ML060620328
| ML060620328 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/28/2006 |
| From: | Crouch W Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC3812, TVA-BFN-TS-431 | |
| Download: ML060620328 (182) | |
Text
{{#Wiki_filter:R08 060228 739 Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabarna 35609-2000 February 28, 2006 TVA-BFN-TS-431 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen: In the Matter of ) Docket No. 50-259 Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 1 - TECHNICAL SPECIFICATIONS (TS) CHANGE TS-431 - RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION SPSB-A.11 REGARDING EXTENDED POWER UPRATE - CREDIT FOR NET POSITIVE SUCTION HEAD (TAC NO. MC3812) This letter provides TVA's supplemental responses to the NRC request for additional information SPSB-A.11 regarding an assessment of the credit for containment overpressure against the five key principles of risk-informed decision making. On June 28, 2004 (Reference 1), TVA requested a TS change to allow Unit 1 to operate at extended power uprate conditions. As part of this TS change, TVA requested approval to take credit for containment overpressure in order to provide adequate net positive suction head (NPSH) to the Emergency Core Cooling System (ECCS) pumps. On October 3, 2005 (Reference 2), NRC requested TVA provide additional information regarding the ECCS pumps NPSH, including an assessment of the credit for containment overpressure against the five key principles of risk-informed decision making. The requested additional information is provided as Enclosure 1 to this letter. A
U.S. Nuclear Regulatory Commission Page 2 February 28, 2006 detailed chronology of the correspondence related to the previous approval of NPSH for pre-uprate conditions on Units 2 and 3 is provided in Enclosure 2. A detailed description of plant systems related to the NPSH analysis is provided :..n . The supporting risk assessment is provided as . The use of containment overpressure to ensure adequate NPSH for ECCS pumps during a limited time after a design basis accident is consistent with NRC staff positions, including Revision 3 of Regulatory Guide 1.82, and is part of the current licensing and design basis for Browns Ferry Units 2 and 3. Crediting containment overpressure results in a small increase in core damage frequency (CDF) and large early release frequency (LERF) of 1.53xlO-9/yr. This small increase is well below the guidelines provided in Regulatory Guides 1.174 (10-6/yr for CDF and 10-7/yr for LERF). TVA has determined that the additional information provided does not affect the no significant hazards considerations associated with the proposed TS changes. The proposed TS changes still qualify for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). If you have any questions about this submittal, please contact me at (256) 729-2636. I declare under penalty of perjury that the foregoing is true and correct. Executed on February 28, 2006. Sincerely, William D. Crouch Manager of Licensing and Industry Affairs
References:
- 1. TVA letter, dated June 28, 2004, "Browns Ferry Nuclear Plant (BFN) - Unit 1 - Proposed Technical Specifications (TS) Change TS-431 - Request for License Amendment -
U.S. Nuclear Regulatory Commission Page 3 February 28, 2006 Extended Power Uprate (EPU) Operation."
- 2. NRC letter, dated October 3, 2005, "Browns Ferry Nuclear Plant, Unit 1 - Request for Additional Information for Extended Power Uprate (TS-431) (TAC No. MC3812)."
Enclosures:
- 1. Response to NRC Request for Additional Information Regarding Proposed Technical Specif:Lcation (TS) TS-431 Extended Power Uprate - Credit for Net Positive Suction Head
- 2. Detailed Chronology of Correspondence Related to the Previous Approval of NPSH for Pre-uprate Conditions
- 3. Detailed Description of P:Lant Systems Related to the NPSH Analysis
- 4. BFN Extended Power Uprate Containment Overpressure Credit Risk Assessment cc (Enclosures):
State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, AL 36130-3017
U.S. Nuclear Regulatory Commission Page 4 February 28, 2006 Enclosures cc (Enclosures): U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Su:Lte 23T85 Atlanta, Georgia 30303-34115 Mr. Malcolm T. Widmann, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Sui..te 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant: 10833 Shaw Road Athens, Alabama 35611-6970 Margaret Chernoff, Project: Manager U.S. Nuclear Regulatory Commission (MS 08G9) One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9) One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY (TVA) BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED TECHNICAL SPECIFICATION (TS) TS-431 EXTENDED POWER UPRATE - CREDIT FOR NET POSITIVE SUCTION HEAD NRC REQUEST 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 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 pressure affects the Probabilistic Risk Assessment (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 decision making stated in RG 1.174 and SRP Chapter 19. Specifically, demonstrate that the proposed containment accident pressure credit meets current regulations, is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in an increase in core-damage frequency and risk that is small and consistent with the intent of the Commission's Safety Goal Policy Statement, and will be monitored using performance measurement strategies. With respect to the fourth key principle (small increase in risk), provide a quantitative risk assessment that demonstrates that the proposed containment accident pressure credit meets the numerical risk acceptance guidelines in Section 2.2.4 of
RG 1.174. This quantitative risk assessment must include specific containment failure Mechanisms (e.g., liner fa:lures, penetration failures, primary containment isolation system failures) that cause a loss of containment pressure and subsequent loss of NPSH to the ECCS pumps. TVA RESPONSE INTRODUCTION The proposed change for BFN Unit 1 Extended Power Uprate! (EPU) includes crediting containment overpressure (COP) in ensuring adequate NPSH to Emergency Core Cooling System (ECCS) pumps following limiting events which cause suppression pool temperature increase. These events are Loss of Coolant Accident (LOCA), Anticipated Transients Without Scram (ATWS), Appendix R and Station Blackout (SBO). COP is defined for BFN as containment pressure in excess of 14.4 PSIA. For the Design Basis Accident (DBA) LOCA, the need to credit COP is due only to consideration of a number of worst case assumptions. More realistic analyses show that elimination of worst case assumptions that have reasonable probability distributions would eliminate the need for COP credit. Results of realistic analyses are presented along with associated probability distributions. Parameters affecting NPSH were included in a modified PEA model along with probability distributions to show the risk impact associated with reliance on containment integrity and overpressure for ECCS pump NPSH. RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" was utilized as a guide for providing risk insights and more realistic analyses to supplement the deterministic analyses and worst case assumptions used in the licensing basis LOCA analysis. These risk insights are used to characterize the degree to which COP is relied upon in the safety design basis. BACKGROUND The following provides an abbreviated background for ECCS strainer issues and the use of COP. An in-depth discussion of the regulatory background is provided in Enclosure 2. E1-2
Previously, BFN Units 2 and 3 installed new large capac:Lty ECCS strainers to meet the requested actions of NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Sirainers by Debris in Boiling-Water Reactors." As part of the resolution of Bulletin 96-03, credit for available COP to maintain adequate NPSH following a LOCA was required. BFN requested a change to the licensing basis for Units 2 and 3 in Reference 1 (as supplemented by Reference 2) and received NRC approval for the requested change in Reference 3. At the time Bulletin 96-03 was resolved for BFN Units 2 and 3, BFN Unit 1 was in an extended shutdown and no actions were taken to resolve Bulletin 96-03 for Unit 1. As part of the restart for Unit 1, large capacity ECCS strainers of the same design as previously installed on Units 2 and 3 have been installed on Unit 1. Credit for available COP to maintain adequate NPSH following a LOCA is also required for Unit 1, the same as Units 2 and 3. Since the intent is to restart Unit 1 at a licensed power level of 120% of original licensed power, NPSH margin analyses were not specifically performed for pre-EPU power levels. For EPU, BFN is proposing a change in the licensing basis to extend the existing approved credit for COP to provide adequate NPSH following a LOCA for Units 2 and 3 and to apply the same credit for COP to provide adequate NPSH following a LOCA for Unit 1. Currently for BFN Units 2 and 3, Reference 3 approves the crediting of 3 psi COP for the Residual Heat Removal (RHR) pumps for the first 10 minutes following a LOCA (short-term requirement) and 1 psi COP for the core spray pumps from approximately 5500 to 35000 seconds (about 8.2 hours) fc'llowing a LOCA (long-term requirement). For EPU, BFN is requesting for all three units approval of 3 psi COP for the RHR pumps for the first 10 minutes following a LOCA (short-term requirement) and 3 psi COP for the core spray pumps from approximately 4,lCO to 52,300 seconds (about 13.4 hours) following a LOCA (lonc-term requirement). As part of the EPU effort, BFN has also given more consideration for NPSH requirements during Appendix R, ATWS, and SBO events. These events (designated as Special Events at BFN) were not addressed in response to Generic Letters 96-03 and 97-04 and are not addressed in Regulatory Guide (RG) 1.82. Conservative evaluation of these events determined that BFN will credit EI-3
available containment pressure for the RHR pumps following an SBO, ATWS, and Appendix R events. SYSTEM DESCRIPTION The following provides an abbreviated system description. An in-depth description of the BFN containment and ECCS systems is provided in Enclosure 3. The BFN units are BWR-4s with Mark I containments, which incorporate a large torus shaped suppression pool. Four RHR pumps and four Core Spray pumps take suction from the suppression pool through a common ring header which connects to the torus at four locations through a stacked disc strainer mounted on each nozzle. The ECCS ring header is also the alternate suction for the High Pressure Core Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) system pumps. The normal suction path for the HPCI and RCIC system pumps is the condensate storage tank (CST). The four strainers are not associated with individual pump suctions but direct suppression pool water to the common ECCS ring header. Therefore, interaction between operating pumps is considered when determining pump suction pressures. LOCA EVENT DESCRIPTION SHORT TERM (T<10 minutes) The bounding design basis event for determining NPSH margin is a double ended recirculation discharge line break. This event results in maximum suppression pool temperature and maximum total pump flow. The discharge line break is chosen because the low system resistance on the broken line produces the most limiting flow and NPSH for two RHR pumps which are assumed to be pumping into the broken line inside containment. At the beginning of the event, four RHR pumps and four Core Spray pumps start automatically and align to inject to the Reactor Pressure Vessel (RPV). Two RHR pumps inject to the RPV at 10,000 gpm each, two RHR pumps inject through the broken line into the containment at 11,000 gpm each (greater than design flow), and four Core Spray pumps inject to the RPV at 3,125 gpm each. This mode of operation is assumed for 10 minutes consistent with not crediting operator action for 10 minutes. ECCS strainers are assumed to accumulate the maximum equilibrium debris load. During this time suppression pool temperature reaches 155.4 0 F E1-4
and only the RHR pumps require credit for COP in order to have sufficient NPSH margin as shown in Figure 1. LONG TERM (T>10 minutes) At 10 minutes, operator action. is assumed which places the minimum complement of ECCS pumps into modes required for long term cooling. Two Core Spray pumps (one loop) at design flow of 3,125 gpm each are assumed for core cooling, and two RHR pumps in one loop in containment cooling mode at 6,500 gpm each are assumed for pool cooling. Containment spray mode of containment cooling is chosen to minimize available containment pressure. Only two of four RHR pumps are assumed for pool cooling due to single failure considerations. ECCS strainers are assumed to accumulate the maximum equilibrium debris load. During this time suppression pool temperature reaches 187.4 0 F and only the two Core Spray pumps require credit for COP in order to have sufficient NPSH margin as shown in Figure 2. REGULATORY GUIDE 1.174 ASSESSMENT RG 1.174, Section 2, provides the set of five key principles that licensing basis changes are expected to meet:
- 1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change, i.e., a "specific exemption" under 10 CFR 50.12 or a "petition for rulemaking" under 10 CFR 2.802.
- 2. The proposed change is consistent with the defense-in-depth philosophy.
- 3. The proposed change maintains sufficient safety margins.
- 4. When proposed changes result in an increase in core damage frequency or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement (Ref. RG 1.175).
- 5. The impact of the proposed change should be monitored using performance measurement strategies.
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- 1. CURRENT REGULATIONS On June 28, 2004, TVA requested a TS change to allow Unit 1 to operate at extended power uprate conditions. As part of this TS change, TVA requested approval to take credit for post-accident COP in order to provide adequate NPSH to the ECCS pumps.
TVA has reviewed the requested credit for COP against those aspects of the BFN licensing basis that may be affected by the proposed change, including rules and regulations, the Updated Final Safety Analysis Report (UFSAR), TSs, License Conditions, and licensing commitments. As previously discussed, NRC previously approved the use of COP to maintain adequate ECCS pump NPSH on BFN Units 2 and 3. The use of COP does not: invalidate TVA's compliance with 10 CFR 50.54(o), Appendix J to 10 CFR 50, 10 CFR 50.46 and Appendix K to 10 CFR 50. The use of COP is discussed in UFSAR Section 6.5.5. The approval of credit for post-accident COP is consistent with the NRC's Final Policy Statement on the Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities, is consistent with NRC staff positions, including Revision 3 of Regulatory Guide 1.82, and is part of the current licensing and design basis for Browns Ferry Units 2 and 3. The credit is supported by the BFN PRAs and the results satisfy the numerical targets contained in NRC Regulatory Guide 1.174. Alternatives which would preclude the need for the use of COP, such as the replacement of pumps or heat exchangers are not practical.
- 2. DEFENSE-IN-DEPTH Defense in depth philosophy is maintained by avoiding over reliance on specific features, human actions and assumptions to ensure plant safety. By preserving the function of the ECCS, multiple barriers of fuel cladding and primary containment are maintained. The ECCS functions are being preserved by the proposed plant design and operation. For a LOCA, reliance on COP is only necessary assuming low probability combinations of worst case assumptions governing heatup of the suppression pool.
RG 1.174 provides guidance for acceptable methods to asEess defense in depth principles. The following addressed the aspects of defense in depth that are potentially impacted by the requested change. Capability of Containment to Provide Containment Overpressure E1-6
The containment is designed to withstand conditions well in excess of those associated with a DBA. Pre-existing containment leakage is well below that which could defeat maintenance of required COP. At the end of 24 hours, 2 percent leakage results in an approximate 0.3 psi decrease in the 3.4 psig available containment pressure compared with no leakage. The containment is equipped with automatic containment isolation which is designed t:o single failure criteria. The COP available is the thermodynamic result of the event itself and does not depend on operator actions or systems other than the containment.
- Excess Containment Cooling Capability Long-term suppression pool temperature in design basis events is determined crediting only two of the four RHR pumps and heat exchangers. Emergency Operating Instructions (EOIs) dictate using all available RHFL pumps for suppression pool cooling. Single failures such as loss of a power supply or failure of containment cooling valves, failure of a service water pump or RHR heat exchanger valves can disable one or two RHR pumps for containment cooling. If no such single failure is assumed in the long term analysis (>10 minutes) then suppression pool temperature remains below 166.40 F with four RHR pumps or 1750 F with any three pumps and positive NPSH margin would be maintained long term without COP. These analyses were performed using the same conservative assumptions for input parameters as the licensing basis analysis. Core Spray pumps require credit for COP above 175.8 0 F. The RHR pumps do not require COP at the peak pool temperature of 187.4 0F.
The likelihood of failing any two RHR pumps is 8.2E-3. It can be concluded that defense in depth philosophy is preserved following the proposed change since multiple failures of safety related features would have to be postulated in order to impact ECCS functions. Credit for COP does not rely upon new operator actions or changes to the accident analysis methodologies. E1-7
- 3. SAFETY MARGINS Analyses for design basis events are performed with established margins added to important parameters to account for uncertainty. Significant parameter margins included in the NPSH analysis were examined and analysis results were obtained using more realistic values. This demonstrates that there is ample margin to ECCS pump functional. failure in design basis LOCA events without credit for COP. The following table provides the parameters of interest, the values used in the safety analysis and the associated realistic values.
LICENSING REALISTIC PARAMETER BASIS VALUE VALUE COMMENT Initial Power 102% Licensed 100% Licensed Probability of Thermal Power Thermal Power 102% power is 5.OE-- Decay Heat ANSI 5.1 (plus ANSI 5.1 (w/o Model 2a) 2a) Service Water 95 0 F 920 F Exceedance Temperature probability for 92 0 F is less than 6.OE-2 Initial 95-F (TS 920 F Exceedance Suppression maximum) probability Pool for 92 0 F is Temperature 8.25E-2 Heat Exchanger 223 BTU/Hr-0 F 241 BTU/Hr-0 F Based on K Value realistic fouling factor of O.C020 vs 0.0025 and maximum number of tubes plugged (1.5%) 225 BTU/Hr-0 F 1.5% tube _ _ plugging only Initial 121,500 ft3 (TS 125,640 ft3 Nominal value Suppression minimum) Pool Volume Containment Assumes no Includes Heat sinks are Heat Sinks heat sinks realistic heat always present sinks but not normally credited E1-8
Sensitivity analyses were performed (with selected analyses verified), which are summarized in Table 1. The purpose of these analyses was to identify input parameter combinations where COP was not required (e.g., suppression pool temperature below 175.8 0 F). Sensitivity to RHR Service Water (RHRSW) Temperature Suppression pool temperature response was examined as a function of RHRSW temperature which is a seasonal variable. Figure 3 shows Suppression Pool temperature as a function of RHRSW temperature using both licensing basis input values and realistic values. These analyses show that COP is not required for RHRSW temperatures 700 F or below assuming all design basis inputs and 860 F using realistic inputs. The probability of exceeding 700 F is 3.97E-l and for 860 F is 1.40E-1.
- Realistic Values Suppression pool temperature for the DBA-LOCA was evaluated by altering the input parameters to reflect the realistic values given above. Defense in depth assumptions such as RHR pump availability were not changed. This evaluation shows that suppression pool temperature remains below that which COP is required (175.80 F). This is indicate[ as Case 4a in Table 1 and shows that credit for COP is not required when realistic input values are assumed.
- Margin in Manufacturers Curves for NPSH Required (NPSHR)
The licensing basis need for COP is based on the conservative assumption in NPSH calculations that the RHR and Core spray pumps will not perform their functicn at NPSH Available (NPSHa) values less than the manufacturers NPSHR. The values used were derived from manufacturers testing for each pump. Suction pressures were reduced with 3 percent reduction in total dynamic head (TDH) to establish minimum NPSH. At this value, the pumps will operate without degradation. BFN RHR pumps are Sulzer-Bingham model 18x24x28 CVIC. Assuming no credit for COP in the limiting short-term LOCA scenario, RHR pumps would be required to be operated for less than 10 minutes at 24.3 feet NPSHa (broken loop) versus 30 feet NPSHR or 25.2 feet: NPSHa (intact loop) versus 26 feet NPSHR. Negative NPSH margin of this magnitude for E1-9
short periods of time will not prevent the RHR pumps from performing long-term in the event. Additional NPSH testing was performed on a BFN RHR pump in 1976 and reported to NRC in Reference 4. In this test, the RHR pump was operated 10,000 GPM (design flow) at approximately 24 feet of NPSH without cavitation and as low as 16 feet without damage. This is compared to 26 feet assumed to be the NPSH limit for the short-term COP requirements for the intact loop at design flow. This demonstrated that the RHR pumps can be operated below the manufacturers curve for at least 10 minutes without damage. This data demonstrated that the RHR pumps have NPSH margin assuming COP is not available. Therefore, in the unlikely event that COP was lost in the short-term LOCA, the function of the RHR pumps would not be affected for the short- and long-term. By comparison to the RHR pumps, the Core Spray pumps would be challenged in the long-term scenario in the event that COP was lost. Core Spray pumps do not require COP in the short-term (Refer to Figure 1). BFN Core Spray pumps are Sulzer-Bingham model 12x16x14.5 CVDS. Assuming no credit for COP, the Core Spray pumps used for long-term core cooling (>10 minutes) would be expected to operate between 27 feet and 22.6 feet of NPSH verses 27 feet used in NPSH calculations for approximately 13.4 hours as Suppression Pool temperature peaks above 175.8 0 F during the LOCA. In the unlikely event they become degraded, there is a reasonable likelihood that the affected pumps would. still be able to function. In addition, only one of the two Core Spray loops is required tD be operated for adequate core cooling and the non-operating Core Spray loop would be available to operators if the operating loop failed after some time period. RHR pumps would also be available in the LPCI mode for core cooling in conjunction with their suppression pool cooling function should all Core Spray pumps become unavailable. COP is not required for RHR pumps in the long-term scenario. Therefore in the unlikely event that COP was lost in the long-term LOCA, the decay heat removal and core cooling functions would be maintained. It can be concluded that safety margins are preserved following the proposed change. Sensitivity analyses show that COP is not required if realistic inputs are utilized without any changes to the accident analysis methodologies. El-10
- 4. RISK ASSESSMENT TVA has evaluated the risk impact of utilizing COP to satisfy the NPSH requirements for RHR and Core Spray pumps to mitigate the consequences of a DBA LOCA. The risk assessment evaluation used the current BFN Unit 1 PRA internal events (including internal floods) model. The evaluation is provided as to this letter. The steps taken to perform this risk assessment evaluation were:
- 1. Evaluate sensitivities to the DBA LOCA accident calculations to determine under what conditions credit for COP is necessary to satisfy low pressure ECCS pump NPSH requirements;
- 2. Revise all large LOCA accident sequence event trees to make low pressure ECCS pumps dependent upon containment isolation when other plant pre-conditiDns exist (i.e., Service Water initial high temperature, Suppression Pool initial high temperature);
- 3. Modify the existing Containment Isolation System fault tree to include the probability of pre-existing containment leakage;
- 4. Quantify the modified PRA models and determine the change in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF); and
- 5. Perform modeling sensitivity studies and a parametric uncertainty analysis to assess the variability of the results.
Crediting COP resulted in a small increase in CDF and LERF of 1.53 E-9/yr. This small increase was well below the guidelines provided in RG 1.174. ATWS, SBO, and Appendix R are highly unlikely event scenarios which are defined by failure of multiple features. Failure assumptions in these events are beyond design basis. Additional failures such as loss of containment integrity need not be assumed. Deterministic analyses have shown that COP will be available as thermodynamic result of the event itself provided that containment integrity is maintained. This is acceptable given the low probability of the events. El-li
- 5. MONITORING Performance monitoring is performed for parameters important to ECCS NPSH analyses to ensure that assumptions remain valid and that corrective actions are initiated for deficiencies.
Containment Integrity Monitoring During normal power operations, the containment is inerted with nitrogen and maintained at greater than or equal to 1.1 psi positive pressure relative to the suppression chamber in accordance with TS 3.6.2.6. Technical Requirements Manual 3.6.5 limits nitrogen makeup to 542 scfh and is determined every 24 hours. This would identify any pre-existing leak in the drywell portion of containment. 10 CFR 50.54(o) and 10 CER Part 50 Appendix J require leak rate testing of the containment structure, penetrations and isolation valves at the maximum predicted LOCA pressure. Containment leak rate testing tests containment penetrations and limits total leakage to < 0.6La. La is two weight percent per day at 50.6 PSIG. Available containment pressure is calculated assuming two weight percent per day throughout the event which is conservative. 10 CFR 50.55a(ii)B requires periodic in-service examination of the containment structure in accordance with the American Society of Mechanical Engineers Code.
- NPSH Monitoring The EOIs include precautionary statements warning the operator that continuous operation of the low pressure injection system pumps with inadequate NPSH may result in pump damage or pump inoperability and that reducing containment pressure may affect pump NPSH. The operator is instructed to monitor NPSH using an NPSH limit curve, showing pump flow versus suppression pool temperature for various suppression pool pressures. The EOIs also list additional indications of inadequate NPSH. Operators are trained on these procedures as part of their periodic re-qualification program.
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RG 1.174 CONCLUSION The use of COP to ensure an adequate NPSH for ECCS pumps during a limited time after a design basis accident is consistent with NRC staff positions, including Revision 3 of RG 1.82, and is part of the current licensing and design basis for BFN 'Units 2 and 3. Alternatives which would preclude the need for the use of COP, such as the replacement of pumps or heat exchangers are not practical. Deterministic evaluations and analyses, which were performed in accordance with regulatory requirements, have demonstrated that an adequate level of protection is maintained. Even though the use of COP was requested on a deterministic basis, a risk-informed assessment was performed in accordance with the guidelines contained in RG 1.174, Revision 1. In summary, a defense-in-depth philosophy is maintained by avoiding an over reliance on specific features, human actions, or assumptions to ensure safety. Safety margins are maintained since realistic analyses demonstrate that adequate NPSH exists for the ECCS pumps without crediting COP. Crediting COP results in a small increase in CDF anc. LERF of 1.53xlO 9 /yr. This small increase is well below the guidelines provided in RG 1.174 (10-6/yr for CDF and 10-7/yr for LERF). The integrity of the primary containment and the associated primary containment isolation valves are monitored. using diverse performance measurement strategies that ensure the detection and correction of adverse conditions. REFERENCES
- 1. TVA letter, T.E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - License Amendment Regarding Use of Containment Overpressure for Emergency Core Cooling System (ECCS) Pump Net Positive Suction Head (NPSH) Analyses,"
September 4, 1998.
- 2. TVA letter, T.E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Response to Request for Additional Information (RAI)
Relating to Units 2 and 3 License Amendment Regarding Use of Containment Overpressure for Emergency Core Cooling System (ECCS) Pump Net Positive Suction Head (NPSH) Analyses," November 25, 1998. E1-13
- 3. NRC letter, W.O. Long to TVA, "Browns Ferry Nuclear Plants, Units 2 and 3 - Issuance of Amendments Regarding Crediting of Containment Overpressure for Net Positive Sucticn Head Calculations for Emergenc y Core Cooling Pumps (TAC Nos. MA3492 and MA3493)," September 3, 1999.
- 4. TVA letter, J.E. Gilleland to NRC, "Browns Ferry Nuclear Plant Unit 3 - Reportable Deficiency - Potential for RHR Pump Operation in Excess of Design Runout - IE Control No.
H01172F2," May 21, 1976. E1-14
TABLE 1 SENSITIVITY ANALYSES OF VARIOUS REALISTIC INPUT PARAMETERS DURING THE DESIGN BASIS ACCIDENT LOSS OF COOLANT ACCIDENT LONG-TERM PHASE
.E 0 ~ 0 - E b0 E u CIO E L_
0 a.i 9i Ec ~.
- - - - - -6 - -
Base EPU Licensing Calculation - 102% ANSI 5.1 95 95 2 Full 2 2 4000 223 2 Minimum Yes No 187.3 Yes Case* DBA LOCA EPU wu2a dFgn Case 1* INo Single Failure E_. ANSI5.1 95 95 Full 223 Yes_ INo I 1 166.4 1 No I___ _ _ __ _ __ EPU JwI~c Case la* 3 Pumps inSPC 102% IANSI 5.1 95 95 223 Yes No 175.0 No EPU I wf2a desgn Case 2 DBA Calculation but Inihial SW Temperature = 85°F 102% ANSI 5.1 95 Full L. L 4000 1223 2 .... Minimum . Yes No
--. AAA 182.0 Yes EPU w/2a desion Case 2a DBA Calculation but Initial SW 102%-o ANSI 5.1 95 2 l Full l 2 2 4000 223 2 Minimum Yes I No 177.6 Yes Case 2b*
Temperature = 75°F DBA Calculation but Initial SW EPU 102% w2ou ANSI 5.1 5 design +/- I II K Temperature = 70°F EPU w/2o - 2I designFull 2 I 2 I4000 223 2 Minimum Yes No 175.9 l No Case 2c DBA Calculation but Initial SW 102% ANSI 5.1 2 Full 2 2 4000 223 2 Minimum Yes No 174.3 No Temperature = 65-F EPU w/2or desian 1-1-5 1. Case 3 DBA Calculation but Initial SP 2 Full 2 2 4000 223 2 Minimum Yes No 183.8 Yes Temoerature = 850F design Case 4 100%initial Powe Minimum SP 2 Full 2 2 4000 2 Minimum I I Yes No I .__A 177.0 I . . Yes Level, and No Heat Sink Credit desirn Case 4a 100%InitialPower Nominal SP - Full 2 2 4000 . 2 m Yes m 174.7 No Level, and Heat Sink Credit design _ Case 4b* 100% Initial Power, Minimum SP 2 Full 2 2 4000 lMinimum Yes m 178.9 Yes Level, and Heat Sink Credit desian Case 4c* 100% Initial Power, Minimum SP 92 1 2 Full 2 2 4000 Minimum Yes No Heat Sink Credit, and SW WLMvei, 1 Design Temp. that results inPeak SP Temp. equal toAess than 176 0F
- - Case verified by formal analyses. ** - This value is acceptable for demonstrating sensitivity analysis results.
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FIGURE 1 NET POSITIVE SUCTION HEAD REQUIREMENTS FOR DESIGN BASIS LOSS OF COOLANT ACCIDENT - SHORT TERM 160 -20 19 150 18 140- 17 16 LI: 30 l15 5 120 j GOP
'14 110 - -SP TEMPERATURE Deg.F i I / /SP PRESSURE psia
/ -BKN LOOP RHR PRESS REQ psia 12 100 , - LPCI LOOP RHR PRESS REQ psia
-/-CS PUMPS PRESS REQ psia 11
-ATMOSPHERIC PRESSURE (psia) 90- 1 400 1 10 ion0 200 300 400 500 600 SECONDS E1-16
FIGURE 2 NET POSITIVE SUCTION HEAD REQUIREMENTS FOR DESIGN BASIS LOSS OF COOLANT ACCIDENT - LONG TERM 200 24 190 22 180 170 20 160 SP TEMPERATURE Deg.F ASP-b VKR-S:SUHE psia 10 IL lATMOSPHEREIC PRESS psia
--- RHR PUMP COP REQUIRED 140G-CS PUMP COP REQUIRED - 16 140 130 14 120 - ', ~ ~ ~ ~ ~ ~ ~ ~ ,1 1202 1110 100 10 0.00 2.00 4.00 6.00 8.00 10.00 12.00 14.00 16.00 HOURS E1-17
FIGURE 3 CONTAINMENT OVERPRESSURE SENSITIVITY TO RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) TEMPERATURE AND NET POSITIVE SUCTION HEAD SENSITIVITY TO RHRSW TEMPERATURE 190 I............................................................................... I....................................................................................................................................................................... 0.8 100 0.7
- LICENSING BASIS VALUES 186 - UREALISTIC VALUES CASE
- COP CREDIT LIMIT 0.6 184- z5 N CURVE FIT U.
-0,-
-WI- RHRSW TEMP PROBABILITY
- 0 0.5 a 182
'u 182 CASE 2 180 - 0.4 3 m
0.W \ / CASE 4b 0 w 178 - a SE 2a/ C- 0.3 co COP REQUIRED 176-COP NOT REQUIRED CASE 4c 0.2
~~ CASE 2b 174 -
~ CASE2c 0.1 172 170l
,. 0 50 55 60 65 70 75 80 85 90 95 100 RH RSW TEMPERAT URE Deg F E1-18
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY (TVA) BROWNS FERRY NUCLEAR PLANT UNIT 1 DETAILED CHRONOLOGY OF CORRESPONDENCE RELATED TO THE PREVIOUS APPROVAL OF NPSH FOR PRE-UPRATE CONDITIONS Following a postulated Loss of Coolant Accident (LOCA), the Residual Heat Removal (RHR) and Low Pressure Core Spray (LPCS) pumps operate to provide the required core and containment cooling. The use of containment overpressure to maintain adequate pump net positive suction head (NPSH) is required to ensure essential pump operation. The limiting NPSH conditions occur during either short-term or long-term post-LOCA pump operation depending on the total pump flow rates, debris loading on the suction strainers, and suppression pool temperature. As chronicled below, credit for containment overpressure (up to 3 psi short-term for the RHR pumps and 1 psi long-term for the LPCS pumps) was extensively reviewed and subsequently approved by NRC. On May 6, 1996, the NRC issued NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors," (Reference 1). That bulletin was issued following events at several operating reactors where clogging of containment cooling pump suction strainers adversely impacted pump operation. As a result the NRC requested licensees to take actions to protect Emergency Core Coo-ing System (ECCS) pump strainers from clogging, and ensure pumps have adequate NPSH to fulfill their function. By letter dated July 25, 1997, TVA responded to NRC Bulletin 96-03 (Reference 2). That letter outlined its proposed actions for resolution of NRC'S concerns for loss of ECCS following a Design Basis LOCA. To ensure adequate ECCS NPSH during and following accidents, TVA stated it planned to install larger capacity passive strainers and credit for a containment pressure in excess of atmosphere for a short period of time. TVA indicated that it would implement appropriate modifications to BFN Unit 1 prior to its restart. By letter dated August 25, 1997, TVA supplemented its July 25, 1997 response to NRC Bulletin 96-03 (Reference 3). TVA indicated that pursuant to discussions with the NRC staff, it was preparing a license amendment request to allow crediting
containment overpressure to ensure adequate ECCS pump NJSH during and following accidents. TVA also indicated tha,: the NRC had previously approved crediting containment overpressure for ensuring ECCS NPSH as part of the BFN original licensing basis. By letter dated October 7, 1997, the NRC issued Generic Letter (GL) 97-04, "Assurance of Sufficient Net Positive Suction Head (NPSH) for Emergency Core Cooling and Containment Heat Removal Pumps," (Reference 4). GL 97-04 requested that licensees review their design basis analyses used to determine the available NPSH for the ECCS and containment heat removal pumps that take suction from the containment following a design basis LOCA, and to provide specific information used therein. GL 97-04 requested, in part, that licerLsees specify whether credit is taken in their ECCS NPSH analyses for containment overpressure, and if so, identify the amount. of overpressure needed and the minimum overpressure available. TVA provided its 90-day response to GL 97-04 with a letter dated January 5, 1998 (Reference 5). In that letter, TVA ind cated that BFN Unit 1 was at that time shut down and defueled. Accordingly, TVA indicated that it would evaluate BFN Unit 1 ECCS and containment cooling pump NPSH prior to its restart. TVA summarized actions taken and planned in response to NRC Bulletin 96-03, provided a description of containment debris analyses performed for BFN Units 2 and 3, and reiterated its intent to submit a license amendment request to support credit for containment overpressure. That submittal also provided required and available BFN Units 2 and 3 ECCS pump NPSH, and assumed a containment overpressure of 2 psig for the limiting case. By letter dated June 11, 1998, the NRC closed GL 97-04 for BFN Units 2 and 3 (Reference 6). On September 4, 1998, TVA submitted a request to change the BFN Units 2 and 3 license basis to permit the use of available containment overpressure for ECCS pump NPSH (Reference 7). On November 25, 1998, in response to a verbal NRC request for additional information, TVA provided (Reference 8):
- The short- and long-term NPSH calculations for the RHR and LPCS pumps;
- Supporting information for these calculations; E2-2
- An explanation as to how the analysis at pre-power uprate conditions bounds the uprated conditions;
- A rationale for why the analysis assumed a desicn flow rate for the LPCS pumps when one RHR pump is in a runout condition;
- A discussion of the requested overpressure value; and
- Graphs showing the NPSEH required for the RHR and. LPCS pumps versus time and available containment pressure.
On September 3, 1999, NRC approved the use of containment overpressure to maintain adequate ECCS pump NPSH on BFN Units 2 and 3 (Reference 9). The NRC approved 3 psi for the short-term and 1 psi for the long-term period from 5,500 to 35,000 seconds (approximately 92 minutes to 91.7 hours). By letter dated November 15, 1999, the NRC closed Bulletin 96-03 for BFN Units 2 and 3 (Reference 10). That closure acknowledged actions taken by TVA to address the potential for ECCS suction strainer clogging, and acknowledged closure of the containment overpressure issue for BFN Units 2 and 3 with issuance of corresponding amendments on September 3, 1999. By letter dated May 6, 2004 (Reference 11), TVA submitted its response to NRC Generic Letter 97-04 for BFN Unit 1. In its response, TVA provided a description of the BFN Unit 1 ECCS pump NPSH analyses performed, key assumptions used, and the ECCS NPSH requirements assuming operations at Extended Power Uprat:e (EPU) conditions, and modification to the ECCS suction strainers in response to NRC Bulletin 96-03. As stated in Reference 11, BFN Unit 1 requires a credit of 3 psig of containment overpressure to ensure adequate NPSH. On June 28, 2004 (Reference 12), TVA requested a TS change to allow Unit 1 to operate at extended power uprate conditions. As part of this TS change, TVA requested approval to take credit for containment overpressure in order to provide adequate NPSH to the ECCS pumps. Specifically, TVA requested approval to credit 3 psi containment overpressure for the RHR pumps for the first 10 minutes following a LOCA (short-term requirement) and 3 psi containment overpressure for the LPCS pumps from approximately 4,100 to 52,300 seconds (about 13.4 hours) following a LOCA (long-term requirement). E2-3
REFERENCES
- 1. NRC letter, NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors," May 6, 1996.
- 2. TVA letter, T.E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - NRC Bulletin 96-03, Potential Plugging of Emergency Core Cooling Suction (ECCS) Strainers by Debris in Boiling Water Reactors (TAC Nos. ]Ml96135, M96136, M96137),"
July 25, 1997.
- 3. TVA letter, T.E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) Unit 2 - NRC Bulletin 96-03, Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling Water Reactors (TAC No. M96136)," August 25, 1997.
- 4. NRC Letter, J. W. Roe to All Licensees, NRC Generic Letter 97-04, "Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps,"
October 7, 1997.
- 5. TVA letter, T. E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Response To NRC Generic Letter (GL) 97-04, Assurance of Sufficient Net Positive Suction Head (NPSH) for Emergency Core Cooling and Containment Heat Removal Pumps," January 5, 1998.
- 6. NRC letter, A. W. De Agazio to 0. J. Zeringue, "Browns Ferry Nuclear Plant, Units 2 And 3-Completion of Licensing Action For Generic Letter 97-04 (TAC NOS. M99964 AND M99965),"
June 11, 1998.
- 7. TVA letter, T.E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - License Amendment Regarding Use of Containment Overpressure :-or Emergency Core Cooling System (ECCS) Pump Net Positive Suction Head (NPSH) Analyses,"
September 4, 1998. E24
- 8. TVA letter, T.E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Response to Request for Additional Information (RAI)
Relating to Units 2 and 3 License Amendment Regarding Use of Containment Overpressure for Emergency Core Cooling System (ECCS) Pump Net Positive Suction Head (NPSH) Analyses," November 25, 1998.
- 9. NRC letter, W.O. Long to TVA, "Browns Ferry Nuclear Plants, Units 2 and 3 - Issuance Df Amendments Regarding Crediting of Containment Overpressure for Net Positive Suction Head Calculations for Emergency Core Cooling Pumps (TAC Nos. MA3492 and MA3493)," September 3, 1999.
- 10. NRC letter, W. 0. Long to J. A. Scalice, "Browns Ferry Nuclear Plant Units 2 and 3, Completion of Licensing Actions for Bulletin 96-06, 'Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors,' Dated May 6, 1996 (TAC NOS. M96135, M96136 and M96137)," November 15, 1999.
- 11. TVA Letter, "Browns Ferry Nuclear Plant (BFN) Unit 1 -
Response to NRC Generic Letter (GL) 97-04, Assurance of Sufficient Net Positive Suction Head (NPSH) For Emergency Core Cooling And Containment Heat Removal Pumps," May 6, 2004.
- 12. TVA letter, "Browns Ferry Nuclear Plant (BFN) - Unit 1-Proposed Technical Specif-cations (TS) Change TS-431 -
Request for License Amendment - Extended Power Uprate (EPU) Operation," dated June 28, 2004. E2-5
ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY (TVA) BROWNS FERRY N4UCLEAR PLANT UNIT 1 DETAILED DESCRIPTION OF PLANT SYSTEMS RELATED TO THE NPSH ANALYSIS Each BFN unit employs a pressure suppression containment system which houses the reactor vessel, the reactor coolant recirculation loops, and other branch connections of the Reactor Primary System. The pressure suppression system consists of a drywell, a pressure suppression chamber (alternatively referred to as the torus or wetwell) which stores a large volume of water, a connecting vent system between the drywell and the suppression chamber, isolation valves, containment cooling systems, equipment for establishing and maintaining a pressure differential between the drywell and pressure suppression chamber, and other service equipment. The drywell is a steel pressure vessel with a spherical lower portion 67 feet in diameter, ELnd a cylindrical upper portion 38 feet 6 inches in diameter. The overall height is approximately 115 feet. In the event of a process system piping failure within the drywell, reactor water and steam would be released into the drywell air space. The resulting increased drywell pressure would then force a mixture of air, steam, and water through the vents into the pool of water which is stored in the suppression chamber. The steam would condense rapidly and completely in the suppression chamber, resulting in rapid pressure reduction in the drywell. Air that is transferred to the suppression chamber pressurizes the chamber and is subsequently vented to the drywell to equalize the pressure between the two vessels. The pressure suppression chamber is a steal pressure vessel in the shape of a torus below and. encircling the drywell, with a centerline diameter of approximately 111 feet and a cross-sectional diameter of 31 feet. Large vent pipes form a connection between the drywell and the pressure suppression chamber. A total of eight circular vent pipes are provided, each having a diameter of 6.75 feet. A 30-inch diameter Emergency Core Cooling System (ECCS) suction header circumscribes the suppression chamber. Four 30-inch diameter tees are used to connect the suction header to the
suppression chamber. Four strainers on connecting lines between the suction header and the suppression chamber have beern provided. The suction lines from the Residual Heat Removal (RHR), High Pressure Coolant Injection (HPCI), Low Pressure Core Spray (LPCS), and Reactor Core Isolation Cooling (RCIC) systems are supplied from this header. The four strainers are not individually associated with separate pump suctions but direct suppression pool water to the common ECCS ring header. Therefore interaction between operating pumps are considered when determining suction losses. The normal suction path for the HPCI and RCIC system pumps is the Condensate Storage Tank. Figure 1 provides a general overview of the primary containment. As shown in Figure 2, the BFN ECCS consists of the following:
- HPCI;
- Automatic Depressurization System (ADS);
- LPCS; and
- Low Pressure Coolant Injection (LPCI), which is an operating mode of RHR.
The ECCS subsystems are designed to limit clad temperature over the complete spectrum of possible break sizes in the nuclear system process barrier, inclucing the design basis break. The design basis break is defined as the complete and sudden circumferential rupture of the largest pipe connected to the reactor vessel (i.e., one of the recirculation loop pipes) with displacement of the ends so that blowdown occurs from both ends. The low-pressure ECCS consists of LPCS and LPCI. The LPCS consists of two independent loops. Each loop consists of two pumps, a spray sparger inside the core shroud and above the core, piping and valves to convey water from the pressure suppression pool to the sparger, and the associated controls and instrumentation. When the system is actuated, water is taken from the pressure suppression pool. Flow then passes through a normally open motor-operated valve in the suction line to each 50 percent capacity pump. E3-2
The RHR System is designed for five modes of operation (i.e., shutdown cooling; containment spray and suppression pool cooling; LPCI; standby cooling; and supplemental fuel pool cooling). During LPCI operation, the four RHR pumps take suction from the pressure suppression pool and discharge to the reactor vessel into the core region through both of the recirculation loops. Two pumps discharge to each recirculation loop. An important consideration in the operation of the LPCS and RHR pumps is the available net positive suction head (NPSH). Adequate available NPSH is important in ensuring that the pump will deliver the flow assumed in the safety analyses at the expected discharge pressure. In order to ensure acceptable flow and discharge pressure, the available NPSH must be equal to or greater than the required NPSH. The required NPSH is a function of the pump design and is determined by the pump vendor. The available NPSH is calculated from the equation: Available NPSH = hatm + hstatic - hioss - hvapor where: hat. = head on the surface of the suppression pool hstatic = the head due to the difference in elevation between the suppression pool surface and the centerline of the pump suction his. = the head loss due to fluid friction, fittings in the flow path from the suppression pool to the pump, and the suction strainers which prevent ingestion of debris into the pumps hvapor = head due to the vapor pressure of the suppression pool water at the suppression pool water temperature E3-3
The increase in power from extended power uprate results in increased decay heat, and a subsequent increase in the suppression pool temperature following the design basis Loss of Coolant Accident. The increased water temperature reduces the available NPSH of the RHR pumps and the LPCS pumps since the vapor pressure of the suppression pool water (or hvapor) increases. The reduction in available NPSH is mitigated, where necessary, by crediting the containment accident pressure, that is, by increasing hatm. E3-4
FIGURE 1 GENERAL CONTAINMENT LAYOUT
¶ REACTOR PRESSURE VESSEL DRYWELL VENT CRADLE E3-5
FIGURE 2 LAYOUT OF THE EMERGENCY CORE COOLING SYSTEM
°2 T or p Dip@ De r --------- 4I I TO lI I I I TO II iiC] TORUS I l I L - - --l- -
IR e bmm II II I ledSIXADS II I _INJC m VALE S I INJECTION INJECTION VALVE a LPC8 VALVE FCV-74-53 .)IINJECTION~ FCV-74-47 RECIEC ECURC PIw PUMP A E3-6
ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY (TVA) BROWNS FERRY NUCLEAR PLANT UNIT 1 EXTENDED POWER UPRATE CONTAINMENT OVERPRESSURE CREDIT RISK ASSESSMENT
I BFIN EPU Containment Overpressure (C:OP) Credit Risk Assessment Performed for: Tennessee Valley Authority Performed by: ERIN Engineering and Research, Inc. February 27, 2006
BFNEPUCOPProbabilisticRiskAssessment
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C13205034-924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment Table Of Contents Section Pane EXECUTIVE
SUMMARY
....................... ii
1.0 INTRODUCTION
.............. 1-1 1.1 Background ., ............ 1-1 1.2 Scope ............ 1-3 1.3 Definitions ............ 1-4 1.4 Acronyms...............................................................................................1-6 2.0 APPROACH .2-1 2.1 General Approach .2-1 2.2 Steps to Analysis .2-3 3.0 ANAYSIS .3-1 3.1 Assessment of DBA Calculations ................................ 3-1 3.2 Probability of Plant State 1 and Plant State 2 ................................ 3-3 3.3 Pre-Existing Containment F-ailure Probability ................................ 3-5 3.4 Modifications to BFN Unit 11PRA Models ................................ 3-6 3.5 Assessment of Large-Late Releases .......... ...................... 3-8 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 ................................ i 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 River Water and SP Water Temperature Variation Appendix D Large-Late Release Impact Appendix E Revised Event Trees Appendix F Fault Trees i C1320503-5924 - 2/27/2006
BEN EPU COPProbabilisticRisk Assessment EXECUTIVE
SUMMARY
The report documents the risk impact of utilizing containment accident pressure (containment overpressure) to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps during DBA LOCAs.
The risk assessment evaluation use600La.
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 UICOP2-9 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.
3-8 3-8 C1320503-5924 - 2/27/2006
BFNEPUCOP ProbabilisticRiskAssessment Table 3-1
SUMMARY
OF COP DETERMINISTIC CALCULATIONS Ch.
- 0 0 0 0 0 0.0co ~COP co
. a... E Peak SP Credit Case Case Description 0 o zO0 D ~zO _ .E wIL..J uz Temp (F) Required Base Case EPI Licensing Cakulation - 102% IANS5.1 95 95 2 Full 2 2 4000 223 2 IMinimum Yes No 187.3 Yes
_ DBA LOCA I PL w/2cr design Case 1 No Single Failure 1 102% ANSI 5.1 9 95 Full 4000 223 Yes INo 166.4 No _ _ _ _ __u _ _ _ Iwr_ design Case la j3 Pumps in SPC 102% ANSI 5.1 95 95 Full 4000 223 YesI No
. . I . . I ._
175.0 I . . No EPU w/2a design Case 2 DBA Calculation but SW 102% ANSI 5.1 95 2 Full 4000 223 num YesI No 182.0 Yes Temperature = 85F EPU w/2a design Case 2a DBA Calculation but SW 102% ANSI 5.1 95 2 Full 2 Temperature = 75F EPU w/2G design 2 4000 223 I 2 Minimum IYes No 177.6 Yes Case 2b DBA Calculation but SW 102% ANSI 5.1 95 2 Full 2 2 4000 223 2 Minimum Yes No 175.9 No Temperature = 70F EPU w/2a design . - .. . .. . Case 2c DBA Calculation but SW 102% ANSI 5.1 2 Full 2 2 140001 223 2 lMinimumnYes INo 174.3 No Temperature- 65F EPU wI2a design Case 3 DBA Calculation but SP 2 Full 2 2 j4000 223 2 Minimum Yes No 183.8 Yes Temperature = 85F design Case 4 1100%Initial Power, Minimum 2 Full 2 1 2 14000 2 Yes I No 177.0 Yes SP LavW, and, VNUI .eaL SAr,k Minimum I Credit Case 4a I 100%1nitial Power, Nominal SP Level, and Heat Sink 2 Full design 2 I j 2 4000 174.7X No Credit 3-9 C1320503-6924 - 2/27/2006
BFN EPU COPProbabilisticRiskAssessment Table 3-1
SUMMARY
OF COP DETERMINISTIC CALCULATIONS 7 i - - r = =--- .2 = - -, _ to C
- a. a.
- 2) X
- a. .P if
- 0) U) E
- 00. ai. LL f c Id 2 -8 E
12 a. 0. C a) a, VaD zo C') E Co U E)
'a a) a) a.
'a f2 a
0~ to a) 0 C) ID COP C0e a) I- .0 :r Co Co Peak SP Credit co Case Case Description 0) z 0 z 0 0 w -j Temp (F) Required Case 4b 4 q- 4-4 - I- * - 100% Initial Power, Minimum 92 2 Full 2 2 4000 Minimum Yes 178.9 Yes SP Level, and Heat Sink design Credit Case 4c I nitial Power, Minimum 92 2 Full 2 2 4000 2 IMinimum I Yes 175.8 I No SP Level, Heat Sink Credit design and SW Temp. that results in Peak SP Temp. equal tWless
. than 176F
- I 3-10 C1320503-6924 - 2127r2006
BFNEPUCOPProbabilisticRisk Assessment Section 4 RESULTS 4.1 QUANTITATIVE RESULTS The results of the base quantification of this risk assessment for the 35 La case are as follows:
. deltaCDF: 1.42E-9/yr
. deltaLERF: 1.42E-9/yr As discussed in Section 3, the additional CDF contributions created by this model manipulation are also all LERF release sequences (i.e., deltaCDF equals deltaLERF).
These very low results are expected and are well within the RG 1.174 guidelines (refer to Figures 2-1 and 2-2) for "very small" risk impact. If greater detail was included to address some of the conservative assumptive assumptions in this risk assessment (e.g., 2 sigma decay heat assumed with a probability of 1.0 given 102% EPU power exists; refer to Section 3.2), the deltaCDF and deltaLERF would be even lower. 4.2 UNCERTAINTY ANALYSIS To provide additional information for the decision making process, the risk assessment provided here is supplemented by parametric uncertainty analysis and quantitative and qualitative sensitivity studies to assess the sensitivity of the calculated risk results. Uncertainty is categorized here into the following three types, consistent with PRA industry literature:
- Parametric
- Modeling 4-1 C1320503.6924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment
- Completeness Parametric uncertainties are those related to the values of the fundamental parameters of the PRA model, such as equipment failure rates, initiating event frequencies, and human error probabilities. Typical of standard industry practices, the parametric uncertainty aspect is assessed here by performing a Monte Carlo parametric uncertainty propagation analysis. Probability distributions are assigned to each parameter value, and a Monte Carlo sampling code is used to sample each parameter and propagate the parametric distributions through to the final results. The parametric uncertainty analysis and associated results are discussed further below.
Modeling uncertainty is focused on the structure and assumptions inherent in the risk model. The structure of mathematical models used to represent scenarios and phenomena of interest is a source of uncertainty, due to the fact that models are a simplified representation of a real-world system. Model uncertainty is addressed here by the identification and quantification of focused sensitivity studies. The model uncertainty analysis and associated results are discussed further below. Completeness uncertainty is primarily concerned with scope limitations. Scope limitations are addressed here by the qualitative assessment of the impact on the conclusions if external events and shutdown risk contributors are also considered. The completeness uncertainty analysis is discussed further below. 4.2.1 Parametric Uncertainty Analysis The parametric uncertainty analysis for this risk assessment was performed using the RISKMAN computer program to calculate probability distributions and determine the uncertainty in the accident frequency estimate. 4-2 C1 320503-5924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment RISKMAN has three analysis modules: Data Analysis Module, System Analysis Module, and Event Tree Analysis Module. Appropriate probability distributions for each uncertain parameter in the analysis is determined and included in the Data Module. The System Module combines the individual failure rates, maintenance, and common cause parameters into the split fraction frequencies that will be used by the Event Tree Module. A Monte Carlo routine is used with the complete distributions to calculate the split fraction frequencies. Event trees are quantified and linked together in the Event Module. The important sequences from the results of the Event Tree Module are used in another Monte Carlo sampling step to propagate the split fraction uncertainties and obtain the uncertainties in the overall results. The descriptive statistics calculated by RISKMAN for the total core damage frequency of the plant caused by internal events include:
- Mean of the sample
- Variance of the sample
- 5th, 50th, and 95th percentiles of the sample The parametric uncertainty associated with delta core damage frequency calculated in this assessment is presented as a comparison of the RISKMAN calculated CDF uncertainty statistics for the two cases (i.e., the Unit 1 base EPU PRA and the EPU COP Credit base case quantification). The results are shown in Table 4-1. Table 4-1 summarizes the CDF uncertainty distribution statistics for the Unit 1 PRA and for the COP credit base quantification.
As can be seen from the parametric uncertainty results summarized in Table 4-1, even when considering the parametric uncertainty the risk impact is small. The statistics show that CDF has not changed while the distribution of CDF for the COP study has narrowed slightly: the 5%ile increased slightly while the 95%ile decreased slightly. 4-3 C1320503-6924 - 2/27/2006
BFN EPUCOPProbabilisticRisk Assessment It should be cautioned that this distribution is developed via Monte Carlo (random) sampling, and as such it is dependent upon the number of samples and the initial numerical seed values of the sampling routine. Neither the initial seeds nor the number of samples used for the model of record are known. Consequently, some variation from the base model statistics is expected. Taking these cautions into consideration, a comparison of the distributions by percentiles shows little if any change. 4.2.2 Modeling Uncertainty Analysis As stated previously, modeling uncertainty is concerned with the sensitivity of the results due to uncertainties in the structure and assumptions in the logic model. Modeling uncertainty has not been explicitly treated in many PRAs, and is still an evolving area of analysis. The PRA industry is currently investigating methods for performing modeling uncertainty analysis. EPRI has developed a guideline for modeling uncertainty that is still in draft form and undergoing pilot testing. The EPRI approach that is currently being tested takes the rational approach of identifying key sources of modeling uncertainty and then performing appropriate sensitivity calculations. This approach is taken here. The modeling issues selected here for assessment are those related to the risk assessment of the containment overpressure credit. This assessment does not involve investigating modeling uncertainty with regard to the overall BFN PRA. The modeling issues identified for sensitivity analysis are:
- Pre-existing containment leakage size and associated probability
- Calculation of containment isolation system failure
- Assessment of power and water temperature pre-conditions
- Number of RHR pumps and heat exchangers in SPC 4-4 C13205036924 - 2/27/2006
BFN EPU COPProbabilisticRisk Assessment Pre-Existing Containment Leakage Size/Probability The base case analysis assumes a pre-existing containment leakage pathway leakage size of 35La that would result in defeat of the necessary containment overpressure credit during a DBA LOCA. The following two modeling sensitivity cases are identified to assess the variability of the risk results to the assumed pre-existing containment leakage size:
- A smaller, even more conservative, pre-existing leak size of 20La is assumed in this sensitivity to result in defeat of the necessary COP credit.
From EPRI 1009325, the probability of a pre-existing 20La containment leakage pathway is 1.88E-03.
- A larger pre-existing leak size of lOOLa, consistent with the EPRI 1009325 recommended assumption for a "large" leak, is used in this sensitivity to defeat the necessary COP credit. From EPRI 1009325, the probability of a pre-existing 10OLa containment leakage pathway is 2.47E-04.
Calculation of Containment Isolation Sys tem 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 all the split fraction solutions for the containment isolation system fault tree. 4-5 C1 320503-6924 - 2/27/2006
BFN EPUCOPProbabilisticRisk Assessment Assessment of Power and Water Temperature 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 (e.g., river water temperature greater than 860 F) exist with a probability of 1.0. Number of RHR pumps and heat exchanaers 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 is approximately 8.2E-3. The likelihood of an unisolated containment is approximately 1.4E-3 and the likelihood of other necessary extreme plant conditions (e.g., high river temperature, high reactor power) existing at the time of the LLOCA is approximately 0.14. As such, the base quantification results in an approximate 1.6E-6 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 negligible, as shown by the sensitivities discussed here. As shown in Table 3-1, even with design basis 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 negligible compared to the 2 RHR pump case explicitly modeled in this analysis. 4-6 c1320503.6924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment An estimate of the deltaCDF risk contribution for the scenario with 3 RHR pumps in SPC operation can be approximated as Follows:
- Sum of BFN PRA Large LOCA initiator frequencies: 3.1OE-5/yr
- Likelihood of failure of 1 RHR pump or 1 RHR heat exchanger: 1.OOE-2 (nominal estimate)
- Probability of 102% EPU initial power level: 5E-3 (same as base analysis)
. Probability of containment isolation failure: 7E-3 (nominal from base analysis)
- Probability of river water temperature >-96cF: 9E-3 (nominal value based on Table C-1. Although the river temperature has not exceeded 900 F based on the collected plant data, statistically there is a non-zero likelihood of such a temperature). 960 F is assumed here as the temperature at which COP credit is required (refer to Case la of Table 3-1).
- deltaCDF contribution for 3 RHR pump case: 3.1 E-5 x 1E-2 x 5E-3 x 9E-3
= -1 E-1 3/yr This additional contribution to the calculated deltaCDF from a 3 RHR pump case is negligible in comparison to the 2 RHR pump case.
An estimate of the deltaCDF risk contribution for the scenario with 4 RHR pumps in operation can be approximated as follows:
- Sum of BFN PRA Large LOCA initiator frequencies: 3.1OE-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: 7E-3 (nominal from base analysis)
- Probability of river water temperature >-1000F: 1E-3 (estimate based on Table C-1. Although the river temperature has not exceeded 900 F based 4-7 C1 320503.6924 - 2/27/2006
BFNEPUCOP ProbabilisticRistAssessment on the collected plant data, statistically there is a non-zero likelihood of such a temperature). 1000 F is assumed here as the temperature at which COP credit is required (refer to Case 1 of Table 3-1). deltaCDF contribution for 3 RHR Ipump case: 3.1 E-5 x 1.0 x 5E-3 x 7E-3 x 1E-3 = -1 E-12/yr Similar to the 3 pump case discussed previously, this additional contribution to the calculated deltaCDF from a 4 RHR pump case is negligible in comparison to the 2 RHR pump case. Summary of Modeling 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 risk contributors are also considered. 4-8 C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment Table 4-1 PARAMETRIC UNCERTAINTY ANALYSIS RESULTS Statistic 1i sCOP BFN Unit 1 Base CDF Risk Assessment CDF 5% 4.71 E-7 4.73E-7 50% 1.23E-6 1.21 E-6 MEAN 1.77E-6 1.77E-6 95% 4.72E-6 4.69E-6 4-9 4-9 C1320503-6924 - 2/27/2006
BFNEPUCOP ProbabilisticRiskAssessment Table 4-2
SUMMARY
OF SENSITIVITY QUANTIFICATIONS Description CDF LERF ACDF ALERF Base Case Quantification 1.77E-06 4.41 E-07 1.42E-09 1.42E-09 Pre-Existing Containment Leakage Sufficient to Fail COP Credit 1.77E-06 4.41 E-07 1.33E-09 1.33E-09 Defined by 100La (probability = 2.47E-4) Pre-Existing Containment Leakage Sufficient to Fail COP Credit 1.77E-06 4.41 E-07 1.53E-09 1.53E-09 Defined by 20La (probability = 1.88E-3) Expansion of Containment Isolation fault tree to Encompass Smaller 1 .77E-06 4.42E-07 2.05E-09 2.05E-09 Lines (approximate by multiplying Cont. Isol. failure probability by 1Ox) Assume Initial Power Level and Water Temperature Pre-Conditions 1.77E-06 4.42E-07 2.66E-09 2.66E-09 Exist 100% of the Time Combination of Cases #2, #3 and #4 1.77E-06 4.48E-07 8.33E-09 8.33E-09 Incorporation of "3-RHR pumps in SPC" and "4-RHR pumps in SPC" 1.77E-06 4.41 E-07 1.42E-09 1.42E-09 loss of NPSH scenarios Notes: (1) Senaris Wiii-l faiure of 2 ur iire RHR Puumps aUld assuUidted heat exUc;arlyeis in SPC are expijjiciy aniayzed in miese cases. As shown in Case 6, explicit incorporation of scenarios with 0 or 1 RHR pumps in SPC failed has a negligible impact on the results. (2) Case 2, 20L 8 containment leakage size, is the case used as the basis for the Conclusions of this study (refer to Section 5). 4-10 C1320503-6924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment Seismic The BFN seismic risk analysis was performed as part of the Individual Plant Examination of External Events (IPIEEE). BFN performed a seismic margins assessment (SMA) following the guidance of NUREG-1407 and EPRI NP-6041. The SMA is a deterministic evaluation process that does not calculate risk on a probabilistic basis. No core damage frequency sequences were quantified as part of the seismic risk evaluation. The conclusions of the SMA are judged to be unaffected by the EPU or the containment overpressure credit issue. The EPU has little or no impact on the seismic qualifications of the systems, structures and components (SSCs). Specifically, the power uprate results in additional thermal energy stored in the RPV, but the additional blowdown loads on the RPV and containment given a coincident seismic event, are judged not to alter the results of the SMA. The decrease in time available for operator actions, and the associated increases in calculated HEPs, is judged to have a non-significant impact on seismic-induced risk. Industry BWR seismic PSAs have typically shown (e.g., Peach Bottom NLJREG-1150 study; Limerick Generating Station Severe Accident Risk Assessment; NIUREG/CR-4448) that seismic risk is overwhelmingly dominated by seismic induced equipment and structural failures. Seismic induced failures of containment are low likelihood scenarios, and such postulated scenarios are moot for the COP question because they would be analyzed in a seismic PRA as core damage scenarios directly. Based on the above discussion, it is judged that seismic issues do not significantly impact the decision making for the BFN EPU and containment overpressure credit. 4-11 C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment Internal Fires The BFN fire risk analysis was performed as part of the Individual Plant Examination of External Events (IPEEE). BFN performed a screening methodology using the EPRI FIVE (Fire Induced Vulnerability Evaluation) methodology. Like most plants, BFN currently does not maintain a fire PRA. However, given the very low risk impact of the COP credit, even if fire risk was explicitly quantified the conclusions of this risk assessment are not expected to change, i.e., the risk impact is very small. Other External Hazards In addition to seismic events and internal fires, the BFN IPEEE Submittal analyzed a variety of other external hazards:
- High Winds/Tomadoes
- External Floods
- Transportation and Nearby Facility Accidents
- Other External Hazards The BFN IPEEE analysis of high winds, tornadoes, external floods, transportation accidents, nearby facility accidents, and other external hazards was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards.
Based upon this review, it was 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 C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment 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 explicil:ly 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 riot identify any differences that would make the Unit 1 PRA results and conclusions not applicable to Units 2 and 3. As further evidence, the Unit 3 PRA was modified in a similar manner as the Unit 1 sensitivity Case #2 and quantified to determine the ACDF impact. The result for Unit 3 was a deltaCDF of 1.9E-9/yr. The revised BF:N 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. 4-13 C1320503-6924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment Section 5 CONCLUSIONS The report documents the risk impact of utilizing containment accident pressure (containment overpressure) to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps during DBA 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 this risk assessment are based on the conservative 20La assumed containment leakage size (refer to Case 2 of Table 4-2). The conclusions of the plant internal events risk associated with this assessment are as follows.
- 1) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 104 /yr. Based on this criteria, the proposed change (i.e., use of COP to satisfy the net positive suction Ihead (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in CDF (1.53E-09/yir).
- 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 Ihead (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in LERF (1.53E-09tyr).
These results are well within the guideline of RG 1.174 for a "very small" risk increase. Even when modeling uncertainty and parametric uncertainty, and external event scenarios are considered, the risk increase is small. As such, the credit for COP in 5-1 C1320503-8924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment determining adequate NPSH for low pressure ECCS pumps during DBA, LOCAs is acceptable from a risk perspective. 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 C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment REFERENCES [1] "Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specifications (TS) Change 448 - One-Time Frequency Extension For Containment Integrated Leakage Rate Test (ILRT) Interval", TVA-BFN-TS-448, July 8, 2004. [2] Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI Report 1009325, Final Report, December 2003. [3] "Project Task Report - Browns Ferry Units 1, 2 & 3 EPU, RAI Response - NPSH Sensitivity Studies", GE Nuclecar 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 C1320503-6924 - 2/27/2006
BEN EPUCOPProbabilisticRisk Assessment Appendix A PRA QUALITY The BFN Unit 1 EPU PRA was used in this analysis for the base case quantification as it was recently updated consistent with the ASME PRA Standard and it is representative of each of the three BFN unit PRAs. The following discusses the quality of the BFN Unit 1 PRA models used in performing the risk assessment crediting containment overpressure for RHR and Core Spray pump NPSH requirements:
- Level of detail in PRA
- Maintenance of the PRA
- Comprehensive Critical Reviews A. 1 LEVEL OF DETAIL The BFN Unit I PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events.
The PRA model (Level 1 and Level 2) used for the containment overpressure risk assessment was the most recent internal events risk model for the BFN Unit 1 plant at EPU conditions (BFN model U1050517). The BFN PRA models adopts the large event tree / small fault tree approach and use the support state methodology, contained in the RISKMAN code, for quantifying core damage frequency. The PRA model contains the following modeling attributes. A.1.1 Initiating Events The BFN at-power PRA explicitly models a large number of internal initiating events: A-1 C1320503-3924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment
- General transients
- LOCAs
- Support system failures
- Internal Flooding events The initiating events explicitly modeled in the BFN at-power PRA are summarized in Table A-1. The number of internal initiating events modeled in the BFN at-power PRA is similar to or greater than the majority of U.S. BWR PRAs currently in use.
A.1.2 System Models The BFN at-power PRA explicitly models a large number of frontline and support systems that are credited in the accident sequence analyses. The BFN systems explicitly modeled in the BFN at-power PRA are summarized in Table A-2. The number and level of detail of plant systems modeled in the BFN at-power PRA is equal to or greater than the majority of U.S. BWR PIRAs 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 6924 - 2/27/2006
BFN EPUCOP ProbabilisticRisk Assessment The human error probabilities for the actions are modeled with accepted industry HRA techniques. The BFN PRA includes an explicit assessment of the dependence of post-initiator operator actions. The approach used to assess the level of dependence between operator actions is based on the method presented in the NUREG/CR-12713 and EPRI TR-1 00259. The number of operator actions modeled in the BFN at-power PRA, and the level of detail of the HRA, is consistent with that of other U.S. BWR PRAs currently in use. A.1.4 Common Cause Events The BFN at-power PRA explicitly models a large number of common cause component failures. Approximately two thousand common cause terms are included in the BFN Unit 1 PRA. Given the large number of CCF terms modeled in the BFN at-power internal events PRA, a summary table of them is not provided here. The number and level of detail of common cause component failures modeled in the BFN at-power PRA is equal to or greater than the majority of U.S. BWR PRAs currently in use. A.1.5 Level 2 PRA The BFN Unit 1 Level 2 PRA is designed to calculate the LERF frequency consistent with NRC Regulatory Guidance (e.g. Reg. Guides 1.174 and 1.177) and the PRA Application Guide. The Level 2 PRA model is a containment event tree (CET) that takes as input the core damage accident sequences and then questions the following issues applicable to LERF: A-3 A-3 C1320503- 6924 - 2127/2006
BEN EPUCOPProbabilisticRisk Assessment
- Primary containment isolation
- RPV depressurization post-core damage
- Recovery of damaged core in-vessel
- Energetic containment failure phenomena at or about time of RPV breach
- Injection established to drywell for ex-vessel core debris cooling/scrubbing
- Containment flooding
- Drywell failure location
- Wetwell failure location
- Effectiveness of secondary containment in release scrubbing The following aspects of the Level 2 model reflect the more than adequate level of detail and scope:
- 1. Dependencies from Level 1 accidents are carried forward directly into the Level 2 by transfer of sequences to ensure that their effects on Level 2 response are accurately treated.
- 2. Key phenomena identified by the NRC and industry for inclusion in E3WR Level 2 LERF analyses are treated explicitly within the model.
- 3. The model quantification truncation is sufficiently low to ensure adequate convergence of the LERF frequency.
A.2 MAINTENANCE OF PRA The BFN PRA models and documentation are maintained living and are routinely updated to reflect the current plant configuration following refueling outages and to reflect the accumulation of additional plant operating history and component failure data. The PRA Update Report is evaluated for updating every other refueling outage. The administrative guidance for this activity is contained in a TVA Procedure. A-4 A-4 C1320503-6924 - 2/27/2006
BFN EPU COPProbabilisticRisk Assessment In addition, the PRA models are routinely implemented and studied by plant PRA personnel in the performance of their duties. Potential model modifications or enhancements are itemized and maintained for further investigation and subsequent implementation, if warranted. Potential modifications identified as significant to the results or applications may be implemented in the model at the time the change occurs if their impact is significant enough to warrant. A.2.1 History of BFN PRA Models The current BFN Unit 1 PRA is the model used for this analysis. The BFN Unit 1 PRA was initially developed in June 2004 using the guidance in the ASME PRA Standard, and to incorporate the latest plant configuration (including EPU) and operating experience data. The Unit 1 PRA was then subsequently updated in August 2005. The Unit 1 PRA was developed using the BIFN 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 C1320503-5924 - 2/27/2006
BFN EPUCOPProbabilisticRisk Assessment A.3 COMPREHENSIVE CRITICAL REVIEWS As described above, the BFN Unit 1 PFA 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 BVVROG 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 rrianagement oversight were utilized to develop the Unit 1 PRA. Ax6 A-6 C1320503-43924 - 2/27/2006
BEN EPUCOPProbabilisticRisk Assessment A.4 PRA QUALITY
SUMMARY
The quality of modeling and documentation of the BFN PRA models has been demonstrated by the foregoing discussions on the following aspects:
- Level of detail in PRA
- Maintenance of the PRA
- Comprehensive Critical Reviews The BFN Unit 1 Level 1 and Level 2 PRAs provide the necessary and sufficient scope and level of detail to allow the calculation of CDF and LERF changes due! to the risk assessment requiring containment overpressure for sufficient NPSH for the low pressure ECCS pumps.
A-7 C1320503-3924 - 2/27/2006
BFN EPUCOPProbabilisticRisk Assessment Table A-1 INITIATING EVENTS FOR BFN PRA Initiator l Mean Frequency Category J (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 IA 4.27E-3 Loss of Instrumentation and Control Bus I B 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-I Inadvertent Opening of SRV ATWS 1.36E-2 Loss of Feedwater ATWS 3.02E-I LOCA Initiator Categories _ Breaks Outside Containment 6.67E-4 Excessive LOCA (reactor vessel failure) 9.39E-9 Interfacing Systems LOCA 3.15E-5 A-8 C1320503-43924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment Table A-1 INITIATING EVENTS FOR BFN PRA Initiator Mean Frequency Category (events per year} Large LOCA - Core Spray Line Break Loop I 1.68E-6 Loop II 1.68E-6 Large LOCA - Recirculation Discharge Line Break Loop A 1.18E-5 Loop B 1.18E-5 Large LOCA - Recirculation Suction Line Break Loop A 8.39E-7 Loop B 8.39E-7 Other Large LOCA 8.39E-7 Medium LOCA Inside Containment 3.80E-5 Small LOCA Inside Containment 4.75E-4 Very Small LOCA Inside Containment 5.76E-3 Internal Flooding Initiator Categories, EECW Flood in Reactor Building - shutdown units 1.20E-3 EECW Flood in Reactor Building - operating unit 1.85E-6 Flood from the Condensate Storage Tank 1.22E-4 Flood from the Torus 1.22E-4 Large Turbine Building Flood 3.65E-3 Small Turbine Building Flood 1.65E-2 A-9 A-9 C1 320503~324 - 2/27/2006
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 A-I0C1320503-43924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment Table A-2 BFN PRA MODELED SYSTEMS Suppression Pool / Vapor Suppression Turbine Bypass and Main Condenser A-11 C1 320503.6924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment Table A-3 PRA PEER REVIEW TECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT l CERTIFICATION SUB-ELEMENTS Initiating Events
- Guidance Documents for Inibiating Event Analysis
- Groupings
- Transient
- LOCA
- Support System/Special
- ISLOCA
- Break Outside Containment
- Internal Floods
- Subsumed Events
- Data
- Documentation Accident Sequence Evaluation
- GuidanoD on Development of Event Trees (Event Trees)
- Event Trees (Accident Scenario Evaluation)
- Transients
- SBO
- LOCAK
- ATVWS
- Special
- ISLOCAIBOC
- Internal Floods
- Success Criteria and Bases
- Interface with EOPs/AOPs
- Accident Sequence Plant Damage States
- Documentation A-12 A-12 C1320503-~3924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment Table A-3 PRA PEER REVIEW TECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Thermal Hydraulic Analysis
- Guidanoe Document
- Best Estimate Calculations (e.g., MAAP)
- Generic Assessments
- FSAR - Chapter 15
- Room Heat Up Calculations
- Documentation System Analysis
- System Analysis Guidance Document(s)
(Fault Trees)
- System Models
- Structure of models
- Level of Detail
- Success Criteria
- Nomenclature
- Data (see Data Input)
- Dependencies (see Dependency Element)
- Assumptions
- Documentation of System Notebooks A-13 C1320503-5924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment Table A-3 PRA PEER REVIEWTECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT CERTIFICATION SUB-ELEMENTS Data Analysis
- Guidance
- Component Failure Probabilities
- SystemrTrain Maintenance Unavailabilities
- Common Cause Failure Probabilities
- Unique Unavailabilities or Modeling Items
- AC Recovery
- Scram System
- EDG Mission Time
- Repair and Recovery Model
- SORV
- LOOP Given Transient
- BOP Unavailability
- Pipe Rupture Failure Probability
- Documentation Human Reliability Analysis
- Guidanoe
- Pre-Initiator Human Actions
- Idenlification
- Anablsis
- Quantification
- Post-Initiator Human Actions and Recovery
- Idenlification
- Analysis
- Quantification
- Dependence among Actions Documentation A-14 C1320503-5924 - 2/27/2006
BFNEPUCOPProbabilisticRisk-Assessment Table A-3 PRA PEER REVIEW TECHNICAL ELEMENTS FOR LEVEL 1 PRA ELEMENT J 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
- Guidanoc
- RPV Capability (pressure and temperature)
- ATWS
- Transient
- Containment (pressure and temperature)
- Reactor Building
- Pipe Overpressurization for ISLOCA
- Documentation Quantification/Results
- Guidance Interpretation
- Computer Code
- Simplified Model (e.g., cutset model usage)
- Dominant Sequences/Cutsets
- Non-Dominant Sequences/Cutsets
- Recovery Analysis
- Truncation
- Uncertainty
- Results Summary A-15 A-15 C1320503-5924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment Table A-4 PRA CERTIFICATION TECHNICAL ELEMENTS FOR LEVEL 2 F-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
- LERF Definition
- CE'Ts
- Documentation A-16 C1320503-5924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment Table A-5 PRA CERTIFICATION TECHNICAL ELEMENTS FOR MAINTENANCE AND UPDATE PROCESS I-PRA ELEMENT CERTIFICATION SUB-ELEMENTS Maintenance and Update Process
- Guidance Document
- Input - Monitoring and Collecting New Information
- Model Control
- PRA Maintenance and Update Process
- Evaluation of Results
- Re-evaluation of Past PRA Applications
- Documentation
=
A-17 C1320503-5924 - 2/2712006
BEN EPUCOPProbabilisticRisk Assessment Appendix B PROBABILITY OF PRE-EXISTING CONTAINMENT LEAKAGE Containment failures that may be postulated to defeat the containment overpressure credit include containment isolation system failures (refer to Appendix D) and pre-existing unisolable containment leakage pathways. The pre-existing containment leakage probability used in this analysis is obtained from EPRI 1009325, Risk Impact of Assessment of Extended Intearated Leak Rate Testing Intervals.[2] This is the same approach as used in the recent 2005 Vermont Yankee EPU COP analyses, and accepted by the NRC and ACRS. [4] EPRI 1009325 provides a framework for assessing the risk impact for, extending integrated leak rate test (ILRT) surveillance intervals. EPRI 1009325 includes a compilation of industry containment leakage events, from which an assessment was performed of the likelihood of a pre-existing unisolable containment leakage pathway. A total of seventy-one (71) containment leakage or degraded liner events were compiled. Approximately half (32 of the 71 events) had identified leakage rates of less than or equal to 1La (i.e., the Technical Specification containment allowed leakage rate). None of the 71 events had identified leakage rates greater than 21 La. EPRI 1009325 employed industry experts to review and categorize the industry events, and then various statistical methods were used to assess the data. The resulting probabilities as a function of pre-existing leakage size are summarized here in Table B-1. The EPRI 1009325 study used 10OLa as a conservative estimate of the leakage size that would represent a large early release pathway consistent with the LERF risk measure, but estimated that leakages greater than 600La are a more realistic representation of a large early release. B-1 C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment This analysis is not concerned per se about the size of a leakage pathway that would represent a LERF release, but rather a leakage size that would defeat the containment overpressure credit. Given the low likelihood of such a leakage, the exact size is not key to this risk assessment, and no detailed calculation of the exact hole size is performed here. The recent COP risk assessment for the Vermont Yankee Mark I BWR plant, presented to the ACRS in November and December 2005, determined a leakage size of 27La using the conservative 10CFR50, Appendix K containment analysis approach. Earlier ILRT industry guidance (NEI Interim Guidance - see Ref. 10 of EPRI 1009325) conservatively recommended use of 10-La to represent "small" containment leakages and 35La to represent "large" containment leakages. Given the above, the base analysis here assumes 35La as the size of a pre-existing containment leakage pathway sufficient to defeat the containment overpressure credit. Such a hole size does not realistically represent a LERF release (based on EPRI 1009325) and is also believed (based on the W 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 the 35La pre-existing containment leakage used in this base case analysis is 9.86E-04. 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 C1320503-5924 - 2/27/2006
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 [2] recommends these values for use for both BWRs and PWRs. Reference [2] makes no specific allowance for the fact that inerted BWRs, such as BFN, could be argued to have lower probabilities of significant pre-existing containment leakages. B-3 B-3 C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment Appendix C ASSESSMENT OF RIVER WATER AND SP WATER TEMPERATURE VARIATION The BFN river and torus water temperatures were analyzed to statistically model variability in temperature. The purpose of this data assessment is to estimate for use in the risk assessment the realistic probability that these temperatures will exceed a given value, i.e. the probability of exceedance. C.1 BFN EXPERIENCE DATA The following sets of river water inlet and torus water daily temperature data were obtained and reviewed: Unit I Data Period Years 2 O1/01100 - 01/31/06 6.1 3 02/01/03 - 01/31/06 3.0 Data for suppression pool water level for the above time periods were also obtained. However, statistical assessment of the variation in pool level was not pursued as the small variation in pool level has a non-significant impact on the COP / NPSH calculations. 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 C1320503-6924 - 2/27/2006
BEN EPUCOPProbabilisticRisk Assessment As the torus water temperature has a high dependence on river water temperature for most of the year, the assessment of the torus temperature variability also is based on the 6.1 year data set from Unit 2. 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, CO) to approximate a distribution of the expected range in temperature. The Crystal Ball software automatically tests a number of curve fits. The best fit for the temperature data is a normal distribution that is truncated at user-defined upper and 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 0F degrees). To constrain the distributions, the following user-defined upper and lower bounds were used: C-2 C132050346924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment
- 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 90 0F)
- 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°lF)
The Crystal Ball software statistical results for the river water temperature, and torus water temperature variations are provided in Figures C-4 and C-5, respectively. The statistical results are also summarized in the form of exceedance probability as a function of temperature in Figures C-6 and C-7. The information is also presented in tabular form, Tables C-1 and C-2. As discussed previously, the river wal:er and the torus water temperature variations are not independent; as such, the exceedance frequencies are not independent (i.e., they should not be multiplied together directly to determine the probability of exceeding a particular temperature in the river AND at the same time exceeding particular temperature in the torus). C-3 c-3 C1320503-5924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment Figure C-1 CHRONOLOGICAL VARIABILITY IN RIVER WATER AND TORUS WATER TEMPERATURES Pool TempI - River Temp AS td% k K - 85- - rAl fix %- 9AOIb (INI -1 A\ -- A\ v i I -- 75 - - la fillI 1.IrI , )i aII III I 2 ART, it Ii E 65 - - I I- V1 I.ilIII 11 - S2 I11' It I', ?i 55 11 I - fU 11.1 I I i I I II - I 19 45
- I
- 1n III I II I 01/01/99 01101/00 12131100 12131101 12131102 12/31/03 12/30/04 12/30/05 12/30/06 Date C-4 C4C1C1020303924 - 2127/2006
BFN EPU COP ProbabilisticRisk Assessment Figure C-2 RIVER WATER TEMPERATURE HISTOGRAM 400 350 300 250 8 200 la 150 100 50 0 32.5 37.5 42.5 47.5 52.5 57.5 62.5 67.5 72.5 77.5 82.5 87-5 92.5 Temperature C-5 C1320503-6924 - 2127/2006
BFN EPU COP ProbabilisticRisk Assessment Figure C-3 TORUS TEMPERATURE HISTOGRAM 700 600 500 A40 4 co (U a 300 200 100 0 - MO. -- 1 S.. C _,75 _2 .5 U, .U T emprat ..5 -2.5 D2.5 07.5 Temperature Cal C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment Figure C-4 STATISTICAL RESULTS FOR RIVER WATER TEMPERATURE VARIATION Crystal Ball Report Simulation started on 2/6/06 at 7:09:56 Simulation stopped on 2/6/06 at 7:11:44 Forecast: Pool Temperature C:ell: C15 Summary: Display Range is from 55.00 to 95.00 F Entire Range is from 55.00 to 95.00 F After 50,000 Trials, the Std. Error of the Mean is 0.05 Statistics: Value Trials 50000 Mean 75.75 Median 76.06 Mode Standard Deviation 11.30 Variance 127.65 Skewness -0.08 Kurtosis 1.85 Coeff. of Variability 0.15 Range Minimum 55.00 Range Maximum 95.00 Range Width 40.00 Mean Std. Error 0.05 Forecast Pool Terwperatwe 60,000 Trials Frequency Chart 0 Outliers Am. 4M. 6.01 l600 75.00 ai0 0000 C-7 C132050343924 - 2/27/2006
BEN EPUCOPProbabilisticRisk Assessment Figure C-5 STATISTICAL RESULTS FOR TORUS WATER TEMPERATURE VARIATION Crystal Ball Report Simulation started on 2/6/06 at 7:09:56 Simulation stopped on 2/6/06 at 7:11:44 Forecast: River Temperature Cell: G18 Summary: Display Range is from 30.00 to 100,00 F Entire Range is from 32.00 to 95.00 F After 50,000 Trials, the Std. Error of the Mean is 0.08 Statistics: Value Trials 50000 Mean 63.50 Median 63.41 Mode Standard Deviation 18.07 Variance 326.51 Skewness 0.00 Kurtosis 1.81 Coeff. of Variability 0.28 Range Minimum 32.00 Range Maximum 95.00 Range Width 63.00 Mean Std. Error 0.08 Forecast River Tom~eratire 50.000 lHats Frequency Chart 0 Ouliers
.012 613 MG9 4697
.6 .e696, .
69 3 163.2~ A0 30L69 47.3 GSW 82.60 I..0 F C-8 C1 320503-6924 - 2/27/2006
BFNEPU COP ProbabilisticRisk Assessment Figure C-6 RIVER WATER TEMPERATURE EXCEEDANCE PROBABILITY 1.OE+O 1.OE-1 4 0! IL wU C. 2 4C awU wU C. xwU 1.OE-2 1.OE-3 25 34 37 41 44 48 51 55 58 62 65 69 72 76 79 83 86 90 93 97 100 RIVER WATER TEMPERATURE (F) C-9 0-9 C1 320503-M94 - 2127r2006
BFNEPU COP ProbabilisticRisk Assessment Figure C-7 TORUS WATER TEMPERATURE EXCEEDANCE PROBABILITY 1.OEiO -A0NVAVffi--
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Lu1.OE-3 50 57 59 61 63 65 67 69 71 73 75 77 79 81 83 85 87 89 91 93 95 TORUS WATER TEMPERATURE (F) C-1 0 c-I320503-6924 oCl - 2r27/2006
BFNEPUCOPProbabilisticRiskAssessment Table C-1 RIVER WATER TEMPERATURE EXCEEDANCE PROBABILITIES Temperature (F) 1 Exceedance Probability 30 1.OOE+OO 35 9.55E-01 40 8.80E-01 45 8.02E-01 50 7.24E-01 55 6.45E-01 60 5.64E-01 65 4.74E-01 70 3.97E-01 75 3.17E-01 80 2.41 E-01 85 1.64E-01 86 1.40E-01 90 8.46E-02 95 9.15E-03 100 O.OOE+00 C-1 1 c-li C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRisk Assessment Table C-2 TORUS WATER TEMPERATURE EXCEEDANCE PROBABILITIES Temperature (OF) Exceedance Probability 30 1.OOE+00 35 1.OOE+OO 40 1.OOE+00 45 1.OOE+O0 50 1.OOE+O0 55 1.OOE+00 60 8.90E-01 65 7.79E-01 70 6.63E-01 75 5.28E-01 80 4.01 E-01 85 2.62E-01 90 1.35E-01 92 8.25E-02 95 1.01 E-02 100 O.OOE+00 C-12 c-I2C1320503-5924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment Appendix D LARGE-LATE RELEASE IMPACT In the November-December 2005 ACRS meetings concerning the Vermont Yankee EPU and COP credit risk assessments, the! ACRS questioned the impact on Large-Late releases from EPU and COP credit. The following discussion is provided to address this question for the BFN COP credit risk assessment. D.1 OVERVIEW OF BFN PRA RELEASE CATEGORIZATION The spectrum of possible radionuclide release scenarios in the BFN Level 2 PRA is represented by a discrete set of release categories or bins. Typical of industry PRAs, the BFN release categories are defined by the following two key attributes:
- Timing of the release
- Magnitude of the release D.1.1 Timing Categorization Three timing categories are used, as follows:
- 1) Early (E) Less than 6 hours from accident initiation
- 2) Intermediate (I) Greater than or equal to 6 hours, but less than 24 hours
- 3) Late (L) Greater than or equal to 24 hours.
The definition of the timing categories is relative to the timing of the declaration of a General Emergency and based upon past experience concerning offsite accident response: D-1 C1 320503-43924 - 2/27/2006
BFN EPUCOPProbabilisticRisk Assessment
- 0-6 hours is conservatively assumed to include cases in which minimal offsite protective measures have been observed to be performed in non-nuclear accidents.
- 6-24 hours is a time frame in which much of the offsite nuclear plant protective measures can be assured to be accomplished.
* >24 hours are times at which the offsite measures can be assumed 1:o be fully effective.
Magnitude Categorization The BFN Level 2 PRA defines the following radionuclide release magnitude classifications:
- 1) High (H) - A radionuclide release of sufficient magnitude to have the potential to cause prompt fatalities.
- 2) Medium or Moderate (M) - A radionuclide release of sufficient magnitude to cause near-term health effects.
- 3) Low (L) - A radionuclide release with the potential for latent health effects.
- 4) Low-Low (LL) - A radionuclide release with undetectable or minor health effects.
- 5) Negligible (OK) - A radionuclide release that is less than or equal to the containment design base leakage.
The definition of the source terms levels distinguishing each of these release severity categories is based on the review of existing consequence analyses performed in previous industry studies, PRAs and NRC studies containing detailed consequence modeling. The BFN Level 2 PRA uses cesium as the measure of the source term magnitude because it delivers a substantial fraction of the total whole body population dose. This approach is typical of most industry PRAs. In terms of fraction of core inventory Csl released, the BFN release magnitude classification is as follows: D-2 C1320503M924 - 2/27/2006
BFNEPUCOPProbabilisticRistAssessment Release Magnitude F:raction of Release Csl Fission Products High greater than 10% Medium/Moderate 1 to 10% Low 0.1 to 1.0% Low-Low less than 0.1% Negligible much less than 0.1%
= =
D.2 LLOCA COP CREDIT IMPACT ON LARGE-LATE Based on the preceding discussions, it can be seen that "Large-Late" scenarios are termed High-Late releases in BFN Level 2 PRA terminology and are defined as releases occurring after 24hrs and with a magnitude of >10% Csl. For this risk assessment it is not necessary to perform any explicit quantification of the Level 2 PRA to determine the effect on large-late releases, i.e., the scenarios of interest in this analysis are never late releases, in fact they are all always Early releases. The scenarios of interest in this risk assessment are very low frequency postulated scenarios that were not explicitly incorporated into the BFN base PRA. These scenarios are defined by containment isolation failure at t=0, leading to assumed loss of NPSH to the ECCS pumps in the short term and leading to core damage in approximately one hour. In summary, there is no change in the frequency of Large-Late releases due to the credit of COP in DBA LOCA scenarios. D-3 D-3 C1 320503-6924 - 2/27/2006
BFNEPUCOP ProbabilisticRisk Assessment Appendix E REVISED EVENT TREES This appendix provides print-outs of the BFN Unit 1 PRA modified event trees used in this analysis. In addition, the RISKMAN software event tree "rules" and "macros" for these revised event trees are also provided in this appendix. E.1 MODEL CHANGES The following are details of the changes made to the BFN Unit 1 PRA RISKMAN models for this risk assessment. The BFN Unit 1 PRA model of record was modified for this risk assessment to question the status of containment integrity first in the Level 1 large LOCA event trees. In addition, a second node was added to the large LOCA event trees to question the probability of extreme plant conditions (e.g., high river water temperature). These nodes are then used to fail the RHR and CS pumps for scenarios with 2 or less RHR pumps in SPC. The scope of the analysis is limited to large LOCA accidents. In order to ensure that only the large LOCA initiators are affected by the event tree changes, several of the existing event trees were renamed. In addition, because the containment isolation top event CIL is located in the containment event tree CET1, it too was renamed. The event tree names were revised as follows: Original Event New Event Tree Tree Description CETI CETN1 Containment event tree 1 LLCS LLCSN Core spray LLOCA event tree LLRD LLDSN Recirc: discharge LLOCA event tree LLO LLON Other large LOCA event tree LLRS LLSN Recirc suction LLOCA event tree E-1 C1320503-5924 - 2/27/2006
BFN EPUCOPProbabilisticRisk Assessment In the containment event tree, top event CIL was replaced with a dummy top event, CILDUM, which is a switch whose branches depends on CIL, now moved into the large LOCA event trees. Two split fractions were developed for CILDUM, one for success (CILDS) and one for failure (CILDF). The branches of CILDUM depend on CIL, which is traced via macro CILFAIL. Macro CILFAIL is a logical TRUE if top event CIL=F, otherwise it is FALSE. If CILFAIL is TRIJE, that is if CIL fails, then the failed branch of CILDUM is assigned via split fraction CILDF (1.OOE+00). Otherwise, the success branch is assigned via split fraction CILDS (O.OE+O0). The purpose of installing dummy top event CILDUM is to preserve the containment event tree structure (i.e., the RISKMAN software allows use of a specific top event name only once in an accident sequence structure). All top events that are asked in the base model if CIL fails are still asked; those that are not normally asked are riot asked in this sensitivity case. In each of the large LOCA event trees, top event CIL was added as the left most top event. Top event NPSH was added as the next top event to the right. In this way, the original event tree structure is preserved because CIL transfers to NPSH which transfers to the original first top of each event tree. CIL models containment isolation penetrations greater than 3 inches, and top event NPSH models the probability of reactor power at 102% as well as river water temperature greater than 86F. Top event NPSH has two split fractions NPSH1 and NPSHS (success, equal to O.OOE+00). The latter is applied for all initiators other than those modeling large LOCAs. The existing CIL fault tree was modified to add the probability of a pre-existing containment leak; a basic event was inserted just under the top 'OR' gate of the CIL fault tree. The basic event is set to different values depending on the size of the leak rate assumed. See Table 4-2 for the sensitivity cases and associated pre-existing leak size. The values used and the resultant CIL split fraction values are listed below: E-2 C1320503-6924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment Sensitivity Leak CIL Split Case Leak Size Proabili Fractions(" Base 35 La 9.86E-04 1.36E-03 1 100 La 2.47E-04 6.22E-04 2 20 La 1.88E-03 2.25E-03 3 Base CIL split fractions X 10, 9.86E-04 6.37E-03 plus pre-existing leak 35 La 4 35 La 9.86E-04 1.36E-03 5 Base CIL split fractions X 10, 1.88E-03 7.37E-03
._. plus pre-existing leak 20 La Note:
T SI support split fraction. Degraded stal:e split fraction is also affected but not shown. Top event NPSH models the probability that the plant is at 102% reactor power with 86F river water, 'OR' the reactor is at the nominal 100% reactor power level with river water greater than 70F. The probability that the plant is at 102% power is modeled using a miscalibration human error probability taken from a similar action documented in the existing BFN Unit 1 PRA Human Reliability Analysis (see event ZHECCL, instrument calibration error, Control Room). The probability that the river water is either greater than 70F or greater than 86F is developed in the data analysis (refer to Appendix C). Top event NPSH has two split fractions, NPSH1 and NPSHS. The latter is used to filter out sequences where greater than 3 RHR pumps are running. This latter pass-through split fraction is used to exclude the cases where sufficient RHR pumps are cooling the torus such that containment overpressuire is not necessary (per DBA calculations) for the success of the RHR and CS purrips. The status of the RHR pumps and heat exchangers is tracked via an existing macro in the event tree RHRET. Split fraction NPSHI is the default split fraction. Refer to Section 4.2.2 where scenarios with more than 2 RHR pumps in SPC are analyzed as a sensitivity case. When both top events CIL and NPSH fail, conditions are present such that the model assumes there is insufficient NPSH for the low pressure pumps to operate during a large LOCA. RISKMAN rules were added to assign guaranteed failure split fractions for E-3 E-3 C1320503-5924 - 2/27/2006
BFNEPUCOPProbabilisticRiskAssessment top events: CS, LPCI, LPCII, SPI and SPII. A macro was created (NPSHLOST, defined as CIL=F*NPSH=F) and defined in each large LOCA event tree. The macro was then added to the split fraction rule for each guaranteed failed split fraction for the desired top event. Note that drywell spray failure is captured by the event tree structure (i.e., if LPCI loops I and 11are failed, then drywell spray is never asked in the event trees). Ed4 E-4 C1320503-6924 - 2/27/2006
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BPC LOGIC SWITCH EOR SMU~?SSO' W OOD COOLIN WITH (11 U oDws. OPUAPO1R AUMGS mfyWELL SPRAY DRYWElLL SPRAY WMA-REW E-1 5
BFNEPUCOPProbabilisticRiskAssessment Model Mua : o1ico2-9 Split Fraction Acsins.evt Rule for Event Tree: 'LLO 5:07 RX t9/*/OO6. page I. sr Split Frction lssignment RuIs CXLI POXfSfl2S + LVPS) CIL2 PtXP(Wn84LV18)i eWPSIS Rft~ltlxEE2"t3Jtl3 4- flfflaltrBhI2tflR4 +F !IR1'R*Aa3'RnR4 t PSlR2*Rn3n1naIRA' Comernnts Xr 3 OF.KORS KM MPS AlAMhALABILE Wi DON"'T RCO COP FM =5 WPSSI. XK7'sbLO 4 tf + -i2LDA+4-11flO + XfltLO + XNEXP-SL2 +
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BFN EPUCOPProbabilisticRisk Assessment Model lime: tVllOf2-9 Split Frsotiogn Aia mrixt Rule for Gent Trees LIN, St6l pM 2/9/2X64 Pae 2 MOW_ SF -Split 3Nraottior Lauigmet tl@ Ccments my t Uo PUgMS OR CS twrae UN==ctLO S~y 1 OSPC1 8iPSW.B*S>SS osscr 1 SP't R-F 4 OPCW KPSHIOST R SPX24 + SPIYF snz~r s'-r 7i-F t8iS t nsnwsi S9C51 (RflstRXF32sS = RP 5 -SI4W4 flfls5fBS~S{'~s-fS 4 OPCS mns-+aostx~sj +a p.?U~ SPMlT I SPC 1 awsr 9 fl-*P-X + tRPA-..*P!.Wp +Rr-F+tIOmf)
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DWoS 1 E-1 7
BFNEPUCOP ProbabilisticRiskAssessment Model blme: ¶7lCOP2-9 Mlao for Ivent Tree: LLCW 5:0-' N 2Sf/toot Macro ~~Macro tie /Cmust ALU Mmo is NEEDED IN4THE cSs ALtINJU2X RPS
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BFNEPUCOP ProbabilisticRisk Assessment M~odel Vu=ma UICOP2-9 Kacso fox I1.vuxit Traee: LLW2
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BFNEPUCOPProbabilisticRisk Assessment
- Model M.s=e, VU1002-9 W&=0fr iEvent TX",. LLO=
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MODEL Neme: UIERIN Page No. I of 2 Event Tree: LLRDN.ETI 1337:46 Fe ary 15, 2008 IE CIL NPSH RPSM RPSE TOR UP IVC DVI 0V2 LPCI LPCII CS Si _ ..... .. _ .. . . .................................. _.__ A
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MODEL Name: UERIN Page No. 2 of2 Evert Tree: LLRDN.ETI 13:37:46 February 16, 2X6 os----c- -- - S-- OlSPC Bpi SPI_ PC O" MS I X# I 2 2 3 3 4 4 Xi 5 xi 6 9-12 Xi 7 13-16 8 17 9 18
............... . . ... ........ . ......... .... ....... . . X2 10 19-36
.... ..... .... ... ..... . .... .... ...................... . .................. X2 11 37-54 X2 12
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. ........ . ......... ...... . ...................... X2 14 91-108 X2 1I 1M026 CZ
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................. .................... .. . .............. ..................... X4 17 X4 18 28432 -k X3 19 43348 Xs 20 46 36 21 937 22 938 23 939
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BFNEPUCOPProbabilisticRisk Assessment Model I= : U1C22-5 Top Events for Event Tree: LL1ODT
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CIL PRIMARY COKAINMNT LSOLATIGJ FAIUR= E - LAPSE (C-3INCU NPSK COEWTIOCS PRAMTINO $"I liTOR LLOCA MES'4 CiW;3 PORTIO4 or RP:I SUOOflUSL RPSE . LOECRICAL ?ORICKI or RP8 tCUG-S3a00 BSis) TOR IPRatSu SUPPfl5SION POIA W tetURaINE TRI P Icw CsUrR OFr WM DVI LOOw I RaTt:Afl'rCa4 a1smu VALV CLOS* S DV2 LOO nT RaXcuLWnm DzaatFIO&E VAIVS GLoUU3 LPX lPCI LOOP E LPCIX lam LOOP I CS CORE SPRAY SYSM St LOGIC SWITC MR SUFTlCIIR4T IXJICTIOW S9COPERATOR 0 LI4S sltPSS rON POOL OLING 21 WJ??SSSC PooL COOLNG mAARE - LOr I SP1$ BUPPRhSSX POOL OOOL:X#4 hxnvm - Loop II SPO -?; OCl SWrrcir MR SUPPRBS3iCE POOL COOtNG WITH W! RAE COWS OPERATO AUrGN DXThLbi -SPRAY mYmtL SpRi mafRm DAUS E-23
BFNEPUCOP ProbabilisticRisk Assessment Model Nauai.: VICOP2-9 Split i'ractioxa Aaig2=Lo3t PRul. for Event Treeat UALD3 CILI + LYP'SL CILP wF~LPr x63RM*RH3iI2*RaS3 + PJXEU*WW*ERKa4 +$ Zxil~3PmR + mm*mrmu4 cowcwnts !Er 3 OR 110H PUMPS ARS &VAILR"LE WS DON'T KtvD COP FOR =cS V1KZ7IITLLCA + !241-LLC 4. ZVI-T.=LD 4. IMrI-OLL + EM4IT@-=Lf. ISXT-VLA 4.
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BFNEPUCOPProbabilisticRisk Assessment Model MIaa: U1CP2-9 Split Fraction Msigmaent tle, for Event Tree: LLWLN 5:CX 1K 2/9/2004 Page 3 Sc Spi>t Cort e spay t t >;14.TFgue: CSZS (RWint+Mir+DaF+A-Sf RPS-lnirP)n+W+rLPvFtFttrS CF# EAn4'&354*tK-? + EA]l>F'ED-F 4 cXss!g'G4OW-rDv-r)a*-1+I"FA-rnBrEC-PsDwgtfi
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MODEL Nae:. UERIN Page No. 1 nf 4 EverA Time LLtSN.ETn 13:38:2 Fedrum 16, 20O IE CfI NPSH RPSM RPSE TOR TTIP NG Dvi 0V2 LPCI LPCII Cs el OSPC SP(
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- 5:09 as 21912008 Mar 1ai. / Sm~
Kch Rule Xau> ooCratex THIS MAMO 15 B N9a D IN ThE CZTS PLRBRSOOL 5YM5 LY'S1C ISCA P.% 7iAYE. SPWSS'XIZ3 E-40
MODEL Now, U1iEPN Pag mNm1 of Event Tre:'CE`N.ETI I3.3t5 Fmry 18 200 IE L. . AL : CoUM 01 IR C TO X o ow W RM I 2 a 4 5 S 7 8 9 10 II 12 13. 14 rp 18 17 la 18 19 tv 20 21 Q 22 23 24 25 20 26 (Z3. 2.7 28 9. 29 Z-cl 3o
.31 R 32 o as L 34 L 35 36 gft 112 m
- 3
MODEL NamVe: UIERN Pae No. 2ot4 EwiA T.e= CE ETM 133B:50 FebmWry ia 20E So 2 3 4 5 6 7 8 9 10 11 42 13 14 ab. 15 16 17 z 18
*9 20 0 21 N 22 23 24 25 26 E2.
27 28 29 t8 rb 30 El CY, 31 32 9t 33 z C-1 34 35 38
MODEL N~mw UIERIN Page No. 3of 4 Even TM.e CEThIETI FbWaty 16, 2001 13:3:605 I E 12 AL CILDUM oa IR CZ TO OwlD .WR RMS I # B# I i I 37 38 39 40 41 rn (,j
-.w I-t
MODEL Name: UiERN Page No. 4 f 4 Event Tree: CETNI.ETI 13:35:0 Fenuay 16, 2e St 37 38 39 40 41 4k 40,
.to 0
C) ZI-v ci Eg C.., m
- s w-4
BFNEPUCOPProbabilisticRisk Assessment M~odel lNume tflCOP2-9 Top Evemts for Even~t Trae: GRT~Tl 5:09 WX21912006 L2 UVEL 2 /LUJ RZWF.?a A.CEIT U)SC iNor FRc CLASS 2 Oz AST (11LOURCIL m We4Y TOP ox OPERRA'OM MPEVURLZ21 AIN IL~2) I-VESSML fSACMY T!) es-ABTsID AWE~~ [~~INODIRE~CT MRYwU RZEA8I IATH WI.T AIR 99CE tAILUJR a~l cQ'~AZN~? 9ULDZNG- Z'ElClVI E-45
BFN EPUCOPProbabilisticRis.k Assessment Model Name: UlCOP2-9 Split WnaOtion Assignment Rule for Event Tzee CEfl1f 5:09 PM 2/9/2006 SW 3it Frlactio As uant Atul. L20 Cowents L20-0 INPLXmr$LkVY 1; D2041 IPLWS LEVEL2; USE MKr TO CHANGE ALr CLASIA 4 OLASsliE 4 CLUSIC + CLASSID + CLassiR 4 CLASS3A + CLASsS3 + ALO ROCO + CLASS2aL CLASS + CLSS2L t =SS2t + CLSSIV 4 (CLASS3D + CLASSA
+ CL43E5) + sucAxT Cow~ents CLASS 3D AN!) CLaS3 4 AM EV£W&WATEO FR D.SF C ? CILVAIL CILDS X 013 CLASS3A + CLASS3t CLASS3C + LW O_' CLSS2A + CLASS2T + NRV4 (CLASSIA + MASSISS + CLASSIS+ CLASSIC, +
CLASSI1 (NPOACIC t bOrC 014 CLASSIS 013 -OPDPLI'(cLRBSI& + CLASSIC 4 CASSiC)D comnents changq I AlGI PRESSUZ LES1r 012 OPDEPL1* (CLASSIA + CMSSIC + CLASSI) comzents claugel MGCA PRESSUPR LAS? 131 OZ1l t ( CLR5S1A 4 OLASI£C) TM3 CLASSISZ IR4 S813t IRS , b0-FCUaSSID XRS Ot- 4 CMLASSID Covoaer~rs the izginal Ul 22 model 131 orwrfc^asslr' IRI 0IlS&CLAS8iE eUnB L Passsoc xIECTIc MPLIIT= CZ2 XR"FTOIofs, C52 CZ1 tRS*OKaF CZ3 lR.5*0OF E-46
BFNEPUCOPProbabilisticRisk Assessment Model game: UlCOP2-9 Split .Faction Assignment Rule for Event Tree:. Cirul 5:09 1*2/912006 PagS 2 SS Splt Frzatio £siget t S TD1. CLASSI TDS Qf ASI TOR I~lSM TD4 -(OF.B)'OL;ASSISL TD4 OluW*CLABSI1A Vol AEaTf S + I5 To-S(CLAsSIA + atAssIs
- CLASSIBL + C tB 4 cLSSSA + CLASSSB 4 OLLSS3C)
I) tD-rs(CLEsSta + CLASSIC CU"SlD 4 aSS3A 4. cLassa 4 CASS32C FD4 TD-F" CLASSIBE + CAS1SEL we Xgi-S'.
- CLASSISL Coter.ts T-3VWSPR2'Mt*RUSPCO& This uaw an asrwap;iOn ibst zeaultad in W%7 ox-i R456 OV.8T>S*S.4090-S 0ZeS*T0..S*N'-*Df F
- L464 OX-S~iD'.f'ID.'
L2O I C--.usnts L200 IMP LZS LrU: LEWV WC- ILM V2 BSKZ W0 C ALE? 411la + cLSBsz 4 'cIaSSiC 4 CLASSID 4 CASSlt + CLASSZA + CLASS33 + CLASS3C AVLJ SOCW + CLASS1ISL 4 CIASSA + CLSS2L + CIa2 + CLS2Y + LS1I + MA554
-+ CLASSS? + A 0CosUttS CLASS 30 aa3 CZ"$ 4 Mg EtUATSD FM 5t1 CILDr 111 eChOS C' E-47
BFN EPU COPProbabilisticRisk Assessment Model. Wn: UCOFP2-9 Split Fraction Assignment Rule for EYQnt Tree: CETN
*9/g/2006 r x:01 Ptage 3 sw Split iracton Msigmett tRle cis CASS3A + CLPJS3S i C1S3C + 2OW 01 C;ASS2a + CtAS$2f + tOR'v* caSslA
- COLSIhh + CTL5S13+ CLASSIC)
C:LISSlr CNOAORC + N1DC0 014 =$SIl 013 O0PDDL1'*(CLASSIA + LASSIC + CLASS)) trments change? hCidh PUtSSOUt LSit? t 012 OPDCPULv CLMS1 + CLASSIC + CLASSiD) Comaxts change I hGP PRSSUnv Lr IR1 ClaP tOLAS55A + CLASSLC! flU CLASSIBE R4 CLASSISL IRS 0lStFCW52D Cotnnta the itqinal VI 7W4 wo4i nri 01-r*CLASSIE IRS OtwSflLSSIE 1:R2 tXS Cow.ents LOW PMSSS0IL INJ3E00t4R DWLICIZ XRF' za2 054 CZi 1R-S*O1-S CZ3 XRt} CE! - i TD1 . CLaSSII TD3 -IOX-B) CLSSIE3 704 -tO£1-3) *CfLfAS~tB - 7013 Ot-rICLASS~lA 701 3 E-48
BFNEPUCOPProbabilisticRiskAssessment Model kamc: tflCOP2-9 Split ratio.Asaiwlmnnt Rfle foar Event free,: CCNET t:age 4
?DI 5L09 fl4 2/9/2006 spw rcotc
_ flugnat 1wi BM +1rIS + - +U2A t02 T-SHUS:ASSlA + CL&853PP3 + CLSSIEL + CLASSIC + CLSI3A + CLASESS
- CLJSSIC)
VF3 Ttr F*(CLASSIA + £CASSIC 4 ctLASSW . CLASSIA + CLASSIX 4 CLASSC) 204 2t-FCC1SlE + CLASIEL) ffRl Dw"S BXES CLASBIBL Coments TD?-SCw9SP9UY*RMSCOOL This was an assumption that resulted In 100 llBE PIS7 Or*Fr RME:S 01a8*TD-r*FvD S P1:JAF 1 LtDO 1 Coments .20-0 fQLS LEVEL 1; L2O1 IKPLISS LEVEL2: USE MFF TO CHAWE ALT Cisa + CASSliI 4 CLASSiC
- CASSID 4 CLASSIX 4 CLSS3IA CWlSS3S '
CLASSIc A=O ,. C 4 CLASSI L i CLASSZA + CLASS8T + CLASS2V + (CLASSI + CLASS4 4 CLASSSJ 4 S. Cwins=aats CLASS 30 P.13 CLASS 4 ABtE RVALtATEID r LEMV CILUF CCILAIL 01X CLASSIIL 4. CLKSS3E + CLASS3C 4 011 CLASS2A + CLASs2A 4' NORVw(CXSSIA. + CLASSIBE i- tASS+ CT-CLASSIC)+ cLASSIOA(ROAwC + Now0J OI3 -PDOILI (CLASSLA 4 CLASSIC 4 CLASSLD) Cc ents change blot PORESSUS LawU' o02 OPflSni4*(CLASSIA + CSUC + ClASSID) Coments change hIt:4 PMESSURS LRF ' E-49
BEN EPUCOPProbabilisticRisk Assessment Model Name: UlCOP2-9 Split Fraction AAsignmlent Rule for Event Tree: Tl 5 LOS'PH 2/9/2006 Page 5 St Split Fraction Assignment Ruleu IR1 01-F*(CIABSSA + CLASSIC] !a3 CLASSBEE TR4 CLA9S1BL 1R5 0I-*JLA3SJD ZR6 OI-S*CLA3Sl10 Conunents the izgizial U1 .L2 model 1R7 01-F*LASS13 IRS OI-S*CLASSIE IR2 az-s Comments LOW PRESSURE INJEOCTION frILICIT IRS' I Ca2 *IR-F*OI-9 CZ4 IR=F*01-F CZ1 IR-S*0I-S CZ3 IR-s*01I=? CZF 1 TDI CLASS1E TD2 015S*DWSPRAY T03 -(OI-BV*CLASS113E TD4 - {OX-B) *CLASB1BL TD8 OI=F*CLASSIA TDF1 S'Dl ALTINJRHS8X + DWSPRAY FD2 TD-S*' CLASS1A + CLASSIE1D + CLASS1BL + CLASS1O + CLASS3A + CLiSS3SB + Ct.AS83C) F03 TO-r*(cfLAsSiA + CLAssic + CLA~SSI + CYASS3A + CLASS3.B + CLASS3C) F04 TD=F* (CLASSIBE + CLASS1lEL) DWIF' WRl DW-S PM8 CLASSlEL E-50
BFNEPUCOPProbabilisticRiskAssessment Model Name: UlCOP2-9 Split Fraction Assignment Rule for Event Tree. CETNI
*5+/-0t YX 2/9/2006 Page 6 SF Split Fraction Assignment Rule Comments TD-S*DWSPMAY*RHRSPCOOL This was an assumption that resulted in 100 RBE RME7 oI-tI
- b5 O6 OIS*TD rD S*DWS-S RPME5 OZ-S*TD-S*FD-S*DWS-F DM54 O-3*TD-S*rD=F RME3 0I-S*TD-F*FD-F RYEF 1 L20 1 Co03nents L20-0 ZXPLXES LEVEL 1; L20-1 IMPLIES LEVEL2; USE FDr TO CHANGE ALF CLASS1A + CLASS1BE + CLASSIC + CLASS1D + CLASS1E + CLASS3A + CLASS:IB +
CLASS3C ALO NOCD + CLASS1BL + CLASSSA + CLASS2L + CLASS2T + CLASS2V + (CLASS3D + CLASS4
+ CLASS5) + BUCKET Comments CLASS 3D AN)4CLASS 4 ARE EVALUATED FOR LERF CILD? CILFAIL CILDS 1 OIS CLASS3A + CLASS3B + CLASS3C.+ LOW OIl CLASS2A + CLASS2T + NORV'*(CLAS1A 4+CLASSiBE + CLASS18L+ CLASSIC) +
CLASSlB*(NOACREC + NODC) OI4 MCASSIE 013 -OPDEPLI*(CLA8SIA + CLAssIC + CLASS1D) Comments changel hIGA PRESSURE LERF 02 OPDEPLI*(CLASSIA + CLASSIC + CLASS1D) Comments changel hIGli PRESSURE LERF IR1 OI-F*(CLASSIA + 'CLASSIC) tR3 CLASS1BE IR4 CLASSlBL. IR5 Ox-r*CLASS1D IR6 OI-S*CLASS1D Comments the irginal U;1 L2 model Z5R7 OI-F*CLASS1E IRS OI-S*CLASS1E E-51
BFNEPUCOP ProbabilisticRisk Assessment Model lYams: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI S:CI9 PK 2/9/2006 Page 7 SP SBp~liArzacticn kssignlmant. Rule 1R2 01-S Comments LO0WPRESSURE INJECTION MIMLICIT ARF C22 XER=F*0I..r CZ4 cz1 IR-S*'01-S CZ 3 IR..S*01-F Czr TDl CLASS1U TD2 01-S*DWSPRAi T03 -(01-B) *CLA9S1BE'
- (OX-B) *CLPJSlBL TD4 OI-rt*CLASS1A TD8 FDl ALTIN~JRSW + DWSPRAY TD=S*(CLASSIA + CLASSLIE + CLAS81BL + CLASS1D + CLASS3A + CLASS3B4 CLASS3C)
WD2 TD-F*(CLA9s1A + CLASSV: + CLASSID + CLAS93A + CLASS3E + CLASS3C) r04 TD-F* (CLASSIBEa + CLASS IBL) aWl'. WA' Dw-s CLASSIBL Comrments TD-S*DWSPIMAY*BHRSPCOOL This was an assumption that reaulted in 100 RBE RME7 oi-r RME6 OIS~9TD--S*F.D5 *DWS-B 01=S.*TD-.S*Efl-S*DWS-F 01-S*TD-S*WFl-F AM3 O01-S*TD-F*WD-F KEMS' 1 1.20 1 E-52
BFNEPUCOPProbabilisticRiskAssessment Modal Name: UlCOP2-9 Split Fraction Assignsrert Rule for Event Tree: CETNI 5:09 WM 2/9/200f Wage a Sp Split Fraction Asaignmkent Rule Comments L20-0 IMELIES LEVEL 1; L20-1 IMPLIES LEVEL2; USE MFF TO CHANGE ALF CLASS1A + CLASS1BE + CLASSIC + CLASS1D + CLASSlE + CLASS3A + CLASS3B + CLASS3C ALO NOCD + CLASS1SL + CLASS2A + CLASS2L + CLASS2T + CLAS52V + (CLASS3D + CLASS4
+ CLASSS) + BUCKET Comments CLASS 3D AND CLASS 4 ARE EVALUATED FOR LERF CILDF CILFAIL CILDS 1 OIS CLASS3A + CLASS3B + CLASS3C + LOW OIl CLASS2A + CLASS2T + NbRV*(CLASSIA + CLASSIBE + CLASS1BL+ CLASSIC) +
CLASS18*(NOACREC + NODC) 014 VLASS1E 013 -OPDEPL1*(CLASS1A + CLASSIC + CLASS1D) Comments change! hIGH PRESSURE LERr OI2 OPDEPL1* CLASSIA + CLASSIC + CLASSlD) Comnments chagige hIGH PRESSURE LERF TR2 OI-3 *(CLASSLA + CLASSIC) _R3 CLASSIBE IR4 CLASS1BL R5 01-F*CLASS1D IR6 OI-S*CLASS1D Comments the irginal Ul L2 model IR7 O1-F*CLASS1E IRS OI-S*CLASS1E IR2 01-S Comments LOW PRES:3URE INJECTION IMPLICIT IRF 1 CZ2 IR-E*OI-S CZ4 IR-F*OI-F CZ1 IR-S*OI-S CZ3 IR-S*OI-F CZF 1 E-53
BFNEPUCOPProbabilisticRiskAssessment Model 9aime: UlCOP2-9 Split Fraction Assignment Rule for Event Tree: CETNI 5:0!9 NX 2/9/2006 Page 9 SF Split Fraction assigzmwx.t Rul-TD1 CLASSIE TD2 0I-S*DWSPRAY TD3 -(OI-B)*CLASS1BE TD4 - (01=B) *CLASS1BL TD8 OIF*CLASS1A TDF 1 FDt ALTIVZWJSW + DWSPSAY FD2 TD-S*CCLASS1A + CLA6S1BE + CLASS1BL + CLASSlD + CLASS3A + CLASS3B + CLASS3C) FD3 T6-F*(CLASSIA + CLASSIC + CLASSID + CLASS3A + CLASS33 + CLASS3C) FD4 TD-F*(CLASSlBE + CLASS1IL) DWIF 1 WR1 DW=S RME8 CLASSlBL Conments TD-S*DVSPEAY*UHRSPCOOL This was an assumption that resulted in 100 RBE RME7 O1-F RMES OI-S*TD-S*FD-S*DWS-S RM35 OI-S*TD-S*FD-S*DWS-F Rfii4. OI-S*TD-S*FD-F VE33 6OI-S*TD-F*FD-F RMBr1 E-54
BFNEPUCOPProbabilisticRiskAssessment Model Nape: UlCOP2-9 Macro for Evmzt Tree: CETN1 8:03 PE 2/1/2006 Page I mtacro Macro Rule / Commets CIC3LERF CZ=F + RM-F* (CILFgL+XI=F+IR-F*TDS*F>S) CZ-F + RME-F* (CILFAIL+nwI"r+IR-F*T>DS*FD-S) CZ-: + RME-Er (CILFIL+"IF+IR-F*TDS*FDS) CZ-F + RME-F (CILrAIL+DWI-F+IR-F*TDBS*FFDS) CZ-F + RME-F* CILFAIL+DWI-F+IR-Fi*TD-S*rD-) E-55
BFNEPUCOPProbabilisticRiskAssessment Appendix F FAULT TREES This appendix provides print-outs of the BFN Unit 1 PRA modified containment isolation fault tree and the NPSH fault tree used in this analysis. F-1 C1320503-6924 - 2/27/2006
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2 BAS~iCFT GS3MDA I TRANWMJ4 S OLAS?!F~ 4 Ram-m 0G1r 2 ORJGATE 4 m mOAT! IOVA 2 ANOIGATE AOYXOIFCWM 4 B&M-FR!f en 2 OlROATE CHA . S TRh4EROlWI 4' OlRGATE _BSA 2 TASORJ 3 (IIU3ATE. 034II 3 O -GATE CIJCY1IS 4 OR-GATE GMSNA 3 TRMIWSOIUT O"S CLCVw71 4 OR-GATE ODOMA 2 TRIAIMFRIU CEFCWTM 4 ORLCAT! eOIMIS 2 OR.GATE 4 OIALATE WP 1 CUMELEVEST 3 TRANSEROU PCA 2 HCUIEJEVT usa CL7X 1 TRANMLaN PCA 2 HOUBRJVENT CONP 4 AID-GATE 4 TANh LCT nsw C*ns 1 1 TPNSFEUM comcwl~m Co-R 4 A40-CATE mtWI(uR .
- C~tPII ' 4 AtlDGATE I TRSAIR IN I BASCFMff 3 BASCEVDT
CONDMONS PREVENTING ECCS NPSH FOR LLOCA CASES NPSH CASE 1: RX POWER ATI 102%
! RIVER WATER IGREATER THAN 89F R[VER89 KVSCAUBRATIO-N RIVER WATER l ERROR RESULTING IN' GREATER THAN 70F ACTUAL POWER 102%
71 ZHECCL l R!ER70 _& i I
SYMBOL NAME P# SYMBOL TYPE 1 ANDGATE NPSH 1 ORPGATE RIVER70 1 BASIC.EVENT RIVER89 1 BASIC-EVENT ZHECCL 1 BASIC EVENT '1 N%)}}