Leeds
| |
| ; 7/14/86 Inadvertent Draining of Reactor Vessel E609 P. Lam 8/8/86 During Shutdown Cooling Operation i 8/14/86 Loss of Low Pressure Coolant Injection E610 E. Leeds Loop Selection Logic at Millstone Unit 1 l
| |
| i N. Thomasson 10/16/86 Deficiencies in Seismic Anchorage for E611 Electrical and Control Panels
| |
| ! 12/17/86 Emergency Diesel Generator Component E612 C. Hsu 1
| |
| Failures Due to Vibration 12/23/86 Localized Rod Cluster Control Assembly E613 E. Brown Wear at PWR Plants i
| |
| 4 j B-3 j
| |
| | |
| Reactor Engineering Evaluations and Technical Reviews (Continued)
| |
| Date Subject No. Author 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 Water T602 E. Leeds System Design Deficiencies at Maine Yankee and Haddam Neck 4/30/86 Inadvertent Pump Suction Transfer and T603 R. Tripathi Potential Auxiliary Feedwater Pump Cavita-tion at Davis-Besse 5/7/86 Events Resulting from Deficiencies in T604 E. Trager Labeling and Identification Systems 6/17/86 Failure of Main Steam Safety Valves T605 R. Freeman to Troperly Reseat 8/7/86 Inadvertent Recirculation Actuation T606 T. Cintula Signals at Combustion Engineering Plants 9/19/86 Occurrence of Events Involving Wrong T607 E. Trager Units / Wrong Train /Wrono 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- T609 E. Leeds i
| |
| 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 Pressure 12/31/86 Degradation of Safety Systems Due to T612 R. Tripathi Component Misalignment and/or Mispositioned Control / Selector Switches B-4
| |
| | |
| Nonreactor Engineering Evaluations Date Subject No. Author 6/17/86 Report on 1985 Nonreactor Events and N601 K. Black Five-Year Assessment for 1981-1985 Medical Misadministrations Reported for N602 S. Pettijohn 6/25/86 1985 and Five-Year Assessment of 1981-1985 Reports Incident Investigation Program Reports i
| |
| 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 System NUREG-1195 Power and Overcooling Transient at Rancho Seco on December 26, 1985 8/86 Incident Investigation Manual --
| |
| 12/86 Incident Investigation Manual - --
| |
| Revision 1 f
| |
| B-5
| |
| | |
| 4 TABLE B-2.
| |
| REPORTS ISSUED IN CY 1985 4
| |
| Case and Special Studies Date Subject No. Author i
| |
| 6/85 Safety Implications Associated with C501 H. Ornstein In-Plant Pressurized Gas Storage and Distribution Systems in Nuclear Power Plants 9/85 Overpressurization of Emergency Core C502 P. Lam Cooling Systems in Boiling Water Reactors a 12/85 Decay Heat Removal Problems at U.S. C503 ;H. Ornstein
| |
| ] Pressurized Water Reactors 12/85 Loss of Safety System Function Events C504 E. Trager 12/85 Therapy Misadministrations Reported to C505 S. Pettijohn
| |
| , NRC Pursuant to 10 CFR 35.42 6/85 Trends and Patterns Analysis of 1981. P502 B. Brady through 1983 LER Data (NUREG/CR-4129) 7/85 Feedwater Transient Incidents in P501 R. Dennig-Westinghouse PWRs 8/85 Engineered Safety Feature Actuations at P503 T. Wolf Commercial U.S. Nuclear Power Reactors January 1 through June 30, 1984 1
| |
| ' 8/85 Trends and Patterns Report of Unplanned P504 L. Bell Reactor Trips at U.S. Light Water Peactors in 1984 3/85 Review of Operational Experience From S501 D. Zukor Non-Power Reactors
| |
| ] 4/85 AE00 Semiannual Report for July-December S502 J. Heltemes 1984 l
| |
| 9/85 Evaluation of Recent Valve Operator Motor S503 E. Brown-
| |
| ) Burnout Events l B-6 i
| |
| | |
| l s
| |
| .R: actor Engineering Evaluations and Technical Reviews Date Subject No. Author i
| |
| 1/17/85 Motor 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 Operate 3/4/85 Partial Failures of Control Rod Systems to E503 M. Chiramal i Scram _
| |
| 3/29/85 Loss or Actuation of Various Safety-Related E504 F. Ashe Equipment Due to Removal of Fuses or Opening of Circuit Breakers 3/29/85 Service Water System Air Release Valve E505 S. Salah Failures 5/13/85 Valve' Stem Susceptibility to Intergranular E506 C. Hsu Stress Corrosion Cracking Due to Improper Heat Treatment i 5/17/85 Electrical Interaction Between Units E507 M. Chiramal During Loss of Offsite Power Event of August 21, 1984 at McGuire Units 1 and 2 1
| |
| I 5/24/85 Nuclear Plant Operating Experience E508 S. Rubin
| |
| : Involving Safety System Disturbances Caused i By Bumped Electro-Mechanical Components 7/25/85 Salem Unit 2 Depressurization Event E509 R. Freeman-1 7/30/85 Disabling of a Shared Diesel Generator Set E510 F. Ashe Due to Electrical Power Supply Arrangement for Support Auxiliaries 8/9/85 Closure of Emergency Core Cooling System _ E511 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 Valve Failures 10/8/85 Core Damage Precursor Event at Trojan E514 D. Zukor B-7 t
| |
| | |
| Reactor Engineering Evaluations and Technical Reviews (Continued)
| |
| Date Subject No. Author 12/11/85 Inadvertent Actuation of Safety System E515 M. Chiramal-Due to Cross-Talk 1/22/85 - Failure of Automatic Protection for Boron T501 R. Freeman Dilution Event at Callaway Unit 1 I
| |
| 3/18/85 Comparative Analysis of Recent Feedline T502 E. Leeds Water Hammer Events at Maine' Yankee, Calvert
| |
| - Cliffs, Palisades, and Salem 5/2/85 Pressurizer Level Instrumentation of T503 M. Chiramal Combustion Engineering Reactor Units 3 5/17/85 - Loss of Instrument Air and Subsequent T504 R. Freeman Pressure Transient at Callaway Unit 1 7/17/85 Beaver Valley Component Cooling Water Pump T505 C. Hsu Damage 1
| |
| 7/25/85 Primary System Release Due to Pressurizer -T506 T. Cintula Degas Relief Valve Lifting 8/13/85 Standby Liquid Control System Pressure 1507 E. Brown-Relief Valves Lift at a Pressure Lower than Reactor Coolant Pressure 8/14/85 Browns Ferry Nuclear Plant High Pressure T508- E. Leeds Coolant Injection System Performance i
| |
| Assessment
| |
| . 8/29/85 Inadequate Surveillance Testing Procedures T509 F. Ashe for Degraded Voltage and Undervoltage Relays Associated with 4160-Volt Emergency Buses 9/4/85 Xenon Induced Power Oscillations at T510 R. Freeman Catawba 9/16/85 Technicians Perform Work on Wrong Control T511 E. Trager Rod Drive Mechanism 1
| |
| 10/24/85 Incorrect Plugging of Steam Generator T512 R. Freeman Tubes
| |
| -11/7/85 Flooding of Safety-Related Valves in Pits T513 D. Zukor i
| |
| 4 4
| |
| B-8
| |
| -- , w-,m- - -
| |
| - - - .,- , - 3 -
| |
| - - - , , - - - , w.
| |
| | |
| R: actor Engineerina Evaluations and Technical Reviews (Continued)
| |
| Date Subject No. Author 4
| |
| 11/25/85 Potential Loss of Component Cooling Water T514 D. Zukor Due to Maladjustment of Relief Valves 12/5/85 Residual Heat Removal Service Water T515 S. Salah Booster Pur.p Air Binding at Brunswick Unit 1 12/11/85 High Pressure Coolant Injection Overspeed T516 E. Trager Trip Loss Events and Subsequent Damage Due to Water Hammer Nenreactor Engineering Evaluations 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 Report Date Subject Designation 7/85 Loss of Main and Auxiliary NUREG-1154 Feedwater Event at the Davis-Besse Plant on June 9, 1985 B-9 1
| |
| 1 j
| |
| | |
| TABLE B-3 REPORTS ISSUED IN CY 1984 Case and Special Studies Date Subject No. Author 3/84 Low Temperature Overpressure-Events at C401 W. Lanning
| |
| , -Turkey Point Unit 4 6/84 Operating Experience Related to Moisture C402 M. El-Zeftawy
| |
| .i Intrusion in Electrical Equipment at Commercial Power Reactors 5/84 Hatch Unit 2 Plant Systems Interaction C403 S. Rubin Event on August 25, 1982 7/84 Steam Binding of Auxiliary Feedwater C404 W. Lanning
| |
| . Pumps 9/84 Breaching of the Encapsulation of Sealed C405 S. Pettijohn Well Logging Sources i
| |
| 2/84 Operating History Overview for Diesel P401 R. Dennig Generators in Nuclear Service M. Chiramal 3/84 AE0D Trends and Patterns Program Plan P402 R. Dennig 5/84 AE0D Trends and Patterns Evaluation Report, P403 F. Hebdon
| |
| " Preliminary Assessment of LER Reporting Under 10 CFR 50.73" i.
| |
| : 3/84 LER Data on Personnel Errors P404 F. Hebdon ,
| |
| ; 11/84 Draft Trends and Patterns Analysis of P405 M. Harper Feedwater Transients at Westinghouse PWRs 11/84 Trends and Patterns Analysis of Reactor P406 L. Bell Scrams (Pilot Study) i 1/84 Human Error in Events Involving Wrong S401 E. Trager j Unit or Wrong Train I
| |
| ] B-10 3
| |
| | |
| Case and Special Studies Continued i
| |
| Date Subject No.- Author 7/84 Pressure Locking of Flexible-Disk Wedge- S402 S. Rubin Type Gate Valves 6/84 Annual Report of U.S. NRC Participation in S403 J. Crooks the Nuclear Energy Agency Incident Reporting System During 1983 6/84 Analysis of Foreign IRS Reports Submitted S404 D. Zukor During CY 1983 9/84 Semiannual Report on AE0D Activities S405 J. Heltemes 10/84 Application of Risk Perspectives: A S406 P. Lam Procedures Guide i
| |
| Reactor Engineering Evaluations and Technical Reviews Date Subject No. Author 1/4/84 Temporary loss of All AC Power Due to E401 H. Chiramal Relay Failures in Diesel Generator Load Shedding Circuitry at Fort St. Vrain 1/10/84 Water Hammer in Boiling Water Reactor E402 S. Rubin High Pressure Coolant Injection Systems 1/17/84 Deficiency in Automatic Switch Company E403 F. Ashe (ASCO) Spare Parts Kits for Scram Pilot Solenoid Valves 2/28/84 Failures in the Upper Head Injection E404 D. Zukor System
| |
| ~
| |
| 3/22/84 Common Mode Failure of HPCI Steam Flow E405 M. El-Zeftawy Isolation Capability at Browns Ferry 3/22/84 Mechanical Snubber Failure E406 C. Hsu i
| |
| 3/26/84 Initiation and Indication Circuitry for E407 F. Ashe
| |
| , High Pressure Coolant Injection Systems B-11
| |
| | |
| Reactor Engineering Evaluations and Technical Reviews (Continued)
| |
| -Date Subject No. Author-3/27/84 Load Reduction Transient at the Salem E323 N. Trehan Unit 2 on January 14, 1982 (Revision 1) 4/13/84 Reversed Differential Pressure Instrument E408 S. Rubin Sensing Lines 5/16/84 Operating Experience Involving Air in E409 S. Salah Instrument Sensing Lines 5/21/84 Operational Experiences Involving Standby E410 F. Ashe Gas Treatment Systems which Illustrate Potential Common Cause Failure or Degrada-tion Mechanisms 5/22/84 Failure of Anti-Cavitation Device in E411 C. Hsu Residual Heat Removal Service Water Heat Exchanger Outlet Valve 5/25/84 Adverse System Interaction with Domestic E412 T. Cintula Water Systems 5/25/84 Natural Circulation in Pressurized Water E413 W. Lanning Reactors t
| |
| 5/31/84 Stuck Open Isniation Check Valve on the E414 P. Lam Residual Heat kemoval System at Hatch Unit 2 6/6/84 Overcooling Transient E415 E. Imbro 6/11/84 Erosion in Nuclear Power Plants E416 E. Brown 7/2/84 Loosening of Flange Bolts on Residual Heat E417 C. Hsu Removal Heat Exchanger Leading to Primary to Secondary Side Leakage 7/24/84 Feedwater Transients During Startup at E418 D. Zukor Westinghouse Plants 7/84 Failures of Fischer-Porter Transmitters E419 M. Chiramal i Used in Safety-Related Systems 8/23/84 Operational Experiences Involving Shorted E420 M. Chiramal Lamp Sockets of Indication Lights B-12 i
| |
| | |
| l
| |
| \
| |
| R actor Engineering Evaluations'and Technical' Reviews (Continued) l Date Subject No. Author '
| |
| ~ 8/27/84 Loss of Pressurizer Heaters During Precore E421 T. Cintula Hot Functional Testing j '8/27/84 'High Pressure Coolant Injection System E422 T. Wolf l.
| |
| Performance at Hatch Units l' and 2 9/20/84: Failure of Large Hydraulic Snubbers to E423 E. Brown
| |
| [.
| |
| Lock Up
| |
| .\
| |
| i 10/1/84~ Failure of Anchor Bolt on Diesel Generator E424 C. Hsu Day Tank at Davis Besse Unit-1 ,
| |
| l 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 l
| |
| i' 11/6/84 Licensee Event Reports that Address .
| |
| E427 F. Ashe Situations Which Potentially Could Result in Overloading Electrical Equipment in t
| |
| i the Emergency Power System or Prevent Operation of the Onsite Power System Sequencer 3/2/84 Failures of Containment Air Monitors at T401 'D. Zukor-Farley Units 1 and 2
| |
| .i 3/22/84 Chemical Contamination of Primary and T402 M. El-Zeftawy
| |
| ;- Secondary Systems in Light Water Reactors 3/23/84 Setpoint Drift of Barton Model 288 T403 M. Chiramal Switches 7
| |
| 4/13/84 Cable Fire and Loss of Control Power to T404 M. Chiramal Engineered Safeguards Valves i
| |
| 4/25/84 Cold Weather Events 1983-1984 T405 T. Cintula i 4/25/84 Improper Spare Parts Procurement Event at .T406 T. Wolf-Grand Gulf Unit 1 4
| |
| 4/30/84 Failure of 4 kV Circuit Breaker to Trip T407 -M. Chiramal' F
| |
| l B-13
| |
| | |
| s L
| |
| Reactor Engineering Evaluations and Technical Reviews (Continued)
| |
| ; Date Subject No. Author i
| |
| $ Diesel Generator Inoperability Due to 5/7/84. T408 M. Chiramal Overheating of Ventilation Cowling j
| |
| 5/1/84 ' Multiple Failures 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 1
| |
| Coolant Injection System Failure to Open During a Surveillance Test i 6/18/84 Contamination of the Nitrogen System at T411 M. El-Zeftawy Sacramento Municipal Utility District 6/18/84 Failure of an Access Door Between the. T412 T. Wolf
| |
| ; Drywell and the Wetwell 3 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 4
| |
| 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 f
| |
| ; 8/1/84 Loss of Engineered Safety Feature Auxiliary' T416 D. Zukor Feedwater Pump Capability at Trojan on '
| |
| January 22, 1983 f- 8/2/84 Excessive Cooldown Rate Event at LaSalle T417 S. Salah
| |
| ! Unit 1 j 8/6/84 Events Involving Fires or Other Related T418 F. Ashe l Abnormalities in Motor Control Centers -
| |
| l with Aluminum Bus Bars i 8/20/84 Contamination of Snubber Bleed Screw and T419 C. Hsu l 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-t 4
| |
| B-14
| |
| | |
| Reactor Engineering Evaluations and Technical Reviews (Continued) ,
| |
| l Date Subject No. ' Author !
| |
| Design Deficiency in Standby Gas T421 M. Chiramal 8/23/84 Treatment System Inoperability of Safety-Injection Pump T422 D. Zukor 8/29/84 at Salem Unit 1 on October 17, 1983 10/25/84- Inoperability of Helium Circulator Over- T423 E. Imbro i speed _ Trip Channels Due to Impedance Variations.in Speed Sensing Cables Exposed to Steam Leak i
| |
| i 11/20/84 Fire Water Main Leakage Into 4 kV T424 T. Cintula Switchgear Room at San Onofre Unit 1 4
| |
| Nonreactor Engineering Evaluations Date Subject No. Author i
| |
| Report on Medical Misadministrations for N204D S. Pettijohn 5/8/84 January 1983 through June 1983 6/11/84 Nonreactor Event Report Database for the N401 K. Black Period July-December 1983 Events Involving Undetected Unavailability N402 E. Trager 6/26/84 of the Turbine-Driven Auxiliary Feedwater
| |
| ; Train 7/3/84 Report on Medical Misadministrations for N403 S. Pettijohn July 1983-December 1983 9
| |
| 1 9
| |
| 4 i
| |
| B-15
| |
| . _ , , , _ , - . . - , _ ,m. --- , = , , - - , . - - , - - - _ - _ . . , , , .--.___,..m. _ ,
| |
| | |
| c-L TABLE B-4 REPORTS ISSUED IN CY 1983
| |
| ; Case and Special Studies Date Subject No. Author 4/83 Failures of Class IE Safety-Related C301 M. Chiramal i Switchgear Circuit Breakers to.Close on Demand 9/83 Potentially Damaging Failure Modes of NUREG/ M. Chiramal 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 and Analysis of Operational Experience J.
| |
| l 2
| |
| Reactor Engineering Evaluations and Technical Reviews d
| |
| Date Subject No. Author l
| |
| j 1/19/83 Fuel Degradation at Westinghouse Plants E301 D. Zukor 1/31/83 Potential Loss of Service Water Flow E302 E. Imbro
| |
| , Resulting from a Loss of Instrument Air 2/16/83 Valve Flooding Event at Surry E303 D..Zukor
| |
| -3/11/83 Investigation of Backflow Protection in E304 T. Cintula Common Equipment and Floor Drain Systems -
| |
| to Prevent Flooding of Vital Equipment in Safety-Related Comp 3rtments 4/13/83 Inoperable Motor Operated Valve Assemblies E305 E. Brown :
| |
| Due to Premature Degradation of Motors F. Ashe and/or Improper Limit Switch / Torque Switch Adjustment i
| |
| , 4/14/83 Cooldown During Loss of Control Room Test E306 D. Zukor i at McGuire Unit I 4/14/83 Update to AE0D/E301 (Fuel Degradation at E301 D. Zukor j Westinghouse Plants) (Revision 1) l B-16 i
| |
| | |
| Reactor Engineering Evaluations and Technical Reviews (Continued)
| |
| Subject No. Author Date l
| |
| Degradation of Safety-Related Batteries E307 F. Ashe 4/18/83 Due to Cracking of Battery Cell Cases and/or Other Possible Aging-Relating Mechanisms Cracks and Leaks in Small Diameter Piping E308 E. Brown 4/19/83 The Potential for Water Hammer During the E309 S. Rubin 4/21/83 Restart of Residual Heat Removal Pumps at BWR Nuclear Power Plants Loss of Shutdown Cooling and Subsequent E310 T. Cintula 4/25/83 Boron Dilution at San Onofre Unit 2 Loss of Salt Water Flow to the Service E311 T. Cintula 4/25/83 Water Heat Exchangers for 23 Minutes at Calvert Cliffs Unit 2 Operability of Target Rock Safety Relief E312 J. Pellet 5/18/83 Valves in the Safety Mode with Pilot Valve Leakage Potential Contamination of the Spent Fuel E313 E. Brown 6/15/83 Pool and Primary Reactor System Loss of All Three Charging Pumps Due to E314 T. Cintula 6/28/83 Empty Common Reference Leg in the Liquid Level Transducers for the Volume Control Tank at St. Lucie 1 Misuse of Valve Resulting in Vibration and E315 E. Brown 7/5/83 Damage to the Valve Assembly and Pipe Supports Frozen Ice Condenser Intermediate Deck E316 D. Zukor 7/11/83 Doors Loss of High Pressure Injection System E317 N. Trehan 8/1/83 Biofouling at Salem Units 1 and 2 E318 E. Imbro 8/15/83 Loss of Drywell-Torus Pressure Differen- E319 S. Rubin 9/8/83 tial During Residual Heat Removal Pump Flow Testing at Cooper Power Operated Relief Valve (PORV) Actua- E320 E. Imbro 9/8/83 tion Resulting in Safety Injection Actua-tion at Calvert Cliffs B-17
| |
| | |
| A Reactor Engineering Evaluations and Technical Reviews (Continued)
| |
| ! Date Subject No. Author 1
| |
| i 9/12/83 Three Similar Events of a Loss of Shutdown E321 T. Cintula
| |
| : Cooling Flow at Combustion Engineering Plants 9/16/83 Damage to Vacuum Breaker Valves as a E322 .C. Hsu Result of Relief Valve Lifting at Peach Bottom Unit 2 i l 9/19/83 Load Reduction Transient at Salem Unit-2 E323 on January 14, 1982 N. Trehan t
| |
| 9/21/83 Review of Events Involving Failures of E324 M. Chiramal Power Supply in Instrumentation and Control
| |
| ;. Systems '
| |
| i 11/21/83 Vapor Binding of Auxiliary Feedwater Pumps E325 W. Lanning s
| |
| at Robinson Unit 2 11/28/83 Steam Voiding in Oconee Unit 3 on June 13.- E326 H. Ornstein 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 Seat Degradation in Henry Pratt Butterfly T303 E. Brown Valves 3/23 Cause of Containment Isolation Valve F042A T304 S. Salah to Close at Brunswick Unit 1 3/28 Flow Blockage in Essential Raw Cooling T305 E. Imbro i
| |
| Water System Due to Asiatic Clam Intrusion j at Sequoyah Unit 1 i
| |
| i 4
| |
| B-18
| |
| - -m_., . _ , _ _ _ - , , , , . _ . . - , , .-.m .,,,-,,,.__m. , . - _ _ . _ , , _ ,
| |
| | |
| Reactor Engineering Evaluations and Technical Reviews (Continued)
| |
| Sub. ject No. _ Author.
| |
| Date Scram Discharge Volume Level Switch Failure T306 J. Pellet 4/13/83 at Hatch Unit 2 Condensate Demineralizer Resin Migration T307 J. Pellet 4/19/83
| |
| .through the Plant Vent and the Standby Gas Treatment System at Pilgrim Unit 1 Undetectable Failure in Westinghouse Solid T308 M. Chiramal l 4/20/83 State Protection System Air in Reactor Water Cleanup System T309 S. Salah 4/25/83 4
| |
| Instrument Sensing Lines at Brunswick Unit 2 i
| |
| Blocking of Automatic Safety Injection T310 M. Chiramal 4/25/83 Signals l
| |
| Rod Control Urgent Failure on June 25, T311 N. Trehan 5/5/83 1982 at Surry Unit 2 T312 M. Chiramal 5/9/83 Failure of 5 kV Cable Terminations Capped Containment Pressure Sensing Lines T313 S. Rubin 5/11/83 Improper Size of Inlet Piping to Primary T314 E. Imbro 5/24/83 l Safety Valves Events Involving Losses of or Perturbations T315 F. Ashe 1
| |
| 5/24/83 l
| |
| in a Single 120 Volt AC Vital Power Supply Inverter and Attendant Distribution Bus which Resulted in Inadvertent Actuations l of Safety Systems Thermal Nonrepeatability Problem with T316 M. Chiramal 5/31/83 Barton Model 763 and 764 Electronic Transmitters T317 D. Zukor 6/13/83 Problems with Diesel-Driven Containment i Spray Pump at Zion Unit 2 on December 16, 1982 1
| |
| Failure of Recirculation Spray Service T318 D. Zukor ,
| |
| 6/13/83 Water Motor Operated Valves l l
| |
| I l
| |
| l i
| |
| B-19
| |
| . A
| |
| | |
| Reactor Case Studies and Engineering Evaluations (Continued)
| |
| Date- Subject No. Author 6/13/83. Design Deficiency in Control Circuits of T319 M. Chiramal Feedwater Isolation Valves and Boron Injection Tank Recirculation Valves 1
| |
| 6/14/83 Inadvertent Safety Injections Attributed T320 F. Ashe to Personnel Error at Sunener 6/15/83 Check Valve Installed Backward: in Instru- T321_ D. Zukor ment Air Line to the Power Operated Relief Valve at Surry Unit 1 6/15/83 Gouges .in Main Coolant System Piping at T322 _D. Zukor Diablo Canyon on April 19, 1983 6/17/83 Turbine Trip Bypass Delay at Grand Gulf T323 S. Salah Unit 1 7/27/83 Events Involving Two or More Simultaneously T324 F. Ashe Dropped Rod Control Cluster Assemblies 8/1/83 Leakage in Static-0-Ring Pressure Switches T325 M. Chiramal 8/2/83 Safety / Relief Valve Corrosion at a Foreign T326 Reactor E. Brown i 8/11/83 Auxiliary Feedwater Header Problems at T327 H. Ornstein j
| |
| Babcock & Wilcox Plants i
| |
| 8/12/83 Two of Three Emergency Core Cooling System T328 D. Zukor Accumulators Inoperable at Surry Unit 1
| |
| ; 8/24/83 Leak in Reactor Water Cleanup System "B" T329 C. Hsu i Regenerative Heat Exchanger Relief Line 8/29/83 Steam Generator Tube Rupture at Oconee T330 M. El-Zeftawy Unit 2 k 8/29/83 Review of Events at Operating Nuclear T331 M. Chiramal Plants Involving Plant Computers 10/7/83 Reactor Vessel Drainage at Grand Gulf T332 S. Salah Unit 1 i
| |
| B-20
| |
| | |
| R; actor Engineering Evaluations and Technical Reviews (Continued)
| |
| Subject No. Author Date 10/31/83 Degradation of Saltwater Cooling System T333 H. Ornstein at San Onofre Unit 1 Caused by a Loss of l Instrument Air 11/15/83 Reactor Vessel Drainage at Grand Gulf T334 S. Salah Unit 1 11/15/83 Simultaneous Safety 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 at Nine Mile Point Unit 1 11/21/83 Water Hammer in the Main Feedwater System T337 E. Imbro Resulting in a Feedwater Line Crack at Maine Yankee 11/28/83 Water Leak through Containment Spray Block T338 D. Zukar Valves at San Onofre 1 11/29/83 Redundant Emergency Core Cooling System T339 T. Cintula Pump Room Air Coolers Out of Service for 22 Hours at Calvert Cliffs Unit 1 12/2/83 Evaluation of a Control Rod Mismanipulation T340 T. Wolf Event at Hatch Unit 2 12/19/83 Corrosion of Carbon Steel Pipe in Service T341 E. Brown Water Headers Nonreactor Engineering Evaluations and Technical Reviews Date_ Subject No. Author 1/11/83 Nonreactor Event Report Database for the N209A E. Trager Period January-June 1982 3/18/83 1125/1131 Effluent Releases by Material N301 S. Pettijohn Licensees B-21
| |
| | |
| Nonreactor Engineering Evaluations and Technical Reviews i Date Subject No. Author
| |
| ,1 6/10/83 Mound Laboratory Fabricated PuBe Sources N302 K. Black 6/10/83 Americium Contamination Resulting from N303 K. Black Rupture of Well-Logging Sources 6/14/83 Nonreactor Event Report Database for the N2098 K. Black Period July-December 1982
| |
| ! 7/14/83 Americium-241 Sources N304 7/14/83- Report on Medical Misadministrations for -N204C S. Pettijohn
| |
| ] January 1981-December 1982 8/4/83 Human Factors Contributions to Accident N305 E. Trager Sequence Precursor Events
| |
| : 12/1/83 Potentially Leaking Americium-241 Sources N306 S. Pettijohn Manufactured by Amersham Corporation i 12/28/83 Nonreactor Event Report Database for the N307 K. Black Period January-June 1983 3/10/83- Internal Exposure to An-241 NT301 K. Black 4/5/83 Kay-Ray, Inc. Reports of Suspected Leaking NT302 S. Pettijohn Sealed Sources Manufactured by General Radioisotope Products 8/24/83 Possession of Unauthorized Sealed Source / NT303 S. Pettijohn Exposure Device Combinations by Mid-Con Inspection Services, Inc.
| |
| 11/4/83 Human Factors Involvement in Events at NT304- K. Black Oconee Units 1, 2, and 3 l B-22 I
| |
| | |
| TABLE B-5 REPORTS ISSUED IN CY 1982 l l
| |
| Case Studies Date Subject No. Author Safety Concern Associated with Reactor C201 M. Chiramal 1/82 Vessel Level Instrumentation in Boiling Water Reactors Report on Service Water System Flow C202 E. Imbro 2/82 Blockages by Bivalve Mollusks at Arkansas Nuclear One and Brunswick Survey of Valve Operator-Related Events C203 E. Brown 5/82 Occurring During 1978, 1979, and 1980 I
| |
| 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 i
| |
| as Applied to the April 1981 Overfill Event at Arkansas Nuclear One, Unit 1 r
| |
| i Inadvertent Loss of Reactor Coolant Events C206 W. Lanning 10/82 4
| |
| at the Sequoyah Nuclear Plant, Units 1
| |
| ; and 2 4
| |
| l Reactor Engineering Evaluations Date Subject No. Author Methodology for Vital Area Determination E201 W. Lanning i 1/12/82 l 1/13/82 Loss of High Pressure Injection Lube Oil E202 J. Pellet Cooling at Rancho Seco i
| |
| ~
| |
| Inadvertent Isolation of Containment Fan E203 W. Lanning 1/21/82 i Units at Salem Unit 1 1/28/82 Effects of Fire Protection System Actua- E204 M. Chiramal tion on Safety-Related Equipment i
| |
| i 8-23 a
| |
| | |
| 1 L Reactor Engineering Evaluations
| |
| [ Date ~ Subject No. Author i
| |
| 2/16/82 Potential Consequences of Heavy Load Drop E205 M. El-Zeftawy Accidents in LWRs i
| |
| 2/22/82- Load Reduction Transient on January 14,- E206 N. Trehan.
| |
| 1982 at Salem Unit 2
| |
| . 2/22/82 LER 50-336/81-26: Investigation of the E207 E. Imbro Relative Frequency of Valve Overtravel l- Anomalies that Could Resuit-in a Potential i
| |
| Centrifugal Pump Runout Exceeding Net Positive-
| |
| ; Suction Head 2/22/82 An Observed Difference in Lift Setpoint E208 E. Imbro i
| |
| for Steam Generator and Pressurizer Safety Valves l
| |
| 2/23/82- Generator Rotor Retaining Ring as a E209 M. Chiramal Potential Missile (Incident at Barseback
| |
| : Unit 1 on April 13,1979) 1 2/23/82 Inadequate Switchgear Cooling at Beaver E210 j Valley Unit 1 W. Lanning 2/24/82 Repetitive Failures of Emergency Feedwater E211 E. Imbro Flow Valves at Arkansas Unit 2 Because of 1
| |
| Valve Operator Hydraulic Problems 2/24/82 Spurious Trip of the Generator Lockout. E212 i Relay Associated with a Diesel Generator F. Ashe i
| |
| Unit t
| |
| : 2/24/82 Trip of Two Inservice Auxiliary Feedwater i
| |
| 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 j Loop E214 T. Wolf i
| |
| ! 3/18/82 Engineering Evaluation of the. Salt Service E215 j E. Imbro-Water System Flow Blockage at the Pilgrim-Nuclear Power Station by Blue Mussels .
| |
| t l 3/28/82 A Recently Evaluated Preoperational Test i
| |
| E216 .H. Ornstein Precursor of the TMI-2 Accident I
| |
| i B-24
| |
| | |
| l R: actor Engineering Evaluations Subject No. Author Date I E217 M. Chiramal 3/31/82 . Scram Pilot Solenoid Valve Failures Due to Low Voltage - Grand Gulf Unit 1 Potential for Air Binding or Degraded E218 S. Rubin
| |
| .3/31/82 Perfomance of BWR Residual Heat-Removal j System Pumps During the Recirculation l
| |
| Phase of a Loss-of-Coolant Accident Proposed Circular: Contamination of Air E219' H. Ornstein l 4/1/82 Serving Safety-Related Equipment Water in the Fuel Oil Tank at Surry Power E220 N. Trehan 4/6/82 '
| |
| Station Unit 2 Indian Point Unit 2 Flooding Event E221 W. Lanning y a/22/82 N. Trehan 5/10/82 Loss of Reserve Station Service Transfomer E222 "B" on January 18, 1982 at Surry Unit 2 3
| |
| E223 W. Lanning 5/11/82 Inadvertent Loss of Coolant Events at Sequoyah Units 1 and 2 i
| |
| E224 W. Lanning i
| |
| 5/21/82 Generic Concerns Associated with the Ginna Steam Generator Tube Rupture Event Degradation of BWR Scram Pilot Solenoid E225 M. Chiramal 6/1/82
| |
| < Valves Due to Abnormal Power Supply Voltage 4
| |
| Inoperability of Instrumentation Due to E226 M. Chiramal
| |
| : 6/18/82
| |
| ;! Extreme Cold Weather Failure of Engineered Safety Features E227 F. Ashe i 6/24/82 Manual Initiation Pushbutton Switches Repetitive Overspeed Trips of the Steam E228 E. Imbro
| |
| ,- 6/25/82 i
| |
| Driven Emergency Feedwater Pump on Initial
| |
| : Start at Arkansas Nuclear One, Unit 2 a
| |
| ' Potential for Flooding in Control Room at E229 E. Imbro 6/29/82 San Onofre Units 2 and 3
| |
| )
| |
| Water in the Fuel Oil Tank at Surry Power E230 N. Trehan 7/7/82 i Station, Unit 2 - Additional Information-i B-25
| |
| - . - . , - . - . _ _ . . - - - , - - - _ - - - - - - . . , - - - . - . ~ . - - - - . - _ , - - , . . . , - . . _ . . , , . - - _ . . . - . . . . - -
| |
| | |
| J
| |
| ! l Reactor Engineering Evaluations
| |
| ! Date Subject No. Author i
| |
| 7/19/82 Millstone Unit 2 Loss of Shutdown Cooling E231 M. Chiramal Due to Trip of Low Pressure Safety Injection Pump 7/19/82 Potential Deficiency in the Sigma E232 Lumigraph Indicators Model Number 9270. F. Ashe 7/28/82 Carbon Dioxide Systems Used for Fire E233
| |
| : Protection in or Adjacent to Critical M. Chiramal a
| |
| Areas 8/11/82 Failure in a Section of 4 kV Bus Cable E234
| |
| !. Manufactured by Okonite F. Ashe i
| |
| 8/11/82 Wiring Error in Handswitch for Solenoid E235 j S. Rubin
| |
| ; Control valves Associated with High Pressure Coolant Injection System Steam Condensing Mode Pressure Control- Valve at Duane Arnold 8/25/82 Brunswick Steam Electric Plant Unit 2 I E236 T. Wolf j Loss of Residual Heat Removal Service Water on January 16, 1982
| |
| , 8/25/82 Power Operated Relief Valve Failure at E237 j Robinson E. Brown l
| |
| { 8/25/82 Water in the Lube Oil in Safety Injection E238 N. Trehan i Pump IA-A at Sequoyah - LER 81-076 9/24/82 Main Steam Isolation Valve Closures and Pressurizer Safety Valve Actuations at E239 T. Cintula .
| |
| St. Lucie Unit 1 on December 19, 1981 {
| |
| 9/29/82 Preliminary Account of Events Associated E240 with a Reactor Trip at Hatch Unit 2 on S. Rubin
| |
| ; August 25, 1982 ;
| |
| i 10/1/82 Emergency Dierel Generator System Problems E241 ' M. Chiramal at Fitzpttrick l 10/21/82 Fuel Assembly Degradation while in the E242 Spent Fuel Storage Pool E. Brown i
| |
| B-26 t
| |
| .. ,..;,- .,.,..-..,..a. - . . . , . --- - ,. n .,.-n, , . - . , , -.._-.-,n., ., . . , . . , , - , , ,,.n - , , . , ,,-..n -.- -~n
| |
| | |
| Reactor Engineering Evaluations Subject No. Author Date 10/21/82 Plant Trip Followed by a Safety Injection E243 T. Cintula l Caused by Loss of "A" Cooling Tower Pump
| |
| ' at Palisades on February 4,1982 10/21/82 Loss of Residual Heat Removal System Event E244 T. Wolf at Pilgrim Nuclear Power Station on December 21, 1981 10/21/82 Failure of Westinghouse Type SC-1 No. E245 F. Ashe 1876-072 Relays 10/21/82 Events Involving Loss of Electrical E246 F. Ashe Inverters Including Attendant Inverters to Vital Instrument Buses 10/26/82 Engineering Evaluation of Turbir.e/ Reactor E247 J. Pellet Trip at Rancho Seco on August 7,1981 Engineering Evaluation Report on McGuire E248 D. Zukor 11/2/82 Overpressurization Event of August 27, 1981 11/4/82 Engineering Evaluation Memorandum, E249 H. Ornstein Licensee Reporting of the Turbine / Reactor Trip at Rancho Seco on August 7, 1981 11/8/82 Quad Cities Unit 2 Loss of Auxiliary E250 M. Chiramal Electrical Power Event on June 22, 1982 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 Diesel Generator Units Under the Conditions of loss-of-Coolant Accident and Loss of Station 11/17/82 Review of Prairie Island Unit 1 E254 M. Chiramal LER 82-015/01T on Diesel Generator Operability B-27
| |
| | |
| Reactor Engineering Evaluations Date Subject No. Author 11/17/82 Failure of the Vent Line on the Common E255 T. Cintula Discharge of the Two Motor-Driven Auxiliary Feedwater Pumps at San Onofre Unit 2 from an Improper Valve Lineup 11/24/82 Loss of Shutdown Cooling and Subsequent E256 T. Cintula Boron Dilution at San Onofre Unit 2 12/2/82 Insufficient Net 0 0sitive Suction Head for E257 D. Zukor Charging Pump Service Water Pumps at Surry Nuclear Power Station Nonreactor Engineering Evaluations i Date Subject No. Author
| |
| )
| |
| I 2/1/82 Report on Medical Misadministrations for N201 S. Pettijohn the Period November 10, 1980 - September 30, 1981 1/21/82 Buildup of Uranium-Bearing Sludge in Waste Retention Tanks N202 K. Black 2/18/82 Lost Plutonium - 238 Source N203 K. Black 3/5/82 Report on Medical Misadministrations for CY 1981 N204 S. Pettijohn 4/27/82 Preliminary AE0D Review of Iodine-125 Sealed Source Leakage Incidents N205 E. Trager 5/6/82 Eberline Instrument Corporation - Part 21 Report N206 K. Black 5/25/82 AEOD Review of Iodine-125 Sealed Source Leakage Incidents N207 E. Trager 8/2/82 Potentially Leaking Plutonium-Beryllium Neutron Sources N208 S. Pettijohn B-28
| |
| | |
| Nonreactor Engineering Evaluations (Continued)
| |
| Date Subject No. Author A Sumary of the Nonreactor Event Report N209 K. Black 8/30/82 Database for 1981 11/15/82 Leaking Hoses on Self Contained Breathing N210 K. Black Apparatus (SCBA) Manufactured by MSA i
| |
| t 1
| |
| I B-29
| |
| | |
| 4 TABLE 8-6 REPORTS ISSUED IN CY 1981 Case Studies Date Subject No. Author 3/81 Report on the St. Lucie Unit 1 Natural C101 E. Imbro Circulation Cooldown on June 11, 1980
| |
| ; 3/81 Robinson Reactor Coolant System Leak C102 W. Lanning on January 29, 1981 3/81 AE00 Safety Concerns Associated with Pipe C103 S. Rubin Breaks in the BWR Scram System 4/81 Millstone Unit 2 Loss of 125 V DC Bus C104 M. Chiramal Event on January 2,1981 12/81 Report on the Calvert Cliffs Unit 1 Loss C105 E. Imbro of Service Water on May 20, 1980 i
| |
| Reactor Engineering Evaluations i
| |
| ; Date Subject No. Author 1/19/81 Degradation of Internal Appurtenances in E101 E. Brown LWR Piping 1/30/81 Sequoyah Unit 1 Loss of Annunciation E102 M. Chiramal 3/2/81 Engineering Evaluation of Feedwater E104 S. Sands Transient and System Pipe Break at Turkey f
| |
| Point 3
| |
| . 3/31/81 Water Hammer During Restart of Residual E105 J. Huang Heat Removal Pumps 3/31/81 Water Hanmer in the Steam Condensing Mode E106 J. Huang of the Residual Heat Removal System Operation 4/17/81 Peach Bottom Unit 3 Occurrence on E107 F. Ashe February 25, 1981 E
| |
| B-30
| |
| | |
| Reactor Engineering Evaluations Date Subject No. Author l 4/21/81 Hatch Units 1 and 2 - Alternate Offsite E108 M. Chiramal Source Interlock with Emergency Diesel Generators l
| |
| l 4/24/81 Potential Common Mode Failure of Diesel
| |
| ~
| |
| E109 M. Chiramal l Generators 4/29/81 Requirements of the Preferred or Offsite E110 F. Ashe Power System 5/22/81 Evaluation of High Pressure Safety Elli E. Imbro Injection Pump Operability without Service Water 6/15/81 Inoperability of Instrumentation Due to E112 M. Chiramal Extreme Cold Weather 6/24/81 Deliberate Pump Trip at Browns Ferry Unit 2 E113 W. Lanning on April 6, 1981 6/24/81 Control System Failures that Could Cause E114 F. Ashe or Exacerbate Nuclear Power Plant Accidents 7/8/81 Additional Information on Events at TMI-2 E115 H. Ornstein During Preoperational Testing (9/5-12,1977) 7/14/81 Failure of 8 Phase Main Trasformer and E116 H. Chiramal Subsequent Fire in the Transformer Area -
| |
| North Anna Unit 2 7/16/81 Events at TMI-2 During Preoperational E117 H. Ornstein Testing 7/20/81 Setpoint Drift Occurrences for the Barton E118 F. Ashe Model 288 Instrument 7/22/81 Loss of Residual Heat Removal Capability E119 E. Imbro at Brunswick Units 1 and 2 8/6/81 Ignition of Gaseous Waste Decay Tank at E120 H. Ornstein San Onofre Unit 1 - July 17, 1981 8/28/81 Crystal River 3 Engineered Safeguards E121 M. Chiramal Relay Failures I
| |
| 1 B-31 i
| |
| | |
| Reactor Engineering Evaluations
| |
| , Date Subject No. Author 9/4/81' AEOD Concern Regarding Inadvertent Opening E122 H. Ornstein of Atmospheric Dump Valves on 88W Plants During Loss of Integrated Control System Nonnuclear/ Instrumentation i
| |
| 9/15/81 Immediate Action Memo: Common Cause E123 H. Ornstein
| |
| ! Failure Potential at Rancho Seco -
| |
| Desiccant Contamination of Air Lines l
| |
| 9/24/81 Review of Information on Purge Valves E124 E. Brown 10/15/81 Engineering Evaluation Report on Shutdown E125 G. Lanik Cooling System Heat Exchanger Failures at Oyster Creek, August 1981 4
| |
| 10/16/81 Event Sequences Not Considered in the E126 F. Ashe Design of Emergency Bus Control Logic 10/28/81 Pressure Boundary Degradation Due to Pump E127 W. Lanning Seal Failure at Arkansas Nuclear One 11/10/81 Inoperable Teledyne Solenoid Valves E128 F. Ashe 12/7/81 Brunswick Unit 2 Diesel Generator Jacket E129 M. Chiramal 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 with E131 F. Ashe Inverter Output Due to Lack of Circuit Tuning 12/23/81 Abnonnal Wear Encountered on Aloyco Swing E132 T. Cintula Check Valves Installed in the Low Pressure Safety Injection System at Palisades 4/15/81 Inadequacies in Periodic Testing of E133 M. Chiramal Combustion Engineering PWR Reactor Protection System B-32
| |
| | |
| Nonreactor Engineering Evaluations Date Subject No. Author Interim Report on Brown Boveri Betatron N101 E. Trager
| |
| ) 3/16/81 Calibration Check Source 3/26/81 Irradiator Incident at an Agreement State N102 K. Black Licensee's Facility (Becton-Dickinson, Broken Bow, Nebraska) 4/13/81 Interim Report on the October 1980 Fire at N103 E. Trager the Sweetwater Uranium Mill 4/30/81 Interim Report on the January 2,1981 Fire N104 E. Trager at the Atlas Uranium Mill 5/18/81 Interim Report on Tallings Impoundment N105 E. Trager Liner Failure at the Sweetwater Uranium Mill 1
| |
| 8/12/81 Review of Reports of Leaking Radioactive N106 E. Trager Sources 12/10/81 Engineering Evaluation of Fire Protection N107 E. Trager 3
| |
| at Nonreactor Facilities i 12/16/81 Notes on AEOD Review of Emissions from N108 E. Trager Tritium Manufacturing and Distribution i Licensees l
| |
| 1 B-33
| |
| | |
| TABLE B-7 REPORTS ISSUED IN CY 1980 Case Studies Date Subject No. Author 7/80 Report on the Browns Ferry Unit 3 Partial C001- S. Rubin 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 at C003 W. Lanning l Arkansas Nuclear One, Units 1 and 2 11/80 AEOD Actions Concerning the Crystal River C004 H. Ornstein Unit 3 Loss of Nonnuclear Instrumentation and Integrated Control System Power on February 26, 1980 12/80 AE0D Observations and Recommendations C005 E. Imbro Concerning the Problem of Steam Generator Overfill and Combined Primary and Secondary Side Blowdown
| |
| : Reactor Engineering Evaluations Date Subject No. Author 4
| |
| 3/20/80 Crystal River Nuclear Power Plant Decay E001 H. Ornstein Heat Closed Cycle Cooling Water Pumps /DCP-IA and DCP-1B S/23/80 BWR Jet Pump Integrity E002 S. Rubin 6/19/80 Comparison of Reactor Coolant Pump Events E003 E. Brown Contained in LERs, NPRDS, RECON, and Plant Records 7/10/80 Data Sumaries of Licensee Event Reports of E004 H. Ornstein Pumps at U.S. Comercial Nuclear Power Plants, January 1, 1972 to April 30, 1978 3
| |
| B-34 i
| |
| i
| |
| | |
| R actor Engineering Evaluations (Continued)
| |
| Subject No. Author
| |
| +
| |
| Date Operational Restrictions for Class IE 120 E005 M. Chiramal 7/15/80 V AC Vital Instrument Buses E006 W. Lanning 8/4/80 Loss of Residual Heat Removal at Beaver Valley, LER 80-031 Potential for Unacceptable Interaction E007 S. Rubin
| |
| , 8/18/80 Between the Control Rod Drive System and Nonessential Control Air System at Browns Ferry Operational Restrictions During Surveil- E008 M. Chiramal 8/20/80 lance Testing of Emergency Diesel Generators E009 W. Lanning Failure of Containment Isolation Valves 8/22/80 at Zion Tie Breaker Between Redundant Class IE E010 M. Chiramal 8/27/80 Buses - Point Beach Units 1 and ?
| |
| Concerns Relating to the Integrity of a E011 E. Imbro 8/29/80 Polymer Coating for Surfaces Inside Containment Salem Unit 1 - Solenoid Valve of E012 M. Chiramal 9/12/80 Containment Fan Coil Unit Service Water l Flow Control Valve Excessive Main Feedwater Transient E013 J. Creswell 9/12/80 Transient at Crystal River Unit 3 - E014 H. Ornstein 10/8/80 September 30, 1980 10/20/80 January 3,1977 Quad Cities Unit 1 Loss E015 G. Lanik-of Air Event and its Effects on Scram Capability 10/21/80 Flow Blockage in Essential Equipment at E016 E. Imbro ANOCausedbyCorbicula_sp.(AsiaticClams) 10/29/80 Engineering Evaluation of Steam Generator E017 W. Lanning Overfill B-35
| |
| | |
| Reactor Enoineering Evaluations Date Subject No. Author 12/18/80 Potential Failure of BWR Backup Scram E018 M. Chiranal (ModeSwitchinShutdown) 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 l
| |
| l B-36 e 1 --
| |
| | |
| =,acgoa. an u. =ucLiu muurou co .um~ i .. oar ~uw. .. -e =c v.,n.....,,
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| E',"3#2'' '
| |
| BIBLIOGRAPHIC DATA SHEET NUREG-1272 ne mir ktaianiasi 2 vat L. AND 50s t af 3 Li .v. S L.~.
| |
| Report to U.S. Nuclear Regulatory Commission on I Analysis an Evaluation of Operational Data - 1986 a o ^" a't' w"''" m oo,1. ....
| |
| fj
| |
| * **"oa '5' Apri1 # 1987
| |
| .opiavoarissuto j
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| oo~17 l l May/ 1987 l e n_,oa.,~aoaa.~e..o ~ .~o ..u o .ooa s uo i. c , . ,aoac, oa. u~o ~uuna Office for Anal is and Evaluation of Operational Data , , , ~ oa .~,~u...a U.S. Nuclear Re atory Commission Washington, DC 55
| |
| ,o .o~som,,o oa. ~a .1,o,, .... .~ .. u o . ooa ... ,, ,, c ,
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| g,.nno,agoa, Annual Same as 7. above. . n ~w wv i a w n-- . e CY 1986 i,su,,uon ... ou.
| |
| This annual report of the U. . Nuclear _gulatory Commission's Office for Analysis and Evaluation of 0 rational ata (AE0D) is devoted to the activities performed during c endar ear 1986. Comments and observations are provided on operating experien at uclear power plants and other NRC licensees, including results fr lected AE0D studies; summaries of abnormal occurrences involving U.S. nucle plants; reviews of licensee event reports and their quality, reactor scra perience from 1984 to 1986, engineered safety features actuations, an t trends and patterns analysis program; and assessments of nonreactor and .edi 1 misadministration events. In addition, the report provides the year end.st us of all recommendations included in AEOD studies, and listings f all A reports issued from 1980 through 1986.
| |
| l l
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| . . _ , . ~ . . . . . . . . . ~ ,.o ,ca..oa, ,,...........
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| Nuclear Plant- ]perating Experience; Trends a Patterns; Abnormal Occur-rences; Lice ee Event Report Assessment; New nt Experience; Reactor Unlimited l
| |
| Scrams; Eng' eered Safety Features; Air Systems; urbine Overspeed Trips; j Nonreactor vents; Fuel Cycle Events; Radiation E osure; Nuclear ' * " cu* " c ' ^ u '"c ^' **
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| {
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| Ma erio,a ' AE00 Recommendation Tracking System; AE0
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| . .oi,1,,t,ias eport Listing '
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| k e u. S, C0vtahmEnt pm!NTIN". OFF ICl e1997- 181 692:63114
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