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.

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.

Transparent thin films of pure and nickel-doped ZrO2 are grown successfully by sol-gel dip-coating technique. The structural and optical properties according to the different annealing temperatures (300 oC, 400 oC and 500 oC) are investigated. Analysis of crystallographic properties through X-ray diffraction pattern reveals an increase in crystallite size due to increase in crystallinity with temperature. All fabricated thin films are highly-oriented along (101) planes, which enhances the increase in nickel doping. Scanning electron microscopy and energy dispersive spectroscopy are employed to confirm the homogeneity in surface morphology as well as the doping configuration of films. The extinction coefficient is found to be on the order of 102, showing the surface smoothness of deposited thin films. UV-visible spectroscopy reveals a decrease in the optical band gap with the increase in annealing temperature due to the increase in crystallite size. The variation in Urbach energy and defect density with doping and the change in annealing temperature are also studied.

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.