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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. From the instant behavior of

Journal Articles

Nuclide production cross sections of Ni and Zr irradiated with 0.4-, 1.3-, 2.2-, and 3.0-GeV protons

Takeshita, Hayato; Meigo, Shinichiro; Matsuda, Hiroki; Iwamoto, Hiroki; Maekawa, Fujio; Watanabe, Yukinobu*

JPS Conference Proceedings (Internet), 33, p.011045_1 - 011045_6, 2021/03

To improve accuracy of nuclear design of accelerator driven nuclear transmutation systems, nuclide production cross sections on Ni and Zr, which were candidate materials to be used in ADS, were measured for GeV energy protons. The measured results were compared with PHITS calculations and JENDL/HE-2007.

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:0.01(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

Development of HCl-free solid phase extraction combined with ICP-MS/MS for rapid assessment of difficult-to-measure radionuclides, 1; Selective measurement of $$^{93}$$Zr and $$^{93}$$Mo in concrete rubble

Do, V. K.; Furuse, Takahiro; Murakami, Erina; Aita, Rena; Ota, Yuki; Sato, Soichi

Journal of Radioanalytical and Nuclear Chemistry, 327(1), p.543 - 553, 2021/01

 Times Cited Count:0 Percentile:0.01(Chemistry, Analytical)

A new HCl-free chromatographic separation procedure has been developed for sequential separation of Zr and Mo from concrete matrices. Accordingly, $$^{93}$$Zr and $$^{93}$$Mo could be sensitively and selectively measured by ICP-MS/MS using ammonia reaction gas. The recoveries of greater than 90% for Zr and Mo from concretes could be achieved. The measurement condition was optimized for complete suppression of interferences from $$^{93}$$Nb and peak tailing from abundant isotopes of Zr and Mo in concrete matrices. The removal of interferences was verified by measurement of radio-contamination-free concretes used as a sample matrix blank. Method detection limits of 1.7 mBq g$$^{-1}$$ and 0.2 Bq g$$^{-1}$$ were achieved for $$^{93}$$Zr and $$^{93}$$Mo, respectively, in the concrete matrices. The interference removal factor for Nb (equivalent to the decontamination factor in radiochemical separation) was of the order of 10$$^{5}$$, and the abundance sensitivity was of the order of 10$$^{-8}$$, indicating that the developed method is reliable for verifying the presence of ultralow concentrations of $$^{93}$$Zr and $$^{93}$$Mo. The present method is suitable for the rapid assessment of $$^{93}$$Zr and $$^{93}$$Mo for radioactivity inventory of concrete rubble.

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

Microscopic analyses on Zr adsorbed IDA chelating resin by PIXE and EXAFS

Arai, Yoichi; Watanabe, So; Ono, Shimpei; Nomura, Kazunori; Nakamura, Fumiya*; Arai, Tsuyoshi*; Seko, Noriaki*; Hoshina, Hiroyuki*; Hagura, Naoto*; Kubota, Toshio*

Nuclear Instruments and Methods in Physics Research B, 477, p.54 - 59, 2020/08

 Times Cited Count:0 Percentile:0.02(Instruments & Instrumentation)

Journal Articles

Quantitative analysis of Zr adsorbed on IDA chelating resin using Micro-PIXE

Arai, Yoichi; Watanabe, So; Ono, Shimpei; Nomura, Kazunori; Nakamura, Fumiya*; Arai, Tsuyoshi*; Seko, Noriaki*; Hoshina, Hiroyuki*; Kubota, Toshio*

QST-M-23; QST Takasaki Annual Report 2018, P. 59, 2020/03

Journal Articles

Study on plutonium burner high temperature gas-cooled reactor in Japan; Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.

Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02

 Times Cited Count:1 Percentile:39.17(Nuclear Science & Technology)

The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. 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. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.

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.

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

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

Journal Articles

Structure change of monoclinic ZrO$$_{2}$$ baddeleyite involving softenings of bulk modulus and atom vibrations

Fukui, Hiroshi*; Fujimoto, Manato*; Akahama, Yuichi*; Sano, Asami; Hattori, Takanori

Acta Crystallographica Section B; Structural Science, Crystal Engineering and Materials (Internet), 75(4), p.742 - 749, 2019/08

 Times Cited Count:2 Percentile:38.26(Chemistry, Multidisciplinary)

Monoclinic ZrO$$_{2}$$ baddeleyite exhibits anomalous softenings of bulk modulus and atom vibrations with compression. We have investigated the pressure evolution of the structure by neutron powder diffraction combined with ab-initio calculations. The present results showed that the anomalous pressure response of the bulk modulus is related not to the change in the bonding characters but to the deformation of an oxygen sublattice, especially one of layers made of oxygens in the crystallographic $$a$$* plane. The layer consists of two parallelograms; one is rotating with little distortion and the other is being distorted with increasing pressure. This deformation of this layer makes one of Zr-O distances long, resulting in the softening of some atom vibrational modes.

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

 Times Cited Count:2 Percentile:42.44(Materials Science, Multidisciplinary)

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

High-temperature interaction between zirconium and UO$$_2$$

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

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

High temperature interaction tests between UO$$_{2}$$ and Zr were performed at around 2173 K, to make clear the UO$$_{2}$$/ $$alpha$$-Zr(O) interaction and the mechanism of degradation, for developing the improved models for advanced severe accident analysis codes. A Zr plate was inserted in a UO$$_{2}$$ crucible, and heat treated at 2173 K in stream of Ar. After the heat-treatment, the samples were subjected to surface microanalysis. The middle region of Zr sample shows streak-like structures which are extended towered the top. It is confirmed that the streak-like structures were mainly consist of U from the EDX results, and the structures revealed that the U-rich phase was liquid during the heat-treatment. It seems that the U-rich liquid grew selectively toward the area where the oxygen concentration was low.

Journal Articles

Analysis on adsorbent for spent solvent treatment by micro-PIXE and EXAFS

Arai, Yoichi; Watanabe, So; Ono, Shimpei; Nakamura, Masahiro; Shibata, Atsuhiro; Nakamura, Fumiya*; Arai, Tsuyoshi*; Seko, Noriaki*; Hoshina, Hiroyuki*; Hagura, Naoto*; et al.

International Journal of PIXE, 29(01&02), p.17 - 31, 2019/00

The spent PUREX solvent containing U and Pu is generated from the reprocessing process of spent nuclear fuel. The nuclear material removal is important for the safe storage or disposal of the spent solvent. Our previous study revealed that the adsorbent with the iminodiacetic acid (IDA) functional group is one of the most promising materials for designing the nuclear material recovery process. Accordingly, an IDA-type adsorbent was synthesized by using graft polymerization technology or a chemical reaction to improve the adsorption rate and capacity. The synthesized IDA-type adsorbent was characterized by micro particle-induced X-ray emission (PIXE) and extended X-ray absorption fine structure (EXAFS) analyses. The micro-PIXE analysis revealed that Zr was adsorbed on the whole synthesized adsorbents and quantified the microamount of adsorbed Zr. Moreover, EXAFS suggested that Zr in the aqueous solution and solvent can be trapped by the IDA group with different mechanisms.

Journal Articles

Dissolution and chemical analysis of Zr-based lanthanide nitrides

Hayashi, Hirokazu; Chiba, Rikiya*

Progress in Nuclear Science and Technology (Internet), 5, p.196 - 199, 2018/11

Uranium-free nitride fuel has been chosen as the first candidate for transmutation of long-lived minor actinides (MA: Np, Am, Cm) using sub-critical accelerator-driven system (ADS) under the double strata fuel cycle concept by Japan Atomic Energy Agency (JAEA). Dissolution behavior of ZrN-based nitrides in nitric acid is examined using lanthanides as surrogate materials of TRU elements. Chemical analysis of the ZrN-based lanthanide nitrides dissolved in nitric acid is also carried out.

Journal Articles

Rapid separation of zirconium using microvolume anion-exchange cartridge for $$^{93}$$Zr determination with isotope dilution ICP-MS

Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*

Talanta, 185, p.98 - 105, 2018/08

 Times Cited Count:7 Percentile:45.32(Chemistry, Analytical)

Estimating the risks associated with radiation from long-lived fission products (LLFP) in radioactive waste is essential to ensure the long-term safety of potential disposal sites. In this study, the amount of $$^{93}$$Zr, a LLFP, was determined by ICP-MS after separating Zr from a spent nuclear fuel solution using a microvolume anion-exchange cartridge (TEDA cartridge). The TEDA cartridge achieved highly selective separation of Zr regardless of its small bed volume of 0.08 cm$$^{3}$$. The time taken to complete the Zr separation was 1.2 min with a flow rate of 1.5 mL/min, which was 10 times faster than that for a conventional anion-exchange resin column. Almost all the other elements were removed, leading to accurate measurement of $$^{93}$$Zr. The result connects experimental value to theoretical prediction provided by ORIGEN2, which requires verification. With the measured value, we demonstrated that the theoretical value is reliable enough to estimate radiation risks.

Journal Articles

A Thermodynamic model for ZrO$$_{2}$$(am) solubility at 25$$^{circ}$$C in the Ca$$^{2+}$$-Na$$^{+}$$-H$$^{+}$$-Cl$$^{-}$$-OH$$^{-}$$-H$$_{2}$$O system; A Critical review

Rai, D.*; Kitamura, Akira; Altmaier, M.*; Rosso, K. M.*; Sasaki, Takayuki*; Kobayashi, Taishi*

Journal of Solution Chemistry, 47(5), p.855 - 891, 2018/05

 Times Cited Count:5 Percentile:8.63(Chemistry, Physical)

We have critically reviewed experimental data for Zr hydrolysis constant values for formation of several mononuclear and polynuclear species and a solubility product value for ZrO$$_{2}$$(am). We have determined new/revised values for the formation constants of Zr(OH)$$_{2}$$$$^{2+}$$, Zr(OH)$$_{4}$$(aq), Zr(OH)$$_{5}$$$$^{-}$$, Zr(OH)$$_{6}$$$$^{2-}$$ and Ca$$_{3}$$Zr(OH)$$_{6}$$$$^{4+}$$, and the solubility product for ZrO$$_{2}$$(am) after the critical review.

Journal Articles

The Influence of the air fraction in steam on the growth of the columnar oxide and the adjacent $$alpha$$-Zr(O) layer on Zry-4 fuel cladding at 1273 and 1473 K

Negyesi, M.; Amaya, Masaki

Annals of Nuclear Energy, 114, p.52 - 65, 2018/04

 Times Cited Count:3 Percentile:45.99(Nuclear Science & Technology)

JAEA Reports

Corrosion test of Fugen pressure tube (Zr-2.5wt%Nb alloy) under the sub-surface disposal environment, 2; Examination of long-term corrosion rate by 5 years keeping sample

Sugaya, Toshikatsu; Nakatani, Takayoshi; Sakai, Akihiro

JAEA-Technology 2017-032, 21 Pages, 2018/01

[The article has been found to have a problem about reliability of the corrosion data acquisition, and thus it is unavailable to download the full text in accordance with authors' intentions to retract the report.] For the purpose of the setting of the rate of nuclide elution necessary to safety assessment, we planned the gas-accumulating type corrosion test on Zr-2.5wt%Nb alloy in order to obtain long-term corrosion rate under low temperature, low oxygen and alkaline conditions assuming the disposal environment. A corrosion rate over a testing period of 5 years is acquired with the aim to grasp a long-term corrosion rate behavior in this report. This corrosion rate is compared with the same data that was previously acquired over a testing period of 2 years. As a result, it is confirmed that an evaluation method that is proportional to the minus cubic root of corrosion time squared can be applicable to the corrosion rate behavior acquired this time over a testing period of 5 years, which is the same result in evaluating the corrosion rate behavior acquired over a testing period of 2 years.

214 (Records 1-20 displayed on this page)