RS-17-102, Response to Request for Supplemental Information Regarding Fourth Inservice Inspection Interval Relief Request 14R-01

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Response to Request for Supplemental Information Regarding Fourth Inservice Inspection Interval Relief Request 14R-01
ML17201Q396
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 07/20/2017
From: Simpson P
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF9758, CACMF9759, RS-17-102
Download: ML17201Q396 (17)
v • d • e


Text

4300 Winf1 eid Road Wa1-re1wii le. !L 60555 Exelon Generation 630 657 2000 Offi ce RS-17-102 10 CFR 50.55a July 20, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 N RC Docket Nos. 50-373 and 50-37 4

Subject:

Response to Request for Supplemental Information Regarding LaSalle County Station Fourth lnservice Inspection Interval Relief Request 14R-01

References:

1) Letter from D. M. Gullett (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "LaSalle County Station, Units 1 and 2, Fourth 10-Year Interval Inservice Inspection Program Relief Requests, dated 11 May 30, 2017 (ADAMS Accession No. ML17150A449}
2) Letter from B. Vaidya (U.S. Nuclear Regulatory Commission) to B. C. Hanson (Exelon Generation Company, LLC), "LaSalle County Station, Units 1 and 2, Supplemental Information Needed for Acceptance Review of Requested Licensing Action Re: Relief Request Re: Fourth 10-Year lnservice Inspection Interval Program Relief Request No. 14R-01 (CAC Nos. MF9758 and MF9759}, dated July 10, 2017 (ADAMS Accession No. ML17181A197) 11 In a letter dated May 30, 2017 (Reference 1), Exelon Generation Company, LLC (EGC) requested approval of a request associated with the fourth lnservice Inspection (ISi) interval for the LaSalle County Station (LSCS), Units 1 and 2. The fourth interval of the LSCS ISi Program is currently scheduled to begin on October 1, 2017, and end on September 30, 2027, and will comply with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda. Relief Request 14R-01 of Reference 1 requested approval of alternative risk-informed ISi program and examination criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 pressure retaining piping welds in accordance with ASME Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1.

11 In Reference 2, the U.S. Nuclear Regulatory Commission (NRC) requested supplemental information for the NRC to complete the acceptance review. The requested information is provided in the Attachment of this letter.

There are no regulatory commitments contained within this letter.

July 20, 2017 U.S. Nuclear Regulatory Commission Page 2 Should you have any questions concerning this letter, please contact Ms. Lisa A. Simpson at (630) 657-2815.

Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC

Attachment:

Response to Request for Supplemental Information cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, LaSalle County Station Illinois Emergency Management Agency- Division of Nuclear Safety

ATTACHMENT Response to Request for Supplemental Information LS-LAR-08, Rev. 0 PAA Upgrade Review in Support of Risk Informed-ISi Relief Request No. 14R-01 Request for Supplemental Information 14 pages follow

Exelon Generation RISK MANAGEMENT TEAM RM DOCUMEN TATION NO. REV: 0 PAGE NO. 1 STATION: LaSalle County Station (LSCS)

UNIT(s) AFFECTED: 1 & 2 TITLE: PRA Upgrade Review in Support of Risk Informed-ISi Relief Request No.14R-01 Request for Supplemental Information

SUMMARY

LSCS is pursuing an lnservice Inspection {ISi) Program relief request for continuation of the LaSalle Risk-Inform ed ISi (Rl-ISJ, or RISI) program for another 10 year interval. The NRC requested Supplement al Information.

The purpose of this document is to determine if PRA model changes made post-peer review are considered PRA Maintenanc e or PRA Upgrades in accordance with ASME/ANS 11Standard 1

for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications ," [1], as requested in the NRC Request for Supplemental Information.

This is a Category I Risk Managemen t Document in accordance with ER-AA-600-1012 Risk Managemen t Documentat ion [2], which requires independen t review and approval.

[ ] Review required after periodic update

[ X ] Internal RM Documenta tion [ ] External RM Documenta tion Electronic Calculation Data Files: NIA Method of Review: [ X ] Detailed [ ] Alternate [ ] Review of External Document This RM documenta tion supersedes : NIA Prepared by: Feli~e Gonzalez I

~~ .:;;;;:::: I 7117/17 cfe. t Print Date Reviewed by: John E. Steinmetz Print I

~j Sign I 7/17117 Date Reviewed by: Gran!_Tea garden I ~__.._:?I 7~-c P--- I 7/17/17 Print Sign ' Date Reviewed by:

{Independent Don Vanover I ZJc,~ E.Y~ I 7 /1'if /, 7 Print Sign Date Review)

Approved by: Eugene Kelly Print ' --~17/fPi r

  • S1gn I 7/t(n LS-LAR-08 REV. 0

LaSalle Support Application RI-ISi REQUEST FOR SUPPLEMENTAL INFORMATION NEEDED REGARDING RELIEF REQUEST NO. 14R-01 FOR LASALLE COUNTY STATION (LSCS), UNITS 1AND2 DOCKET NOS. 50-373 AND 50-374 (CAC NOS. MF9758 AND MF9759)

This support application provides a response to the NRC request for supplemental information below. This NRC request was transmitted to Exelon Generation Company (EGC) by letter dated July 10, 2017 [6].

Background

After reviewing the Relief Request (RR) No. 14R-01, it is not clear if there have been any probabilistic risk assessment (PRA) upgrades, since 2008. The licensee has properly used their latest PRA model, but the U.S. Nuclear Regulatory Commission (NRC) staff was not able to verify if the latest PRA model has been peer reviewed based on the information provided in the RR. It appears that no peer review has been conducted since 2008. Additionally, the 2008 peer review used only Rev. O of RG 1.200, rather than the current Rev. 2. The RR states that the EGC, the licensee performed a self-assessment against Revision 1 of RG 1.200, but to date there does not appear to be any documentation verifying that a self-assessment against Rev. 2 of RG 1.200 has been completed.

Supplemental Information Needed from Licensee The licensee is using a risk-informed approach to support the inservice inspection interval program. Provide the documentation to demonstrate the technical adequacy of their PRA against Revision 2 of RG 1.200 as described below:

  • In Enclosure LS-LAR-007 to Attachment 1 of the licensee submission (page 11 of 25), it is stated that "the most recent update of the LaSalle PRA model (designated the LS2014A model) was completed in November 2015 as a regularly scheduled update to the previous LS2011 A model."
  • Provide details of all changes to the PRA model since the version reviewed by the 2008 peer review (you may group changes such as "data refinements," but be specific as to other changes, e.g., changes in model structure or methods), justifying whether or not, the change constitutes PRA upgrade as defined in the ASME/ANS [American Society of 2 LS-LAR-08 REV. 0

LaSalle Support Application RI-ISi Mechanical Engineers/American Nuclear Society] PRA standard - "new should be interpreted as new to the subject PRA even though the methodology in question has been applied in other PRAs" [Section 1-A.1 of ASME/ANS RA-Sa-2009].

  • If any changes constitute an upgrade, perform a focused scope peer review on the affected technical elements (including affected high level requirements as well as all supporting requirements within the affected high level requirements) and provide the results of the review, complete with the Facts and Observations/Findings and dispositions addressing any effect upon the application.

Response

There have been two updates to the internal events PRA model since the performance of the 2008 peer review. The updates were the 2011 and 2014 PRA.

The 2008 PRA peer review was performed using the NEI 05-04 process and the ASME PRA Standard ASME RA-Sc-2007 version along with RG 1.200, Revision 1. A self-assessment was performed during both the 2011 and 2014 PRA updates against RG 1.200, Revision 2. The results of that assessment are summarized in this response.

There has not been an internal events PRA peer review since the 2008 peer review; however, an independent assessment team recently reviewed the 2008 peer review findings for closure. In their review, the 2017 LaSalle County Generation Station PRA Independent Assessment (IA) team used the following standards and references:

  • NEI 05-04, Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard, Nuclear Energy Institute, Rev. 2, November 2008. This document defines the review process used in the BWROG industry peer reviews.
  • NEI Appendix X to NEI 05-04, 07-12 and 12-06, Close Out of Facts and Observations (F&Os), Nuclear Energy Institute, Rev. 0, February 2017.
  • Standard for Level1 /Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sa-2009, February 2009. This document defines the review assessment criteria.
  • Latest ASME PRA Standard interpretations from the ASME website.

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  • U.S. Nuclear Regulatory Commission Memo to Stacey L. Rosenberg, Branch Chief, PRA Licensing Branch, Division of Risk Assessment, U.S.

Nuclear Regulatory Commission Staff Expectations for an Industry Facts and Observations Independent Assessment Process, May 1, 2017.

As part of the IA review, the utility provided information to support that the changes made to address each finding were considered maintenance of the PRA model and not upgrades. The IA team reviewed the findings and associated PRA model changes and PRA documentation supporting their closure and determined that the PRA model changes were maintenance activities and not upgrades. The IA team documented that each change submitted for review was maintenance in Table A-1 of the Independent Assessment Report [3].

Review of changes to the PRA model are grouped according to the model updates (i.e.,

2011 and 2014).

2011 PRA Model Changes The following is a summary of all of the changes made to the model in the 2011 PRA Update. The change descriptions are from LS-PSA-13 LaSalle PRA Summary Notebook [4].

- PRA Maintenance (Examples 2 and 3). Using updated plant-specific and new generic data is considered PRA maintenance. This was not the first time Bayesian updating was performed for such data.

  • Incorporated support system initiating event fault trees for service water (SW), reactor building closed cooling water (RBCCW) and turbine building closed cooling water (TBCCW) into the single top logic.

- PRA Maintenance (Examples 6 & 11 ). Logic model enhancement, no new methodology employed compared to the prior model. The support system fault trees existed as separate files at the time of the Peer Review. Their incorporation into the single top model is considered a logic model enhancement which is less significant than changing from one fault tree linking code to another (Example 11 ).

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- PRA Maintenance (Examples 1& 6). Logic model enhancement, no new methodology employed compared to the prior model. Overall, CDF decreased slightly with the re*moval of bus 241 Y and 242Y as initiating events. Loss of bus 241 X, 242X, and 251 are not significant contributors to CDF or LERF.

- PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.

  • Revised common cause failure (CCF) calculations to incorporate the updated individual random basic event probabilities and 2009 CCF parameters from INEEL (NUREG/CR-6268).

- PRA Maintenance (Examples 3 and 26). Using new data, no new methodology employed.

  • Updated maintenance unavailability data based on the most recent LaSalle operating experience.

- PRA Maintenance (Examples 2 and 19). Using new plant-specific data, no new methodology employed.

  • Deleted most of the coincident maintenance terms that had previously been added because they no longer meet the definition of the coincident maintenance as defined in the ASME/ANS PRA standard as "planned and repetitive."

- PRA Maintenance (Examples 6 & 7). Logic model enhancement, no new methodology employed compared to the prior model.

  • Revised the Turbine Building and Auxiliary Building flooding initiator event trees to credit the use of the VR check dampers as a flood mitigation strategy.

- PRA Maintenance (Example 9). Logic model enhancement (corrects an omission), no new methodology employed compared to the prior model.

  • Created a new event tree for isolated Turbine Building and Auxiliary Building floods (TB-FLO.eta).

- PRA Maintenance (Examples 6 & 7). Logic model enhancement, no new methodology employed compared to the prior model. This event tree is similar to existing isolated Reactor Building event tree (RB-FLD). Prior model also included an ET for unisolated Turbine Building to Reactor Building flood. The new ET had a small quantitative impact.

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  • Developed an HEP for the operators closing the VR check dampers in the event of a Turbine Building or Auxiliary Building flood.

- PRA Maintenance (Example 20 & 23). HEP modeling updated, Internal Flooding operator actions added to model in accordance with existing HRA methodology, no new methodology employed.

  • Converted the Human Reliability Analysis (HRA) calculations to the EPRI HRA Calculator software platform. The HRA Calculator was also used to facilitate the HEP dependence analysis.

- PRA Maintenance (Example 11 ). For the 2011 LaSalle FPIE, the updated HRA was entered into the EPRI HRA Calculator. The use of the HRA Calculator is to ensure consistency among the HEPs and to enhance documentation of the HEPs. The methods (e.g., ASEP, THERP, CBDT) used for calculating the LaSalle HEPs was unchanged from the Peer Review. The use of the HRA Calculator tool is consistent with a PRA update. Similar to Example 11 of Appendix 1-A, changing software while utilizing the same methodology is not considered a PRA upgrade. Additionally, an independent review of the HRA was performed by an experienced HRA analyst to confirm that none of the changes made to the HRA after the peer review would qualify as PRA upgrade.

  • Extensive HRA re-assessment based on operating crew interviews using the latest EOPs and support procedures.

- PRA Maintenance (Examples 20 & 22). HEP modeling updated consistent with existing model methods, no new methodology employed.

  • Added a detailed pre-initiator HEP screening evaluation and added additional pre-initiators to the model.

- PRA Maintenance (Example 20). Pre-initiator HEPs were already included in the model (e.g., mis-calibration of the low pressure permissives). Additional pre-initiators were added to the model (18 new pre-initiators bringing total to 86) as a modeling enhancement, no new methodology employed.

  • The level of detail in the modeling of CSCS was increased to include fans, dampers, etc.

- PRA Maintenance (Examples 6 & 7). Logic model enhancement, no new methodology employed compared to the prior model.

  • ATWS event tree sequences where the ATWS was mitigated (i.e., ATWS scenarios with successful reactivity control) were re-evaluated based on comparisons against other Exelon BWRs. Some end states were changed from Class IV to Class 11 where appropriate.

- PRA Maintenance (Example 6). Logic model enhancement, no new 6 LS-LAR-08 REV. 0

LaSalle Support Application RI-ISi methodology employed compared to the prior model. CDF was unchanged, LERF decreased by 1%.

  • The power supply modeling for the Station Air Compressors (SACs) was modified to reflect plant modifications. The control power is supplied from the same bus as the SAC pre-lube pump (i.e., SACO control power from 1338-2, SAC1 control power from 132A-1 and SAC2 control power from 232A-1 ).

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The RHR water hammer scenarios were re-evaluated as part of the 2011 model update. For HPCS, LPCS and RHR C, if the system is operating in injection mode or on min flow, a drain down scenario is only possible if the pump discharge check valve (CV) fails to close when the pump stops.

The failure of the CV to close was added to the logic.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The probability of a water hammer event causing a rupture was revised downward for consistency with industry experience and other Exelon BWR PRA models.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model. *

  • A change *was made to the small water LOCA Event Tree (ET) to reflect that for some small LOCAs, RCIC may be a viable long term injection source. A node was added to the ET to question whether the small LOCA was "large" or 'small." A 50/50 probability was assigned to each break size. If the small LOCA was smaW and depressurization fails, the event 11 tree then questions the long term availability of RCIC.

- PRA Maintenance (Examples 6 & 7). Logic model enhancement, no new methodology employed compared to the prior model. Small water LOCA sequences make up <1% of total CDF and LERF.

  • The diesel generator recovery factors DGRECOV-4HR and DGRECOV-

?HR were changed to 1.0 due to peer review comments and consistency efforts.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The offsite power recovery factors were corrected in the model to match the values documented in LS-PSA-001 LaSalle Initiating Events Notebook, Appendix E (5) and as given in NUREG/CR-6890.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

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  • Basic Event 1OPPH-RX-ENVIF-- probability was corrected to match the documentation.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The value of 2CNHU-PREINIT was changed from 5E-03 to 2.3E-3 to be consistent with current industry information in EPRI TR101824.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The fault trees related to recovery of AC power in the intermediate timeframe were enhanced by modifying the gate names and descriptors to clarify that the intermediate timeframe is 2 to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. Also, the conditional probability that the battery fails was changed and was or'ed with the operator action to load shed.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Division 3 DC fault tree was reviewed and revised to reflect that failure of the charger alone will fail Division 3 for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time, or if a LOOP/DLOOP event has occurred that the battery needs to be available to support Division 3 EOG starting and then the charger can be re-powered by the EOG and then can fulfill the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Gate 2DC1 OE-PWR was removed from gate MDRFP. This was done based on a review of LOA-DC-201, which led to the conclusion that Bus 2A does not fail the motor driven feedwater pump.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The basic event 2ACHB-2425---K-- (i.e., FTC) under gate 1CB2425-CO-HW was changed to a spurious opening event 2ACHB-2425---U--

because breaker 2425 is normally closed in the normal aux power lineup.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • LOA-AP-201 was reviewed and it was determined that specific instructions are included for cross-tying bus 142Y to 242Y when bus 142Y is being powered from the 1A DG. Therefore, the 1A DG should be credited as a power source if both the Unit 1 UAT and SAT are unavailable.

Functionally, to credit the 1A DG in the model, the OR gate t42Y-PWR-SOURCES was changed to an AND gate. Additionally, gate 142Y-SAT-UAT under 142Y-242Y-PWR was deleted because it was also under 1AP06E-PWR and was failing DG1 A on loss of power improperly.

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- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The logic for gate UAT141-142X (UAT POWER TO 142X) was not failed for dual unit scrams. Gate DUAL-SCRAM, "DUAL UNIT SCRAMS," was added under gate UAT141-142X "UAT POWER TO 142X".

- PAA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The basic events 2SYPPBOCINRB-R-- and 2SYPPBOCSTMTLR-- were revised to reflect the current values of the BOC in the initiating events notebook.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The success and failure nodes of OPISOL added up to 1.05 instead of 1.00. The inputs to gate NODES-ISOL where reviewed and corrected.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • A new basic event was created for the air intake filter for the 2B DG room ventilation. This is included in the logic consistent with the other CSCS ventilation modeling. Note that this is a dependency between Division 2 and Division 3 that was not previously recognized or modeled. This event was also added to the failure modes of the 2B DG.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Basic event ids 2RHMV-65A----U-- and 2RHMV-65B----U-- were changed to 2RHAV-65A----U-- and 2RHAV-65B----U-- to reflect that these valves are AOVs, not MOVs. The descriptions of various other gates were changed to reflect that these are AOVs not MOVs. Gate RR-TAB-DIV was changed to an OR gate and RR-TRB-DIV-SP2 was changed to an AND gate. This was done to correct the logic and make it the same as the A train.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Gate HG-SY-INITIATE was deleted and replaced with SY-INITIATE under gate MDFW-AUTO. This was done because the operator action 2HCOPSTART---H-- is being deleted. This operator action is redundant to and is being replaced by the operator actions for manually initiating ECCS under the SY-INITIATE gate. Additionally, the gate SY-INITIATE-SUC was deleted globally. Similarly, 2HCOPSTART---H-- was replaced by SY-INITIATE under gates HPCS-FAIL-START, DG-2B-INIT and RCIC-START-CKT.

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- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Basic event 2CRAVC11 F001 AU-- was changed to 2CRAMC11 F001 AU--

to reflect that this is a relief valve, not an AOV. Similar change was made for 2CRAVC11 F001 BU--. Additionally, the type code was added to the data spreadsheet and appropriate failure probabilities were generated.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The model was changed to show that 2FC086 and 2FC115 are manual valves and are not automatically closed on a containment isolation signal.

The basic events related to these valves under the gate CN-RW-DRAIN were changed to have XV in the BE name rather than AV since they are not AOVs. Additionally, the failure probabilities were revised to reflect that of manual valves. Additionally, the gate CN-CISIGNALFAIL was deleted under that gate since there is no containment isolation signal for these valves. That gate was replaced with the operator action for closing the containment isolation valves (2CNOPCLOSEPIVH--). Finally, the common cause term was deleted since these are manual valves.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • The gate DEPRESS-COND and the subtree were deleted including 2ADOP-COND---H--. This part of the fault tree was to account for a dependency between depressurizing and manual starting injection systems. This part of the fault tree is unnecessary as the dependency analysis will dictate the dependencies included in the model.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Gate DG1 A-MAI NT was incorrectly an "AND" gate instead of an "OR" gate. It was corrected and is now an OR gate. Additionally, BE 1DGDG-1A-PLANM-- was deleted as it was redundant to the other MUA event and it was assigned a probability of 0.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Basic event 2DGDG-2A-PLANM-- was deleted from gate DG2A-MAINT because it was redundant to 2DGDG-DG2A---M--. Similar change was made for basic event 2DGDG-2B-PLANM-- under gate DG2B-MAINT and BDGDG--0-PLANM-- under gate DGO-MAINT.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Under gate PSW-U1TBCCW, the BE 1WSAVSW030---V-- was revised to reflect that this valve is a manual valve not an AOV. The BE name was 10 LS-LAR-08 REV. 0

LaSalle Support Application RI-ISi changed to 1WSXVSWS030---V-- and the failure probability was changed as well to reflect the failure probability of a manual valve spurious closure.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

  • Addressed several documentation related UREs with no quantitative impact.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

Based on review of the changes during the 2011 update identified above, all changes are found to be maintenance, not PRA upgrades.

2014 PRA Model Changes The 2014A PRA update is a regularly scheduled data update per the Exelon Training and Reference Materials (T&RMs). The following is a summary of the changes made to the model as listed in the 2014 LaSalle PRA Summary Notebook [5]:

- PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.

- PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed. This was not the first time Bayesian updating was performed.

  • The evaluation of containment heat removal and RPV injection success following offsite AC power recovery in the LOOP and DLOOP event trees is added to the model to more explicitly model the recovery/restoration capabilities of the plant.

- PRA Maintenance (Example 10). Similar to Example 10, this modeling change was performed for completeness purposes in response to industry peer review comments.

  • Minor revisions to human error probabilities (HEP). Updated HRA dependency modeling as part of 2014A update 11 LS-LAR-08 REV. 0

LaSalle Support Application RI-ISi

- PRA Maintenance (Example 20). HEP modeling enhancement, no new methodology employed.

  • Updated maintenance unavailability data based on the most recent LaSalle operating experience.

- PRA Maintenance (Examples 2 and 19). Using new plant-specific data, no new methodology employed.

  • Revised common cause failure (CCF) calculations to incorporate the updated individual random basic event probabilities and 2014 CCF parameters from INEEL (NUREG/CR-6268).

- PRA Maintenance (Example 3 and 26). Using new data, no new methodology employed.

  • Addressed several documentation related UREs with no quantitative impact.

- PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

Based on review of the changes during the 2014 update identified above, all changes are found to be maintenance, not PAA upgrades.

2014 PAA Self-Assessment Following the completion of the 2014 PAA update, the PAA self-assessment against the RG 1.200 Revision 2, was updated. The following SRs were identified as "Not Met" for Capability Category 11.

  • Four (4) SRs (i.e., SC-AS, HR-03, DA-C7, and DA-CB) were assessed as meeting Capability Category I only. Per the EPRI guidance for required PAA technical adequacy [7], these SRs only need to meet Capability Category I. Therefore, the technical quality of these SRs is acceptable for this application.
  • Three (3) SRs (i.e., IFSO-A3, IFSN-A7, and IFQU-A3) were assessed as "Not Met". Per the EPRI guidance [7], these SRs are only applicable if the internal flooding study is used to directly the support the development of the RI-ISi program. For LaSalle, the internal flooding study is not used to support RI-ISi. Therefore, the assessment of "Not Met" for these SRs is acceptable.
  • Two (2) SRs (i.e., DA-C6 and DA-C10) related to data were assessed as "Not Met." Per the EPRI guidance [7], DA-C6 should be Met for Capability 12 LS-LAR-08 REV. 0

LaSalle Support Application RI-ISi Category 1-111, and DA-C10 should be Met for Capability Category I. These two SRs are reviewed for this application.

SR DA-C6 specifies estimation of plant-specific demands for standby components on the basis of surveillance tests, maintenance acts, and operational demands. Currently in the PRA, estimates of demands for standby components are based on a mixture of data sources such as plant process computer data, test frequency and associated procedure review (e.g., # cycles I test times the number of tests per year), MSPI basis document data, operator logs, work clearance order database, and system manager estimates. In the PRA self-assessment, these varied approaches were not judged to meet the strict definition of the SR. The plant data sources and developed demand estimates, however, are judged to be reasonable to support the PRA. Pursuing plant demand data per the explicit direction in the SR is not expected to result in significant impacts upon the PRA results. Therefore, DA-C6 is being self-assessed as "Not Met" is not judged to impact this application.

- SR DA-C10 relates to using surveillance test data and reviewing the test procedure to determine whether a test should be credited for every possible failure mode. Currently in the PRA, use of surveillance procedures is only performed a small proportion of the time (e.g.,

standby components) end the procedure are not always reviewed in consideration of every possible failure mode, however, the test estimates developed based on surveillance test data review is judged reasonable to support the PRA. Pursuing more detailed review of surveillance procedures is not expected to result in significant impacts upon the PRA results. Therefore, DA-C1 O is being assessed as "Not Met" is not judged to impact this application.

Summary Conclusion All changes to the PRA model since the version reviewed by the 2008 peer review have been reviewed and assessed with regards to being PRA maintenance or upgrades, as documented above. Each change (or group of changes) was reviewed against examples of PRA Maintenance and PRA upgrades in the 2009 ASME/ANS PRA Standard [1]. Each change was justified as PRA maintenance as defined in the ASME/ANS PRA standard. Additionally, the independent assessment conducted for the purpose of F&O closure also identified all PRA changes associated with F&O resolution involved only PRA maintenance. Therefore, a focused scope peer review on affected technical elements is not required.

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LaSalle Support Application RI-ISi References

[1] ASME/ANS, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME/ANS RA-Sa-2009, March 2009.

[2] ER-AA-600-1012 Risk Management Documentation, Rev. 13.

[3] LSGS Unit 2 PRA Facts and Observations Independent Assessment Report Using NEI 05-04/07- ~ 2/12-06 Appendix X, dated June 2017.

[4] LS-PSA-13 LaSalle PRA Summary Notebook, Rev. 7 dated March 2013.

[5] LS-PSA-13 LaSalle PRA Summary Notebook, Rev. 8 dated November, 2015.

[6] NRC, "LaSalle County Station, Units 1 and 2, Supplemental Information Needed for Acceptance Review of Requested Licensing Action Re: Relief Request Re:

Fourth 10-year lnservice Inspection Interval Program Relief Request No. 14R-01 (CAC NOS. MF9758 and MF9759), July 10, 2017.

[7] EPRI, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs," EPRI TR-1018427, Dece.mber 2008.

14 LS-LAR-08 REV. 0

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