An investigation on recycling the recovered uranium from electro-refining process in a CANDU reactor
Research highlights
► We perform basic calculations for recycling the recovered uranium from electro-refining process into a CANDU reactor. ► We examine impurity effects in the recovered uranium and handling possibility. ► Increasing impurity will change the effects of safety parameters of CANDU reactor with a decreased burnup. ► Increasing impurity will make it difficult to treat the recovered uranium by bare hands. ► Performance of recovered uranium depends on estimated impurities in electro-refining process.
Introduction
Recently, advanced nuclear countries are taking into consideration of a closed fuel cycle using the recycled fuel in order to maximize the utilization of the limited uranium resources. For example, lots of works have been performed to reuse the recovered uranium due to increasing uranium fuel cost. The weight percent of U-235 in a light water reactor spent fuel is still high between 0.8 wt.% and 1.0 wt.% depending on the irradiation histories. Thus, it is sufficiently re-utilized in a heavy water reactor such as CANDU reactor which usually uses a natural uranium fuel. The reuse of recovered uranium will bring not only a large economical profit and save of uranium resources but also an alleviation of burden on the management and disposal of the spent fuel. In several previous researches on recycling of recovered uranium, it was assumed that the recovered uranium was obtained form the aqueous reprocess (Ellis, 2007, Suk, 2001). And they found that there was a sufficient possibility to reuse recovered uranium in terms of a reactor’s characteristics as well as the fuel performance.
Another example is the DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) program, which recycle the uranium with fission products simultaneously. Therefore the DUPIC fuel provides a high proliferation resistance but it requires remote technology. Korea Atomic Energy Research Institute (KAERI) has performed and demonstrated the fundamental technologies for the DUPIC program (Yang et al., 2006). KAERI also has developed pyroprocessing to recycle the spent nuclear fuel. The recovered uranium from electro-refining process of pyroprocessing contains some TRUs and fission products as impurities. Recently, a preliminary works have been done for reuse of the recovered uranium from the electro-refining process and it exhibits a slightly different behavior from the previous recycling options (Park et al., 2008).
In this paper, further investigation on the recovered uranium including the reactor’s characteristics and radioactivity analysis extending the previous work. Especially reactor safety parameters are analyzed based on the lattice calculations which are performed with the WIMS_AECL (version 2-5b-7) for the CANFLEX bundle (Donnelly, 1986). And the handling concern was also treated through the source term analysis was performed by the ORIGEN-S, a subprogram of the SCALE-5.0 (Gauld et al., 2004) and surface dose analysis by MCNPX2.5 (Pelowitz, 2005). From the source analysis, it is quantified that the amount of activities and the neutron and the gamma spectrums which are main data for the surface dose rate analysis. The compatibility of handling the recovered uranium with impurities may be an issue when considering the recovered uranium from the electro-refining process.
In Section 2, the characteristics of the recovered uranium by the depletion analysis of the PWR fuel are described. Lattice calculation and safety parameter analysis are provided in Section 3. The shielding analysis for handling concern is given in Section 4. Finally, the conclusion is presented in Section 5.
Section snippets
Characteristics of the recovered uranium from pyroprocessing
As late of 2008, the spent nuclear fuel of Korea were produced about 10,000 ton and most of them are stored temporarily in the nuclear plant site. The inevitable product of the nuclear power, the spent nuclear fuel, becomes the prior problem in the nuclear industry. Therefore, Korea just planned to recycle the spent nuclear fuel with the preprocessing for the next generation reactor. In order to recover uranium and TRU elements from spent nuclear fuels, the pyroprocessing has been developed
Lattice calculation and safety parameters for the recovered uranium loaded CANDU reactor
Based on the inventories obtained from the ORIGEN-S, a simple lattice calculation was performed with the WIMS_AECL code for the CANDU reactor. The lattice geometry of the CANFLEX bundle which contains 43 rods was taken into consideration for the lattice calculation and reactor characteristics data were obtained including the discharge burnup and the effective multiplication factor, and the relative bundle peak power. The discharge burnup was estimated based on the previous work which modified
Surface dose rate of the recovered uranium
When treating the recovered uranium with hand, it is required to check that a surface dose rate of the recovered uranium with some impurities, such as TRU and fission products. In order to determine the source term of the recovered uranium, the previous results were used. Table 6 shows the activity, gamma and neutron spectra for various cases including natural uranium. The recovered uranium without any impurities (Case 1) exhibits 3.61 Ci which is slightly higher than that of natural uranium due
Conclusion
The recovered uranium from electro-refining process could be recycled into a CANDU reactor from the results of the lattice calculations and safety parameters. From the safety parameter analysis, the temperature coefficients and void reactivity of the recovered uranium fuel do not have a significantly different behavior compared with the normal CANDU fuel. Although the effect of impurities such as TRU and fission products was insignificant to the reactor characteristics, the surface dose rate
Acknowledgements
This work has been carried out under the Nuclear Research and Development program of the Korea Ministry of Science and Technology. And this work was supported by Grant No. KSC-2008-S01-0006 from Korea Institute of Science and Technology Information.
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