• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.547-556
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• 2011

Two computational methods for determining equilibrium states for the Mark-IV electrorefiner (ER) have been assessed to improve the current computational electrorefiner model developed at University of Idaho. Both methods were validated against measured data to better understand their effects on the calculation of the equilibrium compositions in the ER. In addition, a sensitivity study was performed on the effect of specific unknown activity coefficients-including sodium in molten cadmium, zirconium in molten cadmium, and sodium chloride in molten LiCl-KCl. Both computational methods produced identical results, which stayed within the 95% confidence interval of the experimental data. Furthermore, sensitivity to unavailable activity coefficients was found to be low (a change in concentration of less than 3 ppm).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.347-353
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• 2019

The composition of the reduced fuel produced in the electrolytic reduction process of pyroprocessing affects the concentration change of $UCl_3$, an important operating variable of the electrorefining process. In this study, we examined the concentration change of $UCl_3$ in the electrorefiner according to the content of TRU and RE elements in the reduced fuel and the concentration of $Li_2O$ introduced in the electrorefiner accompanied with the reduced fuel. Considering only the TRU and RE elements, the concentration of $UCl_3$ decreased with increasing the number of electrorefining operation batch. In order to operate one campaign (20 batches) of electrorefining process, it was found that additional injection of $UCl_3$ should be conducted more than 3 times. On the other hand, the concentration of $UCl_3$ in the electrorefiner changed significantly depending on the concentration of $Li_2O$ and, accordingly the number of operable electrorefining batches decreased rapidly, showing that the concentration of $Li_2O$ is an important operating variable in electrorefining. Therefore, the results of this study show that to maintain the concentration of $UCl_3$ in the electrorefiner, the operation mode should be set by taking into account the effect of $Li_2O$ as well as the TRU and RE elements contained in the reduced fuel.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.175-182
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• 2008

The results of process development for the blending of waste salt from the electrorefining of spent fuel with zeolite-A are presented. This blending is a key step in the ceramic waste process being used for treatment of EBR-II spent fuel and is accomplished using a high-temperature v-blender. A labscale system was used with non-radioactive surrogate salts to determine optimal particle size distributions and time at temperature. An engineering-scale system was then installed in the Hot Fuel Examination Facility hot cell and used to demonstrate blending of actual electrorefiner salt with zeolite. In those tests, it was shown that the results are still favorable with actinide-loaded salt and that batch size of this v-blender could be increased to a level consistent with efficient production operations for EBR-II spent fuel treatment. One technical challenge that remains for this technology is to mitigate the problem of material retention in the v-blender due to formation of caked patches of salt/zeolite on the inner v-blender walls.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2004-2010
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• 2022

Quantification of sensitive material is of vital importance when it comes to the movement of nuclear fuel throughout its life cycle. Within the electrorefiner vessel of electrochemical separation facilities, the task of quantifying plutonium by neutron analysis is especially challenging due to it being in a constant mixture with curium. It is for this reason that current neutron multiplicity methods would prove ineffective as a safeguards measure. An alternative means of plutonium verification is investigated that utilizes the (𝛼,n) signature that comes as a result of the eutectic salt within the electrorefiner. This is done by utilizing the multiplicity variable a and breaking it down into its constituent components: spontaneous fission neutrons and (𝛼,n) yield. From there, the (𝛼,n) signature is related to the plutonium content of the fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.71-76
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• 2010

The electrorefining process, one of main processes which are composed of pyroprocess to recover the useful elements from spent fuel, and the domestic development of electrorefiner have been reviewed. The electrorefiner is composed of an anode basket containing reduced spent fuel such as uranium, transuranic and rare earth elements, and a solid cathode, which are in LiCl-KCl eutectic electrolyte. Oxidation (dissolution) reaction occurs on the anode and a pure uranium is electrochemically reduced (deposited) on the solid cathode. By application of graphite cathode, which has a self-scrapping characteristics for the uranium deposits, and a recovery of the fallen deposits by a screw conveyer, a high-throughput continuous electrorefiner with a capacity of 20 kgU/day has been developed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.06a
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• pp.178-178
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• 2006

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.11a
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• pp.277-278
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• 2007

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.83-83
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• 2017

Electrorefining is a key step in pyroprocessing. The electrorefining process is generally composed of two recovery steps - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. Distillation process was employed for the cathode processing. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. In this study, a mesh-covered crucible was investigated for the sat distillation of electrorefiner uranium deposits. A liquid salt separation step and a vacuum distillation step were combined for salt separation. The adhered salt in uranium deposits was efficiently removed in the mesh-covered crucible. The salt distiller was operated simply since repeated cooling - heating step was not necessary for the change of the crucible. The operation time could be reduced by the use of the mesh-covered crucible and the combined operation of the two steps. A method to preserve a vacuum level was proposed by double O-rings during the operation of the distiller with the mesh-covered crucible. After the salt distillation, the salt content was measured and was below 0.1wt% after the salt distillation. The residual salt after the salt distillation can be removed further during melting of uranium metal.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1452-1461
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• 2020

A triple bubbler sensor was tested in LiCl-KCl molten salt from 450 to 525 ℃ in a transparent furnace to validate thermal-expansion corrections and provide additional molten salt data sets for calibration and validation of the sensor. In addition to these tests, a model was identified and further developed to accurately determine the density, surface tension, and depth from the measured bubble pressures. A unique feature of the model is that calibration constants can be estimated using independent depth measurements, which allow calibration and validation of the sensor in an electrorefiner where the salt density and surface tension are largely unknown. This model and approach were tested using the current and previous triple bubbler data sets, and results indicate that accuracies are as high as 0.03%, 4.6%, and 0.15% for density, surface tension, and depth, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.1-10
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• 2018

A simplified flowsheet for pyroprocessing commercial spent fuel is proposed in which the only salt treatment step is actinide drawdown from electrorefiner salt. Actinide drawdown can be performed using a simple galvanic reduction process utilizing the reducing potential of gadolinium metal. Recent results of equilibrium reduction potentials for Gd, Ce, Nd, and La are summarized. A description of a recent experiment to demonstrate galvanic reduction with gadolinium is reviewed. Based on these experimental results and material balances of the flowsheet, this new variant of the pyroprocessing scheme is expected to meet the objectives of minimizing cost, maximizing processing rate, minimizing proliferation risk, and optimizing the utilization of geologic repository space.