ML18039A602
| ML18039A602 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/02/1998 |
| From: | ABNEY T E TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-88-20, TAC-M74385, TAC-M74386, NUDOCS 9811100197 | |
| Download: ML18039A602 (28) | |
Text
CATEGORY 1 REGULA ORY INFORMATION DISTRIBUTIO SYSTEM (RIDS)ACCESSjON NBR:9811100197 DOC.DATE: 98/11/02 NOTARIZED:
NO FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee ,.50'-,296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee AUTH.NAME AUTHOR AFFILIATION
'ABNEY,T.E.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Records Management B anch (Document Control Desk)
SUBJECT:
Provides response to RAI re GL 88-20 for Unit 3 individual plant exam (IPE)&Unit 2 multi-unit PRA.DISTRIBUTION CODE: A011D COPIES RECEIVED:LTR ENCL SIZE: TITLE: Generic Ltr 88-20 re Individual Plant Evaluations NOTES: DOCKET 05000260 05000296 E RECIPIENT ID CODE/NAME PD2-3-PD'OPIES LTTR ENCL 1 1 RECIPIENT ID CODE/NAME DEAGAZIO,A COPIES LTTR ENCL 1 1 INTERNAL: ACRS HOUSTON,M CENTER CB RES'/DST/IPEEE RGN 1 RGN 3 1 1 1 1 1 1 1 1 1 1 1 1 AEOD/SPD/RRAB NRR/DRPE/PDl-3 NRR/DSSA/SPSB RES/PRAB RGN 2 RGN 4 1 1, 1 1 1 1 3 3 1 1 1 1 EXTERNAL: LITCO-BRYCE,J.
NRC PDR 1 1 1 1 NOAC 1 1 D NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPZES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)ON EXTENSION 415-2083'TOTAL NUMBER OF COPIES REQUIRED: LTTR 19 ENCL 19 0~4 I II Tennessee Vatley Authority, Post Office Box 2000.Decatur, Alabama 36609 November 2, 1998 U.S.Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.20555 Gentlemen:
In the Matter of Tennessee Valley Authority Docket Nos.50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN)-UNITS 2 AND 3-UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI UNIT PRAr GENERIC LETTER 88-20-RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (TAC NOS.M74385 AND M74386)On August 10, 1998, TVA met with representatives of the NRC staff and management to discuss the approach for resolution of outstanding issues relating to the BFN Generic Letter 88-20 activities.
On September 28, 1994, the staff issued its evaluation of the BFN IPE for single-unit operation of Unit 2.NRC concluded that BFN's IPE submittal was complete with the level of detail consistent with the information requested in NUREG-1335, Individual Plant Examination Submittal Guidance.However, NRC stated that closure of IPE activities for BFN was dependent upon the submittal of an acceptable PRA that considered multi-unit operation.
TVA submitted the Multi-Unit PRA (MUPRA)in April 1995, and addressed containment performance issues as requested.
This submittal closed out all open IPE issues for Unit 2.In June 1997, NRC requested additional information as a result of its review of the MUPRA for applicability to Unit 3.TVA, however, declined to respond stating that further staff review of the MUPRA was unwarranted and that an adequate basis existed for closing out the BFN IPE program without the expenditure of more TVA and NRC resources.
On May 20, 1998, the staff issued a summary evaluation of the GL 88-20 effort at BFN.This evaluation stated that the NRC 98liX00197 98XMZf PDR ADOCK 05000260 P1DR 0 0 II U.S.Nuclear Regulatory Commission Page 2 November 2, 1998 was unable to conclude that the intent of GL 88-20 had been met, and cited areas where information was incomplete.
In the August 10, 1998, meeting, TVA agreed to provide information regarding residual heat removal system modeling, applicability of the Unit 2 containment model to Unit 3, verification of design and operating similarities between Uni'ts 2 and 3, and verification that the containment performance improvements have been implemented and are applicable to Unit 3.The agreed upon information is provided in the Enclosure.
It is TVA's understanding that this information should be sufficient to allow the staff to complete the GL 88-20 reviews for BFN.There are no commitments contained in this letter.If you'ave any questions regarding this information please call me at (256)729-2636.S'ncerely T.E.Abney Manager of L si g and Indu ry Affa'rs Enclosure cc: See page Ib II/
U.S.Nuclear Regulatory Commission
~Page 3 November 2, 1998 Enclosure cc (Enclosure):
Mr.Harold O.Christensen, Branch Chief U.S.Nuclear Regulatory Commission Region II 61 Forsyth Street, S.W.Suite 23T85 Atlanta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr.Albert W.De Agazio, Project Manager U.S.Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr.L.Raghavan, Project Manager U.S.Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 II ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)UNITS 2 AND 3 UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI-UNIT PRA, GENERIC LETTER 88-20 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION Discussion of the Residual Heat Removal S stem Model Staff Issue.Discuss the basis for loss of RHR sequence modeling resulting in a relatively high CDF contribution.
TVA Response.Table 1, Contributions of Functional Failure Groups to CDF (Unit 3 with Unit 2 operating);
Table 2, PSA Importance of Individual Systems (Unit 2 with Unit 3 operating) ahd Table 3, PSA Importance of Individual Systems (Unit 3 with Unit 2 operating) provided below are taken from the current revisions of the Unit 2 and Unit 3 PRAs.It has been confirmed that the importance measures given do not account for double counting of RHR failure during ATWS scenarios.
Normalizing the RHR contribution for Unit 2 shown in Table 2 by dividing the raw contribution (18%)by 1.62, results in a normalized RHR importance of 11%, which is consistent with the expected importance from other RHR systems in BWR IPE submittals.
In a similar fashion, modifying the Unit 3 model to allow for 3 RHR pumps to remove ATWS decay heat (rather than 4)and removing the interference with Unit 2 sharing RHR pumps B and D with Unit 3 for decay heat removal following a multiple Unit trip, the raw importance for RHR drops to 20.3%for Unit 3 (ATWS drops to 35.5%).This results in a Unit 3 normalization factor of 1.45 and a normalized RHR group importance of 14%, which is, again, close to the expected importance value for this group.Therefore, the normalized importance values for Unit 2 and Unit 3 RHR are reflected as follows: Unit 2 Unit 3 11%14%The primary difference between these results is the reflection of the ability of Unit 2 to crosstie to adjacent units (1 or 3), whereas only one Unit 3 RHR loop can crosstie to an adjacent unit (Unit 2).
~I ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI UNI T PRAg GENERIC LETTER 88 20 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED Table 1 Contributions of Functional Failure-Groups To CDF Unit 3 with Unit 2 Operating System ATWS Loss of Residual Heat Removal Transient with Reactor Vessel at Hi h Pressure Transient Followed b Loss of Vital DC Power Blackout of Unit 3 Station Blackout De raded Emer enc E ui ment Coolin Water Mean CDF*(per year)3.52E-06 3.30E-06 7.86E-07 6.85E-07 2.72E-07 2.18E-07 2.28E-08 PSA Importance*
38.6 36.1 8.6 7.5 3.0 2.4 0.2*The mean CDF is determined by examining the dominant sequence file.The sequences in the dominant sequence file represent in excess of 99%of the total CDF and form a convenient database for risk management applications.
The accident sequence groups are defined by specifying success or failure combinations of top events or split fractions.
For example, the"ATWS" accident sequence group is defined as all sequences with the Top Event"RPS" faile'd.Since the dominant sequence file represents less than 1008 of the total CDF, the"percentage of total" for each accident sequence group is determined by dividing the mean CDF for that group by the total CDF represented by the dominant sequence file.In the current model, the total CDF represented by the dominant sequence file is 9.13E-06 (per year).For the ATWS accident sequence group, for example, the"percentage of total" is calculated as: 3.52E-06/9.13E-06=38.6%
Ik 0 ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI-UNIT PRA, GENERIC LETTER 88-20 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED Table 2 PSA Importance of Individual BFN Systems Unit 2 with Unit 3 Operating System Reactor Protection S stem Diesel Generators Hi h Pressure Coolant In'ection S stem Residual Heat Removal S stem Reactor Core Isolation Coolin S stem Residual Heat Removal Service Water S stem Control Rod Drive S stem Standb Li uid Control System Shared Actuation Instrumentation 250V DC Batter Boards Main Steam S stem Includin Turbine Tri Coze S za RBCCW Emer ency E ui ment Coolin Water S stem Condensate and Feedwater System Plant Air PSA Importance*
0.41 0.21 0.18 0.18 0.16 0.15 0.09 0.07 0.05 0.05 0.04 0.02 0.01<0.01<0.01<0.01*The fraction of CDF with sequences in which the failures occur in the indicated system.E-3 41 ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI-UNIT PRA, GENERIC LETTER 88-20 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED Table 3 PSA Importance of Individual BFN Systems Unit 3 with Unit 2 Operating System Reactor Protection S stem Residual Heat Removal S stem Diesel.Generators Residual Heat Removal Service Water System Hi h Pressure Coolant In'ection S stem Reactor Core Isolation Coolin S stem Main Steam S stem Includin Turbine Tri 250V DC Battery Boards Shared Actuation Instrumentation Control Rod Drive S stem Standby Liquid Control S stem RBCCW Condensate and Feedwater S stem Core Spra Plant Air Emer enc E ui ment Coolin Water S stem PSA Importance*
0.39 0.36 0.21 0.16 0.13 0.09 0.08 0.07 0.05 0.05 0.04 0.01<0.01<0.01<0.01<0.01*The fraction of CDF with sequences in which the failures occur in the indicated system.
0 ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI-UNIT PRA, GENERIC LETTER 88-20 RESPONSE'TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED A licabilit of the Unit 2 Containment Model to Unit 3 (Backend Anal sis)Critical thinking was used to determine that a unique Unit 3 Level 2 evaluation was not necessary at the time the Unit 3 Level 1 model was developed.
The following common attributes support this conclusion.
Common Secondar Containment The reactor building exterior walls , roof, floor, penetrations,, and qualified membrane extensions form the secondary containment membrane.The secondary containment is divided into four ventilation zones (Units 1, 2, and 3 reactor buildings, and the reactor building refueling zone.which is common to all three units).While the four zone design provides increased capability of localizing the consequences of an accident or radioactive release to a single zone, if the internal zonal boundaries should fail, the entire reactor building (Units 1, 2, and 3)would still meet the requirements of secondary containment.
Similar Secondar Containment Isolation S stem Each unit's reactor zone isolation is initiated by the same signals, reactor zone high radiation, high drywell pressure, or low reactor vessel water level with the same initiation setpoints.
High radiation localized in the refuel zone isolates only the refuel zone.A reactor zone isolation in any unit isolates the common refuel zone in addition to the affected reactor zone.Common Standb Gas Treatment S stem The function of the standby gas treatment system is to ensure tha't radioactive materials present in secondary containment following a design basis accident are filtered and adsorbed prior to release to the environment.
It consists of three redundant 50%capacity subsystems which are shared between all three units.All three subsystems share a common suction manifold which connect to all three reactor zones and the refueling zone.Upon receipt of an accident signal, all three standby gas treatment subsystems will start.E-5
ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI UN I T PRA~GENERI C LETTER 8 8 2 0 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED Similar Primar Containment S stem for Units 2 and 3 Each unit houses an identical GE Mark I pressure suppression containment system which houses the reactor vessel, reactor coolant circulating loops, and other branch connections of the reactor primary system.The basic operation of the primary containment system for units 2 and 3 is identical~Similar Primar Containment Isolation S stem(PCIS)
The PCIS for Units 2 and 3 are very nearly identical.
This is demonstrated by the BFN FSAR which lists the principal primary'ontainment penetrations and isolation valves identically for all three units on the same table (FSAR Table 5.2.-2).Similar Parameter In uts Used for Primar and Secondar Containment Isolation The primary containment isolation valves at BFN are categorized into seven"groups" based upon the type(s)of systems isolated and the associated isolation signals.Additional evidence of the similarity between the units is the fact that the same PCIS isolation signal setpoints result in the actuation of equivalent isolation valves in Units 2 or Unit 3.Modifications A review of the completed modifications for Units 2 and 3 reveals that the only substantive differences in the modifications programs for the two units are differences caused by implementation schedules which are dependent on refueling outage schedules (e.g., power uprate, 24-month fuel cycle).There are no planned modifications in the current BFN capital projects plan which would make either unit unique.
4i ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI UNI T PRA~GENERIC LETTER 88 20 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED Desi n and 0 eratin Similarities Between Unit 2 and Unit 3 Im roved Technical S ecifications Improved Technical Specifications (ITS)were implemented for BFN Units 2 and 3 on July 27, 1998.The use of ITS imposes similar test and maintenance schedules for safety-related and.major systems and components.
Similar Plant Procedures Both Unit 2 and Unit 3 EOIs/AOIs implement the BWR Owners Group Emergency Procedure Guidelines, Revision 4.Extensive upgrades were made to the Unit 2 AOIs/EOIs to support Unit 2 restart in 1991.The similarities between the units allowed near replicatio~
of the Unit 2 procedures with the addition of Unit 3 unique identifiers.
The minor unit differences which do exist are addressed in the operator training program.Safet S stem Initiation Set pints The safety system initiation setpoints for both units are very nearly identical.
A recent line by line electronic comparison of the Units 2 and 3 Improved Technical Specifications was performed for Section 3.Only minor differences were revealed by this review.Almost all of the differences are attributed to the recent power uprate implementation on Unit 3.These differences will be eliminated with the power uprate implementation planned for the spring 1999 Unit 2 refueling outage.Maintenance Rule Im lementation The Maintenance Rule program is implemented consistent with issued procedures.
These procedures provide guidance for initiation, retrieval, trending, and reporting data relative to"Plant Level" and"Function Specific" indicators of performance required by the Maintenance Rule.The requirements of the procedures are in agreement with 10 CFR 50.65 and NUMARC 93-01.Two aspects of the Maintenance Rule implemented program illustrate and support the fact that the units are essentially symmetrical.
One aspect of the Maintenance Rule implementation process is the monitoring and trending of E-7
ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI-UNIT PRA, GENERIC LETTER 88-20 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED performance of selected Units 1, 2, 3, and common plant systems structures, and components.
Performance criteria is developed and documented for each of the Units 1, 2, 3, and common plant systems structures, and components within the scope of the program.The selection of the performance criteria and associated bases include the technical justification for the performance criteria.The bases: 1)identify the performance criteria;2)identify the function being monitored, its risk significance and operational mode (normally operating or standby);3)define functional failures (for reliability and functional failures)as it relates to its performance criteria;4)include data collection methodology and formulas;and 5)include PSA assumptions relative to this function.The application.
of the above stated performance criteria and associated bases approach for each unit would not be possible without the fact that the units are essentially symmetric.
The other aspect of the Maintenance Rule implementation process is the monitoring of plant level performance indications.
The plant level performance indicators are: unplanned scrams (automatic or manual);unplanned capability loss factor;and unplanned reportable safety system actuations.
~Unplanned scrams (automatic or manual)is the indicator that provides an indication of the success in improving plant safety by reducing the number of unplanned thermal-hydraulic and reactivity transients requiring reactor scrams.It also provides an indication of how well the plant is operated and maintained.
~Unplanned capability loss factor is an indicator monitoring the effectiveness of maintenance in minimizing outage time and power reductions that result from unplanned equipment failures or other conditions.
It reflects the effectiveness of plant programs and practices in maintaining systems available for safe electrical generation.
E-8 ik ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI-UNIT PRA, GENERIC LETTER 88-20 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED~Unplanned reportable safety system actuations criteria is established as no common cause (i.e., repetitive) reportable unplanned ESF actuations per unit or no more than 2 unplanned reportable ESF actuations per system, per unit, or no more than 4 total per unit all measured on a 24-month rolling interval.Again, the application of the above stated monitoring approach for each unit would not be possible without the fact that the units are essentially symmetric.
Containment Performance Im rovement Pro ram Im lementation The results of TVA's evaluation of containment performance improvement (CPI)program recommendations for Unit 2 was submitted to NRC with the Unit 2 individual plant examination on September 1, 1992 (Reference 3).The scope of the Unit 2 CPI program was further described by TVA in Reference 5 in response to NRC's November 19, 1993 request for additional information (Reference 4).The CPI recommendations as described for Unit 2 have been similarly implemented on Unit 3.The Hardened Wetwell Vent has been implemented on Unit 3 and is modeled in the Unit 3 PSA.The BWROG Emergency Procedure Guidelines (EPGs), Revision 4 have been implemented for Units 2 and: 3.Conclusion The common attributes listed above demonstrate the substantial design and operational similarities which exist between Unit 2 and 3.The Browns Ferry Unit 2 specific PRA and the Unit 3 specific PRA effectively reflect the integrated operation of Browns Ferry Nuclear Plant and meet the intent of GL 88-20 for the performance of an Individual Plant Examination.
E-9
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~~g~ENCLOSURE UNIT 3 INDIVIDUAL PLANT EXAMINATION (IPE)AND UNIT 2 MULTI UNI T PRAI GENERIC LETTER 88 20 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONTINUED REFERENCES 1)NRC letter to TVA dated June 19, 1997, Browns Ferry Nuclear Plant Units 2 and 3-Request for Additional Information Regarding the Multi-Unit Probabilistic Risk Assessment (TAC NO.M74386)2)NRC letter to TVA dated May 20, 1998, Browns Ferry Unit 3, Individual Plant Evaluation (IPE)and Multi-Unit PRA-Generic Letter 88-20 (TAC NO~M74384)3)TVA letter to NRC dated September 1, 1992, Browns Ferry Nuclear Plant (BFN)-Response to Generic Letter 88-20"Individual Plant Examination for Severe Accident Vulnerabilities
-10 CFR 50.54(f)" 4)NRC letter to TVA dated November 19, 1993, Request for Additional Information Regarding the Browns Ferry Unit 2 Individual Plant Examination (TAC NO.M74385)5)TVA letter to NRC dated December 23, 1993, Browns Ferry Nuclear Plant (BFN)-Response to Request for Additional Information Regarding the Individual Plant Examination (IPE)(TAC NO.M74385)6)TVA letter to NRC dated April 14, 1995, Browns Ferry Nuclear Plant (BFN)-Multi-Unit Probabilistic Risk Assessment (PRA)E-10
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