ML20035C304
| ML20035C304 | |
| Person / Time | |
|---|---|
| Issue date: | 03/30/1993 |
| From: | Novak T NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | Grimes B Office of Nuclear Reactor Regulation |
| References | |
| AEOD-E93-02, NUDOCS 9304070026 | |
| Download: ML20035C304 (4) | |
Text
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MAR 3 01993 t
MEMORANDUM FOR:
Brian K. Grimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation FROM:
Thomas M. Novak, Director Division of Safety Programs Office for Analysis and Evaluation of Operational Data
SUBJECT:
LOSS-0F-OFFSITE-POWER DUE TO PLANT-CENTERED EVENTS Our Engineering Evaluation report on loss-of-offsite-power (LOOP) at nuclear power plants due to plant-centered events is enclosed for your information and use. We initiated this study as a follow up of the March 20, 1990, event at Vogtle which resulted in loss of all ac power when a truck backed into the support column for an offsite power feed to the reserve transformer.
The study covers plant-centered events (events caused by failure or malfunction of equipment or systems inside the plant) involving LOOP at the medium voltage (between 2 kV and 15 kV) Class 1E buses.
The study included both total and partial LOOP events. This enabled a wider evaluation of the subject and a comparison between the total and partial LOOP events.
About 30 percent of the plant-centered LOOP events resulted in total LOOP.
The study identified a total of 86 events between 1985 and 1989.
The study analyzed these events from the points of view of LOOP under power and shutdown operation, root-cause analysis, and corrective measures to reduce the plant-centered LOOP events. Many of these events involved multiple equipment malfunctions or failures, especially during power operation.
Analysis of the 86 events indicated that 48 percent of these events were caused by personnel errors, 28 percent by equipment malfunctions or failures, 14 percent by design deficiencies, and the remaining 10 percent by inadequate maintenance practices. Analysis of the events caused by personnel errors indicated that most of these events could have been avoided by better personnel awareness.
Furthermore, the duration of the event could have been reduced by preplanning for quick restoration of offsite power especially during plant shutdown conditions.
The findings of this report may be of special interest in station blackout and shutdown risk evaluations and in the review of technical specification submitt al s.
Specifically, some of the findings of this study are significantly different from previous findings reported in other studies on LOOP.
For example, report NUREG-1032, June 1988, " Evaluation of Station Blackout Accidents at Nuclear Power Plants," identified that between 1968 and 1985 there were 46 plant-centered total LOOP events with a median restoration '
time of 18 minutes. Many of the conclusions cf NUP.EC-1109, "Psegulatory/
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sis for the Resolution of Unresolved Safety Issue A-44, Station nmmy n mnn g) ut 7:as
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Brian K. Grimes,
i Blackout," are based on fiUREG-1032.
Our study indicates that between 1985 and 1989, there were 26 plant-centered total LOOP events with a median LOOP duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 minutes.
The plant-centered LOOP events identified in this study were compared with LOOP events evaluated in the Accident Sequence Precursor program as a means to assess risk significance.
Twelve events had been evaluated with conditional core damage probabilities in the E-5 to E-4 range.
Thus, plant-centered LOOP events are considered risk significant and precursors to accidents.
If you need additional information regarding the report, please contact Earl Brown of my staff at 301-492-4491, or Subinoy Mazumdar at 301-504-2917.
%y :j +r rj bj Thomas M. flovak, Director Division of Safety Programs Office for Analysis and Evaluation of Operational Data
Enclosure:
As stated Distribution:
PDR LSpessard CGrimes, f4RR Central File KBrockman CBerlinger, f4RR ROAB R/F SRubin JRichardson, f1RR DSP R/F JJohnson SMazumdar, 14RR EBrown KRaglin, TTC Plewis, If1PO JRosenthal RSavio, ACRS DQueener, NOAC EJordan MTaylor, EDO VChexal, EPRI Dross GMarcus, 11RR TNovak WMinners, f4RR VBenaroya Achaffee, I4RR PBaranowsky FCongel, f4RR
- See previous concurrence-ROAB ROAB C:ROAB DSP
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'INovak SMazumdar:mmk EBrown JRosenthal VBey/93 3/f-330/93 12/17/92*
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Brian K. Grimes..'
l Blackout," are based on NUREG-1032.
Our study indicates that between 1985 nd' 1989 there were 26 plant-centered total LOOP events with a median LOOP duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 minutes.
If you need additional information regarding the report, please conta Subinoy Mazumdar of my staff at 492-4308.
Thomas M. Novak, Director Division of Safety Progr ms Office for Analysis an Evaluation of Operational Data
Enclosure:
As stated Distribution:
PDR LSpessard CGrimes, NRR Central File KBrockman Plewis, INPO ROAB R/F SRubin DQueener, N iC DSP R/F PBaranowsky VChexal, E (I SMazumdar KRaglin, TTC EBrown RSavio, ACRS JRosenthal MTaylor, ED0 EJordan GMarcus, NRR Dross WMinners, NRR TNovak AChaffee, NRR VBenaroya FCongel, NRR
- See pr ious concurrence:
y ROAB ROAB ROAB DSP D:DSP SMaz dar:mmk EBrown Rosenthal VBenaroya TNovak 12/ /92*
12/17/92*
2/3/93 2/ /93 2/ /93
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w Brian X. Grimes l Blackout," are based on 14UREG-1032. Our study indicates that between 1985 and l
1989 there were 26 plant-centered total LOOP events with a median LOOP durati n
'of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 minutes, j
If you need additional information regarding the report, please contact binoy Mazumdar of my staff at 492-4308.
.l Thomas M. Novak, Director Division of Safety Progr.s Office for Analysis and valuation of Operational Data -
7
Enclosure:
As stated Distribution:
PDR LSpessard CGrimes, NRR Central File KBrockman Plewis, INP0 ROAB R/F~
SRubin DQueener, N AC DSP R/F PBaranowsky
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JRosenthal MTaylor, EDO l
EJordan GMarcus, NRR i
Dross WMinners, NRR TNovak AChaffee, NRR l
VBenaroya FCongel, NRR i
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ROAB ROABEM C:ROAB DSP D:DSP~
SMazu. ar:rgz EBrowF JRosenthal VBenaroya TNovak 12//7/92 12// 7/92 12/ /92 12/ /92 12/ /92
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AEOD/E93-02 ENGINEERING EVALUATION REPORT EVALUATION OF LOSS-OF-OFFSITE POWER DUE TO PLANT-CENTERED EVENTS MARCH 1993 Prepared by: Subinoy Mazumdar, Ph.D.
Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission
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CONTENTS ABBREVIATIONS v
ABBREVIATIONS USED IN TABLES si 1.
SDDIARY 1
2.
INTRODUCTION
.. 1 3.
DISCUSSION 2
3.1 Identification of Events.
2 3.2 Event Trends.
3 3.3 Event Duration 3
3.4 Total LOOP Events
. 5 3.5 Events Under Power Operation 5
3.6 Events Under Shutdown Condition 7
3.7 Event Causes 8
3.7.1 Events Caused by Personnel Errors S
3.7.2 Events Caused by Equipment Malfunction or Failure 10 3.7.3 Events Caused by Design Deficiencies 10 3.7.4 Events Caused by Inadequate Maintenance Practices.
10 3.8 Failures of Multiple Safety-Related Equipment During LOOP Events..
!l 3.9 LOOP Event Risk Significance I1 4.
FLNDINGS..
12 5.
CONCLUSIONS......
13 6.
REFERENCES 13 iii
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TABLES Table 1. Total Loss-of-Offsite Power Caused by Plant-Centered Events Under Power Operation 15 Table 2. Partial Loss-of-Offsite Power Caused by Plant-Centered Events Under Power Operation 20 Table 3. Total Loss-of-Offsite Power Caused by Plant-Centered Events Under Plant Shutdown Operation.
28 Table 4.
Partial Loss-of-Offsite Power Caused by Plant-Centered Events Under Plant Shutdown Operation 31 Table 5. Duration of Loss-of-Offsite Power 39 Table 6. Analysis of LER Event Causes 40 Table 7. Plant Centered LOOP Event Significance 40 l
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O ABBREVLATIONS ac alternating current ASP Accident Sequence Precursor CCDP conditional core damage probability de direct current EDG emergency diesel generator H.PCI high-pressure coolant injection IIT Incident Investigation Team LOOP loss-of-offsite power NSST normal station service transformer RCP reactor coolant pump RFP reactor feed pump RSST reserve station service transformer SAT startup auxiliary transformer SCSS Sequence Coding and Search System SST station startup transformer SUT startup transformer v
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I ABBREVIATIONS USED IN TABLES AFW auxiliary feedwater ASW auxiliary salt water BFM breaker failure module CAC containment atmosphere control i
CREFS control room essential filtration system CREV control room essential ventilation CT current transformer DD design deficiency ECCS emergency core cooling system EF equipment failure EFP emergency feedwater pump ESF engineered safety feature FBEVS fuel building essential ventilation system h.
hour HPCS high-pressure core spray I
inadvertent LPCI low-pressure coolant injection m.
minute MFRV main feedwater regulating valve MG motor generator MOV motor-operated valve MSIV main steam isolation valve MSSV main steam safety valve MT maintenance NSSSS nuclear steam supply shutoff system P
procedure PE personnel error PORV power-operated relief valve PT potential transformer vi
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ABBREVIATIONS USED IN TABLES (cont.)
RAT reserve auxiliary transformer l
RCIC reactor core isolation cooling i
RPS reactor protection system 7
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seconds SBGT standby gas treatment SI safety injection
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SP set point SRV safety relief valve ST standby transformer i
SWP service water pump l
t TBV turbine bypass valve UAT unit auxiliary transformer USST unit station service transformer r
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1.
SU3 DIARY This report documents AEOD's investigation of loss-of-offsite power (LOOP) events at f
United States nuclear plants. The study addresses LOOP events at medium voltage (between 2 kV and 15 kV) Class IE buses causea by malfunction or failure of equipment or systems t
inside the plant, henceforward called plant-centered LOOP events.
A data search identitled 86 plant-centered total or partial LOOP events between 1985 and f
1989. Many of these events involved multiple equipment malfunctions or failures, especially
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during power operation. Analysis of these 86 LOOP events indicates the events could be r
grouped in four cause categories with 48 percent of these events related to personnel errors, 28 percent related to equipment malfunctions or failures,14 percent related to design deficiencies, ar.d the remaining 10 percent rela:ed to inadequate maintenance practices.
Analysis of the events caused by personnel errors indicates that most of these events could have been avoided by better personnel awareness. This study indicates that between 1985 and 1989, there were 26 plant-centered total LOOP events with a median LOOP duration time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 minutes (Tables 1 and 3) as compared to previously reported 18 minute median restoration times. The duration of the LOOP could have been reduced by preplanning for quick restoration of offsite power especially during shutdown operations.
Several 5t-centered LOOP events identified in the study had been evaluated in the Accident. quence Precursor (ASP) program. The conditional core damage probabilities (CCDPs) were in the E-4 range indicating the plant-centered LOOP events are considered risk significant and precursors to accidents.
2.
INTRODUCTION The importance of reliable offsite power supplies for safe operation of nuclear plants is well known to the industry. The minimum requirements for c%ite power are specified in 10 CFR 50, Appendix A (Reference 1) while plant technich specifications (TS) cover I
permissible plant operating conditions. The guidance for staff review of offsite power is contained in Section 8.2 of the standard review plan (Reference 2). Tables 1 through 4 contain information about the 86 specific plant-centered LOOP events. Table I lists 15 events where a total LOOP occurred with the nuclear plant at power. The 26 events listed in j
Table 2 were partial LOOPS that occurred when the plants were at power. Table 3 presents i
11 events of total LOOPS that occurred during plant shutdown, and Table 4 lists 34 events j
where a partial LOCP occurred during plant shutdown. Each table lists in Column 2 the j
plant name, docket number /LER c,ambwr, and the event duration. The remaining columns provide a description of the event, the cause, failed equipment, corrective actions and the event significance.
Several reports have been published on LOOP events at nuclear plants. Many of these reports are identified in References 3 through 8. Reference 3 documents total and partial LOOP events that occurred through the end of 1983 categorizing the LOOP events as plant-centered or grid-centere?. Grid centered events are the events caused by malfunctions or failures of equipment or systems outside the plant. Reference 3 also indicates that about 70
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i percent of all LOOP events were caused by plant-centered events. Based on causes of the LOOP events, Reference 3 further categorizes the causes of the LOOP events into weather i
related. human error, design error, and hardware failure. However, Reference 3 did not i
l analyze the LOOP events from the point of view of event duration. event cause, failed i
equipment, corrective action, and safety significance.
Reference 4 repcrts on the risk of core damage from external events such as nearby
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industrial or military facility accidents, onsite hazardous material storage accidents, severe temperature transients, severe weather storms, lightning strikes, external Eres, extraterrestrial activity, volcanic activitt. earth movement, and abrasive windstorms. Reference 5 tabulates i
recovery time of offsite.: power from LOOP events into plant-centered, grid-centered, and weather-related categories. Reference 6 tabulates the LOOP events between 1975 and 1989 classifying them into five categories based upon duration and availability of power sources.
Reference 7 provides an assessment of the major contributor to the frequency of station j
blackout and the probability of subsequent core damage. Reference 7 reported that between 1968 and 1985 there were 46 plant-centered total LOOP events with a median restoration time of 18 minutes. These studies focused on estimating the frequency of events and associated risk. However, these studies did not analyze the LOOP events from the point of view of event duration, event cause, failed equipment, corrective action, and safety j
significance which hase been included in this report.
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An NRC Incident Investigation Team (IIT) report (Reference 8) on failure of all ac power to i
safety-related loads at Vogtle Unit 1 on March 20,1990, and the Ending in Reference 3 that about 70 percent of all LOOP events are caused by plant-centered events provided the impetus for this engineering study on LOOP events caused by plant-centered problems.
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This study covers only plant-centered LOOP events involving total or partial loss of power at l
4 medium-voltage (between 2 kV and 15 kV) Class IE buses; it does not cover grid-centered l
l events that involve LOOPS caused by loss of transmission lines and equipment in the high-l l
soltage switchyard. However, the study includes LOOP events caused by malfunctions or failures of the main transformer, station auxiliary transformers (SATs), and startup i
a transformers (SUTs), including associated protective relays, bus ducts, current transformers, potential transformers, and lightning arresters. At most plants, such equipment is located inside the plant boundary and not in the switchyard. This study analyzes plant-centered j _
LOOP events from the point of view of total and partial LOOP events, LOOP duration under power and shutdown operation, root cause analysis, and corrective measures to reduce the plant-centered LOOP events.
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3.
DISCUSSION 3.1 Identification of Events
-l The Sequence Coding and Search System (SCSS) was used to conduct multiple searches of i
3-the operating experience database to identify the plant-centered LOOP events which occurred from 1985 through 1989. The staff also screened all LERs and daily event reports in 1989 to identify LOOP events caused by plant-centered failures. l i
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Review of the above data identified 86 LERs related to plant-centered LOOP events. A i
listing of these LERs with the plant name and a brief description of the event, event causes, failed equipment, corrective actions, and safety significance is given in Tables 1 through 4 Table 1 lists the total LOOP events that occurred while the plant was at power, Table 2 lists j
the partial LOOP events while the plant was at power, Table 3 lists the total LOOP events during shutdown operation, and Table 4 lists the partial LOOP events during shutdown i
operation. Because of the limitations of the LER search process, the listed LERs are i
representative events of interest, but the lists can not be considered all inclusive.
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1 3.2 Esent Trends Grouping the LOOP events by the year of occurrence indicated that 31 percent of these
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events occurred in 1989, 20 percent in 1988, 20 percent in 1987, 17 percent in 1986, and 12 percent in 1985. The number of LOOP events per year from 1985 through 1988 changed sery little except for events attributed to new plants commissioned during this period.
i However, the increase in LOOP events during 1989 can be attributed to the manual screening of all LERs reported in 1989 as opposed to the SCSS screening for previous years. Another reason for the increase in LOOP events in 1989 is that the newer units experienced higher failure rates. For example, out of 27 events occurring in 1989, four events occurred at South Texas Unit 2 which began commercial power operation in June 1988, and 4 events occurred at Palo Verde Units 1,2, and 3, which began commercial power operation between January 1986 and January 1988. River Bend began commercial power operation in June 1986 and four events occurred at River Bend in 1988 and 1989.
3.3 Event Duration The salient features of LOOP durations recorded in Tables 1 through 4 are reproduced in Table 5. Most of the LOOP durations were extracted from the LERs. When the duration was not stated in the LERs. it was obtair.ed from the licensee. The licensees were also contacted for all LOOP events that lasted more than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to find out why it took that long to restore offsite power.
Out of the 86 events, on 29 occasions (34 percent of the 86 events) the LOOP duration was less than 30 seconds. The average duration for all events was 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 42 seconds. On nine occasions, the LOOP durations were more than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. When these nine events are excluded, the average LOOP duration comes to I hour and 20 minutes. When the 29 momentary losses are excluded, the average LOOP duration was 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 10 minutes.
The longest LOOP event (Table 2, No.13) was a partial LOOP at LaSalle Unit 1, LER 50-373/89-009, for 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> and 28 minutes. It occurred at 4.16 kV Class IE bus 243 i
which is normally fed from offsite power through startup auxiliary transformer (SAT) 2. On i
failure of the power supply from SAT 2, bus 243 is powered by emergency diesel generator 1
(EDG) 2B. There is no provision to energize this bus from any other source. On March 2, 1989, there was a snow storm around the plant area. A lightning arrester on the primary side of SAT 2 failed resulting in opening of the breaker feeding power to SAT 2. The EDG..
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2B started and powered bus 243. The licensee explained that a combination of factors contributed to the long restoration time. This included the inclement weather, the time required to check SAT 2 including high potential test and oil sample tests, the time required to get the replacement lightning arrester from storage about 70 miles away, the time required to mount the new arrester, and allow enough time for the oil inside the arrester to stabilize before the arrester could be energized.
The second longest LOOP event (Table 2, No.14) was also a partial LOOP at 4.16 kV bus 243 at 12Salle Unit 2. LER 50-373/89-007. The duration was 51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br /> and 5 minutes.
l In this event, the SAT 2 was lost because of an inadvertent actuation of the SAT 2 deluge system. Similar to the LaSalle Unit i event, the licensee took time to check the associated I
equipment prior to energizing bus 243 from the offsite power.
The licensee was questioned about the duration of these two events at LaSalle. The licensee explained that during these two events there was li te they could do to significantly reduce the LOOP duration. However. the LOOP durations of these two events were significantly more than the LOOP duration of any other event.
The third longest LOOP event (Table 1, No.11) was a total LOOP for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 51 minutes at Palo Verde Unit 2, LER 50-529/89-01, on January 3,1989. Each Palo Verde Unit has two Class lE buses which are normally fed from the 13.8 kV buses through two 13.8 kV to 4.16 kV transformers. However, because of rain and pollution, two bushings of each transformer flashed over simultaneously resulting in a total LOOP. The licensee attributed the LOOP duration to weather, shift turnover, and the time required to replace the bushings in sequence.
The next longest LOOP event (Table 3, No. 6) was a total LOOP at Pilgrim on February 21, 1989. LER 50-293/89-010. At this plant the Class IE buses can be fed from the 345 kV offsite power through the SUT or from the 23 kV offsite power through the shutdown transformer or from the 345 kV offsite power through the main transformer and the SAT.
At the time of the event, the plant was under shutdown operation, the shutdown transformer was undergoing maintenance, and the auxiliary loads were fed through the SUT. A cable fault occurred on the secondary side of the SUT which tripped the SUT resulting in a LOOP.
It took the licensee 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 20 minutes to restore the offsite power through the main transformer and SAT. The licensee mentioned that the main cause for the delay was the time taken to prepare for equipment tagging, tag inspection, removal of tags, and for shift turnover.
Another event of interest was a total LOOP event (Table 1, No.10) at Palisade:, on July 14, 1987, LER 50-255/87-024. On this occasion, with the reactor operating at 91 percent power, maintenance work was in progress on the transformer deluge system. A sudden actuation of the deluge system occurred which created a fault on a startup transformer resulting in a LOOP. The licensee took 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 26 minutes to tag and inspect the equipment before restoring the offsite power. Subsequently, the licensee added another offsite power feed through an underground cable with automatic bus transfer provision and replaced the generator disconnect link with a mMor operated disconnect. With these
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l modifications the licensee now has three redundant offsite sources with provision for automatic fast and slow bus transfers.
1 3.4 Total LOOP Events The staff review of the 86 LERs indicated that 26 events (30 percent of all plant-centered events) resulted in total LOOP (Tables 1 and 3). Fifteen events occurred with the plant under power operation and 11 events occurred while the plant was under shutdown operation.
The longest total LOOP occurred at Palo Verde Unit 2, LER 50-529/89-001 (Table 1 No.11). This LOOP was for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 51 minutes. The second longest total LOOP occurred at Pilgrim, LER 50-293/89-010 (Table 3, No. 6), LOOP duration 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and 20 minutes. These two events have been described in section 3.3.
A review of the average LOOP duration recorded in Table 5 indicates that under power operation the average total LOOP duration was about twice the average total LOOP duration l
under plant shutdown conditions. This can be justified by the fact that under plant shutdown conditions the plant loads are much less and the plants operate on a simpler power supply configuration.
The total LOOP duration acquired for this study, which covered 1985 to 1989, was compared with the station blackout study report, NUREG-1032 (Reference 7), data which covered 1968 to 1985. There were 26 total LOOP events during the recent 5 year span and 46 total LOOP events during the prior 18 year span. The event median duration from 1985 to 1989 was 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 minutes (time from start of the LOOP to power restoration) while the station blackout study indicated a median duration of 18 minutes. We understand the station blackout study included an evaluation to estimate when power could have been i
restored and that was the event duration reported. Many conclusions of NUREG-1109
" Regulatory /Backfit Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout" are based on NUREG-1032. The recent median LOOP duration data of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 minutes could mean the LOOP events are actually longer duration events than previously thought or thst the prudent course of action involves more effort to assure restoration attempts will succeed. Since operating experience shows the actual licensee restoration time is much longer than the estimated restoration time, a review of this effect on the resolution to unresolved safety issue A-44, Station Blackout, may be warranted.
Fifty-two percent of the total LOOP events were caused by personnel errors, 25 percent by equipment malfunctions or failures,18 percent by design deficiencies, and the remaining 5 percent by inadequate maintenance practices.
3.5 Events Under Power Operation Forty-one LOOP events (48 percent of the 86 LOOP events) occurred with the plant under power operation (Tables 1 and 2). Almost all the events in Tables 1 and 2 involved multiple equipment or system malfunctions or failures including bus transfer failures and severe perturbations in the electrical system.
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One example of multiple equipment failure is the event (Tabie 1, No. 5) at Diablo Canyon Unit 2, LER 50-323/88-008, on July 17, 1988. During this event, with the plant operating at full power, ground alarms were received from one reactor coolant pump (RCP) and two circulating water pumps fed from two 12 kV buses D and E. Soon ground alarms were received from the secondary side of the 25 kV to 12 kV auxiliary transformer feeding these i
two buses from the main generator. The operators transferred the 12 kV buses from the auxiliary transformer to the stanup transformer. Then a phase unbalance alarm was received l
i from the RCP. The operators reduced the reactor power to 50 percent and tripped one l
circulating water pump motor. Soon after that a nre was reponed from the stanup l
transformer grounding resistor. The operators tripped the reactor and the RCP motor. Soon l
after that the 230 kV breaker feeding the startup transformer tripped causing a total LOOP.
The 6re in the vicinity of the grounding resistor was caused by the burning of insulation of a r
cable that connected the grounding transformer to the grounding resistor. A sheet of micana type material had been inadvertently left on the resistor banks directing heat from the resistor
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banks to the cable which had been routed over the resistor banks. The Ere created a ground fault on the secondary side of the grounding transformer which blew two fuses on the primary side of the grounding transformer. The blown fuses left the 12 kV system ungrounded and created a voltage transient that resulted in a phase to phase fault and tripped
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the 230 kV breaker feeding the stanup transformer. The RCP 2-3 was the last RCP to be j
tripped and at that stage its associated steam generator SG 2-3 was the main heat sink. This resulted in SG 2-3 being cooled off more than the other steam generators which ultimately l
depressurized it to the differential pressure set point which initiated safety injection. In addition, a secondary system trausient produced water hammer in the condensate and the feedwater systems, flashing in the feedwater heater inlet and outlet piping, and l
depressurization in lines connected to the condenser hotwell. Also, the coinpressed air system pressure temporarily fell below normal pressure. The first ground fault alarm was caused by a deteriorating RCP motor terminal connectors with galled aluminum threads.
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This event was quantified in the ASP program to have a CCDP of 4.lE-5 (See Section 3.9).
Another example of a LOOP under power condition with mdltiple complications is the event (Table 2, No. 8) at Dresden Unit 3, LER 50-249/89-001, on March 25, 1989. On this i
occasion, a capacitor inside a 345 kV switchyard circuit breaker failed. The breaker 1
received a trip signal but did not trip. The breaker backup logic tripped' additional breakers i
that isolated the 345 kV power feed to the reserve auxiliary transformer, TR 32. The fast bus transfer was inhibited because of dirty breaker contacts. _A slow bus transfer took about 14 seconds which tripped a reactor feed pump (RFP) motor and a reactor recirculation pump motor. The reactor water level rose to the main turbine and the RFP trip set points. The reactor scrammed on turbine stop valve closure. The operators manually closed the main steam isolation valves and used mildly contaminated condensate water because the clean demineralized water shell side supply valve was deenergized. This resulted in low level contamination around the isolation condenser. Furthermore, the power supply to an i
annunciator panel tripped due to a fuse failure requiring declaration of a generating station emergency alen. Also the two feeder breakers to the low pressure coolant injection swing bus tripped spuriously and the reserve feed breaker for the swing bus failed to close. In addition, the high-pressure coolant injection (HPCI) pump experienced HPCI high pressure beanng oil drain high temperature because a HPCI lube oil cooling water normal return valve opened. The HPCI turning gear was unavailable following the trip of the HPCI J
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turbine because of turning gear motor failure. The security multiplexer was unavailable for about I hour because of an open power supply fuse. The security computer was without power for about 17 minutes because the security system uninterruptible power supply failed.
The primary containment oxygen analyzer was unavailable for a short period. Also, the main turbine turning gear was inoperable because of the loss of the instrument air system.
This event was quanti 6ed in the ASP program to have a CCDP of 1.3E-5 (See Section 3.9).
Similar multiple equipment or system malfunctions or failures occurred during most of the LOOP events listed in Tables I and 2. A review of these events indicates that most of the secondary failures during these events were caused by unrecognized design deficiencies, sudden equipment failure on demand, and previously failed components that remained undetected. The secondary failures were at times associated with personnel errors.
Forty percent of the LOOP events under power operation were caused by personnel errors, 30 percent by equipment malfunctions or failures, 20 percent by design de6ciencies, and the remaining 10 percent by inadequate maintenance practices.
3.6 Events Under Shutdown Condition Fifty-two percent of the LOOP events (Tables 3 and 4) occurred while the units were
[
shutdown. In the shutdown condition, the unit auxiliary loads were often fed from one offsite source only. The standard TS permit nuclear unit operation in shutdown modes (modes 5 and 6) with power supplied from only one offsite source and one onsite source (usually an EDG).
i During the shutdown operations or refueling operations the licensees usually did not make contingency plans for quick restoration of alternate offsite power source (s) if the Erst offsite power source was lost. Because of this lack of preplanning for quick restoration of offsite power, in a number of cases under shutdown conditions when the first offsite source was lost, the Class IE loads were powered by the diesel generator (s) for hours and the licensees were slow in restoring the available offsite power. With this approach, the plants were i
without redundant power supply for hours. From a plant safety point of" view, it is desirable to switch over to an alternate offsite power source (s) at the earliest opportunity to minimize relying only on the EDGs.
9 A review of the single line diagrams of all operating nuclear plants indicates that about 60 percent of the plants have more than one offsite power source available under shutdown conditions. With adequate preplanning, most plant licensees. would be able to quickly switch to an alternate offsite power source. On October 28,1991, the NRC issued an Information i
Notice (Reference 10) on this subject.
Furthermore, because the licensees tried to minimize the outage duration and conduct as much maintenance, inspection, testing, and replacement as possible during the shutdown periods, many activities were performed without adequate planning and supervision. At times, people were not adequately quali6ed for the job and, on occasions, vehicle and
- i personnel movement was not adequately controlled. The Vogtle event of March 20, 1990, i
O O
(Reference 8) demonstrates this point. In this event, with the plant in mid-loop operation, all ac power to safety related loads and the residual heat removal system were lost when a truck
[
in the low voltage switchyard backed into the support column for the offsite power feed to the reserve transformer. The event at Diablo Canyon Unit 1 on March 7,1991, (Reference-11), when a mobile crane caused flash-over of a 500 kV line inside the power plant is another example of poor job execution during plant refueling. Another example (Table 3, i
No. 2) is the event at Crystal River Unit 3 on October 16,1987, LER 50-302/87-025, when a worker inside the plant touched a 230 kV feeder with a metal pole. On February 18, 1992, the NRC issued an Information Notice, IN 92-13, (Reference 14) on this subject.
During these events, the licensees had no provision for quick restoration of alternate offsite I
power in the event the available offsite power source was lost due to a mishap.
t Fifty-seven percent of the LOOP events under shutdown condition were caused by personnel errors, 23 percent by equipment malfunctions or failures,10 percent by design deficiencies, and the remaining 10 percent by inadequate maintenance practices.
Thus, adequate preplanning for quick restoration of offsite power and improvement in personnel awareness are likely to sigmticantly reduce LOOP occurrences and their duration during plant shutdown operation.
J 3.7 Event Causes i
Staff analysis of the 86 total and partial LOOP event causes are summarized in Table 6. The event causes reported in the LERs have been categorized into four broad classes, namely (1) personnel errors, (2) equipment malfunction or failures, (3) design deficiencies, and (4) l inadequate maintenance practices. Personnel errors have been further subdivided into inadvertent action, procedure inadequacies, and set point errors. Many of the LERs reported multiple causes for each event. Addition of these multiple causes for the 86 LERs resulted in 111 event causes. Multiple event cause basis has been used for all event cause analysis discussed in this report. Thus, for the 86 events included in this study,53 event causes relate to personnel errors (48 percent of 111 total event causes), 31 event causes to equipment malfunctions or failures (28 percent),16 event causes to design deficiencies (14 t'
percent), and 11 event causes to inadequate maintenance practices (10 percent).
i 3.7.1 Events Caused by Personnel Errors Table 6 indicates that personnel error was the prime cause of all LOOP events (48 percent of all LOOP event causes). A review of the events caused by personnel error (Tables 1 through 4) indicates that about 9 out of 10 times these personnel errors involved bumping j
panels. slamming panel doors, producing vibration while working on an adjoining panel, operating a wrong switch, pulling the wrong fuse, failing to remove temporary jumpers, making wrong connections, and using wrong devices. Another six percent of the personnel errors were caused by procedure deficiency, failure to follow procedure or by poor communication. The remaining three percent of the personnel errors were caused by device set point error, t
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Staff investigation of the LOOP events caused by personnel errors indicates that many of these events could have been avoided oy adequate personnel and procedure controls. For i
example, the event (Table 1, No. 3) at Brunswick Unit 2 (LER 50-324/89-009) could have
~
been avoided if the technician doing the trouble shooting had understood the basic features of the electrical distribution system and if his work had been properly supervised by a qualified person. In this event the technician shorted a neutral grounding transformer because he thought a grounding transformer was the same as a CT. At Brunswick Unit 2, the neutral of the station startup transformer (SST) is grounded through a grounding transformer with a resistor on the secondary side of the trar.sformer. The technician was trouble shooting an t
SST ground fault alarm. In an effort to clear the alarm, the technician decided to put a jumper around the grounding transformer primary which created a low resistance ground path for an existing ground causing severe damage to the 4160 V bus duct. This resulted in reactor scram, turbine trip, and total LOOP for the unit.
A second example of a LOOP event (Table 2, No.1) due to personnel error is discussed in LER 50-313/87-005 for Arkansas Nuclear Unit 1. This event involved testing a generator high voltage breaker. Prior to the test, the relay test personnel discussed the need to take out the breaker failure module. However, during the test, the relay test personnel failed to pull out the breaker failure module. This resulted in opening four 500 kV switchyard breakers, generator trip, and drop out of loads due to slow bus transfers.
l r
A third example (Table 3 No.7) of LOOP due to personnel error is the Shoreham event on March 18,1987, LER 50-322/87-003. In this event, a condensate pump start first tripped c
the normal station service transformer (NSST) which was followed by tripping of the reserve station service transformer (RSST) resulting in a total LOOP. On loss of the offsite power.
l the reactor tripped along with isolation of various safety systems. This event was caused by l
adding jumpers across the NSST and RSST CT secondaries during modification work for l
testing an EDG without studying the consequences of adding these jumpers.
]
Another similar LOOP event (Table 1, No. 9) occurred at Millstone Unit 2 on October 25, 1988, LER 50-336/88-011, because maintenance personnel grounded a 4160 V safety related bus by inserting a wrong " ground and test device."
LER 50-255/87-024, submitted by the licensee for the Palisades Nuclear Plant (Table 1, No.10), discusses an LOOP event caused by a transformer fault when maintenance personnel j
inadvertently actuated the deluge system during planned maintenance.
A fatal accident (Table 3 No. I1) occurred at Wolf Creek Unit 1, LER 50-482/87-048, because the individuals involved failed to follow the procedure to check the power supply before working on a 4.16 kV switchgear.
The Vogtle event of March 20,1990, (Reference 8), and the Diablo Canyon event of March 7,1991, (Reference 11), clearly demonstrate the results of lack of control over vehicle movement inside nuclear plants.
On 16 occasions the relays and other devices malfunctioned from vibrations caused by panel tapping, slamming of panel doors, working on adjoining panels, and accidental bumping. On O
O 14 occasions personnel opened the wrong device (e.g., the fuse or switch) or made incorrect connections or failed to remove jumpers installed during testing.
3.7.2 Events Caused by Equipment Malfunction or Failure As indicated in Table 6,31 of the events were caused by equipment malfunctions or failures.
A review of these events indicated that 12 of these events (36 percent) involved transformers, 6 events (18 percent) involved relays 4 events (12 percent) involved switchgear, 3 events (9 percent) involved generators, and 2 events (6 percent) involved fuses. In the majority of these events, the licensee could not established the exact cause of the equipment failures.
3.7.3 Events Caused by Design Deficiencies LOOP events caused by design deficiencies were often related to inadequate bus transfers.
About 90 percent of the LOOP events under power operation (Tables 1 and 2) were accompanied by bus transfer failures. The different aspects of bus transfer improvements were reported in Reference 9. In 1991 the NRC issued an Information Notice (Reference
- 12) on this subject.
One example of LOOP caused by design deficiency (DD) is the event (Table 4 No. 23) reported in LER 50-499/89-001 for South Texas Unit 2. In this event a LOOP occurred due to an inadvertent actuation of the standby transformer deluge system. Corrective action included repositioning the deluge system nozzles to lessen the deluge water impingement on the transformer bushings.
i A second example of LOOP caused by deluge actuation is the event (Table 1, No.10) reported in LER 50-255/87-024 at Palisades. During this event the LOOP lasted for 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 26 minutes. Other examples of LOOP caused by deluge actuation are the events reported in LER 50-301/89-002 at Point Beach Unit 2 (Table 1, No.12), LER 50-483/
85-011 at Callaway Unit 1 (Table 2, No. 4), and LER 50-456/87-048 at Braidwood Unit 1 (Table 3, No.1).
Another common DD is the degraded voltage or inadequate relay setting as reported in LER 50-311/86-007 at Salem Unit 2 (Table 1, No.14), LER 50-395/89-008 at Summer (Table 1, No.15), LER 50-369/86-011 at McGuire (Table 4, No. I1), LER 50-313/87-005 at Arkansas Unit 1 (Table 2, No,1), and LER 50-249/89-001 at Dresden Unit 3 (Table 2.
No. 8).
3.7.4 Events Caused by inadequate Maintenance Practices Ten percent of all event causes relate to inadequate maintenance practices. The most common cause is the accumulation of dust and water inside bus ducts as reported in LERs 50-324/89-009 at Brunswick Unit 2 (Table 1, No. 3), LER 50-268/89-008 at Browns Ferry Unit 2 (Table 4, No. 3). LER 50-528/88-003 at Palo Verde Unit 1 (Table 4, No.16), i l
1
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i and LER 50-328/88-034 at Sequoyah Unit 2 (Table 4, No. 21). The LER 50-255/87-024 reported on contamination of transformer deluge system at Palisades (Table 1, No.10) and the LER 50-280/89-005 on breaker failure due to accumulation of dust and dirt at Surry Unit 1 (Table 1, No. 27).
3.8 Failures of Multiple Safety-Related Equipment During LOOP Events A review of the LERs on LOOP events indicated that often multiple equipment failed or malfunctioned during the LOOP events, at times causing concern for plant safety. Such multiple equipment failures were more prevalent under power operation (Tables 1 and 2). In many cases such LOOP events involved voltage transients in the plant medium voltage system which prevented bus transfers and caused load shedding. In a number of cases one or more Class lE equipment failed to start on demand. In a few cases, the failed equipment had become inoperable earlier but remained unidentified until the LOOP event occurred.
The LOOP events at Diablo Canyon and Dresden described in Section 3.5 clearly indicate how involved multiple equipment failures can be under power operation. Another similar event is the recent failure of uninterruptible power supplies following a catastrophic failure of the main transformer at Nine Mile Point Unit 2, on August 13,1991, (Reference 13).
One example of such an event under plant shutdown operation was the Vogtle event of March 20,1990, (Reference 8) in which both of the EDGs failed to start on a LOOP resulting in a station blackout. Another example is the LOOP event subsequent to the main e
transformer failure at Diablo Canyon on March 7,1991, (Reference 11). During this LOOP 5
event multiple equipment failed including emergency lighting and the plant communication systems.
Our review of the 86 LERs indicated that about 29 of these events (34 percent) involved LOOP events associated with malfunction of multiple equipment or systems.
3.9 LOOP Event Risk Significance The events identified in this study were compared with previous LOOP events evaluated in the ASP program as a means to establish a measure of risk significance associated with plant-cemered LOOP events. The ASP program evaluated 25 LOOP events of all kinds from 1985 through 1989. Twelve of those events were identified in this study as plant-centered LOOP events. Table 7 identifies the distribution of plant-centerd events that received an ASP evaluation, the plant condition (power or shutdown) when the event occurred, and the maximum ASP conditional core damage probability (CCDP) that was calculated.
Three of the eight ASP evaluations involving plant-centered total LOOP events resulted in CCDP around E-4 with a maximum predicted value of 4.3E-4. Under shutdown conditions, three plant-centered total LOOP event CCDP were around E-5 with a maximum of 7E-5.
The maximum CCDP for any of the 25 ASP LOOP event evaluations was approximately E-3 l
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followed by 9 events with a CCDP in the E-4 range. Thus, plant-centered LOOP events are considered risk significant, and as precursors to accidents.
l 4.
FINDINGS The important findings from evaluation of the 86 total or partial plant-centered LOOP events between 1985 and 1989 are as follows:
A.
LOOP Duration The average duration of the 86 LOOP events included in this study was 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 42 seconds. Out of the S6 events, on 29 occasions (34 percent of the 86 events) the LOOP duration was less than 30 seconds. When the 29 momentary losses are excluded the average duration was 5 hcurs and 10 minutes. On nine occasions the LOOP duration was more than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The longest duration was 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> and 28 minutes.
B.
Total LOOP Events The staff review of the 86 events indicated that 26 events (30 percent of the 86 events) resulted in total LOOP. Under power operation the average total LOOP duration was about twice the average total LOOP duration under plant shutdown condition. The median LOOP duration was 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 minutes.
Fifty-two percent of the total LOOP events were caused by personnel errors, 25 percent by equipment malfunctions or failures,18 percent by design deficiencies, and the remaining 5 percent by inadequate maintenance practices.
C.
Events Under Power Operation Forty-one LOOP events (48 percent of the 86 events) occurred with the plant under power operation. Most LOOP events under power operation (about one-third of the 86 events) involved multiple equipment or system malfunctions or failures including bus transfer failures and severe perturbations in the electrical system.
Forty percent of the LOCP events under power operation were caused by personnel errors, 30 percent by equipment malfunctions or failures, 20 percent by design deficiencies, and the remaining 10 percent by inadequate maintenance practices.
D.
Events Under Shutdown Condition During the shutdown periods, most licensees did little or no preplanning for quick restoration of offsite power on loss of the first offsite power source. Also, during the shutdown periods, many activities were performed without adequate planning and supervision. At times, people were not adequately qualified for the job and, on occasions, vehicle and personnel movement was not adequately controllei
1 O
O-r Fifty-seven percent of the LOOP events under shutdown condition were caused by personnel errors. 23 percent by equipment malfunctions or failures,10 percent by design de6ciencies, and the remaining 10 percent by inadequate maintenance practices.
E.
Event Causes The study identi6ed that 48 percent of the 86 events were caused by personnel errors, 28 percent by equipment malfunctions or failures,14 percent by design de6ciencies, and the remaining 10 percent by inadequate maintenance practices.
A review of the events caused by personnel errors indicates that about 9 out of 10 times these persormel errors involved bumping panels, slamming panel doors,' producing vibration while working on an adjoining panel, operating a wrong switch, pulling the wrong fuse, failure to remove temporary jumpers, making wrong connections, and using wrong devices.
Under power operation the average total LOOP duration was about twice the average total LOOP duration under plant shutdown condition.
5.
CONCLUSIONS The current rate of LOOP caused by plant-centered events can be reduced by:
A.
Improvement in administrative controls to ensure that persons conducting modi 6 cations, maintenance, and testing are knowledgeable and careful about their work and that these activities are properly scheduled, controlled, and supervised.
B.
Improvement in the alertness of people carrying out maintenance and testing of plant equipment.
C.
Improvement in control of personnel and vehicle movement inside the plant area.
D.
Preplanning for quick restoration of offsite power in the event the available power source is lost due to a mishap.
E.
Improvement of the plant auxiliary power distribution and bus transfer schemes.
6.
REFERENCES 1.
Of6ce of the Federal Register, National Archives and Records Administration, Title -
10 of the Code of Federal Regulations, " Energy," revised annually.
2.
U. S. Nuclear Regulatory Commission (U.S. NRC), NUREG-0800, July 1981,
" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition." 1
b
.)
v t
3.
R. E. Battle, NUREG/CR,3992, February 1985, " Collection and Evaluation of
[
Complete and Partial Losses of Offsite Power at Nuclear Power Plants."
4.
C. Y. Yimura and P. G. Prassienos NUREG/CR-5042, February 1989, " Evaluation of External Hazards to Nuclear Power Plants in the United States."
5.
R. L. Iman and S. C. Hora, NUREG/CR-5032, January 1988, "Modeling Time to Recovery and Initiating Event Frequency for oss of Offsite Power Incidents at Nuclear Power Plants."
i 6.
H. Wyckoff, Nuclear Safety Analysis Center, NSAC/147, March 1990, " Losses of Off-Site Power at U. S. Nuclear Power Plants through 1989."
7.
P. W. Baranowsky, U. S. Nuclear Regulatory Commission, NUREG-1032, June 1988, " Evaluation of Station Blackout Accidents at Nuclear Power Plants."
8.
NUREG-1410, June 1990, " Loss of Vital AC Power and the Residual Heat Removal f
System During Mid-Loop Operations at Vogtle Unit 1 on March 20, 1990."
9.
Engineering Report AEOD/E90-05, " Operational Experience on Bus Transfers."
{
10.
Information Notice, IN 91-68, " Careful Planning Significantly Reduces the Potential Adverse Impacts of Loss of Offsite Power Events During Shutdown".
i 11.
Pacific Gas and Electric Company, Licensee Event Report 50-275/91-004, Diablo Canyon Unit 1, March 7,1991.
12.
Information Notice, IN 91-57, " Operational Experience on Bus Transfers."
l 13.
Niagara Mohawk Power Corp., License Event Report, 50-410/91-017, Nine Mile Point Unit 2, August 13, 1991.
14.
Information Notice, IN 92-13, " Inadequate Control Over Vehicular Traffic at Nuclear Power Plant Sites."
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Table I. (cont.)
Failed No.
Plant Data Event Description Cause Equipment Correctise Actions Esens Significance 5.
Diablo Canyon 2 Galled aluminum threads LF: the cause of the Cable and Neutral cable for SU l'
'I he eleventh ECCS 323/88-003 caused a ground fault at galled threads could not connector rerouted. Sur ge actuation resulted in a 2h. 22m.
RCP 2-2. The operator be established. PE-l:
suppressors on 12 LV discharge of water into transferred 12 kV bus D the fire was caused by vacuum breakers the reactor. The EDGs to the SUT. A fire leftover mica sheet on replaced. l he security started and powered developed at SUT grounding resistor.
alarm system will be the Class IE buses.
grounding resistor evaluated.
Some condenser tube causing a ground fault plugs were blown out h which resulted in a total by the secondary LOOP. Si occurred. The transients.
secondary system transients produced water hammers in the condensate and feedwater systems.1 he compressed air system
[
pressure fell below nor mal level.
i 6.
Dresden 2 A fault on R AT TRl2, EF: A fault on the Transformer TR21 repaired. Ilus EDGs started and 237/85-034 tripped R AT TR22. Ilus secondary side of TRl2 TR21 and transfer circuit powered the Class IE 4h. 36m.
transfer to TR21 failed.
initiated this LOOP coil of an modified. New buses.
Italf scram occurred.
event. DD: Bus transfer MOV in computers installed.
RFP 2B tripped and 2C l' ailed because of isolation New replacement failed to start. Reactor defective control condenser radios ordered.
scrammed on low water circuit. Process system.
Emergency power y
level. Process computer computer had poor supply added to alarm printer stopped.
coramunication with telephones.
Lost audio contact all the printer. The radio
~
over the plant and lost and telephone several telephones communication failed Operator could not because the repeater control transfer of transmitters did not condensate water to the have alternate power isolation condenser shell.
supply l
e 9
h
Q\\
/.
d d
d E
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n rf e eeW Vf i n c ef mi t
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s el b
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mcd eas ul f iai s a h e T6 1
l ur o n
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s a
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ai nnvu oe Amt sbt WpoG4 loc Dcpt r n r r r
r E
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l
Table 1. (cont.)
r Failed No.
Plant Data Esent Description Cause Equipment Correctise Actions I:sent hignificance 11.
Palo Verde 2 Two transformers, MT: 'i he tra sformer Tr ansformes Failed bushings
~l he 1:lKis started and 529/89-001 feeding power from a failed because of bushings replaced. Ilushing powered the Class IE 24h.$1m.
13.8 kV bus to two 4.10 particle and moisture creepage distance buses.
kV Class IE buses, accumulation on the increased. Implemented failed causing a LOOP bushings. DD:
bushing washing inadequate creepage program.
distance of the bushings.
12.
Point Ileach 2 Du' ring trouble shooting PE-1: The deluge Transformer Failed bushings 1 he generator stuck 301/89-002 the main transformer system had wiring bushings replaced.
breaker relay acted 3h. 22m.
deluge system, the error. The cause for prematurely and deluge system actuated.
transformer trip could caused the grid The transformer tripped not be established.
degradation. 'l he resulting in a reactor EDGs started and trip. Ilus transfer took powered the Class Ili i
place but the grid buses.
I~o voltage degraded.13us I
undervoltage relay operated and EDGs energized the Class IE loads.
13.
Itobinson 2 While EDG Il was out of EF: Investigation could Nonc Installed larger size Control room 261/88-005 service, Class I E bus E-not establish the cause fuse on the PT primary ventilation failed to 4h. 44m.
2 was lost because of a of the fuse failure. DD.
and copper inseits on align to emergency blown fuse in the SUT tripped because of the secondary side.
recirculation mode, undervoltage relay de saturation of cts.
Fire water supply to circuit. The reactor containment vessel was tripped and the auxiliary isolated. The plant was loads were transferred to cooled by natural SUT. A 115 kV breaker circulation and tripped and de-POlt Vs.
energized SUT. EDG A started and powered Class I E bus E-1. Si and MSIV closure signals were initiated.
e 4
I able I. hont.)
+
l' ailed No.
Plant Data I:sent Description Cause I:quipment Correctisc Actions I:sent Significance 14 Salem 2 While trouble shooting Pl.-1: During None A number of short-term lilXi 214 was out of 311/86-007 steam generator level, troubleshooting steam and long-term service resulting in loss 3h. 47m.
the reactor tripped and generator water lesel, a insestigatioas and design of all ac to sital bus 51 occurred. Multiple technician shorted an modifications initiated.
211. Number of MOVs bus transfers occurred instrument bus DD.
and safeguard due to degraded bus Degraded bus soltage equipment were lost.
voltage. Component was caused by g
cooling water pump and unplanned load growth.
W RCPs were secured.
Additional findings Reactor coolant pressure under investigation.
increased requiring cycling of PORV 2PRI.
15.
Summer Technicians shorted two PE-1: Technicians None Multiple modifications lloth I:DGs started and 395/89-008 power leads inside a inadvertently shorted in relay settings powered Class II:
2 h. 10m.
generator cabinet two power leads inside initiated.
liuses.
i causing a turbine trip the generator stator 7,
and a reactor trip. Three cooling water cabinet.
other offsite generating 1:F Turbine runback units tripped causing a relay failed. DD. Relay degraded grid.
settings in other plants were inadequate
Table 2. l'artial I ow-of-Offsite Power Cansed h3 Plant-Centered 1:sents IJnder l'ower Operation Failed No.
Plant Data Eient Description Cause Equipment Corrective Actions 1:sent Signifitance 1.
Arkansas i During maintenance PE-l: lil:M not pulled None Switchyard maintenance LDGs started but did 313/87-005 tests, generator output out. DD: Syncheck practiec and ilFM not tie to Class IE 20 cycles breaker failed to trip relay set too low. Ill-M circuitry under review.
buses. Lost main resulting in a turbine not set to trip the feedwater pumps. The trip. One bus fast generator lockout relay.
plant was stabilized in transferred, other buses hot shutdown.
slow transferred.
2.
Ileaver Valley 2 During 100 percent load EF: Turbine overspeed None None. 'I he load rejection No. I I DG autostarted 412/87-032 rejection test, bus protection control test was a one-time test and energized class IE 40m.
transfers were inhibited malfunctioned.
and will not be repeated.
bus.
because of phase Elevated pressure difference between experienced in high onsite and offsite power pressure turbine. Steam sources.
dump did not operate.
[
3.
thunswick i During manual control EF: Oxide buildup on Generator Defective parts were Group 2, 3,6. and 8 y
325/86-024 of main generator manual potentiometer voltage replaced.
containment isolation 3hr. 31m.
voltage regulator, the contact wiper regulator occu r red. SR Vs reactor tripped. Two prevented control of opened. Two EDGs Class IE buses tripped generator voltage.
powered two Class I E because of degraded bus buses.
voltage.
4.
Callaway l During a reactor startup, EF: Water accumulated None pull station plug itcactor manually g
483/85-0ll the station SUT was lost inside a deluge pullbox.
resealed. The deluge tripped and one EDG 2h. 26m.
because of deluge MT: A pipe plug on control circuit modified autornatically started actuation.
pullbox was loose and to inhibit deluge and powered Class IE not properly caulked.
operation until the bus.
DD: Inadequate deluge transformer trips.
control circuit.
5.
Callaway i The main generator PC-l: The operator None Additional operator One EDG and AFW 483/88-015 breaker tripped when an unknowingly vibrated training on racking and pump actuated.
30m.
operator racked in an the generator breaker relay sensitivity adjoining breaker, cubicle. MT: Relay initiated. Relay contact contact set too gap readjusted.
sensitive.
9 4
m m.
od o
r 1
t o
d d
wt n 1 a
e le lor G a s s
c iaf aD d
la p
n teC eE I
a ms ic
. a e
a e i
d e
if e e n
ts h r
n pt uO t
s g
ip t
Gd r te ioe.
Dz s
iS t
r e
a d
omt e I ig l
t s
a u
P s et er.
Cieh r
e s c
a n
o n et n u s
Oeb Rt or s I
e r
u e
.d ly d
r s
gcd g d et e o
ina e n e c nt n
lpi n
r p
u vi s o e ra s
or v itc is e ei e
r c a pe q
A m
e t
mc e r
s t
c d
t r
h a n
n s l
i aT s
n e r
a il a n o n n. d.
g n
e r ed d ot o i
t wa c
ai l
n s ei t
e n
e r
le p r i e
r i med biii m
p a v v o
a o d
Ia. M re e e C
i coa r s e
r.
e g t
r e
t t
n i.
n e
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a it Olt
. or n.
w mVs e oh d p ki s
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ow a
r Mir ps N
hb l
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)
tno loe d
d
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s l
g eih a
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d i o a pt t
r n
e 2.
p odk d
a gm r
e n uk t
ir c e a n
V f
t n
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a ble t f eile r
e r
r i s a pe a oS s nu oi l
sf loSiVia a n u
s e gc oer q a
t pt n
tcf M dl n tef tsi phf ah ro r i e i
r r e t
r pl aTr a e eht n aDn n
v o n
- d t
c l
a Ta aid a
u h r f weI y. -
r o
i d f u
pi i EbDr e o cMt os ts dDi et a
f od Ogh. o Dvi n
n g
s e i la s
. gwodt en u
.d mpa l:
n m ah. n og I
gd i
r ojr e. o s
e l
n t
e ai a uid e r
s t
pk a l
ycl i Waait e cl u
- o c
l c s s p k
d ai iol oe n l
luS l
wb oe a eoi l
or h i
a l
r C
l' wr r f b pmt b cf t il P pAtoe a
r r sl e
ip ly t
h p
s n
it t
r e
e u
t u
t r
u a
n to
.d t
s ut e o bA c
r r
nt k
r ur o a Od e r n a
ebi t
e n v cd s e
o c n e a e
.l adVgk esi a
r itp of b
e sP ak n c if h e a r
, n
,n i o l
t ir a
aW.
Ve x
c i4 s
c ur d
L n
l d p p S
n n e e npm s
gl D
ih toai 4 a ah s is e
t f
e n
l u
a a
c t
h ni s u wg
,f geid eTl p
g a r oidV nt r
. meWt.
t r
clleat Sl u int n
l E
- on ileSV i
e l
h e
w s r ra louooS a e u v
i ofaMT Dml e pvf npAt s
I 2
i no e
y a
s 1 n7 a0 s1 t
e0 0
C-a l
D l 7 o8 t
is8 l
8.
v6.
b/m n
/
a3 a
a l
D32 i
4 h 2 4 P
D31 o.
N 6
7.
2 i
l tl 1llll
'l able 2. (cont.)
l' ailed No.
Plant Data Escut liescription Cause l' quip men t Correctise Actions I: vent Significance 8.
Dresden 3 A fault occurred inside a EF: A failed capacitor None Defective components 1.ow lesel 249/89-001 345 kV breaker. A 4 kV inside a 345 kV replaced. Design and contamination in the 7h. 22m.
bus did not fast transfer.
breaker. Pitted relay procedures modified.
isolation condenser RFPs dropped out. T he contacts in bus transfer Contaminated areas sent area.
reactor scrammed.
circuits. Fuse failure, decontaminated.
Several components DD: Inadequate time Mamtenance practices malfunctioned. Low delay in bus transfer.
r e s ie wed.
g contamination in area Feedwater control around the isolation unable to control rapid condenser vent.
water level increase.
The clean water supply to the isolation condenser too low. UPS failed due to inappropriate switching sequence. PE-P; Main turbine turning gear did not engage because of procedure deficiency. Operator failed to control llPCI lube oil cooling because of lack of procedures MT: Improper g-installation of IIPCI turbine gear motor brush and pitting of relay contacts.
9.
Duane Arnold During a power / load EF: Reactor scrammed Hreaker trip Defective components One ElXi started and 331/89-011 unbalance test, the because of a failure of coil and replaced. Power / load powered the Class IE 40m.
reactor scrammed. 't he 11 generator cts. Ilus cts unbalance circuit bus.
Class IE buses did not transfer blocked operating procedures transfer.
because the breaker
- revised, trip coil was burrit out.
6 G
~
w c
y.
dE E
eI d s t
I r s d
au E
s e
eb a s ns t
a c
dI r
a a sl n
l e
ns l
ic st a
toc dC a
ah a s e
ud e
f o
dl a e t
n a
i wt eC r
z a
t s
o n
i ts r e Gg d
d i
g ah Xt ts e
De e
r l
iS t
f l
- n.
2 iz
- c sd.
I e
t I
n Ges o
e Gr.
es g
ne e
d e i
h Da wd Des n
s s
u nu nu o
o E
I c
E l b Tab E eb f
l d
g e5 h
u ni it o a
n cl o
d nS.
w r
i d.
a n a -
n g
s v
e d ia n2 d n o
n i
s e r teTi ei d
o l
t o
l e
nnUa lvd t
n i
r l
i i d.
c s r id noi a o u e tc ht ol r e n m.f iod vbl d e u
t t
A a
n e
el a
e s
ie s
v e s
l s
p cu e
n c
m mv en lei u n
i on r
r o
n t
v u oieitnd nr c u
t s v c
nb o o
c no o
a o
r e
ic n
ee o u r
e k
r c
r r
es c
eepveod r
c s
t r e o o
h a a
pvi mr r n e h C
Tw R
Ogi ppu Pt w tn g
e n
ir g m
lei e
e e
tnon d p t i i u n
n n
h she
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gr s a
u N
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1
)
tno d
e6 l
g k
e hT S
d.
b t S T
le n
e g h n e
r g
esS c
ni pr m
r topk h.
togn3h a a
d t i e
e wl n
2 a i a c
a nit w2f h i nt t
ou oh le b
e r
n e aUs nTot lb b g
r r
or e
r t
e re e
u oUen di r
a pyb om pwl. ai u
ini r si s eel T
ot or o
ct l
t t
n c o e
u c i T:
t l
a u i s e r
es e
l cf eh n o bb pMc o
n om.
ei gb i
h as h
e er t
pr t
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et eh g nh, Du eoo n
Td ad
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Tr s n
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e
- vb Ar
- p aGpc n c
1: mf k
di df e
s l
t I
s ts t
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u d
pi ol ne peDiwr iams' l' f r t
a r
isiw Eoo a r a
Ea e e
Ei l
nh h
r or r wf C
Pi t
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l' bet s r
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t od luf r h c la t
ye I n
p u
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a oat inip e
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t m.
a r
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f a
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la u
l f
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r k
gl o el arCei pd a
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, t k o noi i dUr r
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dd x e ed.
6 n
f o d oe e
a t
9.T ri t it o a e r
nr p d e er n at s nr ni up s
ec er d 6 S u ns le u
a t
e o
o s
r ahi ciep sih c ar c ae pot r b le it r - d o oi abt i e a i e3 t
e r ar u vTs g
oy ir s r u gr t h r
y n peeVa
- c. p f t
mt a o ip r r op n pd T e uwt a
or h fhTnfh ga Ls r
r e
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.h d r c
r l
i r
i t k e
e e r e s m t
ea e e
,toUp ock o imt t
s t
r c
en
.e s
t t r e o n
a h
r o s
s ig 9. T ymed t
eecidoet mmu leh S
r n
n o gi a.
t e
h el eff c o
n t
nwe n eps gepc t
t D
6 Aredi sb io ae2 f gt n z
a oe r r
s oI mms n
i s ne ct oa c
,d r
s wa S s
cys m le t
sf e
.k eh n
iGcobugpe t
n moe r
upo l
,2 e
pcmhi hht s t t
mve a s ns i o a
nia cc r
r r
pl pT o ept ot s pt in a nr r r
s e
a o
e uDs0nu i h n a a h gi oi o io n
v i
r r ir E
DEd1 e pdT Aot w Wt rUlcb mr wp AoUtrUsci r
a r ap t I t
t 2
9 t
0 9
n4 0
a t
1 1
i0 0.
o0 m
a 0
0.
P i 9 I
m m
le8 8 2 -
D y
i 8 7
7 4
n 8 4 l
/ 2 t
8 m8 le/m ia/1 a3.
r 8 t 1/ 1 n
l a4 5 c4h.
d7.
S7h a
33 In4h a32 2l L35 P
I 31 I
o 0
2 3
I N
1 I
1 1
j t
i
'l able 2. (cont.)
l' ailed No.
Plant Data Etent Description
( 'a n se j 1:quipment
('orreclisc Actions I:sent Significance 14 I aSalle 2 The SAT tripped Ei: Inadvertent trip of SAT Cause of the deluge
~l he is and C drywell 374/89-007 following sudden SAT from actuation of bushing actuation couhl not be chillers, the 211 Sih. Sm.
actuation of the deluge the deluge system.
established.
circulating water system. On loss of SAT pump, reactor building bus 243 is normally fed ventilation, reactor from EDG-21). The water cleanup system, EDG 21) was undergoing and the IC feedwater maintenance. Unit 2 heater string tripped.
g remained on line and the EDG 211 was started Ih.
6m. after the SAT tripped.
15.
Nort' A'ina i A failure of the RSST I:ll: Mechanical failure MI:RV and Defective parts of the The source range 338/8ti-20 caused degraded voltage of the MFRV.
RSST Mi RV and RSST nuclear 8s.
at Class IE bus lj. The tapchanger tapchanger replaced, instrumentation was i
Mi~RV closed causing a Initiated evaluation of manually energized.
'd reactor trip.
operator performance, One main feedwater I
cause of RSST isolation valve did not tapchanger failure, fully close. One EDG rnanual supervision of started and energized tapchanger operation, buslj.
and replaced MFRV trim during next refueling outage.
16.
Palo Verde I An electrical fault EF: Faulty CT. MT:
CI Defective CT replaced.
EDGs started and 528/86-003 occurred on the high Water leaked inside the A 10 percent sample of energized the Class IE 20m.
voltage side of a 13.8V -
lomi center.
similar cts tested.
buses.
480V load center transformer. Due to a faulty CT, the SUT tripped resulting in scram of reactors I and 2.
4 e
9
L l'
w e
E E
e nf E
n s
d dI I
i o dI r
u Ge n
cg n
s s o8 ai z
a sf n n a
s e
Di.
dl dl cn i o.
dl s
sa m
c g
a a
e ot it C t
n i
s t
s b
r r luaI a
e u eC nCs lo e
teC t
t ic d nb r
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17 ailed No.
I'lant Data Esent Description Cause I:quipinen t Cur ret-lise Actions inent Si:nificance t
21.
Surry i Material blown from the p t-1: I he material on itelay iteplaced the defectise l he 1:lXi started and 280/89-044 turbine roof caused a the roof was not r ela y, energized the Class IE 16 m.
fault on the RSST. 'l he secured properly. LI:
lius.
operator manually A relay failure tripped the reactor and prevented the tripping the turbine. The of the main generator generator was manually breaker.
tripped after 200 g
seconds.
22.
Vogtle While transferring power DD: A CT was None Coraceted the improper t he LDG started and 425/89-023 supply to a non-lE bus, improperly terminated.
Cl termination and energized the Class IE 4h.17m.
the incoming breater reviewed other CT loads.
tripped resulting in terrninations.
power loss to one Class IE and various non-lE t
buses.
's 23.
Waterford 3 While transferring power DD: The feedwater None T he cup valve seat The I.DG started and i
382/85-040 supply from startup to pump governor cup seplaced. T he breaker energized the Class IE 2m.
auxil:ary transformer, valve seat was control circuit under loads.
the incoming breaker undersized.
review.
first closed and then opened. The main feedwater pump tripped from overspeed.
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i l~ahle 3. (cont.)
Eailed No.
Plant I)ata Esent 1)escription Ca u se Equipment Corecctisc Actions I:sent Significance 6.
Pdgrim A cable fault on El : Cable fault. Cable Cable I: ailed caw replaced.
~l he 1.1)Gs started and 293/89-010 secondary side of SUT damaged during energized the Class IE 15h. 20n..
tripped the transformer.
installation.
buses.
7.
Prairie Island 1 I)uring restoration of El: A protective telay Nonc
~1 he defectise relay
'I he EI)Gs started and 282/85-006 reserve transformer,4.16 malfunctioned.
seplaced.
energized the Class 1E 2h.21m.
kV buses were de-
- buses, energized.
8.
River llend All four preferred PE-l. Iland held radios None Seseral design changes LI)Gs started and 458/86-002 station transformers caused spurious sigt:als.
implemented *, pievent energized the Class IE lb. 49m.
tripped resulting in a spurious operation from buses.
total LOOP.
radio caise.
9.
Shoreham A condensate pump start PE-P. A test procedure None
'i he responsible I:I)Gs started and 322/87-003 tripped the NSST. I ast required installing a individuals counselled.
energized the Class IE 2h. 6m.
bus transfer tripped the jumper across CT buses. Initiated scactor RSST. The reactor terminals. 'I he engineer buildmg standby 7
tripped along with preparing the sentilation system, NSSSS isolation and procedure did not control room air initiation and isolation consider the effects of conditioning, reactor of various safety the changes on the water cleanup systems.
protective system.
isolation, scactor building closed loop cooling water split, and scactor building service water split.
Table 3. (cont.)
failed No.
Plant Data Etent Description Cause 1:quipmen t Currectitc Actions Esent Significance 10.
Shoreham
't he NSST tripped PE-1: I he technician None The event repor t was I:DGs started but did 322/87-026 during a relay failed to pull out a read by I&C not load as slow bus 20 cycles calibration. The fast bus relay prior to the test.
management and relay transfer was transfer was inhibited.
supervisors and successful.
The slow transfer took technicians.
place. Numerous ESFs initiated. The reactor building standby g
ventilation and the control room air conditioning system chillers became inoperable.
II.
Wolf Creek i During execution of PI:-P: 'I he individuals Switchgear Strong recommendations I:DG started but the 482/87-048 maintenance work, a involved failed to X NI102 for prejob briefing 1:DG breaker did no' i
15m.
technician was follow procedure to issued.
close.
8 electrocuted resulting ir.
check the power supply I
a fire in Class IE bus before working on N1102. In trying to clear 4160V switchgear.
the fault, the operator tripped the remaining Class I E bus.
O 4
9 e
e
I able.I. Partial I.oss-of-Of fsite Power Caused h3 Plant-Centered 13ents Linder Plant Shutdow n Operation I
l' ailed No.
Plant Data Event Description Cause Equipment Correctisc Actions L>cnt Significance 1.
Arkansas I LOOP occurred twice PE-l: 1: xact cause could None Senior management and Reactor coolant system 313/89-043 durmg manual bus not be established.
supervisor y personnel temperature increased 15m.
translers. During the Likely cause operator addressed operating from 103 Y to 120 T.
second event, the decay error.
crews on heat removal pump was professionalism, lost for 9 minutes.
formality, attention to details, and safety.
h 2.
Ilrowns Ferry I While inspecting floor PE-l: Unauthorized None Responsible personnel Initiated half scram 259/86-029 penetrations, person nel opening of the PI' received disciplinary signal on RPS, 27m.
erroneously opened PT cabinet door, action.
containment isolation cabinet which tripped groups 2, 3,6, arid 8.
the normal feeder Unit 1,2, and 3 refuel breaker to the bus.
zone isolation, control room emergency ventilation and SilGT.
I EDG started and M
powered Class I E bus.
i 3.
Browns Ferry 2 Ar electrical fault on a DD. Inadequate 4160V bus.
Preventive maintenance 1.ost three RPS MGs 260/89-008 unit USST led to the loss insulation above the bus duct.
practice reviewed, and RPS buses lil,2 A, 16m.
of a shutdown bus. The bus duct. T he design of and design changes initiated, and 211. Unit I half alternate feed to this bus the bus duct allowed transfor mer and corrective action scrammed and Unit 2 was tagged out for collection of water bushing plans initiated for this full scrammed.
maintenance.
from condensation.
connections.
transformer and other Containment isolation MT: Vendor t ransfor mers.
groups 2, 3,6, and 8.
recommended SilGT trains A, II, and preventive maintenance C, CREV trains A and was not performed.
II, and Units I,2, and PE-1: Multiple 3 refuel zone isolation abnormal electrical actuated.
system alignments and inadeouate communications led to inappropriate actions.
J
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Talite 4. (cont.)
Failed No.
Plant Data Event Description Cause Equipment Correctisc Actions Escut Significance 15.
North Anna I,2 During a switchyard PE-l: An electrician None Procedures modified EI)Gs started and 338/89-010 modifica tion, offsite inadvertently grounded powered the Class IE 16m.
power to Class IE buses a control wire causing buses ill and 2J.
til and 2J was lost due the I.OOP. PE-P; 1.ack to a ground fault. RilR of adequate procedure.
pump was lost for about 5 minutes.
16.
Palo Verde i SUT tripped due to a MT: Dust accumulated flus duct Failed components EDGs started and 528/88-003 fault on a 13.8 kV bus inside the failed bus and CT replaced. cts being powered Class IE 2h.41m.
and a CT. ESF actuation duct, and water seeped replaced by CIs with buses. Ilatf leg trips and loss of radiation in. DD. CT insulation better insulation.
received on monitor occurred inadequate.
containment isolation, resulting in actuation of SI, containment spray, control room essential and recirculation filtration, containment actuation signals.
I purge isolation, and fuel U
building essential ventilation signals.
17.
Palo Verde I While replacing a relay, PE-1: The electrician None None EDG started and 528/88-019 power to a Class IE bus caused a ground fault powered the Class IE 37m.
was lost. This actuated while working in a bus.
train 11 FilEVS and restricted area.
CREFS signals, which tripped trains A and B Fi1EVS and CREFS.
18.
Peach llottoia 3 During maintenance of PE-I; Probable cause -
None Cause of the evem EDGs started and 3278-88/009 Class IE bus 23, the electrician error, investigated but could powered the Class IE 30m.
feeder breaker to the bus not be ascertained.
bus.
tripped unexpectedly.
Four outboard Group 11 and 111 prima y containment isolation system isolation valves closed.
e b
o
=
Iable 4. (cont.)
l' ailed No.
plant Data Esent Description Cau se Equipment Correclise Actions Eient Significance 19.
Quad Cities 2 A fault on the reserve pl.-1: 'I he personnel I.ine Replaced damaged I:lXi started and 20$/85-011 U AT developed as lowering the cord were insulators insulators.
powered the Class IE 43m.
personnel lowered a unaware of the danger.
damaged.
bus.
power cord close to the transformer. The fault created a transient on the electrical system resulting in a loss of the RPS bus and a lock-up of a feedwater regulating valve.
20.
Iliver llend While installing a spare PE-l: 1.ack of None All relay department EDG started and 458/89-029 transfor mer, technicians communication personnel retrained.
powered the Class IE 24m.
shorted two live leads between the supervisor panels labeled.
bus.
resulting in loss of a and the technicians.
I preferred transformer.
d ESF actuated resulting in i
start of Division I and 11 SilGT, annulus mixing and fuel building ventilation filters.
Various containment isolation valves closed.
21.
Sequoyah During a thunderstorm, h1T: hiixture of Ilus duct and Repaired damaged l'our EDGs started and 328/88-034 the 6.9 kV start bus moistute and dust breaker components.
powered the Class I E 41m.
flashed over resulting in caused short circuit compartment.
buses.
loss of sever 116.9 kV inside the 6.9 kV start Class IE buses. Lost bus.
cooling tower lights and pumps, and plant security computers.
Table 4. (cont.)
Failed No.
Plant Data Esent Description Canse Eq uipruen t Correctise Actions Esent Si:nificance t
22.
South Texas i Following a fire, PE-1: Operator slipped.
None F ne One I.IKi started and 498/89-006 operators were bleeding powered Class I E bus.
I h. 7m.
air from the generator circuit breaker. One operator slipped and tripped I: e main transformer, Train A Class IE buses, and all non-Class IE buses.
23.
South Texas 2 Following replacement PE-l: An operator None Developed procedure for Two EDGs started and 499/89-001 of a ST heat detectors, restored the deluge fire protection panel powered the Class IE 25m.
the deluge system was system without restoration. Operators buses.
being restored. An resetting ST detector briefed. Deluge nozzles actuation of the deluge alarm. PE-P: Lack of repositioned.
system occurred procedure for resetting 8
resulting in a loss of ST.
alarm. DD: Inadequate settmg of deluge e
i nozzles.
24.
South Texas 2 During a thunderstorm, EF: Failure of lightning Lightning All lightning arresters Two EDGs started and 499/89-005 a ST lightning arrester arrester.
arrester for ST replaced.
powered the Class I E lb. 22m.
failed resulting in a loss buses.
25.
South Texas 2 Four seconds after PE-l: Improper wiring None Generator relay wiring EDG 21 started and 499/89-029 generator of generator relays. EF:
corrected. Ilroken lug powered Class IE bus 6m.
synchronization to the RCP tripped because of repaired.
E2A.
grid, the reactor tripped a broken lug in control causing a LOOP to circuit.
auxiliary buses and Train A ESF bus. RCP ID tripped after energization.
o 4
b e
O A
Table.8. (cont.)
Failed No.
Plant Data Esent Description Cause Equipment Correctiie Actions Eient Significance 26.
South Texas 2 While checking the main EF: A component One meter
~l he meter was needed EDG 21 started and 499/89-014 generator protective failure inside generator for testing only it was powered Class IE bus S i m.
circuit with the phase angle meter removed. The EDG fuel E2A.
genera:or main and field backfed voltage to linkage corrected.
breaker open, the high generator protective voltage breakers feeding relay. EDG failed due the main transformer to misadjustment of g
tripped resulting in an fuel rack actuation W
offsite power loss to ESF linkage.
bus E2A. One nonsafety-related EDG failed to start.
27.
Surry I During a manual bus MT: The incoming flatter y Necessary maintenance, two EDGs started and 280/89-005 transfer, the incoming breaker failed to close repair, and adjustments powered the Class IE 31 m.
breaker failed to close because of dust and per formed.
buses.
i resulting in LOOP at two dirt. EF: llackup D
Class IE and two non-service air compressor Class IE buses.
failed because of low Subsequently, more oil pressure. Loss of buses tripped on memory and SPs to undervoltage. The monitors was due to a running RilR and failure of backup component cooling water battery supply.
pumps tripped. llackup service air compressor failed to start. Memory and SPs for process and ventilation vent monitors were lost. Sample pumps for containment particulate and gas radiation monitors tripped.
Table 4. (cont.)
Failed
[ No.
Plant Data Esent Description Cause I:stuipment Correctise Actions Esent 5ignificance 28.
Th11 1 During a manual bus I:F: The breaker fault 4160V Coeuit breaker replaced.
Offsite power was 289/86-008 transfer, the incoming was due to a circuit quietly restored on all 3h. 25m.
circuit breaker component failure.
breaker.
buses except for the developed a ground fault faulted bus. The EDG resulting in a loss of the started but did not incoming transformer, load because of the bus RCPs, and secondary fault.
g plant components.
29.
TMII During Appendix R PE-P; Inadequate None None
'T he EDG started but 289/87-001 modification work, a instruction to the did not load because of 8m.
technician opened a CT operator.
the operation of the circuit in IE bus which mercurrent relay.
energized a neutral overcurrent relay and tripped the I E bus.
I g
30.
TMII During maintenance, an MT; The relay wires None None
'I he !!DG star ted and i
289/87-002 electrician replaced an were not properly powered the Class t E 13m.
improperly placed relay terminated bus.
cover which tripped the l A auxiliary transformer.
31.
Turkey Point 3 While closing a relay PE-1: Vibration from None Technician cautioned.
I:DG 3 A sic
.d and g
250/85-012 panel door, the main closing the door tripped powered the Class IE Sh 35m transformer, the a lockout relay.
bus.
auxiliary transf ormer, and bus 3 A tripped.
32.
Turkey Point 4 During construction PE-I: Inadvertent None Construction crew 1 DG Il started but did 251/86-007 activity, the lockout jarring of the relay.
cautioned.
. m had because of 59m.
relay for 4160 V bus 411
.jpt i at an of the was actuated resulting in
.ed oe t relay.
loss of the transformer feeding this bus.
I d
l e
s e
o 9
1 able.s. (t ont.)
17 ailed No.
Plant Data Esent Description Cause Equipment Correcline Actions 1:sent Significance 33.
Yankee Rowe While modifying the PE-l: Excessise None None ElXi started and 29/87-005 control room main vibration produced powered the Class IE 3h 50m.
panel, the SST tripped.
while cutting a hole on bus.
control panel.
34.
Yankee Rowe During testing of the PE-P: Deficient test None This tett is a one time EDG started and 29/88-010 generator exciter field, procedure caused this event. Procedure powered the Class IE 2h. 22m.
two Class IE buses were event. EF: The tie checked with the s.
lost. Subsequently, breaker failed because manufacturer.
during return of power of failure of breaker from normal source, the trip linkage.
bus tie breaker did not close.
t
's i
Table 5. Duration of Loss-of-Offsite Power Plant Power Total or Number of No. of Events Total LOOP Aserage I.OOP No. of Events Table Condition Partial LOOP E,ents
< 30 m.
Duration Duration
> 6h.
1.
Power Total 15 2
70h 36m.
4h. 42m.
3 2.
Power Par tial 26 9
164h. 58m.
6h.21m.
5 3.
Shutdown Total iI 4
25h. 28m.
2h.19m.
I 4.
Shutdown Par tial 34 14 35h. 49m.
Ih. 3m.
O TOTAL 86 29 296h.Sim.
3h. 27m.
9
Table 6. Analysis of I EH I:ient Causes Plant Total or I:quipment inadequate Power Partial Malfunctions or 51ain tenance Table No.
Condition LOOP Personal Errors Failures Design Deficiencies Practices No.
Percent No.
Percent No.
Percent No.
Percent h
1.
Power Total 13 48 6
22 6
22 2
8 2.
Power Partial 11 33 13 40 5
15 4
12 3.
Shutdown Total 7
59 4
33 I
8 0
0 4
Shutdown Par tial 22 57 8
20 4
10 5
13 r-l TOTAL 53 48 31 28 16 14 11 100 i
8 Table 7. Plant Centered LOOP Eient Significance Number of Number of Maximuru ASP Table No.
LOOP Type Condition Eients ASP Eialuations CCDP 1.
Total LOOP at Power 15 8
4.30-4 2.
Putial LOOP at Power 26 1
1.3 E-5 3.
Total LOOP Shutdown 11 3
7E-5 4
Partial 1.OOP Shutdown 34 0
4 2
J