RS-17-123, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors.

Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors.

ML17244A093
Person / Time
Site:	Byron, Braidwood
Issue date:	09/01/2017
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-17-123
Download: ML17244A093 (67)

Text

4300 Winfield Road Warrenville, IL 60555 www.exeloncorp.com 10 CFR 50.90 10 CFR 50.69 RS-17-123 September 1, 2017 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. 50-454 and 50-455

Subject:

Application to Adopt 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors" In accordance with the provisions of 10 CFR 50.69, and 10 CFR 50.90, Exelon Generation Company, LLC (Exelon) is requesting an amendment to the license to Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, and Renewed Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, 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 for nuclear power reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, alternative treatment requirements can be implemented in accordance with this regulation. For equipment determined to be of high safety significance, requirements will not be changed or will be enhanced. This allows improved focus on equipment that has safety significance resulting in improved plant safety.

The enclosure to this letter provides the basis for the proposed change to both Braidwood Station and Byron Station Units 1 and 2 Renewed Facility 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.

License Amendment Request Adopt 10 CFR 50.69 Docket Nos. 50-456 and 50-457 (Braidwood)

Docket Nos. 50-454 and 50-455 (Byron)

September 1, 2017 Page 2 Exelon intends to submit a separate license amendment request to revise both Braidwood Station and Byron Station Units 1 and 2 Technical Specifications to adopt TSTF-505, Revision 1, "Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b," within the next six months using the same PRA model described in the Enclosure to this letter. Exelon requests that the NRC coordinate their review of the PRA technical adequacy description in Section 3.2 and 3.3 of this enclosure for both applications. This would reduce the number of Exelon and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

Exelon requests approval of the proposed license amendment by September 1, 2018, with the amendment being implemented within 60 days.

The proposed changes have been reviewed and approved by the Braidwood Station and Byron Station Plant Operations Review Committees in accordance with the requirements of the Exelon Quality Assurance Program.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), Exelon is notifying the State of Illinois of this application for license amendment by transmitting a copy of this application, with attachments, to the designated Illinois Official.

There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Jessica Krejcie at (630) 657-2816.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 1st day of September, 2017.

Respectfully, D~1Ylc;fV David M. Gullatt Manager - Licensing Exelon Generation Company, LLC

Enclosure:

Evaluation of the Proposed Change cc: NRC Regional Administrator, Region Ill, USNRC NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector- Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety

Enclosure Enclosure Evaluation of the Proposed Change TABLE OF CONTENTS 1

SUMMARY

DESCRIPTION ................................................................................. 3 2 DETAILED DESCRIPTION ................................................................................. 4 2.1 CURRENT REGULATORY REQUIREMENTS ................................................... 4 2.2 REASON FOR PROPOSED CHANGE ............................................................. 4

2.3 DESCRIPTION

OF THE PROPOSED CHANGE ................................................ 5 3 TECHNICAL EVALUATION ................................................................................. 7 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i)) ............. 8 Overall Categorization Process........................................................ 8 Passive Categorization Process ..................................................... 10 3.2 TECHNICAL ADEQUACY EVALUATION (10 CFR 50.69(b)(2)(ii)) ................... 11 Internal Events and Internal Flooding ........................................... 11 Fire Hazards ............................................................................... 11 Seismic Hazards .......................................................................... 12 Other External Hazards ................................................................ 12 Low Power & Shutdown ............................................................... 13 PRA Maintenance and Updates ..................................................... 13 PRA Uncertainty Evaluations ........................................................ 13 3.3 PRA REVIEW PROCESS RESULTS (10 CFR 50.69(b)(2)(iii)) ......................... 14 3.4 RISK EVALUATIONS (10 CFR 50.69(b)(2)(iv)) ........................................... 15 4 REGULATORY EVALUATION ............................................................................ 16 4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA .............................. 16 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS ........................... 16

4.3 CONCLUSION

S ....................................................................................... 18 5 ENVIRONMENTAL CONSIDERATION ................................................................ 19 6 REFERENCES ................................................................................................. 20 i

Enclosure LIST OF ATTACHMENTS : List of Categorization Prerequisites ................................................... 23 : Description of PRA Models used in Categorization ............................. 24 : Disposition and Resolution of Open Peer Review Findings ................... 25 and Self-Assessment Open Items ..................................................... 25 : External Hazards Screening .............................................................. 33 : Progressive Screening Approach for Addressing External Hazards........ 47 : Disposition of Key Assumptions/Sources of Uncertainty ...................... 48 ii

Enclosure 1

SUMMARY

DESCRIPTION The proposed amendment would modify the licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, 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.

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Enclosure 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 "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.

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Enclosure 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 places 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 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 Exelon to improve focus on equipment that has safety significance resulting in improved plant safety.

2.3 DESCRIPTION

OF THE PROPOSED CHANGE Exelon proposes the addition of the following condition to the renewed facility operating licenses of both Braidwood Station Units 1 and 2, and Byron Station Units 1 and 2, to document the NRC's approval of the use 10 CFR 50.69.

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Enclosure Exelon 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 September 1, 2018.

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).

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Enclosure 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 § 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 § 50.69(c)(1)(i).

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

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

Exelon intends to submit a separate license amendment request to revise technical specifications to adopt TSTF-505, Revision 1, "Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b," for both Braidwood Station and Byron Station Units 1 and 2 within the next 6 months using the same PRA model described in this Enclosure. Exelon requests that the NRC coordinate their review of the PRA technical adequacy description in Section 3.2 and 3.3 of this enclosure for both applications. This would reduce the number of Exelon and NRC resources necessary to complete the review of the applications. This request should not be considered a linked requested licensing action (RLA), as the details of the PRA models in each LAR are complete which will allow the NRC staff to independently review and approve each LAR on their own merits without regard to the results from the review of the other.

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Enclosure 3.1 CATEGORIZATION PROCESS DESCRIPTION (10 CFR 50.69(b)(2)(i))

Overall Categorization Process Exelon will implement the risk categorization process in accordance with 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 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 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 § 50.69(f)(1) will be documented in Exelon 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.

• 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.

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Enclosure

• Exelon 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, both Braidwood and Byron Stations will not take credit for alternate means unless the alternate means are proceduralized and included in Licensed Operator training.

The technical adequacy of the risk analysis being implemented for each hazard is described in the following:

• Internal Event Risks: Internal events including internal flooding PRA model BB016A, March 2017 (Reference 3).

• Fire Risks: Fire PRA model versions BB11b-FL, updated to reflect the 2017 internal event risk model (References 3 and 4).

• Seismic Risks: Success Path Component List (SPCL) from the IPEEE seismic analysis accepted by NRC SEs dated May 30, 2001 (References 5 and 6).

• Other External Risks (e.g., tornados, external floods, etc.): Using the IPEEE screening process as approved by NRC SEs dated May 30, 2001 (References 5 and 6). The other external hazards were determined to be insignificant contributors to plant risk.

• Low Power and Shutdown Risks: Qualitative defense-in-depth (DID) shutdown model for shutdown configuration risk management (CRM) based on the framework for DID provided in NUMARC 91-06, "Guidance for Industry Actions to Assess Shutdown Management" (Reference 7), which provides guidance for assessing and enhancing safety during shutdown operations.

A change to the categorization process that is outside the bounds specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach) will not be used without prior NRC approval. The SSC categorization process documentation will include the following elements:

1. Program procedures used in the categorization

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

3. Mapping of components to support function(s)

4. PRA model results, including sensitivity studies 9

Enclosure

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 periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members Passive Categorization Process For the purposes of 10 CFR 50.69 categorization, passive components are those components that have a pressure retaining function. Passive components and the passive function of active components will be evaluated using the Risk-Informed Repair/Replacement Activities (RI-RRA) methodology consistent with the Safety Evaluation (SE) by the Office of Nuclear Reactor Regulation "Request tor Alternative ANO2-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 8).

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 are consistent with ANO-2 RI-RRA License Amendment as the rule does not remove the repair and replacement provisions of the ASME Code required by § 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 part of the reactor coolant system or containment. This is further clarified in the rules 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 only evaluates the consequence of a rupture of the SSCs pressure boundary. As stated in the ANO SE, 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.

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Enclosure 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 9). 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 both Braidwood Station Units 1 and 2 and Byron Station Units 1 and 2 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.

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

Fire Hazards The Braidwood Station and Byron Station 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 Exelon risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the Braidwood and Byron units. Attachment 2 at the end of this enclosure identifies the applicable Fire PRA model.

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Enclosure Seismic Hazards The Braidwood Station and Byron Station 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 10) 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 seismic plant equipment list (SPEL) 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 SPEL would identify credited equipment as HSS regardless of their capacity, frequency of challenge or level of functional diversity.

An evaluation was performed of the as-built, as-operated plant against the SMA SPEL.

The evaluation was a comparison of the as-built, as-operated plant to the plant configuration originally assessed by the SMA. Differences were reviewed to identify any potential impacts to the equipment credited on the SPEL. Appropriate changes to the credited equipment were identified and documented. This documentation is available for audit. The Exelon risk management program ensures that future changes to the plant will be evaluated to determine their impact on the SMA and risk categorization process.

Other External Hazards The Braidwood Station and Byron Station categorization process will use screening results from the Individual Plant Evaluation of External Events (IPEEE) in response to GL 88-20 (References 5 and 6) for evaluation of safety significance related to the following other external hazards:

• High Winds

• External Flooding

• Transportation and Nearby Facility Accidents

• Other External Initiating Events (i.e., other "HFO" Events)

All SSCs credited in other IPEEE external hazards are considered HSS. All other external hazards were screened from applicability to both Braidwood Station and Byron Station Units 1 and 2, per a plant-specific evaluation in accordance with GL 88-20 (Reference 10) 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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Enclosure Low Power & Shutdown The Braidwood Station and Byron Station 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.

PRA Maintenance and Updates The Exelon risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant for each of the Braidwood Station and Byron Station 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, Exelon 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.

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 Exelon will utilize a factor of 3 to increase the unavailability or unreliability of LSS components consistent with that approved for Vogtle in Reference 9. Consistent with the NEI 00-04 guidance, Exelon 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 13

Enclosure study, the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in PRAs for all systems that have been categorized are increased by a factor of 3. This sensitivity study together with the periodic review process assures that the potential cumulative risk increase from the categorization is maintained acceptably low. The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed before reaching the rate assumed in the sensitivity study.

Sources of model uncertainty and related assumptions have been identified for the Braidwood Station and Byron Station PRA models using the guidance of NUREG-1855 (Reference 11) and EPRI TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments" (Reference 12).

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

The list of assumptions and sources of uncertainty have been reviewed to identify those which would be significant for the evaluation of this application. If the Braidwood and Byron PRA models use non-conservative treatments, or use methods 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 categorization risk calculations were considered key for this application.

Key Braidwood Station and Byron Station PRA model specific assumptions and sources of uncertainty for this application have been identified and dispositioned and are available for NRC audit. The conclusion of this review is that no additional sensitivity analyses are required to address both Braidwood Station and Byron Station 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 13) consistent with NRC RIS 2007-06.

The internal events PRA model was subject to a self-assessment and a full-scope peer review conducted in July 2013.

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Enclosure The fire PRA model was subject to a self-assessment and a full-scope peer review conducted in October 2015 (Braidwood Station Units 1 and 2) and June 2015 (Byron Station Units 1 and 2).

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 14) as accepted by NRC in the staff memorandum dated May 3, 2017 (ML17079A427) (Reference 15). The closure review was conducted in February 2017.

The results of this review have been documented and are available for NRC audit. provides a summary of:

• Open items and disposition from the Braidwood and Byron RG 1.200 self-assessment.

• Open findings and disposition of the Braidwood and Byron peer review.

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)(1)(i).

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

The Braidwood Station and Byron Station 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 §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.

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Enclosure 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 Exelon proposes to modify the licensing basis to allow for the voluntary implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors." The provisions of 10 CFR 50.69 allow adjustment of the scope of equipment subject to special treatment controls (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). For equipment determined to be of low safety significance, 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.

Exelon 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:

16

Enclosure

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.

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 17

Enclosure 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, Exelon 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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

18

Enclosure 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.

19

Enclosure 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. BB-PRA-014, Byron / Braidwood Probabilistic Risk Analysis Quantification Notebook, Revision BB016a, March 2017.

4. Byron and Braidwood Probabilistic Risk Assessment - Application-Specific Model (ASM), BB-ASM-005 Revision 0

5. Review of Braidwood Individual Plant Examination of External Events (IPEEE)

Submittal (TAC nos. M83593 and M83594), dated May 30, 2001

6. Review of Byron Individual Plant Examination of External Events (IPEEE) Submittal (TAC nos. M83600 and M83601), dated May 30, 2001

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

8. ANO-2 SE, Safety Evaluation 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.

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

10. NRC Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4," USNRC, dated June 28, 1991.

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

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

13. NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Revision 2, March 2009.

14. 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, Accession Number ML17086A431.

15. USNRC Letter to Mr. Greg Krueger (NEI), U.S. Nuclear Regulatory Commission 20

Enclosure 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, Accession Number ML17079A427.

16. Byron/Braidwood Generating Stations Updated Final Safety Analysis Report (UFSAR), Revision 16, December 2016

17. Exelon Generation Company, LLC, RS-14-052, Exelon Generation Company, LLC Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 2, Flood Hazard Reevaluation Report, March 12, 2014 for Braidwood Station

18. Exelon Generation Company, LLC, RS-15-112, "Response to Request for Additional Information Regarding Fukushima Lessons Learned - Flood Hazard Reevaluation Report," June 24, 2015

19. Exelon Generation Company, LLC, RS-14-053, Exelon Generation Company, LLC Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 2, Flood Hazard Reevaluation Report, dated March 12, 2014 for Byron Station

20. Letter from Tekia V. Govan, (US NRC) to Bryan Hanson (Exelon), "Byron Station, Units 1 and 2-Interim Staff Response to Reevaluated Flood Hazards Submitted in Response to 10 CFR 50.54(f) Information Request- Flood-Causing Mechanism Reevaluation (TAC nos. MF3893 and MF3894), September 3, 2015

21. Exelon Generation Company, LLC, RS-16-180, "Mitigating Strategies Flood Hazard Assessment (MSFHA) Submittal, September 30, 2016

22. Letter from Tekia V. Govan, (US NRC) to Bryan Hanson (Exelon), "Byron Station, Units 1 and 2- Flood Hazards Mitigation Strategies Assessment," (TAC nos. MF3893 and MF3894), October 21, 2016

23. NRC Safety Evaluation Report, "Electric Power Research Institute (EPRI) Topical Reports Concerning Tornado Missile Probabilistic Risk Assessment (PRA)

Methodology," October 26, 1983.

24. NRC Regulatory Issue Summary 2008-14, "Use of TORMIS Computer Code for Assessment of Tornado Missile Protection," June 16, 2008.

25. Braidwood Station Technical Specifications

26. Byron Station Technical Specifications

27. Byron/Braidwood Generating Stations Updated Final Safety Analysis Report (UFSAR) Section 2.5 "Geology, Seismology, and Geotechnical Engineering."

28. NUREG/CR-6890, Reevaluation of Station Blackout Risk at Nuclear Power Plants, Analysis of Loss of Offsite Power Events: 1986-2004, December 2005.

29. A. C. Thadani, Transmittal of Technical Work to Support Possible Rulemaking on Risk-Informed Alternative to 10 CFR 50.46/GDC 35," July 31, 2002.

21

Enclosure

30. NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants, February 2007.

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

32. WCAP-15603, WOG 2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs, June 2003.

33. Westinghouse Owner's Group Simplified Level 2 Modeling Guidelines, WCAP-16341-P, 2005.

34. Westinghouse, Joint Westinghouse Owners Group/Westinghouse Program: ATWS Rule Administration Process, WCAP-1192, December 1988.

35. EPRI 1013141, Pipe Rupture Frequencies for Internal Flooding PRAs, Revision 1, 2006.

36. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RAS-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.

37. NUREG-1524, A Reassessment of the Potential for an Alpha-Mode Containment Failure and a Review for the Current Understanding of Broader Fuel-Coolant Interaction Issues, August 1996.

38. NUREG/CR-6850 (also EPRI 1011989), Fire PRA Methodology for Nuclear Power Facilities, September 2005, with Supplement 1 (EPRI 1019259), September 2010.

39. NUREG-2169 (also EPRI 300200936), Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience through 2009, Revision 0, January 2015.

40. BB-PRA-026, Braidwood Uncertainty and Sensitivity Notebook, Revision 2, June 2012.

41. NUREG-2178 (also EPRI 3002005578), Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE-FIRE) - Volume 1: Peak Heat Release Rates and Effects of Obstructed Plume, Revision 0, April 2016.

42. NUREG/CR-7150, Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE), October 2012.

43. BB-PRA-014, Byron / Braidwood Probabilistic Risk Analysis Quantification Notebook, Revision BB011b, September 2012.

22

Enclosure Attachment 1 : List of Categorization Prerequisites Exelon 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 as discussed in Section 3.1.1.

23

Enclosure Attachment 2 Attachment 2: Description of PRA Models used in Categorization Plant Units Model Baseline CDF Baseline LERF Comments Full Power Internal 2016 Full Power Events and Internal Events Internal Flooding 1.1E-05 (Unit 1) 9.0E-07 (Unit 1)

(FPIE) with 1&2 (FPIE)

Internal Flooding 1.1E-05 (Unit 2) 9.0E-07 (Unit 2)

PRA Model of BB016A for both Byron Record (MOR) units Application Specific Fire PRA 5.6E-05 (Unit 1) 3.1E-06 (Unit 1) Model (ASM) to the 1&2 Fire Update which BB11b-FL for both 6.1E-05 (Unit 2) 3.1E-06 (Unit 2) was based on the units 2011 PRA MORs Full Power Internal 2016 Full Power Events and Internal Events Internal Flooding 1.2E-05 (Unit 1) 1.0E-06 (Unit 1)

(FPIE) with 1&2 (FPIE)

Internal Flooding 1.2E-05 (Unit 2) 1.0E-06 (Unit 2)

PRA Model of BB016A for both Braidwood Record (MOR) units Application Specific Fire PRA 5.5E-05 (Unit 1) 3.2E-06 (Unit 1) Model (ASM) to the 1&2 Fire Update which BB11b-FL for both 6.6E-05 (Unit 2) 3.3E-06 (Unit 2) was based on the units 2011 PRA MORs 24

Enclosure Attachment 3 Attachment 3: Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items Finding Supporting Capability Require- Category Finding Number Disposition for 50.69 ment(s) (CC) Description SY-B12- SY-B12 Cat I/II/III Recovery actions are included in the model. PARTIALLY RESOLVED 01 However, a heating, ventilation and air conditioning (HVAC) dependency is not Updated room cooling calculations and survivable included for the switchgear and battery rooms. temperature evaluations have been performed and indicate The room heatup calculations are based on a 4 support for the current modeling assumptions for most hour mission time, with assumption that within scenarios for the Engineered Safety Feature (ESF) and that period, operators would take action to Non-ESF Switchgear Rooms. The only identified potential restore cooling to the rooms. A similar issue impact on the model is for high energy line break (HELB) also was identified as part of the Supporting scenarios, which are an overall small contributor to the Requirement (SR) H2-E2 review where local baseline risk results. A sensitivity calculation that simply operator actions to provide inverter room inserts new HELB-related room cooling requirements shows ventilation cooling (such as opening the door a small increase in baseline CDF and LERF, but these and installing temporary fans) were not increases do not trigger consideration of an emergent model identified. This was identified as a necessary update per the Exelon Risk Management procedures.

operator action in the Equipment Survivability Model changes to incorporate these additional HELB-related Notebook to recover from a failed room room cooling requirements are being tracked in the URE ventilation cooling system. (Updating Requirements Evaluation) database and will be incorporated as part of the normal cumulative model update process.

This open item will have minimal impact on the PRA results and minimal impact on this application.

There is no impact to the FPRA from the outstanding HELB issue.

16-4 CS-B1 Cat I Based on information provided there is lack of OPEN.

details to meet Capability Category II/III as the only statement is in Section 3.8 of the A review of breaker coordination was performed for credited

'Braidwood Fire PRA Cable Selection Notebook FPRA power supplies but not all calculations were available (BW-PRA-021.03), Rev 0' : "The Braidwood at the time of the Fact and Observation (F&O) Finding 25

Enclosure Attachment 3 Finding Supporting Capability Require- Category Finding Number Disposition for 50.69 ment(s) (CC) Description Fire PRA reviewed the electrical coordination Closure Review. These calculations have been identified calculations for applicability to the Fire PRA. and their results will be incorporated into the current FPRA These were reviewed for each of the credited update, so that this issue will have no impact on this power supplies in the model." This did not application.

provide analysis or identify any additional requirements; it only stated that it was reviewed.

18-12 UNC-A1 Not Met Some anomalies were observed in the OPEN database that was used for the uncertainty analysis which was different from that used in A review and update, if needed, of the basic event (BE) the FPRA. probability distributions used in the parametric uncertainty analysis is required. Some potential discrepancies may exist between the database used for the parametric uncertainty, for UNCERT runs, and the probability distributions that should apply to some of the basic events.

It is anticipated that this review may identify some changes required in the parametric uncertainty analysis but a significant change in the probability distribution of the FPRA total CDF and LERF results is not expected. This issue will be resolved in conjunction with the next FPRA model update.

The impact of this issue is limited to the parametric uncertainty analysis. It has no impact on the FPRA results and no impact on this application.

19-8 FQ-E1 Not Met Document the relative contribution of PARTIALLY RESOLVED contributors to LERF.

The documentation and review of results did not include importances by accident progression contributors.

Importances by basic event (BE) and sequence flags as well as for LERF plant damage states were provided. This is a documentation issue only since the importance by accident progression contributors can be extracted from the current model but is not readily available in the current documentation. The importances by BE, sequence flags and plant damage states will likely be the more useful input and were provided and reviewed as part of the model development.

26

Enclosure Attachment 3 Finding Supporting Capability Require- Category Finding Number Disposition for 50.69 ment(s) (CC) Description This is a documentation issue with no impact on this application.

19-9 FQ-E1 Not Met HLR-QU-D7 requires review of importance of PARTIALLY RESOLVED components and basic events to determine that they make logical sense. The documentation and review of results did not include importances by accident progression contributors.

Importances by basic event (BE) and sequence flags as well as for LERF plant damage states were provided. This is a documentation issue only since the importance by accident progression contributors can be extracted from the current model but is not readily available in the current documentation. The importances by BE, sequence flags and plant damage states are the more useful input and were provided and reviewed as part of the model development.

This is a documentation issue with no impact on this application.

19-11 FQ-F1 Not Met There is no documentation of the importance PARTIALLY RESOLVED measures for Braidwood Unit 2 CDF/LERF from QU-F2 (j). The documentation and review of results did not include importances by component. Importances by basic event (BE) were provided. This is a documentation issue only since the importance by component can be extracted from the current model but is not readily available in the current documentation. The importances by BE are the more useful input and were provided and reviewed as part of the model development.

This is a documentation issue with no impact on this application.

19-15 FQ-F1 Not Met Document the process used to identify plant PARTIALLY RESOLVED damage states and accident progression contributors. The documentation and review of results did not include importances by accident progression contributors.

Importances by basic event (BE) and sequence flags as well as for LERF plant damage states were provided. This is a documentation issue only since the importance by accident 27

Enclosure Attachment 3 Finding Supporting Capability Require- Category Finding Number Disposition for 50.69 ment(s) (CC) Description progression contributors can be extracted from the current model but is not readily available in the current documentation. The importances by BE, sequence flags and plant damage states are the more useful input and were provided and reviewed as part of the model development.

This is a documentation issue with no impact on this application.

20-1 IGN-A7 Cat I/II/III During the Peer Review walkdown, three rooms PARTIALLY RESOLVED WITH OPEN DOCUMENTATION were checked and found to have errors in ignition source counting: 11.4A-1, 11.4A-2, and Two panels which were missing for the fire scenarios in the 11.6-0. Diesel Generator Auxiliary Feedwater Pump (DGAFWP) room were added to that room. Subsequently it was determined that the actual location differed from the change that was incorporated in the current FPRA. Panels 1/2PL85JA and 1/2PL85JB were included in their respective Braidwood Unit 1 DGAFWP room (fire area 11.4-1) but were missing from the Unit 2 DGAFWP room (fire area 11.4-2).

These same panels were included in both Byron Unit 1 and Unit 2 DGAFWP rooms (fire areas 11.4-1 and 11.4-2). It has subsequently been determined that panels 1/2PL85JA are located in fire area 11.4-0 and not in fire area 11.4-1 or 11.4-

2. A review of the risk of the quantified scenarios indicates that the impact of the correction in panel locations will result in a net decrease in risk on the order of 1% of the total plant risk. Therefore, the impact of this open item is a small conservatism in the FPRA. This item will be closed in the next revision to the FPRA.

Resolution of this issue will have minimal impact on this application given that the impact is a small conservatism in the FPRA.

20-8 FSS-B2 Cat I Currently there is no credit given for operators OPEN.

safely shutting down the plant from the remote shutdown panel or from actions outside the The Main Control Room (MCR) abandonment scenario MCR. This results in a CCDP of 1.0 being conditional core damage probability (CCDP) was quantified applied to scenarios that require operator using a scaling factor applied to the scenario CCDP /

abandonment or where sufficient functionality is CLERP. A scaling factor was applied for each CCDP /

28

Enclosure Attachment 3 Finding Supporting Capability Require- Category Finding Number Disposition for 50.69 ment(s) (CC) Description lost at the Main Control Board (MCB). CLERP range to reflect the significance of the shift of command and control for the complexity of the shutdown based on the CCDP / CLERP. The un-scaled CCDP /

CLERP values for the abandonment scenario used human error probabilities (HEPs) associated with command and control remaining in the Main Control Room (MCR) as opposed to transfer to the remote shutdown panel. This scaling of MCR abandonment CCDP / CLERP to address the impact of an outside the MCR command and control location has been accepted by the USNRC in Safety Evaluations associated with transition to NFPA 805 at several nuclear plant sites (see the USNRC Safety Evaluations (SEs) for the Turkey Point (ADAMS ML15061A237), St. Lucie (ADAMS ML15344A346) and Farley (ADAMS ML14308A048) NFPA 805 LARs for NRC acceptance of this methodology). This methodology is considered bounding. A revised methodology is being issued under NUREG 1920, Supplement 1.

This revised methodology will be reviewed and addressed in the Byron and Braidwood Fire PRAs during the next update of the models subsequent to final issuance of the new guidance.

Resolution of this issue will have no impact on this application given the limited contribution of control room abandonment on the total Fire risk.

24-12 HRA-D1 Not Met The joint human error probabilities (JHEPs) PARTIALLY RESOLVED WITH OPEN DOCUMENTATION used to recover the risk are not supported by the documented dependency analysis. This finding remains open as a tracking item to provide a means of confirmation of consistency between the final FPRA human reliability analysis (HRA) report and the recovery file used in the final FPRA quantification model.

Several changes were made during the final FPRA quantification and the final documentation will reflect these.

However, a review of the final data to ensure that no inconsistencies exist is the intent of this open item.

29

Enclosure Attachment 3 Finding Supporting Capability Require- Category Finding Number Disposition for 50.69 ment(s) (CC) Description This is a documentation issue with no impact on this application.

24-14 UNC-A1 Not Met A large number of entries in the uncertainty OPEN.

database lack sufficient information for a complete parametric uncertainty analysis. A review and update, if needed, of the BE probability distributions used in the parametric uncertainty analysis is required. Some potential discrepancies may exist between the database used for the parametric uncertainty, for UNCERT runs, and the probability distributions that should apply to some of the basic events (Bes).

It is anticipated that this review may identify some changes required in the parametric uncertainty analysis but a significant change in the probability distribution of the FPRA total CDF and LERF results is not expected.

The impact will be limited to the parametric uncertainty analysis and will have no impact on the FPRA results and no impact on this application.

25-5 FQ-E1 Not Met Some of the top scenarios are found as being OPEN.

not fully developed which may mask the important contributors to fire risk. This Refinements were performed to address conservative joint Supporting Requirement (SR) requires that human error probability (JHEP) values and to refine the significant contributors be identified in treatment for the containment isolation valves for the accordance with HLR-QU-D. HLR-QU-D6 containment mini-purge lines. These changes reduced the requires that significant contributors be overall risk. With these changes completed no specific need identified and HLR-QU-D7 requires review of for additional refinement is considered to be required.

important components and basic events (Bes) to determine that they make logical sense. This These changes were finalized after the F&O Finding Closure is not possible with overly conservative Review and are reflected in the current FPRA results; scenario models. therefore, they have no impact on this application.

25-9 FQ-E1 Not Met HLR-QU-D7 requires review of importance of PARTIALLY RESOLVED components and basic events (Bes) to determine that they make logical sense. The The documentation and review of results did not include information provided did not meet the intent of importances by accident progression contributors.

providing a review of importance of Importances by basic event (BE) and sequence flags as well components and basic events. as for LERF plant damage states were provided. This is a 30

Enclosure Attachment 3 Finding Supporting Capability Require- Category Finding Number Disposition for 50.69 ment(s) (CC) Description documentation issue only since the importance by accident progression contributors can be extracted from the current model but is not readily available in the current documentation. The importances by BE, sequence flags and plant damage states are the more useful input and were provided and reviewed as part of the model development.

This is a documentation issue with no impact on this application.

25-11 PRM-B2 Not Met Internal events finding AS-B3-01 does not PARTIALLY RESOLVED appear to have been addressed for the FPRA analysis. This finding is partially resolved in internal events model but not yet resolved in the FPRA model. The system notebook describes an acceptable approach for sump clogging based on WCAP-16362-NP. The fault tree modeling is consistent with that approach.

Random failure due to containment sump screen clogging will have minimal impact on the FPRA. The logic from the updated FPIE model will be incorporated in the next revision of the FPRA.

Resolution of this issue will have minimal impact on this application.

25-21 FQ-F1 Not Met The documentation for the relative contribution PARTIALLY RESOLVED of contributors to LERF was not addressed.

Importances by basic event (BE) and sequence flags as well as for LERF plant damage states were provided. This is a documentation issue only since the importance by accident progression contributors can be extracted from the current model but is not readily available in the current documentation. The importances by BE, sequence flags and plant damage states are the more useful input and were provided and reviewed as part of the model development.

This is a documentation issue with no impact on this application.

31

Enclosure Attachment 3 Finding Supporting Capability Require- Category Finding Number Disposition for 50.69 ment(s) (CC) Description 25-22 FQ-F1 Not Met The process used to identify plant damage PARTIALLY RESOLVED states and accident progression contributors was not documented. The process used to identify plant damage states and accident progression contributors was not provided in the notebook. Importances by basic event (BE) and sequence flags as well as for LERF plant damage states were provided. This is a documentation issue only since the importance by accident progression contributors can be extracted from the current model but is not readily available in the current documentation. The importances by BE, sequence flags and plant damage states are the more useful input and were provided and reviewed as part of the model development.

This is a documentation issue with no impact on this application 26-9 IGN-A7 Cat I/II/III During the Peer Review walkdown, two fairly OPEN.

large electrical wall-mounted cabinets with 14 switches were not counted in 11.4C-0, which is Wall mounted panels were screened from the FPRA. Based a risk-significant fire zone. on further review of the requirements of NUREG/CR-6850 and associated dispositions to NFPA 805 Frequently Asked Questions (FAQs), it has been determined that only those panels with four or fewer switches should have been screened (per NUREG/CR-6850, p. 6-18, discussion of Bin 15 ignition frequency).

The impact of this open item is expected to be minimal. The generally simple configuration for a wall mounted panel, although it may have more than four switches, is expected to be offset by the reduction in frequency of all other panels which is expected to include many panels with higher risk scenarios than the wall mounted panels. This change will be incorporated in a future FPRA model update.

The resolution of this open item will have a minimal impact on the FPRA results and a minimal impact on this application.

32

Enclosure Attachment 4 Attachment 4: External Hazards Screening Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron From Braidwood UFSAR From Byron UFSAR Section 3.5.1.6, Aircraft Section 3.5.1.6, Aircraft Hazards, the airports and Hazards, the airports and airways in the vicinity of the airways in the vicinity of site are described in the site are described in Braidwood UFSAR Byron UFSAR Subsection Subsection 2.2.2.5 2.2.2.5 (Reference 16).

(Reference 16).

There are no airports with The airports in the vicinity of projected operations the site do not pose a hazard greater than 500 d2 because (1) there are no movements per year within airports within 5 miles of the 10 miles of the site and site, and (2) there are no greater than 1000 d2 airports with projected movements per year operations greater than 500 outside 10 miles, where d d2 movements per year within is the distance in miles 10 miles of the site and from the site. There are no greater than 1000 d2 low altitude federal airways PS2 movements per year outside within 2 miles of the site.

Aircraft Impact Y 10 miles, where d is the PS4 distance in miles from the For the airports and site. seaplane base within 10 miles of the site (shown in Using a refined method as Byron UFSAR Figure 2.2-described in Braidwood 3), an analysis has been UFSAR Section 3.5.1.6, the performed which shows probability of an aircraft that the probability of an crashing into the safety- aircraft crashing into the related plant structures is safety-related plant 3.8E-08 per year, which is structures is 3.7E-08 per less than the threshold value year.

of 1.0E-07 per year for postulating the design-basis Based on this review, the event. Hazards due to aircraft aircraft impact hazard can crashes thus need not be be considered to be considered in the design of negligible.

the Braidwood Station.

Based on this review, the aircraft impact hazard can be 33

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron considered to be negligible.

The mid-western location of The mid-western location Braidwood station precludes of Byron station precludes Avalanche Y C3 the possibility of a snow the possibility of a snow avalanche. avalanche.

Actions committed to and Actions committed to and completed by Braidwood completed by Byron station in response to station in response to Generic Letter 89-13 (Service Generic Letter 89-13 Water System Problems provide on-going control of Affecting Safety-Related biological hazards. These Equipment) provide on-going controls are described in control of biological hazards. Exelon procedure ER-AA-These controls are described 340, "GL 89-13 Program in Exelon procedure ER-AA- Implementing Procedure".

340, "GL 89-13 Program In addition, station actions C3 Implementing Procedure". In taken in response to INPO Biological Event Y addition, station actions taken SOER 07-2 provide an C5 in response to INPO SOER additional layer of 07-2 (Intake Structure biological hazard Blockage) provide an management.

additional layer of biological hazard management. Based on these actions, the potential impact of Based on these actions, the biological hazard events Is potential impact of biological considered negligible and hazard events is considered is screened from further negligible and is screened consideration.

from further consideration.

The mid-western location of The mid-western location Braidwood station precludes of Byron station precludes Coastal Erosion Y C3 the possibility of coastal the possibility of coastal erosion. erosion.

These effects would take These effects would take place slowly allowing time for place slowly allowing time Drought Y C5 orderly plant reductions, for orderly plant including shutdowns. reductions, including shutdowns.

34

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron The external flooding hazard The external flooding at the site was recently hazard at the site was updated and the flood hazard recently updated and the reevaluation report (FHRR) flood hazard reevaluation was submitted to NRC for report (FHRR) was review on March 12, 2014 submitted to NRC for (Reference 17) and revised review on March 12, 2014 to respond to Request for (Reference 19).

Additional Information (RAI) on June 24, 2015 (Reference By letter dated 18). The results indicate that September 3, 2015 Local intense precipitation (Reference 20), the NRC (LIP) was evaluated and staff concluded that the found to have a maximum reevaluated flood hazard water surface elevation mechanisms for Byron are (WSE) of 601.11 ft. This bounded by the current mechanism does not produce design basis.

a WSE greater than the Subsequently, Byron plant's current licensing basis revised the model used to (CLB) or the building entry develop the local intense way elevation of 602.25 ft. as precipitation (LIP) flood C1 indicated in the UFSAR hazard parameters. The External Flooding Y (Reference 16). revision resulted in minor PS2 differences between the The combined flooding LIP parameters described mechanism effects found that in the external flooding the limiting scenario is an all mitigating strategies season probable maximum assessment (MSA) precipitation (PMP) resulting submitted to NRC on in a probable maximum flood September 30, 2016 (PMF) raising the WSE in the (Reference 21) and the cooling pond adjacent to the reevaluated LIP site. This is combined with parameters described in the 2-year wind speed to the September 3, 2015 estimate wave runup effects letter.

and the maximum WSE of 602.34 ft (Reference 18), Specifically, the revised which is greater than plant model resulted in an building entry way at increase in the maximum elevation 602.25 ft. In order flood elevation due to LIP to provide protection for this from 870.9 feet to 870.94 extreme event, the exterior feet, which exceeds the dike adjacent to the lake plant floor elevation at the screen house was built 2.5 ft. east and southwest sides higher than the other 35

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron locations. The top elevation of the turbine building.

of this dike is 602.5 ft. MSL, which is greater than the The NRC staff assessment WSE. No impacts from this letter of the Byron MSA external flood causing dated October 21, 2016 mechanism are expected. (Reference 22) concluded that this minor increase in maximum LIP elevation is:

1) bounded by an internal flood; and 2) does not adversely impact mitigating strategies equipment.

Since this difference is minor and has no impact on implementation of the Byron station mitigating strategy, the NRC concluded that the flood hazards used in the MSA are equivalent to the design-basis of the facility and suitable for use in the MSA. In addition, Byron Station completed an evaluation which concluded that ingress from the revised LIP flood, with higher flood levels and period of inundation, is bounded by an internal flood and that the plant would be able to shutdown safely.

Wind damage is bounded by Wind damage is bounded tornadoes, and the tornado by tornadoes, and the wind speed corresponding to tornado wind speed C1 Extreme Wind or the 1E-6/yr exceedance corresponding to the 1E-Y Tornado frequency is much less than 6/yr exceedance frequency PS4 the Braidwood design value; is much less than the therefore, damage due to the Byron design value; forces associated with therefore, damage due to extreme winds or tornados the forces associated with 36

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron can be screened. extreme winds or tornados can be screened.

For tornado missiles, a plant-specific TORMIS analysis For tornado missiles, a was performed in accordance plant-specific TORMIS with the guidance described analysis was performed in in the 1983 NRC TORMIS accordance with the Safety Evaluation Report guidance described in the (Reference 23), as clarified 1983 NRC TORMIS Safety by Regulatory Issue Evaluation Report Summary (RIS) 2008-14, (Reference 23), as clarified "Use of TORMIS Computer by Regulatory Issue Code for Assessment of Summary (RIS) 2008-14, Tornado Missile Protection." "Use of TORMIS (Reference 24). Computer Code for Assessment of Tornado The CDF associated with Missile Protection."

tornado missiles is less than (Reference 24).

1E-6/yr.

The CDF associated with tornado missiles is less than 1E-6/yr.

The principal effects of such The principal effects of events (such as freezing fog) such events (such as would be to cause a loss of freezing fog) would be to off-site power and are cause a loss of off-site addressed in the weather- power and are addressed Fog Y C1 related Loss of Offsite Power in the weather-related initiating event in the internal Loss of Offsite Power events PRA model for initiating event in the Braidwood. internal events PRA model for Byron.

The site landscaping and lack The site landscaping and Forest or Range of forestation prevent such lack of forestation prevent Y C3 Fire fires from posing a threat to such fires from posing a Braidwood station. threat to Byron station.

The principal effects of such The principal effects of events would be to cause a such events would be to Frost Y C1 loss of off-site power and are cause a loss of off-site addressed in the weather- power and are addressed related Loss of Offsite Power in the weather-related initiating event in the internal Loss of Offsite Power 37

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron events PRA model for initiating event in the Braidwood. internal events PRA model for Byron.

The principal effects of such The principal effects of events would be to cause a such events would be to loss of off-site power and are cause a loss of off-site C1 addressed in the weather- power and are addressed Hail Y related Loss of Offsite Power in the weather-related C4 initiating event in the internal Loss of Offsite Power events PRA model for initiating event in the Braidwood. internal events PRA model for Byron.

The principal effects of such The principal effects of events would result in such events would result in elevated lake temperatures elevated river and SX which are monitored by Cooling Tower basin station personnel. Should temperatures which are the temperature exceed the monitored by station Technical Specification limit, personnel. Should the an orderly shutdown would temperature exceed the be initiated. Another Technical Specification potential initiating limit, an orderly shutdown consequence would be to would be initiated.

High Summer Y C1 cause a loss of off-site Another potential initiating Temperature power. These effects would consequence would be to take place slowly allowing cause a loss of off-site time for orderly plant power. These effects reductions, including would take place slowly shutdowns. At worst, the allowing time for orderly loss of off-site power events plant reductions, including would be subsumed into the shutdowns. At worst, the base PRA model results. loss of off-site power events would be subsumed into the base PRA model results.

The mid-western location of The mid-western location Braidwood station precludes of Byron station precludes High Tide, Lake the possibility of a high tide the possibility of a high tide Level, or River Y C3 condition. condition.

Stage High lake effects would take High river effects would place slowly allowing time for have negligible impact to 38

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron orderly plant reductions, the plant due to the including shutdowns. installation of cooling towers being the ultimate heat sink.

The mid-western location of The mid-western location Braidwood station precludes of Byron station precludes the possibility of a hurricane. the possibility of a In addition, hurricanes would hurricane. In addition, Hurricane Y C4 be covered under Extreme hurricanes would be Wind or Tornado and Intense covered under Extreme Precipitation. Wind or Tornado and Intense Precipitation.

The principal effects of such The principal effects of events would be to cause a such events would be to C1 loss of off-site power and are cause a loss of off-site addressed in the weather- power and are addressed Ice Cover Y C3 related Loss of Offsite Power in the weather-related initiating event in the internal Loss of Offsite Power events PRA model for initiating event in the Braidwood. internal events PRA model for Byron.

Joliet Arsenal is located eight There are no military miles from the site. This facilities within 10 miles of facility has largely ceased the site. Industrial activities except for those of a manufacturing facilities custodial nature. Shipments have also been evaluated of explosive materials have and determined to not essentially been terminated. have an impact to the Byron site as discussed in There are no manufacturers the Byron UFSAR, section Industrial or C1 of hazardous materials 2.2.3 (Reference 16).

Military Facility Y located within five miles of Accident C3 the site. Hazardous The evaluation of chemical chemicals used and/or stored hazards from military or by manufacturers within five industrial facilities is miles of the plant were also performed in accordance evaluated and determined to with Technical either screen from further Specification 5.5.18 evaluation or were (Reference 26).

determined to meet the acceptance criteria associated with Control 39

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron Room operator protection as discussed in the Braidwood UFSAR, section 2.2.3 (Reference 16).

The evaluation of chemical hazards from military or industrial facilities is performed in accordance with Technical Specification 5.5.18 (Reference 25).

The Braidwood Internal The Byron Internal Events Events PRA includes PRA includes evaluation of Internal Flooding N None evaluation of risk from risk from internal flooding internal flooding events. events.

The Braidwood internal Fire The Byron internal Fire Internal Fire N None PRA addresses risk from PRA addresses risk from internal fire events. internal fire events.

The mid-western location of The mid-western location Braidwood station precludes of Byron station precludes the possibility of a landslide. the possibility of a Landslide Y C3 Not applicable to the site landslide. Not applicable to because of topography. the site because of topography.

Lightning strikes are not Lightning strikes are not uncommon in nuclear plant uncommon in nuclear plant experience. They can result experience. They can in losses of off-site power or result in losses of off-site surges in instrumentation power or surges in output if grounding is not fully instrumentation output if effective. The latter events grounding is not fully Lightning Y C1 often lead to reactor trips. effective. The latter events Both results are incorporated often lead to reactor trips.

into the Braidwood internal Both results are events model through the incorporated into the Byron incorporation of generic and internal events model plant specific data. through the incorporation of generic and plant specific data.

40

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron These effects would take These effects would take place slowly allowing time for place slowly allowing time Low Lake Level Y C5 orderly plant reductions, for orderly plant or River Stage including shutdowns. reductions, including shutdowns.

The principal effects of such The principal effects of events would be to cause a such events would be to loss of off-site power. These cause a loss of off-site effects would take place power. These effects slowly allowing time for would take place slowly C1 Low Winter orderly plant reductions, allowing time for orderly Y

Temperature including shutdowns. At plant reductions, including C5 worst, the loss of off-site shutdowns. At worst, the power events would be loss of off-site power subsumed into the base PRA events would be model results. subsumed into the base PRA model results.

The frequency of a meteorite The frequency of a or satellite strike is judged to meteorite or satellite strike Meteorite or be very low such that the risk is judged to be very low Y PS4 Satellite Impact impact from such events such that the risk impact insignificant. from such events insignificant.

There are six natural gas The nearest pipeline to the pipelines and four other site is 2.5 miles away and petro-chemical pipelines is only a 3-inch diameter within five miles of the site. pipe. There is no There is no significant hazard significant hazard to the from toxic releases or site from these events.

explosions involving these pipelines that could interact Chemical Hazards Pipeline Accident Y C1 with the plant. transported using pipelines that are located in the Chemical Hazards vicinity of the plant are transported using pipelines analyzed in accordance that are located in the vicinity with Technical of the plant are analyzed in Specification 5.5.18 accordance with Technical (Reference 26).

Specification 5.5.18 (Reference 25).

41

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron Chlorination of water systems Chlorination of water is performed using a systems is performed hypochlorite system. No using a hypochlorite chlorine gas is stored on-site. system. No chlorine gas is Various acids and caustics stored on-site. Various are stored on-site but pose acids and caustics are no hazard to the plant. stored on-site but pose no Release of hazard to the plant.

Chemicals in Y PS2 Chemical Hazards stored and Onsite Storage transported in the vicinity of Chemical Hazards stored the plant are analyzed in and transported in the accordance with Technical vicinity of the plant are Specification 5.5.18 analyzed in accordance (Reference 25). with Technical Specification 5.5.18 (Reference 26).

There is no historical or Due to the great width of topographical evidence the Rock River and the indicating that flow in the relatively flat surrounding Kankakee River can be terrain, there is little diverted away from its possibility that rock falls, present course. The river is ice jams or subsidence wide, and there are no deeply could completely divert the incised gorges upstream flow away from the where landslides could makeup water intake Refer River Diversion Y C3 entirely cut off river flow. to Byron UFSAR Section Upstream ice jams will not 2.4.9(Reference 16).

divert flow completely, since they do not prevent overbank or subsurface flow. Refer to Braidwood UFSAR Section 2.4.9 (Reference 16).

The mid-western location of The mid-western location Braidwood station prevents of Byron station prevents C1 sandstorms. More common sandstorms. More Sand or Dust Y wind-borne dirt can occur but common wind-borne dirt Storm C5 poses no significant risk to can occur but poses no Braidwood station given the significant risk to Byron robust structures and station given the robust protective features of the structures and protective 42

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron plant. features of the plant.

Seiche was found to not be Seiche was found to not C1 an applicable external be an applicable external Seiche Y flooding mechanism in flooding mechanism in C3 Reference 17. Reference 19.

See information in Section See information in Section Seismic Activity N None 3.2.3 of this application. 3.2.3 of this application.

Snow cover is included as an Snow cover is included as C1 input to the probable an input to the probable Snow Y maximum flood (PMF) WSE maximum flood (PMF)

C4 calculations (Reference 17). WSE calculations (Reference 19).

Based on the discussions Based on the discussions and conclusions reached in and conclusions reached the Braidwood UFSAR in the Byron UFSAR section 2.5, "Geology, section 2.5, "Geology, Seismology, And Seismology, And C1 Soil Shrink-Swell Geotechnical Engineering" Geotechnical Engineering" Y

Consolidation (Reference 27), the impact (Reference 27, the impact C5 from soil shrink or swell from soil shrink or swell (subsidence or uplift) is (subsidence or uplift) is expected to be negligible and expected to be negligible can be screened from further and can be screened from evaluation. further evaluation.

The mid-western location of The mid-western location Braidwood station precludes of Byron station precludes Storm Surge Y C3 the possibility of a sea level the possibility of a sea driven storm surge. level driven storm surge.

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

43

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron Chemical hazards stored and Railroad track approaches transported in the vicinity of no closer than four miles to the plant are analyzed in the plant site. There is no accordance with Braidwood heavy traffic in bulk Technical Specification hazardous materials 5.5.18 (Reference 25) and capable of impacting the Regulatory Guides 1.78, 1.95 site.

and 1.196 to ensure Control Room habitability is The Rock River is not maintained. The latest navigable to barge traffic in analysis was performed in the area of the plant site.

2014.

C3 Chemical Hazards that The analysis concluded that may result from being C4 toxic chemicals transported transported on local roads Transportation or stored within the vicinity of that are located in the Y PS2 vicinity of the plant are Accident the plant do not pose a threat that requires the use of analyzed in accordance PS4 instrumentation. with Byron Technical Specification 5.5.18 The Kankakee River is not (Reference 26).

navigable to barge traffic in the area of the plant site.

Chemical Hazards that may result from being transported on local roads that are located in the vicinity of the plant are analyzed in accordance with Braidwood Technical Specification 5.5.18 (Reference 25).

The mid-western location of The mid-western location Tsunami Y C3 Braidwood station precludes of Byron station precludes the possibility of a tsunami. the possibility of a tsunami.

As noted in section 10.2.3 of As noted in section 10.2.3 the Braidwood UFSAR of the Byron UFSAR Turbine- PS2 (Reference 16, the potential (Reference 16), the Generated Y for turbine generated missiles potential for turbine Missiles PS4 is managed through an generated missiles is ongoing station program to managed through an monitor turbine performance ongoing station program to and integrity. At each monitor turbine 44

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron refueling outage, a performance and integrity.

calculation for total unit At each refueling outage, a missile generation probability calculation for total unit is made. To make this missile generation calculation the operational probability is made. To hours on each of the low make this calculation the pressure turbines is operational hours on each gathered. The missile of the low pressure generation probability for turbines is gathered. The each of the LP turbine rotors missile generation is taken from the individual probability for each of the rotors graph based on the LP turbine rotors is taken operational time on the rotor. from the individual rotors The values for the three graph based on the rotors are then added operational time on the together to determine the rotor. The values for the current missile generation three rotors are then probability for the unit. This added together to value must be below 1.0E-05 determine the current to allow loading the turbine missile generation and bringing the unit on line. probability for the unit. This value must be below 1.0E-Based on this ongoing 05 to allow loading the management of the potential turbine and bringing the for turbine-generated unit on line.

missiles, including a performance related Based on this ongoing threshold for operation of the management of the turbine, this hazard can be potential for turbine-considered negligible and generated missiles, screened from further including a performance evaluation. related threshold for operation of the turbine, this hazard can be considered negligible and screened from further evaluation.

Not applicable to Braidwood Not applicable to Byron Volcanic Activity Y C3 Station Station 45

Enclosure Attachment 4 Screening Result External Hazard Comment Screening Screened?

Criterion (Y/N)

(Note a)

Braidwood Byron Waves are addressed as part Waves are addressed as of the combined-effects part of the combined-flooding in Reference 17, effects flooding in Flood Hazard Reevaluation Reference 19, Flood C3 Report (FHRR). It is shown Hazard Reevaluation Waves Y that waves will not challenge Report (FHRR). It is C4 plant grade of the finished shown that waves will not floor elevation of the power challenge plant grade of block. the finished floor elevation of the power block.

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

46

Enclosure Attachment 5 Attachment 5: Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source Comments Initial Preliminary C1. Event damage potential is NUREG/CR-2300 Screening < events for which plant is and ASME/ANS designed. Standard RA-Sa-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 it. and ASME/ANS Standard RA-Sa-2009 C4. Event is included in the NUREG/CR-2300 Not used to screen.

definition of another event. and ASME/ANS Used only to include Standard RA-Sa- within another 2009 event.

C5. Event develops slowly, ASME/ANS Standard allowing adequate time to RA-Sa-2009 eliminate or mitigate the threat.

Progressive PS1. Design basis hazard ASME/ANS Standard Screening cannot cause a core damage RA-Sa-2009 accident.

PS2. Design basis for the NUREG-1407 and event meets the criteria in the ASME/ANS Standard NRC 1975 Standard Review RA-Sa-2009 Plan (SRP).

PS3. Design basis event mean NUREG-1407 as frequency is < 1E-5/y and the modified in mean conditional core damage ASME/ANS Standard probability is < 0.1. RA-Sa-2009 PS4. Bounding mean CDF is < NUREG-1407 and 1E-6/y. ASME/ANS Standard RA-Sa-2009 Detailed PRA Screening not successful. PRA NUREG-1407 and needs to meet requirements in ASME/ANS Standard the ASME/ANS PRA Standard. RA-Sa-2009 47

Enclosure Attachment 6 Attachment 6: Disposition of Key Assumptions/Sources of Uncertainty Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition LOOP Initiating Event Frequency: SSCs that support The overall approach for the LOOP scenarios LOOP frequency and fail to recover probabilities utilized is The initiating event analysis develops consistent with industry practice.

frequencies for LOOP and DLOOP Therefore, this does not events based on the data in represent a key source of NUREG/CR-6890 (Reference 28) uncertainty for the 50.69 (updated through 2013). LOOP types application.

include plant-centered, switchyard-centered, grid-related, and severe weather events, and all are partitioned into LOOP and DLOOP events. The NUREG provides plant-specific values applicable to both sites using supplemental data through 2013.

Failure to recover probabilities for SSCs that support The LOOP frequencies utilized in LOOP: LOOP scenarios the model are based on NUREG/CR-6890 (Reference 28) as updated with data through The industry wide data in NUREG/CR-2013. Therefore, this does not 6890 (Reference 28) (updated through represent a key source of 2013) is utilized to develop the failure uncertainty for the 50.69 to recover probabilities for the four application.

LOOP categories. The industry wide recovery data is applicable to both sites and is acceptable for the base case analysis.

48

Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Grid stability after a reactor trip: SCCs that support The consequential LOOP LOOP scenarios probabilities utilized provide a reasonably realistic modeling. As The consequential LOOP failure such, this does not represent a probabilities are based on EPRI and significant source of model NRC evaluations with different values uncertainty in this application.

following a reactor trip or LOCA Therefore, this does not represent (References 28, 29). The use of a key source of uncertainty in the generic data for consequential LOOP 50.69 application.

events is assumed to be applicable for both sites. The consequential LOOP events are assumed to be dual-unit and are distributed among grid-related, plant-centered, or switchyard-centered based on data from Reference 30.

Offsite power restoration: SCCs that support The LOOP recovery probabilities LOOP scenarios are realistic with slight conservative bias on the recovery Restoration is possible as the times. This does not represent a switchyard has its own 125V DC significant source of model distribution system to provide breaker uncertainty in this application.

and transformer control power. When Therefore, this does not represent offsite power is available at the a key source of uncertainty in the switchyard, then power is available to 50.69 application.

charge the batteries needed for breaker control to align power to the site. The specific failure modes of the offsite restoration are implicitly included via the use of the generic LOOP recovery probabilities.

49

Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Support System Initiating Events SSCs that support Realistic with slight conservative (SSIEs): SSIEs. bias because MTTR is typically less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This does not represent a significant source of Support System Initiating Event fault model uncertainty in this trees are developed for loss of application. Therefore, this does Component Cooling Water (CC), loss not represent a key source of of Service Water (SX), and loss of uncertainty in the 50.69 Non-Essential Service Water (WS).

application.

The loss of support system success criteria are developed consistent with the post-trip configuration requirements (e.g. 1 of 2 SX pumps) and mission time requirements (i.e. 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Mean Time to Repair (MTTR) assumed consistent with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mitigation mission time).

Support System Initiating Events SSCs that support Slight conservative bias treatment (SSIEs): SSIEs. since alpha factors are known to be high when utilized in an annualized fashion and compared Increasing use of plant-specific models to plant-specific experience. This for support system initiators (e.g. loss does not represent a key source of of SX, CC, or Instrument Air (IA), and uncertainty in the 50.69 loss of AC or DC buses) have led to application.

inconsistencies in approaches across the industry. The common cause failure (CCF) for the fail-to-run terms is based on annualized mission times using generic alpha factors, but with plant-specific information for the independent failure rate. The use of the generic alpha factors based on industry wide experience is applicable for the site.

50

Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Support System Initiating Events CC and SX SSCs Slight conservative treatment (SSIEs): that could contribute since credit for recovery beyond to SSIEs. system failure could reduce the baseline CDF and LERF risk Modeling of recovery to prevent metrics. This does not represent a support system initiating events is key source of uncertainty in the limited to procedurally-directed 50.69 application.

alignments of standby equipment given failure of running equipment, if such alignments can be accomplished prior to loss of the support system. No additional credit for recovery beyond system failure is modeled.

LOCA initiating event frequencies: SSCs that support The LOCA frequency values Safety Injection (SI) represent realistic treatment based on accepted industry data The Large and Medium LOCA initiating sources. This does not represent event frequencies are based on failure a key source of uncertainty in the probabilities from NUREG/CR-1829 50.69 application.

(Reference 31) (interpolated for plant-specific LOCA definitions). Small Non-Isolable LOCA initiating event frequencies are based on failure probabilities in NUREG/CR-6928 (Reference 30), and includes both the pipe break frequency and spontaneous reactor coolant pump seal failure.

Small isolable LOCA initiating event frequencies due to stuck-open PORVs are calculated directly using NRC data.

Operation of equipment after battery SSCs supporting No credit for equipment operation depletion: scenarios that after battery depletion may require DC power represent a slight conservative bias. This does not represent a No credit is taken for continued key source of uncertainty in the operation of any systems without DC 50.69 application.

power that normally require DC power for operation. This includes steam generator (SG) level control.

51

Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition RCP seal LOCA treatment: SSCs that support The operator action timing seal LOCA accident assumptions are based on the scenarios WOG 2000 consensus model and The assumed timing and magnitude of Shutdown Seal model. Therefore, RCP seal LOCAs given a loss of seal this does not represent a key injection and thermal barrier cooling can source of uncertainty for the 50.69 have a substantial influence on the risk application.

profile. The WOG 2000 consensus model (Reference 32) has been implemented, along with the model for the Shutdown Seals (Reference 33).

Battery life calculations: SSCs involved in The modeling is realistic given the scenarios in which relatively long battery life without DC power is recharge. This may be slightly Design basis calculations indicate that required. conservative for SBO scenarios

~8 hours of battery life is available and does not represent a key depending on scenario specifics. Credit source of model uncertainty in the for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is utilized in the model for 50.69 application.

most scenarios without chargers available. Because the SBO coping time is set at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is used in LOOP power-recovery calculations.

Number of PORVs required for bleed SSCs supporting The modeling is realistic and does and feed: bleed and feed not represent a key source of scenarios model uncertainty in the 50.69 application.

Plant-specific success criteria calculations have been performed using MAAP to determine the number of PORVs required to open (and timing of opening) for successful bleed and feed cooling. This has been done as a function of the ECCS pumps available.

Results show that a single PORV opening represents bleed and feed success for the condition where a charging pump is running. Success is also credited where two PORVs and a single safety injection pump is running.

The appropriate success criteria (i.e., 2 PORVs open or 1 PORV opens) are applied depending on the available ECCS pumps for the scenario being modeled.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Impact of failure of pressure relief: SSCs supporting The approach taken is consistent scenarios where with that used in other PWR pressure relief via PRAs. The potential impact on For general transients with reactor trip, PORVs and SRVs CDF due to not explicitly modeling the PORVs provide pressure relief if are required the possibility of overpressure for needed, and the likelihood of a safety non-ATWS events is not relief valve challenge is sufficiently small significant. Therefore, this does that explicit modeling is not required. For not represent a key source of general transients (non-ATWS), it is uncertainty in the 50.69 commonly assumed (and evident in application.

success criteria calculations) that opening of any 1 of the 2 PORVs and 3 SRVs is sufficient to preserve RPV integrity below ASME Service Level C.

Impact of failure of pressure relief: SSC supporting Slight conservative bias treatment ATWS scenarios for in assumption that overpressure which PORVs or failure in ATWS cases goes For ATWS scenarios, the number of SRVs are required directly to core damage, since PORVs and/or SRVs is a function of core RCS piping failures in most reactivity, available AFW capacity, and locations would not result in other parameters as specified in the LOCAs in excess of ECCS WOG ATWS model (Reference 34). Per capability. However, the modeling the WOG model, there may be brief is in accordance with an industry periods of time in which all available recognized model and thus does pressure relief is not adequate to not represent a key source of maintain RCS pressure below the ASME uncertainty in the 50.69 Service Level C pressure. Failure to application.

maintain RCS pressure below the ASME Service Level C pressure is modeled (non-mechanistically) in the PRA as leading to vessel failure and core damage.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Operability of equipment in beyond SSCs supporting Not explicitly accounting for the design basis environments: scenarios for which limited time frames when two DG DGs are required HVAC fans may be required to provide DG cooling is a potential Generally, credit for operation of systems slight non-conservatism. Not beyond their design-basis environment is modeling the proceduralized not taken. However, each DG requires restoration of DG ventilation is a ventilation to operate successfully. This potential conservatism. Given that dependency is modeled to include a ventilation dependency is suction and exhaust dampers (including modeled, this does not represent a CCF terms as applicable) and fan fail-to-key source of uncertainty in the start and fail-to-run terms. There may be 50.69 application.

limited times when 2 DG HVAC fans are required to provide cooling, but this is assumed not to be a significant dependency and is not modeled. Station procedures provide guidance for the emergency restoration of the DG ventilation and for the use of portable ventilation to maintain DG temperatures acceptable, but this option is not credited in the PRA model.

Operability of equipment in beyond SSCs supporting Realistic with slight conservative design basis environments: scenarios for which bias on assumed battery life time.

batteries are This does not represent a key required. source of uncertainty in the 50.69 Generally, credit for operation of systems application.

beyond their design-basis environment is not taken. However, given the typical conservatisms associated with the design-basis battery calculations, and the relatively long battery life, explicit representation of load shedding is not assumed to be required to obtain the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> battery life times for non-SBO scenarios.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Widespread LOOP effects: SSCs supporting Lack of credit for TSC actions is LOOP scenarios slightly conservative.

Credit for TSC actions is not currently Lack of explicit consideration of used for cognitive error contributions in increased stress due to a HRA due to its significant uncertainty. widespread LOOP could be The TSC is implicitly used for execution slightly non-conservative, but its recovery for long-term actions, but this is effect is expected to be low based not directly affected by a widespread on the low likelihood of the event LOOP. and the already present stress during a normal LOOP event.

Increased stress due to communication challenges is recognized as a source of Therefore, this does not represent model uncertainty and is not explicitly a key source of uncertainty in the included in the LOOP-related HEP 50.69 application.

calculations. However, as directed by NEI 00-04, internal events human error basic events are increased to their 95th percentile and also decreased to their 5th percentile values as part of the required 50.69 PRA categorization sensitivity cases.

Piping failure mode: SSCs that support Considered an industry good Internal Flood practice approach, but is not yet a scenarios consensus model approach. This The Internal flood analysis and initiating is not a source of significant model event frequencies for spray, flood, and uncertainty given that a major flood scenarios are developed recognized methodology has been consistent with the EPRI methodology applied using plant-specific piping (Reference 35). The flooding analysis is data. Therefore, this does not integrated into the internal events at represent a key source of power model. The use of generic flood uncertainty in the 50.69 frequencies with plant-specific estimates application.

of pipe lengths is suitable for representation of the flood frequencies at the site.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Piping failure mode: SSCs that support Realistic since such a low flowrate Internal Flood would not affect most systems scenarios needed to mitigate an accident.

The Internal flood analysis and initiating Therefore, this does not represent event frequencies for spray, flood, and a key source of uncertainty in the major flood scenarios are developed 50.69 application consistent with the EPRI methodology (Reference 33). The flooding analysis is integrated into the internal events at Assuming major flood sources power model. Spray flood scenarios with greater than 100 gpm totally less than 100 GPM flow do not totally disable the system they arise from disable the system they arise from. is conservative in that the system may not be totally disabled in all cases. Therefore, this does not Major flood sources greater than 100 represent a key source of gpm are assumed to totally disable the uncertainty in the 50.69 system they arise from.

application.

Core melt arrest in-vessel: SSCs that support Conservative bias treatment in that LERF scenarios in-vessel core melt arrest might be feasible in some scenarios. This In-vessel recovery of the molten core by does not represent a key source of flooding of the reactor cavity and heat uncertainty in the 50.69 transfer through the vessel is not application.

credited in the Level 2 analysis.

Uncertainties due to the ability of the cavity to be flooded to sufficient depth, the effects of lower head insulation and instrument penetrations, and the ability to achieve sufficient heat transfer to prevent vessel failure make in-vessel recovery difficult to justify.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Thermally induced failure of hot leg/SG SSCs that support Approach is consistent with recent tubes: LERF scenarios industry approaches and appropriate for determination of LERF. Therefore, this does not The approach follows Simplified Level 2 represent a key source of Modeling Guidelines, WCAP-16341-P uncertainty in the 50.69 (Reference 33), which many plants are application.

currently using as a basis for updated Level 2 analyses. This WCAP provides a common, standardized method for PWRs with large dry containments to produce an analysis that generally meets capability category II of the ASME PRA Standard (Reference 36). The guidance particularly addresses the latest understanding for induced steam generator tube ruptures and other Level 2 issues.

Vessel failure mode: SSCs that support Approach is appropriate for LERF scenarios determination of LERF. Therefore, RPV catastrophic failure leading to early this does not represent a key containment failure via missiles is source of uncertainty in the 50.69 extremely unlikely based on studies application.

documented in NUREG-1524 (Reference 37)

No explicit impact on model, since failure mode is assumed to be a small contributor to the overall likelihood of containment failure.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Vessel failure mode: SSCs that support Approach is consistent with LERF scenarios general industry approaches and appropriate for determination of The approach follows Simplified Level 2 LERF. Therefore, this does not Modeling Guidelines, WCAP-16341-P represent a key source of (Reference 33), which many plants are uncertainty in the 50.69 currently using as a basis for updated application.

Level 2 analyses. This WCAP provides a common, standardized method for PWRs with large dry containments to produce an analysis that generally meets capability category II of the ASME PRA Standard (Reference 36). The guidance particularly addresses the latest understanding for direct containment heating and other Level 2 issues.

Vessel failure mode: SSCs that support Approach is appropriate for LERF scenarios determination of LERF for Braidwood and Byron. Therefore, Ex-vessel steam explosions noted as this does not represent a key very unlikely based on reference to source of uncertainty in the 50.69 generic studies. Based on WCAP-16341-application.

P (Reference 33), this is a greater issue for free-standing reactor cavities (as opposed to excavated cavities).

Because this is an excavated cavity, steam explosions do not pose a failure mechanism for early containment failure.

Ex-vessel cooling of lower head: SSCs that support No credit for ex-vessel cooling of LERF scenarios the lower head represents a realistic treatment with a slight No credit for ex-vessel cooling.

conservative bias. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Core debris contact with containment: SSCs that support The modeling reflects the plant LERF scenarios design. Therefore this does not represent a key source of This is not considered as an early failure uncertainty in the 50.69 mechanism because there is no direct application.

path for core debris to contact the containment shell.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Containment integrity following vessel SSCs that support Vessel rupture frequency is on the rupture event: LERF scenarios order of E-7, i.e., very small, such that potential impact on LERF is also small. Containment integrity Vessel rupture sequence is assumed to following vessel rupture is not result in concurrent containment therefore not identified as a failure coincident with the vessel rupture.

candidate source of model uncertainty.

Condensate Storage Tank Inventory: SSCs that support Lack of explicit modeling of the scenarios where various sources of decay heat secondary side decay removal should not have a The inventory in the Condensate Storage hear removal is significant impact on CDF or Tank (CST) is normally not sufficient for required. LERF. The CST is the preferred the full 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time modeled in source of water for AF, but service the PRA. CST inventory is adequate for water is the safety related suction approximately 16 hrs; an assumption is source for AF. Therefore, the made that the likelihood of failing to duration the CST is able to provide either refill the CST or align to shutdown suction for AF pumps is not a key cooling within this extended time frame is source of model uncertainty for very small. The service water (SX) applications.

system is the safety related suction source for auxiliary feedwater (AF) pumps. The suction source for AF pumps automatically switches to SX on low CST suction pressure.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Human Error Probabilities (HEPs): SSCs that use an Sensitivity cases for the base Operator Action basic internal events PRA (HEP values Detailed evaluations of HEPs are event as a surrogate of 5th or 95th percentile value performed for the risk significant human for a modeled HEPs) show that the results are failure events (HFEs) using industry component. somewhat sensitive to HRA model consensus methods. Mean values are and parameter values.

used for the modeled HEPs. Uncertainty The BY/BW PRA model is based associated with the mean values can on industry consensus modeling have an impact on CDF and LERF approaches for its HEP results. calculations, so this is not considered a significant source of epistemic uncertainty.

However, as directed by NEI 00-04, internal events human error basic events are increased to their 95th percentile and also decreased to their 5th percentile values as part of the required 50.69 PRA categorization sensitivity cases. These results are capable of driving a component and respective functions HSS and therefore the uncertainty of the PRA modeled HEPs are accounted for in the 50.69 application.

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Enclosure Attachment 6 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Model Sensitivity and Assumption / Uncertainty 50.69 Impact Disposition Dependent HEPs: Potentially all SSCs The BY/BW PRA model is based evaluated during on industry consensus modeling Dependent HEP values are developed approaches for its dependent HEP for significant combinations of HEPs that 50.69 categorization identification and calculations, so have been demonstrated to appear this is not considered a significant together in the same cutsets. source of epistemic uncertainty.

However, as directed by NEI 00-04, internal events human error basic events are increased to their 95th percentile and also decreased to their 5th percentile values as part of the required 50.69 PRA categorization sensitivity cases.

These results are capable of driving a component and respective functions HSS and therefore the uncertainty of the PRA modeled HEPs are accounted for in the 50.69 application.

Common Cause Failure: Potentially all SSCs The BY/BW PRA model is based evaluated during on industry consensus modeling Common cause failure values are approaches for its common cause developed using available industry data. 50.69 categorization identification and value determination, so this is not considered a significant source of epistemic uncertainty.

Therefore, this does not represent a key source of uncertainty and will not be an issue for 50.69.

Inter-system LOCA (ISLOCA): SSCs that support The values utilized provide a LERF scenarios reasonable best-estimate approach, will have only a minor The detailed ISLOCA analysis includes impact on the 50.69 and do not the relevant considerations listed in IE-represent a key source of C12 of the ASME/ANS PRA Standard uncertainty.

(Reference 36) and accounts for common cause failures and captures likelihood of different piping failure modes.

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Enclosure Attachment 6 The table below describes the Fire PRA sources of model uncertainty and their impact.

Fire PRA Sources of Model Uncertainty Description Sources of Uncertainty Disposition Analysis This task poses a limited source of The multi-compartment analysis further boundary and uncertainty beyond the credit taken for reduces this uncertainty by addressing partitioning boundaries and partitions. the potential impact of failure of partition elements on quantification.

Component This task is associated with the The only uncertainty associated with Selection development of the linkage between this task is related to the identification safe shutdown (SSD) analysis of all credible MSO scenarios. This component/cable data to fault tree source of uncertainty is reduced as a failure modes. Also included in this result of multiple overlapping tasks task is the development and including the MSO expert panel and incorporation of multiple spurious industry owners group identification of operation (MSO) scenarios not applicable MSOs. Additional internal addressed in the internal events model reviews of analysis results further fault tree. reduce the uncertainty associated with this task.

Cable Selection No treatment of uncertainty is typically The limited number of components required for this task beyond the without available cable routing (most understanding of the cable selection active components credited in the fire approach (i.e., mapping an active basic PRA have their cables routed) as well event to a passive component for as the crediting by exclusion of these which power cables were not selected). components (where justified) helps to Additionally, PRA credited components reduce unnecessary conservatism. A for which cable routing information was sensitivity analysis for this not provided represent a source of conservatism is addressed in the uncertainty (conservatism) in that these FPRA uncertainty analysis.

components are assumed failed unnecessarily.

Qualitative Qualitative screening was not In the event that a structure which Screening performed; however, structures were could lead to a plant trip was excluded eliminated from the global analysis incorrectly, its contribution to CDF boundary and ignition sources deemed would be small (with a CCDP to have no impact on the fire PRA were commensurate with base risk) and excluded from the quantification based would likely be offset by inclusion of on qualitative screening criteria. The the additional ignition sources and the only criterion subject to uncertainty is resulting reduction of other scenario the potential for plant trip. frequencies. A similar argument can be made for ignition sources for which scenario development was deemed unnecessary.

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Enclosure Attachment 6 Fire PRA Sources of Model Uncertainty Description Sources of Uncertainty Disposition Fire-Induced A reactor trip is assumed as the FPIE and fire PRA peer reviews Risk Model initiating event for all quantification. (including the F&O resolution process)

This is somewhat conservative since and internal assessments are useful in not all fires postulated will result in a exercising the model and identifying plant trip. weaknesses with respect to this assumption. The use of split fractions to define the probability of a reactor trip in areas where a fire is not likely to cause a reactor trip tends to reduce the conservatism associated with this assumption. This split fraction has been added directly to the Byron and Braidwood models, and uncertainty distribution is applied to the uncertainty factor and is included as part of the parametric uncertainty.

Fire Ignition Ignition source counting is an area with The fire PRA utilizes the bin Frequency inherent uncertainty; however, the frequencies from NUREG/CR-2169 results are not particularly sensitive to (Reference 39), which represents the changes in ignition source counts. The most current approved bin frequencies.

primary source of uncertainty for this As such, some of the inherent task is associated with the frequency conservatism associated with bin values from NUREG/CR-6850 frequencies from NUREG/CR-6850 (Reference 38) which result in (Reference 38) has been removed. A uncertainty due to variability among parametric uncertainty analysis using plants along with some significant the Monte Carlo method is provided in conservatism in defining the section 4.1.1 of the FPRA uncertainty frequencies, and their associated heat and sensitivity notebook (Reference release rates, based on limited fire 40).

events and fire test data.

Quantitative Other than screening out potentially Quantitative screening is limited to Screening risk significant scenarios (ignition refraining from further scenario sources), there is no uncertainty from refinement of those scenarios with a this task on the fire PRA results. resulting CDF/LERF below the screening threshold. All of the results were retained in the cumulative CDF/LERF, therefore, no uncertainty was introduced as a result of this task.

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Enclosure Attachment 6 Fire PRA Sources of Model Uncertainty Description Sources of Uncertainty Disposition Scoping Fire The approach taken for this task The employment of generic fire Modeling included: 1) the use of generic fire modeling solutions did not introduce modeling treatments in lieu of any significant conservatism. Detailed conservative scoping analysis fire modeling was only applied where techniques and 2) limited detailed fire the reduction in conservatism was modeling was performed to refine the likely to have a measurable impact.

scenarios developed using the generic The NUREG-2178 (Reference 41) heat fire modeling solutions. The primary release rates are used and they conservatism introduced by this task is constitute the most recent available associated with the heat release rates heat release rate data. Some specified in NUREG/CR 6850 conservatism in this data is believed to (Reference 38). exist. The level of conservatism cannot be quantified at this time.

Detailed Circuit Uncertainty considerations for the No specific uncertainty is associated Failure Analysis circuit failure analysis task are with the performance of the circuit addressed via the use of circuit failure analysis.

mode probability factors in Task 10.

No specific uncertainty is associated with the performance of the circuit analysis.

Circuit Failure The uncertainty associated with the Circuit failure mode likelihood analysis Mode Likelihood applied conditional failure probabilities was generally limited to those Analysis poses competing considerations components where spurious operation primarily due to the assumption that all was expected to be a large contributor spurious operations occur at the same to total risk. The assumption that all time. The hot short probability and the spurious operations (hot shorts) occur hot short duration factors defined in at the same time results in a significant NUREG/CR-7150 (Reference 42) are conservatism in the analysis but is not considered best available data. easily assessed with respect to the impact on the overall results.

Detailed Fire The primary uncertainty in this task is Detailed fire modeling was performed Modeling in the area of target failure only on those scenarios which probabilities. Conservative heat otherwise would have been notable release rates may result in additional risk contributors and only where target damage. Non-conservative heat removal of conservatism in the generic release rates would have an opposite fire modeling solution was likely to effect. provide benefit either via a smaller Credit for fire brigade response and zone of influence or to allow credit for detection are considered bounding automatic suppression. Fire modeling given that the data used for manual was used to evaluate the time to non-suppression probability is based abandonment for control room fire on extinguishment of a fire and not scenarios for a range of fire heat control (prevention of further spread) of release rates. The analysis a fire. methodology conservatism is primarily associated with conservatism in the 64

Enclosure Attachment 6 Fire PRA Sources of Model Uncertainty Description Sources of Uncertainty Disposition heat release rates and manual non-suppression probability data specified in NUREG-2178 (Reference 41).

Post-Fire Human error probabilities represent a Conservative HEP adjustments were Human potentially large uncertainty for the fire made to the nominal HEP values used Reliability PRA given the importance of human in the FPIE model then revisited to Analysis actions in the base model. Since many address unique fire considerations. A of the HEP values were adjusted for detailed analysis was performed for all fire, the joint dependency multipliers fire specific HFEs. A floor value of 1E-developed for the FPIE model also 06 was applied for all combinations.

represent a potential for introducing a Uncertainty in HEP values is degree of conservatism. propagated through the parametric uncertainty analysis.

Seismic-Fire Since this is a qualitative evaluation, Seismic fire interaction has no impact Interactions there is no quantitative impact with on fire risk quantification.

Assessment respect to the uncertainty of this task.

Fire Risk As the culmination of other tasks, most See Quantification Notebook Quantification of the uncertainty associated with (Reference 43) discussion of quantification has already been convergence and truncation limits addressed. The other source of used. No further sensitivity with uncertainty is the selection of the respect to truncation is required truncation limit. (Reference 40).

Uncertainty and This task does not introduce any new N/A Sensitivity uncertainties but is intended to address Analyses how uncertainties may impact the fire risk.

FPRA This task does not introduce any new The documentation task compiles the Documentation uncertainties to the fire risk. results of the other tasks. See specific technical tasks for a discussion of their associated uncertainty and sensitivity.

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