Volume 90, Issue 8 -

Experimental studies on the transmutation of some long-lived radioactive waste nuclei, such as 129 I, 237 Np, and 239 Pu, as well as on natural uranium and lanthanum were carried out at the Synchrophastron of the Laboratory for High Energies at the Joint Institute for Nuclear Research (JINR) in Dubna, Russia. The radioactive targets (I, Np and Pu) were contained in weld-sealed aluminium holders produced by the Institute of Physics and Power Engineering, Obninsk, Russia. Spallation neutrons were produced by relativistic protons with energies in the range of 0.5 GeV≤ E p ≤1.5 GeV interacting with 20 cm long uranium or lead target stacks. The metallic targets were surrounded by 6 cm thick paraffin moderators. The uranium and lanthanum samples were positioned on the outside of the moderator surface and typically contained approximately 0.5 to 1.0 gram of uranium or lanthanum. The highest fluence of spallation neutrons was observed in the region of 5 to 10 cm downstream the entrance of the primary beam into the metallic target, rather independent of the target material or the proton energy. The results obtained by nuclear chemistry methods were supplemented by SSNTD (Solid State Nuclear Track Detector) studies. Consistent and systematic results of B -values and spectral distributions for neutrons have been found. From the experimentally observed transmutation rates one can extrapolate that in a subcritical nuclear power assembly (or "energy amplifier") using a 10 mA proton beam of 1 GeV onto a Pb-target as used here, one can transmute within one month in one gram of sample about 3 mg 129 I, 21 mg 237 Np, 3.3 mg 238 U, and 200 mg 239 Pu. Rather similar results have been found by another group for 129 I and 239 Pu. Observations show that the transmutation rates increase almost linearly with the proton energy in the energy interval 0.5 GeV up to 7.4 GeV. These findings are largely confirmed by model calculations using the LAHET- and DCM/CEM-codes.

Employing the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several ( n , 2 n ), ( n , p ) and ( n , α) reactions on isotopes of dysprosium; our measurements constitute the first systematic studies. Of special interest was the investigation of the 156 Dy( n , α) 153 Gd reaction for the production of no-carrier-added 153 Gd in a nuclear reactor. Using 100% enriched 156 Dy target, 8633 KBq of 153 Gd per batch can be produced which is, however, not sufficient for medical application.

Two different methods – leaching and microwave assisted total dissolution – have been exploited for the treatment of environmental samples for the determination of plutonium and strontium. Leaching applied to reference materials demonstrated the procedure to be applicable for the recovery of technogenic Pu and Sr from environmental samples. For the measurement of the alpha emitters of plutonium, co-precipitation with calcium oxalate and ferric hydroxide and separation with anion exchange has been used. For preparation of α-spectrometry sources, co-precipitation with NdF 3 on a membrane filter or electro-deposition using the (NH 4 ) 2 C 2 O 4 /HCl method have been tested. The beta emitter 241 Pu was measured by liquid scintillation counting. Pu isotope concentrations determined in the reference materials agreed well with the certified concentrations. 90 Sr was measured in the leachate solutions from environmental samples collected close to a nuclear facility and from reference materials, after separation from the other leached elements, by liquid scintillation counting and Cherenkov counting. The 90 Sr-concentrations determined in the reference materials agreed well with the certified concentrations. In the samples collected close a nuclear facility (soil, grass and sheep faeces), 90 Sr was found at higher levels, which could also be correlated with the location of the sampling.

N-octanoyl-N-phenylhydroxamic acid (OHA) has been prepared by partial reduction of nitrobenzene with Zn dust/NH 4 Cl, followed by acylation with n-octanoyl chloride. Structure of OHA was confirmed by 1 H NMR, IR spectra and CHNS microanalysis. The extraction of uranium by OHA has been studied as a function of pH, temperature and concentration of OHA. The extraction of uranium was found to increase with increase in pH. Uranium forms a 1:2 complex with OHA. IR Spectra & NMR Spectra of the uranium-OHA complex and slope analysis obtained from extraction studies indicated that the extraction of uranium was due to the formation of a chelate. The enthalpy change accompanied by the extraction of uranium by OHA has been obtained by temperature variation method. The extraction process was found to be endothermic at pH 2 and exothermic at pH 6.

Electrodialysis has been investigated as a method to enhance the transport of U(VI), Th(IV) and lanthanides through Nafion cation exchange membrane I and impregnated with HDEHP-kerosene II. The recovery factor of U(VI) through Nafion cation exchange membrane with and without the applied electric field was 0.8 and 0.14, respectively. The transport process of U(VI) through ion exchange membrane was studied as a function of variation of nitric acid concentration in the feed solution, HDEHP concentration in the membrane stripping solution concentration, pH of the feed solution and effect of electric field. From this study the cationic flux of U(VI) through Nafion I and impregnated with HDEHP-kerosene II was 5.2 × 10 −8 and 8.7 × 10 −8 g eq cm −1 , respectively. The technique of electrodialysis is efficient in preconcentration of U(VI) to 2.8 fold of the initial concentration. The recovery factors of lanthanide elements La(III), Eu(III) and Er(III) through the systems I and II were 0.8 in system I and 0.86, 0.8 and 0.73 in system II, respectively. The recovery factors of La(III), Eu(III), Er(III) and U(VI) through the system II were 0.82, 0.75, 0.71 and 0.27, respectively. The recovery factor of Th(IV) and U(VI) at pH = 1.5 was found to be 0.99 and 0.3, respectively.

At AECL – Chalk River Laboratories, Ontario, Canada solid, low-level radioactive wastes from industrial, academic and medical applications have been stored in trenches above unconsolidated sandy glacial tills and permeable very-fine to fine-grained sands overlying crystalline bedrock. The sandy aquifer system drains into a swamp comprised of approximately 3 m of sphagnum peat. A comprehensive field and analytical program, involving measurements of total iodine, 129 I, tritium, 14 C and 13 C/ 12 C ratios in groundwater and geologic materials (sands and peats), was initiated at this site to examine the partitioning of 127 I and 129 I amongst the various reservoirs in this system and the controlling factors. The maximum iodine concentration and 129 I inside the groundwater contaminant plume at the recharge and discharge sites were 67.0 ng/ml and ∼8.3 × 10 11 atoms/liter, and 32.4 ng/ml and ∼2.9 × 10 11 atoms/liter, respectively, with positive correlations between iodine, 14 C (0.82), and tritium (0.87). Maximum total iodine concentrations for in-plume, recharge-site sands and discharge-site peats were 190.1 µg/kg and 14100 µg/kg, respectively. 129 I analyses on these same samples showed concentrations of 2.3 × 10 7 and 6.4 × 10 9 atoms/g of soil, respectively. K D values (concentration on porous medium/concentration in co-existing water) calculated from the contaminant plume data for 127 I and 129 I were 1.3 and 1.6 l/kg, respectively, at the recharge site and 486 and 93 l/kg, respectively, at the discharge area, indicating that both stable and radio-iodine are preferentially sorbed to the organic rich, aquifer materials at the discharge sites. Incremental leach experiments on these same geologic materials have borne out these differences, with 127 I being more strongly sorbed than 129 I, probably as the result of kinetically controlled sorption mechanisms and the differing residence times of stable and radio-iodine in this hydrologic regime.

Accurate partial gamma-ray production cross sections were determined for the prompt and radioactive product decay gamma rays following cold neutron capture in 99 Tc. They can be used for non-destructive assay of technetium by prompt gamma activation analysis (PGAA) and neutron activation analysis (NAA), offering orders of magnitude higher analytical sensitivities than passive gamma-ray counting. A lower limit of 21.21 ± 0.17 b was also deduced for the thermal-neutron-capture cross section.