Reprocessing of spent nuclear fuel in India: Present challenges and future programme

A two-site equilibrium model that previously only accommodated monovalent cations has been extended to include divalent and trivalent cations for ion exchange between zeolite-A and molten chloride salts, a process being considered for concentrating nuclear fission products into high level waste forms. Equilibrium constants were determined by fitting the model to equilibrium data sets for ion exchange between zeolite-A and Cs ternary salt (CsCl-LiCl-KCl), Rb ternary salt (RbCl-LiCl-KCl), Na ternary salt (NaCl-LiCl-KCl), Sr ternary salt (SrCl2-LiCl-KCl), and U ternary salt (UCl3-LiCl-KCl). The results reveal a good fit between the experimental data sets and the model. The two ion exchange sites, framework sites and occluded sites, demonstrate different relative selectivities for the cations. It was found that Sr2_ is the preferred cation in the ion exchange site, and Cs_ is the preferred cation in the occlusion site. Meanwhile, Li_ has the highest combined selectivity when both ion exchange and occlusion sites are considered. Interestingly, divalent and trivalent species are more preferred in the ion exchange site than the monovalent species with the exception of Li_.

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Equilibrium Model for Ion Exchange Between Multivalent Cations and Zeolite-A in a Molten Salt

This paper presents, for the first time, a study to analyze the surface morphology of metal extracted from a high temperature molten salt medium in the electrodeposit using x-ray radiography and computed tomography. Widely used methods such as scanning electron microscopy and inductively coupled plasma-optical emission spectrometry/mass spectrometry are destructive and the related processes are often subject to the air condition. The x-ray imaging can provide rich information of the target sample in a non-destructive way without invoking hydrolysis or oxidation of a hygroscopic sample. In this study, the x-ray imaging conditions were optimized as following: tube voltage at 100 kVp and the current exposure time product at 8.8 mAs in our in-house x-ray imaging system. LiCl-KCl and cerium metals used in this work produced substantially distinguishable contrasts in the radiography due to their distinctive attenuation characteristics, and this difference was well quantified in the histograms of brightness. Electrodeposits obtained by chronoamperometry and chronopotentiometry demonstrated a completely different behavior of electrodeposition even at the same applied charge. In particular, computed tomography and volumetric analysis clearly showed the structural and morphological dissimilarity. The structure of cerium metal in the electrodeposit was successfully separated from the chloride salt structure in the CT image by an image segmentation process.

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Selective morphological analysis of cerium metal in electrodeposit recovered from molten LiCl-KCl eutectic by radiography and computed tomography

Application of multivariate analysis techniques to safeguards of the electrochemical treatment of used nuclear fuel

The most advanced methodology for the pyroprocessing of spent nuclear fuel is the electrorefining of uranium metal in LiCl–KCl eutectic, in which uranium is solubilized as U(III). The production of U(III) in LiCl–KCl eutectic by the chlorination of uranium metal using BiCl3 has been performed for research purposes. In this work, this reaction was studied in-situ by visual observation, electronic absorption spectroscopy and electrochemistry at 450 °C. The most likely mechanism has been determined to involve the initial direct oxidation of uranium metal by Bi(III) to U(IV). The dissolved U(IV) then reacts with unreacted uranium metal to form U(III).

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Preparation of uranium(III) in a molten chloride salt: a redox mechanistic study

Pyroprocessing technology has been actively developed at Korea Atomic Energy Research Institute (KAERI) to meet the necessity of addressing spent fuel management issue. This technology has advantages over aqueous process such as less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, and compact equipments. This paper describes the pyroprocessing technology development at KAERI from head-end process to waste treatment. The unit process with various scales has been tested to produce the design data associated with scale-up. Pyroprocess integrated inactive demonstration facility (PRIDE) was constructed at KAERI and it began test operation in 2012. The purpose of PRIDE is to test the process regarding unit process performance, remote operation of equipments, integration of unit processes, scale-up of process, process monitoring, argon environment system operation, and safeguards-related activities. The test of PRIDE will be promising for further pyroprocessing technology development.

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Current Status of Pyroprocessing Development at KAERI

This work studied a way to reclaim uranium from contaminated UO2 oxide scraps as a sinterable UO2 powder for UO2 fuel pellet fabrication, which included a dissolution of the uranium oxide scraps in a carbonate solution with hydrogen peroxide and a UO4 precipitation step. Dissolution characteristics of reduced and oxidized uranium oxides were evaluated in a carbonate solution with hydrogen peroxide, and the UO4 precipitation were confirmed by acidification of uranyl peroxo–carbonate complex solution. An agglomerated UO4 powder obtained by the dissolution and precipitation of uranium in the carbonate solution could not be pulverized into fine UO2 powder by the OREOX process, because of submicron-sized individual UO4 particles forming the agglomerated UO4 precipitate. The UO2 powder prepared from the UO4 precipitate could meet the UO2 powder specifications for UO2 fuel pellet fabrication by a series of steps such as dehydration of UO4 precipitate, reduction, and milling. The sinterability of the reclaimed UO2 powder for fuel pellet fabrication was improved by adding virgin UO2 powder in the reclaimed UO2 powder. A process to reclaim the contaminated uranium scraps as UO2 fuel powder using a carbonate solution was finally suggested.

Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

Kweon-Ho Kang

Papers

Current Status of Pyroprocessing Development at KAERI

Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution