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Fukaya, Yuji
Meruko Kanri Kaikei Kenkyu, (11-2), p.45 - 62, 2019/05
Evaluation of electricity generation cost is important to make energy policy. Levelizing technique for cost and benefit, i.e. discounting technique, plays an important role in the evaluation especially for Nuclear Power Generation (NPG) because the activity related to power generation have been performed beyond generations. There are many arguments for the discounting technique for a long term, e.g. the cost of climate change. However, those are few for the electricity generation cost for NPG. Therefore, I discuss the discounting technique by referring the history of cost evaluation for NPG and the recent discussion of backend problem. Moreover, gamma discounting, which attracts a particular interest because of the gradual decrees of effective discounting rate, and two-good model are attempted and discussed.
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Yan, X.; Nishihara, Tetsuo; Tsubata, Yasuhiro; Matsumura, Tatsuro
Journal of Nuclear Science and Technology, 55(11), p.1275 - 1290, 2018/11
Percentile：100(Nuclear Science & Technology)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.
Fukaya, Yuji; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio; Okamoto, Koji*
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 9 Pages, 2018/10
The research on introduction scenarios of Pu-burner High Temperature Gas-cooled Reactor (HTGR) of Japan has been performed based on the "Long-term Energy Supply and Demand Outlook" released by the Ministry of Economy, Trade and Industry (METI) of Japan in 2015. In the perspective, the electricity generation capacity of nuclear power generation reduces from 50 GWe (peak around 2010) to 30 GWe in 2030. To maintain the capacity, light water reactors (LWRs) should be introduced from 2025 to 2030. After 2030, HTGRs, which are superior to LWRs from the viewpoint of safety and economy, will be introduced to fill the capacity and incinerate plutonium. We assumed introduction of U fueled HTGR as well. The Pu-burner reactor will be introduced with the priority to incinerate separated plutonium by reprocessing. Moreover, we also evaluated hydrogen generation and its effect on CO
reduction. As a result, effective plutonium incineration and CO reduction effect are confirmed.
Fukaya, Yuji; Mizuta, Naoki; Goto, Minoru; Ohashi, Hirofumi; Yan, X.
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10
Conceptual design study of a high performance commercial HTGR has been performed at target output of 165MWt. Requirements for the HTGR are small-sized vessel for transportation, durability of vessel to irradiation damage, fuel reloading scheme to shorten the duration of reloading, low pressure drop fuel element, a small number of fuel enrichments, and so on. To satisfy the requirement, we investigated the core configuration, shielding and reflector configuration, fuel reloading scheme. As a result, we completed the design with the vessel diameter of 4.5m, which can be transported by any means, such as, by load, rail, ship, and air plane, and high load factor over 90%.
Ohashi, Hirofumi; Goto, Minoru; Ueta, Shohei; Sato, Hiroyuki; Fukaya, Yuji; Kasahara, Seiji; Sasaki, Koei; Mizuta, Naoki; Yan, X.; Aoki, Takeshi*
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10
Conceptual design study of an experimental HTGR is performed to upgrade the plant system from Japanese High Temperature engineering Test Reactor (HTTR) to a commercial HTGR. Safety systems of HTTR are upgraded to demonstrate the commercial HTGR concept, such as a passive reactor cavity cooling system, a confinement, etc. An intermediate heat exchanger (IHX) is replaced by a steam generator (SG) for a process heat supply to demonstrate the technology for a commercial use. This paper describes the conceptual design study results of the plant system of the experimental HTGR.
Goto, Minoru; Fukaya, Yuji; Mizuta, Naoki; Inaba, Yoshitomo; Ohashi, Hirofumi; Yan, X.
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10
The HTTR (High Temperature engineering Test Reactor) constructed at JAEA-Oarai R&D center is a block-type experimental HTGR (High Temperature Gas-cooled Reactor) with 30 MW thermal power. It attained the first criticality at 1998 and has yielded very useful data for future HTGR design. Although the HTTR was designed very conservatively because the HTTR is the first HTGR for Japan, future HTGRs can be designed with a reasonable conservativeness based on the HTTR data. Additionally, it is possible to enhance the performance of the reactor core by improving the design and introducing new technologies. This paper describes a concept of an experimental HTGR that is upgraded from the HTTR by the reasonable conservativeness, the design improvement and the new technology introduction.
Goto, Minoru; Aihara, Jun; Inaba, Yoshitomo; Ueta, Shohei; Fukaya, Yuji; Okamoto, Koji*
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10
JAEA has conducted design studies of a Pu-burner HTGR. The Pu-burner HTGR incinerates Pu by fission, and hence a high burn-up is required for the efficient incineration. In the fuel design, a thin ZrC layer, which acts as an oxygen getter and suppresses the internal pressure, was coated on the fuel kernel to prevent the CFP failure at the high burn-up. A stress analysis of the SiC layer, which acts as a pressure vessel for the CFP, was performed for with consideration of the depression effect due to the ZrC layer. As a result, the CFP failure fraction at high burn-up of 500 GWd/t satisfied the target value. In the reactor core design, an axial fuel shuffling was employed to attain the high burn-up, and the nuclear burn-up calculations with the whole core model and the fuel temperature calculations were performed. As a result, the nuclear characteristics, which are the shutdown margin and the temperature coefficient of reactivity, and the fuel temperature satisfied their target values.
Ueta, Shohei; Aihara, Jun; Mizuta, Naoki; Goto, Minoru; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10
The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO) 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 CeO-YSZ kernel have been carried in the Japanese fiscal year 2017. As results of ZrC coating tests by the bromide chemical vapor deposition process, stoichiometric ZrC coatings with 3 - 18 microns of thicknesses were obtained with 0.1 kg of particle loading weight.
Fukaya, Yuji; Kasahara, Seiji; Mizuta, Naoki; Inaba, Yoshitomo; Shibata, Taiju; Nishihara, Tetsuo
JAEA-Research 2018-004, 38 Pages, 2018/06
JAEA-Research-2018-004.pdf:1.81MB
The demand of HTGR to investigate its introduction scenario and heat balance of HTGR have been researched. First, previous studies of HTGR demand were researched. Next, heat balance of GTHTR300, a commercial scale HTGR design, and its characteristics were researched. By using this information, installation number of HTGR to suit for demand in Japan are evaluated. In addition, heat balance evaluation code was developed in this study.
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Nishihara, Tetsuo; Tsubata, Yasuhiro; Matsumura, Tatsuro
Annals of Nuclear Energy, 116, p.224 - 234, 2018/06
Times Cited Count：1 Percentile：38.14(Nuclear Science & Technology)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.
Fukaya, Yuji
Uranium; Safety, Resources, Separation and Thermodynamic Calculation, p.22 - 48, 2018/05
Safety and economics of uranium utilization for nuclear power generation were investigated and discussed. In order to sustain energy supply with nuclear power generation, uranium resources should be abundant. From the viewpoint of depletion of the resources, FBR, which is breeder reactor of plutonium, has been developed, but that has been not diffused as a commercial reactor yet. Instead of obtaining inexhaustible resources by breeding plutonium, it is known that the inherent safety feature becomes weak in the fast neutron spectrum. As the result of the investigation, it is confirmed with concrete FBR designs that the inherent safety feature and breeding ability are related to the transactions. The amount of uranium resources and electricity generation cost with the resources were investigated. It is concluded that the semi-permanently sustainable energy supply can be established with reasonable cost by using seawater uranium. In addition, the significance of P&T, which is one of the advantages of FBR, was also discussed from the viewpoint of environmental burden from radioactive waste.
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
Nuclear Engineering and Design, 326, p.108 - 113, 2018/01
Percentile：100(Nuclear Science & Technology)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 
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 %
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.
Nakagawa, Shigeaki; Sato, Hiroyuki; Fukaya, Yuji; Tokuhara, Kazumi; Ohashi, Hirofumi
JAEA-Technology 2017-022, 32 Pages, 2017/09
JAEA-Technology-2017-022.pdf:3.59MB
As for the design of commercial HTGRs, the fuel design, core design, reactor coolant system design, secondary helium system design, decay heat removal system design and confinement system design are very important and quite different from those of LWRs. To contribute the establishment of the safety standards for commercial HTGRs, the evaluation items to attain safety requirements in fuel and core designs were studied. In this study, the excellence features of HTGRs based on passive safety or inherent safety were fully reflected. Additionally, concerning the core design, the stability to spatial power oscillation in reactor core of HTGR was studied. The evaluation items as the result of the study are applicable to the safety design of commercial HTGRs in the future.
Goto, Minoru; Ueta, Shohei; Aihara, Jun; Inaba, Yoshitomo; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
A PuO-YSZ fuel kernel with a ZrC coating, which enhances safety, security and safeguard, namely: 3S-TRISO fuel, was proposed to introduce to the plutonium-burner HTGR. In this study, the efficiency of the ZrC coating as the free-oxygen getter was examined based on a thermochemical calculation. A preliminary study on the feasibility of the 3S-TRISO fuel was conducted focusing on the internal pressure. Additionally, a nuclear feasibility of the reactor core was studied. As a result, all the amount of the free-oxygen is captured by a thin ZrC coating under 1600
C and coating ZrC on the fuel kernel should be very effective method to suppress the internal pressure. The internal pressure of the 3S-TRISO fuel at 500 GWd/t is lower than that of UO kernel TRISO fuel whose feasibility had been already confirmed and the 3S-TRISO fuel should be feasible. The fuel shuffling allows to achieve 500 GWd/t. The temperature coefficient of reactivity is negative during the operation period and thus the nuclear feasibility of the reactor core should be achievable.
Fukaya, Yuji; Goto, Minoru
Annals of Nuclear Energy, 99, p.19 - 27, 2017/01
Times Cited Count：1 Percentile：64.68(Nuclear Science & Technology)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.
Honda, Yuki; Fukaya, Yuji; Nakagawa, Shigeaki; Baker, R. I.*; Sato, Hiroyuki
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.704 - 713, 2016/11
A high-temperature gas-cooled reactor (HTGR) has superior safety characteristics. A loss of forced cooling (LOFC) test using a high-temperature engineering test reactor (HTTR) has been carried out to verify the inherent safety of an HTGR when forced cooling is diminished without reactor scram. In the test, an all-gas circulator was tripped with an initial reactor power of 9 MW and re-criticality was shown. This study focuses on developing a point kinetics method with RELAP5-3D code for an LOFC accident. There is a large temperature difference between the inlet and outlet of the core in an HTGR, and the temperature fluctuation range has been large in several accidents. We analyze the temperature dependency of xenon-135 reactivity and show that the temperature dependency of xenon-135 microscopic absorption cross-section affected the re-criticality time of the LOFC test.
Fukaya, Yuji; Nishihara, Tetsuo
Nuclear Engineering and Design, 307, p.188 - 196, 2016/10
Times Cited Count：1 Percentile：76.09(Nuclear Science & Technology)Reduction of High Level Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and the features are significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing, and effective waste loading method for direct disposal is proposed by applying the feature in this study. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with LWR representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.
Fukaya, Yuji; Tokuhara, Kazumi; Nishihara, Tetsuo
JAEA-Research 2016-008, 52 Pages, 2016/06
JAEA-Research-2016-008.pdf:2.18MB
To investigate the xenon stability quantitatively, a study on stability criterion of xenon oscillation based on an analysis solution for HTGR design had been performed. Randall developed the stability criterion method of xenon oscillation based on an analysis solution. And, that have been employed for a LWR design. On the other hand, HTGR is also planted to design new type of reactors, such as Pu fueled reactor, and it is necessary to confirm the xenon stability of those new types of reactors. Then, we developed the criterion method based on the Randall's method termed D-XESC/A, and high xenon stability of HTGR and feasibility for Pu fueled reactor is confirmed by comparing with xenon stability of other types of reactors.
Fukaya, Yuji; Nishihara, Tetsuo
JAEA-Research 2016-001, 23 Pages, 2016/05
JAEA-Research-2016-001.pdf:3.31MB
A study on Correlation effect between elements of statistical hot spot factor for High Temperature Gas-cooled Reactor (HTGR) Design had been performed. Both of safety and reactor specification can be remained if the uncertainty is correctly propagated by revising hot spot factor. In this context, it is reported for light water reactor design that the propagated uncertainty can be reduced by statistical hot spot factors with numerical statistical approach, that is Monte Carlo method, because correlation effects for each factor can be considered. For HTGR with sleeve covered fuel, it is expected that the fuel temperature also reduces by employing the same approach because the gap between sleeve and fuel compact, which shows significant temperature increase, have direct correlation. In addition, Monte Carlo method treats correlation effect at the price of evaluating contribution of individual factor. Therefore, improved method based on conventional method has been developed in this study. Then, statistical hot spot factor for fuel temperature of HTGR was evaluated by Monte Carlo method and the improved method. As a result, it is not found significant difference between the result of the conventional method and the improved method. Moreover, usage of hot spot factor is investigated and we proposed new one reflecting the investigation.
Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
JAEA-Research 2015-023, 44 Pages, 2016/02
JAEA-Research-2015-023.pdf:2.13MB
A study on innovative High Temperature Gas-cooled Reactor (HTGR) to reduce generation of potential radiotoxicity had been performed. Unlike the Fast Breeder Reactor (FBR) and Accelerated Driven System (ADS), which can confine radioactive nuclides into its fuel cycle as multi-recycling and transmute, in this study we attempt to reduce the generation of the radiotoxicity itself by preventing the generation of Pu and MA, which is generated with the energy generation. In this context, we proposed the innovative HTGR that employs the Highly Enriched Uranium (HEU) fuel by removing 
U : source of the Pu and MA. However, there are the problems of fuel integrity, nuclear proliferation, nuclear self-regulation characteristics, and economy of electricity generation which are caused by employing HEU. For these problems, we investigated and proposed the solutions. Especially for the nuclear self-regulation characteristics, which were improved by adding Er, the optimized nuclear design was quantitatively determined and elucidated by the Bondarenko approach. As a result, it was confirmed that the proposed reactor can solves these problem for employing HEU fuel and the high specification and economy as same as those of standard HTGR fueled uranium.
