ML20215M915

Us NRC Participation in NEA Incident Reporting Sys During 1983

ML20215M915
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Issue date:	06/30/1984
From:	NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
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ML20215M911	List: ML20215M910 ML20215M915
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TASK-AE, TASK-S403 AEOD-S403, NUDOCS 8611030460
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AE00/S403 Date: June 1984 AE0D Special Report USNRC Participation In The Nuclear Energy Agency Incident Reporting System During 1983 Prepared By:

Data Management Section Program Technology Branch Office for Analysis & Evaluation of Operational Data United States Nuclear Regulatory Commission l

0611030460 ORG 840703 NEXD PDR PR

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Major Contributors John L. Crooks (Lead)

Peggy Cross-Prather James T. Fry Sheryl A. Massaro i

Abstract Since 1980, the Nuclear Regulatory Commission has participated in a central-ized system for internationally exchanging information on nuclear power plant operational experience. The system, sponsored by the international Committee on the Safety of Nuclear Installations (CSNI) and run by its Nuclear Energy Agency (NEA) is called the Incident Reporting System (IRS). In CY83, the NRC submitted 59 IRS reports - 54 individual event reports, four generic reports which included 18 individual event reports, and one report specific to a vendor's component. The rate of reporting was about 0.90 reports per reactor year. The U.S. supplied approximately 46 percent of the IRS reports submitted in CY83 and has about 35 percent of the operating nuclear power plants reporting to the IRS.

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BACKGROUND The NEA-IRS Reporting System In mid-1979, the United States Nuclear Regulatory Commission proposed a centralized system for exchanging information on operational experience in thermal power reactors to the international Committee on the Safety of Nuclear Installations (CSNI), the governing body of the Nuclear Energy Agency (NEA).

The NRC proposal was subsequent to the Three Mile Island accident and was an initiative to further develop the information exchange that was on-going between individual countries. CSNI recognized that the timely exchange of information on operating experience in nuclear power plants is important for improving their safe operation. Such information helps avoid incidents in one country recurring elsewhere, facilitates the analysis of general safety issues, assists in developing larger data banks and not only contributes to the better regulation on nuclear power plants but also provides additional guidance for safety research programs. With this in mind, in early 1980, CSNI established the Incident Reporting System (IRS). Thirteen countries (Belgium, Canada, Finland, France, Federal Republic of Germany, Italy, Japan, Netherlands, Spain, Sweden, Switzerland, United Kingdom and the United States) participated. Six other countries plus the Commission of the European Communities became observers.

The system was established with provision for confidentiality if requested by the data suppliers. The system was to be operated on a trial basis for two years from January 1980 to January 1982. In November 1981, CSNI reviewed the IRS and decided that the exchange of information had such importance that it should be improved and continued. CSNI also adopted revised guidelines for which incidents to report ( Appendix A) and a standard report format ( Appendix B).

Initially, the establishment and operation of the system was placed under

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. a Coordinating Group comprised of designated representatives from each of the participating countries. In November 1982, a reorganization of CSHI activities took place, and the IRS was placed under Principal Working Group Number One, "Operatina Experience and Human Factors."

The Working Group meets annually to dia, cuss significant operating experience and to provide program direction including identifying areas for inprovements.

The IRS reports are currently being added to a computerized data system at the Commission of European Communities Ispra Joint Research Center. In September 1983, the NEA also began exchanging incident infomation with the International Atomic Eneroy Agency (IAEA) through joint meetings on operating experience.

The exchange of formal event reports between these agencies was also initiated on a limited basis in mid-1984.

The NRC's Office for Analysis and Evaluation of Operational Data (AE00) and Office of International Programs (IP) are the principal U. S. participants in the IRS. IP is the official U.S. IRS Coordinator, while AE00 is the principal technical representative and is responsible for reporting U.S.

events to the IRS.

As of May 1984, the NEA has received and, in turn, distributed about 400 IRS reports to the participants in the program.

U. S. Reporting to IRS AE0D screens each operational event report fron U. S. nuclear power plants to determine whether the event should be reported to NEA. This decision is based on the IRS criteria for reportability and a judgement of the event's significance or potential significance. The IRS reporting guidelines are given in Appendix A. Once an event is identified for reporting, AE00 prepares the proposed report and forwards it directly to the NEA-IRS Coordinator in

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Paris who then forwards it to the other IRS participants. Additionally, NEA may, on its own initiative, occasionally select an item from publicly available infomation and prepare and distribute an IRS report.

Most U. S. reports describe an individual significant event and are submitted to NEA several months after the occurrence. IRS reports are also submitted on generic concerns. These consist of a summary IRS report identifying the potentially significant generic cencern, and several IRS reports on events which, although not individually significant, are pertinent and related to the ceneric concern. Thus, these individual events usually numbering from two to six reports, become collectively significant in the context of the generic Concern.

IRS reports generally contain the following infomation:

1. Plant name

2. Unit number

3. Licensee

4. Date of occurrence

5. Type of reactor and manufacturer

6. Authorized electrical power output

7. Systems or components affected

8. Initial plant condition

9. Way in which incident was detected

10. Radiation exposure or radioactivity release

11. Event description

12. Cause of event

13. Lessons learned or safety significance

14. Actions taken or planned

15. IRS guideline (s) under which event is being reported

16. Identification of relevant items (e.g., QA problem, loss of offsite power, interaction between systems) from a 42 item watchlist.

Availability and Analysis of IRS Reports Within the U.S.

NEA/lRS sends the IRS reports directly to IP where additional distribution is made (1) within the NRC to the Offices of Nuclear Reactor Regulation, Nuclear l Regulatory Research, Inspection and Enforcement and AE00; (2) to the Advisory l

4-Committee for Reactor Safeguards; (3) to the Institute of Nuclear Power Operations (INP0) and (4) to the Nuclear Operations Analysis Center (NOAC) at Oak Ridge National Laboratory.

The various recipients review the infonnation for its appifcability to the U.S.

nuclear power program in their particular areas of responsibility. For example, the following specific data review and data processing activities are conducted:

e AE00 reviews each IRS report received for its applicability to U.S. plants and its apparent significance. Events of interest receive more detailed technical review and engineering evaluation, and may lead to actions being taken by other NRC offices.

e NOAC, under contract to AE00, processes the IRS reports (for power plants greater than 200 MWe and similar to U.S. designs) received from foreign countries into a computerized data base along w!th other foreign incident data. This data file, known as the Foreign Events File (FEF), is searchable and data can be retrieved in various formats.

The FEF is available on the Department of Energy's (D0E) RECON network with controlled access. The IRS reports have recently been made available via the FEF to several DOE offices, INP0 and at least one national laboratory.

Copies of the U.S. supplied IRS reports are available to the public and can be obtained from the NRC's Public Document Room,1717 H St. NW, Washington D.C.

20555. Most of the foreign IRS reports, however, are foreign proprietary and thus are only made available to authorized individuals. Once authorized, copies of the original foreign IRS reports and tranlations can be obtained from the NRC's Office of International Programs, Washington D.C. 20555.

Benefit of Foreign Experience Over the years, the operational experience information obtained from foreign reactors has benefited the U.S. program. This information has had value as an early warning of potential problems and as input to related studies of U.S.

operating experience. Information on serious events occuring in foreign light-water reactors is generally provided by foreign governments, plant operators or a U. S. reactor vendor in a reasonably prompt manner. A few examples of significant foreign experience that served as early warning indications for U.S. plants are:

(1) the detection in 1975 of excessive wear due to vibrating incore instrument tubes causing damage to fuel channel boxes in a foreign BWR-4. This experience led to extensive inspections and corrective measures for similar plants in the U. S. ,

(2) the detection of cracking in large diameter (>20 inches) piping in the j reactor coolant and associated safety systems. As a result of this foreign experience, a broad program of piping inspections was conducted in U. S. plants which provided early detection of cracking in similar piping systems.

(3) The detection of vibration problems in a new steam generator design in an overseas plant provided the bases and incentive for a U.S. regulatory position regarding the corrective measures to be taken for similar steam generators in U. S. plants.

(4) foreign experience with the failure of control rod guide tube pins resulted in the inspection and early detection of similar failed pins in U.S. plants.

. (5) foreign experience with the perfomance of a safety valve during a transient (i.e., rapid opening and closing) that resulted in a pipe break provided the necessary infomation to develop and support a U.S.

regulatory position requiring corrective action.

In this regard, IRS reports have proven to be a significant source of oper-ational experience, principally because they have been prescreened from larger data sources. For example, two recent IRS reports that resulted in AE0D

,l l evaluation and subsequent actions are:

(1) an IRS report described several events in which pilot actuated PWR main steam safety valves exhibited opening pressures considerably higher than allowable. The cause of the problem was reported to be a bonding together of the pilot valve seat and disk at their point of contact due to corrosion i

products " growing together" (contact corrosion). This infomation was evaluated by AE0D and the results provided to a BWR Owner's Group investi-gating the causes and needed corrective actions for high set point drift being exhibited by Target Rock 2-stage pilot actuated safety relief valves (SRV). With this infomation, the owner's group subsequently detemined that corrosion induced bonding in the pilot valve disk-to-seat

, interface area was the cause of the domestic BWR SRV setpoint drift.

l (2) an IRS report described several events in which two flexible disk wedge-type gate valves installed in the RHR suction path failed to open. The entrapment of high pressure fluid in the valve bonnets caused the failures.

An AE00 study of this event prompted a review of similar reports of i don.estic valve failures. This AE00 study concluded that gate valve failures caused by entrapment of high pressure fluid in the valve bonnets i

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. could be a significant source of common mode valve failure. An NRC notice is being prepared to inform domestic nuclear plant operators of this concern.

The review, assessment and feedback to U.S. plants of foreign operating data, including the IRS reports, continues to involve some difficulties. These include:

e limited reporting of events by some countries e lack of technical depth and/or detail in some reports e translation costs and delays e restrictions on public dissemination e differences in plant design.

These itens make it difficult at tines to fully assess the sionificance of the infornation and applicability of the foreign experience to U.S. plants.

In terns of costs, obtaining and processing foreign operational experience has not resulted in major costs to the NRC. For example, the resource expen-ditures in manpower, travel, automated data processing, and contractor support associated with the IRS are estimated at less than $150K.

USNRC PARTICIPATION IN THE NEA-IRS DURING 1983 In calendar year 1983, the United States submitted a total of 59 reports covering 73 individual operational events. Included in this total were:

e 54 individual event reports (51 covered new significant events and 3 added significant new information to previously reported events) e four generic reports which included 18 individual event reports on events related to the generic topics. For example, five individual IRS reports were submitted with the IRS report on the generic topic of loss of decay

8 heat renoval capability. The generic reports were concerns that warranted reporting because of their frequency, i.e., combined significance, rather than individual significance. The individual event reports cited for the generic concern are intended to be illustrative and not a comprehensive listing of all events related to the particular concern being reported.

e one IRS report concerning the operation of Westinghouse DS-416 reactor trip breakers which NEA initiated from an NRC information notice.

Appendix C shows the distribution of individual event reports by licensed operating units. Of the 79 units units licensed to operate in 1983, 41 (52%)

were represented in the IRS reports. A iist of units licensed to operate is found in Appendix D.

Although most events during 1983 involved multiple causes, the dominant causes of the 54 individual events were as follows: 52% were attributed to equipment failure; 7% to personnel error; 20% to procedural / administrative problems; 17%

to design error and the remainder to other causes such as explosions, fires and naturally occurring events (floods, snow storms, etc.). Appendix E contains a list of the dominant cause identified for each event reported in an IRS Report.

In sunnary, the United States submitted a' ut 0.9 (73 reports /78 reactor years) event reports per reactor year in CY 1983. Although somewhat below the objec-tive of one report. per reactor year, the United States reported 46 percent of the 127 IRS reports submitted in CY 83. The U.S. has approximately 35 percent of the operating nuclear power plants reporting to the IRS. Appendix F contains a list of the titler of the IRS reports submitted.

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APPENDIX A Reporting Criteria for the NEA Incident Reporting System (IRS)

1. Significant Release of, or Exposure to, Radioactive Material Some examples may be:

An event that results in release of radioactive materials to the environment exceeding authorized limits; An event that results in a member of the public receiving a radiation dose exceeding authorized limits; An event that results in facility personnel receiving a radiation dose exceeding authorized limits.

2. Significant Degradation of Safety-related Systems 2.1 Fuel cladding failure.

Some examples may be:

Fuel cladding failures requiring plant shutdown; Cladding failures in spent fuel in the storage pond.

2.2 Degradation of the primary coolant pressure boundary, main steam line or feedwater line.

Some examples may be:

Through-wall failures of the piping or the significant components of the primary coolant circuit; Welding defects or material defects in the primary coolant circuit; Rap:d tenperature or pressure transient exceeding the authorized limits; Loss of relief and/or safety valve functions during tests or operation.

2.3 Loss of containment function or integrity.

Some examples may be:

Contaimnent leakage rates exceeding the authorized limits; Loss of containment isolation valve functions during tests or operation; Loss of main steam isolation valve functions during tests or operation;

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- Loss of containment cooling capability.

2.4 Degradation of systems required to control criticality.

Some examples may be:

- Failures of the control rod systems;

- Accidental criticality;

- Failures of the boron injection system.

2.5 Degradation of systems required to control the system pressure or temperature.

Some examples may be:

Failures of the emergency core cooling systems such as the high/ low pressure core injection system and the core spray system;

- Loss of core cooling ability including failures of the residual heat removal system;

- Loss of auxiliary feedwater system.

2.6 Loss of essential support system:

Some examples may be:

- Loss of AC/DC power;

- Failures of the emergency generator system;

- Loss of service water, air, gas, etc.

3. Significant Deficiencies in Design, Construction, Operation or Safety Evaluation Some examples may be:

- Deficiencies in design or construction which if not corrected could result in the loss of a required safety function;

- Personnel errors or procedural deficiencies which result in loss of plant capability to perform essential safety functions;

- Discovery of a major condition not specifically considered in the i authorized limits or previously analyzed.

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4. Significant Generic Problems Some examples may be:

Recurring incidents;

- Incidents with implications for similar facilities.

5. Significant Consequential Actions Significant consequential actions resulting from reported events taken by the competent safety authority on licensing, design or operation.

6. Incidents of Potential Safety Significance Events which have no significant consequences but may be considered as approaching "near-misses."

7. Effects of Unusual External Events Either of Man-made or Natural Origin t

Some examples may be:

An earthquake exceeding the safe shutdown earthquake; A flood exceeding the safe shutdown flood; An aircraft crash on a nuclear facility.

8. Events Which Attract Significant Public Interest l

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T&ex 630668 AEN/NEA APPENDIX B Sample IRS Report No. IRS XXX.X RESTRICTED DIFFUSION RESTREINTE Title -

Titre Inadvertent Isolation of Emergency Bus Country - Pays Date of Incident. - Date de 1* incident USA May 24, 1983 Type of Reactor - Type de rdacteur PWR -

Plant -

Centrale Licensee _ ,36 ten t eur du permis d 'exploita tion Beaver Valley - Duquesne Light' Unit N* -

Tranche n o Manufacturer - Fabricant I Westinghouse -

Power -

Puissance First Commercial Operation -

Bio Mwe(net) D te de mise en service October 1, 1976 Systems or Components Affected -

Systemes ou composants affectds Initial Plant Condition -

Eta t initial de la tranche 99% Power Way in which Incident was Detected ?

Comment l ' inci den t a-t-il dtd ddtects ?

T6 sting Radiation Exposure or Racioactivity Release -

Exposition aux rayonnements ou libdration de radioactivits None Date of Receipt Da te de relception -

Date of Distribution - Date 4e distribution

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Evont description, possible causes, actions 'taken or planned and lessons learned (cofety significance of incident) should be included in the following pages.

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Dascription de l' incident, causes possibles, mesures. prises ou projetees et

= - - - . -- ._

i The auther of the 27.5 report is recuested to check of f all items bcx).

relevant to the incident reported (please put an X in the 1

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1.

• Q.A.'pr'oblem (design, construction or installation) .'

E{ 2. ' Q.A. problem (operation or maintenance) including .

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3.

procedural deficiency or insufficient training Design peficiency or error

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4. Fabrication deficiency or error ,'

5. In'stallation deficiency or error

{ [~ _ 6. Operational error s

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__ 7. Violation of Operating technical specification '

__ B. Cause external.to the plant (earthquake, flooding,

, storm, etc.) .

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9. Environmental influence (pressure, temperature, humidity, etc.) ,

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10. Ageing ~

'll. Corrosion - ~

[. (( 12. Other fluid hydraulics effect (cavitation, erosion,

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vibration, gas binding, etc.)

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i 13. Loose parts .

[~ 14. Release of radioactive material '

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15. Exposure to radiation -

_. 16. Reactor trip (manual or a'utomatic) .'

1 2< 17. Actuat' ion of Engineered Safety Feature -

18. '

Chhllenge to safety and/or relief valve in the primary

! .. ' c'ircuit .

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19. Fuel cladding failure ~~

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20. Leak in the reactor coolant pressure boundary 1

- T 21. Defect in the reactor coolant pressure boundary

(( 22. Degradation of contai'nment integrity or leak tightness j '. __ .23. Complete loss of a safety function, such as loss of i

Emer'gency Core Cooling Systems or Containment Spray

'i System

24. Failure or degradation of control rod system or other ,

system required to control criticality

, 25. Loss of off-site power, ~

l . 2( !26. Lgss of on-site power, inc5' j.

8 ' power distribution system o,uding failurePower r Emergency of AC/DC Generation System i

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• ((] l 27. Failure o'r degra'dation of Residual Heat ' Removal System

, i or other system for core cooling during shutdown-i 7 ::2 8. Failure of supply system for service water, air or gas

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.2 9 . Failure of Reactor Protection System j,

30. Failure of instrumentation, supervision or monitoring i __

system j _-_ 31. Failure of' Control System

__ 32 Single failure ,

i __ ,3 3 . Multiple failure i __ '34 Common cause failure '

l __ 35. Interaction be' tween systems' -

l -- 36. Abnormal pressure or temperature transient in the -

i primary circuit l

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3 '3 7 . Fuel handling' incident

' 38. Internal fire .

(( 39.. Internal flooding

40. Radwaste incident l _J Ll. Discovery of major conc'ition not ;reviously considered ,

Event Description At Beaver Valley Unit 1 on May 24, 1983, during normal operation at 99% power, relay testing was being performed on the IB system station service transformer (relay A). Spurious actuation of relay B occurred, causing inadvertent isola-tion of the IDF emergency bus from its normal supply. Diesel generator No. 2 automatically started and picked up the bus, sequencing on all appropriate loads.

Cause(s) of Event n

An investigation into the cause of this event has revealed a number of contrib-uting factors. Lockout relay A has a normally open contact which closes to energize relay B. Relay A cannot be cleared by. itself without isolating a sub- .

stantial portion of the protection circuitry for the 1B system station service transformer. The relay was therefore tested uncleared. This should have posed no problem since relay A is a de relay and should not operate on the low ac

_ test voltages applied to the contacts. Unknown to the test crew, however,

.there was a capacitor installed between the 125 V de control power lines and

. ground for surge suppression. This suppres~sion capacitor inadvertently dis-charged when the test signal was applied to the relay contacts and activated relay B. The capacitor discharged because the test equipment utilized a grounded ac source. This ground provided a potential path for current flow, and. allowed discharging of the capacitor to occur.

When relay B activated, the feeder breaker to the emergency bus tripped open.

This action caused a " dead bus" condition on the emergency 4 kV bus, which intiated an automatic start of the diesel generator. After the diesel generator attained design output voltage, the output breaker closed and sequencing of all appropriate loads occurred. .

The test which led to the spurious activation was rerun. During subsequent performance of the relay test, relay B activated each time, as it had in the i initial event. This verified that it was the test that caused the relay actu-ation, a'nd not the shorting of contacts or carelessness on the part of the i

technician.

Actions Taken Personnel at Beaver Valley who work on relays were reinstructed in the performance of tests on circuits such as this and others which require

- special considerations.

-~ Reason for Reporting This event is reportable pursuant to Criterion 2.6, "Significant Degradation of Safety-related Systems."

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Appendix C Listing of Nuclear Power Plants with Events in Individual IRS Reports in 1983*

Unit Number of IRS Reports Arkansas 1 1**

Beaver Valley 1 1 Browns Ferry 2 1 Browns Ferry 3 2 Brunswick 1 2**

Brunswick 2 1 Calvert Cliffs 1 1**

Calvert Cliffs 2 1 Crystal River 3 1 Dresden 2 1 Farley 2 1 Fitzpatrick 1 Ginna 2 Hatch 1 1 Hatch 2 1 Maine Yankee 1 McGuire 1 1

, Millstone 2 1 North Anna 1 2**

Oconee 2 2 Oyster Creek 1 2 Peach Bottom 2 1**

Pilgrim 1 1 Prairie Island 1 1 Quad Cities 1 1**

Quad Cities 2 1 Robinson 2 1 St. Lucie 1 1 Salem 1 2 Salem 2 2 i San Onofre 1 2**

San Onofre 2 2 San Onofre 3 1 Summer 1 1 Surry 2 1 Trojan 1 Turkey Point 3 1**

Vermont Yankee 1 Yankee-Rowe 1 1 Zion 1 3**

Zion 2 2

• Does not include IRS Reports for generic topics.

• • One IRS report covered an event which involved more than one plant.

Appendix 0 Nuclear Power Plants in U.S. for 1983 Power Comm.

Reactor (MWe Net) Type Manuf. Docket No. Operation Arkansas 1 836 PWR B&W 313 12/74 Arkansas 2 858 PWR CE 368 3/80 Arnold, Duane 545 BWR GE 331 5/74 Beaver Valley 1 833 PWR W 334 4/77 Big Rock Point 63 BWR GE 155 12/62 Browns Ferry 1 1067 BWR GE 259 8/74 Browns Ferry 2 1067 BWR GE 260 3/75 Browns Ferry 3 1067 BWR GE 2% 3/77 Brunswick 1 790 BWR GE 325 3/77 Brunswick 2 790 BWR GE 324 11/75 Calvert Cliffs 1 850 PWR CE 317 5/75 Calvert Cliffs 2 850 PWR CE 318 4/77 Connecticut Yankee 582 PWR W 213 1/68 Cook 1 1054 PWR W 315 8/75 Cook 2 1094 PWR W 316 7/78 Cooper 778 PWR GE 298 7/74 j Crystal River 3 825 PWR B&W 302 3/77 Davis-Besse 1 906 PWR B&W 346 11/77 Dresden 2 794 BWR GE 237 8/70 Dresden 3 794 BWR GE 249 10/71 Farley 1 829 PWR W 348 12/77 Farley 2 829 PWR W 364 7/81 Fitzpatrick 821 BWR GE 333 7/75 -

Ft. Calhoun 1 486 PWR CE 285 9/73 Ft. St. Vrain 343 HTGR GA 267 7/79 Ginna 490 PWR W 244 3/70 Hatch 1 797 BWR GE 321 12/75 Hatch 2 806 BWR GE 366 8/79 Indian Point 2 873 PWR W 247 7/74 Indian Point 3 965 PWR W 286 8/76 Kewaunee 535 PWR W 305 6/74 373

• La Salle 1 1078 BWR GE Lacrosse 50 BWR AC 409 11/69 Maine Yankee 825 PWR CE 309 12/72 McGuire 1 1180 PWR W 369 12/81 W 370

• McGuire 2 1180 PWR Millstone 1 660 BWR GE 245 12/70 Millstone 2 870 PWR CE 336 12/75 Monticello 536 BWR GE 263 7/71 Nine Mile Point 1 610 BWR GE 220 12/69 North Anna 1 865 PWR W 338 6/78

Reactor (MNE Net) Type Manuf. Docket No. Operation North Anna 2 890 PWR W 339 12/80 Oconee 1 860 PWR B&W 269 7/73 Oconee 2 860 PWR B&W 270 9/74 Oconee 3 860 PWR B&W 287 12/74 Oyster Creek 620 BWR GE 219 12/69 Palisades 740 PWR CE 255 12/71 Peach Bottom 2 1065 BWR GE 277 7/74 Peach Bottom 3 1065 BWR GE 278 12/74 Pilgrim 1 670 BWR GE 293 12/72 Point Beach 1 497 PWR W 266 12/70 Point Beach 2 497 PWR W 301 12/72 Prairie Island 1 530 PWR W 282 12/73 Prairie Island 2 530 PWR W 306 12/74 Quad Cities 1 789 BWR GE 254 8/72 Quad Cities 2 789 BWR GE 265 10/72 Rancho Seco 913 PWR B&W 312 4/75 Robinson 2 665 PWR W 261 3/71 St. Lucie 1 804 PWR CE 335 12/76 St. Lucie 2 900 PWR CE 389 8/83 Salem 1 1090 PWR W 272 6/77 Salem 2 1115 PWR W 311 10/81 San Onofre 1 436 PWR W 206 1/68 San Onofre 2 1100 PWR CE 361 362

• San Onofre 3 1087 PWR CE Sequoyah 1 1148 PWR W 327 7/81 Sequoyah 2 1148 PWR W 328 6/82 830 PWR W 395

• Summer 1 Surry 1 775 PWR W 280 12/72 Surry 2 775 PWR W 281 5/73 Susquehanna 1065 BWR GE 387 6/83 Three Mile Island 1 792 PWR B&W 289 9/74 Trojan 1130 PWR W 344 5/76 Turkey Point 3 666 PWR W 250 12/72 Turkey Point 4 666 PWR W 251 9/73 Vermont Yankee 514 BWR GE 271 11/72 Yankee Rowe 175 PWR W 29 6/61 Zion 1 1040 PWR W 295 10/73 Zion 2 1040 PWR W 304 9/74 TOTALS: 53 PWRs, 25 BWRs, and 1 HTGR (79 Units)

• Not declared commerical in 1983 (reference NUREG - 0020, Vol. 8, No.1).

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' Appendix E Dominant Cause of Event Reported *  !

Procedural / , _ ,

1 Equipment Personnel Akinistrative Design Other Failure Errors Problem Error IRS # 102.8 249.1 ' 239.1 234.1 286.1 222.2 292.1 250.1 235.1 334.1 222.3 297.1 252.1 ' 255.1 232.1 338.1 257.1 285.1 232.2 258.1 289.1 233.1 284.1 289.2 236.1 287.1 298.1 .-

237.1 291.1 327.1 242.1 330.1 333.1-251.1 336.1 253.1 337.1 -

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254.1 256.1 259.1 281.1 282.1 .

283.1 288.1 -

290.1 293.1 294.1 295.1 .

296.1 299.1 329.1

, 331.1 332.1 335.1

• Does not include IRS reports for generic topics 1

APPENDIX F U.S. Reports Submitted to the NEA-IRS During 1983 IRS No.* Title 102.8 "J" Tube Degradation And Replacement 222.2 Temporary Loss of Shutdown Heat Removal Capability 222.3 232.1 Failure of The Reactor Protection System To Initiate 232.2 Automatic Shutdown 233.1 Failure of Undervoltage Trip Function of Reactor Trip Breaker 234.1 Loss of Shutdown Cooling Due To Testing 235.1 Potential Deficiency In Sigma Lumigraph Indicator Model 9270 236.1 Failure of Eleven Safety Relief Valves To Actuate At Setpoint 237.1 Inoperable High Pressure Coolant Injection High Steam Flow Switches 238.1 Loss of Decay Heat Removal Capability (Generic Report) e Loss of Suction To Residual Heat Removal (RHR)

Pump e Faulty Microswitch Causes "B" RHR Train To Be Inoperable o "A" Train RHR Service Water Pumps Inoperable From Flooding e RHR Service Water Loops Inoperable o Malfunctioning Static Inverter Caused Interruption of Decay Heat Removal 239.1 Significant Decrease of Steam Generator Feed Pump Suction Pressure 240.1 Moisture Intrusion (Generic Report) e Inoperability of Fire protection System o Inoperable Loop In The Partial Loss of Containment Isolation Valve Indication

• This is the sequential number assigned by the IRS Coordinator in Paris. The decimal indicates either an original report (i.e., 0.1) or an update (i.e.,

0.2 is the first update, 0.3 is the second update, etc.).

IRS No. Title e Failure Of RHR Pump Room Cooling Fan e Failure of Steam Supply Isolation Valve e Failure of Reactor Core Isolation Cooling System Turbine Exhaust Diaphragm Instruments e Inoperable Breakers To a Main Steam Line Drain Yalve 241.1 Inoperability Due to Extreme Cold Weather (Generic Report) e Frozen Instrumentation Lines Cause Inadvertent c Actuation of Engineered Safety Features e Inoperable Main Feed Flow Transnitter And Refueling Water Storage Tank Level Transmitter e High Oxygen Concentration In Drywell e Frozen Water In Fire Hydrants e Crack in Recirculation Line of the Steam -

Generator Drain Tank 242.1 Vital Bus De-energized Due To Noise Generated In System Control Circuitry 249.1 Inoperable Overpressure Protection System 250.1 Overexposure During Entry Into Reactor Cavity With i, Incore Instrumentation Thimbles Retracted 25'l.1 Stuck Tranversing Incore Probe Causes High Local Radiation 252.1 Tenporary Unavailability of Diesel Generator Emergency Pcser 253.1 Failed Dowel Pins Disable Diesel Generator 254.1 Loss of 24 Volt DC Power Common To Two Units 255.1 Common Cause Feedwater Pump Failures 256.1 Miscategorization of Loss of Feedwater Transient 257.1 Test Procedure Inadequacies 258.l' Loss of Auxiliary Power Affects Two Units

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IRS No. Title 259.1 Fuel Assembly Degradation In Spent Fuel Storage Pools 260,1 Potential Deficiencies of DS-416 Reactor Trip Breakers 281.1 Test Failure of Reactor Trip Circuit Breaker 282.1 Plant Trip and Partial Loss of Offsite Power 283.1 Failure of Control Rod Drive Coils 284.1 Inadvertent Release of Radioactivity 285.1 Insufficient NPSH Causes Inoperable Charging Pump Service Water Pumps 286.1 Flooding Incidents at Two Plants 287.1 Inoperable Containment Spray System 288.1 Valve Problems 289.1 Pressurizer Code Safety Valve Problem 289.2 290.1 Emergency Bus Loss Due To Breaker Problems 291.1 Inadv2rtent Release From Open Valve in Waste Gas System 292.1 Recurring Operator Errors Make Equipment Unavailable 293.1 Main Transformer Failures 294.1 Inadvertent Actuation of Engineered Safety Features Actuation System 295.1 Control Rod Drive Failure And Reactor Trip 3

296.1 Reactor Scrams Due To Excessive Valve Leakage

, 297.1 Inadvertent Plant Trip and Subsequent Cooldown 298.1 Unplanned Gasseous Release 299.1 Malfunctioning Isolation Condenser Isolation Valve l

327.1 Main Feedwater Line Break Due To Water Hammer l

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IRS No. Title 328.1 Blocking of Automatic Safety Injection Signals (Generic Report) e Blocking of Automatic Safety Injection Signals e Blocking of Automatic Safety Injection Signals 329.1 RCIC Switch Failure Due To Corrosion 330.1 Inadvertent RPS Trip With PORY Actuation 331.1 Loss of All Charging Pumps Due to Empty Common Reference Leg 332.1 Defective Station Battery Cells At Fitzpatrick 333.1 Broken Holddown Spring in Burnable Poison Rod Assemblies 334.1 Inadvertent Loss of Instrument Air To Salt Water Cooling Systen 335.1 Steam Generator Tube Degradation And Repair 336.1 Improper Use of Lubricant on Valve Seats 337.1 Inadvertent Safety Injection Signals During Testing 338.1 Unavailability of the Auxiliary Feedwater System

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