Hideaki Mineo

Applicability of a Model Predicting Iodine-129 Profile in a Silver Nitrate Silica-Gel Column for Dissolver Off-Gas Treatment of Fuel Reprocessing

The applicability of a mathematical model predicting iodine-129 profile in a silver nitrate silica-gel (hereinafter referred to as Ag-S) column for radioactive iodine treatment was examined using actual off-gas from a dissolution step of a spent nuclear fuel reprocessing test rig, where PWR spent fuels with burnups up to 44,000 MWdt−1 were dissolved. The iodine-129 profile predicted by the model showed good agreement with the iodine profile obtained in the experiment. It was suggested that the model would be useful to predict iodine profiles in an Ag-S column operated at 423 K for the off-gas treatment of actual spent fuel dissolution. Also, the values of the parameters used in the prediction, i.e., the effective diffusion coefficient and the Langmuir coefficient, were supposed to be applicable within the volumetric NO2 concentration range lower than 1%.

A Simple Model Predicting Iodine Profile in a Packed Bed of Silica-Gel Impregnated with Silver Nitrate

Based on a simple adsorption theory, a mathematical model was proposed to predict axial iodine profiles of the column of silica gel impregnated with silver nitrate (hereinafter referred to as Ag-S) in an off-gas treatment system for spent fuel dissolution. The unknown parameters of the model: the effective diffusion coefficient Dea and the Langmuir coefficient K were determined by curve fitting of iodine profile experimentally obtained. At 423 K, Dea and K were found to be 5.60×10-7 m2.s-1 and 1.0×105 m3.kg-1, respectively. Using the parameter values, the model could well predict the iodine profiles obtained at 423 K in the previous works under different experimental conditions. Furthermore, the effect of silver contents on the iodine profiles was reasonably predicted. It was suggested that the proposed model is simple and would be useful to predict the iodine profiles in Ag-S adsorbent columns.

Plutonium management in the medium term

Abstract For many years various countries with access to commercial reprocessing services have been routinely recycling plutonium as UO2/PuO2 mixed oxide (MOX) fuel in light water reactors (LWRs). This LWR MOX recycle strategy is still widely regarded as an interim step leading to the eventual establishment of sustainable fast reactor fuel cycles. The OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR) has recently completed a review of the technical options for plutonium management in what it refers to as the “medium term.” For the purpose of the review, the WPPR considers the medium term to cover the period from now up to the point at which fast reactor fuel cycles are established on a commercial scale. The review identified a number of different designs of innovative plutonium fuel assemblies intended to be used in current LWR cores, in LWRs with significantly different moderation properties, as well as in high-temperature gas reactors. The full review report describes these various options and highlights their respective advantages and disadvantages. This paper briefly summarizes the main findings of the review.

Solvent Extraction Behavior of Minor Nuclides in Nuclear Fuel Reprocessing Process

The solvent extraction behavior of minor nuclides such as neptunium and technetium in a current PUREX process and an advanced PUREX process (PARC process) were studied by chemical flow sheet experiments using spent nuclear fuels. The uranium, plutonium, neptunium and technetium fractions distributed in the products and raffinates of the PARC process showed that n-butyraldehyde was an effective reductant of neptunium(VI) in the presence of uranium(VI) and plutonium(IV). It was also found that scrubbing with high acid concentration was effective for technetium separation.

An investigation into dissolution rate of spent nuclear fuel in aqueous reprocessing

A simple equation was proposed for the dissolution rate of spent LWR fuel, of which the change in the dissolution area was estimated by taking into account of the area of the cracks in the pellet caused by thermal stress during irradiation. The applicability of proposed equation was examined using LWR spent fuel dissolution test results in the present study as well as the results obtained by other workers. The equation showed good agreements with the dissolution test results obtained from spent fuel pellets and pulverized spent fuel. It was indicated that the proposed equation was simple and would be useful for the prediction of dissolution of spent LWR fuels.

Extraction Behavior of TRU Elements in the Nuclear Fuel Reprocessing

The extraction behaviors of uranium, plutonium and other nuclides in the fuel reprocessing have been investigated to study the capability of confining the radioactivity in the PUREX process. Reprocessing tests using spent fuel of 44GWd/t have been carried out in alpha-gamma cell equipped with reprocessing equipment in NUCEF(Nuclear Fuel Cycle Safety Engineering Research Facility). The reprocessing equipment consists of a dissolver, six mixer-settlers, off-gas treatment and aqueous waste treatment systems. The mixer-settlers were assigned to diluent washing, co-decontamination, FP scrub, U recovery and U/Pu partitioning steps. The concentration profiles of nitric acid, U, Pu, neptunium and americium were obtained under normal operation of PUREX flow sheet. The distribution fractions of the nuclides were calculated by the concentrations and the flow rates of organic and aqueous solutions. It was found that U and Pu were recovered from the spent fuel dissolver solution and were well separated from each other. About 10% of Np was distributed into the raffinate solution in the co-decontamination step. Almost ail of the Am migrated to the raffinate solution. The experimental results were used for the evaluation of the simulation code ESSCAR(Extraction process simulation code) developed for the safety analysis of reprocessing.

Long-lived nuclide separation for advancing back-end fuel cycle process

Abstract An advanced PUREX process named PARC process having a partitioning function of long-lived nuclides has been developed for establishing the coming generations fuel cycle technologies, which satisfy a sustainable energy security and global environmental protection. Key technologies applied in the process are: 1) removal of carbone-14 and iodine-129 with adsorption techniques in dissolver off-gas, 2) separation of neptunium and technetium by selective reduction of Np(Vl) by normal-butyraldehyde and the high acid scrubbing of technetium, 3) separation of americium from the raffinate of co-extraction step with adsorption technique, 4) solvent-washing with normal-butylamine compounds. The separation efficiency of long-lived nuclides in the process was measured in chemical flow sheet experiments using spent fuels burned up to 44,000 MWD/tU.