ML20206G661
| ML20206G661 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/04/1999 |
| From: | De Agazio A NRC (Affiliation Not Assigned) |
| To: | Scalice J TENNESSEE VALLEY AUTHORITY |
| References | |
| GL-88-20, TAC-M74384, NUDOCS 9905100107 | |
| Download: ML20206G661 (8) | |
Text
- . -
r Mr. J. A. Scalics Chl:f Nucler Officer and Executive Vice President
! Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennecsee 37402-2801
SUBJECT:
BROWNS FERRY, UNIT 3, INDIVIDUAL PLANT EXAMINATION GENERIC LETTER 88-20 (TAC NO. M74384)
Dear Mr. Scalice- 1 i
On April 14,1995, the Tennessee Valley Authority (TVA) submitted the multi-unit probabilistic !
risk assessment (MUPRA) for the Browns Ferry Nuclear Plant (BFN), Units 1,2 and 3. The MUPRA was submitted to resolve risk implications of operating all three units at the BFN site. '
On May 20,1998, the U.S. Nuclear Regulatory Commission (NRC) staff issued its evaluation
- on the MUPRA. Based on its evaluation of the information submitted by TVA, the NRC staff was unable to conclude that the intent of Generic Letter (GL) 88-20 had been met for the Browns Ferry Nuclear Plant. The staff evaluation (SE) identified a number of issues that required resolution.
On August 10,1998, TVA representatives met with the NRC staff and agreed to provide additional information required to resolve the outstanding issues. TVA submitted the requested additional information by letter dated November 2,1998. This additional information has allowed the NRC staff to conclude that your Individual Plant Examination (IPE) process,
. including the Browns Ferry Unit 3 PRA, is capable of identifying the most likely severe accidents
'and severe accident vulnerabilities at Unit 3. Therefore, the NRC staff concludes that TVA has met the intent of GL 88-20 for Unit 3. However, our evaluation has identified certain weaknesses which may limit the application of the IPE results in areas such as accident management and future applications of the Unit 3 PRA, thus, TVA should consider improvements in these areas. This NRC evaluation is not intended as NRC approval or endorsement of any IPE material for purposes other than those assoc;ated with meeting the intent of GL 88-20. The enclosed SE supercedes the BFN, Unit 3, SE dated May 20,1998.
Sincerely, Original signed by:
Albert W. De Agazio, Senior Project Manager, Section 2 Project Directorate il Division of Licensing Project Management Office of Nuclear Reactor Regulation /
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nclosu e: S a f valuation cc w/ encl: See next page bb DISTRIBUTION:
E Docket File, BFN Reading OGC PUBLIC CNorsworthy (e-mail SE only RCN) PFredrickson, Rii 9 JZwolinski/SBlack . Belayton SPeterson ADe Agazio ACRS HBerkow PFredrickson, Rll JLane, RES DOCUMENT NAME: G:\BFN\743842ND.wpd .
To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with rttachment/ enclosure "N" = No copy CFFICE DII-2/PM g [/[ f PDII-2/LA lG PDII-2/ R W l l l NAME ADeAgaz[o flXJ/# BClayton du SPetersoT DATE 05 g / N 05/ 1 /99 05/H/99 05/ /99 05/ /99 1g02j Official Record Copy 9905100107 990504 PDR ACOCK 05000296 P PDR
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NUCLEAR REGULATORY COMMISSION
- WASHINGTON, D.c. 20666-0001 j
\****p8 May 4,1999 Mr. J. A. Scalice Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
SUBJECT:
BROWNS FERRY, UNIT 3, INDIVIDUAL PLANT EXAMINATION
, GENERIC LETTER 88-20 (TAC NO. M74384)
Dear Mr. Scalice:
On April 14,1995, the Tennessee Valley Authority (TVA) submitted the multi unit probabilistic risk assessment (MUPRA) for the Browns Ferry Nuclear Plant (BFN), Units 1,2 and 3. The MUPRA was submitted to resolve risk implications of operating all three units at the BFN site.
On May 20,1998, the U.S. Nuclear Regulatory Commission (NRC) staff issued its evaluation on the MUPRA. Based on its evaluation of the information submitted by TVA, the NRC staff was unable to conclude that the intent of Generic Letter (GL) 88-20 had been met for the Browns Ferry Nuclear Plant. The staff evaluation (SE) identified a number of issues that required resolution.
On August 10,1998, TVA representatives met with the NRC staff and agreed to provide additional information required to resolve the outstanding issues. TVA submitted the requested additional information by letter dated November 2,1998. This additional information has allowed the NRC staff to conclude that your Individual Plant Examination (IPE) process, including the Browns Ferry Unit 3 PRA, is capable of identifying the most likely severe accidents and severe accident vulnerabilities at Unit 3. Therefore, the NRC staff concludes that TVA has met the intent of GL 88-20 for Unit 3. However, our evaluation has identified certain weaknesses which may limit the application of the IPE recults in areas such as accident
. management and future applications of the Unit 3 PRA, thus, TVA should consider improvements in these areas. This NRC evaluation is not intanded as NRC approval or endorsement of any IPE' material for purposes other than those associated with meeting the intent of GL 88-20. The enclosed SE supercedes the BFN, Unit 3, SE dated May 20,1998.
Sincerely, 1
Albert W e Agazio, Seni 'roject Manager, Section 2 Project irectorate ll Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-296
Enclosure:
Staff Evaluation cc w/ encl: See next page
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STAFF EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE INDIVIDUAL PLANT EXAMINATION TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT 3 DOCKET NO. 50 296
1.0 INTRODUCTION
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On September 1,1992, the Tennessee Valley Authority (TVA) submitted a probabilistic risk j assessment (PRA) for Brown's Ferry, Unit 2, (BFN-2) in response to Generic Letter (GL) 88-20.
The submittal reflected that at the time it was prepared, Units 1 and 3 were in extended shutdowns, i.e., the PRA did not consider multi-unit operation. It consisted of a Level 1 PRA !
along with a limited containment analysis, and it assumed Unit 2 in cperation and Units 1 and 3 shut down. k j
i On September 28,1994, the U.S. Nuclear Regulatory Commission (NRC) issued its staff evaluation (SE) covering the BFN-2 Individual Plant Examination (IPE) submittal. The NRC noted in the SE that closure of the IPE activities for the BFN site was dependent on the receipt and review of the Multi-Unit PRA (MUPRA), which was intended to estimate core damage ]
frequency (CDF) for BFN-2 assuming all three units at the site are operating.
On April 14,1995, TVA submitted the MUPRA which was based on an updated version of the original BFN-2 PRA. It included a model change w hich attempted to account for plant l modifications implemented since the original analysis was performed, and it represented the BFN-2 results with all three uhits in operation. The MUPRA submittal also served to close out two discussion items regarding the containment performance improvements program, as q requested in the BFN-2 SE.
On June 19,1997, NRC issued a request for additionalinformation (RAl) based upon a review of the MUPRA to assess its adequacy as a surrogate for the Brown's Ferry, Unit 3 (BFN 3) IPE submittal. On August 6,1997, TVA docketed a PRA for BFN-3. This PRA assumed Unit 2 and Unit 3 operating and Unit 1 shutdovm, and was based on the BFN-2 model created in the ,
MUPRA but modified to reflect BFN-3 specific characteristics (see Section 2 for details). On !
May 20,1998, the NRC issued an SE on GFN-3 indicating the staff could not conclude, based on the combined review of the BFN-3 PRA and the MUPRA, that GL 88-20 had been satisfied.
TVA met with NRC on August 10,1998, to oiscuss open items in that evaluation. l l
On November 2,1998, TVA submitted additionalinformation in response to the staff's request I for additional information (RAl) issued on June 19,1997. TVA has not indicated an intent to <
Enclosure
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2.0 EVALUATION
- The BFN-3 PRA is an extension of PRAs previously performed for BFN 2. To derive the BFN-3 PRA model, TVA modified the BFN-2 model to reflect BFN-3 specific system configurations such as residual heat removal crossties, as well as BFN-3-specific system dependencies. For example, TVA developed new system success criteria for BFN 3 given the relationship between shared systems at BNF-2 and BFN-3. The selected shared systems which were evaluated i included the electric power system, control and service air system, raw cooling water system, I turbine building, reactor building closed cooling water system, emergency equipment cooling ))
water system, the fire protection system, and the reactor building and control bay ventilation and cooling system.
The quantification of CDF considered the response of BFN-3 to initiating events while operating -
at full power, considering that BFN 2 may also be at full power and either may remain so, may
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be affected by the same initiating event, or may have been previously shut down. The BFN-3 i PRA estimated a CDF of 9 x 104per reactor-year from internally initiated events including #
- flooding. According to TVA's initiating event groupings, transients with the reactor not isolated contributed l27 percent, loss of offsite power contributed 23 percent, transients with the reactor i
- isolated contributed 20 percent, events initiatsd by support system failures (loss of cooling .
water, control air, instrumentation busses, etc.) contributed 15 percent, loss of coolant I accidents (LOCAs) and internal floods each contributed 6 percent, stuck open relief valve contributed 2 percent, and interfacing systems LOCA~ contributed less than 1 percent.
TVA also characterized the results according to BFN-3 functional failure categorizations as follows: Anticipated Transients Without Scram (ATWS) events represented 39 percent, loss of i
residual heat removal (RHR) events represented 36 percent, other general transients leading to ,
core damage represented 16 percent, and blackout (including Unit 3 only and station blackout) represented 5 percent.
The important system / equipment contributors to the estimated CDF in decreasing order of importance are: the reactor protection system, RHR system, diesel generators, RHR service water, high pressure coolant injection system, and the reactor core isolation cooling system.
TVA's Level 1 analysis appears to have examined the significant initiating events and dominant accident sequences.
The BFN-3 MA provided no revision to the vu'nerability definition provided as part of the BFN-2 IPE suomittal. In the BFN-2 IPE submittal, TVA defined a vulnerability as the failure of ;
. critical components, support systems, or operator actions that contribute significantly to the overall CDF or to early release frequency. Using this deimition, no plant vulnerabilities were {
identified as part of the BFN-3 PRA. Similarly, no plant improvements or insights were. :
. provided. 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, An issue raised in the original BFN-3 SE addressed the significant contribution toward CDF resulting from the loss of RHR. In comparing the results from the BFN-2 PRA with those from I the BFN-3 PRA, TVA has shown that a similar result is obtained in both cases. Also, as I indicated in the BFN 2 SE, the failure to remove decay heat contributes 37 percent of CDF (not ]
j
including station blackout scenarios, which also involved loss of RHR). This is consistent with the 36 percent reported for BFN-3. TVA indicated that the primary difference between the BFN-2 and BFN 3 in this area appears to result from the ability of multiple RHR loops at BFN-2 to crosstie to two adjacent units, either BFN-1 or BFN 3, whereas only one BFN-3 RHR loop
.can crosstie to an adjacent unit, BFN 2.
Given the consistency of results between BFN-3 and BFN 2 in this area and noting that TVA's treatment of Unresolved Safety Issue A-45," Shutdown Decay Heat Removal Requirements,"
was found acceptable by the staff, the staff finds that TVA's treatment at BFN-3 is acceptable, j and that the issue is resolved for BFN-3. '
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.
The human reliability analysis (HRA) performed for BFN-3 was based on the methodologies used in the BFN-2 IPE submittal which were found to be acceptable. TVA addressed both pre-initiator actions (performed during maintenance, test, surveillance, etc.) and post initiator
- actions (performed as part of the response to an accident). Quantification of the pre initiator actions was accomplished using the technique for human error rate prediction methodology (NUREG/CR-1278). Quantification of the post-initiator events was accomplished using the Serrass Ukelihood Indicator Methodology.
TVA addressed the appropriateness of using the HRA methodology and results developed for BNF-2 at BFN-3 by comparing the similarities in plant design and operation between the two units. In this regard:
improved technical specifications have been implemented at both BFN-2 and BFN-3 which impose similar test and maintenance schedules for safety related components,
=
tne BWR Owners' Group Emergency Procedures Guides, Rev. 4, are used at both BFN 2 and BFN-3 with minor differences between units being addressed in the training !
program, and the safety systern initiation setpoints are very nearly identical for both units.
According to TVA, the most important operator actions contributing to core damage are:
Failure to manually depressurize the reactor vessel using safety relief valves Failure to manually control low pressure injection during an ATWS Failure to manually align RHR to suppression pcol cooling mode Failure to start standby liquid control following an ATWS I
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= Failure to align Unit 2 RHR crosstie from Unit 2 to Unit 3 j
_ Failure to inhibit automatic depressurization system during an ATWS I Failure to manually start RHR/ core spray l Failure to maintain reactor vessel level using RHR/ core spray
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Failure to maintain reactor level control during an ATWS j 4
The staff finds the HRA portion of the BFN-3 submittal acceptable.
A back-end analysis was provided as part of the BFN-2 IPE submittal, and TVA has l demonstrated that the BFN-2 and BFN-3 containments are similar, but no plant-specific BFN-3 j back-end analysis was performed as part of either the BFN 3 PRA or the MUPRA. Thus, TVA 1 is (1) unable to quantify the containment response to BFN-3 plant specific core damage sequences that are not identical to those of BFN-2, and (2) unable to characterize the radionuclide release frequencies associated with those sequences for BNF3, which would be of assistance in accident mani.gement applications.
For example, the BFN-2 IPE submittal indicated that loss of offsite power, including station blackout, accounts for 69 percent of CDF. Owing to model enhancements and physical plant
. changes implemented since the BFN 2 IPE was performed, the BFN-3 PRA, in contrast, indicated that lose of offsite power accounts for only 23 percent of CDF. Similarly, the BFN-2 IPE submittalindicates that ATWS represents 3 percent of CDF, whereas it represents 39 percent in the BFN-3 PRA. A plant specific BNF-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. l
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In the November 2,1998, letter TVA discussed their evaluation of BFN 3 in terms of the Mark l containment performance improvements (CPI) issues (GL 88-20, Supplement 3). The CPI recommendations that were implemented and found acceptable to the staff for BFN-2 have also been implemented for Unit 3:
- 1. The hardened wetwell vent has been installed and modeled in the BFN-3 PRA
- 2. The BWR Owners Group Emergency Procedure Guide, Rev. 4, has also been ,
implemented for BFN-3. l In addition, the use of a diesel-driven fire pump as an alternate source of low pressure injection is included in the Emergerwy Operating Instructionc for BFN-3. Although it is not credited in the !
PRA because the existing piping limits its ability to successful!y cool the core, it is available as a back-up system to provide limited suppression pool make-up.
The staff finds the CPI issue resolution acceptable, i k
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3.0 CONCLUSION
Based on the above findings, the staff notes that the BFN4 PPA results are reasonable given the BFN-3 design, operation, and history and are consistent with other BWR Mark I nuclear plant IPEs, as outlined in NUREG-1560. Weaknesses, however, have been identified by the NRC staff. These weaknesses, which relate mostly to the lack of plant-specific information requested in NUREG-1335, are summarized below.
The NRC staff believes that TVA has undertaken extensive efforts in the development of numerous PRAs for the Brown's Ferry site reflecting various plant configurations and improvements in sy= tem modeling at Units 2 and 3. The staff believes that these efforts have enabled TVA to address the four specific objectives of GL 88-20: (1) to develop an appreciation for severe accident behavior, (2) to understand the most likely severe accident sequences that could occur at its plant, (3) to gain a more quantitative understanding of the ;
overall probabilities of cora damage and fission product release, and (4) if necessary, to f reduce the overall probability of core damage and fission product release. As a result, the {
staff concludes that TVA's IPE process, including the BFN-3 PRA, is capable of identifying the !
most likely severe accidents and severe accident vulnerabilities at BFN-3, and therefore, BFN- I 3 has met the intent of GL 88-20. In forming this opinion, the staff relied primarily upon TVA j submittals: the BFN-2 IPE, the MUPRA, the BFN-3 PRA, and the November 2,1998 letter to j NRC (TVA response to staff RAl's).
As noted, however, weaknesses were identified in: the front-end analysis in terms of the lack of discussion of the PRA inputs as requested in NUREG-1335, the lack of a back-end analysis despite a significantly different core damage profile as compared to BFN-2, and the lacksof a discussion of the insights gained in the IPE process for BFN-3. These weaknesses rrey limit the application of the IPE results in areas such as accident management and future applications of the BFN-3 PRA. TVA should consider improvements in these areas.
It should be noted that the staff's review focused primarily on TVA's ability to examine BFN-3, for severe accident vulnerabilities. Although certain aspects of the PRA submittal were explored in more detail than others, the review is not intended to validate the accuracy of TVA's detailed findings (or quantification estimates) that stemmed from the examination.
Therefore, this SE does not constitute NRC approval or endorsement of any IPE material for purposes other than those associated with meeting the intent of GL 88-20.
Principal Contributor: John Lane Date: May.4, 1999 l
Mr. J.- A. Scalice BROWNS FERRY NUCLEAR PLANT Tennessee Valley Authority cc:
Senior Vice President Mr. Mark J. Burzynski, Managar
' Nuclear Operations Nuclear Licensing i Tennessee Valley Authorityf Tennessee Valley Authority l 6A Lookout Place - 4X Blue Ridge '
1101 Market Street . 1101 Market Street l
~ Chattanooga..TN- 37402-2801' Chattanooga, TN 37402-2801 j l
Mr.' Jack A.' Bailey, Vice President Mr. Timothy E. Abney, Manager Engineering & Technical Services l Licensing and Industry Affairs l Tennessee Valley Authority- Browns Ferry Nuclear Plant 6A Lookout Place . Tennessee Valley Authority I 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Decatur, AL 35609 Mr. Karl W. Singer, Site Vice President l Senior Resident inspector Browns Ferry Nuclear Plant U.S. Nucbar Regulatory Commission { j Tennessee Valley Authority Browns Ferry Nuclear Plant P.O. Box 2000 10833 Shaw Road Decatur, AL 35609 Athens, AL 35611 General Counsel State Health Officer Tennessee Valley Authority Alabama Dept. of Public Health 4 ET 10H 434 Monroe Street- {
400 West Summit Hill Drive Montgomery, AL 36130-1701 i Knoxville, TN 37902 _ /
Chairman
. Mr. N. C. Kazanas, General Manager
. Limestone County Commission Nuclear Assurance 310 West Washington Street Tennessee Valley Authority Athens, AL 35611 SM Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Robert G. Jones, Plant Manager j
- Browns Ferry Nuclear Plant q
Tennessee Valley Authority . J P.O. Box 2000 '
Decatur, AL 35609
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