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A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

AbstractThis work studied a conceptual process to recover uranium alone from spent nuclear fuel using high-alkaline carbonate media with hydrogen peroxide for the purposes of reducing the volume of high-level active waste and recycling of uranium from the spent fuel with greatly enhanced proliferation resistance, environmental friendliness, and operational safety. The transuranium (TRU) elements were evaluated to be undissolved and precipitated together with other fission products during the oxidative leaching of uranium from the spent fuel. The leaching ratio of uranium dioxide to TRU dioxide from spent fuel in the carbonate solution with H2O2was estimated to be more than about 108. Only Cs, Tc, Mo, and Te among the major fission products in the spent fuel were dissolved together in the carbonate solution. In the carbonate solution with H2O2, UO2was dissolved in the form of uranyl peroxo-carbonato complex ions, which could be recovered in the form of uranium peroxide precipitate with a very low solubility by acidification of the solution in a succeeding step. All the inorganic salts of Na2CO3, NaOH, and HNO3used in the process suggested could be almost completely recovered and recycled into the process again without any generation of secondary wastes.

Although tritium gas is produced a small amount comparing to the other fission products, it should be controlled for long half-life, high residence time, high isotopic exchange rate and ease of assimilation into living matter. A demand for lengthening the life time of disposal site needs treatment of spent nuclear fuel to reduce the volume of high level radwaste. This study is for trapping the tritium gas during dry processes carried out in oxygen condition, having a potential of exposure into operators and environment. The experiments were performed into two different hydrogen concentration ranges, i.e. 1000 ~ 9000 ppm of high range of hydrogen concentration and 25 ~ 100 ppm of low range of hydrogen concentration. At high range of hydrogen concentration, H2O conversion ratio at 400°C indicated above 98 % up to 7000 ppm, and 100% at 450°C. All the results of H2O conversion ratio at the low range of hydrogen concentration are represented close to 99%. Converted H2O vapor was adsorbed at the Molecular sieve close to 100%. More than 98% of H2O conversion ratio was attained up to 4 cm/s of linear gas velocity, whereas over 99% for the low range of hydrogen concentration. Some catalytic effect of ability of conversion hydrogen into H2O on stainless steel was studied.

AbstractQuantitative analysis of the fission gas release characteristics during the voloxidation and oxidation and reduction of oxide fuel (OREOX) processes of spent pressurized water reactor (PWR) fuel was carried out by spent PWR fuel in a hot cell of the DUPIC Fuel Development Facility. The release characteristics of 85Kr and 14C fission gases during voloxidation process at 500°C are closely linked to the degree of conversion efficiency of UO2to U3O8powder, and it can be interpreted that the release from grain boundary would be dominated during this step. Volatile fission gases of 14C and 85Kr were released to near completion during the OREOX process. Both the 14C and 85Kr have similar release characteristics under the voloxidation and OREOX process conditions. A higher burnup spent fuel showed a higher release fraction than that of a low burnup fuel during the voloxidation step. It was also observed that the release fraction of semivolatile Cs was ~16% during a reduction at 1000°C of the oxidized powder, but over 90% during the voloxidation at 1250°C.

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and fuel handling system in the reactor core by both experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design, which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high-power and high-burnup conditions even though some material properties, such as the thermal conductivity, are a little lower compared to the uranium fuel. However, it is required that the current DUPIC fuel design be changed slightly to accommodate the high internal pressure of the fuel element. It is also strongly recommended that more irradiation tests of the DUPIC fuel be performed to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

The thermal diffusivity and the thermal expansion of simulated DUPIC fuel were measured using a laser flash apparatus in the temperature range from room temperature to 1473 K and dilatometer in the temperature range from room temperature to 1900 K, respectively. The relative density of simulated DUPIC fuel is lower than that of UO2fuel. At a low temperature to 600 K, the difference of the relative densities is small and increases as temperature increases. As temperature increases to 1900 K, the relative density of the simulated DUPIC fuel decreases to 93.764 % of the initial density at 298 K. The thermal conductivity of simulated DUPIC fuel is lower than that of UO2fuel. The difference of thermal conductivity between simulated DUPIC fuel and UO2fuel is large at room temperature, and it decreases as the temperature increases. The thermal conductivity of simulated DUPIC fuel decreases from 5.77 W/m.K (78.12 % that of UO2) at 373 K to 2.35 W/nr K (87.79 % that of UO2) at 1473 K.

The change of Np oxidation state in nitric acid and the effect of nitrous acid on the oxidation state were analyzed by spectrophotometry, solvent extraction, and electrochemical methods. The Np extraction with 30 vol.% TBP was enhanced by the adjustment of the Np oxidation state using a glassy carbon fiber column electrode system. The knowledge of electrolytic behavior of nitric acid was important because the nitrous acid affecting the Np redox reaction was generated during the adjustment of the Np oxidation state. The Np solution used in this work consisted of Np(V) and Np(VI) but no Np(IV). The ratio of Np(V) in the range of 0.5M∼5.5 M nitric acid was 32%∼19%. The electrolytic oxidation of Np(V) to Np(VI) in the solution enhanced the Np extraction efficiency about five times higher than without electrolytic oxidation. It was confirmed that the nitrous acid in a concentration of less than about 10−5M acted as a catalyst to accelerate the chemical oxidation reaction of Np(V) to Np(VI).

Abstract: A new static contactor was developed for solvent extraction using capillary phenomena induced among clearances formed within a highly packed fiber bundle. Feeding two immiscible phases cocurrently into the fiber bundle generated a very large liquid-liquid contact area for mass transfer within the fiber bundle without any flow turbulence or drop phenomena. In order to test the characteristics and stability of the fiber bundle contactor, continuous extraction experiments were carried out using the fiber bundle contactor with a TBP-uranyl ion-nitric acid system. The fiber bundle contactor had the same extraction performance as that of an ideal batch extractor with good reproducibility due to the sufficient liquidliquid contact area generated by the packed fiber bundle. A minimum residence time of the aqueous phase within the fiber bundle contactor was required for the extraction system to reach an extraction equilibrium state. In the TBP-uranyl ion-nitric acid system, the residence time was about 1.9 minutes. This contactor was confirmed to be effective enough to perform solvent extraction and to study the extraction kinetics because of the stable and large static liquid-liquid contact area.

Abstract: In order to remove U, Tc, and Np, which are positioning materials or target nuclides for transmutation, from the high-level radioactive waste, conditions of co-extraction and sequential stripping of the nuclides were studied by using 30 vol.% TBP. On the basis of the experiments performed on each element of U, Tc, and Np, a combination of co-extraction of U, Tc, Np rarr Tc stripping rarr Np stripping rarr U stripping was suggested. To enhance the Np extraction yield, the electrolytic oxidation of Np(V) was required at the co-extraction step. For the stripping of Tc 5M HNO3, of Np the electrolytic reduction of Np(VI) to Np(V), and of U 0.3M sodium carbonate were used. Phase ratios (O/A or A/O) were recommended to be of 2-3, for co-extraction and for stripping.

This work studied the anodic dissolution characteristics of a UO2electrode at several potentials which were much higher than the corrosion potential in carbonate solutions of a high concentration at several pHs. The cyclic voltammograms with several wide potential windows and dissolution rates at several constant applied potentials were measured in the carbonate solutions. The corrosion product at the electrode and the current efficiencies of the electrolytic dissolution were evaluated. The peak occurring after the potential region of the oxygen evolution in the voltammogram of the UO2electrode was ascribed to the suppression of the oxygen evolution by a corrosion product of UO2CO3. The effective dissolution of UO2in a carbonate solution could be obtained at an applied potential such as +4V(vs SSE) or more, which was with an overpotential of oxygen evolution high enough to rupture the corrosion product at the electrode surface, rather than the potential for the oxidation of UO2grain as UO22+before the oxygen evolution. The corrosion potential of UO2decreased with pH in the carbonate solution. And the dissolution rate and current efficiency of UO2increased with decrease of pH in the carbonate solution.