Electrochemical behaviors of uranium and plutonium at simultaneous recoveries into liquid cadmium cathodes

Experiments were conducted on simultaneous recovery of uranium and plutonium electrochemically into laboratory scale liquid cadmium cathodes (LCCs) at different U/Pu ratios in the salt phase, and the influence of the salt composition on the recovered amounts of uranium and plutonium, the morphologies of uranium and plutonium in the LCC, and the behavior of americium, which is present as a decay product of plutonium-241, were examined. As a result, it was shown that there is a threshold in the U/Pu ratio in the salt phase between 1/4.3 and 1/1.73 for the successful simultaneous recovery of uranium and plutonium up to 10 wt% in the LCC at high current efficiencies. In the LCC, uranium and plutonium existed in the forms of intermetallic compounds, (U, Pu)Cd₆ and (U, Pu)Cd₁₁, and also pure uranium metal. It was also revealed that americium associates with plutonium according to the separation factor during the LCC operation.     

Characteristic Behavior of Actinides During Electrochemical Reprocessing of Nitride Fuel in Chloride Melts

The results of experiments on anodic dissolution of (U, Pu)N pellets in (LiCl-KCl)_eut-UCl₃-PuCl₃ melt and precipitation of uranium and plutonium on a solid cathode in the presence of a small amount of neptunium and americium are presented. For electrochemical cell voltage 2.1 V, the ratio of uranium and plutonium in the cathode deposit obtained corresponds to the ratio in the electrolyte. Thus the requirement for nonpartitioning of actinides is achieved. Under these conditions, americium and neptunium precipitate simultaneously on the cathode.The conditions for complete precipitation of uranium and plutonium are determined experimentally. A residual concentration of these elements of 0.43 and 0.05 mass %, respectively, in the electrolyte is achieved.__________Translated from Atomnaya Energiya, Vol. 98, No. 3, pp. 201–206, March, 2005.     

Separation of uranium from (U, Th)O2 and (U, Pu)O2 by solid state reactions route

Solid state reactions of UO₂, ThO₂, PuO₂ and their mixed oxides (U, Th)O₂ and (U, Pu)O₂ were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO₂ with NaNO₃ above 500 °C were readily soluble in 2 M HNO₃, whereas ThO₂ and PuO₂ did not react with NaNO₃ to form any soluble products. Thus reactions of mixed oxides (U, Th)O₂ and (U, Pu)O₂ with NaNO₃ were carried out to study the quantitative separation of U from (U, Th)O₂ and (U, Pu)O₂. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.     

Thermodynamic Evaluation of Liquid Cd Cathode Containing U and Pu

In our previous study, a mixture of U and Pu was recovered in liquid Cd cathode from molten salt under various conditions of the U:Pu ratio. Two important things were observed. The first was that three kinds of precipitated phase had been detected in the saturated liquid Cd cathode, such as a U metal and two kinds of U-Pu-Cd compound. The compositions of the compounds were roughly (U+Pu):Cd = 1:11 and (U+Pu):Cd = 1:6. The second was that the U metal had selectively precipitated in the saturated liquid Cd cathode under the condition that the U:Pu ratio is higher than about 0.8 in the liquid Cd phase. In the present study, phase diagrams were evaluated by the CALPHAD method on the liquid Cd cathode containing U and Pu. The U-Pu-Cd compounds were modeled as MCd11-type and MCd6-type, respectively, based on the reported binary phase diagrams of U-Cd and Pu-Cd. The calculated result reasonably agreed with the experimental observations. The variations in the U and Pu activities were estimated along with the U:Pu ratio, which is related to the precipitation of various phases in the liquid Cd cathode.     

Burn up extension with slightly enriched uranium and plutonium fuel in CANDU reactors

A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO₂ fuel, 1.2% enriched UO₂ fuel and for the 0.45% PuO₂-UO₂ fuel. The amount of ²³⁵U and ²³⁸U consumed and ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu produced in the three fuel cycles are also calculated and compared.     

Compression and Decompression Waves in Steam-Water System

The ratio of the γactivities per fission from fission products of ²³⁹Pu and ²³⁵U, and its time dependence were measured by double fission chamber technique. The γ-activity from the fission products of ²³⁹Pu fission was lower than the corresponding activity relevant to ²³⁵U fissions. The ratio varied with the cooling time allowed after irradition.

This ratio was applied to power distribution measurements by γ-scanning method in multi-region cores composed of PuO₂-UO₂ and UO₂ fuels. To obtain the relative power, the measured γ-activities from the fission products in the fuel rods were corrected for the difference between the γ-activities per fission from the fission products.     

Observation of Lithium Isotope Effect Accompanying Electrochemical Insertion of Lithium into Tin

Electrolysis of an organic electrolyte solution containing lithium ions was conducted to observe lithium isotope fractionation accompanying electrochemical insertion of lithium from the electrolyte to tin metal. The experimental setup consisted of a three-electrode electrolysis cell with a tin wire as cathode, lithium foils as anode and reference electrode and 1 M LiPF₆ dissolved in 1:2 volume ratio of ethylene carbonate and methylethyl carbonate as electrolyte and a power supply. The supplied electric energy was mostly consumed for the lithium insertion from the electrolyte to the tin cathode within the range of the cathode potential, relative to the reference electrode potential, from 0.05 V to 0.30 V. The single-stage separation factor increased with increasing cathode potential and seemed to asymptotically approach to the limiting value of 1.015 at 25®C, with ⁶Li being preferentially fractionated into tin metal.     

Criticality Safety Evaluation of Materials Concerning Pyroprocessing

Criticality evaluation of materials concerning pyroprocessing has been performed employing Monte Carlo techniques. Fresh UO₂ fuel and spent PWR fuel have been employed in an electroreducer and, according to the pyroprocessing material flow, their metallic products are also evaluated correspondingly. TRU and Pu metals are evaluated concerning the electrowinner products. Totally, 6 kinds of reflectors and 4 stuffing materials are introduced to simulate the complexity during the pyroprocessing. With the optimized water intrusion models, certain subcritical data have been determined: 33.6 kg for UO₂ fuel in electroreducer, 617.5 kg for spent fuel in electroreducer, 613.6 kg for metallic spent fuel in electroreducer, and 40.0 kg for U in electrorefiner. The conservative cases for electrowinner indicate the subcritical masses of 6.8 kg for Pu and 7.9 kg for TRU. However, with regard to the dilution effect of Cd, the Pu mass needed to reach the subcriticality changes from 11.1 kg with 1.2 kg of Cd to 500.0 kg with 2000.0 kg of Cd. The TRU mass needed to reach subcriticality changes from 13.0 kg with 1.4 kg Cd to 714.0 kg with 2856.0 kg of Cd. TSUNAMI has been employed to clarify the contributions of each nuclide quantitatively. The data obtained would therefore be used in the conceptual design study of pyroprocessing facilities.     

Electrolytic Reduction of Spent Light Water Reactor Fuel Bench-Scale Experiment Results

A series of experiments were performed to demonstrate the electrolytic reduction of spent light water reactor fuel at bench-scale in a hot cell at the Idaho National Laboratory Materials and Fuels Complex. The process involves the conversion of oxide fuel to metal by electrolytic means, which would then enable subsequent separation and recovery of actinides via existing electrometallurgical technologies, i.e., electrorefining. Four electrolytic reduction runs were performed at bench scale using ～500 ml of molten LiCl–1 wt% Li₂O electrolyte at 650°C. In each run, ～50 g of crushed spent oxide fuel was loaded into a permeable stainless steel basket and immersed into the electrolyte as the cathode. A spiral wound platinumwire was immersed into the electrolyte as the anode. When a controlled electric current was conducted through the anode and cathode, the oxide fuel was reduced to metal in the basket and oxygen gas was evolved at the anode. Salt samples were extracted before and after each electrolytic reduction run and analyzed for fuel and fission product constituents. The fuel baskets following each run were sectioned and the fuel was sampled, revealing an extent of uranium oxide reduction in excess of 98%.     

Electrodeposition of Uranium and Transuranic Elements onto Solid Cathode in LiCl-KCl/Cd System for Pyrometallurgical Partitioning

A pyrometallurgical partitioning process is being developed for recovering transuranic elements (TRUs) from high-level liquid waste. In the process, actinides are separated from fission product, especially rare earth elements (REs), by means of an electrorefining technique or a reductive-extraction technique. In this study, electrorefining experiments were carried out in LiClKCl/Cd system to recover actinides from salt bath containing actinides and REs. Uranium and neptunium could be depleted from the salt bath and recovered onto a solid cathode with high collection efficiency and high selectivity. Plutonium and americium, however, were difficult to be recovered at high current efficiency because reduction of Nd³⁺ to Nd²⁺ at about—1.7V consumed cathodic current prior to the deposition of Pu or Am. The rotation of the cathode had rather negative effect against deposition of Am and Pu in case of coexistence of much amount of Nd because Nd²⁺ was removed from the cathode surface quickly and the reaction of Nd³⁺ to Nd²⁺ was promoted. At higher current density, Pu and Am could be recovered onto solid cathode but current efficiency became too low. The result indicated that electrorefining technique in the pyro-partitioning was effective for U and Np but not for Pu and Am.     

On the electrochemical formation of Pu–Al alloys in molten LiCl–KCl

Properties of Pu–Al alloys were investigated in connection with development of pyrochemical methods for reprocessing of spent nuclear fuel. Electroseparation techniques in molten LiCl–KCl are being developed in ITU to group-selectively recover actinides from the mixture with fission products. In the process, actinides are electrochemically reduced on solid aluminium cathodes, forming solid actinide–aluminium alloys. This article is focused on electro-chemical characterisation of Pu–Al alloys in molten LiCl–KCl, on electrodeposition of Pu on solid Al electrodes and on determination of chemical composition and structure of the formed alloys. Cyclic voltammetry and chronopotentiometry were used to study Pu–Al alloys in the temperature range 400–550 °C. Pu is reduced to metal in one reduction step Pu³⁺/Pu⁰ on an inert W electrode. On a reactive Al electrode, the reduction of Pu³⁺ to Pu⁰ occurs at a more positive potential due to formation of Pu–Al alloys. The open circuit potential technique was used to identify the alloys formed. Stable deposits were obtained by potentiostatic electrolyses of LiCl–KCl–PuCl₃ melts on Al plates. XRD and SEM–EDX analyses were used to characterise the alloys, which were composed mainly of PuAl₄ with some PuAl₃. In addition, the preparation of PuCl₃ containing salt by carbochlorination of PuO₂ is described.     

Pu2O3 and the plutonium hydriding process

The role of cubic Pu₂O₃ in the corrosion of PuO₂-coated Pu by H₂ was investigated. Experiments were conducted to demonstrate that nucleation of hydriding is promoted by formation of Pu₂O₃ sites in the oxide layer. The nucleation mechanism based on diffusion of hydrogen through the PuO₂ layer was evaluated and an alternative mechanism based on formation of catalytic Pu₂O₃ sites via the Pu-PuO₂ reaction is proposed. The possibility of active participation of other impurities and inclusions in the dioxide is also discussed.     

The influences of Pu and Zr on the melting temperatures of the UO2–PuO2–ZrO2 pseudo-ternary system

Specimens of (U, Pu, Zr)O₂ were prepared as simulated corium debris that were assumed like debris generated in the severe accident of the Fukushima Daiichi Nuclear Power Plant and their melting temperatures were measured by the thermal arrest technique in order to evaluate the influence of plutonium and zirconium content on the melting temperature of the corium debris. From the evaluation, it was found that the influence of zirconium on the melting temperatures of both (U, Pu, Zr)O₂ and (U, Zr)O₂ was similar and that the melting temperature of (U, Pu, Zr)O₂ had a local maximum value in the Pu-content between 0 and 20 mol%. The UO₂–PuO₂–ZrO₂ pseudo-ternary phase diagram at 2900 and 3000 K was evaluated from the present experimental results and previously reported results.     

Helium behavior in oxide nuclear fuels: First principles modeling

UO₂ and (U, Pu)O₂ solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein.We present density functional theory (DFT) calculations for UO₂, PuO₂ and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO₂ remains semiconducting with He in the octahedral position while UO₂ requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.     

Carbothermic Synthesis of High Purity Plutonium Nitride from Plutonium Oxide

Synthetic conditions of PuN from carbothermic reduction of PuO₂ has been studied in a mixed 8%H₂+92%N₂ stream at a temperature range of 1,270–1,680°C. In the course of both reactions of the carbothermic synthesis of PuN from PuO₂ and the hydrogenation of C, the vaporization loss of Pu was observed. It increased with temperature in the temperature range of 1,350–1,450°C, and reached to a constant value 1.3% of total Pu in the temperature range of 1,450–1,680°C, at which PuN was synthesized at a reaction rate of high enough. The minimum mixing ratio (C/PuO₂, mole ratio) for the formation of high purity PuN depends on temperature. The value is 2.15 for 1,620°C and 2.35 for 1,680°C. The oxygen and carbon impurities in the PuN obtained were found to be 0.095–0.028 and 0.17–0.012w/0, respectively.     

Fundamental Study of the Sulfide Reprocessing Process for Oxide Fuel (I) Study on the Pu,MA and FP Tracer-Doped U3O8

For the recovery of fuel materials from spent nuclear fuel, a novel reprocessing process based on the selective sulfurization of fission products (FP) has been proposed, where FP and minor actinides (MA) are first sulfurized by CS₂ gas, and then, dissolved by a dilute nitric acid solution. Consequently, the fuel elements are recovered as UO₂ and PuO₂. As a basic research of this new concept, the sulfurization and dissolution behaviors of U, Pu, Np, Am, Eu, Cs, and Sr were investigated by γ-ray and α spectrometries in this paper using ²³⁶Pu-, ²³⁷Np-, ²⁴¹Am-, ¹⁵²Eu-, ¹³⁷Cs-, and ⁸⁵Sr-doped U₃O₈ samples. The dependence of the dissolution ratio of each element on the sulfurization temperature was studied and reasonably explained by combining the information of the sulfide phase analysis and the chemical thermodynamics of the dissolution reaction. The sulfurization temperature ranging from 350 to 450°C seems to be promising for the separation of FP and MA from U and Pu, since a clear difference in the dissolution ratio between FP and U was derived by the sulfurization treatment in this temperature range.     

Application of a boron doped diamond (BDD) electrode as an anode for the electrolytic reduction of UO2 in Li2O–LiCl–KCl molten salt

A boron doped diamond thin film electrode was employed as an inert anode to replace a platinum electrode in a conventional electrolytic reduction process for UO₂ reduction in Li₂O–LiCl molten salt at 650 °C. The molten salt was changed into Li₂O–LiCl–KCl to decrease the operation temperature to 550 °C at which the boron doped diamond was chemically stable. The potential for oxygen evolution on the boron doped diamond electrode was determined to be approximately 2.2 V vs. a Li–Pb reference electrode whereas that for Li deposition was around ?0.58 V. The density of the anodic current was low compared to that of the cathodic current. Thus the potential of the cathode might not reach the potential for Li deposition if the surface area of the cathode is too wide compared to that of the anode. Therefore, the ratio of the surface areas of the cathode and anode should be precisely controlled. Because the reduction of UO₂ is dependent on the reaction with Li, the deposition of Li is a prerequisite in the reduction process. In a consecutive reduction run, it was proved that the boron doped diamond could be employed as an inert anode.     

Recovery of Neptunium by Electrolysis of NpN in LiCl-KCl Eutectic Melts

The electrochemical behavior of neptunium nitride, NpN, in the LiCl-KCl eutectic melt containing NpCl₃ at 450, 500 and 550°C was investigated from the viewpoint of the application of electrochemical refining in a fused salt to nitride fuel cycle. The electrochemical dissolution of NpN began nearly at the potential theoretically evaluated, though this reaction was irreversible owing to small partial pressure of N₂ in the salt and the reaction rate was slow. Under the electrolysis in the NpCl₃-LiCl-KCl eutectic melt, NpN was dissolved into the salt as Np³⁺ at the anode, and Np metal was deposited at the cathode. About 0.5 g of Np metal was obtained by heating the deposit containing the salt at 800°C for 3.6 ks.     

A theoretical study of electronic and valence properties of U/PuO2

We consider the basic electronic and thermodynamic properties of the U/PuO₂ phase, paying special attention to the contrasts between the behaviour of this material and that of UO₂. We also report calculated ionisation potentials for Pu which play an important rôle in our analysis.