ML060880460
| ML060880460 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/23/2006 |
| From: | O'Grady B Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC3812, TVA-BFN-TS-431 | |
| Download: ML060880460 (39) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 Bian O'Grady Vice President, Browns Ferry Nuclear Plant March 23, 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 - REQUEST FOR EXTENDED POWER UPRATE OPERATION- RESPONSE TO NRC REQUESTS FOR ADDITIONAL INFORMATION REGARDING CREDIT FOR CONTAINMENT OVERPRESSURE (TAC NO. MC3812)
On June 28, 2004 (ADAMS Accession No. ML041840109), TVA requested a TS change to allow Unit 1 to operate at extended power uprate conditions. As part of this TS change, TVA requested approval to take credit for containment overpressure in order to provide adequate net positive suction head (NPSH) to the low pressure Emergency Core Cooling System (ECCS) pumps. NRC Request SPSB-A.11 from NRC letter dated October 3, 2005 (Accession No. ML052430341),
requested TVA to provide additional information regarding NPSH for the low pressure ECCS pumps, including an assessment of the credit for containment overpressure against the five key principles of risk-informed decision making in Regulatory Guide 1.174.
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U.S. Nuclear Regulatory Commission Page 2 March 23, 2006 TVA's submittal dated February 28, 2006 (ADAMS Accession No. ML060620328) provided the response to the NRC's request.
During preparation and final review of TVA's submittal dated March 7, 2006, a legacy error was discovered in the existing design calculation which determines the available low pressure ECCS pump NPSH requirements. The error has been documented in BFN's Corrective Action Program, and the calculation has been revised. The effect of the error is small; however, it is non-conservative and impacts numerical values that were provided in the original EPU submittal and in the February 28, 2006 submittal. Additionally, the errcr impacted values that were needed to respond to questions ACVB.17, ACVB.18, ACVB.26, and ACVB.32 from the Request for Additional Information (RAI) provided by the NRC in a letter dated December 22, 2005 (Accession No. ML053560120).
This letter provides a revised response to question SPSB-A.11 from the NRC letter dated October 3, 2005, and supersedes the response to this question submitted by the IVA letter dated February 28, 2006.
Additionally, this letter provides the responses to questions ACVB.17, ACVB.18, ACVB.26, and ACVB.32 from the NRC letter dated December 22, 2005, as noted in TVA's submittal dated March 7, 2006.
TVA has determined that the additional information provided does not affect the previously provided no significant hazards considerations associated with the proposed license arid TS changes.
If you have any questions about this submittal, please contact William D. Crouch at (256) 729-2636. I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 23rd, 2006.
cc: See page 4.
U.S. Nuclear Regulatory Commission Page 3 March 23, 2006
Enclosures:
- 1. Response to Question SPSB-A.11 From NRC Letter Dated October 3, 2005
- 2. Detailed Chronology of Correspondence Related to the Previous Approval of Containment Overpressure for Pre-uprate Conditions
- 3. Detailed Description of Plant Systems Related to the NPSH Analysis
- 5. Response To Questions ACVB.17, ACVB.18, ACVB.26, and ACVB.32 From NRC Letter Dated December 22, 2005
- 7. Containment Calculation for NPSH Evaluations
- 8. Markup of PUSAR Section 4.2.5 Indicating Changes Associated With Revised NPSH Calculation
U.S. Nuclear Regulatory Commission Page 4 March 23, 2006 cc (w. 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 Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 Mr. Malcolm T. Widmann, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Unit 1 Restart 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
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 1 RESPONSE TO QUESTION SPSB-A.11 FROM NRC LETTER DATED OCTOBER 3, 2005 On June 28, 2004 (ADAMS Accession No. ML041840109), TVA requested a TS change to allow Unit 1 to operate at extended power uprate conditions. As part of this TS change, TVA requested approval to take credit for containment overpressure in order to provide adequate net positive suction head (NPSH) to the low pressure Emergency Core Cooling System (ECCS) pumps.
NRC Request SPSB-A.11 from NRC letter dated October 3, 2005 (Accession No. ML052430341), requested TVA to provide additional information regarding NPSH for the low pressure ECCS pumps, including an assessment of the credit for containment overpressure against the five key principles of risk-informed decision making in Regulatory Guide 1.174.
TVA"S submittal dated February 28, 2006 (ADAMS Accession No. ML060620328) provided the response to the NRC's request. During preparation and final review of TVA's submittal dated March 7, 2006, a legacy error was discovered in the existing design calculation which determines the available low pressure ECCS pump NPSH requirements. The error has been documented in BFN's Corrective Action Program; the calculation has been revised arid is provided as Enclosure 4. The effect of the error is small; however, it is non-conservative and impacts numerical values that were provided in the original June 28, 2004, EPU submittal and in the February 28, 2006, submittal.
This enclosure (and supporting information in Enclosures 2, 3, and 4) provides a revised response to question SPSB-A.1l, and supersedes the response to this question submitted by TVA letter dated February 28, 2006.
Change bars indicate areas of revision in the body of this Enclosure and Enclosure 4. Due to the number of changes made to Table SPSB-A.11-2 of this enclosure, change bars are not practical. The contents of Enclosures 2 and 3 remain unchanged from Enclosures 2 and 3 of the February 28, 2006, submittal; they are provided herein for completeness.
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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-OO1, 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 chancre 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 assessmentithat demonstrates that the proposed containment accident pressure credit meets the numerical risk acceptance guidelines in Section1 2.2.4 of RG 1.174. This quantitative risk assessment must include specific containment failure mechanisms (e.g., liner failures, penetration failures, primary containment isolation system failures) that cause a loss of containment pressure and subsequent loss of NPSH to the ECCS pumps.
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TVA Response to SPSB-A.11 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 PRA 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 t:o 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 COPy An in-depthidiscussion o:
the regulatory background is provided in Enclosure 2.,
Previously, BFN Units 2 and 3 installed new large capacity ECCS strainers to meet the requested actions of NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors." As part of the 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.
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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 (CS) pumps from approximately 5,500 to 35,000 seconds (a duration of about 8.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) following a LOCA (long-term requirement). For EPU, TVA is requesting, for all three BFN units, approval for crediting 3 psi of COP for RHR pump NPSH for the first 10 minutes following a LOCA (short-term requirement) and 3 psi of COP for CS pump NPSH from approximately 2,600 to 64,000 seconds (a duration of about 17.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) following a LOCA (long-term requirement).
As part of the EPU effort, BFN has also given more consideration to 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 andJ97-04 and are not addressed in Regulatory Guide (RG) 1.82. Conservative evaluation of these events determinedithat BFN will credit available containment pressure for the RHR pumps following an Appendix R, ATWS, and SBO events.
SYSTEM DESCRIPTION The following provides an abbreviatedisystem description. An in-depth description of the BFN containment and ECCS systems is provided in Enclosure 3. The BFN units are BWR-4sJwith Mark -I containments, which incorporate a large torus shaped suppression pool. Four RHR pumps and four CS 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 E1-4
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 CS 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 CS 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 0F, and only the RHF pumps require credit for COP in order to have sufficient NPSH margin as shown iniFigure SPSB-A.11-1.
LONG TERM (T>10 minutes)
At 1() minutes, operator action is assumed which places the minimum complement of ECCS pumps into modes required for long--term cooling. Two CS pumps (one loop) at design flow of 3,125 gpm each are assumed for core cooling, and two RHR pumps (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 tc single failure considerations. ECCS strainers are assumed to accumulate the maximum equilibrium debris load. During this time, suppression pool temperature reaches 187.4 0F, and only the two CS pumps require credit for COP in order to have sufficient NPSH margin as shown in Figure SPSB-A.11-2.
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REGUIATORY 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.
TVAiS assessment of the credit for containment accident pressure against the five key principles is as follows:
- 1. CURRENT REGULATIONS On June 28, 2004 (Accession No. ML041840109), TVA submitted a Technical Specificationsl(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 andregulations, 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 6.5.5 of the BFNT UFSAR. I The approval of credit for post-accident COP is consistent with the NRC's Final Policy Statement on the Use of E1-6
Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities, 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. The credit is supported by the BFN PRAs and the results satisfy the numerical targets contained in RG 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 governirg heatup of the suppression pool.
RG 1.174 provides guidance for acceptable methods to assess defense-in-depth principles. The following addresses the aspects of defense-in-depth that are potentially impacted by the requested change:
- Capability of Containment to Provide Containment Overpressure 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2% 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 to 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 inidesign basis events is determined crediting only two of the four RHR pumps and heat exchangers. Emergency Operating Instructions (EOIs) dictate using all available RHR 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 E1-7
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 166.4 0 F with four RHR pumps operating (Table SPSB-A.11-2, Case 1). This analysis was performed using the same conservative assumptions for input parameters as the licensing basis analysis.
Other single failures that could result in 3 RHR pumps operating would yield a peak suppression pool temperature of 170.50 F [Table SPSB-A.11-2, Case Ic (TVA)], and positive NPSH margin would be maintained long term without COP. This analysis was performed using conservative plant parameters for reactor power, SW temperature (exceedance probability of 8.46E-02 for 90'F per Table C-1 of Enclosure 4), suppression pool initial temperature (exceedance probability of 1.01E-02 for 950 F per Table C-2 of Enclosure 4), and initial suppression pool water volume
[A conservative probability value corresponding to -5.7C0" (123,500 ft3) instead of -5.90" (123,250 ft3) per Table C-3 of Enclosure 4].
CS pumps require credit for COP when suppression pool temperature exceeds 171 0F. The RHR pumps do not require COP at the peak pool temperature of 187.30 F. Per page i-5 of Enclosure 4, "... The likelihood of failing any two RHR pumps during the 24-hr PRA mission time is approximately 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 l:o the accident analysis methodologies.
- 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. Table SPSB-A.11-1 provides the parameters of interest, the values used in the safety analysis and the associated realistic values.
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TABLE SPSB-A.11-1:
PARAMETERS OF:INTEREST, LICENSING BASIS VALUE, AND REALISTIC VALUE LICENSING REALISTIC PARAMETER BASIS VALUE VALUE COMMENT Initial Power 102% 100% Related probability Licensed Licensed information is provided Thermal Thermal in Enclosure 4.
Power Power Decay Heat ANSI 5.1 ANSI 5.1 Model (plus 2a) (w/o 2a)
Service Water 95 65 to 92 Exceedance probabilities Temperature for service water (OF) temperature are provided in Enclosure 4.
Initial 95 ; 85 to 92 Exceedance probabilities Suppression (TS for initial pool Pool maximum) temperature are provided Temperature in Enclosure 4.
(OF) .
Heat 223 241 Based on realistic Exchanger fouling factor of 0.0020 K Value vs 0.0025 and maximum (BTU/sec-OF) number of tubes plugged (1.5%) I 225 1.5% tube plugging onLy Initial 121,500 123,250 Exceedance probabilities Suppression (TS to for initial pool volune Pool Water minimum) 125640 are provided in Volume Enclosure 4.
(ft3 ):
Containment Assumes no Includes Heat sinks are always Heat Sinks heat sinks realistic present but not normally heat sinks credited.
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Sensitivity analyses were performed (with selected analyses verified) which are summarized in Table SPSB-A.11-2. The purpose of these analyses was to identify input parameter combinations where COP was not required (e.g., suppression pool temperature below 171'F).
- 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 (171'F). Table SPSB-A.11-2 provides realistic cases ancd designates those for which COP is not required. 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 CS pumps will not perform their function at NPSH Available (NPSHa) values less than the manufacturer's required NPSH (NPSHR). The values used were derived from manufacturer's 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 23.5 feet NPSHa (broken loop) versus 30 feet NPSHR or 24.6 feet NPSHa (intact loop) versus 26 feet NPSHR. Negative NPSH margin of this magnitude for short periods of time will not prevent the RHR pumps from performing long-term in thelevent.
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 at 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 manufacturer's curve for at least 10 minutes without damage. This data demonstrated that the RHR pumps have El-10
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.
In the long-term, operator action is credited and two RHR pumps (one loop in containment cooling mode) at 6,5000 gpm each are assumed for pool cooling.
By comparison to the RHR pumps, the CS pumps would be challenged in the long-term scenario in the event that COP was lost. CS pumps do not require COP in the short-tern.
(Refer to Figure SPSB-A.11-1). BFN CS pumps are Sulzer-Bingham model 12x16x14.5 CVDS. Assuming no credit for COP, the CS pumps used for long-term core cooling
(>10 minutes) would be expected to operate between 27 feet and 21.2 feet of NPSH verses 27 feet used in NPSH calculations for approximately 17.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> as suppressior.
pool temperature peaks above 171 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 CS loops is required to be operated for adequate core cooling and the non-operating CS 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 CS 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.
- 4. RISK ASSESSMENT TVA evaluated the risk impact of utilizing COP to satisfy the NPSH requirements for RHR and CS pumps'to mitigate the consequences of a DBA LOCA. The risk assessment evaluation used the current BFN Unit lPRA internal events (including internal floods) model. The evaluation is provided as Enclosure 4 to this letter.! The steps taken to perform this risk assessment evaluation were:
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- a. 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;
- b. Revise all large LOCA accident sequence event trees to make low pressure ECCS pumps dependent upon containment isolation when other plant pre-conditions exist (i.e., Service Water initial high temperature, Suppression Pool initial high temperature);
- c. Modify the existing Containment Isolation System fault tree to include the probability of pre-existing containment leakage;
- d. Quantify the modified PRA models and determine the change in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF); and
- e. Perform modeling sensitivity studies and a parametric uncertainty analysis to assess the variability of the results.
Per page iii of Enclosure 4, "The conclusion of the plant internal events risk associated with this assessment is as follows.
- 1) ... the proposed change (i.e., use of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in CDF (1.4E-09/yr).
- 2) ... the proposed change (i.e., use-of COP to satisfy the net positive suction head (NPSH) requirements for RHR and Core Spray pumps) represents a very small change in LERF (1.4E-09/yr)."
This very small increase was well below the guidelines provided in RG;1.174.
ATWS, SBO, and:Appendix R are highly unlikely event scenarios which areidefined by failure of multiple features. Failure assumptions inlthese 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 a thermodynamic result of the event itself provided that containment integrity is maintained.
This is acceptable given the low probability of the events.
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- 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. Per TS LCO 3.6.2.6, "The drywell pressure shall be maintained 1.1 psid above the pressure of the suppression chamber." Per TRM LCO 3.6.5, "When the primary containment is inerted the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements. Nitrogen makeup tc the primary containment, averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for drywell temperature, pressure, and venting operations), shall not exceed 542 scfh." Per TRM Surveillance Requirement (TSR) 3.6.5.1, "When the primary containment is inerted, the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements." The frequency for this TSR is "24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." Satisfying these requirements would identify any pre-existing leak in the drywell portion of containment.
10 CFR 50.54(o) and 10 CFR 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.
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- 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 to5 monitor NPSH using an NPSH limit curve, showing pump flow versus suppression pool temperature fcr various suppression pool pressures. The EOIs also list additional indications of inadequate NPSH; see TVA Reply to ACVB.23 in TVA letter to NRC dated March 7, 2006, for details. Operators are trained on these procedures as part of their periodic re-qualification program.
RG 1.174 CONCLUSION The use of COP to ensure an adequate NPSH for ECCS pumps during a limited time after aldesign 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-informediassessment 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 very small increase in CDF and LERF of 1.4E-09/yr per pace iii of Enclosure 4. IThis very small increase is well below the guidelines provided in RG 1.174 (100 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.
E1-14
REFERENCES
- 1. TVA letter, T. E. Abney to NRC, "Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - License Amendment Regarding Use o:r 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 (RikI)
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
- 3. NRC letter, W. 0. Long to TVA, "Browns Ferry Nuclear Plants, Units 2 and 3 - Issuance of 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
- 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 24, 1976 E1-15
TABLE SPSB-A.11-2 SENSITIVITY ANALYSES OF VARIOUS REALISTIC INPUT PARAMETERS DURING THE DESIGN BASIS ACCIDENT LOSS OF COOLANT ACCIDENT LONG-TERM PHASE U_ 0 E O -. 0 c 0.5 9 0C U 0 .e0 ECE EOD L>Ua 3 0 Base EPU Licensing 102% ANSI 95 95 2 full 2 2 4,000 1 223 2 121,500 Yes No 187.3 Yes Case** Calculation - DBA EPU 5.1_ design (GE) LOCA (wI2a)
Case 1** DBA Calculation but No 102% ANSI 95 95 4,000 121,500 Yes No 166.4 No (GE) Single Failure EPU 5.1 Case la** DBA Calculation but 102% ANSI 95 95 4,000 121,500 Yes No 175.0 Yes (GE) 3 RHR Pumps in EPU 5.1 Suppression Pool - (w/20)
Cooling - -
Case la DBA Calculation but 102% ANSI 95 95 4,000 121,500 Yes No 175.0 Yes (TVA) 3 RHR Pumps in EPU 5.1 Suppression Pool (wI2a)
(This case is benchmarked Cooling against Case la (GE).]
Case lb 100% Initial Power, 95 4,000 121,5001 Yes No 171.0 No (TVA) RHRSW Temp 890F, 3 RHR Pumps in Suppression Pool Cooling, K Value 225, 4 CS Pumps E1-16
TABLE SPSB-A.11-2 SENSITIVITY ANALYSES OF V.ARTftUS REAJJT STCT mipri'r VPPAMrTPE nTTPTKhIG THE DESIGN BASIS ACCIDENT LOSS OF COOLANT ACCIDENT LONG-TERM PHASE I I k 2 eC)
-a 9 U-X.C.0 M W
Zs C 8
- a. 9 CLa Ik 0:0 C,)
W --
0 0 C*
- 0) CL o: I: Cn 0)
E C0
.SEa a0- iE.s Ea
(- LF I a.
c.G 0C o a 3:
0
= zo 0: E 6
?!
CL E C ZN _;
0*
C ta. il M . 'F 'n = Ic 3! a/)
W~a .ea.
co E a--
Case* Case Description Z iB 00 0 = 00 i40 Case Ic 100% Initial Power, 95 full 4,000 Yes No 170.5 No (TVA) RHRSW Temp 900F, design 3 RHR Pumps in Suppression Pool Cooling, K Value 225, 4 CS Pumps, Nominal Suppression Pool Water Level Case 1d DBA Calculation but full 4,000 Yes No 171.0 No (TVA) RHRSW Temp 90°F, design Suppression Pool Initial Temp 91°F, 3 RHR Pumps in Suppression Pool Cooling, K Value 225, 4 CS Pumps Case le DBA Calculation but . 102%- ANSI - full 4,000 121,500 Yes No 171.1*** No (TVA) RHRSW Temp 92°F, EPU 5.1 design Suppression Pool Initial (wI2a)
Temp 9 0°F, 3RHR- -
Pumps in Suppression Pool Cooling, K Value 225, 4 CS Pumps Case 2 DBA Calculation but 102% ANSI 2 4,000 223 I 121,500 Yes No 182.0 Yes (GE) Initial SW Temp 85°F EPU 5.1
_ _ _ __ __ _ _ _ _ __ ___ ___ ____ ___ _ __ _ _ (W I2 ca)
E1-17
TABLE SPSB-A.11-2 SENSITTVTTV JZTM.SES OF VZO-IT7US . RWEALISTIC INPUT PRkIMETERS DULRING THE DESIGN BASIS ACCIDENT LOSS OF COOLANT ACCIDENT LONG-TERM PHASE m
- I - I imp~~ - y y. . m . .
m-- - F a U..
E C) 0.
.S 8.
CL - E Se -8 _= k U)
CLi CL a o is 0 a CL C9
- 0. -m C 0 0 to I~
z U)I a)- coa.. U) a 9i .~ U)-
(U 0. C.) .0 a E As
!2 Case* Case Description z 0 0 0 W q 00 Case 2 DBRACalculation but 102% ANSI 95 2 full 2 2 4.000 223 2 121,500 Yes No 182.2 Yes (TVA) SW Temp 850 F EPU 5.1 design
[This case is benchmarked against Case 2 (GE).] _
Case 2a DBA Calculation but 102% ANSI 95 2 full 2 2 4.000 223 2 121,500 Yes No 177.6 Yes (GE) Initial SW Temp 750 F EPU 5.1-- design (wI2a)
Case 2b** DBA Calculation but 102% ANSI 95 2 full 2 2 4,000 223 2 121,500 Yes No 175.9 Yes (GE) Initial SW Temp 700 F EPU 5.1 design (wI2a)
Case 2c DBA Calculation but 102% ANSI 2 full 2 2 4,000 223 2 121,500 Yes No 174.3 Yes (GE) Initial SW Temp 650 F EPU 5.1 design
__ ___ ____ ___(w /2a)
Case 2d DBA Calculation but 102% ANSI -2 full 2 2 4,000 223 2 Yes No 170.6 No (TVA) SW Temp 65 0 P, - - - EPU 5.1- -- - design .
Suppression Pool Initial (w/2a)
Temp 880F, Nominal Suppression Pool Water Level E1-18
TABLE SPSB-A.11-2 SENSITIVrITY Av-h-LY SES OF 1r7^PDTrnTI 'DVU T TCrPTf' TINU1TT MA'7%.MT'VRVV flT1TV)ITN THE DESIGN BASIS ACCIDENT LOSS OF COOLANT ACCIDENT LONG-TERM PHASE p p-i p p- p- p - w w- * .-
- C p p p 0
.e aR
'a K
3 o Ea 8 E 85 C 0 C 0.
.a'fl *D 0. Q I i , -- = 8 C
o! 9 tn E C
.C W aW 0 a0.
a 0 a o h
- ro E o~ =
0 L
&.e o 8 zo ie*° Q.
,E a E Me as 9 I @_
C CO 0 Ts i C-qay
.2C 'a a; co a. a, a) co
- Cu ole XOc 4- I 0n 00 CK Z., ,E j IE M i IM C. - o 0 Case* Case Description Z O p- p-Z i Z .5 ilo -
00 UJ p p p p p p - p - p - I-Case 2e DBA Calculation but 102% ANSI 2 full 2 2 4,000 223 2 121,500 Yes No 170.7 No (TVA) SW Temp 650F, EPU 5.1 design Suppression Pool Initial (w/2a)
-Temp 87°F-Case 2f DBA Calculation but 102% ANSI 2 full 2 2 4,000 223 2 Yes No 171.1* No (TVA) SW Temp 68°F, -EPU 5.1 - design Suppression Pool Initial (w/2a)
Temp 870F, Nominal -
Suppression Pool Water Level Case 3 DBA Calculation but 2 full 2 2 4,000 223 2 121,50 Yes No 183.8 Yes (GE) Initial Suppression Pool design Temp 85°F Case 4 100% Initial Power, 2 full 2 2 4,000 2 121,500 Yes No 1 (GE) Minimum Suppression design Pool Level, and No Heat Sink Credit El-19
TABLE SPSB-A.11-2 SESTITIVTYV 77.ATVYSVO OF T7V.TrT7Q 1DLIT TaCq'Tt ThOTTr pAD7%ME-P^C DT7MTX)7T TH E a I - - - LOS OF- C.LANT ACe a.E. L - _.- PH AS THE DESIGN BASIS ACCIDENT LOSS OF COOLANT ACCIDENT LONG-TERM PHASE p m . ,- Y -l - I- I- Y -# II-I - Y-
.'C a S-Li. 0. co 0. 8 c E 0 ca- 0 C
- 0. .C 9 S C 3:: Q8 Q C o O ICD, co oa)
.5 0 co.E E ,E o, 00 8 COc C
$0 8 co M- u a- 0
.5 C I I a) ° I C 8D Q) = Cu gn
.9 82 c
W aL 9_ .~
Q 5 8o0 v)
Eu 0'0 co.
C) co O
.U) =
ii Pu c
ot of (D , = off C Case* Case Description .S r_ z 0 W 0 =
- * - b - I - I - i -
Case 4 100% Initial Power, 2 full 2 2 4,000 2 121,500 I Yes No 177.1 Yes (TVA) Minimum Suppression design
[This case is Pool Level, and No benchmarked Heat Sink Credit against Case 4 (GE).]
Case 4a 100% Initial Power, 2 full 2 2 4,000 2 Yes 174.7 1 Yes (GE) Nominal Suppression design Pool Level, and Heat Sink Credit Case 4b** 100% Initial Power, 2 full 2 2 4,000 2 1i2"1,500 Yes 178.9 1 Yes (GE) Minimum Suppression design Pool Level, and Heat Sink Credit Case 4c** 100% Initial Power, 2 full 2 2 4,000 2 121,500 Yes 175.8 1 Yes (GE) Minimum Suppression design Pool Level, Heat Sink Credit, and SW Temp that results in Peak Suppression Pool Temp
_ _176°F E1-20
TABLE SPSB-A.11-2 SENSITIVITY ANALYSES OF VARIOUS REALISTIT INPUT PARAMETERS DURING THE DESIGN BASIS ACCIDENT LOSS OF COOLANT ACCIDENT LONG-TERM PHASE I -- I - I I - I - I Y - Y - Y ~
c
.0 2
- 0. U..Q Q.
Cu C
-a 3/) a a C E.
0.
0 0
.S. f,, - U.. -9
-cn L t3 16 0 fla to U0 U).-
9 (SI co X W COm 0- .U 0C 0.
CL U) a 0 co a--
-0 CD t- al W.G 0 U) CuE Case Case Descrpton Case 4d 100% Initial Power, 2 full 2 2 4,000 2 121,500 Yes No 177.0 Yes (TVA) -RHRSW Temp 860F,- _ design Suppression Pool Initial Temp 920F, KValue 225 Case 4e 100% Initial Power, 2 full 2 2 4,000 2 121,500 Yes No 176.1 Yes (TVA) RHRSW-Temp 860F, design Suppression Pool Initial Temp 900F, K Value 225 Case 4f 100% Initial Power, 2 full 2 2 4,000 2 Yes No 175.6 Yes (TVA) RHRSW Temp 860F, design Suppression Pool Initial Temp 90 0F, K Value 225, Nominal Suppression Pool Water Level Case 4g 100% Initial Power, 2 full 2 2 4,000 2 Yes No 173.1 Yes (TVA) RHRSW Temp 860F, design Suppression Pool Initial Temp 90 0F, K Value 241, Nominal Suppression Pool Water Level E1-21
TABLE SPSB-A.11-2 SENSITrVTrTV hNhT-Yr.S OF VhInrJT7 P'^.TIC INP1UT PAF.MvTE DTjTNG THE DESIGN BASIS ACCIDENT LOSS OF COOLANT ACCIDENT LONG-TERM PHASE pm p- p- y - p- - I - - p - a) I - I- I - I - - I- u-S.g cO a
E .S CL W C E CL I 1= ttm C
- 0. C 8
a-4a) C
.S.2 a 8 E h
= ._ iC, LL Co,
-a a a- 0E 0C3 o: 0 U] t aL-a .5 0 ILC 5 0. O W Q2 E zo 'O t o *. 3E I a)-c CL) =g_
3o 9
Ci)'Z sa c W' o tl_ = -M (0 V5 CO -
o.> a: O E of 3!
c- EJi o E .F 0 = -a Cc) q 'a F
§ c Z sg Case* -Case Description EY .W o
U
'9 C)
Case 4h 100% Initial Power, - _2 full 2 2 4,000 __2 - Yes No 175.1 Yes (TVA) RHRSW Temp 85°F, design Suppression Pool Initial-Temp 900F, K Value 225, Nominal Suppression Pool Water Level -
Case 41 100% Initial Power, 2 full 2 2 4,000 2 Yes No 170.8 No (TVA) RHRSW Temp 85°F, design Suppression Pool Initial -
Temp 860F, K Value 241, Nominal Suppression Pool Water Level Case 4j 100% Initial Power, 2 full 2 2 4,000 2 Yes No 171.0 No (TVA) RHRSW Temp 85°F, design Suppression Pool Initial Temp 88°F, K Value 241, Nominal Suppression Pool Water Level E1-22
- Column information includes designation of organization that accomplished sensitivity analysis.
- Case verified by formal analysis.
- Using a COP break point temperature of 171'F.
- This value is acceptable for demonstrating sensitivity analysis results.
Shaded areas in the Table "highlight" differences from the Base Case.
E1-23
FIGURE SPSB-A.11-1: NPSH REQUIREMENTS FOR DBA LOCA - SHORT TERM Suppression Pool Temperaure N Pre-EPU Peak Suppression Pool Temperture
-Containment Pressure
-- - Atrmospheri Pressure
-- 11RHR Pump Broken Loop Containment Pressure Required A Pre-EPU RHR Pump Broken Loop Containment Pressure Required
-*- RHR Punp LPCI Loop Containment Pressure Required
-I-CS Pump Containment Pressure Required
- Pre-EPU CS Purrn Containment Pressure Reauired a, 140 S.
10 40 E _--
I- I-120 100 I/ I I 1 10 0 100 200 300 400 500 600 Time (seconds)
E1-24
200 i1o j70 II 4E i60-
£I Fe j5°i 0
gA -25
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 1 DETAILED CHRONOLOGY OF CORRESPONDENCE RELATED TO THE PREVIOUS APPROVAL OF CONTAINMENT OVERPRESSURE 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 Cooling 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 tc 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 NPSH E2-1
during and following accidents. TVA also indicated that 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 provide specific information used therein. GL 97-04 requested, in part, that licensees 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 indicated 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):
- Supporting information for these calculations;
- An explanation as to how the analysis at pre-power uprate conditions bounds the uprated conditions;
- A rationale for why the analysis assumed a design flow rate for the LPCS pumps when one RHR pump is in a runout condition; E2-2
- A discussion of the requested overpressure value; and
- Graphs showing the NPSH 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 9.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />).
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 Uprate (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 forlthe LPCS pumps from approximately 4,100 to 52,300 seconds (a durationlof about 13.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) 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 3oiling-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. M96135, 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 for Emergency Core Cooling System (ECCS) Pump Net Positive Suction Head (NPSH) Analyses,"
September 4, 1998
- 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 E2-4
- 9. NRC letter, W. 0. Long to TVA, "Browns Perry Nuclear Plants, Units 2 and 3 - Issuance of 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
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 Specifications (TS) Change TS-431 -
Request for License Amendment - Extended Power Uprate (EP1J)
Operation," dated June 28, 2004 E2-5
ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
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, and 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 steel 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.
E3-1
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 are 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 is 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;
- LPCS; and
The ECCS subsystems are designed to limit clad temperature over the complete spectrum of possible break sizes in the nuclear system process barrier, including 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 looplconsists of two pumps, a spray sparger!inside the coreishroudland abovelthe core, piping and valves to convey waterl 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:
hatm = 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; hlos= 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; and hvrapor = head due to the vapor pressure of the suppression pool water at the suppression pool water temperature.
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.
E3-3
FIGURE 1 GENERAL CONTAINMENT LAYOUT
. REACTOR PRESSURE
! VESSEL DRYNELL IN VENT
-TORUS CRADLE E3-4
FIGURE 2 LAYOUT OF THE EMERGENCY CORE COOLING SYSTEM
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- I LPCI LPCI INJECTION INJECTION VALVE 'VALVE FCV-74-53 FCV-74467 RCIne RECIRC
~1 3 P1W A E3-5