Scalability of inertial particle deposition in bubbles with internal circulation

Motegi, Kosuke; Shibamoto, Yasuteru; Kukita, Yutaka

Annals of Nuclear Energy, 184, p.109679_1 - 109679_10, 2023/05

https://doi.org/10.1016/j.anucene.2022.109679

Nonuniform particle distribution and interference between removal mechanisms during unsteady aerosol deposition from a rising spherical bubble

Motegi, Kosuke; Shibamoto, Yasuteru; Kukita, Yutaka

Journal of Nuclear Science and Technology, 59(8), p.1037 - 1046, 2022/08

https://doi.org/10.1080/00223131.2022.2029724

Spin excitations of the =1/2 one-dimensional Ising-like antiferromagnet BaCoVO in transverse magnetic fields

Okutani, Akira*; Onishi, Hiroaki; Kimura, Shojiro*; Takeuchi, Tetsuya*; Kida, Takanori*; Mori, Michiyasu; Miyake, Atsushi*; Tokunaga, Masashi*; Kindo, Koichi*; Hagiwara, Masayuki*

Journal of the Physical Society of Japan, 90(4), p.044704_1 - 044704_9, 2021/04

https://doi.org/10.7566/JPSJ.90.044704

Effects of magnetic field and pressure on the valence-fluctuating antiferromagnetic compound EuPtSi

Takeuchi, Tetsuya*; Yara, Tomoyuki*; Ashitomi, Yosuke*; Iha, Wataru*; Kakihana, Masashi*; Nakashima, Miho*; Amako, Yasushi*; Honda, Fuminori*; Homma, Yoshiya*; Aoki, Dai*; et al.

Journal of the Physical Society of Japan, 87(7), p.074709_1 - 074709_14, 2018/07

https://doi.org/10.7566/JPSJ.87.074709

Stabilization of MOX dissolving solution at STACY

Kobayashi, Fuyumi; Sumiya, Masato; Kida, Takashi; Kokusen, Junya; Uchida, Shoji; Kaminaga, Jota; Oki, Keiichi; Fukaya, Hiroyuki; Sono, Hiroki

JAEA-Technology 2016-025, 42 Pages, 2016/11

JAEA-Technology-2016-025.pdf:17.88MB

A preliminary test on MOX fuel dissolution for the STACY critical experiments had been conducted in 2000 through 2003 at Nuclear Science Research Institute of JAEA. Accordingly, the uranyl / plutonium nitrate solution should be reconverted into oxide powder to store the fuel for a long period. For this storage, the moisture content in the oxide powder should be controlled from the viewpoint of criticality safety. The stabilization of uranium / plutonium solution was carried out under a precipitation process using ammonia or oxalic acid solution, and a calcination process using a sintering furnace. As a result of the stabilization operation, recovery rate was 95.6% for uranium and 95.0% for plutonium. Further, the recovered oxide powder was calcined again in nitrogen atmosphere and sealed immediately with a plastic bag to keep its moisture content low and to prevent from reabsorbing atmospheric moisture.

Basic study on decontamination of TRU wastes with cerium mediated electrolytic oxidation method

Ishii, Junichi; Kobayashi, Fuyumi; Uchida, Shoji; Sumiya, Masato; Kida, Takashi; Shirahashi, Koichi; Umeda, Miki; Sakuraba, Koichi

JAEA-Technology 2009-068, 20 Pages, 2010/03

JAEA-Technology-2009-068.pdf:2.49MB

At Nuclear Fuel Cycle Safety Engineering Research Facility, the cerium mediated electrolytic oxidation method which is a decontamination technique to decrease the radioactivity of TRU wastes to the clearance-level has been developed for the effective reduction of TRU wastes generated from the decommissioning of a nuclear fuel reprocessing facility and so on. This method corrodes the oxide layer and the surface of metallic TRU metal wastes by the strong oxidation power of Ce in nitric acid. In this study, parameter tests were conducted to optimize the solution condition of Ce initial concentrations and nitric acid concentrations. The target corrosion rate of metallic TRU wastes set to be 24m/h for the practical use of this method. Under the optimized solution condition, a dissolution test of stainless steel simulating wastes was carried out. From the result of the dissolution test, the average corrosion rate was 3.3 m/h during the test time of 90 hours. Based on the supposition that the corrosion depth of metallic TRU wastes was 20 m enough to achieve the clearance-level, the treatment time for the decontamination was about 6 hours. It was confirmed from the result that the decontamination could be performed within one day and the decontamination solution could repeatedly reuse 15 times.

Report on the fuel treatment facility operation

Kokusen, Junya; Seki, Masakazu; Abe, Masayuki; Nakazaki, Masato; Kida, Takashi; Umeda, Miki; Kihara, Takehiro; Sugikawa, Susumu

JAERI-Tech 2005-004, 53 Pages, 2005/03

JAERI-Tech-2005-004.pdf:5.92MB

This report presents operating records of dissolution of uranium dioxide and concentration of uranyl nitrate solution and acid removal, which have been performed from 1994 through 2003, for the purpose of feeding 10% and 6% enriched uranyl nitrate solution fuel to Static Experimental Critical Facility(STACY) and Transient Experimental Critical Facility(TRACY) in Nuclear Fuel Safety Engineering Facility(NUCEF).

Water flow experiments and analyses on the cross-flow type mercury target with perforated plates

Haga, Katsuhiro; Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro; Tagawa, Hisato*; Kukita, Yutaka*

Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04

The cross-flow type mercury target, in which mercury flows crossing the proton beam path, has been developed as the spallation target of the material and life science facility in the high intensity proton accelerator project. As a part of design optimization, we proposed a mercury target using perforated plates aiming to simplify the inner structure, to make the target assembling easier, and to decrease the assembling cost. Then, the effectiveness of the target structure was investigated by no-heat water experiments and computational analyses. A mockup model of the cross-flow type mercury target using perforated plates was fabricated with plexi-glass and the water flow field was measured using PIV technique and the results were compared with the analytical results. The cross-flow field was realized by perforated plates and the analytical results corresponded well with the experimental results in the proton beam path where the cooling of heat generation is important.

Development of simulation code for MOX dissolution using silver-mediated electrochemical method (Contract research)

Kida, Takashi; Umeda, Miki; Sugikawa, Susumu

JAERI-Data/Code 2003-001, 29 Pages, 2003/03

JAERI-Data-Code-2003-001.pdf:6.34MB

MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in NUCEF. A simulation code for the MOX dissolution has been developed for the operating support. In this report an outline of the simulation code is proposed and a comparison with the experimental data and a parameter study on the MOX dissolution rate are described.The principle of this code is based on Zundelevich's model for PuO dissolution using Ag. The influence of nitrous acid on the material balance of Ag and the surface area of MOX powder on the basis of particle size distribution are taken into consideration in this model. A comparison with experimental data was carried out to confirm a validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using the appropriate MOX dissolution rate constant. The parameters affecting the dissolution rate were studied, it was found that MOX particle size was major governing factor on the dissolution rate.

Data report for ROSA-IV LSTF 10% hot leg break experiment Run SB-HL-04

Kukita, Yutaka; Nakamura, Hideo; *; *; *; Anoda, Yoshinari; Kumamaru, Hiroshige; Suzuki, Mitsuhiro; ; Yonomoto, Taisuke; et al.

JAERI-M 91-040, 122 Pages, 1991/03

JAERI-M-91-040.pdf:3.42MB

no abstracts in English