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Production of tritium using a high-temperature gas-cooled reactor (HTGR) has been studied for a prior engineering test with tritium handling and for the startup operation of a demonstration fusion reactor. For this purpose, the hydrogen absorption speed of Zr in a Li-loading rod for the reactor operation is experimentally measured, and an analysis model is presented to evaluate the tritium outflow from the Li rod in a high-temperature engineering test reactor (HTTR). On the basis of the presented model, the structure of the Li-loading rod for the demonstration test using the HTTR is proposed.

Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing the amount of recovered Pu. In the Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO2 (PuO2-YSZ) kernel and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. Japan Atomic Energy Agency (JAEA) has carried out ZrC coatings of kernels simulating PuO2-YSZ kernels. Kernels which simulated PuO2-YSZ kernels were YSZ containing CeO2 (CeO2-YSZ) kernels fabricated by Nuclear Fuel Industry, Ltd. (NFI) or commercially available YSZ kernels. Ce was used as simulating element of Pu. In this manuscript, microstructures of ZrC coated CeO2-YSZ or YSZ kernels are reported, and future direction of development of Pu-burner HTGR is discussed. CeO2-YSZ kernel consisted of Ce-rich grains and Zr-rich grains. Observation results indicated CeO2-YSZ kernel was corroded during ZrC deposition process. Then Ce can be extracted from CeO2-YSZ kernels. If Pu can also be extracted from PuO2-YSZ kernels, that PuO2-YSZ kernel is undesirable in view of nuclear proliferation resistance. Fuel design and/or reactor core design of Pu-burner HTGR must be revised, if it is very hard to fabricate PuO2-YSZ kernels, with composition used in the present design of Pu-burner HTGR, and from which Pu is hard to be extracted. Observation results also indicated the possibility that only Ce-rich grains were corroded. If Pu can be extracted only from Pu-rich grains, PuO2-YSZ kernels containing no or only a small amount of Pu-rich grains, should be fabricated. ZrC layer was bound to YSZ kernel, but was completely debonded after heat treatment simulating sintering of fuel compacts. Remarkable grain growth occurred both in ZrC layer and YSZ kernel, after heat treatment. Voids were distributed in ZrC layer both in grain and on grain boundary.

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., 6Li(n,α)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 6Li(n,α)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.

A high-temperature, gas-cooled reactor (HTGR) is proposed as a tritium production device that has the potential to produce a large amount of tritium using the 6Li(n,α)T reaction without major changes to the original reactor core design. In an HTGR design, generally, boron is loaded into the core as a burnable poison to suppress excess reactivity. In this study, lithium is loaded into the HTGR core aiming to produce thermal energy and tritium simultaneously and is loaded instead of boron as a burnable poison. The nuclear characteristics and fuel temperature were analyzed to confirm the nuclear and thermal feasibility of a lithium-loaded HTGR. It was shown that the analysis results satisfied the design requirements and hence the nuclear and thermal feasibility was confirmed for a lithium-loaded HTGR that produces thermal energy and tritium.

Large quantities of tritium are required to start up fusion reactors and conducts engineering tests using tritium for a fusion blanket system. However, tritium is very rare and kg orders of tritium must be produced artificially. Tritium production, by 6Li(n,α)T reaction using a high temperature gas-cooled reactor (HTGR) has been proposed. This method considers the loading of Li rods into burnable poison holes in the HTGR. In this paper, the Li rod suited for use in the High Temperature engineering Test Reactor (HTTR) was designed, and tritium production and leakage from Li-rod capsules were evaluated by adjusting the thicknesses of LiAlO2, alumina, and Zr layers. An irradiation test scenario to be conducted in the HTTR for demonstration of the Li rod’s tritium production and containment performance was presented.

Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of $^{10}$B, and saturate at a level of 1 \% of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 \% of IG-11 shows ignorable values lower than 0.01 \%$\Delta$k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110., 商用ブロック型高温ガス炉のための黒鉛中の不純物の燃焼特性と臨界性に関する研究を行った。高温ガス炉では、炉内が黒鉛構造で満たされるため、黒鉛中の不純物による臨界性に対する毒作用が無視できない。そこで、商用高温ガス炉であるGTHTR300は高純度黒鉛材料であるIG-110を臨界性の観点から、燃料ブロックに隣接する反射体に用いている。燃料ブロックでもIG-110が用いられるべきであるが、経済性の観点から低純度黒鉛材料であるIG-11を用いている。本研究では、高純度黒鉛材料の商用高温ガス炉に対する必要性を不純物の燃焼特性と臨界性を評価することにより再検討する。不純物の毒作用はホウ素等量であらわされるが、この値は$^{10}$Bの燃焼のように、指数的に減少し、初期値の1\%程度のレベルで飽和する。一方で、IG-11のホウ素等量の1\%に相当する0.03ppmの臨界性は燃料ブロック及び反射体ブロックに装荷した場合の両方において、0.01\%$\Delta$k/kk'以下であり無視できる。不純物は天然ホウ素で代表しても問題はない。そこで、不純物の毒作用を全炉心燃焼計算で行った。その結果、商用高温ガス炉に対しては、不純物の臨界性に対する影響は問題にならないことが分かった。なぜなら、IG-11を用いた場合でもサイクル末期までに十分に燃焼するためである。この結果により、IG-110を排除することにより、より経済的な発電が期待できる。

The production of tritium (T) using high-temperature gas-cooled reactors (HTGRs) has been studied for a prior engineering research with T handling and initial T possession in demonstration fusion reactors. Stable containment of T in Li-loading rods during HTGR operation is a critical issue. This study investigates this for an irradiation test to examine T-containment performance in Li-loading rods and develops an analytical model of evaluating the amount of T outflow to a He coolant. The hydrogen absorption characteristics, including the deterioration of the hydrogen absorption speed after Zr has sufficiently absorbed the hydrogen, is experimentally measured assuming an HTGR setting. We present an analytical model of evaluating the T outflow from a Li rod and, on the basis of this model, estimate the total T outflow, assuming the presence of a gas-turbine high-temperature reactor of 300 MWe with a nominal capacity and a high-temperature engineering test reactor. It is demonstrated that, by loading a sufficient amount of Zr into the Li rod, the T outflow can be suppressed to less than a small percent of the total T produced during 360 days of reactor operation.

At high power operation of the HTTR, the control rod should be kept at the top of the active core for maintaining the optimized power distribution so as to minimize the maximum fuel temperature. It is important to calculate the control rod position each time the operating conditions change in order to ensure the safe operation of the reactor. Since the Monte-Carlo code cannot change the core geometry such as control rod position during criticality and burnup calculation, the critical control rod position was determined by adjusting the control rods manually at each burnup step. This complicates the calculation procedures as well as prolongs the total time including calculation time, handling time, and waiting time. Therefore, this study develops a new utility tool that seeks the control rod position automatically without any further handling procedures and waiting time. As a result, the determination of critical control rod position becomes simpler and the total time was also reduced significantly from about 5 days to less than 2 days. The calculated critical control rod position using the new tool also gives a good agreement with the experiment data.

Performance of tritium production for fusion reactors, using a high-temperature gas-cooled reactor (HTGR) is examined. From the viewpoints of tritium recovery and environmental safety, tritium outflow from Li rods should be suppressed to the same level as the liquid radioactive waste from the pressurized water reactors (PWRs) in Japan. Methods for suppressing tritium leakage from Li rods are studied. The tritium outflow is reevaluated accurately on the basis of non-equilibrium simulations and the influence of coolant temperature on tritium leakage is clarified. The approach using Zr in the Li rod to reduce the tritium pressure and the resulting suppression of tritium leakage are also investigated.The results of the non-equilibrium simulation show that the tritium outflow is approximately 40% lower than the outflow reported in a previous study. Although the electric power generation efficiency is reduced, lowering the coolant temperature to 600 K results in a reduction of the tritium outflow to ~1/...

This work reports the characteristics and tribological properties of both Ag/DLC nanocomposite coatings (RF-Ag-DLC) and Cu/DLC nanocomposite coatings (RF-Cu-DLC) with hydrogen-free DLC matrix deposited by RF magnetron sputtering using a concentric composite target (CCT). The CCT consisted of a C base target and metal tablet, and the tablet was located on the center of the base target concentrically where the etching rate by Ar ions is extremely low. By changing the diameter of Ag or Cu tablets in CCT, RF-Ag-DLC with an Ag concentration ranging from 6 to 65 at.% and RF-Cu-DLC with Cu concentration ranging from 7 to 75 at.% can be prepared. These coatings show a granular structure having Ag or Cu nano-crystals with a diameter ranging from 5 to 10 nm dispersed homogeneously in the hydrogen-free DLC matrix. The friction coefficient of DLC varied depending on the species and content of metal. The transition of the friction coefficient became stable when metal-rich tribofilms formed on the counterfaces.

In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system “GTHTR300”, which is the subject of our previous research, a small modular HTGR plant “HTR50S” and two types of plutonium burner HTGRs “Clean Burn with MA” and “Clean Burn without MA” were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.

Conceptual design study of a commercial High Temperature Gas-cooled Reactor (HTGR) for early introduction has been performed based on the cumulated experience in design, construction, and operation of the High Temperature engineering Test Reactor (HTTR) and design of the commercial Gas Turbine High Temperature Reactor 300 (GTHTR300). The power output is 165 MWt and the inlet and outlet coolant temperatures are 325$^{\circ}$C and 750$^{\circ}$C, respectively, to provide steam for industrial utilization. However, given a requirement for the reactor pressure vessel to be smaller even that of the 30 MWt HTTR, several challenging technical problems have to be dealt with to arrive in a high performance core design that provides extended fuel burnup, prolonged refueling period, improved fuel refueling scheme, improved fuel element and so on from the HTTR., 早期導入を目的とした商用高温ガス炉の概念設計をこれまで積み上げてきたHTTRの設計、建設運転の経験及び、商用炉設計であるGTHTR300の設計の経験に基づき実施した。熱出力は165MWtであり、入り口出口の冷却材温度は325$^{\circ}$Cおよび750$^{\circ}$Cであり、工業用蒸気を供給する。しかしながら、設計要求として炉心出力30MWtのHTTRよりも小さな圧力容器を用いなければならず、高性能な炉心設計を実現するためには、HTTRよりの燃焼度増加、長い燃料交換機関、改良された燃料交換法、燃料要素など挑戦的な技術課題に取り組む必要があった。

The investigation on the erbium loading method to improve reactivity coefficients for Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR) is performed. The fuel employs HEU to reduce toxicity generation from uranium-238. The reactivity coefficients show positive values without any additive. Then, the erbium is loaded in the core to obtain negative reactivity coefficient due to the large resonance peak of neutron capture reaction of erbium-167. The loading methods are investigated. The erbium is mixed into fuel kernel of CPF, loaded by binary packing with fuel particle and erbium particle, and embedded into the graphite shaft deployed center of fuel compact. It is found that the erbium loading causes negative reactivity as a moderator temperature reactivity, and it should be loaded into fuel pin elements for pin-in-block type fuel from the viewpoint of heat transfer. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficient even at EOC. The loading method which effectively causes self-shielding should be selected to avoid to be incinerated with burn-up. The mechanism is elucidated by application of Bondarenko approach. As a result, it is conclude that the erbium embedded into graphite shaft is preferable for LRSF-HTGR to remain the reactivity coefficient negative at EOC., 潜在的有害度低減高温ガス炉の反応度係数を改善するためのエルビウム装荷法の検討を行った。本炉心では、U-238からの有害度発生を低減するために高濃縮ウランを燃料として用いる。対策がなされない状態では正の反応度係数を持ちうる。そこで、負の反応度係数を得るためエルビウムを炉心に装荷する。Er-167の巨大な共鳴捕獲断面積ピークにより反応度係数の改善が可能である。本研究では、その装荷法について検討した。装荷法としては、被覆粒子燃料の燃料核への混合、燃料粒子・エルビウム粒子の二粒子による装荷、燃料コンパクトを支持する黒鉛心棒への埋め込みを検討した。検討の結果として、エルビウムは減速材温度係数としての働きが主であり、熱輸送の観点から、燃料棒要素への装荷が適していることがわかった。さらに、燃焼末期で負の反応度係数を得るためには、エルビウムの燃焼速度は遅い必要がある。これらのことを考慮して、自己遮蔽効果が強くエルビウムの燃焼を避けられる装荷法が適していると言え、その自己遮蔽効果発生のメカニズムはボンダレンコアプローチにより確認した。その結果、黒鉛心棒へのエルビウム装荷が潜在的有害度低減高温ガス炉の反応度係数の改善には適しており、燃焼末期においても負の反応度係数が得られることがわかった。

Tritium production utilizing nuclear reactions by neutron and lithium in a high-temperature gas-cooled reactor is attractive for development of a fusion reactor. From viewpoints of tritium safety a...

Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al 2 O 3 cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300 MW electrical nominal capacity (GTHTR300) with 600 MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600 g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

To start up a fusion reactor, it is necessary to provide a sufficient amount of tritium from an external device. The fusion DEMO reactor is planned to start up in the 2030s. Herein, methods for supplying the reactor with tritium are discussed. For the initial startup of the fusion reactor, use of a high temperature gas-cooled reactor (HTGR) as a tritium production device has been proposed. So far, the analyses have been focused only on the operation in which fuel is exchanged at stated periods (batch) using the block type HTGR. In this paper, to improve the performance of tritium production, properties of the HTGR are studied from the viewpoint of continuous operation for several conditions. In continuous operation, for example, in the pebble bed type HTGR, it is possible to design an operation that has no time loss for refueling. The pebble bed modular reactor (PBMR) and the gas turbine high temperature reactor of 300 MWe nominal capacity (GTHTR300) are assumed as the calculation and comparison targets, ...

JAEA has carried out research and development to establish the technical basis of high temperature gas cooled reactors (HTGRs) using HTTR. In order to connect hydrogen production system to HTTR, it is necessary to ensure the stability of plant dynamics when the thermal-load of the system is lost. Thermal-load fluctuation test is planned to demonstrate the stable reactor dynamics and to gain the test data for validation of the plant dynamics code. It will be confirmed that the reactor become stable state during a part of removed heat at HTTR heat-sink is lost. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations for safety analysis, and changes with burnup because of variance of fuel compositions. Measurement of temperature coefficient of reactivity has been conducted by HTTR to confirm the validity of the calculated temperature coefficient of reactivity. A loss of forced cooling (LOFC) test using HTTR has been carried out to verify the inherent safety of HTGR under the condition of loss of forced cooling while the reactor shut-down system disabled.

Japan Atomic Energy Agency (JAEA) has started a conceptual design study of a small-sized High Temperature Gas-cooled Reactor (HTGR) with 50 MW thermal power (HTR50S) to be deployed in developing countries in the 2020s. The nuclear design of the HTR50S is performed by upgrading that of a High Temperature Engineering Test Reactor (HTTR), which is the Japanese HTGR with 30 MW thermal power. In the HTTR design, 12 kinds of fuel enrichment were used to optimize the power distribution. In the previous study of the HTR50S, we succeeded in reducing the number of the fuel enrichment to 3. The present study challenges the nuclear design for effective use of uranium by utilizing high burn-up fuel and axial fuel shuffling, in which a half of the loaded fuel elements is discharged from the core every 2 years and the remains are reloaded. The core burn-up calculations were performed and the nuclear characteristics were confirmed to satisfy the design requirement.

The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

The high temperature engineering test reactor (HTTR) constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR is a graphite-moderated and helium-gas-cooled reactor. The reactor thermal power is 30 MW, and the reactor maximum outlet coolant temperature is 850 °C in rated operation mode and 950 °C in high-temperature test operation mode. The main objectives of the HTTR are to establish and develop HTGR technology and to demonstrate process heat application. The HTTR was operated for 30 days in rated operation mode and 50 days in high-temperature operation mode to obtain various characteristic data for HTGRs. The main test results are as follows: (1) The coated fuel particles (CFPs) of the HTTR are excellent at confining the fission product and have the highest performance in the world. (2) The measured temperature of the core internals is in good agreement with the design value, which means that they will maintain their structural integrity. (3) The intermediate heat exchanger maintains excellent heat transfer performance from the beginning of operation. In addition, the following two issues were validated using the HTTR burnup data: (1) the effectiveness of rod-type burnable poisons at reactivity control in the HTTR and (2) the whole core burnup calculation method for the nuclear characteristics of the HTTR.

The performance of a high-temperature gas-cooled reactor as a tritium production device is examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target of a typical gas-cooled reactor, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the 3-dimensional entire-core region of GTHTR300 were carried out considering its unique double heterogeneity structure. It is shown that gas-cooled reactors with thermal output power of 3 GW in all can produce 5–8 kg of tritium in a year.

The performance of a high-temperature gas-cooled reactor as a tritium production device was examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) was assumed as th...

This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations.