Present status of the High Temperature Engineering Test Reactor (HTTR)

In Japan, the research and development on the High Temperature Gas-cooled Reactors (HTGRs) had been carried out for more than fifteen years since 1969 as the multi-purpose Very High Temperature gas-cooled Reactor (VHTR) program for direct utilization of nuclear process heat such as nuclear steel making. Recently, reflecting the change of the social and energy situation and with less incentives for industries to introduce such in the near future, the JAERI changed the program to a more basic ‘HTTR program’ to establish and upgrade the HTGR technology basis.The HTTR is a test reactor with a thermal output of 30 MW and reactor outlet coolant temperature of 950°C, employing a pin-in-block type fuel block, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. Since 1986 a detailed design has been made, in which major systems and components are determined in line with the HTTR concept, paying essential considerations into the design for achieving the reactor outlet coolant temperature of 950°C. The safety review of the Government started in February 1989. By request of the Science and Technology Agency the Reactor Safety Research Association reviewed the safety evaluation guideline, general design criteria, design code and design guide for the graphite and the high-temperature structure of the HTTR.The installation permit of the HTTR was issued by the Government in November 1990.     

Numerical simulation of three-dimensional thermal-hydraulic behavior for HTTR (High Temperature Engineering Test Reactor)

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are now in progress in order to verify the inherent safety features and to improve safety designing and analysis technologies for future HTGR (high temperature gas-cooled reactor). Coolant flow reduction test is one of the safety demonstration tests for the purpose of demonstration of inherent HTGR safety features in the case that coolant flow is reduced by tripping of helium gas circulators. If reactor core element temperature and core internal structure temperature during abnormal events are estimated by numerical simulation with high-accuracy, developed numerical simulation method can be applied to future HTGR designing efficiently. In the present research, three-dimensional in- and ex-vessel thermal-hydraulic calculations for the HTTR are performed with a commercially available thermal-hydraulic analysis code “STAR-CD^®” with finite volume method. The calculations are performed for normal operation and coolant flow reduction tests of the HTTR. Then calculated temperatures are compared with measured ones obtained in normal operation and coolant flow reduction test. As the result, calculated temperatures are good agreement with measured ones in normal operation and coolant flow reduction test.     

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.     

高通量工程试验堆（HFETR）的维修管理

介绍了高通量工程试验堆运行１８年所形成的维修策略,维修组织机构,维修程序及维修管理工作,系统地研究与总结了维修人员的培训设计及培训。通过对运行年报及维修统计结果的分析,证明了ＨＦＥＴＲ的维修能满足安全运行的需要。     

高通量工程试验堆物理计算方法的研究

研究了高通量工程试验堆堆芯栅元计算模型 ,特别是多层套管燃料组件的计算模型 ;提出了控制棒扩散系数修正方法 ;开发了栅元计算程序WIMS D4 CNPRI和六角形堆芯扩散计算程序CITATION/SIXTUS 2 / 3软件包 ,并成功的将它们用于高通量工程试验堆的物理计算。     

高通量工程试验堆（HFETR）安全运行十周年

HFETR 自1980年12月16日达到满功率起,至今已安全运行十周年。十年来,本堆累计运行22炉,完成辐照试验任务47项,研制并生产了十多种高比活度同位素,获得了大量科技成果;同时,在反应堆运行与维修、堆物理、热工、辐照试验、辐照效应研究、辐射防护、环境监测与评价、放射化学、辅射加工等领域积累了较丰富的经验。运行十年的历史证实,HFETR 设计是成功的,不仅在辐照试验方面具有很大的潜在能力和灵活性,而且反应堆系统和重要安全设备运行安全可靠。目前,HFETR 正处在壮年时期,应加强开发利用,使其为四化,特别是为发展核电做出贡献。     

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.     

基于微机系统的高温气冷堆工程仿真机

基于微机系统的高温气冷堆工程仿真机（HTRSIMU）由清华大学核能技术设计研究院开发完成HTRSIMU运行于Windows98或Windows2000平台上,采用多进程、多显示器结构,具有人机界面友好、结构紧凑、操作方便、易于扩展等特点。它的模型包括10MW高温气冷堆（HTR－10）的一、二回路主要部件,能够对反应堆堆芯、主回路系统和蒸汽发生器等部件做详细的物理和热工分析计算,可以模拟正常运行和各种事故工况过程,仿真结果和蒸汽发生器等部件做详细的物理和热工分析计算,可以模拟正常运行和各种事故工况过程,仿真结果与HTR－10的设计值和安全分析报告符合得很好。利用HTRSIMU系统不仅可以进行高温气冷堆的工程设计、安全分析和人员培训,而且将来可以对HTR－10主控室的操作人员进行现场支持,给实际运行和各项研究提供帮助。     

HFETR堆芯燃料管理计算方法的研究

研究了高通量工程试验堆（ＨＦＥＴＲ）堆芯燃料管理计算方法,以栅元计算程序ＷＩＭＳ－Ｄ４－ＣＮＰＲＩ和三维节块程序ＳＩＸＴＵＳ－３为基础,研制了ＨＦＥＴＲ堆芯燃料管理计算软件包ＨＦＥＴＲＦＭ。并对高通量工程试验堆首炉堆芯进行了计算,取得了令人满意的结果。     

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.     

商用高温气冷堆联合循环方案研究

本文针对高温气冷堆动力转换单元设计了3种联合循环方案,并将3种循环方案在反应堆出口温度900℃的情况下与闭式Brayton循环进行比较。结果表明:闭式Brayton循环在反应堆出口温度较高时,相应反应堆入口温度也较高,这受到反应堆压力壳材料限制,且所需压气机压比较大;联合循环方案的反应堆入口温度低于370℃,反应堆压力壳可使用SA533钢材,无需内壁冷却,且所需压气机压比较小。方案比较显示,提高联合循环效率需增加下位循环出力。方案3的上位循环是简单Brayton循环,下位循环是再热Rankine循环,循环效率可达50.1%。     

10 MW高温气冷堆包覆燃料颗粒辐照考验

TRISO型包覆燃料颗粒由燃料核芯、疏松热解炭层、内致密热解炭层、碳化硅层和外致密热解炭层组成.在冷态性能检验合格的基础上,进行了10 MW高温气冷堆包覆燃料颗粒的静态辐照试验和动态回路辐照试验.在辐照温度1 000 ℃、累积快中子注量1.28×1025 m-2和燃耗(以金属铀计)达到95 GW·d·t-1时,包覆燃料颗粒的放射性裂变产物85Krm的释放率为1.02×10-6,辐照后检验未发现包覆燃料颗粒破损.辐照考验结果表明,包覆燃料颗粒的性能可以满足我国10 MW高温气冷堆安全运行的要求.     

10MW高温气冷堆热气导管高温性能试验

水平布置的同轴双层热气导管在１０ＭＷ在高温气冷实验反应堆中是连接堆芯和蒸汽发生器的重要部件， 外分别流过高温和低温氦气，在氦气工程试验回路上进行了热气导管热工性能试验，使用氦气介质，在３．０ＭＰａ，９５０℃温度下连续运行时间超过９８ｈ，ｄ３．０ＭＰａ７００℃以上温度条件下的热运行时间超过３５０ｈ，还在０．１－３．４ＭＰａ压力范围内进行了２０闪压力循环；在１００－９５０℃范围内，进行了１８次温度循     

可移动式熔盐冷却高温堆(TFHR)热工水力特性不确定性分析

基于不确定分析软件DAKOTA、反应堆热工水力最佳估算程序RELAP5,编写程序耦合接口,对新型20 MWth棱柱式熔盐冷却高温堆稳态热工水力特性进行不确定性分析。选取关键热工参数(如功率、物性、几何尺寸)作为变量输入,基于现有实验堆安全运行经验,指定各参数概率密度分布,经过大量重复性计算,最终得到在95%置信水平下燃料峰值温度的统计分布,进而分析反应堆安全特性。统计结果表明:传热系数和燃料气隙的不确定性对燃料峰值温度影响最为显著且为负相关;燃料峰值温度有0.5%的概率超过燃料稳态运行极限,现有反应堆设计方案需进一步优化。     

10MW高温气冷堆蒸汽安全阀全性能试验

文章介绍10MW高温气冷堆(HTR-10)二回路超压保护系统中的核二级蒸汽安全阀的设计要求、结构特点及性能要求,并对其性能进行了实验验证。实验结果表明：蒸汽安全阀的性能满足设计要求,达到了核规范的标准。     

高通量工程试验堆考验回路功率分布重组

利用精细注量率重组和注量率形状因子的乘积方法对高通量工程试验堆的考验回路和堆芯燃料组件内的功率分布进行重组,并给出全堆芯的热点因子及其出现的时间(燃耗步)、组件位置、轴向位置、径向环位置和环向角度.     

高通量工程试验堆考验回路功率分布重组及堆芯热点因子计算研究

利用精细注量率重组和注量率形状因子的乘积方法对高通量工程试验堆的考验回路和堆芯燃料组件内的功率分布进行重组 ,并给出全堆芯的热点因子及其出现的时间 (燃耗步 )、组件位置、轴向位置、径向环位置和环向角度     

遗传算法在高通量工程试验堆燃料管理优化中的应用

建立了基于遗传算法的高通量工程试验堆(HFETR)堆内燃料管理优化模型。根据HFETR的实际情况，提出了基于组件位置的二进制编码/解码技术。研究了不同选择策略对算法性能的影响，探讨了遗传算法和专家经验的结合，吸收了自适应遗传算法的思想，得到了可直接应用于HFETR的优化换料方案。     

Preparation of High Specific Activity Strontium-89 Solution in High Flux Engineering Test Reactor

⁸⁹SrCl₂ is an important radioactive drug for the bone metastasis. It is included in the new pharmacopoeia in 2015 and has a promising future in the market. Depending on the high flux engineering test reactor(HFETR), the process of preparation of high specific activity⁸⁹SrCl₂ solution by nuclear reaction ⁸⁸Sr(n, γ)⁸⁹Sr was studied. High purity enriched⁸⁸SrCO₃ was used as target material and irradiated for 56 days under the condition of thermal neutron fluence rate about 2×10¹⁴ n•cm^-2•s^-1. After dissolution and filtration, the colorless⁸⁹SrCl2 solution was obtained. The specific activity of⁸⁹SrCl₂ solution was 7.77×10⁹-1.08×10¹⁰ Bq•g^-1, the activity concentration was 3.59×10⁸-1.21×10⁹ Bq•mL^-1, the gamma impurity content was 0.11%-0.14%, the Al impurity content was much lower than 2 μg•mL^-1(activity concentration 7.4×10⁷ Bq• mL^-1).⁸⁹SrCl₂ solution had been tested to meet the requirements of the industry and could be used as raw material for the production of injection. The development of single 7.4×10¹⁰ Bq level preparation device of high purity and high specific activity of ⁸⁹Sr had been finished. This research is important for localization of isotope products.     

高通量工程试验堆堆芯流量分配计算

建立了合理的高通量工程试验堆（HFERT）堆芯流量分配计算模型,并编制了相应的计算程序。应用此程序可以精确计算和分析HFETR堆芯流量分配和温场。经过中国核动力研究设计院提供的实验数据的验证计算表明,本程序计算结果与相应的实验数据符合很好。