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Separation of platinum-group metals (PGMs: Ru, Rh, Pd) and Mo from high-level liquid waste (HLLW) makes the vitrification process more stable. Aluminum ferrocyanide (AlHCF) has been synthesized, and it is suitable for the simultaneous recovery of PGMs and Mo from HLLW. In this study, an adsorption experiment onto AlHCF was conducted in a spent fuel solution prepared from an irradiated MOX fuel. An adsorption experiment in a simulated HLLW with 26 elements was also conducted as reference. As a result, both Pd and Mo have been successfully adsorbed on AlHCF in both the spent fuel solution and the simulated HLLW. Americium was not adsorbed onto AlHCF as tri-valent lanthanide ions were.

In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FRs were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.

Penetration behavior of radionuclides such as 137Cs into dried concrete material, dried mortar material and epoxy paint for a few dozen days was observed using a solution containing fission products extracted from irradiated fuels to obtain fundamental information on the radionuclide penetration rate and depth. Hardly any radionuclides could penetrate into the epoxy paint. The radionuclide solution penetrated into concrete and mortar materials to a depth of a few millimeters for a few dozen days. The penetration behavior observed near the surface of concrete and mortar materials was similar to the diffusion of nuclides in media such as water-saturated concrete, bentonite and cement materials.

Since the start of the severe accidents at the Fukushima Daiichi Nuclear Power Plant in March 2011, concrete surfaces within the reactor buildings (RBs) have been exposed to radioactive contaminant...

The release behavior of Pu and Am was investigated under the reducing atmosphere expected in sodium cooled fast reactor severe accidents. Irradiated Pu and U mixed oxide fuels were heated at maximum temperatures of 2773 K and 3273 K. EPMA, γ-ray spectrometry and α-ray spectrometry for released and residual materials revealed that Pu and Am can be released more easily than U under the reducing atmosphere. The respective release rate coefficients for Pu and Am were obtained as 3.11 × 10 −4 min −1 and 1.60 × 10 −4 min −1 at 2773 K under the reducing atmosphere with oxygen partial pressure less than 0.02 Pa. Results of thermochemical calculations indicated that the main released chemical forms would likely be PuO for Pu and Am for Am under quite low oxygen partial pressure.