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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.

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 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...

Suppressing the kernel migration rates under normal operation condition is quite important from the viewpoint of the fuel integrity for High-Temperature Gas-cooled Reactors. It depends on both the ...

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 °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.

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.

The following design details of the high temperature engineering test reactor (HTTR) are described: nuclear, thermal-hydraulic seismic and safety performance, metal and graphite components, core components and reactor internals, and cooling, reactivity control, instrumentation and control systems. The new design criteria have been developed for graphite structures, and a super heat-resistant alloy, Hastelloy XR, was newly developed for high temperature components such as heat transfer tubes of an intermediate heat exchanger. The performance and structural integrity of the fuel core components and reactor internals were verified in advance by the research and development described in Chapter 3, R&D on Components. As for the safety design, the strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat, and confinement of fission products shall be well performed to ensure safety.

The following RD fuel, core components and reactor internals, passive cooling system, intermediate heat exchanger, and basic feature of air ingress during primary pipe rupture accident. As for the fuel, the design criteria and specifications for the HTTR, and fabrication process are described. As for core components and reactor internals, thermal-hydraulic tests on the core and core bottom structure of the HTTR were carried out with the helium engineering demonstration loop (HENDEL) under simulated reactor operating conditions. As for the IHX, experimental and numerical studies with partial models were carried out. This chapter explained the performance and structural integrity of the fuel core components and reactor internals were verified in advance by the research and development.

High temperature gas-cooled reactor (HTGR) can be used in various ways such as hydrogen production and process heat supply as well as power generation. For the commercial HTGR, R&D on the system, and safety designs, key components for heat application and advanced fuel have been carried out by JAEA as shown in the following items: system design for power generation, system design for cogeneration, system design for steelmaking, safety design for connection of HTGR and heat application system, gas turbine technology for power generation, iodine–sulfur process technology for hydrogen production, system integration technology for connection of heat application system and HTGR, prevention technology for air ingress during depressurization accident, advanced fuel technology for high burnup, advanced fuel technology for plutonium burner, and advanced fuel technology for reduction of high-level radioactive waste. The final chapter explained research and development on the system and safety designs, key components for heat application and advanced fuel.

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.

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.

To reduce environmental burden and thread of nuclear proliferation, multi-recycling fuel cycle with High Temperature Gas-cooled Reactor (HTGR) has been investigated. Those problems are solved by incinerating TRans Uranium (TRU) nuclides, which is composed of plutonium and Minor Actinoide (MA), and there is concept to realize TRU incineration by multi-recycling with Fast Breeder Reactor (FBR). In this study, multi-recycling is realized even with thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium by reprocessing and natural uranium are enriched and mixed with recovered TRU by reprocessing and partitioning to fabricate fresh fuels. The fuel cycle was designed for a Gas Turbine High Temperature Reactor (GTHTR300), whose thermal power is 600 MW, including conceptual design of uranium enrichment facility. Reprocessing is assumed as existing Plutonium Uranium Redox EXtraction (PUREX) with four-group partitioning technology. As a result, it was found that the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for High Level Waste (HLW) can be reduced by 99.7\% compared with GTHTR300 using existing reprocessing and disposal technology. Suppress plutonium is not generated from this cycle. Moreover, incineration of TRU from Light Water Reactor (LWR) cycle can be performed in this cycle., 環境負荷低減と核拡散の脅威の削減を目的として高温ガス炉を用いたマルチリサイクルに関する研究を行った。これらの問題はプルトニウムとマイナーアクチノイドからなる超ウラン元素を燃焼させることにより解決され、高速増殖炉の多重リサイクルにより超ウラン元素を燃焼させるコンセプトがある。本研究では、増殖の代わりに核分裂性ウランをサイクルの外部から供給することにより、熱中性子炉であってもマルチリサイクルを実現させる。この燃料サイクルにおいて、再処理から得られる回収ウランと天然ウランは濃縮され、再処理・分離から得られる回収超ウラン元素と混合され、新燃料が作られる。その燃料サイクルを600MW出力のGTHTR300を対象に、ウラン濃縮施設の概念設計も含め設計した。再処理は現行PUREXに4群分離技術を付随したものを想定した。結果として、ネプツニウム以外の超ウラン元素のマルチリサイクルの成立を確認した。潜在的有害度が天然ウランレベル以下に減衰するまでの期間はおよそ300年程度であり、高レベル廃棄物の処分場専有面積は、既存の再処理処分技術を用いた場合と比較し99.7\%の削減を確認した。このサイクルから余剰プルトニウムは発生しない。さらに、軽水炉サイクルからの超ウラン元素の燃焼も本サイクルにより可能である。

Optimization of disposal method and scenario to reduce volume of High Level Waste (HLW) and the footprint in a geological repository for High Temperature Gas-cooled Reactor (HTGR) has been performed. It was found that HTGR has great advantages to reducing HLW volume and its footprint, which are high burn-up, high thermal efficiency and pin-in-block type fuel, compared with those of LWR and has potential to reduce those more in the previous study. In this study, the scenario is optimized, and the geological repository layout is designed with the horizontal emplacement based on the KBS-3H concept instead of the vertical emplacement based on KBS-3V concept employed in the previous study. As a result, for direct disposal, the repository footprint can be reduced by 20 \% by employing the horizontal without change of the scenario. By extending 40 years for cooling time before disposal, the footprint can be reduced by 50 \%. For disposal with reprocessing, the number of canister generation can be reduced by 20 \% by extending cooling time of 1.5 years between the discharge and reprocessing. The footprint per electricity generation can be reduced by 80 \% by extending 40 years before disposal. Moreover, by employing four-group partitioning technology without transmutation, the footprint can be reduced by 90 \% with cooling time of 150 years., 高温ガス炉の高レベル廃棄物減容及び処分場専有面積低減のための処分法及び処分シナリオの最適化を行った。高温ガス炉は廃棄物発生体積及び処分場専有面積低減に対し、軽水炉と比較し有利な特徴(高燃焼度、高熱効率、ピンインブロック型燃料)を持つこと、およびこれらの減容が可能であることが先行研究で分かっている。本研究では、シナリオの最適化、地層処分場のレイアウトをKBS-3H概念に基づいた横置きに基づき(先行研究では、KBS-3Vに基づいた竪置き)評価した。その結果、直接処分において、横置きを採用しただけで専有面積の20\%減を確認した。40年冷却期間を延長することにより、専有面積の50\%が低減できる。再処理時は燃料取り出しから再処理までの冷却期間を1.5年延長するだけで廃棄体発生体数の20\%削減ができる。専有面積については、処分までの冷却期間を40年延長することにより80\%の低減が可能である。さらに、核変換を行わずに4群分離技術のみを導入した場合、150年冷却の冷却を想定すると専有面積は90\%削減できることが分かった。

Coated fuel particle (CFP) is one of important factors attributing to the inherent safety feature of high temperature engineering test reactor (HTTR). However, the random arrangement of CFPs makes the simulation more complicated, becoming one of the factors affects the accuracy of the HTTR criticality calculations. In this study, an explicit random model for CFPs arrangement, namely realized random packing (RRP), was developed for the whole core of HTTR using a Monte-Carlo MCNP6 code. The effect of random placement of CFPs was investigated by making a comparison between the RRP and conventional uniform models. The results showed that the RRP model gave a lower excess reactivity than that of the uniform model, and the more number of fuel columns loading into the core, the greater the difference in excess reactivity between the RRP and uniform models. For example, the difference in excess reactivity increased from 0.07 to 0.17 %Δk/k when the number of fuel column increased from 9 to 30. Regarding the control rods position prediction, the RRP showed the results, which were closer to experiment than the uniform model. In addition, the difference in control rods position between the RRP and uniform models also increases from 12 to 17 mm as increasing number of fuel columns from 19 to 30.

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。

To develop Pu-burner high temperature gas-cooled reactor (HTGR) attaining very high burn-up around 500 GWd/t, the security and safety fuel (3S-TRISO fuel) is proposed. The 3S-TRISO fuel employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO2) and yttria stabilized zirconia (YSZ) as an inert matrix. Especially, a zirconium carbide (ZrC) coating is one of key technologies of the 3S-TRISO, which performs as an oxygen getter to reduce the fuel failure due to internal pressure during the irradiation. R&Ds on ZrC coating directly on the dummy kernel made of YSZ which surrogates PuO2-YSZ have been started in the Japanese fiscal year 2015. As results of ZrC coating tests on YSZ particle by the bromide process developed in Japan Atomic Energy Agency, stoichiometric ZrC coatings with 18 - 21 μm of thicknesses were obtained with 0.1 kg of particle loading weight. As indicated by a thermochemical analysis on the chemical stability of YSZ against the ZrC bromide process, no deterioration of YSZ exposed by source gases of ZrC bromide process was observed through the characterization by Scanning Transmission Electron Microscope (STEM) observation. Finally, reproducibility of stoichiometric ZrC coating on YSZ particle was demonstrated by the bromide chemical vapor deposition process.

A reasonable source term of a hypothetical criticality accident in a fuel fabrication facility with a solution system has been proposed. The public exposure must not exceed the limitation of 5 mSv during an accident. Then, we proposed as a reasonable source term of the first burst peak due to the hydrogen gas generation by radiation decomposition of water. With the criticality control system composed of the Criticality Accident Alarm System (CAAS) and a soluble neutron absorber, safety is ensured by the reduced fission number. We confirmed the effect by environmental impact assessment during a criticality accident by using the site condition of a fuel fabrication facility in Tokai-mura, Japan. As a result, the public exposure was reduced at a site boundary from 68 mSv to 0.6 mSv which is under the current regulatory guideline.

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.

This study investigated the random arrangement effect of coated fuel particles (CFPs) on the criticality of the fuel compact of the high-temperature engineering test reactor (HTTR). A utility program coupled with MCNP6, namely the realized random packing (RRP) model, was developed to explicitly model the random arrangement of CFPs in a specified fuel compact of the HTTR. The criticality and neutronic calculations for a pin cell model were performed using the Monte Carlo MCNP6 code with ENDF/B-VII.1 neutron library data. The reliability of the RRP model was confirmed by the insignificant variance of the infinite multiplication factor (kinf) among 10 different random arrangements of the CFPs. Next, the criticality of the RRP model was compared with those of a non-truncated uniform packing (NUP) model and an on-the-fly random packing (ORP) model, which is a stochastic geometry capability in MCNP6. The results indicated that there was no substantial difference between the NUP and ORP models. However, the RRP model presented a lower kinf (by about 0.32–0.52%Δk/k) than the NUP model. In addition, the difference in kinf was increased as uranium enrichment decreased. Investigation of the 4-factor formula showed that the difference in kinf was predominantly a result of the resonance escape probability, with the RRP model showing the smallest value.

Sustainable and safe energy supply with seawater fueled HTGR have been investigated to sustain the nuclear energy safely by electricity generation with HTGR, the uranium resources must be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. The uranium dissolved in seawater is in an equilibrium state with the uranium on surface of sea floor, which is approximately a thousand times of the amount, that is 72 million years. It can be recoverable. In other words, the uranium from seawater is almost inexhaustible natural resource. The cost of extracting uranium from seawater with current technology is still expensive compared with that of conventional uranium. However, the economy of nuclear power generation fueled by seawater uranium should be assessed for entire electricity generation cost. In the present study, the economy of electricity generation using uranium from seawater is assessed using a commercial HTGR. Compared with ordinary LWR using conventional uranium, HTGR can realize lower cost of electricity owing to small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR fueled by seawater uranium with the current technology enables the energy sustainability to be maintained for a long term approximately 70 million years with superior inherent safety features and low cost of 7.28 yen/kWh, which is lower than the 8.80 yen/kWh cost of LWR using conventional uranium., 海水ウランを用いた高温ガス炉による持続的で安全なエネルギー供給について検討した。ウラン資源が無尽蔵に存在すれば、安全な高温ガス炉発電による永続的なエネルギー供給が可能である。海水ウランは大量に海水中に存在するため代替候補となりうる。海水ウランは45億トン存在し、7万2千年のエネルギー消費量に相当する。また、海水ウランは海底表面の岩盤に含まれる4.5兆トンのウランと平衡状態にあるため、これらも回収することができる。この量は7200万年の消費に相当する。現状の技術では海水ウランの回収コストはウランの市場価格よりも高価である。しかし、経済的に成立するか否かの判断はウラン回収コストのみではなく発電原価全体でなされるべきである。本研究では、海水ウランを用いた高温ガス炉による商用発電の発電原価について評価した。高温ガス炉は直接ガスタービンによる簡素なシステムにより、海水ウランを用いても高い経済性が期待できる。結果として、海水ウランを用いた高温ガス炉の発電原価は7.28円/kWhであり、在来型ウランを用いた軽水炉の8.80円/kWhよりも安価と評価された。

Feasibility study on Burnable Poison (BP) credit concept to High Temperature Gas-cooled Reactor (HTGR) fuel fabrication has been performed. By mixing BP into fuel material in the first place of fuel fabrication, criticality safety is ensured in the all fuel fabrication process even with high enrichment fuel such as 14 wt% used in commercial HTGR. However, the poison effect also degrades the criticality even in the HTGR core, and it may shorten cycle length and achievable burn-up of the core. Therefore, the effect is evaluated by whole core burn-up calculation. As a BP, boron, gadolinium, erbium, and hafnium are investigated. As a result, it is found that boron and gadolinium suit this concept, and the 14 wt% fuel can be fabricated even with severe Light Water Reactor (LWR) class safety measure corresponding to 5 wt%.

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.

Objective Sublingual immunotherapy (SLIT) appears to offer practical advantages for the treatment of allergic rhinitis (AR). Based on a review of the scientific literature, we present recommendations as guiding principles to administer SLIT safely. Methods Clinical questions concerning SLIT were prepared. Literature published between January 2003 and December 2012 was searched from PubMed, the Cochrane Library, and Japana Centra Revuo Medicina. Qualified studies were analyzed and the results were evaluated, consolidated, and codified. We answered 17 clinical questions and, based on this, presented evidence-based recommendations. Results Sublingual immunotherapy improved symptoms (e.g., quality of life [QOL]) and reduced medication scores in seasonal AR and perennial AR. Most SLIT-induced adverse effects were local oral reactions, although systemic adverse effects such as gastrointestinal symptoms, urticaria, and asthma are occasionally reported. There have been no reports of lethal anaphylactic reactions by SLIT. When SLIT is continued for 3–4 years, its effect persists long after discontinuation. Conclusion A correct diagnosis of AR and sufficient informed consent from patients are required before initiating SLIT. Sublingual immunotherapy should be continued for 3 years or longer. The initial administration of SLIT during the uptitration of an allergen vaccine and the general condition of patients are critical for the safe performance of SLIT.

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 self-shielding effect of double heterogeneity for plutonium burner High Temperature Gas-cooled Reactor (HTGR) design has been performed. Plutonium burner HTGR designed in the previous study by using the advantage of double heterogeneity to control excess reactivity. In the present study, the mechanism of the self-shielding effect is elucidated by the analysis of burn-up calculation and reactivity decomposition based on exact perturbation theory. As a result, it is revealed that the characteristics of burn-up reactivity are determined by resonance cross section peak at 1 eV of $^{240}$Pu due to the surface term of background cross section, this is, the characteristics of neutron leakage from fuel lump and collision to a moderator. Moreover, significant spectrum shift is caused during the burn-up period, and it enhances reactivity worth of $^{239}$Pu and $^{240}$Pu in EOL., プルトニウム燃焼高温ガス炉のための二重非均質による自己遮蔽効果の研究が行われた。プルトニウム燃焼高温ガス炉設計の先行研究では、二重非均質性の利点を用い余剰反応度の制御を行った。本研究では、自己遮蔽効果のメカニズムを燃焼解析及び厳密摂動論に基づいた反応度分解法により解明した。結果として、燃焼反応度の特性が$^{240}$Puの1eVの共鳴断面積ピークにより決まり、背景断面積の表面項に依存することが明らかとなった。すなわち燃料塊からの中性子の漏洩と減速材への衝突により決定される。さらに、顕著なスペクトルシフトが燃焼期間中に起こり、この効果が、$^{239}$Puと$^{240}$PuのEOLにおける反応度価値が強化されることが分かった。

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.