ML20244D704
| ML20244D704 | |
| Person / Time | |
|---|---|
| Issue date: | 01/31/1986 |
| From: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| TASK-AE, TASK-P601 AEOD-P601, NUDOCS 8603040083 | |
| Download: ML20244D704 (18) | |
Text
Tp ' , c.
g 1
' AE0D/P601
.g.
Trends and Patterns ,.
I Program Plan FY86 - FY88 j 1
Prepared by:
Office for Analysis and Evaluation of Operational Data January 1986 g i
1
)
s d
i y i
'qkfbhNYh0 Yb~
(_ _
t., t,,
Table of Contents I. Introduction ..................................... 1 II. P rog ram Obj ec ti ves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 III. Program Elements ................................. 2 IV. Resources and Milestones ......................... 9 Appendices ............................................ 1?
l l
l l
l l
1 j
l l
l l
l l
i l
t i
I. INTRODUCTION This Plan describes the AE0D program for the periodic analysis of sets of operational event data reported by commercial power reactor licensees. The sets of data covered by the Plan are extracted from Licensee Event Reports (LERs) and the Nuclear Plant Reliability Data System (NPRDS).
The phrase " trends and patterns" is used in this plan to describe a program for analyzing incidents of low individual significance for which frequency is the element which lends significance.
In this Plan the words "tred " and " pattern" are used as follows:
Pattern: The observed distribution of similar occurrences (i.e., incidents),
among a set of given classifications (e.g., plant, component type, root cause).
Trend: A pattern which includes time as a characteristic (e.g., a sustained decrease or increase in occurrence rate as a function of time).
Each incident has certain classifications associated with it. For example, an incident can be described as:
A failure of a Target Rock Relief Valve to remain closed at Brunswick I on April 19, 1983 due tn inadeauate maintenance.
Any reasonable subset of the underlined items (i.e., classifications) can be used to specify what is meant by similar incidents. How this type of incident is distributed across the additional classifications (e.g.,
failures of Target Rock relief valves to remain closed distributed across plants) can be examined.
The program described in this Plan covers FY86 through FY88, taking advantage of the experience gained in executing the first two years of an earlier plan
[ Trends and Patterns Program Plan for the period FY84-FY86 (AE0D/P402) issued .
in March 1984]. I Section II of this Plan lists the program objectives. Section III describes the Program Elements in detail.Section IV summarizes the resources and detailed milestones for planned Program Elements.
II. PROGPAM OBJECTIVES The AE0D Trends and Patterns Program complements the ongoing engineering reviews of operating experience within AE0D and other NRC offices. Unlike an engineering review which usually begins with the formulation of a specific concern, trend and pattern analyses usually assess operational data with limited prior formulation of a concern. Rather, the results are driven by the data and the imbalances, nonuniformities, and changes in frequency of occurrence.
I
.i.
l -.
The Program uses the above approach to meet the following specific ob.iectives:
to provide a perspective nn the performance of the nuclear industry, including the performance of individual plants, through a systematic and broad assessment of reportable events.
. to identify and investigate potential safety concerns associated with events and failures that are of low individual safety significance.
to investigate operational experience in terms of the assumptions and bases used in regulatory analyses, including probabilistic risk assess-ments.
The need and usefulness of such data is becoming increasingly evident. The frequency of operational events and failures, and the changes in tnis frequency are a matter of concern and interest in a number of regulatory activities. Thus, reports from this program will serve to provide input to these activities, such as SALP assessments and analysis of individual plant performance, as well as serving to identify the need for further technical studies and helping to guide the allocation of staff resources.
III. PROGRAM ELEMENTS In order to accomplish the ob.iectives of the Trends and Patterns Program, three separate elements or initiatives have been pursued: (a) analysis of event data, i.e., LER data, for selected operational events; (b) analysis of component data, i.e. , NPRDS, to identify operational experience with selected components; and (c) analysis of failure data from LERs in order to search for unrecognized system and component problems. These three program elements are discussed individually below.
A. Analysis of Event Data The core of the Trends and Patterns Program analysis of LER data consists of a series of four in-depth periodic reports, each of the four treating a different category of operational events covered by the requirements of 10 CFR 50.73. The categories are: (1) Reactor Trips; (2) Engineered Safety Feature (ESF) Actuations; (3) System Unavailability; (4) Technical Specifica-tion Violations and Shutdowns. A fifth report will provide an integrated review of performance for recently licensed plants.
Each edition of these reports will cover events from the most recent calendar year and will compare the latest results with those from earlier years.
Further, each report will contain findings, conclusions, and recommendations for correcting problems found to be the most significant. The schedules for these reports will be established such that preliminary findings for the preceding year will be available for discussion in the AE0D Annual Report published in April of each year.
, - r' '< >3 -
, ge
,n , : .
The, reports 'to be' prepared are discussed in detail below..
- 1. Reactor Trips Paragraph 50.73(a)(2)(i.v) of the LER rule requires reporting'of:
"Anyevent'or.conditionthatresultedinmanuafobautomaticactuation of any Engineered Safety Feature (ESF), including'ihe Reactor Protection- J System (RPS). However, actuation of an ESF, including *the RPS, that- D, .
'resulted from and was part of a preplanned sequence during testing or?t i J reactor operation need not be reported." '
]
1 The Reactor Trip report covers RPS actuatioin hnfbh resulted in control rod motion.-
f The analysis of 1984 data was published as AE0D/P504 in August 1985. This I report serves as the prototype for this topic! The ' table of contents for l the report is provided as Appendix A. The anal,fsis of?1985' data is underway. !
Preliminary findings and conclusions will be avellable by March 1986 for j incorporation in the April 1986 AE0D Annual Report covering 1985. The' final f report will be published in June 1986. Data for 1986 and 1987 will be handled ;
similarly. No program support funds are involved in ,this report since analysis I of the data is done entirely by AE0D staff. Milestones are listed in Section'IV. {
' l
- 2. ESF Actuations <j 'i Paragraph 50.73(a)(2)(iv) requires reporting all Engineered Safety Feature (ESF)actuations. All non-RPS actuations are covered by the ESF report, i
The analysis of the' first six months of 1984 was published as A200/P503 in August 1985. This report serves as the prototype for this topic.. The table of contents for this report is provided as Appendix B. The analysis of the e second half of 1984 is underway. PreU minary findings and conclusions will j be available'by January 1986. Inparallkl,,acontractorwillbegin(January l 1986) data analysis for 1985 data. Some0 preliminary results for 1985 data t j should be available for evaluation andrfor inclusion in the AE0D Annual. ,
Report to'be published in April 1986 Dats foy 1985, 1986 and 1987,wi M be analyzed with contract support; however", ~AE0D staft"w111 be responsible,for evaluating the data and formulating conclusions and reenwendations. '
Milestones are listed in Section IV. t
/
v
- 3. System Unavailability Paragraphs 50.73(a)(2)(v) and (vi) require reporting of: j
"(v) Any event or condition that alone could have prevented the fulfill- l
-ment of the safety function of structures or systems that are needed to:
)
(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; i
i
_ i.______ _ f a
,.; [ ,
-4 (C) Control the release of radioactive material; or
.(D) Mitigate the consequences of an accident.
L (vi) Events covered in paragraph (a)(2)(v) of this section may include one.or more personnel errors, equipment failures, and/or discovery f
of design, analysis, fabrication, construction, and/or procedural T [o ' ' f inadequacies. However, individual component failures need not be reported pursuant to this paragraph if redundant equipment in i the same system was operable and available to perform the required safety function."
i The idtent of these paragraphs is to capture those events where there would h6e been a failure of a system to properly complete a safety function, i regardless of when the failures were discovered or whether the system was needed at the time. In addition, failures of non-safety systems are frequently described in LERs for events that were reportable for other reasons (e.g.,
? loss of main feedoter, which is not reportable, causes a reactor scram, which is reportable).
AE0D has recently cobpleted an analysis of the loss of safety system function. This study (published as AE00/C504, dated December 1985) focused on the 1981-83 time period. The work under this plan continues this effort using data reported by the LER rule, and the analysis of 1984 data is underway. Data analysis is being performed with contract support, as will be the case in subsequent annual analyses.
~
Aspectsofth$eventthatwillbeconsideredintheanalysiswillinclude:
I
- 1) Facility
- 2) Plant condition before event
- 3) Plant power level before event
- 4) System that was unavailable
- 5) Type of unavailability
- 6) Severity of unavailability 7a Mepod used to detect unavailability 8 System failure mode 9 Plant reaction to failure
- 10) Date and time of the beginning, detection, and restoration, of the unavailability
- 11) Rootcause(s)
AE0D staff are responsible for evaluating the data and formulating conclusions and recommendations.
Prelimincy results for 1984 data hill be available by January 1986.
Data analyt.is for 1985 has begun, and some statistics should be available for the April 1986 edition of the AEOD Annual Report. 1986 and 1987 data will be analyzed in accordance with the milestones shown in Sectinn IV.
M 3
$ R $ :.{q (* 3 K , .' 4
'y , ,
'[ .
%g
[t u ig g '[L .
- 4; Technical 1 Specification?Tiolations and Shutdowns Paragraph 50.73(a')(2) i))rNguires . reporting of: -
/
AK 70 "(A) The completion of any n6c1 ear plantishutdewn Wquired by the ew g ,
'W , 4 plant's Technical SpecifU ,tions; or 3
t m_'~
i ff g. [(d)'AnyMeration.prohibithd)bythe. plant'sTechnical$pdifications; a - t a m )~ 4 V ,
1 ,) op 3' gQ y i
Tji 1
- , 1 rr ; (C) 'g9ny deviatios rom M.he plant's Technical Specificatiorts authorized r ant tc ib .f ( ) ofilthis part." '!-
d / 9 wh This paragraph req 0 ires events to be reported where the licensee is requ(red ,
W yMc $& y y toshutdowntheplantbecaushtherequirementsoftheTech11calfpecirtia.
,[ 'tions were For notthemet.
purpose of this paragraph, "sh'utdown". is. defined l
i W as the point in' time where +.he Technical Specifications require that thes ' '
n ,' plantf be in the' firs t3hutdown: condition required by a Limit!np sCendit for.Y for. Operation [e.g., h c sta M 9 (Mode 3) for the PWRs with ine Standard
,i Technica1LSpecificatiou(s]. If the condition is corrected beiWe the' tihe,'L Lf ?, L' ,
limit for>being shb0down (i.e., before completion of the shutd$wn), the9p$nt
!: ednotbegr>orted.% j. 7 s \ lrdd$ditio%L1f :h cMion thab was prohibited.hy thATechr/fOkh.Sphcificatiods i
' exi%ted (i.'c.', The plant was .in/a d,e. graded mode ello@d by thu sTschnical- s m Specifications) for a period of . time longer than t9at pbrmittui hy_ the 9 Technical Specifications, it must be reported even'if thonditicipwas not.
Y discovered until after the cilowable time had elapsed ar6the condition was rectifiedimmediatelyafterldiscovtry, s
4 . : .
The events rqwed per 50.73(a)(2)(i) will be treated fri>thd Technical
~
y Specifications' Violations and yutdowns trends and patterns report. .i
\
This studp. has the followin{ objectives:
.AToidentifyandcatalog[technicalspecification-relatedLERs.
Y
- h. s
. To categor.ize and evaluate the actions rerorted in the LERs emphasizing:
a) the. impact on plant operations, (b) thf frequency of occurrence, c) the' system (s)tinvolv4d.s(d) the specific 1tschnical specification q '
. recyirement, and M the cause of the original problem which led to the pappyicationofthespecifictechnicalspecificationrequirement. j
\ ,,h' 3 a ,
1 a , Tu idedtify my,ie( ues mising from the evaluation which appear to have y n fe '
s
,' 'i ,sa,q ,ty significony,(e. .
,t t
, .W84 dato y this spit 0was analyzedgs ;! art of the Technical Specification i 7 Irytavement' Program S SIP) conducte( hy h'<R.' The AE00 angtysis for~1985 data i
- q. will maintain ecntinu:ty with that effort. InLparticulan,'the 1985 data j base will be consistent with the 1984 data bate in structure. In addition, I the scope of the trends and patterns; report will. be expanded to cover Gray I e Boak data on power reductions.(whictjare not reported under 50.73) in order i
to tontinue the TSIP review of,the impact of technical specifications on
' plant operations.
A q ,
s
- s
-$ % v , i 1 l: .
je 4 h a 1 1 LJ- _li
e 1'
i., ,
The 1985 analysis is underway and will be conducted with contractor support, as will analysis for 1986 and 1987 data. Preliminary results for 1985 should be available for the April 1986 AE0D Annual Report. Milestones are provided in Section IV.
- 5. Analysis of Operational Experience From New Plants In the NRC Policy and Planning Guidance for 1985 is a requirement to "... continue to closely monitor the first two years of operation of new plants coming on line, particularly those licensees who have no prior experience with nuclear plants." To help meet this commitment, AE0D has undertaken a new study which has the purpose of trying to assess the operating performance of recently licensed commercial U.S.
nuclear reactor units.
The objectives of this study include an attempt to identify and catalog 1 reportable events that occur during initial unit operation. Emphasis is being placed on determining operational trends which may be used '.n the analysis of the performance of a nuclear unit or utility. Compari- j sons of performance between units will also be conducted to determine ;
if safety significant differences can be recognized. Also, if indicated 'l by the performance trends, recommendations for further studies, safety l reviews, or corrective actions will be formulated. i All units which received an initial license for power operation on or after January 1, 1983 will be included in the first issue of this study. This will permit the establishment of a two year baseline of performance data for recently licensed units. As of December 1985, the study included data from 17 units. !
l For the initial study, the data sources will be restricted to computerized j reactor event data bases such as 50.72, SCSS (LEPs), Gray Book and NPRDS.
Such data sources are fairly objective and readily searchable. If l additional data appears necessary to support or help determine performance i trends, more subjective data sources (such as inspection r9erts, A0 reports and SALP reports) may be considered.
In large measure, the initial effort is concerned with identifying valid, accessible performance measures. Hence, tha work is exploratory at this time, and no long term schedule has been established.
B. Analysis of Component Data !
The NPRD System is an industry-wide system for tracking the performance of selected systems and components at nuclear power plants. Since almost all U.S. plants in commercial operation supply detailed design data, operating characteristics and performance data, NPRDS provides an extensive data source for analysis of operating experience.
AE00 has worked closely with the Institute of Nuclear Power Operations (INP0),
which manages NPRDS, to monitor the status of NPRDS in order to assure that 4 steady progress is made toward meeting NRC needs for reliability data. Also AE00 has begun the development of a systematic program of analysis of NPRDS data as part of its Trends and Patterns Analysis Program.
{ f
- 4 Prior to beginning actual data analysis AE0D worked with INP0 and other NRC offices to develop a list of critical components on which to focus attention. From the approximately'4000 components in the data base, 300
" key components" have been selected.and flagged in the data base with their " application code" (e.g., MSIV, AFW pump). ;
The'AE0D trends and patterns analysis of NPRDS data will concentrate on these key components for the following reasons:
(1) The. key components, by definition, are those components which j
have the greatest impact on safety system availability and the occurrence of plant transients. Therefore, these components are :
of regulatory. interest, and their importance is. sufficient to l warrant regulatory action if problems are detected.
(2) The key components typically include the components of greatest 3 interest to the licensee. Therefore, the data for key components should be of higher quality and conclusions drawn from statistical analysis will be less diluted by concerns about the cuality of j the data.
l (3) Analysis of component. failure statistics from.high aggregations (e.g., all valves, all pumps) would be plagued by problems with data cuality and problems associated with a lack of homogeneity in the population.
The AE00 trends and patterns analysis of NPRDS data will combine statistical and engineering evaluation in a series of reports on key components in the following fashion:
(1) Statistical analysis will be performed with contractor assistance:
(a) Key components will be ranked using "importance" data from various PRA activities. The key components will then be analyzed in decreasing order of importance.
(b) Target or alarm failure rates will be determined whenever i possible to provide a quantitative basis for judging performance.
(c) Statistical analysis will be performed using methods based on times to failure and failure counts. These analyses will include:
(i) Trend analysis of the failure rate (ii) Detection of a shift in failure rate (iii) Identification of factors (pedigree information) which appear to influence the failure pattern (2) The statistical analysis package for each key component will be used by AE0D engineers as a starting point for developing findings, conclusions and recommendations for the selected key component.
i 1
. ~
Methods development for this analysis program has been underway since October 1984. INP0 is expected to release the final key component labeling for general use in retrieving data in January 1986. Data analysis will begin in January 1986; and four key components will be analyzed and four reports will be prepared (with contract support) by April 1986. Since this is a prototype effort, release dates for final reports are not yet firmly established. Our goal is to issue between 10 and 20 key component reports a year. Milestones are listed in Section IV.
C. Analysis of LER Failure Data i One major activity under the program plan published in March 1984 was the-
- automated analysis of LER data using the data attributes (e.g., cause, effect) stored in the Sequence Coding and Search System (SCSS) format. To that end software was developed and applied as documented in the following reports
NUREG/CR-3824 CONTING Program Guide September 1984 NUREG/CR-4071 Exploratory Trend And February 1985 Pattern Analysis Of 1981 Licensee Event Report Data NUREG/CR-4129 Exploratory Trend And January 1985 Pattern Analysis Of (Draft) 1981 Through 1983 Licensee Event Report Data In performing these analyses, the following conclusions were reached:
. When data was sorted by plant and one or more additional variables (e.g., component) the data tables were very sparse (i.e., contained many very low counts and a lot of zeroes). Data were so sparse when looking at one year of data (i.e., NUREG/CR-4071) that the next analysis was expanded to include three years of data. Even so, the sparsity remained, and statistical analysis beyond identifying two or three high count cells in a table was not productive.
. The investigation of high count cells revealed no previously unrecognized substantive safety issues, however, it served to check and verify other screening activities (i.e., event screening as conducted by AE00 and IE).
. Extensive analysis of high count cells was required to separate potential safety issues from other factors such as variations in reporting require-ments, plant designs and reporting philosophies among plants which can result in high counts. Outliers could not be assumed to represent safety issues without considerable additional analysis.
. Reviewers outside AE00 indicated that the age of the data analyzed (the report on data from 1981-1983 was issued in draft in January 1985) diminished the value of the analysis.
q;- 7 -9 -
q
. i
-.. 'In' January'.1984, the LER rule (10 CFR 50.73) became effective. This l
. rule reduces the~ number of-reportable events which exacerbates the.
l sparsity problem for LER data. .However, approval of 50.73 included a decision.to rely on the NPRD System for component-level failure data.-
Thus, further analysis of LERs for predominantly component' level data (e.g.,
similar to the analysis described in NUREG/CR 4129) has been assigned a lower J p(riority. - The earliest possible date for the next study of this type is FY88 .
i.e., CY 1984-1987 data). 1
.j
'IV. RESOURCES AND MILESTONES j The proaram support funds and AE00 staff years associated with each activity .
.in Section III are shown in Table 1. Table 2 provides the major milestones for each activity through FY 1988.
I i
1
S 8
8 Y
F mt ar ro gp 0 0 0 0 0 0 2 1 1 5 9 op 2 1 1 2 6 ru PS s
r
- a e)
- YE D T
_ 0fF 3 3 3 3 0 0 2
_ Ef(
Aa-
_ 1 1 1 1 2 1 8 t
S
_ 7 8
Y F
mt ar ro gp 0 0 5 5 0 0 1 0 0 5 7 op 2 1 1 2 6 ru PS 0
1 s r
a e)
DYE 0 T 3 0 0 0 5 0 8 EfF Af( 1 1 1 1 1 1 6 1 a t
e 6 S l 8 b Y a F T
mt ar ro gp 0 0 0 0 2 2 0 0 0 4 4 op 2 1 1 3 7 ru PS s
d y n t s l a i n a s Y l o t u T i i o o I b t T h V a a T I s l l s T n i o i n C p o a i s i A i i v V y r t a l s s T a n c a t r u U e n n a r t p A a l o c m S l l t A e S P o c t h D D a F s c R w e S y e P e R E S T N N *
,l<
gm ..',' l
., i i
f
[r Table 2. l l: Program Milestones i o
Reactor Trips 1985 Preliminary Findings March 1986 1985-Final Report' June 1986
-1986 Preliminary Findings March 1987 { 1 1986 Final Report ' June 1987 1987 Pre'liminary Findings. March 1988. ;
! 1987. Final Report June 1988.' !
ESF~Actuations
~1984 Preliminary Findings January 1986.
1984 Final Report March 1986 1985 Preliminary Findings March 1986
.1985 Final Report September'1986 1986 Preliminary F.indings March 1987 1986 Final Report .
July 1987' 1987 Preliminary Findings March 1988
'1987 Final / Report July 1988 System Unavailability 1984: Preliminary Findings -January 1985
'1984 Final Report March 1986-1985 Preliminary Findings March 1986 1985 Final Report. August 1986 1986-Preliminary Findings March 1987 1986 Final Report July 1987 1987 Preliminary Findings March 1988 1987 Final Report July 1988 Technical Specification Violations 1985 Preliminary Findings March 1986 1985 Final ~ Report September 1986
- 1986 Preliminary. Findings March 1987 1986 Final Report July 1987 1987 Preliminary Findings March 1988 1987 Final Report July 1988 New Plants Pilot Study February 1986 NPRDS Analysis Four Key Component Statistical Analysis April 1986
[ Schedule for subsequent AEOD reports to be supplied after experience with first fourcomponents)
I m_______.______._
E7' ? . s !
- n.- _
i
' . l
,,. I
., 1 Trends and Patterns Report Unplanned Reactor Trip at U.S. Light Water Reactors in 1984 Page Number ,
EXECUTIVE
SUMMARY
...........................................'l .
l
1.0 INTRODUCTION
.......................................... 5 !.
t 2.0 '0VERVIEW 0F 1984' UNPLANNED REACTOR TRIP STATISTICS ..... 8 2.1 Reactor Trips Above 15% Power . . . . . . . . . . . . . . . . . . . . 11 2.1.1 Initiating Systems......................... 11 ;
2.1.2 Causes .................................... 17 i 2.1.3 Contribution From Maintenance, Testing and. 23 Calibration
'2.1.4 Plant Trip Rates .......................... 28 ,
2.2 Reactor Trips At or . Below 15% Power . . . . . . . . . . . . . . 32 3
2.2.1 In i tia ting Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . ' 34 2.2.2 Causes .................................... 34 2.2.3 Contribution From Maintenance, Testing and. 38 Calibration 2.3 Summary of Findings and Conclusions Based on 1984 . 39 Reactor Trip Statistics 3.0 REACTOR TRIPS WITH ASSOCIATED FAILURES ................ 42 4.0 COMPARISION OF U.S. AND FOREIGN REACTOR TRIP RATES .... 44 1
e 1
1
< -- -- ------------- - - ---~------ - - - -
[,>j' g3 Ll q Page Number j
1 i
APPENDICES ................................................... 48 1
]
-Appendix A_- Unplanned RPS Actuations at U.S. LWRs .... 49 in 1984
] 4 i
Appendix B - - Suninary of 1984 Unplanned ' Reactor Trip ... 62
-Statistics by Plant Appendix C - Human Error Indu'ced Reactor Trips . . . . . . . . . - 65 Above 15% Power -
. Appendix 0 - Component and Piece-Part Failures ........ 75 Above 15% Power t
Appendix E - Quality of LER Reactor Trip Information .. 76 i Appendix F - Reactor Trips With Associated ici!9res ' . . .. - 78 Appendix G - Trips at Foreign Reactors ................ 89 i i1 1
i.
l i
l l
/
]
r;- ---; .;
e..- . -
1 CONTENTS l
J: Page .
LIST OF FIGURES ..................................................,.....
iv.
LIST OF' TABLES ........................................................., v EXECUTIVE
SUMMARY
..................................,...................' 1
1.0 INTRODUCTION
...................................................... 4 2.0 GENERAL'0 OBSERVATIONS .............................................. 5 i 1
3.0 ACTUATION TYPES ................................................... 8
, 3.1 ' Valid Actuations ............................................. 8 ,
!. 3.1.1 Gene ral Cha rac te ri s tics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- !
3.1.2 Speci fi c Ac tua tion Characteri stics . . . . . . . . . . . . . . . . . . . . 8 1 3.2 False Actuat, ions ............................................. 14 l
3.2.1 Ge n era l C ha ra c te ri s t i cs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.2.2 . Speci fic Actua tion Characteristics . . . . . . . . . . . . . . . . . . . . 18 3.3 Conclusions and Recommendations .............................. 20 3.3.1 Conclusions ........................................... 20 3.3.2 Recommendations ....................................... 21 4.0 EMERGENCY CORE COOLING SYSTEMS - SAFETY INJECTIONS ................ 21 i
4.1 Discussion ................................................... 21 4.2 Conclusions and Recommenda tions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 4.2.1 Conclusions ........................................... 22 4.2.2 Recommendations ....................................... 22 5.0 FAILURES AND PROBLEMS ............................................. 23 5.1 ESF Failures to Actuate Properly ............................. 23 5.1.1 Brunswick 1 - Loss of High Pressure Coolant Injection (LER 84-007) ........................................ 23 5.1.2 Duane Arnold - Loss of Standby Filter Units (LER 84-004) ........................................ 23 5.1.3 LaSalle 2 - Loss of Containment Isolation Valve (LER 84-009) ........................................ 24 5.1.4 Sequoyah 2 - Loss of Turbine-Supplied Auxiliary ,
Fe e dwa te r ( L E R 84 -009 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 1 5.1.5 Yankee Rowe - Loss of Containment Isolation Valve l
} (LER 84-003) ........................................ 26 1
l
,-e, i.
.' =
Page 5.2 Associated. Failures .......................................... 26 5.3. Conclusions and Recomenda tions . . . . . . . . . . . . . . . ' . . . . . . . . . . . . . .
. 27 .
2 i
. s.3.1 Conclusions ..... . 4................................... 27
- 5. 3. 2 . Re come n da t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28-6.O 5UMMARY'0FFINDINGS,1CONCLUSIONSANDRECOMMENDTIONS............. 28 APPENDIX................................................................ 31 I
a e
t S
I I
l
)
l-l t
I
= - - - _ _ . . _ . - . -_-
.j