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{{#Wiki_filter:ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)UNITS 2 AND 3 EPU CONTAINMENT OVERPRESSURE (COP) CREDIT RISK ASSESSMENT See attached. | {{#Wiki_filter:ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) | ||
I BFN EPU Containment Overpressure (COP)Credit Risk Assessment Rev. I I Performed for: Tennessee Valley Authority Performed by: ERIN Engineering and Research, Inc.March 21, 2006 | UNITS 2 AND 3 EPU CONTAINMENT OVERPRESSURE (COP) CREDIT RISK ASSESSMENT See attached. | ||
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I BFN EPU Containment Overpressure (COP) | |||
.... ........... | Credit Risk Assessment Rev. I I | ||
. | Performed for: | ||
.............................................. | Tennessee Valley Authority Performed by: | ||
..... ....... ..... .... .... .. ..... .... .... ........ .... .......... | ERIN Engineering and Research, Inc. | ||
.............................................................................................................................................................. .... .... .... .... .... ......... | March 21, 2006 | ||
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C1320503-6924Rl -3/2212006 I | |||
BEN EPUCOPProbabilisticRisk.Assessment Table Of Contents Section Pane EXECUTIVE | |||
==SUMMARY== | ==SUMMARY== | ||
.. ii | |||
==1.0 INTRODUCTION== | ==1.0 INTRODUCTION== | ||
.1-1 1.1 Background. 1-1 1.2 Scope. 14 1.3 Definitions. 1-4 1.4 Acronyms. 1-6 2.0 APPROACH .. 2-1 2.1 General Approach .2-1 2.2 Steps to Analysis .2-3 3.0 ANAYSIS .. 3-1 3.1 Assessment of DBA Calculations .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 1 PRA Models .3-7 3.5 Assessment of Large-Late Releases .3-9 4.0 RESULTS .4-1 4.1 Quantitative Results .4-1 4.2 Uncertainty Analysis .4-1 4.3 Applicability to BFN Unit 2 and Unit 3.4-13 | |||
. | ==5.0 CONCLUSION== | ||
S ............. 5-1 REFERENCES Appendix A PRA Quality Appendix B Probability of Pre-Existing Containment Leakage Appendix C Assessment of Browns Ferry Data Appendix D Large-Late Release Impact Appendix E Revised Event Trees Appendix F Revised Fault Trees i C1320503-6924R1 -3/2Z2006 I | |||
BFNEPUCOPProbabilisticRiskAssessment EXECUTIVE | |||
==SUMMARY== | ==SUMMARY== | ||
==1.1 BACKGROUND== | The report documents the risk impact of utilizing containment accident pressure (containment overpressure) to satisfy the net positive suction head (NPSiH) requirements for RHR and Core Spray pumps during DBA LOCAs. | ||
The risk assessment evaluation uses the current BFN Unit 1 Probabilistic Risk Assessment (PRA) internal events model (including internal flooding). The BFN PRA provides the necessary and sufficient scope and level of detail to allow the calculation of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) changes due to the crediting of containment overpressure in determining sufficient NPSH requirements for the RHR system and Core Spray system emergency core cooling purmps. | |||
The steps taken to perform this risk assessment evaluation are as follows: | |||
: 1) Evaluate sensitivities to the DBA LOCA accident calculations to determine under what conditions credit for COP is required to satisfy low pressure ECCS pump NPSH. | |||
: 2) Revise all large LOCA accident sequence event trees to make low pressure ECCS pumps dependent upon containment isolation when other plant pre-conditions exist (i.e., SW high temperature, SP initial high temperature, SP low water level). | |||
: 3) Modify the existing BFN PRA Containment Isolation System fault tree to include the probability of pre-existing containment leakage. | |||
: 4) Quantify the modified PRA models and determine the following risk metrics: | |||
* Change in Core Damage Frequency (CDF) | |||
* Change in Large Early Release Frequency (LERF) | |||
: 5) Perform modeling sensitivity studies and a parametric uncertainty analysis to assess the variability of the results. | |||
ii C1320503-6924R1 - 312212006 I | |||
BFN EPU COPProbabilisticRisk Assessm ent 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 (1.4E-09/yr). | |||
: 2) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of Large Early Release Frequency (LERF) below 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 (1.4E-09/yr). | |||
... C1320503-6924R1 - 3/222006 | |||
BFNEPUCOPProbabilisticRiskAssessmwnt Section 1 INTRODUCTION The report documents the risk impact of utilizing containment accident pressure (containment overpressure) to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps during DBA LOCAs. | |||
Revision 1 of this report incorporates the following changes: | |||
: 1) Reduction in the peak SP temperature requiring containment overpressure credit | |||
: 2) Probabilistic credit in the risk analysis of the likelihood of a low SP water volume at the start of the postulated DBA LLOCA accident. | |||
These two changes counteract (i.e., change I increases the calculated ACDF, and change 2 reduces the calculated ACDF), resulting in risk results slightly lower than the Rev. 0 analysis. Both the Rev. 0 and Rev. 1 results represent "very small" changes in risk per RG 1.174. | |||
==1.1 BACKGROUND== | |||
Tennessee Valley Authority (TVA) submitted the BFN extended power uprate (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-001, Revision 0 states that the licensee needs to address the risk impacts of the extended power uprate on functional and system-level success criteria. The staff observes that crediting containment accident 1 1 C1320503-6924R1 - 3/22'2006 I | |||
BENEPUCOPProbabilisticRiskAssessrn nt pressure affects the PRA success criteria; therefore, the PRA should contain accident sequences involving ECCS pump cavitation due to inadequate containment pressure. Section 1.1 of Regulatory Guide (RG) 1.174 states that licensee-initiated licensing basis change requests that go beyond current staff positions may be evaluated by the staff using traditional engineering analyses as 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 traditional engineering analyses, while determining the regulatory acceptability of crediting containment accident pressure. | |||
Considering the above discussion, please provide an assessment of the credit for containment accident pressure against the five key principles of risk-informed decisionmaking stated in RG 1.174 and SRP Chapter 19. | |||
Specifically, demonstrate that the proposed containment accident pressure credit meets current regulations, is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in an increase in core-damage frequency and risk that is small and consistent with the intent of the Commission's Safety,Goal Policy Statement, and will be monitored using performance measurement strategies. With respect to the fourth key principle (small increase in Brisk);} provide a quantitative risk assessment that demonstrates Ithat the proposed containment accident pressure credit meets the numerical risk acceptance guidelines in Section 2.2.4 of RG 1.174. This quantitative risk assessment must include specific containment failure mechanisms!(e.g., liner failures, penetration failures, primary containment isolation system failures) that cause a loss of containment pressure and subsequent loss of NPSH to the ECCS pumps." | |||
Typical of other industry EPU LAR submittals, the BFN EPU LAR includes a request to credit containment accident pressure, also known asicontainment overpressure (CoP), | |||
in the determination of net positive suction head (NPSH) for low pressure EC;CS 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 BFN EPU LAR requests containment overpressure credit for BFN Unit 1 for DBA LLOCA accidents. | |||
1-2 1-2 C1320503-6924R1 - 3/2V/2006 1 | |||
BFNEPUCOP ProbabilisticRiskAssessment 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 accidernts. | |||
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: | |||
"For certain operating BWRs for which the design cannot be practicably altered conformance with Regulatory Position 2. 1.1.1 may not be possible. | |||
In these cases, no additional containment pressure should be included in the determination of available NPSH than is necessary to preclude pump cavitation. Calculation of available containment pressure should underestimate the expected containment pressure when determining available NPSH for this situation. Calculation of suppression pool water temperature should overestimate the expected temperature when determining available NPSH." | |||
The proposed change in the BFN license basis regarding credit for COP meets the approved positions of RG 1.82. However, developments between the NRC staff and members of the Advisory Committee on Reactor Safeguards (ACRS) in 2005 regarding proposed language to Revision 4 of RG 1.82 prompted the NRC to request performance of a 'risk-informed' assessment in accordance with NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis". | |||
1-3 C1320503-6924R1 - 3/22/2006 | |||
BFNEPUCOPProbabilisticRiskAssessment 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 1 at-power internal events PRA resulting from the COP credit requirement for low pressure ECCS pumps during large LOCA scenarios. | |||
External event and shutdown accident risk is assessed on a qualitative basis. | |||
In addition, a review of the BFN Unit 2 and Unit 3 models is performed to show that the results from the Unit 1 BFN PRA apply to Units 2 and 3, as well. | |||
1.3 DEFINITIONS Accident sequence - a representation in terms of an initiating event followed by a combination of system, function and operator failures or successes, of an accident that can lead to undesired consequences, with a specified end state (e.g., core damage or large early release). An accident sequence may contain many unique variations of events that are similar. | |||
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. | |||
Cori 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. | |||
1-4 C1320503-6924R1 -3/22/.006 l | |||
BFNEPUCOP ProbabilisticRiskAssessment 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 RKS, where isolation between the breached system and the RCS fails. An ISLOCA is usually characterized by the over-pressurization of a low-pressure system when subjected to RCS pressure and can result in containment bypass. | |||
Large early release - the rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of off-site emergency response and protective actions. | |||
Large early release frequency - expected number of large early releases per unit of time. | |||
Level I - identification and quantification of the sequences of events leading to the onset of core damage. | |||
Level 2 - evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment. | |||
Plant damage state - Plant damage states are collections of accident sequence end states according to plant conditions at the onset of severe core damage. The plant conditions considered are those that determine the capability of the containment to cope with a severe core damage accident. The plant damage states represent the interface between the Level 1 and Level 2 analyses. | |||
Probability- is a numerical measure of a state of knowledge, a degree of belief, or a state of confidence about the outcome of an event. | |||
Probabilisticrisk assessment - a qualitative and quantitative assessment of the iisk 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 iisk assessment, PRA). | |||
1-5 C1320503-924R1 -3/22/:2006 | |||
BFNEPUCOPProbabilisticRisk Assessment 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 tof 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 prod jct release into the reactor vessel and containment, with potential release to the environment. | |||
Split Fraction - a unitless parameter (i.e., probability) used in quantifying an event tree. | |||
It represents the fraction of the time that each possible outcome, or branch, of a particular top event may be expected to occur. Split fractions are, in general, conditional on precursor events. At any branch point, the sum of all the split fractions representing possible outcomes should be unity. (Popular usage equates "split fraction" with the failure probability at any branch [a node] in the event tree.) | |||
1.4 ACRONYMS ACRS Advisory Committee on Reactor Safeguards ATWS Anticipated Transient without Scram BFN Browns Ferry Nuclear plant CCF Common Cause Failure CDF Core Damage Frequency CET Containment Event Tree COP Containment Overpressure CPPU Constant Pressure Power Uprate 1-6 C1320503-6924R1 -3/22/2006 | |||
BFNEPUCOPProbabilisticRiskAssessment DBA Design Basis Accident DW Drywell ECCS Emergency Core Cooling Systems EPU Extended Power Uprate GE General Electric HEP Human Error Probability HPCI High Pressure Core Injection system HRA Human Reliability Analysis IPE Individual Plant Examination IPEEE Individual Plant Examination for External Events ISLOCA Interface System Loss of Coolant Accident La Maximum Allowable Primary Containment Leakage Rate LERF Large Early Release Frequency LOCA Loss of Coolant Accident LLOCA Large LOCA LOOP Loss of Offsite Power event LPCI Low Pressure Coolant Injection MAAP Modular Accident Analysis Program NPSH Net Positive Suction Head NRC United States Nuclear Regulatory Commission PRA Probabilistic Risk Assessment PSA Probabilistic Safety Assessment RCIC Reactor Core Isolation Cooling System 1-7 C1320503-6924R1 - 3221:2006 I | |||
BFNEPUCOPProbabilisticRiskAssessment RG Regulatory Guide RHR Residual Heat Removal System RPV Reactor Pressure Vessel SMA Seismic Margins Assessment SP Suppression Pool SPC Suppression Pool Cooling SW Service Water TS Technical Specifications TVA Tennessee Valley Authority WW Wetwell 1-8 C1320503-6924R1 - 3/221006 | |||
BFNEPUCOP ProbabilisticRiskAssessment Section 2 APPROACH This section includes a brief discussion of the analysis approach and the types of inpits 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). | |||
2-1 C1320503-6924R1 -3/22/2006 | |||
BEN 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 RISKIMAN 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 spolit 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 end outside containment, support system failure initiators), Imodeled 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. | |||
2-2 C1320503-6924R1 - 3/22/2006 I | |||
BFNEPUCOPProbabilisticRiskAssessment 2.2 STEPS TO ANALYSIS The performance of this risk assessment is best described by the following major analytical steps: | |||
* Assessment of DBA calculations | |||
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 RISKIMAN 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 spolit 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 end outside containment, support system failure initiators), Imodeled 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.2-2 C1320503-6924R1 | |||
-3/22/2006 I | |||
* Estimation of pre-existing containment failure probability | * 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 | * Analysis of relevant plant experience data | ||
* 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 DBA Calculations The purpose of this task is to develop an understanding of the BFN EPU design basis LLOCA calculations that result in the need to credit 3 psig containment overpressure credit.The need for COP credit requests is driven by the conservative nature of design basis accident calculations. | * Manipulation and quantification of BFN Unit 1 RISKMAN PRA models | ||
The DBA LOCA calculations are reviewed and sensitivity calculations performed to determine under what conditions of more realistic inputs is there no need for COP credit in the determination of low pressure ECCS pump NPSH.2-3 C1320503-6924R1 | * Comparison to ACDF and ALERF RG 1.174 acceptance guidelines | ||
-3/22/2D06 l | * Performance of uncertainty and sensitivity analyses | ||
* Assessment of "Large Late" Release Impact | |||
This is the same'approach used in the recent Vermont Yankee EPU COP analyses presented to the ACRS in December 2005).The pre-existing unisolable containment leak probability is combined with the BFN PRA containment isolation failure on demand fault tree (CIL) to develop the likelihood of an unisolated primary containment at t=0 that can defeat the COP credit necessary for the determination of adequate low pressure ECCS pump NPSH.2.2.3 Analysis of Relevant Plant Experience Data An unisolated primary containment is not the only determining factor in defeating low pressure ECCS pump NPSH. The DBA calculations Ishow 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 | * Review of BFN Unit 2 and Unit 3 PRAs Each of these steps is discussed briefly below. | ||
2.2.1 Assessment of DBA Calculations The purpose of this task is to develop an understanding of the BFN EPU design basis LLOCA calculations that result in the need to credit 3 psig containment overpressure credit. | |||
The need for COP credit requests is driven by the conservative nature of design basis accident calculations. The DBA LOCA calculations are reviewed and sensitivity calculations performed to determine under what conditions of more realistic inputs is there no need for COP credit in the determination of low pressure ECCS pump NPSH. | |||
2-3 C1320503-6924R1 -3/22/2D06 l | |||
BFNEPUCOPProbabilisticRiskAssessment 2.2.2 Estimation of Pre-Existing 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 (ILFRT) 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 calculations Ishow 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 | * High initial torus water temperature | ||
* Low initial torus water level 2-4 C1320503-6924R1 | * Low initial torus water level 2-4 C1320503-6924R1 - 3/222006 | ||
-3/222006 | |||
2.2.4 Manipulation And Quantification of BFN Unit 1 RISKMAN PRA Models This task is to make the necessary modifications to the BFN Unit 1 RISKMAN-based PRA models to simulate the loss of low pressure-ECCS pumps during PRA Large LOCA scenarios due to inadequate NPSH caused by an unisolated containment fand other extreme plant conditions (e.g., high service water temperature). | BFNEPUCOP ProbabilisticRiskAssessment This step involves obtaining plant experience data for river water temperature and torus water temperature and level and performing statistical analysis to determine the probabilities of exceedance. | ||
2.2.4 Manipulation And Quantification of BFN Unit 1 RISKMAN PRA Models This task is to make the necessary modifications to the BFN Unit 1 RISKMAN-based PRA models to simulate the loss of low pressure- ECCS pumps during PRA Large LOCA scenarios due to inadequate NPSH caused by an unisolated containment fand other extreme plant conditions (e.g., high service water temperature). | |||
All large LOCA initiated sequences in the BFN PRA are modified as appropriate (except ISLOCAs and LOCAs outside containment, because these LOCAs result in deposition of decay heat directly outside the containment arid 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 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 arid not into the suppression pool). This approach to manipulating only LLOCA scenarios is to mirror the DBA accident calculations requiring COP credit. This is consistent with the ACRS observations during the December 2005 Vermont Yankee EPU COP hearings, in which the ACRS commented that they did not prefer the approach of assigning COP credit to all accident sequence types in the PRA simply for the sake of conservatism. | ||
The modeling and quantification is performed consistent with common RISKMAN modeling techniques. | The modeling and quantification is performed consistent with common RISKMAN modeling techniques. | ||
2.2.5 Comparison to ACDF and ALERF RG 1:174 Acceptance Guidelines The revised BFN Unit 1 PRA models are quantified to determine CDF and LERF. The difference in CDF and LERF between the revised model of this assessment and 'the BFN Unit 1 PRA base results are then compared to 'the RG 1.174 risk acceptance guidelines. | 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 1.1174 ACDF and ALERF risk acceptance guidelines are summarized in Figures 2-1 and 2-2, respectively. The boundaries between regions are 2-5 C1320503-6924R1 -3/22/2006 l | ||
The RG 1.1174 ACDF and ALERF risk acceptance guidelines are summarized in Figures 2-1 and 2-2, respectively. | |||
The boundaries between regions are 2-5 C1320503-6924R1 | BFNEPUCOPProbabilisticRiskAssessment not necessarily interpreted by the NRC as definitive lines that determine the acceptance or non-acceptance of proposed license amendment requests; however, increasing dElta risk is associated with increasing regulatory scrutiny and expectations of compensatory actions and other related risk mitigation strategies. | ||
-3/22/2006 l | 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. | |||
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 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. | ||
This task is performed because the ACRS questioned Entergy on this issue during the recent Vermont Yankee EPU ACRS hearings in December 2005.This aspect of the analysis is for additional information, and does not directly correspond to the RG 1.174 risk acceptance guidelines shown in Figures 2-1 and 2-2.2.2.8 Review of BFN Unit 2and 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 3.2-6 C1320503-6924R1 | 2.2.8 Review of BFN Unit 2and 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. | ||
-3/222006 liability | 2-6 C1320503-6924R1 - 3/222006 | ||
.............................. | |||
......... | liability BFNEPUCOPProbabilisticRiskAssessmont Figure 2-1 RG 1.1 74 CDF RISK ACCEPTANCE GUIDELINES | ||
-3/2212006 l | . .:. .,. .. . '. s... . . . . . . . . . . . . . . . ....... ..................................... | ||
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-3/22/2006 BFN | |||
* Initial reactor power level at 102% EPU* Decay heat defined by 2 sigma uncertainty | BFNEPU COPProbabilisticRisk Assessment Figure 2-2 RG 1.174 LERF RISK ACCEPTANCE GUIDELINES | ||
* 2 RHR pumps and 2 RHR heat exchangers in SPC* All pumps operating at full flow* River water temperature at 95 | ,..... .... .......... ..... ........ ... ......... ... ... ...... .. .... ....... ...... .... ................... | ||
Initial reactor power level 3-1 C1320503-6924R1 | ::::::::: | ||
-3/22/2006 l | :::: : : : : : : : : : :: : : :.: ::. : : : . : .: : : . : : : . . .. . . . . . . | ||
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* Decay heat* Number of RHR pumps and heat exchangers in SPC* River water temperature | ..,:. ...:.i............l................:.. i... | ||
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2-8 C1320503-6924R1 -3/22/2006 | |||
BFN EPU COP ProbabilisticRisk Assessm ent Section 3 ANALYSIS This section highlights the major qualitative and quantitative analytic steps to the analysis. | |||
3.1 ASSESSMENT OF DBA CALCULATIONS The purpose of this risk assessment is due to the fact that the conservative nature of design basis accident calculations result in the need to credit COP in determining adequate low pressure ECCS pump NPSH. Use of more realistic inputs in such calculations shows that no credit for COP is required. | |||
The GE DBA LOCA calculation makes the following conservative assumptions, among others, regarding initial plant configuration and operation characteristics: | |||
* Initial reactor power level at 102% EPU | |||
* Decay heat defined by 2 sigma uncertainty | |||
* 2 RHR pumps and 2 RHR heat exchangers in SPC | |||
* All pumps operating at full flow | |||
* River water temperature at 95 0F | |||
* Initial suppression pool temperature at 95tF | |||
* Initial SP water volume at minimum technical specification level | |||
* No credit for containment heat sinks The GE DBA LOCA calculations were reviewed and the following input parameters were identified as those with a potential to significantly impact the DBA analytic conclusions regarding the need for COP credit in NPSH determination: | |||
Initial reactor power level 3-1 C1320503-6924R1 -3/22/2006 l | |||
BFNEPUCOP ProbabilisticRisk Assessment | |||
* Decay heat | |||
* Number of RHR pumps and heat exchangers in SPC | |||
* River water temperature | |||
* Initial suppression pool temperature | * Initial suppression pool temperature | ||
* RHR heat exchanger effectiveness | * RHR heat exchanger effectiveness | ||
* Initial suppression pool water volume* Credit for containment heat sinks Based on knowledge of the calculations, other inputs such as initial containment air temperature and humidity, have non-significant impacts on the results.It is recognized that there are numerous different combinations of more realistic calculation inputs that show that COP credit is not necessary for maintenance of low pressure ECCS pump NPSH. To simplify the risk assessment, the different combinations of realistic input sensitivities were maintained at a manageable number.A number of sensitivity calculations were performed to identify key input parameters for use in this risk assessment. | * Initial suppression pool water volume | ||
The results of these calculations are shown in Table 3-1 (the shaded cells show those parameters that changed from the base DBA LO'A calculation). | * 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. | ||
[3]From the results of the sensitivity cases summarized in Table 3-1, the following general conclusions can be made:.Initial reactor power, decay heat level, initial water temperatures, suppression pool volume, and the number of RHR pumps/HXs in operation are the key determining factors in the analytic conclusions. | It is recognized that there are numerous different combinations of more realistic calculation inputs that show that COP credit is not necessary for maintenance of low pressure ECCS pump NPSH. To simplify the risk assessment, the different combinations of realistic input sensitivities were maintained at a manageable number. | ||
A number of sensitivity calculations were performed to identify key input parameters for use in this risk assessment. The results of these calculations are shown in Table 3-1 (the shaded cells show those parameters that changed from the base DBA LO'A calculation). [3] | |||
From the results of the sensitivity cases summarized in Table 3-1, the following general conclusions can be made: | |||
. Initial reactor power, decay heat level, initial water temperatures, suppression pool volume, and the number of RHR pumps/HXs in operation are the key determining factors in the analytic conclusions. | |||
These factors are evaluated in this risk assessment. | These factors are evaluated in this risk assessment. | ||
* RHR heat exchanger effectiveness land credit for containment heat sinks also influence the results, but to manage the risk calculation, this assessment takes no probabilistic credit for these issues.3-2 C1320503-6924R1 | * RHR heat exchanger effectiveness land credit for containment heat sinks also influence the results, but to manage the risk calculation, this assessment takes no probabilistic credit for these issues. | ||
-3/22/2006 l | 3-2 C1320503-6924R1 -3/22/2006 l | ||
* 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 371).* 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 Iconservative water temperature and SP volume assumptions (refer to Case ld 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 | BFNEPUCOPProbabilisticRiskAssessment | ||
* 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 | * 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 371). | ||
'of 68 | * 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 Iconservative water temperature and SP volume assumptions (refer to Case ld in Table 3-1). | ||
The probability of being in Plant State I or Plant State 2 is discussed below in Section 3.2.3-3 C1320503-6924R1 | * 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 F, then torus water temperature must be above 87tF to result in the need for COP credit (refer to Case 2f in Table 3-1). | ||
-3/22/2006 l | * 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 850F,, then torus water initial temperature must be above 860 F to result in the need for COP credit (refer to Case 4i in Table 3-1). | ||
The analytic conclusions are used in this risk assessment to define two plant states that will result in failure of low pressure ECCS pumps on inadequate NPSH during large LOCAs if the containment is unisolated: | |||
* 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). | * 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 870F. l | ||
* The probability of river water temperature greater than | * Plant State 2: 100% EPU initial power level, nominal decay heat, 2 RHR pumps and heat exchangers in SPC, iriitial SP volume at 123,500 ft3, river water temperature of 86 0F,; and river water initial temperature above 85 0F. | ||
-3/22/I.M6 l | These two plant states are used in this risk assessment to modelithe LLOCA scenarios that can result in loss of low pressure ECCS pumps',due'to inadequateiNPSH when the containment is unisolated. The probability of being in Plant State I or Plant State 2 is discussed below in Section 3.2. | ||
BFN | 3-3 C1320503-6924R1 -3/22/2006 l | ||
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 | BFNEPUCOPProbabilisticRiskAssessment Scenarios with 3 or 4 RHR pumps and heat exchangers are not explicitly incorporated into the base case quantification because the risk contribution from such scenarios is non-significant (refer to Section 4.2.2). | ||
-3/22/2006 1 | 3.2 PROBABILITY OF PLANT STATE I AND PLANT STATE 2 This section discusses the estimation of the probability of being in Plant State 1 or Plant State 2. This assessment is based on the statistical analysis of BFN experience data. | ||
Refer to Appendix C for the statistical analysis of variations in BFN river water temperature and torus water temperature and level. | |||
The BFN primary containment performance experience shows BFN containment leakages much less than 20La. Per 346 C1320503-6924R1 | 3.2.1 Probability of Plant State 1 The probability of being in Plant State I is determined as follows: | ||
-3/22/2006 | * 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 | ||
[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. | * 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 | ||
3.4 MODIFICATIONS TO BFN UNIT 1 PRA MODELS As discussed in Section 2, all large LOCA initiated sequences in the BFN PRA are modified as appropriate (except ISLOCAs and LOCAs outside containment, because these LOCAs result in deposition of decay heat directly outside the containment and not into the suppression pool). The following Large LOCA initiated sequences in the BFN Unit 1 PRA were modified:* Large LOCA -Loop I Core Spray Line Break (LLCA)* Large LOCA -Loop If Core Spray Line Break (LLCB)* Large LOCA -Loop A Recirc. Discharge Line Break (LLDA)* Large LOCA -Loop B Reciro. Discharge Line Break (LLDB)* Large LOCA -Loop A Recirc. Suction Line Break (LLSA)* Large LOCA -Loop B Recirc. Suction Line Break (LLSB)* Other Large LOCA (LLO)3-7 3-7 C1320503-6924RI | * 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). | ||
-3/22Iroo6I BEN | * The probability of river water temperature greater than 680F is determined from the BFN experience data statistical analysis summarized in Appendix C. | ||
.A top event modeling the additional Plant State pre-conditions (NPSH)was added to the beginning of the Level 1 event tree structures, right after the CIL top event..If top events CIL and NPSH are satisfied (i.e., occur), then the RHR pumps and CS pumps are directly failed Refer to Appendix E for print-outs of the revised large LOCA event trees.The CIL top event is quantified using a fault tree. The fault tree is a modified version of the existing BFN Unit I Level 2 PRA containment isolation fault tree. The BFN Unit 1 Level 2 PRA containment isolation fault tree models failure of the containment isolation system on demand given an accident signal. Hardware, power and signal failures for all primary containment penetrations greater than 3" diameter are modeled in the fault tree.To this fault tree structure was added the probability of a pre-existing containment leak size of 2OLa. Refer to Appendix F for a print-out of the containment isolation fault tree used in this analysis for the CIL node in the large LOCA event trees.The NPSH top event is also quantified using a fault tree. The 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.The quantification of the revised model was performed to produce the new CDF. All the new CDF scenarios are those in which the containment is unisolated at t=O, all R1'V injection is lost early, and core damage occurs at approximately one hour. As such, the additional CDF contributions created by this model manipulation are also all LERF 3-8 C1320503-6924R1 | * Given river water temperature 680F, the conditional probability that the torus water temperature is 870 F is determined from the BFN experience data statistical analysis summarized in Appendix C. | ||
-3/2212D06 l | 34 C1320503-6924R1 -3/22/I.M6 l | ||
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 | BFN EPUCOPProbabilisticRisk Assessment The probability that suppression pool water level is less than 123,500 ft3 is also based on the BFN experience data statistical analysis summarized in Appendix C. | ||
[2] determines that a containment leak representative of LERF is >600La.The quantification results and uncertainty and sensitivity analyses are discussed in Section 4.The revised BFN Unit 1 PRA RISKMAN model for this base case analysis is archived in file UICOP-H and saved on the BFN computers along with the other BFN PRA RISKMAN models.3.5 ASSESSMENT OF LARGE-LATE RELEASES As discussed above in Section 3.3, all the deltaCDF resulting from this risk assessment also results directly in LERF. As such, there is no increase in Large-Late releases due to scenarios modeling in this risk assessment. | 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. | ||
Refer to Appendix D for more discussion. | As can be seen from Table C-1, the probability of river water temperature being greater than 680 F at the time of the DBA LOCA is 5.64E-1. As discussed in Section C.2.1, the conditional probability that suppression pool temperature is greater than 87 0F is 4.42E- | ||
3-9 C1320503-6924R1 | : 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. | ||
-312212006 l | Therefore, the probability of being in Plant State 1 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. | ||
BFN | 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 850 F is determined from the BFN experience data statistical analysis summarized in Appendix C. | |||
* Given river water temperature of 850 F, the conditional probability that the torus water temperature is 87 0F, is taken to be 1.0. This is reasonable (refer to Figure C-1). | |||
3-5 3-5 C1320503.6924Rl - 3/22/2006 1 | |||
BFNEPUCOPProbabilisticRiskAssessment The probability that suppression pool water level is less than 123,500 ft3 is also based on the BFN experience data statistical analysis summarized in Appendix C. | |||
As can be seen from Table C-1, the probability of river water temperature being greater than 850 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 i:he 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., EFPRI Report 1009325, Risk Impact of Extended Intearated Leak Rate Testina Intervals, 12/03). [2] | |||
This assessment is provided in Appendix B of this report. As discussed in Appendix B, a pre-existing unisolable containment leakage path of 20La is assumed in the base case quantification of this risk assessment to result in defeating the necessary C-DP credit. As can be seen from Table B-1, the probability of a 20La pre-existing containment leakage at any given time at power is 1.88E-03. | |||
This low likelihood of a significant pre-existing containment leakage path is consistent with BFN primary containment performance experience. The BFN primary containment performance experience shows BFN containment leakages much less than 20La. Per 346 C1320503-6924R1 -3/22/2006 | |||
BFNEPUCOPProbabilisticRiskAssessment 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. | |||
3.4 MODIFICATIONS TO BFN UNIT 1 PRA MODELS As discussed in Section 2, all large LOCA initiated sequences in the BFN PRA are modified as appropriate (except ISLOCAs and LOCAs outside containment, because these LOCAs result in deposition of decay heat directly outside the containment and not into the suppression pool). The following Large LOCA initiated sequences in the BFN Unit 1 PRA were modified: | |||
* Large LOCA - Loop I Core Spray Line Break (LLCA) | |||
* Large LOCA - Loop If Core Spray Line Break (LLCB) | |||
* Large LOCA - Loop A Recirc. Discharge Line Break (LLDA) | |||
* Large LOCA - Loop B Reciro. Discharge Line Break (LLDB) | |||
* Large LOCA - Loop A Recirc. Suction Line Break (LLSA) | |||
* Large LOCA - Loop B Recirc. Suction Line Break (LLSB) | |||
* Other Large LOCA (LLO) 3-7 3-7 C1320503-6924RI -3/22Iroo6I | |||
BEN EPUCOPProbabilisticRiskAssessmlnt The accident sequence modeling for the above LLOCA initiators was modified as follows: | |||
* A top event for loss of containment integrity (CIL) was added to the beginning of the Level 1 event tree structures | |||
. A top event modeling the additional Plant State pre-conditions (NPSH) was added to the beginning of the Level 1 event tree structures, right after the CIL top event. | |||
. If top events CIL and NPSH are satisfied (i.e., occur), then the RHR pumps and CS pumps are directly failed Refer to Appendix E for print-outs of the revised large LOCA event trees. | |||
The CIL top event is quantified using a fault tree. The fault tree is a modified version of the existing BFN Unit I Level 2 PRA containment isolation fault tree. The BFN Unit 1 Level 2 PRA containment isolation fault tree models failure of the containment isolation system on demand given an accident signal. Hardware, power and signal failures for all primary containment penetrations greater than 3" diameter are modeled in the fault tree. | |||
To this fault tree structure was added the probability of a pre-existing containment leak size of 2OLa. Refer to Appendix F for a print-out of the containment isolation fault tree used in this analysis for the CIL node in the large LOCA event trees. | |||
The NPSH top event is also quantified using a fault tree. The 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. | |||
The quantification of the revised model was performed to produce the new CDF. All the new CDF scenarios are those in which the containment is unisolated at t=O, all R1'V injection is lost early, and core damage occurs at approximately one hour. As such, the additional CDF contributions created by this model manipulation are also all LERF 3-8 C1320503-6924R1 - 3/2212D06 l | |||
BFNEPUCOPProbabilisticRiskAssessment 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. | |||
The quantification results and uncertainty and sensitivity analyses are discussed in Section 4. | |||
The revised BFN Unit 1 PRA RISKMAN model for this base case analysis is archived in file UICOP-H and saved on the BFN computers along with the other BFN PRA RISKMAN models. | |||
3.5 ASSESSMENT OF LARGE-LATE RELEASES As discussed above in Section 3.3, all the deltaCDF resulting from this risk assessment also results directly in LERF. As such, there is no increase in Large-Late releases due to scenarios modeling in this risk assessment. Refer to Appendix D for more discussion. | |||
3-9 C1320503-6924R1 -312212006 l | |||
BFN EPUCOPProbabilisticRiskAssessment Table 3-1 | |||
==SUMMARY== | ==SUMMARY== | ||
OF COP DETERMINISTIC CALCULATIONS(5) co 0 0 U, 4) cm E-C | OF COP DETERMINISTIC CALCULATIONS(5) co 0 0 U, 4) cm E C - C CL > w 4 orW a a. | ||
-3/22/2006 I | : a. )L a ~ - 0o | ||
- -2 C = | |||
Vh 06 0 a . | |||
0 c#0 Case(') Case pionS CaeDes pton . o COi cn.Ezo 0 fr- z Z.S Ca. W> | |||
* s0 w.a - EL t Base Case(2) EPU Licensing Calculation- 102%- ANSI 5.1 FuDl (GE) 95 95 2 2 2- 4000 223 2 1121,500 Yes No 187.3 Ye DBA LOCA EPU w/2a design Case 1(2) DBA Calculation but No 102% ANSI 5.1 _ Ful 4000 121,500 Yes No 166.4 No (GE) Single Failure EPUI_ w/2a 9 9 design Case UaP2 | |||
) DBA Calculation but 3 RHR 102% ANSI 51 . Fufl (GE) Pumps inSuppression Pool EPU wf2a 95 95 4000 121,500 Yes No 175.0 Ye design Coi _ . | |||
Case la (TVA) DBA Calculation but 3 RHR IThis case is 102% ANSI Full benchmarked Pumps in Suppression Pool 5.1 EPU T -wt2a 95 95 design 4000 121,500 Yes No 175.0 Ye Cooling against Case la (GE)] | |||
Case lb 100% Initial Power, RHRSW (TVA) 89F 3 Pumps in- - Full Suppression Pool Cooring, K 95 design 4000 Yes No 171.0 No Value 225, 4 CS Pumps Case Ic 100%lnitial Power, RHRSW (TVA) 90°F, 3 Pumps in Full Suppression Pool Coobng, K 95 4000 Yes No 170.5 No design Value 225, 4 CS Pumps, Nominal SP WL 3-10 C1320503-6924R1 - 3/22/2006 I | |||
BFNEPUCOP ProbabilisticRiskAssessment Table 3-1 | |||
==SUMMARY== | ==SUMMARY== | ||
OF COP DETERMINISTIC CALCULATIONS(5) | OF COP DETERMINISTIC CALCULATIONS(5) | ||
I p-q | I p-q 5 5 5 5 Y ! ! inyin | ||
= 85oF EPU wi2a against Case 2 (GE) | : 0. 2 0 | ||
= 75°F EPUL wl2a Case 2b(2) DBA Calculation but Initial 102% ANSI 5.1 (GE) SW Temperature | 0 E >,4 C 0 | ||
= 70°F EPU wl2cr Case 2c DBA Calculation but Initial 102% ANSI5.1 (GE) SW Temperature 65°F EPU | - - - % 0. E C,, | ||
(fD S | |||
: 0. s 8 I 0 0, E ia E cm9 UC 0 CLI | |||
= O E hi a ..9C: | |||
Z . | |||
i 0-9 hi 0 Ceo 20 | |||
> 06 0 | |||
0 C' | |||
-o 0 0 G EL g. | |||
M - | |||
tC a) 21i z U W_;m_ | |||
a-41 I- Co o- i-- | |||
C = | |||
. tLco Wo- 0 a..- -E O- | |||
- 10 CL | |||
=-. - | |||
Co) .= | |||
o *0 - M | |||
_ _ _ Z 0- cRrc- _ 0 Case0) Case Description ~ ZO- | |||
_-7 -u~ -i 0i o 9 a CS wwO Z.'. w §_ WLI J a. | |||
Case 1d DBA Calculation but RHRSW (IVA) :90F, SP Initial Temp 910F 3 102% ANSI Pumps inSuppression Pool Full 5.1 w/ 4000 121,5W Yes No 171.0 No Coobng, K Value 225, 4 CS EPU 2a design Pumps. | |||
Case le DBA Calculation but RHRSW (TVA) 92*F, SP Initial Temp 90°F, 3 102% ANSI Pumps inSuppression Pool EP% 5.1 w/ Full design 4000 121,500 Yes No 171.1(41 No Cooling, KValue 225, 4 CS EPU 2cr Pumps Case 2 DBA Calculation but Initial 102%- ANSI 5.1 Full (GE) SW Temperature 85°F 2 2 4000 223 2 121,500 Yes No 182.0 Ye-EPU wl2a design Case 2 (TVA) -- -ANSI | |||
[This case is DBA Calculation but SW 102% 5.1 95 2 Full 2 2 4000 223 benchmarked design 2 121,5W0Yes No 182.2 Ye - | |||
Temperature = 85oF EPU wi2a against Case 2 (GE) _ | |||
Case 2a DBA Calculation but Initial 102% ANSI 5.1 (GE) SW Temperature = 75°F 95 2 desin 2 2 4000 223 2 121,500 Yes No 177.6 Ye _ | |||
EPUL wl2a Case 2b(2) DBA Calculation but Initial 102% ANSI 5.1 (GE) SW Temperature = 70°F 95 2 deFsugn 2 2 4000 223 2 121,500 Yes No 175.9 Ye EPU wl2cr design Case 2c DBA Calculation but Initial 102% ANSI5.1 (GE) SW Temperature 65°F EPU 95 2 Fulln 2 2 4000 . 223 2 121 ,Y0n Yes No 174.3 Yel_ | |||
wl2cr 3-11 C1320503-6924R1- 322/2006 I | |||
BFNEPUCOPProbabilisticRiskAssessment Table 3-1 | |||
==SUMMARY== | ==SUMMARY== | ||
OF COP DETERMINISTIC CALCULATIONS(5) | OF COP DETERMINISTIC CALCULATIONS(5) | ||
I | I * - I 7 T i , , I, CL 0 | ||
* I | )- 't 0 0 E 0 y4C 0 | ||
= 65oF, SP 102% 5.1 Initial Temp 88oF,- Nominal- -EPU w120 SP WL Case 2e DBA Calculation but SW 102% ANSI (TVA) Temperature | .5 a. 0 E 0 (t aL_. 0~ I~0 0 W | ||
= 650F, SP EPU 5.1 Initial Temp 87oF w/2r Case 2f DBA Calculation but SW ANSI (TVA) Temperature | : a. = a CL E E 0 | ||
= 68oF, SP 102% ANS Initial Temp 870F, Nominal EPU w12 | .C U) | ||
= 85°F 2 | -o.° LLi. 01 E a a. | ||
-3/22/2006 I | a0a.,E z O Ea. 0 0 3.t -8 BC 0 (t0 il E C 2 oa 0 Q 00 _i6.2. | ||
BEN | -5 0 t 0 0 | ||
0 I u5 E = | |||
a 0 | |||
ICO 0 C-a) t:S U) E, tCL. | |||
CL | |||
.0 0 | |||
= c3 | |||
.05 E . | |||
I x a50.. | |||
. 8 r1 s <o a-U) ne E 0 Q_ ' I c a:ca C ARL e co G AC 0 Caset') Case Description 0 Z O Z .' 0iEL | |||
- | |||
U00 O | |||
U I L3A Case 2d DBA Calculation but SW ANSI (TVA) Temperature = 65oF, SP 102% 5.1 _.- -design Initial Temp 88oF,- Nominal- - EPU w120 2 Fuln 2 2 4000 223 2 Yes No 170.6 No SP WL Case 2e DBA Calculation but SW 102% ANSI (TVA) design Temperature = 650F, SP EPU 5.1 2 FuI 2 2 4000 223 2 No 170.7 No Initial Temp 87oF 121,500 Yes w/2r Case 2f DBA Calculation but SW ANSI (TVA) Temperature = 68oF, SP 102% ANS Initial Temp 870F, Nominal EPU w12 design Yes No 171.1') No SPw/2a Case 3 DBA Calculation but Initial - Full (GE) SP Temperature = 85°F 2 2 2 4000 223 2 121,500 Yes No 183.8 YeJ design 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 Yej Credit- design Case 4 (TVA) 100%Initial Power, Minimum | |||
[This case is SP Level, and No Heat Sink 2 FujN 2 2 4000 2 177.1 bench-marked 121,500 Yes No Ye, Credit design against Case 4 Case 4a 100%Initial Power, Nominal (GE) SP Level, and Heat Sink 2 Full 2 2 4000 2 Yes 174.7 Yes rrft, design | |||
_ __ _ _ _ _ 1 *~~l _ __ I _ __ I _I I_ | |||
3-12 C1320503-6924R1 - 3/22/2006 I | |||
BEN EPUCOP ProbabilisticRiskAssessment Table 3-1 | |||
==SUMMARY== | ==SUMMARY== | ||
OF COP DETERMINISTIC CALCULATIONS(5) | OF COP DETERMINISTIC CALCULATIONS(5) p - - - . - p- p - p- I - Y - - p- p p- - Y - Y - | ||
Casell) Case Description Case | * U | ||
-322/2006 l | .5 0 e c a.5 0 E 0 0 E | ||
E) t 0 pa 8I C | |||
0 0 a.' | |||
E U-- | |||
-8 0 0SC 0= 04) 0 8 3. | |||
0 0FEa | |||
: i. U)- i. ,m a 0 .n 00 0 | |||
F 45 00 E 0 4) U Ca | |||
= 0a. | |||
a: | |||
= | |||
Ea 0: | |||
0 a a. aL co 0t 0 V | |||
.9 Z | |||
,Io oa co 0. ~.Q.' W m) 0 C2I: X-Z 0 U. | |||
W4 = | |||
CL La.. a) .' 0.W =Z E tr _ | |||
-.8 W @ | |||
CO~ U =) | |||
Casell) Case Description S ID Z OL 00 oCL 1=, YeO 0e _ | |||
Case Ab2) 100% Initial Power, Minimum (GE) SP Level, and Heat Sink 2 Full 2 2 4000 2 121,500 Credit design Yes 178.9 2 | |||
Case 4C( ) 100%Initial Power, Minimum (GE) SP Level, Heat Sink Credit, and SW Temp. that results in 2 Full 2 2 4000 design 2 121,500 Yes 175.8 Ye, Peak SP Temp. equal tolless than 176°F Case 4d 100% Initial Power, RHRSW (TVA) 860F, SP Initial Temp 92oF, K 2 Full 2 2 2 design 4000 2 121,5W Yes No 177.0 Ye Value 225---- | |||
Case 4e 100% Initial Power, RHRSW (TVA) 860F,SP Initial Temp 900F, K 2 Full 2 40 2 121,500 Yes 176.1 design 2 2 40 No Ye! | |||
Value 225----- -- - - - | |||
Case4f 100% Initial Power, RHRSW (IVA) 860F, SP Initial Temp 900F, K Full 2 design 2 2 4000 2 Yes No 175.6 Ye Value 225, Nominal SP WL Case 4g 100% Initial Power, RHRSW (TVA) 860F, SP Initial Temp 900F, K 2 deFsun 2 2 4000 2 Yes No 173.1 Ye Value 241, Nominal SP WL-Case 4h 100% Initial Power, RHRSW (VA) 850F, SP Initial Temp 900F, K 2 design 2 2 4000 2 Yes No 175.1 Ye Value 225, Nominal SP WL 3-13 C1320503-6924R1 - 322/2006 l | |||
BFNEPUCOP ProbabilisticRiskAssessment Table 3-1 | |||
==SUMMARY== | ==SUMMARY== | ||
OF COP DETERMINISTIC CALCULATIONS(5) | OF COP DETERMINISTIC CALCULATIONS(5) | ||
*3 | *3 | ||
*Case 4i | = U a . - . | ||
(2) Case verified by formal analysis.(3) COP credit required for peak suppression pool temperature of 171°F.(4) This value is acceptable for demonstrating sensitivity analysis results.(5) Shaded areas in the table "highlight" differences from the Base Case.3-14 C1320503-6924R1 | * 4, | ||
-3/22/2006 l | .0 4, | ||
-a a 0 C | |||
1.4E-9/yr.deltaLERF: | 0 E W -; E i I aE A2 c 14) | ||
1.4E-9lyr As discussed in Section 3, the additional CDF contributions created by this model manipulation are also all LERF release sequences (i.e., deitaCDF equals deltaLERF). | U, 2 | ||
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: | aL.0 Mi r 4) a. .5 E | ||
=E UM | |||
$S 00 | |||
'a. LL-E | |||
.S t | |||
hi C3 | |||
.B U, E a | |||
: 4) - M S hi M Ehi 0coE.= W _2 E cm .C C v E | |||
-0 00. E. .2 0 | |||
: 4) C | |||
'6 = <DCL | |||
= 4 60) | |||
_ .Q | |||
*t .5 V | |||
3n | |||
~-3 U M I-r54) | |||
M r M a- Lt , (.5 U) | |||
C) .5 U) .C W | |||
_ | |||
cc 4 rk g- *e -co | |||
.3 0 E 0 2t CS.- | |||
Case(1) a: a 4): .L- I 2* = | |||
Case Description .E. - 0- U)- U).S Z o wO °. | |||
0 EL | |||
* Case 4i 100% Initial Power, RHRSW (TVA) 85°F, SP Initial Tenp 86oFj K- - FUi 2 2 -4000 - . 2 - Yes No 170.8 No Value 241; Nominal SP WL -- | |||
Case 4j - 100% Initial Power, RHRSW (TVA) 85°F, SP Initial Temp 88°F, K | |||
._.__ JValue 241, Nominal SP WL 2 desi 2 2 4000 2 Yes No 171.0 No Notes to Table 3-1: | |||
(1) Column information includes designation of organization that performed the calculation. | |||
(2) Case verified by formal analysis. | |||
(3) COP credit required for peak suppression pool temperature of 171°F. | |||
(4) This value is acceptable for demonstrating sensitivity analysis results. | |||
(5) Shaded areas in the table "highlight" differences from the Base Case. | |||
3-14 C1320503-6924R1 - 3/22/2006 l | |||
BFNEPUCOPProbabilisticRiskAssessment Section 4 RESULTS 4.1 QUANTITATIVE RESULTS The results of the base quantification of this risk assessment case are as follows: | |||
. deltaCDF: 1.4E-9/yr | |||
. deltaLERF: 1.4E-9lyr As discussed in Section 3, the additional CDF contributions created by this model manipulation are also all LERF release sequences (i.e., deitaCDF 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 | * Parametric | ||
* Modeling* Completeness 4-1 C1320503-6924R1 | * Modeling | ||
-3/22/2006 I | * Completeness 4-1 C1320503-6924R1 -3/22/2006 I | ||
Typical of standard industry practices, the parametric uncertainty aspect is assessed here by performing a Monte Carlo parametric uncertainty propagation analysis. | BFNEPUCOP ProbabilisticRiskAssessment 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. | ||
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. | 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. | ||
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. | 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. | ||
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.RISIKMAN has three analysis modules: Data Analysis Module, System Analysis Module, and Event Tree Analysis Module. Appropriate probability distributions for each uncertain 4-2 C1320503-924R1 | 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. | ||
-3/22/2006 a I l | RISIKMAN has three analysis modules: Data Analysis Module, System Analysis Module, and Event Tree Analysis Module. Appropriate probability distributions for each uncertain 4-2 C1320503-924R1 -3/22/2006 | ||
Event trees are quantified and linked together in the Event Module. The important sequences from the results of the Event Tree Module are used in another Monte Carlo sampling step to propagate the split fraction uncertainties and obtain the uncertainties in the overall results.The descriptive statistics calculated by RISKMAN for the total core damage frequency of the plant caused by internal events include:* Mean of the sample* Variance of the sample* 5th, 50th, and 95th percentiles of the sample The parametric uncertainty associated with delta core damage frequency calculated in this assessment is presented as a comparison of the RISKMAN calculated CDF uncertainty statistics for the two cases (i.e., the Unit i base EPU PRA and the EIDU COP Credit base case quantification). | |||
The results are shown in Table 4-1. Table 4-1 summarizes the CDF uncertainty distribution statistics for the BFN Unit 1 PRA and for the COP credit base quantification. | a I l BFNEPUCOPProbabilisticRiskAssessment parameter in the analysis is determined and included in the Data Module. The System Module combines the individual failure rates, maintenance, and common cause parameters into the split fraction frequencies that will be used by the Event Tree Module. A Monte Carlo routine is used with the complete distributions to calculate t:he 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. | ||
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. | The descriptive statistics calculated by RISKMAN for the total core damage frequency of the plant caused by internal events include: | ||
Taking these cautions into consideration, a comparison of the distributions by percentiles shows little if any change.4-3 C1320503-6924R1 | * Mean of the sample | ||
-3/2212D06 I | * Variance of the sample | ||
* 5th, 50th, and 95th percentiles of the sample The parametric uncertainty associated with delta core damage frequency calculated in this assessment is presented as a comparison of the RISKMAN calculated CDF uncertainty statistics for the two cases (i.e., the Unit i base EPU PRA and the EIDU COP Credit base case quantification). The results are shown in Table 4-1. Table 4-1 summarizes the CDF uncertainty distribution statistics for the BFN Unit 1 PRA and for the COP credit base quantification. | |||
The PRA industry is currently investigating methods for performing modeling uncertainty analysis. | 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. | ||
EPRI has developed a guideline for modeling uncertainty that is still in draft form and undergoing pilot testing. The EF'RI approach that is currently being tested takes the rational approach of identifying key sources of modeling uncertainty and then performing appropriate sensitivity calculations. | 4-3 C1320503-6924R1 -3/2212D06 I | ||
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 invo ve 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 | 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. | ||
* 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 2OLa that would result in defeat of the necessary containment overpressure credit during a DBA LOCA.4-4 C1320503-6924R1 | 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 EF'RI 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. | ||
-3/2212006 l | The modeling issues selected here for assessment are those related to the risk assessment of the containment overpressure credit. This assessment does not invo ve 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 | |||
This modeling sensitivity case expands the scope of the containment isolation fault tree to include smaller lines as potential defeats of COP credit. This sensitivity is performed by increasing by a factor of 10 the failure probability associated with the containment isolation system.Assessment of Power and Water Temperature and Level Pre-conditions This is a conservative sensitivity that assumes that all that is necessary for failure of the low pressure ECCS pumps due to inadequate NPSH during a large LOCA is an unisolated containment. | * Calculation of containment isolation system failure | ||
This sensitivity is performed by assuming the other pre-conditions represented by the top event NSPH exist with a probability of 1.0.Number of RHR pumps and heat exchangers in SPC The base case COP credit quantification addresses the situation in which 2 or less RHR pumps and heat exchangers are operating in SPC mode. The likelihood of failing any two RHR pumps during the i24-hr PRA mission time is approximately 8.2E-3. The likelihood of an unisolated containment given an accident initiator is approximately 2.2E-3, and the likelihood of other necessary extreme plant conditions (e.g., high river temperature, high reactor power, reduced suppression pool water level) existing at the 4-5 C1320503-6924R1 | * Assessment of power and water temperature and level pre-conditions | ||
-3/22/2006 I | * 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 2OLa that would result in defeat of the necessary containment overpressure credit during a DBA LOCA. | ||
4-4 C1320503-6924R1 - 3/2212006 l | |||
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 1d in Table 3-1):* Sum of BFN PRA Large LOCA initiator frequencies: | BFNEPUCOPProbabilisticRiskAssessment A larger pre-existing leak size of 10OLa, consistent with the EPRI 1009225 recommended assumption for a "large" leak, is used in this sensitivity to defeat I:he necessary COP credit. From EPRI 1009325, the probability of a pre-existing I00La containment leakage pathway at any given time at power is 2.47E-04. | ||
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: | 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. | ||
3E-3 (nominal from base analysis)* Probability of river water temperature | Assessment of Power and Water Temperature and Level Pre-conditions This is a conservative sensitivity that assumes that all that is necessary for failure of the low pressure ECCS pumps due to inadequate NPSH during a large LOCA is an unisolated containment. This sensitivity is performed by assuming the other pre-conditions represented by the top event NSPH exist with a probability of 1.0. | ||
> | Number of RHR pumps and heat exchangers in SPC The base case COP credit quantification addresses the situation in which 2 or less RHR pumps and heat exchangers are operating in SPC mode. The likelihood of failing any two RHR pumps during the i24-hr PRA mission time is approximately 8.2E-3. The likelihood of an unisolated containment given an accident initiator is approximately 2.2E-3, and the likelihood of other necessary extreme plant conditions (e.g., high river temperature, high reactor power, reduced suppression pool water level) existing at the 4-5 C1320503-6924R1 -3/22/2006 I | ||
* Conditional probability that suppression pool water temperature | |||
> 91°F given river water temperature | BFNEPUCOPProbabilisticRiskAssessment time of the LLOCA is approximately 2.4E-3. As such, the base quantification results in an approximate 4.3E-8 conditional probability, given a LLOCA, of loss of low pressure ECCS pumps due to insufficient NPSH due to inadequate COP. | ||
> | This sensitivity discusses the risk impact of also explicitly quantifying scenarios with only 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 t:he sensitivities discussed here. As shown in Table 3-1, even with very conservative assumptions, if 3 or more RHR pumps and heat exchangers are operating in SPC, there is no need for containment overpressure. To result in a need for COP credit in such cases would require even more conservative input assumptions than the 2 RHR pump scenario. As such, the additional risk from such scenarios is non-significant compared to the 2 RHR pump case explicitly modeled in this analysis. | ||
-3/22/!006 I | 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 1d in Table 3-1): | ||
* Sum of BFN PRA Large LOCA initiator frequencies: 3E-5/yr | |||
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: | * Likelihood of failure of 1 RHR pump or 1 RHR heat exchanger during the 24-hr PRA mission time: 1.OOE-2 (nominal estimate) | ||
3E-3 (nominal from base analysis)* Probability of river water temperature | * 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) | ||
100OF is assumed here as the river water temperature at which COP credit is required (refer to Case I of Table 3-1).* Conditional probability that suppression pool water temperature | * Probability of river water temperature >900 F at any given time: 9E-2 (nominal value based on Table C-1. Although the' river temperature has not exceeded 900 F based ion 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 > 900F: 1.0 (refer to Figure C-1). | ||
> 100F: 1.0 (refer to Figure C-1)..No probabilistic credit for low suppression pool volume or low heat exchanger effectiveness is taken here.* deltaCDF contribution for 3 RHR pump case: 3.1 E-5 x 1.0 x 5E-3 x 3E-3 x | * No probabilistic credit for low suppression pool volume or low heat exchanger effectiveness is taken here. | ||
-3/22/2006 I | 4-6 C1320503-6924R1 -3/22/!006 I | ||
-BFN | |||
4-8 C1320503-6924R1 | BFNEPUCOPProbabilisticRisk 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 case is non-significant in comparison to the 2 RHR pump case. | ||
-3/22/2006 l | 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): | ||
! E | * Sum of BFN PRA Large LOCA initiator frequencies: 3E-5/yr | ||
-3/2212D6 I | * 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 > 1000 F at any given time: 1E-3 (estimate based on Table C-1. Although the river temperature has not exceeded 900F 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 I of Table 3-1). | |||
* Conditional probability that suppression pool water temperature > 950 F given river water temperature > 100F: 1.0 (refer to Figure C-1). | |||
. No probabilistic credit for low suppression pool volume or low heat exchanger effectiveness is taken here. | |||
* deltaCDF contribution for 3 RHR pump case: 3.1 E-5 x 1.0 x 5E-3 x 3E-3 x 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 case is non-significant in comparison to the 2 RHR pump case. | |||
4-7 C1320503-6924R1 - 3/22/2006 I | |||
-BFN EPU COPProbabilisticRisk Assessment Summary of Modelinq Uncertainty Results The modeling uncertainty sensitivity cases are summarized in Table 4-2. | |||
4.2.3 Completeness Uncertainty Analysis As stated previously, completeness uncertainty is addressed here by the qualitative assessment of the impact on the conclusions if external events and shutdown iisk contributors are also considered. | |||
4-8 C1320503-6924R1 - 3/22/2006 l | |||
! E BFNEPUCOPProbabilisticRiskAssessrr ent Table 4-1 PARAMETRIC UNCERTAINTY ANALYSIS RESULTS Statistic!Bs cBFN Unit 1 D COP RiskCDF Assessment 5% 4.71 E-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 4-9 C1320503-6924R1 -3/2212D6 I | |||
BFNEPUCOP ProbabilisticRiskAssessment Table 4-2 | |||
==SUMMARY== | ==SUMMARY== | ||
OF SENSITIVITY QUANTIFICATIONS Case Description CDF LERF ACDF(2 1 ALERFI | OF SENSITIVITY QUANTIFICATIONS Case Description CDF LERF ACDF( 2 1 ALERFI 2l Base(' Base Case Quantification (20 La leak size) 1.768E-6 4.411 E-7 1.4E-9 1.4E-9 l 1(1) Pre-Existing Containment Leakage Sufficient to Fail COP Credit 1.768E-6 4.411 E-7 1.4E-9 1.4E-9 Defined by 100La 2(l) Assume Low Suppression Pool Water Volume (123,500 ft3) Exists 1.768E-6 4.413E-7 1.6E-9 1.6E-9 100%/ of the Timei 3(1) Expansion of Containment Isolation fault tree to Encompass Smaller 1.768E-6 4.411 E-7 1.4E-9 1.4E-9 l Lines (approximate by multiplying Cont. Isol. failure probability bylOx)_ _ | ||
I 4-10 C13205 924R 3/22/2006 I | 4(1) Assume Initial Power Level and Water Temperature and Level Pre- 1.770E-6 4.432E-7 3.5E-9 3.5E-9 l Conditions Exist 100% of the Time 5e() -Combination of Cases #3 and #4- 1.773E-6 4.463E-7 6.6E-9 6.6E-9 l 6 Incorporation of "3-RHR pumps in SPC" and "4-RHR pumps in SPC' 1.768E-6 4.411 E-7 1.4E-9 1.4E-9 l loss of NPSH scenarios Notes: | ||
(1) Scenarios with failure of 2 or more RHR pumps and associated heat exchangers in SPC are explicitly analyzed in these cases. As shown in Case 6, explicit incorporation of scenarios with 0 or 1 RHR pumps in SPC failed has a negligible impact on the results. | |||
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/COR-4448) that seismic risk is overwhelmingly dominated by seismic induced equipment and structural failures. | (2) The ACDF and ALERF values are with respect to the BFN Unit 1 PRA model of record CDF of 1.767E-6/yr and LERF of 4.397E-7/yr. I 4-10 C13205 924R 3/22/2006 I | ||
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.4-11 C1320503-6924R1 | |||
-3122/2006 | BFNEPUCOP ProbabilisticRiskAssessment 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. | ||
Like most plants, BFN currently does not maintain a fire PRA. However, given the very low risk impact of the COP credit, even if fire risk was explicitly quantified the conclusions of this risk assessment are not expected to change, i.e., the risk impact, is very small.Other External Hazards In addition to seismic events and internal fires, the BFN IPEEE Submittal analyzed a variety of other external hazards:* High Winds/Tomadoes | 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. | ||
* External Floods* Transportation and Nearby Facility Accidents* Other External Hazards The BFN IPEEE analysis of high winds, tornadoes, external floods, transportation accidents, nearby facility accidents, and other external hazards was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards.Based upon this review, it was 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.4-12 C1320503-6924R1 | 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. | ||
-3/22/2006 | Industry BWR seismic PSAs have typically shown (e.g., Peach Bottom NUREG-1150 study; Limerick Generating Station Severe Accident Risk Assessment; NUREG/COR-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. | ||
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. | 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. | ||
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 I PRA results and conclusions not applicable to Units 2 and 3. As further evidence, the Unit 3 PRA was modified in a similar manner as the Unit 1 sensitivity Case #2 and quantified to determine the ACDF impact. The result for Unit 3 was a deltaCDF of 1.9E-9/yr. | 4-11 C1320503-6924R1 - 3122/2006 | ||
The revised BFN Unit 3 PRA RISKMAN model supporting this review is archived in file U3COP2-9 and saved on the BFN computers along with the other BFN PRA RISKMAN models.Given the above, the results for the Unit 1 PRA risk assessment are comparable to the Units 2 and 3 PRAs.The U2/U3 assessment discussed in this sub-section was performed for the Rev. 0 analysis. | |||
Given the similar results obtained in Rev. 1 analysis using the U-1 model, the U2/U3 assessment discussed above was not re-performed as the conclusion would be the same.4-13 C1320503-6924R1 | BFNEPUCOPProbabilisticRisk Assessment 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 EF'RI FIVE (Fire Induced Vulnerability Evaluation) methodology. | ||
-3/22/!006 I | 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: | |||
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 i in increases of core damage frequency (CDF) below 10Q/yr. Based on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in CDF (1.4E-09/yr). | * High Winds/Tomadoes | ||
* External Floods | |||
* Transportation and Nearby Facility Accidents | |||
* Other External Hazards The BFN IPEEE analysis of high winds, tornadoes, external floods, transportation accidents, nearby facility accidents, and other external hazards was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards. | |||
Based upon this review, it was 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. | |||
4-12 C1320503-6924R1 -3/22/2006 | |||
BFNEPUCOP ProbabilisticRiskAssessment 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 I PRA results and conclusions not applicable to Units 2 and 3. As further evidence, the Unit 3 PRA was modified in a similar manner as the Unit 1 sensitivity Case #2 and quantified to determine the ACDF impact. The result for Unit 3 was a deltaCDF of 1.9E-9/yr. The revised BFN Unit 3 PRA RISKMAN model supporting this review is archived in file U3COP2-9 and saved on the BFN computers along with the other BFN PRA RISKMAN models. | |||
Given the above, the results for the Unit 1 PRA risk assessment are comparable to the Units 2 and 3 PRAs. | |||
The U2/U3 assessment discussed in this sub-section was performed for the Rev. 0 analysis. Given the similar results obtained in Rev. 1 analysis using the U-1 model, the U2/U3 assessment discussed above was not re-performed as the conclusion would be the same. | |||
4-13 C1320503-6924R1 - 3/22/!006 I | |||
BFNEPUCOPProbabilisticRiskAssessmont Section 5 CONCLUSIONS The report documents the risk impact of utilizing containment accident pressure (containment overpressure) to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps during DBA LOCAs. | |||
The need for COP credit requests is driven by the conservative nature of design basis accident calculations. Use of more realistic inputs in such calculations shows that no credit for COP is required. | |||
The conclusions of the plant internal events risk associated with this assessment are as follows. | |||
: 1) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting i in increases of core damage frequency (CDF) below 10Q/yr. Based on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in CDF (1.4E-09/yr). | |||
: 2) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting fin increases of Large Early Release Frequency (LERF) below 10-7 yr. Based on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in LERF (1.4E-09/yr). | : 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 fin increases of Large Early Release Frequency (LERF) below 10-7 yr. Based on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in LERF (1.4E-09/yr). | ||
These results are well within the guideline of RG 1.174 for a "very small" risk increase.Even when modeling uncertainty and parametric uncertainty, and external event scenarios are considered, the risk increase is smrall. As such, the credit for COP in determining adequate NPSH for low pressure ECCS pumps during DBA LOCAs is acceptable from a risk perspective. | These results are well within the guideline of RG 1.174 for a "very small" risk increase. | ||
5-1 C1320503-6924R1 | Even when modeling uncertainty and parametric uncertainty, and external event scenarios are considered, the risk increase is smrall. As such, the credit for COP in determining adequate NPSH for low pressure ECCS pumps during DBA LOCAs is acceptable from a risk perspective. | ||
-312212006 l | 5-1 C1320503-6924R1 -312212006 l | ||
BFN | |||
-3/2212006 I | BFN EPU COP ProbabilisticRisk Assessment The general conclusions that the risk impact from the COP credit for DBA LOCAs is very small, applies to BFN Unit 1 as well as BFN Units 2 and 3. | ||
5-2 C1320503-6924R1 - 3/2212006 I | |||
[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-TS448, July 8, 2004.[2] Risk Impact Assessment of Extended Intearated Leak Rate Testing Intervals, EPRI Report 1009325, Final Report, December2003. | |||
[3] "Project Task Report -Browns Ferry Units 1, 2 & 3 EPU, RAI Response -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 C1320503-6924R1 | BFNEPUCOP ProbabilisticRiskAssessment REFERENCES | ||
-3/22J.?O6 | [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-TS448, July 8, 2004. | ||
* Level of detail in PRA* Maintenance of the PRA* Comprehensive Critical Reviews A. 1 LEVEL OF DETAIL The BFN Unit 1 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 iisk assessment was the most recent internal events risk model for the BFN Unit I plant at EPU conditions (BFN model U1050517). | [2] Risk Impact Assessment of Extended Intearated Leak Rate Testing Intervals, EPRI Report 1009325, Final Report, December2003. | ||
The BFN PRA models adopts the large event tree / small fault tree approach and use the support state methodology, contained in the RISIKMAN code, for quantifying core damage frequency. | [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 C1320503-6924R1 -3/22J.?O6 | |||
BFNEPUCOPProbabilisticRiskAssessment 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 Llnit 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 1 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 iisk 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 RISIKMAN code, for quantifying core damage frequency. | |||
The PRA model contains the following modeling attributes. | 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 C1320503-6924R1 | A.1.1 Initiating Events The BFN at-power PRA explicitly models a large number of internal initiating events: | ||
-3/22/2006 | A-1 C1320503-6924R1 -3/22/2006 | ||
BFNEPUCOP ProbabilisticRisk Assessment | |||
* General transients | * General transients | ||
* LOCAs* Support system failures* Internal Flooding events The initiating events explicitly modeled in the BFN at-power PRA are summarized in Table A-1. The number of internal initiating events modeled in the BFN at-power PRA is similar to or greater than the majority of U.S. BWR PRAs currently in use.A.1.2 System Models The BFN at-power PRA explicitly models a large number of frontline and support systems that are credited in the accident sequence analyses. | * LOCAs | ||
The BFN systems explicitly modeled in the BFN at-power PRA are summarized in Table A-2. The number and level of detail of plant systems modeled in the BFN at-power PRA is equal to or greater than the majority of U.S. BWR PRAs currently in use.A. 1.3 Operator Actions The BFN at-power PRA explicitly models a large number of operator actions:* Pre-Initiator actions* Post-Initiator actions* Recovery Actions Dependent Human Actions Approximately fifty operator actions are explicitly modeled in the BFN PRA. A summary table of the individual actions modeled is not provided here.A-2 C1320503-6924R1 | * Support system failures | ||
-3/2,2O6 I BFN | * Internal Flooding events The initiating events explicitly modeled in the BFN at-power PRA are summarized in Table A-1. The number of internal initiating events modeled in the BFN at-power PRA is similar to or greater than the majority of U.S. BWR PRAs currently in use. | ||
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 EFPRI TR-I 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. | 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. | ||
Approximately two thousand common cause terms are included in the EFN 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 F'RA 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 C1320503-6924R1 | A. 1.3 Operator Actions The BFN at-power PRA explicitly models a large number of operator actions: | ||
-3/22'2006 I | * Pre-Initiator actions | ||
* Post-Initiator actions | |||
* 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 | * 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. | ||
* 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 de:ail 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 C1320503-6924R1 - 3/2,2O6 I | ||
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. | BFN EPU COPProbabilisticRisk Assessment The human error probabilities for the actions are modeled with accepted industry HRA techniques. | ||
Ad4 C1320503-6924R1 | 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 EFPRI TR-I 00259. | ||
-3122/,0V6 l | 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 EFN 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. | |||
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. | 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 F'RA Application Guide. | ||
The BFN Unit 1 PRA was initially developed in June 2004 using the guidance in the ASME PRA Standard, and to incorporate the latest plant configuration (including EPU) and operating experience data. The Unit 1 PRA was then subsequently updated in August 2005. The Unit 1 PRA was developed using the BFN Unit 2 and Unit 3 PRAs as a starting point.The BFN Unit 2 and Unit 3 PRAs have been updated numerous times since the original | 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: | ||
The BFN Unit 2 PRA revisions are summarized below: Original BFN IPE Submittal 9/92 Revision to address plant changes and 8/94 incorporate BFN IE and EDG experience data Revision to ensure consistency with the 4/95 BFN Multi-Unit PRA Revision to address PER BFPER 970754 10/97 2002 PRA Update 3/02 2004 PRA Update (includes conditions to 6/04 reflect EPU)2005 Update 8/05 A-5 C1320503-6924R1 | A-3 C1320503-6924R1 - 3/22'2006 I | ||
-3/212006 | |||
The purpose of the peer review process is to establish a method of assessing the technical quality of the PRA for its potential applications. | BFNEPUCOPProbabilisticRiskAssessment | ||
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. | * Primary containment isolation | ||
The review system used allowed the Peer Review team to focus on technical issues and to issue their assessment results in the form of a "grade" of 1 through 4 on a PRA sub-element level.To reasonably span the spectrum of potential PRA applications, the four grades of certification as defined by the BWROG document "Report to the Industry on PRA Peer Review Certification Process -Pilot Plant Results" were employed.During the Unit 2 and 3 PSAs updates in 2003, the significant findings (i.e., designated as Level A or B) from the Peer Certification were resolved, resulting in the PRA elements now having a minimum certification grade of 3. The Unit 1 PRA used in this analysis has incorporated the findings of the Units 2 and 3 PSA Peer Review. The previously conducted Peer Review was effectively an administrative and technical Peer Review of the Unit 1 PRA. Similar models, processes, policies, approaches, reviews, and management oversight were utilized to develop the Unit 1 PRA.Ax6 C1320503-6924R1 | * RPV depressurization post-core damage | ||
-3/22/2006 I | * 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 de:ail 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. | |||
Ad4 C1320503-6924R1 - 3122/,0V6 l | |||
BFNEPUCOPProbabilisticRiskAssessment In addition, the PRA models are routinely implemented and studied by plant PRA personnel in the performance of their duties. Potential model modifications or enhancements are itemized and maintained for further investigation and subsequent implementation, if warranted. Potential modifications identified as significant to the results or applications may be implemented in the model at the time the change occurs if their impact is significant enough to warrant. | |||
A.2.1 History of BFN PRA Models The current BFN Unit 1 PRA is the model used for this analysis. The BFN Unit 1 PRA was initially developed in June 2004 using the guidance in the ASME PRA Standard, and to incorporate the latest plant configuration (including EPU) and operating experience data. The Unit 1 PRA was then subsequently updated in August 2005. The Unit 1 PRA was developed using the BFN Unit 2 and Unit 3 PRAs as a starting point. | |||
The BFN Unit 2 and Unit 3 PRAs have been updated numerous times since the original IPE Submittal. The BFN Unit 2 PRA revisions are summarized below: | |||
Original BFN IPE Submittal 9/92 Revision to address plant changes and 8/94 incorporate BFN IE and EDG experience data Revision to ensure consistency with the 4/95 BFN Multi-Unit PRA Revision to address PER BFPER 970754 10/97 2002 PRA Update 3/02 2004 PRA Update (includes conditions to 6/04 reflect EPU) 2005 Update 8/05 A-5 C1320503-6924R1 - 3/212006 | |||
BFNEPUCOPProbabilisticRiskAssessment A.3 COMPREHENSIVE CRITICAL REVIEWS As described above, the BFN Unit I 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. | |||
Ax6 C1320503-6924R1 - 3/22/2006 I | |||
BFNEPUCOP ProbabilisticRiskAssessment A.4 PRA QUALITY | |||
==SUMMARY== | ==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 I 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 C1320503-6924R1 | |||
-3/22/2006 l | 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 | |||
-3/2Y200O6I | * Maintenance of the PRA | ||
-3/2212006 l | * Comprehensive Critical Reviews The BFN Unit I 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 C1320503-6924R1 -3/22/2006 l | |||
Hardened Wetwell Vent High Pressure Coolant Injection Main Steam System Plant Air Systems Primary Containment Isolation Raw Cooling Water Reactor Building Closed Cooling Water Reactor Core Isolation Cooling Reactor Protection System Recirculation System Residual Heat Removal System RHR Service Water Secondary Containment Isolation Shared Actuation Instrumentation System SRVs/ADS Standby Gas Treatment System Standby Liquid Control System A-1 0 C1320503-6924R1 | |||
-3/22/2006 l | BFNEPUCOP ProbabilisticRiskAssessment 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 A-8 C1320503-6924R1 - 3/2Y200O6I | |||
-3/22oo6 I | |||
* Guidance Documents for Initiating Event Analysis a Groupings-Transient-LOCA-Support System/Special | BFNEPUCOPProbabilisticRiskAssessment Table A-1 INITIATING EVENTS FOR BFN PRA Initiator Mean Frequency Category (events per year) | ||
-ISLOCA-Break Outside Containment | Large LOCA- Core Spray Line Break Loop I 1.68E-6 Loop II 1.68E-6 Large LOCA - Recirculation Discharge Line Break Loop A 1.1 8E-5 Loop B 1.1 8E-5 Large LOCA- Recirculation Suction Line Break Loop A 8.39E-7 Loop B 8.39E-7 Other Large LOCA 8.39E-7 Medium LOCA Inside Containment 3.80E-5 Small LOCA Inside Containment 4.75E-4 Very Small LOCA Inside Containment 5.76E-3 Internal Flooding Initiator Categoriesies EECW Flood in Reactor Building - shutdown units 1.20E-3 EECW Flood in Reactor Building - operating unit 1.85E-6 Flood from the Condensate Storage Tank 1.22E-4 Flood from the Torus 1.22E-4 Large Turbine Building Flood 3.65E-3 Small Turbine Building Flood 1.65E-2 A-9 C1320503-6924R1 -3/2212006 l | ||
-Internal Floods* Subsumed Events* Data* Documentation Accident Sequence Evaluation | |||
BFNEPUCOPProbabilisticRisk 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-1 0 C1320503-6924R1 - 3/22/2006 l | |||
BFNEPUCOP ProbabilisticRiskAssessment Table A-2 BFN PRA MODELED SYSTEMS Suppression Pool / Vapor Suppression Turbine Bypass and Main Condenser A-1 1 C1320503-6924R1 - 3/22oo6 I | |||
BFNEPUCOPProbabilisticRiskAssessment Table A-3 PRA PEER REVIEW TECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Initiating Events | |||
* Guidance Documents for Initiating Event Analysis a 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) | * Guidance on Development of Event Trees (Event Trees) | ||
* Event Trees (Accident Scenario Evaluation) | * Event Trees (Accident Scenario Evaluation) | ||
-Transients | - Transients | ||
-SBO-LOCA-ATWS,-Special-ISLOCA1BOC | - SBO | ||
-Internal Floods* Success Criteria and Bases* Interface with EOPs/AOPs* Accident Sequence Plant Damage States* Documentation A-1 2 C1320503-6924R1 | - LOCA | ||
-3/22J2006 I | - ATWS, | ||
- Special | |||
* Guidance Document* Best Estimate Calculations (e.g., MAAP)* Generic Assessments | - ISLOCA1BOC | ||
* FSAR -Chapter 15* Room Heat Up Calculations | - Internal Floods | ||
* Success Criteria and Bases | |||
* Interface with EOPs/AOPs | |||
* Accident Sequence Plant Damage States | |||
* Documentation A-1 2 C1320503-6924R1 - 3/22J2006 I | |||
BFNEPUCOPProbabilisticRiskAssessment Table A-3 PRA PEER REVIEWTECHNICAL ELEMENTS FOR LEVEL I | |||
_ , | |||
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 | * Documentation System Analysis | ||
* System Analysis Guidance Document(s)(Fault Trees) | * System Analysis Guidance Document(s) | ||
* System Models-Structure of models-Level of Detail-Success Criteria Nomenclature | (Fault Trees) | ||
-Data (see Data Input)-Dependencies (see Dependency Element)-Assumptions | * System Models | ||
* Documentation of System Notebooks A-1 3 C1320503-6924R1 | - Structure of models | ||
-3/2212006 | - Level of Detail | ||
---------- --I --Data Analysis* Guidance* Component Failure Probabilities | - Success Criteria Nomenclature | ||
- Data (see Data Input) | |||
- Dependencies (see Dependency Element) | |||
-Assumptions | |||
* Documentation of System Notebooks A-1 3 C1320503-6924R1 -3/2212006 | |||
BFNEPUCOPProbabilisticRiskAssessment Table A-3 PRA PEER REVIEW TECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT CERTIFICATION SUB-ELEMENTS | |||
- - -------- - - I - - | |||
Data Analysis | |||
* Guidance | |||
* Component Failure Probabilities | |||
* System/Train Maintenance Unavailabilities | * System/Train Maintenance Unavailabilities | ||
* Common Cause Failure Probabilities | * 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 | * Unique Unavailabilities or Modeling Items | ||
-Pipe Rupture Failure Probability | - AC Recovery | ||
* Documentation 4.Human Reliability Analysis* Guidance* Pre-Initiator Human Actions-Identification | - Scram System | ||
-Analysis-Quantification | - EDG Mission Time | ||
* Post-Initiator Human Actions and Recovery-Identification | - Repair and Recovery Model | ||
-Analysis-Quantification | - SORV | ||
* Dependence among Actions* Documentation A-14 C1320503-6924R1 | - LOOP Given Transient | ||
-3/22'2006 l | - BOP Unavailability | ||
- Pipe Rupture Failure Probability | |||
* Guidance Document on Dependency Treatment* Intersystem Dependencies | * Documentation 4. | ||
* Treatment of Human Interactions (see also HRA)* Treatment of Common Cause* Treatment of Spatial Dependencies | Human Reliability Analysis | ||
* Walkdown Results* Documentation Structural Capability | * Guidance | ||
* Guidance* RPV Capability (pressure and temperature) | * Pre-Initiator Human Actions | ||
-ATWS-Transient* Containment (pressure and temperature) | - Identification | ||
* Reactor Building* Pipe Overpressurization for ISLOCA* Documentation Quantification/Results | - Analysis | ||
- Quantification | |||
* Post-Initiator Human Actions and Recovery | |||
- Identification | |||
- Analysis | |||
- Quantification | |||
* Dependence among Actions | |||
* Documentation A-14 C1320503-6924R1 -3/22'2006 l | |||
BFNEPUCOPProbabilisticRiskAssessment 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) | |||
* Reactor Building | |||
* Pipe Overpressurization for ISLOCA | |||
* Documentation Quantification/Results | |||
* Guidance lInterpretation | * Guidance lInterpretation | ||
* Computer Code* Simplified Model (e.g., cutset model usage)* Dominant Sequences/Cutsets | * Computer Code | ||
* Simplified Model (e.g., cutset model usage) | |||
* Dominant Sequences/Cutsets | |||
* Non-Dominant Sequences/Cutsets | * Non-Dominant Sequences/Cutsets | ||
* Recovery Analysis* Truncation | * Recovery Analysis | ||
* Truncation | |||
* Uncertainty | * Uncertainty | ||
* Results Summary A-1 5 C1320503-6924R1 | * Results Summary A-1 5 C1320503-6924R1 - 3/22006 l | ||
-3/22006 l | |||
* Important HEPs* Containment Capability Assessment | BFNEPUCOPProbabilisticRiskAssessment Table A-4 PRA CERTIFICATION TECHNICAL ELEMENTS FOR LEVEL 2 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Containment Performance Analysis | ||
* Guidance Document | |||
* Success Criteria | |||
* L1/L2 Interface | |||
* Phenomena Considered | |||
* Important HEPs | |||
* Containment Capability Assessment | |||
* End state Definition | * End state Definition | ||
* LERF Definition | * LERF Definition | ||
* CETs* Documentation A-16 C1320503-6924R1 | * CETs | ||
-3/222006 l | * Documentation A-16 C1320503-6924R1 - 3/222006 l | ||
* Guidance Document* Input -Monitoring and Collecting New Information | |||
* Model Control* PRA Maintenance and Update Process* Evaluation of Results* Re-evaluation of Past PRA Applications | BFNEPUCOPProbabilisticRiskAssessnr.ent Table AZ5 PRA CERTIFICATION TECHNICAL ELEMENTS FOR MAINTENANCE AND UPDATE PROCESS PRA ELEMENT l CERTIFICATION SUB-ELEMENTS Maintenance and Update Process | ||
* Documentation I ;A-17 C1320503-6924R1 | * Guidance Document | ||
-3/22/2006 | * Input - Monitoring and Collecting New Information | ||
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] | * Model Control | ||
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. | * PRA Maintenance and Update Process | ||
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. | * Evaluation of Results | ||
Approximately half (32 of the 71 events) had identified leakage rates of less than or equal to 11La (i.e., the Technical Specification containment allowed leakage rate). None of the 71 events had identified leakage rates greater than 211La. EFPRI 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 preexisting leakage size are summarized here in Table B-1.The EPRI 1009325 study used 1100La 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 C1320503-6924R1 | * Re-evaluation of Past PRA Applications | ||
-3122/2006 l | * Documentation I ; | ||
A-17 C1320503-6924R1 - 3/22/2006 | |||
Earlier ILRT industry guidance (NEI Interim Guidance -see Ref. 10 of EFPRI 1009325) conservatively recommended use of 10-La 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-exisling containment leakage pathway sufficient to defeat the containment overpressure credit.Such a hole size does not realistically represent a LERF release (based on EP'RI 1009325) and is also believed (based on the VY hole size estimate) to be on the low end of a hole size that would preclude containment overpressure credit. As can be seen from Table B-1, the probability of a 20La pre-existing containment leakage at any given time at power is 1.88E-03.Sensitivity studies to the base case quantification (refer to Section 4) assess the sensitivity of the results to the pre-existing leakage size assumption. | |||
B-2 C1320503-6924R1 | BFNEPUCOPProbabilisticRisk Assessment Appendix B PROBABILITY OF PRE-EXISTING CONTAINMENT LEAKAGE Containment failures that may be postulated to defeat the containment overpresslire 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] | ||
-3/2212006 | 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. | ||
[21 recommends these values for use for both BWRs and PWRs. Reference | 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 11La (i.e., the Technical Specification containment allowed leakage rate). None of the 71 events had identified leakage rates greater than 211La. EFPRI 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 preexisting leakage size are summarized here in Table B-1. | ||
[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 C1320503-6924R1 | The EPRI 1009325 study used 1100La 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. | ||
-3122'2006 l | B-1 C1320503-6924R1 -3122/2006 l | ||
BFN | |||
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. | BFNEPUCOP ProbabilisticRiskAssessment 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 B\W\R 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 EFPRI 1009325) conservatively recommended use of 10-La to represent "small" containment leakages and 35La to represent "large" containment leakages. | ||
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 IData Period Years River Water Temperature and 2 01/01/00 -01/31/06 6.1 Suppression Pool Temperature 3 02/01/03 -01/31/06 3.0 Suppression Pool Level 2 01/01/00 -01/31/06 6.1 3 02/01/03 -01/31/06 3.0 The river water temperature data from the above units is not pooled because river temperature is dependent upon the seasonal, cycle in weather and is not independent between the units. Use of data for SW inlet temperatures from multiple units 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 C1320503-6924R1 | Given the above, the base analysis here assumes 20La as the size of a pre-exisling containment leakage pathway sufficient to defeat the containment overpressure credit. | ||
-312X12006 l | Such a hole size does not realistically represent a LERF release (based on EP'RI 1009325) and is also believed (based on the VY hole size estimate) to be on the low end of a hole size that would preclude containment overpressure credit. As can be seen from Table B-1, the probability of a 20La pre-existing containment leakage at any given time at power is 1.88E-03. | ||
Sensitivity studies to the base case quantification (refer to Section 4) assess the sensitivity of the results to the pre-existing leakage size assumption. | |||
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. | B-2 C1320503-6924R1 -3/2212006 | ||
Also, the river water temperatures and torus temperatures are closely correlated in the warmer months when river water temperature is above approximately 70 | |||
BFNEPUCOPProbabilisticRiskAssessment Table B-1 PROBABILITY OF PRE-EXISTING UNISOLABLE CONTAINMENT LEAK [2] | |||
(as a Function of Leakage Size)(') | |||
Leakage Size Mean Probability of (La) Occurrence 1 2.65E-02 2 1.59E-02 5 7.42E-03 10 3.88E-03 20 1.88E-03 35 9.86E-04 50 6.33E-04 100 2.47E-04 200 8.57E-05 500 1.75E-05 600 1.24E-05 Notes: | |||
(1) Reference [21 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 C1320503-6924R1 - 3122'2006 l | |||
BFN EPU COPProbabilisticRisk 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 IData Period Years River Water Temperature and 2 01/01/00 - 01/31/06 6.1 Suppression Pool Temperature 3 02/01/03 - 01/31/06 3.0 Suppression Pool Level 2 01/01/00 - 01/31/06 6.1 3 02/01/03 - 01/31/06 3.0 The river water temperature data from the above units is not pooled because river temperature is dependent upon the seasonal, cycle in weather and is not independent between the units. Use of data for SW inlet temperatures from multiple units 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 C1320503-6924R1 -312X12006 l | |||
BFNEPUCOP ProbabilisticRiskAssessment 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 0F. | |||
The 6.1 years of temperature data was categorized into 5-degree temperature bins ranging from 500 F to 990 F degrees. The resulting histograms are shown in Figures C-2 and C-3. Figure C-2 presents histogram for the river water temperature and Figure C-3 presents the histogram for the torus water temperature. | |||
The histogram information was then used in a statistical analysis software package (Crystal Ball, a MS Excel add-in, developed by Decisioneering, Inc. of Denver, C0) to approximate a distribution of the expected range in temperature. | The histogram information was then used in a statistical analysis software package (Crystal Ball, a MS Excel add-in, developed by Decisioneering, Inc. of Denver, C0) 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 C13205036924R1 -3/2,Y2006 I | |||
BFNEPUCOPProbabilisticRiskAssessment lower bounds. If upper and lower bounds are not defined, the tails of the curve fit distribution extend to unrealistic values (e.g., river water and torus water temperatures below 0NF degrees). To constrain the distributions, the following user-defined upper and lower bounds were used: | |||
* River water temperature lower bound of 320 F (no data points in the 6.1 years of data reached 320F, only a single data point reached 350F) | |||
* River water temperature upper bound of 950 F (no data points in the 6.1 years of data exceeded 900F) | |||
* Torus water temperature lower bound of 550 F (no data points in the 6.1 years of data reached lower than 570F) | |||
* Torus water temperature upper bound of 950 F (only a single data point in the 6.1 years of data reached 93 0F) | |||
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 | |||
* 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. | * 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. | * 6-24 hours is a time frame in which much of the offsite nuclear plant protective measures can be assured to be accomplished. | ||
* >24 hours are times at which the offsite measures can be assumed to be fully effective. | * >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: | Magnitude Categorization The BFN Level 2 PRA defines the following radionuclide release magnitude classifications: | ||
: 1) Frigh (H) - A radionuclide release of sufficient magnitude to have the potential to cause prompt fatalities. | |||
: 2) Medium or Moderate (M) - A radionuclide release of sufficient magnitude to cause near-term health effects. | |||
: 3) Low (L) - A radionuclide release with the potential for latent health effects. | |||
: 4) Low-Low (LL) - A radionuclide release with undetectable or minor health effects. | |||
: 5) Negligible (OK) - A radionuclide release that is less than or equal to the containment design base leakage. | |||
The definition of the source terms levels distinguishing each of these release severity categories is based on the review of existing consequence analyses performed in previous industry studies, PRAs and NRC studies containing detailed consequence modeling. The BFN Level 2 PRA uses cesium as the measure of the source term magnitude because it delivers a substantial fraction of the total whole body population dose. This approach is typical of most industry PRAs. | |||
In terms of fraction of core inventory CsI released, the BFN release magnitude classification is as follows: | |||
D-2 C1320503-6924R1 -3122/2006 l | |||
BEN EPU COPProbabilisticRisk Assessment Release Magnitude l Fraction of Release CsI Fission Products High greater than 10% | |||
Medium/Moderate 1 to 10% | |||
Low | |||
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Page No. 2of4 MODEL Name: UlERIN 13:38:20 Febuary 16, 200e Event Tree: LLRSN.ETI SPII SPC ODWS DWS XI B5 | |||
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NRC Request ACVB.26 Demonstrate with a "realistic" or best-estimate calculation of available net positive suction head for the RHR or core spray pumps, whichever is most limiting, whether credit for containment accident pressure is needed. All input and assumptions should be, to the extent possible, nominal values. | |||
TVA Reply to ACVB.26 In the reply to NRC Request SPSB-A.11 in Enclosure 1 of this letter, the results of several calculations of NPSH for the RHR and CS pumps are presented which include the use of realistic input parameters. The realistic inputs and assumptions for each case as well as the resultant need for COP are presented in Table SPSB-A.11-2 in Enclosure 1. Using realistic assumption:, | |||
COP is not required to ensure adequate NPSH for a DBA-LOCA. | |||
NRC Request ACVB.32 Provide for staff review the NPSH calculations (including the containment calculations) for the Units 2 and 3 core spray and RHR pumps at EPU conditions. | |||
TVA Reply to ACVB.32 provides a copy of Revision 8 of Calculation MDQ0999970046, "NPSH Evaluation of Browns Ferry RHR and CS Pumps." Enclosure 7 provides applicable information from the containment calculation that supports the NPSH evaluations. | |||
The realistic inputs and assumptions for each case as well as the resultant need for COP are presented in Table SPSB-A.11-2 in Enclosure | |||
TVA Reply to ACVB.32 | |||
E5-10}} | E5-10}} | ||
Revision as of 20:38, 23 November 2019
| ML060880413 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/21/2006 |
| From: | ERIN Engineering & Research |
| To: | Office of Nuclear Reactor Regulation, Tennessee Valley Authority |
| References | |
| Download: ML060880413 (177) | |