P. Mendoza, “Experimental Characterization of Pu Separation by PUREX Process on a Low-Burnup, Pseudo-Fast-Neutron Irradiated DUO2 for Product Decontamination Factors and Nuclear Forensics”, Ph.D. Dissertation, Nuclear Engineering, Texas A&M University, College Station, TX (2018).
Recovery by Extraction (PUREX) process were conducted. PUREX process was performed for a depleted UO2 (DUO2) sample irradiated in a fast neutron environment in the High Flux Isotope Reactor (HFIR). The DUO2 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, and roughly 0.3% attributed to fission products. Two cycles of a four extraction, three back-extraction process achieved 93% of Pu recovered in a product solution with less than 1% of the original U. The mathematical connection between DCs and DFs was derived for the above-mentioned experiment and this relationship 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.rnFurther, a forensic methodology was utilized for determining parameters like fuel burnup, scalar neutron flux, 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 were used in later calculations. The above parameters were determined using single group cross sections and fission yields determined by averaging with four different normalized flux spectra. The four flux spectra were: the initial HFIR sample flux spectra, the Fast Breeder Reactor (FBR) blanket region, an AP1000, and a Pressurized Heavy Water Reactor (PHWR). Concentrations determined at the end of PUREX processing 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 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. The calculated parameters using the AP1000 flux spectra estimated the sample to have a burnup of 4390 [MWd/tHM], a scalar flux of 2.55 x 1015 [n/cm2s], 31 days of irradiation, an initial uranium enrichment of 0.29 wt.% 235U , and 385 days between removal from reactor and the date 5/2/2014. The initial scalar flux for the sample was estimated with MCNP as 1.73 x 1015, there were 50 days of irradiation, the initial enrichment was 0.28 wt.% 235U, and there were actually 355 days between removal from the reactor and the date 5/2/2014. The large difference in scalar flux estimates is likely due to the changing flux spectra during irradiation.