NLS2004081, Response to Request for Additional Information Regarding Risk-Informed Relief Request RI-34
| ML042160125 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/29/2004 |
| From: | Edington R Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2004081 | |
| Download: ML042160125 (7) | |
Text
Nebraska Public Power District NLS2004081 Always there when you need us July 29, 2004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Response to Request for Additional Information Regarding Risk-Informed Relief Request RI-34 Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46
Reference:
- 1. Letter to R. Edington (Nebraska Public Power District) from U.S. Nuclear Regulatory Commission dated May 20, 2004, "Request for Additional Information Regarding Risk-Informed Relief Request RI-34 (TAC No.
MC2351)."
- 2. Letter to U. S. Nuclear Regulatory Commission from S. Minahan (Nebraska Public Power District) dated March 11, 2004, "Risk-Infonned Inservice Inspection Program (Relief Request RI-34)" (NLS2004023).
The purpose of this letter is for the Nebraska Public Power District (NPPD) to respond to the Request for Additional Information provided in Reference 1 by the Nuclear Regulatory Commission (NRC) regarding the previously submitted Relief Request of Reference 2.
Question 1 Regulatory Guide (RG) 1.178, An Approachfor Plant-Specific Risk-Informed DecisionmakingforInservice Inspection of Piping, Revision 1, datedSepteniber 2003, replaced the original "For Trial Use" RG dated September 1998. Revision I of the RG 1.178 includes guidance on what should be included in risk-informed inservice inspection (RI-ISI) submittals,particularlyin dealing with probabilistic risk assessment (PRA) issues. Specifically, on Page 28 of RG 1.178, thefollowing is stated regardingthe information that should be included in a submittal:
A descriptionof the staff and industry reviews perfornedon the PRA.
Limitations, weakness, or improvements identified by the reviewers that could change the results of the PRCshould be discussed. The resolution of the reviewer comments, or an explanation of the insensitivity of the analysis used to support the submittal to the comment, should be provided.
- a. Our review of the NRC StaffEvaluationforyourIPE appears to indicate no weaknesses with that document. Please confirm that this is your understanding, or indicate 1) what weaknesses were identified and 2) what was done to correct the identified weaknesses, or why the uncorrected weaknesses are not relevant to this application.
COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com
NLS2004081 Page 2 of 6 Response: NPPD confirms that the NRC Staff Evaluation Report for the Cooper Nuclear Station (CNS) Individual Plant Evaluation (IPE) submittal did not identify weaknesses with that document. However, there were three plant improvements noted in the IPE submittal (and were acknowledged in the Staff Evaluation) that merited incorporation in the IPE:
- A hardened torus vent system was added to the design configuration of primary containment.
- The Emergency Operating Procedures were revised to permit earlier spraying of the containment to help mitigate drywell liner melt-through accidents.
- CNS has performed extensive modifications to environmentally qualify components to continue functioning in harsh environments under accident conditions.
Changes were made to the IPE to reflect these activities prior to the development of the Risk-Informed Inservice Inspection (RI-ISI) submittal. These improvements resulted in a reduction in the overall CNS risk profile.
Question 2: Section 1.2 notes that as a result of the tvo industry reviews of the CNS PRA, you are currentlyperforming a major revision to the PRA and that this particular revision was not used in the preparationof your submittal Front this discussion, four areasofyour PRA were reported to either be in revision or are "opportunitiesfor improvement ":
S InternalFloodingAnalysis (from the first of the tvo peer reviews)
S InitiatingEvent (7E) Analysis (from the latter ofthe tvo peer reviews) 0 Data Analysis (from the latter of the two peer reviews) 0 Human Reliability Analysis (HRA) (from the latterof the tvo peer reviews)
Forthefirst item you noted that "certain conclusions regardinginternalflooding were consideredqualitatively and reviewed against the most currentplant informationfor potential insights." You contended that the latter three items should not impact the consequence rankings determination,mostly because the "the risk importance of the systems in the RI-ISIprocess is dominated by the LOCA events." As discussed below, the NRC staffrequests additional information on thefirst (flooding) and thefourth (HRA) items.
The NRC staff concurs with yourjudgment that revisions to the IE Analysis should have very little influence on the consequence rankings ofpipe segments, and hence, on this PRA application,because the ConditionalCore Damage Probability(CCDP)for pipe breaks that causeIEs is determined by dividing out the initiatingeventfrequency (IEF) term, leaving CCDP unchanged, regardlessof IEF changes andforpipe breaks that impact mitigatingcapability, the CCDP should not be drasticallyaffected by the "composite" change in IEF terms, since
NLS2004081 Page 3 of 6 the change in CCDPwill be proportionalto the difference in IEFs in each one of the "delta cutsets ". Barringa highly unusual trend,changes in JEFare expected to be either small or negative (current trend of events in the industry continues to decline).
The NRC staffalso concurs with yourjudgment that revisions to the Data Analysis should have very little influence on the consequence rankings ofpipe segments, primarilybecause it is likely that a revisedDataAnalysis will generally reflect both increasesand decreases in the availabilityand reliabilityof equipment, resulting in offsetting impacts on core damagefrequency (CDF) calculations (bothfor the case where thepipe segment 's surrogateBasic Event is set to TRUE as well asfor the base case). However, even if there were a significantchange in the performance of certain equipment in recent years, the impact on the DataAnalysis (i.e., the Basic Event values) would be dampened by the Bayesian Updateprocess. Because of overall improvements in equipment performance in recent years, there is likely a bias toward lower CDFs,and hence, slightly lower CCDPs. Thus, it appears that there is little change that revisions to the DataAnalysis would impact the CCDPof any pipe segment to the extent that its consequence ranking would need to be elevated.
- a. Thefirst area of improvement listed above suggests that you arepreparing to incorporatean upgraded internalfloodingmodel into your PRA.
Please explain the above statement about "certain conclusions regarding internalfloodingwere consideredqualitatively... " (from page 3 of29).
Pleaseprovide a description of the conclusionsyou have considered qualitatively and explain why they have no impact on this RI-ISI application, including a discussion about any newfloodingpropagation or spatialimpact scenariosthat may also be applicable to the consequences ofpipe breaks being consideredin this application.
- b. Sign~ifcant changes to the HRA could be influential. Revised HRA methodology can sometimes cause signifcant revisions to Hunman Error Probabilities(HEPs), which are not dampened by a Bayesianprocess.
When several of the dominant cutsets ofyour PRA contain the same Human ErrorBasic Event and equipment Basic Event used as a surrogate for a specific pipe segment, and a revisedHRA significantly increases this HEP, it could result in a significant increaseto the CCDPof that pipe segment, resulting in an elevated consequence ranking. Pleaseprovide additionalinformation to supportyour contention that these changes should not impact the consequence rankings.
Response: The 1996b Probabilistic Safety Assessment (PSA) model revision used to develop the RI-ISI submittal incorporated improved models and methods as well as relevant plant changes, such as those mentioned in the response to Question 1 above. This is consistent with NPPD's commitment to maintain a living PRA
NLS2004081 Page 4 of 6 model for CNS. It should therefore be noted that the risk effects of future upgrades to the PSA model would be evaluated for impact on the approved RI-ISI Program.
2a. Information provided in the RI-ISI submittal regarding future improvements in the internal flooding evaluation address comments from the PSA Certification. The PSA Certification comments pertained to formally incorporating more recent Probabilistic Risk Assessment (PRA) insights into the Level 1 backup documentation, rather than correcting weaknesses.
Specifically, it indicated that the documentation of the PSA should be enhanced to ensure it is properly maintained and revised as needed to support the living PSA applications in the future. Despite these future changes, NPPD would not expect the risk conclusions for internal flooding used with the RI-ISI submittal to change significantly.
The IPE did not quantify the core damage from internal flooding, but rather qualitatively evaluated the internal flooding risk to CNS, as provided for in NUREG/CR-4767. The following observations support the conclusion that the internal flooding contribution to plant risk is generally insignificant:
- 1) Station flooding calculations concluded that backflow into the Emergency Core Cooling System corner rooms would not result in water levels exceeding the maximum safe operating levels for the credited safety-related equipment in these rooms.
- 2) CNS does not have non-seismic water bearing lines routed over safety related electrical switchgear (in contrast to similar vintage plants).
More recently, a sensitivity study evaluated the relative risk impact of breaks outside of Primary Containment to address specific harsh environment concerns related to the Equipment Qualification Program. These results indicate that individual sequence contributions are not significant when compared to the IPE internal events Level 1 model results.
The RI-ISI Consequence Evaluation is based on an independent spatial evaluation of flooding impact, including walk-downs, for RI-ISI scope of piping. The RI-ISI evaluation did not identify any spatial dependencies that had not been identified in the IPE.
2b. It is understood that HRA methodologies have matured since the IPE was developed, and several have emerged as the preferred approach within the industry. This observation is reflected in the most recent PSA Certification comments, with the recommendation that such methods should be incorporated into a PSA model upgrade, as appropriate, during normal model review and maintenance schedule. NPPD intends to implement these
NLS2004081 Page 5 of 6 recommendations, and thereby enhance significantly the ability to perform absolute risk determinations that take into account the relevant HRA elements in a number of areas. However, it is not expected that these improvements would impact the consequence rankings established in the RI-ISI analysis, mainly because the revised HEP values are expected to remain the same or be reduced. Accordingly, currently planned model revisions are not expected to increase the risk category resulting in new high consequence pipe segments.
Question 3: The paragraphat the bottom ofpage 2 of 29 refers to a 1996b Level 2 PRA model. Please confirm that it was intended to refer to a Level 1 PRA model.
Response: NPPD confirms that the statement in the paragraph at the bottom of page 2 of 29 should refer to the 1996b Level 1 PRA model. The revised statement should read as follows, "The results of the 1998 Level 2 model and 1996b Level 1 are integrated into the updated CNS PRA."
Question 4: Section 3.6.1 indicates that you used the "Simplified Risk Quanti icationMethod as describedin Section 3.7 of the ElectricPowerResearch Institute (EPRI) topical report (TR)-1 12657, in support ofyour overall risk impact assessment.
You selected a value of IE-08 per weld-year as the pressure boundaryfailure (PBF)frequencyfor a weld with no known degradationmechanism (i.e., low failurepotential) and a value of2O times that (i.e., 2E-07) for a weld with medium failurepotential, which is similar to the failure rate used by a couple of the pilot plantsfor the EPRI TR-112657, as noted byyour citation ofReferences 9 and 14 in EPRI TR-112657.
- a. Given this information, as an example, a Category4 weld should have a contribution to CDF of (JE-3) *(JlE-8), or JE-I 1/year. Assuming that the inspections are 100 percent effective infindingflaws before they progress to a rupture, then the decrease of one weld inspection should result in an increase in CDFof IE-I 1/year. Table 3.6-1 (as well as the Table on page 12 of 29) which present the risk impact results, indicates a net decrease of one system NB category 4 weld inspection, resulting in a CDFincrease of SE-12/year. Pleaseclarify this apparentdiscrepancy.
- b. Many of the numerical entries in Table 3.6-1 have the same CDFor large early releasefrequency (LERF) values in the "wIPOD [probabilityof detection]" column as in the "w/o POD column (most of these are Category 4 welds). Some of the entries have different CDFand LERF values. Please explain why sometimes the "wxPOD and w/o POD values are different, andsometimes they are the same.
Response: 4a. The 5E-12 values in Table 3.6-1 include the probability of detection (POD=0.5 for Category 4). Note that POD does not change for Category 4
NLS2004081 Page 6 of 6 for both cases "Nv/POD" and "w/o POD" because there is no degradation mechanism associated with Category 4. POD changes only apply to welds that have Thermal Fatigue and no other degradation mechanism assigned.
4b. As explained above in the Response to Question 4a, POD only changes when Thermal Fatigue applies with no other degradation mechanism type.
Otherwise it is always 0.5. For Thermal Fatigue, POD=0.5 for the "w/o POD" case, which means no credit is taken for improved RI-ISI POD. For Thermal Fatigue "w/POD" case, which credits improved RI-MSI POD due to improved methods, POD=0.3 for the SXI program and 0.9 for the RI-ISI.
Should you have any questions concerning this matter, please contact Mr. Paul Fleming at (402) 825-2774.
all K. Edington Vice President - Nuclear and Chief Nuclear Officer
/wrv cc: Regional Administrator USNRC - Region IV Senior Project Manager USNRC - NRR Project Directorate IV-1 Senior Resident Inspector USNRC NPG Distribution Records
I ATTACHMENT 3 LIST OF REGULATORY COMMITMENTSl Correspondence Number: NLS2004081 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITTED DATE COMMITMENT OR OUTAGE None PROCEDURE 0.42 l REVISION 15 l PAGE 18 OF24