Lattice parameters of fluorite-structured uranium-americium mixed oxides

Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Yokoyama, Keisuke; Murakami, Tatsutoshi

Journal of Nuclear Materials, 584, p.154576_1 - 154576_11, 2023/10

https://doi.org/10.1016/j.jnucmat.2023.154576

Differential pressure rise event for filters of HTTR primary helium gas circulators, 1; Investigation of differential pressure rise event

Nemoto, Takahiro; Arakawa, Ryoki; Kawakami, Satoru; Nagasumi, Satoru; Yokoyama, Keisuke; Watanabe, Masashi; Onishi, Takashi; Kawamoto, Taiki; Furusawa, Takayuki; Inoi, Hiroyuki; et al.

JAEA-Technology 2023-005, 33 Pages, 2023/05

JAEA-Technology-2023-005.pdf:5.25MB

During shut down of the HTTR (High Temperature engineering Test Reactor) RS-14 cycle, an increasing trend of filter differential pressure for the helium gas circulator was observed. In order to investigate this phenomenon, the blower of the primary helium purification system was disassembled and inspected. As a result, it is clear that the silicon oil mist entered into the primary coolant due to the deterioration of the charcoal filter performance. The replacement and further investigation of the filter are planning to prevent the reoccurrence of the same phenomenon in the future.

Materials science and fuel technologies of uranium and plutonium mixed oxide

Kato, Masato; Machida, Masahiko; Hirooka, Shun; Nakamichi, Shinya; Ikusawa, Yoshihisa; Nakamura, Hiroki; Kobayashi, Keita; Ozawa, Takayuki; Maeda, Koji; Sasaki, Shinji; et al.

Materials Science and Fuel Technologies of Uranium and Plutonium mixed Oxide, 171 Pages, 2022/10

https://doi.org/10.1201/9781003298205

Innovative and advanced nuclear reactors using plutonium fuel has been developed in each country. In order to develop a new nuclear fuel, irradiation tests are indispensable, and it is necessary to demonstrate the performance and safety of nuclear fuels. If we can develop a technology that accurately simulates irradiation behavior as a technology that complements the irradiation test, the cost, time, and labor involved in nuclear fuel research and development will be greatly reduced. And safety and reliability can be significantly improved through simulation of nuclear fuel irradiation behavior. In order to evaluate the performance of nuclear fuel, it is necessary to know the physical and chemical properties of the fuel at high temperatures. And it is indispensable to develop a behavior model that describes various phenomena that occur during irradiation. In previous research and development, empirical methods with fitting parameters have been used in many parts of model development. However, empirical techniques can give very different results in areas where there is no data. Therefore, the purpose of this study is to construct a scientific descriptive model that can extrapolate the basic characteristics of fuel to the composition and temperature, and to develop an irradiation behavior analysis code to which the model is applied.

Measurements of thermal conductivity for near stoichiometric (UPuAm)O (z = 0.05, 0.10, and 0.15)

Yokoyama, Keisuke; Watanabe, Masashi; Tokoro, Daishiro*; Sugimoto, Masatoshi*; Morimoto, Kyoichi; Kato, Masato; Hino, Tetsushi*

Nuclear Materials and Energy (Internet), 31, p.101156_1 - 101156_7, 2022/06

https://doi.org/10.1016/j.nme.2022.101156

In current nuclear fuel cycle systems, to reduce the amount of high-level radioactive waste, minor actinides (MAs) bearing MOX fuel is one option for burning MAs using fast reactor. However, the effects of Am content in fuel on thermal conductivity are unclear because there are no experimental data on thermal conductivity of high Am bearing MOX fuel. In this study, The thermal conductivities of near stoichiometric (UPuAm)O solid solutions(z = 0.05, 0.10, and 0.15) have been measured between room temperature (RT) and 1473 K. The thermal conductivities decreased with increasing Am content and satisfied the classical phonon transport model ((A+BT)) up to about 1473 K. A values increased linearly with increasing Am content because the change in ionic radius affects the conduction of the phonon due to the solid solution in U and Am. B values were independent of Am content.

Spark plasma sintering of SiC/graphite functionally graded materials

Watanabe, Masashi; Yokoyama, Keisuke; Imai, Yoshiyuki; Ueta, Shohei; Yan, X.

Ceramics International, 48(6), p.8706 - 8708, 2022/03

https://doi.org/10.1016/j.ceramint.2021.12.004

Previous studies have used various methods for sintering of SiC, carbon, and SiC/carbon functionally graded materials (FGM). However, no experimental studies on SiC/graphite FGM manufacturing using the spark plasma sintering (SPS) method have been reported. In this study, a SiC/graphite FGM specimen has been fabricated using SPS. The interface between the adjacent layers of the sintered specimen exhibits no apparent defects such as gaps or delaminations. The SiC and graphite phases in the specimen show no substantial change before and after sintering.

Irradiation behavior of fast reactor fuel pins, 1; Evaluation of cesium behavior by coupling computer codes

Uwaba, Tomoyuki; Yokoyama, Keisuke; Ikusawa, Yoshihisa; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*

no journal, , 

A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup, showing consistency with post irradiation examinations.

Computer simulation for short-term irradiation behavior of mixed oxide fuels containing Am in a fast reactor

Yokoyama, Keisuke; Ikusawa, Yoshihisa; Uwaba, Tomoyuki; Morimoto, Kyoichi; Tanaka, Kosuke; Hirooka, Shun

no journal, , 

We have developed analytical models for describing the effects of MAs on fuel physical properties such as vapor pressure, thermal conductance and density. These models were incorporated into a computer simulation code CEDAR, which predicts mechanical and thermal behaviors of a fuel pin irradiated in FRs. CEDAR was applied to the short-term irradiation experiments of the fuel pins in the experimental fast reactor Joyo. Mechanisms of these behaviors were investigated through the comparison between the PIEs and simulations and analyzing them.

Thermal conductivity measurement of high Am bearing mixed oxide fuel

Yokoyama, Keisuke; Watanabe, Masashi; Kato, Masato; Tokoro, Daishiro*

no journal, , 

In current nuclear fuel cycle systems, to reduce the amount of high-level radioactive waste, minor actinides (MAs) bearing MOX fuel is one option for burning MAs using fast reactor. However, the effects of Am content in fuel on thermal conductivity are unclear because there are no experimental data on thermal conductivity of high Am bearing MOX fuel. In this work, the thermal conductivity of high Am-bearing MOX fuel samples was measured. In this study, MOX fuel samples containing 10 at% Am were prepared. The oxygen to metal ratio (O/M ratio) of sintered pellet was adjusted to 2.00. Thermal diffusivity was measured from R.T. to 1473 K by the laser flash method. Thermal conductivity was calculated from thermal diffusivity, heat capacity and the density of fuel samples. The measured thermal conductivity values decreased with the increase of Am content. Those for 10 at% Am bearing MOX fuel agreed well with the classical phonon transport model, and the effects of bearing 10 at% Am on MOX fuel samples were in good agreement with those predicted from previous experimental study results.

Oxygen potential and thermal conductivity of highly Am-doped UO

Watanabe, Masashi; Yokoyama, Keisuke; Kato, Masato; Hino, Tetsushi*

no journal, , 

no abstracts in English

Demonstration research on fast reactor recycling using low decontaminated MA-bearing MOX fuels, 10; Fabrication of MA-bearing MOX fuel pellets

Onishi, Takashi; Morishita, Kazuki; Watanabe, Masashi; Yokoyama, Keisuke; Kato, Masato

no journal, , 

no abstracts in English