ML20212N604

AEOD/P602, Trends & Patterns Rept of Unplanned Reactor Trips at Us LWRs in 1985

ML20212N604
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Issue date:	08/31/1986
From:	Bell L NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
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TASK-AE, TASK-P602 AEOD-P602, NUDOCS 8608280302
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AE00/P602 Trends and Patterns Report of Unplanned Reactor Trips at U.S. Light Water Reactors in 1985 August 1985 Program Technology Branch Office for Analysis and Evaluation of Operational Data Prepared by:

Lawrence Bell NOTE:

This report documents the results of a study by the Office for Analysis and Evaluation of Operational Data. The findings and recomendations do not necessarily represent the position or requirements of the responsible program office or the Nuclear Regulatory Comission, y

8608280302 860815 PDR ORG NEXD PDR

Trends and Patterns Report Unplanned Reactor Trips at U.S. Light Water Reactors in 1985 Page Number EXECUTIVE

SUMMARY

.................................................... I

1.0 INTRODUCTION

................................................... 6 2.0 OVERVIEW 0F 1985 UNPLANNED REACTOR TRIP STATISTICS............. 8 2.1 Reac tor T ri ps Above 15% Power............................. 11 2.1.1 Initiating Systems................................. 11 2.1.2 Ca u s e s............................................. 17 2.1.3 Contribution From Maintenance, Testing, Troubleshooting, Calibration and Testing........... 30 2.1.4 Plant Trip Rates................................... 36 2.2 Reactor Trips At Power Level s Below 15%................... 40 2.2.1 Initiating Systems................................. 40 2.2.2 Causes............................................. 40 2.2.3 Contribution From Maintenance, Troubleshooting, Calibration and Testing............................ 46 3.0 REACTOR TRIPS WITli ASSOCIATED FAILURES......................... 50 4.0 QUANTITATIVE SAFETY SIGNIFICANCE MEASURES...................... 51 5.0

SUMMARY

OF FINDINGS AND CONCLUSIONS............................ 56 APPENDICES Appendix A - Unplanned RPS Actuations at U.S. LWRs............. 60 in 1985 Appendix B - Summary of 1985 Unplanned Reactor Trip............ 75 Statistics by Plant Appendix C - Human Error Induced Reactor Trips................. 77 Above 15% Power Appendix D - Reactor Trips With Associated Failures............ 83 Appendix E - 1985 Tri ps a t Sho re ham............................ 96 11 i

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List of Figures Figure Page Number 2.1-1 Power Distribution for Reactor Trips Above 15% Power...... 12 (1984 and 1985) 2.1-2 Initiating Systems Sumary Power Greater Than 15%........14 (1984 and 1985) 2.1-3 Initiating Systems By NSSS Power Greater Than 15% PCS....15 and Non-PCS (1984 and 1985) 2.1 Electrical Distribution System Summary Power Above 15%...16 (1984 and 1985) 2.1-5 Cause Summary Power Greater Than 15%..................... 18 (1984 and 1985) 2.1-6 Generic Component Class Power Greater Than 15% (1985)... 19

2. '. -7 Specific Valve Failures Power Greater Than 15%........... 20 (1984 and 1985) 2.1-8 Specific Valve Part Failures Power Greater Than 15%...... 21 (1985) 2.1-9 Main Feedwater Regulatin Valve Part Failures Power...... 23 Greater Than 15%

(1985 2.1-10 Specific Electrical Component Pump Failures.............. 24 Power Greater Than 15% (1985) 2.1-11 Specific Pump Failures Power Greater Than 15% (1985)...... 25 2.1-12 Specific Pump Part Failures Power Greater Than 15%....... 26 (1985) 2.1-13 Main Feedwater Pump Part Failures........................ 27 Power Greater Than 15% (1985) 2.1-14 Human Error - Type of Personnel.......................... 28 Power Greater Than 15% (1984 and 1985) 2.1-15 Personnel Root Causes Leading to Reactor Trips........... 29 Power Greater Than 15% (1985) 2.1-16 Cause of Power Conversion Systems Initiating Trips....... 31 Power Greater Than 35% (1984 and 1985) 2.1-17 Cause of Electrical Subsystems Initiated Trips........... 32 Power Greater 15% (1984 and 1985) iii

List of Figures Figure Page Number E.1-18 Cause of Reactor Protection System Initiated Trips......... 33 Power Greater Than 15% (1984 and 1985) 2.1-19 Trip Causes During Maintenance, Troubleshooting and........ 34 Calibration (1984 and 1985) 2.1-20 Trips Initiated During Testing Power Greater Than 15%...... 35 (1984 and 1985) 2.1-21 Initit ting Systems Affected by Maintenance,................ 37 Troubleshooting, Calibration and Testing Power Greater Than 15% (1984 and 1985) 2.1-22 Reactor Trip Rates vs. Critical Hours by Plant Power....... 38 Greater Than 15% (1984 and 1985) 2.2-1 Reactor Trips At or Below 15% Power (1984 and 1985)....... 41 2.2-2 Initiating Systems Summary at Power Between 0% and 2%...... 42 (1984 and 1965) 2.2-3 Initiating Systems Summary at Power Between 2% and 15%..... 43 (1984 and 1985) 2.2-4 Cause Summary Power Less Than 15% (1984 and 1985)......... 44 2.2-5 Human Error - Type of Personnel Power Greater Than 0%...... 45 and Less Than 2% and Power Greater Than 2 and Less Than 15% (1985) 2.2-6 Personnel Root Causes Leading to Reactor Trips............. 47 Power Between.0% and 2% and Power Between 2% and 15% (1985) 2.2-7 Activity at Time of Reactor Trips Power Less Than 15%...... 48 (1984 and 1985) 2.2-8 Initiating Systems Af fected By Maintenance,................ 49 Troubleshooting, Calibration and Testing Power At Or Below 15% (1984 and 1985) 3.0-1 S a mpl e' Ev e n t T re e.......................................... 53 iv l

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List of Tables Table Page Number 2.0-1 Reactor Scram Frequency (1984 and 1985).............. 9 2.0-2 1985 U.S. LWR Unplanned Trips........................ 11 2.1-1 Plants With Relatively Poor Trip Experience in....... 39 1985 2.1-2 Plants With Relatively Poor Trip Experience in....... 39 1984 s

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EXECUTIVE

SUMMARY

This report analyzes unplanned trips (i.e., scrams) that occurred at U.S. power reactors in 1985. This study is the second in a series of periodic AEOD trends and patterns analysis reports on scrans which draws upon the more complete and detailed operational experience information required as of January 1,1984 by the Licensee Event Report (LER) rule (10 CFR 50.73).

In this report we define a reactor trip (i.e., scram) as any actuation of the Reactor Protection System (RPS) whether automatic or manual which results in control rod motion.

Plants were included in the statistics contained in this report if they: (1) held a full power operating license and (2) accumulated critical hours for some portion of the calendar yecr 1985. The progression of events leading to reactor trips and the post-trip response of the plant and personnel have obvious safety significance. Further, the Commission has concluded that a reduction in

- the frequency of challenges to plant safety systems should be a prime s

goal of each licensee.

Findinas and Conclusions Based on an analysis of the reactor protection system actuations that occurred in 1985, the following conclusions can be made:

1.

Of the 719 urplanned RPS actuations at U.S. Light Water Reactors (LWRs) in 1985, we identified 552 reactor trips (i.e., resulted in control rod motion).

2.

The scram rate for 1984 and 1985 for the industry was approximately one scram per 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of critical operetion.

In terms of the overall performance of the industry we observed a slight change in the total trip rate from 1984 to 1985, (i.e., 5.9 and 6.0 trips per reactor year, respectively). While overall changes were small,

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significant shifts took place at the NSSS specific level.

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3.

In both years the majority of the reactor trips occurred with the

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reactor power above 15 percent: 66 percent.in 1984, and 75 percent in 1985. We further observed that 31 percent of all trips in 1984 j

and 38 percent of all trips in 1985 occurred while the plant was at i-95 percent power or above. We believe this reflects both the minor i

contribution of startup problems to scram frequency and the very short time spent operating in the lower power regime.

4.

Above 15 percent power, the feedwater system was the single system riost responsible for reactor trips, contributing approximately 27 percent in 1984 and 24 percent in 1985.

5.

Above 15 percent power, the major balance-of-plants systems, collectively

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referred to in this report as Power Conversion Systems (PCS) (i.e.,

feedwater, turbine, condensate, main steam, and main generator),

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contributed 59 percent and 52 percent of all reactor' trips above 15 l

percent power in 1984 and 1985, respeciively.

6.

The Electrical Distribution and Reactor Protection System were the major non-PCS contributors in both 1984 and 1985; in these years they contributed 23 percent and 29 percent, respectively, of trips above 15 percent power.

j Hardware failures were the dominant'cause of unplanned reactor tripe 7.

above 15 percent power.

In 1984 and 1985 hardware failures contributed 60 percent and 55 percent respectively of all trips above 15 percent power. The Feedwater Regulating Valve, Main Steam Isolation Valve, 1

Turbine Control Valve, and Turbine Stop Valves were major. contributors.

These valves were responsible for 8 percent of all trips above 15 percent-power in 1985.

I 8.

Personnel related problems (i.e., human error, manual steam generator level control problems ard procedure deficiencies) accounted for 28 percent of' all reactor trips above 15 percent powe'r in 1984 and 32 percent in 1985, making them a substantial but secondary cause of reactor trips! For 1984 and 1985 unlicensed personnel were responsible for 10 percent and 14.

O percent, respectively, of all trips above 15 percent power, with unlicensed technicians involved in roughly one of every 12 trips.

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Non-cognitive errors due to training deficiencies, inattention, poor communication, and failure to follow procedures occurred at nearly twice the frequency of cognitive errors for licensed operators at power greater than 15 percent in 1985. Cognitive errors by technicians occurred at approximately the same frequency as the total error count due to training, inattention, poor communication, and not following procedures.

10.

In 1985 maintenance, troubleshooting, calibration and testing were found to be causal factors in 40 percent of the reactor trips above 15 percent:

25 percent from maintenance, troubleshooting and calibration, and 15 percent from testing.

For 1984, the corresponding percentages were: 36 percent overall,13 percent for maintenance, troubleshooting and criibration, and 23 percent for testing. Thus, these activities constitute a substantial contribution to the overall reactor trip freqpency.

11. At power levels above 15 percent a trip frequency of 2.0 trips per 1000 critical hours was selected as a breakpoint for examining relatively poor performance (this level roughly corresponds to 10-11 scrams per reactoryear).

In 1984 ten plants exhibited rates at or above the cutoff, with a maximum rate of 5.7 trips per 1000 critical hours.

Five of these plants had initial criticality in 1984.

In 1985, a tctal of eight plants exceeded the criterion and six of these had initial criticality in 1985. The maximum trip frequency above 15 percent power for 1985 was 5.3 trips per 1000 critical hours. Although showing large decreases from 1984 to 1985, Callaway 1 and Grand Gulf 1 are the only two plants that are above the cutoff for 1984 and 1985.

12.

For both 1984 and 1985 the Reactor Protection System was the primary initiating system for reactor trips from power levels less than or equal to 15 percent.

13. A decrease in trips below 15 percent power from 1984 to 1985 came from a decrease in trips below 2 percent power, with the RPS and Main Steam System showing the greatest reduction.

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14. Hardware failures were major contributors to trips below 15 percent power for both 1984 and 1985. However, unlike 1984 the differential between hardware and human error initiated trips was much smaller in 1985.

Similar to 1984, personnel related problems (i.e., human error, manual feedwater level control problem, and procedure deficiencies) accounted for a greater share of the reactor trips at low power than at high power in 1985, 48 percent vs. 32 percent. Licensed operators were the dominant persons responsible for trips at power levels below 15 percent.

15.

In 1984 the contribution from maintenance, troubleshooting, calibration and testing below 15 percent power was 35 percent, essentially the same as above 15 percent power.

In 1985, 30 percent of the trips below 15 percent power were caused by these activities vs. 40 percent of the trips above 15 i

percent power. Thus in 1985 these activities collectively have become less significant (i.e., a smaller percentage of an already smaller total

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number of trips below 15 percent power) than in 1984 The reductions for the Main Steam, RPS and Turbine Systems are most responsible for the decrease.

16.

In ?985, 23 percent, or roughly the same percentage as 1984, of scrams above 15 percent power included associated failures (i.e.,

additional failure or personnel error not directly related to the initiator of the scram). About 25 percent of all reactor trips at power levels of 95 percent or higher included at least one associated failure.

A total of 109 trips comprised a total of 139 separate failures. About 22 percent of these scrams involved multiple associated failures.

17. The Accident Sequence Precursor (ASP) approach to quantifying safety significance results in grouping the vast majority of reactor trips, included those with associated failures, into a conditional probability range of IE-7 to 1E-5.

The loss of ICS at Rancho Seco, which was the subject of an IIT, serves as a useful calibration point for the upper end of the range. The San Onofre 1 IIT event is rated one to two orders of magnitude more significant than the majority of reactor trips, and the Davis Besse IIT event still two order of magnitude higher. The quantitative results appear

5 consistent with engireering judgement based on the rough ordering by I

- safety significance of IIT events.

The quantitive methodology should prove useful in gauging the trend in relative safety significance of thb reactor trips over time.

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1.0 INTRODUCTION

This report analyzes unplanned trips (i.e., scrams) which occurred at U.S.

power reactors in 1985. This study is one of a series of periodic AEOD trends and patterns analysis reports which draw upon the more complete and detailed operational experience information required as of January 1, 1984 by the Licensee Event Report (LER) rule (10 CFR 50.73). This report is the second annual report prepared on this subject.

Safety Significance of Reactor Trips There are generally three phases to a scenario or sequence of events involving a reactor trip.

First, there is the generation of some off-normal plant state which results in operation of the Reactor Protection System (RPS) and portions of the Control Rod Drive System (CRDS). Second, there is the operation of the RPS and CRDS themselves. Third, there is the plant and operator response to the trip. Each phase has safety. significance.

The RPS is designed to sense movement of the plant toward a state wherein fuel integrity or reactor coolant system integrity is potentially threatened.

While margins to damage are routinely incorporated in the design of the RPS, the cause and progression of such incidents have obvious safety significance.

Further, the Conunission has concluded that a reduction in the frequency of challenges to plant safety systems should be a prime goal of each licensee, and the Commission believes that large Anticipated Transient Without Scram (ATWS) risk reductions can be achieved by reducing the frequency of transients which call for the RPS to operate.

In this regard, the Commission is interested both in events where the RPS was needed to mitigate the consequences of a transient, and events where an RPS operated unnecessarily, i

Proper operation of the RPS/CRDS in response to a valid demand is obviously a major safety concern.

Finally, because of the rapid change in the plant state, each trip represents g

a potentially stressful situation during which additional equipment operation and operator actions are generally needed in order to bring the plant to a safe and controlled shutdown condition.

Thus we have a safety interest in knowing

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7 the proportion of unplanned reactor trips where recovery was complicated by additional equipment failures or operator errors, and the nature, extent and causes of thos.e complications.

Source of Data Licensee Event Reports which reported an unplanned RPS actuation were selected for detailed review and subsequent sorting by a number of categories (e.g.,

power level prior to the actuation, initiating system).

Paragraph 6

50.73(a)(2)(iv) of the LER rule requires reporting of:

"Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (PDS). However, actuation of an ESF, including the RPS, that resulted from and was part of the preplanned sequence during testing or reactor operation need not be reported."

Thus, actuations that need not be reported and, therefore are outside the scope of this report, are those initiated at the discretion of the licensee as part of a planned procedure or evolution.

Licensee Event Reports were the sole source of technical details about reactor trips. Licensees were not contacted for additional information or clarification. However, in order to ensure that we had a complete set of reports of unplanned reactor trips, we also reviewed the data on reactor trips available from the NRC Operations Center computer file. A complete listing of the RPS actuations identified for 1985 is provided in Appendix A.

Analysis Methodoloay In this report, we attempt to analyze the reactor trip experience from i

numerous perspectives in order to gain as much insight as possible from this operational experience data.

Initially, the data is viewed in aggregate in order to determine the trends in trip experience. Then the data is," cut" based on a number of variables (e.g., plant, NSSS vendor, cause, system in which the transient initiated, power level) in order to identify any patterns or outliers that may be worthy of further study.

In each section, the most

8 dominant, significant, and/or interesting trends or patterns are described in some detail.

2.0 OVERVIEW 0F 1985 UNPLANNED REACTOR TRIP STATISTICS We identified a total of 719 unplanned RPS actuations at U.S. Light Water Reactor (LWRs) in 1985. These RPS actuations occurred over the whole range of plant status, from shutdown and de-fueled to operation at 100 percent power.

In this report we define a reactor trip as an actuation of the RFS, whether automatic or manual, which results in control rod motion. Plants were included in these statistics if they: (1) held a full power operating license *, and (2) accumulated critical hours for some portion of the calendar year in question.

In 1985 there were a total of 552 unplanned scrams at the 92 U.S. LWRs which were licensed to operate at above 5 percent power and which had accumulated some critical hours. The corresponding figures for 1984 were 494 scrams at 83 LWRs.

Of the 552 scrams in 1985, a total of 67 (12 percent) were manual. This percentage matches that for 1984 In both 1984 and 1985 the scram rate for the industry was approximately one scram per 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of critical operation.

The 1985 industry average did not show much change from 1984. The comparative statistics for 1983 through 1985 are shown below.

Scram Type Averace Rate Per Plant Per Year 1983**

1984 1985 Manual 0.9 0.7 0.8 Automatic 5.6 5.2 5.2 Total 6.5 5.9 6.0 See Appendix E for trips at Shoreham 1 that did not meet our criteria for inclusion in this report.

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• • From the NRR study, "1983 Reactor Trip Statistics," dated August 1984.

g Reactor scram rates per 1000 critical hours and per reactor year for unplanned reactor scrams occurring in 1984 and 1985 are displayed in Table 2.0-1.

While there was little change in the PWR aggregate rate, significant shifts took place at the NSSS-specific level.

Table 2.0-1 Reactor Scram Frequency 1984 1985 Scrams Per Scrams Per Scrams Per 1000 Critical Scrams Per 1000 Critical Reactor Year

• Hours Reactor Year

• Hours PWR W

7.1 1.22 6.8 1.04 CE 5.9 0.96 7.5 1.24 B&W 3.0 0.44 5.0 0.88 Total 6.3 1.04 6.7 1.06 BWR GE 5.5 1.12 5.3 0.94 AC (Lacrosse) 7.0 0.94 9.0 1.16 i

The decrease in the Westinghouse (W) average rate from 1.22 to 1.04 per 1000 critical hours is reflective of a broad-based decrease of the individual plant rates. Ofthe35 Westinghouse-designedplantswithdataforbothyears,24 showed a decrease in scram rate, six showed an increase and five remained the same. The two Westinghouse-designed units that achieved initial criticality in 1984 also showed a large decrease in rates from 1984 to 1985.

The increase in the Combustion Engineering (CE) average from 0.96 to 1.24 was largely due to two units with initial critic'ality in 1985--Palo Verde 1 and Waterford 3.

Eliminating these plants from the average results in a decrease for CE plants fron 0.96 to 0.79 scrams per 1000 critical hours.

• For Table 2.0-1, reactor years were calculated for portions of the calendar year where necessary, based on the date of initial criticality.

10 The Babcock & Wilcox (B&W) average is based on the smallest number of plants (sever, in 1984, eight in 1985), and hence is most sensitive to individusi plant behavior. Ncnetheless, the increase in average rate from 1984 to 1985 reflects an increase in the rate for six of the seven plants operating in both years.

Finally, the Gencral Electric (GE) BWR average shows a sizable decrease from 1.12 to 0.94 scrams per 1000 critical hours.

The decrease reflects a drop in rate for 18 of the 28 GE-designed plants which operated in both 1984 and 1985.

This outweighed the increases for the other 10 plants plus the impact of relatively high scram rates from three new plants (two of the plants had initial criticality in 1985; the remaining plant was initially critical on 12/22/84 and was not included with the 1984 data).

Table 2.0-2 provides another overview of unplanned reactor trip statistics for major NSSS vendors *. Division by power level is driven by the distinctly different operating regimes during startup (defined for this report as equal to or less than 15 percent thermal power), and power operation which we defined as any power level above 15 percent.

Analysis of the data (in Table P.0-1 and Table 2.0-2) by NSSS vendor and by vendor subgroups is prcvided because this is such a widely used basis for classifying reactors. However, the reader is cautioned not to infer a strong causal link between NSSS vendor and the occurrence of reactor trips. The range of trip rates show wide variation among individual plants belonging to the same NSSS vendor class. Appendix B provides reactor trip statistics by plant.

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• Statistics for Lacrosse, designed by Allis Chalmers, are omitted.

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11 Table 2.0-2 1985 U.S. LWR Unplanned Trips Number of Trip Rate Per No. of Startup Trips 1000 Critical Hours NSSS Group Units (Power < 15%)

(Power > 15%)

Total Avg. No.

Range Avg. Rate Range Westinghouse:

2-loop 6

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1. 5 0-5 0.2 0-0.4 Early 4-loop 2

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1. 0 0-2 0.3 0-0.6 3-loop 12 15 1.3 0-5 0.6 0-1.1 4-loop 19 43 2.3 0-10 1.2 0.1-5.3 General Electric 31 37 1.2 0-8 0.6 0-2.6 i

Combustion Engineering 13 23

1. 8 0-13 0.9 0-4.5 Babcock &

Wilcox 8

6 0.8 0-2 0.8 0.2-1.6 i

2.1 Reactor Trips Above 15 Percent Power Figure 2.1.-1 displays trip counts as a function of power level for 1984 and 1985.

In both 1984 and 1985 the majority of reactor scrams occurred with the reactor power above 15 percent:

68 percent in 1984, and 75 percent in 1985.

We believe this reflects both the minor contribution of startup problems to scram frequency and the very short time spent operating in the lower power regime.

In fact, 31 percent of total scrams in 1984 and 38 percent of total scrams in 1985,' respectively, occurred while the plant was at 95 percent power l

or above.

Because of the preponderance of scrams above 15 percent power and the greater decay heat removal needs, our analysis is focused first on this power regime.

2.1.1 Initiating Systems We examined each reactor trip to determine the system containing the root cause of the reactor trip.

That system was designated as the " initiating system" if hardware belonging to that system failed; or if operation, 1

i maintenance, or testing of that system led to the reactor trip.

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12 REACTOR TRIPS ABOVE 15% POWER IN 1985 MIDPOINT POWER FREQ CUM.

PERCENT CUM.

FREQ' PERCENT 20 34 34 8.21 8.21 30 34 68 8.21 16.43 40 10 78 2.42 18.84 50 32 110 7.73 26.57 60 15 125 3.62 30.19 l

15 140 3.62 33.82 70 80 29 160 7.00 40.02 90 37 206 8.94 49.76 103 208 414 50.24 100.00 0

50 100 150 200 250 FREQUENCY NOTEiPOWER-20 MEANS 16X THROUGH 24X 100 NEANS Q5X THROUGH 100X REACTOR TRIPS ABOVE 15% POWER IN 1984 NIDPOINT POWER FREQ CUM.

PERCENT CUN.

FREO PERCENT 20 46 46 13.65 13.65 l

30 28 74 8.31 21.96 l

40 Q

83 2.67 24.63 l

50 17 100 5.04 29.67 h

60 15 115 4.45 34.12 l

70 18 133 5.34 39.47 l

80 18 151 5.34 44.81 90 34 185 10.09 54.90 100 152 337 45.10 100.00 0

50 100 150 200 250 FREQUENCY NOTE: POWER = 20 NEANS 16X THROUGH 24X 100 NEANS 95X THROUGH 100X Figure 2.1-1

13 systems and making assignments of faults to the systems, we used nomenclature and boundaries developed for the Nuclear Plant Reliability Data System (NPRDS) whenever possible. The results of this categorization are shown in Figure 2.1-2.

In 1984, the feedwater system was the single system most responsible for trips, and this continued to be the case in 1985. Moreover, the major balance-of-plant systems which collectively are sometimes referred to as the Power Conversion System (PCS) (i.e., feedwater, turbine, condensate, main generator and main steam) accounted for 59 percent and 52 percent of all trips above 15 percent power in 1984 and 1985, respectively. Figure 2.1-3 provides a visual presentation by NSSS vendor of the contributions from PCS and all non-PCS systems for 1984 and 1985.

As was the case in 1984 the Electrical Distribution System and Reactor Protection System in 1985 are again the major non-PCS initiating systems.

The Electrical Distribution System is actually an amalgamation of safety and i

non-safety. subsystems. Five electrical subsystems are used in this report to classify the electrical systems initiating trips for purposes of our evaluation.

The subsystem classifications are: Instrument 120 VAC, Large Plant Loads, Control Center, DC-Bus, and Switchyard.

The Instrument 120 VAC subsystem defines all sources of power for all instrument and control functions under all plant conditions. The class defined as Large Plant Loads includes the 4.16 KV, 6.9 KV, 12.7 KV, and 13.8 KV buses. Losses and upsets on these buses cause the loss of large electric motors, (such as reactor coolant pumps) which l

subsequently result in reactor trips. Switchyard defines those components I

used to interface the plant with the grid. Failures in the subsystem classified as Switchyard affect availability of offsite power sources for the plant such as the 500 KV, 240 KV, 115 KV, and 345 KV buses. The Control Centers include the 480 V and 600 V buses, and DC Bus includes any problem on any DC bus that i

initiates a reactor trip.

Figure 2.1-4 provides a comparison for 1984 and 1985 of the Electrical Distribution System initiated reactor trips. The In:,trument 120 VAC subsystem was most responsible for the increased contribution from Electrical Distribution Systems in 1985.

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INITIATING SYSTEMS

SUMMARY

POWER GREATER THAN 15X.

IN 1985 SYSTEM FREO CUM.

PERCENT CUH.

FREQ PERCENT

• FEEDWATER 100 lea 24.15 24.15 ELECTRICAL 69 169 16.67 40.82

• TURBINE 50 219 12.08 52.90 RPS 40 268 11.F4 64.73 OTHER 49 316 11.59 76.33

• MAIN GFNERATOR 30 346 7.25 83.57 CTRL ROD DRIVE 30 376 7.25 90.82

• COtOENSATE 21 397 5.07 95.89

• MAIN STEAM 17 414 4.11 100.00 v,.

0 20 40 60 80 100 FREQUENCY

• POWER CONVERSION SYSTEMS CPCS)

INITIATING SYSTEMS

SUMMARY

PO4ER GREATER THAN 15%.

IN 1984 SYSTEM FREQ CUN.

PERCENT CUM.

FREQ PERCENT l

• FEEDWATER 91 Gt 27.00 27.00 OTHER 40 140 14.54 41.54

• TURDINE 40 189 14.54 50.08 ELECTRICAL 47 236 13.95 70.03 RPS 52 268 0.50 70.53 M

• CONDENSATE 21 280 8.23 85.76

• MAIN GENLRAWR 20 309 5.93 9I.60 1

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• MAIN STEAM 18 327 5.34 97.03 l

10 337 2.97 100.00 CTRL ROD DRIVE B

20 40 00 83 100 FREQUENCY

= POWER CONVERSION SYSTEMS CPCS)

Figure 2.1-2

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i 1985 INITIATING SYSTEM BY NSSS PCS AND NON-PCS VENDOR FRED CUM.

PERCENT CUh.

FREQ PERCENT WESTINGHOUSE

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194 194 46.86 46.86

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GENERA *. ELECTRIC 113 307 27.20 74.15 COMBUSTION ENG 72 379 17.39 91.55 BABCCCK WILCOX 28 407 6.76 98.St ALLIS CHALMERS 7

414 I.89 190.00 0

50 100 150 200 FREQUENCY LEGEND: SYSTEM F29995 NON PCS M POWER CONVERSION POWER CONVERSION SYSTEMS (PCS) INCLUDE THE FEEDWATER MAIN STEAM MAIN GENERATOR CONDENSATE AND TURBINE SYSTEMS.

1984 INITIATING SYSTEM BY NSSS PCS AND N0r4-PCS s

VENDOR FREQ CUM.

PERCENT CUM.

FREQ PERCENT WESTINGHOUSE 158 158 46.88 40.98 GENERAL ELECTRIC I10 208 32.84 79.53 COM8USTION ENG 42 Ste 12.40 et.99 8A900CK WILCOX 29 SSe 5.03 07.92 ALLIS CHALMERS 7

337 2.09 lee.Se 9

59 18e 158 200 FREQUENCY LEGENDS SYSTEM NON PCS POWER CO'fVERSION POWER CONVERSION EMS CPCS) INCLUDE THE ATER MAIN STEAM MAIN GENERATOR CONDENSATE AND TUR8INE SYSTEMS.

Figure 2.1-3

16 ELECTRICAL DISTRIBUTION SYSTEM

SUMMARY

POWER GREATER THAN 15X.

IN 1985 SUBSYS FREQ CUH.

PERCENT CUH.

FREO PERCENT INSTRU 120VAC 35 35 51.47 St.47 SWITCH YARD 14 40 20.50 72.06 La PLANT LOADS 11 60 18.18 88.24 CTRL CENTER 5

65 7.35 95.50 DC BUS 3

68 4.41 100.08 0

10 20 30 40 50 FREuuENCY INSTRU I20VAC INCLUDES ALL INSTRUNENT POWER SUPPLIES AND OTHER VITAL OR UNINTERRUPTA9LE 120V POWER SOURCES.

Lo PLANT LOADS INCLUDES THE 4.10KV 6.GKV 13.8KV BUSES.

CTRL CENTER INCLUDES THE 480V AND 000V LOAD CENTERS.

SWITCH YARD DEFINES THE GRID CONNECTIONS TO THE UNIT.

ELECTRICAL DISTRIBUTION SYSTEM

SUMMARY

POWER GREATER THAN ISX.

IN 1984 SUBSYS FREQ CUM.

PERCENT CUN.

FREQ PERCENT La PLANT LOADS 17 17 38.00 36.90 INSTRU 120VAC le 33 34 18 71.74 SWITCH YARD 8

41 17.30 80.13 CTRL CENTER 3

44 6.S2 9S.6S DC BUS 2

46 4.3S 100.08 0

10 20 30 40 50 FREQUENCY INSTRU 120VAC INCLUDES ALL INSTRUHDIT POWER SUPPLIES AND OTHER VITAL OR UNINTERRUPTABLE 120V POWER SOURCES.

La PLANT LOADS INCLUDES THE 4. lSKV 6.WCV 13.8KV BUSES.

CTRL CENTER INCLUDES THE 480V AND 680V LOAD CENTERS.

SWITCH YARD DEFINES THE GRID CONNECTIDNS TO THE UNIT.

Figure 2.1-4

17 2.1.2 Causes The LER description of each trip was reviewed to determine the general classification of root cause or causes as shown in Figure 2.1-5 While most LER descriptions for each reactor trip are self evident, the Manual Steam Generator Level Control category requires some further explanation.

This category was established for situations in which the licensee attributed the trip to general difficulty in controlling steam generator level, while not specifically citing operator error in controlling water level.

Above 15 percent power, hardware failures dominated in both years.

Figure 2.1-6 provides a summary of the generic component classes associated with hardware failures in 1985. Over 80 percent of all hardware failures can be classified as a valve, electrical component, pump, or control circuit

• problem.

These generic classes were analyzed to determined which valves, electrical components, and pumps were the major contributors to unplanned reactor trips.

Valve malfunctions accounted for 10 percent and 17 percent of all trips above 15 percent power in 1984 and 1985, respectively.

Figure 2.1-7 shows that for the past two years the Feedwater Regulating Valve (FRV), Main Steam Isolation Valve (MSIV), Turbine Control Valve (TCV), and Turbine Stop Valve have been the major contributors. These valves were responsible for 8 percent of all trips in 1985 above 15 percent power.

Figure 2.1-8 shows that approximately 60 percent of all valse problems are concentrated in remotely operated valve Control loop.**

Figure 2.1-9 summarizes piece part failures for FRV. Most FRV problems were in the valve's positioner. Most MSIV and TCV problens tended to be related to limit switch faults.

• Control circuit as used here ir.cludes all signal processors or direct control devices that control the operation of main components such as valves, pump, etc., and also includes such items as integrated circuit boards, and other electronicdevices(i.e., resistors, capacitors).

• • The control loop consists of valve positioners, operators, controllers, and limit / torque switches used by valves to control flow, or initiate opening and closing.

18 CAUSE

SUMMARY

POWER GREATER THAN 15%

1985 CAUSE FRE0 CUM.

PERCENT CUN.

FREO PERCENT HARDWARE 228 228 55.07 55.07 HUNAN ERROR 104 332 25.12 80.19 PROCEDURES 26 358 6.28 86.47 UNKNOWN 24 382 5.80 92.27

'3YSTEM DESIGN 11 393 2.00 94.93 ENVIRONMENTAL 9

402 2.17 97.10 SG LEVEL 5

407 I.21 98.31 HUMAN HARDWARE 5

412 1.21 99.52 NOT PROVIDED 2

414 0.48 100.00 0

50 100 150 200 250 l

FREQUEhCY CAUSE

SUMMARY

POWER GREATER THAN 15%

1984 CAUSE FRE0 CU PERCENT M

HARDWARE 204 204 60.53 60.53 HUMAN ERROR 75 279 22.26 82.79 PROCEDURES 12 291 3.56 86.35 HUMAN HARDWARE 11 302 3.26 P9.61 ENVIRONMENTAL 10 312 2.97 92.58 l

UNKNOWN 10 322 2.97 95.55 l

SG LEVEL 8

330 2.37 97.92 NOT PROVIDED 5

335 1.48 09.41 SYSTEM DESIGN 2

337 0.59 100.00 0

50 100 150 200 250 FREQUENCY Figure 2.1-5 i

19 GENERIC COMPONENT CLASS POWER GREATER THAN 15X YEAR 1985 CLASS FREO CUM. PERCOff CUM.

FREO PERCENT VALVE 71 71 31.14 31.14 ELECTRICAL 53 124 23.25 54.39 PUMP 33 157 14.47 68.00 CONTROL CIRCUITS 31 188 13.68 82.46 PIPE 11 199 4.82 87.28 GENERATOR 10 289 4.39 91.67 TURBINE 7

216 3.07 94.74 NUCLEAR INSTRU 6

222 2.63 97.37 CRDM 2

224 8.88 98.25 HEAT EXCHANGER I

225 0.d4 98.08 TANK l

226 0.44 99.12 CONDENSER I

227 af. 44 99.56 PROCESS COMPUTER I

228 8.44 100.00 e

2e de 6e se FREQUENCY Figure 2.1-6

20 SPECIFIC VALVE FAILURES POWER GREATER TNAN 15%

YEAR 1985 COMPS FREQ CUM.

PERCENT CUM.

FREQ PERCENT MAIN FV REG VLV 15 15 21.13 21.13 MSIV G

24 12.88 33.80 TUR CTRL VLV 8

32 1I.27 45.07 RELIEF VALVE 4

35 5.63 50.70 TUR INTERCEPT V 3

39 4.23 54.93 INSTRU ISOL VLY 3

42 4.23 59.15 PRZ SPRAY VLV 3

45 4.23 63.38 FV RECIR VLV 3

48 4.23 67.61 FW ISOL VLV 3

El 4.23 71.83 FL CTRL VLV 3

54 4.23 76.06 TRIP SOLENOID 2

SS 2.82 78.87 STN DUMP VLV 2

58 2.82 81.69 CRO PUMP STOPCK 1

59 1.41 83.1P OIL PUMP SUC Y 1

60 t.41 84.51 SCRAM SOL VLV 1

61 I.41 85.92 MAKEUP VLV 1

62 1.41 87.32 SG BLOVD VLV t

63 1.41 88.73 CIRC WTR DISC V I

64 1.41 90.14 MSR SPILL VLV l

65 1.41 91.55 HEAT DRTK DISCV 1

66 1.41 92.96 SJAE SUC VLV I

67 1.41 94.37 SJAE S*JPPLY VLV 1

68 1.41 95.77 EPR CTRL VLV 1

69 1,41 97.18 CIV 1

70 1.41 98.59 PSI REGULATOR 1

71 1.41 100.00 0

5 10 15 FREQUENCY SPECIFIC VALVE FAILURES POWER GREATFR THAN 1SX YEAR 1984 VALVE FREO CUN.

PERCENT CUN.

FREQ PERCEbT NAIN FW REG VLV 10 10 28.57 28,57 NSIV 7

17 i'0.00 48.57 TURBINE STOP VLV 4

21 11.43 60.80 MSR DRAIN VLV 3

24 8.57 68.57 STM DUMP VLV 1

25 2.86 71.43 FW ISOL VLV I

26 2.86 74.29 CONDENSATE CLEAN 3

1 27 2.80 77.14 FVP CONDENSER 1

28 2.86 60.00 S AFETY RELIEF 1

29 2.06 62.86 SG INLET CHECK 1

30 2.86 85.7f FW BYPASS YLV 1

31 2.86 88.57 PRZ SPRAY VLV 1

32 2.86 91.43 CIRC WTR DISC V 1

33 2.86 94.29 NS BYPASS VLV 1

34 2.86 97.14 INTERNEDIATE VAL 1

35 2.86 100.00 e

5 le 15 j

FREQUENCY Figure 2.1-7

21 S33CIflC VLVE PART FAILU:lES 1

POWER SREATER THAN 15X YEAR 1985 PART FRE0 CUM. PERCENT CUM.

FRE0 PERCENT VALVE SEAT 15 15 21.13 21.13 SVITCH 9

24 12.68 33.80 l

NOT PROVIDED 8

32 11.27 45.07 SCtEN0ID 8

40 11.27 56.34 OPERATOR 7

47 9.86 66.20 POSITIONER 7

54 9.86 76.06 CCNTROLLER 6

60 8.45 84.51 1

IC CTRL CARD ETC 3

63 4.23 88.73 VALVE STEM 3

66 4.23 92.96 ACT'JATOR 2

68 2.82 95.77 RELAY 2

70 2.82 98.59 BUSTLING BEARING 1

71 1.41 100.00

_ p,,,

0 5

10 15 FREQUENCY Figure 2.1-8

22 Specific electrical component failures are summarized in Figure 2.1-10.

Operating conditions such as vibration, humidity, dust and other environmental conditions are responsible for the failure of major contributors, i.e., inverter and transformer. A preliminary AE0D case study report " Operational Experience Involving Losses of Electrical Inverters" expands on the root causes of electrical system failures.

In 1985 main feedwater pumps experienced problems seven times more frequently

)

than the pump with the next highest count. A sumary of all pump failures is i

provided in Figure 2.1-11 and a sumary of specific pump part failures is provided in Figure 2.1-12.

Figure 2.1-13 summarizn piece part failures for the feedwater pumps.

From Figure 2.1-13 it is evident that pump control problems are dominant.

In most cases the problems were associated with pumps that are turbine driven.

personnel related proolems (i.e., human error, manual steam generator level i

control problems, procedure deficiencies) accounted for 28 percent of all reactor trips above 15 percent power in 1984 and 32 percent in 1985, making i

them a substantial, but secondary cause. Appendix C provides further details i

on the 104 trips categorized as human error in Figure 2.1-5 for 1985. A summary of the type of personnel responsible for human error initiated trips is provided in Figure 2.1-14 For 1984 and 1985 unlicensed personnel were responsible for f

, 10 percent and 14, percent, respectively, of all trips above 15 percent pow. ;

furthermore, technicians (unlicensed) were involved in rcughly one of every IP. trips in each of the years evaluated.

The root causes of the 104 human error initiated trips in 1985 are provided in Figure 2.1-15.

Non-cognitive errors due to training deficiencies, inattention, poor connunication, and failure to follow procedures occurred at nearly twice the frequency of cognitive

• errors for licensed operators at power greater

• Cognitive error is defined as an error made by licensed or unlicensed person-nel during the performance of their normal job function. Cognitive errors include all mistakes made by personnel with a broad understanding and high degree of training in relation to the task at hand.

This type of error is primarily the result of a mental lapse and is not correctible with revised procedures or additional training.

e 23 l

MAIN FEEDWATER REGULATING VALVE PART FAILURES i

POWER GREATER THAN 15X YEAR 1985 PART FREQ CUM.

PERCENT CUM.

FREQ PERCENT POSITIONER 7

7 40.07 40.67 NOT PROVIDED 3

te 20.80 60.67 i

OPERATOR 2

12 13.33 88.80 VALVE STEM t

13 0.07 80.07 RELAY l

14 6.87 G3.33 I

CONTROLLER 1

16 6.67 100.00 l

i j

O t 2 S 4 5 6 7 FREQUCNCY l

l l

4 1

l l

e i

Figure 2.1-9

_n..----,-,,..

24 SPECIFIC ELECTRICAL COMPONENT PART FAILURES POWER GREATER THAN 15%

YEAR 1985 PART FREO CUM. PERCENT CUM.

FRE0 PERCENT INVERTER 18 18 18.87 18.87 TRANSFORMER 8

18 15.89 33.96 ELEC CONNECTION 7

25 13.21 47.17 BUS 5

38 9.43 56.60 RELAY 5

35 9.43 86.e4 BREAKER 4

39 7.55 73.58 FUSE 3

42 5.66 79.25 INSULATOR 3

45 5.66 84.91 NOT PROVIDED 2

47 3.77 68.68 RESISTOR I

48.

l.89 90.57 VOLTAGE REG I

49 1.89 92.45 IC CTRL CARD ETC 1

58 f.89 94.34 ARRESTOR 1

SI 1.89 96.23 NG SET l

52 1.89 98.II POWER SUPPLY 1

53 1.89 100.80 0

2 4

6 8

10 FREQUENCY

)

1 l

)

'I Figure 2.1-10

25 SPECIFIC PUMP FAILURES POWER GREATER THAN 15X YEAR 1985 COMPS FRE0 CUM. PERCENT CUM.

FRE0 PERCENT FW PUMP 21 21 63.64 63.64 RX COOLANT PUMP 3

24 9.09 72.73 CIRC WATER PUMP 3

27 9.09 81.82 SS CR TK PUMP l

28 3.03 84.85 RHR PUMP 1

29 3.03 87.88 SJAE DR PUMP l

30 3.03 90.91 GTATOR COOL PUMP I

31 -

3.03 93.94 CO@ENSATE PUMP I

32 3.03 96.97 CMARGING PUMP l

33 3.03 100.00 0

10 20 30 FREQUENCY i

Figure 2.1-11

26 8:33C 7;C PUE3 par 7A]WS POWER GREATER THAN 15%

YEAR 1985 PART FRE0 CUM. PERCENT CUM.

FRE0 PERCENT CONTROLLER 8

8 24.24 24.24 IC CTRL CARD ETC 6

14 18.18 42.42 BUSHING BEARING 4

18 12.12 54.55 NOT PROVIDED 3

21 9.89 63.64 PUMP CASING 2

23 6.06 69.70 STRAINER 2

25 6.06 75.76 SENDING UNIT 2

27 6.06 81.82 ELEC CONNECTION 2

29 6.06 87.88 PUMP TURBINE I

30 3.03 90.91 MOTOR ELEC 1

31 3.03 93.94 SEAL GASKET l

32 3.03 96.97 RELAY I

33 3.03 100.00 012345678 FREQUENCY l

Figure 2.1-12 L.

l 27 l

MAIN FEEDWATER PUMP PART FAILURES POWER GREATER THAN 15%

YEAR 1985 PART FREQ CUM. PERCENT CUN.

FRE0 PERCENT F3 8

8 38.10 38.10 CONTROLLER IC CTRL CARD ETC 3

11 14.29 52.38 TN3HING BEARING 14 14.29 06.67 SENDING UNIT 2

16 9.52 76.19 ELEC CONNECTION 2

18 9.52 85.71 PUMP TURBINE I

19 4.76 90.48 STRAINER I

20 4.76 05.24 NOT PROVIDED 1

21 4.76 100.00 012345678 FREQUENCY Figure 2.1-13

28 4

HUMAN ERROR - TYPE OF PERSONNEL POWER GREATER THAN 15X r

YEAR 1985 i

PERSON FREQ CUM. PERCENT CUM.

FREQ PERCENT

. TECHNICIAN 38 38 38.54 36.54 LIC OPERATOR 29 67 27.88 64.42 NOT DETERMINED 9

78 8.65 73.08

.UNLIC OPERATOR 9

85 8.65 81.73 UNLIC WORKER 6

9I 5.77 87.50 LICENSEE STAFF 6

97 5.77 93.27 CONTRACTOR 3

100 2.86 96,15 STAFF OPS 2

102 1.92 98.08 l

CONTRACT WORKi.T' t

103 0.96 99.04 i

I LIC MANAGEMENT I

104 0.96 100.00

-y.

0 10 20 30 40 50 FREQUCNCY HUMAN ERROR - TYPE OF PERSONNEL POWER GREATER THAN 15X YEAR 1984 PERSON FREO CUM. PERCENT CUM.

FREQ PERCENT TECHNICIAN 28 26 34.67 34.67 LIC (FERATOR 19 45 25.33 60.00 LICENSEE STAFF 10 55 13.33 73.33 NOT DETERMINED 9

64 12.00 85.33 UNLIC OPERATOR 4

68 5.33 90.67 UNLIC WORKER 2

70 2.67 Q3.33 l

LTC MANAGEMENT 2

72 2.67 98.00 CONTRACTOR 1

73 1.33 97.33 VENDOR STAFF 2

74 1.33 98.67 STAFF DPS 1

75 1.33 100.00

.......n,..

0 10 20 30 40 50 FREQUENCY Figure 2.1-14

29 PERSONNEL ROOT CAUSES LEADING TO REACTOR TRIPS POVG GRe.ATER 111AN 15%

YSAR 1985 ACTION FREQ CUM. PERCENT CLN.

FRE0 PERCENT COGNITIVE 28 28 26.92 26.92 ADMIN CTRL 24 52 23.08 50.08 TRAINING 19 71 18.27 88.27 INATTENTION 9

80 8.85 78.92 PROC NOT USED 7

87 8.73 83.65 I

CONMLNICATION 5

92 4.81 88.46 PLANNINS 4

96 3.85 92.31 l

OTHER 3

99 2.88 95.19 I

TEST EQUIP USE 3

102 2.88 98.08 j

HUMAN FACTOR 2

104 1.92 100.00 0

10 20 30 FREQUENCY i

l i

i i

~

f i

h Figure 2.1-15 1

30 thar. 15 percent.

It should also be noted that in the power ranse greater than 15 percent, technicians tended to initiate slightly more trips than licensed operators. Cognitive errors by technicians occurred at approximately the same frequency as non-cognitive errors due to training, inattention, poor communi-cation, and not following procedures.

Section 2.1.1 identified PCS systems, the Electrical Cistribution Systems, and the Reactor Protection System as the systems in which reactor trips are most frequently initiated.

Figures 2.1-16, 2.1-17 and 2.1-18 show the breakdown of causes within each of these system classifications for the two year period 1984 and 1985. There is a gradual decrease in the ratio of hardware-to-personnel causes moving through the figures.

In 1984 and 1985 hardware failure initiated trips predominate for Power Conversion, accounting for 135 of 199, or 68 percent of all reactor trips in 1984 and 131 of 218, or 60 percent in 1985. Hardware failure in PCS system accounted for 31 percent of all trips in 1985 at power levels above 15 percent vice 40 percent for 1984.

In both 1984 and 1985 haroware failure was the largest contributor to trips originating in the Electrical Distribution System, although to a lesser degree than the PCS case'.

Finally, in the case of the Reactor Protection System human error predominates.

However, in both 1984 and 1985 such errors accounted for only 5 percent of all reactor trips above 15 percent power.

2.1.3 Contribution from Maintenance, Testing, Troubleshooting, Calibration and Testing The concern is frequently raised that maintenance, troubleshooting, calibration and testing at power pose significant risks. Figures 2.1-19 and 2.1-20 place that risk in perspective.

In 1985 these activities were found to be causel factors in 40 percent of the reactor trips above 15 percent power: 25 percent from maintenance, troubleshooting and calibration (which are often not elective but part of needed corrective action), and 15 percent from testing.- For 1984 the corresponding percentages were: 36 percent overall,13 percent from maintenance, troubleshooting and calibration and; 23 percent for testing.

i

31 CAUSE OF POWER CONVERSION SYSTEMS INITIATED TRIPS POWER GREATER THAN 15%

IN 1985 CAUSE FREQ CUM. PERCENT CUM.

FREQ PERCENT HARDWARE xxxxxanuu u u u w i 131 131 88.39 88.80 HlNAN ERROR

• N 47 178 21.58 8i.85 PROCEDURES 14 192 8.42 88.07 UNKNOWN 12 204 5.50 93.58 SG LEVEL 5

299 2.29 95.87 SYSTEM DESIGN l

4 213 f.83 97.71 ENVIRONMENTAL 2

215 8.92 98.82 HUMAN MARDWARE 2

217 8.92 99.54 NOT PROVIDED t

218 8.46 100.80 0

58 100 158 FREQUENCY LEGEND: SYSTEM

>*.**a TURBINE CDPOENSATE T N1 MAIN GENERATOR EXX2 MAIN STEAM

== FW POWER CONVERSION SYSTEMS (PCS) INCLUDE THE FEEDWATER MAIN STEAM MAIN GENERATOR CONDENSATE AND TURBINE SYSTEMS.

CAUSE OF POWER CONVERSION SYSTEMS INITIATED TRIPS POWER GREATE*4 TMAN 15X IN 1984 CAUSE FREQ CUM. PERCENT CUM.

FREQ PERCENT HARDWARE 135 135 87.84 87.84 HUMAN ERROR 35 170 17.59 85.43

)

SG LEVEL 8

178 4.92 89.45 h

PROCEDURES 8

186 4.82 93.47 h

HUMAN HARDWARE 6

192 3.82 96.48 l

NOT PROVIDED 4

198 2.8l IA8.49 UNKNOWN 2

198 1.01 99.50 SYSTEM DESIGN 1

199 0.50 198.98 rm, mep.

e 5e 188 iSe j

FREQUENCY LEGEND SYSTEM txxxx1 TURBINE EBER COPOENSATE

• * *

• MAIN GENERATOR

' *

• MAIN STEAM r,ru Fw POWER CONVERSION SYSTEMS (PCS) INCLUDE THE FEEDWATER MAIN STEAM MAIN GENERATOR cot 0ENSATE AND TURBINE SYSTEMS.

Figure 2.1-16 i

32 CAUSE OF ELECTRICAL SUBSYSTEMS INITIATED TRIPS POWER GREATER THAN 15X IN I985 CAUSE FREQ CUM.

PERCENT CLN.

FREQ PERCENT HARDWARE 36 36 52.I7 52.t7 HUMAN. ERROR 24 68 34.78 E6.96 ENVIRONMENTAL 4

64 5.80 92.75 l

UNKNOWN 4

Se 5.80 98.55 PROCEDURES 1

60 t.45 190.98 9

le 20 30 40 50 FREQUENCY ELECTRICAL SUBSYSTEMS INCLUDES INSTRUMENT 120VAC,LARGE PLANT LOADS, CONTROL CENTERS, SWITCH YARD.

CAUSE OF ELECTRICAL SUBSYSTEMS INITIATED TRIPS POWER GREATER THAN 15X IN 1984 CAUSE FREQ CUM. PERCENT CUM.

FREQ PERCENT HARDVARE 23 23 48.94 48.94 HUMAN ERROR 11 34 23.40 72.34 ENVIRONMENTAL 6

40 12.77 85.II HUMAN HARDWARE 3

43 6.38 91.49 PROCEDURES 2

45 4.20 95.74 HOT PROVIDED I

46 2.13 97.87 SYSTEM DESIGN I

47 2.13 100.09 O

le 20 30 40 50 FREQUENCY El.EC1RICAL SUBSYSTEMS INCLUDE: INSTRUHENT 120VAC.LARGE PLANT LOADS, CONTROL CENTERS, SWITCH YARD.

Figure 2.1-17

33 CAUSE OF REACTOR PROTECTION SYSTEM INITIATED TRIPS POWER GREATER THAN 15X IN 1985 CAUSE FREO CUM. PERCENT COM.

FRE0 PERCENT HUHAN ERROR 21 21 42.86 42.86 HARDWARE 14 35 28.57 71.43 PROCEDURES 7

42 14.29 85.71 l

UNKNOWN 3

45 6.12 91.84 SYSTEH DESIGN 3

48 6.12 97.96 ENVIRONMENTAL 1

49 2.04 100.00

.,.nm.......

i 0

10 20 30 FREQUENCY CAUSE OF REACTOR PROTECTION SYSTEM INITIATED TRIPS POVER GREATER THAN 15X IN 1984 CAUSE FREO CUM.

PERCENT Ct.M.

FREQ PERCENT HUMAN ERROR 17 17 53.13 53.13 HARDWARE 9

26 28.13 81.2S UNKNOWN 5

31 15.63 96.88 ENVIRONMENTAL 1

32 3.13 180.Bd

..........,,m,r-0 10 20 30

~

FREQUENCY Figure 2.1-18 i

34 TRIP CAUSES DURING MAINTENANCE, TROUBLESHOOTING,AND CALIBRATION POWER GREATER THAN 15%

IN 1985 CAUSE FREQ CUM. PERCENT CUM.

FREQ PERCENT HUMAN ERROR 46 46 45.54 45.54 HARDWARE 28 74 27.72 73.27 PROCEDURES 14 88 13.86 87.13 UNKNOWN 6

94 5.94 93.07 SYSTEM DESIGN 3

97 2.97 96.04 ENVIRONMENTAL 2

99 1.98 98.02 l

HUMAN MARDWARE 2

101 1.98 100.00 0

10 20 30 40 50 FREQUENCY TRIP CAUSES DURING MAINTENANCE, TROUBLESHOOTING,AND CALIBRATION' POWER GREATER THAN 15%

IN 1984 CAUSE FREQ CUM.

PERCENT CUM.

FREQ PERCENT HUMAN ERROR 20 20 47.62 47.62 i

HARDWARE 17 37 40.48 88.'8 NOT PROVIDED 1

38 2.38 90.48 UNKNOWN 1

39 2.38 92.88 PROCEDURES 1

40 2.38 95.24 HUMAN HARDWARE 1

41 2.38 07.62 SYSTEM DESIGN 1

42 2.38 100.00 0

10 20 30 40 50 FREQUENCY Figure 2.1-19

35 TRIPS INITIATED DURING TESTING POWER GREATER THAN 15%

IN f985 CAUSE FREQ CUM. PERCENT CUM.

FREQ PERCENT HARDWARE 33 33 51.56 51.58 HUMAN ERROR 21 54 32.81 84.38 PROCEDURES 6

60 9.38 93.75 LNKNOWN 4

64 6.25 100.30 0

10 20 30 40 50 FREQUENCY l

TRIPS INITIATED DURING TESTING POWER GREATER THAN 15%

IN 1984 CAUSE FREQ CUM.

PERCENT CUM.

REQ PERCENT HARDWARE 34 34 43.94 43.94 i

HUMAN ERROR 30 64 37.07 81.0I PROCEDURES 6

70 7.59 88.61 HUMAN HARDWARE 5

75 6.33 94.94 NOT PROVIDED 2

77 2.53 97.47 UNKNOWN I

78 1.27 98.73 SYSTEM DESIGN 1

70 t.27 100.00 0

10 20 30 40 Se FREQUENCY Figure 2.1-20 l

1

36 Thus, these activities constitute a substantial contribution to the overall reactor trip frequency.

To gauge whether this is purely a personnel-related problem we examined the caused breakdown within these activities. We noted that in both years human error predominated for maintenance, troubleshooting and calibration, but hardware failure dominated for testing.

Finally, Figure 2.1-21 shows the systems affected by the four activities. Here some changes are evident from 1984 to 1985. Most notably, the turbine has dropped in contribution, while electrical distribution and RPS have gained.

2.1.4 Plant Trip Rates Figure 2.1-22 is a plot of each plant's reactor scram rate for scrams above 15 percent power. The axes are scrams per 1000 critical hours and number of critical hours for the plant in 1984 and 1985. One feature noticeable in Figure 2.1-22 is the general shift of the reactor population to higher per plant critical hours in 1985; the average was 5551 critical hours in 1984 and 5878 critical hours in 1985.

For 1984, a scram frequency of 2.0 scrams (above 15 percent power) per 1000 critical hours was selected as a breakpoint of examining relatively poor performance. Ten plants exhibited rates at or above the cutoff, with a maximum rate of 5.7 screms per 1000 critical hours.

Five of these plants had initial criticality in 1984 Proceeding similarly for 1985, a total of eight plants met the criteria, and six of the eight had initial criticality in 1985.

Although showing large decreases from 1984 to 1985, Callaway 1 (criticality 10/2/84) and Grand Gulf I are the only two plants that are above the cutoff in both 1984 and 1985.

Lastly, we note that the maximum scrau frequency for 1985 is 5.3 scrams per 1000 critical hours, less than the maximum for 1984.

Tables 2.1-1 and 2.1-2 are provided to show those plants with relatively poor performance over the two year period 1984 and 1985.

a

37 INITIATING SYSTEMS AFFECTED BY MAINTENANCE TROUBLESHOOTING CALIBRATION 1 TESTING POWER GREATER THAN 15%

IN 1985 SYSTEM FREQ CUM.

PERCENT CUM.

FREQ PERCENT RPS 34 34 19.77 19.77

• FEEDWATER 31 - - 65 18.02 37.79 ELECTRICAL

- 26 9J 15.12 52.91 OTHER M

21 112 12.21 65.12

• TlRBINE 20 132 11.63 76.74 CTRL ROD DRIVE M

13 145 7.56 84.30 Q

10 155 5.81 90.12

• MAIN STEAM h

9 164 5.23 95.35

• MAIN GENERATOR h

• CONDENSATE 8

172 4.65 100.00 0

10 20 30 40 50 FREQUENCY LEGEND: ACTIVITY rrrm TESTANG nmEtKat TROUBLESHOOTING

2XI3 CALIBRATION rxT2 MAINTENANCE 1

l 1

INITIATING SYSTEMS AFFECTED BY MAINTENANCE TROtmLESHOOTING CALIBRATION & TESTING POWER GREATER THAN 15X IN 1984 SYSTEN FREQ CUM.

PERCENT CUM.

FREQ PERCENT

• TURBINE M

24 24 19.83 19,83 RPS 21 45 17.36 37.10 OTHER 29 65 16.53 53.72

• FEEDWATER 29 85 16.53 70.25

• MAIN STEAM 12 97 9.92 80.17 ELECTRICAL 10 197 8.26 86.43

• CONDENSATE 9

116 7.44 95.37 l

CTRL ROD DRIVE 3

119 2.48 96.J5 l

• MAIN GENERATOR 2

121 1.65 100.00 9

10 29 39 40 50 FREQUENCY LEGENDS ACTIVITY MIII3 TESTING m TROUBLESHOOTING M EI3 CALIBRATION

• w=

MAINTENANCE Figure 2.1-21

38 REACTOR TRIP RATES VS CRITICAL HOURS BY PLANT POWER GREATER THAN 15X 10000:

P L

44 +t * +

A

$+

p+

+ Callaway 1 N

3% 4 4$ t+

+

C

-43 k + + +++,,

R

!50005+

• +

g

+ Grand Gulf 1

+

+

• lf Creek Wo C

t A

2

++

+ Catawba

+Waterford 3

+ +

+ Palo Verde 250ek

+ +

g 3

+

+Diablo Canyon 2 3

U R

+

S 0

e e

1 1

2 2

3 3

4 4

5 5

6 5

5 5

5 5

5 i

5 5

5 5

5 5

PLANT TRIP RATE PER 1000.0 CRITICAL t5tS YEAR 1985 REACTOR TRIP RATES VS CRITICAL HOURS BY PLANT POWER GREATER THAN 15X 4

18000{

P L

75008 +*I

++

N

+i+++

T

+

++ +

4

• 14% t ++ 4 C

R 3

+

+McGuire 2 fseeel*,+,+[++.

[

]

• LaSalle 2 h

+#

2

+WPPSS 2 H

hiem,Susquehanna2 25eaj+

+

1 y

y

+ Diablo Canyon 1 4 Grand Gulf + Callaway 1

]

y j$

+

s M*

0 0

t i

2 2

3 3

4 4

5 5

6 5

5 i

i i

5 5

5 i

5 i

i i

PLANT TRIP RATE PER 1000.8 CRITICAL 15tS YEAlt 1984 Figure 2.1-22

39 Table 2.1-1 Plants With Relatively Poor Trip Experience in 1985 Name Trips Above Critical Trip Rate Per Mean Critical Hours 15% Power Hours 1000 Critical Hours Between Trips (Power > 15%)

(Power > 15%)

Diablo Canyon 2 10 1874.2 5.34 187.4 Waterford 3 15 3343.0 4.49 222.9 Byron 1 15 4656.4 3.22 310.4 Palo Verde 1 9

2888.0 3.12 320.9 Grand Gulf 1 14 5092.1 2.75 363.7 Catawba 8

3612.4 2.21 451.6 Callaway 1 17 8161.0 2.08 480.1 Wolf Creek 9

4471.7 2.01 496.9 Table 2.1-2 Plants With Relatively Poor Trip i

Experience in 1984 Trips Above Critical Trip Rate Per

-- Mean Critical Hours Name 15% Power Hours 1000 Critical Hours Between Trips (Power > 15%)

(Power > 15%)

WPPSS 2 17 2983.0 5.70 175.5 Callaway 1 6

1131.5 5.30 188.6 Grand Gulf 1 4

1010.0 3.96 252.5 Susquehanna 2 7

2145.9 3.26 306.6 Salem 1 7

2672.3 2.62 381.8 McGuire 2 16 6138.3 2.61 383.6 Salem 2 8

3386.0 2.36 423.3 Hatch 2 7

3108.7 2.25 444.1 Diablo Canyon 1 2

967.1 2.07 483.6 LaSalle 2 9

4469.8 2.01 496.6

40 2.2 Reactor Trips at Power Levels Below 15 Percent Reactor trips from power levels below 15 percent are considerably less significant than trips from higher power levels because they generally occur during start-up when the residual decay heat is at a minimum. Consequently, discussion of these trips is abbreviated in comparison to the discussion of reactor trips from above 15 percent power. Also, as noted previously, plants spend very little time in this power regime, thus we have not calculated trip rates per reactor critical hour below 15 percent power.

There were 157 reactor trips in 1984 and 138 in 1985 at power levels less than or equal to 15 percent power.

Figure 2.2-1 shows the distribution of tnese trips by power level.

2.2.1 Initiating Systems As in 1984, the Reactor Protection System (RPS) continued to be the primary initiator of reactor trips from low power.

However, most of the these trips occurred below 2 percent power as snown by Figure 2.2-2 and 2.2-3.

These figures also show that the decrease in trips below 15 percent power from 1984 to 1985 came from a decrease in trips below 2 percent power, with the RPS and Main Steam showing the greatest reduction.

The Feedwater, Control Rod Drive, Turbine and Electrical systems contributed significantly to low power trips in both 1984 and 1985; however, the contribu-tion from Main Steam and Condensate declined. sat power levels between 2 percent and 15 percent,. the Feedwater System was the primary contributor.

2.2.2 Causes Figure 2.2-4 provides an overall summary of causes.

In a manner similar to 1984, personnel related problems (i.e., human error, manual feedwater level control problem, and pro'cedure deficiencies) accounted for a greater share of the reactor trips at low power than at high power in 1985, 48 percent vs.

32 percent.

Figure 2.2-5 shows the distribution of personnel types responsible for the 48 trips due to human error in 1985, split into the power ranges where RPS or l

)

41 REACTOR TRIPS AT OR BELOW 15% POWER IN 1985 MIDPOINT POWER FREQ CUM.

PERCENT CUM.

FREQ PERCENT l

0 54 54 39.13 39.13 2

18 72 13.04 52.17 4

11 83 7.97 00.14 6

10 93 7.25 67.39 8

4 07 2.90 70.29 le 15 112 10.87 81.16 l

12 11 123 7.97 89.13 l

14 15 138 10.87 100.00

...... ' r = ""T" 0

10 20 30 40 50 60 FREQUENCY NOTE POWER = 0 MEANS 0 THROUGH IX REACTOR TRIPS AT OR BELOW 15% POWER IN 1G84 MIDPOINT POWER FREQ CUM.

PERCENT CUM.

FREQ PERCENT 0

54 54 34.39 34.39 2

36 90 22.93 57.32 4

13 103 8.28 65.61 6

8 11I 5.10 70.70 8

10

! ?. I 8.37 77.07 10 17 138 10.83 87.90 12 6

144 3.82 91.72 14 13 157 8.28 100.00 0

10 20 30 40 50 60 FREQUENCY NOTE POWER- 0 MEANS 0 THROUGH 1X Figure 2.2-1 l

,I 42 INITIATING SYSTEMS

SUMMARY

POWER GREATER THAN OX AND LESS THAN OR EQUAL 2X IN 1985 SYSTEN FREQ CUM. PERCENT CUN.

FRE0 PERCENT RPS 29 29 43.28 43.28 CTRL ROD DRIVE 14 43 20.90 04.18 wfEEDWATER 7

50 10.45 74.63 OTHER 5

55 7.46 82.09 l

ELECTRICAL 5

60 7.46 89.55 wTURSINE 3

63 4.48 94.03 l

wMAIN STEAM 2

65 2.99 97.01

• CONDENSATE 2

67 2.99 100.00

....3..

.3..

0 10 20 30 40 50 j

FREQUENCY

• POWER CONVERSION SYSTEMS (PCS)

INITIATING SYSTEMS

SUMMARY

POWER GREATER THAN OX AND LESS THAN OR EQUAL 2%

IN 1984 SYSTEM FREQ CtM. PERCENT CUN.

FREQ PERCENT RPS 38 38 42.22 42.22 OTMER 12 50 13.33 55.56 M

= MAIN STEAM 10 Se 11.11 06.67 h

CTRL ROD DRIVE 9

Se 10.0e 76.67

• FEEDWATER 8

77 8.89 85.56 ELECTRICAL 7

84 7.78 S3.33

• TURBINE 4

88 4.44 97.78 mCONDENSATE 2

90 2.22,t90.90 0

10 20 30 40 50 FREQUENCY POWER CONVERSION SYSTEMS CPCS)

=

Figure 2.2-2

43 INITIATING SYSTEMS

SUMMARY

l POWER GREATER THAN 2X AND LESS THAN OR EQUAL 15X i

IN t986 SYSTEN FREQ CUM.

PERCENT CUM.

FREQ PERCENT

• FEEDWATER 29 29 40.86 40.86 RPS 13 42 18.31 59.t5 mTURBINE 12 54 10.G0 70.06 CTRL ROD DRIVE 6

68 8.45 64.51 OTHER 3

63 4.23 88.73 F.LECTRICAL 3

66 4.23 92.96 l

2 68 2.82 95.77 AMAIN STEAM l

aCONDENSATE 2

70 2.82 98.59 AMAIN GENERATCR 1

71 1.41 180.08 O

10 20 30 40 FREQUENCY

= POWER CONVERSION SYSTEMS <P';D INITIATING SYSTEMS

SUMMARY

POWER GRZATER THAN 2X AND LESS THAN OR EQUAL 154 IN 1984 SYSTEM FREO CUM.

PERCENT CLN.

FREQ PEP. CENT mFEEDWATER 27 27 40.38 40.30 aTLRBINE 12 30 17.01 58.21

• MAIN STEAM G

48 13.43 71.e4 RPS 8

56 11.94 83.58 OTTER 4

60 5.07 89.55 CTRL ROD DRIVE 3

63 4.48 94.93

• CONDENSATE 3

66 4.48 99.51 mMAIN GENERATOR 1

87 1.46 190.00 g..

..g..

.g..

..g..

..g.

O 10 29 38 40 FREQUENCY a POWER CONVERSION SYSTEMS <PCS)

Figure 2.2-3

)

44 CAUSE

SUMMARY

POWER LESS THAN 15%

1985 CAUSE FREO CUN.

PERCENT CUH.

FREQ PERCENT HARDWARE S2 52 37.68 37.08 HUMAN ERROR 40 100 34.78 72.46 PROCEDURES 11 111 7.97 80.43 UNKNOWN 9

120 6.52 86.96 SG LEVEL 8

128 5.80 92.75 SYSTEM DESIGN 0

134 4.35 97.10 ENVIRONMENTAL 2

136 I.45 98.55 HUMAN HARDWARE 2

138 1.45 100.00

..i.

0 20 4n 60 80 FREQUENCY CAUSE

SUMMARY

POWER LESS THAN 15%

1984 CAUSE FREQ CUM.

PERCENT CtM.

FREQ PERCENT HARDWARE 70 79 44.59 44.59 HUMAN ERROR 47

!!7 29.94 74.52 l

I PROCEDURES 13 130 8.28 82.80 SG LEVEL 12 142 7.84 90.45 UNKNOWN 8

150 5.I8 95.54 HUMAN MARDWARE 4

154 2.55 98.99 f

NOT PROVIDED 3

157 1.91 190.00 e

20 4e e0 80 FREQUENCY Figure 2.2-4

45 HUMAN ERROR - TYPE OF PERSONNEL POWER GRFATER THAN O AND LESS THAN 2X YEAR 1985 PERSON FREQ CUM.

PERCENT CUM.

FREQ PERCENT LIC OPERATOR 12 12 48.00 48.00 TECHNICIAN 8

18 24.00 72.00 STAFF OPS 3

21 12.00 84.00 CONTRACTOR 1

22 4.00 88.00 VENDOR STAFF 1

23 4.00 92.00 LIC MANAGEMENT 1

24 4.08 96.00 ENG STAFF 1

25 4.00 100.00

....mi......

0 5

10 15 20 FREQUENCY HUMAN ERROR - TYPE OF PERSONNEL POWER GREATER THAN 2 AND LESS THAN 154 YEAR 1966 PERSON FREQ CUM.

PERCENT CUM.

F.iE0 PERCENT LIC OPERATOR 15 15 65.22 65.22 NOT DETERMINED 4

19 17.39 82.0t STAFF OPS I

20 4.35 86.96 i

ENG STAFF 1

21 4.35 91.30 TECHNICIAN 1

22 4.35 95.85 UNLIC OPERATOR 1

23 4.35 100.00 0

5 le 15 20 FREQUENCY Figure 2.2-5

45 feedwater dominate. Licensed operators had the largest contribution in both ranges.

For the entire power range less than or equal to 15 percent power, inadequate trainiiig is the leading root cause of human error.

Figure 2.2-6 provides a summary of root causes for human error for the two low power intervals evaluated.

2.2.3 Contribution from Maintenance, Trcubleshooting, Calibration and Testing

)

Figures 2.2-7 and 2.2-8 show the impact of maintenance, troubleshooting, calibration and testing on reactor trip statistics below 15 percent power.

In 1984, the contribution from these activities was 35 percent, essentially the sane as above 15 percent power.

In 1985, 30 percent of the trips below 15 l

percent power were caused by these activities vs. 40 percent the trips above 15 percent power. Thus in 1985 these activities collectively have become less significant (i.e., a smaller percentage of an already smaller total number of trips below 15 percent power) than in 1984.

Figure 2.2-8 provides information for the systems impacted by the activities. The Main Steam, RPS and Turbine Systems are most responsible for the decrease.

I i

e

47 PERSONNEL ROOT CAUSES IEADING TO REACTOR TRIPS POWER GREATER THAN O APO LESS THAN 2X YEAR 1986 ACTION FREQ CUM. PERCENT CLPl.

FREQ PERCENT COGNITIVE 7

7 28.00 D3. 00 ADMIN CTRL 5

12 20.00 48.00 PRDC NOT USED 4

to 16.00 64 40 TRAINING 3

19 12.00 70.00 INATTENTION 3

22 12.00 88.00 OTHER I

23 4.00 92.00 TEST EQUIP USE 1

24 4.00 06.00 COMMUNICATION 1

25 4.00 109.90 O

2 4

6 8

10 FREQUENCY i

PERSONNEL ROOT CAUSES LEADING TO REACTOR TRIPS' POWER GREATER THAN 2 AND LESS THAN 15%

YEAR 1985 j

ACTION FREQ CUM. PERCENT CUM.

FREQ PERCENT TRAINING 8

8 34.78 34.78 INATTENTION 5

13 21.74 56.52 CCit1UNICATION 4

17 17.39 73.91 i

OTHER 2

19 8.70 82.81 PLANNING 2

21 8.70 91.30 ADMIN CTRL 1

22 4.35 95.65 PROC NOT USED 1

23 4.35 100.80

,................. rrn, 0

2 4

6 8

18 FRE0VENCY Figure 2.2-6

.--.\\

48 ACTIVITY AT TIME OF REACTOR TRIPS POWER LESS THAN OR EQUAL TO 15X IN 1985 ACTIVITY FREO CUH.

PERCENT CLN.

FREQ PERCENT TRANSITION 89 89 64.49 64.49 TESTING 23 112 18.67 81.18 TROUPLESH30 TING G

121 8.52 87.68 HAINTENANCE 8

129 5.80 93.48 UNDEFINED 7

138 5.07 98.55 CALIBRATION 1

137 0.72 99.28 l

STEADY STATE 1

138 0.72 100.00 0

20 40 60 80 100 FREQUENCY s

ACTIVITY AT TIME OF REACTOR TRIPS POWER LESS THAN OR EQUAL TO 15X IN 1964 ACTIVITY FREQ CUM. PERCENT CUM.

FREQ PERCENT TRANSITION 80 86 54.78 54.78 TESTING 37 123 23.57 78.34

)

UNDEFINED 16 139 10.19 88.54 TROUBLESHOOTING 9

148 5.73 94.27 MAINTENANCE 8

156 5.10 99.33 CALIBRATION 1

157 0.84 100.08 W

W T

w T

W T

y y

y 0

20 40 Se 80 100 FREQUENCY Figure 2.2-7

49 INITIATING SYSTEMS AFFECTED.BY MAINTENANCE TROUBLESHOOTING CALIBRATION & TESTING POWER LESS THAN OR EQUAL TO 16X IN 1986 SYSTEM FREQ CUN.

PERCENT CUM.

FREQ PERCENT RPS 16 15 36.50 36.50

• FEEDWATER r

24 21.95 58.54 5

29 12.20 10.73 OTHER h

4 33 9.78 89.49 CTRL R0D DRIVE

• TURBINE 4

37 9.76 90.24 ELECTRICAL 3

40 7.32 97.56 aCCNDENSATE I

41 2.44 100.00 e

6 to 16 28 25 FREQUENCY LEGEND ACTIVITY musumu TESTING mnm TROUBLESHOOTING axumooc CALIBRATION rrrr1 MAINTENANCC INITIATING SYSTEMS AFFECTED BY MAINTENANCE TROWLESH00 TING CALIBRATION & TESTING POWER LESS THAN OR EQUAL TO 15%

IN 1984 SYSTEM FREQ CUN. PERCENT CUM.

FREQ PERCENT RPS 20 20 36.36 36.36

• MAIN STEAM 6

28 14.55 50.01

• TURBINE E

8 36 14.55 65.45 CTRL ROD DRIVE E

6 42 10.91 76.36 l

OTHER 5

47 9.09 85.45

• FEEDWATER E

5 52 9.09 94.55 ELECTRICAL 2

54 3.64 98.18 aCONDENSATE I

55 1.82 100.00 0

5 10 15 20 25 FREQUENCY I

LEGEND: ACTIVITY M TESTING txxrX1 TROUBLESHOOTING l

Entnam CALIBRATION rYuri MAINTENANCE Figure 2.2-8

50 3.0 REACTOR TRIPS WITH ASSOCIATED FAILURES As previously noted, unplanned reactor trips where the recovery was complicated

~by additional equipment failures or personnel errors can be of concern because of the higher level of stress and demands placed upon the operating personnel. This section addresses this issue.

" Associated failures" are defined in this report as component failures or personnel errors that did not contribute directly to the cause of a reactor trip, but are associated with post-trip activities (e.g., normally the failure was discovered or occurred when the ccmponent was actuated to mitigate the consequences of the trip). Examples of associated hardware failures included failure of AFW pumps and valves, failure of main feedwater regulating valves, failure of electrical systems, failure of steam relief and safety valves, and failure of instranentation.

In 1984 about 20 percent of all trips above 15 percent power included one or more associated failures, and over 25 percent of all reactor trips at power levels of 95 percent or higher included at least one associated failure.

In total, 77 reported trips indicated associated failures. The LERs that included associated failures reported a total of 123 separate failures. Of the 77 trips that involved associated failures, over one-third (27) included multiple associated failures.

In 1985, 23 percent, or roughly the some percentage as 1984, of scrams above

,15 rercent power included associated failures. About 25 percent of all reactor trips at power levels of 95 percent or higher included at least one associated failure. A total of 109 trips comprised a total of 139 separate failures.

About 22 percent involved multiple associated failures. The 109 reactor trips and their associated failures are tabulated in Appendix D.

Associated failures are a concern because they complicate the recovery from the trip, and increase the probability that the recovery from the trip will not proceed satisfactorily. However, within the set of trips with associated failures there is a wide range of immediate significance. Section 4.0 discusses

~

an approach which allows further discrimination with regard to safety significance.

b

51 4.0 QUANTITATIVE SAFETY SIGNIFICANCE MEASURES The oreceding discussion relics on several qualitative assumptions and generalizations concerning the safety significance of a reactor scram.

Specifically, that reactor scram frequency should be minimized; that scram significance is positively correlated with the reactor power level immediately preceding the scram; and that additional failures and personnel errors following the scram add to the safety significance of the trip.

This section describes an approach to obtaining a quantitative measure of the safety significance of an event sequence (transient) involving a reactor scram.

,The specialized probabilistic event tree technique called " Accident Sequence Precursor (ASP)" analysis was used to review the 1985 reactor scram sequences as documented in LERs. Based on a reading of the LER, specific sequences were chosen for more detailed analysis (i.e., supplementing the LER material with FSAR information, system design drawings, etc.) if the scram sequence included one or more of tne following:

Any failure to function of a system that should have functioned as a con-sequence of an off-normal event or accident; Any support system failures, including single component failures in cooling water systems, instrument air and electrical poser systems; Any event or o;erating condition that was not enveloped by or proceeded differently from the plant design basis; Any instance where two or more failures occurred; or Any event that, based on the reviewer's experience, could have resulted in or significantly affected a chain of events leading to potential severe core damage.

D e

52 Thirty-two scram sequences were selected for quantification after this more detailed qualitative review.

The objective of the quantification is to produce a measure of the conditional risk of core damage or core vulnerabilty given the observed sequence.

To do this we overlay the observed sequence on event trees which provide a reasonable set of alternate paths which the sequence might have taken.

An example of such an event tree is shown in Figure 3.0-1.

The end state of a path through an event tree is labeled " core damage" if the path has been formally analyzed and comprises less than proper operation of a minimum set of caponents.

The end state is labeled " core vulnerability" if core protection is believed to be provided but no analytic basis is generally available to directly support that conclusion.

A basic " transient" tree such as Figure 3.0-1 was developed for each of a number of. sub-classes of light water reactors in order to more accurately portray the general responses (paths) for that class.

If the observed scram sequence involved loss of main feedwater, loss of offsite power or loss of coolant

\\

(LOCA), more specialized trees were used.

Finally, further plant specificity was achieved by manipulating the conditional branch probabilities used.

For' example, the likelihood of feedwater system operability following a BWR small LOCA (e.g., a safety relief valve spuriously opening) considered whether the feed pumps at the plant were steam or motor driven.

As stated above, a conditional probability of core damage or core vulnerability was calculated for each of the 32 selected scrams.

In setting the sequence branch'probabilties, this calculation assumed that the failure probabilities for systems that failed during the event were equal to the like-lihood of failing to recover from the observed state.

Failure probabilities for systems observed degraded during the sequence were set equal to the conditional probability that the system would fail and not be recovered within a 20- to 30-minute period.

The failure probability associated with observed successes and the systems unchallenged during an actual sequence were assumed equal or fault tree models consisting of typical (i.e., generic) train and common mode failure probabilities.

4

e 53 SE 988 go, spe 4M

,803.F

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3,,,aEL

.35 y,

Out W

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E D

WM TE8 g

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=

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em vets 1

? M Sur.d4 I

i e em gar.se I

, em ear.se en

.m a

d

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is amt==

4 W

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- 3 a

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r.a as wet terudit a

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=

M due tarust e,

ens mm te giu aur.se 1

i 3.

,==

sa arws c

o.use.c..suerraus ar m a., ass e Figure 3.0-1 Sample Event Tree

b 54 The conditional probabilities of severe core damage for the 32 sequences range from 1.1E-2 [for a loss of feedwater (LOFW) and AFW failure] to 1.7E-6, with many events in the IE-6 to IE-4 range. Because of the uncertainties inherent in the calculationr, it is not appropriate to rank events on an event-by-event basis.

Instead, events can be grouped into conditional probability ranges to identify the more significant events. On an order of magnitude basis, the following numbers of events were identified:

Conditional Probabilities of Severe Core Damace Events IE-1 to' IE-0

- No events IE-2 to IE-1

- LOFW and AFW Failure at Davis-Besse (1.1E-2)*

1E-3 to IE-2

- MSIV Closure and Subsequent SDV Isolation Problems at Oyster Creek

- Stuck Open Relief Valve and HPCI/RCIC Unavailability at Hatch 1 1E-4 to IE-3

- Effective LOOP and AFW System Unavailability at San Onofre 1*

- AFW Pumps Failure on Demand at Trojan

- Reactor Trip, Loss of Feedwater and AFW Train Failure at Davis-Besse

- LOFW and RCIC Trip at Hatch 2

- Loss of Circulating Water and Non-Safety Service Water Due to Expansion Joint Failure at LaSalle 1

- LOFW plus HPCS Failure at Grand Gulf 1 1E-5 to IE-4

- 10 events including Loss of ICS at Rancho Seco*

(1.5E-5) 1E-6 to IE-5

- 13 events The three events above which are identified by an asterisk were investigated by NRC Incident Investigation Teams (IITs).

To provide perspective on the conditional probabilities for 'these events, and to isolate the safety significance of the scram function per se, the l

55 conditional probabilities of a spurious scram (i.e., the plant state does not require the scram) with no subsequent complications and a scram due to loss of main feedwater were calculated for each plant class. A " simple" reactor trip with no complications has an associated conditional probability estimated in the IE-7 to low IE-5 range.

Losses of feedwater without further complications have associated core damage probabilities in the IE-6 to high IE-5 range.

In most cases, the conditional probability associated with loss of feedwater is not substantially greater than that associated with a spurious trip at the same plant, primarily because the likelihood of loss of feedwater during a reactor trip is quite high.

Based on these estimates, it can be concluded that from a core damage probability standpoint, scram sequences with conditional probabilities below IE-4 are not substantially more significant than

" simple" trips and losses of feedwater.

Thus, using this approach to quantifying safety significance results in grouping the vast majority of reactor trips, including those with associated failures, into a conditional probability range of IE-7'to IE-5.

The loss of ICS at Rancho Seco, which was the subject of an IIT, serves as a useful calibration point for the upper end of the range.

The San Onofre 1 IIT event is rated one to two orders of magnitude more significant than the majority of reector trips, and the Davis-Besse IIT event still tvec orders of magnitude higher. The quantitative results appear consistent with engineering judgement based on the rough ordering by safety significance of IIT events, and the quantitative methodology should prove useful in gauging the trend in relative safety significance of the reactor trip sequences over time.

00 5.0

SUMMARY

OF FfNDINGS AND CONCLUSIONS f

Based on an analysis of the reactor protection system actuations that occurred in 1985, the following conclusions can be made:

1.

Of the 719 unplanned RPS actuations at U.S. Light Water Reactors (LWRs) in 1985, we identified 552 reactor trips (i.e., resulted in control rod nation).

2.

The scran rate for 1984 and 1985 for the industry was approximately one scram per 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of critical operation.

In terms of the overall performance of the industry we observed a slight change in the total trip rate from 1984 to 1985, (i.e., 5.9 and 6.0 trips per reactor year, respectively). While overall changes were small, significant shifts took place at the NSSS specific level.

3.

In both years the majority of the reactor trips occurred with the reactor power above 15 percent: 68 percent in 1984, and 75 percent in 1985.

We further observed that 31 percent of all trips in 1984 and 38 percent of all trips in 1985 occurred while the plant was at 95 percent power or above.

5 We believe this reflects both the minor contribution of startup problems to scram frequency and the very short time spent operating in the lower power regime.

4.

Above 15 percent power, the feedwater system was the single system most responsible for reactor trips, contributing approximately 27 percent in 1984 and 24 percent in 1985.

5.

Above 15 percent power, the major balance-of-plants systems, collectively referred to in this report as Power Conversion Systems (PCS) (i.e.,

feedwater, turbine, condensate, main steam, and main generator),

contributed 59 percent and 52 percent of all reactor trips above 15 percent power in 1984 and 1985, respectively.

4 t

s'

57 6.

The Electrical Distribution and Reactor Protection System were the major non-PCS contributors in both 1984 and 1985; in these years they con-tributed 23 percent and 29 percent, respectively, of trips above 15 percent power.

7.

Haraware failures were the dominant cause of unplanned reactor trips above 15 percent power.

In 1984 and 1985 hardware failures contributed 60 per-cent and 55 percent respectively, of all trips above 15 percent power.

The Feedwater Regulating Valve, Main Steam Isolation Valve, Turbine Control Valve, and Turbine Stop Valves were major contributors. These valves were responsible for 8 percent of all trips above 15 percent power in 1985.

8.

Personnel related problems (i.e., human error, manual steam generator level ~ control problems and procedure deficiencies) accounted for 28 per-cent of all reactor trips above 15 percent power in 1984 and 32 percent in 1985, making them a substantial but secondary cause of reactor trips.

For.1984 and 1985 unlicensed personnel were responsible for 10 percent and 14 percent, respectively, of all trips above 15 percent power, with unlicensed technicians involved in roughly one of every 12 trips.

9.

Non-cognitive errors due to training deficiencies, inattention, poor communication, and failure to follow procedures occurred at nearly twice the frequency of cognitive errors for licensed operators at power greater than 15 percent in 1985. Cognitive errors by technicians occurred at approximately the same frequency as the total error count due to training, inattention, poor communication, and not following procedures.

10.

In 1985 maintenance, troubleshooting, calibration and testing were found to be causal factors in 40 percent of the reactor trips above 15 percent power: 25 percent from maintenance, troubleshooting and calibration, and 15 percent from testing.

For 1984, the corresponding percentages were:

36 percent overall, 13~ percent for maintenance, troubleshooting and calibration, and 23 percent for testing. Thus, these activities constitute a substantial contribution to the overall reactor trip frequency.

11. At power levels above 15 percent a trip frequency of 2.0 trips per 1000 critical hours was selected as a breakpoint for examining relatively

58 poor performance (this level roughly corresponds to 10-11 scrams per reactor year).

In 1984 ten plants exhibited rates at or above the cutoff, with a maximum rate of 5.7 trips per 1000 critical hours. Five of these plants had initial criticality in 1984.

In 1985, a total of eight plants exceeded the criterion and six of these had initial criticality in 1985. The maximum trip frequency above 15 percent power for 1985 was 5.3 trips per 1000 critical hour. Although showing large decreases from 1984 to 1985, Callaway 1 and Grand Gulf 1 are the only two plants that afe.above the cutoff for 1984 and 1985.

12. For both 1984 and 1985 the Reactor Protection System was the primary initiating system for reactor trips from power levels less than or equal to 15 percent.

13. A decrease in trips below 15 percent powtr from 1984 to 1985 came from a decrease in trips below 2 percent power, with the RPS and Main Steam System showing the greatest reduction.

14. Hardware failures were major contributo.rs to trips below 15 percent power for both 1984 and 1985. However, unlike 1984 the differential between hardware and human error initiated trips was much smaller in 1985.

Similar to 1984, personnel related problems (i.e., human error, manual feedwater level control problem, and procedure deficiencies) accounted for a greater share of the reactor trips at low power than s

at high power in 1985, 48 percent vs. 32 percent.

Licensed operators were the dominant persons responsible for trips at power levels below 15 percent.

15.

In 1984 the contribution from maintenance, troubleshooting, calibration and testing belcw 15 percent poier was 35 percent, essentially the same as above 15 percent.

In 1985, 30 percent of the trips below 15 percent power were caused by these activities vs. 40 percent of the trips above 15 percent power. Thus in 1985 these activities collectively have become less significant (i.e., a smaller percentage of an already smaller total number of trips below 15 percent power) than in 1984. The reductions for the Main Steam, RP5 and Turbine Systems are most responsible for the decrease.

59 16.

In 1985, 23 percent, or roughly the same percentage as 1984, of scrams above 15 percent power included associated failures (i.e., additional failure or personnel error not directly related to the initiator of the scram). About 25 percent of all reactor trips at power levels of 95 percent or higher included at least one associated failure. A total of 109 trips comprised a total of 139 separate failures. About 22 percent of these scrams involved multiple associated failures.

17. The Accident Sequence Precursor (ASP) approach to quantifying safety significance results in grouping the vast majority of reactor trips, including those with associated failures, into a conditional probability range of IC-7 to IE-5.

The loss of ICS at Rancho Seco, which was the subject of an IIT, serves as a useful calibration point for the upper end of the range. The San Onofre 1 IIT event is rated one to two orders of magnitude more significant than the majority of reactor trips, and the Davis-Besse IIT event still two orders of magnitude higher. The quantitative results appear consistent with engineering judgement based on the rough ordering by safety significance of IIT events. The quantitative methodology should prove ustful in gauging the trend in relative safety significance of the reactor trips over time, i

APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE OBS NAME DOCKET LER MONTH DAY VENDOR TRIPS POWER 1

YANKEE ROWE 29 2

28 W

AUTO O

2 YANKEE ROWE 29 9

12 9

W AUTO O

3 BIG ROCK POINT 155 1

1 5

GE AUTO O

4 BIG ROCK POINT 155 2

1 7

GE MANUAL 12 5

BIG ROCK POINT 155 8

11 14 GE AUTO O

6 BIG ROCK POINT 155 9

12 7

GE MANUAL 15 7

SAN ONOFRE 1 206 14 9

19 W

AUTO 93 8

SAN ONOFRE 1 206 17 11 21 W

MANUAL 60 9

HADDAM NECK 213 6

3 16 W

MANUAL 100 10 HADDAM NECK 213 7

3 12 W

AUTO 50 11 HADDAM NECK 213 11 5

16 W

MANUAL 100 12 HADDAM NECK 213 28 11 10 W

AUTO 100 13 HADDAM NECK 213 28 11 21 W

AUTO 97 14 OYSTER CREEK 219 6

2 24 CE AUTO 20 15 OYSTER CREEK 219 12 6

12 GE AUTO 100 16 OYSTER CREEK 219 15 7

8 GE AUTO 80 17 OYSTER CREEK 219 16 8

9 GE AUTO 8

18 OYSTEP CREEK 219 22 11 20 GE AUTO 77 19 OYSTEN CREEK 219 12 15 GE AUTO 100 20 NINE MILE POINT 1 220 3

3 4

GE AUTO 100 21 NINE M LE POINT 1 220 5

4 16 GE AUTO 100 22 NINE MILE POINT 1 220 14 8

19 GE At*TO 90 23 NINE MILE POINT 1 220 17 8

23 GE MANUAL 99 m

24 NINE MILE POINT 1 220 21 11 1

GE AUTO 98 ca 25 NINE %ILE POINT 1 220 22 11 7

GE AUTO O

26 NINE MILE POINT 1 220 22 11 7

GE AUTO O

27 DRESLE N 2 237 22 5

2 GE MANUAL 40 28 DRESOt.N 2 237 26 5

18 GE AUTO 84 29 DRESCEN 2 237 28 6

19 GE AUTO 1

30 DRESJEN 2 237 31 7

27 GE AUTO 57 31 DRESU.N 2 237 32 8

2 GE AUTO O

32 DRESCEN 2 237 34 8

16 GE AUTO 70 33 DRESDEh 2 237 35 9

29 GE AUTO 1

34 DRESDEN 2 237 41 11 12 GE AUTO 86 35 R.E.GINNA 244 6

4 6

W AUTO 5

36 R.E.GINNA 244 7

4 6

W AUTO 12 37 R.E.GINNA 244 8

4 7

W AUTO 13 38 R.E.GINNA 244 9

4 8

W AUTO 2

39 R.E.GINNA 244 11 4

11 W

AUTO 7

40 R.E.GINNA 244 14 6

6 W

AUTO 100 41 R.E.GINNA 244 18

.9 28 W

MANUAL 50 42 R.E.GINNA 244 19 11 25 W

AUTO 100 43 MILLSTONE 1 245 9

8 13 GE AUTO 100 44 MILLSTONE 1 245 17 9

27 GE AUTO 15 45 MILLSTONE 1 245 20 10 7

GE AUTO 97 46 INDIAN POINT 2 247 1

2 2

W MANUAL 100 47 INDIAN POINT 2 247 2

2 4

W AUTO 100 48 INDIAN POINT 2 247 4

3 6

W AUTO 100 49 INDIAN POINT 2 247 5

3 16 W

MANUAL 25 50 INDIAN POINT 2 247 6

4 16 W

AUTO 100 51 INDIAN POINT 2 747 9

9 20 W

AUTO 95

\\

~

APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE OSS NAME DOCKET LER MONTH DAY VENDOR TRIPS POWER 52 INDIAN POINT 2 247 10 9

23 W

AUTO 12 53 INDIAN POINT 2 247 12 9

28 W

AUTO 24 54 INDIAN POINT 2 247 14 10 24 W

AUTO 13 12 31 W

AUTO 100 55 INDIAN POINT 2 247 12 12 W

AUTO 100 56 INDIAN POINT 2 247 16 57 DRESDEN 3 249 1

1 2

GE AUTO 85 58 DRESDEN 3 249 2

2 1

GE AUTO 99 59 DRESDEN 3 249 10 4

27 GE AUTO O

60 DRESDEN 3 249 18 9

19 GE AUTO 83 61 DRESDEN 3 249 23 10 17 GE AUTO 79 62 TURKEY POINT 3 250 3

1 19 W

AUTO 10 63 TURKEY POINT 3 250 4

1 29 W

AUTO 100 64 TURKEY POINT 3 250 18 7

16 W

AUTO O

65 TURKEY POINT 3 250 19 7

21 W

AUTO 100 66 TURKEY POINT 3 250 22 7

29 W

AUTO 100 67 TURKEY POINT 3 250 23 8

1 W

AUTO 30 68 TURKEY POINT 3 250 32 10 15 W

AUTO 100 69 TURKEY POINT 4 251 3

2 6

W AUTO 27 70 TURKEY POINT 4 251 4

2 7

W AUTO 100 71 TURLfY POINT 4 251 10 5

15 W

AUTO 100 72 TURKEY POINT 4 251 11 5

17 W

AUTO 100 73 TURKEY POINT 4 251 12 5

30 W

nUTO 100 74 TURKEY POINT 4 251 13 6

6 W

AUTO 100 os 75 TURKEY POINT 4 251 17 6

20 W

AUTO 100 76 TURKEY POINT 4 251 19 7

17 W

AUTO 100 77 TURKEY POINT 4 251 21 8

20 W

AUTO 100 78 TURKEY POINT 4 251 25 11 23 W

AUTO O

79 QUAD CUTIES 1 254 6

5 30 GE AUTO 100 80 QUAD CITIES 1 254 11 5

7 GE AUTO 90 81 QUAD CITIES 1 254 19 11 25 GE MANUAL 10 82 PALISADES 255 10 8

11 CE AUTO 98 83 PALISADES 255 16 8

30 CE AUTO 47 84 BROWNS FERRY 1 259 16 1

16 GE AUTO 100 85 BROWNS FERRY 1 259 8

3 19 GE MANUAL 90 86 H.B. ROBINSON 2 261 3

1 8

W AUTO O

87 H.B. ROBINSON 2 261 4

1 8

W AUTO O

88 H.B. ROBINSON 2 261 5

1 9

W AUTO 11 89 H.B. ROBINSON 2 261 9

2 27 W

AUTO 100 90 H.B. ROBINSON 2 261 10 3

1 W

AUTO 52 91 H.B. ROBINSON 2 261 11 3

8 W

AUTO 100 92 H.B.ROBINSCN 2 261 13 5

21 W

AUTO 100 93 H.B. ROBINSON 2 261 15 7

5 W

AUTO 12 94 H B. ROBINSON 2 261 16 8

10 W

AUTO 97 95 H.B. ROBINSON 2 261 17 8

22 W

AUTO 100 96 H.B. ROBINSON 2 261 20 9

11 W

AUTO 100 97 H.B. ROBINSON 2 261 21 9

17 W

AUTO 100 98 MONITCELLO 263 8

4 Il GE AUTO 100 99 MONTICELLO 263 10 6

2 GE AUTO 100 100 MCNTICELLO 263 13 7

29 GE AUTO 100 101 OUAD CITIES 2 265 1

1 25 GE AUTO 77 102 QUAD CITIES 2 265 3

1 16 GE AUTO 100

APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE OBS NAME DOCKET LER MONTH DAY VENDOR TRIPS POWEN 103 OUAD CITIES 2 265 3

1 16 GE AUTO 100 1

104 OUAD CITIES 2 265 5

2 19 GE AUTO 95 105 OUID CITIES 2 265 21 10 15 GE AUTO 100 106 POINT BEACH 1 266 3

6 26 W

AUTO 88 107 OCONEE 1 269 2

1 22 BW AUTO 100 108 OCONEE 1 269 5

4 11 BW AUTO 100 109 OCONEE 1 269 6

4 11 BW AUTO 17 110 OCONEE 1 269 7

4 25 BW AUTO 94 111 OCONEE 2 270 2

4 21 BW AUTO O

112 OCONEE 2 270 4

4 22 BW AUTO 29 113 OCONEE 2 270 5

4 26 BW AUTO 75 114 OCONEE 2 270 6

7 11 BW AUTO 94 115 VERMO*ll YANMEE 271 4

2 6

GE AUTO 100 116 SALEM i 272 12 10 6

W AUTO 99 117 DIABLO CANYON 1 275 1

1 2

W AUTO 49 118 DIABLO CANYON 1 275 2

1 4

W AUTO 10 119 DIABLO CANYON 1 275 9

2 13 W

AUTO 75 120 DIABLO CANYON 1 275 11 2

17 W

MANUAL 48 121 DIABLO CANYON 1 275 12 3

21 W

AUTO 100 122 DIABLO CANYON 1 275 14 5

18 W

AUTO 100 123 DIAB 1.0 CANYON 1 275 15 5

20 W

AUTO 55 124 DIABLO CANYON 1 275 16 5

20 W

AUTO O

125 DIABLO CANYON 1 275 30 8

27 W

AUTO 100 cn 4

PO 126 DIA3.0 CANvCN 1 275 33 10 25 W

AUTO 100 127 PEACn BOTTOM 2 277 7

22 GE MANUAL 25 128 PEACH BOTTOM 2 277 11 8

5 GE MANUAL 100 129 PEACH ROTTOM 2 277 12 8

7 GE AUTO 2

130 PEACH BOTTOM 2 27/

16 8

26 GE AUTO 5

131 PEACM BOTTOM 2 277 22 10 17 GE AUTO 100 132 PEACH BOTTOM 2 277 25 11 29 GE AUTO 31 133 PEACH BOTTOM 2 277 12 26 GE AUTO 44 134 PEACH BOTTOM 3 278 7

3 1

GE AUTO 30 135 SURRY 1 280 2

1 13 W

AUTO 20 136 SURRY 1 280 3

1 26 W

AUTO 100 137 SURRY 1 280 4

1 27 W

AUTO 10 138 SURRY 1 280 6

1 28 W

AUTO 15 139 SURRY 1 280 7

4 29 W

AUTO 17 140 SURRY 1 280 18 9

8 W

AUTO 100 141 SURRY 2 281 13 11 7

W AUTO 22 142 PRAIRIE ISLAND 1 282 9

5 8

W AUTO 100 143 PRAIRIE ISLAND 1 282 10 5

9 W

AUTO 5

144 PRAIRIE ISLAND 1 282 12 9

15 W

AUTO 100 145 INDIAN POINT 3 286 1

1 21 W

AUTO 100 146 INDIAN POINT 3 286 2

1 23 W

AUTO 15 147 INDIAN POINT 3 286 2

1 23 W

AUTO 29 148 INDIAN POINT 3 286 3

3 19 W

MANUAL 85 149 INDIAN POINT 3 286 5

6 7

W MANUAL 55 150 INDIAN POINT 3 286 6

10 4

W AUTO 20 151 INDIAN POINT 3 286 7

10 15 W

AUTO 30 152 INDIAN POINT 3 286 12 11 29 W

AUTO 100 153 INDIAN POINT 3 286 12 11 30 W

AUTO O

o s

+~

APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE OBS NAME DOCKET LER MONTH DAY VENOOR TRIPS POWER 154 INDIAN POINT 3 286 13 11 30 W

AUTO 13 155 OCONEE 3 287 1

7 23 BW AUTO 74 156 OCONEE 3 287 5

10 24 BW AUTO 10 157 THREE MILE IS. 1 289 2

9 7

BW MAP"1AL 0

158 THREE MILE IS. I 289 3

12 1

CE AUTO 75 159 PILGRTM 293 6

3 15 GE AUTO 100 160 PILGRIM 293 9

4 4

GE AUTO 85 161 PILGRIM 293 14 6

15 GE AUTO 10 162 PILGRIM 293 25 9

1 GE AUTO 37 163 ZION 1 295 5

1 21 W

AUTO 20 164 ZION 1 295 24 6

29 W

AUTO 1

165 ZION 1 295 25 6

27 W

MANUAL 2

166 ZION 1 295 44 12 6

W AUTO O

167 BROWNS FERRY 3 296 2

1 7

GE MANUAL 100 168 BROWNS FERRY 3 296 8

3 9

GE MANUAL 0

169 COOPER 298 10 10 5

GE MANUAL 44 170 POINT BEACH 2 301 12 31 W

AUTO 100 171 CRYSTAL RIVER 3 302 15 8

20 BW AUTO 20 172 CRYSTAL RIVER 3 302 16 8

20 BW AUTO 20 173 CRYSTAL RIVER 3 302 20 10 9

BW MANUAL 96 174 CRYSTAL RIVER 3 302 23 10 26 BW MANUAL 95 175 CRYSTAL RIVER 3 302 25 10 29 BW AUTO 96 m

176 CRYSTAL RIVER 3 302 26 11 22 BW AUTO.

22 177 CRYSTAL RIVER 3 302 25 12 3

BW AUTO 93 178 CRYSTAL RIVER 3 302 30 12 7

BW AUTO O

179 ZION 2 304 8

3 17 W

AUTO 99 180 KEWAUNEE 305 3

2 8

W AUTO 5

181 KEWAUNEE 305 10 4

5 W

AUTO O

182 KEWAUKEE 305 17 8

8 W

MANUAL 73 183 KEWAUNEE 305 17 8

8 W

AUTO O

184 KEWAUNEE 305 20 11 13 W

AUTO 100 185 KAWAUNEE 305 23 12 12 W

AUTO 100 186 MAINE YANKEE 309 2

3 10 CE MANUAL 80 187 MAINE YANKE E 309 3

4 30 CE AUTO 96 188 MAINE YANKEE 309 4

5 5

CE AUTO O

189 MAINE YANKEE 309 7

7 1

CE AUTO 95 190 MAINE YANKEE 309 16 10 23 CE AUTO 4

191 MAINE YANKEE 309 17 10 25 CE AUTO 25 192 MAINE YANKEE 309 17 10 26 CE AUTO 30 193 MAINE YANKEE 309 18 11 6

CE AUTO 82 194 MAINE YANKEE 309 19 11 22 CE AUTO 69 195 SALEM 2 311 4

4 13 W

AUTO 25 196 SALEM 2 311 5

4 17 W

AUTO 18 197 SALEM'2 311 6

4 23 W

AUTO 54 198 SALEM 2 311 8

5 2

W AUTO 69 199 SALEM 2 311 9

5 2

W AUTG 20 200 SALEM 2 311 11 7

7 W

AUTO 33 201 SALEM 2 311 12 7

8 W

AUTO 10 202 SALEM 2 311 17 8

8 W

AUTO 100 203 SALEM 2 311 20 9

21 W

NANUAL 25 204 SALEM 2 311 22 10 7

W AUTO 100

APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE OBS NAME DOCKET LCR MONTH DAY VENDOR TRIPS POWER 205 RANCHO SECC 1 312 20 7

3 BW AUTO O

206 RANCHO SWCO 1 312 19 10 2

BW AUTO 14 207 RANCHO SEP3 1 312 12 26 BW AUTO 75 208 RANCHO SECO 1 312 23 12 5

CE AUTO 91 209 ARKANSAS 1 313 2

1 29 BW AUTO 30 210 ARKANSAS 1 313 3

4 9

BW AUTO 100 211 ARKANSAS 1 313 4

5 31 BW AUTO 100 212 ARKANSAS 1 313 5

6 2

BW AUTO 45 213 ARKANSAS 1 313 7

8 11 BW AUTO 98 214 ARKANSAS 1 013 8

8 19 BW MANUAL 88 215 ARKANSAS 1 313 10 9

2 BW AUTO 98 216 ARKANSAS 1 313 11 10 1

BW MANUAL 99 217 D.C. COOK 1 315 65 11 25 W

AUTO 78 218 D.C. COOK 2 316 2

1 12 W

AUTO 2

219 D.C. COOK 2 316 3

1 26 W

AUTO 96 220 D.C. COOK 2 316 35 10 29 W

AUTO 79 221 D.C. COOK 2 316 11 13 W

AUTO 81 222 CALVERT CLIFFS 1 317 2

2 5

CE AUTO 100 223 CALVERT CLIFFS 1 317 8

8 S

CE AUTO 17 224 CALVERT CLIFFS 1 317 9

8 6

CE AUTO 19 225 CALVERT CLIFFS 1 317 10 8

7 (E

AUTO 48 226 CALVERT CLIFFS 1 317 11 9

30 CE AUTO 100 227 CALVERT CLIFFS 1 317 12 10 2

CE AUTO 100 m

228 CALVERT CLIFFS 2 318 1

4

'25 CE MANUAL 100 a

229 CALVERT CLIFFS 2 318 2

5 6

CE AUTO 53 230 CALVERT CLIFFS 2 318 12 12 12 CE AUTO 46 231 HATCH 1 321 6

1 30 GE AUTO 47 232 HATCH 1 321 10 1

16 GE AUTO 61 233 HATCH 1 321 6

2 2

GE AUTO 75 234 HATCH I 321 13 2

16 GE MANUAL 15 235 HATCH 1 321 18 5

15 GE MANUAL 100 236 HATCH 1 321 26 6

27 GE AUTO 64 237 HATCH 1 321 27 7

24 GE AUTO 98 238 HATCH I 321 10 1

GE MANUAL 55 239 DIABLO CANYON 2 323 6

8 24 W

MANUAL 0

240 DIABLO CANYON 2 323 7

8 29 W

AUTO 4

241 DIABLO CANYON 2 323 9

10 22 W

AUTO 30 242 DIABLO CANYON 2 323 11 28 W

AJTO O

243 DIABLO CANYON 2 323 13 11 6

W AUTO 47 244 DIABLO CANYON 2 321 15 11 9

W AUTO 24 245 DIABLO CANYON 2 323 16 11 26 W

AUTO 75 246 DIABLO CANYON 2 323 17 11 28 W

AUTO 1

247 DIABLO CANYON 2 323 12 3

W MANUAL 20 248 DIABLO CANYON 2 323 12 21 W

AUTO 50 249 DIABLO CANYON 2 323 12 25 W

AUTO 90 250 DIABLO CANYON 2 323 16 12 1

W AUTO 75 251 DIABLO CANYON 2 323 18 12 2

W MANUAL 30 252 DIABLO CANYON 2 323 10 10 24 W

AUTO 30 253 BRUNSWICK 2 324 5

9 4

GE MANUAL 100 254 BRUNSWICK 2 324 11 10 15 GE AUTO 99 255 BRUNSWICK 2 324 12 11 22 GE AUTO 70

\\.

APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE OBS NAME DOCEET LER MONTH DAY VENDOR TRIPS POWER 256 BRUNSWICK 1 325 8

1 24 GE AUTO 37 257 BRUNSWICK 1 325 59 11 2

GE AUTO 3

258 SEQUOYAH I 327 29 7

19 W

AUTO 100 259 SEQUOYAH 2 328 1

1 14 W

AUTO 30 260 SEQUOYAH 2 328 2

1 12 W

AUTO 100 261 SEQUOYAH 2 328 4

2 15 W

AUTO 100 262 SEQUOYAH 2 328 4

2 17 W

AUTO 30 263 SEQUOYAH 2 328 9

5 3

W AUTO 100 264 SEQUOYAH 2 328 10 5

22 W

AUTO 100 265 FITZPATRICK 333 6

2 15 GE AUTO 60 266 FITZPATRICK 333 17 6

10 GE AUTO 29 IS7 FITZPATRICK 333 18 6

24 GE AUTO 99 268 FITZPATRICK 333 19 7

19 GE AUTO 100 269 FITZPATRICK 333 21 7

26 GE AUTO 90 270 FITZPATRICK 333 21 8

9 GE AUTO 90 271 FITZPATRICK 333 22 8

20 GE AU60 100 272 FITZPATRICK 333 26 10 31 GE AUTO 100 273 BEAVER VALLEY 1 334 3

1 16 W

AUTO 96 274 BEAVER VALLEY 1 334 6

2 21 W

AUTO 17 275 BEAVER VALLEY 1 334 10 5

6 W

AUTO 36 276 BEAVER VALLEY 1 334 13 7

6 W

AUTO 9

277 BEAVER VALLEY 1 334 15 8

29 W

AUTO 100 278 BEAVER VALLEY 1 334 16 9

16 W

AUTO 100 os 279 BEAVER VALLEY 1 334 18 10 4

W AUTO 40 U'

280 BEAVER VALLEY 1 334 19 10 25 W

AUTO 100 281 ST.LUCIE 1 335 3

3 7

CE MANUAL 100 282 MILLSTONE 2 336 11 7

15 CE AUTO 100 283 NORTH ANNA 1 338 17 9

17 W

MANUAL 16 284 NORTH ANNA 1 338 19 10 24 W

MANUAL 87 285 NORTH ANNA 2 339 5

3 23 W

MANUAL 100 286 NORTH ANNA 2 339 6

4 26 W

AUTO 100 287 FERMI 2 341 29 6

28 GE MANUAL 2

288 FERMI 2 341 40 7

1 GE AUTO 2

289 FERMI 2 341 33 7

5 GE AUTO 4

290 FERMI 2 341 35 7

9 GE AUTO 2

291 FERMI 2 341 59 9

3 GE AUTO 3

292 FERMI 2 341 66 9

27 GE AUTO 5

293 FERMI 2 341 68 10 1

-GE AUTO 1

294 FERMI 2 341 71 10 11 GE AUTO O

295 TROJAN 344 2

3 9

W AUTO 100 296 TROJAN 344 9

7 20 W

AUTO 100 297 TROJAN 344 10 8

26 W

AUTO 100 298 TROJAN 344 12 9

24 W

AUTO 65 299 DAVIS-BESSE 346 2

1 15 BW AUTO O

300 DAVIS-BESSE 1 346 5

3 21 BW AUTO 28 301 DAVIS-BESSE 1 346 9

4 24 BW AUTO 98 302 DAVIS BESSE 1 346 11 6

2 BW AUTO 85 303 DAVIS-BESSE 1 346 13 6

9 BW AUTO 80 304 FARLEY 1 348 2

3 13 W

AUTO 100 305 FARLEY I 348 10 6

8 W

AUTO 22 306 FARLEY 1 348 12 6

23 W

AUTO 99

i, APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE OBS NAME DOChET LER MONTH DAY VENDOR TRIPS POWER 307 FARLEY 1 348 13 7

17 W

AUTO 100 308 LIMERICK 1 352 21 1

31 GE AUTO 4

309 LIMERICK 1 352 73 9

11 GE AUTO 28 310 LIMERICK 1 352 83 10 15 GE AUTO 2

311 LIMERICK 1 352 95 12 8

GE AUTO 65 312 SAN ONOFRE 2 361 24 4

16 CE AUTO 2

313 SAN ONOFRE 2 361 28 4

19 CE AUTO 50 314 SAN ONOFRE 2 361 31 5

18 CE AUTO 80 315 SAN ONOFRE 2 361 18 8

20 CE AUTO 100 316, SAN ONOFRE 2 361 41 8

1 CE AUTO 100 317 SAN ONOTRE 2 361 46 9

12 CE AUTO 100 318 SAN ONOFRE 2 361 50 10 18 CE AU10 100 319 SAN ONOFRE 2 361 51 10 19 CE AUTO 19 320 SAN ONOFRE 2 361 60 11 9

CE AUTO 10 321 SAN ONOFRE 2 361 58 12 10 CE AUTO 100 322 SAN ONOFRE 3 362 8

3 19 CE AUTO 16 323 SAN ONOFRE 3 362 10 3

29 CE AUTO 96 324 SAN ONOFRE 3 362 12 4

4 CE AUTO 100 325 SAN ONOFRE 3 362 13 4

8 CE AUTO 100 326 SAN ONOFRE 3 362 20 5

24 CE AUTO 100 327 FARLEY 2 364 8

3 28 W

AUTO 24 328 FARLEY 2 364 9

3 30 W

AUTO 95 329 FARLEY 2 364 10 7

15 W

AUTO 99 330 FARLEY 2 364 11 7

17 W

AUTO 26 331 FARLEY 2 364 12 8

2 W

AUTO 98 332 HATCH 2 366 1

1 19 GE AUTO 75 331 HATCH 7 366 26 5

24 GE AUIO 2

334 HATCH 2 366 18 6

3 GE AUTO 90 335 HATCH 2 366 30 11 5

GE AUTO 99 336 HATCH 2 366 36 11 7

GE AUTO O

337 ARKANSAS 2 368 4

2 4

CE AUTO 100 338 ARKANSAS 2 368 5

2 7

CE AUTO 7

339 ARKANSAS 2 368 14 7

18 CE AUTO 100 340 ARKANSAS 2 368 15 7

30 CE AUTO 100 341 ARKANSAS 2 368 16 8

5 CE AUTO 1U0 342 ARKANSAS 2 368 17 8

13 CE AUTO 100 343 ARKANSAS 2 368 18 8

6 CE AUTO 100 344 ARKANSAS 2 368 20 9

19 CE AUTO O

345 ARKANSAS 2 368 22 10 8

CE AUTO 98 346 ARKANSAS 2 368 23 10 19 CE AUTO 80 347 MCGUIRE 1 369 4

1 28 W

AUTO 100 348 MCGUIRE 1 369 6

2 5

W AUTO 94 349 MCGUIRE 1 369 34 11 2

W AUTO 100 350 MCGUIRE 1 369 36 11 19 W

AUTO 53 351 MCGUIRE 1 369 12 22 W

AUTO 100 352 MCGbIRE 2 370 12 5

8 W

MANUAL 10 353 MCCUIRE 2 370 13 5

16 W

MANUAL 96 354 MCGUIRE 2 370 14 5

17 W

MANUAL 0

355 MCGUIRE 2 370 17 6

1 W

AUTO 100 356 MCGUIRE 2 370 18 6

24 W

AUTO 100 357 MCGUIRE 2 370 19 7

12 W

AUTO 100

-4

u APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE CBS NAME DOCKET LER MO' NTH DAY VENDOk TRIPS POWER 358 MCGUTRE 2 370 19 7

2*

W AUTO 63 359 MCGUIRE 2 370 21 7

29 W

MANUAL 0

360 MCGUIRE 2 370 26 10 24 W

AUTO 100 361 MCGUIRE 2 370 27 10 2E W

AUTO 91 362 MCGUIRE 2 370 34 11 2

W AUTO 100 363 MCGUIRE 2 370 30 12 11 W

AUTO O

364 LASALLE 1 373 2

1 5

GE MANUAL 99 365 LASALLE 1 373 11 2

2 GE AUTO 96 3b6 LASALLE 1 373 17 2

8 GE AUTO 75 367 LASALLE 1 373 24 3

3 GE AUTO 98 368 LASALLE 1 373 29 3

21 GE AUTO 99 369 LASALLE 1 373 35 4

11 GE AUTO 80 370 LASALLE 1 373 45 5

31 GE AUTO 64 371 LASALLE 1 373 52 6

29 GE MANUAL 92 372 LASALLE 1 373 58 8

17 GE AUTO 93 373 LASALLE 2 374 44 10 21 GE AUTO 97 374 WATERFORD 3 382 4

2 9

CE AUTO O

375 WATERFORD 3 382 7

3 14 CE AUTO O

376 WATERFORD 3 382 8

3 17 CE AUTO 5

377 WATERFORD 3 382 8

3 19 CE AUTO 15 378 WATERFORD 3 382 10 3

21 CE AUTO 20 379 WATERFORD 3 382 10 3

22 CE AUTO 15 380 WATERFORD 3 382 13 4

4 CE AUTO 15 m

381 WATERFORD 3 382 14 4

15 CE AUTO 20

'd 382 WATERFORD 3 382 17 5

5 CE AUTO 17 383 WATERFORD 3 382 18 5

11 CE AUTO 1

384 WATERFORD 3 382 20 5

18 CE AUTO 25 385 WATERFORD 3 382 21 5

23 CE AUTO ES 386 WATERFORD 3 a82 22 5

24 CE AUTO 11 387 WATERFORD 3 382 27 6

26 CE AUTO 91 388 WATERFORD 3 382 28 6

27 CE AUTO 1

389 WATERFORD 3 382 29 6

30 CE AUTO 15 390 WATERFORD 3 3E2 31 7

4 CE AUTO 100 391 WATERFORD 3 382 32 7

4 CE AUTO 6

392 WATERFORD 3 382 33 7

5 CE AUTO 58 393 WATERFORD 3 38?

34 7

/

CE AUTO 90 394 WATERFORD 3 782 35 7

14 CE AUTO 100 395 WATERFORD 3 382 40 9

29 CE AUTO 92 396 WATERFORD 3 382 41 10 2

CE AUTO 100 397 WATERFORD 3 382 42 10 2

CE AUTO 3

398 WATERFORD 3 382 44 10 10 CE AUTO 5

399 WATERFORD 3 382 47 10 28 CE AUTO 100 12 19 CE AUTO 100 400 WATERFORD 3 382 12 6

CE AUTO 100 401 WATERFORD 3 382 51 402 SUSQUEHANNA 1 387 3

1 24 GE AUTO 82 403 SUSQUEHANNA 1 387 30 10 28 GE AUTO 100 404 SUSOUEHANNA 1 387 31 10 30 GE AUTO 64 405 SUSQUEHANNA 1 387 34 12 2

GE AUTO 100 406 SUSQUEHANNA 2 388 3

1 19 GE AUTO 99 407 SUSQUEHANNA 2 388 21 8

30 GE AUTO 100 408 SUSQUEHANNA 2 388 23 8

5 GE AUTO 87

- _ = _

APPENOIX A TABLE A.1 1985 REACTOP TRIPS PAGE OBS NAME DOCKET LER MONTH DAY VENDOR TRIPS POWER 40S SUSQUEHANNA 2 388 25 10 5

GE AUTO 100 410 SUSQUEHANNA 2 388 34 12 2

GE AUTO 81 All ST.LUCIE 2 389 1

4 8

CE AUTO 10 412 ST.LUCIE 2 389 2

4 8

CE AUTO 15 413 ST.LUCIE 2 389 3

4 17 CE MANUAL 99 414 ST.LUCIE 2 389 4

4 17 CE AUTO 20 415 ST.LUCIE 2 389 7

7 18 CE AUTO 99 416 ST.LUCIE 2 389 8

8 8

CE AUTO 99 417 ST.LUCIE 2 389 9

9 10 CE MANUAL 99 418 SUMMER 395 1

2 16 W

MANUAL 3

419 SUMMER 395 1

2 27 W

MANUAL 10 420 SUMMER 395 3

2 28 W

AUTO 6

421 SUMMER 395 5

3 17 W

AUTO 100 422 SUMMER 395 8

4 4

W MANUAL 100 423 SUPMER 395 11 4

18 W

AUTO 100 424 SUMMER 395 13 4

29 W

AUTO 30 425 SUPME R 395 22 8

24 W

AUTO 8

426 SUPMER 395 24 8

20 W

AUTO 100 427 Sup#9ER 395 24 8

24 W

AUTO 10 428 SUMMER 395 27 9

20 W

AUTO 93 429 SUDMER 395 12 25 W

MANUAL 92 430 WPPSS 2 397 2

1 1

GE AUTO 100 431 WPPSS 2 397 3

1 25 GE AUTO 100 m

432 WPPSS 2 397 6

1 17 GE AUTO 100 CD 433 WPPSS 2 397 7

1 31 GE AUTO 100 434 WPPSS 2 397 14 2

14 GE AUTO 14 435 WPPSS 2 397 16 2

13 GE AUTO 100 436 WPPSS 2 397 24 3

22 GL AUTO 100 437 WPPSS 2 397 46 6

29 GE MANUAL 20 438 WPPSS 2 397 47 7

1 GE AUTO 42 439 WPPSS 2 397 53 8

4 GE AUTO 71 440 WPPSS 2 397 59 11 13 GE AUTO 50 441 WPPSS 2 397 61 11 17 GE AUTO 1

442 LA CROSSE 409 8

4 20 AC AUTO 73 443 LA CROSSE 409 9

4 21 AC AUTO 1

444 LA CROSSE 409 11 4

27 AC AUTO 78 445 LA CROSSE 409 12 5

17 AC MANUAL 97 446 LA CROSSE 409 14 7

25 AC AUTO 96 447 LA CROSSE 409 15 7

26 AC AUTO 1

448 LA CROSSE 409 16 9

14 AC AUTO 96 449 LA CROSSE 409 17 10 22 AC AUTO 98 450 LA CROSSE 409 19 10 26 AC AOTO 71 451 CATAWBA I 413 4

1 14 W

MANUAL 0

452 CATAWBA 1 413 8

1 23 W

AUTO 20 453 CATAWBA 1 413 12 2

9 W

AUTO 12 454 CATAWBA 1 413 20 3

14 W

AUTO 49 455 CATAW8A 1 413 25 4

15 W

.tuTO 90 456 CATAWBA 1 413 41 6

13 W

MANUAL 6

457 CATAWBA 1 413 42 6

16 W

AUTO 50 458 CATAWBA 1 413 43 6

20 W

MANUAL 64 459 CATAWBA 1 413 44 6

22 W

AUTO 15 w

L

~

APPENDIX A TABLE A.1 1985 REACTOR TRIPS PAGE OBS NAME DOCKET LER MONTH DAY VENDOK TRIPS POWER 460 CATAWBA 1 413 45 7

10 W

AUTO 100 e

461 CATAWBA 1 413 67 11 20 W

AUTO 62 462 CATAWBA 1 413 12 20 W

AUTO 100 463 GRAND GULF 1 416 2

1 29 GE AUTO 52 464 GRAND GULF 1 416 4

1 27 GE AUTO 69 465 GRAND GULF 1 416 7

2 10 GE AUTO 49 466 GRAND GULF 1 416 8

2 13 GE AUTO 16 467 GRAND GULF 1 416 13 4

3 GE AUTO 57 468 GRAND GULF 1 416 16 4

7 GE AUTO 75 469 GRAND GULF 1 416 18 4

14 GE AUTO 73 470 GRAND GULF 1 416 20 5

24 GE AUTO 74 471 GRAND GULF 1 416 21 6

4 GE AUTO 97 472 GRAND GULF 1 416 24 6

27 GE MANUAL 97 473 GRAND GULF 1 416 27 7

1 GE AUTO 100 474 GRAND GULF 1 416 30 8

7 GE AUTO 89 475 GRAND GULF 1 416 36 9

16 GE AUTO 100 476 GRAND GULF 1 416 12 31 GE AUTO 98 477 BYRON 1 45?

22 2

18 W

AUTO 3

478 BYRON 1 454 71 2

27 W

AUTO 6

479 BYRON 1 454 24 3

1 W

AUTO 13 480 BYRON 1 454 28 3

4 W

AUTO 13 481 BYRON 1 454 29 3

5 W

AUTO O

482 BYRON 1 454 31 3

8 W

AUTO 18 os 483 BYRON 1 454 42 3

29 W

AUTO 18 43 484 BYRON 1 454 39 4

2 W

AUTO 30 485 BYRON 1 454 42 4

10 W

AUTO 28 486 BYRON 1 454 46 4

21 W

AUTO 50 487 BYRON 1 454 51 5

21 W

AUTO 6

488 BYRON 1 454 52 5

22 W

AUTO 50 489 BYRON 1 454 53 5

25 W

AUTO 13 490 BYRON 1 454 54 5

24 W

AUTO 50 491 BYRON 1 454 61 6

24 W

AUTO 97 492 BYRON 1 454 62 6

27 W

AUTO 99 493 BYRON 1 454 63 7

8 W

AUTO 8

494 BYRON 1 454 66 7

12 W

AUTO 100 495 BYRON 1 454 68 7

13 W

AUTO 11 496 BYRON 1 454 69 7

14 W

AUTO O

497 BYRON 1 454 70 7

24 W

MANUAL 100 498 BYRON 1 454 78 8

7 W

AUTO 98 12 27 W

AUTO 92 499 BYROM 1 454 04 21 W

AUTO 50 500 BYRON 1 454 45 501 BYRON 1 454 90 10 9

W AUTO 92 502 RIVER BEND 1 458 34 11 13 GE MANUAL 2

1 503 RIVER BEND 1 458 41 11 21 GE AUTO 3

l 504 RIVER BEk0 1 458 420 11 28 GE AUTO 2

l 505 RIVER BEND 1 458 12 24 GE AUTO 1

12 31 GE AUTO 25 506 RIVER BEND 1 458 507 WOLF CREEK 482 39 6

6 W

AUTO 6

l 508 WOLF CREEK 482 41 6

13 W

AUTO 1

t 509 WOLF CREEK 482 45 6

23 W

MANUAL 29 l

510 WOLF CREEK 482 46 6

24 W

AUTO O

l i

I i

APPEND 1X A TABLE A 1 1985 REACTOR TRIPS PAGE OBS NAME DOCKET LER MONTH DAY VENDOR TRIPS POWER 511 WOLF CREEK 482 42 7

11 W

AUTO 14 512 WOLF CREE %

482 49 7

9 W

AUTO 47 513 WOLF CREEK 482 50 7

10 W

AUTO 11 514 WOLF CREEK 482 54 7

23 W

AUTO 25 515 WOLF CREEK 482 58 7

31 W

AUTO 86 516 WOLF CREEK 482 60 8

7 W

AUTO 92 517 WOLF CREEK 482 64 9

2 W

AUTO 6

518 WOLF CREEK 482 65 9

5 W

AUTO 100 519 WOLF CREEK 482 67 9

23 W

AUTO 100 520 Wotr CREEK 482 69 10 7

W MANUAL 34

.521 WOLF CiEEK A82 72 10 10 W

AUTO 27 522 CALLAWAY 1 483 1

1 2

W AUTO 50 523 CALLAWAY 1 483 2

1 7

W MANUAL 50 524 CALLAWAY 1 483 5

1 31 W

AUTO 100 525 CALLAWAY 1 483 10 2

21 W

AUTO 100 526 CALLAWAY 1 483 11 2

22 MANUAL 1

527 CALLAWAY 1 483 16 3

3 W

AUTO 100 528 CALLAWAY 1 483 22 4

10 W

AUTO O

529 CALLAWAY 1 483 24 5

6 W

AUTO 100 530 CALLAWAY 1 483 26 6

7 W

AUTO 100 531 CALLAWAY 1 483 31 6

20 W

AUTO 100 532 CALLAWAY 1 483 34 7

18 W

AUTO 100 533 CALLAWAY I 483 36 7

31 W

AUTO 100

-a 534 CALLAWAY 1 483 38 8

20 W

AUTO 100 C3 535 CALLAWAY 1 483 39 8

20 W

ALTO 25 536 CALLAWAY 1 483 42 10 2

W AUTO 100 537 CALLAWAY 1 483 48 11 2

W AUTO 100 538 CALLAWAY 1 483 49 11 11 W

AUTO 100 12 26 W

AUTO 100 539 CALLAWAY 1 483 12 8

W AUTO 85 540 CALLAWAY 1 483 51 541 CALLAWAY 1 483 52 12 9

W AUTO 1

542 PALO VERDE 1 528 19 6

14 CE AUTO 19 543 PALO VERDE 1 52 43 7

1 CE AUTO 43 544 PALO VERDE 1 528 49 7

17 CE AUTO 50 545 PALO VERDE 1 528 63 9

12 CE AUTO 53 10 7

CE AUTO O

546 PALO VERDE 1 528 10 3

CE AUTO 52 547 PALO VERDE 1 528 10 24 CE AUTO 100 548 PALO VERDE 1 528 10 24 CE AUTO 31 549 PALO VERDE 1 528 71 550 PALO VERDE 1 528 12 17 CE AUTO 2

551 PALO VERDE 1 528 88 12 4

CE AUTO 54 552 PALO VERDE 1 528 32 20 CE AUTO 41 8

m___

APPENDIX A TABLE A.2 1985 RPS ACTUATIONS WITHOUT ROD MOTION PAGE OSS NAME DOCKET LER MONTH DAY VENDOR TRIPS POWER 1

BIG ROCK POINT 155 3

4 14 GE AUTO O

2 BIG ROCK POINT 155 4

5 25 GE AUTO O

3 OYSTER CREEK 219 12 6

12 GE AUTO O

4 OYSTER CREEK 219 12 6

12 GE AUTO O

5 OYSTER CREEK 219 12 6

12 GE AUTO O

6 OYSTER CREEK 219 12 6

12 GE AUTO O

7 OYSTER CREEK 219 22 11 20 GE AUTO O

8 OYSTER CREEK 211 22 11 20 GE AUTO O

9 DRESDEN 2 237 4

2 6

GE AUTO O

10 DRESDEN 2 237 4

2 6

GE AUTO O

11 DRESDEN 2 237 6

2 17 GE AUTO O

12 DRESDEN 2 237 8

2 22 GE AUTO O

13 DRESDEN 2 237 12 3

9 GE AUTO O

14 DRESDEN 2 237 13 3

10 GE AUTO O

15 DRESDEN 2 237 14 3

15 GE AUTO O

16 DRESDEN 2 237 21 5

3 GE AUTO O

17 DRESDEN 2 237 21 5

3 GE AUTO O

18 DRESDEN 2 237 23 5

2 GE AUTO O

19 DRESDEN 2 237 23 5

2 GE AUTO O

20 DRESDEN 2 237 34 8

16 GE AUTO O

21 DRESDEN 2 237 41 11 12 GE AUTO O

22 R.E.GINNA 244 5

4 5

W AUTO O

y 23 R.E.GINNA 244 10 4

8 W

AUTO O

24 DRESDEN 3 245 24 11 20 GE AUTO O

25 DRESDEN 3 249 11 15 GE AUTO O

26 TURKEY POINT 3 250 40 11 30 W

AUTO O

27 TURKEY POINT 4 251 17 6

20 W

AUTO O

7 17 W

AUTO O

28 TURKEY POINT 4 251 5

30 GE AUTO O

29 QUAD CITIES 1 254 6

30 PALISADES 255 13 8

23 CE AUTO O

31 PALISADES 255 13 8

26 CE AUTO O

32 PALISADES 255 13 3

27 CE AUTO O

33 PALISADES 255 13 3

27 CE AUTO O

34 BROWNS FERRY Z 259 43 8

25 GE AUTO O

35 BROWNS FERRY 1 259 48 29 8

GE AUTO O

36 BROWNS FERRY 2 260 3

2 27 GE AUTO O

37 BROWNS FERRY 2 260

-9 7

12 GE AUTO O

38 Q00 CITIES 2 265 10 3

28 GE AUTO O

39 QUAD CITIES 2 265 10 3

28 GE AUTO O

40 QUAD CITIES 2 265 14 5

31 GE AUTO O

41 VERMONT YANKEE 271 9

10 7

GE MANUAL 0

42 PEACH BOTTOM 2 277 2

5 30 GE AUTO O

43 PEACM BOTTOM 2 277 4

6 22 GE AUTO O

44 PEACH BOTTOM 2 277 4

6 22 GE AUTO O

45 PEACH BOTTOM 2 277 6

6 27 GE AUTO O

46 PEACH BOTTOM 2 277 7

6 29 GE AUTO O

47 PEACH BOTTOM 2 277 7

6 29 GE AUTO O

48 PEACH BOTTOM 2 277 9

6 29 GE AUTO O

49 PEACH BOTTOM 2 277 14 8

20 GE AUTO O

50 PEACH BOTTOM 2 277 15 8

22 GE AUTO O

51 fCACH BOTTOM 2 277 20 9

24 GE AUTO O

L

~

APPENDIX A TABLE A.2 1985 RPS ACTUATIONS WITHOUT ROD MOTION PAGE l OBS NAME DOCKET LER MONTH DAY VENDOR TRIPS

• 0WER 52 PEACH BOTTOM 3 278 21 8

26 GE AUTO O

53 PEACH BOTTOM 3 278 21 8

28 GE AUTO O

54 PEACH BOTTOM 3 278 21 8

29 GE AUTO O

55 PEACH BOTTOM 3 278 21 9

11 GE AUTO O

56 PEACH BOTTOM 3 278 21 9

12 GE AUTO 0

1 t

57 PEACH BOTTOM 3 278 21 9

13 GE AUTO O

58 PEACH BOTTOM 3 278 21 9

13 GE AUTO O

59 PEACH BOTTOM 3 278 21 9

14 GE AUTO O

60 PEACH BOTTOM 3 278 22 9

11 GE AUTO O

61 PEACH BOTTOM 3

'78 16 10 18 GE AUTO O

62 PEACH BOTTOM 3 278 17 10 18 GE AUTO O

63 PEACH BOTTOM 3 278 21 10 10 GE AUTO O

64 PEACH BOTTOM 3 278 12 17 GE AUTO O

65 INDIAN POINT 3 286 4

3 20 W

AUTO O

66 INDIAN POTNT 3 2t!

4 3

20 W

AUTO O

67 INDIAM POINT 3 286 4

3 20 W

AUTO O

68 ZION 1 295 19 6

7 W

AUTO O

69 ZION 11 295 21 6

3 W

AUTO O

70 KEWAUNEE 305 12 4

10 W

AUTO O

71 RANCHO SECO 1 312 11 6

12 W

AUTO O

72 D.C. COOK 1 315 46 1

7 W

AUTO O

73 D.C.Coch 1 315 10 25 W

AUTO O

u t

i 74 D.C. COOK 1 315 54 10 26 W

AUTO O

N i

l 75 D.C. COOK 2 315 57 10 25 W

AUTO O

76 D.C. COOK 1 315 59 11 4

W AUTO O

77 D.C. COOK 2 316 18 3

20 W

AUTO O

78 D.C. COOK 2 316 24 1

6 W

AUTO O

79 HATCH 1 321 3

1 6

GE AUTO O

SO HATCH 1 321 25 6

18 GE AUTO O

31 HATCH 1 321 9

21 GE AUTO O

82 BRUNSWICK 1 325 11 3

30 GE AUTO O

83 BRUNSWICK 1 325 31 5

23 GE AUTO O

84 BRUNSWICK a 325 32 6

2 GE AUTO O

85 BRUNSWICK 1 325 37 7

10 GE AUTO O

86 BRUNSWICK 1 325 45 8

15 GE AUTO O

87 BRUNSWICK 1 325 51 9

21 GE AUTO O

SS DUANE ARNOLD 331 4

2 2

GE AUTO O

l 89 DUANE ARNOLD 331 9

3 15 GE AUTO O

90 DUANE ARNOLD 331 9

3 15 GE AUTO O

91 DUANE AKNOLD 331 11 3

21 GE AUTO O

92 DUANE ARNOLD 331 11 3

21 GE AUTO O

93 DUANE ARNOLD 331 18 6

6 GE AUTO O

94 DUANE ARNOLD 331 20 6

10 GE AUTO O

95 DUANE ARNOLD 331 20 6

10 GE AUTO O

96 DUANE ARNOLD 331 24 7

7 GE AUTO O

97 DUANE ARNOLD 331 24 7

7 GE AUTO O

98 DUANE ARNOLD 331 25 7

8 GE AUTO O

99 DUANE ARNOLD 331 34 7

9 GE AUTO O

100 FITZPATRICK 333 16 5

24 GE AUTO

'O 101 BEAVER VALLEY 1 334 8

5 2

W AUTO O

102 FERMI 2 341 1

3 28 GE AUTO O

APPENDIX A TABLE A.2 1985 RPS ACTUATIONS WITHOUT ROD MOTION PAGE l OSS NAME DOCKET LER MONTH DAY VENDOR TRIPS POWER j

1C3 FERMI 2 341 1

3 31 GE AUTO O

104 FERMI 2 341 3

4 3

GE AUTO O

105 FERMI 2 341 8

4 25 GE AUTO O

106 FERMI 2 341 8

4 25 GE AUTO O

107 FERMI 2 341 10 4

27 GE AUTO O

108 FERMI 2 341 10 4

27 GE AUTO O

109 FERMI 2 341 10 4.

27 GE AUTO O

110 FERMI 2 341 10 4'

27 GE MANUAL 0

111 FERMI 2 341 11 4

27 GE AUTO O

112 FERMI 2 341 11 4

21 GE AUTO O

113 FERMI 2 341 12 4

28 GE AUTO O

114 FERMI 2 341 12 4

28 GE AUTO O

IIS FERMI 2 341 14 5

1 GE AUTO O

116 FERMI 2 341 15 5

6 GE AUTO O

117 FERMI 2 341 16 5

9 GE AUTO O

118 FERMI 2 341 22 6

6 GE AUTO O

119 FERMI 2 341 67 9

28 GE AUTO O

120 FERMI 2 341 10 15 GE AUTO O

121 FERMI 2 341 73 10 14 GE AUTO O

122 FARLEY I 348 9

5 17 W

AUTO O

123 LIMF RICK 1 352 46 4

23 GE AUTO O

124 LIMERICK 1 352 66 8

8 GE AUTO O

y 125 HATCH 2 366 2

4 29 GE AUTO O

00 126 HATCH 2 366 2

4 29 GE AUTO O

127 HATCH 2 366 12 4

9 GE AUTO O

128 HATCH 2 366 13 4

11 GE AUTO O

129 HATCH 2 366 25 5

20 GE AUTO O

9 13 CE AUTO O

s 130 ARKANSAS 2 368 3

27 GE AUTO O

131 tASALLE 1 373 28 132 LASALLE 1 373 55 7

22 GE AUTO O

133 LASALLE 2 374 22 5

10 GE AUTO O

134 LASALLE 2 374 25 6

6 GE AUTO O

135 LASALLE 2 374 25 6

6 GE AUTO O

136 LASALLE 2 374 25 6

7 GE AUTO O

137 LASALLE 2 374 28 6

4 GE AUTO O

138 LASALLE 2 374 28 6

4 GE AUTO O

139 SUSQUEHANNA 1 387 18 5

13 GE AUTO O

140 SUSQUEHANNA 1 387 18 5

14 GE AUTO O

141 SUSOUEHANNA 1 387 19 5

16 GE AUTO O

142 SUSQUEHAsutA 2 388 16 4

27 GE AUTO O

143 WPPSS 2 397 1

17 GE AUTO O

144 WM SS 2 397 13 2

14 GE AUTO O

145 WPP15 2 397 30 5

7 GE AUTO O

146 WPPSS 2 397 31 5

8 GE AUTO O

147 WPPSS 2 397 33 5

25 GE AUTO O

148 LA CROSSE 409 14 7

25 AC AUTO O

149 CATAd6A 1 413 11 15 W

AUTO O

150 GRA:.D GULF 1 416 9

2 14 GE AUTO O

151 MILLSTONE 3 423 2

12 15 W

AUTO O

152 BYRON 1 454 3

1 4

W AUTO O

153 BYRON 1 454 16 1

24 W

AUTO O

I

s

~

APPENDIX A TABLE A.2 1985 RPS ACTUATIONS WITHOUT ROD MOTION PAGE t OSS NAME DOCKET LER MONTH DAY VENDOM TRIPS POWER 154 RIVER BEND 1 458 1

8 31 GE AUTO O

155 RIVER BEND 1 458 2

9 4

GE AUTO O

156 RIVE R BEND 1 458 2

9 4

GE AUTO O

157 RIVER BEND 1 458 2

9 4

GE AUTO O

158 RIVER BEND 1 458 3

9 4

GE AUTO O

159 RIVE

• BEND 1 458 5

9 13 GE AUTO O

160 RIVER BEND 1 458 6

9 16 GE AUTO O

161 RIVER BEND 1 458 3

9 23 GE AUTO O

162 WOLF CREEK 482 23 5

1 W

AUTO O

163 CALLAWAY I 483 25 5

6 W

AUTO O

164 PALO VERDE 1 528 3

12 W

AUTO O

165 PALO VERDE 1 528 2

3 2

CE AUTO O

IM PALO VERDE 1 528 9

3 21 CE AUTO O

167 PALO VERDE 1 528 43 7

1 CE AUTO O

5

-y

--,i

i k

APPENDIX B 1985 REACTOR TRIP RATES AT POWER GREATER THAN 15%

IN 1985 PAGE APPENDIX B 1985 REACTOR TRIP RATES NAME MANUAL PITO LESS THAN GREATER CRITICAL TRIP RATE PER MEAN TIME MATIC OR EQUAL THAN HOURS 1000 HOURS BETWEEN TRIPS 15% POWER 15% POWER POWER GT 15 POWER GT 15%

DIABLO CANYON 2 3

11 4

to 1874.2 5.34 187.4 WATERFORD 3 0

28 13 15 3343.0 4.49 222.9 BYROM 1 1

24 10 15 4656.4 3.22 310.4 PALO VERDE 1 C

11 2

9 2388.0 3.12 320.9 l

GRAND GULF 1 1

13 0

14 5092.1 2.75 363.7 CATAWBA 3

9 4

8 3612.4 2.21 451.6 CALLAWAY 1 2

18 3

17 3161.0 2.08 480.1

. WOLF CREEK 2

13 6

9 4471.7 2.01 496.9 PEACH BOTTOM 2 2

5 2

5 2910.6 1.72 582.1 SALEM 2 1

9 1

9 5231.2 1.72 581.2 RIVER BEND 1 1

4 4

1 589.4 1.70 589.4 CRYSTAL RIVER 3 2

6 1

7 4385.3 1.60 626.5 LASALLE 1 2

7 0

9 5757.5 1.56 639.7 SAN ONOFRE 2 0

10 2

8 5235.8 1.53 654.5 j

DIABLO CANYON 1 1

9 2

8 5295.6 1.51 662.0 i

MCGUIRE 2 4

8 4

8 5490.5 1.46 686.3 WPPSS 2 1

11 2

10 6899.7 1.45 690.0 DAVIS-BESSE 1 0

5 1

4 2346.6 1.41 711.7 FITZPATRICK 0

3 0

t 5799.6 1.38 725.0

-a ARKANSAS 2 0

10 2

8 6377.4 1.25 797.2 on BROWNS FERRY 1 1

1 0

2 1647.7 1.21 823.9 INDIAN POINT 3 2

3 3

7 5901.1 1.19 843.0 TURKEY POINT 4 0

10 1

9 7916.8 1.14 879.6 ARKANSAS 1 2

6 0

8 7005.4 1.14 875.7 SEQUOYAH 2 0

6 0

6 5289.4 1.13 881.6 CALVERT CLIFFS 1 0

0 6

53o7.6 1.12 894.6 SL w R 4

8 5

7 6439.9 1.09 920.0 TFOIAN POINT 2 2

9 2

9 8504.1 1.06 944.9 Seed ONOFRE 3 0

5 0

5 4789.6 1.04 958.0 H.B.RO8INSON 2 0

12 4

3 7859.8 1.02 982.5 DRESDEN 2 1

7 3

5 4961.6 1.01 992.3 HATCH I 3

5 1

7 6907.5 1.01 986.8 MAINE YANKEE 1

8 2

7 7037.1 0.99 1005.3 TURKEY POINT 3 0

7 2

5 5405.0 0.93 1081.0 LA CROSSE I

8 2

7 7757.2 0 90 1108.2 BEAVER VALLE(

0 8

1 7

8245.3 0.85 1177.9 QUAD CITIES 2 0

5 0

5 6361.8 0.79 1272.4 OYSTER CREEK 0

6 1

5 6818.5 0.73 1363.7 FARLEY 2 0

5 0

5 6888.1 0.73 1377.6 MCGUIRE 1 '

O 5

0 5

6842.4 0.73 1368.5 SUSQUEHANNA 1 0

4 0

4 5598.5 0.71 1399.6 RANCHO SECO 1 0

4 2

2 2374.6 0.70 1437.3 SUSQUEHANNA 2 0

5 0

5 7121.2 0.70 1424.2 ST.LUCIE 2 2

5 2

5 7442.7 0.67 1488.5 BROWNS FERRY 3 2

0 1

1 1517.5 0.66 1517.5 DRESDEN 3 0

5 1

4 6718.8 0.60 1879.7 NINE MILE POINT 1 1

6 2

5 3524.0 0.59 1704.8 TROJAN O

4 0

4 6804.7 0.59 1701.2 i

i

,a -

e APPENDIX B 1985 REACTOR TRIP RATES AT POWER GREATER THAN 15%

IN 1985 PAGE,

l HADDAM NECK 2

3 0

5 8682.4 0.58 1736.5 I

LIME RICK 1 0

4 2

2 3420.1 0.58 1710.1 l

FARLEY 1 0

4 0

4 7504.1 0.53 1876.0 SURRY 1 0

6 2

4 7935.4 0.50 1983.9 D C. COOK 2 1

3 1

3 5948.8 0.50 1982.9 COOPER 1

0 0

1 2057.5 0.49 2057.5 THREE MILE IS 1 1

1 1

1 2084.8 0.48 2084.8 OCONEE 1 0

4 0

4 3453.3 0.47 2113.3 OCCNEE 2 0

4 1

3 6740.3 0.45 2246.8 CALVERT CLIFFS 2 1

2 0

3 6834.2 0.44 2294.7 l

BRUMSWICK 2 1

2 0

3 7134.8 0.42 2378.3 KEWAUMEE 1

5 3'

3 7266.5 0.41 2422.2 i

HATCH 2 0

5 2

3 7373.1 0.41 2457.7 D.C. COOK 1 0

1 0

1 2595.6 0.39 2595.6 R.E.GINNA 1

7 5

3 7838.4 0.38 2612.8 MONTICELLO O

3 0

3 8163.0 0.37 2721.0 PILGRIM 0

4 1

3 8159.0 0.37 2719.7 SAN ONOFRE 1 1

1 0

2 6783.8 0.29 3391.9 BRUNSWICK 1 0

2 1

3409.6

.0.29 3409.6 NORTH ANNA 1 2

0 0

2 6938.8 0.29 3469.4 l

MILLSTONE 1 0

3 1

2 7324.4 0.27 3662.2 l

PALISADES O

2 0

2 7490.2 0.27 3745.1

[

PRAIRIE ISLAND 1 0

3 1

2 7363.2 0.27 3681.6 i

SEQUOYAH 1 0

1 0

1 3797.2 0.26 3797.2 LASALLE 2 0

1 0

1 3777.6 0.26 3777.6

-4 PEACH BOTTOM 3 0

1 0

1 4055.7 0.25 4055.7 O'

QUAD CITIES 1 1

2 1

2 8339.0 0.24 4169.5 NORTH ANNA 2 1

1 0

2 8534.4 0.23 4267.2 MILLSTONE 2 0

1 0

1 4460.7 0.22 4460.7 ZION 1 1

3 3

1 5321.2 0.19 5321.2 SURRY 2 0

1 0

1 5936.5 0.17 5936.5 ZICM 2 0

1 0

1 5909.2 0.17 5909.2 OCOMEE 3 0

2 1

1 6140.9 0.16 6140.9 VERMONT YANKEE O

1 0

1 6297.2 0.16 6297.2 POINT BEACH 1 0

1 0

1 6974.4 0.14 6974.4 ST_LUCIE 1 1

0 0

1 7134.7 0.14 7134.7 POINT BEACH 2 0

1 0

1 7576.2 0.13 7576.2 i

SALEM 1 0

1 0

1 8361.9 0.12 8361.9 YANKEE ROWER 0

2 2

0 7598.3 0.00 BIG ROCK POINT 2

2 4

0 6539.5 0.00 j

FT. CALHOUN O

O O

O 6466.1 0.00 l

FERMI 2 1

7 8

0 2400.7 0.00 DUANE ARNOLD 0

0 0

0 4733.2 0.00 PRAIRIE ISLAND 2 0

0 0

0 7408.6 0.00 67 485 118 414 m

k APPENDIX C HUMAN ERROR INDUCED REACTOR TRIPS AT POWER GREATER THAN 15% AT PWR*s IN 1985 PAGE CBS SYSTEM POWER PERSON DESC 1

RPS 30 TECHNICIAN Technicians pulled the wrong power fuses during instrument calibration.

2 RPS 100 LIC OPERATOR During RPS test a licensed operator placed test switch in the wrong position.

3 RPS 25 LIC OPERATOR Poor judgement by operators when synchronizing unit with SG Bbistables tripped.

4 RPS 100 TECHNICIAN Communication problem between maintenance and operations personnel.

5 RPS 100 TECHNICIAN During data collection an improper connection of test leads caused the trip.

6 RPS 100 TECHNICIAN Technicians grounded MIS with a soldering iron during repair.

7 RPS 100 TECHNICIAN During analog channel operational test technicains omitted a procedural step.

I RPS 100 TECHNICIAN Communication problem between technicians during calibration lead to trip D 9

RPS 98 TECHNICIAN Failure to verif y the proper MIS cabinet number lead to the t rip.

10 RPS 100 TECHNICIAN Procedural steps were performed out of order.

11 RPS 100 LICENSEE STAFF Personnel did not follow plant procedures.

12 RPS 24 LIC OPERATOR The operator did not block the power range trip at reacto r powe r above 10%.

13 RPS 78 TECriNICIAN Inapporiate use of a meter probe in the operable power range drawer.

14 TURBINE 17 NGT DETERMINED A valve was not repositioned properly after maintenance was performed.

15 TURBINE 48 NOT DETERMINED An improperly set thrust bearing wear detector deeice initiated the trip.

16 TURBINE 25 TECHNICIAN Valve position was not verified during startup.

17 CONTAINMENT 79 TECHNICIAN Trip was attributed to bumping a radiation monitor during troubleshooting.

13 CONDENSATE 20 LIC OPERATOR Operators improperly returned equipment back to service.

19 COICENSATE 90 LIC OPERATOR The backwash selector switch was not changed from the on line domineralizer.

20 COeOENSATE 100 LIC OPERATOR An operator improperly closed all four condenser inlet valves.

21 CONDENSATE 65 TECHNICIAN There was a lack of understanding of systent respanse with the ctrl valv in autC

APPENDIX C HUMAN EAROR INDUCED REACTOR TRIPS AT POWER GREATER THAN 15% AT PWR's IN 1985 PAGE 22 ELECTRICAL 96 TECHNICIAN A feeder bkr. to a protection ch.was t ripped during a maintenance ef f ort.

l 23 ELECTRICAL 100 LIC OPERATOR Operators made a switching error while t rying to correct a ground condition.

l 24 ELECTRICAL 100 UNLIC OPERATOR Plant personnel removed unit 3 MG set rat he r t han unit 4. Wrong unit.

25 ELECTRICAL 100 LIC OPERATOR Power was lost to e protection sys ch when an inverter was removed for repair.

l 26 ELECTRICAL 99 LIC OPERATOR Circuit drawings were misinterpreted while t rying to locate a ground.

27 ELECTRICAL 100 TECHNICIAN Technicians shorted the 115v vital power supply while testing a level ch on PW5 l

28 ELECTRICAL 100 UNLIC OPERATOR The 120v AC vit al bus was de-energized when t he wrong t-kr. was opened.

29 ELECTRICAL 22 LIC OPERATOR The RCP were not properly aligned to the startup transformers.

30 ELECTRICAL 100 LIC OPERATOR Procedural step was not executed during an attempt to energize an inverter.

31 ELECTRICAL 100 LIC OPERATOR Improper positioning of the alternate bkr before the normal bkr was tripped.

32 ELECTRICAL 100 TECHNICIAN A recorder power supply on the vital bus was accidently grounded.

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33 ELECTRICAL 43 LICENSEE STAFF Due to ventitation lineup air temperatures were allowed to go to high.

34 ELECTRICAL 100 NOT DETERMINED Trip att ributed to a wiring error caused by a lack of clarity on wiring diagram 35 ELECTRICAL 100 UNLIC WORKER The trip was the result of the unintentional t ripping of the inverter bkr.

36 CRDS 80 NOT DETERMINED Trip suspected to have been caused by an installation problem with circuit card 37 CRDS 29 LIC OPERATOR The reactor trip hkr was inadvertently tripped during testing.

l 33 CRDS 100 NOT DETERMINED The trip was the direct result of an improperly wired FW control valve.

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3g CRDS 100 NOT DETERMINED Procedures did not caution against tr'obleshooting CEAC without operable flags.

l 40 CROS 100 LIC OPERATOR An operator entered the wrong constant into the CEAC.

41 MAIN QENERATOR 47 LIC OPERATOR Prior to repair the generator's potential transformer was improperly disabled 42 MAIN GENERATOR 100 LIC OPERATOR Improper t roubleshooting and incomplete system unoerstanding caused the t rip.

43 MAIN GENERATOR 100 UMLIC WORKER Accidental jarring of the t ransformer relay by a const r. worker caused the trip 44 MIN GENERATOR 30 TECHNICIAN Technician inadvertently sctuated the current test block of the main generator.

45 MAIN GENEPATOR 91 LICENSEE STAFF Improperly installed fases because of lack of t raining caused the trip.

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APPENDIX C MUMAN ERROR INDUCED REACTOR TRIPS j

AT POWER GREATER THAN 15% AT PWR's IN 1985 PAGE 46 MAIN STEAM 25 LICENSEE STAFF Disk mounting bolts were not torqued to the required torque.

47 MAIF STEAM 30 LICENSEE STAFF Disk mounting bolts were not torqued to the required torque.

48 AFW 69 NOT DETERMINED A relay was installed incorrectly because the drawnings did not reflect chan'ge'a 49 FW 30 TECHNICIAN The error due to inattention to detail during root valve replacement.

50 FW 17 LIC OPEKATOR This trip was caused by improper operation of the feedwater regulating valve.

51 FW 24 TECHNICIAN While troubleshooting a technician removed the wrong circuit card.

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l 52 FW 90 ENG STAFF Operators were not made aware by testing personnel of key unit parameters.

1 53 fW 96 TECHNICIAN A techaician mistakenly cow!pleted the pumps low suttion t rip path wiih meter.U 54 FW 36 LIC OPERATOR Communication problem between operators caused the trip.

55 FW 45 STAFF OPS The t rip was the result of an improper viv. lineup and eot rssetting a' bistable.

56 FW 64 LIC OPERATOR An operating supervisor did not utilize procedures for tag out evolution.

,57 FW 97 UNLIC OPERATOR An equipment operator inadvertently activated the local overspeed trip bar.

58 Fk 95 TECHNICIAN Technicians failed to recognize the effects of the actions.

5g FW 33 LIC OPERATOR A licensed operator failed to follow procedures.

60 FW 47 TEChMICIAN The trip was caused by a technician improperly removing test equipment.

61 FW 100 TECHNICIAN A technician accidently broke a connection in the protection sys.

62 FW 100 LIC CFERATOR An operator initiated this trip by inadvertenly closing the MFIV.

63 EW 27 LIC OPEPATOR Cognitive personnet error allowed insufficient time for SG 1evel to stabilize.

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64 FW 24 LIC OPERATOR Inadequate t raining of operator while the FW cont rols are in manual.

65 FW 22 LIC MAf*4GEMENT Inadequate training in relation to auto FW ct ri system and OTSG 1evel actuation.

1 66 FW 100 (JNLIC WORKER A painter accidently actuated the local MFP manual t rip switch.

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67 FW 10C TECHNICIAN The technician and operator did not throughly research and plan the evolution.

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_ _ - _ _ _. ~. _.

APPENDIX C HUMAN ERROR INDUCED REACTOP TRIPS AT POWER GREATER THAN 15% AT P'A

• s 1

IN 1985 PAGE 68 FW 100 UNLIC OPERATOP An aux.bidg. operator incorrectly de-energized a part of the Feedwater Ctrl.Sys.

I 6s FW 81 TECHNICIAN Technicians working on an instrument ch. initiated the t rip.

70 ESF 20 UNLIC OPERATOR An ESF relay was accidently bumper by a construction worker.

71 ESF 100 TECHNICIAN The improper assembly of a mat riz pushbutton was responsible f or the t rip.

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APPENDIX C HUMAN ERROR INDUCED REACTOR TRIPS AT POWER GREATER THAN 15% AT BWR's IN 1985 PAGE OBS SYSTEM POWER PERSON DESC 1 RPS 100 UNLIC OPERATOR An unlicensed operator inadvernt antly t ripped MG set supplying power to RPS.

2 RPS 99 CONTRACTOR Unknown persons bumped level instruments that initiated a trip.

3 RPS 100 TECHNICIAN Technicians incorrectly valved unit inst rument ation back into se rvice.

4 RPS 99 UNLIC OPERATOR Communicat ion break down between unlic. ope rator and tech. led to trip.

5 RPS 100 CONTRACTOR Inattention while moving the crane by contract personnel.

6 RPS 100 TECHNICIAN Technician did not understand procedural steps following a delay in execution.

7 RPS 93 TECHNICIAN A technician improperly tried to restore an instrument sensing line to service.

8 RPS 31 LIC OPERATOR A misunderstanding of an alarm and lack of precaution during troubleshootingh3 9 TURBINE 60 NOT DETERMINED Trip was the result of an improper valve position for a circuit being tested.

10 TURBINE 83 TECHNICIAN A circuit card in the EHCS was accidently moved while testing was in progress.

11 TURBINE 44 LIC OPERATOR License J operator did not follow procedures.

12 CONDENSATE 99 LIC OPERATOR While transfering loads on the 6.9KV bus the operator made a switching error.

13 CONDENSATE 98 TECHNICIAN While venting hotwell level instruments technicians tripped booster pumps.

14 CONDENSATE 52 LIC OPERATOR Licensed operatos deviated from an approved procedure.

15 ELECTRICAL 100 TECHNICIAN Technicians inadvertently allowed a power cable to short to ground.

16 ELECTRICAL 100 UNLIC WORKER A non plant worker did not remove grounding cables at an offsite location.

17 ELECTR} CAL 74 TECHNICIAN Poor inst ruction were given to an elect rician to peform maintenance on 6.9kw bus 16 ELECTRICAL 64 UNLIC CPERATOR Plant personnel closed the wrong fire protection valve.

19 ELELTRTCAL 100 LIC OPERATOR Due to operator error power was lost to the 4kw bus.

l 20 ELECTRICAL 96 TECHNICIAN A technician initated short blew a f use which supplied power to ct ri.rm inst ru.

21 ELECTRICAL 98 TECHNICIAN Technician bumped the 6.9kw tie bkr while winterizir.g the bkr.

APPENDIX C HUMAN ERROR INDUCED REACTOR TRIPS AT POWER GREATER THAN 15% AT BWR's IN 1985 PAGE 1 22 ELECTRICAL 50 LICENSEE STAFF Insufficient planning and lack coordination with ctr1.rm caused the trip.

23 ELECTRICAL 47 UNLIC WORKER The trip was the result of construction worker installing the wrong insulators.

24 ELECTRICAL 75 UNLIC WORKER The trip was the result of construction worker installing the wrong insulators.

25 MAIN GENERATOR 90 TECHNICIAN Elect rician grounded the main generator excitation system during an inspection.

26 MAIN STEAM 100 TECHNICIAN The trip was the direct result of an improperly wired MSIV solenoid.

27 MAIN STEAM 29 CONTRACT WORKER MSIV closure resulted from an incomplete valve lineup.

28 ESFAS 100 TECHNICIAN Technician inadvertently turned a level instrument valve in the wrong direction.

29 FW 75 STAFF OPS Poor documentation procedures allowed operating staff to use outdated drawings.

30 FW 20 LIC OPERATOR Power was reduced to fast and operators could not stabilize the unit parameters.

31 FW 87 TECHNICIAN Switch position was confirmed in the wrong position by a technician.

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32 INSTRO AIR 100 UNLIC OPERATOR An auxiliary operator incorrectly isolated the control instru. air supply.

33 FORCED CIRCULATION 100 CON'3 ACTOR The bkr f or lub water f or the circulating water pump was inadvertently opened.

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Appendix D Associated Failures Unit LER Power Description Of Failures San Onofre 014 90 AFW pump turbine tripped due to failure of i

inboard and outboard steam valve.

Steam Lump Valves closed prematurely due to a failed pressure switch and relay.

San Onofre 017 60 Main Feedwater System water hammer caused damage to snubbers and feedwater regulating valve.

Oyster Creek 012 100 A Reactor Water Cleanup System isolation valve failed to open because the valve breaker had tripped.

25 Nine Mile Point 1 005 100 Loss of power to a power board caused the loss of I HPCI train.

014 90 The main feedwater pump tripped due to set point drift and a loose connection.

021 98 A Main Steam System electromatic relief valve failed to open due to failure of its actuator.

Dresden 2 034 70 After the trip, radio contact lost throughout the plant and several plant tejephones would not operate.

034 Valve on the clean demi.7eralized water supply line to the isolation condenser failed to open because the power supply for valve had tripped.

034 Motor operated valve in the condensate transfer supply '

line to the isolation condenser would not open.

.--A-Unit LER Power Description Of Failures 034 The breaker between the RPS bus and the RPS MG set tripped.

R. E. Ginna 007 12 The main turbine failed to trip due to acchanical binding of the trip solenoid.

009 2

Control rod bottom indicating light failed due to relay failure.

009 Incorrect MSIV position indicated due to sticking valve position limit switch.

R. E. Ginna 009 A power fuse in the Nuclear Instrumentation System failed.

011 7

The control rod indicating light failed due to relay problems.

Incorrect MSIV position indicated due to 22 sticking valve position limit switch.

Indian Point 2 006 100-The offsite power source relay for the 6.9KV bus failed.

006 The AFW pump flow switch relay failed.

009 95 The AFW flow control valve controller failed.

009 The moisture separator stop valve failed to open.

016 100 Normal 138KV offsite power was disrupted.

016

. An AFW pump experienced several trips and was declared inoperable.

8

Unit LER Power Description Of Failures Dresden 3 001 85 A group 1 isolation occurred due to vibration of the main steam line Icw pressure switch 018 83 An RPS failure was the result of a stuck contact in the reactor mode switch.

Dresden 3 018 Problem with the SDV was the result of low air header pressure.

Turkey Point 3 023 30 SG level controller problems were encountered and the failure was traced to a pre-existing wiring error, i.e., the power leads feeding the level controller were reversed.

Turkey Point 4 013 100 An inverter supplying power to 120 VAC vital instrument tripped because of a ground in the input filters.

021 100 A pre-existing condition affecting the FW Control Valve contributed to the valve drift condition.

019 100 A failed timer relay in the reactor protection rack generated a continuous turbine runback signal rather than the cyclic on-off runback signal.

Browns Ferry 1 016 100 The RCIC initiated and tripped immediately due to setpoint problems.

Oconee 1 002 100 Main Steam Relief Valves did not reseat after lifting.

005 100 After lifting the plant encountered difficulty reseating the MSSV's.

Unit LER Power Description Of Failures 007 94 After lifting the plant encountered difficulty reseating the MSSV's.

Diablo Canyon 1 001 49 Containment Fan Cooler tripped on thermal overload after an auto start.

011 48 Overpressurization of LP turbine caused perforation of rupture disk.

Leak in the hydrogen cooler led to a 011 small hydrogen burn.

011 During valve lineup a rapid depressuriza-tion occurred in the feedwater system and the water hammer that occurred damaged the feedwater bypass line snubbers.

014 100 Problems were encountered with the diesel b

generator.

(No additional details provided.)

The AFW system piping snubbers were damaged 014 as a result of water hammer.

Surry 1 003 100 A redundant steam generator trip valve did not close.

Olfs 100 The feedwater pump tripped due to failure of the pump recirculation valve.

Pilgrim 006 100 The main generator failed to trip due to a failed control circuit fuse.

Pilgrim 025 37 The reactor high pressure alarm did not annunciate as anticipated.

Unit LER Power Description Of Failures Zion 1 044 0

The Source Range Monitor failed during recove ry.

Crystal River 3 015 20 The main turbine crossover line severed.

023 95 The Main Steam Safety Relief Valve failed to reseat properly after lifting.

025 96 An Atmospheric Dump Valve did not reseat properly.

025 The feedwater control valve would not reseat properly.

026 22 The PORY failed to open on demand.

Main Yankee 007 95 The standby main teedwater pump failed to auto start.

a3 Salem 2 017 100 The atmospheric dump valves did not open automatically as designed.

Rancho Seco 1 50.72 75 The makeup seal pump failed to start.

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D.C. Cook 1 065 78 The feedwater pump turbine would not respond to control signals.

D.C. Cook 2 003 96 The turbine driven AFW pump failed to auto start due to problems with the throttle valve.

035 79 A failed undervoltage assembly caused the reactor trip breaker to fail to open.

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Unit LER Power Description Of Failures Calvert Cliffs 1 008 17 The Main Generator output breaker tripped because of a failed limit switch.

012 100 The Fee 1 water Regulating Valves controller relays failed and caused the valves to take too long to ramp down.

Calvert Cliffs 2 001 100 The Atmospheric Dump Valves opened and failed to reseat.

002 53 An unanticipated Auxiliary Feedwater pump actuation occurred.

Hatch 1 010 61 The RCIC Throttle valve positioner failed in the trip position.

The HPCI Turbine Stop. valve stuck in the mid 010 position.

og 018 100 An errorenous HPCI trip solenoid actuation occurred.

55 An equipment drain pump isolation failed to operate as designed.

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Diablo Canyon 2 007 4

An unanticipated actuation of the containment spray system occurred.

009 30 The Containment Fan Cooling unit tripped on overload due to control position.

010 30 The Feedwater Isolation Valve failed to open.

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Unit LER Power Description Of Failures 010 The Containment Fan Cooling unit tripped on Hi starting current.

016 75 The MSIV's failed to close.

017 1

The reactor trip breaker failed to open.

Diablo Canyon 2 50.72 90 Power was lost to the Reactor Coolant Pumps.

Brunswick 1 059 6

Diesel Generator Number 4 tripped and locked out due to low lube oil pressure due to improperly calibrated diesel start time delay relay.

Brunswick 2 005 100 RCIC Turbine tripped on hi pressure for unidentified reasons.

Sequoyah 2 002 100 A reactor trip breaker failed to open with auto signal.

e 009 100 Reactor trip breaker would not close because of false IRM signal.

Fitzpatrick 019 100 Unanticipated closure of the MSIV's believed caused by faulty signals.

021 90 Due to a failed relay the recirculation pump did not runback to minimum speed.

Beaver Valley 1 015 100 The Safety Injection pumps experienced leakage due to failed 0-rings and gaskets on the control rods in the pump casing.

i 015 An RHR pump failed to start.

018 40 A failure of the 120 VAC bus caused abnormalities in plant systems.

4 Unit LER Power Description Of Failures North Anna 2 006 100 Loss of circulating water caused two main turbine rupture disks to fail and four others to be damaged.

Trojan 002 100 A rupture occurred in the discharge piping of heater drain pump discharge.

009 100 The diesel driven aur.iliary feedwater pump tripped on low suction pressure.

Davis-Besse 1 002 0

The AFW system improperly transferred to the service water supply source.

009 98 The main feedwater pump tripped.

i 011 85 The main feedwater pump tripped for unknown reasons and the auxiliary feedwater pump responded at less than one half capacity.

gg, 013 80 MSIVs closed due to spurious Steam and Feedwater Rupture Control system actuation.

The AFW pumps tripped on overspeed.

013 The PORVs actuated three times and did not 013 reseat at the proper Reactor Coolant System pressure after the third actuation.

t 013 The only operable source range nuclear instrument channel failed to indicate properly when automatically energized after the trip.

Unit LER Power Description Of Failures 013 The AFW system improperly transferred to the service water system.

013 The AFW containment isolation valves did not open automatically.

San Onofre 3 010 96 There was a power loss due to the failure to transfer loads to the reserve auxiliary transformer.

010 The main feedwater pump turbines tripped due to high vibrations; cause unknown.

010 The condensate pump tripped due to improper trip set point.

Farley 095 95 A vacuum signal error caused all atmospheric relief valves to open.

N Hatch 2 030 99 The RCIC system tripped due to turbine mechanical overspeed.

Arkansas 2 004 100 The Steam Dump and Bypass Control System failed to control steam pressure in automatic.

The problem was traced to a loose connection in the master controller.

014 100 A 4.16KV bus failed to transfer because of a failed relay.

015 100 The AFW control valve would not close because of a failed closing coil in the valve operator.

Unit LER Power Description Of Failures 016 100 The AFW control valve failed to open due to a shorted closing coil in the valve operator.

017 100 The AFW control valve failed due to a shorted closing coil in the valve operator.

018 100 The Steam Dump valve did not close at the expected steam pressure.

McGuire 1 004 100 Electronic components failed in the main turbine EHCS. The turbine driven AFW pump discharge check valve failed closed.

004 The PORV failed to open.

[

004 The TDAFW pump discharge check valve failed to close.

McGuire 1 006 94 The feedwater nozzle isolation valve was ES damaged after c20 sing on feedwater isolation.

006 The AFW header check valve failed to close after the pump was secured.

LaSalle 1 017 75 RCIC tripped on overspeed.

017 The Main Steam Safety Relief Valves opened before reactor pressure reached the required set point.

045 64 The RWCS isolated as a result of SRV cycling.

Summer 024 100 The IRM and SRM detectors failed because of an electrical short.

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Unit LER Power Description Of Failures 027 93 A failed air line check valve caused the Feedwater Isolation System not to isolate on demand.

027 93 A pre-existing improper channel calibration caused an IRM channel to fail low.

027 AFW flow transmitter problem traced to improper valve lineup.

WPPSS 2 030 0

Scram Discharge Volume drain valve did not

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close as required.

059 50 Human error caused loss of critical instrument 120 VAC.

Lacrosse

,017 98 The stack monitor and MET equipment lost g,

their nemory due to loss of power.

us 017 Valve stem binding caused circulating water pump discharge valve not to reseat.

Catawba 1 41 6

Three Steam Generator PORV's failed to open due to improper set point calculations.

43 64 SG PORV's did not open at the expected pressure.

45 100 The feedwater heater required isolation because of a failed relief valve.

45 SG PORY did not lift when set point was exceeded.

67 62 Three Main Steam Bypass Valves failed to open.

Unit LER Power Description Of Failures Byron 1 046 50 Voltage fluctuations caused the radiation monitor to enter an interlock condition.

052 50 Pre-existing condition of an improperly set relay on Main Transformer.

052 Unnecessary ESF actuation resulted from voltage transient.

053 13 Loss of power to the instrument power bus caused the isolation of an AFH and MFW train.

Byron 1 078 98 A voltage transient caused the actuation of the Control Room Ventilation System.

River Bend 1 50.72 1

The main feedwater pump discharge valve failed to open due to high delta P.

y; Wolf Creek 041 1

The turbine driven auxiliary feedwater pump tripped on overspeed due to operator error.

049 47 An IRM channel fault would not allow permissive to energize the SRM.

058 86 A failed flow indicator caused the AFW control throttle valve to fail closed.

069 34 The diesel driven fire pump was damaged when the traveling screen collapsed.

Callaway l' 005 100 An auto bus transfer between auxiliary

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transformer and startup transformer failed.

011 1

Startup transformer breaker failed to close due to high contact resistance.

a Unit LER Power Description Of Failures 026 100 A pre-existing undefined condition within the nuclear instrument system was identified by the reactor trip.

034 100 Due to power supply fault one SRM and IRM channel failed.

Palo Verde 1 019 19 The condensate pump mini flow Control Valve caused a pump low suction pressure trip.

063 53 The Volume Control Tank outlet isolation valve closed due to level control system failure.

Palo Verde 1 063 A chiller failed due to a failed solder joint losing refrigant.

E 071 81 A safety injection pump could not be secured due to a failed relay.

Arkansas 1 004 100 The expansion joint for the main feedwater heater cracked at the weld.

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6

t APPENDIX E e

1985 REACTOR TRIPS AT THE SHOREHAM POWER PLANT PAGE OBS NAME DOCKET LER MONTH DAY VENDOR TRIPS POWER 1

SHOREHAM 1 322 5

1 8

GE AUTO O

2 SHOREHAM 1 322 17 4

29 GE AUTO O

3 SHOREHAM 1 322 18

-5 9

GE AUTO O

4 SHOREHAM 1 322 20 5

21 GE AUTO O

5 SHOREHAM 1 322 22 6

6 GE AUTO O

6 SHOREHAM 1 322 24 7

13 GE AUTO 2

7 SHOREHAM 1 322 31 7

24 GE AUTO O

8 SHOREHAM 1 322 35 8

31 GE MANUAL 1

4 9

SHOREHAM 1 322 9

27 GE AUTO O

10 SHOREHAM 1 322 37 9

6 GE AUTO 1

11 SHOREHAM 1 322 42 9

12 GE AUTO 2

12 SHORFHAM 1 322 11 12 GE AUTO O

13 SHCREHAM 1 322 52 11 4

GE AUTO O

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