High-Temperature Gas-Cooled Reactor Fuel Research Articles

Deterministic Transport Calculation Method for Statistical Geometry with Small Fuel Particles

A deterministic transport calculation method is proposed for the treatment of dispersed fuel particles in a fuel compact/fuel pebble of a typical high-temperature gas-cooled reactor fuel. The random distribution of fuel particles was considered using the statistical geometry (STG) method, which is widely used in the Monte Carlo method. A long-ray trace, which represents a neutron flight path, was considered, and the segment lengths and material distributions on the ray trace were randomly sampled using STG. Then a conventional transport sweep, as used in the method of characteristics, was performed along the ray trace. The proposed deterministic statistical geometry (DSTG) method can calculate the flux spatial distribution in a heterogeneous geometry containing randomly dispersed fuel particles and the surrounding graphite matrix, which is consistent with the STG in a Monte Carlo method. The validity of the DSTG method was confirmed through sensitivity calculations and comparisons with a multigroup Monte Carlo method that utilizes STG. The proposed method can be used for the homogenization of heterogeneous structures inside a fuel compact or fuel pebble as an alternative to conventional deterministic unit cell calculations that consider fuel particles and the surrounding matrix in high-temperature gas-cooled reactor fuels.

Development of the Assembly-Level Monte Carlo Neutron Transport Code M3C for Reactor Physics Calculations

The general geometry continuous-energy Monte Carlo code M3C is currently under development at the Bhabha Atomic Research Centre for reactor physics calculations. The development of the Monte Carlo code M3C for reactor design entails the use of continuous-energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. This paper describes the current status of the development of the code. The performance and accuracy of the code in application to a variety of problems have been investigated. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low-energy treatment, probability table treatment in unresolved resonance range, and capability of handling the microscopic fuel particles (TRISO) dispersed randomly, which is very useful in modeling high temperature gas-cooled reactor fuels. Apart from all of the important features in any Monte Carlo code available worldwide, the M3C code has an advanced capability to handle the geometry, which is not described by mathematical equations but only represented by the geometrical points. The code has been validated for its accuracy against a large number of sample problems covering a wide range from simple (like spherical) to complex geometry (like pressurized heavy water reactor lattice) and including randomly dispersed TRISO fuel particle systems. The code is presently restricted to assembly-level calculations.

The electrochemical deconsolidation mechanism of graphite matrix in HTGR spherical fuel elements

In High-Temperature Gas-cooled Reactor (HTGR)s, coated fuel particles are dispersed in a graphite matrix, which forms both compact and spherical fuel elements. As for the post irradiation examination (PIE) of HTGR irradiated fuels, the electrochemical deconsolidation method in which graphite matrix is disintegrated into graphite powder is one of the feasible processing methods. The graphite matrix is composed of natural graphite, artificial graphite and glassy carbon from carbonized resin, which have different effects on electrochemical deconsolidation. In this work, synthesized graphite matrix specimens were produced by intentionally varying the resin binder content in the graphite matrix raw materials from 5% to 50%, in order to investigate the electrochemical deconsolidation mechanism of graphite matrix in HTGR fuel elements. The results showed that increasing the glassy carbon content increased the degree of chemical oxidation, with more pronounced production of graphite oxide. The driving force for graphite matrix deconsolidation is proposed to be a synergetic effect of the expansion stress around closed- and through-hole microstructure originating from oxidatively destroyed graphite layers. This work will not only help to more thoroughly understand the electrochemical deconsolidation procedure of HTGR fuels, but also contribute to the development of a feasible method to manage radioactive graphite waste.

The development of HTGR-TRISO coated fuels in the globe: challenging of Indonesia to be an HTGR fuel producer

A first high quality coating TRISO high temperature gas-cooled reactor (HTGR) fuel has been developed since almost six decades ago. After Germany had invented the highly qualified HTR fuel in 2008, China contributed on the fabrication process of tri-structural isotropic (TRISO) UO2 coated fuel particle. Meanwhile, the University of South Carolina (USA) had also developed the design as well as fabrication of an advanced TRISO fuel with ZrC coating, the INET-China conveyed research on the over-coating process in the manufacture of spherical fuel elements for HTGR and invented the friendly-operated equipment to manufacture HTGR fuel mass in 2014. Since HTGR becomes a trend of nuclear technology in the globe, Indonesia is currently conniving a 10 MW experimental power reactor (RDE) which is HTGR type and the program of the RDE was firstly introduced to the Agency for National Development Planning (BAPPENAS) at the beginning of 2014 and the RDE is expected to be operable in 2022/2023. The RDE program is projected to have positive impacts on community prosperity, self-reliance and sovereignty of Indonesia. Since the country has a lot of small islands especially in the eastern part of the country, the small medium HTGR reactors with the power of 5 MWe to 50 MWe will be very suitable to be built in those areas. Among other GEN IV reactor types, HTGR is one of Gen-IV nuclear reactors to be mostly available by 2050 even till 2100. There are few advanced countries to deal with HTGR reactors, so that Indonesia has a very good opportunity to become one of the HTGR reactor producers and even further an HTGR fuel producer in the globe or in ASEAN countries. This paper pronounces the current status of R&Ds on HTGR fuels in the globe as well as that of nuclear technology perspectives till 2100 and policy implications for the global nuclear power industry. Moreover, the paper also showed some countries emerging their new NPPs by 2050. The country through BATAN has showed the long-term plan to treat all low-level radioactive waste resulted both from industries and from the future NPPs. The national capabilities supported by the strategic state-owned enterprises to support nuclear-energy-based industry are also explained. The participation of local industries to reinforce the nuclear-energy-based industry in the country proved more than 40%. Indeed, to become the HTGR fuel producer by 2030 seems much clearer for Indonesia taking into account the strong, legal sustenance from the government.

The Strategy to Support HTGR fuels for The 10 MW Indonesia's Experimental Power Reactor (RDE)

The Indonesia’s 10 MW experimental power reactor (RDE) is developed based on high temperature gas-cooled reactor (HTGR) and the program of the RDE was firstly introduced to the Agency for National Development Planning (BAPPENAS) at the beginning of 2014. The RDE program is expected to have positive impacts on community prosperity, self-reliance and sovereignty of Indonesia. The availability of RDE will be able to accelerate advanced nuclear power technology development and hence elevate Indonesia to be the nuclear champion in the ASEAN region. The RDE is expected to be operable in 2022/2023. In terms of fuel supply for the reactor, the first batch of RDE fuel will be inclusive in the RDE engineering, procurement and construction (RDE-EPC) contract for the assurance of the RDE reactor operation from 2023 to 2027. Consideration of RDE fuel plant construction is important as RDE can be the basis for the development of reactors of similar type with small-medium power (25 MWe–200/300 MWe), which are preferable for eastern part of Indonesia. To study the feasibility of the construction of RDE fuel plant, current state of the art of the R&D on HTGR fuel in some advanced countries such as European countries, the United States, South Africa and Japan will be discussed and overviewed to draw a conclusion about the prospective countries for supporting the fuel for long-term RDE operation. The strategy and roadmap for the preparation of the RDE fuel plant construction with the involvement of national stakeholders have been developed. The best possible vendor country to support HTGR fuel for long-term operation is finally accomplished. In the end, this paper can be assigned as a reference for the planning and construction of HTGR RDE fuel fabrication plant in Indonesia.Keywords: RDE, Indonesia, HTGR, fuel, strategy.

Fully ceramic microencapsulated fuel in prismatic high temperature gas-cooled reactors: Analysis of reactor performance and safety characteristics

Advanced nuclear reactor technologies have the potential to expand the missions of nuclear energy while reducing carbon emissions. This paper presents scoping reactor physics and thermal hydraulics analysis of a high temperature gas-cooled reactor (HTGR) using the fully ceramic microencapsulated (FCM) fuel form, and demonstrates the feasibility of FCM fueled HTGRs. FCM fuel consists of tristructural isotropic (TRISO) coated fuel particles embedded in a matrix of silicon carbide (SiC). The potential advantages of FCM fuel, which uses a monolithic SiC matrix, over conventional HTGR fuels with a carbon-based matrix include: a long refueling interval; high stability of the SiC matrix under irradiation with limited swelling; high fission product retention of the fuel form, with the SiC matrix acting as an additional barrier to fission product release; and enhanced oxidation resistance during normal operation and air ingress accidents. In addition, the literature shows that the effective thermal conductivity of SiC fuel compacts and conventional HTGR compacts are expected to be similar.The key finding of this study is that FCM fuel, within the form factor of a typical General Atomics prismatic graphite block, exhibits similar fuel cycle performance to conventional HTGR fuel. The reactor cycle length, discharge burnup, and natural resource utilization are similar. However, the reduced moderation in the FCM designs considered here does marginally reduce the discharge burnup, and therefore natural resource utilization, versus the reference HTGR design.The hardened neutron flux spectrum resulting from the SiC matrix, which displaces carbon from the core, requires a slightly higher packing fraction of conventional uranium oxy-carbide (UCO) fuel kernels or the use of higher density uranium mononitride (UN)-based fuel kernels. These options will marginally increase the decay power, because they harden the neutron flux energy spectrum and increase the density of 238U in the fuel. In one case considered, this will increase the absorption of neutrons in 238U, and the resultant impact of 239Np isotope on the decay power. The Doppler coefficients normalized per total fuel heat capacity are weaker in the FCM-fueled designs than in the reference HTGR design. This impacts the energy deposition in a control rod ejection accident, and hence the design of potential transient tests of these fuel forms. In addition, analyses of loss-of-forced cooling accidents indicate that the fuel temperature during these design basis accidents are up to ∼30 °C higher with FCM fuel than with conventional HTGR fuels due to the increased decay power.

3D Nondestructive Visualization and Evaluation of TRISO Particles Distribution in HTGR Fuel Pebbles Using Cone-Beam Computed Tomography

A nonuniform distribution of tristructural isotropic (TRISO) particles within a high-temperature gas-cooled reactor (HTGR) pebble may lead to excessive thermal gradients and nonuniform thermal expansion during operation. If the particles are closely clustered, local hotspots may form, leading to excessive stresses on particle layers and an increased probability of particle failure. Although X-ray digital radiography (DR) is currently used to evaluate the TRISO distributions in pebbles, X-ray DR projection images are two-dimensional in nature, which would potentially miss some details for 3D evaluation. This paper proposes a method of 3D visualization and evaluation of the TRISO distribution in HTGR pebbles using cone-beam computed tomography (CBCT): first, a pebble is scanned on our high-resolution CBCT, and 2D cross-sectional images are reconstructed; secondly, all cross-sectional images are restructured to form the 3D model of the pebble; then, volume rendering is applied to segment and display the TRISO particles in 3D for visualization and distribution evaluation. For method validation, several pebbles were scanned and the 3D distributions of the TRISO particles within the pebbles were produced. Experiment results show that the proposed method provides more 3D than DR, which will facilitate pebble fabrication research and production quality control.

Basis for Fluoride Salt–Cooled High-Temperature Reactors with Nuclear Air-Brayton Combined Cycles and Firebrick Resistance-Heated Energy Storage

The fluoride salt–cooled high-temperature reactor (FHR) with a nuclear air-Brayton combined cycle (NACC) and firebrick resistance-heated energy storage (FIRES) is a new reactor and plant concept. The development of a new reactor is a large undertaking that requires a strong commercial case and a strong basis for government support to enable commercialization. The goals are (1) increase plant revenue by 50% to 100% relative to base-load nuclear plants with capital costs similar to light water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, (3) no potential for major fuel failure and thus no potential for major radionuclide off-site releases in a beyond-design-basis accident, and (4) offer a pathway to more advanced energy systems. The approaches to achieve those goals are described herein.The FHR uses liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel is dissolved in the coolant. However, in the FHR the fuel is not dissolved in the coolant. Instead, the FHR uses graphite-based high-temperature gas-cooled reactor (HTGR) coated-particle fuel. This combination enables delivering heat to the power cycle between 600°C and 700°C that, in turn, enables the FHR to couple to NACC. Using an air-Brayton power cycle enables the FHR to operate as a base-load reactor and produce added electricity in a peaking mode with the addition of auxiliary heat (natural gas, stored heat, or hydrogen). The auxiliary heat-to-electricity production is a thermodynamic topping cycle with efficiencies of 66%. Because an FHR with NACC is more efficient in converting natural gas into peak electricity than a stand-alone natural gas plant (60%), it can economically compete with a natural gas plant for peak electricity production because it uses less fuel. NACC can also incorporate FIRES heat storage that enables buying electricity at low prices to later sell electricity at high prices. For every 100 MW(electric) of base-load capacity, the station output can vary from minus several hundred megawatts to +242 MW(electric) because of the capabilities of the NACC with FIRES. The FHR can be built in different sizes.The use of high-temperature liquid-salt cooling and coated-particle fuel enables near-term reactor designs where large-scale fuel failure cannot occur, and thus, large-scale off-site releases of radionuclides cannot occur. Under extreme accident conditions decay heat can be passively conducted to the environment at temperatures below fuel failure temperatures and thus avoid the potential for large-scale radionuclide releases.The FHR is the gateway technology to several advanced salt-cooled reactor systems that have additional capabilities. The MSR can be designed to enable breeding and actinide waste burning. New high-magnetic-field fusion systems may require liquid-salt cooling. The FHR would provide the required salt and power cycle technology for these advanced reactors. There are significant development challenges. The United States has a competitive advantage in developing the FHR because it leads in gas turbine technology, high-temperature materials, and HTGR fuels.

SiC/SiO<sub>2</sub> Coating on Matrix Graphite Spheres of HTR Fuel Element Produced by a Two-Step Pack Cementation/High-Temperature Oxidation Process

Graphite is one of the most important material for the reactor core and fuel elements of high temperature gas-cooled reactor (HTR). Improving the oxidation resistance of graphite is very essential for the research of new fuel elements and the development of HTR. In this study, a gradient SiC layer of 500~700 μm was prepared on matrix graphite spheres by a two-step pack cementation, and the outer SiO2 layer prepared by the high-temperature oxidation process. The phases, microstructure, bonding strength and oxidation resistance of SiC/SiO2 coated matrix graphite spheres were investigated. The SiC/SiO2 coated matrix graphite spheres were carried on rapid thermal shocking tests from 1773 K to room temperature for 50 times without any cracks. The SiC/SiO2 coated matrix graphite spheres exhibits excellent anti-oxidation properties. No obvious weight loss was found after isothermal oxidation in air at 1273 K for 50 h and the weight gain was less than 1% at 1773 K in air for 50 h due to the oxidation of SiC layer.

Corrigendum to “Carbon Contained Ammonium Diuranate Gel Particles Preparation in Mid-Process of High-Temperature Gas-Cooled Reactor Fuel Fabrication” [Nucl. Eng. Technol. 48 (2016) 175–181

Corrigendum to “Carbon Contained Ammonium Diuranate Gel Particles Preparation in Mid-Process of High-Temperature Gas-Cooled Reactor Fuel Fabrication” [Nucl. Eng. Technol. 48 (2016) 175–181

Carbon Contained Ammonium Diuranate Gel Particles Preparation in Mid-process of High-temperature Gas-cooled Reactor Fuel Fabrication

Abstract This study investigates the dispersibility of carbon in carbon contained ammonium diuranate (C-ADU) gel particles and the characteristics of C-ADU gel liquid droplets produced by the vibrating nozzle and integrated aging–washing–drying equipment. It was noted that the excellent stability of carbon dispersion was only observed in the C-ADU gel particle that contained carbon black named CB 10. ADU gel liquid droplets containing carbon particles with the excellent sphericity of approximately 1,950 μm were then obtained using an 80–100-Hz vibrating nozzle system. Dried C-ADU gel particles obtained by the aging–washing–drying equipment were thermal decomposed until 500°C at a rate of 1°C/min in an air or in 4% H2 gas atmosphere. The thermally decomposed C-ADU gel particles showed 24% weight loss and a more complicated profile than that of ADU gel particles.

Effect of pebble packing on neutron spectrum and the isotopic composition of HTGR fuel

Fission products play an important role in the safety and fuel integrity of high-temperature gas-cooled reactor (HTGR) and they depend on temperature, burnup, neutron energy distribution, and fast fluence. Energy distribution of neutrons in a fuel region determines the isotopic distribution of the fission products to be produced. The local concentrations of these isotopes are considered to be functions of temperature and burnup as well as the amount transported from the kernel to the coating layers where they interact and may degrade layers. Thus, the integrity of the fuel particle may be lost and fission products can be released into the reactor coolant inventory. In this study, it is the main purpose to perform neutron energy spectrum shift in spherical HTGR fuels and to investigate its effect on fission products. Moreover, it is also intended to analyze the effect of unit cell geometries on criticality of the system.The calculations for group fluxes based on ENDF4 library with 27 neutron energy groups are accomplished by the MCNP5 neutron transport code. Burnup and criticality analyses are performed by using the MONTEBURNS2 code (MCNP5 coupled with ORIGENS). To simplify the neutron transport problem, instead of full core modeling, two fundamental unit cell arrangements, body-centered cubic (BCC) and hexagonal close-packed (HCP) lattices, are considered to be as reference geometry models. Unit cells are defined with proper boundary conditions and random packing for TRISO particles provided by the stochastic geometry card specified in MCNP5 for HTGR pebbles is used.

Suggested Failure Probability Calculation Method for Japanese High-Quality High-Temperature Gas-Cooled Reactor (HTGR) Fuel

Suggested Failure Probability Calculation Method for Japanese High-Quality High-Temperature Gas-Cooled Reactor (HTGR) Fuel

111 新型高温ガス冷却原子炉用燃料用ZrC被覆層のTEM・STEM観察(III)(超塑性と材料および造形法,オーガナイズドセッション)

111 新型高温ガス冷却原子炉用燃料用ZrC被覆層のTEM・STEM観察(III)(超塑性と材料および造形法,オーガナイズドセッション)

TEM/STEM Observation of ZrC‐Coating Layer for Advanced High‐Temperature Gas‐Cooled Reactor Fuel

The Japan Atomic Energy Agency (JAEA) has started to study and develop ZrC‐coated fuel particles for advanced high‐temperature gas‐cooled reactors. The ZrC‐coating layer was fabricated using the bromide process at JAEA. In the early stage of the project, however, the deposition temperature was varied. This paper mainly focuses on the microstructures of the ZrC‐coating layer developed in the early stage of the project. Some circumferential stripes were observed in the ZrC‐coating layer on optical micrographs. It was found that the stripes were caused by the nonuniform distribution of the free carbon phase. It was also revealed by means of transmission electron microscope /scanning transmission electron microscope observations that crystal grains of the ZrC were small and columner in shape, and were not equiaxed especially near the surface. It appears that the oscillated deposition temperature results in the nonuniform distribution of the free carbon region. The structure of the free carbon region formed in the ZrC‐coating layer appeared to be such that the c‐plane was roughly parallel to its lengthened direction. The ZrC‐coating layer appeared to be bound to the PyC layer. Fibrous carbon existing at the PyC/ZrC boundary was also observed.

107 革新的高温ガス炉燃料用ZrC被覆層のTEM・STEM観察(超塑性と材料および造形法)

107 革新的高温ガス炉燃料用ZrC被覆層のTEM・STEM観察(超塑性と材料および造形法)

405.高温ガス炉燃料の再処理技術

A High Temperature Gas-Cooled Reactor (HTGR) is particularly attractive due to capability of producing high temperature helium gas and its inherent safety characteristic. Research and development of high temperature gas turbine plant and high temperature heat utilizing technology are now undergoing. The High Temperature Engineering Test Reactor (HTTR) is a research facility constructed by the Japan Atomic Energy Research Institute (JAERI). All rise-topower tests have been successfully carried out and the performance of the HTTR has been evaluated. Now, preparation for the operation with outlet coolant temperature of 950°C and safety demonstration tests are undergoing.This paper describes reprocessing technology of HTGR fuels. Coated fuel particles, consisted of a microsphere of low enriched UO2 with TRISO particles, are used as the HTGR fuels. In order to reprocess HTGR fuels, a head-end process is needed and JAERI had confirmed jet-grind method as basic technologies of the head-end process. Since Purex method can be used after the head-end process, a reprocessing system for the HTGR fuels could be established. Also the preliminary study on the methodology for disposing graphite blocks in a HTGR was carried out, and its evaluation results were briefly presented.

An Investigation of Irradiation Performance of High Burnup HTGR Fuel

In order to investigate fuel behavior under high burnup irradiation condition of high temperature gas-cooled reactor (HTGR), an irradiation test was performed. An irradiation was carried out as a part of a cooperative effort between the US DOE and the Japan Atomic Energy Research Institute. The fuel for the irradiation test was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR). In order to keep fuel integrity up to high burnup over 5%FIMA (% fission per initial metallic atom), thickness of buffer and SiC layers of fuel particle were increased. This report describes the fuel behavior under high burnup condition in the irradiation test.

An Investigation of Irradiation Performance of High Burnup HTGR Fuel.

In order to investigate fuel behavior under high burnup irradiation condition of high temperature gas-cooled reactor (HTGR), an irradiation test was performed. An irradiation was carried out as a part of a cooperative effort between the US DOE and the Japan Atomic Energy Research Institute. The fuel for the irradiation test was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR). In order to keep fuel integrity up to high burnup over 5%FIMA (% fission per initial metallic atom), thickness of buffer and SiC layers of fuel particle were increased. This report describes the fuel behavior under high burnup condition in the irradiation test.

Irradiation experiments on high temperature gas-cooled reactor fuels and graphites at the high flux reactor petten

Because of its favourable design and operational characteristics and the availability of dedicated experimental equipment the High Flux Reactor at Petten has been extensively used as a test bed for HTR fuel and graphite irradiations for more than 20 years. Earlier fuel testing programmes contributed to the development of the coated fuel particle concept by extended screening tests. Now these programmes concentrate on performance testing of reference coated fuel particles and reference fuel elements for the German HTR-Module, the HTR-500 and to a lesser extent for the US HTGR concepts. It is shown with representative examples that these fuels have excellent fission product retention capabilities under normal and anticipated off-normal operating conditions. Extended irradiation programmes in the HFR Petten have significantly contributed to the database for the design of HTR graphite structures. The programmes not only comprise radiation damage accumulation in the temperature range from 570 to 1570 K up to very high fast neutron fluences and its influence on technological properties, but also irradiations under specified load conditions to investigate the irradiation creep behaviour of various graphites in the temperature range 570 to 1170 K.