Scenario analysis of future nuclear energy use in Japan, 1; Methodology of nuclear fuel cycle simulator: NMB4.0

Abe, Takumi; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*

Progress in Nuclear Science and Technology (Internet), 7, p.299 - 304, 2025/05

https://doi.org/10.15669/pnst.7.299

Currently, much research continues on stable energy sources that do not emit CO in order to achieve a carbon-neutral and sustainable society. Nuclear energy is one of the such sources, and various new reactors and reprocessing technologies are being developed. In order to implement the nuclear fuel cycle with these technologies, a nuclear fuel cycle simulator is required to quantitatively evaluate various quantities, such as the distribution of nuclear fuel materials and the scale of waste loading. For this purpose, NMB4.0 was developed in collaboration with Tokyo Institute of Technology and Japan Atomic Energy Agency. This code calculates the material balance of 179 nuclides including actinides and fission products (FPs) from the front-end to the back-end and simulates the nuclear fuel cycle in an integrated manner. Unlike other nuclear fuel cycle simulators, the code is capable of performing precise back-end analyses such as the number of radioactive wastes and the scale of the geological repository considering heat generation of waste package under diverse nuclear energy scenario, and is an open source code that runs on Microsoft Excel. By these features, it is possible to quantitatively study nuclear energy utilization strategies with various stakeholders. The presentation will detail the numerical model used in NMB4.0.

Development of corrosion-stable dual-Si-layered membranes for hydrogen production via thermochemical iodine-sulfur process

Myagmarjav, O.; Tanaka, Nobuyuki; Noguchi, Hiroki; Kamiji, Yu; Ono, Masato; Nomura, Mikihiro*; Takegami, Hiroaki

Progress in Nuclear Science and Technology (Internet), 7, p.235 - 242, 2025/05

https://doi.org/10.15669/pnst.7.235

Achievements and status of the STRAD project for radioactive liquid waste management

Arai, Yoichi; Watanabe, So; Nakahara, Masaumi; Funakoshi, Tomomasa; Hoshino, Takanori; Takahatake, Yoko; Sakamoto, Atsushi; Aihara, Haruka; Hasegawa, Kenta; Yoshida, Toshiki; et al.

Progress in Nuclear Science and Technology (Internet), 7, p.168 - 174, 2025/05

https://doi.org/10.15669/pnst.7.168

The Japan Atomic Energy Agency (JAEA) has been conducting a project named "Systematic Treatment of RAdioactive liquid waste for Decommissioning (STRAD)" project since 2018 for fundamental and practical studies for treating radioactive liquid wastes with complicated compositions. Fundamental studies have been conducted using genuine liquid wastes accumulated in a hot laboratory of the JAEA called the Chemical Processing Facility (CPF), and treatment procedures for all liquid wastes in CPF were successfully designed on the results obtained. As the next phase of the project, new fundamental and practical studies on primarily organic liquid wastes accumulated in different facilities of JAEA are in progress. This paper reviews the representative achievements of the STRAD project and introduces an overview of ongoing studies.

Safeguards approach and design of transuranium fuel cycle with accelerator-driven system based on material attractiveness

Oizumi, Akito; Sagara, Hiroshi*

Progress in Nuclear Science and Technology (Internet), 7, p.331 - 337, 2025/03

https://doi.org/10.15669/pnst.7.331

Research and development of transuranium (TRU) fuel cycle with accelerator-drive systems (ADSs) transmuting minor actinides separated from the commercial cycles has been continuously conducted to reduce the high-level radioactive waste contained in spent fuel discharged from nuclear power plants. Since the chemical form and composition of the ADS fuels are different from those of the current commercial cycles, it is necessary to examine the inspection goal of the safeguards which are required for the ADS cycle. In this study, the of Pu was evaluated assuming the diversion of any items in the ADS cycle in terms of nuclear non-proliferation. They were compared with the of the MOX fuel assemblies (fresh and spent fuels) for a conventional boiling water reactor (BWR). The all items in the ADS cycle, regardless of whether they are fresh or spent fuel, were found to have the same of Pu as the spent fuel assembly of BWR-MOX. Additionally, in order to design and evaluate accounting systems for safeguards, the assumed uncertainty () values of measurements by operators and verification by regulators for the Pu flow rate in the ADS cycle were derived with reference to the accuracy targets for Pu measurement technology. The derived values were compared with the 1 significant quantity (1SQ), which was generally used as a target value, and the spent fuel standard (equivalent to 5%) set based on the evaluation, respectively. As a result, it was clarified that the both values of Pu measurements by the operator and the regulator exceeded 1SQ but the spent fuel standard was generally achievable.

Study on restricted use of contaminated rubble on Fukushima Daiichi NPS site, 1; Estimation of reference radiocesium concentration for recycling materials

Shimada, Taro; Miwa, Kazuji; Takeda, Seiji

Progress in Nuclear Science and Technology (Internet), 6, p.203 - 207, 2019/01

https://doi.org/10.15669/pnst.6.203

Rubbles less than 5 Sv/h of surface dose rate, which are stored outdoor in the Fukushima Daiichi NPS (1F) site, will be recycled and applied in a restricted reuse only within 1F site in the future. In this study, we suggested a concept for establishing the reference radioactive concentration of recycling material for the restricted use in the 1F site. Reference radiocesium concentration is calculated so that increased dose rate by restricted reuse does not exceed 1 Sv/h which is the minimum value of dose rate map in the 1F entire site. In order to justify the restricted reuse under the reference concentration calculated, additional occupational dose, dose rate at the site boundary and groundwater concentration at the outlet to the ocean are evaluated and confirmed that the values are below 2 mSv/y, 1 mSv/y and 1 Bq/cm of Cs and Cs, respectively. And then calculated the reference radiocesium concentrations of the recycling material used for paved roads and the bases of concrete building.

Development of ROV system to explore fuel debris in the Fukushima Daiichi Nuclear Power Plant

Kamada, So*; Kato, Michio*; Nishimura, Kazuya*; Nancekievill, M.*; Watson, S.*; Lennox, B.*; Jones, A.*; Joyce, M. J.*; Okumura, Keisuke; Katakura, Junichi*

Progress in Nuclear Science and Technology (Internet), 6, p.199 - 202, 2019/01

https://doi.org/10.15669/pnst.6.199

As a technology development to investigate the distribution of submerged fuel debris in the primary containment vessel (PCV) of the Fukushima Daiichi Nuclear Power Station, we are conducting development experiments of sonar system to be mounted in a compact ROV. The experiments were conducted in two types of water tanks with different depths, simulating the PCV, using sonar with different sizes, ultrasonic frequencies, and beam scanning method, and simulated fuel debris. As a result, we characterized the shape discrimination performance of the simulated debris, and the noise due to multi-path in narrow closed space.

Proposal of framework for decision making of additional monitoring of eye lens dose for radiation workers

Kowatari, Munehiko; Yoshitomi, Hiroshi; Oishi, Tetsuya; Yoshizawa, Michio

Progress in Nuclear Science and Technology (Internet), 6, p.86 - 90, 2019/01

https://doi.org/10.15669/pnst.6.86

Comparison of the Fukushima radioactive mapping by two different aerial radiation monitoring systems

Kim, B.-J.*; Sasaki, Miyuki; Sanada, Yukihisa

Progress in Nuclear Science and Technology (Internet), 6, p.130 - 133, 2019/01

https://doi.org/10.15669/pnst.6.130

ACE library of JENDL-4.0/HE

Matsuda, Norihiro; Kunieda, Satoshi; Okamoto, Tsutomu*; Tada, Kenichi; Konno, Chikara

Progress in Nuclear Science and Technology (Internet), 6, p.225 - 229, 2019/01

https://doi.org/10.15669/pnst.6.225

Measurement of activation cross sections of aluminum for protons with energies between 0.4 GeV and 3.0 GeV at J-PARC

Matsuda, Hiroki; Meigo, Shinichiro; Iwamoto, Hiroki

Progress in Nuclear Science and Technology (Internet), 6, p.171 - 174, 2019/01

https://doi.org/10.15669/pnst.6.171

Activation cross sections of various materials are strongly required for the improvement of the accuracy of nuclear design and the reduction of the construction costs for spallation neutron sources and transmutation systems. Activation cross sections have been measured in several facilities. However, they have low accuracy and precision. Especially, there are merely experimental data with 3 GeV protons which are used for spallation neutron source (MLF) in J-PARC, the experimental data is required for the improvement of the target materials. Thus, we measured cross sections of tungsten, gold, indium, and beryllium with 0.4 GeV to 3.0 GeV protons. Moreover, ones of aluminium that are set with materials were also measured for a variation of this experiment. It was found that more accurate data than current ones would be measured by using precise beam controls and highly accurate beam monitoring. We compared the experimental data, the evaluated data (JENDL-HE/2007), and the calculations with several intra-nuclear cascade models by PHITS code. Although the experimental data agreed with JENDL-HE/2007, the calculations underestimated about 40%. This could come from the evaporation model (GEM) being included in PHITS code. We found that the calculations agreed with the experimental data by upgrading PHITS code. The cross sections for the other materials have been analysed so far.