External sources of neutron provide stable and sufficient neutron for initial startup of a nuclear reactor. They also provide signals for neutron detectors to monitor the safety of reactor during shutdown. In the high temperature engineering test reactor (HTTR), Cf-252 is used as the external neutron source. However, the Cf-252 sources must be renewed every approximately 7 years because of its relatively short half-life of 2.6 years. The renewal of Cf-252 sources requires a high cost and a very complicated procedure. This study investigated the feasibility of using BeO rods as the secondary neutron sources to avoid renewing the Cf-252 neutron sources periodically. The BeO rods could exist in the reactor for a long time so that if the reactor operates long enough, the neutron flux at the wide-range monitoring detectors remains more than 10n.s(-1).cm(-2) even if the reactor is shutdown for as long as 5 years. The results of this study indicated that using BeO rods as the secondary neutron sources would be an attractive option for the future HTGR design with a long-life fuel cycle.

In the conventional study, neutron time series data required for reactor noise analysis have been measured using Single Channel Analyze (SCA) and Multi-Channel Scaler (MCS). However, those method provides only limited information. Currently, there is a new Multi-Channel Analyzer (MCA) with time-stamping capability which can be used to obtain more information. In this study, using the same neutron time-sequence data obtained from a new MCA, prompt neutron decay constants were determined by the conventional variance-to-mean ratio method and the auto correlation function method. The auto correlation function method was one of reactor noise technique and few experiments with the method have been conducted in the past. The experiments were conducted using the Kyoto University Critical Assembly (KUCA). As the results, the decay constants were agreement between the variance-to-mean method and the auto correlation function method within the error range. In addition, the obtained decay constant by the auto-correlation function method had not the detector position dependency.

An initial tritium inventory is required to start fusion DEMOnstration Power Station (DEMO) reactors. However, a method to supply adequate tritium has not been determined yet. Tritium production via 6Li(n,α)T reaction by loading Li rods into the burnable poison (BP) holes of a high-temperature gas-cooled reactor (HTGR) has been proposed to address this problem (Matsuura et al., Nucl. Eng. Des. 243 (2012) 95 - 101). In previous preliminary studies, Li rods loaded in all BP holes were assumed to have the same design. This study evaluated whether the performance of Li rods can be improved for future optimization by adjusting the Li rod arrangement and the amount of Li compounds in them. The amount of tritium produced for gas turbine high-temperature reactor 300 (GTHTR300) was evaluated, while the total amount of Li compounds was maintained and the amount of loaded Li compounds changed depending on the layers and fuel regions. The maximum amount of tritium produced did not increase during the evaluations when reactor feasibility was satisfied. This implies that it is possible to reduce the number of Li rods while maintaining the amount of tritium produced for optimization, thereby reducing the costs of manufacturing Li rods and tritium recovery. GTHTR300 can produce 800 g of tritium in 360 days of operation using 2160Li rods. The results showed that the same amount of tritium could be produced by loading 720Li rods with the same number of fuel blocks. In addition, the effective multiplication factor, burn up, and power density of GTHTR300 were not significantly influenced during the operation.

Recently, we have been proposing a sleeveless SiC matrix fuel compact design to increase the core power density and enhance the safety for High Temperature Gas-cooled Reactor (HTGR). For that application, the oxidation behavior of Reaction-Sintered Silicon Carbide (RS-SiC) under conditions of air ingress accident needs to be investigated. However, few study focused on RS-SiC oxidation, especially under conditions of interest for HTGR. The detailed RS-SiC oxidation kinetics and passive-to-active transition behavior remain unclear. In this work, aiming at feasibility of sleeveless SiC matrix fuel compact for HTGR application, thermogravimetry (TG) measurements were carried out in the conditions of 800 ~ 1400 ℃ and 1 ppm ~ 20 % oxygen concentration under atmospheric pressure, 30 min real-time monitoring. As a result, we have experimentally determined the passive-to-active transition that might be encountered in case of accidents. The present study successfully collected basic data on RS-SiC oxidation, thus, the material feasibility of sleeveless SiC matrix fuel compact for HTGR use was obtained. We partly concluded that the RS-SiC can be applied to HTGR fuel matrix material in the case of air ingress accident.

Feasibility of reprocessing of High Temperature Gas-cooled Reactor (HTGR) spent fuel by existing Plutonium Uranium Redox EXtraction (PUREX) plant and technology has been investigated. The spent fuel dissolved so-lution includes approximately 3 times amount of uranium-235 and 1.5 times amount of protonium because of the 3 times higher burnup compared with that of Light Water Reactor (LWR). Then, the heavy metal of the spent fuel is planned to be diluted to 3.1 times by depleted uranium to satisfy the limitation of Rokkasho Reprocessing Plant (RRP) plant. In the present study, recoverability of uranium and plutonium with the dilution is confirmed by a simulation with a reprocessing process calculation code. Moreover, the case without the dilution from the economic perspective is investigated. As a result, the feasibility is confirmed without the dilution, and it is ex-pected that the reprocessed amount is reduced to 1/3 compared with a diluted case even though the facility should be optimized from the perspective of mass flow and criticality.

The conceptual design of a demonstration hydrogen production facility using heat supply from the high-temperature engineering test reactor (HTTR) is being researched and developed at Japan Atomic Energy Agency (JAEA). This facility produces hydrogen with a thermochemical water-splitting Iodine-Sulphur (IS) process that requires high-temperature heat to extract hydrogen efficiently. The HTTR could supply the heat to the IS demonstration plant through the secondary helium loop, coupling the IS plant to the HTTR.

This work proposes tritium (T) production using a high-temperature gas-cooled reactor (HTGR) and studies it for the initial T inventory in a demonstration fusion reactor and a prior engineering test with T handling. At this stage, the aim is to investigate the compatibility between electricity and T production, which makes the stable confinement of T in the Li-loading rods during the HTGR operation period a crucial issue. The total T outflow into the helium gas during the reactor operation period was attempted to be reduced using Zr spheres with Ni coating as a T absorption material to avoid an increase in the inner T pressure. The basic hydrogen absorption properties of the Zr spheres with Ni coating, which coexist in an environment with Al2O3 and/or LiAlO2 oxides, were measured. The structures of the Li-loading rod for a typical commercial HTGR and irradiation test were studied using the obtained data. In addition, an outline of the irradiation test is reported.

Estimation of decay gamma distribution in a reactor core is essential for safely conducting various works after reactor shutdown such as periodic maintenance, shuffling fuel, removing spent fuel at the end of cycle, etc. Because of the dependency on the complex operating history of the reactor, attempting to calculate the decay gamma rays distribution in the core remains a challenge. This study shows a method to calculate the shutdown gamma distribution in the HTTR core by coupling a Monte-Carlo transport calculation code MCNP6 and an activation code ORIGEN2 to take advantage of spatial dependence and transport abilities of MCNP6 and the detailed fission products tracking during burnup and cooling of ORIGEN2. As result, the three-dimensional shutdown gamma distribution in the HTTR core for different cooling times and spatial locations could be obtained accurately.

An improvement of the electricity generation cost evaluation method for High Temperature Gas-cooled Reactors (HTGRs) has been performed. Japan Atomic Energy Agency (JAEA) had completed the commercial HTGR concept named Gas Turbine High Temperature Reactor (GTHTR300) and the electricity generation cost evaluation method approximately a decade ago. The cost evaluation was developed on the basis of the method of Federation of Electric Power Companies (FEPC). The FEPC method was markedly revised after the Fukushima Daiichi nuclear disaster. Moreover, the escalation of material and labor costs for the decade should be considered to evaluate the latest cost. Therefore, we revised the cost evaluation method for GTHTR300 and the determined cost was compared with that of the Light Water Reactor (LWR). As a result, it was found that the electricity generation cost of HTGR of 7.9 yen/kWh is cheaper than that of LWR of 11.7 yen/kWh by approximately 30% at the capacity factor of 70%.

Tritium is required for research and development activities for the deuterium-tritium (DT) fusion reactor and fueling the DEMOnstration Power Station (DEMO). However, tritium is a very rare nuclide and must be produced artificially. Tritium production by loading Li compounds (Li rods) into burnable poison holes of a high temperature gas-cooled reactor (HTGR) has been proposed (H. Matsuura, et al., Nucl. Eng. Des. 243 (2012) 95-101.). Al2O3 and Zr are used to prevent tritium leaks. Nuclear reaction heat caused by the nuclear reaction (e. g., Li-6(n,alpha)T reaction) can cause a spatial temperature profile in the Li rods and may change its tritium containment performance, because Al2O3 and Zr performance strongly depend on these temperatures. The effect of nuclear reaction heat by the Li-6(n,alpha)T reaction on the tritium containment performance of the Li rods was evaluated by simulation. The temperatures of the Li rods for the high-temperature engineering test reactor (HTTR) and gas turbine high-temperature reactor 300 (GTHTR300) increased by 36 K and 46 K, and the leaked tritium decreased by 32% and 37% via nuclear reaction heat, respectively.

The experience of Fukushima Daiichi nuclear power plant accident caused by the great earthquake that occurred in eastern Japan in 2011 showed the importance of preparing for the loss of function of the engineered safety features. Increasing the strength of equipment to prevent loss of function in an accident is effective, but the possibility of loss of function remains. Therefore, it is important to have an alternative to lost functions in order to put the accident under control early. Thus, this study designed an alternative shutdown system, namely a portable backup shutdown system (PBSS), to make countermeasures in the event of a loss of shutdown function more robust without impairing economic efficiency of the High Temperature Gas-cooled Reactor (HTGR). The PBSS is portable and capable of being installed manually so that it can operate in a total loss of off-site electricity. Various neutron absorber materials for the PBSS were also considered from the viewpoints of technical and cost-effective properties. As results of optimization, the boron nitride (BN) was selected as it shows a good neutronic property as well as a reasonable cost in comparison with other materials.

During operation of the HTTR, hundreds of technical signals and operating conditions must be observed and evaluated to ensure safe operation of the reactor. The accumulated experiment data of the HTTR is not only important for the HTTR operation, but also for the basic development of the HTGR in general. Artificial intelligence (AI) and particularly machine learning (ML) could give the ability to make predictions as well as allow the extraction of key information about physical process from large datasets. Hence, there is a lot of potential to apply AI and ML to predict the operating and safety parameters of the HTTR. In this study, the control rod position of the HTTR is predicted based on ML without using the conventional neutronic codes. The ML with a linear regression algorithm finds a functional relationship between the input dataset and a reference dataset, constructing a function that predicts control rod position from the other operation conditions. As result, the ML gives a good prediction of the HTTR control rod position with less than 5% difference compared to that in the experiment. With increasingly complicated experiments that create a large amount of data, ML is also expected to improve the design and safety analysis of the HTTR in the future.

Kernel migration of TRi-structual ISOtopic (TRISO) fuel for High Temperature Gas-cooled Reactor (HTGR) has been analyzed to investigate the potential dominating effects. Kernel migration is a major fuel failure mode and dominant to determine the lifetime of the fuel for High Temperature engineering Test Reactor (HTTR). However, the result and reliability depend on the evaluation method. Therefore, we propose the recommendation method by taking into account of actual distribution of Coated Fuel Particles (CFPs) and the resulting heterogeneous fuel temperature calculation with such distribution. As a result, the evident and conservative method can be pro-posed, and the method shows that the Kernel Migration Rate (KMR) is predicted to be about 10% less compared with the most conservative evaluation.

Proposed higher power density HTGR by removing the graphite sleeve and implementing the SiC matrix fuel compact has neutronics problem on criticality. Present study selected Er burnable poison and Pu fissile material as possible candidate to compensate the reactivity. To ensure core performance of HTTR modeled reactors, the various tests were carried out using Serpent 2 code, and nuclear characteristic data were obtained successfully. The obtained results verified the expected characteristics. It was demonstrated that both Er and Pu solved criticality issue. Furthermore, especially Pu-loaded core showed better burn-up performance compared to current HTGR. One possible option that proposed reactor can also burn Pu safely based on the concept of small-Clean Burn-HTGR was concluded.

The first criticality of the high temperature engineering test reactor (HTTR) was achieved on November 10, 1998. After improvement of the reactor system and performance confirmation, the rise-to-power test was started in September 1999. The HTTR achieved the full power of 30 MW at a reactor outlet coolant temperature of about 850°C in December 2001, and the reactor outlet coolant temperature of 950°C was achieved in April 2004. After that, various operations such as high temperature operation at 950°C and safety demonstration tests were conducted until December 2010, and the HTTR has been suspended since the Great East Japan Earthquake in 2011. The high temperature operation at 950°C showed that the HTTR has enough performance to supply high temperature heat to a heat utilization system such as a hydrogen production system. Furthermore, the inherent safety of HTGR was confirmed by the safety demonstration tests.

A power distribution monitoring system by using a moving detector for a core with a long neutron flight path has been proposed. High Temperature Gas-cooled Reactor (HTGR) and Fast Reactor (FR) has a long neutron flight path and the neutrons reach to detector far from fuel assembly in the center of the core unlike Light Water Reactor (LWR). By using the feature, power distribution can be observed with a few detectors by moving the detector and computed tomography technology similar to X-ray Computed Tomography (CT). For a small-sized core, the power distribution can be evaluated only by an ex-core neu-tron detector. For a large-sized core with inner detectors, the power distribution can be observed with a small number of in-core detectors even if the deployment is limited due to material integrity conditions such as temperature environment. The feasibility is numerically confirmed by simulations of the HTGR core and its detector response. It is expected to observe the power distribution in the core of HTGR and FR, which is difficult continuously to deploy in-core detectors because of high temperature and/or high irradiation damage.(c) 2021 Elsevier Ltd. All rights reserved.

Manufacturability of the BP-mixed uranium dioxide (UO2) kernel for burnable poison (BP) credit concept was estimated to ensure criticality safety in 14 wt% enrichment fuel fabrication facility for commercial high temperature gas-cooled reactor (HTGR). The case of the coating process has not been examined, so this study is different from manufacturability estimation on Tri-ISOtropic (TRISO) coated fuel. As a BP, boron, gadolinium, erbium, and hafnium were investigated. It is essential that mixing BP does not affect the shape of the fuel kernel, and that BP does not melt or vaporize during the heat treatment in the reduction process. We determined to add BP as a nitrate by considering the existing kernel fabrication process. The results of thermodynamic equilibrium analysis showed boric acid would melt and vaporize during the heat treatment processes. On the other hand, it was found that gadolinium, erbium, and hafnium nitrate would change to solid oxides that do not melt even at 2000 degrees C. These results indicate the existing external sol-gel method could be applied for the fabrication of the UO2 fuel kernel mixed with gadolinium, erbium, or hafnium.

The concept of Pu-burner high temperature gas-cooled reactor (HTGR) has been proposed to more safely reduce the amount of recovered Pu. In the Pu-burner HTGR concept, coated fuel particles with ZrC-coated yttria-stabilized zirconia containing PuO2(PuO2-YSZ) kernels are employed for very high burn-up and high nuclear proliferation resistance. The role of ZrC layer is that of oxygen getter. CeO2-YSZ kernels were fabricated to simulate the PuO2-YSZ kernel and coated with a ZrC layer. In this study, we clarified that both Ce-rich grains and Zr-rich grains were densely distributed in surface regions of the as-fabricated CeO2-YSZ kernel. However, we have already clarified that the surface region of the CeO2-YSZ kernel coated with a ZrC layer was porous and mainly consisted of Zr-rich grains. These experimental results confirmed that Ce-rich grains were selectively corroded during the ZrC coating process. Then, the chemical stability of Zr-rich grains would be higher than that of Pu-rich grains. Thus, it would be more difficult to extract Pu from PuO2-YSZ kernels (in which almost all grains are Zr-rich) than from PuO2-YSZ kernels (in which many Pu-rich grains are included). Influences of the sintering of fuel compact on the microstructure of the ZrC-coated CeO2-YSZ kernel is also reported.