Ten years after the NPP accident at Fukushima; Review on fuel debris behavior in contact with water

Grambow, B.; Nitta, Ayako; Shibata, Atsuhiro; Koma, Yoshikazu; Utsunomiya, Satoshi*; Takami, Ryu*; Fueda, Kazuki*; Onuki, Toshihiko*; Jegou, C.*; Laffolley, H.*; et al.

Journal of Nuclear Science and Technology, 59(1), p.1 - 24, 2022/01

https://doi.org/10.1080/00223131.2021.1966347

A Project focusing on the contamination mechanism of concrete after the accident at Fukushima Daiichi Nuclear Power Plant

Yamada, Kazuo*; Maruyama, Ippei*; Haga, Kazuko*; Igarashi, Go*; Aihara, Haruka; Tomita, Sayuri*; Kiran, R.*; Osawa, Norihisa*; Shibata, Atsuhiro; Shibuya, Kazutoshi*; et al.

Proceedings of International Waste Management Symposia 2021 (WM2021) (CD-ROM), 10 Pages, 2021/03

Radioactive contamination of several materials following the Fukushima Daiichi Nuclear Power Station accident

Koma, Yoshikazu; Shibata, Atsuhiro; Ashida, Takashi

Nuclear Materials and Energy (Internet), 10, p.35 - 41, 2017/01

https://doi.org/10.1016/j.nme.2016.08.015

Due to the accident of Fukushima Daiichi Nuclear Power Station (NPS) of Tokyo Power Electric Co. in 2011, radioactive nuclides were released to environment and heavily contaminated the NPS site itself. Analytical concentration of radionuclides for contaminated materials including the accumulated water, rubble, soil and vegetation which already opened were utilized to discuss contamination behavior of some radionuclides. The concentration ratio of a radionuclide to that of Cs was normalized with composition of the damaged reactor fuels referred to as "transport ratio" to investigate elemental behavior. The transport ratio of Sr to rubble and soil compared with which was assumed to be via air was approximately 10 to 10, and the successive transport to the accumulated water was comparable to Cs except for the initial release. Transport of Pu, Am and Cm were scarce relative to Cs regardless of transport process by means of air and water, whereas those of iodine, selenium and tellurium were comparable or greater. Contamination with and C was independent of Cs, Sr and TRU nuclides and would be a result of different transport process.

Study on the behavior of radiolytically produced hydrogen in a high-level liquid waste tank of a reprocessing plant; Comparison between actual and simulated solutions

Kinuhata, Hiroshi*; Yamamoto, Masahiko; Taguchi, Shigeo; Surugaya, Naoki; Sato, Soichi; Kodama, Takashi*; Tamauchi, Yoshikazu*; Shibata, Yuki*; Anzai, Kiyoshi*; Matsuoka, Shingo*

Nuclear Technology, 192(2), p.155 - 159, 2015/11

https://doi.org/10.13182/NT15-15

Experiments using a small-scale apparatus with 30 ml actual high-level liquid waste from the Tokai Reprocessing Plant were carried out to show that the hydrogen concentration in the gas phase reaches a steady-state value of much less than 4% (lower explosive limit) in the absence of sweeping-air. The H concentration reached a steady-state value as was expected and it was compared with a value predicted from an equation with parameters which had been obtained using the simulated solution. Satisfactory agreement showed that the Pd-ion catalytic H consumption reaction previously found in the simulated solution proceeded equally well in the actual solution.

Transport behavior of radionuclides into the accumulated water at the Fukushima Daiichi NPS

Koma, Yoshikazu; Shibata, Atsuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

https://doi.org/10.1299/jsmeicone.2015.23._ICONE23-2_53

The radioactive contaminated water has been accumulated at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company (TEPCO). Radioactivity concentrations determined for the water was normalized with the fuel composition and divided by that for Cs in order to investigate behavior of some elements to contaminate the water. Transport of Sr was initially suppressed and gradually increased to the extent comparable to Cs until almost one year. Transport of He, I and Se were greater than that of Cs. Transport of Pu, Am, Cm and Eu were considerably small. Concerning activation product the extent of Co transport was significant than that of Ni.

Shielding study at the Fukui Prefectural Hospital Proton Therapy Center

Satoh, Daiki; Maeda, Yoshikazu*; Tameshige, Yuji*; Nakashima, Hiroshi; Shibata, Tokushi*; Endo, Akira; Tsuda, Shuichi; Sasaki, Makoto*; Maekawa, Motokazu*; Shimizu, Yasuhiro*; et al.

Journal of Nuclear Science and Technology, 49(11), p.1097 - 1109, 2012/11

https://doi.org/10.1080/00223131.2012.730900

At the Fukui Prefectural Hospital Proton Therapy Center, neutron doses behind concrete shields and at maze have been measured by using radiation monitors, DARWIN, Wendi-2, a rem meter, and solid state nuclear track detectors. The measured data were compared with the estimations by analytical models and Monte Carlo code PHITS. The analytical model with the parameters employed in shielding design of the facility gave considerably larger estimates than the measured data. This means that the facility was designed with an enough safety margin. The calculation results of PHITS were less than those of the analytical model, and were about 3 times larger than the measured data. From the view point of a safety policy with conservative estimation for shielding design, Monte Carlo simulation is a better tool for estimating radiation safety at accelerator-based proton treatment facilities.

Behavior of fission products in simplified solvent extraction system for uranium, plutonium and neptunium co-recovery

Nakahara, Masaumi; Shibata, Atsuhiro; Koma, Yoshikazu

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 4 Pages, 2011/12

To elucidate the behavior of fission products, two counter current experiments were carried out with a dissolver solution derived from irradiated fast neutron reactor "JOYO" core fuel, one was a single cycle flow sheet using double scrubbing solution with 9 and 1 mol/dm HNO and the other used a Tc scrubbing solution with 10 mol/dm HNO. Among fission products, the decontamination behavior of Zr and Tc differed according to HNO concentration of the scrubbing solutions. The decontamination factor of Zr and Tc increase to 76.8 and 7.52 with 10 mol/dm HNO in the Tc scrubbing solution. Other fission products such as Cs was well decontaminated and its DF resulted in 10.

FaCT Phase-I evaluation on the advanced aqueous reprocessing process, 4; Solvent extraction simplified for FBR fuel reprocessing

Koma, Yoshikazu; Ogino, Hideki; Sakamoto, Atsushi; Nakabayashi, Hiroki; Shibata, Atsuhiro; Nakahara, Masaumi; Washiya, Tadahiro

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12

Separation of actinide elements by solvent extraction using centrifugal contactors in the NEXT process

Nakahara, Masaumi; Sano, Yuichi; Koma, Yoshikazu; Kamiya, Masayoshi; Shibata, Atsuhiro; Koizumi, Tsutomu; Koyama, Tomozo

Journal of Nuclear Science and Technology, 44(3), p.373 - 381, 2007/03

https://doi.org/10.3327/jnst.44.373

Actinides recovery was attempted by the simplified solvent extraction process using TBP as an extractant for U, Pu and Np co-recovery and the SETFICS process for Am and Cm recovery with a view to decreasing the environmental impact. Uranium, Pu and Np co-recovery was conducted under the condition with high nitric acid concentration in the feed solution or scrubbing solution. High nitric acid concentration in the feed solution availed to the Np oxidation not only in the feed solution, but also at the extraction section. This oxidation reaction permitted the Np extraction with U and Pu. In the SETFICS process, a TRUEX solvent of 0.2M CMPO/1.4M TBP was employed to increase the loading of metals. In place of sodium nitrate, HAN was applied to this experimental flow sheet for "salt-free" concept. This experiment was succeeded in Am and Cm product. On high-loading flow sheet, the flow rate of aqueous effluents and spent solvent was expected to decrease in 47% and 54%, respectively.

Actinides recovery by solvent extraction in NEXT process

Nakahara, Masaumi; Sano, Yuichi; Koma, Yoshikazu; Kamiya, Masayoshi; Shibata, Atsuhiro; Koizumi, Tsutomu; Koyama, Tomozo

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10

Concerning the advanced aqueous reprocessing system named NEXT process, the behavior of actinide elements was investigated in main two extraction processes of NEXT process, i.e. the simplified PUREX process for U, Pu and Np recovery, and SETFICS process for Am and Cm recovery.