Volume 108, Issue 9

The 50 Cr(n,2n) 49 Cr, 52 Cr(n,2n) 51 Cr, 52 Cr(n,p) 52 V, 55 Mn(n,2n) 54 Mn, 55 Mn(n,α) 52 V and 55 Mn(n,γ) 56 Mn reaction cross sections at the neutron energy of 14.54 ± 0.24 MeV were measured by using activation method along with off-line γ-ray spectrometry. The 27 Al(n,p) 27 Mg monitor reaction was used for the 52 Cr(n,p) 52 V and 55 Mn(n,α) 52 V reactions, whereas the 197 Au(n,γ) 198 Au monitor reaction for the 55 Mn(n,γ) 56 Mn reaction. The 27 Al(n,α) 24 Na monitor reaction was used for other three reactions. The neutron beam was generated from the T(d,n) 4 He reaction using the Purnima neutron generator at BARC, Mumbai. The covariance analysis for the uncertainties of reaction cross sections was performed by considering the correlations between different attributes. The present results were compared with the literature data, evaluated data of ENDF, ROSFOND, CENDL, JENDL and JEFF libraries as well as with the theoretical values based on TALYS-1.9 calculation.

The simultaneous separation of Am and Cm from lanthanides is important for atomic energy fields. However, the process is difficult owing to the chemical behavior of trivalent metal ions with similar ionic radii. All lanthanides, Am, and Cm can be extracted by diglycolamide (DGA). In addition, relatively high separation factors between An and Ln were obtained by the extraction system of TODGA, DTPA (diethylenetriamine- N , N , N ′, N ″, N ″-pentaacetic acid) and HNO 3 . In this work, DTPA-BA (diethylenetriamine- N , N ′, N ″-triacetic- N , N ″-bisamide), which is an improved version of DTPA, was employed for the separation of Ln and An. After performing a basic study on DTPA-BA, a relatively high separation factor (approximately 8) for actinides/lanthanides was obtained. Then, the multi-step extraction was performed. Thus, the recoveries of 94.7 % for Nd and 4.7 % for Am and Cm in organic phase, and 5.3 % Nd and 95.3 % for Am and Cm in aqueous phase were obtained.

With the aim of evaluating the corrosion behavior of a Pu evaporator made from Zr in the Rokkasho Reprocessing Plant (RRP), the electrochemical characterization of Zr was investigated in Pu nitrate solutions. Before and after experiments, spectra of Pu nitrate solutions were recorded, and Pu 4+ was oxidized to PuO 2 2+ by heating these solutions. The ratio of PuO 2 2+ in 3 mol dm −3 HNO 3 was higher than that in 7 mol dm −3 HNO 3 in our experiments. The maximum open circuit potential of Zr in the Pu nitrate solution was approximately 1 V in the Pu nitrate solution containing 7 mol dm −3 HNO 3 . However, there were no significant changes at high PuO 2 2+ concentrations, and this value remained below the passivity breakdown potential under our experimental conditions. In this study, Zr showed high corrosion resistance, even though many PuO 2 2+ ions existed in the Pu nitrate solution containing 7 mol dm −3 HNO 3 .

Separation of neptunium by solvent extraction has been based on tributylphosphate (TBP) for decades, but TBP is not fully incinerable, which adds to the burden of long-lived radioactive waste. Alternatives to TBP for uranium and plutonium extraction, such as the N , N -diakylamides, previously have been explored in the hopes of transitioning to an extractant that is incinerable. Four N , N -diakylamides, N , N -dihexylhexanamide (DHHA), N , N -dihexyloctanamide (DHOA), N , N -di(2-ethylhexyl)butanamide (DEHBA), and N , N -di(2-ethylhexyl)-iso-butanamide (DEHiBA) were considered in this work for their potential to extract millimolar concentrations of Np(IV), Np(V), and Np(VI) from nitric acid solutions into organic solutions containing 1 M extractant in Exxsol D60. Under these conditions the branching of the alkyl substituents affects the extractability of Np(VI) and Np(IV), causing three of the dialkylamides, DHHA, DHOA and DEHBA, to extract neptunium in the expected order Np(VI) > Np(IV) > > Np(V). In contrast, branched DEHiBA is so poor an extractant for Np(IV) that the extraction order becomes Np(VI) > > Np(V) > Np(IV) between 0.1 and 5.6 M HNO 3 due to partial oxidation of the Np(V) in nitric acid.

In this study, the hierarchical mesoporous carbon (HMC) was synthesized by the hydrothermal method. The batch adsorption experiments showed that HMC exhibited the ultrafast equilibrium fate (80 % U(VI) capture efficiency within 5 min), high UO 2 2+ capture capacity (210 mg/g, pH = 4.5) and well recyclability. The investigations of XPS techniques indicated the oxygen-containing functional groups were responsible for high efficient UO 2 2+ adsorption. The pH-dependent adsorption was simulated by three surface complexation modellings, revealing that UO 2 2+ adsorption on HMC was excellently fitted by triple layer model using two inner-sphere complexes (i. e. SOUO 2 + and SOUO 2 (CO 3 ) 3 5− species) compared to constant capacitance model and diffuse layer model. These findings are crucial for expanding actual applications of HMC towards the removal of radionuclides under environmental cleanup.

The expanded perlite and base activated perlite (Zeolite-NaA) results from treating perlite has a SiO 2 :Al 2 O 3 ratio ~2 were used as an adsorbents for the removal of 144 Ce and 152+154 Eu. The two adsorbents were fully characterized. The factors affecting the sorption of radionuclides onto expanded and modified perlite including contact time, pH, ion concentration were investigated. The modification process of expanded perlite enhanced the removal of 152+154 Eu from 17.0 to 97.0 %, while for 144 Ce from 22.0 to 91.0 %. Both 144 Ce and 152+154 Eu obey Langmiur isotherm model in case of using modified perlite (Zeolite-NaA) and expanded perlite.

In 1957 Glenn T. Seaborg conceived and advocated for the construction of the High Flux Isotope Reactor (HFIR) and the Transuranium Processing Plant (since then renamed the Radiochemical Engineering Development Center, or REDC) at Oak Ridge National Laboratory. Heavily shielded hot cells, glove boxes, and laboratories allow recovery of transuranium elements produced in substantial quantities. Seaborg’s vision of HFIR and REDC producing milligram quantities of berkelium, californium, and einsteinium has been fulfilled beginning in 1966 through May 2019 with 78 production campaigns yielding a cumulative totals of 1.2 g of 249 Bk, 10.2 g of 252 Cf, 39 mg of 253 Es, and 15 pg of 257 Fm. Notably, 252 Cf is a neutron source used in many industrial applications including oil exploration; process control systems for the cement industry, coal analysis, and power production; sources to start nuclear reactors and perform nondestructive materials analyses; homeland security and national defense detection devices; and medical research. Isotopes made available through transplutonium production at HFIR/REDC have enabled scientists to study the nuclear properties and reactions, chemical properties, optical properties, and solid-state properties of transplutonium elements. Long-lived isotopes have served as targets in heavy ion accelerators to produce heavier elements leading to the discovery of 104 Rf, 105 Db, 106 Sg, 113 Nh, 114 Fl, 115 Mc, 116 Lv, 117 Ts, and 118 Og. This paper reviews the evolution of the processing flowsheets to produce, separate, and purify transplutonium isotopes, which have evolved over 50 years of operation at HFIR and REDC, and summarizes directions of future work to improve the efficiency of the production operations.

In view of several significant discrepancies in the excitation function of the 86 Sr(p,n) 86g+xm Y reaction which is the method of choice for the production of the non-standard positron emitter 86 Y for theranostic application, we carried out a careful measurement of the cross sections of this reaction from its threshold up to 16.2 MeV at Forschungszentrum Jülich (FZJ) and from 14.3 to 24.5 MeV at LBNL. Thin samples of 96.4% enriched 86 SrCO 3 were prepared by sedimentation and, after irradiation with protons in a stacked-form, the induced radioactivity was measured by high-resolution γ -ray spectrometry. The projectile flux was determined by using the monitor reactions nat Cu(p,xn) 62,63,65 Zn and nat Ti(p,x) 48 V, and the calculated proton energy for each sample was verified by considering the ratios of two reaction products of different thresholds. The experimental cross section data obtained agreed well with the results of a nuclear model calculation based on the code TALYS. From the cross section data, the integral yield of 86 Y was calculated. Over the optimum production energy range E p  = 14 → 7 MeV the yield of 86 Y amounts to 291 MBq/μA for 1 h irradiation time. This value is appreciably lower than the previous literature values calculated from measured and evaluated excitation functions. It is, however, more compatible with the experimental yields of 86 Y obtained in clinical scale production runs. The levels of the isotopic impurities 87m Y, 87g Y, and 88 Y were also estimated and found to be <2% in sum.