ML18019A091

Response to Request for Additional Information, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (Sscs) for Nuclear Power Plants

ML18019A091
Person / Time
Site:	Limerick
Issue date:	01/19/2018
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML18019A091 (39)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 January 19, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

Response to Request for Additional Information Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants"

References:

1. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants," dated June 28, 2017 (ADAMS Accession No. ML17179A161).

2. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Supplement to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors," dated August 14, 2017 (ADAMS Accession No. ML17226A336).

3. Electronic mail message from V. Sreenivas, U.S. Nuclear Regulatory Commission, to Glenn Stewart, Exelon Generation Company, LLC, "Limerick 50.69 license amendment request application: Request for Information (RAI)," dated December 6, 2017 (ADAMS Accession No. ML17341A250).

By letter dated June 28, 2017 (Reference 1), as supplemented by letter dated August 14, 2017 (Reference 2), Exelon Generation Company, LLC (Exelon) requested an amendment to the Renewed Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (Limerick), Units 1 and 2, respectively. The proposed amendment would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors.

U.S. Nuclear Regulatory Commission Response to Request for Additional Information Application to Adopt 10 CFR 50.69 Docket Nos. 50-352 and 50-353 January 19, 2018 Page 2 The NRC staff reviewed the information provided that supports the proposed amendment and identified the need for additional information in order to complete their evaluation of the amendment request. The request for additional information (RAI) was sent from the NRC to Exelon by electronic mail message on December 6, 2017 (Reference 3). The NRG email requested a response by January 19, 2018. to this letter provides a restatement of the RAI questions followed by our responses. Attachment 2 contains proposed markups of the Limerick Unit 1 and Unit 2 Renewed Facility Operating Licenses.

Exelon has reviewed the information supporting a finding of no significant hazards consideration, and the environmental consideration, that were previously provided to the NRC in Attachment 1 of the Reference 1 letter. Exelon has concluded that the information provided in this response does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92. In addition, Exelon has concluded that the information in this response does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

There are no regulatory commitments in this response.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), Exelon is notifying the Commonwealth of Pennsylvania of this RAI response by transmitting a copy of this letter and its attachments to the designated State Official.

If you have any questions or require additional information, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 191h day of January 2018.

James Barstow Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments:

1. Response to Request for Additional Information Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants"

2. Markup of Proposed Renewed Facility Operating License (RFOL) Pages cc: Regional Administrator - NRG Region I w/ attachments NRG Senior Resident Inspector - Limerick Generating Station NRG Project Manager, NRR - Limerick Generating Station Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection

ATTACHMENT 1 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Response to Request for Additional Information Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants"

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 1 of 30 Docket Nos. 50-352 and 50-353 By letter dated June 28, 2017 (Reference 1), as supplemented by letter dated August 14, 2017 (Reference 2), Exelon Generation Company, LLC (Exelon) requested an amendment to the Renewed Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (Limerick), Units 1 and 2, respectively. The proposed amendment would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors.

The NRC staff reviewed the information provided that supports the proposed amendment and identified the need for additional information in order to complete their evaluation of the amendment request (Reference 3). Below is a restatement of the questions followed by our responses.

RAI 01 - Internal Events and Internal Flooding Probabilistic Risk Assessment (PRA)

Facts and Observations (F&Os)

a. F&O HR-A1 Review of Procedures and Practices for Test and Maintenance Pre-Initiators Open F&O HR-A1-01 appears to indicate that the test and maintenance pre-initiators were not derived from a review of procedures and practices as prescribed by PRA standard supporting requirement (SR) HR-A1. The disposition to the F&O in license amendment request (LAR) Attachment 3.a, Open and Partially Resolved Peer Review Findings, states that the risk-significant pre-initiators are included in the model without explaining how the test and maintenance pre-initiators were identified. Explain how the test and maintenance pre-initiators were derived (e.g., through a review of procedures) and explain how the conclusion that the risk-significant pre-initiators have been included was reached.

Response

The PRA model, as peer reviewed, includes numerous pre-initiators for a number of risk significant systems but these were not derived from a formal review of procedures and practices.

However, as a part of the current routinely scheduled PRA Update, the HRA pre-initiator study is being updated to fully meet Category II. Upon completion of the current model update, and before implementation of 50.69, a focused scope peer review of this upgrade will be performed. Therefore, this finding will not be relevant to the PRA model to be used in the 10 CFR 50.69 application.

SR HR-A1 is for the identification of activities that can lead to misalignments. The process used for the ongoing Limerick model update is to first identify the plant systems that are included in the PRA model. The implicit assumption is that only the PRA systems have a meaningful impact on plant risk such that misalignment in non-modeled systems can be excluded from consideration. For each PRA system, a review was performed to identify components for which a misalignment could cause failure of the associated system/train/function. Screening rules were developed and documented (e.g., components that automatically re-align can be screened, unless they impact redundant divisions or

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 2 of 30 Docket Nos. 50-352 and 50-353 diverse systems). A review of plant operating experience was also performed to determine if there have been instances of latent errors at the plant. As part of this process, plant events representing failure modes not already captured by the system review process would be added as additional pre-initiator events (none were identified for Limerick). For those events that were not screened, screening probabilities were assigned to the events, and the events were included in the model. The model was then quantified and, for those events that were determined to be risk significant based on their importance measures, a detailed evaluation was performed. The non-risk-significant events were retained in the model with their screening values. As part of the detailed evaluation process, a procedure search was performed to identify all procedures that include the component associated with the risk-significant event. The procedures were then reviewed to identify the subsets that move the component out of its desired position as part of test, maintenance, or surveillance activities.

The process to designate a procedure as the quantification basis for the action is to review the subset of procedures that move the component out of its normal position, and to select the procedure with the fewest recovery mechanisms.

b. F&O IF-B3 Water Volume Not Considered Open F&O IF-B3-01 indicates that internal flooding scenarios may have been inappropriately screened out because the full volume of water that could drain from the cited cooling water systems was not considered. The disposition to this F&O in LAR Attachment 3.a, Open and Partially Resolved Peer Review Findings, states that flood scenarios need to be reviewed in the next update to determine if revisions or additional scenarios are needed. However, any changes are expected to have no material impact on the Title 10 of the Code of Federal Regulations (10 CFR) Section 50.69 application. Given the acknowledgement that scenarios may need to be added, it is not clear why these exclusions can have no material impact on the 10 CFR 50.69 application. Justify why the excluded scenarios cannot have a material impact on the 10 CFR 50.69 application, or alternatively, incorporate the additional water volumes cited in the F&O into the internal flooding PRA.

Response

Contrary to the statement of the F&O IF-B3-01, the Turbine Enclosure Cooling Water (TECW), Reactor Enclosure Cooling Water (RECW), Control Enclosure Cooling Water (CECW), and Drywell Chilled Water (DWCW) systems (noted in F&O IF-B3-01) were not screened out of the Internal Flood (IF) PRA model. The Internal Flood documentation identifies screening of buildings and flood areas. The IF analysis does not screen based on flood sources.

A bounding water volume estimate is being calculated in the current scheduled PRA model update and included in the IF documentation for these closed loop systems. Previous screening was performed on a quantitative basis (initiating frequencies and CCDPs) but subsequent PRA updates have included all identified water sources and their scenario frequencies as calculated in the IF analysis. Therefore, this finding will not be relevant to the internal flood model to be applied in the 10 CFR 50.69 application.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 3 of 30 Docket Nos. 50-352 and 50-353

c. F&O SC-SY-B1 Fire Water Makeup Resolved F&O SC-SY-B1-01 stated that a high probability was used for failure of fire water makeup to the vessel to prevent core damage to include the uncertainty as to whether or not the fire protection system can actually prevent core damage after depressurizing the reactor and within four hours after an initiating event. The reported resolution in Attachment 3.b, Resolved Peer Review Findings, was that a detailed HRA [human reliability analysis]

calculation was performed for aligning fire water makeup to the reactor vessel. Performing a detailed HRA calculation provides confidence that the alignment is feasible and that failure to align is properly quantified, but not necessary, as to whether the flow and amount of water is sufficient to prevent core damage. Confirm that sufficient flow and amount of water was determined as part of the HRA calculation, or alternatively, provide justification that the fire protection system can successfully prevent core damage as credited in the PRA model.

Response

The Limerick PRA RHR System Notebook documents the ability of the fire water system to provide 300 gpm of water to the reactor vessel when the pressure of the reactor vessel is 100 psig or less. This confirms that sufficient flow and amount of water is available as assumed for the HRA calculation. This alternate injection source is modeled as a success path in the PRA in late scenarios when the RPV has been depressurized.

d. Closed F&Os DA-C14-01 and QU-A4-01, related to credit for repair, stated that credit for repair had been removed for some listed systems but implied that credit was retained for the instrument air system.

The internal events gap assessment results and resolutions have previously been provided for an inservice inspection relief request dated April 13, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16104A122), as supplemented by request for additional information (RAI) response dated September 19, 2016 (ADAMS Accession No. ML16263A218). In the September 19, 2016, RAI response, the licensee stated that repair is not credited in the current model.

Identify any structures, systems, and components (SSCs) for which repair is credited in the current internal events PRA and justify how the applicable SRs (e.g., SY-A24, DA-C1, LE-C3, and CA-C15) are met.

Response

Following the current PRA Update, no credit for repair will be taken.

The Limerick PRA Data Notebook provides historical information on repair estimates for EDGs and RHRSW / ESW / RHR pumps. However, the basic events modeling these repairs are set to "T" (True) in the Limerick PRA flag file and are compressed out in the quantification process. Therefore, repair of SSCs modeled by these events is not credited in the PRA model.

The Limerick PRA Data Notebook evaluates the recovery of instrument air only for the purpose of recovering the emergency containment vent. This evaluation is based on

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 4 of 30 Docket Nos. 50-352 and 50-353 judgment, consideration of Limerick accident scenario specifics (MAAP runs), and review of industry studies (EPRI study NSAC-161, WASH-1400). The recovery of instrument air has been included in the model based on this recovery evaluation.

Given that there is uncertainty in the recovery steps that would be taken to recover instrument air to support containment venting, this event is being set to True in the flag file during the current PRA model update and thus will also be compressed out of the quantification process. The current PRA model update will be completed prior to implementation of the 50.69 process.

RAI 02 - Fire PRA Facts and Observations

a. F&O 2-8 against SR PRM-B6 New Success Criteria Open F&O 2-8 is stated to be a documentation issue with no impact on the application; however, this F&O appears to identify potential modeling issues in the PRA model. Both the finding and the disposition to this F&O state that the characterization of a MSIV [main steam isolation valve] spurious opening as a LLOCA [large loss-of-coolant accident] above TAF [top of active fuel] was not supported by T/H [thermal-hydraulic]. Provide justification that appropriate success criteria have been used in the PRA model for scenarios that consider MSIV spurious opening as an LLOCA.

Response

The fire PRA documentation is in the process of being updated to include the justification for the MSIV spurious opening scenario success criteria. The fire PRA update will be completed and a new FPRA model of record reflecting this change will be in place prior to 10 CFR 50.69 implementation.

The PRA injection requirements for large loss of reactor pressure vessel (RPV) inventory scenarios are the same for pipe breaks greater than six inches (i.e., LLOCAs) and for the MSIV spurious opening scenario in which RPV inventory may be lost to the Main Condenser. The difference in the scenarios is containment response and location of a radionuclide release. The MSIV spurious opening scenario includes RPV inventory lost outside containment and may result in radionuclide release outside of containment.

Therefore, the MSIV spurious opening scenario is being revised, in the current model update, to be consistent with the containment bypass success criteria.

b. F&O 2-25 against SR FSS-D7 - Crediting Fire Detection and Suppression Systems The disposition to open F&O 2-25 only addresses the availability and/or reliability of the fire detection and suppression system(s). Both the standard and the finding state that fire detection and suppression systems may only be credited if they are installed and maintained in accordance with applicable codes and standards. Proper installation and maintenance is normally documented in a code compliance calculation. Verify that all credited fire detection and suppression systems have been reviewed for compliance to applicable codes and documented in a code compliance calculation. If not, evaluate all credited fire detection and suppression systems against the original code of construction, and upon completion, adjust credit in the PRA accordingly (remove credit for those systems not installed in accordance

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 5 of 30 Docket Nos. 50-352 and 50-353 with the original code of construction or provide justification that any deviations from the code do not impact system effectiveness).

Response

A review of the fire protection program was performed for the fire PRA credited fire detection and suppression systems. The review included the Fire Protection System Design Baseline Document, the UFSAR Section 9.5.1 (Fire Protection program), and UFSAR Appendix 9A (Fire Protection Evaluation Report). These documents detail the design of the fire protection systems in accordance with the applicable codes. In addition, a review of program health reports for the credited fire PRA systems was performed for the fire PRA. The review concluded that the credited fire PRA fire detection and suppression systems were installed and maintained in accordance with the applicable codes and standards.

Exelon procedures and the station work management processes provide the bases for establishing that the fire PRA credited systems are maintained in accordance with the applicable codes. The design change process procedure for the fire protection program assures that a proposed configuration change involving fire detection or suppression does not adversely impact the licensing basis for fire protection. The work management process prioritizes focus on fire detection or suppression systems and includes restoring inoperable systems under the station priority list. Monitoring and reporting of the effectiveness of the work management process in maintaining high levels of equipment reliability are achieved through the semi-annual health reporting metric for fire system impairments.

c. F&O 4-6 against SR HRA-A3- Undesired Operator Actions The disposition to open F&O 4-6 states that undesired operator actions, such as tripping or isolating equipment, were identified that could result from spurious signals, but such actions were determined to have "no material impact on the 10 CFR 50.69 application. The disposition explains that in such cases, there would be time for recovery if there is no damage to equipment from fire. Undesired operator actions based on spurious signal create additional risk. Moreover, the success of recovery actions can be hampered by fire or fire damage and the difficulty of diagnosing what is happening in the plant when spurious signals have occurred. In light of these observations:

i. Provide justification that the operators would recover from spurious signals (cues), since operators are trained to believe their instrumentation and procedural controls may dictate that the undesired action should be taken. Include discussion of procedural guidance that limits the possibility of undesired actions and the cognitive and execution challenges associated with recovering from undesired operator actions.

ii. Provide further justification that the risk from undesired operator actions is negligible and can be excluded from the fire PRA.

iii. If it cannot be justified that the risk from undesired operator actions is negligible, then incorporate undesired operator actions caused by spurious signals into the fire PRA.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 6 of 30 Docket Nos. 50-352 and 50-353

Response

The impact of potential undesired operator actions is being evaluated in the current fire PRA update. Potential undesired operator actions resulting from spurious indications from failure of a single instrument were identified based on a review of procedures. Operator talk-throughs have been performed to discuss the identified potential undesired operator actions and the training and procedural guidance used in these postulated scenarios. The updated fire PRA will include the impact of undesired operator actions based on the inputs from the operators and conclusions from the evaluation. The fire PRA update reflecting these changes will be completed and a new FPRA model of record reflecting this change will be in place prior to 10 CFR 50.69 implementation.

Since the addition of these impacts could be viewed as a PRA upgrade, upon completion of the current model update, and before implementation of 50.69, a focused scope review of this change will be performed.

d. F&O 4-23 against SR PRM-C1 - New Contributors Due to Spurious Operations Closed F&O 4-23 identifies that MCRAB [Main Control Room Abandonment] Event tree uses existing FPIE [full power internal events] success criteria and T-H [thermal-hydraulic]

analysis, without the proper confirmation or justification of applicability. The reported resolution in Attachment 3.b, Resolved Peer Review Findings, states, In general, the use of internal events and/or fire non-abandonment T/H runs for MCRAB actions is appropriate when the scenario details match closely enough. The use of the terms in general and closely enough seems to imply that there are situations where the use of internal events and/or fire non-abandonment T/H runs are not appropriate. Provide a discussion of how the situations where the internal events and/or fire non-abandonment T/H runs are not appropriate were addressed.

Response

The thermal-hydraulic analysis for in-Main Control Room (MCR) accident sequences and Main Control Room Abandonment (MCRAB) postulated sequences are not different for the fire PRA. The differences are in the operator action response. The fire PRA human reliability analysis for MCRAB operator actions include distinctions accounting for the differences in timing to accomplish credited operator actions for the differing scenarios (i.e.,

non-MCRAB and MCRAB). Specifically, the MCRAB credited operator actions include additional timing delays for the diagnosis, decision, and execution to establish the credited systems for MCRAB postulated scenarios based on procedures and operator interviews.

The time available for operator actions, rather than the system window, is the difference between non-MCRAB and MCRAB sequences. Therefore, the terms in general and closely enough in the F&O resolution were used to indicate that the thermal-hydraulic calculations are appropriate with the understanding that the timing to initiate certain functions differ between non-MCRAB and MCRAB scenarios. These timing differences were addressed in the human reliability analysis.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 7 of 30 Docket Nos. 50-352 and 50-353

e. F&O 4-30 against SR IGN-A7 - Exclusion of Junction Box Modelling The disposition to open F&O 4-30 states that the risk contribution from junction boxes has not been included in the fire PRA, but that the fire PRA will be modeled consistent with guidance in FAQ 13-0006 during the next PRA update. It is not clear to the U.S. Nuclear Regulatory Commission (NRC) staff that the risk associated with junction boxes is negligible for the 10 CFR 50.69 application. Provide justification that the risk from junction boxes has negligible impact on the 10 CFR 50.69 application, or alternatively, incorporate the risk associated with junction boxes into the fire PRA model.

Response

The fire PRA is in the process of being updated. Junction box fires have been added to the in-process updated fire PRA consistent with FAQ 13-0006. The fire PRA update reflecting these changes will be completed and a new FPRA model of record reflecting this change will be in place prior to 10 CFR 50.69 implementation.

f. F&O 4-34 against SR FSS-G1 - Exclusion of Certain Transient Fire Modeling Open F&O 4-34 cites examples of transient fires that were excluded from consideration without justification. The disposition to this F&O states that better documentation is needed for the basis of this screening but does not provide the basis for the exclusion. Identify the guidance used, or explain why transient fires are excluded in specific scenarios such as those cited in the F&O, or alternatively, incorporate the excluded transient fires into the fire PRA model.

Response

F&O 4-34 referencing SR FSS-G1 is related to the fire PRA multi-compartment analysis.

The fire PRA was updated in 2015 to incorporate transient fires. The fire PRA documentation is in the process of being updated to include the justification for screened transient fires in the multi-compartment analysis consistent with the guidance in NUREG/CR-6850. The fire PRA update reflecting these changes will be completed and a new FPRA model of record reflecting this change will be in place prior to 10 CFR 50.69 implementation.

g. F&O N/A against FSS-C6 - Focused-Scope F&Os against THIEF Model The disposition for the open self-identified F&O associated with PRA standard SR FSS-C6 explains that a focused-scope peer review was performed on the implementation of the thermally-induced electrical failure (THIEF) fire modelling tool, which resulted in two F&Os that are being resolved in the current (2017) model update. The impact of the resolution of these two F&Os on the 10 CFR 50.69 program is unknown. Provide justification for why the two unresolved F&Os from the focused-scope peer review associated with a PRA upgrade of the fire modeling have minimal impact on the 10 CFR 50.69 application, or alternatively, incorporate resolution to these two focused-scope fire modeling F&Os into the fire PRA model.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 8 of 30 Docket Nos. 50-352 and 50-353

Response

A focused scope peer review on the implementation of the THIEF model was completed in June 2017 and resulted in two F&Os.

F&O FSS-D4-1 references SR FSS-D4 regarding verification of the implementation of THIEF. A draft verification document was reviewed as part of the focused scope peer review. Subsequent to the focused scope review, the implementation of the THIEF model has been verified to produce the expected results, and this verification is in the process of being incorporated into the fire PRA documentation in the current model update. Therefore, the F&O is resolved and has no impact on the 10 CFR 50.69 application.

F&O FSS-H5-1 references SR FSS-H5 regarding documentation of the uncertainty of the input parameters used for THIEF. The fire PRA documentation is in the process of being updated to include the results of sensitivity studies where THIEF input parameters were varied. Sensitivity studies were performed and the conclusion was that the variation in input parameters would have a negligible impact on the fire PRA results. Therefore, the F&O is resolved and has no impact on the 10 CFR 50.69 application.

RAI 03 - PRA Maintenance versus PRA Upgrade

a. Attachment 3.b, Resolved Peer Review Findings, of the LAR provides information regarding the disposition of fire PRA F&Os that were closed by the July 2016 F&O closure review. The disposition for the F&Os does not include a discussion about whether each change is PRA maintenance or a PRA upgrade. Inspection of the reported change indicates that some changes may use a new methodology and that the change could impact significant accident sequences or the significant accident progression sequences (i.e., the change was an upgrade that should be peer reviewed).

i. For each of the following changes, summarize the original method in the PRA and the new method to demonstrate that the change is not an upgrade because a new methodology was not used and that the changes do not impact significant accident sequences or the significant accident progression sequences, or identify whether the change is determined to be an upgrade:

• changes to the fire PRA event trees to use of the fire initiating event decision tree, as indicated in closed F&Os 2-12 and 4-2;

Response

The fire initiating event decision tree (FIEDT) is a pre-event tree to determine which fire-induced failures cause an initiating event. The use of the FIEDT does not define new initiating events, accident sequences, success criteria, or include new system models. The FIEDT merely aligns fire-induced failures into the system logic that had already existed in the PRA model. The FIEDT is used to explicitly capture the fire-induced failures, and propagate them through the appropriate event tree logic. At the time of the Limerick peer review, this mapping approach for the fire-induced failures was utilized for some initiating events (i.e., those captured by the multiple spurious operation logic additions), but not for all. All of the fire initiators were also

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 9 of 30 Docket Nos. 50-352 and 50-353 assumed to result in a turbine trip initiator. This led to some potential double counting of the fire scenarios. The development of the FIEDT was done to eliminate the potential for double-counting of those MSO initiators. At the same time, the development of the FIEDT allowed for the expansion of the approach to include other potential initiating events as well. Therefore, this does not represent a PRA Upgrade.

• change to use of the Fire Modeling Workbook approach, as indicated in closed F&Os 2-24 and 4-17;

Response

The fire modeling workbook approach is not a new method because the workbook approach and the previous approach both use algebraic correlations. Therefore, this does not represent a PRA Upgrade. For example, the previous approach and the current approach both use the same plume temperature and heat flux correlations to estimate time to damage. These are the same correlations used in the NUREG-1805 spreadsheets.

The previous approach used the correlations to provide pre-solved generic solutions, which were then applied to specific ignition source configurations. The results from this approach were not transparent to the analyst (e.g., the results did not distinguish between target damage from plume temperature or heat flux). Therefore, many iterations were required to fully model the potential fire impacts on surrounding equipment and cables.

The fire modeling workbook approach removed the middle step of pre-solved solutions, and uses the correlations based on user input for the ignition source configuration of interest. The fire modeling workbook approach is Microsoft Excel based and provides results for the desired heat release rate distribution and damage mechanism (e.g., plume temperature or heat flux). Therefore, the analyst may use a single calculation and the ranges of results to glean many insights for the selected fire scenario.

The fire modeling workbook does include the THIEF model, which was subject to a focused scope peer review in June 2017.

• change to replacement of Fire Modelling Treatment notebook-based reduced heat release rates for transient fires of 60 kilowatts (kW) and 140 kW with guidance endorsed by the June 21, 2012, letter from Joseph Giitter to Biff Bradley entitled, Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinets Fires (ADAMS Package Accession No. ML12172A406), as indicated in open F&O 4-35;

Response

The referenced memo endorsed the Fire PRA Methods Review Panel decision that the use of a lower justified transient heat release rate (HRR) is not a new method, but a clarification of the existing method guidance in NUREG/CR-6850. The panel

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 10 of 30 Docket Nos. 50-352 and 50-353 decision included the provision that if the use of the guidance for lower transient HRR was subsequent to the fire PRA peer review then a focused scope peer review is warranted. For the Limerick fire PRA, the use of a lower transient HRR was included in the peer reviewed fire PRA. The updates to the fire PRA since the peer reviewed model only enhanced the justification for the use of lower transient HRRs based on clarifications of the guidance in the description of treatment for transient fires. The justification considers specific aspects in the guidance such as administrative controls and plant location configurations with limited floor space.

Therefore, this does not represent a PRA upgrade.

• change to use of the guidance in Frequently Asked Question (FAQ) 14-0009, Treatment of Well Sealed MCC [Motor Control Center] Electrical Panels Greater than 440V [volts], dated October 20, 2014 (ADAMS Accession No. ML15118A810),

as indicated in closed F&O 4-26; and

Response

The guidance in FAQ 14-0009 clarifies that fire propagation from a well-sealed MCC greater than 440V must be considered similar to other NUREG/CR-6850 Bin 15 electrical cabinet fires. The FAQ provides a factor that represents the fraction of fires assumed to breach a well-sealed MCC. The FAQ provides fire severity factors that may be used based on fire modeling using the plume centerline temperature correlation from NUREG-1805. The FAQ also suggest that other accepted fire modeling methods may be applied for the specific configuration of interest.

Previously, the fire PRA did not include fire propagation outside a well-sealed MCC.

The current fire PRA uses the FAQ 14-0009 factor and provided fire severity factors.

The MCC fire severity factors use the same algebraic correlation applied in the fire models for other electrical cabinets in the fire PRA. The equipment and cable targets were identified using the same approach for MCCs and other electrical cabinets.

The difference in the treatment of MCCs and other electrical cabinets is the application of the FAQ factor for the fraction of fires assumed to breach a well-sealed MCC. Therefore, this does not represent a PRA upgrade.

• change to use of the guidance in FAQ 12-0064, Hot work/transient fire frequency:

influence factors, dated September 5, 2012 (ADAMS Accession No. ML122550050),

as indicated in closed F&O 5-7.

Response

The guidance in FAQ 12-0064 expands upon and provides clarifications to the guidance in NUREG/CR-6850. FAQ 12-0064 does not provide new calculation methods from those in NUREG/CR-6850 which were applied in the fire PRA.

Therefore, this does not represent a PRA upgrade.

The fire PRA previously used lower transient influence factors not in NUREG/CR-6850 to distinguish between locations in the plant that have more transient controls than others. FAQ 12-0064 provided new lower transient influence factors and guidance in using the lower transient influence factors. In applying the guidance in

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 11 of 30 Docket Nos. 50-352 and 50-353 FAQ 12-0064, it was concluded that the lower transient influence factors may not be applicable, and therefore, the lower transient influence factors were removed.

ii. Identify any other changes to the fire PRA not listed above that constitute PRA upgrades.

Response

Two upgrades associated with the Fire PRA were identified. In response to RAI 02, Part c, it was identified that the resolution to F&O 4-6 against SR HRA-A3 was to include undesired operator actions in the model update. The identification and modeling of undesired operator actions is considered an upgrade. In response to RAI 02, Part g, it was identified that the use of the THIEF model was an upgrade.

An upgrade associated with the FPIE PRA was identified. In response to RAI 01, Part a, it was identified that the resolution to the F&O against SR HR-A1 was to update the pre-initiator study.

iii. For each upgrade identified in items i. and ii. above, either provide the results of the focused-scope peer review(s) performed on these upgrades and disposition of any findings, or alternatively, commit to an implementation item in response to RAI 04 to perform a focused-scope peer review of the upgrade and to close all resulting F&Os through a new peer review or through the F&O closure process accepted by the NRC in its letter dated May 3, 2017 to Nuclear Energy Institute (NEI) (ADAMS Accession No. ML17079A427), prior to implementing the 10 CFR 50.69 categorization process.

Response

In response to RAI 02, Part g, it was identified that a focused scope peer review was performed for the use of the THIEF model. The response includes a description of the two resulting findings and the disposition for the applications.

The identification and modeling of undesired operator actions (RAI 02, Part c) will be subject to a focused scope peer review when the model update is complete and prior to implementing the 10 CFR 50.69 categorization process. See RAI 04 response below.

The updated pre-initiator study (RAI 01, Part a) will be subject to a focused scope peer review when the model update is complete and prior to implementing the 10 CFR 50.69 categorization process. See RAI 04 response below.

b. A September 19, 2016, RAI response (ADAMS Accession No. ML16263A218), associated with an earlier relief request dated April 13, 2016 (ADAMS Accession No. ML16104A122),

reported changes in the internal events PRA after the 2005 full scope peer review. The 2008 model LG108A and LG208A changes converted the HRA from a spreadsheet to the EPRI

[Electric Power Research Institute] HRA Calculator. The September 19, 2016, RAI response stated that the change was PRA maintenance because the HRA Calculator uses the same HRA methodologies as were used in the spreadsheets.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 12 of 30 Docket Nos. 50-352 and 50-353

i. Please briefly confirm that, or alternatively summarize how, the previous spreadsheets applied the following HRA methodologies used in the HRA calculator:

• included pre-initiator, post-initiator, and dependency analysis;

• quantified pre-initiators using the accident sequence evaluation program (ASEP) or technique for human error rate prediction (THERP);

• included a cognitive error and an execution error for each post-initiator HFE;

• analyzed the cognitive error using the EPRI cause-based decision tree method and human cognitive reliability/occupational radiation exposure, THERP, or SPAR methods.

• selected the appropriate input value (i.e., median or mean) for THERP or ASEP;

• analyzed the execution error using THERP;

• included or excluded recovery actions; and

• included a dependency analysis that could identify combinations of HFEs in cutsets and analyzed combinations of pre-initiators, postinitiators, and both pre- and post-initiators.

Response

The methods used prior to implementation of the EPRI HRA Calculator (HRAC) at Limerick for Human Error Probability development including pre-initiators, post-initiators, and dependency analysis are the same as those used via implementation of the EPRI HRAC. The transition to the use of the EPRI HRAC from a spreadsheet approach implementing the same methods does not constitute a method change.

• Pre-initiators were quantified using the ASEP methodology and were not transitioned to the EPRI HRAC.

• Post-initiator HFEs included a cognitive and execution error for each human error probability (HEP) both before and after transition to the EPRI HRAC.

• The cognitive error was developed using the EPRI CBDTM methodology and included an ASEP time reliability correlation contribution for short term actions both before and after transition to the EPRI HRAC.

• For each operator action evaluated in detail, the HEP determination spreadsheets worked in tandem with an associated Word document. The Word document listed all critical subtasks for the action. THERP values from NUREG-1278 were selected, recorded in the Word document, and manually summed. The sum was then manually recorded in the spreadsheet as total execution error. The reduction in manual effort afforded by the EPRI HRA Calculator suggests a reduction in the probability of analyst errors.

The spreadsheet THERP values were the median values reported in NUREG-1278.

When the ASME Standard was issued and directed the use of mean values per SR HR-D6, the conversion from median to mean values for the Limerick HRA was achieved via use of the HRAC at the next update opportunity, which was the 2008 PRA Update.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 13 of 30 Docket Nos. 50-352 and 50-353 Regarding the ASEP Pc contribution, the result from a separate ASEP curve spreadsheet was manually entered into the HEP determination spreadsheet.

• The execution error was analyzed using THERP both before and after transition to the EPRI HRAC.

• The spreadsheets are simply a tool to provide a quantitative result after recovery actions are identified. The spreadsheets are not used to identify recovery actions.

• The dependency analysis identified combinations of post-initiator HFEs in the cutsets and used the Swain and Guttmann approach identified in NUREG/CR-1278 for evaluating the dependencies among the events both before and after transition to the EPRI HRAC. Pre-initiator dependencies were identified a priori, and dependent events were quantified using the Swain and Guttmann approach identified in NUREG/CR-1278 based on the independent HEPs derived from the ASEP quantifications both before and after transition to the EPRI HRAC.

• Zero dependence was determined between pre-initiator and post-initiator HFEs both before and after transition to the EPRI HRAC.

ii. Although the estimated human error probabilities (HEPs) are not expected to be identical between the spreadsheet and the EPRI calculator results, confirm that any significant differences were evaluated and are based on known differences between the implementation of the two tools.

Response

The EPRI HRA Calculator was used for the first time for Limerick during the 2008 Internal Events PRA update. HEP differences between the 2004 and 2008 PRA models for independent, risk-significant operator actions were identified as given in the table below. In summary, 17 risk-significant independent actions were present in the 2008 results with 4 HEPs decreasing in value and 13 HEPs increasing in value.

For the majority of actions with HEP increases, the main source of increase was the implementation of mean values from THERP as required per the ASME PRA Standard.

The 2004 Limerick HRA used median values. Note that the change to mean values would have occurred even if the HRAC had not been used. The only significant HEP decreases were for WHUESWDXD0 and WHUESWDXI0 which reflect the same action (ESW pump start) in two different time frames. The main reason these HEPs decreased is due to the elimination of subtasks that were determined to be non-critical. As in the use of mean values, this change is independent of the HRAC.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 14 of 30 Docket Nos. 50-352 and 50-353 Risk Significant HEPs Spreadsheet vs HRA Calculator Value Comparison HEP Description LG104C LG108A Comment Values Values (spreadsheet) (Calculator)

AHUATWDXI FAILURE TO EMERG; DEPRESSURIZE 2.1E-02 3.7E-02 The system window for this action was reduced by 2 AFTER HPI FAILS IN ATWS minutes for the 2008 update as the 2004 update used the time to core damage rather than the time to depressurize in order to avoid core damage. This reduction produced an increase in the ASEP contribution. The CBDTM selections and the execution subtasks remained the same, but execution error increased due to the use of mean THERP values as required by the ASME PRA Standard.

AHUFINDXI OPERATOR FAILS TO INHIBIT ADS 1.9E-03 1.8E-03 The CBDTM selections did not change and the execution subtasks remained the same. As in other actions, execution error increased due the transition to mean values as required by the ASME PRA Standard. However, this increase was offset by the application of a slightly longer system window which reduced the ASEP contribution.

AHUINXDXI FAILURE TO INHIBIT ADS IN ATWS 6.4E-02 6.8E-02 The CBDTM selections did not change, the ASEP W/O FEEDWATER AVAILABLE contribution remained the same, and the execution subtasks remained the same. The HEP increase is primarily due to an increase in execution error caused by the transition to mean values as required by the ASME PRA Standard.

AHUXTRDXI FAILURE TO INITIATE CONTROLLED 3.0E-04 3.6E-04 The system window for this action was reduced by 14 MANUAL DEPRESSURIZATION minutes for the 2008 update as the 2004 update used the time to core damage rather than the time to depressurize in order to avoid core damage. This reduction produced an increase in the ASEP contribution. The CBDTM selections and the execution subtasks remained the same, but execution error increased due to the use of mean THERP values as required by the ASME PRA Standard.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 15 of 30 Docket Nos. 50-352 and 50-353 Risk Significant HEPs Spreadsheet vs HRA Calculator Value Comparison HEP Description LG104C LG108A Comment Values Values (spreadsheet) (Calculator)

BHUMX1DXI CRD FLOW NOT MAXIMIZED PER T-240 1.8E-02 2.7E-02 The HEP increase is primarily due to the transition to (AFTER DEP AT HCTL) mean values as required by the ASME PRA Standard.

DHUSPCDXI FAILURE OF OPERATOR TO INITIATE 5.4E-05 3.9E-04 The CBDTM selections did not change and the RHR/SPC execution subtasks remained the same. The HEP increase is primarily due to the transition to mean values as required by the ASME PRA Standard.

DHUSPCDXD COND. FAILURE OF OPERATOR TO 2.2E-01 5.4E-01 As in the DHUSPCDXI action, the CBDTM selections INIT. SPC/SDC LATE did not change and the execution subtasks remained the same. The HEP increase is primarily due to the transition to mean values as required by the ASME PRA Standard. Note that this later version of the action is applied as a conditional HEP.

JHU002DXD0 CONDITIONAL FAILURE TO BYPASS 5.1E-01 5.5E-01 An increase in execution error caused by the PUMP TRIP (LATER BY PCPL) transition to mean values as required by the ASME PRA Standard was somewhat offset by the change in THERP selection from Table 20-12-9 and 20-12-5 to 20-12-8a as the control switches for the associated equipment were confirmed to be two position switches. There was a minor increase in the CBDTM contribution as a change to Pcg was made to reflect the presence of an OR statement.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 16 of 30 Docket Nos. 50-352 and 50-353 Risk Significant HEPs Spreadsheet vs HRA Calculator Value Comparison HEP Description LG104C LG108A Comment Values Values (spreadsheet) (Calculator)

JHU002DXI0 OPERATOR FAILS TO BYPASS PUMP 7.8E-04 7.7E-04 As this is a conditional HEP that represents the TRIP (EARLY BY HCTL) delayed case of the JHU002 action, the relative increase for the applied HEP is not proportional to the independent HEP increase reflected in the JHU002DXI action. An increase in execution error caused by the transition to mean values as required by the ASME PRA Standard was somewhat offset by the change in THERP selection from Table 20-12-9 and 20-12-5 to 20-12-8a as the control switches for the associated equipment were confirmed to be two position switches. There was a minor increase in the CBDTM contribution as a change to Pcg was made to reflect the presence of an OR statement.

SHULCEDXI FAILURE TO CONTROL LEVEL EARLY IN 4.3E-02 5.1E-02 The CBDTM selections did not change, the ASEP ATWS contribution remained the same, and the execution subtasks remained the same. The HEP increase is primarily due to an increase in execution error caused by the transition to mean values as required by the ASME PRA Standard.

SHULCLDXI FAILURE TO CONTROL LEVEL LATE IN 1.3E-01 2.8E-01 The CBDTM selections did not change, the ASEP ATWS contribution remained the same, and the execution subtasks remained the same. The HEP increase is primarily due to an increase in execution error caused by the transition to mean values as required by the ASME PRA Standard.

VHULCIDXI FAILURE TO CONTROL RPV LEVEL WITH 9.6E-02 1.0E-01 The CBDTM selections did not change, the ASEP LP ECCS W/ SORV contribution remained the same, and the execution subtasks remained the same. The HEP increase is primarily due to an increase in execution error caused by the transition to mean values as required by the ASME PRA Standard.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 17 of 30 Docket Nos. 50-352 and 50-353 Risk Significant HEPs Spreadsheet vs HRA Calculator Value Comparison HEP Description LG104C LG108A Comment Values Values (spreadsheet) (Calculator)

VHULCXDXI FAILURE TO CONTROL RPV LEVEL WITH 9.6E-02 1.0E-01 The CBDTM selections did not change, the ASEP LP ECCS W/ HPCI FAILED contribution remained the same, and the execution subtasks remained the same. The HEP increase is primarily due to an increase in execution error caused by the transition to mean values as required by the ASME PRA Standard.

VHUVTHDXI OPERATOR FAILS TO INITIATE VENT 7.1E-03 2.0E-02 The CBDTM selections did not change and the GIVEN RHR HARDWARE FAILURE execution subtasks remained the same with the exception of the addition of the order to evacuate the Reactor Enclosure in the 2008 update. The HEP increase is primarily due to the transition to mean values as required by the ASME PRA Standard.

WHUESWDXD0 CONDITIONAL FAILURE ESW PUMPS FAIL 7.1E-02 5.0E-02 As this is a conditional HEP that represents the TO BE INITIATED LATE delayed case of the WHUESW action, the relative increase for the applied HEP is not proportional to the independent HEP increase reflected in the WHUESWDXI0 action. While the HEP would have increased due to CBDTM entry errors in addition to the use of mean THERP values in Pe, these increases were offset by the deletion of non-critical subtasks for valve closures.

WHUESWDXI0 ESW PUMPS FAIL TO BE INITIATED 3.1E-02 2.4E-02 While the HEP would have increased due to CBDTM edits made to reflect the presence of an AND statement in the procedure and remove credit for graphically distinct procedure steps in addition to the use of mean THERP values in Pe, these increases were offset by the deletion of non-critical subtasks for valve closures.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 18 of 30 Docket Nos. 50-352 and 50-353 Risk Significant HEPs Spreadsheet vs HRA Calculator Value Comparison HEP Description LG104C LG108A Comment Values Values (spreadsheet) (Calculator)

ZHUHRLDXI OPERATOR FAILS TO TAKE MANUAL 1.0E-01 1.1E-01 The CBDTM selections did not change, the ASEP CONTROL OF HPCI/RCIC EARLY contribution remained the same, and the execution subtasks remained the same. The HEP increase is primarily due to an increase in execution error caused by the transition to mean values as required by the ASME PRA Standard.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 19 of 30 Docket Nos. 50-352 and 50-353 iii. Alternatively, if item i. and ii. above cannot be confirmed, either provide the results of a focused-scope peer review of the HRA elements and disposition of any findings, or alternatively, commit to an implementation item in response to RAI 04 to perform a focused-scope peer review of the HRA and to close all resulting F&Os through a new peer review or through the F&O closure process accepted by the NRC in its letter dated May 3, 2017, to NEI prior to implementing the 10 CFR 50.69 categorization process.

Response

See responses to items i and ii above.

RAI 04 - Implementation Items Table 2 of the LAR supplement dated August 14, 2017, presents dispositions for assumptions and modeling uncertainties that include planned updates to the PRA models after the 10 CFR 50.69 amendment has been issued and before implementation of the 10 CFR 50.69 Table 2 of the LAR supplement dated August 14, 2017, presents dispositions for assumptions and modeling uncertainties that include planned updates to the PRA models after the 10 CFR 50.69 amendment has been issued and before implementation of the 10 CFR 50.69 program. It appears that these future updates can have an impact on the 10 CFR 50.69 categorization.

These updates include accounting for load shedding directed by procedure, updating the internal flooding pipe break frequencies, and removing credit for core melt arrest in vessel.

a. Justify why the updates listed above have no impact on the 10 CFR 50.69 categorization results or include them in response to item b. below.

Response

See response to item b below.

b. Provide a list of activities and PRA changes, including any items from RAI 01, RAI 02, RAI 03, and RAI 04.a, which will not be completed prior to issuing the amendment but must be completed prior to implementing the 10 CFR 50.69 categorization process (i.e.,

implementation items). Propose a mechanism that ensures these activities and changes will be completed and appropriately reviewed and any issues resolved prior to implementing the categorization process, such as a reference to the table of implementation items in a license condition.

Response

The Limerick PRA model is currently undergoing a scheduled model update. This update includes addressing the modifications to the PRA model to address the model uncertainties specified in Table 2 of the Attachment to Exelon letter dated August 14, 2017 (Reference 2),

FPIE and IF PRA F&Os (RAI 01), FPRA F&Os (RAI 02), and PRA Maintenance versus PRA Upgrade items (RAI 03), as they relate to the 50.69 process. Once reviewed and approved, in accordance with the Exelon Risk Management program requirements, the new model will be the Model of Record (MOR). This new MOR will be completed prior to implementation of 50.69 categorization at Limerick.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 20 of 30 Docket Nos. 50-352 and 50-353 Based on the above, Exelon proposes to add the following license condition to Appendix C, Additional Conditions, of the Renewed Facility Operating License Nos. NPF-39 and NPF-85 for Limerick, Units 1 and 2, respectively (see Attachment 2):

Exelon will complete the items listed in the response to RAI 04 in Attachment 1 of Exelon letter to NRC dated January 19, 2018 prior to implementation of 10 CFR 50.69.

RAI 05.a, 05.b, and 05.c - Overall Categorization Process LAR Section 3.1.1, Overall Categorization, process has two different sets of bulleted elements and concludes with an additional list of ten elements. Some of the elements discuss training that will be given, some discuss the different hazard models, and some discuss PRA model results.

It is not clear from these discussions what the sequence of evaluations will be in the categorization process, what information will be developed and used, and what guidance on acceptable decisions by the Integrated Decisionmaking Panel (IDP) will be followed during the categorization of each system. Information on the training and expertise of the IDP team is provided in the LAR and need not be repeated in this RAI response.

a. Please summarize, in the order they will be performed, the sequence of elements or steps that will be followed for each system that will be categorized. A flow chart, such as that provided in the NEI presentation (ADAMS Accession No. ML17249A072) for the September 6, 2017, public meeting with NEI regarding 10 CFR 50.69 LARs (ADAMS Accession No. ML17265A020) may be provided instead of a description. The steps should include:

i. the input from all PRA evaluations such as use of the results from the internal events, internal flooding, and fire PRAs; ii. the input from non-PRA approaches (seismic, other external events, and shutdown);

iii. the input from the responses to the seven qualitative questions in Section 9.2 of NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline; iv. the input from the defense-in-depth matrix; and

v. the input from the passive categorization methodology.

b. In description to item a. above, please clarify the difference between preliminary HSS [high safety significant] and assigned HSS, and identify which inputs can and which cannot be changed from preliminary HSS to low safety significant by the IDP, and confirm that the proposed approach is consistent with the guidance in NEI 00-04, as endorsed by Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Reactor Power Plants According to Their Safety Significance.

c. In description to item a. above, please clarify which steps of the process are performed at the function level and which steps are performed at the component level. Describe how the categorization of the component impacts the categorization of the function and vice-versa.

Describe instances in which the final safety significance of the function would differ from the

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 21 of 30 Docket Nos. 50-352 and 50-353 safety significance of the component(s) that support the function, and confirm that the proposed approach is consistent with the guidance in NEI 00-04, as endorsed by RG 1.201.

Response

The process to categorize each system will be consistent with the guidance in NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, as endorsed by RG 1.201. RG 1.201 states that the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and that all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by §50.69(c)(1)(iv). However, neither RG 1.201 nor NEI 00-04 prescribe a particular sequence or order for each of the elements to be completed. Therefore, the order in which each of the elements of the categorization process (listed below) is completed is flexible and as long as they are all completed they may even be performed in parallel. Note that NEI 00-04 only requires Item 3 to be completed for components/functions categorized as LSS by all other elements. Similarly, NEI 00-04 only requires Item 4 to be completed for safety related active components/functions categorized as LSS by all other elements.

1. PRA-based evaluations (e.g., the internal events, internal flooding, and fire PRAs)

2. non-PRA approaches (e.g., seismic safe shutdown equipment list (SSEL), other external events screening, and shutdown assessment)

3. Seven qualitative criteria in Section 9.2 of NEI 00-04

4. the defense-in-depth assessment

5. the passive categorization methodology Below is an example of the major steps of the categorization process described in NEI 00-04:

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 22 of 30 Docket Nos. 50-352 and 50-353 Figure 5-1: Categorization Process Overview Categorization of SSCs will be completed per the NEI 00-04 process, as endorsed by RG 1.201, which includes the determination of safety significance through the various elements identified above. The results of these elements are used as inputs to arrive at a preliminary component categorization (i.e., High Safety Significant (HSS) or Low Safety Significant (LSS)) that is presented to the Integrated Decision-Making Panel (IDP). Note: the term preliminary HSS or LSS is synonymous with the NEI 00-04 term candidate HSS or LSS. A component or function is preliminarily categorized as HSS if any element of the process results in a preliminary HSS determination in accordance with Table 1 below. The safety significance determination of each element, identified above, is independent of each other and therefore the sequence of the elements does not impact the resulting preliminary categorization of each component or function. Consistent with NEI 00-04, the categorization of a component or function will only be preliminary until it has been confirmed by the IDP. Once the IDP confirms that the categorization process was followed appropriately, the final Risk Informed Safety Class (RISC) category can be assigned.

The IDP may direct and approve detailed categorization of components in accordance with NEI 00-04, Section 10.2. The IDP may always elect to change a preliminary LSS component or function to HSS; however, the ability to change component categorization from preliminary HSS to LSS is limited. This ability is only available to the IDP for select process steps as described in NEI 00-04 and endorsed by RG 1.201. Table 1 summarizes these IDP limitations in NEI 00-04.

The steps of the process are performed at either the function level, component level, or both.

This is also summarized in Table 1 below. A component is assigned its final RISC category upon approval by the IDP.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 23 of 30 Docket Nos. 50-352 and 50-353 Table 1: IDP Changes from Preliminary HSS to LSS Drives Categorization Step - IDP Change Element Evaluation Level Associated NEI 00-04 Section HSS to LSS Functions Internal Events Base Case -

Yes Not Allowed Section 5.1 Fire, Seismic and Other Risk (PRA No Allowable External Events Base Case Component Modeled)

PRA Sensitivity Studies No Allowable Integral PRA Assessment -

Yes Not Allowed Section 5.6 Fire, Seismic and Other Component Not Allowed External Hazards No Risk (Non-modeled)

Shutdown - Section 5.5 Function/Component Not Allowed No Defense-in- Core Damage - Section 6.1 Function/Component Yes Not Allowed Depth Containment - Section 6.2 Component Yes Not Allowed Qualitative Considerations - Section 9.2 Function N/A Allowable Criteria Passive Passive - Section 4 Segment/Component No Not Allowed The mapping of components to system functions is used in some categorization process steps to facilitate preliminary categorization of components. Specifically, functions with mapped components that are determined to be HSS by the PRA-based assessment (i.e., Internal events PRA or Integrated PRA assessment) or defense-in-depth evaluation will be initially treated as HSS. However, NEI 00-04, Section 10.2 allows detailed categorization which can result in some components mapped to HSS functions being treated as LSS; and Section 4.0 discusses additional functions that may be identified (e.g., fill and drain) to group and consider potentially LSS components that may have been initially associated with a HSS function but which do not support the critical attributes of that HSS function. Note that certain steps of the categorization process are performed at a component level (e.g. Passive, Non-PRA-modeled hazards - see Table 1). These components from the component level assessments will remain HSS (IDP cannot override) regardless of the significance of the functions to which they are mapped.

Therefore, if a HSS component is mapped to a LSS function, that component will remain HSS. If an LSS component is mapped to an HSS function, that component may be driven HSS based on Table 1 above, or may remain LSS.

RAI 05.d - Preliminary Safety Significance of Functions

d. Section 7 of NEI 00-04 states that If any SSC is safety significant, from either the PRA-based component safety significance assessment (Section 5) or the defense-in-depth

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 24 of 30 Docket Nos. 50-352 and 50-353 assessment (Section 6), then the associated system function is preliminary safety significant. The NRC staff interprets that the cited guidance applies to all aspects identified in Sections 5 and 6 of NEI 00-04, including Sections 5.3 through 5.5, dedicated to seismic, external hazards, or shutdown risk.

If the licensees categorization process differs from the guidance in Section 7 of NEI 00-04 cited above where functions supported by any HSS component(s) will be assigned HSS, describe and justify the approach.

Response

Section 5 defines categorization process considerations for both PRA-based and non-PRA-based (i.e., deterministic) assessment methods. Section 5.3, for example, describes the process for categorization from seismic risk considerations using either a seismic PRA (i.e.,

PRA-based) or using a seismic margin assessment (SMA, i.e., deterministic and not PRA-based). Section 7 requires assigning the safety significance of functions to be preliminary HSS if it is supported by an SSC determined to be HSS from the PRA-based assessment in Section 5, but does not require this for SSCs determined to be HSS from non-PRA-based, deterministic assessments in Section 5. The interpretation of this requirement is further clarified in the Vogtle SER (ML14237A034) which states if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS.

The reason for this is that the application of non-PRA-based assessments results in the default safety significance categorization of any SSCs associated with the safe shutdown success paths defined in those deterministic assessments to be HSS regardless of its risk significance.

Therefore, there is no risk basis for assigning the SSC-associated functions to be HSS, since the deterministic analyses from which the associated safe shutdown equipment lists are derived do not define functions equivalent to those used in the categorization process. This is the reason that the guidance in Section 7 of NEI 00-04 clearly notes PRA-based in reference to Section 5 of NEI 00-04. The categorization process is consistent with the guidance in NEI 00-04 as endorsed by RG 1.201.

RAI 05.e - Results and Integration of Passive Categorization

e. The industry flow chart presented at the September 6, 2017, public meeting shows that the passive categorization would be undertaken separately from the active categorization.

i. Explain how the results from the passive categorization will be integrated with the overall categorization results.

ii. If the results from the passive categorization can be changed by the IDP, explain and justify the proposed approach.

Response

Please see the response to RAIs 05.a, 05.b, and 05.c above. If the results of the passive categorization are HSS, then the SSC is categorized as preliminary HSS regardless of the other

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 25 of 30 Docket Nos. 50-352 and 50-353 categorization elements. A HSS determination by the passive categorization process cannot be changed by the IDP, as noted in the response to these RAIs.

RAI 06 - SSCs Categorization Based on Other External Hazards NEI 00-04 provides guidance on including external events in the categorization of each SSC to be categorized. Fire (Section 5.2) and seismic (Section 5.3) hazards are discussed in Sections 5.2 and 5.3, respectively. All other hazards are discussed in Section 5.4, Assessment of Other External Hazards. Figure 5-6 in Section 5.4 illustrates the process that begins with the SSC selected for categorization and then proceeds through the flow chart for each external hazard.

LAR Section 3.2.4 states that the Limerick [Limerick Generating Station] categorization process will use screening results from the Individual Plant Evaluation of External Events (IPEEE) in response to GL [Generic Letter] 88-20 for evaluation of safety significance related to the

[following] other external hazards. LAR Section 3.2.4 continues that [a]ll SSCs credited in other IPEEE external hazards are considered HSS. The use of other instead of a more precise description does not allow the NRC staff to compare the licensees proposed process with the guidance.

a. Identify the external hazards that will be evaluated according to the flow chart in Figure 5-6 of NEI 00-04.

Response

The "other" external hazards that will be evaluated according to the flow chart in Figure 5-6 of NEI 00-04 are any hazards listed in Attachment 4 of the LAR (Reference 1), "External Hazards Screening" that have not been screened in accordance with ASME/ANS PRA Standard RA-Sa-2009.

For Limerick, all other external hazards (i.e., other than internal events, internal flood, internal fire, and seismic) have been screened as noted in the LAR. As part of the external hazard screening, an evaluation was performed to determine if there are components that participate in screened scenarios and whose failure would result in an unscreened scenario.

Consistent with the flow chart in Figure 5-6 of NEI 00-04, these components would be considered HSS.

b. Identify which hazards will have [a]ll SSCs credited [] considered HSS instead of using the flow chart.

Response

The statement All SSCs credited in other IPEEE external hazards are considered HSS was intended to be consistent with the flow chart in Figure 5-6 of NEI 00-04. There are no Other External Hazards that will be evaluated using a method other than depicted in the flow chart.

c. Describe and justify any additional method(s) different from (a) or (b) that will be used to evaluate individual SSCs against external hazards and identify the hazards that will be evaluated with these methods.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 26 of 30 Docket Nos. 50-352 and 50-353

Response

There are no additional method(s) different from a. or b. that will be used to evaluate individual SSCs against external hazards.

d. Confirm that all hazards not included in the categorization process (a), (b), or (c) above will be considered insignificant for every SSC and, therefore, will not be considered during the categorization process.

Response

All external hazards not included in the categorization process a., b., or c. above are considered insignificant for every SSC and, therefore, will not be considered during the categorization process.

e. Attachment 4 of the LAR indicates that external flooding and extreme wind or tornado hazards are screened. Justify the basis for screening and explain how the guidance in Figure 5-6 of NEI 00-04 will apply to these hazards and whether these hazards will or will not be considered during the categorization process.

Response

The basis for screening external flooding and extreme wind or tornado hazards in Attachment 4 of the LAR (Reference 1) is discussed below. The screening process followed the guidance in Figure 5-6 of NEI 00-04. The screening process includes an evaluation of whether SSCs participate in screened scenarios; and also considers whether, if credit for SSCs were removed relative to the hazard being evaluated, the hazard would then become unscreened. More specifically, for each external hazard in Attachment 4 of the LAR, an assessment was performed to determine if equipment (i.e., SSCs) is relied upon to mitigate a hazard based on the design basis and severe accident functions of the component. Such SSCs would be considered HSS.

External Flooding The external flooding hazard at the site was recently updated as a result of the post-Fukushima 50.54(f) Request for Information and the flood hazard reevaluation report (FHRR) was submitted to NRC for review on March 12, 2015 (ADAMS Accession No. ML15084A586). The results indicate that flooding from rivers and streams (precipitation based) and dam failure are bounded by the current licensing basis (CLB) and do not pose a challenge to the plant. However, Local Intense Precipitation (LIP) screening relied on Diesel Generator Building exterior doors being closed. Therefore, failure to credit the external diesel doors (doors to the DG Building) could result in an unscreened external flooding scenario and, as such, the external diesel doors would be HSS if the EDG system were categorized.

Extreme Wind or Tornado Hazards Section 3.3 of the Limerick UFSAR describes the capability of safety related structures to withstand wind and tornado loadings. The design basis tornado was reviewed against Table

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 27 of 30 Docket Nos. 50-352 and 50-353 6-1 of NUREG/CR-4461, Rev. 2, and was found to be bounding. Therefore, no additional considerations were necessary.

Section 3.5.1.4 of the Limerick UFSAR describes the capability of safety related structures to protect SSCs against tornado missiles. A comparison was made of tornado frequencies used in the current design basis and those based on data from NUREG/CR-4461, Rev. 2.

The comparison showed that estimated tornado frequencies have been reduced by more than a factor of 10 from the original design basis. Thus, the loss of ability to safely shut down the plant due to tornadoes and tornado missiles is estimated to be very small.

RAI 07 - Shutdown Risk LAR Section 3.2.5 states the Limerick categorization process will use the shutdown safety management plan described in NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management, for categorization of safety significance related to low power shutdown conditions. However, the LAR does not cite the other criteria specified in NEI 00-04, Section 5.5, pertaining to low power shutdown events (i.e., includes defense-in-depth attributes and failures that would initiate a shutdown event). Clarify and provide a basis for how the categorization of SSCs will be performed for shutdown events and how it is consistent with the guidance in NEI 00-04, as endorsed by RG 1.201.

Response

For plants without a shutdown PRA, such as Limerick, NEI 00-04, as endorsed by RG 1.201, allows the use of a modified process based on the NUMARC 91-06 program. Limericks categorization process will follow the guidance and criteria in Section 5.5 in NEI 00-04 to address shutdown risk. Below is a summary of the NEI 00-04 process and requirements.

The overall process for addressing shutdown risk is illustrated in Figure 5-7 of NEI 00-04.

NUMARC 91-06 specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. The key safety functions defined in NUMARC 91-06 are evaluated for categorization of SSCs. NEI 00-04 provides two criteria for SSCs to be considered preliminary HSS.

1. If a system/train supports a key safety function as the primary or first alternate means, then it is considered to be a primary shutdown safety system and is categorized as preliminary HSS. The station's Shutdown Safety Management Program, which is consistent with NUMARC 91-06, is used as a guide to identify primary and first alternative means. NEI 00-04 defines a primary shutdown safety system as also having the following attributes:

• It has a technical basis for its ability to perform the function.

• It has margin to fulfill the safety function.

• It does not require extensive manual manipulation to fulfill its safety function.

2. If the SSCs failure would initiate an event during shutdown plant conditions (e.g., loss of shutdown cooling, drain down), then that SSC is categorized as preliminary HSS.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 28 of 30 Docket Nos. 50-352 and 50-353 As stated in NEI 00-04, If the component does not participate in either of these manners, then it is considered a candidate as low safety significance with respect to shutdown safety.

RAI 08 - Passive Component Categorization LAR Section 3.1.2 states that for the categorization of passive components and the passive function of active components, Limerick will use the method for risk-informed repair/replacement activities consistent with the safety evaluation issued by the Office of Nuclear Reactor Regulation, Request for Alternative ANO2-R&R-004, Revision 1, Request to Use Risk-informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems, Third and Fourth 10-Year Inservice Inspection Intervals, for Arkansas Nuclear One, Unit 2, dated April 22, 2009 (ADAMS Accession No. ML090930246).

The safety evaluation for this methodology states that the methodology is only applied to Class 2 and Class 3 piping. This methodology is a modification to the American Society of Mechanical Engineers Code Case N-660, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities,Section XI, Division 1, and deleted one question from it (Section I-3.1.3(a)(2)), requiring that all Class 1 items (except for some Class 1 parts defined in 10 CFR 50.55a(c)(2)(i) and (ii)) should be classified as HSS.

Pease confirm that only Class 2 and Class 3 equipment will be categorized using this methodology or explain and justify how the methodology will be modified to include Class 1 equipment.

Response

The Statements of Consideration (SOC), 69 FR 68008, dated November 22, 2004, III.4.3 Section 50.55a(f), (g), and (h) Codes and Standards, states that Section 50.69(b)(2)(iv) removes RISC-3 SSCs from the scope of certain provisions of § 50.55a, relating to Codes and Standards. The provisions being removed are those that relate to treatment aspects, such as inspection and testing, but not those pertaining to design requirements established in § 50.55a.

Since this section discusses RISC-3, it acknowledges that the components discussed in this section have been subject to the categorization process.

Further, in the second paragraph of this section, it goes on to state: Section 50.55a(f) incorporates by reference provisions of the ASME Code, as endorsed by NRC, that contains inservice testing requirements. These are special treatment requirements. Through this rulemaking, RISC-3 SSCs are removed from the scope of these requirements and instead are subject to the requirements in § 50.69(d)(2). Note: there is no mention of limiting removal of these requirements to ASME Class 2 and 3 SSCs.

This section further states that Section 50.55a(g) incorporates by reference provisions of the ASME Code, as endorsed by NRC, that contain the inservice inspection, and repair and replacement requirements for ASME Class 2 and Class 3 SSCs. The Commission will not remove the repair and replacement provisions of the ASME Code required by § 50.55a(g) for ASME Class 1 SSCs, even if they are categorized as RISC-3, because those SSCs constitute principal fission product barriers as part of the reactor coolant system or containment.

Therefore, the SOC explicitly acknowledges that Class 1 SSCs can be subject to the categorization process and may result in a RISC-3 classification, but that the alternate treatment

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 29 of 30 Docket Nos. 50-352 and 50-353 for these components must retain the repair and replacement provisions of the ASME Code.

Therefore, the categorization methodology does not require modification in order to appropriately categorize Class 1 SSCs. The ASME classification of the SSC does not impact the methodology as the methodology evaluates the consequence of a rupture of the SSCs pressure boundary (direct and indirect effects). As stated in the Vogtle SER, categorizing solely based on consequence which measures the safety significance of the pipe given that it ruptures is conservative compared to including the rupture frequency in the categorization and the categorization will not be affected by changes in frequency arising from changes to the treatment. Therefore, this methodology is appropriate to apply to ASME Class 1 SSCs, as the consequence evaluation and deterministic considerations are independent of the ASME classification when determining the SSCs safety significance and will maintain this acceptable level of conservatism. The passive categorization process described in the Limerick 50.69 LAR is intended to apply the same risk-informed process accepted in the ANO2-R&R-004 for the passive categorization of Class 2 and 3 components, to Class 1 pressure retaining SSCs in the scope of the system being categorized.

The ANO RI-RRA passive methodology implements the same risk-informed inservice inspection (RI-ISI) consequence evaluation process contained in EPRI TR-112657, Revised Risk-Informed Inservice Inspection Procedure supplemented with additional qualitative considerations. The NRC SER of this EPRI topical report was issued by letter dated October 28, 1999. Section 3.2.1 of the SER describes the scope of the RI-ISI methodology as:

The full-scope option includes ASME Code Class 1, 2, and 3 piping, piping whose failure could prevent safety-related structures, systems, or components (SSCs) from fulfilling their safety functions, and non-safety-related piping that is relied upon to mitigate accidents for whose failure could cause a reactor scram or actuation of a safety-related system.

While many pressure boundary components (passive components) are not modeled in a PRA, the consequence evaluation process of TR-112657, Rev B-A provides an explicit and robust process for determining the importance of pressure boundary components for both moderate and high energy systems. Consistent with the ASME/ANS PRA Standard, this supplementary analysis is used to augment the base PRA information. Further, as discussed above, the methodology uses the consequence portion of EPRI RI-ISI process enhanced with additional considerations which provide an additional layer of confidence for categorizing Class 1 SSCs as well as Class 2, Class 3 and non-class SSCs.

The same process as it pertains to inservice inspection has been approved for use on the full scope and code class designations of pressure retaining piping and welds in nuclear power plants. It has been determined to be sufficiently robust to assess the consequence risk of Class 1 piping and welds in the context of ISI even without the additional qualitative steps. The ANO RI-RRA has also been determined to be sufficiently robust to assess the consequence of all Class 2 and Class 3 SSCs (with the additional qualitative steps) in the context of repair/replacement. Therefore, the ANO RI-RRA methodology should be sufficiently robust to assess the consequence of the full spectrum of pressure retaining components as well as active components with a pressure retaining function regardless of ASME classification.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 30 of 30 Docket Nos. 50-352 and 50-353 References

1. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants,"

dated June 28, 2017 (ADAMS Accession No. ML17179A161).

2. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Supplement to Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors," dated August 14, 2017 (ADAMS Accession No. ML17226A336).

3. Electronic mail message from V. Sreenivas, U.S. Nuclear Regulatory Commission, to Glenn Stewart, Exelon Generation Company, LLC, "Limerick 50.69 license amendment request application: Request for Information (RAI)," dated December 6, 2017 (ADAMS Accession No. ML17341A250).

ATTACHMENT 2 License Amendment Request Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Response to Request for Additional Information Application to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants" Markup of Proposed Renewed Facility Operating License (RFOL) Pages Unit 1 RFOL Pages Appendix C, Page 1 Unit 2 RFOL Pages Page 8 Page 9 Appendix C, Page 1

APPENDIXC ADDITIONAL CONDITIONS OPERATING LICENSE NO. NPF-39 Amendment No. +28, .:t-31-, 447, 184

(13) The license e's UFSAR supple ment submitted pursua nt to 10 CFR 54.21(d), as revised during the license renewal applica tion review process, and as revised in accordance with license condition 2.C.(12), describ es certain programs to be implem ented and activitie s to be comple ted prior to the period of extended operation (PEO).

(a) Exelon Generation Compa ny shall implem ent those new program s and enhanc ements to existing program s no later than Decem ber 22, 2028.

(b) Exelon Generation Compa ny shall comple te those activities design ated for completion prior to the PEO, as noted in Comm itment Nos. 18, 19, 20, 22, 23, 24, 28, 29, 30, 38, 39, 40, 41, 42, 43, and 47, of Append ix A of NUREG -2171, "Safety Evaluation Report Related to the License Renewal of Limeric k Genera ting Station, Units 1 and 2," no later than Decem ber 22, 2028, or the end of the last refueling outage prior to the period of extend ed operation, whiche ver occurs later.

(c) Exelon Generation Compa ny shall notify the NRC in writing within 30 days after having accomplished item (a) above and include the status of those activities that have been or remain to be comple ted in item (b) above.

D. The facility requires exemp tions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include (a) exemption from the require ment of Append ix J, the testing of containment air locks at times when the contain ment integrity is not required (Section 6.2.6.1 of the SER and SSER-3), (b) exemption from the requirements of Append ix J,

the leak rate testing of the Main Steam Isolation Valves (MSIVs) at the peak calculated contain ment pressure, Pa, and exemption from the requirements of Append ix J that the measured MSIV leak rates be include d in the summa tion for the local leak rate test (Section 6.2.6.1 of SSER-3), (c) exemption from the requirement of Append ix J, the local leak rate testing of the Traversing lncore Probe Shear Valves (Section 6.2.6.1 of the SER and SSER-3), and (d) an exemption from the schedule requirements of 10 CFR 50.33(k)(I) related to availability of funds for decommissioning the facility (Section 22.1, SSER 8). The special circumstances regarding exemp tions (a), (b) and (c) are identified in Section s 6.2.6.1 of the SER and SSER 3. An exemption from the criticali ty monitoring requirements of 10 CFR 70.24 was previously granted with NRC materials license No. SNM-1 977 issued Novem ber 22, 1988. The license e is hereby exempted from the requirements of 10 CFR 70.24 insofar as this requirement applies to the handling and storage of fuel assemb lies held under this renewed license.

Renewed License No. NPF-85

E. Deleted F. The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

G. This renewed license is effective as of the date of issuance and shall expire at midnight on June 22, 2049.

FOR THE NUCLEAR REGULATORY COMMISSION

!d k William M. Dean, Director Office of Nuclear Reactor Regulation

Enclosures:

()

r_)

Renewed License No. NPF-85

INSERT 1 Exelon Generation Company, LLC shall comply with the following conditions on the schedule noted below:

Amendment Additional Conditions Implementation Number Date XXX Exelon is authorized to implement 10 CFR Prior to implementation 50.69 subject to the following condition: of 10 CFR 50.69.

Exelon will complete the items listed in the response to RAI 04 in Attachment 1 of Exelon letter dated January 19, 2018.

INSERT 2 (14) The Additional Conditions contained in Appendix C, as revised through Amendment No.

[XXX], are hereby incorporated into this renewed license. Exelon Generation Company shall operate the facility in accordance with the Additional Conditions.

APPENDIX C ADDITIONAL CONDITIONS OPERATING LICENSE NO. NPF-85 Exelon Generation Company, LLC shall comply with the following conditions on the schedule noted below:

Amendment Additional Conditions Implementation Number Date XXX Exelon is authorized to implement 10 CFR Prior to implementation 50.69 subject to the following condition: of 10 CFR 50.69.

Exelon will complete the items listed in the response to RAI 04 in Attachment 1 of Exelon letter dated January 19, 2018.

Amendment No.