Temperature effect on radiolytically generated hydrogen yield from a plutonium nitric acid aqueous solution

Toigawa, Tomohiro; Hotoku, Shinobu; Kumagai, Yuta; Abe, Yuma*; Oyama, Kanichi*; Fukaya, Hiroyuki; Ban, Yasutoshi; Kida, Takashi; Hasegawa, Satoshi*; Nakano, Masanao*; et al.

Journal of Nuclear Science and Technology, 63(3), p.322 - 327, 2026/03

https://doi.org/10.1080/00223131.2025.2541067

The effect of temperature on hydrogen production generated from radiolysis was investigated to determine the associated implications for nuclear fuel reprocessing safety. The hydrogen yield from radiolysis of plutonium nitric acid solution was measured at temperatures up to the boiling temperature of the solution. The results showed no notable temperature dependence even under boiling conditions. The impact of solution agitation on hydrogen production was also assessed, which revealed minor differences in the hydrogen yield between static and agitated conditions at room temperature. These findings suggest that high temperatures or boiling the solution do not considerably enhance hydrogen generation, and provide crucial information for accurately modeling hydrogen risks under severe accidents.

High-speed 3D modeling for nuclear reactor environment based on feature extraction results from video images (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Sapporo University*

JAEA-Review 2025-033, 71 Pages, 2025/11

JAEA-Review-2025-033.pdf:4.48MB

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 FY2023. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F), 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 FY2023, this report summarizes the research results of the "High-speed 3D modeling for nuclear reactor environment based on feature extraction results from video images" conducted in FY2023. The present study aims to develop a 3D model for a workspace that maximizes the amount of information based on the features extracted from video, which is taken when surveying the primary containment vessel and inside the reactor building as part of the decommissioning of 1F, considering within a specified time. In FY2023, we verified extracting effective shooting conditions for obtaining 3D reconstruction based on photogrammetry and the method extracting feature values that can generate 3D restoration results from a small amount of data within a specified time based on deep learning. In addition, we applied point cloud data extracted from video to segmentation and classified it into parts with instance labels.

Impact of fast reactor fuel type on backend processes in the nuclear fuel cycle

Takeshita, Kenji*; Okamura, Tomohiro*; Nakase, Masahiko*; Abe, Takumi; Nishihara, Kenji

Progress in Nuclear Science and Technology (Internet), 8, p.52 - 57, 2025/09

https://doi.org/10.15669/pnst.8.52

Verification of analytical model of MELCOR code for accident of evaporation to dryness by boiling of reprocessed high level liquid waste

Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi

JAEA-Research 2025-003, 24 Pages, 2025/06

JAEA-Research-2025-003.pdf:2.06MB

An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (RuO) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. RuO is expected to be absorbed chemically into water dissolving nitrous acid. Condensation of mixed vapor plays an important role for Ru transporting behavior in the facility building. The thermal-hydraulic behavior in the facility building is simulated with MELCOR code. The latent heat, which is a governing factor for vapor condensing behavior, has almost same value for nitric acid and water at the temperature range under 120 centigrade. Considering this thermal characteristic, it is assumed that the amount of nitric acid is substituted with mole-equivalent water in MELCOR simulation. Compensating modeling induced deviation by this assumption have been assembled with control function features of MELCOR. The comparison results have been described conducted between original simulation and modified simulation with compensating model in this report. It has been revealed that the total amount of pool water in the facility was as same as both simulations.

Trial analysis of chemical behavior in high-level radioactive liquid waste tank at accident of evaporation to dryness by boiling of reprocessing plant

Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi

JAEA-Research 2024-007, 24 Pages, 2024/08

JAEA-Research-2024-007.pdf:2.1MB

An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. It has been observed experimentally that volatility of RuO is suppressed by HNO generated by HNO radiolysis. The analysis of chemical reactions of NO including HNO and HNO in the waste tank is essential to simulate of these phenomena. To resolve this issue, an analytical approach has been attempted to couple dynamically two computer codes SHAWED and SCHERN. The simulation of boiling behavior in the tank is conducted with SHAWED. SCHERN simulates chemical behaviors of HNO, HNO and NO in the tank. A programmatic coupling algorithm and a trial simulation of the accident are presented in this report.

Estimating the corrosion rate of stainless steel R-SUS304ULC in nitric acid media under concentrating operation

Irisawa, Eriko; Kato, Chiaki

Journal of Nuclear Materials, 591, p.154914_1 - 154914_10, 2024/04

https://doi.org/10.1016/j.jnucmat.2024.154914

The amount of corrosion of austenitic stainless-steel R-SUS304ULC was evaluated considering the changes in solution composition and boiling during actual concentration operations. Austenitic stainless-steel R-SUS304ULC is the structural material of the highly radioactive liquid waste concentrator in Japanese spent fuel reprocessing plant, which treats highly corrosive nitric acid solutions during enrichment operations. The study results show that it is necessary to focus on nitric acid concentrations, oxidizing metal ion concentrations, and decompression boiling as factors that accelerate the corrosion rate of stainless steel because of cathodic reaction activation.

Criticality safety evaluation of high active liquid waste during the evaporation to dryness process at Tokai Reprocessing Plant

Miura, Takatomo; Kudo, Atsunari; Koyama, Daisuke; Obu, Tomoyuki; Samoto, Hirotaka

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10

Tokai Reprocessing Plant (TRP) had reprocessed 1,140 tons of spent fuel discharged from commercial reactors (BWR, PWR) and Advanced Thermal Reactor "Fugen" from 1977 to 2007. TRP had entered decommissioning stage in 2018. In order to reduce the risk of High Active Liquid Waste (HALW) held at the facility, the vitrification of HALW is given top priority. HALW generated from reprocessing of spent fuel contains not only fission products (FPs) but also trace amounts of uranium (U) and plutonium (Pu) within the liquid and insoluble residues (sludge). Under normal conditions, concentrations of U and Pu in HALW are very low so that it can not reach criticality. Since FPs with high neutron absorption effect coexists in HALW, even if the cooling function is lost due to serious accident and HALW evaporates to dryness, it is considered that criticality would not been reached. In order to confirm this estimation quantitatively, criticality safety evaluations were carried out for the increase of U and Pu concentrations by evaporation of HALW to the point of dryness. In this evaluation, infinite multiplication factors were calculated for each of solution system and sludge system of HALW with respect to the concentration change through evaporation to dryness. It is confirmed it could not reach criticality. The abundance ratios of U, Pu and FPs were set conservatively based on analytical data and ORIGEN calculation results. Multiplation factors for two-layer infinite slab model of solution and sludge systems of HALW were also calculated, and it was confirmed it could not reached criticality. In conclusion, the result was gaind that there could be no criticality even in the process through evaporation to dryness of HALW in TRP.

First investigations on a UNEX process

Simonnet, M.; Muller, L.*; Sittel, T.*; Weling, P.*; Mllich, U.*; Geist, A.*; Panak, P.*

Radiochimica Acta, 111(8), p.597 - 600, 2023/05

https://doi.org/10.1515/ract-2023-0151

Consideration on roles and relationship between observations/measurements and model predictions for environmental consequence assessments for nuclear facilities

Togawa, Orihiko; Okura, Takehisa; Kimura, Masanori

JAEA-Review 2022-049, 76 Pages, 2023/01

JAEA-Review-2022-049.pdf:3.74MB

Before construction and after operation of nuclear facilities, environmental consequence assessments are conducted for normal operation and an emergency. These assessments mainly aim at confirming safety for the public around the facilities and producing relief for them. Environmental consequence assessments are carried out using observations/ measurements by environmental monitoring and/or model predictions by calculation models, sometimes using either of which and at other times using both them, according to the situations and necessities. First, this report investigates methods, roles, merits/demerits and relationship between observations/measurements and model predictions which are used for environmental consequence assessments of nuclear facilities, especially holding up a spent nuclear fuel reprocessing plant at Rokkasho, Aomori as an example. Next, it explains representative examples of utilization of data on observations/measurements and results on model predictions, and considers points of attention at using them. Finally, the report describes future direction, for example, improvements of observations/measurements and model predictions, and fusion of both them.

Development of simulation program; SHAWED for analysis of accident of evaporation to dryness by boiling of reprocessed high level liquid waste in tank

Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*

JAEA-Research 2022-011, 37 Pages, 2022/12

JAEA-Research-2022-011.pdf:2.88MB

An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents at a fuel reprocessing plant. Two major mechanisms are expected for fission products (FPs) transfer from liquid to vapor phase. One is non-volatiles FPs transfer in the form of mists to the vapor phase in the tank, the other is volatilization of such as Ruthenium. These FPs transferred to the vapor phase in the tank could be released with water and nitric-acid mixed steam and NO gas flow to the environment. NO is generated from denitration of nitrate fission products during dry out phase. These phenomena occurred in this accident originate from the liquid waste boiling in the tank. It is essential for the risk assessment of this accident to simulate thermo-hydraulic and chemical behaviors in the waste tank quantitatively with a versatile computer program. The SHAWED (imulation of igh-level radioctive aste vaporation and ryness) has been developed to realize these requirements. In this report, detailed description of major analytical models is explained based on the features of this accident, and some simulation examples are also described for the use in an actual risk assessment.