Development of high temperature gas-cooled reactor (HTGR) fuel in Japan

This paper describes experiences and present status of research and development works for the high temperature gas-cooled reactor (HTGR) fuel in Japan. Recently, Very High Temperature Reactor (VHTR) is evaluated highly worldwide, and is a principal candidate for the Generation IV reactor systems. In Japan, HTGR fuel fabrication technologies have been developed through the High Temperature Engineering Test Reactor (HTTR) project in Japan Atomic Energy Agency since 1960’s. In total about 2 tons of uranium of the HTTR fuel has been fabricated successfully and its excellent quality has been confirmed through the long-term high temperature operation. Based on the HTTR fuel technologies, SiC TRISO fuel has been newly developed for burnup extension targeted VHTR. For ZrC-TRISO coated fuel as an advanced fuel designs, R&Ds for fabrication and inspection have been carried out in JAEA. The irradiation with the Japanese uniform stoichiometric ZrC coating has been completed in the cooperation with Oak Ridge National Laboratory of the United States.     

Irradiation performance of HTGR fuel rods with diluted thermosetting matrices

Satisfactory irradiation performance of experimental thermosetting HTGR fuel rods whose injected matrices were properly diluted with a low-char-yield additive (fugitive) was demonstrated for a fast-neutron fluence of 5 × 10²¹ n/cm² (E > 0.18 MeV) at a temperature of 1200°C. The addition of the fugitive introduced microporosity throughout the fired matrix; this reduced internal matrix shrinkage during irradiation by limiting the amount of binder char present, and it also reduced the strength and bondability of the matrix. Sufficient fugitive had to be added to reduce the percentage of binder char (PBC) in the graphite-filled matrix to less than 32 wt.% in order to prevent pyrocarbon coatings on close-packed fuel particles from being damaged during irradiation by strong particle-to-matrix bonding in conjunction with large matrix shrinkage, as had previously occurred for undiluted thermosetting rods. At the same time, the PBC had to be maintained at or above 17 wt.% to give the rods strength enough for adequate particle retention. Within the window of acceptability defined above, thermosetting rods performed about as well under irradiation as did the standard pitch-based rod that was included for comparison. Moreover, such rods offer processing advantages over the thermoplastic standard used in the fabrication of fresh fuel in that they can be subjected to free-standing carbonization, and this might be particularly important in the remote fabrication of reprocessed fuel.     

Crushing behaviour of high temperature reactor coated fuel particles

This paper examines in detail the crushing behaviour of high-temperature reactor fuel particles with pyrolytic carbon or silicon carbide outer coatings and discusses their failure mechanisms, in an attempt to relate crushing failure loads to coating strengths, and provide a simple, quick testing technique for quality control or performance assessment. Failure occurs by a series of mechanisms, in varying sequence, initiated by Hertzian cracking. Because the first event detected in a crushing test load/deflection curve is not the formation but the propagation of the Hertzian crack, the crushing load cannot be related to the coating strength; instead, it is governed by the fracture surface energy of the outer coating. A crushing test is therefore not a suitable technique for measurement of the strength of particle coatings. However, through measurement of the size of the contact surface, reliable estimates of the Young's modulus of the outer coating can be made by application of the Hertz theory of contact.     

Metal fission product behaviour in high temperature reactors -UO2 coated particle fuel

The migrational behaviour of a number of metallic fission products in the coated UO₂ particle fuel, proposed for HighTemperature Reactors, is described. The derivation of parameters enabling calculations to be made of the release of selected important isotopes is also discussed, particularly with reference to ‘Triso’ coated particles in which the silicon carbide layer is either defective or absent.     

Fabrication of uniform ZrC coating layer for the coated fuel particle of the very high temperature reactor

Fuel for the very high temperature reactor is required to be used under severer irradiation conditions and higher operational reactor temperatures than those of present high temperature gas cooled reactors. Japan Atomic Energy Agency has developed zirconium carbide (ZrC)-coated fuel particles previously in laboratory scale which are expected to maintain their integrity at higher temperatures and burnup conditions than conventional silicon carbide-coated fuel particles. As one of the important R&D items, ZrC coating process development has been started in the year 2004 to determine the coating conditions to fabricate uniform structure of ZrC layers by using a new large-scale coater up to 0.2 kg batch. It was thought that excess carbon formed in the ZrC layer under the oscillation of coating temperature would cause non-uniformity of the ZrC layer. Finally, uniform ZrC coating layer has been fabricated successfully by adjusting the time constant of the coater and keeping the coating temperature at around 1400 °C.     

Requirements for high temperature reactor fuel particle design assessment

In the framework of a large Research and Development programme devoted to High Temperature Reactors (HTR) and set up in the CEA from 2000 on, we will address ourselves to the issue of coated fuel performance and design. Although HTR fuel main features have been established for a long time, we need today to reassess the fuel design to make sure that it meets the requirements linked to the most recent projects of High Temperature Reactors. Thus, in collaboration with Framatome and in connection with the Gas Turbine - Modular Helium Reactor (GT-MHR) international project, we are planning to perform parametric thermal and mechanical studies, regarding different particle design options (kernel diameter, layers composition and thickness) and seeking optima concerning particle leak tightness and fission product retention. But to initiate such studies, we have first of all to define the design bases and the requirements for HTR fuel, in terms of kernel composition (fissile element, oxide stoechiometry, enrichment), particle and compact geometry (dimensions, particle volume fraction in the graphite matrix), power density, cooling gas temperature and irradiation conditions (burnup, fast fluence).     

Release of short-lived noble gases from HTGR fuel with failed coated fuel particles and contaminated matrix

Fuel compacts of the high-temperature gas-cooled reactor may contain a fraction of exposed uranium as defective coated fuel particles and a contamination of graphite matrix. Releases of short-lived noble gases were measured on the fuel compacts containing artificial failed particles as well as those having a highly contaminated matrix. The results were compared with the prediction by the JAERI model of short-lived gas release, which has been generated from previous irradiation testings. The release from the compacts with artificial failed particles agreed with the prediction except at lower temperatures where the fission-induced diffusion would predominate. The release from the matrix-contaminated compacts was different from the model prediction: The model fairly accurately predicted R/B of Xe, but significantly overpredicted that of Kr.     

Development of high temperature structural materials for the HTGR

In accordance with the HTGR program in Japan, a series of R&D for high temperature structural materials in particular with respect to the HTTR design code has been performed in JAERI for more than 20 years. This paper introduces R&D results of the pressure retaining low alloy steel 2 1/4Cr-1Mo and the high temperature structural alloys Hastelloy XR and Ni-Cr-W superalloy for the design code together with some fruits of recent studies.     

Irradiation performance and modeling of HTR-10 coated fuel particles

The irradiation test of HTR-10 spherical fuel elements was carried out in the Russian IVV-2M research reactor with the irradiation temperature of 1000 ± 50 °C. After the burnup reached 100,000 MWd/t, the irradiation temperature was raised to a higher temperature. The high R/B levels observed during the normal irradiation test were due to manufacture defects of one to four coated particles. Post-irradiation examination indicated that at normal irradiation condition, the pyrolytic carbon (PyC) and silicon carbide (SiC) layers of particles kept their integrity. However, after irradiation at higher temperatures, several types of defects including radial and tangential cracks in SiC layers, cracks in buffer layers, and through coating failure were found, and the failure fraction reached 5.8 × 10⁻². These defects were most likely caused by the higher thermal stresses generated. In this study, PANAMA fuel performance code was used to estimate the heating temperature in the irradiation test. The calculated results showed that when the heating temperature is much higher than 1850 °C, the failure fraction of coated particle can reach the level of 1%.     

Behavior of HTGR coated fuel particles and elements in tests under normal and accident conditions

Scientific-Industrial Organization Ray. Russian Scientific Center Kurchatov Institute. Siberian Division, Scientific-Research Construction Institute of Power Engineering. Translated from Atomnaya Énergiya, Vol. 73, No. 3, pp. 189–195, September, 1992.     

Estimation of maximum coated particle fuel compact packing fraction

This work develops an analytic fuel fraction packing model for a high temperature gas cooled reactor fuel compact fabricated from overcoated particles of a single size. The model includes the effects of one dimensional compression and finite matrix grain size. One dimensional compression limits the maximum fuel packing fraction to about 48% for the pressed compact in this single sized particle system. This limit is due to two effects. The first is that the process of die loading limits the pre-compression packing configuration to one that is stable under gravity, which is not the most space efficient one. The second effect is due to the one dimensional compression which reduces only the axial dimension of the particle lattice rather than uniformly compressing the lattice. The die wall can also limit the maximum packing fraction by preventing the nearby particles from moving into a more space efficient configuration.     

Current status and future development of coated fuel particles for high temperature gas-cooled reactors

The coated particles were first invented by Roy Huddle in Harwell 1957. Through five decades of development, the German UO₂ coated particle and US LEU UCO coated particle represent the highly successful coated particle designs up to now. In this paper, current status as well as the failure mechanisms of coated particle so far is reviewed and discussed. The challenges associated with high temperatures for coated particles applied in future VHTR are evaluated. And future development prospects of advanced coated particle suited for higher temperatures are presented. According to the past coated fuel particle development experience, it is unwise to make multiple simultaneous changes in the coated particle design. Two advanced designs which are modifications of standard German UO₂ coated particle (UO₂^∗ herein) and US UCO coated particle (TRIZO) are promising and feasible under the world-wide cooperations and efforts.     

Experimental and computational study of the pyrocarbon and silicon carbide barriers of HTGR fuel particle

HTGR safety is secured by a system of barriers limiting the emission of fission products from the core into the surrounding environment during normal operation and postulated anticipated accidents. An experimental-computational analysis of two fundamentally important barriers — fuel kernels and their coating, whose function is to contain radionuclides and to protect workers and the environment, is examined. The function of the barriers and the requirements which they must satisfy are examined for HTGR fuel particles. The results of post-reactor studies are analyzed. Mathematical models and computational codes simulating the behavior of fuel particles are analyzed. Probabilistic-statistical models and the GOLT code are being developed to evaluate the behavior of fuel particles under irradiation. Together with other models, this code is used for comparative test calculations of the behavior of particle fuel under normal irradiation conditions (<1300°C). The first results of such calculations are discussed. __________ Translated from Atomnaya énergiya, Vol. 105, No. 1, pp. 14–25, July, 2008.     

Crossflow characteristics of HTGR fuel blocks

Leakage crossflow characteristics in an HTGR core have been studied experimentally and numerically. Two-block crossflow experiments were carried out and the crossflow rate and the pressure difference were measured for various interface gap configurations. A numerical model has been proposed to predict crossflow rates, and the numerical results using the finite element method agreed well with experimental ones. In addition, empirical crossflow equations, which apply to various fuel blocks, were derived for the analysis of the flow distribution in an HTGR core.     

Irradiation behavior of atomized U–10wt.% Mo alloy aluminum matrix dispersion fuel meat at low temperature

In order to examine the in-reactor behavior of very-high-density dispersion fuels for high flux performance research reactors, U–10wt.% Mo alloy dispersions in an aluminum matrix have been irradiated at low temperature in the Advanced Test Reactor (ATR). The alloy fuel dispersant was produced by a centrifugal atomization process. The fuel shows stable in-reactor irradiation behavior to a fission density of 5×10²⁷ m⁻³ at an irradiation temperature of 65 °C. The fuel–matrix interaction layer growth rate is similar to that observed in uranium-silicide fuels. The fuel particles have a fine and a relatively narrow fission gas bubble size distribution. There appears to be features in the microstrucure that are the result of segregation of the microstructure in to molybdenum rich and depleted regions on solidification.     

高温气冷堆包覆燃料颗粒破损机制及失效模型

高温气冷堆的燃料元件的基本构成单元是全陶瓷型的包覆燃料颗粒,其性能决定了高温气冷堆的安全性。除了传统的辐照实验检测外,建立理论模型对其研究具有重要的意义。本文主要介绍了TRI-SO型包覆燃料颗粒的结构及破损机制,以及国外现有的几个主要模型的基本假设,计算原理和特点,通过对比几个模型的优缺点,提出今后研究的方向。     

The modeling of graphite oxidation behavior for HTGR fuel coolant channels under normal operating conditions

The oxidation of graphite in normal operating conditions is a very important factor when evaluating the service time of the graphite structural material in a high temperature gas-cooled reactor (HTGR). This paper deals with the modeling of graphite oxidation by steam in the helium channel of a fuel block. The FEM software COMSOL is used: the turbulent flow of the coolant is simulated by using the k-? model and the chemical reaction is expressed by the Langmuir-Hinshelwood equation. Calculations were carried out for steam pressures around 1 Pa and for different temperature distributions. The influence of burn-off and the diffusion in graphite porosities were both considered in the oxidation. Results show that oxidation mainly occurred on the graphite surface at the bottom of the core because of the higher temperature. The thickness of graphite with a burn-off higher than 8% was about 1 mm at the core base. Less than 15% of steam was consumed in the coolant channel of the fuel assemblies. Calculations also showed that the mean gasification rate in one channel for the second service time was larger than the first service time.     

Irradiation performance of experimental fuel particles coated with silicon-alloyed pyrocarbon: A review

A review has been conducted on the use of silicon-alloyed pyrocarbon (Si-PyC) as an improved coating material for the two types of fuel particles used in the cores of high-temperature gas-cooled reactors. Based on recent data from extensive irradiation testing and postirradiation annealing of such experimental fuel particles, it is concluded that Si-PyC coatings offer considerable promise as replacements for the standard pure pyrocarbon (PyC) coatings used on thorium-based fertile fuels that have BISO coating designs. The primary advantage here is improved retention of fission products from bred U-233, with diffusion coefficients being as much as 100 times smaller for Si-PyC than for PyC. However, there is no significant improvement in mechanical performance of Si-PyC coatings over standard PyC coatings under irradiation. As a result, there is no incentive for using these coatings on TRISO particle designs of the type used on uranium-based fissile fuels, because here a silicon carbide barrier layer provides superior fission-product retention.