Fluoride Salt-cooled High-temperature Reactor Research Articles

Evaluation of mitigation strategies for radioactive fission gases in fluoride-salt-cooled high-temperature reactors

The Fluoride-salt-cooled High-temperature Reactor (FHR) is a promising Generation IV nuclear reactor due to its improved safety design, near atmospheric working pressure, and high electric power conversion efficiency. However, control of radioactive fission gases, such as tritium, is a critical issue in FHRs due to the significantly larger production rate and potentially larger leakage rate of tritium compared to those in Light Water Reactors (LWRs). Therefore, this study proposes two mitigation strategies for tritium in FHRs: Double-Wall Heat eXchanger with a Tritium Carrier (DWHX-TC) option and Single-Wall Heat eXchanger with a Tritium Barrier (SWHX-TB) option. For the DWHX-TC option, two tube surface configurations, i.e., plain and fluted tubes, and four tritium carriers, i.e., helium, FLiBe, FLiNaK, and KF-ZrF4, are initially considered, among which the best tube configuration and promising tritium carrier are then selected and used for design optimization of the DWHX. For the SWHX-TB option, two cases, tritium barriers with and without cracks, are investigated in this study. In addition, the two mitigation strategies are evaluated by an in-house, one-dimensional, coupled HEat and MAss Transfer (HEMAT) code developed to predict the leakage rate of tritium in nuclear systems. Our analysis shows that (1) a double-wall configuration using fluted tubes as both the inner and outer tubes is better than other double-wall configurations investigated; (2) helium, as a tritium carrier, is superior to other candidates, including FLiBe, FLiNaK, and KF-ZrF4; (3) the leakage rate of tritium (into the atmosphere) is reduced from 3400 Ci/day, taking the Advanced High-Temperature Reactor (AHTR) as a reference, to 4.0 Ci/day for an optimum design of the DWHX-TC option using helium as the tritium carrier and 1.7 Ci/day for an optimum design of the SWHX-TB option using 30-µm thick silicon carbide (SiC) as the tritium barrier (no cracks); and (4) potential cracks in tritium barriers may significantly deteriorate mass transfer performance of the SWHX-TB option. For example, a cracking rate of 0.1% leads to an increase of 694.1% in the leakage rate under the conditions investigated.

Demonstration of a random sampling approach to uncertainty propagation for generic pebble-bed fluoride-salt-cooled high temperature reactor (gFHR)

Nuclear reactor safety analysis requires accurate propagation of uncertainty, a research topic that is rather underdeveloped for advanced reactors. We develop and demonstrate a framework for uncertainty analysis (UA) based on random sampling specific to a benchmark generic fluoride salt-cooled high temperature reactor (gFHR) core. This framework effectively encompasses uncertainties from nuclear data, manufacturing tolerances, and operational uncertainties such as thermophysical properties and material input uncertainties. These uncertainties are propagated to steady state figures of merit (k-effective/fuel temperature coefficient) and a single pebble depletion model.The gFHR core is modelled in SCALE 6.3b using core parameters from open-source data. Nuclear data input uncertainties are sampled from ENDF/B-VII.1 using the covariance-based XSUSA method via the Sampler sequence in SCALE. This article uses a standard statistical analysis based on normality and absolute margins to describe output probability distributions. The distribution of both steady state output quantities failed tests of normality; however, graphical inspection shows that the normal and uniform parameterizations sufficiently represent the output uncertainty. In practice, this may require more conservative engineering judgement. The mean and propagated uncertainty for k-effective and FTC were found to be 1.01271 ± 1.44% and −5.156 ± 8.22% (pcm/K) respectively. Distributions of depletion output quantities showed the trend of failing normality tests early and passing most often as burnup increased. Burnup (expressed in %FIMA) passed normality tests at end-of-life discharge (9 passes) with a mean and propagated uncertainty of 0.1877 ± 3.2% with 95% confidence that the true uncertainty falls between 3.012% and 3.312%. Absolute margins, or uniform parameterization, of the uncertainty distributions for steady-state and depletion quantities are also presented in the results section. Unique results from the nominal behavior of this depletion model are reflected in the depletion output distributions and explored further in a case study on the production chain of plutonium-239.

Application target image of a small modular fluoride salt-cooled high temperature reactor based on the analysis of energy demand and environmental conditions in western China

To meet the energy demand in remote areas of western China, a small modular fluoride-salt-cooled high-temperature reactor will be deployed there. The design of this reactor needs an application target image, including the power, lifetime, size, weight, and environmental restrictions. This paper analyzes the energy demand in northwest China to find the possible application scenarios and regions of this reactor. Then according to typical application scenarios, the power requirements of the reactor and the environment and transportation (size and weight) restrictions that need to be adapted can be obtained. The application target image of this reactor would be: a) the single unit capacity of 50MWe; b) the lifetime of at least 60 years; c) the length of less than 15.5 m, the diameter of less than 3.88 m, and weight of less than 60 tons; d) the adaptability to the severe climate and environment in the west, such as cold winter, deep-frozen soil layer, strong wind, arid environment, and complex terrain; e) the site selection to avoid geological disasters, such as landslides, collapses, torrents, mudslides, and ground fissures.

Physics Modeling for Conceptual Designs of Proposed Experiments in the Zero Energy Deuterium-2 Critical Facility for Testing Small Modular Reactor-Type Fuels

Abstract To enable the deployment of small modular reactors (SMRs) in Canada, Canadian Nuclear Laboratories (CNL) is interested in the applicability of zero energy deuterium-2 (ZED-2) experiments for validation of Monte Carlo reactor physics codes and models for SMR reactor fuels. This applicability is investigated here by considering potential ZED-2 experiments for testing fuel assemblies (FAs) for two different SMR technologies: the pressurized-water reactor (PWR), and the fluoride-salt-cooled high-temperature reactor (FHR). Each proposed set of mixed-lattice substitution experiments uses driver fuel channels containing CANDU flexible low enriched uranium/recovered uranium (CANFLEX-LEU/RU) fuel bundles. Simulation results indicate that a number of critical core configurations are possible, and should provide suitable reactor physics measurement data that can be used for physics design verification, and also for benchmarking and validation of computational reactor physics codes used in the design, operations, and safety analysis of SMRs based on PWR or FHR technologies.

Design space exploration for salt-cooled reactor system: Part I – Thermal hydraulic design

Fluoride-salt-cooled High-temperature Reactors (FHRs) are a variant of nuclear reactor offering high power densities and operating temperatures. This paper proposes a new design stream of FHR, leveraging developments and experience from the British Advanced Gas-cooled Reactor (AGR) fleet, the only commercial fleet of high-temperature reactors to-date. The design space for the proposed AGR-like FHR is explored using a thermal-hydraulic simulation and optimisation package, seeking to maximise the reactor’s economic performance. Starting from a salt-cooled AGR assembly and applying reasonable thermal-hydraulic constraints, this exploration includes varying the assembly fuel type and geometry, coolant salt, as well as design of the pumps and heat exchangers. Some of the resulting designs promise substantially higher power densities than AGRs and improved economics.

Assessment of SCALE and MELCOR for a generic pebble bed fluoride high-temperature reactor

This paper presents the development and application of SCALE and MELCOR models for a fluoride salt-cooled high-temperature reactor (FHR) based on publicly available specifications. SCALE version 6.3beta15 was used to generate power distributions and decay heat curves, and MELCOR version 2.2.18019 was used to calculate the thermal hydraulic response of an assumed FHR primary system. An approach was developed for determining the equilibrium state of the core using depletion of a core slice model and blending of fuel compositions at different burnups to provide three-dimensional fuel composition in the core. Results compared between a core with entirely fresh fuel and one with an equilibrium fuel composition revealed that the equilibrium core led to lower steady-state temperatures but slower cooldown during a loss of flow accident (LOFA). A sensitivity study was then conducted to explore the transient response of our FHR system to variations in thermal hydraulic parameters of the system using the equilibrium core. The inlet temperature, graphite thermal conductivity, and SCRAM time all provided significant control over peak fuel and coolant temperatures. Uncertainties in radionuclide decay data in SCALE were used to perform a decay heat sensitivity study, and we found that uncertainties in decay heat led to negligible impact on peak temperatures during the course of the transient. In all cases evaluated, the observed peak fuel temperatures remained approximately 700 K below anticipated failure limits for the LOFA.

Atomistic and cluster dynamics modeling of fission gas (Xe) diffusivity in TRISO fuel kernels

TRISO fuel particles are candidates for use in next generation reactors including gas reactors, fluoride salt-cooled high temperature reactors, and micro-reactors. The UCO fuel kernel consists of a uranium dioxide (UO2) and uranium carbide mixture. The addition of UC2 helps suppress the formation of carbon monoxide gas, which led to failures during initial TRISO development. The addition of uranium carbide alters the chemistry of the UO2 kernel, which is known to influence performance parameters such as fission gas diffusivity, although the impact has not been quantified and no models exist that take the change in chemistry into account. Therefore, better understanding and more accurate models of the impact of chemistry on fuel performance are of high priority. In this paper, a first-principles density functional theory (DFT) and empirical potential based multi-scale study has been carried out to model the diffusivity of fission gas xenon (Xe) in UCO TRISO fuel kernels. The focus is on the UO2 component in the UCO fuel kernels, as that represents the largest volume fraction of the fuel kernels. The study relies on DFT and empirical potential calculations to determine Xe and point defect properties, which are then used in thermodynamic and kinetic models to predict diffusion for intrinsic conditions. In addition, the information is utilized in cluster dynamics simulations using the Centipede code to estimate the impact of irradiation on defect transport. The presence of UC2 or UC2−x in the UCO fuel kernels is shown to have a substantial impact on the UO2 non-stoichiometry by inducing oxygen vacancies and driving UO2 sub-stoichiometric, which causes much slower Xe diffusion in UCO compared to light water reactor UO2 fuel. The application of this model in fuel performance simulations using the Bison code is also demonstrated.

Performance evaluation of DRACS system of molten salt reactors using a transient solidification model

A transient solidification model was developed and coupled to ASYST code to simulate the transient behavior of the direct reactor auxiliary cooling system (DRACS) of the fluoride-salt-cooled high-temperature reactor (FHR) during the loss of heat sink (LOHS) accident. The model was firstly validated by comparing the experimental data with the simulation results that deriving from available heat transfer and friction factor correlations. The simulation results are found to be well-fitted with the experimental data. The transient solidification model successfully predicted the occurrence of solidification and the complete blockage of the heat transfer pipes in DRACSs. The solidified layer is found to be beneficial for the decay heat removal at the early stage of solidification, however, a sudden clog is found at the end stage of the solidification. Based on the simulation results, an optimization attempt was made to extend the failure time of DRACS by reducing the length of heat transfer tubes. The model is applicable to all the reactors where solidification may occur, such as other molten salt reactors (MSRs) and lead-cooled fast reactors (LFRs).

Advanced gas-cooled reactors technology for enabling molten-salt reactors design – Optimisation of a new system

Molten salts present significant advantages as coolants over traditional light water or gas. Salt-cooled nuclear reactor can potentially i allow an increase in core power density and simplify the reactor safety case. Traditionally, molten salt reactors assume molten salt being both the fuel and the coolant. This, however, creates major challenges for the design and operation due to requirements for chemistry control and corrosion behaviour of the core structural materials. The Fluoride salt-cooled High-temperature Reactors (FHR) on the other hand, have conventional solid fuel geometry surrounded by moderator and coolant. Furthermore, some FHR variations share many common characteristics with the Advanced Gas-cooled Reactors (AGR). Thus, using the knowledge accumulated during many decades of successful AGR operation can speed up the FHR development and deployment. However, replacing carbon-dioxide coolant by molten salt significantly changes the reactor performance. Hence, a new core design is needed. In this work, a Multi-Objective Particle Swarm Optimisation is used to identify the most favourable configurations for a new system layout. Initially, the optimisation targeted the beginning of cycle parameters such as criticality and Coolant Temperature Coefficient (CTC). The present work is the next step in this analysis. Each configuration is examined with respect to its thermal-hydraulic performance to assess the power uprate potential which is limited by multiple temperature constraints (e.g. fuel centreline and cladding temperatures as well as the coolant freezing/boiling). The estimated maximum power was then used in the fuel burnup calculations, from which a discharge burnup and cycle average CTC were obtained. As a result of the optimisation process, several families of possible solutions were identified, which form an optimal Pareto front. The most attractive configurations in terms of achievable power density, however, were not necessarily on the Pareto front. Most of the identified design options had only a small amount of graphite moderator or no graphite at all, relying entirely on moderation in the salt. Increasing the graphite volume was consistently found to worsen most of the optimisation parameters. The newly identified design options have the potential to achieve power density which is higher than that of a typical AGR by up to a factor of five, while maintaining negative CTC through the burnup cycle.

Neutronics, thermal-hydraulics, and multi-physics benchmark models for a generic pebble-bed fluoride-salt-cooled high temperature reactor (FHR)

• A generic FHR design benchmark data set. • Neutronic, Thermal-Hydraulic and Pebble flow analysis. • Coupling of Monte Carlo, Porous Media and Discrete Element Modeling. • Multi-pass fuel cycle analysis with Monte Carlo based depletion calculation.

Evaluation of a freeze-tolerant decay heat removal system redundancy for fluoride salt-cooled high-temperature reactors (FHR)

Typical FHRs use a direct reactor auxiliary cooling system (DRACS) for decay heat removal under a variety of accident scenarios where active normal shutdown cooling is unavailable. Response surface method has been used for evaluating the reliability of the passive system. In this paper, a loss of heat sink (LOHS) accident sequence is evaluated using the response surface methodology. Response surfaces on maximum core outlet temperature and DRACS freezing time has been presented for evaluating the effects of friction loss, heat transfer coefficients and partial and full blockages of the DRACS. The present study shows that for one DRACS failure situations, the system is unconditional safe. But, for two DRACSs failure situations, there is probability that peak core outlet temperature (PCOT) will not exceed the allowable temperature. During the LOHS accident, the DRACS would not freeze within 3.33 days. This study will provide an insight for investigating DRACS redundancy and functional failure.

Engineering demonstration reactors: A stepping stone on the path to deployment of advanced nuclear energy in the United States

Nuclear energy is an important ultra-low carbon electricity source. Engineering demonstration reactors, such as Shippingport for Pressurized Water Reactor technology, have been used to advance nuclear reactor technology to deployment. It may be necessary to develop similar engineering demonstration reactors for other reactor concepts that have not yet been built. This paper reviews the historical role of engineering demonstration reactors and highlights the potential for future engineering demonstrations of Fluoride Salt-Cooled High Temperature Reactors (FHRs) and Molten Chloride Fast Reactors (MCFRs). Future engineering demonstration reactors could significantly accelerate deployment of larger-scale FHRs and MCFRs by the advanced nuclear industry via cost and risk reduction. Experience in the United States shows that sustained investment in a public-private partnership, such as the current Advanced Reactor Demonstration Program, is essential for deployment of advanced reactor technologies that have never been built before. The National Reactor Innovation Center at Idaho National Laboratory and Clinch River Site near Oak Ridge National Laboratory are examples of ideal locations for these demonstrations.

Determination of lattice physics properties and uncertainties in a solid fuel molten salt cooled assembly using OpenMC

The desire for reducing our carbon footprint is driving significant interest in advanced nuclear reactors to produce energy. One promising technology is the Fluoride Salt-Cooled High-Temperature Reactor (FHR) which provides both electricity and high temperatures for use in process heat applications. The reactor technology considered in this work is a Generation IV design that uses TRISO particles uniformly embedded in fuel stripes within the fuel plates and liquid salt as coolant. The fuel design and conditions are the same as those used in a recent OECD/NEA FHR Benchmark. This paper reports the multiplication factor, fission reaction rate, and neutron spectrum for different cases from that benchmark study. A unique contribution of this work is the assessment of nuclear data uncertainty impact on the Figures of Merit used in the FHR benchmark. The OpenMC Monte Carlo code and perturbed nuclear data files from the TENDL-2017 library were utilized in these analyses. This work is part of the ongoing international efforts to explore and identify the applicability of current methods and codes for FHR simulations and molten salt reactors in general.

Preliminary design and assessment of a heat pipe residual heat removal system for the reactor driven subcritical facility

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.

The impact of neutron irradiation, graphite oxidation and fluorination on tritium uptake into and desorption from graphite in molten salt environments

Tritium management is necessary in both fission and fusion nuclear reactors. In fusion reactors, tritium is a fuel that needs to be produced in breeding blankets. For fission reactors, and especially Fluoride Salt-Cooled High-Temperature Reactors (FHRs) and Molten Salt Reactors (MSRs), tritium is a contaminant to be separated and removed. The current literature on high-temperature hydrogen-graphite interactions is generated predominantly by the fusion research community and does not yet cover the low-pressure interval of relevance for FHRs and MSRs. Predictions of graphite uptake capacities and uptake rates at FHR conditions of few Pa partial pressure can only be performed using extrapolations. In order to make reliable extrapolations from the available data, the impact of phenomena that take place during operation and accident scenarios, such as neutron irradiation, air oxidation and reactions with molten fluoride salts must be accounted for. This article provides a summary of hydrogen-graphite thermodynamics and kinetics of interaction and discusses the effects of irradiation, air oxidation and fluorination on uptake capacities, uptake and desorption kinetics. We find that all three phenomena increase uptake capacities in graphite. Neutron irradiation and reactions with fluoride salts are expected to reduce tritium uptake rate, while oxidation is expected to increase it. In all three cases, the changes are more pronounced at low tritium partial pressures.

Code validation of SAM using natural-circulation experimental data from the compact integral effects test (CIET) facility

The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) experimental loop. SAM is a modern system analysis code being developed at Argonne National Laboratory for safety analysis of designs for advanced non-light water reactors (non-LWRs), such as sodium-cooled fast reactors, high-temperature gas-cooled reactors, and fluoride salt-cooled high-temperature reactors (FHRs). To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments highly relevant to these reactor concepts. The CIET facility, which was designed to reproduce the thermal-hydraulics response of FHRs under both forced- and natural-circulation conditions, has been identified and selected as one of the benchmark test facilities for SAM code validation. In this study, two sets of available CIET tests were selected for SAM code validation purposes, namely, forced convection cooling transient test and steady-state natural-circulation tests. For all selected tests, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.

Optimisation of AGR-Like FHR Fuel Assembly Using Multi-Objective Particle Swarm Algorithm

Utilising molten salt as coolant instead of carbon dioxide in traditional advanced gas-cooled reactors (AGRs) can potentially increase their core power density, simplify the safety case and shorten the time needed for the development of the fluoride-salt-cooled high-temperature reactor (FHR). However, the change of coolant has a strong impact on the system behaviour. Therefore, a new type of fuel assembly is required. However, the design of a new assembly is affected by a wide range of parameters. Systematic search through all the potential configurations is prohibitively computationally expensive. In this work, a multi objective particle swarm optimisation (MOPSO) algorithm is utilised to identify the most attractive candidate configurations for the hybrid AGR-like FHR assembly. The first optimisation step targets basic design parameters such as radius and enrichment of the fuel pins, their number and arrangement. MOPSO is based on the concept of Pareto dominance, which is used to determine the flight direction of the simulated particles. The outcome of the optimisation process provides insight on families of possible solutions, which described by the Pareto front.

Modeling Tritium Retention in Graphite for Fluoride-Salt-Cooled High-Temperature Reactors

Mitigating the release of tritium produced from neutron irradiation of molten salts containing lithium or beryllium is a technical challenge for several advanced reactor designs. In a pebble bed Fluoride-Salt-Cooled High-Temperature Reactor (FHR), tritium generated in the Li2BeF4 (Flibe) coolant is expected to interact with the large inventory of graphite in the core. The degree to which tritium is retained in the FHR core graphite is important to understand in order to predict the tritium distribution in the reactor, operational dose rates in the plant, tritium source term, and optimal strategies to mitigate environmental release. Tritium retention in graphite is simulated in this work based on a model that considers tritium diffusion from Flibe into graphite pores as well as diffusion and trapping in graphite grains. The retention model was implemented into the TRIDENT model framework to study tritium transport at the FHR system level. Tritium permeation through the FHR primary heat exchanger was the largest source of release from the primary system, followed by tritium retention and recirculation of graphite fuel pebbles.

ROM-Based Surrogate Systems Modeling of EBR-II

The System Analysis Module (SAM), developed and maintained by Argonne National Laboratory, is designed to provide whole-plant transient safety analysis capabilities for a number of advanced non–light water reactors, including sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and molten salt reactor (MSR)/fluoride-salt-cooled high-temperature reactor (FHR) designs. SAM is primarily constructed as a systems-level analysis tool, with the potential to incorporate reduced order models from three-dimensional computational fluid dynamics (CFD) simulations to improve characterization of complex, multidimensional physics. It is recognized that the computational expense associated with CFD can be intractable for various engineering analyses, such as uncertainty quantification, inference, and design optimization. This paper explores the reducibility of a SAM model using recent advances in randomized linear algebra techniques, which attempt to find recurring patterns in the various realizations generated by a model after randomly perturbing all its input parameters. The reduction is described in terms of fewer degrees of freedom (DOFs), referred to as the active DOFs, for the model variables such as input model parameters and model responses. The results indicate that there is significant room for additional reduction that may be leveraged for additional computational gains when employing SAM for engineering-intensive analyses that require repeated model executions. Different from physics-based reduction approaches, the proposed approach allows one to estimate upper bounds on the reduction errors, which are rigorously developed in this work. Finally, different methods for surrogate model construction, such as regression and neural network–based training, are employed to correlate the input and output active DOFs, which are related back to the original variables using matrix-based linear transformations.

Photoneutron production and characterization in fluoride-salt cooled high-temperature reactors

Due to the presence of beryllium-bearing salts, fluoride-salt cooled high-temperature reactors (FHRs) are expected to have higher production of delayed photoneutrons than other common power reactor systems which may impact transient behavior. New methods for computationally quantifying the characteristics of delayed photoneutrons using Monte Carlo particle transport, including absolute fraction, effective fraction, and group decay constants, are outlined. These methods are validated where possible against existing data from CANDU reactors, and then used to make a best-estimate characterization of the delayed photoneutron effect in a prototypical FHR design. It is found that delayed photoneutrons have a worth of approximately 9 pcm in the reference FHR system, and that their importance is slightly higher than that of fission neutrons due mostly to their softer spectrum. Existing emission group structures from literature have been compared to computational data and found to be lacking, particularly in their representation of groups with short half-lives. Therefore, a new 13-group structure is proposed for use in FHR dynamic calculations.