ML25240B584

Supplement to License Amendment Request for a Risk-Informed Resolution to GSI-191

ML25240B584
Person / Time
Site:	Wolf Creek
Issue date:	08/28/2025
From:	Bayer R Wolf Creek
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
001052
Download: ML25240B584 (1)

Text

P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 Robert J. Bayer Vice President Engineering August 28, 2025 001052 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Reference:

1)

Letter dated August 16, 2022, from Samson Lee, USNRC, to C.

Reasoner, WCNOC, Wolf Creek Generating Station, Unit 1 - Issuance of Amendment No. 233 RE: Removal of Table of Contents from the Technical Specifications (NRC ADAMS ML22199A294)

2)

Letter ET 24-000478, dated September 12, 2024, from M. T. Boyce, WCNOC, to USNRC, License Amendment Request for a Risk-Informed Resolution to GSI-191 (NRC ADAMS ML24260A071)

3)

Letter dated January 14, 2025, from Samson Lee, USNRC, to C.

Reasoner, WCNOC, Wolf Creek Generating Station, Unit 1 -

Regulatory Audit in Support of Review of License Amendment and Exemption Requests for Risk-Informed Resolution to Generic Safety Issue-191 (NRC ADAMS ML25010A369)

4)

Letter dated June 10, 2025, from Samson Lee, USNRC, to C.

Reasoner, WCNOC, Wolf Creek Generating Station, Unit 1 - Audit Summary for License Amendment Request and Exemption Request for a Risk-Informed Resolution to Generic Safety Issue-191 (NRC ADAMS ML25143A051)

Subject:

Docket No. 50-482: Supplement to License Amendment Request for a Risk-Informed Resolution to GSI-191 Commissioners and Staff:

By letter dated September 12, 2024 (Reference 2), Wolf Creek Nuclear Operating Corporation (WCNOC) submitted a license amendment request (LAR) for Wolf Creek Generating Station (WCGS). The proposed amendment potentially revises the WCGS licensing basis to allow the use of a risk-informed approach to address Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance.

As part of the LAR review process, by letter dated January 14, 2025 (Reference 3), the Nuclear Regulatory Commission (NRC) scheduled a regulatory audit with WCNOC staff.

001052 Page 2 of 3 Following the audit, the NRC issued an audit summary (Reference 4) which noted which responses need to be placed on the docket to support the NRC safety evaluation.

This letter is a supplement to the LAR in Reference 2.

Attachment I to this letter provides responses to audit questions posed by the NRC staff during the regulatory audit which required docketed information.

Attachment II provides mark-ups of a single Technical Specification (TS) page to be included with the mark-ups provided in the LAR in Reference 2. The mark-up is in response to NRC audit question STSB-Q1 regarding TSTF-567.

In addition to STSB-Q1 item, as noted on the cover page of Attachment II, the TS Table of Contents were inadvertently included in the original LAR application (Reference 2) and are now under licensee control as approved by Amendment 233 (Reference 1). With the TS Table of Contents now under licensee control, WCNOC requests that any changes involving the TS Table of Contents pages be withdrawn from NRC review.

Attachment III provides the proposed (clean) TS page for the added change. The proposed changes within Attachment II and III do not impact the other TS mark-ups and clean pages provided in the Reference 2 LAR.

The supplemental information provided in the Attachments do not impact the conclusions of the No Significant Hazards Consideration provided in Reference 2. In accordance with 10 CFR 50.91, Notice for public comment; State consultation, a copy of this supplement is being provided to the designated Kansas State official.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4015, or Dustin Hamman at (620) 364-4204.

Sincerely, Robert J. Bayer RJB/nwl Attachments I: Responses to NRC Audit Questions Related to the GSI-191 LAR II: Proposed Technical Specification Changes (Mark-Up)

III: Proposed Technical Specification Clean Pages cc:

A. N. Agrawal (NRC), w/a S. S. Lee (NRC), w/a J. Meinholdt (KDHE), w/a J. D. Monninger (NRC), w/a Senior Resident Inspector (NRC), w/a WCNOC Licensing Correspondence ET 25-001052, w/a

001052 Page 3 of 3 STATE OF KANSAS

)

) ss COUNTY OF COFFEY )

Robert J. Bayer, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

. Bayer Vice esident Engineering SUBSCRIBED and sworn to before me this 'Lf day of

< \\J..= u,.S '\\

Pubtlc

• State of lntment Expires Expiration Date l  l \\ * '20 (J,:'1

, 2025.

Attachment I to 001052 Page 1 of 29 Responses to NRC Audit Questions Related to the GSI-191 LAR

Attachment I to 001052 Page 2 of 29 RESPONSES TO NRC AUDIT QUESTIONS NOTE: The U.S. Nuclear Regulatory Commission (NRC) staffs audit questions, as provided in the audit summary dated June 10, 2025 (NRC ADAMS Accession No. ML25143A051), are in italics throughout this attachment to distinguish from the Wolf Creek Nuclear Operating Corporation (WCNOC) responses for Wolf Creek Generating Station (WCGS).

Attachment I Table of Contents NRC Question APLB-Q1 (Related to Audit Plan Initial Item No. 3):........................................... 3 NRC Question NVIB-Q1 (Related to Audit Plan Initial Item No. 2):............................................ 4 NRC Question NVIB-Q2 (Related to Audit Plan Initial Item No. 2):............................................ 5 NRC Question NVIB-Q4 (Related to Audit Plan Initial Item No. 2):............................................ 8 NRC Question NVIB-Q5:......................................................................................................... 10 NRC Question NVIB-Q6:......................................................................................................... 11 NRC Question STSB-Q1:........................................................................................................ 12 NRC Question STSB-Q2:........................................................................................................ 13 NRC Question STSB-Q3:........................................................................................................ 19 NRC Question STSB-Q12:...................................................................................................... 20 NRC Question STSB-Q13:...................................................................................................... 21 NRC Question STSB-Q14.b:................................................................................................... 22 NRC Question STSB-Q16:...................................................................................................... 23 NRC Question NCSG-Q1:....................................................................................................... 24 NRC Question NCSG-Q2:....................................................................................................... 25 NRC Question NCSG-Q3:....................................................................................................... 26 NRC Question APLC-Q1:........................................................................................................ 27 NRC Question APLC-Q2:........................................................................................................ 28 NRC Question APLC-Q5:........................................................................................................ 29

Attachment I to 001052 Page 3 of 29 NRC Question APLB-Q1 (Related to Audit Plan Initial Item No. 3):

The application uses the 25-year loss-of-coolant accident (LOCA) frequencies from NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the Elicitation Process, April 2008, rather than the 40-year values also provided in that document. The provided justification is that the Wolf Creek plant is not yet 40 years old and so the 25-year values are appropriate. However, the 40-year anniversary of the Wolf Creek Operating License will occur prior to the issuance of this license amendment, if the NRC staff determines that it is appropriate to issue it.

Discuss the effect that using the 40-year values would have on the results of the analysis, including on sensitivity studies included in the application.

Discuss whether this frequency change would move the proposed change over the threshold into Region II of Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions of Plant-Specific Changes to the Licensing Basis, figures of merit.

Discuss any plans to update the analysis once the 40-year anniversary is passed.

The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable but needs the response on the docket to finalize its regulatory finding.

WCNOC Response:

The following response was provided to the NRC during the audit:

Using the Geometric Mean 40-year LOCA frequency values gives a CDF value of 1.5E-06/yr and a LERF value of 4.2E-11/yr. The 40-year 5th and 95th percentiles give ranges from 6.7E-09/yr to 5.2E-06/yr for CDF and 1.9E-13/yr to 1.5E-10/yr for LERF.

The mean 40-year LOCA frequency would move the proposed changes slightly over the threshold into Region II of the Regulatory Guide 1.174 figures of merit. However, most of the parameters in the Wolf Creek GSI-191 evaluation are using either conservative or bounding values (see Figure 1 of Attachment VII to the submitted LAR), and the overall risk quantification results are very conservative. Since the sensitivity using the 40-year LOCA frequencies shows that the risk metrics are just barely in Region II, it is reasonable to conclude that the actual risk associated with the effects of debris is very small (i.e., a more refined assessment would show that it is still in Region III).

Wolf Creek entered its period of extended operation on March 11, 2025. The Wolf Creek GSI-191 evaluation uses either conservative or bounding values for most of the parameters resulting in very conservative risk quantification results. Wolf Creek does not intend to update the analysis even though the 40-year anniversary has passed.

Attachment I to 001052 Page 4 of 29 NRC Question NVIB-Q1 (Related to Audit Plan Initial Item No. 2):

Section 2.3.2, RCS [Reactor Coolant System] Pressure Boundary of attachment IX to ET 24-000478 (PDF page 304 of the LAR) states that Wolf Creek developed a program plan to manage the risk of Primary Water Stress Corrosion Cracking (PWSCC) degradation in Alloy 600 components and Alloy 82/182 welds. The plan is in accordance with 10 CFR 50.55a [Title 10 of the Code of Federal Regulations Section 50.55a], ASME Code [American Society of Mechanical Engineers Boiler and Pressure Vessel Code]Casesl N-722-1 ([Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials]) and N-770-2 ([Alternative Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler material With or Without Application of Listed Mitigation Activities]),

and NEI [Nuclear Energy Institute] 03-08 (Guideline for the Management of Material Issues, Revision 3). The plan identifies all Alloy 600/82/182 locations, ranks the locations based on their risks of developing PWSCC, provides inspection requirements, and presents mitigation/replacement options. Wolf Creek has either implemented or planned mitigation measures for the welds of concern. Periodic inspections of the Alloy 600 components and Alloy 82/182 welds are covered in the ISI [inservice inspection]

Program...

The licensee has uploaded the PWSCC degradation management plan in the ePortal. The NRC staff has reviewed the plan. The plan contains valuable information that the staff is seeking as shown in the audit question below. The staff needs the information as requested in the audit questions NVIB-Q2, Q3, and Q4 below to address the PWSCC degradation management issue.

Two options.

Option 1 - the licensee submits to the NRC its PWSCC management plan (the document) on the docket. In this scenario, the NRC staff would not ask audit questions NVIB-Q2, Q3, and Q4 regarding PWSCC in the request for additional information (RAI) questions because the staff could use the plan on the docket to address the PWSCC issue in the safety evaluation (SE).

Option 2 -The licensee does not submit the PWSCC management plan (the document) on the docket. In this scenario, the NRC staff will officially issue the PWSCC-related RAI questions as shown in the audit questions. The staff will use the licensees response to the RAI questions to evaluate the PWSCC issue in the SE.

The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding for Question NVIB-Q2.

WCNOC Response:

WCNOC selected Option 2 during the audit. As a result, responses to audit questions NVIB-Q2 and NVIB-Q4 are provided below. A response to audit question NVIB-Q3 was not requested.

Attachment I to 001052 Page 5 of 29 NRC Question NVIB-Q2 (Related to Audit Plan Initial Item No. 2):

(a) Provide/upload a list (e.g., a table) of and/or documents related to in-scope welds with associated piping system that are fabricated with nickel-based Alloy 82/182 material.

(b) On the list, identify in-scope Alloy 82/182 welds that have and have not been mitigated to reduce their susceptibility to stress corrosion cracking.

(c) Discuss whether the unmitigated in-scope Alloy 82/182 welds will be mitigated in the future.

(d) For those unmitigated in-scope Alloy 82/182 welds, discuss how they are being inspected (e.g., inspection frequency and method) to monitor ensure their structural integrity. (f)

Upload in the portal the NRC-approved relief requests regarding actions on Alloy 82/182 welds, if any.

The NRC staff reviewed the documents provided in response to Audit Plan Initial Item No.

2. The NRC staff requires additional information regarding the table notation in the portal response. This information should be provided on the docket along with the information provided in response to Audit Plan Initial Item 2.

WCNOC Response:

The following response was provided to the NRC during the audit, with additional information as requested:

(a), (b)

The table below contains the in-scope welds fabricated with Alloy 82/182 material. The table is arranged by the relative risk ranking for locations in the first column. The ranking has been replaced with Mitigated / Repaired if the location has been mitigated or repaired. These rankings were determined as part of a study conducted by a vendor for Wolf Creek and reflect conclusions based on Wolf Creek Data.

Rank Component / Location In-Service Cracking Material Pressure Boundary Mitigated/

Repaired Pzr Safety and Relief Nozzle Safe-End Weld Yes (1) 82/182 Yes Mitigated/

Repaired Pzr Surge Line Nozzle Safe-End Weld Yes (1) 82/182 Yes Mitigated/

Repaired Pzr Spray Nozzle Safe-End Weld No 82/182 Yes Mitigated/

Repaired RV Outlet Nozzle Safe-End -

Hot Leg Yes (2) 82/182 Yes Mitigated/

Repaired CRDM Nozzle and Nozzle Weld Yes 82/182 Yes Mitigated/

Repaired RV Bottom Mounted Nozzle and Weld Yes (2) 600 Yes

Attachment I to 001052 Page 6 of 29 Rank Component / Location In-Service Cracking Material Pressure Boundary Mitigated/

Repaired Head Vent Penetration Weld No 82/182 Yes Mitigated/

Repaired RV BMN Yes (2) 82/182 Yes Mitigated/

Repaired RV BMN to Guide Tube Weld No 82/182 Yes Mitigated/

Repaired RV Inlet Nozzle Safe-End (Cold Leg) Weld No 82/182 Yes 35 RV Head Vent to Elbow Weld No 82/182 Yes 36 RV Head Vent Elbow to Piping Weld No 82/182 Yes 37 RV Head Vent Pipe to SS Elbow No 82/182 Yes 38 CRDM to Flange Weld No 82/182 Yes 39 RV Head Vent Nozzle Elbow No 600 Yes 40 RV Head Vent Horizontal Pipe No 600 Yes Unranked RCS Hot Leg Thermowells No 600 Yes Unranked RCS Cold Leg Thermowells No 600 Yes (1) During Volumetric Examination performed prior to Structural Weld Overlay, indications were indicative of PWSCC.

(2) Industry Operating Experience of In-Service Cracking, not detected at Wolf Creek.

(c)

The long term mitigation plan for Alloy 600/82/182 materials is to replace the RCS hot and cold leg thermowells (Alloy 600) and associated fillet welds (Alloy 81/182) in a future outage that is yet to be determined. Currently no other locations have been selected for mitigation.

(d)

Location Exam Type Frequency Source RV Head Vent to Elbow Weld VT-2 (3)

Each Refueling ASME Section XI RV Head Vent Elbow to Piping Weld VT-2 (3)

Each Refueling ASME Section XI RV Head Vent Pipe to SS Elbow VT-2 (3)

Each Refueling ASME Section XI

Attachment I to 001052 Page 7 of 29 Location Exam Type Frequency Source CRDM to Adapter Weld VT-2 (3) (7)

Each Refueling ASME Section XI RV Head Vent Nozzle Elbow VT-2 (3)

Each Refueling ASME Section XI RV Head Vent Horizontal Pipe VT-2 (3)

Each Refueling ASME Section XI RCS Hot Leg Thermowells Visual (VE)

Each Refueling CC N-722-1 RCS Cold Leg Thermowells Visual (VE) (8)

Once per ISI Interval CC N-722-1 NOTES (1) Not required during outages when volumetric examination being performed (2) Visual examination on condition of welds, cladding, and plates performed during entry into S/G for eddy current testing of tubes (see AP 29A-003)

(3) Examined per criteria of ASME Section XI (4) Preemptive Structural Weld Overlays (SWOLs) immediately go into a sampling pool where 25% of this group must be volumetrically inspected each ISI interval.

(5) Repairs via SWOL must be inspected within the next two refueling cycles and if no crack initiation or crack growth is detected go into a sampling pool where 25% of this group must be volumetrically inspected each ISI interval.

(6) Regulation 10 CFR 50.55a requires that Class 1 DM welds be examined in accordance with Code Case N-770-5. This includes welds mitigated by Weld Overlay (WOL). The examinations performed in accordance with I3R-05 also meet the examination requirements of Code Case N-770-5, Items C1 and F1, as applicable.

(7) Volumetric Exam performed of periphery population once per ISI Interval in accordance with ASME Section XI.

(8) Visual (VE) of Cold Leg thermowells performed each refueling outage concurrent with Hot Leg thermowells.

(e) - Not used (f)

I3R-05 Installation and Examination of Full Structural Weld Overlays for Repairing / Mitigating Pressurizer Nozzle-to-Safe End Dissimilar Metal Welds and Adjacent Safe End-to-Piping Stainless Steel Welds.

Attachment I to 001052 Page 8 of 29 NRC Question NVIB-Q4 (Related to Audit Plan Initial Item No. 2):

(a) If the risk analysis contains welds that contain flaws that have not been repaired, discuss whether the LOCA frequency for the degraded welds in the risk analysis is increased from that of NUREG-1829 estimates. If the LOCA frequency is not increased in the analysis, provide justification.

(b) Discuss whether the failure probability value in the risk analysis is increased to account for the in-scope unmitigated nickel-based Alloy 600/82/182 components (welds and pipe components). If a higher probability of failure was not used in the risk analysis, provide justification.

The NRC staff reviewed the documents provided on the portal. In addition to the portal response, the NRC staff requires additional information regarding the break sizes (list all breaks smaller and larger than the 10-inch threshold break size) that were evaluated in the risk-informed analysis. This information should be provided on the docket.

WCNOC Response:

The following response was provided to the NRC during the audit:

(a) The risk analysis does not contain welds with flaws that have not been repaired.

(b) The Wolf Creek GSI-191 evaluation did not include any adjustment of LOCA frequencies or weld failure probabilities based on weld-specific flaws or degradation mechanisms.

Early in the risk-informed GSI-191 pilot project, South Texas Project (STP) had intended to estimate the LOCA frequencies at individual welds based on weld-specific degradation mechanisms (Reference 1). This was referred to as the bottom-up approach for estimating overall LOCA frequency. The NRC staff did not agree with this approach because the total LOCA frequency would not add up to the LOCA frequencies provided in NUREG-1829 (Reference 2). To address the staffs concerns, STP pursued two alternative methods (Reference 3):

The top-down approach essentially allocated the LOCA frequencies from NUREG-1829 equally across welds (as long as the weld was large enough for a given break size to occur).

The hybrid approach allocated the NUREG-1829 frequencies across welds based on the weld-specific degradation mechanisms (i.e., the bottom-up frequencies were used to determine relative probabilities of breaks at specific welds but were normalized to add up to the NUREG-1829 frequencies).

In the final RoverD assessment (Reference 4), STP only used the top-down approach.

This assessment was accepted by the NRC (Reference 5).

For Wolf Creek, allocation of LOCA frequencies to individual welds was not necessary based on the conservative approach taken. As described in the LAR submittal (Reference 6), Wolf Creek is implementing a simplified risk-informed approach where all breaks equal to or greater than the threshold break size are conservatively assumed to fail. The threshold break size was determined to be 10 inches. Therefore, breaks in all pipes that are at least 10-inch diameter are assumed to cause debris-related failures, and the interpolated frequency from NUREG-1829 for a 10-inch break can be used directly for

Attachment I to 001052 Page 9 of 29 the risk quantification without specifying the relative contribution from each weld that is 10+ inches.

References:

1. Fleming, Karl, and Bengt Lydell, LOCA Frequencies for GSI-191 Applications, ANS PSA 2013 International Topical Meeting on Probabilistic Safety Assessment and Analysis, Columbia, SC, September 22-26, 2013.

2. NUREG-1829 Volume 1, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, April 2008.

3. NOC-AE-13003043, South Texas Project Units 1 and 2 Docket Nos. STN 50-498 and STN 50-499, Supplement 1 to Revised STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Resolving Generic Safety Issue (GSI)-191, November 13, 2013, ML13323A183.

4. NOC-AE-16003401, South Texas Project Units 1 and 2 Docket Nos. STN 50-498 and STN 50-499 Supplement 3 to Revised STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Address Generic Safety Issue (GSI)-191 and Respond to Generic Letter (GL) 2004-02, October 20, 2016, ML16302A015.

5. South Texas Project, Units 1 and 2 - Issuance of Amendments Re: Changes to Design Basis Accident Analysis Using a Risk-Informed Methodology to Account for Debris in Containment, July 11, 2017, ML17038A223.

6. Docket No. 50-482: License Amendment Request for a Risk-Informed Resolution to GSI-191, September 12, 2024, ML24260A071.

Additionally, the following list contains the break sizes, in inches, that were evaluated in the risk-informed analysis.

0.375, 0.5, 1.338, 1.689, 2, 2.125, 2.626, 3.438, 4, 5.189, 6, 8, 8.75, 10, 10.5, 11.188, 11.5, 12, 14, 17, 20, 23, 26, 27.5, 29, and 31 inches.

Attachment I to 001052 Page 10 of 29 NRC Question NVIB-Q5:

The NRC staff notes that as part of plant operating procedures, pressurized water reactor owners inspect RCS piping and associated components in addition to the inspections performed per NRC regulations or the ASME Code,Section XI, such as operator walkdowns, opportunistic inspections, the boric acid corrosion program, and the fatigue monitoring program per Materials Reliability Program (MRP)-146, Revision 1, Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines. Discuss any administrative or opportunistic inspections at the Wolf Creek plant that monitor the structural integrity of the in-scope piping and components in addition to the NRC regulations.

The licensee provided a response to this question on the portal. The NRC staff reviewed the information and concluded that additional information regarding the boric acid control program, leak inspection program, and other opportunistic inspections regarding potential reactor coolant pressure boundary leakage be included on the docket.

WCNOC Response:

The following is a revised response from what was provided to the NRC during the audit:

Wolf Creek developed and administered an Outside Diameter Stress Corrosion Cracking program as a result of industry and internal operating experience. The program has evolved since its inception and has been credited for satisfying subsequent license renewal inspections. It is limited to areas under pipe clamps on stainless steel pipes within ASME Class piping. Operating experience at Wolf Creek identified Outside Diameter Stress Corrosion Cracking (ODSCC) on the Auxiliary Spray piping during Refueling 17. During the extent of condition, other locations were identified as being susceptible to ODSCC. This procedure is to perform periodic Dye Penetrant (PT) inspections in accordance with QCP-20-501, VISUAL DYE PENETRANT EXAMINATIONS, of piping/locations identified as most likely susceptible to ODSCC based on Operating Experience. Activities identified in this procedure are credited with the performance of CLR-11, External Surfaces Monitoring of Mechanical Components, aging management activities required by license renewal.

In addition to the ODSCC inspections, Wolf Creek performs visual inspection of the cold leg RTD Thermowells on the same frequency as the hot leg Thermowells instead of once per ISI Interval. ASME Section XI Pressure Test Program monitors the RCS pressure boundary for external leakage by VT-2 qualified examiners during every refuel and accessible class 2 and 3 systems during online operations. The Boric Acid Corrosion Control program is implemented to prevent and manage corrosion caused by borated water leaks by focusing on early detection. Similar to the pressure test examinations, walkdowns are performed periodically to detect leakage for borated systems. Operators and other plant personnel are instructed to report plant equipment leaks or potential leaks in the corrective action program. These are observed during operator walkdowns, general plant maintenance and other activities that occur daily. Operations also frequently monitors and trend RCS leakage to aid in leak detection both internally and externally.

Attachment I to 001052 Page 11 of 29 NRC Question NVIB-Q6:

Nuclear plants have a leakage detection system to monitor leakage from the RCS. The RCS leakage detection system is usually designed in accordance with RG 1.45, Revision 1, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, to provide time for appropriate operator action to identify and address RCS leakage. However, based on operating experience, the RCS leakage detection system in some nuclear plants does not provide the same capability as originally designed. Discuss (a) any changes made to the current RCS leakage detection system that are different from the description in the Final Safety Analysis Report, (b) whether the capability of the sensors and instrumentation of the current RCS leakage detection system is consistent with RG 1.45, Revision 1, and (c) whether the diversity, redundancy, and reliability of the current detection instrumentations still satisfy the guidance in RG 1.45, Revision 1.

The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.

WCNOC Response:

The following response was provided to the NRC during the audit:

a) No changes have been made that created a difference from the USAR description.

b) The equipment utilized for the purpose of RCS leakage detection system is capable and consistent with RG 1.45, Rev 1. Note that Wolf Creek is committed to RG 1.45, Rev. 0.

c) Yes, the equipment still satisfies the guidance in RG 1.45, Rev 1. Note that Wolf Creek is committed to RG 1.45, Rev. 0.

Attachment I to 001052 Page 12 of 29 NRC Question STSB-Q1:

For the adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-567, Add Containment Sump TS to Address GSI-191 Issues, it appears that TS 5.5.15, Safety Function Determination Program, should be revised per the traveler. This was not included in the LAR or addressed as a variation to the traveler. Provide a supplement to include the change or explain why the change is not needed.

The licensee stated that it would provide a supplement to the LAR that will incorporate the change into its TS. The NRC staff concluded that this would provide an acceptable resolution to this issue.

WCNOC Response:

WCNOC proposes to revise the WCGS Safety Function Determination Program (TS 5.5.15) to clarify its application when a supported system is made inoperable by the inoperability of a single Technical Specification (TS) support system. Attachment II of this supplement contains the proposed mark-up TS 5.5.15. Attachment III of this supplement contains the revised (clean) TS page. These TS changes are in addition to those provided in the LAR, and do not affect any previously provided TS mark-up and revised (clean) pages of the LAR. This change, along with the added information described below, is in accordance with TSTF-567 Revision 1 without variation.

The following information is to be added to the first standard of the No Significant Hazards Consideration of Attachment I of the LAR (Page 13 of 17):

The proposed change clarifies the Safety Function Determination Program when a supported system is made inoperable by the inoperability of a single Technical Specification support system. The Safety Function Determination Program directs the appropriate use of TS actions and the proposed change does not alter the current intent of the TS. The actions taken when a system is inoperable are not an assumption in the initiation or mitigation of any previously evaluated accident.

The following information is to be added to the second standard of the No Significant Hazards Consideration of Attachment I of the LAR (Pages 14 and 15 of 17):

The proposed change clarifies the Safety Function Determination Program when a supported system is made inoperable by the inoperability of a single Technical Specification support system. The Safety Function Determination Program directs the appropriate use of TS actions and the proposed change does not alter the current intent of the TS. The proposed change to the Safety Function Determination Program will not result in any change to the design or design function of the containment sump or a method of operation of the plant.

No additions are made to the third standard. The finding of no significant hazards consideration is unaffected by these additions.

Attachment I to 001052 Page 13 of 29 NRC Question STSB-Q2:

Table B 3.6.8-1 in attachment V of the LAR (PDF page 59) lists the containment sump debris limits for breaks that are considered to pass the deterministic criteria. The table values are also explained starting on PDF page 315. Discuss how these values were developed. The bases for the fiber values were not clear and did not appear consistent with the other values. This may also be related to another question regarding the information on PDF page 316 of the LAR.

a. The title implies that these are containment sump debris limits. Would strainer limits be a more accurate description based on the notes for the table?

b. Should the Thermolag particulate value in the table be increased to bound or equal the transported value with a subsequent reduction in coatings margin. This might be a more realistic representation in design basis amounts and eliminate a footnote. The value in the Bases is slightly exceeded by the current plant condition as discussed on page 316, footnote 4. Alternately, should the TS table provide an explanation on this?

c. The fiber fines value in the bases table appears to exceed the full debris load (FDL) tested amount by about 3 pounds (141.78 tested and 144.6 in the table). The fine fiber limit value in the Bases table is compared to a quantity for two train transport (in-vessel analysis?)

on PDF page 315. To reflect the lowest fine fiber amount that can be accommodated by the strainer, should the strainer head loss test debris amount be used?

d. Provide an explanation of what debris sizes and amounts the total fiber value includes and how it will be used in an operability analysis. It appears to be the same total fiber value that was included in the FDL test. The footnote for the Bases table states that it is the maximum amount allowed to transport to a single strainer. The footnote on PDF page 315 states that this is the amount included in the full debris load test. Is erosion considered in this value?

How, during an operability assessment, would it be determined whether discovered fiber exceeds this or the fine fiber limit? If it is determined that the fine fiber limit, including erosion, is not exceeded and all additional fiber plus the fine fiber are less than the total fiber limit that operability would be demonstrated so this value is likely valid. It is just difficult to determine how it would be used.

e. How was the value for degraded paint chips determined and how is the condition assessed if additional chips are identified in the plant?

f.

The footnotes for the Bases table provide some insight into the above, but it is not clear how someone referring to the table would be able to determine whether a plant condition is bounded by the limits or easily determine whether this is the case or find the basis for the values.

The licensee provided a response to this question on the portal and the NRC staff reviewed the information. The information provided adequate clarification for some of the issues identified in this question, but the NRC identified issues that required additional discussion. During the audit discussions, the licensee provided an updated concept for determining operability and margin values for the various types of debris, particularly fibrous debris. The NRC requires this information to be provided as a supplement to the LAR. This response also covers questions related to STSB-Q14b and Q16.

WCNOC Response:

The following is revised from the response provided to the NRC during the audit in order to include the requested clarification regarding determination of operability and margin values:

Attachment I to 001052 Page 14 of 29

a. No. The title Containment Sump Debris Limits is appropriate since the new TS 3.6.8 is specific to the containment sump. Note that the strainer debris limit is used instead of the in-vessel debris limit. However, these debris limits are applicable to both strainer and in-vessel failure mechanisms.

b. The ThermoLag particulate value is the debris limit based on the debris quantities used during the full debris load (FDL) head loss test given in Table 3.f.7-1. As it is based on the tested value, it cannot be increased to bound the transported value.

However, the transported value contains conservatisms, and it is reasonable to round the results for both the tested and transported quantities to 0.5 ft3 to eliminate the footnote.

c. Yes. The quantities of debris generated and transported to a single ECCS strainer for the range of potential break locations, sizes, and orientations are provided in the debris transport quantity summary calculation for both single train and two train operation. The fiberglass fines debris quantity included both generated-as-fines and erosion fines. For single train operation, erosion fines were only calculated for small and large pieces of fiber that do not transport to the strainer (based on the fact that erosion would also occur for small pieces added to the head loss test tank). The debris transport evaluation was revised to remove unnecessary conservatism in the two train cases. Because the strainer fine fiber debris limit is more limiting than the in-vessel fine fiber debris limit, and single train operation is more conservative than two train operation for strainer head loss, the single train results and strainer debris limits are used to determine available margin.

Based on the fiber fines strainer debris limit of 141.8 lbm and a maximum transported fiber fines quantity of 97.9 lbm (106.2 lbm as shown in the figure below minus 8.3 lbm of latent fiber margin) for break sizes 10 inches during single-train operation, the resulting available margin is 43.9 lbm.

d. The total fiber debris limit includes fiberglass fines (either generated by the break or through erosion), small pieces, large pieces, latent fiber, and ThermoLag fiber. The

Attachment I to 001052 Page 15 of 29 value is based on the FDL test quantity.

The debris transport calculation was updated to account for the removal of unnecessary conservatism. The following results in the tables presented below were updated to reflect this approach.

Table 3.e.6-8, the Overall Transport Fractions for LBLOCA with Sprays On in the Steam Generator Compartment is shown below:

The Overall Transport Fractions for LBLOCA with Sprays On in the Pressurizer Compartment is shown below:

The Overall Transport Fractions for LBLOCA with Sprays On in the Annulus Compartment is shown below:

Attachment I to 001052 Page 16 of 29 If a new source of fiber debris is identified in containment, an operability assessment would be required. It would be necessary to analyze the debris quantity, size distribution, and transport (including erosion). The fiber fines limit is important for both strainer head loss and in-vessel effects. However, it is also necessary to consider the total quantity of fiber debris that would be transported and compare this to the total fiber limit. If either the fines quantity or the total fiber quantity exceed the available margin associated with these limits, the sump would be declared inoperable and the requirements of TS 3.6.8 would be followed.

e. The paint chip debris limit is based on the tested quantity of paint chips. Paint chip quantities are derived from the quantity of degraded qualified epoxy coatings inside containment. This is currently 0 ft2 since walkdowns have not identified any degraded qualified epoxy coatings inside containment. However, if a paint chip debris source term is identified in the future, the transportable quantity can be assessed against the paint chip debris limit.

f. The table in the Technical Specifications (TS) Bases provides the debris limits. Note that the debris limits were updated based on the changes discussed in response to Parts b and c. The revised TS Bases table is shown below:

Attachment I to 001052 Page 17 of 29 Table B 3.6.8-1 Containment Sump Debris Limits for Breaks 10 inches Debris Type Debris Limit Fiber Fines*

144.6141.8 lbm Total Fiber Fines, Small Pieces, and Large Pieces**

322.5 lbm Latent Particulate**

122.2 lbm ThermoLag Particulate**

0.50 ft3 Coatings Particulate**

2.43 ft3 Degraded Paint Chips**

158.4 ft2 Miscellaneous Debris (Tags, Labels, etc.)**

20.0 ft2

• Transportable fine fiber debris in the pool that is available to transport to either strainer during single train operation or split between both strainers during two train operation.

• • Maximum quantity of debris allowed to transport to a single strainer.

However, the available margin is maintained outside of the TS Bases document. The following table is an update of Table 2 in LAR Attachment X that incorporates the changes discussed in response to Parts b and c. This table clearly shows the margin value for each type of debris. A newly identified source of debris can be evaluated against the relevant margin to determine whether the total quantity exceeds the debris limit.

Table 1: Debris Margins for Breaks 10 Inches Debris Type Current Quantity Limit Available Margin Fiber Fines (lbm) 119.6 97.9(1) 144.6 141.8 25.0 43.9 Total Fiber Fines, Small Pieces, and Large Pieces (lbm) 235.8(2) 322.5 86.7 Latent Particulate (lbm) 54.2(3) 122.2 68.0 ThermoLag Particulate (ft3) 0.51 0.50 0

Coatings Particulate (ft3) 1.67(5)(4) 2.43 0.76 Degraded Paint Chips (ft2) 0 158.4 158.4 Miscellaneous Debris (ft2) 7.1 20.0 12.9 (1) This is the maximum transported fiber fine quantity for breaks up to 10 inches during single train operation with the built-in margin for transported latent fiber subtracted.

(2) This is the maximum transported total fiber quantity for breaks up to 10 inches during single-train operation (see BB-01-S105-04 listed in Table 3.e.6-10 of Attachment 8) with the built-in margin for transported latent fiber subtracted.

(3) This is the maximum transported latent particulate debris during single-train operation without margin based on generated latent debris load (75 lbm x 85%, see the Response to 3.d.3 in Attachment 8) and a transport fraction of 85% (see the Response to 3.e.6 in Attachment 8).

(4) This is the maximum transported coatings debris load for breaks up to 10 inches (see Table 3.f.7-1 of ).

Attachment I to 001052 Page 18 of 29

Reference:

1. ET 24-000478, Wolf Creek Nuclear Operating Corporation, Docket No. 50-482:

License Amendment Request for a Risk-Informed Resolution to GSI-191, September 12, 2024, (ADAMS Accession No. ML24260A071)

Attachment I to 001052 Page 19 of 29 NRC Question STSB-Q3:

In attachment VII of the LAR (PDF page 96), it is stated that control rod drive mechanism (CRDM) ejections and bottom head instrument penetration breaks are excluded as debris generation events due to the location of the breaks and the direction of any jet from these locations. The discussions in sections 3.a and 3.b of attachment VIII of the LAR (PDF pages 132-143 do not elaborate on why the CRDM housings and instrument penetrations are excluded or describe the insulation materials in the vicinity of the potential breaks. Discuss the basis for the elimination of these breaks. For these break locations, the NRC is primarily concerned with problematic insulation types that are not bounded by the current debris generation evaluation.

The licensee provided clarifying information on the portal. The NRC concluded that the break analysis is acceptable. However, the NRC requested that the licensee provide the type of insulation on the reactor vessel head on the docket.

WCNOC Response:

The reactor vessel top head has NUKON blanket insulation.

Attachment I to 001052 Page 20 of 29 NRC Question STSB-Q12:

In attachment VIII of the LAR (PDF page 234), the weight of debris accumulated on the strainer is discussed as an input for the structural calculation. Provide the weight of debris assumed in the analysis.

The licensee provided a response to this question on the portal that included the debris amount that collects on each strainer module. The NRC staff concluded that the response was acceptable and requested that the information be provided on the docket.

WCNOC Response:

The following response was provided to the NRC during the audit:

The weight of debris accumulated on the strainer is 63 lbm for the 11-disk module and 39.6 lbm for the 7-disk module.

Attachment I to 001052 Page 21 of 29 NRC Question STSB-Q13:

In attachment VIII of the LAR (PDF page 236), the NRC was unsure how to interpret the last sentence of the second bullet in section 3.k.2. Clarify whether the structural analysis uses the greater low temperature differential pressure (dP) of 5.5 feet for all cases or if the reduced dP of 4.0 feet is used for some cases. Both values are listed on PDF page 235. Was the thermal stress at 268 degrees Fahrenheit used for all cases?

The licensee provided a response to this question on the portal. The NRC staff reviewed the question and requested additional clarification during the audit discussions. The clarifications provided the information necessary for the NRC to understand how the calculations were performed. The NRC staff requested that the material properties used for the low and high temperature structural cases be provided on the docket. This information could be added to the second bullet on PDF page 236 of the LAR (section 3.k.2).

WCNOC Response:

The material properties used for the low and high temperature structural cases are provided in the tables below.

Component Material Types Component Material Type Perforated Plate Stainless Steel ASTM A-240, Type 304 Core Tube Stainless Steel ASTM A-240, Type 304 Radial Stiffeners Stainless Steel ASTM A-240, Type 304 Wire Stiffeners Stainless Steel ASTM A-493, Type 304 (Drafted to a higher tensile strength)

Tension Rods Stainless Steel ASTM A-276, Type 304, Condition B Structural Shapes Stainless Steel ASTM A-276, Type 304 Bolting Stainless Steel ASTM A-193 Grade B8, Class 2 Material Properties The strainer and plenum members Material Properties (Material Properties of Type 304 Austenitic Steels)

Cold Case Hot Case Modulus of Elasticity, Es (ksi) 27730 27200 Yield Strength, Sy (ksi) 25.85 23.2 Ultimate Strength, Su (ksi) 72 67.7 ASME Allowable Stress, S (ksi) 20 19.25 Note: These properties are conservatively applied for the Type 304 wire stiffeners which are drafted to a higher tensile strength than standard Type 304 stainless steels.

Bolting Materials (A-193, Grade B8, Class II) and Tension Rods (A-276, type 304, Condition B)

Cold Case Hot Case Yield strength, Sytr (ksi) 86.2 77.3 Ultimate Strength, Sutr (ksi) 120 112.8

Attachment I to 001052 Page 22 of 29 NRC Question STSB-Q14.b:

b. For the evaluation of fiber inside the reactor vessel on PDF page 260, what is the fiber amount assumed in the pool at the beginning of the calculation, including latent fiber? Is all fine fiber assumed. If small or large fiber is assumed, how is erosion treated?

The licensee provided a response to this question on the portal. The NRC staff discussed the response with the licensee. This issue is related to questions STSB-Q2. The licensee will provide information on the docket that will address this issue along with the response to STSB-Q2.

WCNOC Response:

The following response was provided to the NRC during the audit:

The initial sump fiber load used for the in-vessel evaluation was assumed to be 100%

fines and the quantity was assumed to be the same as the fiber fines quantity from the Full Debris Load (FDL) test, 141.78 lbm.

Reference:

1. ET 24-000478, Wolf Creek Nuclear Operating Corporation, Docket No. 50-482:

License Amendment Request for a Risk-Informed Resolution to GSI-191, September 12, 2024, (ADAMS Accession No. ML24260A071)

As discussed in the response to STSB-Q2, the strainer debris limit is more limiting than the in-vessel debris limit. The overall assessment was simplified by using just the strainer debris limit compared to the transportable quantity of fine fiber debris.

Attachment I to 001052 Page 23 of 29 NRC Question STSB-Q16:

The NRC could not verify all the values in the debris margins table (table 2 in attachment X (PDF page 316). The NRC also did not understand all the footnotes for the table.

a. The value for total fiber fines, small pieces, and large pieces appears to include small and fine Nukon fiber and latent fiber. Does this value include ThermoLag fiber? The footnote (2) for this value also states that the built-in margin for latent fiber is subtracted. It is not clear what this means.

b. Should the ThermoLag particulate value in the table be increased to bound or equal the transported value with a subsequent reduction in coatings margin. This might be a more realistic representation in design basis amounts and eliminate a footnote. The current value is slightly greater than the limit in footnote 4. It might be better to use some coatings margin to offset the slight negative margin in this value. (Question also asked related to the TS Bases table on PDF page 59.).

c. The limit values for fine and total fiber were questioned with respect to the TS Bases table on PDF page 59.

The licensee provided a response to this question on the portal. The NRC staff discussed the response with the licensee. This issue is related to questions STSB-Q2. The licensee will provide information on the docket that will address this issue along with the response to STSB-Q2.

WCNOC Response:

The following response was provided to the NRC during the audit:

a. Yes, this value includes Nukon, latent fiber, and ThermoLag fiber. As discussed in the LAR response to 3.d.3 (Reference 1), the five latent debris walkdowns completed at Wolf Creek indicated a maximum of 75 lbm of latent debris. However, a value of 140 lbm was assumed in the debris generation analysis. Therefore, there is 65 lbm of margin in the generated quantity of latent debris. Given an 85% transport fraction for latent debris (see LAR response to 3.e.6) and an 85/15 split between latent particulate and latent fiber (see LAR response to 3.d.3), this is equivalent to 47.0 lbm of margin in the transportable latent particulate quantity and 8.3 lbm of margin in the transportable latent fiber fines quantity. This margin was subtracted from the analyzed maximum transported quantity to properly reflect the available margin in LAR Attachment X Table 2.

b. See response to STSB-Q2 Part b.

c. See response to STSB-Q2 Part c.

Reference:

1. ET 24-000478, Wolf Creek Nuclear Operating Corporation, Docket No. 50-482:

License Amendment Request for a Risk-Informed Resolution to GSI-191, September 12, 2024, (ADAMS Accession No. ML24260A071).

Additional details are provided in the updated response to STSB-Q2.

Attachment I to 001052 Page 24 of 29 NRC Question NCSG-Q1:

Table B.3.6.8-1 (PDF page 59 of the LAR) identifies containment sump debris limits for breaks less than or equal to 10 inches. Aluminum is not included in this table. How is aluminum tracked to ensure assumed analysis amounts are not exceeded?

The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.

WCNOC Response:

The following response was provided to the NRC during the audit:

WCGS has implemented procedures and programs for monitoring, controlling, and assessing changes to the quantity of aluminum in containment. The WCGS design change process requires the use of Design Attribute Review (DAR) forms to identify potential impact on long term core cooling by the proposed modification. The DAR forms include the question Does the modification change the amount of exposed aluminum and/or zinc in containment? (Reference 1, Attachment I, Page 10). Changes to the aluminum quantity in containment are then addressed by reviewing the appropriate design calculations to ensure that the aluminum quantity, the resulting aluminum chemical debris quantity, and resulting aluminum solubility (temperature and timing) are within the tested and analyzed limits.

Reference:

1. ET 24-000478, Wolf Creek Nuclear Operating Corporation, Docket No. 50-482:

License Amendment Request for a Risk-Informed Resolution to GSI-191, September 12, 2024, (ADAMS Accession No. ML24260A071).

Attachment I to 001052 Page 25 of 29 NRC Question NCSG-Q2:

Please clarify the minimum containment sump pool pH. In attachment VIII of the LAR (PDF pages 263 and 274, the licensee states that the Wolf Creek minimum final containment pool pH is 8.78.

Table 3.n.1-2 on PDF page 265 indicates the minimum sump pH (long-term) is 8.5. If the appropriate value is 8.5, discuss how that would affect the precipitation temperatures that were determined for the sump strainer and the reactor vessel.

The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.

WCNOC Response:

The following response was provided to the NRC during the audit:

The minimum final containment pool pH is 8.78 at a reference temperature of 25°C (Reference 1, Attachment VIII, Page 151). The pH at a reference temperature of 25°C is appropriate for use for comparisons with testing and for use in correlations developed from testing where the pH is measured at a reference temperature of 25°C. A conservatively lower pH of 8.5 at 200°F is used to compare to the WCAP-17788 test parameters; however, a pH of 8.78 at a reference temperature of 25°C would also be appropriate for this purpose.

Reference:

1. ET 24-000478, Wolf Creek Nuclear Operating Corporation, Docket No. 50-482:

License Amendment Request for a Risk-Informed Resolution to GSI-191, September 12, 2024, (ADAMS Accession No. ML24260A071).

Attachment I to 001052 Page 26 of 29 NRC Question NCSG-Q3:

Do WCAP-17788-P tests 07-05 and IBOB 07-05 have test conditions that are representative for the Wolf Creek post-LOCA conditions? If so, what is the reason they are not referenced in the LAR?

The licensee provided a response to this issue on the portal. The NRC discussed this issue with the licensee during the audit call and requested clarification on the amounts of e-glass included in the tests. The licensee responded that the amounts of e-glass that would remain in the containment were not fully quantified at the time of the tests. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.

WCNOC Response:

The following response was provided to the NRC during the audit:

The Group 9 and Group 15 tests are referenced to address the full range of sump pH values while bounding the aluminum generating materials (Reference 1, Attachment VIII, Page 142). Tests 07-05 and IBOB 07-05 can be considered representative, best-estimate tests at a low, solubility minimizing pH; however, they were not referenced in the LAR because these tests did not conservatively bound the aluminum and E-Glass parameters for Wolf Creek (Reference 2, Appendix F).

References:

1. ET 24-000478, Wolf Creek Nuclear Operating Corporation, Docket No. 50-482:

License Amendment Request for a Risk-Informed Resolution to GSI-191, September 12, 2024, (ADAMS Accession No. ML24260A071).

2. WCAP-17788, Volume 5, Rev. 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) - Autoclave Chemical Effects Testing for GSI-191 Long-Term Cooling.

Attachment I to 001052 Page 27 of 29 NRC Question APLC-Q1:

In attachment VII of the LAR (PDF page 91), the licensee states, Seismic events can result in direct or indirect LOCAs that generate and transport debris similar to a random pipe break LOCA. A direct seismically induced LOCA occurs when the RCS pressure boundary fails due to seismic force[s]. An indirect seismically induced LOCA occurs when a support or structure fails due to seismic forces, which subsequently causes an RCS pressure boundary failure.

However, in Section 2.3.4, Seismically Induced LOCAs, in attachment VII of the LAR (PDF page 98), it appears that the licensee only addresses indirect LOCAs caused by support failures, not by structure failures.

Provide examples of indirect LOCAs resulting from structural failures and clarify whether such failures were considered in the licensees Generic Safety Issue (GSI)-191 evaluation. If they were not considered, provide justification for their exclusion.

The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.

WCNOC Response:

The following response was provided to the NRC during the audit:

Consistent with NUREG-1903 (Reference 1) and NUREG/CR-3660 (Reference 2),

structural failure refers to the failure of major components within the containment that can result in pipe ruptures during a seismic event, not to specifically a building failure. For steam generators, reactor coolant pumps, and the reactor pressure vessel, the critical failure modes are typically failure of support members. For the overhead crane, failure of crane structural elements such as box girders, leg columns, and seismic lugs are among the critical failure modes.

Seismically induced LOCA frequencies for Wolf Creek were determined using two approaches: (1) using representative LOCA fragility parameters presented in EPRI TR-3002000709 (Reference 3), (2) using site-specific LOCA fragility parameters calculated in accordance with the guidelines provided in NUREG-1903 (Reference 1). The first approach utilized the EPRI representative fragility parameters for small, medium and large LOCAs; thus, the direct and indirect LOCA frequencies were not calculated explicitly. The second approach used the Safe Shutdown Earthquake (SSE) Peak Ground Acceleration (PGA), which is the minimum seismic capacity of a safety-related component, to determine a lower bound for site-specific fragility. Therefore, the second approach accounts for the structural failures implicitly. The results from the second approach serve to confirm the results from the first approach. The LOCA frequencies obtained from the first approach are bounding.

References:

1) NUREG-1903, Seismic Considerations for the Transition Break Size, Feb 2008.

2) NUREG/CR-3660, UCID-19988, Vol. 3, Probability of Pipe Failure in the Reactor Coolant Loop of Westinghouse PWR Plants Vol.3: Guillotine Break Indirectly Induced by Earthquakes, Feb 1985.

3) EPRI TR-3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, Final Report, December 2013.

Attachment I to 001052 Page 28 of 29 NRC Question APLC-Q2:

In attachment VII of the LAR (PDF page 92), the licensee states, High wind events, including tornados, would not generate debris inside containment and therefore are screened from the GSI-191 risk quantification. While containment can withstand the effects of high winds including tornado missiles, high winds may still lead to LOCAs through loss of offsite power (LOOP),

potentially resulting in debris blockage.

Discuss whether LOCAs due to LOOP caused by high winds should be considered in the licensee's GSI-191 risk evaluation.

The licensee provided a response to this question on the portal. The NRC staff concluded that the response was acceptable and needs the response on the docket to finalize its regulatory finding.

WCNOC Response:

The following response was provided to the NRC during the audit:

LOCAs due to LOOPs caused by high winds do not need to be included in the Wolf Creek GSI-191 risk evaluation. This conclusion is based on the type of LOCAs that could result from a LOOP and their size relative to the 10-inch threshold break size. The LOCAs resulting from a LOOP would be associated with either a stuck open primary PORV or an RCP seal failure. These would be small or very small LOCAs, which are significantly smaller than Wolf Creeks threshold break size. Therefore, LOCAs due to LOOPs do not need to be included in the Wolf Creek GSI-191 risk evaluation.

Reference:

1. ML24260A071, Docket No. 50-482: License Amendment Request for a Risk-Informed Resolution to GSI-191, September 12, 2024.

Attachment I to 001052 Page 29 of 29 NRC Question APLC-Q5:

Audit document, WCN021-CALC-008, figure 1, presents the WCGS seismic hazard curves for various spectral frequencies, including the peak ground acceleration (PGA). The figure title suggests that the PGA is defined at 100 hertz (Hz), but the figure itself includes separate curves for both the PGA and 100 Hz. Clarify which frequency corresponds to the PGA.

The NRC staff discussed this issue with the licensee during the audit call and requested that the licensee clarify the spectral frequency associated with the PGA used in the GSI-191 resolution and state whether the seismic-induced LOCA frequencies are impacted by the clarification in spectral frequency. If LOCA frequencies are affected, the licensee will provide revised seismic-induced LOCA frequency estimates accordingly. The NRC also requested that the licensee provide the reference Electric Power Research Institute document on the portal to allow the review of this reference. This information is required on the docket.

WCNOC Response:

The following response was provided to the NRC during the audit, with some minor wording changes:

The PGA is at a very high frequency. In the LOCA frequency calculation (Reference 2),

the PGA hazard curve (dashed line in Figure 1 of Reference 2, excerpt from the first figure in Attachment 2 of Reference 1) is used for convolution, not the 100 Hz hazard curve. Therefore, the LOCA frequencies calculated in Wolf Creek Calculation WCN021-CALC-008 (Reference 2) are not impacted.

References:

1. New Information Report (NIR) R-2023-68, "NGA-East Ground Motion Model - Wolf Creek", Enclosure 1 of LTR-106048.005, Electric Power Research Institute (EPRI),

December 18, 2023.

2. Wolf Creek Calculation WCN021-CALC-008, LOCA Frequency for Seismically-Induced LOCA, February 27, 2024.

Attachment II to 0001052 Page 1 of 2 Proposed Technical Specification Changes (Mark-Up)

Note: The Technical Specification (TS) Table of Contents are now under licensee control per License Amendment (LAR) 233 (cover letter Reference 1). The TS Table of Contents were inadvertently included in the original LAR application. With the TS Table of Contents now under licensee control, WCNOC requests that any changes involving the TS Table of Contents pages be withdrawn from NRC review.

This attachment contains the proposed mark-up for TS 5.5.15 in response to NRC question STSB-01. This change is in accordance with TSTF-567 Revision 1 without variation. Attachment III of this supplement contains the revised (clean) TS page. The TS changes are in addition to those provided in the LAR (cover letter Reference 2), and do not affect any previously provided TS mark-up or revised (clean) pages of the LAR.

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-20 Amendment No. 123, 142, 152, 164, 226, XXX 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Containment Leakage Rate Testing Program

a.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

b.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48 psig.

c.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day.

d.

Leakage rate acceptance criteria are:

1.

Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests; (continued)

Attachment III to 001052 Page 1 of 2 Proposed Technical Specification Clean Pages

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-20 Amendment No. 123, 142, 152, 164, 226, 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Containment Leakage Rate Testing Program

a.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

b.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48 psig.

c.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day.

d.

Leakage rate acceptance criteria are:

1.

Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests; (continued)