Benchmark analyses of criticality calculation on SCALE 6.2.3 code system

Okamoto, Naritoshi; Komeno, Akira; Seya, Atsumasa; Inaba, Hideki*; Terakado, Shinichi*; Higuchi, Masashi*

JAEA-Data/Code 2025-022, 497 Pages, 2026/03

JAEA-Data-Code-2025-022.pdf:18.06MB

The Plutonium Fuel Third Development Laboratory of the Nuclear Fuel Cycle Engineering Laboratories has applied for a change of use permit (hereinafter referred to as "license") for plutonium fuel facilities. For the criticality safety design of gloveboxes and equipment/instruments handling mixed oxide (MOX), various criticality calculation codes are used. The most recent employs the 3D Monte Carlo calculation code KENO-V.a embedded in the SCALE 4.4 code system, along with the 27-group ENDF/B-IV neutron cross-section library. SCALE 4.4 was released by the Oak Ridge National Laboratory (ORNL) in the US in 1998, and has now been in use for 27 years. ORNL has continuously improved its functionality, with SCALE 6.3.2 released in 2024. When designing and constructing new MOX fuel facilities, it is desirable to obtain a license using criticality calculation codes based on the latest knowledge. However, it is necessary to verify that these codes have sufficient reliability. Therefore, in 2018, benchmark calculations were performed using the 252-group ENDF/B-VII.1 neutron cross-section library (v7-252n) for two versions of the criticality calculation sequences KENO-V.a and KENO-VI from SCALE 6.2.3, based on past criticality experimental setups. The estimated critical-limiting multiplication factor was calculated. The results indicate that these codes can be used with sufficient confidence for criticality safety design of MOX fuel facilities.

Summaries of research and development activities by using supercomputer system of JAEA in FY2024 (April 1, 2024 - March 31, 2025)

HPC Technology Promotion Office, Center for Computational Science & e-Systems

JAEA-Review 2025-044, 140 Pages, 2026/01

JAEA-Review-2025-044.pdf:8.77MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. Over the past 10 years or so, the publication of papers utilizing computational science and technology at JAEA has accounted for about 20 percent of the total publications each fiscal year. The supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2024, the system was utilized in R&D activities that were prioritized in the Fourth Medium- to Long-Term Plan, including contributing to carbon neutrality through the development of innovative technologies for improving safety, creating innovation by promoting diverse R&D related to nuclear science and technology, promoting R&D in response to the accident at TEPCO's Fukushima Daiichi Nuclear Power Station, steadily implementing technological developments for the treatment and disposal of high-level radioactive waste, and supporting nuclear safety regulatory administration and nuclear disaster prevention by promoting safety research for these purposes. This report presents a great number of R&D results accomplished by using the system in FY2024, as well as user support, operational records and overviews of the system, and so on.

Pilot study on thermal, physico-chemical, and mechanical behavior of concrete to understand the failure behavior of Fukushima Daiichi Nuclear Power Station reactor pressure vessel pedestals (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokai National Higher Education and Research System*

JAEA-Review 2025-034, 83 Pages, 2025/12

JAEA-Review-2025-034.pdf:6.9MB

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 "Pilot study on thermal, physico-chemical, and mechanical behavior of concrete to understand the failure behavior of Fukushima Daiichi Nuclear Power Station reactor pressure vessel pedestals" conducted in FY2023. The present study aims to examine the mechanism of the collapse of only concrete with rebar remaining in the pedestal in the containment vessel (PCV) of 1F. In verifying concrete-specific factors, (1) to clarify the short-term dissolution mechanism by high temperature, we investigated data acquisition methods in melting experiments, established an analytical framework for determining dissolution, and developed a numerical analysis method for volume change by heating. Additionally, (2) to clarify long-term dissolution mechanism by temperature history, we organized the temperature and water injection history, determined concrete exposure conditions during experiments, and established a method for selecting materials and measuring expansion. Furthermore, we summarized existing knowledge of the expansion phenomenon caused by water supply after high temperature heating. In the verification of special external environmental factors, (1) to evaluate thermal conditions of PCV concrete during an accident, a preliminary heat transfer analysis of fuel debris was conducted. In addition, (2) as elemental behavior tests and comprehensive tests, a preliminary high temperature storage test on concrete materials in a water vapor atmosphere and a preliminary reaction test on the reaction behavior of metal debris and concrete were conducted. Furthermore, uranium-containing suboxides were prepared. This study provided comprehensive insight into the mechanism of concrete failure in 1F Unit 1.

Development of real-time mapping software for wide-area radiation survey

Takahashi, Tone; Koizumi, Mitsuo; Yoshimi, Yuki*; Mochimaru, Takanori*

JAEA-Technology 2025-007, 26 Pages, 2025/11

JAEA-Technology-2025-007.pdf:1.6MB

To prevent the smuggling of nuclear and radioactive materials into event venues for the purpose of terrorism, it is common practice to individually inspect people and vehicles entering and exiting using radiation detectors. However, since there remains a risk of such inspections being bypassed, it is necessary to complement them with a wide-area radiation survey to ensure that no nuclear or radioactive materials have been brought in. Radiation mapping is an effective method for efficiently surveying large areas. In this method, a gamma-ray detector equipped with GPS is used to record location data and radiation dose rates while moving. By utilizing network connectivity, measurement data from multiple detectors can be aggregated at a central command post, allowing real-time monitoring of survey progress. This system helps to prevent both redundant and missing measurements and enables the prompt detection of suspicious radiation sources. Furthermore, by incorporating spectrometers into the gamma -ray detectors, it becomes possible to identify radioactive isotopes, thereby enabling appropriate responses. To enable such wide-area radiation surveys, we developed real-time mapping software. The developed software receives measurement data transmitted from GPS-equipped gamma-ray spectrometers, processes it sequentially in real time, and plots it onto pre -downloaded map data. Additionally, by integrating the spectral data collected from regions showing abnormal radiation levels can be displayed immediately. To enhance information security, the software is designed to function within local networks without requiring internet connectivity. In this report, we introduce an overview of the developed software and provide a simplified version of the source code as an appendix. The provided code is developed using open and free operating systems, libraries, and environments, making it freely available and usable by anyone.

Optimising sodium incorporation into potassium-activated metakaolin-based alkali-activated materials

Chaerun, R. I.; Sato, Junya; Hiraki, Yoshihisa; Yoshida, Yukihiko; Sato, Tsutomu*; Osugi, Takeshi

Construction and Building Materials, 500, p.144270_1 - 144270_10, 2025/11

https://doi.org/10.1016/j.conbuildmat.2025.144270

Alkali-activated materials (AAMs), particularly those derived from metakaolin, have gained significant attention as sustainable binders for hazardous waste immobilisation, owing to their dense microstructure and chemical durability. Their amorphous aluminosilicate framework enables effective encapsulation of hazardous materials and reduces environmental risks. However, maintaining the stability of this amorphous network is challenging, particularly when sodium (Na)-rich precursors are used, as excess Na) promotes crystallisation and compromises matrix integrity. This study systematically investigates the influence of Na) concentration on the structural stability of metakaolin-based AAMs activated primarily with potassium (K)). The objective is to identify the threshold Na incorporation level that preserves the amorphous structure and maintains chemical stability. Transmission electron microscopy (TEM), Raman spectroscopy, and thermodynamic modelling were employed to examine the structural evolution of K-AAMs across a range of Na:K molar ratios. The results reveal that higher Na:K ratios induce nanopore formation and early crystallisation of Na-rich zeolitic phases, which can reduce matrix stability. In contrast, an optimal Na:K ratio was identified that maintains the amorphous network and preserves the aluminosilicate framework. These findings provide valuable insights into optimising K-AAMs for advanced, durable waste encapsulation technologies.

Actinide separation by column method using polyvinylpolypyrrolidone

Wada, Kazuma*; Kazama, Hiroyuki*; Abe, Chikage*; Onishi, Takashi; Yamamoto, Masahiko; Taguchi, Shigeo; Kuno, Takehiko; Maeda, Koji; Idemitsu, Kazuya*; Suzuki, Tatsuya*

Journal of Radioanalytical and Nuclear Chemistry, 334(11), p.8961 - 8968, 2025/11

https://doi.org/10.1007/s10967-025-10539-8

The feasibility of using commercially available polyvinylpolypyrrolidone (PVPP) as a pretreatment for mass spectrometry to separate actinides was investigated. Uranium, thorium, and plutonium were used as actinides in the study. Adsorption data for An(IV) and An(VI) were obtained by adsorption experiments in nitric acid. In addition, An(III)/An(IV)/An(VI) mutual separation was successfully performed by a column separation method using PVPP. Furthermore, the elution efficiency of U(VI) was improved by the introduction of carbonate ions. This method may be applicable as a mutual separation for actinide mass spectrometry.

Operation, maintenance and technical development of cryogenic hydrogen system at MLF (JFY20182022)

Aso, Tomokazu; Ariyoshi, Gen; Muto, Hideki*; Tanaka, Shigeto*

JAEA-Technology 2025-005, 51 Pages, 2025/10

JAEA-Technology-2025-005.pdf:2.47MB

The cryogenic hydrogen system of the J-PARC center is one of the most important pieces equipment, which is a refrigeration system for moderating (cooling) the high energy neutrons generated the spallation neutron source of the Material and Life Science Experimental Facility (MLF) that is used by many users from Japan and abroad. Since the first operation of the MLF for neutron use in 2008, the cryogenic hydrogen system has been continuously operated and maintained, and related technology development has been carried out. This report summarized these activities over the past five years.

Corrigendum to "Neutron diffraction study on the deuterium composition of nickel deuteride at high temperatures and high pressures" [Phys. B Condens. Matter. 587 (2020) 412153]

Saito, Hiroyukki*; Machida, Akihiko*; Hattori, Takanori; Sano, Asami; Funakoshi, Kenichi*; Sato, Toyoto*; Orimo, Shinichi*; Aoki, Katsutoshi*

Physica B; Condensed Matter, 714, p.417234_1 - 417234_3, 2025/10

https://doi.org/10.1016/j.physb.2025.417234

Corrigendum to "Neutron diffraction study on the deuterium composition of nickel deuteride at high temperatures and high pressures" [Phys. B Condens. Matter. 587 (2020) 412153] was reported.

External gelation conditions in fabrication of nitride fuel for transmutation of minor actinides

Iwasa, Toma; Takano, Masahide

Progress in Nuclear Science and Technology (Internet), 8, p.291 - 295, 2025/09

https://doi.org/10.15669/pnst.8.291

We have developed the external gelation technology for the fabrication of MAs nitride particles in high-performance heterogeneous fuel. Although the particle fabrication technology using external gelation methods has been developed, there has been almost no study targeting MAs nitrides fuel. Previously study suggested that the size of particle was required to be smaller than 250 m to avoid the degradation of thermophysical properties. The purpose of this study is to optimize the external gelation conditions for spherical gel particle smaller than 500 m because the particle shrank less than half size by calcination and nitridation. The external gelation tests were performed with the viscosity and pressure of dropping solution as parameters. The results show that the smaller particle with higher sphericity was obtained at the higher pressure of 350-500kPa at each viscosity of 30-50cP with positive correlation.

Achievement of safety demonstration tests using HTTR; Loss of forced cooling test at 100% reactor power (30 MW)

Nagasumi, Satoru; Hasegawa, Toshinari; Nakagawa, Shigeaki; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Saikusa, Akio; Nojiri, Naoki; Saito, Kenji; Furusawa, Takayuki; et al.

JAEA-Research 2025-005, 23 Pages, 2025/07

JAEA-Research-2025-005.pdf:2.68MB

A safety demonstration test under abnormal operating conditions using the HTTR (High Temperature Engineering Test Reactor) was conducted to demonstrate safety features of the HTGRs (High Temperature Gas-cooled Reactors). Under a simulation of a control rod shutdown failure, all primary helium gas circulators were intentionally stopped during a steady-state operation at 100% reactor thermal power (30 MW), temporal changes of the reactor power and temperatures around the reactor pressure vessel (RPV) were obtained after the complete loss of forced heat removal from the reactor core. After the event (primary coolant flow stopped), the reactor power quickly decreased due to the negative reactivity feedback associated with the core temperature rise, and then the reactor power spontaneously shifted to a stable state of low power (about 1.2%) even after a recriticality. Heat dissipation from RPV surface to a surrounding vessel cooling system (water-cooled panels) ensured the amount of heat removal required to maintain the reactor temperature constant in the low power state. In this way, the transition from the event occurrence to the stable and safety state, i.e., inherent safety features of HTGRs, were demonstrated in the case of core forced cooling loss without active shutdown operations.