1CAN011003, Response to Request for Additional Information Request for Alternative - Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716

Response to Request for Additional Information Request for Alternative - Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716

ML100210159
Person / Time
Site:	Arkansas Nuclear
Issue date:	01/20/2010
From:	David Bice Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN011003
Download: ML100210159 (19)

Text

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4710 David B. Bice Acting Manager, Licensing Arkansas Nuclear One 1CAN011003 January 20, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Response to Request for Additional Information Request for Alternative - Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716 Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51

REFERENCES:

1. Entergy letter to the NRC, dated June 11, 2009, Request for Alternative

- Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716 (1CAN060902)

Dear Sir or Madam:

In Reference 1, Entergy Operations, Inc. (Entergy) requested authorization to implement a risk-informed Inservice Inspection program based on the American Society of Mechanical Engineers (ASME) Code Case N-716, as documented in the Request for Alternative ANO1-ISI-014.

Subsequent to the submittal, the NRC determined that additional information was required to complete their review and provided a request for additional information (RAI) in an email dated November 2, 2009. The attachment to this letter contains the RAI and associated responses.

As noted during a November 10, 2009, conference call, the Arkansas Nuclear One, Unit 1 (ANO-1) Probabilistic Safety Assessment (PSA) model was updated and, in August of 2009, an industry peer review of the model and supporting documentation was performed. The results of the peer review were favorable; however, several issues were identified that required further enhancement. The results of the peer review relating to the internal flooding analysis have been evaluated and relevant information provide in attachment to this letter.

1CAN011003 Page 2 of 2 This letter contains no new commitments.

If you have any questions or require additional information, please contact me.

Sincerely, DBB/rwc

Attachment:

Response to Requests for Additional Information cc: Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Kaly Kalyanam MS O-8 B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Attachment to 1CAN011003 Response to Requests for Additional Information

Attachment to 1CAN011003 Page 1 of 16 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION REQUEST FOR ALTERNATIVE ANO1-ISI-014 TO IMPLEMENT RISK-INFORMED INSERVICE INSPECTION (RI ISI)

PROGRAM BASED ON CODE CASE N-716

1. Appendix 1 of the submittal states that a gap analysis was held in June 2007.

The submittal also states that an industry peer review of the ANO-1 Probabilistic Safety Assessment (PSA) was conducted in 2002.

A. Identify the guidance document(s) used for the gap analysis and the peer review (e.g., NEI 00-01).

B. Update Table 2 and the unresolved facts and observations Facts and Observations (F&Os) in Appendix 1 to include a description of the supporting requirement for each corresponding F&O number.

C. Identify the results of the 2002 peer review and explain how unresolved F&Os affect this relief request.

In a conference call with the NRC on November 10, 2009, Entergy explained that the Arkansas Nuclear One, Unit 1 (ANO-1) Probabilistic Safety Assessment (PSA) model had been updated and a peer review of that model occurred in August of 2009. The NRC agreed that, in light of this information, no response to the above question is necessary..

2. The submittal states that the ANO-1 internal flooding analysis (IFA) was significantly upgraded to meet Regulatory Guide (RG) 1.200, Revision 1, in 2008, and a gap analysis was conducted in June 2007. In accordance with the guidance endorsed by RG 1.200, a focused-scope peer review should have been performed for a model upgrade. Please summarize the results of the focused-scope peer review conducted on the ANO-1 internal flooding Probabilistic Risk Analyses (PSA) after the 2008 internal flooding upgrade to the American Society of Mechanical Engineers (ASME) PSA Standard RA-Sb-2005 or latest standard and how the peer review findings have been addressed for this application.

In August of 2009, an industry peer review of the ANO-1 PSA model and supporting documentation was performed. The results of the peer review were favorable; however, several issues were identified that require further enhancement. The results of the peer review relating to the IFA have been evaluated and the supporting requirements (SRs) that were identified as Not Met have been listed in the table below. Each of the SRs listed has been evaluated in relation to its applicability to the Code Case N-716 and dispositioned accordingly. In most instances, the SR was addressed in an update to the IFA for ANO-1.

Attachment to 1CAN011003 Page 2 of 16 TABLE 1 SUPPORTING REQUIREMENTS NOT MET AND DISPOSTION Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met IFPP-B3 Document sources of Cat 1-3 is Sources of model Documenting and model uncertainty and related NOT Met uncertainty and understanding sources of assumptions (as identified in related model uncertainty is critical in QU-E1 and QU-E2) assumptions were ultimately ensuring that associated with the internal not documented application results are flood plant partitioning. for the IFA. understood and utilized in a manner that is defensible in relation to the application.

This SR, although important for developing a thorough understanding of the results, does not affect the ANO-1 results for the following reasons and will be included, as necessary, in future revisions to the IFA:

1) The sizes of flood sources were first determined assuming guillotine breaks of lines or the catastrophic rupture of valves, tanks, gaskets, fittings, and heat exchangers. This approach is consistent with the NRC Standard Review Plan (SRP), Sections 3.6.1 and 3.6.2, for high-energy lines and components but is more conservative than the NRC SRP for the medium energy lines that are of principal concern in this analysis.

2) No flood area was screened out on the basis of operator action PRA-A1-01-002 (ANO-1 IFA) R2 Sec 3.4

Attachment to 1CAN011003 Page 3 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met

3) Flow rates were taken as a maximum flow capacity for the propagation path.

4) A plant scram was assumed for each flood scenario.

5) In general floods were allowed to persist for long periods of time (~ 12 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) unless the flood source is finite.

6) Model uncertainty is not incorporated into this revision of the analysis.

Uncertainty analysis does not affect the results.

IFSO-B1 Document the Cat 1-3 is Although flood This particular comment does internal flood sources in a NOT Met sources were not affect the use of this manner that facilitates PSA documented and analysis in support of applications, upgrades, and discussed within applications because it does peer review. Section 4.2 of not alter or affect the results.

PRA-A1-01-002, they are not The documentation of the amenable to PSA analysis for applications, applications and upgrades and peer review is upgrades. critical for ensuring that the Supplemental integrity of the analysis is Excel preserved and understood.

spreadsheets were obtained and This SR was not met primarily flood sources because the supporting were listed by documents associated with flood zone, but it the flood frequency was confusing as calculations, maximum flood to why different height, flow rates, and floor lengths were used areas were not included within for general and the context of the analysis.

major flood These documents are scenarios. available and have now been included in the revision to the IFA.

Attachment to 1CAN011003 Page 4 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met Also, there was a The specific concern relating lack of clarifying to the different length of piping information as to was used for the flood versus why certain pipe spray or the general versus lengths that were major flood are addressed as considered for follows:

flood scenarios were excluded Different pipe lengths apply to from spray various scenarios as smaller scenarios. diameter piping capable of giving rise to a spray rupture

(< 100 gpm) might not be capable of giving rise to a flood rupture, etc.

Scenarios, in which spray damage (as opposed to submergence damage) arises, result from the rupture of those lengths of piping in proximity to a vulnerable target and within direct line of sight. Many pipe lengths considered in flooding scenarios are therefore not being considered in spray damage scenarios.

IFSO-B3 Document sources Cat 1-3 is Sources of model See IFPP-B3 above.

of model uncertainty and NOT Met uncertainty and related assumptions (as related identified in QU-E1 and QU- assumptions were E2) associated with the not documented internal flood sources. for the IFA.

Attachment to 1CAN011003 Page 5 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met IFSN-A6 For the SSCs Cat 1-2 is See example for zone identified in IFSN-A5, identify NOT Met AB317-5.

the susceptibility of each SSC in a flood area to flood- Cat 3 is While water might accumulate induced failure mechanisms. Not in flood zone AB335-53, Include failure by Assessed Upon review of should flooding be allowed to submergence, spray, jet the internal flood persist, water will force open impingement, pipe whip, document (PRA- the door into flood zone humidity, condensation, A1-01-002) and AB335-20 before temperature concerns, and other submergence damage to the any other identified failure supplemental High Pressure Injection (HPI) modes in the identification files, it was not motor operated valves (MOVs) process. explicitly clear as can occur. Water entering to what SSCs flood zone AB317-5 will flow For the SSCs identified in were susceptible into the other flood zones on IFSN-A5, identify the to a particular the 317 elevation of the susceptibility of each SSC in a flood damage Auxiliary Building, but no flood area to flood-induced category, e.g., submergence damage to failure mechanisms. Include submergence or safety-related systems will failure by submergence and spray. occur on the 317 elevation or spray in the identification in the valve pit. Spray damage process. For the purposes in the filter room would also be of meeting the of no great consequence.

EITHER RG 1.200 qualification, the The current analysis does (a) ASSESS qualitatively the scope of the IFA indeed address the SR as impact of flood-induced did not explicitly required. However, in order to mechanisms that are not include any better address this issue, the formally addressed (e.g., qualitative IFAs has been revised to using the mechanisms assessment of include a more comprehensive listed under Capability the impacts from listing of components Category III of this the mechanisms contained in the zone and the requirement), by using listed in assumed failures for each conservative assumptions; Capability zone. This information has Category III been incorporated as part of OR the IFA for ANO-1.

(b) NOTE that these mechanisms are not included in the scope of the evaluation.

Attachment to 1CAN011003 Page 6 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met IFSN-B1 Document the Cat 1-3 is Documentation of This particular comment is a internal flood scenarios in a NOT Met flood scenarios documentation issue and does manner that facilitates PRA for the Auxiliary not affect the application of applications, upgrades, and Building, Turbine this analysis in support of the peer review. Building, and RI ISI submittal because it Intake Structure does not alter or affect the are explained in results.

sufficient detail and organized by The documentation of the subsection under analysis for applications, Section 4.2 of upgrades and peer review is PRA-A1-01-002. critical for ensuring that the However, the integrity of the analysis is calculational preserved and understood.

details of the reported water This SR relates to the SR heights and flow IFSO-B1 in that the supporting rates reported for documents associated with the the analyzed flow rates and floor areas were scenarios were not included within the context omitted from the of the analysis. These documentation. documents are available and have been included in the revision to the IFA.

IFSN-B3 Document sources of Cat 1-3 is Sources of model See IFPP-B3 above.

model uncertainty and related NOT Met uncertainty and assumptions (as identified in related QU-E1 and QU-E2) associated assumptions with the internal flood were not scenarios. documented for the IFA.

Attachment to 1CAN011003 Page 7 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met IFEV-A3 Group or subsume Cat 1-2 is Reviewed PRA- The ANO-1 IFA has been the flood initiating scenarios NOT Met A1-01-002 Rev 2: revised to calculate the core with an existing plant initiating Unevaluated damage frequency (CDF) and event group, if the impact of Cat 3 is scenarios are large early release frequency the flood (i.e., plant response NOT Met grouped into (LERF) for the scenarios that and mitigating system similar but have been identified. The new capability) is the same as a analyzed internal revision to the analysis does plant initiating event group events, however, not screen or subsume any already considered in the PSA the impact of the scenarios or zones. The issue in accordance with the flood may not be relating to this SR has been applicable requirements of 'the same as the addressed in the revision to 2-2.1. plant initiating the IFA.

event group Do not group and do not already subsume flood-initiating considered in the scenarios with other plant PSA'. For initiating event groups. example, see Section 4.2.1.30.

Because the process of subsuming was used, Capability Category (CC) III was not met.

IFEV-A6 In determining the Cat 1 is PRA-A1-01-002 A review of plant specific data flood initiating vent frequencies MET Rev 2: Page has been performed and the for flood scenario groups, use 25/375 states: data does not reflect a one of the following: Cat 2-3 is 'The pipe and necessary change to the NOT Met equipment generic data presented in the (a) generic operating rupture and leak analysis (Also see response to experience frequencies used RAI #5). The results of this are presented in review have been included in (b) pipe, component, and tank Tables 3.2.1.1 the revised internal events (IE) rupture failure rates from and 3.2.1.2; analysis.

generic data sources failure rates for control loops and The following practices are (c) A combination of (a) or (b) incorporated within the next human error above with engineering revision to determine IE probabilities are judgment frequencies.

those developed Gather plant-specific elsewhere in the information on plant design, PSA. These data a) The use of generic and represent generic plant-specific operating operating practices, and experience conditions that may impact data for the US

Attachment to 1CAN011003 Page 8 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met flood likelihood (i.e., material nuclear industry (b) pipe, component, and tank condition of fluid systems, that might rupture failure rates from experience with water appropriately be generic data hammer, and maintenance- modified, or induced floods). In updated, on the (c) engineering judgment for determining the flood-initiating basis of plant consideration of the plant event frequencies for flood practice or specific information scenario groups, use a experience. In collected combination of the following: this regard, we would note that (a) generic and plant-specific ANO-1 has operating experience implemented both continuous (b) pipe, component, and tank flow and rupture failure rates from continuous generic data sources and chlorination to plant-specific experience address microbiological (c) engineering judgment for (bacterial) and consideration of the plant-macro-biological specific information (clams) fouling collected and corrosion.

Accordingly, we would see no reason to doubt the applicability of the generic pipe and equipment rupture data to ANO-1. 'There is no indication of use of plant specific operating experience or initiator information in the determination of IE frequencies.

Attachment to 1CAN011003 Page 9 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met IFEV-A7 Include consideration Cat 1-2 is The EPRI failure While the need to assess of human-induced floods NOT Met database in human induced floods in during maintenance through TR-1013141 relation to the application of application of generic data. Cat 3 is excluded certain the generic data is important Evaluate plant-specific Not events in the for ensuring that the flood maintenance activities for Assessed calculation of frequency is inclusive, the potential human-induced pipe failure inclusion of these human floods using human reliability frequencies that induced floods has no bearing analysis techniques. Note: appear to be on the susceptible damage this would require related to mechanisms associated with consideration of errors of maintenance ISI examinations. EPRI commission. Subsection 2-2.5 activities. See TR-1013141, states, All piping does not at this time provide Table C-2, e.g., system pressure boundary specific requirements related Crystal River 3 failures have been included to errors of commission event. including failures in pipe base metal, welds, and other metallic pressure boundary components such as valve bodies, heat exchangers, and fittings. Human induced causes of flooding that do not involve piping system pressure boundary failure such as overfilling tanks and inappropriate valve operations that release fluid from the system are not included. Such events are expected to be included in the human reliability analysis for the internal flood PRA as they are design and procedure specific and should not be buried in the component failure data.

Also, a review of the plant-specific experience for ANO-1 identified two instances of maintenance related activities affecting a potential flood. In each of the cases identified, the maintenance related failures were associated with improper valve closure during or after the

Attachment to 1CAN011003 Page 10 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met maintenance activity. There are no instances of maintenance related activities that would affect the susceptible damage mechanisms associated with ISI examinations.

This SR will be addressed, as necessary, in a future update to the IFA. This SR has no affect on the RI-ISI submittal for ANO-1 since the generic data includes the piping system pressure boundary failures including those that are human induced. Also, IFEV-A7 is not considered required per EPRI 1018427 Table 2-2 for technical adequacy of PRAs used to develop RI-ISI programs.

Attachment to 1CAN011003 Page 11 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met IFEV-A8 Screen out flood Cat 1-3 is In reviewing This issue has been scenario groups if NOT Met PRA-A1-01-002, addressed in a revision to the it was noted for IFA via revision of the (a) the quantitative screening several scenario scenarios and quantification of criteria in IFSN-A10, as frequencies that the scenarios in which a flood applied to the flood they were frequency was determined.

scenario groups, are met, 'screened' on a Therefore, this SR has been OR strict comparison resolved.

with the IE (b) the internal flood event initiating affects only components in frequency, which a single system, AND it does not can be shown that the compare with this product of the frequency of particular SR. A the flood and the few examples probability of SSC failure may be found in given the flood is two Sections orders of magnitude lower 4.2.1.50, than the product of the 4.2.1.52, and non-flooding frequency for 4.2.1.36.

the corresponding initiating event in the PSA, AND the random (non flood-induced) failure probability of the same SSCs that are assumed failed by the flood. If the flood impacts multiple systems, do not screen on this basis.

IFEV-B3 Document sources of Cat 1-3 is Sources of model See IFPP-B3 above.

model uncertainty and related NOT Met uncertainty and assumptions (as identified in related QU-E1 and QU-E2) associated assumptions with the internal flood-induced were not initiating events. documented for the IFA.

Attachment to 1CAN011003 Page 12 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met IFQU-A5 If additional human Cat 1-3 is Although it was Although the scenario failure events are required to NOT Met noted in Section description indicates that support quantification of flood 3.3 of PRA-A1- timely operator action may scenarios, perform any human 01-002 that a mitigate the consequences of reliability analysis in review of the the flood, the quantification of accordance with the applicable human actions the scenarios took no credit for requirements described in credited was operator action, except where 2-2.5. done to ensure expressly stated, and assumed that the internal that the continued flooding flooding would would propagate, failing not affect the affected equipment in the actions, several affected zones. Equipment operator actions failures were not dismissed not already due to human reliability to credited in the perform mitigating actions but PRA were rather other parameters in the discussed in scenario such as water Section 4.2 that volume, lack of submergence were assumed to and type of equipment under occur that would consideration.

isolate the source of flooding. For the scenarios crediting operator action for flood mitigation, the process for calculating the human error probability was performed in accordance with applicable requirements.

Attachment to 1CAN011003 Page 13 of 16 Whether Description of ASME SR Met or Disposition Issue If Not Met Not Met IFQU-A10 For each flood Cat 1-3 is Similar to the The ANO-1 PSA model links scenario, review the LERF NOT Met CDF model, the the level 1 and level 2 results analysis to confirm applicability quantification of into a single fault tree file. The of the LERF sequences. If the IF model is top gates for level 1 are CORE appropriate LERF sequences done by DAMAGE, and the top gate for do not exist, modify the LERF application of LERF is level 2. The flag files analysis as necessary to specific flag files are set during the account for any unique flood- in the base quantification for level 1 and induced scenarios or model. This is level 2 results. Since the phenomena in accordance explained in LERF portion of the tree is with the applicable Appendix A of the quantified in a similar manner requirements described in IFA. to that for Level 1, the flag files 2-2.8. used in the quantification are There was no the same, and the LERF objective sequences are not altered by evidence that the the flood analysis (only the LERF analysis flood failures affecting the was reviewed for probability of a level 1 impact by IF. endstate and the flood failures affecting the LERF mitigation components are affected in the flood analysis), the quantification of the Level 2 results are valid. The results of the LERF analysis when related to the corresponding CDF provide additional validation of the relative nature of the results.

IFQU-B3 DOCUMENT Cat 1-3 is Sources of model Same comment as above for sources of model uncertainty NOT Met uncertainty and IFPP-B3.

and related assumptions (as related identified in QU-E1 and QU- assumptions E2) associated with the were not internal flood accident documented for sequences and quantification. the IFA.

Attachment to 1CAN011003 Page 14 of 16

3. The supporting requirement IF-C3 in ASME PRA Standard RA-Sb-2005 identifies the failure mechanisms that shall be evaluated to determine the susceptibility of each safety-related structure, system, and component (SSC) in a flood area to flood-induced failures. Capability Category III identifies failure by submergence and spray as requiring detailed analysis and includes jet impingement, pipe whip, and humidity, condensation, and temperature concerns. As endorsed in RG 1.200, Capability Category II would require all SSC failures induced by a pipe break to be considered. Please demonstrate that all SSC failures that are induced by a pipe break are adequately addressed in your analysis.

The ANO-1 IFA risk model includes a discussion of submergence and propagation paths. Unless specifically evaluated, the PSA modeled components within a source room are assumed failed via spray or other impacts such as those noted in Category III for SR IF-C3. This method is conservative when determining whether a flood could exceed a core damage impact of 1E-06 per Code Case N-716.

The susceptibility to flood and spray damage was explicitly considered in the IFA. Also considered was the impact of jet impingement, pipe whip, temperature and humidity in high energy line breaks. Specifically, jet impingement is considered as a result of main steam line breaks in the piping areas and primary makeup system ruptures in the lower north piping penetration area; pipe whip is considered as a result of primary makeup system ruptures in the upper north piping penetration area; temperature and humidity are considered as a result of Emergency Feedwater (EFW) and primary makeup system ruptures in the upper north piping penetration area, primary makeup system ruptures in the lower north piping penetration area, Main Feedwater (MFW) and EFW ruptures in the upper south piping penetrations area and main steam system rupture in the penthouse, lower south piping penetrations area and chiller rooms.

4. Supporting requirements, IF-C6 and IF-C8, permits screening out of flood areas based on, in part, the success of human actions to isolate and terminate the flood. The endorsed Risk-Informed Inservice Inspection (RI-ISI) methods require determination of the flood scenario with and without human intervention, which corresponds to the Capability Category III (i.e., scenarios are not screened out based on human actions). Therefore, a Category III analysis is consistent with the expected RI-ISI analyses. To provide confidence that scenarios that might exceed the quantitative core damage frequency (CDF) and large early release frequency (LERF) guidelines are identified, please describe how credit is given to human actions and identify if the current application analyses meet Capability Category III for these supporting requirements.

Supporting Requirements IF-C6/C-8:

Capability Category II:

Use potential human mitigative actions as additional criteria for screening out flood areas/flood sources if all the following can be shown:

(a) Flood indication is available in the control room.

(b) The flood sources in the area can be isolated.

(c) The mitigative action can be performed with high reliability for the worst flooding initiator.

Attachment to 1CAN011003 Page 15 of 16 High reliability is established by demonstrating, for example, that the actions are procedurally directed, that adequate time is available for response, that the area is accessible, and that there is sufficient manpower available to perform the actions.

Capability Category III:

Do not screen out flood areas/flood source based on reliance on operator action to prevent challenges to normal plant operations.

The ANO-1 IFA meets Capability Category III of the ASME RA-Sb-2005 for these SRs.

In August of 2009, a PSA peer review of the ANO-1 IE PRA was reviewed against the capability categories defined in ASME/ANS RA-Sa-2009. The IFA was reviewed as part of that effort. The ANO-1 IFA was given a Met for Capability Category III of SRs HLN-IFSN-A14 and A16 (Same SRs as IF C-6 and IF C-8 of ASME/ANS RA-Sa-2008).

In the scenario quantifications, operator actions are credited for plant shutdown and sequence mitigation in order to accurately reflect the risk of the flooding event.

However, the IFA does not screen out flooding areas or flooding sources that prevent the challenges to normal plant operations by crediting operator actions. Although most flooding scenarios would result in alarms due to high drain tank or sump levels, no flooding areas or sources were screened from quantification by crediting such alarms to initiate operator actions.

5. Supporting requirement IF-D5a addresses the development of flood initiating (pipe rupture) frequencies for use during the scenario development. RI-ISI is premised on inspecting locations with the highest risk, driven mostly by failure frequency. The plant-specific information collected and used should include experience related to degradation mechanisms that could indicate increased likelihood of pipe failure at particular locations. Please describe how plant-specific operating experience was used to identify experience related to degradation mechanisms and how this experience was incorporated into the development of pipe failure frequencies.

In August of 2009, a PSA Peer Review of the ANO-1 IE PSA was reviewed against the capability categories defined in ASME/ANS RA-Sa-2009. The IFA was reviewed as part of that effort. The ANO-1 IFA was given a Met for Capability Category I of SRs HLN-IFEV-A6 (Same SRs as IF D-5a of ASME/ANS RA-Sa-2008).

A Findings and Observation (F&O) was generated from the peer review team relating to meeting the capability Category II/III for this SR. This F&O states, There is no indication of use of plant specific operating experience or initiator information in the determination of IE frequencies. It is necessary that plant specific operating experience and/or plant specific information be incorporated with the generic failure rate information to achieve greater that Cat I for this SR.

Although a review of the plant specific operational experience was performed as part of the IFA, there was no documentation or references for the peer review team to review in order to substantiate that the criteria for CC II/III was indeed met.

Attachment to 1CAN011003 Page 16 of 16 In order to better address this issue, a review of the ANO-1 and Common (to both ANO-1 and ANO-2) Condition Reports (CRs) were searched for information relating to flooding and excessive leakage that might be integrated into the generic data. The results of this search have been incorporated into a revision to the IFA.

One CR was identified with a failure that would potentially correlate to the initiating frequency for a flood. This CR is associated with a failed Service Water (SW) pipe to a Decay Heat (DH) room cooler during an outage due to loss of primary means of SW pressure control. The root cause for this failure concludes that the failure was due to the inappropriate installation of a cantilevered drain valve into a cleanout port of the cooling coils. This header drain connection mounted on the vendor intended cleanout plug location is unique in the as-built design of the DH room Marlo cooling coils. This inappropriate installation combined with the ANO-1 outage configuration created a condition leading to failure. Since this failure is attributed to outage conditions, these unique circumstances prohibit the usefulness of this data for updating the generic flood frequency used in the at-power IFA.

This review of the plant operating experience provides the basis for conclusion that no plant specific operational experience was noted that would indicate there are piping locations which should use values above generic frequencies. The analyses use generic pipe break frequencies, and the review failed to note any cases where ANO-1 would be an outlier nor would it require the use of more conservative frequencies. No plant specific update to the generic frequencies used in the IFA for ANO-1 is necessary and this information, and as stated above, has been included in a revision of the IFA.

Therefore, the F&O has been addressed in the IFA revision used in this application

6. It is not clear if new examination locations were identified. If so, using an upper-bound estimate for new locations would overestimate the risk decrease and therefore be not conservative. Please demonstrate that this non-conservative approach, if corrected in the evaluation of your proposed RI-ISI program, would not cause the delta risk guidelines to be exceeded.

A bounding sensitivity case was evaluated where risk impact is estimated using upper bound values for Conditional Core Damage Probability (CCDP) and Conditional Large Early Release Probability (CLERP) in those cases that result in a risk increase and lower bound values for CCDP and CLERP in those cases that result in a risk decrease.

The delta risk impact guidelines are not exceeded.