[en] Japan Atomic Energy Agency (JAEA) has High Temperature Engineering Test Reactor (HTTR) which is 30MW test reactor of HTGR. Although the HTTR was not operated after the Fukushima Daiichi accident in 2011, it restarted operation in July 2021. The HTTR can be used to carry out various tests about safety features of HTGR. In 2010, safety demonstration test about a loss of forced cooling (LOFC) condition was carried out under a framework of OECD/NEA. Two LOFC tests with different conditions will be done in early 2022. The LOFC project is an international joint project. After the LOFC, JAEA is planning to launch a new test project. Test theme concerning Xe build-up/decay behavior after reactor scram is one of the attractive new test themes, but not limited to this. This paper presents possible test themes by using the HTTR. JAEA has room to discuss possible new themes with expecting foreign partners to launch an international joint project. (author)

[en] The High Temperature Engineering Test Reactor (HTTR) can provide very large irradiation spaces at high temperatures for various irradiation tests. The first irradiation test rig for the HTTR, the I-I type irradiation equipment, was developed for an in-pile creep test on a stainless steel with large standard size specimens. The equipment uses the ambient high temperature of the core for the irradiation temperature control. The target irradiation temperatures are 550 and 600 deg. C with the target temperature deviation of ±3 deg. C. In this study, the specimen temperature stability at the irradiation test was assessed by both analytical and experimental approaches. The irradiation temperature changes at transient conditions were analyzed by a finite element method (FEM) code and the temperature controllability of the equipment was examined by a mockup test. The controllability was evaluated with the measured temperature transient data at the core graphite components in the Rise-to-Power tests of the HTTR. The result indicates that the temperature control method of the I-I type irradiation equipment is effective to keep the irradiation temperature stable in the irradiation test.

No abstract available

[en] Worldwide attention has been paid to GaAs and SiC as a kind of the heat-resisting and advanced semiconductor materials. Doping of specific impurities into semiconductor materials is a key technology for producing semiconductor devices. As one of doping technologies, thermal diffusional doping has been successfully applied to Si. Application of thermal diffusional doping to GaAs needs expensive and complicated processes in order to prevent an occurrence of lattice defects by impurities. On the other hand, it is difficult to apply this technology to SiC, because of small diffusion coefficient of impurities to SiC. Therefore, it is of great importance to develop a substitutional doping technology for these materials for realizing heat-resisting and advanced semiconductor. To dope some impurities into Si crystal by using neutron irradiation is a mature technology and is called NTD. The High Temperature engineering Test Reactor(HTTR) has an unique and superior capability to irradiate large-sized specimen, in the order of 10cm in diameter, at high temperature up to approximately 1000degC. This report presents a result of feasibility study of potential applicability and effectiveness of NTD to GaAs and SiC at the HTTR. First of all, advantages and disadvantages were identified by reviewing the state-of-the-art technology of NTD to Si. Potential applicability of NTD to GaAs and SiC are discussed. Based on this discussion, effectiveness and feasibility of NTD to these materials at the HTTR are examined. As a result, NTD is feasible to SiC but not to GaAs. The HTTR provides the capability to produce SiC semiconductor, in particular, to produce the semiconductor with (1) low irradiation damage, (2) uniform distribution of doped impurities and (3) high productivity, if a large-sized SiC crystal is capable to be commercialized. Practical application of NTD at the HTTR will be discussed in the next study. (author)

[en] To maintain the structural integrity of graphite components during plant operation a visual inspection using a TV camera as an in-service inspection is planned in the High Temperature Engineering Test Reactor. In order to verify the in-service inspection method a preliminary analytical and experimental studies were performed. In the analytical study the harmful flaw size was determined from a viewpoint of structural integrity based on the fracture mechanics approach. Furthermore, the visible flaw size was determined by the TV camera performance test with graphite test specimens having several kinds of artificial flaws. This paper presents the analytical result on the harmful flaw size and the experimental result on the visible flaw size. From both results the applicability on the visual inspection by the TV camera as the in-service inspection is discussed in this paper. (author)

[en] Oxidation damage is one of the crucial factors to degrade mechanical properties of graphite components in the HTGRs. The oxidation increases the porosity of graphite and, hence, results in degradation. In order to evaluate the oxidation damage at neutron irradiated conditions, a new analytical method by ultrasonic wave propagation characteristics was developed. Irradiation effects, a dimensional change and a pinning of dislocations in crystals, on the propagation characteristics in graphite are taken into consideration in the method. It was shown that an equivalent velocity of the wave in graphite is increased by the irradiation, and that a signal height of a propagated waveform is increased by the irradiation, and it decreases with increasing porosity caused by the oxidation. The Young's modulus for an ideal graphite polycrystals without pore was evaluated by considering the wave velocity in them in order to evaluate the change of the apparent modulus at simultaneous irradiated and oxidized conditions as an application of the developed method. It was also shown that the oxidation-induced change of the modulus is appropriately evaluated by the method, suggesting that it is possible to evaluate the change for the irradiated conditions. It can be said from this study that the developed method is promising to evaluate the oxidation damage on graphite components in the HTGRs by nondestructive way. (author)

[en] An HTGR-hydrogen production system should be designed to have stable controllability because of a large difference in thermal dynamics between reactor and hydrogen production system and such a control design concept should be universally applicable to a variety of hydrogen production processes by the use of nuclear heat from HTGR. A transient response analysis of an HTGR-steam reforming hydrogen production system showed that a steam generator installed in a helium circuit for cooling the nuclear reactor provides stable controllability of the total system, resulting in avoiding a reactor scram. A survey of control design-related characteristics among several hydrogen production processes revealed the similarity of endothermic chemical reactions by the use of high temperature heat and that steam is required as a reactant of the endothermic reaction or for preheating a reactant. Based on these findings, a system design concept with stable controllability and universal applicability was proposed to install a steam generator as a downstream cooler of an endothermic reactor in the helium circuit of an HTGR-hydrogen production system. (author)

[en] One of the safety issues to couple a hydrogen production system to an HTGR is how the reactor coolability can be maintained against anticipated abnormal reduction of heat removal (thermal disturbance) of the hydrogen production system. Since such a thermal disturbance is thought to frequently occur, it is desired against the thermal disturbance to keep reactor coolability by means other than reactor scram. Also, it is thought that the development of a passive cooling system for such a thermal disturbance will be necessary from a public acceptance point of view in a future HTGR-hydrogen production system. We propose a SG as the passive cooling system which can keep the reactor coolability during a thermal disturbance of a hydrogen production system. This paper describes the proposed steam generator (SG) for the HTGR-hydrogen production system and a result of transient thermal-hydraulic analysis of the total system, showing availability of the SG against a thermal disturbance of the hydrogen production system in case of the HTTR-steam reforming hydrogen production system. (author)

[en] In High Temperature Gas-cooled Reactor (HTGR), graphite materials are used as core internal structural components. The neutron irradiation and thermal gradient induce residual stress of graphite components which is a crucial factor to determine the lifetime of them. It is hence important to measure and assess the stress for lifetime extension of the components. Since the residual stress gives characteristics change to the micro-indentation behavior, it is possible to evaluate the residual stress by means of measurement of indentation depth. Therefore, in order to evaluate the change of residual stress of graphite components non-destructively, we are now developing the indentation method. This report showed the relationship between indentation depth and residual stress based on experimental data obtained by changing the stress condition of graphite specimen parametrically. (author)

[en] Top priority objective of development of the first heat utilization system to be connected to the HTTR is the demonstration of technical feasibility of a nuclear process heat utilization system for the first time in the world. Key features of a nuclear process heat utilization system significantly differing from a conventional fossil-fired process heat utilization system are 1) the heat source of nuclear reactor, 2) heat transferring fluid of pressurized helium and 3) a closed system helium circuit. Major issues to be demonstrated at the HTTR heat utilization system are, therefore, 1) the safety as the nuclear reactor system, 2) helium-heated components and 3) a control system for the closed circuit including start-up and shutdown control system. A steam reforming system was selected as a primary candidate for design of the first HTTR heat utilization system due to the following reasons: - Basic system arrangement and heat of endothermic chemical reactions occurred in steam reformer are similar. - A helium-heated steam reformer has been basically developed. Successful operation of the HTTR steam reforming system will demonstrate the feasibility of control and safety design concepts that are applicable to major nuclear process heat chemical systems. Major design achievements up to now are as follows: (1) Its main system arrangement has been optimized to effectively utilize the nuclear heat. (2) Proposed design concepts of steam reformer will improve the heat utilization efficiency to approximately 80%, comparable to a conventional fossil-fired reformer. (3) Any procedure for start-up and shutdown of steam reforming system has been found to be applicable due to a large heat sink capacity of steam generator installed downstream of steam reformer. (author)