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Journal Articles

Influences of the ZrC coating process and heat treatment on ZrC-coated kernels used as fuel in 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 Science and Technology, 58(1), p.107 - 116, 2021/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The concept of a Pu-burner high temperature gas-cooled reactor (HTGR) has been proposed for purpose of more safely reducing amount of recovered Pu. This concept employs coated fuel particles (CFPs) with ZrC coated PuO$$_{2}$$-YSZ kernel and with tristructural (TRISO) coating for very high Pu burn-up and high nuclear proliferation resistance. In this report, we investigate the microstructure of the region that includes the surface of an as-fabricated CeO$$_{2}$$-YSZ kernel simulating PuO$$_{2}$$-YSZ kernel. We found both Zr-rich grains and Ce-rich grains to be densely distributed in that region including surface of CeO$$_{2}$$-YSZ kernel. On the other hand, it has been reported that there was a porous region near surface of the CeO$$_{2}$$-YSZ kernel of Batch I. This finding confirms that Ce-rich grains near surface of CeO$$_{2}$$-YSZ kernels coated with ZrC layers have been corroded during the deposition of the ZrC layer, whereas the Zr-rich grains were hardly affected.

Journal Articles

Present status and practical issues on dosimetry for the lens of the eye at JAEA MOX Fuel Facilities

Tsujimura, Norio; Yamazaki, Takumi; Takada, Chie

Journal of Nuclear Science and Technology, 58(1), p.40 - 44, 2021/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Evaluation of tritium release into primary coolant for research and testing reactors

Kenzhina, I.*; Ishitsuka, Etsuo; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*

Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain reaction of $$^{9}$$Be is dominant and its calculation results agree well with trends derived from the measured variation of tritium concentration in the primary coolant. It also indicates that the simple calculation method used in this study for the tritium recoil release from the beryllium reflectors can be utilized for an estimation of the tritium release into the primary coolant for a research and testing reactors containing beryllium reflectors.

Journal Articles

Online measurement of the atmosphere around geopolymers under gamma irradiation

Cantarel, V.; Lambertin, D.*; Labed, V.*; Yamagishi, Isao

Journal of Nuclear Science and Technology, 58(1), p.62 - 71, 2021/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The gas production of wasteforms is a major safety concern for encapsulating active nuclear wastes. For geopolymers and cements, the H$$_{2}$$ produced by radiolytic processes is a key factor because of the large amount of water present in their porous structure. Herein, the gas composition evolution around geopolymers was monitored on line under $$^{60}$$Co gamma irradiation. Transient evolution of the hydrogen production yield was measured for samples with different formulations. The rate of its evolution and the final values are consistent with the presence of a chemical reaction of the pseudo-first order consuming hydrogen in the samples. The results show this phenomenon can significantly reduce the hydrogen source term of geopolymer wasteform provided their diffusion constant remains low. Lower hydrogen production rates and faster kinetics were observed with geopolymers formulations in which pore water pH was higher. Besides hydrogen production, a steady oxygen consumption was observed for all geopolymers samples. The oxygen consumption rates are proportional to the diffusion constants estimated in the modelization of hydrogen recombination by a pseudo first order reaction.

Journal Articles

Derivation of ideal power distribution to minimize the maximum kernel migration rate for nuclear design of pin-in-block type HTGR

Okita, Shoichiro; Fukaya, Yuji; Goto, Minoru

Journal of Nuclear Science and Technology, 58(1), p.9 - 16, 2021/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Suppressing the kernel migration rates, which depend on both the fuel temperature and the fuel temperature gradient, under normal operation condition is quite important from the viewpoint of the fuel integrity for High Temperature Gas-cooled Reactors. The presence of the ideal axial power distribution to minimize the maximum kernel migration rate allows us to improve efficiency of design work. Therefore, we propose a new method based on Lagrange multiplier method in consideration of thermohydraulic design in order to obtain the ideal axial power distribution to minimize the maximum kernel migration rate. For one of the existing conceptual designs performed by JAEA, the maximum kernel migration rate for the power distribution to minimize the maximum kernel migration rate proposed in this study is lower by approximately 10% than that for the power distribution as a conventional design target to minimize the maximum fuel temperature.

Journal Articles

A Sensitive method for Sr-90 analysis by accelerator mass spectrometry

Sasa, Kimikazu*; Honda, Maki; Hosoya, Seiji*; Takahashi, Tsutomu*; Takano, Kenta*; Ochiai, Yuta*; Sakaguchi, Aya*; Kurita, Saori*; Satou, Yukihiko; Sueki, Keisuke*

Journal of Nuclear Science and Technology, 58(1), p.72 - 79, 2021/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Experimental study on transport behavior of cesium iodide in the reactor coolant system under LWR severe accident conditions

Miyahara, Naoya; Miwa, Shuhei; Gou$"e$llo, M.*; Imoto, Jumpei; Horiguchi, Naoki; Sato, Isamu*; Osaka, Masahiko

Journal of Nuclear Science and Technology, 57(12), p.1287 - 1296, 2020/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

In order to clarify the cesium iodide (CsI) transport behavior with a focus on the mechanisms of gaseous iodine formation in the reactor coolant system of LWR under a severe accident condition, a reproductive experiment of CsI transport behavior was conducted using a facility equipped with a thermal gradient tube. Various analyses on deposits and airborne materials during transportation could elucidate two mechanisms for the gaseous iodine formation. One was the gaseous phase chemical reaction in Cs-I-O-H system at relatively high-temperature region, which led to gaseous iodine transport to the lower temperature region without any further changes in gas species due to the kinetics limitation effects. The other one was the chemical reactions related to condensed phase of CsI, namely those of CsI deposits on walls with surface of stainless steel to form Cs$$_{2}$$CrO$$_{4}$$ compound and CsI aerosol particles with steam, which were newly found in this study.

Journal Articles

Evaluation of decreasing trend in air dose rate and ecological half-life within an 80 km range from Fukushima Dai-ichi Nuclear Power Plant, using car-borne survey data measured by KURAMA systems up to 2018

Ando, Masaki; Sasaki, Miyuki; Saito, Kimiaki

Journal of Nuclear Science and Technology, 57(12), p.1319 - 1330, 2020/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Air dose rates measured by car-borne surveys within 80 km range of Fukushima Dai-ichi Nuclear Power Plant using KURAMA systems from 2011 up to 2018 were analyzed, and decreasing trends and ecological half-life were evaluated. The decreasing speed of air dose rates became 0.08 over a period of seven years, indicating a much more rapid decrease than estimated by the physical decay of radiocesium (0.3). The fast components of the ecological half-lives evaluated in entire the evacuation order area were about 2 to 3 y, and were considerably larger than those outside the evacuation order area (0.4 to 0.5 y). To model the acceleration in the decrease of air dose rates observed in evacuation order areas, we modified the bi-exponential curve formula of ecological half-life and introduced the two-group model. The fast decreasing components of the ecological half-life evaluated using the two-group model after 2013 were 0.5 to 1 y, and were much shorter than those up to 2013, at 2 to 3 y.

Journal Articles

Voltage drop analysis and leakage suppression design for mineral-insulated cables

Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Otsuka, Noriaki; Tsuchiya, Kunihiko

Journal of Nuclear Science and Technology, 57(12), p.1276 - 1286, 2020/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The influence of materials of mineral-insulated (MI) cables on their electrical characteristics upon exposure to high-temperature conditions was examined via a transmission test, in the objective of achieving the stability of the potential distribution along the cable length. Occurrence of a voltage drop along the cable was confirmed for aluminum oxide (Al$$_{2}$$O$$_{3}$$) and magnesium oxide (MgO), as insulating materials of the MI cable. A finite-element method (FEM)-based analysis was performed to evaluate the leakage in the potentials, which was found at the terminal end. Voltage drop yields by the transmission test and the analysis were in good agreement for the MI cable of Al$$_{2}$$O$$_{3}$$ and MgO materials, which suggests the reproducibility of the magnitude relationship of the experimental results via the FEM analysis. To suppress the voltage drop, the same FEM analysis was conducted, the diameter of the core wires ($$d$$) and the distance between them ($$l$$) were varied. Considering the variation of $$d$$, the potential distribution in the MI cable produced a minimum voltage drop corresponding to a ratio $$d/D$$ of 0.35, obtained by dividing $$d$$ with that of the insulating material ($$D$$). In case of varying $$l$$, a minimum voltage drop was l/$$D$$ of 0.5.

Journal Articles

Safety enhancement approach against external hazard on JSFR reactor building

Yamamoto, Tomohiko; Kato, Atsushi; Chikazawa, Yoshitaka; Hara, Hiroyuki*

Nuclear Technology, 206(12), p.1875 - 1890, 2020/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

This paper gives a detailed evaluation of the countermeasures for the external hazards and severe accidents that could impact the 2010 JSFR design building by lessons learned from the Fukushima Daiichi nuclear power plant (Fukushima I NPP) accident.

Journal Articles

Flexible fuel cycle system for the effective management of plutonium

Fukasawa, Tetsuo*; Hoshino, Kuniyoshi*; Yamashita, Junichi*; Takano, Masahide

Journal of Nuclear Science and Technology, 57(11), p.1215 - 1222, 2020/11

The flexible fuel cycle initiative system (FFCI system) has been developed to reduce spent fuel (SF) amounts, to keep high availability factor for the reprocessing plant and to increase the proliferation resistance for the recovered Pu. The system separates most U from the SF at first, and the residual material called recycle material (RM) which contains Pu, minor actinides, fission products and remaining U will go to Pu(+U) recovery from the RM for Pu utilizing reactor in future. The Pu utilizing reactor is FBR or LWR with MOX fuel. The RM is the buffer material between SF reprocessing and Pu utilizing reactor with compact size and high proliferation resistance, which can suppress the amount of relatively pure Pu. The innovative technologies of FFCI are most U separation and temporary RM storage. They are investigated by the literature survey, fundamental experiments using simulated material and analyses using simulation code. This paper summarizes the feasibility confirmation results of FFCI.

Journal Articles

Decomposition behavior of gaseous ruthenium tetroxide under atmospheric conditions assuming evaporation to dryness accident of high-level liquid waste

Yoshida, Naoki; Ono, Takuya; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi

Journal of Nuclear Science and Technology, 57(11), p.1256 - 1264, 2020/11

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Emphasis has been placed on the behavior of ruthenium (Ru) in the evaporation to dryness accident due to the loss of cooling functions (EDLCF) of high-level liquid waste in fuel reprocessing plants. It is because Ru would form volatile compounds such as ruthenium tetroxide (RuO$$_{4}$$) and could be released into the environment with other coexisting gasses such as nitric acid (HNO$$_{3}$$), water (H$$_{2}$$O). To contribute to the safety evaluation of this accident, we experimentally evaluated the decomposition and chemical change behavior of the gaseous RuO$$_{4}$$ (RuO$$_{4}$$(g)) under the various atmospheric conditions: temperature and composition of coexisting gasses. As a result, the behavior of the RuO$$_{4}$$(g) was diverse depending on the atmospheric conditions. In the experiments with the dry air or H$$_{2}$$O vapor, decomposition of RuO$$_{4}$$(g) was observed. In the experiment with the mixed gas which containing HNO$$_{3}$$, almost no decomposition of the RuO$$_{4}$$(g) was observed, and chemical form of the RuO$$_{4}$$(g) was retained.

Journal Articles

Measurement of displacement cross-sections of copper and iron for proton with kinetic energies in the range 0.4 - 3 GeV

Matsuda, Hiroki; Meigo, Shinichiro; Iwamoto, Yosuke; Yoshida, Makoto*; Hasegawa, Shoichi; Maekawa, Fujio; Iwamoto, Hiroki; Nakamoto, Tatsushi*; Ishida, Taku*; Makimura, Shunsuke*

Journal of Nuclear Science and Technology, 57(10), p.1141 - 1151, 2020/10

 Times Cited Count:2 Percentile:8.37(Nuclear Science & Technology)

To estimate the structural damages of materials in accelerator facilities, displacement per atom (dpa) is widely employed as a damage index, calculated based on the displacement cross-section obtained using a calculation model. Although dpa is applied as standard, the experimental data of the displacement cross-section for a proton in the energy region above 20 MeV are scarce. Among the calculation models, difference of about factor 8 exist, so that the experimental data of the cross-section are crucial to validate the model. To obtain the displacement cross-section, we conducted experiments at J-PARC. The displacement cross-section of copper and iron was successfully obtained for a proton projectile with the kinetic energies, 0.4 - 3 GeV. The results were compared with those obtained using the widely utilized Norgertt-Robinson-Torrens (NRT) model and the athermal-recombination-corrected (arc) model based on molecular dynamics. It was found that the NRT model overestimates the present displacement cross-section by 3.5 times. The calculation results obtained using with the arc model based on the Nordlund parameter show remarkable agreement with the experimental data. It can be concluded that the arc model must be employed for the dpa calculation for the damage estimation of copper and iron.

Journal Articles

Comprehensive analysis and evaluation of Fukushima Daiichi Nuclear Power Station Unit 2

Yamashita, Takuya; Sato, Ikken; Honda, Takeshi*; Nozaki, Kenichiro*; Suzuki, Hiroyuki*; Pellegrini, M.*; Sakai, Takeshi*; Mizokami, Shinya*

Nuclear Technology, 206(10), p.1517 - 1537, 2020/10

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Investigation of high-temperature chemical interaction of calcium silicate insulation and cesium hydroxide

Rizaal, M.; Nakajima, Kunihisa; Saito, Takumi*; Osaka, Masahiko; Okamoto, Koji*

Journal of Nuclear Science and Technology, 57(9), p.1062 - 1073, 2020/09

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The interaction of cesium hydroxide and a calcium silicate insulation material was experimentally investigated at high temperature conditions. A thermogravimetry equipped with differential thermal analysis was used to analyze thermal events in the samples of mixed calcium silicate and cesium hydroxide under Ar-5%H$$_{2}$$ and Ar-4%H$$_{2}$$-20%H$$_{2}$$0 with maximum temperature of 1100$$^{circ}$$C. Prior being mixed with cesium hydroxide, a part of calcium silicate was pretreated at high temperature to evaluate the effect of possible structural changes of this material due to a preceding thermal history and also the sake of thermodynamic evaluation to those available ones. Based upon the initial condition (preliminary heat treatment) of calcium silicate, it was found that if the original material consisted of xonotlite (Ca$$_{6}$$Si$$_{6}$$0$$_{17}$$(0H)$$_{2}$$), the endothermic reaction with cesium hydroxide occurred over the temperature range 575-730$$^{circ}$$C meanwhile if the crystal phase of original material was changed to wollastonite (CaSi0$$_{3}$$), the interaction occurred over temperature range 700-1100$$^{circ}$$C. Furthermore, the X-ray diffraction analyses have indicated on both type of pretreated calsils that regardless of Ar-5%H$$_{2}$$ and Ar-4%H$$_{2}$$-20%H$$_{2}$$0 atmosphere, cesium aluminum silicate, CsAlSi0$$_{4}$$ was formed with aluminum in the samples as an impurity or adduct.

Journal Articles

Application of photoluminescence microspectroscopy; A Study on transfer of uranyl and europium ions on dry silica gel plate

Kusaka, Ryoji; Watanabe, Masayuki

Journal of Nuclear Science and Technology, 57(9), p.1046 - 1050, 2020/09

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Main findings, remaining uncertainties and lessons learned from the OECD/NEA BSAF Project

Pellegrini, M.*; Herranz, L.*; Sonnenkalb, M.*; Lind, T.*; Maruyama, Yu; Gauntt, R.*; Bixler, N.*; Morreale, A.*; Dolganov, K.*; Sevon, T.*; et al.

Nuclear Technology, 206(9), p.1449 - 1463, 2020/09

 Times Cited Count:3 Percentile:24.17(Nuclear Science & Technology)

Journal Articles

Utilizing PUNITA experiments to evaluate fundamental delayed gamma-ray spectroscopy interrogation requirements for nuclear safeguards

Rodriguez, D.; Koizumi, Mitsuo; Rossi, F.; Seya, Michio; Takahashi, Toon; Bogucarska, T.*; Crochemore, J.-M.*; Pedersen, B.*; Takamine, Jun

Journal of Nuclear Science and Technology, 57(8), p.975 - 988, 2020/08

 Times Cited Count:1 Percentile:24.17(Nuclear Science & Technology)

Journal Articles

A Linear Equation of characteristic time profile of power in subcritical quasi-steady state

Yamane, Yuichi

Journal of Nuclear Science and Technology, 57(8), p.926 - 931, 2020/08

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

An equation of power in subcritical quasi-steady state has been derived based on one-point kinetics equations for the purpose of utilizing it for the development of timely reactivity estimation from complicated time profile of neutron count rate. It linearly relates power, $$P$$, to a new variable $$q$$, which is a function of time differential of the power. It has been confirmed by using one-point kinetics code, AGNES, that the calculated points ($$q, P$$) are perfectly in a line described by the new equation and that points ($$q, P$$) calculated from transient subcritical experiments by using TRACY made a line with a slope indicated by the new equation.

Journal Articles

Generation of nuclear data using Gaussian process regression

Iwamoto, Hiroki

Journal of Nuclear Science and Technology, 57(8), p.932 - 938, 2020/08

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

We present a new approach to generate nuclear data from experimental cross section data by Gaussian process regression. This paper focuses on generating proton-induced nuclide production cross sections for nickel target. Our results provide reasonable fitting curves together with their uncertainties and suggest that this approach appears to be effective in generating or evaluating the nuclear data. Besides, our results suggest that our approach could be available for experimental design in terms of reducing the generated nuclear data uncertainty.

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