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During normal operations, the only deposition on a spent nuclear fuel electrorefiner cathode consists of uranium. A computer model, ERAD, was used to simulate conditions under which the electrorefiner salt was progressively contaminated from spent nuclear fuel processing without operator intervention. The model was adjusted to reflect three scenarios: one in which salt removed with every cathode deposit was returned but the UCl 3 concentration was never adjusted, one in which the salt was not removed and UCl 3 concentration was never adjusted, and one in which the Pu/U ratio in the salt was allowed to reach a value of one before starting the first scenario. Under the base conditions simulated, uranium initially deposits alone but eventually plutonium begins to co-deposit along with uranium. Up to 350 g of plutonium was calculated to deposit along with 8.3–9.5 kg of uranium. While precise tracking of changes in the salt composition may not be possible, the model was able to identify markers for plutonium deposition by tracking changes in electrode potential. This type of marker should ultimately provide inspectors with more options for situations when they have limited access to electrorefiner data. [ABSTRACT FROM AUTHOR]

Abstract: A model has been developed and tested for its ability to describe the ion exchange of fission product species between zeolite and a molten salt bath used for pyroprocessing of spent nuclear fuel. The model is based on a system at equilibrium and assumes species interacting with zeolite are chemically independent of each other. This assumption was tested via a set of salt–zeolite contact experiments. Zeolite 4A was batch contacted with Cs ternary salt (CsCl–LiCl–KCl), Nd ternary salt (NdCl3–LiCl–KCl) or Cs–Nd quaternary salt (CsCl–NdCl3–LiCl–KCl) until ion exchange steady state was reached. Results from ternary salt and zeolite chemical analyses were used to determine model parameters for all species present. These parameters, in turn, were applied to compute estimated compositions of zeolite contacted with quaternary salt. The estimated compositions compare favorably with results from chemical analyses of the zeolite–quaternary salt experiment samples. Based on the data collected, the model is modified for predictive capability for a range of salt compositions. [Copyright &y& Elsevier]

Abstract Electrochemical methods have been developed to measure metal ion concentrations in spent nuclear fuel electrorefiner by correlating concentration with current response. Analytes UCl 3 and GdCl 3 were studied in LiCl-KCl eutectic solvent at 773 K. A mixture matrix, which consisted of varying concentrations of UCl 3 and GdCl 3 from 1 to 10 wt% and from 1 to 3 wt%, respectively, was tested. Optimized normal pulse voltammetry (NPV) measurements were made on this mixture matrix and used for the concentration correlations. The accuracy of the prediction was improved by including solution resistance compensation and minimizing the working electrode surface area. The presence of GdCl 3 showed no effect on the U3+ NPV reduction current, but high U concentrations affected the Gd3+ NPV reduction current. This is speculated to be due to increased migrational current in the molten salt mixture. The average relative measurement errors obtained were 1.6% and 2.7% for UCl 3 and GdCl 3 respectively. [ABSTRACT FROM AUTHOR]

Several countries have shown interest in developing the electrochemical treatment of used nuclear fuel (UNF), commonly termed pyroprocessing. From a proliferation perspective, an advantage of pyroprocessing is its inability to isolate pure plutonium. However, because plutonium is present in the process, it needs to be effectively safeguarded to protect against diversion of plutonium-containing materials that could be post-processed elsewhere. The most complicated unit to safeguard in the process is the electrorefiner where UNF is chemically separated into several material entities. Molten LiCl–KCl serves as the electrolyte into which actinides (including uranium and plutonium), rare earths, and other active metals from the fuel are partitioned after being oxidized to chloride salts. Various voltammetric methods are being developed to measure the concentration of actinides in the molten salt in near-real-time. However, these methods have mostly been applied to molten salt mixtures containing a single actinide. Unfortunately, the presence of multiple actinides will create interferences in the electrochemical responses which could make traditional analysis of voltammetric data inaccurate for determining individual species concentrations. It is proposed to use multivariate techniques to more accurately predict concentrations of multiple actinides from voltammetric data. Two techniques, principal component analysis and partial least squares, are demonstrated on experimental and simulated data for molten salt mixtures containing uranium and plutonium. These techniques consistently yielded more accurate predictions of uranium and plutonium concentrations than simply using peak height. Possible methods of employing multivariate techniques in safeguarding an electrorefiner are also demonstrated. [ABSTRACT FROM AUTHOR]

Abstract: Experimentation, data analysis, and modeling of ion exchange kinetics between fission products (cesium and strontium) and zeolite-A beads in molten LiCl–KCl have been performed to support optimization of an electrochemical process to treat used nuclear fuel. Models based on pseudo-first- and pseudo-second-order sorption as well as diffusion have been adapted and compared to experimental data to assess their validity in describing the system. Individual experiments were performed with different concentrations of CsCl or SrCl2 in the LiCl–KCl salt. Zeolite beads were removed from the molten salt at prescribed intervals of time and prepared for analysis by inductively coupled plasma-mass spectrometry. Results indicate maximum cesium (Cs) and strontium (Sr) loading occurred at approximately 31min and 104min of contact, respectively. The rate of loading and maximum loading were found to increase with increasing initial concentration of Cs or Sr. Data analysis included determination of rate constants and diffusion coefficients of the proposed models for each experimental condition. Results reveal that the diffusion model provides the best fit to the experimental data with average diffusion coefficients of 2.0×10−10 m2 s−1 for Cs and 6.3×10−11 m2 s−1 for Sr. This suggests that chemical diffusion is the dominant mechanism of mass transfer. [Copyright &y& Elsevier]

This paper reports the hygroscopic properties of eutectic LiCl-KCl after absorption into zeolite-4A, up to salt loadings of 75 wt%. Samples of the salt occluded zeolite were hydrated in a humidity chamber at constant temperature and relative humidity for up to 100 h. At up to 45 wt% salt loading, the un-occluded phase of salt consisted primarily of NaCl, which forms when the Na+ ions present in the zeolite framework exchange with Li+ and K+ ions from the eutectic LiCl-KCl. This results in minimal water absorption and corrosion of contacted stainless steel. At greater than 45 wt% salt loading, water absorption and corrosion progressively worsened. The mixture has a significant amount of excess LiCl-KCl, making it highly hygroscopic. This study reveals an option for the intermediate treatment of waste salt from spent nuclear fuel electrorefiners that could facilitate it to be stored in a non-inert atmosphere for extended periods of time before final conversion into a permanent waste form. [ABSTRACT FROM AUTHOR]

Molten salt mixtures containing LiCl-KCl-UCl 3 -CeCl 3 were used as surrogates for molten salt reactor fuel in a study of electrochemical concentration analysis. Current responses to low concentration CeCl 3 (0.5 wt% Ce) in molten eutectic LiCl-KCl with high concentration UCl 3 (5 and 10 wt% U) were measured using cyclic voltammetry (CV) and square wave voltammetry (SWV). CV peaks associated with CeCl 3 were impossible to quantify because of dominant reducing current contribution from UCl 3. SWV results featured peaks from UCl 4 , UCl 3 , and CeCl 3 with minimal baseline interference. A potential hold at −1.5 V for 100 s was effective at cleaning the working electrode in between each SWV scan, which enabled generating reproducible (with ±1.10% average error) peak current results for CeCl 3 over a period of 2 h. The peak current density for CeCl 3 increased by 8.3% from doubling the U concentration from 5 to 10 wt%, thus indicating that this method is promising for measuring minor metal ion concentrations in molten salt reactor fuel with high U concentrations. [ABSTRACT FROM AUTHOR]