ML25266A243

Enclosure 1 - Safety Evaluation for the TRISO Fuel Design Methodology Topical Report for the Evinci Microreactor Non-Proprietary

ML25266A243
Person / Time
Site:	99902079
Issue date:	12/19/2025
From:	Alissa Neuhausen NRC/NRR/DANU/UAL2
To:	Westinghouse
Vechioli-Feliciano L
Shared Package
ML25266A244	List: ML25266A241 ML25266A243
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Download: ML25266A243 (0)

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OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION (1)(i) The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof; (ii) Interdependent effects among the safety features of the design are acceptable, as demonstrated by analysis, appropriate test programs, experience, or a combination thereof; and (iii) Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses over a sufficient range of normal operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditions; or (2) There has been acceptable testing of a prototype plant over a sufficient range of normal operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditions. If a prototype plant is used to comply with the testing requirements, then the NRC may impose additional requirements on siting, safety features, or operational conditions for the prototype plant to protect the public and the plant staff from the possible consequences of accidents during the testing period.

Paragraph 52.47(a)(2)(iv) of 10 CFR requires:

(a) The application must contain a final safety analysis report (FSAR) that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole, and must include the following information:

(2) A description and analysis of the structures, systems, and components (SSCs) of the facility, with emphasis upon performance requirements, the bases, with technical justification therefore, upon which these requirements have been established, and the evaluations required to show that safety functions will be accomplished. It is expected that the standard plant will reflect through its design, construction, and operation an extremely low probability for accidents that could result in the release of significant quantities of radioactive fission products. The description shall be sufficient to permit understanding of the system designs and their relationship to the safety evaluations. Such items as the reactor core, reactor coolant system, instrumentation and control systems, electrical systems, containment system, other engineered safety features, auxiliary and emergency systems, power conversion systems, radioactive waste handling systems, and fuel handling systems shall be discussed insofar as they are pertinent. The following power reactor design characteristics will be taken into consideration by the Commission:

(iv) The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Special attention must be directed to plant design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable postulated site parameters, including site meteorology, to evaluate the offsite radiological consequences. The evaluation must determine that:

A) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE);

B) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE; Similarly, 10 CFR 52.157(d) is relevant for an ML application, 10 CFR 50.34(a)(1)(ii)(D) is relevant for a CP application, and 10 CFR 52.79(a)(2)(iv) is relevant for a COL application.

The NRC guidance documents that are applicable to the review of this TR are described below.

DANU-ISG-2022-01, Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications - Roadmap (ML23297A158).

Regulatory Guide (RG) 1.203, Transient and Accident Analysis Methods, (ML053500170) provides the evaluation model (EM) development and assessment process (EMDAP) as an acceptable framework for developing and assessing EMs for reactor transient and accident analyses.

RG 1.253, Guidance for a Technology-Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors (ML23269A222).

NUREG-2246, Fuel Qualification for Advanced Reactors, (ML22063A131) discusses a framework for use in qualification of nuclear fuels. The framework discusses the identification of key fuel manufacturing parameters, the specification of a fuel performance envelope to inform testing requirements, the use of EMs in the fuel qualification process, and the assessment of the experimental data (ED) used to develop and validate models and empirical safety criteria. The framework outlines a set of goals that, when met, can be used to justify that a nuclear fuel design is qualified for use.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Principal Design Criteria Westinghouse TR EVR-LIC-RL-001-P/NP-A, Revision 1, Principal Design Criteria Topical Report, (ML24353A097) dated December 17, 2024, provides principal design criteria (PDC) for the eVinciTM Microreactor design that were reviewed and approved by the NRC staff in the associated safety evaluation (SE) (ML24283A133). The PDCs below are identified as relevant per TR EVR-LIC-RL-003-P, Westinghouse TRISO Fuel Design Methodology Topical Report:

PDC 1, Quality Standards and Records - Safety significant structures, systems, and components (SSCs) shall be designed, fabricated, erected, and tested to quality standards commensurate with the safety significance of the functions to be performed.

Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the safety significant function. A quality assurance program (QAP) shall be established and implemented in order to provide reasonable assurance that these SSCs will satisfactorily perform their safety significant functions. Appropriate records of the design, fabrication, erection, and testing of safety significant SSCs shall be maintained by or under the control of the nuclear power unit licensee for an appropriate period of time.

PDC 10, Reactor Design - The reactor system and associated heat removal, control, and protection systems (along with any SSCs supporting the reactor system and associated heat removal, control, and protection systems safety function(s)) shall be designed with appropriate margin to ensure that specified acceptable system radionuclide release design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).

PDC 12, Suppression of Reactor Power Oscillations - The reactor core; associated structures; and associated coolant, control, and protection systems shall be designed to ensure that power oscillations that can result in conditions exceeding specified acceptable system radionuclide release design limits are not possible or can be reliably and readily detected and suppressed.

PDC 16, Functional Containment - A functional containment shall be provided to control the release of radioactivity to the environment and to ensure that the safety significant functional containment design conditions are not exceeded for as long as licensing basis event (LBE) conditions require.

PDC 26, Reactivity Control - Reactivity control shall be provided. Reactivity control shall provide:

(1) A means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the specified acceptable system radionuclide release design limits and the reactor helium pressure boundary design limits are not exceeded and safe shutdown is achieved and maintained during normal operation, including AOOs.

(2) A means, which is independent and diverse from the other(s), shall be capable of controlling the rate of reactivity changes resulting from planned, normal limits and the reactor helium pressure boundary design limits are not exceeded.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION (3) A means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the capability to cool the core is maintained and a means of shutting down the reactor and maintaining, at a minimum, a safe shutdown condition following a LBE.

(4) A means for holding the reactor shutdown under conditions that allow for interventions such as fuel loading, inspection, and repair.

PDC 64, Monitoring Radioactive Releases - Means shall be provided for monitoring the functional containment performance, effluent discharge paths and facility environs for radioactivity that may be released from normal operations and LBEs.

TECHNICAL EVALUATION Scope of NRC Review Westinghouse intends to follow NUREG-2246 for accelerated fuel qualification, relying on modeling and simulation to inform fuel performance where deviations in the fuel design depart from the historic irradiation testing. Westinghouse states that the methodology for the modeling and simulation follows the EMDAP outlined in RG 1.203.

As requested in TR section 1.4, Request for NRC, this SE covers the entire Westinghouse TRISO fuel design methodology for the eVinciTM microreactor core under normal operation, AOOs, and design basis accidents (DBAs) with the understanding that further development and licensing activities will occur. Westinghouse has identified key elements to software and analysis methods described in TR section 4.1, Bison Governing Equations, section 4.2, Bison TRISO Fuel Particle Material Properties, section 4.3, Bison TRISO Physical Models, and section 5.0, TRISO FUEL PERFORMANCE ANALYSIS METHODOLOGY,; the code validation and verification plan described in TR section 4.4, Verification and Validation Plan; the uncertainty quantification plan described in TR section 4.5, Uncertainty Quantification; and the transient fuel capsule testing described in TR section 3.5, Transient Fuel Capsule Testing.

The SE is structured in alignment with TR table 1.3-1, Mapping of NUREG-2246 Goals, wherein each goal from NUREG-2246 is cross-referenced to the section(s) or information in the TR that is intended to support the accomplishment of that goal. The information provided by Westinghouse in TR section 6.0, Technical Specifications Framework and Failed Fuel Monitoring, and appendix A, Physical Constants and Conversion Factors, was also considered by the NRC staff to inform aspects of the review but is not part of the NRC staffs evaluation. The methods described in the TR are preliminary and V&V of these models has not been completed at the time of this review. As such, the NRC staff imposes Limitation 1, which withholds staff findings on the sufficiency of the results from testing, modeling, and analyses performed in accordance with the methodology to demonstrate conformance with PDCs or regulations. The NRC staff will review those, as requested, as part of future TRs or license applications. The NRC staff further imposes Condition 1, which requires an applicant referencing this TR to provide justification that parameters affecting the fuel performance, including the composition, dimensions, and operating envelope are within the scope of those to which the methodology can be applied and its results validated.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION addressed, as requested by Westinghouse, in the review of future submittals, consistent with Limitation 1 and Condition 1.

G1.3 - End State Attributes G1.3 states that the final attributes of the manufactured fuel should be specified or otherwise justified. These attributes may include microstructure, thicknesses, sphericity, coating coverage, or phase composition of the fuel after fabrication. Examples of how historic TRISO fuel has defined end state attributes can be found in Electric Power Research Institute (EPRI)-AR-1(NP)-

A, Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance (ML20336A052). TR section 3.2.2, TRISO Fuel Compact Specification, discusses nominal TRISO fuel compact parameters including ((

)) TR section 3.3 discusses fuel manufacturing quality control. Westinghouse is using an end state specification adherence that will rely upon out-of-pile testing to ensure quality, rather than a process-based adherence. This testing includes ((

.)) Westinghouse identifies the primary manufacturing defects and plans to assess each in a statistical manner through sampling and analysis of batches. The identification of end state attributes and the end state testing meets goal G1.3 of NUREG-2246 through a combination of specification and justification. NRC staff makes no findings on the qualification of fuel with these specific end state attributes, nor on the conclusions to be drawn from the tests, but these will be addressed in the NRC staffs review of future submittals, consistent with Limitation 1 and Condition 1.

G2 - Safety Criteria The elements of G2, Safety Criteria, involve design limits and the performance of the fuel under both normal and accident conditions to assess how safety criteria are satisfied. To adequately assess safety criteria, fuel failure mechanisms should be clearly defined and understood. TR section 3.4.1, Failure Mechanisms, recounts fuel failure mechanisms from PNNL-31427 in TR table 3.4-1, A Brief Summary of Failure Mechanisms. The elements of NUREG-2246 goal G2 are to ensure a margin to design limits can be demonstrated under normal operation and AOOs (G2.1, Design Limits During Normal Operation and AOOs), to demonstrate a margin to radionuclide release limits under accident scenarios (G2.2, Radionuclide Release Limits), and to demonstrate the ability to achieve and maintain safe shutdown (G2.3, Safe Shutdown). Portions of these may reference or require an EM or ED.

The associated NUREG-2246 goals for EMs and EDs are covered below.

G2.1 - Design Limits During Normal Operation and AOOs G2.1 establishes that the fuel is expected to remain intact or adhere to Specified Acceptable System Radionuclide Release Design Limits (SARRDLs) under conditions of normal operation, including the effects of AOOs. This goal comprises two sub elements: defining a fuel performance envelope (G2.1.1, Definition of Fuel Performance Envelope), and the specification of means of evaluating fuel for performance, failure, and degradation (G2.1.2, Evaluation Model). TR section 3.4.2, AGR TRISO Qualification Envelope vs. eVinci Microreactor Operating Conditions, covers operating conditions including temperatures, burnups, fast fluence, and power density. The section also states that it is the design goal for eVinciTM to have an operational envelope bounded by the AGR test data. The intended operating envelope for eVinciTM is summarized in figure 3.4-2, ((

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC staff finds that Westinghouse has identified the necessary models to meet EM G.1.2 following benchmarking and code validation and verification. The NRC staff imposes Limitation 2, that the application of the BISON framework will be limited to UCO TRISO fuels encompassed by experimental data, both in design and operational envelopes, used for validation of the model, or to where the scalability of both the UCO TRISO fuel system and the model has been adequately demonstrated.

EM G1.3 - Modeling Relevant Physics EM G1.3 emphasizes that the model must simulate the key physical processes affecting TRISO fuel, such as fission product migration, chemical interactions with coolant, and mechanical stresses.

TR section 4.3, Bison TRISO Physical Models, covers the relevant physical models including burnup, fission gas release, fission product recoil, Booth model for diffusion, fission gas production, internal gas pressure, fission product transport, fission product release and birth ratios, carbon monoxide production, kernel migration, and Palladium penetration.

Historical data, particularly those identified in the TR for benchmarking and validation assessment, are capable of informing the validity of burnup and fission gas release models included in the topical report. Direct recoil of ((

)) are fundamental physics equations derived from literature and are not easily, nor expected to be, benchmarked and validated directly in full scale testing. However, a benchmarking and validation process consistent with EMDAP and RG 1.203 would be capable of quantifying model sensitivities.

Similarly, TR section 4.3.2.2, Booth Model, proposes a model to simulate diffusive release of fission products through the kernel.

Fission product diffusivity will impact internal pressures and subsequent fuel kernel swelling, which may be directly analyzed in the test data, likely limiting the impact that this model directly has on fuel performance prediction. Coupling this with the model for fission gas production, pressure, and fission product transport in TR section 4.3.3, Fission Gas Production, section 4.3.4, Internal Gas Pressure, and section 4.3.5, Fission Product Transport, it is possible for an assessment to be made against the identified data sets.

Westinghouse intends to submit future licensing documentation on the eVinciTM microreactor mechanistic source term that will further support conformance with this goal. As such, the NRC staff finds the identified physical models and identified test base to be sufficient to formulate a mechanistic source term assessment and demonstrate that the modeling efforts conform to EM G1.3 from RG 1.203, subject to Limitation 1 and Condition 1.

EM G2 - Evaluation Model Assessment EM G2 states that the EM should be rigorously assessed to ensure it produces reliable predictions. This includes validating the model against ED to confirm its accuracy, as proposed in TR section 4.4. EM G.2 is subdivided into the assessment of ED and the fuel performance prediction over test envelope, which includes quantification of model error, data coverage of performance envelope, justification of sparse data regions, and restricted application domain.

EM G2.1, Experimental Data, is further divided into ED goals. EM G2.1 states that the model should be validated using data relevant to the proposed TRISO fuel, ensuring that predictions match observed performance. Westinghouse has identified data for such a validation in TR

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION conservative input assumptions. The benchmarking of modeling, V&V, and transient testing supports conformance with the conservative criteria outlined in G2.2 of NUREG-2246.

Additionally, TR section 6.0, Technical Specifications Framework and Failed Fuel Modeling, describes proposed active monitoring and associated TS to limit the release of fission products from failed fuel particles during normal and off-normal conditions. This will include the development and implementation of an online monitoring system, capable of detecting potential fission product releases from in-service fuel failures or manufacturing defects. As the fuel design is not directly correlated with historical TRISO fuel designs, this active monitoring is a critical component of the eventual fuel qualification efforts. As such, the NRC staff imposes Condition 3, that any application of this methodology will rely on the development of a failed fuel monitoring system that is capable of monitoring for adherence to radionuclide release limits and associated quantifiable TS.

G2.3 - Safe Shutdown G2.3 addresses the ability of the reactor to achieve and maintain safe shutdown following an accident, ensuring that the fuel maintains a coolable geometry and does not compromise core cooling. G2.3.1, Maintaining Coolable Geometry, states that the fuels structural integrity should allow it to maintain a coolable geometry during and after an accident. For TRISO fuel, this means ensuring that fuel particle integrity is maintained, preventing the release of fission products that could impede cooling. Westinghouse has stated that there are no phenomena that could cause the loss of coolable geometry because fuel channels are physically isolated from coolant channels (i.e., heat pipes). As such, there are no applicable EMs necessary to assess the coolable geometry margin. The NRC staff will assess this during the review of future submittals, consistent with Limitation 1.

G2.3.2, Negative Reactivity Insertion, states that the reactor design should ensure that negative reactivity can be inserted during an accident. TRISO fuels behavior under such conditions must support this reactivity control, contributing to safe shutdown. This goal has two supporting subgoals: G2.3.2(a) states that, in part, Criteria should be provided to ensure that the means to insert negative reactivity is not obstructed during conditions of normal operation or accident conditions.

The TR EVR-LIC-RL-001-A, Revision 1, Principal Design Criteria Topical Report, (ML24353A097) dated December 17, 2024, provides PDCs for the eVinciTM Microreactor design that were reviewed and approved by the NRC staff in the associated SE (ML24283A133).

PDC 26 states, in part, that, Reactivity control shall be provided. Reactivity control shall provide a means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the specified acceptable system radionuclide release design limits and the reactor helium pressure boundary design limits are not exceeded, and safe shutdown is achieved and maintained during normal operation, including AOOs. Reactivity control shall provide a means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the capability to cool the core is maintained and a means of shutting down the reactor and maintaining, at a minimum, a safe shutdown condition following a licensing basis event. G2.3.2(b) states, An evaluation model is available to assess geometry changes as a result of normal operation and accident conditions.

The NRC staff notes that the execution of this methodology will inform geometric evolution sufficient to assess negative reactivity insertion. Moreover, negative reactivity insertion is design

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION LIMITATIONS AND CONDITIONS The NRC staff imposes the following limitations and conditions on this TR:

Limitation 1:

The NRC staffs approval of the methods in this TR is limited to the general use of BISON, the described model assumptions, and verification and benchmarking approaches. Staff make no findings on the sufficiency of the results from testing, modeling, and analyses performed in accordance with the methodology to demonstrate conformance with PDCs or regulations.

Condition 1:

An applicant referencing this TR must provide justification that parameters affecting the fuel performance, including the composition, dimensions, and operating envelope, are within the scope of those to which the methodology can be applied and its results validated.

Condition 2:

Future submittals referencing this TR must quantify manufacturing tolerances in a manner that can demonstrate adherence to potential SARRDLs or otherwise specified radionuclide release limits.

Limitation 2:

Application of the BISON framework will be limited to UCO TRISO fuels encompassed by experimental data, both in design and operational envelopes, used for validation of the model or to where the scalability of both the UCO TRISO fuel system and the model has been adequately demonstrated.

Condition 3:

Any application of this methodology will rely on the development of a failed fuel monitoring system that is capable of monitoring for adherence to radionuclide release limits and associated quantifiable technical specifications.

CONCLUSION The NRC staff has determined that Westinghouses topical report, EVR-LIC-RL-003-P/NP, Westinghouse TRISO Fuel Design Methodology," Revision 0 (ML24214A277), provides an acceptable approach for TRISO fuel qualification, subject to the limitations and conditions discussed above, by outlining a methodology that follows NUREG-2246 and RG 1.203. The NRC staff notes that while the fuel design and operating conditions provided in the TR are relatively mature, as is the current modeling framework proposed in the TR, the methodology is preliminary, and this TR defers significant technical details to future submittals. Successful implementation of this methodology will rely on the benchmarking and validation used to evaluate test distortions, an activity that is deferred to future submittals.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION REFERENCES

1. U.S. Nuclear Regulatory Commission (NRC), EVR-LIC-RL-003 P/NP Westinghouse TRISO Fuel Design Methodology Topical Report, Revision 0 (ML24214A277), July 31, 2024.

2. U.S. NRC, EVR-LIC-RL-001-A, Revision 1 Principal Design Criteria Topical Report, (ML24353A097), December 17, 2024.

3. U.S. NRC, Final Safety Evaluation for eVinci' Principal Design Criteria Topical Report, (ML24283A133), October 16, 2024.

4. U.S. NRC, Regulatory Audit Report regarding Westinghouse Electric Companys TRISO Fuel Design Methodology Topical Report, (ML25190A670), August 21, 2025.

5. U.S. NRC, Regulatory Guide (RG) 1.203, Transient and Accident Analysis Methods, (ML053500170), December 2005.

6. U.S. NRC, DANU-ISG-2022-01, Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications - Roadmap, (ML23297A158), March 2024.

7. U.S. NRC, NUREG-2246, Fuel Qualification for Advanced Reactors, (ML22063A131),

March 2022.

8. U.S. NRC, EPRI (2020) Transmittal of Published Topical Report EPRI-AR-1(NP)-A, Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance, (ML20336A052), November 30, 2020.

9. U.S. NRC, Final Safety Evaluation for Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance Topical Report EPRI-AR-1(NP)

(ML20210M241), August 4, 2020.

Principal Contributors: Ayesha Athar Walter Williams Date: December 19, 2025