Development of advanced HTR fuel elements

Spherical fuel elements with coated particles are the essential constituents of the integrity of a pebble-bed reactor, particularly with respect to fission product retention. While all gas-cooled reactors exhibit an extremely low primary circuit activity, recent developments have reduced the source term to virtually zero. This entails an enormous effort in the characterization of the fuel behaviour in manufacture, during irradiation, and in accident testing.With modern SiC-coated particles it can be shown that all radiologically relevant fission products are completely retained inside the silicon carbide layer of intact fuel particles. The dominant source term for fission product release will, therefore, be the small number of particles with defective coatings. Permanent damage to the fission product retaining SiC layer can only start in accidents involving temperatures significantly beyond 1600°C.     

Revisiting the concept of HTR wallpaper fuel

As early as the 1970s, attempts were made to reduce the peak fuel temperature by means of so-called “wallpaper fuel”, in which the fuel is arranged in a spherical shell within a pebble: By raising particle packing fraction, fuel kernels are condensed to the outer diameter of the fuel zone, leaving a central part of the pebble free of fuel. This modification prevents power generation in this central fuel-free zone and decreases temperature gradient across the pebble.Besides particle temperature reduction, the wallpaper concept also enhances neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burn-up. To assess such improvements, calculations were performed using Monte Carlo neutron transport and depletion codes MCNP/MCB. Among others, investigations of conversion ratio, temperature coefficient of reactivity, spent fuel composition and neutron multiplication (for which a method to determine the six-factor formula was developed), were conducted.It is demonstrated that this fuel type impacts positively on the fuel cycle, reduces production of minor actinides (MA) and improves the safety-relevant parameters of the reactor. A comparison of these characteristics with PBMR-type fuel is presented: By comparison with PBMR fuel, the “wallpaper design” results in an effective neutron multiplication coefficient increase (by about 1750 pcm), which is combined with a decrease of between 4.6 and 17.5% in MA production. An improved neutron economy of the heterogeneous design enables enrichment of the “wallpaper type” of fuel to be reduced by more than 6%.The fuel changes suggested in this paper offer more versatility to the HTR concept: Conversion ratio can be decreased (leading to lower MA build-up and fuel reprocessing cost) or raised (leading to lower fuel consumption and fuel cost). Variations around this concept also enable higher reactivity, thus higher achievable burn-up, improving sustainability of HTRs.     

EBSD investigation of SiC for HTR fuel particles

Electron back-scattering diffraction (EBSD) can be successfully performed on SiC coatings for HTR fuel particles. EBSD grain maps obtained from thick and thin unirradiated samples are presented, along with pole figures showing textures and a chart showing the distribution of grain aspect ratios. This information is of great interest, and contributes to improving the process parameters and ensuring the reproducibility of coatings.     

Surface analysis of HTR fuel particles by X-ray photoelectron spectroscopy

An experimental technique is described which allows X-ray photoelectron (X-pe) spectra to be obtained from small samples with surface area ~ 1 mm² (in this case an HTR fuel particle), using an AEI ES 300 spectrometer. It is shown that the spectra obtained from these particles are of sufficient quality to provide chemical information concerning the nature of the surface. Silicon is found to be present in the surface of the pyrocarbon layer deposited on the fuel particles.     

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。     

Research on the incineration of plutonium in a modular HTR using thorium-based fuel

Large stockpiles of plutonium have accumulated globally. In order to avoid the production of second-generation plutonium during the reuse of this nuclear fuel, it can be incinerated in combination with the fertile material thorium instead of uranium. The coated-particle fuel of high-temperature reactors (HTR) allows a very high heavy metal burnup and thus it can achieve a very high incineration fraction of the initially loaded plutonium (86%) already during the first reuse in a modular pebble-bed HTR. Only a negligible amount of second-generation plutonium is bred from the small quantity of highly enriched uranium, which is used as make-up fissile material. An optimization of the fractions of Pu, U and Th contained in the fuel elements has been done on condition that a negative temperature coefficient of the reactivity has to be achieved over the whole range of operating temperature.     

Considerations pertaining to the achievement of high burn-ups in HTR fuel

The satisfactory irradiation performance of coated fuel particles up to burn-ups of, say 10%, has been demonstrated in the past. However, it is shown that some of the most important physical properties of their constituents, which determine this good performance, are only poorly known. This ignorance is likely to become more serious when particles are required to attain higher burn-up values. For example, neutron dose limits above which the anisotropy of the pyrocarbon layers become unacceptably high need to be established. Again, the particle must be designed with adequate voidage to accommodate the gas released by fission, and kernel-coating mechanical interaction must be avoided. Experiments are proposed, in addition to the irradiation and post irradiation examination of particles, which should address these issues. The way this new information can be used in fuel performance codes is discussed, thereby providing a scientific understanding of the behaviour of coated particles—as of interest to researchers, by industry, by utilities and, perhaps most importantly, by regulators.     

Thermal-fluid characteristics of the transuranics fuel in a deep-burn HTR core

The Deep Burn Project is evaluating the feasibility of the DB-HTR (Deep Burn High Temperature Reactor) to achieve a very high utilization of transuranics (TRU) derived from the recycle of LWR spent fuel. This study intends to evaluate the thermal-fluid and safety characteristics of TRU fuel in a DB-HTR core using GAMMA+ code. The key design characteristics of the DB-HTR core are more fuel rings (five fuel-rings), less central reflectors (three rings) and decay power curves due to the TRU fuel compositions that are different from the UO₂ fuel. This study considered three types of TRU kernel compositions such as 100%(PuO₂ + NpO₂ + Am), 99.8%(PuO_1.8, NpO₂) + 0.2%UO₂ + 0.6 mole SiC getter, and 70%(PuO_1.8, NpO₂) + 30%UO₂ + 0.6 mole SiC getter. The first fuel type of TRU kernel produces higher decay power than the UO₂ kernel. For the second and the third fuel types, removing the initial Am isotopes and reducing the volumetric packing fraction of TRISO particles will reduce the decay power. The flow distribution, core temperature and TRISO temperature profiles at the steady state were examined. As a safety performance, this study mainly evaluated the peak fuel temperature during LPCC (low pressure conduction cooling) event with considering the impact of decay power, the annealing effect of the irradiated thermal conductivity of graphite, and the impact of the FB (fuel block) end-flux-peaking. For the 600 MW_th DB-HTR core, the peak fuel temperature of 100%(PuO₂ + NpO₂ + Am) TRU was found to be much higher than the transient fuel design limit of 1600 °C due to the lack of heat absorber volume in the central reflector as well as to the increased decay power of the TRU fuel compositions. For a 0.2%UO₂ mixed or a 30%UO₂ mixed TRU, the peak fuel temperature was decreased due to the reduced decay power, however, it was still higher than 1600 °C due to the lack of heat absorber volume in the central reflector.     

Modeling of fission product release from HTR fuel for risk analyses

In the United States and Federal Republic of Germany, methodologies have been developed to determine the performance of and fission product release from TRISO-coated fuel particles under postulated accident conditions. The paper presents a qualitative and quantitative comparison of models used by General Atomics (GA) and by the German Nuclear Research Center at Jülich (KFA/ISF). A benchmark calculation was performed for fuel temperatures predicted for the U.S. Department of Energy (DOE) sponsored Modular High-Temperature Gas-Cooled Reactor (MHTGR). Good agreement in the benchmark calculations supports the on-going efforts to verify and validate the independently developed codes of GA and KFA/ISF.     

Long time experience with the development of HTR fuel elements in Germany

The development of spherical fuel elements for HTR-designs in Germany is discussed. Special attention is given to the development, production and characterization (incl. kernel and coatings) as well as to the irradiation and post-irradiation examination of the different coated particle systems. It has been demonstrated in various irradiation tests which were supplemented by heating tests that for a modular HTR power plant (with a thermal output of 200 MJ s⁻¹) during the specified normal operation as well as in the case of incidents and even accidents, where the maximum fuel temperature will be below 1620 °C, the fission product release is very low. In this context, it must be mentioned that the present coated particle design has not yet been optimized for the combination of high burn-up and high temperature resistance under accident conditions. The TRISO fuel available is a result from fuel development for large HTR's with steam turbines in a time when the modular concept was not yet been invented although its capabilities inspired the design of modular reactors. Thus, there is still a huge potential for improvement of coated particles especially when plutonium or actinide burning is also taken into account.     

Comments on americium volatilization during fuel fabrication for fast reactors

The physical processes relevant to the fabrication of metallic nuclear fuels are analyzed, with attention to recycling of fuels containing U, Pu, and minor volatile actinides for use in fast reactors. This analysis is relevant to the development of a process model that can be used for the numerical simulation and prediction of the spatial distribution of composition in the fuel, an important factor in fuel performance.     

Fission product migration in the coatings of irradiated uranium dioxide HTR fuel particles

Electron probe microanalysis has been used to determine quantitative profiles of the fission products caesium, xenon and barium in the coatings of UO₂ HTR fuel particles irradiated for up to 140 days to burn-ups of between 5% and 11% heavy atom fission at temperatures of 1100–1500°C. The xenon profiles were typical of recoil. Barium profiles showed a peak due to recoil but also a low level ‘tail’ extending to the silicon carbide layer indicating a fast diffusing barium species. In contrast to barium, strontium profiles were not observed. Caesium was the most abundant fission product in the inner pyrocarbon coating. The profiles often extended to the silicon carbide layer where a marked fall in caesium concentration was observed. Some profiles displayed zones of fairly constant caesium concentration which it is suggested correspond to caesium/pyrocarbon ‘compound’ formation as opposed to physical aspects of the manufacturing process.     

Operation of an experimental facility for fabrication of fuel elements and fuel assemblies of the bor-60 containing vibrocompacted fuel

Translated from Atomnaya Énergiya, Vol. 67, No. 5, pp. 320–323, November, 1989.     

Build-up of plutonium isotopes in HTR fuel elements a comparison between computed prediction and chemical analysis

Irradiation experiments with HTR fuel elements in the AVR reactor are described. It will be shown how fuel elements are inserted into the fuelling process so selectively that, after a single core passage, it is possible to uniquely identify them by means of the burn-up measuring system so as to enable a selective discharge. The method of chemical digestion of the fuel elements is described and the results of mass spectrometric analyses are presented. Computer analyses with the HTR-2000 reactor simulation program are described. The experimental results will be compared with the results from computer simulation.     

Block-type HTGR spent fuel processing: CEA investigation program and initial results

Considering the criteria put forward by the Generation IV Forum to select future nuclear systems (preservation of resources, minimization of final waste impact, economics, etc.), the impetus for the very high-temperature He-cooled reactor raises the issue of processing the fuel in the context of a closed cycle with actinide recycling. Due to the unique structure of the particle fuel used for this class of reactors, the difficulties essentially involve accessibility to the uranium kernels coated by carbon and SiC layers and dispersed in a large volume of graphite. Starting from past experience in this field, a research program has been recently undertaken by the CEA to propose attractive solutions. The mechanical extraction of compacts from the spent fuel blocks appears to be a promising approach, as well as removing the graphite from the compacts by pulsed currents to free the particles. Subsequent removal of the carbon and silicon carbide layers by high temperature oxidation or by carbochlorination to access the kernels is assessed. For actinide recycling, gelation appears to be a suitable process for fabricating the kernels. This paper provides a brief overviews of the developments currently in progress at the CEA.     

Evolution of defects in micropellet fuel during the fabrication of HTGR compacts

The stress distribution arising in micropellets and cylindrical fuel compacts during fabrication, the stress concentration in micropellets located near the surface of a compact, and the evolution of defects in micropellets as a function of the type of stress state are investigated.It has been found that an ensemble of micropellets with a large number of particles contains a continuous spectrum of defects in the range 10^–4–10² µm. Mechanical stresses engender evolution of the defects according to the scheme accumulation of microdefects microcracks cracks through defects.Recommendations are formulated for lowering the number of defects in micropellets during deposition of coatings on the micropellets and compaction.Research Institute of the Luch Scientific and Industrial Association.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 44–50, January, 2005.     

Key differences in the fabrication, irradiation and high temperature accident testing of US and German TRISO-coated particle fuel, and their implications on fuel performance

Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally has displayed gas release values during irradiation three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.     

A US perspective on fast reactor fuel fabrication technology and experience. Part II: Ceramic fuels

This paper is Part II of a review focusing on the United States experience with oxide, carbide, and nitride fast reactor fuel fabrication. Over 60 years of research in fuel fabrication by government, national laboratories, industry, and academia has culminated in a foundation of research and resulted in significant improvements to the technologies employed to fabricate these fuel types. This part of the review documents the current state of fuel fabrication technologies in the United States for each of these fuel types, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.