In recent years, there has been renewed interest in Molten Salt Reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, many methods and codes have been developed to capture the unique features of MSRs. Among them, the System Analysis Module (SAM) is a modern system analysis tool that provides fast-running, modest-fidelity, whole-plant transient analysis capabilities, essential for fast-turnaround design scoping and engineering analyses of advanced reactor concepts. For liquid-fuel MSRs, the complex physics and chemistry involved in MSR operation—such as reactor kinetics, fluid flow, heat transfer, and salt composition dynamics—pose significant challenges for system-level modeling. Specific modeling capabilities are needed for system-level transient simulation. This paper presents recent advancements in SAM capability enhancements for system-level modeling of MSRs, focusing on improved simulation fidelity, computational efficiency, and multi-physics integration. Key enhancements include the development of species transport, Delayed Neutron Precursor (DNP) drift, modified Point Kinetics Equations (PKE), decay heat modeling, key fission product behavior, salt corrosion, and thermal-hydraulic coupling, as well as code robustness and performance enhancements for MSR applications. The code enhancement allows for better predictive accuracy in safety analysis, transient behavior, and operational optimization, thus supporting the design and licensing of next-generation MSRs. Results from case studies are presented to demonstrate the benefits of these enhancements in accurately capturing key reactor transient behaviors.

The molten salt reactor is one candidate among the Generation IV nuclear reactor designs, with its deployment relying on advanced computational tools to capture the unique behavior of the circulating fuel system. The Molten Salt Reactor Experiment (MSRE) provides valuable experimental data for validating these computational tools. This work develops a reactor transient benchmark based on the MSRE pump transient tests.Two computational models are evaluated in the benchmark: a simplified one-dimensional (1D) system-level model and a more detailed R-Z axisymmetric model using the porous medium approximation. The models are used to evaluate the impact of spatial resolution on predicted reactivity responses during the transient. Several impactful factors are examined during the benchmark evaluation, including the neutron diffusion multigroup energy structure, delayed neutron precursor (DNP) diffusion, DNP group structure, bypass flow, and transient flow rates.The reactivity predictions using the computational models are compared to the experimental data. The mean errors in the predicted reactivity responses ranged from 11 to 21 pcm (1 pcm = 10-5) for the pump startup transient and 5 to 13 pcm for the pump coastdown transient. These results indicate that the 1D model can provide adequate accuracy on MSRE pump transients with limitations in predicting the rate of reactivity at the early stage of the transient, while the higher-order model improves this capability by incorporating the influence of radial salt flow distribution and bypass flow on transient reactivity.4

The Molten Salt Reactor (MSR) Multiphysics Applications technical area in the Nuclear Energy Advanced Modeling and Simulation program has supported the development of 3D Monte Carlo models of the Molten Salt Breeder Reactor (MSBR) over the last couple of years. This MSBR model was previously run with the Shift Monte Carlo code to perform radiation shielding calculations in the reactor cell area. The MSBR is a 2250 MWth (1000 MWe) liquid-fueled molten salt reactor design developed at Oak Ridge National Laboratory in the 1970s. Determining the source terms from activated primary heat exchanger (HX) components is important because delayed and prompt neutron fluxes incident on these components affect the dose rate in the primary HX maintenance areas. This information can be used in the development of remote handling procedures required during shutdown for maintenance. A methodology has been developed and is proposed in this paper to quantify the activated source term from the primary HX components as a result of the movement of the delayed neutron precursors in flowing primary fuel salt through the primary HXs in the MSBR. The goal of this research is to evaluate the gamma dose rates in the maintenance hatches above the primary HX using the activated HX source terms. The study showed that the gamma dose rates are approximately two orders of magnitudes higher when accounting for the neutron activation from the movement of delayed neutron precursors through the HXs than when flowing fuel is not considered. Thus, the movement of delayed neutron precursors must be taken into account for accurately predicting the neutron activation of primary loop components.

sectors are identified by the letters A through F around the perimeter of the core diagram. A table at the end of the report summarizes the pertinent information for every EBR-II run through run 67. (auth)

Advanced microreactors are expected to play an indispensable role in reliable energy solutions for civilian/military applications in off-grid regions and space applications such as power supplies for Lunar and Mars bases. An innovative fast neutron spectrum heat pipe microreactor (HP-MR) concept was recently proposed with two unique features: molten U-Mn fuel and heat pipes as the heat removal mechanism. The microreactor was designed to be self-regulated, solely relying on reactivity feedbacks from fuel temperature. The reactor design has been further optimized to provide sufficient engineering safety margin and effective heat removal. In this study, the self-regulating performance of the optimized molten metal fuel microreactor was investigated by multiphysics simulations (neutronics, heat-transfer and heat pipe modeling) based on the MOOSE MultiApps system. The responses of the reactor to a series of transient scenarios were simulated to demonstrate the exceptional and inherent safety features of the novel microreactor concept.

This report documents the benchmarks being developed and simulations performed using tools and codes developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, utilizing MSRE experimental data. In FY23, three main work scopes were investigated under the NEAMS MSR work package at ANL. The first scope investigated the Griffin-SAM coupling model for simulating the pump startup transient experiment of MSRE. The analyses start with a simple model (single-channel, single-lattice), gradually adding more details (multi-channel, full-core) into the model. The results show that the reactivity loss curve is very sensitive to the axial boundary conditions and the radial core discretization. The simple model can predict a similar reactivity trend as that of the more sophisticated model, which is likely due to error cancellation. Accurately modeling the axial boundary condition may further improve the reactivity trend but would require significant efforts to generate the mesh of the MSRE inlet and upper plenum. The core channel radial discretization for the Griffin-SAM coupled model also depends on the flow distribution. Given the complex geometry in the inlet plenum, the flow distribution needed to be calculated from CFD analysis, which was performed using the NekRS code. This analysis employed a MSRE CAD model developed by Copenhagen Atomics. The CAD model was disassembled to keep the inlet plenum region only, which was subsequently cleaned and modified so that the mesh generated is under the memory limit. The results are merged to a few radial regions to show that the flow rate is highest in the central region. This would be useful for future improvement of the Griffin-SAM coupling model of the MSRE core. The last task investigated is tritium transport modeling using the standalone SAM code. This task aimed to initiate the effort to demonstrate and validate the tritium transport model implemented in SAM. The preliminary investigation employed an MSRE model consisting of the primary loop. Three tritium transport pathways were examined including the retention in the graphite, the permeation through the HX tube wall, and the removal from the off-gas system. The results compare well with the MSRE data, but improvements are still needed on the initial conditions (i.e., the present state may not have reached equilibrium), the boundary conditions, the off-gas system modeling, and a better numerical strategy to reach the equilibrium state.

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the U.S. Department of Energy, Office of Nuclear Energy. Safety analysis of the conceptual VTR design is being performed using the SAS4A/SASSYS-1 fast reactor safety analysis code with a model representing the reactor core, primary and intermediate heat transport systems, reactor vessel auxiliary cooling system, and reactor protection system. The system’s response and safety performance has been evaluated for a wide spectrum of event initiators and accident sequences. This paper focuses on the results for several unprotected transient scenarios where the reactor protection system is assumed to fail to take any action. Even without the reactor protection system, the strong reactivity feedback response of the core reduces power to safe levels matching the available heat rejection. In the station blackout transient, the primary heat transport system is able to transition quickly and effectively to natural circulation. At the current stage of design, transient simulation results for the VTR indicate that large safety margins exist for many event initiators, including the unprotected transients presented in this paper.

be statistically validated again the overall structural evolution. The GB energy, specific GB mobility, grain growth kinetic, GB interaction with pores, and effects of large temperature gradients will be specifically studied using experiments. The isothermal grain growth, grain growth under large temperature gradient, and pore and GB interactions will be modeled using 2-D or 3-D simulations. Four questions will be answered at the conclusion of this project: Is the anisotropic grain boundary energy model able to better represent a real view the grain property and would it have a significant impact on the grain growth behavior? Is the classical Zener pinning model in MARMOT sufficient for representing the retarding effect on grain gain boundary migration from pores? Does the temperature gradient have a minimum impact on the average grain growth behavior at varied average temperatures and temperature gradients? Does the pore migrate towards the high temperature domain at the velocities predicted by MARMOT code?

are also important to determining inelastic response to longer term flux gradients which cause irradiation creep and swelling. While core bowing provides a strong negative reactivity feedback when the restraint system is designed properly, it also results in additional forces between assemblies which increase the loads required to extricate them during refueling or control rod movement. Therefore, the restraint system must be designed with these tradeoffs in mind. The first stage of the work, which commenced in FY21 and continues through FY22, assesses thermo-mechanical modeling tools for producing core bowing predictions consistent with conventional tools. The Multiphysics Object Oriented Simulation Environment (MOOSE) Tensor Mechanics and Contact Modules are employed. This status report describes work on additional thermo-mechanical benchmark verification problems with increased complexity from the examples demonstrated in FY21. Several benchmark verification examples were selected from the IAEA verification and validation report. These examples involve clusters of ducts representative of a sector of a hexagonal reactor core which bow into each other and cause contact and load pad elevations, as well as single ducts subjected to irradiation fields undergoing swelling and subsequent bowing. The MOOSE-based results were compared to both IAEA benchmark participants’ results, analytic equations as available, and NUBOW-3D, a beam model code developed by Argonne National Laboratory. In every case, the MOOSE results agreed with other simulations results, providing additional verification basis of the tools for this particular physics application.

The Versatile Test Reactor (VTR) is currently under development by the U.S. Department of Energy. It will provide very high fast neutron flux irradiation capabilities that are currently unavailable in the United States. Given the increasingly large number of advanced reactor concepts being pursued in recent years, this irradiation testing capability will be essential to support maturation of these designs. Radiation protection is an important part of the VTR design. High neutron fluxes can pose a challenge for radiation protection of the structures and equipment near the reactor core. This paper provides a summary on the status of the radiation protection considerations and shielding analysis performed for VTR under a nominal operating condition. The main radiation sources identified and examined in the study are applicable only under this operating condition. The paper focuses on three areas of radiation protection and shielding: secondary sodium activation in the intermediate heat exchanger, air activation in the reactor vessel auxiliary cooling system, and dose rate above the head access area due to primary sodium activation. VTR design and development are continuously progressing, and as such, the shielding considerations discussed in this paper will evolve alongside the overall VTR design.

limit state is optimal for expansion. When the parameter space is transformed into a standard Gaussian space, the optimal expansion point, known as the lost Probable Point (MPP), has the property that its location on the Limit State surface is closest to the origin. Additionally, the projections onto the parameter axes of the vector from the origin to the MPP are the sensitivity coefficients. Once the MPP is determined and the Limit State surface approximated, formulas (see Equations 4-7 and 4-8) are available for determining the probability of the performance measure being less than the limit state. By choosing a succession of limit states, the entire cumulative distribution of the performance measure can be detemined. Methods for determining the MPP and also for improving the estimate of the probability are discussed in this report.

The microreactor concept has received significant attention in the United States for its lower capital investment, siting flexibility and high mobility. A microreactor aims to provide reliable electricity in remote communities, mining areas, or isolated islands, which will save the high fuel transportation cost. Thus, minimizing the staff level for operating and maintaining a microreactor is desirable. The objective of the study is to design a self-regulated Molten Metal fueled microReactor core (MMR) and to demonstrate the self-regulating capability of the MMR core. The control of the reactor is solely relying on reactivity feedbacks from fuel temperature perturbation. Molten UMn (liquid form at the temperature over 720 degrees C) was utilized as the fuel material in MMR due to the larger thermal expansion coefficient compared to the conventional solid fuel. The MMR core has a fast neutron spectrum, which corresponds to a small burnup reactivity swing. The core design parameters were selected based on DAKOTA, REBUS, and ANLHTP calculations. The MMR is designed to run for 10 years at a power level of 15 MWth. No refueling is needed during the reactor lifetime. The system dynamic analyses were performed, and it was found that the reactor can be self-regulating within a temperature range of 800 +/- 80 degrees C for a loss of heat sink transient.

Successful operation of the Versatile Test Reactor (VTR) relies on satisfying stakeholder requirements and guaranteeing customer timelines. Although the VTR leverages proven sodium fast reactor technology, its mission as a test reactor introduces unique design and operational requirements. This paper covers the preliminary analysis and methodology development for two areas of the VTR core design related to the operational flexibility necessary for the testing mission. The first of these introduces a framework for assessing the feasibility of storing used driver fuel in the VTR shield region, which offers potential benefits to operations but may affect core reactivity and increased cooling time. A methodology to assess these impacts using neutronics and depletion calculations is demonstrated on three in-shield storage configurations. The second focus area highlights operational considerations and maximum residence time of the VTR control assemblies, which are critical to maintaining the irradiation environment necessary to deliver on the VTR mission. A preliminary methodology that assesses B-10 depletion and absorber rod swelling is demonstrated with the goal of informing future development. Together, these research activities illustrate how the early-stage VTR design is guided by anticipation of operating objectives.