ML24207A019

SE Addendum Related to the License Amendment No. 338 for Implementation of the Alternative Source Term (DPO-2021-001)

ML24207A019
Person / Time
Site:	FitzPatrick
Issue date:	10/09/2024
From:	Hipolito Gonzalez Plant Licensing Branch 1
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
Poole J
References
Download: ML24207A019 (1)

Text

October 9, 2024 David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - SAFETY EVALUATION ADDENDUM RELATED TO THE LICENSE AMENDMENT FOR IMPLEMENTATION OF THE ALTERNATIVE SOURCE TERM

Dear David Rhoades:

On August 25, 2023, the Executive Director for Operations (EDO) directed the Office of Nuclear Reactor Regulation (NRR) staff to take appropriate regulatory action to address issues identified in the Differing Professional Opinion (DPO) Case File for DPO-2021-001, FitzPatrick Amendment Concerning an Alternate Source Term for Calculating LOCA Accident Dose Consequences (Agencywide Documents and Access System (ADAMS) Accession No. ML23263A639). Specifically, NRR staff was directed to ensure that the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) is in compliance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, and to resolve any licensing basis clarity issues related to the previously issued alternative source term (AST) license amendment (ADAMS Accession No. ML20140A070). The DPO case file identified issues in the staffs review regarding the impact of the outboard main steam line isolation valve packing leakage, the basis for the limiting break location, and the aerosol deposition credit for the main condenser.

In conducting its review, consistent with the EDOs direction, NRR staff concluded with reasonable assurance that FitzPatrick is in compliance with 10 CFR 50.67. The staffs bases for reaching this finding are documented in the enclosed safety evaluation addendum. NRR staff found that the licensee followed the Nuclear Regulatory Commissions (NRC) guidance, supporting documentation, and relevant precedents in preparing and submitting its AST LAR.

NRR staff followed modern licensing procedures/approaches for current commercial operating light-water nuclear power reactors in conducting this review to ensure reasonable assurance of adequate protection of public health and safety. However, the information that the staff relied on in making its findings was not clearly articulated in its original safety evaluation. As such, NRR staff is providing an addendum to the original safety evaluation, as enclosed, to provide for additional clarity of its reasonable assurance finding.

A copy of the safety evaluation addendum is enclosed with this letter. If you have any questions, please contact Project Manager, Richard Guzman at 301-415-1030 or Richard.Guzman@nrc.gov.

Sincerely, Hipólito González, Branch Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosure:

As stated cc: Listserv HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2024.10.09 08:34:40 -04'00'

Enclosure SAFETY EVALUATION ADDENDUM BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 338 FOR IMPLEMENTATION OF THE ALTERNATIVE SOURCE TERM CONSTELLATION FITZPATRICK, LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59

1.0 INTRODUCTION

On August 25, 2023, the Executive Director for Operations (EDO) directed the Office of Nuclear Reactor Regulation (NRR) staff to take appropriate regulatory action to address issues identified in the Differing Professional Opinion (DPO) Case File for DPO-2021-001, FitzPatrick Amendment Concerning an Alternate Source Term for Calculating LOCA Accident Dose Consequences, (Agencywide Documents and Access System (ADAMS) Accession No. ML23263A639). Specifically, the staff was directed to ensure that the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) is in compliance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, and to resolve any licensing basis clarity issues related to the previously issued alternative source term (AST) license amendment No. 338 (ADAMS Accession No. (ML20140A070). The DPO case file raised issues related to the staffs review regarding the impact of the outboard main steam line isolation valve packing leakage, the basis for the limiting break location, and the aerosol deposition credit for the main condenser.

As described in Section 2.1 of the Technical Evaluation below, NRR has completed its assessment and has verified that FitzPatrick is in compliance with 10 CFR 50.67. During its assessment, the staff determined that no additional docketed information was needed to verify that FitzPatricks licensing basis provides a sufficient technical basis to demonstrate compliance with the regulatory requirements. This addendum to the safety evaluation for the FitzPatrick alternate [alternative] source term license amendment request (LAR) provides additional clarifications to the staffs evaluation and documents its verification of FitzPatricks compliance.

The process of taking appropriate regulatory action based on the EDOs direction involved a thorough review of pertinent information gathered from the DPO-2021-001 package. The information provided in the DPO package, along with the information that was already on the docket regarding the FitzPatrick AST LAR, provided the analysis and findings related to the subject matter and served as the foundation for the decision-making process, offering insights into the nature and extent of the regulatory concerns.

The regulatory action taken by NRR staff is to addend the original NRC staff safety evaluation for the FitzPatrick Amendment No. 338, to uphold and clarify its original conclusion that FitzPatrick is in compliance with 10 CFR 50.67 and resolve and disposition each of the licensing basis clarity issues discussed in the EDOs memo dated August 25, 2023. This addendum documents the details of the NRR staffs determination that the licensees AST analysis of record demonstrates compliance with 10 CFR 50.67.

When assessing each issue discussed in the Appeal Panel report enclosure to the August 25, 2023, memo from the EDO, NRR staff did so against applicable regulations and supporting information in the Federal Register notices, such as statements of consideration related to the pertinent rule; regulatory guidance such as Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors Revision 0; office instructions; the LWR Standard Review Plan (SRP); other supporting Agency documents; and licensing precedents established by the NRC staff when reviewing other AST submittals. In conducting these actions, NRR staff assessed the relevant regulatory, policy, and technical information to inform decisions regarding the most appropriate regulatory actions. The staffs actions ensured that the regulatory requirements and the broader objectives of safety are being met, and that there is reasonable assurance of adequate protection of public health and safety1.

A qualified technical subject matter expert independently peer-reviewed this addendum and pertinent supporting information gathered from the DPO-2021-001 package. This peer reviewer was not part of the original staff review of the FitzPatrick AST license amendment request (LAR) and any subsequent actions related to that review, including the DPO-2021-001. The purpose of having an independent, qualified technical subject matter expert perform the peer review was to provide an unbiased assessment of the technical accuracy and quality of the staffs addendum, ensuring that it meets the established requirements and guidance.

The original NRC staff safety evaluation was written consistent with many other AST safety evaluations completed for similar types of LARs for other plants. Therefore, certain information and context were deemed not necessary to be included in the original staff safety evaluation, which may have contributed to the confusion regarding issues identified in the DPO-2021-001 package. NRR provides this addendum to better communicate the staffs bases and conclusions that the licensees AST analysis of record is in compliance with 10 CFR 50.67 and to address, as appropriate, the identified licensing basis clarity issues.

Consistent with the SRP, NRR staffs review of the licensees AST analysis of record was composed of the entirety of the licensees docketed submittal, which constitutes its analysis of record. The submittal, referenced documents, and updated final safety analysis report contain all the necessary information for the staffs review of the licensees application of the AST.

NRR staffs review found reasonable assurance that FitzPatrick is in compliance with 10 CFR 50.67, and the bases for reaching this finding are documented in the following technical evaluation sections of this document. The staff found that the licensee followed the appropriate NRC guidance, supporting documentation, and relevant precedents in preparing and submitting its AST LAR. The staff followed modern licensing procedures for current commercial operating light-water nuclear power reactors to conduct its review by weighing safety margin and defense-in-depth against the uncertainties and identified non-conservatisms in the licensees 1 Discussions on reasonable assurance can be found in memorandums available at ADAMS Accession Nos. ML18240A410 and ML19015A290.

submittal with other information in an integrated, risk-informed manner. This encompasses the Commissions comprehensive framework that ensures reasonable assurance of adequate protection of public health and safety.

2.0 TECHNICAL EVALUATION

NRR staffs technical evaluation is categorized by four sections that correspond to the issues identified in DPO-2021-001 and the EDOs conclusions regarding these issues. The following sections address the four items discussed in the EDO memo, as follows:

Section 2.1-Compliance with 10 CFR 50.67: Provides the methodology and results from a margin analysis and defense-in-depth assessment consistent with SRP Section 15.0.1 to provide clarification to the staffs original safety evaluation.

Section 2.2-Packing Leakage: Provides an overview of the regulatory treatment of the main steam isolation valves (MSIVs), the assessment of the design-basis accident radiological consequences analyses when the licensees voluntary adopt 10 CFR 50.67 in which RG 1.183 supersedes corresponding radiological analysis assumptions provided in other regulatory guides and SRP chapters, and an assessment of precedents for the approval of the removal of the main steam line leakage control system. This section provides clarification to the staffs original safety evaluation.

Section 2.3-Limiting Break Location Approvals for an AST: Provides results of the justification for the licensees selection of the limiting break location and an assessment of precedents for approval to provide clarification to the staffs original safety evaluation.

Section 2.4-Aerosol Deposition Credit for the Main Condenser: Provides clarification as to whether aerosol deposition credit for the main condenser was either assumed by the licensee or within the staffs original safety evaluation, and the staffs use of NRRs Office Instruction LIC-206, Revision 1, Integrated Risk-Informed Decision Making for Licensing Reviews, to provide clarification to the staffs original safety evaluation.

Section 2.1-Compliance with 10 CFR 50.67 In the Case File for DPO-2021-001, the EDO concluded that:

The staffs evaluation of the AST amendment did not demonstrate that the application complied with the criteria specified in 10 CFR 50.67(b)(2).

Specifically, the applicants analysis did not demonstrate with reasonable assurance that calculated control room dose would not exceed 5 rem [total effective dose equivalent] TEDE for the duration of the accident.

NRR staff noted that the FitzPatrick safety evaluation was written consistent with many other AST safety evaluations that were completed for similar types of LARs for other plants.

Specifically, the AST safety evaluations were typically written to follow a select set of review guidance provided in SRP Chapter 15.0.1, Radiological Consequences Analyses Using Alternative Source Terms, including RG 1.183, Revision 0, issued in 2000, which contains many conservative sets of assumptions. NRC staff employed a streamlined review process that followed a select set of review guidance provided in SRP Chapter 15.0.1 that reflected lessons learned and experience gained since 2000. This efficient and risk-informed approach includes consideration in areas such as risk significance, safety margins, and past precedent. In addition, the AST FitzPatrick review was the first to utilize the LIC-206 process. LIC-206 provides NRR reviewer guidance on a graded approach for using risk insights in licensing so as to better integrate complementary insights from traditional engineering and risk assessment approaches and to foster a broadened understanding of the benefits that risk-informed decision-making can bring to the overall regulatory approach. In response to the EDOs decision, NRR staff has addressed a more comprehensive set of review guidance, as provided in SRP Chapter 15.0.1, to better clarify issues identified in the EDO memo.

The regulation at 10 CFR 50.67(b)(2) is a performance-based rule. It requires a licensee, or applicant, to provide a control room habitability design that meets a specified dose-based criterion. By doing so, the licensee or applicant demonstrates the minimum necessary requirements for structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to public health and safety. To demonstrate compliance, licensees perform traditional deterministic radiological consequence analyses, which are reviewed and approved by the NRC staff. Performance-based regulations do not provide prescriptive methodologies to determine that the regulations are met and, therefore, do not require licensees to use specific designs or methodologies to comply with the regulations.

Rather, performance-based regulations allow licensees and applicants to demonstrate how their particular facility design and licensing basis can meet the specified acceptance criteriain this case, the criteria of 10 CFR 50.67(b)(2). Thus, while NRC RGs and SRPs provide acceptable methodologies that licensees can use to perform the analyses, which are then incorporated as appropriate into the licensing basis for the facility, the use of these guidance documents is optional, and licensees may propose alternative means of complying with the NRCs regulations.

The SRP 15.0.1 guides the NRC staff in reviewing applications with respect to 10 CFR 50.67.

An application to replace the traditional Technical Information Document (TID) 14844, Calculation of Distance Factors for Power and Test Reactor Sites, (TID-14844)-based design-basis accident source term with an AST is acceptable if the plant, as modified, will continue to provide sufficient margin of safety with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameter inputs. SRP 15.0.1 provides guidance for staff reviewers in performing safety reviews to assure the quality and uniformity of staff reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of the reviews.

SRP Chapter 15.0.1,Section II., Acceptance Criteria, states that:

An application to replace the current DBA source term with an AST is acceptable if the plant, as modified, will continue to provide sufficient margin of safety with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameter inputs. The staff should allow licensees to pursue technically justifiable uses of an AST in the most flexible manner compatible with maintaining a clear, logical, and consistent design-basis.

An implementation of an AST is acceptable when the radiological consequences of analyzed design-basis accidents calculated in terms of rem total effective dose equivalent (TEDE) at the exclusion area boundary, the outer boundary of the low population zone, and the control room do not exceed the corresponding acceptance criteria.

Under Section III., Review Procedures, the SRP guides the reviewer to select and emphasize specific aspects of the SRP that are appropriate for the particular application. The review areas to be given attention and emphasis are based on (1) the material presented and its similarity to recently reviewed applications for other plants, (2) the scope of the proposed AST implementation, that is, full or selective, (3) the nature and extent of associated plant modifications, and (4) whether the application is for an initial AST implementation or is based on a previously accepted implementation. NRR staff followed these review procedures and determined:

The material presented by the licensee and supporting documentation is sufficient and that there is ample precedent due to similar recently reviewed applications for other plants requesting to revise their accident source term by adopting 10 CFR 50.67.

That since the scope of the proposed AST is a full implementation, with revisions to several technical specifications, emphasis is placed on all aspects of the application.

That the nature and extent of the associated plant modifications are of particular importance since the associated changes involve safety-related systems that are relied upon, as part of the primary success path, to function, or actuate to mitigate the dose consequences of a design-basis accident.

That the application is for the initial full scope AST implementation, which requires a thorough review of the licensees AST analysis of record and supporting documentation to ensure the licensee satisfies certain basic regulatory requirements such as diversity, redundancy, defense-in-depth, safety margins, and the General Design Criteria of 10 CFR 50, Appendix A, as applicable.

In response to the EDOs direction, NRR staff performed an integrated assessment of the models, assumptions, and parameter inputs provided in the docketed information to ensure that conservative design and licensing basis information was applied, as outlined in various regulatory documents. The staff assessed licensee-proposed alternatives to guidance, with applicable precedents, and found them to be of an appropriate level of conservatism. For significant departures from guidance, the staff performed additional reviews. The staff also accepted the licensees continued use of several previously approved assumptions in the FitzPatrick licensing basis, unless the assumptions were technically inconsistent with the AST or TEDE methodology, or if the use of an assumption in conjunction with the proposed modification created a concern regarding adequate protection of the public health and safety, as it relates to 10 CFR 50.67. Based on its assessment, NRR staff found that the licensee applied sufficiently conservative safety factors within the radiological consequence analysis models.

NRR staff reviewed the original safety evaluation for the FitzPatrick AST LAR to understand how differences between the licensees methods and assumptions and those deemed acceptable to the NRR staff from various guidance documents were evaluated by the staff during the review of the LAR. NRR staff found that these differences were resolved adequately as documented in the LAR, the licenses AST analysis of record, and the original safety evaluation, as demonstrated by the subsequent independent analyses performed by NRR.

Using independent analyses and calculations, NRR staff assessed the licensees margin of safety with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameter inputs. It is NRR staffs continued position that considering safety margin and defense-in-depth is consistent with demonstrating compliance with the regulations, as discussed in SRP 15.0.1. NRR staffs assessments of the margin of safety and defense-in-depth in the licensees analyses are as follows:

Margin of Safety Consistent with SRP 15.0.1, NRR assessed whether there is a sufficient margin of safety within the licensees application of the AST. A margin analysis is a critical step in assessing the level of conservatism in the licensees model. This process systematically evaluates how the variations or sensitivities in each input parameter, and its associated safety factor, can lead to results that fall either above or below the established figures-of-merit or acceptance criteria.

This is a valuable exercise, as it helps quantify the systems robustness and sensitivity to key variables, while also identifying the amount of margin inherent to the model.

When performing such an analysis, it is acceptable for results to occasionally deviate from the established acceptance criteria, provided these deviations are within an acceptable range when balanced with other parameter and modeling conservatisms also assumed in the model. This is because real-world systems often encounter uncertainties, changing conditions, or unexpected events that may lead to degraded performance. By understanding and quantifying the behavior of sensitivities impacting acceptable margins, decision-makers gain valuable insights into the systems behavior and can make informed choices regarding the determination of reasonable assurance of adequate protection. Likewise, the use of a margin of safety analysis allowed the staff to determine an appropriate balance between safety and efficiency with optimized use of resources, ensuring that the licensees AST analysis of record is robust enough to withstand uncertainties or unexpected variations, while avoiding unnecessary overdesign.2 Before discussing the NRR staffs Margin of Safety analysis and results, a discussion of the methodology of traditional design-basis accident radiological consequence analyses is presented. These analyses utilize a system of coupled first-order differential equations to describe the behavior of the source term traveling from the reactor core, through the containment, and other plant systems, to the environment. Mathematically, this approach represents a conservative simplification that is intended to bound the complexities of reality for an engineered system. The models that are used in these analyses can be very simplistic to very complex, ranging from a few dozen to several hundred input parameters. These input parameters not only model the design of the engineered system itself, but other aspects as well, such as meteorological conditions, human performance, and health effects of exposure to ionizing radiation, various physical processes to mitigate the source term, as well as representative testing data. As such, the uncertainties in these models are quite large and, typically, unquantified; however, it is common practice to present results as point estimates in LAR submittals.

A significant consideration in these models is the inclusion of safety factors placed on input parameters to maintain a conservative model and account for uncertainty. The need for safety factors arises from the inherent uncertainties that exist in real-world engineering applications.

These uncertainties can stem from various sources, such as variations in environmental conditions, operator actions, unexpected system failures, or unforeseen events. Safety factors help account for these uncertainties by introducing several layers of conservatism into the analysis results. This conservatism ensures that the system not only meets its intended 2 Analyses performed in support of DPO-2021-001 assessed select input parameters and modeling assumptions from the licensees submittal to support the staffs completion of the margin of safety analysis.

functions but also maintains a high level of reliability to withstand conditions that may exceed original design expectations.

NRR staff reviewed and analyzed the entirety of the licensees docketed submittal, which constitutes its AST analysis of record. The submittal, referenced documents, and updated final safety analysis report contain all the necessary information for the review of the licensees application of the alternative source term.

These supporting documents that are referenced in the material that was provided on the docket include important sources of data, testing, and inspections results, control room operator simulation results, and analyses to develop input parameters into the design-basis accident radiological consequence analysis. Supporting data and test results include meteorological data from onsite meteorological towers, operating procedures to assess operator response timings, and control room habitability tracer gas leak testing to assess the leak-tightness of the control room design. These datasets are of high-quality in which several-to tens-of-thousands of measurements are used by the licensee to derive conservatively biased parameter input assumptions. The staff also assessed the licensees parameter input assumptions based on as-built structures, systems, and components.

NRR staff found the licensee generally followed guidance on the assignment of numerical inputs to values with the objective of determining conservative results. For deviations from guidance, the licensee provided supporting justification, which was assessed by the staff during the original review and documented in the original safety evaluation. NRR staff continues to find deviations from guidance acceptable, where appropriately justified by the licensee and that there is sufficient margin within its AST analysis of record. While the NRC staff acknowledged in its initial review and in the DPO documentation package that some of the licensees assumptions were non-conservative, as noted below, the staff discusses that the safety margin and defense-in-depth in the licensees submittal more than compensate for these areas.

NRR staffs Margin of Safety analysis found that the licensee incorporated appropriate safety factors and generally applied conservative modeling assumptions. The model has inherent layers of conservatism, ensuring the model remains robust and that assumptions for sensitive key variables with large uncertainties are compensated. The amount of margin is best illustrated with an example that provides a discussion of how the licensee developed and utilized the unfiltered in-leakage input parameter.

Example: Applied Input Parameter Safey Factors and Assessment of Margin The design-basis accident radiological consequence analysis models include a parameter to account for unfiltered in-leakage into the control room. This parameter is based on the facilities Control Room Envelope Habitability Program, which is controlled by the technical specifications.

The licensees control room leak-tightness is controlled by Technical Specification 5.5.14, Control Room Envelope Habitability Program, which has a limit of 100-cfm. The licensee described the results of their tracer gas testing covering 14 years of operation from 2004 through 2018 as 0-, 17(+/-4)-, and 20(+/-4) cfm, respectively. Therefore, the control room has been found to be essentially leak tight from unfiltered in-leakage during emergency ventilation mode based on tracer gas test results.

The licensee applied a safety factor of four to the 100 cfm Technical Specification limit, based on the assumption of an additional Single Failure of a motor-operated valve.

Based on this assumption, the licensee conservatively applies an additional 300 cfm to the unfiltered in-leakage input parameter. The licensee, therefore, very conservatively combined the leakage under a failed condition (300 cfm) with the leakage under an un-failed condition (100 cfm). Removing this particular margin decreases the control room rem TEDE results by a factor of ~ 4. This margin alone more than compensates for the issues noted in the DPO Appeal Panel report. The licensees AST analysis of record margin also compensates for the licensees use of an elemental iodine removal constant greater than specified in SRP Section 6.5.2, Containment Spray as a Fission Product Cleanup System, which impacted the final result by a small fraction, among other modeling assumptions with associated uncertainties.

The NRR staff assessed the unfiltered in-leakage parameter in relation to the regulatory guidance mentioned above, as well as compared it to relevant precedents. The NRR staff found that the applied safety factor, which is based on the inclusion of an additional component failure, amounts to multiple failures beyond the Single Failure criteria and exceeds the single active component failure assumption regulatory position that is used for design-basis accidents analyses. The subsequent NRR staffs evaluation found licensing precedents for licensees to include such additional margin within their AST analysis of record to provide operational flexibility.

Consistent with many other AST safety evaluations completed for similar types of LARs for other plants, the staffs original safety evaluation did not include specific safety margin discussions throughout the licensees AST analysis of record and their impact on the final results. The SRP Section IV, Evaluation Findings, discusses that evaluation findings of acceptability provide a consistent, scrutable basis that is derived from the information submitted by the licensee on the docket. As part of the NRR staffs subsequent integrated assessment of the models, assumptions, and parameter inputs used by the licensee, through the NRR staffs margin analysis, the staff concluded that the licensee satisfies the regulatory requirements of including sufficient safety margin in its analysis and that there is reasonable assurance that the licensee is in compliance with 10 CFR 50.67.

Defense-in-Depth Consistent with SRP 15.0.1, the NRR staff assessed whether there is adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameter inputs. The staff assessed the modeling penalty assumptions by the licensee. Typically, these modeling penalty assumptions conservatively credit only those accident mitigation features that are classified as safety-related, required to be operable by technical specifications, powered by emergency power sources (assuming a loss-of-offsite-power), and either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in the emergency operating procedures.

The NRR staff applied the Single Failure Criterion as it serves well as a licensing review tool to promote and assure system reliability as one element of the defense-in-depth approach to reactor safety. Application of the Single Failure Criterion involves the systematic search for potential Single Failure points and their effects on prescribed missions and achievement of safety functions.

The NRR staff has found the licensee assumed several single active component failures, and a passive pipe failure, in its radiological consequence analysis. Except for operator actions, all of these active component failures are assumed to occur at the beginning of the accident in the calculation.

The licensees assumptions include:

Primary Single Failure:

One of four inboard MSIVs in a main steam line failing to close. This failure complies with a single active component failure requirement that results in the most limiting radiological consequences (RG 1.183, Section 5.1.2).

Additional Examples of the Licensees Single Failure Assumptions Include the Following Examples:

Loss of a division of emergency power, which results in the failure of one control room emergency ventilation air supply system filter train, one standby gas treatment system filter train, and one drywell spray pump.

Crediting only the lower containment spray nozzle elevation to mitigate 20% of the containment atmosphere volume, when it is expected that 90% of the cross-sectional area would be sprayed, if the upper spray elevation was also credited.

Failure of a motor-operated valve, while in the control room emergency ventilation mode, resulting in additional unfiltered in-leakage (as discussed above).

Passive failure of the reactor recirculation pump, consistent with the 10 CFR 50.46 loss-of-coolant accident (LOCA) analysis, defined as the instantaneous guillotine rupture of the recirculation pipe with displacement of both ends of that blowdown that occurs from both ends.

The remaining three main steam lines that are not assumed to have the primary single failure are modeled with the outboard main steam isolation valves open (i.e., assumed to fail to close). While this was done for modeling simplicity, this assumption results in less credit for source term holdup.

Modeling only two of the four main steam lines for crediting aerosol deposition. This was done for modeling simplicity and to accommodate for modeling limitations of the NRC code itself used to perform the analyses, resulting in less credit for source term deposition. The licensee explains that the remaining two intact steam lines are assigned a leakage of 0 scfh in the analysis. These remaining two intact steam lines are therefore not explicitly modeled in the radiological consequence analysis but are represented by the leakage through the second shortest main steam line. In conducting this modeling, NRR staff notes that the licensee simplified its RadTrad model, which inherently includes analytical margin within the calculation that does not fully credit fission product mitigation for safety-related and seismically qualified piping. This is an additional conservatism, which is allowed by RG 1.183.

These defense-in-depth, or conservative, modeling assumptions impact results by increasing the radiological consequences, thus reducing the calculated margin to the regulatory acceptance criteria. Limiting the number of failure assumptions to reasonably reflect the facility as designed would significantly lower calculated radiological consequences, resulting in more margin to the criteria.

The staffs original safety evaluation did not discuss specific defense-in-depth modeling assumptions that the licensee made throughout its AST analysis of record and their impact on the final results. The staffs subsequent evaluation found that the information submitted by the licensee on the docket was sufficient to demonstrate that there is reasonable assurance, supported by risk and engineering insights, that the licensees estimates for the exclusion area boundary, low population zone, and control room results comply with the established acceptance criteria. This conclusion is supported by NRR staffs integrated assessment of the models, assumptions, and parameter inputs used by the licensees evaluation, where the staff determined that the licensee satisfied the regulatory requirements of including sufficient diversity, redundancy, and defense-in-depth, which provides adequate safety margin in its analysis. Overall, the staff determined that the licensee has effectively integrated the appropriate levels of defense-in-depth into its AST analysis of record to ensure reliability of the control room design. Additionally, the NRR staff has found the assessed safety functions of the FitzPatrick design demonstrate that multiple independent and redundant layers of defense are available to compensate for potential human and mechanical failures so that no single layer, no matter how robust, is exclusively relied upon.

Section 2.2-Packing Leakage In the Case File for DPO-2021-001, the EDO concluded that:

The LAR review process did not adequately address potential leakage from the MSIV packing. Control of packing leakage from the outboard MSIV was a safety function described in the FitzPatrick licensing basis, but an assessment of the impact of the removal of the Main Steam Leakage Collection System (MSLCS) on this function was not provided by the licensee nor documented in the staffs safety evaluation. In addition, outboard MSIV packing leakage into the steam tunnel area represents a potentially unfiltered release path to the environment.

The NRR staff has assessed this issue and is responding with two discussions regarding packing leakage to supplement the staffs original safety evaluation. The first is regarding the regulations and operational programs that address packing leakage pursuant to 50.54(o) and 10 CFR 50 Appendix J, which controls primary reactor containment leakage and testing. The second is regarding the regulations and regulatory guidance for performing the design-basis radiological consequence analysis pursuant to 10 CFR 50.67 and RG 1.183, Revision 0, which sets forth the acceptance criteria for the control room and guidance on performance DBA radiological consequence analyses. This regulation and RG supersede previous regulatory guidance analysis assumptions that were once found necessary but are no longer necessary to ensure adequate protection.

With respect to compliance with 10 CFR 50.67 and RG 1.183, Revision 0, NRR staff has assessed this issue and determined that the licensees assessment of the MSIV leakage pathway is facility-dependent due to the seismic qualifications of the downstream main steam line piping. RG 1.183, Revision 0, treats MSIV leakage, based on the technical specification leakage, as an unmitigated atmospheric release at the valve location for non-seismically qualified piping. This conservative approach discounts the effect of the non-safety-related main steam piping, the steam tunnel enclosing the outboard main steam isolation valve, and the turbine building ventilation on the release characteristics to the environment. In essence, in the generalized sense, leakage is assumed to occur at the valve location, which is bounding.

However, the FitzPatrick design is different because the downstream main steam line piping is safety-related and seismically qualified to function. Therefore, the licensee modeled MSIV leakage through the MSIV seat to travel through the seismically qualified main steam line piping to the turbine stop valve location. The staff found that the licensee appropriately reflected the FitzPatrick design in its AST analysis of record, pursuant to 10 CFR 50.36 and 10 CFR 50.34, in support of its 10 CFR 50.67 application.

Discussion Item 1: 10 CFR 50.54(o) Regulations and Operational Programs Valve packing leakage is a well-understood and acknowledged characteristic of valve design.

10 CFR 50.54(o) requires that primary reactor containments for water cooled power reactors be subject to the requirements of Appendix J to 10 CFR Part 50. Appendix J specifies the leakage test requirements, schedules, and acceptance criteria for tests of the leak tight integrity of the primary reactor containment and systems and components, which penetrate the containment.

Normal valve maintenance, combined with leak testing, in accordance with the technical specifications of 10 CFR 50.36, Technical Specifications, ensures continued performance throughout the plant life. There is no regulatory requirement in Appendix J to specifically assess leakage through valve stem packing.

To understand the significance of MSIV leakage, it is important to understand how the Appendix J tests are performed. Two isolation valves are provided in series in a horizontal run of each main steam line, as close as practical to the primary containment, one inside containment (inboard) and the other outside (outboard). The valves, when closed, form part of the primary containment barrier for the Reactor Coolant System breaks inside the containment and part of the Reactor Coolant Pressure Boundary for main steam line breaks, outside the primary containment. Steam flows through the inboard valve and then the outboard valve. This flow would work in conjunction with the valve actuator to close and seat (seal) the valve.

Appendix J states that before testing leak tightness, no repairs or adjustments shall be made so that the containment can be tested as close to as is conditional as practical. Thus, when the leak tightness of the valves is tested, the valves are closed without any other treatment.

During testing, the main steam line between the MSIV is pressurized to containment design pressure or another accepted value and held for a period of time. The pressure acts on the stem side of the outboard valve seat disk, in the same direction as steam flow, which increases seating pressure. However, the pressure acts under the seat of the inboard valve, which conservatively decreases seating pressure and is counter to the direction of steam flow. Thus, the pressure is working to open the inboard value. This opening pressure on the inboard valve, in conjunction with a small allowable leakage rate, can result in technical specification leakage limit violations. Also, there are other connections between the MSIV,s which can also leak.

While some licensees have made changes to improve the tested conditions, e.g., closing the valves under some steam flow, replacing the valve stem with a stouter one, and increasing actuator pressure, the testing still involves pressurizing the opposite side of the valve, as well as the other connections. As such, it is very difficult to test the leak tightness of each MSIV under conditions representative of post-accident differential pressures. Therefore, the total test leakage results represent a very conservative estimate of total main steamline leakage.

Considering the conservative contribution of inboard valve leakage and other connections to the overall main steamline leakage value determined by the test, it is difficult to correctly apportion leakage values among the inboard valve seat, the outboard valve seat, and the outboard valve stem. For FitzPatrick, tests are performed in accordance with Technical Specification 3.6, Containment Systems, Surveillance Requirement 3.6.1.3.10. The Technical Specification limits the combined leakage value per line, without partitioning between the two valves, their seats, and the stems. This testing practice is generally consistent with other similar BWR reactor designs. Since Appendix J testing does not partition the leakage between the possible flow paths (i.e., stem packing leakage or valve seat leakage), the staffs action of determining what fraction leaks through the packing as opposed to the valve seats, as described by DPO 2021-001, would be a new regulatory position.

Additionally, the staffs review of current operational experience has not demonstrated an issue with these valves that would indicate a modern generic issue. NRR staff has found that current operating experience, routine maintenance practices, and the licensees AST analysis of record support the information in the licensees LAR for the respective Technical Specification leakage limits. The information detailed in the discussion items below support the NRR staffs finding that the licensees AST analysis of record is conservative and consistent with the regulatory guidance, and thus compliant with 10 CFR 50.67.

Discussion Item 2: 10 CFR 50.67 and Regulatory Guidance Superseding Previous Assumptions The purpose of this LAR, in part, was for the licensee to obtain the staffs approval to adopt the AST, in accordance with 10 CFR 50.67 and RG 1.183. As discussed above, 10 CFR 50.67 allows licensees to voluntarily replace the traditional TID-14844 source term with an AST and update the acceptance criteria from 10 CFR Part 100, Reactor Site Criteria. As part of adopting 10 CFR 50.67, RG 1.183 supersedes the licensees original licensing basis assumptions for calculating design-basis accident radiological consequences.

As discussed in the AST analysis of record submittal, the licensee requested removal of the Main Steam Leakage Collection (MSLC) system from the technical specifications when adopting 10 CFR 50.67 and, in part, the RG 1.183 assumptions. The FitzPatrick licensing basis utilizes a combination of two outdated regulatory guidance documents to perform the design basis radiological consequence analyses. The first is TID-14844 and the second is RG 1.3, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, published in June 1974, for evaluating the potential radiological consequences for boiling water reactors.

Consistent with its licensing basis, the licensee processes the effluent from the MSLC via the Standby Gas Treatment System and then it is exhausted through the stack. This is consistent with the RG 1.96, Revision 1, Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants, published June 1976. RG 1.96 describes a basis acceptable to the NRC staff for implementing General Design Criterion 54 regarding the design of a LCS for the MSIVs of boiling water reactor nuclear power plants to ensure that total site radiological effects do not exceed the 10 CFR Part 100 limits. RG 1.96 notes that a reduction and control of steam packing leakage or other direct leakage to the steam tunnel from the outboard isolation valve should be a design objective of the leakage control system or of other systems provided for this purpose. RG 1.96 describes a basis acceptable to the NRC staff for implementing General Design Criterion 54 with regard to the design of a leakage control system for the main steam isolation valves of boiling water reactor nuclear power plants to ensure that total site radiological effects do not exceed guidelines of 10 CFR Part 100.

RG 1.183, Section B, Discussion, provides assumptions and methods acceptable to the NRC staff for performing design-basis radiological analyses using an AST. Specifically, providing guidance that it supersedes corresponding radiological analysis assumptions provided in other regulatory guides and SRP chapters when used in conjunction with an approved AST and the TEDE criteria provided in 10 CFR 50.67. It specifies RG 1.3 as an affected Regulatory Guide.

Therefore, when adopting 50.67 and utilizing RG 1.183, the original TID-14844 accident source term and corresponding radiological analysis assumptions of RG 1.3 are superseded.

10 CFR 50.67 has no specific regulatory requirement to model MSIV leakage through the valve stem packing. Additionally, 10 CFR 50.36 requires that technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. The analysis and evaluations provided by the licensee, in its AST analysis of record, modeled MSIV leakage to travel through the seismically qualified main steam line piping to the turbine stop valve location, then released. As discussed in the LAR, this is consistent with the FitzPatrick Updated Final Safety Analysis Report, Section 4.11.4, which provides additional details in that the main steam lines up to the turbine stop valves have been dynamically analyzed and meet the requirements of a Class I Seismic System. This main steam line qualification ensures the main steam lines are available during a LOCA, including one effectively initiated by a seismic event, where the turbine building structure and support systems could be compromised, and the release occurs directly from the turbine stop valve to the environment. Therefore, the licensee reflected the FitzPatrick design in its analysis and evaluations, pursuant to 10 CFR 50.36 and 10 CFR 50.34, in support of its 10 CFR 50.67 application, as treating the release of MSIV leakage to occur at the turbine stop valve location. This approach is consistent with RG 1.183, Appendix A, Section 6.5, which allows for credit for all main steam lines in the MSIV leakage release pathways. since they are seismically designed and supported to withstand the Safe Shutdown Earthquake. Specifically, RG 1.183, Appendix A, Section 6.5 states:

A reduction in MSIV releases that is due to holdup and deposition in main steam piping downstream of the MSIVs and in the main condenser, including the treatment of air ejector effluent by off gas systems, may be credited if the components and piping systems used in the release path are capable of performing their safety function during and following a Safe Shutdown Earthquake (SSE)

For this reason, NRR staff continues to find the licensees presented AST analysis of record to be conservative and consistent with the regulatory guidance and thus compliant with 10 CFR 50.67, as the main steam line piping is seismically qualified to the turbine stop valve location and that previous licensing basis assumptions and methods have been appropriately superseded. NRR continues to apply modern licensing policies and procedures, which may supersede previous regulatory guidance analysis assumptions that were once found necessary but are no longer necessary to ensure adequate protection.

In addition, NRR staff surveyed past LARs and safety evaluations for boiling water reactor related MSIVs and associated leakage control systems. The purpose of this survey was to understand precedence when assessing the removal of the leakage control systems as addressed by other licensees and the staff. To the extent of the review, there was no indication that the NRC staff requested this assessment, in the context of the approved LARs for removal of a leakage control system.

Further, the staff notes that requiring an assessment of the impact of packing leakage for a 10 CFR 50.67 analysis, regardless of characteristics of a specific facility, would be a new regulatory position.

Section 2.3-Limiting Break Location In the Case File for DPO-2021-001, the EDO concluded that:

[t]he AST amendment did not consider the most limiting LOCA break location.

NRR staff reviewed the basis for the limiting LOCA break location.

Consistent with precedents set forth by previously approved LARs for 10 CFR 50.67, the licensee also assumed a deterministic passive failure of piping referred to as the limiting break location, within its design-basis accident radiological consequence analysis. This additional assumption is not a regulatory position in RG 1.183 Rev. 0 and is beyond design-basis because this piping is seismically qualified. The licensee acknowledged in its AST analysis of record that its assumption of the limiting break location does not result in the highest dose results but is consistent with its licensing basis LOCA analysis. Additionally, NRR finds from the licensees sensitivity study, and with consideration of the safety margins and defense-in-depth assessments, as discussed in Section 2.1, that the impact of the limiting break location on the licensees AST analysis of record will result in a small fractional change and compensated for by its inherent conservatism.

The licensee derived its limiting break location based on Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, which defines LOCAs as those postulated accidents that result from a loss-of-coolant inventory at rates that exceed the capability of the reactor coolant makeup system. Therefore, the licensee assessed the radiological consequences based on this description of a large-break LOCA for evaluating the performance of release mitigation systems and the containment response. The licensing bases for the selection of the break location are from the FitzPatrick Updated Final Safety Analysis Report, Section 6.5.3.1, Liquid Line Breaks, which defined the design-basis accident as the instantaneous guillotine rupture of the recirculation pipe with displacement of both ends so that blowdown occurs from both ends. Also, postulating a main steam line break upstream of the inboard MSIV during a LOCA is considered another design basis accident event of a main steam line break inside the containment, which is neither credible nor feasible based on the qualification of the main steam lines.

The licensees AST analysis of record assumed all four main steam line piping sections between the reactor pressure vessel nozzle and the turbine stop valves remain intact and is capable of performing their safety function during and following a Safe Shutdown Earthquake.

This assumption is consistent with RG 1.183, Appendix A, Regulatory Position 6.5, which allows for credit, if the components and piping systems used in the release path are capable of performing their safety function during and following a Safe Shutdown Earthquake downstream of the MSIVs. The licensee assumes the active failure of the inboard MSIV in one of the shortest main steam lines to close and subsequently remain open during the accident. This complies with a single active component failure requirement that results in the most limiting radiological consequences, described in RG 1.183, Regulatory Position 5.1.2. As modeled, the licensees AST analysis of record only credited deposition within two main steam lines as opposed to four, further demonstrating additional margin to safety and defense-in-depth. The NRC staff found this assumption conservative, since the licensee takes less credit for aerosol deposition mitigation than what is otherwise allowable.

The licensee adjusted its model to overcome limitations with the now outdated version of the NRC code RadTrad (Version 3.03). RadTrad Version 3.03 is the software that is used to perform radiological consequence calculations. Previous versions of the software limit the users ability to more accurately model the facility as designed and necessitating several simplifying modeling assumptions. The code was developed for the NRC staff and is used by applicants in licensing reviews. Subsequent versions of the code have improved on these limitations but did not fundamentally change the models. Consistent with SRP Chapter 15.0.1, the staff compared the results estimated by the licensee to the results estimated by the staff in its confirmatory calculations based on docketed information submitted by the licensee. NRR staff finds these simplifying modeling assumptions result in a large amount of analytical margin embedded in the licensees submitted AST analysis of record.

During the original NRC staff LAR review, at the request of the staff through a request for additional information (see response at ADAMS Accession No. ML20090E279), the licensee performed additional sensitivity analyses by modeling a main steam line break as opposed to a break in the recirculation line. The staff requested this analysis since the licensee acknowledged in its AST analysis of record submittal that postulating a main steam line break in one main steam line inside the drywell would maximize the radiological consequence. This is irrespective of Regulatory Guide 1.183 Rev. 0, which does not request licensees to assume, or assess, a passive pipe break.3 To quantify the impact of assuming a main steam line break on the resulting radiological consequences, the licensee adjusted its AST analysis of record model in a sensitivity analysis to explicitly model all four main steam lines to credit these additional design features not modeled in the original LAR submittal with the simplified steam line modeling. As discussed by the licensee, this was done to overcome the limitations of the RadTrad code (Version 3.03). Also, the licensee took no credit for aerosol deposition or elemental iodine plate out in the line segments assumed to break from the reactor pressure vessel nozzle to the inboard MSIV. Results of the sensitivity analysis demonstrate that the AST analysis of record for the recirculation line break has sufficient margin and bounds the two cases of a main steam line break, based on a more representative model of the FitzPatrick design-basis. As these analyses were done for the purpose of a sensitivity analysis, the use of assumptions that were not consistent with the licensing basis was sufficient.

NRR staff finds the sensitivity analysis is consistent with RG 1.183 Revision 0, Section 1.3.3, Use of Sensitivity, or Scoping Analysis. As discussed in the guide, the use of a sensitivity study is to demonstrate that existing analyses have sufficient margin and need not be recalculated. As used in this guide, a sensitivity analysis is an evaluation that considers how the overall results vary as an input parameter is varied. The guide states that [w]here present, arbitrary designer margins may be adequate to bound any impact of the AST and TEDE criteria. The staff finds that the licensees sensitivity analysis is consistent with RG 1.183 Revision 0, as appropriate.

NRR staff reviewed previously approved LARs for 10 CFR 50.67 with respect to the limiting break location to understand precedence and found that most licensees have been approved for assuming the limiting break location to be based on 10 CFR 50.46 Emergency Core Cooling System LOCA analysis, which is a recirculation line.4 3 To comply with precedents set forth by previously approved LARs for 10 CFR 50.67, licensees would assume a deterministic passive failure of piping referred to as the limiting break location, within their design-basis accident radiological consequence analysis. This additional assumption of another single failure is not a regulatory position in RG 1.183 Revision 0 and is beyond design-basis for FitzPatrick.

4 RG 1.183 Revision 1 now applies the MHA concept (based on the DPO-2020-002), where a licensee no longer needs to assume a specific pipe break scenario when analyzing the MHA.

Section 2.4-Aerosol Deposition Credit for the Main Condenser In the Case File for DPO-2021-001, the EDO concluded that:

There is a lack of appropriate regulatory documentation and justification for crediting the power conversion system for aerosol deposition. Crediting aerosol deposition in the condenser, a capability not requested by the licensee, was necessary to address known non-conservatism in the licensees analysis supporting the AST LAR. Absent this credit, the staffs evaluation is not sufficient to demonstrate compliance with the control room dose limit.

The NRR staff reviewed the documentation regarding the issue pertaining to aerosol deposition credit for the main condenser. Credit of the condenser is not necessary to determine compliance with 10 CFR 50.67 due to the licensees applied safety margin and defense-in-depth modeling assumptions. The licensee did not credit aerosol deposition for the condenser in its LAR submittal to demonstrate compliance with 10 CFR 50.67. During the original staff LAR review, at the request of the NRC staff through a request for additional information (see response at ADAMS Accession No. ML20090E279), the licensee performed additional sensitivity analyses due to uncertainties with gravitational settling credit in the main steam lines in association with credit for containment sprays. The sensitivity analysis evaluated the impact of sprays on the aerosol settling velocity and was used to identify other inputs with well-defined uncertainty or conservatism that could be used to offset the uncertainty associated with the current aerosol deposition model.

One of these sensitivity analysis cases considered the impact of potential aerosol removal in the condenser on the licensees AST analysis of record. In its response to the staff, the licensee stated that these modeling assumptions are not credited in its AST analysis of record submittal.

As previously discussed, the licensee adjusted its model to overcome limitations in the endorsed outdated version of the NRC code RadTrad (Version 3.03). Adjustments were also made to the parameters describing the aerosol particle size distribution that have been typically applied to other BWR-type AST reviews and endorsed by the NRC through Regulatory Issue Summary, RIS 2006-04, NRC Regulatory Issue Summary 2006-04, Experience with Implementation of Alternate Source Terms, published on March 7, 2006, to account for non-conservatisms in the aerosol model.

For this sensitivity case, the leakage is assumed to travel to the condenser through the drain lines from the main steam line piping between the MSIVs. This case conservatively neglects any holdup and deposition in the outboard main steam line piping, which is a genuine, measurable effect. The licensee explained the approach to modeling the condenser was generally consistent with precedent for other plants in the Exelon fleet (e.g., LaSalle and Limerick). The licensee also provided additional context with operating experience associated with the North Anna earthquake and post-Fukushima reevaluated hazard evaluations in which components and piping systems typically used in this release path are sufficiently rugged to ensure they are capable of performing some level of radioactivity removal during and following a Safe Shutdown Earthquake. The net effect produced from this sensitivity case resulted in offsite and control room radiological consequence results by a factor of about 3.5 lower than the licensees submitted AST analysis of record.

The NRC staff considered risk and engineering insights, as addressed in Section 3.5, NRC Staff Risk and Engineering Insights, of the staffs original safety evaluation based, in part, on the licensees sensitivity analysis. As with Section 2.3 of this addendum for the limiting break location, these analyses were conducted for the purpose of a sensitivity analysis and found to be sufficient. The conclusion of this addendum is not dependent on this sensitivity case. The additional information discussed above provides clarifying information of potential plant-specific margins.

As discussed above, the licensee applied appropriate safety factors throughout its analysis with the inclusion of several defense-in-depth modeling assumptions, including additional component failures beyond the Single Failure criteria. However, consistent with the staffs original evaluation, the licensee's sensitivity analysis results demonstrate that a BWR condenser can be very effective in substantially reducing the radiological consequences from MSIV leakage.

3.0 CONCLUSION

NRR staff has reviewed its original safety evaluation for the AST implementation LAR proposed by the licensee for FitzPatrick. NRR staff also reviewed the proposed plant modifications associated with this proposed AST implementation. In performing this review, the staff relied upon information placed on the docket by the licensee, staff experience in performing similar reviews and, where deemed necessary, on confirmatory calculations. Additionally, NRR staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological impacts of the proposed plant modifications in the context of the proposed AST.

NRR staff finds that the licensee used analysis methods and assumptions consistent with the conservative guidance of RG 1.183, with the exceptions discussed, and accepted in the NRC staffs original safety evaluation. NRR staff also finds the methods and assumptions used by the licensee to be in compliance with applicable requirements. The staff compared the results estimated by the licensee to the applicable acceptance criteria and to the results estimated by the staff in its confirmatory calculations. NRR staff finds with reasonable assurance that the licensees estimates of the total effective dose equivalent due to design-basis accidents will comply with the requirements of 10 CFR 50.67 and the guidance of RG 1.183.

NRR staff finds reasonable assurance that the licensee will continue to provide sufficient safety margins with adequate defense-in-depth for FitzPatrick, which will adequately address unanticipated events and compensate for uncertainties in accident progression and analysis assumptions and parameters at FitzPatrick. Therefore, the staff concludes that the proposed AST implementation and the associated plant modifications are acceptable.

Principal Contributors: E. Dickson, NRR M. Hart, NRR Date: October 9, 2024

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