Mendoza, Paul Michael (2018-02). Experimental Characterization of Pu Separation by Purex on a Low-Burnup, Pseudo-Fast-Neutron Irradiated DUO2 for Product Decontamination Factors and Nuclear Forensics. Doctoral Dissertation.
Thesis
Experimental investigations to determine fission product decontamination factors while employing the Plutonium Uranium Recovery by Extraction (PUREX) process were conducted. PUREX process was performed for a depleted UOv2 (DUOv2) sample irradiated in a pseudo-fast neutron environment in the High Flux Isotope Reactor (HFIR). The DUOv2 sample (0.26 wt% ^235U) was covered with gadolinium to absorb thermal neutrons and irradiated to a low-burnup (4.43 +- 0.31 GWd/tHM) and PUREX process was performed 538 days after the neutron irradiation. Decontamination factors (DF), with respect to Pu, for the elements U, Mo, Ru, Ce, Sm, Sr, Pm, Eu, Nd, Pd, and Cd were measured with mass spectrometry. DFs as well as distribution coefficients (DC) were determined with gamma spectroscopy for Cs, Ru, Ce, and Eu. 30 vol.% tri-n-butyl phosphate (TBP) in a kerosene diluent was used for U/Pu extraction and 0.024 M iron(II) sulfamate in HNO3 with concentrations ranging from 0 to 4 M was used for Pu back-extraction. The Pu in the irradiated fuel was characterized as near-weapons-grade (89.3%^ 239Pu), with 1.5% of the total fuel mass attributed to Pu, 86% to U, roughly 0.3% attributed to fission products and the rest of the mass attributed to oxygen. Two cycles of a four extraction, three backextraction process achieved 93% of Pu recovered in a product solution with less than 1% of the original U. The mathematical expression between DCs and DFs was derived for the PUREX experiment and this expression was used to calculate DFs from DCs. The ratio between DFs determined with DCs as opposed to direct measurement was 1.5, 1.0, 0.99, and 0.91 for Cs, Ru, Ce, and Eu, respectively. Further, a forensic methodology was developed for determining parameters like fuel burnup, scalar neutron flux, neutron irradiation time in reactor, initial^ 235U enrichment, and time since removal from the reactor. Each parameter was determined in the order listed because information from earlier calculations was used in later calculations. The above parameters were determined using single group neutron cross sections and fission yields determined by averaging with five different normalized neutron flux spectra. The five reactor flux spectra were: the initial HFIR sample flux spectrum, the Fast Breeder Reactor (FBR) blanket region, an AP1000, a Pressurized Heavy Water Reactor (PHWR), and an average HFIR flux spectrum. Concentrations determined at the end of the PUREX experiment were used with DF values to calculate unprocessed concentrations, which were used as inputs for the forensic calculation. The flux spectra input which most correctly determined sample history among the non-averaged spectra was the AP1000, which indicates that the sample received the majority of its burnup while in a thermal spectrum. Although the sample was covered with gadolinium to remove the thermal flux, investigations with MCNP revealed that the gadolinium burned out some time near the end of irradiation, and consequently the fuel sample was irradiated for an unknown time in a thermal spectrum. This is why the forensic calculations were completed with an average HFIR spectra, which most correctly estimated the sample history. The calculated parameters using the average HFIR flux spectrum estimated the sample to have a burnup of 4.42 GWd/tHM, a scalar flux of 1.62_10^15 n/cm^2s, 45 days of irradiation, an initial uranium enrichment of 0.265 wt.% ^235U, and 356 days between removal from reactor and the analysis date. The initial scalar flux for the sample was estimated with MCNP as 1.73_10^15 n/cm2s, there were 50 days of irradiation, the initial uranium enrichment was 0.26 wt.% ^235U, and there were actually 355 days between removal from the reactor and the analysis date. The 6% difference in scalar flux estimates is likely due to the changing flux spectra during irradiation.
