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Journal Articles

Thermal conductivity measurement of uranium-plutonium mixed oxide doped with Nd/Sm as simulated fission products

Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01

The thermal conductivities of near-stoichiometric (U,Pu,Am)O$$_{2}$$ doped with Nd$$_{2}$$O$$_{3}$$/Sm$$_{2}$$O$$_{3}$$, which is major fission product (FP) generated by a uranium-plutonium mixed oxides (MOX) fuel irradiation, as simulated fission products are evaluated at 1073-1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneity of Nd/Sm are prepared using two kinds of powder made by ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the classical phonon transport model (A+BT)$$^{-1}$$. The dependences of the coefficients A and B on the Nd/Sm content (C$$_{Nd}$$ and C$$_{Sm}$$, respectively) are evaluated as: A(mK/W)=1.70 $$times$$ 10$$^{-2}$$ + 0.93C$$_{Nd}$$ + 1.20C$$_{Sm}$$, B(m/W)=2.39 $$times$$ 10$$^{-4}$$.

Journal Articles

Computational and experimental examination of simulated core damage and relocation dynamics of a BWR fuel assembly

Hanus, G.*; Sato, Ikken; Iwama, Tatsuya*

Proceedings of International Waste Management Symposia 2016 (WM2016) (Internet), 12 Pages, 2016/03

JAEA plans a large-scale test to evaluate damage and relocation behavior of BWR core materials consisting of fuel rods, channel boxes, control blade and lower support structures. Its purpose is to contribute to understanding of core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing, JAEA desires preliminary investigations to examine melting test pieces. The purpose of such tests is to verify the materials and test piece will be heated by plasma to the target temperature (ca.2900K) and to collect data about the material relocation behavior. Results from preliminary computational simulations are presented illustrating the effectiveness of a 150 kW non-transferred plasma jet. An experimental test program using the computational analyses as a basis and a plasma torch is described.

JAEA Reports

Development of pellet melting temperature measuring technique; Melting temperature measuring technique for small sample

Harada, Katsuya; Nakata, Masahito; Harada, Akio; Nihei, Yasuo; Yasuda, Ryo; Nishino, Yasuharu

JAERI-Tech 2004-034, 13 Pages, 2004/03

JAERI-Tech-2004-034.pdf:0.69MB

The Department of Hot Laboratories has been aiming the establishment of the melting temperature measuring technique for small samples obtained from the micro-region of irradiated fuel pellet. Due to the modification of the shape of tungsten capsule contained sample and the improvement of the detection method for melting temperature from indistinct thermal arrest point owing to small sample, it is possible to determine the melting temperature of small sample and to utilize effectively for the irradiated fuel pellet by using the existing apparatus. This paper describes the technique of the melting temperature measurement for small sample and the experimental results by using tantalum, molybdenum, hafnium oxide and un-irradiated UO$$_{2}$$ pellet.

JAEA Reports

Study on high-performance fuel cladding materials; Joint research report in FY 1999-2000 (Phase 1) (Joint research)

Kiuchi, Kiyoshi; Ioka, Ikuo; Tachibana, Katsumi; Suzuki, Tomio; Fukaya, Kiyoshi*; Inohara, Yasuto*; Kambara, Shozo; Kuroda, Yuji*; Miyamoto, Satoshi*; Ogura, Kazutomo*

JAERI-Research 2002-008, 63 Pages, 2002/03

JAERI-Research-2002-008.pdf:7.85MB

no abstracts in English

JAEA Reports

A Study on density, melting point, thermal expansion, creep, thermal diffusivity and thermal conductivity of the simulated rock-like oxide (ROX) fuels

Yanagisawa, Kazuaki; Omichi, Toshihiko*; Shirasu, Noriko; Muromura, Tadasumi; *

JAERI-Tech 99-032, 65 Pages, 1999/03

JAERI-Tech-99-032.pdf:3.23MB

no abstracts in English

JAEA Reports

Fuel melting and mechanical energy generation during power burst experiments with aluminum-cladding uranium silicide fuel plate

Fuketa, Toyoshi; Ishijima, Kiyomi; ; Soyama, Kazuhiko; Ichikawa, Hiroki; Kodaira, Tsuneo

JAERI-Research 95-077, 28 Pages, 1995/10

JAERI-Research-95-077.pdf:2.49MB

no abstracts in English

Journal Articles

Observation of FBR-type fuel rod melting in void under power transients

Sobajima, Makoto; Katanishi, Shoji;

SMiRT 11 Transactions,Vol. C, p.191 - 194, 1991/08

no abstracts in English

Journal Articles

Failure behavior of stainless steel clad fuel rod under simulated reactivity initiated accident condition

;

Journal of Nuclear Science and Technology, 23(12), p.1051 - 1063, 1986/12

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

JAEA Reports

Fuel Pin Behavior During UO$$_{2}$$ Pellet Melting

JAERI-M 7503, 10 Pages, 1978/01

JAERI-M-7503.pdf:0.61MB

no abstracts in English

JAEA Reports

An Examination of the Standard Fuel Behavior in NSRR Experiments

JAERI-M 7267, 36 Pages, 1977/09

JAERI-M-7267.pdf:2.01MB

no abstracts in English

Oral presentation

Characterization of melt-solidified (U, Gd, Zr)O$$_{2-x}$$ as simulated corium debris

Morimoto, Kyoichi; Hirooka, Shun; Akashi, Masatoshi; Watanabe, Masashi

no journal, , 

The influence of Gd on characteristics of debris is important for removing the debris from the reactors of Fukushima Daiichi Nuclear Power Plant because subassemblies of nuclear fuels containing Gd$$_{2}$$O$$_{3}$$ were loaded in the some reactor cores. Additionally, it is important to assess the distribution state of Gd from the anxiety of re-criticality caused by the relocation of debris while removing them. In this study, sintered pellets of (U$$_{0.95-y}$$Gd$$_{0.05}$$Zr$$_{y}$$)O$$_{2-x}$$ (y=0,0.5, 2-x=1.989-2.000) were melted and solidified to prepare specimens of simulated corium debris. Phase states and fundamental properties of them were evaluated.

Oral presentation

Visualization of the boron distribution in simulated melted core material by neutron energy resolving method

Kai, Tetsuya; Abe, Yuta; Matsumoto, Yoshihiro*; Parker, J. D.*; Shinohara, Takenao; Oishi, Yuji*; Nagae, Yuji; Sato, Ikken

no journal, , 

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