Diffusion and Evaporation of Fission Products in Coated Fuel Particles

Fractional releases of ¹³³Xe, ¹⁴⁰Ba and ⁸⁹Sr from slightly-irradiated pyrolytic-carbon-coated and SiC-coated particles were measured over a temperature range of 1,200°–1,750°C. The results are analyzed mathematically in order to obtain the diffusion and evaporation coefficients relevant to PyC and SiC. The resulting expressions for the coefficient of diffusion in PyC are 2.9x10-⁷ exp(-61x10³/RT) for ¹³³Xe and 4.7x10-² exp(51x10³/RT) for ¹⁴⁰Ba. For the coefficients of evaporation of ¹⁴⁰Ba from PyC, the expression is 3.5x10³ exp(-42x10³ /RT). As for SiC, the diffusion and evaporation coefficients of these nuclides are given for a temperature of 1,750°C. A high diffusivity path for the diffusion of ¹⁴⁰Ba is postulated to explain the difference in diffusion behavior between ¹³³Xe and ¹⁴⁰Ba in PyC.     

HTGR燃料芯核及包覆颗粒的物性测试

本文介绍用图象分析技术对高温气冷堆燃料芯核的直径、密度和球形度,以及包覆燃料颗粒的疏松碳层,致密碳层和碳化硅层的涂层厚度和涂层密度的定量测试方法。本方法快速、准确,测量相对标准偏差为1%。     

Detection of Failed Coated Particles in HTGR Fuels by Acid Leaching

By acid leaching of several kinds of irradiated coated particle fuels, uranium-and ¹³⁷Cs-leaching fractions were measured. These fractions were compared with the surface failure fraction of the coated particles by a visual inspection in the post-irradiation examination. From the measurements it was found that ¹³⁷Cs was unsuitable as the objective nuclide for acid leaching, because this nuclide escaped from failed coated particles during irradiation or trapped in the graphite grains of the fuel-compact matrix, and therefore, it was difficult to relate the leaching fraction of ¹³⁷Cs to the failure fraction of the coated particles. Cesium-137 leaching fraction was less than 10% of the uranium fraction. The uranium leaching fraction agreed fairly well with the surface failure fraction. Therefore, uranium is regarded as the suitable objective nuclide for acid leaching. Also, combination of the acid leaching and the surface inspection is an useful method for detection of failed coated particles.     

Improvements in Quality of As-Manufactured Fuels for High-Temperature Gas-Cooled Reactors

The mechanisms of coating failure of the fuel particles for the high-temperature gas-cooled reactors during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the defective particle fraction of the as-manufactured fuels. Through the observation of the defective particles, it was found that the coating failure during the coating process was mainly caused by the strong mechanical shocks to the particles given by violent particle fluidization in the coater and by unloading and loading of the particles. The coating failure during the compaction process was probably related to the direct contact with neighboring particles in the fuel compacts. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the as-manufactured fuel compacts was improved outstandingly.     

Failure of Coated Fuel Particles during Thermal Excursion above 2,000°C

In an attempt to obtain information on survival temperature limit for TRISO particles, out-of-pile heating experiment was carried out on unirradiated, unbonded particles. Catastrophic failure resulting in the formation of a hole like a volcano crater occurred during heating at 2,250~2,300°C for 20~35 min. Although the failure temperatures were lower than the melting point of the kernel material (UO₂), visual and microscopic observations of failed particles suggests kernel melting. Failure of SiC layer was detected after heating at temperatures as low as 2,130°C. At 2,560°C, SiC layer disappeared completely. A similar heating experiment on particles embedded in graphite matrix was carried out for the purpose of comparison. The embedded particles survived higher temperature excursion than in the case of the unbonded particles. Effects of heating on OPTAF of PyC and on crushing strength of particles are also reported.     

Research and Development of HTTR Coated Particle Fuel

The coated particle fuel has been developed within a framework of the HTTR (High Temperature engineering Test Reactor) Development Program at the Japan Atomic Energy Research Institute. The HTTR fuel is a prismatic block type containing TRISO-coated U0₂ particles. Research and development on the fuel has been progressed in three categories; a work for fuel production technology, a proof test of fuel performance and a safety-related research. In the present report the concept and outline of the fuel in the HTTR design are firstly described, and then fuel fabrication technology including recently developed methods for improving fuel quality is followed. Tests for proving fuel performance have been carried out extensively on the reference fuel of the HTTR design by irradiation in an in-pile gas loop and capsules, and typical results are presented in this report. Concerning the safety-related research, fuel failure and ¹³⁷Cs release at abnormally high temperature are described.     

UO2燃料颗粒涂层工艺

本文简述了UO_2燃料的涂层工艺和涂层原理,评价了某些涂层设备,讨论了涂层工艺参数(涂层温度、进料参数、涂层时间等)对涂层过程及结果的影响,简单介绍了涂层产品质量的检验方法。     

10MW高温气冷堆包覆燃料颗粒的研制

我国10MW高温气冷堆采用全陶瓷型包覆颗粒球形燃料元件。TRISO型包覆燃料颗粒由燃料核芯、疏松热解碳层、内致密热解碳层、碳化硅层和外致密热解碳层组成。采用丙烯和乙炔混合气体制备致密热解碳层以及四层连续包覆的新工艺,开展生产工艺条件试验,系统地研究了生产工艺和性能之间的关系,摸索出最佳生产工艺条件。用化学气相沉积方法在150mm流化床沉积炉系统中批量生产出TRISO型包覆燃料颗粒。用扫描电镜观察分析了包覆燃料颗粒的微观结构,包覆燃料颗粒的制造破损率为3.4×10-6,冷态性能达到我国10MW高温气冷堆设计要求。包覆燃料颗粒辐照考验结果(放射性裂变产物释放率R/B为1×10-6左右)表明,包覆燃料颗粒的质量可以满足10MW高温气冷堆安全运行的要求。     

水冷反应堆燃料元件的在役检测和处理

水冷反应堆包括轻水堆和重水堆,轻水堆分为压水堆和沸水堆;重水堆分为加压重水堆和加拿大的氘铀堆。国际上把它们归为一类进行研究。本文涉及的破损燃料元件的在役检测和处理包括：反应堆运行时的检测;换料时或换料后的检测;在燃料组件内鉴别破损的燃料棒;燃料组件的监测、拆卸和修复;破损燃料棒拆出后的检测,破损定位与修补。     

水冷反应堆燃料元件的在役检测和处理

水冷反应堆包括轻水堆和重水堆,轻水堆分为压水堆和沸水堆;重水堆分为加压重水堆和加拿大的氘铀堆。国际上把它们归为一类进行研究。本文涉及的破损燃料元件的在役检测和处理包括：反应堆运行时的检测;换料时或换料后的检测;在燃料组件内鉴别破损的燃料棒;燃料组件的监测、拆卸和修复;破损燃料棒拆出后的检测,破损定位与修补。     

Behavior of Coated Fuel Particle of High-Temperature Gas-Cooled Reactor under Reactivity-Initiated Accident Conditions

In order to clarify the failure mechanism and determine the failure limit of the High-Temperature Gascooled Reactor (HTGR) fuel under reactivity-initiated accident (RIA) conditions, pulse irradiations were performed with unirradiated coated fuel particles at the Nuclear Safety Research Reactor (NSRR). The energy deposition ranged from 0.578 to 1.869 kJ/gUO₂, in the pulse irradiations and the estimated peak temperature at the center of the fuel particle ranged from 1,510 to 3,950 K. Detailed examinations after the pulse irradiations showed that the coated fuel particles failed above 1.40 kJ/gUO₂, where the peak fuel temperature reached over the melting point of UO₂ fuel. It was concluded that the coated fuel particle was failed by the mechanical interaction between the melted and swelled fuel kernel and the coating layer under RIA conditions.     

Fabrication of the First-Loading Fuel of the High Temperature Engineering Test Reactor

The High Temperature Engineering Test Reactor (HTTR). which is the first high temperature gas-cooled reactor (HTGR) in Japan, attained its first criticality in November 1998. The fabrication of the first-loading fuel started June 1995 and in December 1997, 150 fuel assemblies were completely formed. A total of 66,780 fuel compacts, corresponding to 4,770 fuel rods, were successfully produced through the fuel kernel, coated fuel particle and fuel compact processes. Fabrication technology for the fuel was established through a lot of research and development activities and fabrication experiences of irradiation samples. As-fabricated fuel compacts contained almost no through-coatings failed particles and few SiC-defective particles. Average through-coatings and SiC defective fractions were as low as 2 × 10^–6 and 8 × 10^–6, respectively. This paper describes (1) characteristics of as-fabricated fuel, (2) the experiences obtained from the first mass-production and (3) prediction of irradiation performance of the fuel in the HTTR.     

Real-Time High-Sensitivity Fuel Failure Detection for HTGR

A real-time high-sensitivity fuel failure detection (FFD) method has been developed, where a wire precipitator radiation detector measures noble-gas fission products (FPs) released from a High Temperature Gas-Cooled Reactor (HTGR). By changing the reference counting rate of the precipitator between the normal state and the failed fuel state in real time in response to reactor operation conditions, i.e. reactor power, fuel temperature, coolant-gas flow rate and so on, fuel failure with an extremely low failure fraction (Release-to-Birth ratio <5×10^?6) can be detected. The reference counting rate is obtained by adding an operational tolerance to the background counting rate that is estimated by a diagnostic equation. The diagnostic equation consists of a release equation for estimating the release rate of noble-gas FPs, a gas circulation equation for calculating concentrations of noble-gas FPs in the primary coolant system and a response equation for determining the detection efficiency of the wire precipitator. The feasibility of the method was evaluated by irradiation experiments using gas swept capsules and the Oarai Helium Gas Loop (OGL-1) in the Japan Material Testing Reactor (JMTR). The background counting rate was estimated with an error of about 20% in real time by the diagnostic equation.     

Verification of Fission Product Release Model from High Temperature Engineering Test Reactor Fuel

A fuel assembly of the High Temperature Engineering Test Reactor (HTTR) is composed of fuel rods and a hexagonal graphite block. A fuel rod is composed of the fuel compacts and a graphite sleeve. The coated fuel particles are incorporated into a graphite matrix to form a fuel compact. The fuel consists of microspheres of low-enriched U02 with a TRISO coating. The TRISO coatings consist of a porous pyrolytic carbon (PyC) buffer layer followed by an isotropic PyC layer, a SiC layer and a final (outer) PyC layer.

In order to evaluate amounts of fission products released from the HTTR fuel rods during normal operation, analytical models have been developed. Fractional releases of noble gases and iodine are calculated based on release data of ⁸⁸Kr which are obtained by irradiation tests with failed coated fuel particles. The transport of the metallic fission products through the kernel, coatings, fuel compact and graphite sleeve is modeled as a diffusion process. These analytical models have been verified by comparison with measured fractional releases in in-reactor tests and have been concluded to be applicable to the HTTR fuel condition.     

Small PWR Using Coated Particle Fuel of Thorium and Plutonium

An innovative concept of PFPWR50 for district heating has been studied, which is a small PWR of 50MWt capability using coated particle fuels with conventional zircaloy cladding. This concept takes advantages of fuel integrity against fission products release of coated particle fuels and a high reliability of PWR technology based on the long history of a successful operation. We have investigated burnup characteristics of fuel rods, assemblies, and reactor cores by the calculation code SRAC95 in order to establish a core concept of long life without on-site refueling. The loading pattern of assemblies with various concentrations of burnable poison is optimized to obtain a flat excess reactivity during the core life in order to eliminate a soluble boron control system. The core life of a cycle is about 8.9 equivalent full power years. And we have also studied the applicability of SiC/SiC composite cladding in place of zircaloy cladding, which is now under development for gas cooled fast reactor fuels. It could be applicable to high burnup fuel rods for a long term operation. From the calculation results, it is found out that the burnup characteristics do not change significantly with SiC cladding and contribute to elongate the core life to 9.2 equivalent full power years.     

高温气冷堆燃料元件发展现状和趋势

本文介绍了高温气冷堆燃料元件的发展历史，现状和趋势。经过３０多年的研究和发展，燃料元件的设计，制造工艺和质量鉴证技术已相当成熟，燃料元件可在１２５０℃长期工作。２１２０００个ＴＲＩＳＯ颗粒辐照试验的时没有一个破损，１６００℃下辐照后退火５００ｈ，阻挡裂变产物释放的能力没有下降。     

Validity of Constant Collision Density Assumption in Intermediate Energy Region

Several kinds of coated fuel particles, with their coating either intact or artificially cracked, were heated out-of-pile in such manner as to create a sharp temperature gradient across the particles (60°120°C per particle), at temperatures from 1,500° to 1,950°C. The purpose was to obtain information on the displacement of the kernel material relative to the coating. To examine this amoeba effect, the particles were observed, after heating, by both ceramography and ×-ray radiography. The results revealed that:

(1) In the case of UO₂ kernel with artificially impaired coating, their kernels were found to move more readily toward the crack, regardless of the temperature gradient, as compared with UC₂.

(2) The amoeba effect is observed even in out-of-pile heating on intact coated particles with UO₂kernel which moves down the temperature gradient. This UO₂ movement was given a new explanation based on the evaporation and subsequent condensation of the UO₂ within the particle, when the coating is intact.

(3) In case of UC₂ kernel, which moves up the temperature gradient, the sealing-in of the kernel by the intact coatings appears to assume a controlling factor, and the occurrence of evaporation is negligible.     

Measurements of Doppler Effect of Coated Particle Fuel Rod in SHE-14 Core Using Sample Heating Device

Reactivity decrease due to temperature rise of a single fuel rod sample was measured in the SHE-14 core using a sample heating device with purpose to verify the calculation accuracy of the Doppler effect for resonance neutron absorption in a Very High Temperature Reactor. The fuel rod sample was a stuck of coated particle fuel compacts containing 4% enriched UO₂ kernels, and it was heated up to about 700°C in a sample heating tube which was placed along the axis of the core.

The difference of reactivity decrease between the two same size samples of fuel rod and graphite rod due to temperature rise can be interpreted as the increased resonance neutron capture of ²³⁸U, i.e. Doppler effect. The SRAC code system was applied to the Doppler effect calculation where the collision probability method was used in the cell calculation and the one-dimensional, multi-group diffusion approximations were adopted in the core calculation. The results gave a ratio of the calculated to the measured Doppler effect of 0.93.