Volume 90, Issue 7 -

The reaction yields of 63 radionuclides with the mass number A = 56 − 135 produced in the photofission of 209 Bi by bremsstrahlung of end-point energies ( E 0 ) from 450 to 1100 MeV have been measured using a catcher foil technique with the aid of intensive chemical separation. The charge distribution was well reproduced by a Gaussian function with the most probable charge ( Z p ) expressed by a linear function of A , i.e. , Z p = RA + S , and with the A -independent full width at half maximum (FWHM CD ). The charge distribution parameters R , S and FWHM CD were independent of E 0 above 600 MeV, reflecting the resonance nature in photonuclear reactions at intermediate energies. The weighted mean values at E 0 ≥ 600 MeV were R = 0.421 ± 0.001, S = 0.6 ± 0.1 and FWHM CD = 2.1 ± 0.1 charge unit (c.u.). The numbers of pre- and post-fission neutrons were deduced to be ν pre = 12 ± 1 and ν post = 1.4 ± 0.3, respectively, assuming the Unchanged Charge Distribution (UCD). Based on the charge distribution parameters, the symmetric mass yield distributions with the most probable mass A p of 96 ± 1 mass unit (m.u.) and the width FWHM MD of 33 ± 1 m.u. were also obtained. The characteristics of the charge and mass yield distributions are discussed by referring to those for 197 Au based on the results of calculations using the Photon-induced Intranuclear Cascade Analysis code combined with the Generalized Evaporation Model (PICA3/GEM).

The extraction behavior of uranium(VI) from nitric acid solutions has been investigated using mixtures of bis(2,4,4-trimethylpentyl)phosphinic acid (HBTMPP) and trialkylphosphine oxide (Cyanex 923 = TRPO), triphenylphosphine oxide (TPhPO) or tributylphosphate (TBP). The results demonstrate that uranium(VI) is extracted into xylene as UO 2 (BTMPP) 2 with HBTMPP alone and as UO 2 (BTMPP) 2 ·S (where S = TRPO, TPhPO or TBP) in the presence of neutral oxo-donors. The extraction equilibrium constants of the above extracted complexes have been deduced by non-linear regression analysis with the aid of suitable chemically based model developed taking into account aqueous phase complexation of metal ion with inorganic ligands and all plausible complexes extracted into the organic phase. The addition of a neutral oxo-donor to the metal chelate system significantly enhances the extraction efficiency of uranium(VI). Complexation strength of uranium(VI) with neutral oxo-donors follows the order: TRPO > TPhPO > TBP, which is also the basicity sequence of these ligands. The IR spectral studies of the extracted complexes were used to further clarify the nature of the extracted complexes.

An experimental procedure has been developed for the analysis of neptunium in input solutions of spent nuclear fuel reprocessing. The method consists of selective chromatographic extraction of neptunium and plutonium using TEVA resin followed by determination of the ratio of [Np]/[Pu] on the column by high-resolution gamma spectrometry. Knowing the concentration of plutonium from routinely performed K -edge densitometry and X-ray fluorescence measurements, the concentration of Np can be evaluated. The method has been tested on simulated solutions and validated on spent fuel of known [Np]/[Pu] ratio.

The emission of uranium and ruthenium oxides from overheated nuclear fuel is a complex phenomenon which depends on fuel matrix, temperature, oxygen potential of the atmosphere in contact and on emission sequence. Furthermore, secondary annealing (“maturation”) reactions of aerosol particles in the gas phase, immediately after emission, strongly influence the properties of particles before their dispersion in the reactor containment building and/or in the environment. Maturation influences surface and bulk gradient composition, particle size, and, as a consequence, weathering rates in the environment. This paper describes how emission and maturation conditions may affect the solubility rates of condensation particles released by overheated U and Ru oxides mixtures, in small-scale laboratory simulations. It will be demonstrated that the presence of Ru decreases the solubilisation rate of U.

Radiolysis of water is a phenomenon which alters the prevailing conditions in the nearfield of a final geological repository for high-level nuclear waste (HLW), because the nominally anoxic conditions in the repository may change due to the production of oxidants close to the solid-liquid interface of the fuel, caused by the alpha radiolysis of water. The influence of water chemistry and oxidant concentration in the liquid phase was tested by experiments which simulate chemical radiolysis effects. Leaching experiments with solutions of de-ionized water (DI) and natural groundwater (GW) containing the water radiolysis product H 2 O 2 in various concentrations were performed contacting pellets of depleted UO 2 as a solid phase. After the experiment, U concentration was measured and the solid surfaces was examined with scanning electron microscopy and electron-dispersive X-ray analysis (SEM-EDX). Uranium concentrations showed an unexpected behaviour; at low H 2 O 2 concentrations in solution, U concentrations were highest. U concentrations in the groundwater leachates were always lower than in pure water. The SEM-EDX examination showed the presence of different alteration phases on the sample. The sample surfaces treated in pure water showed optically a yellow discolouration and were covered with a thick layer of oxidation products. In the case of the treatment in groundwater, the samples remained black, and no dense layer was formed. Instead, crystals of octahedral shape were found on the surface, while the matrix grain structure remained intact. All alteration phases were found to contain only U (and possibly O, H, and C also). A reaction mechanism is proposed for the oxidation reactions taking place under the various conditions. Comparisons with natural analogues and previous work are made, and geochemical solubility limits of possible newly formed substances are examined. The results suggest that different polymorphs of uranium(IV) oxyhydroxides (para-, meta-schoepite, hemihydrate) and of uranium(VI) peroxide (Studtite, meta-studtite) have formed, and that in the case of the samples treated with groundwater, a scavenging mechanism, possibly controlled by an ion present in groundwater, leads to the deactivation of H 2 O 2 in the solution.

Several charred plant and charcoal samples from various stratigraphic levels of the Neolithic Site, Çatalhöyük – Turkey, were dated in the AMS facility of Purdue University (PRIME Lab). Radiocarbon dates reveal a complicated chronology, as was foreseen from archeological investigations. Our measurements suggest that this unique Neolithic town may have been initiated at the East mound around 8390 BP.

The radioisotope 90 Nb is a positron emitter with a relatively high positron branching of 51% and a rather low β + -energy of E mean = 662 keV E max = 1.5 MeV) with potential for application in positron emission tomography (PET). Its half-life of 14.6 h makes it especially promising for the quantitative investigation of biological processes with slow distribution kinetics. An advantageous production route of 90 Nb is the nat Zr( p , n ) 90 Nb process. To separate nca (no-carrier-added) radioniobium from macroscopic amounts of irradiated zirconium targets, an effective two step separation procedure was developed. A liquid-liquid extraction using N-benzoyl-N-phenylhydroxylamine (BPHA) in CHCl 3 was applied first to separate the bulk of the zirconium matrix with a Zr separation coefficient of about 10 4 , followed by anion-exchange chromatography for high grade purification of nca radioniobium. A sufficiently high decontamination factor of > 2 × 10 7 and an overall radiochemical yield of nca 90 Nb of about 70% was achieved within about 3 hours.

A novel method for the recovery of 99m TcO 4 − from (n, γ) 99 Mo using conventional anion-exchange chromatography over Dowex-1 × 8 column, followed by elution and subsequent novel purification over AgCl-alumina column is reported. 99 MoO 4 2− solution (7 − 14 GBq/mg) in NaOH was passed down a Dowex-1 × 8 column (6 − 7 × 2 mm), collected and preserved for reuse on the subsequent days. 99m TcO 4 − preferentially retained on the Dowex-1 column, was recovered quantitatively by elution with 4 ml of 4 M HNO 3 or 4 ml 0.2 M HClO 4 or 6 ml of 0.2 M NaI solution. Further purification to deliver 99m TcO 4 − in a physiological vehicle has been achieved. Use of NaI eluent and post elution purification over a bed of 1 g AgCl + 1 g alumina developed has been adapted in preference to the other two eluents. The overall yield of 99m TcO 4 − was 75%−85% and the quality features of 99m TcO 4 − were compatible to radiopharmaceutical use. The method has applicability for decontamination of 99g Tc from spent 99 Mo waste solution and recovery of 99g Tc for research use. The procedure should also be equally applicable for recovery of 188 ReO 4 − from 188 WO 4 2− in a radioisotope laboratory.

Several derivatives of MAG 3 , in which one or both of the first two amide bonds are inverted and contain an additional methyl or carboxyl-substituent on the ligand backbone, have been synthesized and evaluated. This was realized by substituting a glycyl with a D- or L-alanyl moiety or by replacing cysteamine with an L-cysteine to investigate the renal characteristics of the resulting derivatives versus the reference 99m Tc-MAG 3 . Labelling with 99m Tc resulted in two main radioactive species, which probably are the isomeric oxo-technetium complexes. These complexes were separated by HPLC and evaluated in mice. The data suggest that the substitution in the amide functions and addition of a methyl or carboxyl moiety results in poor renal handling of these compounds. However, one of the derivatives 99m Tc- C - L B, exhibits renal characteristics comparable to 99m Tc-MAG 3 .