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Journal Articles

Evaluation of power distribution calculation of the very high temperature reactor critical assembly (VHTRC) with Monte Carlo MVP3 code

Simanullang, I. L.*; Nakagawa, Naoki*; Ho, H. Q.; Nagasumi, Satoru; Ishitsuka, Etsuo; Iigaki, Kazuhiko; Fujimoto, Nozomu*

Annals of Nuclear Energy, 177, p.109314_1 - 109314_8, 2022/11

Journal Articles

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

Lu, K.; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.

International Journal of Pressure Vessels and Piping, 199, p.104706_1 - 104706_13, 2022/10

 Times Cited Count:1 Percentile:0.01(Engineering, Multidisciplinary)

Journal Articles

Characterization of bremsstrahlung and $$gamma$$-rays of fuel debris

Matsumura, Taichi; Okumura, Keisuke; Fujita, Manabu*; Sakamoto, Masahiro; Terashima, Kenichi; Riyana, E. S.

Radiation Physics and Chemistry, 199, p.110298_1 - 110298_8, 2022/10

Journal Articles

Study on the relation between the crystal structure and thermal stability of FeUO$$_{4}$$ and CrUO$$_{4}$$

Akiyama, Daisuke*; Kusaka, Ryoji; Kumagai, Yuta; Nakada, Masami; Watanabe, Masayuki; Okamoto, Yoshihiro; Nagai, Takayuki; Sato, Nobuaki*; Kirishima, Akira*

Journal of Nuclear Materials, 568, p.153847_1 - 153847_10, 2022/09

FeUO$$_{4}$$, CrUO$$_{4}$$, and Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ are monouranates containing pentavalent U. Even though these compounds have similar crystal structures, their formation conditions and thermal stability are significantly different. To determine the factors causing the difference in thermal stability between FeUO$$_{4}$$ and CrUO$$_{4}$$, their crystal structures were evaluated in detail. A Raman band was observed at 700 cm$$^{-1}$$ in all the samples. This Raman band was derived from the stretching vibration of the O-U-O axis band, indicating that Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ was composed of a uranyl-like structure in its lattice regardless of its "x"' value. M$"o$ssbauer measurements indicated that the Fe in FeUO$$_{4}$$ and Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ were trivalent. Furthermore, Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ lost its symmetry around Fe$$^{mathrm{III}}$$ with increasing electron densities around Fe$$^{mathrm{III}}$$, as the abundance of Cr increased. These results suggested no significant structural differences between FeUO$$_{4}$$ and CrUO$$_{4}$$. Thermogravimetric measurements for UO$$_{2}$$, FeUO$$_{4}$$, and CrUO$$_{4}$$ showed that the temperature at which FeUO$$_{4}$$ decomposed under an oxidizing condition (approximately 800 $$^{circ}$$C) was significantly lower than the temperature at which the decomposition of CrUO$$_{4}$$ started (approximately 1250 $$^{circ}$$C). Based on these results, we concluded that the decomposition of FeUO$$_{4}$$ was triggered by an ``in-crystal'' redox reaction, i.e., Fe$$^{mathrm{III}}$$ $${+}$$ U$$^{mathrm{V}}$$ $$rightarrow$$ Fe$$^{mathrm{II}}$$ $${+}$$ U$$^{mathrm{VI}}$$, which would not occur in the CrUO$$_{4}$$ lattice because Cr$$^{mathrm{III}}$$ could never be reduced under the investigated condition. Finally, the existence of Cr$$^{mathrm{III}}$$ in FexCr$$_{1-x}$$UO$$_{4}$$ effectively suppressed the decomposition of the Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ crystal, even at a very low Cr content.

Journal Articles

Calculation of shutdown gamma distribution in the high temperature engineering test reactor

Ho, H. Q.; Ishii, Toshiaki; Nagasumi, Satoru; Ono, Masato; Shimazaki, Yosuke; Ishitsuka, Etsuo; Goto, Minoru; Simanullang, I. L.*; Fujimoto, Nozomu*; Iigaki, Kazuhiko

Nuclear Engineering and Design, 396, p.111913_1 - 111913_9, 2022/09

Journal Articles

Measurement of nuclide production cross sections for proton-induced reactions on $$^{rm nat}$$Ni and $$^{rm nat}$$Zr at 0.4, 1.3, 2.2, and 3.0 GeV

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

Nuclear Instruments and Methods in Physics Research B, 527, p.17 - 27, 2022/09

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

Journal Articles

Calculating off-axis efficiency of coaxial HPGe detectors by Monte Carlo simulation

Omer, M.; Shizuma, Toshiyuki*; Hajima, Ryoichi*; Koizumi, Mitsuo

Radiation Physics and Chemistry, 198, p.110241_1 - 110241_7, 2022/09

JAEA Reports

Radiation tolerant rapid criticality monitoring with radiation-hardened FPGAs (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Okayama University*

JAEA-Review 2022-017, 56 Pages, 2022/08

JAEA-Review-2022-017.pdf:6.39MB

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 FY2019, this report summarizes the research results of the "Radiation tolerant rapid criticality monitoring with radiation-hardened FPGAs" conducted in FY2020. This research is developing a radiation-hardened optoelectronic FPGA with a 1 Grad total-ionizing-dose tolerance on which optical technologies are introduced onto a semiconductor technology and a radiation hardened FPGA with a 200 Mrad total-ionizing-dose tolerance not using any optical component. Moreover, Japanese research group will support hardware acceleration on FPGAs used for neutron-detection system developed by UK team. Finally, we will provide our radiation-hardened FPGA for the UK neutron-detection system.

JAEA Reports

Document collection of the Special Committee on HTTR Heat Application Test

Aoki, Takeshi; Shimizu, Atsushi; Iigaki, Kazuhiko; Okita, Shoichiro; Hasegawa, Takeshi; Mizuta, Naoki; Sato, Hiroyuki; Sakaba, Nariaki

JAEA-Review 2022-016, 193 Pages, 2022/08

JAEA-Review-2022-016.pdf:42.06MB

Aiming to realize a massive, cost-effective and carbon-free hydrogen production technology utilizing a high temperature gas cooled reactor (HTGR), Japan Atomic Energy Agency (JAEA) is planning a HTTR heat application test producing hydrogen with High Temperature Engineering Test Reactor (HTTR) achieved 950$$^{circ}$$C of the highest reactor outlet coolant temperature in the world. In the HTTR heat application test, it is required to establish its safety design realizing highly safe connection of a HTGR and a hydrogen production plant by the Nuclear Regulation Authority to obtain the permission of changes to reactor installation. However, installation of a system connecting the hydrogen production plant and a nuclear reactor, and its safety design has not been conducted so far in conventional nuclear power plant including HTTR in the world. A special committee on the HTTR heat application test, established under the HTGR Research and Development Center, considered a safety design philosophy for the HTTR heat application test based on an authorized safety design of HTTR in terms of conformity to the New Regulatory Requirements taking into account new considerable events as a result of the plant modification and connection of the hydrogen production plant. This report provides materials of the special committee such as technical reports, comments provided from committee members, response from JAEA for the comments and minutes of the committee.

JAEA Reports

Study of natural fracture topography in 50cm scale granodiorite rock block including natural fracture intersection measured by precision grinder

Tetsu, Keiichi*; Takayama, Yusuke; Sawada, Atsushi

JAEA-Research 2022-005, 84 Pages, 2022/08

JAEA-Research-2022-005.pdf:10.32MB
JAEA-Research-2022-005-appendix(CD-ROM).zip:35.68MB

Nuclide migration analyses are conducted for the safety assessment in geological disposal system of HLW. Nuclide migration evaluation of fractured host rock mainly uses a model that approximates the fracture with parallel flat plates. However, the actual fractures in the host rock are different from the parallel flat plates, the fractures have complex characteristics such as roughness of fracture surface and the fillings in the fracture. In approximating a fracture model, the methodology development how to set parameter values such as the transmissivity coefficient and the fracture aperture is an issue. One of the issues is to investigate the geometrical features of actual fractures in the host rock. In this study, for the purpose of understanding the geometrical features of fractures including fracture intersection, the internal fracture shape was measured in detail using the method of surface grinding on 50cm scale granodiorite with natural fracture intersection. Thus, the fracture width, fracture aperture, and the shape of the fracture surfaces were obtained. From the obtain data, characteristics such as the average value of the fracture width, the roughness of the fracture surfaces, and the distribution of the fracture aperture were evaluated. As a result, it was confirmed that one of the fractures near the fracture intersection has a particularly large fracture width and fracture aperture as compared with the other part of fractures. Thus, in the granodiorite rock block used in this study, it was inferred that the most permeable path is not the fracture intersection itself, but the particularly large fracture aperture in the vicinity of the fracture intersection.

Journal Articles

BWR lower head penetration failure test focusing on eutectic melting

Yamashita, Takuya; Sato, Takumi; Madokoro, Hiroshi; Nagae, Yuji

Annals of Nuclear Energy, 173, p.109129_1 - 109129_15, 2022/08

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

Journal Articles

Structure, stability, and actinide leaching of simulated nuclear fuel debris synthesized from UO$$_{2}$$, Zr, and stainless-steel

Kirishima, Akira*; Akiyama, Daisuke*; Kumagai, Yuta; Kusaka, Ryoji; Nakada, Masami; Watanabe, Masayuki; Sasaki, Takayuki*; Sato, Nobuaki*

Journal of Nuclear Materials, 567, p.153842_1 - 153842_15, 2022/08

 Times Cited Count:0

To understand the chemical structure and stability of nuclear fuel debris consisting of UO$$_{2}$$, Zr, and Stainless Steel (SUS) generated by the Fukushima Daiichi Nuclear Power Plant accident in Japan in 2011, simulated debris of the UO$$_{2}$$-SUS-Zr system and other fundamental component systems were synthesized and characterized. The simulated debris were synthesized by heat treatment for 1 to 12 h at 1600$$^{circ}$$C, in inert (Ar) or oxidative (Ar + 2% O$$_{2}$$) atmospheres. $$^{237}$$Np and $$^{241}$$Am tracers were doped for the leaching tests of these elements and U from the simulated debris. The characterization of the simulated debris was conducted by XRD, SEM-EDX, Raman spectroscopy, and M$"o$ssbauer spectroscopy, which provided the major uranium phase of the UO $$_{2}$$-SUS-Zr debris was the solid solution of U$$^{mathrm{IV}}$$O$$_{2}$$ (s.s.) with Zr(IV) and Fe(II) regardless of the treatment atmosphere. The long-term immersion test of the simulated debris in pure water and that in seawater revealed the macro scale crystal structure of the simulated debris was chemically very stable in the wet condition for a year or more. Furthermore, the leaching test results showed that the actinide leaching ratios of U, Np, Am from the UO$$_{2}$$-SUS-Zr debris were very limited and less than 0.08 % for all the experiments in this study.

JAEA Reports

Calculation of nuclear core parameters for HTTR; Report of summer holiday practical training 2021

Isogawa, Hiroki*; Naoi, Motomasa*; Yamasaki, Seiji*; Ho, H. Q.; Katayama, Kazunari*; Matsuura, Hideaki*; Fujimoto, Nozomu*; Ishitsuka, Etsuo

JAEA-Technology 2022-015, 18 Pages, 2022/07

JAEA-Technology-2022-015.pdf:1.37MB

As a summer holiday practical training 2021, the impact of 10 years long-term shutdown on critical control rod position of the HTTR and the delayed neutron fraction ($$beta$$$$_{rm eff}$$) of the VHTRC-1 core were investigated using Monte-Carlo MVP code. As a result, a long-term shutdown of 10 years caused the critical control rods of the HTTR to withdraw about 4.0$$pm$$0.8 cm compared to 3.9 cm in the experiment. The change in critical control rods position of the HTTR is due to the change of some fission products such as $$^{241}$$Pu, $$^{241}$$Am, $$^{147}$$Pm, $$^{147}$$Sm, $$^{155}$$Gd. Regarding the $$beta$$$$_{rm eff}$$ calculation of the VHTRC-1 core, the $$beta$$$$_{rm eff}$$ value is underestimate of about 10% in comparison with the experiment value.

JAEA Reports

Safety design philosophy of HTTR Heat Application Test Facility

Aoki, Takeshi; Shimizu, Atsushi; Iigaki, Kazuhiko; Okita, Shoichiro; Hasegawa, Takeshi; Mizuta, Naoki; Sato, Hiroyuki; Sakaba, Nariaki

JAEA-Technology 2022-011, 60 Pages, 2022/07

JAEA-Technology-2022-011.pdf:2.08MB

Japan Atomic Energy Agency is planning a High Temperature Engineering Test Reactor (HTTR) heat application test producing hydrogen with the HTTR which achieved the highest reactor outlet coolant temperature of 950$$^{circ}$$C in the world to realize a massive, cost-effective and carbon-free hydrogen production technology utilizing a high temperature gas cooled reactor (HTGR). In the HTTR heat application test, it is required to establish its safety design for coupling a hydrogen production plant to HTGR through the licensing by the Nuclear Regulation Authority (NRA). A draft of a safety design philosophy for the HTTR heat application test facility was considered taking into account postulated events due to the plant modification and coupling of the hydrogen production plant based on the HTTR safety design which was authorized through the safety review of the NRA against New Regulatory Requirements. The safety design philosophy was examined to apply proven conventional chemical plant standards to the hydrogen production plant for ensuring public safety against disasters caused by high pressure gases. This report presents a result of a consideration on safety design philosophies regarding the reasonability and condition to apply the High Pressure Gas Safety Act for the hydrogen production plant, safety classifications, seismic design classification, identification of important safety system.

JAEA Reports

Development of semantic survey map building system using semi-autonomous mobile robots for surveying of disaster area and gathering of information in nuclear power station (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Polytechnic University*

JAEA-Review 2022-011, 80 Pages, 2022/07

JAEA-Review-2022-011.pdf:5.42MB

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 semantic survey map building system using semi-autonomous mobile robots for surveying of disaster area and gathering of information in nuclear power station" conducted from FY2018 to FY2021 (this contract was extended to FY2021). Since the final year of this proposal was FY2021, the results for four fiscal years were summarized. The present study aims to research and develop semi-autonomous mobile robot systems (multi-sensor fusion system, semantic simultaneous localization and mapping (SLAM), system for traversable-route learning and safe traversable-route presentation, etc.) that simply, safely, and rapidly make semantic survey maps including multiple information

JAEA Reports

Analysis of the radioactivity concentrations in low-level radioactive waste generated from JPDR, JRR-3 and JRR-4 Facilities

Tsuchida, Daiki; Mitsukai, Akina; Aono, Ryuji; Haraga, Tomoko; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Data/Code 2022-004, 87 Pages, 2022/07

JAEA-Data-Code-2022-004.pdf:6.73MB

Radioactive wastes generated from nuclear research facilities in Japan Atomic Energy Agency are planning to be buried in the near surface disposal field. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes until by the beginning of disposal. In order to contribute to this work, we collected and analyzed samples generated from JPDR, JRR-3 and JRR-4. In this report, radioactivity concentrations of 20 radionuclides ($$^{3}$$H, $$^{14}$$C, $$^{36}$$Cl, $$^{60}$$Co, $$^{63}$$Ni, $$^{90}$$Sr, $$^{94}$$Nb, $$^{99}$$Tc, $$^{rm 108m}$$Ag, $$^{129}$$I, $$^{137}$$Cs, $$^{152}$$Eu, $$^{154}$$Eu, $$^{234}$$U, $$^{238}$$U, $$^{238}$$Pu, $$^{239+240}$$Pu, $$^{241}$$Am, $$^{244}$$Cm) were determined based on radiochemical analysis and summarized as basic data for the study of evaluation method of radioactive concentration.

Journal Articles

Numerical simulation of sodium mist behavior in turbulent Rayleigh-B$'e$nard convection using new developed mist models

Ohira, Hiroaki*; Tanaka, Masaaki; Yoshikawa, Ryuji; Ezure, Toshiki

Annals of Nuclear Energy, 172, p.109075_1 - 109075_10, 2022/07

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

In order to evaluate the mist behavior in the cover gas region of Sodium-cooled Fast Reactors (SFRs) in good accuracy, turbulent model for Rayleigh-B$'e$nard convection (RBC) was selected, and the Reynolds-averaged number density and momentum equations for mist behavior were developed and incorporated into the OpenFOAM code. In the first stage, the RBC in a simple parallel channel was calculated using Favre-averaged k-$$omega$$ SST model. The average temperature and flow characteristics agreed well with results from DNS, LES, and experiments. Then the basic heat transfer experiment simulating the cover gas region of SFRs was calculated using this turbulent model and new mist models. The calculated average temperature distribution in the height direction and the mist mass concentration agreed well with the experimental results. We developed a method that could simulate the mist behavior in turbulent RBC environments and the cover gas region of SFRs with high accuracy.

Journal Articles

Using CO$$_{2}$$ Reactions to Achieve Mass-spectrometric Discrimination in Simultaneous Plutonium-isotope Speciation with Inductively Coupled Plasma-Tandem Mass Spectrometry

Matsueda, Makoto; Kawakami, Tomohiko*; Koarai, Kazuma; Terashima, Motoki; Fujiwara, Kenso; Iijima, Kazuki; Furukawa, Makoto*; Takagai, Yoshitaka*

Chemistry Letters, 51(7), p.678 - 682, 2022/07

 Times Cited Count:0

New methodology for a simultaneous isotope speciation of various Pu isotopes without complicated isobaric interferences is developed by using inductively coupled plasma-mass spectrometry (ICP-MS). In analyzing ICP tandem MS (ICP-MS/MS), CO$$_{2}$$ gas reactions in a dynamic reaction cell (DRC) almost eliminated the background noise intensity produced by isobaric interference from isotopes originating from actinides such as Am, Cm, and U at the locations (m/z) of significant Pu isotopes ($$^{239}$$Pu, $$^{240}$$Pu, $$^{241}$$Pu, $$^{242}$$Pu, $$^{244}$$Pu).

Journal Articles

Experimental study on the localization and estimation of radioactivity in concrete rubble using image reconstruction algorithms

Takai, Shizuka; Namekawa, Masakazu*; Shimada, Taro; Takeda, Seiji

IEEE Transactions on Nuclear Science, 69(7), p.1789 - 1798, 2022/07

 Times Cited Count:0

To reduce a large amount of contaminated concrete rubble stored in the Fukushima Daiichi Nuclear Power Station site, recycling low-radioactivity rubble within the site is a possible remedy. To promote recycling while ensuring safety, not only the average radioactivity but also the radioactivity distribution of concrete rubble should be efficiently evaluated because the details of rubble contamination caused by the accident remain unclear and likely include hotspots. However, evaluating inhomogeneous contamination of thick and/or dense materials is difficult using previous measurement systems, such as clearance monitors. This study experimentally confirmed the potential applicability of image reconstruction algorithms for radioactivity distribution evaluation in concrete rubble filled in a chamber. Radiation was measured using plastic scintillation fiber around the chamber (50 $$times$$ 50 $$times$$ 40 cm$$^{3}$$). Localized hotspots were simulated using standard sources of $$^{137}$$Cs, which is one of the main nuclides of contaminated rubble. The radioactivity distribution was calculated for 100 or 50 voxels (voxel size: (10 cm)$$^{3}$$ or 10 $$times$$ 10 $$times$$ 20 cm$$^{3}$$) constituting the chamber. For 100 voxels, inner hotspots were undetected, whereas, for 50 voxels, both inner and surface hotspots were reconstructible. The distribution evaluated using the maximum likelihood expectation maximization algorithm was the most accurate; the average radioactivity was estimated within 70% accuracy in all seven cases.

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