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.     

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.     

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.     

Achievement of Reactor-Outlet Coolant Temperature of 950°C in HTTR

A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high-temperature helium gas and to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved its rated thermal power of 30 MW and reactor-outlet coolant temperature of 950°C on 19 April 2004. During the high-temperature test operation which is the final phase of the rise-to-power tests, reactor characteristics and reactor performance were confirmed, and reactor operations were monitored to demonstrate the safety and stability of operation. The reactor-outlet coolant temperature of 950°C makes it possible to extend high-temperature gas-cooled reactor use beyond the field of electric power. Also, highly effective power generation with a high-temperature gas turbine becomes possible, as does hydrogen production from water. The achievement of 950°C will be a major contribution to the actualization of producing hydrogen from water using the high-temperature gas-cooled reactors. This report describes the results of the high-temperature test operation of the HTTR.     

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.     

Multiobjective Optimization for Expansion Planning of Rad-Waste Management System

In order to investigate fuel behavior under high burnup irradiation condition of high temperature gas-cooled reactor (HTGR), an irradiation test was performed. An irradiation was carried out as a part of a cooperative effort between the US DOE and the Japan Atomic Energy Research Institute. The fuel for the irradiation test was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR). In order to keep fuel integrity up to high burnup over 5%FIMA (% fission per initial metallic atom), thickness of buffer and SiC layers of fuel particle were increased. This report describes the fuel behavior under high burnup condition in the irradiation test.     

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.     

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.     

Safety Analysis of Abnormal Reactivity Events in HTTR

Abstract

To confirm the safety of the High Temperature Engineering Test Reactor (HTTR) facility which is being constructed as the first high temperature gas cooled reactor in Japan, the representative abnormal reactivity events assumed in the safety analysis of the HTTR were analyzed. The HTTR is a graphite moderated and He-gas-cooled reactor with thermal power of 30 MW, inlet coolant temperature of 395°C and outlet coolant temperature of 950°C.

This report presents the analytical results of two representative events, “Abnormal control rod withdrawal from a subcritical condition” and “Abnormal control rod withdrawal during the full power operation”, showing that the safety of the HTTR is secured in conformity with the unique features of the HTTR with respect to the maximum fuel temperature, which is a key factor for the safety criteria.

The results of the safety analysis could demonstrate the safety of the HTTR facility with respect to abnormal reactivity events postulated in the HTTR, showing that the maximum fuel temperature is lower than the limit of the maximum fuel temperature of 1,600°C.     

Reactor core design of Gas Turbine High Temperature Reactor 300

Japan Atomic Energy Research Institute (JAERI) has been designing Japan’s original gas turbine high temperature reactor, Gas Turbine High Temperature Reactor 300 (GTHTR300). The greatly simplified design based on salient features of the High Temperature Gas-cooled Reactor (HTGR) with a closed helium gas turbine enables the GTHTR300 a highly efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the High Temperature Engineering Test Reactor (HTTR) and existing fossil fired gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original design features of this system are the reactor core design based on a newly proposed refueling scheme named sandwich shuffling, conventional steel material usage for a reactor pressure vessel (RPV), an innovative coolant flow scheme and a horizontally installed gas turbine unit. The GTHTR300 can be continuously operated without the refueling for 2 years. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200,000 yen (1667 US$)/kW e, and the electric generation cost is close to a target cost of 4 yen (3.3 US cents)/kW h.

This paper describes the original design features focusing on the reactor core design and the in-core structure design, including the innovative coolant flow scheme for cooling the RPV. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.     

Present status of energy in Japan and HTTR project

An amount of primary energy supply in Japan is increasing year by year. Much energy such as oil, coal and natural gas is imported so that the self-sufficiency ratio in Japan is only 20% even if including nuclear energy. An amount of energy consumption is also increasing especially in commercial and resident sector and transport sector. As a result, a large amount of greenhouse gas was emitted into the environment. Nuclear energy plays the important role in energy supply in Japan.Japan Atomic Energy Research Institute (JAERI) has been carried out research and development of a hydrogen production system using a high temperature gas cooled reactor (HTGR). The HTTR project aims at the establishment of the HTGR hydrogen production system. Reactor technology of the HTGR, hydrogen production technology with thermochemical water splitting process and system integration technology between the HTGR and a hydrogen production plant are developed in the HTTR project.     

高温和超高温气冷堆动力转换方案研究

动力转换单元是高温和超高温气冷堆的重要组成部分。本文对高温和超高温气冷堆的动力转换单元进行研究。从4个关键参数（反应堆出口温度、反应堆入口温度、压缩比和主蒸汽参数）入手,对5个循环方案进行比较分析。综合考虑各种工程因素,上位循环为简单氦气透平循环、下位循环为有再热的蒸汽轮机循环的联合循环方案是具有竞争力的,其中下位循环在高温气冷堆范围是亚临界参数循环,在超高温气冷堆范围是超临界参数循环。联合循环可实现高温和超高温气冷堆热量的高效率转化,且反应堆入口温度在反应堆压力壳材料允许的范围内,具有足够的安全性。     

Development of control technology for HTTR hydrogen production system with mock-up test facility: System controllability test for loss of chemical reaction

The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data.     

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.     

Research and development on nuclear hydrogen production using HTGR at JAERI

JAERI has been conducting R&D on HTGR and on hydrogen production using HTGR. The reactor technology has been developed using HTTR installed at Oarai site of JAERI. HTTR reached its full power operation of 30MW in 2001 and demonstrated reactor outlet helium temperature of 950°C in April 2004. As for the hydrogen production technology, the thermo-chemical IS process is under study. The process control method for continuous hydrogen production has been examined using a bench-scale apparatus. Also, studies are underway on process improvement and on materials of construction to be used in the corrosive environment. As for the system integration of HTGR and the hydrogen production plant, R&D is underway aiming to develop technologies for safe and economical connection. It covers safety technology against explosion, safety technology against radioactive materials release, control technology to prevent the thermal disturbance from hydrogen production plant to reactor, etc.     

Interpolation and Extrapolation Method to Analyze Irradiation-Induced Dimensional Change Data of Graphite for Design of Core Components in Very High Temperature Reactor (VHTR)

Graphite materials are used as core components in the High-Temperature Gas-Cooled Reactor (HTGR) and Very High Temperature Reactor (VHTR). The authors prepared technical documents for design, material, products, in-service inspection and maintenance of the graphite components for the HTGR/VHTR, which were summarized as a draft of standard for the graphite components through discussion made in a “Special committee on research on preparation for codes for graphite components in HTGR” set up within AESJ. The draft of standard contains graphical expressions for the irradiated material properties of IG-110 graphite. It is possible to use the graphical expressions for the components design of VHTR. The graphs were obtained based on the interpolation and extrapolation of the irradiation data. The irradiation-induced dimensional change of IG-110 graphite was obtained through the interpolation and extrapolation of the irradiation data with a quadratic equation of fast neutron fluence. The irradiation data for H-451 and ATR-2E graphites were used for the evaluation of the interpolation and extrapolation of irradiation data for IG-110. It was shown in this study that the proposed interpolation and extrapolation method is reasonable for IG-110 with regard to the database available at present.     

HTR-10燃料元件装卸系统冷态调试

10MW高温气冷实验堆（HTR－10）是一座球床堆，由燃料元件装卸系统实现燃料元件的装卸和循环，且不需要停堆，为保证HTR－10的正常运行，燃料元件装卸系统必须安全，可靠，为此，必须对燃料元件装卸系统进行周密，细致的调试试验和验证，本文介绍了燃料元件装卸系统冷调试的主要调试项目，调试方法和调试结果。     

世界核电发展趋势与高温气冷堆

核能的发展面临经济竞争力、核安全、核废物的最终处置及防止核武器材料扩散的挑战。为改善公众的可接受性 ,核电厂的安全性进一步改进。电力市场体制的非管制化改革加剧了电力技术的竞争。环境保护意识增强使核废物的处置倍受关注。 80年代中期以来发展的先进轻水堆核电厂如ABWR ,System 80 ,EPR ,AP60 0等是今后一段时期内商用核电的主力堆型。进入 2 0 0 0年之际 ,美国能源部正在规划发展第四代先进核能系统 ,目标是在 2 0 2 0年或之前 ,向市场提供经过验证的成熟的第四代核电厂技术 ,以替代美国退役的核电容量。球床高温气冷堆被认为是第四代先进核能系统的优选技术。南非ESKOM电力公司选择了球床高温气冷堆作为今后核电发展的堆型。清华大学承担设计和建设的 10MW高温气冷实验堆计划在 2 0 0 0年内临界。通过10MW高温气冷堆的建造 ,我国已形成了高温气冷堆技术的自主知识产权 ,初步具备了自主设计、制造和建造的能力     

Core Burnup Calculation of Plutonium Burning Inert-Matrix Fueled High Temperature Gas Cooled Reactor

For the efficient reduction of excess plutonium amount, Japan Atomic Energy Research Institute (JAERI, now Japan Atomic Energy Agency) has studied a concept of rock-like oxide (ROX) fuel as a kind of inert matrix fuel (IMF). In the JAERI study, ROX fuel is burnt in existing light water reactors (LWRs), while in this study, pebble bed type high temperature gas cooled reactor (HTGR) is studied, mainly because of its high neutron economy and softer neutron spectrum than LWRs. Here, PuO₂-yttria stabilized zirconia (YSZ: (Zr,Y)O_2-x) particles are dispersed in graphite matrix. In the ROX fueled LWR study, it was necessary to improve fuel temperature reactivity coefficients by adding such additives as ²³⁸U and Er. Here in HTGR, although the negative temperature coefficient is much larger than that in LWR without any improvements, temperature coefficient was improved as large as possible to the level of UO₂ HTGR case by adding Er in the fuel. Burnup calculations on PuO_2-YSZ fueled HTGR, by simulating the continuous refueling of fuel pebbles with the batch fuel loading, showed almost complete transmutation for ²³⁹Pu and more than 80% for the total plutonium. As the maximum power density of the fuel pebble obtained by the core burnup calculation was very large in comparison with the UO₂ HTGR, the maximum temperature in YSZ fuel particle was also evaluated. Despite the low thermal conductivity of YSZ, the evaluated YSZ temperature was well below the melting point, thanks to the high thermal conductivity of graphite and small YSZ particle size. Here, the high power density in the Pu-YSZ fueled core might become a problem, but is possible to be reduced by adjusting the initial plutonium enrichment.     

Experimental validation of effectiveness of rod-type burnable poisons on reactivity control in HTTR

In block-type high temperature gas-cooled reactors (HTGRs), insertion depth of control rods (CRs) into a core should be retained shallow to keep fuel temperature below 1495 °C through a burnup period, and hence excess reactivity should be reduced through a different method. Loading burnable poisons (BPs) into the core is considered as a method to resolve this problem as in case of light water reactors (LWRs). Effectiveness of BPs on reactivity control in LWRs has been validated by experimental data, however, this has not been done yet for HTGRs, because there was not enough burnup characteristics data for HTGRs required for the validation. The High Temperature Engineering Test Reactor (HTTR) is a block-type HTGRs and it adopts rod-type BPs to control reactivity. The HTTR has been operated up to middle burnup, and thereby the experimental data was expected to show effect of the BPs on the reactivity control. Hence, in order to validate effectiveness of rod-type BPs on reactivity control in the HTTR, we investigated on the HTTR results whether the BPs have functioned as designed. As a result, the CRs insertion depth has been retained shallow within allowable range, and then effectiveness of rod-type BPs on reactivity control in the HTTR was validated.