High Temperature Reactor Design Research Articles

High uranium loading TRISO particle for microreactor applications

Tristructural isotropic (TRISO) fuel is currently being considered for novel high-temperature reactor (HTR) designs that deviate significantly from the prismatic or pebble bed modular high-temperature gas-cooled reactors (MHTGRs) that have been the focus of development efforts for the last four decades. These new designs include microreactors with compact cores that can benefit from increasing the core fissile density beyond that conventionally required in an MHTGR using TRISO fuel to enable longer fuel in-core residence time or higher power density. One means of increasing the fissile density in an HTR core is to adjust the coated particle design to increase the ratio of kernel volume to total particle volume. This would increase the volume of fissile phase in a fuel compact with no increase in particle packing fraction beyond that used in fuel compacts tested by the AGR program. Results show that the end-of-life failure probabilities of these high uranium loading particles and those of the validated reference particles are similarly low (∼2 × 10-5) for burnups of at least 9 % FIMA.

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.

Peridynamic modelling of cracking in TRISO particles for high temperature reactors

A linear-elastic computer simulation (model) for a single particle of TRISO fuel has been built using a bond-based peridynamic technique implemented in the finite element code ‘Abaqus’. The model is able to consider the elastic and thermal strains in each layer of the particle and to simulate potential fracture both within and between layers. The 2D cylindrical model makes use of a plane stress approximation perpendicular to the plane modelled. The choice of plane stress was made by comparison of 2D and 3D finite element models. During an idealised ramp to normal operating power for a kernel of 0.267 W and a bulk fuel temperature of 1305 K, cracks initiate in the buffer near to the kernel-buffer interface and propagate towards the buffer-iPyC coating interface, but do not penetrate the iPyC and containment of the fission products is maintained. In extreme accident conditions, at around 600% (1.60 W) power during a power ramp at 100% power (0.267 W) per second, cracks were predicted to form on the kernel side of the kernel-buffer interface, opposite existing cracks in the buffer. These were predicted to then only grow further with further increases in power. The SiC coating was predicted to subsequently fail at a power of 940% (2.51 W), with cracks formed rapidly at the iPyC-SiC interface and propagating in both directions. These would overcome the containment to fission gas release offered by the SiC ‘pressure vessel’. The extremely high power at which failure was predicted indicates the potential safety benefits of the proposed high temperature reactor design based on TRISO fuel.

Characterisation of neutron field in large graphite insertion in special core of the LR-0 reactor

Graphite is important reactor material used as a neutron moderator in various reactor designs. It is not only structural material used in previous reactors design but even material to which attention is focused due to the very high-temperature reactor design, which uses graphite as moderator and reflector. Due to the well-characterised cross-section of graphite, it is also a good candidate for neutron shaping experiments. In this kind of experiment, the neutron spectra are modified to the unconventional shape of neutron spectra which can be used to validate cross-section. This paper deals with validation of neutron field in graphite block surrounded by driver core arranged in large special core in LR-0 reactor. The experimentally determined shape of neutron spectra and spatial distribution of neutron flux is compared with the calculation. It is worth noting that the agreement for flux distribution is satisfactory in regions below the moderator level.

Anisotropic temperature-dependent elastic constants and thermal conductivities of TRISO particle coatings

Tristructural isotropic (TRISO) particles show great promise as a candidate fuel for use in several next-generation high-temperature nuclear reactor designs due to their structural integrity and fuel performance at high temperatures and burnups. Computational codes exist that can simulate TRISO fuel performance characteristics and failure probabilities under extreme conditions which require knowledge of the TRISO coatings’ thermophysical properties. The thermophysical descriptions of the TRISO particle's layers (i.e., buffer, pyrolytic carbon, and silicon carbide) currently used in fuel performance codes, however, assume that many of these properties are constant with respect to temperature or texture. In this paper, we obtain the full elastic stiffness tensors of the carbon and silicon carbide layers, which have transversely isotropic symmetry. The calculated elastic properties of each of the coating layers are in remarkable agreement with the current models at room temperature and correct orientations. Additionally, the textured 3C-SiC layer was found to exhibit novel auxetic behavior above 1500 °C. The anisotropic temperature-dependent thermal conductivities of the carbon layers were calculated using acoustical Grüneisen-Debye theory which are in excellent agreement with current models at room temperature and correct orientations. These texture- and temperature-dependent relationships can be incorporated into the thermophysical description of TRISO particles in order to more accurately model fuel performance and failure probabilities under extreme conditions in forthcoming high-fidelity computational simulations.

Comparison of neutronic aspects in high‐temperature gas‐cooled reactor using ZrC and SiC Triso particle with 50 and 100 MWt power

The use of zirconium carbide (ZrC) as a substitute for silicon carbide (SiC) has been proposed to improve the performance of coated particles for high-temperature gas-cooled reactor (HTGR) fuel. Irradiation test on ZrC has proved that it has excellent characteristics for HTGR fuel compared with SiC, such as fission product retention capabilities and better resistance on fission product corrosion. Neutronic analysis on 30 MWt HTGR using ZrC and SiC has been performed in many previous studies. The research presented in this paper was conducted as a further continuation of the aforementioned studies. The present study was directed to analyze tristructural isotropic (TRISO)-coated particles with ZrC as a substitute for the SiC layer on an HTGR with different power rates. Here, we used high-temperature test reactor design, an experimental scale prismatic HTGR built and operated in Japan. The power was varied at the range of 50-100 MWt, while the reactor geometry was modified by applying an additional fuel layer in the axial and radial directions. Neutronic analysis of the reactor was investigated by calculating the k-eff, k-inf, power peaking factor, burn-up level, neutron spectrum, and Doppler effect using ZrC and SiC layers for TRISO-coated fuel particles. The result shows the coated fuel particle using the ZrC layer has the same performance as SiC with an insignificant difference in the values on neutronic aspects calculation. Neutronic calculations are performed using the SRAC code with the Japanese Evaluated Nuclear Data Library 4.0 nuclear data library.

Thermochemical reduction modeling in a high-temperature moving-bed reactor for energy storage: 1D model

The design of robust and efficient high-temperature thermochemical reactors and determination of operating conditions are critical steps toward enabling high-efficiency long-duration solar energy storage. This work presents a computational model for the thermal reduction of a metal oxide material (Mg-Mn-O) up to 1450 °C and the coupled complex transport phenomena in a novel tubular thermal reactor design that features the capability for a high extent-of-reduction (high energy storage density) and inherent heat recuperation. A one-dimensional model coupling counter-current gas–solid flow, two-phase heat transfer, thermochemical redox reactions, and species transport in a moving-bed reactor is developed. Simplified versions of the model are validated with published results in the literature for packed beds with both inert and reactive particles; the fully coupled model is also validated with experimental measurements of a moving-bed reactor in terms of local temperatures and oxygen release at the exit. Detailed comparisons on the effects of different boundary conditions in the reaction zone (prescribed wall temperature vs. heat flux conditions) and formulations based on a simple uniform flow assumption vs. plug flow using Ergun equation for gas flow are investigated. The results are compared with experimental measurements, and for all cases, the energy flow components in the reactor system and the thermal to chemical conversion efficiency and overall system efficiency are computed. The predicted high thermal-to-chemical efficiency ∼95% and system efficiency ∼30% agree with experimental measurements.

Preliminary three-dimensional neutronic analysis of IFBA coated TRISO fuel particles in prismatic-core advanced high temperature reactor

In this study, preliminary neutronic calculations were performed in order to provide a basic design of a prismatic-core advanced high temperature reactor (PAHTR) that adapts a mixture of lithium fluoride and beryllium fluoride (Li2BeF4, known as FLiBe) as a coolant. This work investigates the effectiveness of utilizing an integral fuel burnable absorber (IFBA) as an additional coating layer for the outer surface of the fuel kernel of the tri-structural isotropic (TRISO) fuel particle. A parametric study was performed to attain a targeted cycle length of 60 months and a fuel burnup greater than 90 MWd/kgU for fuel unit cell, and 70 MWd/kgU and 45 MWd/kgU for 2D and 3D full reactor core, respectively, with a uranium enrichment of 19.75 w/o to support non-proliferation goals. This paper discusses the idea of applying homogeneous distribution of IFBA as an additional coating layer with different thicknesses through the reactor core and perform preliminary neutronics analysis on the reactivity control mechanism for both two- and three-dimensionals core. The Monte Carlo Serpent code has been utilized to estimate the core reactivity balance using the cycle length by reducing the burnup swing between the beginning of cycle (BOC) and end of cycle (EOC). This can be done by adopting the innovative idea of covering the outer surface of the fuel kernel with a thin coating layer of IFBA. This technique is utilized in the 3D evaluation of the PAHTR reactor. Additionally, the neutron spectrum, peaking factor (axial and radial), and multiplication factor for the 3D PAHTR reactor core at the BOC, middle of the cycle (MOC), and EOC were compared by taking into account the double heterogeneity nature of the TRISO fuel particles embedded in the graphite matrix.

Continuous Fly-Through High-Temperature Synthesis of Nanocatalysts.

The conventional thermal treatment systems typically feature low ramping/cooling rates, which lead to steep thermal gradients that generate inefficient, nonuniform reaction conditions and result in nanoparticle aggregation. Herein, we demonstrate a continuous fly-through material synthesis approach using a novel high-temperature reactor design based on the emerging thermal-shock technology. By facing two sheets of carbon paper with a small distance apart (1-3 mm), uniform and ultrahigh temperatures can be reached up to 3200 K within 50 ms by simply applying a voltage of 15 V. The raw materials can be continuously fed through the device, allowing the final products to be rapidly collected. As a proof-of-concept demonstration, we synthesized Pt nanocatalysts (∼4 nm) anchored on carbon black via this reactor at ∼1400 K. Furthermore, we find it features excellent electrocatalytic activities toward methanol oxidation reaction. This work offers a highly efficient platform for nanomaterials synthesis at high temperatures.

Low-Cost Radiant Heater for Rapid Response, High-Temperature Heating

High-temperature processing has an irreplaceable role in many research and industrial applications. Despite remarkable development spanning over a century, the pursuit of even higher thermal flux density and more rapid thermal transients has not slowed down. As part of the ongoing energy evolution, many industrial applications are transitioning from direct combustion of fossil fuels as primary energy sources to increasing electrification, capable of adapting to renewable power grids. Thus, there is an emerging need for electrical heaters that can replace burners and supply the heat demand, especially at the highest temperatures. In this study, we report on a radiant heater design that can achieve cyclic heating/cooling rates of up to 400 K min–1 and a temperature range in excess of 1,800 K, comparable to those of commercial infrared gold image furnaces, at high surface and volumetric power densities. The heater consists of a modular unit of incandescent tungsten filament and is enclosed in an evacuated ceramic envelope, chemically inert, tolerant of thermal shock, and impervious to gasses. The material and manufacture cost of such heaters, which is estimated at ∼$0.05/W, is less than 0.03% of that for infrared gold image furnaces, which is at >$2/W. Tests of more than 10,000 demanding cycles (high temperature and high heating/cooling rate) over 350 h of total operational time and in different temperature ranges confirm the robust performance of radiant heater prototypes. The design is widely applicable to high-temperature reactor and furnace designs. In thermochemistry research and practice, these radiant heaters could offer multiple benefits compared to solar simulators, lasers, infrared gold furnaces, ceramic heaters, or direct concentration of solar input.

Part 3: Evaluating a small modular high temperature reactor design during control rod withdrawal and a depressurised loss of coolant accidents

With safety, economics and financial risk playing a large role within the deployment of small modular reactors, it is vital that passive safety features are implemented and the maximum operational temperatures within the design are well understood. This article within the series evaluates the designed 10 MWth U-Battery concept's safety features with respect to a control rod withdrawal and a depressurised loss of coolant accident. A high temperature reactor's passive safety features depend on the Tristructural-isotropic (TRISO) fuels mechanical and thermal integrity and through experimental data this is shown to remain intact up until ~1600 °C (1873 K). This article provides the methodology of how to perform both transients, with the control rod withdrawal being undertaken using point kinetics and the depressurised loss of coolant accident with thermal feedback for both transients being provided via a 1D heat transfer model. Overall, this work has shown that during the control rod withdrawal the fuel temperature rises by 110 K and at this point the excess reactivity is compensated by the negative temperature coefficient of the fuel. During the depressurised loss of coolant accident, the maximum fuel temperature reached 1455 K after 60 h. This concludes that during both transients the temperatures maintained well below the maximum fuel operating temperature.

Coupled Computational Fluid Dynamics–Discrete Element Method Study of Bypass Flows in a Pebble Bed Reactor

In the high-temperature reactor design, it is common practice to leave gaps between the graphite blocks of the reflectors to accommodate thermal dilatation and material swelling, as well as to provide an additional cooling source during operation. These gaps give rise to bypass flows entering the reactor core. The bypass flows can change friction factors and heat exchange coefficients obtained in the bulk of the pebble bed. In this paper, a coupled computational fluid dynamics–discrete element method (CFD-DEM) model is proposed. In this model, the pebbles are resolved by the CFD grid and the turbulent field is partially captured using a detached eddy simulation method. The DEM model is first validated against empirical correlations for the packing of the pebbles, and the coupled model is then tested against thermal measurements in the SANA experiment. Then the model is used to perform three-dimensional studies of the effects of the bypass flows in a representative pebble bed configuration. It is determined that the effect of cross flows can be approximately bounded to the first two layers of pebbles next to the reflector wall. Additionally, an increase of ~12% in the Nusselt number in the pebbles next to the reflector is predicted, with a maximum local increase in the pebble of ~100%.

Development of a structural integrity evaluation program for elevated temperature service according to ASME code

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

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.

Preferred core conceptual design of pebble bed advanced high temperature reactor

Key parameters of Pebble Bed Advanced High Temperature Reactor (PB-AHTR) core are studied, and then the preferred core conceptual design is proposed. Both evacuating molten salt and injecting poison into coolant can be used as the auxiliary system of the second shutdown system in PB-AHTR. Compared with injecting poison into coolant, the effect of evacuating molten salt on the core is smaller and more beneficial to the engineering realization. Compared to one-time loading scheme, batch loading scheme can ensure that a small core excess reactivity in whole life-time which is easier to be controlled, but more complex to operation. In one-time loading scheme, the second shutdown system gains enough fast shutdown margins by increasing the number of second control rod, instead of reducing the stack height of core activity. The shutdown margins of both the first shutdown system and the second shutdown system meet the design requirements.

Determining the acute oxidation behavior of several nuclear graphite grades

The acute oxidation behaviors for a number of nuclear graphite grades pertinent to gas-cooled high-temperature reactor (HTR) designs are compared using established American Society for Testing and Materials oxidation testing standards. In the unlikely event of an air-ingress accident, graphite components within the HTR core region are anticipated to oxidize if oxygen is allowed to enter the hot core region. Specimens from six disparate nuclear graphite grades were exposed to air/helium mixtures (100% air and 10% air-in-helium) within and beyond the kinetic oxidation temperature regime to ascertain the oxidation performance for each grade. All specimens were oxidized to a prescribed mass loss within a constant total gas flow of 10 ℓ/min at multiple isothermal temperatures for calculation of oxidation rate and Arrhenius parameters for each graphite grade. A comparison of the oxidation performance among the different grades, as well as the effects resulting from high impurity levels, are discussed.

Inelastic Material Models of Type 316H for Elevated Temperature Design of Advanced High Temperature Reactors

In this paper, the inelastic material models for Type 316H stainless steel, which is one of the principal candidate materials for elevated temperature design of the advanced high temperature reactors (HTRs) pressure retained components, are investigated and the required material parameters are identified to be used for both elasto-plastic models and unified viscoplastic models. In the constitutive equations of the inelastic material models, the kinematic hardening behavior is expressed with the Chaboche model with three backstresses, and the isotropic hardening behavior is expressed by the Voce model. The required number of material parameters is minimized to be ten in total. For the unified viscoplastic model, which can express both the time-independent plastic behavior and the time-dependent viscous behavior, the constitutive equations have the same kinematic and isotropic hardening parameters of the elasto-plastic material model with two additional viscous parameters. To identify the material parameters required for these constitutive equations, various uniaxial tests were carried out at isothermal conditions at room temperature and an elevated temperature range of 425–650 °C. The identified inelastic material parameters were validated through the comparison between tests and calculations.

Development of creep elastic follow-up factor solutions for piping design of high-temperature reactor under combined primary and secondary loadings

In this study, creep elastic follow-up (EFU) factor solutions are developed for the piping design of high-temperature reactors. A straight pipe under combined primary and secondary loadings is considered for high-temperature piping. Axial tension, internal pressure, and bending moment are applied to the straight pipe as a primary loading condition. The thermal stress caused by the temperature gradient through the thickness is considered as a secondary loading. Applying the similarity between the EFU factor curve and primary-secondary creep strain curve, an exponential form is employed to obtain a scheme of the EFU factor solutions. The ratio of the primary stress to the total stress, and the creep exponent are selected as the variables that affect the EFU. Elastic-creep finite element (FE) analyses are performed, and the R5 Option 3 method is used for estimating the EFU factors. Based on the FE results, the creep EFU factor solutions are proposed by providing the values of the parameters in the scheme. The proposed creep EFU factor solutions are validated by comparing the stress relaxation curves predicted by the present EFU solutions with those predicted by the FE results.

Conceptual design study of a high performance commercial HTGR for early introduction

Conceptual design study of a commercial High Temperature Gas-cooled Reactor (HTGR) for early introduction has been performed based on the cumulated experience in design, construction, and operation of the High Temperature engineering Test Reactor (HTTR) and design of the commercial Gas Turbine High Temperature Reactor 300 (GTHTR300). The power output is 165MWt and the inlet and outlet coolant temperatures are 325 °C and 750 °C, respectively, to provide steam for industrial utilization. However, given a requirement for the reactor pressure vessel to be smaller even that of the 30 MWt HTTR, several challenging technical problems have to be dealt with to arrive in a high performance core design that provides extended fuel burnup, prolonged refueling period, improved fuel refueling scheme, improved fuel element and so on from the HTTR.

Numerical Simulation of Isothermal Flow Across Slant Five-Tube Bundle with Spectral Element Method Code Nek5000

The helical coil steam generator (HCSG) is a specific type of shell-and-tube heat exchanger known for having a higher heat transfer coefficient than many other designs. For this reason, they are considered in small modular reactor and high temperature reactor designs. Investigation of flow behavior in HCSGs is important for better design. In this paper we present our study for modeling the coolant flow in the primary side of the HCSG. We used Nek5000, an open source, high-order spectral element computational fluid dynamics code developed in Argonne National Laboratory. Nek5000 accepts only hexahedral mesh, which makes the meshing process for the complicated HCSG geometry very challenging. A tetrahedral-to-hexahedral meshing strategy was applied to bypass the geometric complexities. In this study large eddy simulation (LES) was performed at the Reynolds number of 9000 based on the inlet velocity and the tube diameter. The employed subgrid-scale model for LES relies on explicit filtering. First- and second-order statistics were compared to available experimental data. Overall velocity and turbulent kinetic energy showed good agreement with particle image velocimetry data.