ML20199L730

Draft Review of Operating Experience History Through 1984 of Haddam Neck for NRC Isap

ML20199L730
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	07/03/1986
From:	Clemans V, Kimmins A OAK RIDGE NATIONAL LABORATORY
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20199L726	List: ML20199L721 ML20199L730
References
CON-FIN-A-9469 ORNL-NOAC-231, ORNL-NOAC-231-DRFT, NUDOCS 8607090593
Download: ML20199L730 (228)

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MAfI ORNL/NOAC-231 Engineering Technology Division Nuclear Operations Analysis Center REVIEW OF THE OPERATING EXPERIENCE HISTORY THROUGH 1984 OF HADDAM NECK FOR THE NUCLEAR REGULATORY COMMISSION'S INTEGRATED SAFETY ASSESSMENT P,ROGRAM A. D. C. Kimmins V. D. Clemons Draft MAFI Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Under Interagency Agreements DOE 40-554-75 NRC FIN No. A9469 Prepared by the OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37831 l operated by l

MARTIN MARIETTA ENERGY SYSTEMS, INC.

t for the l U.S. DEPARTMENT OF ENERGY j under Contract No. DE-AC05-840R21400 9607090593 860703

PDR ADOCK 05000213 P PDR 1

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• y FOREWORD The work reported here was undertaken by the Nuclear Operations An-alysis Center (NOAC) at Oak Ridge National Laboratory on behalf of the Office of Nuclear Reactor Regulation (NRR) of the Nuclear Regulatory Commission (NRC). The technical monitor for the proj ect was E. M.

McKenna of the NRR Integrated Safety Assessment Directorate.

NOAC performs analysis tasks, as well as information gathering activities, for the NRC. NOAC's activities involve many aspects of nu-clear power reactor operations and safety.

NOAC was established in 1981 to reflect the broadening and refocus-ing of the scope and activities of its predecessor, the Nuclear Safety Infor: nation Center. It conducts a number of tasks related to the analy-sis of nuclear power experience, including summaries of operation for U.S. power reactors, generic case studies, plant operating assessments, and risk assessments.

NOAC has designed and developed a number of major data bases that it operates and maintains for the NRC. These data bases collect diverse i types of information on nuclear power reactors from the construction l

phase through routine and off-normal operation. These data bases make extensive use of reactor-operator-submitted reports, such as the Licensee Event Reports.

NOAC also publishes staff studies and bibliographies, disseminates monthly nuclear power plant operating event reports, and prepares the Nuclear Safety Journal. Direct all inquiries to Joel R. Buchanan, Director, Nuclear Operations Analysis Center, P.O. Box Y, Oak Ridge National Laboratory, Oak Ridge, TN 37831, Telephone 615-574-0393 (FTS: 624-0393).

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= . . a f TABLE OF CONTENTS i

1 Page FOREWORD ....................................................... I t

LIST OF TABLES ................................................. \/

l LIST OF FIGURES ................................................ Vi EXECUTIVE

SUMMARY

.............................................. dU

! ABSTRACT ....................................................... XV l

l 1. INTR 3 DUCTION ............................................... I i

2. REVIEW METHODOLOGY ......................................... 4 2.1 Availability and Capacity Factors ...... ............... JI 2.2 Environmental Events .................................. 6 2.3 Forced Shutdowns and Power Reductions ................. 6 2.4 Reportable Events ..................................... 7 2.5 Evaluation of Operating Experience .................... 9

3. OPERATING EXPERIENCE REVIEW EVALUATION ..................... IO 3.1 General Plant Description ............................. /o 32 Evaluation Findings ................................... 10 33 Availability and Capacity Factors ..................... /1 3.4 Forced Shutdowns and Power Reductions ............... .. 11.

3.4.1 Systems Involved in Forced Shutdowns and l Power Reductions ............................... SLI 1

l Steam and Power Systems LJ 3.4.1.1 ...............

3.4.1.2 Electrical Power Systems .............. %t 3.4.1.3 Instrumentation and Controls ... .... ... Of 3.5 Causes of Forced Shutdowns and Power Reductions ..... .. ?dl l

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Page 3.6 Non-DBE Shutdowns ..................................... %7 3.7 DBE-Initiated Events .................................. 7. f 3.7.1 D2.2 - Feedwater system malfunctions that resulted in an increase in feedwater flow ...... IO 3.7.2 D2.2 - Loss of external electric load .......... Io 3.7.3 D2.3 - Turbine trips ........................... D 3.7.4 D2.6 - Coincident loss of onsite and offsite AC power to the station ........................ R 3.7.5 D2.7 - Loss of normal feedwater flow ....'....... 79 3.7.6 D3.1 - Single and multiple reactor coolant pump trips ..................................... 7I 3.7.7 D4.3 - Control rod maloperation ................ II 3.8 Reportable Events ..................................... 3I 3.9 Systems Involved in Reportable Events ................. If 3.9.1 Reactor Coolant Syssea ......................... 37

• 3.9 2 Electrical Power System ........................ JT 3.9.3 Steam and Power System ......................... 4-8 3.9.4 Instrumentation and Controls ................... fl 3.9.5 Chemical and Volume Control System ............. 42.

3 9.6 Reactog System ...................... 43

y. 9. 7 F,el H,,ggiwg, 5y:; fg,.,g . , ,. , . . . . . . 4.]

3.9 8 Engineered Safety Features Systems ............. Af 3.10 Causes of Reportable Events .......................... Of 3 11 Events of Environmental Importance ................... f7 3.12 Radioactive Release Events ........................... 4~7 3.13 Nonradiological Events ............................... f?

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s e Page 3.14 Review of Significant Events ......................... N 3.14.1 Loss of Offsite Power ........................ b 3.14.2 Unanticipated Reactor Cooldown ............... [8' 3.14.3 Loss of Auxiliary Feedwater .................. [9 3.14.4 Failure of a Feedwater Regulating Valve ...... ET 3.14.5 Blocked Open Blowdown Lines .................. [O 3 14.6 Reactor Coolant Pump Failures ................ S8 3.14.7 Pressurizer PORY Inadvertently Opened ........ S 3.14.8 Boron Recovery Tank Ruptured ................. 52' 3.14.9 Radioactive Spill Resulted in Unplanned Release ......................................

3.14.10 Control Rod Drives ..........................

3.14.11 Contaminated Moisture Separator Tube Bundles Released From Site ..................

3.14.12 Loss of Containment Control Air ............. SI 3.14.13 Failure of Refueling Pool Seal .............. [8 3.15 Trends and Safety Implications of Forced Shutdowns and Power Reductions .................................

3.16 Trends and Safety Implications of Reportable

'71 Events

4. CONCLUSIONS ................................................ U APPENDIX A: Review of Forced Shutdowns and Power Reductions ........................................ 87 APPENDIX 3: Review of Reportable Events ......................./Ii REFERENCES ..................................................... h edt.~)

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LIST OF IABLES Number 3.1 Availability and Capacity Factors for Haddam Neck .... 13 32 Forced Shutdown Summary for Haddam Neck .............. / 4 ct 3.3 Power Reduction Summary for,Haddam Neck .............. IS~

3.4 NSIC Primary Category Summary for Non-DBE Shutdowns at Haddam Neck ....................................... SL7

, 3.5 DBE Initiated Events Summary for Haddam Neck ......... 1.8 3.6 Summary of Systems Involved in Reportable Events at Haddam Neck ....................................... $7 3.7 Causes of Reportable Events for Haddam Neck .......... 44 3.8 Summary of Radioactivity Released from Haddam Neck .......................................... f'8 3.9 Unplanned Radioactive Releases at Haddam Neck ........ fk 3.10 Tabulation of Reports Categorized as Significant at Haddam Neck ....................................... S'/

3.11 Summary of Significant Event Categories at Haddam Neck ....................................... IlL-I

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• LIST OF FIGURES Number 3.1 Haddam Neck Plant .................................... Il 3.2 Availability and Capacity Factors for Haddam Neck .... Ik 3.3 Yearly Totals of Forced Shutdowns .................... Ib 3.3A Interpreted Forced Shutdown Rate . . . . . . . . . . . . . . . . . . . . . I7 3.4 Yearly Total of Power Reductions ..................... I8 3.4A Interpreted Power Reduction Rate ..................... 17 3.5 Steam and Power System Shutdowns ..................... 28 3 5A Interpreted Shutdown Rate with Steam and Power Systems as Causes .............................. ll 3.6 Electric Power System Shutdowns ...................... L1L 3.7 Forced Shutdowns due to Maintenance and Testing ...... 16~

3.8 DBE Initiated Shutdowns .............................. 1L9 3.9 Turbine Trip Shutdowns ............................... Il 3.10 Yearly Totals of Reportable Events ................... 35 3.11 Shutdowns Due to Equipment Failures .................. 08' 3.11A Interpreted Shutdown Rate with Equipment Failure as Cause ..................................... Ik 3 12 Yearly Totals of Significant Events ..................

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3.13 Yearly Totals of Conditionally Significant Events .... <J 3.14 Yearly Totals of Significant and Conditionally Significant Events ................................... 2h 7-3 15 Yearly Totals of Losses of Electric Power ............ 75 i 3.16 Yearly Totals of Control Rod Drives Problems ......... 7 5P i

3.17 Yearly Totals of Charging Pump Problems .............. 7G 1

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REVIEW OF THE OPERATING HISTORY OF HADDAM NECK THROUGH 1984 EXECUTIVE

SUMMARY

The Systematic Evaluation Program Branch of the Nuclear Regulatory Commission (NRC) is conducting a pilot program for the Integrated Safety Assessment Program (ISAP), in which all pending licensing actions and safety issues for selected operating reactors will be evaluated. A new approach to addressing a growing need for order and efficiency in the

! implementation and resolution of licensing requirements for operating nuclear power plants is being evolved under this program.

f The new ISAP approach provides a structure for the regulatory man-agement of licensing requirements on a plant-specific basis. One objec-tive is to assure the implementation of the most ef fective safety mea-sures in the near-term, while using both NRC staff and licensee re-sources efficiently. To accomplish this obj ective , ISAP will:

(1) evaluate all applicable issues related to plant safety in accordance with a pre-established scope; (2) identify cost-effective corrective ac-tions, where necessary, to enhance safety on a plant-specific basis; (3) establish a technical basis to judge implementation schedules; and (4) document the results of the evaluation so that the implementation schedule can be periodically updated, as necessary, to incorporate cor-rective actions for issues that may arise in the f..ture.

Tools utilized in the ISAP evaluation process include: (1) a de-terministic review of all pending licensing actions and safety issues; (2) a plant-specific Probabilistic Safety Analysis (PSA); and (3) an evaluation of plant operating experience and reliability data, including Yll

B S licensee performance (e.g., from existing Safety Assessments of Licensee Performance (SALP) evaluations].

As part of the pilot program for ISAP, the NRC contracted with the Oak Ridge National Laboratory Nuclear Operations Analysis Center (NOAC) to perform operating history reviews for two plants that volunteered to be included: Millstone 1 and Haddam Neck Plants. These reviews will be used as an integral part of the ISAP evaluation process. Each review includes the collection and the evaluation of data on availability and capacity factors, forced shutdowns, forced power reductions, reportable events, environmental events, and radiological release events. The data is analyzed and evaluated to identify any trends and symptoms that will be important in the resolution of regulatory actions to be applied to the plant. Observations and conclusions which focus on the key findings l of the review are provided.

The review of the Haddam Neck operating history incorporates the findings previously presented in the report generated under the Syste-matic Evaluation Program (SEP) NUREG-0826, " Integrated Plant Safety Assessment, Systematic Evaluation Program - Haddam Neck Plant". This new report updates those findings to include data for the years 1982 through 1984.

The operating history review focuses principally on evaluations of l

data which is divided into two segments: (1) data on forced shutdowns and power reductions and (2) data on reportable events. In the forced shutdown and power reduction segment, the review identifies Design Basis Events (DBEs) and recurring events that may be used as indicators of po- ,

tential operating ct.ncerns. DBEs are defined as tnose events delineated Vlil~

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i in the NRC's Standard Revieu Plan, and which result from failures in systems, equipment and operator actions that initiate system transients and challenge engineered safety features. In the reportable event seg-ment, which includes environmental events and radiological release events, the review identifies significant events and recurring events that may also be used as indicators of potential operating concerns.

Significant events are classified as those events in which a degradation of safety margin occurs, such that safe operation cannot continue to be assured or in which challenges to the safety protection features of the plant resulted. Selection of significant events was accomplished using a set of criteria pre-established in the SEP reviews. The findings of the review are summarized below.

Conclusions The Haddam Neck Plant has been generally operated in a safe and orderly manner. However the operating hist.ory reveals that there are a few areas with demonstrated deficiencies or indications of conditions relating to safety concerns. There are no indications in the operating history that there are extended or repetitious problems that have re-suited in any significant safety consequence.

A visit to the plant confirmed inferences derived from the opera-tional review data that attention to safety is an operating character-istic. A tour of the plant left an impression of a clean and well main-tained operation. There was evidence of positive approaches to imple-menting lessons learned from experience.

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One indicator of the quality of performance of plant operations is the ability of a plant to achieve and maintain high levels of reli-ability in operation. This also results in a consequential direct bene-fit to safety. Haddam Neck has a history of higher than average avail-ability and capacity factors. Twice the plant has accomplished extended uninterrupted operating runs, the first for a period of 343 days ending in August 1977 which established a record, and the second for 401 days ending in August 1984. .

Adverse effects of the extended runs were not conclusively evident from analysis of the data. While there was an observable increase in the number of maintenance-generated shutdowns subsequent to the 1977 run, the same characteristic was not observed in the final months of 1984. However, the rate of submittal of LERs af ter each run did in-crease, suggesting an increase in maintenance activities.

Apart from problems in the areas of off-site power, charging pumps, and control rod drives, in general any problems involving nuclear safety systems or equipment were found and resolved without the need of a con-sequent plant shutdown. Sixty-six percent of all shutdowns involving reportable events occurred before 1970, which supports the conclusion that there is an operational focus on preventing nuclear systems problems from resulting in forced shutdowns.

Trends and Symptoms Trends were discerned by examination of the data, and were more evident in the shutdown data than in the reportable event data. It was concluded that the forced shutdown data showed a definite asymptotic W

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trend of improvement. However, this conclusion was dependent upon the j l

premise that certain perturbations to the projected trend can be readily isolated, and correlated to an identifiable set of causes. Accordingly, it was determined that it would be necessary to conduct a broader examination of the operational history. However, the scope of the effort and . the resources available became limiting factors which con-strained the depth to which this examination could be conducted.

It was concluded that the causes of perturbations which occurred in 1974, 1978, and 1981 may be related to (1) extensive turbine problems, (2) an extended operating run in 1977, and (3) efforts in response to TMI, respectively. Actual symptoms of these perturbations were not di-rectly attributed to their theorized causes because of limitations in scope and resources. No direct correlation with reportable event fre-quency was detected. As another conclusion, it might be anticipated that the effects of major efforts applied to the repair or maintenance of plant components in one area may be evidenced in different and unex-pected areas.

In sunsaary, it was concluded from analysis of the trends that:

1. After an initial break-in period the overall performance of the plant was continuously improving with the exception of identified l

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2. Major "refic" efforts resulted in perturbations to the plants' overall performance.

3. Recovery from refits extended to one or two years, and probably be-l l cause of consequent increases in equipment failures.

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4. A perturbation as a result of a refit may not necessarily be evi-danced by problems directly related to that refit. Thus, the more important negative effects of refits may be caused by other reasons such as reduced attention by plant personnel to other areas not affected by the refits. It should be noted, however, that the shutdown data did not provide any strong indications that plant personnel relaxed their attention to nuclear safety systems.

Three important symptoms were identified through examination of the reportable event data. These were determined from analysis of the sig-nificant events.

Loss of offsite power During the early history of the plants the significant events were dominated by the occurrence of five loss-of-offsite-power events. These were all attributed to problems with the design of the plant protective relaying. Subsequently corrective measures to the protective relays were implemented. However, there was a later recurrence of this problem and additional loss-of-offsite-power events occurred later on during the 1980s. Thus it was concluded that this constitutes an outstanding symptom of a problem which was safety significant. Loss of offsite power is a precursor to station blackout, . which is a safety issue.

Haddam Neck sustained one actual station blackout during the period under review.

Control rod drive failures A significant number of control rod drop events were experienced during the first three years of operation. These were determined to have been principally caused by faulty relays. Corrective measures were f ..

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applied and there was no further recurrence for 6 years. However, as the plant grew older, control rod drive failures started to reoccur, for causes including separation of rod cluster control vanes, plastic deformation of coupling fingers, and other random equipment failures.

While no direct correlation to aging was made, it was concluded that this may be a factor in the perpetuation of control rod drive prob-less. The control rods are the principal mechanisms for reactivity con-trol and shutdown of the reactor, and thus serve an important safety function. Failure of the drive mechanisms to operate on improper opera-tion are safety concerns. Thus it was concluded that these failures of the control rod drives constitute an outstanding symptoms of a problem

which has safety significance.

1 Charging pump failures The charging pumps became significantly undependable in the latter half of the review period. For the first 9 years no problems were re-ported. However problems started to occur in 1976 and continued through the later half of the review period, again suggestive that aging may be I

a factor. Equipment failures were reported to have been caused typ-ically by wear induced vibration, erosion, chemical solidification, and leaks. The changing pumps serve to inject chemical additions to the re-actor coolant system and also to provide backup to high pressure coolant inj ection should this be required subsequent to an accident, both of which are safety functions. It was concluded that recurrence of oper-ability problems in the charging pumps constitutes an outstanding symp-tom of a problem which has safety significance.

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Other symptoms

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It was concluded that causes of downtime were principally for main-tenance and testing, occurring mostly in balance of plant systems in-cluding the turbine generator system, main steam systems, and electric power systems. In general nuclear systems did not contribute much to interruptions to plant operations.

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ABSTRACT

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A review of the operating experience- of the Haddam Neck nuclear power plant was performed for the Nuclear Regulatory Commission's Inte-grated Safety Assessment Program (ISAP) by the Nuclear Operations An-alysis Center. The operating history of the plant from 1967 through 1984 was reviewed and atalyzed. The findings of the review identified tre'nds and symptoms in the operating data that can be used as toois in the resolution and prioritization of the Haddam Neck ISAP issues.

The review includes evaluation of data collected on plant avail-ability and capacity factors, forced shutdowns, power reductions, re-portable events (reportable occurrence, licensee event reports, etc.),

and environmental considerations. The methodology used is also dis-cussed. Data and information is presented in appendices.

It was concluded that the operating history shows the Haddam Neck Plant have generally been operated in a. safe and orderly manner, with trends that show continuous improvement. Perturbations to these trends were correlated to theorized causes: (1) an extensive turbine refit, (2) post-effects of an extended operating run, and (3) effects of efforts applied in response to TMI. Three specific symptoms of con-tinuing saiety related problems were identified through: (1) recurrence of loss of offsite power events, (2) control rod drive failures, and (3)

I charging pump failures.

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1. INTRODUCTION In the early 1980's the Nuclear Regulatory Commission (NRC) imple-mented a program of reviews of older operating commercial nuclear power plants. This program, called the Systematic Evaluation Program (SEP),

grew out of a concern that some older plants were licensed under regula-tions that may have been less stringent than those in existence today.

Specifically, SEP accomplished five objectives:

1. The program established documentation showing how, for each operat-ing plant reviewed, the plant specific criteria compare with cur-rent criteria on significant safety issues, and provided a rationale for acceptable departures from these criteria.

2. The program provided the capability to make integrated and balanced decisions with respect to any required backficting.

3. The program was structured for early identification and resolution of any significant deficiencies.

4. The program assessed the safety adequacy of the design and opera-tion of currently licensed nuclear power plants.

5. The program used available resources efficiantly and minimized re-quirements for additional resources by NRC Jr industry.

Through SEP, the NRC recognized a need to provide order and effi-ciency in the implementation and resolution of regulatory requirements for operating nuclear plants. The experience gained from both SEP and the Interim Reliability Evaluation Program (IREP) enabled the NRC to develop a new and integrated approach called the Integrated Safety I

Assessment Program (ISAP). In early 1985, the NRC implemented a pilot program for the ISAP which has examined two volunteer plants - Mill-stone 1 and Haddam Neck.

The features of ISAP lie in the integration of many different pro-grams currently applied to the management and regulation of operating nuclear plants. Through an integrated evaluation of these programs, ISAP will enable a balanced and cost effective approach to the manage-a ment of regulatory requirements on a specific basis for individual oper-ating plants. Evaluation tools for ISAP will typically include:

(1) the deterministic reviews of all pending licensing actions and safety issues; (2) a plant specific Probabilistic Safety Analysis (PSA);

and (3) an evaluation of plant operating experience.

In SEP it was found that plant Operating Experience Reviews contri-buted significant value to the program. These reviews compiled and evaluated data on availability / capacity factors, forced shutdowns and i

reportable occurrences. They provided additional perspective for the integrated assessment of the plant, and it was recognized that their contributions were equally as significant as those made by safety topic evaluations and risk analyses.

l Thus, operating experience reviews have been included as an in-tegral part of the ISAP evaluation process and are used to (1) confirm the adequacy of the data used in the plant specific PSA, and (2) high-light strengths and weaknesses in plant operation and maintenance which could be considered as factors for judging the adequacy of any proposed corrective actions to resolve an issue.

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The NRC contracted with the Nuclear Operations Analysis Center, of the Oak Ridge National Laboratory, to perform the operating experience reviews for ISAP. This report presents the results of the operating ex-perience review of the Haddam Neck Nuclear Plant, and updates an earlier report published under the Systematic Evaluation Program, NUREG-0826,

" Integrated Plant Safety Assessment, Systematic Evaluation Program -

Haddam Neck Plant." The review includes evaluation of data collected on availability and capacity factors, forced shutdowns, forced power reduc-tions, reportable events, environmental events, and radiological release events.

Section 2 of this report contains a brief description of the methodology employed in the review. The technical approach is de-scribed, including identification of information sources, data collec-tion techniques, and data review procedures.

Section 3 contains the evaluations of the data.

Section 4 contains the conclusions derived from the evaluations of the data.

Appendices A and B contain the data presented both in tabular form with coded data, and in narrative form containing brief descriptions of I

! event data as yearly sununaries.

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2. REVIEW METHODOLOGY The objective of the review is to provide insights into the actual strengths and weaknesses of the design, operation and maintenance of the Haddam Neck plant. The results of the review will be used with other ISAP tasks, in an integrated approach to establish a manageable regula-tory baseline. From this baseline a plant specific "living schedule" of plant mdifications can be generated.

The evaluation of the operational histor? consisted of a four-step -

procets: (1) compiling information on plant operating events, including forced shutdowns and reportable occurrences, (2) screening the events to determine their significance, using selected criteria and guidelines (3) evaluating and categorizing events to facilitate a search for trends or symptoms in operating characteristics, and (4) preparing findings based on any observed trends or patterns identified in the data.

Data was compiled on the following tspects of operation: avail-ability and capacity factors, events of environmental importance includ-ing radioactivity releases, forced shutdown and power reduction events, and reportable events.

The focus of this evaluation was on forced shutdowns and power re-ductions, and on reportable events. Availability and capacity factors, and information about environmental events are used to establish an overall perspective on plant operations. Procedures ,that assured con-sistency in the review were applied to the screening and categorizing of the information about these events. Af ter the screening and categoriz-ing, a safety significance assessment of the events was made , and the existence of any determinable trends or relationships was established.

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v 21 Availability and Capacity Factors Both reactor and unit availability factors were collected for '.1 years of operation through 1984. Starting with 1974, the unit capacity factors, using the Design Electrical Rating (DER) in net megawatts (electric), and the Maximum Dependable Capacity (MDC) in net megawatts (electric), were compiled as well. Data on capacity factors was not -

available for earlier years.

Two availability and two capacity factors used in this report are defined as follows:

1. Reactor Availability =

hours reactor critical + reactor reserve shutdown hours 100 period hours

2. Unic Availability =

hours generator on line + unit reserve shutdown hours 100 period hours

3. DER Unit Capacity = not electrical enerEY Eenerated 100 period hours DER not e ec r ca energy generated

4. MDC Unit Capacity = ne 100 period hours MDC net The term " reactor reserve shutdown hours" represents the length of time the reactor is not critical (e.g. , the unit is shutdown for administra-tive or other similar reasons) when operation could have been contin-ued. The term " period hours" represent the total number of hours of the period under consideration.

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a O 2.2 Environmental Events Significant or recurring environmental problems were summarized based on the review of forced shutdowns, power reductions, reportable events (environmental LERs), and other operating reports. Routine ra-dioactivity releases were tabulated. Those releases where established limits were exceeded were reviewed in more detail.

2.3 Forced Shutdowns and Power Reductions Data on forced shutdown and power reduction events were collected and reviewed. Forced shutdowns are generally caused by equipment fail-ures or human errors that result in abnormal challenges to the unit 's operation. Power reductions in general result from some need for main-tenance or operations upgrade which does not require a full shutdown.

The power reductions and forced shutdowns are included in chronological sequence in Appendix A.

Each shutdown or power reduction was placed in one of two sets of significance categories. The shutdown and power reductions were first evaluated against criteria for Design Basis Events (DBEs) delineated in Chap. 15 of the Standard Revisu Plan.2 If the shutdown or power reduc-tion could not be categorized as a DBE initiator, then it was placed in one of the Nuclear Operations Analysis Center (NOAC) categories. The l

l method of assigning significance to and the coding of events is des-

! cribed in more detail in Appendix A.

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2.4 Reportable Events Information on operating events was collected from Licensee Event ,

Reports (LERs) and LER predecessors (e.g., Abnormal Occurrence Reports (AORs), unusual event reports , and Reportable Occurrences (R0s)] . This information was retrieved from the NOAC operational data files. Any documents that contained LER-type information (such as equipment fail-ures or abnormal events) were coded so that they could be reviewed in the same manner as an LER. Primarily, this involved various types of operating reports and general correspondence for the early 1970s. Other 3

sources of information such as the NucIsar Safety Journal and reports from the NRC Office for Analysis and Evaluation of Operational Data were also used.

Two sets of criteria, originally established in the SEP reviews, were used in determination of the significance of reportable events.

The first set addresses those events whose results include chal-1enges to the safety protection features of the plant. These events are termed " safety significant," and satisfy one or more of the following criteria:

Two or more failures occur in redundant systems during the same event

- Two or more failures due to a common cause occur during the same event

- Three or more failures occur during the same event

- Component failures occur that would have easily escaped detection by testing or examination

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- An event proceeds in a way significantly different from what would be expected

- An event or operating condition occurs that is not enveloped by the plant design bases

- An event occurs that could have been a greater threat to plant safety with (1) different plant conditions, (2) the advent of another credible occurrence, or (3) a different progression of occurrences

- Administrative, procedural, or operational errors are committed that resulted from a fundamental misunderstanding of plant per-formance or safety requirements Other The second set addresses events that have the potential to challenge the safety protection features of the plant. These events, which might re-quire additional information or evaluation to determine their full im-plication, were termed " conditionally significant," and satisfy one or more of the following criteria:

- A single f ailure occurs in a nonredundant system

- Two apparently unrelated failures occur during the same event

- A problem results in a offsite radiation release or exposure to personnel

- A design or manufacturing deficiency is identified as the cause of i a failure or potential failure A problem results in a long ourage or major equipment damage T

An engineering safety feature actuation occurs during an event A particular occurrence is recognized as having a significant re-currence rate Other The methods for assigning significance to and for the coding of events are described in detail in Appendix B.

2.5 Evaluation of Operating Experience -

The operating history of the plant was evaluated based on a review that involved screening, compiling, and categorizing data. Judgments and conclusions were made regarding safety problems, operations, trends (recurring problems), or potential safety concerns. Events were an-alyzed to determine their safety significance from the information pro-vided through the various operating reports and the review process. The Final Safety Analysis Report (FSAR) and conversations with plant person-nel provided specific plant and equipment details where necessary.

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3. OPERATING EXPERIENCE REVIEW EVALUATION 31 General Plant Description The Haddam Neck Plant (also called Connecticut Yankee Nuclear Power Station) is a Westinghouse Electric Corporation pressurized water reac-tor plant of 582 W(e) net maximum dependable capacity (1,825 megawatts-thermal (MWt)] (Fig. 3.1). The licensee and operator of the plant is Northeast Utilities, which is a company formed by three of the eleven partners comprising the Yankee Atomic Power Company that originally built the plant. Stone and Webster was the architect-engineer and also the constructor of the plant.

The plant is located on the Connecticut River about twenty-one miles south of Hartford, Connecticut. A " Facility Description and Safety Analysis Report" was submitted to the AEC on July 19, 1966, and the interim facility license, DPR-61, was issued on June 30, 1967. On December 31, 1969, the licensee applied for a Full Term Operating Li-conse, which was granted by the NRC on December 27, 1974.

32 Evaluation Findings The operational history data was evaluated as described in Sect. 2,

" Review Methodology". After the data was categorized, coded, and screened, searches were conducted to identify trends and symptoms or any other patterns in operating characteristics. The findings from these evaluations are set forth below, and are presented in the following order:

1) Availability and capacity factors

2) Forced shutdowns and power reductions 10

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3) Reportable events, including events of environmental importance and radioactivity releases. -

3.3 Availability and Capacity Factors The plant availability and capacity factors are shown for each year of operation through 1984 in Table 3.1 and on Fig. 3.2. Unit availabil-ity was good for all years except 1967, the first year of operation, and 1973, when extensive repairs to the turbine were performed. The cumula-tive unit availability (82.5%) was well above the industry average of

. 63.0% at 1984 years end.

i 3.4 Forced Shutdowns and Power Reductions The Haddam Neck Plant experienced a total of 181 forces shutdowns and reported 64 power reductions from the start of operations in 1967 through 1984. Figures 3.3 and 3.4 present in graphic format the yearly i totals of forced shutdowns and the yearly totals of power reductions re-spectively.

t 3.4.1 Systems Involved in Forced Shutdowns and Power Reductions The majority of forced shutdowns over the operating history of the plant occurred principally in three groups of systems: steam and power i

conversion systems, electrical power systems and instrumentation and controls systems. Tables 3.2 and 3.3 summarize the systems involved in the forced shutdowns and the power reductions, respectively. Figure 3.5 displays steam and power system shutdowns, and Fig. 3.6 displays elec-l trical power systems shutdowns.

I

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.. . - . ~ .

WP ,

Table 3.1. Availability and capacity f actors for Haddas Neck 1967 1968 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 1979 1980 1981 1982 1983 1984 Cumulatt Reactor availability 87.6 90.7 94.4 80.8 89.5 90.8 58.1 96.2 88.7 87.3 87.2 98.4 89.2 77.1 86.5 99.4 79.3 74.2 86.5 Outt availability 44.3 78.7 86.5 78.7 86.6 87.7 50.5 91.2 86.1 82.5 83.9 98.2 87.5 75.0 84.3 93.4 77.8 71.7 82.2 Unit capacity (tsc)(a ND 8 ND ND ND ND ND 48.1 92.0 87.9 83.4 83.3 97.7 85.4 73.1 83.2 93.1 75.9 67.3 81.9d past capacity (DEk)b ND ND ND ND ND ND 44.3 84.8 82.6 79.8 79.7 93.5 81.7 69.9 79.9 89.0 74.2 65.8 76.Sd

1. MIC = Haminima dependable capacity.

bbER = Design electrical rating.

1. ND = No data available.

JCalculated with a weighted average.

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s t I AVAILABILITY AND CAPACITY FACTORS 100-A s ***

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0 e i i i i i i i a i i i i 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.2 l

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i Table 3 3. Feuer reductlene sammeery for unddam leeck 1967 8%8 IM9 1970 8978 1972 1971 1974 1975 1976 1977 4978 1979 1980 1981 1982 8983 1984 Totale

1. Power reJactions

1. Total number 1 5 8 5 8 9 $ 2 S 3 3 8 I I 64

2. Cause A. Elecis ic power (EA) i 2 7 8 IS

s. nelatenance or testine 1 5 8 5 8 9 5 I S 3 I I I S3

3. system involved A. Elect ric power (EA) 2 8 1 5 32 B. Electric power (ES) 5 I

- C. Reactor (Cs) 2 I I I $

seg D. Steam anJ Power Con-version (HJ) 4 E. Turblue generator (HA) 8 I I 3 F. m in steam supply (HB) 1 I 2 8 8 6 G. Main condensers (HC) 3 3 5 8 2 I 6 39 H. Cisculating water 8 I system (HF)

3. ConJeumate and feeJ- 2 1 1 4 3 I I B3 water (HH)

.l . Nin stema system and controls (CC)

K. Reactus (kB) 1

5. . Caeeous radioactive 3 I waste (MR)

4. Total ns I,=r uf DBE related 6 power reJus tous (lac tuJeJ la totals in Part 1) if

YEARLY TOTALS OF FORCED SHUTDOWNS 30, 1

24,,

H 19 U 18, 12 11 11 9 8

llliilliillY 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84

, YEAR OF OPERATION FIGURE 3.3 l

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. YEAR OF OPERATION FIGURE 3.3A 17

o YEARLY TOTALS OP POWER REDUCTIONS 10, 8_ 8 H

U 6, I

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0 0 0 0 0 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.4 i$

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SHUTDOWNS CAUSED BY ELECTRIC POWER SYSTEMS 10_ .

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67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.6 11 O

3.4.1 1 Steam and power systems. Two-thirds (66%) of the total downtime reported by Haddam Neck resulted from problems with the steam and power conversion group of systems. Ninety-nine forced shutdowns re-sulted from problems with these systems (Fig. 3.5). Plant systems in-cluded in this group were turbine-generator system, condensers, steam generators, and feedwater systems. Most troublesome of these was the turbine generator system. In the first three years, problems with the turbine alone caused 1574 h downtime during 24 forced shutdowns. From 1970 to 1973 this race fell to only ten outages involving turbine fail-

. ures. In June 1973, the reactor shut down twice for a total of 4194 h to replace turbine blades. Again more blades were replaced in a March 1974 outage (660 h). Other periodic outages (6) occurring from 1975-1980 were for turbine balancing only. Overall, problems with the tur-bine generator and its controls accounted for 7533 h of downtime, over half of all outage time at Haddam Neck. The 56 outages for the turbine generator averaged over 120 h each, while the average outage for all other systems in the plant was about 60 h.

The condensers at Haddam Neck also caused some significant outage time. Fourteen forced shutdowns were required to plug tubes in the con-densers totaling 210 h. Ten of these shutdowns occurred from 1970 l

through 1972.

The condensate and feedwater systems caused 720 h of outage in 22 shutdowns. Most shutdowns resulted from repairs to the steam generators and their feed pumps. Feedwater regulating valves also required minor l repairs and adjustments. Since loss of feedwater control is classified as a DBE initiating event , these failures are discussed in more detail in Sect. 3.14.4.

l 13 l

5

3.4.1.2 Electrical Power Systems. The outages due to problems in the electrical power systems were generally short in duration. Most shutdowns involved maintenance on the offsite transmission lines, and repairs to vital bus equipment. Problems in the electrical power sys-tems were the cause of 29 forced shutdowns at Haddam Neck (Fig. 3.6),

resulting in 648 h downtime.

The most significant shutdowns, from a safety point of view, re-l sulted from faults in the protective relaying which isolates the plant from offsite power. These events resulted in seven reportable events which were classified as significant. They are discussed in more detail in Sect. 3.14.1.

3.4.1.3 Instrumentation and Controls. Problems with the instru-mentation and controls systems caused 24 shutdowns at Haddam Neck total-ing 332 h downtime. Most of these outages occurring in early years of operation were caused by spurious scrans for unknown reasons. Thirteen shutdowns caused by instrumentation and controls systems problems re-sulted in reportable events, and according by additional consideration was given to them in the reportable event reviews (see Sect. 3 9.4).

3.5 Causes of Forced Shutdowns and Power Reductions Tables 3 2 and 3.3 summarize the causes of forced shutdowns and power reductions, respectively. Of the 12,681 h total downtime at Haddam Neck, 79% (9966 h) was caused by maintenance and testing (Fig.

3.7). As discussed previously, the primary cause of plant shutdowns was maintenance and testing. The majority of shutdowns due to this cause were experienced as results of problems in the turbine generator, the M

f

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y

• U S e SHUTDOWNS DUE TO MAINTENANCE AND TESTING 20_

16, 16 S

H U 12-T 11' D

0 W 8_

N S 6 6 6 4_ 4 si 2 2 0 I 5 0 0 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.7

'5

i. .

feedwater aystems, and the offsite power systems. Maintenance and test-ing accounted for seventy-nine percent of the total downtime. The second largest cause of shutdowns or power. reductions was equipment failure. Seventeen percent of the downtime was attributed to equipment failure (2193).

The Haddam Neck Plant shut down twice (300 h total) for regulatory restrictions. On June 17, 1978, the plant shut down for 66 h to install heat shrunk sleeves on electric penetrations as requested by the NRC.

On September 29, 1979, the plant shut down for 234 h to inspect welds on the steam generator feed lines, also at the NRC's request.

There were 51 forced shutdowns which resulted in an LER. These events were furthered revie,wed as reportable events (Sect. 3.8). A total of 35 of these 51, or 65%, occurred in the first 3 years of opera-tions.

1 3.6 Non-DBE Shutdowns Table 3.4 summarizes the NSIC categories assigned to non-DBE shut-downs. Only the major categories are listed in this table. Equipment failure accounted for 85% of the non-DBE shutdowns. For the early oper-ating years (1968-1973), equipment failures were responsible for 93% of )

l the shutdowns or power reductions. All other NSIC categories con-l tributed to less than 5% of the non-DBE shutdowns. No discernible time trends appeared in the compilation of non-DBE shutdowns.

3.7 DBE Initiated Shutdowns Of the total of forced shutdowns and powe re etions reviewed for Haddam Neck, thirty-five were identified as Doc initiated events 15

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DBE INITIATED SHUTDOWNS i 6. 0, 5 5 4.8_

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3

(Fig. 3.8). None of these events initiated any sequence that led to a significant safety hazard at the plant. Table 3 5 lists the number of f

DBE initiating events at Haddam Neck for each year reviewed. There are no discernible time trends in the number of DBE initiating events oc-curring at Haddam Neck. The following sections present a brief discus-sion of those shutdowns which were categorized as DBE initiating events.

3.7.1 D1.2 - Feedwater system malfunctions that resulted in an increase in feedwater flow only 2 of 35 DBE initiating events fell into this category. In both instances, feedwater regulating control was lost and the water level rose in the steam generators.

On June 10, 1969, a broken feedwater regulating valve plug caused loss of control of feedwater and subsequent flooding of the No. 3 steam generator. The plug had dropped open allowing full feedwater flow to steam generator No. 3. (See Sect. 3.14.4 for further details.)

On February 22, 1973, the connector between the No. 2 feedwater i

control valve stem and actuator loosened, permitting the valve to go fully open. This resulted in an uncontrolled increase in the No. 2 l

l steam generator level. No serious transient resulted from this DBE initiating event.

3.7.2 D2.2 - Loss of external electric load i

On December 3, 1977, a generator voltage regulator malfuaction caused the plant to shut down for 7 h.

3.7.3 D2.3 - Turbine trips. Out of a total of 35 DBE initiating events, 21 were attributed to turbine trips (Fig. 3.9) . These outages resulting from turbine trips were all relatively short. Only 525 h of downtime accunulated from all the turbine trip outages.

$0 l

l

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SHUTDOWNS DUE TO TURBINE TRIPS 5_

4_ .

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N 2_ !j 2 2 ,

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OO 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.9 I

Six of the category D2.3 events were caused by failures of the tur-bine stop valves. On August 2, 1968, both turbine stop valves closed during a test of the lef t hand stop valve, tripping the turbine and then the reactor. A week later, on August 9, the right hand stop valve closed following a planned closure of the left hand stop valve. The first trip was caused by a stuck-open check valve in the oil discharge line from the right hand valve. This caused oil from the left hand valve to drain through the right side, resulting in a closure of both valves. The plant was shut down to repair the valves. On August 5, the plant returned to service until August 9, when the right hand valve in-advertently closed following closure of the lef t hand valve. Failure of the right hand valve was attributed to loss of oil pressure in the valve auto-stop system.

On January 8 and March 29, 1969, problems with the right hand stop valve resurfaced. The first failure resulted from leaking oil in the auto-stop system. The second trip was due to failure of servo-motor cup valve. On July 14, 1975 the auto stop oil system began leaking, forcing

. the reactor down for three hours.

Additional turbine trips were caused by a variety of reasons. The l turbine tripped three times on overspeed (August 2 and November 20, l

1980, and December 11, 1981). The turbine governor system malfunc-tioned, tripping the reactor on two occasions (November 11 and 12,

! 1969). Maintenance errors during wiring modifications caused the tur-bine to trip on March 27, 1980. Oa March 28, 1968, the turbine tripped on an inadvertent activation of the turbine low vacuum trip signal. On July 5, 1975, the turbine tripped due to a broken oil pressure gauge

~

3L

• t

-- - -r --- ,- -

--r

line on the turbine. A constant voltage transformer failed causing a turbine trip to occur on September 21, 1972. Repairs on the unit re-quired the plant to shut down for 85 h. All of these turbine trips re-suited in relatively minor consequences and short outages.

On July 29, 1981, while reducing load, a turbine and reactor trip occurred due to a low pressure steam dump malfunction caused by OPC relay 63. While reducing power on December 11 to plug the B waterbox J condenser tubes and for scheduled turbine maintenance, erratic operation of the turbine control valves caused a turbine trip due to overspeed.

On December 22, 1981, a turbice and reactor trip occurred due to high pressure heater drain-tank level due to a failed fitting on control air system.

In 1982 there were two occurrences of a turbine trip. On January 31, six blown fuses resulted in the loss of the generator field, causing the turbine to trip. It was determined tha't the fuses had blown as a l result of short circuits in the exciter. Investigation showed that i

shims used for leveling inside the exciter casing had partially sheared l

off and th'at the excess metal had been blown into the wiring by the cooling fans, causing shorting between circuits. The shims were trimmed to prevent reoccurrence.

3.7.4 D2.6 - Coincident loss of onsite and offsite AC power to the station On April 27, 1968, a switching error activated a transfer trip re-lay, causing site feeder breakers to open. This action isolated the plant from offsite power. The reactor subsequently tripped and the three diesel generators started but failed to load. Thus, the plant was N

4

-,y, ,,w,--,..,. , ,_ - - - - - - - - - _ _ _ _ - _ _ _ _ _ _ - - -

without AC power for a brief period of time (25 min.) . The resulting outage lasted only nine hours; however, the plant returned to full power without thoroughly testing the AC power system, a breach of regulatory specifications. Due to the safety implications of a total station blackout, this event is discussed in further detail as a reportable event, in Sect. 3 14.1.

3.7.5 D2.7 - Loss of normal feedwater flow Seven loss-of-normal-feedwater events occurred over the period of the operating history review. Five of the seven failures involved loss of feedwater control.

On August 23, 1968, a feedwater regulating valve failed closed due to loss of control air. The shutdown lasted less than two hours. A loss of feedwater control to steam generator No. 4 resulted from the failure of a solenoid on August 21, 1971. The solenoid was replaced and the unit returned to service in four hours. No cause was found for failure of the solenoid. On September 7, 1971, a momentary ground of a vital bus resulted in loss of control of feedwater to one of the four steam generator feedwater solenoids. Again, no specific cause was found for failure of the solenoid. On February 1,1975, the plant shut down for 16 h to remedy low feed pump suction pressure. On February 29, 1981 l

the separation of a fitting on the control air header caused the loss of feedwater control. Low feed pump suction pressure was again the cause of a feedwater trip on November 8' 1982. The plant was returned to ser-vice af ter system components were verified to be operating properly.

i

. o 3.7.6 D3.1 - Single and multiple reactor coolant pump trips Only one reactor coolant pump trip was experienced during the pe-riod under review. On September 10, 1976, a coolant pump shut down as a result of an operator error. The plant was shut down for seven hours.

3.7.7 D4.3 - Control rod maloperation on November 18, 1980, Haddam Neck experienced problems with two rod gripper coils, resulting in two dropped rods. The plant shut down for 14 h to replace the coils. Several control rod drive failures occurred; however, none caused serious problems. Another occurrence of r'od drop was experienced on November 17, 1982. The bank "C" rods dropped during rod motion checks for an undetermined cause. Control rod drive anom-alles are discussed in more detail in Sect. 396 in the reportable events discussion.

38 Reportable Events A total of 342 reportable events that occurred at the Haddam Neck l plant from the beginning of operations in 1967 through 1984 were re-l viewed. Figure 3.10 presents in graphic format the yearly totals of re-portable events.

3.9 Systems Involved In Reportable Events A compilation of all reportable events by system and year is pre-l sented in Table 3.6. The systems listed in this table represent groups l

of systems related by function. Systems which had no reports filed are omitted. Most reportable events involved the following systems: reac-tor coolant, electrical power, steam and power, instrumentation and con-trol, chemical and volume control, reactivity control, and engineered N

1 l _ _ _ .

O 4 YEARLY TOTALS OF REPORTABLE EVENTS i

40_ ,

! 31 32_

! 24, D 23 i

16 16 11 11 E i t i 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.10 t

Table 3.6 Summary of systems involved in reportable events at fladdas Necit 1976 1977 1978 1979 1980 1981 1982 1983 1984 Total 1967 1968 1969 1970 1971 1972 1973 1974 1975 System I I I I 2 22 2 1 kerctor 1 4 6 2 4 3 39 I 3 2 2 3 3 3 2 i 2 5 3 2 60 ke.sctor cool.snt 4 9 3 I I I 2 6 Instrinsentation an.1 controls 7 9 4 5 1 I 2 4 4 2 6 3 53 4 4 1 3 2 2 6 2 3 t:lec t r ic powei 7 3 1 2 3 2 6 2 5 6 3 9 to 56 I 3 3 1 Engineered safety teatures I 1 1 3 Fuel lianJilng .uul stos.sge 2 3 1 10

! I I 2 1 A Auxiliary water 2 4 3 2 4 72 2 3 I I 7 to 16 2 Ste.a ai:1 power 1 5 7 2 C 2 1 1 Radiation protec tion 2 6 2 1 19 3 2 1 kaJioactive w.a s te m.au .sgemen t 2 4 2 7 41 3 2 2 6 4 5 1 1 2 1 1

! 10 10 Auntliary process i  !

I 2 6 O t tier atuzillery systeins 1 I i 1 No ayutem applicatale 41 21 27 24 11 36 40 400 6 9 18 15 6 28 37 Total 17 24 27 16 s.1 J

l

O O safety features. Problems with these various systems are discussed as follows.

391 Reactor Coolant System TM designation of reactor coolant system encompasses a broad range of heat transfer related equipment in the reactor. This includes the reactor coolant pumps, pressurizers, relief valves, and support struc-tures for equipment in all four main cooling loops.

! Reactor coolant pump (RCP) seal failures occurred three times at the plant (AO's 70-05, 77-19, 80-12). All of the failures affected only one of four reactor coolant pumps and the first two are considered sig-nificant (see Section 3 14.6).

Three inadvertent openings of the pressurizer PORV occurred at Haddam Neck. On August 13, 1979 (LER 79-10) a PORV and its isolation valve failed open due to failure of a bistable in the pressurizer pres-sure controller. Reactor pressure dropped from 2000 to 1950 psig. The signal was overridden. This action closed both valves and prevented further depressurization. This event is discussed further in Sect. 3.14.7. On February 4, 1980 the same valves opened again (LER 80-04). The PORV was open for about 2 min before it was manually closed. No cause was reported for the second event. On April 3,1981, l the pressurizer relief and blocking valves opened due to a loose elec-trical connection (LER 81-03).

Control of the pressurizer spray valves and the PORV's was lost twice during 1983 due to the loss of containment control air. On Novem-ber 1 the containment control air was lost due to a maintenance error on 4

a filter (LER 33-20) . The filter was isolated and repaired. Control P

9 - v e O air was again lost on November 28 due to a broken air filter canister (LER 83-21). In both events, control air pressure was restored in less than one hour.

On August 19, 1970, a small fire was discovered and extinguished at the juncture of an RCP and the suction piping ( AO 70-08). Reactor power was reduced to 65% so that personnel could enter the loop areas to make repairs. Oil had dripped from the thermocouple conduit attached to the motor thrust bearing. The dripping oil contacted the 520'F coolant pipe and vaporized. As the v'apor emerged from the insulation around the pipe

'it ignited. All insulation from the pump outlet to the center of the stop valve was replaced and the oil leak was stopped with epoxy. (For further details , see Sect. 3.14.6).

3.9.2 Electrical Power System Fifty-three failures were reported for the electrical power sys-tems. Systems included in this designation are AC onsite power, offsite power, and energency generators. Failures of the onsite AC power ac-counted for fourteen reportable events over the operating history of Haddam Neck. Eight of these events involved reactor trips due to fail-l ure of the AC vital bus, five of which occurred in the first five months l of operation. Component failure was responsible for two of . the eight bus failures. The remaining failures were caused by workers inadver-tently grounding the bus during maintenance. The bus and related equip-ment were replaced several times, the failures then ceased af ter 1968.

Seven loses of offsite power also occurred at Haddam Neck. These fail-ures are discussed further in Sect. 3.14.1.

73

3.9 3 Steam and Power System Most failures in the steam and power conversion systems were attri-buted to problems with the steam generators (HB system). Four failures of the steam generators were due to failures of hold down bolts and seismic supports. In all instances no cause was reported and the sup-ports were replaced.

Four leaks were reported at the junction of' the waste liquid steam generator blowdown discharge piping and the service water effluent line. The leaks were all caused by a rapid deterioration of the piping material where the hot blowdown water contacted the relatively cold ser-vice water effluent. In each of the first two failures (A0 76-13, A0 77-01) a small amount of tritium was releared, well below the allowable limits. The leak recurred !n March 1978 (LER 78-03) and again in February 1980 (LER 80-07). Small amounts of tritium were released in the 1980 leakage. In each instance, the affected areas were replaced.

Repetitive failure of the piping is attributed to corrosion due to ther-mal shock.

Six reportable events occurred due to main steam isolation valve (MSIV) failures.

Four of the failures were caused by binding of the l valve gland packing. In each case the packing was adj usted and the valves operated satisfactorily. The fif th MSIV failure occurred on June 6, 1982 (LER 82-04). This valve failed to cycle during a shutdown due to a warped tail link. It was concluded that the tail link was damaged by a jacking device used during corrective maintenance in the September-November 1981 outage. The final MSIV failure occurred on August 23 ,

1984 (LER 84-16). An isolation valve in the main steam drain line i

bT 4 J

, e failed to close during a test. During subsequent maintenance of the valve, it was discovered that the valve suffered from a bent stem. No cause for the damaged stem was determined. The valve was repaired and ratested satisfactorily.

3.9.4 Instrumentation and Controls The instrumentation and controls system was responsille for 60 re-portable events at Haddam Neck. This high number of events is due to the function of the system as a safeguard for more serious occur-rences. Twenty-six of the events resulted in reactor trips. The causes of the reactor trips are split evenly between equipment malfunction (9 events), maintenance or operator error (8 events), and unknot.n causes (9 events). Most of the reactor trips (21 events) occurred during the first four years of the plant operation.

Two of the events that involved the instrumentation and controls system were classified significant to safety. The first event occurred l

l

' on October 12, 1970 (A0 70-09). A reactor trip occurred due to an er-roneous loss of flow signal. Following the reactor trip, one of the steam dump valves failed to close due to the valve positioner being out of adj ustment. The second event occurred on August 25, 1978 (LER 78-18). An FM transceiver caused a rod drop alarm. Following the rod drop alarm, the turbine load runback alarm failed to initiate due to a closed pressure switch isolation valve. The isolation valve disabled the turbine load runback feature. Further details on these two events can be found in Sects. 3 14.2 and 3.14.10. On January 18, 1974, ins tru-ment sensing lines froze due to cold weather ( A0 74-02). The frozen lines caused two main steam line high flow alarms and subsequent reactor

$ D trip. Plant personnel implemented a number of changes to prevent recur-rence. This event was classified conditionally significant because ac-curate and reliable information from instrumentation is essential for safe plant operation. Frozen sensing lines could produce misleading and erroneous information to an operator. The remaining failures reported for this system produced no serious consequences.

3.9.5 Chemical and Volume Control System Thirty-six reportable events were caused by failures of the chem- -

ical and volume control system (CVCS). Eighteen of the failures re-ported for this system involved leaks while two of the leaks resulted in unplanned releases of radioactive material. 5Act of the events reported on the CVCS were due to failures of the charging pumps. Since the charging pumps also serve as the high pressure injection pumps, these failures are discussed in greater detail in Sect. 3.16.

On May 3, 1969 the Boron Recovery Distillate tank ruptured due to an improper valve lineup (A0 69-06). Approximately 200 gallons of radioactive distillate were spilled in the surrounding area. This event was classified significant to plant safety and is discussed further in Sect. 3 14.8.

on July 15 , 1969 all offsite power was lost during switching i

l changes involving the 115 kV supply lines. Subaequently, one diesel generator failed to start and one charging pump failed to run. Due to the safety significance of this event, it to discussed in detail in Sect. 3.14.1.

.12.,

• s 3.9 6 ReactC./t . _ System The reactog . system Wa.S involved in 22 of the report-able events that occurred over the operating history of Haddam Neck.

Fifteen of these reports involved failures of the control rod drive ,

(CRD) systa.a. Due to the safety implications of problems with the CRD system, these failures are all discussed in more detail in Sect. 3 15.

During the May 1970 refueling outage two radial vanes in a control assembly were found to be cracked. Each of the radial vanes to which the control rods were attached had broken loose from the hub assembly.

This is discussed in greater depth in Sect. 3.14.10.

3. 9. 7 he / bd/MG W!kM -

On August 21, 1984 the reactor cavity seal ring failed (LER 84-13). The failure occurred 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> before refueling was scheduled to begin. During the 20 minutes following the seal failure, 200,000 gal-lons of radioactive water drained from the reactor cavity to the lower levels of the containment building. The original all-metal seal had been replaced with an inadequately

• designed seal that included flexible rubber boots. Although the seal failure did not pose an actual threat to safety, the consequences of the seal failure could have been severe had refueling operations already begun. The event is discussed in de-tail in Sect. 3.14.12.

3.9, Engineered Safety Features Systems The engineered safety features (ESF) systems were involved in 56 of the reportable events that occurred at Haddam Neck. The ESF systems in-clude both structural and mechanical equipment which perform both pas-sive and active safety functions. Included under the ESF designation are the reactor containment structure, containment isolation and control equipment, emergency core cooling system, and other saf ety equipment.

2

o Five containment isolation valve failures involved check valves for the Component Cooling Water (CCW) system. All of the failures occurred during the last four years of the operating history under review. In each event the check valves failed to seal properly due to rust / scale deposits in the valves.

Failures involving the containment air recirculation system ac-counted for t.lu4 events. Half of the events ' involved the recirculation dampers and their associated control linkage. Four other events in-volved the service water supply to the fan coolers. The air recircula-l tion system is needed for heat removal from the containment atmosphere I

during both normal operation and following a loss of coolant incident.

l 3.10 Causes of Reportable Events Table 3.7 pres _ents a summary of causes of reportable events at Haddam Neck. No causes were reported for several events in 1967-1973.

Over 60% (20t}) of all reportable events at the plant were attributed to incipient failure. Incipient failures are failures in which only the component itself is held accountable. This includes set point drifts, wear out, and many failures for which no cause could be determined.

Over the operating experience reviewed, human error was responsible l for 36% (126) of all reportable events. Human error can be classified i

into two categories: (1) in-plant personnel (maintenance, operator, and installation errors), and (2) out-of-plant personnel (administrative, design, and fabrication errors). In-plant personnel accounted for 65%

of the human errors (82 events) while out-of-plant personnel accounted for 35% (44 events). The mos t common human error found in the report-able events was naintenance error.

hb

Table 3.7 Causes of reportable events for Haddam Neck cause 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 Total Adrainis t ra t ive 1 2 I 1 1 1 3 4 1 3 j8 Design 1 ,1 1 2 1 5 2 4 1 1 2 3 24 Fabrication 2 1 3 Inherent failure 9 13 15 8 4 4 8 9 3 12 24 23 10 11 12 7 19 18 ;209 Installation 1 1 2 1 2 3 3 4 2 1 do Lightning 1 l A is Maintenance 6 6 2 1 3 1 1 2 3 2 1 1 1 3 3 36 Operator 1 4 1 1 1 5 1 4 3 1 5 27 7

Weather 1 2 3 1 Unknown g 3 1 4 Total 16 23 24 11 6 8 14 13 6 23 33 39 20 25 21 10 26 33 350

• Sac Eyty fS lAny'ohsf /H2C YlMN 046 C/ll/Jf.

3 11 Events of Environmental Importance Haddam Neck reported 53 events of environmental significance through the 18 years of operating experience. The reports covered un-planned releases of radioactivity (24 events), excessive doses to work-ers (4 events), fish impingement on intake filter screens (18 events), a high rate of change of the discharge temperature (6 events), and a high discharge canal PR (1 event).

3.12 Radioactive Release Events A summary of radioactivity releaces for Haddam Neck is shown in Table 3.8. The table lists the airborne and liquid releases and the solid waste ' shipped for the years 1967 to 1984.

Haddam Neck reported 2f unplanned releases of radioactive gases or liquids, shown in Table 3.9. In addition, a shipment of solid waste was released prior to a health physics review. Of all releases, solid, liquid, and gaseous, human errors caused 16 of them. Operator errors caused six releases, administrative caused four, installa-C^used IfuttC pad MAtaf&avec cwowa' oNF '

tiond j design errors caused two.f' In all releases, the workers received less than the maximum permissible dose.

Four events , though not releases, dealt with the overexposure of plant personnol. In the fourth quarter of 1973, two men had quarterly exposure readings of 3.03 and 3.66 rem. During the 1975 refueling out-

. age, a maintenance worker was overexposed while installing a scismic re-straint system. A dosimeter read 2.19 rem exposure while a film badge read an exposure of 3.Il rem. Another overexposure occurred due to inadequate recording procedures in F. arch 1979. The final personnel 47

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Table 3.9. Unplanned radioactive releases at Haddam Neck Report Event Cause Description of release Number Date A06906 050369 H Baron recovery tank ruptured due to improper valve lineup, small release resulted.

A06907 050669 D 500 gal rad liquid spilled on baron recovery area, small release resulted.

041871 H Radioactive iodine released due to operator error with penetration seals. Workers exposed to a release of 700 nC1 1-131.

A01202 051972 N Unplanned release (1.9 Ci of Xe) from domineraliser due to operator error.

A07306 062173 D Unplanned airborne release (0.238 C1) due to leak in purification system valve.

A07307 062373 A Unplanned release of radioactive liquid when letdown system placed in service.

A073tl 110l73 D Valve on RWST thermosiphon leaked radiontive water through a bad diaphragm.

Leaked into a local storm sewer.

A07407 042674 D Unplanned radioactive release from auxiliary building exhaust f an due to leak in diaphragm in hydrogen supply regulatory.

1.ER7608 033076 E Unplanned release of radioactive gas (11.59 C1). Two rupture disks in vaste gas tank failed due to damage during installation.

LER7613 061576 D Small leak in wall of safety injection cubicle due to deterioration of steam generator blowdown piping.

LER7701 121476 D Small leak in liquid waste line. Tritium activity in external sump increased.

ETS7708 110477 C A river effluent monitor sample pump was not operating while 21 tanks were drained.

LER7721 091877 D Unplanned release'of 7.44 C1 radioactive gas Shen a diaphragm ruptured in a waste M ETS7810 090078 D gas decay tank.

Tritium level in river sample exceeds limit.

EER7906 121679 B A level control valve relay failed, 15.8 C1 released in 10 minutes via che stack.

LER8003E 042880 D Unplanned radioactive release occurred when when a desassifier rupture disk flange cracked. Instantaneous rolesse limits were not exceeded.

IER8004E 051980 A Tritium activity exceeds release limit due to large volume of processing water for upcoming refueling outage.

LER8002E 050480 A Radioactive release limite exceeded when an ion exchanger was replaced. Cases released were I.7 and 2.6 times the tech spec limits.

1.ER8005E 052880 B A waste gas system valve opened releasing I.7 C1 of Xenon in etz minutes.

LER800?E 092680 H An operator opened the wrong valve, 1.3 C1 released via the stack.

LER8013 092680 N Technician made a sampling error allowing radioactive gas to be released.

Technician received a 9 mres exposure.

1.ER8100S 042283 A Contaminated tube bundles were released f rom ette ptf or to health physica review.

LER8100E 081681 N Noble gas release rate exceeded limite due to operator error.

LER8tl5 091781 E Radioactivity (0.533 oci) released via cracked exhaust duct to stack.

PNo-I-63 090683 D 3.3 C1 of Noble asses released to stack when relief valve on the waste gas surge

-95 tank lifted.

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overexposure occurred on October 13, 1984. One worker received a high exposure (quarterly reading of 2.8 rem) because an unqualified health physics technician was assigned to the work area. Table 3.9 provides a list of the unplanned radioactive releases at Haddam Neck.

3.13 Nonradiological Events From 1976 to 1981, Haddam Neck experienced 11 instances of exces-sive fish impingement on the intake screens. The problem was attributed to an unusually large fish population during those years. Other events of environmental concern include a rapid change of the discharge temper-ature (6 events), hypochlorite inadvertently released into the river (1 events), chlorine discharged into the river ( event), and the dis-charge canal reaching a high PH level (1 event).# No events of environ-g ene nepaa/rd mental importance during the last three years (1982-1984) of the review.

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3 14 Review of Significant Events -

Each reportable event was screened, using various criteria, as an additional step in the evaluation process. A tabulation of the signi-ficant events by year at Haddam Neck are given in Table 3.10. Events with serious safety implications are described in detail in the follow-ing sections. Table 3.11 presents a summary of these events. The events which were classified as significant are:

1. losses of offsite power (7),

2. unanticipated reactor cooldowns (2),

l . _ _ . .__ . _ _ . _ - _ _ _ _

Table 3.10. Tabulation of reporte categorized as significant at Haddam Neck Significance N date Event description Section A06707 9/22/67 S3 Inadvertent opening of all lo steam dump valves A06807 4/27/68 S3,57 Maintenance crew inadvertently opened site feeder breakers -

All three DCs loaded then tripped of f - Total station

. blackout A06906 5/3/69 S8 8oron recovery tank ruptures A06907 5/6/69 58 500 gal rad liquid spilled in boron recovery area A06938 6/10/69 57 High level in steam gen No. 3 due to broken feedvater flow-control valve-reactor tripped A06909 7/15/69 53jS7 Loss of offette power, I DC f ailed to load. I charging pump failed to run A06910 8/2/69 S7 Complete loss of offsite power due to lightning strike on ta'ephone relay A07008 8/19/70 S7 Small fire near RCP due to leaking oil A07009 10/12/70 5),S7 Loss of feedwater flow and steam dump valve out of positidA f - reactor tripped A07403 1/19/74 53jS7 Loss of offsite power during ice storm - Su pumps on oc did u- not start automatically d LER7634 6/24/76 S7 Totaljoesofoffsitepower-RHRflowlost3 times l LER7616 7/5/76 S2 80th aux feedwater pumps f all - Common cause failure of both pumpe due to faulty check valve LER1789 8/21/77 S3 RCP seal f alla during operation, causing other 2 seals to fail LER7818 8/25/78 S8 FM transceiver caused dropped rod alare, followed by failure of load runback signal LER7822 12/29/78 53j56.S8 Air supply valves to isolation valves blocked open, loss of l blowdown and loss of cc.itainment high pressure isolation l LER7910 8/13/79 S7 Pressuriser /80RV inadvertently opens LER8100S 4/22/81 S8 Contaminated tube bundles released from site prior to health physica review LER8320 11/1/83 S2 Loss of containment control air due to incorrect o-ring LER8321 11/28/83 S2 loss of containment control air due to broken air filter canister PNO-I- 3/15/83 S7 Improper latching between control rod dri,ve shaf ts and rod l 8320 control cluster assemblies LER8409 8/l/84 S2,55 Total lose of offsite power due to inadvertent closing of breaker LEB8413 8/21/84 S7 Failure of the refueling pool seal due to improper design LER8414 8/24/84 S7 Total loss of offsite power due to maintenance error i

e

=

)

t l

I Table 3.lf Summary of significant event categories at Haddam Neck l

D "III"""'" 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 Total category St O S2 I 1 2 1 $'

S3 1 1 l l 1 1 6 S4 0

, vg SS e 1 1 in S6 1 1 l

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I l

I i

j t *

)

3. loss of auxiliary feedwater (1),

4. failure of feedwater regulating valve (1),

5. blocked open steam generator blowdown lines (1),

6. reactor coolant pump (2),

7. pressurizer PORV inadvertently opened (1),

8. boron recovery tank ruptured (1),

9. radioactive spill resulted in unplanned release (1),

10. control rod drive problems (2)

11. shipped moisture separator tube bundles with contamination in ex-cess of limits to unlicensed metal processing center (1)

12. loss of containment control air (2), and

13. failure of refueling pool seal (1).

3.14.1 Loss of Offsite Power. Offsite power was lost seven times at Haddam Neck. Five failures involved the design of the protective re-laying system for offsite power to the plant. The loss of offsite power event are individually discussed below:

(1) Loss of offsite nower, April 27, 1968. While restoring a 115 kV outside line on April 27, 1968, a switching procedure was used which degraded a transfer trip relay, causing both offsite feeder breakers to open (A0 68-07). All offsite power to the plant was lost. The trip signals were not cancelled for some time because the operators responsi-ble for the task were occupied with the diesel generators and could not hear the telephone ringing outside the diesal room.

Upon loss of offsite power all three diesels started but failed to load and had to be shut down. The reactor was held just suberitical while the diesels were autonatically phased and loaded, and the reactor

$~3

=

was brought critical several hours later. The reactor was restarted without a test of the diesels or an adequate review of the significance of this event.

The most significant feature of this event was the failure of the utility to report the simultaneous loss of all three diesels during a loss of offsite power. No cause for the tripping of all three diesels was found. To help prevent recurrence of this event, diesel performance tests during a simulated loss of offsite power are carried out during planned shutdowns.

The loss of communication between the grid operators and auxiliary operators could have been avoided if the telephone would have had a visual signal besides the normal ring. A flashing light was placed in the diesel room to indicate incoming calls and alleviate this problem.

(2) Loss of offsite power, July 15, 1969. On July 15, 1969, the Haddam Neck reactor was shutdown during switching changes involving 115 kV offsite power (A0 69-09). Normally two 115 kV lines are available, each as a backup for the other. Both are equipped with devices that re-spond to a line fault by disconnecting the other line, which scrams the reactor and trips the turbine. Also, each device has t trip defeat switch. The switching order did not contain instructions to defeat the trip. Therefore, when one line was removed from service, the other line tripped and offsite power was lost. One diesel generator started and loaded immediately, but the others momentarily delayed loading. As the reactor coolant temperature decreased, the pressurizer level dropped. This fully opened the flow control valve on one of the charging pumps when this pump was starting. The pump was shut down by 5

4

,__ _ ,_ _..,_..._.____.,,,____a.,_,-o_.m_,. ,

its thermal-overload protection system. Shortly af ter shutdown the reactor coolant pressure rose to 2270 psi. Two electromatic relief valves opened to prevent higher pressurization. Even though these problems occurred, all plant systems functioned normally to place the I plant in hot standby condition. After 9 min the offsite lines were reistored and the plant restarted. During startup, one of the four i reactor coolant pumps had to be shut down due to a failed seal. With 4

three reactor coolant pumps in service, the plant load was increased to 425 W(e).

The plant protective relaying scheme was reviewed after the April 27, 1968, loss of offsite power. It was noted that new circuitry had been designed but not yet installed. This change would have pro-hibited loss of power by the switching errors committed. Also, all op-

erating personnel were instructed as to proper line switching procedure.

(3) Loss of offsite power. August 2 1969. During an electrical storm on August 2,1969, the relays for all four power circuit breakers

in the 345 kV switchyard, as well as breakers on other interconnecting lines, opened (A0 69-10). The reactor automatically scrammad and the turbine tripped on the loss of offsite power. All plant systems func-tioned to shut the plant down. The 345 kV switchgear control was shifted from remote to local and all tie breakers were re-closed. A lightning strike in the vicinity of the 345 kV switchyard caused offsite power to be lost. The lightning strike also caused the loss of the telephone lines that are used for transmitting relay signals, control signals, communications, etc. between the station and the switchyard.

((

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(4) Loss of offsite powe r , January 19 , 1974. An ice storm on January 19, 1974 caused a total loss of station service power at Haddam Neck (A0 74-03). One of the transmission lines providing station power tripped due to a faulted lightning arrestor on an adjacent line. The second transmission line then tripped due to improper blocking relay operation. Both diesel generators started, but one failed to run. The diesel generator's service water pump failed to start automatically due to a malfunction of the pump breaker undervoltage , device. The diesel generator's service water pump was manually started.-

(5) Loss of offsite power, June 24, 1976. While the reacror was shutdown for refueling on June 24, 1976, a total loss of 115 kV station service power occurred on three separate occasions (LER 76-14) . In-vestigations revealed that backup protective relaying from one of the lines was getting its potential signal from the wrong line due to a de-sign error. During the loss of offsite power, RHR flow was lost 'three times. The first time flow was lost for 30 s, the second and third times flow was lost for 10 s.

Problems with the protective relaying at Haddam Neck are discussed further in Sect. 3.1.6.

(6) Loss of offsite power, August 1, 1984. While the reactor was critical at 0% power a total loss of normal offsite power occurred.

This event was initiated by the inadvertent closing of a 4 kV circuit breaker during final check out steps prior to removal from its switch-gear cubicle. The operator had disconnected DC control power for the open breaker and was attempting to lift the manual "open" plunger in order to verify the "open" condition. Due to the close proximity of the

open and close plungers behind plexiglass covers and due to the large high voltage glover, the operator inadvertently actuated the close plunger. Closure of this breaker created an overload on the offsite power supplies. The subsequent voltage dip was sufficient to initiate load-shed of the nonsafeguard 4 kV buses by opening the bus tie and supply breakers, thereby disconnecting the overload and the offsite power supplies. The plant was in a total loss of offsite power condi-tion for 10 minutes.

(7) Loss of offsite power, August 24, 1984. While in the refueling mode a total loss of normal of fsite power was initiated by starting a large pump. Power was being supplied by one offsite line and station service transformer. Automatically, both diesel generators started and unnecessary loads were shed. The automatic closure of one diesel gen-erator output circuit breaker was delayed about 20 mins. Differential relay current transformer wire was found pulled from its terminal lug.

l Inrush current of starting the pump appeared as an internal transformer i fault causing isolation of the station service transformer. The wire pull occurred earlier the same day when maintenance activities were per-l

! formed in close proximicy. Also, a diesel voltage re ator was left l

slightly below the breaker voltage permissive relay s, t when it had been previously shutdown. The relay eventual closed due to vibration of resetting nearby relays and/or voltage and frequency operating varia-tions. Corrective actions include: (1) a station directive to limit access near electrical equipment panels, (2) revision of operating pro-l l

cedures to adjust diesel voltage regulator well above the permissive I

setpoint prior to shutdown, (3) inspections for other open terminations, I

(7 l

1-

(4) initiation of procedure and training enhancements, (5) initiation of permissive setpoint evaluations.

3.14.2 Unanticipated Reactor Cooldown The reactor cooldown rate was unanticipated and excessive on two occasions at Haddam Neck. The first occurred on September 22, 1967 when the reactor coolant temperature dropped 90*F in 5 min (A0 67-07).' While pulling vacuum on the main condenser, all 10 steam dump valves opened when the low vacuum trip was cleared. It was later discovered that the average temperature signal was locked at 570*F. Af ter the steam dump valves opened, steam flow increased and the reactor coolant system cooled down 90*F to 525'F in 5 min. The transient was terminated by closing the nonreturn valves in the main steam lines. The average tem-perature signal was locked in at 570*F on September 21. At the time, personnel were calibrating instruments in the reactor control and pro-taction circuit. Once the low vacuum block was cleared, steam dump was initiated due to the 570*F average temperature signal.

The second unanticipated reactor cooldown occurred on October 12, 1970, when the reactor scrammed from full power (A0 70-09). Following the scram, the turbine shut down and ten steam dump valves opened to re-lease steam directly to the turbine condenser. As pressure decreased, all but one of these valves closed. As a result, heat continued to be removed from the system, cooling the reactor 40*F in 9 min. It was determined that no detrimental effects to the reactor vessel or other pressure systems occurred.

The reactor scram was caused by a spurious low-flow signal from one loop. Components were replaced to prevent recurrence of the scram. The g*q JW

t steam dump valve failed to close because the valve-positioner-coil stop was out of adjustment. The stop was repaired and the remaining valves were inspected for the same problem and properly adjusted.

3.14.3 Loss ot Auxiliary Feedwater During plant startup on July 5,1976 both auxiliary feed pumps were

, found to be vapor bound ( A0 76-16). The reactor was shut down immedi-ately. The pumps were vented, tested satisfactorily, and returned to service. ,

This event was classified significant because of the common cause failure attributable te back-leakage from the number three steam genera-tor through a check valvo in the feed line. The faulty check valve was removed, cleaned, and replaced. No further changes to the valve were I

implemented. To prevent recurrence of this incident, temperature sens-ing devices were installed at the check valve to detect back flow into the auxiliary feedwater system. This event was originally submitted as l an LER 76-16 but was withdrawn later because it did not violate a tech-nical specification. Apparently the event was withdrawn as an LER because the reactor was only at 1% power. Operation of the auxiliary feedwater system was not required.

3.14.4 Failure of a Feedwater Regulating Valve On June 10, 1969, low feedwater pump suction pressure caused the 1A i feedwater pump to trip (A0 69-08). Since the specific cause was not l

obvious, a rapid load reduction was made in order to reduce the feed-l water requirements and restore suction pressure. The water level in steam generator No. 3 continued to rise. A =anual override of the feed-water regulator failed to reduce the rate of water level increase in the 57

steam generator. The steam generator isolation valve was given a close signal, however the valve requires 3 min to close completely. Since the water level in the steam generator was approaching a level that would result in gross carryover, the reactor and turbine were manually tripped. The eventual closure of the isolation valve prevented flooding of the steam generator and normal shutdown procedures were followed to j

place the plant in hot standby.

The transient was caused by the valve plug which broke off the stem of the feedwater reguisting valve. Since the valve was reverse seating, the plug dropped open allowing full feedwater flow to the steam gen-erator.

3 14.5 Blocked Open Blowdown Lines On December 29, 1978, an air supply valve failed, closing isolation trip valves on all four steam generator blowdown lines (LER 78-22). The operator subsequently reset and blocked open the air supply valve to re-store blowdown. This action was deemed inappropriate because it would have prevented the valves from closing in the event of a steam generator tube rupture, providing a potential release path for radioactivity. The

! valves were blocked open for 2 h. The operator did not realize the sig-nificance of blocking open the air supply valve. In this condition the plant would have had only one valve available for containment isola-tion. A revision of procedures was implemented to prevent recurrence of this event.

3.14.6 Reactor Coolant Pump Failures On August 19, 1970 (A0 70-06), a small fire was discovered and ex-tinguished at the j uncture of a reactor coolant pump and the suction c

b0

t piping. Since the reactor was. at full power, the fire fighting was done from outside the reactor-coolant-loop area. Reactor power was then re- -

duced to 65% of full power so that personnel could enter the loop area to determine the cause of the fire and make repairs. Oil had dripped from the thermocouple conduit attached to the motor thrust bearing. The permanent insular. ion on the piping immediately below the fire area had

~

recently been removed for in-service inspection and had been replaced

~

with temporary insulation that was not completely sealed to the adjacent insulation. The oil ran down an outside cover of the insulation, en-tered the joint, seeped through the insulation, contacted the 520*F coolant pipe, and vaporized. As the vapor emerged from the insulation, it spontaneously ignited. The oil leak was stopped with epoxy, and all visibly damaged insulation was replaced. Af ter cleanup of the oil in the area, the loop was placed in service, and full-power operation was resumed. Six days later, oil was discovered on the surface of the new insulation. Once more the loop was removed from service. There was no indication of any damage other than to the insulation. To prevent future occurrences drop pans were installed under all four reactor-coolant-pump motors, and checks for oil during in-service inspections are now required.7 On August 21, 1977 (12R 77-19), an operator noticed seal pressure fluctuations in reactor coolant pump (RCP) No. 2. The pump was shutdown due to indications of a seal failure. Visual inspection confirmed the f ailure and an estimated 4020 gal. had leaked from the seals. The No. I seal failed during operation and caused the No. 2 and No. 3 seals to be damaged. The full system pressure and high temperature through seal No.

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I apparently caused flashing between the No. 2 seal runner support and pump shaft. This resulted in a slight bulging of the No. 2 seal runner support sides. Westinghouse modified the runner support by drilling two holes in it to allow any pressure buildup to be relieved. All three seals were replaced. At the time of the vent, the other three RCPs were available.

3.14.7 Pressurizer PORV Inadvertently Opened The pressurizer PORV opened inadvertently on three separate occa-sions at Haddam Neck. The PORV. opened twice as a result of a spurious signal. The resulting pressure drops were 20 psi (April 3,1981) and 24 psi (February 4, 1980). On August 13, 1979, the PORV and its isolation valve opened when the bistable in the pressurizer pressure controller failed (LER 79-10). The pressure dropped 50 psi (from 2000 to 1950 psig) before the PORV could be shut. The operator overrode the signal to the isolation valve stopping the depressurization. A light in the bistable shorted causing the failure. The bistable was changed to a solid state design to improve its reliability.

3.14.8 Boron Recovery Tank Ruptured The boron recovery tank was filled solid on May 3,1969 due to an improper valve lineup (A0 69-06). The tank became pressurized and cracked at the heat seam. Approximately 200 gal. of distillate was re-leased to .the surrounding yard area before the valve line-up was cor-rected. A total of 20,000 gal. of fresh water was used in flushing the area. The run-off collected in a yard storm drain which drains to in the plant discharge canal.

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3 14.9 Spill Resulted in Unplanned Release on May 5, 1969, 500 gal. of radioactive liquid waste was discharged from the boron recovery evaporator (A0 69-07). The borated water drained into the floor drains and emptied into the aerated drain tanks. However, due to a broken nipple on the bottoms pump, some borated water was discharged onto the floor. The discharge was not noticed until 2 h later due to the small size of the gage line and the low discharge pressure of the bottoms pump. Some flashing occurred and caused steam vapor in the boron recovery area. Also, a portion of the l

concentrate cooled suf ficiently to solidify before reaching the floor drains. After the area cooled, decontamination efforts began. All solidified boric acid was collected and drumed. The surrounding area was flushed with water, draining to the aerated drain tanks. Radiation levels in the recovery area were 10-15 mrem. The estimated release of I-131 from the vapor was 1.85 x 10 2 C1. The estimated release of tritium was 2 36 C1.

3.14.10 Control Rod Drives Zero power physics tests on March 15, 1983, (PNO-1-83-2C) discov-ered that four Control Rod Drive Shafts and Rod Control Cluster As-i semblies (CRDS-RCCA) assemblies were unlatched. Investigation showed that the CRDSs had been improperly latched to the RCCAs during reactor reassembly. The coupling fingers on the four CRDSs were plastically de- ,*

formed (spread), which greatly increased the chance of installing the CRDS with one coupling finger outside of the RCCA hub. This would occur if the CRDS is oriented during installation so that the coupling fingers o3

are positioned between the guide sheaths during insertion into the RCCA hub.

All CRDSs were removed from the reactor and inspected. The coupl-ings in the four improperly latched CRDSs were rebuilt. The coupling in a fif th CRDS was rebuilt because of latching difficulties that had been experienced during inspection. CRDSs will be inspected for deformed coupling fingers during future refuelings to ensure proper operation.

Three of the RCCAs that were paired with three of the faulty CRDSs 1

were replaced; the fourth RCCA was acceptable for reuse. Five addi-tional RCCA hubs were inspected and determined to be acceptable for for continued operation.

The CRDSs were reinstalled using a revised installation procedure that emphasizes proper orientation of the CRDS to the RCCA while latch-ing. The Westinghouse Electric Corporation provided appropriate input to these procedures.

The connect / disconnect CRDS buttons were examined for proper height to confirm correct installation and latching. All CRDS were measured for correct height. RCCA drag tests were performed that verified proper latching and alignment.

On August 25, 1978, the use of an FM radio in the control room caused a dropped rod / rod stop alarm (LER 78-18). The alarm should have initiated a turbine load runback followed by a matching reactor power reduction as well as an automatic cod withdrawal stop signal. The auto-matic rod withdrawal stop signal actuated but a turbine load runback reactor power reduction was not observed. Investigation revealed a closed pressure switch isolation va'.ve which disabled the turbine load l

runback feature. The procedures were revised to include this valve on a i l checklist to assure that it is returned to its normal position upon com-l pletion of maintenance and/or calibration. '

The reactor core is protected by redundant signals. One is the I

turbine load runback and the other is the automatic rod withdrawal stop signal. The automatic rod withdrawal stop signal actuated. This would have protected the reactor core if an actual dropped rod had occurred.

3 14.11 Contaminated Moisture Separator Tube Bundles Released From Site On April 22, 1981, two moisture separator reheater tube bundles were shipped from Haddam Neck to a metals waste processor. The bundles consisted of 43,000 pounds of copper-nickel alloy material with an ac-i civity of 10 2 C1. The material was off-site for 18 h. The amount of activity was in excess of the exempt quantities in 10 CFR Schedule B.

Additionally, the metal waste processor did not have a license to re-ceive these materials.

When it was noticed that the tube bundles had lef t the site, the metal processing company was notified and CYAP00 health physics person-nel were dispatched to recover the material. Procedures were revised to include plant management review and any necessary health physics review and authorization prior to the release of all materials that could con-1

~

tain or be contaminated with radioactive materials.

3 14.12 Loss of Containment Control Air The first loss of containment control air occurred on November 1, 1983 (LER 83-20). The result was a loss of control of the pressurizer spray valves and pressurizer power operated relief valves. Maintenance

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had installed an incorrect o-ring on a control air filter canister. The filter canister was improperly installed which led to the loss of con-trol air. The correct o-ring was installed and the filter returned to service. ,

The second loss of containment control air occurred on November 28, 1983. Control of the pressurizer spray valves and power operated relief valves was again lost. A filter canister had failed due to worn threads on the filter cap. A new type of filter canister was used to replace the broken canister.

3 14.13 Failure of Refueling Pool Seal On August 21, 1984, (LER 84-013) the refueling pool seal failed while the reactor refueling cavity was flooded in preparation for re-fueling. This failure resulted in the draining of approximately 200,000 gallons of borated water from the refueling cavity to the reactor vessel flange in about 20 minutes.

Water flooded the lower levels of the Containment Building. At time of the event, the reactor vessel head had been moved to its lay down area, the spent fuel pool in the Fuel Building was isolated from the refueling cavity, the reactor vessel upper internals were still in place, and no movement of fuel was being conducted.

l The cause of the failure was an improper seal design. Corrective actions include: (1) installation of a redesigned seal, (2) installa-tion of a cofferdam in the fuel transfer canal, and (3) development of i emergency operating procedures.

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1 3.15 Trends and Safety Implications of Forced Shutdowns and Power Reductions I i

l Data on both the forced shutdowns and the power reductions was an-

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alyzed with the obj ective of determining any trends on patterns. In general, the data tends to show a time trend towards incremental in-provement each year. However, the data also evidences three distinct perturbations, clearly displayed in the interpreted forced shutdown fre-quency plot presented in Fig. 3.3A. Although examination of the shut-down data itself did not lead to any direct cause or causes for each perturbation, other generalized causes may be related.

The first perturbation occurred around 1973 and 1974. The fre-quency of forced shutdowns increased to twf :e the rate of the previous two years, counter to the trend established 1968 through 1973.* In 1975 and 1976 the rate dropped again to follow the previously established trend. At the same time that this observed perturbation occurred the plant experienced extensive problems with the turbine.

Concurrently, the frequency of forced shutdowns occurring as a re-sult of equipment failures was showing an excellent trend in improved performance, falling from 10 events in 1968 to 1 event in 1973. In 1974 the f requency jumped back up to 6 and then a f avourable trend was re-established. This data is presented in Fig. 3.11, and Fig. 3.11 A shows the interpreted data which suggests a saw-tooth pattern.

A second perturbation can be seen on the shutdown frequency curve occurring about 1977 through 1978. As with the 1974 data the frequency doubled over a period of 2 years and then returned to a rate that can be projected from the initial trend. Relational causes are not as clear in S7

o-SHUTDOWNS DUE TO EQUIPMENT FAILURES 10_ ,

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67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.11 t

INTERPRETED SHUTDOWN RATE WITH EQUIPMENT FAILURE AS CAUSE 10 - -

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YEAR OF OPERATION FIGURE 3.11A

this case, although the plant had just completed an extended uninter-rupted run of 343 days in August 1977. Subsequent to this run a number of BOP problems occurred and then reoccurred, suggesting difficulty in effecting problem solutions. Again, examining the shutdown due to equipment failure frequencies shown in Fig. 3.11A there is seen to be a minor perturbation effect from this cause.

A third perturbation is seen around 1981 and 1982, when the shut-down frequency increased by a factor of two before falling back to a projectable rate. In this time period a number of efforts were being applied in response to the TMI effort. Again the saw tooth pattern in the equipment failure induced shutdown rate may be seen, although in 4

this case with a much enhanced recovery rate.

Other tests for trends or patterns in the shutdown data, such as the DBE turbine trip frequency, or in the average duration of shutdown either gave random results or further emphasized the existence of the perturbations.

The data on reported power reductions generated a completely dif-ferent pattern. This annual rate plots to two distinct " humps" with a zero rate occurring in the years 1974 through 1977. No discernable correlations were found from examination of the interpreted power reduc-I tion rate plot shown in Fig. 3.4A.

The data showed that causes of downtime were principally for main-

tenance and testing, occurring mostly in the balance of plant systems including the turbine-generator system, main steam systems, and electric power systems.

ld

l These findings may be interpreted as follows:

1. The overall performance of the plant is continuing to improve.

2. Major " refit" efforts result in perturbations in the plants' over-all performance.

3. Recovery from refits may extend one or two years, as a result of subsequent increase in equipment failures.

4. A perturbation as a result of a refit may not necessarily be evi-denced by problems directly related to that refit. Thus, the more.

important negative effects of the refit may lie in reduced at-tention by plant personnel to areas not affected by it. It should be noted, however, that the shutdown data attention to nuclear safety systems.

5. Salance of plant systems were the major causes of interruptions to plant operation, mainly for repair and maintenance.

3.16 Trends and Safety Implications of Reportable Events As an additional step in the overall evaluation process, the re-portable events were examined to detect discernible trends that might indicate potential safety problems. Figure 3 12 presents the yearly 4

totals of significant events at Haddam Neck. The plot does not show any discernable trends. Figure 3.13 shows the yearly totals of condi-tionally significant events while Figure 3.14 shows the yearly totals of both significant and conditionally significant events. Both of these plots show a large number of events during the early years, 1967 through 1969 followed by a significant decrease during 1970 to 1972 due to the increase in experience in operating the plant. Peaks in the number of i 71

YEARLY TOTALS OF SIGNIFICANT EVENTS 5, 5 N

U 11 4_

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0 0 0 0 l 0 " 0 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.12

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• m 1EARLY TOTALS OF CONDITIONALLY SIGNIFICANT EVENTS 25_

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g 8 e 4 YEARLY TOTALS OF SIGNIFICANT AND CONDITIONALLY SIGNIFICANT EVENTS 25_

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67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.14 k

events can be seen in 1973, 1978, 1980, and 1984. Specific trends and problems were identified in the following safety-related areas:

(1) offsite power, (2) control rod drives, and (3) charging pumps.

These are discussed below:

(1) Offsite power. Offsite power was lost seven times to the Haddam Neck plant. Figure 3.15 shows the time distribution of these events. No real trend was discerned from this plot. Five failures in-volved the design of protective relaying system for offsite power co the plant. The first two failures (A0 68-07, A0 69-09) were caused by oper~

ators f ailing to block out protective relaying during switching proced-ures. The protective relaying functioned as designed and isolated the plant from the grid. The next two trips (A0 69-10, A0 74-03) were I

caused by adverse weather, that is, lightning and ice storms. Again, the protective relaying functioned as designed but should not have cut off station service power. During the latter loss of offsite power (A0 l

74-03), one of the service water pumps failed to start automatically and had to be manually started. Failure of tne pump to . start automatically was caused by a faulty relay. The fifth loss of offsite power (LER 76-14) occurred during refueling when it was discovered that the protec-tion for one transmission line was getting its power from the other line. This caused protective relaying to trip out one line due to the fault on another. Protective relaying was improved during July 1976 ending this series of events.

l Off site power was los t twice during August 1984. Both events are attributed to personnel errors in the switchgear room. The plant will construct a new switchgear room during the 1989 refueling outage.

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YEARLY TOTALS OF LOSSES OF OFFSITE POWER 5_ ,

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1 1 1 S  !: i II li i 0 0  !! 0 0 0 0 0 i0 0 0 0 0 0 0 l -

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.l YEAR OF OPERATION FIGURE 3.15 i s' 1

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(2) _ Control rod drives. Fif teen events (in Fig. 3.16) involved the control rod drive system. Rods dropped into the core on seven sepa-rate occasions (A0 68-07, A0 68-15, A0 69-11, A0 69-18, A0 70-01, A0 70-03, and LER 80-16). Dropped rods are a safety consideration since the consequences are flux depressions in the core. All but one. of the dropped rods occurred before 1971. No rods were dropped during 1971 A m fo net r/ m tesoefost O rpf.

through 1979. Faulty relays caused five of the rod drops. A Causes were not reported for the remaining events.

Control rods became inoperable on four (A0 68-22, A0 69-05, A0 69-13, and A0 69-14) occasions. Inoperable rods are a serious safety concern since control rod movement is responsible for reactor control.

All instances of control rod inoperability occurred within the first 2-years of operation.

One failure (LER 77-27) involved the separation of a rod cluster control vane. The cause was a faulty bias joint. PNO-1-83-20 reported improper latching between contror rod drive shafts and rod control cluster assemblies. The coupling fingers had plastica 11y deformed. The remaining failure events involved a failure to withdraw rods during a test (LER 83-25) due to a switch failure and a failure of a control rod drive slave cycle.

(3) Charging pumps. Fourteen failures of the charging pumps at Haddam Neck were reported over the operating history reviewed (Fig.

3.17). The plant has two centrifugal charging pumps which are used in the chemical and volume control system to charge the coolant loops dur-ing startup and transient conditions. During loss of coolant accidents the pumps serve as part of a high pressure coolant injection system.

77

YEARLY TOTALS OF CONTROL ROD DRIVE EVENTS 5, 5 N

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0 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OEERATION FIGURE 3.16 78

YEARLY TOTALS OF CHARGING PUMP EVENTS 5_

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o oo oo o oooo  !! eo !L o 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 YEAR OF OPERATION FIGURE 3.17

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. s The charging pumps also provide water to reactor coolant pump seals. In addition to the two centrifugal pumps, there is a positive displacement paens//y no ytu/n~* sd pump *' the system.

On January 24, 1976 a charging pump outboard seal leaked due to an ,

o-ring failure (LER 76-05). The o-ring was replaced.

Seven failures occurred from Apri l 1977 to May 1978 (see Fig. 3.1%, a period of little over a year. On April 4, 1977 the 1A charging pump failed due to cracks on the pump shaf e (LER 77-05). The

. shaft was replaced. The cracks were caused by the thrust collar rot being perfectly square on the shaf t.

On April 26, 1977 a small weep developed on the 1A charging pump seal housing (LER 77-06). The housing was replaced. Two days later the pump was placed back in operation when it failed again (LER 77-07).

This failure was caused by misalignment between pump and motor during replacement.

Later that year, the IB chargitig pump bypass isolation valve failed and was replaced (LER 77-16). On May 4, 1978, the same IB pump valve failed and was replaced (LER 78-07) only to fail again four days later (LER 78-08). The valves were being eroded by high velocity water flow. The valve was replaced with another type and the problems dis-appeared.

On May 31, 1978, the IB charging pump's pressure gauge isolation valve leaked near a weld (LER 78-12). The weld was repaired and the pump placed back in service. This failure brought to an end the first set of events.

O

Two failures occurred in 1981. Both failures were due to excessive vibration. On July 4, 1981, two nipples on the charging pump cracked and leaked oil (LER 81-10). The leak was on the charging pump oil lubricating system which is not capable of being isolated. Charging ,

pump vibration levels became excessive on November 19, 1981 (LER 81-19). The vibration was caused by a worn key and worn parts of the thrust bearing.

6ne failure occurred in 1982 (LER 82-09). A pinhole leak occurred on October 15, 1982 in a pipe joint next to a charging pump throttle valve. This leak resulted in a small release of radioactive coolant to the auxiliary building. The pipe joint was repaired.

Three failures occurred in 1983. In one event an operator was unable to rotate charging pump 1A by hand (LER 83-09). Subsequently, solidified boric acid was removed from the pump. In another event a charging pump was declared inoperable (LE2 83-16) due to broken wires on a motor bearing thermocouple. The*last failure (LER 83-18) involved the IB charging pump which had a leaky outboard pump seal. The seal was re-placed.

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4

4. CONCLUSIONS 4.1 Overall Plant Operating Experience The Haddam Neck Plant has been generally operated in a safe and

! orderly manner. However the operating history reveals that there are a few areas with demonstrated deficiencies or indications of conditions relating to safety concerns. There are no indications in the operating history that there are extended or repetitious problems that have re-suited in any significant safety consequence.

A visit to the plant confirmed inferences derived from the opera-tional review data that attention to safety is an operating character-l 1stic. A tour of the plant left an impression of a clean and well main-tained operation. There was evidence of positive approaches to imple-menting lessons learned from experience.

One indicator of the quality of performance of plant operations is the ability of a plant to achieve and maintain high levels of reli-ability in operation. This also results in a consequential direct bena-fit to safety. Haddas Neck has a history of higher than average avail-ability and capacity factors. Twice the plant has accomplished extended l uninterrupted operating runs, the first for a period of 343 days ending i in August 1977 which established a record, and the second for 401 days ending in August 1984.

Adverse effects of the extended runs were not conclusively evident a

! from analysis of the data. While there was an observable increase in the number of maintenance-generated shutdowns subsequent to the 1977 run, the same characteristic was not observed in the final months of I

i 1

l 1

- ,..--, . -...n,,-,-..-,_,_-,_..-,_,-,,..~........-,.....-.._.,,,..s,.-__,

- +,..n.,- ** ~+--st~~n-- -,-g----------

0 1

1984. However, the race of submittal of LERs af ter each run did in-crease, suggesting an increase in maintenance activities.

Apart from problems in the areas of off-site power, charging pumps, and control rod drives, in general any problems involving nuclear safety a

systems or equipment were found and resolved without the need of a con-4 sequent plant shutdown. Sixty-six percent of all shutdowns involving

! reportable events occurred before 1970, which supports the conclusion that there is an operational focus on preventing nuclear systems l problems from resulting in forced shutdowns.

l

, 4.2 Trends and Symptoms f

l

]

Trends were discerned by examination of the data, and were more 1

evident in the shutdown data than in the reportable event data. It was concluded that the forced shutdown data showed a definite asymptotic trend of improvement. However, this conclusion was dependent upon the i i premise that certain perturbations to the projected trend can be readily isolated, and correlated to an identifiable set of causes. Accordingly, i.

it was determined that it would be necessary to conduct a broader examination of the operational history. However, the scope of the t

l effort and the resources available became limiting factors which con-I strained the depth to which this examination could be conducted.

It was concluded that the causes of perturbations which occurred in 1974, 1978, and 1981 may be related to (1) extensive turbine problems, f (2) an extended operating run in 1977, and (3) efforts in response to TMI, respectively. Actual symptoms of these perturbations were not di-rectly attributed to their theorized causes because of limitations in JD

scope and resources. No direct correlation with reportable event fre-quency was detected. As another conclusion, it might be anticipated that the effects of major efforts applied to the repair or maintenance of plant components in one area may be evidenced in different and unex-pscted areas.

In summary, it was concluded from analysis of the trends that:

1. After an initial break-in period the overall performance of the plant was continuously improving with the exception of identified perturbations.

2. Major "refic" efforts resulted in perturbations to the plants' overall performance.

3. Recovery from refits extended to one or two years, and probably because of consequent increases in equipment failures.

4. A purturbation as a result of a refit may not necessarily be evi-danced by problems directly related to that refit. Thus, the more important negative effects of refits may be caused by other reasons such as reduced attention by plant personnel to other areas not affected by the refits. It should be noted, however, that the shutdown data did not provida any strong indications that plant personnel relaxed their attention to nuclear safety systems.

Three important symptoms were identified through examination of the

,i reportable event data. These were determined from analysis of the significant events.

t.oss of of fsite power D

l During the early history of the plants the significant events were f dominated by the occurrence of five loss-of-offsite-power events. These 5

o ,

were all attributed to problems with the design of the plant protective relaying. Subsequently corrective measures to the protective relays were implemented. However, there was a later recurrence of this problem and additional loss-of-offsite-power events occurred later on during the 1980s. Thus it was concluded that this constitutes an outstanding symp-ton of a problem which was sa'fety significant. Loss of offsite power is a precursor to station blackout, which is a safety issue. Haddam Neck sustained one actual station blackout during the period under review.

Control rod drive failures A significant number of control rod drop events were experienced during the first three years of operation. These were determined to have been principally caused by faulty relays. Corrective measures were applied and there was no further recurrence for 6 years. However, as the plant grew older, control rod drive failures started to reoccur, for causes including separation of rod cluster control vanes, plastic deformation of coupling fingers, and other random equipment failures.

While no direct correlation to aging was made, it was concluded that this may be a factor in the perpetuation of control rod drive prob-l l

1 ems. The control rods are the principal mechanisms for reactivity con-trol and shutdown of the reactor, and thus serve an important safety function. Failure of the drive mechanisms to operate on improper opera-tion are safety concerns. Thus it was concluded that these failures of the control rod drives constitute an outstanding symptoms of a problem which has safety significance.

?S

i Charging pump failures The charging pumps became significantly undependable in the latter half of the review period. For the first 9 years no problems were re-  ;

4 ported. However problems started to occur in 1976 and continued through the later half of the review period, again suggestive that aging may be a factor. Equipment failures were reported to have been caused typ-

~

ically by wear induced vibrat1on, erosion, chemical solidification, and 4

leaks. The changing pumps serve to inj ect chemical additions to the reactor coolant system and also to provide backup to high pressure coolant injection should this be required subsequent to an accident, both of which are safety functions. It was concluded that recurrence of operability problems in the charging pumps constitutes an outstanding

symptom of a problem which has safety significance.

4 Other symptoms j

It was concluded that causes of downtime were principally for main-j tenance and testing, occurring mostly in balance of plant systems in-ciuding the turbine generator system, main steam systems, and electric power systems. In general nuclear systems did not contribute auch to interruptions to plant operations.

i i

I l

i i

l

APPENDIX A REVIEW OF FORCED SHUTDOWNS AND POWER REDUCTIONS This appendix presents the Heddam Neck data on shutdowns and power reductions in narrative and tabular format and lists the information sources. It describes how the information was encoded and how sig-nificance screening criteria were applied.

A.1 Scope

, Da ta collected in this review includes inf orma tion about each

forced shutdown and power reduction that occurred between 1970 and

, 1984. The data is presented in narrative Yearly Summaries (Section A.4) and Tabular Summaries (Table A.1). Forced shutdowns result from equip-i ment failures that present an abnormal challenge to the unit's opera-t tion. Scheduled shutdowns for refueling and maintenance were not in-cluded. However, if the utility scheduled a refueling or maintenance

outage to coincide with a shutdown that resulted from an abnormal event, that shutdown was included even though the utility reported it as sched-1

uled. That portion of the outage time caused by the abnormal event was attributed to the shutdown and included in the compilations.

Information on power reductions provides additional de tail to a previous or subsequent shutdown, or indicates a safety significant trend. The power reductions were included in the proper chronological l

l sequence with the shutdowns in the data tables for the forced shutdowns and power reductions (Table A.1).

37

A.2 De ta Sources The review of the forced shutdowns and power reductions included data from clm following sources.

1. Nuotear Power Plant Operating Experience for 19XX, for the years 197 H 984.

2. NUREG-0020 series (Gray Books).20 ,

3. Annual or semiannual reports from the time of startup through ,

1977. For 1977 through 1984, monthly operating reports were used because the utility was no longer required to file annual re-ports. The review of power reductions used primarily the annual,

, semiannual, and monthly reports.

When LERs describing shutdowns and power reductions were available, their event descriptions gave additional informa tion and helped to sup-port significance screening.

A.3 Significance Screening _

Shutdowns and power reductions were evaluated against design basis events (DBEs) found in Chap. 15 of the Scandaztf Revisu Plan 2 (Table A.5). DBEs are those postulated disturbances in process vari-ables or failures and malfunctions of equipment that plants are designed

, to kithstand. Licensees issue the results of their analyses of these

, events in safety analysis reports.

Generic design-basis initiating events such as " Increase in Hea t Removal by the Secondary System" or " Decrease in Reactor Coolant System Flow Ra te ," were used as primary flags for reviewing the forced shut-downs (and power reductions). Once the generic type of event was 55

iden tified, the particular initiating event was determined from the de-tails associated with the shutdown. For example, if the reactor shuts down as a result of an increase in heat removal because a feedwater reg-ulator valve failed open, the event falls into the category of generic Type 1 DBEs. Based on the specific initia ting event (valve failed open), the event is classified as a 1.2 DBE "Feedwater System Malfunc-tion that Results in an Increase in Feedwater Flow." Some shutdowns were readily identifiable as specific DBEs, such as the tripping of a main coolant pump, classified as a 3.1 DBE. Once categorized as a DBE, the shutdown was considered significant regardless of the resulting effect on the plant (because a DBE had been initiated) .

Loss of flow from one feedwater loop was considered sufficient to qualify as a 2.7 DBE " loss of normal feedwater flow." The closure of a main steam isolation valve in one loop was considered sufficient to qualify as a 2.4 DBE " inadvertent closure of main steam isolation valves."

Those shutdowns that were not classified DBEs were assigned to NOAC defined categories (Table A.6) inorder to provide more information on i

che failure or error associated wich the shutdown. Wich these ca te- 1

gories, more specific types of errors and failures could be examined, 1

I through the tabular stannaries , to focus the reviewer's attention on l

probles areas (safety related or not) that were not revealed through the 1

i DBE categories.

Table A.1 of this Appendix provides a tabular summary of informa-tion about forced shutdowns and forced power reductions at Haddam

! Neck. More information is available about these shutdowns tha t were i

1

- __ _, _ _ _ _ _ _ _ _ . . - _ _ . _ _ _ - , , , . . _ _ . . _ . _ _ . _ _ _ . ___...-_,.,____7.__-._. . . - _ _ _ _ , - . _ - _ . _ - _ _ _ . _ . . . , .

~

also reportable even ts; in such instances, this additional information may be found in Appendix B.

A.4 Yearly Summaries for Heddam Neck A discussion of the shutdowns and power the reductions occurrir.g in each of the years,1967 through 1984, follows:

1967 The Haddam Neck Plant went critical on July 24, 1967. Electrical output of 50 W commenced on August 1. By December the plant reached full power output of 450 W(e) . Nineteen forced shutdowns and one power reduction occurred in the first few months of operation. Five of the shutdowns were necessita ted by maintenance on the turbine control valves, which accounted for 17% of the total downtime in 1967. Problems with power distribution equipment caused nine shutdowns. These repairs accounted for only 10% of the downtime for the year.

Reactor trip systems caused two forced shutdowns, totalling 32 h down time. Repairs to the reactor coolant systems caused three forced shutdowns, accounting for 1184 h of downtime, 1008 h of which occurred on August 21 to repair leaking pressurizar relief valves.

The only forced power reduction that year occurred on October 30 to inspect turbine control valves. Power was reduced for 81 h to perform this inspection.

1968 The greatest number of forced shutdowns (30) for any one year oc-curred in 1968. Thirteen of these involved reportable events. Problems continued with the turbine control system. The plant shut down eight times for a total of 740 h to resolve these problems.

N

a .

Repairs to steam generators required 454 h of downtime in a single outage beginning on April 1. The plant shut down for 272 h on March 1 to replace the reheater drain tank and to repair the moisture separator.

Three power reductions occurred this year (1) for repairs to:

(1) a feed pump discharge valve; (2) a feedwa ter flow orifice, and 1

(3) one of the two transmission lines.

1969 Twenty power reductions and eight forced shutdown occurred in 1969. The plant shut down six times for a total of 513 h for repairs to I the turbine control system. Repairs on the turbine control system ac-counted for about 40% of the total downtime for the year. On April 11 the plant shut down for 456 h for t.urbine valve repairs and modifica-tions. The only other lengthy shutdown occurred on May 1, when the main generator was repaired (258 h).

Power was reduced eight times for total of 177 h to perform main-tenance on the 345 kV transmission lines.

1970 Fifteen forced shutdown for a total of 416 h occurred in 1970.

Over 25 % (109 h) of the downtime was incurred to repair damage stemming from a first on the insulation of one of the four reactor coolant loops. The fire resulted from lubricating oil spilling onto hot piping insulation. This event generated abnormal occurrence A0-70-9-8 and is further discussed in Sect. 3.14.6.

Five of the forced shutdowns were due to problems with the steam generators and with the feedwater control sys tem. The longest shutdown of the year occurred on September 10 when the plant shut down for 65 h to replace the packing glands in the reactor coolant pucips .

.I

Five forced power reductions occurred, two in Augus t , to replace insulation damaged in the oil fire. The remaining power reductions were required for repairs to a circulating water pump, for replacing a ses-tion service transformer, and for repairs to a feedwater control valve.

1971 Nine forced shutdowns and eight power reductions occurred in 1971. The main condenser caused three shutdowns early in the year.

Main condenser problems continued through 1972 and 1973, requiring plug-ging of many tubes. Feedwater control problems forced the plant down on three occasions for a total of 45 h. Feedwater control problems are discussed in more detail in Sect. 3.14,4 .

On June 7, the plant shut down for 52 h to investigate the source of grounding of the 4.16 kV buses. On October 23, repairs to the steam generator feed ptanp check valve required 47 h of downtime for the plant.

The eight power reductions occurring in 1971 were relatively brief. Most of them were caused by problems with one of the condensers.

1972 .

Haddam Neck shut down seven times in 1972 for a total of 279 h.

Three maj or outages occurred during the year. On September 21, the plant went down for 85 h to repair the main trans f ormer. On November 23, the plant shut dom for 73 h to repair the turbine control system.

On July 14, repairs on the leaking primary and secondary safety valves

consumed 63 h of downtime. Several short outages occurred late in the year to plug condenser tubes.

Nine forced power reductions occurred during 1972. Five of them occurred to plug tubes in a leaking condenser, the other four were 31

required to replace a reheater drains tank and to repair a steam genera-i tor feedwater pump.

1973 Seven outages occurred in 1973, resulting in 4338 h downtime for the plant. The plant shut dom for five months, starting July 8, to re-solve turbine vibration problems. The 3828 h outage was the longest in the history of che plant. Both lower pressure turbines required exten-sive modifications. Due to the lengthy maintenance outage, refueling

was done during that shutdom period. This did not lengthen the outage time , however. Just prior to the major turbine repair outage, the plant shut dom for 366 h to replace an entire row of blades on one of the low

pressure turbines.

Five power reductions occurred this year to plug more tubes in the condenser and to repair the reheater drains tank.

1974 Fourteen forced shutdowns occurred in 1974, causing a loss of 771 i operating hours. Turbine problems accounted for 90% of the total down time (695 h) accurred over seven outages. One of these outages occurred l on March 23, when the plant shut dom for 660 h to repair broken turbine blades.

1975 Only five forced shutdowns occurred for a total of 70 h downtime in 1975. The longest outage of the year (39 h), besides refueling, oc-

, curred on March 26 when the packing gland on the letdown system stop i

valve had to be replaced due to excess leakage.

i3

1976 The high plant availability established in 1975 continued into 1976, with only 80 h lost to forced outages. Almost half of the down-time was incurred by repairs to a broken steam baffle in one of the moisture separators. A 17 h outage occurred on January 22 when several instrumenta tion sensing lines froze causing the reactor to trip (LER 76-1).

1977 Twelve forced shutdowns resulting in 289 h of downtime occurred in 1977. No outages occurred until August 19 when the plant shut down for 70 h to repair a leaky feedwa ter heater and to perform general mainten-ance. The only other lengthy outages of the year occurred in le.te December. The plant shut down five times for a total of 83 h for tur-bine balancing during that month.

1978 Fourteen forced shutdowns occurred in 1978 for a total downtime of 332 h. About 90% of the downtime resulted from malfunctions in three systems: (1) the moisture separators; (2) steam generator feedwater system, and (3) the onsite electric AC power system.

Three major outages, totaling 119 h, were needed to effect repairs to the moisture separa tors and reheaters. Most of this time was re-quired for the rewelding of baffle places on the separatoro. Repairs eroded drain piping on the moisture separator reheaters were also com-placed.

The steam generator feedwater system repairs required two shutdowns f or a total of 91 h of down tice. Both outages were needed for repairs to the B steam generator feed pump.

Y

The longest outage of the year occurred on June 17 when the plant shut down for 66 h to install heat shrunk sleeves on the electrical pen-etrations. A 23 h outage was required on March 24 to replace the elec-trical terminal block in the containment.

One power reduction occurred in 1978 for repairs to a steam genera-tor feed pumps.

1979 Four shutdowns resulted in 268 h downtime in 1979. On September 29 the plant shut dem for 234 h to inspect the welds on the steam genera-tor feed line nozzles. The remaining outages were relatively brief and were insignificant with respect to plant operations.

Five power reductions for a total of 243 h occurred in 1979. Ra-pairs on a leaking reactor coolant pump seal accounted for 168 h at re-duced power hours.

1980 In 1980, there were six outages resulting in a total of 188 h of down time. The plant shut dom for 44 h in February to plug tubes in' main condenser unter boxes. Turbine balancing accounted for 65 h of downtime starting on September 27.

No other outages exceeded 20 h in dura clon.

Power was reduced three times for a total of 64 h in 1980. Two re-ductions were required to pitg leaking condenter tubes, and the third was required to repair a steam generator feed ptanp.

95

1981 In 1981, there were 11 forced shutdowns and 3 power reductions re-sulting in a total of 248 h of downtime. On December 11, while reducing power for turbine maia tenance and the plugging of tubes in 'B' con-denser, erratic operation of the turbine control valves caused a turbine overspeed trip. The unit remained shut down for 82 h. the reactor was manually scrammed on August 23 in order to repair a valve packing leak which required 47 h downtime before the unit could be returned to ser-vice. On July 29, while reducing power, a low pressure steam dump mal-function caused by OPC relay 63, resulted in a reactor and turbine trip and a down time of 39 h. The reactor scrammed during APRM functional testing on October 1, requiring 31 h of downtime.

On February 27, the plant tripped on a loss of feedwater control due to a loose control air header fitting. On December 22, the reactor and turbine tripped on high pressure heater drain tank level as a result of a failed fitting on the control air system. On Sep tember 10, a spurious closure of the right hand turbine stop valve resulted in a plant trip.

The other outages and power redoctions, none of which incurred more than 5 h downtime, involved a steam generator low level alarm, flooding of a feedwa ter heater, a steam leak on MOV36, a transfer of reactor coolant pump suction to the station service water, and hood reductions due to valve steam leakage resulting in three loop operation, off-gas activities, and condenser tube plugging.

5

1982 Seven forced shutdowns for a coul of 407.5 h occurred in 1982. Of these seven, five occurred as results of failures in the turbine genera-tor systems and in the electrical power systems accounting for 395.4 h of downtime. Only 12.1 h forced shutdown was caused by problems in the reactor systems.

The turbine and reactor tripped twice as a result of short circuit problems in the generator exciter for 150.7 hs down time. Moisture in the main transformer oil, occurring as a result of leaks, caused another 195.1 h downtime. The pin holding the disc to the stem in a turbine trip valve sheared, so tha t the valve would not open. 39.2 h was re-quired for repairs.

Two events, one involving operation error and one involving a rod drop for unknown reason, occurred in the reactor systems. While the rod drop event was classified as a design basis event, neither event caused any significant result.

Power was reduced 8 times, mostly to repair leaks in the condenser tubes. Also repairs were conducted on a vibrating steam generator pump, and a steam generator leak. A monthly turbine stop valve test was con-ducted.

1983 A total of only 33.6 h of downtime was incurred as a result of four forced shutdowns in 1983. Of these, 9.9 h was required in one shutdown for an overspeed turbine trip test. The other three shutdowns were variously caused an inadvertent loss of power supply as a result of man-ually resulting control switches, spurious instrument cutput and a feed regula ting valve operation. No significant results were experienced.

97

One load reduction was required to repair leaks on a steam genera-tion level indicator.

1984 Again, in 1984 there were only four forced shutdowns, resulting in 86.0 h of downtime 79.0 h accumulated in three shutdowns caused by prob-less in main generator hydrogen system. Two other problems resulting in shutdows were a cycling voltage regulator and flow control of a reactor coolant pump.

One power reduction was required to repair leaks in the feedwater system.

Table A1 1967 Forc2d Sh:tdsens and Power Rodrctions far Heddan Meck DBE (D) /

N SIC (N)

Duration Power Reportable Shutd own System Com ponent Event Dato (lit s) (1) Event Description Cause Method Involv ed Involved Category 7/25/1967 NA A0-67-1 Inadvertent actuation of D 3 ED RELAYI N5.1 l over-voltage relay in the rod power supply j 7/20/1967 NA A0-67-2 Reactor trip due to a spike la the A 3 IA INSTRO N2.4 intermediate range start-up circuit 8/07/1967 174 9 AO-67-3 Replacement of penis on all reactor B 1 CD PURPI N 1.1. 4 coolant pumps, repairs to main steam lian leaks and turbine control valves 8/15/1967 2 9 Turbine trip due to overcurrent A 3 Eb RELAYI D2.3 protective relaying on 309 service station transformer 5/22/1805 2 7 Slippage of asia steam isolation A 3 HB INSTRU N1.1.4 valve off the upper limit switch energizing the main steam line trip valve closure circuit 8/17/1967 6 9 A0-67-4 Accidential gronading of the vital B 3 EB ELECON N5.2 bus during calibration of reactor control system 8/19/1967 11 9 Lightning striking 345 Kr line in H ,3 EA ELECON E 9. 2 s witchyard 8/19/1967 0 7 Excessive vibration at generator B 3 HA GENE 3A N1.1.4 exciter 8/21/1967 1000 22 Repair leaking pressurizer safety B 1 CA VALVEI N 1.1. 4 valves and spray valves 10/02/1967 15 9 Repair bloma gasket on left hand D 1 RA PIPEII N1.1.4 turbino stop valva pressure equalizing line 10/05/1967 130 22 Ao-67-8 No. 1 vital bus inverter power A 3 ED INSTRU N1.1.4 supply to all primary giant instrumentstion was in terrupted 10/10/1967 32 Ao-67-9 Turbine control valve failed to A 3 IA VALVEI N1.1.4 a close 10/13/1967 54 7 A0-67-10 Dlown fuse in power su gply f rom No. A 3 Ep INSTRU N1.1.4 1 vital bus inverter -

kk -

Table A1 1967 Forc:d Shutd2tn3 a:d Powtr Radictic D f r Hcddza Beck-(Continwd) -

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

DDE(D) /- -

NSIC (N)

Duration Power Reportable Shutdown System Co m ponent Event Date (!!rs) (%) Event Description Cause Method Involved Involved Category I

10/16/1967 3 30 10-67-12 Inadvertent grounding of vital bus B 3 EB ELECON N 5.1 while trouble shooting an irregularity in the power supply 10/25/1967 120 43 10-67-13 Blown fuse in the power supply to A 3 ED INSTRU N1.1.4 the coincidenter 10/30/1967 43 Power redu: tion. To inspect all B 5 HA VALVEI N1.1.4 f our turbine control valves 11/04/1967 36 43 Repaired steam leak on No. 1 B 1 HA PIPEII N1.1.4 control valve - sain steam turbine Icad-in line 11/10/1967 25 65 Installed nov level control valves B 1 HH VALVEI N1.1.4 on reheater drain tanks 11/18/1967 2 70 Ao-67-15 Inadvertent granading of the vital B 3 ED ELECON N5.1 bus while troubleshooting 11/20/1967 177 75 Bemoval anS re-lastallation of B 1 HA VALVEI N 1.1. 4 turbine control valves e

L l00

Tahlo A1 1968 Forced Shutdoens and Power Radsctions far Hadden Nick-(Chntinnd)

DBE(D) /

h SIC (N)

Duration Power Reportable Shutdova System Com ponent Eve nt ato (llr n) (1) Event Description Cause Nethod Involved Involved Catego ry 1/12/1960 242 70 Repai ed pressurizer safety, B 1 C1 VALVEI N1.1.4 feedwater check, and steam line isolation trip valves 2/09/1963 0 45 10-68-2 Power red uction, sepairing leak on B 5 EH VALVEI N1.1.4 In feed pump discharge check valve 2/14/1968 5 85 A0-68-4 Reactor coolant pump bus feeder A 3 EC CETBRK N1.1.4 circuit breaker tripped open 3/01/1963 272 85 Installed new coheater drain tank B 1 MH TURBIN N 1.1. 4 and repaired solsture separa to r desister sections 3/12/1963 49 68 Repaired steam leak on high 8 1 HA TU R BIN N 1.1. 4 pressure turbine flange 3/15/19(8 298 17 Repaired steam leak on high , B 1 HE TURBIN N1.1.4 pressure turbine flange 3/20/1963 1 10 A0-68-8 Inadvertent acteation of turbine G 3 Ik INSTRO 31.1.4 low vacuum trip signal 3/29/1963 32 17 Removed bijk pressure turbine upper B 1 HA TURBIN N1.1.4 cylinder for repair 4/01/1963 4b4 0 Installed 4 steam generator seal B 1 CC VESSEL N 1.1. 3 welded aanway diaphrams 4/27/1968 9 65 10-68-7 A switching procedurc bas used G 3 EA CKTBBK D2.6 which unguarded a traarfer trip relay causing the site low side breakers ta open 5/03/1963 0 52 Power reduction. Repairing weld B 5 HH PIP EIR 51.1.4 leak in 82 feedvater line flow o rifico 5/04/1968 4 52 A0-60-0 Automatic reactor trip caused by A 3 IA INSTRU N 1.1. 4 f also overpower trip signals on two power range channels 5/19/1963 0 52 Power reduction. For maintena nce B 5 EA EL ECOM N 1.1. 4 -

on 345 Kr transmission line 6/05/1968 96 43 Repaired turbine control valves and B 1 HA VALVEI N1.1.4 high pressure turbine gland seals

• Inl

Table A1 1968 Forced Shutdowns and Power Redrctio2s for Haddas Neck -(Contixued)

.e ~4

. . c DBE (D) /

N SIC (N)

Duration Po'ver Reporta ble Shutdown System Component Event Date (Ifrs) (X ) Event Description Cause Method Involved Involved Category 6/10/1963 6 85 10-68-10 A salfua: tion la loop 1 flow A 3 IA IN STRU N2.3 transmitter caused the flow signal to fail low 6/16/1963 4 ' h'i Repaired pinhole leak in 83 feed B 1 CD V&LVEI N1.1.4 line check valve 7/05/1963 29 52 Replaced steam generator feed pump B 1 HH PD M PII N1.1.4 nochanical seals 7/19/1968 51 83 Repaired turbine control valve B 1 51 VALVEI N1.1.4 0/02/1968 65 70 10-68-12 Both turbine,,stoo ,waj wrs clo sed - 1 1 51 YALVEI D2.3 tripping tne turonne 0/09/1968 87 70 10-68-13 Right hand turbine stop valve A 3 UA VALVEI D2.3 closed 0/23/1968 2 70 t o- 6 8- 14 & failed closed feedvater A 3 RH VALVEI D2.7 regulating valve resulted in a steam / feed flow mismatch 8/29/1968 55 70 Rebuilding turbine control and stop B 1 BA VALVEI N1.1.4 valves 9/01/1968 115 0 Repair leaking spray valve bonnet B 1 SF VALVEI N 1.1. 4 9/21/1968 6 85 Replaced test solenoid dump valve B 1 HA VALV0P N1.1.4 in the auto-stop oil s ystem from the right hand turbine stop valve 11/17/1968 0 52 Power reduction. For maintenance B 5 EA ELECOM N 1.1. 4 on 345 Ky transmission line 11/10/1968 12 85 A0-6 8- 16 1D scram breaker opened without A 3 In CETBRK N 1.1. 4 apparent cause, tripping the plant l 11/22/1963 0 61 Power redu: Lion. Reactor physics B 5 RB CONROD N 1.1. 4 testing - 3etermination of rod worth l 11/20/1968 71 10 Shutdown ta plug leaking tubes in B 1 HC PIPEII N 1.1.1 l

l main condenser l l O2-1

i Table A1 1968 Farc2d Shttdsens cad Pecer Rsducticas fcr Mcddam Ecck-(Coitirued) l _ __ __ _. _-- -----. _ - -- . . .

DDE (D) /

N SIC (N)

Duration Power Reportable Shutd own System Component Event Date (lic a) (%) Event Descrip tion Cause Nethod Involved Involved Category 12/01/1968 26 0 Installing new circulating water B 1 HP PIP EII N 1.1.1 varming line

~

i 12/09/1969 4 85 A0-68-19 1B scram breaker opened without A 3 IA CKTBRK N1.1.4 apparent reason, tripplag the plant 12/17/1968 10 85 40-68-21 1D scram breaker opened without A 3 IA CKTBRK N1.1.4 apparent reason, tripplag the plant 12/25/1963 14 85 10-68-23 IB scram breaker opened without & 3 IA CET BRK N 1.1. 4 apparent reason, tripping the plant 12/27/1969 50 85 Investigation of the unexplained B 1 IA ZZZZZZ N1.1.4 plant trips e

9 e

105 -

utes. ; e, g

) _

DBE(:)/

N SIC (N ), ,

Duration Power Reportable Shutdown System Co m ponent Event Data (Hrs) (K) , Event Description cause Method Involved Involved Category 1

I 1/08/1969 5 07 A0-69-2 1D trip breaker opened - tripping A 3 IA CKT ERK N1.1.4 the plant ,

1/08/1969 2 100 no-69-3 uhile repairing an oil leak, the A 2 HA VALTOP D2.3 gage line on the turbine auto-stop system was broken, causing the closure of the right hand turbine stop valve 1/18/1969 0 51 Power redaction. Maintenance on B 5 EA ELECON N1.1.4 345 Ky treassission lime e

1/19/1969 0 51 Power reduction. Maintenance on B 5 EA EL ECON N 1.1. 4 l l . 345 KV transmission lite l

l 1/25/1969 0 51 Power reduction. Maintenance on B 5 EA ELECON N1.1.4 345 Ky transmission line l

1 2/02/1969 0 51 Power reduction. Maintenance on b 5 EA ELECON N1.1.4 l l 345 KT transmission lite 2/09/1969 0 51 Power redaction. Maintenance on E 5 EA ELECON N1.1.4 345 KT transmission line 2/14/1969 0 51 Power reduction. Raintenance on B 5 EA ELECON N 1.1. 4 345 Kw transmission lire 2/22/1969 0 51 Power redaction. Maintenance on B 5 EA ELECON N1.1.4 345 KV transmission lite 3/29/1969 7 100 Right hand stop valve closed, due A 2 HA VALTOP D2.3 to failure of servo-motor cup valve 4/11/1969 456 100 Turbine valve modifications and B 1 HA VALTEI N1.1.1 repairs 5/01/1969 250 0 Disassembly and repair of main B 1 HA GENERA N 1.1. 4 genera tor 6/06/1969 53 04 Repaired generator hydrogen leak B 1 HA GENERA N1.1.4 04 An-6 9-8 Broken feedvater regulating valve 2 IIH VALTEI DI.2 6/10/1969 18 A plug causing flooding cf 83 steam generator 7/15/1969 5 03 AO- 6 9- 9 While switching outside the plant H 3 BA CKTDRK M9.1 to remove one 115 Er line froe service, ,

a procedural error allowed both incoming lines to trip IO+

Tahle A1 1969 Forcad Shitdoens aRd Power Rsductions far Haddsa Meck-(Continued) r

, DBE(D)/

N SIC (N)

Dur.ition Power Reportable . Shutdown System Comyonent Event Datu (Hrs) (%) Event Description cause Method Involved Involved Catego ry

, .n-7/10/1969 50 83 Installed new seals in 84 reactor B 1 CD PU N P11 N1.1.4 coolant pump 0/02/1969 7 83 A0-69-10 Electrical stora opened all four H 3 BA CKTBBK N9.2 power circuit breakers in 345 Ev yard 0/10/1969 0 83 10-69-11 Manual trip due to two dropped A 2 RB REL ATI N 1.1. 4 control rods because of malfunction in-out relays 0/30/1969 41 83 A0-69-12 Replaced resistance temperature B 1 ID INSTRO N 1.1. 4 detectors in loops 83 and 84 9/01/1969 15 0 AO-69-13 843 control rod stickieg during A 3 RB COMROD N1.1.4 power increase 9/06/1969 39 83 Installed five thimble plugs in B 1 HH PI P EII N1.1.4 leaky tubes in IB feedsater , heater 10/06/1969 13 03 Repaired relief valve ca turbine B 1 HA VALVEI N1.1.4 governor oil pump 10/10/1969 50 63 Repaired tube leaks in 3B feedwater B 1 HH PIPEII N1.1.4 heater 11/11/1969 3 100 An-69-15 Steam / feed / level mismatch due to G 3 HA VALVE 1 D2.3 inadvertent opening of turbine governor oil system relief valve 11/12/1969 32 100 An-69-16 Steam / feed / level aisaatch due to A 3 HA VALVEI D2.3 malf unction in turbine governor oil system relief valve sten 11/10/1969 3 100 An-69-17 The primary electric protection A 3 EB RELAYI N1.1.4 relay tripped due to scisture in terminal box causing a short circuit ,

11/20/1969 42 100 A0-69-18 Manual trip due to two dropped A 2 RH INSTRU N 1.1. 4

• control rods caused by a faulty capacitor in 811 rod botton .:

bistable drawer

~

l e f."

Tchle A1 1969 FcrcId Shttdat:2a exd Posco Rrdtetio~as frr if addIQ Nock -(Continued) .

DBE (D) /

N SIC (N)

Duration Pow er Reportable Shutdown System Component Event Datu (Itcs) (%) Event Descrip tion Cause Method Involved Involved. Category 12/06/1969 0 56 Power reduction. Maintenance on B 5 EA ELECON N 1.1. 4 345 KT transmission line 4

9 9

0 106

Tablo 11 1970 P:rced Shutdo::sa ccd Pouer Radicticca fcr Bcadoa Neck-(Continusd)

DBE (D) /

sSIc ts)

Duratiou Power Reportable Shatdown Systee Com pon ent Event Dats Ints) (5) Event Descrip tion Cause nothod Involved Involved Category 3/21/1970 16 78 Repair leaking 11 stea s gene ra tor B 1 BB TALTEI N1.1.4 f eed pump discharge valve 3/25/1970 4 78 Repair leak in 83 staae generator B 1 RB PIPEII M1.1.4 feedwater line drain 4/02/1970 3 75 10-70-3 Banual trip due to more than one A 2 BB RELAYI E 1.1. 4 dropped control rod caused by a f aulty contact la a 'ccattel in' relay l 4/14/1970 5 75 Sepair steam leak in 83 feedwater B 1 HH PIPEII N1.1.4 l flow control orifice senslag line 1

I 6/27/1970 al 63 Balancing tarbine and adjusting B 1 BA TU R BIE N 1.1. 2 l turbine controls 7/19/1970 10 100 Loss of condenser vacuma dee to B 1 BC BTEICE N1.1.4 plugged gland seal seal steam s trainer 8/09/1970 47 70 Replaced as turbine stcp valve B 1 HA TALTEI N1.1.4 shaft 8/17/1970 1 100 AO-70-1 Lightning struck terminal box la 1 3 BA CETBBK E9.2 345 Es suitchyard, trigging plant 8/19/1970 0 75 10-70-8 Power reduction. To replace B 5 CB PIPEII N1.1.4 insulation on 84 reactcr coolant

[ loop

\'

8/25/1970 0 65 10-70-8 Power reduction. To replace B 5 CB PIPEII M1.1.4 insulation on 84 reactcr coolant loop .

0/30/1970 4 75 10-70-8 Returning isolated loog 4 to B 1 CB EEEEEZ N1.1.4 service after repairs 9/01/1970 9 100 Replaced valve packing glands la B 1 CB VALTRI E 1.1. 4 reactor coolant loops 9/10/1970 65 100 Replaced valve packing glands in B 1 CB TALTEI N1.1.4 reactor coolant loops 9/24/1970 0 47 Pouer redoction. Replace faulty B 5 EA TRANSF N1.1.4 station service transf creer 102 -

~

Table A1 1970 Forced Shatdowns and Power Reductions for Hadden Neck-(Contis.ued)

DBE (D) /

uSIc(E)

Duration Power Reportable Shetdown System Component Event Date illes) (%) Event Descrip tion Cause Bethod Involved Involved Category 10/12/1970 6 100 10-70-9 Bosentary decrease la reactor 1 3 IA IESTRO 31.1.4 coolant flow indication tripped plant 10/23/1970 13 100 Operator Liceassag exans E 1 BI ZZZZZE E8.3 10/26/1970 8 100 10-70-11 Investigation of pressurizer level B 1 CA INSTRO N1.1.4 probles 11/24/1970 0 70 Power reduction. Dae to loss of B 5 BC PURPII N1.1.4 "Da circulating water gump 11/25/1970 0 70 Power reduction. . To repair ' flange B 5 HH TALTEI 31.1.4 leak in 83 feeduater ccattol valve 12/03/1970 6 100 Plant trip caused by suitching the G 3 EB ELECOE N2.2 semi-vital bus power supply to a f anited source 10hi

Tablo Als 1971 F:rc d Shutdocac c:d Poser Ecoscticte frr Mcdd 0 Neck-(Continu1d)

DBE(D)/

NSIC (E)

Duration Power Reportable Shutdown Systen Co mpon ent Event Datu (Hrs) (1) Event Descrip tion Cause Method Involved Involved Category 1/03/1971 0 70 Power reduction. Identification of B 5 BC HTEICH B1.1.4 malm condenser tube leaks 1/04/1971 0 90 Power redaction. To repair B 5- HH YAL10P N1.1.4 feedwater heater valve positioners 1/31/1971 0 70 Power reduction. . Idealfication of B 5- HC HTEICH N1.1.4 main condenser tube le aks 2/01/1971 0 70 Power reduction. To plug main B 5 BC HTEICH M1.1.4 condenser tube leaks 2/02/1971 0 70 Poser reduction. To repair steam B 5 HB PIPEII E1.1.4 generator feed pump seal cooling water bose connection 2/06/1971 0 70 Poser reduction. To cceplete work B 5 EA EL ECON N1.1.4 on transmission ilmes 3/19/1971 0 70 Power reduction. To perform E 5 BB ELECON M1.1.4 transmission line mala tenance 5/28/1971 8 90 Repair leaking high pressure B 1 HA TUERIN N1.1.4 turbine inspection par t 6/07/1971 S2 100- Shutdova to hot standby to identify B 1 EA ELECOE E2.2 source of 4.16 It ground 6/11/1971 13 100- Failed light source casses reactor A 3 11 INSTRU N 2.1 trip 7/12/1971 36 100 nanual turbine trip to investigate E 2 PC PDHPII M1.1.4 cause of leak in heater drains and pump suction 0/21/1971 4 100 Loss of solenoid valve la 84 A 3 HH TALTEI D2.7 feedvater control systes resulting in closure of 94 feedsater control valve, causlag steam /f eed flow mismatch coincident with low level in steam generator 0/21/1971 25 6 Plugging leaking tubes in 11 and 1B B 2 HH HTEICH N1.1.4 ,

feedvater heaters

( 0'I .

Table A1 1971 Forced Shutdowns and Power Reductions for Hadden Neck -(Continued)

DBE (D) /

i usIc ts) i Duration Power Reportable Shutdous Systen Congonent Event Date (lles) (5) Event Description Cause Bethod Involved Involved Category I

I' 9/07/1971 18 100 A momentery around on the vital 3 3 IA bus, camsd by as adjustment of the INSDEU D2.7 low neutros level alars motpoint needle on source range chamael ett resulted in loss of control of feedwater to steam generator 10/23/1971 47 70 Repairing steam generator feed penp B 1 55 YALTEI N1.1.4 check valve 12/02/1971 23 100 10-71-1 A fallare la the T A 3 EN TALTEI D 2.7 average /feedwater over-ride solenoid operating valve closed 82 f eedvater control valve causing steam / feed flow mismatch coincident with lov level la 82 steam generator 12/16/1971 0 52 Power redaction. Repairing leaking B 5 BM TALTEI N1.1.4 feedvater regulating velve upper bonnet gadget II0

Table A1 1972 Esrcrd Shutdarne and Pouer Reductions for fladdoc Neck -((butinued)

= - - - - - - - - - - - - - - - - - - _ = ---- --

DDE (D) /

H SIC (N)

Duration P ow er Reportable Shutdown System Co m ponent Event Date (itrs) (%) Event Description Cause Method Involv ed Involved Ca tegory 1/22/1972 20 100 automatic reactor / turbine trip A 3 IA INSTRU D2.3 caused by failed light source in the 8 2 reactor coolant loop flow indicator

2/18/1972 0 70 Power reduction. Plugging main D 5 ffC IITEICII N1.1.4 condenser leaking tuber 2/27/1972 21 100 Steam / feed flow mismatch coincident G 3, EB INSTRU 26.1 with lov level in 84 s team generator due to loss cf on-site power caused by error of test department personnel 3/03/1972 0 93 Power reduction. Decrease in i3 D 5 till VALVEI N1.1.4 steam generator feed pump suction due to sheared valve pcshtioner on heater drains tank normal level .

control valve 3/08/1972 0 50 Power reduction. Plugsing main B 5 HC IITEICll N 1.1. 4 condenser tube leaks 3/23/1972 0 62 Power reduction. Plugsing main D 5 IIC N HT EICII N1.1.4 condenser tube leaks 3/25/1972 0 50 Power red uction. Replacing heater D 5 Illi VALVEI N1.1.4 d rains tank normal level control valver packing 4/14/1972 0 50 Power reduction. Repacking heater B 5 Hit VALVEI N 1.1. 4 d rains tank normal level control valve 4/29/1972 0 50 Powder reduct ion. Repacking heater D 5 Illi VALVEI N1.1.4 d rain tank normal level control valve 5/22/1972 12 100 Repaired steam leak on 84 foedvater B 1 lill VALVEI N1.1.4 bypass check valve 7/14/1972 63 0 Ropair leaking primary and D 1 CD V A LVEI N1.1.4 secondary safety valves ,

7/17/1972 5 0 Turbine balancing B 1 HA TURBIN N 1.1. 4 .

(Il .

=

Tablo A1 1972 Forced Shutdowns and Power keductions for Hadden Neck -(Continued)

DBE(D)/

N SIC (N)

Duration Power Reportable Shutdown System Component Event Datu (llrs) (%) Event Description Cause Method Involved Involved Ca tego ry 9/21/1972 85 100 Turbine / reactor trip due to a loss 1 3 ED TRANSF D2.3 of a constant voltage transformer 11/23/1972 73 0 Repaired turbine lef t hand stop D 1 HA VALTEI N1.1.4 valve 12/15/1972 0 70 Reduced power. Plugging leaking B 5 HC HTEICH N 1.1. 4 condenser tubes * -

12/21/1972 0 70 Power reduction. Plugging leaking B 5 ..C HTEICH N1.1.4 condenser tubes 1

O e

ll1-

Table A1 1973 Forc d Shttdm:ns and Power Rc4uctions for Hadds:3 Nec k -(Contlaued)

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

DBE(D)/

NSIC (N)

Duration Power Reportable Shutdown System Co m ponent Event

, Da to (Hru) (%) Event Description cause Hethod Involved Involved Ca tego ry 1/09/1973 0 70 Power reduction. Plugsing main B 5 HC HTEICH N1.1.4 -

condenser tube leaks 1/25/1973 0 52 Power reduction. Replacing level e 5 HH YALVEI N1.1.4 column lower isolation valve on B reheater drains tank 2/11/1973 0 70 . Power reduction. To permit B 5 HP HECFDN N1.1.4 shutdown of BED circulating water pumps while a shear pia was ,

replaced on B traveling water i screen 2/22/1973 9 100 Ao-73-2 Connector between 82 fcedvater A 2 HH YALVEI D1.2 control valve stem and actua tor loosened, permitting tie valve to go fully open resulting in an uncontrolled invrease Ja 82 steam generator level 2/24/1973 47 100 Nelded a broken steam inlet baffle B 1 HH HTEICH N1.1.4 plate on D soisture segarator reheater 4/02/1973 0 52 Power reduction. Repacking heater B 5 HH VALVEI N1.1.4 drains tank normal level control valve 4/08/1973 0 52 Power reduction. Renoied heater B 5 HH VALVEI N 1.1.4 d rains tank normal level control valve and replaced it with a temporary spool piece 6/02/1973 366 100 Planned shutdown due tc turbine B 1 HA TU R BI N N1.1.4 vibration - replaced all sixth row blades 7/08/1973 3023 100 Planned shutdova to in vesti' gate B 1 HA TUDDIN N1.1.4 increased turbino vibration.

Replaced 'both low pressure turbine rotors 12/21/1973 0 0 Turbine balancing B 4- HA TUREIN N 1.1. 4 12/21/1973 16 0 Added balance weights to B 4 HA TU R BIN N 1.1. 4 t urbine/g enerator -

If 3 .

e .

Table A1 1973 Forced Shutdowns and Power Reductions for Hadden Neck-(Continued)

DDE (D) /

N SIC (N)

Duration Power Reportable .

Shutdown System Component Event Date (Hrs) (1) Event Descrip tion Cause Nethod Involved Involved Ca tego ry 12/29/1973 64 100 Performed additionai B 1 HA TUR BIN N 1.1. 4 turbine / generator balancing

)

II+

Tabio A1 1974 Forced Shutdowns and Power Reductions for Hadden Neck -(Continued)

. DDE (D) /

N SIC (N)

Duration Power Reportable Shutdown Systen Coatonent Event Date (Hrs) (%) Event D escrip tion Cause Bothod Involved Involved Category 1/10/1974 7 50 Frozen sensing lines tc steam flow A 3 IA INSTR 0 M 9. 2 transmitters produced spurious signal 1/19/1974 5 50 10-74-2 Ice stora shorted protective relays A 3 EA RELAYI M9.2 causing loss of of f-site poder 1/19/1974 4 50 10-74-3 operator inadvertently shut down G 3 HF PURPIX N6.1 two circulatin) water gunps .

supplying the same con denser 2/16/1974 12 50 Repaired leaking flange on 84 A 1 UH VALVEI N1.1.4

, feedwater control valve 1/23/1974 660 50 Increasing vibration in turbine due A 1 HA TURBIN N1.1.4 to broken blades 4/30/1974 4 50 Vibra tion in turbine B 2 UA TU R BIN N 1.1. 4 4/20/1974 4 50 Vibration in turbine B 2 HA TU B BIN N 1.1.4 e

O/20/1974 4 50 Vibration in turbine B 2 HA TU R BIN N 1.1. 4

,4/20/1974 8 50 Vibration in turbine B 2 HA TU R BIN N1.1.4 4/20/1974 6 50 Vibration in turbine B 2 UA TUR BIN N1.1.4 6/24/1974 9 50 Defective capacitor in Loop 82 flow A 3 IA INSTBU N1.1.4 transmitter initiated reactor trip -

signal 9/08/1974 9 50 Sepaired turbine eccentricity A 1 HA TU R BIN N1.1.4 12/08/1974 12 80 Lightning f aulted tranraission H 3 EA ELECON N9.2 lines causing generator load rejection

~

U5

.)

Table A1. 1975 Forced Skatdowns and Power Redactions for Hadden Neck-((butinued)

---

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

DB E (D) /

MS?C(u)

Duration P ow er Reportable Shutdova System Co mpon ent Event Dats (Hrs) (1) Event Descrip tion Cause Method Involved Involved Category

---

-- - a __- -

2/01/1975 16 0 Trip from low feed pump section 1 3 HH PUR PII D2.7 pressere 3/26/1975 39 50 10-75-1 Packing gland leakage from letdown 1 3 CB TALTEI 31.1.4

, system stop valve to the valve sten leakoff header was in escess of adelaistrative limits -

7/05/1975 5 50 Dait forced off line by broken oil & 3 RA PIPEII D 2. 3 pressure gange line on turbine 7/14/1975 3 50 Onit forcel off ilme by leaking & 3 R1 PIPEII D2.3 auto-stop 11ae on turbine

, 12/06/1975 7 80 Repaired leaking tebes in feeduater B 1 HH HfEICE E 1.1. 4 heaters Ilb

Tablo 11 1976 Forcad Shutdowns and Power Reductions for Hadden Neck-(Continued)

DE E (D) /

N SIC (N)

Du 4 tion P ow er Reportable Shutdown Systea Co m ponent Event Date (Ilcs) (%) Event Description Cause Nethod Involv ed Involved Category

---

_ ==--- __ -- ----- _ _

1/22/1976 17 100 LER-76-1 Frozen senslag line caused steam A 3 IA PIPEII N9.2 line break signals which tripped unit off line 4/02/1976 37 100 Broken steam faffle in D moisure A ,1 HH VESSEL N1.1.4 separator. Welded back to original condition 4/38/1976 6 70 Spurious signal from the low A 3 IA INSTRU N 2.4 pressure scram calcula tors 0/01/1976 13 100 Lightning strike on 345 Ky line H 3 EA ELE CON N9.k resulting in plant trip 9/10/1976 7 100 Unit trip caused by coclant pu mp G 3 CB PUR PII N 6.1 shutdown due to operator error o

O e

II) ,

I i

Table A1 1977 Forced Shutdowns and Power Reductions for Hadden Neck-(Continued)

DDE(D)/

M SIC (N)

Duration Power Reportable Shutdown System Comtonent Event '

Datu (!!r s) (%) Event Description Cause Method Involved. Involved Category

, C/19/1977 140 70 Feedvater heater leak corrected B 1 HH llTEICH N1.1.4 general plant asiatenarce during o utage s 9/10/1977 11 100 Generator voltage rege]ator A 3 HA GENERA N1.1.4 aalfunction 10/09/1977 7 100 Turbine load mismatch caused- by a A 3 HA VALVEI N 2.1 control valve malfunction 12/01/1977 5 0 LER-77-26 Niring error la main steam line A 3 HB ELECON N5.1 trip valve circuitry 4 12/03/1977 7 100 Generator voltage regulator A 3 HA GENERA N1.1.4 malf unction

'12/10/1977 31 100 Generator voltage regulator probten B 4 la GENERA N1.1.4

.; investigation i

! 12/14/1977 14 90 Turbine balance B 4 NA TU R BIN N1.1.4 12/15/1977 18 10 Turbine balance B 4 HA TURBIN N1.1.4 12/17/1977 22 90 Turbine balance B 4 HA TU R BIN N1.1.4 12/27/1977 7 90 Turbine balance B 4 HA TU R BIN N 1.1. 4 12/10/1977 22 100 Turbine balance B 4 HA TU R BIN N 1.1. 4 Its-

Table A1 1970 ForcId Shutdsens anil P:ccr its4uctions fcr Ildloa Neck-(Contirued)

-u DDE(D)/

N SIC (N)

Duration Power Reportable Shutdown System Component Event Date (Ilt s) (%) Event Description Cause Method Involved Involved Category 1/01/1970 12 0 Plant trip due to stuck relay A 3 IA INSTRU N 2.1 1/02/1970 7 100 Balancing turbine B 1 ' HA TURDIN N1.1.4 1/09/1970 27 100 Revelded a loose floor plate in B B 1 Hil llTEICH N1.1.4 moisture separator reheater 1/20/1970 51 100 Repairing B steam generator feed D 1 HD PUHEII N1.1.4 pump bearing 2/13/1978 42 00 Revelded haffle plates in the B 1 Hit VESSEL N1.1.4 moisture separators 3/24/1970 23 100 LER-78-2 Replacing electrical tcrainal block F 1 EU ELECON N1.1.4 enclosure in the containment 3/30/1973 6 50 Reactor was shut down to comply H 1 CD PIPEII N1.1.4 with procedure for placing an isolated loop back in service 4/30/1978 6 100 Nhile bringing an isolated loop B 3 IA INSTRU N1.1.4 into service, t he 8 2 s team generator experienced a mismatch of steam flow with feed flow coinciden t with low ficv in steam genera tor 5/01/1978 3 100 Nhile turning 4160Y bus voltaetor A 3 IA INSTRU N2.2 f ound "of f" to positior 1 5 2 f uses supplying loop Nc. 1 a nd No.

2 flow indication blev causing loss of flow reactor trip 6/04/1978 30 1.00 Nelding looso ,baf fle plate on B B 1 IIA TESSEL N1.1.4 aoisture separator rebeater l 6/17/1970 66 100 Plant design ch ange No. 270 D 1 ED ELECON N 1.1. 4 l required the installation of heat shrunk sleeves on elec trical penetrations 7/30/1978 0 60 Power reduction. Replacing heater D 5 Illi 'UnPII P N1.1.4

• drain tank. pump seal

~

8/20/1970 12 100 Repaired leaking, crodtd dra in A 2 lill PIPIII N 1.1. 4 piping on moistere separator .

reheaters

Table A1 1978 Forced Shutdowns and Power Re luct ions for fladlem Neck-(Continued)

- _ - - - - - - - - = _ ,

. DDE(D)/

N SIC (N)

Duration Power Reportable Shutdown Systen Component Event Date (firs) (%) Event Description Cause Method Involved Involved Category 9/14/1978 0 52 Power reduction. For saintenance A 5 IIB PUMPII. N1.1.4 on B steam generator f eed pu mp 11/02/1978 3 100 Low pressure pressurizer trip due B 3 IA INSTRO N2.4 to false signal while instrumentation check la progress e

4 i l'LD

Table A1 1979 Forced Shutdonna and Power Reductions for Hadden Neck-(Continued)

DBE(D)/

NSIC (M)

Duration Power Reportable Shutdown System Co m ponent Event

, DSte (Hrs) (1) Event Description Cause Method Involved Involved Category 3/12/1979 1 0 saintenance on electrical equipment B 9 ED EL ECON N1.1.4 3/12/1979 3 0 Turbine balance B 9 HA TURBIN N1.1.4 3/14/1979 0 40 Power reduction. Replacing 811 B 5 HH PUN PII M1.1.4 condensate puay 3/17/1979 0 52 Power redaction. Repairing turbine B 5 HA TURBIN N1.1.4 flange leak S/18/1979 0 52 Power redaction. eplacing outboard B 5 HD PDH PII N1.1.4 pump seal on 11 steam generator j feed peep 6/02/1979 0 60 Power reduction. Repairing leak on B 5 HD FESSEL N 1.1. 4 84 steam generator hand hole 7/14/1979 0 62 Power redactica. Shutdbva of 83 B 5 CB PUN PII N3.1 loop because of leaking seal on 83 reactor coolaat pump 7/21/1979 30 60 51ssatch os low pressure steam damp A 3 HE- INSTED N2.4 systen 9/29/1979 234 100 Check weld area of 5/G feed line D 1 HH PIPEII M 8. 3 nozzles for cracks e

Ill/ .

1

i a

1 Table A1 1980 Forced Shutdowns and Power Reductions for Hadden Neck-(Continwd) 1 DBE (D) /

a SIC (N)

Duration power Reportable Shutdown System Com ponent Event i

Date (Hrs) (5) Even t Descrip tion cause Nethod Involved Involved Oategory 3/19/1980 0 60 Power reduction. Plugging two B 5 HC PIPEII 31.1.4 tubes each le mala condenser water i boxes C & D l

l l

3/20/1980 0 22 Power reduction. Plugging two B 5 BC STEIC8 N1.1.4 tubes in main condenser water bor B 3/27/1980 11 100 Reactor and techine trip was B 3 RI ELECON D2.3 experlesced declag/ wiring modification per SUREG 0578 4/26/1980 0 56 , Power redactica. . Nater la lobe B 5 RB PURPII N1.1.4 system of sala steam generator feed pump ,

8/02/1980 19 5 Reactor and turbine trip due to G 3 RA INSTRU D2.3 overspeed trip setting improperly adjusted 1 8/05/1980 15 51 Shutdova in order to tie la Loop 2 F 1 CB PDEPII B1.1.4 i af ter 42 reactor coolant pump l repair

9/27/1980 65 8 Turbine balance B 1 R& TO R BIN N1.1.4 11/18/1983 14 31 LER-80-16 While stopplag in control back A 2 RB CEDRVE D4.3 l rods, one step, both audible and l visual alaras indicated two dropped

! rods - notable gripper coils were at fault 11/20/1983 33 15 Rechanical overspeed device on H.P. H 3 HA INSTRO E5.2 turbine out of adjuntaent -

adjusted device and came back on 11ae 11.lt

i Table A1 1981 Forcad Shutdssns aad Po:ct Esductions fer Haddzo Neck-(Cortinued)

', , DD E (D) /

NSIC (N)

Duration Pow er Reportable Shutdown System Co m ponent Event Date ,

(Hrs), (%) Event. Descrirlion cause Method Involved Involved Category

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

1/04/1981 5 100 Steam generator low level alara A 2 HB TESSEL N3.2 1/05/1981 ~2 Flooding of feedwater beater A 2 MH HTEICH N1.1.4 2/27/1981 10 100 Plant trip due to coattol air A 1 PA PIPEII D 2. 7 header fitting amparation which caused loss of feedwater control 3/09/1981 100-80 Power red action for valve test and B 5 BC HTEICH W1.1.4 tube plugging La condenser "Ba, "C"

) water box 7/25/1981 100-75 Power reduction and three loop A 5 CD TALTEI N 3.1 oleration due to valve sten leak 7/29/1981 39 25 Reactor and turbine trip while H 3 HE RELAY 1 D2.1 coming down la power to bring plant o f f-line. Trip due to low pressure steam dump aalfunction caused by OPC relay 63 8/23/1981 47 Reactor shutdova to repair valve A 2 RI VALTEI N3.1 packing leak

~

9/10/1981 10 95 Plant manually tripped due to A 2 HA VALVEI N1.1.4 spurious right hand turbine stop valve closure 10/01/1981 31 10 acactor scran during AERR B 3 IA INSTRO N1.1.4 f unctional test 10/02/1981 0 10 Power reduction due to offgas A 5 MD UNK NNN N1.1 activities l 11/12/1981 5 Reactor and turbine samually A 1 CB PU R PII N6.1 tripped while attemptire to transfer reactor coolant pumps to station service water 12/11/1981 82 90 Nhile reducing power to plug ' B' A 3 HA T A LYEI D2.3 waterbox candenser tubes and for scheduled turbine main tenance, .

erratic operation of t urbine ontrol valves caused turbine everspeed .

trip 12,3

• 1

l Table 11 1981 Forced Skatdowns and Power Reductions for Haddes Neck-(Continued)

_ =----_-_-___ ____. ______ _______

DDE (D) /

NSIC (N)

Duration Power Boportable Shutdown System Co m ponent Event Date (Itrs) (%) Event Description Cause Bethod Involv ed Involved Category

___-_-------=-_ __

12/22/1981 12 100 Reactor and turbine trip from high 1 3 PA PIPEZI D2.3 pressure heater drain tank level due to failed fitting on control air systen 12/23/1981 5 Turbine generator off line to A 1 HD FALTEI N 3. 2 repair staan leak on MCY36 114

Table A1 1902 Forcsd Shttdowns and Power Reductions fcr Heddsc Neck-(Cortinued)

DB E (D) /

HbIC (N)

Duration Power Reportable Shutdown System Co m pon ent Event

• Date (Hrs) (%) Event Description Cause Method Involved Involved Category 1/12/1982 0 Load reduction due to condenser in A 5 HC HTEICH N1.1.4

- (leakage) 1/21/1982 0 Reduced power to plug tubes in "D" A 5 HC HTEXCH N1.1.4 and "B" waterboxes 1/31/1982 18 Turbine and reactor trip. Trip A 3 ZZ ZZZ ZZZ D2.3 caused by loss of generator field.

Located six blown f uses on main exciter 3/04/1962 0 Reduced power for repair of leaking A 5 HC HTEICH N1.1.4 condensor tubes 4/24/1902 0 Reduce power to repair: (A) A 5 CH PD H PII N 1.1. 4 Vibration on 1B steam Senera tor feed pump; (B) Leaking steam generator inspection hand hold on 8 3 SG; (C) Check for tube leaks in all four waterboxes S/14/1982 0 Load decrease to 400 MUE to check A 5 HC HTEICH K1.1.4 for condenser tube leakage 6/04/19C2 132.7 Reactor 5 turb trip due to short A 3 HA GENERA D2.3 circuit in exciter. Ecpaired short c ircuit. Also, 81 RSIT would not close. Repaired and retested valve 7/01/1982 0 Load decrease to repair extraction A 5 HB TU R BI N N1.1.4 line leak on H.P. turbine 9/17/1982 195.1 82-06/3L Moisture in main trans former oil. B 1' EB TRANSF N1.1.4 Filtered oil, repaired leaks, tested and returned to service.

Also found battery plates had swelled resulting in cracking of casing. Replaced battery 9/20/1902 0 Reduced power to plug condenser A 5 HC HTEICH N1.1.4 t ubes ,

11/08/1982 10.4 Main feel pump trip. Cue to loss A 3 HD PU H PIX D 2.1 ,

of suction. Verified system componemtn operating properly .

IM

l l

l l

Table A1 1982 Forced Shutdowns and Power Reductions for Hadden Neck-(Continued)

_ - __ ----- = __ -------------- --

DBE(D)/

NSIC (N)

Du rat i on Power Reportable Shutdown Systes Co m pon ent Event Dato (Hrs) (1) Event Descrip tion Cause Method Involv ed Involved Category 11/13/1982 39.2 Turbine right hand trit valve would A 1 HA MECFUN N1.1.4 not open. The pia . holding the disc to the sten sheared and was replaced 11/17/1982 9.8 Bank "C" rods dropped during rod A 2 IA CBDRDE D4. 3 action checks. Cause undete rmined 11/17/1982 2.3 FCP transfer greater than 10% G 3 CB PUNPII N6.1 power. Discussed with opera tors 12/12/1982 0 Load reduction for monthly turbine B 5 HA VALTEI N1.2.4 stop valve test 9 / 2_6

\

i l

l 1

Table A1 1903 Forced Shutdowns and Power Reductions for Hadden Neck -(Continued)

DDE(D)/

NSIC(N)

Duration Power Reportable Shutdown System Co m pon ent Event Date (Hrs) (%) Event Description Cause Bethod Involved Involved Category 4/12/1983 9.9 overspeed turbine trip test C 2 HA TUEBIN N1.2.4 4/12/1983 4.0 Inadvertent loss of poher to RCP G 3 CD CKT BRK #2.2 bus attempting to reset control switch position flags 5/31/1983 7.4 Spurious p3ver range instrument H 4 IE INSTRO N 2. 4 spike 6/10/1983 11.5 82 feed rojulating valvo failed A 2 HH HECFUN N1.1.4 open - repaired leakage 7/07/1981 0 Load reduction to repair leak on A 5 CC VALVEI N1.1.4 steam 9enerator level Indication isolation valve I17 If .

~.

Table A1 1984 Forced Shutdovas and Power Roductions for Hadden Neck-(ibntinnd)

DDE (D) /

N SIC (N) i Duration P ow er Reportable Shutdown System Co m ponent Event Date (lles) (1) Event Description Cause Method Involved Involved Category l __---------- . --- --_

! 5/05/1984 0 6" drain line from MSR to heater B 5 HJ PIPEXI N1.1.4 d rains tank developed a leak.

Beduced power to 65% to repair leak

. 11/09/1984 12.0 High hydrogen temperature deviation A 1 EB GENERA N1.1.4 11/10/1984 55.9 Took generator off line for B 1 EB GE N ER A N1.1.4 hydrogen seal repair and turbine overspeed trip test 11/15/1984 11.1 84-026 nanual trip - reduce Icad due to A 2 EB G E N ER A N2.2 i

voltage regulator cycling

! 11/20/1984 7.0 84-025 Loss of f13e laadvertent shutdown G 3 ZZ IIIIII N 3.1 of number three reactor coolant Pump l

I 4

1 e

' l (1L3'

Table A.2. Codes used for classifying the causes of forced shutdowns and power reductions and also for classifying the methods of shutdown Causes A Equipment failure B Maintenance or testing

• C Refueling D Regulatory restriction E Operator training and license exams .

F Adminis tra tive G Operational error H Other Methods 1 Manual 2 Manual scram C[

3 Automatic scram --

4 Continuation 5 Load reduction 9 Other 1

e s a

Table A.3. Systems involved in forced shutdowns and power reductions System Code Reactor RX Reacter vessel internals RA Reactivity control systems RB Reactor core RC Reactor coolant and connected systems CX Reactor vessels and appurtenances CA Coolant recirculation systems and controls CB Main steam systems and controls CC Main steam isolation systems and controls CD Reactor core isolation cooling systems and controls CE Residual heat removal systems and controls CF Reactor coolant cleanup systems and controls CG Feedwater systems and controls CH Reactor coolant pressure boundary leakage detection systems CI Other coolant subsystems and their controls CJ Engineered safety features SX i

~

Reactor containment systems SA o Containment heat removal systems and controls SB en Containment air purification and cleanup systems and controls SC Containment isolation systems and controls SD Containment combustible control systems and controls SE Emergency core cooling systems and controls SF Control room habitability systems and controls SG Other engineered safety feature systems and their controls SH Instrumentation and controls IX Reactor trip systems IA Engineered safety feature instrument systems IB Systems required for safe shutdown IC Safety-related display instrumentation ID Other instrument systems required for safety IE Other instrument systems not required for safety IF Electric power systems EX Offsite power systems and controls EA l AC onsite power systems and controls EB DC onsite power systems and controls EC Onsite power systems and controls (composite ac and de) ED Emergency generator systems and controls EE i Emergency lighting systems and controls EF

! Other electric power systems and controls EG l

l i

. i Table A.3 (continued)

System Code Fuel storage handling systems FX New fuel storage facilities FA Spent-fuel storage facilities FB Spent-fuel pool cooling and cleanup systems and controls FC Fuel handling systems FD Auxiliary water systems WX Station service water systems and controls WA Cooling systems for reactor auxiliaries and controls WB j Demineralized water makeup systems and controls WC Potable and sanitary water systems and controls WD Ultimate heat sink facilities WE Condensate storage facilities WF Other auxiliary water systems and controls WG Auxiliary process systems PX Compressed air systems and controls PA Process sampling systems PB

, Chemical, volume control, and liquid poison systems and PC l

controls l Failed-fuel detection systems PD Os 4

Other auxiliary process systems and controls PE -

Other auxiliary systems AX Air conditioning, heating, cooling, and ventilation sys tems AA and controls Fire protection systems and controls AB Communication systems AC Other auxiliary systems and controls AD Steam and power conversion systems HX Turbine-generators and controls HA Main steam supply systems and controls (other than CC) HB Main condenser systems and controls HC

, Turbine gland sealing systems and controls ED l

Turbine bypass systems and controls HE Circulating water systems and controls HF Condensate cleanup systems and controls HG Condensate and feedwater systems and controls (other than CH) HH Steam generator blowdown systems and controls HI Other features of steam and power conversion systems (not HJ included elsewhere) 1

- - - = ,w.,,w- - ,w.-..-,. ,,.w_,,.__-.

, e - -

a .

Table A.3 (continued)

System Code Radioactive waste management systems MK Liquid radioactive waste management systems MA Gaseous radioactive waste management systems MB Process and effluent radiological monitoring systems MC Solid radioactive waste msnagement systems MD Radiation protection systems BX Area monitoring systems BA Airborne radioactivity monitoring systems BB m

l e

I

. o Table A.4. Components involved in forced shutdowns and power reductions Component type Component type includes Code Accumulators Scram accumulators, safety injection ACCUMU tanks, surge tanks, holdup / storage tanks Air dryers AIRDRY Annunciator modules Alarms, bells, buzzers, claxons, ANNUNC horns, gongs, sirens Batteries and chargers Chargers, dry cells, wet cells, BATTRY storage cells Blowers Compressors, gas circulators, fans, BLOWER Ventilators Circuit closers /interruptors Circuit breakers, contactors, Con- CKIBRK

- trollers, starters, switches (other than sensors), switchgear Control rods Poison curtains CONROD Control rod drive mechanisms CRDRVE wN Demineralizers Ion exchangers DEMINX

{}

Electrical conductors Bus, cable, wire ELECON Engines, internal combustion Butane, diesel, gasoline, natural ENGINE gas, and propane engines Filters Strainers, screens FILTER Fuel eleents FUELXX Generators Inverters GENERA Heaters, electric Heat tracers HEATER Heat exchangers Condensers, coolers, evaporators, HTEXCH regenerative heat exchangers, steam generators, fan coil units Instrumentation and controls Control 11ers, sensors / detectors / INSTRU elements, indicators, dif ferentials integrators (totalizers), power suppl 11es, recorders, switches, transmitters, computation modules Mechanical function units Mechanical controllers, governors, MECFUN gear boxes, varidrives, couplings Mo tors Electric motors, hydraulic motors, MOTORX Pneumatic (air) motors, servomotors

Table A.4 (continued)

Component type Component type includes Code Penetrations, primary containment Air locks, personnel access, fuel PENETR handling, equipment access, elec-trical, instrument line, process P iiP ng Pipes, fittings PIPEXX Pumps PUMPXX Rec 6mbiners RECOMB Relays Switchgear RELAYX Shock suppressors and supports Hangers, supports, sway braces / SUPORT stabilizers, snubbers, antivaibra-tion devices Transformers TRANSF Turbines Steam turbines, gas turbines, hydro TURBIN turbines Valves Valves, dampers VALVEX Valve operators Explosive, squib VALV0P [(~

Vessels, pressure Containment vessels, dry wells, VESSEL pressure suppression chambers, pressurizers, reactor vessels Other components XXXXXX Codes not applicable ZZZZZZ f

l l

1 i

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

. '\

i Table A.S. Initiating event descriptions for DBEs as listed in Chap. 15. Standard Reviso Plan (Revision 3) i

1. Increase in heat removal by the secondary system j 1.1 Feedwater system malfunction that results in a decrease in feedwater temperature {

1.2 Feedwater system malfunction that results in an increase in feedwater flow 1.3 Steam pressure regulator malfunction or failure that results in l l increasing steam flow l 1.4 Inadvertent opening of a steam generator relief or safety valve l l 1.5 Spectrum of steam system piping failures inside and outside of i containment in a pressurized-water reactor (PWR) i l 1.6 Startup of idle recirculation pump a l 1.7 Inadvertent opening of bypass resulting in increase in steam flow"

2. Decrease in heat removal by the secondary system 2.1 Steam pressure regulator malfunction or failure that results in decreasing steam flow 2.2 Loss of external electric load 2.3 Turbine trip (stop valve closure) 2.4 Inadvertent closure of main steam isolation valves N

2.5 Ioss of condenser vacuus 2.6 Coincident loss of onsite and external (offsite) ac power to the station 2.7 Ioss of normal feedwater flow 2.8 Feedwater piping break 2.9 Feedwater system malfunctions that result in an increase in feedwater temperatures'

3. Decrease in reactor coolant system flow rate 3.1 Single and multiple reactor coolant pump trips 3.2 Boiling-water reactor (BWR) recirculation loop controller malfunction that results in decreasing flow race 3.3 Reactor coolant pump shaft seizure 3.4 Reactor coolant pump shaft break

4. Rasetivity and power distribution anomalies 4.1 Uncontrolled control rod assembly withdrawal from a suberitical or low-power start-up condition (assuming the most unfavorable reactivity conditions of the core and reactor coolant system),

including control rod or temporary control device removal error during refueling 4.2 Uncontrolled control rod assembly withdrawal at the particular power level (assuming the most unfavorable reactivity conditions of the core and reactor coolant system) that yields the most severe results (low power to full power) 4.3 Control rod maloperation (system malfunction or operator error), including maloperation of part length control rods

e A.5 (continued) 4.4 Start-up of an inactive reactor coolant loop or recirculating loop at an incorrect temperature 4.5 A malfunction or failure of the flow controller in a BWR loop that results in an increased reactor coolant flow rate 4.6 Chemical and volume control system malfunction that results in a decrease in the boron concentration in the reactor coolant of a PWR 4.7 Inadvertent loading and operation of a fuel assembly in an improper position 4.8 Spectrue of rod ejection accidents in a PWR 4.9 Spectrum of rod drop accidents in a BWR

5. Increase in reactor coolant inventory 5.1 Inadvertent operation of emergency core cooling system during

power operation.

5.2 Chemical and volume control system malfunction (or operator error) that increases reactor coolant inventory 5.3 A number of BWR transients, including items 1.2 and 2.1-2.6

6. Decrease in reactor coolant inventory 6.1 Inadvertent opening of a pressuriser safety valve in either a  %

t PWR or a BWR m 6.2 Break in instrument line or other lines from reactor coolant pressure boundary that penetrate containment 6.3 Steam generator tube failure 6.4 Spectrum of BWR steam system piping failures outside of containment 6.5 ! ass-of-coolant accidents resulting from the spectrue of postulated piping breaks within the reactor coolant pressure boundary, including steam line breaks inside 'of containment in a BWR 6.6 A number of BWR transients, including items 1.3, 2.7, and 2.8

7. Radioactive release from a subsystem or component 7.1 Radioactive gas waste system leak or failure

, 7.2 Radioactive liquid waste system leak or failure

! 7.3 Postulated radioactive releases due to liquid tank failures

! 7.4 Design basis fuel handling accidents in the containment and 1 l spent fuel storage buildings ,

) 7.5 Spent fuel cask drop accidents i 8. Anticipated transients without scram 8.1 Inadvertent control rod withdrawal 8.2 Loss of feedwater 8.3 toss of ac power 8.4 loss of electrical load 8.5 Loss of condenser vacuus j i 8.6 Turbine trip 8.7 Closure of main steam line isolation valves aThese initiating events were added for BWRs to be more specific  !

than DBE events 5.3 and 6.6.

A.6. NOAC event categories for non-DBE shutdowns N 1.0 Equipment failure N 1.1 Failure on demand under operating conditions N 1.1.1 Design error N 1.1.2 Fabrication error N 1.1.3 Installation error N 1.1.4 End of design life / inherent failure / random failure N 1.2 Failure on demand under test condii; ions N 1.2.1 Design error N 1.2.2 Fabrication error N 1.2.3 Installation error N 1.2.4 End of design life / inherent failure / random i

failure .

N 2.0 Instrumentation and control anomalies N 2.1 Hardware failure N 2.2 Power supply problem N 2.3 Setpoint drift N 2.4 Spurious signal N 2.5 Design inadequacy (system required to function outside design specifications g N 3.0 Non-DBE reductions in coolant inventory (leaks) D N 3.1 In primary system N 3.2 In secondary system and auxiliaries iI N 4.0 Fuel / cladding failurs (densification, swelling, failed fuel l elements as indicated by elevated coolant activity)

N 5.0 Maintenance error N 5.1 Failure to repair component / equipment / system N 5.2 Calibration error l N 6.0 Operator error

N 6.1 Incorrect action (based on correct understanding on ,the i

part of the operator and proper procedures, the operator

turned the wrong switch or valve - incorrect action)

N 6.2 Action on misunderstanding (based on proper procedures

and improper understanding or misinterpretation on the

operator's part of what was to be done -

incorrect action)

N 6.3 Inadvertent action (purpose and action not related, for j example, bumping against a switch or instrument cabinet)

N 7.0 Procedural / administrative error (incorrect operating or testing procedures, incorrect analysis of an event -

failure to consider certain conditions in analysis) t

,-- -m.,,_.-,,-,,.,_,,,-----g . , _ , , , - - - , . , - , - - , . - , . , , , . , , ,

. o Table A.6 (continued)

N 8.0 Regulatory restriction N 8.1 Notice of generic event ,

N 8.2 Notice of violation 1 N 8.3 Backfic/ reanalysis

N 9.0 External events N 9.1 Human induced (sabotage, plane crashes into transformer)

N 9.2 Environment induced (tornado, severe weather, floods, earthquake)

N 10.0 Environmental operating constraint as set forth in Technical Specifications s

m l

) i V.-_..-------.,_,--,.,.----- - - . _ _ . - - . . - - . , _ . - . - _ . , - ,_ , , - . - . - - _ , , , , , , , - - , _ _ , , , _ . ~ - - - - , ~.

APPENDIX B REVIEW OF REPORTABLE EVENTS This appendix presents the Haddam Neck data on reportable events in narra tive and tabular format and lists informa tion sources. It des-cribes how the information was encoded and how significance screening criteria were applied.

B.1 Scope 342 reportable events tha t occurred at the Haddam Neck plant from 1967 to 1984 were reviewed. The events include reportable occurrences (R0s), abnormal occurrences ( A0s ) ,

• and licensee even t reports (LERs) e filed by the utility resulting from technical specification viola- ~

tions. In addition, some events that occurred prior to 1975 but which do not fit in the above classifications, are included since they are signilicant from a safety standpoint. The information contained in the reportable events was coded in the format discussed later in this sec-tion. Tables of coded reportable events, arranged by year, are pre-sented in Table B.1.

This studf reviewed information about operating events reported in LERs and LER predessors (e.g., AORs, unusual events reports, reportable occurrences (R0s)] . Da ta on these types of events were retrieved from the NOAC data files. Also, any documents that contained LER-type infor-mation (such as equipment failures or abnormal events) were included in

• These report designations are to be distinguished from the events included in NUREG-0090, Reporta to Congress on AbnormL Occurrences.

the review to ensure that a total picture of the plants operating istory was obtained. Various types of opera ting repor ts and general correspondence were involved this way.

The table of reportable events (Table B.1) contains the following information for each event reviewed:

1. LER number or other means of identification of report type,

2. NSIC accession number (a unique identification number assigned to each document entered into the computer file),

3. date of the event,

4. date of the report or letter transmitting the event description,

5. status of the plant at the time of the occurrence (Table B.2), t

6. system involved with the reportable event (Table B.3),

7. type of equipment involved with the reportable event (Table B.4),

8. type of instrument involved with the reportable event (Table B.4),

9. status of the component (equipment) at the time of the occurrence (Table B.2),

10. abnormal condition associated with the reportable event (e.g., cor-rosion, vibration, leak) (Table B.5),

11. cause of the reportable event (Table B.2),

12. significance of the reportable event, and

13. comments and/or details on the event.

i B.2 Data Sources 1

Thr N6aC files of LERs (including the Sequence Coding and Search System) were the primary source of information for the review of report- l I

able events. When additional informa tion on the event was needed, the ,

J m-- -- . . . - . - - _ _ . - . - , , , , . . . - , - - - ~------m..

original LER (or equivalent) was consulted by examining (1) those full-sized copies on file at NOAC (for the years 1976 -1984); (2) the micro-fiche file of docket material at NOAC; or (3) the appropriate operating report (semiannual, annual, or monthly).

Printouts ob tained from the computer files identified " docket material" other than reportable event reports. This included licensee correspondence with NRC (or the Atomic Energy Commission (AEC)] concern-ing particular events. Licensees are of ten requested to submit addi-I tional information or perform furthar analysis. Before the LERs came into existence in the mid-1970s, it was not unusual for licensees to submit on their own or at the request of NRC or AEC more than one letter transmi tting information on a particular event. Thus, these printouts provided additional sources of information on reportable events.

Several special publications were reviewed for more detailed infor-ma tion on evenes of significance. These publications provided addi-tional details and evaluations or assessments of the events.

1. Reports to Congress on Abnonall Occurrences, NUREG-0090 series; I 2. " Power Reactor Even t Series" (formerly Current Even t Series) published bimonthly by NRC;

3. " Opera ting Experiences," a section of each issue of the Nuclear Safety journal; and

4. the publications of NRC's Of fice of Inspection and Ehforcemen t i

! (IE), such as operating experience bulletins, IE bulletins, IE cir-i culars, and IE informa tion notices.

l

- - - - - - - - - - , --- - - - - , _ -----m

- - - - ,--.-m ---,.m e .y. ..-,m- . - - - . - , ,, - - - , .

a

• B.3 Significance Screening

Two sets of criteria were used in determining the significance of reportable events. The first set of criteria listed in Table B.6 address events whose results include challenges to the safety protection l features of the plant. These events are termed " safety significant".

The second set of criteria listed in Table B.7 address events that have the potential to challenge the safety protection features of the plant. These events, which might require additional information or evaluation to determine their full implic tion, were termed "condi-tionally significant".

All reportable events were reviewed, applying the two sets of cri- y 4-teria for significance rather liberally. A number of significant events and conditionally significant events were noted. The events initially identified as significant or conditionally significant were analyzed and l

evaluated further based on (1) engineering judgment; (2) the systems, equipment, or components involved; or (3) whe ther the- nafety of the ,

plant was compromised. The conditionally significant events were sub-4 sequently " upgraded" to significant or " downgraded" to acr4significant as necessary. The final evaluation for significance considered whether a DBE was initiated or whother a safety funetion was compromised such thae the sys tem as designed could not mi tiga te the progression of even ts.

Thus, the number of events finally categorized as significant was re-duced considerably by these steps in the review process.

I j The reportable events not identified as either significant or con-1 dicionally significant were categorized as not significant (with an 'N' in the significance coluczn of the coding shee ts and in the tables).

.i

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

These events and the events rejected during the additional review step were further reviewed by compiling a tabular summary of the systems to detect trends and recurring problems. Table B.3 provides a listing of the systems.

B.4 Review of Reportable Events from 1967 through 1984 Figure B.1 illustrates the number of reportable events filed per year. The highest number of reportable events submitted by the licensee occurred during the years 1968, 1969, 1977, 1978 and 1984. Most of the abnormal occurrences in the early years were reactor trips due to spuri-l ous signals and due to faults on the AC power vital bus.

e e

B.4.1 Yearly summaries

• The following sections present a yearly stamaries of the reportable events.

1967 Haddam Neck went critical in July 1967. The reportable events con-sisted solely of inadvertent reactor trips. Six of sixteen trips were caused by maintenance errors while repairing or installing equipment on the plant's vital buses. Five other trips were attributed to spurious signals from various relays and power range monitors.

The only reactor trip classified as a significant event occurred on September 22. All ten steam dump valves opened during start-up due to an erronhous test signal and cancellation of trip override ( A0 67-07).

The reactor coolant temperature dropped 90*F (f rom 525'F to 435'F) in five minutes during this transient. The incident . was terminated by closing the nonreturn valves in the main steam lines. See Se c t. B.5.1.2 for a detailed account of this event.

1968

- Twenty-three reportable events were recorded in 1968. The reactor tripped fourteen times during the year, with four of the trips occurring during a five-week period in November and December. Three reportable events were caused by problems with control rods. Control rod problema are discussed in more detail in Sect. B.6.2. Wo of four steam isola-tion valves failed to close during a test. All four were then over-hauled and ratested successfully.

One reportable event was classified significant. On April 27, a switching error caused a loss of all of fsite power to the plant (A0 68-07). This was followed by the failut e of all three diesel generators to run. This s ta tion blackout is discussed in greater detail in 8 Sect. B.5.1.1.

1969 Twenty-three reportable events were reviewed for 1969. ho offsite power losses occurred, one on July 15 ( AO 69-09) and another on August 2 (Ao 69-10). During the July event, one of the charging pianps failed to run. These two events represent 28% of the total number (7) of offsite power losses occurring over the history of the plant. These two events were classified as significant are discussed in Sect. B.5.1.1.

Five reportable events involved control rods. On three occasions

) con trol rods dropped into the core. On the other two occasions they were inoperable. Control rod problems are discussed in Sect. B.6.2.

The reactor was shut down five times due to inadequa te oil pressure in the turbine control sys tem ( A0's 69-03, 69-04, 69-15, 69-16, and 69-19). The low pressure was a ttributed to leaks and pump f ailures. The pump itself was replaced twice (AO's 69-15, 69-16).

1

Two large spills of radioactive liquid which occurred on May 3 and May 6 were classified as significant events. The first spill (A0 69-06) was caused by overpressure in the boron recovery tank due to improper valve alignment. The overpressure cracked the tank at the disked head sean, releasing about 200 gal of radioactive distillate (1.74 Ci tri-tium). The second spill occurred when a nipple broke and discharged 500 gal of radioactive boric acid (A0 69-07) . In both instances the rup-tured vessels were repaired and the con tamina ted areas were cleaned up. Trace amounts of radioactive liquid and vapor were released to the atmosphere. For a further description of these events, see Sects. B.5.1.8 and B.5.1.9, respectively. For other event of environ- g 4-mental significance, see Sect. B.4.4.

Another event classified as significant, (Se ct. B.5.1.4), involved the f ailure of a feedwater regulating valve. On June 10, a feedwater pump tripped on a low suction pressure, but the steam generator water level continued to rise. Since the feeduster regulating valve is re-verse seating, the plug dropped open allowing full feedwater flow to the ste m generator (A0 69-08).

l 1970 l

i Eleven repor table even ts occurred in 1970. Control rod problems con tinued this year with groups of rods dropping into the core on two

! occasions (AO's 70-01, 70-03). Both events were caused by faulty relays r

which were replaced.

j On October 12, a malfunction of one of ten steam dump valves caused l

! an unanticipated cooldown of the reactor ( A0 70-09). This incident was l classified significant and is discussed fur ther in Sect. 3.5.1.2.

o .

On August 19, a small fire was discovered and extinguished at the junction of a reactor coolant pump and the suction piping (AD 70-08).

This incident was also classified significant and is discussed further in Sect. B.5.1.6.

1971 Six reportable events occurred during 1971. None of the events were classified as significant to plant safety.

On April 18, radioactive iodine was released when an operator pre-maturely broke the seals for the in-core thermocouples. The operators had assumed the reactor coolant cleanup had been completed. After breaking the seals, as in-core monitor indicated high particulate activ-icy. All personnel evacuated the containment. The reactor was resealed until April 22 when normal refueling operations resumed. Four personnel involved received 10 90% of the maximum permissible dose for I-131. For other events of environmental significance, see Sect. B.4.4.

1972 __

Eight reportable events occurred in 1972. None of the events were significant to safety. Leaking containment isolation valves resulted three reportable events. A fourth reportable event occurred on May 19, when one of the purification ion-exchange vessels was tested to evaluate the effectiveness of unplugging operations ( A0 72-02). The vessel was tested, then depressurized. The depressurization process evolved gases which were unexpectedly released into the a tmos phere. All activities were below the maximum permissible concentration for the nuclides re-leased.

On February 27, a partial loss of onsite power caused a reduction in feedwater flow followed by a reactor trip. The onsite power loss was caused by test personnel. Following the loss of onsite power, some dif-

ficulty was experienced while separating the main generator from the 1

i grid. Automatic sepcration failed due to two faulty breakers and the generator had to be isolated manually. Routine maintenance was per-formed and the plan t continued operation. On June 10, main tenance i

worker caused a scram by inadvertencly grounding trip inserumentation.

1

on September 21, power from a transformer to a semi-vital bus was lost, causing a reactor trip. Following the trip, the lef t side turbine stop valve failed due to steam cutting.

P-i &

1973 Haddam Neck recceded fourteen reportable events in 1973. For the third consecutive year, none of the events were classified significant to plant safety. Problems with the pressurizer level alarm occurred i

i- early in the year (AO's 73-01, 73-02). Replacement of bad photodiodes corrected the problems. 'No failures of steam generator support struc-cures were reported ( AO's 73-08, 73-12). These failures were attributed to installation errors. The supports were subsequently replaced.

I Unplanned releases of radioactive material continued to occur ( AO's 73-06, 73-07). The first airborne release of radioactive gas occurred due to a leaky diaphragm valve at the ion-exchange outlet. In the second release, a procedural error during maintenance on the purifica-tion system caused the release of 20 gal of coolant. None of the re-leases exceeded limits. A total of 0.238 of Noble gases were released in the events.

o . .

4 1974 Eleven reportable events occurred during 1974. Severe winter a

weather caused problems at the plant. The only event classified as sig-i nificant occurred on January 19, when all of fsite power was lost to the plant due to incorrect protective relaying during an ice storm (A0 l

74-03). This loss of service power (see Sect. B.5.1.1 was followed by

! the failure of a diesel generator service water pump to start. The pump

} was started manually. A faulty relay in the pump motor control circuit caused the pump to fail to start automatically. On January 18 the re-actor tripped due to . frozen instrument sensing lines ( A0 74-02) . Sub-f zero temperatures, strong winds, and f ailures of two heaters caused the g

2 lines to freeze. Backup heaters were installed to prevent recurrence.

An unplanned release of radioactivity occurred when the hydrogen l supply regulator to the volume control tank was leaking through the dia-1 l phraga (A0 74-07). The valve was replaced, no limits were exceeded.

On October 11 one of four main steam isolation valves failed to close due to dried out valve packing (A0 74-10). Following adjustment

of the packing gland, a successful valve movement test was made.

l

{ 1975 i

! Six reportable events were recorded in 1975. None of the events i

! were classified as significant to plant safety. On October 30, broken i

hold-down bolts were found on a steam generator support structure (A0

! 75-02). This was the third failure of steam generator hold down bolts reported at the plant, two others occurring in 1973 (Ao 73-08 and Ao l 73-12). The remaining reportable events involved nonessential equipment l failures and resulted in no consequences.

i, >

i I

I

-_m.._. . , _ _ _ _ . _ , _ _ . . . . _ . . . . . _ . _ . _ . . _ _ . _ _ , _ _ . . , _ _ _ _ . _ _ _ . _ _ _ .

1976 Two of the twen ty-two reportable events which occurred at Haddam Neck in 1976 were classified significant. Offsite power was lost during refueling on June 24 due to poor design of protective relaying (LER 76-14). During the outage, RHR flow was lost three separate times (see Sect. B.5.1.1). -

On July 5, with the reactor at 1% power, both auxiliary feedwater pumpe became vapor bound (LER 76-16). Both pumps failed due to back leakage through the check valve feed line. The ptsaps were vented and 1 returned to service. The faulty check valve was cleaned and repaired.

This incident involved a total loss of the auxiliary feedwater system e --

and is discuased in greater depth in Sect. B.5.1.3.

8 1977 The second largest number of reportable events (33) were recorded

in 1977. Six failures involved the charging pump sys tem. The first three failures occurred in April 1977, with charging pump 1A. During operation, pump 1A was removed from service fc low discharge pres-sure. Invescigation revealed 2 cracks on the shaf t due to installation l errors. When the pump was repaired and replaced, the seal housing leaked. Af ter replacing the seal housing, the bearings failed due to 1" proper housing ins talla tion. The axial alignment was adjus ted and problems ceased. The operability of the charging ptsspa is importan t i

since the pumpe serve as the high pressure injection pumps during a loss I

of coolant accident. Since these ptsaps perform a safety function, a more detailed analysis of their failures is examined in Sect. B.6.3.

One event classified as significant occurred in 1977. On August 21, multiple failures of reactor coolant pump (RCP) seal occurred (LER 77-19). The No. I seal failed due to chipping on its face. The cause of this failure was not found. Leakage from this seal caused the No. 2 and No. 3 seals to fail. The loop was shutdown. Three other coolant loops were available throughout the incident. This event is discussed further in Seec. B.5.1.6.

1979 Of the 37 reportable events occurring in 1978, 2 were classified significant to plant safety. On August 25, an IM radio caused a rod C

drop alarm, following by failure of the load runback signal to initiate D (LER 78-18). Investigation revealed a closed pressure switch isolation valve that prevented a signal from reaching the turbine load limit con-trol oil system. This disabled the turbine load cutback feature. The valve was'in a closed position due to an incorrect procedure. The valve was opened and the system functioned normally. The procedures were re-vised. This event is discussed further in Sect. B.5.1.10.

On December 29, an air supply valve unistched, closing isolation trip valves on all four steam generator blowdown lines (LER 78-22). The operator reset and blocked open the air supply valve to restore blow-down. This action provided a potential atmosphere release path in the event of a steam generator tube rupture. Due to the safety implications of this event, it is discussed further in Sect. B.5.1.5.

1979 Nineteen LERs were submit ted by the utili ty in 1979. Only one event was classified significant to safety. On August 13, a pressuri::er

, ~ '. .

y ._ ,

s -

s

~s 1 ,

PORV , inadvertently opened due to failure of a bistable in a pressurizer-

! pressure controller (LER 79-10). The reactor blowdown stopped when an operator closed the isolation valve. This event is discussed further in Sect. B.5.1.7.

1980 Twen ty-f our licensee event reports were recorded in 1980, none of which were classified significant. On February 4, a pressurizer pres-sure relief valve opened for about 2 min (LER 80-04) .' The valve was

! manually isolated and then closed. Pressure in the reactor dropped dur-ing the transient but was restored within minutes af ter valve isola-tion. The valve actuation was believed to be caused by spurious sig-nals, though no source was found. Leaking continued at the junction of the steam generator blowdown and service water lines (LER 80-07).

1981

- Twen ty-two reportable events occurred during 1981. Three of the events were of environmental concern. On August 16 and again on September 17, the amoun t of radioactivi ty released via the stack ex-ceeded limits. The last environmental event, which was classified sig-l nificant, occurred on April 22. Contaminated tube bundles were released

from the site prior to a health physics review. (See Sect. B.5.1.11 for further details).

i A procedural deficiency caused both auxiliary feedwa ter system

! ( AFWS) pumps to be inoperable (LER 81M)8). On June 16, the 'B' pump was removed from service for repair. The 'A' pump was tested and declared operable. During a valve lineup check, On operator discovered that the i

'A' pump recirculation valve had been closed rather than the 'B' pump

valve. A review of the surveillance procedures showed that the ' A' and '

'B' pump recirculation valve numbers were reversed. The error was not discovered earlier since both recircula tion pa ths are normally opened for operability tests. The procedure was revised.

1982 Ten reportable events occurred during 1982. None of the events were classified significant. Four of the events involve failures of mechanical equipment during a test. Two failures of mechanical equip-ment during a test. Two failures involved main steam isolation valves (MSIVs). The valves were repaired and ratested successfully. The other d

two failures involved dampers on the containment air recirculation D fan. A control linkage was found disconnected in one event and a link-age failure mechanically in another. Af ter repairs were completed, the damper system operated satisfactorily during the ratest.

1983 Twenty-six reportable events were recorded during 1983. Three of the events were significant.

Two of the significant events involved a loss of containment con-crol air which resulted in the loss of control of the pressurizer spray valves and the Power Opera ted Relief Valves (PORV). The first event was due to improper maintenance and the second event was caused by mechan-ical failure of the air filter canister. Both events occurred during the same month.

The third significant event occurred on March 15, 1983 during low power physics tests (LPPT) following a refueling outage. Due to

, abnormalities during the test, it was discovered that four control rods

were unlatched from their control rod drive shaf t assemblies. The unit was brought to a cold-shutdown condition and the reactor head was re-moved to facilitate repairs.

1984 Of the 31 reportable events that occurred during 1984, three were classified significant.

Two of the significant events involved a total loss of offsite power. Both events were caused by human error. The first loss of off-site power occurred on August 1. While performing a check-out proce-dure, an operator selected the wrong circuit breaker. All of fsite power tn was lost for ten minutes. The second loss of offsite power occurred on I August 24 when a large pump was started during a refueling outage. The service station transformer isolated during the attempted start because a transformer wire had been pulled from its terminal lug. The wire pull occurred earlier the same day when maintenance activities were performed in close proximity.

The third significant event of 1984 involved the failure of the re-fueling pool seal during a refueling outage. Approxima tely 200,000 gallons of borated reactor coolant water drained from the reactor cavity to the containment floor in 20 minutes. The original all metal seal had been replaced wi th a seal which included flexible rubber boots. The l

seal failure is attributed to inadequate design. The design error re-sulted in the NRC levying an $80,000 fine.

~ ~

k3010 Y Coding Sheet for 50portabl9 Ev03tc et Hadd33 Neck-1967

~

muabar AccessiO2 E700t RCport Plset comp:nett Abstract S ig nifica sco N uIbe r Da ta Dito StatC3 Syrt3D E;Cipsest I ctr: ment States cetdition Ctase cctegory Cc mest A06701 030200 07/25/1967 08/08/1967 8 RB && P 8 BF G E Ma in te na nce worker in adve rt en tl y trips overwaltage relay in cootrol rod power supply (reactor shutdows)

A06702 030200 07/28/1967 08/l8/1967 B IA H B BF D 5 Spike in the S in term ed ia te ra nge Su t 1 circuit and au tomatic cutout of BF-3s (reactor sh etdo va)

AO6703 030201 08/07/1967 09/18/1967 3 IA T & DJ E N Beactor trip BF when operator bumps botton A06705 030201 08/16/1967 09/18/1967 8 IA 00 T B EB D 5 Reactor trlp on CD BF setpolat drif t of MSIT lou

, pressure switch A06704 030201 08/17/1967 09/18/1967 8 EB G B BD D E Reactor trips on BF in a d ve rt en t groemdiaq of vital bus A06706 030202 09/20/1967 10/25/1967 B IA T B BF D 5 Beactor trip on in ad ve rten t clearing of pe rmissive switch A06707 030202 09/22/1967 10/25/1967 8 NE 00 B AY G S3 In ad ve rt e n t BF opening of alt 10 steam dump valves causes reactor trip 406108 030203 10/05/1967 11/17/1967 B EB G B ED D C7 Beactor trip due

&& BF to loss of vital bus powe r su pp ly A06709 030203 10/10/1967 11/17/1967 8 IA P B BF B B Bapid load reject signal f rom relay trips reactor during load reduction A06710 030203 10/33/1967 11/17/1967 8 EL) G B ED D C7 seactor trip d.e AA SF to loss of vital

[h'Q- bus powe r supply

=

\

Table B2 Coding Sheet for Reportable Events at Haddas Beck-1967 - (Continued)

~

Nu::bef~~~~lccession Event Report Fiant Component Absoraal SkalficaEI Number Date Date Status System Egulpment Instrement Status Condition Cause Category Comment A05718 030203 10/15/1967 11/17/1967 8 In 8 B BF D E Beactor trips If dee to high Sus signal AO5712 030203 10/16/1967 11/17/1967 B EB G A ED G C7 Beactor trips BF dee to inadverten t grounding of vital bus dering salatemance A06713 030203 10/25/1967 11/17/1967 B 'Tg A1 H B EB D E Beactor trip on 3 BF blown fase la power supply A05714 030204 11/05/1967 12/19/1967 B EB G A ED G C7 Beactor trip d ue BF to inadvertent grounding of vital bus daring mainte na nce 406715 030204 11/18/1967 12/19/1967 8 EB G A ED G C7 Beactor trip due BF to laadvertent grounding of vital bus while changing beibs 106716 030204 11/24/1967 12/19/1967 B EB G B ED 3 C7 Beactor telps BF des to maintena nce error while checking lasta11ation of ne w wi ta l bus 9

e (55 .

Table E2 Codisj Shset far Rsicctchls Evasto at Medics Neck-1918 - (Contiomed) ,

" ~

Nuthor Ac ce ssion Event Ptpart Picat Casp4Gert A basr=ol Sij3ificancs Numter Date Date S tatus System Equipmen t Instrument Sta t us Condition Cause Category Comment ,,

406802 041938 31/15/1960 02/09/1968 D HB 00 C BB D C dl two steam isolation valves fail to close 406001 041833 01/15/1960 02/09/1968 B IA && & BG G N Baintonance on BF power supply clears permissive, causing reactor to trip A06803 041e42 02/07/1960 03/12/1968 B CF FF B /9]F D N Escessive DD leakage of BMS pump seal 106804 041842 02/14/1968 03/12/1968 8 EC && & BG G N Loss of control

,BF power to RCP gover bus causes reactor trip 406005 041842 02/20/1960 03/12/1968 2 I&' jh B D N Failure of rod BB position coil stack causes rod bottom alara 106806 041641 31/28/1968 34/11/1968 B IA 8 B BF D N Inadvertent air into low vacuos turbine trip actua tor causes reactor trip A06607 030590 04/27/1968 05/14/1968 8 En F P B AT G S3 maintenance crew

{}ff 57 inadverten t1r tripped site feeder breakers.

All three DGs loaded then tripped off.

Total station

. blackout A06803 041846 05/04/1968 06/12/1968 8 In F C BF G N Beactor trip during test of power range trap set points A06dO9 036295 06/09/1968 07/12/1968 'B RB J B BF G W Beactor trip gg when worker deenergized rod gripper while changing lobe

, oli A06810 036295 06/10/1960 07/12/1968 8 I& E B EF D N salfunction of BF' flow transmitter lg f caused reactor en trin

Tcble 82. Cadiz.J Shact fer angertebis Eisnts at esdico unck-1968 - (certimad)

No bar Accession Ersat 32 Picat - CCfpStoSt Abn3rCAL SiJIific5aco Number Date Dakott e States Systes Egelpeent Instrument Sta tus Condition Cause Category Comment 406811 041840 07/16/1J68 08/13/1968 B PC DD B AU D N Tolosse con trol JJ tank gases leak back into primary water guaps A06812 041839 08/02/1968 09/12/1968 5 NE NN C AC D N peactor trir 00 BF ductag test of j turblae stop l valve 406813 041039 08/09/1968 09/12/1968 3 NA NE ($ AO- E C7 Beactor trip 00 BF during reprogranaing of turbine control valvos due to i installation error 106014 041839 08/22/1968 09/12/1968 9 RR 90 B BF D N Loss of control air closes feedwater rejulatin3 valve and trips reactor A06015 030206 10/13/1968 11/19/1968 E 25 J B BC D N Control rod 016282 slips in 42 steps A06817 030208 11/18/1968 12/20/1968 E NA M5 T C AG D N Beactor tripped SF when metal fillags caused binding in turbine control switch A06316 030208 11/18/1968 12/20/1968 8 IA 8 BF D C7 Beactor trip -

no apparent cause e

106810 030907 12/07/1968 01/14/1968 5 EB G B AC D C7 Vital bus (1/4) l 11 fails due to had transformer in inve r ter A06820 030907 12/09/1968 01/14/1968 E CB DD P C AL G N Coil leads for time dolay relays on RCP -

tound disconnected

• A06819 010907 12/09/1968 01/14/1969 E IA B BF C7 Beactor trip - -

no known cause (see A06816) ,

Table 82 coding Sheet for Bogottable Events at Raddam Neck-1968 - (Continued)

Musber Accession Event Report Plant r;osponent Abnormal Si jaltica nce N umber Date Date Status Systen Rguipment Instrument S ta tes Condition Cause Category Conawat A06821 030907 12/17/1968 01/14/1969 E IA E BF C7 Reactor trig -

no apparent-causo 106022 030907 12/18/1968 01/14/1969 B RB J f C D N Rod control salfunction during critical approach due to relay failure A06023 030907 12/25/1968 01/14/1969 5 IA B BT C7 seactor trip -

no apg>aren t cause 9

- 9c5ns 68

- - - - - - - - - --- --------~

ct'oSlej S& cot. (for Ragartebiz Ev!nts at Hiddr Neck-1969-(Continued) su 55r--- A5ces3In:T Event --~~ iRFirt Plant cotec= sat Abanraaf SIsIIIIcenes Musber Date Dato S tatus Syst0D Egtip2ma t Instrese3t . S ta t us Conditica Cn==e Catogsry Cac00st

= - - - - - - - - - - - - __

Ao6901 019269 01/06/1969 02/11/1969 D EB LL D BF D C7 Inverter f or vital bus fails, causing reactor trip A06903 039269 01/08/1969 02/11/1969 B HA MM B AD G N Beactor trips NN BF due to oil leak of terbine i

control systes.

Line broken by maintenamac .

A06902 339269 01/08/1969 92/11/1969 8 IA B BF D C7 Beactor trips Ef due to spurious i

noise, source unknown A 069 04 03/25/1969 04/18/1969 B HE QQ B AL D C7 Reactor manually BF tripped on failure of 1/2 turbine stop valves A06905 034862 04/12/1969 05/16/1969 8 RB J B OA N Rod drive inoperable (no apparent cause) 314863 05/01/1969 05/16/1969 B Ed N C C OE H k Diesel genera tor fails to run during test due to procedural error 039001 05/01/1969 06/01/1969 ND JJ B OD A C3 Storage drum moved dear the fence (700 aces /hr activity at fence)

A06906 039231 05/03/1969 06/19/1969 B PC BB /h B AH H S8 Boron recovery OH tank ruptures due to improper valve lineup Ao6907 039231 05/06/1969 06/19/1969 B MA Z B Av D 58 500 gal rad OH liquid spilled AD on boron recovery area -

broken pipe ,.

Ao6900 0 39 229 06/10/1969 07/25/1969 B HH 00 B AD D S7 High level in

//g gp steam gen 83 due to broken ,

f eed v ate r I regulating flow cc atrol valve - -

m- LG RK f m ems Uce ksacrtab13 Eveatc at H:ddg3 Ncck-1969-(Continued)

- ~ ~ ~

SuillEr~~~~Icc335 Ion ~57Ent Sepon Picat Cacpanant abastcef Number D5to 04te St&tuS Syct03 Eqtiprea t I ctrument S ta t es SI S TIcanco Conditian Cctso Categsry Ccceent

____________________ _ _ = - - -. -

.

• c A06909 336147 07/15/1969 07/15/1969 8 PC C B BG H 57 Loss of offsite EA N OK JEF power,1 DG ER DD fails to cua, I charging pump fails to run 106910 0 39046 08/02/1969 09/11/1969 B EA f P B BG F S7 Complete loss of g j: offsite power due to lightning strike on telephone relay 1069'? 0 19046 08/18/1969 09/11/1969 B RB J P B AS L M Control rods BF inadvertently drop into core due to relay fallare -

reactor shutdown' A06912 019046 08/20/1969 09/11/1969 B ID G B EG D E Delta-T indicator f ails in loop 4 - unit replaced A06912 039046 08/29/1969 01/11/1969 B ID G B EG D N Del ta-T indicator fails in loop 3 - unit replaced 106913 0 39067 08/31/1969 09/11/1969 8 RB I C AG 019046 J D E Stuck control rod, no cause reported, rod functioned normally after cooldown 106914 0 39345 09/07/1969 10/13/1969 B RB I C AG D E Two control rods J

steck (no cause reported)

A06915 030807 11/11/1969 12/22/1969 8 NA DD B BK I& BF D N Failure of pump B in turbine oil AY system. Beactor, BE shutdown by +ubME.

control system when ope ra tor inadvertently opened a relief valve 106916 038807 11/12/1969 12/22/1969 B H& {y) 1 4G D N Failure of pump in turbine governor -

[$O impeller slee we

Tal. lo 11 2 codinj stest far Warsrttblo Evaa*.s at Ilidd:c Nr.ck-1969-(continued) liu~siiUr"~~~IEEEEsI3n~ E ve nt ~~~ Report Plant Component Abnoraaf SI] Ell'lcance N um ber Data Date Status System Eguipseat Instrument S ta tus Condition Cause Category Comment AOb917 330087 11/18/1969 12/22/1969 B EB P b ED G N Protective BF relayinJ trips plant - caused by noisture in relay box A06919 038887 11/28/1969 12/22/1969 B HE MN C BK D N Left hand 00 turbine stop valve tails closed dus to low pressure in valve hydraulics A06910 038887 11/28/1969 12/22/1969 B SB J P D AS D M Two rods dropped due to taulty relay. Also loss of rod control due to faulty capacator A06920 030007

• 11/29/1969 12/22/1969 B SB J 8 BC D N Control rod out of position (no cause reported)

-k e

t I

e Ibl -

Tabla 82 c: ding sh2ct fer aspet:blo Evan*.a at Ridd;Q Ncck-1970- (Coninued)

~~ ..

Sim5er Accession Event leport Plaat

~~~ ~

Domponent Abnotaal ~~

SigallIcance number Dat e Date Status Systen Egelyneat Instrument S ta tus Condition Cause category Comment A07001 044800 03/21/1970 04/21/1970 B BB J T B AS D. N Five rols slip into core -

faulty microswi tch A07002 344800 03/22/1970 04/21/1970 ta E JB B EG D N Reactor trips on BB BF erroneous signal (no cause reported)

A07003 044 864 04/02/1970 05/27/1970 B SS J P B 15 D C7 Rod subgroup drops in -

caused by faulty contacts on relay.

AO7004 044864 04/14/1970 05/27/1970 B IA L B BF D N Automatic reactor trip during power range test. (N o i cause reported)

A07005 044864 04/20/1970 05/27/1970 B CB 1 SC

/J As D N Air leak f rom containment through RCP seal water retura system (no cause reported) 107006 347866 06/21/1970 07/27/1970 B CB CC B AU B N Steam leak from FF pressurizer due to gasket failure - gasket not rated for application 107007 058514 08/17/1970 09/15/1970 B AC G B BF [I N Reactor trip due to lightning strikes on telephone lines A07008 358514 08/19/1970 09/15/1970 B CB DD B S4 BY D S7 Small fire near '

RCP 84 - leaking oil from pump was ignited by hot piping A07009 357469 10/12/1970 11/24/1970 B IA QQ 5 BF D 57 NB /3 Reactor trip on BB S 6L erroneous loss EG of flow signal -

steam dump valve

/f 1L- failed to close

- nnesesn..e n.. e

T: bis d2 Cading Sk30t frr R0[trtablo Ev3bts at HLdd 3 Meck-1970- (Continued)

~~~~~ ~

Narber ~5$ cession Event Repor[ Plant Component Absoraal Significance uusber Date Date S tatus System Eguipment Instrument S ta tus Coedition Cause Category Coenent A07010 057469 10/24/19 70 11/24/1970 D ID I C ER D M Setpoint drif t CB in pressurizer level aoattors 107011 057469 10/27/1970 11/24/1970 B ID I C BF G E Balatemance crew C8 trips reactor during work on presserizer levet sensor 1

l 1

e O

e t&T -

~.

Tabla d2 Ccdi:3 Shoot far R31crteb13, Ersats at Hrddio 5cck-1971 - (Continued) ,

~ ~

Eumber~ Ecossion Event Report Plant Component Absoraal Sihificance

~~

susber Date Date Status System Equipment Instrument S ta tus Condition Cause Category Comment 363110 04/18/1971 05/10/1971 C CA FF 5 OD M C3 Radioactive OJ lodiae released OG due to operator error with penetration seals 068401 08/01/1971 08/20/1971 B PC 3 B E N Raptured 40 expansion jciat due to poor velds. Wald la L8 heater-draias piping 066476 08/21/1971 09/21/1971 8 CR 5 8 AU D N Leaking tubes la feedwater heaters 067210 09/07/1971 10/19/1971 B EB G C 8F D N Homentary groemd ED on vital has causes reactor trip 340653 12/02/1971 01/24/1972 B CE 00 C D u Solenoid valve A1 failure causes feedvater control valve to close 107101 12/02/1971 01/24/1972 B I1 P C AG D N Reactor trip relays blad due to grit 4

i Tabl6 82 Cadi:g Sh:<t he sagerteb12 Evtsto et andd:s a:ck-1972 - (corti:ued)

Werber Accession Event Report Plant Component Abnormal Significance Humber Date Date Status System Egelpaest Instrupeat S ta tes Coedition Caese Category Comment 407201 069324 01/18/1972 01/28/1972 B SC 00 15 B D 5 Containment leak 80 dae to cracked "

yoke on exhaust bypass isolation valuw 055283 02/27/1972 03/23/1972 B EB G C BF G N Test personnel R3 cause partial loss of orsite  ;

Fower, reactor trip 107202 05/19/1972 06/23/1972 B PC E C OG H C3 Unplanned airborne release DJ from desimeralizer due to Operator error 072845 06/10/1972 07/01/1972 3 11 L 1 BF G E Reactor scrans DJ vhen maintenance i saa grounds -

instrument 076391 09/21/1972 10/20/1972 B EB LL B BF G s Beactor trip on BG loss of power to ,

semi-vital bus f rom transformer  !

076191 09/21/1972 10/20/1972 5 EE 00 8 D E After reactor BB trip, left side k terbine stop .

valve fails, due .

to steam cattlag i A07203 075600 10/05/1972 10/13/1972 B SD 00 C D C7 1/3 containment '

00 Ba isolation valves -

fail to close, -

due to leaking solenoid A07204 077213 12/19/1972 12/28/1972 e SC 00 8 D 5 Contaissent 4 AT Purge ethaust ,

bypass isolation  ;

watve fails ,

open due to -

crack in yoke (see 107201) +

$b

Table B2 Coting Shsot fer Ralert:bla E7en'.s at Middts Ncck-1973 - (Continued)

• Eurbor AcceErios Esset R;ptrt Pic2t Carp 2naat Aba'stmal Sig ificanco N um ber Dat o Dtto S tatus System Eguipment Instrument Sta tus Condition Caqso Category Comment .-

Ao7301 078302 01/11/1973 01/22/1973 8 IA CC I C AA D N 1/3 pressurizer CB .

g level alars fails due to failure of light source A07302 079103 02/22/1973 03/02/1973 5 IA CC A C 4A D N 1/3 pressurizer CS pressure sensor fails due to bad photodiode A07303 080132 03/26/1973 04/04/1973 3 38 N C C AL D N DG' load fluctuates during test due to loose governor Ao7305 081508 06/01/1973 06/22/1973 D SD 00 C AT D N Containment AT 1 solation val ve leaks, is replaced A07304 081514 06/02/1973 06/04/1973 5 00 8 AT D C7 1/3 reactor PC At letdown valves

{same as 407203) f ails to close due to boric acid around valve packing Ao7306 082978 06/21/1973 06/22/1973 8 PC 00 C OG D C3 Uaplanned A dd airborne release due to leak in purif sca tion system valve.

valve had faulty diaphrage.

Valve replaced AO7107 081867 06/21/1973 07/03/1973 D 00 jQ BI & C3 Unplanned .

PC ON release of rad 11guld when letdova systen placed in servico due to procedural error A07308 084215 09/01/1973 09/28/1973 8 NB C AD E C7 Steam gene ra tor EK 04 0 seismic support hold down bolts fail A07109 084871 10/04/1973 10/18/1973 B SC C C BC C C2 Ta mers f or EE 54 service wa ter

, pumps and cont.

air f ans f ail

/ s,e0 during DG test

s Table 82 codinj Sheet for tegottable "Even*.s at Haddas Neck-1973 - (coattaued)

Numbut l' cession Event Repor t Plant N um ber Date Component Aboormal Significa nce Dato Status Systen Eguipseat Instruneat S ta tus Coadition Cause Category Comme nt A07310 084874 10/07/1973 10/19/1973 8 EB 11 T C 88 8 C4 Power supply fails to e

transfer dae to desiga error A07311 085506 11/01/1973 12/13/1973 3 Sr 00 a 18 D C3 Valve on aus?

JJ ON thermosipnoa 3eaks radioactive water (bad diaphraga) into local stora newer a07312 387101 11/26/1973 12/07/1973 3 NB C /hh E C7 Steam gene rator EK 01 a holddova bolt failure 087305 12/01/1973 12/21/1973 a OD & C3 overeuposure of two sea (gae rterly readings of 3.0 3 and 3.66 rea)

A07313 088104 12/28/1973 01/07/1974 5 dO 00 C 88 D u 1/4 RSIV fails to operate, gland packing too tight O

e (67

Tablo 82 Cading Sheet f r Rsytetab13 Ev30tc at N:4420 N;ck-1974 - (Cortinued) *

~~' ~~

35Eb5r'~~~ A3 cession Evott R2 port Plcat sunber Date Date Carpscott abattnal Status System Egelpment Instressat States SiglifIcanca Coaditica Casse Category Comment *

• 807401 088083 01/01/1974 01/28/1974 3 IC y C D E Fa11ere of a BB core cooling lattiation timer to load DGs A074 02 088084 01/18/1974 01/28/1974 5 I4 T E B BF D C7 seactor trip dee as I to frosen lastrument sensing lines (mait heaters not operational) 407403 088451 01/19/1974 01/21/1974 5 E& 5 P S BD as D 57 Total loss of 54 I g3 of f si te po wer dering ice store

- Se peeps on a DG did not start automatically -

loss of line dee to incorrect relay 1ag A07404 089348 04/01/19'74 05/12/1974 D PC e C &C Sif 00 D s Degradation of air charcoal filters (filters replaced) 407405 089744 04/04/1974 04/04/1974 D 00 C BB e 3 PC Failure of reactor letdows stop valve to close completely (packing too tigtil 107406 390648 04/04/1974 04/10/1974 D SA PP C AT D s &acessive leak 40 rate in contatement penetration check valve (grit on valve stem) 107407 091668 04/26/1974 04/29/1974 5 FC 00 B OG D C3 caplassed radioactive release from aux building exhaast faa dee to leak la diaphrage in hydrogen supply regalator A 074 08 093699 05/06/1974 05/06/1974 5 35 C ~BT B C4 DC excitation LL transfoceer fails (improper grounding scheme)

I6f

Table 52 i

coding sheet for negottable 5: eats at saddam sock-1974. (Coottaued) t seabor Accession Event sunber Date brort Plant Cosponent Ahmeraal

~

Significance

~

Date States system Egelpaent Instrument s ta tes condition Casse category Conseat 107409 092190 05/24/1974 06/03/1974 8 It L C 55 D 8 Setpoint drif t la accioac overpower trip lastrementation AO7410 393792 06/14/1974 06/24/1974 5 d@ 00 C AC D 5 asIt fails to BB close dee to dry valve pacting A07411 09437Y 06/20/1974 06/28/1974 5 CS CC E C 35 9 N Setpoint drif t IA T la pressere switch

• I

/G .

Table 52 Codtag Sheet for begottable' sweats a t saddam sock-1975 - (Gatimad) insbei"~~~45cessica sweat soport Flaat u ue ber Date Date component Ah'n ormal States System agalpaeat Instransat Statae Sigallicence Coadition Canse Category Commen t 407501 101154 03/26/1975 04/01/1975 s Cs ~

00 s aI SF e s acs VAL VdI #AtKIN6 glaae fails, reactor skatdown a07502 103051 05/21/1975 05/27/1975 1: se II &

l Es j} Se /9 C7 Steam generator a hotadosa bolts fail 104202 05/26/1975 06/30/1975 C Cs EE & AL 3 s Seismic CC BC 0 restraints (3/a) on press:stser

act properly icatalled 104049 07/02/1975 07/02/1975 C CE EE C As e e Feedwater hanger II a pipe attachments brokea 407501 106 348 Os/08/1975 09/05/1975 s es & C 01 e a missolved osygen IS across plant cooling water ascoeds limit A07504 108249 11/1s/1975 11/25/8975 s y

$s' a OJ e s Service water to O

costalaseat faa cooters laterrupted dae to operator error 1

/78

TIbis 82 Cading SbCt fir R3]ortab17 Events at HIddse ucck-1976-(Continued) suiSEr'~~~i3c3Ei[3a Event Report Plaal Component abnormal Status Systen Egulpaent Instrument Sta tus Condi tion cause SI]allica nce Number Date Date Category Comment lea 7604 110359 01/01/1976 02/05/1976 Rr F B or D C7 High fish impingemen t rate on latake screens LER7603 110941 01/08/1976 02/02/1976 D I& L C EN D N Setpoint drif t la overpower channel lea 7602 110933 01/22/1976 02/03/1976 O d s0 00 T C An I u RSIT 00 se B8 malfunctions due to frozen air opera tor lek 1601 110940 01/22/1976 02/02/1976 8 NS E B OE I B Sensing lines on It Bu G Luo steam flow Br sensors froze L Ek7605 111649 01/24/1976 02/20/1976 B PC *DD E D M Charging pump FF AT outboard seal leaks due to 0-ring failure LER7606 111602 02/03/1976 02/23/1976 B EE 5 C OK G N Energency diesel BF tr1 PS 'h*"

calibration tool is left in unit L ER1607 111603 02/19/1976 02/25/1976 8 11 L C EN D W Selpoint drift in overpower instr u ne at ation L EH7608 113549 03/30/1976 04/29/1976 8 58 ghg B OD E C3 Unplanned OG release of radioactive gas (11.59 Cil - two rupture disks in vaste gas ducar taan 4%:/ cbse do damage during insta lla tion L EH1609 113548 04/03/1976 05/03/1976 B In E 5 n EG I N Steam line break low flow alare received due to chanie in weather affectie)

L ER7610 instrumentation 114170 04/25/1976 05/24/1976 D EC C C D EC D N Dattery charger '

goes into overcharJe due -

to faulty contact on timer .

Table 82 coding Sleet for negottable Events at Haddam Neck-1976-(4betimed)

~~

535EEi'~~~i3cessio$'3 vent Report Plant Composant Absormel Significance u ual.er Date Dato Status Systen Equipment Instrument Status Condition Cause Category Commen t LER7617 116272 07/2G/1976 03/03/1976 8 NB B E B BK E N Steam line flow G AT indicator fails due to leak in i

sensing line -

leak caused by cross threaded titting i LER761b 117665 08/19/1976 09/03/1976 3 SF B S C4 Error discovered is a.ssumption la vendor ECCS analysis ETs760s 118793 10/01/1976 10/13/1976 NF F 5 0F D C7 Nigh fish ispingenea t rate on intake scree ns Lea 7619 I l9 51's 10/01/1976 10/22/1976 3 PC D 8 AB D N Concentrated DD AT boric acid pump fails due to bearing failure L ES7701 121030 12/14/1976 01/14/1977 3 NI E 110100 3 15 D C3 Small leak ta

1. A RF 11guid waste NA 05 line adjacent to area reported la lea 7613 O

9 D3 '

Tabla 82 Cc: ding Stoct tr tsgretab12 E?osts at Midd;o ucck-1974-(&mtimed)

NuEber'~~~i3cozzios EYeat Report Plant Component Absoraal dueber Date Date Status Systen Egelpaent Instrument Sta tus Sidalfica nce - c condition Cause CateJory Commen t Lga7611 05/10/1976 06/14/1976 B S M-C 00 C at D s Lov discharje pressure on ama feed pum p (s fer6A[ driven) due to laadvertest opening of steam supply valve L ER7613 115876 06/15/1976 07/12/1976 C BI 1 8 As D C3 Small leak la 51 05 vall of safety lajection .

cubicle due to deterioration ot steam generator blowdown piping L ER7612 115450 06/17/1976 06/24/1976 C CF S B SG E N Power lost to BB DD BNR pe op-o pe ra tor overloaded bus for pamp Len1614 115875 06/24/1976 07/09/1976 C Et F B BF B 57 Total loss of CF SG of f si te ge wer due to poor design la protective relaying-kNS flow lost 3 times Len1615 1 16 274 06/29/1976 07/27/1976 C 55 00 8 AD D E Waste gas decay OG tank rupture disk fails due to pressere surge - 2nd time 1151.10 07/01/1976 07/15/1976 C A OD A C3 Possible overamposure of maintenance worker (quarterly readings: 2.19 ren dosialter, 3.11 ren tron badge)

LER7610 11621) 07/05/1976 07/27/1976 P SH-C DB A DK D S2 PP Two aux feed AT pumps fail to DJ reach pressure -

both pumps vapor bound due to fault y check l }P 1L-

t a blo 82 Codia) sheet for Regortable Events at Itaddas Neck-1977-(Coettmund)

~~~

id55$r'~~~i3ce55IoE~ Event Roport Plant component abnormal Significa nce numbur Date Date status Systen squipment Instrument S ta tus condition cause Category consent ETs770a 325397 05/25/1977 06/07/1977 8 NF P B 0F D C7 Nigh. fish impingenea t rate on intake acreens L Ek1109 125591 05/31/1977 06/30/1977 B FD L T B && D N Overhead crane control malfunctions due to wear of swatch mechaatse L EB1711 127017 06/23/1977 07/06/1977 8 ID G L 5 AC D N NacLear Er instrument detector current erratic due to deterioration of detector and cables lea 7710 126490 06/27/1977 07/06/1977 8 I& P 5 EG D N 8eactor coolant pg EE low flow relay fails due to open circuit la relay cott ET:;7705 126006 06/29/1977 07/07/1977 9 NF F 8 0F D C7 sigh fish impingemen t rate on intake scree ns LEH7712 le3528 07/05/1977 08/01/1977 5 I& I C EN D N Setpoint drif t in pressurizer level channel LEH7713 143527 07/19/1977 08/09/1977 8 IE ED LL U S EF 5 cy Perturbations in BL loop temperature lastrumentation caused by failure of static inverter la h8.gk temp envitomment LEH1714 1435:6 07/20/1977 08/09/1977 8 IA II B d && D /V Steam generator I EG marrow ranje

• level trannoit be r fails due to amplifier

['[jI failure

• Tablo M2 Coding Sheet for Regottable ivoats at haddas sock-1977-(cutimad) " "

~

W uatear AccessioA~ Eve nt Japort Plant component Abgereal Sijaificance 4 ua ber Date Date Status Systes Equipment Imatrument S ta tus Coadition Cause Category Comme nt ETS1701 122205 01/12/1977 02/09/1977 8 NF P 8 0F D C7 Nigh fish impingement rate on intake screens ko7702 123001 02/24/1977 03/10/1977 8 N& JJ C AB E 5 Leak in recycle 0 & LJ test tank due to corrosion (1000 gal radioactive water spilled) rT57702 123054 03/09/1977 03/31/1977 5 NF F S OF 4 C7 sigh fish impingement rate on intake screens lek 7705 124333 04/04/1977 05/03/1977 3 PC 89 3 &Y E N Decrease of BE charglag pump discharge pressure due to cracks la shaf t 80770s 148001 04/13/1977 04/14/1977 3 PC 1 110101 B Adl D N Leak in boric 15 acid peep

. smetion piping due to chloride stress corrosion LEH770t 125033 04/26/1977 05/12/1977 3 PC DD 8 &T D E Small weep fouse tr in charging pump seal (la) houstag due to porosity la satorial L P.R7 707 125034 04/28/1977 05/12/1977 5 PC DD 5 SL G N Bearing fails in D BC charging pump la doe to misalign me nt after saintenance ETS7703 125015 04/29/1977 05/12/1977 3 NP P B 0F D C7 Nigh' fish 1spingement rate on intake screens L ER1706 125535 05/23/1977 06/22/1977 5 SC D 8 && D N Bearing fails on E AP contaissent air circulation fan j7 due to moraal

, wear

Table B2 Cadtag She:t far sagnetsbis E7ents et Cadd2n 3:ck-1977 -(continued)

~~

Nulbur~~~~5cces5 ion Evect Deprrt Fisst Carpseett Abarreal

~

Sijsifica ncs M us be:r Date Data Status Systen Eguipseat Instrument S ta t us Conditloa Cause Category Consent lea 772) 141420 10/03/1977 10/14/1977 3 as Og g D N Seactor power lin 81 level mones taril y escoeded due to slow response of turbine control valve ETS7707 130910 10/19/1977 11/15/1977 8 NF F B 0F D C7 Righ fish leptagement rate on latake screens EFS7706 113614 10/28/1977 01/13/1978 3 BA D C3 Trities activity 0F in river water OD sample was high L EH7728 113613 10/31/1977 01/20/1978 C FC 3 5 8 10 E 5 Boric acid leaks IA T onto pressurizer pressure switch due to veld failure L Ek7727 149386 11/03/1977 12/06/1977 C 23 I C AO C C7 Bod cluster control spider assembly vahe separates from bub due to faulty brare jola t .

ETS7708 130919 11/04/1977 11/17/1977 8 NL DD a 05 G C3 A river effluent BA F somitor sample pump was not operating while 21 tanks were drained LER7725 144201 11/14/1977 08/25/1978 8 .c5 GG OA C D s right saubbers 3 AW for feedvater piping failed tests due to fuel oil ta saubber LER7726 144127 12/01/1977 12/05/1977 D IA 00 B AN G N Wiring errors in c) G SF RSIV valve circuitry, could have prevented a reactor trip.

Caused by salatemance error .

17 7 -i

T2 bis 82 Coding Shact for R2gart:513 Brsato at Hrdd:a Nick-1977-(Cutiewed) '

suiEir~~~~Ic6eli[EE Ev33t PJpart Pictt ~~

sustier Dato Date C3tPunert absurnaf-~' ~ '~~Illallicasca Status Systen Equipment lastrumeqt Sta tus Condition cause Categogy Comment .

lea 7716 143416 07/30/1977 08/29/1977 8 FC DD l

C A ll C 5 Pinhole leak in 00 charging pump l

bypass salve due to faatty casting ,

L ER7715 143544 08/01/1977 08/18/1977 B SR-C DD C BE D s Aux f eed p um p PP AT pressure below BB limit caused by leaking stuck open check vales L Ek7717 143417 08/11/1977 09/06/1977 B NB 3 5 44 D 5 Steaa leak on SG 00 15 level settling pot vent pleg valve due to steam cutting L Ea7 718 143525 08/20/1977 09/20/1977 3 IA R C AQ D E Brs low flow SS trip matrix relay fails to operate due to grit lea 7719 143418 08/21/1977 09/20/1977 5 CB De 3 43 D S3 BCP loop 1 seal FF AT fails [chippia on seat face))

during operation, causing other 2 SCP seals to fail lea 7720 143419 09/16/1977 09/26/1977 3 4h 00 '

C 4G D C7 RSIT fails to PF move during test due to binding of valve packing gland LER7721 144204 09/18/1977 09/19/1977 3 NB 00 8 OD D C3 Deplanned OG releasp of 7 44 C1 Nd**Cf/V8-gas-d iaphr a ja ruptured in vaste decay tank LER7722 143524 09/21/1977 10/14/1977 a 11 L C :H D N Setpoint drift la nuclear overpower trip chaemel t14

Tabla 82 Codi:3 Sh2ct fic 82gsctibia Evasts at R:ddio 53ck-1978 - (continued) l I

ustbst Acceesion Eeur.t se Plcnt CM s cat Ahaarnal 31 alficance b unaber Date Dahrt States System Eguipment Instreneht St tus Condition Cause hategory Comme n t '

lea 7806 138369 05/08/1973 05/19/1978 8 EE 8 8 EA B C4 Potentini exists for diesel y generators to be overloaded under  ;

certain LOCA [

conditions (

(utong design i assum ption s)

• p L ER7 80s 138939 05/09/1978 06/01/1978 8 FC E B RF D u Leak found La -

AT charging pomp rectre. bypass '

valve due to erosion [Puey 8) t LER7812 141030 05/31/1978 06/21/1978 B PC I C 10 E C7 Leak found la 00 As charjing pump pressure gauge  !

isolation valve '

l due to veld failure [ Pump B) f LER7009 140558 06/05/1978 06/12/1978 8 CS CC & AT D E Leak tound la f SS pressurizer 6 spray isolation valve ETS7806 140556 06/06/1978 06/19/1978 3 NF DD 8 0F M C7 Bate of change a:r of discharge tenge rat ur e '

exceeds limit LER7010 1 39504 06/13/1978 06/21/1978 5 SR 0 C of E N Electrical G penetration FF fails LOCA test j due to landeguately sealed cable end tot LER7813 139025 05/17/1978 07/05/1978 8 gg G A AC D N Hole found in E ED cable sheath for '

CAR fan motor due to electrical arc LEH7811 141029 06/18/1978 06/21/1978 D SN GG C CA E 5 Hydraulic g3 I saubber uncoupled from '

feeduater Piping g

-f 11*l

~

Tabis 82 Coditg Shact for Begsrtablo E7sato at crddan GCck-1978- (Contiamed)

~

• uuaber Accession Event Report Plant Component Abnormal Number Date Date 31Jaitica nce Status System Egelpeemt Instrummat Sta tus Condition Cease' Category Comment r

LER1801 1349f 0 01/26/1978 02/10/1978 3 ZE S C 01 8 C4, Terminal clocks f ail to . pa ss gealification tests +

ETS7801 835892 01/29/1978 02/16/1978 EF F 5

, 0F D C7 Nigh fisk  ;

impingemen t rate on intake screens ETS1802 136380 02/23/1978 03/09/1978 SF F 3 0F D C7 Nigh fish impingemen t rate on intake scree ns LER1803 137248 03/97/1978 03/31/1978 a NI I g. C7 B Leak in steam 34 A0 cf generator 54 BC blowdoun line due to a sold failure caused by tuteraction of hot steam and cold service water LER7802 137340 03/22/1978 04/04/1978 3 se e C '

s C4 tiectrical 08 degradation on terataal block due to interface with aluaalaus bos ETS7804 137853 03/29/1978 04/26/1978 E,F F B 0F D C7 Nigh fish impingemen t rate on intake

, ecree ns LER7805 137854 04/03/1978 05/03/1978 8 S F-C FF A ST D N Cracks found in DD wear rings and tapeller rings of NPSI pumps L ER1804 138255 04/28/1978 05/12/1978 8 FB EK d DJ B C4 Pressere buildup in anautar poison cavity of sient fuel racks. Design deficiency LER7dO7 118797 05/0s/1978 05/26/1978 8 PC a ur D u Leak found in 60 AT charging Pump j recirc. valve due to erosion O[ (Pump B) .

T2 bis 82. Codi:g shoot far asgrrt bis Evrats et Midden W(ck-1978- (Cestimmed)

' ' ~ '

uusher Acesssics Evett R: port Finat campement Abstrcli Sigiificance Number Date Date States Systen Egelpsent Instrument' S ta tus Conditica Cause Category Connea t ETS7809 141286 09/29/1978 10/13/1978 NF F 5 0F 3 C7 Nigh fish lapingemen t rate on intake scree ns lek 7820 141853 10/04/1978 10/17/1978 B Cs CC S && D E treasuriser PC T 41 levet and Lressure crease due to fattere of charging flow cont ro11er suitch ETS7817 151479 10/16/1978 04/02/1979 3 EF F C 0f -

D C7 Rate of change AD of discharge DD . tempe rat ure exceeds limit ETS7011 141812 10/31/1978 11/16/1978 RF F B 0F D C7 Nigh fish impingement rate os intake screens .

ETS7812 142220 11/02/1978 11/30/1978 D EF BF D C7 Reactor trip up caused rate of change of discharge temperature to escoed 11mit ETS7813 142646 11/12/1978 12/13/1978 NF F OC 8 S 5 Fish impiajenent or saapie accid entally discarded L ER181e g 146631 12/01/1978 01/12/1979 3 NF F B OF D C7 Nigh fish impingement rate on intate screens RTS7H15 E 146630 12/15/1978 01/15/1979 8 NF OF D C7 Rate of change of discharge tem pe rat ore escoeds limit -

five times LER7821 144618 12/27/1978 01/27/1979 5 BB J C OC N u gineekiy rod motion test not completed on et

! h

Tab 13 82 CodL:g Sh::t fre R2gertsb12 strats at N dd;o 5:c&-1978 - (Contimad) e

~~

i Number Accession Event Report Plant Component abnormal

~

Sijullica nce j ' N um ber Dat e Date Status Systes Equipment Instrument Sta tus Condition Caeso Category Comment

{

l lea 7814 139810 06/18/1978 07/05/1978 8 31 E C

( AT G N Containmen t electrical

! penetration j

' leaks as a result of malatenasco lea 7816E 146629 07/01/1978 01/15/1979 a ur e a or D C7 sigh fish layingemen t rate on intake screens

( L ER7815 139901 07/11/1978 07/28/1978 5 I& E B A& D N Drain valve on 00 A6 8t**" 11"'

sensor leaks due to moraal usarout L ER7816 139902 07/24/1978 08/04/1978 8 SF-C DD A av D C7 Cracks found on NPSI pump la shaf t sleeve ETS7808 140769 08/01/1978 09/20/1978 D EF- 0F D C7 Bate of change of discharge tem pe rat ure exceeds limit LEB7017 139957 08/04/1978 08/10/1978 3 CB U, 3 D N Decrease la 00 pressurizer 85 pressere d me to EG failure of press urize r spray valve controller LER7818 140161 08/2f/1978 09/08/1978 8 00 A B OE N 58 FM radio caused IC R DJ G dropped rod alata, followed by failure of load rumback signal to ialtiate.

Failure caused by procedural error ETS7813 141456 09/01/1978 10/27/1978 8 BA B ON D C3 Trittua level la river sample exceeded limit lea 7819 141760 09/09/9978 10/06/1978 8 IA S E 8 AA D N High stoam flow 4.1 G BC indicator failed due to failure of power supply I60 f...

V Table $2 Cadi:3 shoct far Battetchio ETents at RIddas ucck-1979 - (coattamni)

Number Accession E ve nt Be et Plant * ' ~~

Number Date 9ake Component &baotaal Status Systen Egelpment Instrument S ta t us 51Juificance Condition Cause Category Coese n t fly 7901 154C72 03/16/1979 03/16/1979 C OC & C3 Escessive socke r CD doses over [

quarter due to leadequate I recordia) I procedures '

E757904 152295 05/17/1979 05/25/1979 pg 8 0# G C7 Hypochlorite was 00 0F inadvertently released to the ri ver lek 7907 150701 07/07/1979 07/23/1979 3 NA RE C AN E E

N fipe support in  ;

service water system f atis to meet seismic criterion LEn7908 152005 08/03/1979 08/14/1979 E CP EE C AL E E RRR pipinj 1 k seismic suppott missin7 due to 5 installation

, error LER7910 152183 08/13/1979 09/07/1979 8 C8 CC  !

a B At 00 EG D S7 Pressurater PCRT [

inadvertently opas due to light source failure Lea 7909 151999 08/31/1979 09/12/1979 3 EE D 0 3

AE C CY Potential probles with DG torbocharJer thrust oearin g lubrication (nanufacturer's error) lea 7911 152237 10/04/1979 10/10/1979 D M& EE C AL E N ~ 35 seismic S F-C 5 restraints  !

PC alssing due to j installation  !

error (3 systems i' involved)

ETs790S 153396 10/16/1979 10/26/1979 a 94 00 a or H C7 Mypochlorite was {

U inadvertently released to the river EEu?90b 154816 12/16/1979 12/26/1979 a as 00 I a 06 o C3 A icvel control P BJ C4 valve relay ,

failed resulting in 15.8 C1 _'

g gf3 relea sed in 10 min via the atack _

Ta ble B2 - coding sheet for kegottable Events at Naddas Neck-1978 - (Continued)

' ~

5NE~uf~~~'I5 cession Event Report Plant

~

Component absoraal

~

s ual.er Date Date $1 Jaificance Status syntes Eguipment Instrument S ta tus Coedition Canse Category Comment L Ek1622 146617 12/29/1978 01/22/1979 8 3D' 00 B OK N 36 Air supply NI MN BG & 58 S

valves to At E3 isolation valves blocked open, loss of blowdova capability and loss of containmen t high pressere isolation ETS7603 1372a9 04/06/1974 04/10/1978 un JJ B 5 C7 Discharge canal 0F . PM level te escoeded limit 05 e

e 9

• lit i

I

)

Table 82 Codiaj Sheet for Begortable Events at gaddam sock-1979 - (Coattaued) 1 N u mbe!E 4ccession Event Besort Plant Component Atmoraal ~

fiumber Date Date Status 3 2stem Egelyment Instrument S ta t us Si.Jaillcance Condition Cause Category Comment I L eatt 002 154453 12/24/1979 01/18/1980 5 EE G & 'E E LL Beutral leads Ok from DG to r N

transformer cut by construction i s

I t

(

9 e

f 175 -

A

Table 82 Ceditg Shict fer negartabla s;ents at Ridd:a ucck-1979 - (Certimed)

BimEir IcciliIii~legat- Deport Plail -

c.omponent IEa5taal sigullica nce uusber Date Date Status Systes Egalpment Instressat S ta t es condition Cause Category Consent ETS7902 150653 01/27/1979 02/12/1979 D Mr or D C7 Bate of chaaje et discharge tem pe rat ure exceeds limit 17S7901 150638 01/31/1979 02/02/1979 NF F B 0F D C7 Nigh tLab impiajemen t rate on intake scree ns LER7906 149251 01/18/1979 04/24/1979 5 SM-C DD C D M 00 L1A flow rate of SL aus feed pump BB due to NS overheating, steam bea tia j supply valve had f ailed open lea 7902 147341 01/28/1979 02/26/1979 D I4 4 E D AD D u Loss of reactor EG coolant flow G

ladicatin-J unit fails due to brokea signal lead LER1905 148745 02/08/1979 04/04/1979 D PC PP C D AT 3 tacessive loot fill check valve leakage due to degradation og seating surf aces lea 1901 150336 02/14/1979 07/24/1979 C R8 8 147340 02/14/1979 02/28/1979 C 19 D M Crack s found in f uel cladding.

All elements examined, rest ok LE87903 148568 02/5 /1979 31/11/1979 C kB DD 151370 02/15/1979 27/1t/1979 C 46 D N Diesel fire peer I se fails to start Ab (1/2) due to burned out coil in starter motor ETS7903 150992 02/19/1979 03/19/1979 C PC JJ B OD D C3 Radioactivity is refueling water storage tant reached 26.1 Ci.

Limit is 10 Ci LEB1904 148744 C3/10/1979 04/05/1979 C CB 1 B L Y, D D Leak in primary drain cooler line due to

• cracked pipe fittings

e Table 82 Cading Shxt fcr psgertibla a:0210 at Hidda N:ck-1980-(Coettaued)

_Nurbar Acce rsio2 Event R2 port Plant

~~

CarpineEt Abatrchl ~

Number Date Date 51 J aificance S tatus Systes Egelpeent Instrement S ta tus Condition Cause Category Coesent L Ea8002 c 158157 05/04/1980 05/12/1980 s1 y & OG & C3 Badioactive OE release limits N& escoeded when an ion exchanjer was removed from service and replaced L ERS 009 158569 05/16/1980 05/30/1980 D 51 FP C D 8 Containment leak AT rates exceeded 33 due to uneven check valve seating LER8004 E 158253 05/19/1980 05/28/1980 R4 5 & C3 Trittua activit y ON escecded release OK limit due to larJe amounts of processtag water for uponia]

l refuellag outage L EB0005 C 158623 05/28/1980 06/06/1980 00 s 11 a C3 AB E

& vaste gas l C4 systen valve OG opened releastag senon to the environment L EkN 010 160253 06/21/1980 07/04/1980 D SF-C EK C AL.

3 E D Seksmic pipe supports on HPSI lines missing -

act installed LER8000E 158255 06/22/1980 07/01/1980 C JJ B . & C3 activity in a OE raw water SF 0D storage taak exceeded limits LE88012 160219 07/17/1980 08/11/1980 D C8 FF C 15 a DD Ca BCP seal supply bypass valve bushing broken due to high torque setting LcRCOstel 170020 07/17/1980 03/12/1981 D CJ 00 3 AM H C4 & throttlinJ valve was designed for i moderato throttling but not low "

throttliaJ for

• which it was (g y .

- i

Tab 13 B2 Cading sheet far asgtetabla E7sato at ntdano u:ck-1980-(Cantimad) -

useber accessiso E7Cet 32 Port Ple:t b5dpsDc;t Abatraal Sig-ificanca

~~

sunber Date Date S tatus Systen Eguipment Instsenent S ta tus Condition cause Category Commen t --

LEna001 154454 01/03/1980 01/15/1980 B ER C i C B u DG load sequence O OA timers for .

se rvice water pumps and containment fans '

fail test L ER800 3 154452 01/29/1980 02/07/1980 8 EE N & R& B C4 Poten tial DG overload duria)

LOOP, LOC &, with SIAS found to exist L ER8004 154451 02/04/1980 02/12/1980 8 CB CC 8 a At D 00 C7 Pressurizer POPV EG ogen 2 aim. -

proba b? y spurious zijnal from press ure controller LER8005 155569 02/07/1980 02/29/1980 3 FC g.7 8 10 D 3 Boron vaste 17" storage tank

thermosig6on heator leaks due to weld cracks LER8001E 155375 02/23/1980 03/03/1980 DD PC 8 0F M C7 Chlorine discharged into river enceaded limit L ER8006 155989 02/26/1980 03/25/1980 8 B& G B B AE E u nata station rad G sonitor falls due to beat plug-la module LER8007 155986 02/27/19f0 03/25/1980 5 31 E B Au D C7 Liquid rad waste EI At line leaks due WL oj) to corrosion of suppo rts L Ena 006 155984 03/06/1980 01/20/1980 5 SC E C AD D u containment 6 , atmosphere recirculat ion f an bypa ss damper fails open due to troken linkage L ER800 i d 158782 04/28/1980 05/12/1980 d, go 11 D C3 an unplanned na OG radioactive release occurred when a dogassifier g puputre disk anse cracked

Table B2 coding Sheet for Re;ortable' Events at uadda: b eck-1980-(cuttemmi) *

~

uuaber tecession F.went Sopor t Plant component Abnormal SigaLcance susbur Date Date Status Systen equipment Instrument S ta tus condition cause category Comme n t LERdO11 160286 08/01/1980 3d/13/1980 B SA FF c oc n a containment batch leak rate not tested on time L ER8014 160042 09/26/1980 10/24/1980 8 45 80 c on D a Diesel fire pung W. SE discharge pressure below tech specs (Low eagine speed) L ER8007 C 161681 09/26/1980 10/08/1980 PC 00 B OG H .C3 As operator OJ opened the wrong valve causing 1.3 ci to be released via the stack LER8013 162400 09/26/1980 09/29/1980 5 FC 00 t OG M c3 Techniciaa makes DJ sampling error, 1.3 ci I radioactive gas relea sed LERH0lb 160512 10/20/1980 10/24/1980 3 51 FF S AT D 5 Aerated draia 1 line leaks la Liguid Gas System due to failed gasket oQ reboiler pump L Endo 16 161762 11/18/1980 12/03/1980 8 R5 J B AC

                                                                                                                                      &S D        5       2 con trol rods drop-ia due to burned out contac to rs      .

LER600s r 16 4= 3 8 12/17/1980 01/11/19s1 mr F r. Or a c7 aigh ispingement rate of fish on the intake screens lWV

Tabis 82 ccding shact frr sagttt:b13 E7s tc at Midd:o N:ck-1981 - (carimma) 1 Eum5ec Accession ~1 ent aeporE riant - component Abaormal sijallica nce~ ~~ ~ Number Date Date Status System Ege1 Paest Instruseet s ta t es Condition cause category conse n t LEE 0101 164334 C2/05/1981 01/02/1981 2 SC 00 C AQ D s A contalement I BB air I tecirculation dealer f ailed to close LESd102 165395 03/22/1981 04/13/1981 3 SF-C DD C OJ N N A safety OE lajection puay DC test was not completed FroPe rlF Les3013 165900 04/03/1981 04/21/1981 9 CS S 3 AL 5 C4 fressurizer reliet and AT blocking valves opened due to a loose electrical comaectton LEB8104 166067 04/07/1981 05/18/1981 3 55-C F 3 BC e a An APUS valve 00 breaker was OJ found to be opea LES$1005 166187 C4/22/1981 05/22/1981 RB 5 OE S8 A Con ta m [A 4 ft4[ bundles were released tros site prior to health physses reviee LER0106 174698 06/03/1981 06/25/1981 B S R-C qgg a AC D 5 Aus feedwater BR pump started due

                                                                                                       ,                                                             to a control valve failure L E B8107  166846                  06/04/1981 06/25/1981                    B    EE                    F          C                  D          N       A diesel EH                          Jesorator load seguenciaJ timer f ailed due to timer drift LERQ103    167720                  06/16/1981 07/10/1981                    5    S M-C      00                   4          OE       A         C2       toth AFWS pumps At                          inoperable due
                                                                                                                                            ,                        to a procedural deficiency LP.38110   167842                  07/04/1981 07/29/1981                    E    PC         DD                   N          AP       D          W       Charjia, pump AW                          vibration caused

• FF- its oti line to crack and leak -

                           ~ ~ ~ ~ ~

if3CLWLJ @KE)g shoot far 333tet:b13 3 coto et ; a4423 N:ck-1981 - (Coettaued) ancher ~~~$~csasica c E7cdi~~~~~'Beytet F155%

                                                                                  ~

Ncobst Dito D.Ito Stetco SyEten Eguipment Instres2atC*rPs555t thastaal Statta Condition sigaificinco ~ CecGo Ca tegtr y Conawat _ _ . e LEB0109 167854 07/23/1981 C8/03/1981 3 OE & N 1he lou population zone was incorrectly calcolated LBaa112 168526 07/28/1981 C8/20/1981 3 NB II e nu D a an air ejector radia tion monitor indicated a primary to secondary steas generator tube leak LEB8111 168527 08/03/1981 08/17/1981 3 EE N C E& D 5

                       ,                                                                                                           During a test, a SC                         diesel generators

• ouput exceeded the easinua LE98111 allowable 168621 08/03/1981 88/31/1981 3 35 u C D M a diesel 2emerator could not be synchronized due to a lact et governor control LBa8100 E 168192 08/16/1981 08/25/1981 3 45 00 5 OG N C3 Bobot gas 03 releaso rate escoeded limits due to oporator ettor LEB3114 168929 09/01/1981 89/14/1981 3 EB O C A LL D u 8 the diesel 9enerator was inoperable due to a leaking coolia) sy ste n LER8115 170001 09/17/1981 10/16/1981 8 && E B 11 E C3 Badioactivity 40 released via OG 40 cracked exhaust L End 11t. 169563 duct to stack 09/27/1981 11/23/1981 D CF 00 8 171561 &L D N 4 valvo an the Av SNR systes leaked due to LER811a loose bolts 170137 172047 11/la /1981 11/2 C/198( 0 50 00 C 11/14/1981 01/26/198 > & N 1he contalasest 40 penetration leak rate exceeded the 11stt l99

i Table 52 Coding Sheet for Begortable Evoets at Naddam Neck-1981 - (Coartaued) Number Accession Event 3eport Plant ';ompcaest Ahmoraal S ignificence n ual.or Date pate status Systen Eguipment Instrument Sta tus Condition Cause CateJory Commer.t _ _ _ . _ _ _ - - = -- _ _ _ _ - ___ L E Rel119 171704 11/19/1981 12/04/1981 B PC DB B AP D W Charjing pump D && vibration was excessive due to a worn out i bearing LEadl20 171821 11/22/1981 12/17/1981 5 NB 00 q 8 BG D N & staae flow IE Au differential pressure transaatter leaked giviaJ falso flow readings Leahl21 171825 12/04/1981 12/15/1981 B NA 00 5 8 5 011 pressure for R5 J turbine valve us at control fluctuated causing power osic114tions y - 14j .

i . Tabla B2 Coding She:t for sagtetchls Ersats et ande)s sick-1982 - (contimmed) a

    .sunber   accesalon Event       Sepe t    Plant                                Composest abnormal         Sigallicence                    ,

number Date Date States Systes Eguipment Instressat S ta tes Conditica cause Category Commen t lea 8201 172191 01/12/1982 02/04/1982 B BC C 8 3G s C4 Dee to desija 5 error both 0 tattery chargers 1 are powered by e one diesel

                                                                                                       .                    generator           -

L Ea8 202 172178 01/20/1992 83/03/1982 3 us LL C a BG 9 8 Failed inverter In I causes f eedvater flow transmitter to f ail high LEBd203 173361 04/23/1982 05/14/1982 8 cp FF C AG D C7 RSIT fails to 00 novo during test due to binding of valve packlag gland-staller to LEB7720 lea 8204 175151 06/06/1982 06/24/1982 D g) 00 ^'C 15 G C7 RSIV failed to cycle durinj test due to damage duriaJ previous maintenance LER8205 175147 06/03/1982 06/24/1982 3 SC E C AL D C7 Contalement air a CA recirculation fan damper f ails test due to disconnected i control linkage LEB8206 179054 09/19/1982 11/02/1982 3 BC C B 15 D s Battery bank 3C decla red O& leoprabie due J to leakin; cell ) Lek 8207 177690 09/27/1982 10/28/1982 5 RB J P B BE D E . Poor relay It contact on a

!                                                                                                                          master cycler j

relay caused a l ' control rod ' drive slave cycler ' f atture/ rod stop 414ra

!   LEB8208    177829   09/24/1982 11/02/1982   D     $4       pp                     &         OC         k       N       Containment
'                                                                                                                          personnel batch test not .

terformed Joe to ongoing fk2-- maintenance

Table 82 Coding Sheet for Regottable Events at Naddaa Beck-1982 '(Cantimad) Number accession Event sepet Plant suaber Conpcaest Absornal 31Jaiticance Date Date Status System Egelpeemt Instressat S ta tus Condition Cause Category Commen t LE23209 179a67 10/15/1982 01/26/1983 E PC Q R AU D N Plehole beak la 4 3 op pipe joint mest to charging peng throttle valve-saall

release of radioactive l coolant to aus building l

LERd210 179421 11/13/1982 12/08/1982 3 SC E & AD a C7 na operating linkage for a bypass desper om a containment air I recirculation faa failed nochanically durin g maintenance i l i 3 I e

                                                                                                                         /?5                                                              .

f

Tabls 82 Codtag shsct far 33girttblo avosto at Haddas teck-1983-(ometamed) . uncher Accattico Evert Dep:rt Flcat Czaproc;t Abairr41 StJilfIcseca l sunber Date Data Status Systes Eguipseat Instressat S ta tus Condition Cause Category Comment . . L ER8 301 130844 01/05/1983 01/26/1983 8 SC E B AR D u Service water W& EN Au leak la coil of j BS containment air i recirc fan cooler due to corrosion and erosion- se rvice water was blanked off L Ek8 332 181080 01/11/1983 02/15/1983 3 BE E B BE D E mechanical pg. failure of air compressor for diesel generator

                                                                                                                                   - alare received on asia control board L Ea8 33 3  190578   01/22/1983 06/25/1984    C     SE-C             DD             C 1

At D E Aux feed pump RB 00 AB fails flow capacity test doe to wore relief valve on steam supply to turbiae lea 8304 183455 01/26/1983 04/12/1983 C 51 FF C 40 0 m contalament aetrations rallleak rate tests LER8305 182157 03/09/1983 04/11/1983 C SC B C AD J) . C7 Rechanical i linkages on two containment air recirculat ion faa dangers found broken LEss 306 182158 03/07/1983 04/11/1983 C CF 00 C AD D 5 j BHR flow controi QQ Bt valve fails l partially opea due to brokaa t actuator ara l t rP9 30 7 182159 03/22/1983 04/22/1983 C SD a C RI D a containment

15 T EN isola tion pressure instrumentation actuation setroint out of calibrat ion d ue to drift la secomtar calibrettui

                                                                                                                                  ,3 Aacc/

19't 1

Tabl2 B2 codtog Sheet far R31ertable Ev20ta at Naddas unck-1983-(Contiamed) - 55 5Zr-~~~I2c3 ZI5a'Ei43Y poport FIs.t Campose:t Abagraal . SIj7IIIsaac3 s ea ber Dato Dato States Systen Eguipment Instrgment S ta tes Condition Casou Category coemen t ___ -= LER8308 183093 04/14/1983 05/20/1983 3 as II 3 47, e a primary to

55 secondary steae i pee ra tor tube leaks in muaber ,

2 steaa  ! i generator i { L ER8 309 183299 05/05/1983 06/17/1983 5 PC DD C AG S W Opera tor was j enable to rotate ' the "A" charging peep by hand la

                                                                                                                            ,                 order to perfore the routiae surveillance test due to solidified boric l                                                                                                                                             acid 1

l i lea 8310 183300 05/07/1983 06/17/1983 3 SC 91 E P B Ag NC

                                                                                                                              )          5    Temporary loss of service water to contalement air recirc f ans dee to a clogged service water filter Lead 311       183301           05/04/1983 06/17/1983   3      S P-C        DD                 C                BK         &          5    NPSI pump fails OK                         to reach the shutoff head of 1400 psi l                                                                                                                                            requireJ by the i

1 test procedure duo to l marealistic regelrement ? I LER8382 183558 05/17/1983 06/27/1983 5 EE 5 C & BE D 5 Governor voltage setting drif ted j when dieari

;                                                                                                                                             generator was shutdown f or aninteaaace at air-star t actors LER8313        183559           06/01/1983 07/05/1983   8      Et           u                  C                SI         e          u    Diesel generator failed to attata moraal idle speed withia j                                                                                                                                            time specified i

la mathly - surve111once test , i 19s- . i

Tablo B2 Cadhg Sheet Er 22prtab13 Emats at Caddas Neck-1933-(Centlaued) , 53ESW~~Ecadon 172E Beperl PEl~ . Ca mpw. vat A ha s:r eal SijillicracJ

                                                                                                                                    ~

uueber Date Date Status Systes Eguipneat Instrqment Status Condition Cause Category Coenest . L ER8 314 185176 07/30/1983 04/25/1983 3 54 S S BG I s one of the two ojr incoming station service supplies was disabled by

                                             ,                                                                           a theaderstors L ER8 315    186378   08/31/1983 10/25/1983     3     CB        CC        I          C            AG      e       a       Pressuriser high In                  F                       BA                      level trip relay stect ductag test - 2 out of 3 logic L Ea8 316    186328   09/01'/1983 10/25/1983    3     PC        G         5          B            AD      8       N       Charging Pump DD                                OA                      declared inoperable due to broken wires os motor bearing thermocouple LER8 318     187111   10/19/1983 11/21/1983     5     PC        DD                   8            As      D       E       the "a" charging C5        FF           .                                            pump was taken        .

out of ser vice due to leakage i on the out board peep seal L ER8 320 187491 11/01/1983 12/13/1983 3 CB P 5 AR G $2 containment P& Fr as control air lost 00 BG due to incorrect BE filte r o-ring - caused loss of pressurizer spray salves and

                                                                          .                                              PORVs LER8311      187253   11/03/1983 12/01/1983     5     EE        3         C          C            EN      D       u       Set point drift S F-8     58                                                        in the energency diesel timer for the "B" LPSI Pe*P lea 8325     187=70   11/28/1983 12/13/1983     3     C5        P                    S            AA      D       S2      Loss of PA        FF                                AD                      containment 00                                AW                      control alt SG                       (broken air BK                      filter canister) caused loss of control of Pressuriser sprey valves and FORVs 146

Tab 13 82 CadL;g Shoct f ar Coggrtab13 EJsata et Nadd22 Nock-1983-(Qattamed) d t 35C5Ic~~~~15c!E2EE3 EviEE I Eer3fE FIta t

• R2aps,omt Ibaarmal sta silIcanca '

sunber Dat e Date Status System Egelpaent Instreneet Sta tus Condition Cause C4tegory Comaea t i b i L ER8 322 188309 11/23/1983 01/16/1984 B PC 3 8 1R S C4 Loss of control JJ BG power to motor 00 operated valve 1 00 on WCT outlet d LL due to f ailed j transformer and 1 fuse 1 l lea 8323 187980 12/01/1983 01/05/1984 8 EE 5 C C BC G N Derlag a test !j EA the diesel - 9enerator . I assened a load greater than the i governor setting des to an out of , adjus t neat - 9overnor assembly E LEB832n 188664 12/30/1983 81/27/1984 3 un 3 EB B C4 Service wa ter SS E flow to

• a contaissent coolers is less con se r va ti ve .

than assumed la current containment j 1 pressure g analysis j L Ea8 325 188387 12/07/1983 01/27/1984 5 RB J T C SD D u Fallere to 8 withdraw rods during rod notion checks due to a t ailed actor starter switch , PMOI6120 03/15/1983 03/17/1983 C RB C I AE G $7 Improper J BC latchia; between C1 control rod drive shaf ts and t rod control cluster assemblies during low power physics tests

!                                                                             19r7                                                            -

i 4

                                                                                                                                                  'l i

s

                         .                                                                                                                              g Table 52     Coding sheet for negortable Events at Naddae sock-1983 - (Continued)                          , e i

los Evea[

             ~

Number Be Plant sue Accesf= r Date Dakaort e Component &bsetaal Status Systee Eguipment Instransat Sta tus 31.alf5C ace Coeditica Cause alegory Comment Puo!8195 09/06/1363 09/06/1983 3 as as a sE D C3 3. 3 Ci of noble 00 . 3J gases released OG to the stack t vbes a cellet valve on the vaste gas surge tant 11fted doe to fallece of gas coeFressors 4 1 1 i e I l 1 - I i I 5 j IVf

T: bis 82 C@ ding Sheet far 23prtsb13 2700to at Nadda2 Cock-1983- (Coettaued) penber Adcassion Event Report Flaat

• Component abnormal Sijaifica nce sunber Date Date Status Systen Eguipseat Instremost Sta t es Condition Cause Category Conseat LEp8401 189204 03/21/1984 04/23/1984 8 At 8 OA D C7 ff Several fire doors were foemd to be Laoperatie LEBd402 189507 03/23/1984 B4/26/1984 5 AB F 8 11 N N Loss of power to 51 BG the fire detection system for the screenwell building due to a manually opened breaker LEs8401 189508 04/04/1984 05/08/1984 3 at ff a 01 D C7 A fire door was inoperable f or

                                                                    ,                                                               two days lea 8404    190589    04/13/1984 06/25/1984           5     SA         FF                    3        01       D         C7       A penetration 18                                                                    tire barrier was found to be inoperable LER8405      190099    04/26/1984 05/30/1984           3     NB         5                     B        AD       5         C4       Errors were I                              OK                         discovered in the design basis evaluation of the steam line treak accident during three loop operatton LER8406     190536    06/11/198e 07/10/1984           3     at       ff                      B        04       0         C7       4 fire door was O                                                         dtscovered with an inoperable latchta}

nochanise lea 0407 190537 06/12/1984 07/10/1984 3 5/l FF a 04 D C7 Several laoperable fire karrier penetration seals were discovered during an inspection L End40d 190973 07/21/1984 08/23/1984 S AB FF 8 04 D C7 A fire door was disco ve red with an taoperable latchinj nochanisa ~ lea 8409 191 329 08/01/1984 09/10/1984 9 En F & A1 N S2 Totat losa of

• SG S5 normal off site EA power due to -

fj4I laadvertent c.

                                                                                                                                   . . losing
                                                                                                                                        . . .. of.a *KW <

                      .                                                                                                                        C Tablo 82 Codia) Sheet for segortable Beasts at Naddas Neck-1984 - (Gunlamed)                             ,
                                                                                                                                           ~

EGa5=c Icce ssI5E~E vill 1 sport Plant component Absoraal sijai? Ice nce States System Eguipment Eastrument S ta t us Condition Cause Category Comment a ua ber Date Date

                                        ==.

L E Ed 410 191266 38/03/1984 09/13/1984 C CB 00 5 B &Y D W During a ref neling shutdown the reactor coolant low pressere overpressure protection system relief valves operated Leadmit 191330 08/17/1984 39/21/1984 C S1 FF C 45 9 5 the containment on penet rations f ailed t he intejratee leak rate test LEBd=12 191311 08/19/1984 09/24/1984 C St PF C AU D 8 Combined leak 00 01 rate from 4 valves escoeds limit for containment penet ration local leak rate test lea 0413 191618 38/21/1984 09/28/1984 C FA FF S 15 3 S7 Fa11ere of the JJ refueling pool seal due to improper destga LgEH414 191741 08/24/1984 09/28/1984 C E1 F 3 3B & $7 total loss of Et g BG G aormal off site LL power due to G maintena nce error on a dif f e ren tial relay current transf ormer - output breaker for one diesel LEB841S 191112 08/20/1984 09/20/1984 C NB II C S W Two steam am generators fall the first level of oddy current test LEB8416 191331 08/23/1984 09/28/1984 C c) 00 C At S 5 Dering a test an SS isolation valve in the aata y steam drain line f ailed to close due to a beat

Nsaber acc2ecTom toe:t seyort plant cocesacte abaarnai sli3IIIcInca mecher Dato Dcto S tatc0 SFCtes 59ei P asat Inctr nont states Condities ccese Categsry conne:t lea 4417 192019 10/04/1984 11/13/1984 C ga G & && D C7 Dejraded cabies AC la the reactor g rotection systes instrumentation that is nousted la the asia controi board Lssa418 192020 10/08/1984 11/14/1944 C 43 ff 5 01 m C7 4 fire door was found to be tropped opea LEEd419 195497 10/M/1984 01/30/1985 a sc 00

• C 08 & s Containment 59 integrity tech spec is violated by operation of post accadest saapie systen -

tech spec witi te revised Lens 420 192275 10/13/1984 11/29/1984 8 OD & C3 cae worker receives high es pos are (guar terly

reading of 2.8 rea) -

angualified health physics technictaa j assi)ned to area i traut21 192201 11/03/1984'12/11/1984 3 11 L C BF M C7 Seactor tripped during start-up physics tests due to operator etror ) tend =22 192276 10/10/1984 12/06/1984 e as FF 8 04 D C7 Besident j 51 laspector j discovered a fire barrier i yenetration ! without a face ! tarrier seai LEade21 192790 10/31/1984 12/06/1984 C F 88 5 C7 gestin jhou se Ce circuit breakers fati to close on demand 6 times j in 5 mon th s d oe J to dust on the relay a

7;ble B2 Codi2J Kheet ist 33gtrtabla Essatu et OcJda2 sock-1983. (custmed) C sesher accession Evuat Deport Plant Component Ahmoraal Si jaliica nce , s ea ber Date Sete States Systee Egelpaent Instremmat S ta t es condition cause Category Coenent # LEade2e 192336 11/10/1984 12/1)/1984 3 In 00 E S SC G s Cae portion of the reactor g rotec tion system was inoperable due to salvin) error during malatemance 192569 11/20/1984 81/04/1985 3 In F T S SF N C7 Opera tor lea 842S CS 99 OJ inadvertently SC tripped a reactor coolant peep - reactor tripped LEke426 192366 11/15/1984 12/24/1984 3 51 T C a SF D C7 Circuit card

                                    -                                                           EC                          defect chaajed sain generator current -

opera tor maneatly tripped outpet breakers and initia ted manual reactor trip 14 L C RR D 5 Set point drif t LEE 8427 192411 11/15/19 64 12/26/1964 3 la overpower trip setpoints for 2 of a power range chamaels s as 00 C At D C7 During a test, lea 8428 192367 11/16/1964 12/26/1984 on 11 of 16 mais steam safety valves lifted at a pressere lower

                                                            .                                                                than the e                    setpoint pressere lea 8429   195498   12/02/1984 D4/12/1985       3     IB                   L          3        sc        G        C6       Load reaback initiated (ESP) due to erroneosa
                                                                                                                              ' dropped rod-rod stop s alara caused by a power raa.je indicator that mas out of ad jus tment Ub 1

Table 52 Ccdisj Sheet for Ragortsbi3 Emots ct Caddae Neck-1984 - (Coetinued) usaber accession Event Report Flaat Component Abnors41 Sijaifica nce number Date Date Status Systee Egsipeest Instrueest Sta t us Coedition Cause Category Comeomt

                                                                                                                                                 ~

2136a14 08/01/1984 12/12/1944 C 34 FF s O& D C7 1mo penetrations as in the safety-relayted cable vault and the aus feed pump roos.were

                                                                                     ~

act scaled and no fire watch was established 2139414 10/30/1984 12/12/1984 c SA FF 5 O& B C7 & toeporary fire AB seal on a penetration was removed and no tire watch was established 4 Los *

     .   .,.~_

e e t Table 8.2. Plant status, component status, and cause of reportable events Code Plant status Component Cause of reportable status event A Construction Maintenan a Administrative error and repair , B Operation Operation Design error Refueling l C Testing Fabrication error - j D Shutdown Inherent error 1 E Installation error j F Lightning . G Maintenance error R Operation error I Weather s i q P e l [ 4 I e 0

l Table B.3. Systems involved with reportable events ( Systes . Code Reactor RX 4 Reactor vessel internals RA I Reactivity control systems RB Reactor core RC Reactor coolant and connected systems CX l Reactor vessels and appurtenances CA Coolant recirculation systems and controls CB Nain steam systems and controls CC Main steam isolation systems and controls CD l Reactor corp isolation coo 11ag systems and controls CE Residual heat removal systems and controls CF Reactor coolant cleanup systems and controls CG i Feedwater systems and controls CE i' Reactor coolant pressure boundary leakage detection systems CI Other coolant subsystems and their controls CJ Engineered safety features SX 1 ( Reactor containment systems SA

;          Containment heat removal systems and controls                 SB Containment air purification and cleanup systems and controls SC             g Containment isolation systems and controls                    SD             g i(         Containment combustible control systems and controls          SE             d l           Imergency core cooling systems and controls                   SF
             . Core reflooding                                           SF-A Imer-pressure safety injection system and controls        SF-5 Righ-pressure safety injection system and controls        SF-C l      ,

Core spray system and controls - SF-D Control room habitability systems and controls SG Other engineered safety feature systems and their controls SH Containment purge system and controls SH-A Containment spray system and controls SH-B Auriliary feedwater system and controls SH-C Standby gas treatment systems and controls SH-D Instrumentation and controls IX Reactor trip systems IA Engineered safety feature instrument systems IB Systems required for safe shutdown IC Safety-related display instrumentation ID Other instrument systems required for safety IE Other instrument systems not required for safety IF L A

4 , l Table 8.3 (continued) ( . ._ . . -- . System

• Code i Electric power systems EX Offsite power systems and controls EA AC onsite power systems and controls EB DC onsite power systems and controls EC

Onsite power systems and controls (composite ac and de) ED l Emergency generator systems and controls EE Emergency lighting systems and controls EF Other electric power systees and controls EG i

Fuel storage handling systems FX , New fuel storage facilities FA

Spent-fuel storage facilities FB Spent-fuel pool cooling and cleanup systees and controls FC Fuel handling systems FD Auxiliary water systems WX Station service water systems and controls WA Cooling systems for reactor auxiliaried and controls WB Domineralized water askaup systems and controls WC g Pocable and sanitary water systems and controls WD Ultimate heat sink facilities WE 3 f Condensate storage facilities WF

      \                                     Other auxiliary water systems and controls                                                             WG Auxiliary process systems                                                                                 FX

Compressed air systems and controls FA Process sampling systems PS 4

                     ,                      Chemical, volume control, and liquid poison

• systems and PC i controls Failed-fuel detection systems PD Other auxiliary process systems and controls PE Other auxilisry systems AX

Air conditioning, heating, cooling, and ventilation systems AA J and controls a Fire protection systems and controls AB l Commsunication systems AC

! Other auxiliary systems and controls AD ) Steam and power conversion systems HX Turbine-generators and controls HA Main staan supply systees and controls (other than CC) HB Main condenser systees and controls BC Turbine gland sealing systems and controls HD

l Table B.3 (continued) ( System . Code Turbine bypass systems and controls HE Circulating water systems and controls HF Condensate cleanup systems and controls HG Condensate and feedwater systems and controls (other than CH) HH Steam generator blowdown systems and controls HI Other features of stesa and power conversion systems (not HJ included elsewhere) Radioaccive waste management systems MK , Liquid radioactive waste management systems MA i Gaseous radioactive waste management systems - MB Process and effluent radiological monitoring systems MC Solid radioactive vaste management systems MD i Radiation protection systems BK Area monitoring systems BA Airborne radioactivity monitoring systems BB Other KK Not applicable ZZ S i 1 I i l l [ l I I 1 l 1 i t 1 l I I I. - , - . - _- - -, . _ . - - . _ - ._ - - - - . - . _ . . _ . _ _ . - - . - - - - - _ .. -. - - - - .

Table B.4. Equipment and instruments involved in reportable events

                    - Code -                                                                Code Equipment A       Accumulator                                                   W  Internal combustion engine B       Air drier                                                     X  Motor C       Battery and charger                                           Y  Noszle D       Bearing                                                       Z  Pipe and pipe fitting E       Blower and dampero                                           AA  Power supply F        Breaker                                                     BB  Pressure vessel G       Cables and connectors                                        CC  Pressuriser H       Condenser                                                    DD  Pump I       Control rod                                                  EE  Recombiner J       Control rod drive                                            FF  Seal K       Cooling tower                                                GG  Shock absorber L       Crane                                                        HK  Solenoid M       Domineralizar                                                II  Steam generator N       Diesel generator                                             JJ  Storage container O       Fastener                                                     KK  Support structure P       Filter / screen                                              LL  Transformer
 . ..___ _             _.Q. _ . Flange                                                       MM  Tubing -          . - - _           _ _ _ . _ . . _ _

R Fuel element NN Turbine S Fuse 00 Valve T ' Generator PP Valve, check b [ U Heat exchanger QQ Valve operator $ V Hester Instrumentation _ A Alarm L Power range instrument

              -         5       Amplifier                                                     M* Pressure sensor C       Electronic function unit                                      N  Radiation monitor D       Failed fuel detection instrument                              0  Recorder E       Flow sensor                                                   P  Relay F       In-core instrument                                            Q  Seismic instrument

• G Indicator R Solid state device t H Intermediate range instrument S Start-up range instrument I Level sensor T Switch J Meteorological instrument U Temperature sensor K Position instrument l

4

_- _ .- - _ __ --.~-._7--_ .

                                                                                              ,      w Table B.5. Abnormal conditions of reportable events

( Mechanical - AA Normal wear / aging /end of life: expected effect of normal usage AB Excessive wear / clearance: component (especially a moving com-ponent) experiences excessive wear or too much clearance or gap exists because of overuse, lack of lubrication AC Deterioration / damage: component is no longer at an acceptable level of quality (e.g., high temperature causes rubber seals to chemically break down or deteriorate, insulation breaks down) i AD Break / shear: structural component physically breaks apart (not l when something " breaks down") AE Warp / bend / deformation: shape of component is physically dis-torted

AF Collapse: tank or compartment has an external pressure exerted l that results in deformation AG Seise/ bind /j as: component has inhibited, movement caused by crud, l foreign asterial, mechanical bonding, another component l AH Excessive mechanical loads: mechanical load exceeds design l Mdu

! AI Mechanical fatigue: failure due to repeated stress AJ Impact: the result of the force of one object striking another AK Improper lubrication: insufficient or incorrect lubrication AL Missing / loose: component is missing from its proper place or is loose or has undesired free movement i AM Wrong part: incorrect component installed in a piece of equip- o-( AN sent Wrong material: incorrect anterial used during fabrication or k i installation A0 Weld-related failure: failure caused by defective veld or

                  , located in the heat-affected zone AP       Vibration other than flow induced: vibration from any cause l                     other than fluid flow AQ      Crud buildup: buildup of foreign material such as dust, sticks, trash (not corrosion or boron precipitation)

AR Corrosion / oxidation: unanticipated attack AS Dropped: component is dropped (includes control rod that is

                     " dropped" into core)

AT lask . internal, within system: leak from one part of a system to another part of the same system AU Leak, internal, between systems: leak from one Epstem to a dif-farent system AV Crack: defect in a component does not result in a leak through I the wall AW Imak, external: defect in a component results in a leak from the system that is contained in an onsite building AX I4ak to environment: leak not resulting from a cracked or broken component AY Was opened / transfers open: component is/vas opened by error or spuriously opens

u

 .,      c.

Table B.5 (continued) I_ AZ Was closed / transferred closed: component is/was - vrongly closed by error or spuriously closes BA Fails to open: component is in the closed state and fails to open on demand (e.g., the circuit breaker " fails to open" when an overcurrent occurs) BB Fails to closes component is in the open state and fails to close on demand BC Malposition or anladjustment: component is out of desired posi-tion (e.g., normally open valve is closed) or adjusted improperly (not for instrument drift or out of calibration) BD Failure to start / turn on: component fails to start on demand BE Stopped / failed to continue to run: componant fails to continua running when it has previously started - BF Tripped: component automatically trips on or off (desired or undesired) (e.g., the turbine tripped because of overspeed, the circuit breaker tripped because of overspeed, or the circuit breaker tripped because of overload) BG Deanergized/ power removed: component on systea loses its driving potential but not necessarily electrical power [e.g., (1) a fuse blows and there is no power to a sensor, and the sensor is de-energized; (2) a valve closes off the steam supply to a turbine, and the turbine has no driving power] BE Energized / power applied: component or system gains its driving potential but not necessarily electrical power (e.g., valve is ( opened allowing steam to turn a turbine) BI Unacceptable response time: component does not respond to a g demand within a desired time frame but does not otherwise fail ;I (e.g., a diesel generator fails to come to full speed within the time constraint) - BJ High pressure: higher than normal or desired pressure exists ic a component or system (does not include instrument aisindica-tions) BK Iow pressure: lower than normal or desired pressure exists in a component or system (does not include instrument aisindication) B1. High temperature: component experiences a higher than normal or desired temperature BM Iow temperature: component (or system) experiences a lower than normal or desired temperature BN Freezing: fluid medium (e.g., water) freezes in or on a com-ponent 50 Excessive thermal cycling: frequent changes in temperature that could result in metal fatigue or cracking BP Unacceptable heatup/cooldown rate: heatup or cooldown rate exceeds limits BQ Thermal transient: system experiences an undesired or unstable thermal transient or thermal change BR Excessive number of pressure cycles: system experiences an un-desired number of significant pressure changes (e.g., pressure pulses as from a positive displacement pump) _. . - , . . . _ . . - . . . . . . - .. . - .. ..-.....v-.-,.-...

se a . . - - .. - - - - u - + . .

                                                                                                          .u      ;

Table B.5 (continued) ( BS High level / volume: higher than normal or des' ired level or volume exists (actual or potential)' in a component, such as tank or sump, or area, such as auxiliary building (not for instrument misindication) BT Iow level / volume: lower than norac1 or desired level or volume exists in a component (not for instrument misindication) BU Abnormal concentration /pH: an abnormal (either high or low) con-centration of a chemical or reagent exists in a fluid system or an abnormal pH exists (does not include abnormal boron concentra-tion) , BV Abnormal boron concentration: process system control rod has an abnormal boron concentration from burnup, dilution, or overaddi-tion SW Overspeed: speed in excess of design limits BX Cladding failure: cladding of a component fails (e.g., the cladding of a fuel pellet is breached, and radioactive fuel leaks occ) BY Burning / smoking: component is on fire or smoking BZ Engaged: component engages or meshes (this is not to be used when a component binds or becomes stuck or jaianed) CA Disengaged / uncouple'd: component disengages, loses required fric-tion, or is no longer meshed (as in gears), for example, the clutch on the actor disengages from the shaft (this should not be used for dropped control rods) ( Electric / instruments u_ EA Excessive electrical loads: electrical loads exceed design - rating E EB Overvoltage/ undercurrent: component failure produces an over-voltage / undercurrent condition other than open circuits

     . EC  Undervoltage/overcurrent: component failure produces                             an under-voltage /overcurrent condition other than shorts ED  Short circuit / arcing / low impedance: electrical component shorts or arcs in the circuit or has a low impedance including shorts to ground EE  Open circuic/high impedance / bad electrical contact:                           electrical component has a structural break, or electrical contacts fail to contact sad fail to pass the desired current

EF Erratic operation

component (especially electrical or instru-ment) behaves erratically or inconsistently (if an instrument produces a bad but constant, signal, use "EG", if an instrument produces an inconsistent signal use "EF")

EG Erroneous /no signal: electrical component or instrument produces an erroneous signal or gives no signal at all (not for out-of-calibration error) EH Drifc: a change in a setting caused by aging or change of physi-cal characteristics (does not include personnel errors or a physical shift of a component) l

u. . .

1 + 9-

• Table 5.5 (continued)

( EI out of calibration: component (particularly instruments) become out of adjustment or calibration (does not include drif t) EJ Electromagnetic interference: abnormal indication or action resulting from unanticipated electromagnetic field EK Instrument snubbing: dampening of pulsating signals to an in-stcument Hydraulie HA High flow: higher than normal or desired flow exists in a com-ponent/ system (does not include instrument aisindication) (see code EG) EB Low flow: lower than normal or desired flow exists in a com-ponent/ system (does not include instrument aisindication) EC No flow or impulse: fluid flowing through a pipe, filter, orifice, or trench or the fluid in an impulse line (e.g., instru-ment sensing line) is blocked completely or decreased due to some foreign material, crud, closed (either partially or completely) valve or damper, or insufficient flow area HD Flow induced vibration HE Cavitation HF Erosion , HG Vortex formation HE Water hammer HI Pressure pulse / surge HJ Air / steam binding ( HK Loss of pump section

• HL Boron precipitation Other

[p

                   . 0A     Declared inoperable:                  component or system is declared inoperable
         ,                  as required by Technical Specification's but any be capable of partially or completely performing its desired duties when re-quested (a component / system that is completely failed should not use this code) 08     Flux anomaly:               flux characteristics of the reactor core are not l                            as required or desired (e.g., flux spika due to xenon burnout)

OC Test not performed: operator or test personnel fails to perform a required test within the required period OD Radioactivity contamination: component, system, or area becomes more radioactive than desired or expected OE Temporary modification: an installation intended for short term use (usually this is for maintenance or modification of installed equipment) I 0F Environmental anomaly OG Airborne release OH Waterborne release OI Operator communication OJ Operator incorrect action OK Procedure or record error l -. .. . . - . . - - . . - - - -

                                                                       -<    ---m  ,,~~-r-   -

t- ~ ~~ ~ ~ r

                                                                                      . r Table 5.6. Significance Criteria for reportable events - Significant
                    ,                              Evenn description 51         Two or more failures occer in redundant systems during the same event S2         Ttso or more failures due to a common cause occur during the same event S3         Three or more failures occur during the same event S4         Component failures occur that would have easily escaped detection by testing or examination
                 $5         An event proceeds in a way significantly different from what would be expected 56          An event or operating condition occurs that is not enveloped by the plant design bases 57         An event occurs that could have been a greater threat to plant safety with (1) different plant conditions, (2) the advent of another credible occurrence, or (3) a different

, progression of occurrences 4 S8 Administrative, procedural, or operational errors are committed that resulted from a fundamental misunderstand-ing of plant performance or safety requirements S9 other (explain)

                                                                                          ~
                                                                                        +

i T( 4, Table B.7. Significance criteria for reportable events -- Conditionally Signifi. cant Category of conditional Event description significance C1 A single failure occurs in a nonredundant system C2 Two apparently unrelated failures occur during the same event C3 A problem results in an offsite radiation release or ex-posure to personnel C4 A design or manufacturing deficiency is identified as the cause of a failure or potential failure C5 A problem results in a long outage or major equipment damage C6 An engineering safety feature actuation occurs during an event C7 A particular occurrence is recognized as having a sig-nificant recurrence rate C8 Other (explain) o-e}}