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Journal Articles

Heavy fermion state of YbNi$$_2$$Si$$_3$$ without local inversion symmetry

Nakamura, Shota*; Hyodo, Kazushi*; Matsumoto, Yuji*; Haga, Yoshinori; Sato, Hitoshi*; Ueda, Shigenori*; Mimura, Kojiro*; Saiki, Katsuyoshi*; Iso, Kosei*; Yamashita, Minoru*; et al.

Journal of the Physical Society of Japan, 89(2), p.024705_1 - 024705_5, 2020/02

 Times Cited Count:0 Percentile:100(Physics, Multidisciplinary)

JAEA Reports

Code-B-2.5.2 for stress calculation for SiC-TRISO fuel particle

Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio

JAEA-Data/Code 2019-018, 22 Pages, 2020/01

JAEA-Data-Code-2019-018.pdf:1.39MB

Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO$$_{2}$$ (PuO$$_{2}$$-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.

Journal Articles

Microstructures of ZrC coated kernels for fuel of Pu-burner high temperature gas-cooled reactor in Japan

Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Mizuta, Naoki; Goto, Minoru; Tachibana, Yukio; Okamoto, Koji*

Journal of Nuclear Materials, 522, p.32 - 40, 2019/08

In order to realize Pu-burner high temperature gas-cooled reactor (HTGR), coated fuel particles (CFPs) with PuO$$_{2}$$-yittria stabilized zirconia (YSZ) fuel kernel coated with ZrC is employed for high nuclear proliferation resistance and very high burn-up. Japan Atomic Energy Agency (JAEA) have carried out ZrC coatings of particles which simulated PuO$$_{2}$$-YSZ kernels (CeO$$_{2}$$-YSZ particles or commercially available YSZ particles). Ce was used as simulating element of Pu. In this manuscript, microstructures of ZrC coated CeO$$_{2}$$-YSZ or YSZ particles were reported.

Journal Articles

Development of compact high field pulsed magnet system for new sample environment equipment at MLF in J-PARC

Watanabe, Masao; Nojiri, Hiroyuki*; Ito, Shinichi*; Kawamura, Seiko; Kihara, Takumi*; Masuda, Takatsugu*; Sahara, Takuro*; Soda, Minoru*; Takahashi, Ryuta

JPS Conference Proceedings (Internet), 25, p.011024_1 - 011024_5, 2019/03

Recently, neutron scattering experiments have been rapidly progressed under high magnetic field. In the J-PARC, proto-type compact pulse magnet system with the power supply, the coil and the sample stick has been developed. Basic specifications of the power supply are as follows; maximum charged voltage with capacitor is 2 kV, maximum current is 8 kA, repetition rate is a pulse per several minutes and pulse duration is several msec. Maximum magnetic field in the coil is more than 30 Tesla. The sample stick is designed for Orange-Cryostat. In this presentation, We report the details of the pulsed magnet system and the performance of it on neutron scattering experiments at MLF beam line (HRC).

Journal Articles

A Probabilistic Approach to Assess External Doses to the Public Considering Spatial Variability of Radioactive Contamination and Interpopulation Differences in Behavior Pattern

Takahara, Shogo; Iijima, Masashi*; Yoneda, Minoru*; Shimada, Yoko*

Risk Analysis, 39(1), p.212 - 224, 2019/01

 Times Cited Count:3 Percentile:13.74(Public, Environmental & Occupational Health)

A dose assessment model was developed based on measurements and surveys of individual doses and relevant contributors in Fukushima City for four population groups: Fukushima City Office staff, Senior Citizens' Club, Contractors' Association, and AgriculturalCooperative. In addition, probabilistic assessments were performed for these population groups by considering the spatial variability of contamination and interpopulation differencesresulting from behavior patterns. As a result of comparison with the actual measurements, the assessment results for participants from the Fukushima City Office, Senior Citizens' Club and the Agricultural Cooperative agreed with the measured values. By contrast, the measurements obtained for the participants from the Contractors' Association were not reproduced well in the present study. To assess the doses to this group, further investigations of association members' work activities and the related dose reduction effects are needed.

Journal Articles

Uranium-based TRU multi-recycling with thermal neutron HTGR to reduce environmental burden and threat of nuclear proliferation

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

 Times Cited Count:0 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.

Journal Articles

Development of security and safety fuel for Pu-burner HTGR; Test and characterization for ZrC coating

Ueta, Shohei; Aihara, Jun; Goto, Minoru; Tachibana, Yukio; Okamoto, Koji*

Mechanical Engineering Journal (Internet), 5(5), p.18-00084_1 - 18-00084_9, 2018/10

To develop the security and safety fuel (3S-TRISO fuel) for Pu-burner high temperature gas-cooled reactor (HTGR), R&D on zirconium carbide (ZrC) directly coated on yttria stabilized zirconia (YSZ) has been started in the Japanese fiscal year 2015. As results of the direct coating test of ZrC on the dummy YSZ particle, ZrC layers with 18 - 21 microns of thicknesses have been obtained with 0.1 kg of particle loading weight. No deterioration of YSZ exposed by source gases of ZrC bromide process was observed by Scanning Transmission Electron Microscope (STEM).

Journal Articles

Conceptual plant system design study of an experimental HTGR upgraded from HTTR

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.

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Design study of fuel and reactor core

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.

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Test and characterization for ZrC coating

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$$_{2}$$) 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$$_{2}$$-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.

JAEA Reports

Excellent feature of Japanese HTGR technologies

Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju; Ohashi, Hirofumi; Kubo, Shinji; Inaba, Yoshitomo; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; et al.

JAEA-Technology 2018-004, 182 Pages, 2018/07

JAEA-Technology-2018-004.pdf:18.14MB

Research and development on High Temperature Gas-cooled Reactor (HTGR) in Japan started since late 1960s. Japan Atomic Energy Agency (JAEA) in cooperation with Japanese industries has researched and developed system design, fuel, graphite, metallic material, reactor engineering, high temperature components, high temperature irradiation and post irradiation test of fuel and graphite, high temperature heat application and so on. Construction of the first Japanese HTGR, High Temperature engineering Test Reactor (HTTR), started in 1990. HTTR achieved first criticality in 1998. After that, various test operations have been carried out to establish the Japanese HTGR technologies and to verify the inherent safety features of HTGR. This report presents several system design of HTGR, the world-highest-level Japanese HTGR technologies, JAEA's knowledge obtained from construction, operation and management of HTTR and heat application technologies for HTGR.

JAEA Reports

Comparison between HTFP code and minory changed FORNAX-A code

Aihara, Jun; Ueta, Shohei; Goto, Minoru; Inaba, Yoshitomo; Shibata, Taiju; Ohashi, Hirofumi

JAEA-Technology 2018-002, 70 Pages, 2018/06

JAEA-Technology-2018-002.pdf:1.46MB

HTFP code is code for calculation of additional release amount of fission product (FP) from fuel rod in high temperature gas-cooled reactor (HTGR) after stop of fission. Minory changed Fornax-A code also can calculate that. Therefore, release behavior of Cs calculated with HTFP code was compared with that calculated with minory modified FORNAX-A code in this report. Release constants of Cs evaluated with minory modified FORNAX-A code are rather different from default values for HTFP code.

Journal Articles

Optimization of disposal method and scenario to reduce high level waste volume and repository footprint for HTGR

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:72.34(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.

Journal Articles

Estimation of radiocesium dietary intake from time series data of radiocesium concentrations in sewer sludge

Pratama, M. A.; Takahara, Shogo; Munakata, Masahiro; Yoneda, Minoru*

Environment International, 115, p.196 - 204, 2018/06

 Times Cited Count:1 Percentile:86.85(Environmental Sciences)

Journal Articles

Burn-up characteristics and criticality effect of impurities in the graphite structure of a commercial-scale prismatic HTGR

Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo

Nuclear Engineering and Design, 326, p.108 - 113, 2018/01

 Times Cited Count:2 Percentile:50.63(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 $$^{10}$$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.

Journal Articles

Experimental observation of temperature and magnetic-field evolution of the 4${it f}$ states in CeFe$$_{2}$$ revealed by soft X-ray magnetic circular dichroism

Saito, Yuji; Yasui, Akira*; Fuchimoto, Hiroto*; Nakatani, Yasuhiro*; Fujiwara, Hidenori*; Imada, Shin*; Narumi, Yasuo*; Kindo, Koichi*; Takahashi, Minoru*; Ebihara, Takao*; et al.

Physical Review B, 96(3), p.035151_1 - 035151_5, 2017/07

AA2017-0611.pdf:0.36MB

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

We revisit the delocalized character of the 4$$f$$ states of CeFe$$_2$$ in the ferromagnetically ordered phase by X-ray magnetic circular dichroism (XMCD) in X-ray absorption spectroscopy (XAS) with improved data quality using single crystals. Surprisingly, the Ce $$M_{4,5}$$ XMCD spectral shape changes significantly as a function of temperature and applied magnetic field, with no concomitant changes in the spectral shape of the Ce $$M_{4,5}$$ XAS as well as the Fe $$L_{2,3}$$ XAS and XMCD. This unusual behavior is characterized by the $$J=7/2$$ states in a 4$$f^1$$ configuration mixed into the $$J=5/2$$ ground state. Such extreme sensitivity of the Ce 4$$f$$ states to the external perturbations can be related to the magnetic instability toward an antiferromagnetic phase in CeFe$$_2$$. Our experimental data presented here provide valuable insights into the underlying physics in strongly-hybridized ferromagnetic Ce compounds.

Journal Articles

Development of security and safety fuel for Pu-burner HTGR, 2; Design study of fuel and reactor core

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$$_{2}$$-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$$^{circ}$$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$$_{2}$$ 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.

Journal Articles

Development of security and safety fuel for Pu-burner HTGR, 5; Test and characterization for ZrC coating

Ueta, Shohei; Aihara, Jun; Goto, Minoru; Tachibana, Yukio; Okamoto, Koji*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 4 Pages, 2017/07

To develop the security and safety fuel (3S-TRISO fuel) for Pu-burner high temperature gas-cooled reactor (HTGR), R&D on zirconium carbide (ZrC) directly coated on yttria stabilized zirconia (YSZ) has been started in the Japanese fiscal year 2015. As results of the direct coating test of ZrC on the dummy YSZ particle, ZrC layers with 18 - 21 microns of thicknesses have been obtained with 0.1 kg of particle loading weight. No deterioration of YSZ exposed by source gases of ZrC bromide process was observed by Scanning Transmission Electron Microscope (STEM).

Journal Articles

Bioaccessibility of Fukushima-accident-derived Cs in soils and the contribution of soil ingestion to radiation doses in children

Takahara, Shogo; Ikegami, Maiko*; Yoneda, Minoru*; Kondo, Hitoshi*; Ishizaki, Azusa; Iijima, Masashi; Shimada, Yoko*; Matsui, Yasuto*

Risk Analysis, 37(7), p.1256 - 1267, 2017/07

AA2015-0445.pdf:0.53MB

 Times Cited Count:4 Percentile:33.18(Public, Environmental & Occupational Health)

Journal Articles

Nuclear thermal design of high temperature gas-cooled reactor with SiC/C mixed matrix fuel compacts

Aihara, Jun; Goto, Minoru; Inaba, Yoshitomo; Ueta, Shohei; Sumita, Junya; Tachibana, Yukio

Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.814 - 822, 2016/11

Japan Atomic Energy Agency (JAEA) has started R&D for apply SiC/C mixed matrix to fuel element of high temperature gas-cooled reactors (HTGRs) to improve oxidation resistance of fuel. Nuclear thermal design of HTGR with SiC/C mixed matrix fuel compacts was carried out as a part of above R&Ds. Nuclear thermal design was carried out based on a small sized HTGR for developing countries, HTR50S. Maximum enrichment of uranium is set to be 10 wt%, because coated fuel particles with 10 wt% uranium have been fabricated in Japan. Numbers of kinds of enrichment and burnable poisons (BPs) were set to be same as those of original HTR50S (3 and 2, respectively). We succeeded in nuclear thermal design of a small sized HTGR which performance was equivalent to original HTR50S, with SiC/C mixed matrix fuel compacts. Based on nuclear thermal design, intactness of coated fuel particles was evaluated to be kept on internal pressure during normal operation.

173 (Records 1-20 displayed on this page)