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This study addressed the problem of measuring the total mass of molten salt in a nuclear system such as a nuclear fuel electrorefiner or a molten salt reactor. In theory, soluble tracers can be add...

Molten eutectic LiCl-KCl salt is a widely used electrolyte for electrorefining uranium from spent nuclear fuel. Due to the hygroscopic nature of this salt, such operations must be performed under c...

It has previously been proposed by safeguards experts that curium will track plutonium through a spent fuel pyroprocessing facility, enabling nondestructive assaying of plutonium via counting neutr...

Signature-based safeguards (SBS) is currently being investigated to assist traditional nuclear material accountancy in tracking special nuclear material (SNM) within a fuel cycle facility. SBS involves the identification and detection of signatures from process monitoring data for off-normal operation scenarios that involve the loss or improper movement of SNM. To determine possible realistic signatures, the electrorefiner (ER) process is modeled using the code Enhanced REFIN with Anodic Deposition (ERAD), and the JCC-31 Neutron Coincidence Counter, a nondestructive assay detector, is simulated using MCNPx-POLIMI. The ERAD model is used to determine the elemental composition of the ER cathode deposit, while the MCNPx model is developed to determine the single and double count rates expected for this deposition using ft8 tallies. For the determination of signatures, changes were made in the ER model for current density and diffusion layer thickness. The signatures in terms of both modeled ER and detector o...

Methods to improve fission product salt sorption into zeolite-A have been investigated in an effort to reduce waste associated with the electrochemical treatment of spent nuclear fuel. It was demonstrated that individual fission product chloride salts were absorbed by zeolite-A in a solid-state process. As a result, recycling of LiCl-KCl appears feasible via adding a zone-freezing technique to the current treatment process. Ternary salt molten-state experiments showed the limiting kinetics of CsCl and SrCl2 sorption into the zeolite. CsCl sorption occurred rapidly relative to SrCl2 with no observed dependence on zeolite particle size, while SrCl2 sorption was highly dependent on particle size. The application of experimental data to a developed reaction-diffusion-based sorption model yielded diffusivities of 8.04 × 10-6 and 4.04 × 10-7 cm2/s for CsCl and SrCl2, respectively. Additionally, the chemical reaction term in the developed model was found to be insignificant compared to the diffusion ...

In this study, a diffusion-based kinetic model essential for design and operational analysis of spent nuclear fuel reduction has been developed. The model considers the cathode side of the system t...

A computational model of the Mark-IV electrorefiner is currently being developed as a joint project between Idaho National Laboratory, Korea Atomic Energy Research Institute, Seoul National Univers...

This paper presents results of experiments and modeling for ion exchange of LiCl-KCl-based molten salts with zeolite-A. The experiments examined the equilibrium distributions of various nuclear fuel fission products between the molten salt and zeolite phases. In addition to data that were collected in previous studies, new experiments were run using ternary salts (LiCl-KCl-YCl 3 , LiCl-KCl-LaCl 3 , and LiCl-KCl-PrCl 3 ) and quaternary salts (LiCl-KCl-CsCl-NdCl 3 and LiCl-KCl-CsCl-SrCl 2 ). All contacting experiments were conducted at 500°C with a salt-zeolite contacting period of 24 h to allow for equilibrium to be reached. The developed equilibrium model assumes that there are ion-exchange and occlusion sites, both of which are in equilibrium with the molten salt phase. A systematic approach in estimating the total occlusion capacity of the zeolite-A was developed. The parameters of the model, including the total occlusion capacity of the zeolite-A, were determined from fitting the entire set of experimental data available between previous studies and the current one. Experiments involving ternary salts were used to estimate the parameters of the model, while those involving quaternary salts were used to validate the model.

The electrochemical processing of spent metallic nuclear fuel has been demonstrated by and is currently in operation at the Idaho National Laboratory (INL). At the heart of this process is the Mark-IV electrorefiner (ER). This process involves the anodic dissolution of spent nuclear fuel into a molten salt electrolyte along with a simultaneous deposition of pure uranium on a solid cathode. This allows the fission products to be separated from the fuel and processed into an engineered waste form. A one-dimensional model of the Mark-IV ER has begun to be developed. The computations thus far have modeled the dissolution of the spent nuclear fuel at the anode taking into account uranium (U3+), plutonium (Pu3+), and zirconium (Zr4+). Uranium and plutonium are the two most important elements in the system, whereas zirconium is the most active of the noble metals. The model shows that plutonium is quickly exhausted from the anode, followed by dissolution of primarily uranium, along with small amounts of ...

The results of a recently reported series of bench-scale actinide recovery experiments with liquid cadmium cathodes (LCCs) are subjected to a more detailed analysis in this paper. It is suggested that separation efficiency (SE), not separation factor (SF), should be used to assess the effectiveness of an LCC to separate actinides from rare earth (RE) elements. The common definition of SF for any pair of actinide and RE elements in the molten salt/liquid Cd system is the ratio of their distribution coefficients, which are measured under equilibrium conditions. The definition of SE is broader than that of SF. For any pair of actinide and RE elements in the molten salt/liquid Cd system, SE is the ratio of their distribution coefficients, such as SE Pu-U = D Pu /D U , where D Pu and D u are measured at either equilibrium or nonequilibrium conditions. The relationship of SE with SF is linear and can be expressed as SE Pu-U = D Pu /D U and D Pu = SF Pu-U D U + b. When D Pu and D u are measured under equilibrium conditions, SE is equal to SF. The physical or chemical meaning of the intercept b is not clear. From a mathematical point of view, the absolute values of b reveal the differences between the measured D Pu /D U or SE and SF. The negative values of b indicate that the SE measurement results are smaller than the associated SF. The values of b may be used to evaluate the SE of LCC on electrochemically recovered actinides from fission product elements. An electrochemical model was developed to investigate the mechanism of RE contamination of the actinides collected by the LCC. It was confirmed that REs were electrochemically transported into the Cd phase. A more negative LCC voltage has a stronger impact on the quantities of REs transported into the Cd than those of the actinides.

This article summarizes the observations and analytical results from a series of bench- scale liquid cadmium cathode experiments that recovered transuranic elements together with uranium from a molten electrolyte laden with real fission products. Variable parameters such as the ratio of Pu3+/U3+ in the electrolyte, liquid cadmium cathode voltage, and feed materials were tested in the LCC experiments. Actinide recovery efficiency and Pu/U ratio in the liquid cadmium cathode product under variable conditions are reported in the article. Separation factors for actinides and rare earth elements in the salt/cadmium system are also presented.

A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be dependent only upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

The electrometallurgical treatment (EMT) process has been designed and developed for stabilizing sodium-bonded, metallic fuel into two high-level waste forms. This process has recently been success...