ML052580169
| ML052580169 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  (DPR-033) |
| Issue date: | 09/15/2005 |
| From: | Crouch W Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GL-88-020 | |
| Download: ML052580169 (16) | |
Text
September 15, 2005 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of )
Docket No. 50-259 Tennessee Valley Authority )
BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 - UPDATE TO THE PROBABILISTIC SAFETY ASSESSMENT (PSA)
This letter provides updated BFN Unit 1 PSA results.
TVA has identified and corrected modeling errors in the BFN Unit 1 PSA. The errors did not apply to the Units 2 and 3 models. While correcting the errors, BFN also incorporated enhancements to the model. This resulted in changes to the calculated Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). The Unit 1 PSA results were previously provided to NRC in the context of six topics:
- 1.
Technical Specification Change TS 431, which requested approval to operate at Extended Power Uprate (EPU) conditions;
- 2.
Technical Specification Change TS 426, which revised the diesel generator allowed outage time;
- 3.
Technical Specification Change TS 435, which revised the Containment Atmosphere Dilution System allowed outage time;
U.S. Nuclear Regulatory Commission Page 2 September 15, 2005
- 4.
Generic Letter 88-20, Supplement 4, Individual Plant Examination of External Events;
- 5.
Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities; and
- 6.
License Renewal - Severe Accident Management Guidelines.
A summary of the previous and revised PSA results and a discussion of why the conclusions previously reached for the above topics were not impacted by the small change in PSA results are provided in the enclosure to this letter.
If you have any questions, please contact me at (205) 729-2636.
Sincerely, Original signed by:
William D. Crouch Manager of Licensing and Industry Affairs
U.S. Nuclear Regulatory Commission Page 3 September 15, 2005 Enclosures cc (Enclosures):
(Via NRC Electronic Distribution)
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 Mr. Stephen J. Cahill, 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 Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 Ms. Margaret Chernoff, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Ms. Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
U.S. Nuclear Regulatory Commission Page 4 September 15, 2005 JEW:SMK:BAB Enclosures cc (Enclosures):
B. M. Aukland, POB 2C-BFN A. S. Bhatnagar, LP 6A-C J. C. Fornicola, LP 6A-C D. F. Helms, BR 4T-C R. F. Marks, PAB 1C-BFN R. G. Jones, NAB 1A-BFN B. J. OGrady, PAB 1E-BFN J. R. Rupert, NAB 1A-BFN K. W. Singer, LP 6A-C E. J. Vigluicci, ET 11A-K NSRB Support, LP 5M-C EDMS WT CA - K S:Lic/submit/subs/PSA Update Rev 2.doc
ENCLOSURE TENNESSEE VALLEY AUTHORITY (TVA)
BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 UPDATE OF PROBABILISTIC SAFETY ASSESSMENT (PSA) RESULTS I.
BACKGROUND TVA has identified and corrected modeling errors in the BFN Unit 1 PSA. While correcting the errors, BFN also incorporated enhancements to the model. This resulted in changes to the calculated Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). The Unit 1 PSA results were previously provided to NRC in the context of six topics:
- 1.
Technical Specification Change TS 431, which request approval to operate at Extended Power Uprate (EPU) conditions;
- 2.
Technical Specification Change TS 426, which revised the diesel generator allowed outage time;
- 3.
Technical Specification Change TS 435, which revised the Containment Atmosphere Dilution System allowed outage time;
- 4.
Generic Letter 88-20, Supplement 4, Individual Plant Examination of External Events;
- 5.
Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities; and
- 6.
License Renewal - Severe Accident Management Guidelines.
A summary of the previous and revised PSA results and a discussion of why the conclusions reached previously for the above topics were not impacted by the small change in PSA results are provided below:
E-2 II.
SUMMARY
OF PREVIOUS AND REVISED RESULTS The CDF and LERF values previously provided to NRC and the current values are provided below:
PARAMETER PREVIOUS VALUE (REFERENCE 2)
REVISED VALUE Total CDF (yr-1, mean value) 1.86 E-6 1.77 E-6 LERF (yr-1, mean value) 1.87 E-7 4.40 E-7 BFN performed a detailed review to understand the reasons and accompanying rational for these CDF and LERF value changes. The following is a description of each type of change made and information regarding impact on CDF and LERF for each of the individual model changes accomplished.
The identified errors were in the logic rules regarding system interdependencies and failure effects associated with the Level 1 / Level 2 interface. Specifically, the Level 1 /
Level 2 interface is accomplished through the use of Level l end states that characterize the core damage sequences. Each end state is treated differently in the Level 2 / LERF event trees.
The end states of the Level 2 / LERF event trees are LERF or no LERF. The total of these sequences is the CDF. There were two reasons for the increase in LERF between the model previously described and the present model.
- 1.
A level 1 end state for containment bypass for general transients had been omitted.
- 2.
The Level 1 end states are assigned by logic rules. Some of the logic rules were not complete in the previous model.
This allowed sequences to be assigned to a more benign state than was warranted.
This resulted in the LERF value changing from 1.87E-7 to 4.40E-7.
The Units 2 and 3 PSA models were reviewed to determine if these errors applied. Because of the basic model structuring approach, the errors did not apply to the Units 2 and 3 models.
E-3 The enhancements incorporated into the Unit 1 PSA model included changes in the Level 1 end state logic rules to ensure that the end states were mutually exclusive. Additional reviews of initiating event frequencies resulted in some minor frequency changes. For example, the Turbine Trip ATWS was revised to allow for better scrutiny and understanding. The slight decrease in the CDF was due to the enhancements made to the model.
No plant vulnerabilities were identified nor did the PSA model reflect the need to accomplish any plant modifications. The CDF and LERF changes were reviewed and it was determined that the conclusions in the BFN PSA model and supported documentation were not changed and continue to be acceptable.
III. TS 431, EXTENDED POWER UPRATE The proposed Unit 1 EPU license amendment and subsequent additional information were provided in References 1 and 2. In Reference 1, as part of the No Significant Hazards Consideration, TVA stated that: An evaluation of the Boiling Water Reactor probabilistic risk assessments concludes that the calculated core damage frequency does not significantly change due to Extended Power Uprate (EPU). Details regarding the supporting PSA evaluation were included in Enclosure 4, Section 10.5. Updated PSA information was provided in Reference 2. As discussed in Section II, the changes in the Unit 1 PSA results were relatively small and do not alter the conclusions stated in the EPU license amendment.
In addition, the following tables containing updated PSA information were provided in Reference 2:
- Table 10-5, Frequency Weighted Fractional Importance to Core Damage of Operator Actions.
Updates to these tables are provided below.
E-4 TABLE 10-4
SUMMARY
OF THE INITIATOR CONTRIBUTIONS TO CDF AND LERF FOR BROWNS FERRY UNIT 1 Initiator Category Previous(1)
Mean frequency (events per year)
Revised Mean frequency (events per year)
Previous CDF Revised CDF Previous LERF Revised LERF Transient Initiator Categories Inadvertent Opening of One SRV 4.36E-2 1.36E-2 1.42E-8 3.70E-9 2.28E-11 4.65E-10 Spurious Scram at Power 8.70E-2 8.76E-2 2.98E-8 2.85E-8 5.83E-11 3.43E-9 Loss of 500kV Switchyard to Plant 9.73E-3 1.02E-2 2.38E-8 2.42E-8 1.70E-10 6.43E-9 Loss of 500kV Switchyard to Unit 2.30E-2 2.37E-2 5.14E-8 5.18E-8 4.48E-10 1.51E-8 Loss of Instrumentation and Control Bus 1A 4.10E-3 4.27E-3 9.73E-10 9.76E-10
<1E-12 1.30E-10 Loss of Instrumentation and Control Bus 1B 4.10E-3 4.27E-3 2.93E-8 3.06E-8 6.79E-11 2.61E-9 Total Loss of Condensate Flow 9.09E-3 9.45E-3 3.76E-8 3.88E-8 1.55E-10 5.96E-9 Partial Loss of Condensate Flow 1.80E-2 1.93E-2 5.55E-9 5.71E-9 7.07E-12 7.11E-10 MSIV Closure 5.70E-2 5.52E-2 9.67E-8 9.34E-8 1.21E-9 3.57E-8 Turbine Bypass Unavailable 1.93E-3 1.95E-3 3.10E-9 3.08E-9 2.82E-11 1.19E-9 Loss of Condenser Vacuum 9.72E-2 9.70E-2 1.67E-7 1.67E-7 2.16E-9 6.41E-8 Total Loss of Feedwater 2.58E-2 2.58E-2 5.59E-8 4.55E-8 5.02E-10 1.67E-8 Partial Loss of Feedwater 2.58E-2 2.47E-2 7.94E-9 8.55E-8 1.12E-11 1.01E-8 Loss of Plant Control Air 1.20E-2 1.20E-2 6.58E-8 6.58E-8 2.23E-10 7.44E-9 Loss of Offsite Power 6.43E-3 7.87E-3 2.17E-7 2.70E-7 9.32E-10 5.03E-9 Loss of Raw Cooling Water 7.95E-3 7.95E-3 1.21E-8 1.18E-7 1.73E-10 4.96E-9 Momentary Loss of Offsite Power 7.17E-3 7.57E-3 1.88E-9 1.90E-9 1.08E-12 2.46E-10 Turbine Trip 5.09E-1 5.50E-1 1.86E-7 1.90E-7 4.88E-10 1.70E-8 High Pressure Trip 4.30E-2 4.29E-2 1.40E-8 1.32E-8 2.25E-11 1.63E-9 Excessive Feedwater Flow 2.60E-2 2.78E-2 8.00E-9 8.20E-9 1.13E-11 1.02E-9 Other Transients 3.70E-1 8.60E-2 1.38E-7 2.79E-8 3.50E-10 3.37E-9 1 See Reference 2.
E-5 Initiator Category Previous(1)
Mean frequency (events per year)
Revised Mean frequency (events per year)
Previous CDF Revised CDF Previous LERF Revised LERF ATWS Categories Turbine Trip ATWS N/A 5.50E-1 1.54E-7 5.58E-8 8.18E-8 5.34E-8 LOSP ATWS N/A 7.87E-3 1.86E-9 1.32E-9 9.73E-10 1.27E-9 Loss of Condenser Heat Sink ATWS N/A 1.52E-1 4.77E-8 6.27E-8 2.52E-8 6.04E-8 Inadvertent Opening of SRV ATWS N/A 1.36E-2 1.21E-8 1.14E-9 6.37E-9 1.07E-9 Loss of Feedwater ATWS N/A 3.02E-1 9.87E-8 1.00E-7 5.23E-8 9.64E-8 LOCA Initiator Categories Breaks Outside Containment 6.67E-4 6.67E-4 3.12E-8 3.12E-8 2.06E-10 6.97E-9 Excessive LOCA (reactor vessel failure) 9.39E-9 9.39E-9 9.09E-9 9.09E-9 4.16E-11 4.16E-11 Interfacing Systems LOCA 3.15E-5 3.15E-5 5.00E-8 5.00E-8 5.20E-9 5.20E-9 Large LOCA - Core Spray Line Break Loop I 1.57E-6 1.68E-6 4.37E-9 4.49E-9 1.44E-10 1.55E-10 Loop II 1.57E-6 1.68E-6 4.36E-9 4.49E-9 1.44E-10 1.55E-10 Large LOCA - Recirculation Discharge Line Break Loop A 1.10E-5 1.18E-5 1.85E-8 1.38E-8 1.13E-9 1.20E-9 Loop B 1.10E-5 1.18E-5 1.85E-8 1.38E-8 1.13E-9 1.20E-9 Large LOCA - Recirculation Suction Line Break Loop A 7.85E-7 8.39E-7 4.72E-9 4.67E-9 6.80E-11 8.11E-11 Loop B 7.85E-7 8.39E-7 4.72E-9 4.67E-9 6.80E-11 8.11E-11 Other Large LOCA 1.57E-6 8.39E-7 2.45E-9 8.56E-10 1.42E-10 7.30E-11 Medium LOCA Inside Containment 4.00E-5 3.80E-5 2.13E-8 2.02E-8 4.21E-9 3.99E-9 Small LOCA Inside Containment 5.00E-4 4.75E-4 1.34E-10 9.05E-11 1.08E-11 1.50E-11 Very Small LOCA Inside Containment 3.38E-3 5.76E-3 7.71E-10 1.38E-9
<1E-12 1.79E-10 EECW Flood in Reactor Building -
shutdown units 1.20E-3 1.20E-3 7.41E-11 7.38E-11
<1E-12 3.19E-11 EECW Flood in Reactor Building -
operating unit 1.85E-6 1.85E-6 1.19E-9 1.19E-9 2.10E-12 2.10E-12 Flood from the Condensate Storage Tank 1.22E-4 1.22E-4 1.39E-9 1.38E-9 4.95E-12 3.63E-10 Flood from the Torus 1.22E-4 1.22E-4 4.01E-8 3.98E-8 8.55E-10 2.89E-10 Large Turbine Building Flood 3.65E-3 3.65E-3 5.51E-8 5.52E-8 7.41E-11 2.29E-9 Small Turbine Building Flood 1.65E-2 1.65E-2 1.65E-8 1.54E-8 3.38E-11 1.50E-9
E-6 TABLE 10-5 FREQUENCY WEIGHTED FRACTIONAL IMPORTANCE TO CORE DAMAGE OF OPERATOR ACTIONS USED IN BROWNS FERRY UNIT 1 PRA Database Variable(1)
Operator Action Description Previous Frequency-Weighted Fractional Importance to Core Damage Revised Frequency-Weighted Fractional Importance to Core Damage HPRVD1 OPERATOR FAILS TO INITIATE DEPRESSURIZATION 2.6734E-1 2.8033E-1 HPWWV1 OPERATOR FAILS TO OPEN WETWELL VENT 2.3142E-1 2.414E-1 HRSPC1 OPERATOR FAILS TO LOCALLY RECOVER SP COOLING FAILURE 1.3861E-1 1.3564E-1 HRRHRX OPERATOR FAILS TO ALIGN THE RHR UNIT 1/UNIT 2 CROSSTIE 4.0047E-2 4.0855E-2 HPHPE1 OPERATOR FAILS TO CONTROL LEVEL WITH HPCI/RCIC - THIS IS A NON ATWS SCENARIO 1.9984E-2 2.4758E-2 HPHPR1 OPERATOR FAILS TO CONTROL LEVEL WITH HPCI/RCIC FOLLOWING LEVEL 8 TRIP 1.7241E-2 1.8028E-2 HOSV1 OPERATOR FAILS TO PREVENT MSIV CLOSURE DURING ATWS (2) 1.7257E-2 HPTAF1 OPERATOR FAILS TO CONTROL LEVEL AT TAF DURING ATWS - UNISOLATED VESSEL 1.5492E-2 1.4605E-2 HOAL2 OPERATOR FAILS TO LOWER AND CONTROL LEVEL DURING ATWS (ISOLATED VESSEL)
(2) 1.2427E-2 HPSPC1 OPERATOR FAILS TO ALIGN SUPPRESSION POOL COOLING - THIS IS A NON ATWS SCENARIO 1.0550E-2 1.0432E-2 ORVD2 (Split fraction)
OPERATOR FAILS TO INITIATE DEPRESSURIZATION GIVEN FAILURE TO CONTROL HIGH PRESSURE LEVEL CONTROL (2) 6.4582E-3 HODWS1 OPERATOR FAILS TO ALIGN FOR DRYWELL SPRAY.
THIS IS A NON ATWS SCENARIO.
6.5391E-3 5.5944E-3 HPTAF2 OPERATOR FAILS TO CONTOL LEVEL AT TAF DURING ATWS-ISOLATED VESSEL 3.3992E-3 4.8371E-3 HREEC1 OPERATOR FAILS TO ALIGN SWING RHRSW PUMPS FOR EECW (SCENARIO REQUIRES 2 PUMPS TO BE ALIGNED) 2.8482E-3 3.7452E-3 HOAL1 OPERATOR FAILS TO LOWER AND CONTROL LEVEL DURING ATWS (NON ISOLATED VESSEL)
(2) 3.6761E-3 HPSLC2 OPERATOR FAILS TO INITIATE STANDBY LIQUID CONTROL - VESSEL IS ISOLATED FROM CONDENSER 1.4147E-3 1.3186E-3
E-7 Database Variable(1)
Operator Action Description Previous Frequency-Weighted Fractional Importance to Core Damage Revised Frequency-Weighted Fractional Importance to Core Damage HOREE2 OPERATOR FAILS TO ALIGN SWING RHRSW PUMPS FOR EECW (SCENARIO REQUIRES 1 PUMP TO BE ALIGNED) 4.3626E-4 5.6522E-4 HPRTB1 OPERATOR FAILS TO PROVIDE BACKUP TRIP SIGNAL (2) 5.1220E-4 HOSL1 OPERATOR FAILS TO INITIATE STANDBY LIQUID CONTROL - VESSEL IS NOT ISOLATED FROM CONDENSER (2) 4.1747E-3 HOX2 OPERATOR FAILS TO CROSSTIE 4 KV SHUTDOWN BOARD 3.1177E-4 4.0820E-4 HOX1 OPERATOR FAILS TO ALIGN BATTERY CHARGER 2B TO 250V DC BATTERY BOARD 2.6481E-4 3.6289E-4 HPADS1 OPERATOR FAILS TO INHIBIT ADS (ISOLATED VESSEL)
(2) 3.4901E-4 HPHPL1 OPERATOR FAILS TO CONTROL HPCI/RCIC LONG TERM (6-24 HOURS) 3.3985E-4 2.3943E-4 HPADS2 OPERATOR FAILS TO INHIBIT ADS (NON ISOLATED VESSEL)
(2) 1.5692E-4 HODSB1 OPERATOR FAILS TO ALIGN DIESEL BOARD FOR DIESEL C 6.4999E-5 8.6858E-5 HOR480 OPERATOR FAILS TO RECOVER 480 SHUTDOWN BOARD 3.8890E-5 5.8085E-5 HPLPC1 OPERATOR FAILS TO CONTROL LPCI/CS INJECTION (3) 1.8690E-5
- 1. A limited number of Database Variable designations were changed to allow easier recognition.
- 2. As part of the revision, separate event trees were developed for ATWS events to allow better scrutiny and understanding.
- 3. This human action was not evident prior to the revision but has now risen to a low importance.
E-8 IV. TS 426, DIESEL GENERATOR ALLOWED OUTAGE TIME In Reference 3, TVA submitted a proposed Technical Specification change which revised the current Unit 1 diesel generator seven day allowed outage time to 14 days. This change had previously been approved on BFN Units 2 and 3.
The results of the previous and current Core Damage Frequency (CDF) / Incremental Conditional Core Damage Probability (ICCDP) and Large Early Release Frequency (LERF) / Incremental Conditional Large Early Release Probability (ICLERP) risk measures calculations are presented below:
CASE PREVIOUS UNIT 1 CDF (REF. 3)
REVISED UNIT 1 CDF PREVIOUS UNIT 1 LERF (REF. 3)
REVISED UNIT 1 LERF 7 Day AOT 1.87E-6 1.7854E-6 1.87E-7 4.3973E-7 14 Day AOT 1.89E-6 1.7999E-6 1.87E-7 4.3969E-7 Change 1.7E-8 1.45E-8 5E-11
-4.00E-11 NRC Guidance CDF/LERF
< 1E-6
< 1E-6
< 1E-7
< 1E-7 ICCDP 6.43E-8 7.96E-8 1.7E-10 6.84E-11 NRC Guidance ICCDP/ICLEWRP
< 5.0E-7
< 5.0E-7
< 5.0E-8
< 5.0E-8 Percentage Change 0.9%
0.87%
0.03%
-0.01%
Notes:
- 1.
The CDF for the revised model of record is 1.7666E-6 and the LERF is 4.3970E-7. The model of record results for CDF and LERF are slightly lower that the 7 day case. This is due to the DG data input reflecting the historical DG out of service durations being used to generate the baseline model of record. This was necessitated based on the shorter DG out of services times experienced following implementation of the 14 day AOT on Units 2 and 3. These changes have no impact on the calculated results.
- 2. The decrease in LERF in the 14 day AOT case is due to truncation and represents no change from the 7 day case for the LERF values.
E-9 TS 426 stated that: The risk-informed assessment concluded the increase in plant risk is small. The proposed change results in a negligible increase in the Unit 1 Conditional Core Damage Probability and the Conditional Large Early Release Probability. It went on to state that: a PSA evaluation concluded that the risk contribution of the DG TS AOT extension is non-risk significant. As demonstrated by the table above, the NRC PSA guidance continues to be met and the conclusions remain valid.
V.
TS 435, CONTAINMENT ATMOSPHERE DILUTION SYSTEM ALLOWED OUTAGE TIME In Reference 4, TVA submitted a proposed Technical Specification change which provided for seven days of continued operation with two inoperable CAD subsystems. This change had previously been approved on BFN Units 2 and 3.
In TS 435, TVA included a qualitative discussion of the effects of the proposed on the PSA. TVA stated that:
The CAD design basis oxygen control function is not required until well after a hydrogen producing LOCA event has occurred because of the time necessary for radiolysis to produce sufficient oxygen inside primary containment.
Since the safety related design function of CAD is not required prior to a core damaging event (the interval evaluated by the BFN Level I PSA), it follows this CAD function cannot impact CDF values.
BFN design basis calculations indicate the CAD function would not be needed sooner than 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> post-accident under anticipated containment conditions. The BFN Level II PSA evaluation for large early release frequency (LERF) evaluates the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-accident. Therefore, the availability of the CAD function does not affect LERF.
This discussion remains valid and is not affected by the updated PSA information.
E-10 VI. GENERIC LETTER 88-20, SUPPLEMENT 4, INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS In Reference 5, TVA provided the remaining portion of its response to Generic Letter 88-20, Supplement 4, when it submitted its Individual Plant Examination of External Events (IPEEE) Unit 1 seismic and internal fires analyses. The internal fires evaluation was performed using a progressive screening analysis based on the EPRI Fire Induced Vulnerability Evaluation methodology, which was previously performed on Units 2 and 3. One phase of the methodology evaluates the likelihood of redundant/alternate shutdown paths being unavailable at the same time a fire occurs within a fire compartment. The PSA model impacts caused by the fires of concern are evaluated generating a conditional core damage probability (CCDP). If the fire related CDF is less than 1E-06, the area can be screened from further consideration. In the next step, the PSA model is further refined by identifying specific plant impacts due to fires in the various areas based on the detailed fire damage assessment.
Following identification of the modeling errors, the BFN model was redone and the results documented and approved. The revised model was then used to re-calculate all of the IPEEE Fire scenarios. This included re-calculating all of the CCDPs and all of the fire areas previously screened out remained screened out. The results did not change any of the conclusions previously determined.
VII. GENERIC LETTER 88-20, INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT VULNERABILITIES In Reference 6, TVA responded to an NRC request for additional information related to Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerability. One of the requests was for the total estimated CDF and LERF. The revised values have been provided in Section II of this letter. As discussed in Section II, the Unit 1 PSA update did not change TVAs conclusion that no vulnerabilities exist and no plant improvements were required.
E-11 VIII. LICENSE RENEWAL - SEVERE ACCIDENT MANAGEMENT GUIDELINES An analysis of severe accident mitigation alternatives (SAMAs) was submitted in support of TVAs application for license renewal for BFN Units 1, 2 and 3 (Reference 7). Additional information was provided in References 8 and 9. A copy of the Unit 1 PSA Summary Report, Revision 2, was attached to Reference 9 for informational purposes. However, the Units 2 and 3 PSAs were the basis for the SAMA evaluation and the conclusions regarding SAMAs were not affected by the corrections to the Unit 1 PSA model or the changes to the CDF and LERF described in Section II.
IX. REFERENCES
- 1.
TVA letter to NRC, dated June 28, 2004, Browns Ferry Nuclear Plant (BFN) - Unit 1-Proposed Technical Specifications (TS) Change TS - 431 - Request for License Amendment - Extended Power Uprate (EPU) Operation.
- 2.
TVA letter to NRC, dated August 23, 2004, Browns Ferry Nuclear Plant (BFN) - Unit 1-Proposed Technical Specifications (TS) Change TS - 431 - Request for License Amendment - Extended Power Uprate Safety Assessment (PSA)
Update.
- 3.
TVA letter to NRC, dated December 6, 2004, Browns Ferry Nuclear Plant (BFN) Unit 1 - Technical Specification (TS)
Change TS 426 - Revision to Diesel Generators Allowed Outage Time.
- 4.
TVA letter to NRC, dated August 2, 2004, Browns Ferry Nuclear Plant (BFN) Unit 1 - Technical Specification (TS)
Change TS 435 - Limiting Condition for Operation (LCO) Time for Containment Atmosphere Dilution (CAD) Subsystem Inoperability.
- 5.
TVA letter to NRC, dated January 14, 2005, Browns Ferry Nuclear Plant (BFN) Unit 1 - Response to NRC Generic Letter (GL) 88-20, Supplement 4 - Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities
- Submittal of Browns Ferry Nuclear Plant Unit 1 Seismic and Internal Fires IPEEE Reports.
E-12
- 6.
TVA letter to NRC, dated August 17, 2004, Browns Ferry Nuclear Plant (BFN) - Unit 1 - Response to Request for Additional Information Related to Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerability (TAC No. MC1895).
- 7.
TVA letter to NRC, dated December 31, 2003, Browns Ferry Nuclear Plant (BFN) - Units 1, 2 and 3 - Application For Renewed Operating Licenses,
- 8.
TVA letter to NRC, dated July 7, 2004, Response to Request for Additional Information (RAI) Regarding Severe Accident Mitigation Alternatives for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3 (TAC Nos. MC1768, MC1769, and MC1770).
- 9.
TVA letter to NRC, dated September 30, 2004, Response to Request for Additional Information (RAI) Regarding Severe Accident Mitigation Alternatives for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3 (TAC Nos. MC1768, MC1769, and MC1770).