-lu
A Study of US Nuclear Power Boiling Water Reactor,
Class IV, Operating Performance, 1992 -1997
by
LCDR David Lester Brodeur, United States Navy
B.S. Optical Engineering, University of Rochester, 1985
Submitted to the Department of Nuclear Engineering in partial
fulfillment of the requirements for the degrees of
Nuclear Engineer
and
Masters of Science in Nuclear Engineering
at the
MASSACHUSETTS INSTITUTE OF TECHNOLOGY
June 1998
@ United States Government, All rights reserved.
The author hereby grants to MIT permission to reproduce and to distribute
publicly paper and electronic copies of this thesis document in whole or part.
Signature of Author:
Department of Npclear Engineering
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Certified by::
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Neil E. Todreas
Professor of Nuclear Engineering
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Thesis Supervisor
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Thesis Reader
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Accepted by:
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Lawrence M. Lidsky
Chairman, Committee on Graduate Students
Department of Nuclear Engineering
A Study of US Nuclear Power Boiling Water Reactor, Class IV, Operating
Performance, 1992 -1997
by LCDR David Lester Brodeur, United States Navy
Submitted to the Department of Nuclear Engineering on May 1, 1998 in partial
fulfillment of the requirements for the degrees of Nuclear Engineer and
Masters of Science in Nuclear Engineering
ABSTRACT
The steady improvement of US Nuclear Utility generation capability observed over the past two
decades has recently halted and somewhat degraded. For the industry to resume its upward trend in
performance a detailed examination must be performed of current performance and new methods
developed to continue the improvement. A detailed study of Boiling Water Reactor, Class IV (BWR/4)
performance over the past five years was conducted to gain insight to the nature of lost generation
capability and develop a methodology to improve capability. Extensive electronic NRC records were
used in conjunction with detailed power plant records and engineering experience at PECO Energy's
Limerick Generating Station and Peach Bottom Atomic Power Station for this research.
Administrative or regulatory shutdowns within the study dominated the lost generation capability
and detracted from the goal of analyzing equipment reliability. Nine of two hundred thirty five
shutdowns were therefore limited to maximum impact of 30 days lost generation. Balance of Plant
system failures were found to initiate 69% of the occurrences of lost generation capability and account
for 59% of the capability loss. The failures of these systems were found to be infrequent events which
correlated poorly to the aggregate industry experience. Approximately fifty percent of the forced outages
were the result of equipment related failures such as weak design or worn parts with the remaining fifty
percent the result of human related failures. Only 19% of the failures were noted to be the result of
component age related failures while 31% of the failures were related to poor equipment design.
The time frame of forced outages with in operating cycles was additionally reviewed. Failures
were found to be more frequent in the early phase of the operating cycle following start up from a
refueling and approximately 400 to 550 days after start up. The impact of these failures was not great
enough to affect the steady state cumulative capability factor of the aggregate BWR/4 utility achieved
after one year of operation. Individual utility sites were found to have opposing strong and weak periods
of performance within their operating cycles. The loss of generation capacity taken for planned
maintenance outages and on line maintenance for minor equipment problems was not found to have a
significant impact on aggregate BWR/4 performance. For plants not involve in lengthy shutdowns, the
strongest impacts on cumulative capacity were forced outages, initial start up and coast down.
The unpredictable and design nature of system failures necessitates a structured effort to
improve the combined performance of all systems at a utility. Balance of Plant systems were found to
all have a 25% probability of causing a single forced outage lasting slightly less than 5 days in length. The
infrequent nature of significant failures necessitates a broad based communication between utilities to
maintain an adequate level of awareness of system vulnerabilities and possible improvements. Two
specific sites examined had opposing and repeatable strong and weak cycle performance traits. The
unique nature of site performance demonstrates the impact that improved communications between
utilities could have on transferring strengths and diminishing weaknesses thus improving overall utility
performance.
Thesis Supervisor: Neil Todreas
Title: Professor of Nuclear Engineering
TABLE OF CONTENTS
Page
8
1. In tro duction .....................................................................................................................................................
8
1.1 Im p etu s....................................................................................................................................................
1.2 Confinement of Study to US BWR/4 Plants over a Five Year Period ............................. 11
13
1.3 R esearch G oal....................................................................................................................................
1.4 Design of Research and Presentation of Report..................................................................14
........................................................................................... 15
1.5 Organization of paper ........
. ................. ....................................................... 16
..
2. Sources of data ................................ .......
2.1 PECO Energy Plant Specific Data.........................................................................................16
19
2.2 NRC Monthly reports, INEEL MORP 2 ................................................................................
2.3 NRC Daily Reports, INEEL MORP 3.........................................................................................20
2.4 INEEL NRC Report Formats ........................................................................................................ 21
22
3. Initiating System Failure Analysis of Plant Unavailability ..... ....................
22
.........................................................
3.1 BWR/4 System Reliability Impact on Plant Capability
3.2 PECO Energy Plant Specific Data.................................................................................................26
3.3 Consistency of System Failures BWR/4 Fleet versus PECO Plants ..................................... 28
.............................................. 37
4. Causal Analysis of System Failure..........................................
4.1 Equipment centered failure analysis.......................................................................................37
.......................................................... 41
4.2 Failure Root Cause Analysis................................
44
5. Operating Cycle Analysis .................................................................................................
46
5.1 Analysis design ......................................................................................................................
46
5.2 Forced Outage Operating Cycle Analysis.....................................................................
5.2.1 Forward looking forced outage operating cycle analysis .............................................. 49
5.2.2 Comparison of PECO sites to BWR/4 operating cycle performance ................... 50
5.2.3 Backward looking forced outage operating cycle analysis............................................. 52
5.2.4 Cumulative impact of forced outages...................................................................................... 53
5.3 Daily Generation Operating Cycle Analysis.................................
54
5.4 Conditional relation of forced outage events upon operating cycle events .......................... 57
5.5 Operating cycle analysis summary .................................................................................................. 58
6. Plant reliability improvement processes.......................................................................................... 59
............................... 62
. ........................................
7. Conclusions.....................................................
Appendices
1. LG S Lost G eneration Data ................................................................................................................... 65
2. PBAPS Lost G eneration D ata..............................................................................................................71
3. BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996....................76
part 2 events within complete operating cycles 1989 - 1997 .............. 100
114
4. NRC monthly report glossary of terms ...................................................................................
5.
6.
7.
8.
9.
10.
11.
12.
13.
BWR/4 Operating Cycle Skylines .......................................
INEEL NRC report formats .........................................................................................................
BWR/4 Forced Outage Data, sorted by failed system ....................................
LGS Unavailability Data, sorted by failed system ....................................
PBAPS Unavailability Data, sorted by failed system ....................................
LGS Unavailability Data sorted by component failure cause............................................
......................
PBAPS Unavailability Data sorted by component failure cause ..................
....................................
failure
of
cause
root
by
sorted
Data
LGS Unavailability
PBAPS Unavailability Data sorted by root cause of failure....................................................
132
140
149
167
170
175
181
186
192
LIST OF FIGURES
Page
Number
8
................................................
1997
1985
Length
Figure 1. US Gross Mean Capacity and Outage
Figure 2. Mean Capacity Factor by Nation, 1981 - 1997..................................................................9
... 10
Figure 3. US Nuclear Power Plant Operational Capability 1980 - 1997...........................
..... 11
Figure 4. US versus US BWR/4 Median Three year capacity ....................................
Figure 5. Plant Gross Capacity Comparison, US vs. BWR/4 vs. PECO...................................12
Figure 6. Sam ple O perating Cycle Skyline.................................................... ..................................... 17
....................... 18
Figure 7. Sam ple PECO plant data...............................................................................
................... 20
ata
Base..........................................................................
D
event
N
RC
ple
Sam
Figure 8.
...... 23
Figure 9. BWR/4 Days of forced outage by failed system .....................................
Figure 10. Comparison - BWR/4 forced outage days and failure frequency by system........... 24
Figure 11. Lost Generation Capacity by System and Occurrence, Limerick Units 1 and 2...........26
Figure 12. Lost Generation by System and Occurrence, Peach Bottom Atomic Power
........................ 27
Station U nits 2 and 3 .....................................................................................
Figure 13. Comparison, Fraction of five year forced outage hours attributed to each system,
Limerick Units 1 and 2, Peach Bottom Units 2 and 3 and average BWR/4
29
industry data....... ......................................................................................................................
Figure 14. Failure profile number of plants with a given number of failures of given
system s in the five year study........................................................ ....................................... 31
33
Figure 15. Frequency of total forced outage days by plant and system failure............................
Figure16. Estimated probability of mean system at mean plant to cause given number of
hours of forced outage in 5.5 yrs................................................................................................34
Figure 17. Excluded operating cycles.......................................................... ......................................... 47
Figure 18. Worst operating cycles included in BWR/4 analysis...................................................48
Figure 19. BWR/4 Operating cycle analysis of forced outage data.........................49
Figure 20. Comparison of average days of forced outage by operating cycle ............................... 50
Figure 21. Comparison of occurrences of forced outages by time in operating cycle ................ 51
Figure 22. BWR/4 operating cycle analysis of forced outage events back from end of cycle ...... 52
Figure 23. BWR/4 cumulative capability factor...............................................................................53
Figure 24. Daily power generation analysis by operating cycle day ...................................... 55
Figure 25. Mean BWR/4 daily operating cycle coastdown modified capacity..............................56
Figure 26. Cumulative BWR/4 daily operating capacity modified for coastdown ...................... 57
LIST OF TABLES
Page
Number
Table 1. Summary of PECO Lost Generation Capacity Data.....................................................19
Table 2. Cummulative days of forced outage attributed to each sydstem ....................................... 29
Table 3. Statistical nature of system failures causing forced outages at each of 19 BWR/4 plants
......................... 30
over the 5.5 yr. study ........................................................................................
Table 4. Percentage of lost generation capacity attributed to six failure causes ............................. 38
Table 5. Percentage of lost generation capacity attributed to equipment and human factors........42
.. 57
Table 6. Conditional relation of PECO forced unavailability events ...................................
ACKNOWLEDGMENTS
The author wishes to recognize the following people who were instrumental in the completion
of this research.
Professor Neil Todreas whose guidance, patience and connections to industry leaders gave life to
the work.
INEEL's Gary Roberts who provided access to all of the automated industrial data for this
research.
NEI's Vince Gilbert for his review of my research findings.
PECO Energy,'s Joe Grimes, Virginnia Angus and Francis Jordan whose generous donation of
engineering resources and explanation of power plant performance were crucial to the
interpretation of the data collected for this work.
To My Wife Catherine who gave me the support to complete this work.
1. Introduction
1.1 Impetus
capacity
steady improvement in its mean
The US Nuclear Power Industry has realized a nearly
to above 75% in the mid 1990's. This
factor from approximately 60% in the mid 1980's
in 1996 with the first reduction in mean
steady upward trend was noted to begin faltering
same period refueling outages were
capacity noted in seven years, Figure 1. During the
in length. It was unclear whether the
decreasing in length and operating cycles were extending
capacity or whether the shifts in operating
industry had begun to see the impact of a ceiling on
A fundamental question to be answered
regimes had created a period of unsettled performance.
a significant extension operating cycle
was whether utility equipment reliability could support
continue to significantly improve their
length. A secondary question was whether utilities could
performance regardless of operating cycle length.
shown in Figure 2, it does not
When compared to other countries performance, as
All nations with ten or more plants in
appear that the US industry is at its maximum capability.
90
-80
100
-i- US Gross Mean Capacity
90
80
US Median RFO length
",--70
0
70
.
60 0
50
A
40
30
-- ' ''-50 1995 1996 1997 1998
1985 1986 1987 1988 1989 1990 1991 1992 1993 1994
Year
- 19971
Figure 1. US Gross Mean Capacity and Outage Length 1985
1US Gross Mean Capacity data from annual report of nuclear generation capacity in Nucleoics
Nuclear Energy Insight, May/June 1996
Week,1986 -1998, RFO data from
100
"
, 90
,.
80
U
70 -
.
-
-
'
....
60
50
S30-40--0
1980
I--
-
1982
1984
1986
1992
1990
1988
1994
1996
1998
Year
-
-
---
Britain
--
Germany
-
South Korea
....---
-- Canada
-- India
Sweden
-- Finland
--
-+-
Japan
--- X-- France
-
""USA
Russia
Ukraine
- 19972
Figure 2. Mean Capacity Factor by Nation, 1981
plants, is additionally shown
operation in 1997 are shown in this figure. Finland, with only four
as a dashed line as it has had such consistently strong performance. Argentina, Belgium, Hungary
Mexico, Romania, Slovenia, Spain, Switzerland and Taiwan all also had less than 10 operating
plants and achieved mean gross capacity factors of greater than eighty percent in 1997. They
were not drawn in the figure to keep from cluttering it and as they are less representative of the
US with its large number of plants. Of interest the country with the highest annual capacity,
Finland, utilizes one year operating cycles with highly efficient, short, refueling outages.
Although there are dramatic differences in the regulatory environments between the nations,
with thirteen nations operating at greater than eighty percent mean capacity, it appears that there
should be room for further improvement within the US industry from its current mean capacity
of seventy percent.
Capacity factor is a good indicator of nuclear power plant performance as it is assigned
without interpretation but has shortcomings when analyzing equipment reliability because the
negative impacts of planned maintenance outages and intentional reductions obscure operational
2 Data to generate graph taken from annual report of nuclear generation capacity in Nucleonics Week, 1982 - 1998
performance. Capability factor disregards these intentional losses and is defined as the
percentage of maximum energy generation that a plant is capable of supplying, limited only by
factors within the control of plant management. Recent US capability factors and unplanned
capability loss factors are provided in Figure 3. The median capability factor flattens out and
does not respond as adversely in the post 1995 period as the mean capacity factor. This reflects
the strong impact of a minority of plants' which experienced poor performance. With the mean
so high and the negative impact of a forced outage so large, this is not surprising.
A collaborative research project was executed with PECO Energy, formerly Philadelphia
Electric Co., to examine the losses of availability at their four Boiling Water Reactor, Class 4
(BWR/4) Power Plants; Limerick Generating Station (LGS) Units 1 and 2 and Peach Bottom
Atomic Power Station (PBAPS) Units 2 and 3. PECO Energy had recently increased the
operating cycles at all four plants from 18 months to two years and was setting industry records
by refueling in less than 30 days. PECO was concerned with the ability of the Balance of Plant
systems to support the extended operating cycles with less time spent in planned maintenance
outages. They had started a program entitled Balance of Plant 700 (BOP-700) with a primary
goal of ensuring that the Balance of Plant systems could support a continuous 700 day run. MIT
.....
100
90
80
70
.
~60 60
-*-US Median Capability Factor
US Mean Capability Factor
a
50
--
o
40
30
-0- US Mean Unplanned Capability Loss Factor
-0- US Median Unplanned Capability Loss Factor
20
10
0
1984
1986
1988
1992
1990
1994
1996
Year
Figure 3. US Nuclear Power Plant Operational Capability 1980 - 19973
3 Data from INPO presentation of Mr.Bill Webster, January 13, 1998
1998
entered into a cooperative research project with PECO Energy centered around BOP-700 The
goals of this research agreement were to: A) Determine single and conditional points of failure,
which would include (1) a review of local records for past sources of lost capacity, (2) a review of
industry records for sources of lost capacity, and (3) an analysis of systems to determine potential
single or conditional points of plant failure. B) Develop a process to utilize the broadest
practical knowledge base to improve Balance of Plant reliability. C) Establishment of a Focus on
Improvement Team (FIT team) process to find engineering solutions for weak system and
components.
1.2 Confinement of Study to US BWR/4 Plants over a Five Year Period
The BWR/4 plant was selected as a case study plant representative of the industry
performance. The research agreement with PECO provided a detailed source of information for
BWR/4 performance at four specific plants. The US BWR/4 performance was seen as
characteristic of US fleet aggregate performance, see Figure 4, and large enough with 19 plants to
provide a significant sample yet small enough in number that every outage record could be
reviewed. The plants designs were similar in nature such that the experiences of one plant could
be applied to others in the base. Plant performance was reviewed over a five year period to
80 75 -
70'
..
65
0
60
US Median DER net capacity
50
1985
US BWR/4 Median DER net cap
I
55-
1986
1987
1988
1989
1990
1991
1992
1993
1994
Year (center year of three year average)
Figure 4. US versus US BWR/4 Median Three year capacity 4
4 Data to generate graph taken from annual report of U.S. capacity factors in NuclearNews, 1987 - 1996, DER is the Design Electnc
Rating
of current
provide sufficient depth to the analysis at each plant and yet remain representative
operation conditions.
Figure 5 presents the relative performance of all US nuclear power plants, the BWR/4
plants used for this study and the PECO plants used for detailed analysis. Histograms of plant
each
gross electrical capacity for the first and last years of the study are presented. The height of
10
70
80
90
to
80
to
90
to
100
70
to
80
40
50
60
to
50
to
60
to
70
40
to
50
10
20
30
to
20
to
30
to
40
0 to
90
to
100
60
to
70
30
to
40
10
to
20
80
to
90
50
to
60
20
to
30
Oto
10
x-axis - Plant annual gross capacity percentage for year of histogram
5
Figure 5. Plant Gross Capacity Comparison, US vs. BWR/4 vs. PECO
s Data to generate graph taken from annual report of U.S. capacity factors in Nucleonics Week, Feb. 11, 1192 and Feb. 12, 1998
5
column represents the number of plants which had the indicated range of gross electrical
capacity for the given year. The green column represents the entire US nuclear industry, blue
BWR/4 and red PECO. The annual gross capacity presented here is more volatile than the
three year average data presented in Figure 4 but has the advantage of presenting a more time
sensitive response to current performance. Note that the overall 1992 performance of the US
industry is bimodal with a normal in appearance distribution of plants centered at 75% gross
electrical capacity and a small number of plants producing no or little power. In 1997 the
distribution has become even more separated. The performance of the better plants has
improved with the median of the upper distribution moving up to 85%. In the same year the
number of plants producing no or little power has dramatically increased. The BWR/4
performance and the PECO performance are seen to be characteristic of the US performance
with the exception of the increased fraction of members producing no or little power. The
issues surrounding the long term shut downs are regulatory and administrative in nature. As this
research is focused on improving plant system reliability it is basically focused on the upper
performance group. A better understanding of the increasing trend towards long term
shutdowns is a topic worthy of further research.
1.3 Research Goal
The primary goal of this research is to explore the capability of existing US Nuclear
Power Plants to reliably support extended operating cycles. A secondary goal is to formulate a
strategy to continue the process of improving industry wide plant reliability. These goals will be
accomplished through a detailed review of on site BWR/4 system failures at the four PECO
plants and a broader review of NRC BWR/4 utility record data. Failures will be categorized by
the primary system responsible, root cause, causal nature, time in the operating cycle of
occurrence and precursor events. This report should provide the reader with a qualitative
understanding for the causes of lost generation capacity, the ability of installed systems to
support extended operating cycles and methods that could by used to correct short comings.
1.4 Design of Research and Presentation of Report
Several studies outlined below were conducted to support the above goals. The sources
of data for these studies will be summarized and then each of the studies reviewed. These results
will then be applied to develop a means for the US nuclear industry to improve its capability and
for the project to assess the impact of extended cycle operation on utility reliability.
A.
Systemic review of failures - Categorize all reductions in generation capacity by the single
system most responsible for the reduction. Develop statistics for the impact of system
failures with respect to the frequency of causing a reduction in generation capacity and
the cumulative lost generation potential attributed.
B.
Causal Failure Analysis
i)
Categorize all reductions in generation capacity by one of five component failure
causal factors; age, design, fabrication, installation, maintenance, and operation.
ii)
Categorize all reductions in generation capacity by the root cause of their failure
as either caused by human or equipment factors. Further subdivide these
categories where data is sufficient
C.
Operating Cycle Forced Outage Analysis
i)
Categorize all reductions in generation capacity by the time in the operating cycle
in which they occur. Develop statistics for the likelihood of failure with time in
the operating cycle and the expected lost generation with time in cycle.
Determine the cumulative effect of forced outages with time in operating cycle.
ii)
Compile the daily average generation of each plant in the study over five years
of completed operating cycle performance. Express the data in terms of
capacity versus day in operating cycle. Modify the data to disregard start up
power ascension and coast down. Use this aggregate data to cross check the
similar forced outage data above and to note the effect of planned maintenance
outages excluded from the above data.
iii)
Analyze forced outages to determine if specific operating cycle events such as
refueling maintenance or planned maintenance outages present a measurable risk
to follow on failures.
D.
Work with PECO Energy in their development the Focus on Improvement Team (FIT)
process and apply lessons learned from this process to a generalized industry
improvement process.
1.5 Organization of paper
Following this introduction chapter two will present an overview of the uniquely
compiled data sources. Chapters three through five will present each of the three broad studies
outlined above. Lessons learned from PECO Energy's Focus on Improvement Team process,
with regard to plant reliability improvement, will then be presented in chapter six. Chapter seven
will summarize the accomplishments of this research and how its findings can answer the
research goals. An attempt has been made to include all original material constructed for this
research in the appendix section to assist others in continuing research on this topic. Definitions
for unfamiliar terms may be found in appendix 4, the NRC glossary of terms.
2. Sources of data
The following three major sources of data were utilized for this research; daily average
power data sent to the NRC by all nuclear utilities and compiled by INEEL, monthly reports of
significant events sent to the NRC by all nuclear utilities and compiled by INEEL, and PECO
Energy plant specific data. The NRC records were valuable in that they are extensive, have strict
reporting requirements and cover all power plants in the United States. The data base work
done with these records by INEEL made the records easily accessible as manageable sections of
detailed data could be located rapidly. Access to the engineering staffs of the four PECO Energy
power plants with their plant records provided a detailed ability to review specific reliability
issues with the engineers and management staff that resolved the equipment problem in
question. This means of directly accessing the root cause of equipment failures and the
vulnerabilities of systems was not possible with the broader NRC records. The cooperative
research agreement with PECO Energy also provided a professional source with which to
review the NRC records. The derivation of each data base is discussed separately below.
2.1 PECO Energy Plant Specific Data
The daily power generation history "skyline" for each of the four PECO plants was
reviewed for the five year period of the study. A skyline for one operating cycle of a sample
plant is shown below in Figure 6. Each reduction of greater than 10% was analyzed in detail.
Examples of forced outages and undesired reductions in power are highlighted in red. Examples
of operationally required reductions in power are highlighted in blue (W - water box cleaning, R
- rod pattern shifts, PMO - planned maintenance period). Forced outages and undesired
reductions in generation capacity were analyzed in detail. An example of the data collected is
provided in Figure 7. The following characteristics assigned to each period of lost generation,
column designations are provided in parentheses were ambiguous;
Effective Outage Days- computed from the generation capacity that the plant would
have been capable of generating had the failure not occurred.
Refuel 9/18193 - 11115193 (58 days), Op period 11115193- 9/21/95 (676 days)
100%-80% -
060%
---
20% 0%
0
100
200
300
400
500
60
Days in operating cycle
Figure 6. Sample Operating Cycle Skyline
System - One system was assigned as responsible for each outage. In some instances this
was a complex decision as a reduction in power may have been taken to repair several systems.
event
If a start up from a forced outage was prevented due to the failure of another system, that
was treated as a separate outage. An attempt was made to limit the number of system categories
such that related failures could be noted. Approximately twenty system categories were
developed for all of the failures noted which resulted in a significant reduction in generation
capability.
Failure details (Date, Type, Component, Failure, Cause, Issue #) - Details of the
component failure and load reduction were noted for future use such as the type of plant loss
of the failure, the
(scram, load drop, run back), the mode of component failure, a description
failed component, the cause of the failure and the PECO failure record number.
Root cause of failure (Category) - Each failure was categorized as either an equipment
failure or a human failure. When possible, these two broad categories were then further
subdivided into the following sub-categories:
Human Factors:
Procedural Inadequacy
Craftsmanship
Operator Actions
Poor Corrective Maintenance
Management Standards
Less than Adequate Corrective Actions
Equipment Factors
Weak Design
Worn Parts
End of Life
Component failure mode (Class) - Each failure was additionally categorized by one of six
of the following failure attributes to better understand the nature of the failure and the possible
impact of different operating strategies.
Age
Operation
Maintenance
Design
Fabrication
Installation
Conditional Relation of Failure (Dependence) - If the failure was significantly
influenced by another operating cycle event such as the previous refueling outage, the previous
planned maintenance outage or an operational power transients this link was noted for the future
correlation of failures.
Time in operating cycle of failure (days for start of cycle, days from end of cycle) - The
number of days following start up from the refueling outage to the event were noted as well as
the number of days from the event to the end of the operating cycle. This data would be later
used to determine the risk of incurring failures with respect to the time in an operating cycle.
The overall statistics for the specific PECO five year plant data collected for the study is
DAYS FROM DAYS FROM
START OF
END OF
DepenCYCLE
dance
CYCLE
RFO
682
1
Date
3/17/93
Type
Load drop
Eff Out Days
030
System
SWC
Component
Valve
Failure
Valve
mispostioned
Cause
LTA procedure
Category
HFPI
Issue #
93-03-24
Class
O
3/26/93
Scram
215
EHC
#6 ISV
Perturbation in
ETS/RETS
Air entrap in
control pack
HF/PI
93-03-38
O
10
673
RFO
4/7/93
Load drop
0 23
FW
FWLCS
'A' level down
spike
Spurious,
Indeterminant
Spurious,
Indeterminant
EF/WD
93-04-04
D
22
661
None
34
649
None
4/19/93
Load drop
008
FW
FWLCS
Master level
controller down
spike
EFIWD
93-04-18
Figure 7. Sample PECO plant data
D
Table 1. Summary of PECO Lost Generation Capacity Data
Site
Total occurrences
Average occurrences per year per plant
Total effective full power outage days
Average outage days per plant per year
LGS
97
9.7
136
13.6
PBAPS
99
9.9
193
19.3
provided in Table 1. The complete data set collected is provided in Appendix 1 for LGS and
Appendix 2 for PBAPS.
2.2 NRC Monthly reports, INEEL MORP 2
This data is historically referred to as "Gray book data". Entries are required in this
record for any significant event and any event which affects reactor power level or operating
conditions. There are over one thousand entries for the BWR/4 plants studied over the five
years of the study. Many of these entries are of little quantitative use as there is no associated
record made of the total effect of the event on generation and many events are included which
do not relate to equipment reliability. One of the specific reporting requirements for this record
is to note any event which removes the main electrical generator from service and the hours that
the generator was off line as a result of the event. The data record was then truncated by only
considering events for which there were recorded generator off line service hours and events
which were the result of an unplanned equipment failure. These remaining entries are
considered equivalent to the accepted definition of a forced outage. It is noted that the recorded
impact of each outage is in lost time of generation not in lost generation power. The net result
was a file of 235 forced outages at the 19 BWR/4 plants over the 5 years of the study or 2.5
forced outage events per plant per year.
While the narrative descriptions of the events in the NRC record were very useful, the
outage causing system and component listings were inconsistent. Each NRC record of events
for the 19 BWR/4 plants were therefore reviewed with a PECO engineer to assign a failed
system consistent with the designations used earlier. The human versus equipment failure
attributes were additionally assigned based on the narrative and the recorded fields of outage
type, method and reason. A sample of the compiled data is provided in Figure 8. The complete
UND
260
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4/15/94
OUT
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OUTG
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14
61
OJTG OUOUTO
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TYPE M
REN
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260
196
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260
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3
13
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H
OTGU
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JJ
TIS
C
T
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ry
rmnentAiM HFACS
DUREGULATORS
UTOMATIC
SCRAMCAUSEDBYBALANCEOFPLANT
AUTOMATIC
SCRAMCAUSEDBYM
DAYS OM DAYSROM
STARTDOPENDOF
LB
OUIT
CO MP
DESCRIP
DURINGPLANNED
MAINTENANCE
ACTITIESONTHE
UNIT2AUTOMATICALLY
AiRHEADER,
SCRAMPILOT
SCRAMMEDON LOW SCRAMAIRHEADERPRESSURE
AND
HBOT
PRMARY
OF
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FOLLOWING
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Y
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315
167
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9
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260950
78
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Y
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EF
Y
26E
EF
Y
CAUSED
BYAFAILED
POWER
SUPPLYTO
BOTH
LEVEL
DRAIN
LOOPS
FORTHE
OFFGASCONDENSER
CONTROL
VALVES
REPLACED
FAILED
ELECTROYTIC
CAPACITOR
CONDENSER
INTHE
POWER
SUPPLY
FORTHEOFrGAS
VALVES
DRAIN
Figure 8. Sample NRC event Data Base
data base is included as Appendix 3. A glossary of the NRC terms used in the data base is
included as Appendix 4.
2.3 NRC Daily Reports, INEEL MORP 3
The daily average power from each power plant is reported to the NRC in a monthly
table and compiled by INEEL as MORP3 data. This data was used to recreate a skyline profile
for each plant in the study for each operating cycle. There was some inconsistency noted in the
reporting of gross versus net electrical generation. To resolve this issue and focus on equipment
reliability concerns, the data from each operating cycle was modified from megawatts of electric
generation to percentage of generation capacity such that the highest output power in the
operating cycle reflected 100 percent generation. Additionally the data was indexed to the day in
a plant's operating cycle vice the calendar day. This would allow the operating cycles to be added
in a parallel operating cycle manner presented later in the report. A separate record was
generated for each operating cycle in which the coast down portion was modified to reflect
generation capability vice capacity. This daily generation capacity and capability data base has
exceptional accuracy with which to characterize the cyclic performance of the utilities owing to
the greater than 30,000 record entries for both capacity and capability. The power generation
profile of Figure 6 was generated by this data base. The capacity and capability skylines for each
of the operating cycles used for this research study are provided as Appendix 5. These skylines
were additionally used to verify that every forced outage was captured by the monthly data bases
described above.
2.4 INEEL NRC Report Formats
The INEEL electronic databases were found to be extremely useful for this research. A
sample of each record has been included as Appendix 6. There are four basic records; MORP1
which provides a monthly summary of information for each plant such as points of contact,
ratings, generation and capacity factors for the month and the year to date, MORP2 which
contains the event reports discussed earlier, MORP3 which provides a monthly summary of the
daily average electrical power generation for each plant, as discussed earlier, and lastly STATUS
which is a summary of the daily status report made by the plant to the NRC.
3. Initiating System Failure Analysis of Plant Unavailability
This portion of the research analyzes the impact of system reliability on BWR/4 utility
performance. BWR/4 system failure data of events which resulted in a forced outages were
retrieved from the broad NRC data base. This information was contrasted to specific plant data
obtained from the four PECO Energy BWR/4 plants. Each forced outage or period of reduced
generation analyzed was attributed to a single system failure. This focused attention on those
system failures which have a direct impact on plant performance. This is contrasted to an
unrestrained study of equipment repairs or system failures which would give equal weight to the
many failures which do not affect overall plant performance.
It is also possible that a maintenance outage could have been taken to repair several
faulty systems and all systems but the one credited with the failure would have their failure rates
underreported. The supposition was made that if a system was weak enough to fail frequently it
would not be able to avoid identification eventually as the system responsible for an outage. As
this study covered 19 plants for a 5 year period, nearly 100 reactor years of operation were
analyzed. This time frame should have been large enough to capture the signature of a system
that is significantly impacting plant reliability. Additionally while many systems may be worked
upon during a mid-cycle outage the supposition was also made that the decision to come off line
was primarily the result of a significant singular system failure. The identification of this critical
failure was a difficult one for which the assistance of the collaborating PECO staff was essential
3.1 BWR/4 System Reliability Impact on Plant Capability
The NRC records for every removal of the main generator from line at each of the 19
BWR/4 plants during the five year study was reviewed to determine whether the outage was the
result of a system failure. The term system failure was broadly interpreted to include any event
which resulted in the undesired or unplanned loss of generation. Many outages reported as
planned outages actually belonged based on our assessment in this category. For example, an
outage planned two weeks in advance to repair a leaking seal on a pump would have been
reported in the NRC data as a planned maintenance outage but was assigned as a forced outage
for this study as the outage would not have been taken had the pump not failed. Some planned
maintenance outages were taken as part of the scheduled operating cycle and were not in this
plant reliability study as a forced outage. These were generally noted to occur prior to a refueling
outage to conduct general plant shut down maintenance.
Two hundred and thirty five outages were identified in the study for a mean outage rate
of 2.5 outages per plant per year. Five thousand four hundred and seventy nine days of outage
were accumulated for a mean lost generation of 58 days per plant per year. This lost capacity is
very large and would account for a loss of 16% capacity. Figure 9 provides a summary of the
annualized days of forced outage per system. It is noted that several significant system outages
dominate the results. Browns Ferry Unit 1 remained shutdown for the entire five years of the
study as the result of TVA and NRC operational safety concerns. Note that the category of
operations was created for operator-related failures. Brown's Ferry Unit 2 was additionally
shutdown for the first 3 /2 years of the study over the same issues. Brunswick Units 1 and 2
were shutdown one year and Cooper Station for nine months as the result of diesel technical
specification discrepancies. These five shutdowns, which are administrative in nature, have a
35
-.
IM
30
30
S251
25-
10
S20
R>15
S5
c
-
Significant Diesel Outages
Brunswick Units 1 & 2
year
Diesel technical specifications
Significant Operation Outages
Browns Ferry Unit 1
entire five year study
TVA and NRC concerns
Cooper Station
Browns Ferry Unit 2
Over 3 1//2 years of study
TVA and NRC concerns
9 months
Diesel testing requirements
10
i-
0
a)
i
a
E
W
7
E::::
"
0
Figure 9. BWR/4 Days of forced outage by failed system
dominant effect on the overall system reliability results. While this risk of large losses due to
regulator action is a significant risk of nuclear power generation, its effects obscure the impact
that individual system reliability have on overall plant capability. Hence, to limit the total impact
of any singular failure, a maximum outage length of one month was assessed for each outage.
This resulted in the truncation of 9 of the 235 outages to 30 days. With a resultant 14.3 days of
modified forced outage per plant per year or an average forced outage length of 5.8 days. This
yielded a mean forced outage rate of 3.9%.
Figure 10 provides a summary of the occurrences of system failures which resulted in a
forced outage and the total days of forced outage accumulated by those system failures, as
modified above. The gray bars represent the occurrence rate of each system failure resulting in a
forced outage. Main Turbine, Electro-Hydraulic Control System (EHC) and Feedwater (FW)
system failures are noted to most frequently be the cause of plant forced outage. Colored bars
represent the mean annual days of forced outage attributed to each system, with the appropriate
scale on the right. The colors of the bars represent the groupings of the systems. Balance of
= Balance of Plant Forced Outage Days
16
14 -=
Primary / Safety Systems Forced Outage Days
Yellow
12
0=
10 -
2.40
2.40
= Forced Outage from External Sources Days
1.90
Forced Outage Frequency Occurence
-
4a
a
-d 1.40
0
0.90
S6
4
0.40
ca
2
I
u
C
I
I
-
I
1
f
1.
-V .1
U
System
Figure 10. Comparison - BWR/4 forced outage days and failure frequency by system
Plant systems, those systems not included in the primary and safety related system category, were
colored red. Primary or safety related systems which have historically received much greater
maintenance and design attention have been colored blue. External system failures such as a loss
of the electrical distribution grid resulting in a forced outage have been colored yellow. Note that
the percentage of the total lost generation capability attributed to one system can be read on the
left vertical scale while the average annual days of unavailability per plant attributed to a given
system is read on the right vertical scale.
Balance of Plant system failures were found to initiate 69% of the forced outage
occurrences and account for 59% of the lost generation. Primary/safety related system failures
initiated 27% of the forced outages and accounted for 38% of the lost generation. It is observed
that the Balance of Plant failures initiated the vast majority of the utility lost generation events
and a lesser majority of the aggregate lost generation capacity. Balance of plant system failures
were found to initiate 1.73 forced outages per year per plant of a mean length of 4.9 days.
Reactor/safety related system failures were found to initiate 0.7 forced outages per year of a
mean length of 8.1 days. It is noted that Balance of Plant system failures more frequently initiated
a forced outage while the resultant forced outage was of a shorter duration than those initiated by
a Primary/safety system failure. This supports the intuitive understanding that Reactor/safety
related systems are much more reliable due to redundancy of design, higher quality components
and more tightly controlled maintenance and operation but that a failure of one of these systems
significant enough to force an outage is more difficult to recover from than a Balance of Plant
system initiated forced outage.
Main Turbine system failures caused the majority of the forced outages and consumed
the majority of the lost generation capacity. These failures were mechanical in nature and
included excessive turbine vibration, failure of the turbine (loss of buckets and related significant
damage) and control valve leakage. The vulnerability of this large piece of capital equipment is
evident. Electro-Hydraulic Control system failures were noted to frequently initiate forced
outages but to be accumulate proportionally less outage time. As this system governs the steam
demand of the reactor it is self evident how its failures could so readily end in a plant outage.
These failures tended to be predominantly of a piping system or control system stability nature.
The operations system category was created for forced outages which were the result of
regulatory or management operator safety concern. This category did not include the
misoperation of a system or regulatory technical specification outages which were attributed to
the system with which there was concern. While four out of the seven outages in this category
were truncated to 30 days, the system grouping still accounted for a significant portion of the lost
generation capacity. It is noted that the relative failure frequency is low in comparison to the lost
generation capacity because these outages are individually very long. One can then characterize
the predominance of a system to cause short or long outages when they do occur by contrasting
the gray with the colored bars of Figure 10 or the frequency of outage with the mean annual lost
capacity. The diesel is noted to pose a significant risk for lengthy outage while the generator is
seen to pose little risk for a lengthy outage. A better understanding of the types of system
failures in each grouping can be gained from a close review of Appendix 7. This appendix is the
compiled NRC forced outage data used in the study sorted by failure system and then length of
outage.
3.2 PECO Energy Plant Specific Data
System failures at the four PECO Energy BWR/4 plants which resulted in a loss of 10%
generation capacity or more were analyzed in a manner similar to that discussed above. It should
be noted that this is a stricter analysis constraint than the forced outage limitation of the BWR/4
= Balance of Plant Lost
20
Primary / Safety Systems Lost Capability
Yel ow
15
= Lost Capability from External
1
= Outage occurrences
o10
0
5
bo
0~
C
1
3
0
Q
1A
F Geneatio
11 L
E
E
L
,
Capcit
~
~A4ystean
by
'
-
=
-lLL
1E
E
1E
E
1EE
U
Figure 11. Lost Generation Capacity by System and Occurrence, Limerick Units 1 and 2
NRC data (100% power reduction). The compiled data for the LGS site is presented in Figure
11. Balance of plant systems failures accounted for 74% of the periods of reduced generation
capacity and 54% of the total lost capacity while Reactor/safety system failures resulted in 24%
of the periods of reduced generation and 42% of the lost capacity. These results were similar to
those found with the greater BWR/4 forced outage data. Balance of plant system failures more
frequently impacted generation capability but had a smaller impact with each failure.
The EHC (Electro-Hydraulic Control) system failures were dominated by piping failures
and control system instabilities. The large generation loss attributed to Main Steam system
failures was the result of three Main Steam Relief valve maintenance shutdowns. This weak
component obviously poses a high risk to efficient operation. The Recirc (Recirculation) system
grouping failures were characterized by control system instabilities, age related failures and
maintenance failures. Details of other LGS system failures can readily be found in Appendix 8
which is a listing of the reduced generation periods sorted by the initiating system failure and
= Balance of Plant Lost Capability
= Primary / Safety Systems Lost
Capabiltiy
= Lost Capability from External
Sources
= Occurrence of Lost Capability
Events
5
o
i
It5
B'
-o
E:
0
0
0
I
.r
B
Co
o
0-
C
.0
E
-o
0
-S
n0
System
Figure 12. Lost Generation by System and Occurrence, Peach Bottom Atomic Power
Station Units 2 and 3
then length of outage.
Periods of reduced generation capability were analyzed at PBAPS in a similar manner.
Figure 12 provides a similar comparison of the lost generation by the initiating failed system.
Again at Peach Bottom the vast majority of the reductions in generation were initiated by a
Balance of Plant system failures (60% vs. 35%). The site experienced numerous outages to
repair mechanical problems with the Feedwater (FW) system as noted by the frequent
occurrence of FW events. It is also noted that Low Pressure Coolant Injection (LPCI) and
Containment experienced one failure in each system which required a shutdown and
containment entry for repairs. The large impact of a single significant failure is noted by the
disparity of the lost power to frequency of occurrence.
3.3 Consistency of System Failures BWR/4 Fleet versus PECO Plants
As a final comparison it was desired to compare the system failure results found at the
PECO plants with the broader NRC BWR/4 data base. To perform this comparison the PECO
data bases had to be reduced to only contain forced outage events, those removing the generator
from service. The cumulative days of forced outage noted in each of the studies is presented in
Table 2. Note that only those systems accounting for greater than 2% of the total lost capacity in
one of the three groupings is reported. All showed a significant effect of Balance of Plant failures
on plant availability causing the significant majority of unavailability periods. Additionally all data
bases demonstrated the difference in the relative impact of a Reactor/safety related failure as
compared to the Balance of plant system failures.
Table 2. Cummulative days of forced outage attributed to each sydstem
NRC ALL BWR DATA
PBAPS FORCED OUTAGE DATA
LGS FORCED OUTAGE DATA
5.5 yrs, 19 plants
1/92 to 6/97 5 yrs
7/92 to 6/97
1/92 to 12/96
5 yrs
Days outage
Days Outage
Days Outage
System
Total
%Total Unit 2
Unit 3
Total % Total Unit 1
Unit 2
Total
% Total
Main Turbine
188.7
13.9
0.0
5.0
5.0
3.0
0.3
0.3
0.6
0.6
9.2
5.1
19.5
24.5
20.8
112.1
8.3
15.0
0.0
15.0
EHC
FW
105.2
7.8
0.0
7.1
7.1
4.3
0.6
3.1
3.7
3.1
Diesel
0.0
104.7
7.7
0.0
0.0
0.0
0.0
0.0
0.0
0.0
3.0
9.3
7.9
12.5
7.6
6.4
4.3
8.9
3.6
Electrical
57.7
Generator
42.3
3.1
9.3
15.2
9.2
0.0
3.2
3.2
5.9
2.7
SWC
34.4
2.5
0.0
0.0
0.0
0.0
2.4
2.7
5.1
4.3
Condenser
31.1
2.3
0.0
0.5
0.5
0.3
0.0
5.0
5.0
4.2
FW heating
23.6
1.7
7.9
0.0
4.8
0.3
7.9
0.7
1.0
0.7
Air Removal
21.4
1.6
0.0
0.0
0.0
0.0
6.5
1.0
7.5
5.6
Condensate
16.8
1.2
14.3
0.0
14.3
8.7
0.7
3.7
4.4
3.2
Offgas
0.4
0.0
1.2
0.9
13.2
8.0
1.2
0.0
1.2
1.0
50.9
22.9
90.6
55.1
23.5
42.1
65.5
BOP sub tot
801.7
59.1
56.7
1.2
8.0
9.2
7.8
122.6
9.0
13.1
5.4
18.5
11.3
Recirc
13.0
11.0
6.5
0.0
12.7
12.7
7.7
13.0
0.0
Reactor
87.7
Operation
4.5
60.8
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
22.4
19.0
0.0
21.3
1.1
0.0
0.0
0.0
57.5
4.2
Main Steam
2.7
2.3
3.0
0.0
0.0
0.0
0.0
2.7
0.0
RPS
40.7
Drywell
32.2
2.4
0.0
0.0
0.0
0.0
4.6
0.0
4.6
3.9
0.0
0.0
7.9
0.0
0.0
0.0
13.0
13.0
29.2
2.2
Containment
DC
10.6
0.8
0.0
4.1
4.1
2.5
0.0
0.0
0.0
0.0
LPCI
0.0
0.0
0.0
14.1
14.1
8.6
0.0
0.0
0.0
0.0
Reactor sub tot
510.3
37.6
13.1
49.3
62.4
38.0
42.8
9.1
51.9
45.1
6.9
1.6
3.1
4.7
4.0
42.9
3.2
2.2
9.0
11.3
Transmission
Grand Total
70.7
93.5
164.2
100.0
1356.9
100.0
67.7
58.4
126.1
100.0
All BWR IV
5
PBAPS 3
0E
PBAPS2
'PZP
w
C
2)
~LGS
E
E11~
2
LGS 1
System
Figure 13. Comparison, Fraction of five year forced outage hours attributed to each system,
Limerck Units 1 and 2, Peach Bottom Units 2 and 3 and average BWR/4 industry data
Figure 13 contrasts the relative days of forced outage between the two Limerick units,
the two Peach Bottom units and the BWR/4 utility average, the data of Table 2. The specific
plant data appear spiked in comparison to the slowly varying BWR/4 data. This spiked
appearance demonstrates that individual plants experience only a small portion of the problems
experienced by the broader industry. Table 3 presents the statistical attributes of the forced
outage data. This table only presents those 14 systems which contributed to greater than 2% of
the total lost generation and was generated solely with the NRC data over a 5 1/2 year period to
cover both of the five year PECO studies. The data represents the rate with which each of the
19 plants encountered forced outages as the result of the indicated system failures. For example,
in the 5.5 year BWR/4 study 14% of the forced outages were the result of Main Turbine system
failures. This system accounted for 18.8 % of the total outage days as modified by the 30 day cap.
The mean number of Main Turbine related forced outages per plant for the 5.5 years of the
study was 1.8 and the mean total number of days of forced outage that a plant experienced
during the study as a result of Main Turbine failures was 13.4 days. Note that in every category
the standard deviation of the distribution was greater than the mean. This implies a very
irregular distribution.
Figure 14 has been constructed to provide a visual image of the probability with which a
given system will cause a forced outage. The horizontal axis is the system listing, the axis out of
Table 3. Statistical nature of system failures causing forced outages at each of 19 BWR/4
plants over the 5.5 yr. study
Accumulated lost days
Frequency of Occurrence
System
Main Turbine
EHC
FW
Diesel
Generator
Electncal
SWC
Condenser
Recirc
Reactor
Main Steam
Operation
Drywvell
Containment
% total
occurence
14 5
132
8.5
2.1
6.0
47
34
38
72
3.4
3.8
0.9
2.1
09
Mean
18
16
11
0.3
0.7
06
04
05
0.9
04
0.5
0.1
03
01
Median
1
1
1
0
0
0
0
0
1
0
0
0
0
0
Max
Min
Mode
1
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
10
8
3
1
3
2
3
3
3
3
3
1
1
1
St dev
2.3
2.1
10
0.5
0.9
0.7
08
0.8
1.0
0.8
1.0
03
0.5
03
%total
modifed
lost days
188
8.3
7.8
73.3
3.1
4.3
25
23
90
67
42
237.6
2.4
22
Mean
13.4
59
55
52.4
2.2
30
18
16
6.5
48
30
1697
1.7
15
Median
21
2.7
2.9
0.0
0.0
00
00
0.0
18
0.0
00
0.0
00
00
Min
Max
0 151.7
20.7
0
20.8
0
0 389.1
7.7
0
24 2
0
253
0
84
0
361
0
33 6
0
24.4
0
0 1825.0
14.7
0
25 4
0
St Dev
35.7
71
67
123.8
29
5.8
57
2.8
10.1
9.1
74
513.2
3.6
58
30
the paper is the number of failures reported by a plant for that system in the study, the vertical
axis is the number of plants which reported the given number of failures of the given system
during the study. All of the columns should add up to 19 in the depth direction, as there were 19
plants in the study. The more frequent that failures are at all plants, the more forward the
columns should move. The more consistently a given failure occurs in the industry the more
even all columns in a given number of failures are. If the average plant experienced one forced
outage per year from a given system then the five outages column would register nineteen plants,
as the study covered a five year period. It is observed that most of the columns crowd the back
wall. This means that most of the plants did not experience any forced outages from these
systems. The system with the worst track record will be explained to ensure that the chart is
understood. For the Main Turbine system, five plants did not experience a forced outage as the
result of the system during the study, six plants experienced one forced outage, three plants
experienced two forced outages, four plants experienced three forced outages and one plant
experienced ten forced outages (six or more category on the plot). The most frequent forced
outage rate of one indicates that it is most likely for the Main turbine to cause one forced outage
in five years and a half years at any plant in the study.
16
2
14
12
No failures
4
One Failure
2-
_ Twofailures
Three failures
"Four failures
0
z
failures
Six or more
oFive
V
2.
.
0
.
2
Figure 14. Failure profile number of plants with a given number of failures of
given systems in the five year study
Figure 15 has been constructed in a similar manner using the total number of days of
forced outage accumulated by each plant for the given system during the study vice the
frequency. The histogram bins have been laid out in five day increments. The back wall should
be identical to Figure 14 as it represents the number of plants which did not experience any
forced outages from that system for the full study. This figure uses the unmodified data, that is
the outages have not been truncated to 30 days. Note that if one plant experienced greater than
45 days of forced outage it is represented by the histogram bars in the foreground. This is the
grouping that takes up all of the extra long outages. The first column away from the back wall
represents all those plants which experienced from 0 to 5 hours of forced outage as the result of
the indicated system failure during the study. The likelihood of large cumulative outage impacts
from a single weak system is noted to be small while the likelihood that many systems will
produce smaller individual losses is noted to be large.
12
10O
8
6-
>0 - 5 days
>5 0- 10 days
>10 to 15 days
>15 to 20 days
S> 20 to 25 days
25 to 30 days
> 30 to 35 days
> 35 to 40 days
>40to45 days
> 45 days
42->
0
0
M
U
Figure 15. Frequency of total forced outage days by plant and system failure
Figures 14 and 15 can provide significant insight to the likelihood of a given system to
initiate a forced outage. Recall that both of these histograms have their systems presented in a
descending order of cumulative lost capacity attributed to the system. Additionally Balance of
Plant systems have all been grouped together and then Reactor/safety systems, see Figure 10.
Balance of plant systems are therefore represented by the first 8 columns from the left edge of
Figures 14 and 15. Note that while the cumulative impact of individual BOP systems is
descending as the columns go to the right, in each case four to six plants reported one forced
outage from each of these systems during the entire study. This equates to a 25% probability of
a single forced outage occurring from each of the BOP systems in five and a half years. It is also
observed that these had nearly the same probability of causing a total loss of 0 to 5 days of
forced outage. This can be taken as the generic minor outage. The FW and EHC systems are
observed to be more problematic. Figure 15 shows plants experiencing a fairly constant, slowly
decreasing distribution of greater cumulative forced outage from these systems. In example, for
the feed water system, 7 plants did not experience a forced outage, 4 plants experienced from 0 5 hours of outage, 3 plants experiencing from 5 - 10 hours of outage, 2 plants 10 - 15 hours, 2
plants 15-20 hours and one plant 20-25 hours of outage. These poorly performing BOP systems
have comparable likelihood for two and three forced outages in the 5 year period of the study.
In summary from Figures 14 and 15 it is observed that each of the eight balance of plant systems
which have historically generated forced outages present a 25% likelihood of causing a single
forced outage of less than five days in a five year operating period. Three BOP systems pose a
greater risk with diminishing likelihood of causing more forced outages and consuming a
commensurately increased amount of generation capability.
Reactor/safety related systems are noted to present a significantly different forced outage
relationship in Figures 14 and 15. Note that as the cumulative forced outage time initiated by
each system failure decreases, see Figure 10, so does the likelihood of failure, Figure 14. The
probability of longer outages is greater than that observed for the BOP systems and the
probability of shorter outages is decreased, Figure 15. The exception to this general observation
0.7
0.7
0.6-
0.6
0.5-
0.5
S 0.4
0.4
0.3
0.3
0.2
0.2 -
0.1
0.1
0
M
L
o
0
E
System failures infive and a half years
0
(o
Figure 16. Estimated probability of mean
system at mean plant to induce given number
of forced outages in 5.5 yrs.
A
A
A
A
A
A
A
Total days of forced outage from systemin 5.5
years
Figure 17. Estimated probability of mean
system at mean plant to cause given number
of hours of forced outage in 5.5 yrs
is the recirc system which could be considered to have many of the attributes of a BOP system.
Figures 16 demonstrates the estimated probability of the mean system to produce one or
more forced outages during the five and a half year study at the mean plant The estimated
probability of failure is the number of failure events experienced in a given system divided by the
total number of events. Note the high probability for a low failure rate and the rapidly
diminishing probability for multiple forced outages from a single system during the study.
Figure 17 demonstrates the estimated probability of the mean system to initiate a given number
of hours of forced outage during the five and a half year study at the mean plant. Again the
highest probability of greater than 60% is for the mean system not to contribute at all to the
forced outage losses. The probability diminishes rapidly as the window of total forced outage
from a particular system increases. These two curves demonstrate the difficulty with using plant
history to improve performance. The greatest likelihood is for different systems to produce
forced outages of fairly short duration. This is observed by the high probability of any system
not to fail and the small probability for single or double failures during the course of the study,
Figure 16, and by the dominance of the short time period of outage by system, Figure 17.
As the plant systems improve in reliability, there will be fewer plant failures, and
therefore individual system failures will become more infrequent. As fewer failures happen at an
individual plant they must use broader sources of failure data to understand plant vulnerabilities
to determine whether they are immune from the failures experienced by others or whether they
could experience the same failure and should modify their system. For example, neither PECO
Generating Stations experienced any diesel related outages, yet that system has a significant
portion of the industry data. Limerick is unique from most of the industry regarding this system
as they have eight independent diesels. Peach Bottom is not unique with only four diesels. As a
further example, it is noted that there are significant forced outages attributed to the Main
Turbine in the industry data but that this system caused few outages at both Limerick and Peach
Bottom. There were 34 industry wide forced outages attributed to the Main Turbine. The
failures were predominately mechanical failures and vibration problems. It is possible that there
are enough differences between the turbines to explain the inconsistency. It is possible that the
time of failure in life is vastly different and that Limericks turbines have not experienced wear
out yet. It is also possible that the failures are infrequent enough that Limerick just has not yet
experienced them during this analysis window. In any case it would be beneficial to investigate
the failures observed at other utilities and apply the lessons learned from those plants to improve
the reliability of the Limerick System against the possible future failure. While a utility is well
aware of their past failures and weaknesses, these plots imply that they must use larger data bases
to understand the potential future weaknesses of their systems.
4. Causal Analysis of System Failure
The goal of this study was to qualitatively understand the nature of system failures which
led to a plant forced outage. The characteristics of failure to study were chosen in a manner that
would answer the two basic questions of this research; whether the present system reliability
could support extended operating cycles and whether utilities could improve overall plant
performance? To examine the first question all system failures were placed into one of six
categories which best descried the root cause of the failure in an equipment centered reference.
This analysis assumes that there are a limited number of basic reasons that any component
unexpectedly fails. The second analysis focused at answering the question of how utilities could
improve performance. This analysis was a deeper root cause analysis of failures which assessed
each system failure resulting in lost plant generation capacity as the result of either a human
failure or an equipment failure. The human failures were seen as functions of how personnel do
their daily jobs while equipment related failures were inherent to the equipment design or its
present maintenance system.
The data used for both analyses in this section was the detailed data taken from the four
PECO plants. This data included all periods of 10% or greater reduction in generation over the
five year study. The assessment of the root cause was very qualitative and required the ability to
interview the system engineers and management staff who corrected the failure. The NRC
records alone were not found to be sufficient to make these judgments for the broader BWR/4
data.
4.1 Equipment centered failure analysis
This first equipment centered failure analysis did not attempt to assess why a failure was
allowed to happen but just the basic nature of the failure. All equipment failures which resulted
in a reduction of generation capacity were placed into one of the following six categories:
age - the component which failed had passed the end of its useful service life. Failures
included worn parts, leaking seals, drifted set points, electronic card failures and
clogged tubes.
Design - component failure was the result of an inherently flawed design. The system
failed with all components operating as designed in the application for which it
was installed. Failures in this category included control system instabilities,
steam relief valve seat leakage (SRV), spurious instrumentation signals and filter
clogging.
Fabrication - the component failure can be traced back to an original fabrication flaw.
The component was not constructed to the required design specifications.
Failures in this category included reactor fuel failures, valve failures and the
failure of recently installed electrical components.
Installation - the component failed as the result of a faulty installation in the system.
Only one failure was noted in this category as the result of foreign material left in
a fluid system.
Maintenance - the component failed as the result of inadequate corrective or
preventative maintenance. Failures in this category included weld failures,
incorrectly assembled components and foreign material left in a system during
maintenance.
Operation - the component was operated outside of its designed range of performance
and therefore induced a plant failure. Failures in this category included operator
error, procedural non-compliance, bumped actuators, the use of equipment
which had already exceeded a maintenance threshold (i.e. continuing to operate
a pump with an alarming vibration sensor).
The cumulative lost generation attributed to each of these component failure causes is
presented in Table 4. The table has been broken down into three major column sections by
each of the PECO sites and then the aggregate for the four PECO units as a whole. Each major
column is then broken down by Balance of Plant systems, Reactor systems and the aggregate for
that category at that site or column grouping. Note that total includes transmission system
failures in addition to BOP and Reactor. As contrasted to the system failure inconsistencies
presented in the last chapter, it appears that the cumulative failure attributes between each of the
Table 4. Percentage of lost generation capacity attributed to six failure causes
Age
Design
Fabrication
Installation
Maintenance
Operation
PECO
PBAPS
LGS
Total
Reactor
BOP
Total
BOP Reactor
BOP Reactor Total
19
14
24
20
21
22
16
5
26
29
21
37
29
24
36
29
18
39
6
9
5
4
0
7
10
22
0
3
0
5
1
0
1
6
0
10
21
23
15
27
31
18
13
11
9
22
32
15
19
23
16
27
44
15
PECO sites is consistent and that one can therefore draw conclusions from the larger data set of
the cumulative PECO data
Component age related failures account for one fifth of the total lost generation
capability. This represents the risk that component wear out poses in the current operating and
maintenance cycle system. It is noted that the Balance of Plant systems have a seventy percent
greater chance than do reactor systems (24% versus 14%) of encountering an age related failure
which impacts generation capability. This is as one would expect with the increased
maintenance standards of the Reactor systems. It is additionally noted that the overall risk posed
by age related failures is relatively small in comparison to the other failure mechanisms. Prior to
extending the maintenance cycle of a component or system this risk should be below an
acceptable level The maintenance cycle can then be cautiously extended based on post
acquisition operating data.. This process is referred to as age exploration by the Naval
Maintenance Command 6 with regard to the extension of shipboard maintenance cycles. There is
not a direct correlation between a components major maintenance cycle and the plants operating
cycle. In example a pump seal which was not examined during the last plant major maintenance
period could wear out during use indicating operation of the component past its useful age but
not excessive plant operating cycle length. If a system were not experiencing any age related
component failures than the process of age exploration could be started. If a component is
experiencing age related failures then alternative maintenance strategies may be developed such
as online maintenance which do not impact the plant operating cycle.
The common failure attribute of weak component design was surprisingly found to be
dominant. Some design issues are long standing and difficult to resolve such as the rapid seat
erosion of SRVs. Balance of Plant design issues were found to be much simpler in nature and
much more prevalent. Control systems instabilities were numerous in which spurious signals
propagated to the controllers resulting in unstable response of the system and an eventual plant
trip. Many of these problems were simple to resolve with minor alterations to the sensor or
6 Naval Sea Systems Command, SEA 915, Age Reliability Analysis Prototype Study, Octobere 21, 1993 prepared by American
Management Systems, Inc pursuant to Contract N00024-92-C4160
controller systems. Another common failure mode in this category was piping rupture due to
equipment vibration. The system reliability was easily improved by isolating the vibration from
the piping. The Balance of Plant systems seem to be inherently flawed in that they were not
designed with adequately high standards for the continuity of power generation. The
improvement of these inherent weaknesses is a good challenge for the engineering staffs and
presents a very quantifiable problem with identifiable corrective actions. The majority of these
design related failures can be eliminated through the conscientious review of system design and
the addition of greater reliability to essential control processes.
The overall impact of fabrication and installation related failures are negligible. This
would be expected with the high quality equipment procured and with the stringent test
programs on initial plant start up. Three events did cause significant impact on the individual
plants within these categories. LGS experienced a 13 day forced outage to replace leaking fuel.
The fuel failure was categorized as a fabrication failure. LGS also had a four day shut down to
repair a leaking steam jet air ejector as a result of improper installation. PBAPS experienced two
fabrication related shutdowns. A turbine control valve stem separated resulting in a 5 day
shutdown and the generator experienced a lock out as the result of a poor solder joint on a
circuit card resulting in a 3 day outage. Improvements to the installation category are effected by
overall maintenance controls while fabrication defects must be overcome by procurement
procedures. The low overall impact of these two categories reflect the high standards in both
categories.
Maintenance related failures accounted for one fifth of the lost generation during the
study. This represents the risk that working a component will cause a subsequent failure not
prevent it. Note that the risk of component wear out from age related effects is identical to the
risk of maintenance failure. This reflects the complex nature of maintenance and the probability
that while it is performed to theoretically restore a component to design specification it may do
harm instead. It is noted that the reactor systems are characteristically more complex and they
have fifty percent more maintenance failures. Much work has been done in this area and it is
understood that the industry works very hard to minimize this effect. The underlying message is
that as much effort should be spent by PECO to eliminate unnecessary maintenance on
satisfactorily performing components at these plants as is spent to improve the maintenance
efforts on those components which are experiencing age related failures.
The operations category directly reflects the complex nature of nuclear power plant
equipment operation. The majority of these failures were the result of the non-procedural
operation of equipment. It is noted that the more complex reactor systems had twice as many
failures. This category represents the greatest challenge to elimination. It is very difficult to
make complex systems easy to operate by a diverse group of people in diverse operating
situations. Human factors engineering efforts should be able to dramatically affect this failure
cause.
In summary the component failure causes noted at the PECO plants during the five year
study reflect great potential for improvement by dedicated engineering actions. The categories
of age, design and maintenance related failures represent 60% of the total lost power. Many of
these failure methods can be easily prevented through the careful analysis of system design,
reliability and the cautious exploration of improved maintenance methods. The operator related
component failures, which also can be affected through improved engineering design, represent
a greater challenge due to the uncertain human factor. The risk posed on component failure by
maintenance periodicity and therefore indirectly operating cycle length is no greater than the risk
posed by maintenance failure. It is predicted that the installed equipment can support a cautious
increase in operating cycle length. The complete list of PECO reductions in generation capacity
have been sorted by component failure cause and are presented as Appendix 10 for Limerick
and Appendix 11 for Peach Bottom for further review.
4.2 Failure Root Cause Analysis
In order to determine the focus of efforts needed to improve power plant capability a
summary was created of the root causes of system failures which initiated periods of reduced
capacity. The root cause was determined by reviewing root cause analysis records at the power
plants and by interviewing engineering and management staff. All failures were initially classified
as either those which were the result of inherent equipment factors or failures that were the
result of a human failure at the power plant. These broad categories were then subdivided into
the following subcategories;
Equipment factors
weak design
worn parts
end of life
fatigued parts
instrument control system
Human factors
procedural inadequacy
craftsmanship
operator action
poor corrective maintenance - repeat failures of a worked component
management standards - management decision to operate with known degraded
equipment
less than adequate corrective action - repeat failure as the root cause of original
failure was not properly diagnosed
It is noted from Table 5 that approximately half of the total lost generation capacity at
each of the PECO sites has been attributed to human vs. equipment factors. It is of interest to
note that the original review of the failures yielded that nearly all were the result of equipment
failures. After detailed review with upper management it was determined that many of the
equipment failures were the result of personnel error. This confirms the well known danger of
conducting a root cause analysis with only records. Note that a subcategory could not be
assigned to every failure and the subcategories do not therefore add up to the parent category.
Table 5. Percentage of lost generation capacity attributed to equipment and human factors
BOP
Equipment Factors
Weak Design
Worn Parts
End of Life
Fatigued Parts
Instrument Control System
Human Factors
Procedural Inadequacy
Craftsmanship
Operator Action
Poor Corrective Maintenance
Management Standards
Less than adequate Corr Action
LGS
Reactor
45
51
17
27
0
13
0
0
0
0
0
0
55
49
9
10
8
13
11
4
0
2
24
0
4
13
BOP
Total
46
22
7
0
0
0
54
9
10
7
1
10
8
PBAPS
Reactor
47
69
20
39
14
9
9
4
0
1
0
0
53
31
20
2
2
4
23
15
0
10
0
0
0
0
Total
BOP
56
29
10
6
0
0
44
10
3
17
10
0
0
PECO
Reactor
46
61
19
34
8
10
5
3
0
1
0
0
54
39
15
5
4
7
18
11
0
7
10
0
2
5
Total
52
26
9
3
0
0
48
10
6
13
7
4
3
42
The greater weakness of the Balance of Plant designs is again noted by the 80% increase
in the weak design root cause of failure. The complexity of the reactor/safety systems is also
noted in the increase in procedural and operator failures. The management standards category
was created to account for the times when engineering staff advised management of a material
condition degradation for which a shutdown was required to repair but management chose to
risk continued operating. This figure does not represent the number of times that management
took the same risk but was successful in making it to the next scheduled maintenance period.
The last category reflects poor troubleshooting of a failure and the restoration of power with an
unresolved problem. These last two categories are singular events for which there was rapid
learning.
The relative weight of categories provides insight to changes that could be made in plant
operation to improve capability. The overall categories shed light as to how much change can
be made by engineering equipment modifications versus improved training or monitoring of
personnel. In the case of PECO both can provide equally promising gains. The complete list of
PECO reductions in generation capacity have been sorted by failure root cause and are
presented as Appendix 12 for Limerick and Appendix 13 for Peach Bottom for further review.
5. Operating Cycle Analysis
This analysis was conducted to identify common strong and weak periods within the
mean operating cycle. The plant operating cycle is defined as the time period from start up
following a refueling outage until shutdown for the subsequent refueling outage. The power
plant is made up of many systems which have many individual components each with unique
maintenance requirements and failure characteristics. While individual components have failure
rates that vary with there time in service since the last major maintenance, it is unclear whether
plant operating cycles have characteristic periods of reduced or improved reliability. Naval ship
system performance studies have found that often the combined effect of many components is a
system with a flat failure probability.7 That is the probability of system failure is nearly constant
with time. Common impacts of the plant operating cycle on components could nonetheless
create observable plant operating cycle failure rate characteristics. This section will determine
whether those common failure characteristics exist and if so their probable impact on extending
operating cycles and the capability to improve overall utility performance.
It is understandable that systems may fail at increasing rates as the time since the last
refueling increases. This would be plausible as the continuos operating period since the last
shutdown maintenance is extending. If there are systems or components which can only be
worked during a shutdown period then their system and component age would increase with
operating cycle age. Additionally these same systems will accumulate maintenance warnings or
indications of impending failure as the operating cycle progresses. Many of these warnings force
plant managers to choose between shutting down to repair the system or to continue operating
until the next scheduled shutdown. In the last section the impact of an incorrect decision to
continue operating was demonstrated. These maintenance warnings will inevitably accumulate as
the operating cycle progresses and could adversely affect the mean late operating cycle
performance more adversely than early operating cycle performance.
7 Naval Sea Systems Command, SEA 915, Age Reliability Analysis Prototype Study, Octobere 21, 1993 prepared by American
Management Systems, Inc pursuant to Contract N00024-92-C-4160
The initial start up period after refueling is seen as another plausible weak performance
period. As the majority of significant system maintenance is conducted during the refueling
outage and each maintenance action caries with it a measured risk of failure on use, this initial
start up period may be a higher risk period of operation. During the shut down period many of
the systems are opened and the state of their contents changed significantly. In example, a steam
system which normally operates at high temperatures and pressures with chemistry controls
could be stagnant with more challenged chemistry controls during a shutdown period. The
increased build up of corrosion products could propagate to many other systems after start up.
Shutdown maintenance periods are additionally very congested events with respect to the high
level of work done in a compressed time frame and space. It is more likely that personnel error
will damage nearby systems during these outages and that this damage could result in degraded
plant performance on initial start up.
Mid-cycle could also have degraded performance as the plant managers are more likely
to shutdown the plant to fix a growing list of systems with maintenance warnings. Plant
managers may be inclined to look for an optimal window for an unscheduled mid cycle
maintenance outage as there is still a long period to operate and the problem list is growing. If a
heavy power usage period is on the horizon a staff may decide to shutdown before hand during
a low power demand period and repair all of the systems which have maintenance flags. This
same decision process is less likely near the end of cycle when there is a limited time until the
scheduled major maintenance period commences. Of interest while a plant staff may try to set
themselves up for the high demand period by maintaining systems, they may actually degrade
performance by creating a new high failure period following the maintenance. Recall that the
last section demonstrated an equal probability of failure from maintenance as from component
wear out.
Transient periods may additionally cause higher failure rates. It is plausible that the plant
dynamics of starting up, shutting down or significantly changing power may impact operational
reliability. Many systems are put on line or taken off line during these periods or have their
parameters significantly altered.
5.1 Analysis design
Three basic analysis methods were constructed to observe these potential operating cycle
characteristics. The first study is an analysis of forced outage events as a function of the time in a
plant's operating cycle at which they occur. This study only recognizes forced outage event
records and is therefore a reflection of plant capability and is very coarse in its time sensitivity.
The second study is an analysis of the daily power generation records as a function of the time in
a plants operating cycle. This study has the benefit of increased time sensitivity as there are 19
reports per day and it captures all plant events. It could not be used to solely analyze equipment
reliability because operationally necessary reductions in power are indistinguishable from system
failures. The second study is used to validate the first studies results and to observe the impact
of variables other than equipment reliability on plant performance. The third study observes the
connection of forced outages to other operating cycle events. This study attempts to capture
the observed impact of events such as shutting the plant down for a mid-cycle outage.
5.2 Forced Outage Operating Cycle Analysis
The BWR/4 and PECO forced outage data discussed earlier were analyzed as a function
of the time in the individual plant operating cycle when the forced outage occurred. Only
complete operating cycle periods were analyzed to ensure that all portions of a cycle were equally
weighted. From 19 BWR/4 plants, 40 complete operating cycles were obtained. Two of these
operating cycles had long technical requirement shutdowns and were therefore excluded from
the study as they were not representative of cyclic plant performance. Had they not been
excluded a method would have been necessary to incorporate the excessively long remaining
cycle with the other 'normal' cycles. The two excluded cycles are shown in Figure 18. The
maximum outage limitation of thirty days from earlier studies was removed. The constraints that
the cycle be complete, not be involved in a lengthy administrative shutdown and have a normal
length removed the excessive shutdowns from the study. The worst of the cycles included in the
study are provided in Figure 19.
Unit 298, Cooper Station, Op period 811/93-10114/95 (805 d)
10004,
80
60
20906
0
0
100
200
300
400
500
600
700
800
Unit 324, Brunswick 2, Op period 1/7/92- 3/26/94 (810 d)
100%
80%
...........
.........
..
.............
60%
40%
20%
OOA
0
100
200
300
400
500
600
700
800
X-axis - Day in operating cycle following start up from refueling
Figure 18. Excluded operating cycles
Two constructions of the data were created. In the first forward looking construction,
all of the outages were analyzed by the day in the operating cycle that the outage occurred. To
understand this method, it is helpful to imagine stacking all of the operating cycles on top of
each other with the same operating cycle day x-axis. This analysis was the most revealing of early
cycle failure attributes and failures directly associated with days in service. It is less useful for the
analysis of end of cycle performance as shorter cycles are continually dropping off line. If the
mean cycle performance was improving at the 700 day mark it cannot be determined whether
that is the effect of shorter poorly performing cycles dropping out or longer cycles improving in
performance. The second backward looking construction was therefore made to better
understand how the mean cycle is completed. All cycles were stacked up on the same cycle end
date and then backed up to the cycle start. In this view plants will start at different points and all
end at the same point. If the mean operating cycle has difficulty completing this would be
apparent as a downward trend at the common end point. By combining these two constructions
one should be able to understand how the average cycle is executed with regard to its engineered
length.
Unit 341, Fermi 2, Op period 11/7/92- 3/12/94 (490 d)
100%
80%
-~l
ilqIl
Il1 - -I
r -I --
II~
I
u
~
60%
40%
20%
0%
i. I
-
.
IF
i I
100
100%
I
200
Ii
300
M
I
400
500
600
Unit 387, Susquehanna 1, Op period 5/16/92- 9/25/93 (498 d)
80%
60%
M
i4 OR
40%
20%
100
200
400
500
600
500
600
Unit 366, Hatch 2, Op period 11/21/92- 3/15/94 (480 d)
100%
80%
ii 'r' ' "
- 'I'1-
60%
40%
20%
100
200
300
400
Figure 19. Worst operating cycles included in BWR/4 analysis
5.2. 1 Forwnvardlookingforced outage operating ycle analysis
A histogram bin was created for each 30 day period within the operating cycle and is
shown as Figure 20. The clear columns represent the number of cycles in the analysis for each 30
day period, as read on the right vertical axes. Note that all 38 plants are in cycle until shortly after
the first year when plants start to finish their operating cycles. The analysis was terminated at
700 days to ensure that there were still sufficient operating cycles remaining, six, to obtain a
meaningful average. The blue columns represent the average occurrences per month as read on
the left vertical scale. The red columns represent the average lost days per month as read on the
same scale. Red and blue sixth order fitted curves of the monthly data have been added to aid
visual trending of the data.
The mean frequency of forced outage occurrences is noted to be fairly constant with the
exception of the first operating cycle month which exhibits twice as many failures as the second
month or the mean frequency of occurrence (0.67 outages during the first month versus 0.29
during the second month and 0.23 mean outages per month for the entire cycle). Early cycle
40
35
30
25 =
20
15 o
10
5
0
LO
-
U)
LO
M-
LO
0)
I--r-
LO
I)
LI)
C
L)
CIO
LI)
L)
14"
1t
c
0)
L)
LO
L
L)
I(.0
Lo
-
CD
Day in operating cycle after refueling
Figure 20. BWR/4 Operating cycle analysis of forced outage data
failures are observed to only impact the first month of operation. There are 3.8 days lost in the
first month of the mean cycle as compared to 1.42 days during the second month of the mean
cycle or 1.33 days for the mean of all 30 day periods. While measurable this is a very small early
cycle failure probability. It is also observed that the mean outage length during the first month
of 5.75 days is consistent with the mean outage length for any month of 5.78 days. The period
from the second month to 390 days is observed to exhibit a very consistent rate of lost days per
month frequency of outage. From 390 days until 570 days it is observed that significant forced
outages are experienced. This is during the same period that many shorter cycles are completing.
The day 405 centered 30 day period increase in days lost is primarily the result of a main turbine
failure at 414 days of operation. The refueling outage was then started 78 days later. This failure
defined the end of the operating cycle as the plant was unrecoverable. The 435 day centered
period is dominated by another main turbine failure at day 422. This failure resulted in a 50 day
outage after which the plant did operate for 26 days before refueling. The 555 day centered
period is composed of five system failures after all of which the plants returned to power.
5.2.2 Comparison of PECO sites to B WR/4 operating yclepetformance
The BWR/4 forced outage operating cycle data was compared to individual PECO site
data to observe the effect that different operating teams can have on plant performance. Figure
21 portrays the days of forced outage experienced by LGS and PBAPS to the BWR/4 average,
65
5.
E
4-
3 -
BWR IV
mLGS
a PBAPS
2
0O-
N-
M
I
I
UO
C0
Day in operating cycle after refueling
Figure 21. Comparison of average days of forced outage by operating cycle
(0
fitted curves have again been added to aid in analysis. It is observed that LGS and PBAPS have
dramatically opposing operating cycle performance. The LGS data was the average of four
cycles and the PBAPS data was the average of five cycles. Close inspection of the individual
cycles revealed that the observed average performance was characteristic of each cycle at the
particular site. LGS repeatedly came out of the refueling experiencing little forced outage early
on but had difficulty with end of cycle performance. PBAPS had difficulty early on but
completed the cycle with very little forced outage.
To fully understand the unique site characteristics the frequency of forced outages were
compared in a similar manner as seen in Figure 22. It is observed that LGS and PBAPS have
nearly identical early cycle forced outage frequencies while PBAPS was observed to have much
greater early cycle days of forced outage in Figure 21. This means that the PBAPS early cycle
outages are of significantly greater length than LGS outages. PBAPS plant management noted
that minor early cycle failures were dramatically amplified by the failure of older nuclear
instrumentation that would not support a rapid restoration of power. To reduce the mean
shutdown time experienced, this older equipment has been replaced with more reliable
instrumentation which will allow a more rapid restoration of power.
Figures 21 and 22 both demonstrate worsening LGS reliability with cycle length in the
1.4
0 BWR IV
1.2 --o
E
H LGS
1 -
c 0.8
oC
0.6
,
0.4
a PBAPS
,
S0.2,.
015
75
135
195
255
315
375
435
495
555
615
675
Day in operating cycle after refueling
Figure 22. Comparison of occurrences of forced outages by time in operating cycle
later half of the mean cycle. While each site had excellent leadership, small differences in there
management philosophies were observed which may have contributed to this observed
difference. LGS had a very strict execution of their refueling plan. It was very difficult to add
new work to the refueling with a strong bias towards deferring repair of new failures until the
next scheduled outage. PBAPS management had a philosophy which opted more towards
immediate correction of newly discovered problems. PBAPS' tendency to promptly correct
discovered failures may have established a greater level of reliability and therefore ability to
support the complete operating cycle. It is possible that the stricter compliance of LGS with the
refueling maintenance plan resulted in a higher quality of maintenance that was conducted but
with more systems operating with marginal performance.
5.2.3 Backward lookingforced outage operatingcycle analysis
To observe the end of cycle effects Figure 23 was constructed which lays out the outage
events in 30 day bins from the end of cycle backwards. The number of plants in cycle are noted
to mirror the last representation with all 38 cycles being represented on the last day of their
: I
mFlllIllilIlIl
.
. .
. .
. .
. .
. .
. .
. ..
_B . .
.
I
n"i
40
1 i B i ;_
I
-- 30
4
-AveOcc/mos
4
0
-
E 3
-
25
25
i Ave Lost days/mos
Cycles in study
20
-
.a
1L
0,
0.
U
15> U
2
10
5
0 -
0
)
-
O
T-
L
L
U')
a)
LO
c)
UO
-
UO
T-
U)
L
UO
,)
UO
C
UO
U)
Days from end of operating cycle
Figure 23. BWR/4 operating cycle analysis of forced outage events back from end of cycle
operating cycle regardless of their individual length. It is observed that the period three months
prior to the end of cycle has a significant increase in average forced outage. This may partially be
a wall effect of the end of cycle. If a significant forced outage, main turbine failure, is
experienced three months prior to the end of the cycle a three month outage will be realized. If
the same failure is experienced one month prior to the end of cycle only one month will be
realized. Overall the plants do exhibit an observable end of cycle degradation in reliability
centered 100 days prior to the end of cycle. This effect is noted to be greater than the early cycle
failures. The ramp in average days lost early in the cycle reflects the infrequent nature of failures
during this good performance period. The number of plants in cycle is steadily decreasing
greater than 400 days before the end of cycle. The denominator used to calculate the average
days of forced outage is therefore decreasing causing the average days of outage to
proportionally increase.
5.2.4 Cumulative impact offorced outages
To better answer the question of whether observed performance supports operating
cycle extension the cumulative capability factor as derived from the BWR/4 30 day forced
outage data presented above was plotted in Figure 24. The blue points represent the mean
capability factor determined from the BWR/4 forced outage data for each operating cycle
month. This is a capability factor as it represents only forced outage events not the power
_
100%
95%
o
CU 90%
LL
-Z
a.
*
85%
80%
Monthly Capabilty Factor from forced outage data
Cumulative Capability Factor
order fit of Cumulative Capability Factor
-3rd
- - - - - 6th order fit of monthly capability factor
75%
0
100
200
300
400
500
Day in Operating Cycle after refueling
Figure 24. BWR/4 cumulative capability factor
600
700
generated. The dashed brown line is a sixth order fitted curve to the monthly capability data.
The effects of the early cycle reduced capability experienced during the first month of operation
and the reduced capability near day 500 can be observed. The red line is a connection of the
discrete monthly cumulative capability factor derived from the monthly capability points. The
cumulative function is noted to be self damping not quickly responding the periods of reduced
or improved performance. The solid black line is a third order fitted curve to the cumulative
capability factor.
It is observed that the cumulative capability is nearly constant at slightly over 9 5% after
300 days of operation. The weak period near day 500 is balanced by the stronger performing
months before and after. The observed cumulative capability is encouraging for operating cycle
extensions as a downward trend in operating cycle capability is not observed with increasing
length.
5.3 Daily Generation Operating Cycle Analysis
This second analysis was conducted by examining the mean daily generation reports
from each of the BWR/4 plants in a stacked operating cycle manner similar to the previous
forced outage operating cycle analysis. The results of this analysis can be used to verify the
previous results as an independent source of data was used. Additionally this analysis can be
used to observe the impact of operational events other than forced outages on cycle
performance. Lastly this data base offers much greater sensitivity as each plant has a data point
for each day in the operating cycle.
The average daily net electrical generation was obtained from the digital NRC data base
maintained by INEEL as discussed in section 2. The daily data was modified such that the
maximum power generated during a single cycle was established as 100% generation capacity.
This removed reporting inconsistencies between plants and allowed an even handed
comparison. The same cycles were analyzed as discussed above for the forced outage operating
cycle analysis. Figure 25 presents the mean generation capacity by day in operating cycle for the
38 cycles analyzed. The solid green line represents the plants in cycle for a given operating cycle
day. The blue points are the daily mean capacity of the all plants for the given operating cycle
day. The red line is a moving 21 day average of the daily mean capacity. The first seven days of
mean capacity produce a linear power ramp up. This is attributed to the careful process of power
accession and required testing following the refueling outage. The moving data average
therefore did not include these points as they would bias the random process following day
seven. The red line prior to day seven is just the daily mean for the first seven days. The 21 day
average also expanded from the eighth day such there would not be any early cycle bias. Day
nine was a three day average, day 10 a five day average until a 21 day average was established by
day 18. The ending mean was tapered down in a similar manner. It can be observed from
Figure 25 that the moving average closely follows the daily means. The cyclic nature of the mean
capacity factor demonstrates the impact of individual forced outages. The solid black line is a
smoothed curve of the moving average from day seven on such that start up effects do not bias
the curve.
BWRs historically coastdown at the end of their operating cycle. During coastdown the
net electrical generation is limited by the declining reactivity of the fuel. Coastdown pulls the
mean daily operating cycle capacity down to 60% by day 700. It is additionally noted that plants
which have been fueled for shorter operating cycles are coasting down earlier. This causes the
large cyclic swings which parallel the cycles in study line. As coasting down plants come off line
the mean moves up to the remaining plants not yet coasting down. Prior to day 300 the mean
100%
40
_
35
S7
80%
25
.
60%
Daily average of 38 cycles
2
o
21 day average of daily averages
-
C
-
15
-
40%
6th order polynomial fit of daily averages, first
7 days excluded
0% -
0
5
I. ,. . .
100
200
n
0
Cycles in study
20%
2
.
CU20
300
400
500
.
.
..
600
Day in Operating Cycle
Figure 25. Daily power generation analysis by operating cycle day
700
capacity is noted to slowly build from day 7 to a gentle maximum at approximately day 150.
Figure 25 demonstrates little with regard to plant reliability late in the operating cycle due to the
dominant effects of fuel coast down. To remove this effect the coastdown period was modified
to reflect the generation capability as limited by fuel reactivity.
Figure 26 provides the coastdown modified mean operating cycle capacity. The daily
modified capacity is observed to be nearly flat over the entire operating cycle. The gradual
increase in daily capacity from 85% at day 7 to 93% at day 200 is observed as before with the raw
data. The end of cycle performance is noted to be similar to that from the forced outage records.
A broad dip is noted for 100 days on either side of day 500. The modified capacity drops from
93% to 89%. The last one hundred days of operating cycles analyzed demonstrate a strong
performance with the daily mean capacity increasing to 95%. The cumulative impact of the daily
modified capacity is illustrated in Figure 27. It is observed that the cumulative capacity slowly
increases from 85% at day 7 to 92% at day 300. The adjusted cumulative capacity is then
observed to be flat at 92% from day 300 on to the end of the analysis. The cumulative function
balances the bimodal nature of the daily mean capacity. The modified daily capacity which
accounts for all operating cycle events such as planned maintenance outages and on line
maintenance demonstrates strong end of operating cycle performance. This result supports the
100%
a
-
40
S35
80%
30
o
(
60%--
E
c
-- 25
25
-
21 day average of daily averages
-- 20 c
2
15b
40%
0
----
-0
"
Daily average of 38 cycles
20%
:
Cylces instudy
- 10
6th order polynomial fit of daily averages,
first 7 days excluded
0
0%
0
0
100
200
300
400
500
600
700
Day in Operating Cycle
Figure 26. Mean BWR/4 daily operating cycle coastdown modified capacity
O
100%
95%
,
CU 90%
LL
".
*,*"-
L
(U
--
. ,a
:
:
, Ave BWR Capability
80%
BWR IV"Cumulative
Capability
.
Cumulative BWR IV Capability (3rd order poly)
'Ave
BWR Capability (6th order poly)
75%
0
100
200
300
400
500
600
700
Day in operating cycle after refueling
Figure 27. Cumulative BWR/4 daily operating capacity modified for coastdown
exploration of longer operating cycles with regard to plant reliability.
5.4 Conditional relation of forced outage events upon operating cycle events
The goal of this analysis was to determine whether specific operating cycle events
demonstrated a measured relationship to forced outage events. If these operating cycle events
are detected than their minimization will have positive impacts on plant reliability. Every
reduction in generation capability noted at the PECO plants was analyzed to determine if the
event was the result of a refueling outage shutdown, an operational plant transient or a planned
maintenance outage. Table 6 summarizes the results. Limerick is observed to have a nearly
constant low risk of a follow on loss of generation capacity event as the result of any of the three
operating cycle events identified, refueling, transient or maintenance outage. Peach Bottom was
observed to have a greater risk of follow on events after a refueling or operational transient.
This may reflect the management policies toward the execution of the refueling shutdown and
Table 6. Conditional relation of PECO forced unavailability events
LGS
PBAPS
Occurrence Lost power Occurrence Lost power
Refueling Outage
9%
6%
14%
21%
Operational Transient
9%
8%
14%
13%
Maintenance Outage
2%
5%
2%
0%
power transients.
5.5 Operating cycle analysis summary
Two independent analysis methods and records were used to demonstrate that the
cumulative operating cycle performance is flat from the mid-cycle point on to the end of the
mean operating cycle. This supports age exploration toward longer operating cycles. Two
periods of lesser system reliability were noted during the first 150 days of the mean operating
cycle and approximately 75 days from the end of an operating cycle. These degraded equipment
reliability periods were of lesser impact on overall cycle performance than the cycle based
operational demands of initial start up following a refueling and the coast down prior to a
subsequent refueling. If the operating cycle were extended one could realize the reduced impact
of the start up and coast down limitations, and the identified fraction of load reduction
associated with refueling in Table 6. The uniqueness of plant performance was demonstrated
with a comparison of the two PECO BWR/4 sites. It is possible that sites could improve weaker
operating cycle performance areas by understanding the successful management strategies of a
site with opposing strengths.
6. Plant reliability improvement processes
This section presents the PECO Balance of Plant reliability improvement process that
was studied and supported as part of this cooperative research project. The genesis for this
research was PECO's Limerick Generating Station BOP-700 project so named to attempt to
improve the reliability of the Balance of Plant such that it could support a continuous 700 day
operating period. From the definition statement of this program, the site management
understood the relative weakness of the BOP systems and felt that they failed with increasing
probability as operating cycle length increased. From sections 3 and 5 these assumption are
noted to be correct for LGS. To improve BOP reliability each system which could effect plant
reliability was studied by a Focus on Improvement Team (FIT). The FIT process was
conceived to be a proactive process. Significant lessons learned were carried over from recent
reactive Tiger Teams created to correct repetitive system failures. The Tiger Teams had a much
more straight forward task to stop recurring failures of the same components. The shift from a
Tiger Team process to the FIT concept can be seen as an evolutionary process. When
components are failing on a frequent basis it is easy to identify the problem and solutions are
more focused. The existing data bases, NPRDS and IPEX are well suited to provided assistance
for correcting problems for supported components because they are cataloged by failing
component. As systems become more reliable, utilities must work harder to find the potential
failure points. This marks a turning point for the industry and may explain the apparent stalling
of the median capability factor noted in Figure 3 of the introduction.
FITs were made up of system managers from both PECO sites, previous system
managers and representatives from Design Engineering, Component Engineering, Operations,
Maintenance, Instrumentation & Controls (I&C), Health Physics and component vendors when
appropriate. An effort was made to optimize the team dynamics from previous Tiger Team
successes. Mentoring relationship were created such as placing a veteran component engineer
with a younger system manager. The teams were gathered for only a short time, such as three
days, to maintain an energetic group process atmosphere.
The teams reviewed all available records of system and component failures and identified
components with a proven negative impact on plant reliability. System dynamics were also
reviewed in a bottom up manner to find reliability choke points. The teams generated lists of
possible design changes, component improvements or changes to operating procedures. The
suggestions were prioritized by each team and handed over to the system manager for
accomplishment. Minor alterations that could be accomplished within the scope of a routine
maintenance period were scheduled for accomplishment. More significant changes requiring a
significant capital investment were submitted by the appropriate system manager in the form a
cost to benefits assessment.
Management is presently saddled with a difficult process of implementing these
proposed alterations. By the nature of the system performance, the probable failure of a given
component is very low. If the capital investment is high, it is hard to rationally justify the
expenditure. Note that none of these issues are directly safety related. It is difficult to measure
the level of confidence that recommended alterations will produce a predicted improvement in
reliability or savings of future costs. If management does not allocate significant resources for a
substantial period of time, it is doubtful that the low failure rates will substantially diminish. It is
recommended that utilities set aside a fixed amount of funds in their annual budget process
towards the continued improvement of plant reliability and that these funds be only altered by
long term plant performance.
The FITs brought together a wide range of experience and focused key personnel on the
issue of a systems reliability. In many cases one member of the team was newly assigned system
responsibilities and learned much of the systems characteristics from the other team members.
PECO expressed an interest in using the same process in the future as part of system manager
turn over. The process also focused more senior levels of management on potential future
problems instead of on current problems.
The capital assets of a system directly affected the function of the FIT. Large capital
systems such as the Main Turbine or the Turbine Generator were managed in great detail by the
vendor. The vendor issued recommended modifications and knew the status of each plants
component. The FIT process for these components degenerated to a review of the vendors
recommended alterations. Unfortunately capital pieces of equipment tended to have
extraordinarily large costs associated with improving their reliability. In example many
alterations for the main generator were proposed which involved purchasing a new rotor. In
contrast the less capital systems such as a support system in general had no external contact to
the plant. For these systems the FITs could brain storm and develop inexpensive alterations that
would significantly improve reliability. Examples of these alterations involved the addition of
relay contacts, the altering of control logic and minor alterations to equipment configuration
such as the replacement of a y-strainer in an essential system with a duplex strainer such that a
clogged filter could be cleaned without shutting down the plant. The process for both types of
systems were found to be beneficial if for no other reason than the transference of system
knowledge.
The large capital systems have excellent technical support systems maintained by the
vendors as mentioned above. The lesser systems need to have the same ability to utilize a larger
base of knowledge. Informal communication groups have started to develop to aid this process.
INPO has tried to foster their creation through system manager conferences. As many of these
groups are autonomous they must be able to exist with little administrative oversight. Efforts in
this area could have a significant affect on plant reliability and ease the duties of system managers
who must oversee systems that rarely fail.
7. Conclusions
The median US nuclear power plant capability factor has recently halted its long standing
upward trend. This is seen as the effect of recent plateaus experienced in efforts to improve
system performance and reduce refueling outage length. In comparing US performance to other
countries performance and small groupings within the US, it is evident that the overall capability
can be improved but only by broadening the community of operational experience.
System mangers need to have access to counterparts at other similar plants to share there
experiences. It is recommended that informal communication networks be established for this
need. Additionally system managers must be able to rapidly access a recorded knowledge base
of system failures. The required NRC reports would easily satisfy this need if better constraints
were placed on the fields all ready required for entry and if all utilities could database search the
record.
Regulatory shutdowns were found to be the dominant source of lost capacity and
warrant further study such that their impact could be minimized. With regard to system
reliability, Balance of Plant systems were found to initiate the vast majority of periods of reduced
generation capability and account for the majority of lost generation in comparison to reactor or
safety related systems. Additionally, Balance of Plant failures were found to have outages of
shorter average length than reactor system failures. All system failures were noted to be
infrequent occurrences as a result of good overall plant performance. The mean system was
found to have a 0.23 probability of failure in five years. It is therefore very difficult to maintain an
adequate understanding of potential system failures unless broader knowledge bases are utilized.
Balance of plant system failures were found to be dominated by design problems
indicative of the historical lower requirements for these systems to sustain a single point failure.
Many design problems with these systems were noted to be easily corrected such as controller
and sensor instabilities. Reactor systems failures were predominantly operator related owing to
the complex nature of the systems. Age related component failures were noted to only account
for 20% of the lost capacity. This was equal in magnitude to the percentage of forced outage
attributed to maintenance failures. This warns against adding on maintenance requirements to
solve age related failures. Instead maintenance requirements should be optimized to lower both
failure causes. The overall causes of component failure were found to be easily impacted by
engineering solutions.
The root cause analysis conducted of the PECO failures indicated that roughly half of
the lost capacity could be attributed to human failures. Craftsmanship issues were found to be
minor contributors while operator error and procedural adequacy issues were found to be
significant. The greatest use of the root cause analysis is for the utility under study to examine
where emphasis should be placed to improve the overall process at that site.
Several analyses were conducted to observe the performance of the mean BWR/4
operating cycle. Minor periods of reduced capability were noted during the first month and
approximately 75 days from the end of cycle. These periods of reduced performance were
adequately balanced by the remainder of the operating cycle such that the cumulative operating
cycle capability and capacity were constant after 300 days of operation. The effects of start up
and fuel coast down were found to be of much greater impact on cycle performance. In
addition a small percentage of forced outage events could be traced to the refueling outage.
Extending the operating cycle can not be seen from this analysis to cause a negative reliability
impact but will reduce the impact of all operating cycle dependent effects (start up, coast down).
The observed system improvement process at PECO was found to be highly beneficial
in terms of raising the level of awareness and transferring knowledge. It was more difficult to
directly impact the systems through modification. It is difficult to justify the expenditure of
significant assets towards accomplishment of costly improvements to systems designed to lower
reliability standards.
In answer to the fundamental questions of this research no impediments are seen
towards extending operating cycle length. On the contrary it appears that overall performance
will improve as the impact of cycle dependent loss will be reduced. For utilities to improve their
performance they must broaden their knowledge base and continue to improve system design
and operation.
27 May 1998
1992 - 1996
Appendix 1. LGS 5 yr unavailability data
Unit
1
Date
111/92
Type
Load drop
MWHr
2,904
Eff Out
Days
0.11
System
Main turbine
Component
Pressure
instrument RV
1
1/21/92
Load drop
7,030
0.27
FW heating
1
7/13/92
Load drop
8,566
0.32
Failure
Cause
Category
Issue #
Remarks
Class
RV closed
Mispositioned
HF/OA
92-01-02
SU, DF, TT
O
Logic
High level
sensed
Spurious
HF/MS
92-01-10
No PM
M
FW
RFP turbine
Insulation fire
Oil soaked
HF/PI
92-07-10
OM
O
Recirc
RX level signal
Low level
sensed
Channel noise
EF/WD
92-10-21
RRB
D
Pressure
switch
Low pressure
Spurious
EF/WD
93-02-16
RRB
D
1
10/16/92
Load drop
990
0.04
1
2/7/93
Load drop
3,510
0.13
FW
1
5/1/93
Load drop
7,313
0.28
Condensate
Motor
Bearing
Misaligned
EF/WD
Realigned
D
1
6/19/93
Load drop
11,050
0.42
Condensate
Motor
Bearing
Misaligned
EF/WD
Realigned
D
1
8/23/93
Load drop
725
0.03
Instrument Air Dryer package
Gasket
Failed
EF
M
Realigned
D
D
1
8/28/93
Load drop
734
0.03
Condensate
Pump
Bearing
Misaligned
EF/WD
1
9/7/93
Scram
146,784
5.56
Electrical
Breaker
Failed to
reclose
Spurious
EF/WD
1
12/1/93
Load drop
930
0.04
FW
Pump
1
12/1/93
Load drop
4020
0.15
Condenser
Waterbox
Cleaning
EF/WP
A
A
1
12/1/93
Load drop
3400
0.13
Main turbine
TCV pressure
switch
Failed
EF/WD
D
1
1/14/94
Scram
59,230
2.24
SWC
Trip circuit
Short
Bulb installation
EF
1
1/17/94
Load drop
940
0.04
Main turbine
Moisture
separator
dump valve
Valve positioner
PM frequency
LTA
HF/MS
1
3/1/94
Load drop
10,800
0.41
FW
LCS
EF
1
7/9/94
Load drop
6,290
0.24
Condenser
Tubes
Cleaning
1
7/29/94
Load drop
570
0.02
FW
Trip lever
Actuated
1
10/7/94
Load drop
31,945
1.21
Offgas
After
Condenser
Cleaning
1
10/8/94
Load drop
3,829
0.15
FW
LCS
1
11/1/94
Load drop
1,200
0.05
Computer
P1 program
Forced Unavailability Data
10000021
Would not run
Page 65
Inadvertent
10001318
OM
M
SU, TT, PM
deferred
M
EF/WD
D
EF/WP
A
HF/OA
O
EF/WP
A
EF/WD
D
D
EF
LGS POC VT ANGUS
27 May 1998
1992-1996
Appendix 1. LGS 5 yr unavailability data
Unit
1
Date
11/1/94
Type
Load drop
MWHr
900
Eff Out
Days
0.03
System
Reactor
Component
Fuel
Failure
Leak
Cause
1
1/2/95
Load drop
18,600
0.70
Recirc
Seal
Leak
Age
EF
1
1/30/95
Load drop
7,000
0.27
Recirc
MG set
Trip
Bumped
HF/OA
1
2/21/95
Scram
41,353
1.57
Transmission
system
Breaker /
relays
Failed to
actuate
PM LTA
HF
1
4/24/95
Load drop
2,230
0.08
FW
Pressure
switch
Failed low
1
5/7/95
Shut down
31,060
1.18
Recirc
Seal
Leak
Age
EF
1
7/19/95
Load drop
700
0.03
Recirc
MG set
Perturbations
Operator
HF/OA
1
7/20/95
Load drop
23,000
0.87
Condenser
Water box
Cleaning
1
8/8/95
Load drop
3,740
0.14
FW
Pressure
switch
Premature
actuation
1
8/20/95
Shut down
342,276
12.97
Reactor
Fuel
Leak
1
8/28/95
Shut down
121,440
4.60
Drywell
Flange
Misaligned
Poor Corr Maint
HF/C
35295006
M
1
9/2/95
Load drop
39,426
1.49
H2
Recombiner
Recorder
Logic
MOD PMT LTA
HF/PI
10004403
O
1
9/11/95
Scram
369,991
14.01
Main steam
SRV
Opened
Pilot seat
erosion
HF/MS
10004442
O
1
3/24/96
Shut down
203,260
7.30
Main steam
SRV
Leak
1
3/31/96
Load drop
47,222
1.70
EHC
Speed control
logic
Speed control
LVG
Sporadic
anomaly
EF/WD
1
4/11/96
Load drop
370
0.01
FW heating
Drain cooler
Tube leak
Poor Design
EF/WD
1
5/21/96
Scram
75,095
2.70
RPS
Logic
No 1/2 scram
alarm
Indeterminate
HF/PI
10005675
1
6/17/96
Load drop
22,121
0.79
Electrical
Improper
installation
HF/C
10005797
Forced Unavailability Data
Output breaker Low pressure
rupture
Page 66
Category
Issue #
Remarks
Class
F
Resealed
A
EF
O
10003606
Dual unit
scram
D
EF/WD
Maintenanc
e outage
1M03
EF/WD
A
O
EF/WP
Out of
calibration
M
10004299
EF
EF/WD
10005451
Normal
Maintenanc
e
A
RRB
D
Maintenanc
e outage
F
1E07
D
Infantile
failure
D
D
O
TT
LGS POC VT ANGUS
27 May 1998
1992 - 1996
Appendix 1. LGS 5 yr unavailability data
Eff Out
Remarks
Class
Unit
1
Date
7/15/96
Type
Manual
scram
MWHr
182,179
Days
6.54
System
Air removal
Component
SJAE
Y-strainer
Failure
Steam leak
Cause
Improper
installation
Category
HF/C
Issue #
10005889
1
7/25/96
Scram
93,663
3.36
EHC
FN card
Failure
infantile failure
EF/WP
10005909
1
8/2/96
Load drop
1,468
0.05
FW heating
Drain valve
Malfunction
1
12/23/96
Load drop
5,427
0.19
SWC
Temperature
CV
Loose adjust
arm screws
Manufacturing
HF/C
2
1/11/92
Load drop
3,430
0.13
FW
Turbine
controller
Malfunction
Relay
EF/WD
D
2
2/10/92
Load drop
FW
Pump
Trip
EF/WD
D
2
4/1/92
Load drop
1,039
0.04
Main turbine
Vac switch
Set point drift
EF/WP
Condensate
Water box
Leaks
EF/WP
I
A
DF
HF/PCM
M
F
10006438
2
4/1/92
Load drop
750
0.03
2
5/9/92
Load drop
36,960
1.40
Condensate
Pump
Bearing
Misaligned
EF/WD
2
6/2/92
Load drop
7,140
0.27
Condensate
Pump motor
Bearing
Misaligned
EF/WD
2
7/18/92
Load drop
1,353
0.05
FW heating
Vent line
Leaks
FAC
EF/WP
A
A
D
D
A
2
8/1/92
Load drop
10,750
0.41
FW heating
Vent line
Leaks
FAC
EF/WP
A
9/23/92
Load drop
58,625
2.22
Condenser
Tube
Leaks
EF/WP
A
EF/WP
A
M
2
2
10/1/92
Load drop
6,375
0.24
Condenser
Tube
Leaks
2
11/19/92
Load drop
44,101
1.67
EHC
#3 CV piping
Leak
2
12/4/92
Manual
scram
157,870
5.98
Recirc
Recirc pump
2
1/3/93
Scram
84,000
3.18
EHC
Relay
Hi pressure
2
3/17/93
Load drop
8,030
0.30
SWC
Valve
2
3/26/93
Scram
56,700
2.15
EHC
2
4/7/93
Load drop
6,000
0.23
2
4/19/93
Load drop
2,000
2
5/1/93
Load drop
6,410
Forced Unavailability Data
HF/C
92-11-20
RF,
10005615
HF/OA
92-12-01
Occurred
during ST
O
Sporadic
anomaly
EF/WD
93-01-01
RF
10004338
D
Valve
mispositioned
LTA procedure
HF/PI
93-03-24
SU, DF
O
#6 ISV
Perturbation in
ETS/RETS
Air entrap in
control pack
HF/PI
93-03-38
FW
FWLCS
'A' level down
spike
Spurious,
Indeterminant
EF/WD
93-04-04
RRB
D
0.08
FW
FWLCS
Master level
controller down
spike
Spurious,
Indeterminant
EF/WD
93-04-18
RRB
D
0.24
Condensate
Pump motor
Bearing
Misaligned
EF/WD
Realigned
D
Weld failure
EOC-RPT logic Breaker tripped
Page 67
O
LGS POC VT ANGUS
27 May 1998
1992 - 1996
Appendix 1. LGS 5 yr unavailability data
Eff Out
Unit
2
Date
5/16/93
Type
Load drop
MWHr
16,493
Days
0.62
System
EHC
Component
#2 TCV servo
Failure
Oil leak
Cause
Category
Issue #
Remarks
Indeterminant
EF
93-05-20
TT
2
6/26/93
Load drop
2,967
0.11
Condensate
Pump
Bearing
Misaligned
EF/WD
Realigned
2
6/28/93
Load drop
2,967
0.11
Condensate
Pump
Bearing
Misaligned
EF/WD
Realigned
2
6/29/93
Load drop
2,967
0.11
Condensate
Pump
Bearing
Misaligned
EFNVD
Realigned
EF/WP
2
10/1/93
Load drop
1,090
0.04
Circ water
Pump
Maintenance
2
2/16/94
Load drop
7,150
0.27
Recirc
Recorder
Mislabelled
LTA MOD
review/ PMT
2
2/24/94
Load drop
10,080
0.38
Recirc
Pump
Trip
Fuses pulled
HF/PI
RRB
O
HF/OA
O
D
EF
A
2
3/1/94
Load drop
1,504
0.06
Condensate
2
9/1/94
Load drop
28,290
1.07
Main steam
Valve
2
9/1/94
Load drop
6,500
0.25
Main turbine
#2 MSV
2
9/8/94
Load drop
28,650
1.09
SWC
Y-strainer
Clogged
2
9/8/94
Load drop
5,220
0.20
FW
Check valve
cap
Leak
2
10/19/94
Scram
77,754
2.95
Electrical
D24 Bus
De-energized
2
1/15/95
Load drop
48,000
1.82
Recirc
MG set
Generator
ground
2
2/21/95
Scram
86,672
3.11
Transmission
system
Breaker /
Relays
Failed to
actuate
PM LTA
HF
2
2/24/95
Load drop
28,000
1.01
Air removal
SJAE nozzle
Broken
2
4/29/95
Load drop
10,000
0.36
Recirc
Coupler
bypass valve
Incorrect
position
2
6/3/95
Load drop
17,150
0.62
EHC
#4 CIV
Leak
Age
EF
A
EF/WP
Generator
hydrogen leak
D
D
D
M
EFNVD
Pump
Leak
10001491
Class
M
EF/WD
10002830
TT
D
EF
A
Inadvertent
HF/OA
Indeterminant
EF/WD
O
D
10003606
EF
LTA
procedures
HF/PI
10003924
Dual unit
scram
M
DF
A
RRB
O
M
A
EF
2
6/17/95
Load drop
4,970
0.18
FW heating
Dump valves
Actuator
2
6/26/95
Load drop
14,115
0.51
Recirc
Temperature
switch
Spiked high
Spurious
EF/WD
10004172
RPT
D
2
6/28/95
Load drop
6,300
0.23
FW
FW UPS
Power
disconnect
switch off
Bumped during
cleaning
HF/OA
10004173
RRB
O
Forced Unavailability Data
Page 68
LGS POC VT ANGUS
27 May 1998
1992-1996
Appendix 1. LGS 5 yr unavailability data
Eff Out
Class
D
Cause
Category
Issue #
Remarks
Spurious
EF/WD
10004172
RPT
Loss of DC
power supply
(2K612)
Loose
connection
EF
10004298
Relay
High
impedance/ NC
contact
LTA design
EF/WD
10004338
H2
Recombiner
Recorder
Logic
MOD PMT LTA
HF/PI
10004403
O
0.01
FW heating
Valve
Leak
15,253
0.55
SWC
Filters
Clogged
38,939
1.40
EHC
#3 CV piping
Leak
Weld repair
failure
HF/PCM
10005615
A
D
M
Scram
89,099
3.20
Generator
Volts/Hz relay
Actuation
inappropriate
MOD package
LTA
HF/PI
10005652
6/1/96
Load drop
12,619
0.45
Condenser
Tube
Leaks
6/6/96
Load drop
378
0.01
Isophase bus
cooling
Fan
Trip
2
10/6/96
Load drop
20,859
0.75
SWC
Y-strainer
Clogged
2
10/12/96
Load drop
37,463
1.35
Condensate
Pump motor
Vibration
FME
HF/PI
10006201
2
12/6/96
Manual
scram
250,560
9.00
EHC
Pressure
switch
Leak
Broken bracket
I severed tubing
HF/LCA
10006385
Repeat
event 2M19
D
2
12/16/96
Shut down
139,200
5.00
Condenser
Expansion
joint
Leak
HF
10006422
2F20, DF
A
2
12/24/96
Manual
scram
55,680
2.00
Recirc
Scoop tube
ball joint
Broke
HF/LCA
10006441
Repeat
Event
M
Total
effective
outage
days:
135.74
Unit
2
Date
7/17/95
Type
Load drop
MWHr
5,600
Days
0.20
System
Recirc
Component
Temperature
switch
Failure
Spiked high
2
8/8/95
Scram
80,794
2.90
FW
FW LCS
2
8/20/95
Scram
69,396
2.49
EHC
2
9/2/95
Load drop
7,470
0.27
2
9/5/95
Load drop
370
2
11/22/95
Load drop
2
4/29/96
Load drop
2
5/14/96
2
2
Forced Unavailability Data
Page 69
A
RF, 93-0101
EF
EF/WND
RF 92-1120
O
A
EF/WP
LTA WO/
procedures
HF/PI
O
10005731
D
EF/WD
Vibration
Induced
D
O
LGS POC VT ANGUS
Unit
Categ
ory
Code
Date
MWHr
Type
I
System
Component
Failure
Category
WP - Worn
Parts
ICS Inadequate
Control
System
EOL - End of
PI - Procedural
Inadequacy
MS Management
Standards
CCraftsmanshi
p
LCA - Less
than adeq.
corr. actions
F - Fabrication
D - Design
Installation
Maintenance
OOperation
DF Dependent
Failure
TT - Turbine
A -Age
I-
Rema
rk
Code
Cause
WD - Weak
Design
EF Equipment
Fators
HFHuman
Factors
Class
Code
Days
27 May 1998
1992 - 1996
Appendix 1. LGS 5 yr unavailability data
RRB Recirc
Runback
Forced Unavailability Data
Trip
RPT - Recirc
pump trip
Page 70
Issue #
Remarks
Class
I
FPFatigued
Parts
OA Operator
Actions
PCM - Poor
Corrective
Maintenance
OM - On line
Maintenance
RF - Repeat
Failure
DFDependent
Failure
SU - During
Start up
LGS POC VT ANGUS
Eff Out
Days
8.9
System
Electrical
Component
3435 breaker
Type
Automatic
Scram
7/27/92 Manual
Shutdown
MWHr
248585
69540
2.5 Recirc
PUMPXX
61988
2
2
Automatic
Scram
9/5/92 load drop
12/13/92 load drop
6974
27598
2 2 Transmissio
n
Recirc
0.3
1 0 Reactor
lock out due to no
bkr
b recirc pump
Iprm 56-41 &5643 cross
connected
2
2
12/16/92 load drop
12/17/92 load drop
6549
10398
02
04
2
12/18/92
Manual
Shutdown
12/21/92 load drop
34334
oscillation
pressure
transimtiter
1 2 Generator stator
69152
2.5
EHC
Manual
Shutdown
load drop
load drop
418199
15.0
EHC
load drop
Automatic
Scram
load drop
4636
398510
Unit
2
2
2
2
Date
7/17/92
8/17/92
Recirc
EHC
2
1/1/93
2
2
1/18/93
1/22/93
2
2
1/29/93
3/2/93
2
3/18/93
2
2
2
3/20/93
4/23/93
4/24/93
load drop
load drop
Manual
Shutdown
5992
21119
128564
2
5/19/93
load drop
678
2
6/24/93
load drop
2120
01
HPCI
2
9/21/93
load drop
7552
03
Reactor
2
2
9/22/93
1/5/94
load drop
load drop
35885
5921
2
1/8/94
load drop
5813
Forced Unavailability Data
893
10171
22741
pressure
transimtiter
pressure
transmitter
2bs018 2b rfp
2cs018 2c rfp
FW
FW
FW
Class
A
Description
rwcu controls
Cause
lighting strike
Category
EF
Issue #
Equipment Failure
EF/WP
27792013
sub sta 205
RECIRC PUMP TRIP AND
VESSEL TEMPERATURE
DIFFERENTIAL
generator lock out
written com to load dispacter
HF/PI
N
O
blind controller
mtce
recirc pump b
Iprm mismatch
gain setting
hooked up wrong
HF/PCM
HF/C
N
SU
M
M
controller
did not work
worn parts
recirc pp controls
turbine control valve oscillations design
EFANP
EF/WD
SU
SU
A
D
h2 leaks
generator h2 leaks
sealent groove seal improper
HF/PI
SU
M
did not work
turbine control vie oscillations
design
EF/WD
SU
D
did not work
turbine control valve oscillations design
EFiWD
N
D
A
2b rfp vibration inspection
rfp c slow responce
unknown
failed
EF
EF/WP
SU
N
M
A
recirc pump atrip
condensate pump a tripped
failed
operator opened bkr
EF/WP
HF/OA
N
N
A
O
recirc pump b(gen hi amps,volts no mtce
EFV/WP
SU
M
recirc pump hi oil level
loss of cooling
recirc mg set
rx instrument mismatch &recirc loss of level
pp
loss of power supply to e22 bus loss of power
EF/ICS
EF/WP
EF/WP
SU
N
N
D
A
A
HF/C
N
M
check valve
broken air line
repair hpci injection check valve scaffold
HF/OA
N
0
fuel
clad
admin precaution increase in off pci
gas level
pci
Flux tilt testing
5a heater extraciton w solenoid end of life
EF/WD
SU
D
EF/WD
EF/EOL
N
N
D
A
HF/PCM
SU
M
1 3 Reactor fuel
0.2 FW Heating solenoid valve
02
Remarks
N
Failure
trpped
rotor imbalance
control valve
bearing seized
coupling trip
tach
0.2
Recirc
loss of pwr to
14.3 Condensate 2c cond pumps
transformer
loss of tach signal loss of contact for
Recirc
08
brushes
dead band
level switch
02
Recirc
vibrated shut
Recirc
vent damper
08
leak
It 73a equilizing
4
Recirc
valve
loss of fw htg
0 0 Electrical e322 trip
00
0.4
5/28/98
JULY 1992 - JUNE 1997
Appendix 2. PBAPS Lost Generation Data
2as018
clad
coil failed
speed controller
siezed
a rfp maintenance
Page 71
improper lubrication
PBAPS POC FL JORDAN
5/28/98
JULY 1992 - JUNE 1997
Appendix 2. PBAPS Lost Generation Data
Eff Out
Component
Failure
Days
System
" 0.1 FW Heating 5a fw htr extration gasket failure
stm valve
Remarks
N
Class
A
EF/WD
N
D
O-ring replaced
EF/WP
N
A
defective
EF/EOL
SU
A
low condenser vacuum
Iaand
IFlux tilt testing
ao-86a repair
chlorine oos an dwarm water
inst at cal imit
pci
small leak
HF/PCM
N
M
EF/WD
EF/WP
N
N
D
A
Description
5a fw htr repair from stm leak
oos
Type
load drop
MWHr
2116
2
2/23/94 1load drop
4577
02
2
4/28/94
load drop
2537
2
5/14/94
165366
N2 leak at charging mnopcontrol rod
block
Ivdt
recirc pump a speed increase
2
Automatic
Scram
6/24/94 load drop
0.1 Control Rod rod 26-15
Drive
5.9
Recirc
a pump
0 2 Circ Water condenser
cleanliness
2
2
9/10/94
1/12/95
load drop
load drop
0 5 Reactor fuel
0 1 Main Steam msiv
clad
packing leak
3/17/95
load drop
2
6/3/95
load drop
2
6/10/95
2
8/16/95
Unit
2
Date
2/10/94
4904
14083
1795
832 I
Reactor
fuel
clad
Cause
bearing cap gasket failure
rod pattern adj due to 5 leakerspci
Issue #
Category
EF/WP
O
00
FW
msc control switch misposition
feedwater transient
operator error
HF/OA
21995
08
FW
power supply fw
control system
failure of PS
power ascension & c rfp
problems
age
EF/EOL
N
A
load drop
835
00
Electrical
e22 bus
HF/PI
N
O
load drop
283
HF/OA
N
2
10/22/95 load drop
340
2
10/26/95 load drop
24869
0 0 Transmissio 220-8 line
n
0.0 Transmissio 220-34 ug line
n
0 9 FW Heating B5 fw heater
loss of pwr to panel loss of power supply to e22 bus Diesel feedback signal during
y-34
&y-34
mod testing
digging into line Unit 1 pl
popped open su 25 220-8 line fault
2
11/8/95
load drop
329
2
2
11/11/95
12/1/95
2
2/3/96
load drop
Manual
Shutdown
load drop
2
3/4/96 load drop
2
3/27/96
2
5/9/96
1
30423
219052
0.01 FW Heating positioner
11
FW
2ap001
7 9 FW Heating 5b fw htr
failed
EF/FP
N
A
unknown
EF/FP
N
A
positioner air
supply
n/a
leak
220-34 line tripped = positive
reactivity
b fw string isolated 95% pwr
limit
3c fw htr drain w broken air line
high level
a rfp vibration 95% pwr limit
5b fw htr repairs
improper support
EF/FP
N
D
foreign material
errosion
HF/C
EF/WP
SU
N
A
broken o-ring
hcu hv-111 broken
O-ring damaged/cut
EF/WP
N
A
probe bumped
HF/OA
N
O
low cal
HF/PI
N
M
vibration
EF/WD
cable fault
tube leaks
3488
0.0 Control Rod hv-111
Drve
01
FW
2bs018 b rfp
vibration
load drop
4763
0 2
cal drifted lo
load drop
2148
01
492
Generator generator
core
monitor
wire lug
EHC
loose wire
b rfp tripped vibration probe
bumped
S generator core monitor alarm
turbine control #2 valve
N
ncllitilnn
j
2
6/4/96
load drop
2
2
10/5/96
10/6/96
load drop
Automatic
Scram
I
Forced Unavailability Data
114
4t263
82012
0 0 Transmissio 220-8 line
n
breaker opened
probe failure
i-vW 40sUi D rip
2 9 Generator negative sequence short/open
relay
0U
220-8 line de energized
b rfp high vibration
scram, gen lock out stator
current unbalance
Page 72
operator at sub opened
incorrectly
worn parts
poor solder joint
_
_
_
I_
_
D
_
I_
_
HF/OA
N
O
EF/WP
EF
SU
SU
A
F
_
PBAPS POC FL JORDAN
JULY 1992 - JUNE 1997
Appendix 2. PBAPS Lost Generation Data
Unit
2
Date
10/9/96
Type
Manual
Shutdown
10/15/96 Automatic
Scram
12/25/96 load drop
MWHr
7
46288
94620
244
Eff Out
System
Component
Days
1.7 Generator bearing
4/1/97
load drop
17756
2
2
4/2/97
4/2/97
load drop
load drop
4271
1793
3
7/4/92
251351
3
3
3
Automatic
Scram
7/14/92 -Manual
Scram
7/23/92 load drop
7/26/92 load drop
8/22/92 load drop
3
3
9/30/92
10/1/92
3
10/15/92 Automatic
Scram
361608
3
13908 0
0 5 Condenser
3
3
12/19/92 Manual
Shutdosn
1/23/93 load drop
3/7/93 load drop
3
3/7/93
1672981
6.0
Reactor
Main
Turbine
FW
3
5/9/93
5/21/93
622
18117
0.0
3
7/4/93
352800
12.7
3
3
3
3
load drop
load drop
Automatic
Scram
load drop
load drop
Manual
Scram
7/30/93 Manual
Scram
9/14/93 load drop
11/19/93 Iloaddrop
Forced Unavailability Data
225039
25191
9791
2385
2171
78134
0.21
0.1
FW
EHC
2as018 a rfp
cooler
9.0 Transmissio 3su feed lost
n
81
Offgas
3239a linkage
failed
0.9
Reclrc
st-r-60a-2
0.4 Reactor detector
IW Heating drain valve
positioner
01
Recirc
3b recirc pump
Recirc
control loop
2.8
Class
M
scram, gen lock out stator
unbalance
4c fw htr level oscillations
poor solder joint
EFIWD
SU
M
electrical comp end of life
EF/EOL
N
A
ehc fluid leak
suspect FME root cause TBD
HF/C
N
M
failed to trnp
restriction
reactor feed pump trouble_
ehc fluid leak
debris in trip dump valve
suspect FME root cause TBD
HF/PI
HF/C
N
N
M
M
e313 cs &343 su
tran
linkage allen set
slippage
calc error
failed
leak
north substation xfmr
173 mtce
HF/PCM
N
M
off gas system
design of air line
EFWD
SU
D
recirc pump
tip machine a
feed water heater
margin
age
out of calibration
HF/PI
EF/EOL
HF/PCM
SU
SU
M
A
M
bkr 2ak34b trip
none mg set lock
up
recirc pump
recirp pump control
unknown
unknown
EF
EF
N
N
A
D
PCIS GROUP I ISOLATION
CAUSED BY BUMPING
INSTRUMENTATION.
Equipment Failure
HF/OA
13 0 Containment VALVEX
HTEXCH
Remarks
SU
Category
HF/C
Description
turbine bearing 12 high temp
Cause
Issue#
electroysis
Failure
hi temp
3 4 Generator negative sequence short/open
relay
0 0 FW Heating level controller
dump valve failed
to open
restriction
0.6
EHC
cooler
2
5/28/98
27892008
O
A
fuel
mts
CLEAN CONDENSER
WATERBOXES
flux tilt
leak
electrical problems mtsv replacement
pcil
failure
EF/WD
EF/EOL
c rfp
hi vibration
poor lubrication to vib sensor
HF/PC M
N
M
Reactor
FW
detector
fp turbine control
cable
mtce left out parts
EF/FP
HF/PC M
N
N
A
M
Reactor
fuel
lost signal
#2 tip machine
parts missing post rework mgu hyd jack solenoid
mtce
sv7
leak
power reduction for fuel repair
pci
EF/NV
N
D
47859
Offgas
stm flow sensor
operator action
HF/OA
FW
Recirc
flow controller
coupling
fm
746
blown fuse
mo99/91
debris in
misaligned
HF/PCM
HF/PCM
N/RRB
SU
M
M
maintenance outage
manual scram due to
recombiner
reactor feed pump trip
recirc pump vibration alarm
Page 73
EF/WP
PBAPS POC FL JORDAN
Appendix 2. PBAPS Lost Generation Data
Unit
Date
11/28/93
3
12/1/93
3
Type
load drop
Manual
Shutdown
12/29/93 iload drop
2/3/94
Manual
MWHr
Eff Out
Days
JULY 1992 - JUNE 1997
Category
Issue #
Remarks
System
Component
Failure
Description
Cause
_____________
_____________
__________I____________________
_______________________________________________________
_________________________________ _____________ ______________....__________
Control Rod 34-31 accumulator failed
EF/WP
o ring failure, rod 3431
accumulator
Drive
o-ring
LPCI
392253
Electrical
102297
37
5/28/98
mo-3-10-025 rhr
bent shaft
3-2a-k004a
Generator breaker
Scram
wrong nut
HF/PI
deenergized when recirc runback a pump
480v load center
30801 was being
restored from the 3
4G4 tie breaker
30b01 Ic being restored
HF/PI
ground resistor left
in place
control valve
actuator binding
control system
stopped working
brush pigtail
shorted inner and
outer collector
field ground resistor-main
generator
rfp control problem
mtce
HF/PCM
not smooth operati on
EFWD
a rfp maintenance
ehc elec cabinet cool fan
b recirc pump brush
replacement
EF/WP
not smooth operati on
unknown
EF/WD
mtce did not stand up leads post HFIC
mtce
Ipci mov 25a inop
Class
M
_________
N/RRB
O
N
M
N/RPT
M
4/27/94
load drop
12881
05
FW
3
5/1/94
load drop
6/12/94
load drop
load drop
FW
EHC
Recirc
rfp
muffin fan
8/6/94
7634
4054
11285
0.3
3
8/8/94
load drop
35702
1.3
Offgas
fe5020
steam leak at flex
recombiner leak troubleshooting flange flex
EFWD
SU
D
99423
3.6
Electrical
main power
transformer for
inverter
shorted winding
scram loss of static inverter y50 break down of insulation
EF/WD
NIRRB
D
wrong pin
HF/PI
N
O
plug design
EF/WD
N
D
3
10/11/94 Automa tic
Scram
0.1
irfp
3ag004
3
1/20/95
loadro p
131941
0.5 Circ Water screens immobile
pin shear
3
3/23/95
93042
3.3
Offgas
ao3466b
failed close
3
8/1/95
29188
1.0
FW
scram-feedwater transient
failed transmitter on card
EF/EOL
NfTT
A
8/26/95
65874
24
3a rfp
speed
controller
limit switch
upscale
3
Manual
Scram
Automatic
Scram
load drop
b screen immobile, b cw pp
removed from service
sjae supply block valve failed
switch failed
environmental conditions
EF/WD
N
D
3
3
9/19/95
11/6/95
load drop
load drop
1946
0.1
01
pm task inadequate
droped holder damage
HF/PI
HF/OA
N/RPT
N
M
O
3
11/17/95 load drop
turbine cv limit switch bad
testing logic
a recirc mg set tripped
13 kv electrical system-loose
fuse
3c fw heater drain closed
wear
EFWP
N
A
3
12/2/95
Automatic
Scram
114994
2nd ground
HF/OA
N
O
3
2/2/96
Manual
Shutdown
61863
worn out
EF/EOL
N
A
Forced Unavailability Data
3336
Main
Turbine
Recirc
Electrical
3-2a-k010a
0 0 FW Heating cv-3043c
cracked terminal
strip
2 21 Generator bushing
loose connection
loose
steam seal
turbine trip - pos & neg ground
2nd ground by
person working on
equipment
main generator hydrogen leak
gasket leak
Page 74
PBAPS POC FL JORDAN
JULY 1992 - JUNE 1997
Appendix 2. PBAPS Lost Generation Data
Unit
3
Date
2/5/96
Type
load drop
MWHr
953
Eff Out
Days
0.0
3
3/27/96
load drop
4441
02
3
6/22/96
load drop
20785
0.7
EHC
servo
leak
3
6/23/96
137576
4.9
stem binding
8/6/96
577
0.0
Main
Turbine
Offgas
valve
3
Manual
Scram
load drop
3
3/8/97
Manual
Scram
150887
5.4
Recirc
3ap034-dr
loss of oil in
a recirc motor low oil level
upper/lower mtr brg
reservoir
3
4/9/97
load drop
27554
10
Recirc
low side
transformer cable
cable insulation
breakdown
3
4/21/97
load drop
35167
1.3
FW
computer dcc-x
power supply
fail/transfer of
control
3
6/13/97
load drop
4238
02
FW
3b rfp speed
controller &hjsv
sol coil burned and 3b fpr speed control problem
controller
degradded
Forced Unavailability Data
System
RPS
Component
pish-3-02-3-055c
Generator alarm setpoint
Failure
defetive logic
switch
drifted low
control valve 9716b failed open
5/28/98
Description
5a k5c relay dropped out
failed
Category
EF/EOL
generator core monitor alarm
cal
repair #4 cv ehc leak and msv
leak
#2 turbine control vv stem
seperated
recombiner isolation
Cause
Remarks
SU
Class
A
HF/PI
N
M
o-ring failure
HF/C
N
M
clearance inadequate
EF
SU
F
EF/WD
N
D
unknown under investigation
HF
SU
M
3b recirc pump trip c phase
cable fault
cable treeing
EF/WP
N/RPT
A
feedwater computer trouble
relay actuation
power supply? transfer - design EF/WD
issue
N/RRB
D
N
A
Page 75
age
EF/EOL
Issue #
PBAPS POC FL JORDAN
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
UNIT
ID
259
OUTG
DATE
1/1/921
OUTG
HRS
43,800
OUTG
OUTG
TYPE
DAYS
S
1,825 0
OUTG
OUTG OUTG
METH REASN SYSTEM
F
4
OUTG
COMP
DESCRIP
ADMINISTRATIVE HOLD TO RESOLVE VARIOUS TVA AND
OUTG LER
System
Operation
Category
HF/NRC
Main Turbine
EF
FW
EF/ICS
RPS
EF
Drywell
Instrument Air
EF
HF/ICS
NRC CONCERNS
260
2/23/92
203
8.5
S
1
B
260
4/27/92
41
17
F
3
A
260
7/28/92
33
14
F
3
A
260
260
9/25/92
4/15/94
102
146
4.3
6 1
F
F
1
3
B
B
UNIT SHUTDOWN TO IDENTIFY AND REPAIR LEAKAGE IN
THE DRYWELL, AND TO REBALANCE THE Main Turbine
GENERATOR
AUTOMATIC FEEDWATER LEVEL CONTROLLER FAILED
REACTOR SCRAMMED ON LOW WATER LEVEL
SCRAM DUE TO A SPURIOUS HIGH WATER LEVEL TRIP,
CAUSED BY A FALSE SIGNAL FROM A NEW ELECTRICAL
SWITCH.
UNIT SHUTDOWN TO REPAIR DRYWELL LEAK
DURING PLANNED MAINTENANCE ACTIVITIES ON THE
SCRAM PILOT AIR HEADER, UNIT 2 AUTOMATICALLY
26094004
SCRAMMED ON LOW SCRAM AIR HEADER PRESSURE
FOLLOWING ISOLATION OF BOTH PRIMARY AND
SECONDARY SCRAM PILOT AIR HEADER PRESSURE
REGULATORS
260
12/2/94
20
08
F
3
A
JJ
TIS
AUTOMATIC SCRAM CAUSED BY BALANCE OF PLANT
EF
26094013
EQUIPMENT FAILURE
Generator
EF
26095004
AUTOMATIC SCRAM CAUSED BY PERSONNEL ERROR
DURING SURVEILLANCE TESTING
26095007
Main Turbine TRIPPED ON LOW CONDENSER VACUUM
CAUSED BY A FAILED POWER SUPPLY TO BOTH LEVEL
CONTROL LOOPS FOR THE OFF GAS CONDENSER DRAIN
VALVES REPLACED FAILED ELECTROYTIC CAPACITOR
IN THE POWER SUPPLY FOR THE OFFGAS CONDENSER
DRAIN VALVES
RPS
HF
Condenser
EF/WP
260/9605
FW
EF/WD
Generator
EF
26095002
260
2/9/95
30
13
F
3
H
AUTOMATIC SCRAM CAUSED BY Main Turbine
260
3/30/95
69
29
F
3
H
260
8/19/95
32
13
F
3
A
WF
JX
260
5/10/96
100
4.2
F
3
H
JB
SK
REACTOR SCRAMMED AUTOMATICALLY ON MAY 10,
1996, DUE TO LOW REACTOR WATER LEVEL DUE TO
ZERO DEMAND SIGNAL THAT RESULTED FROM
REINITALIZATION OF THE REACTOR FEED PUMP
FEEDWATER LEVEL CONTROL SYSTEM ROOT CAUSE
WAS INADEQUATE DESIGN
260
10/29/96
121
50
F
3
A
TL
EXC
THE UNIT 2 MAIN GENERATOR FIELD COLLAPSED DUE TO 26096007
AN EXCITER MALFUNCTION, AND THE RESULTANT
VOLTAGE AND CURRENT CONDITION CAUSED THE
GENERATOR BACKUP RELAYS TO OPERATE
Source - INEEL / NRC MORP2
GENERATOR EXCITER GROUND RELAY TRIPPING
Page 76
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
_,_
,___,_
OUTG
DATE
i/1~I
9711
15/921
_271' -- t/4/6/93
271
12/6/93
271
12/9/93
271
12/17/93
OUTG
DAYS
OUTG
HRS
UNIT
ID
131
131
OUTG
TYPE
0.1
F
0.61
119
OUTG
OUTG OUTG
OUTG
COMP
METH REASN SYSTEM
1
B
ZZ
XXXXXX
I l
B
CONROD
S
1
B
HA
VALVEX
F
1
A
HC
HTEXCH
S
1
A
HC
PIPEXX
EQUALIZING LINE.
2/9/94
4ELECON
F
4/10/94
F
2711
10/1594
48l
271
12/8/95
69
271
11/1/96
271
11/3/96
i
23
54
10/4/94
271
20
S
F
1
1
A
HA
B
EB
DURING A ROUTINE INSPECTION, DISCOVERED NEUTR/
GROUND ON THE MAIN GENERATOR DISCONNECTED
GENERATOR WAS TAKEN OFF LINE TO MAKE THE
CONNECTION.
VALVEX
"C"MOISTURE SEPARATOR HIGH LEVEL Main Turbine
TRIPPED AND A REACTOR SCRAM. REPLACED A FAULT
LEVEL CONTROLLER
VITAL AC AUTO BUS TRANSFER PROBLEM. REPAIRS
MADE TO VOLTAGE REGULATOR
COMBINATION OF SERVICE WATER LEAK ON THE HEAT
EXCHANGER AND "B" RBCCW BYPASS VALVE STUCK
OPEN LINE ISOLATED, BLANKED OFF
TURBINE TRIP/REACTOR SCRAM DUE TO
MALFUNCTIONING FEEDWATER REGULATOR VALVE
TURBINE TRIP DUE TO "A" MOISTURE SEPARATOR HIGH
LEVEL SIGNAL
TURBINE TRIP DUE TO LOSS OF CONDENSER VACUUM;
CAUSED BY ATMOSPHERIC DRAIN TANK LEVEL CONTROL
SYSTEM PROBLEM
INSTRU
-------
B
WB
VALVEX
A
CH
VALVEX
F
i
i
Source - INEEL / NRC MORP2
INSTRU
Page 77
I-
.1_
ELECON
VALVEX
3
OUTG LER
DESCRIP
STEAM JET AIR EJECTOR RUPTURE DIAPHRAGM REPAIR
"
MANUALLY SHUTTING DOWN DUE TO A LEAK ON THE "B
PIPING.
HEADER
DISCHARGE
FEEDWATER
MANUAL SHUTDOWN TO REPAIR "A"MOISTURE
SEPARATOR EMERGENCY DRAIN VALVE
MANUAL SHUTDOWN TO REPLACE EXPANSION JOINT O
THE "A"MAIN CONDENSER DUE TO INCREASED AIR
INLEAKAGE
MANUAL SHUTDOWN TO REPAIR THE CONDENSER
System
Air Removal
FW
Category
EF
EF
Main Turbine
EF
Condenser
EF
{Condenser
EF
Generator
2794
27194004
HF/C
Mai Tubn
Main Turbine
Electrical
27194013
_
JEF
RBCCW
271 ~5021
27195021
EF
Main Turbine
EF
Condenser
EF
L
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
OUTG
OUTG
HRS , DAYS
DAYS
232
277
OUTG
DATE
DATE
3/27/92
277
4/7/92
61
25
277
5/20/92
184
77
7/17/92
219
91
277
7/27/92
60
25
277
8/17/92
47
277
2771
12/18/92
- 1/2/93
3/2/93
4/24/93
277
8/11/93
_____I
..
..
OUTG
TYPE
F
F
OUTG
OUTG OUTG
METH ,REASN SYSTEM
A
1
F
A
i
HH
OUTG
COMP
INSTR
i
PIPEXX
RECOMB
VALVEX
A
XX
XXXXXX
A
CB
PUMPXX
CKTBKR
GENERA
L
81
34
F
F
3
1
A
198
83
S
2
H
HH
ID
PUMPXX
INSTRU
INSTRU
DESCRIP
UNIT SHUTDOWN DUE TO REACTOR WATER LEVEL
IMISMATCH
CONDENSATE VENT LINE FAILURE. STEAM LEAK ON
RECOMBINER FLOW TRANSMITTER.
TWO CIVS CLOSED SIMULTANEOUSLY CAUSING POWER
LOAD IMBALANCE.
LIGHTNING STRIKE - AUTO SCRAM INITIATED BY TCV
FAST CLOSURE ON LOAD IMBALANCE.
RECIRC PUMP TRIP AND VESSEL TEMPERATURE
DIFFERENTIAL
REACTOR SCRAM AS A RESULT OF PROBLEMS
ENCOUNTERED DURING THE BLOCKING OF BREAKERI
REPAIR GENERATOR HYDROGEN LEAK
MAINTENANCE OUTAGE TO REPAIR RECIRC PUMP SEAL.
SECOND CONDENSATE PUMP TRIP
PLANT SHUTDOWN DUE TO REACTOR LEVEL
INSTRUMENT MISMATCH.
MAINTENANCE OUTAGE FOR REACTOR WATER LEVEL
Category
EF
OUTG LER
27792005
System
Reactor
27792006
27792006
Condensate
Condensate
27792009
Main Turbine
EF
27792012
Transmission
Nature
27792013
Recirc
EF
27792015
Electrical
EF
EF
27793004
27793010
Generator
Recirc Condensate
Reactor
Reactor
|EF
1
MODIFICATION
277
5/14/941
277
10/6/96
121
5.01
277
10/8/96
48
20
277
10/15/96
66
2.7
Source - INEEL / NRC MORP2
F
3
F
3
F
F
1
3
PUMPXX
A
HA
RELAYX
A
HA
TURBIN
A
HA
RELAYX
APRM HI HI FLUX AUTOMATIC SCRAM DUE TO RECIRC
PUMP SPEED PROBLEMS.
MAIN GENERATOR NEGATIVE PHASE RELAY OPERATION
TURBINE BEARING (#12) HIGH TEMPERATURE
MAIN GENERATOR NEGATIVE PHASE RELAY OPERATION
Page 78
Reclrc
EF
Generator
EF
Turbine
Main
Main Turbine
Generator
EF
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
OUTG OUTG OUTG
OUTG
OUTG
OUTG
OUTG
DAYS
TYPE METH REASN SYSTEM
DATE
HRS
EA
911 I F i 3 ' A II
219 I
7/4/92 I
MB
A
7/14/92
F
81
194
SH
F
254
A
609
10/15/92
UNIT
IDSI
278
278
278
I
278
_________
______
______
1*
i
i
I 12/19/92 I
i
I
(
3/7/93
i
IHTEXCH 1
i
i
PUMPXX
FUELXX
7/4/93
278
INSTRU
7/30/93
I
OUTG
COMP
' RELAYX
HTEXCH
VALVEX
I
I
I
I -
278
278
2/3/941
10/11/94
81
91
3.4
3.8
F
F
2
3
A
A
HA
CH
VALVEX
GENERA
GENERA
278
3/23/95
84
3.5
F
2
A
HC
VALVEX
278
7/30/95
68
2.8
F
3
A
HC
VALVEX
278
278
12/2/95
2/2/96
99
50
4.1
2 1
F
F
3
1
H
A
HA
HA
TURBIN
GENERA
278
6/23/96
107
4 5
F
2
A
CD
VALVEX
12/2/93
Source - INEEL / NRC MORP2
DESCRIP
AUTO SCRAM - #1 TRANSFORMER FAILURE.
SJAE FLOW CONTROLLER FAILURE
PCIS GROUP I ISOLATION CAUSED BY BUMPING
INSTRUMENTATION
OUTG LER
27892010
27892005
27892008
WA1ER OXES
CLEAN CONDENSER WATERBOXES
REACTOR FEED PUMP TRIPPED, OTHER PUMP FAILED 1TO 27883002
START.
MAINTENANCE OUTAGE FOR REPLACEMENT OF
DEFECTIVE FUEL ASSEMBLIES REACTOR MANUALLY
SHUTDOWN TO 18% AND THEN SCRAMMED FROM
THERE
MANUAL SCRAM DUE TO RECOMBINER ISOLATION AND
SUBSEQUENT LOSS OF CONDENSER VACUUM.
LPCI MOTOR OPERATED VALVE MO-25A INOPERABLE
MAIN GENERATOR FIELD GROUND RESISTOR.
AUTOMATIC SCRAM/HIGH REACTOR WATER LEVEL DUE
TO FEED PUMP CONTROL PROBLEMS CAUSE BY LOSS
OF THE STATIC INVERTER
MANUAL SCRAM, LOSS OF VACUUM DUE TO STEAM
SUPPLY VALVE FAILURE TO AIR EJECTORS.
FEEDWATER TRANSIENT, HIGH REACTOR LEVEL SCRAM
AUTOMATIC SCRAM/TURBINE TRIP.
GENERATOR TAKEN OFF LINE FOR A MAIN GENERATOR
HYDROGEN LEAK.
REPAIR #2 TURBINE CONTROL VALVE STEM
Page 79
System
Electrical
Air Removal
Containment
Category
EF
EF
HF/OA
Condenser
FW
EF/WP
EF
Reactor
EF
H2 Recombiner
I
27894005
RHR
-EF
Generator
FW
EF
EF
Air Removal
EF
FW
EF
Main Turbine
Generator
EF
EF
Main Turbine
EF
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
~~~~
________
-~
OUTG
HRS
33,578
OUTG
DATE
1/1/92
296
296
11/25/95
11/26/95
296
11/27/95
48
20
S
2
B
296
2/29/96
66
27
F
3
A
296
4/21/96
296
5/1/96
296
9/4/96
296
9/15/96
_
I
~ ~-----I---i
I
r
OUTG
COMP
OUTG
OUTG OUTG
METH REASN SYSTEM
OUTG
OUTG
DAYS
TYPE
1,399 1
UNIT
ID
296
I
C
i---t
7
1--
OUTG LER
DESCRIP
ADMINISTRATIVE HOLD TO RESOLVE VARIOUS TVA AND
NRC CONCERNS
-r
TRIPPED MAIN TURBINE DUE TO EXCESSIVE VIBRATION
I
TRIPPED MAIN TURBINE FOR MAINTENANCE ON MAIN
GENERATOR CURRENT TRANSFORMER CIRCUITS.
MANUAL SCRAM OF UNIT 3 REACTOR REQUIRED BY
TESTING SCHEDULE. ALSO MAIN GENERATOR HAD
INSUFFICIENT COOLING FLOW THROUGH THE EXCITER
COOLER.
JJ
A FAILED TURBINE SPEED FEEDBACK CARD IN THE
ELECTRO-HYDRAULIC CONTROL SYSTEM CAUSED
FLUCTUATION INTHE TURBINE CONTROL AND BYPASS
VALVES THIS CAUSED A REACTOR PRESSURE SPIKE,
CAUSING AN AVERAGE POWER RANGE MONITOR HIGH
FLUX SPIKE, SCRAMMING THE REACTOR
CNV
44
F
18
___ _I
Source - INEEL / NRC MORP2
I
2
B
2
A
I
i
AD
MG
__
REACTOR SCRAMMED DUE TO LOW REACTOR WATER
LEVEL FOLLOWING FAILURE OF THE STEAM PACKING
EXHAUSTER A BYPASS VALVE
I
SHUTDOWN FOR SCHEDULED MAINTENANCE AND
REPAIRS
SHUTDOWN BY MANUAL SCRAM FOLLOWING THE 3
RECIRCULATION PUMP TRIP
I
Page 80
Category
HF/NRC
Main Turbine
Generator
EF
EF
Generator
EF
FW
IEF
Air Removal
IEF
29696001
REACTOR SCRAM DUE TO THE INADVERTENT TRANSFER 29696002
OF OIL FROM THE "3C" FEEDWATER PUMP TURBINE OIL
TANK, RESULTING IN A TRIP OF "3C" FEEDWATER PUMP
SJ
System
Operation
296/9603
~------~29960
Reci
Recirc
i
i
29696005
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
r
yr. study, 1992- 1996
BWRI4 Forced Outage Data, part 1 all events used for 5________
-.
OUTG
OUTG
HRS
DATE
2/10/9 21
2/10/94/19/92i
12/14/93
1518
UNIT
ID
298
OUTG
DAYS
298
I
3/2/94
2E57
2981
3/1694
1881
781
298
5/25/94
6,522
271 7
298
1/10/96
.
298
6/1/96
F
Source - INEEL / NRC MORP2
1
S
07
I
2
Ii
-
OUTG
SYSTEM
OUTG
COMP
I .
BTRY
BTRY
'
107
16
I
.I
OUTG OUTG
METH REASN
-I---I---"-
298
_
OUTG
TYPE
A
DESCRIP
.
.
DEGRADED 250V BATTERIES.
OUTAGE.
REPLACEMENT
BATTERY
FEEDWATER LEVEL CONTROL RFC-LC-83 FAILED,
RESULTING IN A REACTOR LOW LEVEL AND
SUBSEQUENT AUTOMATIC SCRAM.
PARTIAL CLOSURE OF Main Turbine GOVERNOR VALVES
DUE TO Main Turbine CONTROL SYSTEM MALFUNCTION
RESULTING IN A REACTOR HIGH FLUX AND SUBSEQUENT
AUTOMATIC SCRAM REPLACED FAILED POWER
SUPPLIES
VALVE RHR-MO-27A FAILED SURVEILLANCE TESTING
REPAIRED RHR-MO-27A
EDG 1 AND EDG 2 DECLARED INOPERABLE DUE TO
INSUFFICIENT UV RELAY TESTING.
TURBINE GENERATOR TAKEN OFF LINE TO REPAIR
TURBINE OIL SYSTEM.
PLANT WAS SHUTDOWN FOR FUEL LEAKER
REPLACEMENT ONE LEAKING FUEL ASSEMBLY WAS
IDENTIFIED AND REPLACED. NORMAL POWER
ASCENSION WAS IMPLEMENTED WITH FULL POWER
ACHIEVED ON 6/15/96.
27 May 1998
OUTG LER
29892003
29892003
29893038
DC
DCF
FW
Category
EF/EOL
EFIEOL
EF
129894004
EHC
EF
RHR
EF
Diesel
HF/PI
29894009
System
Main Turbine
Reactor
I
Page 81
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
OUTG
HRS
OUTG
OUTG
TYPE
DAYS
F
36
86
OUTG
OUTG OUTG
METH REASN SYSTEM
EB
GB
3
OUTG
COMP
CKRBKR
TRANSF
UNIT
ID
321
OUTG
DATE
3/28/92
321
4/30/92
84
35
S
2
B
AA
VALVEX
321
5/22/92
56
23
F
3
A
HB
321
8/27/92
66
2 8
F
3
H
321
9/30/92
16
0 7
F
2
321
6/15/93
42
17
F
321
10/22/93
91
38
321
10/28/93
17
0.7
Source - INEEL / NRC MORP2
FW
Category
HF/OA
THE UNIT WAS SHUTDOWN TO INVESTIGATE THE CAUSE
OF INCREASING TEMPERATURES INTHE UPPER
REGIONS OF THE DRYWELL PERSONNEL DISCOVERED
THE AIR SUPPLY DAMPER TO ONE OF THE COOLING
UNITS HAD CLOSED DUE TO A LOOSE WING NUT ON THE
DAMPER.
Drywell HVAC
EF
FILTER
REACTOR SCRAM WHEN DEBRIS, CAUSED BY MATERIAL 32192014
DEGRADATION OF FILTERS IN THE MAIN TURBINE'S
ELECTRO-HYDRAULIC CONTROL SYSTEM, RESTRICTED
FLUID FLOW DURING WEEKLY TURBINE STOP VALVE
TESTING. FILTER REPLACED.
EHC
EF/WP
HG
DEMINX
RPS
EF
A
HB
INSTRU
AN AUTOMATIC REACTOR SCRAM OCCURRED DUE TO A 32192021
GROUP I ISOLATION CAUSED BY AN UPSCALE SPIKE ON
THE MAIN STEAM LINE RADIATION MONITORS.
32192024
A MANUAL REACTOR SCRAM WAS INITIATED WHEN
VIBRATION AT THE NO. 3 TURBINE BEARING REACHED
APPROXIMATELY 12 MILLS DURING A POWER
REDUCTION FOR A FAILED PRESSURE SWITCH ON
MOISTURE SEPARATOR REHEATERS
Main Turbine
EF
3
H
CH
VALVEX
AN AUTOMATIC REACTOR SCRAM OCCURRED DUE TO A 32193012
FALSE LOW REACTOR WATER LEVEL SIGNAL. THIS
OCCURRED WHEN AN INSTRUMENT LINE
DEPRESSURIZED AFTER A PACKING NUT ON A VALVE IN
THE SENSING LINE DISENGAGED
RPS
EF
F
2
A
HH
INSTRU
A SIMULTANEOUS TRIP OF THREE CONDENSATE PUMPS 32193013
CAUSED A DECREASE IN FEEDWATER FLOW TO
REACTOR VESSEL AND CORRESPONDING DECREASE IN
REACTOR WATER LEVEL SHIFT INSERTED A MANUAL
REACTOR SCRAM ANTICIPATING AUTOMATIC REACTOR
SCRAM ON LOW REACTOR WATER LEVEL
Condensate
EF
F
1
B
HA
PIPEXX
SHIFT REMOVED THE MAIN TURBINE FROM SERVICE TO
REPAIR A STEAM LEAK ON THE ABOVE SEAT DRAIN FOR
CONTROL VALVE NO 4 THE LEAK WAS REPAIRED
Main Turbine
EF
OUTG LER
DESCRIP
32192009
SHIFT PERSONNEL MISTAKENLY OPENED THE SUPPLY
BREAKER TO 600V BUS 1B, CAUSING A MOMENTARY
LOSS OF CONTROL POWER TO THE REACTOR
FEEDWATER PUMPS THIS RESULTED IN AN AUTOMATIC
REACTOR SCRAM ON LOW WATER LEVEL.
"COMBUSTIBLE GAS" ALARM WAS RECEIVED.
Page 82
System
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
OUTG
HRS
OUTG
OUTG
TYPE
DAYS
F
08
OUTG
OUTG OUTG
METH REASN SYSTEM
HH
B
1
OUTG
COMP
VALVEX
UNIT
ID
321
OUTG
DATE
10/29/93
321
11/11/93
64
27
F
2
B
HA
PIPEXX
321
12/7/93
79
33
F
3
A
CH
321
12/26/93
62
2.6
F
2
A
321
3/29/94
84
35
F
3
321
11/19/94
34
14
F
321
1/4/96
87
3.6
321
4/30/96
59
24
19
Source - INEEL / NRC MORP2
System
FW heating
Category
EF
SHIFT REMOVED THE MAIN GENERATOR FROM SERVICE
AND A MANUAL SCRAM WAS INITIATED TO REPAIR AN
EHC FLUID LEAK ON A ONE INCH FLUID SUPPLY LINE
THE LEAK WAS REPAIRED AND THE UNIT WAS
RETURNED TO SERVICE
EHC
EF
CKTBRK
AN AUTOMATIC REACTOR SCRAM OCCURRED DUE TO A 32193016
LOW REACTOR WATER LEVEL SIGNAL.THE LOW WATER
LEVEL WAS CAUSED BY A TRIP OF THE "A"RFP AND
FAILURE OF THE REACTOR RECIRCULATION PUMPS TO
RUN BACK TO THE NO 2 SPEED LIMITER.
Recirc
EF
HH
VALVEX
SHIFT REMOVED THE MAIN GENERATOR FROM SERVICE,
AND A MANUAL SCRAM WAS INITIATED TO REPAIR 1N21F253 THE VALVE WAS REPAIRED AND THE UNIT
RETURNED TO SERVICE.
A
HA
GENERA
LOSS OF MAIN GENERATOR FIELD EXCITATION LED TO
LOAD REJECTION BY TURBINE-GENERATOR SYSTEM
AND RESULTED IN AN AUTOMATIC REACTOR
SHUTDOWN FIELD EXCITATION WAS LOST WHEN
ARCING OCCURRED BETWEEN THE MGE BRUSH
RIGGING AND A COLLECTOR RING ON THE MGE ROTOR
3
H
HJ
VALVEX
32194014
AN AUTOMATIC REACTOR SCRAM OCCURRED FROM
TURBINE STOP VALVE CLOSURE WHEN A TURBINE TRIP
SIGNAL WAS GENERATED DUE TO HIGH WATER LEVEL IN
THE MOISTURE SEPARATOR REHEATER "A/B"
F
3
A
HB
FILTER
AN AUTOMATIC REACTOR SCRAM OCCURRED ON HIGH
REACTOR PRESSURE WHEN ALL FOUR MAIN TCVS
DRIFTED CLOSED DUE TO THE VALVES'SERVO
STRAINERS BECOMING CLOGGED,CAUSING LOSS OF
HYDRAULIC FLUID PRESSURE TO THE SERVO VALVE
SPOOL.
F
9
F
HB
PIPEXX
DESCRIP
SHIFT REMOVED THE MAIN TURBINE FROM SERVICE TO
COMPLETE REPAIRS ON A STUCK CHECK VALVE INTHE
NORMAL DRAIN FROM THE 2ND STAGE OF THE C/D MSRS
TO THE 5TH STAGE "B"FEEDWATER HEATER. THE VALVE
WAS REPAIRED.
OUTG LER
EF
Generator
EF
Main Turbine
EF
32196001
EHC
EF
DURING STARTUP,THE UNIT EXPERIENCED HYDRAULIC 32196008
FLUID LEAKS ON A MAIN TURBINE CONTROL VALVE AND
TURBINE STOP VALVE. MEASURES TAKEN TO ISOLATE
THE LEAKS RESULTED INA PARTIAL LOSS OF
HYDRAULIC FLUID FLOW TO THE MAIN TURBINE BYPASS
VALVES
EHC
EF
Page 83
32194003
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
UNIT
ID
321
OUTG
DATE
5/26/96
321
5/26/96
OUTG
HRS
OUTG
OUTG
TYPE
DAYS
F
1.6
38
10
04
Source - INEEL / NRC MORP2
F
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
ED
INSTRU
A
9
1
A
HA
PIPEXX
OUTG LER
DESCRIP
32196009
SHIFT MANUALLY SCRAMMED REACTOR WHEN BOTH
REACTOR FEEDWATER PUMPS TRIPPED AND REACTOR
WATER LEVEL DECREASED REACTOR FEEDWATER
PUMPS TRIPPED DUE TO FAILED A BOARD. REPLACED A
BOARD
SHIFT MANUALLY TRIPPED THE MAIN TURBINE TO
REPAIR AN EHC SYSTEM FLUID LEAK
Page 84
FW
Category
EF
EHC
EF
System
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
___
UNIT
ID
324!
OUTG
DATE
1/10/921
OUTG
HRS
7
_,
OUTG OUTG
DAYS
TYPE
411
1 71 S
OUTG OUTG
OUTG
METH REASN SYSTEM
OUTG
COMP
OUTG LER
DESCRIP
MAIN GENERATOR REMOVED FROM GRID TO PERFORM 24792002
MAINTENANCE ON ELECTROHYDRAULIC CONTROL
SYSTEM (EHC) AND TO CORRECT EXCESSIVE VIBRATION
ON GENERATOR EXCITER BEARING REGULATOR FOR
EHC WAS REPAIRED AND EXCITER BEARING WORKED
I
2/2/92
324t
4/21/92
324
3/13/96
324
324
3/17/96
4/27/96 --
324
7/11/ 36
7/26/96
9/5/96
9,338
m
-
i
32
.
324
389 1
HB
XXXXXX
CH
IVALVEX
PUMPXX
113
231
.
-
--
1.0
1
F
..
I
_I----c---e
1
I
A
HI
H
---A
I
ZZZZZ
VALVEX
ZZZZZZ
1
c
REACTOR SCRAMMED DURING CONTROL VALVE
TESTING DUE TO ELECTROHYDRAULIC CONTROL (EHC)
SYSTEM FAILURE SUSPECTED CAUSE WAS AIR OR
NITROGEN INTHE SYSTEM CAUSED BY ACCUMULATOR
PERFORMANCE OR VENTING
32492001
DIESEL GENERATOR WALL 9D-1 WAS DECLARED
INOPERABLE AS A RESULT, BUSES E-5 AND E-6 WERI
DECLARED INOPERABLE BECAUSE OF TECH. SPEC
3.0 3 THE UNIT WAS TAKEN TO COLD SHUTDOWN.
REPAIR EHC SYSTEM, MAIN TURBINE, REACTOR
FEEDPUMP AND THE 4A FEEDWATER HEATER.
32492012
Category
Diesel
1
UNIT WAS OFF LINE BECAUSE OF THE TURBINE
OVERSPEED TRIP TEST, EXTENDED DUE TO MSR
MANWAY LEAKS.
Main Steam
EF
MANUAL SCRAM DUE TO SW PUMPS PROBLEMS.
REMOVED FROM SERVICE TO REPAIR
FEEDWATER HEATER LEVEL CONTROL VALVE
'
OUTAGE DUE TO HURRICANE BERTHA
I
FORCED OUTAGE TO REPAIR HD-LV-75.
OUTAGE DUE TO HURRICANE FRAN.
SW
FW heating
EF
EF
Transmission
Nature
Transmission
INature
1TURBINE
4
Source - INEEL / NRC MORP2
System
______________________
Page 85
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
.
...
OUTG
DAYS
OUTG
HRS
UNIT B OUTG
ID
DATE1
1/17/92
3251
69
3251
2/29/92T
121
325
4/21/92
325
325
OUTG
TYPE
F
_
_
...
____
OUTG OUTG
OUTG
OUTG
METH REASN SYSTEM
-L-- COMP
3
DESCRIP
LURE OF THE UNINTERRUPTIBLE POWER SUPPLY
SED THE UNIT 1 REACTOR TO SCRAM THE POWER
'LY WAS REPAIRED AND THE REACTOR WAS
JRNED TO SERVICE.
OUTG LER
32592003
;TOR SCRAMMED WHILE STOP VALVE TESTING WAS 32592005
tOGRESS. CAUSED BY A DEFECTIVE RELAY INTHE
:TROHYDRAULIC CONTROL SYSTEM
32592012
EL GENERATOR WALL 9D-1 WAS DECLARED
ERABLE. AS A RESULT, BUSES E-5 AND E-6 WERE
LARED INOPERABLE BECAUSE OF TECH. SPEC.
THE UNIT WAS TAKEN TO COLD SHUTDOWN. MORE
%IRSMADE
TG
System
Electrical
50
F
3
7,69C
3204
F
1
10/10/94
8
0.31
F
1
H
HA
TUBINE
I TURBINE MANUALLY TAKEN OFF LINE DUE TO
ESSIVE VIBRATION AT THE NO. 3 MAIN TURBINE
RING. THIS VIBRATION WAS EXPERIENCED AS
CTOR POWER WAS BEING REDUCED FOR THE
VEPLANNED MAINTENANCE ACTIVITIES.
Main Turbine
11/17/94
6
0.3
1 T
B
HA
ITURBIN
MAN UALLY TRIPPED MAIN TURBINE TO FACILITATE THE
OVAL OF THE ISOPHASE BUS DUCT COVER
ECTION PLATE
Isophase bus
cooling
TURBIN
REACTOR SCRAM DUE TO AN ERRATIC PRESSURE
ERROR SIGNAL FROM EHC.
1REACTOR SCRAM DUE TO LOW VESSEL LEVEL CAU
BY CONDENSATE SYSTEM TRANSIENT. THE TRANS
WAS A RESULT OF LOSS OF CONDENSATE PUMP
SUCTION PRESSURE CAUSED BY EXCESSIVE AIR BI
ADMITTED TO THE SUCTION HEADER OF THE
CONDENSATE PUMP
7/13/95
S-
9/30/95
61
26
3
1 A 1
HH
1PUMPXX
325
1/23/96
46
19
2
A
HA
ZZZZZZ
325
325
325
3/18/96
7/10/96
9/5/96
156
6.5
2
2
A
WA
ZZ
ZZ
PUMPXX
ZZZZ
ZZZZZZ
----
--
Source - INEEL / NRC MORP2
_
-
MANUAL SCRAM ON HI #5 MAIN TURBINE BEARING
VIBRATIONS.
FORCED OUTAGE DUE SW PUMPS PROBLEMS
OUTAGE DUE TO HURRICANE BERTHA
OUTAGE DUE TO HURRICANE FRAN
Page 86
-
Category
EF
Diesel
EF
------
32595015
'---- 32595018
_
3 25 9 60 0 3
1
-
-
Condensate
L
Main Turbine
rEF
SW
Transmission
Transmission
EF
Naur
Nature
Nature
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
UNIT
ID
331
OUTG
DATE
8/17/92
OUTG
HRS
148
OUTG OUTG
DAYS
TYPE
6.2
F
331
9/3/92
8
04
F
1
B
WF
RCB
331
11/13/92
90
3.7
F
3
A
NN
P
331
1/24/93
140
58
S
1
B
NN
331
10/26/93
50
21
F
3
A
331
5/29/94
180
75
F
1
331
7/10/94
148
6.2
F
331
11/16/94
352
14.7
331
5/14/95
78
331
6/1/95
119
OUTG OUTG
OUTG
OUTG
METH REASN SYSTEM
COMP
A
AD
FT
27 May 1998
DESCRIP
OUTG LER
AUTOMATIC REACTOR SCRAM CAUSED BY PERCEIVED
33192013
HIGH AVERAGE POWER RANGE NEUTRON FLUX,
CAUSED BY ELECTRO-MAGNETIC SIGNAL NOISE, WHICH
REDUCED FLOW BIASED SET-POINTS TO BELOW THE
CURRENT POWER LEVEL
RPS
Category
EF
THE PLANT TWICE SECURED THE GENERATOR TO
CEASE PREMATURE RECOMBINATION OF HYDROGEN
AND OXYGEN IN THE OFF GAS SYSTEM.
HIGH CONDENSER BACKPRESSURE TURBINE TRIP AND 33192018
SCRAM CAUSED BY FAILURE OF CIRCULATING WATER
PUMP SHAFT. FAILURE ALLOWED THE FLOW FROM THE
REMAINING PUMP TO SHORT CYCLE BACK TO THE PUMP
PIT AND CUT OFF FLOW TO THE CONDENSER
Offgas
EF
Circ water
EF
P
OUTAGE TO RECONNECT CIRCULATION WATER
PUMP.VERY COLD HIGH WINDS CAUSED ICING OF
CIRCULATION WATER SPRAY CREATING POTENTIAL FOR
DAMAGE TO THE COOLING TOWER FILL RESTART
FOLLOWING THE 01/24/93 OUTAGE WAS DELAYED UNTIL
THE WINDS DECREASED
Circ water
Nature
TA
FCV
THE REACTOR SCRAMMED DUE TO A MOMENTARY
33193010
GROUND COMBINED WITH AN EXISTING UNDETECTED
ELECTRICAL GROUND INTHE CONTROL CIRCUITRY FOR
THE MAIN TURBINE STEAM CONTROL VALVE
Main Turbine
EF
B
TG
TBG
FATIGUE INDUCED WELD CRACK ON AN
ELECTROHYDRAULIC CONTROL OIL SUPPLY LINE TO THE
#2 TURBINE CONTROL VALVE A 0 5 GPM HYDRAULIC
LEAK WAS DISCOVERED DURING OPERATOR ROUNDS
REPAIR OF VARIOUS BALANCE OF PLANT STEAM LINE
VALVE PACKING LEAKS
EHC
EF
2
A
JI
TBG
CRACK IN FLUID SUPPLY LINE TO TURBINE CONTROL
VALVE ELECTRO-HYDRAULIC SYSTEM REQUIRED
SHUTDOWN TO REPLACE DAMAGED SECTION OF
TUBING AND INSTALLATION OF HYDRAULIC
ACCUMULATORS INTHE SUPPLY LINE
EHC
EF
S
2
B
BJ
ISV
Drywell
EF
3.2
F
3
A
SL
GR
SHUTDOWN FOR DRYWELL ENTRY TO VERIFY SOURCE
OF AND REPAIR UNIDENTIFIED DRYWELL LEAKAGE.
THE TRIP OF THE "B" RFP WAS DUE TO STRIPPING THE
INTERNAL GEARS OF THE COUPLING BETWEEN THE
REACTOR FEED PUMP SHAFT AND LUBE OIL PUMP
FW
EF
5.0
S
1
A
SG
TBG
Condenser
EF
Source - INEEL / NRC MORP2
TUBE LEAK IN LOW PRESSURE CONDENSER
SHUTDOWN TO DRAIN THE WATERBOXES, IDENTIFY THE
LEAKING TUBE, AND PLUG IT
Page 87
33194010
33195005
System
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
UNIT
ID
3331
OUTG
DATE
2/25/93
OUTG
OUTG OUTG
DAYS
HRS
TYPE
27 1 FSF
6511
BW/ Foce
4/21/93 Ouag
3 33
OUTG OUTG
OUTG
METH REASN SYSTEM
HBA
OUTG
COMP
i
I
5/19/93
1
5/25/93
9/24/93
130
64
333
5/30/95
216
333
9/5/95
333
2/22/96
333
9/16/96
333
12/15/96
333
DESCRIP
SHUTDOWN DUE TO BLOCKAGE OF THE INTAKE
STRUCTURE SCHEDULED OUTAGE FOR "B" RECIRC
SEAL REPAIR. SHUTDOWN TO REPAIR LEAK IN CHEMICAL
DECON CONNECTION
OUTG LER
System
Recrc
92-19
4l
4
evnt usdfr5y.suy
XXXX
SHUTDOWN DUE TO LOSS OF FEEDPUMP "A" SPEED
Daa par
333
333
27 May 1998
I
I
3
I
-7-----~-----1
'
'
33393009
CONTROL DUE TO A SHORTED TERMINAL STRIP THE
TERMINAL STRIP WAS REPLACED.
SHUTDOWN DUE TO HPCI CHECK VALVE LEAK CAUSED
BY FAILED PRESSURE SEAL
' SHUTDOWN DUE TO "E"APRM UPSCALE TRIP.
' 33393013
33393020
DURING GROUND FAULT TESTING OF THE TURBINE
CONTROL SYSTEM, A BYPASS VALVE ALARM/TRIP RELAY
LEAD WAS MISTAKENLY LIFTED CAUSING #2 BYPASS
VALVE TO CLOSE AND A REACTOR TRIP ON HIGH
PRESSURE
FW
F
139
F
2
A
' NI/TIPs
DURING PERFORMANCE OF 345KV RELAY CALIBRATION
TWO TERMINALS WERE INADVERTENTLY SHORTENED
CAUSING THE 10042 AND 10052 BREAKERS TO OPEN
LEADING TO A SCRAM. WORK PROCESS IS BEING
REVIEWED FOR IMPACT ON PLANT OPERATIONS
100
Source - INEEL / NRC MORP2
2
TG
SEAL
Page 88
HF/OA
HF/OA
EHC
33396010
EHC HYDRAULIC FLUID LEAK ON NO 1 TURBINE BYPASS 33396014
VALVE ACTUATOR SEAL
'
Recrc
AN INADVERTENT REMOVAL OF A FEEDWATER CONTROL 33395013
FUSE CAUSED A FEEDPUMP TRANSIENT AND PLANT
SCRAM ON LOW WATER LEVEL
WHILE PERFORMING A CONTROLLED REACTOR
33396002
SHUTDOWN DUE TO EXCESSIVE SCRAM TIME, AN EHC
LINE TO TURBINE BYPASS VALVES RUPTURED.
OPERATORS INSERTED A MANUAL SCRAM EHC TUBING
WAS MODIFIED WITH FLEXIBLE TUBING AND SCRAM
SOLENOID PILOT VALVE DIAPHRAGMS REPLACED.
TG
EF
HPCI
33395010
A 3/4" MANUAL VALVE (JET PUMP TO RECIRC PUMP
SUCTION) PACKING LEAK EXCEEDED TECH SPEC LIMITS.
THE PACKING WAS REPLACED WITH A DIFFERENT STYLE
THAT IS LESS PRONE TO GROSS FAILURE.
74
Category
Electrical
-i---
1L
HF/PI
EF
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
UNIT
ID
3411
OUTG
DATE
3/16/921
341
4/7/92
341
11/18/92
OUTG
DAYS
OUTG
HRS
2/19/93
OUTG OUTG
OUTG
METH REASN SYSTEM
OUTG
COMP
46
39
F
1
A
NH
VACB
S
-~------c
321
I
341
341
4/10/93
4/20/93
240
33
341
8/13/93
491
System
Recrc
Category
HF/OA
34192003
DURING PERFORMANCE OF A ROUTINE SURVEILLANCE
ON DRYWELL AND SUPPRESSION CHAMBER VACUUM
BREAKER OPERABILITY, A VACUUM BREAKER DID NOT
CLOSE AFTER BEING OPENED THE VACUUM BREAKER
ACTUATOR (UTILIZED DURING TESTING ONLY) BOUND IN
THE OPEN POSITION
Containment
EF
34192012
Condensate
HF/OA
FW heating
Condenser
EF
DESCRIP
MANUAL REACTOR SCRAM DUE TO OPERATION IN
REGION OF INSTABILITY (HIGH POWER TO FLOW)
FOLLOWING INADVERTENT ACTUATION OF SAFETY
SYSTEMS DURING SURVEILLANCE TEST SAFETY
SYSTEM ACTUATION OCCURRED WHEN TEST METER
SHORTED INTERNALLY
SCRAM DUE TO LOSS OF HEATER FEEDWATER PUMPS
INADVERTENT OPENING OF CONDENSATE
DEMINERALIZER INLET VALVE RESULTED IN LOSS OF
HEATER FEED PUMP (HFP) AUCTION PRESSURE AND
CONSEQUENT TRIP OF HFP.
57
12/5/92
2/10/93
341
OUTG
TYPE
27 May 1998
2.0
9/17/93
Source - INEEL / NRC MORP2
F
COND
REPAIR OF EXTRACTION STEAM LINE RUPTURE
CONDENSER TUBE LEAK CAUSED CONDENSATE
CHEMISTRY TO REACH ACTION LEVEL PLANT
SHUTDOWN FOR TUBE PLUGGING
OUTG LER
34192002
I
ROUTINE PUMP BREAKER PM TESTING INADVERTENTLY 34193004
ACTUATED IN-SERVICE TRIP RELAYS AN IN-SERVICE
PUMP BREAKER TRIP RELAY FAILED TO PROPERLY
ACTUATE, LEADING TO INABILITY TO TRANSFER FEED TO
ALTERNATE SUPPLY.
Electrical
REPAIR OF EXTRACTION STEAM LINE RUPTURE
SCRAM OCCURRED DURING RECOVERY FROM
EXTRACTION STEAM LINE REPAIR OUTAGE SCRAM
CAUSED BY INCORRECTLY INSTALLED TEST
INSTRUMENT WHICH LEAKED STEAM AND WATER ONTO
MAIN STEAM MANIFOLD PRESSURE TRANSMITTERS
FW hea ing
Main Steam
THE REACTOR SCRAM WAS AUTOMATICALLY INITIATED
BY A TRIP OF THE MAIN TURBINE DUE TO A FALSE HIGH
REACTOR WATER LEVEL SIGNAL
34193007
34193010
34193013
WHILE SHUTTING DOWN TO REPAIR A HEATER DRAIN
SYSTEM LEVEL CONTROL VALVE, PRESSURE INTEGRITY
WAS LOST DUE TO MAINTENANCE ACTIVITIES ON THE
VALVE. THIS RESULTED IN LEAKAGE FROM THE
FEEDWATER SYSTEM WHICH WAS TERMINATED AFTER
THE REACTOR WAS SCRAMMED.
Page 89
EF
HF/OA
FW heating
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
OUTG
HRS
OUTG
OUTG
TYPE
DAYS
0.9
F
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
TA
52
1
B
UNIT
ID
341
OUTG
DATE
9/20/93
341
12/25/93
1,835
76 5
F
3
A
TA
TRB
REACTOR TRIPPED FOLLOWING TRIP OF MAIN TURBINE
EXTENSIVE DAMAGE TO LOW PRESSURE TURBINE
NUMBER 3, THE MAIN GENERATOR, AND THE MAIN
GENERATOR EXCITER OCCURRED DURING THIS EVENT
CAUSES OF THIS EQUIPMENT DAMAGE ARE UNDER
INVESTIGATION REFUELED MEANWHILE
341
1/27/95
387
16.1
S
1
B
TA
TRB
341
2/1/95
259
10 8
F
9
B
TD
PSP
341
2/13/95
663
27 6
F
4
B
TD
PSP
341
3/14/95
18
0.8
S
1
B
TA
TRB
341
3/16/95
21
09
S
1
B
TA
TRB
341
3/26/95
48
20
F
1
B
TJ
PSP
341
4/9/95
62
26
S
2
B
TA
TRB
341
4/12/95
42
17
F
1
B
SB
ISV
341
4/25/95
249
10 4
F
3
A
JJ
RG
341
6/2/95
334
13.9
F
3
B
TA
SIS
341
3/27/96
526
21.9
F
2
B
BI
TK
341
4/19/96
108
45
F
2
B
BJ
PC
341
11/20/96
504
21 0
F
4
A
AC
RV
22
Source - INEEL / NRC MORP2
System
Main Turbine
Category
EF
Main Turbine
EF
TURBINE TAKEN OFF LINE TO PERFORM POST OUTAGE
BALANCING. TURBINE REMAINED OFF LINE TO REPAIR #4
JACKING OIL PUMP DISCHARGE PIPING.
Main Turbine
EF
TURBINE REMAINED OFF LINE TO REPAIR #4 JACKING
OIL PUMP DISCHARGE PIPING.
TURBINE TAKEN OFF LINE TO REPAIR TURBINE JACKING
OIL SYSTEM STRUCTURAL CONCERNS
TURBINE TAKEN OFF LINE TO PERFORM POST OUTAGE
BALANCING.
TURBINE TAKEN OFF LINE TO OBTAIN TURBINE
COASTDOWN BEARING VIBRATION DATA.
TURBINE TAKEN OFF LINE TO REPAIR A STATOR
COOLING WATER VENT LINE LEAK
34195004
MANUAL REACTOR/TURBINE TRIP PER SOE 95-10 TO
OBTAIN HOT TURBINE COASTDOWN VIBRATION DATA AT
APPROXIMATELY 80% POWER.
TURBINE TAKEN OFF LINE TO REPAIR N3018F607, MAIN
STEAM TO MSR ISOLATION VALVE.
34195005
AUTOMATIC REACTOR SCRAM ON APRM NEUTRON
UPSCALE TRIP RESULTING FROM REACTOR PRESSURE
REGULATOR TRANSIENT.
34195006
AUTOMATIC MAIN TURBINE TRIP ON MECHANICAL
OVERSPEED TRIP RING #2 WHILE PERFORMING MTG
OVERSPEED TRIP TEST
34196005
TECH SPEC REQUIRED SHUTDOWN DUE TO BOTH
DIVISIONS OF EECW BEING DECLARED INOPERABLE DUE
TO MAKE-UP TANK DESIGN ISSUE MODIFICATION BEING
INSTALLED
Main Turbine
EF
Main Turbine
EF
Main Turbine
EF
Main Turbine
EF
SWC
EF
Main Turbine
EF
Main Turbine
EF
RPS
EF
Main Turbine
EF
ESW
EF/WD
HPCI
EF
Main Steam
EF
DESCRIP
DURING STARTUP, THE MAIN TURBINE TURNING GEAR
CIRCUIT BREAKER FAILED. THE REACTOR WAS
SHUTDOWN TO MINIMIZE DIFFERENTIAL HEATING OF
THE TURBINE SHAFT DURING THE TIME THAT TURNING
GEAR WAS OUT OF SERVICE. TURNING GEAR BREAKER
WAS REPLACED
DURING UNIT STARTUP HPCI AND RCIC DECLARED
INOPERABLE, TECH SPEC REQUIRED SHUTDOWN.
UNIT SHUTDOWN TO REPAIR SRV 'A' TAIL PIPE
PRESSURE SWITCHES
Page 90
OUTG LER
34193014
34196007
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
DESCRIP
UNIT REMAINED SHUTDOWN BEYOND THE 53 DAY
SCHEDULED REFUEL OUTAGE DUE TO EDG 11
AUTOMATIC VOLTAGE REGULATOR FAILURE.
UNIT SHUTDOWN TO REPAIR SRV 'D' SOLENOID
ACTUATOR
UNIT SHUTDOWN TO REPAIR T23-F400J DRYWELL TO
TORUS VACUUM BREAKER.
REACTOR SCRAM DUE TO FALSE LEVEL 2 AND 8
INITIATION WHILE VALVING IN REFERENCE LEG OF RX
WATER LEVEL BACK FILL.
Source - INEEL / NRC MORP2
Page 91
LER
System
_/ OUTG
~i~L~
Diesel
34196023
Main Steam
EF
Drywell
EF
34904Reactor
eco
34196024
Category
HF/OA
I
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
-
UNIT
ID
OUTG
DATE
9/7/93
OUTG
HRS
OUTG
DAYS
2
A
F
1
A
F
3
1/14/94 T-
571
24
F
352
10/8/94
12
0.5
352
2/21/
16
5/7/95
OUTG OUTG
OUTG
METH REASN SYSTEM
OUTG
TYPE
352
352
1
4.11 S
2
HA
OUTG
COMP
CKTBRK
GENERA
TURBIN
CONROD
1
B
HH
1HTEXCH
EF
Main Steam
EF
EHC
EF
Main Steam
HF
Electncal
|EF
9
A
PIPEXX
352
9/2/95
352
9/11/95
338
352
3/24/96
158
6/17/96
352
7/15/96
352
VALVEX
7/25/96
INSTRU
VALVEX
-i----t-
135
71
-r
0.3
5.6
3.0
Source - INEEL / NRC MORP2
~--l---~----------T
CHTBRK
HTEXCH
F
F
SA
HA
EF
Main Steam
46
352
Recirc
35295008
REACTOR WAS SHUTDOWN DUE TO A FAILED OPEN
SAFETY RELIEF VALVE (SRV)
REACTOR WAS SHUTDOWN DUE TO SRV AND TIP
MACHINE MAINTENANCE.
t
MAIN TURBINE WAS TAKEN OFF THE GRID TO REPLACE
THE EHC SPEED CONTROLLER CARD.
35296013
REACTOR SCRAM DUE TO A PRESSURE SPIKE DURING
THE PERFORMANCE OF A MSIV SURVEILLANCE TEST
REACTOR POWER WAS REDUCED AND THE TURBINE
WAS TAKEN OFF THE GRID DUE TO REPAIR OF 220KV A
CIRCUIT BREAKER, REACTOR REMAINED CRITICAL
REACTOR WAS SHUTDOWN TO REPAIR UNISOLABLE
LEAKS IN SJAE ROOM.
I
REACTOR WAS SHUTDOWN DUE TO A ELECTROHYDRAULIC CONTROL TRANSIENT OF TURBINE
CONTROL VALVES
111
5/21/96
EF
VALVEX
8/28/95
352
Transmission
HF/PI
352
I-------e
--
EF
H2 Recombiner
FUELXX
4/2/96
Main Turbine
UNIT SHUTDOWN INACCORDANCE WITH TECH SPEC.
3 0 3 AS A RESULT OF DISCOVERING BOTH POST-LOCA
HYDROGEN RECOMBINER SYSTEMS WERE INOPERABLE
DUE TO IMPROPER WIRING OF CERTAIN RECORDERS
DURING A RECENT RECORDER MODIFICATION.
A
I
EF
RECOMB
2
6.6
SWC
HF/C
77
141
I
Category
EF
Drywell
186
0.6
REACTOR MANUAL SCRAM OCCURRED DUE TO A LOSS 35294001
OF STATOR WATER COOLING
TURBINE TAKEN OFF LINE DUE TO HIGH TURBINE
VIBRATION
REACTOR SCRAM DUE TO ELECTRICAL DISTURBANCE AT
WHITPAIN SUBSTATION.
REACTOR WAS SHUTDOWN TO PERFORM MAINTENANCE
ON "C" DRAIN COOLER, "A" RECIRCULATION PUMP SEAL,
AND THE CONDENSER WATERBOXES
System
Electrical
Reactor
8/20/95
141
OUTG LER
DESCRIP
AUTOMATIC REACTOR SCRAM OCCURRED AFTER LOSS
OF AN OFFSITE POWER SUPPLY DURING THE
AUTOMATIC TRANSFER TO THE SECONDARY POWER
SUPPLY, BREAKER FOR THE 1A FEEDWATER CONTROL
SYSTEM FAILED TO RECLOSE, RESULTING IN A
REDUCTION IN REACTOR WATER LEVEL.
REACTOR WAS SHUTDOWN TO REPLACE A FAILED FUEL
BUNDLE
35295006
REACTOR WAS SHUTDOWN SHORTLY AFTER BEING
CRITICAL DUE TO LEAKAGE INTO THE DRYWELL CAUSED
BY A MISALIGNED REACTOR PRESSURE VESSEL
INSTRUMENT FLANGE CONNECTION.
352
352
7
7
______________
I
VALVEX
Page 92
35295007
Air Removal
EF
1i
I
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
r
___________
UNIT
0ID
353
OUTG
OUTG
DATE
HRS
11/20/921
47
OUTG
OUTG OUTG OUTG
OUTG
OUTG
DESCRIP
DAYS
TYPE - METH REASN SYSTEM I COMP ,
19
THE TURBINE GENERATOR WAS TAKEN OFF LINE TO
PIPEXX
REPAIR AN EHC LEAK ON THE #3 MAIN TURBINE
CONTROL VALVE.
-T---~-----7----~7-------~--
353
12/4/92
353
1/3/93
353
3/17/93
353
3/26/93
353
5/15/93
353
10/19/94
353
2/21/95
~-
i
PUMPXX
INSTRU
5
42
81
02
1.7
F
F
HTEXCH
-----L~------C--~----
031
21
F
3
A
HA
TURBIN
1
A
HA
PIPEXX
3
A
INSTRU
CONROD
79
T~----~----~-------1--------
353
6/3/95
6
S
353
8/8/95
70
353
8/20/95
60
353
11/22/95
353
5/1/96
08
F
353
5/2/96
05
F
353
5/14/96
24
F
3
H
353
10/6/96
F
11
1 A
353
12/6/961
25
3
A
HA
VALVEX
F
3
A
CH
INSTRU
F
3
A
HA
INSTRU
r
--
57
21
185
Source - INEEL / NRC MORP2
i
REACTOR WAS MANUALLY SCRAMMED AFTER BOTH
RECIRCULATION PUMPS TRIPPED DURING
SURVEILLANCE TESTING
REACTOR AUTOMATICALLY SCRAMMED ON HIGH
REACTOR PRESSURE CAUSED BY MAIN TURBINE
CONTROL VALVE CLOSURE DUE TO AN UNDETERMINED
DHC MALFUNCTION.
THE MAIN TURBINE TRIPPED OFF LINE DUE TO LOW
FLOW IN THE STATOR WATER COOLING SYSTEM
THE UNIT AUTOMATICALLY SHUTDOWN DUE TO AIR
ENTRAINED INTHE MAIN TURBINE ELECTRO HYDRAULIC
CONTROL SYSTEM
1
Recirc
Category
---
IEF
EF
35393001
SWC
35393005
REACTOR POWER WAS REDUCED TO 19% AND THE
TURBINE TAKEN OFF LINE TO REPAIR A LEAK INTHE
ELECTRO-HYDRAULIC CONTROL SYSTEM
REACTOR SCRAM ASSOCIATED WITH A RELAY COIL
FAILURE COMBINED WITH AN INAPPROPRIATE ACTION
TAKEN BY AN OPERATOR DURING TESTING OF AN
EMERGENCY DIESEL GENERATOR.
REACTOR SCRAM DUE TO ELECTRICAL DISTURBANCE AT
WHITPAIN SUBSTATION
MAIN TURBINE REMOVED FROM SERVICE TO REPAIR
EHC LEAK AT #4 CIV.
REACTOR WAS SHUTDOWN DUE TO A FAILED POWER
35395008
SUPPLY INTHE FEED WATER CONTROL SYSTEM
REACTOR WAS SHUTDOWN DUE TO A HIGH IMPEDANCE 135395010
ACROSS THE EHC CONTROL RELAY CONTACT
RESULTING IN SPORADIC OPENING AND CLOSING
TURBINE BYPASS VALVES.
IDiesel
HF/OA
Transmission
7
THE GENERATOR WAS TAKEN OFF THE GRID TO
REPLACE THE STATOR WATER COOLING FILTERS
VALVEX
THE TURBINE WAS TAKEN OFF THE GRID DUE TO REPAIR
OF TURBINE EHC LEAK.
-~------T
INSTRU
TURBINE WAS TAKEN OFF THE GRID DUE TO A FAILED
BACKUP OVERSPEED TRIP TEST.
35396004
ZZZZZ
REACTOR SCRAM DUE TO TURBINE TRIP CAUSED BY
GRID INSTABILITY.
'
POWER REDUCTION DUE TO REPLACEMENT OF THE
iFILTER
HA
GENERATOR STATOR COOLING "Y" STRAINER.
REACTOR WAS SHUTDOWN DUE TO A FAILED PRESSURE 35396007
INSTRU
SWITCH ON THE EHC SYSTEM
FILTER
A
System
-
_, OUTG LER _~
Page 93
EHC
EF
SWC
EF
EHC
EF
Main Turbine
EF
Transmission
EF
SWC
EF
EF
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
UNIT
ID
353
OUTG
DATE
12/18/96
OUTG
HRS
115
353
12/24/96
31
OUTG
OUTG
DAYS
TYPE
4.8
F
13
Source - INEEL / NRC MORP2
F
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
HTEXCH
HC
A
2
2
A
CB
MECFUN
OUTG LER
DESCRIP
REACTOR WAS SHUTDOWN DUE TO CRACK IN THE MAIN
CONDENSER NECK SEAL GASKET
35396009
REACTOR WAS SHUTDOWN DUE TO A FAILED SCOOP
TUBE POSITIONER ON THE 2B MG SET.
Page 94
System
Condenser
Category
EF
Recirc
EF
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
OUTG OUTG
DAYS
TYPE
4 7
F
OUTG OUTG
METH REASN
A
2
UNIT
ID
354
OUTG
DATE
5/26/92
OUTG
HRS
112
354
6/13/92
19
0.8
F
1
A
354
12/3/92
52
2.2
F
2
H
354
354
5/16/93
12/1/93
100
132
4.2
5.5
F
F
3
2
A
A
354
5/15/94
153
6.4
F
3
B
354
6/21/94
16
07
F
1
A
354
8/1/94
63
26
F
3
A
354
8/30/94
45
19
F
3
H
354
10/2/94
238
9.9
F
3
A
354
3/20/95
181
75
F
2
354
7/8/95
418
174
F
354
11/1/96
146
61
S
Source - INEEL / NRC MORP2
OUTG
SYSTEM
OUTG
COMP
System
Drywell
Category
EF
EHC
EF
Recirc
HF/OA
UNIT TRIPPED DUE TO FAULTY COMPONENT IN EHC
UNIT SHUTDOWN DUE TO EXCESSIVE ARCING OF THE
MAIN GENERATOR EXCITER BRUSHES
AUTOMATIC SCRAM DURING DIGITAL FEEDWATER
TESTING.
POWER REDUCTION TO REPAIR EHC LEAK ON #2
BYPASS VALVE.
AUTOMATIC SCRAM DURING IRM SURVEILLANCE DUE TO
FAULTY TEST EQUIPMENT
AUTOMATIC SCRAM CAUSED BY GENERATOR RUNBACK
DUE TO LOSS OF STATOR WATER COOLING
AUTOMATIC SCRAM CAUSED BY DESIGN ERROR IN
DIGITAL FEEDWATER CONTROL SYSTEM WHEN
ATTEMPTING RESTART AN EHC SYSTEM PROBLEM
CAUSE A PROBLEM WITH THE TURBINE ROLL.
OPERATOR CLOSED ALL TURBINE VALVES WHICH
RESULTED INAUTO REACTOR SCRAM.
EHC
Generator
EF
EF
G
1
1
DESCRIP
FAILED DRYWELL TO SUPPRESSION CHAMBER DECAY
TEST, POWER WAS REDUCED TO 21% AND THE
REACTOR WAS MANUALLY SCRAMMED.
UNIT WAS TAKEN OFF LINE TO REPAIR EHC LEAK. THE
REACTOR WAS KEPT AT APPROXIMATELY 3% POWER
FOR THE DURATION OF THE OUTAGE.
CONTRACT EMPLOYEE BUMPED CART INTO A MCC,
CAUSING REACTOR RECIRCULATION PUMP M/G SET
VENT FANS TO TRIP, RESULTING IN A DOUBLE
RECIRCULATION PUMP TRIP. CONTROL OPERATOR
MANUALLY SCRAMMED THE REACTOR
OUTG LER
35492006
35492013
FW
EHC
EF
NI/TIPs
EF
SWC
EF
FW
EF/WD
WHILE I&C TECHS WERE PERFORMING A PM ON THE
OPTICAL ISOLATOR FOR THE REACTOR RECIRCULATION
PUMP MG SETS, A LOSS OF BOTH MG SETS OCCURRED
A MANUAL SCRAM WAS INITIATED IN ACCORDANCE WITH
THE PROCEDURE.
Recirc
HF/OA
A
WHEN LCO 3 7.2 A FOR CONTROL ROOM VENTILATION
ACTION STATEMENT EXPIRED A UNIT SHUTDOWN WAS
INITIATED CAUSE WAS A MOMENTARY INTERRUPTION
TO THE CONTROL CIRCUIT COMBINED WITH LENGTHY
CABLE RUNS.
MCR HVAC
EF
B
PLANNED MAINTENANCE TO REPAIR THE REACTOR
RECIRC PUMP SEAL.
Recirc
EF
Page 95
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
OUTG OUTG
TYPE
DAYS
6 9 SF
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
PIPEXX
2
B
CB
TURBIN
UNIT
ID
366
OUTG
DATE
1/24/92
OUTG
HRS
166
366
6/25/92
65
2 7
F
3
G
EB
CKTBKR
366
11/24/92
45
191
F
1
B
HA
366
11/27/92
49
20
F
3
H
366
12/6/92
185
77
S
1
366
3/4/93
807
336
F
366
5/21/93
34
14
366
11/8/93
183
366
8/30/94
52
System
Recirc
Category
EF
PERSONNEL ERROR WHILE SEARCHING FOR A GROUND 36692009
ON LPCI INVERTER 2R24-SO18A, RESULTING IN THE
SUPPLY BREAKER OPENING TO 600V BUS C, WHICH
CAUSED A LOSS OF CONTROL POWER TO THE REACTOR
FEED PUMPS, FOLLOWED BY AN AUTOMATIC REACTOR
SCRAM.
Electrical
HF/OA
VALVOP
THE MAIN TURBINE WAS TAKEN OFF LINE TO REPAIR AN
ELECTRO-HYDRAULIC CONTROL SYSTEM FLUID LEAK AT
THE REHEAT CYLINDER ON COMBINED INTERMEDIATE
VALVE NO 4.
EHC
EF
HA
TURBIN
Main Turbine
EF
B
HJ
GENERA
AN AUTOMATIC REACTOR SCRAM OCCURRED WHEN
VIBRATION AT THE NO. 6 TURBINE BEARING REACHED
APPROXIMATELY 12 MILS
THE UNIT WAS SHUTDOWN DUE TO A HYDROGEN LEAK
AT THE NEUTRAL BUSHING ON THE MAIN GENERATOR.
THE BUSHING WAS REPLACED AND TESTED FOR
LEAKAGE, AND THE UNIT WAS RETURNED TO RATED
THERMAL POWER
Generator
EF
1
A
RC
FUELXX
THE UNIT WAS SHUTDOWN TO IDENTIFY AND REMOVE
THE LEAKING FUEL BUNDLE FROM THE CORE AND
INSPECT OTHER FUEL BUNDLES FOR POSSIBLE
DAMAGE
Reactor
EF
F
2
G
ZZ
Recirc
EF
76
F
2
B
SF
VALVEX
A MANUAL REACTOR SCRAM WAS INITIATED WHEN BOTH 36693005
REACTOR RECIRCULATION PUMPS TRIPPED.
SHIFT REMOVED THE MAIN GENERATOR FROM SERVICE,
AND A MANUAL SCRAM WAS INITIATED TO INVESTIGATE
INCREASED LEAKAGE INTO THE DRYWELL FLOOR DRAIN
SYSTEM. INVESTIGATION REVEALED A BONNET
PRESSURE SEAL LEAK ON CORE SPRAY TESTABLE
CHECK VALVE 2E21-F006B
Core Spray
EF
22
F
3
H
IA
INSTRU
RPS
HF/OA
Source - INEEL / NRC MORP2
DESCRIP
UNIDENTIFIED SOURCE OF FLOOR DRAIN LEAKAGE
INSIDE THE DRYWELL INVESTIGATION REVEALED THE
PACKING LEAKOFF LINE FOR THE "B" REACTOR
RECIRCULATION PUMP'S DISCHARGE ISOLATION VALVE
HAD SEPARATED TURBINE ROTOR REQUIRED
REBALANCING
AUTOMATIC REACTOR SCRAM WHEN RPS ELECTRICAL
BUS 2A WAS BEING TRANSFERRED FROM ITS
ALTERNATE TO ITS NORMAL SUPPLY THE EVENT WAS
CAUSED BY INADVERTENTLY MOVING THE SWITCH
BEYOND ITS CENTER POSITION WHEN TRANSFERRING
FROM "ALT A"TO THE "NORM" POSITION
Page 96
OUTG LER
36692026
36694007
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
-
....
OUTG
HRS
271
.
.
OUTG OUTG
METH REASN
3
G
OUTG
OUTG
TYPE
DAYS
113
F
OUTG
OUTG
SYSTEM L COMP
--INSTRU
CG
UNIT
ID
3661
OUTG
DATE
4/11/95
366
5/4/95
GENERA
THE UNIT WAS MANUALLY SCRAMMED TO REPAIR THE
NO, 9 AND NO.10 BEARINGS ON THE MAIN TURBINE
GENERATOR THE NO 9 AND NO 10 BEARINGS AND
JOURNALS WERE REPAIRED
Main Turbine
9/2/95
HTEXCH
36695003
UNIT WAS MANUALLY SCRAMMED DUE TO A
DECREASING VACUUM ON THE "A"MAIN CONDENSER AS
A RESULT OF "D" WATERBOX BECOMING AIRBOUND
AFTER FILL MATERIAL IN CELL 10 OF COOLING TOWER
NO 5 COLLAPSED AND CLOGGED THE SCREENS AT THE
TOWER
Circ water
MAIN TURBINE TRIPPED ON MOISTURE SEPARATC)R
REHEATER HIGH LEVEL A MOTOR OPERATED VAL.VE IN
THE HIGH LEVEL DRAIN LINE WAS FOUND CLOSE[ D
THE UNIT WAS MANUALLY SCRAMMED TO REPLA( CE
MAIN STEAM LINE SRVS "D"AND "H"
SHIFT MANUALLY TRIPPED THE MAIN TURBINE AN
INSERTED A MANUAL SCRAM TO REPAIR A LEAK C
REACTOR FEEDWATER VENT LINE
Main Turbine
EF
Main Steam
EF
oil0
366
11/21/95
366
3/12/96
83
366
4/25/96
62
I
i
G
HJ
HTEXCH
34
A
CC
VALVEX
2.6
A
HH
PIPEXX
_I ,
Source - INEEL / NRC MORP2
I
F
- i
2
-
1
c
I
i
DESCRIP
AN AUTOMATIC REACTOR SCRAM OCCURRED DUE TO
THE IMPROPER PLACEMENT OF A JUMPER WHILE
ATTEMPTING TO RETURN THE REACTOR WATER
CLEANUP SYSTEM TO SERVICE
-i
Page 97
OUTG LER
36695001
Category
System
-------------------HF/OA
!EF
EF
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
I
I
1
7
1
I _
I
OUTG
DATE
5/17/92
UNIT
ID
387
387
OUTG
OUTG
TYPE
I DAYS
DAYI
OUTG
HRS
OUTG OUTG
OUTG
METH I REASN I SYSTEM
OUTG
COMP
TRB
6/5/92
I
OUTG LER
DESCRIP
UNIT ONE TOOK THE GENERATOR OFF LINE AT 0327
HOURS MAY 17TH DUE TO HIGH VIBRATION ON THE #5
BEARING OF THE MAIN TURBINE
THE UNIT WAS TAKEN OFF LINE TO REPAIR THE "A"
138792010
REACTOR FEED PUMP ISOLATION VALVES. THE VALVES
WERE REPAIRED AND A STARTUP COMMENCED THE
STARTUP WAS HALTED AND THE UNIT MANUALLY
SCRAMMED DUE TO AN IGNITION OF CHARCOAL IN THE
1B OFFGAS GUARD BED.
387
11/12/92
57
THE SCRAM WAS CAUSED BY A FAULTY RELAY IN ONE
DIVISION OF THE RFP TURBINE, MAIN TURBINE HI LEVEL
TRIP CIRCUIT WHILE A SURVEILLANCE WAS BEING
PERFORMED INTHE OTHER DIVISION OF THE HI LEVEL
TRIP LOGIC
38792017
387
7/12/93
1,203
UNIT ONE EXPERIENCED AN AUTOMATIC MAIN TURBINE
TRIP WITH AUTOMATIC REACTOR SCRAM. MAIN
TURBINE TRIPPED ON HIGH VIBRATION CAUSED BY
FAILURE OF TWO TURBINE BUCKETS ON THE C LOW
PRESSURE ROTOR
38793008
387
11/10/951
606j
387
8/1/96
103
387
10/28/961
1321
25 3
5.51
2
F
3
TH
4
1
F
1
GEN
OUTAGE TO REPAIR A HYDROGEN LEAK INTO THE
STATOR WATER COOLING SYSTEM. UNEXPECTED MAIN
GENERATOR BAR TIE, SPACER DAMAGE AND WEDGE
LOOSENESS WAS IDENTIFIED WHICH EXTENDED THE
OUTAGE DURATION
TA
VIS
TURBINE TRIP WAS CAUSED BY A FALSE SPURIOUS
SIGNAL FROM TURBINE #1 BEARING VIBRATION
INSTRUMENTATION LOOP COMPONENTS WERE
REPLACED THAT MOST LIKELY CONTRIBUTED TO THIS
SPURIOUS SIGNAL
SB
IPSF
FORCED OUTAGE ACTIVITIES INCLUDED REPLACEMENT
OF A PORTION OF THE MAIN STEAM LINE DRAIN PIPING,
ALIGNMENT CHECKS ON THE "B" REACTOR
RECIRCULATION PUMP AND INSTALLING A MAIN
TURBINE/GENERATOR BALANCE SHOT
B
S
A
I
I
System
Main Turbine
Category
EF
FW
EF
L
Main Turbine
EF
SWC
38796006
Main Turbine
Main Steam
L
Source - INEEL / NRC MORP2
Page 98
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 1 all events used for 5 yr. study, 1992 - 1996
I
UNIT
ID
388
r
OUTG
DATE
3/18/921
OUTG
HRS
144
1/29/93
186
388
12/10/93
480
388
1/20/94
49
388
2/20/94
388
6/12/94
388
4/15/95
388
7/14/96
I
4
OUTG
DAYS
i
20
0.2
OUTG OUTG
METH REASN
OUTG
TYPE
i
F
F
I
3
1 1
OUTG
SYSTEM
EB
I
i
T
182
Source - INEEL / NRC MORP2
F
DESCRIP
DURING PERFORMANCE OF OP PROCEDURE TO SWAP
IN THE E DIESEL GENERATOR, AN OPERATOR FOUND
THE PROTECTIVE RELAY ON THE C DIESEL PANEL
TRIPPED. FURTHER PROBLEMS RESULTED IN THE
POTENTIAL FOR INBOARD MSIVS TO GO CLOSED,
CAUSING UNIT TO BE MANUALLY SCRAMMED
OUTG LER
38892001
i
SJ
V
CBL1
/
/
System
Electrical
Category
POWER REDUCED TO INVESTIGATE AND REPAIR
CONDENSER TUBE LEAKS CONDENSER
DEMINERALIZER INFLUENT (CDI) CONDUCTIVITY
EXCEEDED ADMINISTRATIVE LIMITS MAIN
TURBINE/GENERATOR WAS MANUALLY TRIPPED
FORCED OUTAGE FOR GENERATOR EXCITER FIELD
GROUND.
Condenser
EF
MANUALLY SHUTDOWN DUE TO HIGH DRYWELL
LEAKAGE. INSPECTION OF DRYWELL REVEALED A
CRACKED WELD ON THE "A" RX RECIRC PUMP RBCCW
OUTLET LINE OTHER WORK INCLUDED INSTALLATION
OF TORQUE COLLARS ON THE MAIN TURBINE AND
INSTALLATION OF RX LEVEL INSTRUMENTATION
Recirc
EF
SWC
EF
AUTOMATIC MAIN TURBINE TRIP WITH AUTOMATIC
REACTOR SCRAM AS A RESULT OF A STATOR COOLING
WATER TEMPERATURE CONTROL VALVE PROBLEM
WHICH CAUSED HIGH STATOR COOLING WATER
TEMPERATURES
F
438
1
OUTG
COMP
RLY-87
i
A
I
38894002
THE MAIN GENERATOR WAS TAKEN OFF LINE TO REPAIR
EHC LEAK ON THE #3 CONTROL VALVE.
REMOVED THE MAIN TURBINE FROM SERVICE TO
REPAIR A STEAM LEAK ON FW HEATER BTV-20210B
38895005
AUTOMATIC REACTOR SCRAM DUE TO A MAIN
GENERATOR LOAD REJECT REPLACED "A" REACTOR
RECIRC PUMP SEAL AND "B" MAIN TRANSFORMER
BUSHING.
REACTOR SCRAMMED WHEN ALL FEEDWATER WAS
LOST DURING POST MAINTENANCE TESTING OF TIE
BUS 0A107, POWER WAS LOST TO AUXILIARY BUS 12A
THIS CAUSED 2 CONDENSATE PUMPS TO TRIP AND DUE
TO LOW SUCTION PRESSURE ALL 3 RFP'S TRIPPED.
Page 99
38896004
EF
FW heating
EF
Generator
Electrical
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
OUTG
OUTG
OUTG OUTG OUTG OUTG
HRS
DAYS TYPE METH REASN SYSTEM
146
61
F
3
B
27 May 1998
DESCRIP
System
DURING PLANNED MAINTENANCE ACTIVITIES ON Instrument Air
THE SCRAM PILOT AIR HEADER, UNIT 2
AUTOMATICALLY SCRAMMED ON LOW SCRAM AIR
HEADER PRESSURE FOLLOWING ISOLATION OF
BOTH PRIMARY AND SECONDARY SCRAM PILOT
AIR HEADER PRESSURE REGULATORS
Category
HF/ICS
OUTO LER
26094004
DAYS
DAYS
FROM
START FROM
OP
OP
OF
ENDOF
CYCLE CYCLE
START LENGTH CYCLE CYCLE
6/4/93
482
315
167
AUTOMATIC SCRAM CAUSED BY BALANCE OF
PLANT EQUIPMENT FAILURE.
AUTOMATIC SCRAM CAUSED BY Main Turbine
Generator
GENERATOR EXCITER GROUND RELAY TRIPPING
AUTOMATIC SCRAM CAUSED BY PERSONNEL
RPS
ERROR DURING SURVEILLANCE TESTING.
Main Turbine TRIPPED ON LOW CONDENSER
Condenser
VACUUM CAUSED BY A FAILED POWER SUPPLY
TO BOTH LEVEL CONTROL LOOPS FOR THE OFF
GAS CONDENSER DRAIN VALVES. REPLACED
FAILED ELECTROYTIC CAPACITOR IN THE POWER
SUPPLY FOR THE OFFGAS CONDENSER DRAIN
VALVES.
EF
26094013
11/23/94
486
9
477
EF
26095002
11/23/94
486
78
408
HF
26095004
11/23/94
486
128
358
EF/WP
26095007
11/23/94
486
269
217
UNIT
ID
260
OUTG
DATE
4/15/94
260
12/2/94
20
08
F
3
A
260
2/9/95
30
13
F
3
H
260
3/30/95
69
29
F
3
H
260
8/19/95
32
1.3
F
3
A
WF
271
4/6/93
234
9.7
F
1
B
RB
CONROD MANUALLY SHUTTING DOWN DUE TO ALEAK ON
THE "B"FEEDWATER DISCHARGE HEADER
PIPING.
271
12/6/93
44
1.8
S
1
B
HA
VALVEX
271
12/9/93
119
50
F
1
A
HC
HTEXCH
271
12/17/93
80
33
S
1
A
HC
PIPEXX
271
2/9/94
6
02
F
1
B
EB
MANUAL SHUTDOWN TO REPAIR THE
CONDENSER EQUALIZING LINE.
ELECON DURING A ROUTINE INSPECTION, DISCOVERED
NEUTRAL GROUND ON THE MAIN GENERATOR
DISCONNECTED. GENERATOR WAS TAKEN OFF
LINE TO MAKE THE CONNECTION
271
4/10/94
51
2.1
F
3
A
HA
VALVEX
271
10/4/94
54
23
S
1
B
EB
271
10/15/94
48
2.0
F
1
B
271
12/8195
69
29
F
3
A
Source INEEL / NRC Morph 2 Data
JJ
OUTG
COMP
TIS
JX
FW
EF
4/20/92
495
351
144
Main Turbine
EF
1 10/25/93
508
42
466
Condenser
EF
10/25/93
508
45
463
Condenser
EF
10/25/93
508
53
455
Generator
HF/C
10/25/93
508
107
401
"C"MOISTURE SEPARATOR HIGH LEVEL. Main
Turbine TRIPPED AND A REACTOR SCRAM
REPLACED A FAULTY LEVEL CONTROLLER.
Main Turbine
EF
10/25/93
508
167
341
INSTRU
VITAL AC AUTO BUS TRANSFER PROBLEM.
REPAIRS MADE TO VOLTAGE REGULATOR
Electrical
EF
10/25/93
508
344
164
WB
VALVEX
COMBINATION OF SERVICE WATER LEAK ON THE RBCCW
HEAT EXCHANGER AND "B" RBCCW BYPASS
VALVE STUCK OPEN LINE ISOLATED, BLANKED
OFF.
EF
27194013
10/25/93
508
355
153
CH
VALVEX
TURBINE TRIP/REACTOR SCRAM DUE TO
MALFUNCTIONING FEEDWATER REGULATOR
VALVE
EF
27195021
5/3/95
492
219
273
MANUAL SHUTDOWN TO REPAIR "A" MOISTURE
SEPARATOR EMERGENCY DRAIN VALVE.
MANUAL SHUTDOWN TO REPLACE EXPANSION
JOINT ON THE "A"MAIN CONDENSER DUE TO
INCREASED AIR INLEAKAGE.
Page 100
FW
27194004
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
27 May 1998
OP
UNIT
ID
277
277
OUTG
DATE
12/18/92
1/2/93
OUTG
HRS
28
357
OUTG OUTG
DAYS
TYPE
12
F
149
S
OUTG OUTG OUTG
METH REASN SYSTEM
1
A
HA
1
H
OUTG
COMP
GENERA
277
277
3/2/93
4/24/93
134
81
5.6
34
F
F
3
1
A
A
HH
ID
PUMPXX
INSTRU
277
8/11/93
198
83
S
2
H
IE
INSTRU
277
5/14/94
121
50
F
3
A
CB
PUMPXX
278
278
278
7/4/92
7/14/92
10/15/92
219
194
609
9.1
8.1
25.4
F
F
F
3
2
3
A
A
A
EA
MB
SH
RELAYX
HTEXCH
VALVEX
278
278
12/19/92
3/7/93
12
119
0.5
50
F
F
1
3
H
A
HF
CH
HTEXCH
PUMPXX
278
7/4/93
293
122
S
2
A
RC
FUELXX
278
7/30/93
33
1.4
F
2
B
HC
INSTRU
278
12/2/93
327
13.6
F
1
A
SF
VALVEX
278
278
2/3/94
10/11/94
81
91
34
38
F
F
2
3
A
A
HA
CH
GENERA
GENERA
278
3/23/95
84
3.5
F
2
A
HC
VALVEX
278
7/30/95
68
28
F
3
A
HC
278
278
12/2/95
2/2/96
99
50
4.1
21
F
F
3
1
H
A
HA
HA
278
278
6/23/96
3/7/97
107
72
45
30
F
F
2
2
A
A
CD
CB
298
1/10/96
16
0.7
S
1
B
DESCRIP
REPAIR GENERATOR HYDROGEN LEAK.
MAINTENANCE OUTAGE TO REPAIR RECIRC PUMP
SEAL
SECOND CONDENSATE PUMP TRIP.
PLANT SHUTDOWN DUE TO REACTOR LEVEL
INSTRUMENT MISMATCH
MAINTENANCE OUTAGE FOR REACTOR WATER
LEVEL MODIFICATION.
APRM HI HIFLUX AUTOMATIC SCRAM DUE TO
RECIRC PUMP SPEED PROBLEMS.
AUTO SCRAM -#1 TRANSFORMER FAILURE
SJAE FLOW CONTROLLER FAILURE
PCIS GROUP I ISOLATION CAUSED BY BUMPING
INSTRUMENTATION.
CLEAN CONDENSER WATERBOXES.
REACTOR FEED PUMP TRIPPED, OTHER PUMP
FAILED TO START.
MAINTENANCE OUTAGE FOR REPLACEMENT OF
DEFECTIVE FUEL ASSEMBLIES REACTOR
MANUALLY SHUTDOWN TO 18% AND THEN
SCRAMMED FROM THERE
System
Generator
Recirc
Category
EF
EF
OUTG LER
Condensate
Reactor
EF
EF
27793004
27793010
Reactor
OP
DAYS
FROM
DAYS
START
FROM
CYCLE
CYCLE
OF
END OF
START LENGTH CYCLE CYCLE
12/9/92
647
9
638
12/9/92
647
24
623
12/9/92
12/9/92
647
647
83
136
564
511
EF
12/9/92
647
245
402
Recirc
EF
12/9/92
647
521
126
Electrical
Air Removal
Containment
EF
EF
HF/OA
27892010
27892005
27892008
1/9/92
1/9/92
1/9/92
618
618
618
177
187
281
441
431
337
Condenser
FW
EF/WP
EF
27883002
1/9/92
1/9/92
618
618
345
423
273
195
Reactor
EF
1/9/92
618
542
76
MANUAL SCRAM DUE TO RECOMBINER ISOLATION H2 Recombiner
AND SUBSEQUENT LOSS OF CONDENSER
VACUUM.
EF
1/9/92
618
568
50
LPCI MOTOR OPERATED VALVE MO-25A
RHR
INOPERABLE
MAIN GENERATOR FIELD GROUND RESISTOR.
Generator
AUTOMATIC SCRAM/HIGH REACTOR WATER
FW
LEVEL DUE TO FEED PUMP CONTROL PROBLEMS
CAUSE BY LOSS OF THE STATIC INVERTER
EF
11/15/93
676
17
659
11/15/93
11/15/93
676
676
80
330
596
346
MANUAL SCRAM, LOSS OF VACUUM DUE TO
STEAM SUPPLY VALVE FAILURE TO AIR
EJECTORS.
VALVEX FEEDWATER TRANSIENT, HIGH REACTOR LEVEL
SCRAM.
TURBIN AUTOMATIC SCRAM/TURBINE TRIP.
GENERA GENERATOR TAKEN OFF LINE FOR A MAIN
GENERATOR HYDROGEN LEAK
VALVEX REPAIR #2 TURBINE CONTROL VALVE STEM
MOTORX 'B' RECIRC PUMP MOTOR TRIP DURING
TROUBLESHOOTING OF 'A' RECIRC PUMP MOTOR
OIL LEVEL.
TURBINE GENERATOR TAKEN OFF LINE TO
EF
EF
27894005
Air Removal
EF
11/15/93
676
493
183
FW
EF
11/15/93
676
622
54
Main Turbine
Generator
EF
EF
1 10/17/95
718
46
672
10/17/95
718
108
610
Main Turbine
Reclrc
EF
HF
10/17/95
10/17/95
718
718
250
507
468
211
Main Turbine
EF
12/30/95
456
11
445
REPAIR TURBINE OIL SYSTEM.
Source INEEL / NRC Morph 2 Data
Page 101
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
OUTO
OUTG
DAYS
TYPE
95
S
OUTG
COMP
DAYS
DAYS
FROM
OP
OP
START
FROM
CYCLE
CYCLE
OF
END OF
START LENGTH CYCLE CYCLE
12/30/95
456
154
302
UNIT
ID
298
OUTO
DATE
6/1/96
OUTG
HRS
228
321
6/15/93
42
1.7
F
3
H
CH
VALVEX
AN AUTOMATIC REACTOR SCRAM OCCURRED
RPS
DUE TO A FALSE LOW REACTOR WATER LEVEL
SIGNAL THIS OCCURRED WHEN AN INSTRUMENT
LINE DEPRESSURIZED AFTER A PACKING NUT ON
A VALVE IN THE SENSING LINE DISENGAGED.
EF
32193012
5/16/93
493
30
463
321
10/22/93
91
38
F
2
A
HH
INSTRU
A SIMULTANEOUS TRIP OF THREE CONDENSATE
PUMPS CAUSED A DECREASE IN FEEDWATER
FLOW TO REACTOR VESSEL AND
CORRESPONDING DECREASE IN REACTOR
WATER LEVEL SHIFT INSERTED A MANUAL
REACTOR SCRAM ANTICIPATING AUTOMATIC
REACTOR SCRAM ON LOW REACTOR WATER
LEVEL
Condensate
EF
32193013
5/16/93
493
159
334
321
10/28/93
17
0.7
F
1
B
HA
PIPEXX
SHIFT REMOVED THE MAIN TURBINE FROM
SERVICE TO REPAIR A STEAM LEAK ON THE
ABOVE SEAT DRAIN FOR CONTROL VALVE NO 4
THE LEAK WAS REPAIRED
Main Turbine
EF
5/16/93
493
165
328
321
10/29/93
19
08
F
1
B
HH
VALVEX
SHIFT REMOVED THE MAIN TURBINE FROM
FW heating
SERVICE TO COMPLETE REPAIRS ON A STUCK
CHECK VALVE IN THE NORMAL DRAIN FROM THE
2ND STAGE OF THE C/D MSRS TO THE 5TH STAGE
"B" FEEDWATER HEATER. THE VALVE WAS
REPAIRED.
EF
5/16/93
493
166
327
321
11/11/93
64
2.7
F
2
B
HA
PIPEXX
SHIFT REMOVED THE MAIN GENERATOR FROM
SERVICE AND A MANUAL SCRAM WAS INITIATED
TO REPAIR AN EHC FLUID LEAK ON A ONE INCH
FLUID SUPPLY LINE THE LEAK WAS REPAIRED
AND THE UNIT WAS RETURNED TO SERVICE
EHC
EF
5/16/93
493
179
314
321
12/7/93
79
33
F
3
A
CH
CKTBRK
AN AUTOMATIC REACTOR SCRAM OCCURRED
DUE TO A LOW REACTOR WATER LEVEL SIGNAL.
THE LOW WATER LEVEL WAS CAUSED BY A TRIP
OF THE "A" RFP AND FAILURE OF THE REACTOR
RECIRCULATION PUMPS TO RUN BACK TO THE
NO 2 SPEED LIMITER.
Recirc
EF
5/16/93
493
205
288
321
12/26/93
62
26
F
2
A
HH
VALVEX
SHIFT REMOVED THE MAIN GENERATOR FROM
SERVICE, AND A MANUAL SCRAM WAS INITIATED
TO REPAIR 1N21-F253 THE VALVE WAS
REPAIRED AND THE UNIT RETURNED TO
SERVICE
5/16/93
493
224
269
Source INEEL / NRC Morph 2 Data
OUTG OUTG
OUTO
METH REASN SYSTEM
2
A
27 May 1998
DESCRIP
PLANT WAS SHUTDOWN FOR FUEL LEAKER
REPLACEMENT ONE LEAKING FUEL ASSEMBLY
WAS IDENTIFIED AND REPLACED NORMAL
POWER ASCENSION WAS IMPLEMENTED WITH
FULL POWER ACHIEVED ON 6/15/96.
Page 102
System
Reactor
Category
EF
EF
OUTG LER
32193016
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
UNIT
ID
321
OUTG
DATE
3/29/94
OUTG
HRS
84
OUTG OUTG
DAYS
TYPE
3.5
F
321
11/19/94
34
1.4
F
3
H
HJ
321
1/4/96
87
3.6
F
3
A
325
10/10/94
8
0.3
F
1
325
11/17/94
6
03
S
325
7/13/95
93
39
325
9/30/95
61
26
OUTG OUTG OUTG
METH REASN SYSTEM
3
A
HA
OUTG
COMP
GENERA
27 May 1998
DAYS
DAYS
FROM
START FROM
OP
OP
CYCLE CYCLE
OF
END OF
START LENGTH CYCLE CYCLE
5/16/93
493
317
176
DESCRIP
System
LOSS OF MAIN GENERATOR FIELD EXCITATION
Generator
LED TO LOAD REJECTION BY TURBINEGENERATOR SYSTEM AND RESULTED INAN
AUTOMATIC REACTOR SHUTDOWN. FIELD
EXCITATION WAS LOST WHEN ARCING
OCCURRED BETWEEN THE MGE BRUSH RIGGING
AND A COLLECTOR RING ON THE MGE ROTOR.
Category
EF
OUTG LER
32194003
VALVEX
AN AUTOMATIC REACTOR SCRAM OCCURRED
Main Turbine
FROM TURBINE STOP VALVE CLOSURE WHEN A
TURBINE TRIP SIGNAL WAS GENERATED DUE TO
HIGH WATER LEVEL INTHE MOISTURE
SEPARATOR REHEATER "A/B".
EF
32194014
1115/94
504
14
490
HB
FILTER
AN AUTOMATIC REACTOR SCRAM OCCURRED ON EHC
HIGH REACTOR PRESSURE WHEN ALL FOUR
MAIN TCVS DRIFTED CLOSED DUE TO THE
VALVES' SERVO STRAINERS BECOMING
CLOGGED, CAUSING LOSS OF HYDRAULIC FLUID
PRESSURE TO THE SERVO VALVE SPOOL.
EF
32196001
11/5/94
504
425
79
H
HA
TUBINE
MAIN TURBINE MANUALLY TAKEN OFF LINE DUE
TO EXCESSIVE VIBRATION AT THE NO. 3 MAIN
TURBINE BEARING THIS VIBRATION WAS
EXPERIENCED AS REACTOR POWER WAS BEING
REDUCED FOR THE ABOVE PLANNED
MAINTENANCE ACTIVITIES
EF
2/11/94
415
241
174
1
B
HA
TURBIN
MANUALLY TRIPPED MAIN TURBINE TO
Isophase bus
FACILITATE THE REMOVAL OF THE ISOPHASE BUS cooling
DUCT COVER INSPECTION PLATE.
OM
2/11/94
415
279
136
F
3
A
HA
TURBIN
EF
32595015
5/15/95
502
59
443
F
3
A
HH
PUMPXX
REACTOR SCRAM DUE TO AN ERRATIC
EHC
PRESSURE ERROR SIGNAL FROM EHC
REACTOR SCRAM DUE TO LOW VESSEL LEVEL
Condensate
CAUSED BY CONDENSATE SYSTEM TRANSIENT
THE TRANSIENT WAS A RESULT OF LOSS OF
CONDENSATE PUMP SUCTION PRESSURE
CAUSED BY EXCESSIVE AIR BEING ADMITTED TO
THE SUCTION HEADER OF THE CONDENSATE
EF
32595018
5/15/95
502
138
364
5/15/95
502
253
249
5/15/95
5/15/95
5/15/95
502
502
502
308
422
479
194
80
23
Main Turbine
PUMP.
325
1/23/96
46
1.9
F
2
A
HA
ZZZZZZ
MANUAL SCRAM ON HI #5 MAIN TURBINE BEARING Main Turbine
EF
VIBRATIONS
325
325
325
3/18/96
7/10/96
9/5/96
156
164
123
65
6.8
51
F
S
S
Source INEEL / NRC Morph 2 Data
2
2
2
A
H
H
WA
ZZ
U
PUMPXX
ZZZZ
ZZZZZZ
FORCED OUTAGE DUE SW PUMPS PROBLEMS
OUTAGE DUE TO HURRICANE BERTHA
OUTAGE DUE TO HURRICANE FRAN
Page 103
SW
Transmission
Transmission
EF
Nature
Nature
32596003
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
27 May 1998
1
UNIT
ID
331
_/
OUTG
DATE
8/17/92
OUTG
HRS
OUTG
DAYS
OUTG
TYPE
OUTG
COMP
FT
OUTG OUTG OUTG
METH REASN SYSTEM
System
THE PLANT TWICE SECURED THE GENERATOR
TO CEASE PREMATURE RECOMBINATION OF
HYDROGEN AND OXYGEN IN THE OFF GAS
SYSTEM.
81
/
I
331
1/24/93 1
1401
331
10/26/93
THE REACTOR SCRAMMED DUE TO A
MOMENTARY GROUND COMBINED WITH AN
EXISTING UNDETECTED ELECTRICAL GROUI
THE CONTROL CIRCUITRY FOR THE MAIN
TURBINE STEAM CONTROL VALVE
5/29/94
FATIGUE INDUCED WELD CRACK ON AN
ELECTROHYDRAULIC CONTROL OIL SUPPLY LINE
TO THE #2 TURBINE CONTROL VALVE A 0.5 GPM
HYDRAULIC LEAK WAS DISCOVERED DURING
OPERATOR ROUNDS REPAIR OF VARIOUS
BALANCE OF PLANT STEAM LINE VALVE PACKING
LEAKS
331 1 11/13/92
04
RCB
I
I
-~---'
IHIGH CONDENSER
BACKPRESSURE TURBINE
TRIP AND SCRAM CAUSED BY FAILURE OF
CIRCULATING WATER PUMP SHAFT FAILURE
ALLOWED THE FLOW FROM THE REMAINING
PUMP TO SHORT CYCLE BACK TO THE PUMP PIT
AND CUT OFF FLOW TO THE CONDENSER
~T~-----~
Category
AUTOMATIC REACTOR SCRAM CAUSED BY
PERCEIVED HIGH AVERAGE POWER RANGE
NEUTRON FLUX, CAUSED BY ELECTROMAGNETIC SIGNAL NOISE, WHICH REDUCED
FLOW BIASED SET-POINTS TO BELOW THE
CURRENT POWER LEVEL
1
9/3/92
F
DESCRIP
I
1
OUTG LER
33192013
ICirc water
i-------
Circ water
Main Turbine
/
DAYS
FROM
OP
START
OF
CYCLE
LENGTH CYCLE
DAYS
FROM
END OF
CYCLE
348
4/27/92
Offgas
OUTAGE TO RECONNECT CIRCULATION WA
PUMP.VERY COLD HIGH WINDS CAUSED ICIN
CIRCULATION WATER SPRAY CREATING
POTENTIAL FOR DAMAGE TO THE COOLING
TOWER FILL RESTART FOLLOWING THE 01/2
OUTAGE WAS DELAYED UNTIL THE WINDS
DECREASED.
OP
CYCLE
START
4/27/92
I
33192018
Nature
-- t----t---
4/27/92
460
200
260
10/12/93
500
271
229
10/12/93
500
400
100
4/27/92
33193010
10/12/93
10/12/93
331
7/10194
148
62
F
2
A
JI
TBG
CRACK IN FLUID SUPPLY LINE TO TURBINE
CONTROL VALVE ELECTRO-HYDRAULIC SYSTEM
REQUIRED SHUTDOWN TO REPLACE DAMAGED
SECTION OF TUBING AND INSTALLATION OF
HYDRAULIC ACCUMULATORS IN THE SUPPLY
LINE
331
11/16/94
352
14 7
S
2
B
BJ
ISV
SHUTDOWN FOR DRYWELL ENTRY TO VERIFY
Drywell
SOURCE OF AND REPAIR UNIDENTIFIED DRYWELL
EHC
EF
EF
33194010
LEAKAGE.
Source INEEL / NRC Morph 2 Data
Page 104
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
27 May 1998
DAYS
FROM
UNIT
ID
331
OUTG
DATE
5/14/95
OUTG
HRS
78
OUTG
OUTG
DAYS
TYPE
32
F
OUTG
OUTG OUTG
METH REASN SYSTEM
3
A
SL
OUTG
COMP
GR
DESCRIP
THE TRIP OF THE "B"RFP WAS DUE TO
STRIPPING THE INTERNAL GEARS OF THE
System
FW
Category
EF
LER
OUTGO
33195005
DAYS
FROM
START
OP
OP
CYCLE
OF
END OF
CYCLE
START LENGTH CYCLE CYCLE
25
516
4/19/95
541
COUPLING BETWEEN THE REACTOR FEED PUMP
SHAFT AND LUBE OIL PUMP.
6/1/95
FSF
HBA
333
2/25/93
333
4/21/93
333
5/19193
F
1
A
BJ
333
5/25/93
F
333
9124/93
3
3
A
A
IG
JI
210
88
XXXX
TUBE LEAK IN LOW PRESSURE CONDENSER.
SHUTDOWN TO DRAIN THE WATERBOXES,
IDENTIFY THE LEAKING TUBE, AND PLUG IT.
Condenser
4/19/95
SHUTDOWN DUE TO BLOCKAGE OF THE INTAKE
STRUCTURE SCHEDULED OUTAGE FOR "B"
RECIRC SEAL REPAIR. SHUTDOWN TO REPAIR
LEAK IN CHEMICAL DECON CONNECTION
Recirc
1/23/93
'---
33393009
SHUTDOWN DUE TO LOSS OF FEEDPUMP "A"
SPEED CONTROL DUE TO A SHORTED TERMINAL
STRIP. THE TERMINAL STRIP WAS REPLACED
SHUTDOWN DUE TO HPCI CHECK VALVE LEAK
CAUSED BY FAILED PRESSURE SEAL.
HPCI
JI
SHUTDOWN DUE TO "E" APRM UPSCALE TRIP.
NI/TIPs
94
DURING GROUND FAULT TESTING OF THE
'--'
43
677
33
644
1/23/93
1/23/93
HF/OA
33393013
1/23/93
33393020
1/23/93
33395010
3/26/95
33395013
3/26/95
33396002
3/26/95
33396010
3/26/95
TURBINE CONTROL SYSTEM, A BYPASS VALVE
ALARM/TRIP RELAY LEAD WAS MISTAKENLY
LIFTED CAUSING #2 BYPASS VALVE TO CLOSE
AND AREACTOR TRIP ON HIGH PRESSURE
5/30/95
333
9/5/95
333
2/22/96
178
F
3
D
AD
ISV
Recirc
A3/4" MANUAL VALVE (JET PUMP TO RECIRC
PUMP SUCTION) PACKING LEAK EXCEEDED TECH
SPEC LIMITS. THE PACKING WAS REPLACED
WITH ADIFFERENT STYLE THAT IS LESS PRONE
TO GROSS FAILURE
F
3
G
JB
SC
AN INADVERTENT REMOVAL OF AFEEDWATER
CONTROL FUSE CAUSED A FEEDPUMP
TRANSIENT AND PLANT SCRAM ON LOW WATER
LEVEL
TBG
HF/OA
WHILE PERFORMING ACONTROLLED REACTOR
SHUTDOWN DUE TO EXCESSIVE SCRAM TIME, AN
EHC LINE TO TURBINE BYPASS VALVES
RUPTURED OPERATORS INSERTED A MANUAL
SCRAM EHC TUBING WAS MODIFIED WITH
FLEXIBLE TUBING AND SCRAM SOLENOID PILOT
VALVE DIAPHRAGMS REPLACED.
333
9/16/96L
1771
7.41
F
13
H
I
EL
DURING PERFORMANCE OF 345KV RELAY
CALIBRATION TWO TERMINALS WERE
INADVERTENTLY SHORTENED CAUSING THE
10042 AND 10052 BREAKERS TO OPEN LEADING
TO A SCRAM WORK PROCESS IS BEING
REVIEWED FOR IMPACT ON PLANT OPERATIONS
Electrical
HF/PI
''
Source INEEL / NRC Morph 2 Data
Page 105
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
OUTG
OUTG OUTG
TYPE
HRS
DAYS
57
2.4
F
OUTG OUTG OUTG
METH REASN SYSTEM
2
G
SF
UNIT
ID
341
OUTG
DATE
11/18192
341
341
12/5/92
2/10/93
210
82
8 7
34
S
F
1
1
A
B
SE
SG
341
2/19/93
32
13
F
3
G
KE
341
341
4/10/93
4/20/93
240
33
10.0
14
S
F
1
9
A
H
SE
341
8/13/93
49
2.0
F
3
G
JE
341
9/17/93
70
29
S
2
B
341
9/20/93
22
09
F
1
B
Source INEEL / NRC Morph 2 Data
OUTG
COMP
ISV
EXJ
COND
System
DESCRIP
SCRAM DUE TO LOSS OF HEATER FEEDWATER
Condensate
PUMPS. INADVERTENT OPENING OF
CONDENSATE DEMINERALIZER INLET VALVE
RESULTED IN LOSS OF HEATER FEED PUMP (HFP)
AUCTION PRESSURE AND CONSEQUENT TRIP OF
HFP
Category
HF/OA
REPAIR OF EXTRACTION STEAM LINE RUPTURE.
CONDENSER TUBE LEAK CAUSED CONDENSATE
CHEMISTRY TO REACH ACTION LEVEL PLANT
SHUTDOWN FOR TUBE PLUGGING
EF
EF
FW heating
Condenser
OUTG LER
34192012
DAYS
DAYS
FROM
START FROM
OP
OP
OF
END OF
CYCLE CYCLE
START LENGTH CYCLE CYCLE
111/7/92 492
11
481
11/7/92
11/7/92
492
492
28
95
464
397
Electrical
ROUTINE PUMP BREAKER PM TESTING
INADVERTENTLY ACTUATED IN-SERVICE TRIP
RELAYS AN IN-SERVICE PUMP BREAKER TRIP
RELAY FAILED TO PROPERLY ACTUATE, LEADING
TO INABILITY TO TRANSFER FEED TO ALTERNATE
SUPPLY.
EF
34193004
11/7/92
492
104
388
REPAIR OF EXTRACTION STEAM LINE RUPTURE. FW heating
SCRAM OCCURRED DURING RECOVERY FROM
Main Steam
EXTRACTION STEAM LINE REPAIR OUTAGE.
SCRAM CAUSED BY INCORRECTLY INSTALLED
TEST INSTRUMENT WHICH LEAKED STEAM AND
WATER ONTO MAIN STEAM MANIFOLD PRESSURE
TRANSMITTERS
EF
HF/OA
34193007
11/7/92
11/7/92
492
492
154
164
338
328
TV
THE REACTOR SCRAM WAS AUTOMATICALLY
INITIATED BY A TRIP OF THE MAIN TURBINE DUE
TO A FALSE HIGH REACTOR WATER LEVEL
SIGNAL
RPS
EF
34193010
11/7/92
492
279
213
SN
LCV
WHILE SHUTTING DOWN TO REPAIR A HEATER
DRAIN SYSTEM LEVEL CONTROL VALVE,
PRESSURE INTEGRITY WAS LOST DUE TO
MAINTENANCE ACTIVITIES ON THE VALVE. THIS
RESULTED IN LEAKAGE FROM THE FEEDWATER
SYSTEM WHICH WAS TERMINATED AFTER THE
REACTOR WAS SCRAMMED
FW heating
HF
34193013
11/7/92
492
314
178
TA
52
11/7/92
492
317
175
EXJ
DURING STARTUP, THE MAIN TURBINE TURNING Main Turbine
GEAR CIRCUIT BREAKER FAILED. THE REACTOR
WAS SHUTDOWN TO MINIMIZE DIFFERENTIAL
HEATING OF THE TURBINE SHAFT DURING THE
TIME THAT TURNING GEAR WAS OUT OF SERVICE
TURNING GEAR BREAKER WAS REPLACED
Page 106
EF
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
UNIT
ID
OUTG
DATE
OUTG
HRS
OUTG
DAYS
OUTG
TYPE
OUTG OUTG OUTG
METH REASN SYSTEM
OUTG
COMP
27 May 1998
DESCRIP
System
Category
OUTG LER
TiT4
DAYS
FROM
DAYS
OP
OP
START FROM
CYCLE CYCLE
OF
END OF
START LENGTH CYCLE CYCLE
11/7/92
492
414
78
341
1/27/95
387
16.1
S
1
B
TA
TRB
TURBINE TAKEN OFF LINE TO PERFORM POST
OUTAGE BALANCING. TURBINE REMAINED OFF
LINE TO REPAIR #4 JACKING OIL PUMP
DISCHARGE PIPING.
Main Turbine
EF
1/18/95
618
10
608
341
2/1/195
259
10.8
F
9
B
TD
PSP
Main Turbine
EF
1/18/95
618
14
604
341
2/13/95
663
27.6
F
4
B
TD
PSP
Main Turbine
EF
1/18/95
618
27
591
341
3/14/95
18
0.8
S
1
B
TA
TRB
Main Turbine
EF
1/18/95
618
55
563
341
3/16/95
21
0.9
S
1
B
TA
TRB
Main Turbine
EF
1/18/95
618
57
561
341
3/26/95
48
2.0
F
1
B
TJ
PSP
SWC
EF
1/18/95
618
67
551
341
4/9/95
62
2.6
S
2
B
TA
TRB
TURBINE REMAINED OFF LINE TO REPAIR #4
JACKING OIL PUMP DISCHARGE PIPING.
TURBINE TAKEN OFF LINE TO REPAIR TURBINE
JACKING OIL SYSTEM STRUCTURAL CONCERNS.
TURBINE TAKEN OFF LINE TO PERFORM POST
OUTAGE BALANCING.
TURBINE TAKEN OFF LINE TO OBTAIN TURBINE
COASTDOWN BEARING VIBRATION DATA.
TURBINE TAKEN OFF LINE TO REPAIR A STATOR
COOLING WATER VENT LINE LEAK.
MANUAL REACTOR/TURBINE TRIP PER SOE 95-10
TO OBTAIN HOT TURBINE COASTDOWN
VIBRATION DATA AT APPROXIMATELY 80%
POWER.
Main Turbine
EF
1/18/95
618
81
537
341
4/12/95
42
1.7
F
1
B
SB
ISV
EF
1/18/95
618
84
534
341
4/25/95
249
10.4
F
3
A
JJ
RG
TURBINE TAKEN OFF LINE TO REPAIR N3018F607, Main Turbine
MAIN STEAM TO MSR ISOLATION VALVE.
AUTOMATIC REACTOR SCRAM ON APRM
RPS
NEUTRON UPSCALE TRIP RESULTING FROM
REACTOR PRESSURE REGULATOR TRANSIENT.
EF
34195005
1/18/95
618
98
521
341
6/2/95
334
13.9
F
3
B
TA
SIS
AUTOMATIC MAIN TURBINE TRIP ON MECHANICAL Main Turbine
OVERSPEED TRIP RING #2 WHILE PERFORMING
MTG OVERSPEED TRIP TEST
EF
34195006
1/18/95
618
135
483
341
3/27/96
526
21.9
F
2
B
BI
TK
TECH SPEC REQUIRED SHUTDOWN DUE TO
BOTH DIVISIONS OF EECW BEING DECLARED
INOPERABLE DUE TO MAKE-UP TANK DESIGN
ISSUE. MODIFICATION BEING INSTALLED.
ESW
EF/WD
34196005
1/18/95
618
434
184
341
4/19/96
108
4.5
F
2
B
BJ
PC
DURING UNIT STARTUP HPCI AND RCIC
DECLARED INOPERABLE, TECH SPEC REQUIRED
SHUTDOWN.
HPCI
EF
34196007
1/18/95
618
457
161
Source INEEL / NRC Morph 2 Data
Page 107
34195004
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
UNIT
ID
UNIT
ID
OUTG
DATE
9/7/93
352
1/14/94
57
352
10/8/94
12
352
2/21/95
38
352
5/7/95
352
8/20/95
186
352
8/28/95
111
352
9/2/95
352
9/11/95
353
3/17/93
353
3/26/93
353
5/15/93
OUTG
DAYS
OUTG
HRS
OUTG
TYPE
OUTG OUTG OUTG
METH REASN SYSTEM
OUTG
COMP
CKTBRK
7
27 May 1998
GENERA REACTOR MANUAL SCRAM OCCURRED DUE TO A
LOSS OF STATOR WATER COOLING.
TURBIN TURBINE TAKEN OFF LINE DUE TO HIGH TURBINE
VIBRATION
CONROD REACTOR SCRAM DUE TO ELECTRICAL
DISTURBANCE AT WHITPAIN SUBSTATION
HTEXCH REACTOR WAS SHUTDOWN TO PERFORM
MAINTENANCE ON "C"DRAIN COOLER, "A"
RECIRCULATION PUMP SEAL, AND THE
CONDENSER WATERBOXES
-" I
0.6
14.1
338
02
5
8
03
F
~-7
RC
FUELXX
PIPEXX
1
2
A
1
1
F
3
F
1
SE
A
CC
I
1
577
554
23
Main Turbine
3/11/94
697
211
486
Transmission
3/11/94
Recirc
3/11/94
Reactor
3/11/94
VALVEX
Main Steam
REACTOR WAS SHUTDOWN DUE TO A FAILED
OPEN SAFETY RELIEF VALVE (SRV)
THE MAIN TURBINE TRIPPED OFF LINE DUE TO
SWC
LOW FLOW INTHE STATOR WATER COOLING
SYSTEM.
THE UNIT AUTOMATICALLY SHUTDOWN DUE TO
AIR ENTRAINED INTHE MAIN TURBINE ELECTRO
HYDRAULIC CONTROL SYSTEM.
REACTOR POWER WAS REDUCED TO 19% AND
THE TURBINE TAKEN OFF LINE TO REPAIR A LEAK
IN THE ELECTRO-HYDRAULIC CONTROL SYSTEM
PIPEXX
EF
35294001
Drywell
H2 Recombiner
UNIT SHUTDOWN INACCORDANCE WITH TECH
SPEC 3 0.3 AS A RESULT OF DISCOVERING BOTH
POST-LOCA HYDROGEN RECOMBINER SYSTEMS
WERE INOPERABLE DUE TO IMPROPER WIRING
OF CERTAIN RECORDERS DURING A RECENT
RECORDER MODIFICATION.
TURBIN
HA
REACTOR WAS SHUTDOWN TO REPLACE A
FAILED FUEL BUNDLE
REACTOR WAS SHUTDOWN SHORTLY AFTER
BEING CRITICAL DUE TO LEAKAGE INTO THE
DRYWELL CAUSED BY A MISALIGNED REACTOR
PRESSURE VESSEL INSTRUMENT FLANGE
CONNECTION.
OUTG LER
7/9/92
SWC
RECOMB
HTEXCH
Category
DESCRIP
System
AUTOMATIC REACTOR SCRAM OCCURRED AFTER Electrical
LOSS OF AN OFFSITE POWER SUPPLY DURING
THE AUTOMATIC TRANSFER TO THE SECONDARY
POWER SUPPLY, BREAKER FOR THE 1A
FEEDWATER CONTROL SYSTEM FAILED TO
RECLOSE, RESULTING INA REDUCTION IN
REACTOR WATER LEVEL
OP
CYCLE
START
7/9/92
DAYS
FROM
DAYS
OP
START FROM
CYCLE
OF
END OF
LENGTH CYCLE CYCLE
I
Ii_:
HF/PI
35295006
3/11/94
35295007
3/11/94
35295008
3/11/94
350
-"-'--'
148
3/16/93
I
I
35393005
3/16/93
i
1
i
3/16/93
i
Source INEEL / NRC Morph 2 Data
Page 108
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
UNIT
ID
OUTG
DATE
10/19/94
353
2/21/95
353
6/3/95
353
OUTG
HRS
OUTG
DAYS
OUTG
TYPE
OUTG
OUTG OUTG
METH REASN SYSTEM
3
3
RB
33
F
6
02
S
8/8/95
70
29
353
8/20/95
60
25
F
353
11/22/95
111
05
F
1
353
5/1/96
191
081
F
1
353
5/2/96
353
5/14/96
57
353
10/6/96
5
F
353
12/6/96
185
F
2
A
353
12/18/96
115
48
F
2
A
353
12/24/96
31
13
F
354
-12/3/92
52
22
F
354
5/16/93
00
42
F
HA
A
HF
OUTG
COMP
INSTRU
------ REACTOR SCRAM DUE TO ELECTRICAL
CONROD
DISTURBANCE AT WHITPAIN SUBSTATION.
1
H
ZZ
HC
MAIN TURBINE REMOVED FROM SERVICE TO
REPAIR EHC LEAK AT #4 CIV.
INSTRU
REACTOR WAS SHUTDOWN DUE TO A FAILED
POWER SUPPLY IN THE FEED WATER CONTROL
SYSTEM.
INSTRU
REACTOR WAS SHUTDOWN DUE TO A HIGH
IMPEDANCE ACROSS THE EHC CONTROL RELAY
CONTACT RESULTING IN SPORADIC OPENING
AND CLOSING TURBINE BYPASS VALVES
EHC
FILTER
THE GENERATOR WAS TAKEN OFF THE GRID TO
REPLACE THE STATOR WATER COOLING
FILTERS
SWC
VALVEX
THE TURBINE WAS TAKEN OFF THE GRID DUE TO IEHC
REPAIR OF TURBINE EHC LEAK.
Main Turbine
TURBINE WAS TAKEN OFF THE GRID DUE TO A
FAILED BACKUP OVERSPEED TRIP TEST
OUTG LER
.
___.__,
12/1/93
1321
55
F
2
A
I
35395008
2/20/95
35395010
2/20/95
---
REACTOR SCRAM DUE TO TURBINE TRIP CAUSED Transmission
BY GRID INSTABILITY.
FILTER
POWER REDUCTION DUE TO REPLACEMENT OF
THE GENERATOR STATOR COOLING "Y"
STRAINER
INSTRU
REACTOR WAS SHUTDOWN DUE TO A FAILED
PRESSURE SWITCH ON THE EHC SYSTEM.
HTEXCH
REACTOR WAS SHUTDOWN DUE TO CRACK IN
THE MAIN CONDENSER NECK SEAL GASKET
Condenser
MECFUN
REACTOR WAS SHUTDOWN DUE TO A FAILED
SCOOP TUBE POSITIONER ON THE 2B MG SET
Recirc
CONTRACT EMPLOYEE BUMPED CART INTO A
MCC, CAUSING REACTOR RECIRCULATION PUMP
M/G SET VENT FANS TO TRIP, RESULTING IN A
DOUBLE RECIRCULATION PUMP TRIP CONTROL
OPERATOR MANUALLY SCRAMMED THE
REACTOR
Recirc
181
Source INEEL / NRC Morph 2 Data
,
1
EF
2/20/951
7111
AI
Page 109
I
97F
2/20/95
35396004
-
35396007
UNIT SHUTDOWN DUE TO EXCESSIVE ARCING OF Generator
THE MAIN GENERATOR EXCITER BRUSHES.
530
2/20/95
2/20/95
449
262
2/20/95
594
117
2/20/95
2/20/95
HF/OA
1UNIT TRIPPED DUE TO FAULTY COMPONENT IN
1
. .
r
2/20/95
35396009
2/20/95
35492013
11/10/92
673
11/10/92
482
11/10/92
482
2951
EHC.
3541
DAYS
FROM
END OF
CYCLE
2/20/95
ZZZZZZ
A1
3
Category
... .
i
HF/OA
Transmission
VALVEX
INSTRU
13
System
DESCRIP
. . . . . . . . . . . .. . .. .. .. . . . . . ..
Diesel
REACTOR SCRAM ASSOCIATED WITH A RELAY
COIL FAILURE COMBINED WITH AN
INAPPROPRIATE ACTION TAKEN BY AN
OPERATOR DURING TESTING OF AN EMERGENCY
DIESEL GENERATOR.
OP
OP
CYCLE
CYCLE
START LENGTH
. . . .. ..
3/16/93
DAYS
FROM
START
OF
CYCLE
1EF
386
96
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
27 May 1998
OP
OP
UNIT
ID
354
OUTG
DATE
5/15/94
OUTO
HRS
153
354
6/21/94
16
0.7
F
1
A
354
8/1/94
63
2.6
F
3
A
354
8/30/94
45
1.9
F
3
H
354
10/2/94
238
9.9
F
3
A
354
3/20/95
181
7.5
F
2
354
7/8/95
418
17.4
F
366
11/24/92
45
1.9
366
11/27/92
49
366
12/6/92
185
$50
$14tj$x":
OUTO OUTG
DAYS
TYPE
6.4
F
OUTG OUTG OUTG
METH REASN SYSTEM
3
B
OUTG
COMP
DESCRIP
System
AUTOMATIC SCRAM DURING DIGITAL FEEDWATER FW
TESTING.
POWER REDUCTION TO REPAIR EHC LEAK ON #2 EHC
BYPASS VALVE.
AUTOMATIC SCRAM DURING IRM SURVEILLANCE NI/TIPs
DUE TO FAULTY TEST EQUIPMENT.
AUTOMATIC SCRAM CAUSED BY GENERATOR
SWC
RUNBACK DUE TO LOSS OF STATOR WATER
COOLING.
Category
OUTG LER
DAYS
FROM
DAYS
START
FROM
CYCLE
CYCLE
OF
END OF
START LENGTH CYCLE CYCLE
4/27/94
564
18
546
EF
4/27/94
564
55
509
EF
4/27/94
564
96
468
EF
4/27/94
564
125
439
AUTOMATIC SCRAM CAUSED BY DESIGN ERROR FW
IN DIGITAL FEEDWATER CONTROL SYSTEM.
WHEN ATTEMPTING RESTART AN EHC SYSTEM
PROBLEM CAUSE A PROBLEM WITH THE TURBINE
ROLL. OPERATOR CLOSED ALL TURBINE VALVES
WHICH RESULTED INAUTO REACTOR SCRAM.
EF/WD
4/27/94
564
158
406
G
WHILE I&CTECHS WERE PERFORMING A PM ON
THE OPTICAL ISOLATOR FOR THE REACTOR
RECIRCULATION PUMP MG SETS, A LOSS OF
BOTH MG SETS OCCURRED. A MANUAL SCRAM
WAS INITIATED INACCORDANCE WITH THE
PROCEDURE.
HF/OA
4/27/94
564
327
237
1
A
WHEN LCO 3.7.2.A FOR CONTROL ROOM
MCR HVAC
VENTILATION ACTION STATEMENT EXPIRED A
UNIT SHUTDOWN WAS INITIATED. CAUSE WAS A
MOMENTARY INTERRUPTION TO THE CONTROL
CIRCUIT COMBINED WITH LENGTHY CABLE RUNS.
EF
4/27/94
564
437
127
F
1
B
HA
VALVOP
THE MAIN TURBINE WAS TAKEN OFF LINE TO
REPAIR AN ELECTRO-HYDRAULIC CONTROL
SYSTEM FLUID LEAK AT THE REHEAT CYLINDER
ON COMBINED INTERMEDIATE VALVE NO. 4.
EHC
EF
11/21/92
480
3
477
2.0
F
3
H
HA
TURBIN
AN AUTOMATIC REACTOR SCRAM OCCURRED
WHEN VIBRATION AT THE NO. 6 TURBINE
BEARING REACHED APPROXIMATELY 12 MILS.
Main Turbine
EF
11/21/92
480
6
474
7.7
S
1
B
HJ
GENERA THE UNIT WAS SHUTDOWN DUE TO AHYDROGEN Generator
LEAK AT THE NEUTRAL BUSHING ON THE MAIN
GENERATOR. THE BUSHING WAS REPLACED AND
TESTED FOR LEAKAGE, AND THE UNIT WAS
RETURNED TO RATED THERMAL POWER.
11/21/92
480
15
465
11/21/92
480
103
377
t
L
l
W
T
:lin-~iii~::I~i
..~ ~lii::::i
..........
a::r~a"~i
f" E~I~ac"5~s,
..........
.............
::::~il~i::~lii~i:
:i:~:::
~ac
...........
.. .. .. ..
..
..........
~ ::::ir~~!f:8:
::~i~lii~~:~l
..
..........
~::: :::
jw:::::i~ ~ g~~~: ::::i~
:::::ji.~ii.i.
:~:1::ia:::::
.....
Source INEEL / NRC Morph 2 Data
.......
Page 110
36692026
EF
Ar
f
:~'~L-~'t P.
"!*' ""' "r~lP:Ill~
-~i~~a?
~,.1
~ -Ill"*'il~i".
Recirc
~
::B:ii:
.....
.........
.....
.......
ilbr:
~-:i;~ ja4~isr
...
,iih
.. ....
.
.i"'a,~.~:'~i81~libr~
.......
~j~Sli
:~~::::
:~::i:::lII:: .....
::)
j~iiii:~~iiii
~ii~ii~
"~~saiai
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
OUTG
OUTG
DAYS
TYPE
1.4
F
OUTG
OUTG OUTG
METH REASN SYSTEM
2
G
ZZ
OUTG
COMP
UNIT
ID
366
OUTG
DATE
5/21/93
OUTG
HRS
34
366
11/8/93
183
76
F
2
B
SF
VALVEX
SHIFT REMOVED THE MAIN GENERATOR FROM
Core Spray
SERVICE, AND A MANUAL SCRAM WAS INITIATED
TO INVESTIGATE INCREASED LEAKAGE INTO THE
DRYWELL FLOOR DRAIN SYSTEM
INVESTIGATION REVEALED A BONNET PRESSURE
SEAL LEAK ON CORE SPRAY TESTABLE CHECK
VALVE 2E21-F006B
EF
366
8/30/94
52
2.2
F
3
H
IA
INSTRU
AUTOMATIC REACTOR SCRAM WHEN RPS
ELECTRICAL BUS 2A WAS BEING TRANSFERRED
FROM ITS ALTERNATE TO ITS NORMAL SUPPLY
THE EVENT WAS CAUSED BY INADVERTENTLY
MOVING THE SWITCH BEYOND ITS CENTER
POSITION WHEN TRANSFERRING FROM "ALT A"
TO THE "NORM" POSITION.
RPS
HF/OA
366
4/11/95
271
113
F
3
G
CG
INSTRU
AN AUTOMATIC REACTOR SCRAM OCCURRED
DUE TO THE IMPROPER PLACEMENT OF A
JUMPER WHILE ATTEMPTING TO RETURN THE
REACTOR WATER CLEANUP SYSTEM TO
SERVICE
RPS
HF/OA
366
5/4/95
120
5.0
F
2
A
HA
GENERA
THE UNIT WAS MANUALLY SCRAMMED TO REPAIR Main Turbine
THE NO 9 AND NO. 10 BEARINGS ON THE MAIN
TURBINE GENERATOR. THE NO 9 AND NO 10
BEARINGS AND JOURNALS WERE REPAIRED
366
9/2/95
57
2.4
F
2
A
HF
HTEXCH
UNIT WAS MANUALLY SCRAMMED DUE TO A
DECREASING VACUUM ON THE "A" MAIN
CONDENSER AS A RESULT OF "D" WATERBOX
BECOMING AIRBOUND AFTER FILL MATERIAL IN
CELL 10 OF COOLING TOWER NO 5 COLLAPSED
AND CLOGGED THE SCREENS AT THE TOWER
366
11/21/95
3
01
F
2
G
HJ
HTEXCH
Main Turbine
MAIN TURBINE TRIPPED ON MOISTURE
SEPARATOR REHEATER HIGH LEVEL A MOTOR
OPERATED VALVE IN THE HIGH LEVEL DRAIN LINE
WAS FOUND CLOSED.
366
3/12/96
83
3.4
S
2
A
CC
VALVEX
THE UNIT WAS MANUALLY SCRAMMED TO
REPLACE MAIN STEAM LINE SRVS "D" AND "H".
366
4/25/96
62
2.6
F
1
A
HH
PIPEXX
387
5/17/92
8
03
F
1
B
TA
TRB
Source INEEL / NRC Morph 2 Data
DESCRIP
A MANUAL REACTOR SCRAM WAS INITIATED
WHEN BOTH REACTOR RECIRCULATION PUMPS
TRIPPED.
System
Recirc
Category
OUTG LER
DAYS
FROM
END OF
CYCLE
480
181
299
11/21/92
480
352
128
36694007
4/30/94
512
122
390
36695001
4/30/94
512
346
166
4/30/94
512
369
143
4/30/94
512
490
22
EF
11/21/95
479
0
479
Main Steam
EF
11/21/95
479
112
367
SHIFT MANUALLY TRIPPED THE MAIN TURBINE
AND INSERTED A MANUAL SCRAM TO REPAIR A
LEAK ON A REACTOR FEEDWATER VENT LINE
FW
EF
11/21/95
479
156
323
UNIT ONE TOOK THE GENERATOR OFF LINE AT
0327 HOURS MAY 17TH DUE TO HIGH VIBRATION
ON THE #5 BEARING OF THE MAIN TURBINE
Main Turbine
EF
5/16/92
498
1
497
Page 111
36693005
OP
CYCLE
START
11/21/92
Circ water
EF
DAYS
FROM
OP
START
OF
CYCLE
LENGTH CYCLE
EF
EF
36695003
MIT POC DL BRODEUR
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
UNIT
ID
387
OUTG
DATE
6/5/92
387
11/12/92
712.
387
OUTG
OUTG
OUTG
HRS
DAYS
TYPE
443
18.5
F
57
2.4
F
OUTG OUTG OUTG
METH REASN SYSTEM
2
B
SJ
3
A
1.2
...........
~~i~~i ~~~52~~i~~sajlr
.....
OUTG
COMP
V
SJ
RLY
TA
..
DESCRIP
THE UNIT WAS TAKEN OFF LINE TO REPAIR THE
"A"REACTOR FEED PUMP ISOLATION VALVES.
THE VALVES WERE REPAIRED AND A STARTUP
COMMENCED. THE STARTUP WAS HALTED AND
THE UNIT MANUALLY SCRAMMED DUE TO AN
IGNITION OF CHARCOAL INTHE 1B OFFGAS
GUARD BED.
System
FW
THE SCRAM WAS CAUSED BY A FAULTY RELAY IN FW
ONE DIVISION OF THE RFP TURBINE, MAIN
TURBINE HI LEVEL TRIP CIRCUIT WHILE A
SURVEILLANCE WAS BEING PERFORMED IN THE
OTHER DIVISION OF THE HI LEVEL TRIP LOGIC.
N.I PI
...
ANAUTM
N
~~axi'iS~~iCti
~
....
.........
:~~::tei:~;~:~~:~IB;;::l~SI::i
ii ~ Zi3jiiiiiiiiiiiii~iH
~ ~ ::-:::~:::w~:~ri::~..~:~~j::~:
:::~:
.ss~:p-~
~ av:::.....
":::
,i.....
;Ip~~;~iiB'"i~gP~......
1/105
66
253
S
2
B
j
GE
11/10/95
606
25.3
S
2
B
TJ
GEN
387
8/1/96
103
4.3
F
3
H
TA
388
1/29/93
186
7.7
F
1
A
388
12/10/93
480
20.0
F
1
A
Category
EF
OUTO LER
38792010
EF
38792017
I.............
T..I..
INSTRUMENTATION.
j~i~i~IX~~~2
387
Source INEEL / NRC Morph 2 Data
27 May 1998
l:~ij
:iiiXiiiiiijii
....
...
.....
..
:: ...
OUAGE O
RPAIRA
HYROGN
STAOR
ATE
COLIN
SYTEMLEK ITO TE SW
UNEXECTE
MAI GENRATO
OUTAGE TO REPAIR
A HYDROGENBARTIE
LEAK INTO THE SWC
STATOR WATER COOLING SYSTEM.
UNEXPECTED MAIN GENERATOR BAR TIE,
SPACER DAMAGE AND WEDGE LOOSENESS WAS
IDENTIFIED WHICH EXTENDED THE OUTAGE
DURATION.
EF5/6/5
VIS
TURBINE TRIP WAS CAUSED BY A FALSE
SPURIOUS SIGNAL FROM TURBINE #1 BEARING
VIBRATION INSTRUMENTATION LOOP.
COMPONENTS WERE REPLACED THAT MOST
LIKELY CONTRIBUTED TO THIS SPURIOUS
SIGNAL.
EF
TL
EXC
POWER REDUCED TO INVESTIGATE AND REPAIR Condenser
CONDENSER TUBE LEAKS. CONDENSER
DEMINERALIZER INFLUENT (CDI) CONDUCTIVITY
EXCEEDED ADMINISTRATIVE LIMITS. MAIN
TURBINE/GENERATOR WAS MANUALLY TRIPPED.
FORCED OUTAGE FOR GENERATOR EXCITER
FIELD GROUND.
XX
ZZZ
MANUALLY SHUTDOWN DUE TO HIGH DRYWELL Recirc
LEAKAGE. INSPECTION OF DRYWELL REVEALED A
CRACKED WELD ON THE "A"RX RECIRC PUMP
RBCCW OUTLET LINE. OTHER WORK INCLUDED
INSTALLATION OF TORQUE COLLARS ON THE
MAIN TURBINE AND INSTALLATION OF RX LEVEL
Page 112
Main Turbine
~I
DAYS
FROM
DAYS
OP
OP
START FROM
CYCLE CYCLE
OF
END OF
START LENGTH CYCLE CYCLE
5/16/92
498
20
478
5/16/92
498
180
318
5/16/92
498
422
76
5/6/2AR..
22
7
42
18
30
5/6/95
492
188
304
5/6/95
492
453
39
EF
11/14/92
518
76
442
EF
11/14/92
518
391
127
EF
38796006
MIT POC DL BRODEUR
27 May 1998
BWR/4 Forced Outage Data, part 2 events within complete operating cycles 1989 - 1997
,
UNIT
ID
388
OUTG
DATE
1/20/94
1998
OUTG OUTO
OUTG
METH REASN SYSTEM
3
A
TJ
OUTG
OUTG
OUTG
TYPE
HRS
DAYS
2.01
F
491
OUTG
COMP
TCV
4
388
2/20/94
388
6/12/94
388
4/15/95
388
7/14/96i
-
1
A
TG
4
1501
i
631
3
F
i
I
Source INEEL / NRC Morph 2 Data
i
A
-- I- ----- -----
FK
DESCRIP
SWC
AUTOMATIC MAIN TURBINE TRIP WITH
AUTOMATIC REACTOR SCRAM AS A RESULT OF A
STATOR COOLING WATER TEMPERATURE
CONTROL VALVE PROBLEM WHICH CAUSED HIGH
STATOR COOLING WATER TEMPERATURES
System
Svstem
Category
6/10/94
458
V
REMOVED THE MAIN TURBINE FROM SERVICE TO
REPAIR A STEAM LEAK ON FW HEATER BTV20210B
FW heating
AUTOMATIC REACTOR SCRAM DUE TO A MAIN
GENERATOR LOAD REJECT. REPLACED "A"
REACTOR RECIRC PUMP SEAL AND "B"MAIN
TRANSFORMER BUSHING.
Generator
38895005
6/09
REACTOR SCRAMMED WHEN ALL FEEDWATER
WAS LOST DURING POST MAINTENANCE
TESTING OF TIE BUS 0A107, POWER WAS LOST
TO AUXILIARY BUS 12A THIS CAUSED 2
CONDENSATE PUMPS TO TRIP AND DUE TO LOW
SUCTION PRESSURE ALL 3 RFPS TRIPPED
Electncal
38896004
10/21/95
CBL1
CYCLE
55
THE MAIN GENERATOR WAS TAKEN OFF LINE TO
REPAIR EHC LEAK ON THE #3 CONTROL VALVE.
38895005
CYCLE
11/14/92
V
BKR
I
OP
OP
CYCLE
LENGTH
START
11/14/92
CYCLE
OUTG LER
38894002
May
DAYS
27
DAYS
FROM
START
FROM
END OF
OF
151
6/10/94
510
267
243
______________________________________
Page 113
MIT POC DL BRODEUR
Appendix 4. NRC monthly report, glossary of terms
Morpl .dbf
Last three digits of the facility's docket number (05000nnn).
DOCKET
Last three digits of the facility's docket number (05000nnn).
DOCKETA
RPT PERIOD Month and year of reporting period. (MMYY)
UTL CONTCT Utility contact concerning monthly operating data.
FIRST NAME Contact's first name.
UTL PHONE Contact's phone number.
Nameplate rating.
NAME RTG
Design electrical rating.
DERNET
MDC GROSS Gross maximum dependable capacity.
Net maximum dependable capacity.
MDC NET
Changes occuring in ratings or capacities.
CHANGES
PWR RSTRCT Power level to which restricted.
RSTRCT RSN Reason for power level restriction.
GBRPTHRS Report period hours.
Year-to-date report period hours.
YTD RPT
Cumulative report period hours.
CUM RPT
CRIT COMM Critical hours.
Year-to-date critical hours.
YTD CRIT
Cumulative critical hours.
CUM CRIT
RESHTD HRS Reactor reserve shutdown hours.
YTD RESHTD Year-to-date reactor reserve shutdown hours.
CUM RESHTD Cumulative reactor reserve shutdown hours.
ONLINE HRS Generator on-line hours.
YTD ONLINE Year-to-date generator on-line hours.
CUM ONLINE Cumulative generator on-line hours.
URESHD HRS Unit reserve shutdown hours.
YTD URESHD Year-to-date unit reserve shutdown hours.
CUM URESHD Cumulative unit reserve shutdown hours.
GROSS THER Gross thermal energy generated (MWH).
Year-to-date gross thermal energy generated (MWH).
YTD THER
Cumulative gross thermal energy generated (MWH).
CUM THER
GROSS ELEC Gross electrical energy generated (MWH).
YTD G ELEC Year-to-date gross electrical energy generated (MWH).
CUM GELEC Cumulative gross electrical energy generated (MWH).
Net electrical energy generated (MWH).
NET ELEC
YTD N ELEC Year-to-date net electrical energy generated (MWH).
CUM N ELEC Cumulative net electrical energy generated (MWH).
SEVICE FAC Unit service factor.
YTD SV FAC Year-to-date unit service factor.
CUM SV FAC Cumulative unit service factor.
AVAIL FAC Unit availability factor.
YTD AV FAC Year-to-date unit availability factor.
CUM AV FAC Cumulative unit availability factor.
Unit maximum dependable capacity (MDC Net) factor.
MDC CAP
YTD MDC CP Year-to-date unit maximum dependable capacity (MDC Net) factor.
CUM MDC CP Cumulative unit maximum dependable capacity (MDC Net) factor.
Unit design electrical rating (DER Net) capacity factor.
DER CAP
YTD DER CP Year-to-date unit design electrical rating (DER Net) capacity factor.
CUM DER CP Cumulative unit design electrical rating (DER Net) capacity factor.
Unit forced outage rate.
FO RATE
114
Appendix 4. NRC monthly report, glossary of terms
YTD FORTE Year-to-date unit forced outage rate.
CUM FORTE Cumulative unit forced outage rate.
FORCED HRS Forced outage hours.
YTD FD HRS Year-to-date forced outage hours.
CUM FD HRS Cumulative forced outage hours.
SCH SHTDWN Scheduled shutdowns over next six months.
EST STRTUP If currently shutdown, estimated startup date.
MOR COMMTS Monthly operating report notes or comments.
POA COMMTS Not used for Gray Book.
THERMALPWR Licensed thermal power.
Not used for Gray Book.
RPT HRS
Not used for Gray Book.
CRIT HRS
115
MORP2.DBF - Unit Shutdown and Power Reduction
Last three digits of the facility's docket number (05000nnn).
DOCKET
RPT PERIOD Month and year of the reporting period (MMYY).
Sequential number assigned to outage.
OUTG SEQ
OUTG DATE Date outage began.
OUTG TYPE Outage type (forced or scheduled). F or S
Outage duration in hours.
OUTG HRS
OUTGREASN Outage reason. A = Equipment Failure
B = Maintenance or Test
C = Refueling
D = Regulatory Restriction
E = Operator Training and License Examination
F = Administrative
G = Operational Error
H = Other
Method of shutting down reactor.
1 = Manual Shutdown
2 = Manual Scram
3 = Automatic Scram
4 = Continuation of Shutdown from previous Month
5 = Power Reduction
9 = Power Hold or Continuation of Shutdown During
same month
Outage LER Number.
OUTG_LER
OUTG SYSTM Outage system.
OUTG COMP Outage component.
Outage description.
DESCRIP
OUTG INDEX Indexing field.
OUTG METH
116
MORP3.dbf -
no longer used, recorded the avg power output (in MW) by a plant over the
previous 24 hour period
AVERAGE DAILY POWER LEVEL
GLOSSARY
AVERAGE DAILY POWER
LEVEL (MWe)
The net electrical energy generated during the day
(measured from 0001 to 2400 hours inclusive) in
megawatts hours divided by 24 hours.
LICENSED THERMAL
POWER (MWt)
The maximum thermal power of the reactor authorized
by the NRC, expressed in megawatts.
DATE OF COMMERCIAL
OPERATION
Date unit was declared by utility owner to be
available for the regular production of electricity;
usually related to satisfactory completion of
qualification tests as specified in the purchase
contract and to accounting policies and practices
of utility.
DESIGN ELECTRICAL
RATING (DER)
The nominal net electrical output of the unit
specified by the utility and used for the purpose
of plant design.
FORCED OUTAGE
An outage required to be initiated no later than the
weekend following discovery of an offnormal
condition.
FORCED OUTAGE HOURS
The clock hours during the report period that a unit
is unavailable due to forced outages.
GROSS ELEC ENERGY
GENERATED (MWH)
Electrical output of the unit during the report
period as measured at the output terminals of the
turbine generator, in megawatt hours.
GROSS HOURS
The clock hours from the beginning of a specified
situation until its end. For outage durations, the
clock hours during which the unit is not in power
production.
GROSS THERMAL ENERGY
GENERATED (MWH)
The thermal energy produced by the unit during the
report period as measured or computed by the licensee
in megawatt hours.
HOURS GENERATOR
ON-LINE
Also, "Unit Service Hours." The total clock hours in
the report period during which the unit operated with
breakers closed to the station bus. These hours
added to the total outage hours experienced by the
unit during the report period, shall equal the hours
117
in the report period.
HOURS IN REPORTING
PERIOD
For units in power ascension at the end of the period,
the gross hours from the beginning of the period or
the first electrical production, whichever comes last,
whichever comes last, to the end of the period.
For units in commercial operation at the end of the
period, the gross hours from the beginning of the
period or of commercial operation, whichever comes
last, to the end of the period or decommissioning,
whichever comes first.
HOURS REACTOR
CRITICAL
The total clock hours in the report period during
which the reactor sustained a controlled chain reaction.
MAX DEPENDABLE
CAPACITY GROSS
(MDC GROSS) (GROSS MWe)
Dependable main-unit gross capacity, winter or summer,
whichever is smaller. The dependable capacity varies
because the unit efficiency varies during the year
due to cooling water temperature variations. It is
the gross electrical output as measured at the output
terminals of the turbine generator during the most
restrictive seasonal conditions (usually summer).
MAX DEPENDABLE
CAPACITY NET
(MDC Net) (Net MWe)
Maximum Dependable Capacity Gross less the normal
station service loads.
NAMEPLATE RATING
(Gross MWe)
The nameplate power designation of the generator
in megavolt amperes (MVA) times the nameplate
rating power factor of the generator. NOTE: The
nameplate rating of the generator may not be
indicative of the maximum dependable capacity, since
some other item of equipment of a lesser rating
(e.g., turbine) may limit unit output.
NET ELEC ENERGY
GENERATED
Gross electrical output of the unit measured at the
output terminals of the turbine generator during the
reporting period, minus the normal station service
electrical energy utilization. If this quantity is
less than zero, a negative number should be recorded.
OUTAGE
A situation in which no electrical production takes
place.
OUTAGE DATE
As reported on Appendix D of Reg. Guide 1.16, the date
of the start of the outage. If continued from a
previous month, report the same outage date but change
"Method of Shutting Down Reactor" to "4 (continuations)"
and add a note: "Continuation from previous month."
OUTAGE DURATION
The total clock hours of the outage measured from
the beginning of the report period or the outage,
118
whichever comes last, to the end of the report period
or the outage, whichever comes first.
119
OUTAGE NUMBER
A number unique to the outage assigned by the
licensee. The same number is reported each month
in which the outage is in progress. One format is
"76-05" for the fifth outage to occur in 1976.
PERIOD HOURS
See "Hours in Reporting Period."
POWER REDUCTION
A reduction in the Average Daily Power Level of more
than 20% from the previous day. All power reductions
are defined as outages of zero hours duration for the
purpose of computing unit service and availability
factors, and forced outage rate.
REACTOR AVAILABLE
HOURS
The total clock hours in the report period during
which the reactor was critical or was capable of
being made critical. (Reactor Reserve Shutdown
Hours + Hours Reactor Critical.)
REACTOR AVAILABILITY
FACTOR
Reactor Available Hours x 100
-----------------------------Period Hours
REACTOR RESERVE
SHUTDOWN
The cessation of criticality in the reactor for
administrative or other similar reasons when operation
could have been continued.
REACTOR RESERVE
SHUTDOWN HOURS
The total clock hours in the report period that the
reactor is in reserve shutdown mode. NOTE: No
credit is given for NRC imposed shutdowns.
REACTOR SERVICE
FACTOR
Hours Reactor Critical x 100
----------------------------Period Hours
REPORT PERIOD
Usually, the preceding calendar month. Can also be
the preceding calendar year (year-to-date), or the
life-span of a unit (cumulative).
RESTRICTED POWER
LEVEL
Maximum net electrical generation to which the unit
is restricted during the report period due to the
state of equipment, external conditions,
administrative reasons, or a direction by NRC.
SCHEDULED OUTAGE
Planned removal of a unit from service for refueling,
inspection, training, or maintenance. Those outages
which do not fit the definition of "Forced Outage"
perforce are "Scheduled Outages."
STARTUP AND POWER
ASCENSION TEST
PHASE
Period following initial criticality during which
the unit is tested at successively higher levels,
culminating with operation at full power for a
sustained period and completion of warranty runs.
120
Following this phase, the utility generally considers
the unit to be available for commercial operation.
UNIT
The set of equipment uniquely associated with the
reactor, including turbine generators and ancillary
equipment, considered as a single electrical energy
production facility.
UNIT AVAILABLE HOURS
The total clock hours in the report period during
which the unit operated on-line or was capable of such
operation. (Unit Reserve Shutdown Hours + Hours
Generator On-Line.)
UNIT AVAILABILITY
FACTOR
Unit Available Hours x 100
Period Hours
Period Hours
UNIT CAPACITY FACTORS
- Using Licensed Thermal Power
Gross Thermal Energy Generated x 100
Period Hours x Lic. Thermal Power
- Using Nameplate Rating
Gross Electrical Energy Generated x 100
----------------------------------------
Period Hours x Nameplate Rating
- Using DER
Net Electrical Energy Generated x 100
--------------------------------------
Period Hours x DER
- Using MDC Gross
Gross Electrical Energy Generated x 100
Period Hours x MDC Gross
Period Hours x MC Gross
- Using MDC Net
Net Electrical Energy Generated x 100
Period Hours x MDC Net
NOTE: If MDC Gross and/or MDC Net have not been determined, the DER Net
is substituted for this quantity for Unit Capacity Factor
calculations.
UNIT FORCED OUTAGE
RATE
Forced Outage Hours x 100
---------------------------------------
Unit Service Hours + Forced Outage Hours
UNIT RESERVE SHUTDOWN
The removal of the unit from on-line operation for
economic or other similar reasons when operation could
have been continued.
121
UNIT RESERVE SHUTDOWN
HOURS
The total clock hours in the report period during
which the unit was in reserve shutdown mode.
UNIT SERVICE FACTOR
Unit Service Hours x 100
Period Hours
Period Hours
UNIT SERVICE HOURS
See "Hours Generator On-Line
Io
Morp 1.dbf
DOCKET
Last three digits of the facility's docket number (05000nnn).
Last three digits of the facility's docket number (05000nnn).
DOCKETA
Month and year of reporting period. (MMYY)
RPT PERIOD
UTL CONTCT
Utility contact concerning monthly operating data.
FIRST NAME
Contact's first name.
UTL PHONE
Contact's phone number.
NAME_RTG
Nameplate rating.
DERNET
Design electrical rating.
Gross maximum dependable capacity.
MDC GROSS
MDC NET
Net maximum dependable capacity.
CHANGES
Changes occuring in ratings or capacities.
PWR RSTRCT
Power level to which restricted.
RSTRCT RSN
Reason for power level restriction.
GBRPT_HRS
Report period hours.
YTD RPT
Year-to-date report period hours.
CUM RPT
Cumulative report period hours.
CRIT COMM
Critical hours.
YTD CRIT
Year-to-date critical hours.
CUM CRIT
Cumulative critical hours.
RESHTD HRS
Reactor reserve shutdown hours.
122
YTD RESHTD
Year-to-date reactor reserve shutdown hours.
CUM RESHTD
Cumulative reactor reserve shutdown hours.
ONLINE HRS
Generator on-line hours.
YTD ONLINE
Year-to-date generator on-line hours.
CUMONLINE
Cumulative generator on-line hours.
URESHD HRS
Unit reserve shutdown hours.
YTD URESHD
Year-to-date unit reserve shutdown hours.
CUM URESHD
Cumulative unit reserve shutdown hours.
Gross thermal energy generated (MWH).
GROSS THER
YTD THER
Year-to-date gross thermal energy generated (MWH).
CUM THER
Cumulative gross thermal energy generated (MWH).
GROSS ELEC
Gross electrical energy generated (MWH).
YTD_G ELEC
Year-to-date gross electrical energy generated (MWH).
CUMGELEC
Cumulative gross electrical energy generated (MWH).
NET ELEC
Net electrical energy generated (MWH).
YTD_N_ELEC
Year-to-date net electrical energy generated (MWH).
CUM N ELEC
Cumulative net electrical energy generated (MWH).
SEVICE FAC
Unit service factor.
YTD SV FAC
Year-to-date unit service factor.
CUM SV FAC
Cumulative unit service factor.
Unit availability factor.
AVAIL FAC
YTD_AV_FAC
Year-to-date unit availability factor.
CUM AV FAC
Cumulative unit availability factor.
MDC CAP
Unit maximum dependable capacity (MDC Net) factor.
YTD MDC CP
Year-to-date unit maximum dependable capacity (MDC Net) factor.
CUM MDC CP
Cumulative unit maximum dependable capacity (MDC Net) factor.
123
DERCAP
Unit design electrical rating (DER Net) capacity factor.
YTD DER CP
factor.
Year-to-date unit design electrical rating (DER Net) capacity
CUM DER CP
factor.
Cumulative unit design electrical rating (DER Net) capacity
FO RATE
Unit forced outage rate.
YTDFORTE
Year-to-date unit forced outage rate.
CUM FORTE
Cumulative unit forced outage rate.
FORCED HRS
Forced outage hours.
YTD FD HRS
Year-to-date forced outage hours.
CUM FD HRS
Cumulative forced outage hours.
SCH SHTDWN
Scheduled shutdowns over next six months.
If currently shutdown, estimated startup date.
ESTSTRTUP
MOR COMMTS
Monthly operating report notes or comments.
POA COMMTS
Not used for Gray Book.
THERMALPWR
Licensed thermal power.
RPT HRS
Not used for Gray Book.
CRIT HRS
Not used for Gray Book.
124
MORP2.DBF - Unit Shutdown and Power Reduction
DOCKET
Last three digits of the facility's docket number (05000nnn).
RPT PERIOD
Month and year of the reporting period (MMYY).
OUTG SEQ
Sequential number assigned to outage.
OUTG DATE
Date outage began.
OUTGTYPE
Outage type (forced or scheduled). F or S
OUTGHRS
Outage duration in hours.
OUTG REASN
Outage reason. A = Equipment Failure
B = Maintenance or Test
C = Refueling
D = Regulatory Restriction
E = Operator Training and License Examination
F = Administrative
G = Operational Error
H = Other
OUTG METH
Method of shutting down reactor.
1 = Manual Shutdown
2 = Manual Scram
3 = Automatic Scram
4 = Continuation of Shutdown from previous Month
5 = Power Reduction
9 = Power Hold or Continuation of Shutdown During
same month
Outage LER Number.
OUTGLER
OUTGSYSTM
Outage system.
OUTG COMP
Outage component.
DESCRIP
Outage description.
OUTG INDEX
Indexing field.
125
GLOSSARY
AVERAGE DAILY POWER
LEVEL (MWe)
The net electrical energy generated during the day
(measured from 0001 to 2400 hours inclusive) in
megawatts hours divided by 24 hours.
LICENSED THERMAL
POWER (MWt)
The maximum thermal power of the reactor authorized
by the NRC, expressed in megawatts.
DATE OF COMMERCIAL
OPERATION
Date unit was declared by utility owner to be
available for the regular production of electricity;
usually related to satisfactory completion of
qualification tests as specified in the purchase
contract and to accounting policies and practices
of utility.
DESIGN ELECTRICAL
RATING (DER)
The nominal net electrical output of the unit
specified by the utility and used for the purpose
of plant design.
FORCED OUTAGE
An outage required to be initiated no later than the
weekend following discovery of an offniormal
condition.
FORCED OUTAGE HOURS
The clock hours during the report period that a unit
is unavailable due to forced outages.
GROSS ELEC ENERGY
GENERATED (MWH)
Electrical output of the unit during the report
period as measured at the output terminals of the
turbine generator, in megawatt hours.
GROSS HOURS
The clock hours from the beginning of a specified
situation until its end. For outage durations, the
clock hours during which the unit is not in power
production.
GROSS THERMAL ENERGY
GENERATED (MWH)
The thermal energy produced by the unit during the
report period as measured or computed by the licensee
in megawatt hours.
HOURS GENERATOR
ON-LINE
Also, "Unit Service Hours." The total clock hours in
the report period during which the unit operated with
breakers closed to the station bus. These hours
added to the total outage hours experienced by the
unit during the report period, shall equal the hours
in the report period.
HOURS IN REPORTING
PERIOD
For units in power ascension at the end of the period,
the gross hours from the beginning of the period or
the first electrical production, whichever comes last,
126
whichever comes last, to the end of the period.
For units in commercial operation at the end of the
period, the gross hours from the beginning of the
period or of commercial operation, whichever comes
last, to the end of the period or decommissioning,
whichever comes first.
HOURS REACTOR
CRITICAL
The total clock hours in the report period during
which the reactor sustained a controlled chain
reaction.
MAX DEPENDABLE
CAPACITY GROSS
(MDC GROSS) (GROSS MWe)
Dependable main-unit gross capacity, winter or summer,
whichever is smaller. The dependable capacity varies
because the unit efficiency varies during the year
due to cooling water temperature variations. It is
the gross electrical output as measured at the output
terminals of the turbine generator during the most
restrictive seasonal conditions (usually summer).
MAX DEPENDABLE
CAPACITY NET
(MDC Net) (Net MWe)
Maximum Dependable Capacity Gross less the normal
station service loads.
NAMEPLATE RATING
(Gross MWe)
The nameplate power designation of the generator
in megavolt amperes (MVA) times the nameplate
rating power factor of the generator. NOTE: The
nameplate rating of the generator may not be
indicative of the maximum dependable capacity, since
some other item of equipment of a lesser rating
(e.g., turbine) may limit unit output.
NET ELEC ENERGY
GENERATED
Gross electrical output of the unit measured at the
output terminals of the turbine generator during the
reporting period, minus the normal station service
electrical energy utilization. If this quantity is
less than zero, a negative number should be recorded.
OUTAGE
A situation in which no electrical production takes
place.
OUTAGE DATE
As reported on Appendix D of Reg. Guide 1.16, the date
of the start of the outage. If continued from a
previous month, report the same outage date but change
"Method of Shutting Down Reactor" to "4 (continuations)"
and add a note: "Continuation from previous month."
OUTAGE DURATION
The total clock hours of the outage measured from
the beginning of the report period or the outage,
whichever comes last, to the end of the report period
or the outage, whichever comes first.
127
OUTAGE NUMBER
A number unique to the outage assigned by the
licensee. The same number is reported each month
in which the outage is in progress. One format is
"76-05" for the fifth outage to occur in 1976.
PERIOD HOURS
See "Hours in Reporting Period."
POWER REDUCTION
A reduction in the Average Daily Power Level of more
than 20% from the previous day. All power reductions
are defined as outages of zero hours duration for the
purpose of computing unit service and availability
factors, and forced outage rate.
REACTOR AVAILABLE
HOURS
The total clock hours in the report period during
which the reactor was critical or was capable of
being made critical. (Reactor Reserve Shutdown
Hours + Hours Reactor Critical.)
REACTOR AVAILABILITY
FACTOR
Reactor Available Hours x 100
-----------------------------Period Hours
REACTOR RESERVE
SHUTDOWN
The cessation of criticality in the reactor for
administrative or other similar reasons when operation
could have been continued.
REACTOR RESERVE
SHUTDOWN HOURS
The total clock hours in the report period that the
reactor is in reserve shutdown mode. NOTE: No
credit is given for NRC imposed shutdowns.
REACTOR SERVICE
FACTOR
Hours Reactor Critical x 100
----------------------------Period Hours
REPORT PERIOD
Usually, the preceding calendar month. Can also be
the preceding calendar year (year-to-date), or the
life-span of a unit (cumulative).
RESTRICTED POWER
LEVEL
Maximum net electrical generation to which the unit
is restricted during the report period due to the
state of equipment, external conditions,
administrative reasons, or a direction by NRC.
SCHEDULED OUTAGE
Planned removal of a unit from service for refueling,
inspection, training, or maintenance. Those outages
which do not fit the definition of "Forced Outage"
perforce are "Scheduled Outages."
STARTUP AND POWER
ASCENSION TEST
PHASE
Period following initial criticality during which
the unit is tested at successively higher levels,
culminating with operation at full power for a
sustained period and completion of warranty runs.
128
Following this phase, the utility generally considers
the unit to be available for commercial operation.
129
UNIT
The set of equipment uniquely associated with the
reactor, including turbine generators and ancillary
equipment, considered as a single electrical energy
production facility.
UNIT AVAILABLE HOURS
The total clock hours in the report period during
which the unit operated on-line or was capable of such
operation. (Unit Reserve Shutdown Hours + Hours
Generator On-Line.)
UNIT AVAILABILITY
FACTOR
Unit Available Hours x 100
Period Hours
Period Hours
UNIT CAPACITY FACTORS
- Using Licensed Thermal Power
Gross Thermal Energy Generated x 100
Period Hours x Lic. Thermal Power
- Using Nameplate Rating
Gross Electrical Energy Generated x 100
Period Hours x Nameplate Rating
- Using DER
Net Electrical Energy Generated x 100
Period Hours x DER
- Using MDC Gross
Gross Electrical Energy Generated x 100
Period Hours x MDC Gross
Net Electrical Energy Generated x 100
- Using MDC Net
Period Hours x MDC Net
NOTE: If MDC Gross and/or MDC Net have not been determined, the DER Net
is substituted for this quantity for Unit Capacity Factor
calculations.
Forced Outage Hours x 100
UNIT FORCED OUTAGE
RATE
Unit Service Hours + Forced Outage Hours
UNIT RESERVE SHUTDOWN
The removal of the unit from on-line operation for
economic or other similar reasons when operation could
have been continued.
UNIT RESERVE SHUTDOWN
HOURS
The total clock hours in the report period during
which the unit was in reserve shutdown mode.
130
UNIT SERVICE FACTOR
Unit Service Hours x 100
-------------------------
Period Hours
UNIT SERVICE HOURS
See "Hours Generator On-Line."
131
27 May 1998
Appendix 5. BWR/4 Operating Cycle Skylines
260, Brown's Ferry 2, RF 1/29/93-9/6/93 (220d), Op 9/6/93-10/1/94 (483d), (74.4, 93.8, 95.3%)
100%
..
80%
60%
40%
20%
0%
100
0
-
100%
200
400
300
500
600
700
(53d), Op 11/23/94-3/22/96 (486d), (83.4, 92 5, 93.9%)
260, Brown's Ferry 2, RF 10/01/94-11/23/94
W-
80%
60%
40%
20%
0%
100
0
-
100%
200
400
300
500
600
700
271, Vermont Yankee, RF 3/6/92-4/20/92 (45d), Op 4/20/92-8/27/93 (495d), (85.4, 93 1, 94 6%)
g
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
271, Vermont Yankee, RF 8/27/93-10/25/93 (59d), Op 1025/93-3/17/95 (508d), (83.1, 92 7, 93 0%)
-
100%
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
271, Vermont Yankee, RF 3/17/95-5/3/95 (47d), Op 5/3/95-9/6/96 (492d), (87.5, 95 9, 96.8%)
-
100%
80%
60%
40%
20%
0%
100
T
200
400
300
X axis - Operating cycle days following refueling
132
500
600
700
Cycle Capacity Factor,
Operating Period Capacity Factor,
Op per CF coast down adjusted
27 May 1998
Appendix 5. BWR/4 Operating Cycle Skylines
100%
.
277, Peach Bottom 2, RF 9/16/94-10/21/94 (35d), Op 10121/94-9/12/96 (693 d), (84.6, 88.9, 95 8%)
.
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
278, Peach Bottom 3, RF 9/13/93-1/9/92 (118d), Op 119/92-9/16/93 (618 d), (67.9, 80.9, 82.8%)
100%
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
278, Peach Bottom 3, RF 9/18/93-11/15193 (58d), Op 11/15/93-9/21/95 (676 d), (79.9, 86.8, 92 9%)
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
278, Peach Bottom 3, RF 9/22/95-10/17/95 (26d), Op 10/17/95-10/3/97 (718 d), (86.7, 89.8, 93 8%)
100%
80%
60%
40%
20%
0
100
200
400
300
X axis - Operating cycle days following refueling
133
500
600
700
Cycle Capacity Factor,
Operating Period Capacity Factor,
Op per CF coast down adjusted
27 May 1998
Appendix 5. BWR/4 Operating Cycle Skylines
298, Cooper Station, RF 10/14/95-12/30/95 (77d),Op 12/30/95-3/30/97 (456 d), (79.8, 93 4, 93 6%)
100%
80%
60%
40%
20%
0%
700
600
500
400
300
200
100
0
321, Hatch 1, RF3/16/93-5/16/93 (61d), Op 5/16/93-9/21194 (493 d), (82.4, 92 6, 92.6%)
100%
80%
60%
40%
20%
0%
0
100
,o'%tr
100
50
200
150
250
350
300
450
400
500
321, Hatch 1, RF9/21/94-11/5/94 (45d), Op 11/5/94-3/22/96 (504 d), (83.9, 91 4, 93.8%)
_...I
-li-rTTTn~-
i _'.._1
on
OU-/
60%
40%
20%
0%
0
100%
100
200
300
400
500
600
324,Brunswick 2, RF 3/26/94-6/30/94 (96d), Op 6/30/94-2/1/96 (582 d (78 8, 91 8 97 4%)
80%
60%
40%
20%
0%
ann
X axis - Operating cycle days following refueling
134
400
-r.,
500
-v-
600
Cycle Capacity Factor,
Operating Period Capacity Factor,
Op per CF coast down adjusted
27 May 1998
Appendix 5. BWR/4 Operating Cycle Skylines
325,Brunswick 1, RF 4/01/95-5/15/95 (44d), Op 5/15/95-9/27/96 (502 d), (81 7, 88.9, 89.4%)
100%
80%
60%
40%
20%
0%
0
600
500
400
300
200
100
331, Duane Arnold, RF 2/27/92-4/27/92 (60d), Op 4/27/92-7/30/93 (460 d , (77.2, 87.2, 91.5%)
100%
80%
60%
40%
20%
0%
0
I
50
100
200
150
250
350
300
450
400
500
331, Duane Arnold, RF 7/29/93-10/12/93 (75d), Op 10/12/93-2/23/95 (500 d), (77 2, 88.8, 88.8%)
100%
80%
60%
40%
20%
0%
0
100
200
300
400
500
600
331, Duane Arnold, RF 2/24/95-4/19/95 (54d), Op 4/19/95-10/10/96 (541 d), (84.9, 93 4, 93 5%)
100%
80%
60%
40%
20%
0%
Inn
qn0
X axis - Operating cycle days following refueling
300
135
400
500
600
Cycle Capacity Factor,
Capacity Factor,
Period
Operating
Op per CF coast down adjusted
27 May 1998
Appendix 5. BWR/4 Operating Cycle Skylines
333, Fitzpatrick, RF 11/30194-3/26195 (116d), Op 3/26/95-10/24196 (579 d), (72.6, 87.2, 87.2%)
100%
80%
60%
40%
20%
0%
600
500
400
300
200
100
0
341, Fermi 2, RF 9/12/92-11/7/92 (56d), Op 11/7/92-3/12/94 (491 d), (63 6,70 8, 70.8%)
100%
80%
60%
40%
20%
0%
0
50
100
300
250
200
150
350
450
400
500
341, Fermi 2, RF 3/12/94-1/18/95 (312d), Op 1/18/95-3/12/96 (618 d), (50 4, 75 9, 77.6%)
100%
80%
60%
40%
20%
0%
0
100
500
400
300
200
600
700
352, Limenck 1, RF 3/20/92-7/9/92 (11 d), Op7/9/92-2/5/94 (577 d), (79 8, 95 1, 92.3%)
- -m
-
100%
80%
60%
40%
20%
0%
0
500
400
300
200
100
600
352, Limerick 1, RF 2/4/94-3/11/94 (35d), Op3/11/94-2/5/96 (697 d), (86.7, 91 0, 91 7%)
100%
80%
60%
40%
20%
0%
inn
nn
400
300
X axis - Operating cycle days following refueling
136
500
600
700
Cycle Capacity Factor,
Operating Period Capacity Factor,
Op per CF coast down adjusted
27 May 1998
Appendix 5. BWR/4 Operating Cycle Skylines
0
20m1h-"
100%
80%
60%
40%
20%
0%
100
0
200
400
300
500
700
600
353, Limerick 2, RF 1/27/95-2/20/95 (24d), Op 220/95-1/30/97 (711 d), (87.0, 89 9, 91.3%)
100%
80%
60%
40%
20%
0%
400
200
500
700
600
354, Hope Creek, RF 9/12/92-11/10/92 (59d), Op 11/10192-3/5/94 (481 d), (84.4, 94.8, 94.8%)
--
100%
80%
60%
40%
20%
0%
0
100
200
400
300
500
700
600
354, Hope Creek, RF 3/5/94-4/27/94 (53d), Op 4/47/94-11/10/95 (563 d), (80 2, 87 8, 88 3%)
100% -
'
80%
60%
40%
20%
0%
n
100nn
200
A
30
300
X axis - Operating cycle days following refueling
400
137
500
600
700
Cycle Capacity Factor,
Operating Period Capacity Factor,
Op per CF coast down adjusted
27 May 1998
Appendix 5. BWR/4 Operating Cycle Skylines
366, Hatch 2, RF9/23/95-11/23/95 (61d), Op 11/23/953/14/97 (478 d), (86.5, 97.5, 97 5%)
/
100%
80%
60%
40%
20%
0%
0
200
100
100%
-
400
300
500
600
700
387, Susquehanna 1, RF 3/6/92-5/16192 (71d), Op 5/16/92-9/25/93 (498 d), (71 2, 81.4, 81.4%)
-
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
387, Susquehanna 1, RF 9/25/93-1/22/94 (119d), Op 1/22/94-3/24/95 (427 d), (75 7, 96.8, 96.8%)
100%
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
387, Susquehanna 1, RF 3/25/95-5/6/95 (42d), 0p516/195-9/8/96 (492 d), (81 8, 88.8, 89 2%)
100%
80%
60%
40%
20%
0%
400
X axis - Operating cycle days following refueling
138
500
600
700
Cycle Capacity Factor,
Operating Period Capacity Factor,
Op per CF coast down adjusted
27 May 1998
Appendix 5. BWR/4 Operating Cycle Skylines
388, Susquehanna 2, RF 9/12/92-11/14/92 (63d), Op 11114/92-4/15/94 (518 d), (81.0, 90.8, 90 8%)
r-
100%
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
388, Susquehanna 2, RF 3/14/94-6/10/94 (88d), Op 6/10/94-9/12/95 (460 d), (79 6, 94.9, 94.9%)
100%
80%
60%
40%
20%
0%
0
100
200
400
300
500
600
700
388, Susquehanna 2, RF 9/13/95-10/21/95 (38d), Op 10/21/95-3/12/97 (510 d), (86.9, 93 4, 94.0%)
100%
80%
60%
40%
20%
0%
500
X axis - Operating cycle days following refueling
139
Cycle Capacity Factor,
Capacity Factor,
Period
Operating
Op per CF coast down adjusted
Sheeti
DOCKET DOCKET RPT_PE
RIOD
A
0195
261
261
282 306
282
306
0195
0195
FIRST_N UTL_PH NAME_R DER_NE MDC_GR MDC_NE CHANGE PWR_RS RSTRCT GB_RPT YTD_RP CUM _RP
T
T
HRS
RSN
TRCT
S
T
OSS
T
TG
ONE
AME
209712
744
744
0
683
700
700
(803) 857- 739
J. S.
1000
SCARBOROUGH
UTL_CONTCT
DALE DUGSTAD
DALE DUGSTAD
305
305
0195
M. L. ANDERSON
315
316
321
366
325
315
316
321
366
325
0195
0195
0195
0195
0195
J. D. JANISSE
J. D. JANISSE
R. M. BEARD
R. M. BEARD
FRANCES
HARRISON
324
373
374
324
373
374
0195
0195
0195
FRANCES
M. J.
M. J.
CIALKOWSKI
382
456
382
456
0195
0195
T. S. BECKER
PAUL STANCZAK
457
461
457
461
0195
0195
PAUL STANCZAK
M. C. HOLLON
155
244
155
244
0195
0195
J. R. JOHNSTON
JOHN V.
WALDEN
254
265
259
254
265
259
0195
0195
0195
KRISTAL MOORE
KRISTAL MOORE
T. R. SMITH
(612) 388- 593
(612) 388- 593
1121
EXT.
(414) 388- 560
2560
EXT.
2453
(616) 465-1152
(616) 465-1133
(912) 367- 850
(912) 367- 850
(910) 457- 867
2756
530
530
545
544
513
512
744
744
0
0
744
744
185208
176326
)r
535
537
511
0
744
744
180842
1020
1090
776
784
821
1056
1100
774
798
791
1000
1060
741
765
767
0
0
0
0
0
744
744
744
744
744
744
744
744
744
744
176064
149760
167303
134929
156696
(910) 457- 867
(815)357- 1146
(815) 357-1146
6761
EXT.
2427
821
1078
1078
782
1146
1146
754
1036
1036
0
0
0
744
744
744
744
744
744
168720
97176
90168
(504) 739- 1200
(815) 458- 1175
2801
(815) 458- 1175
(217) 935- 985
8881
EXT.
3537
(616) 547- 75
(315) 524- 490
4446
EXT. 588
1104
1120
1120
1175
1075
1120
0
0
744
744
744
744
82009
57057
1120
933
1175
973
1120
930
0
0
744
744
744
744
55139
63010
72
470
71
490
67
470
0
0
744
744
744
744
279163
220752
(309) 654- 828
(309) 654- 828
(205) 729- 1152
789
789
1065
813
813
0
769
769
0
0
0
0
744
744
0
744
744
0
199919
198356
95743
Y.,
Page 1
'
- 2ru K ;-15c
,a//
Sheetl
CRIT_CO YTD_CRI CUM_CR RESHTD YTD_RE CUM_RE ONLINE_ YTD ON CUM_ON URESHD YTD UR CUM UR GROSS_ YTD_TH
MM
T
IT
HRS
SHTD
SHTD
HRS
LINE
LINE
HRS
ESHD
ESHD
THER
ER
23.2
1697913 1697913
0
744
744
145063.2 0
0
3314.7
148155.5 0
744
744
CUM_TH GROSS_ YTD_G
ER
ELEC
ELEC
563675
2.97E+08 563675
744
744
744
744
158343.3 0
155854.2 0
0
0
5571.1
1516.1
744
744
744
744
156685
0
154466.7 0
0
0
0
0
1225396
1224824
1225396
1224824
2.48E+08 415460
2.46E+08 414610
415460
414610
744
744
155519
0
0
2330.5
744
744
153590.6 0
0
10
1227125
1227125 2.44E+08 409000
409000
744
744
744
744
744
744
744
744
744
744
132653.9
100760.5
126385.6
104104
96615.3
0
0
0
0
0
0
0
0
0
0
463
0
0
0
1647.1
744
744
744
744
744
744
744
744
744
744
130566.6
97279.1
121244.4
100505.7
92701
0
0
0
0
0
0
0
0
0
0
321
0
0
0
0
2344805 2344805
2533412 2533412
1810481 1810481
1802220 1802220
1797926 1797926
768830
833510
596629
603780
600565
768830
833510
596629
603780
600565
744
744
744
744
744
744
105299.6 0
67384.1
0
65952.1
0
0
0
0
0
1641.2
1716.9
744
744
744
744
744
744
99861.6
65854.9
64729.9
0
0
0
0
0
0
0
1
0
1795891 1795891 2.14E+08 585730
2457885 2457885 1.95E+08 839182
2425194 2425194 1.96E+08 830597
585730
839182
830597
744
744
744
744
69466.8
45342.8
0
0
0
0
0
0
744
744
744
744
68588.3
44710.3
0
0
0
0
0
0
2517937 2517937
2508279 2508279
2.27E+08 838480
1.38E+08 875340
838480
875340
744
715.5
744
715.5
45576.5
46467.2
0
0
0
0
0
0
744
707.5
744
707.5
45186.3
45131.9
0
0
0
0
0
4
2520337 2520337
1930960 1930960
1.37E+08 869340
1.2E+08 634445
869340
634445
744
744
744
744
202458.5 0
176559.2 0
0
0
0
1687.6
744
744
744
744
199275.1 0
173716.7 0
0
0
0
8.5
164538
1095892
38418252 52547
2.47E+08 370072
52547
370072
512.8
744
0
512.8
744
0
152946.3 0
150520.3 0
0
59521
0
0
0
3421.9
2985.8
6997
457
744
0
457
744
0
148279.9 0
146675.7 0
58267
0
0
0
0
909.2
702.9
0
882951.9 882951.9 3.2E+08 281020
1655253 1655253 3.18E+08 524392
1.68E+08 0
0
0
281020
524392
0
Page 2
164538.
1095892
3.84E+08
2.97E+08
2.73E+08
2.22E+08
2.03E+08
Sheet1
CUM G NET ELE YTD NE CUM N SEVICE
FAC
ELEC
LEC
C
ELEC
91117149 100
96426221 537798 537798
YTD SV CUM SV AVAIL F YTD AV CUM AV MDC CA YTD MD CUM MD DER CA YTD DE CUM DE FO RAT
E
R CP
R CP
P
C CP
C CP
P
FAC
FAC
AC
FAC
FAC
0
62.1
103.3
103.3
63.6
105.8
105.8
69.2
100
100
69.2
100
81765110 395542
80126650 394903
395542
394903
76850809 100
75752399 100
100
100
84.6
87.6
100
100
100
100
84.6
87.6
103.6
103.7
103.6
103.7
82.4
85.8
100.3
100.1
100.3
100.1
78.3
81.1
0
0
80736000 389817
389817
76843318 100
100
84.9
100
100
84.9
102.5
102.5
82.9
97.9
97.9
79.4
0
1.25E+08
96140500
87708945
72741830
66760290
741655
807133
572149
579453
584170
741655
807133
572149
579453
584170
1.2E+08
92564568
83469855
69285268
64183209
100
100
100
100
100
100
100
100
100
100
74.2
65
72.5
74.5
59.2
100
100
100
100
100
100
100
100
100
100
74.3
65
72.5
74.5
59.2
99.7
102.3
103.8
101.8
102.4
99.7
102.3
103.8
101.8
102.4
67.9
59.1
66.6
67.2
52.2
97.7
99.5
99.1
99.3
95.6
97.7
99.5
99.1
99.3
95.6
66.2
57.6
64
65.5
49.9
0
0
0
0
0
69235351 569093
65230073 813809
65519166 805795
569093
813809
805795
66402352 100
62582953 100
62990557 100
100
100
100
59.2
67.8
71.8
100
100
100
100
100
100
59.2
67.8
71.8
101.4
105.6
104.5
101.4
105.6
104.5
50.3
62.2
67.4
93.2
101.5
100.5
93.2
101.5
100.5
47.9
59.7
64.8
0
0
0
75572940 804313
47379162 845089
804313
845089
72046677 100
45377909 100
100
100
83.6
78.4
100
100
100
100
83.6
78.4
100.6
101.4
100.6
101.4
81.7
71
97.9
101.4
97.9
101.4
79.6
71
0
0
46889129 839335
39791396 606581
839335
606581
44942065 100
37882738 95.1
100
95.1
81.9
71.6
100
95.1
100
95.1
81.9
71.6
100.7
87.7
100.7
87.7
72.8
64.6
100.7
87.4
100.7
87.4
72.8
64.4
0
4.9
12208252 49903.9
81604124 351805
49903.9
351805
11549637 100
77461035 100
100
100
71.4
78.7
100
100
100
100
71.4
78.7
100.1
100.6
100.1
100.6
61.5
75.8
93.2
100.6
93.2
100.6
57.5
75.8
0
0
1.04E+08 265985
1.02E+08 501116
55398130 0
265985
501116
0
97854151 61.4
96657025 100
53796427 0
61.4
100
0
74.2
73.9
60.9
61.4
100
0
61.4
100
0
74.6
74.3
60.9
46.5
87.6
0
46.5
87.6
0
63.7
63.4
52.8
45.3
85.4
0
45.3
85.4
0
62
61.8
52.8
38.6
0
0
Page 3
.
Sheet1
YTD_FO_ CUM_FO FORCED YTD FD
HRS
HRS
RTE
RTE
0
0
15.1
0
EST_ST
CUM_FD SCHSHTDWN
RTUP
HRS
25843.7 REFUELING
OUTAGE, APRIL
29, 1995, 45
DAYS.
8125.9
REFUELING
4010.7
OUTAGE, MAY
12, 1995.
REFUELING
3273.6
OUTAGE, APRIL
1, 1995, 40 DAYS.
0
0
4.9
2.5
0
0
0
0
0
2.1
0
0
0
0
0
0
0
5.6
15.6
11.3
7
14.8
0
0
0
0
0
0
0
0
0
0
7800.6
18044.3
15378.5
7562.5
16106.4
0
0
0
12
8.2
10.5
0
0
0
0
0
0
13562.5
5868.7
7576.2
0
0
3.1
8.3
0
0
0
0
2210
4030.5
0
4.9
5.8
9.2
0
36.5
0
36.5
2771.3
4560.2
0
0
11
5.6
0
0
0
0
16417.3
10353.8
38.6
0
0
7.6
9.9
25.6
287
0
0
287
0
0
12251.8
16078.1
20022
REFUELING
OUTAGE, APRIL
1, 1995, 55 DAYS.
MOR_COMMTS
CUMULATIVE
CUMULATIVE
UNIT CAPACITY
FACTOR (MDC
CUMULATIVE
UNIT CAPACITY
FACTOR (MDC
NET) IS
CUMULATIVE
CUMULATIVE
CUMULATIVE
CUMULATIVE
CUMULATIVE
UNIT CAPACITY
FACTOR (MDC
NET) IS
CUMULATIVE
REFUELING
OUTAGE,
FEBRUARY 18,
1995, EIGHT
WEEKS.
FEBRUARY 19,
1995.
REFUELING
OUTAGE,
MARCH 12, 1995,
50 DAYS.
REFUELING/MAI
NTENANCE
OUTAGE,
MARCH 26, 1995,
36 DAYS.
CUMULATIVE
CUMULATIVE
REACTOR AND
UNIT RESERVE
SHUTDOWN
HOURS ARE
LICENSEE HAS
Page 4
POA_COMMTS
THERMALPWR RPTHRS CRIT_HRS
2300
744
744
1650
1650
744
744
744
744
1650
744
744
3250
3411
2436
2436
2436
744
744
744
744
744
744
744
744
744
744
2436
3323
3323
744
744
744
744
744
744
3390
3411
744
744
744
744
3411
2894
744
744
744
715.5
240
1520
744
744
744
744
2511
2511
3293
744
744
0
512.8
744
0
Sample NRC/INEEL MORPH 2 data
RPT
DOCKET PERIOD
352
0195
OUTG
DATE
1/2/95
8/5/97
352
0195
1/22/95
352
0195
1/23/95t
352
0195
1/30/95
352
0295
2/21/95
352
0296
2/2/96
352
0296
2/5/96
OUTG OUTG OUTG OUTG OUTG
OUTG LER
TYPE METH REASN SYSTM COMP
DESCRIP
0
5
PUMP REACTOR POWER REDUCED FOR RECIRC
XX PUMP MAINTENANCE.
RB
CONR REACTOR POWER REDUCED DUE TO
S
B
5
OD
CONTROL ROD PATTERN ADJUSTMENTS.
INSTR REACTOR POWER WAS REDUCED DUE TO A
5
CC
0
S
B
CONTROL VALVE PRESSURE SWITCH.
U
0
F
GENE REACTOR POWER REDUCED DUE TO A
CB
5
H
RA
TRIPPED M/G SET.
F
37.8
3
RB
CONR REACTOR SCRAM DUE TO ELECTRICAL
H
OD
DISTURBANCE AT WHITPAIN SUBSTATION.
S
RC
FUELX GENERATOR WAS TAKEN OFF GRID FOR
2
7.4
C
REFUELING OUTAGE, GENERATOR WAS
THEN PLACED BACK ON GRID TO SUPPORT
PJM POWER SHORTAGE.
S
FUELX REFUELING OUTAGE.
2
578
RC
C
352
0396
2/5/96
13.6
S
4
C
RC
352
0396
3/4/96
0
S
5
B
RB
352
0396
3/24/96
158.3
S
2
B
CC
352
0396
3/31/96
0
F
5
A
HE
3521
0495
4/24/95
0
F
5
A
CH
352
0496
4/1/96
0
F
5
A
HA
352
0496
4/2/96
3.1
F
1
A
HA
352
0496
4/3/96
0
S
5
B
RB
352
0497
4/11/97
0
S
5
B
HA
352
0497
4/12/97
75
S
1
B
HA
OUTG
HRS
x
LCDR David L. Brodeur
FUELX
x
CONR
OD
VALVE
X
VALVE
X
PUMP
XX
INSTR
U
INSTR
U
REFUELING OUTAGE CONTINUED.
POWER REDUCED DUE TO CONTROL ROD
PATTERN ADJUSTMENT.
REACTOR WAS SHUTDOWN DUE TO SRV
AND TIP MACHINE MAINTENANCE.
REACTOR POWER WAS REDUCED DUE TO
TURBINE VALVE/EHC CONTROL PROBLEM.
REACTOR POWER REDUCED DUE TO
REACTOR FEEDPUMP TRIP.
POWER REDUCED TO REPLACE EHC SPEED
CONTROLLER CARD.
MAIN TURBINE WAS TAKEN OFF THE GRID
TO REPLACE THE EHC SPEED CONTROLLER
CARD.
CONR POWER REDUCED FOR CONTROL ROD
OD
PATTERN ADJUSTMENT.
GENE POWER REDUCTION FOR TURBINE
BENERATOR MAINTENANCE.
RA
GENE MAIN TURBINE WAS TAKEN OFF LINE FOR
RA
GENERATOR MAINTENANCE.
Page 1
Bwr495-7.xls
8/5/97
Sample NRC/INEEL MORPH 2 data
RPT
DOCKET PERIOD
352
0497
OUTG
DATE
4/16/97
OUTG OUTG OUTG
OUTG OUTG
TYPE METH REASN SYSTM COMP
VALVE
5
S
X
-----------1
1
I
HTEXC
2
99.4
S
H
OUTG
HRS
I
05951
5/7/95
0596
5/21/96
51
F
3
0596
5/25/96
0
S
5
0696
6/17/96
6.9
F
1
0696
6/19/96
0
F
5
0696
6/20/96
0
S
5
0795
7/20/95
S
5
352
0795
7/22/95
0
S
5
352
07961
7/15/96
134.8
F
2
352
0796
7/25/96
70.8
F
3
352
0895
8/8/95
0185F
5
352
0895
8/20/95
S
2
352
LCDR David L. Brodeur
185.6
DESCRIP
OUTG LER
POWER REDUCTION DUE TO MAIN TURBINE
VALVE TESTING.
REACTOR WAS SHUTDOWN TO PERFORM
MAINTENANCE ON "C" DRAIN COOLER, "A"
RECIRCULATION PUMP SEAL, AND THE
CONDENSER WATERBOXES.
35296013
VALVE REACTOR SCRAM DUE TO A PRESSURE
X
SPIKE DURING THE PERFORMANCE OF A
MSIV SURVEILLANCE TEST.
CONR REACTOR POWER WAS REDUCED DUE TO
OD
CONTROL ROD PATTERN ADJUSTMENT.
CHTB REACTOR POWER WAS REDUCED AND THE
TURBINE WAS TAKEN OFF THE GRID DUE
RK
TO REPAIR OF 220KV A CIRCUIT BREAKER,
REACTOR REMAINED CRITICAL.
CONR REACTOR POWER WAS REDUCED DUE TO A
ROD PATTERN ADJUSTMENT.
OD
CONR REACTOR POWER WAS REDUCED DUE TO A
OD
ROD PATTERN ADJUSTMENT.
CRDR REACTOR POWER WAS REDUCED FOR
VE
CONTROL ROD HYDRAULIC CONTROL UNIT
MAINTENANCE
HTEXC REACTOR POWER WAS REDUCED DUE TO
CONDENSER WATERBOX CLEANING.
H
HTEXC REACTOR WAS SHUTDOWN TO REPAIR
H
UNISOLABLE LEAKS IN SJAE ROOM.
VALVE REACTOR WAS SHUTDOWN DUE TO A
X
ELECTRO-HYDRAULIC CONTROL TRANSIENT
OF TURBINE CONTROL VALVES.
INSTR REACTOR POWER REDUCED DUE TO
U
REACTOR FEEDWATER PUMP MIN FLOW
VALVE FAILED OPEN DUE TO A FAILED
POWER SUPPLY.
FUELX REACTOR WAS SHUTDOWN TO REPLACE A
X
FAILED FUEL BUNDLE.
Page 2
Bwr495-7.xls
Sample NRC/INEEL MORPH 2 data
RPT
DOCKET PERIOD
352
0895
OUTG
DATE
8/28/95
8/5/97
OUTG OUTG
HRS
TYPE
89.1
S
352
0896
8/2/96
0
F
352
0995
8/28/95
22
S
352
0995
9/2/95
13.5
F
352
0995
9/11/95
338.2
F
352
1095
10/21/95
S
352
1196
11/16/96
S
352
1295
12/27/95
S
352
1296
12/23/96
F
LCDR David L. Brodeur
OUTG OUTG OUTG OUTG
DESCRIP
METH REASN SYSTM COMP
PIPEX REACTOR WAS SHUTDOWN SHORTLY
AFTER BEING CRITICAL DUE TO LEAKAGE
X
INTO THE DRYWELL CAUSED BY A
MISALIGNED REACTOR PRESSURE VESSEL
INSTRUMENT FLANGE CONNECTION.
VALVE POWER REDUCTION DUE TO A FEEDWATER
HEATER DRAIN VALVE MALFUNCTION.
X
PIPEX CONTINUED IN A SHUTDOWN CONDITION
DUE TO LEAKAGE INTO THE DRYWELL
X
CAUSED BY A MISALIGNED REACTOR
PRESSURE VESSEL INSTRUMENT FLANGE
CONNECTION.
RECO UNIT SHUTDOWN IN ACCORDANCE WITH
MB
TECH. SPEC. 3.0.3 AS A RESULT OF
DISCOVERING BOTH POST-LOCA
HYDROGEN RECOMBINER SYSTEMS WERE
INOPERABLE DUE TO IMPROPER WIRING OF
CERTAIN RECORDERS DURING A RECENT
RECORDER MODIFICATION.
VALVE REACTOR WAS SHUTDOWN DUE TO A
X
FAILED OPEN SAFETY RELIEF VALVE (SRV).
OUTG LER
35295006
35295006
352950071
35295008
CONR
OD
REACTOR POWER WAS REDUCED FOR
CONTROL ROD PATTERN ADJUSTMENT AND
MAIN TURBINE VALVE TESTING.
CRDR POWER REDUCTION DUE TO HCU ON-LINE
VE
MAINTENANCE.
HTEXC POWER REDUCTION TO REMOVE THE 5TH
H
FEEDWATER HEATERS FROM SERVICE AND
TO PERFORM CONTROL ROD SCRAM TIME
TESTING.
VALVE POWER REDUCTION DUE TO A FAILED
X
TEMPERATURE CONTROL VALVE ON THE
STATOR WATER COOLING SYSTEM.
Page 3
Bwr495-7.xls
I
Sheet1
DOCKET
261
282
306
305
315
316
321
366
325
324
373
374
382
456
457
461
155
244
254
265
259
260
296
266
301
271
277
278
289
293
298
317
318
327
328
S333
4
RPT_PERIOD DAY1 DAY2 DAY3 DAY4 DAY5 DAY6 DAY7 DAY8 DAY9 DAY10 DAY11 DAY12 DAY13 DAY14 DAY1 5 DAY16
725
725
671
726
724
725
725
726
725
726
725
722
725
725
724
724
0195
531
532
531
531
531
531
534
533
530
532
530
531
532
532
531
529
0195
530
531
529
529
530
530
531
531
529
530
529
531
531
531
532
532
0195
520
528
528
520
520
524
524
524
524
524
524
524
524
524
524
524
0195
537
639
1008
1037
1033
916
1030
1030 1031 1030 1032 1028 1030 1030 1032 1029
0195
1088
1084
1080
1083
1089
1089
1083
1083 1085 1082 1079 1090 1086 1086 1089 1081
0195
760
768
769
767
756
772
768
767
762
765
774
773
774
770
771
773
0195
782
758
767
782
779
780
784
775
784
755
785
785 780
783
783
786
0195
789
771
791
791
790
792
791
791
791
792
790
792
791
791
791
776
0195
772
770
772
773
772
772
772
772
573
769
773 773
772
773
773
773
0195
1095
1094
1094
1096
1042
1096
1099
1081 1088 1076 1087 1099 1097 1098 1098 1098
0195
1097
1099
1099
1068
1072
1057
1037
956 1074 1100 1094 1091 1104 1102 1104 1076
0195
1083
1084
1081
1082
1083
1083
1082
1084 1082 1082 1082 1082 1081 1042 1080 1084
0195
1142
1145
1139
1143
1143
1142
1139
985 1140 1140 1140 1142 1142 1141 1142 1142
0195
1058
1132
1133
1130
1135
1136
1133
1127 1134 1134 1135 1135 1136 1135 1132 1133
0195
914
910
921
919
923
924
781
912
0
104
259
913
833
916
914
878
0195
67
67
66
67
66
66
66
67
66
67
67
66
67
66
66
66
0195
483
483
483
483
483
483
483
483
483 483
483
483 483
483
483
483
0195
171
142
143
-7
129
-9
-8
-8
-9
-8
-8
-8
-8
-8
-8
-8
0195
720
669
726
629
703
633
709
717
706
710
721
637
721
607
609
610
0195
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0195
1096
1101
1084
1098
1090
1094
1095
1097 1096 1092 1095 1095 1091 1086 1094 1097
0195
0
0
0
0
0
0
0
,0
0
0
0
0
0
0
0
0
0195
495
497
493
495
495
495
494
495
496
495
495
495
496
495
495
419
0195
486
466
473
498
486
486
486
485
486
486 485
486 486
484
485
485
0195
519
516
519
518
5-19
518
519
519
519
518
519
518
519
519
518
517
0195
1136
1127
1131
1131
1050
1131
1131
1126 1130 1130 1126 1126 1130 1082 1127 1131
0195
1072
1064
1067
1068
1063
1064
1068
1068 1064 1068 1064 1068 1064 1064 1068 1064
0195
815
806
811
822
821
823
825
823
823
824
821
825
823
824
821
823
0195
664
664
664
665
665
665
665
665
665 665
665 664
665
664
664
664
0195
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0195
864
864
863
866
866
867
867
867
866
866
865
865
866
865
867
867
0195
-30
-29
-30
376
869
869
868
867
868
868
867
867
867
868
867
867
0195
1152
1156
1153
1152
1152
1156
1157
1154 1154 1154 1151 1152 1154 1158 1157 1151
0195
1139
-37
652
849
1139
1139
1142
795 1139
282
-22 125
1143 1146 1143 1141
0195
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0195
1
9
1
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(
OUTG
OUTG
TYPE
DAYS
F
81
F
5.6
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
HTEXCH
MB
A
2
HTEXCH
HC
A
2
UNIT
ID
278
352
OUTG
DATE
7/14/92
7/15/96
OUTG
HRS
194
135
296
5/1/96
88
36
F
3
A
SJ
FCV
278
3/23/95
84
3.5
F
2
A
HC
VALVEX
271
331
1/15/92
1/24/93
13
140
06
58
F
S
1
1
B
B
ZZ
NN
XXXXXX
P
331
11/13/92
90
37
F
3
A
NN
P
366
9/2/95
57
24
F
2
A
HF
277
321
3/2/93
10/22/93
134
91
561
38
F
F
3
2
A
A
325
9/30/95
61
26
F
3
277
4/7/92
61
2 5
F
1
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
System
Air Removal
Air Removal
Category
EF
EF
Air Removal
EF
Air Removal
EF
Air Removal
Circ water
EF
Nature
HIGH CONDENSER BACKPRESSURE TURBINE TRIP AND 33192018
SCRAM CAUSED BY FAILURE OF CIRCULATING WATER
PUMP SHAFT. FAILURE ALLOWED THE FLOW FROM THE
REMAINING PUMP TO SHORT CYCLE BACK TO THE PUMP
PIT AND CUT OFF FLOW TO THE CONDENSER
Circ water
EF
HTEXCH
36695003
UNIT WAS MANUALLY SCRAMMED DUE TO A
DECREASING VACUUM ON THE "A" MAIN CONDENSER AS
A RESULT OF "D" WATERBOX BECOMING AIRBOUND
AFTER FILL MATERIAL IN CELL 10 OF COOLING TOWER
NO 5 COLLAPSED AND CLOGGED THE SCREENS AT THE
TOWER.
Circ water
EF
HH
HH
PUMPXX
INSTRU
27793004
SECOND CONDENSATE PUMP TRIP
A SIMULTANEOUS TRIP OF THREE CONDENSATE PUMPS 32193013
CAUSED A DECREASE IN FEEDWATER FLOW TO
REACTOR VESSEL AND CORRESPONDING DECREASE IN
REACTOR WATER LEVEL SHIFT INSERTED A MANUAL
REACTOR SCRAM ANTICIPATING AUTOMATIC REACTOR
SCRAM ON LOW REACTOR WATER LEVEL.
Condensate
Condensate
EF
EF
A
HH
PUMPXX
REACTOR SCRAM DUE TO LOW VESSEL LEVEL CAUSED
BY CONDENSATE SYSTEM TRANSIENT. THE TRANSIENT
WAS A RESULT OF LOSS OF CONDENSATE PUMP
SUCTION PRESSURE CAUSED BY EXCESSIVE AIR BEING
ADMITTED TO THE SUCTION HEADER OF THE
CONDENSATE PUMP.
32595018
Condensate
EF
A
HH
PIPEXX
RECOMB
CONDENSATE VENT LINE FAILURE STEAM LEAK ON
RECOMBINER FLOW TRANSMITTER
27792006
Condensate
EF
OUTG LER
DESCRIP
27892005
SJAE FLOW CONTROLLER FAILURE
REACTOR WAS SHUTDOWN TO REPAIR UNISOLABLE
LEAKS IN SJAE ROOM.
296/9603
REACTOR SCRAMMED DUE TO LOW REACTOR WATER
LEVEL FOLLOWING FAILURE OF THE STEAM PACKING
EXHAUSTER A BYPASS VALVE.
MANUAL SCRAM, LOSS OF VACUUM DUE TO STEAM
SUPPLY VALVE FAILURE TO AIR EJECTORS
STEAM JET AIR EJECTOR RUPTURE DIAPHRAGM REPAIR.
OUTAGE TO RECONNECT CIRCULATION WATER
PUMP.VERY COLD HIGH WINDS CAUSED ICING OF
CIRCULATION WATER SPRAY CREATING POTENTIAL FOR
DAMAGE TO THE COOLING TOWER FILL.RESTART
FOLLOWING THE 01/24/93 OUTAGE WAS DELAYED UNTIL
THE WINDS DECREASED.
Page 149
MIT POC DL BRODEUR
Appendix 7. BWR/4 Data, 1992 - 1996
Appendix 7.
UNIT
ID
341
BWRI4 Data, 1992
OUTG
OUTG
OUTG
IRS
D'ATE
1/29/93
331
6/1/95
271
12/9/93
353
12/18/96
115
341
2/10/93
82
260
12/17/93
OUTG
27 May 1998
7
Lost
7 Generation
Capacity by Failed System, Outage length
27 May 1998
OUTG
OUTG
G
2
F
241
50
119
SF
A
1
S
COMP
O
DAS
ISV
SG
TBG
HTEXCH
F
2
A
HC
HTEXCH
3.41
F
1
B
SG
COND
4
4
4-
4
4-
4.81
A
4
~
--
HC
--T-------
8/19/95
4
3
278
10/15/92
609
25 4
341
4/7/92
93
39
1/119
..
4
11/3/96
271
OUTG
Lost Generation Capacity by Failed System, Outage length
TYPEF MFTH RFAN SYSTEM
HRnonsS
EANSYTM
DATEU1EH
-01
27R
7
DESCRIP
YE
SCRAM DUE TO LOSS OF HEATER FEEDWATER PUMPS.
INADVERTENT OPENING OF CONDENSATE
DEMINERALIZER INLET VALVE RESULTED IN LOSS OF
HEATER FEED PUMP (HFP) AUCTION PRESSURE AND
CONSEQUENT TRIP OF HFP.
OUTG LER
1^44^~^4~
34192012
POWER REDUCED TO INVESTIGATE AND REPAIR
CONDENSER TUBE LEAKS. CONDENSER
DEMINERALIZER INFLUENT (CDI) CONDUCTIVITY
EXCEEDED ADMINISTRATIVE LIMITS MAIN
TURBINE/GENERATOR WAS MANUALLY TRIPPED.
FORCED OUTAGE FOR GENERATOR EXCITER FIELD
GROUND.
388
___ 4
OUTG
nAYS
57|
11/18/921
1996
-
.
F
4-
4
4
-.
~
PIPEXX
T-
I........
HTEXCH
1
A
IF
INSTRU
F
3
A
SH
VALVEX
F
1
A
NH
VACB
ii
System
Condensate
Condenser
Condenser
TUBE LEAK IN LOW PRESSURE CONDENSER.
SHUTDOWN TO DRAIN THE WATERBOXES, IDENTIFY THE
LEAKING TUBE, AND PLUG IT.
Condenser
MANUAL SHUTDOWN TO REPLACE EXPANSION JOINT ON
THE "A" MAIN CONDENSER DUE TO INCREASED AIR
INLEAKAGE
Condenser
REACTOR WAS SHUTDOWN DUE TO CRACK IN THE MAIN
CONDENSER NECK SEAL GASKET
t
Condenser
CONDENSER TUBE LEAK CAUSED CONDENSATE
CHEMISTRY TO REACH ACTION LEVEL PLANT
SHUTDOWN FOR TUBE PLUGGING.
Condenser
MANUAL SHUTDOWN TO REPAIR THE CONDENSER
EQUALIZING LINE
~onaenser
r.
~bU~UU(
ZSUO26095007 Condenser
Main Turbine TRIPPED ON LOW CONDENSER VACUUM
LEVEL
BOTH
TO
SUPPLY
CAUSED BY A FAILED POWER
CONTROL LOOPS FOR THE OFF GAS CONDENSER DRAIN
VALVES. REPLACED FAILED ELECTROYTIC CAPACITOR
IN THE POWER SUPPLY FOR THE OFFGAS CONDENSER
DRAIN VALVES
....
CLEAN CONDEN tKR WAI I-KtBUXt~
TURBINE TRIP DUE TO LOSS OF CONDENSER VACUUM;
CAUSED BY ATMOSPHERIC DRAIN TANK LEVEL CONTROL
Category
I rl~
I-HF/OA
'-e
EF
EF
EF
fEF
EF
Condenser
Condenser
EF/WP
EF
Containment
HF/OA
Containment
EF
SYSTEM PROBLEM
27892008
PCIS GROUP I ISOLATION CAUSED BY BUMPING
INSTRUMENTATION.
DURING PERFORMANCE OF A ROUTINE SURVEILLANCE 34192003
ON DRYWELL AND SUPPRESSION CHAMBER VACUUM
BREAKER OPERABILITY, A VACUUM BREAKER DID NOT
CLOSE AFTER BEING OPENED. THE VACUUM BREAKER
ACTUATOR (UTILIZED DURING TESTING ONLY) BOUND IN
THE OPEN POSITION.
Source, INEEL / NRC Morp2
Page 150
MIT POC DL BRODEUR
OUTG
OUTG
TYPE
DAYS
F
76
OUTG OUTG
METH REASN
B
2
OUTG
OUTG
COMP
SYSTEM
SF
VALVEX
UNIT
ID
366
OUTG
DATE
11/8/93
OUTG
HRS
183
298
98
324
2/10/92
4/19/92
4/21/92
131
124
9,338
5.5
52
389 1
F
S
F
2
2
1
A
A
A
EL
EL
NB
325
4/21/92
7,690
320 4
F
1
A
NB
298
5/25/94
6,522
271 7
F
1
H
EK
27
341
11/20/96
273
11 4
F
9
A
EK
RG
353
10/19/94
51
21
F
3
A
CH
INSTRU
331
11/16/94
352
14 7
S
2
B
BJ
ISV
354
5/26/92
112
47
F
2
A
352
8/28/95
111
4.6
S
9
A
CA
PIPEXX
260
341
9/25/92
12/24/96
102
96
4 3
40
F
F
1
9
B
A
BF
VACB
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
BTRY
BTRY
System
Core Spray
Category
EF
29892003
29892003
32492012
IDC
DC
Diesel
EF/EOL
EF/EOL
EF
32592012
DIESEL GENERATOR WALL 9D-1 WAS DECLARED
INOPERABLE AS A RESULT, BUSES E-5 AND E-6 WERE
DECLARED INOPERABLE BECAUSE OF TECH. SPEC.
3 0 3 THE UNIT WAS TAKEN TO COLD SHUTDOWN. MORE
REPAIRS MADE
Diesel
EF
29894009
Diesel
HF/PI
Diesel
EF
REACTOR SCRAM ASSOCIATED WITH A RELAY COIL
FAILURE COMBINED WITH AN INAPPROPRIATE ACTION
TAKEN BY AN OPERATOR DURING TESTING OF AN
EMERGENCY DIESEL GENERATOR
Diesel
HF/OA
SHUTDOWN FOR DRYWELL ENTRY TO VERIFY SOURCE
OF AND REPAIR UNIDENTIFIED DRYWELL LEAKAGE
FAILED DRYWELL TO SUPPRESSION CHAMBER DECAY
TEST, POWER WAS REDUCED TO 21% AND THE
REACTOR WAS MANUALLY SCRAMMED
Drywell
EF
35492006
Drywell
EF
35295006
REACTOR WAS SHUTDOWN SHORTLY AFTER BEING
CRITICAL DUE TO LEAKAGE INTO THE DRYWELL CAUSED
BY A MISALIGNED REACTOR PRESSURE VESSEL
INSTRUMENT FLANGE CONNECTION
Drywell
HF/C
UNIT SHUTDOWN TO REPAIR DRYWELL LEAK
UNIT SHUTDOWN TO REPAIR T23-F400J DRYWELL TO
TORUS VACUUM BREAKER
Drywell
Drywell
EF
EF
DESCRIP
SHIFT REMOVED THE MAIN GENERATOR FROM SERVICE,
AND A MANUAL SCRAM WAS INITIATED TO INVESTIGATE
INCREASED LEAKAGE INTO THE DRYWELL FLOOR DRAIN
SYSTEM INVESTIGATION REVEALED A BONNET
PRESSURE SEAL LEAK ON CORE SPRAY TESTABLE
CHECK VALVE 2E21-F006B
DEGRADED 250V BATTERIES
BATTERY REPLACEMENT OUTAGE
DIESEL GENERATOR WALL 9D-1 WAS DECLARED
INOPERABLE AS A RESULT, BUSES E-5 AND E-6 WERE
DECLARED INOPERABLE BECAUSE OF TECH. SPEC.
3.0 3 THE UNIT WAS TAKEN TO COLD SHUTDOWN.
REPAIR EHC SYSTEM, MAIN TURBINE, REACTOR
FEEDPUMP AND THE 4A FEEDWATER HEATER.
EDG 1 AND EDG 2 DECLARED INOPERABLE DUE TO
INSUFFICIENT UV RELAY TESTING
UNIT REMAINED SHUTDOWN BEYOND THE 53 DAY
SCHEDULED REFUEL OUTAGE DUE TO EDG 11
AUTOMATIC VOLTAGE REGULATOR FAILURE.
Page 151
OUTG LER
34196023
MIT POC DL BRODEUR
UNIT
ID
321
OUTG
DATE
4/30/92
333
2/22/96
OUTG
DAYS
35
OUTG
HRS
84
OUTG
TYPE
SS
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
OUTG OUTG
METH REASN
I
OUTG
SYSTEM
I
!
OUTG
COMP
VALVEX
System
Drywell HVAC
Category
EF
IEHC
EF
PARTIAL CLOSURE OF Main Turbine GOVERNOR VALVES 29894004
DUE TO Main Turbine CONTROL SYSTEM MALFUNCTION
RESULTING INA REACTOR HIGH FLUX AND SUBSEQUENT
AUTOMATIC SCRAM. REPLACED FAILED POWER
SUPPLIES
EHC
EF
REACTOR WAS SHUTDOWN DUE TO A FAILED PRESSURE 35396007
EHC
EF
EHC
EF
DESCRIP
THE UNIT WAS SHUTDOWN TO INVESTIGATE THE CAUSE
OF INCREASING TEMPERATURES INTHE UPPER
REGIONS OF THE DRYWELL PERSONNEL DISCOVERED
THE AIR SUPPLY DAMPER TO ONE OF THE COOLING
UNITS HAD CLOSED DUE TO A LOOSE WING NUT ON THE
DAMPER
OUTG LER
I
-
j
1
324
2/2/921
287
1201
F
13
A
298
3/2/94
257
107
F
3
A
JJ
RJX
353
12/6/96
185
7 7
F
2
A
HA
INSTRU
TG
WHILE PERFORMING A CONTROLLED REACTOR
SHUTDOWN DUE TO EXCESSIVE SCRAM TIME, AN EHC
LINE TO TURBINE BYPASS VALVES RUPTURED
OPERATORS INSERTED A MANUAL SCRAM EHC TUBING
WAS MODIFIED WITH FLEXIBLE TUBING AND SCRAM
SOLENOID PILOT VALVE DIAPHRAGMS REPLACED.
33396002
REACTOR SCRAMMED DURING CONTROL VALVE
TESTING DUE TO ELECTROHYDRAULIC CONTROL (EHC)
SYSTEM FAILURE SUSPECTED CAUSE WAS AIR OR
NITROGEN IN THE SYSTEM CAUSED BY ACCUMULATOR
PERFORMANCE OR VENTING
32492001
32492001
SWITCH ON THE EHC SYSTEM
331
5/29/94
180
75
F
1
B
TG
TBG
FATIGUE INDUCED WELD CRACK ON AN
ELECTROHYDRAULIC CONTROL OIL SUPPLY LINE TO THE
#2 TURBINE CONTROL VALVE. A 0 5 GPM HYDRAULIC
LEAK WAS DISCOVERED DURING OPERATOR ROUNDS
REPAIR OF VARIOUS BALANCE OF PLANT STEAM LINE
VALVE PACKING LEAKS
331
7/10/94
148
6.2
F
2
A
JI
TBG
CRACK IN FLUID SUPPLY LINE TO TURBINE CONTROL
VALVE ELECTRO-HYDRAULIC SYSTEM REQUIRED
SHUTDOWN TO REPLACE DAMAGED SECTION OF
TUBING AND INSTALLATION OF HYDRAULIC
ACCUMULATORS INTHE SUPPLY LINE
33194010
EHC
EF
325
2/29/92
121
50
F
3
A
TG
94
EHC
EF
353
1/3/93
103
43
F
3
A
CC
INSTRU
REACTOR SCRAMMED WHILE STOP VALVE TESTING WAS 32592005
IN PROGRESS CAUSED BY A DEFECTIVE RELAY IN THE
ELECTROHYDRAULIC CONTROL SYSTEM
35393001
REACTOR AUTOMATICALLY SCRAMMED ON HIGH
REACTOR PRESSURE CAUSED BY MAIN TURBINE
CONTROL VALVE CLOSURE DUE TO AN UNDETERMINED
DHC MALFUNCTION
EHC
EF
354
5/16/93
100
42
F
3
A
UNIT TRIPPED DUE TO FAULTY COMPONENT IN EHC
EHC
EF
Source, INEEL t NRC Morp2
Page 152
MIT POC DL BRODEUR
OUTG
OUTG
TYPE
DAYS
F
42
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
SEAL
TG
A
2
UNIT
ID
333
OUTG
DATE
12/15/96
OUTG
HRS
100
325
7/13/95
93
3 9
F
3
A
HA
TURBIN
321
1/4/96
87
36
F
3
A
HB
FILTER
352
7/25/96
71
30
F
3
A
HA
VALVEX
296
2/29/96
66
2 7
F
3
A
JJ
CNV
321
11/11/93
64
2.7
F
2
B
HA
PIPEXX
333
9/24/93
64
2 7
F
3
A
JI
353
8/20/95
60
2 5
F
3
A
321
4/30/96
59
24
F
9
321
5/22/92
56
23
F
3
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
EHC
Category
EF
EHC
EF
EHC
EF
EHC
EF
EHC
EF
SHIFT REMOVED THE MAIN GENERATOR FROM SERVICE
AND A MANUAL SCRAM WAS INITIATED TO REPAIR AN
EHC FLUID LEAK ON A ONE INCH FLUID SUPPLY LINE
THE LEAK WAS REPAIRED AND THE UNIT WAS
RETURNED TO SERVICE
EHC
EF
94
33393020
DURING GROUND FAULT TESTING OF THE TURBINE
CONTROL SYSTEM, A BYPASS VALVE ALARM/TRIP RELAY
LEAD WAS MISTAKENLY LIFTED CAUSING #2 BYPASS
VALVE TO CLOSE AND A REACTOR TRIP ON HIGH
PRESSURE.
EHC
HF/OA
HA
INSTRU
REACTOR WAS SHUTDOWN DUE TO A HIGH IMPEDANCE
ACROSS THE EHC CONTROL RELAY CONTACT
RESULTING IN SPORADIC OPENING AND CLOSING
TURBINE BYPASS VALVES
35395010
EHC
EF
F
HB
PIPEXX
DURING STARTUP, THE UNIT EXPERIENCED HYDRAULIC 32196008
FLUID LEAKS ON A MAIN TURBINE CONTROL VALVE AND
TURBINE STOP VALVE. MEASURES TAKEN TO ISOLATE
THE LEAKS RESULTED IN A PARTIAL LOSS OF
HYDRAULIC FLUID FLOW TO THE MAIN TURBINE BYPASS
VALVES.
EHC
EF
A
HB
FILTER
REACTOR SCRAM WHEN DEBRIS, CAUSED BY MATERIAL 32192014
DEGRADATION OF FILTERS INTHE MAIN TURBINE'S
ELECTRO-HYDRAULIC CONTROL SYSTEM, RESTRICTED
FLUID FLOW DURING WEEKLY TURBINE STOP VALVE
TESTING. FILTER REPLACED
EHC
EF/WP
OUTG LER
DESCRIP
EHC HYDRAULIC FLUID LEAK ON NO. 1 TURBINE BYPASS 33396014
VALVE ACTUATOR SEAL.
32595015
REACTOR SCRAM DUE TO AN ERRATIC PRESSURE
ERROR SIGNAL FROM EHC
32196001
AN AUTOMATIC REACTOR SCRAM OCCURRED ON HIGH
REACTOR PRESSURE WHEN ALL FOUR MAIN TCVS
DRIFTED CLOSED DUE TO THE VALVES' SERVO
STRAINERS BECOMING CLOGGED, CAUSING LOSS OF
HYDRAULIC FLUID PRESSURE TO THE SERVO VALVE
SPOOL
REACTOR WAS SHUTDOWN DUE TO A ELECTROHYDRAULIC CONTROL TRANSIENT OF TURBINE
CONTROL VALVES.
A FAILED TURBINE SPEED FEEDBACK CARD IN THE
ELECTRO-HYDRAULIC CONTROL SYSTEM CAUSED
FLUCTUATION IN THE TURBINE CONTROL AND BYPASS
VALVES. THIS CAUSED A REACTOR PRESSURE SPIKE,
CAUSING AN AVERAGE POWER RANGE MONITOR HIGH
FLUX SPIKE, SCRAMMING THE REACTOR.
Page 153
29696001
System
MIT POC DL BRODEUR
OUTG
HRS
OUTO OUTG
TYPE
DAYS
F
19
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
PIPEXX
HA
A
1
UNIT
ID
353
OUTG
DATE
11/20/92
366
11/24/92
45
19
F
1
B
HA
VALVOP
353
3/26/93
42
17
F
3
A
HA
TURBIN
324
1/10/92
41
1.7
S
1
B
TG
TG
354
6/13/92
19
0 8
F
1
A
353
5/1/96
19
08
F
1
A
354
6/21/94
16
07
F
1
A
321
5/26/96
10
04
F
1
A
HA
PIPEXX
353
5/15/93
8
03
F
1
A
HA
PIPEXX
353
6/3/95
6
02
S
3
A
HA
VALVEX
388
2/20/94
3
01
F
1
A
TG
V
352
4/2/96
3
01
F
1
A
HA
INSTRU
388
7/14/96
438
18.2
F
2
H
EB
CBL1
278
7/4/92
219
91
F
3
A
EA
RELAYX
47
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
HA
VALVEX
EHC
Category
EF
EHC
EF
35393005
THE UNIT AUTOMATICALLY SHUTDOWN DUE TO AIR
ENTRAINED IN THE MAIN TURBINE ELECTRO HYDRAULIC
CONTROL SYSTEM
MAIN GENERATOR REMOVED FROM GRID TO PERFORM 24792002
MAINTENANCE ON ELECTROHYDRAULIC CONTROL
SYSTEM (EHC) AND TO CORRECT EXCESSIVE VIBRATION
ON GENERATOR EXCITER BEARING REGULATOR FOR
EHC WAS REPAIRED AND EXCITER BEARING WORKED.
EHC
HF
EHC
EF
UNIT WAS TAKEN OFF LINE TO REPAIR EHC LEAK THE
REACTOR WAS KEPT AT APPROXIMATELY 3% POWER
FOR THE DURATION OF THE OUTAGE.
THE TURBINE WAS TAKEN OFF THE GRID DUE TO REPAIR
OF TURBINE EHC LEAK.
POWER REDUCTION TO REPAIR EHC LEAK ON #2
BYPASS VALVE
SHIFT MANUALLY TRIPPED THE MAIN TURBINE TO
REPAIR AN EHC SYSTEM FLUID LEAK.
REACTOR POWER WAS REDUCED TO 19% AND THE
TURBINE TAKEN OFF LINE TO REPAIR A LEAK INTHE
ELECTRO-HYDRAULIC CONTROL SYSTEM.
EHC
EF
EHC
EF
EHC
EF
EHC
EF
EHC
EF
MAIN TURBINE REMOVED FROM SERVICE TO REPAIR
EHC LEAK AT #4 CIV.
THE MAIN GENERATOR WAS TAKEN OFF LINE TO REPAIR
EHC LEAK ON THE #3 CONTROL VALVE.
MAIN TURBINE WAS TAKEN OFF THE GRID TO REPLACE
THE EHC SPEED CONTROLLER CARD.
38896004
REACTOR SCRAMMED WHEN ALL FEEDWATER WAS
LOST DURING POST MAINTENANCE TESTING OF TIE
BUS 0A107, POWER WAS LOST TO AUXILIARY BUS 12A
THIS CAUSED 2 CONDENSATE PUMPS TO TRIP AND DUE
TO LOW SUCTION PRESSURE ALL 3 RFP'S TRIPPED
EHC
EF
EHC
EF
EHC
EF
Electrical
HF
Electrical
EF
DESCRIP
THE TURBINE GENERATOR WAS TAKEN OFF LINE TO
REPAIR AN EHC LEAK ON THE #3 MAIN TURBINE
CONTROL VALVE
THE MAIN TURBINE WAS TAKEN OFF LINE TO REPAIR AN
ELECTRO-HYDRAULIC CONTROL SYSTEM FLUID LEAK AT
THE REHEAT CYLINDER ON COMBINED INTERMEDIATE
VALVE NO 4.
AUTO SCRAM - #1 TRANSFORMER FAILURE
Page 154
System
OUTG LER
27892010
MIT POC DL BRODEUR
Appendix~~~~~
96Ls
UNIT
ID
333
OUTG
DATE
9/16/96
OUTG
HRS
177
388
3/18/921
144
9/7/93
T
7I
eeainCpct
aa 92
OUTG
OUTG
OUTG OUTG
SYSTEM
COMP
METH REASN ---------~-C
SSE
3
H
OUTG OUTG
DAYS
TYPE
74
F
I
t
I
I
6.01
F
RLY-87
561
F
CKTBRK
325
1/1
7/92
3(6
6/25/92
271
10/4/94
54
23
S
1
B
277
8/17/92
47
19
F
3
341
2/19/93
32
13
F
352
6/17/96
7
0.3
341
3/27/96
526
21 9
~-
Source, INEEL / NRC Morp2
I2U~
yFie
ytm
uaelnt
DESCRIP
Ci-l-/
DURING PERFORMANCE OF 345KV RELAY CALIBRATION
TWO TERMINALS WERE INADVERTENTLY SHORTENED
CAUSING THE 10042 AND 10052 BREAKERS TO OPEN
LEADING TO A SCRAM WORK PROCESS IS BEING
REVIEWED FOR IMPACT ON PLANT OPERATIONS.
--
i-------~--------~-
;
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
DURING PERFORMANCE OF OP PROCEDURE TO SWAP
INTHE E DIESEL GENERATOR, AN OPERATOR FOUND
THE PROTECTIVE RELAY ON THE C DIESEL PANEL
TRIPPED FURTHER PROBLEMS RESULTED IN THE
POTENTIAL FOR INBOARD MSIVS TO GO CLOSED,
CAUSING UNIT TO BE MANUALLY SCRAMMED
OUTG LER
~--I
er System
33396010
Electrical
38892001
Electrical
Category
I.
HF/PI
I
325 2UU3
32592003
Electrical
EF
Licria
M.c.r.
IIUP,
~.
I---
Electrical
CKTBKR
PERSONNEL ERROR WHILE SEARCHING FOR A GROUND 36692009
ON LPCI INVERTER 2R24-S018A, RESULTING IN THE
SUPPLY BREAKER OPENING TO 600V BUS C, WHICH
CAUSED A LOSS OF CONTROL POWER TO THE REACTOR
FEED PUMPS, FOLLOWED BY AN AUTOMATIC REACTOR
SCRAM.
EB
INSTRU
Electrical
EF
H
EA
CKTBKR
Electrical
HF
3
G
KE
VITAL AC AUTO BUS TRANSFER PROBLEM. REPAIRS
MADE TO VOLTAGE REGULATOR
27792015
REACTOR SCRAM AS A RESULT OF PROBLEMS
ENCOUNTERED DURING THE BLOCKING OF BREAKERS
ROUTINE PUMP BREAKER PM TESTING INADVERTENTLY 34193004
ACTUATED IN-SERVICE TRIP RELAYS. AN IN-SERVICE
PUMP BREAKER TRIP RELAY FAILED TO PROPERLY
ACTUATE, LEADING TO INABILITY TO TRANSFER FEED TO
ALTERNATE SUPPLY
Electrical
EF
F
1
A
EG
CHTBRK
Electrical
EF
F
2
B
BI
TK
REACTOR POWER WAS REDUCED AND THE TURBINE
WAS TAKEN OFF THE GRID DUE TO REPAIR OF 220KV A
CIRCUIT BREAKER, REACTOR REMAINED CRITICAL.
34196005
TECH SPEC REQUIRED SHUTDOWN DUE TO BOTH
DIVISIONS OF EECW BEING DECLARED INOPERABLE DUE
TO MAKE-UP TANK DESIGN ISSUE MODIFICATION BEING
INSTALLED
ESW
EF/WD
Page 155
7My19
Electrical
Electrical
AUTOMATIC REACTOR SCRAM OCCURRED AFTER LOSS
OF AN OFFSITE POWER SUPPLY DURING THE
AUTOMATIC TRANSFER TO THE SECONDARY POWER
SUPPLY, BREAKER FOR THE 1A FEEDWATER CONTROL
SYSTEM FAILED TO RECLOSE, RESULTING IN A
REDUCTION IN REACTOR WATER LEVEL.
A FAILURE OF THE UNINTERRUPTIBLE POWER SUPPLY
CAUSED THE UNIT 1 REACTOR TO SCRAM. THE POWER
SUPPLY WAS REPAIRED AND THE REACTOR WAS
RETURNED TO SERVICE
-
IWI
Electrical
uHF/OA
MIT POC DL BRODEUR
UNIT
ID
387
OUTG
DATE
6/5/92
354
10/2/94
OUTG
HRS
443
OUTG
OUTG
DAYS TYPE
F
185
OUTG
OUTG
COMP
SYSTEM
V
SJ
OUTG OUTG
METH REASN
B
2
.1__
__
9/5/95
298
12/14/93
354
5/15/94
OUTG LER
38792010
Category
EF
System
FW
EF/WD
MANUALLY SHUTTING DOWN DUE TO A LEAK ON THE "B"
FEEDWATER DISCHARGE HEADER PIPING
~
-33393009
XXXX
SHUTDOWN DUE TO LOSS OF FEEDPUMP "A"SPEED
CONTROL DUE TO A SHORTED TERMINAL STRIP. THE
TERMINAL STRIP WAS REPLACED.
AN INADVERTENT REMOVAL OF A FEEDWATER CONTROL 33395013
SC
JB
FUSE CAUSED A FEEDPUMP TRANSIENT AND PLANT
SCRAM ON LOW WATER LEVEL
29893038
FEEDWATER LEVEL CONTROL RFC-LC-83 FAILED,
RESULTING INA REACTOR LOW LEVEL AND
SUBSEQUENT AUTOMATIC SCRAM.
i
------------AUTOMATIC SCRAM DURING DIGITAL FEEDWATER
TESTING.
PUMPXX
REACTOR FEED PUMP TRIPPED, OTHER PUMP FAILED TO 27883002
-1
CONROD
L __
4 __
__
t
_
_
_4
4/21/93
3331
DESCRIP
THE UNIT WAS TAKEN OFF LINE TO REPAIR THE "A"
REACTOR FEED PUMP ISOLATION VALVES THE VALVES
WERE REPAIRED AND A STARTUP COMMENCED. THE
STARTUP WAS HALTED AND THE UNIT MANUALLY
SCRAMMED DUE TO AN IGNITION OF CHARCOAL IN THE
1B OFFGAS GUARD BED.
AUTOMATIC SCRAM CAUSED BY DESIGN ERROR IN
DIGITAL FEEDWATER CONTROL SYSTEM. WHEN
ATTEMPTING RESTART AN EHC SYSTEM PROBLEM
CAUSE A PROBLEM WITH THE TURBINE ROLL
OPERATOR CLOSED ALL TURBINE VALVES WHICH
RESULTED IN AUTO REACTOR SCRAM.
4/6/93
__
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
178
153
----
74
F
661
F
1
3
G
3
A
I
t--~
--
6.4
-
278
3/7/93
119
5.0
F
3
2601
5/10/96
1001
4.21
F
3
~------
L
FW
EF
IF-W
HF/OA
FW
EF
FW
START
H
A
10/11/94
3211
3/28/921
I
6
361
F
13 1G B 1
JB
ISK
CHCH GENERA
EB
1CKRBKR
TRANSF
SI
Source, INEEL / NRC Morp2
I
REACTOR SCRAMMED AUTOMATICALLY ON MAY 10,
1996, DUE TO LOW REACTOR WATER LEVEL DUE TO
ZERO DEMAND SIGNAL THAT RESULTED FROM
REINITALIZATION OF THE REACTOR FEED PUMP
FEEDWATER LEVEL CONTROL SYSTEM ROOT CAUSE
WAS INADEQUATE DESIGN
EF/WD
260/9605
AUTOMATIC SCRAM/HIGH REACTOR WATER LEVEL DUE 27894005
TO FEED PUMP CONTROL PROBLEMS CAUSE BY LOSS
OF THE STATIC INVERTER
32192009
SHIFT PERSONNEL MISTAKENLY OPENED THE SUPPLY
BREAKER TO 600V BUS 1B, CAUSING A MOMENTARY
LOSS OF CONTROL POWER TO THE REACTOR
FEEDWATER PUMPS THIS RESULTED IN AN AUTOMATIC
REACTOR SCRAM ON LOW WATER LEVEL
"COMBUSTIBLE GAS" ALARM WAS RECEIVED
'--------i
'''HF/OA
HF/OA
_
___
Page 156
MIT POC DL BRODEUR
OUTG
HRS
OUTG OUTG
TYPE
DAYS
32
F
OUTG OUTG
METH REASN
A
3
OUTG
OUTG
COMP
SYSTEM
GR
SL
UNIT
ID
331
OUTG
DATE
5/14/95
353
8/8/95
70
29
F
3
A
CH
INSTRU
271
12/8/95
69
29
F
3
A
CH
VALVEX
278
7/30/95
68
2 8
F
3
A
HC
VALVEX
366
4/25/96
62
26
F
1
A
HH
PIPEXX
387
11/12/92
57
24
F
3
A
SJ
RLY
260
4/27/92
41
17
F
3
A
321
5/26/96
38
1.6
F
9
A
ED
INSTRU
296
4/21/96
33
1.4
F
3
A
SJ
341
341
341
4/10/93
12/5/92
9/17/93
240
210
70
100
8.7
2.9
S
S
S
1
1
2
A
A
B
324
4/27/96
23
10
F
1
321
10/29/93
19
08
F
1
78
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
FW
Category
EF
FW
EF
FW
EF
FW
EF
FW
EF
38792017
FW
EF
AUTOMATIC FEEDWATER LEVEL CONTROLLER FAILED
REACTOR SCRAMMED ON LOW WATER LEVEL
32196009
SHIFT MANUALLY SCRAMMED REACTOR WHEN BOTH
REACTOR FEEDWATER PUMPS TRIPPED AND REACTOR
WATER LEVEL DECREASED. REACTOR FEEDWATER
PUMPS TRIPPED DUE TO FAILED A BOARD REPLACED A
BOARD
FW
EF/ICS
FW
EF
FCV
REACTOR SCRAM DUE TO THE INADVERTENT TRANSFER 29696002
OF OIL FROM THE "3C" FEEDWATER PUMP TURBINE OIL
TANK, RESULTING IN A TRIP OF "3C" FEEDWATER PUMP
FW
EF
SE
SE
SN
EXJ
EXJ
LCV
REPAIR OF EXTRACTION STEAM LINE RUPTURE
REPAIR OF EXTRACTION STEAM LINE RUPTURE
34193013
WHILE SHUTTING DOWN TO REPAIR A HEATER DRAIN
SYSTEM LEVEL CONTROL VALVE, PRESSURE INTEGRITY
WAS LOST DUE TO MAINTENANCE ACTIVITIES ON THE
VALVE THIS RESULTED IN LEAKAGE FROM THE
FEEDWATER SYSTEM WHICH WAS TERMINATED AFTER
THE REACTOR WAS SCRAMMED.
FW heating
FW heating
FW heating
EF
EF
HF
A
CH
VALVEX
FW heating
EF
B
HH
VALVEX
TURBINE REMOVED FROM SERVICE TO REPAIR
FEEDWATER HEATER LEVEL CONTROL VALVE
SHIFT REMOVED THE MAIN TURBINE FROM SERVICE TO
COMPLETE REPAIRS ON A STUCK CHECK VALVE IN THE
NORMAL DRAIN FROM THE 2ND STAGE OF THE C/D MSRS
TO THE 5TH STAGE "B"FEEDWATER HEATER THE VALVE
WAS REPAIRED
FW heating
EF
OUTG LER
DESCRIP
33195005
ITHE TRIP OF THE "B"RFP WAS DUE TO STRIPPING THE
INTERNAL GEARS OF THE COUPLING BETWEEN THE
REACTOR FEED PUMP SHAFT AND LUBE OIL PUMP.
35395008
REACTOR WAS SHUTDOWN DUE TO A FAILED POWER
SUPPLY IN THE FEED WATER CONTROL SYSTEM
27195021
TURBINE TRIP/REACTOR SCRAM DUE TO
MALFUNCTIONING FEEDWATER REGULATOR VALVE.
FEEDWATER TRANSIENT, HIGH REACTOR LEVEL SCRAM.
SHIFT MANUALLY TRIPPED THE MAIN TURBINE AND
INSERTED A MANUAL SCRAM TO REPAIR A LEAK ON A
REACTOR FEEDWATER VENT LINE.
THE SCRAM WAS CAUSED BY A FAULTY RELAY IN ONE
DIVISION OF THE RFP TURBINE, MAIN TURBINE HI LEVEL
TRIP CIRCUIT WHILE A SURVEILLANCE WAS BEING
PERFORMED INTHE OTHER DIVISION OF THE HI LEVEL
TRIP LOGIC.
Page 157
System
MIT POC DL BRODEUR
OUTG
HRS
OUTG OUTG
TYPE
DAYS
F
02
4
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
SJ
V
A
1
UNIT
ID
388
OUTG
DATE
6/12/94
366
12/6/92
185
77
S
1
B
HJ
GENERA
388
4/15/95
150
6.3
F
3
A
FK
BKR
354
12/1/93
132
5 5
F
2
A
260
10/29/96
121
50
F
3
A
TL
EXC
321
3/29/94
84
35
F
3
A
HA
GENERA
LOSS OF MAIN GENERATOR FIELD EXCITATION LED TO
LOAD REJECTION BY TURBINE-GENERATOR SYSTEM
AND RESULTED INAN AUTOMATIC REACTOR
SHUTDOWN FIELD EXCITATION WAS LOST WHEN
ARCING OCCURRED BETWEEN THE MGE BRUSH
RIGGING AND A COLLECTOR RING ON THE MGE ROTOR
278
277
2/3/94
10/15/96
81
66
34
27
F
F
2
3
A
A
HA
HA
GENERA
RELAYX
278
2/2/96
50
21
F
1
A
HA
GENERA
296
11/27/95
48
2.0
S
2
B
277
10/6/96
33
14
F
3
A
260
2/9/95
30
13
F
3
H
277
271
12/18/92
2/9/94
28
6
1.2
0.2
F
F
1
1
A
B
296
11/26/95
1
00
F
1
B
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
HA
HA
EB
System
FW heating
Category
EF
Generator
EF
38895005
Generator
EF
UNIT SHUTDOWN DUE TO EXCESSIVE ARCING OF THE
MAIN GENERATOR EXCITER BRUSHES
THE UNIT 2 MAIN GENERATOR FIELD COLLAPSED DUE TO 26096007
AN EXCITER MALFUNCTION, AND THE RESULTANT
VOLTAGE AND CURRENT CONDITION CAUSED THE
GENERATOR BACKUP RELAYS TO OPERATE.
Generator
EF
Generator
EF
32194003
Generator
EF
MAIN GENERATOR FIELD GROUND RESISTOR
MAIN GENERATOR NEGATIVE PHASE RELAY OPERATION.
Generator
Generator
EF
EF
GENERATOR TAKEN OFF LINE FOR A MAIN GENERATOR
HYDROGEN LEAK.
MANUAL SCRAM OF UNIT 3 REACTOR REQUIRED BY
TESTING SCHEDULE ALSO MAIN GENERATOR HAD
INSUFFICIENT COOLING FLOW THROUGH THE EXCITER
COOLER.
Generator
EF
Generator
EF
RELAYX
MAIN GENERATOR NEGATIVE PHASE RELAY OPERATION.
Generator
EF
Generator
EF
GENERA
ELECON
26095002
AUTOMATIC SCRAM CAUSED BY Main Turbine
GENERATOR EXCITER GROUND RELAY TRIPPING.
REPAIR GENERATOR HYDROGEN LEAK
DURING A ROUTINE INSPECTION, DISCOVERED NEUTRAL
GROUND ON THE MAIN GENERATOR DISCONNECTED
GENERATOR WAS TAKEN OFF LINE TO MAKE THE
CONNECTION.
Generator
Generator
EF
HF/C
TRIPPED MAIN TURBINE FOR MAINTENANCE ON MAIN
IGENERATOR CURRENT TRANSFORMER CIRCUITS
Generator
EF
DESCRIP
REMOVED THE MAIN TURBINE FROM SERVICE TO
REPAIR A STEAM LEAK ON FW HEATER BTV-20210B.
THE UNIT WAS SHUTDOWN DUE TO A HYDROGEN LEAK
AT THE NEUTRAL BUSHING ON THE MAIN GENERATOR
THE BUSHING WAS REPLACED AND TESTED FOR
LEAKAGE, AND THE UNIT WAS RETURNED TO RATED
THERMAL POWER
AUTOMATIC REACTOR SCRAM DUE TO A MAIN
GENERATOR LOAD REJECT REPLACED "A"REACTOR
RECIRC PUMP SEAL AND "B" MAIN TRANSFORMER
BUSHING
Page 158
OUTG LER
MIT POC DL BRODEUR
OUTG
HRS
OUTG OUTG
TYPE
DAYS
F
14
OUTG OUTG
METH REASN
B
2
OUTG
OUTG
COMP
SYSTEM
INSTRU
HC
UNIT
ID
278
OUTG
DATE
7/30/93
352
9/2/95
14
06
F
1
A
SE
RECOMB
333
5/19/93
128
54
F
1
A
BJ
V
341
4/19/96
108
45
F
2
B
BJ
PC
260
4/15/94
146
61
F
3
B
325
11/17/94
6
0.3
S
1
B
HA
TURBIN
341
11/20/96
504
21 0
F
4
A
AC
RV
352
9/11/95
338
141
F
2
A
CC
VALVEX
352
3/24/96
158
66~
S
2
B
CC
VALVEX
387
10/28/96
132
55
F
4
A
SB
PSF
366
3/12/96
83
3 4
S
2
A
CC
VALVEX
352
5/21/96
51
2.1
F
3
B
CD
VALVEX
341
12/22/96
48
20
F
9
A
AC
RV
341
4/20/93
33
14
F
9
H
33
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
DESCRIP
MANUAL SCRAM DUE TO RECOMBINER ISOLATION AND
SUBSEQUENT LOSS OF CONDENSER VACUUM
UNIT SHUTDOWN INACCORDANCE WITH TECH SPEC
3.0 3 AS A RESULT OF DISCOVERING BOTH POST-LOCA
HYDROGEN RECOMBINER SYSTEMS WERE INOPERABLE
DUE TO IMPROPER WIRING OF CERTAIN RECORDERS
DURING A RECENT RECORDER MODIFICATION
SHUTDOWN DUE TO HPCI CHECK VALVE LEAK CAUSED
BY FAILED PRESSURE SEAL.
DURING UNIT STARTUP HPCI AND RCIC DECLARED
INOPERABLE, TECH SPEC REQUIRED SHUTDOWN.
DURING PLANNED MAINTENANCE ACTIVITIES ON THE
SCRAM PILOT AIR HEADER, UNIT 2 AUTOMATICALLY
SCRAMMED ON LOW SCRAM AIR HEADER PRESSURE
FOLLOWING ISOLATION OF BOTH PRIMARY AND
SECONDARY SCRAM PILOT AIR HEADER PRESSURE
REGULATORS
MANUALLY TRIPPED MAIN TURBINE TO FACILITATE THE
REMOVAL OF THE ISOPHASE BUS DUCT COVER
INSPECTION PLATE.
UNIT SHUTDOWN TO REPAIR SRV'A' TAIL PIPE
PRESSURE SWITCHES.
REACTOR WAS SHUTDOWN DUE TO A FAILED OPEN
SAFETY RELIEF VALVE (SRV).
REACTOR WAS SHUTDOWN DUE TO SRV AND TIP
MACHINE MAINTENANCE.
FORCED OUTAGE ACTIVITIES INCLUDED REPLACEMENT
OF A PORTION OF THE MAIN STEAM LINE DRAIN PIPING,
ALIGNMENT CHECKS ON THE "B" REACTOR
RECIRCULATION PUMP AND INSTALLING A MAIN
TURBINE/GENERATOR BALANCE SHOT
System
H2 Recombiner
Category
EF
H2 Recombiner
HF/PI
HPCI
EF
34196007
HPCI
EF
26094004
Instrument Air
HF/ICS
Isophase bus
cooling
OM
Main Steam
EF
Main Steam
EF
Main Steam
EF
Main Steam
EF
Main Steam
EF
Main Steam
HF
Main Steam
EF
Main Steam
HF/OA
OUTG LER
35295007
35295008
THE UNIT WAS MANUALLY SCRAMMED TO REPLACE
MAIN STEAM LINE SRVS "D"AND "H"
35296013
REACTOR SCRAM DUE TO A PRESSURE SPIKE DURING
THE PERFORMANCE OF A MSIV SURVEILLANCE TEST.
UNIT SHUTDOWN TO REPAIR SRV'D' SOLENOID
ACTUATOR.
34193007
SCRAM OCCURRED DURING RECOVERY FROM
EXTRACTION STEAM LINE REPAIR OUTAGE. SCRAM
CAUSED BY INCORRECTLY INSTALLED TEST
INSTRUMENT WHICH LEAKED STEAM AND WATER ONTO
MAIN STEAM MANIFOLD PRESSURE TRANSMITTERS
Page 159
MIT POC DL BRODEUR
Appendix 7. BWR/4 Data, 1992 - 1996
OUTG
HRS
OUTG OUTG
DAYS
TYPE
13
F
OUTG OUTG
METH REASN
A
2
OUTG
OUTG
COMP
SYSTEM
XXXXXX
HB
UNIT
ID
324
OUTG
DATE
3/13/96
341
12/25/93
1,835
76 5
F
3
A
TA
TRB
387
7/12/93
1,203
501
F
3
A
TA
TRB
UNIT ONE EXPERIENCED AN AUTOMATIC MAIN TURBINE
TRIP WITH AUTOMATIC REACTOR SCRAM MAIN
TURBINE TRIPPED ON HIGH VIBRATION CAUSED BY
FAILURE OF TWO TURBINE BUCKETS ON THE C LOW
PRESSURE ROTOR.
341
2/13/95
663
27 6
F
4
B
TD
PSP
341
1/27/95
387
16 1
S
1
B
TA
TRB
341
6/2/95
334
13.9
F
3
B
TA
SIS
341
2/1/95
259
10 8
F
9
B
TD
PSP
260
2/23/92
203
85
S
1
B
277
5/20/92
184
77
F
3
A
CD
VALVEX
366
5/4/95
120
50
F
2
A
HA
GENERA
278
387
6/23/96
8/1/96
107
103
45
4.3
F
F
2
3
A
H
CD
TA
278
341
12/2/95
4/9/95
99
62
4.1
2 6
F
S
3
2
H
B
HA
TA
32
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
System
Main Steam
Category
EF
34193014
Main Turbine
EF
38793008
Main Turbine
EF
TURBINE TAKEN OFF LINE TO REPAIR TURBINE JACKING
OIL SYSTEM STRUCTURAL CONCERNS
TURBINE TAKEN OFF LINE TO PERFORM POST OUTAGE
BALANCING TURBINE REMAINED OFF LINE TO REPAIR #4
JACKING OIL PUMP DISCHARGE PIPING.
Main Turbine
EF
Main Turbine
EF
34195006
AUTOMATIC MAIN TURBINE TRIP ON MECHANICAL
OVERSPEED TRIP RING #2 WHILE PERFORMING MTG
OVERSPEED TRIP TEST
TURBINE REMAINED OFF LINE TO REPAIR #4 JACKING
OIL PUMP DISCHARGE PIPING.
UNIT SHUTDOWN TO IDENTIFY AND REPAIR LEAKAGE IN
THE DRYWELL, AND TO REBALANCE THE Main Turbine
GENERATOR.
TWO CIVS CLOSED SIMULTANEOUSLY CAUSING POWER 27792009
LOAD IMBALANCE.
THE UNIT WAS MANUALLY SCRAMMED TO REPAIR THE
NO. 9 AND NO. 10 BEARINGS ON THE MAIN TURBINE
GENERATOR. THE NO 9 AND NO 10 BEARINGS AND
JOURNALS WERE REPAIRED.
Main Turbine
EF
Main Turbine
EF
Main Turbine
EF
Main Turbine
EF
Main Turbine
EF
VALVEX
VIS
REPAIR #2 TURBINE CONTROL VALVE STEM
TURBINE TRIP WAS CAUSED BY A FALSE SPURIOUS
SIGNAL FROM TURBINE #1 BEARING VIBRATION
INSTRUMENTATION LOOP COMPONENTS WERE
REPLACED THAT MOST LIKELY CONTRIBUTED TO THIS
SPURIOUS SIGNAL
Main Turbine
Main Turbine
EF
EF
TURBIN
ITRB
AUTOMATIC SCRAM/TURBINE TRIP.
34195004
MANUAL REACTOR/TURBINE TRIP PER SOE 95-10 TO
OBTAIN HOT TURBINE COASTDOWN VIBRATION DATA AT
APPROXIMATELY 80% POWER
Main Turbine
Main Turbine
EF
EF
DESCRIP
UNIT WAS OFF LINE BECAUSE OF THE TURBINE
OVERSPEED TRIP TEST, EXTENDED DUE TO MSR
MANWAY LEAKS
REACTOR TRIPPED FOLLOWING TRIP OF MAIN TURBINE
EXTENSIVE DAMAGE TO LOW PRESSURE TURBINE
NUMBER 3, THE MAIN GENERATOR, AND THE MAIN
GENERATOR EXCITER OCCURRED DURING THIS EVENT
CAUSES OF THIS EQUIPMENT DAMAGE ARE UNDER
INVESTIGATION REFUELED MEANWHILE
Page 160
OUTG LER
38796006
MIT POC DL BRODEUR
OUTG
HRS
OUTG
OUTG
TYPE
DAYS
F
21
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
HA
VALVEX
A
3
UNIT
ID
271
OUTG
DATE
4/10/94
331
10/26/93
50
2.1
F
3
A
TA
FCV
366
11/27/92
49
2.0
F
3
H
HA
277
325
10/8/96
1/23/96
48
46
2.0
19
F
F
1
2
A
A
271
12/6/93
44
18
S
1
341
4/12/95
42
17
F
321
11/19/94
34
14
271
11/1/96
25
341
9/20/93
341
OUTG LER
DESCRIP
27194004
"C" MOISTURE SEPARATOR HIGH LEVEL Main Turbine
TRIPPED AND A REACTOR SCRAM REPLACED A FAULTY
LEVEL CONTROLLER.
33193010
THE REACTOR SCRAMMED DUE TO A MOMENTARY
GROUND COMBINED WITH AN EXISTING UNDETECTED
ELECTRICAL GROUND IN THE CONTROL CIRCUITRY FOR
THE MAIN TURBINE STEAM CONTROL VALVE
System
Main Turbine
Category
EF
Main Turbine
EF
TURBIN
AN AUTOMATIC REACTOR SCRAM OCCURRED WHEN
VIBRATION AT THE NO. 6 TURBINE BEARING REACHED
APPROXIMATELY 12 MILS.
36692026
Main Turbine
EF
HA
HA
TURBIN
777777
Main Turbine
Main Turbine
EF
EF
B
HA
VALVEX
Main Turbine
EF
1
B
SB
ISV
Main Turbine
EF
F
3
H
HJ
VALVEX
TURBINE BEARING (#12) HIGH TEMPERATURE.
MANUAL SCRAM ON HI #5 MAIN TURBINE BEARING
VIBRATIONS
MANUAL SHUTDOWN TO REPAIR "A" MOISTURE
SEPARATOR EMERGENCY DRAIN VALVE
TURBINE TAKEN OFF LINE TO REPAIR N3018F607, MAIN
STEAM TO MSR ISOLATION VALVE
32194014
AN AUTOMATIC REACTOR SCRAM OCCURRED FROM
TURBINE STOP VALVE CLOSURE WHEN A TURBINE TRIP
SIGNAL WAS GENERATED DUE TO HIGH WATER LEVEL IN
THE MOISTURE SEPARATOR REHEATER "A/B".
Main Turbine
EF
10
F
1
A
Main Turbine
EF
22
09
F
1
B
TA
52
TURBINE TRIP DUE TO "A" MOISTURE SEPARATOR HIGH
LEVEL SIGNAL
DURING STARTUP, THE MAIN TURBINE TURNING GEAR
CIRCUIT BREAKER FAILED THE REACTOR WAS
SHUTDOWN TO MINIMIZE DIFFERENTIAL HEATING OF
THE TURBINE SHAFT DURING THE TIME THAT TURNING
GEAR WAS OUT OF SERVICE TURNING GEAR BREAKER
WAS REPLACED
Main Turbine
EF
3/16/95
21
09
S
1
B
TA
TRB
Main Turbine
EF
341
3/14/95
18
0.8
S
1
B
TA
TRB
Main Turbine
EF
321
10/28/93
17
07
F
1
B
HA
PIPEXX
TURBINE TAKEN OFF LINE TO OBTAIN TURBINE
COASTDOWN BEARING VIBRATION DATA.
TURBINE TAKEN OFF LINE TO PERFORM POST OUTAGE
BALANCING
SHIFT REMOVED THE MAIN TURBINE FROM SERVICE TO
REPAIR A STEAM LEAK ON THE ABOVE SEAT DRAIN FOR
CONTROL VALVE NO 4 THE LEAK WAS REPAIRED
Main Turbine
EF
321
9/30/92
16
07
F
2
A
HB
INSTRU
Main Turbine
EF
298
1/10/96
16
0.7
S
1
B
Main Turbine
EF
51
Source, INEEL / NRC Morp2
VALVEX
A MANUAL REACTOR SCRAM WAS INITIATED WHEN
VIBRATION AT THE NO. 3 TURBINE BEARING REACHED
APPROXIMATELY 12 MILLS DURING A POWER
REDUCTION FOR A FAILED PRESSURE SWITCH ON
MOISTURE SEPARATOR REHEATERS
TURBINE GENERATOR TAKEN OFF LINE TO REPAIR
32192024
TURBINE OIL SYSTEM
Page 161
MIT POC DL BRODEUR
Appendix 7. BWR/4 Data, 1992 - 1996
Appendix 7. BWR/4 Data, 1992 - 1996
UNIT
ID
353
OUTG
DATE
5/2/96
--- T~---- ~---3521
10/8/94
387
5/17/92
10/10/94
296
366
354
11/25/95
11/21/95
7/8/95
OUTG
HRS
OUTG
DAYS
DMaS
OUTG
TYPE
OUTG OUTG
METH REASN
Lost Generation Capacity by Failed System, Outage length
27 May 1998
Lost Generation Capacity by Failed System, Outage length
27 May 1998
OUTG
SYSTEM
OUTG
COMP
INSTRU
DESCRIP
TURBINE WAS TAKEN OFF THE GRID DUE TO A FAILED
BACKUP OVERSPEED TRIP TEST
~T~-----T~---?---~T~---~----~--~~
TURBIN
TURBINE TAKEN OFF LINE DUE TO HIGH TURBINE
12
0.5
VIBRATION
-c----e
~-----~
---UNIT ONE TOOK THE GENERATOR OFF LINE AT 0327
8
03
HOURS MAY 17TH DUE TO HIGH VIBRATION ON THE #5
BEARING OF THE MAIN TURBINE
-- t
MAIN TURBINE MANUALLY TAKEN OFF LINE DUE TO
TUBINE
EXCESSIVE VIBRATION AT THE NO. 3 MAIN TURBINE
BEARING. THIS VIBRATION WAS EXPERIENCED AS
REACTOR POWER WAS BEING REDUCED FOR THE
ABOVE PLANNED MAINTENANCE ACTIVITIES.
7
3
0.3
01
F
F
1
2
B
G
418
174
F
1
A
HJ
HTEXCH
IJI
OUTG LER
System
Turbine
Category
Main Turbine
I
Main Turbine
1
I
Main Turbine
EF
EF
Main Turbine
TRIPPED MAIN TURBINE DUE TO EXCESSIVE VIBRATION
MAIN TURBINE TRIPPED ON MOISTURE SEPARATOR
REHEATER HIGH LEVEL A MOTOR OPERATED VALVE IN
THE HIGH LEVEL DRAIN LINE WAS FOUND CLOSED
WHEN LCO 3.7 2 A FOR CONTROL ROOM VENTILATION
ACTION STATEMENT EXPIRED A UNIT SHUTDOWN WAS
INITIATED. CAUSE WAS A MOMENTARY INTERRUPTION
TO THE CONTROL CIRCUIT COMBINED WITH LENGTHY
CABLE RUNS
Main Turbine
Main Turbine
EF
MCR HVAC
EF
33393013
SHUTDOWN DUE TO "E"APRM UPSCALE TRIP.
AUTOMATIC SCRAM DURING IRM SURVEILLANCE DUE TO
FAULTY TEST EQUIPMENT
THE PLANT TWICE SECURED THE GENERATOR TO
CEASE PREMATURE RECOMBINATION OF HYDROGEN
AND OXYGEN INTHE OFF GAS SYSTEM
NI/TIPs
NI/TIPs
EF
EF
Offgas
EF
333
354
5/25/93
8/1/94
130
63
54
26
F
F
3
3
A
A
IG
331
9/3/92
8
04
F
1
B
WF
259
1/1/92
43,800
1,8250
S
4
F
ADMINISTRATIVE HOLD TO RESOLVE VARIOUS TVA AND
NRC CONCERNS.
Operation
HF/NRC
296
1/1/92
33,578
1,399.1
S
4
F
Operation
HF/NRC
271
10/15/94
48
2.0
F
1
B
WB
VALVEX
ADMINISTRATIVE HOLD TO RESOLVE VARIOUS TVA AND
NRC CONCERNS
COMBINATION OF SERVICE WATER LEAK ON THE HEAT
EXCHANGER AND "B"RBCCW BYPASS VALVE STUCK
OPEN. LINE ISOLATED, BLANKED OFF
RBCCW
EF
366
3/4/93
807
336
F
1
A
RC
FUELXX
THE UNIT WAS SHUTDOWN TO IDENTIFY AND REMOVE
THE LEAKING FUEL BUNDLE FROM THE CORE AND
INSPECT OTHER FUEL BUNDLES FOR POSSIBLE
DAMAGE
Reactor
EF
278
7/4/93
293
12.2
S
2
A
RC
FUELXX
MAINTENANCE OUTAGE FOR REPLACEMENT OF
DEFECTIVE FUEL ASSEMBLIES REACTOR MANUALLY
SHUTDOWN TO 18% AND THEN SCRAMMED FROM
THERE.
Reactor
EF
Source, INEEL / NRC Morp2
RCB
Page 162
27194013
MIT POC DL BRODEUR
OUTG OUTG
TYPE
DAYS
F
9.7
27 May 1998
Lost Generation Capacity by Failed System, Outage length
Appendix 7. BWR/4 Data, 1992 - 1996
OUTG OUTG
METH REASN
A
1
OUTG
SYSTEM
ID
OUTG
COMP
INSTRU
UNIT
ID
277
OUTG
DATE
3/27/92
OUTG
HRS
232
298
6/1/96
228
9.5
S
2
A
277
8/11/93
198
8.3
S
2
H
IE
INSTRU
352
8/20/95
186
7.7
S
2
A
RC
FUELXX
341
12/28/96
157
65
F
9
A
JE
LI
277
4/24/93
81
3.4
F
1
A
ID
INSTRU
333
2/25/93
651
27 1 F S F
1
H BA
388
12/10/93
480
200
F
1
A
XX
ZZZ
277
333
1/2/93
5/30/95
357
216
14 9
9.0
S
F
1
3
H
D
AD
ISV
354
3/20/95
181
75
F
2
G
353
12/4/92
180
75
F
2
G
PSF
CB
PUMPXX
System
Reactor
Category
EF
Reactor
EF
MAINTENANCE OUTAGE FOR REACTOR WATER LEVEL
MODIFICATION
REACTOR WAS SHUTDOWN TO REPLACE A FAILED FUEL
BUNDLE
34196024
REACTOR SCRAM DUE TO FALSE LEVEL 2 AND 8
INITIATION WHILE VALVING IN REFERENCE LEG OF RX
WATER LEVEL BACK FILL.
27793010
PLANT SHUTDOWN DUE TO REACTOR LEVEL
INSTRUMENT MISMATCH
SHUTDOWN DUE TO BLOCKAGE OF THE INTAKE
STRUCTURE SCHEDULED OUTAGE FOR "B"RECIRC
SEAL REPAIR SHUTDOWN TO REPAIR LEAK IN CHEMICAL
DECON CONNECTION
Reactor
EF
Reactor
EF
Reactor
HF/OA
Reactor
EF
Recirc
EF
MANUALLY SHUTDOWN DUE TO HIGH DRYWELL
LEAKAGE INSPECTION OF DRYWELL REVEALED A
CRACKED WELD ON THE "A" RX RECIRC PUMP RBCCW
OUTLET LINE OTHER WORK INCLUDED INSTALLATION
OF TORQUE COLLARS ON THE MAIN TURBINE AND
INSTALLATION OF RX LEVEL INSTRUMENTATION.
Recirc
EF
MAINTENANCE OUTAGE TO REPAIR RECIRC PUMP SEAL
33395010
A 3/4" MANUAL VALVE (JET PUMP TO RECIRC PUMP
SUCTION) PACKING LEAK EXCEEDED TECH SPEC LIMITS.
THE PACKING WAS REPLACED WITH A DIFFERENT STYLE
THAT IS LESS PRONE TO GROSS FAILURE.
Recirc
Recirc
EF
EF
WHILE I&C TECHS WERE PERFORMING A PM ON THE
OPTICAL ISOLATOR FOR THE REACTOR RECIRCULATION
PUMP MG SETS, A LOSS OF BOTH MG SETS OCCURRED
A MANUAL SCRAM WAS INITIATED INACCORDANCE WITH
THE PROCEDURE
Recirc
HF/OA
REACTOR WAS MANUALLY SCRAMMED AFTER BOTH
RECIRCULATION PUMPS TRIPPED DURING
Recirc
EF
DESCRIP
UNIT SHUTDOWN DUE TO REACTOR WATER LEVEL
MISMATCH
PLANT WAS SHUTDOWN FOR FUEL LEAKER
REPLACEMENT ONE LEAKING FUEL ASSEMBLY WAS
IDENTIFIED AND REPLACED. NORMAL POWER
ASCENSION WAS IMPLEMENTED WITH FULL POWER
ACHIEVED ON 6/15/96
OUTG LER
27792005
SURVEILLANCE TESTING.
Source, INEEL / NRC Morp2
Page 163
MIT POC DL BRODEUR
Appendix 7. BWR/4 Data, 1992 - 1996
OUTG OUTG
DAYS
TYPE
6 9 SF
OUTG
OUTG
OUTG OUTG
COMP
METH REASN SYSTEM
PIPEXX
B I CB
2
TURBIN
UNIT
ID
366
OUTG
DATE
1/24/92
OUTG
HRS
166
354
11/1/96
146
61
S
1
B
277
5/14/94
121
50
F
3
A
CB
PUMPXX
352
5/7/95
99
41
S
2
B
HH
HTEXCH
321
12/7/93
79
3.3
F
3
A
CH
277
7/27/92
60
25
F
1
A
CB
354
12/3/92
52
2.2
F
2
H
341
3/16/92
46
19
F
2
B
296
9/15/96
44
18
F
2
A
AD
366
5/21/93
34
14
F
2
G
ZZ
353
12/24/96
31
13
F
2
A
CB
MECFUN
278
298
12/2/93
3/16/94
327
188
13.6
7.8
F
F
1
2
A
A
SF
VALVEX
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
System
Recirc
Category
EF
PLANNED MAINTENANCE TO REPAIR THE REACTOR
RECIRC PUMP SEAL
APRM HI HI FLUX AUTOMATIC SCRAM DUE TO RECIRC
PUMP SPEED PROBLEMS.
REACTOR WAS SHUTDOWN TO PERFORM MAINTENANCE
ON "C" DRAIN COOLER, "A" RECIRCULATION PUMP SEAL,
AND THE CONDENSER WATERBOXES.
Recirc
EF
Recirc
EF
Recirc
EF
CKTBRK
AN AUTOMATIC REACTOR SCRAM OCCURRED DUE TO A 32193016
LOW REACTOR WATER LEVEL SIGNAL THE LOW WATER
LEVEL WAS CAUSED BY A TRIP OF THE "A"RFP AND
FAILURE OF THE REACTOR RECIRCULATION PUMPS TO
RUN BACK TO THE NO 2 SPEED LIMITER
Recirc
EF
PUMPXX
RECIRC PUMP TRIP AND VESSEL TEMPERATURE
DIFFERENTIAL
CONTRACT EMPLOYEE BUMPED CART INTO A MCC,
CAUSING REACTOR RECIRCULATION PUMP M/G SET
VENT FANS TO TRIP, RESULTING INA DOUBLE
RECIRCULATION PUMP TRIP. CONTROL OPERATOR
MANUALLY SCRAMMED THE REACTOR
27792013
Recrc
EF
35492013
Recrc
HF/OA
MANUAL REACTOR SCRAM DUE TO OPERATION IN
REGION OF INSTABILITY (HIGH POWER TO FLOW)
FOLLOWING INADVERTENT ACTUATION OF SAFETY
SYSTEMS DURING SURVEILLANCE TEST SAFETY
SYSTEM ACTUATION OCCURRED WHEN TEST METER
SHORTED INTERNALLY
34192002
Recrc
HF/OA
29696005
SHUTDOWN BY MANUAL SCRAM FOLLOWING THE 3A
RECIRCULATION PUMP TRIP.
A MANUAL REACTOR SCRAM WAS INITIATED WHEN BOTH 36693005
REACTOR RECIRCULATION PUMPS TRIPPED.
35396009
REACTOR WAS SHUTDOWN DUE TO A FAILED SCOOP
ITUBE POSITIONER ON THE 2B MG SET.
LPCI MOTOR OPERATED VALVE MO-25A INOPERABLE.
VALVE RHR-MO-27A FAILED SURVEILLANCE TESTING.
REPAIRED RHR-MO-27A
Recrc
EF
Recirc
EF
Recirc
EF
RHR
RHR
EF
EF
MG
DESCRIP
UNIDENTIFIED SOURCE OF FLOOR DRAIN LEAKAGE
INSIDE THE DRYWELL. INVESTIGATION REVEALED THE
PACKING LEAKOFF LINE FOR THE "B"REACTOR
RECIRCULATION PUMP'S DISCHARGE ISOLATION VALVE
HAD SEPARATED TURBINE ROTOR REQUIRED
REBALANCING.
Page 164
OUTG LER
MIT POC DL BRODEUR
Appendix 7. BWR/4 Data, 1992 - 1996
OUTG OUTG
TYPE
DAYS
11 3
F
27 May 1998
Lost Generation Capacity by Failed System, Outage length
OUTG OUTG
METH REASN
3
G
OUTG
OUTG
COMP
SYSTEM
INSTRU
CG
UNIT
ID
366
OUTG
DATE
4/11/95
OUTG
HRS
271
341
4/25/95
249
10 4
F
331
8/17/92
148
6.2
F
260
3/30/95
69
29
F
3
H
321
8/27/92
66
28
F
3
H
HG
DEMINX
366
8/30/94
52
22
F
3
H
IA
INSTRU
341
8/13/93
49
2.0
F
3
G
JE
TV
321
6/15/93
42
1.7
F
3
H
CH
260
7/28/92
33
14
F
3
A
325
324
387
3/18/96
3/17/96
11/10/95
156
113
606
65
47
25.3
F
F
S
2
2
2
A
A
B
WA
WA
TJ
PUMPXX
PUMPXX
GEN
352
1/14/94
57
24
F
2
A
HA
GENERA
3
A
JJ
RG
A
AD
FT
ALVEX
RPS
Category
HF/OA
RPS
EF
RPS
EF
26095004
AUTOMATIC SCRAM CAUSED BY PERSONNEL ERROR
DURING SURVEILLANCE TESTING
AN AUTOMATIC REACTOR SCRAM OCCURRED DUE TO A 32192021
GROUP I ISOLATION CAUSED BY AN UPSCALE SPIKE ON
THE MAIN STEAM LINE RADIATION MONITORS
36694007
AUTOMATIC REACTOR SCRAM WHEN RPS ELECTRICAL
BUS 2A WAS BEING TRANSFERRED FROM ITS
ALTERNATE TO ITS NORMAL SUPPLY. THE EVENT WAS
CAUSED BY INADVERTENTLY MOVING THE SWITCH
BEYOND ITS CENTER POSITION WHEN TRANSFERRING
FROM "ALT A" TO THE "NORM" POSITION
RPS
HF
RPS
EF
RPS
HF/OA
34193010
RPS
EF
AN AUTOMATIC REACTOR SCRAM OCCURRED DUE TO A 32193012
FALSE LOW REACTOR WATER LEVEL SIGNAL. THIS
OCCURRED WHEN AN INSTRUMENT LINE
DEPRESSURIZED AFTER A PACKING NUT ON A VALVE IN
THE SENSING LINE DISENGAGED.
RPS
EF
SCRAM DUE TO A SPURIOUS HIGH WATER LEVEL TRIP,
CAUSED BY A FALSE SIGNAL FROM A NEW ELECTRICAL
SWITCH.
RPS
EF
DESCRIP
AN AUTOMATIC REACTOR SCRAM OCCURRED DUE TO
THE IMPROPER PLACEMENT OF A JUMPER WHILE
ATTEMPTING TO RETURN THE REACTOR WATER
CLEANUP SYSTEM TO SERVICE
OUTG LER
36695001
34195005
AUTOMATIC REACTOR SCRAM ON APRM NEUTRON
UPSCALE TRIP RESULTING FROM REACTOR PRESSURE
REGULATOR TRANSIENT.
33192013
AUTOMATIC REACTOR SCRAM CAUSED BY PERCEIVED
HIGH AVERAGE POWER RANGE NEUTRON FLUX,
CAUSED BY ELECTRO-MAGNETIC SIGNAL NOISE, WHICH
REDUCED FLOW BIASED SET-POINTS TO BELOW THE
CURRENT POWER LEVEL.
THE REACTOR SCRAM WAS AUTOMATICALLY INITIATED
BY A TRIP OF THE MAIN TURBINE DUE TO A FALSE HIGH
REACTOR WATER LEVEL SIGNAL
System
FORCED OUTAGE DUE SW PUMPS PROBLEMS.
MANUAL SCRAM DUE TO SW PUMPS PROBLEMS
OUTAGE TO REPAIR A HYDROGEN LEAK INTO THE
STATOR WATER COOLING SYSTEM. UNEXPECTED MAIN
GENERATOR BAR TIE,SPACER DAMAGE AND WEDGE
LOOSENESS WAS IDENTIFIED WHICH EXTENDED THE
OUTAGE DURATION
32596003
SW
SW
SWC
EF
EF
EF
REACTOR MANUAL SCRAM OCCURRED DUE TO A LOSS
35294001
SWC
EF
SOFSTATOR WATER COOLING.
Source, INEEL / NRC Morp2
Page 165
MIT POC DL BRODEUR
Appendix 7. BWR/4 Data, 1992 - 1996
OUTG
HRS
OUTG OUTG
TYPE
DAYS
20
F
49
OUTG
OUTG OUTG
OUTG
COMP
METH REASN SYSTEM
TJ
TCV
3
A
UNIT
ID
388
OUTG
DATE
1/20/94
341
3/26/95
48
20
F
1
B
354
8/30/94
45
19
F
3
H
353
11/22/95
11
0.5
F
1
A
HF
353
3/17/93
5
0.2
F
1
A
WG
353
10/6/96
5
02
F
1
A
HA
277
7/17/92
219
9.1
F
3
A
XX
324
325
324
325
353
9/5/96
7/10/96
7/11/96
9/5/96
2/21/95
198
164
154
123
79
83
68
64
5 1
33
S
S
S
S
F
2
2
2
2
3
H
H
H
H
H
ZZ
ZZ
ZZ
ZZ
RB
353
5/14/96
57
2 4
F
3
H
ZZ
352
2/21/95
38
16
F
3
H
RB
296
9/4/96
139
58
S
2
B
321
12/26/93
62
26
F
2
A
HH
324
260
7/26/96
12/2/94
44
20
18
08
F
F
2
3
A
A
CH
JJ
Source, INEEL / NRC Morp2
27 May 1998
Lost Generation Capacity by Failed System, Outage length
TJ
DESCRIP
AUTOMATIC MAIN TURBINE TRIP WITH AUTOMATIC
REACTOR SCRAM AS A RESULT OF A STATOR COOLING
WATER TEMPERATURE CONTROL VALVE PROBLEM
WHICH CAUSED HIGH STATOR COOLING WATER
TEMPERATURES.
OUTG LER
38894002
TURBINE TAKEN OFF LINE TO REPAIR A STATOR
COOLING WATER VENT LINE LEAK.
AUTOMATIC SCRAM CAUSED BY GENERATOR RUNBACK
DUE TO LOSS OF STATOR WATER COOLING.
THE GENERATOR WAS TAKEN OFF THE GRID TO
FILTER
REPLACE THE STATOR WATER COOLING FILTERS
THE MAIN TURBINE TRIPPED OFF LINE DUE TO LOW
HTEXCH
FLOW IN THE STATOR WATER COOLING SYSTEM.
POWER REDUCTION DUE TO REPLACEMENT OF THE
FILTER
GENERATOR STATOR COOLING 'Y' STRAINER.
27792012
LIGHTNING STRIKE - AUTO SCRAM INITIATED BY TCV
XXXXXX
FAST CLOSURE ON LOAD IMBALANCE.
OUTAGE DUE TO HURRICANE FRAN
ZZZZZZ777777
777ZZZZ OUTAGE DUE TO HURRICANE BERTHA
OUTAGE DUE TO HURRICANE BERTHA,
ZZZZZ
OUTAGE DUE TO HURRICANE FRAN
ZZZZZZ
REACTOR SCRAM DUE TO ELECTRICAL DISTURBANCE AT
CONROD
WHITPAIN SUBSTATION
35396004
REACTOR SCRAM DUE TO TURBINE TRIP CAUSED BY
ZZZZZZ
GRID INSTABILITY
REACTOR SCRAM DUE TO ELECTRICAL DISTURBANCE AT
CONROD
WHITPAIN SUBSTATION
SHUTDOWN FOR SCHEDULED MAINTENANCE AND
REPAIRS
SHIFT REMOVED THE MAIN GENERATOR FROM SERVICE,
VALVEX
AND A MANUAL SCRAM WAS INITIATED TO REPAIR 1N21F253. THE VALVE WAS REPAIRED AND THE UNIT
RETURNED TO SERVICE
PSP
VALVEX
TIS
FORCED OUTAGE TO REPAIR HD-LV-75
JAUTOMATIC SCRAM CAUSED BY BALANCE OF PLANT
IEQUIPMENT FAILURE
Page 166
26094013
SWC
Category
EF
SWC
EF
SWC
EF
SWC
EF
SWC
EF
SWC
EF
Transmission
Nature
Transmission
Transmission
Transmission
Transmission
Transmission
Nature
Nature
Nature
Nature
EF
Transmission
EF
Transmission
EF
System
EF
EF
EF
EF
MIT POC DL BRODEUR
Appendix 8. LGS unavailability data, sorted by failed system
27 May 1998
1992 - 1996
A
B
C
D
E
F
G
H
I
J
K
L
M
Unit
Date
Type
MWHr
Eff Out Days
Total effective
outage days
System
#REFI
Component
Failure
Cause
Category
Issue #
Remarks
Class
1
7/15/96
Manual
scram
182,179
654
Air removal
SJAE
Y-strainer
Steam leak
Improper
installation
HF/C
10005889
4
5
2
2
2/24/95
10/1/93
Load drop
Load drop
28,000
1,090
1 01
004
Air removal
Circ water
SJAE nozzle
Pump
OF
6
7
8
9
1
2
2
1
11/1/94
5/9/92
10/12/96
6/19/93
drop
drop
drop
drop
1,200
36,960
37,463
11,050
005
1 40
1 35
0 42
Computer
Condensate
Condensate
Condensate
P1 program
Pump
Pump motor
Motor
Broken
Maintenance
Would not run
Beanng
Vibration
Bearing
Misaligned
FME
Misaligned
10
11
12
13
14
15
16
17
18
1
2
2
2
2
2
2
2
1
2
Load drop
5/1/93
Load drop
6/2/92
Load drop
5/1/93
6/26/93 Load drop
6/28/93 Load drop
6/29/93 Load drop
Load drop
3/1/94
Load drop
4/1/92
8/28/93 Load drop
12/16/96 Shut down
7,313
7,140
6,410
2,967
2,967
2,967
1,504
750
734
139,200
0 28
0 27
0 24
0 11
0 11
0 11
006
0 03
0 03
500
Motor
Condensate
Pump motor
Condensate
Pump motor
Condensate
Pump
Condensate
Pump
Condensate
Pump
Condensate
Pump
Condensate
Water box
Condensate
Pump
Condensate
Condenser Expansion joint
Beanng
Beanng
Bearing
Bearing
Beanng
Beanng
Misaligned
Misaligned
Misaligned
Misaligned
Misaligned
Misaligned
Leaks
Bearing
Leak
Misaligned
A
M
D
D
O
D
D
D
D
D
D
D
D
A
D
A
2
1
9/23/92
7/20/95
Load drop
Load drop
58,625
23,000
2 22
0 87
Condenser
Condenser
Tube
Water box
Leaks
Cleaning
2
2
1
1
1
6/1/96
10/1/92
7/9/94
12/1/93
8/28/95
Load drop
Load drop
Load drop
Load drop
Shut down
12,619
6,375
6,290
4020
121,440
045
0 24
0 24
0 15
4 60
Condenser
Condenser
Condenser
Condenser
Drywell
Tube
Tube
Tubes
Waterbox
Flange
Leaks
Leaks
Cleaning
Cleaning
Misaligned
2
12/6/96
Manual
scram
250,560
9 00
EHC
Pressure switch
Leak
1
7/25/96
Scram
93,663
3 36
EHC
F/V card
2
1/3/93
Scram
84,000
3 18
EHC
2
8/20/95
Scram
69,396
2 49
2
3/26/93
Scram
56,700
1
3/31/96
Load drop
2
11/19/92
2
I
3
Load
Load
Load
Load
EF
EF/WP
EF
EF/WD
HF/PI
EF/WD
EF/WD
EF/WD
EF/WD
EF/WVD
EF/WD
EF/WD
EFWND
EF/WP
EF/WD
HF
10006201
Realigned
Realigned
Realigned
Realigned
Realigned
Realigned
10006422
Realigned
2F20, DF
19
20
21
22
23
24
25
EF/WP
EF/WP
Normal
Maintenanc
A
A
e
A
A
A
A
M
EFVWP
EF/WP
EF/WP
EF/WP
HF/C
35295006
Broken bracket
Isevered
tubing
HF/LCA
10006385
Failure
Infantile failure
EF/WP
10005909
Relay
Hi pressure
Sporadic
anomaly
EFNV/WD
93-01-01
RF
10004338
D
EHC
Relay
High
Impedance/ NC
contact
LTA design
EF/WD
10004338
RF, 93-0101
D
2 15
EHC
#6 ISV
93-03-38
1 70
EHC
EF/WD
10005451
Load drop
44,101
1 67
EHC
Speed control
logic
#3 CV piping
Air entrap in
control pack
Sporadic
anomaly
HF/PI
47,222
Perturbation in
ETS/RETS
Speed control
LVG
Leak
Weld failure
HF/C
92-11-20
RF,
10005615
M
4/29/96
Load drop
38,939
1 40
EHC
#3 CV piping
Leak
Weld repair
failure
HF/PCM
10005615
RF 92-1120
M
2
2
1
5/16/93
6/3/95
9/7/93
Load drop
Load drop
Scram
16,493
17,150
146,784
0 62
062
5 56
EHC
EHC
Electncal
#2 TCV servo
#4 CIV
Breaker
Oil leak
Leak
Failed to
reclose
Indeterminant
EF
EF
EF/WD
93-05-20
TT
M
M
D
2
1
10/19/94
6/17/96
Scram
Load drop
77,754
22,121
2 95
0 79
Electrical
Electrical
D24 Bus
Output breaker
De-energized
Low pressure
rupture
Inadvertent
Improper
installation
2
8/8/95
Scram
80,794
2 90
FW
FW LCS
1
1
2
3/1/94
7/13/92
4/7/93
Load drop
Load drop
Load drop
10,800
8,566
6,000
0 41
0 32
0 23
FW
FW
FW
LCS
RFP turbine
FWLCS
2
6/28/95
Load drop
6,300
023
FW
2
9/8/94
Load drop
5,220
0 20
FW
Poor Corr
Maint
26
Repeat
event 2M19
D
27
A
28
29
30
31
32
33
34
35
36
37
38
Spunous
O
Infantile
failure
10000021
D
O
HF/OA
HF/C
10005797
EF
10004298
EFJWD
HF/PI
EF/WD
92-07-10
93-04-04
OM
RRB
D
O
D
HF/OA
10004173
RRB
O
TT
39
40
41
42
Insulation fire
'A' level down
spike
Oil soaked
Spunous,
Indeterminant
FW UPS
Power
disconnect
switch off
Bumped
dunng cleaning
Check valve
cap
Leak
43
45
Data sorted by system, lost generation
Loose
Loss of DC
powersupply connection
(2K612)
Page167
EF
A
A
I
LGS POC VT ANGUS
Appendix 8. LGS unavailabilrty data, sorted by failed system
46
47
27 May 1998
1992 - 1996
J
K
L
M
EFNVD
EF/WD
10004299
RRB
D
D
Spurious
EFNVD
93-02-16
RRB
D
Relay
EF/WD
D
EFNV/WD
D
H
A
B
C
D
E
F
G
1
1
10/8/94
8/8/95
Load drop
Load drop
3,829
3,740
0 15
0 14
FW
FW
LCS
Pressure switch
1
2/7/93
Load drop
3,510
013
FW
Pressure switch Low pressure
2
1/11/92
Load drop
3,430
0 13
FW
Turbine
controller
Malfunction
1
4/24/95
Load drop
2,230
0 08
FW
Pressure switch
Failed low
2
4/19/93
Load drop
2,000
008
FW
FWLCS
1
1
2
2
1
12/1/93
7/29/94
2/10/92
8/1/92
1/21/92
Load
Load
Load
Load
Load
930
570
004
0 02
10,750
7,030
0 41
027
FW
FW
FW
FW heating
FW heating
Pump
Trip lever
Pump
Vent line
Logic
2
1
2
6/17/95
8/2/96
7/18/92
Load drop
Load drop
Load drop
0 18
005
005
FW heating
FW heating
FW heating
2
1
2
9/5/95
4/11/96
5/14/96
Load drop
Load drop
Scram
370
370
89,099
0 01
0 01
3 20
1
9/2/95
Load drop
39,426
1 49
2
9/2/95
Load drop
7,470
0 27
1
8/23/93
Load drop
725
0 03
2
6/6/96
Load drop
378
0 01
Premature
actuation
Out of
calibration
48
49
50
51
52
53
54
55
56
57
58
59
60
61
drop
drop
drop
drop
drop
4,970
1,468
1,353
Actuated
Trip
Leaks
High level
sensed
Inadvertent
Dump valves
Drain valve
Vent line
Actuator
Malfunction
Leaks
Age
FW heating
FW heating
Generator
Valve
Drain cooler
Volts/Hz relay
Leak
Tube leak
Actuation
inappropnate
Poor Design
MOD package
LTA
H2
Recombiner
Recorder
Logic
Logic
62
63
64
Spunous,
Master level
controller down Indeterminant
spike
FAC
Spunous
EFNVD
EF
HF/OA
EFNVD
EFNV/WP
HF/MS
93-04-18
92-01-10
RRB
D
No PM
A
O
D
A
M
EF
EF/WD
HF/PI
10005652
A
M
A
A
D
O
MOD PMT LTA
HF/PI
10004403
O
MOD PMT LTA
HF/PI
10004403
O
FAC
EF
HF/PCM
EF/WP
DF
Recorder
H2
Recombiner
Instrument Air Dryer package
Gasket
Isophase bus
cooling
Main steam
Fan
Trip
LTA WO/
procedures
HF/PI
10005731
O
SRV
Opened
Pilot seat
erosion
HF/MS
10004442
O
Failed
M
EF
65
66
1
9/11/95
Scram
369,991
1401
1
3/24/96
Shut down
203,260
7 30
Main steam
SRV
Leak
EFNVD
2
2
1
9/1/94
9/1/94
12/1/93
Load drop
Load drop
Load drop
28,290
6,500
3400
1 07
0 25
0 13
Main steam
Main turbine
Main turbine
Valve
#2 MSV
TCV pressure
switch
Leak
EF
EF/WP
EFNVD
1
111/92
Load drop
2,904
0 11
Main turbine
Pressure
instrument RV
2
1
4/1/92
1/17/94
Load drop
Load drop
1,039
940
0 04
004
Main turbine
Main turbine
1
10/7/94
Load drop
31,945
1 21
Offgas
After
Condenser
Cleaning
EF/WP
1
8/20/95
Shut down
342,276
12 97
Reactor
Fuel
Leak
EF
1
2
11/1/94
12/4/92
900
157,870
0 03
598
Reactor
Recirc
Fuel
Recirc pump
2
12/24/96
55,680
200
Recirc
2
1/15/95
Load drop
Manual
scram
Manual
scram
Load drop
48,000
1 82
Recirc
Scoop tube ball
joint
MG set
1
5/7/95
Shut down
31,060
1 18
Recirc
Seal
1
2
1/2/95
6/26/95
Load drop
Load drop
18,600
14,115
070
0 51
Recirc
Recirc
Seal
Temperature
switch
2
2
2/24/94
4/29/95
Load drop
Load drop
10,080
10,000
0 38
0 36
Recirc
Recirc
2
2/16/94
Load drop
7,150
0 27
1
2
1/30/95
7/17/95
Load drop
Load drop
7,000
5,600
1
10/16/92
Load drop
990
67
1E07
D
68
69
70
71
72
73
74
75
Failed
RV closed
Set point drift
Vac switch
Valve
Moisture
positioner
separator dump
valve
Mispositioned
PM frequency
LTA
HF/OA
EF/WNVP
HF/MS
A
A
D
92-01-02 SU, DF, TT
10001318 SU, TT, PM
deferred
78
79
Leak
EOC-RPT logic Breaker tnpped
Maintenanc
e outage
Vibration
Induced
Indeterminant
HF/LCA
Age
EF
Leak
Spiked high
Age
Spurious
EF
EFNVD
Pump
Coupler bypass
valve
Tnp
Incorrect
position
Fuses pulled
LTA
procedures
Recirc
Recorder
Mislabelled
LTA MOD
review/ PMT
027
020
Recirc
Recirc
MG set
Temperature
switch
Trp
Spiked high
0 04
Recirc
RX level signal
Low level
sensed
Broke
Generator
ground
Leak
80
81
82
83
84
85
86
87
88
89
Data sorted by system, lost generation
Page168
F
F
EF
HFIOA
A
M
A
76
77
O
Occurred
during ST
10006441 Repeat
Event
92-12-01
EF/WD
O
M
D
Maintenanc
e outage
1 M03
A
10004172
Resealed
RPT
A
D
HF/OA
HF/PI
10003924
RRB
O
O
HF/PI
10001491
RRB
O
Bumped
Spunous
HF/OA
EF/WD
10004172
RPT
O
D
Channel noise
EF/WD
92-10-21
RRB
D
LGS POC VT ANGUS
Appendix 8. LGS unavailability data, sorted by failed system
90
27 May 1998
1992 - 1996
A
B
C
D
E
F
G
H
I
J
K
1
1
7/19/95
5/21/96
Load drop
Scram
700
75,095
0 03
2 70
Recirc
RPS
MG set
Logic
Perturbations
No 1/2 scram
alarm
Operator
Indeterminate
HFIOA
HF/PI
10005675
1
1/14/94
Scram
59,230
2 24
SWC
Tnp circuit
Short
Bulb
installation
EF
2
9/8/94
Load drop
28,650
1 09
SWC
Y-strainer
Clogged
Generator
hydrogen leak
EF/WD
2
2
2
10/6/96
11/22/95
3/17/93
Load drop
Load drop
Load drop
20,859
15,253
8,030
0 75
055
0 30
SWC
SWC
SWC
Y-strainer
Filters
Valve
Clogged
Clogged
Valve
LTA procedure
mispositioned
EFNVD
EFWID
HF/PI
93-03-24
1
12/23/96
Load drop
5,427
0 19
SWC
Temperature
CV
HF/C
10006438
91
92
93
94
95
96
97
Data sorted by system, lost generation
Loose adjust
arm screws
Pagel69
Manufacturing
10002830
M
L
O
O
OM
M
TT
D
SU, DF
D
D
O
F
LGS POC VT ANGUS
Unit
3
Date
1/20/95
Type
load drop
MWHr
13194
2
6/24/94
load drop
4904
2
3/2/93
3
3
Automatic
Scram
12/19/92 Manual
Shutdosn
10/15/92 Automatic
Scram
Eff Out
Component
System
Days
0 5 Circ Water screens immobile
0 2 Circ Water condenser
14.3 Condensate 2c cond pumps
398510
0.5 Condenser
13908.0
Failure
pin shear
cleanliness
loss of pwr to
transformer
3
11/28/93 load drop
3651
0 1 Control Rod 34-31 accumulator failed
Drive
o-ring
2
4/28/94
load drop
2537
2
2/3/96
load drop
492
3
12/2/95
Automatic
Scram
114994
0 1 Control Rod rod 26-15
Drive
0 0 Control Rod hv-1 11
Drive
DC
cracked terminal
4 1
strip
2
1/1/93
418199
15.0
EHC
2
Manual
Shutdown
12/21/92 load drop
69152
2.5
EHC
3
6/22/96
load drop
20785
0.7
EHC
pressure
transmitter
pressure
transimtiter
servo
2
4/1/97
load drop
17756
0.6
EHC
cooler
2
12/17/92 load drop
10398
04
EHC
3
2
6/12/94
5/9/96
load drop
load drop
4054
2148
0.1
0.1
2
4/2/97
load drop
1793
2
3
3
Automatic
Scram
10/11/94 Automatic
Scram
7/17/92
11/6/95
load drop
condensate pump a tripped
N2 leak at
charging block
broken o-ring
Cause
wrong pin
chlorine oos and warm water
and inst at cal limit
operator opened bkr
Remarks
N
Class
O
HF/PCM
N
M
HF/OA
N
O
Category
HF/PI
Issue #
A
EF/WP
O
27892008
Equipment Failure
HF/OA
o ring failure, rod 3431
accumulator
worn
EF/WP
N
M
Inop control rod
O-ring replaced
EF/WP
N
A
hcu hv-111 broken
O-ring damaged/cut
EF/VP
N
A
HF/OA
N
O
EF/WD
N
D
turbine trip - pos &neg ground 2nd ground
2nd ground by
person working on
equipment
turbine control valve oscillations design
did not work
did not work
turbine control vie oscillations
design
EF/WD
SU
D
leak
o-ring failure
HF/C
N
M
restriction
repair #4 cv ehc leak and msv
leak
ehc fluid leak
suspect FME root cause TBD
HF/C
N
M
did not work
turbine control valve oscillations design
EF/WD
SU
D
EHC
EHC
pressure
transimtiter
muffin fan
wire lug
stopped working
loose wire
unknown
vibration
EF/WD
EF/WD
SU
N
D
D
0.1
EHC
cooler
restriction
ehc elec cabinet cool fan
turbine control #2 valve
oscillation
ehc fluid leak
suspect FME root cause TBD
HF/C
N
M
248585
89
Electrical
3435 breaker
tripped
rwcu controls
lighting strike
EF
N
A
99423
36
Electncal
main power
transformer for
inverter
shorted winding
scram loss of static inverter y50 break down of insulation
EF/WD
N/RRB
D
01
Electrical
fuse
loose
13 kv electncal system-loose
fuse
HF/OA
N
0
3336
. 1
Data sorted by system
Description
b screen immobile, b cw pp
removed from service
low condenser vacuum
CLEAN CONDENSER
WATERBOXES.
PCIS GROUP I ISOLATION
CAUSED BY BUMPING
INSTRUMENTATION.
HTEXCH
13 0 Containment VALVEX
361608
27 May 1998
July 1992 - June 1997
Appendix 9. PBAPS Unavailability Data, sorted by failed system
Page170
droped holder damage
PBAPS POC FL JORDAN
Unit
3
27 May 1998
July 1992 - June 1997
Appendix 9. PBAPS Unavailability Data, sorted by failed system
Type
Date
12/29/93 load drop
MWHr
2266
Eff Out
Days
0.1
System
Electrical
Component
3-2a-k004a
Description
Failure
deenergized when recirc runback a pump
480v load center
30B01 was being
restored from the 3
4G4 tie breaker
loss of power supply to e22 bus Diesel feedback signal during
mod testing
&y-34
loss of power supply to e22 bus loss of power
Cause
30b01 Ic being restored
Remarks
N/RRB
Class
O
HF/PI
N
O
HF/C
N
M
HF/PCM
N
M
N/RRB
D
Category
HF/PI
Issue #
2
6/10/95
load drop
835
00
Electrical
e22 bus
2
5/19/93
load drop
678
0.0
Electncal
e322 trip
loss of pwr to
panel y-34
loss of fw htg
3
3/7/93
167298
60
FW
c rfp
hi vibration
maintenance outage
poor lubncation to vib sensor
3
4/21/97
Automatic
Scram
load drop
35167
1.3
FW
computer dcc-x
power supply
fail/transfer of
control
feedwater computer trouble
relay actuation
power supply? transfer -design EF/WD
issue
2
3
30423
29188
11
1.0
FW
FW
a rfp vibration 95% pwr limit
scram-feedwater transient
foreign material
failed transmitter on card
HF/C
EF/EOL
SU
N/TT
A
21995
08
FW
2ap001
3a rfp speed
controller
power supply fw
control system
n/a
upscale
2
11/11/95 load drop
8/1/95 Automatic
Scram
load drop
6/3/95
failure of PS
power ascension &c rfp
problems
age
EF/EOL
N
A
3
5/21/93
load drop
18117
0.7
FW
fp turbine control
HF/PCM
N
M
4/27/94
load drop
12881
0.5
FW
rfp
rework mgu hyd jack solenoid
sv7
rfp control problem
mtce left out parts
3
not smooth operation
EF/WD
N
D
2
1/22/93
load drop
10171
04
FW
2cs018 2c rfp
rfp c slow responce
failed
EF/WP
N
A
3
2
5/1/94
1/8/94
load drop
load drop
7634
5813
0.3
02
FW
FW
rfp
2as018
a rfp maintenance
a rfp maintenance
not smooth operation
improper lubrication
EF/WP
HF/PCM
N
SU
M
M
2
2
3
4/2/97
10/5/96
6/13/97
load drop
load drop
load drop
4271
4263
4238
02
02
02
FW
FW
FW
2as018 a rfp
2bs018 b rfp
3b rfp speed
controller &hjsv
parts missing post
mtce
control valve
actuator binding
control valve
bearing seized
control system
speed controller
siezed
failed to trip
probe failure
sol coil burned
and controller
degradded
reactor feed pump trouble
b rfp high vibration
3b fpr speed control problem
debns in trip dump valve
worn parts
age
HF/PI
EF/WP
EF/EOL
N
SU
N
M
A
A
2
3/4/96
load drop
3488
0 1
FW
2bs018 b rfp
vibration
probe bumped
HF/OA
N
O
3
2
2
9/14/93
1/18/93
3/17/95
load drop
load drop
load drop
2553
893
832
0.1
0.0
00
FW
FW
FW
debris in
flow controller
rotor imbalance
2bs018 2b rfp
msc control switch misposition
b rfp tripped vibration probe
bumped
reactor feed pump trip
2b rfp vibration inspection
feedwater transient
fm
unknown
operator error
HF/PCM
EF
HF/OA
N/RRB
SU
N
M
M
O
2
Manual
Shutdown
10/26/95 load drop
2
12/1/95
Data sorted by system
219052
24869
7.9 FW Heating 5b fw htr
leak
5b fw htr repairs
errosion
EF/WP
N
A
0 9 FW Heating B5 fw heater
tube leaks
b fw string isolated 95% pwr
limit
unknown
EF/FP
N
A
Pagel71
PBAPS POC FL JORDAN
Unit
2
Date
1/5/94
Type
load drop
MWHr
5921
3
8/22/92
load drop
2385
Eff Out
Component
System
Days
0 2 FW Heating solenoid valve
0 1 FW Heating drain valve
27 May 1998
July 1992 - June 1997
Appendix 9. PBAPS Unavailability Data, sorted by failed system
Remarks
N
Class
A
HF/PCM
N
M
bearing cap gasket failure
EF/WP
N
A
3c fw htr drain w broken air
line high level
4c fw htr level oscillations
improper support
EF/FP
N
D
electrical comp end of life
EF/EOL
N
A
3c fw heater drain closed
wear
EF/WP
N
A
mtce
HF/PCM
N
M
poor solder joint
EF/WD
SU
M
poor solder joint
EF
SU
F
worn out
EF/EOL
N
A
Cause
Failure
coil failed
Description
5a heater extraciton w solenoid end of life
leak
feed water heater
out of calibration
5a fw htr repair from stm leak
oos
Category
EF/EOL
Issue #
positioner
0.1 FW Heating 5a fw htr extration gasket failure
stm valve
2
2/10/94
load drop
2116
2
11/8/95
load drop
329
0.0 FW Heating positioner
2
12/25/96 load drop
244
0 0 FW Heating level controller
3
11/17/95 load drop
143
0 0 FW Heating cv-3043c
3
2
Manual
Scram
10/15/96 Automatic
Scram
10/6/96 Automatic
Scram
2/2/96 Manual
Shutdown
10/9/96 Manual
Shutdown
12/18/92 Manual
Shutdown
3/27/96 load drop
3
3/27/96
load drop
2
6/24/93
load drop
3
12/1/93
2
1/12/95
3
6/23/96
3
positioner air
supply
dump valve failed
to open
steam seal
61863
ground resistor left field ground resistor-main
generator
in place
scram, gen lock out stator
3.4 Generator negative sequence short/open
unbalance
relay
scram, gen lock out stator
2 9 Generator negative sequence short/open
current unbalance
relay
main generator hydrogen leak
gasket leak
2.2 Generator bushing
46288
1.7 Generator bearing
hi temp
turbine bearing 12 high temp
electroysis
HF/C
SU
M
34334
1.2 Generator stator
h2 leaks
generator h2 leaks
sealent groove seal improper
HF/PI
SU
M
cal drifted low
generator core monitor alarm
low cal
HF/PI
N
M
4441
0 2 Generator generator core
monitor
0 2 Generator alarm setpoint
drifted low
generator core monitor alarm
cal
HF/PI
N
M
2120
0.1
HPCI
check valve
broken air line
repair hpci Injection check valve scaffold
HF/OA
N
O
Manual
Shutdown
load drop
392253
14.1
LPCI
mo-3-10-025 rhr
bent shaft
Ipci mov 25a inop
wrong nut
HF/PI
N
M
0 1 Main Steam msiv
packing leak
ao-86a repair
small leak
EFNVP
N
A
137576
4.9
stem binding
EF
SU
F
2 4
environmental conditions
EF/WD
N
D
3
3/7/93
load drop
5666
0.2
mtsv
#2 turbine control w stem
seperated
turbine cv limit switch bad
switch failed
testing logic
electrical problems mtsv replacement
clearance inadequate
65874
Main
Turbine
Main
Turbine
Main
Turbine
valve
8/26/95
Manual
Scram
load drop
failure
EF/EOL
N
A
3
7/14/92
225039
81
Offgas
EFNVD
SU
D
93042
33
Offgas
linkage alien set
slippage
failed close
design of air line
3/23/95
3239a linkage
failed
ao3466b
off gas system
3
Manual
Scram
Manual
Scram
sjae supply block valve failed
plug design
EFMID
N
D
2
2
3
2
2
2/3/94
Data sorted by system
102297
94620
82012
4763
1795
3.7 Generator breaker
limit switch
Page172
PBAPS POC FL JORDAN
MWHr
47859
Eff Out
Days
1.7
System
Offgas
Component
stm flow sensor
8/8/94
Type
Manual
Scram
load drop
35702
13
Offgas
fe5020
8/6/96
load drop
577
00
Offgas
352800
12 7
Reactor
2
2
Manual
Scram
9/22/93 load drop
12/13/92 load drop
control valve
9716b
fuel
35885
27598
1.3
1.0
Reactor
Reactor
2
3
2
9/10/94
7/26/92
9/21/93
load drop
load drop
load drop
14083
9791
7552
05
0.4
0.3
2
2/23/94
load drop
4577
3
3
2
5/9/93
1/23/93
5/14/94
3
3/8/97
load drop
load drop
Automatic
Scram
Manual
Scram
2
4/24/93
3
10/1/92
Class
O
EFNVD
SU
D
EF/WD
N
D
Category
HF/OA
flange flex
Issue #
failed open
leak
power reduction for fuel repair
pci
EF/WD
N
D
fuel
Iprm 56-41 &5643 cross
connected
clad
mtce
Flux tilt testing
Iprm mismatch
pci
hooked up wrong
EF/WD
HF/C
N
SU
D
M
Reactor
Reactor
Reactor
fuel
detector
fuel
clad
failed
clad
pci
age
pci
EF/WD
EF/EOL
EFWD
N
SU
SU
D
A
D
0.2
Reactor
fuel
clad
Flux tilt testing
tip machine a
admin precaution increase in
off gas level
rod pattern adj due to 5 leakers
pci
EF/WD
N
D
622
543
165366
0.0
0.0
59
Reactor
Reactor
Recirc
detector
fuel
a pump
lost signal
leak
Ivdt
#2 tip machine
flux tilt
recirc pump a speed increase
cable
pci
defective
EF/FP
EF/WD
EF/EOL
N
N
SU
A
D
A
150887
54
Recirc
3ap034-dr
loss of oil in
upper/lower mtr
brg reservoir
a recirc motor low oil level
unknown under investigation
HF
SU
M
Manual
Shutdown
load drop
128564
4.6
Recirc
leak
N
A
28
Recirc
rx instrument mismatch &recirc loss of level
pp
unknown
recirp pump control
EFNVP
78134
It 73a equilizing
valve
control loop
EF
N
D
7/27/92 Manual
Shutdown
69540
2 5 Recirc
Date
7/30/93
3
3
2
Remarks
N
Cause
operator action
Description
manual scram due to
recombiner
recombiner leak
troubleshooting
recombiner isolation
Unit
3
3
27 May 1998
July 1992 - June 1997
Appendix 9. PBAPS Unavailability Data, sorted by failed system
7/4/93
Failure
blown fuse
mo99/91
steam leak at flex
none mg set lock
up
PUMPXX
cable insulation
breakdown
3
4/9/97
load drop
27554
10
Recirc
low side
transformer cable
3
2
7/23/92
3/18/93
load drop
load drop
25191
22741
09
08
Recirc
Recirc
2
3
4/23/93
8/6/94
load drop
load drop
21119
11285
08
0.4
Recirc
Recirc
calc error
st-r-60a-2
loss of tach signal loss of contact for
brushes
vibrated shut
vent damper
brush pigtail
3ag004
shorted Inner and
outer collector
2
2
2
9/5/92 load drop
12/16/92 load drop
3/20/93 load drop
6974
6549
5992
0.3
0.2
0.2
Recirc
Recirc
Recirc
Data sorted by system
brecirc pump
oscillation
level switch
blind controller
controller
dead band
27792013
A
RECIRC PUMP TRIP AND
VESSEL TEMPERATURE
DIFFERENTIAL
Equipment Failure
EF/WP
3b recirc pump trip c phase
cable fault
cable treeing
EFWP
N/RPT
A
recirc pump
recirc pump b(gen hi
amps,volts
recirc mg set
b recirc pump brush
replacement
margin
no mtce
HF/PI
EF/WP
SU
SU
M
M
loss of cooling
mtce did not stand up leads
post mtce
EF/WP
HF/C
N
N/RPT
A
M
recirc pump b
recirc pp controls
recirc pump hi oil level
gain setting
worn parts
HF/PCM
EFV/WP
EF/ICS
N
SU
SU
M
A
D
Page173
PBAPS POC FL JORDAN
Appendix 9. PBAPS Unavailability Data, sorted by failed system
Type
load drop
load drop
load drop
load drop
load drop
MWHr
4636
2171
1946
746
953
251351
2
Automatic
Scram
8/17/92 Automatic
Scram
10/22/95 load drop
2
8/16/95
load drop
283
2
6/4/96
load drop
114
Unit
2
3
3
3
3
Date
1/29/93
9/30/92
9/19/95
11/19/93
2/5/96
3
7/4/92
2
Data sorted by system
61988
340
Eff Out
Days
0.2
0.1
0.1
0.0
00
Failure
coupling trip
bkr 2ak34b trip
loose connection
misaligned
defetive logic
switch
e313 cs & 343 su
9 0 Transmissio 3su feed lost
tran
n
2.2 Transmissio lock out due to no sub sta 205
bkr
n
cable fault
0 0 Transmissio 220-34 ug line
n
popped open su 25
0 0 Transmissio 220-8 line
bkr
n
breaker opened
0 0 Transmissio 220-8 line
n
System
Recirc
Recirc
Recirc
Recirc
RPS
Component
tach
3b recirc pump
3-2a-kO10a
coupling
pish-3-02-3-055c
27 May 1998
July 1992 - June 1997
Cause
failed
unknown
pm task Inadequate
Issue #
Remarks
N
N
N/RPT
SU
SU
Class
A
A
M
M
A
Description
recirc pump a trip
recirc pump
a recirc mg set tripped
recirc pump vibration alarm
5a k5c relay dropped out
failed
Category
EF/WP
EF
HF/PI
HF/PCM
EF/EOL
north substation xfmr
173 mtce
HF/PCM
N
M
generator lock out
written com to load dispacter
HF/PI
N
O
220-34 Ine tripped = positive
reactivity
220-8 line fault
failed
EF/FP
N
A
digging into line Unit 1 pl
HF/OA
N
O
220-8 line de energized
operator at sub opened
incorrectly
HF/OA
N
O
Page174
PBAPS POC FL JORDAN
Appendix 10. LGS unavailability data, sorted by component failure cause ( c l,
cw ...
27 May 1998
,)
Category
Issue #
Remarks
Class
HF
10006422
2F20, DF
A
Infantile failure
EF/WP
10005909
A
Loose
connection
EF
10004298
A
Unit
2
Date
12/16/96
Type
Shut down
MWHr
139,200
Eff Out Days
5 00
System
Condenser
Component
Expansion joint
Failure
Leak
Cause
1
7/25/96
Scram
93,663
3 36
EHC
FN card
Failure
2
8/8/95
Scram
80,794
2.90
FW
FW LCS
Loss of DC
power supply
(2K612)
2
9/23/92
Load drop
58,625
2 22
Condenser
Tube
Leaks
EF/WP
A
1
10/7/94
Load drop
31,945
1.21
Offgas
After
Condenser
Cleaning
EF/WP
A
1
5/7/95
Shut down
31,060
1.18
Recirc
Seal
Leak
2
9/1/94
Load drop
28,290
1 07
Main steam
Valve
Leak
2
2/24/95
Load drop
28,000
1.01
Air removal
SJAE nozzle
Broken
EF
DF
A
1
7/20/95
Load drop
23,000
0 87
Condenser
Water box
Cleaning
EF/WP
Normal
Maintenanc
e
A
EF
Resealed
A
1
1/2/95
Load drop
18,600
0.70
Recirc
Seal
Leak
2
6/1/96
Load drop
12,619
0 45
Condenser
Tube
Leaks
2
8/1/92
Load drop
10,750
0 41
FW heating
Vent line
Leaks
0.25
Main turbine
#2 MSV
Tube
Age
EF
Maintenanc
e outage
1M03
A
EF
Age
A
EF/WP
A
EF/WP
A
EF/WP
A
Leaks
EF/WP
A
FAC
2
9/1/94
Load drop
6,500
2
10/1/92
Load drop
6,375
0 24
Condenser
1
7/9/94
Load drop
6,290
0 24
Condenser
Tubes
Cleaning
EF/WP
A
2
9/8/94
Load drop
5,220
0.20
FW
Check valve
cap
Leak
EF
A
2
6/17/95
Load drop
4,970
018
FW heating
Dump valves
Actuator
EF
A
1
12/1/93
Load drop
4020
0.15
Condenser
Waterbox
Cleaning
EF/WP
A
2
7/18/92
Load drop
1,353
0 05
FW heating
Vent line
Leaks
EF/WP
A
2
4/1/92
Load drop
1,039
0 04
Main turbine
Vac switch
Set point drift
EF/WP
A
1
12/1/93
Load drop
930
0.04
FW
Pump
EF
A
2
4/1/92
Load drop
750
0.03
Condensate
Water box
Leaks
EF/WP
A
2
9/5/95
Load drop
370
0 01
FW heating
Valve
Leak
EF
A
Data Sorted by Failure Category
Page 175
Age
FAC
LGS POC VT ANGUS
27 May 1998
Appendix 10. LGS unavailability data, sorted by component failure cause
Cause
Broken bracket
/ severed
tubing
Issue #
10006385
Remarks
Repeat
event 2M19
Class
D
1E07
D
Unit
2
Date
12/6/96
Type
Manual
scram
MWHr
250,560
Eff Out Days
9 00
System
EHC
Component
Pressure switch
Failure
Leak
1
3/24/96
Shut down
203,260
7.30
Main steam
SRV
Leak
1
9/7/93
Scram
146,784
5.56
Electrical
Breaker
Failed to
reclose
Spurious
EF/WD
10000021
2
1/3/93
Scram
84,000
3.18
EHC
Relay
Hi pressure
Sporadic
anomaly
EF/WD
93-01-01
RF
10004338
D
2
8/20/95
Scram
69,396
2 49
EHC
Relay
High
impedance/ NC
contact
LTA design
EF/WD
10004338
RF, 93-0101
D
2
1/15/95
Load drop
48,000
1.82
Recirc
MG set
Generator
ground
Indeterminant
EF/WD
1
3/31/96
Load drop
47,222
1.70
EHC
Speed control
logic
Speed control
LVG
Sporadic
anomaly
EF/WD
2
5/9/92
Load drop
36,960
1.40
Condensate
Pump
Bearing
Misaligned
EF/WD
2
9/8/94
Load drop
28,650
1.09
SWC
Y-strainer
Clogged
Generator
hydrogen leak
EF/WD
Category
HF/LCA
EF/WD
2
10/6/96
Load drop
20,859
0 75
SWC
Y-strainer
Clogged
EF/WD
2
11/22/95
Load drop
15,253
0 55
SWC
Filters
Clogged
EF/WD
2
6/26/95
Load drop
14,115
0.51
Recirc
Temperature
switch
Spiked high
Spurious
EF/WD
1
6/19/93
Load drop
11,050
0.42
Condensate
Motor
Bearing
Misaligned
EF/WD
1
3/1/94
Load drop
10,800
0.41
FW
LCS
1
5/1/93
Load drop
7,313
0.28
Condensate
Motor
Bearing
Misaligned
EF/WD
2
6/2/92
Load drop
7,140
0 27
Condensate
Pump motor
Bearing
Misaligned
EF/WD
2
5/1/93
Load drop
6,410
0 24
Condensate
Pump motor
Bearing
Misaligned
EF/WD
2
4/7/93
Load drop
6,000
0.23
FW
FWLCS
'A'
level down
spike
Spurious,
Indeterminant
EF/WD
2
7/17/95
Load drop
5,600
0 20
Recirc
Temperature
switch
Spiked high
Spurious
EF/WD
1
10/8/94
Load drop
3,829
0.15
FW
LCS
Data Sorted by Failure Category
D
D
10005451
Infantile
failure
10002830
TT
D
Page 176
D
D
D
10004172
RPT
D
Realigned
D
EF/WD
EF/WD
D
D
Realigned
D
D
Realigned
D
93-04-04
RRB
D
10004172
RPT
D
D
LGS POC VT ANGUS
27 May 1998
Appendix 10. LGS unavailability data, sorted by component failure cause
Unit
1
Date
8/8/95
Type
Load drop
MWHr
3,740
Eff Out Days
0 14
System
FW
Component
Pressure switch
Failure
Premature
actuation
Cause
Out of
calibration
Category
EF/WD
Issue #
Remarks
Class
10004299
RRB
D
1
2/7/93
Load drop
3,510
0.13
FW
Pressure switch
Low pressure
Spurious
EF/WD
93-02-16
RRB
D
2
1/11/92
Load drop
3,430
0 13
FW
Turbine
controller
Malfunction
Relay
EF/WD
D
1
12/1/93
Load drop
3400
0.13
Main turbine
TCV pressure
switch
Failed
EF/WD
D
2
6/26/93
Load drop
2,967
0.11
Condensate
Pump
Bearing
Misaligned
EF/WD
Realigned
D
2
6/28/93
Load drop
2,967
0.11
Condensate
Pump
Bearing
Misaligned
EF/WD
Realigned
2
6/29/93
Load drop
2,967
0 11
Condensate
Pump
Bearing
Misaligned
EF/WD
Realigned
D
D
1
4/24/95
Load drop
2,230
0.08
FW
Pressure switch
Failed low
2
4/19/93
Load drop
2,000
0.08
FW
FWLCS
2
3/1/94
Load drop
1,504
0.06
Condensate
Pump
1
11/1/94
Load drop
1,200
0.05
Computer
P1 program
Would not run
1
10/16/92
Load drop
990
0.04
Recirc
RX level signal
Low level
sensed
Channel noise
EF/WD
1
8/28/93
Load drop
734
0.03
Condensate
Pump
Bearinng
Misaligned
EF/WD
1
4/11/96
Load drop
370
0.01
FW heating
Drain cooler
Tube leak
Poor Design
2
2/10/92
Load drop
FW
Pump
Trip
EF/WD
1
8/20/95
Shut down
342,276
12.97
Reactor
Fuel
Leak
EF
1
12/23/96
Load drop
5,427
0 19
SWC
Temperature
CV
Loose adjust
arm screws
1
11/1/94
Load drop
900
0.03
1
7/15/96
Manual
scram
182,179
6 54
Reactor
Air removal
Fuel
SJAE
Y-strainer
Leak
Steam leak
1
6/17/96
Load drop
22,121
0 79
Electrical
Output breaker
Low pressure
rupture
1
8/28/95
Shut down
121,440
4.60
Drywell
Flange
Misaligned
Data Sorted by Failure Category
EF/WD
Master level
Spurious,
controller down Indeterminant
spike
EF/WD
D
93-04-18
RRB
D
EF/WD
Page 177
EF
Manufacturinng
D
92-10-21
RRB
D
Realigned
D
EF/WD
D
D
Maintenanc
e outage
HF/C
10006438
EF
HF/C
10005889
Improper
installation
HF/C
10005797
Poor Corr
Maint
HF/C
35295006
Improper
installation
D
F
F
F
TT
M
LGS POC VT ANGUS
27 May 1998
Appendix 10. LGS unavailability data, sorted by component failure cause
Remarks
Dual unit
scram
Class
OM
M
10006441
Repeat
Event
M
HF/C
92-11-20
RF,
10005615
M
PM LTA
HF
10003606
Dual unit
scram
M
Leak
Weld repair
failure
HF/PCM
10005615
RF 92-1120
M
011 leak
Indeterminant
EF
93-05-20
TT
M
Spurious
HF/MS
92-01-10
No PM
M
DF
M
M
SU, TT, PM
deferred
M
Unit
2
Date
2/21/95
Type
Scram
MWHr
86,672
Eff Out Days
3.11
System
Transmission
system
Component
Breaker/
Relays
Failure
Failed to
actuate
Cause
PM LTA
Category
HF
1
1/14/94
Scram
59,230
2 24
SWC
Trip circuit
Short
Bulb
installation
EF
2
12/24/96
Manual
scram
55,680
2.00
Recirc
Scoop tube ball
joint
Broke
Vibration
Induced
HF/LCA
2
11/19/92
Load drop
44,101
1 67
EHC
#3 CV piping
Leak
Weld failure
1
2/21/95
Scram
41,353
1.57
Failed to
actuate
2
4/29/96
Load drop
38,939
1.40
EHC
#3 CV piping
2
5/16/93
Load drop
16,493
0 62
EHC
#2 TCV servo
2
6/3/95
Load drop
17,150
0.62
EHC
#4 CIV
Leak
1
1/21/92
Load drop
7,030
0.27
FW heating
Logic
High level
sensed
Transmission Breaker/ relays
system
Issue #
10003606
EF
M
M
1
8/2/96
Load drop
1,468
0.05
FW heating
Drain valve
Malfunction
HF/PCM
2
10/1/93
Load drop
1,090
0 04
Circ water
Pump
Maintenance
EF/WP
1
1/17/94
Load drop
940
0 04
Main turbine
Moisture
separator dump
valve
Valve
positioner
PM frequency
LTA
HF/MS
1
8/23/93
Load drop
725
0.03
Gasket
Failed
EF
1
9/11/95
Scram
369,991
14 01
Main steam
SRV
Opened
Pilot seat
erosion
HF/MS
10004442
2
12/4/92
Manual
scram
157,870
5 98
Recirc
Recirc pump
HF/OA
92-12-01
2
5/14/96
Scram
89,099
3.20
Generator
Volts/Hz relay
Actuation
inappropriate
MOD package
LTA
HF/PI
10005652
2
10/19/94
Scram
77,754
2 95
Electrical
D24 Bus
De-energized
Inadvertent
HF/OA
1
5/21/96
Scram
75,095
2.70
RPS
Logic
No 1/2 scram
alarm
Indeterminate
HF/PI
10005675
O
2
3/26/93
Scram
56,700
2.15
EHC
#6 ISV
Perturbation in
ETS/RETS
Air entrap in
control pack
HF/PI
93-03-38
O
1
9/2/95
Load drop
39,426
1.49
H2
Recombiner
Recorder
Logic
MOD PMT LTA
HF/PI
10004403
O
Data Sorted by Failure Category
Instrument Air Dryer package
EOC-RPT logic Breaker tripped
Page 178
10001318
M
O
Occurred
during ST
O
O
O
LGS POC VT ANGUS
27 May 1998
Appendix 10. LGS unavailability data, sorted by component failure cause
Unit
2
Date
10/12/96
Type
Load drop
MWHr
37,463
Eff Out Days
1.35
System
Condensate
Component
Pump motor
Failure
Vibration
Cause
FME
Category
Issue #
HF/PI
10006201
2
2/24/94
Load drop
10,080
0 38
Recirc
Pump
Trip
Fuses pulled
HF/OA
2
4/29/95
Load drop
10,000
0 36
Recirc
Coupler bypass
valve
Incorrect
position
LTA
procedures
HF/PI
10003924
Remarks
Class
0
O
RRB
O
1
7/13/92
Load drop
8,566
0 32
FW
RFP turbine
Insulation fire
Oil soaked
HF/PI
92-07-10
OM
O
2
3/17/93
Load drop
8,030
0 30
SWC
Valve
Valve
mispositioned
LTA procedure
HF/PI
93-03-24
SU, DF
O
2
2/16/94
Load drop
7,150
0 27
Recirc
Recorder
Mislabelled
LTA MOD
review/ PMT
HF/PI
10001491
RRB
O
2
9/2/95
Load drop
7,470
0 27
H2
Recombiner
Recorder
Logic
MOD PMT LTA
HF/PI
10004403
1
1/30/95
Load drop
7,000
0.27
Recirc
MG set
Trip
Bumped
HF/OA
2
6/28/95
Load drop
6,300
0.23
FW
FW UPS
Power
disconnect
switch off
Bumped during
cleaning
HF/OA
10004173
RRB
O
1
1/1/92
Load drop
2,904
0 11
Main turbine
Pressure
instrument RV
RV closed
Mispositioned
HF/OA
92-01-02
SU, DF, TT
O
1
7/19/95
Load drop
700
0 03
Recirc
MG set
Perturbations
Operator
HF/OA
1
7/29/94
Load drop
570
0 02
FW
Trip lever
Actuated
Inadvertent
HF/OA
2
6/6/96
Load drop
378
0 01
Isophase bus
cooling
Fan
Trip
LTA WO/
procedures
HF/PI
Total effective
outage days
135 74
Categ
ory
Code
EF Equipment
Fators
Data Sorted by Failure Category
WD - Weak
Design
WP - Worn
Parts
ICS Inadequate
Control System
EOL - End of
Life
Page 179
O
O
O
O
10005731
0
FP - Fatigued
Parts
LGS POC VT ANGUS
27 May 1998
Appendix 10. LGS unavailability data, sorted by component failure cause
Unit
I
Date
MWHr
Type
Component
Failure
I
Cause
Category
Issue #
Remarks
Class
CCraftsmanship
OA - Operator
Actions
MSManagement
Standards
LCA - Less than
adeq corr
actions
PCM - Poor
Corrective
Maintenance
A -Age
F - Fabrication
D - Design
I-
MMaintenance
O - Operation
DF Dependent
Failure
TT - Turbine
Trip
OM - On line
Maintenance
RF - Repeat
Failure
RRB Recirc
Runback
RPT - Recirc
pump trip
DF - Dependent
Failure
SU - During
Start up
Installation
Rema
rk
Code
System
PI - Procedural
Inadequacy
HFHuman
Factors
Class
Code
Eff Out Days
Data Sorted by Failure Category
Page 180
LGS POC VT ANGUS
Appendix 11. PBAPS Unavailability Data, sorted by component failure cause
Unit
2
2
2
2
2
System
Electrical
MWHr
248585
69540
2 5 Recirc
Manual
Shutdown
Automatic
Scram
load drop
61863
2.2 Generator bushing
29188
10
FW
27554
10
Recirc
10/26/95 load drop
24869
0 9 FW Heating B5 fw heater
08
3
2/2/96
3
8/1/95
3
4/9/97
2
Eff Out
Days
89
Type
Automatic
Scram
12/1/95 Manual
Shutdown
5/14/94 Automatic
Scram
4/24/93 Manual
Shutdown
7/27/92 Manual
Shutdown
Date
7/17/92
Component
3435 breaker
27 May 1998
128564
4 6
Recirc
It 73a equilizing
valve
PUMPXX
leak
gasket leak
rx instrument mismatch &recirc loss of level
pp
Equipment Failure
RECIRC PUMP TRIP AND
VESSEL TEMPERATURE
DIFFERENTIAL
main generator hydrogen leak worn out
upscale
scram-feedwater transient
cable insulation
breakdown
tube leaks
power supply fw
control system
failure of PS
0.8
Recirc
0.5 Condenser
vent damper
HTEXCH
vibrated shut
04
2cs018 2c rfp
control valve
bearing seized
failed
controller
coil failed
21119
13908.0
2
3
2
2
7/26/92 load drop
12/16/92 load drop
load drop
1/5/94
9791
6549
5921
0.4
Reactor detector
oscillation
0.2
Reclrc
0 2 FW Heating solenoid valve
10171
EF/WP
N
A
EF/WP
Issue #
A
27792013
EF/EOL
N
A
failed transmitter on card
EF/EOL
N/TT
A
3b recirc pump trip c phase
cable fault
cable treeing
EF/WP
N/RPT
A
b fw string isolated 95% pwr
limit
power ascension &c rfp
problems
unknown
EF/FP
N
A
age
EF/EOL
N
A
loss of cooling
EF/WP
EF/WP
N
A
A
failed
EF/WP
N
A
EF/EOL
EF/WP
EF/EOL
SU
SU
N
A
A
A
failure
EF/EOL
N
A
recirc mg set
CLEAN CONDENSER
WATERBOXES.
rfp c slow responce
age
tip machine a
worn parts
recirc pp controls
5a heater extraciton w solenoid end of life
mtsv
electrical problems mtsv replacement
tach
2bs018 b rfp
3b rfp speed
controller &hjsv
coupling trp
probe failure
sol coil burned
and controller
degradded
recirc pump a trip
b rfp high vibration
3b fpr speed control problem
failed
worn parts
age
EF/WP
EF/WP
EF/EOL
N
SU
N
A
A
A
0.1 Control Rod rod 26-15
Drive
3b recirc pump
Recirc
01
N2 leak at
charging block
bkr 2ak34b trip
Inop control rod
O-nng replaced
EF/WP
N
A
recirc pump
unknown
EF
N
A
0.1 FW Heating 5a fw htr extration
stm valve
gasket failure
5a fw htr repair from stm leak
oos
bearng cap gasket failure
EF/WP
N
A
3
3/7/93
load drop
5666
02
2
2
3
1/29/93
10/5/96
6/13/97
load drop
load drop
load drop
4636
4263
4238
0.2
0.2
0.2
2
4/28/94
load drop
2537
3
9/30/92
load drop
2171
2
2/10/94
load drop
2116
CL/L.lr
FW
A
defective
recirc pump a speed increase
4/23/93 load drop
12/19/92 Manual
Shutdosn
1/22/93 load drop
SU
errosion
Ivdt
2
3
EF/EOL
5b fw htr repairs
a pump
FW
A
leak
Recirc
21995
N
Category
EF
59
load drop
EFWNP
Cause
lighting stnke
165366
6/3/95
Class
A
Description
rwcu controls
7.9 FW Heating 5b fw htr
3a rfp speed
controller
low side
transformer cable
Remarks
N
Failure
trpped
219052
2
Data Sorted by Failure Category
July 1992 - June 1997
Main
Turbine
Recirc
FW
FW
C-IICII~711)
Page 181
PBAPS POC FL JORDAN
Appendix 11. PBAPS Unavailability Data, sorted by component failure cause
Eff Out
Component
System
Days
0 1 Main Steam msiv
cable
O-ring damaged/cut
EF/FP
EFNVP
N
N
A
A
failed
EF/FP
N
A
electrical comp end of life
EF/EOL
N
A
wear
EF/WP
N
A
failed
detector
hv-1 11
#2 tip machine
hcu hv-111 broken
220-34 ug line
cable fault
level controller
dump valve failed
to open
steam seal
220-34 line tripped = positive
reactivity
4c fw htr level oscillations
3c fw heater drain closed
load drop
953
00
3
2
5/9/93
2/3/96
load drop
Iload drop
622
492
2
10/22/95 load drop
340
2
12/25/96 load drop
244
0 0 Reactor
0.0 Control Rod
Dnrive
0.0 Transmissio
n
0 0 FW Heating
3
11/17/95 load drop
143
0 0 FW Heating cv-3043c
2
Manual
Shutdown
7/4/93 Manual
Scram
7/14/92 Manual
Scram
10/11/94 Automatic
Scram
pish-3-02-3-055c
Cause
Issue #
pressure
transmitter
fuel
did not work
turbine control valve oscillations design
EF/WD
N
D
leak
power reduction for fuel repair
pci
EF/WD
N
D
3239a linkage
failed
main power
transformer for
inverter
linkage allen set
slippage
shorted winding
off gas system
design of air line
EFNVD
SU
D
EF/WD
N/RRB
D
Offgas
ao3466b
failed close
sjae supply block valve failed
plug design
EF/WD
N
D
2 8
Recirc
control loop
recirp pump control
unknown
EF
N
D
pressure
transimtiter
limit switch
none mg set lock
up
did not work
turbine control vie oscillations
design
EF/WD
SU
D
environmental conditions
EF/WD
N
D
pci
flange flex
EF/WD
EFNVD
N
SU
D
D
N/RRB
D
EF/WD
EF/WD
N
N
D
D
418199
15.0
EHC
352800
12.7
Reactor
225039
81
Offgas
99423
3.6
Electrical
93042
3.3
78134
3
3/23/95
3
10/1/92
Manual
Scram
load drop
2
12/21/92
load drop
69152
2.5
EHC
3
8/26/95
load drop
65874
2.4
2
3
9/22/93
8/8/94
load drop
load drop
35885
35702
13
13
Main
Turbine
Reactor
Offgas
3
4/21/97
load drop
35167
13
FW
2
3
9/10/94
4/27/94
load drop
load drop
14083
12881
05
05
Reactor
FW
2
12/17/92 load drop
10398
0.4
EHC
2
9/21/93
load drop
7552
0.3
Reactor
2
2
3/20/93
2/23/94
load drop
load drop
5992
4577
02
0.2
Recirc
Reactor
Data Sorted by Failure Category
A
5a k5c relay dropped out
2/5/96
1
SU
defetive logic
switch
lost signal
broken o-ring
3
3
EFIEOL
Category
EF/WP
MWHr
1795
3
Class
A
small leak
Type
load drop
3
Remarks
N
Description
ao-86a repair
Date
1/12/95
1/1/93
27 May 1998
Failure
packing leak
Unit
2
RPS
July 1992 - June 1997
fuel
fe5020
computer dcc-x
scram loss of static inverter y50 break down of insulation
turbine cv limit switch bad
testing logic
Flux tilt testing
clad
steam leak at flex recombiner leak
troubleshooting
feedwater computer trouble
power supply
relay actuation
fail/transfer of
control
switch failed
power supply? transfer - design EFWD
issue
clad
control valve
actuator binding
did not work
Flux tilt testing
rfp control problem
turbine control valve oscillations design
EF/WD
SU
D
fuel
clad
EF/WD
SU
D
level switch
fuel
dead band
clad
pci
admin precaution increase in
off gas level
recirc pump hi oil level
rod pattern adj due to 5 leakers pci
EF/ICS
EF/WD
SU
N
D
D
fuel
rfp
pressure
transimtiter
Page 182
pci
not smooth operation
PBAPS POC FL JORDAN
Appendix 11. PBAPS Unavailability Data, sorted by component failure cause
Unit
3
2
Date
6/12/94
5/9/96
Type
load drop
load drop
MWHr
4054
2148
Eff Out
Days
01
01
3
8/6/96
load drop
577
0.0
3
2
1/23/93
11/8/95
load drop
load drop
543
329
3
Manual
Scram
10/6/96 Automatic
Scram
11/11/95 load drop
12/1/93 Manual
Shutdown
Automatic
7/4/92
Scram
3/7/93 Automatic
Scram
3/8/97 Manual
Scram
2
2
3
3
3
3
6/23/96
82012
30423
392253
Component
muffin fan
wire lug
Offgas
control valve
9716b
0.0 Reactor fuel
0.0 FW Heating positioner
failed open
167298
150887
54
102297
37
Recirc
3ap034-dr
Description
ehc elec cabinet cool fan
turbine control #2 valve
oscillation
recombiner isolation
N
D
clearance inadequate
EF
SU
F
poor solder joint
EF
SU
F
foreign material
wrong nut
HF/C
HF/PI
SU
N
1
M
e313 cs & 343 su
tran
hi vibration
north substation xfmr
173 mtce
HF/PCM
N
M
maintenance outage
poor lubrication to vib sensor
HF/PCM
N
M
loss of oil in
upper/lower mtr
brg reservoir
a recirc motor low oil level
unknown under investigation
HF
SU
M
mtce
HF/PCM
N
M
poor solder joint
EF/WD
SU
M
electroysis
HF/C
SU
M
46288
34334
1 2 Generator stator
27598
10
Reactor
lprm 56-41 &5643 cross
connected
Generator breaker
load drop
load drop
25191
22741
09
0.8
Reclrc
Recirc
3
6/22/96
load drop
20785
07
EHC
3
5/21/93
load drop
18117
07
FW
2
4/1/97
load drop
17756
06
EHC
recirc pump
calc error
st-r-60a-2
loss of tach signal loss of contact for recirc pump b(gen hi
amps,volts
brushes
repair #4 cv ehc leak and msv
leak
servo
leak
fp turbine control parts missing post rework mgu hyd jack solenoid
mtce
sv7
ehc fluid leak
restriction
cooler
3
8/6/94
load drop
11285
04
Recirc
3ag004
Data Sorted by Failure Category
EF/WD
D
D
7/23/92
3/18/93
2
Class
D
D
Issue #
N
N
3
2
2
Remarks
SU
N
Category
EF/WD
EF/WD
EF/WD
EF/FP
ground resistor left field ground resistor-main
generator
in place
scram, gen lock out stator
3 4 Generator negative sequence short/open
unbalance
relay
turbine bearing 12 high temp
hi temp
1 7 Generator bearing
94620
Cause
unknown
vibration
pci
improper support
2
2
27 May 1998
flux tilt
3c fw htr drain vv broken air
line high level
#2 turbine control w stem
seperated
scram, gen lock out stator
current unbalance
a rfp vibration 95% pwr limit
Ipci mov 25a Inop
leak
positioner air
supply
stem binding
Main
valve
Turbine
2.9 Generator negative sequence short/open
relay
n/a
FW
2ap001
1.1
LPCI
mo-3-10-025 rhr bent shaft
14.1
49
9.0 Transmissio 3su feed lost
n
FW
c rfp
6.0
251351
Failure
stopped working
loose wire
Manual
Scram
10/15/96 Automatic
Scram
10/9/96 Manual
Shutdown
12/18/92 Manual
Shutdown
12/13/92 load drop
3
2/3/94
137576
System
EHC
EHC
July 1992 - June 1997
h2 leaks
generator h2 leaks
sealent groove seal improper
HF/PI
SU
M
mtce
Iprm mismatch
hooked up wrong
HF/C
SU
M
margin
no mtce
HF/PI
EF/WP
SU
SU
M
M
o-ring failure
HF/C
N
M
mtce left out parts
HF/PCM
N
M
suspect FME root cause TBD
HF/C
N
M
mtce did not stand up leads
post mtce
HF/C
N/RPT
M
brush pigtail
shorted inner and
outer collector
b recirc pump brush
replacement
Page 183
PBAPS POC FL JORDAN
Appendix 11. PBAPS Unavailability Data, sorted by component failure cause
Eff Out
Days
0.3
03
0.2
System
FW
Recirc
FW
Component
rfp
b recirc pump
2as018
Remarks
N
N
SU
Class
M
M
M
HF/PCM
N
M
HF/PI
N
M
cal
HF/PI
N
M
reactor feed pump trouble
o ring failure, rod 3431
accumulator
debris in trip dump valve
worn
HF/PI
EF/WP
N
N
M
M
debris in
leak
reactor feed pump trip
feed water heater
fm
out of calibration
HF/PCM
HF/PCM
N/RRB
N
M
M
loose connection
restriction
a recirc mg set tripped
ehc fluid leak
pm task inadequate
suspect FME root cause TBD
HF/PI
HF/C
N/RPT
N
M
M
rotor imbalance
misaligned
loss of fw htg
unknown
2b rfp vibration inspection
recirc pump vibration alarm
loss of power supply to e22 bus loss of power
EF
HF/PCM
HF/C
SU
SU
N
M
M
M
loss of pwr to
transformer
condensate pump a tripped
operator opened bkr
HF/OA
N
O
PCIS GROUP I ISOLATION
CAUSED BY BUMPING
INSTRUMENTATION.
Equipment Failure
HF/OA
2nd ground
HF/OA
N
O
Failure
control system
blind controller
speed controller
siezed
cleanliness
Description
a rfp maintenance
recirc pump b
a rfp maintenance
Cause
not smooth operation
gain setting
improper lubrication
Category
EFNVP
HF/PCM
HF/PCM
low condenser vacuum
cal drifted low
generator core monitor alarm
chlorine oos and warm water
and inst at cal limit
low cal
drifted low
generator core monitor alarm
Unit
3
2
2
Date
5/1/94
9/5/92
1/8/94
Type
load drop
load drop
load drop
MWHr
7634
6974
5813
2
6/24/94
load drop
4904
0 2 Circ Water condenser
2
3/27/96
load drop
4763
3
3/27/96
load drop
4441
0 2 Generator generator core
monitor
0 2 Generator alarm setpoint
2
3
4/2/97 load drop
11/28/93 load drop
4271
3651
failed to trip
FW
2as018 a rfp
02
0.1 Control Rod 34-31 accumulator failed
o-ring
Drive
3
3
9/14/93
8/22/92
load drop
load drop
2553
2385
3
2
9/19/95
4/2/97
load drop
load drop
1946
1793
flow controller
FW
0.1
0 1 FW Heating drain valve
positioner
3-2a-k010a
0.1
Recirc
EHC
cooler
01
2
3
2
1/18/93 load drop
11/19/93 load drop
5/19/93 load drop
893
746
678
2
3
0.0
0.0
00
FW
Recirc
Electncal
2bs018 2b rfp
coupling
e322 trip
Automatic
Scram
10/15/92 Automatic
Scram
398510
14 3 Condensate 2c cond pumps
361608
13 0 Containment VALVEX
3/2/93
4.1
DC
3
12/2/95
Automatic
Scram
114994
2
8/17/92
61988
3
7/30/93
3
1/20/95
Automatic
Scram
Manual
Scram
load drop
2
3/4/96
load drop
3488
0.1
FW
3
11/6/95
load drop
3336
01
Electncal
Data Sorted by Failure Category
cracked terminal
strip
47859
2 2 Transmissio lock out due to no
n
bkr
stm flow sensor
1.7
Offgas
13194
0.5 Circ Water screens immobile
27 May 1998
July 1992 - June 1997
turbine trip - pos & neg ground
2nd ground by
person working on
equipment
Issue #
O
27892008
sub sta 205
generator lock out
wntten com to load dispacter
HF/PI
N
O
blown fuse
mo99/91
pin shear
manual scram due to
recombiner
b screen immobile, b cw pp
removed from service
b rfp tripped vibration probe
bumped
13 kv electrical system-loose
fuse
operator action
HF/OA
N
O
wrong pin
HF/PI
N
O
probe bumped
HF/OA
N
O
droped holder damage
HF/OA
N
O
2bs018 b rfp
vibration
fuse
loose
Page 184
PBAPS POC FL JORDAN
Appendix 11. PBAPS Unavailability Data, sorted by component failure cause
Unit
3
Type
Date
12/29/93 load drop
MWHr
2266
Eff Out
Days
0 1
System
Electrical
Component
3-2a-k004a
Description
Failure
deenergized when recirc runback a pump
480v load center
30B01 was being
restored from the 3
4G4 tie breaker
check valve
broken air line
2
6/24/93
load drop
2120
01
HPCI
2
6/10/95
load drop
835
00
Electrical
2
3/17/95
load drop
832
00
FW
2
8/16/95
load drop
283
2
6/4/96
load drop
114
0 0 Transmissio 220-8 line
n
0 0 Transmissio 220-8 line
n
Data Sorted by Failure Category
loss of pwr to
panel y-34
msc control switch misposition
e22 bus
27 May 1998
July 1992 - June 1997
Cause
30b01 Ic being restored
Category
HF/PI
Issue #
Remarks
N/RRB
Class
O
repair hpci Injection check valve scaffold
HF/OA
N
O
loss of power supply to e22 bus Diesel feedback signal during
mod testing
&y-34
operator error
feedwater transient
HF/PI
N
O
HF/OA
N
O
digging into line Unit 1 pl
HF/OA
N
O
operator at sub opened
incorrectly
HF/OA
N
O
popped open su 25 220-8 line fault
bkr
220-8 line de energized
breaker opened
Page 185
PBAPS POC FL JORDAN
Appendix 12. LGS unavailability data, sorted by root cause of failure
27
9,
2j7. May 11998
( e4
Remarks
Class
Maintenanc
e outage
F
Category
EF
Issue #
Loose
connection
EF
10004298
Short
Bulb
installation
EF
OM
M
Leak
Age
EF
Maintenanc
e outage
1M03
A
Unit
1
Date
8/20/95
Type
Shut down
MWHr
342,276
Eff Out Days
12 97
System
Reactor
Component
Fuel
Failure
Leak
Cause
2
8/8/95
Scram
80,794
2 90
FW
FW LCS
Loss of DC
power supply
(2K612)
1
1/14/94
Scram
59,230
2.24
SWC
Trip circuit
1
5/7/95
Shut down
31,060
1.18
Recirc
Seal
2
9/1/94
Load drop
28,290
1.07
Main steam
Valve
Leak
EF
2
2/24/95
Load drop
28,000
1.01
Air removal
SJAE nozzle
Broken
EF
DF
A
1
1/2/95
Load drop
18,600
0 70
Recirc
Seal
Leak
Age
EF
Resealed
A
2
5/16/93
Load drop
16,493
0 62
EHC
#2 TCV servo
Oil leak
Indeterminant
EF
TT
M
2
6/3/95
Load drop
17,150
0 62
EHC
#4 CIV
Leak
EF
M
2
9/8/94
Load drop
5,220
0.20
FW
Check valve
cap
Leak
EF
A
2
6/17/95
Load drop
4,970
0.18
FW heating
Dump valves
Actuator
EF
A
1
11/1/94
Load drop
1,200
0.05
Computer
P1 program
Would not run
EF
D
1
12/1/93
Load drop
930
0.04
FW
Pump
EF
A
1
11/1/94
Load drop
900
0.03
Reactor
Fuel
1
8/23/93
Load drop
725
0.03
Instrument Air Dryer package
Age
A
A
93-05-20
Leak
EF
F
Gasket
EF
M
Failed
A
2
9/5/95
Load drop
370
0 01
FW heating
Valve
Leak
EF
1
3/24/96
Shut down
203,260
7 30
Main steam
SRV
Leak
EF/WD
1
9/7/93
Scram
146,784
5.56
Electrical
Breaker
Failed to
reclose
Spurious
EF/WD
10000021
2
1/3/93
Scram
84,000
3.18
EHC
Relay
Hi pressure
Sporadic
anomaly
EF/WD
93-01-01
RF
10004338
D
2
8/20/95
Scram
69,396
2.49
EHC
Relay
High
impedance/ NC
contact
LTA design
EF/WD
10004338
RF, 93-0101
D
2
1/15/95
Load drop
48,000
1.82
Recirc
MG set
Generator
ground
Indeterminant
EF/WD
Data Sorted by root cause
Page 186
1E07
D
D
D
LGS POC VT ANGUS
27 May 1998
Appendix 12. LGS unavailability data, sorted by root cause of failure
Failure
Speed control
LVG
Cause
Sporadic
anomaly
Category
Issue #
Remarks
Class
EF/WD
10005451
Infantile
failure
D
Pump
Bearing
Misaligned
EF/WD
Y-strainer
Clogged
Generator
hydrogen leak
EF/WD
10002830
TT
Component
Speed control
logic
MWHr
47,222
Eff Out Days
1 70
System
EHC
Load drop
36,960
1 40
Condensate
Load drop
28,650
1 09
SWC
10/6/96
Load drop
20,859
0 75
SWC
Y-strainer
Clogged
EF/WD
11/22/95
Load drop
15,253
0.55
SWC
Filters
Clogged
EF/WD
6/26/95
Load drop
14,115
0 51
Recirc
Temperature
switch
Spiked high
Spurious
EF/WD
1
6/19/93
Load drop
11,050
0 42
Condensate
Motor
Bearing
Misaligned
EF/WD
1
3/1/94
Load drop
10,800
0 41
FW
LCS
1
5/1/93
Load drop
7,313
0.28
Condensate
Motor
Bearing
Misaligned
EF/WD
2
6/2/92
Load drop
7,140
0.27
Condensate
Pump motor
Bearing
Misaligned
EFM/D
2
5/1/93
Load drop
6,410
0.24
Condensate
Pump motor
Bearing
Misaligned
EF/WD
EF/WD
EF/WD
Unit
1
Date
3/31/96
Type
Load drop
2
5/9/92
2
9/8/94
2
2
2
D
D
D
D
10004172
RPT
D
Realigned
D
EF/WD
D
Realigned
D
D
Realigned
D
93-04-04
RRB
D
10004172
RPT
D
2
4/7/93
Load drop
6,000
0.23
FW
FWLCS
'A' level down
spike
Spurious,
Indeterminant
2
7/17/95
Load drop
5,600
0.20
Recirc
Temperature
switch
Spiked high
Spurious
1
10/8/94
Load drop
3,829
0 15
FW
LCS
1
8/8/95
Load drop
3,740
0.14
FW
Pressure switch
Premature
actuation
Out of
calibration
EF/WD
10004299
RRB
D
1
2/7/93
Load drop
3,510
0 13
FW
Pressure switch
Low pressure
Spurious
EF/WD
93-02-16
RRB
D
2
1/11/92
Load drop
3,430
0.13
FW
Turbine
controller
Malfunction
Relay
EF/WD
D
1
12/1/93
Load drop
3400
0 13
Main turbine
TCV pressure
switch
Failed
EF/WD
D
2
6/26/93
Load drop
2,967
0 11
Condensate
Pump
Bearing
Misaligned
EF/WD
Realigned
D
2
6/28/93
Load drop
2,967
0.11
Condensate
Pump
Bearing
Misaligned
EF/WD
Realigned
D
2
6/29/93
Load drop
2,967
0.11
Condensate
Pump
Bearing
Misaligned
EF/WD
Realigned
D
1
4/24/95
Load drop
2,230
0 08
FW
Pressure switch
Failed low
Data Sorted by root cause
D
EF/WD
Page 187
EF/WD
D
LGS POC VT ANGUS
Appendix 12. LGS unavailability data, sorted by root cause of failure
Unit
2
Date
4/19/93
Type
Load drop
MWHr
2,000
2
3/1/94
Load drop
1
10/16/92
Load drop
1
8/28/93
1
4/11/96
2
2/10/92
Load drop
1
7/25/96
Scram
93,663
2
9/23/92
Load drop
1
10/7/94
Load drop
1
7/20/95
2
2
27 May 1998
Category
Issue #
Remarks
Class
EF/WD
93-04-18
RRB
D
92-10-21
RRB
D
Realigned
D
Eff Out Days
0.08
System
FW
1,504
0 06
Condensate
Pump
990
0.04
Recirc
RX level signal
Load drop
734
0.03
Condensate
Pump
Load drop
370
0 01
FW heating
Drain cooler
FW
Pump
Trip
3.36
EHC
FIV card
Failure
58,625
2.22
Condenser
Tube
Leaks
EF/WP
A
31,945
1.21
Offgas
After
Condenser
Cleaning
EF/WP
A
Load drop
23,000
0 87
Condenser
Water box
Cleaning
EF/WP
6/1/96
Load drop
12,619
0 45
Condenser
Tube
Leaks
8/1/92
Load drop
10,750
0 41
FW heating
Vent line
Leaks
2
9/1/94
Load drop
6,500
0 25
Main turbine
#2 MSV
2
10/1/92
Load drop
6,375
0 24
Condenser
Tube
1
7/9/94
Load drop
6,290
0 24
Condenser
1
12/1/93
Load drop
4020
0 15
2
7/18/92
Load drop
1,353
2
10/1/93
Load drop
2
4/1/92
2
Component
FWLCS
Failure
Cause
Master level
Spurious,
controller down Indeterminant
spike
D
EF/WD
Low level
sensed
Channel noise
EF/WD
Bearing
Misaligned
EF/WD
Tube leak
Poor Design
EF/WD
D
D
EF/WD
Infantile failure
EF/WP
10005909
A
Normal
Maintenanc
e
A
EF/WP
A
EF/WP
A
EF/WP
A
Leaks
EF/WP
A
Tubes
Cleaning
EF/WP
A
Condenser
Waterbox
Cleaning
EF/WP
A
0.05
FW heating
Vent line
Leaks
EF/WP
A
1,090
0.04
Circ water
Pump
Maintenance
EF/WP
M
Load drop
1,039
0.04
Main turbine
Vac switch
Set point drift
EF/WP
A
4/1/92
Load drop
750
0.03
Condensate
Water box
Leaks
EF/WP
2
12/16/96
Shut down
139,200
5 00
Condenser
Expansion joint
Leak
HF
10006422
2F20, DF
A
2
2/21/95
Scram
86,672
3 11
Transmission
system
Breaker /
Relays
Failed to
actuate
PM LTA
HF
10003606
Dual unit
scram
M
1
2/21/95
Scram
41,353
1.57
Transmission Breaker / relays
Failed to
PM LTA
HF
10003606
Dual unit
M
system
Data Sorted by root cause
actuate
Page 188
FAC
FAC
A
scram
LGS POC VT ANGUS
27 May 1998
Appendix 12. LGS unavailability data, sorted by root cause of failure
Class
Unit
1
Date
7/15/96
Type
Manual
scram
MWHr
182,179
Eff Out Days
6 54
System
Air removal
Component
SJAE
Y-strainer
Failure
Steam leak
Cause
Improper
installation
Category
HF/C
Issue #
10005889
1
8/28/95
Shut down
121,440
4 60
Drywell
Flange
Misaligned
Poor Corr
Maint
HF/C
35295006
2
11/19/92
Load drop
44,101
1 67
EHC
#3 CV piping
Leak
Weld failure
HF/C
92-11-20
RF,
10005615
1
6/17/96
Load drop
22,121
0.79
Electrical
Output breaker
Low pressure
rupture
Improper
installation
HF/C
10005797
TT
1
12/23/96
Load drop
5,427
0 19
SWC
Temperature
CV
Loose adjust
arm screws
Manufacturing
HF/C
10006438
2
12/6/96
Manual
scram
250,560
9 00
EHC
Pressure switch
Leak
Broken bracket
/ severed
tubing
HF/LCA
10006385
Repeat
event 2M19
D
2
12/24/96
Manual
scram
55,680
2.00
Recirc
Scoop tube ball
joint
Broke
Vibration
Induced
HF/LCA
10006441
Repeat
Event
M
1
9/11/95
Scram
369,991
14 01
Main steam
SRV
Opened
Pilot seat
erosion
HF/MS
10004442
1
1/21/92
Load drop
7,030
0.27
FW heating
Logic
High level
sensed
Spurious
HF/MS
92-01-10
1
1/17/94
Load drop
940
0 04
Main turbine
Moisture
separator dump
valve
Valve
positioner
PM frequency
LTA
HF/MS
10001318 SU, TT, PM
deferred
M
2
12/4/92
Manual
scram
157,870
5.98
Recirc
Recirc pump
EOC-RPT logic Breaker tripped
HF/OA
92-12-01
O
2
10/19/94
Scram
77,754
2 95
Electncal
D24 Bus
De-energized
Inadvertent
HF/OA
O
2
2/24/94
Load drop
10,080
0.38
Recirc
Pump
Trip
Fuses pulled
HF/OA
O
1
1/30/95
Load drop
7,000
0 27
Recirc
MG set
Trip
Bumped
HF/OA
2
6/28/95
Load drop
6,300
0 23
FW
FW UPS
Power
disconnect
switch off
Bumped during
cleaning
HF/OA
10004173
RRB
O
1
1/1/92
Load drop
2,904
0 11
Main turbine
Pressure
instrument RV
RV closed
Mispositioned
HF/OA
92-01-02
SU, DF, TT
O
1
7/19/95
Load drop
700
0.03
Recirc
MG set
Perturbations
Operator
HF/OA
O
1
7/29/94
Load drop
570
0.02
FW
Trip lever
Actuated
Inadvertent
HF/OA
O
Data Sorted by root cause
Page 189
Remarks
M
M
F
O
No PM
Occurred
during ST
M
O
LGS POC VT ANGUS
27 May 1998
Appendix 12. LGS unavailability data, sorted by root cause of failure
Issue #
10005615
Remarks
RF 92-1120
Class
DF
M
Failure
Leak
FW heating
Drain valve
Malfunction
Generator
Volts/Hz relay
Actuation
inappropriate
MOD package
LTA
HF/PI
10005652
O
2.70
RPS
Logic
No 1/2 scram
alarm
Indeterminate
HF/PI
10005675
O
56,700
2 15
EHC
#6 ISV
Perturbation in
ETS/RETS
Air entrap in
control pack
HF/PI
93-03-38
O
Load drop
39,426
1.49
H2
Recombiner
Recorder
Logic
MOD PMT LTA
HF/PI
10004403
O
10/12/96
Load drop
37,463
1 35
Condensate
Pump motor
Vibration
FME
HF/PI
10006201
2
4/29/95
Load drop
10,000
0 36
Recirc
Coupler bypass
valve
Incorrect
position
LTA
procedures
HF/PI
10003924
Eff Out Days
1.40
System
EHC
Load drop
1,468
0.05
Scram
89,099
3 20
5/21/96
Scram
75,095
2
3/26/93
Scram
1
9/2/95
2
Date
4/29/96
Type
Load drop
1
8/2/96
2
5/14/96
1
Cause
Weld repair
failure
Category
HF/PCM
Component
#3 CV piping
MWHr
38,939
Unit
2
HF/PCM
M
O
RRB
O
1
7/13/92
Load drop
8,566
0.32
FW
RFP turbine
Insulation fire
Oil soaked
HF/PI
92-07-10
OM
O
2
3/17/93
Load drop
8,030
0 30
SWC
Valve
Valve
mispositioned
LTA procedure
HF/PI
93-03-24
SU, DF
O
2
2/16/94
Load drop
7,150
0.27
Recirc
Recorder
Mislabelled
LTA MOD
review/ PMT
HF/PI
10001491
RRB
O
2
9/2/95
Load drop
7,470
0.27
H2
Recombiner
Recorder
Logic
MOD PMT LTA
HF/PI
10004403
O
2
6/6/96
Load drop
378
0 01
Isophase bus
cooling
135 74
Fan
Trip
LTA WO/
procedures
HF/PI
10005731
O
Total effective
outage days'
Categ
ory
Code
EFEquipment
Fators
Data Sorted by root cause
WP - Worn
Parts
WD - Weak
Design
Page 190
FP - Fatigued
Parts
LGS POC VT ANGUS
Appendix 12. LGS unavailability data, sorted by root cause of failure
Unit
Date
Type
MWHr
Rema
rk
Code
System
Component
ICS Inadequate
Control System
Failure
Cause
Category
EOL - End of
Life
Issue #
Remarks
Class
PI - Procedural
Inadequacy
CCraftsmanship
OA - Operator
Actions
MSManagement
Standards
LCA - Less than
adeq. corr
actions
PCM - Poor
Corrective
Maintenance
A- Age
F - Fabrication
D - Design
IInstallation
MMaintenance
O - Operation
DF Dependent
Failure
TT - Turbine
Trip
OM - On line
Maintenance
RF - Repeat
Failure
RRB Recirc
Runback
RPT - Recirc
pump trip
DF - Dependent
Failure
SU - During
Start up
HFHuman
Factors
Class
Code
Eff Out Days
27 May 1998
Data Sorted by root cause
Page 191
LGS POC VT ANGUS
Appendix 13. PBAPS Unavailability Data, sorted by root cause of failure
Type
Automatic
Scram
Manual
Scram
Automatic
Scram
load drop
MWHr
248585
Eff Out
Days
89
137576
49
2171
893
165366
6/3/95
load drop
load drop
Automatic
Scram
Manual
Shutdown
Automatic
Scram
load drop
3
2
7/26/92
1/5/94
load drop
load drop
9791
5921
3
3/7/93
load drop
3
6/13/97
3
Unit
2
Date
7/17/92
3
6/23/96
2
10/6/96
3
10/1/92
3
2
2
9/30/92
1/18/93
5/14/94
3
2/2/96
3
8/1/95
2
System
Electrical
Component
3435 breaker
July 1992 - June 1997
Failure
tripped
27 May 1998
Description
rwcu controls
Cause
lighting strike
Category
EF
#2 turbine control w stem
seperated
scram, gen lock out stator
current unbalance
recirp pump control
clearance inadequate
Issue #
Remarks
N
Class
A
EF
SU
F
poor solder joint
EF
SU
F
unknown
EF
N
D
recirc pump
2b rfp vibration inspection
recirc pump a speed increase
unknown
unknown
defective
EF
EF
EF/EOL
N
A
SU
SU
M
A
gasket leak
main generator hydrogen leak
worn out
EF/EOL
N
A
upscale
scram-feedwater transient
failed transmitter on card
EF/EOL
N/TT
A
failure of PS
power ascension &c rfp
problems
age
EF/EOL
N
A
0 4 Reactor detector
0.2 FW Heating solenoid valve
failed
coil failed
tip machine a
age
5a heater extraciton w solenoid end of life
EF/EOL
EF/EOL
SU
N
A
A
5666
0.2
mtsv
electrical problems mtsv replacement
failure
EF/EOL
N
A
load drop
4238
02
Main
Turbine
FW
3b rfp speed
controller &hjsv
3b fpr speed control problem
age
EF/EOL
N
2/5/96
load drop
953
00
RPS
sol. coil burned
and controller
degradded
defetive logic
switch
5a k5c relay dropped out
failed
EF/EOL
2
12/25/96
load drop
244
0.0 FW Heating level controller
4c fw htr level oscillations
electrical comp end of life
EF/EOL
N
A
2
10/26/95 load drop
dump valve failed
to open
tube leaks
b fw string isolated 95% pwr
unknown
EF/FP
N
A
3
2
5/9/93
load drop
10/22/95 load drop
lost signal
#2 tip machine
cable
EF/FP
N
A
cable fault
EF/FP
N
A
2
11/8/95
load drop
329
EF/FP
N
D
2
2
3/20/93
1/1/93
5992
418199
0.2
15 0
Recirc
EHC
EF/ICS
EFNVD
SU
N
D
D
3
7/4/93
352800
12.7
Reactor
leak
power reduction for fuel repair
pci
EFD
N
D
3
7/14/92
load drop
Manual
Shutdown
Manual
Scram
Manual
Scram
positioner air
supply
dead band
did not work
220-34 line tripped = positive
failed
reactivity
3c fw htr drain w broken air
improper support
line high level
recirc pump hi oil level
turbine control valve oscillations design
225039
81
Offgas
linkage allen set
slippage
off gas system
design of air line
EFVD
SU
D
Data Sorted by root cause
82012
78134
Main
valve
stem binding
Turbine
2 9 Generator negative sequence short/open
relay
2.8
Recirc
control loop
none mg set lock
up
0.1
Recirc
3b recirc pump
bkr 2ak34b trip
0.0
FW
2bs018 2b rfp
rotor imbalance
5.9
Recirc
a pump
Ivdt
61863
22
29188
1.0
FW
21995
0.8
FW
Generator bushing
3a rfp speed
controller
power supply fw
control system
pish-3-02-3-055c
24869
699
622
0.9 FW Heating B5 fw heater
R tlimit
0.0 Reactor detector
340
0.0 Transmissio 220-34 ug line
n
0.0 FNVHeating positioner
level switch
pressure
transmitter
fuel
3239a linkage
failed
Page 192
A
PBAPS POC FL JORDAN
Appendix 13. PBAPS Unavailability Data, sorted by root cause of failure
Eff Out
Days
3.6
Unit
3
Date
Type
10/11/94 Automatic
Scram
MWHr
99423
2
94620
3 4 Generator
93042
3.3
Offgas
2
10/15/96 Automatic
Scram
3/23/95 Manual
Scram
12/21/92 load drop
69152
25
EHC
3
8/26/95
load drop
65874
2.4
Main
Turbine
2
3
9/22/93
8/8/94
load drop
load drop
35885
35702
1.3
1.3
Reactor
Offgas
3
4/21/97
load drop
35167
13
FW
2
3
9/10/94
4/27/94
load drop
load drop
14083
12881
0.5
05
Reactor
FW
2
12/17/92
load drop
10398
0.4
EHC
2
9/21/93
load drop
7552
0.3
Reactor
pressure
transimtiter
fuel
2
2/23/94
load drop
4577
02
Reactor
3
2
6/12/94
5/9/96
load drop
load drop
4054
2148
01
01
EHC
EHC
3
8/6/96
load drop
577
0.0
Offgas
3
2
1/23/93
12/1/95
3
2
2
load drop
Manual
Shutdown
4/24/93 Manual
Shutdown
7/27/92 Manual
Shutdown
System
Electrical
July 1992 - June 1997
Component
Failure
main power
shorted winding
transformer for
inverter
negative sequence short/open
relay
ao3466b
failed close
Description
Cause
scram loss of static inverter y50 break down of insulation
Category
EF/WD
scram, gen lock out stator
unbalance
sjae supply block valve failed
poor solder joint
turbine control vie oscillations
Issue #
Remarks
N/RRB
Class
D
EFNVD
SU
M
plug design
EF/WD
N
D
design
EF/WD
SU
D
turbine cv limit switch bad
testing logic
clad
Flux tilt testing
steam leak at flex recombiner leak
troubleshooting
power supply
feedwater computer trouble
fall/transfer of
relay actuation
control
environmental conditions
EF/WD
N
D
pci
flange flex
EF/WD
EF/WD
N
SU
D
D
N/RRB
D
clad
control valve
actuator binding
did not work
pci
not smooth operation
EF/WD
EF/WD
N
N
D
D
turbine control valve oscillations design
EFVD
SU
D
clad
admin precaution increase in
off gas level
pc
EF/WD
SU
D
fuel
clad
rod pattern adj due to 5 leakers pci
EF/WD
N
D
muffin fan
wire lug
stopped working
loose wire
unknown
vibration
EF/WD
EF/WD
SU
N
D
D
EF/WD
N
D
pci
errosion
EF/WD
EF/WP
N
N
D
A
N
A
pressure
transimtiter
limit switch
fuel
fe5020
computer dcc-x
fuel
rfp
did not work
switch failed
Flux tilt testing
rfp control problem
power supply? transfer - design EF/WD
issue
failed open
543
219052
control valve
9716b
0.0
Reactor fuel
7.9 FW Heating 5b fw htr
ehc elec cabinet cool fan
turbine control #2 valve
oscillation
recombiner isolation
leak
leak
flux tilt
5b fw htr repairs
128564
4.6
leak
rx instrument mismatch &reclrc loss of level
EF/WP
pp
RECIRC PUMP TRIP AND
VESSEL TEMPERATURE
DIFFERENTIAL
Equipment Failure
EF/WP
3b recirc pump trip c phase
cable fault
cable treeing
EF/WP
N/RPT
A
no mtce
EF/WP
SU
M
loss of cooling
EF/WP
N
A
Recirc
69540
2.5 Recirc
It73a equilizing
valve
PUMPXX
3
4/9/97
load drop
27554
10
Recirc
low side
transformer cable
2
3/18/93
load drop
22741
08
Recirc
2
4/23/93
load drop
21119
0.81
Recirc
loss of tach signal loss of contact for recirc pump b(gen hi
brushes
amps,volts
vent damper
vibrated shut
recirc mg set
Data Sorted by root cause
27 May 1998
cable insulation
breakdown
Page 193
27792013
A
PBAPS POC FL JORDAN
Appendix 13. PBAPS Unavailability Data, sorted by root cause of failure
Unit
3
Eff Out
Days
System
0.5 Condenser
July 1992 - June 1997
MWHr
139080
2
Date
Type
12/19/92 Manual
Shutdosn
1/22/93 load drop
3
2
2
2
3
5/1/94
12/16/92
1/29/93
10/5/96
11/28/93
loaddrop
load drop
load drop
load drop
load drop
7634
6549
4636
4263
3651
0.3
FW
02
Recirc
0.2
Recirc
0 2
FW
0.1 Control Rod
Drive
2
4/28/94
load drop
2537
2
2/10/94
load drop
2116
0.1 Control Rod rod 26-15
N2 leak at
Drive
charging block
0.1 FW Heating 5a fw htr extration gasket failure
stm valve
2
1/12/95
load drop
1795
0 1 Main Steam msiv
2
2/3/96
load drop
492
11/17/95 load drop
143
10171
0.4
FW
Component
HTEXCH
Failure
2cs018 2c rfp
control valve
rfp
oscillation
tach
2bs018 b rfp
34-31 accumulator
o-ring
control system
controller
coupling tnp
probe failure
failed
Description
CLEAN CONDENSER
WATERBOXES.
rfp c slow responce
27 May 1998
Cause
Category
EF/WP
Issue #
Remarks
Class
A
failed
EF/WP
N
A
a rfp maintenance
recirc pp controls
recirc pump a trip
b rfp high vibration
o ring failure, rod 3431
accumulator
not smooth operation
worn parts
failed
worn parts
worn
EF/WP
EF/WP
EF/WP
EFWP
EF/WP
N
SU
N
SU
N
M
A
A
A
M
inop control rod
O-ring replaced
EF/WP
N
A
5a fw htr repair from stm leak
oos
bearing cap gasket failure
EF/WP
N
A
packing leak
ao-86a repair
small leak
EF/WP
N
A
0.0 Control Rod hv-1 11
Drive
0 0 FW Heating cv-3043c
broken o-ring
hcu hv-111 broken
O-ring damaged/cut
EF/WP
N
A
steam seal
3c fw heater drain closed
wear
EF/WP
N
A
150887
5.4
a recirc motor low oil level
unknown under investigation
HF
SU
M
46288
1.7
turbine bearing 12 high temp
electroysis
HF/C
SU
M
2
2
Manual
Shutdown
11/11/95 load drop
12/13/92 load drop
loss of oil in
upper/lower mtr
brg reservoir
hi temp
30423
27598
1.1
1.0
FW
Reactor
2ap001
lprm 56-41 &5643 cross
connected
n/a
mtce
a rfp vibration 95% pwr limit
Iprm mismatch
foreign material
hooked up wrong
HF/C
HF/C
SU
SU
M
3
6/22/96
load drop
20785
0.7
EHC
servo
leak
o-ring failure
HF/C
N
M
2
4/1/97
load drop
17756
06
EHC
cooler
restriction
repair #4 cv ehc leak and msv
leak
ehc fluid leak
suspect FME root cause TBD
HF/C
N
M
3
8/6/94
load drop
11285
04
Recirc
3ag004
brush pigtail
b recirc pump brush
shorted inner and replacement
outer collector
mtce did not stand up leads
post mtce
HF/C
N/RPT
M
2
4/2/97
load drop
1793
01
EHC
cooler
restriction
ehc fluid leak
suspect FME root cause TBD
HF/C
N
M
2
5/19/93
load drop
678
00
Electrical
e322 trip
loss of fw htg
loss of power supply to e22 bus loss of power
HF/C
N
M
2
3/2/93
Automatic
Scram
loss of pwr to
transformer
condensate pump a tripped
HF/OA
N
O
3
3
3/8/97
2
10/9/96
Manual
Scram
Data Sorted by root cause
398510
Recirc
3ap034-dr
Generator bearing
14 3 Condensate 2c cond pumps
Page 194
operator opened bkr
PBAPS POC FL JORDAN
Appendix 13. PBAPS Unavailability Data, sorted by root cause of failure
Unit
3
Eff Out
Days
System
Component
13.0 Containment VALVEX
Date
Type
10/15/92 Automatic
Scram
MWHr
361608
3
12/2/95
Automatic
Scram
114994
4.1
DC
cracked terminal
strip
3
7/30/93
47859
1.7
Offgas
stm flow sensor
2
3/4/96
Manual
Scram
load drop
3488
0.1
FW
3
11/6/95
load drop
3336
01
Electrical
2
6/24/93
load drop
2120
01
HPCI
2
3/17/95
load drop
832
00
FW
2
8/16/95
load drop
283
2
6/4/96
load drop
114
3
7/4/92
3
3/7/93
167298
3
2/3/94
102297
3.7
3
5/21/93
Automatic
Scram
Automatic
Scram
Manual
Scram
load drop
0.0 Transmissio
n
0 0 Transmissio
n
9 0 Transmissio
n
6.0
FW
18117
0.7
FW
2
2
9/5/92
1/8/94
load drop
load drop
6974
5813
03
02
Recirc
FW
2
6/24/94
load drop
4904
0.2 Circ Water condenser
3
3
9/14/93
8/22/92
load drop
load drop
2553
2385
3
3
11/19/93 load drop
12/1/93 Manual
Shutdown
8/17/92 Automatic
Scram
12/18/92 Manual
Shutdown
7/23/92 load drop
746
392253
2
2
3
Data Sorted by root cause
251351
July 1992 - June 1997
Failure
2bs018 b rfp
2nd ground by
person working on
equipment
blown fuse
mo99/91
vibration
fuse
loose
check valve
broken air line
msc control switch misposition
220-8 line
220-8 line
3su feed lost
c rfp
Generator breaker
Description
PCIS GROUP I ISOLATION
CAUSED BY BUMPING
INSTRUMENTATION.
27 May 1998
Cause
Equipment Failure
Category
HF/OA
Issue #
27892008
Remarks
Class
O
turbine trip - pos & neg ground 2nd ground
HF/OA
N
O
manual scram due to
recombiner
b rfp tripped vibration probe
bumped
13 kv electrical system-loose
fuse
repair hpci injection check valve
operator action
HF/OA
N
O
probe bumped
HF/OA
N
0
droped holder damage
HF/OA
N
O
scaffold
HF/OA
N
O
feedwater transient
operator error
HF/OA
N
O
digging into line Unit 1 pl
HF/OA
N
O
operator at sub opened
incorrectly
173 mtce
HF/OA
N
O
HF/PCM
N
M
poor lubrication to vib sensor
HF/PCM
N
M
mtce
HF/PCM
N
M
mtce left out parts
HF/PCM
N
M
gain setting
improper lubrication
HF/PCM
HF/PCM
N
SU
M
M
HF/PCM
N
M
HF/PCM
HF/PCM
N/RRB
N
M
M
popped open su 25 220-8 line fault
bkr
breaker opened
220-8 line de energized
e313 cs & 343 su north substation xfmr
tran
hi vibration
maintenance outage
ground resistor left
in place
parts missing post
mtce
blind controller
speed controller
siezed
cleanliness
field ground resistor-main
generator
rework mgu hyd jack solenoid
sv7
recirc pump b
a rfp maintenance
debris in
leak
reactor feed pump trip
feed water heater
chlorine oos and warm water
and Inst at cal limit
fm
out of calibration
misaligned
bent shaft
recirc pump vibration alarm
Ipci mov 25a Inop
wrong nut
HF/PCM
HF/PI
SU
N
M
M
generator lock out
written com to load dispacter
HF/PI
N
O
34334
2.2 Transmissio lock out due to no sub sta 205
n
bkr
1 2 Generator stator
h2 leaks
generator h2 leaks
sealent groove seal improper
HF/PI
SU
M
25191
0.9
recirc pump
margin
HF/PI
SU
M
61988
fp turbine control
b recirc pump
2as018
0 1
FW
flow controller
0.1 FW Heating drain valve
positioner
0.0
Recirc
coupling
14.1
LPCI
mo-3-10-025 rhr
Recirc
st-r-60a-2
calc error
low condenser vacuum
Page 195
PBAPS POC FL JORDAN
Appendix 13. PBAPS Unavailability Data, sorted by root cause of failure
Eff Out
Days
System
Component
0.5 Circ Water screens immobile
cal drifted low
4441
Generator generator core
monitor
0.2 Generator alarm setpoint
Description
b screen immobile, b cw pp
removed from service
generator core monitor alarm
dnfted low
generator core monitor alarm
4/2/97 load drop
12/29/93 load drop
4271
2266
0.2
0.1
FW
Electrical
2as018 a rfp
3-2a-k004a
failed to trp
reactor feed pump trouble
deenergized when recirc runback a pump
480v load center
30801 was being
restored from the 3
4G4 tie breaker
load drop
load drop
1946
835
01
0.0
Recirc
Electrical
3-2a-k010a
e22 bus
loose connection
loss of pwr to
panel y-34
Unit
3
Date
1/20/95
Type
load drop
MWHr
13194
2
3/27/96
load drop
4763
3
3/27/96
load drop
2
3
3
2
9/19/95
6/10/95
Data Sorted by root cause
02
27 May 1998
July 1992 - June 1997
Failure
pin shear
Remarks
N
Class
O
HF/PI
N
M
cal
HF/PI
N
M
debns in trip dump valve
30b01 Ic being restored
HF/PI
HF/PI
N
N/RRB
M
O
HF/PI
HF/PI
N/RPT
N
M
O
Cause
wrong pin
Category
HF/PI
low cal
pm task inadequate
a recirc mg set tripped
loss of power supply to e22 bus Diesel feedback signal during
& y-34
mod testing
Page 196
Issue #
PBAPS POC FL JORDAN