ML060880464
| ML060880464 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/21/2006 |
| From: | ERIN Engineering & Research |
| To: | Office of Nuclear Reactor Regulation, Tennessee Valley Authority |
| References | |
| TAC MC3812 | |
| Download: ML060880464 (177) | |
Text
ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 1 BFN EPU CONTAINMENT OVERPRESSURE (COP) CREDIT RISK ASSESSMENT I
See attached.
- I I I 1,
BFN EPU Containment Overpressure (COP)A Credit Risk Assessment Rev. 1 I
Performed for:
Tennessee Valley Authority Performed by:
ERIN Engineering and Research, Inc.
March 21, 2006
BFNEPUCOP ProbabilisticRisk Assessmnent
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C1320503-6924Rl -3/22/2006
BFNEPUCOPProbabilisticRiskAssessment Table Of Contents Section Page EXECUTIVE
SUMMARY
................ ii
1.0 INTRODUCTION
.i -1 1.1 Background. 1-1 1.2 Scope. 1-4 1.3 Definitions .1-4 1.4 Acronyms. 1-6 2.0 APPROACH .2-1 2.1 General Approach .2-1 2.2 Steps to Analysis .2-3 3.0 ANAYSIS .3-1 3.1 Assessment of DBA Calculations .31 3.2 Probability of Plant State 1 and Plant State 2. -4 3.3 Pre-Existing Containment Failure Probability ..- 6 3.4 Modifications to BFN Unit 1 PRA Models................................................ E-7 3.5 Assessment of Large-Late Releases . 3-9 4.0 RESULTS .4-1 4.1 Quantitative Results................................................................................ .4-1 4.2 Uncertainty Analysis .4-1 4.3 Applicability to BFN Unit 2 and Unit 3.4-13
5.0 CONCLUSION
S 5-1 REFERENCES Appendix A PRA Quality Appendix B Probability of Pre-Existing Containment Leakage Appendix C Assessment of Browns Ferry Data Appendix D Large-Late Release Impact Appendix E Revised Event Trees Appendix F Revised Fault Trees i C1320503-6924R1 -3/2212006 I
BFNEPUCOP ProbabilisticRisk.Assessment EXECUTIVE
SUMMARY
The report documents the risk impact of utilizing containment accident pressure (containment overpressure) to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps during DBA LOCAs.
The risk assessment evaluation uses the current BFN Unit 1 Probabilistic Risk Assessment (PRA) internal events model (including internal flooding). The BFN PRA provides the necessary and sufficient scope and level of detail to allow the calculation of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) chances due to the crediting of containment overpressure in determining sufficient NPSH requirements for the RHR system and Core Spray system emergency core cooling pumps.
The steps taken to perform this risk assessment evaluation are as follows:
- 1) Evaluate sensitivities to the DBA LOCA accident calculations to determine under what conditions credit for COP is required to satisfy low pressure ECCS pump NPSH.
- 2) Revise all large LOCA accident sequence event trees to make low pressure ECCS pumps dependent upon containment isolation when other plant pre-conditions exist (i.e., SW high temperature, SP initial high temperature, SP low water level).
- 3) Modify the existing BFN PRA Containment Isolation System fault tree to include the probability of pre-existing containment leakage.
- 4) Quantify the modified PRA models and determine the following risk metrics:
- Change in Core Damage Frequency (CDF)
- Change in Large Early Release Frequency (LERF)
- 5) Perform modeling sensitivity, studies and a parametric uncertainty analysis to assess the variability of the results.
ii C1320503-6924R1 - 3/22/2006
BFNEPUCOPProbabilisticRiskAssessment The conclusion of the plant internal events risk associated with this assessment is as follows.
- 1) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF)I below 106/yr. jBased on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in CDF (1.4E-09/yr). I
- 2) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changeslin risk as resulting in increases of Large Early Release Frequency (LERF) below 1O7/yr.J Based. on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in LERF (1.4E-09/yr). I ii Ii Ii i
i . .
I 1
I i l',ii ii C1320503-6924R1 -3/22/006
BFNEPUCOP ProbabilisticRiskAssessment Section 1 INTRODUCTION The report documents the risk impact of utilizing containment accident pressure (containment overpressure) to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps during DBA LOCAs.
Revision 1 of this report incorporates the following changes:
- 1) Reduction in the peak SP temperature requiring containment overpressure credit
- 2) Probabilistic credit in the risk analysis of the likelihood of a low SP water volume at the start of the postulated DBA LLOCA accident.
These two changes counteract (i.e., change I increases the calculated ACDF, and change 2 reduces the calculated ACDF), resulting in risk results slightly lower than the Rev. 0 analysis. Both the Rev. 0 and Rev. 1 results represent "very small" changes in risk per RG 1.174.
1.1 BACKGROUND
Tennessee Valley Authority (TVA) submitted the BFN extended power uprate (EF'U) license amendment request (LAR) to the NRC in June 2004. In a October 3, 2005 letter to TVA, the NRC requested the following additional information on the EPU LAR:
"SPSB-A. 11 As part of its EPU submittal, the licensee has proposed taking credit (Unit
- 1) or extending the existing credit (Units 2 and 3) for containment accident pressure to provide adequate net positive suction head (NPSH) to the ECCS pumps. Section 3.1 in Attachment 2 to Matrix 13 of Section 2.1 of RS-001, Revision 0 states that the licensee needs to address the risk impacts of the extended power uprate on functional and system-level success criteria. The staff observes that crediting containment accident 11 C1320503-6924R1 - 3/22/ 006 I
BFNEPUCOPProbabilisticRiskAssessm t pressure affects the PRA success criteria; therefore, the PRA should contain accident sequences involving ECCS pump cavitation due to inadequate containment pressure. Section 1.1 of Regulatory Guide (RG) 1.174 states that licensee-initiated licensing basis change requests that go beyond current staff positions may be evaluated by the staff using traditional engineering analyses as wel as a risk-informed approach, and that a licensee may be requested to submit supplemental risk information if such information is not submitted by the licensee. It is necessary to consider risk insights, in addition to the results of traditional engineering analyses, while determining the regulatory acceptability of crediting containment accident pressure.
Considering the above discussion, please provide an assessment of the credit for containment accident pressure against the five key principles of risk-informed decisionmaking stated in RG 1.174 and SRP Chapter 19.
Specifically, demonstrate that the proposed containment accident pressure credit meets current regulations, is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in an increase in core-damage frequency and risk that is small and consistent with the intent of the Commission's Safety Goal Policy Statement, and will be monitored using performance measurement strategies. With respect to the fourth key principle (small increase in risk), provide a quantitative risk assessment that demonstrates that the proposed containment accident pressure credit meets the numerical risk acceptance guidelines in Section 2.2.4 of RG 1.174. This quantitative risk assessment must include specific containment failure mechanisms (e.g., liner failures, penetration failures, primary containment isolation system failures) that cause a loss of containment pressure and subsequent loss of NPSH to the ECCS pumps."
Typical of other industry EPU LAR submittals, the BFN EPU LAR includes a request to credit containment accident pressure, also known as contain!ment overpressure (COP),
in the determination of net positive suction head (NPSH) for, low pressure ECGS systems following design basis events. Also consistent with oth~er industry EPU LAR submittals, the NRC is requesting risk information from licensees regarding the COP credit request.
BFN Units 2 and 3 already have existing approvals for containment overpressure credit.
The 8FN EPU LAR requests containment overpressure credit for BFN Unit 1 for DBA LLOCOA accidents.
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BFNEPUCOPProbabilisticRiskAssessment The need for COP credit requests is driven by the conservative nature of design basis accident calculations. Use of more realistic inputs in such calculations shows that no credit for COP is required. In any event, the request for containment accident pressure credit is a physical aspect that will exist during the postulated design basis accidents.
The EPU LAR simply requests to include that existing containment accident pressure in the ECCS pump NPSH calculations. The NRC request is to investigate the impact on risk if the containment accident pressure is not present (e.g., postulated pre-existing primary containment failure) during the postulated scenarios.
The Nuclear Regulatory Commission (NRC) has allowed credit for COP to satisfy NP'SH requirements in accordance with Regulatory Guide 1.82 (RG 1.82). Specifically, RG 1.82 Position 2.1.1.2 addresses containment overpressure as follows:
"For certain operating BWRs for which the design cannot be practicably altered conformance with Regulatory Position 2.1.1. 1 may not be possible.
In these cases, no additional containment pressure should be included in the determination of available NPSH than is necessary to preclude pump cavitation. Calculation of available containment pressure should underestimate the expected containment pressure when determining available NPSH for this situation. Calculation of suppression pool water temperature should overestimate the expected temperature when determining available NPSH."
The proposed change in the BFN license basis regarding credit for COP meets the approved positions of RG 1.82. However, developments between the NRC staff and members of the Advisory Committee on Reactor Safeguards (ACRS) in 2005 regarding proposed language to Revision 4 of RG 1.82 prompted the NRC to request performance of a 'risk-informed' assessment in accordance with NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis".
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BFNEPUCOP ProbabilisticRiskAssessmcnt 1.2 SCOPE This risk assessment addresses principle #4 of the RG 1.174 risk informed structure.
Principle #4 of RG 1.174 involves the performance of a risk assessment to show that the impact on the plant core damage frequency (CDF) and large early release frequency (LERF) due to the proposed change is within acceptable ranges, as defined by RG 1.174. The other principles (#143, and #5) are not addressed in this report.
This analysis assesses the CDF and LERF risk impact on the BFN Unit 1 at-power internal events PRA resulting from the COP credit requirement for low pressure EC('S pumps during large LOCA scenarios.
External event and shutdown accident risk is assessed on a qualitative basis.
In addition, a review of the BFN Unit 2 and Unit 3 models is performed to show that the results from the Unit 1 BFN PRA apply to Units 2 and 3, as well.
1.3 DEFINITIONS Accident sequence - a representation in terms of an initiating event followed by a combination of system, function and operator failures or successes, of an accident that can lead to undesired consequences, with a specified end state (e.g., core damage or large! early release). An accident sequence may contain many unique variations of events that are similar. l Core damage - uncovery and heatup of the reactor core to the point at which prolonged oxidation and severe fuel damage is anticipated and involving enough of the core to cause a significant release.
Core damage frequency - expected number of core damage events per unit of time.
End State - is the set of conditions Sat the end of an event sequence that characterizes the impact of the sequence on the plant or the environment. End states typically include:
success states, core damage sequences, plant damage statesfor Level 1 sequences, and release categories for Level 2 sequences.
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BEN EPU COPProbabilisticRisk Assessment Event tree - a quantifiable, logical network that begins with an initiating event or condition and progresses through a series of branches that represent expected system or operator performance that either succeeds or fails and arrives at either a successful or failed end state.
Initiating Event - An initiating event is any event that perturbs the steady state operation of the plant, if operating, or the steady state operation of the decay heat removal systems during shutdown operations such that a transient is initiated in the plant. Initiating events trigger sequences of events that challenge the plant control and safety systems.
ISLCOCA - a LOCA when a breach occurs in a system that interfaces with the RCS, where isolation between the breached system and the RCS fails. An ISLOCA is usually characterized by the over-pressurization of a low-pressure system when subjected to RCS pressure and can result in containment bypass.
Large early release - the rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of off-site emergency response and protective actions.
Large early release frequency - expected number of large early releases per unit of time.
Level I - identification and quantification of the sequences of events leading to the onset of core damage.
Level 2 - evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment.
Plant damage state - Plant damage states are collections of accident sequence end states according to plant conditions at the onset of severe core damage. The plant conditions considered are those that determine the capability of the containment to cope with a severe core damage accident. The plant damage states represent the interface between the Level 1 and Level 2 analyses.
Probability - is a numerical measure of a state of knowledge, a degree of belief, or a state of confidence about the outcome of an event.
Probabilisticrisk assessment - a qualitative and quantitative assessment of the risk assoziated with plant operation and maintenance that is measured in terms of frequency of occurrence of risk metrics, such as core damage or a radioactive material release and its effects Don the health of~the public (also referred to as a probabilistic rsk assessment, PRA).
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BFNEPUCOPProbabilisticRiskAssessment Release category - radiological source term for a given accident sequence that consists of the release fractions for various radionuclide groups (presented as fractions of initial core inventory), and the timing, elevation, and energy of release. The factors addressed in the definition of the release categories include the response of the containment structure, timing, and mode of containment failure; timing, magnitude, and mix of any releases of radioactive material; thermal energy of release; and key factors affecting deposition and filtration of radionuclides. Release categories can be considered the end states of the Level 2 portion of a PRA.
Risk - likelihood (probability) of occurrence of undesirable event, and its level of damage (consequences).
Risk metrics - the quantitative value, obtained from a risk assessment, used to evaluate the results of an application (e.g., CDF or LERF).
ii Severe accident - an accident that involves extensive core damage and fission product release into the reactor vessel and containment, with potential release to the environment.
Split Fraction - a unitless parameter (i.e., probability) used in quantifying an event tree.
It represents the fraction of the time that each possible outcome, or branch, of a particular top event may be expected to occur. Split fractions are, in general, conditional on precursor events. At any branch point, the sum of all the split fractions representing possible outcomes should be unity. (Popular usage equates "split fraction" with the failure probability at any branch [a node] in the event tree.)
1.4 ACRONYMS ACRS Advisory Committee on Reactor Safeguards ATWS Anticipated Transient without Scram BFN Browns Ferry Nuclear plant CCF Common Cause Failure CDF Core Damage Frequency CET Containment Event Tree COP Containment Overpressure CPPU Constant Pressure Power Uprate 1-6 C1320503-6924R1 -3/22/2006 I
BFN EPU COPProbabilisticRisk Assessmcnt DBA Design Basis Accident DW Drywell ECCS Emergency Core Cooling Systems EPU Extended Power Uprate GE General Electric HEP Human Error Probability HPCI High Pressure Core Injection system HRA Human Reliability Analysis IPE Individual Plant Examination IPEEE Individual Plant Examination for External Events ISLOCA Interface System Loss of Coolant Accident La Maximum Allowable Primary Containment Leakage Rate LERF Large Early Release Frequency LOCA Loss of Coolant Accident LLOCA Large LOCA LOOP Loss of Offsite Power event LPCI Low Pressure Coolant Injection MAAP Modular Accident Analysis Program NPSH Net Positive Suction Head NRC United States Nuclear Regulatory Commission PRA Probabilistic Risk Assessment PSA Probabilistic Safety Assessment RCIC Reactor Core Isolation Cooling System 1-7 C1320503-6924R1 -3/22/2X)6 I
BFNEPUCOPProbabilisticRisk Assessment RG Regulatory Guide RHR Residual Heat Removal System RPV Reactor Pressure Vessel SMA Seismic Margins Assessment SP Suppression Pool SPC Suppression Pool Cooling SW Service Water TS Technical Specifications TVA Tennessee Valley Authority WW Wetwell 1-8 C1320503-6924R1 - 3/22/2006
BFNEPUCOP ProbabilisticRiskAssessment Section 2 APPROACH This section includes a brief discussion of the analysis approach and the types of inputs used in this risk assessment.
2.1 GENERAL APPROACH This risk assessment is performed by modification and quantification of the BFN PRA models.
2.1.1 Use of BFN Unit 1 PRA The current BFN Unit 1 PRA models (BFN model U1050517) are used as input to perform this risk assessment. The Browns Ferry PRA uses widely-accepted PRA techniques for event tree and fault tree analysis. Event trees are constructed to identify core damage and radionuclide release sequences. The event tree "top events" represent systems (and operator actions) that can prevent or mitigate core damage.
Fault trees are constructed for each system in order to identify the failure modes.
Analysis of component failure rates (including common cause failures) and human error rates is performed to develop the data needed to quantify the fault tree models.
For the purpose of analysis, the Browns Ferry PRA divides the plant systems into two categories:
- 1. Front-Line Systems, which directly satisfy critical safety functions (e.g.,
Core Spray and Torus Cooling), and
- 2. Support Systems, which are needed to support operation of front-line systems (e.g., AC power and service water).
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BFNEPUCOPProbabilisticRiskAssessment FronL-line event trees are linked to the end of the Support System event trees lor sequence quantification. This allows definition of the status of all support systems For each sequence before the front-line systems are evaluated. Quantification of the event tree and fault tree models is performed using personal computer version of the RISKMAN code.
The Support System and Front-Line System event trees are "linked" together and solved for the core damage sequences and their frequencies. Each sequence represents an initiating event and combination of Top Event failures that results in core damage. The frequency of each sequence is determined by the event tree structure, the initiating event frequency and the Top Event split fraction probabilities specified by the RISKMAN master frequency file. RISKMAN allows the user to enter the split fraction names and the logic defining the split fractions (i.e., rules) to be selected for a given sequence based on the status of events occurring earlier in the sequence or on the type of initiating event.
2.1.2 PRA Quality The BFN PRA used as input to this analysis (BFN model U1050517) is of sufficient quality and scope for this application. The BFN Unit 1 PRA is highly detailed, including a wide variety of initiating events (e.g., transients, internal floods, LOCAs inside and outside containment, support system failure initiators), modeled systems, extensive level of detail, operator actions, and common cause events.
The BFN Units 2 and 3 at-power internal events PRAs received a formal industry PRA Peer Review in 1997. All of the "A"and "B" priority comments have been addressed.
Refer to Appendix A for further details concerning the quality of the BFN PRA.
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BFNEPUCOP ProbabilisticRiskAssessment 2.2 STEPS TO ANALYSIS The performance of this risk assessment is best described by the following major analytical steps:
- Assessment of DBA calculations
- Estimation of pre-existing containment failure probability
- Analysis of relevant plant experience data
- Comparison to ACDF and ALERF RG 1.174 acceptance guidelines
- Performance of uncertainty and sensitivity analyses
- Assessment of "Large Late" Release Impact
2.2.1 Assessment of DBA Calculations The purpose of this task is to develop an understanding of the BFN EPU design basis LLOCA calculations that result in the need to credit 3 psig containment overpressure credit.
The need for COP credit requests is driven by the conservative nature of design basis accident calculations. The DBA LOCA calculations are reviewed and sensitivity calculations performed to determine under what conditions of more realistic inputs; is there no need for COP credit in the determination of low pressure ECCS pump NPSH.
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BFNEPUCOPProbabilisticRiskAssessmmnt 2.2.2 Estimation of Pre-Existina Containment Failure Probability This task involves defining the size of a pre-existing containment failure pathway to be used in the analysis to defeat the COP credit, and then quantifying the probability of occurrence of the un-isolable pre-existing containment failure. The approach to this input parameter calculation will follow. EPRI guidelines regarding calculation of pre-existing containment leakage probabilities in support of integrated leak rate test (ILRT) frequency extension LARs (i.e., EPRI Report 1009325, Risk Imoact of Extended Integrated Leak Rate Testing Intervals, 12/03).[2] This is the same approach used in the recent Vermont Yankee EPU COP analyses presented to the ACRS in December 2005.
The pre-existing unisolable containment leak probability is combined with the BFN PRA containment isolation failure on demand fault tree (CIL) to develop the likelihood of an unisolated primary containment at t=0 that can defeat the COP credit necessary for the determination of adequate low pressure ECCS pump NPSH.
2.2.3 Analysis of Relevant Plant Experience Data An unisolated primary containment is not the only determining factor in defeating low pressure ECCS pump NPSH. The DBA calculations show that other extreme low likelihood plant conditions are required at t=0 to result in'the need to credit COP in the determination of pump NPSH, such as:
- High initial reactor power level
- High river water temperature
- High initial torus water temperature
- Low initial torus water level 2-4 C1320503-6924R1 - 3/2212006 I
BFNEPUCOP ProbabilisticRisk Assessment This step involves obtaining plant experience data for river water temperature and torus water temperature and level and performing statistical analysis to determine the probabilities of exceedance.
2.2.4 Manipulation And Quantification of BFN Unit 1 RISKMAN PRA Models This task is to make the necessary modifications to the BFN Unit 1 RISKMAN-based PRA models to simulate the loss of low pressure ECCS pumps during PRA Large LOCA scenarios due to inadequate NPSH caused by an unisolated containment and other extreme plant conditions (e.g., high service water temperature).
All large LOCA initiated sequences in the BFN PRA are modified as appropriate (except ISLOCAs and LOCAs outside containment, because these LOCAs result in deposition of decay heat directly outside the containment and not into the suppression pool). This approach to manipulating only LLOCA scenarios is to mirror the DBA accident calculations requiring COP credit. This is consistent with the ACRS observations during the December 2005 Vermont Yankee EPU COP hearings, in which the ACRS commented that they did not prefer the approach of assigning COP credit to all accident sequence types in the PRA simply for the sake of conservatism.
The modeling and quantification is performed consistent with common RISKMAN modeling techniques.
2.2.5 Comparison to ACDF and ALERF RG 1.174 Acceptance Guidelines The revised BFN Unit 1 PRA models are quantified to determine CDF and LERF. The difference in CDF and LERF between the revised model of this assessment and the BFN Unit 1 PRA base results are then compared toethe RG 1.174 risk acceptance guidelines. The RG 1.174 ACDF and ALERF risk acceptance guidelines are summarized in Figures 2-1 and 2-2, respectively. The boundaries between regions are 2-5 C1320503-6924R1 - 322/2006
BFNEPUCOPProbabilisticRiskAssessment not necessarily interpreted by the NRC as definitive lines that determine the acceptance or non-acceptance of proposed license amendment requests; however, increasing delta risk is associated with increasing regulatory scrutiny and expectations of compensatory actions and other related risk mitigation strategies.
2.2.6 Performance of Uncertainty and Sensitivity Analyses To provide context to the variability of the calculated deltaCDF and deltaLERF results, a parametric uncertainty analysis was performed using the RISKMAN software.
2.2.7 Assessment of "Large Late" Release Impact This task is to perform an assessment of the EPU COP credit impact on BFN Unil: 1 PRA "Large Late" radionuclide releases. This task is performed because the ACRS questioned Entergy on this issue during the recent Vermont Yankee EPU ACRS hearings in December 2005.
This aspect of the analysis is for additional information, and does not directly correspond to the RG 1.174 risk acceptance guidelines shown in Figures 2-1 and 2-2.
2.2.8 Review of BFN Unit 2 and Unit 3 PRAs The base analysis uses the BFN Unit I PRA models. This task involves reviewing the BFN Unit 2 and BFN Unit 3 RISKMAN PRA models and associated documentation to determine whether the analysis performed for BFN Unit 1 is also applicable to Unit 2 and Unit 3.
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BFNEPUCOPProbabilisticRiskAssessment Figure 2-1 RG 1.174 CDF RISK ACCEPTANCE GUIDELINES
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BFNEPUCOPProbabilisticRiskAssessment Figure 2-2 RG 1.174 LERF RISK ACCEPTANCE GUIDELINES
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BFNEPUCOP ProbabilisticRiskAssessment Section 3 ANALYSIS This section highlights the major qualitative and quantitative analytic steps to the analysis.
3.1 ASSESSMENT OF DBA CALCULATIONS The purpose of this risk assessment is due to the fact that the conservative nature of design basis accident calculations result in the need to credit COP in determining adequate low pressure ECCS pump NPSH. Use of more realistic inputs in such calculations shows that no credit for COP is required.
The GE DBA LOCA calculation makes the following conservative assumptions, among others, regarding initial plant configuration and operation characteristics:
I
- Initial reactor power level at 102% EPU
- Decay heat defined by 2 sigma uncertainty
- River water temperature at 950F
- Initial suppression pool temperature at 950 F a Initial SP water volume at minimum technical specification level
- No credit for containment heat sinks The GE DBA LOCA calculations were reviewed and the following input parameters were identified as those with a potential to significantly impact the DBA analytic conclusions regarding the need for COP credit in NPSH determination:
Initial reactor power level 3-1 C1320503-6924R1 -3/22roo I
BENEPUCOPProbabilisticRisk Assessment Decay heat Number of RHR pumps and heat exchangers in SPC River water temperature Initial suppression pool temperature
- RHR heat exchanger effectiveness
- Initial suppression pool water volume
- Credit for containment heat sinks Based on knowledge of the calculations, other inputs such as initial containment air temperature and humidity, have non-significant impacts on the results.
It is recognized that there are numerous different combinations of more realistic calculation inputs that show that COP credit is not necessary for maintenance of low pressure ECCS pump NPSH. To simplify the risk assessment, the different combinations of realistic input sensitivities were maintained at a manageable number.
A number of sensitivity calculations were performed to identify key input parameters for use in this risk assessment. The results of these calculations are shown in Table 3-1 (the shaded cells show those parameters that changed from the base DBA LOCA calculation). [3]
From the results of the sensitivity cases summarized in Table 3-1, the following general conclusions can be made:
- Initial reactor power, decay heat level, initial water temperatures, suppression pool volume, and the number of RHR pumps/HXs in operation are the key determining factors in the analytic conclusions.
These factors are evaluated in this risk assessment.
- RHR heat exchanger effectiveness and credit for containment heat sinks also influence the results, but to manage the6 risk calculation, this assessment takes no probabilistic credit for these issues.
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BFNEPUCOP ProbabilisticRiskAssessment
- COP credit is not required for NPSH, even with the conservative DBA calculation inputs, if 4 RHR pumps and associated heat exchangers are in operation (refer to Case 1 in Table 3-1).
- COP credit is not required forNPSH when 3 RHR pumps are in operation event with conservative 102% EPU power and 2 sigma decay heat assumptions, and conservative water temperature and SP volume assumptions (refer to Case 1d in Table 3-1).
- If the plant is operating at an unexpected 102% EPU initial power level with an assumed 2 sigma decay heat, only i2 RHR pumps and heat exchangers are placed in SPC operation, initial SP volume at 123,500 ft3, and river water temperature is at 680 F, then torus water temperature must be above 870 F to result in the need for COP credit (refer to Case 2f in Table 3-1).
- If the plant is operating at the expected nominal 100% EPU initial power level (2 sigma decay heat not assumed), only 2 RHR pumps and heat exchangers are placed in SPIC operation, initial SP volume at 123,500 ft3, and river water temperature is at 850F, then torus water initial temperature must be above 860 F to result in the need for COP credit (refer to Case 4i in Table 3-1).
The analytic conclusions are used in this risk assessment to define two plant states that will result in failure of low pressure ECCS pumps on inadequate NPSH during large LOC.As if the containment is unisolated:
- Plant State 1: 102% EPU initial power level, 2 sigma decay heat, 2 RHR pumps and heat exchangers~in SPC, initial SP volume at 123,500 ft3, river water temperature of 680F, and torus water initial temperature above 87 0F.
- Plant State 2: 100% EPU initial power level, nominal decay heat, 2 RHR pumps and hleat exchangers 'in SPC, initial SP volume at 123,500 ft3, river water temperature of 861F,: and river water initial temperature above 850F.
These two plant states are used in this risk assessment to model the LLOCA scenarios that can result in loss of low pressure ECCS pumps due to inadequate NPSH when the containment is unisolated. The probability of being in Plant State 1 or Plant State 2 is discussed below in Section 3.2.
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BFNEPUCOPProbabilisticRiskAssessment Scenarios with 3 or 4 RHR pumps and heat exchangers are not explicitly incorporated into the base case quantification because the risk contribution from such scenarios is non-significant (refer to Section 4.2.2).
3.2 PROBABILITY OF PLANT STATE 1 AND PLANT STATE 2 This section discusses the estimation of the probability of being in Plant State 1 or Plant State 2. This assessment is based on the statistical analysis of BFN experience data.
Refer to Appendix C for the statistical analysis of variations in BFN river water temperature and torus water temperature and level.
3.2.1 Probability of Plant State 1 The probability of being in Plant State I is determined as follows:
- The probability of being at 102% EPU power at the time of the postulated DBA LOCA is modeled as a miscalibration error of an instrument
- If such a miscalibration error occurs, it is assumed that the plant will be operating at 102% and that the operator does not notice other differing plant indications that would cause the operator to re-evaluate the plant condition
- If the plant is operating at 102% power, the decay heat level defined by 2 sigma uncertainty is assumed to occur with a probability of 1.0 (this conservative assumption is to simplify the analysis).
- The probability of river water temperature greater than 680 F is determined from the BFN experience data statistical analysis summarized in Appendix C.
- Given river water temperature 680 F, the conditional probability that the torus water temperature is 870 F is determined from the BFN experience data statistical analysis summarized in Appendix C.
3-4 C1320503-924R1 - 3/22/2006 l
BFNEPUCOPProbabilisticRiskAssessment The probability that suppression pool water level is less than 123,500 ft3 is also based on the BFN experience data statistical analysis summarized in Appendix C.
The probability of being at 102% power at the time of the accident is modeled as the likelihood of a miscalibrated instrument. Based on review of the pre-initiator human error probability calculations in the BFN Unit 1 PRA Human Reliability Analysis, this risk assessment assumes a nominal human error probability of 5E-3 for miscalibration of an instrument. As such, the probability of being at 102% power at t=0 is taken in this analysis to be 5E-3.
As can be seen from Table C-1, the probability of river water temperature being greater than 680 F at the time of the DBA LOCA is 5.64E-1. As discussed in Section C.2.1, the conditional probability that suppression pool temperature is greater than 870 F is 4.42E-
- 1. As can be seen from Table C-3, the probability of suppression pool water volume being below 123,500 ft3 at the time of the DBA LOCA is 1.45E-2.
Therefore, the probability of being in Plant State 1 at the time of the DBA LOCA is 5E-3 x 5.641--1 x 4.42E-1 x 1.45E-2 = 1.8E-5.
3.2.2 Probability of Plant State 2 The probability of being in Plant State 2 is determined as follows:
- The probability of being at 100% EPU power at the time of the postulated DBA LOCA is reasonably assumed to be 1.0
- The probability of river water temperature greater than 85TF is determined from the BFN experience data statistical analysis summarized in Appendix C.
- Given river water temperature of 85 0F, the conditional probability that the torus water temperature is 870 F, is taken to be 1.0. This is reasonable (refer to Figure C-1).
3-5 C1320503-6924R1 - 3/22/2006 I
BFNEPUCOPProbabilisticRisk Assessment The probability that suppression pool water level is less than 123,500 ft3 is also based on the BFN experience data statistical analysis summarized in Appendix C.
As can be seen from Table C-1, the probability of river water temperature being greater than 850F at the time of the DBA LOCA is 1.64E-1. As can be seen from Table C-3, the probability of suppression pool water volume being below 123,500 ft3 at the time of the DBA LOCA is 1.45E-2.
Therefore, the probability of being in Plant State 2 at the time of the DBA LOCA is 1.641_-1 x 1.0 x 1.45E-2 = 2.4E-3.
3.3 PRE-EXISTING CONTAINMENT FAILURE PROBABILITY As discussed in Section 2, the approach to this input parameter calculation follows the EPRI guidelines regarding calculation of pre-existing containment leakage probabilities in support of integrated leak rate test (ILRT) frequency extension LARs (i.e., EFRI Report 1009325, Risk Impact of Extended Integrated Leak Rate Testing Intervals, 12/03). [2]
This assessment is provided in Appendix B of this report. As discussed in Appendix B, a pre-existing unisolable containment leakage path of 20La is assumed in the base case quantification of this risk assessment to result in defeating the necessary COP credit. As can be seen from Table B-1, the probability of a 20La pre-existing containment leakage at any given time at power is 1.88E-03.
This low likelihood of a significant pre-existing containment leakage path is consistent with BFN primary containment performance experience. The BFN primary containment performance experience shows BFN containment leakages much less than 20La. Per 3-6 C1320503-6924R1 - 3/22/1006
BFNEPUCOP ProbabilisticRisk Assessment Reference [1], the BFN Unit 2 and Unit 3 primary containment ILRT results from tie most recent tests are as follows:
Containment Leakage Unit Test Date (Fraction of La) 2 11/06/94 0.1750 2 03/17/91 0.1254 3 10/10/98 0.1482 3 11/06/95 0.4614 Although the above results are for Units 2 and Units 3, given the similarity in plant design and operation and maintenance practices, the results are reasonably judged to be reflective of BFN Unit 1, as well.
Sensitivity studies to the base case quantification (refer to Section 4) assess the sensitivity of the results to the pre-existing leakage size assumption.
3.4 MODIFICATIONS TO BFN UNIT 1 PRA MODELS As discussed in Section 2, all large LOCA initiated sequences in the BFN PRA are modified as appropriate (except ISLOCAs and LOCAs outside containment, because these LOCAs result in deposition of decay heat directly outside the containment and not into the suppression pool). The following Large LOCA initiated sequences in the BFN Unit 1 PRA were modified:
- Large LOCA - Loop I Core Spray Line Break (LLCA)
- Large LOCA - Loop II Core Spray Line Break (LLCB)
- Large LOCA - Loop A Recirc. Discharge Line Break (LLDA)
- Large LOCA - Loop B Recirc. Discharge Line Break (LLDB)
- Large LOCA - Loop A Recirc. Suction Line Break (LLSA)
- Large LOCA - Loop B Recirc. Suction Line Break (LLSB)
- Other Large LOCA (LLO) 3-7 C1320503-6924R1 - 322/2006 l
BFNEPUCOPProbabilisticRiskAssessment The accident sequence modeling for the above LLOCA initiators was modified as follows:
. A top event for loss of containment integrity (CIL) was added to the beginning of the Level 1 event tree structures
. A top event modeling the additional Plant State pre-conditions (NPSH) was added to the beginning of the Level 1 event tree structures, right after the CIL top event.
. If top events CIL and NPSH are satisfied (i.e., occur), then the RHR pumps and CS pumps are directly failed Refer to Appendix E for print-outs of the revised large LOCA event trees.
The CIL top event is quantified using a fault tree. The fault tree is a modified version of the existing BFN Unit 1 Level 2 PRA containment isolation fault tree. The BFN Unit 1 Level 2 PRA containment isolation fault tree models failure of the containment isolation system on demand given an accident signal. Hardware, power and signal failures for all primary containment penetrations greater than 3" diameter are modeled in the fault tree.
To this fault tree structure was added the probability of a pre-existing containment leak size of 2OLa. Refer to Appendix F for a print-out of the containment isolation fault tree used in this analysis for the CIL node in the large LOCA event trees.
The NPSH top event is also quantified using a fault tree. The NOSH incorporates the fault tree logic to model the probability of being in Plant State 1 or Plant State 2. Refer to Appendix F for a print-out of the fault tree used in this analysis for the NPSH node in the Large LOCA event trees.
The quantification of the revised model was performed to produce the new CDF. All the new CDF scenarios are those in which the containment is unisolated at t=O, all RPV injection is lost early, and core damage occurs at approximately one hour. As such, the additional CDF contributions created by this model manipulation are also all LERF 3-8 C1320503-6924R1 -3/22J2006 l
BFNEPUCOPProbabilisticRisk Assessmcnt release sequences (i.e., deltaCDF equals deltaLERF). This is a conservative assumption as it assumes that the pre-existing containment leakage of 2OLa used in the base quantification is representative of a LERF release. Reference [2] determines that a containment leak representative of LERF is >600La.
The quantification results and uncertainty and sensitivity analyses are discussed in Section 4.
The revised BFN Unit 1 PRA RISKMAN model for this base case analysis is archived in file UICOP-H and saved on the BFN computers along with the other BFN PRA RISKMAN models.
3.5 ASSESSMENT OF LARGE-LATE RELEASES As discussed above in Section 3.3, all the deltaCDF resulting from this risk assessment also results directly in LERF. As such, there is no increase in Large-Late releases due to scenarios modeling in this risk assessment. Refer to Appendix D for more discussion.
3-9 3-9 C1320503-6924Rl - 3/22/2006
BFNEPU COP ProbabilisticRisk Assessment Table 3-1
SUMMARY
OF COP DETERMINISTIC CALCULATIONS(5) 0 !4) 4
- 4) -4)t E S _
- (I E c 4)
(0 0 -- 0.
- orE CLa (D -6 5 iR 5 ID 0) - ) ... -o a. 64CD . 4 Base Case( )2 Cae' aeeain Case('-Case-escrition-EPU Licensing Calculation -
.E C 102% ANSI 5.1
~
C 9
coO. z c-c Full li z aL
.~
n: ~ELo
._i.
(GE) DBA LOCA EPU wr2a5 95 2 2 2 4000 223 2 1121,500 Yes No 187.3 Ye, design Case 1(2) DBA Calculation but No 102% ANSI 5.1 95 95 Full (GE) Single Failure EPU w/2c design 4000 223 121,500 Yes I No 166.4 l No 21 Case 1a() DBA Calculation but 3 RHR 102% ANSI 5.1 (GE) Pun-ps inSuppression Pool Full EPIJ w/2 95 95 4000 121,500 Yes No 175.0 Ye_
Cooling design Case 1a (TVA) DBA Calculation but 3 RHR ANSI
[This case is 102% Full Pumps inSuppression Pool 5.1 95 95 4000 121,500 Yes No 175.0 Ye benchmarked - Cooling ---- . EPU, -wl2a design against Case la (GE)}
Case lb 100%Initial PowerRHRSW (TVA) 890F, 3 Pumprs in Full Suppression Pool Cooling, K 95 4000 Yes No 171.0 Nc design Value 225,4 CS Pumps Case Ic 100%Initial Power, RHRSW (TVA) 900F, 3 Pumps in Suppression Pool Cooling, K Full 95 4000 Yes No 170.5 No Value 225,4 CS Pumps, design Nominal SP WL 3-10 C1320503-6924R1 - 3'22/2006 l
BFN EPU COP ProbabilisticRisk Assessment Table 3-1
SUMMARY
OF COP DETERMINISTIC CALCULATIONS(5)
F - - - - p- p- - p p- p- - - . - U - M*E I,,
C 0. 2 e m. 0 0 W 4) vE E E
~2 4e F0
,e E 0 . w
- 4) 4 0.
toM 0
- C C
=2 0., 4) C
= a E~ -5 2 U -a 0C o0 iE wC Q L L~ Q 0 4)
I 4) 4 E oa = c.) E 0 a-0 4 0E)~ C. )a I- E W 8 ml go4)
.0 ' Zi.~ 13D.a X0 U0~
(L U) CL co 4)
.2e 0
- 84) E *' 0 4 CaseC') Case Description
-a a Co 9 4l) I.
Y Y1 = 0 00 CO4-w 6 8
.1 Aa LO) a Case Id DBA Calculation but RHRSW (TVA) 90°F, SP Initial Temp 91F, 3 102% ANSI
_ Full Pumps in Suppression Pool EPU 5.1 w/ 4000 121,5M Yes No 171.0 No Cooling, KValue 225, 4 CS 2a design Pumps Case le DBA Calculation but RHRSW (TVA) 920F, SP Initial Temp 90°F, 3 102% ANSI Full Pumps in Suppression Pool EPU 5.1w/ 4000 121,5C0 Yes No 171.1(4 No design Coobng, K Value 225, 4 CS - 2 Pumps_ _ _ _
Case 2 DBA Calculation but Initial 102% ANSI 5.1 Full (GE) SW Temperature= 85°F EPU w/2a 2 2 4000 223 2 121,500 Yes No 182.0 Yel design Case 2 (TVA) AS
[This case is DBA Calculation but SW 102% ANSI 95 2 Full design 2 2 4000 223 2 121,500 Yes No 182.2 Ye_
benchmarked Temperature = 850F EPU 5.1 wI2a against Case 2 Case 2a DBA Calculation but Initial 102% ANSI 5.1 (GE) SW Temperature = 75°F EPU w/2a 95 2 design 2 2 4000 223 2 121,500 Yes No 177.6 Ye _
Case 2b(2) DBA Calculation but Initial 102% ANSI 5.1 95 2 Full 2 2 4000 223 2 121,500 Yes No 175.9 Ye!
(GE) SW Temperature = 70°F EPU wf2a design Case 2c DBA Calculation but Initial 102% ANSI 5.1 Full (GE) SW Temperature = 65°F design 2 2 4000 223 2 121,500 Yes No 174.3 Ye~
EPU wl2a 3-11 C1320503-6924R1 -3122/2006 I
BEN EPUCOP ProbabilisticRiskAssessment Table 3-1
SUMMARY
OF COP DETERMINISTIC CALCULATIONS(5)
I q - q- q- I- - - I- I- I- I- Y- I- I - I- I - Y- in 0,t 2
- e E 0 4, 0 0 ID
.5 S z0 0 4,
CdC C 1%
- a. 0 y8.dC 0 0 CL E W0.
E 4,
Co C 0Q.
EL -
8, -8 0.
Er 31:E SC ' o 0. _i 0 a.
=0.
C.
I 0 E FE:d C =
I 0Z 3cm
.Z .s .2a 2¢ m
~0~ 8M 4,
ni I E 2: W0 cf) . . to 80 .0 0 _b I
~ Oa 4, . CL
= O. a: -,; 0. 0u_ (I)
Case') - -Case Description - -O -3 ci) - ' = 0 z .2 L,00 W J 0L Case 2d DBA Calculation but SW - -- .
(TVA) Temperature = 65oF, SP 102% ANS Initial Temp 880F, Nominal EPU 5.1 2 Full 2 2 4000 223 2 Yes No 170.6 No design SPWL wr2a Case 2e DBA Calculation but SW 102% ANSI (TVA) Temperature = 65oF, SP EPU 5.1 - 2 Full 2 2 4000 223 2 Yes No Initial Temp 87oF .- design 121,500 170.7 No w2a Case 2f DBA Calculation but SW -- -
(1VA) Temperature~-68oF, SP -
Initial Temp 870F, Nominal 2 deuig 2 2 4000 223 2 Yes No 171.1(W No SP WL design Case 3 DBA Calculation but Initial -- FUR (GE) SP Temperature=85 0F__ -2 design 2 2 4000 223 2 121,500 Yes No 183.8 Ye!
Case 4 100%1nitial Power, Minimum (GE) SP Level, and No Heat Sink 2 dFsign 2 2 4000 2 121,50 Yes No 177.0 Ye{
Credit Case 4 (TVA) 100% Initial Power, Minimum
[This case is SP Level, and No Heat Sink 2 desig 2 2 4000 2 121,500 Yes No 177.1 Ye bench-marked Credit against Case 4 Case 4a 100% Initial Power, Nominal 1
2 Full 2 2 40 (GE) SP Level, and Heat Sink 2 design 2 Yes Cc t 2 2 40 174.7 Yel I I MI - I i 11 3-12 C1320503-6924R1 - 3/22/2006 l
BF7 EPUCOP ProbabilisticRiskAssessment Table 3-1
SUMMARY
OF COP DETERMINISTIC CALCULATIONS(5)
. - S- I - I S S- S - S - S - S- *- S __ ..
0 4)
M 4)
C 0 El ta 0
V 2 0
Ea.
4) 4 0 E, 4) 0 a . S K CC E
4)1:4 c M 0
CL 4)
- 4) 9I0 CL E
8 e8 E B
Bc 0 4) ao 0 0 C at1 Ea- E
- 4) .= 4) 721 0 CL 604 9i .2 C 40 a R 0
04E i0 o 0)co a2-S 'ck 4) c0 W0L .gg 2. -
-e
. r_
- a. O0 =
6 .!2 W -a 4)
UJ CO . .L.
__U i
a-Ci)
E 4)
. C I .L-V~ 8 -V ZoF 8 W CO
- 4) 0..r. 0) -0 (L Case(') Case Description 0 0 c '. z D
zd Z oS ct:>
00t CO~
6w...j c F
-a-o 3:
Case 4b( 2) 100% Initial Power, Minimum -
(GE) SP Level, and Heat Sink 2 Full 2 2 4000 2 121,500 Yes 178.9 Credit design Case 4c(2) 100% Initial Power, Minimum (GE) SP Level, Heat Sink Credit, and SW Temp. that results in 2 Full 2 2 4000 2 121,500 Yes 2 design 2 2 4000 175.8 Yei Peak SP Temp. equal to/less than 176°F Case 4d 100% Initial Power, RHRSW (VA) 860F, SP Initial Temp 92oF, K Full 2 design 2 2 4000 2 121,500 Yes No 177.0 Ye~
Value 225 Case 4e 100% Initial Power, RHRSW _ . _
(TVA) 865F, SP Initial Temp 90°F, K 2 Full 2 2 wd 2 Yes 176.1 121,500 No Yei Value 225 Case 4f 100% Initial Power, RHRSW (VA) 860F, SP Initial Temp 9OoF, K 2 - Yes No 175.6 Ye _
Value 225, Nominal SP WL Case 4g 100% Initial Power, RHRSW 2 design 2 2 40 (VA) 860F, SP Initial Temp 90°F, K 2 Fuld 2 2 4000 2 Yes No 173.1 Yec Value 241, Nominal SP WL- -
2 desin 2 2 4000 Case 4h 100% Initial Power, RHRSW (VA) 850F, SP Initial Temp 90°F, K 2 design 2 2 40 2 Yes No 175.1 Ye Value 225, Nominal SP WL 3-13 C1320503-6924R1 - 3/22/2006
BFNEPUCOP ProbabilisticRiskAssessment Table 3-1
SUMMARY
OF COP DETERMINISTIC CALCULATIONS(5)
I U p p p p p p P I I I U P i I 0
.5 0 0 0 y¢ 0 0 E C
~~co El a
.5 r)C CL aL.
E a. Z a. fit C,
be C
.S c 2 cm is .S a, CL
- a. 0S C
C CL
.- g r o .5 O
- n. 0 F o0. E (00. I 0) 0 E VE0a Q a- .c, 0 u, a 4. E la00 5 0) 2 F _
0 C 4) m a. .0 = iE = a. e i= 1,ED 8 .
.L 0 ceg- 0 va . -- JE. m1 a: m
/) .'
a) 20
= ak. I .c = 0' c 0 Case(') Case Desciption z 0 aW W Z .Q tL w -J 0 =W a.2 Case 4i 100% Initial Power, RHRSW (1VA) 85oF, SP Initial Terrp 86SF, K 2 Full 2 2 4000 2 Yes No 170.8 No
-Value 241; Nominal SP WL- design Case , 100% Initial Power, RHRSW (VA) 850F, SP Iniial Temp 88oF, K 2 Full 2 2 4000 2 Yes No 171.0 No Value 241, Nominal SP WL design Notes to Table 3-1:
(1) Column information includes designation of organization that performed the calculation.
(2) -Case verified by formal analysis. --------- - --- - -
(3) COP credit required for peak suppression pool temperature of 171°F.
(4) This value is acceptable for demonstrating sensitivity analysis results.
(5) Shaded areas in the table "highlight' differences from the Base Case.
3-14 C1320503-6924R1 - 32212006 I
BFN EPU COPProbabilisticRisk Assessment Section 4 RESULTS 4.1 QUANTITATIVE RESULTS The results of the base quantification of this risk assessment case are as follows:
. deltaCDF: 1.4E-9/yr
. deltaLERF: 1.4E-9/yr As discussed in Section 3, the additional CDF contributions created by this model manipulation are also all LERF release sequences (i.e., deltaCDF equals deltaLERF).
These very low results are expected and are well within the RG 1.174 guidelines (re er to Figures 2-1 and 2-2) for "very small" risk impact. If greater detail was included to address some of the conservative assumptions in this risk assessment (e.g., 2 sigma decay heat assumed with a probability of 1.0 given 102% EPU power exists; refer to Section 3.2), the deltaCDF and deltaLERF would be even lower.
4.2 UNCERTAINTY ANALYSIS To provide additional information for the decision making process, the risk assessment provided here is supplemented by parametric uncertainty analysis and quantitative and qualitative sensitivity studies to assess the sensitivity of the calculated risk results.
Uncertainty is categorized here into the following three types, consistent with PRA industry literature:
- Parametric
. Modeling
- Completeness 4-1 C1320503-6924R1 - 3/22)2206 l
BFNEPUCOP ProbabilisticRiskAssessnmnt Parametric uncertainties are those related to the values of the fundamental parameters of the PRA model, such as equipment failure rates, initiating event frequencies, and human error probabilities. Typical of standard industry practices, the parametric uncertainty aspect is assessed here by performing a Monte Carlo parametric uncertainty propagation analysis. Probability distributions are assigned to each parameter value, and a Monte Carlo sampling code is used to sample each parameter and propagate the parametric distributions through to the final results. The parametric uncertainty analysis and associated results are discussed further below.
Modeling uncertainty is focused on the structure and assumptions inherent in the risk model. The structure of mathematical models used to represent scenarios and phenomena of interest is a source of uncertainty, due to the fact that models are a simplified representation of a real-world system. Model uncertainty is addressed here by the identification and quantification of focused sensitivity studies. The model uncertainty analysis and associated results are discussed further below.
Completeness uncertainty is primarily concerned with scope limitations. Scope limitations are addressed here by the qualitative assessment of the impact on the conclusions if external events and shutdown risk contributors are also considered. The completeness uncertainty analysis is discussed further below.
4.2.1 Parametric Uncertainty Analysis The parametric uncertainty analysis for this risk assessment was performed using the RISKMAN computer program to calculate probability distributions and determine the uncertainty in the accident frequency estimate.
RISKMAN has three analysis modules: Data Analysis Module, System Analysis Module, and lEvent Tree Analysis Module. Appropriate probability distributions for each uncertain 4-2 C1320503-6924R1 - 3/22/2006 I
BFNEPUCOP ProbabilisticRisk Assessment parameter in the analysis is determined and included in the Data Module. The System Module combines the individual failure rates, maintenance, and common cause parameters into the split fraction frequencies that will be used by the Event Tree Module. A Monte Carlo routine is used with the complete distributions to calculate the split fraction frequencies. Event trees are quantified and linked together in the Event Module. The important sequences from the results of the Event Tree Module are used in another Monte Carlo sampling step to propagate the split fraction uncertainties and obtain the uncertainties in the overall results.
The descriptive statistics calculated by RISKMAN for the total core damage frequency of the plant caused by internal events include:
- Mean of the sample
- Variance of the sample
- 5th, 50th, and 95th percentiles of the sample The parametric uncertainty associated with delta core damage frequency calculated in this assessment is presented as a comparison of the RISKMAN calculated CDF uncertainty statistics for the two cases (i.e., the Unit I base EPU PRA and the EPU COP Credit base case quantification). The results are shown in Table 4-1. Table 4-1 summarizes the CDF uncertainty distribution statistics for the BFN Unit 1 PRA and for the COP credit base quantification.
It should be cautioned that this distribution is developed via Monte Carlo (random) sampling, and as such it is dependent upon the number of samples and the initial numerical seed values of the sampling routine. Neither the initial seeds nor the number of samples used for the model of record are known. Consequently, some variation from the base model statistics is expected. Taking these cautions into consideration, a comparison of the distributions by percentiles shows little if any change.
4-3 C1320503-6924R1 - 3/22/2006 I
BFNEPUCOP ProbabilisticRiskAssessment 4.2.2 Modeling Uncertainty Analysis As stated previously, modeling uncertainty is concerned with the sensitivity of the results due to uncertainties in the structure and assumptions in the logic model.
Modeling uncertainty has not been explicitly treated in many PRAs, and is still an evolving area of analysis. The PRA industry is currently investigating methods for performing modeling uncertainty analysis. EPRI has developed a guideline -for modeling uncertainty that is still in draft form and undergoing pilot testing. The EFRI approach that is currently being tested takes the rational approach of identifying key sources of modeling uncertainty and then performing appropriate sensitivity calculations. This approach is taken here.
The modeling issues selected here for assessment are those related to the risk assessment of the containment overpressure credit. This assessment does not involve investigating modeling uncertainty with regard to the overall BFN PRA. The modeling issues identified for sensitivity analysis are:
- Pre-existing containment leakage size and associated probability
- Calculation of containment isolation system failure
- Assessment of power and water temperature and level pre-conditions
- Number of RHR pumps and heat exchangers in SPC Pre-Existing Containment Leakage Size/Probability The base case analysis assumes a pre-existing containment leakage pathway leakage size of 2OLa that would result in defeat of the necessary containment: overpressure credit during a DBA LOCA.
4-4 C1320503-6924R1 - 3/22/00 I
BFNEPUCOPProbabilisticRiskAssessment A larger pre-existing leak size of 10OLa, consistent with the EPRI 1009325 recommended assumption for a "large" leak, is used in this sensitivity to defeat the necessary COP credit. From EPRI 1009325, the probability of a pre-existing 100La containment leakage pathway at any given time at power is 2.47E-04.
Calculation of Containment Isolation System Failure The base case quantification uses the containment isolation system failure fault tree logic to represent failure of the containment isolation system. The fault tree specifically analyzes primary containment penetrations greater than 3" diameter. This modeling sensitivity case expands the scope of the containment isolation fault tree to include smaller lines as potential defeats of COP credit. This sensitivity is performed by increasing by a factor of 10 the failure probability associated with the containment isolation system.
Assessment of Power and Water Temperature and Level Pre-conditions This is a conservative sensitivity that assumes that all that is necessary for failure of the low pressure ECCS pumps due to inadequate NPSH during a large LOCA is an unisolated containment. This sensitivity is performed by assuming the other pre-conditions represented by the top event NSPH exist with a probability of 1.0.
Number of RHR pumps and heat exchangers in SPC The base case COP credit quantification addresses the situation in which 2 or less RHR pumps and heat exchangers are operating in SPC mode. The likelihood of failing any two RHR pumps during the 24-hr PRA mission time is approximately 8.2E-3. The likelihood of an unisolated containment given an accident initiator is approximately 2.2E-3, and the likelihood of other necessary extreme plant conditions (e.g., high river temperature, high reactor power, reduced suppression pool water level) existing at the 4-5 C1320503-6924R1 -3/22/2006 I
BFNEPUCOP ProbabilisticRiskAssessmont time of the LLOCA is approximately 2.4E-3. As such, the base quantification results in an approximate 4.3E-8 conditional probability, given a LLOCA, of loss of low pressure ECCS pumps due to insufficient NPSH due to inadequate COP.
This sensitivity discusses the risk impact of also explicitly quantifying scenarios with only I or no RHR pumps failed. Such scenarios are not explicitly included in the base quantification because their risk contribution is non-significant, as shown by the sensitivities discussed here. As shown in Table 3-1, even with very conservative assumptions, if 3 or more RHR pumps and heat exchangers are operating in SPC, there is no need for containment overpressure. To result in a need for COP credit in such cases would require even more conservative input assumptions than the 2 RHR pump scenario. As such, the additional risk from such scenarios is non-significant compared to the 2 RHR pump case explicitly modeled in this analysis.
An estimate of the deltaCDF risk contribution for the scenario with 3 RHR pumps in SPC operation can be approximated as follows (refer to Case Id in Table 3-1):
- Likelihood of failure of 1 RHR pump or 1 RHR heat exchanger during the 24-hr PRA mission time: 1.OOE-2 (nominal estimate)
- Probability of containment isolation failure given an accident initiator: 3E-3 (nominal from base analysis)
- Probability of river water temperature >90OF at any given time: 9E-2 (nominal value based on Table C-1. Although the river temperature has not exceeded 900 F based on the collected plant data, statistically there is a non-zero likelihood of such a temperature).
- Conditional probability that suppression pool water temperature > 91 0F given river water temperature > 900 F: 1.0 (refer to Figure C-1).
- No probabilistic credit for low suppression pool volume or low heat exchanger effectiveness is taken here.
4-6 4-6C1320503-6924Rl -3122/2X06I
BEN EPU COPProbabilisticRisk Assessment deltaCDF contribution for 3 RHR pump case: 3E-5 x 1E-2 x 5E-3 x 3E-3 x 9E-2 x 1.0 = - 4E-1 3/yr This additional contribution to the calculated deltaCDF from a 3 RHR pump case is non-significant in comparison to the 2 RHR pump case.
An estimate of the deltaCDF risk contribution for the scenario with 4 RHR pumps in operation can be approximated as follows (refer to Case 1 of Table 3-1):
- Probability of containment isolation failure given an accident initiator: 3E-3 (nominal from base analysis)
- Probability of river water temperature > 100IF at any given time: 1E-3 (estimate based on Table C-1. Although the river temperature has not exceeded 900 F based on the collected plant data, statistically there is a non-zero likelihood of such a temperature). 100OF is assumed here as the river water temperature at which COP credit is required (refer to Case 1 of Table 3-1).
- Conditional probability that suppression pool water temperature > 950 F given river water temperature > 100F: 1.0 (refer to Figure C-1).
. No probabilistic credit for low suppression pool volume or low heat exchanger effectiveness is taken here.
- deltaCDF contribution for 3 RHR pump case: 3.1 E-5 x 1.0 x 5E-3 x 3E-3 x I E-3 x 1.0 = -5E-1 3/yr Similar to the 3 pump case discussed previously, this additional contribution to the calculated deltaCDF from a 4 RHR pump case is non-significant in comparison to the 2 RHR pump case.
4-7 C1320503-6924R1 -3/22/2006 I
BFNEPUCOP ProbabilisticRiskAssessmmt Summary of Modeling UncertaintV Results I The modeling uncertainty sensitivity cases are summarized in Table 4-2.
4.2.3 Completeness Uncertainty Analysis As stated previously, completeness uncertainty is addressed here by the qualitative assessment of the impact on the conclusions if external events and shutdown risk contributors are also considered.
4-8 C1320503-6924R1 -3/22/2006 l
BFNEPUCOPProbabilisticRisk Assessment Table 4-1 PARAMETRIC UNCERTAINTY ANALYSIS RESULTS StaisticBFN Unit I Base CDF COP Risk Assessment 5% 4.71 E-7 5.15E-7 50% 1.23E-6 1.23E-6 MEAN 1.77E-6 1.77E-6 95% 1 4.72E-6 4.47E-6 4-9 C1320503-6924R1 -3/22/2006 I
BFNEPU COPProbabilisticRiskAssessment Table 4-2
SUMMARY
OF SENSITIVITY QUANTIFICATIONS Case Description CDF LERF ACDF2 ALERFm l Base(') Base Case Quantification (20 La leak size) 1.768E-6 4.411 E-7 1.4E-9 1.4E-9 l 1() Pre-Existing Containment Leakage Sufficient to Fail COP Credit 1.768E-6 4.411 E-7 1.4E-9 1.4E-9 Defined by 1OOLa 2(a) -Assume Low Suppression Pool Water Volume (123,500 ft3) Exists 1.768E-6 4.413E-7 1.6E-9 1.6E-9 100% of the Timel 3(1) Expansion of Containment Isolation fault tree to Encompass Smaller 1.768E-6 4.411 E-7 1.4E-9 1.4E-9 l Lines (approximate by multiplying Cont. Isol. failure probability by lOx) 4(1) Assume Initial Power Level and Water Temperature and Level Pre- 1.770E-6 4.432E-7 3.5E-9 3.5E-9 l Conditions Exist 100% of the Time 5(1) Combination of Cases #3 and #4 1.773E-6 4.463E-7 6.6E-9 6.6E-9 l 6 Incorporation of u3-RHR pumps in SPC" and M4-RHR pumps in SPC" 1.768E-6 4.411 E-7 1.4E-9 1.4E-9 l loss of NPSH scenarios Notes:
(1) Scenarios with failure of 2 or more RHR pumps and associated heat exchangers in SPC are explicitly analyzed in these cases. As shown in Case 6, explicit incorporation of scenarios with 0 or 1 RHR pumps in SPC failed has a negligible impact on the results.
(2) The ACDF and ALERF values are with respect to the BFN Unit 1 PRA model of record CDF of 1.767E-6/yr and LERF of 4.397E-7/yr.
I 4-10 C1320503-6924R1 - 3/22/2006 I
BFNEPUCOPProbabilisticRiskAssessment Seismic The BFN seismic risk analysis was performed as part of the Individual Plant Examination of External Events (IPEEE). BFN performed a seismic margins assessment (SMA) following the guidance of NUREG-1407 and EPRI NP-6041. The SMA is a deterministic evaluation process that does not calculate risk on a probabilistic basic. No core damage frequency sequences were quantified as part of the seismic risk evaluation.
The conclusions of the SMA are judged to be unaffected by the EPU or the containment overpressure credit issue. The EPU has little or no impact on the seismic qualifications of the systems, structures and components (SSCs). Specifically, the power uprate results in additional thermal energy stored in the RPV, but the additional blowdown loads on the RPV and containment given a coincident seismic event, are judged not to alter the results of the SMA.
The decrease in time available for operator actions, and the associated increases in calculated HEPs, is judged to have a non-significant impact on seismic-induced risk.
Industry BWR seismic PSAs have typically shown (e.g., Peach Bottom NUREG-1150 study; Limerick Generating Station Severe Accident Risk Assessment; NUREG/CR-4448) that seismic risk is overwhelmingly dominated by seismic induced equipment and structural failures. Seismic induced failures of containment are low likelihood scenarios, and such postulated scenarios are moot for the COP question because they would be analyzed in a seismic PRA as core damage scenarios directly.
Based on the above discussion, it is judged that seismic issues do not significantly impact the decision making for the BFN EPU and containment overpressure credit.
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BFNEPUCOPProbabilisticRiskAssessmcnt Internal Fires The BFN fire risk analysis was performed as part of the Individual Plant Examination of External Events (IPEEE). BFN performed a screening methodology using the EPRI FIVE (Fire Induced Vulnerability Evaluation) methodology.
Like most plants, BFN currently does not maintain a fire PRA. However, given the very low risk impact of the COP credit, even if fire risk was explicitly quantified the conclusions of this risk assessment are not expected to change, i.e., the risk impact is very small.
Other External Hazards In addition to seismic events and internal fires, the BFN IPEEE Submittal analyzed a variety of other external hazards:
- High Winds/Tomadoes
- External Floods
- Transportation and Nearby Facility Accidents
- Other External Hazards The BFN IPEEE analysis of high winds, tornadoes, external floods, transportation accidents, nearby facility accidents, and other external hazards was accomplished by reviewing the plant environs against regulatory requirements' regarding these hazards.
Based upon this review, it was concl ded that BFN meets the applicable NRC Standard Review Plan requirements and therefore has an acceptably low risk with respect to these hazards. As such, these other external hazards are judged not to significantly impact the decision making for the BFN EPU and containment overpressure credit.
4-12 4-2C1320503-6924R1 -3/2/, I
BFNEPUCOPProbabilisticRiskAssessment Shutdown Risk As discussed in the BFN EPU submittal, shutdown risk is a non-significant contributor to the risk profile of the proposed EPU. The credit for containment overpressure is not required for accident sequences occurring during shutdown. As such, shutdown risk does not influence the decision making for the BFN EPU containment overpressure credit.
4.3 APPLICABILITY TO BFN UNIT 2 AND UNIT 3 This risk assessment was performed using the BFN Unit 1 PRA. To assess the applicability of the Unit 1 results to BFN Units 2 and 3, the BFN Unit 3 PRA was reviewed. The Unit 3 PRA was explicitly reviewed because it has a higher base CIDF than the Unit 2 PRA due to fewer inter-unit crosstie capabilities than Unit 2.
Review of the Unit 3 PRA models did not identify any differences that would make the Unit 1 PRA results and conclusions not applicable to Units 2 and 3. As further evidence, the Unit 3 PRA was modified in a similar manner as the Unit 1 sensitivity Case #2 and quantified to determine the ACDF impact. The result for Unit 3 was a deltaCDF of 1.9E-9/yr. The revised BFN Unit 3 PRA RISKMAN model supporting this review is archived in file U3COP2-9 and saved on the BFN computers along with the other BFN PRA RISKMAN models.
Given the above, the results for the Unit 1 PRA risk assessment are comparable to the Units 2 and 3 PRAs.
The U2/U3 assessment discussed in this sub-section was performed for the Rev. 0 analysis. Given the similar results obtained in Rev. I analysis using the U-1 model, the U2/U3 assessment discussed above was not re-performed as the conclusion would be the same.
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BFNEPUCOPProbabilisticRiskAssessmcnt Section 5 CONCLUSIONS The report documents the risk impact of utilizing containment accident pressure (containment overpressure) to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps during DBA LOCAs.
The need for COP credit requests is driven by the conservative nature of design basis accident calculations. Use of more realistic inputs in such calculations shows that no credit for COP is required.
The conclusions of the plant internal events risk associated with this assessment are as follows.
- 1) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10Q/yr. Based on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in CDF (1.4E-09/yr).
- 2) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of Large Early Release Frequency (LERF) below 10-7/yr. Based on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in LERF (1.4E-09/yr).
These results are well within the guideline of RG 1.174 for a "very small" risk increase.
Even when modeling uncertainty and parametric uncertainty, and external event scenarios are considered, the risk increase is small. As such, the credit for COP in determining adequate NPSH for low pressure ECCS pumps during DBA LOCAs is acceptable from a risk perspective.
5-1 C1320503-6924R1 -3/22/2D06 l
BFNEPUCOPProbabilisticRiskAssessment The general conclusions that the risk impact from the COP credit for DBA LOCAs is very small, applies to BFN Unit 1 as well as BFN Units 2 and 3.
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BFNEPUCOP ProbabilisticRiskAssessmmnt REFERENCES
[1] "Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specificaticns (TS) Change 448 - One-Time Frequency Extension For Containment Integrated Leakage Rate Test (ILRT) Interval", TVA-BFN-TS-448, July 8, 2004.
[2] Risk Impact Assessment of Extended Intearated Leak Rate Testing Intervals, EPRI Report 1009325, Final Report, December2003.
[3] "Project Task Report - Browns Ferry Units 1, 2 & 3 EPU, RAI Response - NP'SH Sensitivity Studies", GE Nuclear Energy, GE-NE-0000-0050-00443-RO-Drmft, February 2006.
[4] Letter from G.B. Wallis (Chairman, ACRS) to N.J. Diaz (Chairman, NRC),
"Vermont Yankee Extended Power Uprate", ACRSR-2174, January 4, 2006.
R-1 C1320503-6924R1 -31221Z006 l
BFNEPUCOPProbabilisticRisk Assessment Appendix A PRA QUALITY The BFN Unit 1 EPU PRA was used in this analysis for the base case quantification as it was recently updated consistent with the ASME PRA Standard and it is representative of each of the three BFN unit PRAs. The following discusses the quality of the BFN Unit 1 PRA models used in performing the risk assessment crediting containment overpressure for RHR and Core Spray pump NPSH requirements:
- Level of detail in PRA
- Maintenance of the PRA
- Comprehensive Critical Reviews A.1 LEVEL OF DETAIL The BFN Unit I PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events.
The PRA model (Level 1 and Level 2) used for the containment overpressure rusk assessment was the most recent internal events risk model for the BFN Unit 1 plant at EPU conditions (BFN model U1050517). The BFN PRA models adopts the large event tree / small fault tree approach and use the support state methodology, contained in the RISKMAN code, for quantifying core damage frequency.
The PRA model contains the following modeling attributes.
A. 1.1 Initiating Events The BFN at-power PRA explicitly models a large number of internal initiating events:
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BFNEPUCOPProbabilisticRiskAssessment
- General transients
. LOCAs
- Support system failures
- Internal Flooding events The initiating events explicitly modeled in the BFN at-power PRA are summarized in Table A-1. The number of internal initiating events modeled in the BFN at-power PRA is similar to or greater than the majority of U.S. BWR PRAs currently in use.
A.1.2 System Models The BFN at-power PRA explicitly models a large number of frontline and support systems that are credited in the accident sequence analyses. The BFN systems explicitly modeled in the BFN at-power PRA are summarized in Table A-2. The number and level of detail of plant systems modeled in the BFN at-power PRA is equal to or greater than the majority of U.S. BWR PRAs currently in use.
A. 1.3 Operator Actions The BFN at-power PRA explicitly models a large number of operator actions:
- Pre-Initiator actions
- Post-Initiator actions
- Recovery Actions
- Dependent Human Actions Approximately fifty operator actions are explicitly modeled in the BFN PRA. A summary table of the individual actions modeled is not provided here.
A-2 C1320503-6924R1 - 3122/2006 I
BFNEPUCOP ProbabilisticRiskAssessmrnt The human error probabilities for the actions are modeled with accepted industry HRA techniques.
The BFN PRA includes an explicit assessment of the dependence of post-initiator operator actions. The approach used to assess the level of dependence between operator actions is based on the method presented in the NUREG/CR-1278 and EFRI TR-1 00259.
The number of operator actions modeled in the BFN at-power PRA, and the level of detail of the HRA, is consistent with that of other U.S. BWR PRAs currently in use.
A.1.4 Common Cause Events The BFN at-power PRA explicitly models a large number of common cause component failures. Approximately two thousand common cause terms are included in the BFN Unit 1 PRA. Given the large number of CCF terms modeled in the BFN at-power internal events PRA, a summary table of them is not provided here. The number and level of detail of common cause component failures modeled in the BFN at-power PRA is equal to or greater than the majority of U.S. BWR PRAs currently in use.
A.1.5 Level 2 PRA The BFN Unit I Level 2 PRA is designed to calculate the LERF frequency consistent with NRC Regulatory Guidance (e.g. Reg. Guides 1.174 and 1.177) and the PRA Application Guide.
The Level 2 PRA model is a containment event tree (CET) that takes as input the core damage accident sequences and then questions the following issues applicable to LERI-:
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BFNEPUCOP ProbabilisticRiskAssessmmt
- Primary containment isolation
- RPV depressurization post-core damage
- Recovery of damaged core in-vessel
- Energetic containment failure phenomena at or about time of RPV breach
- Injection established to drywell for ex-vessel core debris cooling/scrubbing
- Containment flooding
- Drywell failure location
- Wetwell failure location
- Effectiveness of secondary containment in release scrubbing The following aspects of the Level 2 model reflect the more than adequate level of detail and scope:
- 1. Dependencies from Level 1 accidents are carried forward directly into the Level 2 by transfer of sequences to ensure that their effects on Level 2 response are accurately treated.
- 2. Key phenomena identified by the NRC and industry for inclusion in BWR Level 2 LERF analyses are treated explicitly within the model.
- 3. The model quantification truncation is sufficiently low to ensure adequate convergence of the LERF frequency.
A.2 MAINTENANCE OF PRA The BFN PRA models and documentation are maintained living and are routinely updated to reflect the current plant configuration following refueling outages and to reflect the accumulation of additional plant operating history and component failure data.
The PRA Update Report is evaluated for updating every other refueling outage. The administrative guidance for this activity is contained in a TVA Procedure.
A-4 C1320503-6924R1 -3/22/f206 l
BFNEPUCOP ProbabilisticRiskAssessment In addition, the PRA models are routinely implemented and studied by plant PRA personnel in the performance of their duties. Potential model modifications or enhancements are itemized and maintained for further investigation and subsequent implementation, if warranted. Potential modifications identified as significant to the results or applications may be implemented in the model at the time the change occurs if their impact is significant enough to warrant.
A.2.1 History of BFN PRA Models The current BFN Unit 1 PRA is the model used for this analysis. The BFN Unit 1 PRA was initially developed in June 2004 using the guidance in the ASME PRA Standard, and to incorporate the latest plant configuration (including EPU) and operating experience data. The Unit 1 PRA was then subsequently updated in August 2005. The Unit 1 PRA was developed using the BFN Unit 2 and Unit 3 PRAs as a starting point.
The BFN Unit 2 and Unit 3 PRAs have been updated numerous times since the original IPE Submittal. The BFN Unit 2 PRA revisions are summarized below:
Original BFN IPE Submittal 9/92 Revision to address plant changes and 8/94 incorporate BFN IE and EDG experience data Revision to ensure consistency with the 4/95 BFN Multi-Unit PRA Revision to address PER BFPER 970754 10/97 2002 PRA Update 3/02 2004 PRA Update (includes conditions to 6/04 reflect EPU) 2005 Update 8/05 A-5 C1320503-6924R1 -3/22/006
BFNEPUCOPProbabilisticRiskAssessmwnt A.3 COMPREHENSIVE CRITICAL REVIEWS As described above, the BFN Unit 1 PRA used in this analysis was built on more than 10 years of analysis effort and experience associated with the Unit 2 and 3 PRAs.
During November 1997, TVA participated in a PRA Peer Review Certification of the Browns Ferry Unit 2 and 3 PRAs administered under the auspices of the BWROG Peer Certification Committee. The purpose of the peer review process is to establish a metlod of assessing the technical quality of the PRA for its potential applicabons. The elements of the PRA reviewed are summarized in Tables A-3 through A-4.
The Peer Review evaluation process utilized a tiered approach using standardized checklists allowing a detailed review of the elements and the sub-elements of the Browns Ferry PSAs to identify strengths and areas that need improvement. The review system used allowed the Peer Review team to focus on technical issues and to issue their assessment results in the form of a "grade" of 1 through 4 on a PRA sub-element level.
To reasonably span the spectrum of potential PRA applications, the four grades of certification as defined by the BWROG document "Report to the Industry on PRA Peer Review Certification Process - Pilot Plant Results" were employed.
During the Unit 2 and 3 PSAs updates in 2003, the significant findings (i.e., designated as Level A or B) from the Peer Certification were resolved, resulting in the PRA elements now having a minimum certification grade of 3. The Unit 1 PRA used in this analysis has incorporated the findings of the Units 2 and 3 PSA Peer Review. The previously conducted Peer Review was effectively an administrative and technical Peer Review of the Unit 1 PRA. Similar models, processes, policies, approaches, reviews, and management oversight were utilized to develop the Unit 1 PRA.
Ax6 C1320503-6924R1 - 3122 W06I
BFNEPUCOPProbabilisticRiskAssessment A.4 PRA QUALITY
SUMMARY
The quality of modeling and documentation of the BFN PRA models has been demonstrated by the foregoing discussions on the following aspects:
- Level of detail in PRA
- Maintenance of the PRA
- Comprehensive Critical Reviews The BFN Unit 1 Level 1 and Level 2 PRAs provide the necessary and sufficient scope and level of detail to allow the calculation of CDF and LERF changes due to the risk assessment requiring containment overpressure for sufficient NPSH for the Il)w pressure ECCS pumps.
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BFNEPUCOPProbabilisticRisk Assessment Table A-1 INITIATING EVENTS FOR BFN PRA Initiator Mean Frequency Category (events per year)
Transient Initiator Categories Inadvertent Opening of One SRV 1.36E-2 Spurious Scram at Power 8.76E-2 Loss of 500kV Switchyard to Plant 1.02E-2 Loss of 500kV Switchyard to Unit 2.37E-2 Loss of Instrumentation and Control Bus 1A 4.27E-3 Loss of Instrumentation and Control Bus 1B 4.27E-3 Total Loss of Condensate Flow 9.45E-3 Partial Loss of Condensate Flow 1.93E-2 MSIV Closure 5.52E-2 Turbine Bypass Unavailable i.95E-3 Loss of Condenser Vacuum 9.70E-2 Total Loss of Feedwater 2.58E-2 Partial Loss of Feedwater 2.47E-1 Loss of Plant Control Air 1.20E-2 Loss of Offsite Power 7.87E-3 Loss of Raw Cooling Water 7.95E-3 Momentary Loss of Offsite Power 7.57E-3 Turbine Trip 5.50E-1 High Pressure Trip 4.29E-2 Excessive Feedwater Flow 2.78E-2 Other Transients 8.60E-2 ATWS Categories Turbine Trip ATWS 5.50E-1 LOSP ATWS 7.87E-3 Loss of Condenser Heat Sink ATWS 1.52E-1 Inadvertent Opening of SRV ATWS 1.36E-2 Loss of Feedwater ATWS 3.02E-1 LOCA Initiator Categories Breaks Outside Containment 6.67E-4 Excessive LOCA (reactor vessel failure) 9.39E-9 Interfacing Systems LOCA 3.15E-5 A-8 C13205036924R1 -3/22/2006 l
BFNEPUCOP ProbabilisticRisk Assessment Table A-1 INITIATING EVENTS FOR BFN PRA Inrtiator Mean Frequency 2!!220!X l pnser year)
Large LOCA - Core Spray Line Break Loop I 1.68E-6 Loop BI 1.68E-6 Large LOCA - Recirculation Discharge Line Break Loop A 1.1 8E-5 Loop B 1.1 8E-5 Large LOCA -Recirculation Suction Line Break Loop A 8.39E-7 Loop B 8.39E-7 Other Large LOCA 8.39E-7 Medium LOCA Inside Containment 3.80E-5 Small LOCA Inside Containment 4.75E-4 Very Small LOCA Inside Containment 5.76E-3 Internal Flooding Initiator Categories EECW Flood in Reactor Building - shutdown units 1.20E-3 EECW Flood in Reactor Building - operating unit 1.85E-6 Flood from the Condensate Storage Tank 1.22E-4 Flood from the Torus 1.22E-4 Large Turbine Building Flood 3.65E-3 Small Turbine Building Flood 1.65E-2 A-9 C1320503-6924R1 - 3122:206
BFNEPUCOP ProbabilisticRisk Assessm nt Table A-2 BFN PRA MODELED SYSTEMS 120V and 250V DC Electric Power AC Electric Power ARI and RPT Condensate Storage Tank Condensate System Containment Atmospheric Dilution Control Rod Drive Hydraulic Core Spray System Drywell Control Air Emergency Diesel Generators Emergency Equipment Cooling Water Feedwater System Fire Protection System (for alternative RPV injection)
Hardened Wetwell Vent High Pressure Coolant Injection Main Steam System Plant Air Systems Primary Containment Isolation Raw Cooling Water Reactor Building Closed Cooling Water Reactor Core Isolation Cooling Reactor Protection System Recirculation System Residual Heat Removal System RHR Service Water Secondary Containment Isolation Shared Actuation Instrumentation System SRVs /ADS Standby Gas Treatment System Standby Liquid Control System A-1 0 C1320503-6924R1 - 3/221!006
BFNEPUCOP ProbabilisticRisk Assessm ont Table A-2 BFN PRA MODELED SYSTEMS Suppression Pool / Vapor Suppression Turbine Bypass and Main Condenser A-1 1 C1320503-6924R1 - 3/22/;06
BFNEPUCOPProbabilisticRiskAssessmmnt Table A-3 PRA PEER REVIEW TECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Initiating Events
- Guidance Documents for Initiating Event Analysis
- Groupings
- LOCA
- Support System/Special
- ISLOCA
- Break Outside Containment
- Internal Floods
- Subsumed Events
- Data
- Documentation Accident Sequence Evaluation
- Guidance on Development of Event Trees (Event Trees)
- Event Trees (Accident Scenario Evaluation)
- SBO
- LOCA
- ATWS
- Special
- ISLOCA/BOC
- Internal Floods
- Success Criteria and Bases
- Interface with EOPs/AOPs
- Accident Sequence Plant Damage States
- Documentation A-12 C1320503-6924R1 -31221Z006 I
BFN EPU COP ProbabilisticRisk Assessmet Table A-3 PRA PEER REVIEWTECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Thermal Hydraulic Analysis
- Guidance Document
- Best Estimate Calculations (e.g., MAAP)
- Generic Assessments
- FSAR - Chapter 15
- Room Heat Up Calculations
- Documentation System Analysis
- System Analysis Guidance Document(s)
(Fault Trees)
- System Models
- Structure of models
- Level of Detail
- Success Criteria
- Nomenclature
- Data (see Data Input)
- Dependencies (see Dependency Element)
- Assumptions
- Documentation of System Notebooks A-1 3 C1320503-6924R1 - 3122/2006 l
BFNEPU COP ProbabilisticRiskAssessment Table A-3 PRA PEER REVIEWTECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Data Analysis
- Guidance
- Component Failure Probabilities
- System/Train Maintenance Unavailabilities
- Common Cause Failure Probabilities
- Unique Unavailabilities or Modeling Items
- AC Recovery
- Scram System
- Repair and Recovery Model
- SORV
- BOP Unavailability,
- Pipe Rupture Failure Probability
- Documentation Human Reliability Analysis
- Guidance
- Pre-Initiator Human Actions
- Identification
- Analysis
- Quantification
- Post-Initiator Human Actions and Recovery
- Identification
- Analysis :
- Quantification l.
- Dependence among Actons
- Documentation i I A-14 C1320503-6924R1 - 3/221)06
BFNEPUCOPProbabilisticRiskAssessment Table A-3 PRA PEER REVIEWTECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Dependencies
- Guidance Document on Dependency Treatment
- Intersystem Dependencies
- Treatment of Human Interactions (see also HRA)
- Treatment of Common Cause
- Treatment of Spatial Dependencies
- Walkdown Results
- Documentation Structural Capability
- Guidance
- RPV Capability (pressure and temperature)
- ATWS
- Containment (pressure and temperature)
- Reactor Building
- Pipe Overpressurization for ISLOCA
- Documentation Quantification/Results
- Guidance Interpretation
- Computer Code
- Simplified Model (e.g., cutset model usage)
- Dominant Sequences/Cutsets
- Non-Dominant Sequences/Cutsets
- Recovery Analysis
- Truncation
- Uncertainty
- Results Summary A-15 C1320503-6924R1 -3/22/2006 l
BFNEPUCOP ProbabilisticRiskAssessment Table A-4 PRA CERTIFICATION TECHNICAL ELEMENTS FOR LEVEL 2 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Containment Performance Analysis
- Guidance Document
- Success Criteria
- L1/L2 Interface
- Phenomena Considered
- Important HEPs
- Containment Capability Assessment
- End state Definition
- Documentation A-1 6 C1320503-6924R1 - 3/22/1006 l
BFNEPUCOPProbabilisticRiskAssessmont Table A-5 PRA CERTIFICATION TECHNICAL ELEMENTS FOR MAINTENANCE AND UPDATE PROCESS PRA ELEMENT CERTIFICATION SUB-ELEMENTS Maintenance and Update Process
- Guidance Document
- Input - Monitoring and Collecting New Information
- Model Control
- PRA Maintenance and Update Process
- Evaluation of Results
- Re-evaluation of Past PRA Applications
- Documentation A-17 C1320503-6924R1 -3/2212006 I
BFNEPUCOPProbabilisticRisk Assessmsnt Appendix B PROBABILITY OF PRE-EXISTING CONTAINMENT LEAKAGE Containment failures that may be postulated to defeat the containment overpressure credit include containment isolation system failures (refer to Appendix D) and pre-existing unisolable containment leakage pathways. The pre-existing containment leakage probability used in this analysis is obtained from EPRI 1009325, Risk Impact of Assessment of Extended Intearated Leak Rate Testing Intervals.[2] This is the same approach as used in the recent 2005 Vermont Yankee EPU COP analyses, and accepted by the NRC and ACRS. [4]
EPRI 1009325 provides a framework for assessing the risk impact for extending integrated leak rate test (ILRT) surveillance intervals. EPRI 1009325 includes a compilation of industry containment leakage events, from which an assessment was performed of the likelihood of a pre-existing unisolable containment leakage pathway.
A total of seventy-one (71) containment leakage or degraded liner events were compiled. Approximately half (32 of the 71 events) had identified leakage rates of less than or equal to 1La (i.e., the Technical Specification containment allowed leakage rate). None of the 71 events had identified leakage rates greater than 211La. EF'RI 1009325 employed industry experts to review and categorize the industry events, and then various statistical methods were used to assess the data. The result ng probabilities as a function of pre-existing leakage size are summarized here in Table B-1.
The EPRI 1009325 study used 10OLa as a conservative estimate of the leakage size that would represent a large early release pathway consistent with the LERF risk measure, but estimated that leakages greater than 600La are a more realistic representation of a large early release.
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BFNEPUCOPProbabilisticRiskAssessment This analysis is not concerned per se about the size of a leakage pathway that would represent a LERF release, but rather a leakage size that would defeat the containment overpressure credit. Given the low likelihood of such a leakage, the exact size is not key to this risk assessment, and no detailed calculation of the exact hole size is performed here. The recent COP risk assessment for the Vermont Yankee Mark I BWR plant, presented to the ACRS in November and December 2005, determined a leakage size of 27La using the conservative 10CFR50, Appendix K containment analysis approach. Earlier ILRT industry guidance (NEI Interim Guidance - see Ref. 10 of EF'RI 1009325) conservatively recommended use of 10La to represent "small" containment leakages and 35La to represent "large" containment leakages.
Given the above, the base analysis here assumes 20La as the size of a pre-existing containment leakage pathway sufficient to defeat the containment overpressure credit.
Such a hole size does not realistically represent a LERF release (based on EFPRI 1009325) and is also believed (based on the VY hole size estimate) to be on the low end of a hole size that would preclude containment overpressure credit. As can be seen from Table B-1, the probability of a 20La pre-existing containment leakage at any given time at power is 1.88E-03.
Sensitivity studies to the base case quantification (refer to Section 4) assess the sensitivity of the results to the pre-existing leakage size assumption.
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BEN EPUCOPProbabilisticRisk Assessment Table B-1 PROBABILITY OF PRE-EXISTING UNISOLABLE CONTAINMENT LEAK [2]
(as a Function of Leakage Size)(')
Leakage Size Mean Probability of (La) Occurrence 1 2.65E-02 2 1.59E-02 5 7.42E-03 10 3.88E-03 20 1.88E-03 35 9.86E-04 50 6.33E-04 100 2.47E-04 200 8.57E-05 500 1.75E-05:
600 1.24E-05 Notes:
' Reference [2] recommends these values for use for both BWRs and PWRs. Reference [2] makes no specific allowance for the fact that inerted BWRs, such as BFN, could be argued to have lower probabilities of significant pre-existing containment leakages.
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BFNEPUCOP ProbabilisticRiskAssessment Appendix C ASSESSMENT OF BROWNS FERRY DATA Variations in river and suppression pool water temperatures, and the suppression pool level at the Browns Ferry plant were statistically analyzed. The purpose of this data assessment is to estimate for use in the risk assessment the realistic probability that ihe water temperatures and level will exceed a given value, i.e. the probability of exceedance.
C.1 BFN EXPERIENCE DATA The following sets of river water inlet daily temperature, suppression pool water daily temperature, and suppression pool daily level data were obtained and reviewed:
Data Unit Data Period Years River Water Temperature and 2 01/01/00 - 01/31/06 6.1 Suppression Pool Temperature 3 02/01/03 -01/31/06 3.0 Suppression Pool Level 2 01/01/00-01/31/06 6.1 3 02/01/03 - 01/31/06 3.0 The river water temperature data from the above units is not pooled because river temperature is dependent upon the seasonal cycle in weather and is not independent between the units. Use of data for SW inlet temperatures from multiple units woold incorrectly assume the sets of data are independent when in fact they are directly dependent upon weather and the common river source. As such, the statistical assessment of the river water temperature variation uses the largest set of data (i.e., the 6.1 years of data from the Unit 2 river water inlet).
c-1 C1320503-6924R1 -3/22I=0O6 l
BEN EPUCOPProbabilisticRisk Assessment As the torus water temperature has a high dependence on river water temperature for most of the year, the assessment of the torus temperature variability also is based on the 6.1 year data set from Unit 2.
The variation in torus level as experienced by Units 2 and 3 can approximate the level range expected to be seen in Unit 1. As such, the statistical assessment of suppression pool level is based on the level data sets from both units. This creates the largest pool of data and will best approximate the variation in level expected from Unit 1 once it begins operation.
C.2 STATISTICAL ANALYSIS OF TEMPERATURE DATA The chronological variation in river water temperature and torus water temperature is plotted together on the graph shown in Figure C-1. As can be seen from Figure C-1, the torus water temperature is always equal to or higher than the river water temperature. Also, the river water temperatures and torus temperatures are closaly correlated in the warmer months when river water temperature is above approximately 70 0F.
The 6.1 years of temperature data was categorized into 5-degree temperature bins ranging from 500 F to 990 F degrees. The resulting histograms are shown in Figures C-2 and C-3. Figure C-2 presents histogram for the river water temperature and Figure (-3 presents the histogram for the torus water temperature.
The histogram information was then used in a statistical analysis software package Ii (Crystal Ball, a MS Excel add-in, developed by Decisioneering, Inc. of Denver, CO) to approximate a distribution of the expected range in temperature.
The Crystal Ball software automatically tests a numberlof curve fits. The best fit for the temperature data is a normal distribution that is truncated at user-defined upper and C-2 C1320503-6924R1 - 3/22/;0O6 I
BFNEPUCOPProbabilisticRiskAssessment lower bounds. If upper and lower bounds are not defined, the tails of the curve fit distribution extend to unrealistic values (e.g., river water and torus water temperatures below 00 F degrees). To constrain the distributions, the following user-defined upper and lower bounds were used:
- River water temperature lower bound of 320 F (no data points in the 6.1 years of data reached 320F, only a single data point reached 350F)
- River water temperature upper bound of 950 F (no data points in the 6.1 years of data exceeded 900F)
- Torus water temperature lower bound of 550 F (no data points in the 6.1 years of data reached lower than 570F)
- Torus water temperature upper bound of 950F (only a single data point in the 6.1 years of data reached 930F)
The Crystal Ball software statistical results for the river water temperature and torus water temperature variations are provided in Figures C-4 and C-5, respectively.
The statistical results are also summarized in the form of exceedance probability as a function of temperature in Figures C-6 and C-7. The information is also presented in tabular form, Tables C-1 and C-2. As discussed previously, the river water and the torus water temperature variations are not independent; as such, the exceedarice frequencies are not independent (i.e., they should not be multiplied together directly to determine the probability of exceeding a particular temperature in the river AND at the same time exceeding particular temperature in the torus).
C.2.1 Conditional Probability of Torus Water Temrrperature One of the parameters used in this risk assessment is the conditional probability that the torus water temperature is greater than or equal to 870 F given river water temperature is greater than or equal to 68 0F. Plant data for Units 2 and 3 were reviewed to C-3 C1320503-6924R1 - 31221:!0 I
BFNEPUCOPProbabilisticRisk Assessment determine this conditional probability. The same data period used for the river water and torus temperature is used in this calculation and both units worth of data is pooled.
A simple likelihood estimate was performed. The following table lists the number of data records where river water temperature was greater,than 680F, and of those records, the number of records where the torus temperature exceeded 87 0F.
River >= 68F Torus>=87F Cond Prob.
Unit2 1103 512 4.6E-1 Unit 3 566 225 4.OE-1 Combined 1669 737 4.42E-1 As the table shows, the likelihood of the torus being greater than 870F when the river temperature is greater than 680 F is 4.42E-1.
C.3 STATISTICAL ANALYSIS OF SP LEVEL DATA The 9.1 years of Browns Ferry Unit 2 and Unit 3 suppression pool level data was categorized into 0.25 inch water level bins ranging from -1.00 inches to -6.25 inches.
Browns Ferry operating instructions require that suppression pool water level remain between these values. The plant is not allowed to remain at power if suppression pool water level falls outside this range. Data points far outside the -1.00 to -6.25 inch range are not included in the statistical analysis because they reflect levels experienced when the plant was shutdown (which is a plant state inapplicable to this risk assessment).
Approximately 53 level data points were not included.
The resulting suppression pool level histogram is shown in Figure C-8.
The histogram was then input into the Crystal Ball software tool to approximate a distribution of the expected range in suppression pool level. The Crystal Ball software statistical results for suppression pool level variations are provided in Figure C-9.
cœ C1320503-6924R1 -X322/10O6
BFNEPUCOPProbabilisticRiskAssessment The statistical results are also summarized in the form of probability as a function of suppression pool level in Figures C-10. The information is also presented in tabular form in Table C-3.
C-5 C1320503-6924R1 -3/2212006 l
BFNEPU COP ProbabilisticRisk Assessment Figure C-1 CHRONOLOGICAL VARIABILITY IN RIVER WATER AND TORUS WATER TEMPERATURES I
I-Pool Temp -RiverTemp l 95 -
85J1
-45 Si C 35 01/01/99 01101/00 12/31/00 12/31/01 12/31/02 12/31103 12/30/04 12/30/05 12/30/06 Date C-6 C1320503-6924R1 -322/2006
BFNEPUCOP ProbabilisticRiskAssessment Figure C-2 RIVER WATER TEMPERATURE HISTOGRAM 350 300 250 M 200 150 100 50 325 37.5 42.5 47.5 52.5 57.5 62.5 67.5 72.5 77.5 82.5 87.5 92.5 Temperature C-7 C1320503-6924R1 - 322/2006 l
BFNEPUCOP ProbabilisticRisk Assessment Figure C-3 TORUS TEMPERATURE HISTOGRAM 7tYI 600 500 400 0(I 300 200 100 0 ...... - - - - ------ 1 7.
52.5 57.5 62.5 67.5 72.5 77.5 82.5 87.5 92.5 97.5 Temperature C-8 C1320503-6924R1 - 3/22/2006 I
BFNEPUCOPProbabilisticRiskAssessm:nt Figure C-4 STATISTICAL RESULTS FOR RIVER WATER TEMPERATURE VARIATION Crystal Ball Report Simulation started on 216/06 at 7:09:56 Simulation stopped on 2/6/06 at 7:11:44 Forecast: River Temperature Cell: G18 Summary:
Display Range is from 30.00 to 100.00 F Entire Range is from 32.00 to 95.00 F After 50,000 Trials, the Std. Error of the Mean is 0.08 Statistics: Value Trials 50000 Mean 63.50 Median 63.41 Mode Standard Deviation 18.07 Variance 326.51 Skewness 0.00 Kurtosis 1.81 Coeff. of Variability 0.28 Range Minimum 32.00 Range Maximum 95.00 Range Width 63.00 Mean Std. Error 0.08 Forecast RiverTemperature 60,000 Trals FrequencyChart 0 OutlIers
.012 113
.009 4 8f .0_06 g l l > .
.000' 0 in 30X0 47.50 651w 0290 10.00 F
Percentiles:
Percentile F 0.0% 32.00 2.5% 33.60 5.0% 35.25 50.0% 63.41 95.0% 91.69 97.5% 93.32 100.0% 95.00 C-9 C1320503-6924R1 -3/22/:2006
BEN EPUCOPProbabilisticRisk Assessm ont Figure C-5 STATISTICAL RESULTS FOR TORUS WATER TEMPERATURE VARIATION Crystal Ball Report Simulation started on 2/6/06 at 7:09:56 Simulation stopped on 2/6/06 at 7:11:44 Forecast: Pool Temperature Cell: C15 Summary:
Display Range is from 55.00 to 95.00 F Entire Range is from 55.00 to 95.00 F After 50,000 Trials, the Std. Error of the Mean is 0.05 Statistics: Value Trials 50000 Mean 75.75 Median 76.06 Mode Standard Deviation 11.30 Variance 127.65 Skewness -0.08 Kurtosis 1.85 Coeff. of Variability 0.15 Range Minimum 55.00 Range Maximum 95.00 Range Width 40.00 Mean Std. Error 0.05 Forecast Pool Temperature 60,000 Trals Frequency Chart 0 Outliers
.01. C -6 _ 3
.009 429.7 Aft e 100 21 .
143
.000 03 0.0 65.00 75.00 6.00 ".00 F
Percentiles:
Percentile F 0.0% 55.00 2.5% 56.22 5.0% 57.46 50.0% 76.06 95.0% 93.04 97.5% 94.02 100.0% 95.00 C-1 0 C1320503-6924R1 -3/22/2006 I
BFNEPU COP ProbabilisticRisk Assessment Figure C-6 RIVER WATER TEMPERATURE EXCEEDANCE PROBABILITY 1.OE+O 1.OE-1
-J 5:i 0co a.
It z
a w
CM xLU 1.0E-2 1.0E-3 25 34 37 41 44 48 51 55 58 62 65 69 72 76 79 83 86 90 93 97 100
'R!V01 '.A.A^T E R FE 'P E FLAT L!PRE 1rF '
C-11I C1320503-6924R1 - 3/22/2006
BFN EPUCOP ProbabilisticRisk Assessment Figure C-7 TORUS WATER TEMPERATURE EXCEEDANCE PROBABILITY 1.OE+O 1.0E-1 M
m o
0.
WC z
0 Lu u 1.0E-2 1.OE-3 50 57 59 61 63 65 67 69 71 73 75 77 79 81 83 85 87 89 91 93 95 TORUS WATER TEMPERATURE (F)
C-12 C1320503-6924R1 - 3/22/2006
BFN EPUCOP ProbabilisticRiskAssessment Figure C-8 SUPPRESSION POOL LEVEL HISTOGRAM 400-350 300 - 11 250 200 150-I-
0
-6.25 -6.00 -5.75 -5.50 -5.25 -5.00 -4.75 -4.50 -4.25 -4.00 -3.75 -3.50 -3.25 -3.00 -2.75 -2.50 -2.25 -2.00 -1.75 -1.50 -1.25 -1.00 Temperature C-1 3 C1320503-6924R1 - 3/22/2006
BFNEPUCOPProbabilisticRiskAssessmnnt Figure C-9 STATISTICAL RESULTS FOR TORUS WATER LEVEL VARIATION Crystal Ball Report Simulation started on 3/7/06 at 15:33:31 Simulation stopped on 317/06 at 15:58:17 Forecast Normal -Torus Level Cell: F3 Summary:
Display Range is from -6.50 to -1.00 Entire Range is from -7.71 to -0.11 After 50,000 Trials, the Std. Error of the Mean is 0.00 Statistics: Value Trials 50000 Mean -3.68 Median -3.68 Mode Standard Deviation 0.90 Variance 0.81 Skewness -0.01 Kurtosis 3.00 Coeff. of Variability 0.27 Range Minimum -7.71 Range Maximum -0.11 Range Width 7.60 Mean Std. Error 0.00 Fcret: No. d-TorsLmvod GQODDTrbb FmcqumqChat 119 Outl
.02S -1240 A19. A836 3 .C12 -l . $24 =
0 _ n r i .00 - -312
.000 a
.6.80 .5.12 3.750 .237 -1.80 Percentiles:
Percentile Value 0.0% -7.71 2.5% -5.45 5.0% -5.16 50.0% -3.68 95.0% -2.20 97.5% -1.92 100.0% -0.11 C-14 C1320503-6924R1 - 3/22/2006
BFN EPUCOP ProbabilisticRiskAssessment Figure C-1 0 SUPPRESSION POOL WATER LEVEL PROBABILITY (Probability that Level Below Value of Interest) 1.OE+O 1.OE-1
.................................................................. JIF..............................................................................................................................................................................
m 0
- 0. ................................. .....................................................................................................................................................................................................
1.OE-3 I'
-6.55 -6.28 -6.01 -5.73 -5.46 -5.18 -4.91 -4.63 -4.36 -4.08 -3.81 -3.53 -3.26 -2.98 -2.71 -2.43 -2.16 -1.88 -1.61 -1.33 -1.06 TORUS WATER LEVEL (inches)
C-15 ci5C1320503-6924R1l-3/2/00
BFNEPUCOPProbabilisticRiskAssessm nt Table C-1 RIVER WATER TEMPERATURE EXCEEDANCE PROBABILITIES Temperature (F) Exceedance Probability 30 1.OOE+00 35 9.55E-01 40 8.80E-01 45 8.02E-01 50 7.24E-01 55 6.45E-01 60 5.64E-01 65 4.74E-01 70 3.97E-01 75 3.17E-01 80 2.41 E-01 85 1.64E-01 86 1.40E-01 90 8.46E-02 95 9.15E-03 100 O.OOE+00 C-16 C1320503-6924R1 -3/22r;006 I
BFNEPUCOPProbabilisticRiskAssessment Table C-2 TORUS WATER TEMPERATURE EXCEEDANCE PROBABILITIES Temperature ('F) Exceedance Probability 30 1.OOE+00 35 1.OOE+00 40 1.OOE+00 45 1.00E+00 50 1.OOE+00 55 1.OOE+00 60 8.90E-01 65 7.79E-01 70 6.63E-01 75 5.28E-01 80 4.01 E-01 85 2.62E-01 90 1.35E-01 92 8.25E-02 95 1.01 E-02 100 0.OOE+00 I
C-1 7 C1320503-6924R1 - 3/22/2006 l
BFNEPUCOPProbabilisticRiskAssessmlnt Table C-3 SUPPRESSION POOL WATER LEVEL PROBABILITY (Probability that Level Below Value of Interest)
Level (inches) Probability
-6.50 1.1 OE-03
-6.45 1.30E-03
-6.39 1.50E-03
-6.34 1.80E-03
-6.28 2.40E-03
-6.23 3.OOE-03
-6.17 3.60E-03
-6.12 4.20E-03
-6.06 4.90E-03
-6.01 5.80E-03
-5.95 6.80E-03
-5.90 7.90E-03
-5.84 9.1 OE-03
-5.79 1.08E-02
-5.73 1.30E-02
-5.70 1.45E-02(')
-5.68 1.55E-02
-5.62 1.83E-02
-5.57 2.11 E-02
-5.51 2.44E-02
-5.46 2.84E-02
-5.40 3.28E-02
-5.35 3.71 E-02
-5.29 4.24E-02
-5.24 4.78E-02
-5.18 5.38E-02
-5.13 l 6.09E-02
-5.07 6.88E-02
-5.02 7.73E-02
-4.96 8.60E-02
-4.91 9.72E-02 C-18 C1320503-6924R1 - 3122/:2006 I
BFNEPUCOPProbabilisticRiskAssessmant Table C-3 SUPPRESSION POOL WATER LEVEL PROBABILITY (Probability that Level Below Value of Interest)
Level (inches) Probability
-4.85 1.08E-01
-4.80 1.1 9E-01
-4.74 1.31 E-01
-4.69 1.44E-01
-4.63 1.58E-01
-4.58 1.74E-01
-4.52 1.90E-01
-4.47 2.07E-01
-4.41 2.26E-01
-4.36 2.45E-01
-4.30 2.64E-01
-4.25 2.85E-01
-4.19 3.07E-01
-4.14 3.29E-01
-4.08 3.51 E-01
-4.03 3.74E-01
-3.97 3.96E-01
-3.92 4.21 E-01
-3.86 4.44E-01
-3.81 4.69E-01
-3.75 4.93E-01
-3.70 5.16E-01
-3.64 5.41 E-01
-3.59 5.64E-01
-3.53 5.88E-01
-3.48 6.12E-01
-3.42 6.35E-01
-3.37 6.58E-01
-3.31 6.81 E-01
-3.26 7.03E-01
-3.20 7.24E-01 C-19 C1320503-6924R1 -3/2212006 l
BFNEPUCOPProbabilisticRisk.Assessment Table C-3 SUPPRESSION POOL WATER LEVEL PROBABILITY (Probability that Level Below Value of Interest)
Level (inches) Probability
-3.15 7.45E-01
-3.09 7.64E-01
-3.04 7.83E-01
-2.98 8.OOE-01
-2.93 8.17E-01
-2.87 8.32E-01
-2.82 8.47E-01
-2.76 8.60E-01
-2.71 8.72E-01
-2.65 8.85E-01
-2.60 8.96E-01
-2.54 9.07E-01
-2.49 9.18E-01
-2.43 9.27E-01
-2.38 9.35E-01
-2.32 9.42E-01
-2.27 9.48E-01
-2.21 9.55E-01
-2.16 9.61 E-01
-2.10 9.66E-01
-2.05 9.70E-01
-1.99 9.74E-01
-1.94 9.78E-01
-1.88 9.81 E-01
-1.83 9.83E-01
-1.77 9.86E-01
-1.72 9.88E-01
-1.66 9.90E-01
-1.61 9.92E-01
-1.55 9.93E-01
-1.50 9.94E-01 C-20 C1320503-6924R1 - 3/22/2006 l
BFNEPUCOPProbabilisticRiskAssessmntI Table C-3 SUPPRESSION POOL WATER LEVEL PROBABILITY (Probability that Level Below Value of Interest)
Level (inches) Probability
-1.44 9.95E-01
-1.39 9.96E-01
-1.33 9.96E-01
-1.28 9.97E-01
-1.22 9.98E-01
-1.17 9.98E-01
-1.11 9.98E-01
-1.06 9.99E-01
-1.00 1.OOE+00 Note to Table C-3:
(1) A conservative probability value corresponding to -5.70" (123,500 ft) instead of -5.90" (123,250 ft) was used in the base case quantification.
C-21 C1320503-6924R1 -3/22/:OM6 I
BFNEPUCOP ProbabilisticRisk Assessment Appendix D LARGE-LATE RELEASE IMPACT In the November-December 2005 ACRS meetings concerning the Vermont Yankee EPU and COP credit risk assessments, the ACRS questioned the impact on Large-Late releases from EPU and COP credit. The following discussion is provided to address this question for the BFN COP credit risk assessment.
D.1 OVERVIEW OF BFN PRA RELEASE CATEGORIZATION The spectrum of possible radionuclide release scenarios in the BFN Level 2 PRA is represented by a discrete set of release categories or bins. Typical of industry PRAs, the BFN release categories are defined by the following two key attributes:
- Timing of the release
- Magnitude of the release D.1.1 Timing Categorization Three timing categories are used, as follows:
- 1) Early (E) Less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from accident initiation
- 2) Intermediate (I) Greater than or equal to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, but less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
- 3) Late (L) Greater than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The definition of the timing categories is relative to the timing of the declaration cf a General Emergency and based upon past experience concerning offsite accident response:
D-1 C1320503-6924R1 -3/22t2006
BFNEPUCOPProbabilisticRiskAssessment
- 0-6 hours is conservatively assumed to include cases in which minimal offsite protective measures have been observed to be performed in non-nuclear accidents.
- 6-24 hours is a time frame in which much of the offsite nuclear plant protective measures can be assured to be accomplished.
- >24 hours are times at which the offsite measures can be assumed to be fully effective.
Magnitude Catecorization The BFN Level 2 PRA defines the following radionuclide release magnitude classifications:
- 1) High (H) - A radionuclide release of sufficient magnitude to have the potential to cause prompt fatalities.
- 2) Medium or Moderate (M) - A radionuclide release of sufficient magnitude to cause near-term health effects.
- 3) Low (L) - A radionuclide release with the potential for latent health effects.
- 4) Low-Low (LL) - A radionuclide release with undetectable or minor health effects.
- 5) Negligible (OK) - A radionuclide release that is less than or equal to the containment design base leakage.
The definition of the source terms levels distinguishing each of these release severity categories is based on the review of existing consequence analyses performed in previous industry studies, PRAs and NRC studies containing detailed consequence modeling. The BFN Level 2 PRA uses cesium as the measure of the source term magnitude because it delivers a substantial fraction of the total whole body population dose. This approach is typical of most industry PRAs.
In terms of fraction of core inventory CsI released, the BFN release magnitude classification is as follows:
D-2 C1320503-6924R1 -3/22/:2006 I
i::~~i .
. BFN EPUCOPProbabilisticRisk Assessment
,1 is I Release Magnitude; I Fraction of Release Csl Fission Products High greater than 10%
Medium/Moderate I to 10%
Low 0.1 to 1.o%
Low-Low less than 0.1%
Negligible much less than 0.1%
D.2 LLOCA COP CREDIT IMPACT ON LARGE-LATE Based on the preceding discussions, it can be seen that "Large-Late" scenarios are termed High-Late releases in BFN Level 2 PRA terminology and are defined as releases occurring after 24hrs and with a magnitude of >10% Csl.
For this risk assessment it is not necessary to perform any explicit quantification of the Level 2 PRA to determine the effect on large-late releases, i.e., the scenarios of interest in this analysis are never late releases, in fact they are all always Early releases.
The scenarios of interest in this risk assessment are very low frequency postulated scenarios that were not explicitly incorporated into the BFN base PRA. These scenarios are defined by containment isolation failure at t=0, leading to assumed loss of NPSH to the ECCS pumps in the short term and leading to core damage in approximately one hour.
In summary, there is no change in the frequency of Large-Late releases due to the credit of COP in DBA LOCA scenarios.
D-3 C1320503-6924R1 - 3/22/:2006
BFNEPUCOP ProbabilisticRiskAssessment Appendix E REVISED EVENT TREES This appendix provides print-outs of the BFN Unit 1 PRA modified event trees used in this analysis. In addition, the RISKMAN software event tree "rules" and "macros" for these revised event trees are also provided in this appendix. These print-outs are provided at the end of this appendix.
E.1 EVENT TREE REVISIONS The following are details of the changes made to the BFN Unit I PRA RISKMAN models for this risk assessment.
The BFN Unit 1 PRA model of record was modified for this risk assessment to question the status of containment integrity first in the Level 1 large LOCA event trees. In addition, a second node was added to the large LOCA event trees to question the probability of extreme plant conditions (e.g., high river water temperature). These nodes are then used to fail the RHR and CS pumps for scenarios with 2 or less RIHR pumps in SPC.
The scope of the analysis is limited to large LOCA accidents. In order to ensure that only the large LOCA initiators are affected by the event tree changes, several of the existing event trees were renamed. In addition, because the containment isolation top event CIL is located in the containment event tree CETI, it too was renamed. The event tree names were revised as follows:
Original Event New Event Tree Tree Descripbon CETI CETN1 Containment event tree 1 LLCS LLCSN Core spray LLOCA event tree LLRD LLDSN Recirc discharge LLOCA event tree LLO LLON Other large LOCA event tree LLRS LLSN Recirc suction LLOCA event tree E-1 C1320503-6924R1 - 3/22/2006
BFN EPU COPProbabilisticRisk Assessment In the containment event tree, top event CIL was replaced with a dummy top event, CIL[)UM, which is a switch whose branches depends on CIL, now moved into the large LOCA event trees. Two split fractions were developed for CILDUM, one for success (CILDS) and one for failure (CILDF). The branches of CILDUM depend on CIL, which is traced via macro CILFAIL. Macro CILFAIL is a logical TRUE if top event CIL::F, otherwise it is FALSE. If CILFAIL is TRUE, that is if CIL fails, then the failed branch of CILDUM is assigned via split fraction CILDF (1.00E+00). Otherwise, the success branch is assigned via split fraction CILDS (0.OOE+00).
The purpose of installing dummy top event CILDUM is to preserve the containment event tree structure (i.e., the RISKMAN software allows use of a specific top event name only once in an accident sequence structure). All top events that are asked in the base model if CIL fails are still asked; those that are not normally asked are not asked in this sensitivity case.
In each of the large LOCA event trees, top event CIL was added as the left most top event, and top event NPSH was added as the next top event to the right. In this way, the original event tree structure is preserved because CIL transfers to NPSH which transfers to the original first top of each event tree. CIL models containment isolation failure probability, and top event NPSH models the probability of other key plant conditions existing at the time of the accident (i.e., high'reactor power, high RW and SP water temperatures, low SP level).
The existing CIL fault tree was modified to add the probability of a pre-existing containment leak; a basic event (CONDPRE) was inserted justlunder the top 'OR' gate of the CIL fault tree. The CONDPRE basic event is set to different values depending on the size of the leak rate assumed in the base quantification and in sensitivity cases (refer to Table 4-2 and to Appendix F).
E-2 E-2 C1320503-6924R1 - 3/22/;aX I
BFN EPU COP ProbabilisticRisk Assessm 6nt Top event NPSH has two split fractions, NPSH1 and NPSHS. The latter is used to filter out large LOCA sequences where 3 or more RHR pumps are running. The status of the RHR pumps and heat exchangers is tracked via an existing macro in the event tree RHRET. Split fraction NPSHI is the split fraction probability resulting from quantificat on of the NPSH fault tree (refer to Appendix F). Refer to Section 4.2.2 where scenarios with more than 2 RHR pumps in SPC are analyzed as a sensitivity case.
When both top events CIL and NPSH fail, conditions are present such that the model assumes there is insufficient NPSH for the low pressure pumps to operate during a large LOCA. RISKMAN rules were added to assign guaranteed failure split fractions for top events: CS, LPCI, LPCII, SPI and SPII. A macro was created (NPSHLOST, defined as C:IL=F*NPSH=F) and defined in each large LOCA event tree. The macro was then added to the split fraction rule for each guaranteed failed split fraction for the desired l:op event. Note that drywell spray failure is captured by the event tree structure (i.e., if LF'CI loops I and 11are failed, then drywell spray is never asked in the event trees).
E-3 E-3 C1320503-6924Rl -3/22/2006 1
MODEL Name: Ul ERIN Page No. 1 of 2 Event Tree: LLCSN.ETI 122235 Febnrary.14, 2OE JE CL NPSH RPSM RPSE TOR UTP IVC LPCI LPCII CS Si m
MODEL Name: UWERIN Page No. 2 of 2 Event Tree: LLCSN.ETn 12 22:35 Febni4ar14, 200E 1 1 2 2 3 3 4 4
...................................................... X1 5 5-8 i ...................................................... Xi 6 9-12 I ....................... ............................... X1 7 13-16 8 17 9 18 10 19 11 20 X2 12 21-38 m
en1 13 39 X2 .14 40-57 15 58 16 59 X2 17 60-77
_. .......................................... ....................... 18 78 X2 19 79-96 20 97
_............................................................................................................................. 21 98 X2 22 99-116 23 117 24 118 X3 25 119-236 26 237 27 238 28 239 29 240 30 241 31 242
.................................................................................................................................. X5 32 243484 Xe 33 485-968
Model lNam: UlCOP2-9 Top Events for Event Tree: LLCSll 5:06 PH 2/9/2006 Page 1 Sop Zimnt Vxms Desription CIL PRIMARY CONTAINMENT ISOLATION FAILURE - LARGE (->3 INCHES)
NPSH, CONDITIONS PREVENTING NPSH FOR LLOCA RPSM MECHANICAL PORTION OF RPS SUCCESSFUL RPSE ELECTRICAL POAtON' OF R?S (NUREG-SSOO BASIS)
TOR PRESSURE SUPPRESSION POOL TTP TURBINE TRIP IVC CLOSURE OF.MSIVS LPCI LPCI LOOP I LPCII LPC LOOP II CS CORE SPRAY SYSTEM SZ LOGIC SWITCH FOR SUFFICIENT INJSECION.
OSPC OPERATOR ALIGNS SUPPRESSION POOL COOLING SPI SUPPRESSION POOL COOLING 1ARDWARE- LOOP I SPII SUPPRESSION POOL COOLING HARDWARE - LOOP II SPC LOGIC S'PITCH FOR IFPPRF.6STON POOL COOLTNG WITH U1 RHR ODWS OPERATOR ALIGNS DRYWELL SPRAY DWS DRYhELL SPRAY HARDWARE E-6
Model Name: U1COP2-9 Split Fraction Assignment Rule for EVent Tree: LLCSN 5:06 PX 2/9/2006 Paga 1 SF Split Fraotion Assignment Rule CIL1 PCA-S*CDWP-S + LVP-S)
CIL2 PCA-F*(DWP-S + LVP-S)
CILF DWP-F*LVP-F SPSHS RHR*RHR2*RHR3 + RHR1*RHR2*RHR4 + RRR1*RHR3R2HR4 + RZR2*RHR3*aRR4 +
RHP1*RER2*;RH3*RHR4 Comments IF 3 OR MORE PUMPS ARE AVAILABLE WE DON'T NE3rD COP FOR ECCS NPSH NPSH1 INIT-LLCA + INIT-LLCZ + IN:T-LLDA + INIT-LLDB + INIT-LLO + INIT-LLSA +
XNXT-LLSB WNPSHS 1
RPSMS 1
RPSZO
.1, TOR1 TTP1 BB5-S*DI-S TTP2 BB5-S*DI-F
_TTP3 BBS-F*DI-S TTPF 1 ,,
.zVC1 1 LPCIr -LPCISUP + NPSHLOST LPCI2 - LPCISUP Comments YMNUAL LPCI START NOT.CREDITED LLOCAS; ODD SPLIT FRACTION SWOULD APPLY LPCIIF -LPCIISUP + NPSHLOST LPCII2 LPCI-S LPCIS4 -LPCISUP LPCII6 LPCI-F*LPCISUP CSF INIT-LLCA*(RF-F+AC-F+DB-F+AD-F+DD-F+NPII-r + CASSIG +DW-F*LV-F+R3-F+ -EECW)
+ INIT-LLCB* RE-F+AA-r+DA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+RC-F+ -EECW) +
IJPSHLOST CS2 INIT-LLCB*-CR3-F+AA-F+DA-F+AB-F+DC-r+NPI-F+DW-F*LV-F+RC-F+ -EECW)
CS2B
- INIT-LLCA*-(RF-F+AC-r+DB-F+AD-F+DD-r+NP2I-F+ CASSIG+DW-'*LV-F+RB-F+ -EECW)
CSF 1 E-7
Model Name: IlCOP2-9 Split Fraction.Assignment Rule for Event Tree: LLCSN 5:06 H i2/9/2006 Pasm 2 SF Split Vraction Aeignment Rale SIS LPCI-S*RPA-S*RPC-S + LPCIZ-S*RPB-S*RPD-S + LPCI-S*LPCI2-S*( (RPA-S+RPC-S) +
(RPB-S+RPD-S) + CS-S )
Comments WIY TWO RNR PUMPS OR CS FROMTHE UNBROKEN LOOP Sir SIF 1 OSPC1 RPSM-S*RPSE-S OSPCF 1 SPIF OSC-F + RE-F + rPSHmOST SP12 RE-S*RC-S*(RPA-S*lXA-S + R2C;S*paC-S)
SPIF 1 . .~
SPIIF OSPC-F + RF-F + MPSHLOST SPII4 (RPB-S*HXB-S + RPD-S*HXD-S)*52I-S SPII5 (RPB-S*HX3-S + RPD-S*HXD-S)*SPI-F'PE-S 6PII6 (RPS-S*HXB-S + RPD-S*RXD-S)*SPI-F*.RE-F SPIIF 1 SPCF , -(SPI-s)*-CSPII-S)
SPCS SPI-S(5RPA-S*HXA-S + RPC-SIHXC-S) + SPII-S*(RP3-S*HXB-S+RP)DS*HXD-fS)
SPCF 1 ODWS1 1 DWSF PX1-F*PX2-F + (RPA-r*RPC-F +RH-F+NOCG) * (RuB-F*RPD-r+RIF + 1OGD)
DWS1 PX1-S*PX2-S* (RPA-S+RPC-S)*-NOGB*( PB-S+RPD-S)*-IOGD DWS2 (RPP,-F*RPCAF +R-F+NCOGB+PX1-F) * (RPB-F*RPD-F+RI-F + NOGD+PX2-F)
DWSF 1 E-8
Model Namne: TlCOP2-9 Macro for Event Tree: LLCSN 5:06 pH 2/9/2006 Page 1 Xacro Xacro Rale / Co=mants ALTISJRHSW RPSH-B THIS MACRO IS NEEDED IN THE CETS ALTINJU2X RPSM-B THIS MACRO IS NESDED IN THE CETS BUCK.ET RPSM-B CILFAIL CIL-F CLASSiA RPSM-B 6LASS1B RPsM-B CLASSIBE RPSM-D CLASS1BL RPSM-B CLASSiC RPSM-.
CLASSlD RPSM-B CLASSXE RPSX-
- CLASS2 RPSM-B CLASS2A RPSM-B CLASS2L SPC-F + OSPc-F CLASS2T RPSM-B CLASS2V RPSM-B CLASS3A RPS.-B CLASS3B RPSM-B E-9
Model Nama: UlCOP2-9 5acro for Event Tree: LLCSN 5:06 PM 2/9/2006 Page 2 macro Macro RuLe / Comment CLASS3C -(SI-S) +,-(TTE-S+IVC-S)
CLASS3D -(TOR-S)
CLASS4 RPSM-F CLASSS -(TTP-S)*-(IVC-S DWSPRAY D'S-S THIS MACRO IS NEEDED IN TF.E CETS EMDEPHDWR RPSM-3 THIS MACRO IS NEEDED IN THE CETS HIGH API RPSM-B LOW. INIT-LLCA + INIT-LLCB LPCIISUP RF-3*( (NPII-S*DW-S) + LV-S LPCISUP RE-$*( (NPI-S*DW-S) + LV-S LOOP I LPCI SUPPORT LPI SI-S NOACREC RPSM-B THIS MACRO IS NEEDED IN THE CETS NOCD RPSM-S TOR-S*CTTP-S+IVC-S)-SI-S*SPC-S NODC RPSM=B THIS MACRO IS NEEDED IN THE CETS NORV RPSM-B THIS MACRO IS NEEDED IN THe CETS NOSRV RPSM-B THIS MACRO IS NEEDED IN THE CETS NPSHLOST CIL-F*NPSH-F E-10
Model Name: UlCOP2-9 Macro for Event Tree: LLCSN 5:06 X 2/9/2006 Page 3 Mae=o lMaaxo Rule / Coumnts OPDEPLI 'RPSM-B THIS MACRO IS NEEDED IN THE CETS RHRSPCOOL SPC-S SORV RPSM-S LARGE LOCAS MP.EALWAYS DZPRESSURIZEED E-1 1
MODEL Naim.: UIERIN PeeNo I of 2 EventTree LLON.EI1 13:37:12 Feluruy 16.2006 E CAL NPSH RPSM RPSE TOR UP WC dCI IPai Cs S OSPC SPI SPI SPC ODWS DMWS I .
1*3
DMONmne: UlERN Page No.2of 2 Everd Tree: LLON.ETI 13:37:12 Febnny 16. 2006 X# BS St 2 2 3 3 4 4 Xl 5 5-8 Xi 6 9-12 xi 7 13-16 8 17 9 18 10 19 11 20 X2 12 21-38 13 39 X2 14 40-47 171 15 58 16 59 X2 17 60.77 18 78 X2 19 79W96 20 97 21 98 X2 22 99-116 23 117 24 118 X3 25 119-236 26 237 27 238 28 239 29 240 30 241 31 242 X4 32 . 243-484 X5 33 485-968
Model Name: UXCOP2-9 Top Events for Event Tree: LLON 5:07 PH 2/9/2006 Page I .
Top tVent Nam Description CIL PRIMARY CONTAINMENT ISOLATION FAILURE -~ LARGE .(-> INCHES)
NPSH CONDITIONS PREVENTING NPSH FOR LLOCA RPSM MECHANICAL PORTION OF RPS SUCCESSFUL RPSE ELECTRICAL PORTION OF RPS INUREG-5500 BASIb).
TOR PRESSURE SUPP.RESSION POOL.
TTP TUR3INE TRIP IVC CLOSURE OF MSIVS LPCI LPCI LOO' I LPCII LPC LOOP It CS CORE SPRAY SYSTEM SI LOGIC SWITCH FOR SUFFICIENT INJECTION OSPC OPERATOR ALIGNS SUPPRESSION POOL COOLING SPI SUPPRESSION POOL COOLING HARDWARE - LOOP I SPII SUPPRESSION POOL COOLING HARDM1ARE - LOOP II SPC LOGIC SWITCH FOR SUPPRESSION POOL COOLING WITH Ul RHR ODN(S OPERATOR ALIGNS DRYNELL SPRAY DlS DRYWELL SPRAY HARDWARE E-14
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLON 5:07 PH 2/9/2006.
Page I Sr Split Fraction Assignment Rule CIL1 PCA-S*(DWP=.S + LVP"S)
CIL2 PCAX-* (DWP-S + LVP-S)
CILF DWP-F*LVP=F NPSHS RHR1*RHR2*RHR3 + RHR1*RHR2*RHR4 + RHR1*RHR3*RHR4 + RHR2*RHR3*RHR4 +
RHR1*RHR2*RHR3*RHR4 Coxents IF 3 OR MORE PUMPS ARE AVAILABLE WS DON'T NEED COP TOR ECCS NPSR NPSH1 ENIT-LLCA + iNIT-LLCB + INIT-LLDA + INIT-LLDB + INIT-LLO + INIT-LLSA +
_NIT-LLSB SPSHS 1 RPS.MS. 1 RSE0 1 TORi 1 TPl . BB5-S*D!-S TTP2 BB5-S*DI-F TTP3. 8B5-F*DI-S TTPF 1 IvCl 1 LPCIF -LPCISUP + N?SELOST LPC12 LPCISUP.
Comments MANCAL LPCI START NOT CREDITED LLOCASi ODD SPLIT FRACTION SWOULD APPLY LPCIIF -LPCIISUP + NPSHLOST LPCII2 LPCI-S LPCII4 -LPCISUP LPCII6 LPCI-F*LPCISUP CSE' (RF-F+AC-F+DE-F+AD-F+DD-F+NPIX-F+ CASSIG+DW-F*LV-F+RB-F+ -ESCX) *
(RE-F+AA-F+DA-F+AB-F+DC-F+NPI-F+DW-r*LV-r+RC-F+ -EECW) +N.PSHLOST CS2 -RE-r+AA-F+DA -+AB-F+DC-F+NPi-r+DW-r*LV-+RC=F+ -EECW)
CS2B -(RF-F+AC-F'+DB-F+AD-F+DD-F+NPII-F+ CASSIG+DW-F*LV-F+RB-F+ -EECW)
CSF 1 SIS LPCI-S*(RPA-S+R"C-S) + LPCII-S*(RPB-S+RPD-S) + CS-S E-1 5
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLON 5:07 PH 2/9/2006 page 2 sB Split TractiDon' )AiSgnmt RUlu Comments AMYTWO RHR PUMPS OR CS FROMTHE UNBROKEN LOOP SIr OsPC1 RPSM-S*RPSE-S OSPCF 1 SPIF RE-F + OSPC-r + SPSHLOST SPI2 SPIIF OSPC-F + RF-F + NPSaLOST SPII4 CRPB-S*HXB-S +
SPI}5 CRPB-S*HXB-S + RPD-9*HXD-S)*SPI-F*RE-S SPII6 *(RPS-S*EX-s -+ RPD-s*Rxx-s *sPz-r*RE-F SPIIF 1 SPCF -(SPI-S)*-tSPII-S)
SPcs SPI-S*(RPA-S*HXA-S + RPC-S*HXC-SI + SPUI-S*(RCPB-S*HXS-S+RPD-S*HXD-S)
SPCF 1 .
ODWS1 1 .*
DWSF PX1-F*PX2-F + (RPA-F*RPCF. +RH-F+NOGB) * (RPB-F*RPD-F+RFJF + NOGD)
Dws1 PXI-S*PX2-S*(RPA-S+RPC-S)*-X XB*tRP8-S+RPD-S)*-NOGD DWS2 (RPA-FrRPC-F +RH-r+NOGB+PX1-F) * (R?3-F*RPD-F+RI-F + NOGD+PX2-F)
DWSr 1 E-16
Model Name: UlCOP2-9 Macro for Event Tree: LLON 3:07 PM 2/9/2006 Page 1 Macro Macro Rule / Comments ALTINJRHSW RPSM-B THIS MACRO IS NEEDED IN THE CETS ALTINJU2X RPSM-B THIS MAJCRO'IS NEEDED IN THE CETS BUCKET RPSM-a CILFAIL CIL-F CLASSlA R SM-B CLASS1B RPSM-B
.CLASSIBE RPSM-8
- CLASS19L RI.SM-D RPSM-B CLASSIC CLASSlC R PSM-B CLASSlE RPSM-B RPSM-D CLASS2A RPSM-B CLASS2A CLASS2L OSPC-F+ SpC-T CLASS2T RPSM-B CLASS2V RPSM-B CLASS3A RPSM-B CLASS3B3 RPSM-B E-17
Model Name: U1COP2-9 Macro for Event Tree: LLON,
' 5:07 PX 2/9/2006 Page 2 26ac~ro Macro Rule / Comments CLAS53C -(SI-S 1+ -(TTP-S+IVO-3S CLASS3D -(TOR-S)
CLASS4 RPSM-F CLASS5 -I(TP-S)*-[IVC-S)
DWSPRAY DWS-S THIS MACRO IS NEEDED IN THE CETS EMDEPHD'XR RFSM-B THIS MACRO IS NEEDED IN-THE CETS HIGH RPSM-9 HPI RPSM-B .
LOW INIT-LLO 4
LPCIISUP RF-s*l (NPII-S 'DW-Si + LV-S I LPCISUP RE-S*C (NPI-S*D*-S) + LV-S LOOP I LPCI SUPPORT . .
LPI SI-S NOACRZC RPSM-B THIS MACRO IS NEEDED IN THE CETS NOCD RPSM-S
- TOR-S*(TTP-S+IVC-S)*SI-S*SPC-S NODC RPSH-B THIS MACRO IS NEEDED IN THE CETS I i NORV RPSM-8 THIS MACRO IS NEEDED IN THE CETS NOSRV RPSM-B THIS MACRO IS NEEDED IN THE CETS NPSHLOST CIL-F*NPSH-F
.1, E-1 8
' Model Name.; UICOP2-9 Macro for Event Tree: LLONT 5:07 PM 2/9/2006 page 3 Macro Macro Rule / Comments OPDEPL1 RPSM-3 THIS MACRO IS NEEDED IN'THE CETS RHRSPCOOL Spc-S SORV RPSM-S LARGE LOCAS ARE ALWAYS DEPRESSURIZEED E-19
MODEL Name: UIERIN Page No. 1of 2 Event Tree: LLRDN.ETI 13:37:46 February 16,2006 CIL NPSH RPSM RPSE TOR TP [e DVI DV2 LPCI IPCII Cs Si
-X7 X63 X42
'7 CO I.................................................................
I ............................
I ......................................................... ...........................
............................................................................................................................ I....................I....................................................................
MODEL Name: UIERIN Page No.2 of 2 Event Tree: LLRDN.ETI 13:37:46 February 16, 2006 OsPc SPI SPII SPC ODWS DWS X#
1 2 2 3 3 4 4 Xi 5 5-8 Xi 6 9-12 Xi 7 13-16 8 17 9 18 m%)
X2
........................................................................................ .................................................................. 10 19-3&
............................................................................................................... X2 1I 37-54
............................................................................................................... X2 12 55-72
............................................................................................................... X2 13 73-90
......................................................................... ...................................... X2 14 91-108
............................................................ ......................... ......................... X2 15 109-126
............................................................................................................... X2 16 127-144 X4 17 145-288
............................................................................................................... X4 18 289-432
............................................................................................................... X3 19 433-468
.................................... .......................................... ....... , _.......... X5 20 469-936 21 937 22 938 23 939 24 940
................................................................................................................ X6 25 941-1880
............................................................................................................... X7 26 1881-3760
Model Name: UlCOP2-9 Top Events for Event Tree: LLRDN 5:09 PH 2/9/2006 Page 1 Top Zvzt Nam Description CIL PRIMARY CONTAINMENT ISOLATION FAILURE - LARGE (->3 INCHES)
NPSH CONDITIONS PREVENTING NPSH FOR LLOCA RPSM MECHANICAL PORTION OF RPS SUCCESSFUL RPSE ELECTRICAL PORTION OF RPS& NUREG-5500 BASIS)
TOR PRESSURE SUPPRESSION POOL TTP TUR3INE TRIP IVC CLOSURE OF MSIVS DV1 LOOP I RECIRCULATION DISCHARG3 VALVE CLOSURE DV2 LOOP II RECIRCULATION DISCHARGE VALVE CLOSURE LPCI LPCI LOOP I LPCII LPC LOOP II CS CORE SPRAY SYSTEM SI LOGIC SWITCH FOR SUFFICIENT INJECTION OSPC OPERATOR ALIGNS SUPPRESS:ON POOL COOLING SPI SUPPRESSION POOL COOLING HARDARE - LOOP I SPIX SUPPRESSION POOL COOLING HARDWARE - LOOP II SPC *LOGIC SWITCH FOR SUPPRESSION POOL COOLING WITH U1 RXR cOWS OPERATOR ALIGNS DRYWELL SPRAY' DWS DRYWELL SPRAi HARDWARSE E-22
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LL1RDN 3:09 PH 2/9/2006 Page 1 SP Split Fraction assignment .Rul CIL1. PCA-S*(DWP-S + LVP=S)
CIL2 PCA-F*(DWP-S + LVP-S)
CILF DWP-F*LVP-F NPSHS RHR1*RER2*RHR3 + RERI*PF2*RHR4 + RHR1*RHR3*RHR4 + RHR2*RHP.3*RHR4 +
RHR1*RHR2*RHR3*R8ER4 Conments IF 3 OR MORE PUPS ARE AVAILABLE WE DON'T NEED COP FOR ECCS KPSR NPSH1 INIS-LLCA + INIT-JLLCB + INIT-.LDA + INIT-LLDB + INIS-LLO INIT-LLSA +
INIT-LLSB NPSHS 1 RPSMS 1 RPSEO. 1 SORI STP1 BBS-S*DI-S STP2 BB5-S*0I-F STP3 BB5-FkDI=S.
TTPF3 IYC1 IVUi Dvir RE-F+RE-F*RC-F+N81-F*NH2-FiDX-F*LV-F DVll DW-S*LV-S*NH1-S*NH2-S*RB-S*RC-S.
DV12 DW-S*LV-S*NX1-C*NH2-S*CRB-F+RC-F)
DV13 W-S* V-F*NH-S*NH2-S*RB-S*RC-S
- DV14 DW-r*LV-S*Nql-S*N82-S*RB-S*PC-S DV1S5 DW-S*LV-S*(NH1-F+NB2-F)*RB-S*Rc-S DV1F 1 DV2F RF-F+PRB-v*RC-F+NH1-F*NH2-F+DW-F*LV-F DV25 RE-F*DVI-F*DW-S*LV-S*NH1-S*NH2-S*RB-S*RC-S DV21 DVl-S*DW-S*LV-S*NHl-S*NH2-S*RB-S*RC-S DV22 DVl-F*D -S*LV-S *NH1-S NH2-S *RB-S*RC-S DV24 RE-(*DV1-F-DW-S*-V-S*NHI-S*NH2-S*(RB-F+RP-F)
E-23
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LL^;RDI 5:09 PM 2/9/2006
?age 2 Sr Split Fraction Amaignmart Rule DVZ3 DVI-S*DW-S*LV-S*NHl-S*NV -S*SRB-r+Rc-Fr)
DV24 DV1-F*OW-S*LV-S*NH1-5*N32-S*(RB-W+RC-F)
DV27 RE-F*DVl-F*DW-S*LV-F*NHl-S*NH2-S*RB-S*RC-S DV20 DV1-S*DW-S*LV-F*N l3-S*N2-S*RD-S*RC-S DV29 DV1-F*DW-S*LV-F*NH1-S*NH2-S*RB-S*RC-S DV2A RE-r*DV1-F*DW-F*LV-S*NII-S*N32-6S.RB-S*Rr-S DV2B DVI-S*DW-F*LV-S*NH1-S*NH2-S*RB-S*RZ-S DV2C .DVI-*DW-F~tV-S*NHl-S*N32-S*RB-S*RC-S DvMD RE-F*DV1-F*DW-S*LV-S* (NCH-F+NH2-F)*R3-S*RC-S Dv2E Vl-S*DW-S*LV-S*(b1H1?-+NH2-r)*RB-S*RC-S DV2G DVl-r*DW-SLV-S* (NH1-F+NH2-F) *RB-s*RC-S DV2F 1 LPCIF -LCISUP+ DV1-F*DV2-F + NPS3LOST
.LFCI2 LPCISUP LPCIIF -LPCI:SUP +DV1-F*DVZ-r + SPSHLOST LPCII2 hPCI-s LPCI14 -LPCISUP LPCII6 LPCI-F*LPCISUP LPCIIF 1 CSF (RE-F+AA-F+DA-F+AB-F+DC-F+SPI-F+DX-F*LV-F+RC-F+EA-F*EB-r*Ec.r +
EA-r*EB-F*ED-F + EA-F*EC-F*ED-F +
EB-F*EC-F*ED-F)*(RF-F+AC-F+D3-F+AD-F+DDWF+NPII-F+ CSSIG+DW-F*LV-F+RB-F+
EA-F*EB-F*EC-F + EA-F*Es=F*EDEF +.EA-F*EC-F*E5-F + ES-F*EC-F*EC-F' +
NPSRLOST CS1 .- (RE-+AA-F+DA-F+AB-r+DC-r+NPI-F+W-F *LV-F+RC-r+EA-F*EB-t*EC-F +
EArF*3B-r*.D-F + EA-F*EC-r*ED-r +
EB-F*EC-F*ED-F)*-( C-F+AC-F+DB-F+AD-F+DD-F+NPII-F+
CASSIG+DW-F*LV-F+RB-F+EA-FtEB-r'Ec-r + EA-FWEB-FrED-F +. ZA-FIEC-F*ED-F +
EBW-*EC-F*ED-r)
CS2 - -~FAA-F+DA-r+AB-F+CC-F+NPI-F+CX-F*LV-F+RC-F+EA-r*EB-FIEC-F +
EA-r*EB-F*ED-r + EA-F*EC-F*ED-F +
EB-F*Ec-F*ED-F)*(RF-F+AC-F+DB-F+AD=F+DD-F+NPII-F+
CSSIG+DW-F*LV-F+RB-F+EA-F*E8-F*EC-F + EA-F*EZ2F*ED-F + EA-F*EC-F*ED-F +
EB-F*EC-F*E5-FJ E-24
Modlo Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLRDN 5:09 P 2/9/2006 Page 3 sr Split Fraction Assignment Rulu CS2B (RE-F+AA-F+DA F+AB-F+DC-F+NPI-F+DW=r*LV-F+RC-F+EA-F*EB-F*EC-r +
EA-F*EB-F*ED-r + EA-F*EC-F ED-F .+
EB-F*EC-F*ED-F)*-(R -F+AC-F+DB-F+AD-F+DD-r+NPII-r+
CASSXG+Dw-r*LV-F+RB-r+EA-r*EB-r*EC-S + EA-rFEB-F*EDFr + EA-F*EC-FrED-F +
EB-F*EC-F*ED-F)
.CSF Cczmients Core Spray Loop II Pipe Break Large LOCA SIS CS-S + LPCI-S* RPA-i + RPB-S) + LPCII-S* RPB-S + APD-S)
SIT 1 OSPC1 RPSM.-S*RPSE-S OSPCF 1 SPIT E-F + OSPC-F + NPSKLOST; SPI2 1.
SPIIF OSPC-F + RT-F + NPSHLOST SPI14 tRPB-S*HXB.-S + RPDs*HXD-S)*SPIS SPI15 (RPB-S*HXB-S + BPD=S*HXD-S)*SPI-F*RE-S SPI16 (RPB-S*!XB-S + .PD-S*HXD-S)*SPI-F*RE-F SPIIF 1 SPCF -cSPI-S)*-cSPII-S)-
SPCS SPI-S*(RPA-S*NXA-S + RPC-S*EXC-S) + SPII-S*(RPB-S*IXB-S+RPD-S*HXD S)
SPCF 1 ODWS1 1 DWSF PX1-*PX2 T + CRpA-r*Rpc-r +RH-F+NOGX) CRDB-r*RpD-F+R5-F + NOGD)
DW31 PX1-S*PX2-S*(RPA-S+RPC-S)*-N OGB*(RPB-S+RPD-S)*-NXOGD DWS2 CRPA-F*RPc-F +R{-T+soGB+PX1-r)
- CRPB-F*nPD-F+RI-F + NOGD+PX2-F)
DWSF 1 E-25
Model Name: UlCOP2-9 Macro for Event Tree: IRALN 5:09 PX 2/9/2006 Page 1 Macro Macro Rule / Comments ALTINJRHSV RPSM-B THIS MACRO IS NEEDED IN THE CETS ALTINJU2X RPSM=B -
THIS HACRO IS NEEDED IN THE CETS BUCKErT RPSM-B CIL:AIL CIL-F CLASSIA RPSM-B CLASSiB RPSM-B CLASSlBE CLASSIDL RPSM-B CLASSIC RP~SM-B CLASSlD RPSX-B CLASSlE
- RPSF-B CLASS2 RPSM-D CLASS2A CLAS32L OSFC-.T + SpC-F CLASS2T RFSM=.B CLASS2V RPSM-B CLASS3A RPSM-B CLASS3B RPSM-B E-26
Model Name: UlC00P2-9 Macro.fox Event Tree: LLRDN 3:09 PM 2/9/2006 Page 2 Macro Macro Rule / Conrants CLASS3C -(SI-S )+ -LTT?-S+IVC=S)
CLASS3D -(TOR-S)
CLASS4 RPSM-F CLASSS -CTT-S)*-CIVC-SI MK PRAY DWS-S THIS MACRO IS NEEDED IN THE CETS
- EMDEPHrwR RPSM-B THIS MACRO IS NEEDED IN THE CETS HIGH RPSM-B HP- . RPSM-B LOW INIT-LLDA + INIT-LLDB LPC!ISU? RT-S*( (NPII-S*DW-Sj + LV-S 1 t
LPCISUP RE-S*( (PI-S*DW-S) + LV-S LOOP I LPCI SUPORT LPI SI-S NOACREC RPSM-B THIS MACRO IS NEEDED IN THE CETS NOCD RPSM-S
- TOR-S*(TTP-S+IV-S)*SI9-S*SPC.S NODC
- RPS.M-B THIS MACRO IS NEEDED IN THE CETS NORV RPSM-B THIS MACRO IS NEEDED IN THE CETS NOSRV RPSH-B THIS MACRO IS NEEDED IN THE CETS NPSHLOST CIL-F*NPSH-F -
E-27
Modal Name: UlCOP2-9 Macro for Event Tree: LLRDN
- 09 Pn42/9/2006 wag. 3 Hwero Macro Rule / Comments OPDPPL1 RPSIM'-
THIS MACRO IS NUEEDSD I.THE CETS RHRSPCOOL OSPC-F + SPC-F SORV- RPSM.S
-LARGE LOCAS ARS ALWAYS DEPRESStMIZE3D E-28
MODEL Name: UlERIN Page No. I of 4 Event Tree: LLRSN.ETJ 13:38:20 February 16, 200E IE CIL. NPSH RPSM RPSE TOR TTP IYC DVI DV2 LPCI LPCII Cs Si OSPC SP[
X0 I X: ! -X- 1 r71 D . .......... .........
-F -........................
I ..........................................................................................................................
Page No. 2of4 MODEL Name: UlERIN Event Tree: LLRSN.ETI 13:38:20 February 16, 200E SPI SPC ODWS DWS XI 8
. O1 1 2 2 3 3 4 4
.................................... X9 5 5-8
...................................... X9 6 9.12
......................................- X 7 13-16 p8 17 o 9 18
............................................................... X1o 10 19-36
............................. ........... "-.------------....... X1o 11 37-54
............................................................... X1O 12 55.72
......--.----------............. XIO 13 73-90 X1l - -----
- .......................-.......-..;l........................
-- 15 108.124
............................................................... X1i 16 125-141
............................................................... Xi 17 142-282
............................................................... Xi 18 283-423
............................................................... 19 424-458 X2
............................................................... X3 20 459-9116 21 917 22 918
Page No. 3of 4 MODEL Name: UIERIN 13:38:20 February 16, 200E Event Tree: LLRSN.ET1 RPSMI RPSE TOR UIP IVC DVI DV2 UPCI UPCII Cs Si OSPO SPI I IE GIL NPSH I
I I ........................................ I...................................................................................................................................................................
MODEL Name: UIERIN Page No. 4 of 4 Event Tree: ULRSN.ETI 13:38&20 Febuary 16, 200E SPlI SPG ODWS DWS XN# S#
23 919 24 920 X12 25 921-1840 X13 26 1841-3680 m
171
Kodel Name: UlCOP2-9 Top Events for Event Tree: LLRSN 5:09 PK 2/9/2006 Page 1 Top Event Name, Dfesoptiaon CIL PRIMARY CONTAINMENT ISOLATION FAILURE - LARGE (->3 INCHES)
NPSH CONDITIONS PREVENTING NPSH FOR LLOCA RPSM MECHANICAL PORTION OF RPS SUCCESSJUL RPSE ELECTRICAL PORTION OF RPS INUREG-5500 BASIS)
TOR PRESSURE SUPPRESSION POOL TTP TURBINE TRIP IVC CLOSURE OF MSIVS DV1 LOOP I RECIRCULATION bISCHARGE VALVE CLOSURE DV2 LOOP II RECIRCULATION DISCFARGE VALVE CLOSURE LPCI LPCI LOOP I LPCII LPC LOOP II CS CORE SPRAY SYSTEM SI LOGIC SWITCH FOR SUFFICIENT INJECTION OSPC OPERATOR ALIGNS SUPPRESSION POOL COOLING SPI SUPPRESSION POOL CObLING,HARSARE - LOOP I SpII SUPPRESSION POOL COOLING HARDWARE - LOOP II SPC LOGIC SWITCH FOR SUPPRESSION POOL 'COOLING W1TH Ul RHR ODXS OPERATOR ALIGNS DRYWELL SPRAY DWS DRYWELL SPRAY HARDWARE E-33
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tfree: LLRSN
- 5:09 WP 2/9/2006 Page I SF Split Fraction Assignment Rule CILl PCA-S*(DXP.S + LVP-S)
CIL2 PCA-F*(DWP-S + LVP-S)
CILF DOP-F*LVP F XPSHS RMR1*PHR2*RR3 + RMR1*RHR2*RHR4 + RHR1*RHR3*RHR4 + RHR2*RKR3*MR4 +
RMRI*RHR2*'RKR3*RER4*
Comments IF 3 OR MORE PUMPS ARE AVAILABLE WE DON'T NEED COP FOR ECCS NPSH NPSH1 INIT-LLCA + INIT-LLC3 + INIT-LLDA + INIT-LLD3 + INIT-LL0 + INIT-LLSA +
INIT-LLSB NPSHS l RPSMS 1 RPSE0. 1 TOR1 1 TTP1 BB5-S*DI-S TTP2 BB5-S*DI-F TP3 BBS5-F*DI-S TTPF
- 1 IVC1 1 DVlF RE-F+RB-F*RC-F+NH1-F*NH2-F+DW-F*LVF*
Dvil DW-S*LV-S^NE1-S*XH2-S*RB-S*RC-S DV12 DW-S*LV-S*NH1-S*FH2-S*CRB-F+RC-F)
DV13 DW-S*LY-F*sNH-S*NH2-S*RB-S*RZ-S DV14 DW-P*LV-S*NH1-9*SH2-S*RB-S*RC-S DV15 D-WS-LV-S*(NH1-F+NH2-F)*RB-S*RC-s DVlF 1 DV2F . RF-F+R3-F*RC-F+NH1-F*NH2-F+DW-F*LV-F DV25 RE- F*DV1-F*Dr-S*LV-S*NHl-S*Nf2-S*RR-S*RC-S DV21 DV1-S*DW-S*LV-S*NH1-S*NH2-s*RB-s*RC-s DV22 OV1-F*DW-S*LV-9*NH1-S*NH2-S*RB-S*RC-S DV24 RtEt*Dvl-F*DN-S*ZV-S*NH1-S*NH2-S*(RB-F+RC-F)
E-34
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLRSN 5:09 PM 2/9/2006
- age P 2 Br Split rwaction Assignment Rul DV23 DVI-S*DH-S*LV-S*NHl-S*NH2-S*(R3-F+RC-r)
DV2i DVI-F*DW-S*LV-S*NH1-S*NH2-S*(RB-F+RC-F)
DV27 RrD-r*Dvl-F*Dw-t*Lv-r*NHl-s*NH2-S*RB-S*~.C-S DV28 DV1-S*D6 -S-LV-F*N31-S*'N2-S*R3-S*RC-S DV29 DVI-r*Dw-s*LVF*N lS*N2 S*RB-S*RC-S DV2A RE-F*DVI-F*D-rF*LV-S *NH1-S*NH2-S*RB-S*RC-S DV2B Dvi-s*Dw-r*Lv-s*NHl-s*NH2-S*RB-S*RC-S DV2C DV1 -F*DW-F*LV-S*N31-S*N32-S*RB-S*RC-S DV2D R- N*DVI-r*Dw-S*LY-S*(NEI-F+NH2-F)*RB-S*.RC-S DV2E DVl-S*DW.S*Lv.S*(NHl-F+NH2-F)*R^B.S*RC-S DV2G DVI-F*DW-S*LV-S*(NHJ-F4r12-Fl)*B-S*IC-S DV2F 1 LPCIF RE-F + DV1-F + NPSHLOST.
LPCI2 1 LPCIIF RF-F + DV2-F + NPSHLOST LPCII2 LPCI-S LPCII4 RS-F LPCII6 LPCI-F*RE-S LPCIIF 1 CSF (RE--+AA-(+DA-A+AB-F+DC-F+NPI-r+DW-F*LV-F+RC-r+-zECX)* (r-F+AC-F+DB-F+AD-r+D D-r+NpZZ-r+ CASSIG+DW-F*LV-F+R3-F+ -EECW) + NPSHLOST CS 1 - tRE-FAA-r+DA-F+AB-F+DC-F+NPI-F+DW-F*Lv-r+RC-F+
- -EECW)*- (Rr-F+AC-F+D3-r+AD-F+DD-F+NP1I-F+ CASSIG+DW-F*LV-F+RB-F+ -EECW9)
CS2 -(RE-P+AA-F+DA-F+AD-F+DC-F+NP1-F+DW-F*LV-F+RC-F+
-EECW)
- R(F+AC-Fr+DB-+AD-F+DD-F+NPII-F+ CASSZG+DW-F*LV-r+RB-F+ -EECW1 CS2B RE-r+AA-r+DA-F+A3-r+DC-r+NPI-r+DW-F*LV-F+RC-F+-EECW)*-(RF-F+AC-F+DD-F+AD-F+
DC-F+XPIZ-F+ CASSXG+OR-F*LV-F+RS-F+-EECW)
CSF 1 Co~ments Core Spray Loop II Pipe Break Large LOCA 6S1 LPCI-S*RPA-S*RPC-S + LPCII-S*RPB-S*RPD-S + LPCI-S*LPCII-S*
(RPA-S+RPC-S)*(RPB-S+RPD-S)
E-35
Modal Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLRSN 5:09 WH2/9/2006 Page 3 SF Split Fraction Assignment Rule SIF .
OSPCI RPS4-S*RPSE-S OSPCF 1 SPIF LE-F + OSPC-F + NPSHLOST SP12 1.
SPIIF OSPC-F + RF-F + NiSHLOST SPI14 CRPB-S*X3-S + RPD-S.*XD-S)*SPI-S SPI15 (RPE-S*HXB-S + RPD-S*HXD-S)*SI-F*RE-S SPII6 (RPB-S*EXB-S + RPD-S*KXD-S}*SPI-F*Ur-r SPIIF SPcr -CSPI-S)*-(SPII-S.
SPCS SPI-S*CRPA-S*HXA-S + RPFCqSEXC-S) + SP2II-S*(RPB-&*EX-S+RPID-S*ED-S)
- SPCEF 1 CDWS1 1.
DwSF PX1-F*PX2-F + (RPA-F*RPC-F +R3-F+NCGB) * (RPB-F*RPD-F+RI-F + KOGD)
DOWS1 PXI-S*PX2-S*(RPA-S+RPC-s)*-NOGB*(RPFB-S+RPD-S) -tOGD DWS2 (RPA-F*RPC-F +RH-F+NOGB+PXl.rW * (RPB-F*RPD-F+RI-F 4 tOSD+PX2-F)
MrSF 1 E-36
Model Na-me: UlCOP2-9 Macro for Event Tree: LLRSN
- .5:09 PK 2/9/2006 Page 1 Macro
- Macro Rule '/ Coments
. ALTINJRHSW RPS-B.B
- THIS MACRO IS NEEDED IN THE CETS ALTINJU2X RPSN-B THIS MACRO IS NEEDED IN THE CETS BUCKET RPSM!B CILFAIL CIL-F CLASSIA RPSm-B CLASS1B RPsM-B CLASSlDB CLASSIBL CLASSIC RPSM-B
- CLASSID RPSM-B CLASSlE RPSM-B CLASS2 RPSM-B CLAS52A CLASS2L CLASS2T CLASS2V RPSM-B CLASS3A RPSM=B CLASS3B RPSM-B E-37
Model Name: UICOP2-9 Macro for Eveent Tree: LLIiSN S:09 E1d2/9/2006 Pag 2 Idaoro Maoro Rulo / Cosmentz CLASS3C -(SI-S J+ -(TTP-S+IVC-S)
CLASS3D -(TOR-9)
CLASS4 RPSH-F CLASS5 -(TTP-S)*-(IVC-S)
EWSPRAY Dws-S THIS MACRO IS NEEDED IN THE CETS.
EECW EA-S-(EB-S + EC-S + ED-SI + EB-SB*(EC=S + ED-S) + EC-S*ED-S EMDEPHDWR RPSM-B' THIS9MACRO IS NEEDED IN THE CETS EIGH RPSM-B FPI RPSM-B LOW INIT-LLSA + INIT-LLSB IPCIISUP RF-8*C (NPII-S*DW-S) + LV-S I I.PCISUP RE-SI( (NPI-S*D-6S) t LV-S I LOOP I LPCI SUPPORT LPI SI-s NQACREC RPS.M-B THIS MACR0 IS NEEDED IN THE CETS IOCD RPSM-S I TOR-S*(TTP-S+IVC-SI*S-.5*SP-CS IJODC RPSM-B THIS MACRO IS NEEDED TN THE CZETS NORV RPSM-B THIS MACRO IS NEEDED IN THE CETS HOSRV RPSM=B THIS MACRO IS NEEDED IN THE CETS E-38
' Model Name: UICOP2-9 Macro for Event Tree: LLP.SN 5:09 PK 2/9/2006 Pago 3 Macro wao Rule / Comments NPSHLOST CIL-F*NPSH-F OPDSPL1 RPSM-B THIS MACRO aS NEEDED IN THE CETS
- RRRSPCOOL SPC-S SORV RPSM-S LARGE LOCAS ARE ALWAYS DEPRESSLRIZEED E-39
MODEL Name: UIERIN Page No. 1 ol 4 Event Tree: CETN1.ETI 13:36:50 February 16, 200E I IE L2 AL 3. CILDUM 01 JR CZ TD FD DWM WR RME V B#
1 2
3 4
5 6
.7 8
9 10 11 12 13 m
I71 14 0
15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30
.31 32 33 34 35 36
Page No. 2of 4 MODEL Name: UlERIN 13:36:50 February 16, 2W0E Event Tree: CETNI.ETI S#
1 2
3 4
5 6
7 8
9 10 11 12 13 171 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36
MODEL Name: UlERIN Page No. 3 of4 Event Tree: CETN1.ETI 13:36:50 February 16, 200C I IE L2 AL CILDUM of IR CZ TD FD ONW WR RME I X# B#
[
I I 37 L 38 39 40 41 b.
MODEL Nane: UtERIN Page No. 4 of 4 EventTree: CETNI.ETI 13:36:50 February 16, 2006 S#
37 38 39 40 41 mp
Model Name: UlCOP2-9' Top Events for Event Tree: CETN1 5:09 PM 2/9/2006 Page I Top Event Name Description L2 LSVEL 2 /LERF RESULTS AL CIT1 LOGIC NODE FOR CLASS 2 AND CLASSISL CILDUM CIL DUMMYTOe OI OPERATORS DEPRESSURIZE RPV (L2)
IR IN-VESSEL RECOVERY CZ CONTAINMENT ISOLATED AND INTACT TD INJECTION ESTABLISHED FD CONTAINMENT FLOODING DWI NO DIRECT DRYhELL RELEASE PATfl WR WE: AIR SPCE FAILURE RME CONTAINMENT BUILDING EFFECT:VE.
E-44
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETNl 5:09 PM 219/2006 Pag* 1 Si' Split Fraction-Assigiment Rule L20 1 Comments L20-0 IMPLI3S LtVEL 1J L20-1 IMPLIES LEVEL2;.USE MFF TO CFNGE ALF CLASS1A + CIASS1BE + CLASSIC + CLASSID + CLASSlE + CLASS3A + CLASS3B +
CLASS3C PLO NOCD + CLASSlBL + CLASS2A + CLASS2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4
- + CLASS5) + BUCKET Coments CLASS 3D AND CLASS 4 ARE EVALUATED FOR LEXF CILDF CTLFAIL CILDS 1 0os CLASS3A + CLASS3B + CLASS3C + LOW 0:1 CLASS2A + CLASS2T + NORV*(CLASSfA + CZASSlBE + CLASS1BL+ CLASSIC) +
CLASS1B*(NOACREC + NODC) 0:4 CLASS1B 0:3 -OPDEPLl*(CLASSlA + CLASSIC + CLASSID)
Coaments change I hIGR PRESSURE LERF 0-2 OPDEPL1*(CLASSlA + CLASSIC + CLASSlO)
Comments c.angel hIGH PRESSURE LERF OIPF*(CLASSIA + CLASSIC)
IR3 CLASSlBE IR4 CLASSlBL.
IR5 OI-F*CLASS1D IR6 OI-S*CLASSlD Comments the irginal 01 L2 model IIX7 OI.F*CLASS1E IlR8 OI-S*CLASS1E IR2 0I-S Comments LOW PRESSURE INJECTION IMPLICIT IRiF 1 C32 IR-F*OI-S*
C34 IR-F*OI-F C31 IR-S*OI-S CZ3 IR-S*OI-F E-45
Model Name: UlCOP2-9 Split.Fraction Assignment Rule for Event Tre:. CETN1 5:o6 PM2/s/2006 Pag 2 Sr Split Faction Assignment Rule CZF' 1 TDI CLASSlE TD2 OI-S*DXSPRAY TD3 -(OI-B)*CLASSlBE TD4 * -(OS-B) CLASS1BL TD8 OI.-F*CLASS1A SDF I FD1 ALTINJRHSW + DWSPRAY FD2 TD-S*(CLASSIA + CTASSBEB + CLASSIEL + CLASSlD + CLASS3A + CLASS33 + CLASS3C)
F03 TD-F*(CLASS1A + CLASSlC + CLASSlD + CLASS3A + CLASS3B + CLASS3C)
FD4 TD-F*(CLASS1BE + CLASS1BL2 DWIF 1 WRI. DN-S RME3 CLASSlBL Cornents TV-S*DWSPRAY*RIEsPCOOL This was an assumption that resulted in 100 R3S ER'37
- OI-F *.
iME 6 O;-S*TD-S*FD-S*DWS-S RM35 01-3*TD-S*FD-S*DWS-F RM34 OI-S*TD-S*FD-F RMZ3 OIsS*TD-F*FrD-F RMEF 1 L20 1 Comments L20-0 IMLIZES LzVEL 1; L20-1 IMPLIES LEVEL2i USE HFF TO CHANGE ALF CLASSlA + CLASSIBZ 4 CLASSlC + CLASS1D 4 CLASS1E + CLASS3A + CLASS3B +
CLASS3C ALO NOCD + CLASSlBL + CLASS2A + CLASS2L + CLASS2T + CLASSZV + (CLASS3D + CLASS4
+ CLASS5) + BUCKET Comnents CLASS 3D AND CLASS 4 ARE EVALUATED FOR LERr CILDF CILFAIL CILDS. 1 E-46
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event TreeQ: CETNI 5:09 Px 2/9/2006 PAge 3 SF Split Fraction Asaignimni Rule oIS CLASS3A + CLASS3S + CLASS3C + LOW OIl CLASS2A + cLASS2T + NORV*(CLASSlA + CLASS1BE + CLASS1BL+ CLASSIC) +
CLASS1B*(NOACREC + NOOC) 014 CLASS1B 013 -OPDEPL* (CLASSlA + CLASSIC + CLASSiD)
CL2ments changel hIGH PRESSJREZ LERF 012 OPDEPLI*(CLASSUA + CLASSIC + CLASS1D)
Comments change I hIGH PRESSURE LERF IRI O1-F*tCLASSlA + CLASSiC)
IR3 CLASSlBE IR4 CLASSlBL IRS OI-F*CLASSID IR6 OI-S*CLASSID Cocmnents the irginal. Cl L2 model IR7 OI-F*CLASS1E IRS OI-S*CLASSIE IR2 01-S Conmments LOW PRESSURE INJECTION IMPLICIT IRF
- CZ2 IR-F*0I-S CZ4 IR-F*OI-F CZ1 :R-S*OI-S CZ3 IR-S*OI-F CzF *.I TD1 CLASS1E TD2 OI-S*DWSPRAY TD3 -(OI-B)'*CLASSlBE T04 -(OT-B)'CLASSlBL TDO 0I-P*CLASS1A TOF .1 E-47
Model Name: UlCOP2-9 Split Fraction.Assignment Rule for Event Tree: CETN1 5:09 FM 2/9/2036 Page 4 SF Split Fraction Assignment Rule FD1 ALTIqJRHSW ,+ DWSPRAY rD2 TO-S*9 CLASSIA + CLASSIBE + CLASSIBL + CLASSlD + CLASS3A + CLASS3B + CLASS3C)
FD3 TD-F*CCLASsiA + CLASSIC + CLASSlD + CLASS3A + CLASS3B + CLASS3C)
FD4 TD-F*CCLASS13E + CLASSIBL)
DWI? I WRi O.-S RME8 CLASSlBL Comments TDfS*DWSPRAY*RHRSPCOOL This was an assumption that resulted in 100 RBE RMS7 01-F RME6 OI-S*TD-S*FD-S*DWS-S RKE5 OI-S*TD-S*FD-S*DWS-F RRE4 OI-S*TD-S*FD-F RME3 OI-S*TD-F*FD-F L20 1 Comments L20-0 IMPLIES LEVEL 1; L20-1 IMPLIES LEVEL2; US3 14FF TO CHAN¢GE ALF CLASS1A + CLASSlBE + CLASSIC + CLASSID + CLASS1E + CLASS3A +.CLASS3B +
CLASS3C AL0 , NOCD + CLASS18L + CLASS2A +CLASS2L + CLASS2T + CLASS2V + (CLASS30 + CLASS4
+ CLAS55) + BUCXET Comments CLASS 3D AND CLASS 4 ARE EVALUATED FOR LERF CILDF CILFAIL CILDS *1 OIS CLASS3A + CLAS53B + CLASS3C + LOW OIl CLASS2A + CLASS2T + NORV*(CLASSlA. + CLASSIBE + CLASS1BL+ CLASSIC) +
CLASS1B* (NOAC.REC + NODC) 014 CLASSIB OI3 -OPDEPL1*(CLASSIA + CLASSIC + CLASSID)
Comments changet hIG! PRESSURE LERF 012 OPDEPLI* CLASSlA + CLASSlC + CLASSlD)
Comments changel hIGH PRESSURE LERF E-48
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETDi s0os PH 2/9/2006
- Palo5 pay. B Sr Split Fraction Assignment Rule IR1 01-F* CLASSIA + CLASS1C IR3 CLASSIBE SR4 CLASSlBL IR5 0O-F*CLAS31D IRS OI-S*CLASSID Cc~ments the lrginal al L2 mndel IRI OI-F*CLASS1E IRS OI-S*CLASS1E IR2 01-S Conments LOW PRESSURE INJECTION IMPLICIT IRF 1 Cz2 IR-F*oI-s CZ4 IR-F*OI-F CZ1 IR-S*OrS CZ3 IR-S*OI-F CZF TD1 CLASS1E TD2 OI-S*DXSPRAY TD3 -(OI-B)*CLASS1PE TD4 -(0-Bl)*CLASSlBL TD8 OI-F*CLASS1A TDF 1 FDl ALTINJRHSW + DWSPRAY FD2 TD-S*(CLASSlA + CLAS91BE + CLASSlBL + CLASSlD + CLAS53A + CLASS3B + CLASS3C)
FD3 TD-F*(CLASSlA + CLASSIC + CLASS1D + CLASS3A + CLASS3B + CLASS3C)
FD4 TD-F*(CLASS1BE + CLASS1BL)
DWIF1-wR1 DW-S
?.XES CLASS1SL E-49
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CEVN1
- 5:09 PX 2/9/2006 Page 6 SF Split Fraction Assigsent Rule Cormnents TD-S*DWSP3AY*RHRSPCOOL This was an assumption that resulted in 100 RBS RME7 Or-t RMZ6 -OS*TD-SkFD-S*DWS-S
?ME5 0IS*TD-S*FD-S*DWS.I BME4 OI-S*TD-S*FD-F RME3 0I-S*TD-F*FD-F L20 1 Coxrnents L20-O :MPLIES LEVEL 1: L20-1 IMPLIES LEVEL2: USE HFF TO CHANGE ALF CLASS1A + CLASS1BE + CLASSlC + CLASSID + CLASSlE + CLASS3A + CLAS53B +
CLASS3C ALO NOCD + CLASS1BL + CLAS32A + CLASS2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4
+ CLASS5) + BUCKET Cozvments CLASS 3D A-ND CLASS 4 ARE EVALUATED FOR LERF CILDF CILFAIL CILDS 1 OIS CLASS3A + CLASS3B + CLASS3C.+ LOW OIl CLASS2A + CLASS2T + NORV*(CLASSlA + CLASSiBE + CLASSlBL+ CLASSlCI +
CLASS1B*(NOACREC + NOD-)
OI4 CLASSlB 013 -OPDEPL1*(CLASSIA + CLASSIC + CLASS1D)
Coimnents changel hIGH PRESSURE LERF 012 OPDEPL1*(CLASS1A + CLASSIC + CLASS1D)
C=oments changel hIGH PRESSURE LERF IR1 OI-F*(CLASSIA + CLASSICI IR.3 CLASSlBE
_R4 CLASS13L IRS OI-y*CLASSlD IR6 O1-S*CLASS1D Coements the irginal Cl L2 model IR7 0I-F*CLASSlE IRS OI-S*CLASS1E E-50
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETN1 5:09 PH 2/9/2006 Page 7 SF Spplit Fraction Assignment Rule
,IR2 01-S Comments LOW PRESSURE INJECTION fMlLICIT IRF 1 CZ2 IR-F*OI-S CZ4 TR-F*6OIF CZl IR-S*OI-S CZ3 IR-S*OI-F CZF I TD1 CLASS13 TD2 O-S*DWSPRAi T03 (OI-Bl *CLASSlBE.
TD4 -(OI-B) *CLASSIBL TD8 OI-F*CLASS1A TDF 1 FD1 ALTINJRHSW + DWSPRAY FD2 TO=S*(CLASS2A + CLASSIBE + CLASSlBL + CLASSlD + CLAS53A + CLAS93B + CLASS3C)
FD3 Tt-F*(CLASS1l + CLASSIC + CLASSlD + CLASS3A + CLASS3S + CLASS3C)
FD4 TD-F*(CLAsSiBE + CLA5SlBL)
DWIF 1 WR1 DW-S M-78 CLASS1BL Comments TD-S*DWSPRAY*RHRSPCOOL This was an assumption tFhat resulted in 100 RSE RME7 OX-F RME6 OI-S*TD-S*FD-S*DHS-S RME5 OI-S*TD-S*FD-S*DWS-F RME4 O-S*TD-S'FD-F RME3 Ol-S*TD-F*FD-F RMEF 1 L20 1 E-51
- odel Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETN1 5:09 RM 2/9/200.6 Page S SF Split Fraction Assignment Rule Colmmenta L20-0 IMPLIES LEVEL 1; L20-1 IMPLIES LEVEL2; US3 MFF TO CHANGE ALF+ CLASSIA + C S91BE + CLASSIC + CLASS1D + CLASSlE + CLASS3A + CLASS3B +
CLASS3C AL0 NOCD + CLASSIBL + CLASS2A + CLASS2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4
+ CLASSS) + BUEKET Comments CLASS 3D AND CLASS 4 ARE EVALUATED FOR LERF CILDF CILThIL CILDS 1 0IS CLASS3A + CLASS3B + CLASS3C + LOW OIl CLASS2A + CLASS2T + N6RV*CCLASSlA + CLASSIBE + CLASSIBL+ CLASSIC) +
CLASSIB*(NOACREC + NODC)
OI4 CLASSIB 013 -OPDEPL1*(CLASSlA + CLASSIC + CLASSlD)
Comnments change I hIGH PRESSURE LERF 012 OPDEPL1*(CLASSIA + CLASSIC + C:ASSID)
Comments changeI hIGH PRESSURE LERF IR1 0I-F*tCLASSIA + CLASSIC)
IR3 CLASSlBE IR4 CLASSlBL IR5 OI-F*CLASSlD IRS OI-S*CLASSlD Comments the irginal C' L2 model IR7 0I-r4CLASSIE IRS OI-S*CLASS1E IR2 OI-S Comments LOW PRESSURE INJECTION IMPLICIT IRF' 1 CZ2 IR-F*OI-S CZ4 IR-F*O1-F CZ1 IR-S*OS-S CZ3 IR-S*OI-F CZF 1 E-52
Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETN1 5:09 FM 2/8/2006 Page 9 SF, Split Fraction J.ssignmant Rul-TDl CLASSUE TD2 OI-S*DWS2RAY TD3 -(OX-B)'CLASSISE TD4 -(OI-B)*CLASSIDL SDS OI-F*CLASSlA TDF l Eil ALTrlgJRHsw + DwSPRAY ED2 TD-S*(CLASSIA + CLASS1BE + CLASSlUL + CLASSlD + CLASS3A + CLASU3B + CLASS3C)
E'D3 TD-F*(CLASSLA + CLASSlC + CLASSlD 4 CLASS3A + CLASS3B + CLASS3C)
E'D4 TD-F*(CLASSlBE + CLASSlBL)
ERl DW-S EHE8 CLASSlBL CozMents TD-S*DWSPRAY*RHRSPCOOL This was an assurption that resulted In 100 REE ENME7 01-F TMN6 01-xS*TD-S*rn-S*DWS-S EMSES OI-S*TD-S*FD-S*DWS-F.
EME4. OI-S*TD-S*FD-F EW4E3 I6-S-TD-r*rD-r iWEr 1 E-53
Model Name: UlCOP2-9 Macro for Event Tree: CETN1 5:09 Yx 2/9/2006 Page I Macro )aoo Rule I CnentS C1C3LERF CZ-F + RMZ-F* (CILFAIL+DKI-F+IR-F*TD-S*FD-SI CZ-F + RME-F* (CILFAIL+DK-F+IR-r*TD-S'FD-5)
CZ-F + RME-F (CILFAZL+>:F+IR-F*TDS*FDS)
CZ-F + RME-F* CILFAIL+aWI--+IR"FTsD-S*FD-SI CZ=F + RME=F* ICILFAIL+IrAI-F+IR-F*TD-S*FD-S)
E-54
BFNEPUCOPProbabilisticRisk Assessment Appendix F REVISED FAULT TREES This appendix provides print-outs of the BFN Unit 1 PRA modified containment isolation (CIL) fault tree and the NPSH fault tree used in this analysis. These print-outs Eire provided at the end of this appendix.
F.1 FAULT TREE REVISIONS The following two BFN Unit 1 PRA RISKMAN fault tree models were revised for this risk assessment:
- Containment Isolation Failure (CIL)
F.1.1 CIL Fault Tree Revisions The 13FN Unit 1 PRA existing CIL (Containment Isolation Failure) fault tree was modifiad to acid the probability of a pre-existing containment leak; a basic event (CONDPRE) was inserted just under the top 'OR' gate of the CIL fault tree. The remainder of the CIL event tree models containment isolation system failure on demand given an accident.
The CONDPRE basic event probability is based on a 20La leak rate (refer to Table B.-1) for the base case quantification. This event is modified for use in different sensitivity studies.
The containment isolation failure portion of the CIL fault tree is not modified in this risk analysis. Note that one of the quantification sensitivity studies investigates the risk impact if more containment penetrations are explicitly analyzed. However, this F-1 C1320503-6924R1 -3/22/2)06 1
BFNEPUCOPProbabilisticRiskAssessmewnt sensitivity was addressed by modifying the CONDPRE basic event probability to mimic the impact (refer to Table F-1).
The value of the CONDPRE basic event and associated CIL top event frequency for each quantification case is summarized in Table F-1.
F.1.2 NPSH Fault Tree Revisions The NPSH (Conditions Preventing ECCS NPSH for LLOCA Cases) fault tree was created for this risk assessment. The NPSH fault tree models the other (i.e., in addition to containment isolation failure modeled by the CIL fault tree) plant conditions that are necessary in order to require COP credit.
The NPSH fault tree is an "OR" gate structure that models the two Plant States used in this analysis (refer to Sections 3.1 and 3.2). One side of the NPSH fault tree models the probability of plant conditions when the plant is assumed to be at the DBA assumed power level of 102% EPU reactor power. The other side of the NSPH fault tree models the probability of plant conditions when the plant is assumed to be the nominal 100%
reactor power level.
The probability that the plant is at 102% power is modeled using a miscalibration human error probability basic event (ZHECCL) taken from a similar action documented in the existing BFN Unit 1 PRA Human Reliability Analysis for Control Room instrument calibration error.
The NPSH fault tree also includes the following basic events that model the likelihood of exceeding specific river water and suppression pool water temperatures:
"Exceedance Prob for River Water >68F" (RIVER68)
"Given RW>68F, Cond Prob SP Water >87F" (CPSP87RV68)
F-2 C1320503-6924R1 -3/22/206 I
BFNEPUCOPProbabilisticRiskAssessment
- "Exceedance Prob for River Water >85F" (RIVER85)
- "Given RW>85F, Cond Prob SP Water >86F" (CPSP86R\A85)
The NPSH fault tree also includes a basic event (SPLVL123K) that models the probability that the suppression pool water level is at or below 123,500 ft3 at the start of the accident.
The probabilities of the above temperature and level basic events are based on analysis of BFN plant data (refer to Appendix C).
The values of the NPSH fault tree basic events and associated NSPH top event probability for each quantification case are summarized in Table F-2.
F-3 C1320503-6924R1 - 3/22/2X06 l
BFNEPUCOP ProbabilisticRiskAssessment Table F-1 CIL FAULT TREE RESULTS FOR EACH QUANTIFICATION CASE CONDPRE Basic Event CIL Quantification Split Fraction Case Probability Leak Size Probability(')
Base 1.88E-03 2OLa 2.25E-03 I
1 2.47E-04 I OOLa 6.22E-04 I
2 Same as Base Same as Base 2.25E-03 I
3 5.217E-03(2) Same as Base 5.59E-03 I
4 Same as Base Same as Base 2.25E-03 I
5 5.217E-03()2 Same as Base 5.59E-03 I
Notes to Table F-1:
(1) SAIl Support Systems Available" split fraction. 'Degraded Support State" split fraction is also affected but is not shown here.
(2) In these sensitivity cases the pre-existing containment leak rate is maintained at the base value of 2OLa, but the sensitivity issue of increasing the detail of the containment isolation system failure modeling to include smaller lines is addressed here by increasing the CONDPRE basic event probability. This surrogate approach is taken for simplicity. Rather than re-designing the containment isolation system fault tree logic, the probability of the containment isolation system portion of the tree (3.71 E-4) is increased by a factor of 1Ox and the CONDPRE basic event value is modified and used as a surrogate to result in the new top event probability.
F-4 C1320503-6924R1 -322/2006 I
BFNEPUCOP ProbabilisticRiskAssessment Table F-2 NPSH FAULT TREE RESULTS FOR EACH QUANTIFICATION CASE
. Basic Event Probabilities 1 NF'SH Split Quantification Fraction Case ZHECCL RIVER68 CPSP87RWM8 SPLVL123K RIVER85 CPSP86RW85 Pro abili Base 5.OOE 5.64E-01 4.42E-01 1.45E-02 1.64E-01 1.00 2.36E-01 1 Same as Same as Same as Base Same as Same as Same as Base 2.3E;E-0c Base Base am sae Base Base amas s 2 Same Sas ame as Base 1.00 Same as Same s Base 1.64E-01l 2Base Base am saeBase aea s Same as Same as Same as Base Same as Same as Same as Base 2.3-E-0 Base Base Base Base Saeasbs 4 n/a(1) n/a1 n/a(l) n/aM') n/aM1) n/aM1) 1.01E+00 5 n/a(1) n/a(1) n/a(') n/a(') n/a(l) n/a(') 1.01E+00 Notes to Table F-2:
(1) In these sensitivity cases the NPSH split fraction is simply set to 1.0.
F-5 C1320503-6924R1 - 3/22/06 I
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F-1 7
ENCLOSURE 5 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 1 RESPONSE TO QUESTIONS ACVB.17, ACVB.18, ACVB.26, AND ACVB.32 FROM NRC LETTER DATED DECEMBER 22, 2005 On June 28, 2004 (ADAMS Accession No. ML041840109), TVA requested a TS change to allow Unit 1 to operate at extended power uprate conditions. As part of this TS change, TVA requested approval to take credit for containment overpressure in order to provide adequate net positive suction head (NPSH) to the low pressure Emergency Core Cooling System (ECCS) pumps. On October 3, 2005 (Accession No. ML052430341), NRC requested TVA to provide additional information regarding NPSH for the low pressure ECCS pumps, including an assessment of the credit for containment overpressure against the five key principles of risk-informed decision making in Regulatory Guide 1.174.
TVA's submittal dated February 28, 2006 (ADAMS Accession No. ML060620328) provided the response to the NRC's request. During preparation and final review of TVA's submittal dated March 7, 2006, a legacy error was discovered in the existing design calculation which determines the available low pressure ECCS pump NPSH requirements. The error has been documented in BFN's Corrective Action Program, and the calculation has been revised.
The effect of the error is small; however, it is non-conservative and impacts numerical values that were provided in the original EPU submittal and in the February 28, 2006, submittal. Additionally, the error impacted values that were needed to respond to questions ACVBJ17, ACVB.18, ACVB.26, and ACVB.32 from the Request for Additional Information (RAI) provided by the NRC in a letter dated December 22, 2005 (Accession No. ML053560120). TVA'slsubmittal dated March 7, 2006, noted that the response to NRC Requests ACVB.17, ACVB.18, ACVB.26, and ACVB.32 wouldibe deferred. This enclosure provides the responses to these NRCjRequests.
NRC Request ACVB.17 Section 4.2.5 of the PUSAR addresses ECCS net positive suction head (NPSH). This section states that 157 ft2 of unqualified paint was assumed in the calculation of ECCS strainer head loss.
Discuss when this determination was!made and why it is still valid. Include a discussion demonstrating that it bounds the actual unqualified paint for both units. Address how this unqualified paint is distributed between the ECCS suction E5-1
strainers. Verify that there have been no changes to the ECCS suction strainer calculations, including debris generation, transport and head loss. Additionally discuss what temperature is assumed for the suppression pool water in the head loss calculations.
TVA Reply to ACVB.17 Previously, BFN Units 2 and 3 installed new large capacity ECCS strainers to meet the requested actions of NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors." As part of the restart for Unit 1, large capacity ECCS strainers of the same design as previously installed on Units 2 and 3 have been installed on Unit 1. The analytical methodologies for debris source terms contained in the Utility Resolution Guidance (Reference ACVB.17-1) were employed in the calculations.
The assumed 157 ft2 value is a historical number (from the late 1980s) which was chosen to be conservative to the actual values of unqualified paint that would affect the ECCS strainers. The amount of unqualified coating in primary containment is tracked for Units 2 and 3 in the Primary Containment Uncontrolled Coating Logs to ensure that the analyzed value of 157 ft2 is not exceeded.
A Primary Containment Uncontrolled Coating Log is being generated as part of the restart of Unit 1. This log will be like the ones used on Units 2 and 3, and it will be maintained to ensure that 157 ft2 will not be exceeded.
During the initial 10 minutes following the LOCA, evaluation of the system operating modes and the resulting flow rates through the four ECCS strainers on the BFN ring header determined the flow and associated debris (including an assumed 157 ft2 of unqualified coatings) distribution between the suction strainers. The results showed that one of the strainers will receive as much as 26.1% of the flow and will, therefore, accumulate a greater portion of the debris than the other strainers. This amount of debris is then assumed for all strainers when calculating head loss. The amount of the debris filtered was determined as 1-eLT, where L is the filtering time constant at a given flow rate and T is the time that the filtering has been underway. The debris is distributed between the strainers according to the flow fraction and the fraction of debris filtered during the first 10 minutes. A similar process is followed subsequent to the 10 minutes for the remaining fraction of the debris. These terms are then combined to give the final distribution of debris on the ECCS strainers. Since E5-2
the ECCS flows are not changed for EPU containment analysis, there is no change to this methodology and no effect on EPU due to the limiting amount of containment unqualified coatings.
For EPU, there have been no changes to the ECCS suction strainer calculations, including debris generation, transport, and head loss. Strainer head loss calculations were performed at 95, 150, and 1770 F at various flow rates and applied to specific cases (references ACVB.17-2 and 3.).
REFERENCES ACVB.17-1: NEDO-32686, "Utility Resolution Guidance for ECCS Suction Strainer Blockage," dated November, 1996 ACVB.17-2: GENE-E12-00148-01, "ECCS Suction Strainer Hydraulic Sizing Report." This was previously provided to the NRC by TVA letter, T. E. Abney to NRC, "Browns Fer:ry Nuclear Plant (BFN) - Response to Request for Additional Information (RAI) Relating to Units 2 and 3 Overpressure for Emergency Core Cooling System (ECCS) Pump Net Positive Suction Head (NPSH)
Analyses," dated November 25, 1998 ACVB. 17-3: Calculation MDQ0999970046, "NPSH Evaluation of Browns Ferry RHR and CS Pumps." A copy is provided in Enclosure 6.
NRC Request ACVB.18 Provide a figure showing the minimum wetwell pressure and the pressures required to provide adequate available NPSH for the RHR and core spray pumps as a function of time after accident initiation. Discuss the minimum pressure difference between the pressure required to provide adequate available NPSH and the calculated minimum wetwell accident pressure.
TVA Reply to ACVB.18 The requested figures are provided in Figures ACVB.18-1 through ACVB.18-5. Figures ACVB.18-1 and ACVB.18-2 are provided in as Figures SPSB-A.11-1 and SPSB-A.11-2, respectively. They are provided here for completeness. These figures provide the containment pressure required for adequate NPSH and the containment pressure available for each event for which containment overpressure (COP) is credited. These events are DBA LOCA (short-term and long-term graphs are provided),
ATWS, Appendix R, and SBO. COP is required for adequate NPSH when the pump NPSH required line exceeds the atmospheric pressure line shown on the figures. The figures also indicate E5-3
pre-EPU (105% OLTP) values for Units 2 and 3 for peak suppression pool temperature and the associated COP required.
As part of the EPU effort, BFN has given additional consideration for NPSH requirements during non-design basis events. These events (designated as Special Events at BFN) consist of Appendix R, Anticipated Transients Without Scram (ATWS), and Station Blackout (SBO). Regulatory Guide (RG) 1.82 addresses requirements for LOCAs but not for these non-LOCA special events. Conservative evaluation of these special events determined that BFN will require credit for available containment overpressure of up to 2 psi, 7 psi, and 10 psi for the R.HR pumps following an SBO event, ATWS event, and Appendix R event, respectively. A markup of PUSAR Section 4.2.5 indicating changes associated with revised NPSH calculation is provided in .
The pressure difference between the pressure required to provide adequate available NPSH and the calculated minimum wetwell accident pressure can be seen on each graph as the difference between the containment pressure line and the individual ECCS pump pressure required line. As shown on these graphs, adequate margin exists between the COP required and the containment pressure except for the RHR pumps injecting into the broken recirculation system piping during the latter portion of the DBA LOCA Short-Term analysis. Operation of the RHR pumps with a small negative NPSH'margin (approximately 0.3 psi) for a short period of time (< 10 minutes) will not cause significant damage to the RHR pumps injecting into the broken recirculation system piping.
E5-4
FIGURE ACVB.18-1: NSPH REQUIREMENTS FOR DBA LOCA - SHORT TERM a vaayi.UT sup-r- I
- Pre-EPU Peak Suppression Pool Temperature
-Containment Pressure
--- Atmospheric Pressure
- RHR Pump Broken Loop Containment Pressure Required A Pre-EPU RHR Pump Broken Loop Containment Pressure Required
- RHR Pump LPCI Loop Containment Pressure Required 22 160
- I-CS Pump Containment Pressure Required
- Pre-EPU CS Punp Containment Pressure Required 20
-18 140
,-16
- - 01 120 100 I
-12 0 100 200 300 400 500 600 Time (seconds)
E5-5
FIGURE ACVB.18-2: NSPH REQUIREMENTS FOR DBA LOCA - LONG TERM Suppression Pool Temperature
- Pre-EPU Peak Suppression Pool Temperature
-0OContainment Pressure
--- Atmospheric Pressure I CS Pump Containment Pressure Required 200 *22
- Pre-EPU CS Purrp Containment Pressure Required RHR Pump Containment Pressure Required 190- A Pre-EPUJ RH-R Pump Containment Pressure Required 1818 170 X 16 m 0*
E.I 0 4 8 12 16 Time (hours)
E5-6
FIGURE ACVB.18-3: NSPH REQUIREMENTS FOR ATWS 220 40 200 35 180 30 U-eE U
.M. 160 25 f 0
- 0. Li E th 0
140 20 120 15 100 - 10 0 1 2 3 Time (hours)
E5-7
FIGURE ACVB.18-4: NSPH REQUIREMENTS FOR APPENDIX R Suppression Pool Temperature
- Pre-EPU Peak Suppression Pool Temperature
-1Containment Pressure
- - - Atmospheric Pressure
-M*RHR Pump COP Required A Pre-EPU RHR Pump COP Required 30 220-200 - 25 180 -
LA!
2 ~202 140 120 100 -10 0 5 10 15 20 25
.~m
.,,h *t1 E5-8
FIGURE ACVB.18-5: NSPH REQUIREMENTS FOR SBO 220 50 200 40 180 IL-30 Cf a 160 0)
- 0. 0 E th.
0 20 140 120 10 100 0 0 1 2 3 4 Time (hours)
E5-9
NRC Request ACVB.26 Demonstrate with a "realistic" or best-estimate calculation of available net positive suction head for the RHR or core spray pumps, whichever is most limiting, whether credit for containment accident pressure is needed. All input and assumptions should be, to the extent possible, nominal values.
TVA Reply to ACVB.26 In the reply to NRC Request SPSB-A.11 in Enclosure 1 of this letter, the results of several calculations of NPSH for the RHJR and CS pumps are presented which include the use of realistic input parameters. The realistic inputs and assumptions for each case as well as the resultant need for COP are presented in Table SPSB-A.11-2 in Enclosure 1. Using realistic assumptions, COP is not required to ensure adequate NPSH for a DBA-LOCA.
NRC Request ACVB.32 Provide for staff review the NPSH calculations (including the containment calculations) for the Unit 1 core spray and RHR pumps at EPU conditions.
TVA Reply to ACVB.32 Enclcsure 6 provides a copy of Revision 8 of Calculation MDQ0999970046, "NPSH Evaluation of Browns Ferry RHR and CS Pumps." Enclosure 7 provides applicable information from the containment calculation that supports the NPSH evaluations.
E5-10