Extraction of Trivalent Rare Earths and Minor Actinides from Nitric Acid with N,N,N’,N’-Tetradodecyldiglycolamide (TDdDGA) by Using Mixer-Settler Extractors in a Hot Cell

Single-stage batch experiments to reveal the extraction properties of N,N,N’,N’-tetradodecyldiglycolamide (TDdDGA) for Y, La, Eu, Nd, and Am in nitric acid were carried out. The distribution ratios of Y, Eu, Nd, and Am exceeded 10 when the nitric acid concentration was 1–2 mol/dm³ (M), and the distribution ratio of La was 5.5 when the nitric acid concentration was 2 M. A continuous counter-current experiment using 0.1 M TDdDGA diluted with n-dodecane was performed using mixer-settler extractors installed in a hot cell. Nitric acid with a concentration of 2.1 M containing minor actinides (MAs: Am and Cm), rare earths (REs: Y, La, Nd, and Eu), and other fission products (Sr, Cs, Zr, Mo, Ru, Rh, and Pd) was fed to the extractor. TDdDGA effectively extracted MAs and REs from the feed, while other fission products were barely extracted. The extracted MAs and REs were back-extracted by bringing them in contact with 0.02 M nitric acid, and they were collected as the MA–RE fraction. The results indicated that more than 98% of Am and Cm in the feed were recovered in the MA–RE fraction. The proportions of Y, La, Nd, and Eu in the MA–RE fraction were 94.0%, 99.9%, 99.9%, and 86.9%, respectively.     

Chromatographic Separation of Trivalent Actinides by Using Tertiary Pyridine Resin with Methanolic Nitric Acid Solutions

Chromatographic separation of trivalent actinides (Am, Cm and Cf) was performed by using a tertiary pyridine resin embedded in silica beads with methanolic nitric acid solutions. The trivalent actinides were eluted from the resin column in the reverse order of atomic numbers (Cf-Cm-Am). Higher concentration of methanol in the mixed solution accelerated both the adsorption of these elements on the resin and the separability for these elements. Americium was clearly separated from Cm and Cf by using a 1 cm-ø × 10 cm-height column with a 60vol% of methanol/40 vol% of concentrated nitric acid mixed solution at ambient temperature.     

Minor actinide transmutation in fast reactor metal fuels irradiated for 120 and 360 equivalent full-power days

An irradiation experiment on uranium–plutonium–zirconium (U–Pu–Zr) alloys containing 5 wt% or less minor actinides (MAs) and rare earths was carried out in the Phénix fast reactor. The isotope compositions of the fuel alloys irradiated for 120 and 360 equivalent full-power days (EFPDs) were chemically analyzed by inductively coupled plasma–mass spectrometry after 3.3–5.3 years of cooling. The results of chemical analysis indicated that the discharged burnups of the fuel alloys irradiated for 120 and 360 EFPDs were 2.1–2.5 and 5.3–6.4 at%, respectively. The changes in the isotopic abundances of plutonium, americium, and curium during the irradiation experiment were assessed to discuss the transmutation performance of MA nuclides added to U–Pu–Zr alloy fuel. Multigroup three-dimensional diffusion and burnup calculations accurately predicted the changes in these isotopic abundances after fuel fabrication. An evaluation of the MA transmutation ratio based on the results of chemical analysis revealed that the quantity of MA elements in the U–19Pu–10Zr–5MA (wt%) alloy decreased by about 20% during the irradiation experiment for 360 EFPDs.     

Evaluation of Neutron Nuclear Data for Z>88 Minor Nuclides

Neutron nuclear data for 15 minor nuclides (Z>88) have been evaluated in the energy range of 10^?5 eV–20 MeV. Since only few experimental data are available, the present evaluation was mainly based on the systematics of the data from neighboring nuclides and also optical and statistical model calculations. The evaluations have been carried out for neutron cross sections of total, elastic scattering, inelastic scattering, (n, 2n), (n, 3n), (n, 4n), fission and capture reactions. In addition, angular and energy distributions of the emitted neutrons and average number of the emitted neutrons per fission were also evaluated. The results were compiled in the ENDF/B-V format and stored in the JENDL-3.     

Critical and Subcritical Mass Calculations of Curium-243 to -247 Based on JENDL-3.2 for Revision of ANSI/ANS-8.15

Critical and subcritical masses were calculated for a sphere of five curium isotopes from ²⁴³Cm to ²⁴⁷Cm in metal and in metal-water mixtures considering three reflector conditions: bare, with a water reflector or a stainless steel reflector. The calculation were made mainly with a combination of a continuous energy Monte Carlo neutron transport calculation code, MCNP, and the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI and JEF-2.2, were also applied to find differences in calculation results of the neutron multiplication factor originated from different nuclear data files. A large dependence on the evaluated nuclear data files was found in the calculation results: more than 10%Δk/k relative differences in the neutron multiplication factor for a homogeneous mixture of ²⁴³Cm metal and water when JENDL-3.2 was replaced with ENDF/B-VI and JEF-2.2, respectively; and a 44% reduction in the critical mass by changing from JENDL-3.2 to ENDF/B-VI for ²⁴⁶Cm metal. The present study supplied basic information to the ANSI/ANS-8.15 Working Group for revision of the standard for nuclear criticality control of special actinide elements. The new or revised values of the subcritical mass limits for curium isotopes accepted by the ANSI/ANS-8.15 Working Group were finally summarized.     

Extraction Behavior of Americium and Curium on Selected Extraction Chromatography Resins from Pure Acidic Matrices

Due to the similar chemical properties between the neighboring trivalent actinide elements americium and curium, their extraction behavior is often perceived as indistinguishable. In this work, the characterization of seven extraction chromatography resins (TEVA, TRU, DGA(N), Actinide, Ln, Ln2, and Ln3) for these trivalent actinides from acidic matrices (HNO₃, HCl, and HBr) has provided some evidence to the contrary. In most cases, Am(III) and Cm(III) exhibit identical extraction properties. However, separation is possible with TRU and DGA(N) resins as demonstrated in this study. The extraction shows a strong dependency on the specific anion in solution that follows the order NO₃^?>Br^?>Cl^?.     

Self-Irradiation Damage of a Curium-Doped Titanate Ceramic Containing Sodium-Rich High-Level Nuclear Waste

The polyphase titanate ceramic containing sodium-rich simulated high-level nuclear waste was doped with 0.69 wt% of ²⁴⁴Cm to accelerate long-term self-irradiation due to α decays. α autoradiography showed that α emissions were almost uniformaly distributed throughout the curium-doped samples on a >20-μm scale although micropore surfaces and titanium oxide agglomerates were free of α-emitting nuclides. The phase assemblage of the curium-doped titanate ceramic included freudenbergite and loveringite in addition to the more abundant oxide phases: hollandite, perovskite, and zirconolite. Accumulation of α decays was accompanied by a gradual decrease in density. The increment of density was – 1% after an equivalent age of 5000 yr. Leach tests showed a slight trend toward higher total release of curium with equivalent age. The release of soluble nonradioactive elements (e.g., Na, Cs, Sr, and Ca) in the oldest specimens (equivalent age, 2000 yr) varied from specimen to specimen but, on average, were higher than specimens that had suffered a lower radiation dose.     

Vaiidition of ORIGEN Computer Code by Measurements on Nuclear Fuels of JPDR-1

The depletion and production amounts of U, Pu, transplutonium nuclides and fission products (FPs) measured on the fuel of JPDR-1 were corrected to take account of the performance history of irradiation and cooling using the results of three-dimensional nuclear-thermo-hydrodynamic and nuclide depletion and production calculations. Except a few nuclides, the corrected values proved to agree well with the values calculated by the ORIGEN computer code. Further enhancement of calculational accuracy calls for systematic re-evaluation of neutron cross sections on the basis of neutron spectrum in nuclear fuel.     

Process Development and Laboratory-Scale Demonstration of a Regular-SANEX Process Using C5-BPP (2,6-Bis(5-(2,2-dimethylpropyl)-1H-pyrazol-3-yl)pyridine)

An optimized solvent comprising 0.01 mol/L C5-BPP + 0.5 mol/L 2-bromo-hexanoic acid in TPH with 10% 1-octanol was developed for the selective extraction of trivalent actinides from lanthanides. Equilibrium extraction data and kinetics of the system in extraction, scrubbing and stripping modes were evaluated and a 16-stage flow-sheet was successfully tested in a spiked centrifugal contactor demonstration. It turned out that 16 stages were insufficient for a complete recovery of An(III), although a clean An(III) product was obtained. The high selectivity of C5-BPP for An(III) was shown. The results of the spiked centrifugal contactor test are presented and discussed.     

Analysis of Curium in Mixed Oxide Fuel Irradiated in the Experimental Fast Reactor JOYO for the Evaluation of Its Transmutation Behavior

Curium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Curium was isolated from the irradiated MOX fuel by anion-ex- change chromatography using a mixed medium of nitric acid and methanol. The isotopic ratio of curium and its content were determined by thermal ionization mass spectroscopy and alpha-spectrometry, respectively. The curium content was less than 0.004 at% even at high burnup of 120GWd/t, which is much smaller than that of PWR-MOX at 60 GWd/t. On the basis of present analytical results, the transmutation behavior of curium isotopes in a fast reactor was discussed from various viewpoints. Transmutation rates of curium isotopes were estimated; the rate for ²⁴⁶Cm, which is known to be a key nuclide in the transmutation of curium, was larger than the previously reported value. It was concluded from these evaluations that the fast reactor was suitable for the incineration of curium.