Yeong-Keong Ha

LOCAL BURNUP CHARACTERISTICS OF PWR SPENT NUCLEAR FUELS DISCHARGED FROM YEONGGWANG-2 NUCLEAR POWER PLANT

Spent UO₂ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by 238 U in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 μm in the center to 100 μm in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the 236 U to 235 U ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

A STUDY ON OXIDATION TREATMENT OF URANIUM METAL CHIP UNDER CONTROLLING ATMOSPHERE FOR SAFE STORAGE

The U metal chips generated in developing nuclear fuel and a gamma radioisotope shield have been stored under immersion of water in KAERI. When the water of the storing vessels vaporizes or drains due to unexpected leaking, the U metal chips are able to open to air. A new oxidation treatment process was raised for a long time safe storage with concepts of drying under vacuum, evaporating the containing water and organic material with elevating temperature, and oxidizing the uranium metal chips at an appropriate high temperature under conditions of controlling the feeding rate of oxygen gas. In order to optimize the oxidation process the uranium metal chips were completely dried at higher temperature than 300℃ and tested for oxidation at various temperatures, which are 300 ℃, 400 ℃, and 500 ℃. When the oxidation temperature was 400℃, the oxidized sample for 7 hours showed a temperature rise of 60 ℃ in the self-ignition test. But the oxidized sample for 14hours revealed a slight temperature rise of 7℃ representing a stable behavior in the self-ignition test. When the temperature was 500 ℃, the shorter oxidation for 7 hours appeared to be enough because the self-ignition test represented no temperature rise. By using several chemical analyses such as carbon content determination, X-ray deflection (XRD), Infrared spectra (IR) and Thermal gravimetric analysis (TGA) on the oxidation treated samples, the results of self-ignition test of new oxidation treatment process for U metal chip were interpreted and supported.

INDUCTIVELY COUPLED PLASMA MASS SPECTROMETRY FOR THE DETERMINATION OF 237 Np IN SPENT NUCLEAR FUEL SAMPLES BY ISOTOPE DILUTION METHOD USING 239 Np AS A SPIKE

A determination method for 237 Np in spent nuclear fuel samples was developed using an isotope dilution method with 239 Np as a spike. In this method, inductively coupled plasma mass spectrometry (ICP-MS) was taken for the 237Np instead of the previously used alpha spectrometry. 237Np and 239Np were measured by ICP-MS and gamma spectrometry, respectively. The recovery yield of 237 Np in synthetic samples was 95.9±9.7% (1S, n=4). The 237 Np contents in the spent fuel samples were 0.15, 0.25, and 1.06 μg/mgU and these values were compared with those from ORIGEN-2 code. A fairly good agreement between the measurements (m) and calculations (c) was obtained, giving ratios (m/c) of 0.93, 1.12 and 1.25 for the three PWR spent fuel samples with burnups of 16.7, 19.0, and 55.9 GWd/MtU, respectively.

BEHAVIORS OF MOLYBDENUM IN UO₂ FUEL MATRIX

Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of UO₂ fuel. In this study, the distribution of molybdenum in spent UO2 fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a UO₂ matrix and its effect on the oxidation behavior of UO₂ were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.

Studies on the Air-oxidation Behavior of Uranium Dioxide I. Phase transformation from (U1-yGdy)O2 to (Ul-yGdy)3OS

The effect of dopant on the phase transformation from UO2 to U3O8 was investigated using X-ray diffraction (XRD) anlysis. Sample powders of Gd-doped stoichiometric UO2 of various dopant contents (y=0-0.16 in U1-yGdyO2) were oxidized at 250, 305, 340, 420 and 700 °C for 4 hours under continuous air flow and the structural changes were measured. The fractions of (U1-yGdy)O2, (U1-yGdy)4O9, (U1-yGdy)O37 and (U1-yGdy)O38 were identified from the peak intensity of each phase. It is confirmed that a tendency of retardation or inhibition by Gd content in the oxidation reaction from UO2 to U3O8.

Determination of plutonium content in high burnup pressurized water reactor fuel samples and its use for isotope correlations for isotopic composition of plutonium

The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium.

Crystallographic characterization of irradiated U-10Zr and U-10Zr-5Ce metallic fuels

Crystal structures of irradiated U-10Zr and U-10Zr-5Ce metallic fuels at approximately 2.9 at.% burnup were probed by micro-X-ray diffraction, which identified the presence of UO2 and α-U phases in both fuels. Analysis of the lattice parameters of α-U phase region using Pawley refinement showed α-U lattice expansion in the b-axis direction, which was correlated with the irradiation growth of orthorhombic α-U. Furthermore, no FCCI and rim structure were observed in both irradiated metallic fuels.