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Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio
JAEA-Data/Code 2019-018, 22 Pages, 2020/01
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.
Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06
Times Cited Count:11 Percentile:70.33(Nuclear Science & Technology)no abstracts in English
Mori, Takamasa; Kojima, Kensuke*; Suyama, Kenya
JAEA-Research 2018-010, 57 Pages, 2019/02
In order to estimate applicability of the statistical geometry model (STGM) of MVP/GMVP, a parametric study in infinite geometry and criticality safety analyses for direct disposal of spent fuel in simple finite geometry have been carried out by using the MVP Monte Carlo code. It has been found that calculations with STGM for larger fuel spheres give larger thermal utilization factors and larger infinite multiplication factors compared with explicit random models in the range of fuel sphere packing fraction between 6.5 % and 63.3 %. Substantial differences are not observed between the results with two nearest neighbor distributions (NNDs); that given by the MCRDF code and the analytical expression based on a statistically uniform distribution. It is inferred that the overestimation by STGM is caused by the facts that STGM cannot take account of the surroundings of each neutron, whether a fuel sphere rich region or a water moderator rich one, because STGM always uses an NND averaged over such surroundings and that STGM, therefore, cannot take the effect of consecutive scatterings in the water moderator into account.
Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki
JAEA-Data/Code 2018-016, 79 Pages, 2019/01
FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.
Nakamichi, Shinya; Hirooka, Shun; Sunaoshi, Takeo*; Kato, Masato; Nelson, A.*; McClellan, K.*
Transactions of the American Nuclear Society, 113(1), p.617 - 618, 2015/10
Cerium dioxide has been used as a surrogate material for plutonium dioxide. Dorr et al reported the use of hyper-stoichiometric conditions causes the start of shrinkage of (U,Ce)O at low temperature compared with the sintering in reducing atmosphere. However, the precise stoichiometry of the samples investigated was not controlled or otherwise monitored, preventing any quantitative conclusions regarding the similarities or differences between (U,Ce)O and (U,Pu)O. The motivation for the present work is therefore to compare the sintering behavior of MOX and the (U,Ce)O MOX surrogates under controlled atmospheres to assess the role of oxygen defects on densification in both systems.
Kakehi, Isao;
JNC TN9400 2000-054, 84 Pages, 2000/04
This report describes accomplishment of the study on the quality of vipac (vibro-packed) oxide fuel obtained by pyrochemical processing (molten salt electrolytic processing). This study is intended to contribute to the design study of the pyro-reprocessing-vipac fuel recycling system of oxide fuel. In this study, vibro-packing experiment has been conducted using granular U0 obtained by molten salt electrolytic processing (cold experiment). The oxide pyro process developed by Research lnstitute of Atomic Reactors (RIAR) is the method in which the sintered oxide is electrically deposited on the cathode at approximately 600C. 0xide granules for vipac fuel are obtained by crushing the oxide deposited on the cathode. This process is also developed as recycle process because it is capable of FP separation. Also in Japan, this process is studied as one of the new FBR fuel recycling systems. ln this study, we made an effort to clarify the mechanisms of vibro-packing of the electrically obtained granules, which influence on the effective parameters of vibro-packing density and fuel particles size distribution in the fuel cladding in case of non-sphere particles of the granules. As a result of the study, smear density of 75% and almost uniform distribution of U0 particles have been taken in the experiment, and much knowledge for the improvement of the vibro-packing quality has been found. And the possibility of the smear density over 80% and the uniform distribution of U0 particles has been suggested in this study.
Kato, Naoki*; Muta, Hiroaki*; Tanaka, Kosuke; Matsuda, Tetsushi*; Oishi, Yuji*; Kurosaki, Ken*; Yamanaka, Shinsuke*
no journal, ,
Physical properties of sintered UO specimens containing over 10% simulated MA were investigated. Based on the measured data, a prediction method for the properties was introduced.
Muta, Hiroaki*; Kato, Naoki*; Tanaka, Kosuke; Matsuda, Tetsushi*; Oishi, Yuji*; Kurosaki, Ken*; Yamanaka, Shinsuke*
no journal, ,
Effect of MA and FP addition on thermo-mechanical properties of UO were investigated.
Moroi, Yuriko*; Kirishima, Akira*; Akiyama, Daisuke*; Sato, Nobuaki*; Kitamura, Akira; Kimuro, Shingo
no journal, ,
Development of spent nuclear fuel direct disposal system is one of important options in Japan to maintain flexibility of the back-end strategy of nuclear fuel cycle. Other countries like Sweden and Finland advance in research and development of the direct disposal system. However, it is known that some groundwater in Japan contains higher concentration of carbonate ion than that in Sweden or Finland. Therefore, the effect of carbonate ion on the dissolution rate of UO has to be discussed to evaluate feasibility of the direct disposal system in Japan.
Moroi, Yuriko*; Kirishima, Akira*; Akiyama, Daisuke*; Sato, Nobuaki*; Kitamura, Akira; Kimuro, Shingo
no journal, ,
Direct disposal of spent nuclear fuel is considered as an alternative option of geological disposal of high level radioactive wastes. In this case, the dissolution speed of uranium should be one of the most important parameter. In this study, the dissolution behavior of UO in the simulated groundwater contains high concentration of carbonate ion was investigated, then, it was revealed that uranium dissolution was promoted by the carbonate ion.
Serizawa, Hiroyuki
no journal, ,
My investigation on cavities in ceramics was triggered by the unexpected discovery of a polyhedral cavity in a UO matrix. The SEM image that attracted my attention was a cavity observed in the fracture surface of a single crystal of UO that was heat-treated in helium at 90 MPa, followed by annealing at 1573 K for 1 h. It was clear that the cavity was a negative crystal that was formed by the precipitation of helium during heat treatment after Hot Isostatic Pressing (HIP) injection. In a series of experiments, I noticed that the shape of the negative crystal changes depending on the heat-treatment history. A truncated octahedron-type, an octa-triacontahedron-type, and a pentacontahedron-type negative crystal were observed. Our study implies that the shape of the negative crystal should change depending on the helium inner pressure enclosed in the negative crystal.
Serizawa, Hiroyuki
no journal, ,
My investigation on cavities in ceramics was triggered by the unexpected discovery of a polyhedral cavity in a UO matrix. In a series of experiments, I noticed that the shape of the negative crystal changes depending on the heat-treatment history. In general, it is difficult to control arbitrarily the shapes of these polyhedral negative crystals embedded in a solid medium; however, the shape can easily be controlled using the helium injection method. Our research team named the shape controlled negative crystal as image crystal.
Morimoto, Kyoichi; Ogasawara, Masahiro*
no journal, ,
The heat capacity of MOX fuel is one of the important thermophysical properties. To evaluate the heat capacity of MOX fuel, the heat capacity of PuO is required because the heat capacity of MOX fuel is generally calculated from the compositional average of those of UO and PuO. The experimental results of the heat capacity of PuO are very scarce. In this study, the enthalpy of PuO pellet was measured in the temperature range from 980 to 2160 K with a drop calorimeter. In the measurement, the pellet was loaded in a tungsten container and a rhenium inner container was applied to prevent the reaction between the specimen and the tungsten container. It was found that the enthalpy increased at a constant rate with increasing temperature up to about 1900 K, and that above about 1900 K, its rate tended to increase with increasing temperature. It means that the heat capacity is raised when temperature exceeds about 1900 K.
Sato, Isamu; Matsumoto, Taku; Koyama, Shinichi; Arima, Tatsumi*
no journal, ,
Diffusion coefficients of fission product elements, Zr, Nd, Ba and Sr in UO crystal which are substituent cations for U were evaluated with molecular dynamics. The diffusion coefficients of Nd and Ba/Sr which are trivalent and divalent cations are greater than those of U and Zr that are tetravalent cations.
Fujimura, Koji*; Oki, Shigeo; Takeda, Toshikazu*
no journal, ,
no abstracts in English
Kato, Masato; Watanabe, Masashi; Nakamura, Hiroki; Machida, Masahiko
no journal, ,
no abstracts in English