ML20238D391
| ML20238D391 | |
| Person / Time | |
|---|---|
| Issue date: | 08/28/1987 |
| From: | Vissing G Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| PROJECT-675A NUDOCS 8709110149 | |
| Download: ML20238D391 (103) | |
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UNITED STATES E '
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August 28, 1987.
Project No. 675 MEMORANDUM FOR:
The Record FROM:
Guy S. Vissing, Project Manager i
Standardization and Non-Power Reactor Project Directorate j
Division of Reactor Projects III, IV, V j
and Special Projects Office of Nuclear Reactor Regulation f
SUBJECT:
SUMMARY
OF MEETING WITH COMBUSTION ENGINEERING CONCERNING PRA METHODOLOGY FOR THE CE SYSTEM 80+ DESIGN, AUGUST 25, 1987
]
L INTRODUCTION A meeting of the staff with representatives of Combustion Engineering'(CE) was held at the NRC offices in Bethesda, Maryland, on August 25, 1987. The i
purpose of the meeting was to discuss the approach to developing the
]
probabilistic risk analysis (PRA) for the CE System 80+ oesign. This meeting:
was one of a series to keep the staff informed regarding in the development of q
the System 80+ design. Enclosure 1 provides the list of those who were in.
j attendance. Enclosure 2 provides the CE viewgraphs. Enclosure 3 provides a j
narrative description of the approach to the Level I baseline PRA for the i
System 80+ design.
DISCUSSION CE emphasized that it was important to the development of the System 80+
l design to maintain close and frequent communication with the NRC staff. They desired early feedback on all aspects of their program. We indicated that l
substantive feedback would begin when we receive their submittals.
In describing the program for the System 80+ design, CE indicated that the 4
l interface requirements for the balance of plant (80P) would be more detailed i
than those of the System 80 design. The B0P requirements will be identified as functional requirements and will include reliability goals. Chapter 10' will be submitted in mid-October. Chapter 1 will be submitted in mid-September. Chapters 3, 4, 5 and 9 will be submitted at the end of October.
CE indicated that the major part of the review will deal with the severe accident issues.
- j i
The baseline PRA will be a ' Level I PRA - it will calculate core melt j
i
)k M 99 g 9 870828 l
^
675A pg J
s August 28, 1987 2-probabilities only.
The baseline PRA will be based on the System 80 design.
It will be used in the System 80+ design as a design tool to evaluate changes.
It will consider only internal events. The baseline PRA is scheduled for submittal for information by the end of October.
It was agreed that we would meet to discuss the baseline PRA two weeks following the submittal. Enclet,ure 3 provides a comprehensive discussion of the baseline PRA program.
CE will establish a reliability assurance program (RAP) based on the baseline PRA. The RAP will contain reliability goals for external events and the 80P, To assure that the goals for external events and the BOP are achieveable, CE has retained Duke Power Company as the A & E for the B0P. CE will also have the beniff t of reviews from other utilities. External events probabilities for the System 80+ design will be the subject of a topic paper.
The final PRA for the System 80+ design will be a Level II PRA which will calculate the probabilities of releases to the atmosphere. fheircurrentgoal for a large release is for a probability no per year and a coremeltprobabilityofnogreaterthan10~greaterthan10~
per year. The final PRA will be fault tree analysis approach.
CE does not plan to treat sabotage in the PRA.
However, CE will use the PRA to identify areas of importance and then take action to reduce the vulnerabilities.
Concern was expressed about the use of the MAAP code for the deterministic analyses of core melt events. The staff expressed the view that its usefullness depended on the way it was applied. At the present time the staff is considering how the severe accident issues were going to be handled and the codes that would be used.
RES indicated in the meeting with CE on July 16, 1987, that the NRC MELCOR would be available to CE for comparison and benchmarking with the MAAP code.
It now appears that this may be delayed.
It appears that the application of the MAAP code will be subject to the NRC staff review and resources will need to be allocated for this.
Original signed by Guy S. Vissing, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Project III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
DISTRIBUTION:
As stated
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t August 28, 1987 probabilities only. The baseline PRA will be based on the System 80 design.
It will be used in the System 80+ design as a design tool to evaluate changes. It will consider only internal events. The baseline PRA is scheduled for submittal for information by the end of October. It was agreed that we would meet to discuss the baseline PRA two weeks following the submittal. Enclosure 3 provides a comprehensive discussion of the baseline PRA program.
CE will establish a reliability assurance program (RAP) based on the baseline PRA. The RAP will contain reliability goals for external events and the BOP, To assure that the goals for external events and the BOP are achieveable. CE has retained Duke Power Company as the A & E for the B0P. CE will l
also have the benifit of reviews from other utilities. External events probabilities for the System 80+ design will be the subject of a topic paper.
The final PRA for the System 80+ design will be a Level II PRA which will calculate the probabilities of releases to the atmosphere. fheircurrentgoal for a large release is for a probability no per year and a coremeltprobabilityofnogreaterthan10'greaterthan10~
per year. The final PRA will be fault tree analysis approach.
CE does not plan to treat sabotage in the PRA. However, CE will use the PRA to identify areas of importance and then take action to reduce the vulnerabilities.
Concern was expressed about the use of the MAAP code for the deterministic analyses of core melt events. The staff expressed the view that its usefullness depended on the way it was applied. At the present time the l
staff is considering how the severe accident issues were going to be handled and the codes that would be used. RES indicated in the meeting with CE on July 16, 1987, that the NRC MELCOR would be available to CE for comparison and benchmarking with the MAAP code. It now appears that this may be delayed.
It appears that the application of the NAAP code will be subject to the NRC staff review and resources will need to be allocated for this.
y S. Vissing, Proj t Manager Standardization and Non-Power Reactor i
Project Directorate Division of Reactor Project III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
As stated l
t n
Attendee List' for v.
Meeting with CE Concerning -
PRA Methodology-Name Organization Stan RTIIerbusch c-1 George A. Davis C-E David J. Finnicum
- C-E Bob Jaquith C-E Rick Summitt IT-Corp (ARSAP)
Steve Long NRC/NRR/DREP/RAB Dean Houston NRC/ACRS Brad Hardin NPC/RES/PRAB E. Che111ah
_ WRC/RES/PRAB Mark P. Rubin
'"NRC/NRR/ SAD Franklin Coffman NRC/RES/RHFB Dan Giessing DOE' Charlie Brinkman C-E Bethesda Adel El-Bassioni
..NRC/NRR/DREP/RAB Guy S. Vissing
'"NRC/NPR/PDSNP Tom Kenyon NRC/WRR/PDSNP Tsong-Lun Chu BNL Robert G. Fitzpatrick BNL Trevor Pratt
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ENCLOSURE 2 i
CESSAR DESIGN CERTIFICATION PROGRAM
-A MEETING WITH THE NUCLEAR REGULATORY COMMISSION -
"PRA METHODOLOGY" DATE: AUGUST 25,1987 COMBUSTION ENGINEERING, INC.-
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1 LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSCRED BY COlWUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING -
NOR ANY PERSON ACTING ON ITS BEHALP.
A.
MAKES ANY ' WARRANTY OR REPRESENTATION, EXPRESS OR IMPUED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR -
PURPOSE OR -MERCHANTASIUTY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHCO, j
OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY CWNED RIG >fTS:CR
- 8. ASSUMES ANY UASIUTIES WITH RESPECT TO THE USE OF, OR FOR DAMAGES RESULTING PROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD CR PROCESS DISCLCSED IN THIS REPORT.
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PURPOSE APPRISE NRC STAFF 0F C-E's PLANS FOR UTILIZING PRA METHODOLOGY TO ADDRESS SEVERE ACCIDENT ISSUES IN OUR DESIGN CERTIFICATION PROGRAM.
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4 AGENDA 9:30 AM INTRODUCTION S. E. Ritterbusch 9:35 AM OVERVIEW OF C-E's DOE G. A. Davis-DESIGN VERIFICATION PROGRAM 9:45 AM OVERVIEW 0F PRA SCOPE R. E. Jaquith AND SCHEDULE 10:00 AM PRA STATUS AND METHODOLOGY D. J. Finnicum 11:30 AM LUNCH 12:30 AM EXTERNAL EVENTS R. Jaquith R. Summit (ARSAP) 1:00 PM DISCUSSION AND FUTURE S. E. Ritterbusch MEETING PLANS sonsmust:0NhWCINEERING I
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INTRODUCTION I
o CESSAR-DC LICENSING APPROACH l
0 COMMUNICATIONS WITH NRC 2
MR, S. E. RITTERBUSCH LEAD ENGINEER, STANDARD PLANT LICENSING sossaustloN)aNCINEERING
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CESSAR-DC LICENSING APPROACH i
o REVIEW SYSTEM 80 DESIGN WHICH OBTAINED FDA l
0 UPGRADE AREAS IDENTIFIED FOR IMPROVEMENT o
MODIFY DESIGN TO INCORPORATE RESOLUTION OF i
SEVERE ACCIDENT POLICY ISSUES i
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[0f!!$!NICATI0f ' WITH NRC 0
DESCRIBE PROGRAM BEFORE BEGINNING HRC REVIEW 4
0 ESTABLISH COOPERATIVE ATTITUDfi i
o INTERACT WITH ALL COGNIZANT NRC PERSONNEL O
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0 SEVERE ACCIDENT POLICY ISSUES - START EARLY ADDRESSING NEW ISSUES i
RESOLUTION OF USI's AND GI's FIRS, MEETING:
SEPTEMBER 17, 1987 I
l COMPLETION OF PRA FIRST MEETING:
JUNE 2, 1987 SECOND MEETING:
AUGUST 25, 1987 NRC REVIEW (DEGRADED CORE ISSUES)
FIRST MEETING:
AUGUST 12, 1987 o
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2.0 METHODOLOGY l
As stated in Sections 1.2 and 1.3, the objectives of this analysis are to calculate a baseline core damage frequency for a generic System 80* plant, to determine the dominant core damage contributors and to assess potential e.reas J
for design improvement in the sahanced System 80 This analysis is equivalent to the baseline Probabilistic Safety Analysis (PSA) described in the PSA Procedures Guide (5) and the methodologies employed in this analysis are consistent, within the scope and intent of this analysis, with methodologies outlined in the PSA Procedures Guide (5) and methodologies described in the PRA Procedures Guide (6)
This analysis basically used the small event tree /large fault tree approach, Figure 2.0-1 represent the major tasks in this analysis.
The following sections describe each of these tasks and any associated methodology in greater detail.
2.1 PLANT FAMILIARIZATION This first major task in this analysis was plant familiarization. The i
objective of this task is to collect the information necessary for identification of appropriate initiating events, determination of the success criteria for the systems required to prevent er mitigate the transients and accidents (the front line systems) and to identify the dependence between the i
front line systems and the support systems which are required for proper functioning of the front line, systems. This task was primarily an infomation gathering task.
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The information collected in this task included design information.
operational infomation and infomation on plant responses to transients.
I4) was used to provide infomation on the design of systems within CESSAR-F the basic NSSS scope and interface requirements for the support systems.
Where additional design detail was needed for support systems, typical system designs were generated based on support system designs described in the FSARs of recent vintage C-E plants with similar NSSS designs (,8,9, 0)
Chapter 5 of this document contains the appropriate system descriptions as well as an evaluation of the system commonalities and, dependencies.
Chapter 15 of CESSAR-FI4) provided the basic information on plant responses to accidents and transients. This information was supplemented with discussions
{
l with safety analysts and licensed operators in C-E's Training Department and with infomation contained in the report, "Depressurization and Decay Heat i
Removal - Response to NRC Questions"(11} and its associated supplements pertinent to System 80 plants (12,13)
Several transients were also run using CEPAC(I4), C-E personal computer based PWR simulator, to get a better understanding of the plant's physical response to transients.
l Operator actions during plant transients were evaluated and established based i
on C-E's Emergency Procedure Guideline (15) and discussions with licensed i
operators in C-E's Training Department and at Arizona Nuclear Power Project.
I Surveillance requirements and operability definitions were derived from C-E's l
l Standard Technical Specifications (16) and, where more specific detail was needed, from the Palo Verde Technical Specification $II7)
Maintenance w4ces busd infomation, were needed, was vsed on common industry practices.
1 I
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I1O), several other ' Published PRA Stud es(19,20,21, The Reactor Safety Study 22.23), and the IDCOR IPE ProceduresI24) guide were also reviewed as part of the plant familiarization task. The objectives of these reviews was to provide a broad overview of areas to be addressed in this analysis and to identify potential problem areas.
2.2 ACCIDENT SEQUENCE DEFINITI'Al I
1 1
i The second major task in this analysis was the accident sequence definition.
The objective of this task was to qualitatively identify those accident sequences which lead to core melt / core damage. This was accomplished using event tree analysis. Event tree analysis involves defining a set of initiating events and constructing a set of system event trees which relate plant system responses to each defined initiating event. Each system event
' tree represents a distinct set of system accident sequences, each of which consists of an initiating, event and a combination of various system successes and failures that lead to an identifiable plant state. Procedures for i
developing system event trees are described in detail in the PRA Procedures Guide (6),
d e
p i
For this analysis, the small event tree /large fault tree approach In this approach, only the front line systems which respond to mitigate an accident or transient, are addressed on the event tree. The impact of the support systems is addressed within the fault tree models for the front line systems.
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The first step in defining the accident sequences was to select the initial 1
set of initiating events to be addressed in the analysis.
(Note: initial i
selection.of initiating event is considered to be.part of the plant i
familiarizationtaskinthePSAProceduresGuide(5)). A Master Logic Diagram-(MLD) was constructed to guide the selection and grouping of the initiating events. An MLD is essentially a top level fault tree in which the general j
conditions which could lead to the top level event are deductively determined.
For this analysis, the top event on the MLD was defined to be "offsite to release"eventhoughthescopeoftheanalysisislimitedMdentifyingcore I
damage frequency and dominant contributors.- This was done to ensure completeness and to facilitate any later extension of this analysis. The bottom level elements on the MLD established the initial groupings of initiating event types to be considered.
1 The next step was to develop an initial list of event initiators. First, the
]
lists of suggested event initiators were extracted from the PRA Procedures Guide (0), the PSA Procedures Guide (5), the IPE Methodology Manual (24) and the lists of event initiators and final initiating event groups were extracted II9), the Oconee PRA(20) from the Calvert Cliffs IREP Study (21), the Zion PRA 5
I. These lists of event initiators were then and the Arkansas IREP Study condensed into a single list. This list was then reviewed to eliminate event initiators which were not applicable because of plant design features (e.g.,
PORV LOCAs were eliminated because System 80 doesn't have PORVs). The event initiators were then grouped into initial initiating event classes based on the bottom events on the MLD, the transient grouping in Chapter 15 of I4I CESSAR-F and the initiating events analyzed in the other PRAs(19,20,21,22,23),
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An iterative process was then used to select the final set of initiating events and tar define the event sequences. First, a initial draft of an event l
tree was developed for each of the initial initiating event classes based on plant system responses to the specific type of initiator. These event trees l
were then compared and where the system responses with respect to preventing l
core damage were the same or equivalent, the classes were combined. The event trees were then briefly evaluated for the individual initiators within the class. If the system responses to the specific initiator were not covered by the class event tree, the initiator was either transferred to an event class for which the system responses were appropriate, or a new event class was Tt st.
Y created and a new draft event was developed. This process was repeated until l
a set of initiating event classes were defined that included all the l
1 initiators in the original list and the event tree for each event class j
i covered the system responses for each initiator. The final event trees were then prepared and the description and success criteria were defined for each j
element on the event trees.
In general, the cuccess criteria for the event tree elements were based on the system performance assumed in the Chapter 15 i
and Chapter 6 analyser, in CESSAR-FI4)
For a few elements, however, transientswererunusingtheSystem80CEPA((14I to determine success criteria and to evaluate transient timing. {
l l
2.3 SYSTEM MODELING I
Each system event tree, as described in Section 2.2, represents a distinct set i
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of system accident sequences, each of 6 hich consists of an initiating event i
and a combination of various system successes and failures that lead to an identifiable plant state.
I 2-6 j
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1 1339e/(85G2)/ca-7 I
I 1
Quantification of the system accident sequences requires knowledge of the failure probability or probability of occurrence for each element of the i
system accident sequence. The initiating event frequency and the probability of failure for a system accident sequence element involving the failure of a
/
single component can be quantified directly from the appropriate raw data using methods described in Reference 6.
However, if the system accident sequence element represents a specific failure mode for a system or subsystem, a fault tree model of the system or subsystem must be constructed and quantified to obtain the desired failure probability. Construction of the fault tree requires a complete definition of the functional requirements for the system, given the initiating event to which it is respondi,3, and the physical layout of the system. The system fault tree is a graphic model of the various parallel and serial combinations of component failures that would result in the postulated system failure mode (25)
The symbols used in constructing the fault tree models are presented and defined in Figure 2.3-1.
The evaluation of each fault tree yields both qualitative and quantitative information. The qualitative infonnation consists of the "cutsets" of the model. The cutsets are the various combinations of component failures th e result in the top event, i.e., the failure of the system. The cutsets fonn the basis of the quantitative evaluation which yields the failure probability for the system accident sequence element of concern.
The quantitative evaluation of the fault trees yields several numerical i
measures of a systems failure probability, two of which are typically employed in the event tree quantification, i.e., the unavailability and unreliability.
1 2-7 1
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FIGURE 2.3-1 FAULT TREE SYMBOLOGY 1
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1339e/(85G2)/ca-9 i
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The unavailability is the probability that a system will not respond when demanded. TMs value is used when the system accident sequence element represents a system function or action which is performed-quickly, such as the reseating of a previously opened safety valve, or if the element represents a i
particular condition, such as offsite power unavailable at turbine trip. ' The j
unreliability is the probability that a system will fail (at least once) during a given required operating period. This value is typically used when the system accident sequence element specifies a required operating period for a system, such as auxiliary feedwater system fails to deliver feedwater for four hours. The unreliability is usually added to the unavailability when the I
system accident sequence element represents the failure of a standby system to I
actuate and then run for a specified period of time.
'l Two types of human failures are typically included in fault tree analyses.
They are " pre-existing maintenance errors" and failures of the operator to I
respond to various demands., Pre-existing maintenance errors are undetected errors committed since the last periodic test of a standby system. An example i
of this type of error is the failure to reopen a mini-flow valve which was closed for maintenance. A failure of the operator to respond includes the failure of the operator to perform a required function at all or to perform it correctly. An example of this type of error is the failure of. the operator to back-up the automatic actuation of a safety system.
l For this PRA, failure of the operator to respond to various demands where there was a time constraint was quantified using the Human Cognitive 0)
Reliability The human cognitive reliability model is a set of time Mehl 2-9
4 l
1339e/(85G2)/ca-10 D RA
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l dependent functions which describe the probability of a crew response in performing a task. The human cognitive reliability model permits the analyst to predict the cognitive reliability associated with a non-response for a given task or series of related tasks, once the dominant type of cognitive l
processing (skill-based, rule-basedorknowledge-based),themediumresponse time for the task or tasks under nominal conditions and performance shaping' factors such as stress levels or environment are identified. The inherent l
l l
time dependence in this model made it ideal for evaluating operator responses during a transient. The failure probability for " pre-utsting maintenance errors" was quantified using the Handbook of Human Reliability Analysis (27),
i The Handbook of Human Reliability analysis is an extension of the human reliability analysis methodology developed for WASH 1400, the Reactor Safety Study (18) and is intended to provide methods, models and estimated human error probabilities to enable analysts to make quantitative or qualitative assessments of the occurrence of human errors that a+t3ct the availability or operational reliability of engineered safety systems and components. The l
emphasis is on tasks addressed in the Reactor Safety Study, calibration, maintenance and selected control room tasks related to engineered safety features availability. It is the best available source for evaluating human performance with respect to maintenance, calibration, testing and other tasks performed during nonnal plant operation. However, its time dependent model is not as thorough and explicit as that provided by the human cognitive reliability model.
For this PRA, the small event tree -
fault tree approach was selected.
The event trees developed for this PRA addressed the response of the front 2-10
1339e/(85G2)/ca-11 q
J line systems, that is, those systems directly involved in mitigating the various initiating events. The impact of the support systems was modeled within the front line system models. The electrical suppl were fully l
modeled within each front line system model, and simplified comon cause/comon element models were included for the other support systems. The simplified comon cause/comon element models were developed by constructing full models for each support systems. These models were then compared to each other and to the front line system models to identify the comon elements, and I
comon cause failures. The elements thus identified were driven to the top of their respective fault trees and removed. Then, within the appropriate front line system models, the support system was represented by a simplified model consisting of the identified comon elements, the comon cause failures and an undeveloped event representing the remainder of the fault tree model for that system. These undeveloped events were qualified by quantifying the appropriate fault tree mocels. Figure 2.3-2 illustrates the simplified common l
cause/comen element model, I4) contains interface requirements for the support systems but it CESSAR-F does not contain any support system configurations or P& ids. Therefore, in order to develop the support system models described above, representative support system configurations were developed using the CESSAR-FI4) interface requirements, support system configurations for recent vintage C-E plants
} and the typical system configurations in the NPRDS Reportable
' '9' Scope Manual for C-E Plants (28)
The support system configurations used in this analysis are described in Section 5 of this report.
j 2-11 l
l 1
l FIGURE 2.3-2 SAMPLE SAFETY SYSTEN FAULT TREE WITH SIMPLIFIED COMMON CAUSE/COMON ELEMENT SUPPORT SYSTEM MODEL HPSI 3
SYSTEM FAILS J
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9 HPSIPUMP A FAILS TO RUN
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MECHANICAL COMMON N0 CCW TO FAILURE OF CAUSE FAILURE ggg pU p PUMP OF PUMP I-k I
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FAILURE OF COMMON FAILURE OF SERVICE O99 CAUSE CCW-A WATER FAILURE COMPONENTS OF CCW e
se Y
Y RESIDUAL COMMON ELEMENTS COMMON CAUSE l
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2.4 DATA ASSESSMENT d
Reliability data is needed for the quantification of the system fault trees and the system accident sequences which result in severe core damage. - The data needed for this quantification includes:'
1.
initiating event frequencies 2.
component failure rates (demand and time-dependent) 3.
component repair times and maintenance frequencies 4
common cause failure rates 1.'
l 5.
human failure probabilities k
1 l
6.
special event probabilities (e.g., restoration of offsite power) l 7
s.
7.
error factors for the items above D
Because the analysis documented in this report was for a generic System 80 plant, generic reliability data was used in this analysis. The basic i
initiating event frequencies were extracted from the'.PSA Procedure Guide (5),
EPRI NP-2230(29) and the NREP Generic Data Base (30). The initiating event II9),theOconeePRA(20)andtheCalvert frequencies presented in the Zion PRA I
2-13 l
1339e/(85G2)/ca-14 BA:.
I2U Cliffs IREP Report were also used as guidelines. The appropriate basic initiating event frequencies were used to calculate the needed initiating event class frequencies as described in Section 6 of this report.
2 The basic component failure rate data and associated error factors was
^
extractedfromAppendixAoftheEPRIALWRRequirementsDocument(3), which contains,a compilation of generic failure rate data from other nuclear sources. This data was supplemented with data from WASH 1400(18), the NREP Data Base (30) and IEEE Std. 500I3I) as needed. Component maintenance frequencies and repair times were calculated using the procedures outlined in the PSA Procedure Guide (5)
The specific component failure data used in this analysis is documented in Section 6 of this report.
Common cause failures of components were explicitly modeled in the system fault trees. The comon cause failures were calculated using equivalent Beta factors. The quantification process was equivalent to that outlined in Appendix A of the EPRI ALWR Requirements Document ( } with data extracted from several data sources (32,33,34,35,36,37) as appropriate. The common cause I
failure rates used in this report are documented in Section 6 of this report.
As discussed in Section 2.3, two types of human failure; " pre-existing maintenance errors" and failure of the operator to perform various actions during a transient; were modeled in this analysis. " Pre-existing maintenance errors" were evaluated using the methods described in the Handbook of Human Reliability Analysis (27) and the operator responses during an event were modeledusingtheHumanCognitiveReliabilityModel(26)
Quantification data a
2-14
l 1339e/(85G2)/ca-15 3
=
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1 J>
was primarily extracted from the Handbook of Human Reliability Analysis @ ).
Task breakdowns and quantification of the operator actions are described and documented in Section 6 of this report, l
l The methods and data used for quantifying the probability of special events l
(e.g., the restoration of offsite power within a given time period) were dependent on the specific event. Section 6 of this report documents the methodology and data sources used for quantifying each special event.
i i
l l
l 2.5 ACCIDENT SEQUENCE QUANTIFICATION lhe basic ebjective of this analysis was to cetemine a baseline core damage D
frequency for a generic System 80 plant. The total core damage frequency, due to internal events, is the sum of the frequencies of the system accident sequence frequencies for those system accident sequences which result in core damage. As described in Section 2.2, the system accident sequences leading to e.ve.n t core damage were identified using auan tree analysis. Each system accident sequence consists of an initiating event and one or more additional elements, each representing either a front line system failure or a special event such as failure to restore off site power within a given time or the most reactive rod sticking out of the core. The frequency for the system accident sequence is determined by quantifying the individual elements in the sequence and then I
combining the results in the appropriate manner. The frequencies for the initiating events and the special events are directly calculable. The specific calculations are presented in Section 6.
I O
2-15
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1339e/(85G2)/ca-16
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3 i
4 J
4 d
i The front line system failure probabilities were calculated using conditioned fault tree analysis. The first step in this process was to construct a fault
{
tree model for each front line system that appeared as an element in a system accident sequence (see Section 2.3). The models included submodels for the appropriate support systems.
The next step was to perform a base line quantification of each fault tree using generic failure rates. For those front line systems appearing in the LOCA or steam line break sequences, base line quantifications were made with and without offsite power. This quantification provided a list of.cutsets, the system unreliability and the system unavailability for each front line system. This quantification was performed using the CEREC computer code (see Section 2.6 for descriptions of the codes used in this analvach The third step in this process was to identify comon elements in fault tree models appearing in any given event sequence and to calculate conditional i
failure probabilities for these elements. Take for example, an accident j
sequence, S = I x A x B, where I is the initiating event, A is the failure of System A and B is the failure of System B.
The components appearing in the fault tree model for System B are compared to the components appearing in the j
fault tree model for System A, and a list of common components is generated.
Then, for each component on the comon component list, the probability that the component is failed given that System A is failed is calculated. For any j
f 1
given common component, Z, all System A cutsets containing.N are identified and their probabilities are sumed. This sum is then divided by the total 2-16 t
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l 1339e/(85G2)/ca-17 J
Al.
I failure probability for System A to determine the probability of System A being failed due to a cutset containing Z.
The conditional failure probability for Z given that System A failed (P(Z 'A))iscalculatedas:
P(Z l A) = P(A due to a cutset containing Z)
X P(Z a cutset containing Z) l
+ P(A due to a cutset not containing Z)
XP(ZfacutsetnotcontainingZ)
(eqn. 2.5-1)
)
i l
But, P(Z l 8 cutset centaining Z) is 1 by definition and P(Z, a cutset not A
containing Z) is P(?) or, for small' A,1 T, where r is the mission time for 7
System A.
Thus, equation 2.5-1 becomes:
P(Z A) = P(A due to a cutset containing 2)'
]
+P(AduetoacutsetnotcontainingZ)$P(Z)
(eqn.2.5-2)
After all the conditioned component failure rates'were calculated, the system fault trees were requantified using the appropriate conditioned component failure rates, thus yielding a set of system failure probabilities specific to the initiating event classes.
The final step in the quantification of the core damage frequency was to solve j
t each system accident sequence equation using the appropriate initiating event, special event and system failure probabilities. This was done using CESAM, a Monte Carlo sampling code for equation solving.
1 2-17
1339e/(85G2)/ca-18' L
2.6 SPECIFIC ANALYSIS GROUND RULES In performance of this analysis the following ground rules were used:
a) only internal events were addressed; i
1 b) only events with potential for core damage were addressed; c) event sequences were evaluated only with respect to core damage;
]
j d) initiating events were only evaluated for 100% power conditions; i
e) where needed, realistic, best-ei,timate assumptions were used when evaluating plant responses to an initiating event; f) core damage was assumed if one of the following conditions occurred:
l 1) the core was uncovered for more than 60 seconds, or l
2) the core was exposed to an uncontrolled overpower condition.
g) the full event mission time was 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the success criteria were:
2-18
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~
1339e/(85G2)/ca-19 1) the plant was in cold shutdown with the shutdown cooling system in
. operation, or
]
2) the plant was in stable hot shutdown with decay heat removal via the secondary side, 1
l l
or 3) the plant was stable with hot and cold leg recirculation and recirculation cooling established (LOCAs);
i h) generic failure rates and initiating event frequencies were used for i
1 quantification.
)
i
2.7 DESCRIPTION
OF COMPUTER CODES l
1 2.7.1 CEREC The evaluation of the fault trees constructed for this study was aided by the useofthecomputercodeCEREC(C-EReliabilityEvaluationCode). CEREC is an extensively modified version of the PREP and KITT codes (,3_8,).
The PREP 8
portion of the code, which gene:ates the cutsets, has several modifications to l
l 2-19
11339e/(85G2)/ca-20 RA:~
its output fomat. The KITT portion of the code, which performs the
]
quantitative evaluations, has several major additions to the original KITT capabilities. They are as follows:
1
)
1.
The capability of calculating the unavailability for a periodically tested standby system using either the demand failure rate (inhibit condition) or the standby failure rate, test interval and allowable downtime.
1 l
2.
The capability of filtering out cutsets based on cutoff values for any of five calculated reliability parameters.
1 3.
The capability of automatically perfoming sensitivity analyses on any parameter.
l l
4 The capability of detemining the uncertainty of any of the output reliability parameters based on the uncertainty of the component failure l
l data.
CEREC is written in FORTRAN IV for use on the CDC 7600 computer, and FORTRAN 77 for use on a PRIME 550 minicomputer, I
i l
2.7.2 CEDAR I
The CEDAR code (C-E Dependency Analysis Routine) is a utility code designed to automate the identification of shared components and the. calculation of their 3
1 2-20
4 6 1339e/(85G2)/ca-21 DRF conditional failure probabilities. The PREP portion of the CEREC code produces and stores a file containing the cutsets of 6 system fault tree model. CEDAR identifies common components within these files and calculates their conditional failure probability as the ratio of the. sum of the probabilities of the cutsets ontaining the shared components to the. total system failure probability /Plustherandomfailureprobabilitymultipliedby one minus the ratio of the sum of the probabilities of the cutsets containing l
the shared component to the total system failure probability (see Equation 2.5-2).
CEDAR is written in FORTRAN IV for use on a CDC 7600 computer and FORTRAN 77 for use on a PRIME 550 minicomputer.
2.7.3 CESAM l
CESAM, C-E's version of the SAMPLE code used in the Reactor Safety Study, is l
designed to perform uncertainty analysis on any generalized equation. The required input consists of a FORTRAN function subroutine to describe the function of interest, specification of the type of distributions to be used in modeling the variables of the function and the parameters used to define the distributions for each variable.
Monte CLrlo simulation is perfarmed by sampling the variable distributions and evaluating the function numerous times. These trials then define the distribution of the total function values and CESAM provides various descriptions of this distribution, j
)
i 2-21 l
l
~ ' *
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1339e/(85G2)/ca-22
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i In the analyses performed for this task, the generalized equations consisted of individual sequence and total core damage frequency equations.. The probabilities of the sequence elements were represented by log-normal distributions. The parameters of the distributions were obtained from the l
CEREC runs for each element.
l SAMPLE is written in FORTRAN IV for the CDC 7600, and FORTRAN 77' for a PRIME 550 minicomputer.
-l
{
l 2-22
1339f/(83GS)/ca-1 3.0 INITIATING EVENT EVALUATION t
A t
3.1 MASTER LOGIC DIAGRAM DEVELOPMENT Following plant familiarization, the first major step in determining the plant core damage frequency is to identify the initiating events for which event trees will be constructed and quantified. To systematize the identification i
of the initiating events, a master logic diagram was constructed. The master j
logic' diagram (Figure 3.1-1) is a high level fault tree model of the potential causes of a postulated undesirable event and the logical relationships between these potential causes. Although this analysis ida level 1 PRA and thus addresses only core damage, the top event for the master logic diagram was chosen to be "offsite release of radioactive material" to ensure completeness in the evaluation.
As shown in Figure 3.1-1, the two ways to get an excessive offsite release a) a release of core material; or b) a release,cf non-core material such are:
as radioactive wastes or spent fuel material. Releases of non-core material are not included in this analysis.
A release of core material will occur if there is an event resulting in severe core damage, and the containment fails, be containment failure includes containment bypass and consequential containment faileres. Containment failure was not addressed in this analysis.
Severe core damage can occur due to excess core power or loss of core cooling.
In an excess power event, core damage results from energy generation within
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s the fuel at a, orate greater than the energy,can be transferred to the coo.larit, t 1 Conceptually dextss powe', events can be initiated by CEA ejections, r
uncontrolled CEA withdrawals from zero power or a Ioron dilution event at hot zero power early in core 1ife.
For loss of core cooling events, core damage results from core uncovery due to loss of the primary coolant inventory. Primary coolant inventory can be lost as a result of an unisolable breach of the primary pressure boundary (a LOCA-type event) or as ch result of a loss,of heat transfer to the secondary j
side which leads to the primary pressure and temperature increasing to the i
point where the primary safety valves will lift and discharge primary coolant
)
into containment. Loss of heat transfer to the secondary side can be directly initiated by a loss of feedwater flow, a loss of steam flow or an interruption.
of the primary coolant flow through the core. Loss of heat transfer to the-secondary side can be indirectb initiated by events which result in a reactort trip and turbine trip. (Note: The initiators discussed above, and shown on i
Figure 3.1-1, represent unn tigated initiators. Mitjgstion of these v
l i.'
initiators is evaluated in the event tree analysis. Severe core damage l
actually occurs only if the initiator occurs and the mitigating systems do not i
function.)
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1339f/(83G5)/ca-6 3
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j 3.2 SELECTION OF TNITIATING EVENTS J
J l
The master logic diagram (Figure 3.1-1) was used to identify a set of general conceptual core damage initiators. The next step in selecting the initiating events for event tree analysis was to identify specific. detailed initiating g.
events for'each of these general initiator classes.
First, sets of detailed initiating events were extracted from several reference documents. Table 3.2-1 presents the forty-one (41) PWR initiating l
events formulated in EPRI NP-2230(27) and contained in the PSA Procedures Guide (5)
Table 3.2-2 presents the initial list of initiating events I19) presented in the Zion PRA
, Table 3.2-3 presents the initial list of f
initiating events from the Oconee PRA(20), and Table 3.2-4 presents the list
)
I21}
of initiating events from the Calvert Cliffs IREP Study
. Finally, Table
'3.2-5prehentsalistofinitiatingeventsanalyzedinChapters6and15of I4)
CESSAR-F The lists of initiating events presented in Tables 3.2-1 through l
3.2-5 were compared and combined into a single comprehensive list, grouped by 1
the eight general initiator categories on the master logic diagram (Figure j
3.1-1).
This list, as presented in Table 3.2-6, represents the initial set of l
initiating events for this analysis.
1 1
The initial list of initiating events on Table 3.2-6 then went through an iterative process of review and evaluation to reduce this list to the final set of initiating events for event tree analysis. First, the events were reviewed to identify those events which were outside of the scope of analysis or were precluded because of system design features. Based on this review,
_+
3-6 b
1339f/(83G5)/ca-7 L
1
" Fire Within Plant" (Item 14.1) was eliminated because it is normally grouped with external events which are outside of the scope of this study, "Startup of Inactive RCP" (Item 13.3) was eliminated because C-E plants operate with all four RCPs running except at very low power levels, and this analysis is I
limited to events occurring from at or near 100% power. Likewise," Col ater Addition" (Item 15.5) was eliminated because there is no credible way to get I
significant amounts of cold water into the RCS at operating temperatures and
]
I pressures.
i The next step in the condensation process was evaluate the remaining events 1
1 I
with respect to the physical process perturbations involved and the anticipated response of the front line systems. The objective was to group the events according to similarities in the process perturbations and the responses of the front line systems. This iterative process, described in Section 2.2, involved evaluation of the accident and transient analysis in
)
CESSAR-FI4) and the development of preliminary event trees. The results of this process are discussed below. The final set of initiating event groups i
and their constituent initiating events are presented in Table 3.2-7.
i The primary system LOCA class break sizes were established based on the response of the HPSI and LPSI systems, the need for reactor trip for reactivity control, the need for secondary side heat removal, and the long term cooling method (hot and cold leg injection with recirculation cooling or
" normal"shutdowncooling). Vessel failure was removed from the large LOCA class and established as a separate initiator defined to be "any loss of coolant accident in excess of the ECCS capacity". CEA ejections were included l
3-7
7 0' -
..g
,7, g
1339f/(83G5)/ca-8 4
in the small LOCA class. C-E plants typically operate with,all-rods out (ARO) or witia fer. Cf4'1 slightly inserted when at power, ~ thus, power '
perturbations, if any, would be minimal, and the major impact would be thct 1
t breach of the primary pressure bouadary. Failure of a primary safety. valve (PSV) to reseat following a PSV opening as a consequence oEa transient were I
also included in the small LOCA class.
i l
Loss or offsite power / station blackou?. was entracted from the turbine trip l
category, and established as a separate initi) tor because off the impact of j
Lt.ese events on the front iine systems.. Event tress were constructed for both l
" Loss of Offsite Power" and " Station Blackout". " Total loss of RCS flow" was.
also included in this class because loss of offside power is the most 'iikely means of losing RCS flow ano the response to a loss of'offsite power is l
boundir:g. Loss of RCS flow in one loop was included in the transient ettegory.
+
I In Large stea/line breaks, inside or outside rf containment, ac.d large feedwater linebreaksdownstreauofth.1mainfeedwaterhsolationvalves(MFIVs), result
)
in a rapid blowdown of the secondary system wSh the $ttendant rapid cooldcan of the primary. system. The response of the front line systems...:: equivslent i
for all;of the breaks. Therefore, a single evkm! initiator class, large Thit initiatim; group includes large secondary sida breaks, was established.
f
/
i tteam tir:e breaks inside or catsitb of containment,imr.in feedwater line breaks downstbam of the MFIVsJspurious openings of multiple turbine bypass valves.
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1339f/(83G5)/ca-9 l
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k, atmospheric dump valves or main steam safety valves. Small steam losses which do not result
, a significant blowdown of the secondary system were included i
l in the transient category.
i 9
An evaluation of the process perturbations and the front line system responses to the full or partial loss of fadwater events (event category 7 on Table l
3.2-6), the loss of steam flow events (event categories 8 and 9), with the j
exception of loss of offsite power and the indirect initiatc*s-(MLD category G on Table 3.2-6) with the exception of the large steam and feedwater line breaks indicated that all these events had equivalent responses. The events produces a process perturbation which resulted in a reactor trip and turbine trip. Main feedwater flow was ramped back, and with a presumed loss of main feedwater flow on the ramp back, the auxiliary feedwater system was actuated.
Steam removal following the trip was via the turbine bypass valves or l
atmospheric dump valves. Loss of condenser vacuum and loss of circulating water events do affect the turbine bypass system, but this was addressable in the turbine bypass system model. Therefore, all of these events were combined B,20,21,22) into the single initiator class " transients". In other PRA's
" loss of component cooling water" and " loss of service water" were included as l
l special initiators because they could initiate a plant transient and affect the ability of the front line systems to respond to the transients. System E
80 plants have a separate essential cooling water system for the essential systems and the ECCS pumps are air-cooled. Therefore, these two events can be treated as transient initiators in this analysis.
l As previously discussed, C-E plants operate at power with all rods out or with a few rods slightly inserted. A CEA withdrawal at power, if possible, would 3-9
. ______ _ _________ _ a
s' 1339f/(83G5)/ca-30
'l n
1 d
produce only a minor power perturbation'followed by a reactor trip and subsequent secondary system responses. Therefore, CEA withdrawal was included in the transient category.
._ I Strictlyspeaking,"AnticipatedTransientWithoutScram(ATWS)"isnotan initiating event, but rather is a faulted response to an event requiring CEA insertion for reactivity control. However, because of the significant impact that an ATWS has on plant responses, it is included as a separate initiating event class.
l l
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i i
l 1
l 3-10
i 1339f/(83G5)/ca-11 i
TABLE 3.2-1 II)-
GENERAL PWR TRANSIENT CATEGORIES Category Title 1
Loss of RCS Flow (1 Loop)
~
l 2
Uncontrolled Rod Withdrawal
-3 CROM Problems and/or Rod Drop -
. i 4
Leakage from Control Rods 5
Leakage in Primary System 6
Low Pressurizer Pressure 7
Pressurizer Leakage f
8 High Pressurizer Pressure J
9 Inadvertent Safety Injectiori Signal 10 Containment Pressure Problems
)
11 CVCS Malfunction - Boron Dilution l
12 Pressure / Temperature / Power Imbalance-Rod Position Error 13 Startup of Inactive Coolant Pump 14 Total loss of RCS Flow 15 Loss or Reduction in Feedwater Flow (11. cop) 16 Total Loss of Feedwater Flow (Mi Loops) 17 Full or Partial Closure of MSIV (J Loop) 18 Closure of All MSIV 19 Increase in Feedwater Flow (.1 Loop) 20 Increase in Feedwater Flow.(All Logps) 21 Feedwater Flow Instability - Operator Error q
22 Feedwater Flow Instability i Misc. Mechanical Causes 2
23 LossofCondensatePump(1Liop)\\
24 Loss of Condensate Pumps (All ioops) l 25 Loss of Condenser Vacuum I
26 Steam Generator Leakage
)
27 Condenser Leakage t
28 Miscellaneous Leakage in S'econdary System 3-11 i
i 1339f/(83G5)/ca-12
=$
d4 TABLE 3.2-1 (Cant.F-Category Title 29 Sudden Opening of Steam Relief Valves 30 Loss of Circulating Water 31 Loss of Component Cooling 32 Loss of Service Water Systems 33 Turbine Trip, Throttle Valve Closure, EHC Problems 34 Gene tor Trip or Generator Caused Faults 35 Tota oss of Offsite Power 36 Pressurizer Spray Failure 37 Loss of Power to Necessary Plant Systems 38 Spurious Trips - Cause Unknown 39 Auto Trip - No Transient Condition 40 Manual Trip - No Transient Condition 41 Fire Within Plant Note 1: Extracted from PSA Procedures Guide, NUREG/CR-2815.
1 I
I i
I i
3-12
o 1339f/(83G5)/ca-13' l
J%A =T l
TABLE 3.2-2 d
II)
ZION UNIT 1 AND 2 INITIATING EVENT CATEGORIES i
1.
LARGE LOSS _OF COOLANT, ACCIDENT (Blowdown greater than 6-inch pipe rupture) 1.1 Pipe Failures j
l 1.2 Valve Failures l
1.3 Vessel Failures 1.4 Other Large LOCAs
)
2.
MEDIUM LOSS OF COOLANT ACCIDENT (Blowdown in range of 2 to 6-inch pipe l
rupture) 2.1 Pipe Failures 2.2 Pressurizer Safety and Relief Valve Failures (Multiple) 2.3 Other Valve Failures 2.4 Other Medium LOCAs 3.
SMALL LOSS OF COOLANT ACCIDENT [ Blowdown less than 2-inch pipe rupture)
)
3.1 Pipe Failure 3.2 Pressurizer Relief Valve or Safety Valve Failures 3.3 Other Valve Failures 3.4 Control Rod Drive Mechanism Failures 3.5 Reactor Coolant Pump Seal Failure (4 or less) l 3.6 Other Small LOCAs t
4 LEAKAGE TO SECONDARY COOLANT l
4.1 Single Steam Generator Tube Rupture 4.2 Other Steam Generator Leaks d
3-13
____----_-_o
1339f/(83G5)/ca-14 TABLE 3.2-2 (Cont.)
I l
i l
5.
LOSS OF REACTOR COOLANT FLOW 5.1 Loss of Reactor Coolant Flow in One Loop 5.2 Loss of Reactor Coolant Flow in All Loops 5.3 Other Losses of Reactor Coolant Flow 6.
LOSS OF FEEDWATER FLOW 6.1 Feedwater Pipe Rupture Outside Containment 6.2 Loss / Reduction of Feedwater Flow in One Steam Generator 6.3 Loss of Feedwater Flow in All Steara Generators 6.4 Feedwater Flow Instability - Operator Error 6.5 Feedwater Flow Instability - Mechanical Causes 6.6 Loss of One Condensate Pump 6.7 Loss of All Condensate Pumps 6.8 Condenser Leakage
- 6. 5' Other Secondary Leakage 7.
PARTIAL LOSS OF STEAM FLOW 7.1 Full Closure of Main Steam Isolation Valve (MSIV)
{
7.2 Partial Closure of Main Steam Isolation Valve
]
7.3 Other Losses of Steam Flow
-l 8.
TURBINE TRIP 8.1 Turbine Trip (General) 8.1.1 Closure of all main steam isolation valves 8.1.2 Increase in feedwater flow in one or more steam generator's 8.1.3 Loss of condenser vacuum 8.1.4 Loss of circulating water
.{
l 1
)
3-14 J
___-_ D
1
~
1339f/(83G5)/ca-15 3
1 TABLE 3.2-2 (Cont.)
J 8.1.5 Throttle-valve closure / electrohydraulic control problems 8.1.6 Generator Trip or generator-caused faults 8.1.7 Turbine trip due to overspeed
~
8.1.8 Other turbine trips 8.2 Turbine Trip Due to Loss of Offsite Power 8.3 Turbine Trip Due to Loss of Service Water 9.
SPURIOUS SAFETY INJECTION 9.1 Spurious Safety Injection: Charging Pumps Operate
- 10. REACTOR TRIP 10.1 Reactor Trip g
10.1.1 Control rod drive mechanism problems and/or rod drop 10.1.2 High or low pressurizer pressure j
10.1.3 High pressurizer level 10.1.4 Spurious automatic trip -- no transient condition 10.1.5 Automatic / manual trip -- operator error 10.1 6 Manual trip due to' false signal 10.1.7 Spurious trip -- cause unknown 10.1.8 Primary system pressure, temperature, power imbalance 10.1.9 Other reactor trips I
10.2 Reactor Trip Due to Loss of Component Cooling Water 10.3 Reactor Trip Due to Loss of DC or AC Power l
I 1
3-15 l
l
1 a 9
+
1339f/(8305)/ca-15 4
m TABLE 3.2-2(Cont.)
l l
- 11. LOSS OF STEAM INSIDE CONTAINMENT l-l
=~~
l 11.1 Steam Pipe Rupture Inside Containment 11.2 Feedwater Pipe Rupture Inside Containment 11.3 Steam Relief Valve or Safety Valves Open Inadvertently (included'-
with inside containment group for functional reasons -- leak upstream of MSIVS) j 11.4 Other Steam Losses Inside Containment
- 12. LOSS OF STEAM OUTSIDE CONTAINMENT 1l 12.1 Steam Pipe Rupture Outside Containment l
12.2 Throttle-Valve Opening / Electrohydraulic Control Problems l
12.3 Steam Dump Valves Failing Open j
12.4 Other Steam Losses Outside Containment l
- 13. CORE POWER INCREASE ll 13.1 Uncontrolled Rod Withdrawal
)
l 13.2 Boron Dilution -- Chemical Volume Control System Malfunction i
13.3 Core Inlet Temperature Drop 13.4 Other Positive Reactivity Addition i
Note 1: FromZionPR6.
p w
3-16 i
i
o-1339f/(83G5)/ca-17 A.
TABLE 3.2-3 POTrNTIAL INITIATING EVENTS FROM OCONEE PRA(1)
Reactivity Control:
1.
Rod Drop 2.
Inadvertent rod withdrawal 3.
Rod ejection 4
Inadvertent boration or deboration 5.
6.
Cold-water addition Core-Heat Removal:
7.
Reactor coolant pump trip 8.
Reactor coolant pump seizure 9.
Flow channel blockage RCS Heat Removal-l
- 10. Loss of main feedwater
- 11. Excess of feedwater (main or energency) l
- 15. Steamline breaks (inside and outside containment)
- 16. Turbine control valve malfunction
- 17. Turbine bypass valve inadvertent' opening
- 18. Turbine trip or malfunction
- 19. Loss of circulating water 3-17 l
1339f/(83G5)/ca-18 S
}";
TABLE 3.2-3 (Cont.)
j 1
l l
Control of RCS Inventory and Pressure:
l
- 20. Small RCS pipe breaks
- 21. Large RCS pipe breaks
- 22. Inadvertent PORV or safety-valve opening l
- 23. Failure of reactor-coolant pump seals
- 24. Leakage of control-rod drive seals
- 25. ' Interfacing-system loss of coolant
- 26. Reactor-vessel rupture
- 27. Steam-generator-tube leak / rupture
- 28. Charging exceeds letdown 1
- 29. Letdown exceeds charging 30.
Inadvertent high-pressure injection
- 31. Failure on or off of pressurizer heaters
- 32. Failure on or off of pressurizer spray Maintenance of Vital Support Systems:
l l
- 33. Loss of offsite power 1
34.
Loss of power to necessary systems
- 35. Partial losses of power to control systems d6.
Loss of service water
- 37. Loss of component cooling
- 38. Loss of instrument air
- 39. Integrated control system' failures
- 40. Fire affecting necessary systems l
- 41. Internal flooding affecting necessary systems
.I 1
l i
3-18
i o
i 1339f/(83G5)/ca-19 1
i TABLE 3.2-3 (Cont.)
r"'
Maintenance of Nonnal Power Operation:
- 42. Generator faults'
- 43. Grid disturbances'
- 44. Administrative 1y caused shutdown i
Note 1: From Oconee PRA, NSAC/60 1
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1339f/(83G5)/ca-22 5
TABLE 3.2-5 j
EVENTS ANALYZED IN CESSAR-F Number Event Section 2
1 LargeBreakLOCAs(>.5ft) 6.3.3.2
~
2 2
Small Break LOCAs (<.5 ft )
6.3.3.2 3
Decrease in Feedwater Temperature 15.1.1 4
Increase in Feedwater Flow 15.1.2 5
Increased Main Steam Flow 15.1.3 6
Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.4 7
Steam System Piping Failures Inside and Outside Containment 15.1.5 8
Loss of External Load 15.2.1 9
Turbine Trip 15.2.2 i
10 Loss of Condenser Vaccam 15.2.3 1
11 Main Steam isolation Valve Closure 15.2.4 12 Steam Pressure Regulator Failure 15.2.5 13 Loss of Non-Emergency A-C Power to Station Auxiliaries 15.2.6 14 Loss of Normal Feedwater Flow 15.2.7 15 Total Loss of Reactor Coolant Flow 15.3.1 l
16 Flow Controller Malfunction Causing Flow Coastdown 15.3.2 j
17 Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power 15.3.3'
.{
1 18 Reactor Coolant Pump Shaft Break with Loss of
/
Offsite Power 15.3.4 l
19 Uncontrolled CEA Withdrawal from a Suberitical i
or Low Power Condition 15.4.1 1
3-22
m
'1339f/(83G5)/ca-23 TABLE 3.2-5 (Cont.)
k i
w Number E'.nt Section 20 Uncontrolled CEA V' ndrawal at Power 15.4.2 i
21 Single CEA Drop 15.4.3 l
1 22 Startup of an Inactive RCP 15.4.4 23 Inadvert'ent Deboration 15.4.6' 24 Inadvertent Loading of a Fuel Assembly into the Improper Position 15.4.7
)
25 CEA Ejection 15.4.8 26 Inadvertent Operation of ECCS 15.5.1 27 CVCS Malfunction - Pressurizer Level Control System Malfunction with Loss of Offsite Power 15.5.2 l
28 Inadvertent Opening of a Pressurizer Safety /
Relief Valve 15.6.1 29 Double Ended Break of a Letdown Line Outside Containment 35.6.2 30 Steam Ger.erator Tube Rupture 15.6.3 j
i l
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