ML25206A372

Final SE Kairos Core Design Methodology TR (Public)

ML25206A372
Person / Time
Site:	99902069
Issue date:	11/25/2025
From:	Office of Nuclear Reactor Regulation
To:
References
EPID L-2024-TOP-0013
Download: ML25206A372 (0)

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Enclosure KAIROS POWER LLC - FINAL SAFETY EVALUATION OF TOPICAL REPORT KP-TR-024-P, KP-FHR CORE DESIGN AND ANALYSIS METHODOLOGY (EPID L-2024-TOP-0013)

SPONSOR AND SUBMITTAL INFORMATION Sponsor:

Kairos Power LLC (Kairos)

Sponsor Address:

707 W. Tower Ave, Suite A Alameda, CA 94501 Project No.:

99902069 Submittal Date:

April 3, 2024 Submittal Agencywide Documents Access and Management System (ADAMS)

Accession No.: ML24095A255 Revision Letter Date and ADAMS Accession No: Revision 1, June 17, 2025, (ML25168A340)

Brief Description of the Topical Report: On April 3, 2024, Kairos Power LLC (Kairos) submitted Topical Report (TR) KP-TR-024-P, KP-FHR Core Design and Analysis Methodology, Revision 0 (Agencywide Documents Access and Management System (ADAMS)

Accession No.:ML24095A255) for the U.S. Nuclear Regulatory Commission (NRC) staff review.

On June 17, 2025, Kairos submitted Revision 1 of this TR (ML25168A340). The TR provides the methodology for core physics, thermal-hydraulic analysis, and radiation effects on materials for the Kairos Power Fluoride SaltCooled, High Temperature Reactor (KPFHR). The methodology described in the TR is used to calculate reactivity coefficients, control and shutdown rod worths, shutdown margin, flux distribution, power distribution, temperature distribution, kinetics parameters, material depletion, radiation damage and heating, and pebble peaking factor. The TR identifies the computer codes used and discusses their verification and validation. The TR also describes an approach for quantifying uncertainties and determining biases to inform the development of nuclear reliability factors (NRFs).

EVALUATION CRITERIA Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) 50.34(a)(4) for a construction permit (CP) application, 10 CFR 50.34(b)(4) for an operating license (OL) application, and 10 CFR 52.79(a)(5) for a combined license (COL) application require, in part, analysis and evaluation of the design and performance of structures, systems, and components (SSCs) of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents.

The core design and analysis methodology presented in the TR is used to determine safety margin for the KPFHR fuel and core during normal operation. The methodology also provides input to the safety analysis performed to determine adequacy of SSCs designed to prevent or mitigate the postulated events in a KPFHR.

Topical report KPTR003NPA, Principal Design Criteria for the Kairos Power Fluoride Salt Cooled, High Temperature Reactor, Revision 1, dated June 12, 2020, (ML20167A174) provides principal design criteria (PDC) for the KPFHR design that were reviewed and approved by the NRC staff. Section 1.2 of the core design and analysis methodology TR identifies PDC that can be addressed, in part, using the methodology presented in the TR.

These include PDC 10, 11, 12, 16, 25, 26, 28, 31, and 34.

While the NRC staff considered these regulations and PDC in its review of the TR, determinations regarding compliance or conformance will be made during review of licensing applications referencing this TR.

TECHNICAL EVALUATION The Kairos core design and analysis methodology presented in the TR is based on the Serpent 2 computer code for the nuclear design and the STAR-CCM+ computer code for the thermal-hydraulic analysis. The methodology models steady-state neutronic and thermal-hydraulic phenomena and is developed for startup, power ascension, and equilibrium conditions in KPFHR test and power reactors. As described in TR section 1, Introduction, Kairos is requesting the NRC staffs review and approval of the following:

use of the Serpent 2 and STAR-CCM+ based calculational framework to calculate reactivity coefficients, control and shutdown rod worths, kinetics parameters, power distribution, material depletion, shutdown margin, flux distribution, temperature distribution, radiation damage and heating, and pebble peaking factor methodology for quantification of biases and uncertainties in neutronic calculations based on code-tocode benchmarks and literature data methodology for validation of thermal-hydraulic models and quantification of biases and uncertainties in thermal-hydraulic calculations methodology for updating NRFs using operational data

1. Topical Report Overview The TR consists of the following major sections and appendices:

Section 1, Introduction, provides a brief description of KPFHR design features and identifies specific technical areas for which Kairos is requesting NRC staff review and approval. The NRC staff considers the information in section 1 of the TR throughout the technical evaluation in this safety evaluation (SE) but does not make any determinations on the information in TR section 1.

Section 2, KP-FHR Core Design Features, describes the KPFHR core design and introduces KPFHR operational regimes considered in the TR. Limitation 1 in TR section 7.2, Limitations, states that the methodology is applicable to the design as presented in this TR; the NRC staff includes this limitation, and other limitations cited in TR section 7.2, through Limitation and Condition 1 in the Limitations and Conditions section of this SE. The NRC staff considers the information in section 2 of the TR throughout the technical evaluation in this SE but does not make any determination on the information in TR section 2.

Section 3, Core Modeling Paradigms, describes the modeling approach for the KPFHR core, which includes the discrete element method (DEM) for modeling pebble movement, a neutronic model for modeling neutron transport, and a thermal-hydraulic model for modeling coolant flow and temperature distribution. It also summarizes the key steady-state phenomena used to determine the adequacy of the methodology to calculate important figures of merit (FOMs) and provides a list of parameters or outputs calculated by the core methodology.

Section 4, Modeling Tools, provides a summary of the computer codes, STAR-CCM+ and Serpent 2, and several supporting computer programs or tools used to implement the core methodology models. The NRC staffs evaluation of the KPFHR core methodology phenomena, modeling approach, associated computer tools, and adequacy to calculate the parameters in TR section 3.6 is found in SE section 2, Methodology Approach.

Section 5, Validation, Verification, and Uncertainty Analysis, discusses the validation, verification, and uncertainty quantification of the DEM, neutronics, and thermal-hydraulic core models. This section discusses use of computational fluid dynamics (CFD) for numerical validation of the thermal-hydraulic model. It also discusses startup testing and the methodology for calculating and updating NRFs. The NRC staffs review of the validation, verification, and uncertainty analysis approach is found in SE section 3, Methodology Validation, Verification, and Uncertainty Analysis. TR appendix B, Verification and Validation, provides additional discussion on validation of the DEM and thermal-hydraulic models. The NRC staff used the information in TR appendix B to enhance its understanding of the methodology but does not make any determinations on TR appendix B.

Section 6, Applications, discusses applications of the methodology, including how FOMs and certain outputs are used for safety analysis, source term, nuclear design, and thermal-hydraulic analysis. It also discusses startup physics testing and the methodology to calculate the desired core composition. TR appendix A, Example Calculation, illustrates the application of the methodology. The NRC staff considered the information in section 6 and appendix A of the TR throughout the technical evaluation in this SE but does not make any determinations on the information in TR section 6 and appendix A.

Appendix C, Neutronics PIRT Results for the KP-FHR, provides a list of neutronic phenomena ranked by importance and knowledge level. The NRC staff considered the information in the phenomena identification and ranking table (PIRT) in appendix C for evaluation of the neutronic model in SE section 2 but does not make any determinations on the PIRT in TR appendix C.

2. Methodology Approach 2.1 Methodology Overview and Computer Codes The Kairos core design and analysis methodology consists of three models (paradigms): DEM, thermal hydraulics, and neutronics. The modeling domain for each model is shown in figure 3-1, Core Modeling Domains, of the TR. Table 1, Methodology Models and Codes, of this SE summarizes these models. Table 2, Computer Codes Used, of this SE summarizes the computer codes used to implement the models and the wrapper tools that facilitate the coupling and/or data transfer between the models.

DEM calculates the movement of pebbles in the inlet fueling region, cylindrical and conical (converging and diverging) core regions, and defueling chute. The methodology uses a commercially available CFD code, STAR-CCM+, to implement DEM. DEM calculates static pebble center locations, packing fractions (bed porosity), and average pebble tracks and velocity profiles that are used in the neutronic and thermal-hydraulic models. The DEM results are post-processed to generate spectral zones in the core for the neutronic calculations. DEM is coupled with the neutronic model using a wrapper code, KPACS. As described in TR section 4.4.3, KPACS, KPACS can simulate pseudo-steady-state evolution of the KPFHR core. The NRC staffs evaluation of the DEM model is provided in SE section 2.2, Discrete Element Method (DEM).

The neutronic model calculations are performed using Serpent 2, a continuous-energy Monte Carlo (MC) neutron transport code. As shown in TR figure 31, the modeling domain for the neutronic model includes a full three-dimensional (3D) representation of the core (including converging and diverging sections), defueling chute, fueling region, reactor shutdown system (RSS), reactor control system (RCS), graphite reflector, reactor penetrations, core barrel, downcomer, and reactor vessel. The core model explicitly represents different types of pebbles and tri-structural isotropic (TRISO) particles. The reflector is modeled with major engineered penetrations and geometries for the coolant paths. The neutronic model is explicitly coupled with the thermal-hydraulic model using the KPATH wrapper code. The power distribution calculated by the neutronic model is provided as input to the thermal-hydraulic model. The material temperature distributions calculated by the thermal-hydraulic model are used to update the material cross-sections in the neutronic model. The NRC staffs evaluation of the neutronic model is provided in SE section 2.3, Neutronics.

The thermal-hydraulic model uses STAR-CCM+ to calculate the steady-state 3D core material temperature distribution for the coolant (Flibe), fuel and graphite pebbles, TRISO particles, and reflector (TR section 7.2, Limitation 2). The model uses two paradigms: (1) a local thermal non-equilibrium (LTNE) porous media (PM) model for the core region, including fuel and moderator pebbles, and (2) a CFD ((

)) model for the reflector region, including the gaps and coolant flow channels. The NRC staffs evaluation of the thermal-hydraulic model is provided in SE section 2.4, Thermal-Hydraulic Model.

In addition to KPACS and KPATH, TR section 4.4, Wrapper Codes, describes three more wrapper codes. Table 2 of this SE provides a brief description of the HEEDS, KACEGEN, and Zoner wrapper tools. TR section 4.4 states that these wrapper codes perform data transfer and do not contain physical models that need to be validated; however, these codes are numerically verified. TR figure 41, High Level Process Flow Diagram of the Core Design and Analysis Methods, illustrates a simplified process data flow from Serpent 2, STAR-CCM+, and wrapper codes to downstream applications. This figure illustrates the connections between computational modules, not inputs and outputs between computational modules. The NRC staff considered the information on these wrapper tools in the context of the overall calculational framework in SE section 2.5, Overall Methodology Process.

The TR also makes use of ((

)) in section 5.2, Neutronics, to perform code-tocode benchmarking and uncertainty evaluation.

The NRC staff determined that the consideration of sources of uncertainties and methodology to develop QU and bias as described in TR sections 5.2.3 and 5.2.4, respectively, are acceptable because they provide reasonably informed estimates and when directly comparable data is available (e.g., data from the Hermes and Hermes 2 test reactors) NRFs may be updated, as described in TR section 5.4 (TR section 7.2, Limitation 3). The NRC staff also has reasonable assurance that the DCs described in TR section 5.2.5 and table 527, Sources of Unquantified Uncertainty, are acceptable for the Hermes and Hermes 2 test reactors because they provide sufficient additional margin such that core design-related safety limits will be ensured during operation (TR section 7.2, Limitation 5).

As stated in TR section 7, Summary, the predictions of the neutronic module will be confirmed during the fuel loading process and subsequent zero-power testing of Hermes and future KPFHR reactors. TR section 6.6, Startup Physics Testing, describes how the startup process of a KPFHR will use nuclear design calculations. Critical mass predictions from Serpent 2 will be used during fuel loading. After achieving criticality, zero-power testing, including control rod calibration, shutdown rod worth, and isothermal temperature coefficient measurements, will be performed and compared against predictions. If the measured values lie outside of the predicted values with uncertainty bands, testing will be suspended and impacts on the safety analysis will be evaluated before any further testing is performed. Measurements will also be taken and compared to predictions during power ascension. The NRC staff notes that this is a reasonable approach to startup physics testing and is consistent with standard industry practice. The NRC staff will review specific testing plans and procedures as part of future licensing submissions.

The strategy for updating the NRFs based on operational data described in TR section 5.4 uses standard confidence interval analysis and is therefore acceptable to the NRC staff. The NRF update process includes the use of test data to formally quantify uncertainties for a system that is directly applicable to future KPFHRs (TR section 7.2, Limitation 3). The NRC staff expects to review these data prior to approving application of this methodology for future KPFHR power reactors. Therefore, the NRC staff imposes Limitation and Condition 5, requiring the operating data gained as part of validating the methodology described in this TR to be reviewed by the NRC staff and any updates to NRFs to be reviewed and approved by the NRC staff prior to future applications to power reactor systems.

TR section 6 describes calculations performed in nuclear design that support subsequent analyses (e.g., safety analysis, source term analysis). These include such quantities as reactivity coefficients, integral and differential control and safety rod worths, kinetics parameters, power distribution, stability, material depletion and transmutation, and radiation fluence.

Although the NRC staff does not make any determinations on the information in this section, the NRC staff generally found that the list of evaluated quantities and approaches to calculate primary and derived parameters is reasonable.

TR appendix A provides example implementation of the core design methodology for a version of the Hermes design, including detailed demonstrations of the calculations described in TR section 6 and the uncertainty and bias estimation described in section 5. Because this appendix is presented primarily to illustrate the methods and assist the staff in understanding how the methodology described in the TR is implemented, the staff did not make any determinations on TR appendix A.

3.3 Thermal-Hydraulic Model 3.3.1 Thermal-Hydraulic Model Validation

The NRC staff determined that the uncertainty quantification methodology is acceptable because it adequately accounts for the major sources of uncertainty. Furthermore, the proposed values for the bias factor and confidence interval factors provide a conservative estimation of the uncertainty band. Although the methodology proposed for the propagation of input uncertainties is reasonable, as discussed earlier in SE section 2.4.1, the methodology input parameters and geometrical specifications and the uncertainty distributions assigned to these parameters must be justified for each application of the methodology and should account for the impact of fluence, temperature, and burnup, as reflected in Limitation and Condition 2.

LIMITATIONS AND CONDITIONS The NRC staff applies the following additional limitations and conditions on the acceptance of this TR:

1. Any licensing application referencing this TR must demonstrate that the limitations listed in TR section 7.2 are met, subject to NRC staff review and approval.

2. Any licensing application referencing this TR must provide for NRC staff review and approval acceptable justification that the input values for thermophysical properties and geometrical specifications, as well as the uncertainty distributions assigned to these values, for the core materials and reflector adequately account for the impact of fluence, temperature, and burnup.

3. Any licensing application referencing this TR must provide for NRC staff review and approval acceptable justification that the mesh design for the CFD models is performed using the process described in TR section 3.5.4.3 for each family of KPFHR designs and expected operating conditions.

4. Any licensing application referencing this TR must provide for NRC staff review and approval acceptable justification that the selection of turbulence and wall models for the CFD models is performed using the process described in TR section 3.5.4.5 for each family of KPFHR designs and expected operating conditions.

5. Any licensing application for a KPFHR power reactor referencing this TR must provide for NRC staff review supporting test data obtained from operation of the Hermes test reactors or other representative data used for final validation and justification of its applicability for each family of KPFHR designs and expected operating conditions. In addition, if updates are made to nuclear reliability factors per the TR methodology, the changes will be provided for NRC staff review and approval.

CONCLUSION The NRC staff determined that Kaiross KPTR024P, KP-FHR Core Design and Analysis Methodology, Revision 1 provides an acceptable methodology for the calculation of parameters in TR section 3.6 and for the quantification of biases and uncertainties subject to the limitations and conditions discussed above. The evaluation of final compliance or conformance with the identified regulations and PDCs will be performed during the review of a licensing application referencing this TR.

REFERENCES

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A.I.Ch.E., Vol. 5, No. 1, pp. 79-85, 1959

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6. Hu, G., O'Grady, D., Zou, L. and Hu, R., 2020, Development of a Reference Model for Molten-Salt Cooled Pebble-Bed Reactor Using SAM (No. ANL/NSE-20/31). Argonne National Lab

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Martineau R.C., H.D., 2020, Pronghorn Theory Manual (No. INL/EXT-18-44453-Rev001), Idaho National Lab

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9. ASME Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer V&V 20 - 2009(R2021)

Principal Contributors: Pravin Sawant Alex Siwy Ben Adams Andrew Bielen Christopher Boyd Joshua Kaizer Date: November 2025