ML17200D162

License Amendment Request to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors

ML17200D162
Person / Time
Site:	Palo Verde
Issue date:	07/19/2017
From:	Lacal M Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-07546-MLL/TNW
Download: ML17200D162 (36)

Text

10 CFR 50.90 10 CFR 50.69 MARIA L. LACAL Senior Vice President, Nuclear Regulatory & Oversight Palo Verde Nuclear Generating Station P.O. Box 52034 102-07546-MLL/TNW Phoenix, AZ 85072 July 19, 2017 Mail Station 7605 Tel 623.393.6491 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk

Subject:

Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529, and 50-530 Renewed Operating License Nos. NPF-41, NPF-51, NPF-74 License Amendment Request to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors In accordance with the provisions of Section 50.69(b)(2) and 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Arizona Public Service Company (APS) is requesting an amendment to the renewed operating license of Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3.

The proposed license amendment request (LAR) would modify the licensing basis, by the addition of a License Condition, to allow for the implementation of the provisions of 10 CFR, Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has high safety significance resulting in improved plant safety.

The enclosure to this letter provides the basis for the proposed change to the PVNGS Units 1, 2, and 3 renewed operating licenses. The categorization process being implemented through this change is consistent with Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, dated July 2005, which was endorsed by the Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, dated May 2006. Attachment 1 of the enclosure provides a list of categorization prerequisites. Use of the categorization process on a plant structure, system, or component will only occur after these prerequisites are met.

A member of the STARS Alliance LLC Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek

102-07546-MLL/TNW ATTN: Document Control Desk U. S. Nuclear Regulatory Commission License Amendment Request to Adopt 10 CFR 50.69 Page 2 The PVNGS Probabilistic Risk Assessment (PRA) models are described in Attachments 2 through 5 of the enclosure. The PRA models described within this LAR are the same as those described within the APS submittal of the LAR dated July 31, 2015, to revise the PVNGS Technical Specifications (TS) to allow risk-informed completion times [Agencywide Document Access and Management System (ADAMS) Accession Number ML15218A300],

with routine maintenance updates applied. APS has also recently conducted a facts and observations (F&O) closure review of PRA peer review findings in accordance with an NRC letter dated May 3, 2017 (ADAMS Accession Number ML17079A427).

APS requests that the NRC staff utilize insights from their on-going review of the technical adequacy of PRA models in the risk-informed completion times LAR as well as the F&O closure review results to inform their review of the same PRA models for 10 CFR 50.69.

This would reduce the number of APS and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR can be independently reviewed on their own merits without regard to the results from the review of the other.

In accordance with the PVNGS Quality Assurance Program, the Plant Review Board and the Offsite Safety Review Committee have reviewed and approved this LAR. By copy of this letter, this LAR is being forwarded to the Arizona Radiation Regulatory Agency in accordance with 10 CFR 50.91(b)(1).

APS requests approval of the proposed license amendment within one year of the date of this letter, with the amendment being implemented within 90 days of issuance. No new commitments are made by this letter. Should you have any questions concerning the content of this letter, please contact Michael DiLorenzo, Licensing Section Leader, at (623) 393-3495.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: July 19, 2017 (Date)

Sincerely, Digitally signed by Lacal, Lacal, Maria Maria L(Z06149)

DN: cn=Lacal, Maria L(Z06149)

L(Z06149)

Date: 2017.07.19 16:13:27

-07'00' MLL/TNW/NTA

Enclosure:

Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components cc: K. M. Kennedy NRC Region IV Regional Administrator S. P. Lingam NRC NRR Project Manager for PVNGS M. M. Watford OBanion NRC NRR Project Manager C. A. Peabody NRC Senior Resident Inspector for PVNGS T. Morales Arizona Radiation Regulatory Agency (ARRA)

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components TABLE OF CONTENTS 1

SUMMARY

DESCRIPTION .................................................................................. 1 2 DETAILED DESCRIPTION .................................................................................. 1 2.1 Current Regulatory Requirements ......................................................................... 1 2.2 Reason for Proposed Change ................................................................................. 1 2.3 Description of the Proposed Change ...................................................................... 2 3 TECHNICAL EVALUATION ................................................................................. 3 3.1 Categorization Process Description [10 CFR 50.69(b)(2)(i)]................................. 4 3.1.1 Overall Categorization Process ............................................................. 4 3.1.2 Passive Categorization Process ............................................................. 5 3.2 Technical Adequacy Evaluation [10 CFR 50.69(b)(2)(ii)] ...................................... 6 3.2.1 Internal Events and Internal Flooding.................................................... 6 3.2.2 Fire Hazards ...................................................................................... 7 3.2.3 Seismic Hazards ................................................................................ 7 3.2.4 Other External Hazards ....................................................................... 7 3.2.5 Low Power & Shutdown....................................................................... 7 3.2.6 PRA Maintenance and Updates ............................................................. 8 3.2.7 PRA Uncertainty Evaluations ................................................................ 8 3.3 PRA Review Process Results [10 CFR 50.69(b)(2)(iii)]......................................... 9 3.4 Risk Evaluations [10 CFR 50.69(b)(2)(iv)] ......................................................... 10 4 REGULATORY EVALUATION ............................................................................11 4.1 Applicable Regulatory Requirements ................................................................... 11 4.2 No Significant Hazards Consideration .................................................................. 11 4.3 Conclusions ......................................................................................................... 12 5 ENVIRONMENTAL CONSIDERATION ............................................................... 13 6 REFERENCES .................................................................................................. 13 LIST OF ATTACHMENTS:

1. List of Categorization Prerequisites

2. Total Unit 1/2/3 Baseline Average Annual CDF/LERF

3. Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process

4. External Hazards Screening

5. Progressive Screening Approach for Addressing External Hazards i

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components 1

SUMMARY

DESCRIPTION The proposed amendment would modify the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance (LSS), alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance (HSS), requirements will not be changed or will be enhanced. This allows improved focus on equipment that has high safety significance resulting in improved plant safety.

2 DETAILED DESCRIPTION 2.1 Current Regulatory Requirements The Nuclear Regulatory Commission (NRC) has established a set of regulatory requirements for commercial nuclear reactors to ensure that a reactor facility does not impose an undue risk to the health and safety of the public, thereby providing reasonable assurance of adequate protection to public health and safety. The current body of NRC regulations and their implementation are largely based on a deterministic approach.

This deterministic approach establishes requirements for engineering margin and quality assurance in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist (e.g., equipment failures and human errors) and establishes a specific set of design basis events (DBEs). The deterministic approach then requires that the facility include safety systems capable of preventing or mitigating the consequences of those DBEs to protect public health and safety. Those structures, systems and components (SSCs) necessary to defend against the DBEs are defined as safety-related, and these SSCs are the subject of many regulatory requirements, herein referred to as special treatments, designed to ensure that they are of high quality and high reliability, and have the capability to perform during postulated design basis conditions. Treatments include, but are not limited to, quality assurance, testing, inspection, condition monitoring, assessment, evaluation, and resolution of deviations. The distinction between treatment and special treatment is the degree of NRC specification as to what must be implemented for particular SSCs or for particular conditions. Typically, the regulations establish the scope of SSCs that receive special treatment using one of three different terms: safety-related, important to safety, or basic component. The terms safety-related and basic component are defined in the regulations, while important to safety, used principally in the general design criteria (GDC) of Appendix A to 10 CFR Part 50, is not explicitly defined.

2.2 Reason for Proposed Change A probabilistic approach to regulation enhances and extends the traditional deterministic approach by allowing consideration of a broader set of potential challenges to safety, providing a logical means for prioritizing these challenges based on safety significance, and allowing consideration of a broader set of resources to defend against these challenges. In contrast to the deterministic approach, Probabilistic Risk Assessments (PRAs) address 1

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components credible initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for common cause failures. The probabilistic approach to regulation is an extension and enhancement of traditional regulation by considering risk in a comprehensive manner.

To take advantage of the safety enhancements available through the use of PRA, in 2004 the NRC published a new regulation, 10 CFR 50.69. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with the regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has high safety significance resulting in improved plant safety.

The rule contains requirements on how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the SSC, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four risk-informed safety class (RISC) categories. The determination of safety significance is performed by an integrated decision-making process, as described by Nuclear Energy Institute (NEI) 00-04, 10 CFR 50.69 SSC Categorization Guideline (Reference 2), which uses both risk insights and traditional engineering insights. The safety functions include the design basis functions, as well as functions credited for severe accidents (including external events). Special or alternative treatment for the SSCs is applied as necessary to maintain functionality and reliability, and is a function of the SSC categorization results and associated bases. Finally, periodic assessment activities are conducted to make adjustments to the categorization and/or treatment processes as needed so that SSCs continue to meet all applicable functional requirements.

The rule does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility. Instead, the rule enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as high safety significant, existing treatment requirements are maintained or enhanced. Conversely, for SSCs that do not significantly contribute to plant safety on an individual basis, the rule allows an alternative risk-informed approach to treatment that provides a reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

Implementation of 10 CFR 50.69 will allow APS to improve focus on equipment that has high safety significance resulting in improved plant safety.

2.3 Description of the Proposed Change APS proposes the addition of the following condition to the PVNGS renewed operating licenses of Units 1, 2, and 3 to document NRC approval to use 10 CFR 50.69.

APS is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the license amendment dated [Date].

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Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

3 TECHNICAL EVALUATION 10 CFR 50.69 specifies the information to be provided by a licensee requesting adoption of the regulation. This request conforms to the requirements of 10 CFR 50.69(b)(2), which states:

A licensee voluntarily choosing to implement this section shall submit an application for license amendment under 10 CFR 50.90 that contains the following information:

(i) A description of the process for categorization of RISC-1, RISC-2, RISC-3 and RISC-4 SSCs. (See section 3.1 of this enclosure)

(ii) A description of the measures taken to assure that the quality and level of detail of the systematic processes that evaluate the plant for internal and external events during normal operation, low power, and shutdown (including the plant-specific probabilistic risk assessment (PRA), margins-type approaches, or other systematic evaluation techniques used to evaluate severe accident vulnerabilities) are adequate for the categorization of SSCs.

(See section 3.2 of this enclosure)

(iii) Results of the PRA review process conducted to meet § 50.69(c)(1)(i). (See section 3.3 of this enclosure)

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy § 50.69(c)(1)(iv). The evaluations must include the effects of common cause interaction susceptibility, and the potential impacts from known degradation mechanisms for both active and passive functions, and address internally and externally initiated events and plant operating modes (e.g., full power and shutdown conditions). (See section 3.4 of this enclosure)

Each of these submittal requirements are addressed in the succeeding sections.

The PRA models described within this LAR are the same as those described within the APS submittal of the LAR dated July 31, 2015, to revise the PVNGS Technical Specifications (TS) to allow risk-informed completion times (Reference 9), with routine maintenance updates applied. APS has also recently conducted a F&O closure review of PRA peer review findings in accordance with an NRC letter dated May 3, 2017 (Reference 17).

APS requests that the NRC staff utilize insights from their on-going review of the technical adequacy of PRA models in the risk-informed completion times LAR as well as the F&O closure review results to inform their review of the same PRA models for 10 CFR 50.69. This would reduce the number of APS and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR can be independently reviewed on their own merits without regard to the results from the review of the other.

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Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components 3.1 Categorization Process Description [10 CFR 50.69(b)(2)(i)]

3.1.1 Overall Categorization Process APS will implement the risk categorization process in accordance with NEI 00-04, Revision 0 (Reference 2), as endorsed by Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, (Reference 1). NEI 00-04 Section 1.5 states: Due to the varying levels of uncertainty and degrees of conservatism in the spectrum of risk contributors, the risk significance of SSCs is assessed separately from-each of five risk perspectives and used to identify SSCs that are potentially safety-significant. Separate evaluation is appropriate to avoid reliance on a combined result that may mask the results of individual risk contributors.

The following are the clarifications taken to the NEI 00-04 categorization process:

• The Integrated Decision Making Panel (IDP) will be composed of a group of at least five experts who collectively have expertise in plant operation, design (mechanical and electrical) engineering, system engineering, safety analysis, and probabilistic risk assessment. At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has worked on the modeling and updating of the plant-specific PRA for a minimum of three years.

• The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address, at a minimum, the purpose of the categorization; present treatment requirements for SSCs including requirements for design basis events; PRA fundamentals; details of the plant-specific PRA including the modeling, scope, and assumptions, the interpretation of risk-importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the defense-in-depth philosophy and requirements to maintain this philosophy.

• The decision criteria for the IDP for categorizing SSCs as HSS or LSS pursuant to

§ 50.69(f)(1) will be documented in APS procedures. Decisions of the IDP will be arrived at by consensus. Differing opinions will be documented and resolved, if possible. If a resolution cannot be achieved concerning the safety significance of an SSC, then the SSC will be classified as HSS.

• Passive characterization will be performed using the processes described in Section 3.1.2 of this enclosure.

• An unreliability factor of 3 will be used for the sensitivity studies described in Section 8 of NEI 00-04. The factor of 3 was chosen as it is representative of the typical error factor of basic events used in the PRA model.

• APS will require that if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6 of NEI 00-04), the associated system function(s) would be identified as HSS.

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Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components

• Once a system function is identified as HSS, then all the components that support that function are preliminarily identified as HSS. The IDP must intervene to assign any of these HSS function components to LSS.

• With regard to the criterion that considers whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, APS will not take credit for alternate means unless the alternate means are proceduralized and included in licensed operator training.

The risk analysis being implemented for each hazard is described below:

• Internal Event Risks: Internal events including internal flooding PRA model (same as described in Reference 9).

• Fire Risks: Internal Fire PRA model consistent with NUREG/CR-6850 (Reference

10) methodology (same as described in Reference 9).

• Seismic Risks: Seismic PRA model (same as described in Reference 9).

• Other External Risks (e.g., tornados, external floods, etc.): Screened out as not requiring PRA models as described in Reference 9, as the other external hazards were determined to be insignificant contributors to plant risk.

• Low Power and Shutdown Risks: Qualitative defense-in-depth (DID) shutdown model for shutdown configuration risk management (CRM) based on the framework for DID provided in NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management (Reference 3), which provides guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above (e.g.,

change from a seismic margins approach to a seismic probabilistic risk assessment approach) will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization

2. System functions, identified and categorized with the associated bases

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies

5. Hazards analyses, as applicable

6. Passive categorization results and bases

7. Categorization results including all associated bases and RISC classifications

8. Component critical attributes for HSS SSCs

9. Results of period reviews and SSC performance evaluations, and

10. IDP meeting minutes and qualification/training records for the IDP members 3.1.2 Passive Categorization Process For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Risk-Informed Repair/Replacement Activities (RI-RRA) methodology consistent with the Safety Evaluation (SE) by the Office of Nuclear Reactor Regulation for Arkansas Nuclear One, 5

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Unit 2, regarding their Request to Use Risk-informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems, dated April 22, 2009 (Reference 5).

The RI-RRA methodology is a risk-informed safety classification and treatment program previously used for repair/replacement activities (RI-RRA methodology) for pressure retaining items and their associated supports. In this method, the component failure is assumed with a probability of 1.0 and only the consequence evaluation is performed. It additionally applies deterministic considerations (e.g., defense in depth, safety margins) in determining safety significance. Component supports are assigned the same safety significance as the highest passively ranked component within the bounds of the associated analytical pipe stress model.

The use of this method was previously approved to be used for a 10 CFR 50.69 application by the NRC in the final SE for Vogtle Electric Generating Plant dated December 17, 2014 (Reference 6). The RI-RRA method as approved for use at Vogtle Electric Generating Plant for 10 CFR 50.69 does not have any plant specific aspects and is generic. It relies on the conditional core damage and large early release probabilities associated with postulated ruptures. Safety significance is generally measured by the frequency and the consequence of the event. However, this RI-RRA process categorizes components solely based on consequence, which measures the safety significance of the passive component given that it ruptures. This approach is conservative compared to including the rupture frequency in the categorization as this approach will not allow the categorization of SSCs to be affected by any changes in frequency due to changes in treatment. Therefore, the RI-RRA methodology for passive categorization is acceptable and appropriate for use at PVNGS for 10 CFR 50.69.

3.2 Technical Adequacy Evaluation [10 CFR 50.69(b)(2)(ii)]

The following sections demonstrate that the quality and level of detail of the processes used in categorization of SSCs are adequate. The PRA models described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed. The PRA models described within this LAR are the same as those described within the APS submittal of the LAR dated July 31, 2015, to revise the PVNGS TS to allow risk-informed completion times (Reference 9), with routine maintenance updates applied. Changes and plant modifications previously identified in Reference 9 that were required to achieve an overall Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) consistent with NRC Regulatory Guide 1.174 (Reference 16) have been completed.

3.2.1 Internal Events and Internal Flooding The PVNGS categorization process for the internal events and internal flooding hazards will use peer reviewed plant-specific Internal Events and Internal Flooding PRA models in accordance with RG 1.200, Revision 2 (Reference 7). The APS risk management process ensures that the PRA models used in this application reflects the as-built and as-operated plant for each of the PVNGS units. Only industry consensus methods were utilized in the development of the Internal Events and Internal Flooding PRA models.

Attachment 2 of this enclosure identifies the Baseline Average Annual CDF and LERF for the Internal Events and Internal Flooding PRA models.

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Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components 3.2.2 Fire Hazards The PVNGS categorization process for fire hazards will use a peer reviewed plant-specific Internal Fire PRA model in accordance with RG 1.200, Revision 2 (Reference 7).

The APS risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the PVNGS units. Industry consensus methods were utilized in the development of the Internal Fire PRA model.

While APS was not an applicant to implement National Fire Protection Association Standard (NFPA) 805 in accordance with 10 CFR 50.48, the Internal Fire PRA model was developed consistent with NUREG/CR-6850 and only utilizes NRC approved methods. As part of the ongoing PRA maintenance and update process described in Section 3.2.6, APS will address Internal Fire PRA methods approved by the NRC since the development of the Internal Fire PRA. Note that APS does not credit incipient fire detection systems in the Internal Fire PRA model. Attachment 2 of this enclosure identifies the Baseline Average Annual CDF and LERF for the Internal Fire PRA model.

3.2.3 Seismic Hazards The PVNGS categorization process for seismic hazards will use a peer reviewed plant-specific Seismic PRA model in accordance with RG 1.200, Revision 2 (Reference 7). The APS risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the PVNGS units. Only industry consensus methods were utilized in the development of the seismic hazards for the seismic PRA. Attachment 2 of this enclosure identifies the Baseline Average Annual CDF and LERF for the Seismic PRA model.

3.2.4 Other External Hazards The PVNGS categorization process for the external hazards will use a peer reviewed plant-specific screening in accordance with RG 1.200, Revision 2 (Reference 7). Each external hazard was evaluated with respect to applicability and/or risk. The ASME PRA Standard RA-Sa-2009, Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 15) outlines preliminary and progressive screening approaches that are acceptable for this task. The screening started with the top approach and progressed downward until the hazard in question screened with respect to risk. If none of the screening approaches were successful, then the hazard was analyzed using a detailed PRA approach that meets applicable requirements in the ASME PRA Standard RA-Sa-2009. Implicit in these screening criteria (ones that do not present a quantitative measure) is the assumption that successfully meeting a criterion for screening indicates that the bounding CDF from that hazard is considered to be lower than 1E-6 per year. Attachment 4 provides a summary of the other external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for addressing external hazards.

3.2.5 Low Power & Shutdown The PVNGS categorization process will use the shutdown safety management plan described in NUMARC 91-06 (Reference 3), for evaluation of safety significance related to low power and shutdown conditions.

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Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components 3.2.6 PRA Maintenance and Updates The APS risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant for each of the PVNGS units. The process delineates the responsibilities and guidelines for updating the PRA model, and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA model (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience) for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The process will assess the impact of these changes on the plant PRA model in a timely manner but no longer than once every two refueling outages. If there is a significant impact on the PRA model, the SSC categorization will be re-evaluated.

In addition, APS will implement a process that addresses the requirements in NEI 00-04, Section 11, Program Documentation and Change Control (Reference 2). The process will review the results of periodic and interim updates of the plant PRA that may affect the results of the categorization process. If the results are affected, adjustments will be made as necessary to the categorization or treatment processes to maintain the validity of the processes. In addition, any PRA model upgrades will be peer reviewed prior to implementing those changes in the PRA model used for categorization.

3.2.7 PRA Uncertainty Evaluations Uncertainty evaluations associated with any applicable baseline PRA models used in this application were evaluated during the assessment of PRA technical adequacy and confirmed through the self-assessment and peer review processes as discussed in Section 3.3 of this enclosure.

Uncertainty evaluations associated with the risk categorization processes are addressed using the process discussed in Section 8 and in the prescribed sensitivity studies discussed in Section 5 of NEI 00-04 (Reference 2).

In the overall risk sensitivity studies, APS will utilize a factor of 3 to increase the unavailability or unreliability of low safety significant (LSS) components consistent with that approved by the NRC in the Vogtle Electric Generating Plant 10 CFR 50.69 License Amendment Safety Evaluation Report (Reference 6). Consistent with the NEI 00-04 guidance, APS will perform both an initial sensitivity study and a cumulative sensitivity study. The initial sensitivity study applies to the system that is being categorized. In the cumulative sensitivity study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in PRAs for all systems that have been categorized are increased by a factor of 3.

This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low.

The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

Sources of model uncertainty and related assumptions have been identified for the PVNGS PRA models using the guidance of NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making (Reference 11) and 8

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components EPRI TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments (Reference 12).

The detailed process of identifying, characterizing and qualitative screening of model uncertainties is found in Section 5.3 of NUREG-1855 and Section 3.1.1 of EPRI TR-1016737. The process in these references was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups.

The list of assumptions and sources of uncertainty were reviewed to identify those which would be significant for the evaluation of this application. If the PVNGS PRA model used a non-conservative treatment, or methods which are not commonly accepted, the underlying assumption or source of uncertainty was reviewed to determine its impact on this application. Only those assumptions or sources of uncertainty that could significantly impact the risk ranking calculations were considered key for this application.

Key PVNGS PRA model specific assumptions and sources of uncertainty for this application were evaluated and documented. These key assumptions and sources of uncertainty reviewed were previously submitted to the NRC in the application dated July 31, 2015 (Reference 9) for risk-informed completion times. The conclusion of the review for this application is that no additional sensitivity analyses are required to address PVNGS PRA model specific assumptions or sources of uncertainty except for the following:

• In the process of categorizing SSCs into risk-informed safety classifications, APS will include in the risk sensitivity study a sensitivity increasing all the Seismic PRA human events failures (HEFs) derived from the internal events PRA model by a factor of 3 to address the uncertainty associated with main control room actions that might take longer in a seismic event versus an internal initiating event.

3.3 PRA Review Process Results [10 CFR 50.69(b)(2)(iii)]

The PRA models described in Section 3.2 have been assessed against RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 7) consistent with NRC RIS 2007-06, Regulatory Guide 1.200 Implementation (Reference 24).

The Internal Events PRA model was peer reviewed in July 1999 by the Combustion Engineering Owners Group (CEOG) prior to the issuance of Regulatory Guide 1.200 (Reference 19). As a result, a self-assessment of the Internal Events PRA model was conducted by APS in March 2011 (Reference 20) in accordance with Appendix B of RG 1.200, Revision 2 (Reference 7), to address the PRA quality requirements not considered in the CEOG peer review. APS conducted a full scope Internal Flooding PRA model peer review (Reference 21) in November 2010, in accordance with RG 1.200, Revision 2 (Reference 7).

The Internal Events PRA quality (including the CEOG peer review and self-assessment results) has previously been reviewed by the NRC in requests to extend the Inverter Technical Specification Completion Time dated September 29, 2010 (Reference 13), and to implement TSTF-425, Relocate Surveillance Frequencies to Licensee Control RITSTF Initiative 5b, December 15, 2011 (Reference 14). All PRA upgrades (as defined by the ASME PRA Standard RA-Sa-2009 [Reference 15]) implemented since conduct of the CEOG peer review in 1999 have been peer reviewed.

9

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components APS conducted a full scope Internal Fire PRA model peer review (Reference 22) in December 2012 in accordance with RG 1.200, Revision 2 (Reference 7). APS conducted a second focused scope peer review of the Internal Fire PRA in December 2014 (Reference 23), to address ASME PRA Standard (Reference 15) supporting requirements determined not met to Capability Category II in the first peer review, not just the associated facts and observations (F&Os) from the first peer review. Thus, the second peer review generated new F&Os which replaced in their entirety the finding level F&Os from the first peer review.

APS conducted a full scope Seismic PRA model peer review (Reference 4) in February 2013, in accordance with RG 1.200, Revision 2 (Reference 7). APS conducted a full scope External Hazards screening peer review (Reference 25) in December 2011, in accordance with RG 1.200, Revision 2 (Reference 7).

An F&O closure peer review was performed in June 2017, in accordance with NRC letter dated May 3, 2017 (Reference 17) to assess the closure of all finding level F&Os from these peer reviews (Reference 18) that were not otherwise addressed by focused scope peer reviews that re-reviewed the associated ASME PRA supporting requirements in their entirety. The F&O closure review was conducted to ensure the findings had been satisfactorily resolved to meet the ASME PRA Standard RA-Sa-2009 (Reference 15) to Capability Category II, the sub-element criteria for the CEOG from Internal Events PRA peer review (Reference 19), and RG 1.200, Revision 2 (Reference 7).

The F&O closure peer review assessed the sixty finding level F&Os from the prior peer reviews and concluded that all were closed with the exception of eight findings. Of the eight not closed findings, six were assessed as partially closed, and two were assessed as open.

The eight not closed findings and their dispositions are described in Attachment 3.

APS will resolve the eight not closed finding level F&Os listed in Attachment 3 and validate closed by a subsequent F&O closure review conducted in accordance with NRC letter dated May 3, 2017 (Reference 17). These not closed finding level F&Os will be closed prior to utilizing the PRA models for categorization. Resolution of the not closed finding level F&Os is not expected to have a significant impact on overall CDF or LERF based on the dispositions described in Attachment 3.

3.4 Risk Evaluations [10 CFR 50.69(b)(2)(iv)]

The PVNGS 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04 (Reference 2). The overall risk evaluation process described in the NEI guidance addresses both known degradation mechanisms and common cause interactions, and meets the requirements of § 50.69(b)(2)(iv). Sensitivity studies described in NEI 00-04 Section 8 (Reference 2) will be used to confirm that the categorization process results in acceptably small increases to CDF and LERF. The failure rates for equipment and initiating event frequencies used in the PRA include the quantifiable impacts from known degradation mechanisms, as well as other mechanisms (e.g., design errors, manufacturing deficiencies, human errors, etc.).

Subsequent performance monitoring and PRA updates required by the rule will continue to capture this data, and provide timely insights into the need to account for any important new degradation mechanisms.

10

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components 4 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements The following NRC requirements and guidance documents are applicable to the proposed change.

• The regulations at Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors.

• NRC Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, May 2006 (Reference 1).

• Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, April 2015 (Reference 16).

• Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009 (Reference 7).

The proposed change is consistent with the applicable regulations and regulatory guidance.

4.2 No Significant Hazards Consideration APS proposes to modify the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Reactors. The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has high safety significance resulting in improved plant safety.

APS has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The process used to evaluate SSCs for changes to NRC special treatment requirements and the use of alternative requirements ensures the ability of the SSCs to perform their design function(s).

The potential change to special treatment requirements does not change the design and operation of the SSCs. As a result, the proposed change does not significantly affect any initiators to accidents previously evaluated or the ability to mitigate any 11

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components accidents previously evaluated. The consequences of the accidents previously evaluated are not affected because the mitigation functions performed by the SSCs assumed in the safety analysis are not being modified. The SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition following an accident will continue to perform their design functions.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not change the functional requirements, configuration, or method of operation of any SSC. Under the proposed change, no additional plant equipment will be installed.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change will permit the use of a risk-informed categorization process to modify the scope of SSCs subject to NRC special treatment requirements and to implement alternative treatments per the regulations. The proposed change does not affect any Safety Limits or operating parameters used to establish the safety margin. The safety margins included in the analyses of accidents are not affected by the proposed change. 10 CFR 50.69 requires that there be no significant effect on plant risk due to any change to the special treatment requirements for SSCs and that the SSCs continue to be capable of performing their design basis functions, as well as to perform any beyond design basis functions consistent with the categorization process and results.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, APS concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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.

12

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components 5 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6 REFERENCES

1. NRC Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, Revision 1, dated May 2006

2. NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Revision 0, Nuclear Energy Institute, dated July 2005

3. NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management, dated December 1991

4. Westinghouse Letter to APS CVER-13-028, Transmittal of Palo Verde Seismic PRA -

Final Peer Review Report, February 14, 2013

5. NRC Safety Evaluation (SE) Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative AN02-R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (ADAMS Accession Number ML090930246), dated April 22, 2009

6. Vogtle Electric Generating Plant, Units 1 And 2 - Issuance of Amendments Re: Use of 10 CFR 50.69 (ADAMS Accession Number ML14237A034), dated December 17, 2014

7. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, dated March 2009

8. NEI 00-02, Probabilistic Risk Assessment (PRA) Peer Review Process Guidance, Nuclear Energy Institute, dated 2000

9. License Amendment Request to Revise Technical Specifications to implement Risk-Informed Completion Times (ADAMS Accession Number ML15218A300), dated July 31, 2015

10. NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, dated September 2005

11. NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, dated March 2009

12. EPRI TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, dated December 2008 13

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components

13. Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Issuance of Amendments Re:

Changes To Technical Specification 3.8.7, "Inverters-Operating" (ADAMS Accession Number ML102670352), dated September 29, 2010

14. Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Issuance of Amendments Re:

Adoption of TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control RITSTF Initiative 5b" (ADAMS Accession Number ML112620293), dated December 15, 2011

15. ASME/ANS RA-Sa-2009, Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, dated February 2009

16. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, dated April 2015

17. NRC letter (ADAMS Accession No. ML17079A427) dated May 3, 2017

18. ABS Consulting Report R-3882824-2037, Palo Verde Generating Stations PRA Finding Level Fact and Observation Closure Review, June 23, 2017

19. ABB Combustion Engineering letter to APS ST-99-542, PSA Peer Review for Palo Verde Nuclear Generating Station, July 12, 1999

20. Palo Verde Engineering Evaluation 3579223, PRA input to the Risk-Informed Task Force (RITS) 5b license amendment, March 10, 2011

21. Westinghouse Letter to APS LTR-RAM-II-10-082, Rev. 0, Internal Flood Focused Scope RG 1.200 PRA Peer Review Against the ASME/ANS PRA Standard Requirements for the Palo Verde Nuclear Generating Station Probabilistic Risk Assessment, November 2010

22. Westinghouse Letter to APS LTR-RAM-12-13, Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS PRA Standard for the Palo Verde Nuclear Generating Station Fire Probabilistic Risk Assessment, January 2, 2013

23. Hughes Associates Report 001014-RPT-01, Palo Verde Nuclear Generating Station Fire PRA Focused-Scope Peer Review, January 22, 2015

24. NRC Regulatory Issue Summary 2007-06 Regulatory Guide 1.200 Implementation, (ADAMS Accession No. ML070650428) dated March 22, 2007

25. Palo Verde Nuclear Generating Station Other External Hazards PRA Peer Review Report, December 2011 14

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 1 List of Categorization Prerequisites

• APS will resolve the eight not closed finding level facts and observations (F&O) listed in Attachment 3 and validate them closed by a subsequent facts and observation (F&O) closure review conducted in accordance with NRC letter dated May 3, 2017 (Reference 17). The not closed finding level F&Os will be closed prior to utilizing the PRA models for categorization.

• APS will establish procedure(s) prior to the use of the categorization process on a plant system. The procedure(s) will contain the elements/steps listed below:

1. Integrated Decision-Making Panel (IDP) member qualification requirements.

2. Qualitative assessment of system functions. System functions are qualitatively categorized as preliminary high safety significant (HSS) or low safety significant (LSS) based on the seven questions in Section 9 of NEI 00-04 (see Section 3.2). Any component supporting an HSS function is categorized as preliminary HSS. Components supporting an LSS function are categorized as preliminary LSS.

3. Component safety significance assessment. Safety significance of active components is assessed through a combination of PRA and non-PRA methods, covering all hazards. Safety significance of passive components is assessed using a methodology for passive components.

4. Assessment of defense-in-depth (DID) and safety margin. Components that are categorized as preliminary LSS are evaluated for their role in providing defense-in-depth and safety margin and, if appropriate, upgraded to HSS.

5. Review by the Integrated Decision-Making Panel. The categorization results are presented to the IDP for review and approval. The IDP reviews the categorization results and makes the final determination on the safety significance of system functions and components.

6. Risk sensitivity study. For PRA-modeled components, an overall risk sensitivity study is used to confirm that the population of preliminary LSS components results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF) and meets the acceptance guidelines of RG 1.174.

7. Periodic reviews are performed to ensure continued categorization validity and acceptable performance for those SSCs that have been categorized.

8. Documentation requirements per Section 3.1.1 of this enclosure.

15

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 2 Total Unit 1/2/3 Baseline Average Annual CDF/LERF CDF LERF Hazard (per reactor-year) (per reactor-year)

Internal events 1.3E-6 4.3E-8 Internal flooding 4.6E-7 2.1E-8 Seismic 3.1E-5 5.7E-6 Internal Fire 2.9E-5 2.4E-6 Total 6.2E-5 8.2E-6 Notes

1. Total CDF meets the RG 1.174 acceptance guideline of < 1E-4 per year

2. Total LERF meets the RG 1.174 acceptance guideline of < 1E-5 per year 16

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement IE-07 (B) / The Interfacing Systems Status: Partially Closed. The Closure Review Team IE-12 Loss of Coolant Accident recommendation will be Basis: Insufficient justification is provided in Impact 200- 84.pdf addressed by evaluating all (ISLOCA) treatment for the Internal to demonstrate that the frequency of the scenario in question is ISLOCA failure modes of the shutdown cooling suction Events negligible. The resolution of this finding only provides qualitative shutdown cooling system line appears to have some argument that this ISLOCA scenario will require failure of two piping. Justification for questionable assumptions.

Motor Operated Valves (MOVs), failure to open of the LTOP valve, screening out any negligible First, it is assumed that the and failure of the warmup piping or the bypass valve. No scenarios will be provided.

Low Temperature Over quantitative values and LTOP valve capacity were provided to Leakage, spurious operation, Pressure (LTOP) valve demonstrate that the frequency of this scenario is negligible. Note and catastrophic failure modes would always open. While of valves will be considered, as that the likelihood of failure of the piping outside containment is this is the most likely well as the LTOP relief valve relatively high. Furthermore, it is not clear if the capacity of the scenario, the LTOP valve failure to open or exceedance LTOP valve is sufficient to relieve the relatively large flow that may can fail to open. Qualitative of its relief capacity.

result from the catastrophic ruptures of the two upstream MOVs.

arguments were made that Finally, ISLOCA may also result from leakage of both of the two should this happen, the These changes are not upstream MOVs, or a combination of leakage and rupture of the resulting LOCA would be expected to have a significant two upstream MOVs, in conjunction with failure of the LTOP valve inside containment impact on total CDF or LERF to open and failure of the downstream piping. This scenario would since the current internal (primarily based on relative have a greater frequency than catastrophic ruptures of both MOVs events contribution to total pipe lengths). This ignores because the frequency of MOV leakage is significantly greater than CDF and LERF is less than 2%

the fact that the high stress its catastrophic rupture. and 0.5%, respectively. While points and stress concentration points are Recommendation: Provide additional justifications (including the the resulting LERF contribution outside containment. from these scenarios may not capacity of the LTOP valve, all of the possible failure mode Furthermore, the shutdown be negligible, they are combinations and their probabilities of occurrence, LERF value, expected to be minimal based cooling warmup crossover etc.) to demonstrate that the frequency of the scenario in question on industry operating piping was not considered. is indeed negligible compared to the LERF. experience.

These changes will be implemented and the finding verified closed by a subsequent F&O Closure Review as a pre-requisite to categorization.

17

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement AS-03 (B) / There are some differences Status: Open. The Closure Review Team AS-24 between treatment of a recommendation will be Basis: Containment heat removal (CHR) is only asked in the small small LOCA associated with addressed by modeling CHR Internal LOCA event tree when the success path is relying on high pressure a pipe break and an in the small LOCA event tree, Events sump recirculation (HPSR) with a failure of the operators to induced small LOCA and event scenarios with depressurize and cool down with successful SG heat removal. In (pressurizer safety valve failure of the PSV to reseat.

this case the RCS remains at temperature so that there is reclosure) in the transient MAAP analyses will be substantial heat transfer to the containment. Table 4 of 13-NS-event trees. For example: performed to include PSV B065 R007 presents MAAP results for LOCA cases with failure of failure to reseat in the small

• In the small LOCA event CHR from spray recirculation. Based on a reply to the reviewers break sizes to determine the tree, successful high question, Table 4 indicates that <2 diameter breaks may just necessity of CHR for long-pressure injection and require CHR because s1_2_1a with SG cooling and failure of SG term stable end-state.

recirculation lead to depressurization exceeds a containment pressure of 50 psig at just questioning whether 6.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Smaller holes as represented by case s1_1_1a for a 1 These changes are not containment heat removal break do not exceed even 50 psig until 22.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The ultimate expected to have a is successful. In the containment failure pressure is 141 psig (i.e. 50% chance of significant impact on total Transient Type 2 and failure) so assuming failure at 50 psig as a basis for success is CDF or LERF since the Transient Type 3 event conservative. It is therefore also conservative to assume that all current internal events trees, RCS integrity can small LOCA sizes (3/8 to 2.35) require CHR under these contribution to total CDF and be lost if pressurizer circumstances. The small LOCA event tree may also need to ask LERF is less than 2% and safety valves do not reset CHR in cases where the SGs are not depressurized; i.e. sequences 0.5%, respectively. The after lifting. In the 1 and 3. likelihood of a small LOCA sequences from these with a loss of CHR for longer For a loss of main feedwater pumps (Type2), containment heat event trees where high than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is very small.

removal is not asked for any sequences. Either SG cooling prevents pressure injection and the PSV from lifting at all so there is no LOCA or the operators These changes will be recirculation are depressurize the RCS for alternate AFW (at low pressure) though implemented and the finding successful, the question the PSVs are assumed to lift and may fail to reseat. Failure of at verified closed by a relating to containment least one PSV to reseat (equivalent hole size of 2.3) requires HPSI subsequent F&O Closure heat removal is not but the SGs are at low temperature in this scenario so CHR is Review as a pre-requisite to asked.

currently assumed not required for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. categorization.

Consideration should be given to containment failure at later times which may lead to subsequent core damage due to failure of sump recirculation at the time of containment failure.

18

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement AS-03 (B) /

• In the small LOCA event For a loss of condenser vacuum (Type3), its similar to the type2 event AS-24 tree, RCS depressurization tree. Therefore, for LOCA scenarios with a hole size no larger than 2.3 and use of low pressure equivalent in diameter, with or without SG depressurization, further Internal injection and recirculation justification is needed to not require CHR to protect the containment.

Events are considered if high The following information is useful in reviewing the documents pressure injection or (cont.) associated with this F&O. Page 337, Figure 3.27.4 of 13-NS-B061, recirculation fail. In the Revision 5 is the small LOCA event tree. Type 2 initiators with RT Transient Type 2 and before turbine trip only challenge the PSVs if all SG cooling is lost.

Transient Type 3 event Event tree for loss of main feedwater pumps in 3.9.4 on Page 130 trees, consideration of RCS shows that containment heat removal is not asked because if depressurization and use of secondary heat removal is lost, core damage is assumed. Type 3 low pressure systems is not initiators are where turbine trips first and may challenge the PSVs.

included because the Figure 3.6.4 on page 102 shows the loss of condenser vacuum ET likelihood of high pressure which is type 3 initiator. Containment heat removal is not asked. Type injection or high pressure 2 and type 3 initiators presently have about the same contribution to recirculation are small. It CDF for internal events, as does small LOCA; i.e., around 12%.

would seem that this assumption should apply to Recommendation: Perform a set of MAAP sensitivity analyses assuming both cases, or not. a stuck open PSV with equivalent hole diameter of 2.3 to investigate the possibility of success without CHR. Expand the discussions in the MAAP report to better describe the basis for the 2 hole size as the critical break size. Further, the assumed mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> may be too short for consideration of containment failure. A loss of CHR that results in an exceedance of the pressure capacity at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is not a stable state at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and is still of concern. In other words, breaks with sizes smaller than 2 may also require CHR under the circumstances postulated in this F&O.

One possibility is to add the events asking for CHR to the event trees for small LOCA, type2, and type3 initiators to see if the change in assumed success criteria makes any difference to CDF.

The saturation temperatures corresponding to 50 psig is approximately 300°F. If it can be shown that SG cool-down limits the exiting RCS coolant temperature to less than these values prior to reaching 50 psig, then containment integrity should be assured.

19

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement 1-1 / As noted in SRs IFSO-A1, Status: Partially Closed. The Closure Review Team IFSO-B2 IFSO-A3, and IFSO-A5, some recommendation will be areas of the documentation do Basis: The documentation (Sections 4.2.5 and 4.2.6 of Study 13- NS-implemented as written.

not provide sufficient detail C094, Revision 1, and footnote in Table C.1 of 13-NS-C093 Revision 1)

Internal have been revised to address the Findings. Section 4.2.6 of Study 13- These changes are not about the process used.

Flooding Specific items for which NS-C094, Revision 1 discusses the flood sources in the TB and the expected to have a significant improved documentation is impact of these flood sources, if any, on equipment in TB that are impact on CDF or LERF since needed include: modeled in the PRA. The rationale for not including the temperature the current internal flood

a. Documentation of sources and pressure of fluid systems based on Assumption 2 of PRA Study 13- contribution to total CDF and in the Turbine Building. NS-C096 Revision 2 must be supported by the fact that there will be no LERF is less than 0.7% and

b. The basis for screening propagation of steam (due to HELB) from the location of the piping 0.2%, respectively.

sources in the Fuel, system break to the adjacent location(s) and impacting PRA equipment Radwaste, and Turbine Preliminary review indicates in the adjacent location(s). Also, for feedwater line break in the TB, it Buildings (i.e., the way in that steam propagation will must be verified that this event will not impact other PRA equipment which the specified criteria have minimal impact on PRA such as the instrument air system due to steam and humidity.

are met for each source is equipment in adjacent not documented). For Recommendation: Verify and document the fact that propagation of locations.

example, a walkdown steam (due to a HELB) will not occur from the location of piping system during the peer review These changes will be break to the adjacent location(s) and impacting PRA equipment in the revealed that there is implemented and the finding adjacent location(s) and that a feedwater line break in the TB will not section of the wet pipe fire verified closed by a subsequent impact other PRA equipment such as the instrument air system due to protection (FP) system F&O Closure Review as a pre-steam and humidity.

running above the turbine requisite to categorization.

cooling water (TC) pumps that could potentially spray both pumps. It is not clear based on 13-NS-C093 and 13-NS-C094 that this impact was considered and dispositioned. Likewise, feedline breaks in the turbine building are assumed to be bounded by the loss of main feedwater initiating event, but may have different impacts such as loss of instrument air due to humidity impacts.

c. The temperature and pressure of flood sources.

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Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement 1-2 / Potential flooding Status: Partially Closed. The Closure Review Team IFEV-A7 mechanisms are primarily recommendation will be Basis: Section 4.1 (pages 24 - 25) of study 13-NS-C097 Revision 2 limited to failures of implemented as written.

addressed the finding with the discussion of the potential for Internal components. Human-human and maintenance induced flooding events. Maintenance These changes are not Flooding induced flooding is activities/procedures were reviewed and a search of plant expected to have a screened based on plant operating experience (APS PVAR/CRDR database, plant trip history significant impact on CDF or maintenance practices (see and LERs) using flood-related keywords for flooding events was LERF since the current 13 NS-C093, Section 3.2, performed as documented in this section (4.1) of the System internal flood contribution to Item 4 and 13-NS-C097, Study. total CDF and LERF is less Section 3.5). This does not than 0.7% and 0.2%,

indicate that there was any It is stated in Section 3.5 of study 13-NS-C097 Revision 2 that respectively.

search of plant operating maintenance activities, which involve the replacement of pumps or experience and plant cleaning of heat exchangers, have the potential to induce a There have been a very maintenance procedures to significant flooding event are not performed on-line at the plant. limited number of human verify no potential for However, there was a PVNGS event that involved the plugging of induced flood events that human-induced flood the condenser tubes during plant operation. There is also a were screened out.

mechanisms. potential for on-line heat exchanger tube plugging if a heat These changes will be exchanger tube leak is detected. Such events were not considered implemented and the finding in the discussion of maintenance/human induced flooding in verified closed by a Section 3.5 of Study 13-NS-C097 Revision 2.

subsequent F&O Closure Recommendation: To be consistent with the state of the industry Review as a pre-requisite to practice, identify and evaluate human-induced floods during plant categorization.

operation for scenarios that may result from risk-significant maintenance activities (e.g., plugging of the condenser tubes and heat exchanger tubes), and include applicable maintenance-induced failure modes in the next update of the internal flooding PRA to incorporate the revised system pipe break frequency values from EPRI Report 3002000079, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessment, Revision 3.

21

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement 1-3 / RG 1.200 Revision 2 Status: Partially Closed. The Closure Review Team IFSN-A6 documents a qualified recommendation will be Basis: Assumption 2 in Section 3.1.3 of 13-NS-C096 states that acceptance of this implemented as written.

All components within a flood area where the flood originates were Internal supporting requirement assumed susceptible and failed as a result of the flood, spray, These changes are not Flooding (SR). The NRC resolution steam, jet impingement, pipe whip, humidity, condensation and expected to have a states that to meet temperature concerns except when component design (e.g., water- significant impact on CDF or Capability Category II, the proofing), spatial effects, low pressure source potential or other LERF since the current impacts of flood-induced reasonable judgment could be used for limiting the effect. This internal flood contribution to mechanisms that are not assumption is appropriate and effectively bounds the potential total CDF and LERF is less formally addressed (e.g.,

impacts from jet impingement, pipe whip, spray, submergence, than 0.7% and 0.2%,

using the mechanisms etc. However, this assumption of susceptible equipment failure in respectively.

listed under Capability the flood originating room does not always bound the impact of Category III of this These changes will be humidity, condensation, and temperature concerns because of the requirement) must be implemented and the finding potential propagation of the flooding effects (e.g., steam). From qualitatively assessed using verified closed by a this consideration, the assumption is nonconservative.

conservative assumptions. subsequent F&O Closure Recommendation: For the applicable flood scenarios in relevant Review as a pre-requisite to locations, evaluate the effects of humidity, condensation, and categorization.

temperature by considering the possible propagation from the initiating room to all connecting rooms. Also see F&O 1-1.

22

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement 1-4 / There is no evidence in 13- Status: Partially Closed. The Closure Review Team IFEV-A6 NS-C097 that a search was recommendation will be Basis: Section 4.1 (pages 24 - 25) of study 13-NS-C097 Revision 2 made for plant-specific implemented as written.

addressed the finding on lack of search for plant-specific operating Internal operating experience, plant experience, plant design features, and conditions that may impact These changes are not Flooding design features, and flood likelihood with a discussion of the search of flood type expected to have a conditions that may impact events in the PVNGS Site Work Management System database and significant impact on CDF or flood likelihood and no License LERF since the current Bayesian updating was internal flood contribution to performed. However, Event Reports. It also discusses the review of the PVNGS total CDF and LERF is less adjustments are made to maintenance procedures on flood prevention guidelines and the than 0.7% and 0.2%,

some initiating event potential of maintenance induced flooding. No Bayesian update was respectively.

frequencies based on performed on the flood initiating event frequency due to insufficient system run times to flood data at PVNGS. However there were a few events associated There have been a very account for differences with leaks/flooding that were screened based on the impact of limited number of flood between impacts when the those events on safety-related or PRA equipment. The criteria for events that were screened pumps are running or in screening of flooding events based on impact may not be out.

standby. applicable to the screening of flood events performed for the These changes will be purpose of evaluating the likelihood of flooding.

implemented and the finding Recommendation: Develop criteria for screening of flood events for verified closed by a the purpose of evaluating the likelihood of flooding or flooding subsequent F&O Closure frequency. The criteria may be in terms of potential spatial impact Review as a pre-requisite to of the flood event. Use the criteria to re-evaluate the flooding categorization.

events at PVNGS for the purpose of flood frequency update.

Unscreened flooding events should then be used for updating the applicable flooding frequency.

23

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement SHA-E2-01 / The evaluation and Status: Open. The updated seismic hazard SHA-E2 incorporation of curves provided by LCI Basis: Finding SHA-E2 is considered as Open. The technical basis for uncertainties in the site this conclusion is summarized below and is supported by the (2015b) and associated Seismic response velocity profile documentation as found in LCI Report 2211-PR-07-Rev. 4, Seismic impacts on fragilities will be may not be properly Hazard Evaluation for Palo Verde Nuclear Generating Station, dated incorporated into the seismic incorporated because of 8/27/2013 (referred to as LCI, 2013), and LCI Report PC No. PV-001- PRA.

insufficient or unreviewable PC-05, Rev. 0, Soil Hazard and GMRS/FIRS Calculation for Palo Verde These changes are not site-specific data and/or its Nuclear Generating Station, dated 2/27/2015 (referred to as LCI, expected to have a documentation. Also, the 2015b). LCI (2013) provides a set of soil hazard curves fractiles and significant impact on CDF or site response evaluation for peak ground acceleration (pga) a set of 96 aggregate pga hazard curves. It is assumed that the aggregate pga hazard curves were LERF based on the NRC was completed using a derived for use as part of the seismic risk quantification. LCI (2013) review of the updated Senior Seismic Hazard provides an adequate summary of how the aggregate pga hazard seismic hazard curves LCI Analysis Committee curves were developed and compared to the pga hazard curve fractiles. (2015b) which found the (SSHAC) Level 1 (L1)

LCI (2015b) provides an updated set of soil hazard curves in part updated seismic hazard is process which does not based on the updated site response evaluation completed at the Palo bounded by the current meet the ASME general Verde site. LCI (2015b) provides sufficient technical material to design basis safe shutdown Capability Category II understand how the site response results were combined with the soil earthquake at most guidelines.

hazard curves for rock site conditions to produce an updated set of soil frequencies above 1 Hertz hazard curves for the Palo Verde site. Finding SHA-E2 is related to the and minor exceedances assessment of aleatory and epistemic uncertainties in site response, above 1 Hertz to be and their subsequent impact on the derivation of soil hazard curves. considered de minimis The disposition to this finding states, A SSHAC L3 analysis was (ADAMS Accession No.

performed subsequent to the seismic PRA development as part of the ML16221A604).

NTTF response to the NRC 50.54f letter on Fukushima. The SSHAC L3 analysis produced a site hazard curve which is bounded by the SSHAC These changes will be L1 hazard curve developed and used in the Seismic PRA model. implemented and the finding Therefore, the issue is resolved by the updated SSHAC L3 hazard verified closed by a analysis. subsequent F&O Closure While the updated probabilistic seismic hazard analysis completed for Review as a pre-requisite to the Palo Verde site included SSHAC L3 seismic source characterization categorization.

and ground motion characterization models, the site response analysis was not completed using an explicit SSHAC process. It is important to note that there is no mandate to complete the site response analysis following the SSHAC guidance; the approach and method documented in LCI (2015a and 2015b) is consistent with expectation documented as part of EPRI guidance (SPID) which has been endorsed by the NRC.

24

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement SHA-E2-01 / However, the seismic hazard curves provided by LCI (2015b) were SHA-E2 not explicitly compared to those derived in LCI (2013) to corroborate the disposition provided. LCI (2015b) does not include Seismic updated hazard curve fractiles for pga that can be compared to either the pga fractiles or the aggregate pga hazard curves as (cont.)

found in LCI (2013). Comparison of the mean pga hazard curves

[Table 7.1 from LCI (2013) to Table 2 from LCI (2015b)] indicates that the updated pga hazard has increased, thus it is not clear if the seismic risk quantification using the pga hazard curves from LCI (2013) is appropriate. For example at a pga = 0.5g the mean hazard has increased from 1.72E-6 [Table 7.1 from LCI (2013)] to 6.53E-6 Table 2 from LCI (2015b)]. In summary, insufficient information has been provided to demonstrate that the updated pga soil seismic hazard curves is bounded by the pga soil hazard curve used in the Seismic PRA model.

Recommendation: Demonstrate that the updated set of soil pga hazard curves fractiles (mean, and 5th, 16th, 50th, 84th, 95th) is bounded by the soil pga hazard curves used in the Seismic PGA model. If the updated set of soil pga hazard curves is greater than those used in the Seismic PRA model, the impact on Seismic risk quantification should be assessed.

25

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement SFR-F3-01 / The draft report LTR-RAM-II- Status: Partially Closed. The Closure Review Team SFR-F3 12-074 indicates that the recommendation will be Basis: In response to F&O SFR-F3-01, two documents were provided: 1) draft relay assessment uses implemented as written.

Westinghouse Letter LTR-RAM-II-12-074 R002 - PV SPRA - Relay Seismic the IPEEE relay assessment Assessment.pdf, and 2) 11c4043-cal-028 rev0 Seis Frag for Sel Relays.pdf. These changes are not as the starting point but These two documents were reviewed to support the conclusion that the expected to have any impact accounts for the updated previously unaddressed relays have been addressed and are included in the on CDF or LERF since the seismic hazard curve at the SPRA model and quantification. The following paragraphs provide the basis recommendations are site. However, the report for the conclusion that the reported fragilities are reasonable, and provide associated with documentation includes the following recommendations for completeness of documentation.

changes to better explain statement in Section 2.3 Westinghouse Document LTR-RAM-II-12-074 R002 - PV SPRA incorporates modeling rationale.

(Unaddressed Relays):

into the existing relay database 288 additional relays including the 69 previously unaddressed in the IPEEE. This document develops the first These changes will be This list (unaddressed relays) round relay fragilities based on the relay analysis performed as part of the implemented and the finding included 69 such relays. Of plant IPEEE. Several relays, including some of the 69 relays that were not verified closed by a subsequent the relays that have been addressed in IPEEE, are screened because they are not chatter sensitive. F&O Closure Review as a pre-included in the SPRA, their The relays for which chatter is not acceptable are assigned initial HCLPF requisite to categorization.

seismic fragility events are values based on the IPEEE review level earthquake (RLE) PGA of 0.5 scaled found in many of the in accordance with the location specific spectral acceleration ratios of the dominant CDF cutsets.

IPEEE ISRS and the SPRA ISRS. These initial HCLPF values are substantiated with walkdowns of host components. The reported initial quantification using these HCLPF and generic betas identified the dominant relay contributors. Westinghouse Document LTR-RAM-II-12-074 R002 - PV SPRA describes the modelling of the relays in the SPRA.

S&A Calculation 11c4043-cal-028 documents the detailed fragility analysis using separation of variable for 13 relays selected from the top contributors. The median capacity is based on the in-cabinet seismic demand on the relays associated with the SPRA ISRS, and EPRI relay GERS. The associated uncertainty parameters are obtained by using the separation of variables method. The resulting fragility parameters are scaled to conform to the revised seismic hazard and the revised GMRS. We concur with the analysis and feel that the resulting fragilities are sufficiently realistic.

Finding SFR-F3-01 could be resolved on the basis that APS will take actions to implement the recommendations provided below and update WEC Document LTRRAM-II-12-074 Rev. 2, and S&A Calculation 11c4043-cal-028 Rev. 0.

26

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 3 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items from Facts and Observation Closure Review Process Finding Description Closure Review Team Recommendation(s) Disposition F&O ID /

Requirement SFR-F3-01 / Recommendation: The detailed fragility evaluation for dominant relay SFR-F3 contributors should demonstrate that the seismic demand is an appropriate median based response, and that the important Seismic uncertainties are included in obtaining log standard deviations.

(cont.) We recommend that the following be incorporated into the Westinghouse Document LTR-RAM-II-12-074 R002 for completeness:

1. During the closure review, the following information was communicated to the reviewers regarding the unaddressed relays via an e-mail: The initial fragilities for relays were estimated based on the IPEEE analysis for 0.5g RLE. The 69 relays that were not addressed in Rev. 1 of the Relay Fragility Assessment were subsequently incorporated into the Relay Fragility Assessment Rev. 2. Twenty-nine of the 69 relays were screened out from modeling. However, 40 relays were modeled. They were assigned simplified fragilities based on walkdowns that were already performed on the parent cabinets. The above information in the e-mail communication should be documented in WEC Document LTR-RAM-II-12-074 Rev. 2.

2. In Table 34: Detailed Fragility Candidates, document the source of the factor SFbldg.

We also recommend that the following be incorporated into the S&A Calculation 11c4043-cal-028 Rev. 0 for completeness of documentation:

1. Justify the use of Best Estimate ISRS as the median. In our opinion, the SSI analysis using BE soil properties, best estimate structure stiffness and a conservative estimate of best estimate structure damping results in a 84th percentile response.

2. The u associated with SSI is obtained using the BE, UB and LB envelop as the 84th and the BE alone as the median. Please explain the rationale that this results for the same building (Control Building) a wide range of SSI c from 0.09 to 0.22.

3. Explain why the uncertainties associated with structure stiffness and damping, time history simulation and earthquake component combination are ignored in the SOV calculations.

27

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 4 External Hazards Screening Screening Result Screening External Hazard Screened?

Criterion Comment (Y/N)

(Note 1)

Airport hazard meets 1975 Standard Review Plan (SRP)

PS2 Aircraft Impact Y requirements. Additionally, airways PS4 hazard bounding analysis per NUREG-1855 is < 1E-6/y.

Not applicable to the site because of Avalanche Y C3 climate and topography.

Sudden influxes not applicable to the plant design [closed loop systems for Essential Cooling Water System (ECWS) and Component Biological Event Y C3, C5 Cooling Water System (CWS)].

Slowly developing growth can be detected and mitigated by surveillance.

Not applicable to the site because of Coastal Erosion Y C3 location.

Plant design eliminates drought as a Drought Y C5 concern and event is slowly developing.

Plant design meets 1975 SRP External Flooding Y PS2 requirements.

The plant design basis tornado has a frequency < 1E-7/y. The spray Extreme Wind or PS2 Y pond nozzles (not protected against Tornado PS4 missiles) have a bounding median risk < 1E-7/y.

Limited occurrence because of arid Fog Y C1 climate and negligible impact on the plant.

Not applicable to the site because of Forest or Range Fire Y C3 limited vegetation.

Limited occurrence because of arid Frost Y C1 climate.

Limited occurrence and bounded by C1 other events for which the plant is Hail Y C4 designed. Flooding impacts covered under Intense Precipitation.

28

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 4 External Hazards Screening Screening Result Screening External Hazard Screened?

Criterion Comment (Y/N)

(Note 1)

Plant is designed for this hazard.

High Summer Y C1 Associated plant trips have not Temperature occurred and are not expected.

High Tide, Lake Level, Not applicable to the site because of Y C3 or River Stage location.

Covered under Extreme Wind or Hurricane Y C4 Tornado and Intense Precipitation.

Ice blockage causing flooding is not applicable to the site because of location (no nearby rivers and C3 Ice Cover Y climate conditions). Plant is C1 designed for freezing temperatures, which are infrequent and short in duration.

Explosive hazard impacts and Industrial or Military control room habitability impacts Y PS2 Facility Accident meet the 1975 SRP requirements (RGs 1.91 and 1.78).

PRAs addressing internal flooding have indicated this hazard typically results in CDFs 1E-6/y. Also, the Internal Flooding N None ASME/ANS PRA Standard requires a detailed PRA for this hazard which is addressed in the PVNGS Internal Flooding PRA.

PRAs addressing Internal Fire have indicated this hazard typically results in CDFs 1E-6/y. Also, the Internal Fire N None ASME/ANS PRA Standard requires a detailed PRA for this hazard which is addressed in the PVNGS Internal Fire PRA.

Not applicable to the site because of Landslide Y C3 topography.

Lightning strikes causing loss of offsite power or turbine trip are contributors to the initiating event frequencies for these events.

Lightning Y C1 However, other causes are also included. The impacts are no greater than already modeled in the internal events PRA.

29

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 4 External Hazards Screening Screening Result Screening External Hazard Screened?

Criterion Comment (Y/N)

(Note 1)

Low Lake Level or Not applicable to the site because of Y C3 River Stage location.

Extended freezing temperatures are Low Winter C1 rare, the plant is designed for such Y

Temperature C5 events, and their impacts are slow to develop.

The frequency of meteorites greater Meteorite or Satellite than 100 lb striking the plant is Y PS4 Impact around 1E-8/y and corresponding satellite impacts is around 2E-9/y.

Pipelines are not close enough to Pipeline Accident Y C3 significantly impact plant structures.

Release of Chemicals Plant storage of chemicals meets Y PS2 in Onsite Storage 1975 SRP requirements.

Not applicable to the site because of River Diversion Y C3 location.

The plant is designed for such C1 events. Also, a procedure instructs Sand or Dust Storm Y C5 operators to replace filters before they become inoperable.

Not applicable to the site because of C3 Seiche Y location. Onsite reservoirs and C1 spray ponds designed for seiches.

PRAs addressing seismic activity have indicated this hazard typically results in CDFs 1E-6/y. Also, the ASME/ANS PRA Standard requires a Seismic Activity N None detailed PRA or Seismic Margins Assessment (SMA) for this hazard which is addressed in the PVNGS Seismic PRA.

The event damage potential is less than other events for which the C1 Snow Y plant is designed. Potential flooding C4 impacts covered under external flooding.

The potential for this hazard is low at the site, the plant design Soil Shrink-Swell C1 Y considers this hazard, and the Consolidation C5 hazard is slowly developing and can be mitigated.

30

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 4 External Hazards Screening Screening Result Screening External Hazard Screened?

Criterion Comment (Y/N)

(Note 1)

Not applicable to the site because of Storm Surge Y C3 location.

Toxic gas covered under release of chemicals in onsite storage, Toxic Gas Y C4 industrial or military facility accident, and transportation accident.

Potential accidents meet the 1975 SRP requirements. Bounding analyses used for offsite rail PS2 shipment of chlorine gas and onsite Transportation PS4 truck shipment of ammonium Y

Accident C3 hydroxide. Marine accident not C4 applicable to the site because of location. Aviation and pipeline accidents covered under those specific categories.

Not applicable to the site because of Tsunami Y C3 location.

Turbine-Generated Potential accidents meet the 1975 Y PS2 Missiles SRP requirements.

Not applicable to the site because of Volcanic Activity Y C3 location.

Waves associated with adjacent large bodies of water are not C3 Waves Y applicable to the site. Waves C4 associated with external flooding are covered under that hazard.

Note 1 - See Attachment 5 for descriptions of the screening criteria.

31

Enclosure Description and Assessment of Proposed License Amendment for Categorization and Treatment of Structures, Systems, and Components Attachment 5 Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source Comments NUREG/CR-2300 C1. Event damage Initial Preliminary and ASME/ANS potential is < events for Screening Standard RA-Sa-which plant is designed.

2009 C2. Event has lower mean NUREG/CR-2300 frequency and no worse and ASME/ANS consequences than other Standard RA-Sa-events analyzed. 2009 NUREG/CR-2300 C3. Event cannot occur and ASME/ANS close enough to the plant Standard RA-Sa-to affect it.

2009 NUREG/CR-2300 Not used to C4. Event is included in and ASME/ANS screen. Used only the definition of another Standard RA-Sa- to include within event.

2009 another event.

C5. Event develops slowly, allowing adequate time to ASME/ANS eliminate or mitigate the Standard threat.

PS1. Design basis hazard ASME/ANS Progressive Screening cannot cause a core Standard RA-Sa-damage accident. 2009 PS2. Design basis for the NUREG-1407 and event meets the criteria in ASME/ANS the NRC 1975 Standard Standard RA-Sa-Review Plan (SRP). 2009 PS3. Design basis event NUREG-1407 as mean frequency is < 1E- modified in 5/y and the mean ASME/ANS conditional core damage Standard RA-Sa-probability is < 0.1. 2009 NUREG-1407 and PS4. Bounding mean CDF ASME/ANS is < 1E-6/y. Standard RA-Sa-2009 Screening not successful. NUREG-1407 and PRA needs to meet ASME/ANS Detailed PRA requirements in the Standard RA-Sa-ASME/ANS PRA Standard. 2009 32