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Journal Articles

Two-phase flow structure in a particle bed packed in a confined channel

Ito, Daisuke*; Kurisaki, Tatsuya*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.6430 - 6439, 2019/08

In core disruptive accident of sodium-cooled fast reactor, cooling of residual fuel debris formed in the reactor core is one of important factors to achieve in-vessel retention of the fuel. To clarify the feasibility of the cooling which is called "in-place cooling", characteristics of gas-liquid two-phase flow in the debris bed must be well understood. Since the debris bed can be formed in a confined flow channel in the core, effect of the channel wall cannot be neglected. Thus, this study aims to clarify the effect of the wall on two-phase flow characteristics in the debris bed, which was simulated as a particle bed packed in a pipe. The pressure drop was measured and compared with results by previous models, and porosity and void fraction distributions were measured by X-ray radiography. Then, the pressure drop evaluation model was modified considering the wall effect, and the applicability of the models was discussed.

Journal Articles

Development of evaluation method for in-place cooling of residual core materials in core disruptive accidents of SFRs

Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Ganovichev, D. A.*; Baklanov, V. V.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 11 Pages, 2019/05

The cooling of the residual core materials after the fuel discharge from the core in the accident of SFRs can significantly affect the distribution fraction of the core materials, which is an important factor for the in-vessel retention (IVR). For the evaluation of the cooling of the residual core materials which is called "in-place cooling", behavior in a SFR core was analyzed preliminary by SIMMER-III. Based on the analysis result, method of phenomena identification and ranking table (PIRT) was applied. Fundamental experiment focusing on three thermal-hydraulic phenomena those were extracted by PIRT was considered in order to investigate them and utilize it for validation of the SIMMER-III. To achieve continuous oscillation of sodium level which can occur in the phase of in-place cooling of SFRs, analytical survey was conducted by SIMMER-III. As a result of that, the effects of experimental conditions on the oscillation amplitude and the duration time were clarified quantitatively, which are necessary to determine the specific experimental conditions.

Journal Articles

Effect of porosity distribution on two-phase pressure drop in a packed bed

Kurisaki, Tatsuya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

In the evaluation of the in-place cooling which is for the residual core materials in the severe accident of sodium-cooled fast reactors, pressure loss of two-phase flow in debris bed is one of the important factors. Although Lipinski model is already proposed for the pressure loss evaluation, the accuracy would decrease when the porosity is not homogeneous. Thus, experiment to measure the pressure loss in a packed bed of non-homogeneous porosity distribution was conducted, and the Lipinski model was modified dividing the cross section to evaluate the pressure loss in it. As a result, it was confirmed that agreement of the experimental values with the values by modified Lipinski model was better than that with the original Lipinski model.

Journal Articles

Fundamental experiments of jet impingement and fragmentation simulating the fuel relocation in the core disruptive accident of sodium-cooled fast reactors

Imaizumi, Yuya; Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Suzuki, Toru; Emura, Yuki

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04

In order to simulate the typical accident conditions of the fuel relocation phase in SFRs, the molten alloy of low melting point was discharged into a shallow water pool. The distance between the nozzle exit and the bottom plate was set to a value which was indicated to be insufficient to fragment. As the experimental result, the melt jet reached the bottom plate, and dispersed in all directions along the plate, together with the progress of fragmentation. In addition, the melt temperature on the bottom plate decreased rapidly along the radius direction. These results suggest that the fragmentation which would accompany this rapid cooling would be enhanced by the plate. This enhancement would be caused by the extension of the melt-water interface when the melt was dispersed forcibly by the plate. The solidified debris remained after the discharge showed remarkable fragmentation which was assumed to be caused by the formations of small vapor bubbles in the interface.

Journal Articles

Survey on effect of crystal texture of beryllium on total cross-section to improve neutronic evaluation in JMTR

Takemoto, Noriyuki; Imaizumi, Tomomi; Kimura, Nobuaki; Tsuchiya, Kunihiko; Hori, Junichi*; Sano, Tadafumi*; Nakajima, Ken*

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 11 Pages, 2014/09

Neutronic evaluations in JMTR have been performed for irradiation tests by Monte Carlo method with thermal neutron scattering law, S($$alpha$$, $$beta$$), data for beryllium metal, etc. The calculation accuracy of fast and thermal neutron fluxes are $$pm$$10% and $$pm$$30%, respectively. Analytical and experimental investigations to achieve higher calculation accuracy, especially for the thermal neutron flux up to the fast neutron flux level, have been therefore performed to offer higher value data technically to the JMTR users. In order to investigate an effect of fabrication method of beryllium material on the calculation accuracy, total cross-sections of beryllium specimens were measured using KURRI-LINAC, and it was found that the total cross-section was different from the theoretical one, and depended on the crystal texture, etc. The S($$alpha$$, $$beta$$) was adjusted based on the measured data, and the applicability to the neutronic evaluation in the JMTR was verified.

JAEA Reports

Results of the environmental radiation monitoring following the accident at the Fukushima Daiichi Nuclear Power Plant; Interim report (Ambient radiation dose rate, radioactivity concentration in the air and radioactivity concentration in the fallout)

Furuta, Sadaaki; Sumiya, Shuichi; Watanabe, Hitoshi; Nakano, Masanao; Imaizumi, Kenji; Takeyasu, Masanori; Nakada, Akira; Fujita, Hiroki; Mizutani, Tomoko; Morisawa, Masato; et al.

JAEA-Review 2011-035, 89 Pages, 2011/08

JAEA-Review-2011-035.pdf:2.97MB

As a correspondence to the accident at the Fukushima Daiichi Nuclear Power Plant, the environmental radiation monitoring was performed at the Nuclear Fuel Cycle Engineering Laboratories, JAEA. This report presented the measurement results of ambient radiation dose rate, radioactivity concentration in the air and radioactivity concentration in fallout and meteorological observation result until May 31, 2011. The ambient radiation dose rate increased, with the peak dose rate of several thousand nGy/h at 7 o'clock in March 15, at 5 o'clock in March 16, and at 4 o'clock in March 21. The variation on the radioactivity concentration in the air and in fallout showed the almost same tendency as that of the dose rate. The concentration ratio of I-131/Cs-137 in the air increased to about 100. The dose was estimated resulting from internal exposure due to inhalation.

Journal Articles

Preliminary report on a new fault exposure of the Senya fault associated with the Rikuu earthquake in 1896, Northeast Japan

Imaizumi, Toshifumi*; Kagohara, Kyoko*; Otsuki, Kenshiro*; Miwa, Atsushi*; Kosaka, Hideki*; Nohara, Tsuyoshi

Katsudanso Kenkyu, (26), p.71 - 77, 2006/06

no abstracts in English

Journal Articles

Role of the impurities in production rates of radiation-induced defects in silicon materials and solar cells

Khan, A.*; Yamaguchi, Masafumi*; Oshita, Yoshio*; Dharmarasu, N.*; Araki, Kenji*; Abe, Takao*; Ito, Hisayoshi; Oshima, Takeshi; Imaizumi, Mitsuru*; Matsuda, Sumio*

Journal of Applied Physics, 90(3), p.1170 - 1178, 2001/08

 Times Cited Count:50 Percentile:85.51(Physics, Applied)

1MeV-electron and 10MeV-proton irradiations into Si doped various impurities such as B, Ga, O and C were performed and residual defects in the Si were studied using DLTS and C-V measurements.It was revealed that Ci-Oi whose level is Ev-0.36 eV and Bi-Oi whose energy is Ec-0.18eV were generated. In Ga-doped Si, the generation of Ci-Oi was suppressed. Since Ci-Oi acts as scattering center, this result indicates that the radiation resistance of solar cells is improved by using Ga-doped Si substrates.Furthermore, a new defect level (Ev+18eV) was observed in Ga-dpoed Si by irradiation. This defect level was annealed out above 350 C.

Oral presentation

Status of neutronic evaluation in JMTR

Takemoto, Noriyuki; Kimura, Nobuaki; Imaizumi, Tomomi; Hori, Junichi*; Sano, Tadafumi*; Nakajima, Ken*

no journal, , 

no abstracts in English

Oral presentation

Studies on relocation behavior of molten core materials in the core disruptive accident of sodium-cooled fast reactors, 2; Experimental study on molten-metal discharge into a shallow water pool

Imaizumi, Yuya; Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Suzuki, Toru; Emura, Yuki

no journal, , 

For the study of the fuel-coolant interaction and deposition behavior of the molten core material on the in-vessel structure in the relocation phase of the core disruptive accident in SFR, several experiments discharging low-melting-point alloy into a shallow water pool were conducted. The experimental results and the mechanism discussion are reported.

Oral presentation

Studies on relocation behavior of molten core materials in the core disruptive accident of sodium-cooled fast reactors, 1; Overall plan

Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Imaizumi, Yuya; Suzuki, Toru; Emura, Yuki

no journal, , 

no abstracts in English

Oral presentation

Studies on relocation behavior of molten core materials in the core disruptive accident of sodium-cooled fast reactors, 4; Effects of test conditions on dispersion and sedimentation of molten-material in a shallow water pool

Emura, Yuki; Matsuba, Kenichi; Isozaki, Mikio; Imaizumi, Yuya; Kamiyama, Kenji

no journal, , 

A series of simulation tests discharging a molten alloy with low-melting point into a shallow water pool is being conducted in order to investigate fuel-coolant interaction and sedimentation behavior of molten core materials on the structure of core inlet coolant plenum in the core disruptive accident of sodium-cooled fast reactors. The present study shows effects of the initial temperature of the molten alloy, the confined coolant volume, etc. on the test results.

Oral presentation

Study on coolant behavior in damaged core of sodium-cooled fast reactor, 4; Evaluation of pressure drop through simulated debris bed in consideration of porosity distribution

Kurisaki, Tatsuya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

no journal, , 

The pressure loss of two-phase flow in debris bed is one of the important factor for coolability evaluation of the inplace-cooling for the residual core-materials in the severe accident of sodium-cooled fast reactors. Although Lipinski model is already proposed for the pressure loss evaluation, the accuracy would decrease when the porosity is not homogeneous. Thus, the evaluation method by dividing the cross section was developed for the two-phase flow in packed bed of non-homogenious porosity, and compared the calculated values with experimental data for the validation.

Oral presentation

Research and development on preceding processing methods for contaminated water management waste at Fukushima Daiichi Nuclear Power Station, 5; Gamma-ray irradiation test on solidified body of cement and Alkali Activated Material

Taniguchi, Takumi; Imaizumi, Ken*; Namiki, Masahiro*; Osugi, Takeshi; Kuroki, Ryoichiro; Kikuchi, Michio*; Yamamoto, Takeshi*; Kaneda, Yoshihisa*; Haga, Kazuko*

no journal, , 

It is important to understand fundamental solidification characteristics of contaminated water management waste at Fukushima Daiichi Nuclear Power Station. The Solidified bodies are fabricated with cementitious material and Alkali Activated Material, and are irradiated with Gamma-ray.

Oral presentation

Research and development on preceding processing methods for contaminated water management waste at Fukushima Daiichi Nuclear Power Station, 1; Examination of Task on processing

Kato, Jun; Taniguchi, Takumi; Namiki, Masahiro*; Imaizumi, Ken*; Terasawa, Toshiharu*; Osugi, Takeshi; Sone, Tomoyuki; Nakazawa, Osamu; Kuroki, Ryoichiro; Koma, Yoshikazu

no journal, , 

Regarding solidification technologies with applied performance for radioactive waste processing, tasks for applicability evaluation of these technologies to contaminated water treatment secondary waste at Fukushima Daiichi Nuclear Power Station was extracted in FY2017. This report introduces the examination for the extracted tasks implemented in FY2018.

Oral presentation

Research and development on preceding processing methods for contaminated water management waste at Fukushima Daiichi Nuclear Power Station, 6; Numerical simulation analysis of relationships between nuclide inventory and solidified body temperature

Hiraki, Yoshihisa; Terasawa, Toshiharu*; Imaizumi, Ken*; Taniguchi, Takumi; Kato, Jun; Osugi, Takeshi; Sone, Tomoyuki; Nakazawa, Osamu; Kuroki, Ryoichiro

no journal, , 

The relationship between the amount of radionuclide to be solidified and the solidified body temperature was analyzed, when solidify contaminated water management waste at Fukushima Daiichi Nuclear Power Station with cement etc. at low temperature. The analysis code used the radiation transport code and the thermal analysis code. Thus, the limit value by radionuclide concentration during solidification process are evaluated. The summary of the test and some of the obtained results are introduced.

Oral presentation

Study on coolant behavior in damaged core of sodium-cooled fast reactor, 5; Effect of pipe and particle sizes on pressure drop in simulated debris bed

Ishiguro, Akinari*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

no journal, , 

no abstracts in English

Oral presentation

Research and development on preceding processing methods for contaminated water management waste at Fukushima Daiichi Nuclear Power Station, 2; Investigation of approach for selecting solidification techniques applied to contaminated water management waste

Furukawa, Shizue*; Koyama, Tadafumi*; Kikuchi, Michio*; Otsuka, Taku*; Yamamoto, Takeshi*; Imaizumi, Ken*; Osugi, Takeshi; Nakazawa, Osamu; Kuroki, Ryoichiro

no journal, , 

In order to evaluate the application to the solidification of the water treatment secondary waste generated from Fukushima Daiichi Nuclear Power Station, research concerning the approach of applicability evaluation was conducted for solidification technology of practical scale.

Oral presentation

Research and development on preceding processing methods for contaminated water management waste at Fukushima Daiichi Nuclear Power Station, 7; Numerical analysis of temperature rise of solidified waste by radiation applying various container shapes

Hiraki, Yoshihisa; Enomoto, Mayu*; Terasawa, Toshiharu*; Imaizumi, Ken*; Kato, Jun; Osugi, Takeshi; Sone, Tomoyuki; Kuroki, Ryoichiro

no journal, , 

Evaluate the maximum temperature of solidified body, when Secondary Waste Produced from Contaminated Water Treatment at Fukushima Daiichi NPP is solidified with cement etc. Simulate various container shapes, and analysis using the radiation transport code and the thermal analysis code was performed. The summary of the test and some of the obtained results are introduced.

Oral presentation

Research and development on preceding processing methods for contaminated water management waste at Fukushima Daiichi Nuclear Power Station, 13; Gamma-ray irradiation characteristics of solidified materials containing carbonate slurry

Kato, Jun; Imaizumi, Ken*; Osugi, Takeshi; Sone, Tomoyuki; Kuroki, Ryoichiro; Kikuchi, Michio*; Yamamoto, Takeshi*; Otsuka, Taku*; Kaneda, Yoshihisa*; Osawa, Norihisa*

no journal, , 

Solidified cement and Alkali Activated Materials (AAM) containing simulated carbonated slurry was prepared, and the amount of Hydrogen gas generation on gamma-ray irradiation was determined. The obtained data was compared with the data of solidified material without any content of waste, and the effect for Hydrogen gas generation with the material containing simulated carbonated slurry was evaluated.

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