View source for NRC 2018-0044, Supplement to Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, ...

{{Adams
| number = ML18271A114
| issue date = 09/28/2018
| title = Supplement to Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, Risk Informed Categorization and Treatment of Structures, ...
| author name = Craven R
| author affiliation = NextEra Energy Point Beach, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000266, 05000301
| license number = DPR-024, DPR-027
| contact person = 
| case reference number = NRC 2018-0044
| document type = Letter type:NRC, Response to Request for Additional Information (RAI)
| page count = 11
| project = 
| stage = Supplement
}}

=Text=
{{#Wiki_filter:Exrera.
EN ERGY~
NRC 2018-0044 10 CFR 50.90 September 28, 2018 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301

==Subject:==
Supplement to Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, "Risk informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants"

==References:==
: 1. NextEra Energy Point Beach, LLC letter NRC 2017-0043, "License Amendment Request 287, Application to adopt 10 CFR 50.69, 'Risk informed Categorization and Treatment of Stmctures, System, and Components (SSCs) for Nuclear Power Plants,'" August 31, 2017 (ML17243A201)
: 2. NextEra Energy Point Beach, LLC letter NRC 2017-0052, "Supplement toLAR 287, Application to adopt 10 CFR 50.69, 'Risk informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants,'" October 26, 2017 (ML17299A012)
: 3. NRC e-mail "Final- Point Beach- 50.69 Risk Informed LAR 287, Request for Additional Inf01'mation, CAC Nos. MG0196 and MG0197, EPID L-2017-LLA-0284," June 27, 2018 (ML18178A599)
: 4. NextEra Energy Point Beach, LLC letter NRC 2018-0038, "Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk informed Categorization and Treatment of Suuctures, System, and Components (SSCs) for Nuclear Power Plants,'" August 10, 2018 (ML18222A539)
In Reference 1 and supplemented by References 2 and 4, NextEra Energy Point Beach, LLC (NextEra) submitted a license amendment request for the Point Beach Nuclear Plant, Units 1 and 2.
The proposed amendment would revise the licensing basis by adding a license condition to allow for implementation of the provisions of Title 10 of the Code ofFederal Regulations (1 0 CFR) Section 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors."
Next Era Energy Point Beac h, LLC 6610 Nucl ea r Road, Two Rivers , WI 542 14

NRC 2018-0044 Page 2 of2 On September 12, 2018, NextEra participated in a conference call with the NRC staff to discuss the license condition and implementation items provided in Reference 4. This letter clarifies these items as discussed during the conference call. Attachment 1 to this letter lists the items that must be completed prior to implementation of the 10 CFR 50.69 categorization process and supersedes Attachment 1 to the Enclosure in Reference 4. Attachment 2 provides markups of the Point Beach Units 1 and 2 operating licenses containing the proposed license condition and supersedes Attachment 2 to the Enclosure in Reference 4.
This response does not alter the conclusions in Reference 1 that the proposed change does not involve a significant hazards consideration pursuant to 10 CFR 50.92, and there are no significant environmental impacts associated with the change.
This letter contains no new or revised regulatory commitments.
Should you have any questions regarding this submittal, please contact lvlr. Eric Schultz, Licensing Manager, at 920-755-7854.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on ~F       :l Y, 2018 Sincerely, Robert Craven Site Director N extEra Energy Point Beach, LLC Enclosure cc:     Administrator, Region III, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Public Service Commission of Wisconsin

N RC 201 8-0044 A ttachment 1 Page 1 of 5 Attachment 1 Table of10 CFR 50.69 Implementation Items

NRC 2018-0044 Attachment 1 Page 2 of 5 The table below identifies the items that are required to be completed prior to implementation of the 10 CFR 50.69 categorization process at Point Beach Nuclear Plant, 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.
NextEra Energy Point Beach, LLC 10 CPR 50.69 PRA Implementation Items Description                                                              Resolution NextEra will perform an evaluation of the as-built, as-operated plant against    NextEra will perform an evaluation of the as-built, as-the SMA SSEL.                                                                    operated plant against the SMA SSEL. The SSEL updated to reflect the as-built, as-operated plant will be used in the 10 CFR 50.69 categorization.
The disposition to F&O IE-A1-01 presented in the TSTF-425 LAR                    The loss of a 4,160 V AC bus will be added to the PRA indicates that a number of special initiators related to the 4160 volts          model as a special initiator to resolve internal events alternating current ryA C) Vital Switchgear bus were not included in the         finding F&O IE-A1-01, as indicated in response to internal events PRA (IEPRA) model because they were considered not               RAI 02.a.
significant, and estimated the CDF for sequences associated with these initiators as high as 1.9E-07 /year. It is not clear to the NRC staff that excluding these sequences cannot increase the risk importance values for specific system components above the threshold criteria for determining safety significance as discussed in NEI 00-04, Section 5 The dispositions to F&Os AS-B6-01 and SY-A21-01 presented in the                 A new failure mode associated with EDG load TSTF-425 LAR stated that although Emergency Diesel Generator (EDG)               management will be added to the internal events and fire load management is a potential failure mode for EDGs, it is not modeled in      PRA model to resolve internal events findings AS-B6-01 the PRA. The TSTF-425 LAR further provided an estimate of the low               and SY-A21-01 and fire fmding PRM-B2-01, as indicated likelihood that the EDG load management would be needed. The                    in response to RAI 02.b and 02.f.
estimated low likelihood is based on the assumption that the events used in the estimate are independent. The NRC staff notes that loss of coolant accident (LOCA) initiators can induce loss of offsite power (LOOP) events and LOOP initiators can induce LOCAs (i.e., the need for safety injection (SI)) making these events de.e_endent. Additionally, NRC staff notes that

NRC 2018-0044 Attachment 1 Page 3 of 5 NextEra Energy Point Beach, LLC 10 CFR 50.69 PRA Implementation Items Description                                                         Resolution modelling exclusions that cause just small increases to CDF /LERF can impact the categorization of specific structures, systems, and components (SSCs).
The disposition to F&O AS-B7-01 presented in the TS1F-425 LAR states          The treatment of power recovery after LOOP events and that recovery of LOOP events is only credited for station blackout (SBO)      battery modeling in the PRA model will be revised to be scenarios and the direct current (DC) batteries are conservatively assumed    more realistic to resolve internal events finding AS-B7-01, to fail at time zero. Conservative modeling in the PRA can skew the plant's   as indicated in response to RAI 02.c.
risk proftle and impact the SSCs risk importance values determined as part of 10 CFR 50.69 categorization.
The disposition to F&O HR-D1-01 in LAR Attachment 3 states that no            F&O HR-D1-01 will be closed by the licensee using an further changes are required. Also, based on the disposition presented in    NRC-accepted process (e.g., full-scope peer review, the TSTF-425 LAR, the F&O appeared resolved. Yet, the LAR associated         focused-scope peer review, or F&O closure review in with the adoption of 10 CFR 50.69 states regarding the disposition of this    accordance with Appendix X accepted by the NRC, with F&O, "[p]rior to implementation, either this fmding will be closed or a       conditions, in letter dated May 3, 2017).
sensitivity study case will be performed to determine the impact on the CDF and LERF res~lts for those categorizations that could be adversely affected by this finding."
The description for F&O IFQU-A6-01 states, "HFEs from internal events        F&O IF-QU-A6-01 will be closed by the licensee using an are 'adjusted' with inadequate basis for those adjustments." The disposition NRC-accepted process (e.g., full-scope peer review, for this F&O presented in the TSTF-425 LAR states that "stress               focused-scope peer review, or F&O closure review in multipliers" from Table 20-16 ofNUREG/CR-1278, "Handbook of                  accordance with Appendix X accepted by the NRC, with Human Reliability Analysis with Emphasis on Nuclear Power Plant              conditions, in letter dated May 3, 2017) .
Applications, Final Report," 1983 (ADAMS Accession Number ML071210299), which are referred to as "stress modifiers" in NUREG/CR-1278,-were used to account for the stress associated with an internal flooding event. However, Table 20-16 of the cited NUREG lists modifiers to be applied to human error probabilities (HEPs) for different stress levels and they are not specific to internal flooding.

NRC 2018-0044 Attachment 1 Page 4 of 5 NextEra Energy Point Beach, LLC 10 CFR 50.69 PRA Implementation Items                                 I Description                                                        Resolution The description for F&O HRA-B2-01 found that credit for graphically           The HEPs developed for the Fire PRA model will be distinct factors is taken for all HRA events, as opposed to taking credit for updated to remove the graphically distinct credit in the graphically distinct procedural steps that stand out from the other steps. cognitive portion of the HEP. The dependency analysis The disposition for this F&O states, "[o]nly about 10% of the HEPs that       will be updated and the Fire PRA quantified using these credited graphically distinct procedure steps would be increased by more      updated Human Error Probabilities, as indicated in than a factor of 2." The disposition concludes "[b]ased on this review, the   response to RAI 02.g.
impact on the modd from this finding is judged minimal." NRC staff notes that modelling issues that can cause even small increases to CDF and LERF can potentially increase the risk importance values for specific system components above the threshold criteria for determining safety significance
~ecified in NEI 00-04, Section 5 The description for F&O FQ-A1-01 states, "some basic events that have         The basic event mapping tables in the fire PRA will be been mapped to scenarios, components, or cables are not found in the          reviewed and compared to the present basic event CAFTA model." The associated disposition states that "information in the      mapping associated with each equipment or cable. Those mapping table should be reviewed to eliminate the extraneous information      items that are no longer needed will be removed and any and eliminate the discrepancies." The disposition for this F&O presented      incorrect m apping will be updated. The Fire PRA model in the PBNP LAR for transition to the National Fire Protection Association    will be quantified using this updated mapping table, as (NFPA) Standard 805 (ADAMS Accession Number ML13182A353)                     indicated in response to RAI 02.h.
indicates that this review has been performed which identified six failure events that were excluded from the PRA as sociated with the MCR. The PBNP NFPA 805 LAR for this disposition also states that a sensitivity study was conducted which determined that the exclusion of these basic events have a negligible impact on fire risk.

NRC 2018-0044 Attachment 1 Page 5 of 5 Modeling of the Reactor Coolant Pump (RCP) Shutdown Seals  Update the internal events and flre PRA models to credit the Westinghouse Generation III RCP seals using the guidance from PWROG-14001-P, Revision 1 and the limitations and conditions in the associated NRC's safety evaluation (ADAMS Accession Number ML17200A116), as stated in response to RAI 10. The additional failure contribution of the Westinghouse RCP Shutdown Seal Bypass failure mode will be added to the PRA. models, consistent with the limitations and conditions in the NRC safety evaluation for P~!R.OG-140001-P, Revision 1.
PRA maintenance versus PRA upgrade                        All changes performed to the PRA to address the above implementation items will be independently reviewed to determine if the resolution of those items in the PRA model constitutes a PRA upgrade. If the review identifies a change is a PRA upgrade, a focused-scope peer review will be performed for that change, and any resulting F&Os will be resolved in the PRA to meet Capability Category II.

NRC 2018-0044
                                                            .Attachment 2 Page 1 of 4 Attachment 2 Markup of the Point Beach Units 1 and 2 Operating Licenses

NRC 2018-0044 Attachment 2 Page 2 of 4 INSERT CONDITIONE E . Adoption of 10 CFR 50.69, "Risk-Informed categorization and treatment of structures, systems, and components for nuclear power plants"
: 1. NextEra Energy Point Beach 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) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (AN0-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e.,
seismic margin analysis (SMA) to evaluate seismic risk, and a screening of other external hazards updated using the external hazard screening significance process identified in ASlviE/ANS PRA Standard RA-Sa-2009; as specified in License Amendment No.
[NUMBER] dated [DATE] .
: 2. Prior to implementation of the provisions of 10 CFR 50.69, NextEra Energy Point Beach shall complete the items below:
: a. Item A in Attachment 1, List of Categorization Prerequisites, to NextEra Energy Point Beach letter NRC 2017-0043, "License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants,'" dated August 31, 2017; and
: b. Attachment 1, Point Beach 10 CFR 50.69 PRA Implementation Items, in NextEra Energy Point Beach letter NRC-2018-0044, "Supplement to Response to Request for Additional Information Regarding License Amendment Request 287, Application to Adopt 10 CFR 50.69, 'Risk informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants,'" dated September 28, 2018 .
: 3. 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).

NRC 2018-0044 Attachment 2 Page 3 of4 D. Physical Protection NextEra Energy Point Beach shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73 .55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled : "Point Beach Nuclear Plant Physical Security Plan , (Revision 4) ," submitted by letter dated May 10, 2006 . NextEra Energy Point Beach, LLC shall fully implement and maintain in effect all provisions of the Commission-approved Point Beach Nuclear Plant Cyber Security Plan (CSP),
including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The NextEra Energy Point Beach CSP was approved by License Amendment No. 243 as supplemented by a change approved by License Amendment No. 247 and License Amendment No. 252.
E. Deleted      +-<:.___-----JjiNSERT CONDITION E                 I F. NextEra Energy Point Beach Unit 1 shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated June 26, 2013, and supplements dated September 16,2013, July 29,2014, August 28,2014, September 25,2014, November 14,2014, December 19, 2014 , January 16,2015, May 12,2015, August 26,2015, February 22,2016, April 07,2016, and May 3, 2016 ,
and as approved in the safety evaluation report dated September 8, 2016 . Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c) , and provided no other regulation, technical specification , license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or license condition , and the criteria listed below are satisfied.
: 1. Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach , methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated , and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model , methods that have been approved by NRC
* through a plant-specific *license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
Renewed License No. DPR-24 Amendment No. ~

NRC 2018-0044 Attachment 2 Page 4 of 4 D. Physical Protection NextEra Energy Point Beach shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p ). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Point Beach Nuclear Plant Physical Security Plan, (Revision 4)," submitted by letter dated May 10, 2006. NextEra Energy Point Beach, LLC shall fully implement and maintain in effect all provisions of the Commission-approved Point Beach Nuclear Plant Cyber Security Plan (CSP),
including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The NextEra Energy Point Beach CSP was approved by License Amendment No. 247 as supplemented by a change approved by License Amendment No. 251 and License Amendment No. 256.
E. Deleted ..c...c-------,-iiNSERT CONDITION E                       I F. NextEra Energy Point Beach Unit 2 shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated June 26, 2013, and supplements dated September 16, 2013, July 29, 2014, August 28, 2014, September 25, 2014, November 14, 2014, December 19, 2014, January 16, 2015, May 12, 2015, August 26, 2015, February 22, 2016, April 07, 2016, and May 3, 2016 and as approved in the safety evaluation report dated September 8, 2016. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and
_ ~ pr_oyide_d_no__o_tber r..eguJation, technical specllic_CltLon,Jlcense_QQOdition_gr r~uirem~nL _
would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or license condition, and the criteria listed below are satisfied.
: 1. Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in ,NFPA 805 risk assessmt;lnts, or methods that have been demonstrated to bound the risk impact Renewed License No. DPR-27 Amendment No. ~}}