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 i

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

Enclosure ii 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

Enclosure 3

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.

Enclosure 4

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.

Enclosure 5

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.

Enclosure 6

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).

Enclosure 7

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.

Enclosure 8

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.

Enclosure 9

• 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

Enclosure 10

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.

Enclosure 11 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.

Enclosure 12 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.

Enclosure 13 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

Enclosure 14 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.

Enclosure 15 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.

Enclosure 16 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:

Enclosure 17

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

Enclosure 18 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.

Enclosure 19 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.

Enclosure 20 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

Enclosure 21 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.

Enclosure 22

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.

Enclosure 23 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. : List of Categorization Prerequisites

Enclosure 24 Plant Units Model Baseline CDF Baseline LERF Comments Byron 1 & 2 Full Power Internal Events and Internal Flooding (FPIE)

BB016A for both units 1.1E-05 (Unit 1) 1.1E-05 (Unit 2) 9.0E-07 (Unit 1) 9.0E-07 (Unit 2) 2016 Full Power Internal Events (FPIE) with Internal Flooding PRA Model of Record (MOR) 1 & 2 Fire PRA BB11b-FL for both units 5.6E-05 (Unit 1) 6.1E-05 (Unit 2) 3.1E-06 (Unit 1) 3.1E-06 (Unit 2)

Application Specific Model (ASM) to the Fire Update which was based on the 2011 PRA MORs Braidwood 1 & 2 Full Power Internal Events and Internal Flooding (FPIE)

BB016A for both units 1.2E-05 (Unit 1) 1.2E-05 (Unit 2) 1.0E-06 (Unit 1) 1.0E-06 (Unit 2) 2016 Full Power Internal Events (FPIE) with Internal Flooding PRA Model of Record (MOR) 1 & 2 Fire PRA BB11b-FL for both units 5.5E-05 (Unit 1) 6.6E-05 (Unit 2) 3.2E-06 (Unit 1) 3.3E-06 (Unit 2)

Application Specific Model (ASM) to the Fire Update which was based on the 2011 PRA MORs

Description of PRA Models used in Categorization

Enclosure 25 Finding Number Supporting Require-ment(s)

Capability Category (CC)

Finding Description Disposition for 50.69 SY-B12-01 SY-B12 Cat I/II/III Recovery actions are included in the model.

However, a heating, ventilation and air conditioning (HVAC) dependency is not included for the switchgear and battery rooms.

The room heatup calculations are based on a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> mission time, with assumption that within that period, operators would take action to restore cooling to the rooms. A similar issue also was identified as part of the Supporting Requirement (SR) H2-E2 review where local operator actions to provide inverter room ventilation cooling (such as opening the door and installing temporary fans) were not identified. This was identified as a necessary operator action in the Equipment Survivability Notebook to recover from a failed room ventilation cooling system.

PARTIALLY RESOLVED Updated room cooling calculations and survivable temperature evaluations have been performed and indicate support for the current modeling assumptions for most scenarios for the Engineered Safety Feature (ESF) and Non-ESF Switchgear Rooms. The only identified potential impact on the model is for high energy line break (HELB) scenarios, which are an overall small contributor to the baseline risk results. A sensitivity calculation that simply inserts new HELB-related room cooling requirements shows a small increase in baseline CDF and LERF, but these increases do not trigger consideration of an emergent model update per the Exelon Risk Management procedures.

Model changes to incorporate these additional HELB-related room cooling requirements are being tracked in the URE (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 details to meet Capability Category II/III as the only statement is in Section 3.8 of the

'Braidwood Fire PRA Cable Selection Notebook (BW-PRA-021.03), Rev 0' : "The Braidwood OPEN.

A review of breaker coordination was performed for credited FPRA power supplies but not all calculations were available at the time of the Fact and Observation (F&O) Finding

Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items

Enclosure 26 Finding Number Supporting Require-ment(s)

Capability Category (CC)

Finding Description Disposition for 50.69 Fire PRA reviewed the electrical coordination calculations for applicability to the Fire PRA.

These were reviewed for each of the credited power supplies in the model." This did not provide analysis or identify any additional requirements; it only stated that it was reviewed.

Closure Review. These calculations have been identified and their results will be incorporated into the current FPRA update, so that this issue will have no impact on this application.

18-12 UNC-A1 Not Met Some anomalies were observed in the database that was used for the uncertainty analysis which was different from that used in the FPRA.

OPEN A review and update, if needed, of the basic event (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.

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 contributors to LERF.

PARTIALLY RESOLVED 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.

Enclosure 27 Finding Number Supporting Require-ment(s)

Capability Category (CC)

Finding Description Disposition for 50.69 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 components and basic events to determine that they make logical sense.

PARTIALLY RESOLVED 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 measures for Braidwood Unit 2 CDF/LERF from QU-F2 (j).

PARTIALLY RESOLVED 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 damage states and accident progression contributors.

PARTIALLY RESOLVED 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

Enclosure 28 Finding Number Supporting Require-ment(s)

Capability Category (CC)

Finding Description Disposition for 50.69 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 were checked and found to have errors in ignition source counting: 11.4A-1, 11.4A-2, and 11.6-0.

PARTIALLY RESOLVED WITH OPEN DOCUMENTATION Two panels which were missing for the fire scenarios in the 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 safely shutting down the plant from the remote shutdown panel or from actions outside the MCR. This results in a CCDP of 1.0 being applied to scenarios that require operator abandonment or where sufficient functionality is OPEN.

The Main Control Room (MCR) abandonment scenario conditional core damage probability (CCDP) was quantified using a scaling factor applied to the scenario CCDP /

CLERP. A scaling factor was applied for each CCDP /

Enclosure 29 Finding Number Supporting Require-ment(s)

Capability Category (CC)

Finding Description Disposition for 50.69 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) used to recover the risk are not supported by the documented dependency analysis.

PARTIALLY RESOLVED WITH OPEN DOCUMENTATION 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.

Enclosure 30 Finding Number Supporting Require-ment(s)

Capability Category (CC)

Finding Description Disposition for 50.69 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 database lack sufficient information for a complete parametric uncertainty analysis.

OPEN.

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 not fully developed which may mask the important contributors to fire risk. This Supporting Requirement (SR) requires that significant contributors be identified in accordance with HLR-QU-D. HLR-QU-D6 requires that significant contributors be identified and HLR-QU-D7 requires review of important components and basic events (Bes) to determine that they make logical sense. This is not possible with overly conservative scenario models.

OPEN.

Refinements were performed to address conservative joint human error probability (JHEP) values and to refine the treatment for the containment isolation valves for the containment mini-purge lines. These changes reduced the overall risk. With these changes completed no specific need for additional refinement is considered to be required.

These changes were finalized after the F&O Finding Closure Review and are reflected in the current FPRA results; therefore, they have no impact on this application.

25-9 FQ-E1 Not Met HLR-QU-D7 requires review of importance of components and basic events (Bes) to determine that they make logical sense. The information provided did not meet the intent of providing a review of importance of components and basic events.

PARTIALLY RESOLVED 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

Enclosure 31 Finding Number Supporting Require-ment(s)

Capability Category (CC)

Finding Description Disposition for 50.69 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 appear to have been addressed for the FPRA analysis.

PARTIALLY RESOLVED 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 of contributors to LERF was not addressed.

PARTIALLY RESOLVED 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.

Enclosure 32 Finding Number Supporting Require-ment(s)

Capability Category (CC)

Finding Description Disposition for 50.69 25-22 FQ-F1 Not Met The process used to identify plant damage states and accident progression contributors was not documented.

PARTIALLY RESOLVED 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 large electrical wall-mounted cabinets with 14 switches were not counted in 11.4C-0, which is a risk-significant fire zone.

OPEN.

Wall mounted panels were screened from the FPRA. Based 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.

Enclosure 33 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron Aircraft Impact Y

PS2 PS4 From Braidwood UFSAR Section 3.5.1.6, Aircraft Hazards, the airports and airways in the vicinity of the site are described in Braidwood UFSAR Subsection 2.2.2.5 (Reference 16).

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

Using a refined method as described in Braidwood UFSAR Section 3.5.1.6, the probability of an aircraft crashing into the safety-related plant structures is 3.8E-08 per year, which is less than the threshold value of 1.0E-07 per year for postulating the design-basis event. Hazards due to aircraft crashes thus need not be considered in the design of the Braidwood Station.

Based on this review, the aircraft impact hazard can be From Byron UFSAR Section 3.5.1.6, Aircraft Hazards, the airports and airways in the vicinity of the site are described in Byron UFSAR Subsection 2.2.2.5 (Reference 16).

There are no airports with projected operations greater than 500 d2 movements per year within 10 miles of the site and greater than 1000 d2 movements per year outside 10 miles, where d is the distance in miles from the site. There are no low altitude federal airways within 2 miles of the site.

For the airports and seaplane base within 10 miles of the site (shown in Byron UFSAR Figure 2.2-3), an analysis has been performed which shows that the probability of an aircraft crashing into the safety-related plant structures is 3.7E-08 per year.

Based on this review, the aircraft impact hazard can be considered to be negligible. : External Hazards Screening

Enclosure 34 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron considered to be negligible.

Avalanche Y

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

The mid-western location of Byron station precludes the possibility of a snow avalanche.

Biological Event Y

C3 C5 Actions committed to and completed by Braidwood station in response to Generic Letter 89-13 (Service Water System Problems Affecting Safety-Related Equipment) provide on-going control of biological hazards.

These controls are described in Exelon procedure ER-AA-340, "GL 89-13 Program Implementing Procedure". In addition, station actions taken in response to INPO SOER 07-2 (Intake Structure Blockage) provide an additional layer of biological hazard management.

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

Actions committed to and completed by Byron station in response to Generic Letter 89-13 provide on-going control of biological hazards. These controls are described in Exelon procedure ER-AA-340, "GL 89-13 Program Implementing Procedure".

In addition, station actions taken in response to INPO SOER 07-2 provide an additional layer of biological hazard management.

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

Coastal Erosion Y

C3 The mid-western location of Braidwood station precludes the possibility of coastal erosion.

The mid-western location of Byron station precludes the possibility of coastal erosion.

Drought Y

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

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

Enclosure 35 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron External Flooding Y

C1 PS2 The external flooding hazard at the site was recently updated and the flood hazard reevaluation report (FHRR) was submitted to NRC for review on March 12, 2014 (Reference 17) and revised to respond to Request for Additional Information (RAI) on June 24, 2015 (Reference 18). The results indicate that Local intense precipitation (LIP) was evaluated and found to have a maximum water surface elevation (WSE) of 601.11 ft. This mechanism does not produce a WSE greater than the plant's current licensing basis (CLB) or the building entry way elevation of 602.25 ft. as indicated in the UFSAR (Reference 16).

The combined flooding mechanism effects found that the limiting scenario is an all season probable maximum precipitation (PMP) resulting in a probable maximum flood (PMF) raising the WSE in the cooling pond adjacent to the site. This is combined with the 2-year wind speed to estimate wave runup effects and the maximum WSE of 602.34 ft (Reference 18),

which is greater than plant building entry way at elevation 602.25 ft. In order to provide protection for this extreme event, the exterior dike adjacent to the lake screen house was built 2.5 ft.

higher than the other The external flooding hazard at the site was recently updated and the flood hazard reevaluation report (FHRR) was submitted to NRC for review on March 12, 2014 (Reference 19).

By letter dated September 3, 2015 (Reference 20), the NRC staff concluded that the reevaluated flood hazard mechanisms for Byron are bounded by the current design basis.

Subsequently, Byron revised the model used to develop the local intense precipitation (LIP) flood hazard parameters. The revision resulted in minor differences between the LIP parameters described in the external flooding mitigating strategies assessment (MSA) submitted to NRC on September 30, 2016 (Reference 21) and the reevaluated LIP parameters described in the September 3, 2015 letter.

Specifically, the revised model resulted in an increase in the maximum flood elevation due to LIP from 870.9 feet to 870.94 feet, which exceeds the plant floor elevation at the east and southwest sides

Enclosure 36 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron locations. The top elevation of this dike is 602.5 ft. MSL, which is greater than the WSE. No impacts from this external flood causing mechanism are expected.

of the turbine building.

The NRC staff assessment letter of the Byron MSA dated October 21, 2016 (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.

Extreme Wind or Tornado Y

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

Enclosure 37 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron can be screened.

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

(Reference 24).

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

extreme winds or tornados can be screened.

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

(Reference 24).

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

Fog Y

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

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

Forest or Range Fire Y

C3 The site landscaping and lack of forestation prevent such fires from posing a threat to Braidwood station.

The site landscaping and lack of forestation prevent such fires from posing a threat to Byron station.

Frost Y

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

Enclosure 38 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron events PRA model for Braidwood.

initiating event in the internal events PRA model for Byron.

Hail Y

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

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

High Summer Temperature Y

C1 The principal effects of such events would result in elevated lake temperatures which are monitored by station personnel. Should the temperature exceed the Technical Specification limit, an orderly shutdown would be initiated. Another potential initiating consequence would be to cause a loss of off-site power. These effects would take place slowly allowing time for orderly plant reductions, including shutdowns. At worst, the loss of off-site power events would be subsumed into the base PRA model results.

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

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

High Tide, Lake Level, or River Stage Y

C3 The mid-western location of Braidwood station precludes the possibility of a high tide condition.

High lake effects would take place slowly allowing time for The mid-western location of Byron station precludes the possibility of a high tide condition.

High river effects would have negligible impact to

Enclosure 39 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron orderly plant reductions, including shutdowns.

the plant due to the installation of cooling towers being the ultimate heat sink.

Hurricane Y

C4 The mid-western location of Braidwood station precludes the possibility of a hurricane.

In addition, hurricanes would be covered under Extreme Wind or Tornado and Intense Precipitation.

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

Ice Cover Y

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

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

Industrial or Military Facility Accident Y

C1 C3 Joliet Arsenal is located eight miles from the site. This facility has largely ceased activities except for those of a custodial nature. Shipments of explosive materials have essentially been terminated.

There are no manufacturers of hazardous materials located within five miles of the site. Hazardous chemicals used and/or stored by manufacturers within five miles of the plant were also evaluated and determined to either screen from further evaluation or were determined to meet the acceptance criteria associated with Control There are no military facilities within 10 miles of the site. Industrial manufacturing facilities have also been evaluated and determined to not have an impact to the Byron site as discussed in the Byron 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 26).

Enclosure 40 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment 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).

Internal Flooding N

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

The Byron Internal Events PRA includes evaluation of risk from internal flooding events.

Internal Fire N

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

The Byron internal Fire PRA addresses risk from internal fire events.

Landslide Y

C3 The mid-western location of Braidwood station precludes the possibility of a landslide.

Not applicable to the site because of topography.

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

Lightning Y

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

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

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

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

Enclosure 41 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron Low Lake Level or River Stage Y

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

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

Low Winter Temperature Y

C1 C5 The principal effects of such events would be to cause a loss of off-site power. These effects would take place slowly allowing time for orderly plant reductions, including shutdowns. At worst, the loss of off-site power events would be subsumed into the base PRA model results.

The principal effects of such events would be to cause a loss of off-site power. These effects would take place slowly allowing time for orderly plant reductions, including shutdowns. At worst, the loss of off-site power events would be subsumed into the base PRA model results.

Meteorite or Satellite Impact Y

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

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

Pipeline Accident Y

C1 There are six natural gas pipelines and four other petro-chemical pipelines within five miles of the site.

There is no significant hazard from toxic releases or explosions involving these pipelines that could interact with the plant.

Chemical Hazards transported using pipelines that are located in the vicinity of the plant are analyzed in accordance with Technical Specification 5.5.18 (Reference 25).

The nearest pipeline to the site is 2.5 miles away and is only a 3-inch diameter pipe. There is no significant hazard to the site from these events.

Chemical Hazards transported using pipelines that are located in the vicinity of the plant are analyzed in accordance with Technical Specification 5.5.18 (Reference 26).

Enclosure 42 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron Release of Chemicals in Onsite Storage Y

PS2 Chlorination of water systems is performed using a hypochlorite system. No chlorine gas is stored on-site.

Various acids and caustics are stored on-site but pose no hazard to the plant.

Chemical Hazards stored and transported in the vicinity of the plant are analyzed in accordance with Technical Specification 5.5.18 (Reference 25).

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

Chemical Hazards stored and transported in the vicinity of the plant are analyzed in accordance with Technical Specification 5.5.18 (Reference 26).

River Diversion Y

C3 There is no historical or topographical evidence indicating that flow in the Kankakee River can be diverted away from its present course. The river is wide, and there are no deeply incised gorges upstream where landslides could entirely cut off river flow.

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

Due to the great width of the Rock River and the relatively flat surrounding terrain, there is little possibility that rock falls, ice jams or subsidence could completely divert the flow away from the makeup water intake Refer to Byron UFSAR Section 2.4.9(Reference 16).

Sand or Dust Storm Y

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

Enclosure 43 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron plant.

features of the plant.

Seiche Y

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

Seiche was found to not be an applicable external flooding mechanism in Reference 19.

Seismic Activity N

None See information in Section 3.2.3 of this application.

See information in Section 3.2.3 of this application.

Snow Y

C1 C4 Snow cover is included as an input to the probable maximum flood (PMF) WSE calculations (Reference 17).

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

WSE calculations (Reference 19).

Soil Shrink-Swell Consolidation Y

C1 C5 Based on the discussions and conclusions reached in the Braidwood UFSAR section 2.5, "Geology, Seismology, And Geotechnical Engineering" (Reference 27), the impact from soil shrink or swell (subsidence or uplift) is expected to be negligible and can be screened from further evaluation.

Based on the discussions and conclusions reached in the Byron UFSAR section 2.5, "Geology, Seismology, And Geotechnical Engineering" (Reference 27, the impact from soil shrink or swell (subsidence or uplift) is expected to be negligible and can be screened from further evaluation.

Storm Surge Y

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

The mid-western location of Byron station precludes the possibility of a sea level driven storm surge.

Toxic Gas Y

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

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

Enclosure 44 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron Transportation Accident Y

C3 C4 PS2 PS4 Chemical hazards stored and transported in the vicinity of the plant are analyzed in accordance with Braidwood Technical Specification 5.5.18 (Reference 25) and Regulatory Guides 1.78, 1.95 and 1.196 to ensure Control Room habitability is maintained. The latest analysis was performed in 2014.

The analysis concluded that toxic chemicals transported or stored within the vicinity of the plant do not pose a threat that requires the use of instrumentation.

The Kankakee River is not 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).

Railroad track approaches no closer than four miles to the plant site. There is no heavy traffic in bulk hazardous materials capable of impacting the site.

The Rock River is not 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 Byron Technical Specification 5.5.18 (Reference 26).

Tsunami Y

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

The mid-western location of Byron station precludes the possibility of a tsunami.

Turbine-Generated Missiles Y

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

Enclosure 45 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron refueling outage, a calculation for total unit missile generation probability is made. To make this calculation the operational hours on each of the low pressure turbines is gathered. The missile generation probability for each of the LP turbine rotors is taken from the individual rotors graph based on the operational time on the rotor.

The values for the three rotors are then added together to determine the current missile generation probability for the unit. This value must be below 1.0E-05 to allow loading the turbine and bringing the unit on line.

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

performance and integrity.

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

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

Volcanic Activity Y

C3 Not applicable to Braidwood Station Not applicable to Byron Station

Enclosure 46 External Hazard Screening Result Screened?

(Y/N)

Screening Criterion (Note a)

Comment Braidwood Byron Waves Y

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

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

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

Enclosure 47 Event Analysis Criterion Source Comments Initial Preliminary Screening C1. Event damage potential is

< events for which plant is designed.

NUREG/CR-2300 and ASME/ANS Standard RA-Sa-2009 C2. Event has lower mean frequency and no worse consequences than other events analyzed.

NUREG/CR-2300 and ASME/ANS Standard RA-Sa-2009 C3. Event cannot occur close enough to the plant to affect it.

NUREG/CR-2300 and ASME/ANS Standard RA-Sa-2009 C4. Event is included in the definition of another event.

NUREG/CR-2300 and ASME/ANS Standard RA-Sa-2009 Not used to screen.

Used only to include within another event.

C5. Event develops slowly, allowing adequate time to eliminate or mitigate the threat.

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

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

NUREG-1407 and ASME/ANS Standard RA-Sa-2009 PS3. Design basis event mean frequency is < 1E-5/y and the mean conditional core damage probability is < 0.1.

NUREG-1407 as modified in ASME/ANS Standard RA-Sa-2009 PS4. Bounding mean CDF is <

1E-6/y.

NUREG-1407 and ASME/ANS Standard RA-Sa-2009 Detailed PRA Screening not successful. PRA needs to meet requirements in the ASME/ANS PRA Standard.

NUREG-1407 and ASME/ANS Standard RA-Sa-2009

Progressive Screening Approach for Addressing External Hazards

Enclosure 48 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition LOOP Initiating Event Frequency:

The initiating event analysis develops frequencies for LOOP and DLOOP events based on the data in NUREG/CR-6890 (Reference 28)

(updated through 2013). LOOP types 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.

SSCs that support LOOP scenarios The overall approach for the LOOP frequency and fail to recover probabilities utilized is consistent with industry practice.

Therefore, this does not represent a key source of uncertainty for the 50.69 application.

Failure to recover probabilities for LOOP:

The industry wide data in NUREG/CR-6890 (Reference 28) (updated through 2013) is utilized to develop the failure to recover probabilities for the four LOOP categories. The industry wide recovery data is applicable to both sites and is acceptable for the base case analysis.

SSCs that support LOOP scenarios The LOOP frequencies utilized in the model are based on NUREG/CR-6890 (Reference 28) as updated with data through 2013. Therefore, this does not represent a key source of uncertainty for the 50.69 application. : Disposition of Key Assumptions/Sources of Uncertainty

Enclosure 49 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Grid stability after a reactor trip:

The consequential LOOP failure probabilities are based on EPRI and NRC evaluations with different values following a reactor trip or LOCA (References 28, 29). The use of generic data for consequential LOOP 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.

SCCs that support LOOP scenarios The consequential LOOP probabilities utilized provide a reasonably realistic modeling. As such, this does not represent a significant source of model uncertainty in this application.

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

Offsite power restoration:

Restoration is possible as the switchyard has its own 125V DC distribution system to provide breaker and transformer control power. When offsite power is available at the switchyard, then power is available to 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.

SCCs that support LOOP scenarios The LOOP recovery probabilities are realistic with slight conservative bias on the recovery times. This does not represent a significant source of model uncertainty in this application.

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

Enclosure 50 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Support System Initiating Events (SSIEs):

Support System Initiating Event fault trees are developed for loss of Component Cooling Water (CC), loss of Service Water (SX), and loss of Non-Essential Service Water (WS).

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).

SSCs that support SSIEs.

Realistic with slight conservative 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 model uncertainty in this application. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Support System Initiating Events (SSIEs):

Increasing use of plant-specific models for support system initiators (e.g. loss of SX, CC, or Instrument Air (IA), and loss of AC or DC buses) have led to 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.

SSCs that support SSIEs.

Slight conservative bias treatment since alpha factors are known to be high when utilized in an annualized fashion and compared to plant-specific experience. This does not represent a key source of uncertainty in the 50.69 application.

Enclosure 51 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Support System Initiating Events (SSIEs):

Modeling of recovery to prevent support system initiating events is limited to procedurally-directed 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.

CC and SX SSCs that could contribute to SSIEs.

Slight conservative treatment since credit for recovery beyond system failure could reduce the baseline CDF and LERF risk metrics. This does not represent a key source of uncertainty in the 50.69 application.

LOCA initiating event frequencies:

The Large and Medium LOCA initiating event frequencies are based on failure probabilities from NUREG/CR-1829 (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.

SSCs that support Safety Injection (SI)

The LOCA frequency values represent realistic treatment based on accepted industry data sources. This does not represent a key source of uncertainty in the 50.69 application.

Operation of equipment after battery depletion:

No credit is taken for continued operation of any systems without DC power that normally require DC power for operation. This includes steam generator (SG) level control.

SSCs supporting scenarios that require DC power No credit for equipment operation after battery depletion may represent a slight conservative bias. This does not represent a key source of uncertainty in the 50.69 application.

Enclosure 52 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition RCP seal LOCA treatment:

The assumed timing and magnitude of RCP seal LOCAs given a loss of seal injection and thermal barrier cooling can have a substantial influence on the risk profile. The WOG 2000 consensus model (Reference 32) has been implemented, along with the model for the Shutdown Seals (Reference 33).

SSCs that support seal LOCA accident scenarios The operator action timing assumptions are based on the WOG 2000 consensus model and Shutdown Seal model. Therefore, this does not represent a key source of uncertainty for the 50.69 application.

Battery life calculations:

Design basis calculations indicate that

~8 hours of battery life is available depending on scenario specifics. Credit 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 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.

SSCs involved in scenarios in which DC power is required.

The modeling is realistic given the relatively long battery life without recharge. This may be slightly conservative for SBO scenarios and does not represent a key source of model uncertainty in the 50.69 application.

Number of PORVs required for bleed and feed:

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.

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

Enclosure 53 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Impact of failure of pressure relief:

For general transients with reactor trip, the PORVs provide pressure relief if needed, and the likelihood of a safety relief valve challenge is sufficiently small that explicit modeling is not required. For general transients (non-ATWS), it is commonly assumed (and evident in 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.

SSCs supporting scenarios where pressure relief via PORVs and SRVs are required The approach taken is consistent with that used in other PWR PRAs. The potential impact on CDF due to not explicitly modeling the possibility of overpressure for non-ATWS events is not significant. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Impact of failure of pressure relief:

For ATWS scenarios, the number of PORVs and/or SRVs is a function of core reactivity, available AFW capacity, and other parameters as specified in the WOG ATWS model (Reference 34). Per the WOG model, there may be brief periods of time in which all available pressure relief is not adequate to maintain RCS pressure below the ASME Service Level C pressure. Failure to 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.

SSC supporting ATWS scenarios for which PORVs or SRVs are required Slight conservative bias treatment in assumption that overpressure failure in ATWS cases goes directly to core damage, since RCS piping failures in most locations would not result in LOCAs in excess of ECCS capability. However, the modeling is in accordance with an industry recognized model and thus does not represent a key source of uncertainty in the 50.69 application.

Enclosure 54 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Operability of equipment in beyond design basis environments:

Generally, credit for operation of systems beyond their design-basis environment is not taken. However, each DG requires ventilation to operate successfully. This dependency is modeled to include suction and exhaust dampers (including CCF terms as applicable) and fan fail-to-start and fail-to-run terms. There may be 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.

SSCs supporting scenarios for which DGs are required Not explicitly accounting for the limited time frames when two DG HVAC fans may be required to provide DG cooling is a potential slight non-conservatism. Not modeling the proceduralized restoration of DG ventilation is a potential conservatism. Given that a ventilation dependency is modeled, this does not represent a key source of uncertainty in the 50.69 application.

Operability of equipment in beyond design basis environments:

Generally, credit for operation of systems 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.

SSCs supporting scenarios for which batteries are required.

Realistic with slight conservative bias on assumed battery life time.

This does not represent a key source of uncertainty in the 50.69 application.

Enclosure 55 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Widespread LOOP effects:

Credit for TSC actions is not currently used for cognitive error contributions in HRA due to its significant uncertainty.

The TSC is implicitly used for execution recovery for long-term actions, but this is not directly affected by a widespread LOOP.

Increased stress due to communication challenges is recognized as a source of model uncertainty and is not explicitly included in the LOOP-related HEP calculations.

SSCs supporting LOOP scenarios Lack of credit for TSC actions is slightly conservative.

Lack of explicit consideration of increased stress due to a widespread LOOP could be slightly non-conservative, but its effect is expected to be low based on the low likelihood of the event and the already present stress during a normal LOOP event.

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

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:

The Internal flood analysis and initiating event frequencies for spray, flood, and major flood scenarios are developed consistent with the EPRI methodology (Reference 35). The flooding analysis is integrated into the internal events at power model. The use of generic flood frequencies with plant-specific estimates of pipe lengths is suitable for representation of the flood frequencies at the site.

SSCs that support Internal Flood scenarios Considered an industry good practice approach, but is not yet a consensus model approach. This is not a source of significant model uncertainty given that a recognized methodology has been applied using plant-specific piping data. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Enclosure 56 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Piping failure mode:

The Internal flood analysis and initiating event frequencies for spray, flood, and major flood scenarios are developed consistent with the EPRI methodology (Reference 33). The flooding analysis is integrated into the internal events at power model. Spray flood scenarios with less than 100 GPM flow do not totally disable the system they arise from.

Major flood sources greater than 100 gpm are assumed to totally disable the system they arise from.

SSCs that support Internal Flood scenarios Realistic since such a low flowrate would not affect most systems needed to mitigate an accident.

Therefore, this does not represent a key source of uncertainty in the 50.69 application Assuming major flood sources greater than 100 gpm totally disable the system they arise from is conservative in that the system may not be totally disabled in all cases. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Core melt arrest in-vessel:

In-vessel recovery of the molten core by flooding of the reactor cavity and heat transfer through the vessel is not 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.

SSCs that support LERF scenarios Conservative bias treatment in that in-vessel core melt arrest might be feasible in some scenarios. This does not represent a key source of uncertainty in the 50.69 application.

Enclosure 57 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Thermally induced failure of hot leg/SG tubes:

The approach follows Simplified Level 2 Modeling Guidelines, WCAP-16341-P (Reference 33), which many plants are 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.

SSCs that support LERF scenarios Approach is consistent with recent industry approaches and appropriate for determination of LERF. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Vessel failure mode:

RPV catastrophic failure leading to early containment failure via missiles is extremely unlikely based on studies 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.

SSCs that support LERF scenarios Approach is appropriate for determination of LERF. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Enclosure 58 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Vessel failure mode:

The approach follows Simplified Level 2 Modeling Guidelines, WCAP-16341-P (Reference 33), which many plants are 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 direct containment heating and other Level 2 issues.

SSCs that support LERF scenarios Approach is consistent with general industry approaches and appropriate for determination of LERF. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Vessel failure mode:

Ex-vessel steam explosions noted as very unlikely based on reference to generic studies. Based on WCAP-16341-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.

SSCs that support LERF scenarios Approach is appropriate for determination of LERF for Braidwood and Byron. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Ex-vessel cooling of lower head:

No credit for ex-vessel cooling.

SSCs that support LERF scenarios No credit for ex-vessel cooling of the lower head represents a realistic treatment with a slight conservative bias. Therefore, this does not represent a key source of uncertainty in the 50.69 application.

Core debris contact with containment:

This is not considered as an early failure mechanism because there is no direct path for core debris to contact the containment shell.

SSCs that support LERF scenarios The modeling reflects the plant design. Therefore this does not represent a key source of uncertainty in the 50.69 application.

Enclosure 59 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Containment integrity following vessel rupture event:

Vessel rupture sequence is assumed to not result in concurrent containment failure coincident with the vessel rupture.

SSCs that support LERF scenarios Vessel rupture frequency is on the order of E-7, i.e., very small, such that potential impact on LERF is also small. Containment integrity following vessel rupture is therefore not identified as a candidate source of model uncertainty.

Condensate Storage Tank Inventory:

The inventory in the Condensate Storage Tank (CST) is normally not sufficient for 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 the PRA. CST inventory is adequate for approximately 16 hrs; an assumption is made that the likelihood of failing to either refill the CST or align to shutdown cooling within this extended time frame is very small. The service water (SX) 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.

SSCs that support scenarios where secondary side decay hear removal is required.

Lack of explicit modeling of the various sources of decay heat removal should not have a significant impact on CDF or LERF. The CST is the preferred source of water for AF, but service water is the safety related suction source for AF. Therefore, the duration the CST is able to provide suction for AF pumps is not a key source of model uncertainty for applications.

Enclosure 60 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Human Error Probabilities (HEPs):

Detailed evaluations of HEPs are performed for the risk significant human failure events (HFEs) using industry consensus methods. Mean values are used for the modeled HEPs. Uncertainty associated with the mean values can have an impact on CDF and LERF results.

SSCs that use an Operator Action basic event as a surrogate for a modeled component.

Sensitivity cases for the base internal events PRA (HEP values of 5th or 95th percentile value HEPs) show that the results are somewhat sensitive to HRA model and parameter values.

The BY/BW PRA model is based on industry consensus modeling approaches for its HEP 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.

Enclosure 61 Internal Events / Internal Flooding PRA Assumptions / Sources of Uncertainty Assumption / Uncertainty 50.69 Impact Model Sensitivity and Disposition Dependent HEPs:

Dependent HEP values are developed for significant combinations of HEPs that have been demonstrated to appear together in the same cutsets.

Potentially all SSCs evaluated during 50.69 categorization The BY/BW PRA model is based on industry consensus modeling approaches for its dependent HEP identification and 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.

Common Cause Failure:

Common cause failure values are developed using available industry data.

Potentially all SSCs evaluated during 50.69 categorization The BY/BW PRA model is based on industry consensus modeling approaches for its common cause 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):

The detailed ISLOCA analysis includes the relevant considerations listed in IE-C12 of the ASME/ANS PRA Standard (Reference 36) and accounts for common cause failures and captures likelihood of different piping failure modes.

SSCs that support LERF scenarios The values utilized provide a reasonable best-estimate approach, will have only a minor impact on the 50.69 and do not represent a key source of uncertainty.

Enclosure 62 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 boundary and partitioning This task poses a limited source of uncertainty beyond the credit taken for boundaries and partitions.

The multi-compartment analysis further reduces this uncertainty by addressing the potential impact of failure of partition elements on quantification.

Component Selection This task is associated with the development of the linkage between safe shutdown (SSD) analysis component/cable data to fault tree failure modes. Also included in this task is the development and incorporation of multiple spurious operation (MSO) scenarios not addressed in the internal events model fault tree.

The only uncertainty associated with this task is related to the identification of all credible MSO scenarios. This source of uncertainty is reduced as a result of multiple overlapping tasks including the MSO expert panel and industry owners group identification of applicable MSOs. Additional internal reviews of analysis results further reduce the uncertainty associated with this task.

Cable Selection No treatment of uncertainty is typically required for this task beyond the understanding of the cable selection approach (i.e., mapping an active basic event to a passive component for which power cables were not selected).

Additionally, PRA credited components for which cable routing information was not provided represent a source of uncertainty (conservatism) in that these components are assumed failed unnecessarily.

The limited number of components without available cable routing (most active components credited in the fire PRA have their cables routed) as well as the crediting by exclusion of these components (where justified) helps to reduce unnecessary conservatism. A sensitivity analysis for this conservatism is addressed in the FPRA uncertainty analysis.

Qualitative Screening Qualitative screening was not performed; however, structures were eliminated from the global analysis boundary and ignition sources deemed to have no impact on the fire PRA were excluded from the quantification based on qualitative screening criteria. The only criterion subject to uncertainty is the potential for plant trip.

In the event that a structure which could lead to a plant trip was excluded incorrectly, its contribution to CDF would be small (with a CCDP commensurate with base risk) and would likely be offset by inclusion of the additional ignition sources and the resulting reduction of other scenario frequencies. A similar argument can be made for ignition sources for which scenario development was deemed unnecessary.

Enclosure 63 Fire PRA Sources of Model Uncertainty Description Sources of Uncertainty Disposition Fire-Induced Risk Model A reactor trip is assumed as the initiating event for all quantification.

This is somewhat conservative since not all fires postulated will result in a plant trip.

FPIE and fire PRA peer reviews (including the F&O resolution process) and internal assessments are useful in exercising the model and identifying 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 Frequency Ignition source counting is an area with inherent uncertainty; however, the results are not particularly sensitive to changes in ignition source counts. The primary source of uncertainty for this task is associated with the frequency values from NUREG/CR-6850 (Reference 38) which result in uncertainty due to variability among plants along with some significant conservatism in defining the frequencies, and their associated heat release rates, based on limited fire events and fire test data.

The fire PRA utilizes the bin frequencies from NUREG/CR-2169 (Reference 39), which represents the most current approved bin frequencies.

As such, some of the inherent conservatism associated with bin frequencies from NUREG/CR-6850 (Reference 38) has been removed. A parametric uncertainty analysis using the Monte Carlo method is provided in section 4.1.1 of the FPRA uncertainty and sensitivity notebook (Reference 40).

Quantitative Screening Other than screening out potentially risk significant scenarios (ignition sources), there is no uncertainty from this task on the fire PRA results.

Quantitative screening is limited to refraining from further scenario refinement of those scenarios with a 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.

Enclosure 64 Fire PRA Sources of Model Uncertainty Description Sources of Uncertainty Disposition Scoping Fire Modeling The approach taken for this task included: 1) the use of generic fire modeling treatments in lieu of conservative scoping analysis techniques and 2) limited detailed fire modeling was performed to refine the scenarios developed using the generic fire modeling solutions. The primary conservatism introduced by this task is associated with the heat release rates specified in NUREG/CR 6850 (Reference 38).

The employment of generic fire modeling solutions did not introduce any significant conservatism. Detailed fire modeling was only applied where the reduction in conservatism was likely to have a measurable impact.

The NUREG-2178 (Reference 41) heat release rates are used and they constitute the most recent available heat release rate data. Some conservatism in this data is believed to exist. The level of conservatism cannot be quantified at this time.

Detailed Circuit Failure Analysis Uncertainty considerations for the circuit failure analysis task are addressed via the use of circuit failure mode probability factors in Task 10.

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

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

Circuit Failure Mode Likelihood Analysis The uncertainty associated with the applied conditional failure probabilities poses competing considerations primarily due to the assumption that all spurious operations occur at the same time. The hot short probability and the hot short duration factors defined in NUREG/CR-7150 (Reference 42) are considered best available data.

Circuit failure mode likelihood analysis was generally limited to those components where spurious operation was expected to be a large contributor to total risk. The assumption that all spurious operations (hot shorts) occur at the same time results in a significant conservatism in the analysis but is not easily assessed with respect to the impact on the overall results.

Detailed Fire Modeling The primary uncertainty in this task is in the area of target failure probabilities. Conservative heat release rates may result in additional target damage. Non-conservative heat release rates would have an opposite effect.

Credit for fire brigade response and detection are considered bounding given that the data used for manual non-suppression probability is based on extinguishment of a fire and not control (prevention of further spread) of a fire.

Detailed fire modeling was performed only on those scenarios which otherwise would have been notable risk contributors and only where removal of conservatism in the generic fire modeling solution was likely to provide benefit either via a smaller zone of influence or to allow credit for automatic suppression. Fire modeling was used to evaluate the time to abandonment for control room fire scenarios for a range of fire heat release rates. The analysis methodology conservatism is primarily associated with conservatism in the

Enclosure 65 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 Reliability Analysis Human error probabilities represent a potentially large uncertainty for the fire PRA given the importance of human actions in the base model. Since many of the HEP values were adjusted for fire, the joint dependency multipliers developed for the FPIE model also represent a potential for introducing a degree of conservatism.

Conservative HEP adjustments were made to the nominal HEP values used in the FPIE model then revisited to address unique fire considerations. A detailed analysis was performed for all fire specific HFEs. A floor value of 1E-06 was applied for all combinations.

Uncertainty in HEP values is propagated through the parametric uncertainty analysis.

Seismic-Fire Interactions Assessment Since this is a qualitative evaluation, there is no quantitative impact with respect to the uncertainty of this task.

Seismic fire interaction has no impact on fire risk quantification.

Fire Risk Quantification As the culmination of other tasks, most of the uncertainty associated with quantification has already been addressed. The other source of uncertainty is the selection of the truncation limit.

See Quantification Notebook (Reference 43) discussion of convergence and truncation limits used. No further sensitivity with respect to truncation is required (Reference 40).

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

N/A FPRA Documentation This task does not introduce any new uncertainties to the fire risk.

The documentation task compiles the results of the other tasks. See specific technical tasks for a discussion of their associated uncertainty and sensitivity.