Text

BFN EPU Containment Overpressure (COP) Credit Risk Assessment, Rev. 2
	ML062090482
Person / Time
Site:	Browns Ferry
Issue date:	07/10/2006
From:	; ERIN Engineering & Research
To:	; Office of Nuclear Reactor Regulation, Tennessee Valley Authority
References
	TAC MC3743, TAC MC3744, TAC MC3812, TVA-BFN-TS-418, TVA-BFN-TS-431 C1320503-6924R2
	Download: ML062090482 (199)
	v • d • e

BFN EPU Containment Overpressure (COP)

Credit Risk Assessment Rev. 2 Performed for:

Tennessee Valley Authority Performed by:

ERIN Engineering and Research, Inc.

July 10, 2006

BFNEPUCOP ProbabilisticRisk Assessment Tennessee Valley Authority Browns Ferry Nuclear (BFN)

BFN EPU Containment Overpressure (COP)

Credit Risk Assessment Rev. 2 Prepared by: Dale: July 10, 2006 Reviewed by-. Date: .July10,2006 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Table Of Contents Section Page EXECUTIVE SUM MARY ............................................................................................. ii

1.0 INTRODUCTION

............................................................................................... 1-1 1.1 Background ......................................................................................... 1-1 1.2 Scope ...................................................................................................... 1-3 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 Assessm ent of NPSH Calculations ......................................................... 3-1 3.2 Probability of Plant State 1 and Plant State 2 ......................................... 3-4 3.3 Pre-Existing Containment Failure Probability .......................................... 3-6 3.4 Modifications to BFN Unit I PRA Models ................................................ 3-8 3.5 Assessment of Large-Late Releases .................................................... 3-11 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 ................................................... 4-10

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 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk 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 LOCA, ATWS and SBO accident scenarios.

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) changes 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 accident calculations to determine under what conditions credit for COP is required to satisfy low pressure ECCS pump NPSH.

2) Revise 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.

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BFN EPUCOPProbabilisticRisk Assessment 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) below 106/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 (2.4E-08/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 107/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 (2.4E-08/yr).

iii iiiC1320503-6924R2 - 7/1 0/2006

BFN EPUCOPProbabilisticRisk Assessment 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 LOCA, ATWS and SBO accident scenarios.

1.1 BACKGROUND

Tennessee Valley Authority (TVA) submitted the BFN extended power uprate (EPU) 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-O01, 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 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-initiatedlicensing basis change requests that go beyond current staff positions may be evaluated by the staff using traditionalengineering analyses as well 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 traditionalengineering analyses, while determining the regulatory acceptability of crediting containment accidentpressure.

Considering the above discussion, please provide an assessment of the credit for containment accident pressure againstthe five key principles of risk-informed decisionmaking stated in RG 1.174 and SRP Chapter 19.

Specifically, demonstrate that the proposed containment accident 1-1 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment 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 numericalrisk 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 containmentpressure 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 containment overpressure (COP),

in the determination of net positive suction head (NPSH) for low pressure ECCS systems following design basis events. Also consistent with other 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 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 and special event 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 NPSH 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:

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BFN EPUCOP ProbabilisticRisk Assessment "Forcertain 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 availableNPSH."

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".

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 (#1-#3, and #5) are not addressed in this report.

This analysis assesses the CDF and LERF risk impact on the BFN Unit I at-power internal events PRA resulting from the COP credit requirement for low pressure ECCS pumps during large LOCA, ATWS and SBO accident scenarios.

External event and shutdown accident risk is assessed on a qualitative basis.

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BFN EPU COPProbabilisticRisk Assessment In addition, a review of the BFN Unit 2 and Unit 3 models is performed to show that the results from the Unit I 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.

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 at 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 states for Level 1 sequences, and release categories for Level 2 sequences.

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.

ISLOCA - 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.

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BFNEPU COPProbabilisticRisk Assessment Large early release frequency - expected number of large early releases per unit of time.

Level 1 - 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 associated 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 on the health of the public (also referred to as a probabilistic risk assessment, PRA).

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).

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.

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BFN EPUCOPProbabilisticRisk Assessment 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 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 1-6 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment 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 RG Regulatory Guide RHR Residual Heat Removal System RPV Reactor Pressure Vessel SBO Station Blackout SMA Seismic Margins Assessment SP Suppression Pool SPC Suppression Pool Cooling SW Service Water TS Technical Specifications 1-7 C1320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment TVA Tennessee Valley Authority WW Wetwell 1-8 C1320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment 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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BFN EPUCOPProbabilisticRisk Assessment Front-line event trees are linked to the end of the Support System event trees for 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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BFNEPU COPProbabilisticRisk Assessment 2.2 STEPS TO ANALYSIS The performance of this risk assessment is best described by the following major analytical steps:

Assessment of NPSH calculations

Estimation of pre-existing containment failure probability

Analysis of relevant plant experience data

Manipulation and quantification of BFN Unit 1 RISKMAN PRA models

Comparison to ACDF and ALERF RG 1.174 acceptance guidelines

Performance of uncertainty and sensitivity analyses

Assessment of "Large Late" Release Impact

Review of BFN Unit 2 and Unit 3 PRAs Each of these steps is discussed briefly below.

2.2.1 Assessment of NPSH Calculations The purpose of this task is to develop an understanding of the BFN EPU NPSH calculations that result in the need to credit containment overpressure for LLOCA, ATWS, and SBO accident scenarios.

The need for COP credit requests is driven by the conservative nature of the accident calculations. The NPSH 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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BFN EPU COPProbabilisticRisk Assessment 2.2.2 Estimation of Pre-Existinq 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 Impact 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 LLOCA NPSH 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 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk 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, ATWS, and SBO scenarios due to inadequate NPSH caused by an unisolated containment coincident with other 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.

All ATWS sequences in the BFN PRA (i.e., transients, LOOP, and IORV initiated ATWS scenarios) are modified to model the COP credit impact.

SBO accident sequences in the BFN PRA are modified to require COP credit for adequate ECCS NPSH upon recovery of AC power after 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The modeling and quantification is performed consistent with common RISKMAN modeling techniques.

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BFNEPUCOPProbabilisticRisk Assessment 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 to the 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 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 Unit 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.

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BFNEPU COP ProbabilisticRisk Assessment 2.2.8 Review of BFN Unit 2 and Unit 3 PRAs The base analysis uses the BFN Unit 1 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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BFN EPUCOPProbabilisticRisk Assessment Figure 2-1 RG 1.174 CDF RISK ACCEPTANCE GUIDELINES t

RegiogioI!I 1

Region Ii 10-6 10D3 ID' CDF -P 2-8 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Figure 2-2 RG 1.174 LERF RISK ACCEPTANCE GUIDELINES

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-io-7 10-11 10-5 LERF10 2-9 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Section 3 ANALYSIS This section highlights the major qualitative and quantitative analytic steps to the analysis.

3.1 ASSESSMENT OF NPSH CALCULATIONS The purpose of this risk assessment is due to the fact that the conservative nature of the 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 can show that no credit for COP is required.

Special events such as ATWS and SBO are not necessarily analyzed using the conservative assumptions as for design basis events such as LOCA. Review of the BFN ATWS and COP NPSH calculations shows that some level of containment overpressure credit would be required even if more realistic inputs are used in the calculations. As such, increasing the degree of realistic treatment in the special events NPSH calculations is not expected to eliminate the need for containment overpressure credit. This is not true for the LLOCA NPSH calculations, which show that use of more realistic values for a variety of input parameters result in showing no need for COP credit. As such, the following discussions address sensitivities to the LLOCA NPSH calculations. The ATWS and SBO scenarios analyzed in this risk assessment assume that COP credit is always required.

The GE DBA LOCA calculation makes the following conservative assumptions, among others, regarding initial plant configuration and operation characteristics:

Initial reactor power level at 102% EPU

Decay heat defined by 2 sigma uncertainty 3-1 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment

2 RHR pumps and 2 RHR heat exchangers in SPC

All pumps operating at full flow 0 River water temperature at 95 0 F

Initial suppression pool temperature at 95 0 F 0 Initial SP water volume at minimum technical specification level 0 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

" 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 during LLOCA accidents. 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 3-2 C1320503-6924R2 -7/10/2006

BFNEPU COPProbabilisticRisk Assessment 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 LLOCA NPSH 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 the risk calculation, this assessment takes no probabilistic credit for these issues.

" 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 for NPSH 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 2 RHR pumps and heat exchangers are placed in SPC operation, initial SP volume at 123,500 ft3, and river water temperature is at 68 0 F, then torus water temperature must be above 87°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 SPC operation, initial SP volume at 123,500 ft3, and river water temperature is at 85 0 F, then torus water initial temperature must be above 86 0 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 LOCAs if the containment is unisolated:

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BFNEPU COP ProbabilisticRisk Assessment 0 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 68 0F, and torus water initial temperature above 870 F.

Plant State 2: 100% EPU initial power level, nominal decay heat, 2 RHR pumps and heat exchangers in SPC, initial SP volume at 123,500 ft3 , river water temperature of 86°F, and river water initial temperature above 85 0 F.

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.

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 during LLOCA scenarios. 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 1 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 3-4 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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 68 0 F is determined from the BFN experience data statistical analysis summarized in Appendix C.

Given river water temperature 68 0F, the conditional probability that the torus water temperature is 870 F is determined from the BFN experience data statistical analysis summarized in Appendix C.

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 68 0 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 I at the time of the DBA LOCA is 5E-3 x 5.64E-1 x 4.42E-1 x 1.45E-2 = 1.8E-5.

3-5 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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 85 0 F 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 87 0 F, is taken to be 1.0. This is reasonable (refer to Figure C-1).

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 85°F 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 1.64E-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., EPRI Report 1009325, Risk Impact of Extended Integrated Leak Rate Testing Intervals, 12/03). [2]

3-6 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment 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-I, 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 Reference [1], the BFN Unit 2 and Unit 3 primary containment ILRT results from the 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.

3-7 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 3.4 MODIFICATIONS TO BFN UNIT I PRA MODELS 3.4.1 PRA Model Modifications for LLOCAs 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)

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

" LPCI and LPCS inter-unit crossties are defeated because the pumps crosstied from the Unit 2 would be aligned to the Unit 1 suppression pool and would experience the same NPSH conditions as the Unit 1 pumps.

3-8 C1320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment 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 20La. 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 NPSH 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.

3.4.2 PRA Model Modifications for ATWS and SBO For the ATWS scenarios, COP is modeled as always required for LP ECCS pump NPSH; if COP is unavailable, all LP ECCS pumps drawing from the torus are modeled as failed due to insufficient net positive head. For the SBO scenarios, overpressure is modeled as required after AC power is recovered at t=4 hours.

The following ATWS and SBO initiated sequences in the BFN Unit 1 PRA were modified:

" Turbine Trip ATWS (TTA)

LOSP ATWS (LOSPA)

Loss of Condenser Heat Sink ATWS (LOCHSA)

Inadvertent Opening of SRV ATWS (IOOVA) 3-9 C1 320503-6924R2 -7110/2006

BFN EPUCOPProbabilisticRisk Assessment

" Loss of Feedwater ATWS (LOFWA)

Loss of Offsite Power (LOSP)

Loss of 500kV Switchyard to Plant (L500PA)

Loss of 500kV Switchyard to Unit (L500U)

Similar to the event tree model changes for LLOCA, the ATWS and SBO event trees were modified in order to determine the status of containment integrity (node CIL) prior to questioning the status of low pressure systems drawing from the torus. In the ATWS event trees, failure of the CIL node leads directly to failure of LP ECCS pumps without questioning additional NPSH pre-conditions as is done for the LLOCA scenarios. The same is true for the SBO scenarios, but the scenarios also require COP only after AC power is recovered.

In addition, as discussed previously for the LLOCA scenarios, LPCI and LPCS inter-unit crossties are defeated.

Refer to Appendix E for print-outs of the revised ATWS and SBO event trees.

3.4.3 Quantification of Revised 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=0 and all RPV injection is lost in the PRA "Early" time frame. Core damage occurs at approximately one hour for the LLOCA and ATWS COP accidents, and in approximately 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the SBO COP accidents. As such, the additional CDF contributions created by this model manipulation are also all LERF release sequences (i.e., deltaCDF equals deltaLERF). This is a conservative assumption as it assumes that the pre-existing containment leakage of 20La used in the base quantification is representative of a LERF release. Reference [2] determines that a containment leak representative of LERF is >600La.

3-10 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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.

3-11 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Table 3-1

SUMMARY

OF LLOCA NPSH DETERMINISTIC CALCULATIONS(5)

CL)

E -( CL 4) U e.

0 o. C. ) 0 2 C). M. - CI-,

coca .0 a-0c EC . 0.

E) -G ~C QE FC =30 I=u -.

6 -a) C 6 [2 Io (00.

'0 0 (0 C Case( 11

~~C Caees.pto a - =o U) 0 u> - zL

.0 A" ZL C2 g. 2 U)~ ES 4J U)L~0- o: U)

Base Case(2) EPU Licensing Calculation - I102% ANSI5.1 Full (GE) DBALOCA EPU w/2a 95 95 2 2 2 4000 I 223 2 1121,500 Yes No 187.3 Yes design Case 1(2) DBA Calculation but No 102% ANSI 5.1 Full (GE) Single Failure EP w/2 95 4000 121,500 Yes No 166.4 No design 2

Case la( ) DBA Calculation but 3 RHR 102% ANSI 5.1 (GE) Pumps inSuppression Pool EPU wa 95 Full 121,500 4000 Yes No 175.0 Yes Cooling design Case la (TVA) DBA Calculation but 3 RHR ANSI

[This case is Pumps inSuppression Pool 102% Full EPUL 5.1 95 95 design 4000 121,500 Yes No 175.0 Yes benchmarked Cooling w/2cr against Case la (GE)1 Case lb 100% Initial Power, RHRSW (TVA) 890F, 3 Pumps in Full Suppression Pool Cooling, K 95 4000 Yes No 171.0 No design Value 225,4 CS Pumps Case 1c 100% Initial Power, RHRSW (TVA) 90oF, 3 Pumps in Suppression Pool Cooling, K 95 Full design 4000 Yes No 170.5 No Value 225,4 CS Pumps, Nominal SP WL 3-12 C1320503-6924R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk Assessment Table 3-1

SUMMARY

OF LLOCA NPSH DETERMINISTIC CALCULATIONS(5)

=BFN -P -O roaiisi- -esmnis Cx 00 d) 0 0 W

0L. CM

- F 2L C 8-0.

E~0 E~- 0) 06 EL

'd d) EDo; E2a Q) 0.

E CMo E0. 0X.

4) 4). ci) 0)

CO)

Z Lo cc) 4)

E4 cr.~

=- CU CL'

.~2 C',

0 K 80. 0

CL

'U Case(0) Case Description zo0 z .-

E)

Case ld DBA Calculation but RHRSW (TVA) 900F, SP Initial Temp 91OF, 3 102% ANSI Pumps in Suppression Pool EPU 5.1 W/ Full design 4000 121,500 Yes No 171.0 No Cooling, KValue 225,4 CS 2a Pumps Case le DBA Calculation but RHRSW (TA) 920F, SP Initial Temp 900F, 3 102% ANSI Pumps in Suppression Pool EPU 5.1 w/ Full design 4000 121,500 Yes No 171.1(0) No Cooling, K Value 225,4 CS 2a Pumps Case 2 DBA Calculation but Initial 102% ANSI 5.1 Full (GE) SW Temperature = 85OF EPU w/2a design 2 2 4000 223 2 121,500 Yes No 182.0 Yes Case 2 (TVA) AS

[This case is DBA Calculation but SW 102% ANI 95 2 Full 2 2 4000 223 2 121,500 design Yes No 182.2 Yes benchmarked Temperature = 85oF EPU w/2(5 against Case 2 (GE)]

Case 2a DBA Calculation but Initial 102% ANSI 5.1 (GE) 95 2 Full 2 2 4000 223 2 121,500 Yes No 177.6 Yes SW Temperature = 75OF EPU w/2a design Case 2b(2) DBA Calculation but Initial 102% ANSI 5.1 (GE) SW Temperature = 70OF EPU k 95 2 Full 2 2 4000 223 2 121,500 Yes No 175.9 Yes w"2a design Case 2c DBA Calculation but Initial 102% AN~51 Full (GE) SW Temperature = 65OF EPU design 2 2 4000 223 2 121,500 Yes No 174.3 Yes 3-13 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Table 3-1

SUMMARY

OF LLOCA NPSH DETERMINISTIC CALCULATIONS(5) r r Y F F Y = = , -

ca.

') U)

El U)

Q.

E Wo U. E E

o.S CL I2-C,_=

0. CL2 ca CLI cc~ '2o 0.

(U Ca

'2) IU '2E2 L0~ (D 9 Z 0.1 0.E 3: )CM

'2) cc~ E -0 U) WCr ca 0 EC4

'%2)- do. a~ )

ci) .

Case(') Case Description zo0 z G~ CU Case 2d DBA Calculation but SW (TVA) Temperature = 650F, SP 102% ANS Initial Temp 880F, Nominal EPU 5.1 2 Full design 2 2 4000 00 223 223 2 2 Yes No 170.6 No SPWL I Case 2e DBA Calculation but SW (TVA) Temperature = 650F, SP 102 ANS Full 2 design 2 2 4000 1 223 2 121,500 Yes No 170.7 No Initial Temp 870F EPU 5.

Case 2f DBA Calculation but SW (ITVA) Temperature = 680F, SP 102/ ANS Initial Temp 870F, Nominal EPU 5.1 2 Full 2 2 4000 223 2 Yes No 171.1(') No design SPWl_

Case 3 DBA Calculation but Initial Full (GE) SP Temperature = 850F 2 design 2 2 4000 223 2 121,500 1 Yes No 183.8 Yes Case 4 100% Initial Power, Minimum (GE) SP Level, and No Heat Sink 2 Full 2 2 4000 2 121,500 Yes No 177.0 Yes Credit design Case 4 (TVA) 100% Initial Power, Minimum

[This case is SP Level, and No Heat Sink 2 Full 2 2 4000 bench-marked 2 design 2 121,500 Yes No 177.1 Yes Credit against Case 4 (GE)]

Case 4a 100% Initial Power, Nominal (GE) SP Level, and Heat Sink 2 Full 2 2 4000 2 0 Yes 174.7 Yes Credit design 3-14 C1320503-6924R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk Assessment Table 3-1

SUMMARY

OF LLOCA NPSH DETERMINISTIC CALCULATIONS(')

BFN EPU COP Probabilistic* Risk

- Assessment p = p -

0. -

2L 0 C0 w

0 2:

E U) CL z0.

ccCL a)

Ix 2-5 S O)C.

LJ Q~. >ý L..

(D U) 0:

CE (D

E U-0..

0)r E0

-C -

~= 0'f) U) .

O.*o 1 (r e CL Case( ) Case Description co IT U) .7 z . 17.

Case 4b(2) 100% Initial Power, Minimum Full 2 2 4000 178.9 Yes (GE) SP Level, and Heat Sink 2 design 2 121,500 Yes Credit Case 4c(2) 100% Initial Power, Minimum (GE) SP Level, Heat Sink Credit, and SW Temp. that results in Full 2 2 4000 2 design 2 121,500 Yes 175.8 Yes Peak SP Temp. equal to/less than 176 0F Case 4d 100% Initial Power, RHRSW Full 2 2 4000 (TVA) 860F, SP Initial Temp 920F, K 2 design 2 121,500 1 Yes No 177.0 Yes Value 225 Case 4e 100% Initial Power, RHRSW (TVA) 860F,SP Initial Temp 900F, K 2 Full 2 2 4000 2 121,500 Yes No 176.1 Yes design Value 225 Case 4f 100% Initial Power, RHRSW Full 2 2 4000 (TVA) 860F, SP Initial Temp 900F, K 2 design 2 Yes No 175.6 Yes Value 225, Nominal SP WL Case 4g 100% Initial Power, RHRSW Full 2 2 4000 (TVA) 860F, SP Initial Temp 900F, K 2 design 2 Yes No 173.1 Yes Value 241, Nominal SP WL.

Case4h 100% Initial Power, RHRSW (TVA) 850F, SP Initial Temp 900F, K Full 2 2 2 4000 2 Yes No 175.1 Yes Value 225, Nominal SP WL design 3-15 -C1320503-69241R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Table 3-1 5

SUMMARY

OF LLOCA NPSH DETERMINISTIC CALCULATIONS( )

- F- - - ~- ~ = p (6

CL 0C C E 0

=.-

IL 0) F E

-g 2 28a. (26 J2 CL Ce 4) 0 0E. :3-o.E w6 01. E Ow 0L. Cs 0 0 .2CL E~~ .0 2 E) 0.

0 .. O C0 in Q0 Z Lc Ix0 0.'

ME I-.

LU) 1 0 0 809 Case( ) Case Description Cl) I-Case 4i 100% Initial Power, RHRSW Full (TVA) 850F, SP Initial Temp 86 0F, K 2 2 2 4000 2 Yes No 170.8 No Value 241, Nominal SP WL design Case 4j 100% Initial Power, RHRSW Lil (TVA) 850F, Value SP 241,Initial Temp 880F, K Nominal SIDWL 2

design 2 2 4000 2 Yes No 171.0 No 0

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 0 F.

(4) This value is acceptable for demonstrating sensitivity analysis results.

(5) Shaded areas in the table "highlight" differences from the Base Case.

3-16 C1320503-6924R2 -7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Section 4 RESULTS 4.1 QUANTITATIVE RESULTS The results of the base quantification of this risk assessment case are summarized in Table 4-1.

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 (refer 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 C1320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment 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 Event Tree Analysis Module. Appropriate probability distributions for each uncertain parameter in the analysis is determined and included in the Data Module. The System 4-2 C1320503-6924R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk Assessment 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 Unit 1 base EPU PRA and the Unit 1 EPU COP Credit LLOCA quantification. The results are shown in Table 4-2.

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. Based on this result, parametric uncertainty analysis for the ATWS and SBO accidents is not necessary as the conclusion would be the same (i.e., very little distribution change, such that delta CDF and delta LERF results would remain well within the RG 1.174 guidelines for "very small" risk impact).

4-3 C41320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment 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 EPRI 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-Existinq Containment Leakage Size/Probability The base case analysis assumes a pre-existing containment leakage pathway leakage size of 20La that would result in defeat of the necessary containment overpressure credit.

4-4 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment 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. Refer to Table F-1 for the CIL event tree node failure probability used.

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 LLOCA 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 C1320503-6924R2 - 7/10/2006

BFNEPUCOP ProbabilisticRisk Assessment 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 LLOCA scenarios with only 1 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 mode during a LLOCA, 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):

" Sum of BFN PRA Large LOCA initiator frequencies: 3E-5/yr

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 102% EPU initial power level: 5E-3 (same as base analysis)

" Probability of containment isolation failure given an accident initiator: 3E-3 (nominal from base analysis)

" Probability of river water temperature >90°F at any given time: 9E-2 (nominal value based on Table C-1. Although the river temperature has not exceeded 90°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°F given river water temperature > 90°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 C1320503-6924R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk 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-13/yr This additional contribution to the calculated deltaCDF from a 3 RHR pump LLOCA case is non-significant in comparison to the 2 RHR pump LLOCA 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):

Sum of BFN PRA Large LOCA initiator frequencies: 3E-5/yr

" Likelihood of 4 RHR pumps and 4 heat exchangers in SPC during Large LOCA: 1.0 (nominal estimate)

Probability of 102% EPU initial power level: 5E-3 (same as base analysis)

Probability of containment isolation failure given an accident initiator: 3E-3 (nominal from base analysis)

" Probability of river water temperature > 100°F at any given time: IE-3 (estimate based on Table C-1. Although the river temperature has not exceeded 90°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 > 95 0 F given river water temperature > 100°F: 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 1E-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 LLOCA case is non-significant in comparison to the 2 RHR pump LLOCA case.

4-7 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Summary of Modeling Uncertainty Results The modeling uncertainty sensitivity cases are summarized in Table 4-3.

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.

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 basis. 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-4-8 C1 320503-6924 R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment 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.

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/Tornadoes

External Floods

" Transportation and Nearby Facility Accidents

Other External Hazards 4-9 C1320503-6924R2 -7110/2006

BFNEPUCOP ProbabilisticRisk Assessment 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 concluded 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.

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 CDF 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 the Unit 3 LLOCA scenarios were quantified to determine the ACDF impact. The result for Unit 3 was a deltaCDF of 1.9E-9/yr, which is comparable to the U-1 LLOCA COP delta CDF contribution of 1.5E-9/yr for sensitivity case #2 . The revised BFN Unit 3 PRA RISKMAN model supporting this review is archived in file 4-10 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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. 2 analysis using the U-1 model, the U2/U3 assessment discussed above was not re-performed as the conclusion would be the same.

4-11 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Table 4-1 BASE CASE RESULTS TYPE DESCRIPTION Total CDF Total LERF ACDF(2), (3) ALERF(2), (3)

LLOCA(1) Large LOCAs. All large LOCA initiated scenarios (except ISLOCAs 1.77E-06 4.41 E-07 1.39E-09 1.39E-09 and LOCAs Outside Containment, because these result in deposition of decay heat directly outside the containment and not into the suppression pool).

ATWS°l) Transient without SCRAM. All PRAATWS scenarios (i.e., transients, 1.77E-06 4.48E-07 8.17E-09 8.17E-09 LOOP, and IORV ATWS scenarios) modified to require COP credit.

Low pressure ECCS pumps failed ifcontainment isolation is failed.

SBO°l) Station black out with recovery of power after 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Low pressure 1.78E-06 4.54E-07 1.47E-08 1.47E-08 ECCS pumps failed when AC power recovered if containment isolation is failed.

TOTAL Results for LLOCA, ATWS and SBO 1.79E-06 4.64E-07 2.43E-08 2.43E-08 Notes:

(1) The results in the top three rows are for each identified group of accident scenarios quantified in isolation and the resulting impact on CDF and LERF. The combined CDF and LERF impact for all three accident scenario types Is provided in the bottom row.

(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.

(3) The results presented above are conservative due to the nature of the RISKMAN quantification. The addition of new nodes or top events to event trees (as is done in this analysis) causes previously existing sequences to split into two or more new sequences. The quantification initiator cutoff limit in the COP calculations was reduced (from the base cutoff of 1E-1 2 to 1E-1 3) to capture the new sequences added to the model. The reduced cutoff limit in the revised model captures the new low frequency sequences, but also results in capturing sequences that are truncated in the base BFN model; as such, the resultant ACDF and ALERF values (which are calculated as the new PRA value minus the base PRA value) shown here are overstated.

4-12 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Table 4-2 PARAMETRIC UNCERTAINTY ANALYSIS RESULTS BFN Unit I BFN Unit I Statistic Base CDF COP LLOCA CDF(1 )

5% 4.71E-7 5.15E-7 50% 1.23E-6 1.23E-6 MEAN 1.77E-6 1.77E-6 95% 4.72E-6 4.47E-6 Notes:

(1) Parametric uncertainty analysis performed on the LLOCA accident sequence impact. Similar results expected for ATWS and SBO sequences (i.e., little change, such that delta CDF and delta LERF results would remain well within the RG 1.174 guidelines for "very small" risk impact).

4-13 C1320503-6924R2 - 7/1012006

BFNEPUCOPProbabilisticRisk Assessment Table 4-3

SUMMARY

OF SENSITIVITY QUANTIFICATIONS Case Description CDF LERF ACDF(2)- (3) ALERF(2). (3)

Base(1 ) Base Case Quantification (20 La leak size) 1.791 E-06 4.640E-07 2.4E-08 2.4E-08 1(1) Pre-Existing Containment Leakage Sufficient to Fail COP Credit 1.771 E-06 4.441E-07 4.4E-09 4.4E-09 Defined by 100La 2(l) Assume Low Suppression Pool Water Volume (123,500 ft3) Exists 1.791E-06 4.642E-07 2.4E-08 2.4E-08 100% of the Time 3(1) Expansion of Containment Isolation fault tree to Encompass Smaller 1.793E-06 4.656E-07 2.6E-08 2.6E-08 Lines (approximate by multiplying Cont. Isol. failure probability by 10x) 4(1) Assume Initial Power Level and Water Temperature and Level Pre- 1.793E-06 4.661 E-07 2.6E-08 2.6E-08 Conditions Exist 100% of the Time 5(1) Combination of Cases #3 and #4 1.798E-06 4.708E-07 3.1 E-08 3.1 E-08 6 Incorporation of "3-RHR pumps in SPC" and "4-RHR pumps in SPC" 1.791 E-06 4.640E-07 2.4E-08 2.4E-08 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.

(3)

The results presented above are conservative due to the nature of the RISKMAN quantification. The addition of new nodes or top events to event trees (as is done in this analysis) causes previously existing sequences to split into two or more new sequences. The quantification initiator cutoff limit in the COP calculations was reduced (from the base cutoff of 1E-12 to 1E-13) to capture the new sequences added to the model. The reduced cutoff limit in the revised model captures the new low frequency sequences, but also results in capturing sequences that are truncated in the base BFN model; as such, the resultant ACDF and ALERF values (which are calculated as the new PRA value minus the base PRA value) shown here are overstated.

4-14 C1320503-6924R2 - 7110/2006

BFN EPUCOPProbabilisticRisk Assessment 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 LOCA, ATWS and SBO accident scenarios.

The need for COP credit requests is driven by the conservative nature of 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 10S6 /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 (2.4E-08/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 (2.4E-08/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 5-1 C1 320503-6924 R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment determining adequate NPSH for low pressure ECCS pumps during DBA LOCA, ATWS and SBO accidents is acceptable from a risk perspective.

The conclusion that the risk impact from the EPU COP credit is very small, applies to BFN Unit I as well as BFN Units 2 and 3.

5-2 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment REFERENCES

[1] "Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specifications (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 Integrated Leak Rate Testing Intervals, EPRI Report 1009325, Final Report, December 2003.

[3] "Project Task Report - Browns Ferry Units 1, 2 & 3 EPU, RAI Response - NPSH Sensitivity Studies", GE Nuclear Energy, GE-NE-0000-0050-00443-RO-Draft, 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 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk 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 I 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 risk assessment was the most recent internal events risk model for the BFN Unit I 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:

A-1 C1320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment

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-I. 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 C1320503-6924R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk Assessment 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 EPRI 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 1 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 LERF:

A-3 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment

" 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 Cl1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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 I 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 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment 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 method of assessing the technical quality of the PRA for its potential applications. 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.

A-6 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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 low pressure ECCS pumps.

A-7 CA1320503-6924 R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk 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 1.95E-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 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Table A-1 INITIATING EVENTS FOR BFN PRA Initiator Mean Frequency Category (events per year)

Large LOCA - Core Spray Line Break Loop I 1.68E-6 Loop II 1.68E-6 Large LOCA - Recirculation Discharge Line Break Loop A 1.18E-5 Loop B 1.18E-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 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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-10 C1320503-6924R2 - 7/10/2006

BFNEPU COP ProbabilisticRisk Assessment Table A-2 BFN PRA MODELED SYSTEMS Suppression Pool / Vapor Suppression Turbine Bypass and Main Condenser A-11 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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

- Transient

- 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)

- Transients

- SBO

- LOCA

- ATWS

- Special

- ISLOCA/BOC

- Internal Floods

Success Criteria and Bases

Interface with EOPs/AOPs

Accident Sequence Plant Damage States

Documentation A-1 2 C1320503-6924 R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Table A-3 PRA PEER REVIEW TECHNICAL 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-13 C1320503-6924R2 - 7110/2006

BFN EPUCOP ProbabilisticRisk Assessment Table A-3 PRA PEER REVIEW TECHNICAL 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

- EDG Mission Time

- Repair and Recovery Model

- SORV

- LOOP Given Transient

- 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

" Dependence among Actions

Documentation A-14 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Table A-3 PRA PEER REVIEW TECHNICAL 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

- Transient

Containment (pressure and temperature) 0 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-1 5 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment Table A-4 PRA CERTIFICATION TECHNICAL ELEMENTS FOR LEVEL 2 PRA ELEMENT CERTIFICATION SUB-ELEMENTS PRA ELEMENT CERTIFICATION SUB-ELEMENTS Containment Performance Analysis Guidance Document Success Criteria LI/L2 Interface Phenomena Considered Important HEPs Containment Capability Assessment End state Definition LERF Definition CETs Documentation A-16 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment 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 Integrated 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. EPRI 1009325 employed industry experts to review and categorize the industry events, and then various statistical methods were used to assess the data. The resulting probabilities as a function of pre-existing leakage size are summarized here in Table B-1.

The EPRI 1009325 study used 100La 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.

B-1 C1320503-6924R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk Assessment 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 EPRI 1009325) conservatively recommended use of 1OLa 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 EPRI 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-I, 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.

B-2 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Table B-1 PROBABILITY OF PRE-EXISTING UNISOLABLE CONTAINMENT LEAK [2]

(as a Function of Leakage Size)(I)

Leakage Size Mean Probability of (La) Occurrence I 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.

B-3 C1 320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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 the 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 3 020/3 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 would 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 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk 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 closely correlated in the warmer months when river water temperature is above approximately 70 0 F.

The 6.1 years of temperature data was categorized into 5-degree temperature bins ranging from 50°F to 99°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 C-3 presents the histogram for the torus water temperature.

The histogram information was then used in a statistical analysis software package (CrystalBall, 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 number of curve fits. The best fit for the temperature data is a normal distribution that is truncated at user-defined upper and C-2 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment 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 O°F degrees). To constrain the distributions, the following user-defined upper and lower bounds were used:

River water temperature lower bound of 32 0 F (no data points in the 6.1 years of data reached 32 0 F, only a single data point reached 35 0 F)

" River water temperature upper bound of 95 0 F (no data points in the 6.1 years of data exceeded 90 0 F)

" Torus water temperature lower bound of 55 0 F (no data points in the 6.1 years of data reached lower than 57 0 F)

Torus water temperature upper bound of 95 0 F (only a single data point in the 6.1 years of data reached 93 0 F)

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 exceedance 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 Temperature One of the parameters used in this risk assessment is the conditional probability that the torus water temperature is greater than or equal to 87 0 F given river water temperature is greater than or equal to 68 0 F. Plant data for Units 2 and 3 were reviewed to C-3 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk 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 68 0F, and of those records, the number of records where the torus temperature exceeded 87 0 F.

River >= 68F Torus>=87F Cond Prob.

Unit 2 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 87 0 F when the river temperature is greater than 68 0 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.

C-4 C1320503-6924R2 - 7/1012006

BFN EPUCOPProbabilisticRisk Assessment 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 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Figure C-1 CHRONOLOGICAL VARIABILITY IN RIVER WATER AND TORUS WATER TEMPERATURES Pool Temp - River Temp 95 -

85 55--

45 35 I 01/01/99 01/01/00 12/31/00 12/31/01 12/31/02 12/31/03 12/30/04 12/30/05 12/30/06 Date C-6 C1320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment Figure C-2 RIVER WATER TEMPERATURE HISTOGRAM 400 350 300 4DU 200 150 150 Ii I/

.- U 100 50 I I

32.5 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 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Figure C-3 TORUS TEMPERATURE HISTOGRAM 700 600 500 400 a,

0 200 100 0

52.5 57.5 62.5 67.5 72.5 77.5 82.5 87.5 92.5 97.5 Temperature C-8 Cl 320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Figure C-4 STATISTICAL RESULTS FOR RIVER 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: 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 River Temperature 50.000 Trials Frequency Chart 0 Outliers

.012 613 Al 459.1 va 0.0 30.00 47.50 65.00 12.50 100.00 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 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment 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 50,000 Trials Frequency Chart 0 Outliers

.01153

.00 429..

- ~28&-5 000 -0 0,00 65.00 75.00 8&00 95.00 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-10 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Figure C-6 RIVER WATER TEMPERATURE EXCEEDANCE PROBABILITY 1.OE+O 1.OE-1 0

w z

0 xi 1.OE-2 1.OE-3 25 34 37 41 44 48 51 55 58 62 65 69 72 76 79 83 86 90 93 97 100 RIVER WATER TEMPERATURE (F)

C-1l1 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Figure C-7 TORUS WATER TEMPERATURE EXCEEDANCE PROBABILITY i.0E+0 1.OE-1 Z

0 i.

w 0

z w

w xw 1.OE-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 C-12 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Figure C-8 SUPPRESSION POOL LEVEL HISTOGRAM 450 400 350 300 I-I-

'I 250 jI-,

200 150 100 50 Ij 0 - =1 11-11

-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-13 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment 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 3/7/06 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 Fomcwt: Nomn -Toras Levd S0O00Trks Fsq-Wohjt 1190utaes

.025 1248 5..06 -53.

000- 1, A 11 0

-6.50 -5.12 --3375 37 -6.00 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 C1320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment Figure C-10 SUPPRESSION POOL WATER LEVEL PROBABILITY (Probability that Level Below Value of Interest) 1.0E+0 1.0E-1 0

1.OE-2 A 1.OE-3 K

-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 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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 2.41E-01 85 1.64E-01 86 1.40E-01 90 8.46E-02 95 9.15E-03 100 0.OOE+00 C-16 C1320503-6924R2 -7/10/2006

BFNEPUCOP ProbabilisticRisk Assessment Table C-2 TORUS WATER TEMPERATURE EXCEEDANCE PROBABILITIES Temperature (OF) I Exceedance Probability 30 1.OOE+00 35 1.OOE+00 40 1.OOE+00 45 1.OOE+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 1.01E-02 100 0.OOE+00 C-17 C1320503-6924R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk Assessment Table C-3 SUPPRESSION POOL WATER LEVEL PROBABILITY (Probability that Level Below Value of Interest)

Level (inches) Probability

-6.50 1.10E-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 9.10E-03

-5.79 1.08E-02

-5.73 1.30E-02

-5.70 1.45E-02(1 )

-5.68 1.55E-02

-5.62 1.83E-02

-5.57 2.11E-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 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 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment 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 1.19E-01

-4.74 1.31E-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 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk 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 8.00E-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 9.61E-01

-2.10 9.66E-01

-2.05 9.70E-01

-1.99 9.74E-01

-1.94 9.78E-01

-1.88 9.81E-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 C1320503-6924R2 -7110/2006

BFN EPUCOPProbabilisticRisk Assessment 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 1.00E+00 Note to Table C-3:

(1) A conservative probability value corresponding to -5.70" (123,500 ftW) instead of -5.90" (123,250 ft 3) was used in the base case quantification.

C-21 C1320503-6924R2 - 7110/2006

BFN EPUCOPProbabilisticRisk 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from accident initiation

2) Intermediate (I) Greater than or equal to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , but less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

3) Late (L) Greater than or equal to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The definition of the timing categories is relative to the timing of the declaration of a General Emergency and based upon past experience concerning offsite accident response:

D-1 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment

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.

0 >24 hours are times at which the offsite measures can be assumed to be fully effective.

Magnitude Categorization 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 Csl released, the BFN release magnitude classification is as follows:

D-2 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment Release Magnitude Fraction of Release Csl Fission Products High greater than 10%

Medium/Moderate 1 to 10%

Low 0.1 to 1.0%

Low-Low less than 0.1%

Negligible much less than 0.1%

D.2 EPU 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 24 hrs 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 (for the LLOCA and ATWS accidents) to approximately six hours (for the SBO accidents).

In summary, there is no change in the frequency of Large-Late releases due to the credit of COP in DBA LOCA, ATWS and SBO scenarios.

D-3 C1320503-6924R2 - 7110/2006

BFN EPU COPProbabilisticRisk Assessment 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 1 PRA RISKMAN models for this risk assessment.

E.1.1 LLOCA Event Tree Changes The Level 1 large LOCA event trees were modified for this risk assessment to question the status of containment integrity first in the tree. 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 RHR pumps in SPC.

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 CET1, it too was renamed. The event tree names were revised as follows:

Original Event New Event Tree Tree I Description CET1 CETN1 Containment Event Tree I 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 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment In the containment event tree, top event CIL was replaced with a dummy top event, CILDUM, 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.OOE+00). Otherwise, the success branch is assigned via split fraction CILDS (O.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 just under 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 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment 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 NPSH1 is the split fraction probability resulting from quantification 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 CIL=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 top event. Note that drywell spray failure is captured by the event tree structure (i.e., if LPCI loops I and II are failed, then drywell spray is never asked in the event trees).

In addition, LPCI and LPCS inter-unit crossties are defeated because the pumps crosstied from the Unit 2 would be aligned to the Unit 1 suppression pool and would experience the same NPSH conditions as the Unit 1 pumps.

E.1.2 ATWS and SBO Event Tree Changes For the ATWS scenarios, COP is modeled as always required for LP ECCS pump NPSH; if COP is unavailable, all LP ECCS pumps drawing from the torus are modeled as failed due to insufficient net positive head. For the SBO scenarios, overpressure is modeled as required after AC power is recovered at t=4 hours.

Similar to the event tree model changes for LLOCA, the ATWS and SBO event trees were modified in order to determine the status of containment integrity prior to E-3 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment questioning the status of low pressure systems drawing from the torus. Each of the original event trees was copied and renamed (appending an N) for use in the analysis.

Original Event New Event Tree Tree Description ATWS3 ATWS3N ATWS Event Tree ATWS4 ATWS4N ATWS Event Tree LPGTET LPGTETN Low Pressure General Transient Event Tree Note:

1. Event trees ATWSI and ATWS2 exist in the BFN PRA, but they do not contain nodes for LP ECCS pumps and thus do not require modification for this risk assessment.

2. It was not necessary to modify event tree HPGTET, High Pressure General Transient, for this risk assessment.

The same revised containment event tree discussed previously for the LLOCA scenarios is also used for the ATWS and SBO scenarios.

The containment isolation top event (CIL) added to the above revised ATWS and SBO event trees is the same one discussed previously for the LLOCA scenarios. The event tree split fraction rules were modified to fail the low pressure systems (top events) if the containment isolation top event fails (CIL).

In addition, as discussed previously for the LLOCA scenarios, LPCI and LPCS inter-unit crossties are defeated.

An additional requirement was used for modeling COP credit impacts for SBO scenarios. Two top events model recovery of AC power: 1) one top event (EPR30) models AC recovery at t=30 minutes; and 2) another top event (EPR6) models AC recovery at t=6 hours. The BFN is not currently designed with 4-hr SBO scenarios; as such, the 6-hr SBO sequences are used as a surrogate to model the COP impact on SBO sequences after AC power recovery at t=4hrs. Event tree node EPR6 (included in E-4 C1 320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment event tree HPGTET) is checked for success prior to requiring COP (i.e., SBO sequences with failure of AC recovery are not modified to question COP issues).

To quantify the impact of the COP requirement on ATWS and SBO, the following ATWS and SBO initiating events were quantified through the revised event tree structures discussed above:

ATWS Initiators LOSP Initiators IOOVA LOSP LOCHSA L500PA LOFWA L500U LOSPA TTA E-5 C1 320503-6924 R2 - 7/10/2006

MODEL Name: UIERIN Page No. 1 oft Event Tree: LLCSN.ETI 12:22:35 February 14, 200O IE CIL NPSH RPSM RPSE TOR TnP IVC LPCI LPCII CS SI 0) 0 C> 0 C.,

C)

MODEL Name: UIERIN Page No. 2 of 2 Event Tree: LLGSN.ETI 12:22:35 February 14, 2006 OSPC SPI SpIl SPC ODWS DWS B B#

1 1 2 2 3 3 4 4 X1 5 5-8 X1 6 9.12 X1 7 13-16 8 17 9 18 10 19 I 11 20

......................................................................................................................... X2 12 21-38 13 39

......................................................................................................................... X2 14 40-57 rn 15 58

-4 I 16 59

......................................................................................................................... X2 17 60-77 18 78 X2 19 79-96 20 97 I 21 98 Cz)

......................................................................................................................... X2 22 99-116 23 117 24 118

.................................................................................................................................. X3 25 119-236 C) 26 237 C) 27 238

(.n 28 239 (D'

29 240 30 241 31 242 X5 32 243-484 X6 33 485-968

BFNEPUCOP ProbabilisticRisk Assessment Model Name: UICOP2-9 Top Events for Event Tree: LLCSN 5:06 PH 2/9/2006 Page 2 Top Event Name 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 TURBINE :TRIP IVC CLOSURE OF'MSIVS 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 COOLING HARDWARE - LOOP I SpII SUPPRESSION POOL COOLING HARDWARE - LOOP II SPC LOGIC SWITCH FOR SUPPRESSION POOL COOLING WITH Ul RHR ODWS OPERATOR ALIGNS DRYWELL SPRAY DWS DRYWELL SPRAY HARDWARE E-8 C1320503-6924 R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLCSN 5:06 PM 2/9/2006 Page I SF Split Fraction Assignment Rule CILl PCA-S*(DWP-S + LVP-S)

CIL2 PCA-F*(DWP-S + LVP-S)

CILF DWP-F*LVP-F NPSHS RHRl*RHi2*RHR3 + RRR1*RHR2'*RHR4 + RHRl*RHR3*RHR4 + RHP2*RHR3*RHR4 +

RHRI*RHR2*RHR3*RHR4 Comuents IF 3 OR MORE PUMPS ARE AVAILABLE WE DON'T NEED COP FOR ECCS NPSH NPSHI INIT-LLCA + INIT-LLCB + INIT-LLDA + INIT-LLDB + INIT-LLO + INIT-LLSA +

INIT-LLSB NPSHS 1 RPSMS 1 RPSEO 1 TORI 1 TTP1 BB5-S*DI-S TTP2 EB5=S*DI-F

.TTP3 BB5-F*DI-S TTPF 1

.IVCl LPCIF -LPCISUP + NPSHLOST LPCI2 LPCISUP Conmnents MANUAL LPCI START NOT.CREDITED LLOCAS; ODD SPLIT FRACTION SWOULD APPLY LPCIIF -LPCIISUP + NPSHLOST LPCII2 LPCI-S LPCII4 -LPCISUP LPCII6 LPCI-F*LPCISUP CSF INIT-LLCA*(RF-F+AC-F+DB-F+AD-F+DD-F+NPII-F + CASSIG +DW-F*LV-F+RB-F+ -EECW)

+ INIT-LLCB*(RE-F+AA-F+DA-F÷AB-F+DC-F+NPI-F+DW-F*LV-F+RC-F+ -EECW) +

NPSHLOST CS2 INIT-LLCB*-(RE-F÷AA-F÷DA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+RC-F+ -EECW)

CS2B INIT-LLCA*-(RF-F+AC-F4DB-F+AD-F+DD-F+NPII-F+ CASSIG+DW-F*LV-F+RB-F+ -EECW)

CSF 1 E-9 C1320503-6924 R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: LLCSN 5:06 PH 2/9/2006 Page 2 SF Split Fraction Assignment Rule SIS LPCI-S*RPA-S*RPC-S + LPCII-S*RPB-S*RPD-S + LPCI-S*LPCII-S*( (RPA-S+RPC-S) +

(RPB-S+RPD-S) + CS-S )

Comments ANY TWO RHR PUMPS OR CS FROMTHE UNBROKEN LOOP SIF 1 OSPC1 RPSM-S*RPSE-S OSPCF 1 SPIF OSPC-F + RE-F + NPSHLOST SPI2 *E-S*RC-S* (RPA-S*HXA-S + RPdCS*HXC-S)

SPIF 1 .. .

SPIIF OSPC-F + RF-F + NPSHLOST SPII4 (RPB-S*HXB-S + RPD-S*RXD=S)*SPI-S SPII5 (RPB-S*HXB-S + PPDS*HXD-S)*SPI-F*R'-S SP116 (RPB-S*HXB-S + RPD-S*HXD-S)*SPI-F*RE-F SPIIF 1 SPCF -(SPI-S)*-(SPII-S)

SPCS SPI-S*((RPA-S*HXA-S + RPC-S*XC-S) + SPII--S(PB-S*HpB-S+RPD=S*pXl-S)

SPCF 1 ODWS1 1 DWS . PXI-F*PX2-F + (RPA-F*RPC-F +RH-F+NOGB) * (RPB-F*RPD-F+RI-F + NOGD)

DWSI PXI-S*PX2-S* RPA=S+RPC-S)*-NOGB*(RPB-S+RPD-S)*-NOGD DWS2 (RPA-F*RPC-a +RH-F+NOGB+PX1-F) * (RPB-F*RPD-F+RI-F + NOGD+PX2-F)

DWSF 1 E-10 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name*- UlCOP2-9 Macro for Event Tree: LLCSN 5:06 PM 2/9/2006 Page 2 Macro Macro Rule / Coments ALTINJRHSW RPSM-B THIS MACRO IS NEEDED IN THE CETS ALTINJU2X RPSM-B THIS MACRO IS NEEDED IN THE CETS BUCKET RPSH-B CILFAIL . CIL-F CLASSiA RPSM-B CLASSIB RPSM=B CLASSIBE RPSm-B CLASSIBL RPSI4=B CLASSIC RPSN=B CLASSID RPSM=B CLASSIE RPSSN-B CLASS2 " RPSM-B CLASS2A RPSM-B CLASS2L SPC-F + OSPC=F CLASS2T RPSm-B CLASS2V RPSM-B CLASS3A RPSM=B CLASS3B RISM-B E-1 1 C1320503-6924R2 - 7110/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LLCSN 5!06 PM 2/9/2006 Page 2 Macro Macro Rule / Comments CLASS3C -(SI-S)+ -(TTP-S+IVC-S)

CLASS3D - (TOR-S)

CLASS4 RPSM-F CLASS5 (TTP-S*- (IVC-S)

DWSPRAY DWS-S THIS MACRO IS NEEDED IN THE CETS LMDEPHDWR RPSM-B THIS MACRO IS NEEDED IN THE CETS HIGH . RPSM-B BPI RPSM-B LOW . INIT-LLCA + INIT-LLCB LPCIISUP RF-S*( (NPII-S*DW-S) + LV-S LPCISUP RE-S*( (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(TTP-S÷IVC-S}*SI-S*SP-S NODC RPSM-B THIS MACRO IS NEEDED IN THE CETS NORY 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-12 C1320503-6924R2 - 711012006

BFNEPU COPProbabilisticRisk Assessment Model Name: tUlCOP2-9 Macro for Event Tree: -LLCSN 5:09 PM 2/9/2006

.Page 3 Macro Macro Rule / Comments ODEP'Ll RPSM-B THIS MACRO IS NEEDED IN THE CETS RKRS PCOOL SPC-S SORV RPSM=S LARGE IOCAS ARE ALWAYS DEPRESSURIZEED E-13 C1320503-6924R2 - 7/10/2006

MODEL NanewUIERIN Page No. I of 2 Event Tree: LLON.ET1 13:37:12 Februay 16,2006 IE CIL NPSH RPSM RPSE TOR TiP IVC LPCI LPCII CS SI OSPC SPI SPII SPC COWS DWS L ...........

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MODEL Name: UIERIN Page No.2 of 2 Event Tre LLON.EfE 13:37:12 Februmy 16, 2006 Xn Bf St 1 I 2 2 3 3 4 4 X1 5 5-8 Xi 6 9-12 Xl 7 13-16 8 17 9 18 10 19 II 20

)a 12 21-38 13 39 X2 14 40-57 15 58 16 59 X2 17 60-77 18 78 X2 19 79-96 m 20 97 21 98 CA X2 22 99-116 23 117 24 118 MO 25 119-236 26 237 27 238 28 239 29 240 C) 30 241 0 31 242 bb X4 32 243-484 ~1 X5 33 485-968

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BFN EPUCOP ProbabilisticRisk Assessment Model Name: UICOP2-9 Top Events for Event Tree: LLON S:07 PM 2/9/2006 Page 1 Top Event Sam. 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 BASIb.

TOR PRESSURE SUPPRESSION POOL TTP TURBINE TRIP IVC CLOSURE OF MSIVS 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 COOLING HARDWARE - LOOP I SPII SUPPRESSION POOL COOLING HARDWARE - LOOP II SPC LOGIC SWITCH FOR SUPPRESSION POOL COOLING WITH UI RHR ODWS" OPERATOR ALIGNS DRYWELL SPRAY DWS DRYWELL SPRAY HARDWARE E-16 C1320503-6924R2 - 7/1012006

BFNEPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: LLON 5:07 PM 2/9/2006.

Page 1 SF Split Fraction Assignment Rule CIL" PCA-S*(DWP-S + iVP-S)

CIL2 PCA-F*(DWP-S + LVP-S)

CILF DWP-F*LVP-F NPSHS RHRI*RHR2*RHR3 + RHR1*RHR2*RHR4 + RHRI*RHR3*RHR4 + RHR2*RHR3*RHR4 +

RHRI *RHR2* RHR3* RHR4 Comments IF 3 OR MORE PUMPS ARE AVAILABLE WE DON'T NEED COP FOR ECCS NPSH NPSHI INIT-LLCA + INIT-LLCB + INIT-LLDA + INIT-LLDB + INIT-LLO + INIT-LLSA +

INIT-LLSB NPSHS 1 RPSMS 1 RPSEO I TORI 1 TTPl BB5-S*DI-S TTP2 BB5-S*DI-F TTP3 BB5-F*DI-S TTPF 1 IVC1 1 LPCIF -LPCISUP + NPSHLOST LPCI2 LPCISUP Corments MANUAL LPCI START NOT CREDITED LLOCAS; ODD SPLIT FRACTION SWOULD APPLY' LPCIIF -LPCIISUP + NPSHLOST LPCII2 LPCI-S LPCII4 -LPCISUP LPCII6 LPCI-F*LPCISUP" CSF (RF-F+AC-F+DB-F+AD-F+DD-F+NPII-F+ CASSIG+DW=F*LV=F+RB=F+ -EECW) *

(RE-F+AA-F÷DA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+RC-F+ -EECW) + NPSHLOST CS2 -(RE-F+AA-F+DA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+RC-F+ '-EECW)

CS2B -(RF-F+AC-r+DB-F+AD-F+DD-F+NPII-F+ CASSIG+DW-F*LV-F+RB-F+ -EECW)

CSF 1 SIS LPCI-S*(RPA-S+RPC=S) + LPCII-S*(RPB=S+RPD-S) + CS-S E-17 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLON 5:07 PM 2/9/2006 Page 2 Sr Split rraction Assignrment Rule Comments ANY TWO RHR PUMPS OR CS FROMTHE UNBROKEN LOOP SIF OSPC1 RPSM-S*RPSE-S OSPCF SPIF RE-F + OSPC-F + WPSHLOST s P12 SPIIF OSPC-F + RF-F + NPSHLOST SPI14 (RPB-S*HXB-S + 'RP6;S*HXD-S)*SI-

SPI15 (RPB-S*HXB-S + RPO-S*HXfl-S)*SPI-F*RE-S SpIX6 * (RPB-S*HXB-S + RPD-S*HXD-S)*SPI-F*RE-F SPIIF SPCF - CSPI-S) *- (SPII-S)

SPCS SPI-S'ý(EPA-S*HXA-S + RPC=S*HXC-S) + SPII-S* CRPB-S*HXB-S+RPD-S*HXO=S)

SPCF ODWsl DWSF PX1-F*PX2-F + (P.PA-F*RPC-F +RII-F+NOGB) * (RPB-F*RPD-F+RI-F + NOGD)

OWS1 PX1-S*PX2-S* (RPA-S+RPC-S *-NOGB* (RPB-S5+RPD-S) *.NOGO DWS2 (RPA-F*RPC-F +PRi-F+NOGB+PX1-F) * (RPB-F*RPD-F+RI-sF 4 NOGD+PX2-F)

OWSF 1 E-18 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LLON 5:07 PH 2/9/2006 Page I Macro Macro Rule / Comnents ALTINJRHSW RPSM-B THIS MACRO IS NEEDED IN THE CETS ALTINJU2X RPSM-B THIS MACRO IS NEEDED IN THE CETS BUCKET RPSM-B CILFAIL CII.F CLASSIA RPSM-B CLASSIB RPSM-B CLASSiBE RPSM-B CLASS1BL RPSMýB CLASSIC RPSM-B CLASSID RPSM-B CLASSIE RPSM-B CLASS2 RPSM-B CLASS2A RPSM-B CLASS2L OSPC-F+ SPC-F CLASS2T RPSM-B CLASSZV RPSM-B CLASSiA RPSM-B CLASS3B RPSM-B E-19 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Macro for Event Tree: LLON B:07 Po 2/9j2006 Page 2 Macro Macro Rule / Comments CLASS3C -(SI-S )+ -(TTP-S+IVC-S)

CLASS3D -(TOR-SS)

CLASS4 RPSM=F CLASS5 -(TTP=S) *(IVC=S)

DWSPRAY DWs-S THIS MACRO IS NEEDED IN THE CETS EMDEPHDWR RPSM-B THIS MACRO IS NEEDED IN'THE.CETS HIGH .RPSM-B HPI RPSM-B.

LOW INIT-LLO LPCIISUP RF-S*( (NPII-S*DW=S) + LV=S LPCISUP " RE-S*( (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*(TTP-S+IVC-SSI-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 T.HIS MACRO IS NEEDED IN THE CETS NPSHLOST CIL-F*NPSH-F E-20 C1320503-6924R2 - 7/10/2006

BFNEPUCOPProbabilisticRisk Assessment Model Name: U1COP2-9 Macro for Event Tree: LLON 5:07 PM 2/9/200'6 Page 3 Macro Macro Rule / Cormments OPDEPLI RPSM=B THIS MACRO IS NEEDED IN'THE CETS RXRSPCOOL SPC-S SORV RPSM-S LARGE LOCAS ARE ALWAYS DEPRESSURIZEED E-21 C1320503-6924R2 - 7/10/2006

MODEL Name: UIERIN Page No. 1 of 2 Event Tree: LLRDN.ETI 13:37:46 February 16, 2006 IE CIL NPSH RPSM RPSE TOR TTP PC DV1 DV2 LPCV LPCII CS SI

- 7

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MODEL Name: Ul ERIN Page No. 2 of 2 Event Tree: LLRDN.ETI 13:37:46 February 16, 2006 OSPC SPI SPII SPC ODWS DWS I X# B#

xiq 1 1 2 2

........................ 3 3 4 4 Xi 5 5-8 Xi 6 9-12 X1 7 13-16 8

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9 18 X2 10 19-36

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,m X2 11 37-54

... °°.°°°.°...°°°.oo°°.°o°°°°.°°°ooo.o.ooooooooo°°.o..o.o°°.........................°.°°°.....°°°°°o.o°°°°.o°°°°° X,2 12 55-72 X2 13 73-90

..*....°°°°°°°°°°.........°...° °.°

... ..°° ..°.° .. °..°.° .......... °.....°..........°°°° ° °.°

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X2 14 91-108 X2 15 109-126 X2 16 127-144 QZ X4 17 145-288 C)

°°°°°o.oo.°°.....°...°°°°.°°°°.o°°°°°°°°°.o°°°°.°.°o.,°.°.°...°°°oo.o........ °.o.°°.°.oo.oo.o.o°°.o°°.°o°...°°° Q)

X4 18 289432

....... °..°..°......°.. °°..°.... °.........°..°..°°.. °.°.. °.. ; °.

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X3 19 433-468

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X5 20 469-936 21 937 C.)

...° ....

° °..........°....°°......°....°......°°

....... ° ° .° -. °°..°°.. °.. °. °°..°..°° 22 938 23- 939 24 940 X6 25 941-1880

.. °...°.. °°°.....° ..°° ....... °°.........°...°..°... °.. °...°..... °... °...° °.....

X7 26 1881-3760 io3 C)

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BFN EPUCOPProbabilisticRisk Assessment Model Name: U1COP2-9 Top Events for Event Tree: LLRDN 5:09 PH 2/9/2006 Page 2 Top zvent Name 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 INUREG-5500 BASIS)

TOR PRESSURE SUPPRESSION POOL TTP TURBINE TRIP IVC CLOSURE OF MSIVS DVI LOOP I RECIRCULATION DISCHARGE 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 SUPPRESSION POOL COOLING SPI SUPPRESSION POOL COOLING HARDWARE - LOOP I SPII SUPPRESSION POOL COOLING HARDWARE - LOOP II SPC' -LOGIC SWITCH FOR SUPPRESSION POOL COOLING WITH Ul RHR ODW1S OPERATOR ALIGNS DRYWELL SPRAY' DWS DRYWELL SPRAYk HARYDWARE E-24 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: t-lCOP2-9 Split Fraction Assignment Rule for Event Tree: LLRDN 5:09 PM 2/9/2006, Page I SF Split Fraction Assignment .Rule CIL" PCA-S*(DWP-S + LVP-S)

CIL2 PCA-F* (DWP-S + LVP-S)-

CILF DWP-F*LVP-F.

NPSHS RHRl*RHR2*RHR3 + RHRlRHR2iRHR4 + RHRi*RHR3*RHR4 + RHR2*RH3*RHR4 +

RHR1*RHR2*RHR3 *RmR4 Comments IF 3 OR MORE PUMPS ARE AVAILABLE WE DON'T NEED COP FOR ECCS NPSH NPSHl INIT-LLCA + INIT-LLCB + INIT-LLDA + INIT-LLDB + INIT-LLO + INIT-LLSA +

INIT-LLSB NPSHS 1 RPSMS 1 RPSEO 1 TORI .'-

TTP1 BB5-S*DI-S TTP2 BB5-S*DI-F TTP3 BB5-F*DI-S TTPF 1 IVCi 1 DVIF RE-F+RB-F*RC-F+NHI-F*NH2-F÷DW-F*LV-F DV1i DW-S*LV-S*NHI-S*NH2-S*RB-S*RC-S DV12 DW-S*LV-S*NHI-S*NH2-S* (RB-F+RC-F)

DV13 DW-S*LV-WF*NHI-S*NH2-S*RB-S*RC-S DV14 Dw-r*LV-S-NH1-S*NH2-S *RB-;SRC-S DVIS DW-S*LV.S- (NHl-F+NH2-F) *RB-S*RC-S DVlF 1 DV2F RFPF+RB-F*RC-F+NH1-F*NH2-F+DW-¥FLV-F DV25 RE-F*DV1-F* DW-S*LV-S*NHIS*NH2-S*RB-S*RC-S DV21 DVlmS*DW-S*LV-S*NH1-S*NH2-S*RB-S*RC-S DV22 DVI*F*DW-S*LV-S*NH1-S*NH2-S*RB-S*RCmS DV24 RE-F*DV1-F*DW S*LV-S*NHI-S*NH2-S*(RB-F+RC-F)

E-25 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLZRDN 5:09 PM 2/9/2006 Page 2 SF Split Fraction Assignment Rule DV23 DV1-S*DW-S*LV-S*NH1l-S*NH2-S* (RB-F+RC-F).

DV24 DVIF*DW-S*LV-S*NHIS*NH2-S* (RB-F+RC-F)

DV27 RE=F*DVI-F*DW-S*LV-F*NH1-S*NH2-S*RB-S*RC-S DV28 DVI-S*DW-S*LV-F*NH1-S*NH2-S*RB-S*RC-S DV29 DV1-F*DW-S*LV-F*NHI-S*NH2-S*RB-S*RC-S DV2A RE-F*DVI-F*DW-F*LV-S*NHIS*NH2..S*RB-S*RC=S DV2B .DVI-S*DW-F*LV-S*NH1-S*NH2-S*RB-S*RC-S DV2C DV1-F*DW-F*tV-S*NHI-S*NH2-S*RBIS*RC-S DV2D RE-F*DV1-F*DW-S*LV=S*(NH1-F+NH2-F)*RB=S*RC-S DV2E "DV-S*DW-S*LV-S*(NH1-F+NH2-F}*RB-S*RC-S DV2G DV1-F*DW-S*LV=S*(NH1-F+NH2-F)*RB-S*RC-S DV2F I LPCIF -LPCISUP+ DVI=F*DV2=F + NPSHLOST LPCI2 LPCISUP LPCIZF -LPCIISUP +DV1-F*DV2-F + NPSHLOST LPCII2 LPCI-S LPCII4 -LPCISUP LPCII6 LPCI-F*LPCISUP LPCIZF 1 CSF (RE-F+AA-F+DA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+RC=F+EA=F*EB=F*EC*F +

EA-F*EB-F*ED=F 4 EA=F*EC-F*ED-F +

EB-F*EC-F*ED-F) * (RF-F+AC-F+DB=F+AD-F+DD-F+NPII-F+ CASSIG+DW-F*LV-F+RB-F+

EA-F*EB-F*EC-F + EA-F*EB-F*ED-F +.EA-F*EC=F*ED=F + EB=F*EC=F*ED-F)" +

NPSHLOST CSI -(RE-F+AA-F+DA-F+AB-F+DC-F+NPI F+DW-F*LV=F+RC=F+EA=F*EB=F*EC=F +

EA-F*EB-F*ED-F + EA=F*EC=F*ED-F + .-

EB-F*EC-F*ED=F)*-(RF-F+AC-F+DB-F+AD-F+DD-F+NPII-F+

'CASSIG+DW=F*LV-F+RB-F+EA-F* EB-F*EC=F + EA=F*EB-F*ED-F +. EA-F*EC-F*ED-F +

EB-F*EC-F*ED-F)

CS2 - (--F+A-F+DA-F+A-F+DC-F+NPI-F+DW-F*LV-F+RC-F+A-F*EB-F*EC-F +

EA-F*EB-F*ED-F + EA-F*EC-F*ED-F +

EB-F*EC-F*ED-F)*(RF-F+AC-F+DB-F+AD-F+DD-F+NPII-F+

CASSIG+DW=F*LV-F+RB-F+EA-F*EB-F*EC-F + EA-F*EB-F*ED-F + EA-F*EC-F*ED-F +

EB-F*EC-F*ED-F)

E-26 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLRDN 5:09 P* 2/9/2006 Page 3 SF Split Fraction Assignment Rule CS2B (RE=F+AA=F+DA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+RC-F÷EA-F*EB-F*EC-F +

EA-F*EB-F*ED-F + EA-F*EC-F*ED-F +

EB=F*EC-F*ED=F) *- (RF-F+AC-F+DB-F+AD-F+DD-F+NPII-F+

CASSIG+DW-F*LV-F+RB-F+EA-F*EB-F*EC-F + EA-F*EB-F*ED-F + EA-F*EC-F*ED-F +

EB-F*EC-F*ED-F)

.CSF 1 Comments Core Spray Loop .II Pipe Break Large LOCA SIS CS-S + LPCI-S*(RPA-S + RPB-S) + LPCII-S*(RPB-S + RPD-S)

SIF 1 OSPC1 RPSM-S*RPSE-S OSPCF 1 SPIF RE-F + OSPC-F + NPSHLOST SPI2 1 SPIIF OSPC-F + RF-F + NPSHLOST SP114 (RPB-S*HXB-S + RPD-S*HXD-S)*SPI-S SP115 (RPB-S*HXM-S + RPD-S*HXD-S)*SPI-F*RE-S SPI16 (RPB-S*(XBS" + RPD-S*HXD-S)*SPI-F*RE-F SPIIF 1 SPCF .- (SPI=S)* (SFII-S).

SPCS SPI=S* (RPA-S*HXA-S + RPC-S*HXCýS) + SPII-S* (RPB-S*HXB-S+RPD-S*HXD=S)

SPCF 1 ODWS1 1 DWSF PXI©F*PX2-F + (RPA-F*RPC-F +RH-F+NOGB) * (RPB-F*RPD-F+RI-F + NOGD)

DWS1 PXI-S*PX2-S* (RPA-S+RPC-S) *-NOGB* (RPB-S+RPD-S) *-NOGD -

DWS2 (RPA=F*RPC=F +RH-F+NOGB+PX1-F) * (RPB-F*RPD-F+RI=F + NOGD+PX2-F)

DwSF 1 E-27 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UlCOP2-9' Macro for Event Tree: LLRDN 5:09 PH 2/9/2006 Page 1 Macro Macro Rule / Corments ALTINJRHSW RPSM=B THIS MACRO IS NEEDED IN THE CETS ALTINJU2X RPSM-B THIS MACRO IS NEEDED I N THE CETS BUCKET RPSM-B CILFAIL CIL-F CLASSlA. RPSM=fB CLASSlB RPSM-B CLASSlBE RPSM-B CLASSIBL RPSM-B CLASSIC RPSM-B CLASSiD RPSM-B CLASSlE RPSM=B CLASS2 RPSM-B CLASS2A. RPSM-B CLASS2L OSPC-F + SPC-F CIASS2T RPSM-B CLASS2V RPSM-B CLASS3A. RPSM-B CLASS3B RPSM-B E-28 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Macro .for Event Tree: LLRDN 5:09 PH 2/9/2006 Page 2 Macro Macro Rule / Co-ments CLASS3C -(SI=S )+ -(TTP=S+IVC=S) aLASS3D -ITOR-S)

CLASS4 RPSM-F CLASS5 -(TTPS)*- (IVC-S)

DWSPRAY DWS-S THIS MACRO IS NEEDED IN THE CETS EMDEPHDWR RPSM-B THIS MACRO IS NEEDED IN THE CETS HIGH RPSM=B HPI RPSM=B LOW - INIT-LLDA + INIT-LLDB LPCIISUP RF-S*( INPII-S*DW-S) + LV-S )

LPCISUP RE-S*(. (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* (TTP-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-29 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LLRDN 5:09 PH 2/9/2006 Page 3 Macro Macro Rule / Comnents OPDEPLi RPSM=B THIS MACRO IS NEEDED IN THE CETS RHR§PCOOL OSPC-F + SPC=F SORV' RPSM-S LARGE LOCAS ARE ALWAYS DEPRESSURIZEED E-30 C1320503-6924R2 - 7/10/2006

MODEL Name: U1ERIN Page No. 1 of 4 Event Tree: LLRSN.ETI 13:38:20 February 16. 200DE JE CIL NPSH RPSM RPSE TOR TUP IVC DVI DV2 LPCI LPCII CS SI OSPC SPI m

0 C.

001 0

IC,

-S

MODEL Name: U1ERIN Page No. 2 of 4 Event Tree: LLRSN.ETI 13:38:20 February 16, 200C SPII SPC ODWS DWS I X#

AW i 1 2

3

.................................... X9 X9 4

5-8 9-12

...................................... X9 13-16 17 18

............................................ X10 19-36

.. ... .. .... . ...................... X10 37.54

°°°°°.°°°°°.......................°.o°°°°.~..... X10 55-72

.°,ool~oo°°°°°°°..

73-90

°°o Xl0

° .

°. .....................

°. x li 91-107 108-124

°.° . ..................

° ° Xil1 125.141 0

°°oo°o° °.°.o

............. .. o.* *.°...... °................°*

...................°.. ° .*D

.. .XXi 3

° *..................................°°,.°°°°°°°.*°.............° 142-282 xi 283-423

°-°°°°=°*

xi

°° *.°°-.*°.........................................

............ Ux 424-458 0 459-916 C-,

C., 917 918 C-,

0 C-,

0)

Il C-,

C-1 (b

MODEL Name: U1ERIN Page No. 3 of 4 Event Tree: LLRSN.ETI 13:38:20 February 16, 200e I IE CIL NPSH RPSM RPSE TOR TP IVW DVI DV2 LPCl LPCII CS SI OSPC SPI I

I I

i................ .....................................................................................................

Q) 0 (P)

-D 0

C) 0)

MODEL Name: UlERIN Page No. 4 of 4 Event Tree: LLRSN.ETI 13:38:20 February 16, 200C SPII SPC ODWS DWS X#

23 919 24 920

................................ *...... *........................ X12 25 921-1840

..-.............-.....-...-.........-...-.-.-.-.-.-..-.......-..-. X13 26 1841-3680 C?,

C.,

.0 C.)

(0 Cal N3 C) a)

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Top Events for Event Tree: LLRSN 5:09 PH i/9/2006 Page I Top Event Namm Dsxcription 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 TURBINE TRIP IVC CLOSURE OF MSIVS Dvi LOOP I RECIRCULATION DISCHARGE VALVE CLOSURE.

DV2 LOOP II RECIRCULATION DISCHARGE VALVE CLOSURE LPCI LPCI LOOP I LPCII LPC LOOP I'I CS CORE SPRAY SYSTEM SI LOGIC SWITCH FOR SUFFICIENT INJECTION OSPe OPERATOR ALIGNS SUPPRESSION POOL COOLING SPI SUPPRESSION POOL COOLING.HARDWARE - LOOP I SPIX SUPPRESSION POOL COOLING HARDWARE - LOOP.II SPC LOGIC SWITCH FOR SUPPRESSION POOL'COOLING WITH Ul RHR ODWS OPERATOR ALIGNS DRYWELL SPRAY

.DKS DRYWELL SPRAY HARDWARE E-35 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLRSN 5:09 PM 2/9/2006 Page 1 SF Split Fraction Assignment Rule CILl PCA-S* (DWP-S + LVP-S)

CIL2 PCA-F*(DWP-S + LVP-S)

CILF DWP-F*LVP-F NPSHS RHR1*RHR2*RHR3 + RHR1*RHR2*RHR4 + RHR1*RHR3*RHR4 + RHR2*RHR3*RHR4 +

RHRI*RHR2*RHR3*RpR4 Cozmments IF 3 OR MORE PUMPS ARE AVAILABLE WE DON'T NEED COP FOR ECCS NPSH NPSHI INIT-LLCA + INIT-LLCB + INIT-LLDA + INIT-LLDB + INIT-LLO + INIT-LLSA +

INIT-LLSB NPSHS 1 "RPSMS 1 RPSEO. 1 TORI 1 TTP1 BB5-S*DI-S TTP2 BB5-S*DI-F TTP3 BB5-F*DI-S TTPF 1 IVC1 1 DVIF RE-F÷RB-F*RC-F+NHIF*NH2-F+DW-F*LV-F" DviI DW-SiLV-S*NH1-S*NH2-S*RB-S*RC-S DV12 DW-S*LV-S*NH1"S*NH2"S*(RB-F+RC-F)

DV13 DW-S*LV-F*NHI-S*NH2-S*RB-S*RC-S DV14 DW-F*LV-S*NHI'S*NH2"S*RB-S*RC-S DV15 DW-S*LV-S*(NH1-F+NH2-F)*RB-S*RC-S DV1F 1 "

DV2F rF+RB-F*RC-F+NH1-F*NH2-F+Dw-r*LV-F DV25' RE--*DVI-F*DW-S*LV-SiNHI-S*NH2-S*RB-S*RC-S DV21 DV1-S*DW-S*LV-S*NHI-S*NH2-S*RB-S*R-S" DV22 DV1-F*DW-SýLV-S*NHI'S*NH2-S*RB-S*RC-S DV24 RE-F*DVI"F*DW-S*LV.S*NHI=S*NH2-S*(RB-F+RC-F)

E-36 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9.*

Split Fraction Assignment Rule for Event Tree: LLRSN

'5:09 PM.2/9/2006

Page 2 SF Split Fraction Aasignmant Rule DV23 " VI-S*DW-S*LV-S*NHI-S*NH2-S*(RS-F+Rc-).

DV24 DV1-F*DW-S*LV-S*NH1-S*NH2-S*(RB-F+RC-F)

DV27 RE-F*DVI-F*DW-5*LV-F*NHI-S*NH2-S*RB-S*RC-S DV28 DVI-S*DW-S*LV-F*NHI-S*NH2"S*RB-S*RC-S DV29 DVI-F*DW-S*LV-F*NHI-s*NH2-S*RB-S*RC-S DV2A RE-F*DV1-F*DW-F*LV-S*NH1-S*NH2-S*RB-S*RC-S DV2B DV1-S*DW-F*LV-S*NH1-S*NH2-S*RB-S*RC-S DV2C DVl F*DW-F*LV=S*NH1*S*NH2"S*RB-S*RC-S DV2D RE-F*DVI F*DW-S*LV-S*(NH1-'F+NH2-F)*RBS*RC-S DV2E DV1-S*DW-S*LV-S*(NHI-F+NH2-F)*RB-S*RC-S DV2G DVIF*DW-S*LV-S*(NHI-F+NH2-F)}*B-S*PC-S DV2F1 LPCIF RE-F + DVI-F + NPSHLOST.

LPCI2 1 LPCIIF RF-F + DV2-F + NPSHLOST LPCII2 LPCI-S" LPCII4 RE-F LPCII6 LPCI-F*RE-S LPIIF 1I CSF (RE-F(AA-F+DA-F+AB-F+DC-F4NPI-F+DW-F*LV-F+Rc-F+-EECW) (RF-F+AC-F+DB-F+AD-F+D D-F+NPII-F+ CASSIG+DW-F*LV-F+RB-F+ -EECW) + NPSHLOST CS1 -(RE-F+AA-F+DA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+RC-F+

-EECW)*-(RF-F+AC-F+DB-F+AD-F+DD-F+NPII-F+ CASSIG+DW-F*LV-F+RB-F+"-EECW)

CS2 -(RE-F+AA-F+DA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+RC-F+.

-EECW)* (RF-F+AC-F+DB-F+AD-F+DD-F+NPII-F+ CASSIG+DW-F*LV-F+RB-F+ -EECW)

CS2B (RE--+AA-FDA-F+AB-F+DC-F+NPI-F+DW-F*LV-F+ C-F+-EECW)-(RF-F+AC-F+DB-F+AD-F+

DD-F+NPII-F+ CASSIG+DW-F*LV-F+RB-r+-EECW)

CSF 1 Corments Core Spray Loop II Pipe Break Large LOCA SIS 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-37 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: LLRSN 5:09 PM. 2/9/2006 Page*3 SF Split Fraction Assignment Rule SIF I OSPCl RPSM-S*RPSE-P OSPCF 1 SPIF RE=F + OSPC-F + NPSHLOST SPI2 1 SPIIF OSPC-F + RF-F + NPSHLOST SPI14 (RPB-S*HXB-S + RPD-S*HXD-S)*SPI-S SPI15 (RPB-S*HXB-S + RPD-S*HXD-S)*SPI-F*RE-S SPII6 (RPB-S*HXB-S + RPD-S*JXD-S).*SPI-F*RE-F SPIIF 1 SPCF -(SPI-S),*-(SPII-S)

SPCS SPI-S* (RPA-S*HXA-S + RPC-S*HXC-S) + SPII-S* (RPB-S.*HXB-S+RPD=S*HXD-S)

SPCF 1 ODWSI 1 DWSF" PX1=F*PX2=F + (RPA-F*RPC-F +RE-F+NOGB) * (RPB-F*RPDýF+RI-F + NOGD)

DWS1 PX-S-*Px-S* (RPA-S+RPC-S) *-NOGB* (RPB-S+RPD-S) *-NOGD DWS2 (RPAiF*RPC-F +RH-F+NOGB+PX1-F) * (RPB-F*RPD-F+RI-F + VOGD+PX2-F)

DwsF 1 E-38 C1 320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LLRSN 5:09 PM 2/9/2006 PageI Mar-ro XMacro Rule -/ Coznents ALTINJRHSW9 RPSM-B THIS MACRO IS NEEDED IN THE CETS ALTINJU2X RPSN-B THIS MACRO IS NEEDED IN THE CETS BUCKET Rpsmz-B CILFAIL CIL-F CLASSlA RPSM.B CLASSIB. RPSM-B CLASSIBE RPSM-B CLASS1BL RPSM-B" CLASSiC RPSM-B CLASSlD RPSM-B CLASSlE RPSM=B CLASS2 RPSM-B CLASS2A RPSM=B CLASS2L OSPC-F + SPC-F CLASS2T RPSM-B CLASS2V RPSM-B CLASS3A RPSM=B CIAsSS3B RPSM-B E-39 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Macro for Event Tree: LLRSN 5:09 PM 2/9/2006 Page 2 Macro Macro Rule / Comments CLASS3C - (SI-S )+ -(TTP-S+IVC-S)

CLASS3D -(TOR-S)

CLASS4 RPSM-F CLASS5 -(TTP-S)*-(IVC-S)

DKSPRAY DWS=S THIS MACRO IS NEEDED IN THE CETS.

EECW EA-S*(EB-S + EC-S + ED-S] + EB-S8*(EC-S + ED-S) + EC-S*ED=.S EMDEPHDWR RPSM=D THIS.MACRO IS NEEDED IN THE CETS HIGH RPSM-B HPI RPSM=B LOW INIT=LLSA + INIT-LLSB LPCIISUP RF-S*( (NPII-S*DW-S) + LV-S LPCISUP RE-S*( (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* CTTPS+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 RPSI-B THIS MACRO IS NEEDED IN THE CETS E-40 C1320503-6924 R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LLRSN 5:09 PH i/9/2006 Page 3 Macro Macro Rule / Coments NPSHLOST CIL-F*NPSH-F OPDEPLI RPSM-B THIS MACRO IS NEEDED IN THE CETS RHRSPCOOL SPC-S SORV RPSM-S LARGE LOCAS ARE ALWAYS DEPRtSSURIZEED E-41 C1320503-6924R2 - 7/10/2006

MODEL Name: UlERIN Page No. I of 4 Event Tree- CETN1.ETl 13:36:50 February 16, 200E IR CZ TM FD DWl WR RME X 1B#

I IE ' L2 AL CILDUM 01 1

2 4

6 7

'8 10 12 L_-- 13 14 m

15 16 17 18 19 20 21 22 23 0 24 25 26 27 28 CA) 29 C-,

Nl) 30

.31 C) 33 34 C-,

il0 36 C-,

C) 0D

MODEL Name: UIERIN Page No. 2 of 4 Event Tree: CETNI.ETI 13:36:50 February 16, 200E Cz C) 0 0

0

MODEL Name: U1ERIN Page No. 3 of 4 Event Tree: CETNI.ETI 13:36:50 February 16, 200f L2 AL CILDUM 01 IR CZ TD FD DW WR RME X#

IE I~

I 37 38 39 40 41 to Z-.

C) 0)

MODEL Name: UIERIN Page No. 4 of 4 Event Tree: CETN1.ETI 13:36:50 February 16, 200E SN 37 38 39 40 41 m

-4 Q)

St, 0

01 0,

Z!

a 0

0)

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Top Events for Event Tree: CETNI 5:09 PM 2/9/2006 Page I Top Event Name Description L2 LEVEL 2 /LERF RESULTS AL CETI LOGIC NODE FOR CLASS 2 AND CLASSIBL CILDUM CIL DUMMY TOP OI OPERATORS DEPRESSURIZE RPV (L2)

IR IN-VESSEL RECOVERY CZ CONTAINMENT ISOLATED AND INTACT TD INJECTION ESTABLISHED FD CONTAINMENT FLOODING DWI NO DIRECT. DRYWELL RELEASE PATH WR WET AIR SPCE FAILURE RME CONTAINMENT 'BUILDING EFFECTIVE.

E-46 C1320503-6924R2 - 7/1012006

BFNEPU COPProbabilisticRisk Assessment Model Name:' UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI 5:09 PH 2/9/2006 Page I SF Split Fraction .Assignimant Rule L20 1 Conments L20-0 IMPLIES LEVEL 1; L20-1 IMPLIES LEVEL2; USE MFF TO CHANGE ALF CLASSlA + CLASSIBE + CLASSIC + CLASSiD + CLASSlE + CLASS3A + CLASS3B +

CLASS3C ALO NOCD + CLASSIBL + CLA9S2A + CLASS2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4

+ CLASS5) + BUCKET Cornuents CLASS 3D AND CLASS 4 ARE EVALUATED FOR LERF CILDF CILFAIL CILDS 1 OIS CLASS3A + CLASS3B.+ CLASS3C + LOW Oil CLASS2A + CLASS2T + NORV* (CLASSlA + CLASSIBE + CLASSIBL+ CLASSIC) +

CLASSlB*(NOACREC + NODC) 014 CLASSIB 013 -DPDEPL1* (CLASSlA + CLASSIC + CLASSID)

Comments change l hIGH PRESSURE LERF 012 OPDEPLI* (CLASSlA + CLASSIC + CLASSID)

Comments change l hIGH PRESSURE LERF IR1 OI=F*(CLASS1A + CLaSsiC)

IR3 CLASSIBE IR4 CLASSlBL IR5 '0I-F*CLASSID IR6 01-S*CLASSID Comments the irginal Ul L2 model 1R7 OI-F*CLASSIE IR8 OI-S*CLASSlE IR2 O-S Comments LOW PRESSURE INJECTION IMPLICIT IRF 1 CZ2 IR-F*OI-S CZ4 IR-F*OI-F CZ1 IR-S*OI-S CZ3 IR-S*OI-F E-.47 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI 5:09 PM:2/9/2006 Page 2 SF Split Fraction Assignment Rule CZF TDI CLASSlE TD2 OI-S*DWSPRAY TD3 -(0-IB) *CLASSlBE TD4 -(OI-B)*CLASSlBL TD8 0I-F*CLASSlA TDF 1 FDI ALTINJRHSW + DWSPRAY FD2 TD-S* (CLASSlA + CLASSlBE + CLASSlBL + CLASSID + CLASS3A + CLASS3B + .CLASS3C)

FD3 TD-F* (CLASSIA +"CLASSIC + CLASSID + CLASS3A + CLASS3B + CLASS3C)

FD4 TD=F* (CLASSIBE + CLASS1BL)

DWIF 1 WRI. DW-S PME8 CLASSIBL Comments TD-S*DWSPRAY*RHRSPCOOL This was an assumption that resulted in 100 RBEE RME7 OI-F RME6 0IS*TD-S*FD-S*DWS-S RME5 OI-S*TD-S*FD-S*DWS-F RP1E4 0I-S*TD-S*FD-F TME3 OI=S*TD-F*FD=F RMEF 1 L20 1 Comments L20-0 IMPLIES LEVEL 1; L20-1 IMPLIES LEVEL21 USE MFF TO CHANGE ALF CLASSlA + CLASSIBE + CLASSIC + CLASSiD + CLASSlE + CLASS3A + CLASS3B +

CLASS3C AL0 NOCD + CLASSlBL + CLASS2A + CLASS2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4

+ CLASS5) + BUCKET Cosents CLASS 3D AND CLASS 4 ARE EVALUATED FOR LERF CILDF CILFAIL CILDS.

E-48 C1320503-6924 R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI 5:09 PM 2/9/2006 Page 3 SF Split Fraction Assignment Rule OIS CLASS3A + CLASS3B + CLASS3C + LOW Oi1 CLASS2A + CLASS2T + NORV* (CLASSlA + CLASSIBE + CLASSIBL+ CLASSIC) +

CLASSlB* (NOACREC + NODC1 014 CLASSIB 013 -OPDEPL1*(CLASSIA + CLASSIC + CLASSID)

Comments change l hIGH PRESSURE LERF 02 OPDEPLI*(CLASSlA + CLASSIC + CLASSID)

Comments change l hIGH PRESSURE LERF IRi OI-F*(CLASSlA + CLASSIC)

IR3 CLASSlBE IR4 CLASSIBL IR5 OI-F*CLASSID IR6 OI1S*CLASS1D Comments the irginal U1 L2 model IR7 OI-F*CLASS1E IR8 OI-S*CLASSiE IR2 0-S Comrments LOW PRESSURE INJECTION IMPLICIT IRF 1

CZ2 IR-F*OI-S CZ4 IR-F*0I-F CZI IR-S*0I-S CZ3 IR-S*0I=F CzF *1 TD1 CLASSIE, TD2 OIS*DWSPRAY TD3 -(OI-B)*CLASSlBE TD4 -(OI-B)*CLASSiBL TD8 OI-F*CLASSlA TDF I E-49 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction.Assignmnent Rule for Event Tree: CETN1 5:09 PM 2/9/2006 Page 4 SF Split Fraction Assignment Rule FD1 ALTINJRHSW t DWSPRAY

'D2 TD-S* (CL.ASSlA + CLASSIBE + CLASSlBL 4 CLASSID + CLASS3A + CLASS3B + *CLASS3C)

FD2 TD-F* (CLASSIA + CLASSIC + CLASSiD + CLASS3A + CLASS3B + CLASS3C)

FD4 TD-F*(CLASSlBE + CLASSIBL)

DWIF 1 WRi DW-S RMES CLASSiEL Comments TD-S*DWSPRAY*RHRSPCOOL This was an assumption that resulted in 100 RBE RME7 OI-F RME6 OI-S*TD-S*FD-S*DWS-S RME5 OI-S*TD-S*FD-S*DWS-F RME4 OI-S*TD-S*FD-F RME3 OI-S*TD-1,*FD-F RME?*

L20 I Comments L20-D IMPLIES LEVEL 1; L20-1 IMPLIES LEVEL2; USE MFF TO CHANGE ALF CLASSlA + CLASSIBE + CLASSIC + CLASSID + CLASSlE + CLASS3A +. CLASS3B +

CLASS3C ALO NOCD + CLASSIBL + CLASS2A + CLASS'2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4

+ CLASS5) + BUCKET Comments CLASS 3D AND CLASS 4 ARE EVALUATED FOR LERF CILDF CILFAIL CILDS "1 OIS CLASS3A + CLASS3B + CLASS3C + LOW oIl CLASS2A + CLASS2T + NORV*(CLASSIA + CLASSiBE + CLASSIBL+ CLASSIC) +

CLASSIB*(NOACREC + NODC) 014 "CLASSIB 013 -OPDEPLl* (CLASSlIA + CLASSIC + CLASSID)

Comnments changel hIGH PRESSURE LERF 012 OPDEPLI*(CLASSIA + CLASSIC + CLASSID)

Comments changel hIGH PRESSURE LERF E-50 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI s:09 PM 2/9/2006 Page 5 SF Split Fraction Assignment Rule IRI OI-F* (CLASSlA + CLASSIC)

IR3 CLASSIBE IR4 CLASSlBL IR5 OI-F*CLASSlD IR6 OI-S*CLASSID Comments the irginal U1 L2 model IR7 0I=F*CLASSIE IR8 0I-S*CLASSIE 1R2 01-S Comments LOW PRESSURE INJECTION IMPLICIT IRF 1 CZ2 IR-F*OIS CZ4 IR-F*0I-F CZl IR-S*OI-S CZ3 IR-S*OI-F CZF I TDI CLASSlE TD2 0I-S*DWSPRAY TD3 -(OI-B)*CLASSIBE TD4 -(OI-B) *CLASSlBL TDO 0I-F*CLASSlA TDF 1 FDl ALTINJRHSW + DWSPRAY FD2 TD-S* (CLASSlA + CLASSIBE + CLASSIBL + CLASSID + CLASS3A + CLASS3B + CLASS3C)

FD3 TD-F* (CLASSlA + CLASSIC + CLASSiD + CLASS3A + CLASS3B + CLASS3C)

FD4 TD-F* (CLASSIBE + CLASSIBL)

DWIF I WRI DW-S RMEB CLASSlBL E-51 C1320503-6924 R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI 5:09 PM 2/9/2006 Page 6 SF Split Fraction Assignment Rule Comments TD-S*DWSPRAY*RHRSPCOOL This was an assumption that resulted in 100 RBE RME7 OI-F RME6 OI-S*TD-S*FD-S*DWS-S RME5 OI-S*TD-S*FI-S*DWS-F RMEd OI-S*TD-S*FD-F RME3 OI-S*TD-F*FD-F RMEF I L20 1 Comments L20-0 IMPLIES LEVEL 1; L20-1 IMPLIES LEVgL2; USE MFF TO CHANGE ALF CLASSlA + CLASSiBE + CLASSIC + CLASS1D + CLASSlE + CLASS3A + CLASS3B +

CLASS3C ALO NOCD + CLASSlBL + CLASS2A + CLASS2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4

+ CLASS5) + BUCKET Comments CLASS 30 AND CLASS 4 ARE EVALUATED FOR LERF CILDF CILFAIL CILDS 1 OIS CLASS3A + CLASS3B + CLASS3C + LOW Oil CLASS2A + CLASS2T + NORV* (CLASSIA "÷ CLASSiBE + CLASS1BL+ CLASSIC) +

CLASS1B* (NOACREC + NODC) 014 CLASSlB 013 -OPDEPLI*(CLASSlA + CLASSIC + CLASSOD)

Comments change l hIGH PRESSURE LERF 012 OPDEPL1 (CLASSlA + CLASSIC + CLASSID)

Comments change l hIGH PRESSURE LERF IRI 0I-F* (CLASSlA +'CLASSIC)

IR3" CLASSlBE IR4 CLASSiBL IR5 61=r*CLASSID IR6 O1.S*CLASSID Comments the irginal U1 L2 model IR7 OI-F*CLASSIE IRS OI-S*CLASSIE E-52 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI 5:09 PH 2/9/2006 Page 7 Sr" Split Fraction Assignment Rule

.IR2 01-S Comments LOW PRESSURE INJECTION IMPLICIT IRF I CZ2 IR-F*OI-S CZ4 IR-F*OI-F CZi IR-S*OI-S CZ3 IR-S*OI-F CZF 1 TDI CLASSlE TD2 0I-S*DWSPRAY TD3 -(0I-B)*CLASSlBE TD4 -(OI-B)*CLASSlBL TD8 OI-F*CLASSlA TDF 1 FDl ALTINJRHSW + DWSPRAY FD2 TD=S* (CLASSIA + CLASSiBE + CLASSIBL + CLASSID + CLASS3A + CLASS3B + CLASS3C)

FD3 TD-F* (CLASSlA + CLASSIC + CLASSID + CLASS3A + CLASS3B + CLASS3C)

FD4 TD-F* (CLASSiBE + CLASSIBL)

DWIF 1 WRI DW-S RMEB CLASSIBL Comments TD-S*DWSPRAY*RHRSPCOOL This was an assumption that resulted in 100 RBE RME7 OI-F RME6 OI-S*TD-S*FD-S*DWS-S RMIE5 OI-S*TD-S*FD-S*DWS-F RME4 OI-S*TD-S*FFDF RME3 OI-S*TD-F*FD-F RMEF 1 L20 E-53 Cl 320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI 5:09 PH 2/9/2006 Page 8 SF Split Fraction Assignment Rule Corments L20-0 IMPLIES LEVEL 1; L20-1 IMPLIES LEVEL2; USE MFF TO CHANGE ALF CLASSIA + CLASSiBE + CLASSIC + CLASSID + CLASSIE + CLASS3A + CLASS3B +

CLASS3C AL0 WOCD + CLASSlBL + CLASS2A + CLASS2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4

+ CLASS5) + BUCKET Coments CLASS 3D AND CLASS 4 ARE EVALUATED FOR LERF CILDF CILFAIL CILDS 1 OIS CLASS3A + CLASS3B + CLASS3C + LOW OIl CLASS2A + CLASS2T + N6RV*(CLASSlA + CLASSlSE + CLASS1BL+ CLASSIC) +

CLASSlB*(NOACREC + VODC) 014 CLASSiB 013 -OPDEPLI* (CLASSlA + CLASSIC .+ CLASSID)

Comments change l hIGF PRESSURE LERF 012 OPDEPLI* ICLASSIA + CLASSIC + CLASSID)

Coments change . hIGH PRESSURE LERF IRi OI-F* (CLASSlA + CLASSIC)

IR3 CLASSIBE IR4 CLASSiBL IR5 OI-F*CLASS=D iR6 OI-S*CLASSID Co*ments the irginal Ul L2 model IR7 OI-F*CLASSlE IR8 OI-S*CLASSIE IR2 OI-S Comments LOWPRESSURE INJECTION IMPLICIT IRF 1 CZ2 IR-F*OI-S CZ4 IR-F*OI-F CZ1 IR-S*01-S C23 IR-S*OI-F CZF E-54 C1320503-6924 R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: CETN1 5:09 PM 2/9/2006 Page 9 SP Split rraction .Assignment Rule TD1 CLASSIE TD2 OI-S*DWSPRAY TD3 -(OI-B)*CLASSlBE TD4 -(OI-B)*CLASSlBL TD8 OI-F*CLASSIA TDF 1 FD1 ALTINJRHSW + DWSPRAY FD2 TD-S* (CLASSIA + CLASS1BE + CLASSIBL + CLASSMD + CLASS3A + CLASS3B + CLASS3C)

FD3 TD-F* (CLASSIA + CLASSIC + CLASSID + CLASS3A + CLASS3B + CLASS3C)

FD4 TD-F*(CLASSlBE + CLAS81BL)

DWIF 1 WRI DW-S RMEB CLASSIBL Coznments TD=S*DWSPRAY*RHRSPCOOL This was an assumption that resulted in 100 RBE RME7 OI-F RME6

O-S*TD-S*FD-S*DWS-P RXS5 OI-S*TD-S*FD-S*DWS-F RME4 OI-S*TD-S*FD-F "RME3 OI-S*TD-F*FD-F RMEF 1 E-55 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: CETN1 5:09 PM 2/9/2006 Page I Macro Macro Rule / Comments CIC3LERF CZ-F + RMEF* (CILFAIL+DWI-F+IR-F*TD-S*FD-S)

CZ-F + RME-F* (CILFAIL4DWI-F+IR-F*TD-S*FD-S)

CZ-F + RME-F* ICILFAIL+DWI-F+IR-F*TD-S*FD=S)

CZ-F. + RME-F* (CILFAIL+DWI-F+IR-F*TD=S*FD=S)

CZ-F +÷RME-F* ICILLFAIL+DWI-F+IR-F*TD-S*FD=S)

E-56 C1320503-6924 R2 - 7/10/2006

MODEL Name: Page No. 1 of 4 Event Tree: LPGTETN.ETI 11:33:48 June 27, 2006 IE PCs CIL NDMGE LVPRES CS 'PC ORHXr U2X XrV PCS, ODWS DWS SP OSPR

-X4-X3-Xl m

10 0 Q1, C)

U, 0U C.)

0)

MODEL Name: Page No. 2 of 4 Event Tree: LPGTETN.ETI 11:33:48 June 27, 2006 SPR OWWV VNT CRD OAI AVA] ),# S#

1 1 2 2 3 3 4 4 5 5 6 6 7 7 8 8 9 9

.10 10 m 11 11 01 0, 12 12

...................................................................................................... x1 13 13-22

...................................................................................................... X3 14 23-42

...................................................................................................... X4 15 43-82

...................................................................................................... X4 16 83-122

...................................................................................................... X4 17 123-162

...................................................................................................... X5 18 163-203

...................................................................................................... X4 19 204-243

...................................................................................................... X4 20 244-283 Cz

...................................................................................................... X4 21 284-323 C) ...................................................................................................... X5 22 324-364 0

01 ...................................................................................................... X4 23 365-404 0

...................................................................................................... X4 24 405-444 Co ...................................................................................................... X4 25 445-484

U ...................................................................................................... X4 26 485-524 C.'

27 525 r.3 I I I C")

0 C) 03

MODEL Name: Page No. 3 of 4 Event Tree: LPGTETN.ETI 11:33:48 June 27, 2006 IE PCS CIL NDMGE LVPRES CS LPC ORHXT U2X XTV PCSI ODWS DWS SP OSPR Cii J11 CD St CY)

(JD Ca Ni 0

C0 Li 00 0,

MODEL Name: Page No. 4 of 4 Event Tree: LPGTETN.ETI 11:33:48 June 27,2006 SPR OWWV VNT CRD OAI AVI IX# B#

I I 28 526 I- 29 527 30 528

...................................................................................................... X6 31 529-1055

,m 03 0

CS 0

0 0-0 0-01 0

lo)

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LPGTETN 12:10 PM 6/27/2006 Page 1 Macro I Macro Rule / Comments ALTINJRHSW RPSM-B BUCKET -CLASSIA*.-IB*-CLB*CLASSBE*-CLASSIBLC-CLASSlC*-CLASSlD*-CLASSIE*-CLASS2A*-C LASS2L*-CLASS2T*-CLASS2V*-CLASS3A*-CLASS3B -CLASS3C *-CLASS3D

-CLASS4*-CLASS5 THIS WAS SEPARATED FROM NOCD TO TEST WHAT IS HISSING. IT WILL BE USED IN THE DAMAGE STATES CILFAIL CIL-F CLASSIA -NOCDI*NEMGE-S*(ORVD-F+ RVD-F + OHPC-F*OHPR-F)*LVPRES-F*-CLASS5 NGE-S*VRPSM-S*RPSE-SR*E-( -S)*(-(HPI-S)*-(RCI-S +-(OHPC-S)*-(OHPR-S))*LVPRE S-F CLASSiB RPSM-B CLASSiBE -NOCD1*RPSM=S*RPSE-S*OG5-F*DGC-FEPR30-F*- (TTP-F)*-(IVC-F)* IHPI-F*RCI-F +

-(OH PC-S)*-(OHPR-S) )*-CLASS5.-CIASSIA CLASSIBL -NOCDI*RPSM-S*RPSE-S*OG5-F*DGC-F*EPR6-F*-I TTP-F) *- (IVC-F) *- IHPI-F*RCI-F +

-(OHPC-S)- (OHPR-S) ) *-CLASS5*-CLASS1A*-CLASSlBE CLASSiC RPSH-B CLASSID -NOCDI*RPSM-S*RPSE-S*LVPRES-S* (- (LC-S) *- (CS-S )+

- IOLPC-S) *-CLASS5*-CLASSiA*-CLASSIBE*-CLASSlBL CLASSIE -NOCDI*RPSC-S*RPSELS*LVPRES-F*DE-F*DH-F*DG-F*-CLASS5*-CLASSIA* -CLASSIBE*-CLA SS1BL*-CLASS1D CLASS2A -NOCDi*RPSM-S*RPSE-S*NDMGE-S*- (SP-S*SPRHR+SPR-S*SPRHR) *- (CND-S+PCSR-S) *-CLAS S5*-CLASSlA*-CLAsSlBE*-CLASSlBL*-CLASSID*-CLASSIE CLASS2L -NOCDI*RPSM-S*RPSE-S* lINIT-SLOCA + RVC-SORVi +

RVC-SORV2) *NNMGE-S*- lSP-S*SPRHR+SPR-S*SPRHR) *- (CND-S+PCSR-S) *-CLASS5*-CLASSI A*-CLASS1BE*-CLASSlBL*-CLASSlD* -CLASSlE*-CLASS2A CLASS2T RPSM-B CLASS2V RPSM-B CLASS3A -NOCD1*RPSM-1B E-61 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LPGTETN 12:10 PH 6/27/2006 Page 2 Macro Macro Rule / Comments CLASS3B -NOCDI*RPSM-S*RPSE-S* t INIT-SLOCA+RVC-SORV1 ) *LVPRES-F* -CLASS5*-CLASSIA* -CLA.SS IBE*-CLASSIBL*-CLASSlD--CLASSIE*-CLASS2A* -CLASS2L CIASS3C -NOCD1*RPSM-S*RPSE-S* (RVC-SORVI*HPI-S +

RVC-SORV2 ) * (CS-S+LPC-S) * ( (SP-S+SPR-S) *SPRHR) *-CLASSlA*-CLASSlBE*-CLASSlBL*-C LASSID*-CLASlSE*-CLASS2A*-CLASS2L*-CLASS3B CLASS3D -NOCD1*RPSM=S*RPSE-S -IINIT-SLOCA-*RVC-SORV1+RVC-SORV2) *- (TOR-S) *-CLASSIA*-CLA

.. . SSBE*-CLASSIBL* -C SID*-CLAS$I*-LAS 2A*-CLASS2L*-CIASS3B*- L5SS3................

CLASS4 RPSM-B CL*SS5 -NOCD1*TTP-F*IVC-F CSASUP AAOK*DA-S*RERCVRY* (EECW-S+EECWR-S)

Core Spray pump A support (sans actuation)

CSBSU P ASOK DC-S*RFRCVRY* (EECW-S+EECWR-S) *- (CASTRAN*RVD-S)

Core Spray pump B support loans actuation)

CSCSUP ACOK*DB-S*RERCVRY* (EECW-S+EECWR-S)

Core Spray pump C support loans actuation)

CSDSUP ADOK*DD-S*RFRCVRY* (EECW-S+EECWR-S) *- (CASTRAN*RVD-S)

Core Spray pump B support (sans actuation)

DWSPRAY DWS-S EMDEPHDWR -(RVD-S)*(RB-F*RC-F*RD-F + EPR6-r)

HIGH -LOW HR6ONLY (RCI-S + HP1-Sl-L8F-S (OHPC-S+OHPR-S)

HRLPT RVC-SORV1 + RVO-SORV2 + INIT-SLOCA HPCI/RCIC LOW PRESSURE TRIP; FJMSW SW1+SW2+SW3+SW4+SW5+SW6+SW7+SWB+SW9+SWl 0+SWII+SW12+SW13+SWI 4+SW15+SW16+SW17+

SW18+SW19+SW20+SW21+SW22+SW23+SW24+SW25+SW26+SW27+SW28+SW29+SW30+SW31+SW32+S W33+SW34+SW35+SW36 HX(BSW SWI+SW2+SW7+SW8+SW9+SWIO+SWJI+SW12+SW13+SW14 +SW15+SW22+SW23+SW24+SW25+SW26+S W27+SW28+SW29+SW30+SW37+SW38+SW39+SW40+SW41+SW42+SW43+SW44+SW45+SW46+SW47+SW 48+SW49+SW5O+SW514SW52 E-62 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LPGTETN 22:10 PM 6/27/2006 Page 3 Macro Macro Rule / Comments HXCSW SW3+SW4+SWS+SWg+SW92+SW13+SW16+SW17+SWl8+SW19+SW20÷SW23+SW24+SW27+SW28+SW31+

SW32+SW33+SW34+SW35+SW37+SW38+SW4 1+SW42÷SW4 3+SW4 4+SW4 5+SW4 7+SW4 8+SW4 9+SW50+S WSI+SW53+SW54+SW55+SW56 HXDSW SWSSW6+SWI0+SWI 1+SWI 4+SW15+SW17+SWl B+SWl 9+SW20++SW2l4-SW25+SW26+SW29+SW3O+SW3 2+SW33÷SW34 +SW35+SW36+SW39+SW40+SW42+SW43+SW44+SW45+SW4 6+SW4 B+SW4 9+SW50+SW51

+SW52+÷SW531.SW54+SW55÷SW56

...- OW.........-.......---..----VPRES-S -.....

LPCI2 RVC-SORVX + RVC-SORV2 + INIT-SLOCA Conditions where 2 RXR Pumps/hXs required for suppression pool cooling LPI LPC-S NCDCRDLT (RCI-S+HPI-S)* (OHPR-S+OHPC-S)*CRDIS* ( (SP-S+SPR-S)*SPRPjR+VNT-S)

At six hours, CRD is capable of removing decay heat (kevel control)

NCDHRLT -HRLPT* ( (HPL-S+RCL-S) *(SP-S+SPR-S)*SPRHR)

HPCI/RCIC used for shutdown, HPCI nor RCI tripped due to low pressure; long term injection with suppression pool pressure control NCDLVPRES (LVPRES-S+RVD-S)*tCND-S + PCSR-S + (CS-S+LPC-S +

-MULTIT*XTV-S)*((SP-S+SPR-S)*SPRHRf)

Rmoved OLPC-S; CRD can be used at 4 bra or OLPC is recoverable NCDSORV (RVC-SORV1*CHPI-S+RCI-S) + RVC-SORV21

(CS-S+LPC-Sl* ((SP-S+SPR-SJ *SPRIR)

NOACREC EPR6-F NOCD . NOCD1 NOCDI RPSM-S*RPSE-S* - CNIT-IOOVA+ INIT-LOCHSA+INIT-LOFWA + INIT-LOSPA+INIT-TTA) *

(NCDHRLT + NCDSORV + NCDLVPRES + NCDCRDLT + FWSD-S + PCSR-S)

NONATWS TRANSIENTS NOCDU2X -MULTI T*LVPRES-S*XTV-s*VNT-S NODC DE-F* DG-F*DH-F NORV RVC-SORVO*- (RVD-S)

RVC-SORVO*-(RVD-S)

OPDEPLI ORVD-S E-63 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Model Name: UICOP2-9 Macro for Event Tree: LPGTETN 12:10 PH 6/27/2006 Page 4 Macro Macro Rule / Comments RHRSPCOOL SP-S + SPR-S SORV RVC=SORV1 + RVC-SORV2 SPRHR RPA-S*HXA-S* ( HXASW+OGS-F*EPR6-S) +RPB-S*HXB-S* (HXBSW+OGS-F*EPR6-S) +RPC-S*HXC-S* (HXCSW+OGS-F*EPR6=S)+RPD-S*HXD-S* (HXDSW + OGS-F*EPR6-S) + XTV-S*-MULTIT

. i.. SW.A.S*SW2A-S*SWlB*.

SW10 SWlA-S*SW1B-S* SWlD-S SWil SWIA-S*SWIB-S*SW2D=S SW12 SWlA-S-SW2B-S*SWlC-S SW13 SWIA-S*SW2B-S*SW2C-S SW24 SWIA-S*SW2B0S*SWID-S SWis SWlA-S*SW2B-S*SW2D.S SW16 SWIA-S*SWIC-S*SW2C-S SWi, SWIA-S* SWlC-S*SWlD-S SWI8 SWlA=S*SWIC-S*SW2D-S SWI 9 SWIA-S*SW2C-S*SWI D-S 5W2 SWIA-S*SW2A-S* SW2B-S SW20 SWIA-S*SW2C-S*SW2D*S SW21 SWIA-s*SWID=S*SW2D-s SW22 SW2A=S*SWflBS*SW2B=S E-64 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: LPGTETN 12:10 PM 6/27/2006 Page 5 Macro Macro 1ule / Comnenta SW23 SW2A-S*SWIB-S*SWIC-S SW24 SW2A-S* SWIB-S*SW2C-S SW25 SW2A-S*SWIB-S*SWID-S SW26 SW2A-S*SWIB-S*SW2 D-S SW27 SW2A-S*SW2B-S*SWIC=S SW28 SW2A-S*SW2B-S*SW2C-S SW29 SW2A-S*SW2B-S* SWID-S SW3 SWIA-S SW2A-S* SWIC-S SW30 SW2A-S*SW2B-S*SW2D-S SW31 SW2A-S*SW1C-S*SW2C-S SW32 SW2A=S*SWIC-S*SWID-S SW33 SW2A-S*SWIC-S* SW2D-S SW34 SW2A-S*SW2C-S*SWID-S SW35 SW2A-S*SW2C-S*SW2D-S 3W36 SW2A-S*SWID-S* SW2D-S SW37 SWIB-S*SW2B-S*SWIC-S SW38 SWlB-S*SW2B-S*SW2C-S SW39 SW1B-S*SW2B-S* SWID-S E-65 C 1320503-6924 R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk Assessment model Name: t31C0P2-9 Macro for Event Tree: LPGTETN 12:10 PM 6/27/2006 Page 6 Macro Macro Rule /Comments SW4 SW1A-S*SW2A-S*SW2C-S SW4 0 SWlB-S*SW2B-S*SW2D-S SW41 SWlB-S*SWlC=S*SW2C-S SW~43 SWlB-S*SWlC-S*SW2D-S SW44 SWlB-S*SW2C-S*SWlD-S SW45 SW1B-S*SW2C-S*SW2D-S SK4 6 SWlB=S*SWID=S*SW2fl=S S%147 SW2B=S*SWlC=S*SW2C=S GW48 SK2B-S* SW1C-S*SWID=S SW149 SW2B-S*SWIC-S-SW2D-S SM5 WI1A-S*SW2A-S*S1'l f-S 51450 SW2B-S*SW2C-S* SWID-S 51451 SWBSS2CSS2 SW452 5142B-S*SW1O.S*SW2D-S SW53 SWlc-S*SW2C-S*SWlD-S SW54 gWlC5*SK2C-S*SW2D-S 51455 SWIC-S*SWID-S*SW2D-S E-66 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Model Name: UICOP2-9 Macro for Event Tree: LPGTETN 12:10 PM 6/27/2006 Page 7 Macro Macro Rule / Coimnents SW56 SW2C-S*SWlD-S*SW2D-S SW6 SWIA-S*SW2A-S*SW2D-S SW7 SWIA-S*SWIB=S*SW2B-S

...... - SW...............- SWIAS SB-S*sWICS ......

5W9 SWIA-S* SWIB-S* SW2C-S E-67 E-67C1320503-6924R2 -7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Ansignment Rule for Event Tree: LPGTETN 12:10 PM 6/27/2006 Page I sP Split Fraction Assignment Rule PCSS FWSD-S + (INIT-IOOVA+ INIT-LOCHSAMINIT-LOFWA + INIT-LOSPA+INIT-TTA)

Comments Successtul it the main condenser, condensate, and teedwater hardware (sans level control) are functional, and the feedwater controller is functional or operators control level or a Level 8 trip occurs with successful operator acti PCSF FWSD-F NDMGES 1 FWH-F*(RCI-F*HPI-F+OHPC-F*OHPR-F)*(ORVD-F + RVD-F)

PCSRb INIT-LbUUU+ INIT-LbUUPA PCSRb INIT-TBU PCSR4 INIT-LOPA PCSR3 INIT-IMSIV PCSR2 INIT-TLFW + INIT-TLCF PCSRl INIT-FLRBJS PCSRF 1 LVPRS (HPI-S + RCI-S)-(OHPC-S+OHPR-S) + RVC-SORVI + RVC-SORV2 + RVD-S Comments The vessel is at low pressure it HPCI ran tor sIx hours, emergency depressurization, or a stuck open SRV or 2 SORVs LVPRF 1 CSF (-CSASUP--CSCSUP + EECW-F-EECWR-F +TOR-F +

-(LV-S)*(-(DW-S)+-(NPI-S))*-(OLPC-S)) * (-CSBSUP*-CSDSUP+(EECW-S+EECWR-S)

+TOR-F+ -(LV-S)*(-(DW-F )+ NPI-S)*-(OLPC-S)) + INIT-FLRB2 + INIT-FLRB3S +

CILFAIL Comments CS bails due to NPSH It there is a containment breach (CIL-F)

CS3 (CSASUP-CSCSUP*(LV-Ss+DW-S*NPI-S+OLPC-S)) *

(CSBSUP*CSDSUP*LV-S* (DW-S*NPII-S+OLPC-S)) * (EECW-S+EECWR-S)*TOR-S CS4 (CSASUP*CSCSUP*LV-S-DW-S*NPI-S) * (CSBSUP-CSDSUP*LV-S-DW-S-NPUI-S)

(EECW-S+EECWR-S)*TOR-S *OLPC-F CS! (CSASUP-CSBSUP-CSCSUP--CSDSUP + CSASUP-CSBSUP-CSCSUP-CSDSUP +

CSASUP*-CSBSUP*CSCSUP*CSDSUP + -CSASUP*CSBSUP*CSCSUP*-CSDSUP) *

(LV-S+DW-S*NPI-S*NPII-S + OLPC-S)

Comments 3 pumps supported with OLPC-S CSbA (CSASUP*CSBSUP*CSCSUP*-CSDSUP + CSASUP*CSBSUP*-CSCSUP-CSDSUP +

CSASUP*-CSBSUP*CSCSUP*CSDSUP + -CSASUP*CSBSUP*CSCSUP*-CSDSUP) *

(LV-S+DW-S*NPI=S*NPII-S)*OLPC-F Comments 3 pumps supported with OLPC-S CSb CSASUP*CSCSUP*(LV-S+DW-S-NPI-S + OLPC-S) + CSBSUP-CSDSUPC(LV-S+DW-S*NPII-S +

OLPC-S)

E-68 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: LPGTETN 12:10 PH 6/27/2006 Page 2 SF Split Fraction Assignment Rule Comments Conditions for all support already asked so LOOPI*-LOOPII +

-LOOPI*LOOPII not necessary CS7 CSASUP*CSCSUP*(LV-S+DW-S*NPI-S)*OLPC-F +

CSBSUP*CSDSUP*(LV-S+DW-S*NPII-S)*OLPC-F CS8 (LV-S + DW-S*NPI-S*NPII-S + OLPC-S) * (CSASUP + CSCSUP)*(CSBSUP + CSVSUP)

Comments Support only for 2 pumps in different loops; heirarchy used -

negatives not shown CSaA OLPC-F*(LV-S + DW-S*NPI-S*NPII-S) * (CSASUP + CSCSUP)-(CSBSUP + CSDSUP)

CS9 (LV-S + DW-S*NPI-S+ OLPC-S) * (CSASUP + CSCSUP) + (LV-S + DW-S*NPII-S +

OLPC-S) * (CSBSUP + CSDSUP)

Comments Support for 1 pump only; heirarchy used - negatives not shown CS9A OLPC-F*{LV-S + DW-S*NPI-S) * (CSASUP + CSCSUP) + OLPC-F*(LV-S +

DW-S*NPII-S) * (CSBSUP + CSDSUP)

Comments Support for I pump only; heirarchy used - negatives not shown LPCF CS-B + LV-F*DWIF*NPI-F*NPII-F*OLPC-F + -RERCVRY*-RFRCVRY + CILFAIL Comments LPC fails due to NPSH if there is a containment breach (CIL-F)

LPCI RERCVRY*RFRCVRY*OLPC-S LPC2 RERCVRY*RFRCVRY*-(OLPC-S)

LPC3 RFRCVRY*-RERCVRY*OLPC-S LPC4 RFRCVRY*-RERCVRY*-(OLPC-S)

LPC5 RERCVRY*-RFRCVRY*OLPC-S LPC6 RERCVRY*-RFRCVRY*-(OLPC-S)

LPCF I ORHXTS I=A-S*HXB-S*HXC-S+ HXA-S*JXB-S*HXD-S + HXA-S*HXC-S*HXD-S + HXB-S*HXC-S*HXD-S OPWTS RF-F+RH-F+EECW-F-EECWR-r+RF-F*RI-F+ (AB-F+DC-F+SWlC-F*SU2C-F) * (AA-F+DA-F+SWIA

-F*SW2A-F) + INIT-FLRBI Comments PASS THROUGH IF SUPPORT FOR XTIE NOT AVAILABLE OREXT1 I U2XF RF-F+RH-F+EECW=F*EECWR-F+RF-F*RI-F+(AB-F+DC-F+-HXCSW)*(AA-F+DA-F+HXASW) +

INIT-FLRB1 + CILFAIL Comments U2X crosstie fails due to NPSH if there is a containment breach (CIL-F)

U2X5 (AA-F+DA-F+ -HXASW) *RI-F U2X6 (AB-F+DC-F+ -HXCSW) *RI-F U2X3 (AA-F+DA-F+ -HXASW)*RI-S U2X4 (AB-F+DC-F+ -HXCSW)*RI-S E-69 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: LPGTETN 12:10 PM 6/27/2006 Page 3 SF Split Fraction IAssignmant Rule U2X2 RI-F U2X1 1 XTVS HXA-S*HXB-S*HXC-S+ HXA-S*HXB-S*HXD-S + HXA-S*HXC-S*HXD-S + HXB-S*RXC-S*HXD-S XTV1 RF-S XTVF 1 PCSRF INIT-LOSP + RVC-SORVI + RVC-SORV2 +

FWH-F* (RCI-F*BPI-F+OHPC-F*OHPR-FI (ORVD-F + RVD-F) 4 CS-F*LPC-F PCSR6 INIT-L500U+ INIT-L500PA PCSR5 INIT-TBU PCSR4 INIT-LOPA PCSR3 INIT-IMSIV PCSR2 XNIT-TLFW + INIT-TLCF PCSRI INIT-FLRB3S PCSRF 1 ODWSI 1 DWSF PXi-F*PX2-F + ((RPA-F+HXA-F)* (RPC-F+HXC-F)

+RE-F+NOGA) * ( (RPB-F+HXC-F) *(RPD-F÷HXC-F) +RF-F+ NOGC) +ODWS-F + CILFAIL Comments DWSfails due to NPSH if there is a containment breach (CIL-F)

DWSl PX1-S*PX2-S * (RPA-S*HXA-S +

RPC-S*HXC-S) *RERCVRY*GA-S (RPB-S*HXB-S+RPD-S*XD-S) *RFRCVRY*GC-S*ODWS-S DWS2 (PXI-S* (RPA-S*HXA-S + RPC-S*HXC-S)*RERCVRY*GA-S +

PX2-S (RPB-S*11XB-S+PRPD-S-HXD-S) RTRCVRY*GC-S) *ODWS-S DWSF 1 SPF OSPC-F + CILFAIL Comments SP fails due to NPSH if there is a containment breach (CIL-F)

SPI - (LPC-S)*RERCVRY*RFRCVRY* (RPA-S*HXA-S + RPC-S*HXC-S)* (RPB-S*HXB-S +

RPD-S*HXD-S + XTV-S*-MULTIT)

Comments Only RMOV 1A and 2B boards are needed SP2 - (LPC-S)* (- (RERCVRY* (RPA-S*HXA-S + RPC-S*HXC-S) )*RFRCVRY* (RPB-S*HXB-S +

RPD-S*HXD-S *XTV-S*-MULTIT) + RERCVRY* (RPA-S*)Xk-S +

RPC-S*HXC-S)*-(RFRCVRY*(RPB=S*HXB-S + RPD-S*HXD-S + XTV-S*-MULTIT)))

SP3 LPC-S*RERCVRY*RFRCVjY*RCOK*RBOK* (RPA-S*HXA-S + RPC=S*HXC-S)* (RPB-S*HXB-S +

RPD-S*HXD-S + XTV-S*-MULTIT)

E-70 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: LPGTETN 12:10 PM 6/27/2006 Page 4 SF Split Fraction Assignment Rule SP4 LPC-S C-CfRERCVRY*RCOK* (RPA-S*1xA-S + RPC-S*HXC-S) )-RFRCVRY*RBOK- (RPB-S-HXB-S

+ RPD-S*HXD-S + XTV-S*-MULTIT) + RERCVRY*RCOK* (-(rRFCVRY*RBOK* (RPB-S*HXB-S

+ RPD-S*HXD=S + XTV-S*-MULTIT) )* (RPA-S*HXA-S + RPC-S*HXC-S)))

SP!) RBOK*RCOK-RERCVRYPRFRCVRY* (RPA-S-HXA=S + RPC-S*HXC-S)* (RPB-S*HXB-S +

RPD-S*HXD-S + XTV-S*-MULTIT)

SPF 1 OSPRF (RPA-F+HXA-F)* (RPB-F+HXB-F)* (RPC-F+HXC-F) (RPD-F+HXC-F) + OSPC-F Comments Support for SPR not available OSPRI 1 SPRF (RPA-F+HXA-F) * (RPB=F+HXB-Fl - CRPC-F+HXC-F *-(RPD-F+HXC-F) + OSPC-F + CILFAIL Comments SPR tails due to NPSH it there is a containment breach ICIL-F)

SPRI RPSM-S Comments This has been simplified and is slightly conservative SPRF I owwvS (HXASW+HXBSW+HXCSW+HXDSW) (SP-S+ SPR-S) + OG5-FIEPRb-S owwvi 1 VNTNN (HXASW+HXBSW+HXCSW+HXDSW) (SP=S+ SPR-S) + OGb-F*EPRb-S VNTF OWWV-F VNT1 RBOK*RCOK*PCA-S VNT2 OWWV-S VNTF 1 CRD2 (CST-S + RCW-S)*Us41C-S + OGb-F*EPRb-S*CST-S CRD3 (CST-S + RCW-S)*UB41C-F*AA-S + OGb-F*EPR6-S*CST-S CRDF 2.

OAIF I AVIF SWID-F+SW2D-F+RF-F AVII 1 E-71 C1320503-6924R2 - 7/10/2006

MODEL Name: Page No. 1 of 3 Event Tree: ATWS3N.ETI 11:42.21 June 27,2006 CIL A3S A3D FWSD EECW OREE EECWR REPWR RFPW" OLPC OSPC RPA HXA I IE

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MODEL Name: Page No. 2 of 3 Event Tree: ATWS3N.ETI 11:42.21 June 27,2006 RPC HXC LPCI SPI RPB HXB RPD HXD LPCII SPitI ORHXT U2X XTV CS

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MODEL Name: Page No. 3 of 3 Event Tree: ATWS3N.ETI 11:42:21 June27, 2006 X# B3# ,S#

1 1 2, 2 3 3 4 4 X2 5 5-7 X3 6 8-13 X4 7 14-25 X5 8 26-49 X6 9 50-97 X7 10 98-193 m X8 11 194-385 X9 12 386-769 X1o 13 770-1537 X1l 14 1538-3073 X12 15 3074-6145 X13 16 6146-12289 X14 17 12290-24577 X15 18 24578-49153 X16 19 49154-98305 X17 20 98306-196609 X18 21 196610-393217 X19 22 393218-786433 X20 23 786434-1572865 o 24 1572866 25 1572867 26 1572868 27 1572869 X21 28 1572870-3145738 or.

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BFN EPUCOP ProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: ATWS3N 1:02 PH 6/27/2006 Page 1 SF Split Fraction Assignment Rule A3SY A2S-S- (OSV-S*MCD-S*CND-S*FWH-S) TOR-S*OAL-S*OTAF-S*OHPC-S + A2S-F Comments RECALL SUCCESS IS DOWN BRANCHI A3SN 1 Comments SUCCESS CANNOT BE DETERMINED IN ATWS2 AJDY A2D-P + HPI-F- (OSV-F + MCD-F 4CND-F + FWH-F + LVF-F) + TOR-F + OAL-F +

OTAF-F + OHPC-F Comments UNCHANGED FROM ATWS2 A3DN 1 FWSDF INIT-LOCHSA FWSDS (RVC-SORVU+RVC-SORV1) MCD-S'CND-S*FWH-S FWSDF 1 EECWF EA-F* EB-FEC-F+EA-P-EB-F*ED-F+EA-F-EC-F*ED-F+EB-F-EC-F*ED-F EECWS 1 OREENN GA-F- GB-F*GC-F GD-F*GE-F*OF-F*GG-F*GH-F-SWINGIC*-SWING1D OREE2 SWING2C-SWINGID OREE1 SWINGIC+SWINGID EECWRS SWINGIC-SWINGID-SWIC-S*SWID-S + SWINGIC*-SWINGID'SWlC-S +

-SWING1C*SWINGID*SWID-S EECWRF 1 REPWRS ROOK Comments 4BU V RMOV BOARD 1A IS RECEIVING PWR FROM 4UU V SD ED 1A REPWRF -ROOR Comments 4tU V RMOV BOARD IA IS NOT RECEIVING PWR FROM 4UU V SD BD 1A RFPWRS RROK Comments 48U V RMOV BOARD IA IS RECEIVING PWR FROM 4VU V SD BD IB RFPWRF -RROK Comments 48U V RMOV BOARD 2A IS NOT RECEIVING PWR FROM 48U V SD E6 IB OLPCNN HPI-S + RCI-S OLPCl I Comments OPERATOR INITIATES LPC (STARTS PUMPS) HUGE QUESTION HERE ON PWER WXCURSION OSPC2 1 RPAF -RPASUP+ INIT-FLRBZ + INIT-FLRBJS + CILFAIL E-75 C1320503-6924 R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: U1COP2-9 Split Fraction Assignment Rule for Event Tree: ATWS3N 1:02 P* 6/27/2006 Page 2 OF Split Fraction 1Auignment Rule Comments RPA fails due to NPSH if there is a containment breach (CIL-F}

RPA1 RPASUP RPAF 1 HXAF -HXASUP 1DIA1 HXASTJP RPCF -RPCSUP + INIT-FLRB2 + INIT-FLRB3S + CILFAIL Comnents RPC fails due to NPSH if there is a containment breach (CIL-F)

RPCl RPCSUP*RPA-S RPC3 RPCSUP*-RPASUP RPC2 nPCSUP*RPASUP*RPA-F RPCF 1 HXC1 HXA-S HXC2 HXA-F*XSUP HXC3 HXA-F*.-HXASUP HXCF I SPIF OSPC-F + REPWR-F SPIl NOLPCI*REPWR-S*(RPA-S*HXA-S + RPC-S*HXC-S)

Comments Only RMOV 1A and 1B boards are needed SP12 -NOLPCI*REPWR-S*RC-S*(RPA-S*HXA-S + RPC-S*HXC-S)

SPIF 1 LPCIF -LPCISUP + CILFAIL Comments LPCI fails due to NPSH if there is a containment breach (CIL-F)

LPCI2 LPCISUP Comments MANUAL LPCI START NOT CREDITED LLOCAS; ODD SPLIT FRACTION SWOULD APPLY LPCIIF -LPCIISUP + CILFAIL Comments LPCII fails due to NPSH if there is a containment breach (CIL-F)

LPCII2 LPCI-S LPCII4 -LPCISUP LPCII6 LPCI-F*LPCISUP E-76 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: ATWS3N 1:02 PM 6/27/2006 Page 3 Sp Split Fraction Assignment Rule RPBF -RPBSUP+ INIT-FLRB2 + INIT-FLRB3S + CILFAIL Comments RPB fails due to NPSH if there is a containment breach (CIL-F)

RPB6 RPBSUP* (-RPASUP*RPC-P*RPCSUP+RPA-r*RPCSUP* (RPC-B+-RPCSUP))

RPB5 RPBSUP* (-RPASUP*RPC-S+RPA-S* (RPC-B+-RPCSUP))

RPB4 RPBSUP*-RPASUP* (RPC-B+- (RPCSUPl RPB2 RPBSUP* (RPA-S*RPC-F+RPA-F*RPC-S)

RPBI RPBSUP*RPA-S*RPC-S RPBF I HXBF -HXBSUP HXB1 HXA-S*HXC-S HXM6 -HXASUP*-HXCSUP HXB5 HXASUP*HXA-F*HXCSUP*HXC-F HXB4 HXA-F*-HXASUP*HXC-F*HXCSUP + HXA-F*HXASUP*HXC-F*-HXCSUP HXB3 RXA-F*-HXASUP*RXC-S + HXA-S*HXC-F*-HXCSUP HXB2 HXA-F*HXASUP*HXB-S + HXA-S*HXB+F*HXBSUP HXBF 1 RPDF -RPDSUP+ INIT-FLRB2 + INIT-FLRB3S + CILFAIL Comments RPD fails due to NPSH if there is a containment breach (CIL-F)

RPDI0 RPDSUP* (RPA-F* (RPC-F*-RPBSUP+ (RPC-B+-RPCSUP) *RPB-F) +-RPASUP*RPC-F*RPB-F)

RPD9 RPDSUP* (RPA-S* (RPC-F*RPCSUP*-RPBSUP+ (RPC-B+-RPCSUP) *RPB-F*RPBSUP) +RPC-S* (RPA

-F*RPSUP*-RPBSUP+-RPASUP*RPB-F*RPBSUP +RPB-S* (RPA-F*RPASUP* CRPC-B+- (RPCSfP)

+- (RPASUP *RPC-F*RPCSUP)))

RPD8 RPDSUP* CRPA-S* (RPC-S*- (RPBSUP) + (RPC-B+- (RPCSUP) *RPB-S) +- (RPASUP) *RPC-S*RPB-S)

RPD7 RPDSUP* (- (RPASUP) * (RPC-F*RPCSUP*- (RPBSUP) + (RPC-B+- (RPCSUP) ) *RPB-F*RPBSUP) +RP A-F*RPASUP* (RPC-B+- (RPCSUP)) *- (RPBSUP))

RPD6 RPDSUP* (- (RPASUP)* (RPC-S*-(RPBSUP) + (RPC-B+- (RPCSUP))*RPB-S) +RPA-S* (RPC-B+- (R PCSUP) )*-(RPBSUP) )

RPD5 RPDSUP*- LRPASUP) * (RPC-B+- (RPCSUP) )*- (RPSSUP)

RPD4 RPDSUP*RPA-F*RPASUP*RPC-F*RPCSUP*RPB-F*RPBSUP E-77 C1320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: ATWS3N 1:02 PM 6/27/2006 Page 4 BF Split Praction *Asignment Rule RPD3 RPDSUP* (RPA-F*RPASUP* (RPC-F*RPCSUP*RPB-S+RPC-S*RPB-F*RPBSUP) +RPA-S*RPC-F*RPC SUP*RPB-F*RPBSUP)

RPD2 RPDSUP* (RPA-F*RPASUP*RPC=S*RPB=S+RPA-S*RPC=F*RPCSUP*RPB-S+RPA-S*RPC=S*RPB-F*

RPBSUP)

RPD1 RPDSUP*RPA-S*RPC-S*IRPB-S RPDF 1 HXDF -HXDSUP HXD10 -HXASUP*-HXCSUP*-HXBSUP HXD9 -HXASUP*-HXCSUP*HXBSUP*HXB-F + -HXASUP*HXCSUP*HXC-F*-HXBSUP +

HXASUP*HXA-F*-HXCSU P*-HXBSUP HXDB -HXASUP*-HXCSUP*HXB-S + -HXASUP*HXC-S*-HXBSUP + HXA-S*-HXCSUP*-HXBSUP HXD7 HXASUP*HXCSUP*HXBSUP*HXA-F*HXC-F*HXB-F HXD6 HXASUP*HXA-F*HXCSUP*HXC=F*-HXBSUP + HXASUP*HXA=F*-HXCSUP*HXBSUP*HXB-F +

-HXASUP*HXCSUP*HXC=F*HXBSUP*HXB-F HXD5 HXASUP*HXA-F*HXCSUP*HXC-F*HXB-S+ HXASUP*HXA-F*HXC-S*HXBSUP*HXB-F +

HXA-S*HXCSUP*HXC-F*HXBSUP*HXB=F HXD4 -MXASUP*HXCSUP*HXC-F*HXB-S + HXASUP*HXA-F*HXC-S*-HXBSUP +

HXA-S*HXCSUP*HXC-F* -HXBSUP HXD3 HXASUP*HXA-F*HXC-S*HXB-S + HXA-S*HXC-S*HXBSUP*HXB-F +

HXA-S*HXCSUP*HXC-F*HXB-S HXD2 -HXASUP*HXC-S*HXB=S + HXA-S*HXC-S*-HXBSUP + HXA-S*-HXCSUP*HXB-S HXD1 HXA-S*HXC-S*HXB-S HXDF 1 SPIUF OSPC-F + RFPWR-F + CILFAIL Corments

SPIZ fails due to NPSH if there is a containment breach (CIL-F)

SPill NOLPCI*RFPWRS* (RPB-S*HXB=S + RPD-S*HXD=S) *SPI-S SPII2 NOLPCI*RFPWR-S* (RPB=S*HXB-S + RPD-S*HXD-S) *SPI-F*REPWR-S SPII3 NOLPCI*RFPWR-S* (RPB-S*HXB-S + RPD-S*HXD-S) *SPI-F*REPWR-PF SPII4 -NOLPCI*RFPWR-S* (RPB-S*HXB-S + RPD-S*HXDfS) *SPI-S SPI15 -NOLPCI*RFPWR-S* (RPBS*HXB-S + RPD-S*HXD-S *SPI-F*REPWR-S SPI16 -NOLPCI*RFPWR-S*(RPB-S*HXB-S + RPD-S*HXD-S)*SPI-F*REPWR-P SPZIF 1 LPCIIF -LPCIISUP E-78 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: ATWS3N 1:02 PH 6/27/2006 Page 5 SF Split Fraction Assignmealt Rule LPCII2 LPCI-S LPCI4 -LPCISUP LPC116 LPCI-F*LPCISUP ORHXTS HXA-S*HXB-S*HXC-S+ HXA-S*HXB-S*HXD-S + HXA-S*HXC-S*HXD-S + HXB-S*HXC-S*HXD-S ORHXTS RF-F+RH-F+EECW-F*EECWR-+RF-F*RI-F+ (AB-F+DC-P+SWlC-F*SW2C-F) * (AA-F+DA-F+SWIA

-F*SW2A-F) + INIT-FLRB1 Comments PASS THROUGH IF SUPPORT FOR XTIE NOT LA t =L ORHXTi 1 U2XF RF-F+RH-F+EECW-F*EECWR-F+RF-F*R-F+ (AB-F+DC-F+-HXCSW) * (AA-F+DA-F+HXASW) +

INIT-FLRB1 + CILFAIL Comments U2X fails due to NPSH if there is a containment breach (CIL-F)

U2X5 (AA-F+DA-F+ -HXASW)*RI-F U2x6 (AB-F+DC-F+ -HXCSWJ*RI-F U2X3 (AA-F+DA-F+ -HXASW)*RI-S U2X4 (AB-F+DC-F+ -HXCSW)*RI-S U2X2 RI-P U2X1 1 XTVS HXA-S*HXB-S*HXC-S+ HXA-S*HXB-S*HXD-S + HXA-S*HXC-S*HXD-S'+ HXB-S*HXC-S*HXD-S XTV1 RF-S XTVF 1 CSF (-CSASUP*-CSCSUP + EECW-F*EECWR-F +TOR-F +

-(LV-S}*(-(DW-S)4-(NPI-S)l*-COLPC-S) * (-CSBSUP*-CSDSUP+(EEcw-S+EECWR-S)

+TOR-F+ -(LV-S)*(-(DW-F )+ NPI-S)*-(OLPC-S)) + INIT-FLRB2 + INIT-TLRB3S +

CILFAIL Comments CS fails due to NPSH if there is a containment breach CCIL-F)

CS3 (CSASUP*CSCSUP*(LV-S+DW-S*NPI-S+OLPC-S)) *

(CSBSUP*CSDSUP*LV-S*(DW=S*NPII-S+OLPC-S)) * (EECW-S+EECWR-S)*TOR-S CS4 (CSASUP*CSCSUP*LV-S*DW-S*NPI-S) * (CSBSUP*CSDSUP*LV-S*DW=S*NPII-S) *

(EECW-S+EECWR=S)*TOR-S *OLPC-F CS5 (CSASUP*CSBSUP*CSCSUP*-CSDSUP + CSASUP*CSBSUP*-CSCSUP*CSDSUP +

CSASUP*-CSBSUP*CSCSUP*CSDSUP + -CSASUP*CSBSUP*CSCSUP*-CSDSUP) *

(LV-S+DW-S*NPI-S*NPII-S.+ OLPC-S)

Comments 3 pumps supported with OLPC-S CS5A (CSASUP*CSBSUP*CSCSUP*-CSDSUP + CSASUP*CSBSUP*-CSCSUP*CSDSUP +

CSASUP*-CSBSUP*CSCSUP*CSDSUP + -CSASUP*CSBSUP*CSCSUP*-CSDSUP) *

(LV-S+DW-S*NPI-S*NPII-S)*OLPC-F E-79 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: ATWS3N 1:02 PM 6/27/2006 Page 6 SF Split Fraction Assignmant Rule Comments 3 pumps supported with OLPC-S CS6 CSASUP*CSCSUP* (LV-S+DW=S*NPI-S + OLPC-S) + CSBSUP*CSDSUP* (LV-S+DW=S*NPII-S +

OLPC-S)

Comments Conditions for all support already asked so LOOPX*-LOOPIX +

-LOOPI*LOOPII not necessary CS7 CSASUP*CSCSUP*(LV-S+DW-S*NPI-S)*OLPC-F +

CSBSUP*CSDSUP*(LV-S+DW-S*NPII-S)*OLPC-F S .- S + OLPC-S *JCSSUP + CSCSUoP * (CSBSUP + CSDSUP)

Comments Support only for 2 pumps in different loops; heirarc- Tu-ed negatives not shown CSaA OLPC-F*(LV-S +'DW-S*NPI-S*NPII-S) * (CSASUP + CSCSUP)*(CSBSUP + CSDSUP)

CS9 (LV-S + DW-S*NPI-S+ OLPC-S) * (CSASUP + CSCSUP) + (LV-S + DW-S*NPII-S +

OLPC-S) * (CSBSUP + CSDSUP)

Comments Support for 1 pump only; heirarchy used - negatives not shown CS9A OLPC-F*(LV-S + DW-S*NPI-S) * (CSASUP + CSCSUP) + OLPC-F*(LV-S +

DW-S*NPII-S) * (CSBSUP + CSDSUP)

Comments Support for I pump only; heirarchy used - negatives not shown E-80 C1320503-6924R2 - 7/10/2006

BFNEPU COPProbabilisticRisk Assessment Model Name: UlCOP2-9 Macro for Event Tree: ATWS3N 1:02 P*M6/27/2006 Page 1 Macro Macro Rule / Comments CILFAIL CIL-F CSASUP AAOK*DA-S*REPWR-S* (EECW-S+EECWR-S)

Core Spray pump A support Jsans actuation)

CSBSUP ABOK*DC-S*RFPWR-S* (EECW-S+EECWR-S)

Core Spray pump B support (sans actuation)

Core Spray pump C support (sans actuation)

CSDSUP ADOK*DD-S*RFPWR-S* (EECW-S+EECWR-S)

Core Spray pump B support (sans actuation)

HRLPT RVC-SORV1 + RVC-SORV2 + RVD-F + INIT-SLOCA HPCI/RCIC LOW PRESSURE TRIP; HWFHXA HXA-F*HXASUP HWEEXB HXB-F*HXBSUP HWFHXC HXC-F*HXCSUP HWMXD HXD-F*HXDSUP

[sans actuation)

HWFRPA RPA-F*RPASUP HWFRPB RPB-F*RPBSUP HWFRPC RPC-F*RPCSUP HWFRPD RPD-F*RPDSUP HXAB RH-F+SW2A-F* SWA-F÷NOGB+HXA-B+RPA-F HXASUP REPWR-S*HXASW*RPA-S HXEB RI-F+SW2B-F*SWIB-F+NOGD+HXB-B+RPB-F HXBSUP RFPWR-S*HXBSW*RPB-S E-81 C1320503-6924R2 - 7/10/2006

BFN EPU COP ProbabilisticRisk Assessment Model Namce: U1COP2-9 Macro for Event Tree: ATWqS3N 1:02 PM 6/27/2006 Page 2 Macro Macro Rule / Co~ments HXCB RH-F4SW2C-F SWlO-F+NOGB+HXC-B+I'PC-F IVCCSLIP RE PWR -S*HXCSW*RPc-S HXDSUP RPWR-S*HXflSW*RPD-S LOOPIIR1~R 1UR2+RIHR4+INIT-LLDA*CS-F____

LOOPIRHR RHR1+RRR3+INIT-LLDB*CS-F LPCIISUP RE--S* ( (NPII-S*DW-S) + LV-S)

LPCISUP RE-S*( INPI-S*DW-S) + LV=S LOOP I LPCI SUPPORT NOGA GA-S*- (EECW-S+EECWR-s)

NOGB GB-S*- (EECW-S+EECWR-S)

NOGC GC-S*-(EECW-S+EECW~R-S)

NOGD GD-S*- (EECW-S+EECWR-S)

NOGE GE.-S*-(EECW-S+EECWR-S)

NOGF GF-S*- (EECW-S+EECWR-S)

NOGG OGG-S*- (EECM-S+EECWR-S)

NOGH GH-S*- IEECM=S+EECWR-S)

NOLPCI MCD-F* (RVC-SCRVO+RVC-SORV11*HI-S RHOK (ABOK+AC-S+7ADOK) *XNITLOSP+RH-S+EPR6..S*MS5 RHfRSW1 SW2B-S+SW1E-S+SW2D-S+SWlO-S E-82 C1 320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Macro for Event Tree: ATWS3N 2:02 PM 6/27/2006 Page 3 Macro Macro Rule / Comments RIOK (ABOK+ACmS+ADOK3

INIT-LOSP+RI=S+EPR6-S*OGS-F RPASUP AAOK*DA-S* (RCW-S+EECW-S+EECWR-S)*REPWR-S* ISIGI*RCOK+OLPC-S + OSPC-S)

RPBSUP ACOK*DB-S* (RCW-StEECW-S+EECWR-S) *RFPWR-S* (SIGII*RBOK+OLPC.S+ OSPC-S)

RPCSUP ABOK*DC-S* (RCW-S+EECW-S+EECWR-S) *REPWR-S* (SIGI* (RBOK+RCOK) +OLPC-S+OSPC-S)

RPDSUP ADOK*DD=S* (RCW-S+EECW-S+EECWR-S) *RFPWR-S* (SIGII* (RBOK+RCOK) +OLPC-S+OSPC-S)

SIG3 (LV-S+DW-S)

SWINGiC SWC-S* (EA-F*EB-F*EC-F*ED-F + EA-F*EB-F*EC-F*ED=S + EA-F*EB-S*EC-F*ED-F +

EA-S*EB-F*EC-F*ED-F)

SWING1D SWlD-S* CEA-F*EB-F*EC-F*ED=F + EA-F*EB-F*EC-S*ED-F + EA-F*EB-S*EC-F*ED-F +

EA-S-EB-F*EC-F*ED-F)

E-83 C1 320503-6924R2 - 7/10/2006

BFN EPU COPProbabilisticRisk Assessment ATWS4N Event Tree

No ATWS4N event tree structure included here, as no tree structure modifications were made for this analysis (i.e., tree structure same as base BFN PRA ATWS4 event tree).

No ATWS4N macros print-out provided here as no new macros were defined for this tree for this analysis.

E-84 C1320503-69241R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk Assessment Model Name: UICOP2-9 Split Fraction Assignment Rule for Event Tree: ATWS4N 2:29 PH 6/27/2006 Page I SF Split Fraction Assignment Rule A3S-F + CLPCZ-S + LpcIr-S) * (SPI-S + S~IIS + XTVS)

A4SY At3S-F + (LPCI-S + LPCII-S)* (SPI-S + SPII-S + XTV-S)

Comments RECALL SUCCESS IS DOWNBRANCH!

A4SN I A4DY A3D-F A4UN I UWNVJ.

'iNT1 RBOK*RCOK*PCA-S VNT2 ONWW-S VN'rF 1 CR02 (CST-S + RCW-S)*UB41C-S + OG5-F*EPR6-S*CST-S CR03 (CST-S + RCW-S) UB41C-F*AA-S + OG5-F*EPR6-S*CST-S CRDr 1 OniqS1 014SF PX1-F*PX2-F + ( (RPA-F+HXA-F)* (RPC-F+HXC-F)

+RE-F+NOGA) * ((RPB-F+NXC-F) * (RPD-F+HXC-F) +RF-F+ NOGC) +ODWS-F + CILFAIL Comnents DWS fails due to NPSH if there is a containment breach (CIL-F) 0W145 PX1-S*PX2-S * (RPA-S*HXA-S +

RPC-S*HXC-S) *REPWR-S*GA-S* (RPB-S*lXB-S+RPD-S*HXD-S) *RFPWR-S*GC-S*ODWS-S (PX1-S*(RPA-S*HXA-S + RPC-S*HXC=S)*REPWR-S*GA-S +

PX2-S* (RPB-S*HXB-S+RPD-S*HXD-S) *RFPWR-S*GC-S) *ODWS-S DWSF 1 E-85 C1320503-6924R2 - 7/10/2006

BFN EPUCOPProbabilisticRisk 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 are 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)

Conditions Preventing ECCS NPSH for LLOCA Cases (NPSH)

F.1.1 CIL Fault Tree Revisions The BFN Unit I PRA existing CIL (Containment Isolation Failure) fault tree was modified to add 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-I) 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 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment 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-I.

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 for LLOCA scenarios.

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" (CPSP87RW68)

F-2 C1 320503-6924R2 - 7/1012006

BFN EPUCOPProbabilisticRisk Assessment

"Exceedance Prob for River Water >85F" (RIVER85)

" "Given RW>85F, Cond Prob SP Water >86F" (CPSP86RW85)

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 Cl1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Table F-1 CIL FAULT TREE RESULTS FOR EACH QUANTIFICATION CASE CONDPRE Basic Event CIL Quantification Probability Leak Size Split Fraction Probability(i)

Case Base 1.88E-03 20La 2.25E-03 1 2.47E-04 100La 6.22E-04 2 Same as Base Same as Base 2.25E-03 3 5.217E-03(2) Same as Base 5.59E-03 4 Same as Base Same as Base 2.25E-03 5 5.217E-03(2) Same as Base 5.59E-03 Notes to Table F-1:

(1) "All 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 20La, 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 C1320503-6924R2 - 7/10/2006

BFN EPUCOP ProbabilisticRisk Assessment Table F-2 NPSH FAULT TREE RESULTS FOR EACH QUANTIFICATION CASE Basic Event Probabilities NPSH Split Quantification Fraction Case ZHECCL RIVER68 CPSP87RW68 SPLVL123K RIVER85 CPSP86RW85 Probability Base 500E 5.64E-01 4.42E-01 1.45E-02 1.64E-01 1.00 2.38E-03 03 1 Same Baseas Same Baseas Same as Base Same Baseas Same Baseas Same as Base 2.38E-03 Same as Same as Same as 2 Base Base Bas Base Same as Base 1.00 Bas Base Same as Base 1.64E-01 3 Same as Same as Same as Base Sameas Sameas Same as Base 2.38E-03 Base Base Base Base 4 n/a(') n/a(1 ) n/aM1) n/a(1) n/a(1) nia(1) 1.OE+00 5 n/aM1 n/a(l) n/a(1) n/a(1 ) n/a(1) n/a(1) 1.OE+00 Notes to Table F-2:

(1) In these sensitivity cases the NPSH split fraction is simply set to 1.0.

F-5 C1320503-6924R2 - 7/10/2006

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(3)

v t e Letter Sequence Other
TAC:MC3743, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (Approved, Closed) TAC:MC3744, (Approved, Closed) TAC:MC3812, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting Questions 19 September 2006 29 December 2004 30 December 2004 6 January 2005 3 October 2005 3 October 2005 22 December 2005 22 December 2005 19 January 2006 7 March 2006 25 May 2006 2 June 2006 26 June 2006 26 June 2006 19 July 2006 19 July 2006 8 August 2006 10 August 2006 7 August 2006 10 August 2006 1 September 2006 6 September 2006 1 September 2006 15 September 2006 21 September 2006 27 September 2006 20 December 2006 17 November 2006 17 November 2006 26 January 2007 Answers 28 February 2006 28 February 2006 23 March 2006 23 March 2006 31 March 2006 19 October 2006 Results Approval, Approval, Approval Other: ML041980222, ML042370100, ML050590036, ML051150191, ML051160036, ML051320148, ML051430036, ML051570381, ML051660602, ML052290397, ML060530185, ML060670304, ML060680586, ML060680590, ML060680596, ML060680599, ML060720253, ML060720272, ML060720273, ML060720281, ML060720293, ML060720295, ML060720298, ML060720303, ML060720354, ML060880464, ML060880469, ML061040217, ML061660002, ML061770163, ML061840027, ML061910705, ML062050056, ML062090482, ML062090555, ML062120411, ML062130163, ML062130343, ML062200277, ML062210102, ML062220647, ML062270240, ML062360160, ML062400472, ML062510022, ML062510371, ML062510374, ML062510382, ML062510385, ML062620064... further results
MONTHYEAR2005-02-28 [Table View]

ML062090482

Text

1.0 INTRODUCTION

5.0 CONCLUSION

SUMMARY

1.1 BACKGROUND

SUMMARY

SUMMARY

0 K 80. 0

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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