Oxidation and reduction behaviors of a prototypic MgO-PuO inert matrix fuel

Miwa, Shuhei; Osaka, Masahiko

Journal of Nuclear Materials, 487, p.1 - 4, 2017/04

https://doi.org/10.1016/j.jnucmat.2017.01.048

Oxidation and reduction behaviors of prototypic MgO-based inert matrix fuels (IMFs) containing PuO were experimentally investigated by means of thermogravimetry. The oxidation and reduction kinetics of the MgO-PuO specimen were determined. The oxidation and reduction rates of the MgO-PuO were found to be low compared with those of PuO. It is note that the changes in O/Pu ratios of MgO-PuO from stoichiometry were smaller than those of PuO at high oxygen partial pressure. From these results, it can be said that MgO matrix lower the oxygen supply and release of PuO, which is preferable as the minor actinides incineration devices, since the high oxygen potentials of minor actinide oxides can cause certain problems in terms of thermochemical aspects such as enlarged cladding inner-surface corrosion.

Behavior of YSZ based rock-like oxide fuels under simulated RIA conditions

Nakamura, Takehiko; Kusagaya, Kazuyuki*; Sasajima, Hideo; Yamashita, Toshiyuki; Uetsuka, Hiroshi

Journal of Nuclear Science and Technology, 40(1), p.30 - 38, 2003/01

https://doi.org/10.3327/jnst.40.30

Pulse irradiation tests of three types of ROX fuel, i.e. YSZ single phase, finely mixed two phase composite of YSZ and spinel, and the other composite of larger YSZ particles dispersed in spinel matrix, were conducted in the NSRR to investigate their behavior under RIA conditions. Owing to their lower melting temperatures than that of UO fuel, melting of ROX fuel occurred while the cladding was still solid and intact in the accident conditions. Therefore, consequence of the ROX fuel failure was quite different from that of UO fuel. When the ROX fuels failed, a considerable amount of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below 12 GJ/m. In terms of the failure threshold, on the other hand, the ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m, which was comparable to that of UO fuel. The results highlighted controlling parameters on the fuel behavior under the RIA conditions.

Energy transfer and thermal conductivity through inert matrix and nuclear fuel analogous materials

C.Degueldre*; Takano, Masahide; Omichi, Toshihiko; Fukuda, Kosaku; P.Heimgartner*; T.Graber*

JAERI-Research 97-087, 19 Pages, 1997/11

JAERI-Research-97-087.pdf:0.87MB

no abstracts in English

Pyroprocessing of ZrN-based nitride fuels

Hayashi, Hirokazu; Sato, Takumi

no journal, , 

Transmutation of long-lived radioactive nuclides including minor actinides (MA: Np, Am, Cm) is effective to reduce the burden of high level radioactive wastes and using repositories efficiently. Uranium-free nitride fuel has been chosen as the first candidate fuel for MA transmutation using accelerator-driven system (ADS) in Japan Atomic Energy Agency (JAEA) under the double strata fuel cycle concept. To improve the transmutation ratio of MA, reprocessing of spent MA fuel and reusing the recovered MA is necessary. Our target is to transmute 99% of MA arisen from commercial power reactor fuel cycle, with which the period until the radiotoxicity drops below that of natural uranium can be shorten from about 5000 years to about 300 years. Each reprocessing process is required to recover 99.9% of MA to meet the target. Typical composition of the solid solution type (MA,Pu,Zr)N fuel is considered as 30 wt.% of MA nitride, 20 wt.% of Pu nitride, and 50 wt.% of ZrN (dilution material to adjust the power density). Pyroprocessing has been proposed to adopt for reprocessing of the spent MA nitride fuel. This paper summarizes the status of our study on pyroprocessing of ZrN-based nitride fuels.