ML042380239
| ML042380239 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/17/2004 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GL-88-020, TAC MC1895 | |
| Download: ML042380239 (20) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 August 17, 2004 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 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)
Reference:
NRC letter, K.N. Jabbour to K.W. Singer, dated June 21, 2004, "Browns Ferry Nuclear Plant, Unit 1 -
Request for Additional Information Related to Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerability" (TAC No. MC1895)."
This letter provides TVA's responses to the referenced NRC request for additional information regarding closure of Generic Letter 88-20 for Unit 1.
The stated NRC use of the Individual Plant Examination (IPE) results was to obtain reasonable assurance that the licensee has adequately analyzed the plant design and operations to discover instances of particular vulnerability to core melt or unusually poor containment performance given a core melt accident.
The BFN facility has been extensively reviewed by the staff, beginning with an interim reliability evaluation in 1982.
This was followed by the subsequent submittal of the November 20, 1986 BFN Unit 1 PRA and its subsequent NRC audit.
TVA subsequently submitted the BFN IPE in 1992 and the Multi-unit PRA (MUPRA) in 1995.
Since then, TVA has performed individual Unit 2 and Unit 3 PRAs.
Neither the MUPRA, nor the subsequently performed individual Unit 2 and Unit 3 PRAs have identified plant
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U.S. Nuclear Regulatory Commission Page 2 August 17, 2004 vulnerabilities when single or multiple units are in operation.
These analyses have provided the staff with more than reasonable assurance that TVA has adequately analyzed the plant design and operations.
TVA and NRC met on February 27, 2004, to discuss closure of Generic Letter 88-20 for BFN Unit 1. As documented in the staff's meeting notes, NRC agreed to identify in a letter to TVA the specific information needed to close out this issue for BFN Unit 1. Without prejudice to its position that additional work is not required to close Generic Letter 88-20 for Unit 1, TVA has agreed to provide a response to the NRC's June 21, 2004 request for additional information in order to allow the staff to complete its review.
A response to each request in the staff's June 21, 2004 letter is enclosed.
There are no regulatory commitments associated with this submittal.
If you have any questions about this amendment, please contact me at (256)729-2636.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on August 17, 2004.
Sincerely,
U.S. Nuclear Regulatory Commission Page 3 August 17, 2004 Enclosures Cc: (Enclosures)
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 Kahtan N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
ENCLOSURE BROWNS FERRY NUCLEAR PLANT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING GENERIC LETTER 88-20, INDIVIDUAL PLANT EXAMINATION FOR SEVERE ACCIDENT VULNERABILITY BACKGROUND Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities, requested all licensees:
- 1.
Perform a systematic examination to identify any plant specific vulnerabilities to severe accidents, and
- 2.
Identify and review proposed plant improvements (design changes and changes to operating procedures, maintenance, surveillance, training, or staffing).
The stated NRC use of the Individual Plant Examination (IPE) results was to obtain reasonable assurance that the licensee has adequately analyzed the plant design and operations to discover instances of particular vulnerability to core melt or unusually poor containment performance given a core melt accident.
The Generic Letter did not request individual IPEs be performed for each unit at a multi-unit site.
The BFN facility has been extensively reviewed by the staff, beginning with an interim reliability evaluation in 1982.
This was followed by the subsequent submittal of the November 20, 1986 BFN Unit 1 PRA and its subsequent NRC audit.
TVA subsequently submitted the BFN IPE in 1992 and the Multi-unit PRA (MUPRA) in 1995.
Since then, TVA has performed individual Unit 2 and Unit 3 PRAs.
Neither the NUPRA, nor the subsequently performed individual Unit 2 and Unit 3 PRAs have identified plant vulnerabilities when single or multiple units are in operation.
These analyses have provided the staff with more than reasonable assurance that TVA has adequately analyzed the plant design and operations.
TVA and NRC met on February 27, 2004, to discuss closure of Generic Letter 88-20 for BFN Unit 1. As documented in the staff's meeting notes, NRC agreed to identify in a letter to TVA the specific information needed to close out this issue for BFN Unit 1. Without prejudice to its position that additional work is not required to close Generic Letter 88-20 for Unit 1, TVA has E-1
agreed to provide a response to the NRC's June 21, 2004 request for additional information in order to allow the staff to complete its review. A response to each request in the staff's June 21, 2004 letter is enclosed.
NRC REQUEST
- 1.
Please discuss in detail how the individual plant examination (IPE)/probabilistic risk assessment (PRA) performed for Browns Ferry Plant (BFN), Unit 1, will properly reflect the unit's anticipated configuration at commercial operation.
Please provide the version designation or date of the current Unit 1 IPE/PRA, the total estimated core damage frequency, and the large early release frequency.
TVA RESPONSE The anticipated configuration at Unit 1 restart for each system is described by a combination of the existing plant documentation and Design Change Notification (DCN) packages.
The existing plant documentation along with each issued Unit 1 DCN, through the design freeze date of February 6, 2003, was reviewed by the staff performing the Probabilistic Safety Analysis (PSA) for Unit 1 to ensure that the Unit 1 system models represent the anticipated configuration.
Other available Unit 1 DCNs in preliminary stages of development were also reviewed during the initial preparation of the Unit 1 PSA system notebooks to evaluate their effects on the Unit 1 system models.
If the Unit 1 DCN was not available at the time of this review, the Unit 2 system configuration was assumed applicable.
This is acceptable due to the design of the two units being essentially identical.
These assumptions are noted in the individual system analysis notebooks.
Following completion of the Unit 1 DCN closure process, which confirms the DCN packages are consistent with the Unit 1 as-built configuration, closed DCN packages will be reviewed to assure consistency with the Unit 1 PSA model. Any differences will be reviewed and required revisions to the Unit 1 PSA will be made.
Applicable Unit 1 operating and test procedures have not been developed, so the PSA model assumed the Unit 2 procedures are applicable. Again, this is acceptable due to the design of the two units being essentially identical and their operating procedures also being functionally duplicative.
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The Unit 1 PSA model has been given the unique designator U1040727 which reflects its issuance on July 27, 2004.
The Unit 1 PSA was developed using "RISKMAN for Windows Software,"
Version 7.0, 2004.
The total estimated mean core damage frequency is 1.86 x 10-6 per year.
The mean large early release frequency is 1.87 x 10-7 per year.
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NRC REQUEST
- 2.
Based on the results of the Unit 1 IPE/PRA, please discuss the specific vulnerabilities to severe accidents which are identified, if any, for BFN Unit 1. Also, please identify and discuss the proposed plant improvements, such as design changes and changes to operating procedures, maintenance, surveillance, training or staffing, resulting from the IPE/PRA.
TVA RESPONSE In the September 1, 1992 BFN Unit 2 IPE submittal"', TVA established the criteria to identify vulnerabilities.
In the September 28, 1994 NRC Staff Evaluation of the BFN IPEI2", the NRC accepted the criteria to define "vulnerability".
These criteria continue to be used for the Unit 1 PSA.
Based on these criteria being satisfied by the Unit 1 PSA results, as with the Units 2' and 3 PSAs, no vulnerabilities and no plant improvements were identified.
1 TVA letter to NRC, dated September 1, 1992, "Browns Ferry Nuclear Plant (BFN) - Response to Generic Letter (GL) 88 "Individual Plant Examination for Severe Accident Vulnerabilities -
2 NRC letter to TVA, dated September 28, 1994, "Browns Ferry Nuclear Plant Unit 2 - Individual Plant Examination Submittal for Internal Events (TAC No. M74385)."
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NRC REQUEST
- 3.
Please discuss any methods and processes used in BFN Unit 1 IPE/PRA but not used in the BFN Units 2 and 3 IPE/PRA, that would potentially affect the results (e.g., mask a vulnerability).
TVA RESPONSE In summary, the methods and processes employed in the BFN Unit 1 PSA do not adversely affect the results nor mask a vulnerability.
The Unit 1 PRA was built upon the Units 2 and 3 PRAs and their supporting documentation.
The Unit 1 and Units 2 and 3 PRA models were developed using the following similar processes and approaches, which included:
Riskman Software, Bayesian updates combining the plant specific and generic data failure rates and initiating event frequencies, Unit 1 used the Units 2/3 plant specific data as input, Interviews were conducted with Operations and Simulator Training personnel, who are responsible for all three units (included running simulator events for modeling inputs),
System success criteria are identical, with shared system success criteria based on three unit operation The Modular Accident Analysis Program (MAAP) was used for thermal-hydraulic analysis response characteristics, event timing and Level 2 parameters However, there were two method/process differences between the Unit 1 PSA and the Units 2 and 3 IPE/PRAs:
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- 1)
Unit 1 used ASME-RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," for guidance.
Following the ASME Standard resulted primarily in the development of better supporting documentation than that assembled for the Units 2 and 3 PSAs.
The use of ASME-RA-S-2002 does not adversely affect the results nor mask a vulnerability.
- 2)
The quantification of human reliability.
Over time, as new guidance was provided, the Units 2 and 3 IPE/PRAs used several valid methodologies in the human reliability quantification process (including the EPRI Human Reliability Analysis method, although the Calculator was not available at the time those analyses were performed).
The Unit 1 PSA uses the EPRI Human Reliability Analysis Calculator in the quantification process for the pre-and post-initiator operator actions.
This was done to satisfy the desire to use a single valid methodology.
The consistent use of this methodology would not adversely affect the results nor mask a vulnerability.
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NRC REQUEST 4a.
Please list and discuss the weaknesses and deficiencies identified by the staff in the staff evaluation reports and the Nuclear Regulatory Commission contractor's Technical Evaluation Report for the BFN Units 2 and 3 IPE/PRA dated September 28, 1994, and May 4, 1999, respectively.
In addition please identify all Units 2 and 3 peer review "facts and observations" that were judged by the peer review team to be necessary to address (Significance level A and B in NEI 00-02), Industry PRA Review Guidance).
TVA RESPONSE The Unit 1 PSA incorporated the findings of the Units 2 and 3 PRAs Peer Review.
The previous conducted Peer Review was effectively an administrative and technical Peer Review of the Unit 1 PSA.
Similar models, processes, policies, approaches, reviews, and management oversight were utilized to develop the Unit 1 PSA.
Each identified weakness or deficiency and their disposition is discussed in the attachment.
NRC REQUEST A.
NRC IDENTIFIED CONTAINMENT PERFORMANCE IMPROVEMENTS NRC letter of September 28, 1994 stated: "TVA is requested to address the feasibility of evaluating the potential benefit of two containment performance improvement items in the multiunit PRA.
These items are (1) the use of diesel-driven fire protection system pump to inject water into the reactor vessel upon loss of AC power, and (2) the need for power to the automatic depressurization system solenoid valves to permit depressurization of the reactor following loss of AC power and depletion of batteries."
TVA RESPONSE TVA addressed the two containment performance improvement (CPI) items in the Multi-unit PRA 3 ). TVA stated: "These improvements were evaluated in conjunction with the hardened wetwell vent 3
TVA letter to NRC, dated April 14, 1995, "Browns Ferry Nuclear Plant (BFN) - Multi-unit Probabilistic Risk Assessment (PRA)."
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because of the synergistic interaction each improvement has on the other.
Sensitivity studies were performed using the diesel-driven fire pump, in conjunction with functional safety relief valves and the hardened wetwell vent path, to provide an open loop mode of core cooling following loss of AC power.
Based on the resulting small changes in core damage frequency, TVA has no plans to provide an alternate source of power to the automatic depressurization system solenoid valves. Use of the diesel driven fire pump and the hardened wetwell vent are already discussed in BFN Emergency Operating Instructions."
NRC's Staff Evaluationt 4) stated: "The staff finds the CPI issue resolution acceptable".
The same modeling approaches were used for Unit 1 and the PSA Unit 1 results are consistent with the Units 2 and 3 results.
B.
WEAKNESSES AND DEFICIENCIES -
SEPTEMBER 28, 1994 NRC LETTER NRC ISSUE
- 1.
Technical Evaluation Report, Section 2.2.2, Item 1.1:
No list of nomenclature can be found.
TVA RESPONSE The back-end analysis was redone for Browns Ferry Unit 1 PSA.
The updated analysis is documented in a new separate analysis notebook.
Nomenclature is defined in the text or tables in the new notebook.
4 NRC letter to TVA, dated May 4, 1999, uBrowns Ferry, Unit 3, Individual Plant Examination, Generic Letter 88-20 (TAC No. M74384)."
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NRC ISSUE
- 2.
Technical Evaluation Report, Section 2.2.2, Item 1.2:
Pages 4.4-3 and 4.4-4 appear twice, each from a different version with different contents, (i.e., once from SECT44.BFN.8/31/92 and again from SECT44.bfn.9/l/92).
TVA RESPONSE A back-end analysis was accomplished for Browns Ferry Unit 1 PSA.
The back-end analysis is documented in a new separate analysis notebook.
Page numbers and associated contents are correct in the new Unit 1 analysis notebook.
NRC ISSUE
- 3.
Section 2.2.2 item 1.3:
The description of Table 4.7-2 on page 4.7-6 of the submittal does not reflect the actual contents of Table 4.7-2:
The next six columns describe the conditions of the primary containment at the time of its failure, the drywell failure area assumed in the evaluation, and whether post-vessel breach water was available to the drywell floor.
The next three columns indicate the in-pedestal concrete ablation depth (feet), and the amount of in-vessel and ex-vessel hydrogen produced (pounds) at the problem end time (typically 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after event initiation).
The last three columns show the cesium iodide mass fractions remaining in the reactor building and released into the environment up to the problem end time.
TVA RESPONSE A back-end analysis was accomplished for Browns Ferry Unit 1 PSA.
The back-end analysis is documented in a new separate analysis notebook.
The text contents and actual table contents are consistent in the Unit 1 analysis notebook, with the exception of in-pedestal concrete ablation depth.
The discussion is included in the Large Early Release Frequency notebook.
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NRC ISSUE
- 4.
Section 2.2.2 item 2:
No response to CPI recommendations appears to have been made.
TVA RESPONSE As stated above, TVA addressed the two containment performance improvement (CPI) items in the April 14, 1995 TVA letter and stated: "These improvements were evaluated in conjunction with the hardened wetwell vent because of the synergistic interaction each improvement has on the other.
Sensitivity studies were performed using the diesel-driven fire pump, in conjunction with functional safety relief valves and the hardened wetwell vent path, to provide an open loop mode of core cooling following loss of AC power.
Based on the resulting small changes in core damage frequency, TVA has no plans to provide an alternate source of power to the automatic depressurization system solenoid valves.
Use of the diesel driven fire pump and the hardened wetwell vent are already discussed in BFN Emergency Operating Instructions."
NRC letter of May 4, 1999 stated: "The staff finds the CPI issue resolution acceptable".
The same modeling approaches were used for Unit 1 PSA and the Unit 1 PSA results are consistent with the Units 2 and 3 results.
NRC ISSUE
- 5.
Section 2.2.2 item 3:
Independent in-house peer review performed for the BF2 IPE appears to be limited.
TVA RESPONSE The BFN Units 2 and 3 IPE 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 PRA 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 E-10
the BWROG document "Report to the Industry on PRA Peer Review Certification Process -
Pilot Plant Results" were employed.
The Browns Ferry Unit 2 and 3 Peer Review resulted in a consistent evaluation across all elements and sub-elements.
During the Unit 2 and 3 PRA updates in 2003, the significant findings (i.e., designated as Level A or B) from the Peer Certification were resolved, resulting in all PRA elements now having a minimum certification grade of 3. A copy of the significant findings and their disposition is attached to this letter.
Since the Unit 1 PSA was built from the Units 2 and 3 PSAs, which incorporate the resolution of the peer review comment, the Unit 1 PSA has incorporated the findings of the Units 2 and 3 PRAs Peer Review.
Thus, the previously conducted Peer Review was effectively an administrative and technical Peer Review of the Unit 1 PSA.
Similar models, processes, policies, approaches, reviews, and management oversight were utilized to develop the Unit 1 PSA.
NRC ISSUE
- 6.
Section 2.2.2, Item 4:
Front-end to back-end dependencies are not explained in the submittal.
TVA RESPONSE The staff's comments addressed the June 1996 Unit 3 PSA.
TVA did not docket the Unit 3 PSA for staff review in response to Generic Letter 88-20.
The staff obtained a copy of the Unit 3 PSA during a Maintenance Rule inspection and placed it on the docket without TVA's knowledge.
Since TVA did not intend for the Unit 3 PSA to be reviewed by the staff in response to Generic Letter 88-20, it did not contain the format or content recommended in NUREG-1335, Individual Plant Examination: Submittal Guidance.
In the Unit 1 PSA back-end analysis, the dependencies from top events in the front-end analysis are now accounted for on a sequence by sequence basis in the back-end analysis.
The two models are directly linked for sequence quantification.
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NRC ISSUE
- 7.
Section 3:
We believe that more attention could have been directed to the causes of early failure; the role of the operators in mitigating the consequences and, ways of reducing the probability and/or consequences of the early failures.
TVA RESPONSE The Unit 1 back-end analysis was accomplished consistent with the NRC contractor and Peer Certification "facts and observations.
For the Unit 1 PSA, additional attention was given to the causes of early containment failure and the role of the operators in mitigating the consequences and reducing the probability of the early failures.
The back-end analysis now considers the possibility of the operator depressurizing the vessel following the onset of fuel damage.
If successful, this action decreases the frequency of large early releases.
This vessel depressurization increases the likelihood that the molten fuel is cooled sufficiently to maintain containment integrity.
C.
WEAKNESSES AND DEFICIENCIES - MAY 4, 1999 NRC LETTER NRC ISSUE
- 1.
Section 2.0:
The staff finds the Level 1 results consistent with other boiling water reactors (BWR) plants, however, the lack of a discussion of the insights gained is considered a weakness in TVA's approach.
TVA RESPONSE The staff's comments addressed the June 1996 Unit 3 PRA.
TVA did not docket the Unit 3 PRA for staff review in response to Generic Letter 88-20.
The staff obtained a copy of the Unit 3 PRA during a Maintenance Rule inspection and placed it on the docket without TVA's knowledge.- Since TVA did not intend for the Unit 3 PSA to be reviewed by the staff in response to Generic Letter 88-20, it did not contain the format or content recommended in NUREG-1335, Individual Plant Examination:
Submittal Guidance.
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Since that time, a separate summary report for the latest PSAs has been prepared for each unit at Browns Ferry; i.e. for Units 1, 2 and 3. These summary reports for the quantification results contain an evaluation of the respective unit results and include the specific insights gained from the PSA quantifications performed for each unit.
NRC ISSUE
- 2.
Section 2.0:
The BFN-3 PRA is a summary document and was not intended originally as the primary document for the IPE program.
Thus, it did not provide input on the front-end topics requested in NUREG-1335 such as Sections 2.1, 2.2, or 2.3.
For example, no details were provided for common cause failure modes, equipment failure rates employed in the analysis, the use of generic versus plant specific data, or internal flooding.
The staff finds the lack of this information a weakness in TVA's front-end portion of the submittal, and notes that this information was requested in the staff's RAI referenced above.
TVA RESPONSE The staff's comments addressed the June 1996 Unit 3 PSA.
TVA did not docket the Unit 3 PRA for staff review in response to Generic Letter 88-20.
The staff obtained a copy of the Unit 3 PSA during a Maintenance Rule inspection and placed it on the docket without TVA's knowledge.
Since TVA did not intend for the Unit 3 PSA to be reviewed by the staff in response to Generic Letter 88-20, it did not contain the format or content recommended in NUREG-1335, Individual Plant Examination:
Submittal Guidance.
The current Unit 3 model makes use of many of the same modeling features as for Unit 2. They share the same system, data, human reliability, and back-end analysis notebooks.
However, separate sequence quantification and summary result notebooks for the latest PSAs have been prepared for Units 2 and 3.
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For Unit 1, separate system, data, human reliability, back-end analysis, sequence quantification, and summary results notebooks were developed.
These notebooks address the topics requested in NUREG 1335, with the CPI issues addressed in the Unit 1 PSA consistent with TVA resolution as described in the April 14, 1995 TVA letter.
NRC ISSUE
- 3.
Section 3.0:
A plant-specific BFN-3 back-end analysis, even using the BFN-2 containment model, would have provided additional insights concerning accident progression issues and overall containment performance at BFN-3, including an evaluation of large, early release frequency.
For this reason, therefore, the staff finds the lack of the back-end portion of the IPE a weakness.
TVA RESPONSE The Unit 1 PSA contains a back-end analysis.
NRC REQUEST 4b.
Please describe how the weakness, deficiencies and significant peer review observations have been resolved in the current Unit 1 IPE/PRA.
TVA RESPONSE As previously discussed, the BFN Units 2 and 3 IPE 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 PRA 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.
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The Browns Ferry Unit 2 and 3 Peer Review resulted in a consistent evaluation across all elements and sub-elements.
During the Unit 2 and 3 PRA updates in 2003, the significant findings (i.e., designated as Level A or B) from the Peer Certification were resolved, resulting in all PRA elements now having a minimum certification grade of 3.
A copy of the significant findings and their disposition is attached to this letter.
The resolution of one "Fact and Observation" was not documented in this report.
For Element MU, sub-element 6 states "The control of the RISKMAN model is not discussed'. Common practice at TVA is to keep the official copy of all electronic PSA model versions in the in-file-keeper electronic system.
Additionally, each identified potential enhancement or problem with an existing model is kept on a "Problem Evaluation Report (PER)" list.
The list of issues on the PER list is systematically evaluated for significance and disposition in model revisions.
Since the Unit 1 PSA was built from the Units 2 and 3 PSAs, which incorporate the resolution of the peer review comment, the Unit 1 PSA has incorporated the findings of the Units 2 and 3 PRAs Peer Review.
Thus, the previously conducted Peer Review was effectively an administrative and technical Peer Review of the Unit 1 PSA.
Similar models, processes, policies, approaches, reviews, and management oversight were utilized to develop the Unit 1 PSA.
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ATTACHMENT BROWNS FERRY NUCLEAR PLANT PROBABILISTIC SAFETY ASSESSMENT CERTIFICATION AND PER RESOLUTION
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT PROBABILISTIC SAFETY ASSESSMENT CERTIFICATION AND PER RESOLUTION Revision 1
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