The separation of zirconium and hafnium using tributyl phosphate (TBP)-Dodecane extractants in nitric acid medium was performed. Zirconium oxychloride, used as extraction feed, was obtained from the synthesis of Kalimantan zircon sand concentrate smelted using NaOH. The extraction process was carried out by dissolving chloride-based metals in nitric acid media in the presence of sodium nitrate using TBP-Dodecane as an extractant. Some of the extraction parameters carried out in this study include variations in organic phase and aqueous phase (O/A), variations in contact time, and variations in nitric acid concentration. Extraction was carried out using a mechanical shaker according to the parameter conditions. X-ray fluorescence (XRF) was used for elemental (Zr and Hf) composition analysis of the aqueous solution. The results showed that zirconium was separated from hafnium at optimum conditions with an organic/aqueous ratio of 1:5, contact time of 75 min, and an HNO3 concentration of 7 M. The resulting separation factor of zirconium and hafnium using TBP-Dodecane was 14.4887.

It has been known that as oxide layer (ZrO2) forms on the nuclear fuel cladding during irradiation in nuclear power plants, the corrosion kinetics are influenced by various parameters such as chemical environments. One of those environments, crud deposition driven by coolant chemistry has an adverse effect on the formation of oxide (ZrO2) and leads to increase thickness of the layer. In this study, crud formation was performed through loop experiment equipment on the surface of intentionally-made oxide layer (ZrO2) on cladding tubes and then the composition and characteristics of cruds were examined for the investigation of nuclear power plant environment. As a result, various cruds in composition and microstructure were formed depending on the exquisite methods and conditions such as metal ion concentration.

Hydride reorientation is widely known as one of the major degradation mechanisms in Zirconium cladding during dry storage. Some previous theoretical models for hydride reorientation used assumption of an ideal radial basal pole orientation for HCP structure of Zirconium cladding. Under this assumption, circumferential hydride was considered to precipitate in the basal plane while radial hydride was considered to precipitate in the prismatic plane, thereby giving energetical penalty on thermodynamical precipitation of radial hydrides. However, in reality, reactor-grade Zirconium cladding exhibits average 30° tilted texture, adding complexity to the hydride precipitation mechanism. In this study, reactor-grade Zirconium cladding was charged with hydrogen and hydride reorientation -treated specimens were fabricated. Microstructural characterization of hydrides was conducted via following three methods in terms of interface and stored energy. And this study aimed to compare these characteristics between circumferential and radial hydrides. Using Electron Back Scattered Diffraction (EBSD), the interface was investigated assuming that interface lies parallel to the axial axis of the tube. These were further validated with Transmission Electron Microscope (TEM). In addition, Differential Scanning Calorimetry (DSC) analysis was conducted to calculate the stored energy. This investigation is expected to establish fundamental understanding of how hydrides precipitate in Zirconium cladding with different orientations. And it will also increase the predictability of radial hydride formation and help understanding the mechanical behavior of Zirconium cladding with radial hydrides.

Change in chemical constitution of a zirconium sample treated at the simulated condition of a pressurized water reactor (water at 315°C and 15.5 MPa) was investigated using X-ray diffraction (XRD) and Raman spectroscopy. We observed swelling of the zirconium sample as well as the change in its color from silver to gray treated after the pressurized water at high temperature. On the basis of XRD and Raman data, we confirmed that the variation in composition of zirconium specimen from hexagonal Zr to monoclinic ZrO2 occurred at the simulated PWR condition. Therefore, we suggested that the oxidation of zirconium appeared due to the reaction with water at high temperature and pressure, as shown in the following reaction: Zr + 2H2O → ZrO2 + 2H2.

Pyroprocessing is a promising technique for the treatment of damaged fuel debris (corium) generated by severe nuclear accidents. The debris typically consists of (U, Zr)O2 originating from the UO2 fuel and Zr alloy-based cladding. By converting the corium to a metallic form, the principal components of the fuel can be recovered through subsequent electrorefining, allowing for long-term storage or final disposal. A study investigated the reduction of zirconium oxide compounds by Li metal as a reductant in molten LiCl salt. This research explored the feasibility of treating damaged nuclear fuel debris, which mainly consists of (U, Zr)O2. The results showed that ZrO2 was successfully reduced to Zr metal by Li metal in LiCl salt at 650C without the formation of Li2ZrO3. In particular, Zr metal was produced without the formation of Li2ZrO3 when LiCl salt containing a high concentration of Li metal was used. However, Zr metal was produced with Li2ZrO3 when LiCl salt containing both Li metal and Li2O was added. This suggests that the concentration of Li metal in the LiCl salt is an important factor in determining the formation of Li2ZrO3. The study also demonstrated that Li2ZrO3 was partially reduced to Zr metal by Li metal in LiCl salt. This finding suggests that Li metal may be effective in reducing other oxide compounds in molten LiCl salt, which could be useful in the treatment of corium. Overall, the research provides valuable insights into the feasibility of using pyroprocessing for the treatment of corium. The ability to recover and store the principal components of the fuel through electrorefining could have important implications for the long-term management of nuclear waste.

The lattice thermal expansion of zirconium-based samples containing tin, niobium, and iron elements at a temperature range of 30–870°C with intervals of 40°C was studied by in situ hightemperature X-ray diffraction (HT-XRD). The a- and c-axes lattice constants of the hexagonal Zr crystal structure for the zirconium-based samples were calculated by Pawley refinement using the in situ HT-XRD spectra. The a-axis lattice parameters for the zirconium-based samples with tin element overall decreased, whereas those for the samples containing niobium or iron elements are not declined, as compared to those for a pure zirconium sample. It suggests that the lattice thermal expansion along the a-axis direction of the hexagonal Zr crystal structure for zirconium-based samples was suppressed by the tin element. This effect is the greatest when the content of tin element added in zirconiumbased sample is 3wt%. On the other hand, the c-axis lattice parameters for all the zirconium-based samples overall increase as compared to the pure zirconium, indicating no suppression effect by tin, niobium, and iron elements, in contrast to the a-axis lattice constants.

The origin of Fe oxide deposition on zirconium oxide with UV irradiation has been investigated in this study. After 7 day corrosion in the flowing autoclave, Fe based oxide is formed on the zirconium oxidewith UV irradiation at 260°C, 6 MPa DI water. Zircaloy-4 coupon is irradiated with a 200 mW·cm−2 UV, and the dissolved oxygen level is maintained below 100 ppb, and dissolved hydrogen concentration is maintained as 2.5 ppm. Zircaloy-4 coupon supplied from Westinghouse is used for this study. MULTEQ version 4.0 developed by EPRI is adopted to simulate how ions dissolved in water can generate deposits on the zirconium oxide with UV irradiation. ICP-OES data after 30 d corrosion in the flowing loop experiment is used for input file for MULTEQ simulation. The system temperature is set as 260°C, and 2,592 L of water is considered the total amount water into the autoclave (0.06 mL·min−1, 30 d). Total numbers of simulation run is set as 8, and the system pH at 260°C is 6.06. Oxidation potential after run #8 is −0.44 V. From MULTEQ simulation, most Fe is existed as Fe(OH)3 and Fe(OH)2, and Fe ions can also exist, but no Fe metal observed. 5.09 × 10−6 ppm (9.73 ppb) of Fe2+, 2.81 × 10−6 ppm FeOH+, and 3.77 × 10−9 ppm Fe(OH)3are in the system. It can be concluded Fe is existed as ion or hydroxide form in the solution. Two precipitates are found from MULTEQ simulation, First, NiO(s) = 5.21 × 10−5 g (52.1 μg), NiFe2O4 = 8.06 × 10−5 g (80.6 μg), and still they are negligible amount. The total concentration of Fe in the electrolyte is the summation of each Fe species concentration and it is equal to 2.69×10−4 ppm. This value is equivalent to 0.269 μg·kg−1 in the solution. The total water volume of the 30 d experiment is 2,592 L (considering water flow from high-pressure pump), so the amount of Fe from ICP-OES data and MULTEQ results in 2,592 L electrolyte is 697.2 μg. This value is order of magnitudes higher than the mass of Fe from the deposits, which was already an upper estimate based on the assumptions. This clearly shows that Fe ions dissolved in the electrolyte can be the source of Fe3O4 on Zr oxide during corrosion with UV irradiation.

Zirconium(Zr) nuclear fuel cladding tubes are made using a three-time pilgering and annealing process. In order to remove the oxidized layer and impurities on the surface of the tube, a pickling process is required. Zr is dissolved in HF and HNO3 mixed acid during the process and pickling waste acid, including dissolved Zr, is totally discarded after being neutralized. In this study, the waste acid was recycled by adding BaF2, which reacted with the Zr ion involved in the waste acid; Ba2ZrF8 was subsequently precipitated due to its low solubility in water. It is very difficult to extract zirconium from the as-recovered Ba2ZrF8 because its melting temperature is 1031 oC. Hence, we tried to recover Zr using an electrowinning process with a low temperature molten salt compound that was fabricated by adding ZrF4 to Ba2ZrF8 to decrease the melting point. Change of the Zr redox potential was observed using cyclic voltammetry; the voltage change of the cell was observed by polarization and chronopotentiometry. The structure of the electrodeposited Zr was analyzed and the electrodeposition characteristics were also evaluated.

The effects of Nb and Cr addition on the microstructure, corrosion and oxide characteristics of Zr based alloys wereinvestigated. The corrosion tests were performed in a pressurized water reactor simulated-loop system at 360oC. Themicrostructures were examined using OM and TEM, and the oxide properties were characterized by low-angle X-ray diffractionand TEM. The corrosion test results up to 360 days revealed that the corrosion rates were considerably affected by Cr contentbut not Nb content. The corrosion resistance of the Zr-xNb-0.1Sn-yCr quaternary alloys was improved by an increasing Nb/Cr ratio. The crystal structure of the precipitates was affected by a variation of the Nb/Cr ratio. The Zr-Nb beta-enrichedprecipitates were mainly formed in the high Nb/Cr ratio alloy while Zr(NbCr)2 precipitates were frequently observed in the lowNb/Cr ratio alloy. The studies of oxide characteristics revealed that the corrosion resistance was related to the crystal structureof the precipitate.

The production of nuclear fuel cladding tube is expected to increase with the nuclear power plant expansion. Zirconium(Zr) scrap that is generated during manufacturing is also expected to increase. Zr electrorefining experiment was carried out in the fluoride salt of LiF-KF-ZrF4 using multiple electrode for scale up and improving throughput Zr electrorefiner development. The Zr reduction peak observed at -0.8 V(vs.Ni). Polarization behavior showed that the amount of applied current increases because of decreasing cell resistance as the number of cathode increases. Experimental results showed the highest recovery rate about 98% at lowest current density of 25.64 mA/cm2 using 6 electrodes. XRD and TG analysis result show that pure Zr was recovered 99.92% and ICP analysis shows that lower impurity content than conventional impurity content of the Anode(97.8%). Electrorefining consumes energy about 7.15 kWh/kg less than 39.7% compared to the Kroll process using 6 electrode width of 20 mm and height of 65 mm. Because of increasing cell efficiency and recovery rate, using multiple cathode is determined as an efficient technique for scale up electrorefining Zr scrap.