ML041270504
| ML041270504 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 04/08/2004 |
| From: | Exelon Generation Co, Exelon Nuclear |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML041270504 (29) | |
Text
1
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E ke c,,, EM Nuclear Risk-iinformed Technical Specifications
- Initiative 5b Relocation of Surveillance Test Intervals April 8, 2004 Presentation to NRC*
Attachment 2
'aI Agenda Ixe Nuclear
- Overview of Initiative 5b - David Helker
- Methodology/Guidance & Pilot Program -
Philip Tarpinian
- Process Considerations - Eugene Kelly
- PRA Model Quality - Donald Vanover
- Scope of Submittal - Glenn Stewart
- Next Steps - David Helker 2
'I Initiative 5b I Nuclear
- Relocate to licensee-controlled document
- Retain surveillance requirements in Tech Specs
- Surveillance test interval (STI) adjustment
- Change interval based on risk-informed process
- Tempered by performance and commitments
-bj MethodologylGuidance to e taM Nuclear
- NEI/BWROG methodology document
- Qualitative and quantitative reviews
- Risk significance of SSC's (high or low)
- Addresses modeled and un-modeled SSC's
- Candidates validate all "legs" of methodology
- Integrated Decision-making Panel (IDP)
- Addresses five criteria of Reg.Guide 1.177 4
Pilot Project Exe n Nuclear
- IDP charter established
- Evaluation record/panel minutes
- Commitment reviews
- Six candidate surveillances
- Observed by NRC, NEI, BWROG
- Equipment and surveillance history
- System manager/subject matter expert presentations
- PRA risk information provided 5
.1 Candidate Surveillances ESM Nuclear
- CRD notch testing
- SGTS/RERS flow
- 4kV under-voltage relays
- LOCA/LOOP logic
- Main steam isolation valve position (RPS)
- Redundant reactivity control system 6
.1 Process Considerations 14ce SM Nucl ear
- Analogous to Maintenance Rule, IDP concept
- Reliable and repeatable
- PRA model revisions ...effect on extensions
- Shortening the interval (two-way street)
- Extrapolation limit to data
- Phased implementation . ."baby steps" 7
S ;
Process Considerations BEEP& Wtty-mm@
L~xe jSnM Nuclear
- Checking for cumulative effects
- Resources, safety, and reliability
- Test methods or practices
- NRC SER for 5b methodology supercedes previous commitments to Regulatory Guides and Topical Reports for "frequency" (TS Bases)
- Interdependence of Maintenance Rule, PRA, and Surveillance Frequency Control Program 8
Process Considerations a@l..n Nuclear
- Capturing surrogate monitoring
- STI Evaluation Minutes
- Some surveillances confirmed as appropriate (e.g., main turbine valves)
- Other surveillances require additional information to adjust (e.g., LOCA/LOOP relays)
- Alternate testing and defense-in-depth ...other tests which exercise the component 9
PRA Model Qua ity lxe nHM Nuclear
- BWROG Peer Certification Report issued in 1999
- One A, 59 B, 97 C, 19 D & 21 S Facts & Observations
- 2001 Model update addressed majority of findings
- Model evaluated against ASME Standard and Regulatory Guide 1.200 -- Gap Analysis
- Areas not fully meeting Category IIof standard evaluated for impact on the application
- Uncertainty associated with 'gaps' explored in sensitivity studies
- Gap Analysis findings in 2004 PRA update 10
- Examine for appropriate level of detail
- Enhance as necessary
- Explicitly address At/2 failure modes
- Identify appropriate 'equivalent' basic events
- Common Cause Failure of relevant components
- Include uncertainty evaluation as input
- Specific assessments for each STI 1.1
Example Analysis Nuclear
- MSIV position switches and relays in Model
- Incorporate standby failure terms and CCF
- "De-modularized" related portions (address Gap)
- Verified no significant change to risk profile
- Re-run with adjusted intervals as input for the standby failure rates and associated CCF terms 12
c ft PRA Input to IDP a e tn SM Nuclear
- Best estimate ACDF and ALERF values
- Include uncertainty based on 95 th percentile
- Re-run 95th percentile value standby failure rates
- Extremely small changes in risk metrics
- Not unexpected ... multiple failures necessary to fail MSIV position logic; only one input to RPS logic
- Two-year interval approved
- Experiential information needed for relays at similar intervals; otherwise, phased approach 13
Archival Requirements 1n>
Nuclear
- RG-1 .200, Section 4
- Documentation provided at NEI
- LGS PRA Model and supporting documentation
-1999 Peer Review Report and responses
- GAP Analysis documentation
- Analysis files including sensitivity studies
- Detailed PRA quality assessment report
- Exelon PRA maintenance and update procedure 14
Submittal Documentation E t n A SM Nuclear
- Discussion of plant changes since last update, and impact on risk metrics
- Gap Analysis findings that impact application
- "Key assumptions" derived from sources of uncertainty
- Resolution of peer review comments and identification of open items that impact application
- Discussion of sensitivity study results 15
Scope of Submittal P Axe?9ISMrn lj Nuclear
- Relocate to licensee-controlled document
- Add Surveillance Frequency Control Program to LGS Technical Specifications
- Allow internal approval of candidate adjustments Overview of STI selection and revision process
- Discussion of the six candidate examples
- Summary of PRA Model quality (Reg. Guide 1.200) 16
Scope of Submittal Ixe n.M Nuclear
- Exclusion Rules
- Event-based with no time component
- Event-based with time component
- References to pre-established programs
- Certain values or parameters
- Surveillance Tables
- ST/STI exceptions (footnotes) and applicability (Operational Conditions) remain in TS
- STIs listed in TS by exception versus new "frequency notation" 17
Next Steps 3 nSM Nuclear
- Remaining STI candidates (Complete)
- Finalize submittal, methodology, TSTF (April 2004)-
- Submit Limerick Amendment Request (May 2004)
- Submit 5b Methodology and TSTF-425 (May 2004)
- Gap analysis (complete)
- NRC Audit (July 2004)
- Integral to application
- Develop administrative controls (later in2004) 18
RI-TS Surveillance Test Interval (STI) Evaluation Procedure # TBD BWROG RI-TS Initiative Sb Pilot (Ref. TSTF-425) Exhibit 1,Rev. D I (Exelon BRIM 132, LGS 2004 Bus. Plan Goal PR.05.LIM.03) Page I of II Station: Limerick Generating Station Unit(s): I & 2 I urvelilance Test Number (s): Revision Number(s):
ST-2-041 -616-1 25 ST-2-041 -61 7-1 25 ST-2-041 -618-1 27 ST-2-041 -619-1 24 ST-2-041 -616-2 15 ST-2-041 -617-2 15 ST-2-041 -618-2 18 ST-2-041 -619-2 19 Technical Specification Surveillance Requirement (SR) Number(s): 4.3.1.1-1, 4.0.5 Technical Specification SR (Text): Reactor Protection system instrumentation surv ance requirements - Main Steam Line Isolation Valve - Closure; Surveillance Requirements for Inservice don and Testing of ASME Code Class 1 2 and 3 Components: _
3.3.1 As a minimum, the reactor protection syst n channels shown in Table 3.3.1-1 shall be OPERABLE with thl N SYSTEM RESPONSE TIME as shown in Table 3.3.1-2. -
TABLE4 19l ItUMrlM I-- m fII -1NA~n t -t .?FRntn K1!15 I A I XU! 5VIIVLI LLAILZ ncx UIItCMCA I a awaaNE CHANNEL FUNCTIONAL CANNEL CONDITIONS FOR WHICH FUNMONALUNIT CHECK TZEST CAUBATIO~iII SIIRJLLOANCE RE01JJIED 1.2 S. Main Steamo Lisie Isolation Valve - Closure N.A. Q R l Technical Specification SR Bases (and Intent):
- 5. Main Steam LIne Isolation Valve-Closure The main steam line Isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events. The MSIVs are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature, and low steam line pressure. The MSIVs closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermaVhydraulic Safety Umits.
The MSIV position scram Is not credited in the design basis RPV overpressure analysis, nor Is It required for mitigation of limiting design basis events analyzed in the reload licensing analysis.
The MSIV/RPS functional test exercises both the MSIV RPS limit switch, and the associated RPS logic via the K03" relay.
Attachment 3
RI-TS Surveillance Test Interval (STI) Evaluation Proce .D BWROG RI-TS Initiative 5b Pilot (Ref. TSTF-425) Exhibit 1, Rev. D (Exelon BRIM 132, LGS 2004 Bus. Plan Goal PR.05.LIM.03) Page 2 of 11 Recommended STI Change: Extend current quarterly (92 days) functional testing, to refuel (24 months) frequency, using a phased approach.
Station Benefit: Station load reductions are taken on a quarterly basis for turbine valve testing and MSIV/RPS testing. This change reduces the length of these on-line load drops (saves MegaWatts), and reduces potential for Inadvertent MSIV closures.
(NOTE: Future Exelon T.S. revision reest will ursue relocation of STI from T.S. to TRM, and STI extension) 1 lm 'wrap.,VIN, 0111% _N;-11 1.10C r
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RI-TS Surveillance Test Interval (STI) Evaluation Procedure # TBD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF-425) Exhibit 1, Rev. D I (Exelon BRIM 132, LGS 2004 Bus. Plan Goal PR`05.LlM.03) Page 3 of II A. SYSTEM & MAINTENANCE RULE (MRule) INFORMATION SYSTEM NUMBER: 041 C: Nuclear Boiler (FW & MSIVs) 071: RPS Current MRule R-S Classification: HSS Current MRule R-S Basis: (from System 41C & 71): "The SSC Is explicitly modeled by the PSA and is quantitatively risk significant. The SSC was assessed by the Expert Panel to be risk significant."
Current PRA Model: LLGSIOIRI/LGS201RI Current PRA R-S Classification (System) _Risk significant _ [values below pertain to System 071]
Current PRA RAW (System): 21.92 (MRULE R-S threshold: 2.0)
Current PRA RRW (System) 1.047 (MRULE R-S threshold: 1.005)
Current PRA Limiting Cutset (System) 4418 (MRULER-SAb'hold: top 90%)
NEI 00-04 R-S Insights MRule HSS classification retainedas allowe g Sb Methodoloev guideline document.
RI-TS Surveillance Test Interval (STI) Evaluation Proe. .BD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF425) Exhibit 1, Rev. D (Exelon BRIM 132, LGS 2004 Bus. Plan Goal PR.05.LIM.03) Page 4 of 11 COMMITMENT REVIEW (Is STI credited in any commitments?) Commitments for MSIV ST ST-2-041-616-1 Main Steam Line Isolation Valve - Closure; Division IA,Channel Al Functional Test T03612 LER 2-94-06, TS 4.0.5 requirement T03878 LER 1-96-013, Scram during MSIV test TS 3.3.1 Reactor Protection System Instrumentation 4.3.1.1 Channel Check, Functional and Calibration per Table 4.3.1.1-1 MSIV Func=Quarterly TS 4.0.5 IST program ST-2-041-616-1 Ref. 6.9 Specification ML-008, Pump & Valve Inservice Testing (IST) Program.
Not full open alarm and test box channel trip I steps In ST-2-041-616-1.
UFSAR (page 1.13-99) li.K.3.16 REDUCTION OF CHALLENGES AND FAILURES OF RELIEF VALVES - FEASIBILITY STUDY AND SYSTEM MODIFICATIONS
- 3) Reduce MSIV testing frequency (see section 6.3.1.4.4 of Reference A number of Isolation events occur when the MSIV closure tests are being conducted. Redu test frequency would result Ina reduction in the number of Isolation events. Appropriate reduLn made to the frequency of testing for the LGS MSIVs.
7.2.2.1.2.3.1.9 RPS - IEEE 279 (1971), Paragraph 4.9 - C r h The MSIV position switches are te d n at caus e limit switches to operate at the setpoint value of the valve positi SER SER for power uprate 2-22-95, 3.2.7 Main Steam Isolation Val MSIV performance will be moni d t rveillance requirements Inthe technical specification to ensure original licensing basis f areserved. Maintenance of MSIV performance to existing licensing basis standards Is acceptable t if.
CT database: No PLS regardin frequency were identified.
Regulatory Guides Regulatory Guide 1.22 (1972) - Periodic Testing of Protection System Actuation Functions (Safety Guide 22)
Regulatory Guide 1.118 (June 1978) - Periodic Testing of Electric Power and Protection Systems IEEE 338 (1975) - Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems (also ref. 1977 and 1987) (Protection Systems changed to Safety Systems In1977) 6.5 Test Intervals 6.5.2 Change of test Interval IEEE 577-1976 Standard Requirements for Reliability Analysis In the Design and Operation of Safety Systems for Nuclear Power Generating Stations 4.4.3 changes in test intervals per IEEE 338 and IEEE 352 IEEE 352-1987 Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Safety Systems IEEE 279 (1971), Criteria for Protection Systems for Nuclear Power Generating Stations RPS meets Intent (per TS Bases).
NEDC-30851P-A, TS Improvement Analysis for BWR RP.S Used to determine RPS surveillance Intervals (per TS Bases)
GL 93-05 Line-item TS Improvements to reduce Surveillance Requirements for Testing During Power Operation RPS functional test frequency may be changed from monthly to quarterly Conclusion Based upon review of the above, no commitments were identified prohibiting this STI change.
RI-TS Surveillance Test Interval (STI) Evaluation Procedure # TBD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF425) Exhibit 1, Rev. D I (Exelon BRIM i 32, LGS 2004 Bus. Plan Goal PR05.LIM.03) Page 5 of II 2 SURVEILLANCE TEST HISTORY OF THE COMPONENTS AND SYSTEM ASSOCIATED WITH THE STI EXTENSION:
The RPS/MSIV surveillance test history was reviewed. In a history of over 800 tests, there was apparent occurrence of failure of the MSIV-RPS input logic. This resulted In a reactor scram. See item 5 below.
3 RELIABILITY REVIEW:
PERFORMANCE (OPERATION & MAINTENANCE) HISTORY OF THE COMPONENTS AND SYSTEM ASSOCIATED WITH THE S1 EXTENSION:
NOTE: THE MSIV-RPS LIMIT SWITCHES ARE DONATED TO THE MR SYSTEM 71 (RPS).
MRule Train Actual Unreliability: not monitored, MRule Unreliability Perfonnance Criteria: N/A MRULE MPFF and FF history here for MSIVs and RPS logic:
A review of the Maintenance Rule functional failures was performed for system 41C. One functional failure was identified, but it was due to a failed outboard DC solenoid. Since It involved an unplanned load drop greater than 20%, It was determined to be a functional failure. There have been no MPFFs on system 41C since Maintenance Rule monitoring has been in effect. (July 1996). _
A review of Maintenance Rule functional failures was also perfoi P171. One functional failurewas identified, but it was due to failure of PS-0010202D to actuate. j ien no MPFFs on system 71 since Maintenance Rule monitoring has been in effect (July 1j) I Additional PIMS component history review:
A review of all maintenance histos th P vlEvis was pWrmed. Those corrective maintenance activities perfor h her wiring issues, or occurrences where the RPS limit switch did not reset t A review of all maintenance hi y Ih P (K03) was performed. Since 1993, 5 corrective maintenance items were crea T e s re for buzzing relays, one was for a failed response time test, and the last one replaced two ys part of the 5/22/96 scram Investigation discussed in item B-6.
Failure Rate Extrapolation Data ( RA Use In Section C-2)
Limit Switches:
Surveillance Test history of other, identical limit switches In other applications was reviewed to support extension of the ST Interval.
Results Other LGS plant equipment that use the NAMCO EA740-501 00 limit switch are:
ZS-051-*51A(B)
ZS-051-*41A(B, C, D)
The above limit switches are exercised during cold shutdown testing (typically I R frequency). Since 1994, 112 tests/partial tests were performed; 20 had unsatisfactory results; of those 20 tests, none were the result of failure of the NAMCO limit switches.
Secondly, a corrective maintenance history review for the above limit switches was performed. No corrective maintenance action requests were created as a result of failure of the NAMCO switches.
K03 Relays:
Later 4 UNAVAILABILITY REVIEW:
MRule Train Actual Unavailability. N/A MRule Unavailability Performance Criteria: NIA No unavailability Performance Indicators exist for MSIVs or MSIVs/RPS. RPS train unavailability Performance Indicators apply to entire RPS train, not a single RPS input such as MSIV Closure.
RI-TS Surveillance Test Interval (STI) Evaluation Procm BD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF-425) Exhibit I, kev. D I
(Exelot BRIM 132, LGS 2004 Bus. Plan Goal PR.05.LIM.03) Page 6 of 11 5 PAST INDUSTRY AND PLANT-SPECIFIC EXPERIENCE WITH THE FUNCTIONS AFFECTED BY THE PROPOSED CHANGES Limerick had previous operating experience associated with test failures of the MSIV-RPS limit switches. The problem was that the limit switches did not reset following actuation. This problem was resolved with a design change to the limit switch roller size Implemented In 1995. No similar failures have occurred since the design change was implemented.
On 5/21/96, during quarterly MSIV-RPS testing on Limerick Unit 1, a reactor scram occurred. The scram was initiated by high reactor pressure. There was a failure of the RPS logic to Initiate the "MSIV Not Fully Open" scram signal, although no specific failed components were found.
On 2/19185, Quad Cities unit 2 scrammed from 96 percent power while performing the biweekly MSIV surveillance test. During this test, each valve is cycled 10 percent closed using a dedicated test push button.
The purpose of the test is to verify operability of the MSIV not full open' reactor protection system (RPS) logic.
This logic utilizes a limit switch mounted on each MSIV which is actuated at the 10 percent closed position.
The logic, which was designed to reopen the MSIV, failed. A similar event oqred at Dresden on 7/12/86.
Brunswick has had numerous problems with exceeding the TS a ir MSIV-RPS limit switch location. Limerick's calibration history has shown very good rem S function actuates within allowable limits). _A On 7/12/82, Fitzpatrick failed a quarterly h At sup being out of adjustment.
The above operating experience s enhancements, differences In'I proposed change does not alt I I1 iterval elision, based on internal design r plants, and in-plant performance. The
-4 6 VENDOR-SPECIFIED M, The LGS Environmental Qua Pm lctates replacement of the MSIV-RPS limit switches.
Maintenance of these switcho ed by the change to the frequency of functional testing.
7 ASME AND OTHER CODE- TEST INTERVAL The Limerick Inservice Testing (IST) program has a cold shutdown test justification for the MSIVs. The justification states that full stroke exercising of the MSIVs requires Isolation of one of the four main steam lines, which results In reactor power fluctuations, primary system pressure spikes, and increased steam flow In the non-isolated steam lines.
Section 4.2.4 of NUREG 1482 discourages MSIV testing at power, including partial stroke testing.
Nevertheless, LGS had continued to perform partial stroke testing as a means to verify valve performance.
Based on this existing IST documentation, a change will also be required to Specification ML-008. This change will be made to delete the statement that partial stroke testing satisfies the LGS IST program. This change can be made via ECR to post ML-008. Based on discussions with the IST Program Manager, the IST change will be an internal document change only, and will not require a relief request.
RI-TS Surveillance Test Interval (STI) Evaluation Procedure # TBD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF-425) Exhibit I, Rev. D I _
(Exelon BRIM i32, LGS 2004 Bus. Plan Goal PR.05.LIM.03). Page 7 of 11 8 OTHER QUALITATIVE CONSIDERATIONS (include (a) Comparison to Improved T.S., (b) Alternate ST Test List [retained], (c) LCO Review is optional)
(a) Comparison to ITS: Improved Technical Specifications Include a functional test of the MSIV-RPS feature on a 92 day frequency. There is no discussion with regard to the basis of the frequency.
(b) Related Tss on MSIVs or MSIVIRPS logic: There are no other tests that exercise the MSIV-RPS logic for which credit can be taken on a 92-day frequency.
Additional considerations: As a prerequisite to the current 92-day ST, reactor power must be reduced to below 95%. The purpose of this power reduction Is to Amit the plant impact in the event of an Inadvertent MSIV closure.
Recently, a single point vulnerability concern due to MSIV testing was raised. Mechanical failure of the MSIV test solenoid would result In the complete closure of the MSIV. A recent qualitative evaluation concluded, "Based on LGS operating history, test, and simulator case results, Inadvertent reactor scram resulting from MSIV closure can be avoided with high confidence if initial power level Is reduced to q or below. '
-I 9 IMPACT ON DEFENSE-IN-DEPTH PROTECTION.
The operation of the RPS relays for MSIV closure is L eext ion of the surveillance test interval. The diversity of scram Inputs also remains u IexteU n of the STI, given that reliability remains the same does not affect the def ept he nse In depth protection of the plant remains unchanged. _
10 THE IMPACT OF SYSTEMS EVENT PRA All relevant systems were Inclui Itmi iandtication thus there Is no Impact of non-modeled systems.
II THE IMPACT OF SYSTEMS PFIC17LERF RESULTS ARE NOT AVAILABLE The Large Early Release Frei (LERF) Is calculated for full power internal events and Is addressed quantitatively In section C.
12 THE IMPACT OF SYSTEMS FOR WHICH EXTERNAL EVENTS AND SHUTDOWN PRA ARE NOT AVAILABLE There Is no quantitative external events (fire and Seismic) or shutdown PRA available for this analysis. For fire risk, the fire areas that would cause an MSIV closure plant scram remain unchanged and given that component reliability Is maintained the fire risk Is unchanged. For Seismic risk, the RPS Input would generally be related to a loss of offsite power not MSIV closure, so seismic risk is unaffected. For shutdown risk, the reactor Is subcritical and RPS Is not required to shutdown the reactor so there Is no Impact on shutdown risk.
The K4SIVs are also generally closed during most shutdown plant operating states and the RPS Input for closure of the MSIVs Is irrelevant.
13 UNCERTAINTY ASSOCIATED WITH THE QUANTITATIVE (PRA) PROCESS A parametric uncertainty analysis was performed on the quantitative internal events CDF and LERF I calculation. The 9 5th percentile values compared to the point estimate mean values were shown to be about 3 for CDF and about 4 for LERF. A factor of 3 or 4 would not significantly Impact the conclusions from this analysis. Additionally, a focused uncertainty assessment which Increased the relevant standby failure terms by a factor of 3 also showed that there would be a very negligible change in the calculated change in CDF and LERF risk metrics.
RI-TS Surveillance Test Interval (STI) Evaluation ProceUz !BD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF-425) Exhibit 1, Rev. D I (Exelon BRIM 132. LGS 2004 Bus. Plan Goal PR.05.LJM.03) Paae 8 of I1 14 QUALITATIVE ANALYSIS - CONCLUSIONS Based on the need for reactivity manipulations to perform this test, the single point vulnerability of test solenoid failures, and the excellent recent performance of the MSIV-RPS limit switches, revision of this TS surveillance Interval Is warranted. NUREG-1482, Section 4.2.4 quotes from the revised standard TS bases, MSIVs should not be tested at power, since even a part-stroke exercise Increase the risk of a valve closure when the unit Is generating power.'
15 PHASED IMPLEMENTATION RECOMMENDATIONS Based on the surveillance test, maintenance rule, and availability history, It Is recommended that quarterly surveillance testing be extended to a refueling outage(1 R)frequency. Calibrationtfunctional testing will continue on a refueling outage (lR) frequency. A phased approach to this extension Is needed to address the lack of performance data at greater than the current quarterly frequency. The first phase will be to a 6 month Interval with further extension to 1 year after 2 successful performances at six-month Intervals, followed by extension to I Rafter two successful performances at 1year Intervals. _
16 PROPOSED SURROGATE MONITORING RECOMMENDATIONS: (CD pof Existing MRule monitoring)
No other testing of the RPS K03" relays Isperformed and the m, rformance criteria would not Indicate degradation of the tested components so a n on uld be developed.
Suggested monitoring isof tracking of ST functional testinfol
RI-TS Surveillance Test Interval (STI) Evaluation Procedure # TBD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF-425) Exhibit I, Rev. D V (Exelon BRIM 132, LGS 2004 Bus. Plan Goal PR.05.LIM.03) Page 9 of 11 I- *i- 4 C. PRA (QUANTITATIVE) ANALYSIS QI check if not modeled in PRA 1 OVERVIEW OF PRA MODELLING of STI (include bounding risk analysis techniques if used, and PRA Quality Issues)
The subject surveillance requirement verifies the capability of RPS to initiate a scram on MSIV closure. This Is modeled In the Limerick PRA. For the purposes of this analysis a standby periodically tested failure was added to the model for the MSIV position switches. The failure rate used was from the EPRI ALWR database using a level switch as a surrogate component since no failure rate data Is available.
The Limerick PRA has been reviewed against the ASME PRA standard and DG-1 1221R.G.1.200. The identified gaps that affect this analysis were addressed as follows.
Existing modules that related to the RPS logic were expanded.
The parameter file was fully populated for performance of the parametric uncertainty assessment.
There is very limited use of plant-specific data In the LGS models. The Incorporation of more plant-specific data and the Incorporation of revised common cause failure probabilities are anticipated, for the most part, to reduce the failure probabilities used in the model. This means that the currently calculated Impacts are most likely conservative.
Other remaining Issues from the GAP analysis are considered as b ed by the uncertainty assessment performed for this analysis.
I The above changes created the application specific PRA m 01R3 and LGS201R3 >
.m! ).% The Unit I application model base Cd s 4. -/ . w Is equivalent to the base Unit 1 level I PRA model CDF. The Unit 1 base applicatiorm el 4.4 /yr. which Is equivalent to the base Unit 1 model LERF. The Unit 2 a p m F alen the Unit 2 base model CDF.
There Is no quantified Unit 2 level 2 m oc
-ddgk __ I 2 FULL POWER INTERNAL E I E=L IMPACTS (CDF Comparison against R. . r The change in CDF Is less th . I igeiitest Interval from once per quarter to once per 2 years.
This Is much less than the R I I llof -/r Note: CDF = Core Damage ncy 3 FPIE LEVEL 2 PRA MODEL IMPACTS (LERF Comparison against R.G 1.174 limits)
The change in LERF Is less than 1E-1 1/yr when extending the test Interval from once per quarter to once per 2 years. This Is much less than the R.G. 1.174 limit of IE-7/yr.
Note: LERF = Large Early Release Frequency 4 FIRE RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)
Based on the very small Impact on the FPIE model, it is judged that the impact from fire risk Is also very small though no quantitative assessment Is available.
5 SEISMIC RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)
Based on the very small Impact on the FPIE model, It is judged that the impact from seismic risk is also very small though no quantitative assessment is available.
6 SHUTDOWN RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)
RPS is not required during shutdown so there is zero Impact on shutdown risk.
RI-TS Surveillance Test Interval (STI) Evaluation Proceaure # TBD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF-425) Exhibit 1, Rev. D I (Exelon BRIM 132. LGS 2004 Bus. Plan Goal PR.05.LIM.03) Page lO of Il 7 OTHER PRA ISSUES (ex. Impacts from Other External Events excluding Seismic & Fire Risk Impacts)
Sensitivity cases with Increased failure probabilities for the tested component Indicate a negligible Increase In risk even when 3-times the extended Interval failure rates are utilized. The parametric uncertainty assessment Indicates that the 95t percentile CDF value Is about a factor of 3 times the base point estimate mean value, and the 95w' percentile LERF value Isabout a factor of 4 times the base point estimate mean value. The low change InCDF and LERF for this assessment indicates that even at the 95kt percentile values for CDF and LERF, the change Inrisk would be of very low risk significance.
I No other external events are applicable. Internal flooding is contained In the Internal events PRA model.
8 CUMMULATIVE EFFECT OF ALL RI-TS STI EXTENSIONS ON INTERNAL, EXTERNAL & SHUTDOWN PRAs. (CDF & LERF Comparison against R.G 1.174 limits)
The previous cumulative Increase InCDF was 2 . With this STI the cumulative Increase in CDF remains approximately By. The previous cumulative Increase In LERF was B e With this STI the cumulative Increase In LERF remains approximately, T_ These cumulative values remain within the RG 1.174 limits.
9 QUANTITATIVE (PRA) ANALYSIS - CONCLUSIONS The change to the surveillance test Interval Is of very low risk si a CDF and LERF perspective as evidenced by the changes remaining within RG 1.174 limit The change Is within R.G. 1.174 limits when quanUfied th a onth fr ency. However this Interval may be extrapolating the existing failure rate beyond curr vai Ota as not recommended unless good performance can be noted from similar corn nts e tes at a 2 year interval. Otherwise, a phased implementation Is recommend e for of confice Inthe failure rate at Intervals beyond 3 months. _ X ^ k 10 PREPARER (SECTION C -
Prepared by: Victoria WI ate 01/19/04 (Risk Manag Revised by: -Don VanovE P. Date: 04/05/04 I/i!^I & A_ ._ _.ws~
RI-TS Surveillance Test Interval (STI) Evaluation Procedure # TBD BWROG RI-TS Initiative 5b Pilot (Ref. TSTF425) Exhibit 1, Rev. D I (Exelon BRIM 132, LGS 2004 Bus. Plan Goal PR.05.LIM.03) Page 11 of II D INTEGRATED DECISION-MAKING PANEL (IDP, a/kIa EXPERT PANEL) MEETING I Presenter(s): -
2 Meeting Discussion: (Review of Qualitative and Quantitative Analyses, and Cumulative Impact) 3 Meeting Results / Recommendations I Bases: (Consider phased implementation, additional performance monitoring of failure rates) 4 Approval I Disapproval: Check one of the following:
o STI Approved O STI Approved with Comments o STI Disapproved IDP I Expert Panel Members: ti ees:
es n required - see MRule Expert Panel / IDP meeting minutes)
I. Engineering Manager *
- 2. Maintenance Manager _ _ ____ _
- 3. Operations Manager *
- 4. Risk Management (PRA) Engi__
- 5. Maintenance Rule Coordinator
- 6. Surveillance Test Coordinator
- 7. System Manager or Component Engineer
_ also Maintenance Rule Expert Panel Member 5 IDP / Expert Panel Coordinator Final Review I Closure:
._ Date:
(IDP Coordinator)