Fluoride Salt-cooled High-temperature Reactor Research Articles

Molten Salt Pump Journal-Bearings Dynamic Characteristics Under Hydrodynamic Lubrication Conditions

Abstract A reliable high-temperature molten salt pump is critical for the development of Fluoride-salt-cooled, High-temperature Reactors (FHRs). By supporting the rotating journal, the appropriate journal bearing can ensure that the high-temperature molten salt pump runs smoothly and efficiently in the high-temperature fluoride salt over a long period of time. However, many bearing candidates served well for only a short period and experienced several issues. Moreover, the alignment of the molten salt pump journal bearings is a key factor for the molten salt pump's long-term steady running. In the long-term operation, a misalignment in the journal bearing can result in vibrations and excessive wear on the bearing surface of the molten salt pump. The journal bearing dynamic characteristics can be used to accurately assess the journal misalignment. Therefore, it is necessary to investigate the detailed journal-bearing dynamic behavior under the high-temperature hydrodynamic fluoride salt lubrication conditions for FHR applications. In this study, a small amplitude vibration is superimposed on a steady-running journal bearing to simulate the molten salt operating conditions. A Fortran 90 code has been written for the journal-bearing dynamic behavior analysis. The code was verified using the numerical data reported in the literature. The code is then employed to predict the dynamic coefficients of high-temperature fluoride salt hydrodynamic lubricated journal-bearing with various Sommerfeld numbers. These journal-bearing dynamic coefficients can be used to provide guidelines in the design of molten salt pumps.

Study on tritium transport characteristics in fluoride-salt-cooled high-temperature advanced reactor

Fluoride-salt-cooled high-temperature reactors (FHRs) are a promising reactor class that combines prominent attributes from the Gen-Ⅳ reactors, raising attractive interests of the international community. However, tritium control has been a major challenge in FHRs because of the neutron reactions of lithium. It's necessary to study the transport characteristics of tritium in the fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) and to provide theoretical support for appropriate tritium control measures and safe operation of the FHRs. In this paper, the inclusive transport characteristics of production, graphite adsorption, diffusion, dissolution, and permeation on tritium in the primary loop are analyzed. The mathematical and physical model of the tritium transport phenomenon is established, and an analysis model of tritium transport characteristics is also developed. Then, the tritium transport characteristics in the FuSTAR primary loop are analyzed. The results show that the tritium in the FuSTAR primary loop mainly exists in the form of Diatomic Tritium (T2), and it's found that Tritium Fluoride (TF) and T2 are adsorbed by graphite after being generated in the core, and this process reaches saturation in about 150 Effective Full Power Days (EFPDs). It also indicates that almost 48% of tritium penetrates the secondary loop through the heat exchanger, and only about 3.8% of tritium is adsorbed on the core graphite. TF makes the corrosion of the heat exchanger and downcomer more serious. Finally, the safety of tritium transport in FuSTAR is conservatively estimated, with a tritium emission dose of 0.1674 mSv/a. This study can provide a theoretical reference for the structural design, tritium recycling, and radiation protection of FHRs.

System Thermal-Hydraulics Model for Fluoride Salt-Cooled Reactor Based on Small Modular Advanced High Temperature Reactor (SmAHTR) Design Concept

Abstract A system thermal-hydraulics model for a fluoride-salt-cooled high-temperature reactor (FHR) based on the small modular advanced high-temperature reactor (SmAHTR) design concept is developed, using RELAP5-3D. The SmAHTR components modeled in the simulations include the reactor core, lower plenum, upper plenum, top plenum, three primary heat exchangers (PHXs) equipped with three primary pumps, and three director reactor auxiliary cooling system (DRACS) equipped with three fluid diodes. Flows through the reactor core are represented by 19 individual fuel channels, one reflector-hole channel, and one downcomer channel. In each of the 19 SmAHTR fuel block channels, the fuel elements are modeled in five groups using five heat structures, each with their corresponding power level. The total reactor power is 125 MWth. Using representative core power distributions for the SmAHTR at beginning-of-cycle (BOC) and at end-of-cycle (EOC), two steady-state system thermal-hydraulic model simulations with RELAP5-3D were performed using a nominal pressure drop loss factor of 1.5 for all nine junctions in each of the 19 fuel channels modeled. Exit coolant temperatures ranged from 688 °C to 739 °C (BOC) and from 696 °C to 721 °C (EOC), while peak fuel centerline temperatures in the highest power block were 1249 °C (BOC) and 1029 °C (EOC). By adjusting the loss factors to modify coolant flow rates in each channel, a more uniform exit coolant temperature was possible, ranging from 701 °C to 709 °C (BOC) and 698 °C to 714 °C (EOC), while peak fuel centerline temperatures were reduced to 1204 °C (BOC) and 988 °C (EOC), giving results more consistent with earlier studies of the SmAHTR.

Recommendations for seismic analysis of a molten salt reactor

A Fluoride-salt cooled High-temperature Reactor (FHR) that uses circulating solid TRISO pebbles, as a fuel and positively buoyant graphite reflector blocks as a moderator is at an advanced stage of development. Seismic base isolation supports a pathway for cost competitive and rapid deployment of this molten salt reactor (MSR) in regions of different seismic hazard. This paper presents recommendations for analysis and modelling of different components of a base-isolated MSR, based on results of earthquake-simulator tests on a scale model of the reactor. Recommendations are provided for evaluating the sloshing behavior of the coolant in an annular region inside the reactor and for estimating hydrodynamic loads for use in analysis of structural and mechanical components. Hydrodynamic loading on the reflector blocks is discussed and a practical approach to estimate forces in the connections between them is presented. Approaches for modelling and analysis of two 2D horizontal base-isolation systems, utilizing spherical sliding bearings, are validated using results from experiments. The recommendations apply to reactor vessels with similar fluid-structure systems.

TRISO Burnup-Dependent Failure Analysis of a HTGR Design-Basis Accident Using BISON

This work assesses the failure behavior of the silicon carbide (SiC) layer in TRIstructural ISOtropic (TRISO) fuel particles with BISON during both steady-state and transient conditions for the MHTGR-350 design by General Atomics. A one-dimensional BISON model for a uranium oxycarbide–bearing TRISO is developed to simulate a power level of 50 mW/particles up to 12.4% fissions per initial metal atom (FIMA). Stress predictions for the materials of interest are presented as a function of burnup, along with the SiC failure probability computed using Weibull statistics under the assumption of realistic SiC quality. A design-basis accident (DBA) that involves control rod withdrawal is then simulated in BISON at various burnup levels. Boundary conditions during the DBA are obtained from RELAP5-3D. The predicted SiC failure probability is 3 × 10−14%. This value does not increase during the transient as a function of burnup since irradiation induces a compressive hoop stress state that reaches a maximum absolute value of around 300 MPa. This compressive stress compensates the tensile loading conditions introduced by the transient. The tensile components appear amplified at higher burnups since they increase from 100 to 140 MPa going from 6% FIMA to 12.4% FIMA. Nonetheless, burnup provides a negligible impact on the overall failure probability predictions for SiC, as transient conditions do not translate to a stress increase toward tensile values at any of the considered burnups. Additionally, a two-dimensional (2-D) exploratory BISON model is developed to determine the effects of inner pyrolytic carbon (IPyC) crack formation on SiC. IPyC failure is predicted using Weibull statistics at 105 days from the beginning of irradiation, which is set as the time for crack initiation using XFEM in the 2-D BISON model. Stress concentrations induced by the crack cause the SiC failure probability to increase with respect to the intact particle case by approximately 10 orders of magnitude. Maximum stress concentrations are found at the time of crack formation, after which stress relaxation is observed during remaining steady-state irradiations and subsequent transient simulations. Considerations are made on the results of the 2-D model being contingent on both mechanical and anisotropy properties for pyrolytic carbon. Recommendations are provided for future work involving higher dimensionality models, further investigations on uncertainties, material properties, and additional designs, such as the fluoride salt-cooled high-temperature reactor, that operate at higher burnups.

Neutronic and thermal-hydraulic calculations under steady state and transient conditions for a generic pebble bed fluoride salt cooled high-temperature reactor

Here, we develop multiphysics coupled neutronics/thermal-hydraulics models for a recently available generic Fluoride-salt-cooled High-temperature Reactor (FHR) at equilibrium core burnup conditions with reactivity control mechanisms. We utilized the pebble bed reactor transient tool KP-AGREE with multigroup homogenized cross-sections generated by the continuous energy Monte Carlo neutron transport code Serpent to simulate transient events for an FHR. Comparing global reactor parameters and safety characteristics between Serpent and published results to standalone neutronic calculations in KP-AGREE verifies our initial conditions for our transient models. This work explores three separate transients: a loss of flow accident, an overcooling accident, and a total control rod withdrawal accident. The boundary conditions of these scenarios come from either transient benchmark definitions of a gas-cooled pebble bed reactor or transient analysis of a similar FHR design. For the loss of flow accident, our predicted peak fuel and coolant outlet temperatures show a similar response to published values, albeit slightly higher, with differences between values attributed to our modeling assumptions. The overcooling transient simulation predicts a 52 K temperature increase in the maximum fuel kernel temperature and an approximately 32.6% increase in total core power due to the negative coolant temperature feedback of the FHR design. Our control rod removal transients predicted a 235 K increase in maximum fuel kernel temperature and a 107.5% increase in total core power. For all transient cases, the fuel temperature remained significantly lower than the temperature limit, but coolant outlet temperatures for the reactivity-initiated accident exceeded external core material temperature limits.

TRISO burnup-dependent failure analysis in FHRs using BISON

In this study, we perform a study on TRIstructural ISOtropic (TRISO) Silicon Carbide (SiC) layer failure using the BISON fuel performance code. SiC thermomechanical behavior is analyzed during both steady-state and accident scenarios pertaining to a Fluoride-Salt-Cooled High-Temperature Reactor (FHR). A monodimensional single particle model is adapted from a previous analysis and run at 140 mW/particle for 500 days. The maximum burnup reached is 18.49% FIMA. A total of three accident scenarios are simulated at both fresh fuel conditions and following multiple irradiations at 2% FIMA increments up to 18.49% FIMA. The instances include an overcooling transient, a Control Rod Withdrawal (CRW) and a Loss of Forced Cooling Accident (LOFA). SiC hoop stress and failure probability predictions are evaluated for all simulated cases. A consistent tangential compressive stress state is found as a function of burnup for SiC. The maximum compressive hoop stress reaches approximately -400 MPa. The maximum SiC failure probability as a function of burnup is 2⋅10−13% at steady-state and no increases are predicted during transient simulations. It is concluded that the selected accident scenarios, under our modeling assumptions, are not predicted to pose a challenge to SiC integrity due to their relatively slow progression and low temperature increments, for both fresh fuel and at burnup values up to 18.49% FIMA. A 2D model is used, along with XFEM, to simulate a crack in the Inner Pyrolytic Carbon (IPyC). Conservative IPyC failure predictions are obtained with Weibull statistics approximately 38 days after beginning of irradiation. After crack formation, hoop stress concentrations are predicted in SiC using conservative assumptions, with failure probability increasing to 3.3⋅10−4%. 2D BISON simulations of the base irradiation are then carried out to the maximum burnup and transient conditions are subsequently applied. SiC maximum tensile hoop stress is predicted at the time of crack initiation, after which crack expansion reduces the predicted stress magnitude in the material. The results presented in the work are interpreted under the selected modeling assumptions and conclusions are based on a comparison with a parallel analysis on a High-Temperature Gas-Cooled Reactor design to evaluate the effects of power density and burnup on SiC performance.

A neutron tomography study to visualize fluoride salt (FLiNaK) intrusion in nuclear-grade graphite

Manufactured graphite is a preferred material for in-core components of molten salt reactors and fluoride salt-cooled high-temperature reactors, which are in permanent contact with liquid salts. However, owing to the porous nature of nuclear graphite, under certain conditions, molten salts may intrude graphite's pores and affect graphite's properties and functionality. Therefore, a better understanding of molten salt intrusion (distribution across sample cross section and penetration depth) is needed to assess its effects. In this work, we have demonstrated the use of neutron imaging (computed tomography) in the evaluation of salt penetration and distribution of a wide range of graphite grades with diverse microstructures that have been subjected to FLiNaK (LiF–NaF–KF) intrusion at 750 °C and 5 bar pressure for 12 h. Because of the great neutron attenuation contrast from scattering and adsorption between Li (from FLiNaK) and the graphite matrix, we have obtained direct visualization of FLiNaK salt distribution in the salt-impregnated graphites for the first time. Three-dimensional reconstructed images and cross-sectional concentration profiles demonstrate that salt penetration and density distribution are greatly dependent on the microstructural properties of the graphite grade.

Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms

Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. We present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.

Analysis of the Brayton cycle coupled with a small fluoride salt-cooled high-temperature reactor

Considering the environmental conditions and transportation conditions of remote areas, an inherently safe integrated energy conversion system featuring miniaturization, modularization, and high environmental adaptability is needed. The small fluoride salt-cooled high-temperature reactor (FHR) coupled with the Brayton cycle is a promising design. In this paper, the efficiency, exergy efficiency, and exergy loss of four different configurations of the supercritical carbon dioxide (S-CO2) Brayton cycle coupled with a new small fluoride salt-cooled high-temperature reactor are compared. The S-CO2 recompressor Brayton cycle has the best overall performance. Meanwhile, the effects of the cooling conditions on the thermal efficiency and exergy efficiency of different cycle configurations are discussed. When the core outlet temperature is 700°C, the efficiency of the designed S-CO2 recompressor Brayton cycle is approximately 42–44% when the cycle minimum temperature is 20–40°C. In conclusion, the designed small FHR coupled with the Brayton cycle system offers interesting performances in power generation, mineral mining, industrial steam supply, molten salt energy storage, and high-temperature hydrogen production in remote areas.

A Novel Core Design with Movable Moderator for a Fluoride Salt–Cooled High-Temperature Reactor

The Fluoride salt–cooled High-temperature Reactor (FHR) is a Generation IV reactor concept that can operate under near atmospheric pressure circumstances and further enhance inherent safety. In this study, an FHR core design with 165 MW of thermal output [MW(thermal)] is proposed. The reactor core employs tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks as the basic fuel form, FLiBe [lithium-beryllium fluoride (2 7LiF-BeF2)] as the primary coolant, and a three-batch fuel cycle scheme. Sensitivity analyses on various parameters were performed to optimize the cycle length and neutronic parameters. The fuel cycle of this core design was evaluated in detail from four aspects: cycle length, power peaking factor (PPF), discharge burnup, and temperature coefficient. It was found that a larger fuel channel pitch would have a relatively harder neutron spectrum and yield a relatively longer cycle length, lower PPF, and better fuel temperature coefficient and moderator temperature coefficient (MTC). In addition, burnable poison (BP) (Er2O3) can effectively reduce PPF, hold down the multiplication factor, and more importantly it can improve the MTC. The preliminary design of control blades is also presented in this paper. Furthermore, on the basis of the proposed 165-MW(thermal) core, we propose a novel core design that incorporates “fuel inside radial moderator (FIRM)” assemblies, movable moderator, and movable BP. This new design can extend the fuel cycle length by approximately 45 days for an 18-month fuel cycle. In addition, improvements were also found in PPF, discharge burnup, and temperature coefficients.

Scaled validation test for high prandtl number fluid mixed convection between parallel plates

The Kairos Power fluoride salt-cooled high temperature reactor (KP-FHR) passive decay heat removal system performance strongly depends on heat transfer characteristics in the reactor vessel’s down comer region. This region is an annular gap where the salt coolant flows downward between the core barrel and reactor vessel before entering the core region at the bottom of the vessel. Kairos Power’s principal safety analysis code, KP-SAM, extrapolates from existing heat transfer correlations for this geometry and applicable flow configurations. Main flow regimes of interest for safety analysis of the KP-FHR, when the system transitions from forced to natural circulation in the vessel upon loss of forced cooling, are the laminar and transition regimes. These regimes are not very well characterized in the literature, so in order to validate the inclusion of appropriate correlations in KP-SAM, scaled separate effects tests are conducted using a test facility that matches the Reynolds, Rayleigh and Prandtl numbers between the prototypical reactor coolant, Flibe, and a surrogate fluid, Therminol VP-1. The specific boundary conditions investigated are both symmetric heating and asymmetric heating. The data generated from this test serves to develop a validated heat transfer correlation for the laminar and transition regimes for implementation in KP-SAM, in support of safety analysis and licensing of the KP-FHR.

Infiltration of molten fluoride salts in graphite: Phenomenology and engineering considerations for reactor operations and waste disposal

The possibility and consequences of salt-infiltration in graphite must be evaluated for graphite used in molten salt reactors (MSRs) and fluoride-salt-cooled high-temperature reactors (FHRs), which can be subjected to salt pressures as high as 500 kPa. The volume of graphite porosity infiltrated by salt can be measured by direct infiltration and it can be predicted from the graphite pore size distribution, the surface tension of the salt, and the contact angle between the graphite and the salt. While these three properties are believed to be insensitive to irradiation, the former can be impacted by chronic or acute oxidation, and the latter two are highly sensitive to the chemistry of the salt and to events such as air ingress. For MSRs, predictions based on nominal properties of salt and graphite reveal that few graphite grades would satisfy the 4 vol% limit set in the Molten Salt Reactor Experiment, and even fewer would satisfy the 0.5 vol% design target. For FHRs, infiltration limits have not been defined and depend on the effect of infiltration on graphite properties, which are discussed. A hypothesis is presented for properties that may be impacted by infiltration and for which future studies are needed.

Development, Verification, and Validation of an Advanced Systems Code KP-SAM for Kairos Power Fluoride Salt–Cooled High-Temperature Reactor (KP-FHR)

To meet the Kairos Power (KP) Fluoride Salt-Cooled High-Temperature Reactor (FHR) (KP-FHR) development and commercialization schedules, the System Analysis Module (SAM), which is an advanced systems code for Generation IV liquid-cooled reactors developed at Argonne National Laboratory (ANL), has been selected as the basis for the development of the KP-FHR systems code KP-SAM. This allows for an accelerated joint development effort between the KP and ANL teams. This paper presents a general overview of the KP-SAM development process, its current status, completed verification, and ongoing validation efforts. KP-SAM development follows the U.S. Nuclear Regulatory Commission Evaluation Model Development and Assessment Process framework. SAM is a high-order fully implicit transient systems code written in C++. The SAM software design, major physical models, and Jacobian Free Newton Krylov–based numerical methods are briefly discussed. KP-SAM has matured enough to be used for the unvalidated demonstration safety analysis for the low-power KP-FHR test reactor (Hermes) as part of Preliminary Safety Analysis Report work. By following the guidance of an internal KP-FHR thermal fluid Phenomena Identification and Ranking Table report, some of the most important separate-effects-test validations were completed for the first iteration. A scaled integral-effects test is under detailed design and will be built in 2022 to provide key data to validate KP-SAM for licensing safety analysis.

Advanced Reactor Control and Operations (ARCO): A University Research Facility for Developing Optimized Digital Control Rooms

The Advanced Reactor Control and Operations (ARCO) facility was constructed in January 2018 to serve as a test bed for advanced reactor control rooms and operator support systems. Since then, it has supported human-machine interface user experience research, fault detection and mitigation technology development, control room concept of operations development, and remote operations research. ARCO serves as the control room for the Compact Integral Effects Test (CIET) facility, which replicates the primary-side flow paths and thermal-hydraulic behavior of a fluoride-salt-cooled high-temperature reactor (FHR) using simulant fluids and scaling principles. New reactor designs feature different operating conditions and scenarios than those in existing reactors. ARCO supports the research and development of digital tools for operator communications, intuitive real-time data analysis, online health monitoring and prognostics, and control room cybersecurity. By integrating these different technologies, ARCO acts as a prototypical control system to iteratively develop methods and tools of operation in advanced small modular nuclear reactors. This paper describes the features of and challenges to operating advanced small modular reactors underlying the design basis for ARCO and its operator support systems.

Numerical Investigations of Molten Salt Pump Journal Bearings Under Hydrodynamic Lubrication Conditions for FHRs

Fluoride salt-cooled, High-temperature Reactors (FHRs), featuring particle fuel, graphite moderator, and molten fluoride salt coolant, are used for electricity generation and process heat applications. The primary loop of an FHR is a closed loop that operates slightly above the atmospheric pressure with the fluoride salt temperature over 600°C. Reliable high-temperature molten salt pumps are critical to the successful deployment of FHRs. To stabilize rotating shafts and reduce the associated friction coefficients, well-designed bearings are required for molten salt pumps. Therefore, it is necessary to investigate the detailed hydrodynamic performance of bearings under high-temperature molten salt conditions. In this study, a computational fluid dynamics software package, i.e., STAR-CCM+, was used to predict the performance of fluoride salt–lubricated bearings. The numerical models were verified and validated respectively based on an analytical solution derived from the Reynolds equation and experimental data published in the literature. Good agreement was observed between the simulation results and the analytical solution and experimental data with a maximum relative discrepancy of less than 5%. The validated numerical model was then employed to predict the pressure distributions, applied static loads, and power losses of high-temperature fluoride salt–lubricated bearings with various Sommerfeld numbers. In addition, a parametric analysis was performed to investigate the influence of the axial and helical grooves of bearings on applied static load and power loss. It is found that under the same salt lubrication conditions, the bearings with helical grooves and axial grooves respectively yield 20% off and 14% off power loss compared with the bearing without grooves.

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.