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Journal Articles

Crystal structures of ReO$$_3$$ under hydrostatic pressure; A Combined neutron, X-ray, Raman, and first-principles calculation study

Efthimiopoulos, I.*; Klotz, S.*; Kunc, K.*; Baptiste, B.*; Chauvigne, P.*; Hattori, Takanori

Physical Review B, 111(13), p.134103_1 - 134103_13, 2025/04

https://doi.org/10.1103/PhysRevB.111.134103
 Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)

We present a comprehensive study of the high pressure behaviour of ReO$$_3$$ using X-ray and neutron diffraction, Raman scattering and first-principles calculations to 15 GPa. We show that the ambient pressure $$Pm$$$=3$m$$ structure converts at 0.7 GPa in a continuous phase transition directly to a cubic phase with space group $$Im$$$=3 which is then stable up to at least 15 GPa. We show that previous reports of monoclinic $C2/c$$ and rhombohedral $$R$$$=3$c$$ structures in this pressure range are an artifact due to an alteration of the sample by high-flux synchrotron X-ray radiation. The structural pressure dependence of the $$Im$$$=3 phase is reported as well as the precise equation of state. Raman scattering data of both natural and isotopically enriched $^{18}$$O samples are presented. The data shed light onto the unusual transition and densification mechanism due to progressive tilting of essentially rigid ReO$$_6$$ octahedra.

Journal Articles

Updating fission product chemistry database based on recent investigation in Fukushima-Daiichi Nuclear Power Station, 3; High-temperature thermochemistry of CaCO$$_{3}$$-CsOH

Rizaal, M.; Luu, V. N.; Nakajima, Kunihisa; Miwa, Shuhei

Proceedings of International Topical Workshop on Fukushima Decommissioning Research 2024 (FDR2024) (Internet), 4 Pages, 2024/10

Journal Articles

Sintering and microstructural behaviors of mechanically blended Nd/Sm-doped MOX

Hirooka, Shun; Horii, Yuta; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*; Vauchy, R.; Hayashizaki, Kohei; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Science and Technology, 60(11), p.1313 - 1323, 2023/11

https://doi.org/10.1080/00223131.2023.2192725
 Times Cited Count：5 Percentile：80.64(Nuclear Science & Technology)

Additive MOX pellets are fabricated by a conventional dry powder metallurgy method. Nd$$_{2}$$O$$_{3}$$ and Sm$$_{2}$$O$$_{3}$$ are chosen as the additive materials to simulate the corresponding soluble fission products dispersed in MOX. Shrinkage curves of the MOX pellets are obtained by dilatometry, which reveal that the sintering temperature is shifted toward a value higher than that of the respective regular MOX. The additives, however, promote grain growth and densification, which can be explained by the effect of oxidized uranium cations covering to a pentavalent state. Ceramography reveals large agglomerates after sintering, and Electron Probe Micro-Analysis confirms that inhomogeneous elemental distribution, whereas XRD reveals a single face-centered cubic phase. Finally, by grinding and re-sintering the specimens, the cation distribution homogeneity is significantly improved, which can simulate spent nuclear fuels with soluble fission products.

Journal Articles

Extraction of $$^{99}$$Mo hot atoms made by a neutron capture method from $$alpha$$-MoO$$_{3}$$ to water

Quach, N. M.*; Ngo, M. C.*; Yang, Y.*; Nguyen, T. B.*; Nguyen, V. T.*; Fujita, Yoshitaka; Do, T. M. D.*; Nakayama, Tadachika*; Suzuki, Tatsuya*; Suematsu, Hisayuki*

Journal of Radioanalytical and Nuclear Chemistry, 332(10), p.4057 - 4064, 2023/10

https://doi.org/10.1007/s10967-023-09125-7
 Times Cited Count：4 Percentile：62.75(Chemistry, Analytical)

Technetium-99m ($$^{99m}$$Tc) is the most widely used medical radioisotope in the world and is produced from molybdenum-99 ($$^{99}$$Mo). Production of $$^{99}$$Mo via the neutron capture method draws attention as an alternative to fission-derived $$^{99}$$Mo due to non-proliferation issues, but the specific radioactivity of $$^{99}$$Mo is extremely low. In this work, a porous $$alpha$$-MoO$$_{3}$$ wire was prepared as an irradiation target in order to improve the specific activity by extracting $$^{99}$$Mo. Porous $$alpha$$-MoO$$_{3}$$ wire is synthesized from Mo metal wire by a two-step heating procedure. The hot atom effect of $$^{99}$$Mo was confirmed by activity and isotope measurements of the porous $$alpha$$-MoO$$_{3}$$ wire after neutron irradiation and the water used for extraction. In term of the extraction effectiveness, the effectiveness of $$^{99}$$Mo extraction in the porous $$alpha$$-MoO$$_{3}$$ wire was comparable to that of commercial $$alpha$$-MoO$$_{3}$$ powder.

Journal Articles

$$beta$$-MoO$$_{3}$$ whiskers in $$^{99}$$Mo/$$^{rm 99m}$$Tc radioisotope production and $$^{99}$$Mo/$$^{rm 99m}$$Tc extraction using hot atoms

Ngo, M. C.*; Fujita, Yoshitaka; Suzuki, Tatsuya*; Do, T. M. D.*; Seki, Misaki; Nakayama, Tadachika*; Niihara, Koichi*; Suematsu, Hisayuki*

Inorganic Chemistry, 62(32), p.13140 - 13147, 2023/08

https://doi.org/10.1021/acs.inorgchem.3c02125
 Times Cited Count：5 Percentile：54.36(Chemistry, Inorganic & Nuclear)

Technetium-99m ($$^{rm 99m}$$Tc) is one of the most important radioisotopes for diagnostic radio-imaging applications. $$^{rm 99m}$$Tc is a daughter product of the $$^{99}$$Mo isotope. There are two methods used to produce $$^{99}$$Mo/$$^{rm 99m}$$Tc: the nuclear fission (n,f) and the neutron capture (n,$$gamma$$) methods. Between them, the (n,f) method is the main route, used for approximately 90% of the world's production. However, the (n,f) method faces numerous problems, including the use of highly enriched uranium, the release of highly radioactive waste, and nonproliferation problems. Therefore, the (n,$$gamma$$) method is being developed as a future replacement for the (n,f) method. In this work, $$beta$$-MoO$$_{3}$$ whiskers prepared by the thermal evaporation method and $$alpha$$-MoO$$_{3}$$ particles were irradiated in a nuclear reactor to produce $$^{99}$$Mo/$$^{rm 99m}$$Tc via neutron capture. The irradiated targets were dispersed into water to extract the $$^{99}$$Mo/$$^{rm 99m}$$Tc. As a result, $$beta$$-MoO$$_{3}$$ whisker yielded higher $$^{99}$$Mo extraction rate than that from $$alpha$$-MoO$$_{3}$$. In addition, by comparing the dissolved $$^{98}$$Mo concentrations in water, we clarified a prominent hot-atom of $$beta$$-MoO$$_{3}$$ whiskers. This research is the first demonstration of $$beta$$-MoO$$_{3}$$ being used as an irradiation target in the neutron capture method. On the basis of the results, $$beta$$-MoO$$_{3}$$ is considered a promising irradiation target for producing $$^{99}$$Mo/$$^{rm 99m}$$Tc by neutron capture and using water for the radioisotope extraction process in the future.

Journal Articles

Study on chemical interaction between UO$$_{2}$$ and Zr at precisely controlled high temperatures

Shirasu, Noriko; Sato, Takumi; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki

Journal of Nuclear Science and Technology, 60(6), p.697 - 714, 2023/06

https://doi.org/10.1080/00223131.2022.2138602
 Times Cited Count：2 Percentile：46.61(Nuclear Science & Technology)

Interaction tests between UO$$_{2}$$ and Zr were performed at precisely controlled high temperatures between 1840 and 2000 $$^{circ}$$C to understand the interaction mechanism in detail. A Zr rod was inserted in a UO$$_{2}$$ crucible and then heat-treated at a fixed temperature in Ar-gas flow for 10 min. After heating in the range of 1890 to 1930 $$^{circ}$$C, the Zr rod was deformed to a round shape, in which the post-analysis detected the significant diffusion of U into the Zr region and the formation of a dominant $$alpha$$-Zr(O) matrix and a small amount of U-Zr-O precipitates. The abrupt progress of liquefaction was observed in the sample heated at around 1940 $$^{circ}$$C or higher. The higher oxygen concentration in the $$alpha$$-Zr(O) matrix suppressed the liquefaction progress, due to the variation in the equilibrium state. The U-Zr-O melt formation progressed by the selective dissolution of Zr from the matrix, and the selective diffusion of U could occur via the U-Zr-O melt.

JAEA Reports

Development of technology to simultaneously measure viscosity and surface tension of molten materials in reactor core (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Osaka University*

JAEA-Review 2021-046, 77 Pages, 2022/01

JAEA-Review-2021-046.pdf:2.92MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2018, this report summarizes the research results of the "Development of technology to simultaneously measure viscosity and surface tension of molten materials in reactor core" conducted from FY2018 to FY2020. Since the final year of this proposal was FY2020, the results for three fiscal years were summarized. Since (U, Zr)O$$_{2}$$ and boride, molten materials in reactor core, exist at extremely high temperature, chemical reactions between the vessel and these molten materials are unavoidable. Therefore, it is difficult to measure the thermophysical property of these materials. In the present study, droplets are produced by heating and melting the samples levitated by a gas levitation method, then the droplets are collided with a substrate.

JAEA Reports

Development of fundamental technologies for domestic production of medical radioisotope (technetium-99m); The First and second phase report (FY2014-2020)

Project 6 Meeting Members for Tsukuba International Strategic Zone

JAEA-Review 2021-016, 102 Pages, 2021/11

JAEA-Review-2021-016.pdf:12.76MB

In December 2011, the Prime Minister designated Tsukuba and some areas in Ibaraki Prefecture as "Comprehensive Special Zones". In the Tsukuba International Strategic Zone, nine advanced research and development (R&D) projects are underway with the goal of promoting industrialization of life innovation and green innovation utilizing the science and technology in Tsukuba. In these projects, the domestic production of medical radioisotope (Technetium-99m, $$^{rm 99m}$$Tc) was certified as a new project in October 2013, and R&D have been performed in collaboration with related organizations with Japan Atomic Energy Agency (JAEA) as the project leader. Japan is the third largest consumer of molybdenum-99 ($$^{99}$$Mo) after the United States and Europe, and all $$^{99}$$Mo are imported. Supply will be insufficient if overseas reactors are shut down due to trouble or if transportation (air and land transportations) is stopped due to volcanic eruptions and some accidents. Thus, early domestic production of $$^{99}$$Mo is strongly required. This project is a technology development aimed at domestic production of $$^{99}$$Mo, which is a raw material of $$^{rm 99m}$$Tc used as a diagnostic agent. This report summarizes the activities carried out in the first and second phase of the domestic production of medical radioisotope ($$^{rm 99m}$$Tc) (here referred to as the "Project 6") in Tsukuba International Strategic Zone (FY2014-2020).

Journal Articles

Fabrication, permeation, and corrosion stability measurements of silica membranes for HI decomposition in the thermochemical iodine-sulfur process

Myagmarjav, O.; Shibata, Ai*; Tanaka, Nobuyuki; Noguchi, Hiroki; Kubo, Shinji; Nomura, Mikihiro*; Takegami, Hiroaki

International Journal of Hydrogen Energy, 46(56), p.28435 - 28449, 2021/08

https://doi.org/10.1016/j.ijhydene.2021.06.079
 Times Cited Count：4 Percentile：13.75(Chemistry, Physical)

Journal Articles

Preventing nuclear fuel material adhesion on glove box components using nanoparticle coating

Segawa, Tomoomi; Kawaguchi, Koichi; Ishii, Katsunori; Suzuki, Masahiro; Tachihara, Joji; Takato, Kiyoto; Okita, Takatoshi; Satone, Hiroshi*; Suzuki, Michitaka*

Mechanical Engineering Journal (Internet), 8(3), p.21-00022_1 - 21-00022_9, 2021/06

https://doi.org/10.1299/mej.21-00022

To reduce the hold-up of the nuclear fuel materials in the glove box and the external exposure dose, the technology of the MOX powder adhesion prevention by the nanoparticle coating to the acrylic panels of the glove box has been developed. The surface analysis by means of atomic force microscopy (AFM) showed that the acrylic test piece surface coated with nanoparticles had a higher root mean square roughness value than that non-coated with nanoparticles. Due to the formation of nano-sized tiny rugged surface, the nanoparticle coating reduced the minimum adhesion force between the UO$$_{2}$$ particles and the acrylic test piece surface with the smallest particle size of about 5 $$mu$$m where desorption was observed, by about one-tenth. Moreover, the nanoparticle coating reduced the amount of the MOX powder adhering to the acrylic test piece to about one-tenth. In this study, it was found that applying the nanoparticle coating to the acrylic panels of glove box can prevent the adhesion of nuclear fuel materials. This method is effective for reducing the hold-up of the nuclear fuel materials in the glove box, the external exposure dose and improving the visibility of the acrylic panels.

JAEA Reports

Development of technology to simultaneously measure viscosity and surface tension of molten materials in reactor core (Contract research); FY2019 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Osaka University*

JAEA-Review 2020-038, 41 Pages, 2020/12

JAEA-Review-2020-038.pdf:3.28MB

JAEA/CLADS had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project in FY2019. Among the adopted proposals in FY2018, this report summarizes the research results of the "Development of Technology to Simultaneously Measure Viscosity and Surface Tension of Molten Materials in Reactor Core" conducted in FY2019.

Journal Articles

Evaluation of oxidation efficiency of hydrophobic palladium catalyst for $$^{3}$$H monitoring in radioactive gaseous waste

Furutani, Misa; Kometani, Tatsunari; Nakagawa, Masahiro; Ueno, Yumi; Sato, Junya; Iwai, Yasunori*

Hoken Butsuri (Internet), 55(2), p.97 - 101, 2020/06

https://doi.org/10.5453/jhps.55.97

Herein, an oxidation catalyst was introduced after heating it to 600$$^{circ}$$C to oxidize tritium gas (HT) existing in exhaust into tritiated water vapor (HTO). This study aims to establish a safer $$^{3}$$H monitoring system by lowering the heating temperature required for the catalyst. In these experiments, which were conducted in the Nuclear Science Research Institute, Japan Atomic Energy Agency, cupric oxide, hydrophobic palladium/silicon dioxide (Pd/SiO$$_{2}$$), and platinum/aluminum oxide (Pt/Al$$_{2}$$O$$_{3}$$) catalysts were ventilated using standard hydrogen gas. After comparing the oxidation efficiency of each catalyst at different temperatures, we found that the hydrophobic Pd/SiO$$_{2}$$ and Pt/Al$$_{2}$$O$$_{3}$$ catalysts could oxidize HT into HTO at 25$$^{circ}$$C.

JAEA Reports

Development of technology to simultaneously measure viscosity and surface tension of molten materials in reactor core (Contract research); FY2018 Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development

Collaborative Laboratories for Advanced Decommissioning Science; Osaka University*

JAEA-Review 2019-025, 36 Pages, 2020/01

JAEA-Review-2019-025.pdf:2.57MB

CLADS, JAEA, had been conducting the Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development (hereafter referred to "the Project") in FY2018. The Project aims to contribute to solving problems in nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2018, this report summarizes the research results of the "Development of Technology to Simultaneously Measure Viscosity and Surface Tension of Molten Materials in Reactor Core". Since (U,Zr)O$$_{2}$$ and boride, molten materials in reactor core, exist at extremely high temperature, chemical reactions between the vessel and these molten materials are unavoidable. Therefore, it is difficult to measure the thermophysical property of these materials. In the present study, droplets are produced by heating and melting the samples levitated by a gas levitation method, then the droplets are collided with a substrate. From the instant behavior of the collision, a new technology to simultaneously derive the viscosity and surface tension will be developed.

Journal Articles

Evaluation of tritium release behavior from Li$$_{2}$$TiO$$_{3}$$ during DT neutron irradiation by use of an improved tritium collection method

Edao, Yuki; Kawamura, Yoshinori; Hoshino, Tsuyoshi; Ochiai, Kentaro

Fusion Engineering and Design, 112, p.480 - 485, 2016/11

https://doi.org/10.1016/j.fusengdes.2016.06.013
 Times Cited Count：8 Percentile：51.02(Nuclear Science & Technology)

The accurate measurement of behavior of bred tritium released from a tritium breeder is indispensable to understand the behavior for a design of a tritium extraction system. The tritium collection method combined a CuO bed and water bubbles was not suitable to measure transient behavior of tritium released from Li$$_{2}$$TiO$$_{3}$$ during neutron irradiation because tritium released behavior was changed to be delayed due to adsorption of oxidized tritium on the CuO. Hence, the tritium collection method with hydrophobic catalyst instead of the CuO was demonstrated and succeeded the accurate release measurement of tritium from Li$$_{2}$$TiO$$_{3}$$. With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li$$_{2}$$TiO$$_{3}$$ as the form of HT as much as possible from the view point of the fuel cycle. Our results indicated; promotion of isotopic exchange reaction on the surface of Li$$_{2}$$TiO$$_{3}$$ by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li$$_{2}$$TiO$$_{3}$$ irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li$$_{2}$$TiO$$_{3}$$ was the dominant parameter to control the chemical form of tritium released from the Li$$_{2}$$TiO$$_{3}$$.

Journal Articles

Evaluation of oxidation efficiency of hydrophobic palladium catalyst for $$^{14}$$C monitoring in gaseous radioactive waste

Ueno, Yumi; Nakagawa, Masahiro; Sato, Junya; Iwai, Yasunori

Hoken Butsuri, 51(1), p.7 - 11, 2016/03

https://doi.org/10.5453/jhps.51.7

In the Nuclear Science Research Institute, Japan Atomic Energy Agency (JAEA), in order to oxidize $$^{14}$$C, which exists in various chemical forms in exhaust, into $$^{14}$$CO$$_{2}$$, a copper oxide (CuO) catalyst is introduced after heating to 600$$^{circ}$$C. Our goal was to establish a safer $$^{14}$$C monitoring system by lowering the heating temperature required for the catalyst; therefore, we developed a new hydrophobic palladium/silicon dioxide (Pd/SiO$$_{2}$$) catalyst that makes the carrier's surface hydrophobic. In these experiments, catalysts CuO, platinum/aluminum oxide (Pt/Al$$_{2}$$O$$_{3}$$), palladium/zirconium dioxide (Pd/ZrO$$_{2}$$), hydrophobic Pd/SiO$$_{2}$$, and hydrophilic Pd/SiO$$_{2}$$ were ventilated with standard methane gas, and we compared the oxidation efficiency of each catalyst at different temperatures. As a result, we determined that the hydrophobic Pd/SiO$$_{2}$$ catalyst had the best oxidation efficiency. By substituting the currently used CuO catalyst with the hydrophobic Pd/SiO$$_{2}$$ catalyst, we will be able to lower the working temperature from 600$$^{circ}$$C to 300$$^{circ}$$C and improve the safety of the monitoring process.

Journal Articles

JAEA-ISCN development programs of advanced NDA technologies of nuclear material

Seya, Michio; Kobayashi, Naoki; Naoi, Yosuke; Hajima, Ryoichi; Soyama, Kazuhiko; Kureta, Masatoshi; Nakamura, Hironobu; Harada, Hideo

Book of Abstracts, Presentations and Papers of Symposium on International Safeguards; Linking Strategy, Implementation and People (Internet), 8 Pages, 2015/03

https://www.iaea.org/safeguards/symposium/2014/home/eproceedings/sg2014-papers/000069.pdf
https://www.iaea.org/safeguards/symposium/2014/home/eproceedings/sg2014-slides/000069.pdf

JAEA-ISCN has been implementing basic development programs of the advanced NDA technologies for nuclear material (NM) since 2011JFY (Japanese Fiscal Year), which are (1) NRF (Nuclear resonance fluorescence) NDA technology using laser Compton scattered (LCS) $$gamma$$-rays (intense mono-energetic $$gamma$$-rays), (2) Alternative to $$^{3}$$He neutron detection technology using ZnS/B$$_{2}$$O$$_{3}$$ ceramic scintillator, and (3) NRD (Neutron resonance densitometry) using NRTA (Neutron resonance transmission analysis) and NRCA (Neutron resonance capture analysis). These programs are going to be finished in 2014JFY and have demonstration tests in February - March 2015.

JAEA Reports

Fabrication technology development and characterization of irradiation targets for $$^{99}$$Mo/$$^{rm 99m}$$Tc production by (n,$$gamma$$) method

Nishikata, Kaori; Kimura, Akihiro; Ishida, Takuya; Shiina, Takayuki*; Ota, Akio*; Tanase, Masakazu*; Tsuchiya, Kunihiko

JAEA-Technology 2014-034, 34 Pages, 2014/10

JAEA-Technology-2014-034.pdf:3.26MB

As a part of utilization expansion after the Japan Material Testing Reactor (JMTR) re-start, research and development (R&D) on the production of medical radioisotope $$^{99}$$Mo/$$^{99m}$$Tc by (n, $$gamma$$) method using JMTR has been carried out in the Neutron Irradiation and Testing Reactor Center of the Japan Atomic Energy Agency. $$^{99}$$Mo is usually produced by fission method. On the other hand, $$^{99}$$Mo/$$^{99m}$$Tc production by the (n, $$gamma$$) method has advantages for radioactive waste, cost reduction and non-proliferation. However, the specific radioactivity per unit volume by the (n, $$gamma$$) method is low compared with the fission method, and that is the weak point of the (n, $$gamma$$) method. This report summarizes the investigation of raw materials, the fabrication tests of high-density MoO$$_{3}$$ pellets by the plasma sintering method for increasing of $$^{98}$$Mo contents and the characterization of sintered high-density MoO$$_{3}$$ pellets.

Journal Articles

Introduction to development of advanced safeguards and security NDA technologies by JAEA-ISCN

Seya, Michio; Kureta, Masatoshi; Soyama, Kazuhiko; Nakamura, Hironobu; Harada, Hideo; Hajima, Ryoichi

Proceedings of INMM 55th Annual Meeting (Internet), 10 Pages, 2014/07

JAEA has been implementing development programs of basic technologies of the following advanced NDA (non-destructive assay) of nuclear material (NM) for nuclear safeguards and security. (1) Alternative to $$^{3}$$He neutron detection using ZnS/B$$_{2}$$O$$_{3}$$ ceramic scintillator, (2) NRD (neutron resonance densitometry) using NRTA (neutron resonance transmission analysis) and NRCA (neutron resonance capture analysis), (3) NRF (nuclear resonance fluorescence)-NDA using laser Compton scattered (LCS) $$gamma$$-rays (intense mono-energetic $$gamma$$-rays). The development program (1) is for NDA systems that use ZnS/B$$_{2}$$O$$_{3}$$ ceramic scintillator as alternative neutron detector to $$^{3}$$He for coming shortage of its supply. The program (2) is for a NDA system of isotopic composition measurement (non-destructive mass spectroscopy) in targets such as particle-like melted fuel debris using NRTA and NRCA. The program (3) is for NDA systems using a specific NRF reaction of certain Pu/U isotope caused by mono-energetic LCS $$gamma$$-ray with energy tuned to the specific excited state of the isotope. This paper introduces above three programs.

Journal Articles

High energy heavy ion induced structural disorder in Li$$_{2}$$TiO$$_{3}$$

Nakazawa, Tetsuya; Naito, Akira*; Aruga, Takeo; Grismanovs, V.*; Chimi, Yasuhiro; Iwase, Akihiro*; Jitsukawa, Shiro

Journal of Nuclear Materials, 367-370(2), p.1398 - 1403, 2007/08

https://doi.org/10.1016/j.jnucmat.2007.04.003
 Times Cited Count：45 Percentile：92.74(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Evaluation of contact strength of Li$$_{2}$$TiO$$_{3}$$ pebbles with different diameters

Tsuchiya, Kunihiko; Kawamura, Hiroshi; Tanaka, Satoru*

Fusion Engineering and Design, 81(8-14), p.1065 - 1069, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.07.031
 Times Cited Count：11 Percentile：59.03(Nuclear Science & Technology)

no abstracts in English

191 (Records 1-20 displayed on this page)