Fukushima Daiichi Nuclear Power Plant Unit 2 Accident analysis considering the thermal stratification and containment leakage

Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya

Journal of Nuclear Science and Technology, 62(12), p.1226 - 1230, 2025/12

https://doi.org/10.1080/00223131.2025.2570371

Reactivity estimation based on the linear equation of characteristic time profile of power in subcritical quasi-steady state

Yamane, Yuichi

Journal of Nuclear Science and Technology, 59(11), p.1331 - 1344, 2022/11

https://doi.org/10.1080/00223131.2022.2053221

The reactivity was estimated from a time profile of neutron count rate or a simulated data in a quasi-steady state after sudden change of reactivity or external neutron source strength. The estimation was based on the equation of power in subcritical quasi-steady state. The purpose of the study is to develop the method of timely reactivity estimation from complicated time profile of neutron count rate. The developed method was applied to the data simulating neutron count rate created by using one-point kinetics code, AGNES, and Poisson-distributed random noise and to the transient subcritical experiment data measured by using TRACY. The result shows that the difference of the estimated and reference value was within about 5% or less for ( -1) for simulated data and within about 7% or less for -1.4 and -3.1 for the experimental data. It was also shown that the possibility of the reactivity estimation several ten seconds after the status change.

Development of the unified cross-section set ADJ2017R

Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2021-019, 115 Pages, 2022/03

JAEA-Data-Code-2021-019.pdf:6.21MB
JAEA-Data-Code-2021-019-appendix(CD-ROM).zip:435.94MB

In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61

Thermal-hydraulics to risk assessment; Roles of thermal-hydraulics simulation to risk assessment

Maruyama, Yu; Yoshida, Kazuo

Nihon Genshiryoku Gakkai-Shi ATOMO, 63(7), p.517 - 522, 2021/07

https://doi.org/10.3327/jaesjb.63.7_517

no abstracts in English

Inner structure and inclusions in radiocesium-bearing microparticles emitted in the Fukushima Daiichi Nuclear Power Plant accident

Okumura, Taiga*; Yamaguchi, Noriko*; Dohi, Terumi; Iijima, Kazuki; Kogure, Toshihiro*

Microscopy, 68(3), p.234 - 242, 2019/06

https://doi.org/10.1093/jmicro/dfz004

Radiocesium-bearing microparticles (CsMPs), consisting substantially of silicate glass, were released to the environment during the Fukushima nuclear accident in March 2011. We investigated a total of nine CsMPs using transmission electron microscopy (TEM) and inferred the atmosphere in the reactors during the accident. From elemental mapping using energy-dispersive X-ray spectrometry, Fe and Zn showing radial inhomogeneities were found in the CsMPs, in addition to the Cs that had been previously reported. Four of the CsMPs included submicron crystals, which were identified as chromite, franklinite, acanthite, molybdenite, and hessite. The chromium-containing spinels, chromite and franklinite, indicated the presence of ferrous iron (Fe), suggesting that the inside of the reactors was reductive to some extent. Electron energy-loss spectroscopy also confirmed that the CsMPs did not contain boron, and therefore the atmosphere in which they were formed might be boron-free.

Analysis of natural circulation tests in the experimental fast reactor JOYO

Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08

Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.

Investigation of multi-dimensional effect in sodium leak and fire behavior

Ohno, Shuji

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 9 Pages, 2014/12

This paper presents the typical characteristics of sodium combustion and subsequent reaction heat transfer behaviors observed and investigated in sodium columnar leak and fire experiment which was conducted in an enclosed steel vessel with large inner volume of about 100 m. Especially the experiment was carried out with the main focus on the burning phenomenon within a limited spatial area in the case of large sodium leak rate as well as on the multi-dimensional thermal-hydraulics both near a sodium burning zone and in a whole region in the vessel. The investigated experimental results show us that the sodium combustion of columnar leak and its splashed droplets would lead to important oxygen deficiency behavior near the burning region, and be followed by the limitation or saturation of maximum sodium burning rate.

Annual report of Naka Fusion Research Establishment from April 1, 2004 to March 31, 2005

Naka Fusion Research Establishment

JAERI-Review 2005-046, 113 Pages, 2005/09

JAERI-Review-2005-046.pdf:24.35MB

This annual report provides an overview of research and development activities at Naka Fusion Research Establishment, including those performed in collaboration with other research establishments of JAERI, research institutes, and universities, during the period from 1 April, 2004 to 31 March, 2005. The activities in the Naka Fusion Research Establishment are highlighted by researches in JT-60 and JFT-2M, theoretical and analytical plasma researches, research and development of fusion reactor technologies towards ITER and fusion power demonstration plants, and activities in support of ITER design and construction.

Measurement and evaluation of isotope effect between tritium and deuterium on diffusion and surface recombination in/on nickel using ion driven permeation method (Cooperative research)

Nakamura, Hirofumi; Nishi, Masataka; Sugisaki, Masayasu*

JAERI-Research 2003-018, 32 Pages, 2003/09

JAERI-Research-2003-018.pdf:1.32MB

Tritium transport behavior in materials, which is essential for the safety evaluation of the fusion reactor, has to be evaluated by either tritium properties or extrapolated value from protium or deuterium (D) to tritium (T) using the isotope effect theory. However, there are still some uncertainties on estimation of T behavior in materials, because there are only a few T transport properties data in materials, and it is not completely proven the application of the isotope effect theory to T due to the lack of T data. Therefore, in order to understand the tritium transport properties in materials, isotope effects on diffusion and surface recombination between T and D in/on nickel, whose hydrogen transport properties were well known, were investigated by comparing the obtained properties of T with those of D measured under the same conditions with the ion driven permeation method. Though obtained diffusion coefficient of T was larger than that of D, and activation energy of diffusion of T was smaller than that of D as the contrary to the classical diffusion theory, those were shown to be explained with a modified diffusion theory by introducing higher vibration temperatures in nickel than previous reported values. In addition, the isotope effect on surface recombination coefficient between D and T was shown to be explained using a modified solution model as well as diffusion.

Exchange of tritium implanted into oxide ceramics for protium by exposure to air vapors at room temperature

Morita, Kenji*; Suzuki, Hironori*; Soda, Kazuo*; Iwahara, Hiroiku*; Nakamura, Hirofumi; Hayashi, Takumi; Nishi, Masataka

Journal of Nuclear Materials, 307-311(2), p.1461 - 1465, 2002/12

https://doi.org/10.1016/S0022-3115(02)01286-2

no abstracts in English