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

https://doi.org/10.1016/j.anucene.2025.111433

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

https://doi.org/10.1016/j.anucene.2025.111256

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 (,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.

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

https://doi.org/10.1016/j.anucene.2025.111229

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

https://doi.org/10.1016/j.anucene.2024.110977

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.

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

https://doi.org/10.1016/j.anucene.2024.110962

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

https://doi.org/10.1016/j.anucene.2024.110783

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 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.

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

https://doi.org/10.1016/j.anucene.2024.110591

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

https://doi.org/10.1016/j.anucene.2023.110243

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.

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

https://doi.org/10.1016/j.anucene.2023.110144

Experiment and new analysis model simulating in-place cooling of a degraded core in severe accidents of sodium-cooled fast reactors

Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Akaev, A.*; Mikisha, A.*; Baklanov, V.*; Vurim, A.*

Annals of Nuclear Energy, 194, p.110107_1 - 110107_11, 2023/12

https://doi.org/10.1016/j.anucene.2023.110107