ML18113A870

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

ML18113A870
Person / Time
Site:	Limerick
Issue date:	04/23/2018
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML18113A870 (21)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com

10 CFR 50.90 April 23, 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).

4. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "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'," dated January 19, 2018 (ADAMS Accession No. ML18019A091).

5. 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 March 26, 2018 (ADAMS Accession No. ML18081B068).

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 April 23, 2018

Page 2 By letter dated June 28, 2017 (Reference 1), as supplemented by letter dated August 14, 2017 (Reference 2), Exelon Generation Com pany, 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.

In an email dated December 6, 2017 (ADAMS Accession No. ML17341A250) (Reference 3),

the U.S. Nuclear Regulatory Commission (NRC) staff requested that Exelon provide additional information. In a letter dated January 19, 2018 (ADAMS Accession No. ML18019A091) (Reference 4), Exelon replied to the NRC staff's request.

The NRC staff reviewed the information provided and identified the need for additional information 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 March 26, 2018 (ADAMS Accession No. ML18081B068) (Reference 5). The NRC email requested a response by April 27, 2018.

to this letter provides a restatement of the RAI questions followed by our responses. Attachment 2 is a list of PRA implementation items which must be completed prior to implementation of the 10 CFR 50.69 categorization process at Limerick. Attachment 3 contains proposed markups of the Limerick Unit 1 and Unit 2 Renewed Facility Operating Licenses. The proposed markups supersede in their entirety the markups proposed in the Reference 4 letter.

Exelon has reviewed the information supporting a finding of no significant hazards consideration, and the environmental considerat ion, 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.

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 April 23, 2018 Page 3 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 23rd day of April 2018. Respectfully, James Barstow Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments:

1. Response to Request for Additional Information, Application to Adopt 1 O CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants" cc: 2. Limerick 50.69 PRA Implementation Items 3. Markup of Proposed Renewed Facility Operating License (RFOL) Pages Regional Administrator

-NRC Region I NRC Senior Resident Inspector

-Limerick Generating Station NRC Project Manager, NRR -Limerick Generating Station Director, Bureau of Radiation Protection

-Pennsylvania Department of Environmental Protection wl attachments II II II ATTACHMENT 1 License Amendment Request Limerick Generating Stat ion, Units 1 and 2 Docket Nos. 50-352 and 50-353

Response to Request for Additional Information Application to Adopt 10 CFR 50.69, "Ri sk-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 8 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.

In an email dated December 6, 2017 (ADAMS Accession No. ML17341A250) (Reference 3), the

U.S. Nuclear Regulatory Commission (NRC) staff requested that Exelon provide additional information. In a letter dated January 19, 2018 (ADAMS Accession No. ML18019A091)

(Reference 4), Exelon replied to the NRC staff's request.

The NRC staff reviewed the information provided and identified the need for additional information 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 March 26, 2018 (ADAMS Accession No. ML18081B068) (Reference 5). Below is a restatement of the questions followed by our responses.

PRA 02.01 - Modeling Undesired Operator Actions in the Fire PRA

The response to request for additional information (RAI) 02.c states that potential undesired operator actions resulting from spurious indications from fire-induced damage of an instrument were identified. It further states that the impact from undesired operator actions will be incorporated into the fire probabilistic risk assessment (FPRA) but provides no additional explanation. The NRC notes that NUREG-1921, "EPRl/NRC-RES Fire Human Reliability

Analysis Guidelines," July 2012 (ADAMS Accession No. ML12216A104) provides guidance on how the complexities associated with identifying and modeling undesired operator actions will be addressed. In light of this:

a. Confirm that the methodology in NUREG-1921 will be applied, or

b. Describe in detail the method to be used including:

i. How Limerick will ensure that the range of possible erroneous indications caused by fire damage are identified and considered.

ii. How Limerick will ensure that the range of possible undesired operator actions are identified given possible erroneous indications. Include discussion of the possibility of undesired actuation of a pump as well [as] the undesired shutdown of a pump. Also, include discussion of the potential for undesired flow diversion as well as flow isolation.

iii. How the failure probabilities for these human error events will be determined and provide a basis for that determination supported by NRC or industry guidance.

Response The NUREG-1921 methodology was applied in identifying undesired operator actions and will be used to incorporate the actions into the fire PRA.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 2 of 8 Docket Nos. 50-352 and 50-353 PRA 02.02 - Sensitivity Study on THIEF Input Parameters

The response to RAI 02.g provided two findings and observations (F&Os) that resulted from the focused-scope peer review on the implementation of the thermally-induced electrical failure (THIEF) fire modeling tool. In disposition to one of the F&Os, FSS-H5-1, the licensee states that sensitivity studies were performed to address the uncertainties of the input parameters used for THIEF, and that the results of sensitivity studies indicate that the THIEF parameter inputs have a negligible impact on the FPRA and therefore the F&O has no impact on the 10 CFR 50.69 application. Summarize the sensitivity studies conducted on the THIEF parameter inputs and provide their results demonstrating the negligible impact on the 10 CFR 50.69 categorization.

Response The THIEF model predicts the temperature profile within a cable as a function of time to simulate the delay in damage to a cable, which allows additional time for manual suppression, thereby improving the manual suppression probabilit

y. Sensitivity studies were performed on the THIEF calculation radial increment, the cable properties, and the conduit size to determine the impact on manual suppression probability. Varying each of these parameters to a range of other inputs shows a relatively small impact on manual suppression probability. However, the categorization process includes the sensitivity studies from NEI 00-04 Table 5-3. One of the sensitivity studies specified in Table 5-3 removes all credit for manual suppression. Another sensitivity study will be performed as part of the categorization process which assumes credit for immediate manual suppression (see Attachment 2). These sensitivity studies bound any uncertainties related to THIEF input parameters.

PRA 04.01 - Implementation Items The response to RAI 04 presents changes to the licensing basis that relate to completing items listed in the response to RAI 04 prior to implementing 10 CFR 50.69 categorization process.

The response to RAI 04 refers generally to: (1) addressing PRA F&Os cited in RAI 01 and RAI 02, (2) performing focused-scope peer reviews on the two PRA upgrades identified in the response to RAI 03, and (3) future actions di scussed in the license amendment request (LAR) supplement dated August 14, 2017 to perform PRA updates to address uncertainties. The response to RAI 04 does not explicitly list the PRA updates that are required prior to implementation of the 10 CFR 50.69 categorization process as requested in RAI 04.

The NRC staff identified the following items that should be completed prior to implementation of the 10 CFR 50.69 categorization process:

i. Update the HRA pre-initiators in the internal events PRA model to meet Capability Category II of the ASME/ANS RA-Sa-2009 as endorsed by RG 1.200 Revision 2, conduct a focused-scope peer review of the pre-initiator analysis, and resolve any resulting F&Os, as indicated in response to RAI 01.a. ii. Remove credit for recovery of instrument air from the internal events PRA model, as indicated in response to RAI 01.d. iii. Update the success criteria for main steam isolation valve (MSIV) spurious opening, as indicated in response to RAI 02.a.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 3 of 8 Docket Nos. 50-352 and 50-353 iv. Modeling undesired operator actions in the FPRA, conduct a focused-scope peer review, and resolve any F&Os, as indicated in response to RAI 02.c, subject to follow up RAI 02.01 above. v. Update the FPRA model to model junction box fires consistent with frequently asked question (FAQ) 13-0006, as indicated in response to RAI 2.e. vi. Update the FPRA model to incorporate transient fires in the multi-compartment analysis, as indicated in response to RAI 2.f. vii. Update the pipe rupture frequencies in the internal flooding PRA to the most recent EPRI pipe rupture frequencies, as indicated on page 7 of supplement dated August 14, 2017. viii. Remove credit for core melt arrest in-vessel at high reactor pressure vessel (RPV) pressure conditions from the internal events PRA model, as indicated on page 7 of supplement dated August 14, 2017. ix. Update the PRA model to account for load shedding when crediting serial operation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) in loss of offsite power (LOOP) and station blackout (SBO) scenarios, as indicated on page 6 of the supplement dated August 14, 2017. To fully address the above issues, provide the following:

a. For each of the nine issues above, please indicate how the issue will be addressed. If the issue has been addressed please state the issue has been addressed. A table of "implementation items" has been used in previous risk-informed licensing actions to formally identify issues requiring resolution before implementation of the amendment.

Response Attachment 2 identifies items that are required to be completed prior to implementation of the 10 CFR 50.69 risk categorization process. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the

10 CFR 50.69 categorization process.

b. Please provide a method to assure that all issues will be addressed and any associated changes will be made, that focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard, and that any findings will be resolved and reflected in the PRA prior to implementation of the 10 CFR 50.69 categorization process (for example, a license condition that all applicable implementation items will be completed prior to categorization).

Response Based on the response to a. above, Exelon proposes to add the following to the proposed license condition in 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 3):

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 4 of 8 Docket Nos. 50-352 and 50-353 Exelon will complete the implementation items listed in Attachment 2 of Exelon letter to NRC dated April 23, 2018 prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

RAI 05.01 - Qualitative Function Categorization

Figure 5-1 provided in the response to RAI 5 show that the evaluation of the seven qualitative criteria in Section 9.2 of Nuclear Energy Institute (NEI) 00-04 is performed at the function level and prior to the Integrated Decision-making Panel (IDP). The response also states that "NEI 00-04 only requires [the seven criteria] to be completed for components/functions categorized as LSS." Table 1 in the RAI 5 response has a column "IDP change HSS to LSS" and has the entry

"[a]llowable" in the qualitative criteria row, which appears to contradict the premise that the seven criteria are only applied to LSS functions. The guidance in NEI 00-04 states that the IDP "should consider the impact of loss of the function/structure, system, and component (SSC) against the remaining capability to perform the basic safety functions." Please clarify the guidance that will be provided to assess the safety significance of a function when there is an impact on, or even loss of, the capability described in each of the seven criteria (e.g., is a single false response sufficient to assign the function HSS).

Response Table 1 below has been modified, as indicated below, to further clarify how guidance in NEI 00-04, Section 9.2, is applied to the Exelon categorization process. Changes are represented in bold italic and strikethrough.

Table 1:

Categorization Evaluation Summary IDP Changes from Preliminary HSS to LSS Element Categorization Step - NEI 00-04 Section Evaluation Level Drives Associated Functions IDP Change HSS to LSS Risk (PRA Modeled) Internal Events Base Case - Section 5.1 Component Yes Not Allowed Fire, Seismic and Other External Events Base Case No Allowable PRA Sensitivity Studies No Allowable Integral PRA Assessment - Section 5.6 Yes Not Allowed Risk (Non-modeled) Fire, Seismic and Other External Hazards Component No Not Allowed Shutdown - Section 5.5 Function/Component No Not Allowed Core Damage - Section 6.1 Function/Component Yes Not Allowed Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 5 of 8 Docket Nos. 50-352 and 50-353 Element Categorization Step - NEI 00-04 Section Evaluation Level Drives Associated Functions IDP Change HSS to LSS Defense-in-Depth Containment - Section 6.2 Component Yes Not Allowed Qualitative Criteria Considerations - Section 9.2 Function N/A Allowable 1 Passive Passive - Section 4 Segment/Component No Not Allowed Notes: 1 The assessments of the qualitative considerations are agreed upon by the IDP in accordance with Section 9.2. In some cases, a 50.69 categorization team may provide preliminary assessments of the seven considerations for the IDP's consideration; however, the final assessments of the seven considerations are the direct responsibility of the IDP.

The seven considerations are addressed preliminarily by the 50.69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step. Each of the seven considerations requires a supporting justification for confirming (true response) or not confirming (false response) that consideration. If the 50.69 categorization team determines that one or more of the seven considerations cannot be confirmed, then that function is presented to the IDP as preliminary HSS. Conversely, if all seven considerations are confirmed, then the function is presented to the IDP as preliminary LSS.

The System Categorization Document, including the justifications provided for the qualitative considerations, is reviewed by the IDP. The IDP is responsible for reviewing the preliminary assessment to the same level of detail as the 50.69 team (i.e., all considerations for all functions are reviewed). The IDP may confirm the preliminary function risk and associated justification or may direct that it be changed based upon their expert knowledge. Because the Qualitative Criteria are the direct responsibility of the IDP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the IDP. If the IDP determines any of the seven considerations cannot be confirmed (false response) for a function, then the final categorization of that function is HSS.

RAI 06.01 - SSCs that Participate in Screened Scenarios The response to RAI 06e for extreme wind and tornadoes states that "[t]he 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." However, the discussion in 6e then simply states that "[t]he design basis tornado was reviewed against Table 6-1 of, "Tornado Climatology of the Contiguous United States," NUREG/CR-4461, Rev. 2, and was found to be bounding. Therefore, no additional considerations were necessary." The Updated Fi nal Safety Analysis Report (UFSAR) indicates design rotational wind speed of 223.5 miles per hour (mph) at the 1E-7/year frequency, not significantly below the wind speed of 202 mph indicated in Table 6-1 of NUREG/CR-4461, Rev. 2 at the 1E-7/year frequency. Further, Table 3.3-2 in the UFSAR identifies "Tornado Protected Systems and Tornado-Resistant Enclosures." The RAI response does not appear to address high straight winds.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 6 of 8 Docket Nos. 50-352 and 50-353

a. Provide the expected frequency of the wind hazard (e.g. straight line winds, tornadoes, hurricanes, etc.) which could cause damage from wind effect or wind-borne missiles to SSCs credited for safe shutdown to demonstrate that this frequency is sufficiently low that the hazard can be screened out without credit for mitigating SSCs.

Response Section 3.3 of the LGS UFSAR describes the capability of safety related structures to withstand wind and tornado loadings. Structures that directly affect the ultimate safe shutdown of the plant are designed to resist applicable design basis tornado forces, which bound other winds. Per the UFSAR, the design basis tornado considers the forces associated with 300 mph winds applied over an entire structure. Per Table 6-1 of NUREG/CR-4461, Rev. 2, the 1E-7 probability tornado wind speed is 256 mph, based on the F-scale, and 202 mph, based on the more recent EF-scale.

Section 3.5.1.4 of the LGS UFSAR describes the capability of safety related structures to protect SSCs against tornado missiles and protection of the ESW and RHRSW system yard piping by burial. LGS is in conformance with Regulatory Guide 1.117 regarding systems to be protected from tornado missiles, except for unprotected parts of the ESW

and RHRSW systems and spray pond networks.

For these unprotected parts of the ESW and RHRSW systems and spray pond networks, a hazard analysis was performed in 1984 (Ref. NUS-4507, Limerick Generating Station - Ultimate Heat Sink Extreme Winds Hazard Analysis, March 1984) to determine the likelihood of the loss of ultimate heat sink (ESW and RHRSW) by tornados and tornado missiles. That hazar d analysis was reviewed by the NRC in NUREG-0991, "Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2," Supplement 4, dated May 1985. The analysis showed that the likelihood of a loss of all UHS was approximately 8E-7/yr. In addition, a comparison of tornado frequencies used in the 1984 hazard analysis against the tornado frequencies from NUREG/CR 4461 shows that estimated tornado frequencies are reduced by at least a factor of 10. This is illustrated in the table below.

Wind Speed (mph) Frequency from NUS-4507 Frequency from NUREG/CR-4461 (Curve Fit) 200 4E-5/yr 1E-6/yr 250 5E-6/yr 1E-7/yr 300 2E-7/yr 2E-8/yr In conclusion, the frequency of wind effect or wind-borne missiles to SSCs credited for safe shutdown is sufficiently low that the extreme wind or tornado hazard can be screened out without credit for mitigating SSCs.

b. Confirm that during categorization SSCs will be evaluated to determine if they participate in screened hazard scenarios and to determine whether their failures would result in an unscreened scenario.

Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 7 of 8 Docket Nos. 50-352 and 50-353 Response Not applicable to the extreme wind or tornado hazard. The hazard has been screened without taking credit for SSCs to mitigate the hazard.

c. Provide an alternative evaluation of the risk from high winds and tornado hazards, demonstrating how the guidance in NEI 00-04 Section 5.4 is met.

Response An alternative evaluation is not required. The guidance in NEI 00-04 Section 5.4 is met.

PRA 08.01 Passive Component Categorization

LAR Section 3.1.2 stated 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 (ANO-2), dated April 22, 2009 (ADAMS Accession No.

ML090930246).

The NRC staff notes that this methodology has been approved for Class 2 and Class 3 SSCs.

Because Class 1 SSCs constitute principal fission product barriers as part of the reactor coolant system or containment, the consequence of pressure boundary failure for Class 1 SSCs may be different than for Class 2 and Class 3, and therefore the criteria in the ANO-2 methodology cannot automatically be generalized to Class 1 SSCs without further justification. Therefore, in RAI 08 the NRC staff asked the licensee to confirm that this methodology will be applied to Class 2 and Class 3 SSCs or to justify how the methodology will be modified to include Class 1

SSCs.

The justification provided in the response to RAI 08 does not justify how the ANO-2 methodology can be applied to Class 1 SSCs and how sufficient defense in depth and safety margins are maintained. A technical justification for Class 1 SSCs should address how the methodology is sufficiently robust to assess the sa fety significance of Class 1 SSCs, including, but not limited to: justification of the appropriateness of the conditional core damage probability (CCDP) numerical criteria used to assign 'High', 'Medium' and 'Low' safety significance to these loss of coolant initiating events; identification and justification of the adequacy of the additional qualitative considerations to assign 'Medium' safety significance (based on the CCDP) to 'High' safety significance; justification for crediting operator actions for success and failure of pressure boundary; guidelines and justification for selecting the appropriate break size (e.g. double ended guillotine break or smaller break); and include supporting examples of types of Class 1 SSCs that would be assigned low safety significance, etc.

As mentioned in the meeting summary from the February 20, 2018, Risk-Informed Steering Committee (RISC) meeting (ADAMS Acce ssion No. ML18072A301), the NRC staff understands that the industry is planning to limit the scope to Class 2 and Class 3 SSCs, consistent with the Response to Request for Additional Information Attachment 1 Application to Adopt 10 CFR 50.69 Page 8 of 8 Docket Nos. 50-352 and 50-353 pilot Vogtle Electric Generating Plant, Units 1 and 2 license amendment (ADAMS Accession No. ML14237A034).

Please provide the requested technical justification or confirm the intent to apply the ANO-2 passive categorization methodology only to Class 2 and Class 3 equipment.

Response The passive categorization process is intended to apply the same risk-informed process

accepted by the NRC in the ANO2-R&R-004 for the passive categorization of Class 2, 3, and non-Code class components. This is the same passive SSC scope the NRC has conditionally endorsed in ASME Code Cases N-660 and N-662 as published in Regulatory Guide 1.147, Revision 15. Both code cases employ a similar risk-informed safety classification of SSCs in order to change the repair/replacement requirements of the affected LSS components. All ASME Code Class 1 SSCs with a pressure retaining function, as well as supports, will be designated as high-safety-significant, HSS, for passive categorization which will result in HSS for its risk-informed safety classification, and cannot be changed by the IDP. Therefore, this methodology and scope for passive categorization is acceptable and appropriate for use at Limerick for 10 CFR 50.69 SSC categorization.

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).

4. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "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'," dated January 19, 2018 (ADAMS Accession No. ML18019A091).

5. 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 March 26, 2018 (ADAMS Accession No.

ML18081B068).

.

ATTACHMENT 2 License Amendment Request Limerick Generating Stat ion, Units 1 and 2 Docket Nos. 50-352 and 50-353

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

Response to Request for Additional Information Attachment 2 Application to Adopt 10 CFR 50.69 Page 1 of 2 Docket Nos. 50-352 and 50-353 The table below identifies the items that are required to be completed prior to implementation of the 10 CFR 50.69 categorization process at Limerick Generating Station, Units 1 and 2. All issues identified below will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Limerick 50.69 PRA Implementation Items Description Resolution i. Update the HRA pre-initiators in the internal events PRA model to meet Capability Category II of the ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Revision 2, conduct a focused-scope peer review of the pre-initiator analysis, and resolve any resulting F&Os, as indicated in response to RAI 01.a contained in Exelon letter dated January 19, 2018. The HRA pre-initiators in the internal events PRA model will be updated to meet Capability Category II of the ASME/ANS RA-Sa-2009 as endorsed by RG 1.200, Revision 2. A focused-scope peer review will be conducted of the pre-initiator analysis, and any resulting F&Os will be resolved, as indicated in response to RAI 01.a contained in Exelon letter dated

January 19, 2018. ii. Remove credit for recovery of instrument air from the internal events PRA model, as indicated in response to RAI 01.d contained in Exelon letter dated January 19, 2018.

Credit for recovery of instrument air will be removed from the internal events PRA model, as indicated in response to RAI 01.d contained in Exelon letter dated January 19, 2018. iii. Update the success criteria for main steam isolation valve (MSIV) spurious opening, as indicated in response to RAI 02.a contained in Exelon letter dated January 19, 2018. The success criteria for main steam isolation valve (MSIV) spurious opening will be updated, as indicated in response to RAI 02.a contained in Exelon letter dated

January 19, 2018. iv. Model undesired operator actions in the FPRA, conduct a focused-scope peer review, and resolve any F&Os, as indicated in response to RAI 02.c contained in Exelon letter dated January

19, 2018. Undesired operator actions will be modeled in the FPRA. A focused-scope peer review will be conducted, and any F&Os will be resolved, as indicated in response to RAI 02.c contained in Exelon letter dated

January 19, 2018. v. Update the FPRA model to model junction box fires consistent with frequently asked question (FAQ) 13-0006, as indicated in response to RAI 2.e contained in Exelon letter dated January 19, 2018. The FPRA model will be updated to model junction box fires consistent with frequently asked question (FAQ) 13-0006, as indicated in response to RAI 2.e contained in Exelon letter dated January 19, 2018.

Response to Request for Additional Information Attachment 2 Application to Adopt 10 CFR 50.69 Page 2 of 2 Docket Nos. 50-352 and 50-353 Limerick 50.69 PRA Implementation Items Description Resolution vi. Update the FPRA model to incorporate transient fires in the multi-compartment analysis, as indicated in response to RAI 2.f contained in Exelon letter dated

January 19, 2018. The FPRA model will be updated to incorporate transient fires in the multi-compartment analysis, as indicated in response to RAI 2.f contained in Exelon letter dated January 19, 2018. vii. Update the pipe rupture frequencies in the internal flooding PRA to the most recent EPRI pipe rupture frequencies, as indicated on page 7 of Exelon supplement letter dated August 14, 2017. The pipe rupture frequencies will be updated in the internal flooding PRA to the most recent EPRI pipe rupture frequencies, as indicated on page 7 of Exelon supplement letter dated August 14, 2017. viii. Remove credit for core melt arrest in-vessel at high reactor pressure vessel (RPV) pressure conditions from the internal events PRA model, as indicated on page 7 of Exelon supplement letter dated August 14, 2017. Credit for core melt arrest in-vessel at high reactor pressure vessel (RPV) pressure conditions will be removed from the internal events PRA model, as indicated on page 7

of Exelon supplement letter dated August

14, 2017. ix. Update the PRA model to account for load shedding when crediting serial operation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) in loss of offsite power (LOOP) and station blackout (SBO) scenarios, as indicated on page 6 of the Exelon supplement letter dated August 14, 2017. The PRA model will be updated to account for load shedding when crediting serial operation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) in loss of offsite power (LOOP) and station blackout (SBO) scenarios, as indicated on page 7 of Exelon supplement letter dated August 14, 2017. x. There are several parameters used in the THIEF model that may affect the calculated time available for manual suppression, and therefore, the probability of manual suppression in fire

PRA scenarios where manual suppression is credited. Although the impact on the relative importance of modeled components is expected to be small, there is uncertainty associated with these parameters. As part of the categorization process for the fire PRA, in addition to the list of fire PRA categorization sensitivities specified in NEI 00-04, Table 5-3, a sensitivity will be performed in which credit is taken for immediate manual suppression in scenarios in which manual suppression is already modeled, as indicated in Exelon letter dated April 23, 2018.

ATTACHMENT 3 License Amendment Request Limerick Generating Stat ion, Units 1 and 2 Docket Nos. 50-352 and 50-353

Response to Request for Additional Information Application to Adopt 10 CFR 50.69, "Ri sk-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 (13) The licensee's UFSAR supplement submitted pursuant to 10 CFR 54.21 (d), as revised during the license renewal application review process, and as revised in accordance with license condition 2.C.(12), describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO). (a) Exelon Generation Company shall implement those new programs and enhancements to existing programs no later than December 22, 2028. (b) Exelon Generation Company shall complete those activities designated for completion prior to the PEO, as noted in Commitment.

Nos. 18, 19, 20, 22, 23, 24, 28, 29, 30, 38, 39, 40, 41, 42, 43, and 47, of Appendix A of NUREG-2171, "Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2," no later than December 22, 2028, or the end of the last refueling outage prior to the period of extended operation, whichever occurs later. (c) Exelon Generation Company 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 completed in item (b) above. D. The facility requires exemptions from certain requirements of 1 O CFR Part 50 and 10 CFR Part 70. These include (a) exemption from the requirement of Appendix J, the testing of containment air locks at times when the containment integrity is not required (Section 6.2.6.1 of the SER and SSER-3), (b) exemption from the requirements of Appendix J, the leak rate testing of the Main Steam Isolation Valves (MSIVs) at the peak calculated containment pressure, Pa, and exemption from the requirements of Appendix J that the measured MSIV leak rates be included in the summation for the local leak rate test (Section 6.2.6.1 of SSER-3), (c) exemption from the requirement of Appendix 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 exemptions (a), (b) and (c) are identified in Sections 6.2.6.1 of the SER and SSER 3. An exemption from the criticality monitoring requirements of 10 CFR 70.24 was previously granted with NRC materials license No. SNM-1977 issued November 22, 1988. The licensee is hereby exempted from the requirements of 1 O CFR 70.24 insofar as this requirement applies to the handling and storage of fuel assemblies 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 idk William M. Dean, Director Office of Nuclear Reactor Regulation

Enclosures:

Renewed License No. NPF-85 INSERT 1 Exelon Generation Company, LLC shall comply with the following conditions on the schedule noted below:

Amendment

Numbe r Additional Conditions Implementation Date [XXX] Exelon is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the license amendment dated [DATE] subject to the following condition:

Exelon will complete the implementation items

listed in Attachment 2 of Exelon letter to NRC dated April 23, 2018 prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Prior to implementation of 10 CFR 50.69.

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.

Renewed License No. NPF-85 Amendment No.

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

Numbe r Additional Conditions Implementation Date [XXX] Exelon is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the license amendment dated [DATE] subject to the following condition:

Exelon will complete the implementation items

listed in Attachment 2 of Exelon letter to NRC dated April 23, 2018 prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Prior to implementation of 10 CFR 50.69.