ML20199L192

Office for Analysis and Evaluation of Operational DATA.1997 Annual Report (Reactors)

ML20199L192
Person / Time
Issue date:	11/30/1998
From:	NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:
References
NUREG-1272, NUREG-1272-V11-N01, NUREG-1272-V11-N1, NUDOCS 9901270152
Download: ML20199L192 (325)

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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications NRC publications in the NUREG series, NRC regu- NRC Public Document Room lations, and Title 70, Energy, of the Code of Federal 2121 L Street, N.W., Lower Level Regulations, may be purchased from one of the fol- Washington, DC 20555-0001 lowing sources:

< http://www.nrc. gov /NRC/PD R/pd r1.htm >

1 -800-397-4209 or locally 202-634-3273

1. The Superintendent of Documents U.S. Government Printing Office Microfiche of most NRC documents made publicly RO. Box 37082 available since January 1981 may be found in the Washington, DC 20402-9328

<http://www. access.gpo. gov /su docs > the v{icinity of nuclear power plants. The loc 202-512-1800 of the LPDRs may be obtained from the PDR (see previous paragraph) or through: i

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<http://www.nrc. gov /NRC/NUREGS/

Springfield, VA 22161-0002 SR1350/V9/lpdr/html>

<http://www.ntis. gov /ordernow> l 703-487-4650 Publicly released documents include, to name a few, NUREG-series reports; Federal Register no-The NUREG series comprises (1) technical and ad- tices; applicant, licensee, and vendor documents ministrative reports, including those prepared for and correspondence; NRC correspondence and international agreements, (2) brochures, (3) pro- internal memoranda; bulletins and information no-ceedings of conferences and workshops, (4) adju- tices; inspection and investigation repc ts; licens-dications and other issuances of the Commission 88 event reports; and Commission papers and and Atomic Safety and Ucensing Boards, and their attachments.

l (5) books.

Documents available from public and special tech-nical libraries include all open literature items, such A single copy of each NRC draft report is available as books, journa! articles, and transactions, Feder-free, to the extent of supply, upon written request al Register notices, Federal and State legislation, as follows: and congressional reports. Such documents as theses, dissertations, foreign reports and transla-Address: Office of the Chief Information Officer tions, and non-NRC conference proceedings may Reproduction and Distribution be purchased from their sponsoring organization.

Services Section U. S. Nuclear Regulatory Commission Copies of industry codes and standards used in a Washington, DC 20555-0001 substantive manner in the NRC regulatory process E-mail: <GRW1 @NRC. GOV > are maintained at the NRC Ubrary, Two White Flint North, 11545 Rockville Pike, Rockville, MD l Facsimile: 301 -415-2289 ,

l 20852-2738. These standards are available in the i A portion of NRC regulatory and technicalinforma- library for reference use by the public. Codes and tion is available at NRC's World Wide Web site: standards are usually copyrighted and may be ,

purchased from the originating organization or, if  ;

<http://www.nrc. gov > they are American National Standards, from-l American National Standards Institute All NRC documents released to the public are avail- 11 West 42nd Street able for inspection or copying for a fee, in paper, New York, NY 10036-0002 microfiche, or, in some cases, diskette, from the <http://www. ansi.org >

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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications NRC publications in the NUREG series, NRC regu- NRC Public Document Room lations, and Title 10, Energy, of the Code of Federal 2121 L Street, N.W., Lower Level Regulations, may be purchased from one of the fol- Washington, DC 20555-0001 lowing sources: < http://www.nrc. gov /N R C/PD R/pdr1.h'.m >

1 -800-397-4209 or locally 202-634-3273 SG e nt rinti e Microfiche of most NRC documents made publicly RO. Box 37082 available since January 1981 may be found in the Washington, DC 20402-9328 Loca Public Document Rooms (LPDRs) located in

<http://www. access.gpo. gov /su docs >

the v{icinity of nuclear power plants. The locatio 202- 512 -1800 of the LPDRs may be obtained from the PDR (see previous paragraph) or through:

2. The National Technical information Service <http://www.nrc. gov /NRC/NUREGS/

Springfield, VA 22161-0002 SR1350/V9/lpdr/html>

<http://www.ntis. gov /ordemow>

703-487-4650 Publicly released documents include, to name a few, NUREG-series reports: Federal Register no- i The NUREG series comprises (1) technical and ad- tices; applicant, licensee, and vendor documents J ministrative reports, including those prepared for and correspondence; NRC correspondence and international agreements, (2) brochures, (3) pro- Internal memoranda; bulletins and information no-ceedings of conferences and workshops, (4) adju- tices; inspection and investigation reports; licens-dications and other issuances of the Commission ee event reports; and Commission papers and and Atomic Safety and Licensing Boards, and the,r i attachments. 1 (5) books. Documents available from public and special tech-nical libraries include all open literature items, such A single copy of each NRC draft report is available as books, journal articles, and transactions, Feder-free, to the extent of supply, upon wntten request al Register notices, Federal and State legislation, as follows: and congressional reports. Such documents as theses, dissertations, foreign reports and transla-Address: Office of the Chief Information Officer tions, and non-NRC conference proceedings may ,

Reproduction and Distribution be purchased from their sponsoring organization.  !

Services Section U. S. Nuclear Regulatory Commission Copies of industry codes and standards used in a Washington, DC 20555-0001 substantive manner in the NRC regulatory process E-mail: <GRW1 @NRC. GOV > are maintained at the NRC Library, Two White Flint Facsimile: 301 -415-2289 North, 11545 Rockville Pike, Rockville, MD 20852-2738. These standards are available in the A portion of NRC regulatory and technical informa- library for reference use by the public. Codes and ,

tion is available at NRC's World Wide Web site: standards are usually copyrighted and may be i purchased from the originating organization or, if i

<http://www.nrc. gov >

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Previous Reports in Series ,

t The following semiennual or annual reports have been prepared by the Office for Analysis and Evaluation of Operational Data (AEOD).

e SemiannualReport, January-June 1984, AEODIS405 September 1984 e SemiannualReport. July- December 1984, AEODIS502, April l985 )

e AnnualReport 1985, AEODIS601, April 1986 l

)

e Rej> ort to the U.S. Nuclear Regulatory Commission on Analysis and Evaluation ofOperational Data  :

1986, NUREG.1272, AEODIS701, May 1987  :

e Report to the US. Nuclear Regulatory Commission on Analysis and Evaluation of Operational Data 1987, NUREG-1272, AEOD/S804 Vol. 2, No.1, Power Reactors, October 1988 l Vol. 2, No. 2, Nonreactors, October 1988 l

e Officefor Analysis and Evaluation of Operational Data 1988 Annual Report, NUREG-1272 Vol. 3. No.1, Power Reactors, June 1989 Vol. 3, No. 2, Nonreactors, June 1989 ,

e Officefor Analysis and Evaluation ofOperational Data 1989 Annual Re, ort, NUREG-1272 Vol. 4, No.1. Power Reactors, July 1990 Vol. 4, No. 2, Nonreactors, July 1990 e Officefor Analysis and Evaluation of Operational Data 1990 Annual Report, NUREG- 1272  :

Vol. 5, No.1, Power Reactors, July 1991  !

Vol. 5, No. 2, Nonreactors, July 1991 .

e Officefor Analysis and Evaluation ofOperational Data 1991 Annual Report, NUREG-1272 ,

Vol. 6, No. I, Power Reactors, July 1992, Vol. 6, No. 2, Nonreactors, August 1992 e Officefor Analysis and Evaluation ofOperationalData 1992 Annual Report, NUREG-1272  ;

Vol. 7, No.1. Power Reactors, July 1993 Vol. 7, No. 2, Nonreactors, October 1993 '

e Officefor Analysis and Evaluation of Operational Data I993 Annual Report, NUREG-1272  :

Vol. 8, No.1, Power Reactors, November 1994 Vol. 8, No. 2, Nuclear Materials, May 1995 e Officefor Analysis and Evaluation ofOperational Data 1994-FY95 Annual Report, NUREG- 1272 ,

Vol. 9, No.1, Power Reactors, July 1996 '

Vol. 9, No. 2, Nuclear Materials, September 1996 l Vol. 9, No. 3, Technical Training, September 1996  !

e Officefor Analysis and Evaluation ofOperational Data 1996 Annual Report, NUREG-1272 Vol.10, No.1, Power Reactors, December 1997 Vol.10, No. 2, Nuclear Materials, December 1997 Vol.10, No. 3, Technical Training December 1997

Reactors l

l l ABSTRACT l

l The United States (U.S.) Nuclear Regulatory In this annual report, NUREG-! 272, Vol. I 1, the Commission's (NRC) Office for Analysis and staff describes activities conducted during FY 1997.

Evaluation of Operational Data (AEOD) has The report is published in three parts. In NUREG-published reports of its activities since 1984. The 1272, Vol. I1 No. I, covering power reactors, the first report covered January through June of 1984, staff presents an overview of the operating experi-and the second report covered July through Decem- ence of the nuclear power industry from the NRC ber of 1984. After those first two semiannual perspective. In NUREG-1272, Vol. I 1 No. 2, repons. AEOD published annual reports ofits covering nuclear materials, the staff presents a activities. Beginning with the repon for 1986, review of the events and concerns associated with AEOD Annual Repons have been published as the use of licensed material in applications other NUREG-1272. Beginning with the report for 1987, than power reactors. In NUREG-1272, Volume 11.

NUREG-1272 was published in two parts, No. I No. 3, covering technical training, the staff presents cavering power reactors and No. 2 covering the activities of the TechnicalTraining Center in nonreactors (changed to " nuclear materials" with support of the NRC's mission. Throughout these the 1993 report). AEOD changed its annual report reports, whenever information is presented for a from a calendar year (CY) to a fiscal year (FY) calendar year, it is so designated. Fiscal year report, and added part 3 covering technical training, information is designated by the four digits of the beginning with the combined Annual Repon for CY fiscal year.

1994 and FY 1995, NUREG-1272, Vol. 9.

l l

l

Reactors CONTENTS Page Abstract. . . . .. . . . . . . .iii Abbreviations. . . . . . .ix Executive Summary . . . . . xiii i Introduction. . . .. I 2 Operating Experience Feedback . . .5 2.1 Operating Performance . .. . .5 2.1.1 Reactor Scrams . . . .5 2.1.2 Engineered Safety Features Actuations . . .5 2.1.3 Significant Eveats . .

.8 2.1.4 Safety System Failures. . . 9 2.1.5 Forced Outage Rate . . . . .9 2.1.6 Equipment Forced Outages per 1000 Commercial Critical Hours . . .9 2.1.7 Collective Radiation Exposure . . .9 2.1.8 Cause Codes. . . . . .. . .9 2.1.9 Unit Operating Factors. . ... .10 2.1.10 Statistical Analysis of Some Trends. . . . .10 2.1.11 Nuclear Reactor Safety Performance . . . .14 2.2 Abnormal Occurrences . . . . . . 14 2.3 Radiation Exposures From Reactors and Nonreactors . . .14 2.3.1 Sources of Radiation Exposure. . .. . .14 2.3.2 Exposures for Reactor and Nonreactor Applications. . . . . . .15 2.3.3 Overexposures in Reactor and Nonreactor Applications . . . .15 2.4 Allegations at Commercial Nuclear Power Plants . .I8 3 AEOD Reliability and Risk Activities . . . .19 3.1 Accident Sequence Precursor Program. . . .19 3.1.1 Results for CY 1995 . .20 3.1.2 Analysis of CY 1982-1983 Events . .. .22

,22 3.1.3 Results for CY 1996.

3.2 System Reliability Studies . .

.22 3.2.1 Reactor Core Isolation Cooling System Reliability. , . .25 3.3 Common-Cause Failure Database... .28 4 AEOD Repons. . . .33 i 4.1 AEOD Activities to Identify and Address Safety Issues . .33 4.2 Case Studies . . .33 4.2.1 Grid Performance Factors . .

. .33 4.3 Special Studies . . . . . .34 4.3.1 Oconee Electrical System Design and Operation . .34 4.3.2 Fire Events - Feedback of U.S. Operating Experience . . .35 V

1997 AEOD Annual Report i

I 4.4 Engineering Evaluations . .. .. . . . .36 4.4.1 Review of Industry Effons to Manage Pressurized-Water Reactor Feedwater Nozzle. l Piping, and Feedring Cracking and Wall Thinning. .. . . . 36 {

4.4.2 Review of Undetected Failures of Safety Systems . . .. . .37 4.4.3 Nuclear Power Plant Cold Weather Problems and Protective Measures . .38  ;

4.5 Technical Reviews . ^

. . .. .. .. . . . . . . .38 4.5.1 Design Errors in Nuclear Power Plants . . .. . . . . 38 5 Operating Experience Data. . . . .. . . . .

.41 5.1 Licensee Event Reporting .. .. . . . . . . .41 5.2 U.S. Operational Experience Databases . . .41 5.3 Reliability and Availability Data. )

. . . . . . .43 6 Incident Response. . .. . . . . . .

.45 6.1 NRC Operations Center . . . . . . .45 6.2 Emergency Response . . .

. .45 6.3 NRC Operations Center Data for 1997. . . .. . . .46 6.4 Emergency Exercises . .. . . .. . .46 6.5 State Outreach .. .. . . . .48 i 6.6 Coordination with Other Federal Agencies. .

.48 l i

7 incident Investigation Program . . . . . . .53 7.1 Incident Investigation Teams . . . . . . . . . .53 7.2 Augmented Inspection Teams . .. . .

. .54 7.2.1 Zion 1. .. .

. .54 7.2.2 Oconee 3 . . . .

. . . . . .55 7.2.3 Oconee 1 . . . . . . 55 7.2.4 Clinton . .. . . . . . . . . . . . .56 8 Diagnostic Evaluation Program . .. . . . . . . 59 8.1 Diagnostic Evaluation Teams.. . .. . . .

.59 8.2 Special Evaluation of Clinton Power Station. . . . . .. . .59 9 Committee to Review Generic Requirements. .. . . . . .. . . .63 9.1 CRGR lssues. . . . . . . .. . . . .64 9.2 Value Added by the CRGR Review . . . . . .

. .65 10 International Exchange ofInformation. . . .71 10.1 incident Reporting System. .. .

.71 10.2 International Suppon Activities . .71 10.3 Lisbon Initiative Activities . . .

.72 10.4 Limited Participation in the International Nuclear Event Scale . .73 Appendices A Plant Operational Experience Data. .. .

. .A B Summary of 1997 Abnormal Occurrences . .

. .B C Reports Issued in 1997. .

. .C ,

D Reports issued From CY 1980 Through CY 1996. . .D NUREG-1272, Vol. I 1, No. I vi

Reactors E AEOD Technical Reports by Category . .E F Status of AEOD Recommendations.. .. F G Status of NRC Staff Actions for Reactor Events Investigated by incident Investigation Teams . . .G H Status of NRC Staff Actions involving Potential Generic issues Resulting From Diagnostic Evaluation Team Findings . . .H I Status of NRC Staff Actions involving Potential Generic Issues Resulting from the NRC/INPO Team Review of the Effects of l Hurricane Andrew on Turkey Point Units 3 and 4. .1 J Status of NRC Staff Actions Resulting from the Independent Safety Assessment of the Maine Yankee Atomic Power Station . . .J Figures 2.1 Performance Indicators - Annual Industry Averages . .6 2.2 Trends in Reactor Scram Causes . .7 2.3 1997 Reactor Scrams by Actisity in Progress . . .7 2.4 Trends in ESF Actuations of Selected Systems . . .8 2.5 Nonlinear Regression Fit for Automatic Reactor Scrams (y = Ac8 '). .11 8

2.6 Nonlinear Regression Fit for Safety System Actuations (y = Ac '). . 1I 8

2.7 Nonlinear Regression Fit for Significant Events (y = Ac ') . .12 2.8 Two-Value Fit for Safety System Failures (y = A,, or A,) . .12 2.9 Constant Fit for Forced Outage Rate (y = A). .13 8

2.10 Two-Model Fit for Equipment Forced Outage Rate (y = A,c ' or A,) . .13 3.1 Distribution of ASP Conditional Core Damage Probabilities by Calendar Year . . .20 3.2 Plant-Specific Estimates of RCIC System Unreliability for 24 Hour Missions . .26 3.3 Plant-Specific Estimates of RCIC System Unreliability for Short Term and Long Term Missions . .

.27 3.4 Plant-Specific RCIC System Unreliabilities With Recovery for Short Term Missions vs. Low-Power License Date . .29 3.5 RCIC System Unreliabilities With Recovery for Short Term Missions by Calendar Year. .29 3.6 RCIC Unplanned Demands Per Plant Operating Year . .

.30 3.7 RCIC Failures Per Plant Operating Year. .

.30 3.8 Plant-Speci'ic RCIC System Failures Per Operating Year vs. Low-Power License Date . .

.31 6.1 Communicators for the Protective Measures Counterpart Link and Health Physics Network. .49 6.2 Executive Team Member and Protective Measures Team Director . .49 6.3 NRC Status Officer Briefing the NRC Executive Team . .50 6.4 Discussica in the Executive Team Room . .50 vii

- .- , . . - . - - . . . ~ . . . - . . - - . - - . - . _ - - . _ . - --

1997 AEOD Annual Report Tables t,

2.1 Summary of PI Regression Analysis Trends.. . .. ... . . .. . . ..... . ... . . ... . . . . .. 10 ;

2.2 Abnormal Occurrences per Year at U.S. Nuclear Power Plants .. ... . .. .. . . . . . . . . . . . . . . .15

~

2.3 Annual Occupational Exposure Data for Commercial Reactors CY 1973 and CY 1991 to CY 1996. . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . .. 16 ,

2.4 Occupational Exposure Data for NRC Licensees CY 1996. ..... . . . . . . . . . . . . . . . .. . . .. 16 2.5 Annual Occupational Overexposures for NRC Licensees CY 1991 to cy 1996 . . . ... .. . .. ... .. . .. . .. . . . . . . . . . . . . . . . . . . . . . .... . . . . . . .17 2.6 Annual Occupational Overexposure Rate at NRC Reactor '

and Radiography Licensees CY 1991 to CY 19% . . . . . . . . . .. .. . .. .. . . . . . . . . 17  ;

3.1 Accident Sequence Precursors for CY 1995 . .. . . . . . .. . . . . . . . . . . . . . . . .. . 21 >

3.2 Accident Sequence Precursors for CY 1996 . ... .. .. . . . . . . . . . .. .. . . . . . . . . . . . . . .23  :

3.3 System Performance Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. 24  ;

5.1 Number of LERs Submitted by Year . . . . . . . . . . . . ..... . . . . . . . . . . . . . 42  ;

5.2 Percentage of LERs Submitted in CY 1997

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by 10 CFR 50.73 Requirement . . . . . . .

. . . . .. ... . . . . . . . . . . . . . ... . 42  :

6.1 Events Reported to the NRC Operations Center in 1997 . . .. ..... . . . . . . . . . ... . . . 4 7 6.2 Classification of Events Under Licensee Emergency Plans CY 1989 to 1997. . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . .. 47 6.3 Alerts Reponed by Power Reactor Facilities in 1997. . . . . . .. .. . . . . . . .. . 48 7.1 Reactor Incidents for Which AITs Were Established in 1997 .... . . . . . . . .. .. . . . . . . .53 4

1 1

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i NUREG-1272, Vol. I1, No. I viii

Reactors ABBREVIATIONS A D ac alternating current DBR design basis reconstitution ADS automatic depressurization system de direct current AEOD Analysis and Evaluation of Operational DE diagnostic evaluation / Division of Data (NRC Office for) Engineering (NRR)

AFW auxiliary feedwater DEP Diagnostic Evaluation Program (NRC)

AISI American Iron and Steel Institute DER design electrical rating AIT augmented inspection team DET diagnostic evaluation team ALARA as low as reasonably achievable DG diesel generator AMS- Allegations Management System DRPM Division of Reactor Program (NRC) Management (NRR) )

ANSI American National Standards Institute DSA diagnostic self-assessment i AO abnormal occurrence ARG Accident Review Group (NRC) E j ASME American Society of Mechanical ECCS emergency core cooling system l Engineers EDG emergency diesel generator ASP accident sequence precursor EDO Executive Director for Operations ATWS anticipated transient without sera (NRC)

EFO equipment forced outage B EFW emergency feedwater B&W Babcock & Wilcox Company EHC electrohydraulic control BWR boiling-water reactor EMEB Mechanical Engineering Branch (NRR)

.BWRO BWR Owners Group EOP emergency operating procedure EPIP emergency plan implementing C procecure CDP core damage probability EPIX Equipment Peformance and Information CCDP conditional core damage probability Exchange System (INPO)

CCF common-cause failure EPRI Electric Power Research Institute CCW component cooling water ERDS Emergency Response Data System CFR Code of Federal Regulations (NRC)

CIV containment isolation valve ESF engineered safety feature CRD control rod drive ESW essential service water CRGR Committee to Review Generic Requirements (NRC) F CS core spray FOR forced outage rate CSNI Committee on the Safety of Nuclear FR Federal Register Installations (NEA) FRERP Federal Radiological Emergency CST condensate storage tank Response Plan cSv centisievert FRP Federal Response Plan  !

CVCS chemical and volume control system FSAR final safety analysis report l CW circulating water FW feedwater l CY calendar year ix

1997 AEOD Annual Report G mrem millirem GE General Electric Company MSB multi-assembly sealed basket GL generic letter MSIV main steam isolation valve MSL main steamline H MSR moisture separator reheater MTC multi-assembly transfer cask HOO headquarters operations officer (NRC)

HP MW megawatt high pressure

. MWe megawatts-electric HPCI high-pressure coolant injection HVAC .. MYAPS Maine Yankee Atomic Power Station heating, ventilation, and air condition-mg HX N heat exchanger NEA Nuclear Energy Agency (OECD) g NMAC Nuclear Maintenance Assistance Center IAEA (EPRI)

International Atomic Energy Agency IC isolation condenser NMSS Nuclear Material Safety and Safeguards I&C - instrumentation and control (NRC Office of)

ICS integrated control systern NOUE notification of unusual event IEEE

. NPRDS Nuclear Plant Reliability Data System '

Institute of Electrical and Electrome Engineers (INPO)

IIP NPSH net positive suction head Incident Investigation Program (NRC)

IIT NRC 'U.S. Nuclear Regulatory Commission meident investigation team ILRT integrated leak rate test NRR Nuclear Reactor Regulation (NRC IN Office of) information notice (NRC)

NSSS nuclear steam supply system INES International Nuclear Event Scale NTSB National Transportation Safety Board

( A AI INEX 2 NUDOCS Nuclear Documents System Second International Emergency Exercise O

INPO Institute of Nuclear Power Operations IPE individual plant examination OECD Organization for Economic Cooperation IPEEE individual plant examination for and Development external events OIG Office of the Inspector General (NRC)

IRM intermediate range monitor OP Office of Personnel (NRC)

IRS Incident Reporting System (NEA)

ISA independent safety assessment P ISI inservice inspection PDR Public Document Room (NRC)

PECB Events Assessment and Generic Com-L munications Branch (NRR)

LER licensee event report PI performance indicator LOCA loss-of-coolant accident PIM plant issues matrix LOOP loss of offsite power PORY power-operated relief valve LP low pressure PPR plant performance review PRA probabilistic risk assessment M PTM plant temporary modification PWG principal working group (NEA)

MCC motor control center PWR pressurized-water reactor MDC maximum dependable capacity MG motor-generator MOOS maintenance-out-of-service R MOV motor-operated valve RCIC reactor core isolation cooling NUREG-l 272. Vol. I 1, No.1 x

Reactors l RCM Response Coordination Manual (NRC) SE significant event (NRC) l RCP reactor coolant pump SER Safety Evaluation Report (NRC)

RCS reactor coolant system SET Special Evaluation Team (NRC)

REIRS Radiation Exposure Information and SFP spent fuel pool Reporting System (RES) SG steam generator I rern roentgen equivalent man SI safety injection l RES Nuclear Regulatory Research (NRC SOV solenoid-operated valve Office of) SRM staff requirements memorandum / source RG regulatory guide range monitor l

l RHR residual heat removal SRV safety / relief valve )

RI Region I(NRC) SSA safety system actuation (NRC) )

King of Prussia. PA SSES Susquehanna Steam Electric Station I Ril Region II (NRC) SSF safety system failure (NRC)

Atlanta, GA SW service water Rlli Region 111 (NRC)

Lisle, IL T RIV Region IV (NRC) T/C thermocouple Arlingten, TX TDP turbine-driven pump RPS reactor protection system TEDE total effective dose equivalent RPV reactor pressure vessel TGIS toxic gas isolation system RTD resistance temperature detector TM1 Three Mile Island RTM Response Technical Manual (NRC) TS technical specifications RWCU reactor water cleanup U

S U.K. United Kingdom SALP systematic assessment oflicensee U.S. United States

' performance (NRC)

SBGT standby gas treatment V SDLC standby liquid control VAT vulnerability assessment team  ;

SCSS Sequence Codmg and Search System V&V validation and verification l l

(NRC) l l

l xi Abbreviations

Reactors I

EXECUTIVE

SUMMARY

General Nuclear Reactor Safety The Office for Analysis and Evaluation of Opera- Performance tional Data ( AEOD) was created in 1979 to provide Through the many activities of AEOD, trends in a strong, independent capability to analyze and overall safety performance of power reactors may evaluate operational safety data associated with be inferred. The Performance Indicator (PI) and i

activities licensed by the United States (U.S.) Accident Sequence Precursor (ASP) Programs of Nuclear Regulatory Commission (NRC). AEOD i: AEOD have been applied to analyze data and also responsible for the NRC's Incident Respon:e, information in a consistent manner over a number Incident Investigation. Diagnostic Evalusion, and of years. As measured by these indicators, U.S.

TechnicalTraining Programs. In addition AEOD industry average safety performance has improved i l supplies administrative and technical support to the steadily. The number of precursors occurring

( NRC's Committee to Review Generic Require- annually has declined significantly since 1984, and monts. AEOD also obtains industry feedback on five of the performance indicators - automatic l

these activities. scrams, safety system actuations, significant events.

The AEOD programs constitute the essential equipment forced outages per 1000 critical hours,

! independent review and assessment of power and collective radiation exposure - show statisti-reactor and nuclear materials safety performance, cally significant improvement since CY 1985. The and complement the reviews of operating events number of initiating events resulting in scrams has performed by the regions, the Office of Nuclear declined significantly, and this is reflected in fewer Reactor Regulation, and the Office of Nuclear and less complicated plant transients (safety system Material Safety and Safeguards. AEOD performs a actuations, significant events, and accident se-quality verification function that ensures feedback quence precursors). However, equipment problems ofimportant operational safety lessons. AEOD persist, as evidenced by the percentage of scrams findings and recommendations continue to be caused by equipment failure (the leading cause of addressed through generic correspondence, in the all scrams), the relatively constant values for safety resolution of generic issues, and in initiatives taken system failures and forced outage rate since CY by industry. 1985, and the lack of improvement in equipment forced outages per 1000 critical hours since CY AEOD has published annual reports of its activities 1994. Industry average unit availability and since 1985. AEOD changed its annual repon from a capacity factors also improved considerably calendar yrv (CY) to a fiscal year (FY) report between CY 1985 and CY 1995. However, this beginning uth the combined Annual Report for CY was due not to fewer forced outage hours, but to 1994 and FY 1995, NUREG-1272, Vol. 9. In this greatly reduced scheduled outage hours.This is a report, NUREG-1272, Vol. I 1, No.1, covering consequence of longer fuel cycles, which result in power reactors, the staff presents an overview of the greater intervals between refueling outages, and of FY 1997 operating experience of the nuclear power shorter refueling outages. Industry average avail-industry from the NRC perspective. Throughout ability and capacity factors declined in CY 1996 this report, whenever information is presented for a and 1997 primarily bme there were a number calendar year, it is so designated. Fiscal year of plants in extended shutdown.

3 information is designated by the four digits of the

! fiscal year, i

4 l

xiii

__ - _ _ _ _ _ . _ . . . _.-- _~-- - - - - - - -

1 1997 AEOD Annual Report I l

l l

Operating Experience Feedback required. Almost all radiation doses from nuclear j power plants are occupational doses, that is, doses ,

l Performance Indicators. The PI program includes to nuclear power plant employees and contractors l l eight indicators: automatic scrams while critical, who work at the plant. The economics of operating j safety system actuations, significant events, safety a plant creates a strong impetus to reduce exposures i l system failures, forced outage rate, equipment- and achieve ALARA (as low as reasonably achier i forced outages per 1000 critical hours, collective able) objectives. As a result, utility violations of l l radiation exposure, and cause codes. PI reports are NRC limits on personnel exposure are rare, and the {

issued annually in January of each year with data vast majority of nuclear power plant personnel have l through the previous fiscal year; they are distributed annual exposures far below NRC regulatory limits widely within the NRC and to Mi operators of l specified in 10 CFR Part 20. This is believed to I commercial nuclear power plants, and they are result primarily from the licensees' extensive dose- [

available to the public in the NRC's Public Docu- reduction efforts. Some measures that reduce j ment Room and the Local Public Document collective exposure are an effective maintenance  !

Rooms, as well as from the Government Printing program, experienced and well-trained personnel, a  !

Office and the National Technical Information good water chemistry control program, effective Service. They are used in various NRC programs decontamination and cleanup practices, good fuel ,

such as the Senior Management Meeting process cladding integrity, effective radiation exposure  !

and in plant-specific analyses of safety perfor- control programs, good housekeeping, and an alert mance. Industry average Pls have been used for the health physics staff. The average dose per worker past nine years to monitor trends in the safety has declined from 0.94 centisievert (cSv [ rem])in  ;

performance of the commercial nuclear power CY 1973 to 0.28 cSv (rem) in CY 1996 (the latest  ;

industry. The PI report is issued as NUREG-1187, year for which data are available). [

beginning with the report for 1997.  !

Abnormal Occurrences. AEOD administers the AEOD Reliability and Risk Activities  :

Commission's program for reporting abnormal  ;

occurrences ( AOs) to Congress. AOs are incidents Accident Sequence Precursor Program. The ASP Program uses probabilistic risk assessment

{

or events that the Commission determines are ,

significant from the standpoint of public health and (PRA) techniques to evaluate the conditional core l safety. Beginning in 1996. AO reports are issued damage probabilities of nuclear power plant events  ;

annually with data through the previous fiscal year. and equipment unavailabilities. It serves as one of The AO report for 1997 (NUREG-0090, Vol. 20) several tools to ensure that important operating  ;

contains one AO at a nuclear power plant. It in- lessons are not overlooked. The program uses a j volved the loss of two of three high pressure rig r us method that integrates actual initiating l injection pumps at Oconee Nuclear Station Unit 3. events, plant conditions, and the reliability and  ;

The number of AOs at nuclear power plants since availability of standby safety equipment into an 1988 has remained low, averaging two per year. verall quantitative assessment, which is ex-  !

pressed as a conditional core damage probability l Radiation Exposures.The NRC regulates both (CCDP) for initiating events and an increase in l reactor and nonreactor applications of nuclear core duge probability (ACDP) for conditions i materials. All NRC licensees are required to and equ pment unavailabilities. Results of the ASP j monitor employee exposure to radiation and Program are peer-reviewed by outside consultants,

{

radioactive materials at levels sufficient to demon- other NRC offices, and the affected licensees.  !

strate compliance with the occupational dose limits They are use<i in NRC initiatives such as the i specified in Part 20 of Title 10 of the Code of Senior Management Meeting process. There were }

, Federal Regulations (CFR). Licensees of power ten events in CY 1995 that met the criterion for an l l reactors are required by 10 CFR 20.2206 to provide ASP event (CCDP or ACDP greater than 10 6). l

) to the NRC annual reports of exposure data for two caused by initiating events and eight due to i individuals for w hom personnel monitoring is conditions or equipment unavailabilities. In j 2 '

l NUREG-1272, Vol. I I, No. I xiv

(

Reactors i

l i

CY 1996 there were 11 events that resulted in and data records contained in the Nuclear Plant l 14 precursors,6 caused by initiating events Reliability Data System. These events are contained (1 occurred while the unit was shut down) and 5 in the common-cause failure database.This data-due to conditions or equipment unavailabilities. base represents the most complete collection of common-cause failure events in the world. A .

System Reliability Studies. AEOD uses opera-software system stores CCF events in a fomiat that tional data to determme the reliability of risk allows PRA analysts to review the events and sigmficant systems m U.S. commercial reactors.

develop understanding of how they occurred. A The data are obtained from licensee event reports report of the results of this effon is presented in (LERs), special reports, and monthly operating four volumes.

experience reports. Each of the four studies com-pleted to date covers the period from CY 1987 through CY 1993. Repons on the reliability of the Results of AEOD Studies high-pressure coolant injection (HPCI) system in .

l the 23 boiling-water reactors (BWRs) with HPCI AEOD studies of operational experience are systems, the emergency diesel generator (EDG) broadly dissemmated throughout the nuclear l

c mmunity and to the public.They provide a basis trains in all plants with EDGs, and the isolation ]

condenser (IC) system at the five BWRs with that f r decision-making based on operational experi-  :

system were completed in prior years. The founh ence. AEOD used a systematic process to nominate, j l

study, on the reliability of the reactor core isolation pri ritize, and select safety issues to be studied, and J cooling (RCIC) system was completed in 1997. The c ntinued its effons to more effectively communi-RCIC system calculated unreliability (including cate the lessons of operating experience through a recovery) based on operating experience in which V 'I#IY II *'

RCIC was required to inject to the reactor vessel for In 1997 the AEOD staff reviewed a broad spectrum less than 15 minutes is 0.04. Failures attributed to of data and issued one case study, eight special j the stan sequence were the leading contributor studies, three engineering evaluations, and one i (48%) to this unreliability. The RCIC system technical review. These reports coven d a wide - I calculated unreliability (including recovery) based range of subjects, from relatively broad evaluations  ;

on operating experience in which RCIC was of grid performance and fire events, to in-depth required to inject to the reactor vessel for more than reviews of subjects such as the Oconee electrical 15 minutes is 0.08. Failure to restan the system for system design and the Reactor Core Isolation subsequent injection was the leading contributor to Cooling System reliability. '

this unreliability (41%). Comparing the estimates of RCIC system unreliability calculated from the information contained in PRA/IPEs to the estimates Operating Experience Data i (with recovery) calculated from the operating The average number of LERs per plant (excluding exp rience data revealed that most (approximately supplemental, canceled, proprietary, voluntary, and 75%) of the PRA/IPE point estimates he withm the safeguards LERs) declined from about 26 in CY uncertainty interval associated with the operating 1987 to about 11 in CY 1995, then increased again experience estimate. However, about 21% of the

-to about 14 in CY 1997. AEOD uses the Sequence PRA/IPE estimates predict better performance than identified by the estimates calculated from the "8 " # ## * ' "I ""E ""

retrieving LER information. In 1997 AEOD staff operating experience data. It was also fw d that used the SCSS data to support NRC activities such most of the PRA/IPEs do not model the RCIC

"*'"*'" "*P"' " E*E'#"" "" **" Y system in the way it is observed to be opecated in management meetmgs. The SCSS database is also a the operating experience data. .

pnmary source of operating experience information Common Cause Failure Database. AEOD has for the Office of Nuclear Reactor Regulation, the compiled common-cause failure events from LERs Office of Nuclear Regulatory Research, and the in the NRC's Sequence Coding and Search System regions. AEOD also maintains data on LERs, xv Executive Summary

1997 AEOD Annual Report monthly operating reports, and plant outages to material releases. During 1997 AEOD also sup-generate the NRC's Performance Indicator Reports. poned the Federal Emergency Management Agency (FEMA) in providing accident assessment training

. to Federal, State, and local officials on four occa-Incident Response sions. In addition, AEOD participated m. the design Operations Center, The NRC Operations Center and conduct of an FRERP workshop sponsored i provides the focal point for NRC communications by FEMA.

with NRC licensees, State agencies, and other Federal agencies about operating events. The Incident Investigation Program center contains a state-of-the-art information management system which integrates voice, video. In 1997 the NRC conducted four Augmented and data systems to provide timely and effective Inspection Team inspections at nuclear power information flow. reactors. There were no events that were judged to have a level of safety significance high enough to l In 1997 the NRC Operations Center received 1938 warrant an Incident investigation Team investiga-immediate notifications, primarily from nuclear tion. Examples of problems found and communi-power plant licensees. Of the 1509 nuclear power cated to licensees from these Augmented {

plant immediate notifications,49 were classified ,

Inspecti n Team inspections included . improper as " Unusual Events" and 2 as " Alerts." None were classified as " Site Area Emergency" or " General f .ntr I r d manipulation, degraded high pressure ,

inFcti n Pumps, and 4160 vac circuit breaker j Emergency." The NRC entered the Monitoring

• ^I"'#"""C# I S*"** '

Phase of the Nonnal Mode for three of the l Unusual Eve its.

Emergency E,xercises. Emergency exercises are Diagnostic Evaluation Program held periodically to ensure that response organiza- In 1997, as an alternative to an NRC diagnostic tions of the NRC, the licensees, the States, and evaluation, the EDO approved the conduct of a other Federal agencies are proficient in dealing with I censee-initiated independent self assessment (ISA) each type of emergency. In 1997 the NRC head- of the Clinton Power Station. In association with i quarters and regional offices participated in four this approval, the EDO directed the Office of NRR full-scale emergency exercises and three limited to form a Special Evaluation Team (SET). The panicipation exercises at nuclear power plants. The goals of the SET were to evaluate the effectiveness NRC initiated a new exercise element in 1996 of the licensee's ISA effort, evaluate the actions of called ingestion exercises, and conducted three of licensee management and staff with respect to safe them at nuclear power plants. plant operation, determine the root cause(s) of the State Outreach. AEOD continued its aggressive safety-related hardware and performance problems, State Outreach Program designed to increase and and obtain additional information on safety perfor-improve interactions with States during events and mance to allow NRC senior managers to male an exercises. Outreach sessions were conducted with informed assessment of plant safety performance.

22 states and numerous licensees. AEOD also The SET, which was led by an NRC manager, was provided Emergency Response Data System approximately half the size of a DET and evaluated training for all states within 10 miles of a nuclear the same functional areas as a DET; in combination power plant. with the IS A, the SET provided for a diagnostic-leve1 evaluation of the facility. The conclusions of Coordination With Other Federal Agencies.The the SET were similar to and consistent with the root NRC continued its participation with other Federal

, causes identified by the IS A. Specifically, the SET agencies in the completion of annexes to the made the following findings:

Federal Response Plan (FRP). AEOD, in coopera-tion with other Federal agencies, began to develop

• Management generally did not establish and the assessment tools needed for handling emergen- implement effective performance standards.

cies that involve chemical as well as radioactive NUREG 1272 Vol.11, No. I xvi

Reactors l

l = CPS programs, processes, and procedures did A EOD is also the principal U.S. technical represen-not consistently provide defense in depth to tative on reactor operating experience to the NEA's assure plant activities were conducted in a Committee on the Safety of Nuclear Installations' safe manner. (CSNI) Principal working Group 1, " Operating

. . Experience and Human Factors." In addition.

• Problem identification was inconsistent and AEOD is a participant in the Expen Group on evaluation and corrective actions were generally Nuclear Emergency matters, established to improve

' " II*** **

the quality of national and intemational nuclear l

• Management did not ensure that the infrastruc- emergency arrangements.

ture was suitable to support major changes.

Lisbon ?nitlativ' Activities. AEOD is continuing to assist the cegulatory authorities of Russia and International Exchange of Ukraine in the improvement of their own capabili-Infortnation ties to respond to nuclear power plant emergencies.

The AEOD staffis helping to establish basic but The Incident Reporting System. The incident functional emergency response systems of plans, Reporting System (IRS) is an international system procedures, facilities, communications, and training jointly operated by the Nuclear Energy Agency of in each country. AEOD coordinates its activities the Organization (or Economic Cooperation and with those of the Department of Energy and other Development (OECD/NEA) and the International agencies of the U.S. Government as well as with Atomic Energy Agency (IAEA) of the United related activities of other countries and organiza-Nations. Through the IRS, NEA and IAEA member tions. AEOD also assisted Ukraine in establishing countries exchange information on safety-signifi- an incident reporting and operating experience cant operational events at nuclear power plants that feedback system, which was completed in 1997.

are of generic interest. In 1997 AEOD prepared and . . .

submitted 29 IRS repons that addressed individual 1 imited Participation in the International operational events and various generic concerns.

Nuclear Events Scale.The NRC has participated in a l.imited manner in the International Nuclear AEOD also reviewed approximately 100 IRS Ennts & ale (WES) since Decemy N WES ,s i j repons from other countries and disseminated the ,

applicable information to the NRC staff and to the a ranking system that, m pnnciple, is used to j pr mptly and consistently communicate to the Institute of Nuclear Power Operations. l public the safety sigmficance of reported events at j International Support Activities. AEOD ex- nuclear installations worldwide. After a two year '

changes information and ideas on a variety of topics trial period, the NRC decided to continue indefi-ofintemational interest, such as high burnup fuel nitely its limited participation in INES. There were and control rod insenion problems, undetected no events at nuclear power reactors during 1997 for safety system failures, and common cause failures. which INES reports were submitted.

1 1

l xvii Executive Summary

Reactors

! 1 INTRODUCTION

< The Office for Analysis and Evaluation of Opera- AEOD reviews and evaluatic... include the follow-l tional Data ( AEOD) was created in 1979 to ing specific functions:

l provide a strong, independent capability to analyze

• .identifying operational safety data needed :o l and evaluate operational safety data associated with activities licensed by the United States (U. supp rt s fety analyses, and developing agency-S.) Nuclear Regulatory Commission (NRC). The wide reporting of these data and the methods and I

office serves as the focal point for the assessment systems to retrieve them of operational events through the collection, .

ana!yzing operational safety data and identifying review, analysis, and evaluation of the safety safety issues that require new or additional NRC performance of both reactor and nuclear materials staff action facilities. To accomplish this mission for commer-cial nuclear power reactors, AEOD (1) collects,

• developing a coordinated system for feedback of analyzes, and disseminates operational data; (2) operational safety information to NRC offices, identifies important events and their associated licensees, and other organizations, as appropriate safety concerns and root causes; (3) assesses the a serving as the focal point for coordinating adequacy of corrective actions taken to address I

. generic operational safety information and data safety concerns; (4) determines the generic appli-systems with industry, foreign governments, and cability of events to other nuclear power plants; other agencies involved with the collection, (5) assesses trends in performance; (6) evaluates analysis, and feedback of operational data operating experience to quantify and to improve the understanding of the risk-significance of a developing and implementing the agency events; (7) conducts reliability studies of risk- program on reactor performance indicators for important systems;(8) analyzes human perfor- use by senior managers mance in operating events; and (9) produces

+

periodic Performance Indicator, Abnormal Occur- analyzing selected operating events using the Accident Sequence Precursor Program to gain rence, and Accident Sequence Precursor reports.

insight into events and to improve the under-AEOD is also responsible for the NRC's Incident standing of events from a risk perspective Response, Incident Investigation, Diagnostic Evaluation, and Technical Training Programs. The

+ estimating the reliability of risk-significant Incident Response Program produces a coordinated systems and identifying the dominant contribu-NRC emergency response to ongoing events t rs to unreliability through the NRC Operations Center. The Incident .

estimating common-cause failure rates, loss of Investigation Program offers a stmetured NRC offsite power frequencies, and initiating event investigative response to significant operational frequencies based on operating experience events according to their safety significance. The Diagnostic Evaluation Program independently

• studying the impact of human performance assesses licensee performance at selected reactor

• preparing the Abnormal Occurrence Report to facilities. The Technical Training Program develops Congress and presents initial and continuing technical training for NRC staff and contractors. In addition.

• continuously staffing the NRC Operations AEOD supplies administrative and technical Center to screen reactor and nuclear materials j support to the NRC's Committee to Review Generic events, and any other information reported to the Requirements. AEOD also obtains industry feed- ccer, to ensure appropriate NRC response to back on t'hese activities. reported events 1

/

1997 AEOD Annual Report

+ developing, in consultation with other NRC FY 1997 operating experience of the nuclear power offices, the NRC policy for response to incidents industry from the NRC perspective, including and emergencies, as well as assessing the NRC trends of some key performance measures. The response capabilities and performance repon also includes the principal findings identified in AEOD studies of power reactor events and issues

+ developing policy, procedures, and program during the year as well as a summary ofinforma-requirements for NRC mcident investigations of tion from licensee event reports, independent safety sigmficant operational events assessments, and the NRC Operations Center.

- tracking the recommendations and staff actions Throughout this report, whenever information is contained in AEOD studies, incident investiga- presented for a calendar year, it is so designated.

tion team reports, and independent safety Fiscal year information is designated by the four assessments until they are resolved digits of the fiscal year. The report also includes the f H wing appendices:

a developing an agency-wide technical qualifica-tion program for a broad range of technical . Appendix A contains data from 1997 to support positions within the NRC staff, and operating the the section on operational experience NRC's Technical Training Center at Chatta-

+ Appendix B lists and summarizes the 1997 nooga, Tennessee, to provide the technical abnormal occurrences training needed by NRC personnel The AEOD programs, taken as a whole, constitute

• Appendix C lists AEOD reports issued in 1997 the essential independent review and assessment of . Appendix D lists AEOD repons issued from CY power reactor and nuclear materials safety perfor- 1980 through CY 1996 mance, and complement the reviews of operating events performed by the regions, the Office of a

Appendix E presents all AEOD reports from CY Nuclear Reactor Regulation, and the Office of 1980 through 1997 sorted by subject Nuclear Material Safety and Safeguards. AEOD

• Appendix F presents the status of recommenda-performs a quality verification function that ensures tions contained in AEOD studies feedback of imponant operational safety lessons.

AEOD findings and recommendations continue to a Appendix G presents the status of NRC staff be addressed through generic correspondence,in actions resulting from the findings of NRC the resolution of generic issues, and in initiatives Incident Investigation Teams taken by industry. -

Appendix H presents the status of NRC staff AEOD consists of three divisions organized as actions involving potential generic issues follows: the Incident Response Division, compris- resulting from the findings of NRC Diagnostic ing the Response Operations Section, the Response Evaluation Teams Coordination Section, and the Operations Officer -

Appendix I presents the status of NRC staff Section; the Safety Programs Division, comprising actions invoM,g potential generic issues the Reactor Analysis Branch and the Reliability and resulting from the NRC/ Institute of Nuclear Risk Assessment Branch; and the Techmcal Tram-Power Operations team review of the effects of ing Division, comprising the Reactor Technology llurricane Andrew on Turkey Point Units 3 and 4 Traimng Branch, the Specialized Technical Training Branch, and the Technical Training Support Branch.

• Appendix J presents the status of NRC staff actions involving potential generic issues AEOD changed its annual report from a calendar resulting from the findings of the NRC Indepen-year (CY) to a fiscal year (FY) report beginning dent Safety Assessment of Maine Yankee Atomic with the combmed Annual Report for CY 1994 and Power Station FY 1995, NUREG-1272, Volume 9. In this report, NUREG-1272 Vol. I1, No.1, covering power In the report on nuclear materials, NUREG-1272, reactors, the staff presents an overview of the Vol 11, No. 2, the staff reviews the events and NUREG-1272, Vol. I1. No. I 2 l

L

, .-. -._ _.- . .~ .. .- - . . .- - _ -

Reactors concerns associated with the use of licensed mate- staff covers technical training and presents the rial during 1997 in applications other than power activities of the Technical Training Center in 1997 reactors. In NUREG-1272, Volume i 1, Pan 3, the in support of the NRC's mission.

1 1

l l

l l

l 1

I l 3 Introduction r

i

Reactors 2 OPERATING EXPERIENCE FEEDBACK 2.1 Operating Performance presented in Appendix A-1 and other plant opera-AEOD collects, analyzes, and disseminates a wide range of operational data, obtained primarily from 2.1.1 Reactor Scrams immediate notifications to the NRC Operations AEOD monhors reactor scrams that occur while the Center in accordance with Section 50.72 of Title 10 ,

affected reactor is entical. Reactor scrams can be of the Code of federal Regulations (CFR), licensee caused by initiating events that range from rela-event repons (LERs) submitted in accordance with tively mm r mcidents to precursors of accidents.

10 CFR 50.73, monthly operating reports submitted Automatic scrams are included m the PI Program in accordance with plant Technical Specifications (see Table A-l.1 of Appendix A-1). AEOD also (TS), and the database of component failures in the tr eks manual scrams and total scrams per 1000 Nuclear Plant Reliability Data System managed by critical hours (see Table A-2.1 of Appendix A-2).

the Institute of Nuclear Power Operations (INPO).

Tables A-2.2 through A-2.5 of Appendix A-2 Other operational data include 10 CFR Part 21 summarize statistical data on combined automatic reports, NRC regional inspection repons, prelimi, and manual reactor scrams.

nary notifications of events or unusual occurrences issued by the NRC, quarterly collective radiation Figure 2.2 shows the industry trend in scram causes exposure data from INPO, and allegations of for CY 1993 through 1997. Equipment failures impropriety or inadequacy received by the NRC. A remain the leading cause of scrams. Of those subset of this information is monitored in the NRC caused by equipment failures during 1997, more Performance Indicator (PI) Program: (1) automatic than half were initiated by problems in four sys-scrams while critical, (2) safety system actuations, tems: feedwater, main turbine and control, main (3) significant events,(4) safety system failures,(5) generator, and electrical. Figure 2.3 shows that, in forced outage rate,(6) equipment forced outages 1997, more than half of all scrams occurred during per 1000 commercial critical hours, (7) collective normal plant operation.

radiation exposure, and (8) cause codes.  !

. Autmnstic Reactor Scrams. Over two-thirds of the Figure 2.1 presents m . dustry-wide annual averages ,a;a occurred in the last 5 years were since CY 1985 for seven of the Pls that AEOD w. m mms.The leading causes of automatic monitors as indicators of plant performance. With scw ns, %. 6 minant initiating systems of those the exception of collective radiation expgsure' activities in progress are the same plants m extended shutdown which require Com- .c, those given for total scrams. The number of mission approval for either restart or operation aut mat scrams has decreased since CY 1993 and hve low power are excluded from the calculation

" # "'*#d '* d##'#"*#

• I 997' ofindustry average Pls. Radiation exposure can be significant during extended outages, hence these Manual Reactor Scrams. Almost one-third of all data are not excluded. Additionally, plants are scrams during the past 5 years were manual scrams

' excluded after they are permanently shut down. The r.vaber has fluctuated from year to year but has averaged about 41 per year. Since total scrams This section presents the results of analyses of have declined, the percentage of manual scrams has selected operational experience data for CY 1993 increased over the past 5 years.

through FY 1997 (note that fourth quarter 1996 data are included in both CY 1996 and 1997).This 2.1.2 Eng.ineered Safety Features Actuat. ions section also presents statistical analyses of trends over the past 10 years in six of the P!s.The opera- AEOD monitors actuations of all engineered safety tional data collected within the PJ Program are features (ESFs), a subset of which are included as 5

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Figure 2.3 1997 Reactor Scrams by Activity in Progress 7 Operating Experience

1997 AEOD Annual Report i

i safety system actuations (SS As) in the PI Program. 2.1.3 Significant Events The SSA PI includes manual or automatic actua-S gnificant events (SEs) are those events that the tions of certain emergency core cooling systems (ECCS) and actuations of the emergency ac power NRC staffidentines for the PJ Program as meeting g  ;

system in response to low voltage on a vital bus.

Data for SSAs are d in Table A-1.2 of Appendix A- +

degradation of important safety equipment  !

1.The number of SSAs has declined steadily since + l CY 1993. a maj r transient r an unexpected plant re-sponse to a transient Figure 2.4 shows the industry trend in total ESF +

degradation of fuel integrity, the primary coolant actuations for the past 5 years, including trends in i pressure boundary, or important associated i actuations of heating, ventilation, and air condition-structures  !

ing (HVAC) systems, emergency power systems, and ECCS. The number of ESF actuations has

• a reactor scram with complications l decreased each year since CY 1993. Tables A-2.6 ,

an unplanned release of radioactivity exceeding I through A-2.9 of Appendix A-2 present industry the TS or regulations  ;

data for all ESF actuations.

+

operation outside the TS limits Boiling-water reactor (BWR) plants have more . .

+

safety systems that are included in ESF counts than ther events considered sigmficant do pressurized-water reactor (PWR) plants. For Figure 2.1 shows the industry trend in the average example, an additional row is provided in Table A- number of SEs since CY 1987. The number of SEs 2.7 for the reactor water cleanup (RWCU) system in decreased from 28 in CY 1993 to 10 in 1997. Table BWRs. As shown in Table A-2.7, the isolation of A 1.'s of Appendix A-1 describes the SEs that this system accounts for a significant percentage of occurred during 1997. Table A-1.4 of Appendix A-1 ESF events. Overall, the number of isolations of the contains SE data for the fourth quarter of CY 1995 RWCU system declined during the last 5 years. (95-4) through the third quarter of CY 1997 (97-3).

4 ETotal DHVAC

, 5 Emergency Power c DECCS o

16 3

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1993 1994 1995 1996 1997 Year Figure 2.4 Trends in ESF Actuations of Selected Systems NUREG-1272, Vol. I 1, No. I 8

Reactors 2.1.4 Safety System Failures 2.1.6 Equipment Forced Outages per 1(M)0 The safety system failure (SSF) PI includes any Commercial Critical Hours  ;

actual event or condition that could prevent the The equipment forced outage (EFO) Pl is the j fulfillment of the safety function of any of 26 safety number of forced outages caused by equipment  !

systems, subsystems, or components. For a system failures in each 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of operation with the that consists of multiple redundant subsystems or reactor critical after the plant is placed into com-trains, inoperability of all trains constitutes an SSF. mercial operation. (This information is contained in An SSF may be indicative of a plant's readiness to the monthly operating reports.) The EFO rate is the respond to anticipated events and postulated inverse of the mean time between forced outages accidents. SSFs include unconditional failures caused by equipment failures. AEOD monitors the (those events or conditions that render the system EFO rate as an indicator of the effects of equipment j incapable of performing its safety function in all problems on overall plant performance.  !

situations) and conditional failures (conditions that 1 could, in cenain specific situations, e.g., a high- Figure 2.1 shows that the .mdustry average EFO rate energy line break or seismic event, prevent the rem med relatively constant over the past 4 years fter decreasing in CY 1992 and CY 1993. Table A-system from performing its safety function). Table A-l.5 of Appendix A-1 provides quarterly plant- 1.8 f Appendix A-1 contains quarterly EFO rates specific SSF data for quaners 95-4 through 97-3. f r qu rters 95-4 through 97-3.

Table A-1.6 of Appendix A-1 contains annual SSF 2.1.7 Collective Radiation Exposure i data for each plant for CY 1993 through 1997.

Licensees of power reactors are required by 10 CFR The same four system groups as reported in the last 20.2206 to submit annual repons to the NRC AEOD Annual Report continued to be the predomi- containing exposure data for each individual for ,

nant contributors to SSFs: the ECCS group, the whom monitoring is required. The PI Program containment and containment isolation systems initially included annual collective radiation group, the emergency power systems group, and the exposure for each nuclear plant, derived from the control room emergency ventilation systems group. data reported as required by 10 CFR 20.2206.  !

These four groups accounted for more than 60 13eginning in CY 1989, the PI Program included ,

percent of the failures. quarterly collective radiation exposure received from INPO, which routinely receives collective 2.1.5 Forced Outage Rate radiation exposure data from each plant on a l i

He forced outage rate (FOR) PI is calculated by  ;"arterly basis. AEOD uses the INPO data to dividing the number of forced outage hours in a provide more timely information without duplicat-period by the sum of the generator on-line hours and ing INPO's effort.

the forced outage hours. (This information is con-Figure 2.1 shows that the industry average collec-tamed m monthly operating repons.) Forced outages tive radiation exposure reported by commercial are defined as those outages required to be imtiated reactors has declined slowly since CY 1992. Table by the end of the weekend following the discovery of A-1.9 of Appendix A-1 shows quarterly collective the off-normal condition. The trend in FOR can radiation exposure data for quarters 95-4 through provide a perspective on overall plant performance.

97-3 for each plant.

Figure 2.1 shows that the FOR decreased in CY 1995 after remaining relatively constant the previ. 2.1.8 Cause Codes ous 8 years.The decrease in CY 1995 did not Tables A-1.10 through A-1.15 of Appendix A-1 continue, as FOR increased in both CY 1996 and show quarterly cause code data for each plant for 1997. Table A-1.7 of Appendix A-1 presents plant- quarters 95-4 through 97-3. The cause codes specific FOR data for quarters 95-4 through 97-3. indicator is intended to identify possible program-matic deficiencies. Cause codes are developed from data in the Sequence Coding and Search System database. The indicator captures the trends in 9 Operating Experience

1997 AEOD Annual Report administrative control problems (Table A-1.10); Report, and adds data for 1997.

licensed operator errors (Table A 1.1 l); other personnel errors (Table A-l.12); maintenance Automatk reactor scrams problems (Table A-l.13); design, construction. .

SSAs installation, or fabrication problems (Table A-! .14);

• SEs and miscellaneous (randem failures of electronic piece-parts or those due to external events)(Table . Sgps A-1.15). Industry averages are not calculated for this indicator.

• FOR

+ EFO 2.1.9 Unit Operating Factors Within the context of its safety mission, the NRC is To perform the analysis, an exponential model,y = l not usually concerned with the availability and Ac8' + C, was fitted to each Pl. In this model,y is capacity factors of nuclear power plant operations. the P1 value; x is a time increment index in years However, because good availability and capacity (with CY 1988 = 1); and A, B, and C are parameters factors require close managerial involvement in estimated from the data. Statistical tests were then day-to-day operations, efficient and effective outage perf rmed to determine if the estimated parameters management, and attention to detail, which are also A, B, and C were non-zero. In all cases except EFO, important in safe plant operation, they can be the C term was not significantly different from zero, indirect indicators of safety performance. Availabil- and was therefore set to zero. Thus the simpler ity, capacity, and outage statistics for the U.S. n nlinear modely = Ac 8' was fitted to all data commercial nuclear industry for 1997 are presented except EFO. For comparison, a linear model y = A in Tables A-2.10 through A-2.12 of Appendix A-2. + Bx was also fitted; the two models produced very The industry average unit availability increased similar fits, but the nonlinear model has the concep-from 66.2 percent in CY 1986 to 73.8 percent in tual advantage of never being negative.

1997, excluding the Browns Ferry units when they The rc ;ults of the regression analysis are summa-were in long-term regulatory shutdown.

rized m Table 2.1 and depicted in Figures 2.5

. through 2.10. Unplanned scrams while critical, 2.1.10 Statistical Analysis of Some Trends . . . .

SSAs, and SEs have statistically sigmficant expo-As part of an assessment of PI trends, AEOD nential model fits, indicating that a trend is discern-perfccmed a regression analysis to evaluate the rate ible. The EFO PI had a nonlinear trend with C equal of change in the following Pls. This analysis to zero for CY 1988 to CY 1993, and was then updates a similar analysis in the 1996 Annual constant for the years CY 1994 to 1997. The SSF Table 2.1 Summary of Pi Regression Analysis Trends Performance Indicator 1988 to 1997 Automatic reactor scrams Slowly decreasing nonlinear trend SSAs Slowly decreasing nonlinear trend SEs Slowly decreasing nonlinear trend SSFs Level over each of two periods FOR Level EFO Slowly decreasing nonlinear trend over one period and then level NUREG-1272 Vol. I1 No.1 i0

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I1 Operating Experience

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NUREG-1272, Vol. I1, No.1 I2

Reactors .

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13 Operating Experience

1997 AEOD Annual Report and FOR Pls have neither a nonlinear nor a linear caused by equipment failure (the leading cause of trend over the 10-year period; FOR was constant all scrams), the relatively constant values for safety (level) during this period and SSFs were modeled system failures and forced outage rate since CY as having two different mean values over the 1985, and the lack of improvement in equipment j periods before and after January 1,1994. As seen in forced outages per 1000 critical hours since CY l

Figure 2.8, the number of SSFs is consistently 1994. Industry average unit availability and capac-lower after CY 1994 but increased again in CY ity factors also improved considerably between CY 1996. This drop was due primarily to the applica- 1985 and CY 1995. However, this was due not to tion of an improved SSF definition in 1994 and a fewer forced outage hours, but to greatly reduced decrease in the number of SSFs reported in LERs. scheduled outage hours. This is a consequence of Using the improved SSF definition, a safety system longer fuel cycles, which result in greater intervals  ;

declared inoperable per Technical Specifications, between refueling outages, and of shorter refuel-which in the past would have been classified as an SSF, would not be counted as an SSF if the licensee ing outages. Industry average availability and {

capacity factors declined in CY 1996 and 1997 [

produces and documents in the LER an analysis primarily because there were a number of plants in which demonstrates that the system is capable of extended shutdown.

performing its safety function. The decrease m  !

LERs reporting SSFs appears to be a consequence of a reduction in the number of SSFs discovered 2.2 Abnormal Occurrences i during design basis reconstitution efforts. Because ,

of this notable change, a model different from the Section 208 of the Energy Reorganization Act of nonlinear model described previously was fit to 1974 (PL 93-438) identifies an abnormal occur- .

the data. Different means were calculated for the rence (AO) as an unscheduled incident or event that  ;

periods from CY 1988 through CY 1993 and from the NRC determines to be significant from the CY 1994 through 1997. These means and their standpoint of public health and safety. The Federal associated confidence intervals are shown in Reports Elimination and Sunset Act of 1995 (PL Figure 2.8. 104-66) requires that AOs be reported to Congress annually. Consequently, the NRC publishes the  :

2.1.11 Nuclear Reactor Safety Performance annual AO report on a fiscal year basis.

Through the many activities of AEOD, trends in i overall safety performance of power reactors may The AO report for 1997 (NUREG-0090, Vol. 20) be inferred. The Performance Indicator (PI) and c ntains one event at a nuclear power plant. Tius -

Accident Sequence Precursor (ASP) Programs of event involved the loss of two of three high pressure AEOD have been applied to analyze data and injection pumps at Oconee Nuclear Station Unit 3.  ;

information in a consistent manner over a number This AO is described in Appendix B to this repon.

of years. As measured by these indicators, U.S.

Table 2.2 shows the number of AOs that have i mdustry average safety performance has improved occurred at nuclear power plants from CY 1987 to steadily. The number of precursors occurrmg 1997. The number has remained low, averaging two annually has declined sigmficantly since 1984, and five of the seven performance indicators - automatic P#' I#"

scrams, safety system actuations, significant events,  !

equipment forced outages per 1000 critical hours, 2.3 Radiation Exposures From l; and collective radiation exposure - show statisti-cally significant improvement since CY 1985. The Reactors and Nonreactors number ofinitiating events resulting in scrams has declined significantly, and this is reflected in fewer 2.3.1 Sources of Radiation Exposure ,

and less complicated plant transients (safety system According to the National Council on Radiation I actuations, significant events, and accident se- Protection and Measurements, the average total 1 quence precursors). However, equipment problems effective dose equivalent (TEDE) to a person in the persist, as evidenced by the percentage of scrams United States is approximately 0.36 centisieverts l

NUREG 1272,Vol. I1, No. I 14

Reactors Table 2.2 Abnormal Occurrences per Year 2.3.2 Exposures for Reactor and at U.S. Nuclear Power Plants Nonreactor Applications Year No. of AOs The NRC regulates both reactor and nonreactor CY 87 3 applications of nuclear materials. All NRC licens-CY 88 3 ees are required to monitor employee exposure to CY 89 4 radiation and radioactive materials at levels suffi-CY 90 I cient to demonstrate compliance with the occupa-CY 91 0 ti n I d se limits specified in 10 CFR Part 20.

CY 92 3 Licensees f p wer re ctors, and those involved in CY 93 1 industri i radi gr phy, the manufacture and distri-CY 94 bution of radioactive materials, fuel fabrication and 2*

1995 3* P' C*SSI"E' I *~I"**I f"di "C'I"" * **" DISP S I' 1996 2 and independent spent fuel storage, are required by

997 3 10 CFR 20.2206 to give the NRC annual reports of exposure data for individuals for whom personnel

• Includes one event from the fourth quarter of CY 94 monitoring is required.

Table 2.3 summarizes the information reported by (cSv)(360 millirem [ mrem]) per year, mostly from licensees of commercial reactors from CY 1991 to natural sources of radiation. The average person in CY 1996. For purpose of comparison, CY 1973 has the United States receives a TEDE of about 0.05 also been included. Table 2.4 compares the occupa-cSv (50 mrem) per year from medical applications. ti nal exposure of reactor licensees with that of The entire fuel cycle, including operation of reac- nuclear materials licensees for CY 1996. For more tors, contributes less than 0.001 cSv (1 mrem) mf rmation on radiation exposures in nuclear per year. All other human-controlled sources of materials applications, see the AEOD Annual radiation combined add up to a TEDE of approxi- Report, Nuclear Materials, NUREG-1272, Vol. I 1, mately 0.006 cSv (6 mrem) per year. No. 2.The data are subject to change as more information becomes available; this may cause  !

The economics of operating a nuclear power plant minor changes in the data published from one year creates a strong impetus to reduce exposures to to the next.

plant employees and contractors who work there and to achieve ALARA (as low as reasonably 2.3.3 Overexposures in Reactor and achievable) objectives. As a result, utility violations Nonreactor Applications of NRC limits on personnel exposure are rare, and Although commercial reactor occupational expo-the vast majority of plant personnel have annual sures have been maintained at a low level, a few exposures far below the NRC regulatory limits verexp sures c ntinue t ccur.The number of specified in 10 CFR Part 20. The average measur-ccup tional overexposu@es m NRC-licensed able TEDE per reactor worker has been reduced re ctor and nuclear materials facih, ties for CY 1991 from 0.94 cSv (940 mrem) per worker in CY 1973 through CY 1996 is given in Table 2.5. In every to 0.28 cSv (280 mrem) per worker in CY 1996 (the ye r shown, the number of individuals overexposed latest year for which data are available). This is

. in nuclear materials applications has equaled or believed to result primarily from the licensees, exceeded the number overexposed at reactor sites.

extensive dose-reduction efforts. Some measures For more information on overexposures in that reduce collective exposure are an effective nonreactor applications, see the AEOD Annual maintenance program, experienced and well-trained Report on Nuclear Materials, NUREG-1272, Vol.

personnel, a good water chemistry control program'

11. No. 2.

effective decontamination and cleanup practices, good fuel cladding integrity, effective radiation The number of overexposures and the number of exposure control programs, good housekeeping, and workers with measurable doses for reactors and an alert health physics staff. NRC-licensed radiographers (the nuclear materials 15 Operating Experience

_. . - - = . .

1997 AEOD Annual Report 4

Table 2.3 Annual Occupational Exposure Data for Commercial Reactors CY 1973 and CY 1991 to CY 1996 No. of Average Collective Workers Measurable TEDE with TEDE per l

No. of (person- Mcasurable Worker i Year Reactors cSv [ rem]) TEDE (cSv [ rem])

CY l973 24 13,962 14,780 0.94 CY 1991 115 28,528 91,085 0.31 ,

CY l992 114 29,298 94,317 0.31 CY 1993 II4 26,365 86,187 0.31 CY 1994 109 21,695 73,780 0.29 CY 1995 109 21,674 70,986 0.31 CY 1996 109 18,874 68,182 0.28 Source: Radiation Exposure Information Reporting System, funded by the Office of Nuclear Regulatory Research. All reactor data are adjusted to account for multiple counting of transient reactor workers.

Table 2.4 Occupational Exposure Data for NRC Licen ,ees CY 1996 No. of Average  !

Collective Workers Measurable  ;

No. of TEDE with TEDE per  ;

Licensees (person- Measurable Worker i Category Reporting cSv [ rem]) TEDE (cSv [ rem]) [

Reactors 109 18,874 68,182 0.28 Industrial Radiography 144 1385 2537 0.55 i

Manufacture and Distribution 36 556 1239 0.45 Fuel Fabrication and Processing 8 878 3061 0.29 Low-Level Waste Disposal 2 8 67 0.12  !

i Independent Spent -

Fuel Storage 1 54 53 1.02 Source: Radiation Exposure Information Reporting System f

NUREG-1272, Vol. I 1. No. I 16

Reactors  !

Table 2.5 Annual Occupational Overexposures for NRC Licensees CY 1991 to CY 1996 CY 1991 CY 1992 CY 1993 CY 1994 CY 1995 CY 1996 l Reactors 0 5 0 1 0 1 Industrial Radiography 2 1 1 2 1 1 Medical Facilities 2 5 3 0 0 0 Manufacture and Distribution 1 0 5 1 2 0 Other 1 3 3 0 0 0 Source: Radiation Exposure Information Reporting System category of most concern because of the high rate for the purpose of reducing their overexposures. In and magnitude of overexposures) are shown in addition, AEOD has prepared a videotape on good Table 2.6. The special radiological problems of safety practices in industrial radiography. The tape industrial radiography have been recognized for is entitled,"Taking Control: Safety Procedures for some time.The NRC has special guidance and Industrial Radiography," and was released in training document, NUREG/CR-0024, " Working December 1993.

Safely in Gamma Radiography," for radiographers Table 2.6 Annual Occupational Overexposure Rate at NRC Reactor and Radiography Licensees CY 1991 to CY 1996 I

Reactors Radiography No. of No. of Workers Over. Workers Over-with No. of exposures with No. of exposures Measurable Workers per 1,000 Measurable Workers per 1,000 Year TEDE Overexposed Workers TEDE Overexposed Workers CY 1991 91,085 0 0.00 4,649 2 0.43 CY 1992 94,317 5 0.05 4.265 1 0.23 CY 1993 86,187 0 0.00 3,007 1 0.33 CY 1994 73,780 1 0.01 2,351 2 0.85 CY l995 70,986 0 0.00 2,465 1 0.41 CY 1996 68,182 1 0.015 2,537 1 0.39 Source: Radiation Exposure Information Reporting System 17 Operating Experience

1997 AEOD Annual Report 2.4 Allegations at Commercial Appendix A-2 lists the number of allegations that

"'ere received from each site, those that remain Nuclear Power Plants open, those that have been substantiated in any The NRC receives allegations from individuals or manner, and those that contain harassment and organizations who assert some impropriety or intimidation concerns. Caution should be used in inadequacy in activities segulatcJ by the NRC. interpreting the table because a definitive break- i Allegations may be received at NRC headquaners down between fully and partially substantiated l or the regional offices. Allegations are entered into allegations is not provided, no differentiation is  !

the Allegation Management System (AMS), which made in the data between allegations having  !

is managed by the Office of Nuclear Reactor varying levels of safety significance, and each l Regulation (NRR). NRR and regional staffjointly allegation may contain one or many individual  ;

collect the allegations, determine their validity, and concems. The AMS database structure has been  !

track their resolution. AEOD analyzes trends in the recently modified to improve its capability to track l numbers of allegations received from each nuclear and analyze allegations. Starting with the 1998  ;

plant site and publishes the data in such a manner as Annual Report, the staff will differentiate between l to protect the identity of the alleger. Table A-2.13 of fully and partially substantiated allegations. i 5

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NUREG-1272, Vol. I 1, No. I 18

Reactors 3 AEOD RELIABILITY AND RISK ACTIVITIES 3.1 Accident Sequence Precursor distribution of ASP events for U.S. nuclear power pl nts f r CY 1984 through CY 1996.

Program

. The ASP Program began in 1979. Since then the The Accident Sequence Precursor ( ASP) Program staff has evaluated and documented nearly 600 uses probabilistic risk assessment (PRA) techniques precursors from reported experience for CY 1969 toevaluate the cond. .itional core damage probabih. . ties through CY 1996. Over the years, the ASP Pro-of nuclear power plant events and equipment gram has evolved such that the methodology and unavailabilities. The purpose of the program .is t results are now used routinely by the NRC.The provide a structured and systematic means of methodology continues to be improved to better quantitatively evaluating the safety sigmficance of account for plant design and operational differ-nuclear plant operating experience. The principal ences, human reliability, and changes in equip-objectives of the program are to identify and rank ment, and to provide user-friendly analytical tools.

the nsk sigmficance of operating reactor events, t Other planned improvements include incorpora-deternune their generic implications, to characterize tion of modeling and data uncertainty in each risk insights, and to document and disseminate the event analysis, a more complete set of accident evaluations for feedback to plant operators t sequences, and better containment response and promote learning from experience. ASP Program consequence evaluation.

results are published annually m the NUREG/CR-4674 series. To identify potential precursors, licensee event

. reports (LERs) or other documentation (e.g.,

An accident sequence precursor is an operational inspection reports, Incident Investigation Team event or plant condition that is an important ele- reports) of plant problems, equipment failures, or ment of a postulated core-damaging accident other operational incidents are reviewed. Event sequence. Accident sequences considered in the trees model plant responses to such challenges as ASP Program are those associated with inadequate transients, loss-of-coolant accidents (LOCAs), loss core cooling, which would be expected to result in of offsite power (LOOP) events, steam generator core damage. Precursors can be mfrequent imtiatmg tube ruptures, and anticipated transients without events or equipment failures that, when coupled scram. Operational occurrences that involve por-with one or more postulated events, could result in tions of these postulated core-damage sequences are a plant condition mvolvmg inadequate core cooling. identified. Plant equipment and human responses ,

The ASP methodology evaluates disparate elements that could affect the progression of an accident are I

of operational experience by assuming random evaluated, including actual failures that have failures for other branches of the event tree models. occurred and the probability that other failures These evaluations account for all actual or potential could occur. Fault tree linking techniques are used concurrent failures, degradations, or outages of to produce a quantitative estimate of the signifi-safety systems. The evaluations also melude cance of the reported data.

estimates of the likelihood of equipment failures and human errors and of the probability of recovery The results of the ASP analyses are considered should they occur. The figure of merit for ASP indications of the level of risk associated with analyses is conditional core damage probability operating nuclear power plants based on direct (CCDP) for initiating events and increase in core assessment of actual operating experience.The damage probability ( CDP) for conditions and precursor events from the ASP Program form a equipment unavailabilities. Events with CCDPs or unique database of historical system failures, CDPs greater than 1.010-6 are considered Accident multiple losses of redundancy, and infrequent core-Sequence Precursors. Figure 3.lshows the yearly damage initiators. Several of the recorded precursor 19

1997 AEOD Annual Report l

CONDITIONAL CORE DAMAGE PROBABILl1Y RESULTS FROM ASP PROGRAM 20 - -- - --- -

---

.0=10'-

b 1.0=10* m 1 -E- 1.0=108b

- - - - - 1.0=108

--- b 1.0=10' b 1.0a10*

1 Ou10* b 1.0=10* E Greater than 1.0=10*

15- - -- - --- -- - - - - - - - - - -

E

=

r i '

E 8 l  : W

! I IE { l g 10 - - --- : ---- - g-- ,

7- -- - - - - - - -

d_

- l g

i i fi. -

ri =

s$$i i  ! i fin  ! 9 5- -

- -- b -- -j j--- r- ---------b--

1 Q h E . '*

I q I5 I: -

l l:

o .. _

_ , L. ,___1.m,__ ,

_% _ . .%L_ %A_ %_. _

1984 1985 1986 1987 1988 1989 1990* 1991 1992 1993 1994 1995 1996 YEAR

'THE VoGTLE EVENT HAS BEEN ROUNDED UP FRoM 9 7:10 AND PLOTTED AS 1.0=1D*.

Figure 3.1 Distribution of ASP Conditional Core Damage Probabilities by Calander Year events involved equipment failure caused by dated April 1997, and are shown in Table 3.1. Ten factors, conditions, or phenomena that affected the events or conditions that occurred in CY 1995 ability of safety equipment to perform its function. resulted in ten precursors (ten units were affected).

These mechanistic failures are different from While this is consistent with the CY 1994 results, it

" random" failures or unavailabilities of equipment. is fewer than in years prior to CY 1994, and the Commercial nuclear power reactors in the United reduction is due in part to specific mitigating States now have a combined total of more than equipment and recovery meayures that were not 2000 years of operating experience. The ASP P'*** " Sly credited. The prehminary ASP analyses Program uses information gained from this experi- were reviewed by the NRC staff and the affected ence to produce an ongoing assessment of nuclear licensees. They were also mdependently reviewed plant operation.This assessment helps to identify by the Sandia National Laboratory under contract to how well plant designs and capabilities can cope the NRC. On the basis of comments received from ,

with actual operational events or conditions. the reviewers, the analyses were revised to help the NRC provide more accurate risk assessments of 3.1.1 Results for CY 1995 the events.

The results of the ASP analyses of CY 1995 events Of the ten precursors for CY 1995, eight involved are documented in NUREG/CR-4674, Vol. 23, discovered conditions or unavailabilities of equip-NUREG-1272, Vol. I 1, No. I 20

"Iti>Ie 3.1 Accident Sequence Prectrsors for CY 1995 CCDP/

ACDP Plant Type LER No. Description Date Precursors involving initiating Events Sorted by Conditional Core Damage Probability l 2.9 x 105 Comanche PWR 445/95-003,-004 Reactor trip, auxiliary feedwater (AFW) pump trip, second AFW pump 10/11/95 PeakI unavailable h 2.0 x 10 5 Arkansas PWR 313/95-005 Reactor tip with emergency feedwater (EFW) problems 04/20/95 Nuclear One, Unit I Precursors involving Equipment Unavailabilities Sorted by increase in Core Damage Probability St. Lucie i PWR 335/95-004.-005, Failed power-operated relief valves, multiple reactor coolant pump seal 08/02/95 9.3 x 10 5

-(X)6 stage failures, relief valve failure, shutdown cooling unavailable and other td problems ,

3.1 x 10' Millstone 2 PWR 336/95-002 Containment sump isolation valves susceptible to pressure locking 01/25/95 Waterford 3 PWR 382/95-002 Reactor trip and fire in turbine building 06/10/9 1.7 x 10 5 St. Lucie 2 PWR 389/95-005 Failure ef one emergency diesel generator (EDG) with common cause i I/20/95 1.3 x 10 5 failure implications i

~

Arkansas PWR 368/95-005 Loss of de bus could fail both EFW trains 07/19/95 y 1.1 x 10 5 E. Nuclear One,

$ Unit 2 E

]c-9.0 x 10' Limerick i BWR 352/95-008 Safety / relief valve failed open, scram, suppression pool strainer fails 09/11/95 09/12/95 E 7.7 x 10 6 D.C. Cook l PWR 315/95-01i One safety injection pump unavailable for six months  ;

k k 4.7 x 10

• Haddam Neck PWR 213/95-010 Multiple safety injection valves susceptible to pressure locking during 03/09/95 f!

I large break loss-of-coolant accident (LOCA) $l

[

l 1997 AEOD Annual Report ment and two involved initiating events. All but one Catawba 2 LOOP while an EDG was out of service ,

of them occurred at PWRs. Six of the precursors for maintenance had a CCDP greater than 1.0x10), j involved problems with electrical systems, although which was the highest CCDP of any of the CY none involved a total LOOP. This is consistent with 1996 precursor events. The degradation of the the results from the preceding five calendar years, essential service water (ESW) system at Wolf Creek for which about 60 percent of the precursor events caused by the formation of frazil ice under severe involved electric power issues. cold weather conditions would have resulted in the loss of all decay heat removal capability if the only 3.1.2 Analysis of CY 1982-1983 Events remaining operable ESW pump had failed. In The review and analysis of CY 1982 and CY 1983 addition, with the turbine-driven AFW pump out of events for precursors began in October 1994 to service f r m intenance when the event occurred, obtain the two years of precursor data that had the plant was vulneratie to a LOOP. A condition previously been missing. More than 10.000 LERs that was discovered at the Haddam Neck plant on were systematically screened for potential precur- August 8,1996, involved potentially madequate sors and 435 were identified for further analysis. As re clor heat removal system pump net positive a result of this analysis,109 precursors were su n w ng a p stulad largo or identified, almost equally distributed between the medium-break loss-of-coolant accident.

two years. The final report was published in NUREG/CR-4674. Vol. 24, dated April 1997. 3.2 System Reliability Studies 3.1.3 Results for CY 1996 AEOD uses operational data to determine the reliability of risk-significant systems in U.S.

The results of the ASP analyses of CY 1996 events are documented in NUREG/CR-4674, Vol. 25, commercial reactors. The data are obtained from dated December 1997, and are shown in Table 3.2. '*E# '*E"*'"*" Y E" "E reports. Four studies, each covering the period from Eleven events or conditions that occurred in CY CY 1987 through CY 1993, have been completed as 1996 resulted in fourteen precursors (thirteen of September 1997. A report on the reliability of the different units were affected). This is slightly higher than for the previous two years (9 in CY 1994 and high-pressure coolant injection (HPCI) system in the 23 BWRs with HPCI systems was completed in 10 in CY 1995).The preliminary ASP analyses 1995. The results of this study are summarized in were reviewed by the NRC staff and the affected AEOD Annual Report,1994-FY95 (NUREG-1272, licensees. They were also mdependently reviewed by the Sandia National Laboratories under contract Vol. 9, No.1). Studies of the reliability of the EDG power system in all plants with EDGs, and of the i to the NRC. On the basis of comments received isolation condenser (IC) system at the five BWRs from the reviewers, the analyses were revised t with that system, were completed in 1996 and are j help the NRC provide more accurate risk assess-summarized in AEOD Annual Report,1996 ments of the events.

(NUREG-1272, Vol.10, No.1). A report for the l

Of the 14 precursors for CY 1996,6 involved reactor core isolation cooling system was com-discovered conditions or unavailabilities of equip- pleted this year and is summarized below. Table 3.3 ment and 8 involved initiating events. Only one of summarizes the studies completed to date.

them occurred at a BWR (affecting both units at the .

site). Five (~36 percent) of the precursors involved "E "* ' ""* *"E#"'Y ## "'"

problems with electrical systems. This is different *Y**#** .

n s n tM Nghenure com spmy system in BWRs were completed this year. A study from the results of the previous 6-year period (CY ,

is pl nned of the high pressure safety injection 1990 to CY 1995), for which about 60 percent of system at PWRs. AEOD is also analyzing the the precursor events mvolved electric power issues.

i i systems for both PWRs Three of the CY 1996 precursor events were and BWRs to estimate their reliability on the basis important precursors (CCDP21.0x10 d). The of actual operating experience.

NUREG-1272,Vol.11, No. I 22

Tcht: 3.2 Accident Sequence Precursors for CY 1996 CCDP/

ACDP Plant Type LER No. Description Date At-Power Precursors invoiring initiating Events Sorted by Conditional Core Damage Probability

! 2.1 x 10 ' Catawba 2 PWR 414/96-001 LOOP with EDG B unavailable 2/6/96 2.I x 10

• WolfCreek PWR 482/96-001,-002 Reactor trip with loss of Train A of ESW and the turbine-driven AFW 1/30/96 pump unavailable 5.3 x 10' Prairie Island 1 PWR 282/96-012 Loss of power to safeguards buses on both units 7/29/96 and 2 7.0 x 10' LaSalle i BWR 373/96-007,-008 Concrete sealant fouls cooling water systems 6/28/96 and 2 5.6 x 10' Arkansas PWR 313/96-005 Reactor trip and subsequent steam generator dryout 5/19/96 Nuclear One, ej Unit i Shutdown Precursors involving Initiating Events 1.7x 10 ' Byron 1 PWR 454/96-007 Transformer bus fault causes a loss of offsite power 5/23/96 At-Power Precursors inrotring Unavailabilities Sorted by increase in Core Damage Probability 1.1 x 10 4 IIaddam Neck PWR 213/96-016 Potentially inadequate residual heat removal (RIIR) pump net positive 8/1/96 suction head following a large- or medium-break LOCA

{ 4.6 x 10 ' Seabrook PWR 443/96-003 Turbine-driven EFW pump unavailable because of a mechanical 5/21/96 g seal failure g 5.8 x 10

• Salem 1 and 2 PWR 272/96-002 Chargirg pump suction valves from the refueling water storage tank I/10/96 potentially unavailable because of pressure locking Q_ . .

E 2.9 x 10

• Iladdam Neck PWR 213/96-024 After an RilR pump seized, it was determined to be susceptible to failun- 9/1/96 D since being overhauled in 1987 @

E M 1.8 x 106 McGuire 2 PWR 370/96-002 2B EDG inoperable due to slow instrumentation response 3/6/96 g:

Z Table 3.3 Systern Performance Summary C }

c w-f Unplanned Failure g Demand Rate Unreliability Un. ;iiability Demand vs. Test o d System Unreliability Trend 'IYend Trend Consistency with PRAs/IPEs vs. Mant Age Failure Differences U

$ o I

IIPCI 0.06 Decreasing Decreasing Steady General agreement with exceptions requiring investigation None Yes lg h S EDG 0.04 Decreasing Decreasing Steady General agreement except actual None Yes $

data better after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> IC 0.02 Steady Steady Steady General agreement but different None No contributors RCIC Decreasing Steady Decreasing General agreement, but restart None Not observabic duc different in PRAs to lack of data 2

Run time 0.04

<l5 min.

Run tine >l5 0.18 min  :

I

i I

Reactors 3.2.1 Reactor Core Isolation Cooling mject to the reactor vessel for missions that are System Reliability longer than 15 minutes and up to several hours is

. 0.08. The long-term missions typically follow a Th.is study .myolved a performance evaluation of the reactor scram where feedwater is not available or l reactor core isolation cooling (RCIC) system at 29 the reactor vessel is isolated, or both. If recovery '

U.S. commercial BWRs. The evaluation was based is excluded, the long-term mission unreliability )

j on operating experience from CY 1987 through CY is 0.16. This unreliability is primarily due to '

1993, as reported in LERs. The objectives of the hardware failures associated with restaning the study were (1) to estimate the system unreliabihty turbine or the cycling of motor-operated valves.

on the basis of operating experience and to compare l these estimates with the assumptions, models, and -

The difference between the calculated data used in PRAs and individual plant examina- unreliability estimates for the long-term and tions (IPEs), and (2) to review the operating data short-term missions is due mainly to restarting from an engineering perspective to determine trends the turbine and maintaining water level in the l and patterns seen in the data and present insights reactor vessel.

into the failures and failure mechanisms associated . The estimate of RCIC system unreliability with the operation of the RCIC system.

(including recovery) for the 24-hour missions The RCIC system unreliabilities were estimated typically modeled in PRAs is 0.18. If recovery is i using a fault tree model to associate event occur- excluded, the mission unreliability is 0.43. This rences with broadly deGned failure modes, such as unreliability is dominated by failure to run (24- I failure to stan or failure to run. The probabilities for hour mission time), failure to restart, and failure l

the individual failure modes were calculated by during the recirculation mode of operation.

reviewing the failure information, categorizing each =

Figures 3.2 and 3.3 display plant-specific esti- j event by failure mode, and then estimating the m tes of RCIC system unreliability for three corresponding number of demands (both successes l specific sets of mission requirements. hgure 3.2 and failures), Twenty-one plant risk source reports (i.e., PRAs, IPEs, and NUREGs) were used for est m tes are based on operating experience data j extrapolated to the 24-hour mission typically comparison with the RCIC reliability results obtained in this study. The information extracted m deled in PRA/IPEs. Figure 3.3 displays plant- ,

speci6c estimates with separate estimates for from the source documents contain RCIC statistics for all but one of the 29 plants. The major findings shon-term (shorter than 15 minutes in duration) and long-term (longer than 15 minutes) missions.

are the following:

. For the short-term mission unreliability, failures

• The RCIC system calculated unreliability attributed to the start sequence (other than the (including recovery) on the basis of operating injection valve) are the leading contributor (48 experience data in which RCIC was required to I percent). The leading contributor to the long-inject to the reactor vessel for shon-term mis-term mission unreliability is the failure to restart sions (less than 15 minutes)is 0.04.The short-the RCIC system for subsequent injection of term missions typically follow a reactor scram in c lant (41 percent). Failure to run (FTR) is the which feedwater is available and the main steam 1 rgest contributor (36 percent), based on a 24-l isolation valves are open. If recovery is ex-hour mission time, for the RCIC system PRA-cluded, the short term mission unreliability is based unreliability. For the FTR failure mode, l 0.06. This unreliability is primarily due to the failures found during unplanned demands l failures to start, typically as a result of problems wm cauwd by human errors in operation of the in controlling turbine speed caused by either 0 w c ntroller and a spurious isolation of the human error or hardware problems that result in turbine steam supply. The spurious isolation of turbine overspeed trips.

the turbine steam supply was a failure mecha-

• The RCIC system calculated unreliability nism not identified as a major contributor to the (including recovery) on the basis of operatiry system failure probability in the PRA/IPEs.

experience data in which RCIC was te ; . a to l

25 Reliability and Risk Activities

1997 AEOD Annual Report m Operating experience unreliability & uncertainty interval i

+ PRAllPE approximate unreliability & uncertainty interval Browns Ferry 2  ; , , , ,  !

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i i i i i , , .i i i . ..ii industry mean (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) -

i i i * ' i ' ' ' ' ' ' '

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O.01 0.10 1.00 Unreliability Figure 3.2 Plant-Specific Estimates of RCIC System Unreliability for 24 Ilour Missions i NUREG-1272 Vol. I 1. No. I 26 .

l

Reactors  ;

i HH Long term mission ti'1 reliability & uncertainty interval t-A-i Short term mission unr9 liability & uncertainty interval  !

Browns Fmy 2 -

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o.001 0.010 0.100 1.000 Unreliability Figure 3.3 Plant-Specific Estimates of RCIC System Unreliability for Short Term and Long Term Missions 27 Reliability and Risk Activities

1997 AFOD Annual Report

+

Comparing the estimates of RCIC system +

For the short-term missions, a decreasing trend unreliability calculated from the information in RCIC system unreliability with respect to contained in PRA/IPEs to the estimates (with calendar year was identified by statistical recovery) calculated from the operating experi- analysis of the operating data. In addition, some ence data revealed that most (approximately 75 indication of a trend was seen in the short-term l percent) of the PRA/IPE point estimates lie unreliability with regard to low-power license )

within the uncenainty interval associated with date, but it is not a strong indication. More data the operating experience estimate. However, (i.e., more operating experience) are needed about 21 percent of the PRA/IPE estimates before this trend can be statistically verified or ,

predict better performance than is identi6ed by disproved. No statistical trends were identi6ed )

the estimates calculated from the operating with regard to long term RCIC unreliability. I experience data. These plan's fall below the 5th Figures 3.4 and 3.5 provide plots of the short-percentile of the distribution computed from the term RCIC unreliability.

operating experience data. .

When plotted against plant operating year (see

+

lt was found that most of the PRA/IPEs do not Figure 3.6), the unplanned demand frequency model the RCIC sy stem as it is observed to be exhibits a statistically signi6 cant decreasing operated in the operating experience data. trend. This is likely a result of a corresponding Specifically, the maintenance of reactor vessel decrease in unplanned plant trips, which typi-water level by either restart or recirculation, or cally inchide a RCIC system actuation. Failure both, following initial injection is generally not frequency exhibits no trend when plotted against modeled. For the PRA/IPEs that model the plant operating year (Figure 3.7). No correlation j system with the restart or recirculation or both was observed between the plant's low-power 4 modes or RCIC, the failure probabilities as- license date and the frequency of failures per signed to these modes of operation appear to be operating year (Figure 3.8). The average number too optimistic. For example, the initial failure to of failures per operating year was 0.62. This start (other than the injection valve) probabilities average frequency was observed for plants and the restart failure probabilities differ by licensed from CY 1970 through CY 1990.

about a factor of 2.6 according to the operating Two plants licensed in the 1970s and two experience data. However, the PRA/IPEs use the plants licensed in the 1980s had relatively high same probabilities for restart as for initial start, failure frequencies.

According to the operating experience data, the failure to restart contribution to overall unreliability is about a factor of 2 greater than 3.3 Common-Cause Failure Database the failure to start (other than the injection valve) AEOD and the Idaho National Engineering and contribution (27 percent versus 12 percent, Environmental Laboratory (INEEL) have developed respecth ely). and are maintaining a common-cause failure (CCF)

The operating data contained five instances in database for the U.S. commercial nuclear power which multiple systems (RCIC, HPCI, and industry. Presently, the data cover CY 1980 through CY 1995. The latest effort provides a method for sometimes reactor water cleanup) either had identifying CCF events, a collection of events from failed or had the potential to fail concurrently: .

these instr.;,ces may be common-cause failures. mdustry failure data, and a computc."ized sy3 tem for quantifying PR A param-ters and uncenainties.

The events mvolved motor-operated valves, the steam leak detection circuitry, and the turbine A CCF event consists of component failures that governors. In two of the five instances, the RCIC meet four criteria: (1) two or mon. individual and HPCI systems were affected during an components fail or are degraded, incieding failures unplanned demand. The other events were during demand, in-service testing, or deheiencies discovered during surveillance testing (2 events) that would have resulted in a failure if a demcod and other routine plant operations (1 event). signal had been recei ed;(2) components fail NUREG-1272 Vol. I1. No.1 28 1

Reactors

[ Plant-specific unreliability & uncertainty interval

- 90% conf. band on the fitted trend - Fitted trend line 0.20 0.15 ---- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

.h

.c ..

E 0.10 r-- - - -- ----

7---

---

T--' .r- - -----

E o e

s o o o o

---

y-0.05 2 '-

7--

1 -- -

_- - - __g _- - .

.- o--

+g ,

-o - -- _ _ .s---yo p-,e--

l

- .e I

i I

0.00 , / ,

1970 1974 1978 1982 1986 1990 Low-power license date Figure 3.4 Plant-Specific RCIC System Unreliabilities With Recovery for Short Term Missions vs. Low-Power License Date

[ Year-specific unreliability & uncertainty interval

- 90% conf. band on the fitted trend - Fitted trend line 0.15 0.12 - ==-----------------------------------

g- 0.09 --- ------

---

y ------ .-

---

]-----

's s E 0.06 ---"------

-s,-

e ------ ----- ------ ----- ---

g N' p~ o

"---..-. ___ u 0.03 --- --- - .----- ------

---4.---------. - "

~ ~~

0.00 1987 1988 1989 1990 1991 1992 1993 Year Figure 3.5 RCIC System Unreliabilities With Recovery for Short Term Missions by Calendar Year 29 Reliability and Risk Activities

1997 AEOD Annual Report

{ Year.epecific frequency and uncertainty interval

-- 90% conf. band on the fitted trend Fitted trend line A 2.60  !

y _ _ _

co 3 s 2.00 -- -

--v--- -------------------------------

K, t s

- s C 1.50 --- - ----

-s--- --- ___--------------------

E, ,

V 1.00 ---- I' g >--l----- 's '~h- 1" ~

~ .- -l _- - - - r - - --= - - - - -

'A~

~ ~

~

c L

' ~

g s

e x*

g 0.60 ------- - ---

2-

- - - - - - - - - i ~ . - jw- -

g--

c --

D ' ' ' ' ' '

0.00 1987 1988 1989 1990 1991 1992 1993 Year Figure 3.6 RCIC Unplanned Demands Per Plant Operating Year

{ Yeer-opecific frequency and uncertainty interval

- 90% conf. band on the fitted trend Fitted trend Ene 2.00 f

x __

o 1.60 --- ----------- --------------------- -~ -

C e

3 ~

gy s. -

e v ~

Jll: 1.00 -------- - 2 _, e .-_-

_-.: - c- ==--

-r-2 3

j

• ~

• - - - - - - . _ .,_ g 0.50

---

I----- -- E --- '

-C ---

- -__ _ h. __ __ _ _

i L '

~ .

0.00 J ~

1987 1988 1989 1990 1991 1992 1993 Year Figure 3.~ RCIC Failures Per Plant Operating Year NUREO 1272.Vol. I1 No. I 30

Reactors

[ Plant-specific failure frequency and 90% interval

- Fitted trend line -- 90% Conf. band on the fitted trend 4.00 n.

es e

M c 3.00 - - - - - - - ---------------------------------

• D E

8. .

O 2.00 ----

-~ ---,-------------- -------- -------

8.

e ,

e ~

g 1.00 ---q g - o .

C C

_ -,gd tpT ai e- .. o 0.00 r"- v$ "

,t. --

1970 1974 1978 1982 1986 1990 Low-power license date Figure 3.8 Plant-Speci5c RCIC System Failures Per Operating Year vs. Low-Power License Date within a selected period of time so that success of and identify CCF events. The CCF event coding the PRA mission would be uncertain; (3) compo- process contains guidance for the analyst to consis-nent failures result from a single shared cause and tently code CCF events. Sufficient information is coupling mechanism; and (4) a component failure recorded to ensure accuracy and consistency, occurs within the established component boundary. Additionally, the CCF events are stored in a format that allows PRA analysts to review the events and Two data sources aie used to select equipment develop an understanding of how they occurred.

failure repons to be reviewed for CCF event identification: INPO's Nuclear Plant Reliability A software system stores CCF events and indepen-Data System (NPRDS), which contains component dent failure counts, and automates PRA parameter failure information, and the NRC's Sequence estimations. The system employs two quantification Coding and Search System (SCSS), which contains models: the alpha factor and the multiple Greek LERs. These sources served as the developmental letter. These models are used throughout the nuclear basis for the CCF data collection and analysis industry. In addition, these parameter estimations system, which consists of CCF event identification can be used in a PRA to estimate basic event methodology, event coding guidance, and a soft- probability and uncertainty.

ware system to estimate CCF parameters. .

A report of the results of th.is study will be pub-The CCF event identification process entails lished as a NUREG/CR document in four volumes:

reviewing failure data to identify CCF events and Overview, Event Definition and Classification, Data counting independent failure events. The process Collection and Event Coding, and Software Refer-allows the analyst to consistently screen failures ence Manual.

31 Reliability and Risk Activities 1

1

Reactors 4 AEOD REPORTS In 1997 the AEOD staff continued to analyze and signiGeance, typically concluding that the i evaluate operating experience and publish studies of licensees' or industry's planned or scheduled l equipment problems and events as well as the corrective actions are adequate. '

analyses of the reliability ofimportant safety I P"'"

systems describe in Section 3. The staff reviewed a #P "* I ^ "" '#8 " '#*P#'#"##

are broadly disseminated.The AEOD staff contin-broad spectrum of operating experience data, .

ued eff ns to more effectively communicate the including reports submitted to the NRC by licens-ees in compliance with 10 CFR 50.72 and 10 CFR less ns of opgating expe&nce through a variety of ther fomms, mcludm.g participation m mdustry 50.73, the database of component failures in the e de committees, presentation of papers at profes-Nuclear Plant Reliability Data System (NPRDS) si n I meetings, anu attendance at owners groups maintained by the Institute of Nuclear Power and international meetmgs.

Operations, and repons of foreign reactor events.

On the basis of the staff's review and analysis of l these data, AEOD in 1997 issued one case study, 4.1 AEOD Activities To Identify '

two special studies, three engineering evaluations, and Address Safety Issues and one technical review. (In addition, AEOD  !

published its 1996 Annual Repon, one Performance AEOD uses a systematic process to nominate.

Indicator Report described in Section 2, three prioritize, and select safety issues to be studied. Six Accident Sequence Precursor Reports described in attributes are considered: risk significance, issue Section 3, and the Reactor Core Isolation Cooling complexity, requirement factors, review factors, System Reliability Study also described in Section industry initiatives, and other considerations. Infor-3.) Appendix C lists the reports issued in 1997, and mation is extracted from various databases, including Appendix D lists those issued from CY 1980 the NRC Sequence Coding and Search System, the through CY 1996. Sections 4.2 through 4.4 below Incident Reponing System of the Nuclear Energy summarize the 1997 repons, which are categorized Agency (NEA) and the International Atomic Energy as follows: Agency (see Section 10.l), the NPRDS, the NRC Allegation Management System (AMS), and NRC

• Case studies involve m.-depth analyses of

. . generic communications. Th.is approach strengthens sigmficant safety issues and document the bases AEOD's independent means of identifying and for AEOD recommendations for regulatory or

. studying generic lessons learned from operating mdustry actions. Each case study goes through a experience. Many of the areas studied involve rig'omus peer review process to ensure technical longstanding design issues that have recently been adequacy.

I addressed. In addition, AEOD has expended consid-

• Special studies involve either (1) accelerated erable effort on a series of reliability studies of rist-i investigations that result in suggestions or significant systems in U.S. commercial reactors (see recommendations for regulatory actions that are Section 3.2 of this volume).

! to be completed expeditiously, or (2) substantial efforts that report the results of significant 4.2 Case Studies

, AEOD programs.

I 4.2.1 Grid Performance Factors

• Engineering evaluations document assessments of significant operating events and suggest (AEOD/C97-01) remedial actions,if appropriate.

The reliability of offsite power is imponant to

. Technical reviews document AEOD studies of nuclear safety. Accident sequences initiated by the issues that the staff determines have little safety loss of offsite power are important contributors to 33 l

1997 AEOD Annual Report .

l l risk for most nuclear plants, in 1979, the Nuclear line; and old equipment is retired. In addition, two l Regulatory Commission identified the loss of all ac relatively recent developments are affecting the  ;

-electrical power to the nuclear plant, called station grid: independent power producers and restructur- .

blackout, as aa unresolved safety issue. Station ing of the electric industry.The North American ,

blackout was shown to be an important contributor Electric Reliability Council has adopted programs l to the total risk from nuclear power plant accidents, and procedures to deal with forecasting, normal ,

Task Action Plan A-44 was issued in July 1980.The operations, emergency conditions, and recovery  ;

Onal report on Unresolved Safety issue A-44 was from system collapse. These programs and proce-contained in NUREG-1032, " Evaluation of Station dures appear to provide the basis for assurance of Blackout Accidents at Nuclear Power Plants," June orderly operation. The grid, as managed by the l

j 1988. In NUREG-1032, the grid was assumed to be Nonh American Electric Reliability Council ,

stable and reliable. member utilities, has adequate resources to give i reasonable assurance that the reliability of the ,

This study was initiated to collect operating experi- system will be maintained under normal conditions.

ence on grid disturbances that rnay have impacted On the whole, the grid is stable and reliable, nuclear power plant operation and the availability of However, there are problems, as well as the uncer-offsite power from 1985 to the present. In order t tainties introduced by restructuring of the electric communicate the findings of the study,it became industry, that indicate the need to monitor grid necessary to inquire into the nature of the grid and conditions on a regular basis.

its operation. The grid is defined and some of its {

basic characteristics are addressed. An event in ,

1989 at the Virgil Summer Nuclear Plant caused a 4.3 Special Studies severe 1,crturbation of the Eastem Grid which resulted in the loss of offsite power to Virgil Sum. 4.3.1 Oconec Electrical System Design and mer.The Los Angeles earthquake of January 1994 Operation (AEOD/S97-01)  ;

l '

resulted in a severe perturbation of the Western In response to a request from the NRC's Executive Grid. In December 1994 the Western Grid experi- Director for Operations, AEOD performed an enced another severe perturbation that caused both ndependent evaluation of the design and operation Diablo Canyon units to scram. In 1995 a series of of the Oconee Nuclear Station emergency electrical equipment failures at a nearby substation caused the system.The evaluation provides qualitative and two Limerick units to scram. The resultant pertur- quantitative discussions of safety concerns and bation was seen throughout the Eastern Grid. In potential associated risks. The evaluation is based '

1996 the Western Grid experienced two major on operating experience, recognizing the unique disruptions, one on July 2 and the second on August design and reliance on a combination of the

10. The August 10 event led to the scramming of Keowee hydroelectric units, the Lee gas turbine  !

l Palo Verde Units 1 and 3 and both Diablo Canyon units, and the standby shutdown facility. .

units. Diablo Canyon declared its 500 kV offsite power system inoperable. The repon presents a comprehensive description of l operating experience and the emergency electrical i Four reports of potential problems related to grid power design as it existed in June 1996. It does not stability were found in which licensees reviewed provide an evaluation of the electrical system as stability analyses and discovered instabilities. The modined by changes after that time. Consequently, reviews were done for a variety of reasons, includ- only information, analyses, and testing that describe ing following up on an informat i on notice, in tr . electrical system as it was before June 1996 are l response to a question raised during an electrical contained in the report.  ;

distribution functional inspection, and follow-up on actual events. Oconee has initiated numerous activities to l address issues raised by the draft reports issued on The condition of the grid is dynamic: capacity, July 8,1996, by AEOD and the Office of Nuclear demand, and transmission patterns change over Reactor Regulation. Those activities include '

time; equipment ages; new equipment is brought on modifications to hardware and enhancements to i

NUREG-1272, Vol. I1 No. I 34 l

I J

Reactors operating procedures relating to the emergency component failure histories from the Nuclear Plant power system. In early January 1997, Oconee also Reliability Data System (NPRDS) database (also took advantage of a three unit shutdown to per- proprietary).The purpose of the study was to form a series of tests of the emergency power characterize the frequency and nature of fire event system. The activities as presented to the NRC by data from U.S. operating plants and to examine the Oconee address the concerns listed in the conclu- potential impact this updated data could have on sions of this report. Satisfactory completion of fire risk assessments.The consolidated fire events those activities should resolve these concerns, database and report appendices developed for this study provide a comprehensive and up-to-date AEOD reviewed operating experience from many compilation of information on fire events, their sources, including licensee event repons, inspection calculated frequencies, and their severity as they reports, event notification reports, the Oconee hve affected U.S. nuclear power plants.

electrical distnbution system functionalinspection report, an Oconee augmented inspection team For the study, AEOD staff reviewed and compared repon, the Keowee reliability analysis, and the plant location fire frequencies - including the Oconee individual plant evaluation. Analysis of this control room, the cable spreading room, the auxil-information was integrated with information iary building (pressurized-water reactor plants gathered irom several site visits to the Oconee only), the reactor building (boiling-water reactor station, meetings with the licensee, and input from plants only), the switchgear room, and the turbine the Committee to Review Generic Requirements, building - with those used in selected probabilistic risk assessments (PRAs).The 1965 to 1994 data Much of the AEOD review addresses issues affect" were used to provide the following:

ing the capability of the emergency electrical system to perform its intended functions following

• a listing of fire event data a loss of offsite power. The capabilities of the ,

apportionment of fire events by number, major standby shutdown facility and the Lee station als cause, and plant location were reviewed because of their use for certain scenarios. The staff's review determined that

- evaluation of the duration and frequency of fire improved system testing, selected design changes events that occurred during power and shutdown and protective features, and improved operator operations procedures and training were needed to ensure that ,

the grouping of fire events by severity during the emernney power system at Oconee w Il power operations functic.. as intended.

- the comparison of calculated fire frequencies The AEOD staff's conclusion regarding the emer- with PRA data and recent industry and NRC-

~

gency electrical system at Oconee is that a level of sponsored studies for the potential effect on fire safety comparable to that of a plant with diesel induced core damage frequency estimates generators may be achieved, assuming that the technical issues or actions raised in the AEOD

• a comparison of the duration and frequency of repon are satisfactorily resolved or completed. shutdown fire events with power operations fire events 4.3.2 Fire Events - Feedback of U.S. - a listing of smoke events, includm.g attnbutes Operatm.g Experience (AEOD/S97-03) . .

similar to f. ire events data, as input to a smoke This study updates an earlier Sandia National events database Laboratory fire events database containing operat- . . .

h m j r findings and conclusions identified in the ing experience data from 1965 through mid-1985.

rep rt are desenbed below.

For the update, fire event data from CY 1965 through CY 1994 was obtained from Licensee - Fire event frequencies and the safety signifi.

Event Reports (LERs), a proprietary fire events cance of these events have both declined slightly database (1965 through 1988) maintained by the from the 1965 through 1985 period to the 1986 Electric Power Research Institute, and fire-related through 1994 period. The most significant fire 35 Reports

1997 AEOD Annual Report event occurred at Browns Ferry ni March 1975, combustible loading, and means of detection and an event that was pivotal to the resognition of suppression. The data in this report were not safety concerns associated with firet . This event suitable for addressing these issues.

caused a reactor scram and the fire piopagated

• Forthe 1986-1994 period, the shutdown fire without suppression to affect multiple redundant j trains of safety equipment. Because c f the frequencies varied in comparison to the fire implementation of modifications and other frequencies at power for most risk significant industry activities, there have been nc. fire events plant locations used in PRAs. Since some plant locations had higher fire frequencies (contain- I of similar safety significance m the 1986 through 1994 period. There were cnly two fim events ment, reactor building, auxiliary building, I switchgear room, and diesel generator building), I that caused a scram and loss of one safety-shutdown fire events in those areas were re-  !

related train or loss of offs te power (LOOP) viewed in greater detail. The following conclu- I during this penod, compard to 10 events in the sions were drawn: I earlier period. Other fires have been severe in j terms of the magnitude and duration of combus-

• Containment fires were predominantly caused by tion (such as some turbine building fires), but welding operations and did not affect decay heat their severity in terms of challenges to the removal.

operability of safety systems has been limited. ,

A limited number of fire events affected the llowever, such fires could be important if functional operability of residual heat removal, redundant safety trains or decay heat removal decay heat removal, and emergency diesel systems were dependent on the equipment generator system trains located there.

, In general, the operating experience indicates that The fire durations durm, g power operations were the frequency and duration of shutdown fire events I generally short (less than 10 minutes). The is similar to, or less significant than, fire events information available on these short-duration occurring during power operation.

fires was not suf6cient to evaluate the probabil-ity of fire detection and suppression used in '

recent PRAs. The fire durations during shutdown 4.4 Engineering Evaluations l

were also generally short (less than 10 minutes).

Fire durations in plant locations that contain 4.4.1 Review ofIndustry Efforts to Manage systems necessary for decay heat removal during Pressurized-Water Reactor Feedwater shutdown were the same as or shorter than Gre Nozzle, Piping, and Feedring Crack-durations for the same plant locations during ing and Wall Thinning NUREG/CR-power operations.

6456 (AEOD/E97-01)

The 1986 through 1994 fire event frequencies at The objective of this study was to develop a com-power operations were lower for the control prehensive overview document pertaining to PWR room and the cable spreading room, approxi- feedwater nozzle, piping, and feedring cracking and mately the same for the auxiliary building (for wall thinning; the safety considerations; and pressurized-water reactors (PWRs)) and for the

.in ustry a ns ta en to m n ge these issues.The reactor building (for boiling-water reactors (BWRs)), and higher for the switchgear room "P 'I " "** "" #"I "E'""' "E' and the turbine building than those values used and evaluating program options to manage these m most PRAs reviewed for this study. A sensitiv- concerns. The study covers the conditions initially discovered in 1979 through 1996. Feedwater nozzle ity study, based solely on changes to the mitiator cracking was discovered in 18 PWRs from 1979 frequencies, did not indicate the potential for through 1983 as a result ofIE Bulletin 79-13, substantial changes to the overall core damage

.. Cracking in Feedwater System Piping "

frequency due to fires. Other aspects of fire analyses may be more critical to a risk assess- The study addressed Geld experience with PWR ment, including the mechanics of combustion, feedwater systems, the factors causing the damage, NUREG-1272 Vol. I1, No.1 36

Reactors design modifications, operating procedure changes, events. There were 28 events reported during the augmented inspection programs, and reprir and 1970s,6 during the 1980s, and none after that.

replacement activities carried out because of the The Gndings indicate that appropriate analysis, degradation that occurred. The effon focused on the inspection, monitoring, mitigation (including opera-feedwater system adjacent to the feedwater nozzle' tional procedures), and replacement techniques have

- where fatigue cracking and wall thinning have been been developed so that thennal fatigue and Dow-reported, meluding the main and auxiliary accelerated corrosion damage to feedwater nozzles, feedwater pipmg adjacent to the feedwater nozzle, piping, and feedrings can be managed effectively.

and the thermal sleeve, feedring, and J-tubes. The This simply means that the tools to manage them are principal areas reviewed were (1) feedwater system available. However, the management process re-design;(2) safety significance of feedwater line quires detailed knowledge of component and system rupture; (3) fatigue cracking experience; (4) flow- design, constmetion, und materials; cognizance of accelerated corrosion-induced wall thinning experi- operating procedures (e:pecially the potential for ence;(5) steam generator water hammer damage extended operation at startap or hot standby with experience; (6) degradation mechanisms; (7) automatic auxiliary feed vater control); in-depth inservice inspection methods; and (8) mitigation, understanding of factors tnat cause thermal fatigue monitoring, and replacement activities. and Dow-accelerated corrosion; and adequate training in the use of predictive analysis methods and Emphasis was placed on understanding the techni-cal aspects cf each area in order to assess the dvanced inspection techniques.The staff at several impact on managing (establish confidence tuat PWR plants have been proactive in managing this type of damage.

limits for safe operation are maintained) degrada-tion of these components. In this context, the study 4.4.2 Review of Undetected Failures of concentrated on causes, mechanisms, conditions Safety Systems (AEOD/E97-02)

(temperature, pressure, environment, etc.), inspec-tion procedures, and corrective actions to determine The purpose of this evaluation was to identify whether current technology is sufficient to manage undetected failures of safety systems in nuclear the problem. power plants. By a searching the 70 events found in the Accident Sequence Precursor ( ASP) database The operating experience review addressed for the time period 1991 through 1993, a set of 33 feedwater nozzle cracking caused by thermal fatigue; such events was identiGed. The undetected failures Dow-accelerated corrosion wall thinning of carbon were analyzed and evaluated with respect to their steel J-tubes, feedrings, and thermal sleeves in top- discovery methods, failure rate, failure causes, feed steam generators, and auxiliary feedwater lines corrective and preventive actions by licensees, and in preheat steam generators; and water hammer resulting regulatory actions.

damage in both types of steam generators.

The event data for 1991 through 1993 show that There was approximately one event per year from nearly 50 percent of the ASP events involved an 1983 until the bulletin was closed in 1991. The undetected failure. Some failures remained undetec-nozzle cracking event frequency increased to six ted for a long time - four for a period of I to 10 per year for 1992 and 1993, but no additional events years and another four for 10 to 18 years. An addi-were detected from 1994 through 1996. It appears tional four events may have gone undetected from that licensee action has been sufficient to minimize initial plant stanup, and two others were undetected flow-accelerated wall thinning in J-tubes and from the time of plant modification. This illustrates auxiliary feedwater lines and, although wall thining that unrecognized deficiencies existed and were not in feedrings and thermal sleeves was not specifi- detected by programmatic activities. These deficien-cally addressed, there was evidence of visual cies either existed since original constmetion or were inspection and repair when needed. Similarly, it introduced during plant modifications performed appears that licensees have taken sufficient action sometime after the commencement of operation.

(primarily design modifications and operating Consequently, plant modifications could be a source procedure changes) to minimize water hammer of undetected failures.

37 Reports

1997 AEOD Annual Report More than 75 percent of the failures were discov- series of reports by an industry group on cold ered during testing or during analysis and evalua- weather operating experience at nuclear power tion of operational problems. Component failures, plants. Recent operating experience suggests that, design deficiencies, or inadequate testing or mainte- despite NRC and industry communications on this nance procedures caused about 70 percent of the subject, some licensees have not effectively pro-failures. The more frequent corrective or preventive tected components whose failure could degrade actions taken by licensees were design changes safety systems. Extremely cold weather conditions (plant modifications), additional training or guid- continued to affect intake structures, process lines, ance for plant personnel, new or modified operating instrument lines, emergency diesel generator oil and procedures, and modified maintenance procedures. grease viscosities, essential chillers, electrical Although the licensees' actions appeared appropri- systems, and heating, ventilation and air conditioning ate for the speci0c events, it is not clear that their systems. Inadequate design oversight, incomplete efforts would apply to other plants. review of operating experience, and insuf6cient Testing, the most frequent method of discovery, was aftention to cold weather preparations were respon-most often associated with one of three failure sible for most of the events. Important approaches causes: component failure, design deficiency, or and measures that licensees have used to protect inadequate testing or maintenance. However, the against cold weather operational events appear m the data indicate that it would be inaccurate to ascribe "Less ns Learned" section of this report.

the discovery of the failure to the successful completion of a planned sequence to determine 4.5 Technical Reviews success or failure. For example, in approximately 60 percent of the events associated with these three 4.5.1 Design Errors in Nuclear Power failure causes, the failure or related deficiency was Plants (AEOD/T97-01) not within the anticipated or defined purpose of the test. This illustrates the need for an improved This report presents observations about how design de6nition of post-work actions (tests, procedures, err rs at operating nuclear facilities have been etc.) intended to conGrm or ensure component or discovered, and the potential impacts of regulatory system operability. r industry actions. The results of the 1985 through 1995 review indicate the followina , onclusions:

4.4.3 Nuclear Power Plant Cold Weather .

The number of reported design errors has been Problems and Protective Measures steadily dropping since the early 1990s, even (AEODIE97-03) though the number of NRC inspector-hours in A significant event on January 30,1996, at the Wolf performing engineering related inspection Creek Generating Station involved icing of cooling modules has increased during the same time water intake trash racks and traveling screens, period.

subsequent loss of an essential service water system

• The number of reported design errors discovered train, and other plant complications. This led AEOD as a result of design-basis review effons steadily to evaluate the extent of cold-weather-related decreased from 1985 to 1995, presumably due in problems at other nuclear power plants over the past pan to diminishing licensee resources allocated 6 years. A search of the operating experience data to this effon and the declining number of  ;

from 1991 through April 1997 involving ice, undiscovered latent design errors.

freezing, and low ambient temperature problems was completed. The events discussed include both

• The number of design errors discovered at any actual failures due to cold weather and some design given time was dependent on the extent of ,

or configuration vulnerabilities that could lead to initiatives taken by the NRC and the industry. l failures under postulated conditions. This report Major NRC initiatives included design-related describes 37 such events at 23 different sites. The team inspections, which began during the mid-repon also identifies previously issued cold- 1980's, and many design-related generic com-weather-related NRC information notices and a munications. During this same time period, NUREG-1272, Vol. I 1, No.1 38

Reactors l

l l

I indiistry had initiated in-depth design-basis .

Approximately 2 percent of the reviewed reviews which led to an increase in the discovery licensee event reports containing design errors of design problems. were found significant enough to be evaluated

. under the Accident Sequence Precursor (ASP)

• There appeared to be a relationsh.ip between the program. Of those design error events that were number of design-related generic communica-in the ASP database, the majority had condi-tions issued by the NRC and the number of tional core-damage probabilities between 10" licensee reported design errors. However, and 106 licensee event reports infrequently cited generic communications as the initiator for the discovery of design errors.

1 39 Reports

Reactors l

5 OPERATING EXPERIENCE DATA 5.1 Licensee Event Reporting events to support trend analyses. This database was placed on a new platform in August 1997 and is i The primary source of information about an opera- now available to NRC staff and NRC contractors tional event is the licensee event repon (LER) via the Internet to the ORNL SCSS web site, submitted as required by 10 CFR 50.73. Safety Access is also available to the general public on a performance is only one of several factors that cost recovery basis.

affects the number of LERs submitted by a licensee.

Therefore, the NRC staff does not base its assess- The SCSS database is the primary source of operat-ment of safety performance of a plant on the ing experience information for AEOD studies and number of LERs that have been submitted. Rather, f r the NRC Offices of Nuclear Reactor Regulation judgments about safety performance are based on nd Nuclear Regulatory Research, and for the an evaluation of the signi6cance of operational regions. In 1997 the AEOD staff also continued to events. For completeness, however, we have use the LER information from the SCSS database to included Table 5.1, which shows the total number supp n cenain other NRC activities, such as of LERs (excluding supplementag, canceled, peratmg expenence repons to support mspect ons and senior management meetings.

proprietary, voluntary, and safeguards LERs) submitted each year since CY 1987 by commercial In addition to the SCSS, AEOD also maintains data nuclear power reactor licensees. The overall de- on LERs, monthly operating reports, and plant crease in the number of LERs from CY 1987 to CY outages at the Idaho National Engineering &

1 1995 appears to be due to a reduction in the number Environmental Laboratory to support the NRC's of events and to changes to the reporting require- Performance Indicator (PI) Program. This P1 ments. Table 5.2 shows the percentage of LERs database contains plant-specific information on submitted in accordance with specific sections of 10 reactor scrams, safety system actuations, significant CFR 50.73. The increased number of LERs since events, safety system failures, forced outage rate,  ;

CY 1995 is attributed to the increased NRC atten- equipment forced outages per 1000 commercial l tion to design basis issues. critical hours, collective radiation exposure, and I cause codes (see Section 2.1 of this volume for a

. . detailed discussion of the 1997 Pls). In addition, 5.2 U.S. Operational Experience AEOD uses these databases to prepare special Databases studies, evaluations of selected plants, and briefing p ekages for Commission site visits.

AEOD uses the Sequence Coding and Search System (SCSS) for storing and retrieving LER Since the early 1980s, AEOD has used component information. This system, developed in the early engineering and failure data from commercial 1980s and maintained under contract with Oak nuclear power plants contained in the Nuclear Plant Ridge National Laboratory (ORNL), contains an Reliability Data System (NPRDS), a proprietary average of 150 items of information for each of the database managed by the Institute of Nuclear Power nearly 44,000 LERs submitted since 1980. The Operations (INPO). In December 1996, INPO LER descriptive text is coded into computer- discontinued reporting to NPRDS, and in January searchable sequences, with each sequence identified 1997 commercial nuclear power plants began by categories such as components, systems, person- collecting data for INPO's new database, the nel errors, causes, and corrective actions. Coding Equipment Performance and Information Exchange the LER in sequences facilitates searches.The (EPIX) System that will replace NPRDS. Data in SCSS, given a series of failures or errors for an EPIX, including archived NPRDS records, will be event or event type, can identify previous similar available with proprietary restrictions in 1998.

41

.- - - . = =--. - - - . . _ _ - _- -.. . - . -

1997 AEOD Annual Rcport Table 5.1 Number of LERs Submitted by Year

• Year No.of LERs No. of Units LERs per Unit CY 1987 2895 Ill 26 i

CY 1988 2479 110 23 l CY 1989 2356 112 21  ;

CY 1990 2128 111 19 CY 1991 1858 111 17 CY 1992 1774 111 16 t CY 1993 1400 109 13 CY 1994 1279 109 12 CY 1995 1178 109 11 CY 1996 1274 109 12 CY 1997 1473 109 14

• Counts do not include Dresden Unit 1; Ilumboldt Bay Unit 3; Three Mile Island Unit 2; Fort St. Vrain after August 29, 1989; Lacrosse after April 30,1987; Rancho Seco after June 7,1089; Shoreham after June 6,1987; Yankee Rowe after February 26,1992; San Onofre Unit I after November 30,1992; and Trojan after January 4,1993. Supplemental, canceled, proprietary, voluntary, and safeguards LERs were excluded from all counts.

Table 5.2 Percentage of LERs Submitted in CY 1997 by 10 CFR 50.73 Requirement Percent of 10 CFR Section Requirement LERs 50.73(a)(2)(i) Technical Specification shutdown or violation 52 50.73(a)(2)(ii) Unanalyzed condition 28 50.73(a)(2)(iv) Engineered safety feature actuation (including reactor trip) 14 50.73(a)(2)(v) Real/ potential safety system loss 10 50.73(a)(2)(vii) Failures in multiple systems 2 t 50.73(a)(2)(iii) External threat <1 50.73(a)(2)(x) Internal threat 0 t

NUREG-1272 Vol. I1, No. I 42

Reactors )

i 5.3 Reliability and Availability Data -

Alert - a condition involving actual or potential substantial degradation of the level of plant On Febmary 12,1995, the Commission published a safety where any offsite radiological releases are proposed mie,10 CFR 50.76," Reporting Reliabil- expected to be limited to small fractions of the ity and Availability Information for Risk-Significant Environmental Protection Agency protective Systems and Equipment"(61 FR 5318). Industry action guideline exposure levels, proposed a voluntary alternative to the mie, and INPO provided a sample of voluntary data from its

• Site Area Emergency - a condition involving

. Safety System Pe formance Indicator (SSPI) actual or likely major failures of one or more database for NRC to evaluate. The purpose of the plant functions required for protection of the evaluation was to determine if the SSPI data, along public or involving conditions with the potential with data from INPO's Equipment Performance and for a significant offsite radiological release but Information Exchange (EPIX) System and other where a core melt situation is not indicated.

data available to the NRC, would be adequate to l .

General Emereeney - a condition involving l estimate the nsk parameters required for PRA and actual or imminent substantial core degradation nsk-informed applications. On May 7,1997, the or melting with potential for loss of containment.

staff reponed to the Commission that, although l there were limitations in the data, methods were in the event of an emergency at a nuclear power l l available to overcome these problems and make plant, the licensee would place an emergency reasonable estimates of the basic PRA parameters. telephone call to the NRC Operations Center, after The staff recommended that the Commission accept notifying appropriate State and local agencies and the voluntarily approach. On June 13,1997, the within one hour of the emergency declaration. For Commission approved the staff's recommendation Site Area Emergencies and higher declarations and and directed the staff to work with industry to for events for which an NRC response may be improve the content of the voluntary data. appropriate, the Regional Administrator and an Executive Team member (typically the Director of In September 1997 NRC and INPO agreed on a the Office of Nuclear Reactor Regulation) will revision to the INPO-NRC Memorandum of discuss the event in a conference call with the Agreement to provide access to and usage of EPIX licensee, data. Late in 1997 INPO sent to the NRC for comment a draft of the proposed guidance for The NRC's response to an event may range from reporting reliability information to EPIX. AEOD routine followup to a complete activation of both communicated to INPO the staff's suggestions to the regional incident Response Center and the NRC i enhance the quality and usefulness of the data. Operations Center. The NRC utilizes the following modes for responding to events: Normal, Standby,

• Unusual Event - a condition involving potential Initial Activation, and Expanded Activation, degradation of the level of plant safety that does not represent an immediate threat to public health and safety.

i

[

43 Operating Experience Data

Reactors l l

l 6 INCIDENT RESPONSE AEOD maintains and implements the NRC's centers with a few select experts to monitor the Incident Response Program with the support of event. The latter is referred to as the Monitoring other headquaners and regional of6ces. This Phase of the Normal Mode.

program includes the receipt of data and reports for .

both emergency and non-emergency events from Standby Mode, the next higher level of response, is licensees, followed by an appropriate NRC re- entered when an event is judged to be sufficiently sponse.The response for the more serious emergen-uncenam r c mplex that the situation needs to be cies is through an incident response organization c ntinu usly m nitored from the headquarters and that includes representatives from several headquar- regi n I resp nse centers by teams of experts. At ters offices and the affected regional office. The this point, the primary responsibility of the affected NRC's response program also includes coordination regi n is t prep re t dispatch a team to the site.

with other Federal agencies as well as State and During the Standby Mode, the NRC response is led local governments, fr m the NRC Operations Center.

If an event threatens public health and safety, the 6.1 NRC Operations Center NRC will enter the Initial Activation Modt and i will promptly send a team from the regional office  :

The NRC Operations Center, located at Two White to the site to lead the NRC response. Until the Site Flint North in Rockville, Maryland, provides the Team is in place, the NRC response will continue to focal point for NRC communications with its be led from the NRC Operations Center. Within the  !

licensees, State agencies, and other Federal agen- NRC Operations Center, teams of specialists will cies about events that occur in the commercial evaluate the status of reactor critical safety func-nuclear sector. It is continuously staffed by a tions and will independently evaluate protective Headquarters Operations Officer who is a nuclear actions recommended by the licensee for imple-systems engineer trained to receive, evaluate, and mentation by State and local authorities. All com-respond to all types of events. The NRC Operations munications with the media, State and Federal Center features a state-of-the-art information of6cials, international organizations, Congress, and j management system that integrates voice, video, the White Ifouse will also be coordinated from the and data subsystems to enhance the timely and NRC Operations Center. During this mode, repre-effective flow of information during the NRC's sentatives from six other Federal agencies report to response to an incident. the NRC Operations Center to lend direct support to the NRC as the Lead Federal Agency.

6.2 Emergency Response Once the NRC site team arrives on the scene and is prepared to accept the operational authority and Commerc.ial nuclear power plants are required t responsibility for the Federal response, the NRC maintam plans for responding to emergencies that enters the Expanded Activation Mode.The could affect the health and safety of the public.

Director of Site Operations, typically the Regional There are four classes of emergencies, in order of Administrator, will report to the licensee's Emer-increasing severity, as follows:

gency Operations Facility near the site. The lead For the Normal Mode, the lowest level of response, responsibility for performing assessments of reactor the NRC will not fully staff the NRC Operations safety, protective measures, and liaison activities Center at headquarters or the incident Response then shifts from headquarters to the NRC team at Center in the applicable regional office. However, the site. The NRC Operations Center will then sorne other action may be taken, such as dispatching furnish logistical and technical support to the NRC a special inspection team or sta'fing the response Site Team as necessary.

45

1997 AEOD Annual Report 6.3 NRC Operations Center Data indication problems while the unit was in Hot Shutdown (Event Number 32271) for 1997

+ Three Mile Island - June 21,1997 - reactor trip in addition to emergency event notifications, the from 1009c' power and loss of offsite power due NRC Operations Center receives many notifications to the failure of a ceramic insulator on a genera-of events that do not meet the threshold for emer-tor output breaker (Event Number 32521) gency classification. Actions taken by the Head-quarters Operations Of6cer in response to such notifications range from entering reports into a 6.4 Emergency Exercises computer and keeping a log, followed by appropri-ate notifications, to establishing emergency confer- Emergency exercises are held periodically to ensure that NRC, licensee, local. State, and other Federal ence calls between licensee representatives and resp nse rganizations are proficient in dealing senior NRC regional and headquaners representa-with each type of emergency. The NRC's primary tives. For very significant events, conference calls r le m these exercises is to independently assess may result in the activation of the agency's Incident licensee actions, assist the licensee when requested, Response Plan.

review the protective action recommendations that Table 6.1 shows the total number of events reported the licensee makes to State and local authorities, to the NRC Operations Center during 1997. These and facilitate communications between the notifications were submitted primarily by nuclear licenseee and other response organizations. Prepa-power plant licensees. A small subset of these ration for these exercises includes the licensee's notifications involved events classified by licensees development of a postulated accident scenario that into ord of the four emergency classes. Table 6.2 normally goes well beyond the f acility's design shows the number of each type of emergency es ent basis and that results in the simulated release of reponed annually from CY 1989 through 1997. The some radioactivity outside the facility's boundary.

number of Unusual Events reported to the NRC NRC responders follow the reactor safety and Operations Center has decreased by 75 percent protective measures aspects of the simulated event; since CY 1989. This can be partially attributed to communicate with the licensee, State, and Federal the fact that many licensees have implemented responders; and make recommendations to an NRC revised procedures for emergency action levels that Executive Team in the NRC Operations Center or to better reflect the severity of events. the NRC Director of Site Operations at the site. In 1997 the NRC headquarters and regional offices Table 6.3 lists the emergency events reported by p rticipated in full-scale emergency exercises with power reactor facilities to the NRC Operations the following nuclear power plants: Farley on Center during 1997 that were classified at the Alert December i1,1996; Turkey Point on May 13, level. (No power reactor events were reported at a 19 Callmy on August 12,1997; and Nine Mile level higher than Alert.) The NRC did not enter the P int n September 24,1997. The Turkey Point Monitoring Phase of Normal Mode for either of these two Alerts. However, the NRC did enter the exercise inv lved ingestion phase activity (dis-cussed m more detail below). The Callaway Monitoring Phase of Normal Mode for the follow-exercise inv Ived a large Federal component during ing three Unusual Events:

the ingestion phase, and the Nine Mile Pomt

+ Pilgrim Unit 1 - April 1,1997 - loss of all offsite exercise involved activation of the NRC News power due to an offsite fault and continued Center at headquarters.

inclement weather in the vicinity of the site . .

Lima d participation exercises are conducted as while the unit was in Refueling (Event Number pan of the State Outreach Program. One objective 32059) of this program is to participate more frequently m

+

Oconee Unit 3 - May 3,1997 - two of three high exeteises with State organizations. During 1997 pressure injection pumps declared inoperable as limited participation exercises were conducted with a result of letdown storage tank water level the following nuclear power plants: Diablo Canyon NUREG 1272,Vol. I1, No.1 46

Reactors Table 6.1 Events Reported to the NRC Operations Center in 1997 Non- Well Logging / ,

Emergency Power Fuel Power Transport /

Class Reactor Facility Reactor Hospital Materials Other Total Non-Emergency 1,458 129 2 54 140 102 1,885 Unusual Event 49 0 0 0 0 0 49 Alert 2 2 0 0 0 0 4 Site Area Emergency 0 0 0 0 0 0 0 General i Emergency 0 0 0 0 0 0 0 Total 1,509 131 2 54 140 102 1,938 Table 6.2 Classification of Events Under Licensee Emergency Plans CY 1989 to 1997 i Emergency Class CY89 CY90 CY91 CY92 CY93 CY94 CY95 FY96 FY97 Unusual Event 197 151 170 135 103 97 66 67 49 Alen 13 10 9 20 8 4 8 10 4 Site Area Emergency 0 1 2 1 1 0 0 1 0 General Emergency 0 0 0 0 0 0 0 0 0 Total 210 162 181 156 112 101 74 78 53 on October 9,1996; Grand Gulf on September 17, tion phase exercises. A small team of experts from 1997; and Kewaunee on September 23,1997. AEOD and the regional offices (the Ingestion Team, r Ue m) p n p ted in ingestbn phase exerches Site-team-only exercises are also conducted with at the following nuclear power plants during 1997:

Headquarters Operations Officer participation.

Anificial Island (Salem / Hope Creek) on October During 1997 site-team-only exercises were con- j99 pg; 3997 ;

ducted at Crystal River on October 16,1996, and c njunction with a full-scale exercise); and Surry on August 26,1997.

Callaway on August 13,1997 (in conjunction with a In addition to the full-scale, limited, and site-team- full-scale exercise). Panicipation in ITeam exercises only emergency exercises. AEOD conducts inges- gives the NRC and other Federal agencies an 47 Incident Response

1997 AEOD Annual Report Table 6.3 Alerts Reported by Power Reactor Facilities in 1997 Plant Name Event Event Event (Vendor / Type) No. Date Description of Event Duration Waterford 3 31619 01/18/97 Toxic chemical (pyrolysis gasoline) spill =10 miles 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and (CE/PWR) north of the site on the Mississippi River (NOTE: 53 minutes The Alert was downgraded to an Unusual Event after i l'our and 44 minutes.)

Waterford 3 31979 03/20/97 Toxic gaseous ammonia release from a chemical 39 minutes (CE/PWR) plant located less than 1 mile from the site Key: CE- Combustion Engineering l PWR - Pressurized Water Reactor V

opportunity to assist State and local organizations South Carolina, and Georgia; m NRC Region 3 for during latter-phase response actions. Immediately Michigan, Indiana, Wisconsin. and Ohio; and in preceding these exercises, AEOD also conducts NRC Region 4 for Califomia, Washington, Oregon, ingestion phase training for NRC, licensee, Federal, Arizona, and Missouri. Participants included both State, and local organizations. radiation health and emergency services organiza-tions within the States, the licensees' counties, and Figures 6.1 through 6.4 show participants m a the regional Federal organizations.

typical exercise as they receive and evaluate the emergency situation, facility status, and licensee ERDS training was also provided for all States actions to determine the appropriate NRC response, within 10 miles of a nuclear power plant. Each including the appropriate guidance for State and NRC regional office coordinated the training of its local governments. respective States. The NRC conducted the training and helped pay for the States to attend.

6.5 State Outreach 6.6 Coordination With Other In 1997 AEOD continued an aggressive State Outreach Program designed to mcrease and im-Federal Agencies prove the NRC's interaction with States during In 1997 AEOD continued to participate with other events and exercises. It included briefings of State Federal agencies in the completion of annexes to officials on the NRC and Federal emergency the Federal Response Plan (FRP), such as the response program, the Emergency Response Data Radiological Incident Annex, which describes how System (ERDS), NRC/ State liaison during an the FRP and the Federal Radiological Emergency emergency, and financial assistance available to Response Plan (FRERP) are integrated. Further-responders. AEOD also expanded the program to more, significant interactions took place in the area include training on NUREG/BR-0230, " Response of preparedness for nuclear, biological, and chemi-Coordination Manual (RCM-96)" and the recently cal terrorist events. As a result of the NRC regula-updated NUREG/BR-0150," Response Technical tory oversight of the gaseous diffusion plants, Manual (RTM-96)". Outreach sessions were AEOD staff worked with other Federal agencies conducted for responders from 22 States and numer- through the Environmental Protection Agency's ous licensees. RCM/RTM Outreach sessions were (EPA's) National Response Team to ensure that conducted in NRC Region I for Maine, New Hamp- proper coordination takes place among Federal shire, Vermont, Massachusetts, Connecticut, New agencies with statutory responsibility to respond to York, New Jersey, Pennsylvania, Maryland, and chemical emergencies. As a result of these interac-Delaware; in NRC Region 2 for North Carolina, tions, the EPA and National Oceanic and Atmo-NUREG-1272, Vol. I 1. No. I 48

Reactors i

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49 Incident Response

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i j NUREG-1272 Vol. I1 No.1 50 4

Reactors 1

spheric Administration have been more thoroughly the protective action guideline subcommittee; i integrated into the NRC response procedures. AEOD participated in coordination of related l Additionally, AEOD began to develop the assess- activities with other Federal agencies. As a member i ment tools needed for handling emergencies that of the Emergency Support Function Leaders Group, involve chemical as well as radioactive material AEOD participated with other Federal agencies in releases. In cooperation with other Federal agen- discussing issues relating to Federal coordination I cies, the role of the principal agencies and their and response.

associated responsibilities in responding to chemi-cal as well as to radiological emergencies at fuel AEOD provided training to o:her Federal and State cycle facilities, such as the EPA and Department of representatives on public information, severe Energy, were better defined. weather response, and emergency dose projection using the RASCAL program.These training In 1997 AEOD supported the Federal Emergency sessions were provided in-house as well as at the i Management Agency (FEMA)in providing acci- Harvard School of Public Health, dent assessment training to Federal, State, and local officials on four occasions. In addition, AEOD AEOD also presented the NRC concept of opera-participated in the design and conduct of an FRERP tions at various forums, such as the American workshop sponsored by FEMA. As a member of the Nuclear Society conference, the Conference of Federal Radiological Preparedness Coordinating Radiation Control Program Directors' meeting, the Committee AEOD participated in meetings to National Emergency Management Agencies discuss issues relating to State and local govem- conference, Nuclear Energy Institute meetings, ment response. Furthermore, as a member of the and the National Radiological Emergency Pre-training subcommittee, exercise subcommittee, and paredness conference.

1 1

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51 Incident Response

Reactors 7 INCIDENT INVESTIGATION PROGRAM The Incident Investigation Program (IIP) ensures identify needed actions from AIT Gndings. AEOD that NRC investigations of significant events are independently reviews AIT reports to gain addi-timely, thorough, well coordinated, and formally tional assurance that potential generic lessons are administered.The scope of the IIP includes investi- learned and communicated to the industry. For gations of significant operational events involving reacter events, the Director of the Office of Nuclear reactor and materials activities licensed by the Reactor Regulation (NRR) is responsible for NRC. Under the llP, the NRC responds to an reviewing AIT reports for generic safety implica-operational event according to its safety signifi- tions, initiating follow-up actions, and tracking cance. For an event of extraordinary safety signifi- issues affecting more than one plant, as appropriate.

cance, the Commission may establish an Accident In this way, indus:n-wide safety is enhanced by Review Group (ARG) led by an individual from including the signi6 cant lessons learned from AITs outside the NRC and comprising experts from with those from engineering studies and reviews of within and outside the NRC.The ARG reports operating experience in generic communications to directly to the Commission and is independent of licensees. As described in NUREG-1303," Incident NRC management. For an event of potentially Investigation Manual," AEOD has overall responsi-major safety significance, the Executive Director bility for administration of the IIP, and NRR is for Operations (EDO) establishes an Incident responsible for maintaining the procedures for an Investigation Team (llT) to investigate the event. AIT response.

For an event of less safety significance, the respon-sible NRC regional administrator may establish an 7.1 Incident Investigation Teams Augmented inspection Team ( AIT) to investigate the event. Both IITs and AITs are assigned to There were no power reactor events in 1997 that determine the circumstances and causes of an were judged to have a level of significance suf5-operational event and to assess the safety signifi- ciently high to warrant an IIT investigation. The cance of the event so that appropriate follow-up status of actions associated with previous IIT actions can be taken.The EDO assigns staff actions findings assigned by the EDO to various NRC arising from IITs, and the regional administrators of6ces is documented in Appendix G.

l Table 7.1 Reactor Incidents for Which AITs Were Established in 1997 Event Date Plant Event 02/21/97 Zion i Improper control rod manipulation 05/03/97 Oconee 3 Degraded high pressure injection pumps 06/20/97 Oconee 1 Failure of Keowee Unit 1 08/05/97 Clinton 4160- volt circuit breaker maintenance issues l

l l

53

(

1997 AEOD Annual Report 1

7.2 Augmented Inspection Teams e a total breakdown in command and control by l Operations supervision Four AITs were established in 1997 to investigate e

significant incidents at power reactor facilities, as inadequate communications between operators, shown in Table 7.1. These AITs helped to improve Operations supervision, Operations manage-safety at the affected plants by providing detailed ment, and Nuclear Engineering Department investigations of the problems experienced and personnel identifying their root causes. .

the failure of Operations supervision, Operations management, and plant management to provide 7.2.1 Zion I clear direction to the operating crew regarding On February 21,1997, at 12:09 p.m., control room the planned shutdown operators began reducing power in preparation for a ,

the failure to pre-plan the shutdown evolution plant shutdown due to an moperable containment and licensed operator training deficiencies spray (CS) pump. At 25 % power, operators entered General Operating Procedure (GOP) 4, " Plant

• a number of control room distractions during Shutdown and Cooldown." At about 2:05 p.m., with shift activities Unit I at 7 7c power, the shift engineer (SE) di-rected the unit supervisor (US) to maintain the The actions of the primary NSO in continuously reactor critical since it was expected that the CS withdrawing control rods to re-establish power at the pump would be returned to service withm. the next POAH reflected a sigmficant lack of understanding few minutes. The US and the primary nuclear f reactor physics and proper control rod manipula-station operator (NSO) reviewed the steps in GOP 4 t ns f r a c ntrolled approach to criticality. The fer taking the turbine otT line and inserting contro actions of the primary NSO were also contrary to rods to establish power at or below the point of instmcti ns in the plant startup procedure.

adding heat (POAH), defined as 0.025 % power. A Regarding command and control, the following few minutes later the main turbine was tripped. The deficiencies were noted:

US then read step 5.21.f aloud to the primary NSO and, after acknowledging this directive, the primary The shutdown briefing was informal, poorly NSO inserted control rods continuously for 3 planned, and ineffective. Operations supervision minutes and 48 seconds until power indicated 0.025 did not give any direction to the operating crew 7c. About one minute later, with the reactor substan, during the briefing regarding the decision point tially suberitical and power continuing to decrease for proceeding to hot shutdown, because of the large amount of negative reactivity .

Despite a number of control room indications added by the control rods, the primary NSO in-and communications, Operations supervision formed the US that he intended to withdraw control was unaware that the primary NSO had continu-rods to stabilize power at 0.025 7c. The primary ously inserted control rods a total of 232 steps, NSO then proceeded to withdraw control rods which placed the reactor in a substantially continuously for I minute and 45 seconds until the suberitical condition, and then withdrew control US directed him to trip the reactor. The SE in-rods 84 steps in an attempt to re-establish power structed the US to trip the reactor because the CS at the POAH.

pump had not been restored and Unit I had to be in hot shutdown within the next 6 minutes in order to .

Operations supervision failed in its responsibil-comply with the Technical Specifications. At 2:15 ity to minimize control room distractions, which p.m., the Unit I reactor was manually tripped, had the potential to adversely impact the ability placing the unit in hot shutdown. of operators to safely conduct plant evolutions.

During its review of the February 21 shutdown Operations management did not appreciate the event, the team developed a number of findings and significance of the actions of the primary NSO in conclusions. The more significant root causes for continuously withdrawing control rods in an this event included the following: attempt to take the reactor critical. This was evident NUREG-l 272, Vol. I 1. No.1 54

l 1

l Reactors >

in the deliberate decision by Operations manage- were management inefTectiveness in organizational l ment to return the involved licensed operators to response to regulatory challenges (unawareness licensed duties, errors), inadequate potential hazards analysis, in dequate performance monitoring and trending, The AIT identified a number of precursor events less an uh p qu te root-cause and common-with root causes related to poor communications, weak command and control, and poor reactivity caus anayses, ina&quate managanent expeta-tions, and madequate networking (complacency).

management. Because of the absence of corrective u t t the licensee's response to the actions in some cases, and because of ineffective event after pump degradation had occurred was corrective actions in other cases, the licensee failed c nsen e emagency plan was used to help l to correct the underlying problems which contrib-m nage the HPI system recovery efforts in order to uted to these events.

continue with the cooldown.

1 7.2.2 Oconee 3 7.2.3 Oconee 1 l

I On May 3,1997, Oconee Unit 3 was being cooled The NRC formed a special AIT, which arrived a down to conduct an inspection of high-pressure e nee t I on June 30 to review th companfs injection (HPI) piping at its interface with the

" ".s in resp nse t the failure of an emergency reactor nozzle. During the cooldown, operators did ectnca supply (Keowee Unit 1, a hydro unit) to not properly perform the fundamental activities of pr perly resp nd during a loss of offsite power test monitoring ami balancing reactor coolant inventory; n June 20, and the failure of electrical breakers therefore, they did not question non-responsive during testing restoration on June 23.

indications ofletdown storage tank level. The loss of inventory in the letdown storage tank caused The AIT reviewed the circumstances surrounding degradation of two of the three HPI pumps. the failure of Keowee Unit 1 to achieve rated v tage following a loss of the Lee Steam Station The AIT conducted an independent inspection of dedicated electrical power path, the failure of air the event and of the circumstances surrounding it, circuit breakers ( ACBs) 5 and 7 during testing, and monitored the licensee's investigations into the the licensee's response and investigation of the incident, and reviewed other related data. The AIT events and the recovery. In addition, the AIT concluded that the proximate causes of the event assessed generic aspects of Oconee and Keowee l were (1) a lack of redundancy in the design of the perations and inspections to evaluate the applica-letdown storage tank level instmmentation system bility of the events to the other units.

and (2) a loss of reference leg inventory resulting in non-conservative and non-responsive level indica- The AIT concluded that the licensee's investiga-tion on both letdown storage tank level instruments. tion of the problems was effective.The AIT The result was that Oconee was operated outside its reached the same conclusions as the licensee's design basis during the period that the level instru- failure investigation.

mentation error caused the letdown storage tank level to be lower than allowed. With a low letdown The root cause of the failure of the field flash storage tank level, the ability of the HPI system to breaker for the emergency generator was most perform its safety function of mitigating a design likely a random failure in a fuse or breaker. There basis accident was not assured. was a recent modification (setpoint change) that may have increased the probability of the failure The AIT determined that the root cause of the event leading to the defeat of the generator start.

was inappropriate use of plant and industry operating experience to ensure that plant design, maintenance, The investigation of the two ACBs concluded that and operation were focused on reliable operation of the root cause of their failure was the effect of an the HPl system and components involved in this encapsulating epoxy expansion on the timer poten-event.The AIT determined that some common error tiometer. The AIT concluded that the licensee's categories that may have contributed to this event choice of timer and timer setpoint may not be t

55 Incident Investigation Program

1997 AEOD Annual Report appropriate for their current use, even though the 1997, IP informed the NRC that in addition to manufacturer verified the accuracy of the timer and identifying this error, a method to identify the the in-use setting. problem that led to the safety-related breaker failure of July 22 had been found, and that all necessary The AIT review of the licensee's maintenance corrective actions had been taken for the other procedure for the ACBs found that the procedure safety related Westinghouse 4160-volt breakers. .

was generally in agreement with the Westmghouse '

, During this meeting IP informed the NRC that Technical Manual, except that one sigmficant '

reasonable assurance existed that safety-related specific preventive maintenance step was omitted. '

breakers would function when called upon.

This omission arose from a deficiency in the process for translating infernation from the techni- At approximately 4 a.m. on August 5,1997, another cal manual to Oconee procedures. safety-related 4,160-volt,1200-amp Westinghouse circuit breaker failed to open on demand during the The failure of Operations personnel to follow a Lee planned change of pumps in the residual heat Steam Station operating procedure caused the loss '

removal (RHR) system.The licensee was shifting of the Oconee main feeder fuses on June 20,1997.

from the *"A" RHR pump to the "B" RHR pump, Further, the AIT noted several procedural deficien-and the "A" pump circuit breaker should have cies as well as training and knowledge deficiencies opened to interrupt power to the pump. The RHR on the part of the Lee staff with regard to suppon of system provides reactor cooling while the plant is activities required by technical specifications. '

shut down. This breaker had been subjected to the The AIT reviewed the basis of the reponability inspections, tests, and corrective actions discussed decision for the events of June 20 and 23, under- above. As a result of this failure and earlier circuit stood the basis, and agreed with all decisions except breaker failures, the licensee declared all safety-  !

for that regarding the emergency start of the related 4,160-volt Westinghouse circuit breakers Keowee units on June 20, for which a voluntary inoperable. The licensee had planned to begin notification was submitted. starting up the unit, which has been shut down since  ;

September 5,1996.

7.2.4 Clinton The NRC formed a special AIT that conducted an i

In February 1997 NRC inspectors identified defi- inspection at Clinton Power Station (CPS) from  !

ciencies in Illinois Power's (IP's) maintenance August 5 through August 15 to review the circuit program for safety-related electrical circuit break- breaker failure, the company's actions in response ers. Following this finding, IP performed inspec- to the failure, including inspection plans and tions and tests in an effort to establish reasonable methods, as well as the root cause determination for assurance that safety-related breakers will work the July 22 and August 5 failures.

when called upon. During a public meeting on July 3,1997, IP reported the completion of sufficient On the bas.is of this inspection, the AIT concluded inspections and tests for the establishment of such that both circuit breaker failures were caused by reasonable assurance. inadequate and inappropriate maintenance activi-  ;

ties, and by deficiencies in CPS' corrective action '

On July 22,1997, a safety-related Westinghouse system. CPS' preventive maintenance program did [

4160-volt,1200-amp breaker failed to open when not lubricate all vendor-recommended areas of the called upon in service. This breaker had been RHR Pump A breaker, most notably the main and subjected to the inspections and tests discussed arcing contacts. In addition, CPS occasionally used  !

above. On the same day, a non-safety-related unapproved cleaning agents, which had the poten-Westinghouse 4160-volt,1200-amp breaker failed tial to remove vendor-applied lubricants, without ,

to close when called upon. As a result of the non- relubricating the areas. Funher, CPS had not safety related breaker failure, IP identified an error

(

effectively evaluated the findings from NRC {

in the applicable maintenance procedure, which Inspection Repon 50-461/97003 regarding breaker [

resulted in mis-adjustment of a critical internal maintenance, had not developed effective immedi-mechanism. During a public meeting on July 31, ate actions to address known maintenance deficien-NUREG-1272. Vol. I 1. No.1 56

Reactors cies, and had not effectively evaluated and identi- the July 22,1997, breaker failure. Although CPS' fled a root cause for the July 22,1997, circuit investigation of the August failure was much breaker failure.The combined deficiencies resulted improved NRC prompting was necessary at times in a potential common mode failure being intro- to ensure the investigation was thorough. For duced into at least all 4.16-kV circuit breakers at example, CPS' initial root cause determination CPS manufactured by Westinghouse Corporation. concentrated on a lack of lubrication and did not initially consider that springs used to open the The AIT concluded that CPS' corrective actions breaker could have contributed to the failure to addressing previously identified breaker preven- open. CPS' review confirmed that although the tive maintenance deficiencies were neither thor- major contributor to the failure was lack of lubri-ough nor effective.Thus the AIT investigation into cant, a bent and short kick-out spring played a the August 5,1997, breaker failure was signifi- considerable role in the breaker's failing to open.

cantly more rigorous than the investigation into i

57 Incident investigation Program

Reactors 8 DIAGNOSTIC EVALUATION PROGRAM The Diagnostic Evaluation Program (DEP) safety-related hardware and performance problems, provides for an independent assessment of licensee and obtain additional information on safety perfor-performance at selected reactor facilities. A mance to allow NRC senior managers to make an diagnostic evaluation assessment augments informed assessment of plant safety perfonnance.

information provided by the Systematic Assess- The evaluation included an assessment of the ment of Licensee Performance (SALP) program, licensee's IS A and, as appropriate, built on any the Performance Indicator (PI) program, and the independently validated findings from the licensee's NRC's inspection program implemented by the effort to arrive at an overall evaluation of the NRC's headquarters and regional offices. performance of the licensee and of the facility.The

. SET, which was led by an NRC manager, was When a diagnostic evaluation (DE) is approved for

, approximately half the size of a DET and evaluated a specific reactor fr.cility, the Executive Director the same functional areas as a DET; in combination for Operations (EDO) authorizes and establishes a with the ISA, the SET provided for a diagnostic-Diagnostic Evaluation Team (DET). The DET level evaluation of the facility.

comprises experienced evaluators supplemented by expert technical staff members from headquar- The major findings of the Clinton special evaluation ters and the regional offices (such as experienced are summarized below.The status of actions resident inspectors), as well as contractors, if involving potential generic issues associated with appropriate. The selected DET manager and team previous DET or SET findings assigned by the members will not have had recent significant EDO to various NRC offices are documented in involvement in the licensing, inspection, or Appendix H.

enforcement process at the selected facility. The onsite evaluation process involves observations of .

plant and corporate activities, in-depth technical 8.2 Special Evaluation of Clinten reviews, employee interviews, equipment walk- Power Station down inspections, and programmatic reviews in a number of functional areas important to safety The Clinton Power Station (CPS) is located on performance. Areas evaluated generally include b".ke Clinton,6 miles east of the city of Clinton, maintenance, surveillance and testing, manage-Illm is.The plant consists of one General Electric ment effectiveness, operations, engineerirg, and (GE) Type 6 boiling-water reactor with a Mark 3 quality programs. c ntainment. The plant's rated electrical output is 930 megawatts. The facility was designed by Sargent and Lundy and was constructed by Baldwin 8.1 Diagnostic Evaluation Teams Associates.

In 1997, as an alternative to an NRC diagnostic Following the January 1997 NRC Senior Manage-evaluation, the EDO approved the conduct of a ment Meeting, the EDO directed the staff to per-licensee-initiated independent self assessment form an evaluation to obtain information needed to (ISA) of the Clinton Power Station. In association make an informed decision on overall performance with this approval, the EDO directed the Office of at CPS and to determine the root causes of identi-Nuclear Reactor Regulation (NRR) to form a tied problems. The Illinois Power (IP) Company, Special Evaluation Team (SET). The goals of the recognizing that significant deficiencies existed in SET were to evaluate the effectiveness of the the overall performance of CPS. committed in a licensee's IS A effort, evaluate the actions of letter dated June 30,1997, to perform an Integrated licensee management and staff with respect to safe Safety Assessment (ISA) in order to prepare for a plant operation, determine the root cause(s) of the diagnostic assessment of CPS performance. This 59

l 1997 AEOD Annual Report  !

ISA was an independent assessment similar in ISA. Specifically, the SET made the following scope and depth to an NRC Diagnostic Evaluation. findings:

Subsequently, the EDO instructed the staff to form a Special Evaluation Team (SET) to assess both the f an gement generally did not establish and implement effective perf rmance st ndards.

efficacy of the licensee's ISA through direct obser-vation and independent assessment and the overall The SET concluded that the failure of IP and performance of CPS. CPS management to establish and implement From August 25 through December 11,1997,an ety ymmana standae was a mot caum f the sigmficant decim, e m safety performance.

SET from the Nuclear Regulatory Commission Management failed to establish and communi-(NRC) cvaluated the performance of IP in ensuring cate appropriate, clearly defined expectations the safe operation of CPS. The SET consisted of a and priorities, and failed to monitor their imple-team manager,10 NRC team members, a manage-ment an rganiz tion contractor, and an admmis- mentation for the desired performance. Manage-ment decisions that were inconsistent with stated trative assistant. In addition, the Illmois Department expectati ns c ntributed to decl. .mmg perfor-of Nuclear Safety (IDNS) sent an observer to the SET, who facilitated communication between the manc . In addition, management did not give the staff sufficient feedback and failed to establish team and the State. The team was organized into the accountability.

areas of Operations, Engineering, Maintenance, Plant Support, and Management and Organization. 2. CPS programs, processes, and procedures did The CPS ISA was performed by a well-qualified n t c nshendy ppe deknw in dep6 to team of industry experts. The ISA team conducted "" *"

• E'""' "#U "" * * *# " # " "'I# ' " "

an effective diagnostic assessment of the station's performance and reached substantive conclusions. The SET concluded that programs, processes, which the licensee recognized as generally valid. and procedures failed to integrate activities The root causes identified by the ISA were con- across departments, incorporate industry firmed by the SET's independent evaluation and information, and clearly delineate ownership were consistent with the SET's determination of and accountability. Program implementation ,

root causes, was not effective in attaining the intended

1. ' #8' "'# 8 * * "" I The ISA used a barrier analysis technique to anive at root causes and focused its root cause effort on a me y cum s me a , E*#Y""****.&

a ng to pmn recent and relatively short time frame. in contrast. "EE" E" "' #' " " " #" * * #' "#"

the SET's observations, validation, and causal w rkers performing en activity. Programs and processes did not provide effective momtormg analysis approach was more historically based and and feedback.

focused on a longer performance period than considered by the ISA. Although the time frames 3. Problem identification was inconsistent and considered were different in length, the SET evaluation and corrective actions were generally confirmed the ISA's process and found its root ineffective.

causes to be consistent with those of the SET. . .

The SET concluded that the m.abih.ty to identify.

Further, the SET found some additional examples of problems observed by the IS A in the Operations, gvalu te, and correct problems was a major ,

impediment to improvement. Inconsistencies in Engineering, and Plam Support functional areas.

problem identification resulted in failure to Potenual safety and compliance issues that were identified during the ISA and SET assessments ensum tp t pmWms wem egco.vely captumd.

Ineffective evaluation of identified problems were subm.itted to Region Ill for appropriate contributed to failure to develop effective regulatory follow-up.

ive actions. Failure to monitor and ensure The conclusions of the SET were similar to and implementation of corrective action plans consistent with the root causes identified by the contributed to recurring problems and an attitude of living with problems.  ;

NUREG-1272. Vol. I 1. No.1 60 i

l

l Reactors t 1 i

l j 4. Management did not ensure that the infrastruc- without appropriately considering the longer term "are was suitable to support major changes. and integrated effects of the decisions. Management i did not ensure that there were appropriately quali.

The SET concluded that management did not

. fled staff, integrated programs and processes, and i recognize that the infrastructure at CPS was insuffi- . i appropriate resources to support implementation of a l cient to support major changes. As a result, man-the reengineering and downsizing effort.

agement made organizational, programmatic, and resource decisions in the context of reengineering l

l i

i I

l i

l i

I l

61 Diagnostic Evaluaion

Reactors l

l l

l 9 COMMITTEE TO REVIEW GENERIC REQUIREMENTS The Committee to Review Generic Requirements the EDO transmitted to the Commission SECY (CRGR) reviews all generic requirements proposed 109, proposing to reduce the basic scope of CRGR by the NRC staff that involve one or more classes of review to include only "high impact" and "contro-power reactors.The CRGR comprises senior versial" generic correspondence and rules before managers from various headquarters program public comment, issues which the staff has diffi-offices and, on a rotational basis, from one of the culty resolving after public comment, emergency NRC regional of6ces. The AEOD Director serves and urgent generic correspondence, and significant as the CRGR Chairman, and the AEOD staff proposals with highly expedited schedules. A June provides support for all of the Committee's activi- 15,1994, staff requirements memorandum (SRM) ties. The AEOD Director also oversees directed the staff to maintian the scope of the plant-specific backfit activities of the NRC staffin CRGR Charter and to consider and to recommend a the headquarters program ofHees and the regional course of action for enlarging the scope of CRGR offices. In this period, one new member from a review to include proposed generic requirements in region was appointed to the CRGR. The member- the nuclear materials area. The SRM also directed ship of the CRGR in 1997 was as follows: the staff to look at measures that would lessen the time spent on CRGR reviews by individual CRGR Denwood E Ross, Director, AEOD (Chairman) members. The Committee evaluated this option and Frank J. Miraglia, Deputy Director, NRR agreed to address, on a 1-year trial basis, selected nuclear m teri is issues identified by the NMSS Malcolm R. Knapp, Deputy Director, NMSS Director or by the EDO. The Committee will assess Joseph A. Murphy, Director, Division of Regula- whether or not the nuclear materials issues that are tory Applications, RES presented by the staff for CRGR review warrant CRGR attention and, if they do, whether the CRGR Dennis C. Dambly, Assistant General Counsel for Materials, review adds sigmficant value. On the basis of that assessment, the Committee will make appropriate Antitrust and Special Proceedings, OGC recommendations to the EDO regarding continua- l tion of the CRGR review of nuclear materials James E. Dyer, Deputy Regional Administrator, issues. This assessment will be included in the Region IV CRGR meeting minutes during the trial period, and While performing the CRGR review function, a it will also be reported to the EDO in the CRGR j CRGR member expresses an individual profes- Weekly items of Interest to be reported to the  !

sional opinion about each item considered, rather Commission.This aspect of the expanded scope of than representing the view of his or her respective CRGR review was included in the ongoing CRGR office. The members of the CRGR determine Charter revision process.

whether proposed new generic requirements hav On February 9,1996, in SECY-96-032, the EDO suf6cient merit in terms of safety and are justi6ed requested Commission approval for this I year trial in terms of cost (where appropriate) before reaching y gram ' include selected nuclear materials a consensus recommendation about each issue issuo.1he Commission was also mformed that the considered. Each independent CRGR recommenda-CRGR has considered and adopted measures to tion is given to the EDO for consideration.

lessen the time spent by members on CRGR In 1994 a staff proposal was submitted to the reviews. When appropriate, based on lack of Commission to reduce the scope of the CRGR controversy, low expected impact, or small potential review and to evaluate various means of reducing for error related to the proposed generic actions, the l the burden on CRGR members. On April 21,1994 CRGR Chairman may agree to one of three courses 63

1997 AEOD Annual Report of action:(l} defer the CRGR's review pending purpose of the CRGR annual visits to various public comment on the proposal; or (2) agree to a facilities is to have a candid dialogue with licensees negative r onsent approach which, in essence, is an to get their feedback on the NRC's regulatory abbrevia.ed review; or (3) forgo a second CRGR backfitting xtivities.There were no meetings this review. thus reducing the number of dual reviews fiscal year of either CRGR members or staff with (i.e., review at both the proposed and final stage). the Nuclear Utilities Backfitting and Reform Group.

All other staff proposals will be scheduled for regular CRGR review.

9.1 CRGR Issues On March 22,1996, the Commission approved i Revision 6 to the CRGR Charter, which expanded In 1997 the CRGR held 18 meeting during which it reviewed 29 issues, all but one ulated to nuclear the scope of CRGR reviews to include, on a 1-year trial basis, selected nuclear materials issues re- p wer pl nts only; one item concerned both nuclear as well as materials facilities. In addition, there quested by the NMSS Director or the EDO. The Staff Requirements Memorandum,"SECY-97-052 - were seven briefings to the CRGR by the staff.

Committee to Review Generic Requirements .

Proposed SER on WCAP-144416-P (credit for (CRGR)- Scope of Resiew and Periodic Review soluble boron issue)

Activities," dated April 18,1997, asked the CRGR h> continue the review of nuclear materials issues Briefing on Source Term Options (Commission for another year, and to also review inspection p per) guidance at the staff's request or at the Committee's .

Proposed steam generator integrity rule self-initiative. Also in response to this SRM, a report containing an annual evaluation of the a Briefing on risk-informed regulatory guides and CRGR's activities and its contributions in achieving the associated Standard Review Plans the agency's mission (COMSECY-96-028) was .

Proposed general Regulatory Guide - a guidance submitted to the Commission in August 1997.This document for risk-informed regulation report contained an assessment of the value added by the CRGR review of various generic staff a Proposed general Standard Review Plan proposals, which included the Committee's own ,

Briefing on proposed mie on shutdown and low-self-assessment, assessments by the program offices power operations and spent fuel pools sponsoring the proposals, and assessments by some of the cognizant staff who had the primary respon-

• Briefing on historic perspectives ofin-service sibility for the proposals reviewed by the Commit- inspections tee. Also included in this report were the highlights -

Proposed generic letter on effectiveness of of regional backfit training and audits done by the ultrasonic testing systems in in-service inspec-CRGR staff, the feedback received from licensees tion programs during the CRGR annual site visit to a nuclear l power plant and to a nuclear materials facility, and a Proposed generic letter on steam generator tube l direct feedback received by the Committee from inspection techniques industry-supported organizations. Another report .

Proposed generic letter on degradatmn of steam containing proposals for i.uture CRGR review of generator intemals j nuclear materials issues selected as part of the 1 l

year continuation of the trial program was also +

General regulatory guide (excluding the Appen-submitted to the Commission. dices) and the associated Standard Review Plan

" # '#E" # "

During this fiscal year the CRGR members and l staff also visited the Callaway nuclear power plant -

Proposed application-specific regulatory guides in Columbia, Missouri, as well as the Gaseous and Standard Review Plans - guidance docu-Diffusion Plant in Paducah, Kentucky, and the ments for risk-informed regulation Allied Signal Chemicals plant in Illinois. The NUREG-1272, Vol. I1, No. I 64

Reactors In-service Testing

• Proposed generic letter on problems with

. medium-voltage circuit breakers Technical Spec.fications i

Draft standard review plan Chapter 7,"Instru-Graded Quality Assurance mentation and Controls," Update

• Urgent generic letter on assurance of sufficient ,

Safety Evaluation Report on EPRI Topical net positive suction head for emergency core Report," Guideline on Evaluation and Accep-cooling and containment heat removal pumps tance of Commercial Grade Digital Equipment

• Generic letter on modifica: ion of the NRC staff's for Nuclear Safety Applications," EPRI-TR-recommendations for the post-accident sampling 106439 system

• S.ix proposed regulatory guides (l.168-1.173) for

• Generic letter on degradation of control rod computer software to be used in safety systems drive mechanism and other vessel head penetra. of nuclear power plants tions

• Briefing on operational experience related to e Proposed Revision 2 to Regulatory Guide 1.160, primary system leaks which endorses Revision 2 to NUM ARC 93-01, ,

Revised application-specific risk-informed

" Industry Guideline For Mom,toring The Effec- guidance document - draft regulatory guide and tiveness Of Maintenance At Nuclear Power standard review plan on in-service inspections Plants"(April 1996 version), and provides certain clarifications

• Briefing by the NMSS staff identifying nuclear

• Proposed generic letter " Potential for Degrada-a ci Fmergency Core Cooling System Recir-Proposed final Revision 3 of Regulatory Guide

ulation due to Construction Deficiencies and 5.44," Perimeter Intrusion Alarm Systems" Foreign M.>terial in the Containment Following

• Proposed generic letter on steam generator tube Loss-of-Coobnt Accident" nspection techniques

• Revised general regulatory guide and the Stan-

• Proposed revision to Generic Letter 91-18, dard Review Plan for risk-informed regulation "Information to Licensees Itegarding NRC

• Revised application-specific (in-service testing, laspection Manual Section on Resolution of technical specifications, and graded quality Degraded and Nonconforming Conditions," and assmance) regulatory guides and the accompa. brbfing on the scope and schedule of related nying Standard Review Plans for risk-informed eff0"S regulation. .

Proposed amendments to 10 CFR 50.55(a),

a Revised proposed generic letter dealing with " Codes and Standards" degradation of the emergency core cooling .

Proposed generic letter on laboratory testing of system and the coatamment spray system due to the nuclear-grade activated charcoal foreign material r ' antainment and con-struction deficiencie, 9.2 Value Added by the CRGR

• Supplement I to Bulletin 96-01," Control Rod ggyjgg insertion Problems."

The CRGR agrees with the general office percep-

• Briefing on 10 CFR 50.67 " Shutdown Opera-ti n th t in ddition to preventing improper tions and Operations Involving Fuel Storage ,

baMts, a nWor role that the CRGR plays is m Pools at Nuclear Power Plants" affording them an mdependent review of various

• Review of Proposed 10 CFR 50.67. " Shutdown proposed actions. The Committee considered value operations and operations imotving fuel storage added by CRGR review and identified various pooh at nuclear power plants" topics reviewed during this assessment period 35 CRGR

1997 AEOD Annual Report where significant value was added to the incoming well-coordinated effort in developing the exten-staff proposals by the CRGR review. Some of the sive guidance for the industry and the staff on a staff proposals required review at more than one complex subject. In the subsequent rounds of the CRGR meeting, involved extensive staff effort to CRGR review, the Committee made only minor re-write the proposals, and involved staff effort to comments.

address the Committee's comments and recommen- I dations. The value added by the CRGR interaction In reviewing the first version of a proposed was reflected in an improved focus on the safety generic letter dealing with foreign material concerns and backfit considerations, as well as in intrusion, the CRGR commented that the previ .

the quality of the product, including scope, con- usly issued several relevant NRC generic tents, tone, completeness and consistency with the c mmunications had not been fully effective.

Commission's policies, rules and regulations. Incidents involving debris and potential degrada-tion of the emergency core cooling system In certain cases the CRGR believed that the review continue to occur. Therefore, the Committee process would have been more efficient with better recommended that the NRC take a stronger stance staff preparation for the CRGR review. For some on the subject, such as targeted inspections and review items, the Committee supported the staff's escalated enforcement. The staff was asked to requests for prompt action by scheduling special modify the generic letter (1) to provide greater CRGR meetings, and,in select cases, also by emphasis on the licensees' responsibility to ensure accepting products still under development or even the operability of the stmetures and components partial submittal of the review material. In such in order to comply with the provisions of the cases, the Committee's endorsement was condi-ECCS Rule and the Maintenance Rule, as appli-tional on acceptability of the post-CRGR-review cable;(2) to combine this generic letter with changes made to the documents. Although these another generic letter under preparation on the considentions by the CRGR appears to have helped related subject of potential inadequacies in the the staff meet the schedule pressures, the Commit-licensees' protective coating programs, especially tee conuded that u was not the most efficient use the problems associated with the use of unquali-of its resources, as it necessitated repeat informal fled paints and coatings in the containment;(3) to review both by the Committee and the CRGR staff.

make it clear that NRC believes these lapses in Specific review items to which significant value the foreign material exclusion (FME) programs to was added by the CRGR review are summarized be serious, and is willing to focus necessary below: inspection resources to verify licensee compliance with the existing regulations; and (4) to state

=

During various meetings, while reviewing the that escalated enforcement actions will be taken preliminary and later still-evolving guidance when warranted.

documents for risk-informed regulation, the CRGR commented extensively on the scope and At a subsequent meeting, the CRGR reviewed the contents of these documents. The Committee the re-drafted, re-titled proposed generic letter, commented in detail on the specific documents, which retained some original 50.54(f) informa-recommended conforming changes in sister tion request features. The Committee recom-documents, and proposed overall improvement mended that the differences between the FME in all these documents. The Committee made and the coatings issues (i.e., why the staff does several general observations which were in. not request information on FME but does so on cluded in the Commission paper (SECY-97 the coatings) should be clearly stated. Also, the 077). The CRGR asked the staff to state message in the generic letter should remain explicitly that the risk informed decision process focused on the main safety concerns, with is voluntary, and that viable alternates or ap- separate discussions of the concerns related to proaches remain available to the regulated debris and those related to the coatings. The industry. Also the Commitice complimented the staff was also asked to include in an appendix various office staff w ho Lad demonstrated a the examples of related enforcement actions.

NUREG-1272, Vol. I 1. No. I 66

Reactors Furthermore, the Committee recommended that the reactor protection system and maintaining discussion of ANSI standards and Regulatory adequate shutdown margin. The Committee Guide 1.54-1973 should clearly point out that agreed with the staff in invoking the compliance these are not requirements and that they are exception to the backfit mle.

being included for information only. Addition-a in the case of the proposed rulemaking on ally, the CRGR concluded that the staff should include a stronger statement about the applica- shutdown and spent fue! pool operations, c nsiderable staff-hours were expended in i bility of the Maintenance Rule. Finally, the addressing the CRGR comments and recommen- )

staff should make it clear that the lack of dations. At a preliminary brie 6ng last year, the conformance with the requirements applicable CRGR identified several signi6 cant technical to both debris and coatings are subject to and policy issues which needed better consider-enforcement actions.

ation (e.g., the need for a back6t analysis for the

• Several items reviewed by the CRGR during this spent fuel portion of the rule; the need for the assessment period (e.g., the proposed supple- rule to address con 6guration control; the need rant to Bulletin 96-01," Control Rod Insertion for critical pmbabilistic risk assessment (PRA)

Problems" and the proposed generic letter on peer res :w; use of objective, measurable or i

assurance of sufficient net positive suction head), calculat. e, and enforceable performance criteria; referenced the General Design Criteria (GDC), applicability of the Safety Goals; consideration Appendix A to 10 CFR 50, as " regulatory of ultimate hea.t sink and of spent fuel pool requirements." There was a topic-specific criticality aspects) which prompted the staff to discussion on the applicability of the GDC. establish a task force. This collegial effort by the There may be as many as 60 plants licensed staff then resulted in developing a shutdown and during the pre-GDC days. The Committee spent fuel pool operations rule which will be observed that although over the years the intent performance-based, less prescriptive, and of the design requirements articulated under the consistent with the Commission's policy of GDC have been included in the current licensing risk-informed and performance-based regulation.

basis of various facilities, GDC may not be The impact in this case on staff resources was generically enforceable. By contrast, the plant welljustified from the stand-point of the issues Technical Specifications and the requirements being identified early in the development of included in Appendix B to 10 CFR 50 are. For rulemaking. The impact was also commensurate proposed generic communications, the Commit- with the significance of the issues raised.The tee asked the staff to delete the reference to GDC expenditure of staff resources was ultimately as requirements. reflected in the quality of the Gnal product. At a later CRGR meeting, the CRGR reviewed the

• In the case of the proposed generic letter on proposed rule 10 CFR 50.67 and endorsed it degradation of control rod drive mechanisrn while recognizing dif6culties in treating volun-(CRDM) and vessel head penetrations (VHP), tary licensee actions.

and the supplement to Bulletin 96-01 on control rod insertion problems, the CRGR recognized . While reviewing the proposed generic letter that the requested actions were very prescriptive. titled," Problems With Medium-Voltage Circuit In the case of the proposed CRDM/VHP generic Breakers," the Committee noted that the licens-letter, the generic communication was also being ees are obligated under the existing regulations used as a vehicle to impose new requirements, to conduct tests and maintain appropriate (i.e., clear backfits). The staff was asked to limit records. Therefore, the Committee determined the 50.54(f) request to the items currently that the generic letter was not the appropriate required by regulations. In the case of the vehicle to correct the breaker problem, and bulletin supplement, the CRGR believed that the instead recommended that the staff issue Tempo-proposed actions did have definite merit, be- rary Instmetions and conduct targeted inspec-cause they related to ensuring the operability of tions to ensure licensee ampliance with the 67 CRGR

l 1997 AEOD Annual Report regulations, especially, the provision of 10 CFR -

The Committee made extensive comments on 50, Appendix B, and the Maintenance Rule, as the proposed amendments to 10 CFR 50.55(a).

appropriate. The Committee did not endorse the The Committee believes that this was one of the proposed generic letter for issuance. must significant staff proposals reviewed during The Committee reviewed and endorsed, with this assessment period. The value added is reDected in the final product as follows:

comments, the Safety Evaluation Report on EPRI Topical Report, " Guideline on Evaluation The rulemaking package was voluminous and and Acceptance of Commercial Grade Digital was extremely complicated as it addressed Equipment for Nuclear Safety Applications," ASME Code construction, in-service inspection, EPRI-TR-106439, and the six proposed Gnal and in-service testing activities. The CRGR regulatory guides (RG 1.168 through 1.173) for made extensive comments which improved the computer software to be used in safety systems backfit arguments on several issues. The staff's of nuclear power plants. Specific comments and rationale for the same became stronger and more recommendations were offered on each regula- defensible. The bases for various technical issues tory guide. Additionally, the staff was recom- were strengthened. The staff's approach became mended to use modified wording used in the more focused. Very specific changes were

" Implementation " section of the previously recommended to the Statement of Consider-approved Regulatory Guide 1.152 and 1.153. ations (SOC) and the Regulatory Analysis, and The staff was also asked to make this and other resulted in a major restructuring of the SOC.

confonning changes in all sister regulatory This item was a classic case where the CRGR guides. and the program offices recognized the advan-tage that the Committee offers in terms of The NMSS staff presented to the Committee the scope and schedule of various ongoing activities, individual views of the members (rather than

. Weir office positions) and the recommendations and identified various areas where future CRGR being made by the Committee as a collegial review may be bene 6cial. No inspection guid-body. This aspect of CHGR review is one of the ance or procedures related to nuclear materials most appreciated features of the CRGR review facihties were identified. The Committee process.The CRGR provided full support for commented on the topics prcposed by the staff timely review and endorsement once the finished where CRGR review could add value. The product was received by the Committee. The Committee beheved that CRGR review will be Committee, however, notes that it had to review most beneficial in certain focused areas. Topics a number of signi6 cant documents related to this such as backfitting procedures for the gaseous diffusion plants, issues related to spent fuel rulemaking. Given the volume and the complex-ity of the package, the CRGR believes that in storage and transportation, and those concerning large fuel cycle facihties could benent from h h value added was more than being commensurate with the associated cost imposed.

CRGR review. The Committee noted that there are certain inherent differences in the risks -

The Committee reviewed and endorsed the associated with the operation of the power proposed revision to Generic Letter 91-18, reactors and those related to the operation of "Information to Licensees Regarding NRC nuclear material facilities (e.g., mixed risk from Inspection Manual Section on Resolution of uranium-bearing toxic compounds). The mem-. Degraded and Nonconforming Conditions," and bers also noted the inherent differences in the briefing on the scope and schedule of related established regulatory framework for the nuclear efforts. The Committee noted that the staff had power reactors and nuclear materials facilities. not requested CRGR review / endorsement of They recognized the need for establishing some NUREG-1606, " Proposed Regulatory Guidance ground rules to ensure consistency and unifor- Related to implementation of 10 CFR 50.59 mity in the review nuclear materials topics, if the (Changes. Tests, or Experiments)" Draft for CRGR were to continue in this role. Comments, April 1997, and expressed an interest NUREG-l272 Vol. I 1. No. I 68

Reactors in reviewing related staff efforts in the future. represented in this document, the Committee However, the Committee supported issuance of recommended that the staff develop plans to this revised generic letter and the attached clearly communicate the inspection guidance to inspection guidance, without public comments. the regions to ensure consistent implementation.

Because of the clarification of the staff position 69 CRGR

Reactors 10 INTERNATIONAL EXCHANGE OF INFORMATION 10.1 Incident Reporting System core damage accidents, evaluations of human factors influencing operator performance during The incident Reporting System (IRS) is an interna- operating events, assessment of spent fuel pool tional system jointly operated by the Nuclear cooling, and review ofindustry efforts to manage Energy Agency of the Organization for Economic pressurized-water reactor feedwater nozzle, piping, Cooperation and Development (OECD/NEA) and and feedring cracking and wall thinning.

the International Atomic Energy Agency (IAEA) of the United Nations. The concept of the IRS goes AEOD also reviews reports of selected foreign back to 1978, when the OECD/NEA took the reactor events and identifies those that are safety-initiative to establish an international system for significant that could be applicable to U.S. plants.

exchanging information on safety-related events it then disseminates those reports to the appropri-that occur in nuclear power plants. The accident at ate NRC staff mernbers. In 1997 AEOD received Three Mile Island (TMI) Unit 2 in 1979 highlighted and reviewed approximately 100 reports from in a dramatic way the importance of an effective f reign c untries.

operational experience feedback process, both On October 1,1997, the IRS database was re-nationally and internationally. The TMI accident placed by the Advanced Incident Reporting accelerated the process of establishing an interna- System ( AIRS) database, which is managed and tional IRS. By the end of 1981, OECD countries operated by the IAEA. AIRS is a very powerful formally approved the operation of the system. In text database that is distributed quarterly on April 1983 the I AEA extended the IRS to all of its compact disc read-only memory, member countries with nuclear power programs.

The fundamental objective of the IRS is to contrib- 10.2 InternationalSupport Activities ute to improving the safety of commercial nuclear power plants, which are operated worldwide. As pan of the NRC's international programs, Through the IRS, NEA and IAEA member coun- AEOD exchanges information and ideas on a tries exchange information on safety-significant variety of topics of international interest. For operational events at nuclear power plants that are example, the AEOD staff provided assistance to of generic interest. In 1997 AEOD continued its foreign countries, to the Nuclear Energy Agency efforts to maintain and improve the exchange of (NEA), and to the International Atomic Energy information on operational experience with the Agency (IAEA) in a number of safety-related areas, international community. These efforts have pro- including high burnup fuel and control rod insertion vided valuable data for AEOD studies I id support problems, undetected safety system failures, and the for regulatory actions, extended task force on human factors. In addition,

"" # *E " "I ##

in 1997 AEOD prepared and submitted 29 IRS repons. These reports addressed individual opera-

"' ' "E"" # " #" "

. deselop a common-cause failure database.

tional events and various generic concerns involv-ing nuclear power plants in the U.S., which were AEOD is also the principal U.S. technical representa-identified by the NRC's operational experience tive on reactor operating experience to the Principal feedback program. The reports were based on Working Group 1 (PWG-1)," Operating Experience generic communications sent to U.S. nuclear power and Human Factors," of the NEA's Committee on the plant licensees in the form of NRC reports Safety of Nuclear Installations (CSNI). The fifteenth (NUREG-series as well as AEOD studies), informa- annual meeting of PWG-1 was held September 17 tion notices, bulletins, and generic letters. The through 19,1996, in Paris, France. At this meeting, report topics included precursors to potential severe PWG-1 decided to complete the Nuclear Power Plant 71

1997 AEOD Annual Report L)ata Collection database as soon as possible and to Russia and the Ministry for Environmental Protec-provide the data on CD ROM to member countries in tion and Nuclear Safety of Ukraine, the AEOD staff early 1997. PWG-1 also agreed to recommend to is helping to establish basic functional emergency CSNI to approve the distribution of the IRS database response systems of plans, procedures, facilities, to the World Association of Nuclear Operators. In communications, and training in each country, addition, the Group approved the development of the . .

Advanced Incident Reporting System with a number The first phase of th.is work is complete. Facih. . ties of recommendations. The Study on Undetected in each country have been renovated for an emer-Failures in Safety Systems, led by France with input gency center at the headquaners of the regulatory from Belgium, Finland, Spain, and the U.S., was also authority, computer systems and other essential recommended to CSNI for approval and distribution. components have been mstalled, and telephone communications link the centers with the nuclear AEOD is a participant in the Expert Group on power plants. Russian and Ukrainian regulators Nuclear Emergency Matters. This group was have drafted initial concepts of emergency opera-established by the Committee for Radiation Protec- tions, plans, and procedures.

tion and Public Health in 1989 to improve the . .

quality of national and international nuclear emer- The final phase will emphasize close coordination gency arrangements. Since then, the Expert Group with other emergency response organizations, has sponsored a series of international tabletop impr vement of analytical tools, and exercises. The exercises in 16 countries and 3 workshops on intent is to leave each country self-suf6cient in its speci6c issues identined during those exercises. ability to maintain and improve its emergency The Expert Group is also in the process of planning resp nse system.

for another series of international exercises, the AEOD coordinates its activities with those of the Second International Nuclear Emergency Exercise U.S. Department of Energy and other agencies of (INEX 2). INEX 2 is a series of regional exercises the U.S. Government, as well as with related simulating an accident in one country during which activities of other countries and organizations.

other participating countries and the IAEA respond following their own emergency plan. The simula- AEOD also assisted Ukraine in establishing an tions, called command post exercises, will use incident reporting and operating experience feed-actual nuclear plants, emergency centers, and real- back system. This effort was completed in 1997. 1 time communications. Assistance with the system included developing strategies for operating experience data collection, The U.S. participated in the first INEX 2 exercise events analysis and evaluation, regulatory response sponsored by Switzerland in November 1996. The to events, and experience feedback to nuclear U.S. Environmental Protection Agency (EPA) plants, as well as information exchange between coordinated the U.S. response as the Lead Federal countries of the former Soviet Union with similar Agency for a radiological emergency occurring in a reactors. Six information exchange meetings took foreign country. The NRC also panicipated in this place, including two trips to Ukraine with associ. j response effon. The NRC is working with the EPA ated visits to operating nuclear plants in Ukraine. i and other Federal agencies to incorporate lessens An on-the-job training effort at NRC included J

learned.The NRC has been an active participant in AEOD, NRR, and RES activities. Corrective action 1 each of these activities, all of which mesh with (at NRC Headquarters and Region II) and root  !

areas currently being worked on in this country. cause/ event investigation (in Kiev, Ukraine) semi- I nars were held with participation by several Ukrai- l 10.3 Lisbon Initiative Activities ni n speci lists. Training for four groups of Ukrainian specialists was provided by Idaho AEOD is continuing to assist the regulatory authori- National Engineering and Environmental Labora-i ties of Russia and Ukraine in the improvement of tory (INEEL) in probabilistic risk assessment of their own capabilities to respond to nuclear power operating events, NRC performance indicators, plant emergencies. Working with counterparts in the equipment reliability, and human performance Federal Nuclear and Radiation Safety Authority of issues in operating events. This training was NUREG-1272, Vol. I 1. No. I 72

Reactors completed in early CY 1997. Software to perform December 1992. INES is a ranking system that is P1 and ASP calculations was provided by the NRC used to promptly and consistently communicate to and was made operationalin Ukraine. A reactor the public the safety significance of reported events ,

events database was established using equipment at nuclear installations worldwide. INES was l supplied by the U.S. Agency for International designed by an international group of experts l Development and based on training received at convened jointly by the IAEA and the NEA. The l NRC and INEEL. Appropriate performance international scale is currently in use in 54 countries l

indicators for Ukrainian reactors are being deter- throughout the world.

mined and the use of Accident Sequence Precur- . . .

The NRC usually limits its participation in the sor- type event analysis is being actively INES to rating only events at nuclear power plants - <

developed. These efforts complete activities that are classified as Alerts or higher on the emer-agreed upon for this effort.

gency response scale used m the United States.

However, additional events can be rated on the basis  :

10.4 Limited Participation in of management discretion. After a trial period of i the International Nuclear more than two years, the NRC decided to continue  !

indefinitely its limited participation in the INES.  !

Event Scale There were no events at power reactors during 1997  :

The NRC has participated in a limited manner in for which INES reports were submitted. l the International Nuclear Event Scale (INES) smee  !

)

I i

i I

l l

73 International Exchange

APPENDIX A Plant Operational Experience Data A-1 Performance Indicator Program Data A-2 Other Operational Experience Data l

\

APPENDIX A-1 Performance Indicator Program Data i

1 e

1 1

I

]

9 e yf y-- m - e mei-- -- -- , - . , .

em.

Reactors CONTENTS Performance Indicator Program Data . .. . . A-1-v Introduction . . . . . . . A-1-v Background.. . . . A-I-v Definitions of the Indicators . . . .. . . . . A-1-vi Tables A-1.1 Number of Automatic Scrams While Critical-Quarterly PI Data . . . A-1-1 A-1.2 Number of Safety System Actuations-Quarterly PI Data . . A- 1 -4 A- 1.3 Descriptions of Significant Events for 1997. . A-1-7 A-1.4 Number of Significant Events-Quarterly PI Data .. . A-1-9 A- 1.5 Number of Safety System Failures-Quanerly PI Data . . . . A-1-12 A-l .6 Annual Safety System Failures . . . A-1-15 A-1.7 Forced Outage Rate (Percent)-Quarterly PI Data . . A-1-18 A- 1.8 Equipment Forced Outages /1000 Commercial Critical Hours-Quarterly PI Data . . A-1-21 1

A-1.9 Collect;ve Radiation Exposure (Person-Centisievert [ Person-Rem])-

Quarterly PI Data. . . .. . . A-I-24 A-1.10 Cause Codes-Administrative Control Problems-Quarterly PI Data. . A-I-27 A-1.11 Cause Codes-Licensed Operator Errors-Quarterly Pl Data . . . A-1-30 A-1.12 Cause Codes-Other Personnel Errors-Quarterly PI Data . . A-1-33 A-1.13 Cause Codes-Maintenance Problems-Quanerly PI Data . . A-1-36 A-1.14 Cause Codes-Design / Construction / Installation / Fabrication Problems-Quarterly PI Data... . . A-1-39 A-1.15 Cause Codes-Miscellaneous-Quanerly PI Data. . . A-I-42 l

l A- 1-iii Appendix A-1

.__m._ . - _ _ _ _ _ _ _ _ - _ _ . _ _ . _ _ _ _ _ _ _ _ _ . . - .

Reactors i

PERFORMANCE INDICATOR PROGRAM DATA Introduction Program, the task group report, and the proto- type  ;

report in SECY-86-317, " Performance Indicators," l The NRC program for monitoring Performance dated October 28,1986. The staff briefed the Indicators (PIs) for operating commercial nuclear Commission on the recommended program in }

power reactors includes the following eight indica- November 1986. The Commission approved the tors: (1) the number of unplanned automatic scrams implementation of the program in December 1986, while a reactor is critical, (2) the number of selected instructing the staff to delete the enforcement safety system actuations (3) the number of signifi' action index from the proposed set of indicators.

cant events,(4) the number of safety system fail- Beginning in February 1987, the AEOD staff ures, (5) the forced outage rate,(6) the number of provided quarterly PI reports to the Commission equipment forced outages per 1000 commercial and to NRC senior managers. From March 1993 to critical hours. (7) the collective radiation exposure September 1995, PI reports were issued semiannu-per plant, and (8) cause code trends. ally, in June and December, with data through The data for significant events are provided by the March arid September respectively. The reports are  !

NRC's Office of Nuclear Reactor Regulation n w published annually in December with data (NRR), and the data for collective radiation expo _ through September. Reports are also placed in the NRC's Public Document Room. Beginning with

~

sure are obtained from the Institute of Nuclear the PI report for the fourth quarter of calendar year Power Operations. The data for cause code trends 1989, the staff has routinely provided plant are obtained from the Sequence Coding and Search specific mformatmn extracted from each PI report System database maintained at the Oak Ridge '

to Heensee managers.

National Laboratory. The data for the remaining five Pls are obtained from Trends and Patterns The Commission approved the use of cause code databases maintained at the Idaho National Engi- trends in the P1 report in SECY 89-211, dated neering Laboratory. August 10,1989. At that time the Commission did not approve the use of cause code deviations from the nuclear steam supply system (NSSS) average, Background but instructed the staff to assess the validity of In May 1986 an interoffice task group began to comparing plants to their NSSS average in light of develop an NRC program for using quantitative plant-to-plant variations within NSSS groups. Early indicators of nuclear power plant safety perfor. in the effort to develop suitable peer groups for mance. In July and August 1986, the group con- c mparison of plant performance, it was found that ducted a trial program with 50 operating plants, a plant's operating phase could also have an effect testing 17 prospective performance indicators. For n the occmence of reportable events.To address the most pan, this trial program used data through this issue, the staff imtiated a study to identify l calendar year 1985.The group then selected eight phases of operation in which the frequency of l performance indicators as candidate for initial reponable events varies significantly. The result was implementation. After considering industry com- the development of the operational cycle / peer group i ments, the staff deleted one of the candidate Pls, the methodology. The interoffice task group was corrective maintenance backlog. rec nvened (with new members) in 1992 to assess the proposed changes. The staff's recommendations In October 1986 the NRC prepared a prototype were sent to the Commission for approval in SECY-report of an expanded trial program on 100 92-425 " Performance Indicator Program - Peer operating reactors using data for the Erst half of Group and Operating Cycle Phase Enhancements",

1986. The staff discussed the recommended dated December 23,1992.

A-1v Appendix A-1

1997 AEOD Annual Report Definitions of the Indicators fulfillment of the safety function of structures or systems.The AEOD staff monitors 26 safety Automatic Scrams While Critical. This indicator systems, subsystems, and components for this is the number of all unplanned automatic scrams indicator. If a system consists of multiple redun-that occur while the reactor is critical. A reactor dant subsystems or trains, failure of all trains scram means any actuation of the reactor protection constitutes a safety system failure. Failure of one system that results in control rod motion. The PI of two or more trains is not counted as a safety program also monitors the number of automatic system failure.

scrams that occur while the reactor is critical at or Fored Outage Rate.This is the number of forced below 15 percent power and the number of auto.

matic scrams per 1000 critical hours that occur utage hours in a period divided by the sum of the while the reactor is above 15 percent power.

f reed outage hours and the generator on-line hours. This indicator is used only for plants that are Safety System Actuations.This is the number of in commercial operation.

manual and automatic actuations of the logic for certain emergency core cooling systems, and Equipment Forced Outages per 1000 Commer-actuations of the emergency ac power system that cial Critical Hours. This indicator is the number of f reed outages caused by equipment failures per are caused by loss of power to a vital bus.

1000 cntical hours of commercial reactor operation.

For pressurized-water reactors, only actuations of the it is the inverse of the mean time between forced high-pn ssure injection system, low-pressure injec- outages caused by equipment failures. This indica-tion system or safety injection tanks are counted. For tor is used only for plants that are in commercial boiling-water reactors, only actuations of the high- operation.

pressure coolant injection system, the low-pressure ..

Collective Radiation Exposure.This mdicator is coolant injection system, the high-pressure core the total radiation dose accumulated by unit spray system, or the low-pressure core spray system pers nnel. Prior to the third quarter of 1992, all are counted. Actuations of the reactor core isolation mulu-unit sites except Indian Point and Millstone cooling system are not counted.

reported site total values, which were divided by Significant Events. This is the number of events the number of units at the site to obtain unit l

that the NRC staffidentifies as meeting certain values. Since that time some multi-unit sites have selection criteria. Examples of these events reported mdividual unit values.

include the degradation of important safety ..

Cause Code Trends. This indicator captures the equipment; an unexpected plant response to a pl nt's trends for admimstrative control problems; transient or a major transient itself; a reactor trip licensed operator errors; other personnel errors; with complications; or a degradation of fuel maintenance problems; design, construction, integrity, the primary coolant pressure boundary, inst 11 tion, or fabrication problems; and miscella-or important associated structures.

neous electronic piece-part or environmentally Safety System Failures.This indicator includes related failures.

any event or condition that could prevent the L

NUREG-1272, Vol. I 1, No.1 A-1-vi

Reactors Table A 1.1 Number of Automatic Scrams While Critical- Quarterly PI Data Calendar Year-Quarter Pirnt Name 95-4 96 1 96-2 96-3 96-4 97 1 97-2 97-3 Arkansas 1 0 0 1 1 0 0 0 0 Arkansas 2 0 0 0 0 0 0 0 0

, Beaver Valley 1 0 0 1 0 0 1 0 1 Beaver Valley 2 0 0 0 0 0 2 0 0 Big Rock Point 1 0 0 1 1 0 0 0 Braidwood 1 0 0 0 0 0 0 0 0 Braidwood 2 0 0 0 0 0 0 0 0 Browns Ferry 1 0 0 0 0 0 0 0 0 Browns Ferry 2 0 0 1 0 1 0 1 0 Browns Ferry 3 0 1 2 0 0 0 0 0 Brunswick 1 0 0 0 0 0 0 0 0 Brunswick 2 0 0 0 0 0 0 0 0 Byron 1 0 0 0 1 0 0 0 0 Byron 2 0 0 0 0 0 0 0 0 Callaway 0 0 0 0 0 0 0 0 Calvert Cliffs 1 0 0 0 0 0 0 0 0 Calven Cliffs 2 0 1 0 0 1 0 0 0 Catawba 1 0 0 0 0 0 0 0 0 Catawba 2 0 1 0 0 0 0 1 0 Clinton 1 0 0 1 0 0 0 0 0 Comanche Peak 1 1 1 0 1 0 0 0 0 ,

I Comanche Peak 2 1 0 0 1 1 0 0 0 Cook 1 0 1 0 1 0 0 0 0 l Cook 2 0 0 1 0 0 1 0 0 Cooper Station 0 0 0 0 0 0 0 0 Crystal River 3 0 0 1 0 0 0 0 0 i 0 0 0 l Davis-Besse 0 0 0 0 1 Diablo Canyon 1 0 0 0 1 1 0 0 0 Diablo Canyon 2 0 0 0 1 0 1 0 0 '

Dresden 2 0 0 0 0 0 0 0 0 Dresden 3 0 0 1 0 0 0 0 0 Duane Arnold 0 0 0 0 0 0 0 0 Farley 1 0 0 0 0 0 0 0 0 Farley 2 1 0 0 0 0 0 0 0 Fermi 2 0 0 0 0 1 0 0 0 FitzPatrick 0 0 0 1 0 0 0 0 Fon Calhoun 0 0 0 0 0 0 0 0 Ginna 0 0 0 1 0 0 0 0 Grand Gulf 0 0 0 0 0 0 0 0 Haddam Neck 0 0 0 0 0 PSD PSD PSD PSD means the plant was permanently shutdown.

A-1-1 Appendix A-1

1997 AEOD Annual Report Table A-1.1 Number of Automatic Scrams While Critical- Quarterly PI Data Calendar Year-Quarter Plant Name 95 4 96-1 96 2 96-3 96-4 97 1 97 2 97 3 Harris 2 0 1 0 0 1 1 1 Hatch 1 0 2 1 0 0 0 0 0 Hatch 2 0 0 0 0 0 0 1 0 '

Hope Creek 0 0 0 0 0 0 0 0 Indian Point 2 0 1 1 2 0 1 0 2 Indian Point 3 0 0 0 0 0 0 0 2 Kewaunee 0 1 1 0 0 0 0 0 LaSalle 1 0 0 1 0 0 0 0 0 LaSalle 2 0 0 0 0 0 0 0 0 Limerick 1 0 0 1 I O O O O Limerick 2 0 0 1 0 0 0 0 0 Maine Yankee 0 1 0 0 1 0 0 0 McGuire 1 1 0 0 0 0 0 0 1 McGuire 2 0 0 1 0 0 0 0 2 Millstone 1 0 0 0 0 0 0 0 0 Millstone 2 0 0 0 0 0 0 0 0 Millstone 3 0 0 0 0 0 0 0 0 Monticello 0 0 2 0 0 0 0 0 Nine Mile Pt.1 0 0 1 0 1 0 0 0 Nine Mile Pt. 2 0 0 0 0 0 0 0 0 North Anna 1 0 0 0 1 1 0 0 0 North Anna 2 1 0 0 0 1 0 0 0 Oconee1 0 1 0 0 0 0 0 0 Oconee 2 0 0 0 0 0 0 0 i Oconee 3 0 1 0 0 0 1 0 0 ,

l Oyster Creek 1 0 1 1 0 0 0 0 Palisades 0 0 0 0 0 0 0 0 Palo Verde 1 2 0 0 1 0 0 1 0 Palo Verde 2 0 1 0 0 0 0 0 0 Palo Verde 3 0 0 0 1 0 0 1 0

! Peach Bottom 2 0 0 0 0 2 0 0 0 l Peach Bottom 3 1 0 0 0 0 0 0 0 l l Pen 0 0 1 0 0 1 1 0  !

l Pilgrim 0 0 1 0 0 0 0 0 Point Beach 1 0 0 0 0 0 0 0 0 Point Beach 2 0 0 1 0 0 0 0 0 Prairie Island 1 0 0 1 0 0 0 1 0 Prairie Island 2 0 1 2 0 0 0 0 0

- Quad Cities 1 0 0 0 0 0 0 0 0 Quad Cities 2 0 0 0 0 0 0 0 0 NUREG-1272, Vol. I 1. No.1 A- t -2

Reactors Table A 1.1 Number of Automatic Scrams While Critical-Quarterly PI Data Calendar Year-Quarter PI:nt Name 95-4 96-1 96 2 96 3 96-4 97 1 97-2 97-3 River Bend 0 0 0 0 0 0 0 1 Robinson 2 0 0 0 0 1 0 0 0 Salem i 0 0 0 0 0 0 0 0 Salem 2 0 0 0 0 0 0 0 0 San Onofre 2 0 0 0 0 0 0 0 0 San Onofre 3 0 0 0 0 0 0 0 0 Seabrook 0 1 0 0 0 0 1 0 Sequoyah1 0 0 1 0 0 0 0 0 Sequoyah 2 ' 0 0 0 0 1 0 0 0 South Texas 1 1 0 0 0 0 0 0 '0 South Texas 2 1 0 0 0 0 1 0 0 St. Lucie 1 0 0 0 0 0 1 0 0 St. Lucie 2 0 0 0 0 0 0 0 0 Summer 0 0 0 0 0 0 1 0 Suny1 0 0 0 0 0 0 0 0 Surry 2 1 0 1 0 1 0 0 0 Susquehanna 1 0 0 0 1 0 0 0 0 Susquehanna 2 0 0 0 0 0 0 0 0 Three Mile Isl 1 0 0 0 0 0 0 1 0 Turkey Point 3 0 1 0 0 0 0 0 1 Tbrkey Point 4 0 0 0 0 0 0 1 0 Vermont Yankee 1 0 0 0 0 0 1 0 j Vogtle 1 0 0 0 0 1 0 0 0 Vogtle 2 0 0 0 0 1 0 0 0 Wash. Nuclear 2 0 0 0 0 0 0 0 0 Waterford 3 0 0 1 0 0 0 0 0 Watts Bar 1 NYC 0 2 0 0 2 1 0 Wolf Creek 0 0 1 0 0 0 0 0 Zion ! 0 1 1 1 0 0 0 0 Zion 2 0 0 0 0 0 0 0 0 Total 16 17 34 19 17 13 15 12 NYC means the plant was not yet critical.

A- t -3 Appendix A-1

1997 AEOD Annual Report Table A 1.2 Number of Safety System Actuations-Quarterly PI Data Calendar Year-Quarter i

Plant Name 95-4 %1 96-2 96-3 96 4 97 1 97-2 97-3  !

Arkansas 1 0 0 0 0 0 0 0 0 Arkansas 2 0 0 0 0 0 0 0 0 )

Beaver Valley 1 0 0 0 0 0 0 0 0 Beaver Valley 2 0 0 0 0 0 0 0 0 l Big Rock Point 0 0 0 0 0 0 0 0 j Braidwood 1 0 0 0 0 0 0 0 0 Braidwood 2 0 1 0 0 0 0 0 0 Browns Ferry 1 0 0 0 0 0 0 0 0 Browns Ferry 2 0 1 2 0 0 0 1 0 Browns Ferry 3 0 0 2 0 0 1 0 0 Brunswick 1 0 0 0 0 1 0 1 0 Brunswick 2 0 0 0 0 0 0 0 0 ,

Byron 1 0 0 1 0 0 0 0 0 Byron 2 0 0 0 0 0 0 0 0 Callaway 0 0 0 0 0 0 0 0  ;

Calvert Cliffs 1 0 1 0 0 0 0 0 0 Calvert Cliffs 2 0 1 0 0 0 0 0 0 Catawba 1 0 0 0 0 0 0 0 0 t Catawba 2 0 2 0 0 1 0 0 0 l Clinton 1 C 0 0 0 0 0 0 0 Comanche Peak 1 0 1 0 0 0 0 0 0 Comanche Peak 2 0 0 0 0 0 0 0 0 Cook 1 0 0 0 0 0 0 0 0 Cook 2 0 0 0 0 0 0 0 0 Cooper Station 0 0 0 0 0 0 0 0 Crystal River 3 0 0 0 0 0 0 0 0 Davis-Besse 0 0 0 0 0 0 0 0  :

Diablo Canyon i 1 0 0 0 0 0 1 0 '

Diablo Canyon 2 0 0 0 0 0 0 0 0 Dresden 2 0 0 0 0 0 0 0 0 Dresden 3 0 0 1 0 0 0 0 0  !

Duane Arnold 0 0 0 0 0 0 0 0 Farley 1 0 0 0 0 0 0 0 0 ,

Farley 2 0 0 0 0 0 0 0 0 ,

Fermi 2 0 0 0 0 0 0 0 0 ,

FitzPatrick 0 0 0 1 0 0 0 0 Fon Calhoun 0 0 0 0 0 0 0 0 ,

Ginna 0 0 0 0 0 0 0 1 ,

Grand Gulf 0 0 0 0 0 0 0 0  !

Haddam Neck 0 0 0 0 0 PSD PSD PSD PSD means the plant was permanently shutdow n.

NUREG-1272, Vol. I1. No. I A- 1 -4 1

Reactors l

Table A-1.2 Number of Safety System Actuations- Quarterly PI Data l

l Calendar Year-Quarter Pirnt Name 95-4 96 1 96-2 96-3 96-4 97 1 97-2 97-3 i

Harris 2 0 1 0 0 0 1 0 Hatch l 0 0 0 0 0 0 0 0 l

Hatch 2 0 0 0 0 0 0 2 0 Hope Creek 1 0 0 0 0 0 0 0 Indian Point 2 0 0 1 0 0 0 0 1 Indian Point 3 0 2 0 0 0 0 2 0 Kewaunee 0 0 0 0 0 0 0 0 l LaSalle l 0 0 0 0 0 0 0 0 LaSalle 2 0 0 0 0 0 0 0 0 '

Limerick 1 0 0 0 0 0 0 0 0 Limerick 2 0 0 0 0 0 0 0 0 ,

Maine Yankee 0 0 0 0 0 0 0 0 McGuire 1 0 0 0 0 0 0 0 0 McGuire 2 0 0 0 0 1 0 1 0 Millstone 1 0 0 0 0 0 0 0 0 Millstone 2 0 0 0 0 0 0 0 0 Millstone 3 0 0 1 0 0 0 0 0 Monticello 0 0 0 0 0 0 1 0 Nine Mile Pt 1 0 0 0 0 0 0 0 0 Nine Mile Pt. 2 0 0 0 0 0 0 0 0 Nonh Anna l 0 0 0 0 0 0 0 0 Nonh Anna 2 0 0 0 0 0 0 0 0 Oconee 1 0 0 0 0 0 0 0 0 Oconee 2 0 0 0 0 0 0 0 1 Oconee 3 0 1 0 0 0 0 0 0 Oyster Creek 0 0 0 0 0 0 0 1 ,

Palisades 0 0 0 0 0 0 0 0 l Palo Verde 1 1 0 0 0 0 0 0 0 Palo Verde 2 1 0 1 0 0 0 0 2 Palo Verde 3 0 0 0 0 0 0 0 0 Peach Bottom 2 0 0 0 0 0 0 0 1 l Peach Bottom 3 0 0 0 0 0 0 0 0 Perry 0 1 0 0 0 1 1 2 Pilgrim 0 0 0 0 0 1 0 0 Point Beach 1 0 0 1 0 0 0 0 0 t

l Point Beach 2 0 0 0 0 0 0 0 0 l

Prairie Island 1 0 0 1 0 0 0 0 0 Prairie Island 2 0 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 Quad Cities 1 0 0 0 0 0 1 0 0 Quad Cities 2 A-15 Appendix A-1 1

1997 AEOD Annual Report Table A 1.2 Number of Safety System Actuations - Quarterly PI Data Calendar Year-Quarter l

Plant Name 95-4 96-1 96-2 96 3 96-4 97-1 97-2 97-3 River Bend 0 0 0 0 0 0 0 0 Robinson 2 0 0 0 0 0 0 0 0 Salem ! 0 0 0 0 0 0 0 -0 I Salem 2 0 0 0 0 2 0 0 0 San Onofre 2 0 0 0 0 0 0 0 0 i San Onofre 3 0 0 0 0 0 0 0 0 Seabrook 0 0 0 0 0 0 0 0 Sequoyah1 0 0 0 0 0 0 2 0 Sequoyah 2 0 0 1 0 1 0 0 0 South Texas 1 I O O O O O O 0 1 i

South Texas 2 0 0 0 0 0 0 0 0 St. I,ucie 1 0 0 3 0 0 0 0 0 St. Lucie 2 0 0 0 0 0 0 0 0 Summer 0 0 0 0 0 0 0 0 Surry 1 0 0 0 0 0 0 0 0 Surry 2 0 0 0 0 0 0 0 0 Susquehanna 1 0 0 0 1 0 0 0 0 Susquehanna 2 0 0 0 1 0 0 0 0 Three Mile Isl 1 0 0 0 0 0 0 1 0 Turkey Point 3 0 0 0 0 0 0 0 0 Tbrkey Point 4 0 1 0 0 0 0 0 0 Vermont Yankee 0 0 0 0 0 0 0 0 Vogtle 1 0 0 0 0 0 0 0 0 Vogtle 2 0 0 0 0 0 0 0 0 Wash. Nuclear 2 0 0 0 0 0 0 0 0 l Waterford 3 0 0 0 0 0 0 0 0 Watts Bar 1 0 0 0 0 0 3 0 0 I WolfCreek 1 0 0 0 0 0 0 0 Zion 1 1 1 0 0 0 1 0 0 '

Zion 2 0 0 0 0 0 0 0 0 Total 9 13 17 3 6 8 14 9 i

i

. l l

l l

NUREG-1272. Vol. I 1, No,1 A-I-6 l

t

.-_.- . . . -. _- . . - ~ - - - - - - - . . - - .- .- ---

Reactors Table A 1.3 Descriptions of Significant Events for 1997 Event Rx NRC Plant Name Date Type Region Description of Event Beaver Valley 2 12/02/96 PWR I The reactor head vent system was not capable of perform-ing its design functions of venting the head or providing an alternate letdown How path. Leak sealant material had migrated from a repair site and fouled valves and piping.

Calvert Cliffs 2 04/03/97 PWR I Several communication breakdowns. mopounded by several failures in the licensee's conuvi programs, resulted in a diver getting close to a very high radiation area in the spent fuel storage pool.

l Cook I and 2 09/08/97 PWR 111 Multiple deficiencies were found involving the emergency core cooling and containment heat removal systems. The safety margin of these systems to perform their recircula-  ;

tion and containment heat removal safety functions following a loss of coolant accident was significantly diminished.  ;

Oconee 2 04/21/97 PWR 11 Licensee programmatic failures, including weaknesses in i their high pressure injection line augmented inspection i program, resulted in an unisolable pressure boundary leak ,

on the high pressure injection to reactor coolant system nozzle. ,

1 Occonee 3 05/03/97 PWR 11 Two high pressure injection pumps were damaged from inadequate net positive suction head as a result of an inaccurate letdown storage tank level indication. The I inaccuracy was caused by a drained reference leg that was shared by both trains of lesel instmmentation.

Sequoyah1 03/24/97 PWR II An excessive amount of primary coolant was drained while lowering pressurizer level. An erroneous pressurizer level indication resulted from a partial voiding in the instrument reference leg. Level could have decreased to the point of affecting shutdown cooling.

Sequoyah 2 10/11/96 PWR 11 A manual scram resulted from a malfunction of the turbir.e impulse pressure switches. Complications included a feedwater isolation with one isolation valve failing to close ,

and the inability to manually control the motor driven  !

auxiliary feedwater pump level control valves.

Zion 1 02/21/97 PWR 111 While shutting down the reactor, the reactor operator ,

continuously inserted control rods 232 steps, taking the  !

reactor substantially suberitical, and then continuously l withdrew the rods 84 steps in an attempt to reestablish power at the point of adding heat.

A- 1 -7 Appendix A-1

1997 AEOD Annual Report Table A-1.3 Descriptions of Significant Events for 1997 Event Rx NRC Plant Name Date Type Region Description of Event Zion 2 03/07/97 PWR III Nitrogen gas from the volume control tank accumulated in the reactor vessel head, displacing approximately 6900 gallons of water. The potential existed to form a void that could have impacted decay heat removal.

NUREG-1272, Vol. I 1. No.1 A- 1 -8

-- _ ~ ~ . . . .- .- -. . - . . - - . . .

1 Reactors  !

Table A 1.4 Number of Significant Events -Quarterly PI Data Calendar Year-Quarter PI:nt Name 95-4 96 1 96-2 %3 %4 97 1 97 2 97-3 Arkansas 1 0 0 1 0 0 0 0 0 Arkansas 2 0 0 0 0 0 0 0 0 Beaver Valley 1 0 0 0 ) 0 0 0 0  ;

Beaver Valley 2 0 0 0 0 1 0 0 0 Big Rock Point 0 0 0 0 0 0 0 0 1

Braidwood 1 0 0 0 0 0 0 0 0 Braidwood 2 0 0 0 0 0 0 0 0 i Browns Ferry 1 0 0 0 0 0 0 0 0 l Browns Ferry 2 0 0 0 0 0 0 0 0 i Browns Ferry 3 0 0 0 0 0 0 0 0  !

Brunswick 1 0 0 0 0 0 0 0 0 ,

Bnmswick 2 0 0 0 0 0 0 0 0 l Byron ! 0 0 0 0 0 0 0 0 l Byron 2 0 0 0 0 0 0 0 0 l Callaway 0 0 0 0 0 0 0 0 Calven Cliffs 1 0 0 0 0 0 0 0 0 Calvert Cliffs 2 0 0 0 0 0 0 1 0 Catawba 1 0 0 0 0 0 0 0 0 Catawba 2 0 1 0 0 0 0 0 0 i Clinton 1 0 0 0 0 0 0 0 0 l Comanche Peak I O O O O O O O O Comanche Peak 2 0 0 0 0 0 0 0 0 Cmkl 0 0 0 0 0 0 0 1 Cook 2 0 0 0 0 0 0 0 1 Cooper Station 0 0 0 0 0 0 0 0  ;

Crystal River 3 0 0 0 0 0 0 0 0 Davis-Besse 0 0 0 0 0 0 0 0 Diiblo Canyon 1 0 0 0 0 0 0 0 0 Diablo Canyon 2 0 0 0 0 0 0 0 0 Dresden 2 0 0 0 0 0 0 0 0 Dresden 3 0 0 1 0 0 0 0 0 Duane Arnold 0 0 0 0 0 0 0 0 Farley l 0 0 0 0 0 0 0 0 Farley 2 0 0 0 0 0 0 0 0 Fenni 2 0 0 0 0 0 0 0 0 FitzPatrick 0 0 0 1 0 0 0 0 Fort Calhoun 0 0 0 0 0 0 0 0 Ginna 0 0 0 0 0 0 0 0 Grand Gulf 0 0 0 0 0 0 0 0 Haddam Neck 0 0 0 1 0 PSD PSD PSD PSD means the plant was permanently shutdown.

A- t-9 Appendix A-1

i 1997 AEOD Annual Report 1

1 Tahic A-1.4 Numher of Significant Events - Quarterly P1 Dr.ta  !

Calendar Year-Quarter Plant Name 95-4 96-1 96-2 96-3 96-4 97-1 97-2 97-3 1

'm%

liarris 0 0 0 0 0 0 0 0 r 11atch I O O O O O O O O l liatch 2 0 0 0 0 0 0 0 0 11 ope Creek 0 1 0 0 0 0 0 0 '

Indian Point 2 0 0 0 0 0 0 0 0  ;

Indian Point 3 0 0 0 0 0 0 0 0 Kewaunee 0 0 0 0 0 0 0 0 LaSalle 1 0 0 0 0 0 0 0 0 LaSalle 2 0 0 0 0 0 0 0 0  :

Limerick l 0 0 0 0 0 0 0 0 t Limerick 2 0 0 0 0 0 0 0  !

Maine Yankee 0 0 0 0 0 0 0 0 hkGuire 1 0 0 0 0 0 0 0 0 McGuire 2 0 0 0 0 0 0 0 0 Millstone 1 0 0 0 0 0 0 0 0 Millstone 2 0 0 0 0 0 0 0 0 Millstone 3 0 0 0 0 0 0 0 0 Monticella 0 0 0 0 0 0 0 0  ;

Nine Mile Pt.1 0 0 0 0 0 0 0 0 i Nine Mile Pt. 2 0 0 0 0 0 0 0 0 North Anna 1 0 0 0 0 0 0 0 0 North Anna 2 0 0 0 0 0 0 0 0 Oconee 1 0 0 0 0 0 0 0 0 Oconec 2 0 0 0 1 0 0 1 0 l Oconee3 0 0 0 0 0 0 1 0 i Oyster Creek 0 0 0 0 0 0 0 0 Friisades 0 0 0 0 0 0 0 0 Palo Verde 1 0 0 0 0 0 0 0 0 Palo Verde 2 0 0 0 0 0 0 0 0 Palo Verde 3 0 0 0 0 0 0 0 0 Peach Bottom 2 0 0 0 0 0 0 0 0 Peach Bottom 3 0 0 0 0 0 0 0 0 '

Perry 0 0 0 0 0 0 0 0 i Pilgrim 0 0 0 0 0 0 0 0 Point Beach 1 0 0 0 0 0 0 0 0 Point Beach 2 0 0 0 0 0 0 0 0 Prairie Island 1 0 0 0 0 0 0 0 0 Prairie Island 2 0 0 0 0 0 0 0 0  ;

Quad Cities 1 0 0 0 0 0 0 0 0  !

0 0 Quad Cities 2 0 0 0 0 0 0 l

! NUREG-1272, Vol. I 1, No.1 A-1-10

Reactors Table A 1.4 Number of Significant Events-Quarterly PI Data Calendar Year-Quarter PI:nt Name 95-4 96-1 96-2 96-3 96 4 97 1 97-2 97 3 River Bend 0 0 0 0 0 0 0 0 Robinson 2 0 0 0 0 0 0 0 0 Salem l 0 0 0 0 0 0 0 0 Salem 2 0 0 0 0 0 0 0 0 San Onofre 2 0 0 0 0 0 0 0 0 l l

San Onofm 3 0 0 0 0 0 0 0 0 l Seabrook 0 0 0 0 0 0 0 0 SequayahI O O O O O 1 0 0 Sequoyah 2 0 0 0 0 1 0 0 0 South Texas 1 1 0 0 0 0 0 0 0 Scuth Texas 2 0 0 0 0 0 0 0 0 St. Lucie 1 0 0 0 0 0 0 0 0 St. Lucie 2 0 0 0 0 0 0 3 0 Summer 0 0 0 0 0 0 0 0 Surry 1 0 0 0 0 0 0 0 0 .

Surry 2 _ 0 0 0 0 0 0 0 0 Susquehanna 1 0 0 0 0 0 0 0 0 Susquehanna 2 0 0 0 0 0 0 0 0 Three Mile Isl 1 0 0 0 0 0 0 0 0 l hrkey Point 3 0 0 0 0 0 0 0 0 hrkey Point 4 0 0 0 0 0 0 0 0 Vermont Yankee 0 0 0 0 0 0 0 0 Vogtle 1 0 0 0 0 0 0 0 0 Vogtle 2 0 0 0 0 0 0 0 0 Wash. Nuclear 2 0 0 0 0 0 0 0 0 Waterford 3 0 0 0 0 0 0 0 0 Watts Bar 1 0 0 0 0 0 0 0 0 Wolf Creek 0 1 0 0 0 0 0 0 Zion 1 0 0 0 0 0 1 0 0 l Zion 2 0 0 0 0 0 1 0 0 Total 1 3 2 3 2 3 3 2 A-!.11 Appendix A-1

1997 AEOD Annual Report Table A 1.5 Number of Safety System Failures -Quarterly P1 Data

(

Calendar Year-Quarter Plant Name 95-4 96-1  %-2 96 3 96-4 97 1 97 2 97-3 Arkansas 1 0 0 1 1 0 0 0 0 Arkansas 2 0 0 0 1 0 1 0 0 Beaver Valley 1 0 0 1 1 2 0 2 3 Beaver Valley 2 0 0 0 0 2 1 0 1 Big Rock Point 0 1 0 0 0 0 1 0 Braidwood i 1 1 2 0 0 0 1 0 Braidwood 2 0 2 2 0 0 0 1 0 ltrowns Ferry 1 0 0 0 0 0 0 0 1 Browns Ferry 2 I i 1 0 0 0 0 1 Browns Ferry 3 0 1 0 2 2 0 0 1 Brunswick i 1 1 1 0 0 0 0 i Brunswick 2 0 1 0 0 0 0 0 2 Byron 1 0 0 2 0 1 2 1 0 Byron 2 0 0 2 0 1 1 0 0 Callaway 0 0 0 0 0 0 0 0 Calven Cliffs 1 0 1 1 1 0 0 0 0 Calvert Cliffs 2 0 1 0 0 0 0 i 0 Catawba 1 0 0 0 0 0 0 1 0 Catawba 2 0 0 0 1 0 0 1 0 Clinion 1 0 1 1 0 2 2 3 4 Comanche Peak i O 2 0 0 0 0 0 0 Comanche Peak 2 0 1 0 0 0 0 0 0 Cook 1 0 0 0 0 0 0 1 5 Cook 2 0 0 0 0 0 0 1 5 Cooper Station 2 2 1 3 1 0 1 1 Crystal River 3 2 6 2 0 1 2 2 4 Davis-Besse 1 0 0 0 2 0 0 2 Diablo Canyon 1 2 2 3 0 0 0 0 0 Diablo Canyon 2 0 2 2 0 0 0 0 0 Dresden 2 1 1 1 1 3 1 2 0 Dresden 3 1 2 0 1 2 0 2 2 Duane Arnold 1 0 0 0 1 0 1 0 Farley 1 0 0 1 0 0 1 0 1 Farley 2 0 0 0 0 0 1 0 1 Fermi 2 0 3 0 0 1 1 2 0 FitzPatrick 0 3 1 1 0 0 0 1 Fon Calhoun 1 0 0 0 5 0 0 1 Ginna 0 1 0 0 2 0 0 0 Grand Gulf 1 0 0 0 0 0 0 0 Haddam Neck 3 1 3 6 1 PSD PSD PSD PSD means the plant was permanently shutdown.

NUREG 1272.Vol. I1, No. I A-I12

Reactors i

Table A 1.5 Number of Safety System Failures-Quarterly PI Data Calendar Year-Quarter Plant Name 95-4 96 1 96-2 96 3 96-4 97 1 97-2 97 3 l Ihrris 0 0 1 0 0 0 0 1 11atch 1 0 0 1 0 0 0 0 1 liatch 2 1 0 1 0 0 1 0 1 liope Creek 1 3 I I 2 3 2 2 Indian Point 2 2 1 0 1 0 0 1 0 Indian Point 3 2 3 1 1 0 0 2 4 Kewaunee 0 1 0 1 1 0 0 0 LaSalle 1 0 0 1 0 3 2 0 1 I,aSalle 2 1 0 1 0 4 2 0 0 Limerick 1 0 2 1 2 0 0 0 i Limerick 2 0 2 1 1 0 0 1 1 Maine Yankee 2 0 1 1 2 1 1 0 McGuire 1 0 0 1 0 0 0 0 0 McGuire 2 0 1 0 0 0 1 0 0 Millstone 1 2 11 7 0 3 2 2 1 Millstone 2 1 10 3 1 3 1 5 2 Millstone 3 0 3 7 6 5 5 4 1 Monticello - 1 1 1 0 1 0 0 1 Nine Mile Pt.1 0 0 0 0 0 0 1 2 Nine Mile Pt. 2 0 0 0 1 2 0 1 0 Nenh Anna l 0 1 0 1 0 0 1 1 Nonh Anna 2 0 0 0 1 0 0 0 1 Oconee 1 0 1 1 0 0 1 0 0 Oconee 2 0 1 1 0 1 1 0 0 Oconee 3 0 1 1 0 1 1 2 3 Oyster Creek 0 0 1 0 1 0 0 0 Palisades 1 3 0 1 1 0 0 1 Palo Verde 1 0 0 1 0 0 0 0 1 Palo Verde 2 0 0 1 1 0 0 0 1 i Palo Verde 3 0 0 1 0 0 0 0 1 Peach Bottom 2 0 0 1 0 1 0 1 1 Peach Bottom 3 0 0 1 0 0 0 0 1 i Perry 0 0 0 0 1 0 0 0 l Pilgrim 1 0 1 0 0 0 2 1 Point Beach 1 0 0 1 1 1 2 8 2 Point Beach 2 0 0 1 0 1 2 7 i Prairie Island i 1 0 1 0 0 0 2 0 Prairie Island 2 1 0 1 0 0 0 2 0 Quad Cities 1 I 2 1 1 1 1 4 0 Quad Cities 2 3 1 3 1 2 0 1 1 A- 1-13 Appendix A-1

1997 AEOD Annual Report I

Table A 1,5 Number of Safety System Failures- Quarterly PI Data Calendar Year-Quarter i

Plant Name 95-4 96-1 96 2 96-3 96-4 97-1 97 2 97-3  :

River Bend 2 0 1 0 0 0 0 3 Robinson 2 0 0 0 1 0 0 1 1 Salem 1 4 2 2 4 1 1 1 0 Salem 2 4 2 3 5 1 1 1 0

- San Onofre 2 , 1 0 0 0 1 0 0 1 l

San Onofre 3 1 0 0 0 0 0 0 1 Seabrook 0 0 1 0 0 0 0 0 ,

Sequoyah1 0 0 0 0 0 1 0 0 t Sequoyah 2 0 0 0 0 0 0 0 0 South Texas 1 0 0 0 0 0 0 0 0 South Texas 2 0 0 0 0 0 0 0 0 I

St. Lucie 1 0 1 0 2 0 0 0 0

.St. Lucie 2 1 0 1 0 0 0 0 0 Summer 0 0 0 0 0 0 0 0 Surry 1 1 0 1 0 0 0 0 l Surry 2 1 0 0 0 0 0 0 1 Susquehanna 1 2 0 0 1 2 0 0 1 ,

Susquehanna 2 1 0 0 1 1 1 0 1 Three Liile 1st 1 0 0 0 0 2 0 1 0 Turkey Point 3 0 1 0 0 0 0 0 0 ,

'Ibrkey Point 4 0 1 0 0 0 1 0 0 Vermont Yankee 0 1 2 1 1 3 0 i Vogtle 1 0 1 1 0 1 1 0 0 Vogtle 2 0 0 1 0 0 1 0 0 Wash. Nuclear 2 0 0 0 0 1 0 0 0

. Waterford 3 1 1 2 0 1 2 3 0 Watts Bar 1 0 2 1 1 0 0 1 0 Wolf Creek 1 3 1 0 1 0 0 0 Zion 1 0 2 1 1 0 2 1 1 Zion 2 0 3 2 1 0 1 1 1 Total 59 105 95 60 79 54 84 82 i

NUREG-1272 Vol. I1, No.1 A- 1 -14

Reactors l

l Table A-1.6 Annual Safety System Failures j 1

Plint Name Rx Type CY93 CY94 CY95 CY96 1997 Arkansas 1 PWR 2 0 0 2 0 Arkansas 2 PWR I i 1 1 1 Beaver Valley 1 PWR I I O 4 7

~ Beaver Valley 2 PWR 3 1 0 2 4 ,

Big Rock Point BWR 2 1 1 1 1 t

Braidwood 1 PWR 2 4 1 3 1 Braidwood 2 PWR 2 3 2 4 1 Browns Ferry i BWR 0 0 0 0 1 Browns Ferry 2 BWR I 3 2 2 1 ,

Browns Ferry 3 BWR 0 0 0 5 3 Brunswick 1 BWR 1 3 3 2 i Brunswick 2 BWR 3 3 1 1 2 Byron i PWR 1 3 0 3 4 Byron 2 PWR 1 2 1 3 2 Callaway PWR 3 1 0 0 0 Calven Cliffs 1 PWR 2 2 0 3 0 Calvert Cliffs 2 PWR 2 0 1 1 1 Catawba 1 PWR 3 1 2 0 1 Catawba 2 PWR 3 1 2 1 1 Clinton 1 BWR 0 2 1 4 11 Comanche Peak i PWR 0 1 1 2 0 Comanche Peak 2 PWR 0 1 1 1 0 Cookl PWR 0 2 3 0 6 ,

Cook 2 PWR 0 1 0 0 6  !

Cooper Station BWR 11 8 4 7 3 Crystal River 3 PWR 1 3 7 9 9 Davis-Besse PWR 1 1 1 2 4 Diablo Canyon 1 PWR 3 2 4 5 0 Diablo Canyon 2 PWR 5 2 1 4 0 Dresden 2 BWR 8 9 2 6 6 Dresden 3 BWR 7 5 9 5 6 Duane Arnold BWR 4 3 2 1 2 Farley 1 PWR 1 0 2 1 2 Farley 2 PWR 1 0 1 0 2 Fermi 2 BWR 3 1 1 4 4 FitzPatrick BWR 9 2 1 5 1 Fort Calhoun PWR 6 2 2 5 6 Ginna PWR 2 1 0 3 2 Grand Gulf BWR 6 1 1 0 0 Haddam Neck PWR 7 4 9 11 1 A-I- 15 Appendix A-1

1997 AEOD Annual Report  !

l l

Table A-1.6 Aminual Safety System Failures Plant Name Rx Type - CY93 CY94 CY95 C'/96 199"'

- Harris PWR 3 1 0 1 1 >

Hatch i BWR 2 3 2 1 1 11atch 2 BWR 4 2 1 1 2 Hope Creek BWR 4 1 5 7 9 indian Point 2 PWR 4

{

0 3 2 1  ;

Indian Point 3 PWR 13 4 5 5 6 Kewaunee PWR 2 0 0 3 1 LaSalle 1 BWR 8 3 2 4 6 LaSalle 2 BWR 6 5 2 5 6 Limerick i BWR 1 0 1 5 i Limerick 2 BWR 1 0 0 4 2 Maine Yankee PWR 2 1 7 4 4 McGuire i PWR 6 0 1 1 0 McGuire 2 PWR 4 0 1 1 I Millstone i BWR 6 7 7 21 8 Millstone 2 PWR 7 8 11 17 11 Millstone 3 PWR 4 1 4 21 15 Monticello BWR 5 2 2 3 2 Nine Mile Pt. I BWR 0 1 2 0 3 Nine Mile Pt. 2 BWR 3 0 1 3 3  ;

North Anna 1 PWR 3 0 0 2 2  !

North Anna 2 PWR 3 0 0 1 1 Oconee 1 PWR 5 2 1 2 1 Oconee 2 PWR 6 1 1 3 2 Oconee 3 PWR 6 2 1 3 4 Oyster Creek BWR 1 3 1 2 1 Palisades PWR 4 6 6 5 2 Palo Verde 1 PWR 2 0 1 1 1 Palo Verde 2 PWR 2 3 2 2 1 i Palo Verde 3 PWR 2 0 1 1 1 i Peach Boitom 2 BWR 6 3 0 2 3 Peach Bottom 3 BWR 5 4 1 1 1 Perry BWR 4 4 0 1 1 Pilgrim BWR 8 4 4 1 3 Point Beach 1 PWR 4 3 0 3 13 Point Beach 2 PWR 5 2 1 2 11 Prairie Island 1 PWR 0 1 2 1 Prairie Island 2 PWR 1 1 2 1 2 Quad Cities 1 BWR 8 7 4 5 6 Quad Cities 2 BWR 13 4 5 7 4 NUREG-1272, Vol,11, No.1 A-1 -16

Reactors Table A-1.6 Annual Safety System Failures Pirnt Name Rx Type CY93 CY94 CY95 CY96 1997 River Bend BWR 4 4 4 1 3 Robinson 1 PWR 2 3 0 1 2 Salem l PWR 2 4 11 9 3 Salem ? PWR 3 4 9 11 3 San Onofre 2 PWR I & I I 2

) San O iofre 3 PWR 2 0 3 0 1 Seabrook PWR 4 0 0 1 0 f

Sequoyah1 PWR PWR 4

4 0 0 0

0 0

I 0

Sequoyah 2 1 South Texas 1 PWR 5 4 0 0 0 South Texas 2 PWR 4 2 0 0 0 St. Lucie i PWR 1 1 2 3 0 St. Lucie 2 PWR 0 0 1 1 0 Summer PWR 1 0 0 0 0 Surry 1 PWR 2 0 1 1 1 Surry 2 PWR I O 2 0 1 Susquehanna l BWR 0 2 2 3 3 Susquchanna 2 BWR 0 3 1 2 3 Three Mile 1st 1 PWR 1 2 0 2 3 Turkey Point 3 PWR 2 2 3 1 0 Turkey Point 4 PWR I I 2 1 1 Vermont Yankee BWR 6 4 3 5 5 Vogtle 1 PWR I 1 1 3 2 Vogtle 2 PWR 2 0 2 1 1 Wash. Nuclear 2 BWR I1 3 0 1 1 "Waterford 3 PWR 3 1 1 4 6 Watts Bar 1 PWR NYL NYL 0 4 1 Wolf Creek PWR 3 2 1 5 1 Zivi. ! PWR 1 4 2 4 4 Zion 2 PWR I i 1 6 3 Total All Plants 353 219 213 339 299 Nurr'ser of Plants 109 109 110 110 110 Total BWR Plants 161 110 78 128 l19 Number of BWR Pir.rt: 37 37 37 37 37 Total PWR Plants 192 109 135 211 180 Number of PWR Plants 72 72 73 73 73 3

NYL means the plant was not yet licensed for low power operation. Calendar year values are shown for 1993 through 1996. Fiscal year values are used for 1997. Data for October through December 1996 are included in both niendar year 96 and fiscal year 97.

A- 1-17 Appendix A-1 1- -

1997 AEOD Annual Report Table A 1.7 Forced Outage Rate (Percent)-Quarterly PI Data Calendar Year-Quarter Plant Name 95 4 96-1 96-2 96-3 96-4 97-1 97-2 97 3 Arkansas 1 0 0 7 3 10 0 0 2 Arkansas 2 27 0 0 0 33 0 0 0 Beaver Valley I 8 0 4 19 -0 14 18 37 Beaver Valley 2 0 1 0 0 79 23 0 14 Big Rock Point 7 0 0 7 38 68 38 0 Braidwood 1 0 21 0 2 0 0 0 0 Braidwood 2 0 0 0 0 0 0 0 0 '

Browns Ferry 1 0 0 0 0 0 0 0 0 Browns Ferry 2 0 0 6 0 5 0 3 0 Browns Ferry 3 1 3 6 2 0 0 0 0 Brunswick 1 2 9 0 0 0 0 0 0 Brunswick 2 0 12 1 2 0 0 0 0 Byron 1 0 0 0 2 0 0 4 0 Byron 2 0 0 7 22 0 0 0 0 Callaway 10 1 1 0 8 0 0 0 Calvert Cliffs 1 7 0 100 36 0 1 4 0 Calvert Cliffs 2 0 7 0 0 3 0 0 0 Catawba 1 0 4 4 0 4 0 0 0 Catawba 2 27 13 0 10 8 0 2 4 Clinton 1 0 0 7 27 100 0 0 0 Comanche Peak i 2 18 3 1 0 0 0 0 Comanche Peak 2 4 1 27 3 1 0 0 0 Cook 1 22 2 0 8 0 0 0 24 Cook 2 0 0 1 0 0 8 7 24 Cooper Station 0 0 0 0 0 0 0 6 Crystal River 3 0 40 13 31 100 100 100 100 L) avis-Bekse 0 0 0 0 0 0 26 0 Diablo Canyon i 19 0 2 6 9 0 0 0 Diablo Canyon 2 6 0 0 7 0 3 6 3 Dresden 2 0 0 52 66 0 0 38 4 Dresden 3 25 0 41 92 72 35 11 0 Duane Arnold 0 0 0 0 0 0 0 5 Farley I 7 0 0 0 0 0 0 0 Farley 2 4 0 0 0 0 0 0 0 Fermi 2 0 5 24 0 100 96 39 0 FitzPatrick 0 15 0 8 9 6 7 0 Fort Calhoun 0 2 16 0 1 0 25 4 Ginna 0 4 0 15 21 0 0 0 Grand Gulf 0 0 3 0 17 0 0 0 Haddam Neck 0 0 0 53 0 PSD PSD PSD PSD means the plant was permanently shutdown.

NUREG-1272. Vol. I 1. No 1 A-l-18

Reactors

\

Table A 1.7 Forced Outage Rate (Percent)-Quarterly PI Data Calendar Year-Quarter Plant Name 95-4 96 1 96 2 96 3 96-4 97-1 97 2 97-3 ihrris 21 9 3 4 3 2 5 10 Hatch i 0 4 7 0 0 0 0 0 llatch 2 0 0 3 0 0 0 4 0 Hope Creek 0 0 0 0 0 0 0 3 Indian Point 2 4 4 2 4 0 53 0 19 Indian Point 3 100 100 9 0 1 7 5 26 Kewaunee 0 1 1 0 0 0 0 0 LaSalle 1 0 0 17 24 100 0 0 0 LaSalle 2 0 8 6 19 0 0 0 0 Limerick 1 0 0 3 9 0 0 0 0 Limerick 2 1 0 4 0 16 0 6 0 Maine Yankee 0 3 0 48 32 100 0 0 l McGuire 1 2 4 0 0 11 0 6 3 McGuire 2 9 0 77 2 13 0 16 16 Millstone 1 36 100 100 100 100 100 100 100 Millstone 2 6 37 100 100 100 100 100 100 Millstone 3 0 1 100 100 100 100 100 100 Monticello 0 0 18 0 5 0 57 33 Nine Mile Pt.1 0 0 3 14 7 3 29 20 Nine Mile Pt. 2 0 0 0 0 7 0 0 7 North Anna 1 0 0 0 2 1 0 0 0 North Anna 2 1 0 0 0 56' O O O Oconee1 0 2 0 0 0 7 31 3 Oconee 2 0 0 2 14 100 38 35 9 Oconee 3 0 11 0 0 0 72 41 9 Oyiter Creek 11 3 6 3 21 0 14 14 Palise ' :s 0 15 0 3 0 31 0 1 Palo %Me 1 5 8 0 2 0 0 5 0 Palo Veroe 2 0 3 0 0 0 0 0 0 Palo Verde 3 0 1 2 1 0 0 1 0

?cach Bottom 2 0 0 0 0 7 0 0 0 Peach Bot;om 3 5 2 5 0 0 3 0 0 Perry 8 0 17 0 0 4 17 1 Pilgrim 0 0 0 14 0 6 22 0 Point B each 1 0 0 0 0 0 47 100 100 P< int Beh 2 0 0 1 0 0 0 0 36 Pairie island 1 0 0 2 1 0 0 16 0

'trairie Island 2 0 2 3 1 0 48 1 1 Quad Cities 1 24 0 0 16 0 15 17 0 Quad Cities 2 37 0 57 49 5 0 6 4 A 19 Appendix A-1

1997 AEOD Annual Report Table A l.7 Forced Outage Rate (Percent)- Quarterly PI Data Calendar Year-Quarter Plant Name 95-4 96-1 96-2 96-3 96-4 97 1 97 2 97 3 River Bend 3 0 10 10 0 0 12 8 Robinson 2 0 0 0 0 1 0 0 0 Salem 1 0 0 100 100 100 100 100 100 Salem 2 100 100 100 100 100 100 100 65 San Onofre 2 2 0 0 0 0 0 1 17 San Onofre 3 0 0 0. 0 0 0 0 0 Seabrook 0 4 0 0 0 0 0 0 Sequoyah1 26 3 2 0 0 0 0 2 Sequoyah 2 4 0 0 0 30 0 0 1 South Texas 1 4 0 0 0 0 0 0 0 South Texas 2 4 3 0 0 0 5 3 0 St. Lucie 1 15 3 0 8 0 3 5 0 St. Lucie 2 0 2 9 0 0 0 0 0 Summer 0 0 0 0 0 0 6 0 Suny1 0 0 0 0 0 18 0 0 Surry 2 16 4 0 2 0 3 0 0 Susquehanna 1 0 0 0 6 8 22 0 0 Susquehanna 2 0 0 0 20 0 0 0 9 Three Mile 1st 1 0 0 0 0 0 0 8 0 Tbrkey Point 3 1 9 0 4 0 2 0 4 TLrkey Point 4 0 0 1 0 0 0 4 0 Verrnont Yankee 3 0 0 0 2 0 12 0

. Vogtle 1 0 0 24 0 0 1 16 0 Vogtle 2 0 0 0 0 3 0 0 0 Wash. Nuclear 2 0 0 53 0 0 1 0 0 Waterford 3 0 0 3 22 0 0 0 0 Watts Bar i NYC NYC 1 0 0 18 2 0 Wolf Creek 0 14 2 0 0 0 5 0 Zion 1 0 9 10 27 0 0 0 0 Zien 2 3 0 3 0 0 0 0 0 Average 5 6 10 12 14 13 15 10 NYC means the plant was not yet critical.

NUREG-1272, Vol. I 1 No.1 A. i -20

1 Reactors Table A.I.8 Equipment Forced Outages /1000 Commercial Critical Hours - Quarterly PI Data Calendar Year-Quarter Pirnt Name 95-4 96-1 96-2 96 3 96-4 97-1 97-2 97-3 Arkansas 1 0.00 0.00 0.49 0.57 0.60 0.00 0.00 0.45 Arkansas 2 3.93 0.00 0.00 0.00 0.67 0.00 0.00 0.00 Beaver Valley 1 0.00 0.00 0.00 0.55 0.00 1.62 0.00 0.74 Beaver Valley 2 0.00 0.00 0.00 0.00 9.72 1.78 0.00 0.00 Big Rock Point 0.96 0.00 0.00 0.48 0.72 1.43 0.00 0.00 Braidwood 1 0.00 0.00 0.00 0.46 0.00 0.00 0.00 0.00

' Braidwood 2 0.00 0.00 0.00 0,00 0.00 0.00 0.00 0.00 Browns Ferry 1 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Browns Ferry 2 0.00 0.00 0.00 0.00 0.48 0.00 0.00 0.00 Browns Ferry 3 0.00 0.47 0.96 0.49 0.00 0.00 0.00 0.00 Brunswick l 0.00 0.99 0.00 0.00 0.00 0.00 0.00 0.00 Brunswick 2 0.00 1.69 0.46 0.53 0.00 0.00 0.00 0.00 Byron 1 0.00 0.00 0.00 0.00 0.00 0.00 0.47 0.00 ,

Byron 2 0.00 0.00 0.00 I.08 0.00 0.00 0.00 0.00 Callaway 0.50 0.92 0.46 0.00 0.71 0.00 0.00 0.00 Calvert Cliffs 1 0.95 0.00 0.00 0.66 0.00 0.46 0.47 0.00 Calvert Cliffs 2 0.00 0.00 0.00 0.00 0.46 0.00 0.00 0.00 Catawba 1 0.00 0.00 0.59 0.00 0.47 0.00 0.00 0.00 Catawba 2 8.49 0.52 0.00 0.50 0.48 0.00 0.70 0.94 Clinton 1 0.49 0.00 0.49 0.62 0.00 0.00 0.00 0.00 Comanche Peak 1 0.46 1.11 0.47 0.00 0.00 0.00 0.00 0.00 Comanche Peak 2 0.47 0.78 0.76 0.00 0.00 0.00 0.00 0.00 Cook 1 0.00 0.46 0.00 0.00 0.00 0.00 0.00 0.00 Cook 2 0.00 0.00 0.00 " CD 0.00 0.49 0.00 0.00 Cooper Station 0.00 0.00 0.00 '0>. 0.00 0.00 0.00 0.48 Crystal River 3 0.00 1.45 0.00 v.>4 0.00 0.00 0.00 0.00 Davis-Besse 0.00 0.00 0.00 0.00 0.00 0.00 0.60 0.00 Diablo Canyon 1 2.58 0.00 0.00 0.00 0.49 0.00 0.00 0.00 Diablo Canyon 2 0.00 0.00 0.00 0.48 0.00 0.48 0.00 0.47 Dresden 2 0.00 0.00 2.02 0.00 0.00 0.00 1.40 0.00 Dresden 3 0.53 0.00 1.50 0.00 1.59 0.00 2.87 0.(X)

Duane Arnold 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.50 Farley 1 0.68 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Farley 2 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Fermi 2 0.00 0.00 0.00 0.00 37.57 6.12 0.00 0.0)

FitzPatrick 0.00 0.52 0.00 0.00 0.91 0.00 0.49 0.(X)

Fort Calhoun 0.00 0.51 0.54 0.00 1.02 0.00 1.18 0.47 Ginna 0.00 0.48 0.00 1.06 0.57 0.00 0.00 0.00 Grand Gulf 0.00 0.00 0.47 0.00 0.79 0.00 0.00 0.00 Haddam Neck 0.00 0.00 0.00 0.00 0.00 PSD PSD PSD PSD means the plant was permanently shutdown.

A- 1 -21 Appendix A-l

1997 AEOD Annual Report Table A-1.8 Equipment Forced Outages /1000 Commercial Critical Hours - Quarterly PI Data Calendar Year-Quarter Plant Name 95-4 96 1 96 2 96 3 96-4 97-1 97 2 97 3 Harris 1.35 0.00 0.94 0.48 0.46 0.47 0.00 0.49 Hatch 1 0.00 0.52 1.34 0.00 0.00 0.00 0.00 0.00 llatch 2 0.00 0.00 0.47 0.00 0.00 0.00 0.00 0.00 Hope Creek 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.58 Indian Point 2 0.92 0.00 0.00 0.93 0.00 0.98 0.00 1.73 Indian Point 3 0.00 0.00 0.48 0.00 0.00 0.70 0.95 2.20 Kewaunee 0.00 0.00 0.46 0.00 0.00 0.00 0.00 0.00 LaSalle 1 0.00 0.00 0.67 0.59 0.00 0.00 0.00 0.00 LaSalle 2 0.00 0.96 0.48 0.00 0.00 0.00 0.00 0.00 Limerick 1 0.00 0.00 0.93 0.98 0.00 0.00 0.00 0.00 Limerick 2 0.45 0.00 0.93 0.00 2.57 0.00 0.00 0.00 Maine Yankee 0.00 1.05 0.00 0.00 0.66 0.00 0.00 0.00 McGuire i 1.14 1.25 0.00 0.00 0.00 0.00 2.85 0.46 McGuire 2 0.49 0.00 6.47 0.00 0.00 0.00 0.55 1.05 Millstone 1 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Millstone 2 0.48 0.82 0.00 0.00 0.00 0.00 0.00 0.00 Millstone 3 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Monticello 0.00 0.00 2.74 0.00 1.40 0.00 0.00 0.00 Nine Mile Pt.1 - 0.00 0.00 0.49 1.55 0.49 0.69 4.25 1.12 Nine Mile Pt. 2 0.(X) 0.00 0.00 0.00 0.00 0.00 0.00 0.48 North Anna 1 0.00 0.00 0.00 0.46 0.46 0.00 0.00 0.00 North Anna 2 0.46 0.00 0.00 0.00 2.29 0.00 0.00 0.00 Oconee 1 0.00 0.47 0.00 0.00 0.00 0.94 0.66 0.00 Oconee 2 0.00 0.00 0.74 1.57 0.00 0.72 0.70 0.98 Oconee 3 0.00 0.51 0.00 0.00 0.00 3.43 1.33 0.47 Oyster Creek 0.50 0.00 0.00 0.63 1.39 0.00 0.52 0.00 Palisades 0.00 0.54 0.00 0.46 0.00 2.45 0.00 0.46 Palo Verde 1 0.95 0.00 0.00 0.00 0.66 0.00 0.48 0.00 Palo Verde 2 0.00 0.56 0.00 0.00 0.00 0.00 0.00 0.00 Palo Verde 3 0.00 0.00 0.92 0.00 0.00 0.00 0.00 0.00 Peach Bottom 2 0.00 0.00 0.00 0.00 1.39 0.00 0.00 0.00 Peach Bottom 3 0.00 0.46 0.48 0.00 0.00 0.48 0.00 0.00 Perry 0.48 0.00 1.73 0.00 0.00 0.48 0.54 0.57 Pilgrim 0.00 0.00 0.00 0.52 0.00 0.93 0.00 0.00 Point Beach 1 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Point Beach 2 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Prairie Island 1 0.00 0.00 0.00 0.00 0.00 0.00 0.55 0.00 Prairie Island 2 0.00 0.46 0.47 0.00 0.00 0.00 0.00 0.46 Quad Cities 1 0.00 0.00 0.00 2.29 0.00 1.06 0.54 0.00 Quad Cities 2 1.38 0.46 0.00 0.00 0.9 ) 0.00 4.I 1 0.47 NUREG-1272, Vol.11. No.1 A-I-22

Reactors l Table A 1.8 Equipment Forced Outages /1000 Commercial Critical Hours -Quarterly PI Data Calendar Year-Quarter PI:nt Name 95-4 96 1 96 2 96-3 96-4 97 1 97-2 97 3 River Bend 0.00 0.00 0.50 0.00 0.00 0.00 0.00 1.22 Robinson 2 0.00 0.00 0.00 0.00 0.57 0.00 0.00 0.00 Salem 1 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Salem 2 0.00 0.00 0.00 0 00 0.00 0.00 0.00 0.00 San Onofre 2 0.46 0.00 0.00 0.00 0.00 0.00 0.46 0.00 San Onofre 3 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Seabrook O.00 0.93 0.00 0.00 0.00 0.00 0.00 0.00 SequoyahI 3.78 0.00 0.46 0.00 0.00 0.00 0.00 0.46 l Sequoyah 2 0.92 0.00 0.00 0.00 1.24 0.00 0.00 0.45 South Texas 1 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 South Texas 2 0.64 0.00 0.00 0.00 0.00 1.18 0.93 0.00 St. Lucie 1 0.52 0.47 0.00 1.88 0.45 0.48 0.48 0.00 St. Lucie 2 0.00 0.47 1.49 0.00 0.00 0.00 0.00 0.00 l Summer 0.00 0.00 0.00 0.00 0.00 0.00 0.96 0.00 '

Surry 1 0.00 0.00 0.00 0.00 0.00 1.54 0.00 0.00 Surry 2 0.54 0.48 0.00 0.00 0.00 0.48 0.00 0.00 Susquehanna 1 0.00 0.00 0.00 0.00 0.62 0.00 0.00 0.00 Susquehanna 2 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.49 Three Mile Isl I 0.00 0.00 0.00 0.00 0.00 0.00 0.50 0.00 Tbrkey Point 3 0.49 0.99 0.00 0.47 0.00 0.69 0.00 1.85 Turkey Point 4 0.00 0.00 0.49 0.00 0.00 0.46 0.47 0.00 Vermont Yankee 0.46 0.00 0.00 0.00 1.34 0.00 0.00 0.00  ;

Vogtle 1 0.00 0.00 0.74 0.00 0.00 0.00 0.00 0.00 Vogtle 2 0.00 0.00 0.00 0.00 0.52 0.00 0.00 0.00 Wash. Nuclear 2 0.00 0.00 2.99 0.00 0.00 0.00 0.00 0.0 )

Waterford 3 0.00 0.00 0.47 0.58 0.00 0.00 0.00 0.00 Watts Bar 1 NYC NYC 0.00 0.00 0.00 1.65 0.47 0.00 Wolf Creek 0.00 1.43 0.50 0.00 0.00 0.00 0.47 0.00 Zion 1 0.00 0.66 0.47 1.23 0.00 0.00 0.00 0.00 Zion 2 0.46 0.00 0.47 0.00 0.00 0.00 0.00 0.00 Average 0.27 0.20 0.30 0.19 0.31 0.23 0.24 0.19 NYC means the plant was not yet critical.

A- 1 -23 Appendix A-1

1997 AEOD Annual Report l

Table A-1.9 Collective Radiation Exposure (Person-Centislevert [ Person-Rem])--Quarterly PI Data Calendar Year-Quarter l

Plant Name 95-4 96 1 96 2 96 3 96-4 97 1 97 2 97 3 1 1

Arkansas 1 2 4 4 83 63 3 3 2 '

Arkansas 2 103 3 2 1 38 4 104 2 Beaver Valley 1 4 56 207 9 3 17 19 25 Beaver Valley 2 1 2 1 128 45 1 1 10 Big Rock Point 14 172 25 30 9 12 1i PSD Braidwood 1 22 64 68 4 167 18 147 3 Braidwood 2 22 64 68 4 167 3 4 17 Browns Ferry i NLA NLA NLA NLA NLA NLA NLA NLA 1 Browns Ferry 2 27 126 133 15 22 15 17 83 Browns Ferry 3 57 12 12 21 16 75 10 12 Prunswick 1 38 49 16 34 128 20 19 118 Brunswick 2 38 308 40 13 128 20 19 118 Byron 1 43 3 122 96 228 9 6 6 Byron 2 43 3 122 96 228 9 6 6 Callaway 20 3 3 6 237 5 3 3 Calvert Cliffs 1 5 4 100 18 5 16 68 7 Calvert Cliffs 2 5 4 100 18 5 64 68 7 Catawba 1 75 5 32 109 5 15 47 5 Catawba 2 75 5 32 109 5 15 47 5 Clinton 1 23 9 21 20 307 50 93 70 Comanche Peak 1 1 53 8 4 75 2 2 2 Comanche Peak 2 1 53 8 4 75 2 2 2 Cooki 21 25 74 5 4 92 29 12 Cook 2 21 25 74 5 4 92 29 12 Cooper Station 174 10 22 12 11 26 136 7 Crystal River 3 3 257 69 16 11 22 32 85 Davis-Besse 1 2 155 1 1 1 7 1 Diablo Canyon i 141 52 80 3 3 3 99 2 Diablo Canyon 2 141 52 80 3 3 3 99 2 Dresden 2 95 79 54 51 226 30 119 16 Dresden 3 95 79 54 51 226 30 119 16 Duane Amold 25 17 13 16 153 18 12 13 Farley 1 74 2 1 2 111 48 87 2 Farley 2 74 2 1 2 111 48 87 2 Fermi 2 9 9 14 10 135 8 17 6 FitzPatrick 19 27 19 35 276 25 28 18 Fort Calhoun 2 13 24 11 188 5 13 14 Ginna 3 6 152 48 5 3 3 3 Grand Gulf 10 16 17 17 291 26 37 28 Haddam Neck 6 12 13 112 42 PSD PSD PSD NLA means the data were no longer available: PSD means the plant was permanently shutdown.

NUREG-1272, Vol. I 1, No.1 A-1-24

Reactors T ble A 1.9 Collective Radiation Exposure (Person-Centislevert [ Person-Rem])-Quarterly PI I)ata ,

Calendar Year-Quarter Pl:nt Name 95-4 96 1 96-2 96-3 96-4 97 1 97 2 97 3 Harris 9 4 7 5 4 5 135 5 llatch 1 164 76 102 19 24 97 71 33 llatch 2 164 76 102 19 24 97 71 33 Hope Creek 143 138 9 8 16 17 10 95 Indian Point 2 21 25 12 8 10 28 300 17 Indian Point 3 9 6 8 5 7 26 121 110 Kewaunee 2 2 2 29 89 23 29 2 LaSalle i 15 176 28 62 406 29 23 38 LaSalle 2 15 176 28 62 406 29 23 38 i

Limerick i 16 83 11 9 13 76 17 9 Limerick 2 16 83 11 9 13 76 17 9 Maine Yankee 150 17 7 21 12 49 81 PSD McGuire 1 43 35 76 3 2 76 40 9 McGuire 2 43 35 76 3 2 76 40 9 Millstone 1 552 180 156 39 41 79 82 5 Millstone 2 9 35 23 10 11 31 36 36 Millstone 3 55 4 18 15 12 14 17 27 Monticello 11 18 177 19 25 13 44 39 Nine Mile Pt. I 8 10 18 12 18 255 74 15 Nine Mile Pt. 2 15 16 13 25 175 13 21 16 North Anna 1 3 94 2 53 12 2 47 2 North Anna 2 3 94 2 53 12 2 47 2 (3conee 1 27 6 37 6 37 15 15 12 Oconee 2 27 6 37 6 36 15 15 12 Oconee 3 27 6 37 6 36 15 15 12 Oyster Creek 22 21 18 308 100 14 15 11  ;

Palisades 9 14 6 12 268 17 12 7 Palo Verde 1 4 4 17 32 90 5 4 3 Palo Verde 2 3 59 95 3 2 2 4 56 Palo Verde 3 177 7 4 4 6 139 8 5 Peach Bottom 2 81 18 29 160 23 25 32 24 Peach Bottom 3 81 18 29 20 23 25 32 29 i Perry 27 279 11 9 8 7 15 152 l Pilgrim 19 13 29 37 46 459 61 18 Point Beach 1 38 3 23 2 21 16 19 8 Point Beach 2 38 3 23 2 21 16 19 8 Prairie Island 1 3 52 52 2 2 47 1 1 Prairie Island 2 3 52 52 2 2 47 1 1 Quad Cities 1 42 274 185 28 52 123 152 25 Quad Cities 2 42 274 185 28 52 123 152 25 PSD means the plant was permanently shutdown.

A-1-25 Appendix A-1

1997 AEOD Annual Report Table A 1.9 Collective Radiation Exposure (Person-Centisievert [ Person-Rem])-Quarterly PI Data Calendar Year-Quarter i Plant Name 95-4 96 1 96-2 96-3 96-4 97 1 97-2 97 3 River Bend 28 335 35 24 13 13 43 159 Robinson 2 4 5 5 96 61 4 3 4 Salem I 41 62 13 21 120 82 23 19 Salem 2 23 57 42 11 15 10 10 8 San Onofre 2 6 2 3 5 55 53 92 17 San Onofre 3 6 2 3 5 55 53 92 17 Seabrook 92 4 2 1 3 4 169 4 Sequoyah1 149 13 11 3 5 77 165 3 Sequoyah 2 2 6 213 3 10 6 9 5 South Texas 1 2 1 190 4 1 4 3 111 South Texas 2 150 2 1 2 2 132 2 1 St. Lucie i 152 7 158 28 4 4 81 3 St. Lucie 2 152 7 158 28 4 4 82 3 Summer 2 12 91 2 2 1 2 9 Suny1 43 9 74 8 8 79 45 4 Suny 2 43 9 74 8 8 79 45 4  :

Susquehanna i 37 11 11 86 36 94 59 21 Susquehanna 2 37 11 11 86 36 94 59 21 Three Mile Isl 1 37 5 3 4 4 4 3 115 Turkey Point 3 8 80 6 4 4 70 12 96 ,

Turkey Point 4 8 80 6 4 4 70 12 96 Vermont Yankee 25 25 48 177 82 15 18 14 I Vogtle 1 4 88 41 84 59 4 4 33 l Vogtle 2 4 88 41 84 59 4 4 33 Wash. Nuclear 2 23 35 '299 14 19 19 200 14 Waterford 3 101 4 3 13 4 6 132 10 l Watts Bar i NYB NYB NYB NYB NYB 3 3 86 Wolf Creek 2 99 8 5 4 3 5 8 Zion 1 221 25 4 31 157 21 22 1I )

Zion 2 221 25 4 31 157 21 22 11 l Total 5292 5288 5355 3249 6907 4111 4988 2640 These data were obtained from the Institute of Nuclear Power Operations (INPO).

I NYB means the plant had not yet begun its first full calendar year of commercial operation.

NUREG-1272, Vol. I1 No.1 A- 1-26

Reactors 1

l Table A-1,10 Cause Codes- Administrative Control Problems--Quarterly PI Data Calendar Year-Quarter Plant Name 95-4 96-1 96-2 96-3 96-4 97 1 97 2 97-3 i 1

Arkansas 1 0 4 1 0 1 0 2 0 Arkansas 2 2 2 2 1 1 2 3 0 Beaver Valley 1 2 3 3 1 1 7 6 2 Beaver Valley 2 0 2 1 2 4 6 3 0 Big Rock Point 1 I 3 2 0 2 0 0 Braidwood 1 2 2 3 1 0 0 2 0 Braidwood 2 0 3 3 2 0 0 1 0 Browns Ferry 1 0 0 1 0 0 1 0 0 Browns Ferry 2 1 0 2 0 1 1 0 0 Browns Ferry 3 1 0 1 0 2 1 0 0 Brunswick i 2 1 2 1 2 3 2 i Brunswick 2 2 0 2 2 1 3 1 0 Byron 1 3 2 1 2 2 3 4 1 Byron 2 1 1 1 4 2 1 3 1 l

Callaway 0 0 0 0 3 2 1 0 i l

Calvert Cliffs 1 0 0 1 1 0 2 1 0 Calveit Cliffs 2 0 0 0 1 0 2 2 0 Catawba 1 0 1 3 1 3 0 2 i Catawba 2 0 1 2 1 3 1 1 1 l

Clinton 1 2 1 1 1 7 5 2 1 1

Comanche Peak 1 0 2 0 0 1 0 1 I l Comanche Peak 2 0 2 0 0 1 0 1 0 l l

Cook 1 2 1 0 1 0 2 3 1 Cook 2 0 1 0 0 0 1 1 1 Cooper Station 6 1 2 3 1 0 5 2 Crystal River 3 2 2 5 1 3 5 0 3 Davis-Besse 1 I 2 0 4 6 0 1 Diablo Canyon 1 1 1 2 0 1 5 1 2 Diablo Canyon 2 0 1 2 0 1 4 0 0 Dresden 2 0 2 2 1 4 4 3 2 Dresden 3 2 1 5 5 6 3 5 0 Duane Arnold 0 0 0 0 1 4 2 1 Farley l 2 0 O O 3 2 1 Farley 2 0 0 , 0 0 4 3 1 Fermi 2 0 2 1 1 7 5 3 0 FitzPatrick 0 1 1 1 1 1 1 1 Fort Calhoun 1 2 2 0 6 0 6 1 Ginna 0 1 1 2 1 0 0 1 Grand Gulf 1 1 0 0 2 1 0 0 Haddam Neck 2 4 3 6 2 PSD PSD PSD PSD means the plant was permanently shutdown.

' A-1-27 Appendix A-1

1997 AEOD Annual Report Table A.I.10 Cause Codes - Administrathe Control Problems-Quarterly PI Data l Ca!cadar Year-Quarter Plant Name 95-4 96 1 96 2 96 3 96-4 97-1 97-2 97 3 liarris 6 3 4 3 2 2 5 1 Ilatch I O 1 3 1 2 3 1 0 llatch 2 1 0 1 1 2 5 2 0 llope Creek 11 8 3 2 4 2 2 2 Indian Point 2 0 3 0 2 2 3 6 2 Indian Point 3 1 2 2 1 0 1 5 0 Kewaunee 1 1 0 0 4 1 1 0 LaSalle 1 2 1 4 4 5 11 11 3 LaSalle 2 1 1 2 4 6 10 10 2 Limerick i O 3 3 2 1 2 0 0 Limerick 2 0 2 2 1 4 3 0 0 Maine Yankee 1 0 5 4 5 3 1 0 McGuire 1 0 0 0 0 2 2 1 0 McGuire 2 1 1 1 0 0 1 2 0 Millstone 1 4 12 10 0 6 13 5 3 Millstone 2 6 10 4 1 11 7 10 2 Millstone 3 3 3 5 7 11 21 10 3 Monticcilo 1 0 1 0 2 2 2 0 Nine Mile Pt.1 0 2 2 2 2 0 3 0 Nine Mile Pt. 2 0 2 3 1 4 1 0 t

{ 0 0 0 0 0 l North Anna 1 1 1 1

! Nonh Anna 2 1 0 0 1 1 1 0 0 Oconee 1 0 1 1 0 0 1 1 0 Oconee 2 0 0 2 2 1 1 2 1 Oconee 3 0 1 1 0 1 1 2 0 Oyster Creek 2 3 2 1 1 2 2 0 l Palisades 0 1 0 3 1 2 1 0 I- Palo Verde 1 2 0 1 1 1 0 1 0 Palo Verde 2 1 1 2 0 1 0 0 1 Palo Verde 3 1 0 1 0 1 0 0 0 l l

Peach Bottom 2 0 1 0 0 1 2 0 2 I Peach Bottom 3 1 1 0 0 0 2 1 2 Perry 0 0 1 1 0 1 1 1 Pilgrim 2 0 2 1 1 3 1 0 Point Beach ! 0 0 3 2 3 8 7 0 Point Beach 2 1 0 0 1 2 7 8 0 Prairie Island 1 0 4 i 1 2 2 5 0 Prairie Island 2 0 2 0 1 2 2 3 0 Quad Cities 1 1 3 2 7 2 2 3 1 Quad Cities 2 1 1 2 5 3 3 2 2 NUREG 1272 Vol. I1, No.1 A-1-28

Reactors ,

Table A 1.10 Cause Codes- Administrative Control Problems --Quarterly Pl Data Calendar Year-Quarter PI:nt Name 95 4 96-1 96-2 96 3 96-4 97-1 97 2 97 3 l

River Bend 2 2 2 2 1 0 1 1 Robinson 2 1 2 0 1 1 3 2 1 Salem i I 2 3 11 12 4 4 2 i Salem 2 3 3 4 11 14 5 5 3 San Onofre 2 1 0 1 1 3 2 1 i San Onofre 3 1 0 1 0 2 2 1 1 Scabrook 2 1 2 2 0 2 3 0 Sequoyah1 3 1 0 2 1 5 2 0 Sequoyah 2 1 1 1 3 1 5 1 0 South Texas 1 0 0 0 0 2 2 2 1 South Texas 2 0 1 0 1 0 2 2 i St. Lucie 1 0 3 4 3 1 1 2 1 St. Lucie 2 0 0 1 1 3 1 5 i Summer 0 1 0 1 1 0 1 0 Surry i I O 4 0 1 2 0 0 Suny 2 1 0 3 0 1 2 0 0 Susquehanna 1 4 0 1 3 4 6 2 4 Susquehanna 2 3 0 1 5 2 5 2 3 Three Mile Isl 1 1 0 0 0 1 2 1 0 hrkey Point 3 0 2 1 1 1 1 1 0 Turkey Point 4 0 2 2 1 0 0 1 0 Vennont Yankee 3 6 3 6 3 2 3 0 Vogtle 1 2 0 3 0 2 1 0 1 Vogtle 2 0 0 2 1 3 2 0 0 Wash. Nuclear 2 0 0 2 1 3 3 1 1 Waterford 3 2 3 1 2 0 5 6 1 l' Watts Bar 1 3 5 3 2 2 6 3 1 Wolf Creek 0 2 2 2 11 1 7 3 i Ziont 3 4 6 3 2 5 4 2 Zion 2 2 2 6 3 3 2 4 2 Total 131 167 201 177 256 303 257 87 A- 1 -29 Appendix A-1

1997 AEOD Annual Report Table A 1.11 Cause Codes - Licensed Operator Errors -Quarterly PI Data Calendar Year-Quarter Plant Name 95-4 96-1 96 2 96 3 96-4 97 1 97 2 97 3 Arkansas 1 0 0 0 0 0 0 0 0 Arkansas 2 0 1 0 0 0 0 1 0 Beaver Valley 1 1 0 0 0 1 0 0 0 Beaver Valley 2 0 0 0 0 1 0 1 0 Big Rock Point 1 0 1 0 0 0 0 0 Braidwood 1 2 0 0 0 0 0 0 0 Braidwood 2 1 0 1 0 0 0 0 0 Browns Ferry 1 0 0 0 0 0 0 0 0 Browns Ferry 2 0 1 0 0 0 0 0 0 Browns Ferry 3 1 0 0 0 0 0 1 0 Brunswick 1 0 0 0 0 0 0 0 0 Brunswick 2 0 0 0 1 0 0 0 0 Byron 1 0 0 0 0 0 0 0 0 Byron 2 0 0 0 1 0 1 0 0 Callaway 0 0 0 0 1 0 0 0 Calvert Cliffs 1 0 0 0 0 0 0 1 1 Calvert Cliffs 2 0 0 0 0 0 0 1 1 Catawba 1 0 0 0 0 0 0 0 0 Catawba 2 1 0 0 0 1 0 0 0 Clinton 1 0 0 1 0 0 1 0 0 Comanche Peak l 0 1 0 0 0 0 0 0 Comanche Peak 2 0 0 0 0 0 0 0 0 Cook 1 0 0 0 0 0 0 0 0 Cook 2 0 9 2 0 0 0 0 0 Cooper Station 2 0 0 0 0 0 1 1 Crystal River 3 0 2 1 0 0 0 0 0 Davis-Besse 0 0 0 0 0 1 0 0 Diablo Canyon i 2 1 2 0 0 0 0 0 Diablo Canyon 2 0 1 1 0 0 0 0 0 Dresden 2 1 1 1 1 0 0 0 3 Dresden 3 0 0 1 2 0 1 0 0 Duane Arnold 0 0 0 0 0 0 0 0 Farley 1 0 1 0 0 0 0 0 1 Farley 2 0 0 0 0 0 0 0 0 Fermi 2 0 1 0 1 2 0 0 0 FitzPatrick 0 0 0 1 0 0 0 0 Fort Calhoun 0 0 0 0 2 0 2 0 Ginna 0 0 0 0 0 0 0 0 Grand Gulf 0 0 0 0 1 0 1 0 liaddam Neck 0 0 0 2 0 PSD PSD PSD PSD means the plant was permanently shutdown.

NUREG-1272. Vol. I 1, No.1 A- 1 -30

Reactors .

Table A 1.11 Cause Codes- Licensed Operator Errors-Quarterly PI Data Calendar Year-Quarter PI:nt Name 95-4 96-1 %2 96 3 96-4 97 1 97 2 97-3 Harris . 3 1 2 2 0 1 2 0

}{atch 1 0 1 1 0 0 0 1 0 Hatch 2 I I O O O O 1 0 Hope Creek 3 0 1 0 1 0 0 1 Indian Point 2 0 0 1 0 1 0 1 0 Indian Point 3 1 2 0 0 0 0 0 0 Kewaunee 0 1 0 1 1 0 1 0 LaSalle 1 0 0 1 1 0 0 2 0 LaSalle 2 1 0 0 1 0 0 1 0 ,

Limerick l 0 1 0 0 0 0 0 0 Limerick 2 0 0 0 0 1 0 1 1 Maine Yankee 1 1 0 0 0 0 0 0 McGuire 1 0 1 0 0 0 0 0 0 McGuire 2 0 0 0 0 0 0 0 0 Millstone 1 3 0 0 0 0 0 0 0 Millstone 2 2 2 0 0 0 0 0 1 Millstone 3 1 1 0 0 0 0 0 0 Monticello 0 0 0 0 1 0 0 0 Nine Mile Pt.1 0 0 0 1 1 0 0 0

- Nine Mile Pt. 2 0 0 2 0 0 0 1 0 Nonh Anna l 0 1 0 0 0 0 1 0 Nonh Anna 2 0 0 0 0 0 0 0 0 Oconee1 0 1 0 0 0 0 0 0 Oconee 2 0 1 1 0 0 0 0 0 ,

Oconee 3 0 2 0 0 0 0 1 0 i

Oyster Creek. 0 0 1 0 0 0 0 0 i Palisades 0 2 0 1 0 1 0 0 Palo Verde 1 0 0 0 0 0 0 0 0 Palo Verde 2 0 1 2 1 0 0 0 0 Palo Verde 3 0 0 0 0 0 0 0 0 Peach Bottom 2 0 0 0 0 1 0 0 0 Peach Bottom 3 0 0 0 0 0 0 0 0 Peg 0 1 1 0 0 0 0 0 Pilgrim 0 0 0 0 0 1 0 0 Point Beach 1 0 0 1 0 0 0 0 0 Point Beach 2 0 0 0 0 0 0 0 0 Prairie Island 1 0 0 0 2 0 0 1 0 Prairie Island 2 0 0 0 1 1 0 0 0 0

Quad Cities 1 0 0 0 1 0 0 1 Quad Cities 2 0 0 0 1 0 0 1 0 A-1-31 Appendix A-1

i 1997 AEOD Annual Report l

l Table A 1.11 Cause Codes- Licensed Operator Errors--Quarterly PI Data Calendar Year-Quarter Plant Name 95-4  %-1 96 2 96-3 96-4 97-1 97 2 97 3 l River Bend 0 0 0 0 0 0 0 0 Robinson 2 - 1 1 0 0 0 1 0 0 <

Salem i 0 0 0 0 0 0 1 0 Salem 2 0 0 0 0 0 0 0 0 San Onofre 2 0 1 0 1 0 1 0 1 San Onofre 3 0 1 0 2 0 0 0 l Seabrook 0 0 0 0 0 0 1 0 Sequoyah1 3 0 0 0 0 0 0 0 Sequoyah 2 1 0 0 0 0 0 0 0 South Texas 1 2 0 0 0 0 0 0 0 South Texas 2 1 0 0 0 0 0 0 0 St. Lucie i I O 1 2 0 0 0 0 St. Lucie 2 0 0 0 0 0 0 0 0 Summer 0 0 0 0 0 0 1 0 Suny1 0 1 0 0 0 0 1 0 Surry 2 0 0 0 0 1 0 0 0 Susquehanna 1 0 0 0 2 1 0 0 0 Susquehanna 2 0 0 0 2 0 0 0 0 Three Mile Isl 1 0 0 0 0 0 0 0 0 Turkey Point 3 0 1 0 0 1 0 0 0 Turkey Point 4 0 0 0 0 0 0 0 0 '

Vermont Yankee 1 0 0 0 0 0 0 0 Vogtle i 1 1 0 0 0 0 0 0 ,

Vogtle 2 0 1 0 1 0 0 0 0 Wash. Nuclear 2 0 0 0 0 0 0 0 0 Waterford 3 0 0 0 0 1 0 1 0 Watts Bar i 1 2 2 1 0 2 0 0 Wolf Creek 0 3 0 0 1 0 0 0 Zion 1 0 1 0 0 0 1 1 0  ;

Zion 2 2 1 1 I O O 1 0 Total 42 45 29 34 22 12 32 12 NUREG-1272, Vol. I 1, No. I A-1 -32

Reactors i

l Table A l.12 Cause Codes- Other Personnel Errors --Quarterly PI Data Calendar Year-Quarter Pirnt Name 95-4 96 1 96-2 96-3 96-4 97 1 97 2 97-3 Arkansas 1 0 0 1 0 0 0 0 0 i Arkansas 2 1 0 0 0 1 2 0 0 1 Beaver Valley 1 0 2 0 0 2 1 1 1 Beaver Valley 2 1 2 0 0 2 1 1 0 Big Rock Point i 1 1 1 1 0 0 0 Braidwood 1 0 1 0 1 1 0 0 0 Braidwood 2 0 0 2 0 1 0 0 0 Browns Ferry 1 0 1 0 0 0 0 0 0 Browns Ferry 2 0 1 0 0 0 0 1 0 Browns Ferry 3 2 0 1 0 0 1 1 0 Brunswick l 0 0 l 1 1 0 0 0 Brunswick 2 1 0 0 0 0 0 0 0 Byron i 1 0 1 1 1 1 1 0 Byron 2 0 1 1 1 1 0 0 0 Callaway 0 0 0 0 1 1 1 0 Calvert Cliffs 1 0 1 0 1 0 1 1 0

' Calvert Cliffs 2 0 1 2 0 1 1 3 0 Catawba 1 0 0 1 0 0 0 0 0 Catawba 2 0 1 1 0 0 1 1 0 Clinton 1 0 0 3 0 3 0 1 0 Comanche Peak 1 0 1 0 0 0 0 0 0 Comanche Peak 2 0 3 1 0 0 0 0 0 Cook 1 0 0 0 2 2 1 2 0 Cook 2 -0 2 0 0 1 0 1 0 Cooper Station 1 0 2 0 0 0 2 0 Crystal River 3 0 3 4 0 0 1 0 1 Davis-Besse 0 0 1 0 0 0 0 1 Diablo Canyon 1 1 1 2 1 0 2 1 i Diablo Canyon 2 1 0 3 1 0 2 1 1 3 1 0 1 1 0 Dresden 2 1 1 0 4 3 0 1 5 0 Dresden 3 1 0 0 0 2 0 0 0 Duane Arnold 1 Farley 1 0 0 1 1 0 3 0 1 0 0 1 4 0 0 Farley 2 1 1 0 0 1 2 0 1 0 Fermi 2 1 FitzPatrick 0 1 0 2 2 0 3 1 i 0 3 0 0 0 Fort Calhoun 1 1 0 0 3 1 0 0 0 Ginna 1 0 0 0 0 2 0 0 0 Grand Gulf 0 2 0 3 i PSD PSD PSD Haddam Neck PSD means the plant was permanently shutdown.

A-1-33 Appendix A-1

i 1997 AEOD Annual Report i

Table A-1.Il Cause Codes - Other Personnel Errors --Quarterly PI Data Calendar Year-Quarter Plant Name 95-4 96-1 96-2 96 3 96-4 97 1 97-1 97 3 i l

i liarris 1 2 0 1 1 0 3 0 11atch 1 2 0 0 0 0 2 0 0 liatch 2 2 0 0 2 0 1 0 0 I Ilope Creek 3 2 0 2 0 2 2 1 i Indian Point 2 2 0 1 0 1 3 0 2 Indian Point 3 0 1 0 0 0 0 4 0 Kewaunee 0 0 0 0 0 1 0 0  ;

LaSalle 1 0 2 4 1 1 2 1 0 LaSalle 2 2 1 2 3 1 2 1 1 Limerick l 0 6 1 2 2 0 0 0 ,

Limerick 2 0 3 0 1 1 0 2 0  !

Maine Yankee 0 0 3 3 2 1 0 0 McGuire 1 0 0 0 0 0 0 2 0 McGuire 2 0 0 0 0 0 0 0 0 Millstone 1 4 4 3 0 0 1 1 0 Millstone 2 0 4 0 1 1 I i 0 Millstone 3 0 0 4 3 0 5 3 0 Monticello 0 0 1 1 1 0 1 0 Nine Mile Pt.1 0 0 1 1 1 0 0 0 Nine Mile Pt. 2 1 1 0 1 1 0 1 0 North Anna 1 0 0 0 0 0 0 2 0 North Anna 2 1 0 0 1 0 0 0 0 Oconee 1 0 1 0 1 0 0 0 0 Oconee 2 0 1 0 0 2 0 0 0 Oconee 3 0 1 0 1 2 0 2 0 Oyster Creek 0 1 1 0 3 1 0 1 Palisades 1 0 0 0 1 0 1 0 Palo Verde 1 1 0 0 1 0 0 0 0 Palo Verde 2 0 0 0 0 0 0 0 0  ?

Palo Veide 3 0 0 1 0 0 0 0 0 Peach Bottom 2 0 1 2 0 0 0 1 0 Peach Bottom 3 1 1 2 0 0 0 0 0 Perry 0 0 0 0 0 1 1 0 Pilgrim 0 0 2 0 0 1 1 0 Point Beach 1 0 0 0 0 1 1 2 0 Point Beach 2 2 0 0 0 0 1 2 0 4 Prairie Island 1 0 2 1 1 0 1 1 0 '

Prairie Island 2 0 0 0 1 1 3 1 0 Quad Cities 1 0 0 0 1 1 2 1 0 Quad Cities 2 0 0 0 0 1 2 0 1 NUREG-1272, Vol. I 1 No.1 A-1 -34

= . _ - _ . - - .-_ - _ _ . - -

Reactors Table A-1.11 Cause Codes - Other Personnel Errors -Quarterly PI Data Calendar Year-Quarter PI:nt Name 95 4 96 1 96-2 96-3 96-4 97 1 97 2 97-3 River Bend i 4 1 1 1 0 0 0 Robinson 2 0 1 0 0 1 1 0 1 Salem 1 0 1 0 2 3 1 2 0 Salem 2 0 2 2 1 3 1 2 0 San Onofre 2 0 1 1 0 l 1 0 0 San Onofre 3 0 2 1 0 0 1 0 0 1 Seabrook 0 0 0 0 1 0 1 l l Sequoyah1 0 3 0 1 0 1 1 1 l 0 2 1 2 0 2 0 l Sequoyah 2 1 0 2 0 1 0 0 0 j South Texas 1 1 0 0 1 0 1 0 0 South Texas 2 1 0 2 0 1 1 1 0 0 l St. Lucie 1 0 0 0 0 0 1 1 0 l St. Lucie 2 0 0 1 0 0 0 l Summer 1 1 0 1 1 0 0 0  ;

Surry 1 1 1 0 0 0 2 1 0 0 0 Surry 2 Susquehanna1 0 1 0 6 3 6 1 1 Susquehanna 2 0 1 1 3 1 6 0 1 0 0 0 0 1 0 0 0 Three Mile Isl 1 0 2 0 0 0 1 1 0 Turkey Point 3 i 0 0 0 0 1 0 0 l

'Ibrkey Point 4 1 2 0 0 2 1 0 1 0 ,

Vermont Yankee 0 0 2 3 0 0 Vogtle 1 1 1 0 0 0 1 2 2 0 0 Vogtle 2 0 0 0 0 1 0 0 l Wash. Nuclear 2 1 2 0 1 0 1 0 0 Waterford 3 1 2 2 2 0 1 3 0 0 l Watts Bar 1 0 1 1 0 2 0 l Wolf Creek 1 1 0 3 1 3 0 1 0 1 Eion 1 0 2 0 2 1 1 0 1 Zion 2 4 1

47 95 87 83 88 94 80 20 Total l

l

.i A-1-35 Appendix A-1

1997 AEOD Annual Report Table A 1.13 Cause Codes - Maintenance Problems-Quarterly PI Data Calendar Year-Quarter Plant Name 95-4 96-1 96-2 96 3 96-4 97 1 97-2 97-3 Arkansas 1 0 3 2 1 1 0 1 0 Arkansas 2 3 1 2 2 2 2 2 0 Beaver Valley 1 3 3 3 0 0 9 3 4 Beaver Valley 2 2 3 1 1 3 8 3 0 Big Rock Point 1 4 1 2 2 2 0 0 Braidwood I .5 4 1 2 1 0 4 0 Braidwood 2 0 2 4 3 1 0 2 0 Browns Ferry 1 I i 1 0 0 1 0 0 Browns Ferry 2 2 1 2 0 2 1 1 1 Browns Ferry 3 3 0 1 I 3 3 2 0 Brunswick 1 3 2 3 2 3 3 1 i Brunswick 2 3 0 2 3 0 3 1 0 Byron i 4 1 2 2 2 3 4 1 Byron 2 0 2 1 3 2 1 3 1 Callaway 1 0 0 0 4 2 2 1 Calvert Cliffs 1 2 2 0 1 0 2 0 0 Calvert Cliffs 2 0 1 1 0 2 2 2 0 Catawba 1 0 1 3 3 4 0 2 i Catawba 2 1 2 2 2 3 2 3 1 Clinton 1 0 2 5 2 7 5 2 1 Comanche Peak 1 0 3 0 0 1 0 1 1 Comanche Peak 2 1 3 1 1 1 0 1 0 Cook 1 1 1 0 2 1 3 3 0 Cook 2 0 4 1 0 1 1 2 0 Cooper Station 8 1 3 3 0 0 3 2 Crystal River 3 2 4 7 1 2 4 1 2 Davis-Besse 1 1 1 0 3 5 0 1 Diablo Canyon i 1 4 3 1 1 3 1 2 Diablo Canyon 2 0 3 5 2 1 4 0 1 Dresden 2 1 3 4 3 5 4 3 0 Dresden 3 3 1 3 5 5 4 6 0 Duane Arnold 2 0 1 0 4 4 2 1 Farley 1 1 1 3 0 0 2 2 2 Farley 2 1 0 3 0 1 3 2 1 Fermi 2 0 1 2 1 9 5 4 0 FitzPatrick 0 3 1 2 4 0 2 1 Fort Calhoun 0 1 1 0 6 0 3 1 Ginna 0 3 3 3 0 0 0 1 Grand Gulf 1 1 0 0 3 1 0 0 Haddam Neck 3 2 1 7 3 PSD PSD PSD PSD means the plant was permanently shutdown.

NUREG-1272. Vol. I 1. No. I A- 1 -36

1 Reactors Table A 1.13 Cause Codes- Maintenance Problems---Quarterly PI Data Calendar Year-Quarter Pirnt Name 95-4 96 1 96 2 96-3 96-4 97-1 97-2 97-3 Harris 5 2 5 5 3 1 7 i Hatch 1 2 1 4 1 3 3 1 0 Hatch 2 6 0 1 3 2- 4 2 1 Hope Creek 12 7 3 3 4 3 5 2 Indian Point 2 2 2 2 3 4 6 5 3 Indian Point 3 1 2 2 1 1 1 5 2 Kewaunee 2 2 1 2 4 1 1 0 LaSalle 1 2 3 4 3 6 9 10 2 LaSalle 2 2 1 2 7 7 8 9 1 Limerick 1 2 5 3 3 2 3 0 0 Limerick 2 1 4 3 1 2 3 2 0 Maine Yankee 1 2 3 6 7 2 2 0 McGuire 1 0 1 1 0 4 1 1 0 McGuire 2 0 1 1 0 2 0 1 1 Millstone 1 11 10 10 0 5 14 4 3 4

Millstone 2 5 7 3 1 9 6 8 3 Millstone 3 4 3 8 8 7 16 9 3 Monticello 1 1 3 1 3 0 3 0 Nine Mile Pt.1 O 1 0 3 2 0 1 0 Nine Mile Pt. 2 1 2 3 1 2 1 0 1 North Anna 1 0 2 0 1 2 1 3 0 North Anna 2 2 0 0 2 2 1 0 0 Oconee 1 1 1 1 0 0 0 1 0 Oconee 2 0 0 1 2 2 0 2 i Oconee 3 0 0 1 1 2 0 4 0 Oyster Creek 3 3 2 1 6 2 3 i Palisades 1 2 0 2 2 2 1 0 Palo Verde 1 4 0 0 1 1 0 1 0 Palo Verde 2 2 1 1 1 1 1 0 2 Palo Verde 3 1 0 1 0 1 1 0 0 Peach Bottom 2 1 1 2 0 1 1 0 2 -.

Peach Bottom 3 5 1 1 0 0 2 0 2 Perry 0 0 1 1 0 0 3 1 Pilgrim 2 1 4 0 1 6 1 1 Point Beach 1 0 0 2 2 4 7 2 0 Point Beach 2 2 0 1 1 4 6 2 2 Prairie Island 1 0 6 1 3 2 2 3 0 Prairie Island 2 0 2 1 2 4 2 1 0 Quad Cities 1 2 2 1 5 2 3 6 1 Quad Cities 2 3 0 1 4 4 3 4 1 A-1-37 Appendix A-1 f

1997 AEOD Annual Report Table A-1.13 Cause Codes-M enance Problems-Quarterly PI D:na t.'alendar Year-Quarter Plant Name 95-4 96-1 96 2 96-3 96-4 97-1 97-2 97-3 River Bend 5 5 2 2 0 1 2 i Robinson 2 1 1 0 1 1 3 '

1 1 Salem i 1 3 2 7 13 4 3 2 Salem 2 2 4 o 7 16 5 4 3 San Onofre 2 1 1 1 1 3 4 1 0 San Onofre 3 1 2 2 2 2 3 2 1 Seabrook 3 1 1  ?. I 2 3 1 Sequoyah1 3 3 3 1 2 7 3 1 Sequoyah 2 1 2 3 2 3 6 3 0 South Texas 1 2 0 2 0  ! 2 3 1 South Texas 2 1 1 0 1 0 6 3 1 St. Lucie 1 2 4 4 5 2 3 2 0 St. Lucie 2 2 1 1 3 1 1 5 0 '

Summer 0 2 2 0 1 1 1 0 Suny1 3 0 4 1 1 3 1 0 Suny 2 3 1 4 0 1 2 0 Susquehanna 1 4 0 1 5 9 S 2 4 Susquehanna 2 5 1 0 0 3 10 1 3 Three Mile Isl 1 1 0 0 0 1 2 3 3 Turkey Point 3 0 5 1 1 1 1 1 0 Turkey Point 4 0 2 2 0 1 1 1 0 {

Vermont Yankee 2 4 1 6 4 0 3 0 l Vogtle 1 1 0 3 1 3 4 0 1 Vogtle 2 0 0 4 4 0 1 3 0 Wash. Nuclear 2 0 0 3 1 1 2 1 2 Waterford 3 4 3 1 3 1 4 5 i 1

Watts Bar 1 5 7 3 1 3 7 4 0 Wolf Creek 1 4 1 2 10 1 6 i I Zion 1 3 6 4 4 1 7 3 4 Zion 2 2 5 5 5 2 4 3 4 Total 208 222 227 223 303 320 M7 97 i 1

t l

NUREG-1272, Vol. I 1, No. I A-1-38

Reactors ,

Table A-1.14 Cause Codes - Design / Construction / Installation / Fabrication - Quarterly PI Data l Calendar Year-Quarter Plant Name 95 4  %-1  %-2 96-3 96-4 97 1 97 2 97 3 Arkansas 1 0 0 1 0 1 0 0 0 Arkansas 2 0 0 0 0 0 0 1 0 Beaver Valley 1 0 0 2 1 0 3 3 4 Beaver Valley 2 0 0 1 0 2 3 1 1 Big Rock Point 0 0 0 1 0 1 1 0 Braidwood i 1 0 0 2 1 0 0 0 Braidwood 2 1 1 0 2 0 0 0 0 Browns Ferry 1 0 0 0 0 0 1 0 0 Browns Ferry 2 0 1 0 0 0 1 0 0 Browns Ferry 3 0 1 0 0 0 2 0 0 Brunswick 1 1 2 0 0 0 0 1 0 Brunswick 2 1 2 0 1 0 0 1 1 Byron i 1 0 0 2 1 1 1 0 Byron 2 1 0 0 0 1 2 1 0 Callaway 0 0 0 0 0 1 1 0 Calvert Cliffs 1 1 0 1 0 0 0 2 0 i Calvert Cliffs 2 0 0 1 0 0 0 1 0 Catawba 1 0 0 1 0 2 1 1 0 Catawba 2 0 0 1 0 2 1 1 0 ,

Clinton 1 1 0 0 0 1 3 2 0 l Comanche Peak 1 1 1 0 0 0 2 1 0 Comanche Peak 2 0 1 0 0 0 2 2 0 '

0 0 0 0 4 0 0 I Cook 1 1 Cook 2 0 1 1 0 0 4 0 0 Cooper Station 2 2 1 2 2 1 1 0 Crystal River 3 6 5 3 0 2 5 6 7 Davis-Besse 0 2 1 0 2 4 1 0 Diablo Canyon 1 5 1 0 0 1 2 1 0 Diablo Canyon 2 1 1 0 0 0 2 0 0 Dresden 2 2 0 2 0 5 2 1 0 Dresden 3 3 1 2 0 6 3 1 0 Duane Arnold 0 0 0 0 0 0 0 0 Farley 1 '

0 0 1 0 2 2 3 0 Farley 2 1 0 1 0 2 2 2 0

)

Fermi 2 1 1 1 0 2 5 0 0 l l ,

FitzPatrick 1 2 0 1 1 2 1 0 i Fort Calhoun 2 0 2 0 2 1 3 1 Ginna 0 0 0 1 2 1 0 0 .

Grand Gulf i 1 1 0 0 0 1 0

, Haddam Neck 1 1 3 7 2 NA NA NA PSD means the plant was permanently shutdown.

A-1-39 6.poendix A-1 I

1997 AEOD Annual Report Table A 1.14 Cause Codes - Design / Construction / Installation / Fabrication - Quarterly PI Data Calendar Year-Quarter

! Plant Name 95-4  %-1  %-2 96-3 96-4 97-1 97-2 97-3 f

Harris 1 2 0 1 2 3 1 0 l Hatch ! , 0 0 2 0 2 1 0 1 i Hatch 2 2 0 1 0 2 2 0 0 l Hope Creek 1 3 2 0 2 3 1 0 l Indian Point 2 0 1 0 3 1 0 1 0 Indian Point 3 1 3 1 1 0 1 3 0 Kewaunee 0 0 0 0 2 1 3 0 LaSalle 1 0 0 1 0 5 5 3 0 LaSalle 2 0 0 0 1 5 4 3 0 Limerick I I 2 0 0 1 0 0 0 Limerick 2 0 1 1 0 1 1 0 0  !

l Maine Yankee 2 1 5 8 4 4 1 0 McGuire 1 0 0 1 0 0 2 0 0 McGuire 2 0 1 1 0 0 2 0 1 Millstone 1 2 14 8 3 7 9 3 1 Millstone 2 2 10 4 1 5 2 5 1 Millstone 3 0 2 8 9 7 8 6 i Monticello 0 0 4 0 1 1 0 0 Nine Mile Pt.1 1 1 2 0 3 2 3 0 -

Nine Mile Pt. 2 2 0 1 1 2 0 1 0 North Anna 1 0 0 1 1 1 1 0 0 Nonh Anna 2 0 0 1 1 1 1 0 0 ,

! Oconee ! 2 2 1 0 0 2 1 0

()conee 2 0 1 1 1 1 2 1 0 Oconee 3 0 2 1 0 1 3 2 0 .

Oyster Creek 0 0 1 0 2 1 0 0 Palisades 2 2 0 2 1 0 0 0 Palo Verde 1 2 0 1 0 1 0 0 1 Palo Verde 2 2 0 1 0 0 0 0 1  !

Palo Verde 3 1 0 1 0 0 0 1 1 Peach Bottom 2 0 1 1 0 0 0 0 0 Peach Bottom 3 1 1 2 0 0 0 0 0 '

Perry 0 0 1 2 1 1 0 1 Pilgrim 1 0 1 1 0 1 2  !

Point Beach 1 0 0 1 5 2 7 9 2  ;

Point Beach 2 0 0 0 4 2 7 9 2 Prairie Island i 1 1 1 1 1 1 1 0 Prairie Island 2 1 1 1 1 1 i 1 0 Quad Cities 1 1 0 3 2 I I 3 0 Quad Cities 2 1 0 1 2 1 1 3 0 NUREG-1272, Vol. I 1. No.1 A-1 -40

Reactors Table A-1.14 Cause Codes- Design / Construction / Installation / Fabrication-Quarterly PI Data Calendar Year-Quarter Pirnt Name 95-4 96 1 96-2  %-3 96-4 97-1 97-2 97-3 River Bend i 1 1 0 0 0 0 2 Robinson 2 0 0 0 0 0 0 2 0 Salem l 1 2 2 8 5 2 1 0 Salem 2 1 2 2 8 6 2 2 0 San Onofre 2 1 0 1 0 1 1 1 0 San Onofre 3 0 1 0 0 1 1 0 Seabrook 0 0 1 0 2 3 4 0 Sequoyah! l 0 0 1 1 0 0 0 Sequoyah 2 1 0 0 1 0 0 1 0 South Texas ! l 0 0 0 0 2 0 0 South Texas 2 0 1 0 0 0 3 0 0 St. Lucie 1 0 0 0 1 0 0 2 1 St. Lucie 2 0 0 1 0 0 0 4 i Summer 0 1 0 2 0 0 0 0 Surry 1 0 0 0 0 0 1 0 0 Surry 2 0 0 1 0 0 0 0 0 Susquehanna l 1 1 1 0 2 1 0 2 Susquehanna 2 1 1 0 1 2 2 0 2 Three Mile Isl 1 0 0 0 0 0 1 1 1 Turkey Point 3 1 1 0 0 0 0 1 0 Turkey Point 4 0 0 0 0 0 0 1 0 Vermont Yankee 1 5 5 5 3 6 2 0 Vogtle 1 1 1 1 0 1 0 0 0 Vogtle 2 0 0 1 0 1 0 0 0 Wash. Nuclear 2 0 0 1 0 3 1 0 0 Waterford 3 1 1 2 1 1 3 3 0 Watts Bar 1 1 4 4 0 1 1 0 1 Wolf Creek i 1 1 0 0 0 0 0 Zion 1 3 2 0 1 1 1 1 0 Zion 2 2 3 0 1 2 1 1 0 Total 86 107 116 101 143 177 137 38 A-1-41 Appenaix A-1

=~ _. . _ - ._ . ._ . .. - _- -. -

1997 AEOD Annual Report Table A-1.15 Cause Codes - Miscellaneous - Quarterly PI Data Calendar Year-Quarter Plant Name 95-4  %-1 96-2 %3 96-4 97 1 97 2 97 3 Arkansas 1 0 0 0 1 0 0 0 1 Arkansas 2 0 0 0 0 0 0 0 0 Beaver Valley 1 0 0 0 0 0 0 0 0 Beaver Valley 2 0 0 0 0 0 0 0 0 Big Rock Point 0 0 0 0 0 0 0 0 Braidwood 1 2 0 0 0 0 0 1 0 i Braidwood 2 0 0 0 0 0 0 0 0 Browns Ferry 1 0 1 0 0 0 0 0 0 Browns Ferry 2 0 0 0 0 0 0 0 0 Browns Ferry 3 0 1 0 1 1 0 0 0 Brunswick 1 0 0 0 0 0 0 0 0 Brunswick 2 0 0 0 0 0 0 0 0 Byron 1 0 0 1 0 0 0 0 0 Byron 2 0 0 0 0 0 0 0 0 Callaway 0 0 1 0 0 0 0 0 Csivert Cliffs 1 0 1 0 0 0 0 0 0 Calvert Cliffs 2 0 1 0 0 0 0 0 0 Catawba 1 0 0 2 0 0 0 0 0 Catawba 2 0 0 0 1 0 0 0 2 Clinton 1 0 0 0 0 0 0 0 0 Comanche Peak 1 1 0 1 1 0 0 0 0 Comanche Peak 2 0 1 0 0 0 0 0 0 Cook 1 0 1 0 1 0 0 0 0 Cook 2 0 1 0 0 0 0 0 0 Cooper Station 0 0 0 0 0 0 0 0 Crystal River 3 0 0 0 0 0 0 0 0 Davis-Besse 0 0 0 0 0 1 1 0 Diablo Canyon i 1 0 0 1 0 0 1 0 Diablo Canyon 2 1 0 0 1 0 0 0 0 Dresden 2 0 0 0 0 0 0 0 0 Dresden 3 0 0 0 0 0 0 0 0 Duane Arnold 0 0 0 1 0 0 0 0 Farley 1 0 0 0 0 0 0 0 0 Farley 2 1 0 0 0 0 0 0 0 Fermi 2 0 1 0 0 0 0 1 0 FitzPatrick 0 0 1 1 0 0 0 0 Fort Calhoun 0 0 0 0 0 0 0 0 Ginna 0 0 0 0 0 0 0 1 Grand Gulf 0 0 1 0 0 0 0 0 Haddam Neck 0 0 0 0 0 NA NA NA PSD means the plant was permanently shutdown.

NUREG-l 272, Vol. I 1, No. I A-1 -42

Reactors Table A 1.15 Cause Codes-Miscellaneous-Quarterly PI Data Calendar Year-Quarter Plant Name 95 4 96-1 96-2 96-3 96-4 97 1 97 2 97 3 liarris 0 0 0 0 0 0 0 0 liatch 1 0 1 1 0 0 0 0 0 llatch 2 0 0 0 0 0 0 0 0 liope Creek 0 0 2 0 0 0 1 0 Indian Point 2 0 0 0 0 0 0 0 1 Indian Point 3 0 1 0 0 0 0 1 0 Kewaunee 0 0 0 0 0 0 0 0 LaSalle 1 0 0 0 0 0 0 0 0 LaSalle 2 0 0 0 0 0 0 0 0 Limerick 1 0 0 0 0 0 0 0 0 Limerick 2 0 0 0 0 0 0 0 1 Maine Yankee 0 0 0 0 2 0 0 0 McGuire 1 1 0 0 0 0 0 0 0 McGuire 2 0 0 0 0 0 0 0 0 Millstone 1 0 0 0 0 0 0 0 0 Millstone 2 0 0 0 0 0 0 0 1 Millstone 3 0 0 0 1 0 0 0 0 Monticello 0 0 0 0 0 0 0 1 Nine Mile Pt.1 0 0 0 0 0 0 0 0 Nine Mile Pt. 2 0 0 1 0 0 0 0 0 North Anna 1 0 0 0 1 0 0 0 1 North Anna 2 0 0 0 0 0 0 0 0 Oconee 1 0 1 0 0 0 0 0 1 Oconee 2 0 0 0 0 0 0 0 0 Oconee 3 0 0 0 0 0 0 0 0 Oyster Creek 0 0 0 0 0 0 0 0 Palisades 0 0 0 0 0 0 0 0 Palo Verde 1 0 0 0 1 0 0 1 0 Palo Verde 2 0 0 0 0 0 0 0 0 Palo Verde 3 0 0 0 1 0 0 0 0 Peach Bottom 2 0 0 0 0 1 0 0 0 Peach Bottom 3 0 0 0 0 0 0 0 0 Pen 0 1 0 0 0 1 0 0 Pilgrim 0 0 0 0 0 0 0 0 Point Beach 1 0 0 0 0 0 0 0 0 Point Beach 2 0 0 0 0 0 0 0 0 Prairie Island 1 0 0 1 0 0 0 0 0 Prairie Island 2 0 0 1 0 0 0 0 0 Quad Cities 1 0 1 0 0 0 0 0 0 Quad Cities 2 0 0 0 0 0 0 0 0 A- 1 -43 Appendix A-1

1997 AEOD Annual Report Table A.I.15 Cause Codes - Miscellaneous - Quarterly PI Data Calendar Wer-Quarter Plant Name 95 4 96-1 96 2 96-3  %-4 97 1 97-2 97 3 River Bend 0 0 0 0 0 0 0 0 Robinson 2 2 0 0 0 1 0 0 0 Salem 1 0 0 0 1 0 0 0 0 Salem 2 0 0 0 0 0 0 0 0 San Onofre 2 0 1 0 0 0 0 0 0 San Onofre 3 0 1 0 0 0 0 0 0 Seabrook 0 1 0 0 0 0 0 0 Sequoyah1 0 0 0 0 0 0 0 0 Sequoyah 2 0 0 0 0 0 0 0 0 South Texas 1 0 0 0 0 0 0 0 0 Sorth Texas 2 0 0 0 0 0 0 1 0 St. Lucie 1 0 0 0 0 0 0 0 0 St. Lucie 2 0 0 0 0 0 0 0 0 Summer 0 0 0 0 0 0 0 0 Surry 1 0 0 0 0 0 0 0 0 Surry 2 0 0 0 0 0 0 0 1 Susquehanna 1 0 0 0 0 0 0 0 0 Susquehanna 2 0 1 0 0 0 0 0 0 Three Mile Isl 1 1 0 0 0 0 0 0 0 Turkey Point 3 0 0 0 1 0 1 0 2 Turkey Point 4 0 0 0 0 0 0 0 0 Vermont Yankee 0 0 0 0 0 0 0 0 '

Vogtle 1 0 0 1 0 0 0 0 0 Vogtle 2 0 0 0 0 0 0 0 0 Wash. Nuclear 2 0 0 0 0 0 0 0 0 Waterford 3 0 0 0 0 0 1 0 0 Watts Bar 1 0 0 0 0 0 0 0 0 Wolf Creek 0 0 0 0 0 0 0 0 Zion 1 0 1 0 1 0 0 0 0 '

Zion 2 0 0 0 1 0 0 0 0 Total 10 18 14 17 5 4 8 13 i

)

l NUREO 1272, Vol. I1, No.1 A- 1-44

_ . . _ _ . . . . . . . . . . . . - ~ . . . . - . - . _ . . . . . -_. .- -

l APPENDIX A-2 Other Plant Operational Experience Data

)

1 I

l l

i i

r i

e

I Reactors CONTENTS Other Plant Operational Experience Data . . .. . . . A-2-iv Tables A-2.1 Automatic and Manual Reactor Scrams While Critical ani Reactor Scrams /1000 Critical Hours . . A-2-1 A-2.2 Reactor Scram Initiating Systems . . . . A-2-5 A-2.3 Activities at Time of Reactor Scram.. . . . A-2-6 A-2.4 Reactor Scram Causes . . A-2-7 A-2.5 Reactor Scram Signals. . . A-2-8 A-2.6 Engineered Safety Feature Actuations. . . A-2-9 A-2.7 Engineered Safety Feature Actuations of Selected Systems . . . A-2-12 A-2.8 Engineered Safety Feature Actuation Activities . . A-2-13 A-2.9 Engineered Safety Feature Actuation Causes. . . A-2-14 A-2.10 Critical, On-Line, Outage, and Availability Data for 1997 . . A-2-15 A-2.11 Capacity Factors for 1997 . . . . A-2-18 A-2.12 Industry Critical, On-Line, Outage, Availability, and Capacity Data . . . A-2-21 A-2.13 Allegations at Commercial Nuclear Power Plant Sites for CY 1993 through CY 1997. . . A-2-22 I

a i

1 i

i A-2-iii Appendix A-2

1997 AEOD Annual Report OTHER PLANT OPERATIONAL EXPERIENCE DATA This appendix presents selected licensee event factors and outage data. Table A-2.13 summarire:

report (LER) and plant operational experience data on allegations at commercial nuclear plant sites.

data. This information is referenced in Section 2 ,

of this report. Note that in Tab:es A-2.2 through A-2.5. because of round-off in scme individual entries in the " Scrams /

Tables A 2.1 through A-2.5 present reactor scram 1000 Critical Hours" columns, the sum of the data. Tables A-2.6 through A-2.9 contain data on individual entries may not equal the total shown for engineered safety features actuations. Tables A-2.10 that column.

through A-2.12 provide annual unit operating s

}

NUREG-1272, Vol. I 1, No.1 A-2-iv

Tsde A-2.1 Automitic cnd Min =1 Reactor Screms While Critical stnd React:r Scrtms/1000 Criticci H=rs CY93 CY94 CY95 CY% 1997 Total. Total Total Total Total Crit Scram Crit Scram Crit Scram C.9 Scram Crit Scram 1 Rate Auto blan lleurs !Ltc Auw Afai llours Rate Auto Stan Ilours Rate Plant Name Rx Type Aute hian Ilours Rate Aute klan fleurs O 7599 4 0 13 0 8657.8 0.12 2 7575.9 0 40 2 0 7663.0 0.26 0 0 8204.8 0.00 Arkansas ! PWR I 1 1 0 0 8390 4 0 00 0 0 7739.7 0(0 2 0 6909.5 0 29 0 0 80615 0 06 0 0 7323 2 0 00 Arkansas 2 PWR 5980.6 0.17 2 0 7025 9 0 28 0 0 6895.0 0 00 I O 71 % I 0.14 2 0 7223.3 0 28 Beaver Valley l PWR I O 0 6191.5 0.32 PWR 0 6829.t 0.15 0 8494.2 0 12 1 0 7657.3 0.13 0 1 6239.2 0.16 2 Bee er Valley 2 1 1

'320 0 2 6598.9 0 30 1 0 8318.7 0.12 2 0 5667.6 0 35 1 0 4742.6 Big Rock Pomt BWR 0 0 6v58 8 O (X) 0 7000 6 0 14 0 6379.2 0.16 0 0 7(M7.5 0.00 0 0 6127.1 D iK)

Braidmond i PWR I O 8081.0 0.12 1 1 BraidwoW 2 PWR I O 7151.7 0.14 2 0 6518.3 0.31 0 0 8589 0 0.00 0 0 7367.7 om 0 0 86M O 0 00 0 0 00 ESD 0 0 00 ESD 0 0 o t* ESD 0 0 00 ESD 0 0 00 ESD Browns Ferry i BWR 0.17 3 0 7310 0 0.41 3 0 8652.0 0J5 2 0 7871 0 0.25 2 0 Ts505 0 0 24 Bruans Ferry 2 BWR 0 1 5853 9 0 0 00 ESD 0 0 00 ESD 0 0 988 8 0 00 3 0 8459.2 0.35 0 0 8320 0 0 00 Browns Ferry 3 BWR 0 0 0.0 ESD 0 0 7990 0 0 00 2 1 7521.3 0 40 0 1 7607.4 0.13 0 0 7997.5 0 00 BrunswkL l BWR 0 0 00 0 0 6549.0 0 00 0 0 8760 0 0.00 0 0 7475.9 0 00 0 0 P.i315 0.00 Brunsunk 2 BWR 0 5915.3 0 0 7152.2 0 00 0 7174 8 0.14 0 0 7234 0 0 00 l I E66.2 0.30 0 0 M318 0.00 Byron i PWR 1 0 7470.3 0.27 0 8709.5 0.11 0 0 7739.5 On) 0 1 7267.8 0.14 0 0 85761 He y Byron 2 PWR 2 1 0.00 0 0 8760 0 0 00 2 I U ? 4.8 0.40 0 1 7951.4 0. I 3 0 0 7954.4 ON y Callaway PWR 0 0 7569 0 0 5911.7 0.51 0 3 85 5.4 0.35 0 0 5869.4 0 00 0 0 8490 4 0 00 Calvert Chffs ! PWR I I 8619 0 0.23 3 6072,4 0.16 4 0 8000 2 0.50 2 1 T205 8 0 42 2 0 8599 6 0.23 1 0 7090 4 0 14 Calvert Cliffs 2 PWR I O 0 7782.1 0 00 0 0 5940.1 0 00 0 0 8678.4 0 00 Catawba I PWR 2 0 6991.4 0.29 1 0 8733 6 0.11 0 0 7294.5 0.14 3 7068.5 0.57 2 1 7157.1 0.42 1 0 8202 9 0.12 1 2 75441 0 40 Catsuba 2 PWR 1 1 0 29 0 0 8308.1 O (x) 0 2 7273 6 0.27 I I 5858.3 0.31 0 0 00 ESD Cimton I BWR 0 2 6970 0 0.57 0 8674 0 0.23 2 1 7539 0 0.40 . 2 7342.0 0.P O O 7767.0 0 00 s Comanche IYak I PWR 3 1 7020.8 2 '

3 5828.0 0 69 2 0 8427.0 0.24 2 I 6960.0 0.43 1 0 86510 0.12 Comanche Peak 2 PWR 2 1 5189.2 0 58 1 0 0 6256.5 0 00 I I 6081.4 0 33 2 0 8595.0 0 23 0 0 t. %0 0 00

• Cooki PWR 0 0 8760 0 0.00 0 5167.5 0.58 4 0 8307.8 0.48 1 0 7687 6 0.13 1 0 796 C O I)

Cook 2 PWR 2 0 ?491.5 0.24 3 0 3076 4 0 33 0 0 5851 2 0 00 0 0 8603.3 0 00 0 0 7395 0 0 00 Cooper Starmn BWR I O 5146 8 0.19 1 0 0 7382.0 0.00 0 0 8760 0 0 00 1 0 3307.3 ' J) 0 0 00 ESO Crystal River 3 PWR I O 7445.8 0.13 0 0 8760 0 0 00 0 0 7490: 0 00 1 0 8236 2 0.12 Davis-Besse PWR 2 0 7305.4 0.27 0 0 7705.0 0 00 0 7(Mt.0 0.14 I 2 7267.2 04l 2 I 8M8 0.36 1 0 7563.5 0 13 Diablo Canyon i PWR I O 8631.I 0.12 1 0 8492.2 0.12 0 759u6 0 13 1 1 8520 4 0.23 thablo Canyon 2 PWR I O 7384.8 0.14 1 1 7560.2 0 26 1 1 0.17 0 3012.0 0 33 0 1 4040 0 0.25 0 1 7924 4 0.13 thesien 2 BWR 0 0 4886.7 0 (.X) 0 1 5980 6 1 0 3085 0 0 32 3 5708 0 0.70 1 0 4357.7 0.23 0 0 4580.0 O(M)  !

y thesden 3 BWR 2 l 7816.7 0.42 1 1 0

0 14 0 8236 0 0,12 0 7345.2 0. I4 0 0 7c68.3 0 00 0 7456.6 0 (X) c Duane Arnold BWR I O 6%3.4 1 1 0 00 0 0 00 0 7432.6 0.27 0 0 8755.3 0 (x) 0 0 6816.9 3

{

Farley i Farley 2 PWR PWR 0

1 0

0 8542 6 6931.8 0 00 0.14 0

2 I 7592.9 87(R 0 0.34 2

2 2 7245.7 0 53 0 0 7213.6 0 00 0 0 0

7189.6 0 00 4114 0 0.24 k

4 8141.8 0 61 0 0 190 6 0 00 2 0 7616 0 0 26 1 0 6073.1 0.16 1 E Ferm: 2 BWR I 6 '

> E b ESD means the plant was in an extended shuhtown (rero critical hours for the year).

Z C Table A-2.1 Automatic and Manual Reactor Scrams While Critical and Reactor Scrams /1000 Critical flours Qi

c e CY93

-a

[.

CY94 CY95 CY% 1997 >

! C Total m

sotal Total Total Total O

(

• d Crit Scram Crit Scram Crit Scram Crit Scram Crit Scrant U Plant Nome Rx Type Auto Man Hours o

Rate Auto Man Iloun Rate Auto Man linurs Rate Auto Man lleurs Rate Auto Man lleurs Rate >

U I

[ FitiPatrick BWR 3 1 7157.5 0 56 0 0 7291.9 0.00 0 6528.7 0.15 I 8

1 2 7297.8 0.41 0 3 7437.5 0.40 E.

• Fort Calhoun PWR 2 0 7081.4 0.28 1 0 8726.2 0.11 1 2 7290.1 0.41 0 I 6983.6 0.14 0 I Z 6972.7 0.14 y Ginna PWR 2 0 7561.8 0.26 I O 7288.8 0.14 0 7851.4 0.13 0 0 9 1 l l 6247.3 0 32 82 % 9 0 00 0 Grand Gulf BWR I 7140.5 0.14 Iladdam Neck PWR I O

I 7145.9 0 28 1

I I

I 8464.5 t ' io.2 0.24 0.29 5

0 0

1 7039.9 6808.9 0.71 0.15 0 0 2 0

7788 0 4895.3 0.26 0 00 0

0 0 1 7815 8 0.0 0.13 ESD 3*2 Ilarris PWR 0 0 8733.4 O (x) 0 0 7247.6 0.00 2 0 7337.4 0.27 I I 8378.9 0.24 3 0 6995 6 0 43 Itatch l BWR 3 2 7099.4 0.70 2 0 7638.1 0.26 0 0 8760 0 0 00 3 1 7832.3 0.51 0 0 8667.1 0 00 llatch 2 BWR 0 1 7873.9 0.13 1 0 7619.7 0.13 I I 7121.7 0.28 0 0 8677.1 0 00 1 0 7817.3 0.13 Ikye Creek BWR I I 8567.4 0.23 5 0 7112.9 0.70 0 6988.0 0 14 0 0 1 6799.2 0 00 0 1 8144 0 0.12 Indian Point 2 PWR 0 0 6630.7 O (x) 0 0 8760 0 0 00 2 0 5884.7 0.34 4 0 8325 6 0 48 3 0 5687.0 0 53 Indian Point 3 PWR 0 0 1303.5 0 00 0 0 0.0 ESD 0 2 1873 4 0 Kewaunce PWR 0 1.07 1 6569.1 0.15 2 1 5135I O 58 2 7607.6 0.26 0 0 7750 8 0.00 2 0 7690.5 0.26 2 0 6305.4 0.32 0 0 2697.7 0 00 LaSalle i BWR I 2 7402.3 0.41 3 0 5313 0 0.56 I I 8302.2 0.24 I I 3777.5 0 53 0 0 00 ESD y LaSalle 2 BWR 0 0 5912.2 0 00 4 0 8282.4 0.48 0 0 6081.5 0.00 0 5755.2 0.17 0 0 1 00 ESD g Limerick I BWR I O 8649.9 0.12 0 1 7909.2 0.13 1 2 8115.4 0 37 2 0 7903 6 0.25 0 0 8659.0 0 00 Limernk 2 BWR 2 0 7401.8 0.27 I O 8720.4 0.11 3 0 8169 8 0 37 I 2 8494 4 0.35 0 2 7853 2 0.25 Maine Yankee PWR 0 0 6991.8 0.00 I I 79601 0 25 0 320.7 3 12 2 0 68M.5 0.29 i 0 McGuire i PWR 1

1508.2 0 66 I O 5164.3 0.19 1 0 6338.9 0.16 1 2 8079.9 0.37 0 7952.6 0.13 0 1 1 6228 4 0.16 McGuire 2 PWR I 3 M25.7 0.62 0 0 7710.8 0 00 0 0 8202.9 0 00 I O 6586 8 0 15 2 0 7821.8 0.26 i Mdistone i BWR I O 8481.2 0.12 0 0 5575 0 0 00 0 0 70nt.3 0 00 0 0 00 ESD 0 0 0.0 ESD Mdistone 2 PWR 5 0 7689.9 0 65 0 1 4349.2 0.23 0 1 3391.9 0 29 0 0 1223 6 0 00 0 0 0.0 ESD Millstone 3 PWR I O 6275.8 0 16 0 8454.7 0 12 0 0 7152.7 0.00 0 0 1 2158 8 0 00 0 0 00 ESD Monticello BWR 2 1 7391.0 0.41 2 0 7624.2 0.26 0 0 8760 0 0 00 2 1 7633 0 0.39 0 1 6729 0 0 15 Nme Mile Pt. I BWR 2 0 7442.3 0.27 4 0 5428.I 0.47 0 7412.4 0. I 3 0 1 2 8200.3 0.24 1 0 6239 0 0. I6 Nme Mile Pt. 2 BWR I O 7377.0 0 14 1 1 8374.3 0 24 0 4 7042.6 0.57 0 0 78 %.I 0.00 0 0 7689.9 0.00 Nonh Anna I PWR 0 0 M74.9 0.00 0 0 8042.1 0 (x) 1 0 8738 6 0 11 2 0 80308 0 25 0 1 8017.2 0 12 Nonh Anna 2 PWR I I 7329.4 0 27 1 0 8559.9 0.12 1 0 7124.1 0.14 0 6897.6 014 I O 1 7423 8 0.13 (konee l PWR I I 7928 0 0.25 1 0 7371.5 0 14 0 0 7.* 45 O(X) I O 6745 4 0.15 0 0 4M4.3 0 00 Oconee2 PWR 2 0 7422.5 0 27 2 0 7387.2 0 27 0 L M.4 0.12 0 1 1 5349.8 0.19 I O 4853 8 0 21 Ownec 3 PWR I O 8655.4 0 12 3 0 6835.7 0 44 0 7650.3 0.13 0 o 1 1 M 34.3 0 16 I 4287.5 0 23 Oyster C:rek BWR 0 0 7690.6 0.00 2 0 6201 6 0.32 0 8532.2 1 0.12 2 1 7252.0 0.41 0 2 7487.5 0.27 Palisades PWR 0 0 4707.4 0 00 0 0 5872 0 0.00 0 1 6639.3 0.15 0 0 7i45.9 0 00 0 0 0 00 6942.3 Palo Verde 1 PWR 0 1 67818 0.15 0 0 8675.4 0 00 3 0 7330.2 0.41 0 7362.2 0.14 0 1 1 7974.5 0.13 Palo Verde 2 PWR I I 47231 0 42 2 0 6l03 2 0 33 0 7483.4 0.13 0 0 0 1 1 7598.6 0.13 8160 0 O lx)  !

Palo Verde 3 PWR I i 8008.2 0.25 2 0 5998.1 0 33 0 0 1679.0 0.00 0 8763.8 0 1 0. I I 1 7868 9 0 13  !

I ESD means the plant was m an estended shutdom n (zero cntical hours for the year) L f

Table A-2.1 A: tom tic and Minual Reactor Scrcms While Criticci rnd R2 actor Scrams /1000 Critical Hours CY95 CY96 1997 CY93 CY94 Total Totat Totat Totat Totat Crit Scram Crit Scram Crit Scram Crit Scrum Crit Scram Rate Auto Man Hours Rate Auto Man Ilours Rate Auto Man Hours Rate Plant Name Rs Type Auto Man fleurs Rate Auto Man flours 0 7851 0 0.13 0 0 8632.4 0 00 2 0 8320.0 0.24 2 0 8706 0 0.23 Itach Bottom 2 BWR I o 7728 0 0.13 1 6613 0 0.30 I i 8588.0 0.23 2 I 8027.8 0.37 0 0 8695.0 0 00 0 1 8700.5 0 Il Itach Bottom 3 BWR I I 0 79041 0 25 0 8378 0 036 0 6780.9 0 15 2 Ittry BWR 0 2 4219 2 0 47 0 1 4398.9 0.23 3 1 0.12 0 7348.7 0.14 0 7082.9 0.42 0 6258 6 0.16 0 1 7066.0 0.14 1 0 8380.8 1 Pilgnm BWR 3 1 3382.6 0 00 0 00 0 0 8134.6 0.00 I 7814.8 0.26 0 0 8219.5 0 00 0 0 Inint Beach I PWR 0 0 7835.6 1 0 00 0 7275.5 0.14 0 6660.5 0.15 0 0 895 4 0 00 Ptunt Beach 2 PWR I O 7924 7 0.13 0 0 7851.0 1 1 0 8381.2 0 12 I O 8507.9 0.12 0 0 7292.4 0.00 0 0 8760.0 0 00 1 0 7381.4 0 14 1 Praine Island i PWR 0 8684.2 0 35 0 0 7233 0 0 00 0 0 7381.4 0.00 0 8743.2 0.11 0 0 7698.8 0 00 3 Praine lsland 2 PWR 1 0 00 0.14 0 2651.2 0.38 0 0 "MID 0 00 0 1 4045.7 0.25 0 0 8141.1 Quad Cities t BWR I O 7020.4 1 I I 5874.2 0.34 1 0 42w.6 0.23 0 0 6516.1 0.00 0 0 5989.5 0.00 Quad Cities 2 BWR 4 I 4725.8 1.06 0.53 0 I 8724.5 0.11 0 2 7495.8 0.27 I I 7999.1 0.25 River Bend BWR 2 0 6272.4 032 2 1 5684.4 I O 7421.1 0.13 I I 77833 0.26 1 0 8312.1 0 12 Rohmson 2 PWR 0 0 6191.2 0 00 0 2 6963 6 0.29 4 I 6587.7 0.76 0 0 2660.9 0 00 0 0 00 ESD 0 0 00 ESD '

Salem I PWR 3 1 5949.9 0.67 0 32 0 2468.4 0.41 0 0 0.0 ESD 0 0 1003 0 om 036 6335 6

> Salem 2 PWR I 0

I 5513.9 0 00 0 I

0 I

8760 0 0 00 0 1

0 6613 6 0.00 0 0 8016.7 0.00 0 0 5497.8 0 00 ja San Onofre 2 PWR 0 7280.2 th 030 0 0 8750 0 0.00 0 0 7250 3 0 00 0 0 8355.5 0 00 0 0 6244 4 0 00 San Onofre 3 PWR 2 0 6726.6 0.18 0 7663.0 0.13 0 8744.9 0.11 I O 7618.7 0.13 Seabrook PWR 3 2 8203,7 0 61 1 0 5559.9 1 1 0.50 2 3 6842.4 0.73 1 1 8428 2 0.24 0 2 7492.9 0.27 Sequoyahi PWR I i 12813 1.56 2 I 6021.2 8237.7 0.49 2 7008.4 0.43 1 1 8164.2 0.24 Sequoyah 2 PWR I 1 2545 9 0.79 0 0 5597.5 0 00 3 1 1 I I 70803 0.28 3 0 7684 4 039 0 0 8242.2 0.00 0 0 8255 2 0 00 South Texas i PWR 0 0 720.1 0.00 0 19 2 0 8064 0 0.25 0 0 8453.7 0 00 1 2 8262.2 036 South Texas 2 PWR 2 0 739 5 2.70 1 0 5280 6 6859.5 0.44 4 I 7D M 0 64 I I 6716.1 0 30 0 2 6627.7 030 1 0 8610.5 0.12 St_ Lucie i PWR 0 3 0 0 7795.3 0.00 aa, 0 - wi 014 ' O 6602.9 0 15 0 2 8575 6 0.23 St Lucie 2 PWR 0 2 67592 t 8656.4 0.23 u.M 0 0 8516 6 0.00 0 0 7928 9 0.00 I I Summer PWR I O 7357.9 0.14 0 9 w90.5 7581 0 0.26 0 0 8784.0 0.00 0 1 7198.1 0.14 Surry l PWR 2 0 8432.2 0 24 0 1 6662.5 0.15 4 2 7165.4 0.56 2 7572.7 0 40 I I 8454.5 0.24 Surry 2 PWR 5 0 6389.4 0 78 0 0 8261.2 0 00 2 1 0 0 71761 0.00 I O 7544 8 0 13 0 1 7823 3 0.13 Susquehannat BWR I O 5275.4 0.19 0 0 8292.4 0.00 0 0r I O 6673.8 0.15 I o 7776 7 0.13 0 1 8394 4 0.12 0 0 7278 3 0 00 Susquehanna 2 BWR 0 0 8275.5 0 7979.2 0.13 0 0 8362.6 0 00 0 0 7954 4 0 00 0 0 8784.0 0 00 1 Three Mile ist i PWR I O 7749 5 0.13 0.13 0 7927.6 0.13 I i 8545.5 0 23 I I 7566 3 0.26 Turkey Point i PWR 0 0 8501.0 020 1 0 7718.1 1 0 7567.5 0 26 0 0 86381 0 00 0 I 7968.1 0.13 1 0 8134 0 0.12 y Turkey Pomt 4 PWR 2 0 7441.7 0.27 2 0 8M62 0.12 0 7618.4 0.13 0 0 7485 6 0 00 1 0 7735 4 0. I 3

• C Vennont Yankee ilWR 0 0 7021.0 0 00 1 1 I I 7248 6 0.28 0 7721.1 0 13 0 0.1 I k

8701.6 1 3 Vogtle i PWR 2 0 7673 2 0 26 2 2

0 0

7889.5 8107.4 0.25 0 25 1

0 7%8.8 0.13 I I 79473 0.25 I I 8474 3 0.24 7794.7 0.26 1

{ vorsle 2 PWR I l 6%I .5 0 57 0 6590.4 0.15 3 0 6935.2 0.43 0 1 6236.8 0.16 0 1 6372.7 0 16 @

X Wash IQ:: lear 2 BWR 3 1 1 6

. U tJ ESD means the plant was in an extended shutdown (zero entical hours for the year).

Z C Table A-2.1 Automatic and Manual Reactor Scrams While Critical and Reactor Scrams /1000 Critical Hours 3 W e CY93 CY94 w

$. CY95 CY96 1997 >

C Total Total

!T1 Crit Scram Crit Scram Total Total Totat O

-d Plant Name Rx Type Crit Scram Crit Scram Crit Scram U Auto Man fleurs Rate o

Auto Man fleurs Rate Auto Man Ilours Rate Auto Man lloun Rate Auto Men fleurs Rate >

.~ Waterford 3 C PWR 2 0 8707.0 0.23 I Watts Bar l PWR O 7622.8 0.13 1 0 7309.9 0.14 1 0 8271.5 0.12 0 0 6238 6 0 00 E

- NYC - - -

NYC NYC Wolf Creek PWR 0 0 7059.8 0 00 0 2 2 69J8 6 0.58 3 0 7413 7 04c M 0 7605.7 0 00 0 8648 7 0.12

. Zmni PWR I O 6987.6 0 14 2 0 4274.0 0 47 1

0 0 6344.8 0 00 1

3 1

0 7135.2 7494.1 0.28 0 40 0

0 1

0 8690.0 0 12 3447.3 0 00 Zen 2 PWR 0 0 5427.4 0.00 0 0 6218 8 0 00 0 0 6348.4 0.00 0 0 5846.4 0.00 0 0 0.9 ESD Total All Plants 121 42 726224 I 0.22 111 31 751613.2 0 19 102 53 778247_2 0.20 87 49 765013.5 0.18 57 32 706777.1 0.13 Numberof Plants 109 109 109 I10 110 Total BWR Plants 42 18 234735.5 0.26 39 13 233389 0 d22 37 17 259566.2 0.21 29 19 249174.2 0.19 12 16 238104 6 0.12 Number of BWR Plants 37 37 37 37 37 Total PWR Plants 79 24 493488.6 0.21 72 18 518224.2 0.17 65 36 518681.0 0.19 58 30 515839 3 0.17 45 16 468672.5 0.13 Numberof PWR Plants 72 72 72 73 73 ESD rneans the plant u as m an extended shutdown (irro cntical hours for the year); NYC means the plant was not yet cntical.

tb L

.- -- - . . . . . . _ _- - .-._=_ -

Reactors 4

Table A 2.2 Re actor Scram Initiating Systems i BWR Plants Total Scrams Scrams /1000 Critical Hours' System CY93 CY94 CY95 CY96 1997 CY93 CY94 CY95 CY96 1997 Main Generator 5 7 7 10 8 0.02 0.03 0.03 0.04 0.03 Turbine 12 13 10 14 4 0.05 0.06 0.04 0.06 0.02 I Electrical 5 6 8 4 4 0.02 0.03 0.03 0.02 0.02 Support 7 5 8 3 3 0.03 0.02 0.03 0.01 0.01 )

RPS 9 5 2 2 3 0.04 0.02 0.01 0.01 0.01 RCS 5 4 4 2 3 0.02 0.02 0.02 0.01 0.01 Feedwater 8 4 9 8 2 0.03 0.02 0.03 0.03 0.01 Condensate 5 3 4 3 1 0.02 0.01 0.02 0.01 0.00 l Main Steam 4 3 1 2 0 0.02 0.01 0.00 0.01 0.00 Ctrl Rod Drive 0 2 1 0 0 0.09 0.01 0.00 0.00 0.00 Total 60 52 54 48 28 0.26 0.22 0.21 0.19 0.12 Number of BWR Plants 37 37 37 37 37

'Cntical hours: 1993 = 234,735.5; 1994 = 233,389.0; 1995 = 259,566.2; 1996 = 249,174.2; and 1997 = 238,104.6 PWR Plants Total Scrams Scrams /1000 Critical Hours 2 System CY93 CY94 CY95 CY96 1997 CY93 CY94 CY95 CY96 1997 ]

Feedwater 28 20 24 25 16 0.06 0.04 0.05 0.05 0.03 Electrical 6 9 16 16 13 0.01 0.02 0.03 0.03 0.03 i RPS 16 8 5 7 7 0.03 0.02 0.01 0.01 0.01 Main Generator J5 19 15 12 7 0.03 0.04 0.03 0.02 0.01 Support 6 10 8 8 6 0.01 0.02 0.02 0.02 0.01 Turbine 12 11 9 10 5 0.02 0.02 0.02 0.02 0.01 Main Steam 7 5 6 1 4 0.01 0.01 0.01 0.00 0.01 RCS 7 2 5 1 2 0.01 0.00 0.01 0.00 0.00 Ctrl Rod Drive 5 4 12 4 1 0.01 0.01 0.02 0.01 0.00 Condensate 1 2 1 4 0 0.00 0.00 0.00 0.01 0.00 Total 103 90 101 88 61 0.21 0.17 0.19 0.17 0.13 Number of PWR Plants 72 72 72 73 73 2 Critical hours: 1993 = 491,488.6; I994 = 51R,224.2; I995 = 518,681.0; I996 = 515.839.3; and I997 = 468,672.5 A-2-5 Appendix A-2

1997 AEOD Annual Report Table A-2.3 Activities at Time of Reactor Scram BWR Plants 1

Total Scrams Scrams /1000 Critical Hours' Activity CY93 CY94 CY95 CY96 1997 CY93 CY94 CY95 CY96 1997 Normal Operation 28 25 29 25 16 0.12 0.11 0.11 0.10 0.07 Maintenance 9 8 6 8 5 0.04 0.03 0.02 0.03 0.02 Testing 15 10 12 5 4 0.06 0.04 0.05 0.02 0.02 Power Change 8 9 7 10 3 0.03 0.04 0.03 0.04 0.01 Total 60 52 54 48 28 0.26 0.22 0.21 0.19 0.12 Number of BWR Plants 37 37 37 37 37

' Critical hours: 1993 = 234,735.5; I994 = 233,389.0; 1995 = 259.566.2; 1996 = 249,174.2; and 1997 = 238,104.6 PWR Plants Total Scrams Scrams /1000 Critical Hours 2 Activity CY93 CY94 CY95 CY96 CY97 CY93 CY94 CY95 CY96 1997 Normal Operation 59 52 64 46 34 0.12 0.10 0.12 0.09 0.07 Maintenance 11 14 16 9 10 0.02 0.03 0.03 0.02 0.02 Power Change 8 14 7 17 9 0.02 0.03 0.01 0.03 0.02 Testing 25 10 14 16 8 0.05 0.02 0.03 0.03 0.02 l Total 103 90 101 88 61 0.21 0.17 0.19 0.17 0.13 Number of PWR Plants 72 72 72 73 73 2 Critical hours: 1993 = 491,488.6; 1994 = $ 18,224.2; 1995 = 518,681.0; 1996 = 515.839.3; and 1997 = 468.672.5 i

NUREG-1272. Vol. I 1, No.1 A-2-6

Reactors Table A 2.4 Reactor Scram Causes BWR Plants Total Scrams Scrams /1000 Critical llours' Cause CY93 CY94 CY95 CY96 1997 CY93 CY94 CY95 CY96 1997 Equipmem 37 37 40 32 16 0.16 0.16 0.15 0.13 0.07 Personnel Error 14 9 8 8 7 0.06 0.04 0.03 0.03 0.03 Other 9 6 6 8 5 0.04 0.03 0.02 0.03 0.02 Total 60 52 54 48 28 0.26 0.22 0.21 0.19 0.12 Number of BWR Plants 37 37 37 37 37

' Critical hours: 1993 = 234,735.5; 1994 = 233,389.0; 1995 = 259,566.2; I996 = 249,174.2; and 1997 = 238,104.6 PWR Plants Total Scrams Scrams /1000 Critical llours2

, C9use CY93 CY94 CY95 CY96 1997 CY93 CY94 CY95 CY96 1997 Equipment 72 63 74 65 53 0.15 0.12 0.14 0.13 0.11 Personnci Error 20 14 15 7 5 0.04 0.03 0.03 0.01 0.01 Other 11 13 12 16 3 0.02 0.03 0.02 0.03 0.01 Total 103 90 101 88 61 0.21 0.17 0.19 0.17 0.13 Number of PWR Plants 72 72 72 73 73 2 Critical hours: 1993 = 491,488.6; 1994 = 518,224.2; 1995 = 518.681.0; 1996 = 515,839.3; and 1997 = 468.672.5 A-2 7 Appendix A-2

1997 AEOD Annual Report Table A-2.5 Reactor Scram Signals BWR Plants Total Scrams Scrams /1000 Critical Hours' Signals CY93 CY94 CY95 CY96 1997 CY93 CY94 CY95 CY96 1997 i Manual 18 13 17 19 17 0.08 0.06 0.07 0.08 0.07 Turbine Trip 12 13 20 12 8 0.05 0.06 0.08 0.05 0.03 Low Reactor Water Level 12 5 6 6 2 0.05 0.02 0.02 0.02 0.01 Other 18 21 11 11 1 0.08 0.09 0.04 0.04 0.00 Total 60 52 54 48 28 0.26 0.22 0.21 0.19 0.12 i Number of BWR Plants 37 37 37 37 37

' Critical hours: 1993 = 234,735.5; 1994 = 233,389.0; I995 = 259,566.2; I996 = 249,174.2; and 1997 = 238,104.6 PWR Plants Total Scrams Scrams /1000 Critical liours2 Signals CY93 CY94 CY95 CY96 1997 CY93 CY94 CY95 CY96 1997 Other 34 29 28 28 22 0.07 0.06 0.05 0.05 0.05 Munual 23 18 38 30 16 0.05 0.03 0.07 0 06 0.03 Turbine Trip 29 27 23 22 15 0.06 0.05 0.04 0.04 0.03 Low SG Level 17 16 12 8 8 0.03 0.03 0.02 0.02 0.02 Total 103 90 101 88 61 0.21 0.17 0.19 0.17 0.13 Number of PWR Plants 72 72 72 73 73 2 Critical hours: 1993 = 491,488.6; 1994 = 518,224.2; 1995 = 518,681.0; 1996 = 515,839.3; and 1997 = 468.672.5 l

NUREG-1272, Vol. I 1, No. I A-2-8

Reactors Table A 2.6 Engineered Safety Feature Actuations Plant Name Rx Type CY93 CY94 CY95 CY96 1997 Arkansas i PWR 2 1 0 1 1 Arkansas 2 PWR 0 0 1 0 1 BeaverValley 1 PWR 4 3 5 4 5 Beaver Valley 2 PWR 7 4 1 1 0 Big Rock Point BWR I 2 0 0 0 Braidwood 1 PWR 0 2 0 0 0 Braidwood 2 PWR 1 1 1 2 0 Browns Ferry 1 BWR 2 1 1 1 0 Browns Ferry 2 BWR 6 5 3 4 2 Browns Ferry 3 BWR 3 2 6 4 4 Brunswick I BWR 7 10 6 3 4 Brunswick 2 BWR 6 5 1 1 1 Byron 1 PWR I 1 0 1 0 Byron 2 PWR 2 2 0 1 0 Callaway PWR 1 0 1 3 3 Calvert Cliffs 1 PWR 1 3 1 1 0 Calvert Cliffs 2 PWR 2 3 1 1 0 Catawba 1 PWR 2 0 0 1 0 Catawba 2 PWR 8 1 1 2 3 Clinton i BWR 3 1 2 4 2 Comanche Peak 1 PWR I O O 2 0 Comanche Peak 2 PWR 5 6 0 4 2 Cookl PWR 0 0 1 0 1 Cook 2 PVVR 1 1 1 3 0 Cooper Station BWR 6 7 5 1 4 Crystal River 3 PWR 2 0 1 1 0 Davis-Besse PWR 0 0 0 0 0 Diablo Canyon 1 PWR 1 6 4 3 3 Diablo Canyon 2 PWR 0 3 1 1 2 Dresden 2 BWR 9 9 1 2 3 Dresden 3 BWR 8 10 7 6 1 Duane Arnold BWR 8 5 9 4 3 Farley 1 PWR 0 0 4 1 0 Farley 2 PWR 1 0 2 0 0 Fermi 2 BWR 7 4 4 9 6 FitzPatrick BWR 5 3 3 5 6 Fort Calhoun PWR 4 1 3 3 1 Ginna PWR 1 4 5 6 1 Grand Gulf . BWR 12 1 7 2 1 Haddam Neck PWR 3 0 1 0 0 l

l A-2-9 Appendix A-2

1997 AEOD Annual Report Table A-2.6 Engineered Safety Feature Actuations Plant Name Rx Type CY93 CY94 CY95 CY96 1997 I

liarris PWR 1 1 5 2 1 Hatch 1 BWR 13 10 3 7 2 l

Ilatch 2 BWR 6 6 6 2 4 .

Hope Creek BWR 4 8 6 5 8  !

Indian Point 2 PWR 3 2 12 9 10 l

Indian Point 3 PWR 0 0 2 4 6  !

Kewaunee PWR 7 2 1 3 1 LaSalle 1 BWR 6 6 3 2 0 LaSalle 2 BWP, 4 3 6 3 0 Limerick I BWR 10 11 7 8 4 ,

Limerick 2 BWR 9 5 10 5 4 Maine Yankee PWR 0 0 0 0 0 '

McGuire 1 PWR 1 2 0 2 4 McGuire 2 PWR 4 0 1 2 3 Millstone 1 BWR 1 2 1 1 4 Millstone 2 PWR 0 6 5 0 2 Millstone 3 PWR 0 0 3 1 0 Monticello BWR 3 11 1 4 3 Nine Mile Pt. I BWR I 1 0 0 0 Nine Mile Pt. 2 BWR 5 3 4 1 1 North Anna 1 PWR 3 1 0 0 0 North Anna 2 PWR 0 0 0 0 0 Oconee 1 PWR 0 0 0 0 1 Oconee 2 PWR 0 2 0 0 1 Oconee 3 PWR 0 1 0 1 0 Oyster Creek BWR 2 6 0 2 3

Palisades PWR 1 2 3 0 0 l Palo Verde 1 PWR 1 0 2 1 0 Palo Verde 2 PWR 2 0 2 1 2 Palo Verde 3 PWR I 2 0 0 1 l

Peach Bottom 2 BWR 1 3 3 6 4 Peach Bottom 3 BWR 2 3 5 1 0 Perry , BWR 4 8 5 2 10 l Pilgrim BWR 12 4 4 2 4 Point Beach 1 PWR 2 0 2 1 0 Point Beach 2 PWR 0 2 2 0 0 Prairie Island 1 PWR 1 6 0 4 0 Prairie Island 2 PWR 1 2 1  ?- 2 Quad Cities 1 BWR 5 0 0 1 2 Quad Cities 2 BWR 5 2 4 0 1 NUREG-1272, Vol. I 1, No.1 A-2-10

Reactors Table A 2.6 Engineered Safety Feature Actuations Plant Name Rx Type CY93 CY94 CY95 CY96 1997 River Bend BWR 7 11 3 5 3 Robinson 2 PWR 1 0 1 0 0 Salem i PWR 9 2 0 0 0 Salem 2 PWR 4 3 2 2 3 San Onofre 2 PWR 0 0 0 0 0 San Onofre 3 PWR 0 1 0 2 1 Seabrook PWR 3 1 1 0 1 Sequoyah1 PWR 7 6 2 1 3 Sequoyah 2 PWR 3 2 0 7 4 South Texas l P W P, 2 3 2 0 0 South Texas 2 PWR 4 7 4 0 0 St. Lucie i PWR 3 4 2 4 i St. Lucie 2 PWR 2 0 0 0 1 Summer PWR 1 1 1 1 0 Surry 1 PWR 2 0 1 0 0 Surry 2 PWR 0 0 5 0 1 Susquehanna 1 BWR 4 3 2 2 1 Susquehanna 2 BWR 3 3 5 2 2 Three Mile Isl 1 PWR 0 1 1 0 1 Turkey Point 3 PWR 0 2 0 2 2 Turkey Point 4 PWR 0 0 0 1 0 Vermont Yankee BWR 3 4 4 1 2 Vogtle 1 PWR 4 2 0 2 i Vogtle 2 PWR 0 0 2 0 0 Wash. Nuclear 2 BWR 5 1 1 1 0 Waterford 3 PWR 0 1 1 0 0 Watts Bar 1 ,

PWR NYL NYL 2 3 4 Wolf Creek PWR 2 4 2 1 0 Zion 1 PWR 4 2 1 1 1 Zion 2 PWR 0 1 0 2 2 Total All Plants 327 295 254 214 182 Number of Plants 109 109 110 110 110 Total BWR Plants 198 179 148 109 99 Number of BWR Plants 37 37 37 37 37 Total PWR Plants 129 116 1% 105 83 Number of PWR Plants 72 72 73 73 73 NYL means the plant was not yet licensed for low power operations.

A-2-11 Appendix A-2

1997 AEOD Annual Report Table A 2.7 Engineered Safety Feature Actuations of Selected Systems BWR Plants System CY93 CY94 CY95 CY96 1997 HVAC 109 65 82 44 34 RWCU 53 46 48 24 32 Emergency Power 37 18 16 14 20 ECCS 25 8 19 10 8 Total 224 147 165 92 94 Number of BWR Plants 37 37 37 37 37 PWR Plants System CY93 CY94 CY95 CY96 1997 Emergency Power 45 44 37 34 27 ,

HVAC 31 20 19 22 12  ;

ECCS 21 15 11 7 8  !

Total 97 79 67 63 47 Number of PWR Plants 72 72 73 73 73 ECCS - systems include: BWR - high pressure coolant injection, high pressure core spray, isolation condensers, low pressure core spray, and low pressure coolant injection.

PWR - high pressure safety injection, accumulators, and low pressure safety injection.

P Emergency Power - includes all unplanned emergency diesel generator starts, including high pressure core spray diesel.

RWCU - BWR reactor water cleanup system.

HVAC - systems include: standby gas treatment, containment fan cooling, containment combustible gas control, containment purge, reactor building environmental contml, drywell environmental control, shield annulus return and exhaust, access corridors environ-mental control, auxiliary building environmental control, fuel building environmental control, radwaste building environmental control, control building environmental control, emergency onsite power supply building environmental control, turbine building environmental control, and plant exhaust.

NUREG 1272. Vol. I 1, No.1 A-2-12

Reactors

, Table A-2.8 Engineered Safety Feature Actuation Activities BWR Plants Activity CY93 CY94 CY95 CY96 1997 Other 198 179 148 109 99  !

Normal Operation 115 85 84 67 56 Testing 55 57 43 19 25 '

Maintenance 27 34 20 22 18 i Total .

395 355 295 217 198 Number of BWR Plants 37 37 37 37 37 j l

I l

l PWR Plants Activity CY93 CY94 CY95 CY% 1997 Normal Operation 55 62 52 53 50 Testing 52 30 32 35 19 ,

Maintenance 21 23 21 17 14 1 Total 257 231 211 210 166 Number of PWR Plants 72 72 73 73 73 I

a a

A 2-13 Appendix A-2

1997 AEOD Annual Report Table A 2.9 Engineered Safety Feature Actuation Causes BWR Plants Cause CY93 CY94 CY95 CY96 1997 ,

1 Personnel Error 60 71 44 43 41 Equipment 92 72 67 47 39  !

Procedure 20 14 24 10 0 Other 26 22 13 9 9 Total 198 179 148 109 99 Number of BWR Plants 37 37 37 37 37 i

PWR Plants Cause CY93 CY94 CY95 CY96 1997 Equipment 64 61 37 49 40 Personnel Error 43 26 32 28 17 Other 3 11 23 15 14 Procedure 19 18 4 13 12 Total 129 116 106 105 83 Number of PWR Plants 72 72 73 73 73 1

NUREG-1272, Vol. I 1, No.1 A-2-14

Reactors l I

Table A-2.10 Critical, On Line, Outage, and Availability Data for 1997 l Reactor Generator Forced Scheduled Unit l Reactor Critical On Line Outage Outage Availability '

Pl:nt Name Type Hours Hours liours flours Factor I

Arkansas 1 PWR 8204.8 8138.8 215.6 405.6 92.9 Arkansas 2 PWR 7323.2 7280.3 734.I 745.6 83.1 Beaver Valley 1 PWR 7223.3 7195.9 1496.1 68.0 82.1  ;

Beaver Valley 2 PWR 6191.5 6107.5 2181.7 470.8 69.7  ;

Big Rock Point BWR 4942.6 4818.7 3159.8 781.5 55.0 Braidwood 1 PWR 6127.1 6078.6 0.0 2681.4 69.4 ,

Braidwood 2 PWR 8664.0 8664.0 0.0 96.0 98.9  !

Browns Ferry 1 BWR 0.0 0.0 0.0 8760.0 0.0 ,

Browns Ferry 2 BWR 8505.0 8486.7 184.3 89.0 96.9  !

Browns Ferry 3 BWR 8320.0 8302.2 0.0 457.8 94.8 l Brunswick 1 BWR 7997.2 7955.3 0.0 804.7 90.8 ,

Brunswick 2 BWR 8331.5 8328.5 0.0 431.5 95.1 l Byron 1 PWR 8231.8 8185.1 79.0 495.9 93.4 l

Byron 2 PWR 8576.7 8513.2 0.0 246.8 97.2 Callaway PWR 7954.4 7893.3 123.8 742.9 90.1  !

Calvert Cliffs 1 PWR 8490.4 8442.1 113.4 204.5 96.4 Calven Cliffs 2 PWR 7090.4 7033.5 67.6 1658.9 80.3 Catawba 1 PWR 8678.4 8586.2 93.2 80.6 98.0 Catawba 2 PWR 7544.1 7451.5 289.3 1019.2 85.1 Clinton i BWR 0.0 0.0 288.0 8472.0 0.0 Comanche Peak l PWR 7767.0 7733.0 0.0 1027.0 88.3 Comanche Peak 2 PWR 8655.0 8626.0 31.0 103.0 98.5 Cookl PWR 6906.0 6820.3 530.2 1409.5 77.9  :

Cook 2 PWR 7967.1 7913.8 846.2 0.0 90.3 I Cooper Station BWR 7395.0 7337.2 129.8 1293.0 83.8

)

Crystal River 3 PWR 0.0 0.0 8760.0 0.0 0.0 Davis-Besse PWR 8236.2 8184.4 575.6 0.0 93.4 Diablo Canyon 1 PWR 7563.5 7507.1 194.2 1058.7 85.7 Diablo Canyon 2 PWR 8520.4 8489.4 270.6 0.0 96.9 Dresden 2 BWR 7924.4 7835.0 925.0 0.0 89.4 Dresden 3 BWR 4580.0 4423.0 2344.0 1993.0 50.5 Duane Arnold BWR 7456.6 7251.5 105.7 1402.8 82.8 Farley 1 PWR 6836.9 6803.7 0.0 1956.3 77.7 Farley 2 PWR 7189.6 7136.6 0.0 1623.4 81.5 Fermi 2 BWR 4114.0 3621.3 3862.5 1276.2 41.3 FitzPatrick BWR 7437.5 7243.0 367.4 1149.6 82.7 )

Fort Calhoun PWR 6972.7 6848.4 632.9 1278.7 78.2 0.0 94.6 Ginna PWR 8296.9 8289.3 470.7 '

Grand Gulf BWR 7815.8 7747.5 242.4 770.1 88.4 Haddam Neck PWR 0.0 0.0 0.0 2209.0 0.0 A 15 Appendix A-2

1997 AEOD Annual Report Table A 2.10 Critical, On Line, Outage, and Availability Data for 1997 Reactor Generator Forced Scheduled Unit Reactor Critical On Line Outage Outage Availability Plant Name Type liours liours llours llours Factor liarris PWR 6995.6 6872.0 361.4 1526.6 78.4 liatch 1 BWR 8667.1 8639.7 0.0 120.3 98.6 11ateh 2 BWR 7817.3 7709.0 70.0 981.0 88.0 11 ope Creek BWR 8144.0 8121.0 61.0 578.0 92.7 Indian Point 2 PWR 5687.0 5511.7 1530.5 1717.8 62.9 t

Indian Point 3 PWR 5135.1 4962.8 287.2 3510.0 6.7 Kewaunee PWR 2697.7 2657.1 0.0 6102.9 30.3 LaSalle 1 BWR 0.0 0.0 673.0 8087.0 0.0 '

LaSalle 2 BWR- 0.0 0.0 0.0 8760.0 00 Limerick i BWR 8659.0 8534.0 0.0 226.0 97.4 Limerick 2 BWR 7853.2 7508.3 466.4 785.3 85.7 Maine Yankee - PWR 1508.2 1502.1 1714.9 4823.0 18.7 McGuire 1 PWR 6228.4 6109.7 374.0 2276.3 69.7 McGuire 2 PWR 7821.8 7771.7 988.3 0.0 88.7 Millstone i BWR 0.0 0.0 8760.0 0.0 0.0 Millstone 2 PWR 0.0 0.0 8760.0 0.0 0.0 Millstone 3 PWR 0.0 0.0 8760.0 0.0 0.0 Monticello BWR 6729.0 6673.6 2086.4 0.0 76.2 Nine Mile Pt.1 BWR 6239.0 6152.0 994.7 1613.3 70.2 Nine Mile Pt. 2 BWR 7689.9 7537.1 251.1 971.8 86.0 North Anna 1 PWR 8017.2 7964.9 28.7 766.4 90.9 North Anna 2 PWR 7423.8 7389.3 1069.5 301.2 84.4 Oconee1 PWR 4644.3 4471.5 814.9 3473.6 51.0 Oconee 2 PWR 4853.8 4768.2 3991.8 0.0 54.4  ;

Oconee 3 PWR 4287.5 3765.3 2122.9 2871.8 43.0 l Oyster Creek BWR 7491.5 7306.9 951.3 501.8 83.4 Palisades PWR 6942.3 6747.1 684.6 1328.3 77.0 Palo Verde 1 PWR 7974.5 7937.1 103.2 719.7 90.6 Palo Verde 2 PWR 8160.0 8160.0 0.0 600.0 93.2 Palo Verde 3 PWR 7868.9 7830.3 30.4 899.3 89.4 Peach Bottom 2 BWR 8706.0 8562.0 147.0 51.0 97.7  !

Peach Bottom 3 BWR 8700.5 8688.0 72.0 0.0 99.2 l Perry BWR 7904.1 7823.7 480.3 456.0 89.3 l Pilgrim BWR 7348.7 7195.2 548.7 1016.1 82.1 j Point Beach 1 PWR 3382.6 3360.0 5400.0 0.0 38.4 Point Beach 2 PWR 895.4 788.5 382.7 7588.8 9.0 Prairie Island 1 PWR 8381.2 8370.0 353.6 36.4 95.5 Prairie Island 2 PWR 7233,0 7I82.8 617.2 960.0 82.0 Quad Cities 1 BWR 8141.1 8052.5 707.5 0.0 91.9 Quad Cities 2 BWR 5989.5 5829.5 193.6 2736.9 66.5 NUREG-1272, Vol. I 1, No. I A-2-16

Reactors Table A-2.10 Critical, On Line, Outage, and Availability Data for 1997 Reactor Generator Forced Scheduled Unit Reactor Critical On Line Outage Outage Availability Plant Name Type llours llours llours llours Factor River Bend BWR 7999.I 7906 5 398.5 455.0 90.3 Robinson 2 PWR 8312.1 8274.2 16.2 469.6 94.5 Salem i PWR 0.0 0.0 8760.0 0.0 0.0 Salem 2 PWR 1003.0 765.0 7995.0 0.0 8.7 San Onofre 2 PWR 5497.8 5430.0 402.0 2928.0 62.0 San Onofre 3 PWR 6244.4 6080.2 0.0 2679.8 69.4 Seabrook PWR 7618.7 7561.7 0.0 1198.3 86.3 Sequoyah1 PWR 7492.9 7431.5 37.1 1291.4 84.8 Sequoyah 2 PWR 8164.2 8085.3 674.7 0.0 92.3 South Texas I PWR 8255.2 8229.4 0.0 530.6 93.9 South Texas 2 PWR 8262.2 8185.9 152.0 422.1 93.4 St. Lucie i PWR 8610.5 8589.9 170.1 0.0 98.1 St. Lucie 2 PWR 7795.3 7756.4 0.0 1003.6 88.5 Summer PWR 8656.4 8628.6 131.4 0.0 98.5 Surry 1 PWR 7191.1 7068.5 285.5 1406.0 80.7 Surry 2 PWR 8454.5 8438.7 70.9 250.4 96.3 Susquehanna 1 BWR 7823.3 7596.7 616.8 546.5 86.7 Susquehanna 2 BWR 7278.3 7212.5 187.8 1359.7 82.3 Three Mile Isl 1 PWR 7979.2 7972.2 181.8 606.0 91.0 Turkey Point 3 PWR 7666.3 7571.8 111.9 1076.3 86.4 Turkey Point 4 PWR 8134.0 8047.0 87.2 625.8 91.9 Vermont Yankee BWR 7735.4 7645.1 279.4 835.5 87.3 Vogtle i PWR 7721.1 7655.5 348.9 755.6 87.4 Vogtle 2 PWR 8474.3 8427.2 48.2 284.6 96.2 Wash. Nuclear 2 BWR 6372.7 6248.9 14.8 2496.3 83.7 Waterford 3 PWR 6238.6 6161.4 0.0 2598.6 70.3 Watts Bar 1 PWR 7413.7 7361.2 422.9 975.9 84.0 Wolf Creek PWR 8690.0 8655.0 105.0 0.0 98.8 Zion 1 PWR 3447.3 3447.3 0.0 5312.7 39.4 Zion 2 PWR 0.0 0.0 0.0 8760.0 0.0 Total All Plants 706780.8 698160.1 105651.0 152517.9 73.1 Total BWR Plants 238108.3 234292.1 29569.2 60258.7 72.6 Total PWR Plants 468672.5 463868.0 76081.8 92259.2 73.4 Unit Availability Factor (Generator On.Line Hours + Unit Reserve Shutdown Hours) x 100 Penod flours Unit Reserve Shutdown llours The hours the unit was removed from on-hne operation for economic or other similar reasons w hen operation could hase continued. For 1997, this equals 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> for all plants except Wash. Nuclear 2.

w nich had 1081.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Period flours The gross hours from the beginning of the year or commercial operation, whichever comes last, to the end of the year or permanent shutdown, whichever comes first. For 1997, this equals 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> for all plants except fladdam Neck, which had 2209 hours0.0256 days <br />0.614 hours <br />0.00365 weeks <br />8.405245e-4 months <br />. and Maine Yankee. which had 8040 hours0.0931 days <br />2.233 hours <br />0.0133 weeks <br />0.00306 months <br />.

A-2-17 Appendix A-2

1997 AEOD Annual Report Table A-2.11 Capacity Factors for 1997 Net Electrical Capacity Capacity Reactor Energy h1DC DER Factor Factor Plant Name Type (GWil) (Net htWe) (Net h1We) (51DC Net) (DER Net)

{ Arkansas 1 PWR 6731.1 836.0 850.0 91.9 90.4 I Arkansas 2 PWR 6264.2 858.0 907.0 83.3 78.8 Beaver Valley 1 PWR 5850.4 810.0 835.0 82.5 80.0 Beaver Valley 2 PWR 4834.0 820.0 836.0 67.3 66.0 Big Rock Point BWR 287.4 67.0 72.0 49.0 45.6 Braidwood 1 PWR 6537.3 1120.0 l I 20.0 66.6 66.6 Braidwood 2 PWR 9613.9 1120.0 1120.0 98.0 98.0 Browns Ferry 1 BWR 0.0 0.0 1065.0 0.0 0.0 Browns Ferry 2 BWR 8786.0 1065.0 1065.0 94.2 94.2 Browns Ferry 3 BWR 85i8.7 1065.0 1065.0 91.3 91.3 Brunswick I BWR 6320.6 767.0 821.0 94.1 87.9 Brunswick 2 BWR 6366.0 754.0 821.0 96.4 88.5 Byron 1 l PWR 8754.1 1105.0 1120.0 90.4 89.2 Byron 2 PWR {

9187.8 1105.0 1120.0 94.9 93.6 Callaway l PWR 8587.2 1125.0 1171.0 87.1 83.7 Calven Cliffs 1 PWR 7203.4 835.0 845.0 98.5 97.3 3 Calven Cliffs 2 PWR 5882.9 840.0 845.0 79.9 79.5 Catawba 1 PWR 9733.3 1129.0 1I45.0 98.4 97.0 Catawba 2 PWR 8370.9 1129.0 1145.0 84.6 83.5 3

Clinton 1 BWR 0.0

)

930.0 933.0 0.0 0.0 l Comanche Peak 1 PWR 8426.2 1150.0 1150.0 83.6 83.6 Comanche Peak 2 PWR 9368.6 1150.0 1150.0 93.0 93.0 Cooki PWR 6636.1 1000.0 1020.0 75.8 74.3 Cook 2 PWR 8238.7 1060.0 1090.0 88.7 86.3 Cooper Station BWR 5460.4 764.0 778.0 81.6 80.1 Crystal River 3 PWR 0.0 8 I 8.0 825.0 0.0 0.0 Davis-Besse PWR 7181.0 873.0 906.0 93.9 90.5 Diablo Canyon 1 PWR 7983.8 1073.0 1086.0 84.9 83.9 Diablo Canyon 2 PWR 8953.9 1087.0 1119.0 94.0 91.3 Dresden 2 BWR 5564.8 772.0 794.0 82.3 80.0 Dresden 3 BWR 2843.2 773.0 794.0 42.0 40.9 Duane Arnold BWR 3606.5 520.0 538.0 79.2 76.5 Farley 1 PWR 5460.1 819.0 829.0 76.1 75.2 Farley 2 PWR 5862.0 822.0 829.0 81.4 80.7 Fermi 2 BWR 3641.9 946.0 1116.0 43.9 37.3 FitzPatrick BWR 5630.2 790.0 816.0 81.4 78.8 Fort Calhoun PWR 3116.7 478.0 478.0 74.4 74.4 Ginna PWR 4045.2 477.0 470.0 96.8 98.3 Grand Gulf BWR 9441.6 1195.0 1250.0 90.2 86.2 liaddam Neck PWR 0.0 560.0 582.0 0.0 0.0 NUREG-1272 Vol. I1.No. I A 18

! Reactors Table A-2.11 Capacity Factors for 1997 Net Electrical r aorcity Capacity Reactor Energy MDC DER P. -w Factor PI:nt Name Type (GWH) (Net MWe) (Net MWe) (MDt. '? il (DER Net)

Harris PWR 5804.7 860.0 900.0 77 i 73.6 Hatch 1 BWR 6864.9 301.0 822.0 C7.. 95.3 Hatch 2 BWR 6184.0 816.0 803.0 S 6..'- 87.9 Hope Creek BWR 8181.8 103).0 1067.0 4. 87.5 Indian Point 2 PWR 5000.4 94!.0 986.0 o , 57.9 Indian Point 3 PWR 4502.3 965.0 9650 .1 53.3 )

Kewaunee PWR I248.6 511.0 5.35.0 7.9 6.6 LaSalle i BWR 0.0 1036.0 1078.0 0.0 0.0 LaSalle 2 BWR 0.0 1036.0 1078.0 0.0 0.0 Limerick I BWR 9281.9 1105.0 1105.0 )5.9 95.9 l

78.7 78.7 i Limerick 2 BWR 7688.7 1115.0 1115.0 '

PWR I174 2 860.0 870.0 17.0 16.8 Maine Yankee  !

McGuire i PWR 6510.7 1129.0 1180.0 65.8 63.0 PWR 8513.1 1129.0 1180.0 86.1 82.4 McGuire 2 BWR 0.0 641.0 660.0 0.0 0.0 Millstone 1 Millstone 2 PWR 0.0 871.0 870.0 0.0 0.0 Millstone 3 PWR 0.0 1137.0 1154.0 0.0 0.0 BWR 3685.8 544.0 553.0 77.3 76.1 Monticello BWR 3636.4 565.0 613.0 73.5 67.7 Nine Mile Pt.1 BWR 8145.3 1105.0 1143.0 84.1 81.3 Nine Mile Pt. 2 PWR 7127.3 893.0 907.0 91.1 89.7 North Anna 1 PWR 6611.4 897.0 907.0 84.1 83.2 North Anna 2 I Oconee 1 PWR 3240.8 846.0 886.0 43.7 41.8 PWR 4003.6 846.0 886.0 54.0 51.6 Oconee 2 PWR 3053.9 846.0 886.0 41.2 39.3 Oconee 3 BWR 4515.0 619.0 650.0 83.3 79.3

. Oyster Creek PWR 5019.8 730.0 805.0 78.5 71.2 Palisades PWR 9745.5 1241.0 1257.0 89.6 88.5 Palo Verde i PWR 10058.4 1241.0 1257.0 92.5 91.3 Palo Verde 2 PWR 9487.0 1243.0 1261.0 87.1 85.9 Palo Verde 3 BWR 9371.8 1093.0 1119.0 97.9 95.6 Peach Bottom 2 BWR 8633.5 1093.0 1119.0 90.2 88.I Peact 'sottom 3 BWR 9018.5 1160.0 1191.0 88.8 86.4 Perry 4553.4 670.0 655.0 77.6 79.4 Pilgrim F.W R

?WR 1587.4 485.0 497.0 37.4 36.5 Point Beach I PWR 323.1 485.0 497.0 7.6 7.4 Point Beach 2 PWR 4307.6 513.0 530.0 95.9 92.8 Prairie Island 1 PWR 3646.6 512.0 530.0 81.3 78.5 Prairie Island 2 BWR 5839.5 769.0 789.0 86.7 84.5 Quad Cities 1 60.0 BWR 4147.8 769.0 789.0 61.6 Quad Cities 2 A-2-19 Appendix A-2

1997 AEOD Annual Report i

Table A 2.11 Capacity Factors for 1997 Net Electrical Capacity Capacity Reactor Energy MDC DER Factor Factor Plant Name Type (GWH) (Net MWe) (Net MWe) (MDC Net) (DER Net)

River Bend BWR 7335.7 936.0 936.0 89.5 89.5 Robinson 2 PWR 5880.3 683.0 700.0 98.3 9f.9 l Salem i PWR 0.0 1106 0 1115.0 0.0 R0 Salem 2 PWR 458.5 1106.0 1115.0 d .7 4.7 San Onofre 2 PWR 5792.3 1070.0 1070.0 61.8 61.8 San Onofre 3 PWR 6474.9 1080.0 1080.0 68.4 68.4 Seabrook PWR 8682.I 1158.0 1148.0 85.6 86.3 Sequoyahl PWR 8206.3 1117.0 1148.0 V,.9 81.6 Sequoyah 2 PWR 8887.9 1117.0 1148.0 90.8 88.4 South Texas 1 PWR 10248.2 1251.0 1251.0 93.5 93.5 South Texas 2 PWR 10106.I 12's.0 1251.0 92.2 92.2 St. Lucie i PWR 7184.0' 539.0 830.0 97.7 98.8 St. Lucie 2 PWR 6467.7 839.0 830.0 88.0 89.0 Summer PWR 8155,1 946.0 955.0 98.4 97.5 Surry 1 PWR 5639.7 601.0 788.0 80.4 81.7 Surry 2 PWR 6841.0 801.0 788.0 97.5 99.1 Susquehanna 1 BWR 8149.3 1090.0 1100.0 85.3 84.6 Susquehanna 2 BWR 772).1 1094.0 1100.0 80.6 80.1 Three Mile Isl 1 PWR 6451.2 786.0 819.0 93.7 89.9 Turkey Point 3 PWR 5245.0 693.0 720.0 86.4 83.2 Turkey Point 4 PWR 5621.1 693.0 720.0 92.6 89.1 Vermont Yankee BWR 3923.1 510.0 522.0 87.8 85.8 Vogtle i PWR 8885.7 1162.0 1169 0 87.3 86.8 Vogtle 2 PWR 9827.2 1162.0 1169.0 96.5 96.0 Wash. Nuclear 2 BWR 6203.I i107.0 1153.0 64.0 61.4 Waterford 3 PWR 6637.5 1075.0 1104.0 70.5 68.6 Watts Bar i PWR 7632.5 1113.0 1155.0 78.3 75.4 Wolf Creek PWR 10184.6 1163.0 1170.0 100.0 99.4 Zion 1 PWR 3378.2 1040.0 1040.0 37.1 37.1 Zion 2 PWR 0.0 1040.0 1040.0 0.0 0.0 Total All Plants 636224.2 907.7 936.6 71.7 70.2 <

Total BWR Plants 199547.9 840.9 894.3 70.4 68.4 Tetal PWR Plants 436676.3 942.0 958.2 72.3 71.1 MDC and DER Annual average valucs.

Capacity factor Net Electri, al Enerev x 100.000 or Net Electrical Encrev x 100.000 Period flours x MDC Net Period flours x DER Net Period flours The gross hours from the beginning of the year or commercial operation. whichever comes last, to the end of the year or permanent shutdown, whichever comes first. For 1997, this equals 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> for all plants except liaddam Neck, w hich had 2209 hours0.0256 days <br />0.614 hours <br />0.00365 weeks <br />8.405245e-4 months <br />, and Maine Yankee, which had 8040 hours0.0931 days <br />2.233 hours <br />0.0133 weeks <br />0.00306 months <br />. l l

l NUREG-1272, Vol. I1, No. I A-2-20

Reactors Table A-2.12 Industry Critical, On-Line, Outage, Availability, and Capacity Data Industry Data CY93 CY94 CY95 CY96 1997 Period Hours 949753.0 454840.0 954396.0 963M ?.0 956329.0 Reactor Critical Hours 724323.0 751614.2 778247.3 62864.0 706780.8 Generator On-Line llours 713214.3 741181.1 766413.7 753309.3 698160.1 Unit Reserve Shutdown Hours 4.0 12.9 578.1 997.4 1081.3 Forced Outage Hours 66907.6 749'7.2 48277.1 87020.1 105651.0 Scheduled Outage Hours 169631.1 i 'ad : 1.7 139705.2 123007.6 152517.9 Net Electrical Energy (GWH) 611207.7 6420'11.7 674478.6 674785.4 636224.2 Average MDC (Net MWe) 885.9 806.8 890.3 902.2 907.7 Average DER (Net MWe) 925.0 9 27.0 928.7 932.3 936.6 Availability Factor 75.1 77.6 80.4 78.3 73.1 Capacity Factor (MDC Net) 72.4 75.1 78.8 76.6 71.7 Capacity Factor (DER Net) 70.7 73.2 76.4 74.9 70.2 Period Hours The gross hours from the beginning of the year or commercial operation, whichever comes last, to the end of the year or petmanent shutdown, whichever comes first.

Unit Reserve Shutdown Hours The hours the unit was removed from on-line operation for economic or other similar reasons when operation could have continued.

Net Electrical Energy (GWH) Gross electrical output of the unit measured at the output terminals of the turbine generator during the reporting period, minus the normal station service electrical energy utilization. Negative quantities should not be used. The unit of measurement for this table is gigawatt-hours.

Maximum Dependable Capacity Dependable main-unit gross capacity, winter or summer, whichever is smaller, less the normal (MDC Net)(Net MWe) station service loads. The dependable capacity varies because the unit efficiency varies during the year due to cooling water temperature variations. It is the gross electrical output as measured at the output terminals of the turbine generator during the most restrictive seasonal conditions, less the normal station service loads. The unit of measurement for this table is megawatts.

Design Electrical Rating The nominal net electrical output of the unit specified by the utility and used for the purpose of plant design.The umt of measurement for this table is megawatts.

Availability Factor (Generator On.I.ine Hours + Unit Reserve Shutdown Hourd x 100 Pe.iod Hours Capacity Factor (MDC Net) Net Electrical Enerev x 10020 Penod Hours x MDC Net Capacity Factor (DER Net) Net Electrical Encrev x 100.000 Period Hours x DER Net A-2-21 Appendix A-2

Z Table A-2.13. Allegations at Commercial Nuclear Plant Sites for CY 1993 Through CY 1997 C Gj;

c e w

@ 1993 1994 1995 1996 1997 >

~ m Site Rc'd Open Sub Dise Mc'd Open Sub Disc Rc'd Open Sub Disc Rc'd Open Sub Disc O Rc'd Open Sub Disc c Arkansas 4 0 1 0 4 0 I 7 0 4 3 0 0 5 1 2 2 7 3 1 0 Heaver Valley 9 0 1 0 7 0 3 0 7 0 I 4 0 0 Braiduood 3 0 1 0 11 0 6 4 3 0 2 5

10 0 4 1 8 1 2 1 5

- ~

1 2 17 7 2 3

'z p

Browns Ferry 39 0 14 4 15 0 8 4 13 0 5 3 23 2 7 4 20 7 3 3  %

Brunswick 25 I lo 6 0 6 7 0 i 11 1 5 I 8 3 3 3 25 17 5 4 g '

Byron 3 0 0 0 8 0 1 4 0 2 0 6 0 4

• 1 I 12 0 2 0 '

Callaway 5 0 3 1 2 0 0 1 5 0 3 . 7 0 3 1 4 2 0 1 Calvert Clifffs 6 0 3 1 7 0 1 0 8 0 0 3 7 0 I 6 3 3 3 1 Catawba 3 0 1 0 7 0 3 0 2 0 0 5 0 0 1 2 7 1 2 0 Clinton 12 0 4 0 18 0 9 2 8 0 5 0 20 9 3 33 1 12 11 6 Comanche Peak 19 0 7 0 5 0 3 I 14 0 4 1 19 0 6 4 13 5 5 1 Cook 6 0 2 3 10 0 6 4 0 0 1 2 1 3 1 1 6 6 0 2 Cooper 3 0 2 0 15 0 5 5 11 0 7 0 i4 I 7 0 4 19 5 3 Crysta1 River 5 0 1 2 3 0 2 1 6 3 I 4 21 1 12 3 20 7 3 0 Davis-Beste 10 0 2 2 10 0 4 0 2 0 I 0

> I 11 2 1 5 1 0 1 (J Diablo Canyon 12 0 I i 13 0 8 2 17 1 8 1 21 2 5 4 17 8 4 to I Dresden 12 0 7 3 12 0 5 12 0 6 0

'" t; 1 12 o 2 18 4 4 2 Duane Arnold 2 0 2 0 2 0 1 1 2 0 1 I O O O O 3 2 1 0 Farley 1 0 0 0 2 0 0 0 3 0 1 0 6 0 2 0 2 1 1 0 Fermi 9 0 5 4 19 0 5 3 8 0 I 3 7 4 18 3 8 2 1 0 Fitipatrick 8 0 0 0 3 0 2 2 7 0 3 5 0 0 0 0 1 5 2 i Fort Calhoun 0 I 2 4 0 2 0 4 0 2 2 4 I 2 10 1 1 3 I Ginna 1 0 0 0 0 0 0 0 0 4 0 0 1 1 1 2 1 0 1 0 Grand Gulf 3 0 0 0 2 0 1 1 2 0 2 0 9 I 3 2 4 I I O ilarris 2 0 1 0 4 0 0 1 3 0 1 0 2 0 0 0 3 I I O Hatch 4 0 2 1 8 0 3 0 6 0 2 0 4 2 0 1 17 4 7 0 Indian Point 2 6 0 2 0 1 0 0 0 6 1 2 3 6 0 3 I il 3 1 0 indian Point 3 25 0 10 3 14 0 4 3 11 0 2 5 6 I i 2 1 15 2 2 Kewaunee 1 0 1 0 0 0 0 0 1 0 0 0 1 0 1 0 1 1 0 0 LaSalle 5 0 2 0 15 0 6 2 7 0 5 0 10 0 3 3 13 i

5 1 2 Limerick 16 0 7 2 12 0 4 0 4 0 1 0 5 0 4 I 9 5 0 i McGuire 0 0 0 0 8 0 1 4 3 0 1 0 4 0 1 0 3 2 0 0 Millstone 42 0 20 11 45 I 18 15 46 3 10 12 72 22 16 27 73 40 4 20 Monticello 2 0 2 0 0 0 0 0 0 0 0 0 6 0 0 0 2 1 1 0 Nine Mde Point 10 0 1 2 7 0 1 I 11 0 3 4 12 I I I 8 1 2 I North Anna 12 0 2 I 2 0 0 0 8 0 2 1 4 0 2 0 3 2 1 0 Oconee 1 0 0 0 4 0 2 0 1 0 0 1 8 1 2 1 2 1 1 I Oyster Creek 9 0 2 1 3 0 1 0 5 0 3 2 4 0 0 1 10 5 0 i Pahsades 5 0 1 1 6 0 2 1 8 0 2 0 5 0 2 0 3 2 0 1 Palo Verde 40 0 5 14 23 1 9 11 18 0 8 4 22 0 9 3 19 7 6 2

Table A-2.13. Allegations at Commercial Nucle:r Plant Sites for 1993 Through 1997 1993 1994 1995 1996 1997 Site Rc'd Open Sub Disc Rc'd Open Sub Disc Rc'd Open Sub Disc Rc'd Open Sub Disc Rc'd Open Sub Dise 6 0 8 0 2 2 3 0 1 0 4 1 2 1 5 2 0 1 Peach Bottom 1 1 8 0 2 2 7 0 3 2 2 0 1 1 10 2 3 3 8 6 0 2 Perry 0 4 2 9 0 2 8 0 1 2 5 0 1 0 2 2 0 2 Pilgrim 19 1 I 4 0 3 0 3 0 2 1 4 1 2 1 10 5 6 I Point Bexh I O O 0 2 0 0 3 0 0 7 0 2 0 5 2 1 0 Prairie lstand 4 0 1 1 1 0 0 9 0 5 7 0 3 0 13 1 5 1 14 4 2 3 Quad Cities 5 1 1 0 6 12 0 6 21 0 9 10 17 1 4 7 8 2 4 i River Bend 21 11 1 4 0 2 3 3 I I 3 0 0 0 1 0 1 0 2 1 0 0 Robinson 1 0 3 3 24 0 6 21 1 7 8 44 5 12 10 19 6 5 3 Salem /Ilope Creek 13 il 4 0 4 30 0 12 3 33 9 6 29 10 7 5 San Onofre 8 0 1 16 8 1 4 0 0 5 0 1 1 3 0 0 0 6 I 1 0 3 I I O Seabrm k 1 0 2 19 0 8 3 28 5 10 6 14 3 5 2 Sequoyah 31 0 7 11 7 1 0 16 16 21 0 9 7 15 0 6 5 25 6 5 7 11 5 3 1 South Texas 38 3 46 0 26 2 69 23 26 I

> St. Lucie 4 0 1 0 11 0 2 2 21 0

1 12 0 9 0 6 I 2 1 0 0 Summer 2 0 0 0 3 0 1 0 2 1 th d Surry 5 0 4 1 5 0 1 0 3 0 1 0 3 0 1 0 1 1 0 0 0 0 8 0 34 I 17 1 25 9 7 2 Susquehanna 7 0 2 2 8 2 1 1 0 0 0 0 0 0 2 1 0 0 4 0 2 1 Dree Mile lsland 4 0 1 2 1 1 4 9 0 4 0 12 0 5 0 16 3 6 2 19 7 5 I Turkey Point 13 0 7 0 2 0 0 5 0 2 2 9 I I I 4 2 0 0 Vermont Yankee 3 0 2 1 ,

4 0 0 0 3 0 6 0 5 0 11 4 2 I Vogtle 7 0 2 11 4 11 0 0 7 4 18 2 5 2 20 6 4 4 Washington Nuclear 8 0 4 4 6 0 4 14 0 2 0 16 0 11 5 21 0 7 7 11 3 1 2 Waterford 5 0 3 2 3 M 0 30 50 16 16 27 5 10 6 13 5 4 3 Watts Bar 69 0 33 18 18 3 6 0 5 0 12 0 7 0 6 0 1 1 13 I 5 0 20 12 1 2 Wolf Creek 0 0 17 0 5 4 8 0 4 I 14 2 6 0 51 32 13 11 Zion 7 1 Re'd. The total number of allegations receised during the year. Each allegation may contain mo:e than one concern.

go Open: %c number of allegations received during the ycar with one or more concerns remaining open.

Sub: The number of allegations fully or partially substantiated for that year. Partially substantiated means that not all the concerns were substantiated.

Diw: De number of allegations that include discrimination issues without regard to w hether they are substantiated. f

{

p- Note: De data are current as of December 31.1977.

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t' a I

APPENDIX B Summary of 1997 Abnormal Occurrences

Reactors CONTENTS Loss of Two of Three High Pressure Injection Pumps at Oconee Nuclear Station Unit 3. . B-1 B-iii Appendix B

Reactors NUREG-0090, VOLUME 20 Report No. 97-1 Loss of Two of HPI pumps would have automatically started and become inoperable very quickly. In addition, the Three High Presmre Injection Pumps pumps may have become air bound and unavailable ct Oconee Nuclear Station Unit 3 when the pump suction was transferred to the On May 3,1997, the Oconee Unit 3 reactor was Borated Water Storage Tank to inject into the RCS.

shut down and the reactor coolant system (RCS) This would have significantly complicated recovery fr m the accident, but would have been within the was being cooled down for inspection of the high pressure injection (HPI) discharge piping. The need Emergency Operating Procedure guidance and for the inspection resulted from RCS leakage from training provided to the operators. It would, how-a weld crack in the HPI makeup piping on Unit 2. ever, increase the probability of core damage. The Reactor pressure was approximately 270 psig, RCS length of time that Unit 3 was in this degraded temperature was approximately 205 F, one reactor status could not be accurately determmed, but the coolant pump (RCP) was running, and the Low c ndition may have existed since start-up in March Pressure Injection System was being used to cool 1997, when plant conditions required that the HPI down the RCS. Makeup water to the RCS to system be operable.

compensate for the temperature decrease was being The loss of the HPI pumps occurred when all of the supplied from the letdown storage tank (LDST) by water was inadvenently pumped from the LDST one of the three HPl pumps. Makeup to the LDST because of faulty level indication. The erroneous consisted of periodic batch additions as needed. level indication was caused by the loss of approxi-These plant conditions were below the point where mately one-half of the water in the level detector the technical specifications required that the HPI reference leg because of a slight leak in the instru-system must be operable; that is, required to ment fitting. This loss of the reference leg water mitigate a small-break loss-of-coolant accident. caused the tank level instrument to indicate a water l

level higher than the actual level, a condition that Plant cool-down evolutions appeared to be normal n2 y have existed since February 1997, the last time until the "B" HPI pump staned to cavitate and the reference leg was verified to be full. It a caused makeup flow to the reactor coolant system was lost. i the bss of the low-level alarm. As a result of these A RCP seal water (which is also supplied by the c nditions, the operators did not provide makeup HPI pump) low-flow signal automatically started _ ,

water t the tank when it was needed, resultmg m the "A" HPI pump. However, it also began to the HPI pump continuing to run until the tank was cavitate. (The third HPI pump is not designed to empty. The LDST level detection system consists of automatically stan on this signal and remained in two level instruments connected to a common the standby condition.) The operators stopped both reference leg. Thus, the condition affected both pumps and began troubleshooting the problem. A . level detectors equally.

Notification of Unusual Event was declared when it was recognized that the pumps would be inoperable In addition, the control room aperators did not past the shift that was on duty. Unit 3 pressure and properly monitor and detec inaccurate LDST temperature were stabilized and there was no level indications. They did n_. .iotice that for a  !

immediate concern that conditions would worsen. shon period of time the indicated level stopped decreasing and continuously showed the tank to be Later investigations revealed that the potential for a approximately half-full at the same time water was more serious situation existed if there had been a being pumped from the tank.

small break loss-of-coolant accident, which is the i design basis for the HPI system, prior to this event. Corrective actions ircluded (1) the addition of a If such an accident had occurred, all three of the second reference leg to the LDST to provide B-1 Appendix B

1997 AEOD Annual Report separate level indications, (2) enhanced operator Level 11 violation existed, and the licensee was j training and procedures, and (3) the performance of assessed a $330,000 civil penalty. Information i an HPI System Reliability Study that is to be Notice 97-38," Level-Sensing System Initiates completed by December 31,1997. Common-Mode Failure of liigh-Pressure-Injection Pumps," was issued on June 24,1997, to alert other An escalated enforcement, which incorporated this licensees to this event.

issue, resulted in the determination that a Severity c

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l NUREG 1272, Vol.11, No.I B-2 I

APPENDIX C Reports Issued in 1997

1 Reactors l l

Table C 1 Reports Issued in 1997 D te Title No. Author Case Studies 03/97 Grid Performance Factors C97-01 M. Wegner SpecialStudies 12/97 Office for Analysis and Evaluation of Operational Data NUREG-1272, Annual Report,1996 Vol.10 04/97 Precursors to Potential Severe Core Damage Accidents: NUREG/CR-4674, ,

1995, A Status Report Vol.23 l 04/97 Precursors to Potential Severe Core Damage Accidents: NUREG/CR-4674, 1982 - 83, A Status Repon Vol.24 12/97 Precursors to Potential Severe Core Damage Accidents: NUREG/CR-4674, 1996, A Status Report Vol.25 01/98 Performance Indicators for Operating Commercial Nuclear Power Reactors, l J

03/97 Oconee Electrical System Design and Operation S97-01 G. Lanik }

H. Ornstein W. Raughley J. Thompson 06/97 Reactor Core Isolation Cooling System Reliability, S97-02 1987 - 1993 1

06/97 Fire Events-Feedback of U.S. Opeiating Experience S97-03 J. Houghton l

I Engineering Evaluations 04/97 Review of Industry Effons to Manage Pressurized Water NUREG/CR-6456 Reactor Feedwater Nozzle, Piping, and (E97-01)

Feedring Cracking and Wall Thinning INEL-96/0089 09/97 Review of Undetected Failures of Safety Systems E97-02 S. Pullani E. Brown 12/97 Nuclear Power Plant Cold Weather Problems and E97-03 M. Padovan Protective Measures C-1 Appendix C

l 1997 AEOD Annual Report j Table C-1 Reports Issued in 1997 Date Titic No. Author I TechnicalReviews 01/97 Design Errors in Nuclear Power Plants T97-01 S. Pullani l

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NUREG-1272, Vol. I1 No. I C-2

i APPENDIX D Reports Issued From CY 1980 Through CY 1996

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Reactors CONTENTS Table D 1 Reports issued in CY 1996 . . D- 1 Table D-2 Reports Issued in CY 1995 . . D-3 Table D-3 Reports Issued in CY 1994 . . D-4 Table D-4 Reports issued in CY 1993 . .D-5 Table D-5 Reports issued in CY 1992 . . D-6 Table D-6 Reports Issned in CY 1991 . D-8 1 1

Table D-7 Reports Issued in CY 1990 . . D-9 Table D-8 Reports issued in CY 1989 . . . D-11 Table D-9 Reports issued in CY 1988 . . D-14 i l

Table D-10 Reports issued in CY 1987 . . D-16 Table D-11 Renorts Issued in CY 1986. . D-18 Table D-12 Reports Issued in CY 1985 . . D-21 Table D-13 Reports issued in CY 1984 . . . D-24 Table D-14 Reports issued in CY 1983 . . D-28 Table D-15 Reports issued in CY 1982 . .D-33 l

Table D-16 Reports Issued in CY 1981 . D-37 Table D-17 Reports issued in CY 1980. . D-40 D-iii Appendix D

l Reactors l

Table D-1 Reports Issued in CY 1996 Date Title No. Author SpecialStudies 12/95 Precursors to Potential Severe Core Damage Accidents: NUREG/CR-4674 l 1994 A Status Report Vols. 21 and 22 i 01/96 Performance Indicators for Operating Commercial Nuclear Power Reactors Data: Through September 1995 01/96 Evidence of Aging Effects on Certain Safety-Related NUREG/CR-6442 Components 08/96 Office for Analysis and Evaluation of Operational Data - NUREG-1272, .

Annual Report,1994-FY 95 Vol. 9, No.1 08/96 Isolation Condenser System Reliability, 1987-1993 S96-01 (INEL-95/0478) 09/96 Assessment of Spent Fuel Cooling NUREG-1275, J. Ibarra Vol.12 W. Jones (S96-02) G. Lanik i H. Ornstein S. Puilani 02/96 Emergency Diesel Generator Power System S95.03 Reliability 1987-1993 (INEL-95/0035)

Engineering Evaluations 03/96 Motor-Operated Valve Key Failures E96-01 C. Hsu 04/96 Analysis of Allegation Data E96-02 S. Israel 06/96 Analysis of Allegation Data E96-02 Supplement 1 S. Israel 04/96 Steam Generator Tube Failures NUREG/CR-6365 (E96-03)

TechnicalReviews 03/96 Technical Review Repon - AEOD Technical Reports T96-01 S. Israel by Category 03/96 Technical Review Repon - AEOD Technical Report T96-01 S. Israel by Category Revision 1 04/96 Technical Review Report - Target Rock Two-Stage T96-02 M. Wegner SRV Performance Update D-1 Appendix D

1997 AEOD Annual Report l

Table D-1 Reports Issued in CY 1996 Date Title No. Author 08/96 Technical Review Report - Response of Babcock & T96-03 W. Raughley Wilcox Company Plants Following a Loss of Nonemergency AC Power 1

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NUREG-1272, Vol. I1, No.1 D-2

Reactors i

Table D-2 Reports Issued in CY 1995 Date Title No. Author SpecialStudies 02/95 liigh-Pressure Coolant Injection (HPCI) System INEL-94/0158 Perfonnance, 1987-1993 Final 03/95 Reactor Coolant System Blowdown at Wolf Creek S95-01 J. Kauffman on September 17,1994 S. Israel Engineering Evaluations 07/95 Operating Events With Inappropriate Bypass or Defeat E95-01 J. Kauffman of Engineered Safety Featu es TechnicalReviews 03/95 Major Disturbances on the Westem Grid and Related Events T95-01 M. Wegner 10/95 Potential Damage to Low-Pressure Injection Valves T95-02 E. Brown During Surveillance Testing 10/95 Review of the National Tre.nsportation Safety Board's T95-03 J. Kauffman Safety Study NTSB/SS-9&Ol, A Review of Flightcrew-involved, Major Accidents of U.S. Carriers, 1978 Through 1990 D-3 Appendix D

1997 AEOD Annual Report Table D-3 Reports Issued in CY 1994 Date Title No. Author SpecialStudies i1/94 Office for Analysis and Evaluation of Operational Data NUREG-1272, 1993 Annual Report, Power Reactors Vol. 8, No. I 10/94 Operating Experience Feedback Repon - Reliability of NUREG-1275, Safety-Related Steam Turbine-Driven Standby Pumps Vol.10 (S94-01) J. Boardman 09/94 Operating Experience Feedback Report - NUREG-1275, H. Ornstein i Turbine-Generator Overspeed Protection Systems Vol.11 (S94-02)

TechnicalReviews 03/94 The Electrical Transient Which Followed the T94-01 M. Wegner Los Angeles Earthquake -January 17,1994 05/94 Review on Mispositioned Equipment Events T94-02 S. Israel 07/94 Computer-Based Digital System Failures T94-03 E. Lee 12/94 Potential for Boiling Water Reactor Emergency T94-04 J. Boardman Core Cooling System Strainer Blockage Due to Loss-of-Coolant Accident Generated Debris

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NUREG-1272, Vol. I 1. No. I D-4

Reactors Table D 4 Reports Issued in CY 1993 Date Title No. Author SpecialStudies 07/93 Office for Analysis and Evaluation of Operational Data NUREG-1272, 1992 Annual Report, Power Reactors Vol. 7, No.1 04/93 Review of Auxiliary Feedwater System Reliability S93-01 J. Houghton D. Rasmuson J. Boardman Not Issued S93-02 04/93 Operating Experience Feedback - Service Water S93-03 J. Houghton System FaiLires and Degradations Not Issued S93-04 04/93 Operational Data Analysis of Shutdown and Low S93-05 R. Prato Power Licensee Event Reports 12/93 Potter & Brumfield Model MDR Rotary Relay Failures S93-06 R. Spence Engineering Evaluations 02/93 aights from Common-Mode Failure Events E92 02 S. Israel applement 1 02/93 .iuman Factors Aspects of Boiling Water E93-01 J. Kauffman Reactor Reactivity Management Events During Power Operations 03/93 Evaluation of Loss of Offsite Power Due to E93-02 S. Mazumdar Plant-Centered Events 12/93 Electrical Inverter Operating Experience 1985 - 1992 E93-03 J. Ibarra s

TechnicalReviews 06/93 Primary Systera Integrity, Pressurized Water Reactor T93-01 J. Kauffman Coolant System Leaks J. Stuller 08/93 Tardy Licensee Actions T93-02 S. Israel 12/93 Loss of Annunciator and Computer System Events T93-03 J. Ibarra 12/93 U.S. Nuclear Regulatory Commission Review of T93-04 H. Ornstein Operating Experience D-5 Appendix D

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1997 AEOD Annual Report Table D 5 Reports Issued in CY 1992 Date Title No. Author Case Studies 12/92 Operating Experience Feedback Report - Human NUREG-1275, J. Kauffman Performance in Operating Events Vol.8 G. Lanik (C92-01) R. Spence E. Trager SpecialStudies 07/92 Office for Analysis and Evaluation of Operational Data NUREG-1272, 1991 Annual Report, Power Reactors Vol. 6, No.1 Not issued S92-01 04/92 Safety and Safety / Relief Valve Reliability S92-02 M. Wegner 06/92 Review of Operational Experience with Molded S92-03 J. lioughton Case Circuit Breakers in U.S. Commercial Nuclear W. Leschek Power Plants P. O'Reilly D. Rasmuson Not issued S92-04 Not issued S92-05 Not issued S92-06 09/92 Operating Experience Feedback Report - Experience NUREG-1275, with Pump Seals Installed in Reactor Coolant Pumps Vol.'

Manufactured by Byron Jackson 12/92 Operatirig Experience Feedback Report - Pressure 1iUREG-1275, C. Hsu Locking and Thermal Binding of Gate Valves V il. 9

(!,92-07)

Engineering Evaluations 05/92 Inadequate Management Control of Snubber Surveillance E 92-01 C. Hsu 06 92 Insights From Common-Mode Fai!ure Events 7 2-02 S. Israel TechnicalReviews 01/92 Enhanced Setpoint Testing Procedures for Pressurizer T92-01 M. Wegner Safety Valves at Oconee imd Catawba 01/92 BWR-5 and 6 Events Applicable to Laguna Verde T92-02 J. Kauffman N. Casas 06/92 Solenoid-Operated Valves and Related Equipment - T92-03 H. Ornstein A Status Report NUREO.!272, Vol.11. No.1 D-6

Reactors f Table D-5 Reports Issued in CY 1992 Date Title No. Author TechnicalReviews (Continued) 1 06/92 Recent Solenoid-Operated Valve Experiences involving T92-04 H. Ornstein Maintenance and Testing Deficiencies 06/92 Errors in Effective Reactor Trip Settings or Monitoring T92-05 S. Israel Associated with Excore Instrumentation 09/92 Water Intrusion into Sensitive Control Room Equipment T92-06 J. Kauffman 09/92 Inope ability of the Standby Liquid Control System T92-07 L. Gundrum During Surveillance Testing at Nine Mile Point Unit 2 10/92 Emergency Diesel Generator Start Frequency T92-08 T. Cintula a

11/92 Review of Manual Valve Failures T92-09 S. Salah 12/92 Prospective Trend of Low R.eliability Emergency T92-10 T. Cintula Diesel Generators D-7 Appendix D

1997 AEOD Annual Report Table D 6 Reports Issued in CY 1991 Date Title No. Author SpecialStudies 07/91 Office for Analysis and Evaluation of Operational Data NUREG-1272, 1990 Annual Report, Power Reactors Vol. 5, No. I 1

09/91 Performance of Emergency Diesel Generators in S91-01 T. Cintula Restoring Power to Their Associated Safety Buses-A Review of Events Occurring at Power Engineering Evaluations 02/91 A Review of Water Hammer Events After 1985 E91-01 E. Brown TechnicalReviews 02/91 Causes ofIncorrect Sp tem Flows T91-01 S. Israel 02/91 Incorrect Rotation of PDP T91-02 T. Cintula 03/91 Overloaded Emergency Buses T91-03 S. Israel 04/91 Turbine Overspeed Trip Due to Steam Valve Leakage T91-04 C. Hsu and Condensate 05/91 Setpoint Testing of Pressurizer Safety Valves With T91-05 M. Wegner Water-Filled Loop Seals 06/91 Deficiencies in External Flood Protection T91-06 S. Israel 07/91 Evaluation of Partial Loss of Station Power Events at T91-07 F. Manning Prairie Island Unit No. 2 on December 21 and December 26,1989 NUREG-1272, Vol. I 1, No. I D-8 )

Reactors Table D 7 Reports Issued in CY 1990 Date Title No. Author Case Studies 10/90 Operating Experience Feedback Report - NUREG-1275. H. Ornstein Solenoid-Operated Valve Problems at U.S. Light Vol. 6 Water Reactors (C90-01)

SpecialStudies 07/90 Office for Analysis and Evaluation of Operational Data NUREG-1272, 1989 Annual Report, Power Reactors Vol. 4, No.1 03/90 Review of Thermal Stratification Operating Experience S902 T.Su 08/90 Recurrence of Important Safety Issues Reported in LERS S90-01 S. Israel Engineering Evaluations 02/90 Failures of Electrical Supply and Power Generation E90-01 M. Wegner Equipment Which Disrupted Plant Function at Nuclear Power Plants 02/90 Crosby Low Pressure Relief Valves E90-02 S. Israel 05/90 Overpressurization of Auxiliary Feedwater Systems E90-03 C. Hsu 04/90 Swelling and Cracking in Hafnium Control Rods E90-04 M. Wegner 05/90 Operational Experience on Bus Transfer E90-05 S. Mazumdar 07/90 Potential for Rcsidual Heat Removal System Pump E90-06 C. Hsu Damage 07/90 Effect of Internal Flooding of Nuclear Power Plants on E90-07 N.Su Safety-Equ:pment 09/90 Low Temperature with Overpressure Protection: Testing E90.08 S. Israel PORVs with Alternate Pneumatic Supply S. Salah 10/90 Additional Factors Affecting the Lift Setpoint of E90-09 L. Padovan Pressurizer Safety Valves 12/90 Evaluation of Boiling Water Reactor Mode Switch Events E90-10 W. Jones TechnicalReviews 01/90 PNO's Issued in First Quarter of 1989 (Proprietary) T90-01 R. Dennig T. Wolf 01/90 insights Regarding Commonwealth Edison Plant T90-02 N. Thomasson Root-Cause Determinations Related to Maintenance Effectiveness (Proprietary)

D-9 Appendix D

- 1997 AEOD Annual Report Table D-7 Reports Issued in CY 1990 Date Title No. Author j TechnicalReviews (Continued) 03/90 Improper Installation of Heat Shrinkable Tubing T90-03 S. Mazumdar 03/90 Reverse (Backward) Acting Valve Manual Handwheels T90-04 T. Cintula  :

03/90 Association Between Nuclear Plant Utilization and i Incentive Regulation by Station Public Utility Commissions T90-05 S. Stern 05/90 Aquatic Life in Emergency Cooling Ponds T90-06 L. Padovan ,

05/90 Reversed Sensing Lines Connections T90-07 B. Kaafer 06/90 Turbine Bypass Malfunctions T90-08 B. Kaufer i 06/90 Inadvertent Partial Draining of Condensate Storage Tanks T90-09 T. Cintula 07/90 Evaluation of Maintenance Trends at Five Selected Sites T90-10 P. O'Reilly (Proprietary) 07/90 Evaluation of Safety Equipment Outages for Signincance T925A F. Manning ,

at Zion (Revised) 08/90 Effect of High Energy Line Breaks on Chilled Water T90-1 i L. Padovan Systems at Nuclear Power Plants 09/90 Loss of Offsite Power to Comply With NRC Regulations T90-12 T. Cintula ,

10/90 Corrosion and Failure of Service Water Pump Impeller T90-13 C. Hsu Snap Rings 10/90 Seal Problems in Boric Acid Transfer Pumps T90-14 S. Israel 10/90 Salem I and 2 Evaluation of Operating Experience T90-15 P. O'Reilly (Proprietary) 11/90 Impact of Pipe Liner Failure of Pump Operation T90-16 S. Israel 12/90 Inadvertent Containment Spray Actuations T90-17 M. liarper

- NUREG-1272,Vol. I1, No. I D-10 l

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Table D-8 Reports issued in CY 1989 Date Title No. Author SpecialStudies 06/89 Office for Analysis and Evaluation of Operational Data NUREG 1272, 1988 Annual Report, Power Reactors Vol. 3 No. I 03/89 Operating Experience Feedback Report -Technical NUREG-1275, P. O'Reilly Specifications Vol.4 G. Plumlee 03/89 Operating Experience Feedback Report - Progress in NUREG-1275, L. Bell Scram Reductions Vol.5 P. O'Reilly 08/89 Operating Experience Feedback Report - Progress in NUREG-1275, L. Bell Scram Reduction Vol. 5 Addendum 01/89 Application of the NPRDS for Effectiveness Monitoring S804B P. O'Reilly (Appendices A and B are Proprietary) T. Wolf P. Cross-Prather )

02/89 Maintenance Programs at Nuclear Power Plants S901 M. Chiramal (Table 2 is Proprietary) Revision 1 S. Israel M. Wegner S. St,:rn Engineering Evaluations 02/89 Problems With Oils, Greases, Solvents and Other E901 S. Israel Chemical Materials 03/89 Fire and Explosive Mixtures Resulted from introduction E902 H. Ornstein of Hydrogen Into Plant Systems Not issued E903 04/89 On Demand Malfunctions of HPCI and RCIC E904 T. Cintula 06/89 Electrical Bus Bar Failures E905 M. Padovan 03/89 Failure of Steam Generator Isolation Check Valve E906 T. Cintula 09/89 Diversion of Seal Cooler Flow for RHR Pumps E907 S. Israel 10/89 Excessive Valve Body Erosion at Brunswick E908 E. Brown 12/89 Operator Actions During Operational Events E909 S. Israel 12/89 Potential for Gas Binding of High Head Safety E910 M. Padovan injection Pumps Resulting From Inservice Testing of VCT Outlet Isolation Valves D-11 Appendix D

1997 AEOD Annual Report Table D-8 Reports Issued in CY 1989 Date Title No. Author TechnicalReviews 01/89 Millstone Unit 1-Safety / Relief Valve Discharge Lin:- T901 T.Su l Vacuum Breakers Failed Open  ;

02/89 Inadvenent Reactor Trips Due to RCS Flow T902 M. Padovan l Instrumentation Maintenance Activities '

03/89 Generic implication of Browns Ferry Fire on T903 T.Su November 2,1987 04/89 Design Deficiency of Safety Injection Block Switch '1904 S. Mazumdar 04/89 Broken Lifting Beam Bolts in HPCI Terry Turbine T905 S. Mazumdar 04/89 Broken l'ifting Beam Bolts in HPCI Terry Turbine T905 T. Cintula 04/89 Component Degradation Due to Indiscriminate Painting T907 M. Padovan A nonreactor report (see NUREG-1272, Vol. 4, No. 2) T908 05/89 Operating Events involving Dampers T909 S. Israel 06/89 Investigation of Cracked Control Rod Drive Seal T910 W. Jones Housing at Palisades 06/89 Evaluation of Individually Reponed Safety System LFRs T911 F. Manning for Their Combined Significance 06/89 Selected Maintenance Rework T912 S. Israel 07/89 Comparison of the Proposed Maintenance Effectiveness T913 N. Thomasson (ME) Indicator With Catawba and Farley Nuclear Plants T. Wolf Regarding Inspections (Proprietary) M. Harper 09/89 Overview of Design / Installation Fabrication Errors in T914 S. Israel 1988 09/89 EDG Ground Fault Detection and Trip-Circuit at Perry T915 S. Mazumdar Unit 1 09/89 Debris in Containment Recirculation Sumps T916 M. Padovan Not issued (refer to E908) T917 09/89 Check Valve Failure Rates From NPRDS Data T918 E. Brown 09/89 Failure of Overcurrent Protective Device at Palisades Unit 1 T919 S. Mazumdar Not issued T920 10/89 Inadequaa Capacity of 4160V Switchgear at FitzPatrick T921 S. Mazumdar 11/89 Failure of HPCI Turbine Due to High Moisture in Lube Oil T922 C. Hsu NUREG-1272 Vol. I1, No.1 D-12

Reactors Table D 8 Reports Issued in CY 1989 Date Title No. Author TechnicalReviews (Continued) 1I/89 Delaminating Foil Insulation in Primary Containment T923 T. Cintula Not issued T924 12/89 Evaluation of Safety Equipment Outages for Significance T925 F. Manning ,

= at Zion i 12/89 Evaluation of Two Beaver Valley 2 Nuclear Plant T926 F. Manning Equipment Degradation Events forTheir Combined Significance 12/89 Follow-up on Steam Binding of AFW Pumps T927 C. Hsu 12/89 Inadequate Overpressure Protection for Auxiliary Steam T928 S. Salah Headers at the Oconee Plants D-13 Appendix D

1997 AEOD Annual Report Table D 9 Reports Issued in CY 1988 Date Title No. Author Case Studies 08/88 Operating Experience Feedback Report - Service NUREG-1275, P. Lam I Water System Failures and Degradations in Light Water Vol.3 E. Leeds Reactors i

SpecialStudies 03/88 Significant Even s That Involved Procedures S801 E. Trager 03/88 Operational Experience Feedback Evaluation Rancho S802 G. Plumlee Seco Nuclear Generating Station, Restart 06/88 AEOD Concems Regarding the Power Oscillation Event SD3 J. Kauffman at LaSalle 2 (BWR-5) 08/88 Preliminary Results of the Trial Program for S804A Maintenance Performance Indicators 09/88 Repon to the U.S. Nuclear Regulatory Commission on NUREG-1272 Analysis and Evaluation of Operational Data - 1987 Vol. 2, No. I Power Reactors (S804)

Not issued S805 Not issued S806 A nonreactor report (see NUREG-1272. Vol. 3, No. 2) S807 Engineering Evaluations 04/88 BWR Overfill Events Resulting in Steam Line Flooding E801 J. Kauffman 05/88 Design and Operating Deficiencies in Control Room E802 S. Israel Emergency Ventilation systems 08/88 Inadequate NPSH in High Pressure Safety Injection E803 S. Israel Systems in PWRs 08I88 Reliability of Recirculation Pump Breaker During E804 T.Su '

an ATWS

• 09/88 Potential LOCA Due to Energized Uncovered E805 T. Cintula Pressurizer Heaters '

' 10/88 Loss of Decay Heat Removal Due to Rapid Refueling E806 M. Padovan Cavity Pumpdown 10/88 Pump Damage Due to Low Flow Cavitation E807 C. Hsu 12/88 Operational Experience Review of Potential Large E808 T. Cintula Openings in Containment NUREG-1272. Vol. I 1, No.1 D-14

Reactors

- Table D 9 Reports Issued in CY 1988 Date Title No. Author lr TechnicalReviews 01/88 Perry Nuclear Power Plant Unit 1 - Unexpected MSIVs T801 T.Su l Closure and Reopening Not issued T802 )

05/88 Summary of Early Operational Experience of Foreign T803 P. O'Reilly l

Commercial Nuclear Reactors (Proprietary) 2 05/88 " Precursor" Operational Events That Occurred From T804 F. Manning }'

November 1,1987,Through March 1988 05/88 Insights From Significant Events in 1987 T805 S. Israel 05/88 Recent Operational Experience Trends at Fermi 2 T806 T. Wolf 06/88 Recent Operational Experience Trends at Indian Point 2 T807 T. Wolf 06/88 A Technical Basis for Granting Test Frequency Relief T808 G. Plumlee 06/88 Blocked Thimble Tubes / Stuck Incore Detector T!'09 M. Wegner 07/88 An Analysis of NPRDS Data for flatch Plant T810 T. Wolf (Proprietary) P. Cross-Prather 11/88 Degradation of Ice Condenser Containment T811 F. Manning  !

Functional Capability incident investigation Program Reports 02/88 Incident Investigation Manual NUREG-1303 l l

k D-15 Ar,pendix D

1997 AEOD Annual Report Table D 10 Reports Issued in CY 1987 Date Title No. Author Case Studies 03/87 Operating Experience Feedback Report - Air Systems NUREG-1275, H. Ornstein l Problems at U.S. Light Water Reactors Vol.2 (C701)

SpecialStudies 05/87 Report to the U.S. Regulatory Commission on Analysis NUREG-1272 and Evaluation of Operational Data - 1986 (S701) 05/87 Loss of Decay Heat Removal Function at Pressurized S702 H. Ornstein Water Reactors With Panially Drained Reactor Coolant Systems A nonreactor report (see NUREG-1272, Vol. 2, No. 2) S703 Engineering Evaluations 01/87 Potential Containment Airlock Window Failure Due E701 S. Israel to Radiation 03/87 MOV Failure Due to Hydraulic Lockup From Excessive E702 E. Brown Grease in spring Pack 03/87 Loss of Offsite Power Due to Unneeded Actuation of E703 F. Ashe Startup Transformer Protection Differential Rely 03/87 Discharge of Primary Coolant Outside of Containment at E704 S. Israel PWRs While on RHR Cooling 03/87 RWCU System Automatic Isolation and Safety E705 N. Thomasson Considerations 03/87 Inadequate Mechanical Blocking of Valves E706 T. Cintula 03/87 Design and Construction Problems at Operating E707 C. Hsu Nuclear Plants 08/87 Depressurization of Reactor Coolant Systems at PWRs E708 S. Israel 08/87 Auxiliary Feedwater Pump Trips Due to Low E709 C. Hsu Suc9on Pressure 10/87 Irodequate NPSH in Low-Pressure Safety Systems E710 S. Israel iz PWRs Program Support Feports 07/87 Operational Experiences at Newly Licensed Nuclear NUREG-1275, Power Plants Vol.1 R. Dennig ,

1 I

NUREG-1272, Vol. I 1, No. I D- 16

, . . . - - . = . - . - - - - . . . . . - .

Reactors Table D 10 Reports Issued in CY 1987 Date Title No. Author TechnicalReviews 01/87 Compression Fitting Failures T701 H. Ornsteic 03/87 Leaking Pulsation Dampener Leads to Loss of T702 T. Cintula Charging System 03/87 Potential for Loss of Emergency Feedwater Caused by T703 M. Wegner Pump Runout During Certain Transients 03/87 Pressurizer Code-Safety Valve Reliability T704 M. Wegner 03/87 Pressurizer Code-Safety Valve Reliability T704 M. Wegner 05/87 Occurrence of Events Involving Wrong Unit / Wrong T705 E. Trager Train / Wrong Component-Update Through 1986 06/87 Recent Events involving Turbine Runbacks at PWRs T706 E. Leeds 08/87 Undetected Loss of Reactor Water T707 S. Israel l

08/87 Problems With M-Fh Pressure Safety Injection Systems T708 S. Israel in Westinghouse PWRs 10/87 Recent New P! ant Operational Experience T709 T. Wolf 11/87 Heating Ventilating, and Air Conditioning System Problems T710 M. Chiramal A nonreactor report (see NUREG-1272, Vol. 2, No. 2) T711 11/S7 Unplanned Criticality Events at U.S. Power Reactors T712 T. Wolf Similar to That at Oskarshamm Unit 3 on 07/30/87 12/87 Mispositioning of" Reverse Acting" Valve Controliers T713 J. Stewart A nonreactor ieport (see NUREG-1272, Vol. 3, No. 2) T714 i

l D-17 Appendix D

1997 AEOD Annual Report Table I)-11 Reports issued in CY 1986 Date Title No. Author Case Studies A nonreactor report (see NUREG-1272, May 1987) C601 08/86 Operational Experience involving Tan.. : Overspeed Trips C602 C. Hsu 12/86 A Review of Motor-Operated Valve Performance C603 E. Brown 12/86 Effects of Ambient Temperature on Electronic C604 M. Chiramal l Components in Safety-Related Instrumentation and Control Systems 12/86 Operational Experience Involving Losses of C605 F. Ashe Electrical Inverters SpecialStudies 04/86 AEOD Annual Report for 1985 S601 J. Heltemes 05/86 An Overview of Nuclear Power Plant Operating S602 J. Crooks l Experience Feedback Programs 06/86 ' Adequacy of the Scope ofIE Bulletin 86-01 S603 E. Leeds Engineering Evaluations 05/86 Core Damage Precursor Event at Trojan E514 Revision 1 D. Zukor 01/86 Deficient Operator Actions Following Dual Function E601 E. Leeds j Valve Failures 01/86 Unexpected Criticality Due to Incorrect Calculation and E602 E. Leeds Failure to Follow Procedures

• 02/86 Delayed Access to Safety Related Areas During E603 T. Cintula i Plant Operation .

03/86 Spurious System Isolations Due to the Panalarm Model E604 E. Leeds  !

86 Thermocouple Monitor l

l:  ;

04/86 Lightning Events at Nuclear Power Plants E605 M. Chiramal 05/86 Loss of Safety Injection Capability at Indian Point Unit 2 E606 R. Tripathi !

l 07/86 Degradation or Loss of Charging Systems With Swing E607 F.Ashe Pump Designs i

07/86 Reexamination of Water Hammer Occurrences E608 E. Leeds l

NUREG-1272, Vol. I 1, No. I D-18

Reactors Table D-11 Reports Issued in CY 1986 DIte Title No. Author Engineering Evaluations (Continued) 08/86 Inadvenent Draining of Reactor Vessel During E609 P. Lam Shutdown Cooling Operation 08/86 Loss of Low Pressure Coolant Injection Loop Selection E610 E. Leeds Logic at Millstone Unit 1 10/86 Deficiencies in Seismic Anchorage for Electrical and E611 N. Thomasson Control Panels 12/86 Emergency Diesel Generator Comporent Failures Due E612 C. Hsu to Vibration 12/86 Localized Rod Cluster Control Assembly Wear at E613 E. Brown PWR Plants Program Support Reports 01/86 Trends and Patterns Program Plan - FY86-FY88 P601 R. Dennig 08/86 Trends and Patterns Report of Unplanned Reactor P602 L. Bell Trips at U.S. Light Water Reactors in 1985 08/86 Trends and Patterns Report of Engineered Safety P603 M. Harper Feature Actuations at Commercial U.S. Nuclear Power Plants 08/86 Trends and Patterns Report of the Operational P604 T. Wolf Experieri of Newly Licensed U.S. Nuclear Power Reactors TechnicalReviews 01/86 Pressure Sensitive Temperature Switch Results in T601 T. Cintula Spurious Actuation of Fire Suppression System 04/86 Emergency Diesel Generator Cooling Water System T602 E. Leeds Design Deficiencies at Maine Yankee and Haddam Neck 04/86 Inadvertent Pump Suction Transfer and Potential T603 R. Tripathi Auxiliary Feedwater Pump Cavitation at Davis-Besse 05/86 Events Resulting From Deficiencies in Labeling and T604 E. Trager Idemification Systems 06/86 Failure of Main Steam Safety Valves to Properly T605 R. Freeman Rescat 08/86 Inadvertent Recirculation Actuation Signals at T606 T. Cintula Combustion Engineering Plants D-19 Appendix D

1997 AEOD Annual Report Table D-11 Reports Issued in CY 1986 Date Title No. Author TechnicalReviews (Continued) 09/86 Occurrence of Events Involvis g Wrong Units / Wrong T607 E. Trager Train / Wrong Component-Update Through June 1986 11/86 Hydrogen Fire and Failure of Detection System T608 M. Chiramal 12/86 Foreign Material and Debris in Safety-Related T609 E. Leeds  :

Fluid Systems 12/86 ADS /RCIC System Interaction Events at River Bend T610 E. Leeds Unit i 12/86 Denied Access Due to Negative Room Pressure T611 T. Cintula 12/86 Degradation of Safety Systems Due to Component T612 R. Tripathi ;

Misalignment and/or Mispositioned Control /

Selector Switches Incident investigation Program Reports 01/86 Loss of Power and Water Hammer Event at NUREG-1190 San Onofre, Unit I on November 21,1986 02/86 Loss of Integrated Control System Power and NUREG-1195 Overcooling Transient at Rancho Seco on December 26,1985 08/86 Incident Investigation Manual

• 12/86 Incident Investigation Manual, Revision 1 *

• Superseded by NUREG-1303 (" Incident Investigation Manual"). published 2/88 (see Table D-9).

l l

l l

NUREG-1272, Vol. I 1, No. I D-20 l

l

Reactors Table D 12 Reports Issued in CY 1985 Date Title No. Author Case Studies 09/85 Licensee Event Report System, Evaluation of First Year NUREG-1022.

Results and Recommendations for Improvements Supplement 2 l

06/85 Safety Implications Associated With In-Plant C501 H. Ornstein Pressurized Gas Storage and Distribution Systems in Nuclear Power Plants 09/85 Overpressurization of Emergency Core Cooling In C502 P. Lam Boiling-Water Reactors j 12/85 Decay Heat Removal Problems at U.S. Pressurized C503 H. Ornstein Water Reactors 12/85 Loss of Safety System Function Events C504 E. Trager A nonreactor report (see AEOD Annual Report for 1985 [S601])

SpecialStudies 03/85 Review of Operational Experience From Non-Power S501 D. Zukor Reactors 04/85 AEOD Semiannual Report for July-December 1984 S502 J. Heltemes 09/85 Evaluation of Recent Valve Operator Motor Bumout Events S503 E. Brown Engineering Evaluations 01/85 Motor-Operated Valve Failures Due to E501 M. Chiramal Hammering Problem 01/85 Failure of Residual Heat Removal Suppression Pool E502 C. Hsti Cooling Valve to Operate 03/85 Partial Failures of Control Rod Systems to Scram E503 M. Chiramal 03/85 Loss or Actuation of Various Safety-Related Equipment E504 F. Ashe Due to Removal of Fuses or Opening of Circuit Breakers 03/85 Service Water System Air Release Valve Failures E505 S. Salah 05/85 Valve Stem Susceptibility to Intergranular Stress E506 C. Hsu Corrosion Cracking Due to Improper Heat Treatment 05/85 Electrical Interaction Between Units During Loss of E507 M. Chiramal Offsite Power Event of August 21,1984 at McGuire Units 1 and 2 D-21 Appendix D

1997 AEOD Annual Report Table D-12 Reports Issued in CY 1985 j l

Date Title No. Author Engineering Evaluations (Continued) 05/85 Nuclear Plant Operating Experience involving Safety E508 S. Rubin System Due to Bumped Electro 41echanical Components 07/85 Salem Umt 2 Depressurization Event E509 R. Freeman 1 07/85 Disabling of a Shared Diesel Generator Set Due E510 F. Ashe to Electrical Power Supply Arrangement for Support Auxiliaries 08/85 Closure of Emergency Core Cooling System Minimum E511 E. Leeds Flow Valves 09/85 Failure of Safety-Related Pumps Due to Debris E512 R. Freeman  !

09/85 High Pressure Core Spray System Relief Valve Failures E513 S. Salah 10/55 Core Damage Precursor Event at Trojan E514 D. Zukor 12/85 Inadvertent Actuation of Safety System Due to Cross Talk E515 M. Chiramal i Program Support Reports 07/85 Feedwater Transient Incidents in Westinghouse PWRs P501 R. Dennig 1

06/85 Trends and Patterns Analysis of 1981 Through 1983 P502 B. Brady LER Data (NUREG/CR-4129) 08/85 Engineered Safety Feature Actuazions at Commercial P503 T. Wolf U.S. Nuclear Power Reactors-January 1 Through June 30,1984 08/85 Trends and Patterns Report of Unplanned Reactor Trips P5N L. Bell j at U.S. Light Water Reactors in 1984 ~

i TechnicalReviews 01/85 Failure of Automatic Protection for Boron Dilution Event T501 R. Freeman at Callaway Unit 1 03/85 Comparative Analysis of Recent Feedline Water T502 E. Leeds Hammer Events at Maine Yankee, Calvert Cliffs, Palisades, and Salem 05/85 Pressurizer Level Instrumentation of Combustion T503 M. Chiramal Engineering Reactor Units 05/85 Loss cfInstrument Air and Subsequent Pressure T5N R. Freeman Transient at Callaway Unit 1 J

NUREO.! 272, Vol. I1. No. I D-22

Reactors .

Table D-12 Reports Issued in CY 19tl5 Dite Title No. Author TechnicalReviews (Continued) 07/85 Beaver Valley Component Cooling Water Pump T505 C. Hsu Damage 07/85 Primary System Release Due to Pressurizer Degas T506 T. Cintula Relief Valve Lifting 08/85 Standby Liquid Control System Pressure Relief Valves T507 E. Brown Lift at a Pressure Lower Than Reactor Coolant Pressure 08/85 Browns Ferry Nuclear Plant High Pressure Coolant T508 E. Leeds Injection System Performance Assessment 08/85 Inadequate Surveillance Testing Procedures for T509 E Ashe Degraded Voltage and Undervoltage Relays Associated With 4160-Volt Emergency Buses 09/85 Xenon Induced Power Oscillations at Catawba T510 R. Freeman 09/85 Technicians Perform Work on Wrong Control Rod T511 E. Trager Drive Mechanism 10/85 incorrect Plugging of Steam Generator Tubes T512 R. Freeman 11/85 Flooding of Safety-Related Valves in Pits T513 D. Zukor 11/85 Potential Loss of Component Cooling Water Due to T514 D. Zukor j Maladjustment of Relief Valves  !

12/85 Residual Heat Removal Service Water Booster Pump T515 S. Salah Air Binding at Brunswick Unit i 12/85 High Pressure Coolant Injection Overspeed Trip Loss T516 E. Trager Events and Subsequent Damage Due to Water Hammer incident investigation Prog.~am Reports 07/85 Loss of Main and Auxiliary Feedwater Event at the NUREG-1154 Davis-Besse Plant on June 9,1985 D-23 Appendix D

1997 AEOD Annual Report  ;

Table D-13 Reports issued in CY 1984 Date Title No. Author Case Studies  :

02/84 Licensee Event Repon System, Description of System NUREG-1022  ;

and Guidelines for Reponing Supplement 1 03/84 Low Temperature Overpressure Events at Turkey Point C401 W. Lanning -

Unit 4 06/84 Operating Experience Related to Moisture Intrusion in C402 M. El-Zeftawy Electrical Equipment at Commercial Power Reactors <

05/84 Hatch Unit 2 Plant Systems Interaction Event on C403 S. Rubin August 25,1982 07/84 Steam Binding of Auxiliary Feedwater Pumps C404 W. Lanning A nonreactor report (see AEOD Semiannual Repon, C405  !

September 1984)

SpecialStudies L

01/84 Human Error in Events Involving Wrong Unit or S401 E. Trager Wrong Train 07/84 Pressure Locking of Flexible Disk Wedge Type S402 S. Rubin Gate Valves 06/84 Annual Report of U.S. NRC Participation in the Nuclear S403 J. Crooks Energy Agency Incident Reponing System During 1983 06/84 Analysis of Foreign IRS Reports Submitted During S4M D. Zukor l CY 1984 1 09/84 Semiannual Report on AEOD Activities S405 J. Heltemes 10/84 Application of Risk Perspectives: A Procedures Guide S406 P. Lam Engineering Evaluations  ;

01/84 Temporary Loss of All AC Power Due to Relay Failure E401 M. Chiramal 3 in Diesel Generator Load Shedding Circuitry at Fon St Vrain 01/84 Water Hammer in Boiling Water Reactor High-Pressure E402 S. Rubin q Coolant Injection Systems 01/84 Deficiency in Automatic Switch Company (ASCO) Spare E403 F.Ashe Parts Kits for Scram Pilot Solenoid Valves 02/84 Failures in the Upper Head Injection System E404 D. Zukor NUREG-1272, Vol. I 1, No.1 D-24

Reactors Table D 13 Reports Issued in CY 1984 Date Title No. Author Engineering Evaluations (Continued) 03/84 Common Mode Failure of HPCI Steam Flow Isolation E405 M. El-Zeftawy Capability at Browns Ferry 03/84 Mechanical Snubber Failure E406 C. Hsu 03/84 Initiation and Indication Circuitry for High Pressure E407 F. Ashe Coolant Injection Systems 04/84 Reversed Differential Pressure Instrument Sensing Lines E408 S. Rubin 05/84 Operating Experience Involving Air in Instmment E409 S. Salah Sensing Lines 05/84 Operational Experiences Involving Standby Gas E410 F. Ashe Treatment Systems That Illustrate Potential Common-Cause Failure or Degradation Mechanisms >

05/84 Failure of Anti-Cavitation Device in Residual Heat E411 C. Hm Removal Service Water Heat Exchanger Outlet Valve 05/84 Adverse System Interaction With Domestic Water Systems E412 T. Cintula 05/84 Natural Circulation in Pressurized Water Reactors E413 W. Lanning 05/84 Stuck Open Isolation Check Valve on the Residual Heat E414 P. Lam Removal system at Hatch Unit 2 06/84 Overcooling Transient E415 E. Imbro 06/84 Erosion in Nuclear Power Plants E416 E. Brown  !

07/84 Loosening of Flange Bolts on Residual Heat Removal E417 C. Hsu Heat Exchanger Leading to Primary to Secondary Side Leakage 07/84 Feedwater Transients During Startup at E418 D. Zukor Westinghouse Plants 07/84 Failures of Fischer-Porter Transmitters Used in Safety E419 M. Chiramal Related Systems i

03/84 Operational Experiences Involving Shorted Lamp E420 M. Chiramal l Sockets ofIndication Lights

. 08/84 Loss of Pressurizer Heaters During Precore Hot E421 T. Cintula Functional Testing 03/84 High Pressure Coolant Injection System Performance at E422 T. Wolf Hatch Units I and 2 09/84 Failure of Large Hydraulic Snubbers to Lock Up E423 E. Brown D-25 Appendix D

1997 AEOD Annual Report Table D-13 Reports Issued in CY 1984 Date Title No. Author Engineering Evaluations (Continued) 10/84 Failure of Anchor Bolt on Diesel Generator Day Tank at E424 C. Hsu l Davis-Besse Unit 10/84 High Pressure Coolant Injection System Lockout at E425 M. Chiramal Vermont Yankee 10/84 Single Failure Vulnerability of Power-Operated Relief E426 E. Imbro l Valve Actuation Circuitry for Low Temperature i Overpressure Protection 11/84 Licensee Event Reports That Address Situations That E427 E Ashe Potentially Could Result in Overloading Electrical Equipment in the Emergency Power System or Prevent Operation of the Onsite Power System Sequencer Program Support Reports 02/84 Operating History Overview for Diesel Generators in P401 R. Dennig Nuclear Service M. Chiramal 03/84 AEOD Trends and Patterns Program Plan P402 R. Dennig 05/84 AEOD Trends and Patterns Evaluation Report, P403 E Hebdon Preliminary Assessment of LER Reporting Under 10 CFR 50.73 5

03/84 LER Data on Personnel Errors P404 E Hebdon 11/84 Draft Trends and Patterns Analysis of Feedwater Transients at Westinghouse PWRs P405 M. Harper 11/84 Trends and Patterns Analysis of Reactor Scrams P4% L. Bell (Pilot Study)

TechnicalReviews 03/84 Failures of Containment Air Monitors at Farley Units I and 2 T401 D. Zukor 03/84 Chemical Contamination of Primary and Secondary T402 M. El-Zeftawy Systems in Lights Water Reactors ,

03/84 Setpoint Drift of Barton Model 288 Switches T403 M. Chiramal i

04/84 Cable Fire and Loss of Control Power to Engineered T404 M. Chiramal Safeguards Valves 04/84 Cold Weather Events 1983 - 1984 T405 T. C ntula 04/84 Improper Spare Parts Procurement Event at T406 T. Wolf Grand Gulf Unit NUREG-1272, Vol. I 1. No. I D-26

Reactors l

1 l

l Table D 13 Reports Issued in CY 1984 D:te Title No. Author TechnicalReviews (Continued) 04/84 Failure of a 4 kV Circuit Breaker to Trip T407 M. Chiramal l

05/84 Diesel Generator Inoperability Due to Overheating of T408 M. Chiramal Ventilation Cowling 05/84 Multiple Failure of Bell and Howell Dua: Potentiometer T409 F. Ashe Modules That Occurred at the Fort Calhoun

Nuclear Station 05/84 Failure of Injection Valve for the High Pressure Coolant T410 E. Brown

! Injection System to Open During a Surveillance Test 06/84 Contamination of the Nitrogen System at Sacramento T411 M. El-Zeftawy l Municipal Utility District 06/84 Failure of an Access Door Between the Drywell and T412 T. Wolf the Wetwell 06/84 Failure of Fire Damper in Safeguards Ventilation System T413 W. Lanning 07/84 Station Operating Restrictions for Loss or Out-Of-Service T414 F. Ashe Power Transformers Through Which Electrical Power is Supplied to the Emergency Buses 07/84 Destruction of Charging Pump T415 W. Lanning 08/84 Loss of Engineered Safety Feature Auxiliary Feedwater T416 D. Zukor Pump Capability at Trojan on January 22,1983 08/84 Excessive Cooldown Rate Event at LaSalle Unit 1 T417 S. Salah 08/84 Events Involving Fires or Other Related Abnormalities in T418 F. Ashe Motor Control Centers with Aluminum Bus Bars 08/84 Contamination of Snubber Bleed Screw and Lockup T419 C. Hsu Poppet Valve j 08/84 Failure of an Isolation Valve of the Reactor Core T420 P. Lam Isolation Cooling System to Open Against Operating Reactor Pressure 08/84 Design Deficiency in Standby Gas Treatment System T421 M. Chiramal 08/84 Inoperability of Safety Injection Pump at Salem Unit 1 T422 'D. Zukor on October 17,983 10/84 Inoperability of Helium Circulator Overspeed Trip T423 E. Imbro Channels Due to Impedance Variations in Speed Sensing Cables Exposed to Steam Lee.k Fire Water Main Leakage into 4 kV Switchgear Room at T424 T. Cintula l j 11/84 i San Onofre Unit 1 i

D-27 Appendix D

1997 AEOD Annual Report Table D-14 Reports Issued in CY 1983 Date Title No. Author Case Studies 09/83 Licensee Event Report System, Description of System NUREG-1022 and Guidelines for Reporting 09/83 Potentially Damaging Failure Modes ofliigh and NUREG/CR-3122 M.Chiramal Medium Voltage Electrical Equipment 04/83 Failures of Class 1 E Safety-Related Switchgear Circuit C301 M. Chiramal Breakers to Close on Demand Engineering Evaluations 01/83 Fuel Degradation at Westinghouse Plants E301 D. Zukor 04/83 Update to AEOD/E301 (Fuel Degradation at E301 D. Zukor Westinghouse Plants) Revision 1 01/83 Potential Loss of Serv;ce Water Flow Resulting From E302 E. Imbro a Loss of Instrument Air 02/83 Valve Flooding Event at Surry E303 D. Zukor 03/83 Investigation of Backflow Protection in Common E304 T. Cintula Equipment and Floor Drain systems to Prevent Flooding of Vital Equipment in Safety-Related Compartments 04/83 Inoperable Motor-Operated Valve Assemblies Due to E305 E. Brown Premature Degradation of Motors and/or Improper Limit F. Ashe Switch / Torque Switch Adjustment 04/83 Cooldown During Loss of Control Room Test at E306 D. Zukor McGuire Unit 1 l

i 04/83 Degradation of Safety-Related Batteries Due to E307 E Ashe Cracking of Battery Cell Cases and/or Other Possible Aging-Related Mechanisms 04/83 Cracks and Leaks in Small-Diameter Piping E308 E. Brown 04/83 The Potential for Water Hammer During the Restart of E309 S. Rubin Residual Heat Removal Pumps at BWR Nuclear '

Power Plants 04/83 Loss of Shutdown Cooling and Subsequent Boron E310 T Cintula Dilution at San Onofre Unit 2 04/83 Loss of Salt Water Flow to the Service Water Heat E318 T. Cintula Exchangers for 23 Minutes at Calvert Cliffs Unit 2 05/83 Operability of Target Rock Safety Relief Valves in the E312 1. Pellet Safety Mode with Pilot Vulve Leakage NUREG-1272, Vol. I 1, No. I D-28

Reactors Table D 14 Reports Issued in CY 1983 Date Title No. Author Engineering Evaluations (Continued) 06/83 Potential Contamination of the Sp:at Fuel Pool and E313 E. Brown Primary Reactor System 06/83 Loss of All Three Charging Pumps Due to Empty E314 T. Cintula j

~

Common Reference Leg in the Liquid LevelTransducers for the Volume Control Tank at St. Lucie 1 07/83 Misuse of Valve Resulting in Vibration and Damage E315 E. Brown  !

to the Valve Assembly and Pipe Supports 07/83 Frozen Ice Condenser Intennediate Deck Doors E316 D. Zukor 08/83 Loss of High Pressure Injection System E317 N. Trehan 08/83 Biofouling at Salem Units 1 and 2 E318 E. Imbro 09/83 Loss of Drywell Torus Pressure Differential During Residual Heat Removal Pump Flow Testing at Cooper E319 S. Rubin 09/83 Power-Operated Relief Valve Actuation Resulting in E320 E. Imbro Safety injection Actuation at Calvert Cliffs 09/83 Three Similar Events of a Loss of Shutdown Cooling E321 T. Cintula )

Flow at Combustion Engineering Plants 09/83 Damage to Vacuum Breaker Valves as a Result of E322 C. Hsu Relief Valve Lifting at Peach Bottom Unit 2 09/83 Load Reduction Transient at Salem Unit 2 E323 N. Trehan I on January 14,1982 09/83 Review of Events involving Failures of Power Supply in E324 M. Chiramal Instrumentation and Control Systems 11/83 Vapor Binding of Auxiliary Feedwater Pumps at E325 W. Lanning Robinson Unit 2 11/83 Steam Voiding in Oconee Unit 3 on June 13,1975: E326 IL Ornstein A Precursor Event to the TMI-2 Accident l1/83 Gaseous Releases From Waste Gas Disposal System E327 N. Trehan i1/93 Human Factors involvement in Events at Oconee NT304 K. Black Units 1,2, and 3 08/83 Human Factors Contributions to Accident Sequence N305 E. Trager Precursor Events D-29 Appendix D

1997 AEOD Annual Report .

1 I

Table D-14 Reports Issued in CY 1983 Date Title No, Author Program Support Reports I

07/83 Report on the Implications of the Anticipated Transient P301 J. Crooks Without Scram Events at the Salem Nuclear Power Plant on the NRC Program for Collection and Analysis of Operational Experience i

TechnicalReviews t 01/83 Diesel Generators Load Sequencer Design T301 M. Chiramal Deficiency-LER 82-025/OIT 02/83 Postulated Loss of Auxiliary Feedwater System T302 E. Imbro Resulting From a Turbine Driven Auxiliary Feedwater Pump Steam Supply Line Rupture 03/83 Seat Degradation in lienry Pratt Butterfly Valves T303 E. Brown 03/83 Cause of Containment Isolation Valve F042A to Close T304 S. Salah 03/83 Flow Blockage in Essential Raw Cooling Water System Due to Asiatic Clam Intrusion at Sequoyah Unit i T305 E. Imbro t

04/83 Scram Discharge Volume Level Switch Failure at Hatch T306 J. Pellet Unit 2 l

04/83 Condensate Demineralizer Resin Migration Through the E307 J. Pellet Plant Vent and the Standby Gas Treatment System at Pilgrim Unit 1 04/83 Undetectable Failure in Westinghouse solid State T308 M. Chiramal Protection System 04/83 Air in Reactor Water Cleanup System Instrument T309 S. Salah Sensing Lines at Brunswick Unit 2 04/83 Blocking of Automatic Safety Injection Signals T310 M. Chiramal 05/83 Rod Control Urgent Failure on June 25,1982, T311 N. Trehan at Surry Unit 2 05/83 Failure of 5 kV Cable Terminations T312 M. Chiramal i 05/83 i j Capped Containment Pressure Sensing Lines T313 S. Rubin '

05/83 Improper Size of Inlet Piping to Primary Safety Valves T314 E. Imbro 05/83 Events Involvmg Losses of or Perturbations in a Single T315 E Ashe 120 Volt AC Vi;al Power Supply Inverter and Attendant Distribution 20s Which Resulted in inadvertent Actuation 5 of Safety Systems NUREG-1272,. Vol. I 1, No. I D-30

I Reactors Tabk D 14 Reports Issued in CY 1983 Date Title No. Author TechnicalReviews (Continued) 05/83 Thermal Non-Repeatability Problem With Barton Models T316 M. Chiramal 763 and 764 Electronic Transrnitters 06/83 Problems With Diesel Driven Containment spray Pump T317 L).Zukor at Zion Unit 2 on December 16,1982 06/83 Failure of Recirculation spray Service Water Motor- T318 D. Zukor Operated Valves 06/03 Design Deficiency in Control Circuits of Feedwater T319 M. Chiramal l Isolation Valves and Boron Injection Tank J Recirculation Valves l 1

06/8i Inadvertent Safety injections Attribute < to Personnel T320 F. Ashe Error at Summer )

06/83 Check Valve Installed Backwa.-ds in '4nstrument Air Line T321 D. Zukor to the Power-Operated Relief Valve at Surry Unit 2 )

06/83 Gouges in Main Coolant System Piping at Diablo T322 D. Zukor  ;

Canyon on April 19,1983 06/83 Turbine Trip Bypass Delay at Grand Gulf Unit 1 T323 S. Salah 07/o3 Events Involving Two or More Simultaneously Dropped T324 F.Ashe Rod Control Cluster Assemblies )

08/83 Leakage in Static-O-Ring Pressure Switches T325 M. Chiramal 08/83 Fafety Relief Valve Corrosion at a Foreign Reactor T326 E. Brown 08/83 Auxiliary Feedwater Header Problems at Babcock T327 H. Ornstein

& Wilcox Plar's 08/83 Two of Three Emergency Core Cooling System T328 D. Zukor Accumulators Inoperable at Surry Unit 1 08/33 Leak in Reactor Water Cleanup System "B" T329 C. Hsu Regenerative Heat Exchanger Relief Line 08/83 Steam Generator Tube Rupture at Oconee Unit 2 T330 M. El-Zeftawy 08/03 Review of Events at Operating Nuclear Plants Involving T331 M. Chiramal Plant Computers 10/83 Reactor Vessel Drainage T332 S. Salah 10/83 Degradation of Saltwater Cooling System at San Onofre T333 H. Ornstein Unit 1 Due to a Loss ofInstrument Air 11/83 ReactorVessel Drainage at Grand Gulf Unit i T334 S. Salah D-31 Appendix D

1997 AEOD Annual Report Table D 14 Reports Issued in CY 1983 Date Title No. Author TechnicalReviews (Continued) ,

i 11/83 Simultaneous Safety Injection Actuation Signal and T335 T. Cintula Recirculation Actuation Signalat San Onofre Unit 3 11/83 Design Deficiency Resulting in Isolatiori of Both Loops of T336 M. Chiramal the Emergency Condenser System at Nine Mile Point Unit I  ;

11/83 Water Hammer in the Main Feedwater System Resulting T337 E. Imbro in a Feedwater Line Crack at Maine Yankee 11/83 Water Leak Through Containment spray Block Valves at T338 D. Zukor San Onofre 1 11/83 Redundant Emergency Core Cooling System Pump T339 T. Cintula  :

Room Air Coolers Out of Service for 22 Hours at Calvert Cliffs Unit 1  :

12/83 Evaluation of Control Rod Mismanipulation Event at T340 T. Wolf Hatch Unit 2 12/83 Corrosion of Carbon Steel Pipe in Service Water T341 E. Brown Headers i

I i

NUREG-1272, Vol. I1. No. I D-32

Reactors Table D 15 Reports Issued in CY 1982 Date Title No. Author Case Studies 01/82 Safety Concern Ansociated With Reactor Vessel Level C201 M. Chiram.d Instrumentation in Boiling Water Reactors 02/82 Report on Service Water System Flow Blockages C202 E. Imbro by Bivalve Mollusks at Arkansa Nuclear One and Brunswick 05/82 Survey of Valve Operator Related Evene. Occurring C203 E. Brown During 1978,1979 and 1980 07/82 San Onofre Unit I Loss of Salt Water Cooling Event on C204 11. Ornstein March 10.1980 08/82 Abnonnal Transient Operating Guidelines as Applied to C205 J. Pelle' I the April 1981 Overfill Event at Arkansas Nuclear One, I Unit 1 l 10/82 Inadvertent Loss of Reactor Coolant Events at the C206 W. Lanning 1

Sequoyah Nuclear Plan, Units 1 and 2 Engineering Eraluations

. 01/82 Methodology for Vi:al area Determination E201 W. Lanning 01/82 Loss e;f High Pressure Injection Lube Oil Cooling at E202 J. Pellet Rancho Seco 01/82 Inadvertent isolation of Contrinment Fan Units at Salem E203 W. Lanning Unit i 1 01/82 Effects of Fire Protection System Actuation on Safety E204 M. Chiramal l' Related Equipment l

02/82 Potential Consequences of Heavy Load Drop Accidents E205 M. El-Zeftawy l in LWRs l l 02/82 Load Reduction Transient on January 14,1982, at E206 N. Trehan Salem Unit 2 02/82 LER 50-336/81-26: Investigation of the Relative E207 E. Imbro Frequency of Valve Overtravel Anomalies that could Result in a Potential Centrifugal Pump Runout Exceeding Net Positive Suction Head 02/82 An Observed Difference in Lift Setpoint for Steam E208 T. Cintula Generator and Pressurizer Safety Valves 02/82 Generator Rotor Retaining Ring as a Potential Missile E209 M. Chiramal (Incident at Barseback Unit 1 on April 13,1979) 02/82 Inadequate Switchgear Cooling at Beaver Valley Unit 1 E210 W. Lanning D-33 Appendix D

1997 AEOD Annual Report l

Table D 15 Reports Issued in CY 1982 Date Title No. Author Engineering Evaluations (Continued) 02/82 Repetitive Failures of Emergency Feedwater Flow E211 T. Cintula Vahes at Arkansas Unit 2 Because of Valve Operator Hydraulic Problems i 02/82 Spurious Trip of th: Generator Lockout Relay E212 F.Ashe Associated With a Jiesel Generator Unit 02/82 ':. rip of Two Inservice Auxiliary Feedwater Pumps From E213 D. Zukor Low Suction at Zion Unit 2 on December 11.1981 03/82 Duane Arnold Loss of River Water System Loop E214 T. Wolf 03/82 Engineering Evaluation of the Salt Service Water E215 E. Imbro System Flow Blockage at the Pilgrim Nuclear Power Station by Blue Mussels 03/82 A Recently Evaluated Preoperational Test Precursor of E216 H. Omstein the TMI-2 Accident 03/82 Scram Pilot Solenoid Valve Failures Due to Low E217 M. Chiramal Voltage-Grand Gulf Unit 1 03/82 Potential for Air Binding or Degraded Perfonnance of E218 S. Rubin i BWR Residual Heat Removal System Pumps During the Recirculation Phase of a Loss-of-Coolant Accident 04/82 Proposed Circular: Contamination of Air Serving Safety E219 H. Ornstein Related Equipment 04/82 Water in the Fuel Oil Tank at Surry Power Station Unit 2 E220 N. Trehan 04/82 Indian Point Unit 2 Flooding Event E221 W. Lanning 05/82 Loss of Reserve Station Service Transformer "B" E222 N. Trehan on January 18,1981, at Surry Unit 2 1

05/82 Ir. advertent Loss-Of Coolant Events at Sequoyah E223 W. Lanning Units 1 and 2 05/82 Generic Concerns Associated With the Ginna Steam E224 W. Lanning {

Generator Tube Rupture Event 06/82 Degradation of BWR Scram Pilot Soler6d Valves Due E225 M. Chiramal to Abnormal Power Supply Voltage 1

06/82 Inoperability of Instrumentation Due to Extreme E226 M. Chiramal Cold Weather 06/82 Failure of Engineered Safety Features ManualInitiation E227 F. Ashe Pushbutton Switches NUREG-1272, Vol. I1, No. I D-34

Reactors Table D-15 Reports Issued in CY 1982 Date Title No. Author Engineering Evaluations (Continued) 06/82 Repetitive Overspeed Trips of the Steam Driven E228 E. Imbro Emergency Feedwater Pump on Initial Start at Arkansas Nuclear One, Unit 2 06/82 Potential for Flooding in Control Room at San Onofre E229 T. Cintula Units 2 and 3 07/82 Water in the Fuel Oil Tank at Surry Power Sta' ion, E230 N. Trehan Unit 2-Additional Information 07/82 Millstone Unit 2 Loss of Shutdown Cooling Due to Trip of E231 M. Chiramal Low Pressure Safety Injection Pump 07/82 Potential Deficiency in the Sigma Lumigraph Indicators E232 F.Ashe Model Number 9270 07/82 Carbon Dioxide Systems Used for Fire Protection in or E233 M. Chiramal Adjacent to Critical Areas 08/82 Failure in a Section of 4 kV Bus Cable Manufactured E234 F. Ashe by Okonite 08/82 Wiring Error in Handswitch for Solenoid Control Valves E235 S. Rubin Associated With High-Pressure Coolant Injection System Steam Condensing Mode Pressure Control Valve at Duane Arnold 08/82 Brunswick Steam Electric Plant Unit 2 Loss of Residual- E236 T. Wolf Heat Removal Service Water on January 16,1982 08/82 Power-Operated Relief Valve Failure at Robinson E237 E. Brown 08/82 Water in the Lube Oil in Safety Injection Pump IA-A at E238 N. TreSan Sequoyah-LER 81-076 09/82 Main Steam Isolation Valve Closures and Pressurizer E239 T. Cintula Safety Valve Actuations at St. Lucie Unit 1 on December 19,1981 09/82 Preliminary Account of Events Associated With a E240 S. Rubin ReactorTrip at Hatch Unit 2 on August 25,1982 10/82 Emergency Diesel Generator System Problems E241 M. Chiramal at FitzPatrick 10/82 Fuel Assembly Degradation While in the Spent Fuel E242 E. Brown Storage Pool 10/82 Plant Trip Followed by a Safety injection Due to Loss E243 T. Cintula of"A" Cooling Tower Pump at Palisades on February 4,1982 D-35 Appendix D l

L ..

1997 AEOD Annual Report Table D 15 Reports Issued in CY 1982 Date Title No. Author Engineering Evaluations (Continued) 10/82 Loss of Residual Heat Removal System Event at Pilgrim E244 T. Wolf Nuclear Power Station on December 21,1981 10/82 Failure of Westinghouse type SC-1 No.1876-072 Relays E245 F.Ashe 10/82 Events Involving Loss of ElectricalInverters Including E246 F.Ashe Attendant Inverters to Vital Instrument Buses 10/82 Engineering Evaluation of Turbine / Reactor Trip at E247 J. Pellet Rancho Seco on August 7,1981 11/82 Engineering Evaluation Report on McGuire E248 D. Zukor Overpressurization Event of August 27,1981 11/82 Engineering Evaluation Memorandum-Licensee E249 H. Ornstein Reporting of the Turbine / Reactor Trip at Rancho Seco on August 7,1981 11/82 Quad Cities Unit 2 Loss of Auxiliary Electrical Power E250 M. Chiramal Event on June 22,1982 11/82 Salem Unit 2 Loss of Vital Bus No. 2A E251 M. Chiramal 11/82 Potential Control Logic Problem Resulting in Inoperable E253 F. Ashe Auto-Start of Diesel Generator Units Under the Conditions of Loss-of-Coolant Accident and Loss of Station Power (LOSP) 11/82 Review of Prairie Island Unit i LER 82-015-OlT on E254 M. Chiramal Diesel Generator Operability 11/82 Failure of the Vent Line on the Common Discharge of E255 T. Cintula the Two Motor-Driven Auxiliary Feedwater Pumps at San Onofre Unit 2 From an Improper Valve Lineup i1/82 Loss of Shutdown Cooling and Subsequent Boron E256 T. Cintula Dilution at San Onofre Unit 2 12/82 Insufficient Net Positive Suction Head for Charging E257 D. Zukor Pump Service Water Pumps at Surry Nuclear Power Station I

i NUREG-1272. Vol. I 1. No. I D-36

Reactors Table D 16 Reports Issued in CY 1981 Date Tit! No. Author Case Studies 03/81 Report on the St. Lucie Unit 1 Natural Circulation C101 E. Imbro Cooldown on June 11.1980 03/81 Robinson Reactor Coolant System Leak on C102 W. Lanning January 29,1981 03/81 AEOD Safety Concems Associated With Pipe Breaks NUREG-0785 in the BWR Scram System (C103) S. Rubin 04/81 Millstone Unit 2 Loss of 125 V DC Bus Event on C104 M. Chiramal January 2,1981 12/81 Report on the Calvert Cliffs Unit i Loss of Service C105 E. Imbro Water on May 20,1980 Engineering Evaluations 01/81 Degradation ofInternal Appurtenances in LWR Piping E101 E. Brown 01/81 Sequoyah Unit 1 Loss of Annunciation E102 M. Chiramal 02/81 Davis-Besse Nuclear Power Station, Unit I- E103 M. Chiramal Engineered Safety Features Actuation System (ESFAS) 03/81 Engineering Evaluation of Feedwater Transient and E104 S. Sands System Pipe Break at Turkey Point 3 03/81 Water Hammer During Restart of Residual Heat E105 J. Huang Removal Pumps 03/81 Water Hammer in the Steam Condensing Mode of the E106 J. Huang Residual Heat Removal System Operation 04/81 Peach Bottom Unit 3 Occurrence on Febmary 25,1981 E107 E Ashe 04/81 Hatch Units I and 2-Alternate Offsite Source Interlock E108 M. Chiramal With Emergency Diesel Generators 04/81 Potential Common-Mode Failure of Diesel Generators E109 M. Chiramal 04/81 Requirements of the Preferred or Offsite Power System El10 E Ashe

05/81 Evaluation of High Pressure Safety injection Pump El11 E. Imbro Operability Without Service Water 06/81 Inoperability of Instrumentation Due to Extreme El12 M. Chiramal Cold Weather 06/81 Deliberate Pump Trip at Browns Ferry Unit 2 on El13 W. Lanning April 6,1981 D-37 Appendix D

1997 AEOD Annual Report Table D-16 Reports Issued in CY 1981 Date Title No. Author Engineering Evaluations (Continued) 06/81 Control System Failures that could Cause or Exacerbate El14 E Ashe Nuclear Power Plant Accidents 4 07/81 Additional Information on Events at TMI-2 During El15 H. Ornstein Preoperational Testing (September 5-12,1977) 07/81 Failure of B Phase Main Transformer and Subsequent El16 M. Chiramal Fire in the Transformer Area-North Anna Unit 2 07/81 Events at TMI-2 During Preoperation Testing El17 H. Omstein 07/81 Setpoint Drift Occurrences for the Barton El18 E Ashe Model 288 Instrument 07/81 Loss of Residual Heat Removal Capability at Brunswick El19 E. Imbro Units 1 and 2 08/81 Ignition of Gaseous Waste Decay Tank at San Onofre E120 E. Imbro Unit 1-July 17,1981 08/81 Crystal River 3 Engineered Safeguards Relay Failures E121 M. Chiramal 09/81 AEOD Ccacern Regarding Inadvertent Opening of E122 H. Ornstein Atmospheric Dump Valves on B&W Plants During Loss of Integrated Control System Nonnuclear/ Instrumentation 09/81 Immediate Action Memo: Common Cause Failure E123 H. Ornstein Potential at Rancho Seco-Desiccant Contamination of Air Lines 09/81 Review ofInformation on Purge Valves E124 E. Brown 10/81 Engineering Evaluation Report on Shutdown Cooling E125 G. Lanik System Heat Exchanger Failures at Oyster Creek.

August 1981 10/81 Event Sequences Not Considered in the Design of E126 E Ashe Emergency Bus Control Logic 10/81 Pressure Boundary Degradation Due To Pump Seal E127 W. Lanning Failure at Arkansas Nuclear One 11/81 Inoperable Teledyne Solenoid Valves E128 E Ashe 12/81 Brunswick Unit 2 Diesel Generator Jacket Water E129 M. Chiramal Tempemture Control Valve and Manual Bypass Valve 12/81 Davis Besse LER 79-062 on Auxiliary Feedwater System E130 M. Chiramal Pressure Switches 12/81 High Circulating Current Associated With Inverter Output E131 E Ashe Due to Lack of Circuit Tuning NUREO I272,Vol. I1 No. I D-38

Reactors n

Table D 16 Reports Issued in CY 1981 ,

Date Title No. Author Engineering Evaluations (Continued) 12/81 Abnormal Wear Encountered on Aloyco Swing Check E132 T. Cintula Valves Installed in the Low Pressure Safety injection System at Palisades 04/81 Inadequacies in Periodic Testing of Combustion E133 M. Chiramal Engineering PWR Reactor Protection System L

t D-39 Appendix D

1997 AEOD Annual Report Table D-17 Reports Issued in CY 1980 Date Title No. Author Case Studies 07/80 Report on the Browns Ferry Unit 3 Partial Failure to C001 S. Rubin Scram Event on June 28,1980 09/80 Report on Interim Equipment and Procedures at C002 O. Lanik Browns Ferry Unit 3 to Detect Water in the Scram Discharge Volume 10/80 Report on Loss-of-Offsite Power Event at Arkansas C003 W. Lanning Nuclear One, Units 1 and 2 11/80 AEOD Actions Concerning the Crystal River Unit 3 Loss C004 H. Ornstein of Nonnuclear Instrumentation and Integrated Control System Power on February 26,1980 12/80 AEOD Observations and Recommendations Concerning C005 E. Imbro the Problem of Steam Generator Overfill and Combined Primary and Secondary Side Blowdown Engineering Evaluations 03/80 Crystal River Nuclear Power Plant Decay Heat Closed E001 H. Ornstein Cycle Ceoling Water Pumps /DCP-I A and DCP-1 B 05/80 BWR Jet Pump Integrity E002 S. Rubin 06/80 Comparison of Reactor Coolant Pump Events Contained E003 E. Brown i in LERs, NPRDS, RECON, and Plant Records 07/80 Data Summaries of Licensee Event Reports of Pumps at E004 H. Ornstein U.S. Commercial Nuclear Power Plants, January 1,1971 to April 30,1978 1 07/80 Operational Restrictions for Class IE 120V AC Vital E005 M. Chiramal Instmment Buses 08/80 Loss of Residual Heat Removal at Beaver Valley, E006 W. Lanning LER 80-031 08/80 Potential for Unacceptable Interaction Between the E007 S. Rubin Control Rod Drive System and Nonessential Control Air Systems at Browns Ferry 08/80 Operational Restrictions During Surveillance Testing of E008 M. Chiramal Emergency Diesel Generators 08/80 Failures of Containment Isolation Valves at Zion E009 W. Lanning 08/80 Tie Breaker Between Redundant Class IE E010 M. Chiramal Buses-Point Beach Units 1 and 2 08/80 Concerns Relating to the Integrity of a Polymer Coating E011 E. Imbro for Surfaces Inside Containment NUREG-1272, Vol. I 1, No. I D-40

Reactors Table D 17 Reports Issued in CY 1980 Date Title No. Author Engineering Evaluations (Continued)

/

09/80 Salem Unit 1-Solenoid Valve of Containment Fan Coil E012 M. Chiramal Unit Service Water flow Control Valve ,

09/80 Excessive Main Feedwater Transient E013 J. Creswell ,

10/80 Transient at Crystal River Unit 3-September 30,1980 E014 H. Ornstein 10/80 January 3,1977, Quad Cities Unit i Loss-of-Air Event E015 G. Lanik and its Effects on Scram Capability 10/80 Flow Blockage in Essential Equipment at ANO Caused E016 E. Imbro by Corbicula sp. (Asiatic Clams) 10/80 Engineering Evaluation of Steam Generator Overfill E017 W. Lanning 12/80 Potential Failure of BWR Backup Scram (Mode Switch E018 M. Chiramal in Shutdown) Capability 12/80 Davis Besse Unit 1-Emergency Core Cooling System E019 M. Chiramal Actuation During Hot Shutdown on December 5,1980 5 o

12/80 Intemal Appurtenances in LWRs E020 E. Brown 1

D-41 Appendix D

APPENDIX E AEOD Technical Reports by Category E-1 PWR Plant Systems E-2 BWR Plant Systems E-3 Activity / Human Factor Deficiency E-4 Topics l

l i

l

Reactors j i

l l

INTRODUCTION Brckground the review of operating experience. Case studies document the bases for AEOD recommendations Appendices C and D list approximately 500 operat- for regulatory or industry actions. Before being ing experience mports published by AEOD since published each se study report goes through a 1980. These reports cover a broad spectrum of rigorous peer review process to ensure technical operating experience data. Some of them have also adequacy.

been published as NUREGs, including the SPecial Studies. Special studies are designated by NUREG-1275 series of Operating Experience Feedback Reports. AEOD reports have been an S prefix and document accelerated im;stigations and suggest or recommend regulatory actions that broadly disseminated throughout the nuclear are to be completed expeditiously.

community and to the public. Most reports can be found in the NRC's Public Document Room. the Engineering Evaluations. Engineering evaluations local public document rooms, and the Nuclear are designated by an E prefix. They document Documents (NUDOCS) database under the Task assesrments of significant operating events and Identifier AE, followed by the report number. suggest remedial actions, if appropriate.

This appendix has been prepared as an aid to more Technical Reviews. AEOD technical reviews are effectively communicate the lessons of operating designated by a T prefix and document studies in experience, and to help ensure that those lessons are which the statT concludes there is little safety not forgotten. It contains tables of AEOD report significance. These studies typically conclude that numbers sorted by topic. A report may be listed in the licensees' or industry's planned or scheduled more than one topical area, depending upon its correctise actions are adequate.

scope. To find the title for any report, refer to Appendix C or D, as appropriate. Prop am Support Reports. Program support reports are designated by a P prefix and document Tables E-1 and E-2 use system descriptions which studies of trends and patterns in a variety of sys-contain material copyrighted by Nuclear Power tems, components, events, and programs.

Experience. This material is reproduced by permis-sion of Hagler Bailly Consulting, Inc. Report Number. For reports issued in the 1980s, the report number is a three digit number. The first Definitions digit is the last digit of the year of publication (i.e..

a "0" indicates 1980 and a "1" indicates 1981). The AEOD reports are designated by an alphanumen. c

, remaining two digits are the sequence number, .

sequence. The first character is a letter pirefix which representing the sequential order of publication m denotes the type of report. The remaining characters that year. For reports issued .in the 1990s, the report comprise the report number, which m. dicate the year .. .

number consists of four digits with a hyphen in the of publication and the sequence number of the middle. The first two digits (before the hyphen) are l report. These designators are described below. the last two d. .igits of the ye:r of publication (.i .e .,

Ct.se Studies. Case studies are designated by a C "90" indicates 1990 and "95" indicates 1995). The i prefix and involve substanove, in-depth analyses of last two digits are the sequential number of publica-significant safety issues that are identified through tion in that year.  !

l E-iii Appendix E

APPENDIX E-1 PWR Plant Systems s

I

, i r --

Reactors CONTENTS

1. Fuel . .. . . . .. . .. . .. E-1-1 II. Reactor Intemals. . . .. . E-i-1 III. Reactor Vessel . . . . .. . . . . . E-1-1 IV. Control Rods and Drives . . . . . E-1-1 V. Reactor Coolant System (RCS) . .. . . . .. E-1-1 VI. Turbine Cycle Systems . .. . . E-1-1 VII. Safety Systems. ... . . . .. . E-1-2 ,

VIII. Auxiliary Systems . . . .. .. E-1-2 IX. Instrumentation and Control (I&C) . .. .. E-1-2 X. Fuel Handling Facilities and Systems ... . . . . . . . . . . E-1-3 XI. Electrical Systems .. . . . . . . . E-1-3 #

XII. Liquid Radwaste System . .. . . . . . . . . . E-1-3 XIll. Gaseous Radwaste System . .. . . . . E-1-3 XIV. Buildings and Containment . . .. . .. .. . E-1-3 X V. Miscellaneous Systems .. . . .. . E-1-3 XVI. Operational Problems . .. E-1-4 i

E-1-iii Appendix E

Reactors l

I PWR PLANT SYSTEMS' l

I l

I. Fuel - E205, E242, E301, E313, E326 E. Pressuri:er - C102, E208, E237, E239, E320, I includes uranium fuel pellets and cladding, fuel E421,T506,E708,T704,E805,E909,S902,E90 I c.ssemblies, holdout springs, guide tubes 08, E90-09 Includes pressurizer shell, internal heaters, H. Reactor Internals T809 manway, nozzles, pressurizer relief tank, PORV Includes upper guide structure, thermal shield, core block valves, resistance temperature detectors barrel, supports for core instrumentation (RTDs), manifold valves III. Reactor Vessel- E114 F. Miscellaneous - None Includes additional RCS Includes reactor pressure vessel (RPV), nozzles, loop valves not associated with above categories head bolts, seals VI. Turbine Cycle Systems IV. Control Rods and Drives A. Turbine None includes main turbine,high A. Control Rods - T324, E613, T712, E90-04 pressure (HP) and low pressure (LP) cylinders, Includes absorber and poison rods, rod control including casings, rotors, shafts, blades, bearings, cluster assemblies, control element assemblies stop and control valves, drain and crossover lines,

'" "Y' #*

D. Drives - E206, T311, E323 Rev.1, S503, T511,

'IHO, T91-07 B. Generator - E209 hiain generator system Includes magnetic jack control rod drive mecha- includes rotor, stator, exciter, brushes, bearings, nisms, housings, drive shafts, motors, clutches, coils, voltage regulator, armature, commutator, litches, grippers windings, generator cooling, seal oil systems V. Reactor Coolant System (RCS) C. Condensers - None hiain condenser includes A. Pumps - E003, E127, E326, E415, T707, tubes, baffles, vacuum pump, air ejector, hotwell, NUREG-1275 Vol. 7 shells, water boxes Includes main reacter coolant pumps (RCPs),

c:: sings, flanges, shafts, bearings, seals, impellers, D. Steam - C005, E017, E122, E128, E208, E239, speed controls E415, E502, E514, T90-04, T90-08, S92-02 Includes turbine bypass and atmo- spheric steam D. Piping - E101, T322, T701 dump valves, SRV, main steam isolation valves Includes main coolant lines, welds, fittings (h1SIV), moisture separator re- heaters (htSRs),

main steam line (htSL) piping C. Relief and Safety Falves - T314, T321, C401, E426, T91 05, S92-02, T92-01, T93-01, T96-02 E. Condensate and Feedwater - C003, C005,  !

Includes safety / relief valves (SRVs), including E013,E017,E020,E104,El15,E117,E206, pressurizer SRV and power-operated relief E211, E213, E228, E248, E255, E323. E325, valves (PORV) T302, T319, T327, T337, C404, P405, E323 Rev.  ;

I, E415, E418, T402, T416, P501, E502, E514, D. Steam Generators (SG) C003, C005, C101, T5d2, NUREG-1154, C602, T603, E709, P701, E101, E224, T330, E413, E423, T512, E708, T703, E906, T927, E90-03, T91-04, S93-01, E906, E909, E96-03 NUREG 1275 Vol.10 Includes SG shell, internal tubing, support plates, Includes condensate, booster, feedwater (FW) and nozzles, manways, blowdown lines auxilkrv feedwater ( AFW) pumps, condensate 8 System descriptions contain meerial copyrighted by Nuclear Power Experience. Material reproduced by permission of Hagler BaiUy Consulting. Inc.

E-1-1 Appendix E-1

1997 AEOD Annual Report storage tank (CST), demineralizer system, LP and system, boric acid storage tank, boron recycle HP heaters, associated valves and piping :y: tem, letdown lines and valves F. Circulating Water E016, C204, E243, E311, B. Auxiliary Cooling - E001, E005, E006, E012, T318, T333, T804, T90-06, T90-16 C105,Elli,El15,E132,C202,E223,E231, Includes intake structures, screens, cooling towers, E256,E257,E302,E303,E304,E310,E311, 1 discharge gates and canals, associated pumps and E315,E321,E202,T303,T305,T341,E411, i valves, saltwater system T403,T415,C503,E502,E506,T505,T514, '

T602, E704, E710, S702, NUREG 1275 Vol. 3, i G. Electrohydraulic Control (EHC) System -

E806, E807, T804, E907, T919, T926, E90-02, C102, E247, E249 E90-06, T90 11, T90-13, T91 0.? , T91-03, E91 01, includes EHC Duids, auto- stop oil, interface S93-03 valves, va!ve operators for HP and LP turbine.

Includes residual heat removal O HR), component pumps and associated controls cool ng water (CCW), service er (SW) and >

H. Miscellaneous E416 Includes heater drain essential raw cooling water sy 1s, pumps, heat system, extraction steam exchangers (HX), shutdown coeing, associated valves and piping VI1. Safety Systems A. Emergency Core Cooling System (ECCS) - C. Miscellaneous - None E111, E207, E238, E317, T310, T319, T328, includes RCS drains, containment sump valves T335,E404,T403,T422,S603,E606,T708,E803, IX. Instrumentation and Control (I&C)

T90 14, NUREG-1275 Vol. 9, T93 01, T95-02 A. NuclearInstrumentation - T92-05 I&C for Includes safety injection (SI), upper- head injection incore neutron Hux monitoring, including source systems, accumulators, boron injection tank and intermediate range monitors (SRMs and IRMs),

power range monitors, related amplifiers and B. Containment Pressure Suppression - E316, indicators T317, T338, TS 13, E710, T811, T926, T90-17 Includes containment spray, ice condensers, recircu. 3. Reactor Protection System (RPS) - E103, E014, lation spray, chemical addition tanks E133, E245, P301, T316, T320, P406, E323 Rev.1, E419, E421, P504, T503, T90-07,94-03 C. Containment Atmosphere Cooling - E012, I&C for manual or auto reactor trip channel actua-E203,E221 tion, including RTD, reactor trip breakers, pressur-Includes cc,ntaiu ent fans, containment air izer pressure and level transmitters, SG level recirculatte transmitters and FW Cow transmitters. Includes D. ContainmentIsolation - E009, E011, E221 Includes containment isolation valves (CIV), C. Reactor Control- C004, E323, E507, containment NUREG 1195 E. Miscellaneous - E204, E229, E230, E233, Includes the integrated control system (ICS), axial T331, T339, T424, T608 flux monitors, control rod positioning, and other rod and core performance monitoring and control Includes fire systems, containment hydrogen I&C venting, purge and recombiners, security systems, respirators D. Turbine Cycle - E017, E228, T706, Vill. Auxiliary Systems NUREG 1275 Vol.11 A. Coolant Volunu, Purification, Chems. cal I&C for manual or automatic turbine trip channel Sampling - C102, E308, E314, T415, E512, T501, actuation and turbine generator and FW control, T504, E607, T702, E910, T91-02, NUREG 1275 including EHC, vibration and wear probes, gover-  !

Vol. 9, T93-01, T95-02 nors, FW and AFW Cow Includes chemical and volume control system (CVCS), charging pumps, post-accident sampling NUREG-1272,Vol. I1 No.1 E-1-2

I Reactors l

E. Safety Systems - E019, E102, E103, E112, E251, NUREG/CR 3122, C301, E320, T301, E114,E121,E226,E227,E321,T308,T310, T311, T315, E401, E412, T404, T418, T424, l T313, T320, T335, C402, E404, E409, E419, E504, C605, NUREG-1190, E703, E905, T919, l T405, P503, E508, E515, C604, E605, T606, E90-01, E90-05, T90-03, T91-07, S91 01, S92 03, .

I T612. T904, NUREG-1275 Vol. 8, T93-03 S93-06, E93-02, T96-03 Includes I&C for actuation of ECCS, engineered Electricti Jistribution systems include buses, safety features (ESFs), solid state protection breakers, inverters, transformers, motor control ,

system, fire systems, containment pressure suppres- centers (MCC), switchgear, on- and offsite distribu-sion and isolation, and main steam isolation, tion lines, and associated I&C refueling water storage tank level, toxic gas isola-H. Liquid Radwaste System - None l tion system (TGIS), TGIS butane monitor, contain-includes liquid and solid radwaste tanks, evapora-ment sump level, borated water storage tank level, tors, filters, valves, chemical drains, piping, associ-steam line differential pressure, RPV level, SG level ated I&C and flow, auxiliary alarm annunciator X10. Gaseous Radwaste System - E120, F. Process Systems - C004, E314, T331 E327,T411 I&C for process computer, RCP pressure seal includes waste gas processing, auxiliary building sensing, CVCS tank level, heat tracing controls, accumulator level, containment fan coil unit SW gas treatment, waste gas decay tank, compressor, gase us hydrogen recombiner, filters, stack moni-llow, and acidity or alkalinity instruments tors, associated I&C  ;

G. Reactor Coolant Control- T902 XIV. Buildings and Containment I&C for RCS flow, subcooling monitors, pressur-A. Anetrations - E701, E808, T804, T916 izer level (B&W)

Includes airlocks, hatches, manways, electrical ar d H. Miscellaneous - E123, C204, E302, T333, piping penetrations, fire doors, seals, gaskets to T401, T504, T506, T804, T92-06 containment and among plant buildings  ;

I&C for containment sampling and monitoring,  !

B. Room - E229, E306, E611, T611, E802, T909, general area radiation monitoring, I&C air, incore E90-07 thermocouples (T/Cs), gaseous nitrogen system valves and loose parts monitor Control room, remote shutdow n panel, cor. trol room ventilation, auxiliary building, turbine building X. Fcel Handling Facilities and Systems -

E242, E313, S96 02 C. Afiscellancous - E124, E304, T413, includes reactor cavity, refueling canal, fuel transfer T710,T909

! system, spent fuel pool, and racks, new fuel storage, Includes heating, ventilation aad air conditioning cranes and lifting devices, tools and fixtures, and (HVAC), fire dampers, charcoal absorbers, contain- ,

l associated I&C ment purge butterfly valves and purge isolation l valves l XI. ElectricalSystems  ;

l A. Eme gency Power- E008, E126, E220, E302, XV. Afiscellaneous Systems - E219, E406, E412, E251, E253, E254, E302, E307, E318, E334, E423, C501, S503, E501, C603, E702, NUREG.

P401,E424,E427,E510,E514,E612,T914, 1275 Vol. 2, NUREG-1275 Vol. 6, E902, T914, T925, S91-01, T92-08, T92-10, E93-03, S96-03 T928. E90-08, T90-04, E92-01, T92-03, T92-04, f E96-01 i

Includes batteries, diesel generators (DG), battery i chargers, motor generator (MG) sets, and associ- Includes plant air synems, snubbert pipe and ated I&C building supports, ponradioactive w aste neutraliz-ing systems, general valve operator problems, B. Other Electrical- C003, E004, E008, E010, rupture discs ard rescue breathing apparatus, C104, E102, E110, E116, E131, E102, E11re, auxiliary systems E116, E131, E210, E212, E222, E234, E246,

, - E-1-3 Appendix E-1

1997 AEOD Annual Report j i

XVI. Operational Problems C. Miscellaneous - E411, E704, NUREG-1275 l

~

A. Inscivice inspection - T327, E612, T805, E906, Vol. 8, S%-02, T95-03 E910. T95-02 Includes operator errors and procedural problems Includes operational problems arising from sched- relating to the full range of plant systems, uled inservice inspections (ISIS) especially those involving. radiation exposure ,

or contamination  !

H. Refueling - E205, E806, S96-02 includes operational errors occurring during initial  !

fuel load, refueling or spent fuel handling i I

t t

I NUREG-1272.Vol. I1 No.1 E-1-4

APPENDIX E-2 BWR Plant Systems P

f 1

Reactors I l

CONTENTS c

I. Fuel................................................................... .......................................E-2-1 II. Reactor Intern als ... ...... ............. ... .. . . . ... . ... . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-2-1

' III.' Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . ... ........... E-2-1

_ IV. Control Rods and Drives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-2-1 V. Recirculation, Steam and Relief.. ........ ..... ......... .. .. ...... ..... ..... ....... ... ..... . . .. . E-2-1 I VI. Turbine Cycle Systems.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . . E-2-1 VII. Safety Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . .... .... E 1 VIII. Auxiliary Systems . .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. E-2-2 IX. Instrumentation and Control (I&C).. . . .. . ....... ........ .. . ... ......... ..................... . E-2-2  :

1 l

X. Fuel Handling Facilities and Systems ... . . . . . . . . . . . . . . . . . . . . . . . . . . . ..............................E-2-2 XI. Electrical Systems . ....... . . .. ....... ... .. . .. .. . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . E-2-2 XII. Liquid Radwaste System ... . .......... ...... . .. . . .. . ... ....... ... . . .. . . . . . . . . . . . . . . E-2-3

]

XIII. Gaseous Radwaste System . . . ....... ....... ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ....... .. E-2-3 XIV. Buildings and Contain ment .... . . ..... . . .................. ... .... .... . . .... ............................E-2-3 XV. Miscellaneous Systems . ..... ............._....................................E-2-3 XVI. Operational Problems.... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-2-3

)

i I

i 1

1 E-2-iii Appendix E-2

.-_ -. - -- - . = .- .- -- _ - -- . . . - . . - .

l Reactors l BWR PLANT SYSTEMS 2 I. Fuel- E205 B. Generator E209 Includes uranium fuel pellets and cladding, fuel Includes rotor, stator, exciter, bearings, voltage l assemblies spacers, tie plates and channels regulator, core monitor, generator cooling systems II. Reactor Internals - E002 C. Condensers None Incindesjet pumps, FW and core spray spargers, Includes tubes, baffles, spargers, shell, water box, steam dryer assembly, core support and guide, core hotwell vacuum systems (air ejector, vacuum shroud pump), expansion joint ,

l III. Reactor Vessel - E114 D. Steam - T323, T417. T605, E801, T801 l

Includes RPV, lines and nozzles - FW and core Includes turbine bypass system, reheaters, moisture spray, control rod drive (CRD) retum, recirculation separators IV. Control Rods and Drives E. Condensate and Feedwater-T90-09 A. ControlRods -T340,T510,T712,S803 Includes pumps, LP and HP heaters, condensate includes rods, sheaths, blades demineralizer system, CST B. Drives - C001, C002, E007. E015, C103, E225, i F. Circulating Water - E113, E214, E215, T323, E240, T306, C403, E403 6

includes CRDs, hydraulic control units, hydraulic Includes intake structure, discharge canal, circulat-supply system, scram discharge header ing water pumps, dilution pumps, cooling towers, V. Recirculation, Steam and Relief cooling water pumps A. Pumps - E107 Includes recirculation pumps, drives, and seals, G. Miscellaneous - E416 speed controls, recirculation manifold includes extraction steam pipes and valves, heater drain system B. Piping - None includes main steam lines, suction and discharge VII. Safety Systems risers, flow restrictors, bypass lines A. Reactor Core Isolation Cooling (RCIC)-

T420, T610, E904, NUREG-1275 Vol. 9, T93-01, C. Relief and Safety Faires - E240, E312, E322, NUREG-1275 Vol.10, T95-02 E502, T610, S92-02,T96-02 Includes RCIC pump, drive and speed controls, and includes SRV, Includes SRV, MSIVs, automatic associated piping and valves depressurization system (ADS) valves l E. Standby liquid Coistrol(SBLC)-T507, D. Miscellaneous - None T91-02, T92-07 includes recirculation loop valves (dra.in valves, includes SBLC pumps, tank, explosive valves, sample isolation valves, flow control valves)

Pii P ng VI. 'Ibrbine Cycle Systems C. Core Spray (CS)- E511, E513, NUREG 1275 A. Turbine - None Vol. 9, T95-02 Includes rotor. shaft, bearings, blades, casing.

valves (admission, stop, control, intercept), cross- Inc{udes HP and LP core spray pumps, valves, Piping over piping, lube oil system 2 System descriptions contain material copyrighted by NucIcar hwer Experience. Material reproduced by permission of flagler Bailly Consulting, Inc.

E-2-1 A Ppendix E-2

~ . ._ _. - _ ____ _ _ . . _ . _ _ _ _ . _ _ _ __ .

1997 AEOD Annual Report D. Residual Heat Removal (RHR) - E105, E106, position indication system, I&C for core perfor-El19,E125,E218,E236,E244,E309,T332, mance, power, mode changes T334, E411, E414, E417, E502, T515, S603, E601, E608, E609, E908, NUREG-1275 Vol. 9, D. Turbine Cycle -T417, NUREG 1275 Vol. I1 T95-02 EHC system, including electric pressure regulators, Includes low pressure coolant injection, contain. mechanical pressure regulators, FW flow control-ment coolers and shutdown cooling systems ica s, condenser hotwell and heater level controls (including HXs), associated valves and piping E. Safety Systems - E109, El14, El18, C201, E. High Pressure Coolant Injection (HPCI)- E226, E227, T325, T336, C402, E405, E407, E235,E402,E407,E422,E425,T410,T508, E408, E409, E425, T403, P503, E508, C604, T516, E904, T906, T922, NUREG 1275 Vol. 9, 21. E605, E610, T612, NUREG-1275 Vol. 8, NUREG-1275 Vol.10, S95-02, T95-02 T93-03 Includes HPCI turbine, pumps, drives and speed I&C for ECCS, ESF, and other safety system controls, associated valves and piping actuations, including rod worth minimizer, isolation condenser, RCIC, SBLC, CS, RHR, HPCI, standby F. Miscellar:cous - E204, E229, E233, E240, gas treatment (SBGT), ADS, torus, main steam line, E319, T304, T331, T336, C502, E511, E601, and fire protection systems T601, T608, T923, T90 16, NUREG 1275 Vol. 9, S96-01 F. Process Systems -T309, T331 Includes isolation condenser systems, CIVs, fire I&C for process computer, RWCU, flow, level, and protection systems, containments, drywell pressure detectors, transmitters, and recorders VIII. Auxiliary Systems G. 3fiscellaneous-E232,T92 06 A. Reactor H'ater Cleanup (R WCU) - T307, T329, includes I&C for containment sampling and E705 monitoring, leak detection, data acquisition, seismic Includes regenerative and nonregenerative HXs, and sonic detection and instrument air systems filter-demineralizer units, RWCU pumps i X. Fuel Handlir.g Facilities and Systems S96 02 B. Reactor Building Closed Cooling Water - None Includes refueling bridge platform, grapple, spent Includes pumps, surge tank, coolers, HXs fuel pool and racks, and assochted I&C, ventilation C. Afiscellaneous C202, E505,NUREG-1275 XI. Electrical Systems Vol 3, E807, T90-13, T91-01, S93-03 A. Emergencypower E108, E109, E126, E129, 1 Includes SW systems, steam line drains, sumo drains E241,E307.E324,T336,P401,E401,E427,

, 10, IX. Instrumentation and Control (I&C) A. , , , , ,

Nuclear Instrumentation -S803 9 3, N3 Incore neutron flux detection I&C, including Includes DG, gas turbine generators, alternating traversing incore probes, SRMs, local power range current (ac) uninterruptible power supply UPS, monitors, IRMs, average power range monitors direct current (de) backup, MG sets, safety buses, batteries, and battery charger  ;

B. Reactor Protection System (RPS)- EI10, T306, P406, E412, T403, P504. 7905, T'3 0 07, B. Other electrical- E107, E108, E246, E250, T94-03 NUREG/CR 3122, C301, T312, T336, E420, Trip channel systems for manual or automatic T407, T414, E504, T509, E605, E804, T903, control rod scramming, safe reactor shutdown. T915, T921, E90-01, E90 05, E90 10, T90-12, including ATWS backfits S91-01, S92-03, S93-06, E93-02, S96-03 Includes main unit transformer, auxiliary trans-C. Reactor Control- E018 former, safeguards inverters, MCCs, buses, break-Includes rod sequence control system, manual rod ers, relays, fuses, switchgear, on- and offsite control system, rod block monitor system, rod distribution lines NUREG ?2, Vol. I 1, No.1 E-2-2

-. _ _ -. . . _ . - _ - . - - . - . ~ _ - __. . _ _ _ .- --

Reactors i

XII. Liquid Radwaste System - None XV. Miscellaneous Systems includes concentrator, demineralizer, filters, collec- E007, E219, E406, E412, E414, T419, C501,

tor tanks, drain tanks, sample tanks, surge tank, S503, E501, C603, E702, NUREG 1275 Vol. 2, CST, spent resin tank, solid radwaste separators, NUREG 1275 Vol. o, T914, E92-01, T92-03, T92-l centrifuges, and hopper, and associated I&C OA E %-01 I n;.udes plant air systems, auxiliary boilers, seismic XIII. Gaseous Radwaste System - None and component restraints (hangers, snubbers, etc.),

Includes stack gas and offgas charcoal absorbers, general valve operator problems

cryogenic distillate systems, sample pumps, recombiners, high-efficiency particulate air filters, XVI. Operational Problems monitors, analyzers, and other I&C A. Inservice Inspection -T805,T95-02 Includes operational problems arising from sched-3 XIV. Buildings and Containment uled ISIS A. Penetrations - C103, T412, E808 includes ,

~

airlock, manway, hatch, electrical and tubing B. Refueling - E205, E612, S96-02 l penetrations, seals, and gaskets to containment and Includes chiefly errors arising from mishandling of

] among plant buildings equipment during periods of removal of RPV head  ;

i for initial fuel loading, refueling and spent fuel  !

B. Rooms - E229, T406, E603, E611, T909, handling ,

E90-07 Control rooms, remote shutdown panel,  !

control room ventilation, auxiliary building, turbine C. Miscellaneous - C92-01, S96 02,T95-03 building Includes operator and personnel errors, procedural problems relating to the full range of plant systems, C. Miscellaneous -E322,T307, E410,T421, particularly those concerning exposure to radiation T710, T713, E802, T903, T909 Includes HVAC or radioactive contamination systems, suppression chamber (torus) pressure

- suppression systems, containment atmosphere

> dilution systems, SBGT systems, vacuum breakers, gaseous nitrogen systems, cranes

}

L E-2-3 Appendix d-2 e

)

e APPENDIX E-3 Activity / Human Factor Deficiency I

J

. . . . . . _- _ . . _ . - - . ..- - - - . - . . - - _ _ - ~ . - _ . . _ _ , . _ . - -

1 Reactors  ;

i i

CONTENTS  !

I. . Administrative /Procedmes ... .... .. . . ..... . . .. .. . .. . . . . . . . . . . . . . . . . . . . . . . . . E-3-1 II. Con struction .. .......... ... .. ... ... . . .. . . . . . . . . . . . . . . . . . . .... ... .... . . ... .. E-3-1 1

III. Design................................................ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. E-3-1 l i

- IV. Fabrication, Part 21, Quality Assurt.u . -- ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-3-1  :

V. Installation ... . . . . . . . . . . . . . . . . . . . . . . . . . . .... ... . .... ... .............. ... . ... ....... E-3-1 VI. Maintenance ..... ... . . ... . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... E-3-1 VII. Operation . . . .. . . . . . . . ....... . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . E-3-1 j VIII. Radiation Pic::ction...... . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . E-3-1 I i

IX. - Test and Calibration ..... . . ..... ......... .......... ....... . . . . . . . . ... . . . .. . . . .. . ... . .. . . . E 1 1

X. 1

. Licensee Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E-3-1 t.

t I

l t

i k I

,- i 4s  !

e 1 lj' I i

i

, 1 I l i

E-3.iii Appendix E-3 4,

Reactors ACTIVITY / HUMAN FACTOR DEFICIENCY I. Administrative /Procedmes VL Maintenance C002 E004, E008, E010,::.213. E306. C204, E237, S401, E401, E403, E410. E414, E323, T306, T313, T320, T321. TM8, IM 25, C503, E504, T511, C605, E607, E608, E426, T416, C503, T502, T5P ;, T510, T512, T612, E707, E708, T701, T704, S804A, C607, C603, E602, EQF, Tul2, E705, S804B, E802, T809, NUREG-1275 Vol. 6, T710,T713,b,DI,S80L7Fdi,T806,E90- S901 Rev.1, E901, T902. T912,1913, E90-02, E90-07. T90-03,190- 12, NUREG-1275 03 E90-07, S91-01, S92-02, NUREG-1275 Vol. 8 E92-01. T92-07, S93-05, E93-01, V61. 9, E92-01, T92-01, 'I92-04, T92-09, T93-01, S95-01, E95-01, S96-02 s?3-05, T93-01, T94-02, T94-04, S95-01, M6-01, E96-03, T96-02 II. Construction 5707, T91-06 VII, Operwlon E221,1223.T328,T340,E602,T708,T712, Ili. Design S803,E801,E802,E803,E901,E909,T909, E012,2013. EJ17 E018, E213, E225 E235, NUREG-1275 Vol. 8, S93-05, E93-01, S95-E308.T301,T302,T303,T308,T319,T325, 01, E95-01, T05-03, S96-02 T329,T336,E407,E408.E410,T408,T421, E502,E511,C602,C604,C605,E604, Vill. Radiation Protection E607, E611 E707, E708, E709, E710, T703, None T708,T710,S803,E802,E803,T805,T904, l T909 T914, E90-07, T91-01, NUREG-1275 IX. Test and Calibration Vol. 9. T94-02, S95-01, E96-0l, E96-03, EI29,E318,E320,T304,T305,T313,E410, S96-02 E414,E420 E421 E425,T410.T424, C503, E512, E515, T510 C605, E90-03, IV, Fabrication,Part 21, Quality Assurance E90-07, E90-08, S92-02, E92-01, T92-01, ,

S401, E403, T410, T805, T914 E96-03 T92-04, T92-05 T92-07, NUREG 1275 Vol. l 11, T95-02, E96-03, T96-02 l V. Installation E408,E424,E611,T701,T704,T805,T914, X. Licensee Program 190-03, E96-01, E96-03 E96-02, E96-02, Supplement 1 E-3-1 Appendix E-3

APPENDIX E-4 Topics s

_ _ . - . . _ _ - . ~ . - . . - . _ _ . . - - . - . - .- - . - _ .- -- . _ - - . .

i J

t Reactors '

s r

t P

CONTENTS I. Anticipated Transient Without Scram (ATWS) ..... .. .. ........ ..... ....... ... . .... . ... .... .... .. E-4-1  !

r

~ II. Stress Corrosion Cracking and Variations .... .. ........ ....... ..... ....... . ..... . .. .. ...... ....... ... . .. .... E 1 i i

III. Loss of Ofisite Power (LOOP) .. ... .............. .... .. ... .. .. ...... ... .. . .... .. .. . .. . ... .. . . .. .. . . .. . ...... .... E 1  ;

IV. Unplanned Criticality . ... .... . . .. .. ......... .. .. . ... . .... .. . ..... . . . .. .. . . . . . . . . . .. . .. .. ... ... .. ...... . . E 1  !

t L

V. Fore ig n Reac t or . . . . . . . .. . . . . ... . . . . . . . . .. . . . . . . . . .. . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . .-. . . . . . . ... .. . .. . . . .. . . . ... . . . . E 1 VI. Weather Related ..-............ . ... . ....... . ........................................................E-4-1 VII. Natural Circulation .. ..... ........ . .. . . ..................................................................E-4-1 VIII. Tran sien t . .. . . . ... . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . .. .. .. . .. . . . .. . . .. ... . . . . . . . . .. . . . . . . . . . . . .... ...... E 1 IX. Loss-of-Coolant Accident (LOCA) ..... ..... .. ... .. .... . .. .... .. ........ .. ... . . .. .. . .. . .. .. . . .. E 1 X. Flooding............................................................... .... . . . . . . . . . . . . . . . . . . . . . . ... ... E 1 i s

\

XI. Water Hammer.... . . . .. .... . . . . . . . . . . . . . . . . . . . . . . .. ... .. . .. . . . . . ....... E-4-1 -

. t

- XII. Flooding of Steam Lines and Steam Generator Overfill .... . ... . . ...... .. .. .. . .. . .. ..... .. ... . . E-4-1 l XIII. Single Failure, Common Cause, Common Mode .. ............. ............. ...... ....... ..... ............... ... . E 1 ,

XIV. Paralleling....................................................................................................................E-4-1 XV. Valves.................................................. . . . . . . .... . . . . . .. . . . .. .. .. .. . .... . E 1 XVI. Clams, Bivalves, and Debris.... .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ... . ....... .. E-4-2 XVII. M ode S witc h . ... ... . . . . . ... . . . . . . . .. .. .. . .... . .. . . . . . . . . .. . . . . .. . . . . . . . . . . . . . . . .. . ... . . . . . . ... .. E-4-2 i

r j XVIII. Shared Systems . .... .. . ......... . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-4-2 l

4

-XIX. Recirculation Actuation Signal ... .. .. .. ........ . .. .. . . ... . ..... .. . . . . . ...... . ..... . E-4-2 i

t

. XX. Blowdown ....... .. . .. ... . .... . ... .. . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . .... .. .. ..... ... E-4-2 i

j XXI. TMI Precursor Event .. ...... ........ ................ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . . . . . . ..... E-4-2 l J

i i

i

l. E-4-iii Appendix E-4  ;

- y ~- y - -i----.,w-F -q

. . . ... .- . . .. .- _- - . _ . , - - - . _ - - . . ~ - . .-

1997 AEOD Annual Report CONTENTS (cont.)

XXII. . Sabotage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. .. . . . . E-4-2 >

f XXIII. Fire.,............ . . . . . . . . . . . . . . ... . . . . . . . ... . . . . . .. .. . .. . E-4-2 XXIV. Corrosion and Erosion .. ....... .. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . E-4-2  ;

XXV. Steam Generator Tube Rupture.. .. .. . . . . . . . . . . . . . . . . . .. .. .... .. E-4-2  :

XXVI. NET Positive Suction Head (NPSH) and Pump Runout... . ..... .. .. ..... . E-4-2 XXVII. Overpressu re . . ... . .. ... ... . . .. . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . E-4-2 i

XXVIII. Piping . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-4-2 XXIX. Stratification. .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . ... . . E-4-2 XXX. Safety Injection Actuation Signal Bypassed or Blocked . . .. . . . .. .. ..... ...... .... E-4-2 XXXI. Explosion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . E-4-2 XXXII. Harsh Environment . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... ...... .... E-4-2 XXXIII. Final Safety Analysis Report (FSAR)..... ...... .... .. . . . . . . . . . . . . . . . . . . . . . . . . E-4-2 XXXIV. Fastener ...... ..... . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . ... E-4-2 -

XXXV. Operational Experience - General .._.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... . .. .. . E-4-2 XXXVI. Risk Assessment.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... E-4-2 XXXVII. Component Aging .. . .. . .. ... . .. . .... . .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . ... .... .. E-4-2 XXXVIII. System Reliability .. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... E-4-2 i

l 1

NUREG-1272, Vol. I 1. No.1 E-4-iv  !

l 2

Reactors TOPICS I. Anticipated Transient Without Scram XI, Water Hammer (ATWS) C005,E104,E105,E106,E309,T327 j C001, C002, C103, El33, P301, E323 Rev. T329, T337, E402, T502, T516, NUREG-1,E503,S803,E804 1190, E91-01 II. Stress Corrosion Cracking and Variations XII. Flooding of Steam Lines and Steam E242 E313,T402,E506,E613,T906,T910, Generator Overfill E96-03 C005, E013. E017, E303, E801 III. Loss of Offsite Power (LOOP) XIII. Single Failure, Common Cause, Common C003, E253 E302, E401, E413, E605, Mode E610, NUREG-1190, E703, T915, T925, E004, E109, E116, E125, C204, E219, E90-01, E90-05, T90-12, T91-03, T91 -07, E230,E302,E304,E311,E325,T302,T304, S91-01, T92-08. E93-02, T94-01, T95-01, T313, T336, T339, C404, E403, E405, T96-03 E408,E410.E426,T410,T418,T421,E503, T505, T507, T509, T513, T515, C604, S603, l IV. Unplanned Criticality E702, E709, T703, T708, NUREG-1275 Vol.

T712,S803 3, E802, NUREG-1275 Vol. 6, E907, E910, T909, T919, E90-05, E90-07, T90-11, l V. Foreign Reactor NUREG-1275 Vol. 9, E92-02, T92-05, T94-T712,E706,E805,T803 04, T95-02, E96-01 VI. Weather Related C003,El12,E226,E401,T405,E605, XIV. Paralleling T90-09 E008,E010,T921 VII. Natural Circulation XV. Valves C003, C101, E413, S96-01 E009, E012, C102, C103, E104, E120, E122,E124.E128,El32,E208,E211,E225, VIII. Transient E235,E237,E239,E248,E304,E305,E311, C004, E014 E104, E114, C205 E206, E312,E315,E320,E325,T303,T314.T318, E221 E238,E240,E246,E247,E249,E306, T319,T321,T326,C404,S402,E403,E414, E323 E326, C403, P405, E323 Rev.1, T410,T420,C502,E502,E505,E506, E413,E415,E418,T417,P501,E509,E514, E511, E513, T507. T513, T514, C603, E601,  !

T605, NUREG 1195, E708, E801, T80!, E702, E904, E905, E909. E90-09, T90-08, NUREG-1275 Vol. 8, T92-02, T96-03 XV. Valves (cont.)

E705, E706, T801, NUREG-1275 Vol. 6, IX. Loss of Coolant Accident (LOCA) E906, E908, E909 T918, T927, T928, E90-C004, El 12, E223 E253, E302, E322, 02, E90-06, E90-09, T90-04, T91 -04, T91 -

T301.T318,C403,E417,C502,T506 05, S92-02, NUREG 1275 Vol. 9. T92-01, E704, E705, E710, T707 E805, S902, T9I- T92-03, T92-04, T92-09, NUREG-1275 Vol.

03, T94-04, S95-01 10 NUREG-1275 Vol. I 1. T95-02, E96-01, X, Flooding E221, E225, E229, T514, E705, E90-07, T91-06, T92-06 E-4-1 Appendix E-4

. _ _ . . .~

1997 AEOD Annual Report XVI. Clams, Bivalves, and Debris XXVII. Overpressure E016. El 11, El19, E123, C202, C204, E248, C401, EB03, E90-09 E202,E215,E219,E220,E318.T305, T307,T402,T419,T422,E512,T513, XXVIII. Piping T609, NUREG-1275 Vol. 3, E905, T9I6, E255,E308,T314,T322,T337 T923, T90-16, S93-03 T341,E612,E705,E902,S902, T90-16, E92-01 XVII. Mode Switch E018, E90-10 XXIX. Stratification E256,E415,S902 XVIII. Shared Systems E507,E510 XXX. Safety Injection Actuation Signal Bypassed or Blocked XIX. Recirculation Actuation Signal E326 T310, E909, NUREG-1275 Vol.

E019,T335,T606,E710,E803,T916 8, E95-01 XX. Blowdown XXXI. Explosion C103, E218, E239, E706, E909, E90-02, E327,E902 S95-01 XXXII. Harsh Environment XXI. TMI Precursor Event T302, T92-06 El15,El17,E120.E216,E326,P402, E909 XXXIII. Final Safety Analysis Report (FSAR)

T319. T323, E423, E612, E90-08 XXII. Sabotage El13 T322,T903 XXXIV. Fastener E424,T906 XXIII. Fire El16,E120 T404,T418,T608,E902, XXXV. Operational Experience - General E905 T903, T915, E90-01 NUREG-1275 Vol.1, NUREG-1275 Vol. 4, NUREG-1275 Vol. 5, NUREG/

XXIV. Corrosion and Erosion CR-4674 Vol.1-22, S96-02 El30 E312,T318,T341,E411,E416, E908, T90-13, S93-03, E96-03 XXXVI. Risk Assessment NUREG/CR-4674 Vol.1-22, S96-02 XXV. Steam Generator'Ibbe Rupture E024, T330, E708, E909, E96-03 XXXVII. Component Aging NUREG/CR-6442 XXVI. NET Positive Suction Head (NPSH) and Pump Runout XXXVIII. System Reliability E213, E214, E218, E256, E257 E302, S95-02,S96-01,S96-03  !

E314, E323, E325, E326, C404, E323 Rev.1, E411 T515, E606 T603, E709, E710,T703,S702,E803,E806,E807, E910,1916, T927, E90-06 NUREG-1272, Vol. I 1, No.1 E-4-2

APPENDIX F Status of AEOD Recommendations

I Reactors l l

1 1

l 1

STATUS OF AEOD RECOMMENDATIONS l l

This appendix summarizes the year-end status of all AEOD recommendations that are either new or not resolved since the last report. In the 19% AEOD Annual Report, all recommendations that had been outstand-ing were resolved. Because no new recommendations have been added, there are no AEOD recommendations outstanding.

1 i

1 l

)

l l

l l

i F-iii Appendix F

- -.- _-. . .- . -- _ _ = - - - . - - . _ . --

1 1

' APPENDIX G l l

Status of NRC Staff Actions for 4 Reactor Events Investigated by Incident investigation Teams J

1

\

4 4

Reactors

! l CONTENTS i

l Staff Actions Resulting from the IIT Repon on Loss of Vital AC Power and the Residual Heat Removal l System During Mid-Loop Operations at Vogtle Unit I on March 20,1990. . . . .. ... .. G- 1 l

Staff Actions Resulting from the IIT Repon on the Unauthorized Forced Entry into the Protected Area at Three Mile Island Unit 1 on February 7,1993. . . . .. G-4 l

l l

l l

l l

G-iii Appendix G

1997 AEOD Annual Report STATUS OF NRC STAFF ACTIONS FOR REACTOR EVENTS ,

l INVESTIGATED BY INCIDENT INVESTIGATION TEAMS In accordance with NRC Management Directive of the actions taken by the responsible office (s) to 8.3, "NRC Incident Investigation Program," da'ed confirm tht.t pertinent aspects of each IIT finding August 12,1992, upon receipt of an Incident are addressed in the implemented resolution, and Investigation Team (IIT) repon, the Executive documenting the resolution of all staff actions.

Director for Operations (EDO) shall identify and Actions whose resolution are reviewed and ap-l assign NRC office responsibility for potentially proved by the Commission are not subject to j industry-generic and plant-specific actions resulting independent revie's oy AEOD.The independent .

I from the investigation that are safety significant and assessment should be completed by the end of the warrant additional attention or action. Office calendar year following the year in which the staff Directors designated by the EDO as having respon- action was reported as resolved by the responsible sibility for the resolution of issues or concerns are responsible for providing written status repons on office (s). The EDO resolves any conflicts between )

AEOD and the responsible office (s) regarding the

, the disposition of assigned actions. Follow-up adequacy of the actions taken by the staff.

( actions associated with the IIT report do not .

This appendix summarizes the disposition of each necessarily include all licensee actions, nor do they f the action items that the EDO assigned to the cover NRC staff activities associated with normal v ri us NRC offices as a result of the findings event follow-up, such as authorization for restart,

} associated with completed IITs at reactor facihties.

plant inspections, or possible enforcement actions.

The descriptions are as of the end of FY 1997. This These items are expected to be defined and imple-appendix presents the status of staff actions that mented through the normal organizational structure were n t documented as resolved in the 1996 and procedures.

AEOD Annual Report, NUREG-1272, Vol.10, AEOD is responsible for monitoring the status of No.l .

the assigned staff actions, evaluating the adequacy l

f NUREG-1272 Vol. I1. No. I G-iv

Reactors AEOD IIT TRACKING SYSTEM Action Source: IIT Report on Loss of Vital AC Power and the Residual Heat Removal System During Mid-Loop Operations at Vogtle Unit 1 on March 20,1990 (Reference 1)

Item 1: Adequacy of Shutdown Risk Management Action (a): Review existing regulatory guidance related to shutdown risk control and issue such new guidance as may be needed. Include the following in the assessment of shutdown risk management: normal and standby electrical systems and sources, including switchyard equipment; normal and alternate cooling systems; special alternate plans for loss of forced circulation; fission product barriers, including primary and containment systems; and special activities such as movement of heavy loads or constmetion activities. (Responsible Office: NRR)

Status: Resolved (Pending AEOD Independent Review)

Although this action was documented as resolved in the 1993 AEOD Annual Report, based on publication of NUREG-1449 (Reference 2), it was subsequently reopened and remained recategorized as ongoing by AEOD pending Commission final decision on a rule to manage risk during shutdown operations.

In SECY-97-168 (Reference 3) the Commission was informed of the staff's plans to re-issue for public comment a proposed rulemaking package addressing shutdown and spent fuel pool storage operations at nuclear power plants. The proposed rule involved significant i changes from a previously proposed rule that had been published in the Federal Register (Reference 4)in October 1994. The proposed new rule was developed following numerous comments and Commission guidance regarding the use of a risk-informed and perfor-mance-based approach for new regulations. In a Staff Requirements Memorandum for  ;

SECY-97-168 the staff was directed by the Commission to not proceed further with the i proposed rulenaking package. Alternatively, the Commission requested the stafT to evalu- f ate whether further codification of current industry practice related to ensuring that risk-significant safety functions are maintained during shutdown operations beyond that contemplated in the revised maintenance rule would provide sufficient benefit to warrant the costs imposed on licensees and the NRC by such codification. After the modification to the Maintenance Rule,10 CFR 50.65, is completed the staff is to interact with stakeholders on this matter and present their conclusions. The staff was expected to continue to monitor licensee performance, through inspections and other means, in the area of shutdown operations to ensure that the current level of safety is maintained.

In a Staff Requirements Memorandum dated December 17,1997,(Reference 5) for SECY-97-173, on a potential revision to the Maintenance Rule, the Commission approved the staff's recommendation to develop rulemaking to revise 10 CFR 50.65(a)(3) of the rule to require licensees to perform safety assessments prior to performing maintenance activities.

The SRM stated that the ste.ff should consider adding as a preamble to the Maintenance Rule language that states that the requirements of the Maintenance Rule are applicable to all plant conditions of operation, including normal shutdown operations. The Commission also directed the staff to designate the final sentence of section 50.65(a)(3) as (a)(4) and revised to read as follows: " Prior to performing maintenance activities on SSCs within the G-1 Appendix G

1997 AEOD Annual Report scope of this section (including, but not limited to surveillance testing, post maintenance testing, corrective maintenance, performance / condition monitoring, and preventive mainte-nance) an assessment of the current plant configuration as well as expected changes to plant 5 configuration that will result from proposed maintenance activities shall be conducted to determine the overall effect on performance of safety functions. The result of this assess-I 2

ment shall be used to ensure that the plant is not placed in risk-significant. Finally, by memorandum dated February 2,1998, the staff issued supplemental staffinspection guid-l ance (Reference 6) for licensee maintenance risk assessments that are performed under the

current maintenence rule. The inspection guidance is used to determine whether licensees

, have adequately assessed the overall affect on the performance of safety functions when J

SSCs are removed from service. The responsible office considers this item closed.

Action (c): Review the present regulatory requirements, such as standard technical specifications for j shutdown conditions, and revise as needed based on the results of Action (a) above. De- l l

velop guidance regarding revision of documents such as EOPs, accident management l l

J procedures, and plant technical specifications as necessary. (Responsible Office: NRR)

Status: Resolved (Pending AEOD Independent Review)

Although this action was documented as resolved in the 1993 AEOD Annual Report based on publication of NUREG-1449 (Reference 4), it was subsequently reopened and remained recategorized as ongoing by AEOD pending Commission final decision on a rule to manage i risk during shutdown operations.

As discussed in Item 1(a) above, in a Staff Requirements Memorandum for SECY-97-168 the staff was directed by the Commission to not proceed further with a proposed new rule for shutdown risk management. Alternatively, the Commission approved the staff's recom-mendation to develop rulemaking to revise 10 CFR 50.65(a)(3) of the Maintenance Rule so as to require licensees to perform safety assessments prior to performing maintenance activities. Additionally, the staff was directed by the Commission to consider adding as a preamble to the Maintenance Rule language that would state that the requirements of the Maintenance Rule are applicable to all plant conditions of operation, including normal shutdown operations. The responsible office considers this item closed.

Item 4: Adequacy of Emergency Preparedness .

Action (a): Evaluate and revise as necessary the guidance included in NUREG-0654 to classify events that could occur in cold shutdown and loss-of-electrical power events. Evaluate the NRC guidance to licensees on classification procedures and revise as appropriate. Evaluate the guidance to licensees for personnel accountability during outages. Revise and follow up as appropriate. Evaluate guidance to licensees regarding the availability of notification sys- -

tems (and alternates) during a loss-of-offsite power event. Consider the priorities and requirements for notifications to ofisite authorities. Follow up as appropriate. (Responsible Office: NRR)

Status: Resolved (Pending AEOD Indepndent Review) On September 16,1997, Nuclear Energy -

Institute submitted, for NRC rpproval, emergency action level guidance for classifying events. This guidance expuds upon the current guidance (NUREG-0654 and NUMARC/

NESP-007, Revision 2) fot classifying events. The staffis in the process of reviewing this NUREG-1272, Vol. I1, No. I G-2

Reactors guidance and expects to complete its review and to endorse the guidance,if appropriate, by the end of 1998. Staff action on this pan of this item is ongoing.

Accountability of personnel during outages was evaluated as pan of the NRC's evaluation of shutdown and low power operations reponed in NUREG-1449. As stated in Section 6.12.2 of NUREG-1449, licensees' emergency plans must address the evacuation and accountability of the large number of non-essential personnel on site should an accident occur during plant shutdown or refueling. Staff action on this pan of this item is complete.

The staff assessed the availability of notification systems and alternates during a loss-of-offsite power event as pan of its assessment of offsite communication systems performed for Item 2 (a) of the staff actions involving potential generic issues resulting f;om the NRC/

INPO team review of the effects of Hurricane Andrew on Turkey Point Units 3 and 4 (see Appendix I of this repon). Licensees are required (in accordance v ith Appendix E to 10 CFR Pan 50) to notify State and local govemment officials of an adrgency event within 15 minutes of the declaration of the event and to notify the NRC (in accordance with 10 CFR 50.72) immediately after notification of these officials. The priorities and requise-ments for notification were found to be adequate. Actions on this pan of this item is complete. The responsible office considers this item closed.

Summary: All of the staff actions resulting from the !!T repon on the loss of vital AC power and the heat removal system during mid-loop operations at Vogtle Unit I are closed according to the responsible offices. This is the last update on this subject pending completion of AEOD Independent Review of Action Items 1(a),1(c) and 4(a).

References:

1. " Loss of Vital AC Power and the Residual Heat Removal System During Mid-Loop Operations at Vogtle Unit 1 on March 20,1990," NUREG-1410, June 1990.

2. NUREG-1449," Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, Final Repon," September 1993.

3. SECY-97-168," Issuance for Public Comment of Proposed Rulemaking Package for Shutdown and Fuel Storage Pool Operation," July 30,1997.

4. " Shutdown and Low-Power Operations for Nuclear Power Reactors," Federal Register (59 FR 52707), October 19,1994.

5. " Staff Requirements: SECY-96-173 - Potential Revision to 10 CFR 50.65(a)(3) of the Maintenance Rule to Require Licensees to Perform Safety Assessments,"

December 17,1997.

6. Memorandum from R. Corria to R. Conte, et.al., " Draft Supplemental Maintenence Rule Inspection Guidance," February 2,1998.

G-3 Appendix G

1997 AEOD Annual Report AEOD IIT TRACKING SYSTEM Action Source: IIT Report on the Unauthorized Forced Entry into the Protected Area at Three Mile 1 Island Unit 1 on February 7,1993 (Reference 1). I Ite m 1: Adequacy of Regulations and Guidance for Protected Area Barriers, Entry Modes, and Design Basis Threat l

Action (b): Evaluate the need for guidance for respone to unauthorized forced entry into the protected area. Issue new guidance as appropriate. (Responsible Office: NRR/RES)

Status: Resolved A review of current guidance, procedures, and regulations, and interviews with licensee j were completed, with findings and conclusions documented by memorandum dated Octo-ber 28,1994 (see Reference 2). The staff determined that additional guidance regarding  ;

response to unauthorized / forced entry into the protected area could be beneficial to licens-ees. The staff has issued letters to the Security Managers of all nuclear power plants attaching a summary of insights and fmdings, both positive and negative, gained from the Operational Safeguards Response Evaluations (OSREs) that have been completed to date, and applicable summary information from two NUREGs that were issued by the NRC for tactical training and exercise planning at fuel cycle facilities. In response to questions raised during the AEOD independent review, NRR provided additional information regarding ,

resolution of the Action Item (see Reference 3), and described improvements to the OSRE program (see Reference 4). The responsible office considers this item resolved. AEOD ,

performed an independent review of this issue and considers it closed.

Summary: All of the staff actions resulting from the IIT report on the unauthorized forced entry into the protected area at Three Mile Island Unit I are resolved. This is the last update on this

, subject.

References:

1. NUREG-1485. " Unauthorized Forced Entry into the Protected Area at Three Mile Island Unit I on February 7,1993," dated April 1993.

2. Memorandum dated October 28,1994, from David N. Orrik to LeMoine Cunningham, j

" Staff Action Plan Resulting from the Investigation of the Unauthorized Forced Entry l into the Protected Area at Three Mile Island Unit I on February 7,1993 (NUREG-1485)."

3. Memorandum dated January 29,1997, from David B. Matthews to Frank J. Congel, "TMI IIT StafT Action Item Ib."

4. Memorandum dated January 16,1997, from Thomas T. Martin to Frank J. Miraglia,

" Review of Operational Safeguards Response (OSRE) Program."

NUREO 1272.Vol i1.No. I G-4

l 1

APPENDIX H )

Status of NRC Staff Actions Involving Potential Generic Issues Resulting From Diagnostic Evaluation Team Findings l

1 l

l l

l l

l I

l' Reactors CONTENTS Staff Actions Resulting From The Special Inspection of Cooper Nuclear Station. . H-1 l

l 1

l l

l l

l H-iii Appendix H

1997 AEOD Annual Report STATUS OF NRC STAFF ACTIONS INVOLVING POTENTIAL GENERIC l ISSUES RESiULTING FROM DIAGNOSTIC EVALUATION TEAM FINDINGS In accordance with Management Directive 8.7,"NRC Diagnostic Evaluation Program [DEP]," dated June 7, 1991, upon receipt of a Diagnostic Evaluation Team (DET) report, the EDO assigned NRC office responsibil-ity for generic and plant-specific staff actions resulting from the Diagnostic Evaluation (DE). Office Directors designated by the EDO as having responsibility for resolving issues or concerns are responsible for providing written status reports on the disposition of assigned actions. The AEOD Director will maintain the status of the staff actions involving generic issues and will repon them in the AEOD Annual Report.

I This appendix preser.ts the status of each of the open generic action items that the EDO has assigned to various NRC offices as a result of the completed des.

)

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1 NUREG-1272, Vol. I 1, No.1 H-iv

_m . _ _ _ . . . _ _ _ _ _

. _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ m _ _ _ _. __

Reactors '

I AEOD DET ACTION TRACKING SYSTEM Action Source: Memorandum from J. Taylor to Office Directors and Region IV Administrator,

" Staff Actions Resulting From The Special Inspection of Cooper Nuclear Station,"

dated December 22,1994 (Reference 1).

Item 6: Safety-related equipment testing did not always assure operability.

Significant weaknesses were recently identified in the licensee's testing and surveillance programs for safety-related systems and components. Deficiencies were found by the SET, regional inspectors, the licensee, and the DSA team. Identified weaknesses included pre-conditioning of equipment to assure passage of tests, and incomplete functional testing of safety-related system actuation logic. Additionally, surveillance procedures did not contain all required TS attributes, post-modification and post-maintenance testing was incomplete or not effectively planned, and preventive maintenance was ineffective in assuring equip-ment operability. Excessive testing resulted in plant challenges or degraded equipment while ineffective test result trending obscured declining equipment performance and the need for actions to correct problems before failure occurTed. The SET repon documents a number of testing weaknesses which substantially degraded the licensee's systcm operabil-ity assurance process. The SET results, together with previous DET findings for other facilities, indicate that licensee testing and surveillance programs vary significantly in their ability to detect or predict non-functionality or failures of systems and components. This situation appears to continue despite considerable operational experience feedback in the form of Information Notices, Bulletins, Generic Letters, and industry correspondence.

Action: Review the SET and previous DET repons to evaluate testing weaknesses in assuring operability. Identify any changes that could be made to improve the effectiveness of testing programs for assuring operational safety. (Responsible Office: AEOD) i Status: Resolved AEOD reviewed (Reference 2) 13 DET reports, the Cooper SET report, the Maine Yankee Independent Safety Assessment (ISA) repon, the Millstone Task Force Lessons Learned report, and the AEOD Engineering Evaluation Report," Review of Undetected Failures of Safety Systems." The DET, SET, and ISA reports identified numerous examples of testing program weaknesses. Licensee implementation and NRC inspection of inservice testing programs were identified weaknesses in 9 of 13 DET/ SET inspections. AEOD reviewed NRC generic documents for the past 3 years. Certain documents, such as Generic Letter (GL) 96-01," Testing of Safety-Related Logic Circuits," addressed testing issues. In reviewing the background of GL 96-01, it was noted that the NRC staff had previously issued five Information Notices regarding problems with testing safety-related circuits.

AEOD concluded that testing to confirm plant design should be part of design bases determination as addressed in the October 199610 CFR 50.54 (f) letters to licensees. NRR and the Regional reviews of these programs should determine that testing to assure oper-ability was considered in the program in response to the licensees' 10 CFR 50.54 (f) letters.

The inclusion of testing as part of design-based inspections would assist in determining the degree to which licensees demonstrate through testing that their plant meets appropriate H-1 Appendix H

1997 AEOD Annual Report design requirements. AEOD's review of undetected failures of safety systems identified that more than 75 percent of the undetected failures were discovered via testing or analysis and evaluation of operational problems. Testing was the most frequent discovery method.

AEOD's review indicated a need for the staff to reassess NRC programs for ensuring that licensee test programs meet regulatory requirements. A followup action was assigned (Reference 3) to NRR to assess the effectiveness of the NRC inspection program and other I

regulatory programs for evaluating licensee test programs and to make changes as appropri-ate. The results of the NRR assessment (Reference 4) indicated a need to revise the core inspection program to shift emphasis from programmatic reviews of engineering to systems based reviews, such as those conducted during Safety System Functional Inspections. NRR plans to incorporate the results of this assessment into the overall assessment of the inspec-tion program to be conducted by the Inspection Program Branch as Phase 2 of the Inte-grated Regulatory Assessment Process. This item is considered closed.

Summary: All of ibe staff actions resulting from the special inspection of Cooper Nuclear Station are resolved. This is the last update on this subject.

References:

1. Memorandum from J. Taylor to W. Russell, L. Callan, E. Jordan, R. Bernero, and P.

Bird. " Staff Actions Resulting From The Special Evaluation At Cooper Nuclear Sta-tion," dated December 22,1994.

2. Memorandum from Thomas T. Martin to L. Joseph Callan, " Status of Cooper Nuclear Station Special Evaluation Issue 6 Staff Action Item," dated January 30,1998,

3. Memorandum from L. Joseph Callan to Samuel J. Collins," Staff Action Resulting From AEOD Review of Issue 6 From The Special Evaluation At The Cooper Nuclear Sta-NUREG-1272 Vol. I1 No.1 H-2

APPENDIX I Status of NRC Staff Actions Involving Potential Generic Issues Resulting from the NRC/INPO Team Review of the Effects of Hurricane Andrew on Tbrkey Point Units 3 and 4

Reactors AEOD HURRICANE ANDREW ACTION TRACKING SYSTEM Action Source: Memorandum from J. Taylor to Office 91 rectors and Regional Administrators,"Re-port on the Effect of Ilurricane Andrew on the 'Ibrkey Point Nuclear Generating Station from August 20 30,1992," dated May 28,1993 (Reference 1).

Item 2: Adequacy of Licensee Offsite Communications for Natural Disasters Within the Plant Design Basis Although diverse and redundant communications equipment existed, offsite communica-tions were lost during the storm due to a common vulnerability to wind damage. Normal telephone service failed because the storm blew down the lines near the station. The dedi-cated commercial telephone lines servicing the telephones installed in the control room, the Technical Support Center, and the Emergency Operations Facility, used to give initial noti 6 cation and status to the State in an emergency, also failed. The Federal Telecommuni-cations System - 2000 lines used for the Emergency Noti 6 cation System failed, cutting off normal communications with the NRC Operations Center. The cellular telephone systems also did not function because the storm damaged the on-site antennas and the offsite repeating stations. Except for the Security Department's one hand held radio for the com-pany FM radio system, the licensee's radio systems did not function during and immedi-ately following the storm. Overall, all offsite communications were lost for about four hours during the storm, and reliable communications were not restored for about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the storm. The NRC's temporary satellite communications system considerably aided recovery efforts and would have been more beneficial if it had been onsite before the storm.

Action (a): Review the existing regulatory guidance and requirements related to normal and backup offsite communications system design capabilities for hurricanes. Based on this review, consider the adequacy of the guidance for other external events. Issue revised guidance or requirements as may be needed. (Responsible Of6ce: NRR/AEOD)

Status: Resolved NRR has reviewed current regulations and regulator' guidance and identified rules and guidance that apply to offsite communications syst< ms. In coordination with the Federal Emergency Management Agency and the NRC te;hnical branches responsible for require-ments on licensee communications, NRR reviewej the identified rules and guidance to determine whether they adequately account for e ternal events.

NRR staff concluded that the requirements and guidance are suf6ciently detailed to provide licensees with the staff's expectations of the capability of offsite communications to function during and following severe natural events. Notwithstanding this conclusion, there was insufficient information on the existing offsite communications capabilities at nuclear power plant sites to conclude that the problems identi6ed with the loss-of-offsite communi-cations at Turkey Point were not pervasive in the industry. Therefore, to make a determina-tion whether generic action was warranted to ensure compliance with the regulations, information was obtained on the offsite communication systems at a sampling of sites.

Information was gathered during routine regional inspections scheduled between February I-1 Appendix I

1997 AEOD Annual R.eport and June 1996 using Temporary Instruction (TI) 2515/131 " Licensee Offsite Communica-tion Capabilities," issued on January 18,1996. NRR evaluated the inspection findings to determine whether guidance to the licensees in the form of a generic communication was necessary to ensure either survivability or rapid recoverability of these circuits from a severe natural event. On the basis of its review of the results of TI , NRR concluded that additional guidance to the industry is warranted. Consequently, on Febmary 27,1997, NRC issued Information Notice 97-05: "Offsite Notification Capabilities," to inform licensee of the results of the TI. Additionally, on December 31,1996, NRC Inspection Procedure 82701, " Operational Status of the Emergency Preparedness Program," was revised to provide guidance for the review of licensees' offsite communication circuits as part of the core inspection program.

Summary: All of the staff actions resulting from the repon on the effect of Hurricane Andrew on the Turkey Point Nuclear Generating Station are resolved. This is the last update on this subject.

References:

1. Memorandum from J. Taylor to Office Directors and Regional Administrators,"Repon on the Effect of Hurricane Andrew on the Turkey Point Nuclear Generating Station from August 20-30,1992," dated May 28,1993.

NUREG-1272, Vol. I1, No. I 1-2

APPENDIX J Status of NRC Staff Actions Resulting from the Independent Safety Assessment of the Maine Yankee Atomic Power Station

Reactors AEOD MYAPS SAFETY ASSESSMENT TRACKING SYSTEM Action Source: Memorandum from J. Taylor to Office Directors and the Region I Administrator,

" Staff Actions Resulting From The Independent Safety Assessment of the Maine Yankee Atomic Power Station," dated November 27,1996 (Reference 1).

Item 1: Adequacy of Analytic Code Validation The ISA team noted that the plant specific validation of RETRAN by Maine Yankee to known industry benchmarks for integral and separate effects test data was deficient. This validation is important to assure that the plant-specific application of the code effectively models known physical effects. The team found the NRC requirement for this validation to be vague. The single document which states NRC policy on this issue is Generic Letter 83-11," Licensee Qualifications for Performing Safety Analysis in support of Licensing Actions," issued on Febmary 8,1993, which states, in part:

. some licer.w. planning to perform their own safety analyses may not intend to demonstrate their ability to use the code by performing their own code verification.

Rather, they plan to rely on the code verification work previously performed by the code developer or others.

NRR does not consider this acceptable and each licensee or vendor who intends to use a safety analysis computer code to support licensing actions should demonstrate their proficiency in using the code by submitting code verification performed by them, not others.

Additionally, the team found that the NRC has acted inconsistently relative to its expecta-tions in this area. In some cases, computer codes have been endorsed for use with little or no validation accomplished.

Action: Evaluate the agency's expectations and policy relative to code validation. Develop and issue additional guidance and requirements if appropriate, and develop and implement inspection methodology to verify licensee conformance as appropriate. (Generic: NRR/RES)

Status: Ongoing The staff has reviewed existing documents to evaluate the adequacy of staff requirements and guidance for code validation in light of the Maine Yankee IS AT findings and found that no guidance to exist on code validation. The results of this evaluation will be documented in February 1998. The staff plans to initiate efforts to draft staff requirements and guidance for code validation. A joint NRR-RES study group will meet by the summer of 1998 to define the scope, effort, schedule and resources required to complete this task (Reference 2).

J-l Appendix J

1997 AEOD Annual Report Item 2: Adequacy of NRC Review of Analysis Codes The ISA formed an expert panel of consultants with extensive experience in the area of analysis code development to assess and critique the results of the ISA's efforts. The reports submitted by these consultants (attached) included observations and suggestions for im-proving the NRC's process of reviewing analytic codes.

Action: Review the attached consultant repons and evaluate the need to make changes to the existing NRC processes as suggested in the reports. Implement changes as appropriate.

(Generic: NRR/RES)

Status: Ongoing The staff has conducted a review of existing staff processes for reviewing analytical codes and found that there is no documented staff process for reviewing analytical codes. The staff typically perfomis a review which is similar to a previous review in lieu of following a specified template. The results of the review will be documented in February 1998. The staff has not yet undertaken efforts to revise the process for reviewing analytical codes.

This issue will be resolved concurrent with Action iteml above (Reference 2).

Item 3: Compliance With Safety Evaluation Reports During the Maine Yankee ISA, compliance with safety evaluation repon (SER) conditions imposed on the use of analytic codes was verified for 66 conditions effecting 13 codes.

While compliance was confirmed, an audit trail to assure compliance was not always available, necessitating, in come cases, additional analyses to verify compliance. The team found that the Regulatory status of an SER condition was unclear.

Additionally, the ISA team found that the quality of NRC code reviews was mixed. This may have stemmed from the fact that there was no standard review plan for code reviews.

Consequently, no guidance or requirements existed for: development of an agreed upon set of identified and ranked phenomena, processes, or key parameters; validation; code model-ing detail; sensitivity studies; or peer review by experts in the field.

Action (a): Evaluate the agency's expectations regarding the tracking and closeout of SER conditions relative to compliance, auditability, and reportability. Issue appropriate industry and inspec-tion guidance as needed. (Generic: NRR)

Status: Ongoing The staff has not yet undertaken effons to develop a standard format and content guide for topical reports which document ECCS methodologies. However. the staff has modified the Core Performance Action Plan to place additional emphasis on verification of conformance to SER conditions. Results of this activity were implemented in an inspection at Siemens Power Corporation (Reference 2).

Action (b): Evaluate the need to develop a standard review plan for code reviews. Develop and issue appropriate guidance. (Generic: NRR)

_ ~ _ _ - .-

Reactors Status: Ongoing Because of resource constraints, the staff has not yet undertaken effons to develop a standard review plan to accompany the standard format and content guide (Reference 2).

Item 4: Adequacy of Licensing Reviews for Power Uprates The IS A team identified a number of mechanical components for which confirmation of operability at the upgraded power level of 2700 MWt could not be confirmed. Additionally,  ;

the team noted that documentation of NRC actions on parameters related to the design and licensing bases for Maine Yankee was not identifiable and retrievable from NRC sources.

Action (a): Evaluate the Agency's technical review and approval of the power uprates for Maine Yankee in light of information developed by the IS A team. Take appropriate actions to confirm or rescind approval of the Maine Yankee power uprates approved in 1978 and 1988, 1

Status: Complete This action was predicated on the Maine Yankee plant continuing to operate. On August 7, 1997, the licensee certified permanent cessation of power operation for the Maine Yankee facility. With the cessation of operation, the staff has determined not to pursue this action (Reference 2).

Action (b): Evaluate the Agency's process for technical review and approval of licensee requested power uprates. Implement changes to the process as appropriate. Based on the results of this review, determine whether any previously approved power uprates should be reevalu-ated and to what extent. (Generic: NRR)

Status: Ongoing The NRR Project Manager's (PM) Handbook has been revised to provide guidance on reviewing UFSAR updates to ensure that changes to the UFSAR have been properly entered and have received appropriate NRC approval. The staff will further revise the PM Handbook by February 1998 to specifically address the review of power uprates. The staff has not yet undertaken efforts to develop a standard power uprate review procedure. A j revised target due date will be determined (Reference 2).

The staff is currently performing an initial screening analysis to identify any significant (

power uprate issues needing prompt attention. A report documenting the results of the initial screening will be issued in March 1998. The issuance of this repon will complete the staff's effort on this action.

Action (c): Evaluate the need and the feasibility of establishing an NRC licensing and design bases database for all plants to centrally collect all documentation necessary to support plant licensing. Take actions as appropriate. (Generic: NRR)

Status: Complete Based on the insights gained in reviewing licensees' responses to the October 9,19%

request for information pursuant to 10 CFR 50.54(f) regarding the adequacy and availabil-ity of design bases information, the staff has concluded that it is not necessary or feasible to J-3 Appendix J

\s

1997 AEOD Annual Report i

establish a design bases database to centrally collect all documentation necessary to support plant licensing (Reference 2).

Item 5: Clarity and Intent of NRC Safety Guide i During the ISA review of containment spray system and high pressure safety injection system net positive suction head (NPSH), the team found the guidance provided by NRC Safety Guide 1, " Net Positive Suction head for Emergency Core Cooling and Containment Heat Removal System Pumps," issued on November 2,1970, to be problematical with regard to relying on containment pressure for assuring NPSH for emergency core cooling and containment heat removal pumps.

NRC Safety Guide I states,"NPSH for emergency core cooling and containment heat removal system pumps caused by increases in temperature of the pumped fluid under loss of coolant accident conditions can be accommodated without reliance on the calculated increase in containment pressure." Furthermore it states: " Emergency core cooling and containment heat removal systems should be designed so that adequate net positive suction head is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated loss of coolant accidents."

Maine Yankee assened that they were not committed to Safety Guide I. Consequently, they assumed containment to be at or above the saturation pressure for the sump fluid temp rather than at pre-accident containment pressure (nominally atmospheric). The issue of whether or not the containment can be assumed to be pressurized at the saturation pressure i

for the sump fluid temperature should be addressed.

Action: Review and clarify the staff's criteria relative to relying on containment overpressure for ensuring appropriate NPSH for emergency core cooling and containment heat removal pumps. The staff is already conducting a sepamte program to determine if and how all plants, including Maine Yankee, meet these criteria. Upon review of this information, the staff will determine the measures to be taken for those plants not in compliance with the criteria. (Generic: RES/NRR)

Status: Ongoing A draft generic letter, which requests information from licensees concerning available NPSH for emergency core cooling system and containment heat removal pumps, was recently issued by RES, and comments are due by late March 1998.The statTplans to review the responses to that generic letter and then develop a revised Regulatory Guide, if needed. Although the changes likely needed to the guide should be relatively straightfor-ward, the staff wishes to assess the magnitude of the problem, as identified by licensees

• responses, before making the change. A proposed schedule of activities for resolution of this issue will be developed and coordinated with NRR following review of the responses to the generic letter (Reference 3).

Item 6: Adequacy of the NRC Inspection Program The ISA team observed that the licensee and NRC staff failed to recognize and/or appropri-ately evaluate degraded / nonconforming conditions. A number of potential operability issues existed; however, neither the licensee nor the staff aggressively pursued resolution of the NUREG-1272, Vol. I 1, No.1 J-4

Reactors issues. Funher, although staff inspections and oversight reviews of Maine Yankee con-ducted prior to the ISA identified significant performance issues, they did not fully convey the broad performance problems and weaknesses identified by the ISA team. These issues included problems with safety system testing programs, licensee-developed technical specification interpretations, and design basis adequacy.

Action (a): Evaluate the inspection program and inspector training and guidance with regard to testing programs for safety systems relative to its scope, rigor, and analysis of results. Implement inspection program changes and develop new guidance as appropriate. (Generic: NRR)

Status: Ongoing The staffis reviewing safety system inspection procedures and inspection guidance for scope, rigor, and analysis of results with respect to Maine Yankee ISAT findings. The staff is sending selected core inspection procedures to responsible technical divisions for review and revision as appropriate. The staff is sending selected regional initiative procedures to responsible technical divisions for review and revision as appropriate (Reference 2).

Action (b): Evaluate the inspection program and guidance with regard to review oflicensee developed technical specification interpretations to assure consistency with the intent of the approved technical specifications. Implement inspectiori program changes and develop new guidance as appropriate. (Generic: NRR)

Status: Complete New inspection guidance on technical specification interpretations is contained in Pan 9900 of the Inspection Manual. The revision was issued in February 1997. In addition, new inspector guidance on technical specification interpretations was communicated to the inspection staff by a memorandum dated 8/23/97 and by an Information Notice dated 11/

21/97. In light of the actions taken above, the staff determined that additional guidance did not need to be included in the Fundamentals of Inspection Refresher Course (Reference 2).

Action (c): Evaluate the inspection program and guidance with regard to the assessment of the ad-equacy of plant design basis, including a review of the disposition of significant findings from previous licensee effons such as design basis documentation or design basis reconsti-tution programs. (Generic: NRR)

Status: Ongoing The staffissued a change to the Inspection Manual in June 1997 to allow the use of inspec-tion procedure (IP) 93809, " Safety System Engineering Inspection," or IP 93801, " Safety System Functional Inspection," to substitute for the programmatic review described in IP 37550, Engineering. In addition, founeen design mspections have been completed through December 1997. Eight additional inspections are planned through program completion in October 1998 (Reference 2).

Item 7: Adequacy of Agency Expectations Regarding Licensee Performance The ISA relied on the existing agency benchmark for assessing performance utilized in the NRC Systematic Assessment of Licensee Performance Program (SALP). Although SALP category rating definitions, functional areas, and assessment criteria have evolved over J-5 Appendix J

1997 AEOD Annual Report time, the Commission raised questions about the SALP definitions. In addition, a number of questions were raised during the October 10,1996, public meeting on the ISA findings at Wiscasset Maine.

Action: Evaluate the appropriatertess of the existing SALP definitions of superior, good, and acceptable performance in light of the NRC's contemporary expectations for licensee performance. Revise thee definitions as necessary. (Generic: NRR)

Status: Ongoing The staff obtained Commission approval to conduct an Integrated Review of Assessment (IRA) to evaluate all licensee assessment processes and develop a revised process. The IRA schedule calls for providing preliminary results to the Commission, a public comment period, and final Commission decision on implementation of a new assessment process.

Final Commission decision is scheduled for July 30,1998 (Reference 2).

Item 8: Cumulative Effect of Operator Workarounds The ISA found that operators at Maine Yankee were required to take numerous actions to compensate for weaknesses in plant design. Some of these would require operators to take time consuming manual actions such as donning steam suits and deploying a 350 foot extension cord during significant plant transients. Additionally, the team found that Maine Yankee had been slow to resolve a work allocation issue which appeared to direct the two on-shift senior operators to leave the control room in the event of a fire coincident with a medical emergency. The cumulative effect of all these actions had not been evaluated by the licensee or the NRC.

Current NRC policy which would restrict credit for operator action or define the time which l may be available for operators to take action is limited. The staff has typically relied on l

guidance provided in ANSI /ANS $8.8, " Time Response Design Criteria for Safety Related Operator Actions." However, the staff has allowed deviations from this guidance when licensee's have provided empirical evidence that operators can take the required actions within the required time constraints.

Action (a): Evaluate the current guidance and po!icies with regard to the cumulative effect of operator workarounds. Develop and issue additional or revised guidance as appropriate. (Generic:

NRR/RES)

Status: Ongoing The staff has reviewed current guidance and policies and has developed a draft Temporary Instruction (TI) to gather the information necessary to complete the evaluation. The staff anticipates soliciting comments from the Regions conceming the draft TI in January 1998.

Completion of the operator workaround TI inspections is anticipated by the end of fiscal year 1998. Additional guidance will be issued based on the results of the evaluation (Refer-ence 2).

Action (b): Evaluate the need to develop inspection policy and guidance directed at assessing the cumulative effect of operator workarounds. Develop and issue guidance as appropriate.

(Generic: NRR)

NUREG-1272 Vol. I1 No. I J-6

Reactors Status: Complete j The staffissued : vised inspection guidance (IMC 71707) to highlight consideration of the

cumulative effect of operator workarounds in September 1997 (Reference 2).

l l Item 9: Agency Policy Regarding Licensee Design Basis Recovery Efforts The ISA team found that the licensee had identified significant design bases issues involv-l ing safety-related systems as part of their Design Basis Reconstitution (DBR) program.

DBR reviews had been performed in ten functional areas. Another nine functional areas had been scheduled, but had been delayed due to resource limitations and priority changes.

l These areas included the emergency diesel generator, electrical distribution, and ventilation.

The licensee and the ISA team found design weaknesses in each of these areas.

Action: Evaluate the current Agency policy regarding licensee design basis recovery efforts. Con-sider the need to require or encourage licensees to accelerate and complete efforts to recover and reconstitute their design basis, especially older facilities where some informa-tion may be missing, difficult to find, or inaccurate. (Generic: NRR)

Status: Complete )

1 This issue was addressed by the staff's review of the information provided by licensees in response to the October 9,1996 request for informatien pursuant to 10 CFR 50.54(f) regarding the adequacy and availability of design bases information (Reference 2). ,

Item 10: Public Involvement in the Assessment Process l The planning and conduct of ti te ISA included extensive State participation through three team members, two process rr tiewers, a five member citizen's group and periodic briefings with the Govemor, a public of servation entrance meeting and a public panicipation meet-ing to convey the findings. He wever, the team received complaints during the public meeting and via written corrr spondence that there was insufficient opportunity for "public participation."

Action (a): Evaluate the need to provide guidance for public participation (via a two part meeting) at the beginning of a review to explain and discuss the scope and objective of the review.

Develop and issue guidance as appropriate. (Generic: AEOD/OPA/NRR)

Status: Complete This issue was addressed by the staff's review of the information provided by licensees in response to the October 9,1996 request for information pursuant to 10 CFR 50.54(f) regarding the adequacy and availability of design bases information (Reference 2).

Action (b): Evaluate the need to allow at least one week from issuance of a report to a "public partici-patior, meeting" on the findings. Revise or issue new guidance as appropriate. (Generic:

AEOD/NRR) i I

Status: Complete This issue was addressed by the staff's review of the information provided by licensees in response to the October 9,1996 request for information pursuant to 10 CFR 50.54(f) regarding the adequacy and availability of design bases information (Reference 2).

J-7 Appendix J

1997 AEOD Annual Report I Action (c): Evaluate the need to make additional copies of the entire report available by sending multiple copies to the local PDR in addition to Internet access. Revise or issue new guid-ance as appropriate. (Generic: AEOD/NRR)

Status: Complete This issue was addressed by the staff's review of the information provided by licensees in response to the October 9,1996 request for information pursuant to 10 CFR 50.54(f) regarding the adequacy and availability of design bases information (Reference 2).

i

References:

1. Memorandum from J. Taylor to Office Directors and Region 1 Administrator," Staff Actions Resulting From The Independent Safety Assessment of the Maine Yankee l Atomic Power Station," dated November 27,1996. l 1

2. Memorandum from Samuel J. Collins to L. Joseph Callan, " Status Of NRR Staff  ;

Actions Resulting From The Independent Safety Assessment of Maine Yankee Atomic l Power Company," January 27,1998.  ;

l

3. Memorandum from David Morrison to L. Joseph Callan,"RES Response to Issue 5 of l Staff Actions Resulting From the Indepenent Safety assessment at Maine Yankee,"  ;

February 27,1997. I i

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2. tule mo sUenTLE Vol.11, No.1 8 oATE REPOHT PUDUSMED Office for Analysis and Evaluation of Operational Data 1997 Annual Report " "

• l YEAR November 1998 4 FIN OR GRANT NUMBER

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11 ABSTRACT 900 word.or i.e.) l This annual report of the U.S. Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational  !

Data (AEOD) describes activities conducted during 1997. The report is published in three parts. NUREG-1272, Vol.11, No.1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about trends of some key performance measures. The re-port also includes the principal fm' dings and issues identified in AEOD studies over the past year and summarizes mformation from such sources as licensee event reports and reports to the NRC's Operations Center.

NUREG-1272, Vol.11, No. 2, covers nuclear materials and presents a review of the events and concerns during 1997 associated with the use of licensed material in ronreactor applications, such as personnel overexposures and medical misadministrations. Both reports also contain a discussion of the Incident Investigation Team program and summarizes both the Incident Investigation Team and Augmented Inspection Team reports. Each volume contains a list of the AEOD reports issued from CY 1980 through 1997. NUREG-1272, Vol.11, No. 3, covers technical training and presents the activities of the Technical Training Centerin support of the NRC's mission in 1997.

l 12 KEY WORDS/DESCRIPTORS (LJet word. or phr.e that will .a.st r e-th.rs ,rilocang th. f. port ) 13 AVAILABiUTY STATEMENT Unlimited nuclear plants safety perforamance 14 SECURITYCLASSIFICATON l operating experience AEOD recommendations m'**8" abnormal occurrences AEOD studies Unclassified significant events Incident Investigation Program (rni, arpori>

reactor trips Diagnostic Evaluatioin Program Unclassified performance indicators Corrmittee to Review Generic Requirements 16 huMBER OF PAGES NRCstaff actions system reliability operating factors accident sequence precursors incident response ""

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