ML17258A230

Review of Operating Experience History of Ginna Nuclear Power Plant for NRC Sep.

ML17258A230
Person / Time
Site:	Ginna
Issue date:	10/31/1981
From:	OAK RIDGE NATIONAL LABORATORY
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NUDOCS 8110150496
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REVIEW OF THE. OPERATING -EXPERIENCE. HISTORY OF THE GINNA NUCLEAR POWER PLANT FOR THE NUCLEAR REGULATORY COMMISSION'S SYSTEMATIC EVALUATION PROGRAM Performed by the Staff of the Nuclear Safety Information Center (NSIC)

Oak Ridge National Laboratory (ORNL)

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CONTENTS

~Pa e INTRODUCTION SCOPE OF REVIEW 1-2 Availability and Capacity Factors...................... 1-2 1.2 Review of Forced Shutdowns and Power Reductions........ 1-3 1.3 Review of Reportable Events............................ 1-S 1.4 Events of Environmental Importance and Radioactivity R eleases............................................. 1-16 1.5 Evaluation of Operating Experience..................... 1-16

2. SOURCES OF INFORMATION UTILIZED IN THE REVIEW.......... 2-1 2.1 Availability and Capacity Factors...................... 2-1 2.2 Forced Reactor Shutdowns and Power Reductions.......... 2-1 2.3 Reportable Events....................... 2-2

'2.4 Environmental Events. and Radioactivity Releases........ 2-2 2.5 Use of Computer Files on RECON and Special P ublications......................................... 2-2

3. CRITERIA AND CATEGORIZATION FOR THE EVALUATIONS OF THE OPERATING HISTORY................................ 3-1 3.1 Significant Shutdowns and Power Reductions............. 3-5 3.1.1 Criteria for significant shutdowns and power reductions........................................... 3-5 3.1.2 Use of criteria for determining significant shutdowns and power reductions................................. 3-5 3.1.3 Non-DBE shutdown and power reduction categorization.... 3-6 3.2 Significant Reportable Events.......................... 3-9 3.2. 1 Criteria for significant reportable events............. 3-9 3.2.2 Use of criteria for determining significant reportable events1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3-9 3.2. 3 Non-significant reportable events......................

~,

CONTENTS (continued)

~Pa e OPERATING EXPERIENCE REVIEW OF GINNA 4-1.

4.l Summary of Operational Events of Safety Importance..... 4-1 4.2 General Plant Description.............................. 4-2 4.3 Availability and Capacity Factors...................... 4-2 4.4 Review of Reactor Shutdowns and Power Reductions....... 4-4 4.4.1 DBE initiating events..................................

4.4.2 Trends and safety implications of shutdowns and power reductions........................................... 4-10 4.5 Review of Reportable Events............................ 4-21 4.5.1 Significant events at Ginna............................ 4 4.5.1.1 .Loss of offsite power loads to excessive cooldown rate. 4-26 4.5.1.2 Both MSIVs close spuriously............................ 4-28 4.5.1.3 Conditionally Significant Events of Importance......, .. 4-28" 4.5.1.3.1 Safety in]ection train B fails to actuate on demand.... 4-28 4.5.1.3.2 Loss of concentrated boric acid injection capability... 4-29 4.5.2- Trends and safety implications of reportable events in addition to those categorized as significant...... 4-30 4.5.2.1 Causes of reported events......... 4-32 4.5.2.2 Trends in conditionally significant events.............

4.5.2.2.1 Recurring failures C7................................ 4-34 4.5.2.2.2 Emergency bus breaker failures......................... 4>>38 4.5.2.2.3 Common cause failures............................. ~ ~ ~ ~ ~ ~ 4%2 4.5.2.2.4 Other conditionally significant events...,....... 4-42 4.6 Events of Environmental Importance......... 4%6 4.7 Evaluation of Operating Experience......... 4-51 RE FERENC E S ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I 6 1 APPENDIX A. GINNA

1. Shutdown and Power Reduction Tables.........

2. Reportable Event Coding Sheets,...,.......,,,.......,...,...., A>>28

1 1'1 Tables No. Title ~Pa e 1.1 Cause of Forced Shutdown or Power Reduction and Method of S hut, down ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~

~

C

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.2 Systems Involved 'With the Forced Shutdown or Power Reduction... 1-6 1.3 Components Involved With the Forced Shutdown or Power Reduction ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-9 1.4 Data Collected for Reportable Events 'lant Status and System Involved.............................................. 1>>13 1.5 Data Collected for Reportable Events Equipment Involved and Instrument Involved......................................

Ml 1-14 1.6 Data Collected for Reportable Events Component Status, Abnormal'ondition, and Cause................................ 1-15 3.1 Standard Review Plan Chapter 15 Initiating Event for Design Basis (Events (Revision 3)............,....... ........... ~ 3-3 I

3.2 NSIC Event Categories for Non DBE Shutdowns................. 3-7 3.3 Reportable Event Criteria Significant.................,.... 3-10 3.4 Reportable Event Criteria, Conditionally Significant....... 3-1.1 4.1 Availability and Capacity Factors for Ginna....................

4.2 Forced Shutdown Summary for Ginna....'.......................... 4-5 4.3 Power Reduction Summary for Ginna.............................. 4-7 4,4 DBE Initiating Events at Ginna................................. 4-9 4' Ginna FSAR Chapter 14 Events for Safety Analysis (DBEs)........ 4-24 4.6 Significant Events for Ginna................................... 4-25 4.7 Summary of Systems Involved in Reportable Events by Year....... 4-31 4.8 Causes of Reported Events...................................... 4-33 4.9 Conditionally Significant Events at Ginna...................... 4-35 4 10 Recurring Failures C7 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4-36 Summary of Failures by Year.................................

4. 11 4-40 4.12 Emergency Bus Breaker Failures'.............................. 4-41 4.13 Common Cause Failures.........................,............,.... 4-'43 4.14 Summary of Radioactivity Released from Ginna...............,... 4-47 4.15 Radiation Over-Exposure and Releases........................... 4-49

lV Figures No. Title ~Pa e 4.1 Number of Reported Events at Ginna............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 4.2 Accident Sequence'Precursor Study Loss of Offsite Power Event Tree for Event 10/21/73......,............ ~ ~ ~ ~ ~ ~ e ~ ~ ~ e ~ ~ 4 27 4.3 Safety ln]ection Emergency Power Supp1y..........

INTRODUCTION The Systematic Evaluation Program Branch (SEPB) of the Nuclear Regulatory Commission (NRC) has the responsibility for conduct of the Systematic Eval-uation Program (SEP) in an effort to determine the safety margins of the design and operation of the eleven oldest operating commercial nuclear power plants in the United States. These eleven plants are being reevaluated in terms of present NRC licensing requirements and regulations. The,SEP program as well is to:

establish documentation which, shows how these operating plants

'compare with current acceptance criteria and guidelines safety issues, and provide a technical rationale on'ignificant for acceptable departures from these criteria and guidelines; provide the capability for making integrated and balanced decisions with respect to any required backfitting; and provide for the early identification and resolution of any potential safety deficiency.

The SEP Program is evaluating specific safety topics (called the Topic List) and is based on an integrated review of the overall ability of a plant to respond to certain design basis events (challenges), including normal operation, transients and postulated accidents. The evaluation will result in a reassessment of the overall safety margins for each facility and documentation of the reassessment on the basis of current criteria.

In this report the operating experience of the Ginna nuclear power plant fs reviewed for the purpose of compiling and interpreting data on plant opera-C tional occurrences and events for application and input to the SEP Program.

The results of this report will be used by SEPB in performing the integrated assessment of overall plant safety for this'lant.

1-2 The review approach with respect to operational events (forced shutdowns and reportable occurrences) consists primarily of a three-step process:

1) compile information on the events, 2) screen the events for significance using selected criteria and guidelines, and 3) evaluate the significance and importance of the events from a safety standpoint. Trends in equipment failures and events where systems failed to perform their intended function are identified.

Other types of operating information as noted in the "scope of the review" section is compiled ro provide an overall view of the plants'perating histories.

1. SCOPE OF REVIEW The assessment of the operating experience review for Ginna covered the time from initial criticality through and including 1979. The review included the following aspects of operation: availability and capacity factors; review of forced shutdowns and power reductions; reportable events; events of environmental importance and radioactivity releases; an evaluation of the operating experience in total.

1.1 Availabilit and Ca acit Factors Both reactor availability and unit availability factors were compiled for all years. Starting with 1974, the unit capacity factors using the design electrical rating (DER - net MWe) and the maximum dependable capacity (MDC net MWe) were compiled as well. Data for the capacity factors was not available from earlier years.

The two availability and two capacity factors are defined as follows:

1. Reactor availability ~

hours reactor critical + reactor reserve shutdown hours x lpO period hours

1-3

2. Unit availability ~

,hours enerator on-line + unit reserve shutdown hours 1OO period hours

3. Unit capacity (DER) ~

net electrical en'er enerated x lpp period hours x DER

4. Unit capacity (MDC) ~

net electrical ener enerated ~ happ period hours x MDC net 1.2 Review of Forced Shutdowns and Power Reductions The forced shutdowns and power reductions were reviewed,and data collected on each incident. Scheduled shutdowns for refueling and maintenance were not included in the review. However, if a utility had a refueling outage scheduled, the plant experienced a shutdown as a result of an abnormal event prior to the I

scheduled refueling, the uti1ity reported that the refueling.was being resched>>

P uled to coincide with the current shutdown, and the utility reported the cause of the shutdown as refueling, then this shutdown was considered as forced.

Only that portion of the outage time concerned with the abnormal event, not the refueling time, was included in the compilations.

The power reductions were included to provide information and details that may have been associated with a previous or subsequent shutdown. The power reductions are included in the proper chronological sequence with the shutdowns in the data tables for the forced shutdowns and power reductions (see Appendices).

The following data was compiled annually for the forced. shutdowns and power reduction:

1. date of occurrence;

2. duration in hours; I

3. noting if the shutdowns were al'so a reportable event, e.g., a licensee event report (LER) or abnormal occurrence report (AOR);

4. a summary description of the events associated with the shutdown or power reduction;

5. cause of the shutdown (Table 1.1);

6. method of shutdown (Table 1.1);

7. the system directly involved with the shutdown or power reduction (Table 1.2);

1-5 Table 1.1 Cause of Forced Shutdown or Power Reduction and Method of Shutdown Cause Equipment failure Maintenance or testing Refueling C Regulatory restriction D Operator training and license exams Administrative Operational error Other Method of Shutdown Manual Manual Scram Automatic scram 3 Continuation 4 Load Reduction 5 Other 9

Table 1.2 Systems Involved With the Forced Shutdown or Power Reduction S stem Descri tion Code, Reactor RX Reactor Vessel Internals RA Reactivity Control Systems RB Reactor Core RC Reactor Coolant System & Connected Systems CX Reactor Vessels & Appurtenances CA Coolant Recirculation Systems 6 Controls CB Main Steam Systems & Controls CC Main Steam Isolation Systems & Controls CD Reactor Core Isolation Cooling Systems & Controls CE Residual Heat Removal Systems & Controls CF Reactor Coolant Cleanup Systems & Controls CG Feedwater Systems & Controls CH Reactor Coolant Pressure Boundary'Leakage Detection Systems CI Other Coolant Subsystems & Their Controls CJ Engineered Safety Features SX Reactor Containment Systems SA-Containment Heat Removal Systems & Controls SB Containment Air Purification & Cleanup Systems & Controls SC Containment Isolation Systems & Controls SD Containment Combustible Control Systems & Controls SE Emergency Core Cooling Systems & Controls SF Control Room Habitability Systems & Controls SG Other Engineered Safety Feature Systems & 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 Po~er Systems & Controls EA AC Onsite Power Systems 6 Controls EB DC Onsite Power Systems 6 Controls EC,.

(Composite AC 6 DC)

Onsite Systems & Controls (Composite ED AC & DC)

Emergency Lighting Systems & Controls EF Other Electric Power Systems & Controls EG Fuel Storage and Handling Systems, FX New Fuel Storage Facilities FA Spent Fuel Storage Facilities FB Spent Fuel Pool Cooling 6 Cleanup Systems & Controls FC Fuel Handling Systems FD

1-7 S stem Descri tion (Cont'd.) Code Auxiliary Water Systems WX Station Service Water Systems 6 Controls WA Cooling Systems for Reactor Auxiliaries & Controls WB Demineralized Water Make-up Systems & Contro'ls WC

. Potable & Sanitary Water Systems & Controls WD Ultimate Heat Sink Facilities WE Condensate Storage Facilities WF Other Auxiliary Water Systems & Their Controls WG Auxiliary Process Systems PX Compressed Air Systems & Controls PA Process Sampling Systems PB Chemical, Volume Control, & Liquid Poison Systems & Controls PC Failed Fuel Detection Systems PD Other Auxiliary Process Systems & Their Controls PE Other Auxiliary Systems Air Conditioning, Heating, Cooling & Ventilation Systems & .Controls AA Fire Protection Systems & Controls AB Communication Systems AC Other Auxiliary Systems & Their Controls AD Steam and Power Conversion Systems HX Turbine-Generators & Controls HA Main Steam Supply System & Controls (Other Than CC) HB Main Condenser Systems & Controls HC Turbine Gland Sealing Systems & Controls HD Turbine Bypass Systems & Controls HE Circulating Water Systems & Controls HG Condensate and Feedwater Systems & Controls (Other Than CH) HH Steam Generator Blowdown Systems & Controls HZ Other Features of Steam & Power Conversion Systems (Hot Included Elsewhere)

Radioactive Waste Management Systems MX Liquid Radioactive Waste Management Systems MA Gaseous Radioactive Waste Management Systems MB Process 6 Effluent Radiological Monitoring Systems MC Solid Radioactive Waste Management Systems MD Radiation Protection Systems BX, Area Monitoring Systems BA Airborne Radioactivity Monitoring Systems ~

BB

8. the component"directly involved with the shutdown or power reduction (Table 1.3); and

9. categorization of the shutdown or power reduction. Each shutdown or powex reduction was placed into one of two sets of categories. The shut'downs and power xeductions were first evaluated against design bases events (DBE) as described in Chap. 15 of the Standard Review, Plan.~ If the shutdown or power reduction could not be categorized as a DBE initiating event, then it was placed into one of a series of NSIC categories. For further discussions of these two sets of categories, use of the categories, and a listing of them, see Sect. 3.1 .and following.

The listings for the cause, shutdown method, system involved,. and component involved along with their respective codes axe those used in NRC's Gray Book sexies for shutdowns. Note that the information listed under the "system involved" column in the data tables in the appendices indicates (1) a general classification of systems (fully written out) and (2) a specific system within the genera1 classification which is coded with two letters.

1.3 Review of Re ortable Events The operating events as repoxted in LERs and LER predecessors, e.g., AORs unusual event reports, reportable occurrences (ROs), were reviewed.

These types of reportable events were retrieved from the NSIC computer file.

Approximately five years ago, operating experience information for oper-ating nuclear power plants in the NSIC file for the time period predating LERs

1-9 Table 1.3 Components Involved With the

. Forced Shutdown or Power Reduction Com onent T e Com anent T e Includes:

Accumulators Scram Accumulators Safety Injection Tanks Surge Tanks Air Dryers Annunciator Modules Alarms Bells Buzzers Claxons Horns Gongs Sirens Batteries & Chargers Chargers Dry Cells Wet Cells Storage Cells Blowers Compressors Gas Circulators Fans Ventilators Circuit Closers/Interrupters Circuit Breakers Contactors Controllers Starters Switches (other than sensors)

Switchgear Control Rods Poison Curtains Control Rod Drive Mechanisms Demineralizers Ion Exchangers Electrical Conductors Bus Cable Wire Engines, Internal Combustion Butane Engines Diesel Engines Gasoline Engines Natural Gas Engines Propane Engines

0 1-10 Com onent T e Includes Engines, Internal Combustion Butane Engines Diesel Engines Gasoline Engines Natural Gas Engines Propane Engines Filters Strainers Screens Fuel Elements Generators Inverters Heaters, Electric Heat Exchangers Condensers Coolers Evaporators Regenerative Heat Exchangers Steam Generators Fan Coil Units Instrumentation and Controls Mechanical Function Units Mechanical Controllers Governors Gear Boxes Varidrives Couplings Motors Electric Motors Hydraulic Motors Pneumatic (Air) Motors Servo Motors Penetrations, Primary Containment Air Locks Pipes, Fittings Pumps Recombiners Relays Shock Suppressors and Supports Transformers Steam Turbines Gas Turbines Hydro Turbines

1-11 Com onent T e Includes Valves Valves Dampers Valve Operators Vessels, Pressure Containment Vessels Dzywells Pressure Suppression Pressurizers Reactors Vessels

1-12 was reviewed. Any documents that contained LER-type information (equipment failure, abnormal event, etc.) were coded or indexed so that they could be xetrieved in the same manner as an LER. Primarily, this involved various types of operating reports and general correspondence for the late 1960s and early 1970s.

The following information was recorded for each reportable event reviewed:

1. an LEE report number or other means of identification of report type;

2. NSIC accession number (a unique identification number assigned to each document entered into the HSIC 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 1.4);

6. system involved with the reportable event (Table 1.4)';

7. type of equipment involved with the reportable event (Table 1.5);

8. type of instrument involved with the reportable event (Table 1.5);

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

1. 6);

10. abnorma3. condition associated with the reportable event, e.g., corrosion, vibration, leak, etc. (Table 1.6);

11. cause of the event (Table 1.6); .and

12. significance of the reportable event. Each reportable event was screened using criteria as a step in the evaluation process (See Sect. 3.2 and following for fuxther discussion of the criteria, the use of the criteria, and a listing of the criteria.)

0

'1-13 Table 1.4 Data Colle'cted for Reportable Events Plant Status and System Involved PLANT STATUS A Construction B Operation C Refueling D Shutdown SYSTEM A Chemical and Volume Control B Component Cooling C Condensate Purification D Condenser Cooling E Containment F Containment Air Cooling G Containment Filtering H Containment Hydrogen Control I Containment Isolation J Containment Purge K Containment Spray L Core Reflooding M Electric Power N Emergency Cooling/LPSI 0 Emergency Electric Power P Engineered Safety Features Q Fire Protection R Hydraulic S Main Cooling T Pneumatic U Radiation Monitoring V Reactor Control W Reactor Protection X Safety Injection/HPSI Y Secondary Cooling/Aux.

Z,. Secondary Cooling/Feedwater AA Secondary Cooling/Steam BB Service Water CC Shutdown Cooling DD Waste Disposal .

EE Ventilation FF Reactor Internals

1-14 Table 1.5 Data Collected for Reportable Events Equipment Involved and Instrument Involved EQUIPMENT INSTRUMENTATION A Accumulator A Alaxm B Air Drier B Amplif C Battery and Charger C iex'lectronic Function Unit D Beaxing D Failed Fuel Detection Instrument E Blower and Dampexs E Flow Sensor F Breaker F In-Core Instrument G Cables and Connectox's G Indicator H Condenser H Intermediate Range Instrument I Control Rod I Level Sensor J Control Rod Drive J Meteorological Instrument K Cooling Tower K Position Instrument L Crane L Power Range Instrument M Demineralizer M Pxessure Sensor N Diesel Generator N Radiation Monitor

'0 Pastener'ilter/Screen 0 Recorder P P Relay Q Plange Q Seismic Instrument R Fuel Element R Solid State Device S Fuse S Start-Up Range Instrument T Generator T Switch U Heat Exchanger U Temperature Sensor V Heater W Internal Combustion Engine X Motor Y Nozzle z Pipe and Pipe Fitting AA Power Supply BB Pressure Vessel CC Pressurizer DD Pump EE Recombinex FF Seal GG Shock Absorber HH Solenoid II Steam Generator JJ Stoxage Container KK Support Structure LL Transformer MM Tubing Turbine 00 Valve PP Valve, Check Valve Operator

"1-15 Table 1.6 Data Collected for Reportable Events Component Status, Abnormal Condition, and Cause COMPONENT STATUS A Maintenance and Repair B Operation C .Testing ABNORMAL CONDITION A Age B Airborne Release-C Concentration D Corrosion E Crack F Crud G Environmental Anomaly H Erosion I Exposure J Fatigue K Fire L Instrument Calibration M Instrument Set Point Drift N Leak 0 Liquid Level p Lubrication Open/Short Circuit R Operator Communication S Operator Incorrect Action T Procedures U Records V Sampling W Smoke X Stress Y Stress Corrosion Z Vibration AA Waterborne Release BB Wear CC Weld CAUSE A Adminis trative Error B Design Error C Fabrication Error D Inherent Failure E Installation Error F Lightning G Maintenance Error H Operator Error I Weather

1-16 1.4 Events of Environmenta1 Importance and Radioactivi Releases Based upon reviewing fo'reed shutdowns, power reductions, reportable events (environmental LERs), and operating reports, any significant or recur-ring environmental problems were summarized.

The routine radioactivity releases were tabulated. as well, and releases where limits were exceeded were reviewed and discussed.

1.5 Evaluation of cretin Experience Based upon the review involving screening, categorizing, and compiling data, the operating history of the plant was evaluated. Judgments and conclusions were made regarding safety problems, operations, trends (recur-ring problems), or potential safety concerns.

From the information provided through the various operating reports and the review process, events were analyzed to determine their safety significance, using the final safety analysis report to provide specific plant and equip-ment details when necessary.

I' 2-1

2. SOURCES OF INFORMATION UTILIZED IN THE REVIEW Several sources of information aad periodic (annual, quarterly, and, monthly), NRC publications were used in the review. Some sources contained information relative to more than one area within the scope of the review.

2.1 Availabilit and Ca acit Factors The availability and capacity factors were either extracted or calculated from data given in the Gray Books from 1974 through 1979 (the first Gray Book was issued ia May 1974) . Prior to 1974,.the aanual or semiannual reports were used to compile the availability factors only.

2.2 Forced Reactor Shutdowns and Power Reductions The review of the forced power'eductions involved checkiag the following sources for completeness of details and'ccuracy:

1. Nuclear Pauer Plant Opera'.ing Experience for 19', for the years 1973, 1974-75, 1976, 1977, aad 1978 (Refs. 3, 4, 5, 6, and 7). The report for 1979 has not been published. However, since the work on the section of these reports on outages has been perfoxmed by NSIC since 1973, the draft copy of this report for 1979 was available;

2. The Gray Book - NUREG-0020 Series;

3. Annual ox semiannual reports from the time of startup thxough 1977. For 1977 through 1979, monthly operating reports were used because the utilities were no longer required to file annuals. The review of power reductions involved primaxily the annuals, semiannuals, aad monthly reports.

2-2 2.3 Re ortable Events The NSIC computer file of LERs was the primary source of information in reviewing the reportable events. The material on the NSIC computer file consists of the appropriate bibliographic material, title, 100-word abstract, and keywords. When it was necessary to obtain additional information on the event, the original LER (or equivalent) was consulted by (1) examining those full-size copies on file at NSIC (for the years 1976 through 1979); (2) the microfiche file of docket material at NSIC or (3) the appropriate operating report (semiannual, annual; or monthly report).

2.4 Environmental Events and Radioactivit Releases i Events of environmental importance were obtained as a result of con-ducting the overall review of the plant's operating history, and the sources of information involve all types of documents listed thus far.

The data for radioactivity releases were compiled primarily from the report Radioactive MateriaEs ReEeased from NucEear P~er PEants AnnuaE Report 2977, NUREG-0521. This report presents year-by-year comparisons for plants in a number of different categories (solid, gas, liquid, noble gas, tritium, etc.). The data for 1978 was taken from the report Radioactive MateriaEs ReEeased from NucEear Power Plants - Annual Report 2878, NUREG/CR-1497, which was published in March 1981. The data for 1979 was compiled from the annual environmental reports submitted by the licensees.

2.5 Use of Co uter Files on RECON and S ecial Publications Two computer files on RECON (a computer retrieval system containing <35 data bases operated at ORNL) were used extensively for another purpose in

2-3 addition to those indicated thus far. Printouts were obtained from the files I

fox Ginna to provide coverage on other types of "docket material" besides reportable events where the licensee may have been in correspondence with NRC [or the Atomic Energy Commission (AEC)] concerning a particular, event. Licensees are often requested to submit additional information or perform further analysis. Before the LERs came into existence in the mid-1970's, it was not unusual for licensees to submit on their own or at NRC (AEC's) request more than one letter transmitting information on a particular event.

Thus these printouts provided additional sources of information on reportable events.

Several special publications were xeviewed to provide details on events of significance. Events described in the following publications often con-tained details, evaluations, or- assessments other than those provided in the reportable event (or shutdown) as a result of furthex analyses and examination:

1. Reports to Congress on Abnozmal Occurrences, NUREG-0020 series;

2. "Power Reactor Event Series" (formerly Current Event Series) published by NRC;

3. "Operating Experience Section" of the A'ucEecu Safety journal; and

4. NRC's Office of Inspection and Enforcement's (IGE) publications

a. Operating Experience Bulletins 1

b. IE Bulletins

c. IE Circulars

d. IE Infoxmation Notices

3. CRITERIA AND CATEGORIZATION FOR THE EVALUATIONS OF THE OPERATING HISTORY In reviewing the opera"ing history of the plant of interest, the two areas focused on were forced shutdowns (and power reductions) and reportable events. Given the large number of both shutdowns and reportable events, it was necessary to develop consistent review procedures that involved screening and categorizing of both occurrences. Following screening and categorization, the study then assessed the safety significance of events and analyzed the categories of events for various trends and recurring problems.

The shutdowns were evaluated against the design basis events (DBE's) as set forth in Chap. 15 of the SRP. The DBE's are those postulated diseur-bances in process variables or postulated malfunctions or failures of equip-ment for which the plants are to be designed to withstand and for which the licensees are expected to analyze and include in safety analysis reports (SAR). In the SAR, the effects of anticipated process disturbances and post-ulated component failures are to be examined to determine their consequences and to evaluate the capability built into the plant to control or accomodate such failures and situations (or .to identify the limitations of expected performance).

The intent is to organize the transients and accidents considered by the licensee and presented in the SAR in a manner that will:

l. Ensure that a sufficiently broad spectrum of initiating events has been considered,

2. Categorize the initiating events by type and expected frequency of occurrence so that only the limiting cases in each group need to be quanti-tatively analyzed, and

3-2

3. permit the consistent application of specific acceptance criteria for each postulated initiating event.

Each postulated initiating event is to be assigned to one of the following categories:

l. Increase in heat removal by the secondary system (turbine plant),

2. Decrease in heat removal by the secondary system (turbine plant),

3. Decrease in reactor coolant system flow rate,

4. Reactivity and power distribution anomalies, 5.'ncrease in reactor coolant inventory,

6. Decrease in reactor coolant inventory,

7. Radioactive release from a subsystem or component, or.

8. Anticipated transients without scram.

Typical initiating events that are representative of those that are to be considered by the licensee in the SAR are presented in Table 3.1 Those shutdowns identified as DBE initiating events were categorized as such. If the shutdown was not a DBE, then it was assigned a category from a list developed by NSIC to indicate the nature and type of error or failure. The NSIC categories for non-DBE shutdowns were examined as part, of a trends analysis.

The reportable events were screened using the criteria presented in,.

Sect. 3.2, (and following) and.were categorized according'to their significance.

The information collected on the reportable events (as outlined in Tables 1.4 through 1.6) was used to analyze trends for all reportable events those identified as significant or non-significant.

3-3 Table 3;1 Initiating Event Descriptkons for Design Basis Events as Listed in Standard Review Plan-Chapter 15 (Revision 3)

l. Increase in Heat Removal b the Seconda S stem 1.1 Peedwater system malfunctions that result. in a decrease in feedwater temperature 1.2 Feedwater system malfunctions that result in an increase in feedwater flow 1.3 Steam pressure regulator malfunction or failure that results in increas-ing steam flow 1.4 Inadvertent opening of a steam generator relief or safety valve 1.5 Spectrum of steam system piping failures inside and outside of contain-ment in a PWR

2. Decrease in Heat Removal b the Seconda S stem

'I 2.1 Steam pressure regulator malfunction or failure that results in decreasing steam flow 2.2 Loss of external electric load 2.3 'urbine trip (stop valve closure) 2.4 Inadvertent closure of main steam isolation valves 2.5 Loss of condenser vacuum 2.6 Coincident loss of onsite and external (offsite) a.c. power to the station 2.7 Loss of normal feedwater flow 2.8 Feedwater piping break

3. Decrease in Reactor Coolant S stem Plow Rate 3.1 Single and multiple reactor coolant pump trips 3.2 BWR recirculation loop controller malfunctions that result in decreas-ing flow rate 3.3 Reactor coolant pump shaft seizure 3.4 Reactor coolant pump shaft break

4. Reactivit and Power Distribution Anomalies 4.1 Uncontrolled control rod assembly withdrawal from a subcritical or low power startup 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 (1bw power to full power) 4' Control rod maloperation (system malfunction or operator error),

including maloperation of part length control rods 4.4 Startup 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

P Table 3. 1 (continued) 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 Spectrum of rod ejection accidents in a PWR 4.9 Spectrum of rod drop accidents in a BWR

5. Increase in Reactor Coolant Invento 5.1 Inadvertent operation of ECCS during powex opezation 5.2 Chemical and volume contx'ol system malfunction (or operator error) that increases reactox coolant inventory 5.3 A number of BWR transients, including items 2.1 through 2.6 and item 1.2

6. Decrease in Reactor Coolant Invento 6.1 Inadvertent opening of a pressurizer safety or relief valve in a PWR or a safety ox relief valve in a BWR 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 Loss-of-coolant accidents resulting from the spectrum 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 2.7, 2.8, and 1.3

7. Radioactive Release from a Subs stem or Com onent 7.1 Radioactive gas waste system leak oz 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 spent fuel storage buildings 7.5 Spent fuel cask drop accidents

8. Antici ated Tzansients Without Scram 8.1 Inadvextent control zod withdrawal 8.2 Loss of feedwater 8.3 Loss of a.c. power 8.4 Loss of electrical load 8.5 Loss of condenser vacuum 8.6 Turbine trip 8.7 Closure of main steam line isolation valves

3-5 The review approach with respect to operational events (forced shut-downs and 'reportable occuziences) consisted primarily of a three-seep process:

1) compile information on the events, 2) screen the events for significance using selected criteria and guidelines, and 3) evaluate the'ignificance and importance of the events from a safety standpoint. The evaluations were to determine those areas where safety problems existed in terms of systems, equipment, procedures, and human error.

The reviewers worked semiindependently (brief exchanges of ideas and information) and then were brought together periodically for discussion and final resolution as to how events were to be categorized and how the criteria were to be used consistently.

3.1 Si ificant Shutdowns and Power Reductions For the purpose of compiling information and for evaluation, the power reductions were treated in the same manner as the forced shutdowns.

3.1.1 Cziteria for si ificant shutdowns and ower reductions As indicated previously, the occurrences identified as design basis events were used as criteria to categorize and note significant shutdowns.

These events are li,sted in Table 3. 1 as they are found in SRP Chap. 15.

3'1.2 Use of criteria for dete'zminin si ificant shutdowns and ower reductions The generic DBE initiating event types, e.g., "increase in heat removal by the secondary system" or "decrease in reactor coolant system," were used as primary flags for. reviewing the forced shutdowns (and power reductions).

Once the genezic type of event was identified, the particular initiating event was determined from the details associated with the shutdown. For example, if the reactor shuts down because of an increase in heat removal

3-6 due to a feedwater regulator valve failing open, the shutdown is a DBE generic type 1 event. Specifically, based upon the initiating event (valve failed open), it is a 1.2 DBE - feedwater system malfunction that results in an

'increase in feedwater flow. Some shutdowns were readily identifiable as specific DBE's, such as tripping of a main coolant pump - a 3.1 DBE.

Once categorized as a DBE," the shutdown was considered significant regardless of the resulting effect on the plant (because a design basis event 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 closures of main steam isolation valves.

3.1.3 Non-DBE shutdown and ower reduction cate orization Those non-DBE shutdowns were assigned NSEC categories (Table 3.2) to provide more information on the failure or error associated with the shut-down. With these categories, more specific types of errors and failures could be examined through tabular summaries to focus the reviewer's attention on problem areas (safety-related or not) that were not revealed by the DBE categories .

The causes for non-DBE shutdowns taken from the Gray Book (listed in Table 1.1) are limited and very general, while NSIC cause categories

3~7 Table 3.2 NSIC Event Categoxies 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 'nstallation Error N 1.1.4 End of design life/inherent failure/random failure N 1.2 Failure on demand under test conditions 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 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 requized to function outside design specifications)

N 3.0 Non-DBE Reductions in Coolant Inventory (Leaks)

N 3.1 In primary system N 3.2 In secondary system and auxiliaries 4.0 Fuel/Cladding Failure (densification, swelling, failed fuel 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 N 6.0 Operator Error N 6.1 Incorrect action (based upon correct understanding on the part of the operator and proper procedures, the operator turned the wrong switch or valve ~ incorrect action)

N 6.2 Action on misunderstanding (based upon proper procedures and improper understanding or misinterpretation on the part, of the operator of what is to be done ~ incorrect action)

N 6.3 Inadvertent action (puzpose and action not related, e.g.,

bumping against a switch or instrument cabinet)

3-8 N 7.0 Procedural/Administrative Error (Incorrect operating or testing procedures.

Incorrect analysis of an event-failure to consider certain conditions in analysis)

N 8.0 Regulatory Restriction N 8.1 Notice of generic event N 8.2 Notice of violation N 8.3 Backbit/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 l0.0 Environmental Operating Constraint as Set Forth in Tech Specs

3-9 are more specific. Thus, as an ex~le, the number of Gray Book- causes noted as eqixLpment failure should" not be expected to equal those identified as.

equipment failures with the NSIC categories. Other NSEC categories, such as component failure; co'uld be classified as an equipment failure if the. only available designations for cause were those listed in the Gray'. Book.

3.2 Si ificant Re ortable Events 3';2.1 Criteria for si ificant re ortable events Two groups of criteria were used in determining significant reportable events. The first=set of criteria (Table 3.3) indicates those events that are'definitely'significant in terms of safety and are termed significant.

Those criteria in'Table 3.4 indicate events that may be of potential con-cern. These events,which. might require additional information or evaluation to determine their full implication, were noted as conditionally significant.

3.2.2 Use of criteria for determinin si ificant re ortable events The reportable events were all reviewed applying the two sets of criteria for significance rather liberally. A number of significant events and conditionally significant events were noted. The events initially identified as significant ox conditionally significant were ana1yzed and evaluated further based upon (1) engineering )udgment; (2) the systems equipment, or components involved; or (3) whether the safety of the plant was compromised. The final evaluation for significance considered whether a DBE was initiated or whether a safety function was compromised such that the system could not mitigate the propagation of events for which it was designed. Thus, the number of events categorized finally as significant was reduced considerably by these steps in the review process.

3-10 Table 3.3 REPORTABLE EVENT CRITERIA SIGNIPICANT SIGNIFICANCE CATEGORY EVENT DESCRIPTION Two or more failures occur in redundant systems during the same event.

S2 Two or more failures due to a common cause occur during the same event.

S3 Three or more failures occur during the same event.

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

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

S6 An event or operating condition occurs that is not enveloped by the plant design bases.

S7 An event occurs which could have been a greater threat to plant safety with different plant the advent of another credible occurrence, condi-'ions, or a different progression of occurrences.

S8 Administrative, procedural or operational errors are committed that resulted from a fundamental mi'sunderstanding of plant performance or safety requirements.

S9 Other (explain).

1

3-11 Table 3.4 REPORTABLE EVENT CRITERIA CONDITIONALLY SIGNIFICANT CATEGORY FOR CONDITIONAL SIGNIFICANCE:" EVENT-DESCRIPTION

/

Cl- -- A single failure occurs in a non-redundant system.

CZ Two apparently unrelated failures occur during the same event.

C3, . ... ... A problem results in an.off-site-radiation release or personnel exposure.

C4 A design or manufacturing deficiency's identified.

as the cause of a failure or potential failure.

C5 A problem results in a long outage or major equip-ment damage.

C6 An ESP actuation occurs during an event.

C7 . A particular occurrence is recognized as having a significant recurrence rate.

C8 Other.

3-12 Those events involving radioactivity releases were automatically cate-gorized as a conditionally significant 3 event and held for discussion in the respective 'environmental and release"'ections of the report.

3.2.3 Non-si ificant re ortable events Those reportable events not identified as significant or conditionally significant were categorized as non-significant (with an "N" in the signi-ficance column in the coding sheets in the appendices). These events and the events re]ected during the additional review step as noted above were further reviewed by compiling a tabular summary of the systems (Table 1.4) to detect trends and recurring problems. The systems selected yield meaningful information concerning the system's ability to mitigate accident sequences or mitigate the effects of such accident sequences.

'(

4. OPERATING EXPERIENCE REVIEW OF GINNA 4.1 Summa of 0 erational Events of Safet Importance" P

The operational history.. of Ginna has been reviewed to indicate those g

areas of plant opexation that compromised plant safety.. The revie~ included a detailed examination o'f pl'ant shutdowns, power reductions, reportable events, and 'events of special environmental importance. The criteria used to show degradations in plant-safety were (I) events that initiated a DBE

'I and (2) events that compromised safety functions designed to mitigate the propagation of the initiating events.

Shutdowns and power reductions indicated the number and types of DBEs entered. The repoxtable events and special environmental events indicated the number of times each engineexed safety function was compromised. The results of the analyses.ideritified 23 DBE's entered. -Additionally, events were identified where loss of safety system function occurred in some engineered safety features.

0 4-2 4,2 General Plant Descri tion R. E. Ginna Nuclear Station is a Westinghouse pressurized water reactor (PWR) of 470-MWe net maximum dependable capacity, owned by Rochester Gas &

Electric Corporation (RGE) and located in Ontario, New York. The Architect/

Engineer was Gilbert Associates, and the constructor was Bechtel. The con-denser cooling method is once-through,'nd Lake Ontario is the condenser coo&a'g water source. The Plant is subject to license DPR-18, issued September 19, 1969, pursuant to Docket Number 50-244. The date of initial reactor criticality was November 9, 1969, and commercial generation of power began July 1, 1970.

The nearest city is Rochester, New York, 17 miles away. The population within 30 miles is 840,000 and within 50 miles, 1,200,000.

I

'4.3-Availabilit "and Ca acit Factors Table 4.1 contains Ginna's availability and capacity factors. The rated net power was boosted from 420 to- 470 K<e on March 3, 1972, to moie fully utilize the installed generating capacity. From mid '1972 .until mid 1973 there was a xegulatoxy limit of 1266 MWt (83% of design value of 1520 MWt). The reactor availability from 1970 through 1979 stayed above 70% except for 2 years, 1974 and 1976, when major outages were necessary for repairs (see Sect. 4.4).

The ten full years of operation, 1970 through 197$ , averaged 78.1% reactor availability and 74.6% plant availability. Capacity factors were not avail-able prior to 1975. The. Ma C and DER capacity factors from 1975 through 1979 averaged 69.5 and 68;5%,, xespectively.

Table 4.1 Availability and Capacity Factors for Ginna 1969 1970 1971 1972 1973 1974 1975 1976 1977, 1978 1979 Reactor Availability 69.3 72.9 77.6 72.0 95.3 63.9 81.5 69.0 87.0 86.9 74.8 Unit Availability 34.3 69.4 75.9 69.2 95.0 62.4 76.7 58.2 85.5 80.6 72.8 Unit Capacity (MDC)* ND~ ND ND ND 73.9 49.9 73.6 78.2 71.9

.Unit Capacity (DER) ~ ND ND ND ~

ND ND ND 73-9 47.9 70-6 '8.2 71.9 HDC ~ Hmdmum Dependable Capacity

't ND ~ No Data

~DER ~'esign Electrical Rating

4.4 Review of Reactor Shutdowns and Power Reductions Table A.l provides a comprehensive summary of information concerning shutdowns and power reductions at Ginna. Some information is still missing (denoted by blanks), however, and some was assumed (denoted by "A/"). More complete information was provided when events generated reports; in such instances, more detailed descriptions are in Sect. 4.5.

Tables 4.2 and 4.3 of forced shutdowns and power. reductions summarize Table A.l. Causes of forced shutdowns, item l.3 in Table 4.2 and item 'Z.2 in Table 4.3, are dominated at Ginna by equipment'failures. Shut-

\

downs reported to be caused by operator errors amount to only 10K of the total, and no power reductions are attributed to operator error. More than one system is often involved in a shutdown or power.'eduction and in some cases the cause of the shutdown is not ascertainable. Therefore, the totals for cause, shutdown method, and system in Tables 4.2 and 4.3 are not compar-able.

Table 4.2 Forced Shutdown Sumaary'for Oinna I

~

, 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 1979 Total I. Forced Shutdowns

1. Total Nunber 11 ~

27 14 12 ~ 5 14 118

2. Total hours down 79.4 1435 169 86.8 434 3291 392

'026 380 281 1107.5 9681.7

3. Cause*

A. Equipment Failure 6.(70.7) 17(613.7) 9(127.4) 10(62.5) 4 (292) 7(3000) 11(336) 12 (1486) 6(380) 3(249) 1(416) 86(7029.3)

B. Haintenance or 6(820.7) 1(17.3) 1(24) 2 (291) 1(44) 3(540) 1(.5) - 15(1737.5)

Testing I

D. Regulatory Restric- 1(691) l(691) tion E. Operator Training/

License Exan F. Adainistrative I 1(32) 1(32)

O. Operational Error 5(8.7) 4 (1.0) 2 (1.8) 1(.3) 12 (11. 8)

N. Other 1(142) 1(12) 2(154)

4. Shutdown Hethod

1. Hanual 2 10 4 7 10 2 2 50

2. Hanual Scree I I 2 12

3. Autoaatlc Scran 8 15 9 4 5 2 2 I 53

4. Continuation 1 1 II. Total nunber of SRP Related Shutdowns (These arc in- 23 cluded in Totals of Part I)

Nunber of hours associated with cause of shutdown is in parentheses.

Table 4.2 (continued) Forced Shucdown Summary For Cinna 1969 1970 1971 1972 1973 1974'975 1976 1977 1978 1979 Total III. System Involved

l. Reactivity Control Systems 1 1 2 14

2. Coolant Recirculacion Systems and Controls (CB) l. i 4 2 .3 14 30 Hain Steam Isolation Systems 6 Controls (CD) 2 2

4. Residual Heat Removal Systems 6 Controls (CF) 1

5. ECCS 6 Concrols (SF) 1: ) 2 4

6. Reactor Trip Systems (Ih) 2 5 I

7. Engineered Safety Feature Instrument Systems (IB)

8. Safety-Related Display Inscrumentation (ID) 1 2

9. Other Instrument Syscems Required for Safety (IE) 1 2

10. Offsice Power Systems and Controls (Eh) 1 4

11. hC Onsice Power Systems & Controls (EB) 1 1

12. DC Onsite Power Syscems & Controls (EC) 1 1

13. Composite hC 6 DC Syscems (ED) 1 1

14. CVCS 6 Liquid Poison Systems 6 Controls (PC) 2

15. Turbine Generators & Controls (llh) 3 1 4 5 19

16. Hain Steam Supply System 6 Controls (HB) 1 2 4 2 15

17. Hain Condenser Syscems & Controls (HC) 1 3

18. Circulating Mater Systems 6 Controls (HC) 1 1

19. Condensate and Feedwster Systems 6 Controls (IIH) 7 ~

11 2 32

~ ~

Table 4.3 Power Reduction Summary for Ginna C I 1969 1970 1971 1972 1973 1974 1975 '976 1977 1978 '979 Total I. Power Reductions

1. Total number 7 ll, 3

2. Cause .

Failure ~

1 2 8 5 23

h. Equipment .

Haintenance or Testing 4 3 1 18 8~

D. Regulatory Restriction Operator Training/License Exam 1 E.

Operational Error ~ I F. I 1 G. Other ~ ~ ~ I u

3. System Involved

~ I Reactivity Control System (RS) 1 1 1.

2~ Other Coolant Subsystems & Their Controls (CJ) 1 1 1

3. Emergency Core Cooling Systems & Controls (SF) I

'I ~

4. Offsite Power Systems & Controls (Eh) 1 1 AC Onsite Po~er Systems & Controls (EB) 1 2 5.

6. Air Cond., llcating, Cooling, & Ventilation (Ah) ~

1 1

7. Turbine-Generators & Controls (Rh) 1 1 a ~

5 I

8. Hain Stcam Supply System & Controls (88) \ 3 V

9. Hain Condenser Systems & Controls (RC) ~g 3 3 I", 25

10. Condensate and FM Systems & Controls (88) 1 1

4-8 4.4.1 DSE Initiatin Events Of the 118 total forced shutdowns and power reductions accumulated at 23 fell into DBE initiating event categories as shown in Table 4.4.

None of these events initiated any sequence that led to any significant economic loss or safety hazard to the plant or the environs.

The categorized events were dominated by control rod malfunctions, steam generator tube leaks, and spurious closures of the main steam isolation valves (MSIVs). Seven of the eight control rod malfunctions generated individual reports (the exception is the first one, July 5, 1971), as did all five steam generator tube failures and one HSIV spuxious closure (June 23, 1975}. Control rod failures and steam generator tube failures are thoroughly discussed as recurring reportable events in Sect. 4.5.2.

The only turbine trip that caused shutdown occurred on January 27,'1970 because of "the loss of the electro-hydraulic (EH) governor pump pressure."

The only loss-of-offsite-power event occurred on October 21, 1973, and is described in detail in .Sect. 4.5.1.

All three loss-of-normal-feedwater events occurred within 2 months of each other on July 11 and 13 and September 5, 1971. The two in July most likely are related (instrument bus, 81 inverter failures), but not enough information is given in the semiannual report to be conclusive. The September event occurred when the containment ventilation was reset causing the feedwater valves to close by mistake.~

III',I'

~ ~

I

~ t I t 'I )

I

' ~

~ I Table 4.4 DBE Initiating Events at I

.Ginna

~

I l

Description DBE Category 1969 1970 1971 1972 1973, 1974 1975 1976 1977, 1978 1979 . Total Loss of Offsite Power D2.2 Turbine Trip D2.3 Ii 1

~,

II Spurious Closure of D2.4 2 ~ 1'. 2 I" .5 I

I'i Steam Isolation Valves

'ain I ~ I Loss of Normal Feedwater D2.7 3 I Plow I~ I'. i ~. t I\ I

~ ~

Control Rod Malfunction D4.3 I ~

II 8". '-

I r I )

,Steam Generator Tube Pailure D 6.3 I I 2 1 II 5 iI II r

I~ I

~

I Total 0 1 6 0 lI 3 0 23 I

t",

I

' II

~ I ~ I Ir.

I

~, I lr II

~ ~

I~

II I ~ ~

I I i ~

~ ~

II I

I':

I I I

~ I I' I I I I I~ ~

t I

~

I ~ I,'I I I ~

.I '

~

j ~ I ~ I~

I ~

I t I" ~ ' ',

iII I ~ t I

I' It:

I~ ~ ~

Ii

4l 0

4-10 Note that the trend of total number of DBEs per year bears no correlation with other trends, such as plant performance as measured by total number .of shutdowns per year or total downtime pex year.

4.4.2 Trends and safet implications of shutdowns and ower reductions Over the moxa than 10 years of operation at Ginna, forced downtime averaged nearly 40 d/year. Annual downtime was dominated by single, large events such as the 114-d shutdown in 1974 to overhaul the /f2 low-pressure turbine. Total annual'hours of forced down-time followed no trend. Years 1970, 1974, 1976, and 1979 each recorded more than 1100 h forced out of service; the other 6 years each. recorded less than 450 h forced out of service.

Pour out of every five. forced shutdowns were caused by equipment, failures.

Operator exrors that were reported to have caused shutdowns fell from five in the 2 months in 1969 and foux in the first full year of operation (1970) to one in 1972 and none thereafter. Some fine tuning of the feedwater system and its controls made it easier to operate after,'970,. a year in.

which all four operator-error shutdowns were because of feedwater regulation problems. Today, the general availability and use of simulators helps alleviate this problem.

All of the 118 forced shutdowns that did not fall intoSBE categories were put in NSIC categories + (see Sect. 3.1.1). Tn NSIC event categories, leaks curtailed operations most often, with 31 shutdowns caused by leaks in the secondary system and 14 caused by leaks in the primary system. Thirty eight 12/15/69, 12/30/69 (2),

• Except eight forced shutdowns [12/6/69, 12/14/69,sufficiently to enable 1/3/70, 1/4/72, and 7/28/72] that were not documented categorization.

4-11 equipment failures- in normal operation (equipment other, than instruments .

and controls) forced. shutdowns or power reductions.. The third major type

.of NSXC event was instrumentation and control anomalies, with 31 leading to shutdowns or forced reductions. 'I No fuel/cladding failures required shutdowns or foxced reductions, and no environmental operating constraints curtailed operations either. Nine events had an undetermined category.

~ ~ . r Causes of the forced reductions in,reactor power were evenly split

~ ~ ~ ~

between equipment failures and maintenance or testing restrictions. Forced L'

xeductions in power were reported only once at Ginna befoxe the fourth year of ooeration. After 1973 began, however, forced reductions were reported I

nearly five times per year. They usually lasted only a few hours, and seldom was power reduced more than 5QX. Power reductions under 30/ were if the P

included only associated failure had safety significance.

A discussion of shutdowns and power reductions for each year 1969 through 1979 follows.

1969 The Ginna PWR went critical at 0530 on November 9, 1969, and the genera-tor was first synchronized on December 2. Low-power physics testing in December showed excellent agreement between experimental and predicted values (Ref. 8, p. 2).

Operatox errors dominated the causes for forced outages, inducing 5 of the 11 recorded incidents (Table A.~,]. ). The steam and power system was involved nine times.

i

~ ~

4-12 Two shutdowns were caused by Technical Specification restrictions.

On December 16, safeguard valve 850B on the RHR pump suction line from the containment sump failed to open during a test. Both discs of this double-disk valve were bowed from internal pressure. When the RHR system is put in service, the water between the disks is. heated to about 285 0 7, creating a large internal pressure. (The valveis tested for 750 psig in this space.}

The recommendation was made that any double-disc valves that are closed during RHR operation be tested immediately after shutting down the system.

The other shutdown caused by a Technical Specification restriction occurred when the d'ischarge valve on the 1B motor-driven auxiliary .feedwater pump (AZWP) was "found to be inoperable" (Ref. 8, p. 17). No cause has been documented.

During the automatic scram on. December 3, one shutdown rod .(E-11) failed to drop. Exercise of the shutdown bank freed the rod. This situation did not occur again and is not considered safety significant.

Two, recurring problems first surfaced during this report period. A condenser hotwell problem was investigated on December 14, but no results were reported (Ref. 8, p. 17). Feedwater control problems caused eight forced shut-downs in the first 7 months of Ginna's operation. The first four shutdowns because of feedwater control problems occurred in this report period. On December 3, three shutdowns resulted from operator errors in aligning feed-I water valves. The shutdown on December 30 occurred in switching to the auxiliary feedwater. After June, 1970, manual control of the feedwater flow caused no problems; some fine tuning af the feedwater control system made it easier to operate.

I 4-13 1970 In 1970, the reactor and control system proved its capacity to operate under the most severe transient conditions: four automatic scrams from 100K power (420 MRe) occurred, and the plant responded as designed each time.

~

This year had 27 forced shutdowns (plus 1 power reduction), the most in the 10 years of operations at Ginna (Table 4.2). The shutdowns axe attxibuted partly to the operator's lack of familiarity with the new plant; ten of the 27 forced'hutdowns were caused by maintenance, testing, and operator errors.

Only 12 of these errors occurred over the following 9-year period (1971 through 1979). The remaining 17 shutdowns were caused by failures of new equipment.

Two lengthy shutdowns accounted for 80K of the forced downtime this year.

On May 4, the unit went down fox'4- d to repair the 82 low-pressuxe turbine blades (Table A.l.2). On September 30, the'nit went down for 14 d to repair, among othex things, a leak in a reactor coolant temperature detector.

One incident at the beginning of 1970, only 2 months after the initial criticality, exemplified the problems in maintenance and testing that plagued Qiana in its early years. Following a spurious safety in]ection signal on January 3, one train of the high-pressure coolant in]ection system did not operate. After dirty contacts on a relay were found and cleaned, a test signal activated the train. However, at this point, motor operator valve (MOV)-878B failed closed, thereby stopping the SI flow from entering the cold leg. After adjusting the valve's arms and, contacts, operators stroked it successfully seven times.

'I 11 The contxol rod step counters malfunctioned on January 17, requiring a manual shutdown. The ability to safely shut down the reactor was not compromised.

4-14 The first of six 'LLSIV failures over the 10 years of operation at Ginna occurred during the low-pressure turbine failure on May 14, when MS'-1B did not close on a manual signal from the control room. None of these failures compxomised the safety of the plant.

A recurring problem with the packing of the pressurizer spray valves first surfaced on May 2, at which time the reactor was shut down to repack these Valves. On July 5,'acking was added to pressurizer control valve (PCV)-431A, but design capability of these valves was never compromised.

The first of many shutdowns to repair'eaking pressurizer relief valves was required on July 12; others followed on November 15 and December 12.

12 These leaks never jeopardized the safe operation of Ginna.

Another pxoblem recurring over the life of the plant first occurred on January 27 when the-EH turbine governor lost pressure, causing the turbine to trip (this was an DBE Category 2.3 initiating event). The manual feed-water contxol was too slow in responding to the consequent steam generator leve1 transient, and the reactor scrammed on lo-lo steam generator level and E

was down for 25 min (Ref. 8, p. 18). On June 19, a manual turbine trip to repair the governor caused an unintentional reactor trip. During the September 6 shutdown, unspecified repairs on the EE governor control system were made.

None of the 12 EK governor-relate'd shutdowns over the life of Ginna compromised the safe operation of the plant.

Feedwater manual contxol problems and leaking condenser tubes continued to plague Ginna in 1970. Brief shutdowns caused by manual feedwater control errors occurred on January 27 (discussed previously) and 28, March 30, and

i 4-15 June 19. Condenser tubes were plugged during two shutdowns in December, at which time other necessary maintenance was concurrently accomplished.

1971 The number of forced outages dropped to 14 in 1971 (Table A.1.3); with no reported power reductions. The unit was generally base loaded at full load during this pex'iod, as it was thx'oughout the following years. Total forced downtime was only 169 h, second lowest in Qiana's history." On February 3 and November 12, the previously mentioned recurring problem with the packing in the pressurizer spxay valves occurred causing forced outages for repairs of 27.2 h total duration. The sexies of failures leading to a shutdown on June 30 is described in detail in Sect. 4.5.2. On July ll, the inverter on the 125-V dc connection to instrument bus Ol led .to an automatic scram. This problem surfaced'our more times in the period 1974 through 1976.

During the x'efueling outage commenced February 27, all,fuel'ssemblies were tested using the wet-dry sipping technique. Defective fuel was found in region 3.

1972 The Atomic Energy Commission (AEC) granted permission to increase power from 1300 to 1520 NRt on March 1. The 1520-HWt level was reached on April 12.

This year saw- the least forced downtime in Ginna's history: 12 forced shutdowns for 87 h total (Table A.1.4), The EH tuzbine governor system caused shutdowns again on Pebruazy 24 and March 8. Shutdowns accounting for 287 of this year's total-downtime were necessary on March 20 and September.

5 to repair'acking on pressurizer spray valves. Malfunctions in the drive

4-16 and .logic supply power for the contxol rods'caused shutdowns on June 23 and July 27, respectively. A leak in the block valve downstream of the power-operated relief valve necessitated a shutdown on December 14; 1972.

Duxing the refueling outage commenced April 14, inspections found defective fuel: some rods were bowed, some were collapsed and some leaking.

Sixty-one new fuel assemblies replaced all of region 3, part of region 2, and 3 'assemblies in region l.

On June 24 the turbine generator returned to service using an interim set of conditions. The maximum nuclear power was not to exceed 1266 MWt (83/ of rated power). Other conditions.,required various tests to be performed at various power levels.

1973 Only five forced shutdowns occurred in 1973, Five power reductions also occurred, all for required maintenance (Table A.l.5).

A new problem surfaced on June 9 when the bearing shoes on the 1A main feedwater pump needed to be replaced. Power reductions were also necessary on August 19 and September 10 to repair the B main feedwater pump.

Tubes in the main condenser leaked again and caused a shutdown for repairs on January 12 and a 38K power reduction on March 9.

Further repairs on the turbine EH contxol system necessitated a 50K power reduction in November.

Two lengthy shutdowns occurred thi.s year; On July 22, an 8.5-d shut-down began to repair a disconnected flow transmitter on the B auxiliary feed-water pump control valve. The disconnection occurred during the only water-hammer event in Ginna's history severe enough to cause a shutdown. 13 In

4-17 October, Ginna experienced its sole loss-of-offsite-power event, described in detail in Sect. 4.5.1.

Ginna was restarted 6 d later.

1974 This year Ginna suffered the moat. forced downtime in its history. Of the total 3291 h out of sexvice, 83K'ere caused by a blade failuxe'.in the. same low-pressure turbine that caused a 5-week shutdown in May and June 1970. The maintenance overhaul began January 1 and ended April 25 (Table A.l.'6). Refuel-ing was completed during this outage.

Instrument bus inverters failed twice,. on April 27 and July 26, both causing a low steam generator level and low feedwater flow. The same event happened in July 1971.

A repair was attempted on June 29 to stop a leak on a charging pump filter vent line. After the 43-h repair, the reactor was restarted. It was shut down 14 h later; and the repair gob was redone correctly.

The 1A main feedwater pump impeller failed on May 20, forcing a power reduction. This was the fifth power reduction or shutdown caused by mechanical failure of a main feedwater pump.

On November 2, the reactor was manually shut down to begin an" inspection of the steam generator tubing. This inspection, lasting more than 11 d, was the first of seven outages forced by steam generator tube degradations or inspections. RG&E's practice throughout the life of Ginna has been to plug leaks in the steam generators and condenser as soon as they are observed as opposed to continuing operation with leakage. Though there had been no leakage at this time, two tubes were plugged.

0 The recurring pxoblem 'with the condenser tubes forced three power reductions of 52% in December. Freon checks of the 1B condenser tubes were made. Since changing the secondary chemistry from phosphate to all volatile treatment (AVT) during the last shutdown for steam generator tube inspection, condenser tube integrity problems have caused more frequent shutdowns. Whereas in the five years of operation previous to November 1974 in which Ginna has been shutdown five times for condenser tube inspections, plugging, and repairs, future years saw an average of three shutdowns per year because of condenser tube problems. Ho~ever, no technical evidence has been documented to demonstrate this link.

1975 Condenser tube leaks continued to plague Ginna in 1975 (Table A.1.7).

Power- reductions .were necessary on July 30 and December 21 and 23 to do freon checks for leaks. On December 23, one leak was found in the B condenser.

Another familiar problem caused three shutdowns in May 1975. The EE turbine control sys'em malfunctioned and forced two shutdowns in May. The xeactor was taken down to repair the control system on May 31. Two more shutdowns were to occur because of this system in 1976.

The unit was at 95% power on June 17 when a reactor trip was caused by

'he loss of the IA inverter and the corresponding loss of the 1A instrument bus; The inverter failed. when the pulse drive printed circuit card failed.

1976 An unusually difficult year for Ginna was 1976. Twenty-six forced outages and power reductions were recorded, only two less than during 1970, the first full year of operation. Equipment failures accounted for 20

i 0

4-19 of these events. A total of 2026 h forced out of service this year was second only to 1974, when 2737 h were needed for one job alone the mainten-ance overhaul of the 82 low-pressure turbine; 36K of the downtime in 1976 was attributed to blade failures on the same turbine [January 29, July 29, and August 7 (Table A.1.8)]. Another large contributoX was steam 'generator tube leaks, accounting for 30/ of the forced downtime this year (continued from the December 30, 1975, outage and on April 24).

Three shutdowns wexe.caused by EH turbine control problems. On-April 18 water leaking into the oil from the oil cooler forced two shutdowns. On May 22 another fluid leak in the system forced another shutdown.

Other recurring problems affecting Ginna in 1976 were an instrument bus inverter failure on June 3 and a main feedwater pump impeller failure on September 10.', 46K power reduction for 1 week was necessary to. repair the main feedwater pump.

A total of ten condenser tubes were plugged during four forced outages.

Erosion from the bottom side of the tubes was caused by steam impingement from the steam dump lines located below the affected tubes (September 25 and December 11, 14, and 17).

1977 This year was the first of three successive years of few forced shutdowns and power xeductions and high unit availability. Only six forced shutdowns and three forced power reductions occurred in 1977 (Table A.1.9).

Condenser tube leaks forced two 50/ power reductions one in January and one in March. The recurring problem with steam generator tube leaks forced an 8-d shutdown beginning July 5 for inspections and repairs.

4-20 A new problem surfaced in 1977. Shaft seals on steam generator snuobers were found leaking. Inoperable snubbers forced shutdowns on August 2 and November 2.

1978 Pour forced shutdowns and -four forced power reductions occurred in 1978, totaling only 281 h down. A steam generator tube leak on January 25 accounted for 73K of this downtime (Table A.l.10}.

Other recurring problems were condensex'ube leaks (January ll and 14 and September 15) and packing leaks ar'ound the pressurizer spray and block valves (February 20) .

1979 Three forced shutdowns and seven forced power reductions occurred in 1979. Repairs beginning December 2 on a steam generator tube leak accounted for 38% of the total downtime this year (Table A.l.ll).

A gasket leak around, the B steam generator handhole forced brief power reductions on October 27 and Hovember 10 and.24.

Nearly a month was spent out of service starting July 6 because of an NRC requirement to inspect welds on the feedwater nozzles at the steam generators.

No condenser tube leaks caused shutdowns or power reductions in 1979; this problem may have been alleviated, although 1980 'data are necessary for confirmation. Note that of the 22 shutdowns caused by condenser tube leaks or degradation, 17 occurred after the switch to all volatile treatment (AVT) was made on Novembex 2, 1974. Similarly, all shutdowns caused by steam

4-21 generator tube degxadations occurred after. this switch.'owever, there is no conclusive technical evidence to link these events.

4.5 Review of Re ortable Events Appendix A.Z is a table of all reportable events at Ginna. A few pieces of information, usually dates, axe missing; blanks in Table A.2 indicate this.

Dashes indicate "not applicable." Occasionally reviewers were able to deduce facts from the sketchy reports; these. assumptions are denoted "A/." Shen the reportable event involved a forced shutdown, it is denoted in the "comment" column by the words "reactor shutdown." These events, then, appear in the shutdown Table A.l as well.

Of the 23 D>E -related shutdowns in Table A.l, 14 generated reports and thus appear in Table A.2 also. Of these, seven involved control rod malfunc-tions, five involved steam generator tube leaks, one loss of offsite power, and one spurious closuxe of the MSZVs. These are of varying degrees of significance, so the overlap between Tables A.l and A.2 does not in itself indicate particularly significant events.

Prom the plant startup in November 1969 through 1973, events were simply reported in letters to the AEC Directorate of Licensing and, fox the purposes of Table A.2, are numbered chronologically. Starting in 1974 the events wexe reported as "Abnormal Occurrences" labeled chronologically with an "AO" pre-Qx. Xn 1975, events were repoxted as "Unusual Events" (UEs) and "Reportable Occurrences" (ROs) as well as Abnormal Events. Tn 1976 and 1977, events wex'e reported as ROs only. Ginna began reporting events as Licensee Event Reports in 1978, and the other designations were dropped.

4-22 The reporting pattern over the life of Ginna is shown in Figure 4.1, with the trend toward more reported events. The years 1970 through 1973 averaged 10 reported events per year whereas. the years 1974 through 1979 averaged over 23 per year. The recurring failures discussed in Sect.

4.5.2.2.1 contribute to this trend. But even if these recurring failures are subtracted from the total number of reported events, the upward trend is still marked. (The year 1978 set capacity factor records and, thus, is below this trend.)

4.5.1 Si ificant events at Ginna As the initial step is identifying significant, events among Ginna's reported events, the reported events were screened by the significance categories listed in Section 3.2. The events meeting at least one of the significance criteria in the first step were reviewed relative to either (1)'he initiation of a DBE from Gi'nna's FSAR Chapter 14 (listed in Table 4.5), or (2) the compromise of a safety function design to mitigate one or more of the DBEs.

The events identified from this review were termed significant.

Two significant events, summarized in Table 4.6 resulted from the review i'f Ginna's reportable events.

These events were:

1) loss of offsite power, and

2) MSIVs closure.

0 30 20 15 10 0

1969 1970 1971 1972 '1973 1974 1975 1976 1977 1978 1979 Year

~ ~

Figure 4.1 Number of Reported Events

", at Ginna

~ I ~

4-24 Table 4.5 GINNA FSAR CHAPTER 14 DESIGN BASIS .EVENTS (DBZs) 14.1.1 Uncontrolled RCCA Withdrawal (k < 1) .

14. 1. 2 Uncont'rolled RCCA Withdrawal (at power).

14.1.3 Malpositioning of the Part Length Rods.

14.1.4 RCCA Drop.

14.7.5 CVCS Malfunction.

14.7.6 Loss of Reactor Coolant Flow.

14. 7. 7 'tartup of an Inactive Reactor Coolant Loop.

14.7.8 Loss of External Electrical Load.

14.1.9 Loss of Normal Feedwater.

Exc'essive Heat Removal Due to Feedwater Temperature Decrease.

14.1.11 Excessive Load Increase.

14.1.12 Loss of All AC Power to the Station Auxiliaries.

14.2.5 Steam Pipe Rupture.

14.2.6 Rupture of CRD Housing (RCCA Ejection) .

14.3 Primary System Pipe Rupture.

4-25 Table 4.6 SIGMTFZCANT EVENTS FOR GONNA Number of Significant Events'nd Number of Times Category Assigned Total S ignificance Times ed-Cate o 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 1979 Assi Sl S3 S5 S6 S7 S8 S9 Total Total No. of S ignificant Events

1. See Table 3.3 for significance criteria descriptions.

2. Blanks indicate 0.

3. See text for explanation.

i 4-26 The total number of significant events, tallied in the bottom row in Table 4.6, does not equal the total in the far right column because two events, 71-02 and 73&9, required multiple significance categories.

4.5.1.1 Loss of offsite ower leads to excessive cooldown rate. On October 21, 1973, the reactor was at 91X power, with one of four 115-kV transmission circuits'out of service because of construction. Auxiliary power was provided from the station switchyard and an external 34.5 kV line. A second 115-kV line, sagging from the increased load, flashed over to an underbuilt 34.5 kV line, and its circuit opened. The consequent 230-MRe power swing on the remaining two 115-kV lines caused them both to trip, resulting in complete loss of generator output abi3.ity. The loss of all 115-kV transmission lines overloaded a supply circuit'or the remaining auxiliary power source, causing it to trip and resulting in a total loss of offsite power plus a turbine trip and a reactor trip (Fig. 4.2).

According to an RGE letter on October 23, 1973, to the AEC Directorate of Regulatory Operations, Region 1 (NSIC 87031), disturbances on the instru-ment buses generated a double low level in both steam generators wnich started the steam driven auxiliary feedwater pump. This injected cold water into the steam generators and caused a rapid cooldown of the reactor coolant system (RCS) (an 85'F drop in the cold leg in 10 min, see NSZC 85369) which exc'ceded the Technical Specification limit. (TS 3.1.2.1.b states "for temperature above an indicated temperature of 290'F the [cooldown] rate shall not exceed 1QO'F/hr.") The rapid cooldown lowered the RCS pressure, and Sl was automatically initiated on low. pressure/low level in the pressurizer.

Io s tae Seaccste y toteatlel'evere

~ actoc ec tec ~ OI ~ 'I I IIILY toe ~ ot oualll ~ ry Coactor tf tp Sae res net Atu Setety la)ac Itewatsry lees Serlc *cl4 TIC ~ I Iastcu tslle4-lov socle Acl4 Stocste,.

ull tovcr CCeacaI ~ alen Ilae ~ ~ ewc courts 4w 4ue ts tales ~ Ieeel taupe Clos Aue to ot pover to Storcte scat bus IA yank level ln4lcstocs bcloce Core Aw to nclatcacaco, to overlos4lot overpovcc/At Cencrstoro ~ I ~ cc ~ rceeelvo vital lacteal yenb level powc reccoraa Iaclcstors cecurae4 to aornel Ocaets o Srova4 teult, ea4 ~ Itas I ~ I ~ Ct coo l4ova ~ cac bes IA la4lc crore veluesc tesulclaS In yrcaetuco poverc4 by

~ ewr aviate bus IA tell ~ ctety la)ection pwp crsnctor la ~ ltael ss4 subsequent setety 4ovncce la)set los peep prewture treater

~

troa boric Acl4 Srorsts teak to AOST, go coaccacrsce4 boric

~ cl4 soc cctulre4 So cetua4cat Iactrwcat pevec evetleblo No go osC ~ e ~ Ive lost Cern cool4ova prcvcate4 Iu ~ Tee Ilo

{turbine 4r Ives sill puup)

AFQ I <<~p ATCCS llo I

Cool4eva, Sctety ln)cctloa, failure ot ~ Tlrsl Iastcvnest Suo ot A. 5 Class ICIIC 55110, 410)I Actual Occurceaco toc tssc ot Ottslc ~ tover, Secoeelva ACS aa4 PSEuto 4.2 hCClDENT SEQUENCE PRECURSOR STUDY U)SS OF OPF-SlTE POMER EVENT TREE FOR EVENT 10/21I73

~ ~

4-28 At this point, RGE's letter of October 31, 1973 (NSIC 85370), picks up the sequence of events: power was lost for 38 s to instrument bus lA, causing a premature transfer of the SI pump suction from the boric acid storage tanks to the refueling water storage tank. Operator error was suspected, but no malfunction was found. The concentrated boric acid was not necessary for maintaining subcriticality, and no core damage occurred.

4.5.1.2 Both MSIVs close s uriousl . On June 23, 1975, the unit was operating at 88% of full power when a reactor trip occurred caused by a double low level in the "B" steam generator. The "A" and "B" main steam isolation valves had failed closed. The steam flow impinged on the leading. edge of the MSIV's disks and forced the valves closed. This was a common cause failure.

The same event had occurred 17 d earlier (June 6, 1975) and caused a shutdown (see Appends< A.l), but no failure cause was found at that time and no report was made, After the June 23, 1975, event, the rubber piston seat in the MSIV actuators was removed to increase the piston stroke, and 1 in. of metal was removed from the valve disk stops to permit positioning of the disk's leading edge out of the steam flow path. This event was mitigated without incident both times.

4.5.1.3 Conditionall Si ificant Events of Im ortance. The following two conditionally significant events are described in further detail because they represent partial loss of important safety system functions.

4.5.1.3.1 Safet in ection train B fails to actuate on demand.

~

Pour failures occurred after. an accidental SI signal on January 3, 1970. The reactor was operating at less than 1X thermal power, and repair personnel were calibrating one of six pressure transmitters associated with the safety injection logic system when the accidental signal was given. The reactor scrammed as designed,

4<<29 but the B SI train did not respond. A rel'ay failed in the open position.

After relay contacts were. cleaned,

~ s '

.the~ ~

.B

~

train actuated properly. The second and third failures in this event were malfunctions of two -different position-indicators on isolation valves inside containment. Finally, the motor-operated valve on the. 2-in. SI line to the B cold leg inside containment did not operate because of an open limit switch contact. -This event was

,flagged S3 because the four failures constituted a breach of the second line of assurance: the B SI t'rain failures reduced. the capability to miti-gate loss-of-coolant accident (LOCA) as analyzed in FSAR Chap. 14.3.

4.5.1.3.2 Loss of concentrated boric acid in ection ca abilit . On April 27, 1971, two concurrent tests turned up a design error that was not detectable by standard tests. A safety injection signal was given during a simulated station blackout. These two conditions were not normally tested concurrently. Shen station blackout occurs, the voltage is momentarily lost on all 480-V safeguard buses (Nos. 14, 16, 17, 18). Instrument bus 1B (1 of 4) is tied to 480-V bus 14 [with a backup supply from the main lighting panel (see FSAR p. 8.2-5 and Figure 8.2-5)]. Instrument bus 1B feeds the concentrated boric acid tank .level channels LC 102 and LC 172, (NSIC 66927, RO-71-02). Therefore, when the 1B bus voltage was lost, a downscale indica-tion was shown on LC 102 and LC 172. Low-level indication from these two channels actuates an interlock which prevents the concentrated boxic acid

\

tank outlet valves (826A, B, C, and D in FSAR Figure 6.2-1) from openingy thereby preventing all direct flow from the chemical volume control system (CVCS) concentrated boric acid. tanks to the suction of the SI pumps. The power supply for LC 102 and LC 172 was changed to a battery-supplied uninterrupt'ed bus.

'-30 The design purpose of the concentrated boric acid injection system is to provide protection from a main steam pipe break and potential return to criticality and power. During hot functional testing prior to initial

~

criticality, the fact that the momentary 1oss of bus 14 would cause this problem was not discovered because two different tests were involved.

4.5.2 Trends and safet im lications of re ortable events in addition to those cata prized as si ificant An overview of all the reportable events is given in Table 4.7. The system categories are those used in coding these events, and the number of times per 'year that a system is involved in a reported failure is tabulated.

The most frequently reported systems are high-pressure safety injection (HPSI) (24), main cooling. (23), chemical and volume control (19), reactor protection (16)., and emergency electric power (14). Fourteen of the twenty-four safety injection/HPSI failures were caused by recurring failures during start'up tests of the HPSI pumps .powered;by the emergency buses (detailed in Sect. 4.5.2.2.1). The CVCS failures did not begin occurring until 1974;

Table 4.7 Summary of Systems Involved in'eportable Events by Year n ~ ~

I System 1969 1970, 1971 1972 1973 '974 1975 1976 1977 I

1978 1979 Total

, A. Chemical and Volume Control 2 ~ 20

, B. Component Cooling 1 1

'. C. Condensate Purification 0 D. Condenser Cooling 0 Containment 6 F, Containment Air Cooling 1

. G. Containment Filtering Containment llydrogen Control Containment Isolation I

1 1 1

0 3

J. Containment Purge I t3 4 K. Containment Spray 2-l 3

)

. L. Core Reflooding 2, "5 M. . Electric Power 2 2 6 N. Emergency Cooling/LPSI 1 jO'. Emergency Electric Power 3 2 .2 2 14 Engineered Safety Features I t ~ 1 3 2' 6 .13

.q0 Fire Protection ~" 1 1 R. Hydraulic 1 I 1 =

Main Cooling 4' 2 2 '5 23 S.

. T. Pneumatic 5'5 0

U. Radiation Monitoring 1 1=

'. V.

X.

Reactor Control Reactor Protection Safety ln)ection/llPSI 1

3 1

1 3

I 3

4 4

t

-7 1

2

\

i 14 10 37 24 Y. Secondary Cooling/Aux FW 1 . 1' 1. 4 ll t I ~ I~ t I

Z0 Secondary Cooling/Feedwater 1 AA. Secondary Cooling/Steam 1 1 2 5

'2 BB. Service Water 1 4 1CC Shutdown Cooling 1 1.. 1 7 0

'DD. Waste Disposal =5 3

', EE. Ventilation ~

~

1 1

'FF Reactor Internals 2 t 2 5 TOTAL 4 10 6 12 '13; 21 27 32 31 10 31 197

=.1

'ore-than one system category was often assigned to each event. Thus, the totals exceed the Jt I I

actual number of events. reported. I ~

~

(": t I

4-32 since then they have averaged over three pez year. Fourteen of the nineteen CX'CS failures involved cxacks in piping and associated welds, all apparently caused by vibrations of the charging pumps.

4.5.2.1'auses of reported events, Causes of th'e 183 reported events are shown in Table 4,8. Only one cause was attributed to each event.

Inherent failures account for 44% of 'all reported events. Roughly 50%

of these were equipment failures caused by age and normal wear, and the other 50% had undetermined causes, which means that nearly a quarter of a11 reported events had undetermined causes. These events with undetex-mined causes represent a significant part of the operating e~erience data base at Ginna and should not be simply ignored by labeling them "inherent failure" with cause unknown.

The 30 administrative errors were generally.attributed to inadequate ox nonexistent procedures oz training and inadequate records and adminis--

trative control. Note that preopexational errors, design and installation erxors, account fox'early 23% of reported events. Greater care taken during constxuction can significantly reduce failures during operation.

Also, note that failures attributed directly to operator exror were not reported until 1975 and then averaged 2.2 per yeaz thxu 1979. The LER reporting system formally implemented in 1975, introduced new cause categories for reporting purposes operator error was one of these. How-ever, all human errors administrative, design, fabzication, installation, maintenance, and operational accounted for 56% of the total number of zeported events. Each year'hey consistently accounted fox about one half I

of all events reported. No significant trend in the reporting of human-induced errors is found.

4-33 Table 4.8 CAUSES OF REPORTED EVENTS 4

CAUSE 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 '979 ~

TOTAL

'L ~ Added.ni.s trative Error 1 3 2 5 6 3 .2 1 .5 31 Design Error 2 1 1 6 2. 16

~ ~ Fabrication Error 1 1 1 1 3

). inherent Failure 3 4 6 5 10 12 14 9 4 11 80 Installation Error 2 3 4 4 3 ~

25 Lightning 0 Maintenance Error 2 1 3 1" 2 '3 12 Operator Error 3. -" 4 3 1 11 Heather 0 Total 2 11 7 12 12 22 27 30 26 9 25 183

4-34 4.5.2.2 Trends in conditionall si ificant events. Though only two events in Ginna's life were labeled "significant," many events portended

\

problems that possibly could adversely affect the safe operation of the plant.

Eighty-four of these were 1abeled conditionally significant (C) and are summarized and discussed in Table 4.'9 and text that fo11ows. Two of these conditionally significant events were discussed earU.er in Sect. 4.5.1.3.

The thirteen C3 events offsite radiation releases are discussed sepa-rately in Sect. 4.6. The fifteen C4 events design, manufacturing, or installation deficiences, might have been prevented by diligently applied, routine quality assurance measures. The six CS events forced outages lasting longer than 10 d', a11 occurred in the second half of Ginn'a life.

Four were shutdowns to plug failed steam generator tubes (AO 75-07, RO 75-13, 76>>15, and 79-22). Both C6 events engineered safety feature (ESF) actuation, did not involve any loss of coolant; event 7i0-01 was an SI subsystem failure in a test, and event 70-01. called on the SX because of overcooling of the primary'ystem by the auxiliary feedwater in this loss-of-offsite-power incident.

The 34 C7 events recurring failures, are discussed in detail in Sect.

4.5.2.2.1. The 12 CS events other conditionally significant events, are discussed in Sect. 4.5.2.2.3 and 4.5.2.2.4.

4.5.2.2.1 Recurrin Failures C7. Failures recurring over the life of the-plant were labeled C7. Four problems are identified with a historical recurrent rate greater than or equal to 0.5 per year; and a fifth area, leaking snubber reservoirs (Table 4'.10) is included because of a recurrence rate of 1 per year for the last 3 years.

4-35 I

Table. 4.'9 CONDITIONALLY SIGNIFICANT EVENTS AT GINNA

'onditionally Number of Events Significant Cate ories 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 1979 Total Cl C2 C3 3 3 3 1 1 1 13 C4 1 2 2 1 4 2 4 16 2 2 2 6 C6 3 C7 1 1 1 1 2 2 4 5 9'. 3- 5' 34 C8 1 2 2 2 1 1 2 12 Total~ 2 3 .6 4 ' 8 10 13 11 4 14 '84

1. See Table 3.4 for conditionally significant criterion descriptions.

2. Blanks indicate 0.

3. The two totals here are equal, unlike for the significant events in section 4.5.1. No attempt. was made here to differentiate the total number of conditionally significant events from the total number of times the CS category was used (84), as was done for the significant events. More than one of the conditionally significant categories could be assigned to each event.

Table 4.10 Recurring Failures C7 Event 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 1979 . Total

' -10 failure to start 2 3 1 SI Pump(s) 1 Steam generator tube failures 9 1

Drive Malfunctions 8 Control Rod I Rod (s) fail to drop Rod(s) fail dropped Automatic control fails Diesel Generator(s) fail -to start Snubber Reservoir Leaking 3 Total C7 Events 35 I

Cl 4-37 All ten of the SI pumps'ailures to start were caused by the emergency bus breaker problems discussed in Sect. 4,5.2.2.2.

Steam generator tube failures'ecame a problem in 1975 concurrent with-the switch to all volatile chemical contxol (system modification 74-45, started during shutdown November 2, 1974}; tube thinning and corrosion problems- have not yet been solved.

None of .the eight contxol rod drive mechanism (CRDM} malfunctions reported over Ginna's life compromised its safety function. Watez from a leaking fitting in a feedwater flow indicator entered the power cabinet and grounded out the control power, causing rods G-5 and G-9 to drop on June 23, 1972 (72-09); the cabinet was dried and the leaking fitting was replaced.

Rod G-5 did not move with 'its bank during a xod control system test in i.- 1975, though the rod drop was not impeded upon a had an open circuit, preventing tzip signal; the normal driving {UR75-08}. Then lift coil in April and July of 1976, rods G-5 and G-9 dropped: the alarm indication was a station-ary "A" regulation failure', but in both. instances no cause could be deter-mined and the rods tested normally. Based on similar instances at other Westinghouse facilities, certain electronic components in the zod control circuitry were replaced with upgraded components (76-13 and 76-18). In August of this year rods G-3 and G-11 partially dropped. into the core. This was the first occurrence with these two rods, but the fifth such occurrence associated with 'the 2BD power cabinet. Again, no abnormalities were found (76-21}. Polarity checks conducted in January, 1978, indicated that 27 movable coils and 2 lift coils, had polarity reversals, resulting in too short an interval between stationary coil mechanism latch-in and movable coil mechanism drop-out. Since these polarity reversals were corrected, Ginna has.had'o further rod drop problems,

4-38 Either of the two station diesel generators failed during tests only five times over Ginna's life; they never failed on demand. This failure rate of 0.25 per year per diesel is remarkably low.

Steam generator or main steam line snubbers were found with low or no reservoir level in 77-13, 78-06, and 79-20, Excessive seal leakage on these vibration-damping components had not yet been remedied.

Finally, note the trend over the years at Ginna of increasing numbers of xecurring failures. These five recurring failures over the period 1975 through 1979 occurred at a combined rate of over 5 per year. This fact focuses future maintenance efforts on recurring failures.

4.5.2.2.2 'Emer enc bus breaker failures. Breakers in three positions marked 1, 2, and 3 in Fig. 4.3 failed 14 times 'in Ginna's life during tests of safety injection pump "C" (see FSAR Fig. 8.2 fox details), as shown in Table 4.11.

Table 4.12 lists the 14 breaker failures between the diesel generator "B", buses 14 and 16, and HPSi pump "C".

From this series of failures, apparently the conservative design of three 50% HPSI pumps is partially nullified by the unreliability of the cir-cuit breakers. From 1973 through 1979, the three breakers failed an avexage of two tests per year. The HPSl pumps are tested monthly (TS 4.5.2.1) this is the relevant test for breaker position numbers 1 and 2. The diesel generators are started in a monthly test (TS 4.6.l.a) this is the relevant test for breaker position 3 only. The diesel generators are also tested with a simulated

4-39 I

~ =

Emergenc Bus 14 Emergency Bus 16 2

DG A SI Pump A SI Pum C SI Pump, B DG B Figure 4.3 Safety Injection Emergency Power Supply

4-40 Table 4.11 Summary of Failures by Year Reportable Event Numbers Total Breaker No. 1 77-01 79-01 3 77-07 Breaker No. 2 73-03 74-04 AO 75-02 73-04 74-14 AO 75-03 AO 75-05 Breaker No. 3 UE 75>>05 77-19 78-07 79-18 4 TOTAL 14

4%1 Table 4. 12 Emergency Bus Breaker Failures Report Event Date Event Descri tion and Problem Solution 6/11/73 SI pump "C" fails to start in test. Problem not found (Reportable event 73-04).

73-04 6/12/73 SI pump "C" fails to start in test from bus 16. The cell switch in the breaker cubicle (breaker position No. 2 in Figure 4.2) was bent. Breaker was adjusted.

74-04 4/6/74 SI pump "C" fails to start in test from bus 16. The trip bar switch in the breaker control relay was out of adjustment. It was adjusted and the capacitor in the control circuit was replaced.

74-14 8/7/74 SI pump "C".fails to start from bus.16 in test. No cause found.

e AO 75>>02 2/14/75 SI pump "C" fails to start from bus 16 in test. Weak (Report date) spring in breaker found and replaced..

AO 75-03 2/14/75 SI. pump "C" fails to start from bus 16 in, test. Loose (Report date) wire in breaker found and repaired.

AO 75-05 2/26'/75 SI pump "C" fails to start from bus 16 in test. Lockout (Report date) solenoid was sluggish, did not complete its stroke. No repair done.

UE 75-05 9/15 Breaker in position No. 3 fails open during test of (Report date) DG "B". No cause found.

77-01 1/3/77 SI pump "C" fails to start in test fxom bus 14. The contact assembly in the breaker in position No. 1 (a W model DB-50, 600 V AC) was replaced.

77-07 6/29/77 SI pump "C" fails to start in test for bus 14. The en-tire breaker in position No. 1 (W DB-50, 600 V AC) was replaced.

77-19 9/14/77 Breaker in position No. 3 fails open during test of DG "B". Secondary contact finger was bent. A new secondary contact section was installed. The bent secondary contact finger is believed.to be caused by improperly returning the breaker to service after being in the "rolled out" position. Amd.liary operators retrained.

78-07 8/16/78 Breaker in position No. 3 fails closed and when finally opened it fails open during test of DG "B". Adjusted the W model DB-75, 600 V AC 3000 amp breaker.

79-01 SI pump "C" fails to start in test from bus 14. Breakers intermittently failed open. Alarm switch in the breaker's contxol wiring was replaced.

79-18 Breaker in position No. 3 fails open in test of DG "B".

Though a relay (W BFDL-22) was suspected, no change was made.

SI signal during every refueling outage (TS 4.6.1.b) this is the relevant test for all three breaker positions. Therefore, each breaker is tested 13 times per year, or, over 7 years, 91 times. Assuming tests were done per Technical Specification frequency and assuming one refueling outage per year, test failure probabilities for 1973 through 1979 are:

Total No. Total No. Test failure Breaker o f failures of tests probabilf.ty 2osisioo in 7 ears X)

A 3 91 3 2 7 91 3 4 91 The trend of the data indicates that this problem is not yet solved.

4.5.2.2.3 Common cause failures. Table 4.13 lists ten common cause failures that were reported at Ginna. Two failures are considered significant events (71-02 and UE 75-03) and are described in detail in Sect. 4.5.1.

Two control rods dropped when water leaked into their shared cabinet on June 23, 1972, and again on March 5, 1975. That problem was solved by welding a leak-tight cover to that cabinet. Three common cause events (72-09, UE 75-01, and UE 75>>03) turned up during operations, and three common cause failures t

(78-06, 79-12, and 79-15) were identified during inspections. With the excep-tion of 71-2, discussed in Sect. 4.5.1, none of these common cause failures significantly affected the safety of the plant.

4.5.2.2.4 Other conditionall si ficant events. Twelve conditionally significant events, flagged C8, portended significant problems but did not qualify as S or C1-7 events.

4-'43 Table 4. 13 Common Cause 'Failures Report Event Significance Numb er Date Event Descri tion and Problem Solution 71-02 4/27/71 C4 Redundant level channels in BAT fail down-scale due to momentary loss of instrument Bus lB voltage during coincident station blackout and SI tests, causing BAT-to-SI valves to close. Level channels'ower supply switched to DC source.

72-09 . 6/23/72 C7 Two control rods drop due to a water leak into a sh'ared power cabinet.

73-11 12/21 C4, C8 Both motor-driven au&liary feedwater pumps (report airbound in test due to bubble in common data) header from condensate supply. Cause of bubble uncertain. The pumps were vented every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until corrective action (unknown) was taken.

74-11 6/26/74 C8 Five of the eight solenoid valves which con-trol the MSIV's fail in test due to over-heating. Could not close MSIV's. Over-

,heating due to solenoids being left energized too long in test. No corrective action taken.

UE 75-01 3/5/75 C7 See 72-08 above. Cover welded onto cabinet.

UE 75-03 6/23/75 S2 Both MSIV's fail closed due to steam im-pinging on leading edge or valves'isks.

Valves ground down so they sit out of steam flow path.

78-06 7/13/78 C7, 8 Two snubbers on main steam system fail due to constant vibration of system. No

• corrective action was taken other than re-placing failed scrubbers with tested spares.

79-12 3/27/79 C8 Six anchor bolts for piping supports for equipment" not up to specifications. 'safety Corrective action see IE Bulletin 79-02.

79-15 7/24/79 C4, 8 Ten piping supports not properly installed Dx containment spray, residual heat re-moval, and service ~ater systems. Supports re-worked and re-analyzed (See IE Bulletin 79-14)-

4&4 Five of the CS events were common cause failures and were discussed in Sect. 4.5.2.2.3. Of the other seven, two similar events were reported in 69-02 and 74-06, both involving a failed valve in the 10-in. line between the containment sump and the residual heat removal pumps. On December 16, 1969, both disks of a valve were bowed outward from internal pressure caused by heatup of water trapped inside the valve by ambient reactor coolant water.

Similarly, on April 22, 1974, water trapped between two valves in the same line was heated, and the consequent pressure seated the valve so tightly that it would not open on a signal from the control room. Operating procedures should preclude this problem.

Qn June 30, 1971, when a thixd condensate pump was put in service, a series of failuxes occurred that led to a condensate line water hammer and a reactor scram (71-04). The safety of the plant and pubs.c was not endangered, but the event sequence deserves some attention. Increased condensate pump outlet pressure, resulting from the added condensate pump, caused the hydrogen cooler temperature control valve superstructuxe to buckle, and the valve failed closed. The hydrogen cooler temperature rose, and the normal condensate bypass valve consequently closed. This caused an emergency feed valve to open, supplying relatively cold condensate directly to the main feedwater pumps'MFA} suction, which induced severe vibrations in the MFA's suction line and tripped HFrP A. MFA B was tripped manually, tripping the turbine and then the reactor in the normal logic sequence. This is the only condensate line water hammer reported at Ginna.

4-45 On July 22 1973, a water hammer in the feedwater line to the B steam generator damaged several supports and insulation at'various".locations the line (73-06). Failure of a small pin which prevents rotation 'long of the control plug on the B feedwater control valve caused the control plug to separate from the threaded stem of the valve, inducing rapid flow variations, causing the water hammer. According to a RG&E letter to the AEC Directorate of Licensing dated August 21, 1973, the valve manufacture was consulted and subsequent investigation revealed that the original stem on the B feedwater control valve had been damaged during construction.

A replacement stem had been installed, but the recommended torquing pro-cedure had not been followed during reassembly of the new plug on the stem.

Had the plug been torqued properly, the holding pin would not have failed.

RG&E has since stated that a design change has been made in this valve operator.

On June 26, 1974, following maintenance on the main steam isolation valves, the four solenoids on each of the MSIVs were left in the energized position. During subsequent testing, five of the eight solenoids failed; but, after they were allowed to cool, they operated correctly. 'nadequate maintenance procedures were responsible for this failure (74-11).

Both discharge valves for one of three 50%-capacity SI pumps were found closed in a test on June 25, 1975 (AO 75-10). The recognized unre-liability of the HPSI emergency bus breakers (Sect. 4.5.2.2.2) makes this operator error more significant.

The administrative oversight reported in UEK 75-06 is significant because it represents.a deficiency in quality assurance procedures.

A modification was made in which a steam generator level signal input was transferred to a bus that was already part of the lo-lo steam generator level protection system and the reactor trip logic for steam flow-feedwater flow mismatch. On discovery of the error only hours later, the system was returned to its original configuration.

An actuator in a control room ventilation damper was originally installed incorrectly and was not discovered until March 8, 1976 (76-11).

Previous tests'ad assumed that actuators had been properly installed.

Proper tests verify functionability of a system and not gust its components.

4.6 Events of Environmental Importance A summary by year of the total radioactivity released from Ginna is shown in Table 4.14, organized by airborne and U.quid.effluents and by off-i site shipments of solid waste. Because of varying reporting requirements in 1969, 1970, and 1978, some data are not available. The 1979 data has not been published as of April 1981. An overall increase over 10 years is apparent in the activity released as airborne tritium and as airborne noble gases, with the notable exception of 1972 which had the largest releases of noble gases.

Note the spike of mixed fission and activation products released via the U.quid pathway in 1970. Approximately 88% of this ad<ed activity was discharged in March and April 1970 because of increased activity in the primary coolant leaking from the plunger leakoffs of the charging pumps.

Increased primary loop activity detected at 100X power in March was first evidence of leaky fuel. On March 25, during full power operation, and

~ ~

Table 4. 14 Summary of Rsdioaccivicy Released From Ginna EFFLUENT (CURIES) 1970 1971 1972 1973 . 1974 1975 1976 1977 ~ 1978 1979 AIRBORNE:

Total Noble Gases 1.00Et012 3.20Et01 1.18Et04 5.76Et 02 7.57Et02 1.04Kt04 5.52Et03 3.20Et03 9.27Et02 6. 84Et01-Total 1-131 Nh Nh 2.85E-02 5. 31E-04 2.82E-04 Nh 3.17E-02 2.54K<<02 1.02E-02 1.91K-O2, Tocal Nalogcns (Including I-131) HAI Hhl 3.00E-02 9.8 E-O'4 . 4.46E-04 2.69E-02 3.718-02 '.63E-02. NA 2.98E-02 1

Total Particulaces (TII>>8 d'ay) 5.00E-02 1.708-01 . 7.81E-OS 3.3 E-05 4.19E-05 2.008-02 8.958-05 7.038-05 1.43Kt04 5;50E-03 Tocal Tricium Nh Hh 8.8 K-03 1,15Et 00 3.56E-01 HA 2.36KtOI 5.00Et01 4. 38Etol 5.67EtOO LIt)UID:

Total Mixed Fission 6 1.0 K+01 9.0 E-01 3.0 8-01 7.5 E-02 1.00E-01 4.21E-01 6.908-01 6.47E-02 6.078-02 1.278-01 t

Activation Products Total Tritium 1.08Et02 1.54Et02 1.19Et02 2.86Et02 1.95Et02 2.6IEt02 2.42K+02 1.19Et02 '.42Et02 1,26Et02 Dissolved Noble Gases 'HA HA HD 3.038-04 ND Hh 3. 93Et02 .HD Hh 9. 05 K-03 SOLID:

Total 4.64Et00 4.73Et01 1.41Kt03 . 5.99Et02 6.14Kt02 1,07Et02 9.788%01 6.90Kt02 6 '78t02 3 '2Et02 l From HUREG-0521, Radioactive Haccrials Released from ltuclcar Poucr.Plants.

The rase of che data uas compiled from RGSK's semi-annual snd annual operating reports.

2 Hoc avail able.

'3 Hone detected.

~ ~

4-48 again on April 24, during startup after a month-long scheduled maintenance outage, the maximum concentration of liquid released (2.4 E-07 uCi/cc, excluding tritium) exceeded the techni'cal specification limit of 1.0 E"07 pCi/cc but did not exceed the maximum permissible concentration (MPC) isotopic limits; isotopic analysis showed Iodine-131 to be the only isotope discharged at a concentration greater than one-tenth of its MPC. (This value for I-131 could not be found.) At no time did the concentration of tritium in the discharge canal exceed the technical specification limit of 3 E-03 pCi/cc.

Four radiation exposure incidents and nine offsite releases are sum-marized in Table 4.15. Thirteen of these appear in Table A.2, all marked C3.

One incident was reported in a letter to the AEC on June ll, 1971, in which two concurrent yet independent leaks were surmised to have exceeded technical specification release limits (71-03). Another incident was reported in a letter to the AEC on February 9, 1973: a maintenance error on the RCS letdown line caused a 47-h unplanned release of radioactive gases on June 10, 11, and 12, 1972 (72-07). Note that 1972 airborne noble gas releases (Table 4-14) were the highest ever at Ginna, exceeding 1971 and 1973 releases by over a factor of 100.

The four overexposures aU. resulted from administrative errors and lax health physics procedures. No equipment failures caused overe~osures.

Of the nine offsite releases, five occurred during operations and four P

during maintenance. The CVCS was involved in all but one of these incidents.

4-49 Table 4.15 Radiation Over-exposures and Releases Event NST.C Event Retort . Number 'Date ..

70-07, 49878, 8/19270 ,' chemist received 4e2 rems in a 2-month (report period. Procedures shortened.

date)

Letter to , 6/1-615/71 Xenon leak from a pressure control AEC (6/11/71) valve (on the volume control tank) to the auxiliary building. Waste evaporator vent concurrently vented I-131 to auxiliary

'building during shutdown of charging system.

A continuous recording iodine monitor was installed on the stack vent. Amounts re-leased uncertain.

71-04 64850 ., 7/21 /71 A steam fitter received 3.09 rems in one (report quarter.

date) 71-05 '7990 10J6J71

{report 6.7 mCi of I-131 released to atmosphere through the plant vent during a 17 hr.

date) period. Monitor failed. Boric acid con-centrator drain flushed.

72-05 70029 4/20/72 I-131 releases in excess of Tech. Specs.

coincident with, processing of reactor coolant from the C-holdup tank through the boric acid evaporator.

Letter to 6/10-6J12/72 Krypton-85 released for 47 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br /> (unplanned)

AEC (2/9/73) due to RCS drain valves left open during maintenance. Maximum release rate was 8.4 EM4 CiJsec, maximum concentration was 2.0 E-05 pCi/cc.

72-11 77411 12J28/72 228 Ci released in 24 'hours, 6X of Tech.

(report Spec. limit, during the first spent resin date) removal operation (from spent resin tanks).

Procedures modified.

Notification 89263 2 J20J74 40 contractor maintenance personnel re-No. 107 (report ceived over 3 rems in January while re-Let:ter to date) ' pairing 1eaks in the spent fuel pit, AEC 90662 J25 774 74-08 94198 5/11J74 Leak'n one of two mixed-bed demineralizers (CVCS) to aux'iliary bui1ding. The highest release rate was 1.63% of Tech. Spec. limit.

4-50 Table 4.15 (continued) Radiatioa Over-exposures and Releases Eveat NSIC Event

~Re urr. Number Date t

74-10 94522 6/2U74 A primary sample system flow indicator (report was reinstalled incorrect1y,.leak filled date) chemical drain tank, high level alarm failed. In 2 ho'urs, 0.9 Ci of noble gases was re1eased.

Letter to 109653 5/19/75 Oae maintenance worker's uptake above NRC (report limits. The dose to the lungs from date) Co and Zr was 9.6 rems for the first.,

Letter to 105548 8/23/75 year. Improper use of masks due to RGE (report administrative error. $ 10,000 fine date) proposed, 76-19 115728 7/16/76 11 low level releases were made from (report the main water treatment demineralizer date) (CVCS) neutralizing tanks without the continuous use of a gross activity mon-itor due to a desiga modification omission.

Maximum conceatration released was 2.25 E-07 pCi/cc.

79-02 146698 1/4/79 Cracked charging pump cylinder leaks to plant vent. Amount unspecified.

4-S 1 4.7 Evaluation of" eratin Experience Reactor availability at Ginna has been among the industry's highest, averaging 78.1% over the ten full years of operation. Equipment failures dominated the shutdowns, 1

accounting for 737 of the total causes; however, equipment failures accounted for only 44X of the'licensee event reports and their predecessors (hereafter referred to as zeportable events), a difference that vill be discussed in more detail in this section.

Whexeas trends in shutdowns and power reductions generally followed trends 'in reactor ave.ability, reportable events had two distinct phases (Fig. 4.1). Through 1973, reportable events were generated consistently at a rate of about ten per year; then, in 1974, these rates doubled and averaged about 25 pez year through 1979. The exceptional year, 1978, with only nine reportable events, set the record for capacity factor at 78.27.. This is the only reason found for the low number of reportable events in 1978.

Tvo points of view provided by the analysis of shutdowns and power reduc-tions and of the reportable events complement one another'n assessing the operational history of the plant. The analysis of the shutdowns is most use-ful in identifying accident sequences, whereas the analysis of xepoztable events is most useful in providing specific information and in identifying specific failures tuithin accident sequences. The distinction arises from the

4-52 fact that the plant is designed to shut down nlhenevez an unsafe operating condition exists, such as when an P

- accident sequence has been initiated; whereas an event report is generated to fulfilla specific technical specifica-tion requirement, such as a limit bounding the accident sequence or a com-ponent failure that limit the plant's capability to respond to an accident sequence. Reportable events tend to be more component oriented, identifying single events or occurrences. Whereas Ginna has ezperie'aced 118 shutdowns, reportable events totaled 181 ~ Reportable events cover a wider range of events, including causes not categorized in the shutdown tables (such as test-induced failures during a shutdown and all design and fabrication errors).

Reportable events cover all actual failures, including releases, ma)or outages and ma5or damages, and all SIs, plus potential failures identified through operations or reanalyses with impact on plant safety. Nearly one in four reportable events had undetermined causes (Sect. 4.5.2.1).

Conclusions about the overall operational safety of the plant do not vary when taken from shutdowns and from reportable events. Safety significant events consistently are identified in both ways. The two points of view are complementary in individual cases. For e~mple, the MSIV spurious closures were reported by shutdowns on both June 6 and 23, 1975, but only by reportable event on the latter date, when the cause was surmised and corrective action was taken. For review purposes, the complementary aspects of the two points of view are apparent from the interwoven nature of Sects. 4.4 and 4.5 on shut-downs and reportable events, respectively.

4-53 One result of the reportable events different from that of the shutdowns concexns preoperational sources of error; design" and installation errors.

-- account for 23X 'of all reportable events (Sect.. 4.5'.2.2). Shutdowns do not acknowledge this source'f error explicitly.

Human factors cause a higher percentage of reportable events, by a factor of 2, than of shutdowns . This difference in human causes 56X of the repoxtable events to 27X of the shutdowns is discussed below...

For reportable events, human'actors 'include administrative, design, fabrication, installation, mainten'ance, and operator errors -'n short,. all those causes listed in Table 4.8 ezcept inherent failures, lightning, and other weather. For -shutdowns, human factors include 'maintenance or .testing, administrative, or operational errors (item I.3 of Table 4.2). This categoriza-tion scheme itself tends to attribute human factors as cause to more reportable events than to shutdowns. In spite of this inherent bias in the categoriza-tion scheme; same valid conclusions can be drawn.

Operator errors as.repoxted causes for shutdowns were reduced to one in 1972 and none thereafter, and no evidence exists in the shutdown reports to suggest that operator-error-induced shutdowns occurred after 1972 but were reported as having other causes. Human factors as defined above caused shut-downs throughout the plant's life, though again the first years of operation were the worst here, with over one-third (10 of 28) occurring in 1970 alone.

4-54 Reportable events, on the other hand, attribute over half of their causes to human factors. No reportable events were attributed to operator errors thxough 1974; however, in 1975 through 1979, they avexaged ovex'wo per year (Table 4.8), Human factors averaged 7 per year'hxough 1974 and from 1975 through 1979 averaged 13.4 per year. 'This increase, however, simply reflects the overall trend of increased numbers of reported events (Fig. 4.1); human factors accounted for roughly 50Z of all reportable event causes each year from 1970 thxough 1979. The difference in causes for shutdowns and reportable events indicates that while human factors can frequently incapacitate com-ponents and can influence tests and maintenance operations that in turn generate reportable events, they are much less likely to incapacitate systems that in turn generate reactor shutdowns. Also, as indicated before, 27Z of reportable events were generated by preoperational causes (design, fabrication, and installation errors) all of which are caused by human factors.

=Tests helped identify many potential safety hazards at Ginna. Tests uncovered significant fault's as reported in (1) AO 75-10, when both of SI

,pump C's outlet valves were found closed, and (2) 76-11, when 25Z of the control room ventilation filter capability was found inoperable. Concurrent tests run on April 27, 1971, uncovered a highly significant fault in the emergency power supply for all four boric acid tank (BAT)-to-SI valves.

Never did failures resulting from tests jeopardize the safety of the plant or nullify the design capability of a safety system.

4-55 Inspections turned up the recurring main steam and steam generator snubber leak problems as reported in 77-13, 78-06, and 79-20. Inspections also found potential common cause failures involving safety equipment anchor bolts (79-12) and piping supports (79-15). System modifications did not cause problems except for one in 1975 when an. inspection found that no separation of redundant function was provided for a modified'afety system (UE 75W6).

Although 1979 was a very good year at Ginna, recurring problems with condenser and steam generator tube failure are not yet solved. Problems here have consistently caused shutdowns from 1975 through the present. Condenser tube failures were identified through power reductions; reportable events never were involved.

Other recuxxing problems leading to shutdowns seem to be solved: EH turbine-generator governor control problems, not a factor since May 22, 1976; instrument bus inverter failures, not a factor since June 3, 1976; manual feedwater control prqblems, never a factor except in the first 8 months of operation; main feedwater pump impeller failures, no occurrence since September 10, 1976; MSXV spurious closuxes, not a factor since June 23, 1975; pressurizer spray valve packing leaks, not a factor since September 5, 1972; and turbine blade failures, not a factor since August 7, 1976. Contxol rod malfunctions, though they have not caused a shutdown since November 17, 1977, have not been resolved.

No enyironmental considerations ever constrained operations at, Ginna.

Only once, in March and April 1970, did the liquid effluent require isotopic analysis and comparison with maximum permissible concentrations befoxe being xeleased to the discharge canal.

5-1 REFERENCES

1. "Standard Review Plan," Chapter 15, U.S. Nuclear Regulatory Commission.

2. Operating Units Sta~ Beport lricensed Operating Reactors, NUREG-0020 Hay 21, 1974.issue through January 1980 issue, Vol. 4, No. 1.

3. Nuclear Peter PEant Operating Experience During 2978, OOE-ES-004, U.S.

Atomic Energy Commission, December 1974.

4. Nuclear Pomr Plant Operating Experience 1974-2975, NUREG-0227, U;S.

Nuclear Regulatory Commission, April 1977.

5. NucEear Pover Plant Operatin~ Experience 2976, 'NUREG-0366, U.S. Nuclear Regulatory Commission, December 1977.

6. Nuclear Pover Plant Operating Experience 2977, NUREG-0483, U.S. Nuclear Regulatory Commission, February 1979.

7. Nuc2ear Pover PEant Operating Experience 2978, NUREG-0618, U.S. Nuclear Regulatory Commission, December 1979.

8. Semiannual Report No. 1, Rochester Gas & Electric Corporation, Ginna Station, Docket 50-244, July 13, 1970, p. 18.

9. Semiannual Report No. 4, Rochester Gas & Electric Corporation, Ginna Station, Docket 50-244, February 25, 1972, p. 5.

10. Nuclear Safety, Vol. 11(3), May-June 1970, p. 241.

11. Nuclear Safety, Vol. 11(4), July-August 1970, p. 328.

12. The block valve leaked enough to cause shutdowns, however, on December 14, 1972, and February 20, 1978.

>3 ~ nuclear Safe+, Vol. 15(1), January-February 1974, .p. 88.

14. Semiannual Report No. 10, Rochester Gas & Electric Corporation, Ginna Station, Docket 50-244, (no date), pp. 3, 23.

~

Appendix A: Ginna Part 1. Shutdown and Power Reduction Tables

CIHHh SEP Table hl.l PORCED OIIThCES hHD POHER REDuCTIOHS R<<portable DBE(D) /

Dace Duration Powur Shutdown Sys tea Coaponent HSIC(H)

Evenc Description lnvulved 'Eve nc (19bp) (lira) ktothod: involved Category

1) l 1-29 l. 3 High cond<<user pressure. Valve bistable left closed after calibra-to, C Stone 6 I &C H6.4 Power tion. (IIC).

2) 12-3 5.1 20 Ltr Double-low SC lovel, Succion volvo C Scone & Valves H6.1 12-12-69 to 1B IIFHP niatakenly closed. (One =

1'ower shutdown rod failed to drop- (IIII) ~

soo letter.)

3) 12-3 ,8 20 Double lnw SC levul, PM bypass Scuse 6 Valves . H6.4 valve waa nor. nanually ruset. Power (IIII)

4) 12-3 .7 20 Double low SC luvol. PM bypass Stean 6 Circuit ' H6.2 volvos wor<< preaLaturaly put in Power cont roll<<rs/

aucueatic. (IIII) interruptura

5) 12-4 .8 24 Loss of both RCPs. fiuid fron gon- C Roaccor Puaps H5.1 crater was nanually rcvxovcd before, Coolant 4160 V busses weru switched. (CB)

6) 12-6 .7 22 Double low SG lev<<l. Excosslvo ~ Stean & ~ ~ I Puaps vibration of seal water differen- Power.

tial pressure nercoid tripp<<d .(IIII)

)IFIIP.

I

7) 12-14 15. 0 30 To clwck condenser hotwell Stuaa and lloac "prob ion", Power Exchanger (IIC)

8) 12-15 .7 2 Low seal water differential pressure h Stone and Puops co Hflll', Pow<<r (IIII)

ClNNh SEP Tablo hl. 1 FDRCFD 00rhces hND PDNSR RSDDCTIDNS DDE(D) /

Duration Pouur Ruportablu Sbutduun Systula Couponunt NS1C(N)

Datu Duscrlpt ion Causa gvunc (1969) (Nrs) (Z) I I d lllvolvud lnvolvud Cacugory

9) 12-16 36. 9 Tucll. Spuc. ruqulruuunc to slruc doun Knginuurud Va lvus Nl.2.1 duu to falluru of sufi.guurd volvo f Si'I u t y NOV-8500 to upun fully during Fuaturu nontbly survulllancu lust. Dutb (SF) vulva disks uuru bouud out duu to prussuru lntl.mal to tleu valvu.

10) 12-30 'FF/SBOI Illth lou SC luvul. Ih AFIIP Stuaa and Puups uas uot uvullablu and stuau drlvun Pouur hllN'aa uluu storting. ID (NN) supply unougll uut'ur fur hHII'id nelc tbu loud.

11) 12-30 17.0 TCCD ~ Spuc ~ fuelull'vaunt to alultduun IL 1 Scuuu aud Valvuu duu cn falluru uf valvu on Jls-, Pouur cbargu of 1D uotor-drlvuu *k'MP. (NN)

~ ~

ClIHIA SEP Tablo h1,2 FORCED Our&CES AHD IOWER REouCTIONS DBE(D)/

Date Duration Poucr Reportable Shutdoun Systea Cooponent HSIC(N),

Event Descrfptfon Event (I 970) (Nrs) .

(2) Huthod Involved Involved Category,

1) 1-3 1.8 SI duc to 2/3 sl8nals of lou'prcs- ,8 l&C I&C c N5.2 sure fn 8 aaln stcaa lfne, Durfng (18) calfbratfon of prcssure channel 483 I&C uorkers closed a leakinS volvo fn pressure channel 478. (Notus 8 ~~ '.i a trafn SI fallcd to operate. Scu f I I,

~ ~~

Nonl.bly Report ll, p. 52.) ~ 1 I

2) 1-3 15.7 NOV-8783 on SI lfne to 8 cold le8 En8fnccred Valves uas found fnoperablu. Safety Features ')

. (SF) 1-17 Steaa lunk thru Basket on turblnu Stcaa and Valves stop valve. I'nuu

~

(HA) r, . ~, -:I I N3.2 I I.

4) 1-17 1.6 2 Haffunctfon nf thu rod st<

Coaponent IISIC(tt)

Date Duration Pouer Kvcnt (1970) (Urs) (2) Description ttethod Involved Involvvd Category

7) 1-28 .9 23 Overteupuraturu delta T duo to dirty 8 IAC 16C II2 . I contact on Channel IV delta T relay (Ih) ln 8 ruactor trip logic.

1-28 .3 40 FFISFISI and lo SG lovi>l. Ilt valve 0 Stuau and 16C tt6. 1 8)

A closed on high SG lovul and Puuer operator cuuld nut e>anually control (UII) tl><'<.sul Clog I<'vel drop r

9) 2-22 .2 10 During 4160 V bus uudorfro<tuoncy 8 Klectrtc Kl<>ccricsl tt6.2 toots, both busses uuro lnadver" Pouer Systeuo Conductors tently tied together. (Kg)

10) 3-2 40 70 To repack clio PZII spray valves. Rvector Valves tll.I. I CnulanC (CS) 3-30 100 FF/Stlgt and lo SG level. Opurator Stvau end 16C II6. I 11)

<'Iuild >nit e>:w><nilly enurr<>l SG li'vol 1'<>u<ir noel I 1st lu>>:I liub>I>'<Ill>'I>l tu IN>>uua I (UU) tnkonvur of IQ contrul.

12) 5-14 816 85 Fol loulng a IOUX full por trip teat und a turbine lnspvctlon, plsnC reuato>.d chutdu<>n for i a) Turblnv r<ipnlra un l2 Ll'<<ctlun Stean and Tranofor<>ors Nl. 1.4 l>lading (roplac>.d all 10th etage Pouor bladee al>d 4 11th stage bladoll ~

(IIA) dauago duv. to furolgu ub)oct) ~

b) Replaceuout uf all RTDa. 16C 16C tt2. I

( ID)

TIIK 5-18-70 c) tISIV-IS did uot closu upon ueuual Stone and Valves tt 1. 1. 4 l.tr,5-22-70 signal fi'nu control ruuu Tho ~

Poucr packing gl>u>d uas ad)usted a<>d Che (UII) valvu closed.

~ ~

CINNA SEP Table hl.2 PORCED OUTACES AND POIIER REDUCTIONS DBE(D)/

Duration Pouor Rvportablu Shutd nun System Coaponont NSIC(N)

Date Description (1970) (lira) (2) .Nuthod ~ Involved lnvolvvd Event Category Sl relay comical serous uore louse, h Engineered I Relays N2:1

13) 6-18 14. 9 0 relay flul.cured causing 8 trip Safety Feature it i bruaker to opvn. (SP) ~

'Stean snd '1&C ' H6. 1

14) 6-19 .3 15 Double-lo SG level. Nhile the PII 0 uas on Dsnual control the turbine Pouer ~

uas intentionally crippvd to prepare (NII) for Ell govvrnor rvpair. Thv. result-ing pressure transient caused the double-lou level. '5) 7-5 . 4 83 Ltr 7-15-70 Pouer reduction. Turbine takvn out Stean and Pipesi N3;2 Ltr 2-17-71 to repair a stean leak in a turbine Pouer fittings high prvssurc gage line. Other (HA) na lntonanco included: added packing to I'ZR spray valve PCV-43lh. checked thv stroke on P28 spray valve PCV-431D ~ and ad)usted tha turbine trip pilot valve. I

16) 7-12 24.7 0 PZR pressure relief valve NOV-516 Reactor Coolant Valves N3. 1 leakaRv rate rvqulred shutdoun.

(CB)'ilva>a 7-25 2. Oi 27 Pp/SBLtl neil liiu SO Ivvol duo to nnd 'Puapn <

I 7) i llllIP A trip. I'nuvr'HH) i, ~,

I u ~

~

~

CIIIIIA SEP Tab)a h1.2 roIICED DDTACES hlln I DMEII KKDUCTIDIIS Ilvportablo Coaponvnc

~ DSK(D) /

Uacu Duration Power Shutdoun Sys co<<i IISIC(II)

Event Doscrlpclon Causu lnvo lvvd Involved Evanc (1970) (Ilrs) (Z)

Category LS) 9-5 14.5 37 Lou SC Levvl plus lou EII f lou Stean und Va Lvu II L. l. Ai occurred after curblnu trip, uhlch Pouor uas done iu<nuslly duu lu appar<inc (IDI) turbine vlbrat luna. I'urtbor lnvoa tlgatlow roveuled tlui canoe Co be fII osclllal lou to tbu 0 SC iluu to

~

loosu packing uf tliu I-g IIEM coact'ol valve.

19) 9-6 N 50 Ilops lrs onl

h. KII control aysteia. Stean and llecbsnical Piiiior function

~ (IIA) units II. IIAIL' aucclon relief valvus. Stuau and Valves. Ill.L.A I'ouvr (IDI)

20) 9-30 335.1 100 To ri..pair a leak on a reactor coolant h IlC IAC II3 ~ 1 ti:uporacuri: detector, plus: replaco- (IE) uonc uf DTlin, riipnlr of utes<<i awd uatur lvaks, correctlvu s<alntunanco of 0 )II'III'. replace>>out of pl<niger In IQ i<<id II, cliiifgliigUn&ps leplncl;

~

uvnt iif cwnnvct lwg riid nnd r<ipwck sil'ill>> lii IA is<i><sill<:Ilii pwap< ~

I<<nial lnl Inn i<1 dii<ira an<I Dali<>> lw hlgli r.i<Ill<Ii<>>i iirii<<i, cl<.<<<II< I Ill< rn

inil r<<lwfiircu (lllur frau< unrk iin plant v< wl >>yntvu IIEUA flllcr biink.

i<<id f<<<i<id 5 fr Icllii<.'0 nllcli<<r I<<ill>>

nn clio 0 IICI'iiiplliirtis<i<i ri plncvil all iif lb<i 56 anclnir biilln for clio

<<tva<<< gvnori<tnra anil IICI'a (iii'<i report ullh Ivllvr 12-21-70).

CIHHh SEP Table h1,2 FORCED OUThCES hHD POllER REDUCTIONS DSE(D)/

Dace Duration Power Reportable Shutdown System Component NS IC(H)

(1970) (Il re) Deecrlpcion Causa Involved involved Event

~ (2)

Category

21) 10-29 1.6 100 Fault on trunk line l23 cnused back h 31 Electric Power Relay Hl. l. 4 distance relay to trip, causing 'p Sy st {ace turbine trip. (Eh)

22) 11-1 ~ 4 100 Generator automaclc voltag<< regula- h Steam an{I Circuit N2.1 cor failure caused turbine trip. Power closure/

(Ilh) intarrupcurs

23) 11-2 7 6 100 C<<n<<racor automatic volcnga regula- h 3 Stun{a and Circuit N2.1 tor failure caused curbluc trip. Power clos{:rs/ ~

(IIh) 'ncerrupters

{

24) 11-15 8~ 7 PER relief valve leakage, Valve Reactor Coolant Valv{:s was repackcd, ~

(CS)

'1.2.4

25) 11-23 1. 3 0 Tame for ba/.tery ground. 3 Electric Power Datcari{:s on PCV-4316 control power. I and chargcrs Ground'ocar{.d Systems 1 (EQ)

26) 12-12 38.7 Hafntcnenca and r{.pairs:

h. P7R power relief guard valve l)eaccor Coolant Va1've" H3. 1 HO V-515 r{:pa c k{1d. (CS) {

S. Plugg{.J cubes ln cond{.necr h. St{.am and Heat'"1 H3. 2 Pow<<.r e){changers

, (IIC)

27) 12-31 76. 8 0 A<<pair of: I~ {

h. Hoisture scparatog ruhaacers. Scca{a and Power )lest Nl. l.4 (HC) exchangars C

S. Condenser tube leaks. Steals und Haec N3. 2 Power <<xchang<<rs (IIC)

Unlc left shutdown for scheduled maintenance.

t Scgedulad tact/unecl{edu)ad repairs uncauncur{xl, Turbine generator r{eaoved from s{.rvlca only. Reactur was controlled by boration with thu D control rod bank at 140-180 steps out, Powur level corruepond{.d to 10(exp <<7) amps on tha incarnadine<<power range charm<<l.

CLIUIA SEL'able A1.3 L'ORCFAI OUTAI'ES AND I'OWER REDUCTIONS Dgf(D)I Reportable Slu! td nun Systeca Couponc'.nt NSICOI)

Datu Duration Pouer Event Descrlptlou Causo Event (19 71) (Nrs) (2) Ilethod involved 1nvo 1 vc! d Category 1" 13 100 Nanual trip of LISLVs. Reactor Coolant D2.4

1) 12. 3 (CD)

2) 2-3 9.9 100 To r<<pack YZR spray valve PCV-4316. A Reactor Coolant Valves Nl.l.l (Cg) 9-L4 10.0 80 Lou SC Love!l. Rcpalred luak at 20 A Stoa!a acul Valves N3.2 3) rehenter turbluu'intercept valve Puu<<r flange. (NII)

4) $ -28 61.6 0 Repairs!

A, Clcarglng pcaap filter drain l<<ak. huxiLLary Pro- Pipes, N3. 1 cess Systuns f Lttlngs (PC)

5. Repacked BI valvus. Stean and Valves II 3. 2 Pouer (IIII)

S) 6-30 7. 7 100 Ltr 7-9-71 Upon boosting tho LIAIP auction prus- A sure by acanuaLLy placing tice third coudensotu pcaap ln ec!rvlcu:

A. Tic>> hydrccgen cuolur tuwpernturu St esca encl Valves Nl.l.l cc>>I!.rul valvu sup!:retructuru Pouer fnl leal under tlcu Increased cun- (NI)

~ Ic'Icuntc! hencl<<l Icrecceuc'u clclcl thcl v>>lv<< fulled cine!.d. Tho nunsal cond>>nance bypass valvu uc:ut clc)seel duu tu hl fill Icyclrccg<<n Cuulc:r teup<<I'utulc:. Thu euc!r-gc:ncy feed valve open!!d, supply-ing relict lvc!Ly culcl cond>>nant>>

dlruct ly tu thu NILII'uctluu, lnclwclng sc:vc:rc: vlbratfuns ln the tlnll'uctlo>> LInua, trlpplug thc:

NIWI'. Tice U NI:WP cene tr lppc:d

!annually, trlpplnfi tbrblue and then =

rc!actor ln thu no!Ical logic sequence.

CIRNA SEP Table hl13 FORCED OUTACES A!ID POIIER REDUCTIO!IS ~

DRE(D) /,

Dace (1971)

Duration (Ilrs)

Pouer Reportable Ev nt Description Cause'ystem Shucdoun Involved:

Coaponent'nvolved i

NSIC(N),

Event Category

3) 0. Ih-IISIV failed to close upon I~ Reactor Coolant Circuit ll2. 1 (cont'd) signai fron control rooa.  : (CD) c losers/

Solunold operators ui:ri. dirty. I 'ncerrupcers

6) 7-S 7.6 100 Voltage uas lost to control Reactor Control rod D4. 3 rods dropped, caused Ion '

rods.'ontrol (Rg) drive P2R pressuru and reactor trip. auchanisas

h. Voltage regulator fulled on tlie Relays h rod drive generator sec, plus B, A reverse currant relay felled. Circuit c losurs/

inturrupcers

7) 7-11 4 ' 100 Lou SC Ii.vul and lou FII f lou caused Scuna and ILC D2.7 by loss of inscruaeucacion co 0 FII Paver control logic (lnverccr on inscru- (IIII) aent bus II failed) causing PM con-  ; ~:1 ~ . I ~ . 1 ~ I trol valve to fail closed;

8) 7-13 10.2 100 SC h lou level and luu feed f lou. 3 Sceaa and D2.7 Pouer i 'ice . II1 i. ~

~ 1 ~

( UR )

9) 7-14 1.4 100 Spurious indication of RCP breaker h 14C Clrcuic N2. I trip due to faulty cuii, (lh) closers/

incerrupcers

10) 9-9 19. 0 10D

'Mhcu contnlnacnt ventilation uas rc- h Stean nnd Circuit D2. 7 set the fccduater valves alstakcnly Pouer c losers/

closed, resulting in u SC low level/ (IIII) incerrupcers lou feed flou trip and a spurious safcguards actuation" (Seal-Annual Report Pi, pp. 5) ~

KA

CtUUh SEI Tablo hl.3 FORCED OUThCES httO I'OIIEK AFU)UUTIOIIS DUE(D)/

Duration Pouur Auportsblu SI<<utatoun Systusa Cosponunt HS1C(tt)

Datu Duscrtptton Caaasaa (1971) (Itrs) (2) lnvolvcd lnvo1v<<d Evunt CatuAory ll) 9-6 S. 1 )S Spurious closuru of ttSlVs caaasud St casa aaNI Clrcult D2.6 doublu lou SC luvul. Pouur closurs/

(UU) lnt<<rrupturs

12) 5-6 .3 DaartnS lnv<<utlSatlon for batta.ry 1hc CI tcul t tt6. 3 Urowada ruactur cuolont f lou trip ( Ih) c lus<<.rs/

utgaaal Was gtvcal lnturrupturs

13) 11-12 0 Tu rutaack I'ZK spray valves. Rosctor Cuolant Valvus ttl.l. 1 (CU)

Ita) 12-1S 1. S 100 DurlnS tustlnS of A ruactor trip 160 Circ ult tt6 ~ 3 bra:akor, bypass brcak<<r uas not ( I E) closurs/

cluar uf bra:ak<<r tost sultcla aud taata:rruptura and taartaiaau trtp clrcuttry aaa>>

ca)etatutud.

~ ~

~

CIHHA SEP Table h1.4 FORCED OICTACES AND POMER REDUCTIONS I ~

Reportable DBE(D) /

Date Duration Power Shutdown Syst<<n Cosponent NSIC(H)

Description Cause (19 72) (Ilrs) (2) 'Nethod Involve/ I Involved Event'at<<Rory 0 ~ ~ ~ ~ I

1) 1-4 <<24 100 I

I'

2) 2-24 2. 7 100 Bloun fuse in EN control caused tur- h St<<usa aad Neclulnlcai N2.1 blnu Bovernor valvus to start to Poucr function

,close. Operator then tripped tur- I' (IIA) Lul it s b lnu.

3) 3-8 ,8 60 Ilot shutdown due to "probleeu with 2 Stean and Nec Ila nice 1 N2.1 EII syst<<sa.

I Paw<<I function (Hh)

I units'alves

4) 3-20 17.5 92 Nut shutdown to ra.pair packinB on Reactor Coolant

.~

N l. l. I

-'.I -

spray valve PCV-4318.

a PZR (CB)

'J

5) 4-14 .3 86 FF/S'FNII plus lou SC level duu Stllan,and IAC D2.4 to accidental closln8 of NSIVs. Pouer (IIB)

~ ~ I

6) 6-23 6.2 36 Ltr 6-30-72 Roals 0-5 and C-9 dropp<<d into cora.. A 3 Reactnr Cnntrnl rod N5. 1 Ilater had Icaka.d into CIUIH cubi- I (RB) il r iva:

nl ts RcuunalluB oalt their Iiower.

~

nechanislas

7) 7-13 2. 5 83 Faulty spa:a:d channel card ou turbine A 0 2 Stean and Circ ult N2. I cnntrol systeu caused contrnl I'ower'(IIA) clos<<Is/

vnlvus to closu. inta.rrupters

8) 7-14 1.6 83 ff/SFIBI plus lnu SC level caused by h Steals'and Pueps Nl.1.4 tIFHP trip due to loss of AC oil Power puup ~ .. (NII) *~ ~

0 4

~,0 ~ I.. ~ I

~

0

CIAUA SEl'ablo Al.4 FORCED OUTACES htlO l'OUER AEDUCTlallS DRE(O) I Datu Duration Pouur Aepurtahlu Shul J own Systuu Couponunt llslc(N)

Duscrl pt ion lluthod lnvolvud lnvolvud Event (19 72) (Ilrs) (2) Category

9) 7-27 4.S 83 Ltr 8-7-72 llauual trip Juu to loss of loalc Auactot'RR)

Control rod ll2.2 supply pouur fur RCCS. drlvu wuchanlses

10) 7-28 2.6 30 loss of 4160 V supply durlna Eluctrlc l'ouur Gu<<orators aynchroulratlou proccJurue. Syut<<us (EA)

11) '-S 7.0 83 PZA spray volvo luak rupalr. Ruactor Coolant (ca)

Va lvus Ill.l. 1

12) 12-14 17.1 82. S 1'ZR PORV block valvu luak rupalr. Ruactur Coolant Valves ll3. 1 (CR)

ClNNh SEP Table hl.5 roRCED 00TRCES AND POIIER REDuCV10IIS DBE(D) /

Duration Pou<<r Reportable Shutdoun Systen NSIC(N)

Date Cause Cooponent'nvolved Event D<<scription Involved (1973) (lira) (2) Category I

8) 10-21 142 91 73-9 lwss of uflnlte paver. Mbeu onu of II Electric Power ILC D2.2 73-10 the four 115 kv trans!sisnlon lines Systens I.tr 10-31-73 uas deus for construction of a neu , (EA)

Ltr 11-15-79 substation, a second 115 kv line. ~ ~

sagging due to the incr<<ased load (statlun output uas 435 NIle net or 912 poucr), flashed over to an underbullt 34.5 kv lln<<and its cir-cuit opened. Tbe consequent 230 IMe I pouer suing on the remaining tuo l 15 kv u lln<<s caus<<d tIua both tu trip, causing cosplete loss of generotos output ability. h turbine and reactor trip ~ ~

folloued laaedlstely. I~ ~ I

9) 4 20 91-4S Youer reduction. Repair on EII control g Stean and tlechan ice 1 N2.1 0 systela ~ Power function ~ I (IIA) units <<

10) 12-11 55 91 Repair ueld leak on charging puup huxlllary Pro filters Nl.l.3 discharge filter. cess SystemLs (PC)

ClNNS Table hl.5 fORCKO OUThGVb hNU POllgb REDUCT10NS SKI'aportublu obv.(o)/

Sbutdoun Syst<<n Coupon<<nc NSlC(N)

Data Duration Pouar Descript lon Cnuau luvulvad lnvul va4 V.vcuc (197 3) (Ilrs) (X) Category 1-12 83 To repair cundanaur Cuba Iauks alai Scents Aud Neat N3. 2 1) condenser baffle pinto. Pouar uxcltangurs (NU) 3-9 13 S3-45 Pouar rl.duccion. Tubu rl.pair on Stean nnd N<<at N3. 2 2) h and B cond<<us<<ra.

1 Poual'NII) axcltangars 6-9/10 12 S3-45 Puuar reduction. Replaced oucbuard Stana nn4 Puup . N1.1.4 3) bearing slwau un lh NfUP. Pouar (IUI)

4) 7-22 203 95 73-6 ff/SVDII plus lou SG level dua to St<<<<A al'ld I'lpas, N2.1 73-7 disconnected f lou cransultcar on Pou<<r ficclngs b hlTIP control vulva. (NN)

5) 7-31 20 Clrcttlatlng uatar puults tripp<<4 dua Stunt<<aod. Clrculc NT. 1 to faulty trip clrcult. Poller c lose ra (NG)

6) S-19 96 90-47 Pouar reduction. B NfNP b<<urlng Steat<< an4 I'uup Nl. 1.4 repair. Pouer (NII)

7) '9-10 24 90-50 Pouar raducclon. b NAIP repair. Scant<<and Nl. 1.4 I'ou sr (NII)

a CIHHA SEP Tab)a A1.6 FORCED OUTACKS ANII I'OIIER REDUCTIONS DRE(D) /a Duration Poucr Rvportable ~

Shutdown System Componvnt NS IC(N)

Data: ~

(Urs) '(2) Evenc Duse ripe ion Ncthod Invo)va:d Involved Event (1974)

Category I) I-) 2 737 91 glade failure on Ho. 2 LI'urbine. ~

A' I Scaaua ond Turb isa:s

~ I Ni.).'4 Refueling uas accompl)alavd uhl)u Pouc[

')

mulncunanca: avvrhaul of turb)ne (ii )

u.ls Iu progra:ss. ~ I

~

4-27 9 43 Instrument bus lnvurter faila!d Electric Cvnvrators H2.2.

resulting In trip. Poua:r

(KD) ~ ~ ~

3) 5-)8 9 71 Sca:am la:ak on main stcam to )A Sca:am and Pipus, N3. 1 r <<ha a cur. Power fittings (Ng)

4) 5-20 102 70-45 Pouvr ra!dua:cion. Ful lure of Steam and N).1.4

)A-NIWP lmpc) Icr. Pouc r (NII)

5) 6-21 139 71 Rvpalr gasket la.uk on PZR manuay. Ruactor Coolanc Vassals ~ N3.1 (Cg) prcssure I I I

6) 6-29 43 70 R<<pair la:ak ln cllarglng paamp filta.r Ruaccoc Pipes,. ~

N3. 2 vvnt Ilna. (RD) fittings

7) I 7-2 47 70 Repair )vak In claurg)ng pump filtvlr A Reactor I Plpa!BE N).2, vane plpa: socket ueld. (RD) f lctlngs

8) 7-26 l6 70 Inscrumvnc luau Invcrta:r felled Electric Poucr Cenvratora N2. 2 ra.cult lug In trip. (KD) '. I '\, ~ I !i a I ~

9) 8-24! 21 91 Repair fa:<<dueler Iacacvr tuba )oaks. Stoa!a and Ha!ac -" N3. 2 Poua! r uschangers I(UU) 'I ~ II

10) 9-) 8 6 92-75 Poucr rcduct lon. Accumulator boron 8 Kalg)nacred Accumulator N7.0 concentrat lon bvlow tuel!a Ical I 'afety a)lac I f I call lulls ~ I Fa.'ilc lares ii. Ia

,(SF) ~ ~ I ~

CIUHh Table hl.6

}'OILCI'.0 OLFfhGES h!IO I'OLIEIL IIEUUCTIOIIS SEI'atu DUE(D)/

Aupaartablu IISIC(H)

Dwrat lon pouur Shut4oun Syatuax Cowponunc Duscrlpc lon Causa Event (19) 6) (llru) (2) Invn1 vu4 lnvo I vuJ Category I I) 11-2 270 IOO lwspalct sta'aws Sulaa.'latnl'aablwg. Scuaas an4 Iluac 113. I Pnuur uxclwwlgura (IIU)

12) 12-11 6 100-48 I'ouur ra:Juct low. Era:un chuck of Stuaaaa un4 lieut II3. 2 I<<N culwluwsur for Cwbu luaka. puuur Uxclalalallura (UC)

I I) 12-1 S IU IOO-hU I'awla.r rualaact lnw. Pra.un a:lauck uf Stuaaas un4 llual. 113. 2 I-U conJa:wsur fur taabu la:aks. laouur uxclaanSal ra (ILC) lh) 12-21 I aa IOO-a'aS I'aauur ruJucclon. Eruow chuck nt Stallus un4 lluac II'1. 2 1-0 a.wwJuwaua fur tubu Iuukaa. I'uuur uxclallaagura (ILC)

Table A1.7 PURGED OUTACES AND POWER REDUCT10NS DBE(D)/

Reportable Shutdoun System ComponI.nt NSIC(N)

Date Duration Pouer D<<scription Cause (l975) (Ilrs) (2) Hethod Involv<<d Involved EVIrnt CatI.'gory 1-4 <24 100-71 R<<pair<<d steam supply line of 18 Steam and Neat N3.2 1) r<<lreat<<r. Pou<<r ., uxcl>angers

~, I

2) 3-5 9 100 UE 75-01 Rods droppI.d due to uater leakage Reactor j&C '4.3'r into rod control cabinets during (RB) maintenance.

3) 3-10 100 AO 75-07 SC tube leak prior to refueling and A St<<am and lleat D6.3 Pouur I.xchangers I maintenance outage.

• I

~

(IIB) Co

4) 5-19 9 25 Ilotuull level control fail<<d causing A Stcam and I&C N2.1 lou SG level. Pouer (IIC) I

5) 5-21 44 55 E.II. control failure: vlbration- Stuhm and El<<ctr ice 1 N2.1 induc<<d wear caused short circuit. Pouer conductors

~ "(IIA)

6) 5-26 6 55 E.II. control valve position limiter h St<<am and Circuit N2. 1 malfunction-control c'lrcuir card ~ PouI. t'  : closurs 'I. ~

failed duu to short on 5/19/75. ~

(IIA)

7) 5-31 24I 50 h. E.II. unit rI.pair. A 3team and Hec ha n ical N2. 1

~ rpouer function I

~ (IIA) units

8. HOV on main feeduat<<r line Steam and Valves Nl.1.4 rI.pair. Pou<<r (IIII)

8) 6-6 12 HS1V malfunction (Sco similarI h I 3 Steam and Valves D2.4 event on 6"23"75 ' .Pouer I OIB) ~r ~

~ I I ~ 'I ~

' ~

~ r

CIPOIA SFI'able Al.7 FORCED OUTAGES ANO POMER REDUCTIONS DBE(D)/

Duration Rc:purtablv. 'Coaponc:nt Dot u Power 'dlcutdown Systen NSIC(N)

(l975) (Nrs)

Dvscrlptlon 'Caccsu

'Nuthod 'involved 'involved Event (X)

Category

9) 6-17 4 95 Loss of lnstruaent bus invvrter clue A 'Elc:ctric Pouer G<<nerators N2.1 ~

to failure of the pulse drive printed Sy a t'e ceca card circuit. '(ED)

10) 6-17 97 L'I'R 7-17-75 Hanual turbbce trip due to exec:salve Stuaccc accII Transforgcurs N2. 1 vibration of thu FM plplng du<< to Power

~

hunting by the FM control valves.c (NN)

11) 6-23 87 88 UE 75-03 HS1Vs nalfunctloned and u<<nt closvd. Stean and Valves D2.4 Pou<<r I, '

-(UB) i ~ c ~ c

'I

12) 7-20 6 100-14 Power reduction. Repaired leak in St uaccc 'and Transforcccers H3.2 th<<drain line on the hlglc preasuie Power turb lnv. c (NA) L ~

I

13) 7-24 12 100 Llghl.ning strike In aultclcyarcl Fluctrlc I'ower R<<lay N9. 2 caus<<cl turlclnu trip. Syatecccs

'i(EA) ~ I I .>c I-

14) 7-30 14 99-4c8 Powc:r reduction. Freon clcvck St,c!accc and Neat N3.2 lc:aku In lA condc:nser water fnr'ube.

0 Power uxclcangvrs boxes. (Ref. 1) (UN)" -'i 8-3

~ ~ ~ iI c

15) <24 100-56 Power rc.'duction. I.oacl Elc:ctrlc Power Hone H9.0 rvlll>ack to ualntain dispatcher'equestc:d Syateus cranxulsslon lluv capalillltles.i 'EA): c'

16) 10-10 44 100 To rcplucu lcowe( cablus for lake intake Icuaters.<

Stuaca and Pouu r (IIF) c

'l.1.4=

Elc:ctrlcal conductors

~ ~ i,,c I~ i c,i,c I ~

~

17) 10-26 <24 100-70 Power reduction. l.oss of 11 genera- Powvr Auxi I- Blaue.rs ci H1.1.4 tor transfuraer cooling fans. fary Syat<<mcs (AA)

18) 12-21 15 100-46 Puuer reduct lun. Free>cc clcvck for St<<accc and Nvat N3. 2 tube: leaks In 1-D-1 cond<<navr water Power vxchangurs bux. No leaks found. (IIII)

GIUZIA SSP Table A l. 7 FogcfD Du.rAGSS AIDI PDIIfg IIFdIUCTLDIIS Dgf (U) I Uurat ion Power IIapo r tab le Shutdown Systa<x Conponant IISLC(II)

Uate Descrlptlon (19)5) (IIIs) (Z) tlathod Involved Invulvvd fvcnt Category L9) l2-23 15 100-46 Power ra<luctlun. Vrunn check for Stoa<A and Ilvnt ll3. 2 tubu leaks ln 1-0-1 coudawuer water Pnwvr exchang<<rs I nx. One leuk found. (IIII)

20) 12-30 .44 100 RO 75-13 SG tubv leak. l Stean a<el Iieet 06 '

~L Power e xc I Ia I I gu r s CD (llg)

A Liefeel I wg outage started 4 days eac'Lier than scl<aduled d<ia to I<.'nk.

t Ilu looks found. Tha source of sndtu<x uas tracud tu n Ivaklug I<uatlng coll lw tha audtw<x hydruxld<. tank.

Other nalntenance Included< L. Lank Ln g AVII Iaulwtl<w< vulva svul uas svaled uwlng thv Vere'<wite process.

2, IIeptaced scored and luaklug accwewlutwr cylinder uf hydrnwllc swubbur un 0 fevduater llwv.

lteplaced cable uwd c<w<nactur ut the reoccur hvnd tu bank 0 CDUII.

GIHNA SEI'able A1.8 fORCED Olfl'ACES AND I'ONER RGNCTIONS Date Duration PowLyr Itcportnbl>> ~ ~ Sloltaluwn Sys Cuan Conpoaiuot DBE(D) /

Event. D<<scrlptlon C HSIC (N)

(l976) (Ila's) (2) IIL' I le 8 Involved lvvd 'nvo Event'ay a yy HA 307 HA, Continuatlon frona 12/30/75 outage Stance and Ilvat 'ontinued) duv Co 8 stean generator tube Power exchangvrs la.skag<<. (Ho) c I) 1-29 66 100 Ho. 2 low pfc:ssura: tufblllu blnde Stance and Turhinqs .~, H1.1.4 failure. Power (HA)

I I~

2) 4-7 <24 Reactor trip due co lanlntvnnnce error 3 1&C I&C N5. I on uvurpowcr delta"T Iaastralaentntloo. (ID),

3) 4-12 <24 40-27 I'ower reductloo. Turb!no raaaabnck Electric Power Clrcult N2.2 due to fnllura. of I'nw<<r lhinge N-44 (Kg) c Ialsl' s/

High Voltage'owef Sllpply, Intc.rrupta:rs

4) 4-14 7D-5D I'uwa.r redaactfun. Vlbrntiun ln HFHI'.

5 I Stones and Pena!a HI. I.ra Puwur (HII)

5) 4-16 5 50 76-13 Plant plnced ln hut shutdown to re- h ~ I c Renctor Contr'ot,rod trieve two dropped ccnltrol rods> '(ICD) dr iva.

C-5 nod C-9. Voua'rlotvd circuit Qc:chanlsns cnrds r<<plncLcl ~ buC oo alc fa:ct ~ '2.1 faaaand.

6) 4-18 5 50 E.N. Ruvcrour prc>bleaas dal<< I.o .wntvr Sc Lllln anal CI fcult y lo oil fruaa ull cuola:r leak. I llwL f c luau rs/

(IIA)., Incur'fllplurs

7) 4-18 . 40 E.N. Rov<<roar probleaas duu to waI.er h, Stu:ceca and,, C I reu it. H2.,1 ln nil froaa oil cooler leak. Pew ur. closets/

(HA) Iaaturrupturs

8) 4-24 308 99 76-35 8 sconce gvnerator tube lunknge. St Lani olid Ilvat D6.3 IuwLyf cxclaangurs (Hn)

~ GINtth St>I'ublo h1.8 FOIICFD OtffhGFS htttt I'ONER IIEDUCTII)IIS ORE(D)l Dutu Pouur Rupur cubi u Sbutdoun Systua Cusponcnc IISIC(N)

Dura cion Doser lpclon Ca>>su Evpot tta:cbnd lovlilvud l>>vo I va:al Evunt (197f.) (Itrs) (2)

Cacogory

9) 5"7 IU 45 ttuul uc fulluru I>> I-A uotataaru Stuuau ua>d I'l pun, . N3. 2 uuparutor rubuutur stuuau 1 lou.. I'ouur f It Inns C

(ltn)

10) 5-12 159 90-50 Pouur ra:duction. Ih IIFIIP Input)sr Scuaua und Pauap Nl. 1.4 fa I luru. I PoliL r (IIII)

11) 5-22 100 Fluid tuuk Io Bl syuccn. Sta:uu und ttocbuu Ical ttl. 1.4 Poua.r function 0th) >>>>les

12) 6-3 8 100 I-B luvurCur falluru cuusa.d lou SG EI Lect'lc Youur Ca:ourators III.'1.4 luvul. (ED)

13) 7-4 7 100 76-18 Oruppud 2 cvaacrot rods (G-5 uoal G~9, Reactor C>otrol rod D4.3 analn) . ([tn) air I vl.

nuc bun ines

14) 7"6 <24 100-81 I'ouur ruductlun. Conda.os>>to pauap Stuava and Puli p Ill.1.4 Islilotud for ra:pairs. Puuur (NII)

15) 7-29 <24 100-50 I'a>uur ruduccluo. F2 luu prussuru Staviu und Turb l>>u Nl. 1.4i tul'b lou ILld uNcusstvu ba'ill lnn Yuuur v it>rut Ilia Itlitnr Ia>uc uaauu uauclll. (NA)

16) 8-4 50 100 76-21 Dfaatltaud 2 clast ful lulls ~ 0 3 aod Kal>>Ctnr Cuutrul rnd 04.3 0-11. (Rn) drlvu uauL'b Iolauas

17) 8-7 668 100 nludu fulluru un f2 lou pra:>>aural Sta:;ill und Tulbtous tt l. 1.4 turbtnu. Puuur (I!h)

GINNA SEP Tab I c Al. 8 FORCED OIITACES AND I'OIIER REDUCTIONS Reportable DBE(D)/

Date Duration Pouer Shutdoun Systca Couponent NSIC(N)

(Nrs)

Dcscrlptlon llutlx!d Involv<<d Involved

~

(19 76) (2) Event

'Category

18) 9- IO 168 96-50 I'neer red!!ctlun. D IINP lapel 1ur Stean and Puup Nl. l. 4 failure. l'nuu r (NN)

19) 9-25 '<24 95-50 Tube leak uas found and plugg<<d ln A Stean and ll<<a t N3.2 182 cond<<user. Caus!.'as stean Pol!or i.xchang<<rs > !. I

. Iuplngcuent fron Lhe stean duup lines 'NN) located balue the nffect<<d tubua. I

20) 10-8 511 100 76-24 leak ln stnlnluss steel pipe betueun 8 Elll',lnoered J'I pea, ~, N3. 2 boric acid tanks and Sl puups..Five Safety flttlngs '7. I, sections uf pip<<uere rl.placed. FI.'aturua

'(SF)

21) 12-11 <24 100-45 Pouer reduction. 18-1 cunduaser 8. 5 Stean and Neat N3. 2 checked for leak. Pouur excllangers (INI)

22) 12-12 <24 100-45 I'ouer r<<duction. lass of condunser Sti an anal Ya lvu .Nl.l.4 vacu!ex duu to u)uctor stean supply Pou<<r valve "prub le!a". (INI)

23) 12"14 <24 100-46 Pouer reduction. 18-2 condeus<<r Stuan and N<<at N3. 2 checked for leaks. Pouur uxchangers (NII)

24) 12-17 100-46 Puuer reduction. 18-2 condensur Stean a!xi lieut N3. 2 checked fur leaks, Pouer uxchangers (INI)

25) 12-17 12 46 76-28 Dropped control rod 7-12. The Reactor Control rod D4 3 stationary hold cull uas fulxld open (Rg) drive nt Lhu reactor h<<ad aas<<ably pluf. uech!<<slews a

Lh!tll the next shutdoun.

t.Furced reduction ran over Into th<< folloulng lncldcnt.

GIIDIA SKI'ublo hl.9 POSCG) OUThCSS h)ID I'OUSS IIKDUCTIQNS

>>bp.(O) /

Diitu Duration Puua:r bubo) table SI!at doun System Coilpona:AC NS1C(N)

(Ilrs) (2) Event Descript la!I Cuusu lavolvud Involved Eve! t (1977)

Cstebory I) 1-11 14 100-46 Pou<<r ra:iluct Ion. 1-8-2 condensur Steno und Ilu at II3. 2 I.ubu leak. Power uxchanburs (IUI)

2) 1-25 14 100-50 Pa!our reluct lon. I-b-1 coudunser Stuaxl And Neat N3.2 tube leak. Pc>our usa:>>auburn (IUI)

3) 3-21 12 100-45 I'<wur rvdoctlun. I-b-l coi(denser Stoa(xi A!Id Ilu;IC N3.2 tuliu la:Ak. P<>uer uxchunburs (DN)

4) 5-23 15 35 77-04 1nopvrabla: contrul rods, K-7 and Svactof Control rud >>4.3 L-8 duu to ln!ctullatlun urrur. (SD) alt lvu uvchanfsus

5) 7-5 188 100 77-08 b stean Svnurator tube leak. Sta!c(! ~ And Ilvat >>6. 3 I'uuur exch!la!Dcf s (Ub)

6) 8- 2 SI 100 Tui! snubburs lnopuriltlva: on stean Stl'iug And Shuck sup Nial 4 bvnuratur h aluc! Cu slaaft oval I'uuo r prussors nnd luakubu. (IID) sIIIIpurt s

7) . I I- 2 48 100 C)n>u unubbvr liloP<<rat lvu on steul Sta!A(s and Sho(.'k sup- NI. I.4 bunvrutor 8 aluu tn shaft sual Pouuf pri.ssua's saul Ii:a kiu ~ (llb) supports

8) 11-17 30 100 77-23 Iu(!puriible control aods duu Co ran- l)uuctor Control rud dom fallura! uf tuu p c cilrds. (SD) air 1va'.

l>l('I'h iolaus

9) 12-3 lb 100 I.i!<iklnb pipe tap - rvsldoal heat Suactcir Coolant I lpu!I~ 03. I ra.(<oval systea. (CF) f It t labs

i CIHIIA SEP Table A1.10 FORCED OUTACES htU) I'OVER REDUCT10tlS Date Duration Pouer Reportable Shutdoun System Cumponc.nt DBE(D) /

tto. Dvscrlpt ion Cnusu HSIC(H)

(1978) (Hrs) (2) 'vc:lit Nethod 1nvolved Involve J Event "

CateRory I) 1-1 I <24 100-48 Pui!er rciductlon. Cond!.nsur tubu Steam and IILat lt 3. 2 leaks uere rupalrcd. Pone Qxc bande ra c'IIII)

2) 1-14 <24 100-48 I'ouer reduction. ConJenser tubu Steam and liest tl3.2 lvaks uvre repaired. Poi!er vxchan8urs (HII)

3) 1-25 212 100 78-03 8 steam Bvnurator tube Ivnk. (Also, Stc.am and Hc.at D6.3 CRDHs feutl'I.cl CQplllred tile 8 Iuop Pouer QxclianRQrs sc!mptu valve Raukc:t l<<akl SC snubber (HD)

A-7 rc.built; rc.ptacvd ttie PZR prea-suru enid lvvel control triic!uiuttturs ulth It tvfmlwal blacks.)

4) 2-20 19 100 I'icktnp lciak on PZR spray vcclvcv Reactor Coolant Valves tt 1. 1. 1 431-8. (Also, I'ZR block valvci {CD) rc! packed . )

I

5) 7-15 <24 100-46 I'uuur rc:duction. Turb lnu runback. ScLam and Turblnvs Hl. l. 4 Pouer (UA)

6) 8-26 18 100 Burned uut coll ln rc.actor trip I A 16C 16C lt2. I lup I c. (IA)

7) 9-IS <24 100-48 Pili'L f c'c'dilct IIIII~ Cuililciiisiil' cibLi Stc!am and It<<at H3. 2 Ival'c pair. P<1QI' vxchanpurs

{IDI)

8) 12-7 32 IOO IAIM ~ II I IevLil ln cQactclf ecccllant RL ac t or C~iu I ant ILC H7.0 pclmp mill elf fc.'IIIII f\'cl mcclccic I I cccb I!1Q (CD) trlP. Rc.irtuf IrlPPI'\I !ill III lci SG Iilvc'Ii kilbL cll I ilcbIL'J tcl I!Cl mlituf ~

GIIUIA SEI'ab I o h1. 11 LONCED OUTACES A!ID I'Oiigg NEUUCTIOiiS DNE(D) I gcprrtai>>lu Shutaloun Syscca CowpuAI7>>t HS1C(II) lb>>to Dul ation I'nualr Dvscrlptlon Causu (Ilrs) (2) Ev<<>>- 1nvolvsd Invo I vud Event (1929)

Catogory I'ouur roil>>et lon. I.oss of lndlca- guactor Coolant ILC ll2. I I) 1-10 <24 C lu>> of >>la>>ill f Iuu on thu IA NCP (CO)

J Pc>>ua'r raI'al>>CCLIIA Dciriug iul>>ual Stuaw a>>d Valvu II I. 1.4

2) 4-9 512 uvurhaul uf A condo>>a>>tu p>>iap, Puuur 0 co>>ala'ascii U passp d I Jch>>rgu v ~ Ivcl (UU)

Qp>>l IUU:I1y ruchlcucl CU>>du>>,'Iatu

. pruss>>ro ut Ilklil'uet lon, cauvvd, rvducvd Ill f lou.

II-1L 5.5 I'oua.r rucl>>ctla>>l. Noisy fsllual ln Kla:ctric Pouur Rulay Iig. I 3) thu wal>> transfurwvr cooling fa>>. (KN)

4) 2-6 691 100 79-13 LINC ruailllri7>>l Shall>>lou>> tal I>>l patllt PII 0 Stoclw aud P I pa.s ~

AUtslu Vi;lalS AC SiiS ~ I'oua: r f II'C lugs (IUI)

5) 0-5 .5 20 lt>>ss uf cundvnsc7r vscucss alurlng Coot 0 Stools snd liuis t ill.2.4 ca>>sud ra.actnr trip. Pouvr uxclls>>gers (UO)

6) 8-31 <24 100 29-17 Pnuur r<<duct la>>l. To bring boric Rvactur hccucsulators ii).0 auld ca>>k cu>>co>>trot lo>> back tc>> (00)

Ml thin Spalolf loiltll>>>>$~

1) 10-2) .3 100-1 I'iauvr rvduct lac>>, 0 $0 lu>>>>ilullu gas- h St>>law r>>ld Iia:a C kvi lvak. Nvac:ia>>r all>>illlnval crit I- I'c>>uur U xc 0;I A gui rs r>>l;Ii IZ piia:a r in acid rvp:Ilr l>>y (Ii0) w:lint>>lining vlcc>>>>caw in rvacinihlrr

~ Iclda>>.

I I- IO '4 IOU-20 Pciuur ra."luct l>>n ~ 0 SC h>>>>alba>>lcl Pss- A St<<ac l>>ld lla"I i II ). I

0) <

ka t la..'lk. I'aaua r I'.ll'h>>'ll;I'rs (IIN)

< ii ~ ~ I, I' I ~

'I I I ~ ~ I I ~

I ~

I ~

CINNA Al.ll SEl'able PURGED OUTAUES ANi3 PO)lER REDUCTIONS DBE(D)/

Date Duration . Pover Reportable Shutdoun Systesl Cosponent NSIC(H)

Ho.

(1929) (Nrs) Event Descript lua Hethod Involved Involved (X) Event Category

9)11-24I <24 100-20 l'over reduction. 0 SG handliolu gas- A St coal al'Id Neat N3. 1 ket leak. Poucr excliangers (HB)

10) 12-2 416 100 79-22 Repairs on:

A. Tube luak ln B SG. Stuaxl and Ruat N3. 1 Pouur uxchangurs (NB)

B. Nortlu on PZR relief line. Ruactor Coolant Heat Nl. 1.4 (CB) uxchangurs C. Correct lou boron concentra- Reactor Accunulators N).0 tion ln boric acid tanks. (RB)

e A-28 I

Appendix A: Ginna Paxt 2. Repoxtable Event Coding Sheets

Ianlu Ac ~ I \ value u<lu<'\, <<<l <<vpv ~ v ~ < ~ ><a ~ < o< ~ ~ ~~

IIumber NSIC hccesslon Event 'cport plant System Equlpmcnt '. nstrument Component hbnorsnl

,l lion Cause C

SlSnlflganpe Date Date Status Stntus Conn Number 69-Ol 40S09 12/3 12/12 5 N J,S ",I 0 D<F D C7 l. ro<l falls to drop on scram (reactor sln<tdoMn).

69-02 40510 12/16 12/23* 0 E,N,CC 00 C X . D Valve falls sIn<t between sump and RIIR pump on test.

(scu 74-06 and NS ll(3) I 241) ~

~ 1

" ' ~

4~

~ g < ~

~ 1~

~

<Ii '< ~ . ~

I ~ ~.

From Semiannual Report No. l.

~<

I 0 * ~ ~ I I ~

I .~ ~

I NS IC Event Report I'ln<<L . Casaba>ncu C Al>laurQal I S I bn IfI cauce t Nasskalr hccoss lou Nassl>er Date 'ate StntUS Sy4Lesl KIIUlpwcnt lnsLI'asselaL Statuo Condlc luu CnUQQ Cntebory Cosssea!

70-01 8

8 S3,C6 "8" Lral<<SI dial aot 1/3 aa:Luate Qu La'ut lllgnal (see Nk I I (4) 328) ~

llc-G relay contactu fall Qpu<<.

70-02 1/8 2',V N Rl 8 Le>speflatllru exceeds Tecla. Spa:c. cold QI<<IL-douu l llslt 70-03 1/30 8 'y bb 00 Servlcc Uater valven to 8 AFP Ilspfol>air ly I<<spa.a:

Lcdiu A/ left closed.

70-04 3/G 0 N FSreuulVa: leakabe tbru coutnlu>sent personnel nlr Iolk.

70-05 46960 5/18 C . Ah FF,DO N IISIVS fall upcu ~ Iurlnb test (SQQ 70-G).

70-06 47292 5/14 5/22 8 AA FF,OO C7 NSIV f.llIS open alalrlnli Lese (reactor QI<<atda>ula; Qco 70-5).

70-07 49878 8/19 8 Ovcfdusc.

70-08 57237 10/14 0 I. 00 N Coro ale luba'alve ful lo Qpun >Iaaf Iaaf 1>llaatal4141 ~

70-09 56044 10/14 D 1, P 00 N Co<<tallssent lnulatlo<<

valve f 8 I 18 open during 8 lau t do>>aa.

70-10 56975 10/23 N 1 Qf 2 Qalaap valves fall>I UPQU alall'lng SLal'CUp 70-11 61003 12/10 S O,DIL N Naia I.'a>>>lant pus>p st uals cracked.

8 Fron Senlaunual Report No. l.

~ ~

Table A2.3 Coding Sile cpot table Events for Clone 1971 Hunbcr HS lc Accession Event Rcport 'lnnt Systcsi Eilulpacnt lnstruaent i Coeponcnt Abner' C, "Signif leance Consent Nuabcr Date Dato Status 71-01 61321 2/24 0 N,X ~

C7 DC fails to start

.during t ra lnlng.

71" 02 63222 4/27 4/29

' Sl, 2, All 4 BAT-to-Sl 3s 4 66927 C4 valves fall gloscd during Lest.

71-03 64820 6/30 7/9 00 B X CS Startup of a third coinlensatc piuip lcd to hi failures ln series and a=shut-dolJn, 71-0II 64850 7/21 ~

C3 Overdose.

71-05 67990 10/6 N C3 Iodine uustc gas.

I iIonltor falls to idctcct release.

71-06 68302 12/14 8 N N . Sl piuiip falls to start on test.

Faulty relay.

I I

~ I - I

,I ~

I I II ~

~

~ ~ '

~~

Also ideatlficd as a precursor to potential severe core damuige in a HSIC study "Accident ScIluence Prccursors" for Hatt Taylor of NRC/PAS.

I I I I ~ ~ ~

II I ~ ~ I I1

~ ~ jj ~ ~ jj~ ~ ~ u j ~ ~ ~ ~ ~ jj ~ ~ ~

NS IC I'.vewc I4!port I'lnnt ~ Ca>aapunuwt hba>ors >I Cia>lsd C

Slgnlflcance Cua>j>j>u w c .

Ralulplxuwt II'Intra>>a>eaat Nwsba:r hceesalun Nuaab sr lb>I u 'ste Status Sjl>ste>s C dl I 72-01 69330 I/27 D N C Q N rrlp suit>.b fnILs co nscuntu uw tost.

72-02 f>5012 .2/7 0 N 0 N Sa:t point air Ift.

'2-03 58877 2/7 C Q 0 Owu of tuu under fr>'!>Iua:a>c7 pfotea tlun sbniuiajls folio un ta:st.

72-04 697 L2 3/20 3/30 0 00 N NSIV flails i>pun (h/ du>lug test).

72-05 70029 4!28 D C3 Kxuuss ra:leone tbru uwx. bldg. fluur drn lns.

72-06 71407 5/15 D CC O,T h 808 pals>p wlc'jalawlal alurlng culd sliwtaloun.

72-07 ~ 72421 6/30 8 J.N,X D Full<<d fwa! I fwuwal during exiwslnntluw.

72-08 72422 6/23 6/30 8 N,Q C C7 Ta>u rods alrop aluu tu usta!r leuk ln suue unblnut (runccur nb>at ab>aja>)

72-09 73245 7/27 8/7 J >hh N Va>linga. a'plku ln LSI Dgj I>a>a>uj r salpp) 2 ra:naia!real rawls wut.

air lvnbla: (reoccur

>abutalua>n) ~

72- IO 75908 LO/30 C N f l.ua>vu alai I I> 1alg fuwwal alui lwg>

I cf ale I lng 72- 1 I 77411 12/28 DD C3 gula!nsu during resin bewail lng.

N<w<bcr 73-01 73-01A 73-02 NSIC Accession 8<saber 81270 81477 81464 Event Bate

~,

Rcport

'Iato 5/31 6/8 6/18 B

8 Table h2.5 plant Status S

Systew V

Ah

~

Coding Sbctt

~

portable Events for Glnna 1973 Equlpwcnt'nsLrlslcnt

- C,L E,T

'Cuwpuncnt S

8 I

hhnorlL11,C ~

K Sign lflcance C7 N

N N

~ ~-

Cowwun t I

IAIosc col<I<.'f Juldt ln liucl car. paver nveraglng unit uhllc <<t pou<.r-rods sLartcd ulthdrnving Arwaturc found felled du r I ng t car. of solcuo ld

<.oil undcrvnlLabc dcvlcc.

Incorrect trip point found during rccbcck.

Sl puwp falls to start 73-03 81505 6/11 B X on tost ~

. I 73-04 81505 6/12 B X C7 Sl puwp falls to sLart on L'cst ~

73-05 82680 7/27 n N,X I,R,T Lou I'Zll level test signal fails to actuate Sl.

73-06 83262 7/22 8/21 8 Z G,OO C4, CB kit control valve falls.

Consequent uatcr ha<<<<<<cr dawagcs supports (sec NS 15(l)88).

73-07 83162 7/22 8/1 E,H C4 liat er boa<war dlsconnccts Il flow orifice nnd 0/P--

gage frow llI piping (reactur shutdoun; scc 73-06).

73-08 83834 9/14 B E,H N SG flou channel falls dounscalc.

IISIC Slbullflcanca tvcnt Rapul t I loot Cuiapuilcot hbour!swl Causa Cucoauol Nwaber hcccss lon lluab er Ilute 0<<tu Status Systca'a Hqolpiauut loslruulCOL 0 lt I C

<<tugory 73-09 85369 10/21 10/23 8 S3,SI IXH)I'l. IIIII pnucr 851)0 laa logs uia hVM ullla:II 87031 iivarcuuls prl¹ary syste¹ .

73-10 8)031 10/21 10/23 Z G,hh S5 llhY cut uut tuo suuo C6 (uliu!I Sl <<fta!r IIOI') duo tu lus:I uf ~ioucr Iu Insta!i

¹Cut luis lh (aui!CIur idiut douo) ~

73-11 87030 12/21 8 Cii, C8 2 of 3 hlllI'ir-Luuiul lo test due to ru¹¹oo hcudar fru¹ cuodeos<<l.a supply.

A hlsu ldentlf lad aa e precursor tu putcntl<<l severe euro da¹<<ge ln o IISIC study uhccldcnt Sequence precursorou for II<<tt Taylor of I(RC/I'hS.

Table A2.6 Coding Sheet'lrtablc Events for Cln<<a l974 Nuwber NSLC Access ion Event Sopor t 'lant Equlpaent Instrugcnt

~ Culpa<<cnt Ahnuranl Cause C

Sign if lcandc'ystea CdaIxbn t N!Xxbcr Date Date Stntus S C di I 74-0L 88090 I/18 I/28 D N Sll !Inter dral<<cd back L!itn rcactbr Ju'ring sl!utdovn 'for aal<<tu-

<<uncc (see NS 15(3) 337).

74-02 88534I 2/19 0 8 I I.HH N SC tube wastagIz.

74-03 89162 2/22 D I 00 F,N N Excessive leak'lgc Ln lest o( contaL<<ment purge a<<d exhaust doupers.!

Notlflca- 89263 dan. 3/25 D C3 lb!lntcnddcc uorkcrs tlon No. 90662 received doses Ln 107 pxcccs of 3 rcalqtr.

74-04 89868 4/6 4I/16 D F,DD C7 F;ill<<rn of SI puap 91662 to start'aa<<ually f rool one bus Ln test ~

74-05 90623 4/18 4/29 NIT N Leak Ln drain linc on thc CVCS Letdovn I

systcIa ~

74-06 93704 4/22 4/23 8 CC A CS Valve Ln RNR systcn froa suxIp fall>>

closed (sce'9-02) 74-07 93705 4/26 4/26 8 Vndcrfrcqucncy relay fa l 1s.

74-08 94198 5/11 5/21 8 DD C3 Radioactivity Leaks

~ l Into auxiliary bulldlng frIxx x!Lxcd bcd Jcx!Lncralizcr.

74-09 94594 5/30 6/10 8 N lh!dcrfrcquency relay fuils'n test.

74-10 94522 6/13 6/21 8 00 A,E C3 llaJloactlvc role!!sc uffsltc due tu faulty alar!x Ln chcnlcal drain tank.

Tab (('a>>IL I nin<d)

IIS I 0 hccess lou Evc<<t . Rcpor t I'laiit Systeae Fqiilpaicnt Instr>a<<a>at Cu><>poa>ant hbnuranL Causa C

818>> If lcanca Cn>><<<a<< I I L Ihe<>bar II<<a>bar Date Data Slnlus Stnlus 8 Cun d, C It 1Ion iice8ory 74-11 94784 6/26 7/8 8 hh UU >OO C X 0 CB 5 uf 8 s<>le>>old viilvi.'s ulilcli cui>trul ilia 2 IISIVs f>>ll durlnB ti>sl due tn uverhcnt fng - ci>uld nut clurc LILVs.

74-12 94753 6/29 7/18 8 8 < II CC -I.uak In i.VLS pfpc a>alai.

N-13 7/3 7/18 8 D II,CC 8 I.uak l<<ueld <<n ch>>CBI>>8 pu><>p flllcr bypass lnc.

74-14 95033 8/7 8/15 B C7 SI pu>>p lal ls tu nlurt un unu biis ln taint.

haPl',

N-15 95594 8/15 8/23 8 C F S<C bluvd<<M>i Is<<lail lu>1 valve falls open In tCSL aluu tu furelrn ub)cct I<<v.>lva seal.

N-16 95922 9/18 9/30 8 II I'rI><cary cuuluut leaks Into accauaiilaturs Lhru a:heck v>>lvu.

N 17 96363 9/25 10/4 B O,U Va>l<<u<<pcr I<<i'I In ~

>><'.ca><<><<la>in> u less tli<iii Ii> Tecll Sp<<cs

~ ~

N-18 97717 ll/19 D l ui>l '> Ia>ua'a>t purl<a 97807 uxliausl valve le<<keg<<

Iilrh d<<rln8 lust.

N-19 98164 L2/13 8 F Undcrv<>llngc ra<L>>y fnlls d<<u Lu scale (:rud)

< n r>as t IIr c< ~

0 Tnbl .(Con t Inucll)

NSIC Conponent hl>normal ISnif lcance Event Rcport Plant Systea Instruucnt Cause S

'Cowaent Nuaber Accession Date . Date Status Equipment Status Condition CatcBory Nuabcr N-20 94523 12/18 B T DD l C T 'h N AAI'tcbt deflclcnt 98719 procedure.

N-21 12/11 12/11 B X I ~ B D N BAT level transnlttcr falls lou u

11 ~ . I tl f ~

gable fi2.'j I'.udt>>R iiliceL I>>irtalitli Yvhnls iny et>>no" l97S ttuab or tiS IC Access lon Niisdi e r Kvoug Once Rcliur t pape I'laul Sl>>cus Sy st en S<lu I tiueac Instruucuc, . Cisstuuiont hbnoriswl dl C u Slonlf le!!lieu Ccsvaun l 75-01 93279 I/17 on C P,T N l.ubo ull supply royu-AOR I at <>r dr I f t s. o ll lig essilrc llflillsi iiiiil turbine ilrlvln

<Iui'lng fest.

Akltl'rips AOR 75-02 100041 2/14 R C7 Onc Sl puup fnllu L'o rlart oiauiwilly In f I rst le it chic stirl>>S In breaker culllact i!au iicilk ~

AOR 75-03 100041 2/14 C7 l.iiuso circ cacaos Sl IIuIDI! Lil fiill tu scarc un tciil AOR 75-04 10004iO 2/21 N Scil'Iiillil Vi!IVcii filLI lii nporalu a ttSIV ln li:sc (sec 70-Si4) ~

AOR 75-05 100276 2/26 0 C7 Rreiiki!r Inlls lu puucr iiilllply tu unu Sl Iiuup i!i li!i'l ~ .

AOR 75-06 100924 3/25 0 00 RCS deburaled durluy

!!I III ld iluli, AOR 75-07 LOL 700 3/10 4/7 S 11,1RI C4!, 5, 7 S(: la!bc call thlniil>>0 ilII a Lo pliuspliali! currus lou cia fulsld ilui li!R liisllci cion (reoccur sliul-iliiun) .

AOR 75-0S 101732 4/L7 C cliarcual Cllcors fur

.102290 pusl ace ldiiuc Iodine shou aging.

{Continued)

NS IC Number Access lou Event. Rcport . I'lant

' System gqulpmcnt 1nstrumcnt 'Component Abnormal,C'Cause 'Slgnl f l cence Comment Number 0Date Date 0 Status Status Condltlon Category AOR 75-09 102555 4/22 ~

8 F 'N ~ ' Boratcd uater 'leaks but caps on CRD rod travel housings.

AOR 75-10 103687 . 6/25 8 00 CD Doth discharge valves 104052 for Sl pump C found closed ln test (pump C ls 1 of 3 50Z capacity pumps).

AOR 75-11 103686 6/30 8 . S N RI q camp l lug schedule Incorrect.

AOR 75-12 104954 8/5 D 0 P F D f'ndcrvultagc relay I I falls test.

AOR 75-13 105297 8/22 8 S N Operator mlstakcnly nil gncd 2 valves tu dump rcfu'cling uatur In to sump during I

I testing '(see NSlC-144,

p. 52).

I ~

UER 75-01 100898 3/5 3/17 8 G C7 Tuo control rods drop duc to a utter leak Into thu'Ir common cabi-'et (reactor shutdoun; I., scc 72-08).

'C UER 75-02 102795 5/8 Held In RCS letdoun linc leaks.

105548 5/19 ,'03 6 uurkcrs used Inade-109653 quate masks. 810K fine proposed.

T< I'il:< Rt lau cd)

Ilua>>bur UER 75-03 NS IC hcccsslon Lb<<aber 104209 Event Bate 6/23 Ilupurt Ualc 7/15 I'lant Status B

Syscc>x AA E>palp>>>c>ac 0>I lust rueunt ,,

Cuaapununt hbnursa>I J

0 Causa SIR<<if Ic<<ucu S2 Cuu>ac<<

B>>lh II.'>IV<< clo>Ic t

IR>ug'I>>>>sly dna lo ail>>ala hxplnBI<<R a:JBO uf 0>>'u<<JI>>R v;Ilvcu'isks (rci>ctur

>I ha t dna>n)

VEIL 75-04 105549 8/21 B U,T Zi>nur J IuJc falls In Lust. Tie> ovug pua>cr dull>a T trip blslablc 6<!l p>>I>>l Ilr tud lf n>>nn>>nscrvnl Ivuly.

UER 75-05 106344 9/15 8 0 YIII ~ C7 UU suppllaial bus lf>

brcak<> ~ f<<II<< upun

>>pc>v>>1 ladle<<Cur <<nd coul>'>>I slB<<a I t>'an> ~

fcrgaiil tu ncu bu>a ullh n>> Iil'pari>i lun uf ra'ala>>>-

al<<nl. lan>cl lun pguv lalcd.

UER 75-07 106619 9/29 8 h,l, K LIT N Alar>x sa!Cp<>lnt hlCh I>>r dcvlat lun 1.>n ba.luna.n I:au>lrol aud I<<die<<Cur <<nd bank I'>>III>L c r ~

UQL 7S-08 106620 10/1 B V Un< rud nul x>uvlnb ulth liank ln L<.st ukR 75-09 10776S 1 1/4 8 S, YY ludlnc i>cL'lvlly ln YES aixc caid;I 1 h>a It dun L u Lh<>riaiil-aa>d prcssuru-Induccd sCresscd duri>>8

>itarCUpi

0 Tabl (I!onl.laced)

NS IC Coeponcnt Abnormal S18nIElca>>ge Niaabcr Acccsslon Event .Rcport plant System KSulpnent Iqstruucng Cause Coauac at Date . Date Status St,atua, S C dl 1 Condition Catcgoly thsab e r UKR 75-10 108519 . 12/2 8 0 N DO uattwctcr Indicates nu response Lo Duvcr-nor control sultch.

RO 75-11 108802 12/15 ~ D II,CC N. Ncld leaks ln charging punp dlscharDe drain linc ~

RO 75-12 109455 1/8/76 8 N,CC N IIcld leak>> In charDlng pimp dlsclrarpe drnln 1 lno (scc 75 11) ~

RO 75-13 110324 12/30 1/26/76 8 11, NN N D CS,C7 Leaking SG tubas actl-vatc bloudoun lsolntlon ulth high lodluc count (reactor situ tdoun) .

t Table A2.8 Cudlng Slc<<ut rlsblu Kv<u<LS Ior Clans '9/6 DSIC Nuuber hccess Ion Kvu<lt Rupof t L lilclt Systole Kqulp<aent lnsLrceaent

'Cacecpunent hb nor en) Cause S I gulf 1cance Co<ca<aunt Date Date Status Status Conc)It loll Category

)bulb or 76-01 110356 1/11 2/6 Ll J 00 C CiN C ~ N K:<cons Ivu I <<aklq)u

< hl <I LunLalusILnt pclrgu SU)l)cly and exhaust valves during Iaull I up ~

76-02 L 10355 I /12 2/6 N Cunpunvut I!<all lug Iclcecp f:Ills clurlng cold shu<duuu ~

76-03 1 )0354 I/15 2/6 CC N,T N Dcl)LS <<n fluugo uw I)l)i) uuc!I lun plplng fr<as bc<cap selt t Iplbt<<uc:d ln ccc'<Inta:nance 76-04 I/29 S,hh D ~

Q N Duu uf four puuer.

I ange channels tll ) Is during laud decrease for slu<tduuu (seu 74" 14) .

76-05 2/8 N Ksca.salvo leakage Lhru Cleat I)new!uL )<urge I'ccp ply anal exhaust valves I>> La:SL.

76-06 111I<19 2/6 2/19 C N Duc)urvu)L'Igo rulay falls ln Lacer.

76-07 111785 2/7 3/I 0 Scvcl) ~ a)lac)cargo valve 111975 falls upon In t<<SL dual Lu rulay latch failure.

" tubes 76-08 112137 2/27 3/LL C S LI,NI C7 SC plugged'C 76-09 112165 2/27 3/22 C 0 F,N )croaker tu bus )8 trips 6 seconds Luu early 76-10 112164 3/22 6/21 D FF,OO N N Dun<a)anent purge 115062 sclpply valva.'s in) )

o)ceo

Tsbl (Continued)

I NSIC trident Abnorsal SIEnlflcance Event Rcport, Plant Equliwcnt ills Couponent Cause C Coiiiuenr.

Nixiibcr Accession Date Date Status System St C dlt lan Cat or Nuaber 76-11 112138 3/S 3/19 C EE qq C - U CCII CB Control ruoiii ventlla-tlun dniiipcr jircc ludes uso of onc of I

filters.four'harcoal 76-12 113287 4/23 5/25 B I FF,KK N Exccsslve leakaBc 114205 ~,

tllrll Iiorsollliel hatch air luck.

76-13 113785 Ii/16 5/11 8 BB D C7 Tuo rod s d rop s pur-i Ioualy durlnB poucr escalat Ion (reactor sliutdoun) .

'I I

76-14 113784 4/12 5/10 C N Ouu channel uf four pouir-'ranRc hlRh voltapc 'puucr supplies falls at 'poucr (scc

~

1 74i-4) .

76-15 114204 4/24 5/24 B S lI,W N,X C5, C7 SC tube leak trips nlr cJcctor nnd bloudoun act Iv Ity ilarns (reactor shutduun).

Ii 76-16 5/20 C N Rod posit Iiin devlntloii cliannel ifuund Jnopcr-ablc (scc UER 75-07).

I 76-17 111864 6/18 7/15 0 C,T A N Accuuulatur level falls 115729 belou Tech. Spec.

1Iiiilts.

I is" lc .i ~ ~

I~ ~

76-18 115S95 7/4 7/19 8 I,S 8 C7 Tuu rolls drop (reactor shutdoun;, sec previous events; 76-13, ii,i ~

UER'75-01; nnd 72-08) ~

'I'.ib I a. (0>>>> I Inucil)

NSIC Evc>>L Kapurt I'llani Instruvacaat Cuupunent Abnursall, SLgaa l(lcsncc I osvscnt Nw bca Accession Status Systcaa E la!IF+clat Na!aber Ustc , Date 76-19 115728 7/8 7/16 8 8 Ah 8 C3 I I ra! II!>>sas candu llllhiiil!propel'li>>ll Lurl>>g.

76-20 L15727 7/9 7/L6 U 0 F,N N Sc<<unil UC nut al.artcil uh<<n (Irst UC >>as lnopcrabla.

76-2L L16779 8/4 8/L9 D C7 'I'a!aa ruds drop (ra actor sI!Utalaiunj suc sauc I:vaults ln slaca CKUII cablncts76-18i 76-13 75-01, 'aa J 72-8) . 'EII 76-22 116890 8/5 8/19 h P,FF II C4 I.OCA Ugh changes by Ilu!'Ll>>gliousa.

76-23 LL7969 8/19 9/14 C N Nrung schcalula plpl>>g lnstallcal alurllig last l'a(lial lug slaaatilulin ~

76-24 118969 10/8 10/22 8 EIN,Y C4, C5 l.u'iks In Sl su<<t la>>

pip lllg (a'cilctur shut-alua>>a I sc<<. NS IC-lhiii~

p. 100).

76-25 119758 10/10 Ll/9 0 CC N,CC E N I.cak I>> KIIK ratua'n I I>>i!

ail!I la sli>>Ldulal (uaau 7ia-5) .

4 76-26 120427 I I/16 12/8 8 0 CC N lj>>i! u( tuu rcalunalaut Na!!8 vol va cu>>trull<<rs (all!i Lu resp!>>lil 76-27 121059 12/16 1/14/77 D N,CC- E N I'lllllaIIu Icilk lll a!alii In I<<!diiun illvurtcr I lnu Lo CVCS luatalup ta>>k.

76-28 12/17 Ca 3 N Onu ruil alfi)ps ihli! Lal i!pc>> sL>>L Lail lry hold roll >>hi la at pa>>or (rcactur shutdoun).

(Continued)

NSIC Event 'cport plant Coaponcnt hbnorlLII SibuI iflcance B:

NIxaber hccass ion Date Date Status SyateIe Bqulpnent Instruacnt Status Condition Causa Category Consent Nuabcr 76-29 121060 12/17 1/1 1/77 B Ml FF R .T ~

h ' gvcnt I76 28 caused Iiuadrant-tp-average puucr tilL ratio to exceed Tccil. Spec.

I ieII t; 76-30 121033 12/17 1/12/77 B li,CC l.cak in lctduun linc ucld to calxcd bcd dcalneraliscr.

~ I

.~

7 ~ i ~

i ~

I- i I Ce-. l

~ g~ ~

~

I ~ 0 I <"

I"I il .~

~ ~ ~ ~ ~ ~

I' I i a >- ~ ,

I ~ I ia

~ ~ I

~ ~ l

'l.

~

b I ' l,l

~

I ol I ll I. II ~

II II ~

• I- ul 6 lE I l.l'Iu>

r, 4

Tzbto A2.9 Cudlng Sliaut i>rtabla Kvanls for tilunu 191l Ns lr.

Kvent Rcport Plant instr us>ant t:uul>anent hbnura>>I Causa SIR>>if tea>>ca Cn>>vaunt tt>xaber hcca>>>>lun Data Status Systa>x K>tutpx>ant Status 8 dl I 4Condlllun Cut>>gary Date ttu>xbnr

', 0 . C7 Sl tluup C fol I>> Lu 11-D I 121663 I/3 I/27 8 start on bu>> 14 Jua tu u foil I t y t>r<<akar 71-02 122135 2/17 2/23 8 V C4 Ix>ss-uf-f lou tron>>lant l>UA ann ly>>ls cl>aogad liy lta>>L loUhuosa.

77-03 125040 5/2 5/L6 C S Ll,tttl C7 SU lubun pluRUail ul tur Inspect lnn (scu tlUlllti-OU90, Vuli> I, lt>>. I ~

p. 17).

77-04 132719 5/23 6/20 D Ci I C>ibtall an L>>u caotrlil roil>> ravars<<d dur ln8

~ ufucll>>8 (r<<actur.

>lo>tiluun).

71-U5 125593 5/20 6/27 C Itutaa>>o i>f cl>lnl'lite fr>xa uld lied uf rcsln uxcacils I:CS Tach.

Span. Ltx>tt.

17-06 126094 6/19 7/7 N Chal'ULUR puwp var lilfLvc

>>>>u>kl>>8; 2 uf 3 puups uut uf safvlcc ~

71-01 132120 6/29 7/27 8 C1 SI p>xap falls tu>>tort lin i>us II>.

71-08 126884 7/5 7/18 8 11,tttt C7 Sr, Lubu Lailks trip RCS

>>el lvlly ill>>i'>xs (ral>ctur sbuLduun; s<<c 77-tt3) .

71-09 143450 7/L2 8/2 tl,cc r N I auk lli ui:ld ln charUln8 pu>>>p illscbar8c ral laf pip lo8.

Tabl (Conttuucd) tIS tc Nuaber hcccsslon Kvcnt'cpor't . Plant Systc>a K>tuft>>>>ent ~ Instru>>>cut Co>at>onent hbnors>>t > Cause

. Sign lfl cence oa>>>>ent ttu>>>bar Date .Date Status :Status C dl

>Condition ~

Category C

P>

I 77-10 '14345I:7/13 '7/27 h $$ Z>KK >B IB >C4 ~ Destgn error tn-redun-

'dant- scr>t lcc uatcr . j

>dtschargc linc 'support.

7 7-11 143455 7/13 . 7/27 5 T 'h Change ln contatno>cnt fu>> cooler Tech. Specs.

77-12 143460 7/13,7/29 5 r>x h,PP ~

tt . Sl pout> dlsct>arpe check valve leaking.

77-13 143461 8/2 8/25 5 Il>KK tt,0,55 D <H 2 uf B.SG snubbcrs shou cxccuslvc seal leakage 77-14 143463 8/9 S/18 5 S,V C >

Flux d l ffc r<<neo (con-

'trol rods plus boron

,dttutton) targe( band dcl>>rturc 77-15 143462 8/22 9/2 PP N >Ct>cck valve leakage

, f to>a RMST,to ttqOtl tank.

77-16 143464 8/23 9/2 00 0 N 'l of 2 boric nctd Clou

>patt>s to RCS Isolated, 77-17 -143465 8/24 9/2 8 :8 >C,T 'h ~

H 'Ct>ange'n cnllbr.>Lion procedure-for ttaOtt Iconccut ractun.

77-18 143466 9/4 9/30 5 '5 H,CC 'C Chargtrig t>uup discharge ~'

rcllcf valve pinhole leak ~

~~ ~

77-19 143545 9/14 10/6 5 0 ~ 'C 8 >N :C7 'DC bus l6 breaker 'falls to 'close In test (seo

'77-01, '77-07) .

77-20 VI3467 9/20 10/19 8 H,CC ~

K H ~ Charging puap drain tltnc plnhule leak.

~ ~ ',crt ",

I t.

~

llS IC l;vl>>>r R>.'pol't pl >>lt Col%pa>>la:>>t AL>>>>r!e!1 Ceuse Slg>> ll lce>>ce Coe>>>e!at Ua>aaL er hcccss le>> U<<tc D<<to Status System Rap! tl>>acut 1<<etrulaCnt . d lluebcr I%I 77-21 lo/7 U 0>R C7 UI: gnV<<l'>>nl'lydr!Iul1<<

>>Ctauata>r I'pCCd SCtll>>g

>sf <<>ill $ >>>>l la<<leal Jusl cd.

77-22 11/2 S FF,KK ~

N C7 0>>c ul <<lgllt SC Snad>L<<ill rl!SerVOlru lul>ud empty.

77-23 144122 11/16 I/21/78 U K,R C7 Cu>>trnl rnd>>rga>>t lail lura! rod stop. 2

p. c. co rale l o l 1 ed.

77-24 11/lg K,R C7 Cu>>tl'ul rod urgcut full>>ra: rual stup. 2 p.c. cords f!!lied.

77-25 11/29 ll ~ CC C l'l>>llallu leak lu uo>>>>

rega:>>er!It lvu Ilx nut la!t pip l>>g.

77-26 12/2 CC ll,CC C RUR ll>>alp leek on 1 luu orll lc<< inlet ucld ln tost.

~t CO

0 Table h2.10 Coding Shen eportablo Events for Clnna 1978 NS IC SL)n'lt Event 'cport I'lant Ctuttponc>>t I Abn'oriaal' ... I):anCc' t Nisaber hccess ion Date . Date Status Syateta gclulj>leant'ttstklaae>>LI Stat>>'s'ondition'at'e'gbt S

C dl l ' S C

btytI C>>'tatttc n 78-01 Nutttbc r 134505 1/11 2/8 ' h . Z B'N CC 0 = N'.cttk I .. ~ ."

'ln'cllarglttg dlschac'gc'elief veld.'

~ ~ I putttp plpc 78-02 134488 1/11" 2/10 . B S

'T'" h N failure'lf verify rod stvp pcltttltlslvu anti luu ra>>pc poucr range

~

I trip as rcqulred by Tech. Specs.

78-03 134506 1/25 2/8 I 1, HN C7 SG Luhu leak trips nlr e3ect>tl turu. I I (reactor shutdoun; aec 77-3) ~: ~, p 78-04 137333 3/4 3/31 8 0 GIN Dt> fua'.1 transfer plttttp una'va1)a(tie duc to I~ poor .cuttnpc1lun. I 78-05 137856 4/19 5/2 C A.S P,OO C,S R'doctor taakcult uatcr fct lttto reactor'rruneuusly during rcfue ling.l, Posit lvc . rcacL lvity insert ion. 78-06 139904 7/13 8/3 Z Z,KK,PP Otgg C7,CB Tuo sttul>bors, on NS systcal fitll duu tu 1 I IIctlstant vlbrat ion. Itt (pce 79-,20),t 78-07 140738 8/16 9/14 8 0 F,N C7 D(. bus l6 breaker falls on Lest (scc 79-lg). This ls bus 16 s LenLN such failure. 'fob la (Co<IT lllu<4<lp 8~bar NSIC hcccos lull Nuubcr Evcut l)atc IICI<<>rt Oatu I'lout Slutus Syotcu 8<lull<scut Illotruucut... >IILIIL Stutus 'oudlt lou C<4<>4I> ~ hl>I><>rw'll Couoc S ll'lll I IC>IIICC Cot<>8<>ry 4 t:U<><<><<<<It 18-08 l40235 8/23 9/5 8 K It h N hu><ll lory l>ul I<Ill<0 <II><>l'>I>c>le<I <lurlu8 fuQl Il I<I<II lll0 )8-09 145262 l2/26 l/9/79 8 N I'ruI>cr flro u<ltcll u<>t <l<>UU <lurlu8 I>>lluo oyotc<a rcI>olr. Table h2.11 Cudlng Slice[ Portable Events for Glnna 1979 HSIC Slaaber Accession Event Report .Plant Systela Equip+cut Instru4aent Collponent hbnorunl Causa C Slgnlf l cence CoaIuea t Nuuber Date Date Status Stntua Condl cion Category 79-01 146697 1/3 1/24 8 X 0 h C E C7 Sl puup falls co start frais bus l4 breaker lu l.cst duu tu loose ulre ln clrcult breaker control ulrlng (see 78-7). 79-02 146698 I/I4 1/24 8 E,J C3 Crack found ln charging plulp pluton cylinder. 79-03 146699 1/4 1/24 S P 4MM C4 Second charging plvap (during 79-02 repair) varldrlve speed con-trol lost nt lou spccds. 79-04 147367 2/6 3/6 0 0 DG crlps duc tu luu level ln fuel tank ln case ~ 79>>05 )47368 2/6 3/16 B C4 Cental ltucnt roclrculu-Llou suup outlet valve fulls tu open ln test. 79-06 148569 3/21 4/3 S 1I,HN ll4Y C7 SG cubes plugged after lnspcctlon (sce 77-3) ~ I 79-07 149252 4/2 4/30 8 In to rued I at c range 64onltors not tested Mhen rLIiulrcd 79-08 152012 4/6 8/21 8 CC4H4Y D I.cak found ln boric 152297 ucld f lou control vulva nlpplc. 79-09 149253 4/16 4/30 8 P PP,JJ C,lf Leak frola RCS thru blender boric acid supply check vulvc (seo 79-10) ~ 'I'lab I e (C<<AI lou<!al) ttS IC ttuaber hera U>>hln Event Bepurt I'lnnt Syatca Ealaltpwcut lnct ra>>aleut Laaiatluucnt hbnorcail CQIIUC 8 II II I I Ic aloe e tb>>use nt It>>to I!ate SLutus St>>to>> CQQIIILloll Cu L QKIIl y tlaaabcr 79-10 149254 Ii /20 4/30 8 Yl',J J Catt L.unk friaia t(CS Lliru lila.iul<<r l>>irtc acid t>>ipply I:h<<ck valve (I'cc 7a)-9). ) 9-11 l4970$ 5/ I I 5/25 8 I'%tt I'ul luf vnlvc a:up>>city <<veri!Luteal/ uuala fdu>>tttneat )9-12 1504i I 8 3/27 )/10 h Z,KK C4,CB Stx nnchur liulls for 154821 plpl>>8 Uuppl>rta for uanf Qty catulpilcntu lallinal not up Lo apcl.'a ~ 79-13 150702 7/6 7/23 0 C$ sc tlt Iiozr.lc to cltauu ua.'lalu ln>>pcct lou pair IE Bulletin )9-13. (reactor staaitduun) )9-14 150904 7/9 8/8 CC Unc of Llll'Qc rcahulal lnt uvccpruaIIur I!ant lun <<Inrclu ha>> s<<L point. drift. )9-15 150905 7/24 .8/1 h PaK.AAa KaKK C4,CB I Qn p I lilll8 alllapul to CC nut prup<<rly lnatnl led Ln CS, Rtlt, olid Stl p Uy a t a'Ila O, ~ Ln ) 9-16 150944 8/4 8/16 . 0 L. It. I' S2 Al.'na:Luf QlilirQLLal lllth A(: puu Qr Incur re<< I Iy Uupp I Ia'8 Lo QQIEQ va I vea. 79-17 152178 8/3 I 9/14 8 tthN'una'a:ilt rat liul QIILQCd!I 3<<CII Ataa'1. ~ 1I!>>IL (pouur ra:duc-L lon) 79-18 152194 9/13 LO/l0 8 0 I!Ii output brc>>kcr tu bua 16 Calla Open (Ucu 78-7) ~ ~ ~ Tab (Continued) NS IC Event - Rcport Plant Cocoponcnt Abnursul SlBnl f lcance Co!omen t Number Accession Syste!a Equlpwent instruuent Status Con dltl t on C t C atcBory y Date , Date Status C Number h Z,KK h T E ~ N I.uouc nuts fuund on 79-19 152357 9/19 10/19 BB service uatcr plpln8 supports to AA/PE 0,88 llS snubber foun!l ultll 79-20 152755 10/9 ll/7 A Z,KK C7 nu accuuulator level (sec 78-6). 10/18 11/l6 8 P,Z 00 C T Crossover valves 79-21 152982 between AllN's not stroked ln test. (ncu rcqulrcacnt). I I, it( D,N D CS,C7 SC tube leaks trip air 79-22 153682 12/1 12/14! B S eJcctnr and bluudoun actlvlty s!onltors. (reactor shutduun; scc 77-3). 79-23 154315 12/7 12/21 D Y,CC E,CC D N l.lacer lndlcatluna found in PZR relief noeale-to-safe-end ueld. 154294 12/17 . '2/28 X 33,00 Back(lou tbru blender 79-24 B boric acid check valvu betueen reactor Isakcllp (scc 79-9, -10) and'AS'l'. 79-25 153941 12/23 1/22/80 B N Contains!ent airlock not tested ubcn required. a l