NRC 2017-0043, License Amendment Request 287, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, System, and Components (Sscs) for Nuclear Power Plants

License Amendment Request 287, Application to Adopt 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, System, and Components (Sscs) for Nuclear Power Plants

ML17243A201
Person / Time
Site:	Point Beach
Issue date:	08/31/2017
From:	Coffey R Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2017-0043
Download: ML17243A201 (55)

Text

August 31, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 NRC 2017-0043 10 CFR 50.90 License Amendment Request 287, Application to adopt 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, System, and Components (SSCs) for Nuclear Power Plants" In accordance with the provisions of 10 CFR 50.69 and 10 CFR 50.90, NextEra Energy Point Beach, LLC (NextEra) is requesting an amendment to the licenses of Point Beach Nuclear Plant (PBNP), Units 1 and 2.

The proposed amendment would modify the licensing basis, by the addition of a License Condition, to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants." 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 safety significance resulting in improved plant safety.

The enclosure to this letter provides the basis for the proposed change to the PBNP, Unit 1 and 2 Operating Licenses. The categorization process being implemented through this change is consistent with NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline," Revision 0, dated July 2005, which was endorsed by the 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 system will only occur after these prerequisites are met.

NextEra Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk Page 2 The NRC has previously reviewed the technical adequacy of the PBNP, Units 1 and 2 Probabilistic Risk Assessment (PRA) model identified in the application for transition of the Point Beach fire protection program to a risk-informed, performance-based program based on National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, in accordance with 10 CFR 50.48(c) in License Amendment Request 271. The PRA model technical adequacy was reviewed by the NRC, June 26, 2013(ML13182A353 and ML13182A350). The NRC has also previously reviewed the technical adequacy of the Point Beach Nuclear Plant PRA models in "Application for Technical Specification Change Regarding Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program,"

License Amendment Request 273, July 3, 2014(ML14190A267). NextEra requests that the NRC utilize the review of the PRA technical adequacy for these applications when performing the review for this application.

NextEra requests approval of the proposed license amendment by September 1, 2018. The amendment will be implemented within 90 days of approval by the NRC.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Wisconsin Official.

This letter contains a License Condition as described in the Enclosure.

If you should have any questions regarding this submittal, please contact Eric Schultz, Licensing Manager, at 920-755-7854.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 31, 2017.

Sincerely, Robert Coffey Site Vice President NextEra Energy Point Beach, LLC

Enclosure:

Evaluation of the Proposed Change cc:

Regional Administrator, USNRC, Region Ill Project Manager, USNRC, Point Beach Nuclear Plant Resident Inspector, USNRC, Point Beach Nuclear Plant Public Service Commission of Wisconsin

Enclosure Evaluation of the Proposed Change TABLE OF CONTENTS 1

SUMMARY

DESCRIPTION................................................................................... 3 2

DETAILED DESCRIPTION................................................................................... 3 2.1 CURRENT REGULATORY REQUIREMENTS..................................................... 3 2.2 REASON FOR PROPOSED CHANGE...............................................................4

2.3 DESCRIPTION

OF THE PROPOSED CHANGE................................................. 5 3

TECHNICAL EVALUATION................................................................................... 6 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i)).............. 7 3.1.1 Overall Categorization Process......................................................... 7 3.1.2 Passive Categorization Process......................................................... 9 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii)).................... 10 3.2.1 Internal Events and Internal Flooding............................................. 10 3.2.2 Fire Hazards................................................................................. 11 3.2.3 Seismic Hazards............................................................................ 11 3.2.4 Other External Hazards................................................................. 11 3.2.5 Low Power & Shutdown................................................................ 12 3.2.6 PRA Maintenance and Updates...................................................... 12 3.2.7 PRA Uncertainty Evaluations.......................................................... 12 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii)).......................... 13 3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv))............................................ 14 4

REGULATORY EVALUATION.............................................................................. 15 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA............................... 15 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS............................ 15

4.3 CONCLUSION

S.......................................................................................... 17 5

ENVIRONMENTAL CONSIDERATION.................................................................. 18 6

REFERENCES................................................................................................... 19 Page 1of53

LIST OF ATTACHMENTS : List of Categorization Prerequisites.................................................... 21 : Description of PRA Models Used in Categorization.............................. 23 : Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items.................................................................. 24 : External Hazards Screening..............................................................40 : Progressive Screening Approach for Addressing External Hazards.......47 : Disposition of Key Assumptions/Sources of Uncertainty...................... 49 Page 2 of 53

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 (SSCs) for Nuclear Power Plants." 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 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 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. Treatment includes, but is 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 Page 3 of 53

"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 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 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 NEI 00-04, "10 CFR 50.69 SSC Categorization Guideline" (Reference 1), 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 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 Page 4 of 53

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 reasonable, though reduced, level of confidence that these SSCs will satisfy functional requirements.

Implementation of 10 CFR 50.69 will allow NextEra Energy Point Beach, LLC (NextEra) to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE NextEra proposes the addition of the following condition to the renewed operating licenses of Point Beach Nuclear Plant (PBNP), Units 1 and 2 to document the NRC's approval of the use 10 CFR 50.69.

NextEra 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].

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).

NextEra shall complete item 1, Modifications Associated with the NFPA 805 Fire PRA, listed in Attachment 1, List of Categorization Prerequisites, of NextEra letter to the NRC, "Point Beach Nuclear Power Plant License Amendment Request 271 Associated with NFPA 805, Updated Attachment Sand Clean License Pages" dated May 3, 2016, prior to implementation.

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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.

(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.

(iii) Results of the PRA review process conducted to meet 10 CFR 50.69(c)(1)(i).

(iv) A description of, and basis for acceptability of, the evaluations to be conducted to satisfy 10 CFR 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).

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

The NRC has previously reviewed the technical adequacy of the PBNP PRA model identified in this application for the application for transition of the Point Beach fire protection program to a risk-informed, performance-based program based on National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, in accordance with 10 CFR 50.48(c) in License Amendment Request 271. The PRA model technical adequacy was reviewed by the NRC, June 26, 2013 (ML13182A353 and ML13182A350). The NRC has also previously reviewed the technical adequacy of the PBNP PRA models in "Application for Technical Specification Change Regarding Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program/' License Amendment Request 273, July 3, 2014 (ML14190A267). NextEra requests that the NRC utilize the review of the PRA technical adequacy for these applications when performing the review for this application.

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3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR S0.69(b)(2)(i))

3.1.1 Overall Categorization Process NextEra will implement the risk categorization process in accordance with the NEI 00-04, Revision 0, as endorsed by RG 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance," (Reference 2). 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 clarifications to be applied 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 a minimum of three years of experience in the modeling and updating of the plant-specific PRA.

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 safety significant or low safety-significant pursuant to 10 CFR 50.69(f)(1) will be documented in NextEra 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 safety-significant.

Passive characterization will be performed using the processes described in Section 3.1.2.

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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.

NextEra will require that if any SSC is identified as high safety significant (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.

Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The Integrated Decision-making Panel (IDP) must intervene to assign any of these HSS Function components to LSS.

With regard to the criteria that considers whether the active function is called out or relied upon in the plant Emergency/Abnormal Operating Procedures, NextEra 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:

Internal Event Risks: Internal events and internal flooding PRA models accepted by NRC for NFPA 805 (References 11 and 15) and Risk-informed TSTF-425 Initiative Sb (Reference 12).

Fire Risks: Fire PRA model used in LAR submittal to NRC for NFPA 805 (References 11 and 15).

Seismic Risks: Safe Shutdown Equipment List (SSEL) used in the IPEEE seismic analysis (References 13 and 21).

Other External Risks (e.g., tornados, external floods, etc.): Screening process in IPEEE submitted to NRC (References 13 and 21). Attachment 4, External Hazards Screening, summarizes an update of the disposition of each hazard. This screening was updated using criteria from ASME/ANS RA-Sa-2009, Part 6 (Reference 10).

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, "Guidance for Industry Actions to Assess Shutdown Management" (Reference 3), which provides guidance for assessing and enhancing safety during shutdown operations.

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A change to the categorization process that is outside the bounds specified above (e.g., change from a seismic margins approach to a seismic PRA 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

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 Report (SER) by the Office of Nuclear Reactor Regulation "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, Third and Fourth 10-Year In-service Inspection Intervals', dated April 22, 2009 (ML090930246) (Reference 5).

The RI-RRA methodology is a risk-informed safety classification and treatment program 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 requirements of 10 CFR 50.69 is consistent with AN0-2 RI-RRA License Amendment as the rule does not remove the repair and replacement provisions of the ASME Code required by 10 CFR 50.55a(g) for ASME Class 1 SSCs, even if they are categorized as RISC-3, since those SSCs constitute principal fission product barriers as Page 9 of 53

part of the reactor coolant system or containment. This is further clarified in the rule's Statement of Considerations. However, since the scope of 10 CFR 50.69 addresses additional requirements, this methodology will be applied to determine the safety significance of ASME Class 1 SSCs, some of which may be evaluated to be RISC-3. The ASME classification of the SSC does not impact the methodology as it is only evaluates the consequence of a rupture of the SSC's pressure boundary. As stated in the ANO SER, "categorizing solely based on consequence which measures the safety significance of the pipe given that it ruptures is conservative compared to including the rupture frequency in the categorization and the categorization will not be affected by changes in frequency arising from changes to the treatment." Therefore, this methodology is appropriate to apply to ASME Class 1 SSCs, as the consequence evaluation and deterministic considerations are independent of the ASME classification when determining the SSCs safety significance and will maintain this acceptable level of conservatism.

The use of this method was previously approved to be used for a 10 CFR 50.69 application by NRC in the final Safety Evaluation for Vogtle dated December 17, 2014 (Reference 14). The RI-RRA method as approved for use at Vogtle 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 PBNP 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. All the PRA models described below have been peer reviewed and there are no PRA upgrades that have not been peer reviewed.

The PRA models credited in this request are the PRA models used in the NFPA 805 application (References 11 and 15) and the Surveillance Frequency Control Program application (Reference 12), with routine maintenance updates applied.

3.2.1 Internal Events and Internal Flooding The PBNP categorization process for the internal events and flooding hazard will use the plant-specific PRA model. The NextEra risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the PBNP units. Attachment 2 at the end of this enclosure identifies the applicable internal events and internal flooding PRA models.

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3.2.2 Fire Hazards The PBNP, Units 1 and 2 categorization process for fire hazards will use a peer reviewed plant-specific fire PRA model. The internal Fire PRA model was developed consistent with NUREG/CR-6850 and only utilizes methods previously accepted by the NRC. The NextEra risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the PBNP units. at the end of this enclosure identifies the applicable Fire PRA model.

3.2.3 Seismic Hazards The PBNP, Units 1 and 2 categorization process will use the seismic margins analysis (SMA) performed for the Individual Plant Evaluation-External Events (IPEEE) in response to GL 88-20 (Reference 6) for evaluation of safety significance related to seismic hazards. No plant specific approaches were utilized in development of the SMA. The NEI 00-04 approved use of the SMA SSEL as a screening process identifies all system functions and associated SSCs that are involved in the seismic margin success path as HSS. Since the analysis is being used as a screening tool, importance measures are not used to determine safety significance. The NEI 00-04 approach using the SSEL would identify credited equipment as HSS regardless of their capacity, frequency of challenge or level of functional diversity.

An evaluation will be performed of the as-built, as-operated plant against the SMA SSEL. The evaluation will compare the as-built, as-operated plant to the plant configuration originally assessed by the SMA. Differences will be reviewed to identify any potential impacts to the equipment credited on the SSEL. Appropriate changes to the credited equipment will be identified and documented. This documentation will be available for audit. The NextEra risk management program will ensure that future changes to the plant will be evaluated to determine their impact on the SMA and risk categorization process.

3.2.4 Other External Hazards The PBNP, Units 1 and 2 categorization process will use screening results from the Individual Plant Evaluation-External Events (IPEEE) in response to GL 88-20 (Reference 6) for evaluation of safety significance. The High Winds hazard was originally screened from applicability in the IPEEE. A High Winds PRA was subsequently developed and is being revised to more realistically reflect the as-built, as-operated plant.

All other external hazards were screened from applicability to PBNP, Units 1 and 2 per a plant-specific evaluation in accordance with GL 88-20 and updated to use the criteria in ASME PRA Standard RA-Sa-2009. Attachment 4 provides a summary of the other external hazards screening results. Attachment 5 provides a summary of the progressive screening approach for external hazards.

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3.2.5 Low Power & Shutdown The PBNP, Units 1 and 2 categorization process will use the shutdown safety management plan described in NUMARC 91-06, for evaluation of safety significance related to low power and shutdown conditions.

3.2.6 PRA Maintenance and Updates The NextEra risk management process ensures that the applicable PRA model(s) used in this application continues to reflect the as-built and as-operated plant for each of the PBNP units. The process delineates the responsibilities and guidelines for updating the PRA models, and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (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, NextEra will implement a process that addresses the requirements in NEI 00-04, Section 11, "Program Documentation and Change Control." 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 model(s) 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 process are addressed using the processes discussed in Section 8 and in the prescribed sensitivity studies discussed in Section 5 of NEI 00-04.

In the overall risk sensitivity studies, NextEra will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference 14. Consistent with the NEI 00-04 guidance, NextEra 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 Page 12 of 53

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 PBNP PRA models using the guidance of NUREG-1855 (Reference 8) and EPRI TR-1016737 Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessment (Reference 9).

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. In addition, the evaluation of uncertainties associated with Fire, Seismic, LPSD, and Level 2 is provided in EPRI TR1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty (Reference 4).

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 PBNP 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 calculations were considered key for this application.

Key PBNP PRA model specific assumptions and sources of uncertainty for this application are identified and dispositioned in Attachment 6. The conclusion of this review is that no additional sensitivity analyses are required to address PBNP PRA model specific assumptions or sources of uncertainty.

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. Specifically, the models were subject to the following industry peer reviews:

November 2010 full-scope peer review of the Point Beach Internal Events PRA model, including Internal Flooding.

June 2011 full-scope peer review of the Point Beach Fire PRA model.

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August 2011 focused-scope peer review of the Point Beach Internal Flooding PRA model.

October 2011 focused-scope peer review of the Point Beach Internal Events PRA model.

May 2013 focused-scope peer review (FSS) of the Point Beach Fire PRA model.

June 2013 focused-scope peer review (FQ) of the Point Beach Fire PRA model.

Closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, "Close-out of Facts and Observations" (F&Os)

(Reference 19) as accepted by NRC in the staff memorandum dated May 3, 2017 (ML17079A427) (Reference 20). The results of this review have been documented and are available for NRC audit. provides a summary of the remaining findings and open items, including:

Open findings and disposition of the PBNP peer reviews.

Identification of and basis for any sensitivity analysis needed to address open findings.

This information demonstrates that the PRA is of sufficient quality and level of detail to support the categorization process, and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC as required 10 CFR 50.69(c)(l)(i).

3.4 RISK EVALUATIONS {10 CFR 50.69{b){2){iv))

The PBNP 10 CFR 50.69 categorization process will implement the guidance in NEI 00-04. The overall risk evaluation process described in the NEI guidance addresses both known degradation mechanisms and common cause interactions, and meets the requirements of 10 CFR 50.69(b )(2)(iv). Sensitivity studies described in NEI 00-04 Section 8 will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (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.

Page 14 of 53

4 REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA 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.

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.

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

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

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS NextEra proposes to modify the licensing basis to allow for the voluntary implementation of the provisions of 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components (SSCs) for Nuclear Power Plants."

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 safety significance resulting in improved plant safety.

Page 15 of 53

NextEra 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. 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 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.

Page 16 of 53

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 analyses of accidents are not affected by the proposed change. The regulation 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, NextEra 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 CONCLUSION

S 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 Commission's 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.

Page 17 of 53

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.

Page 18 of 53

6 REFERENCES

1.

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

2.

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.

3.

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

4.

EPRI TR1026511, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty."

5.

ANO SER 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 (TAC NO. MD5250) (ML090930246), April 22, 2009.

6.

Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4," USNRC, June 1991.

7.

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

8.

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

9.

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

10. ASME/ANS RA-Sa-2009, Standard for Level I/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, February 2009.

11. Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (CAC NOS. MF2372 and MF2373)

(ML16196A093), September 8, 2016.

12. Point Beach Nuclear Plant Units 1 and 2 - Issuance of Amendments Regarding Relocation of Surveillance Frequencies to Licensee Control (TAC NOS MF4379 and MF4380) (ML15195A201), July 28, 2015.

Page 19 of 53

13. Dockets 50-266 and 50-301, Generic Letter 88-20, Supplement 4 (TAC NOS. 74452 and 74453) Summary Report on Individual Plant Examination of External Events for Severe Accident Vulnerabilities Point Beach Nuclear Plant, Units 1 and 2, June 30, 1995.

14. Vogtle Electric Generating Plant, Units 1 and 2 -Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC NOS. ME9472 AND ME9473), December 17, 2014 (ML14237A034)

15. NRC 2016-0013, Point Beach Nuclear Plant, Units 1 and 2 - Supplement 4 to License Amendment Request 271 Associated with NFPA 805, April 7, 2016.

16. NextEra Energy Point Beach, LLC, Response to NRC 10 CFR 50.54(f) Request for Information Regarding Near-Term Task Force Recommendation 2.1, Flooding-Submittal of Flooding Hazards Revaluation Report, March 12, 2015 (ML15071A413)

17. NextEra Energy Point Beach, LLC, Focused Evaluation for Local Intense Precipitation, June 22, 2017 (ML17173A082)

18. NextEra Energy Point Beach, LLC, Mitigating Strategies Assessment (MSA) Report Submittal, November 22, 2016 (ML16327A099)

19. NEI Letter to USNRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," February 21, 2017, (ML17086A431).

20. USNRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017, (ML17079A427).

21. Point Beach Nuclear Plant Units 1 and 2 - Review of Individual Plant Examination of External Events (IPEEE) Submittal (TAC NOS M83661 and M83662)

(ML112030452), September 15, 1999.

Page 20 of 53

List of Categorization Prerequisites A. The PRA model to be used for categorization credits the following modifications to achieve an overall CDF and LERF consistent with NRC Regulatory Guide 1.174 risk limits. Use of the categorization process on a plant system will only occur after the modifications are completed.

1) Modifications associated with the NFPA 805 Fire PRA are documented in Attachment S of Reference 15. These modifications, credited in the Point Beach NFPA 805 PRA model, are currently scheduled to be completed during the fall 2017 Unit 1 refueling outage, prior to applying the categorization process.

B. NextEra 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.

Integrated Decision Making Panel (IDP) member qualification requirements.

Qualitative assessment of system functions: System functions are qualitatively categorized as preliminary HSS or 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.

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.

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.

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.

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.

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

Documentation requirements per Section 3.1.1.

Page 21of53

C. NextEra will perform an evaluation of the as-built, as-operated plant against the SMA SSEL. The evaluation will compare the as-built, as-operated plant to the plant configuration originally assessed by the SMA. Differences will be reviewed to identify any potential impacts to the equipment credited on the SSEL.

D. Prior to implementation, the noted findings in Attachment 3 will either be closed or a sensitivity study case will be performed to determine the impact on the CDF and LERF results for those categorizations that could be adversely affected by this finding.

Page 22 of 53

Description of PRA Models Used in Categorization Unit Model Baseline CDF Baseline LERF Comments 1

Internal Events 5.1E-06 3.7E-08 1

Internal Flood 3E-07 2E-08 Will be used for categorization3 1

Internal Fire 5.9E-05 9.0E-07 1

High Wind 1.74E-06 5.73E-08 1

Seismic 6.24E-06 1.21E-06 Not used for categorization 1

Other External

<1E-06

<1E-07 Hazards 1

Total 7.3E-051 2.3E-062 Unit Model Baseline CDF Baseline LERF Comments 2

Internal Events 5.1E-06 3.6E-08 2

Internal Flood 3E-07 2E-08 Will be used for categorization3 2

Internal Fire 6.9E-05 1.1E-06 2

High Wind 1.16E-06 5.25E-08 2

Seismic 6.24E-06 1.21E-06 Not used for categorization 2

Other External

<1E-06

<1E-07 Hazards 2

Total 8.3E-051 2.SE-062 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.

3. Models to be used for categorization were previously submitted in Reference 15.

Page 23 of 53

Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications The loss of a single 4.16 kV AC bus does not result in a unit trip; therefore, this is not an initiating event.

Loss of HVAC was evaluated in PRA S.2S notebook. The evaluation for some critical areas was revised, and for some PRA lacked systematic approach and areas fault tree models were IE-Al documentation for treatment of special developed to evaluate the IE-AS Not Met initiating events. Examples given impact of the loss of HVAC.

IE-Al-01 IE-82 included loss of a 4kV bus, loss of HVAC.

These calculations provide a IE-D2 Discussion in the PRA documentation quantitative basis that these needs more explanation for why not all HVAC systems do not contribute special initiators were included.

and need not be modeled.

Prior to implementation, either this finding will be closed or a sensitivity study case will be performed to determine the impact on the CDF and LERF results for those categorizations that could be adversely affected by this findinq.

AS-86 Not Met Electrical limitations, e.g., load Some of the electrical load AS-86-01 SY-AS management failures, may need to be limitations that existed at the SY-A21 considered in PRA model.

time this findino was written are Page 24 of S3

Finding Supporting Capability Number Requirement( s)

Category Description Disposition for 50.69 And for (CC)

Other Applications SY-86 SY-815 no longer in place. The remaining electrical load limitations were reviewed during the preparation of the Surveillance Frequency Control Program LAR (Reference 12) where it was determined that no limitations were needed in the PRA model.

Prior to implementation, either this finding will be closed or a sensitivity study case will be performed to determine the impact on the CDF and LERF results for those categorizations that could be adversely affected by this findinq.

The current DC battery model allows for only limited recovery of offsite power, i.e., the recovery of offsite power does Inadequate treatment of time-based not account for the extra time AS-87-01 AS-87 Not Met dependencies, e.g., recovery of offsite afforded by battery depletion.

power, HVAC treatment, and battery This method could result in depletion treatment.

conservative CDF and LERF values and potentially underestimate the delta-risk associated with batteries. It I

should not have a similar impact on the delta-risk calculated for Page 25 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement{ s)

{CC)

Other Applications other SSCs.

Prior to implementation, either this finding will be closed or a sensitivity study case will be performed to determine the impact on the CDF and LERF results for those categorizations that could be adversely affected by this findinq.

See response to AS-86-01.

Prior to implementation, either this finding will be closed or a SY-A21-01 SY-A21 Not Met Excessive electrical loading concerns sensitivity study case will be performed to determine the impact on the CDF and LERF results for those categorizations that could be adversely affected by this findinq.

The treatment of pre-imitators was reviewed during the preparation of the Surveillance Frequency Control Program LAR Not Met Screening pre-initiators values were used (Reference 12) where it was HR-Dl-01 HR-Dl in the model. Use of screening values for determined that the significant all pre-initiators only meets CC I.

pre-initiators were evaluated properly and no further model changes were required.

Prior to implementation, either Page 26 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement( s)

(CC)

Other Applications this finding will be closed or a sensitivity study case will be performed to determine the impact on the CDF and LERF results for those categorizations that could be adversely affected by this findinq.

This finding does not affect the use of the PRA model for LE-C9 calculating large early release LE-ClO Not Met No justification provided for equipment frequency. It may affect LE-C9-01 LE-Cll survivability or human actions credited calculations for a full Level 2 LE-C12 under adverse environments.

PRA analysis, which is beyond I

LE-D3 the scope of PRA model used in this application. Finding has no impact on this aoolication.

Internal flood model inconsistently Flag files were reviewed and propagates initiating event data into the determined to be complete.

IFQU-Al Not Met flag files used for quantification. Flag Documentation updates are IFQU-Al-01 IFEV-82 files need to be reviewed for needed to close this finding.

completeness and documentation The documentation updates should not affect the results.

updated to reflect this.

No impact on PRA applications.

The treatment of operator Model does not adequately develop actions in the internal flood model was reviewed.

Not Met human failure events specific to internal IFQU-A6-01 IFQU-A6 flood scenarios. HFEs from internal Documentation updates are events are "adjusted" with inadequate basis for those adjustments.

needed to close this finding.

The documentation updates should not affect the results.

Page 27 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications No impact on PRA applications.

CS-C3-01 CS-C3 Not Met Add the assumed cable routing for the The cable routing for the turbine stop valves and steam dump condenser steam dump valves valves in the Turbine Building and not was performed and documented documented in the Fire PRA Notebook as an example. The same Circuit Selection and Cable Analysis.

process would apply for the turbine stop valves.

I Document the cable routing in the Fire PRA Notebook Circuit Selection and Cable Documentation updates are Analysis of the turbine stop valves and needed to close this finding.

steam dump valves in the turbine The documentation updates building.

should not affect the results.

No impact on PRA applications.

FSS-81-01 FSS-81 Not Met Control room abandonment is considered The current methodology is only in case of loss of habitability or loss judged to be conservative.

of control due to a fire in the Main Control Room. Main Control Room abandonment Documentation updates are due to loss of control (for a fire in another needed to close this finding.

room) is however a plausible cause. While The documentation updates not crediting control room abandonment should not affect the results.

due to loss of control may be No impact on PRA applications.

conservative, a justification should be provided.

FSS-E3-01 FSS-E3 Not Met Uncertainty of the parameters used for Documentation updates are modeling the significant fire scenarios needed to close this finding.

was evaluated qualitatively, consistent The documentation updates with the requirements of SR FSS-E3 should not affect the results.

Capability Category I. However, no No impact on PRA applications.

statistical representation of uncertaintv Page 28 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications intervals was given, as required by See also IGN-Al0-01 and Capability Category IL SR FSS-HS is also UNC-Al-01 assessed at Capability Category I because SR FSS-E3 is assessed at Capability Cateqory I.

FSS-G2-01 FSS-G2 Not Met In the Multi-Compartment Fire Analysis Documentation updates are (P2091-2900-04, Revision 1, May 2013),

needed to close this finding.

Subtask 4 discusses screening multi-The documentation updates compartment fire scenarios based on hot should not affect the results.

gas layer dilution in the exposed fire No impact on PRA applications.

compartment. A finding is created to address potential cases where a Fire PRA target located in the connected compartment near a failed fire barrier or fire barrier element could be damaged by hot gases before hot gas layer dilution.

These multi-compartment interactions could have been improperly screened from further consideration FSS-Hl-03 FSS-Hl Not Met Several documents submitted to the peer Documentation updates are reviewers were draft versions. Examples needed to close this finding.

are: "Detailed Fire Modeling in Selected The documentation updates Point Beach Nuclear Plant Fire Zones" should not affect the results.

(1RCG27064.000.001), "Compartment No impact on PRA applications.

Analysis Notebook" (P2091-2900-01, Draft Rev. 2, May 2013), "Main Control Room Analysis" (P2091-2700-01, Revision 1, May 2013). Finding FSS-Hl-03 is created to ensure that such draft documents are reviewed, signed off, and their outputs are verified to be correctly Page 29 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement{ s)

{CC)

Other Applications implemented for quantification. Also, the 2011 peer review created a finding (FSS-82-01) related to the modeling of human failure events in case of control room abandonment. While this finding is deemed resolved in the 2013 focused peer review based on explanations from EPM, the documentation of the resolution in Report P2091-2910-01 ['Post-Fire Human Reliability Analysis) was not completed at the time of the peer review.

Accordingly, Finding FSS-Hl-03 calls for proper documentation of the resolution of I

2011 Finding FSS-82-01. Finally, Finding I

FSS-Hl-03 calls for documenting, in the relevant Fire PRA document, the basis for the fire resistance of wraps credited in the Fire PRA. This could be done, for example, in Report P2091-2900-01 (Compartment Analysis Notebook).

PRM-82-01 PRM-82 Not Met A review of the findings and resolutions As of the time of this submittal, from the Internal Events Peer Review the only remaining open internal indicated several findings against Accident events peer review findings Sequence, Success Criteria or Human identified in this fire PRA finding Reliability Analysis. Findings AS-81-01, are AS-86-01 and SY-A21-01.

AS-82-01, AS-86-01, SY-A21-01, SY-A22-01, HR-G7-01, and QU-83-01 could Prior to implementation, either potentially impact the fire PRA evaluation this finding will be closed or a considerably. Other findings, such as SY-sensitivity study case will be 83-01 and DA-C14-01, are associated with performed to determine the common cause or common mode svstem impact on the CDF and LERF Page 30 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications failures that could be important for fire results for those categorizations risk. Some or all of these findings could that could be adversely affected have an impact on the results of the by this finding.

FPRA.

Evaluate (qualitatively or quantitatively) these findings to determine the possible imoact on the FPRA.

PRM-89-01 PRM-89 Not Met P2091-2500-0l documents the PBNP Documentation updates are Plant Response Model development. PBNP needed to close this finding.

developed a stand-alone fault tree to The documentation updates evaluate the fire non-suppression should not affect the results.

probability for sequences where the No impact on PRA applications.

electric-driven fire pump was failed by a fire. This was based on an internal events model for the fire protection system where it was used as a backup system for cooling the auxiliary feed water pumps.

The new model used a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> mission time and was used to calculate a point estimate for the non-suppression probability that was to be added to the appropriate fire scenarios for quantification. The internal events model for the fire protection system where it was used as a backup system for cooling the auxiliary feed water pumps used a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time. The structure of the overall model is such that the non-suppression probability and failure of the fire protection system may show uo in the Page 31 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement( s)

(CC)

Other Applications same scenarios. However, it is not possible to identify the dependency between the non-suppression probability and the random failure of the fire protection system. Failure to suppress the specific fire within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> will result in the failure of the fire protection system within the first hour of its 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time but this will not be caught because the dependency is not modeled Revise model to include a factor to address the dependency between the new model used to evaluate the fire non-suppression probability for sequences where the electric driven fire pump was failed by a fire and the model for the fire protection system where it was used as a backup system for cooling the auxiliary feed water pumps. This needs to be carefully documented.

IGN-A10-01 IGN-AlO Not Met Propagating of the uncertainty intervals to Documentation updates are the Fire PRA results has not been done needed to close this finding.

under supporting requirement UNC-A2.

The documentation updates I

The uncertainty has not been evaluated.

should not affect the results.

No impact on PRA applications.

Propagate the uncertainty intervals to the Fire PRA model.

See also FSS-E3-01 and UNC-Al-01.

Page 32 of 53

I Finding Supporting Capability Disposition for 50.69 And for I Category Description Number Requirement(s)

(CC)

Other Applications I

HRA-82-01 HRA-B2 Not Met A review of the HRA calculator events The fire PRA HEPs were I

using the Caused Based Decision Tree reviewed to determine potential methodology (CBDTM) shows that credit impact. Only about 10% of the for graphically distinct is taken for all HRA HEPs that credited graphically events. As discussed in EPRI TR 101259 distinct procedure steps would and in the calculator documentation, be increased by more than a credit for graphically distinct is only factor of 2. Of these HEPs, only applicable in a flow chart if the shape, two are risk-significant HEPs color, etc. make the item standout as with risk achievement worth more important than other steps or in a values greater than 2. Based on procedure if the item is separated from this review, the impact on the other steps by a caution statement. If all model from this finding is events are graphically distinct, then in judged minimal.

effect none are graphically distinct.

Prior to implementation, either Credit graphically distinct factors only for this finding will be closed or a those events that stand out from the sensitivity study case will be other procedural or flow chart actions.

performed to determine the Credit for graphically distinct actions impact on the CDF and LERF reduces pee tree by a factor of 3. In some results for those categorizations cases, pee is the dominant contributor to that could be adversely affected the cognitive decision. Therefore, the HRA by this finding.

would be increased by a factor of 3.

HRA-Cl-01 HRA-Cl Not Met A median response time is chosen as 5 Documentation updates are minutes for all fire response actions.

needed to close this finding.

While five minutes is generally acceptable The documentation updates to respond to the fire alarms and send a should not affect the results.

fire brigade out to the area, it may take No impact on PRA applications.

significantly more time to go through the verifications and system requirements for that fire area before reaching the Page 33 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement( s)

(CC)

Other Applications operator manual action needed to respond to the event. The timing for these fire actions should be verified for feasibility, particularly for beyond Appendix R accidents with hardware failures in addition to the fire damage.

See F&O HRA-Al-01 for performing an HRA walkdown FQ-Al-01 FQ-Al Not Met A review of the FRANX database shows The only remaining issue is that some basic events that have been reconciliation of discrepancies mapped to scenarios, components, or found between the FRANX cables are not found in the CAFTA model.

mapping table and the CAFTA In particular:

model and its documentation.

The information in the mapping

1) 111 basic events mapped to cables table should be reviewed to are not in the CAFTA model. For example:

eliminate the extraneous FI-P23--CAB--SO and ESF-PT--N0-00469.

information and eliminate the

2) 21 basic events mapped to discrepancies.

components are not in the CAFTA model.

For example: 416-BKRC01A5215 and R--

Documentation updates are AOV-CC-00371 needed to close this finding.

3) 2 basic events mapped to scenarios The documentation updates are not found in the CAFTA model: FI-should not affect the results.

1CV110BCABSO and FI-1CV111-CAB-SO.

No impact on PRA applications.

A finding is created to ensure that basic events that are mapped to scenarios, components, or cables, are included in the CAFTA model, as appropriate. It could be that the basic events identified above Page 34 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications are "leftover" from a previous model revision that needs to be cleaned up and that they have no impact on the results.

Thus, this finding does not necessarily point to an incorrect model and rather points to a transparency issue.

FQ-A4-02 FQ-A4 Not Met Fire PRA quantification is understood to Documentation updates are be performed using the Fire PRA plant needed to close this finding.

response model that meets Technical The documentation updates Element PRM (plant response model) of should not affect the results.

the ASME/ANS Standard (see Note 2 of No impact on PRA applications.

SR FQ-A4). In that context, some multiple spurious operations (MSOs) appear to not have been adequately dispositioned and/or modeled in the Fire PRA. Examples are: loss of reactivity control, excessive RCS injection, and RCS overcooling.

Expert Panel discussion listed in Document P2092-110A-001 Rev. 0 (dated August 2010) included recommendations to revise the PRA model to address the potential impact due to fire. However, a later document (P2091-2500-02 Rev. 0, dated May 2011) appears to not address all MSOs nor provided justification for not being included in the PRA model as the earlier document has recommended. A finding is created to ensure that the generic list of MSOs from the industry owner groups be reviewed to verify that all MSOs relevant to Point Beach are Page 35 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications considered and their disposition properly documented.

FQ-81-01 FQ-81 Not Met Supporting Requirement FQ-81 calls for Documentation updates are CDF quantification in accordance with needed to close this finding.

HLR-QU-8. Supporting Requirements QU-The documentation updates 86, QU-87, and QU-88 address the should not affect the results.

treatment of mutually exclusive events in No impact on PRA applications.

the Fire PRA. A review of these mutually exclusive events appears to indicate that some important combinations may be missing or inconsistently applied. See for example Gate MEX-DC. A finding is created to ensure that mutually exclusive event combinations are systematically reviewed for aooropriateness.

FQ-El-01 FQ-El Not Met Supporting Requirement FQ-El calls for a Documentation updates are review of the CDF and LERF quantification needed to close this finding.

results in accordance with HLR-LE-F.

The documentation updates Supporting Requirements LE-Fl should not affect the results.

associated with that HLR requires, for No impact on PRA applications.

Capability Category II, a quantitative evaluation of the relative contribution to LERF from plant damage states and significant LERF contributors from Table 2-2.8-9 in the ASME/ANS standard (note:

the table number given here corrects an apparent typo given in the SR text). While the Quantification Notebook provides LERF contribution by compartment, scenarios, and equipment failure modes, Page 36 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications the LERF contributors from Table 2-2.8-9 do not appear to be evaluated in the notebook.

FQ-Fl-05 FQ-Fl Not Met Supporting Requirement FQ-Fl calls for Documentation updates are documentation of the CDF and LERF needed to close this finding.

analyses in accordance with HLR-QU-F.

The documentation updates Supporting Requirement QU-F4 should not affect the results.

associated with that HLR requires the No impact on PRA applications.

I characterization of sources of uncertainties and related assumptions.

However, several assumptions listed in the Quantification Notebook do not provide such characterization. For example:

1) In Section 2.4.3, instrument air is assumed failed in the Fire PRA. It should be clarified that instrument air failure is not credited as a success in the Fire PRA.

2) In Section 2.4.4, the effect of not crediting the charging pump low pressure trip modification is qualitatively evaluated, but no characterization of the potential adverse impacts that this modification may have is given.

3) In Section 2.4.6, the failure of the MSIVs to close is eliminated from further consideration, but the justification given Page 37 of 53

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications for this assumption is not substantiated with sufficient details.

4) In Section 2.S.3.6.1, an assumption is made that the probability of a hot short lasting greater than 10 minutes is 0.1. But the basis for this assumption is not clearly stated. If none is available, a sensitivity study to characterize the impact of this assumption is needed.

UNC-Al-01 UNC-Al Not Met Contrary to the requirements of QU-E3, Documentation updates are PBNP has not calculated a mean CDF and needed to close this finding.

uncertainty. Also, because they have The documentation updates currently not completed final should not affect the results.

quantification of CDF and LERF, they are No impact on PRA applications.

not really able to review contributors for reasonableness (e.g., to assure excessive See also FSS-E3-01 and conservatisms have not skewed the IGN-Al0-01.

results, level of plant specificity is appropriate for significant contributors, etc.) as required by LE-F2. The current results are definitely driven by conservatisms in the model and thus preclude performing a representative review of the contributors.

PBNP needs to complete the quantification of their model to the point where they have a fire CDF that is acceptable (e.g., of the order of SE-OS/year). As part of the final Page 38 of S3

Finding Supporting Capability Disposition for 50.69 And for Category Description Number Requirement(s)

(CC)

Other Applications quantification, PBNP needs to calculate the mean CDF and the associated uncertainty interval. PBNP should then review both the CDF and LERF contributors for reasonableness.

Page 39 of 53

External Hazards Screening Screening Resµlt Externa.1 Hazard Screening*

Screened?

(Y/N)

Criterion Comment (Note a)

Screened based on low probability of Aircraft Impact y

PS4 aircraft crash and small target size of SR structures.

Excluded due to site topography that Avalanche y

C3 would not support snow buildup that would lead to an avalanche.

Biological Event -

C4 Included implicitly in LOOP initiator.

y Animal Infestation cs Slow developing with limited impact.

Organic Material in Water is a more Cl credible scenario to cause intake Biological Event -

y C3 blockage than normal aquatic growth.

Aquatic Grown cs Slow developing hazard, can be detected and managed.

Biological Event -

C3 Slow developing hazard, can be Organic Material in y

cs detected and managed.

Water Coastal Erosion y

C3 Excluded based on design of plant.

Excluded since the capacity of the Drought y

C3 Ultimate Heat Sink (UHS) is not impacted by drought.

Page 40 of S3

External Hazards Screening Screening Result External Hazard Screening Screened?

(Y/N)

• criterion Comment

{Note a)

The external flooding hazard at the site was recently evaluated as a result of the post-Fukushima 50.54(f) Request for Information and the flood hazard reevaluation report (FHRR) was submitted to NRC for review on March 12, 2015 (Reference 16). The results indicate that flooding from all hazards, except local intense precipitation, are Y (see bounded by the current licensing External Flooding Cl basis (CLB) and do not pose a challenge comments) to the plant. Flooding from local intense precipitation was subsequently evaluated (Reference 17). Point Beach's focused evaluation and Mitigating Strategies Assessment (MSA) for flooding (Reference 18) conclude that the current station procedures for implementing the FLEX strategy provide an acceptable method of assuring safe shutdown.

The High Winds hazard was originally screened from applicability in the IPEEE.

Extreme Wind or N

n/a A High Winds PRA was subsequently Tornado developed and is being revised to more realistically reflect the as-built, as-operated plant.

Fog and mist may increase the frequency of accidents involving aircraft, Fog y

C4 ships, or vehicles. This weather condition is included implicitly in the accident rate data for these Transportation Accidents.

Page 41 of 53

External Hazards Screening Screening Result External Ha~artE! a)

Seismic margins analysis (SMA)

Seismic Activity N

n/a performed for the Individual Plant Evaluation-External Events (IPEEE).

Plant design includes snow loads and Cl other bounding loads.

Snow y

C4 cs Included implicitly in weather-related LOOP initiator.

Soil Shrink-Swell y

C3 Excluded based on structures founded Consolidation on bedrock and/or engineered fill.

Storm Surge N

n/a Included in External Flooding PRA documented in IPEEE.

There are no hazardous chemicals on or Toxic Gas y

C3 near the site which would cause control room habitability issues.

Cl Conservative bounding assessment Transportation y

C2 shows that these events can be Accident C3 screened.

C4 Tsunami y

C3 Not applicable to Point Beach since Point Beach is not located on an ocean.

Page 45 of 53

External Hazards Screening Screening Result External Hazard Screening

. Screened?

(Y/N)

Criterion Comment (Note a)

Screened based on low probability of Turbine-Generated y

PS4 turbine wheel failure and low probability Missiles of missile impacting safety-related equipment.

Volcanic Activity y

C3 Excluded due to distance from nearest cs potentially active volcano.

Waves N

n/a Included in External Flooding PRA documented in IPEEE.

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

Page 46 of 53

Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source Comments Cl. Event damage potential NUREG/CR-2300 Initial Preliminary and ASME/ ANS Screening is < events for which plant is Standard RA-Sa-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 C3. Event cannot occur close NUREG/CR-2300 enough to the plant to affect and ASME/ANS Standard RA-Sa-it.

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

Standard RA-Sa-to include within 2009 another event.

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

PSl. Design basis hazard ASME/ANS Progressive Screening cannot cause a core damage Standard RA-Sa-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-5/y modified in and the mean conditional ASME/ANS core damage probability is <

Standard RA-Sa-0.1.

2009 Page 47 of 53

Progressive Screening Approach for Addressing External Hazards Event,Analysis Criterion Source Comments NUREG-1407 and PS4. Bounding mean CDF is ASME/ANS

< lE-6/y.

Standard RA-Sa-2009 Screening not successful.

NUREG-1407 and Detailed PRA PRA needs to meet ASME/ANS requirements in the Standard RA-Sa-ASME/ANS PRA Standard.

2009 Page 48 of 53

Disposition of Key Assumptions/Sources of Uncertainty Assumption/

Uncertainty In the evaluation of the frequency of a stuck open safety valve on the pressurizer (which is a contributor to medium LOCA) there have been no cases in which the safety valves have opened during a trip. Therefore, it is conservatively assumed that 0.5 openings occurred in all the trip events in the historical record.

The SBO event tree assumes SI is required for all success sequences to address RCP seal leakage, even if power is restored early.

For the ventilation system of the PAB electrical equipment rooms, the model assumes that loss of air inlet to the air handling unit from the turbine building duct or recirculation line will lead to system failure.

Discussion This is a typical assumption used in data analysis. If no failures are in the plant-specific database, it is assumed that there are 0.5 failures in the next demand. This assumption has a conservative but minor impact in the resulting mean value. This does not require evaluation.

It is assumed in the SBO event tree that SI is required for all success. While this is conservative for very short SBO events, in general SBO events are of sufficient duration that an RCP Seal LOCA would occur, requiring primary makeup. Since SBO is dominated by failure to recover power, this has essentially no discernable impact and does not require evaluation.

This ventilation system is new to the PRA. It was added based on recent room heatup calculations that were evaluated in the HVAC PRA Notebook. The air handling units take suction from the turbine building duct and supply air to the battery charger rooms. Heat is removed by the air passing through chiller units that are cooled by service water. Loss of makeup air from the turbine Page 49 of 53 Disposition This assumption conservatively estimates the frequency of a valve demand during a unit trip.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This assumption is conservative in that SI is required for all SBO accident scenarios with out the consideration for recovery of power.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This modeling assumption conservatively over-estimates the failure of the PAB electrical room HVAC system and the equipment in those areas.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

Assumption/

Uncertainty During the injection phase of RHR system operation, it is assumed that component cooling water to the RHR pumps and heat exchange rs is not required for successful system operation.

During the recirculation phase of RHR system operation, it is assumed that component cooling to the RHR pumps and heat exchange rs is required for successful system operation. The RHR heat exchange rs are also required to remove decay heat from the containment sump water to provide core cooling.

Discussion hall will decrease the quantity of air flow due to leaks and also increase the heat load on the chiller units. Normal operation of the system relies on cool air from the turbine hall being mixed with hot air being exhausted from the rooms. Continuous recirculation without make-up will result in increasing room tern peratu res and eventual equipment failure.

The assumption states that the RHR pumps do not require CCW during the injection phase, when suction is taken from the refueling water storage tank. The Loss of CCW event tree does not credit RHR since CCW is required for the recirculation phase. In the remaining event trees, even if RHR injection is successful, RHR recirculation mode is failed without CCW and this is directly modeled in the system fault tree. This is realistic and does not require evaluation.

This assumption for the CCW pumps to provide cooing to the RHR pumps when in recirculation mode is conservative. The RHR pump seals would not fail immediately on a loss of CCW.

RHR pump seal failure without CCW would occur sometime between the start of RHR recirculation and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This level of modeling is consistent with the development of all of the accident sequences (PRA Page 50 of 53 Disposition This assumption realistically models the plant system design; therefore, no sensitivity analysis is required for this application.

Not crediting RHR for the recirculation phase when CCW is unavailable is conservative.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

Assumption/

Uncertainty Operator actions to control AFW flow later in an accident sequence are not explicitly modeled in the AFW system fault trees.

The HEP for operator failure to close containment isolation valves following the failure of the valves to automatically close is modeled in the PRA.

Failures of expansion joints are not modeled in the fire protection system fault tree.

The cycle time for the instrument air dryers was assumed to be 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Demand failures occurring with the cycling include fail-to-close and fail-to-Discussion mission time is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) and does not require evaluation.

Current HRA practice is to assume that if the operators successfully control flow initially they will do so correctly for the remainder of the mission time; therefore, failure to control feedwater flow late in an accident sequence is not modeled.

This assumption impacts only the LERF model. The timing of this operator action was developed based on Large LOCA and applied to all LOCAs.

For vessel rupture LOCA, there would be insufficient time to perform this action. A sensitivity analysis performed by failing this action resulted in negligible change to LERF.

An estimation of expansion joint failure rate is negligible when compared to other mechanic al failure modes of components in the fire protection system. Therefore, the expansion joint failures are considered to be captured within the error factor of the failure rate coefficients for those other components.

Inclusion of expansion joint failures would have a negligible impact on the results.

The assumption states that for the instrument air dryers it is conservatively assumed that the cycle time for the dryers is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Page 51 of 53 Disposition A sensitivity analysis was performed evaluating the impact of not controlling AFW flow for the full PRA mission time of the system. Based on the small changes observed, no additional sensitivity analysis is required for this application.

A sensitivity analysis was performed to assess this modeling assumption. Based on the negligible changes observed, no additional sensitivity analysis is required for this application.

A sensitivity analysis was performed to assess this modeling assumption. Based on the negligible changes observed, no additional sensitivity analysis is required for this application.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

open.

Assumption/

Uncertainty The mission time for battery-operated components is one hour.

RCS pressure is considered to be high at the time of core damage for transient and SGTR events.

Stuck open PORV/PSV and Large RCP seal LOCA sequences have depressurization capability.

However, these sequences have conservatively been assumed to have high RCS pressure in the PBNP LERF model.

RCS pressure is considered to be high at the time of core damage for Transient and SGTR events.

All core damage accident class sequences in which core damage occurs at high reactor pressure, and Discussion Additionally, the system would not be failed immediately if moist air were to pass through the system. This is a realistic estimate.

Actual battery life is expected to be greater than one hour, which conservatively supports the one hour mission time for the batteries. Within this time frame the diesel generators need to be successfully supplying their design basis loads such that the buses can be transferred to operate off of the diesels.

For the transient and SGTR event trees, core damage is only reached when depressurization is successful but long term cooling is not successful and there is no primary system injection. In these cases, the primary system would re-pressurize.

Conservative. LERF would be reduced because assuming high pressure will minimize containment pressurization time and maximize containment pressure.

Conservative. Assuming high pressure will minimize containment pressurization time and maximize containment pressure.

Conservative. A PI-SGTR is only of concern if the RCS is at high pressure and a SG is depressurized to atmospheric Page 52 of 53 Disposition Crediting actual higher capacities of the batteries would result in addition al mitigation capabilities made available.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This assumption is conservative since some of the accident sequences have successful depressurization without injection or cooling.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

Assumption/

Uncertainty the steam generators are dry at the time of core damage are assumed to have the potential to lead to pressure-induced SGTR.

For simplicity, it is assumed that all Steam Generator Tube Rupture (SGTR) and Interfacing Systems LOCA (ISLOCA) initiated events are containment bypass scenarios.

Systems without cable tracing are failed unless further analysis was performed to assure systems are not compromised by the transient or fire (credit by exception).

Instrument Air system is assumed failed for the fire PRA model.

HEAF and bus duct fires are assumed to damage fire wrap and fire rated cables, negating fire rating for these fire scenarios.

For fires in a single electrical cabinet, it is assumed that a fire wou Id damage all cables and components within the cabinet. No probability of non-suppression is calculated for single cabinet fires as damage is assumed at ignition time.

Discussion pressure. Not all sequences have SG depressurized to atmospheric pressure.

This is conservative since LERF would be reduced if scrubbing by AFW was credited or ISLOCAs credited auxiliary building room flooding.

This assumption introduces uncertainty. While this is a conservative approach, the tracing of cables is an intense process requiring additional effort. This assumption is reviewed during the quantification process.

This assumption over-predicts CDF and LERF.

This assumption is conservative with respect to CDF and LERF, and contributes uncertainty to the analysis.

Although electric al ca bi net fires that spread to adjacent cabinets can be manually suppressed to inhibit the spread of the fire, it is assumed that there is insufficient time for suppression of a single cabinet fire.

Page 53 of 53 Disposition This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.

This is a conservative assumption; therefore, no sensitivity analysis is required for this application.