TRISO coated fuel particles with enhanced SiC properties

The silicon carbide (SiC) layer used for the formation of TRISO coated fuel particles is normally produced at 1500-1650 °C via fluidized bed chemical vapor deposition from methyltrichlorosilane in a hydrogen environment. In this work, we show the deposition of SiC coatings with uniform grain size throughout the coating thickness, as opposed to standard coatings which have larger grain sizes in the outer sections of the coating. Furthermore, the use of argon as the fluidizing gas and propylene as a carbon precursor, in addition to hydrogen and methyltrichlorosilane, allowed the deposition of stoichiometric SiC coatings with refined microstructure at 1400 and 1300 °C. The deposition of SiC at lower deposition temperatures was also advantageous since the reduced heat treatment was not detrimental to the properties of the inner pyrolytic carbon which generally occurs when SiC is deposited at 1500 °C. The use of a chemical vapor deposition coater with four spouts allowed the deposition of uniform and spherical coatings.     

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.     

On mathematical models for calculating the mechanical behaviour of coated fuel particles

Coated fuel particles, as used in high-temperature gas-cooled reactors, consist of a fuel kernel and a spherically symmetric multilayer coating. During their lifetime they undergo varying stress states, caused by the simultaneous occurrence of a number of basic physical effects.     

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%.     

Reaction rates in blanket assemblies of a fusion-fission hybrid reactor

To validate neutronics calculation for the blanket design of fusion-fission hybrid reactor,experiments for measuring reaction rates inside two simulating assemblies are performed.Two benchmark assemblies were developed for the neutronics experiments.A D-T fusion neutron source is placed at the center of the setup.One of them consists of three layers of depleted uranium shells and two layers of polyethylene shells,and these shells are arranged alternatively.The ²³⁸U capture reaction rates are measured using depleted uranium foils and an HPGe gamma spectrometer.The fission reaction rates are measured using a fission chamber coated with depleted uranium.The other assembly consists of depleted uranium and LiH shells.The tritium production rates are measured using the lithium glass scintillation detector which is placed in the LiH region of the assembly.The measured reaction rates are compared with the calculated ones predicted using MCNP code,and C/E values are obtained.     

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.     

Progress in understanding fission-product behaviour in coated uranium-dioxide fuel particles

Supported by results of calculations performed with two analytical tools (MFPR, which takes account of physical and chemical mechanisms in calculating the chemical forms and physical locations of fission products in UO₂, and MEPHISTA, a thermodynamic database), this paper presents an investigation of some important aspects of the fuel microstructure and chemical evolutions of irradiated TRISO particles. The following main conclusions can be identified with respect to irradiated TRISO fuel: first, the relatively low oxygen potential within the fuel particles with respect to PWR fuel leads to chemical speciation that is not typical of PWR fuels, e.g., the relatively volatile behaviour of barium; secondly, the safety-critical fission-product caesium is released from the urania kernel but the buffer and pyrolytic-carbon coatings could form an important chemical barrier to further migration (i.e., formation of carbides). Finally, significant releases of fission gases from the urania kernel are expected even in nominal conditions.     

On the safety of conceptual fusion-fission hybrid reactors

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium.As a result of these studies, it appears that highly reliable and even redundant decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered.     

聚变裂变混合发电堆水冷包层热工水力学设计分析

一种以能量倍增为目标的聚变裂变混合发电堆(FDS-EM)概念已被提出,FDS-EM初步设计为可以产生约1.0 GW的电功率,并能实现氚自持。本文对FDS-EM水冷包层进行了热工水力学设计与分析。设计采用了压水堆的成熟技术,并给出了典型的热工设计参数,通过对典型参数下包层的数值模拟分析,得出了温度场和应力场分布,初步证明了设计的工程可行性。     

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.     

A summary of EPRI's fusion-fission hybrid evaluation activities

A summary is given of recently completed and planned fusion-fission hybrid projects. Electricity supply/demand projections and estimates of future uranium requirements for several different combinations of nuclear systems, including hybrids, are discussed.     

聚变裂变混合发电堆水冷包层中子学设计分析

主要针对聚变裂变混合发电堆FDS-EM水冷包层的能量倍增因子M和氚增殖率TBR等中子学参数进行优化计算。FDS-EM包层主要设计目标是在氚自持的基础上获得约1 GW的电功率,并且尽可能长时间连续运行不换料。通过初步设计分析给出一个使用核废料(压水堆卸出的废料钚、锕系加上贫铀)作为裂变燃料,能够实现氚自持、能量倍增因子约为90等设计目标,且连续运行至少10年不换料的中子学方案。     

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.     

多功能聚变裂变混合实验堆FDS-MFX氦冷包层三维中子学初步设计与分析

FDS-MFX(Multi-Functional eXperimental fusion-fission hybrid reactor)是一个基于现实可行技术的多功能聚变裂变混合实验堆概念,分3个阶段相继开展实验研究,分别采用纯氚增殖包层、铀燃料包层和乏燃料包层.本文重点对其中铀燃料包层后期阶段中高浓缩铀模块的摆放方式...     

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.     

基于GDT的聚变裂变混合堆堆芯参数初步设计研究

基于Gas Dynamic Trap(GDT)装置的实验进展,提出了用于驱动聚变裂变混合堆包层的聚变堆芯参数设计。基于零维堆芯物理模型,计算分析给出了一套聚变功率为50MW的初步堆芯参数方案。利用GDT装置的实验结果对该物理模型进行计算对比校验,显示该物理模型和设计参数的可靠性。     

A cost/benefit analysis of commercial fusion-fission hybrid reactor development

A simple algorithm was developed that allows rapid computation of the ratio,R, of present worth of benefits to present worth of hybrid R&D program costs as a function of potential hybrid unit electricity cost savings, discount rate, electricity demand growth rate, total hybrid R&D program cost, and time to complete a demonstration reactor. In the sensitivity study, these variables were assigned nominal values (unit electricity cost savings of 4 mills/kW-hr, discount rate of 4%/year, growth rate of 2.25%/year, total R&D program cost of $20 billion, and time to complete a demonstration reactor of 30 years), and the variable of interest was varied about its nominal value. Results show thatR increases with decreasing discount rate and increasing unit electricity savings and ranges from 4 to 94 as discount rate ranges from 5 to 3%/year and unit electricity savings range from 2 to 6 mills/kW-hr.R increases with increasing growth rate and ranges from 3 to 187 as growth rate ranges from 1 to 3.5%/year and unit electricity cost savings range from 2 to 6 mills/kW-hr.R attains a maximum value when plotted against time to complete a demonstration reactor. The location of this maximum value occurs at shorter completion times as discount rate increases, and this optimal completion time ranges from 20 years for a discount rate of 4%/year to 45 years for a discount rate of 3%/year.The views expressed in this paper are solely those of the author and do not necessarily represent the views of the U.S. Department of Energy.     

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.     

Embedding of coated particles in SiC

The disposal of spent HTR fuel elements requires a relatively large volume due to the integration of fuel particles and moderator graphite. This can be reduced by separating the coated particles from the graphite matrix. However, the coated particles cannot be disposed without a suitable embedding or backfill. Silicon carbide has been identified as a potential embedding material with less porosity as is the case with graphite. Therefore, a method has been developed to proof the technical feasibility of embedding coated particles in silicon carbide.