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DOE OSTI2025

SAM Theory Manual

The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR reactor safety analysis. It aims to provide fast-running, modest-fidelity, whole-plant transient analyses capabilities, which are essential for fast turnaround design scoping and engineering analyses of advanced reactor concepts. While SAM is being developed as a system-level modeling and simulation tool, advanced modeling techniques being implemented include a reduced-order three-dimensional module, pseudo 3-D conjugate heat transfer modeling in reactor core, flexible and multi-scale modeling of heat transfer between fluid and structures, in addition to the advances in software environments and design, and numerical methods. SAM aims to be a generic system-level safety analysis tool for advanced non-LWRs, including Liquid-Metal-cooled fast Reactors (LMR), Molten Salt Reactors (MSR), Fluoride-salt-cooled High-temperature Reactors (FHR), and High-Temperature Gas-cooled Reactors (HTGR). SAM takes advantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented computational framework (MOOSE), and its underlying meshing and finite-element library and linear and non-linear solvers, to leverage the modern advanced software environments and numerical methods. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As new code capabilities and features are added, the SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

SAM User's Guide

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for Sodium-cooled Fast Reactors (SFR), Lead-cooled Fast Reactors (LFR), and Molten Salt Reactors (MSR) or Fluoride-cooled High-temperature Reactors (FHR). SAM takes advantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage the modern advanced software environments and numerical methods. This document provides a user’s guide, which will help users understand the input description and core capabilities of the SAM code. A brief overview of the code is presented, as well as how to obtain and run it. The input syntax for various parts of the code is provided. Additionally, a number of example problems, starting with simple unit component problems to problems with increasing complexity, are provided. Because the code is still under active development, this SAM User’s Guide will evolve with periodic updates.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Preliminary Analysis of Nuclear-Powered Data Center Scenarios

This report provides a comprehensive analysis of the potential for nuclear energy to meet the growing energy demands of data centers (DCs). It evaluates the technical, economic, and socio-environmental implications of coupling Nuclear Power Plants (NPPs) with DCs, providing initial responses to several key research questions: What is the potential increased energy demand from DCs in the U.S., in the short, medium and long term? The U.S. is experiencing a rapid increase in energy demand from DCs, with projections indicating a total increase of 24-74 GWy(e) by 2028. Meeting this demand with nuclear energy would require 27–85 GWe of installed capacity. While this surge is expected to slow in the long term, the DC industry needs reliable, scalable, and clean energy sources. How much nuclear capacity can be deployed to meet DC demand and in which timeframe? Several pathways for increasing nuclear capacity were identified, including uprates, restarts of recently retired reactors, power purchase agreements with existing fleet, and new construction. Approximately 20‒28 GWe of nuclear capacity could be dedicated to DCs by the early 2030s. How much High Assay Low Enriched Uranium (HALEU) would be needed to support some nuclear deployment scenarios for DCs? Meeting the deployment targets announced by Google and Amazon for the Kairos Power Fluoride-Salt-Cooled High-Temperature Reactor or KP-FHR (~500 MWe by 2035) and the Xe-100 (~1 GWe by 2040), respectively, requires ramping up 19.75% enriched HALEU production to ~6 t/yr by 2040. What types of nuclear energy/DC coupling options exist, and what are the different benefits/challenges? Five coupling options were analyzed, ranging from grid-connected configurations to colocated, behind-the-meter setups. Key design considerations include the proximity to high- and/or medium-voltage transmission lines, the desired internal fault tolerance, and the sources of alternative/backup power during outages. Each coupling option offers unique benefits and challenges in terms of reliability, system costs, regulation, timeline, etc. A list of NPP/DC deployment scenarios was developed, considering existing or newly built NPP or DC projects. Colocated DCs with new small modular reactors or large reactors on greenfield and brownfield sites are the focus of this report. What types of reactors, especially what size, may be incentivized by DCs? Reactor sizing optimization revealed that the ideal reactor size and number of units depend on DC demand, coupling configurations defined in this report, and other economic factors. Larger reactors are preferred for high-demand DCs and grid-connected systems, while larger number of smaller reactors are better suited for DC configurations without grid backup. Which sites may be compatible with co-located nuclear-powered DCs? Siting those projects is a complicated evaluation factoring local water resources, grid connection availability and reliability, IT infrastructure, local work force, proximity to population zones, etc. For this effort greenfield and brownfield sites such as retired coal-fired plants were used to evaluate this question. This evaluation is not meant to recommend any particular site but it highlights key siting criteria and demonstrates large-scale site availability. What are the socio-economic impacts of co-located nuclear-powered DCs? Those projects generate substantial economic benefits to the local economy, particularly in urban settings. Hyperscale DCs colocated with nuclear power plants (sized around 1 GW of power) can create nearly 1,700 jobs for annual operations and more than 7,300 jobs among the supply chain and local businesses as a result of increased household spending. Rural projects also provide significant benefits, but at lower magnitudes compared to urban deployments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Radiological Releases from Novel Fuel Forms in Advanced Reactors During Severe Accidents for Consequence Analyses

Various advanced reactor developers are exploring the potential for reductions in the size of physical security forces and emergency planning zones. These reductions are based on robust fuel forms and inherently safe reactor designs. However, such reductions in physical protection measures could increase the risk of sabotage. To assess the possibility of reducing these measures, sabotage-induced radiological consequence analyses were carried out. These analyses considered accident scenarios that were beyond design basis accidents and overly conservative (Shah, 2025a; Shah, 2025b; Shah and Hartanto, 2026), yielding very large release fractions. These fractions, which can be used to evaluate physical protection and emergency planning requirements, have been crudely determined and applied as demonstrations for a sodium-cooled fast reactor (SFR) (Shah and Hartanto, 2025a), a high-temperature gas-cooled reactor (HTGR) (Shah and Hartanto, 2025b), a heat pipe–cooled reactor (HPR) (Shah and Hartanto, 2025c), and a molten salt–cooled reactor (MSR) (Shah et al., 2026). A Sandia National Laboratories (SNL) team used MELCOR—a fully integrated severe accident analysis code—to demonstrate the code’s capability to analyze advanced (i.e., not light water–cooled) reactors (including a fluoride salt–cooled high-temperature reactor [FHR]) and calculate radiological releases to the environment during severe accidents (Wagner et al., 2022a, 2022b, 2022c, 2023a, and 2023b). Although the analyses were carried out to demonstrate MELCOR’s growing capability, the release source terms were estimated for advanced reactors, providing valuable insights into the accident progression and radiological releases. These findings from prior SNL studies, including estimated source terms and related sensitivity studies, were leveraged to derive source terms for postulated sabotage-induced accidents. Insights from these sensitivity studies informed the scaling of SNL’s estimated source terms for the defined accident scenarios. The derived release fractions for the severe accident scenarios for the respective reactor designs can be used to perform more nuanced dose consequence analyses to evaluate the reactors’ physical protection and emergency planning zone requirements. These analyses are in accordance with the risk-informed, performance-based approach proposed under 10 CFR Part 53. This study builds on the prior source term analyses and associated sensitivity studies by SNL to derive time-dependent and design-informed release fractions. Section 2 describes the diverse advanced reactor designs analyzed by the SNL team. Section 3 discusses the severe accident analyses, the release fractions calculated, and the limitations and assumptions of the demonstration project. Section 4 presents the release percentages derived for the hypothetical sabotage-induced severe accidents at the advanced reactors. Section 5 summarizes the study’s findings and conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Charge Reduction and Performance Analysis of a Heat Pump Water Heater Using R290 as a Refrigerant—A Field Study

Heat pump water heaters (HPWHs) are a proven technology for water heating that has been commercialized. The adoption of HPWHs for domestic and commercial water heating is growing rapidly because of their superior performance compared with alternative water heating methods. Whereas most existing systems use R-134a as a working refrigerant, R290 has gained major attention owing to its superior thermodynamic properties. The goal of the current study is to assess the performance of residential HPWH with R290 as a direct refrigerant replacement for R134a. Two units of a 50 gal HPWH were used in this experimental study. A baseline unit contained R134a refrigerant, and a prototype unit contained R290 refrigerant. The prototype unit was developed through the modification of a commercially available HPWH unit to achieve a low charge of R290 refrigerant. Another major modification was the replacement of the baseline compressor with a compressor designed for R290. Tests were conducted in a field environment (a research and demonstration house) using programmed drawn profiles daily. The prototype that reduced the charge by 43–47% provided displayed performance comparable to the baseline unit regarding first-hour rating (FHR) and the uniform energy factor (UEF).

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2024

Evaluating Nuclear Forensic Signatures for Advanced Reactor Deployment: A Research Priority Assessment

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Fluoride Salt-Cooled High-Temperature Reactor Development Roadmap

Fluoride salt-cooled high-temperature reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics and fully passive safety. This paper provides an overview of a technology development pathway for expeditious commercial deployment of first-generation FHRs. The paper describes the principal remaining FHR technology challenges and the development path needed to address the challenges. First-generation FHRs do not appear to require any technology breakthroughs, but will require significant technology development and demonstration. FHRs are currently entering early phase engineering development. As such, the development roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant; the lack of an approved licensing framework; the lack of qualified, salt-compatible structural materials; and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

Holcomb, David Eugene↗
DOE OSTI2015

Improving Nuclear System Economics using Firebrick Resistance-Heated Energy Storage (FIRES)

As the electricity grid continues to implement unstable and intermittent energy resources in place of conventional dispatchable fuels, highly responsive energy storage systems will become an essential part of grid operation, to assure grid reliability, and reconcile electricity supply and demand. FIRES coupled with high-temperature heat systems such as NACC or industrial furnace enables different operating modes capable of yielding economic advantages to those who make the investment in storage, while also bringing relief to the grid by responding to the highly variable supply caused by wind and solar installations: In the case of the industrial furnace, by varying its own demand, while in the case of the FHR, by switching from a large supplier of electricity to a large consumer at times of low electricity prices. It should also benefit base-load nuclear plants by eliminating very low electricity prices as seen in some areas of the U.S. with significant installed wind or solar capacity—while lowering greenhouse gas emissions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Radiative Heat Transport and Optical Characterization of High Temperature Molten Salts

The goal of this project was to advance the field of molten fluoride radiative heat transfer (RHT), through optical property measurements, heat transfer loop construction and operation, and computational fluid dynamics simulations. The following outcomes have been accomplished. A comprehensive review has been conducted to compile all available RHT-relevant optical absorption data for molten fluorides. The various measurement methods have also been compiled and critiqued. Thermal emission measurement methods have been refined and documented. A unique thermal emissivity measurement facility has been constructed and used to measure the emissivity of flowing FLiBe-exposed SS316 samples. A slight increase has been observed in the emissivity compared to unexposed SS316. A dual reflectance-emission setup was partially demonstrated for the purpose of molten fluoride absorption measurement. Preliminary infrared edge data for FLiNaK was obtained. A visible near-infrared (NIR) measurement setup has been constructed and will be used to measure absorption of corrosion products in FLiBe. A new multi-band thermal radiation method using blackbody-weighted spectral bands for the absorption-coefficient was deployed to investigate the thermal performance of the UW Natural Circulation FLiBe Loop under various power levels and operating conditions. The results showed an increasing and decreasing trend in the overall heat transfer with optical thickness with a maximum at optical thickness τ D ≈Ο(1). A forced convection FLiBe loop was constructed for the purpose of studying corrosion and thermal hydraulic phenomena. A vertical cantilever pump is used to achieve up to 5 gpm and 5 m/s flow velocity in hot and cold test sections. A methodology has been established for modeling radiative heat transfer (RHT) in molten salt coolants for the purpose of quantifying the contribution of thermal radiation in conditions relevant to FHR or MSR systems. The comparison between the results obtained for forced and mixed convection showed that in the studied ranges of optical thickness, the maximum effect of RHT on overall heat transfer was 60% improvement observed at z/D = 250 for forced convection; whereas for mixed convection, that value was higher at 170% observed at z/D = 30. The highest effect of RHT on the overall energy distribution was observed at τ D ≈3.4. Negligible effect (<20%) was observed above τ D =18 and below τ D ≈0.002. A CFD code-to-code comparison was conducted using COMSOL Multiphysics at UW-Madison and STAR-CCM at MIT, for a forced and mixed convection laminar flow for a grey medium in a vertical heated using both COMSOL Multiphysics at UW-Madison and STAR-CCM at MIT. The relative error was the largest at τ D =40 for the radiative heat flux, bulk temperature, and Nusselt number. On the other hand, the relative error was the least at τ D =0.002 for the conductive heat flux, bulk temperature, and Nusselt number.

Anderson, Mark↗
DOE OSTI2022

Model-based Co-Simulation of Heat Pump Water Heater with Phase Change Materials Thermal Energy Storage

The study analysis the behavior of a new developed heat pump water heater technology which integrates a phase change materials storage with a standard heat pump water heater to maximize the performance parameters of the Unified Energy Factor (UEF) and First Hour Rating (FHR). A model-based control development co-simulation platform is developed to include equipment models, such as heat pump, standard water tank, phase change materials storage tank, and integrate them with control model to support controls design, analysis, verification, and validation. Simulation results are compared with lab test results to validate the accuracy of developed co-simulation platform

Sun, Jian↗
DOE OSTI2023

Fission Product Transport in TRISO Particles and Pebbles

This document demonstrates completion of the goals described in the technical narrative of the FOA project titled: ”Modeling and Simulation Development Pathways to Accelerating KP-FHR Licensing” regarding fission product transport in the Kairos-proposed fuel pebble by INL and Kairos Power. Showcased in this report are code developments and simulations in BISON that extend the state of the art in computation and understanding of fission product transport in a TRISO fuel particle and pebble. These enhancements lay the foundation for making predictions of fission product transport that can be used as input in the fuel licensing process. This was achieved by installing existing fuel material models originally used in PARFUME, developing a new failure probability method that is efficient and multi-dimensional, employing material homogenization, and expanding verification and validation simulations to demonstrate the efficacy of the work. All this work is leveraged to spotlight the main deliverable; a three-dimensional model and corresponding demonstration simulation of a pebble, which will serve as the starting point for models used to predict fission product release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Current status of the advanced high temperature reactor

The Advanced High Temperature Reactor (AHTR) is a design concept for a central station type [1500 MW(e)] Fluoride salt-cooled High-temperature Reactor (FHR) that is currently under development by Oak Ridge National Laboratory for the U. S. Dept. of Energy, Office of Nuclear Energy's Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR design option exploration is a multidisciplinary design effort that combines core neutronic and fuel configuration evaluation with structural, thermal, and hydraulic analysis to produce a reactor and vessel concept and place it within a power generation station. The AHTR design remains at the notional level of maturity, as key technologies require further development and a logically complete integrated design has not been finalized. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)

A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Measurement of neutron spectra in the experimental reactor LR-0

The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important task is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Optimizing Short-Term Operation of a FIRES-like Industrial Heat Supply Thermal Energy Storage System

The Firebrick Resistance-heated Energy-storage System (FIRES) is under development at MIT for two applications: (1) integration with the Nuclear Air-Brayton Combined Cycle (NACC), a high-performance power conversion system (PCS) that may be used with high- temperature advanced reactors; and (2) industrial heat supply. This paper describes some key features of FIRES and equivalent systems configured for industrial heat supply and presents an assessment of the revenue achievable by a system operating under the conditions expected for early demonstration FIRES installations. The greatest technical challenge of the 21. century is likely to be the effort to minimize the damage caused by global climate change. This effort will require massive reductions in greenhouse gas (GHG) emissions throughout our economy and a transition to an electric power system driven entirely by very-low-carbon generators. Many investigators agree that this will require a substantial increase in electricity generation and in the use of nuclear power, and there is growing awareness that this will also require new technologies and systems for managing and storing energy and matching intermittent supply of renewable resources with electricity demand. Thermal energy storage systems (TESS) will, in some applications, be much more cost effective than the more popular and widely discussed battery electric storage systems and may provide unique capabilities. The Firebrick Resistance-heated Energy-storage System (FIRES) is under development at MIT for two applications: (1) integration with the Nuclear Air-Brayton Combined Cycle (NACC), a high-performance power conversion system (PCS) that may be used with high-temperature advanced reactors like the Fluoride-salt-cooled High-temperature Reactor (FHR), molten salt reactor (MSR) and sodium-cooled fast reactors (SFR); and (2) industrial heat supply. FIRES installations include a stack of refractory firebrick with imbedded electric heaters and airflow channels, a compressor or air blower, and a duct system to connect the firebrick with the heat demand (turbine or industrial customer). Analysis of the FIRES design space has shown that the concept is suitable for a wide range of peak operating temperatures (from about 200 deg.C up to at least 1800 deg. C) and pressures (from ambient to at least 20 atmospheres) and offers very high energy density (up to 1 MWh per cubic meter of firebrick for high temperature systems) at very low cost (as low as $5/kWh). These results favor early demonstration of FIRES. If the cost target of $5/kWh for FIRES systems were achieved, even the lowest expected operating income calculated here could cover the capital cost of a 10 MWh system in a single year. If the FIRES system were assumed to operate for at least 10 years, this level of operating income could readily recover a capital investment up to $50/kWh, which is simultaneously much cheaper than the costs expected for near-future battery systems (in excess of $200/kWh) and much more expensive than the small, low temperature firebrick systems already available for domestic heating. This provides a large margin for unexpected development costs and other uncertainties. FIRES would set a minimum price of electricity near that of natural gas-improving nuclear, wind, and solar economics while reducing greenhouse gas emissions from the industrial sector. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Preliminary neutronic assessment of a novel fluoride-salt-cooled high-temperature reactor design

The Fluoride-Salt-Cooled High-Temperature Reactor (FHR) is a molten salt cooled, graphite moderated high temperature reactor. In this work, single-batch cycle length, block power distributions, and temperature-related reactivity effects were assessed using the SCALE code suite. Two types of parametric studies were performed. The first consisted of fixed temperature (isothermal) cases in which to determine the relationship between operation power and cycle length as well as pin power distributions. This second study calculates various isothermal reactivity coefficients based on heterogeneous temperature distributions. Reactivity coefficients with respect to the temperature of the fuel, coolant, and moderator were performed separately to determine their individual contributions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Benchmark Study of Infinite Lattice of Pebble Fuel within FLiBe Coolant for TMSR-SF1 Design

Solid Fueled Thorium Molten Salt Reactor (TMSR-SF1) is Chinese Fluoride-salt-cooled High Temperature Reactor (FHR) designed by the TMSR group. The construction of this test reactor is expected in the near years. Using TRISO particle based pebble fuel leads to various challenges in neutronics modeling, such as double heterogeneity effect, preserving exact number of particles, their random distributions in pebble, and etc. For the purpose of verifying the calculations, benchmark study of infinite lattice of pebble fuel within FLiBe coolant for the TMSR-SF1 design has been performed. Back-to-back comparisons have been made between TMSR and MIT researchers using different Monte-Carlo based codes. Very good agreements have been found and the results quantitatively evaluate the neutronic effect of various modeling approaches. In addition, sensitivity study has been carried out for the reference pebble-bed fuel lattice. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

New Czech R and D Program on Fluoride Salt-cooled Nuclear Reactor Systems

A technology of nuclear reactor systems with liquid molten salt fuel has been investigated in the Czech Republic since 1999. The original effort came from the national Partitioning and Transmutation concept based on the subcritical Accelerator Driven System for incineration of transuranium elements with liquid fluoride fuel and pyrochemical partitioning fuel cycle technology. After 2005, the original R and D intentions were gradually converted to classical Molten Salt Reactor (MSR) technology and to thorium - uranium fuel cycle. The aim of this choice was to contribute to the development of an advanced nuclear technology, which can minimize environmental impact of nuclear power, save the natural resources and which has some potential to be deployed also in a non-superpower country. The basic technological development of selected areas of MSR and Th - U fuel cycle technology was realized in the frame of the SPHINX project solved in 2005 - 2012. In addition to the domestic activities of the MSR technology development, in 2012 the Czech Ministry of Industry and Trade and the US - Department of Energy concluded an agreement about the collaborative R and D on Molten Salt Reactor and Fluoride-salt-cooled High-temperature Reactor (FHR) technologies. These facts and results created the principal background for the present national program on the MSR/FHR technology development which was approved by the Ministry of Industry and Trade in 2016 and is granted by the Technological Agency of the Czech Republic.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗