Text

Issuance of Amendment No. 246 to Implement GSI-191
	ML25317A782
Person / Time
Site:	Wolf Creek
Issue date:	01/14/2026
From:	Samson Lee; NRC/NRR/DORL/LPL4
To:	Reasoner C; Wolf Creek
References
	EPID L-2024-LLA-0125
	Download: ML25317A782 (0)
	v • d • e

January 14, 2026 Mr. Cleveland Reasoner Chief Executive Officer and Chief Nuclear Officer Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION, UNIT 1 - ISSUANCE OF AMENDMENT NO. 246 TO IMPLEMENT GENERIC SAFETY ISSUE-191 (EPID L-2024-LLA-0125)

Dear Mr. Reasoner:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 246 to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station, Unit 1 (Wolf Creek.) The amendment consists of changes to the technical specifications (TSs) in response to your application dated September 12, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24260A071), as supplemented by letter dated August 28, 2025 (ML25240B584).

Wolf Creek Nuclear Operating Corporation (the licensee), submitted a license amendment request, exemption request, and an updated response to Generic Letter (GL) 2004-02 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors (ML042360586) for Wolf Creek. The amendment allows the use of a risk-informed approach to address safety issues discussed in NRC Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on PWR [Pressurized Water Reactor] Sump Performance. The letter dated September 12, 2024, as supplemented constitute the licensees final response to GL 2004-02.

The amendment revises Wolf Creek TS 3.5.2, ECCS [Emergency Core Cooling System] -

Operating; TS 3.5.3, ECCS - Shutdown; and adds a new TS 3.6.8, Containment Sumps, to section 3.6, Containment Systems. The amendment incorporates Technical Specifications Task Force (TSTF) Traveler TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues, dated August 2, 2017 (ML17214A813). The NRC issued a final safety evaluation (SE) approving TSTF-567, Revision 1, on July 3, 2018 (ML18109A077).

In addition to the license amendment, the licensee requested an exemption pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.12, Specific exemptions, from certain requirements of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

A copy of the related SE is enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA - T. Byrd for/

Samson S. Lee, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482

Enclosures:

1. Amendment No. 246 to NPF-42

2. Safety Evaluation cc: Listserv

WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 246 License No. NPF-42

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Wolf Creek Generating Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated September 12, 2024 as supplemented by letter dated August 28, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-42 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 246, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael Mahoney, Acting Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 14, 2026 MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.01.14 10:20:46 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 246 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482 Replace the following pages of Renewed Facility Operating License No. NPF-42 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License REMOVE INSERT 4

4 Technical Specifications REMOVE INSERT 3.5-6 3.5-6 3.5-8 3.5-8 3.6-22 3.6-22 3.6-23 3.6-23 5.0-20 5.0-20

4 (5)

The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The Operating Corporation is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100%

power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 246, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 229, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Antitrust Conditions Evergy Kansas South, Inc. and Evergy Metro, Inc. shall comply with the antitrust conditions delineated in Appendix C to this license.

(4)

Environmental Qualification (Section 3.11, SSER #4, Section 3.11, SSER #5)*

Deleted per Amendment No. 141.

The parenthetical notation following the title of many license conditions denotes the section of the supporting Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed License No. NPF-42 Amendment No. 246

ECCS - Operating 3.5.2 Wolf Creek - Unit 1 3.5-6 Amendment No. 123, 155, 168, 227, 246 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each mechanical position stop is in the correct position.

Valve Number EM-V0095 EM-V0107 EM-V0089 EM-V0096 EM-V0108 EM-V0090 EM-V0097 EM-V0109 EM-V0091 EM-V0098 EM-V0110 EM-V0092 In accordance with the Surveillance Frequency Control Program

ECCS - Shutdown 3.5.3 Wolf Creek - Unit 1 3.5-8 Amendment No. 123, 227, 246 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs are applicable for all equipment required to be OPERABLE:

SR 3.5.2.1 SR 3.5.2.7 SR 3.5.2.3 SR 3.5.2.4 In accordance with applicable SRs

Containment Sump 3.6.8 Wolf Creek - Unit 1 3.6-22 Amendment No. 246 3.6 CONTAINMENT SYSTEMS 3.6.8 Containment Sump LCO 3.6.8 Two containment sumps shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more containment sumps inoperable due to containment accident generated and transported debris exceeding the analyzed limits.

A.1 Initiate action to mitigate containment accident generated and transported debris.

AND A.2 Perform SR 3.4.13.1.

AND A.3 Restore the containment sumps to OPERABLE status.

Immediately Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 90 days (continued)

Containment Sump 3.6.8 Wolf Creek - Unit 1 3.6-23 Amendment No. 246 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One or more containment sumps inoperable for reasons other than Condition A.

B.1 Declare affected Emergency Core Cooling System train(s) inoperable.

AND B.2 Declare affected containment spray train(s) inoperable Immediately Immediately C.

Required Action and associated Completion Time of Condition A not met.

C.1 Be in MODE 3.

AND C.2 Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.8.1 Verify by visual inspection, the containment sumps do not show structural damage, abnormal corrosion, or debris blockage.

In accordance with the Surveillance Frequency Control Program

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-20 Amendment No. 123, 142, 152, 164, 226, 246 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)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 246 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482

1.0 INTRODUCTION

By letter dated September 12, 2024 (Reference 1), as supplemented by letter dated August 28, 2025 (Reference 2), Wolf Creek Nuclear Operating Corporation (the licensee), submitted a license amendment request (LAR), exemption request, and an updated response to Generic Letter (GL) 2004-02 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors (Reference 3) for Wolf Creek Generating Station, Unit 1 (Wolf Creek).

The amendment would modify the Wolf Creek Technical Specifications (TSs). Specifically, the amendment would allow the use of a risk-informed approach to address safety issues discussed in U.S. Nuclear Regulatory Commission (NRC, the Commission) Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on PWR [Pressurized Water Reactor] Sump Performance. The letter dated September 12, 2024, as supplemented, constitutes the licensees final response to GL 2004-02. This safety evaluation (SE) reviews the LAR and the licensees responses to GL 2004-02.

In addition to the LAR, the licensee requested an exemption pursuant to Title 10 of the Code of Federal Regulations (10 CFR 50.12), Specific exemptions, from certain requirements of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, to allow use of a risk-informed methodology instead of the traditional deterministic methodology to resolve the concerns associated with GSI-191 and respond to GL 2004-02 for Wolf Creek.

The supplemental letter dated August 28, 2025, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on November 26, 2024 (89 FR 93365).

1.1 Background

1.1.1 Challenges to Function of Safety Systems from Debris in Containment The function of the emergency core cooling system (ECCS) is to cool the reactor core and provide shutdown capability following a loss-of-coolant accident (LOCA). The primary functions of the containment spray system (CSS) are to reduce containment pressure and reduce the concentration and quantity of fission products in the containment building after a LOCA.

Nuclear plants are designed and licensed with the expectation that they are able to remove reactor decay heat following a LOCA to prevent core damage. Long-term core cooling (LTCC) following a LOCA is also a basic safety function for nuclear reactors. The recirculation sump located in the lower areas of the reactor containment structure provides a water source to the ECCS for extended cooling of the core in a PWR once the initial water source has been depleted and the systems are switched over to recirculation mode.

If a LOCA occurs, piping thermal insulation and other materials located in containment may be dislodged by the two-phase (steam and liquid) coolant jet emanating from the broken reactor coolant system (RCS) pipe. This debris may transport by the flow of water and steam from the break or from the CSS to the pool of water that collects in the containment recirculation sump.

Once transported to the sump pool, the debris could be drawn toward the ECCS sump strainers, which are designed to prevent debris from entering the CSS and the ECCS. If this debris clogs the strainers, the ECCS could fail, resulting in core damage, or the CSS pumps could fail, resulting in containment pressure or radiation dose increasing beyond deterministic limits. It is also possible that some debris could bypass the sump strainers and get lodged in the reactor core. This could result in reduced core cooling and potential core damage.

1.1.2 Generic Safety Issue-191 In 1996, the NRC identified an issue associated with the effects of debris accumulation on PWR sump performance during design-basis accidents (DBAs).

Findings from research and industry operating experience raised questions concerning the adequacy of PWR sump designs. Research findings demonstrated that the amount of debris generated and transported by a high-energy LOCA could be greater than originally anticipated.

The debris from a LOCA could also be finer, and thus, more easily transportable, and could be comprised of debris consisting of fibrous material combined with particulate material that could result in a substantially greater flow restriction than an equivalent amount of either type of debris alone. These research findings prompted the NRC to open GSI-191.

The two distinct but related safety concerns are: (1) potential clogging of the sump strainers that results in ECCS or CSS pump failure, and (2) potential clogging of flow channels within the reactor vessel because of debris bypassing the sump strainers, often referred to as in-vessel effects. Clogging at either the strainers or in-vessel channels can result in loss of the LTCC safety function.

1.1.3 GL 2004-02 As part of the actions to resolve GSI-191, in September 2004, the NRC issued GL 2004-02 to holders of operating licenses for PWRs. In GL 2004-02, the NRC staff requested that licensees perform an evaluation of their ECCS and CSS recirculation functions, considering the potential

for debris-laden coolant to be circulated by the ECCS and the CSS after a LOCA or high-energy line break (HELB) inside containment, and, if appropriate, take additional actions to ensure system function. GL 2004-02 required, per 10 CFR Section 50.54(f), that licensees provide the NRC a written response describing the results of their evaluation and any modifications made, or planned, to ensure ECCS and CSS system function during recirculation following a design-basis event, or any alternate action proposed, and the basis for its acceptability.

The NRC Staff Requirements Memorandum (SRM) associated with SECY-10-113, Closure Options for Generic Safety Issue 191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance, dated December 23, 2010 (Reference 4), directed the NRC staff to consider a risk-informed approach for resolution of GSI-191. In 2012, the NRC staff developed three options to resolve GSI-191. These options were documented and proposed to the Commission in SECY-12-0093, Closure Options for Generic Safety Issue-191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance, dated July 9, 2012 (Reference 5). The options are summarized as follows:

Option 1 allows licensees to demonstrate compliance with 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, through approved models and test methods.

Option 2 requires implementation of additional mitigating measures and allows additional time for licensees to resolve issues through further industry testing or use of a risk-informed approach.

Option 3 involves separating the regulatory treatment of the sump strainer and in-vessel effects so that strainer issues can be treated deterministically, and in-vessel issues can be risk-informed.

These options allowed industry alternative approaches for resolving GSI-191. The Commission issued SRM-SECY-12-0093 on December 14, 2012 (Reference 6), approving all three options for closure of GSI-191.

In a public meeting on November 18, 2020 (Reference 7), the licensee stated it would pursue a full risk-informed resolution path (i.e., Option 2 of SECY-12-0093), to resolve GL 2004-02 and GSI-191 for Wolf Creek, using methods similar to other PWRs, but implementing a Threshold Break Size approach that was stated to be more conservative and easier to implement than other PWR risk-informed submittals. The licensee submitted its LAR on August 12, 2021 (Reference 8). The LAR was not accepted for NRC review as described in the NRC non-accept letter dated September 23, 2021 (Reference 9). The reason for the non-acceptance of the LAR was that the core damage frequency (CDF) and large early release frequency (LERF) values reported in the LAR could erode margin to NRC guidelines. The NRC letter also cited that the licensees evaluation did not aggregate values when calculating changes in CDF and LERF from various hazards without sufficient justification. The licensee withdrew the LAR from NRC review by letter dated October 20, 2021 (Reference 10), and also stated that a new submittal may be made in the future. The licensee submitted a revised LAR (Reference 1), which is the topic of this SE. The licensee cited the South Texas Project, Units 1 and 2 (STP) pilot assessment (Reference 11) as a precedent.

1.2 Licensees Approach The licensees risk-informed approach to evaluate the effects of debris on the sump strainer and pumps of the ECCS is documented in attachment VII to its LAR (Reference 1). Effects referred to as downstream effects (including in-vessel effects) were addressed using methods in topical report (TR) WCAP-17788-NP, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090), Volume 1, Revision 1 (Reference 12). The NRC staff found that the TR provided significant insights into the response of PWRs to the effects of debris that transport to the vessel, but the NRC did not approve it. This SE documents the NRC staffs evaluation of the licensees risk-informed approach to resolve GL 2004-02 at Wolf Creek.

The licensees LAR used guidance from draft Regulatory Guide (RG) 1.229, Risk Informed Approach for Addressing the Effects of Debris on Post Accident Long Term Cooling (Reference 13), addressing key principles of risk-informed integrated decision-making such as defense-in-depth and safety margins. The licensees overall evaluation of risk attributable to debris for Wolf Creek is based on physical models that have been used in the past in similar and precedent GSI-191 risk-informed assessments and generally accepted by the NRC for GSI-191 resolution. The licensee provided a summary of the plant-specific conditions and models (related to GSI-191, as well as a description of the risk quantification methodology, relying on computer-assisted design (CAD) 3-dimensional (3D) modeling of the distribution of insulation and coatings in the reactor containment, and the BADGER software to compute debris amounts as a function of break sizes and break orientations at multiple postulated break locations. The licensee used loss-of-coolant break frequencies from NUREG 1829, Estimating Loss of Coolant Accident (LOCA) Frequencies Through the Elicitation Process, Volumes 1 and 2, dated April 2008 (Reference 14), to estimate changes in risk to the plant. Based on strainer testing, the licensee used an approach that demonstrated all breaks smaller than the threshold break size could be mitigated and assumed all breaks exceeding the threshold break size would cause core damage. The licensee considered limited outputs of the probabilistic risk assessment (PRA) model, such as the CDF and the LERF conditional on LOCA events in its evaluation.

The licensee determined that most break scenarios were mitigated successfully when evaluated using deterministic methods. All breaks greater than the threshold break size were conservatively assumed to result in a core damage event. These failures, of total frequency equal to the NUREG-1829 exceedance LOCA frequency of the threshold break size, were assumed to contribute to the change in plant risk and were compared against RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis, acceptance guidelines (Reference 15). The licensee concluded that the change in risk is very small.

1.3 Method of NRC Staff Review The purpose of the NRC staffs review is to evaluate the licensees assessment of the impact of debris on ECCS and CSS functions following postulated LOCAs at Wolf Creek. The NRC staff evaluated the licensees LAR, conducted a regulatory audit, and performed confirmatory calculations in areas deemed appropriate by the NRC staff. The NRC staff performed an audit and issued a summary of the audit (Reference 16), and the licensee revised and updated its request in the supplement to the LAR dated August 28, 2025 (Reference 2), partially to address additional information requests and partially to address other issues identified by the licensee.

In areas where the licensee used NRC approved or widely accepted methods in performing analyses related to the proposed methodology, the NRC staff reviewed relevant material to ensure that the licensee used the methods consistent with the limitations and conditions placed on the methods. Details of the NRC staff review, audit, and confirmatory calculations are provided in section 3.0 of this SE.

2.0 REGULATORY EVALUATION

2.1 Applicable Regulatory Requirements The NRC staff assesses proposed remedial actions in accordance with the general standards for license amendments. Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards for licenses and construction permits in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public.

The NRC staffs acceptance criteria for ECCS performance following a LOCA are based on 10 CFR 50.46. LOCAs are postulated accidents that would result in the loss of reactor coolant from piping breaks in the reactor coolant pressure boundary at a rate exceeding the capability of the normal reactor coolant makeup system to replenish it. Loss of significant quantities of reactor coolant would prevent heat removal from the reactor core unless the water is replenished. The reactor protection and ECCS systems are provided to mitigate these accidents. The NRC staffs review covered the acceptance criteria based on 10 CFR 50.46, insofar as it establishes standards for the calculation of ECCS performance and acceptance criteria, considering the effects of debris as specified in GL 2004-02.

The NRC requirements for TSs are in 10 CFR 50.36, which state that TSs are to include items in, among other things, the following five specific categories related to station operation:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.

2.2 Applicable Regulatory Guides, Review Plans, and Guidance Documents Volume 1 of Nuclear Energy Institute (NEI) 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, dated December 2004, and Volume 2, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, dated December 6, 2004 (References 17 and 18, respectively), describe a method acceptable to the NRC staff, with limitations and conditions for performing the evaluations requested by GL 2004-02. Taken together NEI 04-07 and the associated NRC staff SE are often referred to as the guidance report/safety evaluation (GR/SE).

The industry developed the following additional TRs to aid licensees in responding to GL 2004-02.

TR WCAP-16530-NP-A, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191, dated March 2008 (Reference 19) and the associated NRC SE (Reference 20).

TR WCAP-16406-P-A, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, Revision 1, dated March 2008 (Reference 21) and the associated NRC SE (Reference 22).

TR WCAP-17788-P Volume 1 (Reference 12), and Volume 5, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) - Autoclave Chemical Effects Testing for GSI-191 Long-Term Cooling (Reference 23)

The reports listed above, subject to the limitations and conditions contained in the NRC SEs for those TRs, describe methods acceptable to the NRC staff for performing the evaluations and analyses within the scope stated in those documents (References 20 and 22).

To more clearly communicate the NRC staffs expectations for the level of technical detail in the licensees submittals, the NRC staff issued documents entitled Revised Content Guide for Generic Letter 2004-02 Supplemental Responses, dated November 21, 2007 (Reference 24),

and Revised Guidance for Review of Final Licensee Responses to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized Water Reactors, dated March 28, 2008 (Reference 25). The content guide describes the information necessary to be submitted to the NRC for review.

RG 1.82, Revision 4, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, dated March 2012 (Reference 26), provides guidance for an evaluation of the effects of debris on ECCS strainers and, more generally, guidance for the evaluation of water sources for long-term recirculation following a LOCA. Although the licensee used Revision 4, the NRC staff notes that Revision 5 was published in August 2022 (Reference 27). However, Revision 4 continues to be one way to meet the NRCs regulations.

Accordingly, the NRC staff used Revision 4 during its review.

RG 1.174, Revision 3 (Reference 15), provides guidance on the use of PRA findings and risk insights in support of licensee requests for changes to a plants licensing basis. This RG provides risk acceptance guidelines for evaluating the results of such evaluations. RG 1.174 also provides the five key principles of risk-informed integrated decision-making.

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, and RG 1.200 Revision 3, Acceptability of Probabilistic risk Assessment Results for Risk-Informed Activities (References 28 and 29, respectively), endorses, with clarifications, the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS) PRA Standard ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S 2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (ASME/ANS PRA Standard) (Reference 30). Revision 3 of RG 1.200 was issued in December 2020, and Revision 3 did not supersede Revision 2. The licensee cited both Revision 2 and Revision 3 of the RG as guidance for its submittal; however, the licensees risk-informed evaluation (attachment VII to Reference 1) cites the updated guidance in Revision 3, to identify information demonstrating the technical adequacy of the PRA when used in a risk-informed application.

General guidance for evaluating the technical basis for proposed risk-informed changes is provided in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR (Light-Water Reactor) Edition (SRP), Section 19.2, Review of

Risk Information Used to Support Permanent Plant Specific Changes to the Licensing Basis:

General Guidance, dated June 2007 (Reference 31).

2.3 Proposed Changes In the LAR, as supplemented, the licensee proposed changes to the Wolf Creek TSs. The changes incorporate Technical Specifications Task Force (TSTF) Traveler TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues, dated August 2, 2017 (Reference 32). The NRC issued a final SE approving TSTF-567, Revision 1, on July 3, 2018 (Reference 33). The amendment would revise TS 3.5.2, ECCS - Operating, and TS 3.5.3, ECCS - Shutdown. The proposed changes would also add a new TS 3.6.8, Containment Sump, to section 3.6, Containment Systems. Although the proposed changes are based on TSTF-567, the licensee and NRC staff identified variations from the TS changes described in TSTF-567. These variations are described in section 2.3.5 of this SE and evaluated in section 3.6.5.

2.3.1 Proposed Changes to TS 3.5.2 TS 3.5.2 currently contains SR 3.5.2.8, which states the following at a Frequency in accordance with the Surveillance Frequency Control Program (SFCP):

Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and suction inlet strainers show no evidence of structural distress or abnormal corrosion.

The licensee proposed to modify and move SR 3.5.2.8 from TS 3.5.2 and include it in the new containment sump TS.

2.3.2 Proposed Changes to TS 3.5.3 TS 3.5.3 currently contains SR 3.5.3.1, which refers to applicable SRs under TS 3.5.2. One of those referenced SRs is SR 3.5.2.8, as described in section 2.3.1 of this SE.

Because the licensee proposed to modify and move SR 3.5.2.8 from TS 3.5.2 and include it in the new containment sump TS, the licensee also proposed to delete the reference to SR 3.5.2.8 in SR 3.5.3.1.

2.3.3 Proposed Changes to TS 5.5.15 The NRC staff noted that the licensee did not propose the TSTF-567 proposed change for TS 5.5.15, Safety Function Determination Program (SFDP). The NRC staff requested that the licensee incorporate this change or explain why the change is not required for Wolf Creek (Reference 16). To resolve the issue, the licensee proposed adding the following sentence at the end of TS 5.5.15 in the supplement to the LAR (Reference 2):

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.

2.3.4 Proposed Addition of a New Containment Sump TS The licensee proposed to add new TS 3.6.8 requiring two containment sumps to be operable during Modes 1, 2, 3, and 4. Condition A specifies that if the containment sump is inoperable due to containment accident generated and transported debris exceeding the analyzed limits, then the licensee is required to: (1) initiate action to mitigate the containment accident generated and transported debris immediately, (2) perform SR 3.4.13.1 once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and (3) restore the containment sump to OPERABLE status within 90 days (Required Actions A.1, A.2, and A.3, respectively). SR 3.4.13.1 requires verification that the RCS operational leakage is within limits by performance of an RCS water inventory balance.

Condition B specifies that if one or more containment sumps are inoperable for reasons other than Condition A, then the licensee is required to declare the affected ECCS and CSS trains inoperable immediately (Required Actions B.1 and B.2).

Condition C specifies that if required actions and associated completion times under Conditions A and B are not met, then the licensee is required to be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (Required Actions C.1 and C.2, respectively).

The licensee proposed to modify and move SR 3.5.2.8 currently located in TS 3.5.2. The new SR 3.6.8.1 requires the licensee to [v]erify, by visual inspection, the containment sumps do not show structural damage, abnormal corrosion, or debris blockage [i]n accordance with the Surveillance Frequency Control Program.

The containment sump design for Wolf Creek includes two containment sumps. The system includes two strainers: one for each train of ECCS and CSS. For Condition A, the sumps are considered part of a single support system because containment accident generated and transported debris issues that would render one sump inoperable would render all of the sumps inoperable.

2.3.5 Proposed Variations from TSTF-567, Revision 1 The licensee identified the following variations from the traveler.

1. The Wolf Creek TSs utilize different numbering than the Standard TSs (STS) on which TSTF-567 was based. Specifically, TS 3.6.19 in TSTF-567 is TS 3.6.8 in the Wolf Creek TSs. This difference is administrative and does not affect the applicability of TSTF-567 to the Wolf Creek TSs.

2. The Required Action and notes of the proposed Condition B in the traveler are revised to require declaring the affected ECCS and CSS trains inoperable immediately. The proposed Required Actions require immediate entry into the appropriate ECCS and CSS TSs. The licensee stated that this change ensures that the appropriate restrictions are implemented in accordance with the required actions of the ECCS and CSS TSs. The licensee stated that these changes are considered administrative since the proposed requirements will result in equivalent actions for the condition.

3. The licensee also stated that the debris limits for the ECCS strainers are provided in the TS Bases, table B 3.6.8-1, instead of the Updated Safety Analysis Report (USAR).

2.3.6 Proposed USAR Changes In attachment VI to the LAR (Reference 1), the licensee provided changes to the Wolf Creek USAR. The proposed changes were provided for information only. These changes describe the treatment of debris with respect to operation of the ECCS and CSS during sump recirculation.

The licensee included a new USAR appendix 6A to describe the methodology used in its risk-informed evaluation of the effects of debris on ECCS and CSS operation.

3.0 TECHNICAL EVALUATION

There are no physical modifications needed or planned in support of this application at Wolf Creek. Operating procedures at Wolf Creek have actions that prevent and mitigate strainer blockage based on indications available to operators such as instrumentation to monitor core water levels, sump water levels, and containment temperatures. When the issue regarding the effects of debris on strainer performance were initially being addressed, the licensee replaced the original strainers with a design that has a much larger strainer area and improved filtering performance.

The NRC staff performed an integrated review of the proposed risk-informed approach, considering the five key principles of risk-informed decision-making set forth in RG 1.174, Revision 3.

3.1 Key Principle 1: The Proposed Change Meets Current Regulations Unless it is Explicitly Related to a Requested Exemption or Rule Change The proposed change requested to modify the Wolf Creek licensing basis analyses to show compliance with 10 CFR 50.46 considering the effects of debris using both deterministic and risk-informed methodologies.

NEI 04-07; RG 1.82, Revision 4; TR WCAP-17788-P; NRC Staff Review Guidance for In-Vessel Effects (Reference 34); and SRP Section 19.2 are the primary guidance documents used to show regulatory compliance with 10 CFR 50.46, considering the effects of debris using deterministic criteria. As described previously, the Wolf Creek method uses both deterministic and risk-informed criteria. Most of the break scenarios are shown to meet deterministic acceptance criteria. For scenarios where deterministic acceptance criteria are not satisfied, the licensee proposed an exemption to 10 CFR 50.46. The requested exemption from 10 CFR 50.46(a)(1) was evaluated by the NRC staff against the criteria of 10 CFR 50.12, Specific exemptions, and found to be acceptable in the related exemption issuance. A successful demonstration that all break scenarios are bounded by the deterministic criteria or fall within the bounds of the exemption demonstrates that regulations have been met for this request.

The criteria to evaluate compliance with 10 CFR 50.46 using a risk-informed methodology are provided in the SRP Section 19.2; RG 1.200, Revision 2; and RG 1.174, Revision 3. However, the NRCs position has historically interpreted and applied the current regulations in 10 CFR 50.46 as requiring a deterministic approach and bounding calculations to show compliance. Thus, the NRCs longstanding practice may be regarded as a legally binding requirement from which an exemption is the appropriate means of granting dispensation from compliance. The licensee stated that, as allowed by SRM-SECY-12-0093, it chose to use a risk-informed method to resolve GSI-191 and to respond to GL 2004-02 (Reference 1). Thus, in accordance with 10 CFR 50.12, the licensee requested an exemption to 10 CFR 50.46(a)(1) in

attachment II to the LAR. The licensee concludes that for Wolf Creek the risk for the effects of debris is less than the threshold for Region III (Very Small Changes) of RG 1.174, and no additional physical changes to the facility or changes to the operation of the facility were proposed.

The NRC staff determined that special circumstances exist to grant the proposed exemption and that granting the exemption would not result in a violation of the Atomic Energy Act of 1954, as amended. Therefore, since the NRC staff has granted the exemption and the change explicitly relates to the requested exemption, the proposed change to use the risk-informed methodology meets the first key principle of RG 1.174.

3.2 Key Principle 2: The Proposed Change is Consistent with the Defense-in-Depth Philosophy As described in Regulatory Position C.2.1.1, Defense in Depth, of RG 1.174, Revision 3, the defense-in-depth philosophy consists of a number of considerations and consistency with the defense-in-depth philosophy is maintained if seven considerations addressed in sections 3.2.1 through 3.2.7 below are met.

In attachment IX to the LAR (Reference 1), the licensee explicitly addressed defense-in-depth and the seven considerations for Wolf Creek. The licensee also addressed safety margin in attachment IX. Items associated with defense-in-depth that were included in the licensees analysis are evaluated in sections 3.2.1 to 3.2.7 of this SE.

3.2.1 A Reasonable Balance is Preserved Among Prevention of Core Damage, Prevention of Containment Failure, and Consequence Mitigation The licensee highlighted physical and procedural changes such as installation of new strainers with increased surface areas and a reduced opening size, installation of flow diverters to prevent debris-laden fluid directly reaching the sumps, implementation of the standard design change process that identifies potential impact to GL 2004-02 compliance by planned modifications, and program controls to ensure the debris load limits are not exceeded. The licensee also stated that the risk-informed elements of the analysis showed a very small increase in risk of containment or reactor failures related to GSI-191 phenomena.

The NRC staff reviewed the licensees rationale and concluded that a reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation because of the following:

1. There is a robust plant design to survive hazards and minimize challenges that could result in the occurrence of an event, and the change to adopt a risk-informed approach for assessing the effects of debris does not increase the likelihood of initiating events or create new significant initiating events;

2. Prevention measures are in place with adequate availability and reliability of structures, systems, and components (SSCs), providing the safety functions that prevent plant challenges from progressing to core damage;

3. Existing measures are in place to contain a source term if a severe accident occurs; and

4. The change does not reduce the effectiveness of the emergency preparedness program, including the ability to detect and measure releases of radioactivity, notify offsite agencies and the public, and shelter or evacuate the public as necessary.

3.2.2 Over-Reliance on Programmatic Activities as Compensatory Measures Associated with the Change in the Licensing Basis is Avoided This defense-in-depth consideration evaluates if design features are substituted by programmatic activities to an extent that significantly reduces the reliability and availability of design features to perform their safety functions. The licensee identified that the change would not adversely affect any of the programmatic activities already in place at Wolf Creek, such as the inservice inspection (ISI) program, plant personnel training, RCS leakage detection program, and containment cleanliness inspection activities. The proposed change does not rely heavily on programmatic activities as compensatory measures nor propose any new programmatic activities that could be heavily relied upon.

The NRC staff reviewed the licensees description of programmatic activities and concludes that this defense-in-depth consideration is met because the proposed change does not affect how safety functions are performed, nor does it reduce the reliability or availability of the SSCs that perform those functions. Existing programmatic activities are maintained, and therefore, there is not an excessive reliance on programmatic activities as compensatory measures related to the risk-informed approach.

3.2.3 System Redundancy, Independence, and Diversity are Preserved, Commensurate with the Expected Frequency, Consequences of Challenges to the System, and Uncertainties (for Example, No Risk Outliers)

The licensee highlighted the redundancy, independence, and diversity of the ECCS and CSS equipment and asserted that the proposed change does not require any design change to these systems. Therefore, system redundancy, independence, and diversity are preserved. These systems were analyzed relative to their contribution to nuclear safety through the Wolf Creek plant-specific PRA (which meets industry PRA standards for risk-informed applications),

accounting for a full range of LOCA events and uncertainties, and no risk outliers were identified. The licensee also stated that the risk contribution due to the change has been evaluated for the full spectrum of LOCA events, including uncertainties, and it was concluded with high confidence that the risk associated with the effects of debris at Wolf Creek are very small.

The NRC staff reviewed the licensees evaluation of this defense-in-depth consideration and concludes that it is met because the risk-informed analysis is consistent with the assumptions in the safety analysis for Wolf Creek and does not significantly increase the expected frequency of challenges to the systems, or consequences of failure of the system functions as a result of a decrease in redundancy, independence, or diversity. The licensee performed a comprehensive risk assessment and demonstrated that reductions in redundancy, independence, or diversity of systems resulting from postulated LOCAs do not cause a significant increase in risk, as evidenced by a margin to RG 1.174 risk acceptance guidelines. The licensee included sensitivity cases to assess uncertainty. Although some sensitivity cases yielded higher risk increases, those alternatives were considered as part of the uncertainty of the risk estimates and controlled by different sets of assumptions. See section 3.4.2.9 of this SE.

3.2.4 Defenses Against Potential Common-Cause Failures are Preserved, and the Potential for the Introduction of New Common-Cause Failure Mechanisms is Assessed The licensee examined common-cause failure mechanisms in the context of GL 2004-02; specifically, the primary failure mechanism of concern is recirculation strainer clogging limiting adequate flow to any of the ECCS and CSS pumps. The defenses against potential strainer clogging are not changed by the risk-informed methodology; there are no design changes to the equipment or changes to emergency operating procedures. The licensee noted that defenses against strainer clogging mechanisms are strengthened by the changes made by the site during its response to GL 2004-02.

The NRC staff reviewed the evaluation of this defense-in-depth consideration and concludes that it is met because the risk-informed evaluation does not introduce a new potential common-cause failure or event for which a defense is not in place; does not increase the probability or frequency of a cause or event that could cause simultaneous multiple component failures; does not introduce a new coupling factor for which a defense is not in place; and does not weaken or defeat an existing defense against a cause, event, or coupling factor. Even though the strainer blockage failure mechanism is not deterministically eliminated, the risk analysis shows that the risk is very small and that additional mitigative and defense-in-depth measures exist.

3.2.5 Independence of Barriers is Not Degraded The three barriers to a radioactive release are the fuel cladding, RCS pressure boundary, and reactor containment building. The licensee stated that in its evaluation of a LOCA, the RCS barrier is postulated to be breached, and the proposed change does not affect the design and analysis requirements for the fuel. The licensee noted that during recirculation, the post-LOCA fluid collecting in the containment sump pits is mobilized by the residual heat removal (RHR) pumps and recirculated back to the containment. The auxiliary building has dedicated filters in the ventilation system to limit offsite releases. The licensee stated that the proposed change does not alter the design or operating requirements for the ECCS or the auxiliary building.

The licensee stated that containment cooling and pressure control can be accomplished considering a single failure that results in the loss of one air cooling train and a train of containment spray (CS), and that sufficient heat removal would occur to prevent containment pressure from exceeding its design limit. The licensee stated that the licensing basis change does not alter the design or operating requirements of these systems and concluded that the independence of the barriers is maintained and not degraded by the licensing basis change.

The licensee provided an additional evaluation of defense-in-depth for the barriers to the release of radioactivity (attachment IX of Reference 1, page 10). The licensee stated that the severe accident mitigation management guidelines (SAMGs) are designed to protect these barriers under conditions that warrant entry into the SAMGs as discussed above. The licensee also provided a description of the programs that prevent and detect pipe breaks. The licensee discussed ASME Boiler and Pressure Vessel Code (ASME Code) requirements that are intended to prevent RCS pressure boundary failures and the leak detection program that can identify small leaks so that actions can be taken to address RCS leakage. The licensee also discussed reactor containment integrity. The containment is designed to accommodate the pressure from the most limiting RCS pipe breaks with margin while considering the most challenging single active failure of ECCS, CSS, and the containment cooling units during injection and the worst active or passive single failure during recirculation. The licensee stated that the methodology and acceptance criteria used to evaluate the acceptability of the

containment is not changed by the proposed amendment and concluded that containment integrity is not affected. Some of these topics are discussed in greater detail in this SE in section 3.3.

The NRC staff reviewed the licensees evaluation of this defense-in-depth consideration and concludes that it is met because implementation of the methodology does not result in a significant increase in the frequency of existing challenges to the integrity of the barriers or in the failure probability of any individual barrier. Moreover, implementation of the methodology does not introduce new or additional failure dependencies among barriers that significantly increase the likelihood of failure.

3.2.6 Defenses Against Human Errors are Preserved This consideration evaluates if implementation of the proposed methodology significantly increases the potential for or creates new human errors that might adversely impact one or more layers of defense. The licensee stated that the proposed change does not involve any additional operator actions or increase the complexity of any operator actions. The licensee concluded that the defenses that are already in place with respect to human errors are not impacted by the proposed licensing basis change.

The NRC staff reviewed the evaluation of this defense-in-depth consideration and concludes that it is met because the implementation of the proposed methodology does not reduce the ability of plant staff to perform actions. Specifically, the methodology does not create new human actions that are important to preserving any of the layers of defense, or significantly increase the probability of existing human errors by affecting performance shaping factors, including mental and physical demands and level of training.

3.2.7 The Intent of the Plants Design Criteria is Maintained The licensee stated that the proposed license change does not involve any change to the physical design of the current plant equipment associated with GL 2004-02. The licensee stated that the acceptance criteria of 10 CFR 50.46 are not changed. The proposed license change revises the licensing basis for acceptable containment emergency sump strainer design and performance in support of ECCS and CSS operation in recirculation mode following postulated LOCAs by demonstrating that the effect of debris on LTCC results in very small risk. The licensee concluded that the intent of the plants design criteria is maintained.

The NRC staff reviewed the licensees evaluation of this defense-in-depth consideration and concludes that the proposed change maintains the intent of the plants design criteria, because an alternate risk-informed evaluation method provides an acceptable approach that demonstrates that LTCC will be maintained following a LOCA, and thus, does not result in a reduction in the effectiveness of one or more layers of defense.

3.2.8 Additional Defense-in-Depth Considerations-Detecting and Mitigating Adverse Conditions The licensee provided additional information on defense-in-depth measures in support of the response to GL 2004-02. In attachment IX to the LAR, the licensee stated that adequate defense-in-depth is maintained by ensuring the capability for operators to detect and mitigate inadequate flow through recirculation strainers and inadequate flow through the reactor core due to the potential impacts of debris blockage.

3.2.8.1 Prevention of Strainer Blockage The licensee identified that the sump strainers are monitored for blockage and that actions are directed if blockage occurs as specified in Wolf Creeks Emergency Management Guidelines (EMGs). The licensee stated that for smaller LOCAs, depletion of the refueling water storage tank (RWST) can be delayed by cooling the RCS to reduce break flow and therefore the injection flow rate. This can bring the plant to cold shutdown before the RWST is drained and recirculation is required. The licensee also described measures taken to control or reduce the debris source term inside containment (attachment IX of Reference 1, page 6).

3.2.8.2 Detection of Strainer Blockage The licensee stated that Wolf Creek has operational procedures to monitor operating parameters related to the RHR pump flow, discharge pressure, and amperage. These procedures allow control room personnel to properly diagnose the occurrence of cavitation as an indication of sump clogging. The licensee also monitors core exit thermocouples and a reactor vessel level indication system. Changes in these indications are used to detect potential blockage in the sump (attachment IX of Reference 1, page 7).

3.2.8.3 Mitigation of Strainer Blockage The licensee highlighted that the Wolf Creek EMGs contain steps to reduce flow through the system and provide alternate injection flowpaths. The EMGs provide guidance for refilling the RWST to allow alternate injection flowpaths. The licensee also stated that temporarily terminating recirculation flow may allow some debris to be dislodged from the strainer thus reducing the head loss across the strainer.

The licensee also noted the diverse and flexible coping strategies (FLEX) to maintain fuel cooling and containment integrity developed in response to the NRC Order EA-12-049, Mitigation Strategies for Beyond-Design-Basis External Events (BDBEE) (Reference 35). The licensee asserted that various modifications have been implemented such that non-emergency equipment can be credited during an event (attachment IX of Reference 1, page 8).

3.2.8.4 Prevention of Inadequate Reactor Core Flow The licensee stated that the actions discussed above for reducing flow through the strainers have a similar positive effect on reducing the potential for fuel blockage. The licensee also discussed the capability to initiate simultaneous cold-leg and upper plenum injection after the RHR pumps have been realigned to the sump. This provides an alternate flowpath (AFP) for coolant to enter the core (attachment IX of Reference 1, page 8).

3.2.8.5 Detection of Inadequate Reactor Core Flow The licensee stated that there are multiple methods for detection of core blockage resulting in inadequate core cooling or RCS inventory. Core blockage would be indicated by an increase in core exit thermocouple temperature or a reduction in reactor water level as monitored by the reactor vessel level indication system. The licensee noted an additional method for detection of core blockage is monitoring containment radiation levels (attachment IX of Reference 1, page 9).

3.2.8.6 Mitigation of Inadequate Reactor Core Flow The licensee stated that multiple methods are available to mitigate inadequate core flow conditions. The Wolf Creek EMGs include direction to reestablish cooling flow to the RCS, reduce RCS pressure, restart reactor coolant pumps (RCPs), and open power-operated relief valves. The licensee also stated that the SAMGs provide additional actions to protect fission product boundaries if the EMGs become ineffective. The SAMGs provide alternate injection means, including flooding the containment. FLEX equipment may also be used to mitigate inadequate core flow (attachment IX of Reference 1, page 9).

3.2.8.7 NRC Staff Review of Additional Defense-in-Depth The NRC staff reviewed the licensees additional defense-in-depth actions and programs and concludes that the licensees measures to prevent, detect, and mitigate adverse conditions (such as inadequate strainer flow or inadequate core flow), barriers for release of radioactivity, emergency plan actions, and SAMGs provide additional defense-in-depth measures beyond the seven factors defined in RG 1.174, Revision 3.

3.

2.9 NRC Staff Conclusion

Regarding Key Principle 2: Defense-in-Depth The NRC staff finds that the philosophy of defense-in-depth is maintained under the analysis described in attachment IX to the LAR, because the licensee appropriately addressed each of the seven factors in section 2.1.1 of RG 1.174, Revision 3, and provided additional information, specifically for issues that have been identified as critical areas where defense-in-depth can be beneficial.

3.3 Key Principle 3: The Proposed Change Maintains Sufficient Safety Margins As described in RG 1.174, safety margins are maintained when codes and standards or their alternatives approved for use by the NRC are met, and when the safety analysis acceptance criteria in the licensing basis (e.g., final safety analysis report, supporting analyses) are met, or proposed revisions provide sufficient margin to account for analysis and data uncertainty.

The licensee examined the safety margins and stated that there are numerous conservatisms included in the risk-informed GL 2004-02 evaluation. The licensee described barriers for release of radioactivity in attachment IX, section 2.3 to the LAR (Reference 1). The licensee cited the fuel cladding, RCS pressure boundary, and the reactor containment building as the barriers.

The licensee summarized a list of conservatisms in table 9-1, Description of Safety Margin, of attachment IX to the LAR, related to several areas in the risk-informed analysis to show that the proposed approach maintains sufficient safety margins.

3.3.1 Barriers for Release of Radioactivity The licensee concluded that the proposed change maintains sufficient defense-in-depth for current barriers (cladding, RCS boundary, containment, and emergency plan actions) against the release of radioactivity. The NRC staff determined that some of the information provided by the licensee also contributes to margin in the analysis. Section 3.2 of this SE discusses the defense-in-depth aspects of these barriers. The licensee provided information regarding measures taken to assure that the RCS pressure boundary will not fail. These measures follow ASME Code and other requirements that ensure margin in the assumptions for failure of the RCS pressure boundary.

The licensee evaluated aspects contributing to safety margin such as the fuel cladding, emergency core cooling and long-term cooling, RCS pressure boundary, ISI program, RCS weld mitigation, RCS leakage detection, containment integrity, containment testing, and operator actions. The licensee highlighted the following aspects in attachment IX, sections 2.3 and 2.4 to its LAR.

Although the RCS pressure boundary is assumed to be failed for the GL 2004-02 sump risk-informed evaluation, the proposed change does not make any change to the previous analyses and testing programs that demonstrate the integrity of the RCS.

The Wolf Creek ISI program plan provides verification that structural integrity of ASME Class 1, 2, and 3 piping components are within the limits specified in the ISI program, and verification that the structural integrity of the main feedwater piping is within the limits specified in the augmented ISI program. The licensee stated that the Class 1 welds, piping, and components are maintained at a high level of reliability through the inspection program.

The licensee cited RCS overpressure protection provided by ASME Code safety valves.

The leak detection program at Wolf Creek is capable of early identification of RCS leakage to provide time for appropriate operator action prior to a large-break.

The containment remains a low leakage barrier against the release of fission products for the duration of the postulated LOCAs. The containment cooling units and CSS are designed to reduce the containment pressure and temperature after a DBA including a loss of offsite power and a single failure.

The proposed change to the licensing basis does not involve any changes to the emergency plan; the use of the risk-informed approach does not impose any additional operator actions or complexity.

During the regulatory audit (Reference 16), the NRC staff requested that the licensee provide additional information regarding defense in depth for the RCS pressure boundary. The licensee responded in its supplement to the LAR (Reference 2). The NRC staff evaluated the response.

The NRC conclusions are provided in the following evaluation.

The evaluation of defense-in-depth for the RCS pressure boundary addresses whether the impact of the proposed licensing basis change (individually and cumulatively) is consistent with the defense-in-depth philosophy, as outlined in RG 1.174. The evaluation also presents the measures available for preventing, detecting, and mitigating conditions that could challenge LTCC due to strainer blockage and inadequate cooling flow to the reactor core. The measures discussed here contribute to both defense-in-depth and safety margins in the analysis.

The licensee stated that the integrity of the RCS pressure boundary is assumed to be compromised for the GL 2004-02 sump performance evaluation. The licensee stated that the proposed licensing basis change does not modify the previous analyses or testing programs that demonstrate the integrity of the RCS. The licensee considers all Class 1 welds in the GSI-191 analysis.

To demonstrate defense-in-depth, the licensee stated that Wolf Creek developed a plan to manage the risk of primary water stress corrosion cracking (PWSCC) degradation in Alloy 600 components and Alloy 82/182 welds that are used in Class 1 RCS piping as discussed below.

Inservice Inspection Program In section 2.3.2 of attachment IX to the LAR (Reference 1), the licensee states that Wolf Creek has developed a program to manage the risk of PWSCC degradation in components that are made of Alloy 600 material and Alloy 82/182 welds. The licensee stated that its ISI plan is in accordance with 10 CFR 50.55a, Codes and standards; ASME Code Case N-722-1, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials,Section XI, Division1; Code Case N-770-2, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N060882 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1; and NEI 03-08, Guideline for the Management of Materials Issues (Reference 36).

The NRC staff noted that the regulation under 10 CFR 50.55a requires nuclear power plants to perform ISI of ASME Code Classes 1, 2 and 3 components. The subject LAR is related to Class 1 and Class 2 piping components (i.e., in-scope components). As such, the licensee is required to inspect Class 1 and Class 2 piping components in accordance with the ASME Code,Section XI, articles IWB-2500 and IWC-2500, respectively. The NRC staff noted that most of the in-scope components are Class 1 because they are required to maintain the primary system pressure boundary. The ASME Code,Section XI, tables IWB-2500-1 and IWC-2500-1 specify the examination method, specific components to be examined, frequency of examination, acceptance criteria of the examination results, and the configuration of the component that needs to be examined. The licensee is required to perform periodic inspections of the in-scope components in accordance with the ASME Code,Section XI, to monitor their condition during refueling outages. Also, the licensee is required to perform system leakage test and associated visual examination of the Class 1 and Class 2 components prior to plant restart after the completion of each refueling outage.

In addition to the required ISIs in accordance with ASME Code,Section XI, the licensee performs additional inspections based on industry guidance as part of plant operating procedures. In the supplement to the LAR, the licensee stated that Wolf Creek developed and administers an Outside Diameter Stress Corrosion Cracking Program as a result of the industry and plant-specific operating experience. The program has evolved since its inception and has been credited for satisfying subsequent license renewal inspections. The inspection 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 outage 17. During the extent of condition review, the licensee identified other locations as being susceptible to ODSCC. The licensee performs periodic dye Penetrant inspections in accordance with its procedures, Visual Dye Penetrant Examinations, of piping/locations identified as most likely susceptible to ODSCC based on operating experience. The licensee stated that activities identified in this procedure are credited with the performance of External Surfaces Monitoring of Mechanical Components, aging management activities required by license renewal. In addition to the ODSCC inspections, the licensee performs visual examination of the cold-leg resistance temperature detector thermowells on the same frequency as the hot-leg every refueling outage, instead of once per ISI interval. The licensee stated that ASME Section XI Pressure Test Program monitors the RCS pressure boundary for external leakage by visual examination VT-2 qualified examiners

during every refueling outage, and accessible Class 2 and 3 systems during online operations.

The licensee stated that it has implemented a boric acid corrosion control program 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. The licensee explained that 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. The licensee also frequently monitors and trends the RCS leakage to aid in leak detection both internally and externally.

The NRC staff finds that the licensees ISIs per the ASME Code,Section XI and industry guidance provide adequate defense-in-depth measures to minimize the probability of component failure.

The NRC staff noted that the ASME Code,Section XI required ISIs are adequate for the nuclear plant components that are not degraded. However, for the degraded components that remain in service or those components that are known to be susceptible to degradation based on operating experience, additional inspections may be required. The NRC staff noted that the operating experience in PWRs has shown that Alloy 600 base metal and welds fabricated with Alloy 82/182 filler material are susceptible to PWSCC.

As stated in section 2.3.2 of attachment IX to its LAR, the licensee stated that it performs ISI according to ASME Code Case N-722-1, Code Case N-770-2, and NEI 03-08. Section 2.3.2 of attachment IX states that the Wolf Creek ISI 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. For example, periodic inspections of the Alloy 600 components and Alloy 82/182 welds are covered in the ISI program.

The NRC staff evaluated how the licensee monitors those unmitigated in-scope Alloy 600/82/182 components to ensure their structural integrity. In the supplement to the LAR, the licensee identified the welds in the RCS piping and vessels that are fabricated with nickel-based Alloy 82/182 weld material in a table. This table also indicates the welds that are either mitigated or repaired and the welds that had inservice cracking. For example, the table shows that during volumetric examination prior to installing structural weld overlays, the licensee detected indications that were caused by PWSCC at the safe-end weld of the pressurizer safety and relief nozzles and surge line.

The licensee indicated that some PWR owners have detected cracking in their bottom mounted instrumentation nozzle and welds of the reactor vessel. The licensee stated that it mitigated the reactor vessel bottom mounted nozzle and welds. The NRC staff noted that from the information provided, in-scope components that had indications/flaws have been repaired or mitigated.

In the supplement to the LAR, the licensee also presented a table showing the Alloy 82/182 welds that have not been mitigated. The table provides inspection strategies for the steam generator drain pipe, reactor bottom mounted nozzle to guide tube welds, welds associated with the reactor vessel head vent-to-elbow, vent elbow-to-piping, vent pipe-to-stainless steel elbow, control rod drive mechanism-to-adapter, and RCS hot and cold-leg thermowells. Many of these unmitigated welds are being examined during each refueling outage such that the condition of these unmitigated components is being monitored.

For those unmitigated in-scope Alloy 600/82/182 components, the licensee stated that the long-term mitigation plan 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. The NRC staff determined that the licensee has plans to replace the thermowells in the hot and cold-leg, and the licensee performs appropriate inspections to monitor the unmitigated Alloy 600/82/182 components.

The NRC staff questioned whether the risk analysis contains welds that contain flaws that have not been repaired or mitigated, and if so, whether the LOCA frequency for the degraded welds in the risk analysis should be increased from the estimates from NUREG-1829 (Reference 14).

In the supplement to the LAR, the licensee stated that the risk analysis does not contain welds with flaws that have not been repaired.

The NRC staff questioned whether the LOCA frequency in the risk analysis was increased to account for the in-scope unmitigated nickel-based Alloy 600/82/182 components. In the supplement to the LAR, the licensee stated that the Wolf Creek GSI-191 evaluation did not include any adjustment to LOCA frequencies based on weld-specific flaws or degradation mechanisms. The licensee stated that early in the risk-informed GSI-191 pilot project, STP estimated the LOCA frequencies at individual welds considering weld-specific degradation mechanisms. This was referred to as the bottom-up approach for estimating LOCA frequencies. At the time, the NRC staff did not agree with this approach because bottom-up LOCA frequencies would not be consistent with the overall LOCA frequencies in NUREG-1829.

To address the NRC staffs concerns, STP pursued two alternative methods. The top-down approach allocated the LOCA frequencies from NUREG-1829 equally across welds large enough to host given break sizes. The hybrid approach allocated the NUREG-1829 frequencies across welds with a bias to account for weld-specific degradation mechanisms (i.e., the bottom-up frequencies were used to determine relative and uneven probabilities of breaks at specific welds, but those frequencies were normalized to add up to the total NUREG-1829 exceedance frequencies). In its final assessment, STP only used the simpler top-down approach, which was accepted by the NRC (Reference 11).

The licensee stated that for Wolf Creek, allocation of LOCA frequencies to individual welds was not necessary for the conservative threshold break size approach. The licensee explained that Wolf Creek implemented a simplified risk-informed approach where all LOCA breaks equal or greater than the threshold break size were conservatively assumed to cause strainer failure.

The licensee defined the threshold break size to be 10 inches. Therefore, the licensee assumed that weld breaks equal or greater than 10 inches would cause strainer failures and core damage, independently of debris amounts generated by the LOCA break. The licensee directly interpolated the exceedance frequency from NUREG-1829 corresponding to 10-inch LOCA breaks to conservatively compute the CDF, without needing to consider the relative contribution from individual welds.

The NRC staff finds the assumption that all breaks exceeding 10 inches cause core damage to be conservative. The NRC staff recognized that there are breaks greater than 10 inches that would not generate enough debris to cause strainer failure and much less core damage. The NRC staff noted that the licensee evaluated debris generation by pipe breaks less than 10 inches to verify that these breaks did not exceed acceptance criteria. The NRC staff determined that, based on the licensees evaluation, the empirical strainer debris load limits, and in-vessel fiber limits, no break smaller than 10 inches would generate enough debris to cause failure of the sump strainer or failure of the ECCS to perform its function.

The NRC staff finds that Wolf Creek identified the components that have been mitigated or replaced with Alloy 690 material and Alloy 52/152 filler material, which are less susceptible to PWSCC than the Alloy 600/82/182 material. In addition, Wolf Creek identified in-scope RCS components and welds that are being monitored. The NRC staff finds that Wolf Creek will follow the inspection and monitoring requirements of the ASME Code,Section XI, ASME Code Cases N-722-1, N-729-X, N-770-2, and inspections based on industry operating experience. Therefore, the monitoring of the in-scope RCS piping as part of defense-in-depth is acceptable.

RCS Leakage Detection Capabilities The NRC staff noted that nuclear plants have leakage detection systems to monitor leakage from the RCS. The concept is that cracks in steel pipes will result in leakage that increases over time before a rupture occurs. The leakage detection systems will notify the control room operators who can take timely corrective actions to shut down the plant to prevent catastrophic failures. In section 2.3.2 of attachment IX to its LAR (Reference 1), the licensee stated that the leak detection program at Wolf Creek is capable of early identification of RCS leakage in accordance with RG 1.45 [Guidance on Monitoring and Responding to Reactor Coolant System Leakage, May 2008 (Reference 37)] to provide time for appropriate operator action to identify and address RCS leakage. The licensee stated that the effectiveness of this program is not reduced by the proposed licensing basis change to the risk-informed approach for GSI-191.

The NRC staff noted that based on operating experience, the RCS leakage detection systems in some nuclear plants do not provide the same capability as originally designed. The NRC staff questioned whether any changes made to the current RCS leakage detection systems at Wolf Creek result in differences from the description in the USAR, and whether the capability of the sensors and instrumentation of the current RCS leakage detection systems at Wolf Creek is consistent with RG 1.45, Revision 1.

In the supplement to its LAR (Reference 2), the licensee stated that it has not changed the RCS leakage detection systems from the USAR. The licensee stated that the RCS leakage detection systems at Wolf Creek were designed to RG 1.45, Revision 0, published in 1973, and that the equipment used for the RCS leakage detection systems is capable consistent with RG 1.45, Revision 1.

The NRC staff noted that Wolf Creek TS 3.4.13, RCS Operational Leakage, contains an LCO which requires, in part, no pressure boundary leakage, a maximum of 1 gallon per minute (gpm) unidentified leakage, and a maximum of 10 gpm identified leakage. If leakage outside the LCO limits occurs in the in-scope RCS pipes, the licensee is required to perform corrective actions within specified times in accordance with TS 3.4.13.

The NRC staff evaluated the Wolf Creek RCS leakage detection systems in section 5.2.5 of the Wolf Creek USAR, Revision 34, 2021 (Reference 38). Section 5.2.5 of the USAR states that one of the design bases for the RCS leakage detection systems is that for leaks of 1 gpm or greater, other than identified leakage sources, the reactor coolant boundary leakage detection systems are designed to detect leaks and determine the leakage rate (in accordance with RG 1.45 and 10 CFR Part 50, Appendix A, General Design Criterion 30, Quality of reactor coolant pressure boundary). A comparison with the RG requirements is provided in USAR table 5.2-6.

The leakage detection equipment is also designed to continuously monitor the environmental conditions within the containment so that a background level is identified, which is indicative of

the normal level of leakage from primary systems and components. Significant upward deviation from normal containment environmental conditions provides positive indication in the control room of increases in leakage rates.

The identified leakage is piped to the reactor coolant drain tank whose level is indicated and alarmed in the control room. For example, the annular gap between the O-rings in the reactor vessel head flange is tapped and piped to a temperature indicator and then to the reactor coolant drain tank. Reactor coolant leakage from this source gives a high temperature indication and alarm. Additionally, the controlled leakage shaft seal system for the RCPs is monitored by reactor coolant drain tank level indication and alarm.

For the unidentified leakage, the leakage detection system consists of the sump level and flow monitoring system, the containment air particulate monitoring system, the containment cooler condensate measuring system, and the containment humidity monitoring system. The sump level and flow monitoring system indicates leakage by monitoring increases in sump level. The containment cooler condensate measuring system and the containment humidity measuring system detect leakage from the release of steam or water to the containment atmosphere. The air particulate gas monitoring system detects leakage from the release of radioactive materials to the containment atmosphere. The containment gaseous radioactivity monitor provides additional indication of leakage if significant reactor coolant gaseous activity is present from fuel cladding defects. Section 5.2.5 of the Wolf Creek USAR provides additional details of the RCS leakage detection systems.

The NRC staff determined that the RCS leakage detection systems as described in section 5.2.5 of the Wolf Creek USAR are consistent with RG 1.45, Revision 1 and the LCO of TS 3.4.13 are consistent with the STS. Therefore, the NRC staff finds that the RCS leakage detection systems and TS requirements provide adequate defense-in-depth to ensure that the structural integrity of the in-scope RCS piping is monitored.

RCS Over-Pressure Protection The licensee stated that at Wolf Creek, the RCS over-pressure protection is provided by means of pressure relieving devices, as required by Section III of the ASME Code. The RCS system is also protected from overpressure at low temperatures by the Low Temperature Overpressure Protection (LTOP) System. The NRC staff finds that the RCS piping is protected from overpressure using the relief and safety valves. In addition, the LTOP system has specific limits on the pressure in the RCS system as part of the defense-in-depth effort to mitigate any potential for leakage or failure of the RCS piping.

3.3.2 Debris Generation The licensee referred to approved guidance used in support of the debris generation analysis such as NEI 04-07 and the associated NRC SE. In table 9-1 of attachment IX of its LAR, the licensee highlighted aspects related to conservatism such as:

The debris generation analysis does not take credit for shielding within the zone of influence (ZOI) by equipment (e.g. steam generators, RCPs) and large piping. The main loop piping breaks did not credit shielding by the pressurizer wall. Shielding by these components would reduce debris generation.

Failure of 100 percent of the unqualified coatings inside containment as particulates, for all breaks, is a conservative assumption considering that much of the unqualified coating may not fail or may fail as chips.

All unqualified coatings outside of the reactor cavity are assumed to fail at the start of recirculation. They would actually fail gradually and may not transport to the sump strainers after spray is secured.

The ZOIs assumed for loop piping breaks were grouped by loop resulting in the overprediction of debris for some breaks.

3.3.3 Debris Transport The licensee stated that debris transport analysis was performed in accordance with NRC-approved methods in NEI 04-07 and cited the following conservatisms in the risk-informed analysis and as discussed in table 9-1 of attachment IX of its LAR (Reference 1):

When CSS is operating, all fine debris is assumed to wash down to the sump pool elevation with no holdup on structures; however, some fine debris would be expected to be retained on walls and structures above the containment pool due to incomplete spray coverage and hold up on structures. This contributes to an overestimate in the amount of fine debris reaching the strainer. When spray is not operating, a high washdown fraction of 10 percent was used.

All fine debris in the pool is assumed to transport to the strainer surface. It is expected that debris will be trapped in stagnant pools. Therefore, the amount of fine debris reaching the strainer is overestimated.

The erosion fraction of large and small pieces of fibrous debris used a high estimate resulting in greater amounts of fine fiber predicted to reach the strainer.

3.3.4 Chemical Effects The licensee stated that the chemical effects analysis was performed in accordance with NRC-approved guidance in TR WCAP-16530-NP-A (Reference 19), which includes the following levels of conservatism and as discussed in table 9-1 of attachment IX of the licensees LAR:

One hundred percent of chemical species in solution are assumed to precipitate after the solubility limit has been reached. Some breaks would not result in precipitate formation, and some would result in less precipitate than assumed in the analysis.

The maximum pH was assumed for chemical release and the minimum pH was assumed for solubility. This results in an overprediction of precipitate amounts.

All insulation debris was assumed to be available for reaction in the sump pool even though a significant amount of insulation would be held up above the pool. This contributes to an overprediction of the amount of aluminum available to precipitate.

3.3.5 Strainer and Pump Failure Evaluations Strainer testing guidance has been developed to ensure that headlosses predicted from testing reasonably represent the most limiting values for the plant conditions being tested. The guidance also directs that the application of the test results be performed conservatively. The licensees test program used the maximum debris loads for all debris types, except fiber, that could be generated by any break smaller than the threshold break size.

The licensee stated that strainer headloss testing was performed in 2016 at Alden Research Laboratories (Alden) in accordance with the NRC guidance, NRC Staff Review Guidance Regarding Generic Letter 2004 02 Closure in the Area of Strainer Head Loss and Vortexing, dated March 2008 (Reference 39), which included the following several levels of conservatism and as discussed in, attachment IX, and sections 3.f and 3.g of attachment VIII of the licensees LAR (Reference 1):

All breaks equal to or larger than the threshold break size are assumed to cause strainer failure and core damage regardless of the amount of debris generated by the break.

Many breaks at or greater than the threshold break size would not necessarily generate enough debris to cause strainer failure. The simplified approach using the threshold break size overpredicts the number and type of breaks that would cause strainer failure and increase the CDF.

The threshold break size was based on the concept of only one type of debris exceeding the tested amounts to define a strainer failure state. One type of debris may exceed the tested amount while several other debris types are below the tested amounts and a strainer failure will not occur, yet in the risk-informed analysis that debris load scenario is postulated to cause strainer failure and core damage. With this approach, the number of breaks causing strainer failure are overpredicted.

Miscellaneous debris is assumed to transport and arrive at the strainer before all other debris types, thus reducing the active strainer area for the remaining debris. Some miscellaneous debris would not transport or would transport later to the strainer. This assumption effectively results in lower allowable debris limits for the strainer and contributes to overpredicting the number of breaks causing failures.

The threshold break size determined for the most limiting equipment configuration was assumed to be applicable to all equipment configurations. For debris loads and headloss, single train failure was assumed (debris is assumed to accumulate in a single strainer), while for in-vessel, failure of both CS pumps was assumed under two functional trains (two active strainers would allow more fiber bypass and accumulation inside the vessel). This results in an overprediction of the failure scenarios. This point also relates to the Core Failure Evaluation in section 3.3.6.

Secondary side break scenarios that require recirculation are assumed to automatically result in failures independently of debris amounts. It is likely that secondary side breaks would not result in failures because of the low strainer flow rates and debris loads. This assumption would result in increased failure predictions.

In the evaluation of the effect of gas voids on pump performance, it was assumed that all voids that form transport to the pumps without compression. The pump net positive

suction head (NPSH) required was adjusted for gas voids using conservative RG 1.82 guidance. In practice, some voids would collect in the strainer and vent back to the containment; transported voids would compress thus reducing their volume.

No credit was taken for containment accident pressure when calculating NPSH margin or voiding due to degasification. This results in a reduction in NPSH margin, partially due to the overprediction of void formation.

The design basis sump temperature was used for all break scenario evaluations. Sump temperatures would be lower for smaller breaks. Therefore, chemical amounts and degasification are overpredicted.

The pump and strainer failures were evaluated using several conservative methods as listed in Table 9-1 of attachment IX and section 3.f and 3.g of attachment VIII of the LAR (Reference 1).

3.3.6 Core Failure Evaluation The core failures were evaluated using several conservative methods as listed in table 9-1 of attachment IX of the licensees LAR such as:

Fiber penetration testing, and the correlation used to predict fiber penetration ignored the effect of particulates and strainer loads of small and large pieces of fibrous debris on the filtration of fine fibers by debris-laden strainers. The penetration of fiber fines would be reduced by the combined effects of fiber and particulate loads on strainers. Therefore, the extent of fiber penetration is overpredicted in the in-vessel fiber accumulation analyses.

During fiber penetration testing, the number of strainer disks was reduced to increase spacing between adjacent disks to prevent the bridging of debris. Bridging is expected to occur at the plant strainer which could decrease fiber penetration. This test design overpredicts the extent of fiber penetration.

Fiber limits developed for core blockage are based on bounding tests and analyses. It is likely that significantly larger quantities of debris could accumulate in the reactor core without adversely affecting cooling. Therefore, states of core failure based on these limits would be overpredicted.

All breaks were evaluated against the core fiber limits for a hot-leg break, while cold-leg breaks result in lower debris amounts transported to the core. Therefore, the corresponding analyses would overpredict core failures.

All of the ECCS pumps (i.e., RHR, safety injection pump, and centrifugal charging pumps) are assumed to start taking suction from the sump at the time when the RHR pumps automatically switch over. There is actually a delay for some pumps. This increases the time during which only the CS pumps are assumed to inject from the RWST and maximizes the recirculation duration when the fiber that penetrates the strainer transports to and accumulates in the reactor via the ECCS which is swapped to recirculation first. Therefore, in-vessel fiber load is overpredicted.

3.

3.7 NRC Staff Conclusion

Regarding Key Principle 3: Safety Margins The NRC staff concludes that the licensees evaluation of defense-in-depth satisfactorily addresses the defense-in-depth philosophy, as outlined in RG 1.174, because the RCS piping is monitored for degradation, the RCS leakage detection systems monitor RCS leakage, and the LTOP and Code relief valves ensure RCS piping will not be overpressurized. The NRC staff further concludes that the selected pipe break locations satisfy the guidance of GL 2004-02 and are, therefore, acceptable. The defense-in-depth evaluation also provides information regarding the safety margins associated with the risk-informed analysis.

The NRC finds that the RCS piping considered is fabricated or mitigated with material that is resistant to cracking such that catastrophic pipe breaks would not likely occur. If cracking does occur, the RCS leakage detection system will be able to detect leakage, and the operator will take corrective actions in accordance with the requirements of the Wolf Creek TS. The NRC staff determined that the subject piping maintains defense in depth and safety margin because it satisfies the regulations of 10 CFR 50.55a, GDC 1, Quality standards and records; GDC 14, Reactor coolant pressure boundary; GDC 30; and GDC 31, Fracture prevention of a reactor coolant pressure boundary. The NRC staff concludes that the use of a risk-informed approach would not result in any changes to the response requirements for plant personnel during a LOCA. Therefore, the NRC staff concludes that the piping considered in the debris generation analysis maintains sufficient safety margin to minimize the potential for a large-break to significantly affect the containment sump performance.

The NRC staff notes that the debris limits correspond to cases of successful strainer operation.

Debris loads may exceed the limits, and the strainer may still be successful. In other words, there is margin in the debris limits to correspond to cases of strainer failure. The debris limits are conservative. Identifying cases of strainer failure based on these limits would overestimate the CDF.

The licensee used approved guidance such as TR WCAP-16530-NP-A, the March 2008 NRC guidance on strainer headloss and vortexing, and NEI 04-07 to develop tests and analyses concerning debris generation, debris transport, chemical effects, and headloss testing. Some of the conservatisms highlighted by the licensee may not have contributed significantly to the risk values in the submittal individually, but when combined the margin is regarded as significant.

The NRC staff concludes that the proposed approach maintains safety margins and that the licensees evaluation included independent margins that help assure that the analysis results in a conservative prediction of risk associated with the impact of debris on LTCC.

3.4 Key Principle 4: When Proposed Changes Result in an Increase in Risk, the Increases Should be Small and Consistent with the Intent of the Commissions Safety Goal Policy Statement This section discusses the licensees base PRA model for Wolf Creek, including the calculated total risk values (CDF and LERF) for each unit, and the licensees risk-informed assessment of debris. A review of this information was necessary to determine whether the risk attributable to debris is very small and consistent with the Commissions Safety Goal Policy Statement.

3.4.1 Acceptability of the Base PRA Model Regulatory Position C.2.3 of RG 1.174, Revision 3, establishes that the scope, level of detail, and technical adequacy (technical elements) of the PRA are to be commensurate with the

application for which it is intended, and the role the PRA results play in the integrated decision process.

The acceptability of the PRA is commensurate with the safety implications of the change being requested and the role that the PRA plays in justifying that change. That is, the more the potential change in risk or the greater the uncertainty in that risk from the requested change, or both, the more rigor is placed into ensuring the acceptability of the PRA.

The objective of the NRC staffs review of the Wolf Creek base PRA model was to determine whether the PRA used in evaluating the risk attributable to debris was of sufficient scope, level of detail, technical elements, and plant representation for this application. The licensee stated in section 2.5 of attachment VII to its LAR, that the Wolf Creek Internal Events PRA was developed and is maintained in accordance with the ASME/ANS PRA standard (Reference 30) and RG 1.200, Revision 3 (Reference 29). The licensee also stated that the internal events PRA model satisfies Capability Category (CC) II requirements and non-conforming aspects of the model were addressed with respect to GSI-191 risk assessment. Therefore, the licensee concluded that the Wolf Creek internal events PRA meets the requirements of RG 1.200, Revision 3, and therefore is acceptable to support the assessment of the risk of internal events associated with GL 2004-02.

The Wolf Creek internal events PRA model was used in a very limited extent to support the risk-informed GSI-191 analysis. The use of the PRA was limited to:

Determining the overall (baseline) internal events, internal flooding, and internal fire events CDF and LERF.

Calculating the frequency of secondary side breaks inside containment (SSBIs) that require ECCS recirculation.

Calculating the change in LERF (LERF) for SSBIs that require ECCS recirculation.

Calculating the change in LERF (LERF) using the conditional large early release probability associated with a large LOCA that results in core damage.

The NRC staffs review focused on the above uses of the licensees PRA models to evaluate the risk of debris in containment. It is noted that the licensee did not use the PRA models, but separate external hazard analyses, and bounding estimates of seismic effects to compute corresponding CDF and LERF contributions to the baseline or total CDF and LERF.

3.4.1.1 Scope of the Base PRA (Modes/Hazards)

Regulatory Position C.2.3.1 Scope of a Probabilistic Risk Assessment to Support an Application, in RG 1.174, Revision 3, states:

The scope of a PRA is defined in terms of the causes of initiating events and the plant operating modes it addresses. The causes of initiating events are classified into hazard groups, which are defined as groups of similar hazards that are assessed in a PRA using common approaches, methods, and likelihood data for characterizing the effect on the plant.

Although all plant operating modes and hazard groups should be addressed, a qualitative treatment of some modes and hazard groups may be sufficient when the licensee can demonstrate that its risk contributions would not affect the decision. However, when the risk associated with a particular hazard group or operating mode would affect the decision being made, it is the Commissions policy (SRM to SECY 04-0118, Plan for the Implementation of the Commissions Phased Approach to Probabilistic Risk Assessment Quality, dated October 6, 2004 (Reference 40)) that, if a staff-endorsed PRA standard exists for that hazard group or operating mode, the risk will be assessed using a PRA that meets that standard.

The licensee stated that the PRA is compliant with RG 1.200, Revision 3. The PRA was peer reviewed and determined to be appropriate for use for risk-informed applications. The few non-conforming aspects of the model with the ASME/ANS PRA standard were determined to have no impact to the GSI-191 risk assessment. The Wolf Creek PRA model was not modified to support the risk-informed GL 2004-02 evaluation.

The NRC staff reviewed the licensees information regarding the scope of its base PRA and concludes that the risks associated with hazards and operating modes that would affect this application were considered using a PRA that meets the applicable PRA standard. Specifically, the NRC staff reviewed the licensees assessment regarding the scope of the PRA used to support this application and concludes that (1) the at-power risk bounds the shutdown risk of debris because the debris ZOI is either approximately the same, or significantly higher, at full power RCS pressure and temperature, the flow rate required to cool the core is significantly reduced for low power or shutdown modes, and the pressure of LOCA water jets at full power would generate more debris; and (2) the use of the internal events PRA model is adequate because the risk contributions from other external hazards do not affect the evaluation of the risk attributable to debris.

3.4.1.2 Level of Detail of the Base PRA Regulatory Position C.2.3.3 in RG 1.174 states, "The level of detail in the PRA should be sufficient to model the impact of the proposed licensing basis change. The characterization of the problem should include establishing a cause-effect relationship to identify portions of the PRA affected by the issue being evaluated.

The licensee stated that very few aspects of the PRA model do not conform to requirements of the ASME/ANS PRA standard and the NRC staff agrees that those aspects have no impact on the GL 2004-02 evaluation. The NRC staff reviewed the licensees description of its base PRA and concludes that the level of detail in the licensees base PRA is sufficient and consistent with its limited use to evaluate the risk attributable to debris from sump strainer and core blockage failures. The licensee-implemented peer reviews following the ASME/ANS standards and NEI guidance and these reviews did not identify issues that would affect the risk-informed GL 2004-02 evaluation.

3.4.1.3 Base PRA Technical Elements RG 1.200, Revision 3, describes one approach for determining whether the PRA, in total or the parts that are used to support an application, is acceptable such that the PRA can be used in regulatory decision-making for LWRs. RG 1.200 endorses, with comments and qualifications, the use of the ASME/ANS PRA Standard (Reference 30); NEI 00-02, Revision 1, Probabilistic Risk Assessment Peer-Review Process Guidance, dated May 2006 (Reference 41); and

NEI 05-04, Revision 2, Process for Performing Internal Events PRA Peer Review Using the ASME/ANS PRA Standard (Reference 42).

The NRC staff relied on the peer-review findings and reviewed the key assumptions in the licensees PRA in its determination of the acceptability of the technical elements of the base PRA model. The ASME/ANS PRA Standard provides technical supporting requirements in terms of three CCs. The delineation of the CCs within the supporting requirements indicates that the degree of scope and level of detail, the degree of plant specificity, and the degree of realism increase from CC-I to CC-III. Current good practice is indicated by CC-II of the ASME/ANS PRA Standard. Consistent with the guidance in RG 1.200 and RG 1.174 for this application of the Wolf Creek PRA to assess the risk associated with GL 2004-02-related phenomena, the NRC staff considers CC-II to be adequate.

The internal events PRA model was peer reviewed and assessed against RG 1.200, Revision 3, endorsed guidance consistent with NRC Regulatory Issue Summary (RIS) 2007-06, Regulatory Guide 1.200 Implementation, dated March 22, 2007 (Reference 43). The review and closure of finding-level facts and observations (F&Os) were performed by an independent assessment team using an NRC endorsed process. The reviews also met the requirements of NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, dated August 2019 (Reference 44). The licensee stated that there are no open F&Os for the Wolf Creek PRA model as of November 2023. Therefore, the licensee concluded that the model is acceptable for the evaluation of risk associated with GSI-191 phenomena. The licensee further stated that 100 percent of the supporting requirements in the internal events PRA model satisfy CC-II requirements of the PRA standard (table 5 of attachment VII of the LAR).

The licensee stated that the PRA scope and technical adequacy is met for this application because the ASME/ANS PRA Standard requirements are generally met at CC-II.

Based on the licensee-implemented peer reviews following ASME/ANS standards, endorsed NEI guidance, and the CC-II classification of the internal events PRA, the NRC staff concludes that the internal events PRA is adequate to support the risk-informed GSI-191 assessment.

3.4.1.4 Plant Representation Regulatory Position C.2.3.4, Plant Representation in a Probabilistic Risk Assessment to Support an Application, in RG 1.174, Revision 3, states, the PRA results used to support an application are derived from a base PRA model that represents the as-built and as-operated plant to the extent needed to support the application.

That is, at the time of the application, the PRA should realistically reflect the risk associated with the plant.

The NRC staff concludes that the licensees PRA model adequately represents the as-built and as-operated plant to the extent needed to support the GL 2004-02 risk assessment because the licensees PRA maintenance procedures include an ongoing review of design and procedure changes for their impact on the PRA model, and PRA data or inputs are reviewed and updated, as necessary, on a periodic basis.

3.4.

1.5 NRC Staff Conclusion

Regarding the Base PRA Model The NRC staff concludes that the Wolf Creek base PRA model used in support of the licensees GL 2004-02 risk assessment is acceptable (e.g., has the appropriate scope, level of detail, technical elements, and plant representation) to evaluate the risk attributable to debris because the licensee applied approaches consistent with the guidance in RG 1.174, Revision 3, and RG 1.200 Revision 3.

3.4.2 Risk-Informed Approach for Addressing the Effects of Debris on LTCC The licensee implemented simplified computations, using limited information from the internal events PRA, and combined those computations with traditional engineering analyses to estimate the risk attributable to debris. This integrated analysis is referred to as the systematic risk assessment.

3.4.2.1 Scope of the Systematic Risk Assessment This section describes the specific approach used by the licensee to determine relevant initiating events for which debris could adversely affect the CDF or LERF. This includes how relevant scenarios (i.e., an initiating event followed by a plant response leading to a specified end state, such as event prevention, core damage, or large early release) that could be mitigated by the activation of sump recirculation were identified and considered in the systematic risk assessment.

In attachment VII to its LAR (Reference 1), the licensee provided information regarding the scope of its systematic risk assessment that employed a screening process to eliminate scenarios that were deemed not relevant, not affected by debris, not requiring sump recirculation, or having an insignificant contribution based on the identified failure modes.

Screening is a common practice in quantitative risk assessments, and one acceptable approach is discussed in NUREG-1855, Volume 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk Informed Decision Making, dated March 2009 (Reference 45).

Specifically, NUREG-1855, Volume 1, describes assessment of model and completeness uncertainty, including the identification of sources of uncertainty that are not related to either the parts of the PRA used to generate the results or the significant contributors to the results, and the use of screening and conservative analyses to address non-significant contributors.

RG 1.174 recognizes that a screening approach allows the detailed analysis to focus on the more significant contributions. Information pertaining to the licensees initial plantwide and location-specific screening approach is described in the following subsections.

3.4.2.1.1 Initial Plant-Wide Screening The initiating events considered in the licensee risk-informed analysis included those with the potential to (1) generate debris inside containment, (2) require sump recirculation for mitigation of the event, and (3) result in debris transport to the containment sump (section 2.1 of attachment VII to Reference 1).

The licensee considered only scenarios that required recirculation through the ECCS or CSS strainers, since without recirculation, there is no potential for debris-related failures of the strainers, pumps, downstream components, or core. The licensee considered the following

initiating internal events and excluded them from GSI-191 risk assessment because they do not generate debris inside containment:

Steam generator tube rupture Interfacing systems LOCAs that discharge outside containment Anticipated transients including inadvertent safety injection (SI), inadvertent or stuck open power operated relief valves that discharge to the pressurizer relief tank, and loss of offsite power Secondary side breaks outside containment Initiating events due to loss of component cooling water, loss of service water, and loss of alternating current or direct current power The licensee excluded the following hazards, also because they do not generate debris or generate minor debris amounts inside containment:

Internal and external flood Internal fires and fire-induced LOCAs Aircraft impacts Lightning strikes Local intense precipitation High winds The licensee excluded initiating events based on (1) small equivalent break sizes, (2) reduced flows required to compensate the water inventory compared to equivalent flows in strainer tests, and (3) location of breaks away from significant insulation sources.

The licensee evaluated water hammer induced LOCAs to demonstrate that their risk is acceptably low. The licensee discussed that the RCS piping is designed for loading under normal and transient conditions. The licensee stated that the RCS is maintained water solid during operation so that only an attached system could induce a water hammer event. The attached systems are charging, SI, and RHR. The licensee has a program to address gas accumulation in systems, which was made in response to GL 2008-01 Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, dated January 11, 2008 (Reference 46). The Wolf Creek response was accepted by the NRC in a letter dated October 28, 2009. (Reference 47). The licensee also stated that a search of the corrective action program did not identify any water hammer induced LOCAs. The licensee concluded that the relevance of water hammer events in the context of its analysis is irrelevant because the piping is robust and designed to ASME III, Class 1 standard, no previous water hammer induced LOCAs were identified, and the plant has implemented a program to control gas accumulation, which is a large contributor to water hammer events. The licensee also concluded that LOCA frequencies are not impacted at Wolf Creek due to water hammer considerations.

Secondary side breaks inside containment (SSBIs) were included in the risk analysis by performing a simplified bounding analysis that assumed that all SSBIs that require ECCS recirculation cause failure due to the effects of debris. These scenarios were assumed to result in core damage, independent of debris amounts. The licensee asserted that this is a conservative assumption because SSBIs would generate less debris than equivalent primary breaks due to lower pressure on the secondary side, and flow rates through the strainer required for feed and bleed cooling are significantly lower than the ECCS flow rate for a large-

break LOCA. The licensee calculated the CDF contribution from SSBIs equal to 3.9x107 yr-1, considering the stated conservative assumptions. In addition, the licensee calculated LERF associated to SSBIs equal to 6.2x1010 yr-1.

For seismic events, the licensee used representative LOCA fragility parameters from table H-1 of Electric Power Research Institute (EPRI) Report 3002000709, Seismic Probabilistic Risk Assessment Implementation Guide, Final Report December 2013 (Reference 48). To validate the generic EPRI-based results, the licensee also calculated site-specific LOCA fragility parameters following the guidelines in NUREG-1903, Seismic Considerations for the Transition Break Size, dated February 2008 (Reference 49). In both approaches, the fragility functions were convolved with site-specific seismic hazard curves to estimate the annual frequency of exceedance of various-sized LOCAs. The licensee assumed that the conditional probabilities for LOCAs due to support failures at STP were applicable to Wolf Creek based on similarity in RCS design. The licensee excluded seismic-induced small and medium LOCAs because they were not found to cause failures due to debris effects. For large-break LOCAs, the licensee conservatively assumed failure of all breaks, including those smaller than the threshold size of 10 inches. The resulting large-break seismic LOCA frequency was estimated to be 6.2x107 yr-1 using the EPRI 3002000709 parameters, and 5.7x107 yr-1 using the site-specific NUREG-1903 approach. The licensee conservatively used the higher value (6.2x107 yr-1) in its risk analysis.

The licensee concluded that the only initiating events of relevance to the GSI-191 risk-informed assessment were (1) small, medium, and large LOCAs due to pipe breaks, including seismically induced LOCAs, and (2) SSBIs that cause a consequential LOCA (e.g., due to failure to terminate SI, loss of auxiliary feedwater, or a stuck open power operated relief valve) and require sump recirculation. The NRC staff reviewed the licensees screening approach and concluded that the approach is technically sound and consistent with state-of-practice approaches. Furthermore, the NRC staff concludes that the results of the plantwide screening adequately reflect initiating events relevant to the licensees systematic risk assessment of GL 2004-02 phenomena.

The licensee also provided justification for not aggregating the results from the initiating events included in the risk-informed assessment, noting varying levels of conservatism and bounding methods used to calculate independent contributions to the CDF from the different initiating events. The NRC staff concludes that this approach is in alignment with the guidance in section C.2.5 of RG 1.174, Revision 3 (Reference 15), and NUREG-1855 Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making (Reference 45).

3.4.2.1.2 Location-Specific Screening For LOCA events, the effects of debris may be dependent on the location of the initiating event.

Therefore, the licensee completed a location--specific analysis to identify accident sequences that could be adversely impacted by debris. The licensee stated that non-pipe LOCAs were not explicitly evaluated because it is reasonable to assume that these breaks would be bounded by the pipe breaks analyzed at each Class 1 weld location. The licensee stated that the non-pipe locations in the reactor cavity were not close to welds that contribute risk in the analysis. This exception is discussed below.

Breaks at the reactor nozzle were not included in the risk-informed analysis. This issue is discussed in section 3.a.1 of attachment VII to the licensees LAR. The NRC staff accepted this

exclusion based on the inability of the pipes to separate enough to result in a jet that could damage insulation. The NRC staff discusses this in more detail in section 3.4.2.6.1 of this SE.

The licensee stated that jets resulting from control rod drive mechanism (CRDM) housing breaks or ejections would not cause damage to insulation and were screened from the analysis.

Similarly, breaks at the instrument penetrations at the bottom of the vessel were screened because the jet would be directed toward the bottom of the reactor cavity and would not result in significant debris generation. During the regulatory audit (Reference 16), the NRC staff requested that the licensee provide additional information to justify these assumptions. The licensee provided clarifying information on the electronic portal. The NRC staff also referred to its SE for Amendment No. 228 dated October 21, 2022 for Callaway Plant, Unit No. 1 (Callaway) (Reference 50), which found that these breaks would not contribute to debris generation. The Callaway and Wolf Creek plants have the same reactor design so the analysis is applicable to both plants. The NRC staff discusses this in more detail in section 3.4.2.6.1 of this SE.

The licensee proposed that any LOCA break greater than or equal to the threshold break size would be assumed to result in core damage. The threshold break size (10 inches) was selected because strainer testing and in-vessel analysis showed that, based on the applicable limiting equipment configuration, breaks smaller than the threshold would not produce enough debris to cause any type of failure. All breaks smaller than the threshold break size were considered to result in successful operation of the ECCS strainers and acceptable LTCC. The licensee also examined the headloss from the tested debris load to exclude failure from other causes (for example deaeration, flashing, or mechanical collapse).

The debris loading used to establish the threshold break size was determined by testing (see section 3.4.2.8.3 of this SE). The fiber fines debris from all breaks less than the threshold size was also shown to result in acceptable in-vessel fiber accumulation amounts. To determine the amount of debris generated from each potential break location, break size, and break orientation, the licensee developed a CAD model of containment (referred to as the BADGER model) to keep track the distribution of debris sources in the system, and to compute debris generation based on the ZOI concept. The licensee used NRC approved guidance and computational fluid dynamics (CFD) models to compute debris transport fractions associated with processes such as blowdown, washdown, pool fill, and recirculation. The licensee applied those transport fractions to debris generated computed using BADGER to estimate debris generated, transported, and loaded on the strainer. The CAD and BADGER model included the locations of each potential break location (welds in the RCS) and locations of debris sources that could be damaged by a LOCA jet.

Based on initial screening results, the licensee performed a quantitative analysis of LOCA RCS pipe breaks ranging from half-inch partial breaks to double ended guillotine breaks (DEGBs) on every Class 1 ISI weld inside the first isolation valve. The licensee explicitly considered ISI welds within the first isolation valve in its risk-informed analysis. The licensee concluded that breaks outside the first isolation valve are risk insignificant, due to similarity of debris amounts for breaks upstream and downstream of the valve combined with the dependence of failure of that valve (e.g., failure to close, spurious opening, or leaking valve) for a LOCA to arise (section 2.6.3 of attachment VII to the licensees LAR (Reference 1)). Also, the licensee examined SSBIs (e.g., a large break in a main steam or feedwater line) using bounding analyses considering the IE PRA model to compute the frequency of SSBIs.

The location-specific screening process refined the quantitative analysis to breaks in ISI welds in the unisolable portion of the Class 1 pressure boundary (i.e., inside the first isolation valve),

and SSBIs. The NRC staff reviewed the licensees location-specific screening evaluation and concludes that the licensee identified all locations that could result in a failure of the ECCS LTCC and CSS functions, because the full spectrum of possible break locations was considered and systematically assessed for potential effects on the calculation of debris amounts generated and transported to the sump.

NRC Staff Conclusion Regarding the Scope of the Systematic Risk Assessment The NRC staff reviewed the scope of the systematic risk assessment and finds it adequate, because the licensee employed a systematic screening process using initial plantwide and location-specific screening approaches to identify relevant scenarios and eliminate scenarios that do not affect the GL 2004-02 risk assessment, in a manner consistent with state-of-practice approaches described in NUREG-1855, Revision 1. The licensee included all scenarios and initiating events relevant to the GL 2004-02 evaluation.

3.4.2.2 Initiating Event Frequencies The licensee implemented a simplified computation of the CDF, based on the large LOCA exceedance frequency of the threshold break size determined via log-linear interpolation of the geometric mean (GM) LOCA frequencies in NUREG-1829, table 7.19 (25-year frequencies)

(Reference 14). Since the threshold break size was established at 10 inches, very small, small, and medium break LOCAs do not contribute to plant risk. Those breaks would not generate enough debris to affect the performance of strainer, nor produce enough fiber to accumulate inside the vessel preventing core cooling.

The exceedance frequency for the 10-inch threshold break size was calculated to be 6.6 x 107 per year using the 25-year GM aggregated LOCA frequencies from NUREG-1829.

The licensee did not use the IE PRA model for the CDF computation. For sensitivity/uncertainty and margin analyses, the licensee used LOCA frequencies in NUREG-1829, Volume 1, 25-year GM values for the 5th and 95th percentile. The licensee also performed a sensitivity study using the 25-year arithmetic mean (AM) values and 40-year GM values.

The guidance in NUREG-1829 contains 25-year or current LOCA frequencies and 40-year or end of license period LOCA frequencies. For most LOCA types, the 40-year values are slightly higher due to anticipated aging effects and the possibility of new degradation mechanisms. In some cases, however, the 40-year values are lower, reflecting an expectation that improved mitigation techniques will lower LOCA frequencies. Wolf Creek was initially licensed in 1985, with its licenses renewed in 2008. Wolf Creek has been operating for about 40 years. The licensee stated that the 25-year frequencies are applicable until the plant exceeds 40 years of age. (This anniversary occurred prior to issuance of the amendment.) The NRC staff noted that the age of the plant is currently over 40 years, and during the audit, requested that the licensee discuss why the 25-year frequencies from NUREG-1829 were used in the risk-informed analysis instead of the 40-year frequencies. The licensee provided updated values during an audit and in its supplement (Reference 2) considering 40-year frequencies, with CDF = 1.52x106 1/yr. The licensee stated that although the CDF is in RG 1.174 Region II, the analysis includes enough conservatisms and margin to conclude the overall risk to be very

small, in Region III. Some of these conservatisms are described in section 3.0 of Attachment IX of the licensees LAR (Reference 1).

The NRC staff found acceptable the use of 25-year frequencies in the baseline analysis and consideration of 40-year frequencies in sensitivity analyses. The NRC staff has examined the overall analysis and concludes there are conservatisms and margin embedded in the analysis, and that a more realistic analysis would reduce CDF estimates.

The NRC staff also finds acceptable defining the Base Case using the GM aggregation frequencies, which is consistent with the STP GSI-191 pilot assessment (Reference 11) and risk-informed submittals for Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle)

(Reference 51) and Callaway (Reference 50), and recommendations in NUREG-1829.

The licensee summarized CDF results in table 11 of attachment VII of the licensees LAR (Reference 1).

During the audit (Reference 16), the NRC staff questioned why the licensee considered it appropriate to use the 25-year LOCA frequencies in its Base Case. The NRC staff asked the licensee to discuss the effect that using the 40-year LOCA frequency values in the Base Case would have on the analysis. The NRC staff also requested that the licensee discuss whether the use of the 40-year frequencies would move the proposed change into Region II of RG 1.174 figures of merit. The NRC staff also asked the licensee to discuss any plans to update the analysis to use the 40-year LOCA frequency values in the future.

In its supplement to the LAR, the licensee used GM aggregation 40-year LOCA frequencies to compute a CDF value of 1.5x106/yr and a LERF value of 4.2x1011/yr. The 40-year 5th and 95th percentiles define a range from 6.7x109/yr to 5.2x106/yr for CDF, and 1.9 x1013/yr to 1.5x1010/yr for LERF.

In its supplement, the licensee also stated that using the 40-year LOCA frequency would move the proposed change slightly over the threshold into Region II of the RG 1.174. However, most of the parameters used in the analysis use either conservative or bounding values and the calculated risk values are very conservative. The licensee concluded that since the CDF based on the 40-year GM aggregation LOCA frequencies is barely in Region II, it is still reasonable to conclude that the risk associated with the effects of debris is very small (i.e., in Region III of RG 1.174).

The licensee also stated in its supplement that Wolf Creek entered its period of extended operation on March 11, 2025. Because the analysis uses conservative parameters, the licensee stated that it does not intend to update the analysis even though the 40-year anniversary has passed.

Based on the above, the licensee concluded that the proposed change results in very small risk.

The licensee also examined the effect of assuming every break to be a DEGB (as opposed to the baseline approach of considering breaks of variable size less or equal than the pipe diameter) and concluded that the corresponding CDF is less than the baseline CDF, because the equivalent DEGB size based on the pipe inner diameter of failing breaks is greater than 10 inches. The licensees approach for evaluating the impact of the aggregation method (AM and GM aggregation) and the continuum break and DEGB alternatives is consistent with the recommendation in NUREG-1829. The NRC staff reviewed the licensees uncertainty and

sensitivity analysis, with results in section 2.6 of attachment VII to the LAR, and concluded that the licensee identified and dispositioned key assumptions and sources of uncertainty in its systematic risk assessment consistent with the guidance in RG 1.174. The NRC staffs review of this topic is discussed in section 3.4.2.9 of this SE.

NRC Staff Conclusion Regarding Initiating Event Frequencies The NRC staff reviewed the licensees information on initiating events and concludes that the initiating event frequencies selected by the licensee for this evaluation are acceptable because:

LOCA break frequencies were based on NUREG-1829, which is the most current source of information available.

The licensee interpreted and used the NUREG-1829 data in a manner consistent with the guidance in NUREG-1829 and precedents addressing GL 2004-02.

The licensee performed sensitivity analyses to address the selection of LOCA frequencies from NUREG-1829 using the AM and GM aggregated frequencies, continuum break and DEGB models, and 25-year and 40-year frequencies.

Even if the use of the 40 year frequencies pushed the results into Region II of RG 1.174, this would be acceptable because the baseline plant risk reported in the LAR (5.90x105/yr CDF and 8.92x107/yr LERF) are below the thresholds identified in RG 1.174 (1.0x104/yr CDF and 1.0x106/yr LERF).

3.4.2.3 Scenario Development For the purposes of this SE, the term scenario means an initiating event followed by a plant response such as a combination of equipment successes, failures, and human actions leading to a specified end state, such as successful event mitigation, core damage, or large early release.

The licensee considered system response to a LOCA break in the presence of debris. Strainer testing was used to define debris amounts to postulate strainer failure or in-vessel failure. The licensee determined that no break of 10-inches or smaller would result in a scenario failure (i.e.,

exceed debris test loads with successful strainer performance) and assigned 10 inches as the threshold break size.

The Wolf Creek system includes two independent ECCS trains connected to two independent sumps with strainers. Each train includes an RHR pump, which supplies suction to SI and CS pumps. In the risk-informed analysis, the licensee assumed only one train would operate when evaluating strainer failure, and that two trains would operate when evaluating in-vessel failures.

The assumption of a single train in service maximizes debris amounts loaded on a single strainer, which provides a more challenging strainer condition. The assumption of both trains in service for the in-vessel scenarios provides the most strainer area and lowest strainer fiber loads, which permit the largest amount of fiber to penetrate the strainer and maximize fiber accumulation in the reactor vessel.

The licensee used the BADGER model to compute generated debris for postulated breaks using ZOI concepts, independently of train and pump configurations. The licensee applied

offline computations using Excel, considering constant transport fractions, and fractions of small and large pieces of fiber eroding as fiber fines and transporting 100 percent to strainers, to compute generated and transported debris amounts loading the strainers. The licensee selected constant transport fractions based on CFD computations, which depend on flow rates, pump configurations, location of debris sources, and obstacles that could capture, for example, small and large pieces of fibrous debris. The licensee used the BADGER model outputs of debris generation for postulated breaks, and the fractions of transported debris, and eroded and transported fibrous debris for one-and two-train systems, to demonstrate that breaks smaller than 10 inches would not exceed successful strainer test loads.

The licensee implemented independent mass-balance computations considering only fiber fines in the pool after a postulated break, a two-train configuration, strainer fiber penetration and fiber shedding from fiber accumulated on strainers, and conservative flow rates to compute fiber accumulation per fuel assembly to compare to limits in TR WCAP-17788 (Reference 12). The licensee assumed two operational trains of ECCS and considered a design basis case where a single CS pump was operating and a beyond design basis case where no CS pumps were operating. The operation of two strainers with ECCS and minimizing the amount of flow to the CSS tends to maximize fiber build-up in the vessel (more fiber would penetrate two operational strainers, and less fiber is recirculated back to the pool by the CS system). The in-vessel mass-balance computations were complementary to the BADGER debris generation computations. In those computations, the licensee assumed an amount of fiber fines initially in the pool equal to the full load strainer test, which bounded amounts of fiber fines generated by breaks up to the threshold break size.

The licensee excluded other initiating events associated with different scenarios (see section 3.4.2.1.1.1 of this SE).

NRC Staff Conclusion Regarding Scenario Development The NRC staff evaluated the licensees scenario development process and results and concludes that the licensee adequately evaluated the relevant scenarios potentially causing strainer failure and in-vessel fiber buildup. The licensee considered models greatly simplifying the description of the system response. The licensee used a systematic process to identify germane operating components and states and properly considered the period of performance in the risk-informed analysis. The NRC staff concluded that the licensee consideration of a single ECCS strainer configuration for strainer debris bed buildup analysis and two functional strainers for in-vessel buildup analysis is adequate because it maximizes the probability of system failure.

3.4.2.4 Failure Mode Identification The following are potential debris-related failure modes for the ECCS LTCC function. Each of these failure modes should be considered and specifically evaluated, or shown to be irrelevant, to the risk-informed evaluation. Other potential failure modes should be evaluated, as necessary, for plant-specific conditions. The licensee evaluated each of the failure processes below and did not identify additional failure modes (section 2.0 of attachment VII to the licensees LAR (Reference 1). These failure modes are only those related to debris.

a. Flow paths upstream of the strainer would not be sufficiently blocked to prevent water from reaching the strainer.

b. Strainer headloss does not result in negative pump NPSH margins.

c. Strainer headloss does not exceed the strainer structural limits.

d. Strainer headloss does not result in a void fraction at the pump suctions that exceeds the acceptance criteria given in NEI 09-10.

e. Strainer headloss does not result in flashing immediately downstream of the strainer.

f.

Strainer headloss does not exceed half of the submergence depth of a partially submerged strainer. This consideration is not applicable for Wolf Creek because the strainer is fully submerged even for a small break LOCA.

g. Strainer flow conditions do not result in air ingestion due to vortexing.

h. Blockage and wear of components downstream of the strainer do not exceed the limits given in WCAP-16406.

i.

In-vessel fiber loads and other relevant parameters do not exceed the limits given in TR WCAP-17788.

The licensee evaluated relevant failure modes in its risk-informed analysis in attachments VII and VIII to the licensees LAR (Reference 1). The licensee evaluated failure modes by inspection of the containment to assure flow paths would remain open considering the potential for debris blockage. Additionally, the licensee has programs that control maintenance and design changes to assure these activities will not result in upstream blockage (section 3.i of attachment VIII to the licensees LAR. The licensee evaluated items b through e by confirming that the greatest amount of debris that is predicted to be generated and transported by the threshold break size would not cause strainer failure due to excessive headloss. Headloss testing with conservative amounts of debris was used to verify that headloss induced failure modes would not occur (section 3.f of attachment VIII of Reference 1). All scenarios not bounded by the threshold break debris amounts were conservatively assumed to cause failure, core damage, and contribute to CDF. The strainer tests were successful strainer performance tests and established that the limiting debris amounts would not cause insufficient NPSH, mechanical collapse, or degassing and excessive void fractions. The licensee showed that failure mode f did not apply to Wolf Creek, because limiting sump level calculations indicate that the strainers will be submerged under all conditions that require recirculation (section 3.f and 3.g of attachment VIII of Reference 1). The licensee concluded that failure mode g, vortexing, would not occur based on bounding testing (response 3.f.3 in attachment VIII of Reference 1). Failure mode h, ex-vessel downstream effects, was concluded not to occur (response 3.m in attachment VIII), based on guidance in WCAP-16406-P-A Revision 1 (Reference 21). The licensee concluded failure mode i (in-vessel downstream effects) would not occur based on fiber penetration testing and bounding mass-balance computations concluding that fiber accumulation inside the vessel per fuel assembly would not exceed acceptance limits (response 3.n in attachment VIII of Reference 1).

NRC Staff Conclusion Regarding Failure Mode Identification The NRC staff evaluated the licensees analysis and compared the licensees failure modes to those established by the NRC staff and determined that the failure modes evaluated by the licensee include all those that could reasonably lead to debris-induced failure of LTCC.

Therefore, the NRC staff concludes that the licensee included the appropriate failure modes in its evaluation.

3.4.2.5 Changes to the Base PRA Model The licensee used the Wolf Creek PRA model of record as the source of the base CDF and LERF values. Also, the licensee used the internal events PRA for a bounding assessment of SSBIs. The licensee stated that the Wolf Creek PRA model was not modified to incorporate initiating events for the GL 2004-02 risk-informed analysis. The licensee performed the risk quantification outside the PRA model, and conservatively assuming specific equipment configurations.

NRC Staff Conclusion Regarding Changes to the Base PRA Model The NRC staff reviewed the information provided by the licensee and concluded that the use of the Wolf Creek PRA model of record, without changes, is acceptable to provide supplementary information required by the risk-informed assessment implemented by the licensee.

3.4.2.6 Debris Source Term Submodel This section describes the debris that may be generated during an initiating event or may be present in the containment prior to the event. It includes a description of the debris types and characteristics that may transport to the strainers and affect the ability of the ECCS and CSS to perform their functions. Additionally, this section evaluates the parts of the deterministic analyses that deal with debris source term to determine whether the licensee used appropriate inputs to the risk-informed analysis.

The licensee conducted strainer headloss tests that included amounts of debris that bounded the debris generated by LOCA initiating events equal to or greater than the threshold break size.

The individual debris amounts of each type of debris included in the testing bounded the amounts predicted to be generated and transport to the strainer for LOCA breaks smaller than the threshold break size. The licensee assumed that all breaks greater than or equal to the threshold break size cause strainer failure and core damage and contribute to the increase in CDF. The licensee included debris accounting for low density fiberglass, dirt/dust, coatings, and chemical debris in its strainer test program. Table 3.f.4-2 in attachment VIII to the licensees LAR contains the debris amounts at the test scale. The NRC staff verified that the test amounts bound the amounts predicted to transport by breaks of sizes smaller than the threshold break size. See section 3.4.2.8.3 of this SE for further discussion on this topic.

The licensee conducted a risk-informed debris generation evaluation that considered the sources of debris that may affect system performance. The risk-informed debris generation evaluation included thousands of postulated breaks of different location, size, and orientation at all welds that could result in a LOCA. The evaluation considered 395 welds with breaks of varying size and orientation at each weld. The licensee verified that no break smaller than the threshold break size was computed to generate and transport more of any debris type than was included in the strainer test. All breaks greater than or equal to the threshold break size were conservatively assigned to strainer failure and core damage.

3.4.2.6.1 Break Selection This section describes the licensees process to identify the break sizes and locations that present the greatest challenge to post-accident sump performance. The licensee provided a summary of the break selection process and the method to address debris generation and ZOI in sections 3.a and 3.b of attachment VIII to its LAR (Reference 1). The licensee also considered other potential initiating events (debris generation locations). Some of these initiating events were excluded from the systematic risk assessment or were not explicitly considered in the break selection process, as discussed in section 3.4.2.1 of this SE The licensee stated that the debris generation calculation was performed using the methodology of NEI 04-07 and the associated NRC SE (References 17 and 18, respectively). However, instead of focusing on limiting breaks, the licensee evaluated a full range of breaks, including breaks at all Class 1 pressure boundary ISI welds (section 3.4.2.1). The licensee also considered secondary side breaks inside containment (large main steam and feedwater line breaks). The licensee conservatively assumed secondary line breaks that required recirculation to lead to failure scenarios, independently of generated debris amounts.

The licensee implemented a simplified risk-informed analysis relying on the BADGER software, considering a range of potential break sizes and orientations on welds in Class 1 piping. The BADGER model uses CAD models for Wolf Creek to identify weld locations and insulation and qualified coating distributions. Debris amounts are computed in the BADGER software based on a ZOI concept. Debris sources that are not break-dependent, such as latent debris and unqualified coatings debris, were included in the debris generation model.

The licensee assembled a 3D CAD model of the Wolf Creek containment building, tracking the as-built insulation configuration and qualified coating distribution. The CAD model was used as input to the BADGER software to calculate debris quantities for each weld, and for each break size and break orientation. BADGER implemented a ZOI concept to compute the debris amounts for each debris type. The ZOI represents the zone or volume in space where a two-phase jet from a HELB can generate debris that may be transported to the sump. The size of the ZOI is defined proportional to the break size and is empirically determined based on the system pressure and the destruction pressure of the insulation material impacted by the jet.

Higher system pressures result in increased ZOIs. Robust insulation materials have smaller ZOIs than fragile materials. The licensee considered each circumferential butt weld as a postulated break location, with partial breaks1 of different sizes up to the pipe diameter. Breaks equal to the pipe diameter were considered DEGBs.2 For each partial break size, the licensee considered different orientations of the hemispherical ZOI, in 45-degree angular increments, to evaluate a range of debris sources located around a break. Partial break sizes were evaluated at discrete values equal to 0.375, 0.5, 2, 4, 6, 8, 10, 12, 14, 17, 20, 23, and 26 inches (section 3.a.1 in attachment VIII of Reference 1). The amounts of different types of generated debris were computed and compiled in a database output by BADGER.

Strainer testing included debris amounts that bound the maximum amounts of debris for all breaks greater than or equal to the threshold break size. The licensee listed four 10-inch breaks (i.e., a specific break location, size, and orientation) that produced the largest debris loads 1 The licensee defined a partial break as a break of diameter less than the pipe diameter. The ZOI was assumed to be of hemispherical shape for partial breaks and centered at the outside of the pipe circumference (section 3.b.1 of attachment VIII to the LAR).

2 A DEGB is a break of size equal to the pipe diameter, with a full pipe offset. The ZOI was assumed to be spherical and centered at the axis of the pipe at the break location (section 3.b.1 of attachment VIII to the LAR).

(largest amount of fiber fines, total fiber, and coating particulates on strainers), in table 3.b.4-1, of attachment VIII to its LAR, and verified that those breaks do not exceed the acceptance criteria. The licensee also listed the minimum, average, and maximum debris amounts computed to be generated by small, medium, and large DEGBs in table 3.a.3-1, of attachment VIII to its LAR. This table shows the range of generated debris amounts for each debris type that BADGER predicted for small, medium, and large LOCAs; the table does not account for erosion of small and large pieces of fiber into fiber fines. In the BADGER analyses, the strainer fibrous debris amounts include ZOI-dependent (debris from insulation and qualified coatings) and ZOI-independent (latent debris and unqualified coatings) amounts. The ZOI-independent debris amounts are assumed to be the same for all breaks.

The licensee excluded reactor nozzle breaks, CRDM ejection, and instrument nozzle breaks from the analysis. This is discussed further in the debris generation section of this SE (section 3.4.2.6.2).

NRC Staff Conclusion Regarding Break Selection The NRC staff concludes that the break selection evaluation is acceptable because the licensee evaluated all welds on ASME Code Class 1 pipes that can result in a LOCA. Although the NEI 04-07 guidance approved by the NRC states that the licensee should evaluate all pipe locations for potential rupture, the staff concludes that the licensees evaluation of piping only at welds is acceptable because the weld locations adequately represent the potential debris generation of all breaks and are more likely break locations, consistent with recommendations in NUREG-1829.

The NRC staff concludes that the break selection process and criteria are acceptable because it identifies a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated as part of an acceptable evaluation model as required, in part, by 10 CFR 50.46.

The NRC staff concludes that the licensee has provided sufficient information as requested by GL 2004-02 and is further described in the revised content guide for GL 2004-02 concerning the break selection criteria. In its submittal the licensee:

Described and provided the basis for the break selection criteria used in the evaluation.

Examined the potential contribution of secondary line breaks inside containment (e.g.,

MSLs and feedwater lines).

Discussed the basis for reaching the conclusion that the break size(s) and locations chosen to present the greatest challenge to post-accident sump performance.

The licensee provided a basis for the use of the break selection process in the overall evaluation of change in risk due to LOCAs. The licensee evaluated all Class 1 welds as potential break locations. The NRC staff concludes that the break selection methodology is acceptable to support estimates of risk.

3.4.2.6.2 Debris Generation and ZOI Submodels The licensee defined the ZOIs for NUKON' (low density fiberglass (LDFG)), using NRC-approved guidance in NEI 04-07 (GR/SE) (References 17 and 18). The licensee also used

a debris specific ZOI as allowed by the GR/SE. The licensee adopted debris size distributions for NUKON' based on testing. The licensee calculated the amount of debris for each debris type and fibrous debris size distributions that could be generated from each postulated break.

To compute the frequency of sump strainer failure, the licensee compared the computed strainer debris loads for each postulated break less than 10 inches (the threshold break size) to test debris limits and verified that computed debris loads did not exceed those test limits. The 10-inch threshold break size was selected based on strainer tests and evaluation of in-vessel debris limits. The licensee conservatively computed the frequency of sump strainer failure equal to the exceedance frequency of a 10-inch LOCA break, considering NUREG-1829 LOCA frequencies.

The licensee also defined ZOIs for antisweat insulation, FOAMGLAS, fire barrier material, and lead blankets. For antisweat insulation, FOAMGLAS, and fire barrier material, the licensee assigned the largest ZOI specified in the GR/SE, even though these materials were not specifically addressed in those documents.

The ZOIs for each material are summarized in table 3.b.1-1, ZOI Radii for Wolf Creek Insulation Types, in attachment VIII to the licensees LAR. In the following discussion, the symbol D is used to represent the break size. The ZOI is assumed of hemispherical shape for partial breaks on circumferential welds, and of spherical shape for DEGBs on circumferential welds. The size of the ZOI is defined by the radius (L) of the sphere or hemisphere, and expressed as multiples of the break size D. More robust materials have higher damage pressures and smaller ZOI radii.

The licensee credited robust barriers to shield potential debris sources from LOCA jets as allowed by the GR/SE.

NUKON' For its risk-informed analysis, the licensee used a 17D ZOI radius for NUKON', which is consistent with the guidance in NEI 04-07 (References 17 and 18). For NUKON', the licensee used a centroid model to estimate the debris size distribution (amounts of fiber fines, small fiber, large fiber, and intact fiber blankets). The centroid is the average distance from the break to the multiple insulation locations within the ZOI. Closer to the break, the debris produced is mainly fiber fines and small fiber pieces. Further from the break, the debris is mostly large pieces of fiber. Some insulation is considered to remain intact (within its protective cover) if far from the break, even if within the ZOI. The licensee defined fractions or percentages for a four-category fiber size distribution (e.g., 10 percent intact fiber blankets, 20 percent large fiber pieces, 30 percent small fiber pieces, and 40 percent fiber fines) as functions of the centroid distance.

The fractions were selected to be consistent with the NEI 04-07 guidance. The licensee used a similar approach to Vogtle, which was found acceptable by the NRC staff in the staff evaluation dated September 30, 2019 (page 47 of Reference 51).

Thermo-Lag Fire Barrier Material The licensee used a 28.6D ZOI for Thermo-Lag. This is the largest ZOI defined for any material in the GR/SE (References 17 and 18). The licensee assumed Thermo-Lag to generate fine fibrous debris and fine particulates. The licensee stated that the fire barrier material is both covered/jacketed and uncovered. The covered fire barrier was stated to be shielded by robust barriers and would therefore not become debris. The uncovered Thermo-Lag was all assumed to be rendered into fine debris for all LOCA breaks.

FOAMGLAS The licensee used a 28.6D ZOI for FOAMGLAS. This is the largest ZOI defined for any material in the GR/SE (References 17 and 18). The size distribution for FOAMGLAS was assumed to be 100 percent fine particulates. However, the licensee predicted that no FOAMGLAS would become debris because it is not installed within any ZOI.

Antisweat Material The licensee used a 28.6D ZOI for antisweat material. This is the largest ZOI defined for any material in the GR/SE (References 17 and 18). The size distribution for antisweat material was assumed to be 100 percent fiber fines. However, the licensee predicted that none of this material would become debris because it is not installed within any ZOI.

Cerablanket The licensee did not define a ZOI or size distribution for Cerablanket material. However, the licensee predicted that none of this material would become debris because it is not installed within any ZOI.

Min-K The licensee concluded Min-K blankets cannot be impacted by HELB jets. Min-K is installed within the reactor annulus and could be damaged by a jet from a reactor nozzle break. The NRC staff found this conclusion to be acceptable due to limited lateral offset and no separation of the piping at postulated breaks in the reactor cavity, which results in no jet impingement on these materials. The pipe restraints prevent pipe separation. The evaluation of the potential for debris generation also cited the geometry of equipment in the cavity that would preclude a jet from striking the insulation should one occur (References 52 and 53, Issue 7). Based on this description and the conservatism from treating all material within the ZOI as fines, the NRC staff determined that the licensees debris generation analysis for Min-K is acceptable.

The licensee stated that CRDM ejections and instrument penetration on the reactor vessel would not result in significant debris generation. The licensee stated, in attachment VII to the LAR, section 2.3.1, that CRDM ejections and instrument penetration breaks would result in jets that are not directed toward debris sources. The NRC staff asked clarifying questions regarding the potential for debris generation from breaks at these locations during the audit. The licensee provided clarifying information. The potential for debris generation from these locations was also addressed by an analysis that is also applicable to the Callaway Plant (Reference 52 and 53, Issue 7). The NRC staff concluded that these locations would not result in the generation of significant debris. In its supplement to the LAR (Reference 2), the licensee stated that the insulation vessel top head consists of NUKON blankets.

The licensee stated that the area of miscellaneous debris (signs, tags, placards, etc.) identified via walkdowns was 7.1 square feet (ft2). However, the licensee assumed a conservative surface area of 20 ft2. This results in an effective sacrificial strainer area of 15 ft2 in the analysis, because the GR/SE allows a reduction in this area for overlap of miscellaneous debris.

Risk-Informed Analysis The licensee followed the debris generation calculation methodology specified in NEI 04-07 and justified adopted deviations from the guidance. The licensee evaluated a full range of breaks instead of assessing only the limiting breaks, as recommended in NEI 04-07. All unisolable welds within the Class 1 ISI pressure boundary (i.e., welds inside the first isolation valve) were evaluated, including DEGBs and partial breaks. To evaluate thousands of break scenarios, the licensee used a CAD model and the BADGER software to describe the insulation configuration, qualified coating distribution, and location of robust barriers within the containment, and to automate computation of insulation and qualified coating amounts within the ZOI of each postulated break. The licensee calculated debris amounts for breaks in each circumferential weld, ranging from 3/8 inch to the full pipe diameter, and considered multiple orientations for each break size and location in 45° increments. For DEGBs, a spherical ZOI was assumed centered at the axis of the pipe in the plane of the weld. For partial breaks, the ZOI was a hemisphere oriented normal to the pipe axis and centered at the edge of the pipe at the break location (section 3.b.1 of attachment VIII to the LAR (Reference 1). The licensee applied simple transport fractions, and fractions of fibrous debris eroded and transported as fiber fines, in offline computations to establish strainer debris loads associated with LOCA breaks, including latent debris and unqualified coatings. The licensee used BADGER outputs and offline transport fractions to verify that breaks less than 10 inches would cause strainer debris loads less than test limits, as described in section 2.4.1 of attachment VII of its LAR. The licensee defined 10 inches as the threshold break size; and assumed that all breaks greater than or equal to 10 inches would lead to core damage. The licensee demonstrated that LOCA breaks smaller than 10 inches would be successfully mitigated.

The BADGER code algorithms have been evaluated previously by the NRC staff as part of the Vogtle staff evaluation of its response letter related to GL 2004-02 (Reference 51), supported by audits and independent calculations sponsored by the NRC staff to explore the adequacy of the BADGER software to identify insulation and coating sources and compute generated debris by postulated LOCA breaks. The NRC staff concluded that use of a detailed CAD model in BADGER, and offline transport and erosion fractions, is a reliable approach to quantify potential debris amounts within the ZOI of a postulated break, and to quantify strainer debris loads.

The licensee assumed that constant amounts of unqualified coatings would detach immediately after the LOCA event and form particulate debris that could be transported to the strainer for any break. The licensee also assumed that unqualified coatings protected by insulation (protected unqualified coatings) would only be damaged and transported to the strainer if the insulation covering the coatings is within the ZOI for that insulation type. The protected unqualified coatings were therefore assumed to contribute to the debris source term for each LOCA break scenario only if they were within the covering insulations ZOI of the break. The licensee considered the presence of 140 pounds (lbs) of latent debris, of which 15 percent was assumed in the form of fibrous debris (21 lbs) and 85 percent in the form of particulate (119 lbs).

The estimated total amount of particulate from unqualified coatings was 1.24 cubic feet (ft3). The unqualified coatings values are provided in table 3.h.1-2 of attachment VIII of the licensees LAR. Except for the protected unqualified coatings, these values were constant for all the postulated breaks, independent of the radius of the ZOI.

The licensee also considered ZOI-dependent sources of particulates such as qualified inorganic zinc and qualified epoxy. The amounts of qualified coatings and protected unqualified coatings from the four worst-case 10-inch breaks are provided in table 3.h.5-1 of attachment VIII of the licensees LAR. Table 3.a.3-1 of attachment VIII of the licensees LAR provides amounts of

qualified and protected unqualified coatings for small, medium, and large LOCA DEGBs. For each break size, the minimum, maximum, and average amount of each of these coating types is provided.

For chemical debris, the licensee assumed a bounding amount of chemical debris for all analyzed breaks. For example, the licensee assumed that the amount of NUKON contributing to chemical effects was from the largest LOCA break with the most limiting NUKON amount. This amount is much greater than the largest amount of NUKON from a threshold break. The chemical debris assumptions are described in section 3.o.2.3 of attachment VIII of the licensees LAR Strainer tests were conducted with chemical, particulate, and fiber amounts that bound all breaks smaller than the threshold break size. The licensee evaluated each postulated break size, location, and orientation for the Class 1 welds to assure that the fibrous, particulate, and chemical debris amounts generated and transported from break smaller than the threshold break size did not exceed the equivalent amounts in strainer testing at the plant scale. Strainer failures were assumed to occur for all breaks greater than or equal to the threshold break size.

The NRC staff concludes that the licensee properly quantified amounts of debris that could be generated within the Wolf Creek containment by the postulated LOCA breaks. The analysis included ZOI-dependent (e.g., fibrous debris from different insulation types and qualified coatings) and ZOI-independent (e.g., dust and latent debris, unqualified and degraded coatings) debris. For the ZOI-dependent debris, the licensee computed debris amounts using BADGER, which relied on a CAD model capturing the location and distribution of insulation and debris sources within the containment. Bounding inputs to the chemical debris analysis were also obtained from the CAD model. For each break (of specific location, size, orientation, and ZOI),

BADGER used CAD model information to determine debris amounts for each material type. For each break, the CAD model clipped the ZOI to account for robust barriers. The NRC staff previously concluded that algorithms in the BADGER code for the computation of generated debris were properly implemented (Reference 51). The licensee adequately considered random factors such as the break size and jet orientation and identified debris amounts to compare to strainer tests. The NRC staff concludes that the licensees methodology to calculate debris loads for each postulated break is acceptable.

NRC Staff Conclusion Regarding Debris Generation and ZOI Submodels The NRC staff notes that the licensee considered guidelines in the NEI 04-07 report to (1) define ZOIs; (2) account for robust barriers; (3) compute debris amounts of NUKON insulation and fire barrier material, and particulate sources such as qualified coatings; (4) compute debris size distributions; and (5) estimate debris amounts associated with latent fiber, latent particulate, and unqualified coatings.

The NRC staff verified that the licensees debris generation calculations were performed accurately and used acceptable assumptions. The NRC staff used a combination of confirmatory calculations, engineering judgement, and review of the licensees software outputs to perform the verifications. The Wolf Creek method to compute debris amounts relies on BADGER software, which was examined in detail as part of the NRC staffs evaluation for Vogtle to risk-inform its treatment of the effects of debris on LTCC (Reference 51). This approach allows the NRC staff to conclude, with a high level of confidence, that the calculations for debris generation were conducted and applied properly.

The NRC staff reviewed the licensees evaluation against the NRC staff-accepted guidance and concludes that the licensee adequately determined for each postulated break location, size, and orientation, the zone within which debris would be generated by a two-phase jet. The NRC staff also concludes that the amount and characteristics of debris predicted to be generated are acceptable. The licensee calculated amounts for all types of debris generated by up to 10-inch breaks (i.e., the threshold break size) and compared these values to the limiting amounts established by strainer testing. All breaks equal to or larger than the threshold break size were assumed to lead to strainer or in-vessel failure and contribute to the plant risk. The threshold break size was established by comparing debris amounts to strainer tests. The debris amounts included in the strainer tests bound the maximum amounts of debris from LOCA breaks up to threshold break size. The licensees methods are consistent with NRC guidance. Therefore, the NRC staff concludes that the licensees evaluation of the ZOI and debris generation is acceptable. The amounts of debris from each postulated break scenario were determined appropriately.

The NRC staff concludes that debris generation and ZOI analysis and methodology are acceptable because they identify a number of postulated LOCAs of differing properties sufficient to provide assurance that the most severe postulated LOCAs are calculated. Also, the NRC staff concludes that the debris generation and ZOI submodel described in the licensees LAR is acceptable for use in an assessment or evaluation model of the effects of debris on long-term cooling of ECCS, as required, in part, by 10 CFR 50.46.

The NRC staff concludes that the licensee provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning the debris generation and ZOI, because the licensee:

Described the methodology used to determine the ZOIs for generating debris.

o Identified which debris analyses used approved methodology default values.

o For materials with ZOIs not defined in the guidance report, discussed methods used to determine ZOI and the basis for each or described why the material would not contribute to the debris source term.

Provided destruction ZOIs and the basis for the ZOIs for each applicable debris constituent.

Identified destruction testing conducted to determine ZOIs.

Quantified the amount of each debris type generated for each break location, size, and orientation evaluated.

3.4.2.6.3 Debris Characteristics The licensee generally used the NRC staff SE on NEI 04-07 (Reference 18) to evaluate the debris characteristics.

The licensee listed the debris characteristics for NUKON and Thermo-Lag, in table 3.c.1-1 of attachment VIII to its LAR (Reference 1). Since testing was used to determine the headlosses associated with debris, the microscopic characteristics are not important to the risk-informed evaluation. On the other hand, debris densities are important inputs to the evaluation.

NUKON' For its risk-informed analysis, the licensee used a 17D ZOI for LDFG and NUKON', which is consistent with the NEI 04-07 guidance (Reference 18). The licensee then further analyzed the 17D ZOI using the centroid methodology discussed in section 3.4.2.6.2 of this SE. The centroid distance was used to determine the size distribution of the NUKON' that was damaged within the ZOI.

Thermo-Lag All Thermo-Lag material predicted to be damaged was assumed to be destroyed into 100 percent fiber fines and fine particulates.

The licensee assumed that Min-K did not require evaluation because it does not become a debris source as discussed in section 3.4.2.6.2 of this SE.

Other non-coating and non-latent debris sources did not contribute to the debris source term.

NRC Staff Conclusion Regarding Debris Characteristics The NRC staff concludes that the NUKON debris characteristics were defined per the applicable guidance. The NRC staff concludes that the debris characteristics for the Thermo-Lag were assigned conservatively because the licensee assumed that the Thermo-Lag within the ZOI would be rendered into 100 percent fine debris.

The NRC staff concludes that the licensee provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning the debris characteristics, because the licensee:

Identified debris characteristics using an approved methodology and default values where available.

For the Thermo-Lag that is not defined in the guidance, the licensee assumed that all unprotected material was damaged and rendered into fine debris. These properties conservatively bound the effects of the debris.

Reflective metal insulation (RMI) was not included in debris generation analysis. The licensee stated that RMI is solely installed on the reactor pressure vessel and is shieled by the breaks outside the reactor cavity by the primary shield wall. The NRC staff concluded that the relatively small amount of RMI would not result in increased headloss if transported to the strainer. The NRC staff also concluded that it would be unlikely for the RMI to transport out of the annulus to the strainer.

3.4.2.6.4 Latent Debris The licensee followed the guidance in the GR/SE (References 17 and 18) to evaluate latent debris. A bounding value of 140 pound-mass (lbm) of latent debris was assumed in the analysis, with the recommended 15 percent being latent fiber. The remaining 85 percent of the latent debris was assumed to be particulate debris.

The licensee sampled containment to determine the actual amount of latent debris present, following accepted guidance. The licensee performed sampling that determined that the maximum amount of latent debris in containment is 75 lbm. Based on this, the amount of latent debris was conservatively assumed to be 140 lbm.

The licensee provided the assumed characteristics for the fibrous and particulate latent debris and stated that the characteristics are consistent with the GR/SE.

The licensee stated that 20 ft2 of miscellaneous debris (tags and labels) were assumed to transport to the strainer with 25 percent overlap, per NRC guidance. This results in a total assumed area of 15 ft2. The actual miscellaneous debris amount was determined via walkdowns, which identified 7.1 ft2. Therefore, the amount of miscellaneous debris assumed in the analysis is conservative. The miscellaneous debris is discussed in the licensees response to 3.b.5 in attachment VIII to its LAR.

The NRC staff concludes that the licensee provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning latent debris, because the licensee:

Identified the amounts of latent debris and miscellaneous debris in containment using an approved methodology.

Used conservative values in the headloss analysis compared to the actual values in containment.

Identified latent debris characteristics using an approved methodology and default values.

3.4.2.6.5 Coatings The licensee stated that the qualified coatings debris within the ZOI were treated in accordance with the guidance in the GR/SE (References 17 and 18). The licensee defined its DBA qualified coatings systems used in the containment and provided the coating systems and manufactures of those systems considered to be qualified. Table 3.h.1-1 of attachment VIII to the licensees LAR listed these systems, including pertinent information for each coating system. Table 3.h.5-1 of attachment VIII to the licensees LAR listed the amounts of qualified coatings for the four worst-case breaks that do not fail any acceptance criteria.

The licensee described the unqualified coatings that are within containment. The licensee listed the generic types and properties of unqualified coatings in table 3.h.1-2 of attachment VIII of its LAR. The licensee also provided table 3.h.5-2 of attachment VIII to its LAR that listed the constant amounts of each type of unqualified coating included in the BADGER model for all breaks.

The licensee assumed that protected unqualified coatings, unqualified coatings that are covered by insulation that is not within a ZOI (and therefore assumed to remain intact), would not be available to transport to the strainer. The protected unqualified coatings were stated to be on the steam generators and the pressurizer. Any protected unqualified coatings installed under insulation were assumed to fail and be available to transport if the insulation was within the break ZOI. Coatings installed under insulation outside the insulation ZOI were assumed to

remain in place. Table 3.h.5-1 of attachment VIII to the licensees LAR lists the amounts of protected unqualified coatings for the four worst-case 10-inch breaks.

The licensee discussed the assumptions used for coatings transport. The licensee assumed that the unqualified coatings fail as 10-micron particulates in lower containment, and those are evenly distributed in the pool at the start of recirculation and transport 100 percent to the strainer. The licensee stated that settling is not credited for any fine debris. The licensee also assumed the qualified coatings in the ZOI fail as 10-micron particulate and transport 100 percent to the strainer except a small percentage held up in inactive volumes.

The licensee stated that headloss testing used pulverized acrylic, with a mean size distribution of 10 microns, as a surrogate for qualified coatings and unqualified coating particulates.

The licensee stated, in section 3.f.4 of attachment VIII to its LAR, that paint chips were added to the headloss tests. Paint chips were created using a food processor or blender with a final size between 0.004 to 0.008 inches. The paint chips were stated to represent degraded qualified coatings, but the NRC staff was unable to locate a description of the degraded qualified coatings or determine the amounts or transport assumptions for this debris type. During the audit (Reference 16), the NRC staff requested that the licensee discuss why the coating chips were included in the strainer test. The licensee stated that the chips were added to the test to provide margin in case degraded qualified coatings were discovered within containment at a future time. The licensee stated that the characteristics and transport of any degraded qualified coatings would be evaluated at the time of discovery and compared against the coating chips used in the test. The NRC staff found this response acceptable.

The licensee based the ZOI sizes for qualified coatings on NRC accepted jet impingement testing and staff guidance for reviewing the coatings evaluation. The licensee assumed the ZOI for epoxy topcoat to be 4D. Exposed inorganic zinc coatings were assumed to have a ZOI of 10D. The licensee determined the qualified coating debris amounts by using a 3D model of containment that modeled the orientation of coated surfaces exposed to postulated break jets.

The licensee assumed the coatings in areas within the appropriate ZOIs fail and calculated the debris amounts using the dry film thicknesses and densities for the coating system on the impacted surface.

Except for protected unqualified coatings, the licensee assumed that unqualified coatings in containment fail regardless of location. The quantities of unqualified coatings were calculated similarly to the qualified coatings. As discussed above, tables of unqualified coating amounts were provided. The licensee stated that the unqualified coating amounts are based on logs maintained by the plant.

The licensee stated that the plant conducts coating condition assessments at each refueling outage to ensure that the coatings are performing their intended design function and that they remain qualified. This ensures that the debris source term remains bounded by the design basis assumptions. If the inspection or assessment identifies degradation, further investigation is performed. Inspection reports are reviewed by engineering. The licensee stated that the assessments and resulting repair activities assure that the amounts of coatings that may detach from the substrate during a LOCA is minimized.

NRC Staff Conclusion Regarding Coatings The licensee performed its evaluation in accordance with NRC approved guidance. The guidance includes the GR/SE and subsequent NRC guidance, Revised Guidance Regarding Coatings Zone of Influence for Review of Final Licensee Responses to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors(Reference 54), for coatings evaluations. The debris generation amounts were determined using appropriate ZOIs and the debris volume was properly preserved for testing by correcting for density differences between the coatings and the test surrogates. The transport metrics used for the coatings debris were also based on the approved guidance or plant-specific testing.

The licensee maintains a coatings condition assessment program that appropriately monitors, tests, and repairs coatings as required.

The NRC staff concludes that the licensee provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning coatings, because the licensee:

Identified the amounts of coatings materials in containment that can become debris using approved methodologies.

Used appropriate characteristics for the coatings as applied in the transport and headloss analyses.

Identified an appropriate coatings surrogate for testing and used conversions to assure that the volume of debris is preserved for the headloss tests.

Described a coatings assessment program to provide ongoing inspection and repair of qualified coatings systems.

3.4.2.6.6 Containment Material Control The licensee described four procedures at Wolf Creek that control materials taken into containment and perform inspections to ensure that the sumps are not blocked by debris. The Containment Entry and Material Control procedure includes requirements for transient materials brought into containment and establishes controls for trash and debris that can be generated during maintenance activities. The procedure is applicable to all materials taken into containment and is applicable in all modes of operation. The Housekeeping Control procedure addresses inspections made to ensure that all debris is removed prior to startup. In Modes 1 through 4, the containment is considered a system. In these modes the procedure requires thorough cleaning of the work area and surrounding areas that could have been affected by the activities in containment. This cleaning is followed by an inspection. In Modes 5, 6, and defueled, more general containment cleaning is required. The Housekeeping Control procedure directs the inspections made prior to startup from lower modes.

The Containment Inspection procedure documents the inspection requirements for ensuring that debris will not impair the recirculation sumps from performing their required functions. The Containment Sump Inspection procedure is implemented to ensure that the sump inlets are not restricted by debris.

The NRC staff noted that the licensee is adopting TSTF-567(Reference 32) as part of this LAR, to implement a containment sump specific TS. This TS includes a SR (SR 3.6.8.1) to inspect the sumps, strainers, and flowpaths to ensure that current or potential debris blockage will not occur, and structural damage or abnormal corrosion are not present, to assure operability of the containment sumps.

The licensee described programmatic controls used to ensure that design changes inside containment do not result in unanalyzed debris sources. Materials added to containment are assessed for potential debris generation including that from insulation, coatings, and exposed aluminum.

The licensee developed an engineering specification that defines the design basis for insulation debris amounts for the strainer. The licensee stated that procedures are in place to control maintenance activities and track transient materials in containment. During normal operation, all materials taken inside containment are logged, and then accounted for at the completion of the activity. Items left in containment are either stored in a container with a latchable door or cover or secured to a structural member to prevent transport to the sump.

The licensee stated that procedures are in place to control maintenance activities and temporary modifications that may affect the debris source term. Guidance for design changes is also applied to temporary modifications.

The licensee stated that there are no planned insulation modifications or replacements to reduce the source term at the sump strainer. The licensee stated that there are no actions planned to reduce the debris source term from insulation, coatings, or other sources.

The licensee also cited that flow diverters had been installed in the containment, but that credit was not taken for these to reduce the transport of debris to the strainers. The licensee stated that it is likely that the diverters would result in some reduction in transport.

NRC Staff Conclusion Regarding Containment Material Control The NRC staff concludes that the licensee provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning containment material control, because the licensee:

Identified the programs in place to reduce the introduction of debris during maintenance and modification activities, including temporary modifications.

Described significant actions taken to reduce the debris source term that may occur at the sumps.

NRC Staff Conclusion Regarding Debris Source Term Submodel Each of the aspects of the debris source term was evaluated. The NRC staff concluded that the debris source term submodel, including break selection, debris generation and ZOI, debris characteristics, latent debris, coatings, and containment material control were adequately addressed. Based on the evaluations for each of these subsections, the NRC staff concludes that the debris source term evaluation is acceptable.

The NRC staff concludes that the debris source term submodel (including break selection, debris generation and ZOI, debris characteristics, latent debris, coatings, and containment material control) is acceptable because it identifies a number of postulated LOCAs of sufficiently differing properties to provide assurance that the most severe postulated LOCAs are calculated.

Also, the NRC staff concludes that the debris source term submodel described in the licensees submittal is acceptable for use in an assessment or evaluation model of the effects of debris on long-term cooling of ECCS, as required, in part, by 10 CFR 50.46.

The NRC staff concludes that the licensee has provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning the debris source term.

3.4.2.7 Debris Transport Submodel 3.4.2.7.1 Strainer Transport The licensee stated that the transport evaluation was based on the methods in the GR/SE (References 17 and 18). The blowdown, washdown, pool fill, and recirculation phases were modeled in the evaluation. The licensee used a transport logic tree methodology, to account for blowdown, washdown, pool fill, and recirculation, to calculate the debris amounts reaching the strainer for each break. Erosion of larger pieces of debris was included in the evaluation.

The licensee provided the bounding debris transport fraction values for each phase of the transport evaluation for various break locations. For blowdown, the fractions minimized the amount of debris retained in the break compartments, and for washdown the bounding values maximized the washdown to lower containment. The transport fractions for the phases of transport were provided in tables 3.e.6-1 through 3.e.6-7 of attachment VIII to the licensees LAR (Reference 1).

For pool fill transport, the licensee stated that the transport to inactive volumes was limited to 15 percent as directed by the GR/SE.

For the recirculation phase, the licensee used Flow-3D (CFD software) to perform transport simulations. Settling of fines was not credited in the transport analysis. The recirculation transport analysis compared transport metrics for the different debris types and sizes, determined via testing, to the local pool hydraulic conditions. If the velocity or turbulent kinetic energy in any specific area of the pool exceeded a metric determined by testing, the debris in that region was considered to transport to the strainer. The licensee stated that for the recirculation phase of the transport analysis all particulate fines and fiber fines transport 100 percent and do not settle. The licensee analyzed four cases of recirculation transport. All breaks were modeled in RCS Loop C because that loop is closer to the two open loop door exits and causes greater turbulence in the annulus than the other loops. One small-break LOCA case with two trains was run. Three large-break LOCA cases were run, one with 2 trains operating and one each of a single loop operating on the 1A and 1B trains. Maximum ECCS and CSS flow rates and minimum pool levels were used to maximize the amount of transported debris.

The licensee assigned an erosion fraction of 10 percent to larger pieces of fiber in the pool and 1 percent for fiber held up on gratings. The pool erosion fraction based on 30-day erosion testing and the erosion for fiber held on gratings is justified by the Drywell Debris Transport Study. The 10 percent erosion fraction for fiber fines deviates from the recommended guidance in NEI 04-07 (which recommends a 90 percent erosion fraction for fiberglass debris). The

licensee described that the 10 percent erosion fraction was based on generic 30-day erosion testing. The NRC staff reviewed the testing that was performed by ALION Science and Technology (ALION) Proprietary report ALION-REP-ALION-1006-04, Revision 0, in 2010 (Reference 55). In LDFG erosion testing conducted by ALION, it was determined that small and large pieces of fiber in the sump pool eroded at a rate below 10 percent. The NRC staff reviewed and developed conclusions regarding this report that are documented in a letter dated June 30, 2010 (Reference 56). The NRC staff concluded that plants that could demonstrate the testing was conducted under conditions that represented or bounded their plant could assume a 30-day erosion value of 10 percent for fiber settled in the sump pool. The licensee stated that the generic testing is applicable to Wolf Creek. These erosion fractions are consistent with NRC staff guidance.

The licensee stated that debris interceptors are not installed in the Wolf Creek containment.

However, the licensee stated that flow diverters are installed. The flow diverters are made of perforated plate with 1/8-inch holes, so flow can pass through them. The diverters were not credited with holding up debris even though it is likely that they would capture some debris.

However, they increase the distance that debris would have to travel from a break to the strainers. These barriers were not credited for the capture of debris in the transport analysis. For the recirculation evaluation the flow diverters were assumed to be blocked resulting in diverted flow through open flowpaths to the annulus on Loops B and C. This increases the flow velocity, and turbulence and therefore the transport amounts.

The licensee stated that the overall transport fractions were determined by incorporating the different phases of transport for each break location, which were summarized in tables 3.e.6-8 and 3.e.6-9 in attachment VIII to its LAR (Reference 1). These transport fractions for two-train transport scenarios were revised for fibrous debris as described in the licensees supplement to the LAR (Reference 2).

The licensee used the maximum transport fractions for all break locations evaluated in the analysis. During the regulatory audit, and by conducting independent examination of the transport results, the NRC staff reviewed the calculations and logic trees and confirmed that the calculations were performed correctly, consistent with the transport fractions in tables 3.e.6-8 and 3.e.6-9 in attachment VIII to the licensees LAR, and as revised in the supplement to the LAR.

The debris transport amounts for 10-inch breaks with the greatest amounts of debris were provided in table 3.e.6-10 of attachment VIII to the licensees LAR. The debris amounts shown in this table remain valid as the maximum amounts transported for single train cases because the licensees supplement only revised (reduced) two-train case transport fractions. The single train case is limiting for strainer failure because there is only half the strainer area available for debris deposition compared to having two trains (strainers) in service.

3.4.2.7.2 In-Vessel Transport The licensee calculated the amount of debris that could reach the reactor core after penetrating the strainers and transporting through the ECCS. The fiber penetration rate, as a function of strainer load, was characterized by the licensee based on plant-specific testing for Wolf Creek conducted in 2016. The testing used strainer modules representing a portion of the plant strainer and included scaled amounts of debris based on the ratio of plant to the test strainer areas. Empirical functions, developed from the testing, for fiber penetration rates and fiber shedding as a function of strainer fiber loads were used in the calculation of in-vessel debris

amounts for various ECCS pump configuration and pump flow scenarios. During testing, the licensee used conservative methods to ensure that the penetration was maximized. For example, the spacing between strainer disks and between the disk edges and the test flume wall was increased to ensure that fiber did not bridge between the disks which would provide additional filtration and less penetration. Only fiber fines were used in the testing. The licensee performed a single penetration test using NUKON fiber. The testing was conducted using plant-specific attributes where possible. For example, the water chemistry was prototypical for the plant, as was the approach velocity during the testing. Efforts were taken to ensure that all fiber reached the strainer. Settled fiber was collected and determined to be a negligible amount (1 gram settled out of more than 13,000 grams introduced during the test). The scaled amount of fiber added during the test was much greater than the largest amount of fiber generated by a threshold break size at the limiting location for fiber debris generation.

The licensee developed a low fiber curve and a high fiber curve based on the testing. The low fiber curve assumed that a maximum of 568.8 lbm of fine fiber transported to each strainer. This amount was represented by the first four batches of fiber in the test. The high fiber curve assumed that up to 1068.4 lbm of fiber reached each strainer. The licensee only used the low fiber curve for its in-vessel analysis since it bounded the amount of fiber that could be generated and transported from the limiting threshold break. The low fiber curve also predicted a greater amount of fiber penetration than the high fiber curve. During testing, the low fiber test amount was completed at 22,000 seconds (about 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) from the beginning of the test. Therefore, the licensee stated that the curve fit shown in figure 3.n.1-2 of attachment VIII of its LAR was only used up to 22,000 seconds. This is acceptable because the amount of fiber in the low fiber portion of the test bounded the threshold break fiber amount, and because the fiber penetration decreases as the amount of fiber buildup on the strainer increases.

The licensee performed the in-vessel debris loading calculation for different scenarios with different pump configurations to assure that the maximum amount of debris reaching the core was determined. The analysis found that the safeguards condition with no spray pumps available and two RHR pumps running resulted in the largest amount of debris reaching the vessel. The NRC staff discussed the methodology with the licensee during the regulatory audit because it was not clear that the licensee had assumed the maximum amount of fine fibrous debris that could be transported to the strainer. The licensee stated that the fiber fines initially in the pool considered in the computations was 141 lbm, consistent with strainer tests. For that condition, the licensee computed in-vessel buildup below TR WCAP-17788P limits. The licensee noted that 10-inch threshold size breaks would produce much less than 141 lbm of fiber fines.

The NRC staff considered empirical equations to compute fiber penetration through the strainer for different pump configuration and flow scenarios, and concluded that the in-vessel fiber loads reported in table 3.n.1-1 of attachment VIII of the licensees LAR (Reference 1) are reasonable.

The licensees calculations predicted a maximum in-vessel fiber amount of 92.54 grams per fuel assembly (g/FA) for the limiting design basis case and 94.15 g/FA for a beyond design basis case. The limiting design basis case has two RHR pumps and one CS pump running. The beyond design basis case has two RHR pumps running and no CS pumps in operation. Both cases included maximum fine fibrous debris amounts generated and transported from breaks 10 inches and smaller.

The results of the transport evaluation are important inputs for the strainer and in-vessel evaluations that are further evaluated in this SE.

NRC Staff Conclusions Regarding the Debris Transport Submodel The licensees approach to evaluating debris transport was consistent with the NEI 04-07 guidance and its associated NRC staff SE, and the licensee provided information requested in the content guide for GL 2004-02. For the in-vessel evaluation, the licensees approach followed the NRC staff accepted guidance from TR WCAP-17788-P.

The NRC staff reviewed the licensees transport evaluation against the NRC staff-accepted guidance in NEI 04-07 and verified the consistency of the computed debris amounts. The NRC staff concludes that the licensee appropriately estimated the fraction of debris that would transport from debris sources within containment to the ECCS strainers. Therefore, the NRC staff concludes that the licensees evaluation of debris transport is acceptable.

The NRC staff concludes that the debris transport submodel described in the LAR is acceptable for use in an assessment or evaluation model of the effects of debris on long-term cooling of ECCS, as required, in part, by 10 CFR 50.46.

The NRC staff concludes that the licensee has provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning debris transport, because the licensee:

Used approved and accepted guidance to perform the majority of the calculations.

Provided the technical basis for assumptions and methods used in the analysis that deviate from the approved guidance.

Provided a summary of, and supporting basis for, credit taken for reduction of debris amounts.

Provided the calculated debris transport fractions and the total quantities of each type of debris transported to the strainers.

Provided the limiting amount of fibrous debris that could accumulate in the reactor core.

3.4.2.8 Impact of Debris Submodel This section evaluates the potential effects that the debris, as described in section 3.4.2.6, of this SE, may have on the operation of equipment important to LTCC. This section examines the operation of the ECCS strainer, the ECCS and CSS pumps, and other equipment downstream of the strainer, including the fuel and vessel. This section also evaluates the potential for the holdup of water in containment such that it may not reach the sump pool.

For this section, all descriptions attributed to the licensees submittal are taken from the licensees LAR (Reference 1), and supplement (Reference 2). The majority of the information is from sections 3.f, 3.g, 3.j, 3.k, 3.l, 3.m, 3.n, and 3.o of attachment VIII of the LAR.

3.4.2.8.1 Upstream Effects The licensee evaluated the containment for the potential for blockage or impedance of the transport of water to the sump. In the LAR, the licensee stated that the containment was divided

into seven areas for evaluation of upstream effects. The following areas of containment were specifically listed for consideration in the evaluation:

Upper Containment Operating Floor Annulus and Inside Secondary Shield Refueling Pool Ground floor Inside Secondary Shield Gound Floor, Annulus Reactor Cavity and Instrumentation Tunnel and Sump The licensee evaluated the layout of the structures and equipment and determined that significant blockage would not occur. The licensee used its containment drawings and photos of the containment to ensure that areas would not become blocked with debris and significant water volumes would not be held up during recirculation.

The licensee identified a potential for blockage of the refueling canal drains and installed a strainer designed to ensure the drain will not become blocked and flow of water to the sump from the refueling canal will be unimpeded.

The licensee evaluated the containment for structures that could prevent or delay the flow of water to the sumps. Areas that could retain water were identified.

Some areas of containment, like the reactor cavity (under the vessel) and instrument tunnel were identified as water holdup volumes. Other areas of the containment will not hold up water from reaching the sump. The holdups are accounted for in the sump level calculations.

The licensee stated that the refueling pool has two 10-inch diameter drains in the floor. There are trash rack cages installed over the drains to prevent debris from blocking them. The trash rack cages are 33 inches square and 15-inches high. The cages have five-inch square openings that allow debris that can flow freely through the pipe to pass and will exclude very large pieces that may block the pipe. A photograph of one of the cages is provided as figure 3.l.4-1 of attachment VIII of the licensees LAR.

NRC Staff Conclusion Regarding Upstream Effects The NRC staff reviewed the licensees evaluation against the NRC staff-accepted guidance in the GR/SE (References 17 and 18) and concludes that the licensee has appropriately evaluated the flow paths upstream of the containment sump for holdup of inventory that could reduce flow to the sump and possibly starve the pumps that take suction from the sump. Therefore, the NRC staff concludes that the licensees evaluation of upstream effects is acceptable.

The NRC staff concludes that the licensee has provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning upstream effects, because the licensee:

Summarized the evaluation of the flow paths from the postulated break locations and CS washdown to identify potential choke points in the flow fields upstream of the sump.

Summarized measures taken to mitigate potential choke points.

Summarized the evaluation of water holdup at installed curbs and/or debris interceptors.

Described how potential blockage of reactor cavity and refueling cavity drains has been evaluated, including likelihood of blockage and amount of expected holdup.

3.4.2.8.2 Screen Modification Package The licensee described the new strainers that replaced the original strainers at Wolf Creek.

The new strainers have an area of about 3,311 ft2 per train. There are two strainers that consist of 72 modules each. Each strainer supplies coolant to one train of the ECCS. The new strainers incorporate a uniform flow design and were supplied by Performance Contracting, Inc. In section 3.j of attachment VIII to its LAR (Reference 1), the licensee describes the strainers. At the maximum flow rate, the approach velocity will be less than 0.01 ft/sec. The strainers are installed in separate sump pits that extend below the containment floor. The pits are surrounded by a 6-inch curb. The strainers are constructed of support structures covered by perforated plate. The filtration surface holes are 0.045 inches in diameter. Because the strainers are ruggedly constructed, trash racks were not installed during the modification. The strainers are constructed of stainless steel.

The licensee stated that debris barriers were installed in all openings through the secondary shield wall near the sumps. Debris barriers were also installed in the drain trenches and other openings in the secondary shield wall near the sumps. The debris barriers are designed to restrict the passage of debris to the strainers while allowing water to pass through.

The licensee also stated that the sump level instrumentation was replaced as a result of the new strainer design. The new instruments provide differential pressure indication across the strainer to provide operators with a qualitative indication of strainer performance.

NRC Staff Conclusion Regarding Screen Modification Package The NRC staff reviewed the design changes made by the licensee in response to GL 2004-02.

The information from the design was appropriately included in the licensees submittals. Based on its review, the NRC staff finds the licensee has provided sufficient information, as requested by GL 2004-02, and used appropriate inputs for its evaluation of LTCC, considering the effects of debris because the licensee:

Provided a description of the major features of the sump screen design modification.

Described modifications necessitated by the sump strainer installation.

3.4.2.8.3 Headloss and Vortexing The licensee stated that the headloss and vortexing evaluations were revised based on updated testing completed in 2016. The headloss testing demonstrated that all breaks smaller than the threshold break size resulted in small headlosses so that the strainers would continue to perform their design functions.

The licensee stated that the minimum strainer submergence is 0.03 ft for the small-break LOCAs and 1.17 ft for large-break LOCAs at ECCS sump switchover. The minimum

submergence occurs at swapover, and submergence increases due to continued injection from the RWST via the CSS as discussed in section 3.g.5 of attachment VIII to the licensees LAR.

Details of the strainer submergence calculations were provided in section 3.g.2 of attachment VII to the licensees LAR. The values discussed here are for limiting calculations based on assumptions described in section 3.g.2.

The licensee stated that strainer testing was conducted at Alden Research Laboratories (Alden).

Two tests, a full debris load (FDL) and a thin bed (TB) test were performed. The FDL test used debris loading that bounded the amount of debris that is predicted to be generated and transported by the most limiting threshold breaks. The TB test was a typical high particulate to fiber ratio test. The test description provides an overview of the important aspects of the testing and indicates that NRC staff review guidance in RG 1.82, Revision 4 (Reference 26), was followed. Specifically, the licensee stated that the 2008 Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing (Reference 39) was followed. The test used prototypical plant strainer modules, and the test scaling accounted for area reduction due to miscellaneous debris. The licensee stated that debris amounts in the testing were derived from the amounts predicted to reach the strainer based on the debris generation and transport analyses. These test results demonstrated that all breaks smaller than the threshold break size can be successfully mitigated considering the deterministic criteria for strainer and ECCS performance. For breaks equal to or larger than the threshold break size, the licensee conservatively assumed that strainer failure and core damage occurred in the risk-informed analyses.

The licensee stated that prototypical testing and analyses were used to evaluate the strainer for air ingestion due to vortex formation. During testing the strainer was monitored for vortex formation under clean and debris-laden conditions. During the regulatory audit (Reference 16),

the NRC staff requested that the licensee provide the strainer submergence that was maintained during the testing. The licensee explained that the strainer in the plant is submerged by a minimum of 14 inches at ECCS swap over and that the submergence increases to a minimum of 18 inches due to continued CS injection from the RWST. During vortex testing of the strainer, the submergence was maintained between 14 and 18 inches and was temporarily reduced to 6 inches after debris had collected on the strainer. The strainer approach velocity was maintained at its bounding value during the tests. The licensee stated that they observed vortex formation during TB testing after the sixth batch of fiber debris. The licensee stated that the vortex was only on the surface of the water and did not entrain air. The licensee stated that no air entraining vortex formations were observed during testing and that the test conditions were more likely to produce a vortex than plant conditions.

The licensee used guidance from RG 1.82, Revision 4, to calculate the Froude number (Fr) for the plant condition and the tested condition. The lower the Fr, the less likely vortex formation will occur. RG 1.82 sets a limit of Fr less than 0.25 to ensure no air ingestion and less than 0.5 to ensure air ingestion less than 2 percent. The licensee calculated a Fr of 0.35 for the plant condition using conservative inputs. For example, the minimum small break LOCA submergence was used along with a bounding velocity. In addition to the plant Fr calculation, the licensee compared the test Fr to the plant Fr. The test Fr was calculated to be 0.45 demonstrating that the test condition was more conservative than the plant condition.

During testing no air ingesting vortex occurred. Therefore, the licensee concluded that vortex formation would not lead to air entrainment at the plant. Based on the LAR and the licensees clarifications regarding test conditions the NRC staff concluded that vortexing would not occur at the plant strainer under the design-basis condition.

The licensee provided a description of the tests including the debris surrogates and amounts used in the testing. The fibrous material used in testing was NUKON', which represented NUKON fiber, latent fiber, and fiber from Thermo-Lag. Particulate debris was pulverized acrylic, silica flour and paint chips to represent qualified and unqualified coatings, degraded epoxy, and a dirt/dust mixture to represent latent particulate debris. The debris and surrogates and the preparation of the materials used in testing were described in detail. The licensee stated that no debris settlement was credited during testing. All non-transported debris was captured and reintroduced to the test so that it could transport, or quantified and subtracted from the debris amount added to the test. The test setup had sufficient turbulence to assure that most debris transported to the strainer surface without disturbing the debris bed on the strainer.

For each test, the licensee provided the amounts of each debris type used including the debris generated directly from materials in containment, and chemical debris predicted to form over time as materials dissolve into the sump pool and later precipitate. The licensee also provided plots to describe the debris additions, flows, and temperatures over the course of the test along with the associated headlosses. The test plots were divided into two figures with one depicting the addition of conventional debris and the other showing the effects of chemical debris additions. Each plot also noted the peak headloss during the test for both conventional and chemical debris. The headloss due to conventional debris is applied with all flow through the strainer including ECCS and CSS. The higher headloss due to conventional plus chemical debris is applied to a flow rate resulting only from ECCS operation because it was demonstrated that chemical precipitation would not occur until the CS pumps are secured. The flow rates used in the testing were scaled to bound the maximum potential plant flow rates.

The FDL test headloss was limiting for both conventional and chemical debris. The conventional debris headloss was stated to be 1.487 pounds per square inch differential (psid) at 122.7 degrees Fahrenheit (°F) and a test flow rate of 982.3 gallons per minute (gpm). The FDL test chemical headloss was 1.920 psid at 95.5 °F and a flow rate of 544.4 gpm. The headlosses that the licensee examined in the risk-informed analysis were based on the results of the FDL test.

The test headloss results are presented in table 3.f.4-7 of attachment VIII to the licensees LAR.

The licensee stated that the impacts of the maximum debris amounts that may occur from breaks smaller than the threshold break size are represented by the headloss test results. In section 3.f.5 of attachment VIII to its LAR, the licensee stated that the test strainer arrangement modeled the plant and that the test debris amounts were scaled based on the area of the plant to the test strainer. The licensee also discussed the basis for the total maximum headlosses determined in the analysis. In section 3.f.7 of attachment VIII to its LAR, the licensee states that the maximum test headlosses, adjusted from test flow rates and temperatures to plant conditions, were combined with the clean strainer headloss to determine the total headloss. The tests were performed under limiting flow conditions as discussed above. This results in higher headlosses because the debris bed is compacted more than it would occur at lower flow rates.

Table 3.f.7-1 shows a comparison of the maximum debris amount for each type of debris that could occur from any break smaller than the threshold break size and the amounts of debris included in the TB and FDL tests. The tested amounts in the table were scaled up to the plant strainer size. Note that each plant debris type is the maximum for that particular type that can result from any break smaller than the threshold break size. The table shows that all debris types are bounded by the test amounts. Based on the above, the licensee concluded that the headlosses provided in table 3.f.4-7 of attachment VIII to the LAR are bounding for all breaks smaller than the threshold break size.

The licensee provided the criteria used to determine strainer failure. The criteria considered the headloss acceptance criteria based on the RHR/CS pump NPSH margin, strainer structural limit and air entrainment due to degasification, flashing and vortexing. For debris loads that occur for scenarios up to the threshold break size, none of these criteria were exceeded. Therefore, all breaks smaller than the threshold break size do not contribute to increases in plant risk. During the audit (Reference 16), the NRC staff requested that the licensee provide the debris mass assumed in the structural calculations to assure consistency with the headloss calculations. The licensee provided information during the audit that the largest possible debris load from all breaks up to and including the largest breaks were used in the structural evaluation. The debris mass assumed for the structural calculation is much greater than that assumed for the threshold break size. The licensee also provided the differential pressures used for the structural evaluation and demonstrated that they bound the strainer headlosses determined by testing.

Therefore, the mass and differential pressures used in the structural analysis are conservative.

The licensee provided this information in its supplement to its LAR (Reference 2). The structural evaluation is discussed in more detail in section 3.4.2.8.4 of this SE.

The licensee described conservatisms associated with the testing in section 3.f.8 of attachment VIII of its LAR. The licensee stated that the approach velocity used during testing was greater than expected to occur in the plant due to a higher strainer flow rate based on conservative test flow rates. The licensee also stated that it assumed only one strainer/train of ECCS/CSS in service. This is conservative because two trains would generally be available resulting in double the strainer area to accommodate the same amount of debris. Finally, the debris loading considered the maximum amount of each debris type generated by any break smaller than the threshold break size. This approach combines the maximum amount of all debris types that would be generated by a break smaller than the threshold break size.

For the vortex evaluation, the licensee stated that conservatism in the analysis included using higher flow rates and lower minimum water levels. The licensee also stated that several of the test inputs used were conservative and would not occur simultaneously in the plant. Assuming that they occur at the same time adds conservatism.

The NRC staff recognizes that the assumptions and test practices used by the licensee provide margins that help to ensure that the test results are bounding of the conditions in the plant, for breaks smaller than the threshold break size. In some cases, the margins and assumptions add significant conservatism.

The licensee stated that the clean strainer headloss (CSHL) (strainer headloss with no debris) was calculated by the strainer vendor using test data from prototypical strainers and using standard calculations. The test results were corrected to the plant strainer dimensions and were determined using a flow rate higher than the plant flow rate. The CSHL was stated to be 0.642 ft at 212°F and a flow rate of 8,830 gpm. At 140°F and 8,830 gpm the headloss was calculated to be 0.651 ft. At 140°F and 4,880 gpm the headloss was calculated to be 0.205 ft. These CSHL values were provided in table 3.f.9-1 of attachment VIII to the licensees LAR.

The licensee corrected the headloss values from test conditions to plant flows and temperatures based on the flow sweeps performed during the tests. Both the chemical and conventional headlosses were scaled in the same manner. The headlosses were corrected to the flow rates and the temperatures of the water at the plant conditions considered. These headlosses were added to the clean strainer headloss to determine the total headloss across the strainer at each plant condition. The total strainer headlosses were provided in table 3.f.10-1 of attachment VIII to the licensees LAR. The NRC staff concluded that the scaling was conducted appropriately.

The licensee stated that the strainer is fully submerged during responses to both small break LOCAs and large break LOCAs, and no partially submerged strainer analysis is needed.

The licensee provided a description of the flashing evaluation performed. Flashing would occur if the pressure downstream of the strainer was calculated to be less than the vapor pressure of the fluid at the sump temperature using a containment pressure curve based on DBA conditions. For the flashing evaluation, under limiting conditions, the licensee stated that a credit of 1 pound per square inch (psi) of containment accident pressure was required to suppress flashing. The minimum margin between the containment pressure and the pressure required to suppress flashing was stated to be slightly greater than 1 psi and is present for only a short time. The licensee also provided an example of a more likely, smaller break scenario and stated that analysis of smaller breaks have much greater minimum margins between the containment pressure and the pressure needed to prevent flashing. The results of the limiting and a more realistic scenario are presented in tables 3.f.14-1 and 3.f.14-2, of attachment VIII to the licensees LAR. The licensee stated that the analysis used a model that was biased to maximize sump temperature and minimize containment pressure. The NRC staff reviewed the response and concluded that the credit for the use of containment accident pressure in the flashing analysis included significant margin and is needed for a relatively short time.

The licensee evaluated the degasification of fluid as it passes through the strainer and debris bed. For the degasification evaluation, no containment pressure was credited. The licensee estimated degasification using the midpoint of the strainer for strainer submergence. The licensee assumed that any gasses liberated due to the pressure drop across the strainer transported to the pump suction without compression or reabsorption. The NRC staff determined that the use of the strainer midpoint for the degasification analysis could be non-conservative if degasification did not occur over the entire height of the strainer. The NRC staff asked that the licensee discuss this issue during the audit (Reference 16) to provide additional information justifying the use of the strainer mid-height. During the audit the licensee stated that using the mid-height of the strainer is reasonable considering the degasification analysis contains considerable conservatism. The licensee cited the assumptions of maximum headloss, uniform flow, minimum sump water level, minimal containment pressure, and no compression or reabsorption of gas bubbles as they travel to the pump suction. The licensee provided a table of predicted gas fractions for minimum and maximum safeguards conditions in tables 3.f.14-3 and 3.f.14-4 of attachment VIII of its LAR. The minimum safeguard conditions resulted in greater void fractions. The licensee stated that the NPSH required values were corrected in accordance with RG 1.82 (Reference 27) when determining NPSH margins for the ECCS and CSS pumps. Based on the cited conservatism and the small void fractions calculated, the NRC staff concluded that the degasification analysis is acceptable.

NRC Staff Conclusion Regarding Headloss and Vortexing The NRC staff reviewed the licensees evaluation against the NRC staff-accepted guidance and concludes that the licensee has appropriately determined the headloss across the sump strainer for the debris loads tested. The debris loads tested are bounding and contain margin compared to the maximum debris amounts that can be generated and transported from breaks smaller than the threshold break size. The licensee has shown that the potential for formation of a vortex at the strainer does not exist under the plant-specific conditions at Wolf Creek for debris loads limited by the threshold break size. The licensee has demonstrated that the strainer will perform acceptably under postulated LOCA conditions, limited by the amount of debris that can be generated by breaks up to the threshold break size and represented in the testing.

Therefore, the NRC staff concludes that the licensees evaluation of headloss and vortexing is acceptable.

The NRC staff concludes that the licensee has provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning headloss and vortexing, because the licensee:

Provided the minimum submergence of the strainer under small-break LOCA and large-break LOCA conditions.

Provided a summary of the methodology, assumptions, and results of the vortexing evaluation and bases for key assumptions.

Provided a summary of the methodology, assumptions, and results of prototypical headloss testing for the strainer, including chemical effects and provided bases for key assumptions.

Addressed the ability of the design to accommodate the maximum volume of debris that is predicted to arrive at the strainer for breaks limited by the threshold break size.

Addressed the ability of the screen to resist the formation of a thin bed.

Provided the basis for the strainer design maximum headloss.

Described significant margins and conservatisms used in the headloss and vortexing calculations.

Provided a summary of the methodology, assumptions, bases for the assumptions, and results for the clean strainer headloss calculation.

Provided a summary of the methodology, assumptions, bases for the assumptions, and results for the debris headloss analysis.

Showed that the sump is fully submerged for all accident scenarios.

Stated that near-field settling was not credited for the headloss testing.

Used flow sweep results from testing to scale the results of the headloss tests to actual plant conditions.

Stated that a small amount of the available containment accident pressure was credited in evaluating whether flashing can occur across the strainer surface and summarized the methodology used to determine the available containment pressure.

3.4.2.8.4 Sump Structural Analysis The Wolf Creek structural evaluations were performed using manual calculations and finite element analysis by employing GTSTRUDL and ANSYS software. The support structures for the strainers were analyzed using American Institute of Steel Construction (AISC), Manual of Steel Construction, 7th Edition and supplemented by American National Standards Institute

(ANSI)/AISC N690-1994, Specification for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities; Structural Engineering Institute/American Society of Civil Engineers (SEI/ASCE) 8-02, Specification for the Design of Cold-Formed Stainless Steel Structural Members; and American Welding Society (AWS) D1.6, Structural Welding Code -

Stainless Steel. The AISI Specification for the Design of Cold-Formed Steel Structural Members, 1996 Edition, was used where the ASCE Code does not provide specific guidance.

The perforated plates and core tube end cover plate stiffeners were stated to have been modeled in accordance with appendix A, article A-8000 of the ASME Code,Section III, 1974 Edition through Winter 1974 addenda.

The licensee provided the load combinations for the design of the strainer assembly and its components such as tubing, bracing, channels, angles, cover plates, perforated plates, connections for various components, and baseplate and anchors. The combinations include consideration of dead load, live load weight of debris, differential pressure, and seismic loads including earthquake induced sloshing. The bounding load combinations for the strainer design are provided on table 3.k.1-1 of attachment VIII to the licensees LAR.

The licensee stated that the maximum debris load for all breaks was used for the structural calculation. During the audit, the NRC staff requested that the licensee provide the debris load used in the analysis. The NRC staff also noted that the licensee stated that the maximum thermal stress was combined with the maximum headloss at lower temperatures. However, both the maximum headloss and a lower value were provided as inputs to the structural analysis.

The NRC staff asked the licensee to clarify which pressure or pressures were used in the analysis during the audit (Reference 16).

During the audit, and in its supplement to the LAR (Reference 2), the licensee provided the debris mass used in the analysis. The licensee stated that the 11 disk modules were assumed to have 63 lbm of debris and the 7-disk modules were assumed to have 39.6 lbm of debris collected on the strainer surfaces for the structural analysis. The total debris loading for the strainer used in the structural analysis is about 4,349 lbm. This is significantly greater than the strainer debris mass calculated for the threshold break size limiting debris amount of approximately 1,285 lbm. The NRC staff verified that the mass used in the structural analysis was significantly greater than the mass of debris that results from the limiting threshold break.

Therefore, from a debris mass perspective, all scenarios would result in a strainer headloss failure before a structural failure would occur. The NRC staff concluded that the debris mass assumed in the structural analysis is conservative.

The licensee also discussed the differential pressures used in the structural analysis during the audit and provided clarifying information in its supplement to the LAR. The licensee stated for the hot case that includes temperatures between 268°F and 175°F, the structural evaluation used 4 ft of differential pressure (dP), which is greater than the maximum strainer headloss of about 3.4 ft for the hot condition. For the cold case (less than (<) 175°F), the structural evaluation was performed by iterating the pressure across the strainer to determine the maximum dP that could occur before structural failure. The maximum allowable pressure calculated for the cold case is 5.5 ft. This pressure exceeds the maximum cold headloss of about 4.15 ft. The licensee also provided materials properties tables in the supplement to the LAR for the cold case and hot case strength and allowable stresses of each material used in the strainer assembly

The licensee provided the design interaction ratios for the sump strainer components, which show that at all locations the calculated stress is less than the allowable stress.

Based on the information provided, the NRC staff concluded that structural failures of the strainer would not occur at debris loads or differential pressures caused by the limiting threshold break. The structural analysis contains margin in both the allowable differential pressures and the debris mass assumed.

To address the issue of potential dynamic effects due to a HELB, the licensee stated that the strainers are installed inside the sump pits and are protected from pipe whip and jet impingement from a LOCA because they are outside the secondary shield wall and protected by a concrete slab above.

The licensee stated that a backflushing strategy is not credited in the analysis.

NRC Staff Conclusion Regarding Sump Structural Analysis The NRC staff concludes that the sump strainer is structurally acceptable for the assumed design-basis loads for which it is deterministically qualified. The differential pressure assumed across the strainer is valid for loads up to the threshold break size. The debris mass on the strainer considered in the structural analysis is much greater than that occurring from the limiting threshold break. The NRC staff finds that the licensee has provided the information requested in item k (Sump Structural Analysis) of the NRCs revised content guide for GL 2004-02 Supplemental Responses because the licensee:

Summarized the design inputs, design codes, loads, and load combinations utilized for the sump strainer structural analysis.

Summarized the structural qualification results and design margins for the various components of the sump strainer structural assembly and demonstrated that code allowable stresses are not exceeded.

Demonstrated that dynamic effects such as pipe whip, jet impingement, and missile impacts associated with HELBs are not applicable due to the use of leak-before-break methodology.

Stated that a backflushing strategy is not used at Wolf Creek.

3.4.2.8.5 Net Positive Suction Head The licensee performed the NPSH margin evaluation in accordance with the guidance in the GR/SE (References 17 and 18). The licensee considered the strainer, ECCS, and CS pump flow rates, sump temperatures, and the containment water levels used in its NPSH analysis.

The licensee calculated water levels for the small-break LOCA and large-break LOCA using conservatisms to establish the minimum values for the vortexing evaluation and the NPSH calculations. The licensee provided the assumptions used in the water level calculation in section 3.g.2 of attachment VIII of its LAR.

The licensee provided the flow rates for both the ECCS and CSS in its analysis. The ECCS flow is 4,760 gpm and the CSS flow is 3,950 gpm. To calculate strainer headloss, the total flow was

conservatively assumed to be 9,100 gpm when both systems are operating. For ECCS operation only, the flow rate for strainer headloss is assumed to be 4,900 gpm. The CS pumps are secured after containment pressure is adequately reduced and before the sump temperature reduces enough to allow chemical precipitates to form and contribute to strainer headloss.

The NPSH for Wolf Creek pumps is calculated using the maximum sump temperatures predicted to occur for the limiting LOCA. The maximum temperature is calculated to be about 270°F. For the NPSH calculation, at temperatures greater than 212°F the containment pressure is set equal to the vapor pressure at the analyzed temperature. The limiting NPSH is calculated with the sump temperature assumed to be 212°F because at that temperature there is no subcooling of the fluid and headlosses are greater at lower temperatures due to increased fluid viscosity. Below 212°F the subcooling adds vapor pressure that more than offsets any increase in headloss due to higher water viscosity. The licensee stated that as temperature decreases below 212°F, NPSH margin increases rapidly. The water levels used in the NPSH calculations are the lowest levels calculated for the small break LOCA and large break LOCA. The water level calculations contain conservatism to assure that related analyses contain margin. The assumptions used in the analysis are listed in section 3.g.2 of attachment VIII to the licensees LAR.

The licensee stated that the required NPSH values for the pumps were determined by the pump performance curves provided by the vendor and that these values were determined using the Standards of the Hydraulic Institute. The licensee stated that the RHR pump NPSH required is 21.01 ft at the maximum calculated flow rate of 4,760 gpm. The CS pumps have an NPSH required value of 16.5 ft at 3,950 gpm. The licensee stated that the NPSH required values were corrected for void fractions based on the void fractions calculated to occur at 212°F. The licensee used a method from NUREG-0897, Containment Emergency Sump Performance, Revision 1 (Reference 57), to correct the NPSH required values for the pumps for void fractions.

The NRC staff confirmed that the method used is consistent with that in RG 1.82, Revision 5.

The licensee stated that the bounding pump NPSH margin for the RHR pumps was determined using Fathom software for the ECCS piping frictional losses. The CSS piping was analyzed via a hand calculation using the Darcy-Weisbach method and the Darcy formula. The component headlosses were calculated from standard industry handbooks.

The licensee described the response of the system to large-break LOCAs and small-break LOCAs. The response includes injection and recirculation modes. For a large-break LOCA, the RHR pumps, SI pumps, and centrifugal charging pumps start in response to a SI signal. These pumps all take suction from the RWST and discharge to the RCS cold legs. This is referred to as the injection mode. When the containment pressure reaches about 30 pounds per square inch gauge (psig), the CS pumps inject via spray headers to the containment. The CS pumps also take suction from the RWST during injection. CS injection cannot be stopped until the injection phase is complete. When RCS pressure decreases to below the accumulator pressure the accumulators inject into the RCS cold legs.

When the RWST level reaches a low setpoint, the RHR pump suction sump valves are automatically opened followed by automatic closure of the RHR RWST suction valves. This realigns RHR to the recirculation mode, taking suction from the sump instead of the RWST. The centrifugal charging and SI pumps are also aligned to take suction from the RHR pump discharge instead of the RWST and continue injecting to the cold legs. The switchover to recirculation for the CS pumps occurs when RWST level is further decreased to about

12 percent of indicated level. At that point the operators manually realign the CS pump suctions from the RWST to the containment sump.

After about 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months following the start of the LOCA, the ECCS lineup is changed to hot-leg recirculation. The centrifugal charging pumps continue supplying coolant to the cold legs, but the injection of the RHR and SI pumps is redirected to the hot legs.

For the SBLOCA, the licensee stated that the RCS pressure may stabilize at a relatively high value so that RHR and accumulator injection does not occur. The licensee also stated that it is likely that containment pressure will remain low enough so that CS does not initiate. If the RWST is not depleted, the plant may be cooled down before the RWST is depleted so that recirculation is not required.

The licensee stated that the worst single failure for NPSH margin is the failure of one train of ECCS and CSS leaving only one train of these systems in service both taking suction through a single strainer. The reason that this is the limiting failure is that flow in a single strainer is maximized, which increases headloss and NPSH required. The failure also results in all transported debris collecting on a single strainer and resultant increases in debris headloss.

The licensee provided an overview of the methods used to calculate the sump level. The methods are described in sections 3.g.8, 3.g.10, 3.g.11, and 3.g.12 of attachment VIII to the licensees LAR. The licensee provided the sources that contribute and those that are hold ups when calculating the sump volume. The licensee calculated the sump level using a correlation that outputs sump level based on the volume of water in the sump. The licensee also provided a list of structures and components that will displace water resulting in a higher pool level.

The licensee stated that no containment accident pressure was credited for the NPSH evaluation. For sump temperatures equal to or less than 212°F, a containment pressure of 14.7 pounds per square inch absolute (psia) was assumed. For sump temperatures greater than 212°F, the containment pressure was assumed equal to the vapor pressure of the fluid. No containment accident pressure is credited for the NPSH margin analysis.

The licensee provided the NPSH margin results for the limiting fiber case for Wolf Creek. For the RHR pumps in recirculation, the minimum NPSH margin was stated to be 1.21 ft at a sump temperature of 212°F and a pump flow rate of 4,760 gpm. For the CS pumps in recirculation the minimum NPSH margin was stated to be 2.14 ft at a sump temperature of 212°F and a pump flow rate of 3,950 gpm. The NPSH margin results are reported in section 3.g.16 of attachment VIII to the licensees LAR.

NRC Staff Conclusion Regarding Net Positive Suction Head The NRC staff reviewed the licensees NPSH evaluation against the NRC staff-accepted guidance and concludes that the licensee has appropriately validated that the plant design provides adequate margin between the NPSH available and the NPSH required for the RHR and CS pumps, running in recirculation mode, for all scenarios in which the break is smaller than the threshold break size. These cases are not considered to result in an increase in plant risk while all scenarios with breaks greater than the threshold break size are assumed to increase plant risk. Therefore, the NRC staff concludes that the licensees evaluation of NPSH is acceptable.

The NRC staff concludes that the licensee has provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning NPSH, because the licensee:

Provided applicable pump flow rates, the total recirculation sump flow rate, sump temperature(s), and minimum containment water level.

Described the assumptions used in the calculations for the above parameters and the sources/bases of/for the assumptions.

Provided the basis for the required NPSH values for the pumps.

Described how friction and other flow losses are calculated.

Described the system response scenarios for large break LOCA and small-break LOCAs.

Described the operational status for each ECCS and CSS pump before and after the initiation of recirculation.

Described the limiting single failure assumptions relevant to pump operation and sump performance.

Described how the containment sump water level is determined.

Provided assumptions that are included in the analysis to ensure a minimum (conservative) water level is used in determining NPSH margin.

Described how the volumes associated with empty spray pipe, water droplets, condensation and holdup on horizontal and vertical surfaces were accounted for in pool level calculations.

Provided assumptions (and their bases) for equipment credited to displace water resulting in higher pool level.

Provided assumptions (and their bases) as to the water sources that are credited to provide pool volume, and the volume from each source.

Provided description of the calculation of containment accident pressure used in determining the available NPSH.

Provided assumptions made, which minimize the containment accident pressure and maximize the sump water temperature.

Specified that the containment accident pressure is set at the vapor pressure corresponding to the sump liquid temperature.

Provided the NPSH margin results for pumps taking suction from the sump in recirculation mode.

3.4.2.8.6 Chemical Effects The objective of the chemical effects section is to evaluate chemical precipitate effects on the Wolf Creek sump strainer headloss. The evaluation of chemical effects on the reactor vessel is contained in section 3.4.2.8.8 of this SE.

The overall chemical effects evaluation methodology for Wolf Creek included:

Quantification of chemical precipitates using the methodology in WCAP-16530-NP-A (Reference 19).

Introduction of pre-mixed chemical precipitate solution into the strainer test fluid after the formation of a stable particulate and fibrous debris bed.

Application of an aluminum solubility correlation to determine the maximum precipitation temperature for strainer head loss evaluations.

Use of TR WCAP-17788-P autoclave test results to determine the minimum precipitation timing for chemical effects in the reactor vessel (section 3.4.2.8.8 of this SE).

Wolf Creeks plant-specific ECCS sump strainer headloss testing was performed at Alden in 2016. The test strainer consisted of two prototypical strainer stacks that matched the key design parameters of the plant strainer stacks. Each strainer stack had 40 strainer disks for a total of 80 strainer disks. The strainer stacks were located in a pit region within the test loop to represent the plant sump pit configuration. The test flume construction and piping promotes debris transport to the strainer. Both TB and FDL headloss tests were performed by first adding particulate and fibrous debris (NUKON' fiberglass) to build a representative post-LOCA debris bed on the strainer. Details of the test debris preparation are provided in the Head Loss and Vortexing, section (section 3.4.2.8.3 of this SE). Once headloss from the particulate and fibrous debris bed stabilized, pre-mixed chemical precipitate solution was added to the upstream end of the test flume and testing continued until a maximum head loss value was established. Wolf Creek performed an FDL headloss test and a TB debris headloss test. Headloss analyses were performed at strainer flow rates of 9,100 gpm and 4,900 gpm for conventional debris and chemical precipitates, respectively, since CS pumps would be secured well before chemical precipitates would form.

Chemical precipitates were prepared according to the WCAP-16530-NP-A instructions (Reference 19). After preparation, the aluminum oxyhydroxide precipitate settlement testing was performed to verify that the precipitate settlement met the WCAP-16530-NP-A acceptance criterion. Settlement testing verifies that the chemical precipitate settles in a representative manner. The aluminum oxyhydroxide precipitate solution was added to the test flume in seven batches. A total of 1,835 gallons aluminum oxyhydroxide precipitate solution was added to the FDL headloss test. The TB headloss test added 1,830 gallons of aluminum oxyhydroxide precipitate solution. Testing continued until peak headloss was attained according to the test plan. The headloss tests demonstrated there was adequate pump NPSH margin for the fiber and chemical effects break size limits assumed in the Wolf Creek evaluation. For all breaks greater than the assumed 10-inch diameter pipe break, the risk-informed analysis conservatively assumes failure in the risk quantification.

As part of the risk-informed evaluation of strainer debris load, Wolf Creek quantified chemical effects. A single bounding chemical precipitate quantity was determined for all analyzed break locations. The amount of chemical precipitate was determined using the WCAP-16530-NP-A methodology. The NRC staff has previously reviewed and approved WCAP-16530-NP-A as one method to calculate the amount of chemical precipitate and to prepare precipitates for strainer testing (Reference 19). The amount of calculated chemical precipitate was maximized by the licensee by applying conservative plant-specific inputs such as pH, temperature, and aluminum quantity. A maximum sump pool pH was conservatively used for chemical release and a minimum sump pool pH was conservatively assumed for determining when aluminum precipitates would form. A maximum temperature profile was assumed to conservatively maximize aluminum release. A minimum temperature profile was used to show precipitation would not occur prior to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Chemical precipitation was calculated using both maximum and minimum sump pool volumes to conservatively determine the precipitate quantity and precipitation temperature.

During the regulatory audit of the LAR for Wolf Creek, the NRC staff and licensee discussed why aluminum was not included in the LAR table B.3.6.8-1 that identifies containment sump debris limits for breaks less than 10 inches. The staff asked how aluminum is tracked to ensure the assumed amounts are not exceeded. In response, the licensee indicated that the Wolf Creek design change process requires the use of certain forms to identify potential impact on LTCC by the proposed modification. Changes to the aluminum quantity in containment are addressed by reviewing the appropriate design calculations to ensure the aluminum quantity remains within the tested and analyzed limits.

The Wolf Creek chemical effects approach relies on an aluminum solubility correlation to determine the maximum chemical precipitation temperature for each break. If precipitation is not predicted for a given pipe break before 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the analysis assumes precipitation occurs at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In accordance with the WCAP-16530-NP-A SE for strainer headloss analysis that applies a time-based NPSH margin acceptance criteria, the aluminum release rate used must not underpredict the initial 15-day aluminum concentration from Integrated Chemical Effects Test #1 in NUREG/CR-6914. Therefore, the licensee developed a spreadsheet that doubled the aluminum release during the initial 15 days post-LOCA. The aluminum solubility equation used by Wolf Creek was developed by Argonne National Laboratory (ANL) based on laboratory testing (Reference 58). As a conservative assumption, Wolf Creek assumed a constant 8.78 sump pool pH for aluminum solubility calculation purposes. Based on the ANL solubility equation, the maximum temperature for precipitation of aluminum in the projected post-LOCA environment for Wolf Creek was calculated to be 117.6°F. Assuming a 9.45 pH for aluminum dissolution while simultaneously assuming an 8.78 pH for aluminum solubility provides conservatism that accounts for uncertainty in the ANL aluminum solubility correlation. Use of the ANL equation for post-LOCA sump pool aluminum solubility calculation is acceptable to the NRC staff. After development of the ANL solubility correlation, additional chemical effects testing was performed by the PWR Owners Group and documented in TR WCAP-17788-P, Volume 5 (Reference 23). Autoclave tests in TR WCAP-17788-P that were representative for the Wolf Creek pH (with sodium hydroxide) and post-LOCA materials showed that the maximum precipitation temperature of 117.6°F was conservative.

NRC Staff Conclusion Regarding Chemical Effects The NRC staff finds the overall Wolf Creek chemical effects sump strainer evaluation acceptable for the following reasons. TR WCAP-16530-NP-A (Reference 19) approach was used to determine the amount of precipitate for each break. The licensee made conservative

assumptions related to sump pool volume, pH, sump pool temperature and aluminum quantity to maximize the amount of predicted precipitate. Strainer chemical effects testing was performed with aluminum oxyhydroxide precipitates prepared according to TR WCAP-16530-NP-A instructions and precipitate settling testing met the requirements. TR WCAP-16530-NP-A was previously reviewed and approved by the NRC staff as an acceptable method to evaluate plant-specific chemical effects. Wolf Creek also credited aluminum solubility in its sump strainer chemical effects evaluation using the ANL equation that is acceptable to the NRC staff. In addition, the licensee simultaneously analytically assumed the sump pool was 9.45 pH to maximize aluminum corrosion and 8.78 pH to minimize aluminum solubility. These conservative pH assumptions, along with doubling the aluminum release during the initial 15-day post-LOCA period account for the uncertainty associated with an aluminum solubility methodology.

The NRC staff concludes that the licensee has provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning chemical effects, because the licensee:

Provided a summary of evaluation results for the assumed pipe breaks that showed that chemical precipitates formed in the post-LOCA containment environment, either by themselves or combined with debris, do not deposit at the sump screen to the extent that an unacceptable head loss results, or deposit downstream of the sump screen to the extent that LTCC is unacceptably impeded.

Conservatively assumed that those pipe breaks that produce greater amounts of chemical precipitates than were found to be acceptable by strainer headloss testing cause strainer failure in the risk quantification.

3.4.2.8.7 Downstream Effects - Components and Systems The licensee stated that debris effects on components downstream of the sump screen were addressed using NRC staff-approved methods. The licensee used TR WCAP-16406-P-A, Revision 1 (Reference 21) for the analysis, including the limitations and conditions provided in the NRC SE (Reference 22). The licensee took no exceptions to this methodology. The evaluation determined that no modifications or operational changes were necessary for components or instrumentation due to ex-vessel downstream effects.

The licensee stated that the analysis assumed that the maximum transported debris amounts loaded the strainer. All the particulate coatings that transport to the strainer were assumed to penetrate the strainer. The licensee used debris concentrations for its downstream evaluations.

The concentrations of particulate, chemical, and fiber debris are listed in section 3.m.2 of attachment VIII to the licensees LAR. For the erosive wear evaluation, the debris concentration was not assumed to deplete over time. For blockage evaluations, the width of the particulate debris was assumed to be 10 percent larger than the ECCS strainer hole size and the thickness of the particulates was assumed to be one half of the hole size. The maximum length of the particulates was assumed to be 2 times the hole diameter.

The licensee summarized the methodology used to identify the components that required evaluation. The licensee stated that both cold-leg and hot-leg alignments were evaluated. The licensee considered the CSS and ECCS in its evaluation. The evaluation included erosive wear, abrasion, and potential blockage that could affect the recirculation mode of ECCS and CSS.

The components evaluated included valves, spray nozzles, orifices, heat exchanger tubes, instrument tubing, and the ECCS and CSS pumps.

The ECCS and CSS pumps were evaluated for erosive and abrasive wear. Some pumps did not have wear ring or impeller material with a Brinell hardness value greater than 400, so additional analysis was needed. The pumps requiring evaluation were the RHR and CSS pumps. The licensee calculated the effects of wear on the impellers to wear ring clearance on these pumps and determined that the clearance limit in TR WCAP-16406-P-A would not be exceeded.

Therefore, no action was required. The other pumps (i.e., CS and SI), had adequate hardness values for the impellers and wear rings and were determined to not need further evaluation. In addition, other aspects of pump performance were evaluated with respect to the effects of debris. The analysis found that the cyclone separators for the CS pumps would continue to operate properly and that the pump seals would not wear excessively. The licensee also determined that the mechanical performance of the pumps would remain acceptable.

For debris blockage, the licensee evaluated the ECCS and CSS components, including valves, piping, instrument tubing, and heat exchangers, and determined that all of these can accommodate sump bypass debris without blockage.

The licensee evaluated the erosive wear of heat exchangers, orifices, and spray nozzles for a mission time of 30 days and determined that wear of these components would not affect the system operation.

Throttle valves in the ECCS system were evaluated against wear criteria from TR WCAP-16406-P-A, and all valves passed the acceptance criteria for wear. Other valves in the system were screened from further wear analysis.

All the components passed the acceptance criteria in TR WCAP-16406-P-A, Revision 1.

The licensee stated that no design or operational changes are being implemented to manage downstream effects.

NRC Staff Conclusion Regarding Downstream Effects Components and Systems The NRC staff reviewed the evaluation methods and results, and finds that the licensee followed the NRC staff-accepted guidance contained in TR WCAP-16406-P-A, Revision 1, including its associated NRC SE. The NRC staff concludes that the licensee performed an adequate downstream effects evaluation of components and systems and that the components are capable of performing their safety-related design functions for the required mission time after a LOCA.

The NRC staff concludes that the licensee has provided sufficient information as requested by GL 2004-02 and further described in the revised content guide for GL 2004-02 concerning downstream effects components and systems, because the licensee:

Summarized the application of NRC-approved methods and stated that the NRC-approved methods were used for the evaluation without exception.

Provided a summary and the conclusions of the downstream effects evaluations.

Stated that no design or operational changes are required as a result of the downstream evaluations.

3.4.2.8.8 Downstream Effects - In-vessel The licensee stated that it assessed the in-vessel effects using the methodology from TR WCAP-17788-P (Reference 23) and followed the latest NRC staff review guidance for in-vessel effects (Reference 34). The NRC staff previously performed a thorough review of the TR and performed its own analyses and confirmatory calculations to gain insights into the relevant phenomena. The NRC determined that TR WCAP-17788-P provides significant insights into the response of PWRs to the effects of debris that transports to the vessel but did not approve it.

The NRC staff concluded that the debris limits developed by TR WCAP-17788-P could be used by licensees for the evaluation of in-vessel fibrous debris. The NRC staff guidance provides additional information on the staffs views of the TR. The PWR Owners Group developed companion guidance (Reference 34) that was also used by the licensee in its in-vessel analysis.

The amount of fiber assumed to transport to the core was based on strainer penetration test results. The transport of fiber to the core is discussed in section 3.4.2.7.2 of this SE. The maximum amount of fiber calculated to reach the reactor was 92.54 grams of fiber per fuel assembly (g/FA) for the worst design-basis case with two RHR pumps and one CS pump in operation, and only fiber fines initially present in the pool. The licensee also determined that for a beyond-design-basis case with both RHR pumps operating and no CS pumps in service, the amount of fiber reaching the reactor was calculated to be 94.15 g/FA. These quantities are 2.5 percent higher than the calculated amounts to account for potential uncertainty in the curve fit of the penetration data.

The in-vessel evaluation was based on penetration testing that included a total fibrous debris source term of 568.6 lbm per strainer. These tests were used to derive fiber penetration equations (referred to as the low fiber curve). The tested amount bounds the transported fibrous debris from breaks smaller than the threshold break size that would not cause strainer failure, as demonstrated by strainer headloss testing. Therefore, the NRC staff concluded that the penetration model was acceptable to estimate penetration in the plant for all conditions up to the threshold break size. The penetration model was used to compute time dependent debris arrival in the core using the guidance in TR WCAP-17788-P.

During the audit (Reference 16), the NRC staff questioned how much fibrous debris was assumed to be in the pool at the beginning of recirculation. The licensee stated that to compute in-vessel fiber buildup, it was assumed that the amount of fine fiber in the pool was 141.78 lbm, which is equal to the maximum fiber fines in the FDL test. The licensee also determined that assuming 144.6 lbm of fiber fines in the pool would result in a calculated value of fiber at the core inlet equal to the TR WCAP-17788-P limit. This information was also provided in the licensees supplement to the LAR (Reference 2). The strainer test limit of 141.78 lbm is much greater than the amount of fiber fines that would be produced by breaks up to the threshold break size. In its supplement to the LAR, the licensee provided updated values for the limiting amount of fiber that would transport to the strainer. The value for fine fiber was reduced by eliminating some double counting of transported fiber. The amount of fine fiber available to transport to and penetrate the strainers is significantly less than the amount that would result in exceeding the assumptions used in the in-vessel analysis.

The licensee stated that Wolf Creek used the Box 4 Path in the NRC in-vessel review guidance to evaluate the effects of debris in the reactor. The Box 4 Path requires a plant-specific

evaluation because Wolf Creek does not fall within all the assumptions and parameters used in the TR WCAP-17788-P analyses. The licensee provided table 3.n.1-1 in attachment VIII to its LAR to provide a comparison of Wolf Creek parameters to the values assumed in the TR WCAP-17788-P analysis.

The licensee stated that the reactor is a Westinghouse upflow design and that the core contains fuel assemblies that are of Westinghouse design. Hot-leg injection is initiated at about 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months following the LOCA.

Because the licensee could not demonstrate that sump switchover (SSO) occurs in greater than 20 minutes, as was assumed in the TR WCAP-17788-P analysis, a plant-specific evaluation was performed for Wolf Creek. The licensee provided comparisons between the TR WCAP-17788-P acceptance criteria and plant-specific parameter values in table 3.n.1-1 of attachment VIII to the LAR and discussed them in its plant-specific evaluation. The NRC staff evaluated LTCC for Wolf Creek with respect to in-vessel debris by reviewing the licensees information and TR WCAP-17788-P.

The licensee stated that TR WCAP-17788-P, Volume 5 (Reference 23) autoclave testing determined that chemical precipitates would not occur prior to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for containment sump temperatures above 87°F. The licensee determined this by using an aluminum solubility curve, shown in figure 3.n.1-6 in attachment VIII to compare the Wolf Creek sump pool aluminum concentration estimated from TR WCAP-16530-NP-A (Reference 19) chemical release methodology with all TR WCAP-17788-P Volume 5 sodium hydroxide (NaOH) group autoclave test results and the TR WCAP-17788-P precipitation boundary equation. This evaluation did not include autoclave tests performed at pH values greater than 10, since they were not considered representative of the Wolf Creek post-LOCA sump pool pH. The licensee stated that TR WCAP-17788, Volume 5 Test Groups 9 and 15 used NaOH buffer and reflect the minimum and maximum pH values for the post-LOCA sump pool at Wolf Creek. For these test groups, chemical precipitation filtration testing was conducted to minimum temperatures between 120°F and 160°F. No precipitation was detected during the 24-hour autoclave test duration.

The NRC reviewed the licensees in-vessel chemical effects evaluation, including the use of the precipitation boundary and autoclave test results from Volume 5 of TR WCAP-17788-P. The NRC staff verified that the TR WCAP-17788-P autoclave Groups 9 and 15 test conditions were representative for the Wolf Creek plant-specific maximum and minimum sump pool pH values, respectively. Both autoclave Groups 9 and 15 used NaOH to control the simulated post-LOCA pool pH, which is representative for Wolf Creek. Test Group 9 contained slightly more aluminum but slightly less E-Glass compared to Wolf Creek. Test Group 15 contained less E-Glass but approximately 50 percent more aluminum than Wolf Creek. Since TR WCAP-16530-NP-A testing has shown that aluminum metal releases significantly more aluminum than E-Glass, the use of data from these test groups to represent Wolf Creek is acceptable to the NRC staff.

During the autoclave tests, no chemical precipitation was detected for Test Groups 9 and 15, through 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and for minimum test filtration temperatures of 120°F and 160°F, respectively.

Based on all the tests with precipitation in the autoclave testing performed for WCAP-17788-P, a precipitation boundary was developed as part of this WCAP. The boundary was created by taking the aluminum concentrations, pH values, and temperature from all samples with precipitation and fitting a 3D surface. This surface was then adjusted to bound all precipitation events that were representative of plant conditions. The licensee used the maximum Wolf Creek aluminum concentration at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the minimum possible sump pool pH (which minimizes aluminum solubility) and the TR WCAP-17788-P precipitation boundary function to determine

the plant-specific 87°F precipitation temperature. The use of the TR WCAP-17788-P precipitation boundary is acceptable to the NRC staff for short term solubility (up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) since it enclosed all precipitation events that were observed in the TR WCAP-17788-P testing.

By using representative autoclave test results in TR WCAP-17788-P and by determining the plant-specific precipitation temperature using the WCAP-17788-P Volume 5 precipitation boundary function, Wolf Creek has shown that chemical precipitation will not occur until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The NRC staff reviewed the licensees evaluation and confirmed that testing supports the Wolf Creek conclusion that no chemical effects will occur before 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a LOCA.

Therefore, the NRC staff finds the licensees chemical effects evaluation for the reactor vessel acceptable.

Following a LOCA, hot-leg switchover (HLSO) is a change in injection flowpath that most PWRs employ to prevent boric acid precipitation in the core. This change in the injection flowpath also bypasses any debris that may be accumulated at the core inlet. If HLSO is performed prior to the formation of chemical precipitates, LTCC is assured by this action that bypasses the core inlet. At Wolf Creek, HLSO is performed at about 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , which is significantly before the earliest time that chemical precipitates are predicted to form at Wolf Creek.

The licensee stated that core cooling can be maintained via AFPs as long as the maximum allowable quantity of fibrous debris for the plant-specific fuel type is not exceeded prior to tblock.

TR WCAP-17788-P defines tblock as the earliest time at which the core inlet can be fully blocked for a specific reactor design but still allow adequate flow through the AFP to ensure LTCC. The licensee evaluated several cases and found that the total in-vessel fiber limit for Wolf Creek would not be exceeded even if AFPs are credited. In the absence of early chemical effects, some coolant flow is likely to be maintained through the fiber bed at the core inlet along with flow through the AFP. As decay heat decreases, less flow is required to maintain LTCC. Plants with lower resistance AFPs have lower tblock times. If chemicals do not precipitate prior to tblock, the tblock value is valid for the plant. Chemical effects testing demonstrated that chemical precipitation will not occur for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months under Wolf Creek plant-specific conditions. This is greater than the proprietary tblock time calculated by TR WCAP-17788-P for the Wolf Creek reactor design. Therefore, even if chemical precipitates form after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the core inlet were to be completely blocked, AFPs would be capable of LTCC.

The licensee stated that the Wolf Creek maximum thermal power is less than the TR WCAP-17788-P assumed power. The Wolf Creek power is 3,565 megawatts thermal (MWt) and the analyzed power is 3,658 MWt in TR WCAP-17788-P for Westinghouse upflow plants.

The licensee stated that the Wolf Creek AFP resistance is less than the TR WCAP-17788-P analyzed value. These values are proprietary. The NRC staff confirmed that the Wolf Creek AFP resistance is less than the analyzed value in TR WCAP-17788-P.

The licensee stated that the Wolf Creek ECCS flow rate per FA exceeds the value used in the TR WCAP-17788-P analysis for design basis cases but is within the analyzed values for beyond design basis scenarios. The analyzed values in TR WCAP-17788-P range from 8 to 40 gpm/FA.

With both RHR pumps operating, the flow rate can be as high as 52.9 gpm/FA. For beyond design basis cases where ECCS flow is reduced to maximize the fiber load in the vessel, the flow rate is 37.8 gpm/FA. The licensee stated that the TR WCAP-17788-P FA testing showed that the highest resistance debris bed was created at low flow rates. At higher flow rates the debris bed became unstable and had lower flow resistance caused by debris bed breakthrough.

The licensee concluded that it is expected that debris beds formed at higher flow rates will have

lower resistance than was assumed in the analysis and that for this parameter, the TR WCAP-17788-P analysis is bounding. The NRC staff concluded that the licensee observations are correct based on review of the TR test results and the way they are used in the licensees analysis regarding flow rates and debris bed resistance.

As discussed in section 3.4.2.7 of this SE, the licensees calculations predicted a maximum in-vessel fiber amount of 92.54 g/FA for the limiting design basis case and 94.15 g/FA for the beyond design basis case. These values were calculated assuming the maximum amount of fine fiber included in the FDL test was initially present in the pool; this initial amount of fiber highly bounds the total fiber fines generated by breaks up to the threshold break size. The calculated in-vessel fiber values are greater than the fuel/RCS design specific acceptance criteria for the fuel inlet specified in TR WCAP-17788-P, but less than the total core fiber acceptance limit. The NRC guidance requires that it be assumed that all fiber that reaches the vessel accumulates at the core inlet. The licensee stated that the assumption in TR WCAP-17788-P is that the fiber accumulates uniformly at the core inlet. However, the flow at the core inlet is not uniform, so the fiber deposition at the core inlet would be skewed to the areas with higher flow. The non-uniform distribution of debris would result in a greater amount of fiber required to block the core inlet and activate the AFPs. TR WCAP-17788-P determined that if the amount of fiber at the core inlet does not exceed the total core fiber limit, LTCC will be maintained. The NRC staffs review guidance for in-vessel effects states that plants that exceed the fuel inlet limit but are within the total core limit will maintain adequate LTCC. The licensee concluded that the maximum core fiber amount is less than the acceptance limit. The NRC staff concluded that the licensees assertion that LTCC will be maintained is sound if other required parameters are within the bounds of the analysis or evaluated to be acceptable.

The licensee stated that the time to SSO assumed in the TR WCAP-17788-P analyses is 20 minutes and that the plant-specific switchover is at 12.18 minutes. This is non-conservative from an in-vessel debris loading perspective because the debris is delivered to the core earlier and blockage may occur earlier resulting in the core having higher decay heat than was assumed in the TR WCAP-17788-P analysis. If blockage occurs when the core is at a higher decay heat value, the temperature used as an acceptance criterion in the thermal hydraulic analysis could be exceeded.

The licensee stated that the switchover time for the plant is calculated in a conservative manner because maximum ECCS flows for all trains were assumed and the RWST was assumed to be at its minimum TS level.

The licensee stated that the WCAP-17788-P analysis decay heat value at 20 minutes was calculated using the 1971 ANS-5.1 decay heat curve plus 20 percent uncertainty and is 87.4 MWt. For the SSO time of 12.18 minutes, the licensee calculated a plant-specific decay heat value of 80.0 MWt using a more realistic decay heat model (ANSI/ANS-5.1-1979 standard with 2 uncertainty). The licensee concluded that the plant decay heat is bounded by the value used in the analysis. The NRC staff concluded that using the more realistic standard removes some margin from the analysis but is an acceptable conclusion for Wolf Creek since margin is maintained in other areas of the analysis.

The licensee described other conservatism associated with assumptions used in TR WCAP-17788-P for debris transport to the reactor vessel. The assumption that all debris reaches the core inlet within 60 seconds of the beginning of SSO to recirculation has been demonstrated to result in significantly conservative predictions of peak fuel temperatures, which are the basis for acceptability of debris amounts calculated in the WCAP. Sensitivity studies

performed in TR WCAP-17788-P showed significant decreases in the maximum calculated peak cladding temperatures with slower arrival of debris at the core inlet. The licensee stated that it takes 428 seconds for the core inlet debris amount at Wolf Creek to reach the plants core inlet fiber limit which is significantly longer than 60 seconds. The NRC staff has concluded that the TR WCAP-17788-P methodology contains conservative margins, including the assumption that the debris arrives at the core inlet in 60 seconds instead of a longer period that would actually occur at the plant.

NRC Staff Conclusion Regarding Downstream Effects - In-vessel The NRC staff reviewed the licensees in-vessel evaluation against the NRC staff-accepted guidance for the topic. The NRC staff concludes that the licensee has appropriately evaluated the ability of the ECCS to ensure LTCC considering the potential for buildup of debris at the core inlet and inside the reactor vessel. Therefore, the NRC staff concludes that the licensees evaluation of in-vessel downstream effects is acceptable.

The NRC staff concludes that the licensee has provided sufficient information as requested by NRC staff review guidance for in-vessel effects and further described in TR WCAP-17788-P concerning in-vessel effects, because the licensee demonstrated that:

The ECCS is realigned to prevent boric acid precipitation within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the start of the event.

Although the SSO occurs earlier than assumed in TR WCAP-17788-P, the plant-specific decay heat at the time of sump switchover is bounded by the WCAP-17788-P analysis assumptions. Other margins in the methodology are maintained.

The total amount of fiber reaching the core will not exceed the total in-vessel fiber limit defined by TR WCAP-17788-P.

NRC Staff Conclusion Regarding Impact of Debris Submodel Each of the aspects of the impact of debris area has been evaluated above. The NRC staff concludes that the sub-areas of upstream effects, screen modification package, headloss and vortexing, sump structural analysis, NPSH, chemical effects, downstream effects - components and systems, and downstream effects - in-vessel were adequately addressed. Based on the evaluations for each of these subsections, the NRC staff finds that the licensees impact of debris evaluation is acceptable.

3.4.2.8.9 Submodel Integration This section provides an overview of how the submodels were combined to obtain the final results of the risk analysis.

The licensee used supporting computations to demonstrate that the amount of debris generated and transported by breaks smaller than the threshold break size would not result in ECCS or CSS failure and the LTCC would be maintained. Use of the threshold break size results in a relatively simple integration because it uses only two failure criteria, strainer debris amount and in-vessel debris amount, to determine whether scenarios are successful or not.

The licensee used conservative assumptions regarding the equipment states for strainer and in-vessel debris failures. That is, the analysis assumed a single train of ECCS and CSS in service for strainer evaluations and two trains in service for the in-vessel evaluations. This maximizes the debris headloss for the strainer analysis and maximizes the debris buildup in the core for the in-vessel analysis. The licensee considered strainer parameters (e.g., buildup of fiber, chemical precipitates, particulates, headloss) associated with debris loads up to the threshold break size and compared those parameters to the failure criteria (e.g., NPSH, flashing, etc.).

The licensee used a program called BADGER to compute the amounts of each type of debris generated for each postulated break at pipe weld locations for a range of break sizes. The licensee used NRC accepted methods to determine quantities such as debris transport, dissolution rates and chemical precipitation, strainer headloss, flow rates, etc.

The licensee used BADGER model outputs to verify that the breaks smaller than the threshold break size cannot generate more debris greater than shown to be successfully mitigated by tests. BADGER was also used to calculate the maximum debris loads that could be generated by any LOCA break, to perform some of the analyses using more conservative inputs. BADGER includes a 3-D description of the distribution of debris sources, robust barriers, and weld locations. For each weld location, BADGER varies the break size and orientation, and outputs debris amounts for each postulated break. The BADGER outputs are post-processed to define a database of debris amounts for each of the weld locations, break sizes, and orientations. The break and debris generation database is post-processed to compute the amount of debris transported to strainers. The post-processing computations were based on blowdown, washdown, pool fill, and recirculation transport fractions, as well as erosion fractions, to define debris amounts that transport to the strainers. The debris generation computations account for debris sources that are ZOI independent (e.g., latent debris and unqualified coatings) and assumed constant for all breaks. Based on the amount of debris transported, the licensee determined that breaks smaller than the threshold break would be successfully mitigated and assumed that breaks greater than the threshold break size would cause strainer failure and core damage. Strainer headloss, NPSH margins, strainer deaeration rates, strainer structural margin, and local pressures (to determine whether flashing may occur) were all shown to be acceptable for debris loads associated with breaks smaller than the threshold break size.

Failures due to debris accumulating in the core were deterministically shown not to result in a loss of core cooling for breaks smaller than the threshold break size. Therefore, in-vessel failures do not result in additional scenario failures beyond those accounted for in the threshold break size methodology.

The licensee assumed that all breaks equal to or greater than the threshold break size resulted in core damage and increase in plant risk. The licensee calculated the CDF as the frequency of breaks exceeding the threshold break size, interpolating 25-year NUREG-1829 LOCA break frequencies based on GM)aggregation.

The NRC staff verified that the licensees calculations were performed accurately and used acceptable assumptions and methods. The NRC used a combination of confirmatory calculations, engineering review, and review of the licensees software outputs to perform the verifications. The verifications included balancing debris amounts output by BADGER with transported debris considering transport and erosion fractions, exploring trends in debris amounts vs. break size and orientation for numerous welds, and identifying the consistency of failure conditions. The verifications also included examining that the in-vessel fiber accumulation would not cause core damage, and alternative computations of the CDF crediting orientation

and larger break sizes not causing strainer failure. The approach allows the staff to conclude, with a high level of confidence, that the calculations for debris generation were conducted and applied properly, and are, therefore, acceptable.

NRC Staff Conclusion Regarding Submodel Integration The NRC staff concluded that the licensees submodel integration was acceptable based on its review of the methodology, the BADGER debris generation outputs, and post-processing computations. The NRC staff concludes that the approach for integrating submodels described in the technical report is acceptable for use in an assessment or evaluation model of the effects of debris on long-term cooling of ECCS, as required, in part, by 10 CFR 50.46.

3.4.2.9 Systematic Risk Assessment RG 1.174 states that, one key principle in risk-informed regulation is that proposed increases in risk are small and are consistent with the intent of the Commissions Safety Goal Policy Statement. In attachment VII to the LAR (Reference 1), the licensee described the risk-informed basis, including the systematic risk assessment. The licensee accounted for potential failure modes addressed in section 3.4.2.8 of this SE. In the debris computations, various breaks generated and transported more debris to the strainer than was included in headloss strainer tests. The licensee assumed that all breaks greater than 10 inches, the threshold break size, resulted in scenarios of strainer failure and core damage. Debris amounts by breaks less than the threshold break size were shown to not result in any failures of the systems needed to assure LTCC. The licensee quantified the CDF equal to the exceedance frequency of 10-inch LOCA breaks, interpolating 25-year LOCA frequencies considering GM aggregation from NUREG-1829.

The licensee screened break locations and GL 2004-02 relevant scenarios, which are evaluated in section 3.4.2.1 of this SE. The licensee concluded that the only breaks contributing to the CDF are breaks in Class 1 welds. Debris generating models are evaluated in Section 3.4.2.6 of this SE. The licensee considered LOCA break frequencies from NUREG-1829 and related sources to estimate CDF and LERF; the frequency selection approach is evaluated in section 3.4.2.2 of this SE. LOCA break frequencies from an approach considering GM aggregation of expert elicited frequencies, and 25-year plant life frequencies, were used by the licensee in the Baseline computations. The licensee reported Baseline CDF and LERF values in table 3 of attachment VII to its LAR.

The licensee evaluated the impact of key assumptions and sources of uncertainty in the systematic risk assessment in sections 2.5 and 2.6 of attachment VII to the LAR. The licensee stated that consensus models and inputs were used in many cases to reduce uncertainty because these are considered conservative or bounding. For inputs that could have an effect on the analysis the licensee evaluated the result of maximizing or minimizing the inputs in table 6 of attachment VII to its LAR. The licensee examined non-consensus inputs such as the pool volume, pool temperature, ECCS flow rate, and CS flow rate. The licensee stated that the parameters listed in table 6 could all be considered consensus inputs considering the level of conservatism and the consideration of competing effects. The licensee stated that the only parameter that needs uncertainty quantification is the LOCA frequencies.

For LOCA frequency, which can have a major effect, the licensee performed a sensitivity considering the 5th and 95th percentile of the 25-year GM aggregation LOCA frequencies and presented those results in table 7 of attachment VII to its LAR. The licensee also performed a

sensitivity study considering the effects of using AM aggregation values from NUREG-1829 (Reference 14). The results of the AM aggregation sensitivity are provided in table 9 of attachment VII to the licensees LAR. The AM aggregation sensitivity values result in about an order of magnitude increase in plant risk over the Base Case values for pipe LOCAs for a threshold break size of 10 inches. The licensee also performed a sensitivity analysis for the calculation of LOCA frequency due to seismic events. The baseline seismic LOCA frequency was calculated to be 6.2x107/yr. A sensitivity study was performed using site-specific fragility parameters and the guidance contained in NUREG-1903 (Reference 49) with a CDF value equal to 5.7x107/yr. This sensitivity study information for seismic LOCA frequencies is provided in table 10 of attachment VII to the licensees LAR.

The licensee examined changes in the CDF magnitude when considering AM aggregation LOCA break frequencies. The sensitivity analysis revealed that the CDF estimate increased by close to a factor of 10 with respect to the baseline CDF (the baseline CDF is based on GM aggregation LOCA break frequencies). Also, the licensee considered 40-year LOCA frequencies from NUREG-1829 and computed a value CDF slightly greater than 106/yr, reported in a supplement. The NRC staff notes that the licensee identified features of the analysis in support of defense-in-depth and safety margins which would ensure inputs, assumptions, and conclusions of the LAR remain valid under uncertainties of the analysis, and that there is margin to lower the CDF estimates considering a more realistic evaluation. The NRC staff concluded that the sensitivity analysis does not change the licenses conclusion of very low risk (that is, the change lies in RG 1.174 Region III) because conservatisms exist in the licensees assessment that can offset the impact of single or combined uncertainties, including uncertainties in LOCA break frequencies. The NRC staff views the aggregation method of the LOCA break frequencies as a key assumption and source of uncertainty for the systematic risk assessment. The staff is not endorsing any specific aggregation method of LOCA expert elicited frequencies for general use.

The licensee discussed completeness uncertainty and stated that the risk-informed evaluation considered enough scenarios and the phenomena being evaluated are relatively well understood. This is based on the significant body of knowledge available on the related topics from research and testing over the last few decades. Therefore, it is unlikely that there are unknown phenomena associated with the parameters and processes related to the risk-informed analysis. The licensee concluded that uncertainty associated with unknown phenomena is small.

The licensee provided table 11 in attachment VII of its LAR to summarize the uncertainty quantification results. In summary, the licensee noted that three of the five uncertainty sensitivities show that CDF and LERF would decrease and two of the uncertainty sensitivities show that CDF and LERF would increase compared to the baseline values. The two sensitivities that show increases in risk are associated with the inputs for break frequency. The licensee also highlighted significant conservatisms that are included in its risk-informed analysis, that if removed (i.e., replaced by more realistic analyses) would result in significant decreases in calculated risk.

The NRC staff concluded that the licensee addressed dominant uncertainties in sensitivity analyses. The licensees approach to compute the CDF only includes a few factors. Other factors, such as uncertainty in fiber generated and transported, were addressed for example by using guidance with safety margin in the ZOI and consideration of bounding transport fractions.

Other failure modes and secondary break sources were properly screened and excluded or conservatively considered to contribute to plant risk. For example, the licensee concluded that

fiber buildup in the vessel would not compromise heat dissipation of the core, considering reasonable sources of uncertainty in the analysis (section 3.n of attachment VIII to the LAR].

The licensee also assumed that SSBIs that lead to recirculation would lead to core damage.

The NRC staff also agrees that the use of consensus models in the analysis along with the use of a threshold break size include conservatism. Therefore, uncertainties are adequately addressed.

The NRC staff performed verification calculations, considering data provided during an audit, to ensure breaks less than the threshold break size would not exceed test limits. The NRC staff also reproduced the CDF in tables 3, 7, and 9, of attachment VII to the licensees LAR, using LOCA frequencies in NUREG-1829. The NRC staff developed sufficient confidence that the licensee properly computed the CDF and LERF.

NRC Staff Conclusion Regarding the Systematic Risk Assessment The NRC staff evaluated the systematic risk assessment methodology and concluded it was acceptable because the inputs and assumptions (e.g., initiating event frequencies for critical welds) were derived using state of practice data and approaches, scenarios that affect the GL 2004-02 risk assessment were adequately identified and included in the risk evaluation, elements of the risk evaluation were developed in a systematic and acceptable manner, and key assumptions were appropriately considered and described. The NRC staff verified selected computations in support of the CDF. Therefore, the licensee used a verifiable and robust methodology to calculate the risk attributable to debris.

The licensee properly considered sources of uncertainty in the computation of the CDF and LERF and concluded that a dominant source of uncertainty is the input LOCA frequencies.

The NRC staff found the conclusion that CDF and LERF belong in RG 1.174 risk Region III acceptable, based on the licensees baseline computations, as well as the licensees factors contributing to safety margins evaluated in section 3.3 of this SE, and factors contributing to defense in depth evaluated in section 3.2 of this SE.

3.

4.3 NRC Staff Conclusion

Regarding Key Principle 4: Risk Assessment The licensee used a method to quantify the CDF and LERF, outside of the PRA model, of the appropriate scope, level of detail, and technical elements and plant representation. The risk-informed approach used by the licensee to address the effects of debris on LTCC is acceptable.

Alternative assumptions were considered as sensitivities for each key assumption employing non-consensus approaches. The increase in risk is very small and in accordance with the Region III acceptance guidelines defined by RG 1.174. Therefore, Principle 4 of integrated risk-informed decision-making is acceptable.

3.5 Key Principle 5: The Impact of the Proposed Change Should Be Monitored Using Performance Strategies RG 1.174, Regulatory Position C.3, Element 3: Define Implementation and Monitoring Program, states:

The primary goal of Element 3 is to ensure that no unexpected adverse safety degradation occurs because of the change(s) to the licensing basis. The [NRC]

staffs principal concern is the possibility that the aggregate impact of changes that affect a large class of SSCs could lead to an unacceptable increase in the

number of failures from unanticipated degradation, including possible increases in common-cause mechanisms. Therefore, an implementation and monitoring plan should be developed to ensure that the engineering evaluation conducted to examine the impact of the proposed changes continues to reflect the actual reliability and availability of the SSCs evaluated. This ensures that the conclusions drawn from the evaluation remain valid.

In the LAR (section 4.0 of Attachment VII and section 3.i of Attachment VIII), the licensee stated that it has implemented programs and procedures to evaluate and control changes to the plant that could have an impact on performance related to the effects of debris on ECCS recirculation.

The licensee stated that they have programs to control potential sources of debris in containment. As part of the TS change included with the LAR, the licensee adopted a TS SR that requires visual inspections to assure that the sumps do not have structural damage, abnormal corrosion, or debris blockage. Procedures require inspections of containment to check for loose debris and control all unattended materials. The licensee stated that its design change control procedure includes provisions for managing potential debris sources such as insulation, qualified coatings, exposed aluminum or zinc, and potential effects of post-LOCA debris on recirculation flow paths and downstream components. The licensees change process explicitly requires changes that involve any work or activity inside the containment be evaluated for the potential to affect the following:

High or moderate line break analysis Add or remove equipment inside containment Quantity of aluminum or zinc inside containment Quantity or type of coatings inside containment Change or addition of thermal insulation Addition of materials that could affect sump performance or cause equipment degradation, affect the design, performance, or operation of pumps As described above, a design evaluation is required to be completed for all design changes, which ensures that new insulation material that may differ from the initial design is evaluated for concerns related to the effects of debris. The licensee also stated that it has implemented procedures to inspect and repair protective coatings used inside containment are maintained remain qualified. The licensee noted that the Maintenance Rule program includes performance monitoring of high safety significant functions associated with ECCS and CSS (section 3.0 of attachment I to the LAR). The inclusion of ECCS and CSS into the Maintenance Rule program, and the assessment of acceptable system performance, provide continued assurance of the availability for performance of the required functions.

The licensee also stated that periodic updates to the risk-informed analysis will be performed to capture the effects of changes that may affect the results of the key elements of the analysis (Section 7.0 of attachment VII to the LAR). The licensee stated that changes to any of the hey elements will require review in accordance with 10 CFR 50.59, Changes, tests, and experiments, and updates will include plant changes, procedure changes, or new information regarding the risk-informed analysis. The licensee stated that as PRA inputs related to reliability data, unavailability data, initiating event frequency data, human reliability data, and other similar inputs are reviewed, the base PRA model for Wolf Creek will be updated to be consistent with the as-built, as-operated plant.

Procedures for operators to detect and respond to sump blockage issues related to GL 2004-02 have been developed and incorporated in the plant emergency operating procedures and emergency contingency actions.

The Wolf Creek and vendor quality assurance (QA) programs are compliant with the requirements of 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. The licensee stated that the analyses and testing for the risk-informed analysis were performed as safety related under these QA programs. The PRA evaluations were not considered to be safety related but were performed under the vendor QA program.

The NRC staff reviewed the licensees information and concludes that the licensees monitoring program is acceptable because it is consistent with the guidance in RG 1.174, Regulatory Position C.3.

3.

5.1 NRC Staff Conclusion

Regarding Key Principle 5: Performance Monitoring The Wolf Creek Maintenance Rule program includes performance monitoring of functions associated with ECCS and CSS, including sump recirculation. Technical Requirements implemented by Wolf Creek procedures require visual inspections of all accessible areas of the containment to check for loose debris, and each containment sump to check for debris.

Licensed Operators are trained on indications of and actions in response to sump blockage issues related to GL 2004-02, and performance is evaluated during training scenarios designed to simulate plant responses. Therefore, Principle 5 of performance monitoring is acceptable.

3.6 TS Changes for Implementation of TSTF-567 3.6.1 Proposed Changes to TS 3.5.2, ECCS - Operating The licensee proposed to modify and move SR 3.5.2.8 from TS 3.5.2 to the new containment sump TS. Therefore, the licensee proposed deletion of SR 3.5.2.8.

The new SR 3.6.8.1 does not limit the visual inspection to the suction inlet, trash racks and screens as currently required by the TSs but instead requires inspection of the entire containment sump system. The containment sump system consists of the containment drainage flow paths, any design features upstream of the containment sump that are credited in the containment debris analysis, the containment sump strainers (or screens), the pump suction trash racks, and the inlet to the ECCS and CSS piping.

The NRC staff concludes the proposed change is acceptable since the existing requirements are either unchanged or expanded and continue to ensure the containment sump is unrestricted (i.e., unobstructed) and stays in proper operating condition. The proposed change meets the requirements of 10 CFR 50.36(c)(3) because it provides an SR to assure the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and that the LCOs will be met. The NRC staff finds the change to TS 3.5.2 acceptable.

3.6.2 Proposed Changes to TS 3.5.3, ECCS - Shutdown The licensee proposed to delete the reference to SR 3.5.2.8 in SR 3.5.3.1.

The NRC staff concludes that the proposed change is acceptable since SR 3.5.2.8 was modified and moved to the new containment sump TS. The existing SR on the containment sump is augmented (by requiring inspection of additional sump components) and moved to the new specification, and a duplicative requirement to perform the SR in TS 3.5.3 is removed. The new specification retains or expands the existing requirements on the containment sump and the actions to be taken when the containment sump is inoperable with the exception of adding new actions to be taken when the containment sump is inoperable due to containment accident generated and transported debris exceeding the analyzed limits. The new action provides time to evaluate and correct the condition instead of requiring an immediate plant shutdown. The proposed change meets the requirements of 10 CFR 50.36(c)(3) because it provides SRs to assure the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and that the LCOs will be met. The NRC staff finds the change to TS 3.5.3 acceptable.

3.6.3 Proposed Changes to TS 5.5.15 Limiting Condition for Operation 3.0.6 states:

When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.15, Safety Function Determination Program (SFDP).. 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 support systems Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

When a loss of safety function is determined to exist, the SFDP requires entry into the appropriate conditions and required actions of the LCO in which the loss of safety function exists. When a loss of function is solely due to a single TS support system the appropriate LCO is the LCO for that support system. When the loss of function is the result of multiple support systems, the appropriate LCO is the LCO for the supported systems.

The licensee proposed to add the following sentence to TS 5.5.15:

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.

The NRC staff finds that the proposed addition to TS 5.5.15 clarifies the intent of the allowance (not to enter the Conditions and Required Actions) provided by LCO 3.0.6 and the SFDP for

single-train support systems. The NRC staff concludes the proposed change is acceptable since the actions for the support system LCO adequately address the inoperability of that system.

Therefore, as required by 10 CFR 50.36(c)(5), the proposed change continues to provide adequate administrative controls to assure safe operation.

3.6.4 Proposed Addition of Containment Sumps TS LCO 3.6.8 3.6.4.1 Evaluation of New LCO 3.6.8 The licensee proposed adding a new LCO to address operability requirements of the containment sump. The numbering for this proposed TS is TS 3.6.8.

The containment sump supports the post-accident operation of the ECCS and CSS. However, only the current ECCS TS contains SRs related to the containment sump and the TS does not specify required actions that specifically address an inoperable containment sump. If the containment sump was found to be inoperable, as an ECCS and CSS support system, those respective LCOs would not be met. In order to address concerns related to containment sump operability due to debris accumulation described in GSI-191, the licensee proposed to add a new specification to address containment sump inoperability and create a condition for when the sump is inoperable due to analyzed containment accident generated and transported debris.

Based on the below evaluation, the NRC staff determined that the proposed TS satisfies the requirements of 10 CFR 50.36(c)(2)(i) because the LCO specifies the lowest functional capability or performance levels of equipment required for safe operation of the facility. There is reasonable assurance that the required actions to be taken when the LCO is not met are adequate to protect the health and safety of the public.

3.6.4.2 Evaluation of the Applicability The new TS requires the containment sump to be operable during Modes 1, 2, 3, and 4. The current ECCS and CSS TSs are applicable during Modes 1, 2, 3, and 4.

The NRC staff finds the proposed applicability is acceptable because the applicability is consistent with the applicability of the ECCS and CSS TSs, the containment sump supported systems.

3.6.4.3 Evaluation of Condition A The licensee has analyzed the susceptibility of the ECCS and CSS to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. The licensee has established limits on the allowable quantities of containment accident generated debris that could be transported to the containment sump based on its current plant configuration. In the current TSs, if unanalyzed debris sources are discovered inside containment, if errors are discovered in debris-related analyses, or if a previously unevaluated phenomenon that can affect containment sump performance is discovered, the containment sump, and the supported ECCS and CSS, may be inoperable and the TSs would require a plant shutdown with no time provided to evaluate the condition.

In order to address this situation and to provide sufficient time to evaluate the condition, the licensee proposed Condition A, which is applicable when the containment sump is inoperable due to containment accident generated and transported debris exceeding the analyzed limits.

Under Condition A, the operability of the containment sump with respect to debris is based on a quantity of debris evaluated and determined to be acceptable by the licensee. Conditions not evaluated under Condition A (containment accident generated and transported debris) and that affect the quantity of analyzed debris will be evaluated using a deterministic process.

Under Condition A, Required Action A.1 mandates immediate action to be initiated to mitigate the condition. The licensees proposed TS Bases for Required Action A.1 provided the following examples of mitigating actions:

Removing the debris source from containment or preventing the debris from being transported to the containment sump; Evaluating the debris source against the assumptions in the analysis; Deferring maintenance that would affect availability of the affected systems and other LOCA-mitigating equipment; Deferring maintenance that would affect availability of primary defense-in-depth systems, such as containment coolers; Briefing operators on LOCA debris management actions; or Applying an alternative method to establish new limits.

The NRC staff finds the proposed Required Action A.1 and its CT are acceptable because they place urgency on the initiation of the appropriate actions that could mitigate or reduce the impact of the identified conditions.

Concurrently, Required Action A.2 mandates SR 3.4.13.1, the RCS water inventory balance, to be performed at an increased frequency of once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . An unexpected increase in RCS leakage could be indicative of an increased potential for an RCS pipe break, which could result in debris being generated and transported to the containment sump.

The NRC staff finds the proposed Required Action A.2 and its CT are acceptable because the more frequent monitoring allows operators to act in a timely fashion to minimize the potential for an RCS pipe break while the containment sump is inoperable.

In addition, Required Action A.3 requires the inoperable containment sump to be restored to operable status in 90 days.

The NRC staff finds the proposed Required Action A.3 and its CT are acceptable because they provide a reasonable amount of time to diagnose, plan and possibly reduce the severity of, or mitigate the unanalyzed debris condition and prevent a loss of ECCS and CSS safety function.

In addition, 90 days is adequate given the conservatisms in the containment debris analysis and the proposed compensatory actions required to be implemented immediately by Required Action A.1. Also, as discussed later in this SE section, the new SR will require visual inspection of the containment sump system (including the containment drainage flow paths, any design features upstream of the containment sump that are credited in the containment debris analysis, the containment sump strainers, the pump suction trash racks, and the inlet to the ECCS and CSS piping for evidence of structural degradation, potential for debris bypass, and presence of

corrosion or debris blockage) to ensure no loose debris is present and there is no evidence of structural distress or abnormal corrosion.

For Condition A, a plant with multiple sumps is treated equivalently to a plant with a single sump, because multiple sumps are considered to be part of a single support system.

3.6.4.4 Evaluation of Condition B Condition B specifies the required actions for when the containment sump is inoperable for reasons other than containment accident generated and transported debris exceeding the analyzed limits.

The licensees proposed Required Actions B.1 and B.2 are variations from the traveler and require declaring the affected ECCS and CSS trains inoperable when Condition B is entered.

The variation also does not incorporate the two notes proposed in the TSTF-576 traveler.

The proposed CT for Required Actions B.1 and B.2 is immediately. This requires the licensee to take the actions appropriate for the inoperability of any affected ECCS or CSS train as directed by the ECCS and CSS TSs.

The TSTF-576 Required Action B.1 is the only required action under Condition B. The traveler required action is modified by two notes, which direct entering the applicable conditions and required actions of LCOs 3.5.2 and 3.5.3 for ECCS trains made inoperable by the containment sump(s) and entering the applicable conditions and required actions of LCO 3.6.6, Containment Spray and Cooling Systems, for CSS and cooling system trains made inoperable by the containment sump(s). These actions, directed by the note, are not required for Wolf Creek since they are required directly by the proposed Actions B.1 and B.2. Since proposed Required Actions B.1 and B.2 direct entry to the corresponding ECCS and CSS TSs, the inoperability of the ECCS and CSS are appropriately addressed by those system TSs. The proposed Required Action B.1 and B.2 appropriately control conditions resulting in an inoperable containment sump for reasons other than containment accident generated and transported debris exceeding the analyzed limits. They are acceptable as proposed, and the notes, as proposed in TSTF-576, are not required.

The NRC staff finds the proposed change acceptable since it continues to provide remedial actions for when the containment sump is inoperable for reasons other than Condition A and ensures safe operation of the plant. In addition, the proposed CT is acceptable since it requires entry into the appropriate ECSS and CSS LCOs immediately.

3.6.4.5 Evaluation of Condition C If operators are unable to restore the affected containment sump to operable status under Conditions A or B, Required Action C.1 requires the unit to be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months followed by Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , as required by Required Action C.2.

The NRC staff finds this proposed condition and its required actions are acceptable because the condition is consistent with the STS and the required action requires the operators to place the unit in a condition in which the LCO no longer applies. In addition, the proposed CTs allow a reasonable amount of time to decrease from full power conditions to the required plant conditions in an orderly manner and without challenging plant systems.

3.6.4.6 Evaluation of the New SR The licensee proposed a new SR in the new containment sump TS. This SR is currently located in TS 3.5.2 and referred to in TS 3.5.3. The licensee proposes that the numbering for this new SR be SR 3.6.8.1. The frequency of the proposed SR is in accordance with the SFCP.

The proposed SR requires verification, by visual inspection, that the containment sump does not show structural damage, abnormal corrosion, or debris blockage.

The new SR is stated in generic terms and expands the scope of the required visual inspection to include the entire containment sump system. The entire containment sump system consists of the containment drainage flow paths, the containment sump strainers (or screens), the pump suction trash racks, and the inlet to the ECCS and CSS piping.

The NRC staff finds the proposed new SR is acceptable since it expands the scope of inspection of the original SR. In addition, the proposed frequency is acceptable since it is the same as that currently required by the TSs. Therefore, the NRC staff finds that, as required by 10 CFR 50.36(c)(3), the necessary quality of systems will be maintained in accordance with the associated LCOs.

3.6.4.7 Evaluation of Changes to Table of Contents The licensee also proposed a conforming change to the Table of Contents to include the new containment sump TS. This conforming change is acceptable since it is an editorial change to support the inclusion of the new containment sump TS. The NRC staff finds the changes to the Table of Contents acceptable.

3.6.4.8 Conclusion Regarding Proposed Containment Sump TS The new containment sump TS retains and expands the existing TS requirements with the exception of the addition of Condition A. Condition A provides a condition for an inoperable containment sump due to containment accident generated and transported debris exceeding the analyzed limits.

The NRC staff reviewed the proposed changes against the regulations and concludes that the changes continue to meet the relevant requirements of 10 CFR 50.36 for the reasons discussed above and thus provide reasonable assurance that adoption of these TSs will have the requisite requirements and controls to operate safely. Therefore, the NRC staff concludes that the proposed TS changes are acceptable.

3.6.5 Variations The Wolf Creek TSs utilize different numbering than the STSs on which TSTF-567 was based.

Specifically, TS 3.6.19 in TSTF-567 is proposed TS 3.6.8 in the Wolf Creek TSs. These numbering differences are editorial, and do not affect the applicability of TSTF-567 to the proposed LAR.

As the Wolf Creek TSs contain a SFCP, the frequency for SR 3.6.8.1 is [i]n accordance with the Surveillance Frequency Control Program. The SFCP was previously incorporated into the TSs and applied to SR 3.5.2.8 that is proposed to be replaced by SR 3.6.8.1. Although the requirements are somewhat expanded, SR 3.6.8.1 will perform the same function as SR 3.5.2.8,

and the intent of the proposed SR is the same. Therefore, the NRC staff finds it acceptable to apply the SFCP to the proposed SR 3.6.8.1.

The TSTF-567 Required Action B.1 allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the sump to operable for ECCS trains made inoperable by the containment sump(s), but also includes two notes which direct entering the applicable conditions and required actions of LCOs 3.5.2 and 3.5.3 and entering the applicable conditions and required actions of LCO 3.6.6, Containment Spray and Cooling Systems, for CSS and cooling system trains made inoperable by the containment sump(s). The licensee variation to Required Actions B.1 and B.2 requires immediate entry into the applicable ECCS and CSS TSs. This is discussed above in section 3.6.4.4, Evaluation of Condition B.

The NRC staff found the licensees proposed changes to Condition B acceptable as described in SE section 3.6.4.4.

3.7 Technical Evaluation Conclusion

The proposed changes referenced in section 2.3 of this SE, describe how the effects of debris are evaluated and how these effects are incorporated into other calculations like NPSH margin for the pumps taking suction from the ECCS sumps. The NRC staff concluded that the description of the key methods, identified in the USAR markup, used in the risk-informed evaluation were acceptable. Any change in these methods is to be evaluated by the licensee to determine whether a departure from an approved methodology as described in the USAR would arise.

The USAR markup accurately describes the various aspects of the risk-informed analysis and provides references to important guidance documents and calculations in a new Section (6A), of the USAR. Section 6 of the USAR was revised to refer to the information in the new section.

A table depicting the debris limits for the Wolf Creek analysis (table B 3.6.8-1), is added to the TS Bases as shown on page 11 of attachment V to the licensees LAR (Reference 1). This table is referred to as shown in the USAR markup in section 6.A.9 of attachment VI to the licensees LAR. The USAR and TS Bases contain information regarding the analyzed debris limits for various debris types included in the risk-informed analysis as determined by plant-specific testing for Wolf Creek.

The NRC staff concluded that the proposed USAR changes are acceptable.

The licensees submittal contains information regarding reporting and corrective actions associated with the changes made via the LAR. In section 6.0 of attachment VII to the LAR, the licensee describes the criteria for reporting in the event the containment sump is inoperable for longer than the TS completion time as proposed in the LAR. The licensee also stated, in attachment I to the LAR, that changes to the debris limits are subject to the reporting requirements of 10 CFR 50.71(e).

Related to the debris amounts contained in TS Bases table B 3.6.8-1, the licensees submittal contained information regarding how compliance with the risk-informed analysis would be maintained in attachment X of its LAR. The information was provided for information only, but the NRC considers the licensees methods for compliance determination to be important. During the regulatory audit (Reference 16) the NRC staff discussed the methodology that will be used to determine compliance with the regulations and operability associated with the debris effects on the ECCS performance during long-term core cooling. In its supplement to the LAR (Reference 2), the licensee clarified that the debris limits in table B 3.6.8-1 are based on

transport to a single strainer. In addition, the licensee stated that chemical debris limits are not included in this table because they are calculated based on other debris in containment. The licensee provided the actions taken to track the amount of aluminum in containment, because aluminum is the major contributor to chemical debris. The NRC requested clarification on the assumptions for debris margin tables 1 and 2 in attachment X of the licensees LAR.

In response to audit question STSB-Q2 (Reference 16), the licensee provided an updated margin table. The margin table was updated to simplify the determination of strainer operability.

The NRC staff reviewed table 1 provided in the LAR supplement along with the updated erosion and transport fractions used to compute the fine fiber margin values. Upon review of the supplemental information, the NRC staff concluded that the two-train case would be limiting for margin of fiber fines instead of the one-train case as reported in table 1 of the LAR supplement.

This is because the licensee used erosion fractions for the two-train case that result in conservative estimates of fiber fines strainer loads. Nonetheless, the NRC staff concluded that it is acceptable to use the margin table provided in the LAR supplement, as described below.

The two-train case should be considered to establish the fiber fines margin based on the acceptable limit for in-vessel fibrous debris accumulation. The one-train case should be considered to establish fiber fines margin based on strainer debris load limits.

The NRC staff determined that the erosion fractions for the two-train case that the licensee used in the supplemental analysis (Reference 2) are overly conservative, because the licensee accounted for erosion of large and small debris pieces that are calculated to reach the strainer.

Per NRCs guidance, pieces of debris that reach the strainer do not have to be evaluated for erosion, but the licensee conservatively included erosion for these fiber pieces for the two-train case. The fractions of fiber pieces that are assumed to erode and transported as fiber fines in the updated table 3.e.6-8 of the LAR supplement (Reference 2) for the two-train case are 9 percent and 8 percent for small and large pieces of fiber, respectively. If the fiber pieces that reach the strainer are not assumed to erode, the erosion fractions into fiber fines would be 2 percent and 5 percent for small and large pieces of fiber, respectively. Use of the less conservative erosion fractions would result in fiber fines strainer loads for the two-train system less than loads of fiber fines for the one-train system, and the one-train case being limiting, as presented in table 1 of the LAR supplement. The licensee chose to use the one-train margin (strainer) instead of the two-train margin (in-vessel) to simplify operability determinations for the operators.

The NRC staff concluded that the licensee could have used the lower erosion transport fractions for the two-train case because this would be consistent with NRC guidance. The NRC staff noted that the margin for the two-train case would be about 9 lbm of fiber fines less than listed in table 1 if the conservative transport fractions for the two-train case, listed in table 3.e.6-8, were used. The NRC staff also noted that adding 9 lbm of fine fiber to the pool would result in only a slight increase in the amount of fiber reaching the core because most of it would be filtered out by the strainer. For example, simplified computations by the licensee for fiber penetration to establish in-vessel fiber accumulation, assumed that only fiber fines would be present in the pool, to maximize penetration of fiber. Instead small and large pieces of fiber as well as any other debris on the strainer would be expected to filter and capture a large proportion of fiber fines. Also, if 9 lbm of fiber were added as insulation in the containment, it would be located such that it would be very unlikely for all of this fine fiber source to be within the zone-of-influence of the weld producing maximal amounts of fine fibrous debris. The NRC staff also noted that the licensee computed limiting in-vessel debris amounts assuming both RHR pumps

are in service and no CS pumps running, which is a beyond design basis condition leading to conservative estimates of in-vessel fiber accumulation.

In the LAR (Reference 1), the licensee described the steps that would be taken if an operability determination is needed due to the identification of an unanalyzed debris source. If the debris source is determined to exceed the established margins, the sump would be declared inoperable, and the licensee would take the actions required by the plant technical specifications. In addition, the licensee would perform a refined debris analysis or update its risk analysis. This would require detailed calculations which would recalculate the strainer loads and in-vessel debris accumulation and identify required actions to assure that the strainers remain operable. If the sump is determined to be operable, the licensee would be required to remove the debris source during the next outage or update the margin calculation. Any update to the margin calculation would require a detailed analysis that would verify that both the one and two-train cases retain margin.

The current plant condition, as reflected in table 1 of the LAR supplement is acceptable for the following reasons:

1) The table presents a simplified tool to assess sump operability due to discovered debris sources.

2) All debris types are well described and consistent with other information presented in the LAR, except for fine fiber margins.

3) The fine fiber margins presented in the table reflect the transport fractions in updated table 3.e.6-8 for a one-train system and strainer load limits. The one-train margin bounds the margin of a two-train system considering in-vessel fiber limits if fine fibrous debris is calculated using NRC staff guidance for debris transport.

4) If additional fiber that may form debris and transport to the strainer is discovered, an assessment involving detailed calculations will be performed that will identify whether margins remain. If the margins identified in table 1 are exceeded, actions required by technical specifications will be taken to place the plant in a safe condition and additional actions will be taken to assure the plant is returned to operation within acceptable limits.

Therefore, the NRC staff finds table 1 of the LAR supplement to be an acceptable listing of margin for the containment sump.

The margin examples concentrate on single train operation because it is more limiting due to all debris collecting on one strainer instead of two. The margin calculation also provides adequate assurance that the in-vessel limit will not be exceeded during two-train operation.

Based on the clarifications in the supplement, the NRC staff concluded that the margins provided in the LAR are acceptable.

As discussed in this SE, the NRC staff reviewed the licensees LAR, as supplemented. The NRC staff finds that the information provided by the licensee demonstrates that there is reasonable assurance that debris will not adversely affect LTCC at Wolf Creek. The NRC staffs conclusions are described in detail in previous sections of this SE. The NRC staff finds that the licensee adequately addressed each technical area of GL 2004-02 using deterministic methods.

For a limited set of scenarios that do not meet the acceptance criteria using the accepted

deterministic methods, the licensee demonstrated that the cumulative risk of these scenarios is very small. To complete its evaluation of these low-risk scenarios, the licensee demonstrated that the five key principles of RG 1.174, Revision 3, were met. Therefore, the NRC staff concludes that the licensees risk-informed methodology for assessing the effects of debris on LTCC at Wolf Creek (including submodels and integration of those submodels) demonstrates that the requirements of 10 CFR 50.46 for LTCC will not be adversely affected by debris.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Kansas State official was notified of the proposed issuance of the amendment on December 2, 2025. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, published in the Federal Register on November 26, 2024 (89 FR 93365), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1.

Boyce, M. T., Wolf Creek Nuclear Operating Corporation, letter to NRC, Docket No. 50-482: License Amendment Request for a Risk-Informed Resolution to GSI-191, dated September 12, 2024 (Agencywide Documents Access and Management System Accession (ADAMS) Accession No. ML24260A071).

2.

Bayer, R. J., Wolf Creek Nuclear Operating Corporation, letter to NRC, Docket No. 50-482: Supplement to License Amendment Request for a Risk-Informed Resolution to GSI-191, dated August 28, 2025 (ML25240B584).

3.

NRC, NRC Generic letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, dated September 13, 2004 (ML042360586).

4.

Vietti-Cook, A. L., NRC, memorandum to R. W. Borchardt, NRC, Staff Requirements -

SECY-10-0113 - Closure Options for Generic Safety Issue-191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance, SRM SECY-10-0113, dated December 23, 2010 (ML103570354).

5.

NRC, Closure Options for Generic Safety Issue-191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, SECY-12-0093, dated July 9, 2012 (ML121310648).

6.

Vietti-Cook, A. L., NRC, memorandum to R. W. Borchardt, NRC, Staff Requirements -

SECY-12-0093 - Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, SRM SECY-12-0093, dated December 14, 2012 (ML12349A378).

7.

Lee, S. S., NRC, letter to Wolf Creek Nuclear Operating Corporation, Summary of November 18, 2020, Pre-Submittal Teleconference with Wolf Creek Nuclear Operating Corporation regarding a Risk-Informed Generic Safety Issue ((GSI)-191 License Amendment Request for Wolf Creek Generating Station, Unit 1 (EPID L-2020-LRM-0099), dated November 30, 2020 (ML20332A046).

8.

Smith, S. L., Wolf Creek Nuclear Operating Corporation, letter to NRC, Docket No. 50-482: License Amendment Request for a Risk-Informed Resolution to GSI-191, dated August 12, 2021 (ML21224A118).

9.

Lee, S. S., NRC, letter to C. Reasoner, Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Unit 1 - Supplemental Information Needed for Acceptance of Requested Licensing Actions Re License Amendment Request and Regulatory Exemption for a Risk-Informed Approach to Address Generic Safety Issue 191 and Respond to Generic Letter 2004-02 (EPID L-2021-LLA-0152 and EPID L-2021-LLE-0039), dated September 23, 2021 (ML21253A090).

10.

Smith, S. L., Wolf Creek Nuclear Operating Corporation, letter to NRC, Docket No. 50-482: Withdrawal of License Amendment Request for a Risk-Informed Resolution to GSI-191 (EPID L-2021-LLA-0152 and EPID L-2021-LLE-0039), dated October 20, 2021 (ML21294A051).

11.

Regner, L. M., NRC, letter to G. T. Powell, STP Nuclear Operating Company, 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 (CAC Nos. MF2400 and MF2401), dated July 11, 2017 (Package ML17019A001).

12.

Framatome Inc. and Westinghouse, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090), Topical Report WCAP-17788-NP, Volume 1, Revision 1, December 2019 (ML20009F699).

13.

NRC, Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling, Draft Regulatory Guide 1.229, March 2016 and January 2017 (ML16062A016 and Package ML17025A060, respectively).

14.

NRC, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, NUREG-1829, Volume 1, Main Report, and Volume 2, Appendices A through M, April 2008 (ML082250436 and ML081060300, respectively).

15.

NRC, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018 (ML17317A256).

16.

Lee, S. S., NRC, letter to C. Reasoner, Wolf Creek Nuclear Operating Corporation, 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 (EPID L-2024-LLA-0125 and EPID L-2024-LLE-0026), dated June 10, 2025 (ML25143A051).

17.

Nuclear Energy Institute, Pressurized Water Reactor Sump Performance Evaluation Methodology, NEI 04-07, Revision 0, Volume 1, December 2004 (050550138).

18.

NRC, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report (Proposed Document Number NEI 04-07), Pressurized Water Reactor Sump Performance Evaluation Methodology, NEI 04-07, Revision 0, Volume 2, December 6, 2004 (ML050550156).

19.

Westinghouse Electric Company LLC, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191, TR WCAP-16530-NP-A, March 2008 (ML081150379).

20.

Nieh, H. K., NRC, letter to G. Bischoff, Westinghouse Electric Company, Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR)

WCAP-16530-NP, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191 (TAC No. MD1119), dated December 21, 2007 (ML073521072).

21.

Buschbaum, D. E., PWR Owners Group, letter to NRC, PWR Owner Group Transmittal of NRC Approved Topical Report WCAP-16406-P-A, Revision 1, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, dated March 28, 2008 (ML081000025).

22.

NRC, Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report (TR) WCAP-16406-P, Revision 1, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, Pressurized Water Reactor Owners Group Project No. 694, dated December 20, 2007 (ML073520295).

23.

Westinghouse Electric Company LLC, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) - Autoclave Chemical Effects Testing for GSI-191 Long-Term Cooling, TR WCAP-17788-NP Volume 5, Revision 1, December 2019 (ML20009F670).

24.

Ruland, W. H., NRC, letter to A. Pietrangelo, Nuclear Energy Institute, Revised Content Guide for Generic Letter 2004-02 Supplemental Responses, dated November 21, 2007 (Package ML073110389).

25.

Ruland, W. H., NRC, letter to A. Pietrangelo, Nuclear Energy Institute, Revised Guidance for Review of Final Licensee Responses to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, dated March 28, 2008 (Package ML080230234).

26.

NRC, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, Regulatory Guide 1.82, Revision 4, March 2012 (ML111330278).

27.

NRC, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, Regulatory Guide 1.82, Revision 5, August 2022 (ML22152A114).

28.

NRC, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Regulatory Guide 1.200, Revision 2, March 2009 (ML090410014).

29.

NRC, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 3, December 2020 (ML20238B871).

30.

American Society of Mechanical Engineers / American Nuclear Society, Addenda to ASME/ANS RA-S 2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, PRA Standard ASME/ANS, February 2009, New York, NY.

31.

NRC, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, NUREG-0800, Section 19.2, Revision 3, June 2007 (ML071700658).

32.

Technical Specifications Task Force, letter to NRC, Transmittal of TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues, dated August 2, 2017 (ML17214A813).

33.

Cusumano, V., NRC, letter to Technical Specifications Task Force, Final Safety Evaluation of Technical Specification Task Force Traveler TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (EPID: L-2017-PMP-0005), dated July 3, 2018 (Package ML18109A077).

34.

Cusumano, V. G., NRC, memorandum to J. E. Marshall, NRC, U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel downstream Effects Supporting Review of Generic Letter 2004-02 Responses, dated September 4, 2019 (ML19228A011).

35.

Leeds, E. J., and Johnson, M. R., NRC, letter to All Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, Order EA-12-049, dated March 12, 2012 (Package ML12054A736).

36.

Nuclear Energy Institute, Guideline for the Management of Materials Issues, NEI 03-08, Revision 4, dated October 2020 (ML20315A536).

37.

NRC, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, Regulatory Guide 1.45, Revision 1, May 2008 (ML073200271).

38.

Wolf Creek Nuclear Operating Corporation, Revision 34 to Updated Safety Analysis Report, Chapter 5, Reactor Coolant System and Connected Systems for Wolf Creek Generating Station, June 2021 (ML21168A117).

39.

NRC, NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing, March 2008 (ML080230038).

40.

Vietti-Cook, A. L., memorandum to L. A. Reyes, NRC, Staff Requirements -

SECY-04-0118 - Plan for the Implementation of the Commissions Phased Approach to Probabilistic Risk Assessment Quality, SRM-SECY-04-0118, dated October 6, 2004 (ML042800369).

41.

Nuclear Energy Institute, Probabilistic Risk Assessment (PRA) Peer Review Process Guidance, NEI 00-02, Revision 1, May 2006 (ML061510621 and ML061510614).

42.

Nuclear Energy Institute, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, NEI 05-04, Revision 2, November 2008 (ML083430462).

43.

NRC, Regulatory Guide 1.200 Implementation, Regulatory Issue Summary 2007-06, dated March 22, 2007 (ML070650428).

44.

Nuclear Energy Institute, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, NEI 17-07, Revision 2, August 2019 (ML19231A182).

45.

NRC, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, NUREG-1855, Volume 1, Main Report, March 2009 (ML090970525).

46.

NRC, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, Generic Letter 2008-01, dated January 11, 2008 (ML072910759).

47.

Singal, B. K., NRC, letter to R. A. Muench, Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station - Re: Closure Letter for Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems (TAC No. MD7896), dated October 28, 2009 (ML092600247).

48.

Electric Power Research Institute, Seismic Probabilistic Risk Assessment Implementation Guide, EPRI Report 3002000709, Palo Alto, CA, December 2013.

49.

NRC, Seismic Considerations for the Transition Break Size, NUREG-1903, February 2008 (ML080880140).

50.

Lingam S. for Chawla, M. L., NRC, letter to F. Diya, Union Electric Company, doing business as Ameren Missouri, Callaway Plant, Unit No. 1 - Issuance of Amendment No. 228 Re: Revise Technical Specifications to Address Generic Safety Issue-191 and Respond to Generic Letter 2004-02 Using a Risk-Informed Approach (EPID L-2021-LLA-0059), dated October 21, 2022 (ML22220A132).

51.

Markley, M. T., NRC, letter to C. A. Gayheart, Southern Nuclear Operating Co., Inc.,

Final Staff Evaluation for Vogtle Electric Generating Plant, Units 1 and 2, Systematic Risk-Informed Assessment of Debris Technical Report (EPID L-2017-TOP-0038), dated September 30, 2019 (ML19228A011).

52.

Westinghouse Electric Company, Transmittal of Responses to NRC Request for Additional Information Re: Pressurized Water Reactor Owners Group Bases for Licensee Debris Generation Assumptions for GSI-191, dated March 5, 2010 (ML100710710).

53.

Rowley, J., NRC, letter to Mr. A. Nowinowski, Westinghouse Electric Company, Nuclear Regulatory Commission Conclusions Regarding Pressurized Water Reactor Owners Group Response to Request for Additional Information Dated January 25, 2010.

Regarding Licensee Debris Generation Assumptions for GSI-191, Dated March 31, 2010 (ML100570364).

54.

Ruland, W. H., NRC, letter to A. Marion, Nuclear Energy Institute, Revised Guidance Regarding Coatings Zone of Influence for Review of Final Licensee Responses to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, dated April 6, 2010 (ML100960495).

55.

Choromokos, R., ALION Science and Technology, letter to NRC, ALION Science and Technology - GSI-191 Low Density Fiberglass Erosion Testing, ALION-REP-ALION-1006-04, Revision 0, dated April 8, 2010 (Package ML101090490; not publicly available, proprietary information).

56.

Ruland, W. H., NRC, letter to R. Choromokos, ALION Science and Technology, Proprietary Erosion Testing of Submerged NUKON Low-Density Fiberglass Insulation in Support of Generic Safety Issue 191 Strainer Performance Analyses, dated June 30, 2010 (ML101540221).

57.

NRC, Containment Emergency Sump Performance, NUREG-0897, Revision 1, October 1985 (ML112440046).

58.

Bahn, C. B., Kasza, K.E., Shack, W. J., and Natesan, K., Argonne National Laboratory, Aluminum Solubility in Boron Containing Solutions as a Function of pH and Temperature, September 19, 2008 (ML091610696).

Principal Contributors: S. Smith O. Pensado C. Moulton S. Park J. Robinson J. Tsao S. Lai P. Klein R. Beaton Y. Chu M. Yoder A. Russell Date: January 14, 2026

ML25317A782

via memo dated OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DRA/APLB/BC NRR/DRA/APLC/BC NAME TByrd PBlechman EDavidson RPascarelli DATE 11/19/2025 12/11/2025 w/comments 12/12/2025 1/13/2026 OFFICE NRR/DNRL/NCSG/BC NRR/DEX/ESEB/BC NRR/DNRL/NVIB/BC NRR/DSS/SNSB NAME SBloom ITseng ABuford NDifrancesco DATE 12/11/2025 1/12/2026 1/13/2026 1/12/2026 OFFICE NRR/DSS/STSB/BC NRR/DORL/LPL4/ABC NRR/DORL/LPL4/PM NAME SMehta MMahoney TByrd DATE 12/15/2025 1/14/2026 1/14/2026

v t e Letter Sequence Approval
EPID:L-2024-LLA-0125, (Open)
Initiation Request Acceptance Administration Questions Answers Results Approval, Approval, Approval Other: ML25143A051
MONTHYEAR2025-01-14 [Table View]