New filter concept for removal of fine particle generated in high level radioactive solution

Takahatake, Yoko; Watanabe, So; Watanabe, Masayuki; Sano, Yuichi; Takeuchi, Masayuki

Progress in Nuclear Science and Technology (Internet), 7, p.195 - 198, 2025/03

https://doi.org/10.15669/pnst.7.195

Extraction chromatgraphy technology for trivalent minor actinide (MA(III) ; Am(III) and Cm(III)) recovery from the solution generated by an extraction process in reprocessing of spent nuclear fuel has been developed. A fine particle is generated in the solution. The fine particle must be removed before MA recovery operation, because that leads clogging of the extraction chlomatography column. In order to prevent clogging the column, filtration system utilizing porous silica beads packed column has been designed. In this study, a fine particle trapping system was developed and particle removal performance of the system was experimentally evaluated using alumina particles as simulated fine particle. Column experiments revealed that the fine particle with the particle size from 0.12 to 15 m is cause of clogging of the filtration column. Since simulated fine particles were trapped on filtration experiments, a filtration system using the porous silica beads column is practical,

Development of high performance clarification system for spent MOX fuel reprocessing

Takeuchi, Masayuki; Takata, Takeshi; Saito, Keita*; Chikazawa, Takahiro*

Progress in Nuclear Science and Technology (Internet), 7, p.135 - 141, 2025/03

https://doi.org/10.15669/pnst.7.135

Control of insoluble sludge from fuel dissolution process is one of the important issues to secure the safety in reprocessing plant operation. In order to achieve the higher sludge recovery in the clarification stage, new clarification system which is an integrated technology of centrifugal device and filter unit has been developed for spent MOX fuel reprocessing. In this study, over-all clarification performance of this integrated system was evaluated on engineering scale. As results, total sludge recovery rate of more than 99.5 % was achieved in all test conditions by this system from the engineering test and this new clarification system showed excellent clarification performance. The sludge recovery rate of centrifugal device was influenced by test conditions and was in the range from 87 to 98 %. This study showed this technology is one of the promising clarification systems to improve the performance of sludge recovery greatly.

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, 2025/03

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.

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, 2025/03

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.

Mobile radiation measurement system by multiple small gamma-ray detectors for radioisotope detection and identification supporting responders in the field of nuclear detection and nuclear security

Kimura, Yoshiki; Yamaguchi, Tomoki

Progress in Nuclear Science and Technology (Internet), 7, p.60 - 66, 2025/03

https://doi.org/10.15669/pnst.7.60

Mo(VI), Zr(IV) and Pd(II) ion recovery mechanism onto baker's yeast from nitric acid medium

Hasegawa, Kenta; Arai, Yoichi; Watanabe, So; Watanabe, Masayuki; Matsuura, Haruaki*; Hagura, Naoto*; Minowa, Kazuki*; Konishi, Yasuhiro*

no journal, , 

CFD analysis of thermal radiation effects on large containment CIGMA vessel with Weighted Sum of Gray Gases (WSGG) Model

Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru

no journal, , 

During a postulated accident, the containment thermal-hydraulics phenomena will necessarily include radiation heat transfer since it has a significant role in the buoyancy-driven flow for large facilities. The radiation heat transfer might occur within the gas mixture and between the gas mixture and the surrounding structures due to the high absorption and emission of radiation from steam. A numerical computational fluid dynamic (CFD) simulation with a radiation model and different parameters on the component of three gases, i.e., helium, air, and steam, was performed in this study. The numerical simulation was carried out using open source CFD code OpenFOAM. A Weighted Sum of Gray Gases (WSGG) model was newly implemented in the OpenFOAM solver. At first, to achieve a significant temperature difference, the initial condition inside the vessel was dry, and steam content was set to 0.1 percent. The helium stratification was initiated with a molar fraction of 50 percent. The initial temperature and pressure were set to 20C and 1 atm. A transient simulation was started by injecting pure helium through a nozzle from the top vessel with a mass flow rate 5 g/s. Furthermore, the transient simulation was carried out using three different models. Results showed that the simulation without the radiation model showed a noticeable temperature difference compared with the radiation model. In addition, these preliminary results indicate that thermal radiation should be considered even though the steam content inside the containment vessel is low.