Integrity Confirmation Tests and Post-irradiation Test Plan of the HTTR First-Loading Fuel

Since the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is the first mass-production High Temperature Gas-cooled Reactor (HTGR) fuel in Japan, their quality should be carefully inspected. For the quality control related to the fabrication process, Japan Atomic Energy Research Institute (JAERI) carried out the tests to certify the fuel integrity during operation. The tests comprise (1) as-fabricated SiC failure fraction measurement, (2) high-temperature heatup test of irradiated fuel and (3) accelerated irradiation test. For (1), the SiC failure fraction was measured independently in JAERI in addition to the measurement in the fabrication process. The measured failure fractions agreed within 95% confidence limit. In order to confirm the integrity of the SiC layer with respect to the 1,600°C criterion, the high-temperature heatup test of irradiated fuel compact was carried out. The result showed that no failed particle was present in the fuel compact after heating. The diffusion coefficient of metallic fission products in SiC layer was also examined in a series of post-irradiation heating tests. The measured diffusion coefficient of ¹³⁷Cs showed a good holding ability as those obtained for research and development fuel specimen. The measured fission gas release rate in accelerated irradiation test showed no additional failure up to 60 GWd/t which was about two times higher than 33 GWd/t of the maximum burnup in the HTTR core. Through the tests, integrity of as-fabricated first-loading fuel of the HTTR was finally confirmed. The future post-irradiation test plan, which will be carried out to confirm the fuel irradiation performance and to obtain the data on its irradiation characteristics in the core, is also described.     

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.     

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.     

碳化硅基新型包覆燃料颗粒的设计及制备

TRISO型包覆燃料颗粒可将核裂变产生的气体、固体裂变产物束缚在燃料颗粒内部,是高温气冷堆安全性的重要保障。为满足未来超高温气冷堆在更高温度及更高燃耗条件下对燃料元件的要求,需对传统TRISO颗粒进行优化和改进。基于包覆颗粒的破损机制,设计了两种SiC基新型包覆颗粒,一种采用疏松SiC层替代疏松热解炭层,包覆层由内而外依次为疏松SiC层、内致密热解炭层、致密SiC层、外致密热解炭层;另一种为全SiC包覆结构,包覆层由内而外依次为内层疏松SiC层、SiC过渡层、外层致密SiC层。根据结构设计,采用流化床化学气相沉积法实验探索了疏松SiC的形成机制及包覆工艺条件,并利用SEM、XRD等进行材料分析,最终成功实现了两种新型包覆颗粒的大规模制备。更进一步,提出了全SiC基燃料元件的概念,并制备了球形和柱形全SiC基模拟燃料元件。     

Derivation of ideal power distribution to minimize the maximum Kernel migration rate for nuclear design of pin-in-block type HTGR

ABSTRACT

Suppressing the kernel migration rates under normal operation condition is quite important from the viewpoint of the fuel integrity for High-Temperature Gas-cooled Reactors. It depends on both the fuel temperature and the fuel temperature gradient. The presence of the ideal axial power distribution to minimize the maximum kernel migration rate allows us to improve the efficiency of design work. Therefore, we propose a new method based on Lagrange multiplier method in consideration of thermohydraulic design in order to obtain the ideal axial power distribution to minimize the maximum kernel migration rate. For one of the existing conceptual designs performed by JAEA, the maximum kernel migration rate for the power distribution to minimize the maximum kernel migration rate proposed in this study is lower by approximately 10% than that for the power distribution as a conventional design target to minimize the maximum fuel temperature.     

包覆燃料颗粒生产线压力平衡分析及系统优化

高温气冷堆使用的TRISO型包覆燃料颗粒采用流化床化学气相沉积法制备。实验研究发现,在包覆燃料颗粒的生产过程中,化学气相沉积过程的压力显著影响包覆燃料颗粒的性能指标。研究了包覆过程中反应压力的变化对包覆层的显微形貌及密度的影响规律。基于对化学气相沉积反应压力的控制需求,进行生产线全系统压力监测及分析,获得整个生产线压力平衡及稳定操作的值。对生产线进行了系统优化,提出从后端调节压力的方法,通过系统的压力平衡,实现了对反应压力的有效控制,保证了生产过程稳定性。研究结果指出,全系统压力平衡分析是一种有效的生产线评估手段,可用于故障分析排除、分级连锁设计以及系统优化等。     

包覆燃料颗粒制备的自动化控制系统设计与研制

针对高温气冷堆包覆燃料颗粒生产逐渐规模化大批量发展的趋势,原制备工艺的手动控制体系已不能适应,需发展现代化工业级别的包覆燃料颗粒制备自动化控制系统。针对TRISO型包覆燃料颗粒4层连续包覆工艺进行分析,将包覆炉系统划分为5个子系统,将整个包覆过程分解为9个操作状态,提出建立分布式控制系统（DCS）自动化控制系统的思路。根据对包覆颗粒制备工艺的严格要求提出了控制系统设计原则,包括连锁控制、安全可靠、集成规范、实用易用、开放和易更新原则,并在具体建设过程中实现了这些原则要求,建立起一套完整的包覆燃料颗粒制备工艺自动化控制系统。该系统在我国高温气冷堆示范工程项目辐照样品的生产中投入运行,经实践检验,证明该系统可较好地实现包覆燃料颗粒制备工艺控制,满足工厂规模的生产要求。     

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。     

包覆颗粒燃料分布随机性的影响研究

高温气冷堆采用弥散在石墨基体中的包覆颗粒燃料。包覆颗粒在燃料球内的离散分布及燃料球在堆芯内的离散分布共同导致了燃料分布的双重不均匀性,其分布还具有随机性,由此可能对反应堆的某些参数造成影响。建立适当燃料颗粒随机分布的几何模型,并用MCNP对模型进行相关计算,并与规则分布模型相比较,分析了分布的随机性对有效增殖因数的影响。结果显示,燃料颗粒随机分布会使全堆有效增殖因数较规则模型的稍大,两种模型偏差的主要原因在于两种颗粒排列方式在空间和角度分布的不同。     

包覆燃料颗粒应力的有限元分析

高温气冷堆的燃料元件由包覆燃料颗粒弥散在石墨基体中组成。在反应堆运行过程中,辐照及各复杂的物理化学反应产生的应力会使包覆燃料颗粒发生破损,对包覆燃料颗粒进行应力分析是评价燃料元件和反应堆运行安全性能的主要内容之一。本文基于压力壳模式,主要考虑内压作用下的球形壳层应力及包覆燃料颗粒的非球形因素,用有限元法对应力进行了分析。     

Release Behavior of Fission Gas from Coated Fuel Particles under Irradiation

TRISO coated fuel particles for HTGR were irradiated by two sweep gas capsules in order to study the release behavior of the fission gas and try to predict the failure fraction of the particles on the basis of the measurement. For verification of the predicted failure fraction, post irradiation examination was conducted, and failure fraction in a visual inspection and acid leaching fraction were measured. Agreement between the predicted failure fraction and the acid leaching fraction was good for these samples except one. From the release behavior from the intact particles, in-pile diffusion coefficients of Kr in LTI-PyC were estimated and expressed as D=(2.9–6.0)×10⁴exp(-2.55×10°/RT) (cm²/s), where R ids the gas constant (=8.314 J/K) and T the absolute temperature. It was recognized that the release from failed particles was controlled by diffusion at 1,600°C and that from intact particles, predominantly by recoil at 1,400°C.     

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.     

Evaluation of Hot Spot Factors for Thermal and Hydraulic Design of HTTR

Abstract

High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950°C in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly.     

HTGR燃料元件包覆颗粒的穿衣主要工艺研究

高温气冷堆的商业化发展对燃料元件穿衣颗粒的规模化生产提出了更高要求。本文采用自主研发的穿衣系统,研究了穿衣鼓转速对动态表面倾角、安息角、粒径分布等参数的影响。确定了穿衣鼓转速的分段设置和最佳状态,并分析了不合格颗粒产生的原因。批量实验结果显示,穿衣颗粒成品率高且稳定,平均成品率达到93-94%,且设备易于操作控制,完全能满足规模生产的需要。     

Application of X-Ray Computer Tomography for Observing the Deflection and Displacement of Fuel Pins in an Assembly Irradiated in FBR

A nondestructive method making use of X-ray computer tomography (X-ray CT) has been applied to post irradiation examination of fast breeder reactor (FBR) fuel assemblies. In the study, an examination is made of the deflection and displacement of fuel pin in a fuel assembly irradiated to 74.2GWd/t peak burnup in the fast reactor “JOYO.”

In the examination, X-ray CT images of transverse cross sections of fuel pin were obtained at different heights of fuel pin along its axis. Analysis of the resulting images indicated that:

1. The hexagonal wrapper tube had its lateral wall faces slightly bulged outward;

2. The fuel pins loaded in the outermost array were markedly displaced in the direction of wrapper tube, particularly in portions of fuel pin intermediate between positions constrained by wrapping wire.

The latter behavior of fuel pins was substantiated by the contours of fuel pin along its axis, which were derived from cross section images obtained at different levels along axis.

Such fuel pin displacement is surmised to have been caused by thermal stressing of the affected fuel assembly cladding.     

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.     

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.     

颗粒尺寸分析仪法测量包覆颗粒疏松层密度的研究

包覆燃料颗粒的质量对于高温气冷堆安全运行起着重要作用。低密度热解炭层作为包覆的第一层非常关键,关系到包覆燃料颗粒和燃料元件的性能质量。本文介绍一种用颗粒尺寸分析仪测量疏松热解炭层密度的方法,该方法采用颗粒尺寸分析仪测量包覆前后颗粒的直径,再结合天平称得包覆前后颗粒的质量,经过计算得到包覆燃料颗粒疏松热解炭层的密度。对该方法测量包覆燃料颗粒疏松层密度的测量精度进行了验证。结果表明,该方法的测量精度满足测试要求,且该方法快速、便捷,适于工程应用。     

ZrCl4蒸汽沉积法制备高温气冷堆包覆燃料颗粒ZrC涂层

采用ZrCl₄蒸汽、H₂和C₃H₆作为化学反应体系,以Ar为载气,在流化床沉积炉中制备高温气冷堆包覆燃料颗粒ZrC涂层。对所制备涂层进行了分析表征,结果显示：ZrC涂层剖面均匀光滑,无明显孔洞;与内致密热解炭层的界面清晰,厚度约为35μm;涂层主要成分为Zr和C,Zr/C摩尔比接近化学计量比1∶1,其主要相组成为面心立方的ZrC;晶粒生长无明显的择优取向。