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Journal Articles

Rational physical protection design of transuranium fuel cycle site with accelerator-driven system by using material attractiveness

Oizumi, Akito; Sagara, Hiroshi*

Annals of Nuclear Energy, 223, p.111677_1 - 111677_12, 2025/12

This study aims to provide a new rational physical protection (PP) design method by using ${it Material Attractiveness}$ (${it Attractiveness}$) and to design a rational PP system for a site of the transuranium fuel cycle with accelerator-drive systems (ADSs cycle) using the new method. First, the new rational PP design method with different PP design requirements for each ${it Attractiveness}$ was generalized based on the definitions of a national standard method defined by the US Department of Energy, the joint US-Japan study, and the International Atomic Energy Agency. A new PP categorization of Uranium (U), including U-234, which is abundant in the ADS cycle, was also developed based on ${it Attractiveness}$. Second, a PP design was conducted for a general BWR site with MOX fuel and the ADS cycle site by using the new rational method. It was clarified that the highest overall ${it Attractiveness}$ of the items within the ADS cycle site was lower than that of the MOX fuel assembly within the BWR site. The BWR site was determined to be Category I requiring the inner area. The PP design requirement level of the ADS cycle site was determined to be Category II, which does not require an inner area, while the ADS cycle site would have been classified as Category I if the PP design had been conducted using the conventional method.

Journal Articles

Experimental study on light gas transport during containment venting by using the large-scale test facility CIGMA

Soma, Shu; Ishigaki, Masahiro*; Shibamoto, Yasuteru

Annals of Nuclear Energy, 219, p.111455_1 - 111455_12, 2025/09

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Nuclear hydrogen demonstration project using the HTTR; Demarcation of nuclear-industrial laws and design standards

Aoki, Takeshi; Shimizu, Atsushi; Ishii, Katsunori; Morita, Keisuke; Mizuta, Naoki; Kurahayashi, Kaoru; Yasuda, Takanori; Noguchi, Hiroki; Nomoto, Yasunobu; Iigaki, Kazuhiko; et al.

Annals of Nuclear Energy, 220, p.111503_1 - 111503_7, 2025/09

 Times Cited Count:0

Aiming to establish coupling technologies between a high temperature gas cooled reactor and a hydrogen production plant, JAEA has initiated the HTTR Heat Application Test Project and is conducting the safety design and the safety analysis for the licensing of the HTTR Heat Application Test Facility. The present study proposed a relative evaluation methodology for the demarcation of applicable laws and design standards for the nuclear hydrogen production system and applied it to the HTTR Heat Application Test Facility. The evaluation results showed that a candidate applying the High Pressure Gas Safety Act to the Heat Application Test Facility (hydrogen production plant) and design standards established under the High Pressure Gas Safety Act to the steam reformer did not show the lowest category in any of the metrics, and was proposed as the most superior demarcation option for the HTTR Heat Application Test Facility.

Journal Articles

Impact of molybdenum on iodine chemistry during fission product transport phenomenology

Rizaal, M.; Nakajima, Kunihisa; Suzuki, Eriko; Miwa, Shuhei

Annals of Nuclear Energy, 218, p.111433_1 - 111433_10, 2025/08

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Integrated thermal power measurement in the modified STACY for the performance inspections

Araki, Shohei; Aizawa, Eiju; Murakami, Takahiko; Arakaki, Yu; Tada, Yuta; Kamikawa, Yutaka; Hasegawa, Kenta; Yoshikawa, Tomoki; Sumiya, Masato; Seki, Masakazu; et al.

Annals of Nuclear Energy, 217, p.111323_1 - 111323_8, 2025/07

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

JAEA has modified the STACY from a homogeneous system using solution fuel to a heterogeneous system using fuel rods in order to obtain criticality characteristics of fuel debris. The modification of the STACY was completed in December 2023. A series of performance inspections were conducted for the start of experimental operations. A new thermal power calibration is required for the performance inspections in order to operate at less than 200 W, which is the permitted thermal power. However, the thermal power measurement method and calibration data used in the former STACY is no longer available due to the modification of the modified STACY. We measured the thermal power of the STACY using the activation method that was improved to adapt to the measurement condition and calibrated the power meter system. Since the positions where activation foils could be installed were very limited, the thermal power was evaluated using numerical calculations supplemented by experimental data. Neutron flux data at the positions of the activation foil was measured by the activation method. Neutron distribution in the core was calculated by the Monte Carlo code MVP. A response function of the activation foil was calculated using the PHITS. The uncertainty of the thermal power measurement was conservatively estimated to be about 15%. Four operations were conducted for the thermal power measurement. The power meter was calibrated by using three operational data and tested with the one operational data. It was found that the indicated value of the meter adjusted by the STACY before the modification work would tend to overestimate the actual output by about 40%. In addition, the current calibration was able to calibrate the meter to within 3% accuracy.

Journal Articles

Prediction of composite neutron source spectra by combination of JENDL-5 and PHITS

Ogawa, Tatsuhiko

Annals of Nuclear Energy, 216, p.111256_1 - 111256_12, 2025/06

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

A novel robust method has been developed to simulate the performance of composite neutron sources composed of an alpha-emitting actinide and a light nucleus with low neutron separation energy. This method is based on the JENDL-5 cross-section data library and the Monte-Carlo radiation transport code PHITS. In contrast to previously devised methods, this approach can predict various quantities of the sources, such as actinide grain size dependence, absolute neutron emission intensity, energy spectra of neutrons and parasitic photons, neutron multiplicity, and time structure, with little approximation. The accurate calculation of stopping power of alpha rays in actinide grains and light elements, as well as the use of ($$alpha$$,n) reaction evaluated cross sections, which is one of the unique features of PHITS Ver.3.34 and its later versions, are the essences of the method. This method allows for the calculation of quantities important for practical applications, such as detection signal frequency, coincidence event rate, and the impact of parasitic gamma-rays.

Journal Articles

Analysis of dissolved radionuclides trapped into corrosion products formed on carbon steel and the corresponding increase in radioactivity

Aoyama, Takahito; Ueno, Fumiyoshi; Sato, Tomonori; Kato, Chiaki; Sano, Naruto; Yamashita, Naoki; Otani, Kyohei; Igarashi, Takahiro

Annals of Nuclear Energy, 214, p.111229_1 - 111229_6, 2025/05

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Numerical analysis of a potential Reactor Pressure Vessel (RPV) boundary failure mechanism in Fukushima Daiichi Nuclear Power Station Unit-2

Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya

Annals of Nuclear Energy, 214, p.111217_1 - 111217_13, 2025/05

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Evaluation of uranium-233 neutron capture cross section in keV region

Otsuka, Naohiko*; Tada, Kenichi; Cabellos, O.*; Iwamoto, Osamu

Annals of Nuclear Energy, 212, p.110977_1 - 110977_9, 2025/03

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

The uranium-233 neutron capture cross section between 3 keV and 1 MeV was evaluated considering the recent new alpha-value measurement performed at the Los Alamos National Laboratory LANCE facility. The obtained capture cross section is systematically lower than the capture cross section in the JENDL-5 library and the reduction is close to 50% around 20 keV. The newly evaluated cross section was validated against 166 criticality experiments chosen from the ICSBEP handbook by performing Monte Carlo neutron transport calculation with the JENDL-5 library, and slight reduction of the chi-square value was achieved by adoption of the newly evaluated capture cross section.

Journal Articles

Built-in physics models and proton-induced nuclear data validation using MCNP, PHITS, and FLUKA; Impact on the shielding design for proton accelerator facilities

$c{C}$elik, Y.*; Stankovskiy, A.*; Iwamoto, Hiroki; Iwamoto, Yosuke; Van den Eynde, G.*

Annals of Nuclear Energy, 212, p.111048_1 - 111048_12, 2025/03

 Times Cited Count:1 Percentile:57.00(Nuclear Science & Technology)

Journal Articles

Uncertainty quantification for severe-accident reactor modelling; Results and conclusions of the MUSA reactor applications work package

Brumm, S.*; Gabrielli, F.*; Sanchez Espinoza, V.*; Stakhanova, A.*; Groudev, P.*; Petrova, P.*; Vryashkova, P.*; Ou, P.*; Zhang, W.*; Malkhasyan, A.*; et al.

Annals of Nuclear Energy, 211, p.110962_1 - 110962_16, 2025/02

 Times Cited Count:6 Percentile:93.24(Nuclear Science & Technology)

Journal Articles

Neutron transmission measurements for silica glass at the KURNS-LINAC

Lee, J.; Rossi, F.; Kodama, Yu; Hironaka, Kota; Koizumi, Mitsuo; Sano, Tadafumi*; Matsuo, Yasunori*; Hori, Junichi*

Annals of Nuclear Energy, 211, p.111017_1 - 111017_7, 2025/02

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Study on the specifications of the basic core configurations of the modified STACY

Gunji, Satoshi; Araki, Shohei; Izawa, Kazuhiko; Suyama, Kenya

Annals of Nuclear Energy, 209, p.110783_1 - 110783_7, 2024/12

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Since the compositions and properties of the fuel debris are uncertain, critical experiments are required to validate calculation codes and nuclear data used for the safety evaluation. For this purpose, the Japan Atomic Energy Agency (JAEA) has been modifying a critical assembly called "STACY." The first criticality of the modified STACY is scheduled for spring 2024. This paper reports the consideration results of the specifications of the basic core configurations of the modified STACY at the first criticality. We prepared two types of gird plates with different neutron moderation conditions (their intervals are 1.50 cm and 1.27 cm). However, there is a limitation on the number of available UO$$_{2}$$ fuel rods. The core configurations for the first criticality satisfying these experimental constraints were designed by computational analysis. A cylindrical core configuration with a 1.50 cm grid plate close to the optimum moderation condition needs 253 fuel rods to reach criticality. As to the 1.27 cm grid plate, we considered core configurations with 2.54 cm intervals by using doubled pitches of the grid plate. It will need 213 fuel rods for the criticality. In addition, we considered the experimental core configuration with steel/concrete simulant rods to simulate fuel debris conditions. This paper shows these core configurations and their evaluated specifications.

Journal Articles

Possible criticality scenario and its mechanism of the Windscale Works criticality accident in 1970 analyzed by computational fluid dynamics and Monte Carlo neutron transport

Fukuda, Kodai

Annals of Nuclear Energy, 208(1), p.110748_1 - 110748_10, 2024/12

 Times Cited Count:1 Percentile:57.00(Nuclear Science & Technology)

Journal Articles

New analysis model of solid body formation in particle method for jet impingement and solidification in severe accidents of SFRs

Imaizumi, Yuya; Kamiyama, Kenji; Matsuba, Kenichi

Annals of Nuclear Energy, 206, p.110658_1 - 110658_10, 2024/10

 Times Cited Count:1 Percentile:57.00(Nuclear Science & Technology)

Journal Articles

Quantifying uncertainty induced by scattering angle distribution using maximum entropy method

Maruyama, Shuhei; Yamamoto, Akio*; Endo, Tomohiro*

Annals of Nuclear Energy, 205, p.110591_1 - 110591_13, 2024/09

 Times Cited Count:1 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

The Behavior of a jet passing through a grid-type obstacle; An Experimental investigation

Abe, Satoshi; Shibamoto, Yasuteru

Annals of Nuclear Energy, 202, p.110461_1 - 110461_16, 2024/07

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Integration of level 3 probabilistic risk assessment for nuclear power plants with transportation simulation considering earthquake hazards

Shimada, Kazumasa; Sakurahara, Tatsuya*; Farshadmanesh, P.*; Reihani, S.*; Mohagehgh, Z.*

Annals of Nuclear Energy, 197, p.110243_1 - 110243_12, 2024/03

 Times Cited Count:1 Percentile:25.62(Nuclear Science & Technology)

This research improves the realism of Level 3 probabilistic risk assessment (PRA) for nuclear power plants (NPP) to avoid subjective expert judgment when setting evacuation behavior for residents. Therefore, the evacuation speed output by the traffic simulation code MATSim was input to the level 3 PRA code MACCS. Furthermore, to set the priority of the places where road closure is to be considered, a method to evaluate the road closure risk due to the earthquake using the natural disaster risk assessment code HAZUS was developed. Then, the relationship between the evacuation routes and the radiation dose was evaluated for the case study of the Sequoyah NPP adopted in the SOARCA study conducted by the US NRC. As a result, the present study found an evacuation route with low closure risk but causing high radiation dose of residents when it is closed. This showed effectiveness of the proposed Level 3 PRA methodology for supporting decision-makers to enhance evacuation routes.

Journal Articles

Development of a formulation to predict molten core spreading in an LWR severe accident

Sahboun, N. F.; Matsumoto, Toshinori; Iwasawa, Yuzuru; Wang, Z.; Sugiyama, Tomoyuki

Annals of Nuclear Energy, 195, p.110145_1 - 110145_12, 2024/01

 Times Cited Count:2 Percentile:46.61(Nuclear Science & Technology)

Journal Articles

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka

Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

171 (Records 1-20 displayed on this page)