Determination of 232U and 236Pu Weight Contents of Spent Nuclear Fuel

Radiochemistry - A procedure has been developed for the determination of 232U and 236Pu content by alpha spectrometry by ion-exchange isolation and purification of uranium and plutonium fractions...     

Determining the Specific Activity of Tritium in Samples of Spent Nuclear Fuel

Radiochemistry - Procedure for determining the specific activity of tritium in samples of spent nuclear fuel (SNF) was developed and methodologically certified. The procedure was used to determine...     

AnSn Coordination Polyhedra (An = Th,U, Np,Pu, Am,Cm, or Cf) in Crystal Structures

Radiochemistry - Voronoi–Dirichlet polyhedra (VPD) were used to perform a crystal-chemical analysis of 216 sulfides containing in their crystal structures 296 coordination polyhedra AnSn (An...     

A Precision Spectrophotometric Method for Determining U,Pu, Nd,and Rh,Based on the Internal Standardization Principle

The principle of spectrophotometric determination of the concentrations of elements based on internal standardization of the sample analyzed is presented. As the function of the parameter to be determined, namely, the concentration of element in solution, we suggest the ratio of two signals, one of which is the light intensity at the wavelength of an absorption band of an element being analyzed and the other is the reference signal of a sample, namely, the light intensity at a wavelength at which the solution of the element being analyzed is virtually transparent. The analysis is performed with a specially designed two-channel spectrophotometer in which fairly strict correlation between the signals is provided by measuring their intensities virtually simultaneously, with the same detector and amplification system. The signals are isolated by a monochromator and interference light filters. The spectrophotometer operation control and data processing are effected by a PC. The metrological characteristics of the method proposed were experimentally determined. The relative root-mean-square deviation of the uranium, plutonium, neodymium, and rhodium concentrations in pure solutions of their salts was within 0.1-0.25%.     

Coordination of Pa,Pu, Am,Cm, Bk,and Cf in Oxygen-containing Compounds in Terms of the Concept of the Voronoi-Dirichlet Polyhedra

The features of the environment of actinide atoms (Pa, Pu, Am, Cm, Bk, Cf) in the crystal structure of their oxygen-containing compounds were analyzed in terms of Voronoi-Dirichlet (VD) polyhedra. The volume of the VD polyhedra of actinide atoms surrounded by the oxygen atoms is determined by their valence state and is practically independent of their coordination number. In going to higher oxidation states, the volume of the VD polyhedra of actinide atoms regularly decreases.     

Determination of the Gravimetric Content of Neptunium in Spent Nuclear Fuel Samples

Radiochemistry - The reduction of penta- and hexavalent Np ions to the tetravalent state with Fe(II) sulfamate in HNO3 solutions was studied. The dependence of the degree of the Np reduction on the...     

Features of Determining the Isotope Composition of Carbon in Gaseous,Liquid, and Solid Media

Measurement Techniques - The main issues and recent achievements in the field of metrological assurance of carbon isotope-ratio analysis are considered. An overview of the existing methods of...     

Standard Reference Data on the Nuclear Characteristics of the Alpha-Radiating Radionuclides 226Ra, 233U, 238Pu,and 239Pu

Estimated values of the half-life, the energy and absolute emission probability of alpha-particles for radionuclides, forming part of the standard spectrometric alpha-radiation sources: ²²⁶Ra with daughter decay products ²³³U, ²³⁸Pu, and ²³⁹Pu, are presented as a draft of standard reference data. The recommended values are obtained by analyzing and selecting published estimated and experimental data (up to December 2000).     

A Laser System for Real-Time Measurement of the 129I and NO2 Concentrations in a Spent Nuclear Fuel Reprocessing Medium

A laser system was developed for real-time detection of the global radionuclide ¹²⁹I and NO₂ during spent nuclear fuel reprocessing. This system was used for determining the concentrations of these substances in a gas stream under actual conditions of nitric acid dissolution of spent nuclear fuel at the Khlopin Radium Institute, Research and Production Association. The tests showed that the laser system is suitable for measuring both the ¹²⁹I and NO₂ concentrations directly in the gas phase during spent nuclear fuel reprocessing (this is essential for monitoring this process and controlling the dissolution cycle safety) and in the gases emitted by radiochemical enterprises. The sensitivity with respect to ¹²⁹I, achieved with the laser system developed, also makes it suitable for real-time monitoring of ¹²⁹I and NO₂ at the maximum permissible concentration (MAC) level. This will allow evaluation of the environmental condition of air in the working areas of radiochemical enterprises, as well as in residential areas.     

Dissolution of U,Np, Pu,and Am Oxides and Their Mixtures in Tributyl Phosphate Saturated with HNO3

It was shown for the first time that weighable amounts of uranium dioxide and its mixtures with Np, Pu, and Am in the form of solid solutions can be efficiently and quantitatively dissolved in the presence of tributyl phosphate saturated with HNO₃. Individual PuO₂ and NpO₂ are not dissolved under these conditions. In treatment of a mechanical mixture of UO₂ with PuO₂ or NpO₂, uranium is completely dissolved, while plutonium and neptunium remain in the precipitate.     

Analysis of Pu, Cm and Am alpha spectra using the unfolding code Colégram without prior use of a nuclear data library

This paper shows the unfolding capabilities of a code developed in our laboratory when applied to alpha spectrometry. The code, called ‘Colégram’, uses the Levenberg–Marquardt algorithm. Colégram has some particularly interesting features when compared to other codes using χ² as the convergence criterion. Low statistics peaks are treated more successfully. No nuclear data library is needed. The emission probabilities obtained with Colégram from ²³⁹Pu, ²⁴⁴Cm and ²⁴³Am alpha spectra characterized by high statistics and good energy resolution are presented. These emission probabilities and the alpha-particle energies deduced from them are compared to those obtained in the most recent works. We show that Colégram also offers the possibility of identifying and quantifying contaminants present in the spectra. Finally, with Colégram, it is necessary to identify electron conversion lines and include them in the unfolding process to increase the quality of the fits.     

Association of 90xSr, 137Cs, 239,240Pu, 241Am,and 244Cm with Soil Absorbing Complex in Soils Typical of the Vicinity of the Chernobyl NPP

Association of the main long-lived radionuclides ¹³⁷Cs, ⁹⁰Sr, ^239,240Pu, ²⁴¹Am, and ²⁴⁴Cm with various components of the soil absorbing complexes from soil samples collected along the western, northwestern, and northern tracks of radioactive fallout in the vicinity of the Chernobyl NPP was studied by the sequential leaching. In the samples of the sandy soil collected in the floodplain of the Pripyat river along the northwestern radioactive track, more than 85% of ⁹⁰Sr, 55% of ^239,240Pu, and 75% of ²⁴¹Am and ²⁴⁴Cm are associated with various components of the soil absorbing complex and are potentially mobile species. In the soil samples collected along the narrow western track, 80-85% of ¹³⁷Cs, ⁹⁰Sr, ^239,240Pu, ²⁴¹Am, and ²⁴⁴Cm are incorporated in hot particles. The degree of ¹³⁷Cs, ⁹⁰Sr, ^239,240Pu, ²⁴¹Am, and ²⁴⁴Cm association with different components of the soil absorbing complex is a function of the radionuclide type and physicochemical features of soil.     

Separation of U,Pu, and Am recovered from mixed oxide (MOX) fuel by countercurrent chromatography

The separation of U, Pu, and Am recovered from MOX fuel with the adduct of HNO₃ with N,N′-dimethyl-N,N′-dioctylhexylethoxymalonamide by countercurrent chromatography (CCC) was studied. Solutions of N,N′-dimethyl-N,N′-dibutyldodecylethoxymalonamide in dodecane were used as stationary phase. The separation of U, Pu, and Am was carried in both isocratic and stepwise elution modes. The better separation of actinides and their higher radionuclidic purity are reached with stepwise elution. The first eluate fraction contained only Am (100%). The second eluate fraction contained U (100%) and Pu (0.7%). The third eluate fraction contained 99.3% of Pu.     

Thermodynamic modeling of thermal processes involving actinides (U,Am, Pu) in the course of heating radioactive graphite in steam

The behavior of actinides (U, Am, Pu) in the course of combustion of radioactive graphite in steam was studied by thermodynamic modeling. Thermodynamic modeling was performed using TERRA program in the temperature interval from 373 to 3273 K to determine the possible composition of volatile actinide species formed in the course of graphite utilization by heating in steam. The modeling shows that the actinides at high temperatures are present in the following forms: U, as UO₃ and UO₂ vapors and as UO₃ ^– and UO₂ ⁺ ions; Am, as Am vapor; and Pu, as PuO₂ and PuO vapors and as ionized PuO⁺. The main reactions within separate phases and at the interface were revealed, and their equilibrium constants were determined.     

Determination of 90Sr, 238Pu, 239,240Pu, 241Pu, and 241Am in soils of chernobyl region using various sample preparation techniques

Radiochemical analysis is made of soddy-podzolic sandy and peaty gley soils collected from the Chernobyl zone. Radionuclides were separated by extraction chromatography or ion exchange, and then determined using an α-ray spectrometer and liquid scintillation and proportional gas counters. Leaching from the preliminarily calcined sample with 8 M HNO₃ does not ensure 100% recovery of radionuclides from peaty gley soil.     

Ionic Composition of Hot Particles Containing U,Cs, and Sr,Evaluated from the Parameters of X-ray Photoelectron Spectra of Outer U5f and Inner U4f,Cs3d, 4d,and Sr 3d Electrons

The elemental and ionic composition of pellets produced from reactor fuel (UO₂) containing 0.1 wt % Cs and 0.5 wt % Sr relative to U and also of hot particles generated by heating of the fuel to 2000°C and then subjected to additional heating to 900°C in air or argon and condensed on aluminum support was analyzed by the X-ray photoelectron spectroscopy. It was shown that within the first 20 s of heating U and Cs sublime predominantly. In the subsequent 300 s of heating U, Cs, and Sr sublime. For example, it was found that hot particles collected in the first 20 s of heating and subjected to additional heating at 900°C in air flow contain 68% U and 32% Cs, whereas particles collected in the subsequent 360 s and subjected to the same additional heating contain 51% U, 13% Cs, and 36% Sr. It is assumed that these hot particles incorporate uranyl compounds of the following types: UO₂CO₃, Cs₂UO₄, Cs₄UO₂(CO₃)₃, CsUO₂(OH)₃, SrUO₄, Sr₃UO₆, and SrUO₂CO₃(OH)₂. Treatment of the surface of hot particles with Ar⁺ ions produces changes in their composition.     

Fuel “Hot” Particles of the Chernobyl Accident in the Retrospective Evaluation of the Emergency Processes at the 4th Block of the Chernobyl Nuclear Power Plant

Solid particles fallen out after the Chernobyl accident were studied by electron microscopy and X-ray diffraction analysis. The morphological and physicochemical characteristics of the dispersed fuel particles ejected into the environment and their nature were evaluated for different fallout directions, which supplements the existing version of the active accident stage at the 4th block of the Chernobyl Nuclear Power Plant (CNPP).     

Efficacy of 3,4,3-LI(1,2-HOPO) for decorporation of Pu, Am and U from rats injected intramuscularly with high-fired particles of MOX

This study aimed to assess the efficacy of 3,4,3-LI(1,2-HOPO) for reducing uranium, plutonium and americium in rats after intramuscular injection of (U-Pu)O2 particles (MOX). Sixteen rats were contaminated by intramuscular injection of a 1 mg MOX suspension and then treated daily for 7 d with LIHOPO (30 or 200 micromol kg(-1)) or DTPA (30 micromol kg(-1)). LIHOPO was inefficient for removing Pu, Am and U from the wound site. However, it reduced Pu retention in carcass and liver by factors of 2 and 6 respectively, and Am retention in carcass and liver by factors of 10 and 30. In contrast, the effect of LIHOPO on U was to decrease the retention in kidneys by a factor of 75. These results confirm that LIHOPO is a good candidate for use after contamination with MOX, in combination with localised wound lavage or surgical treatment aimed at removing most of the contaminant at the wound site.     

Specific Features of Nuclear and Thermophysical Calculations of (Zr,U)N Nuclear Fuel Prepared Using an Oxidation-Assisted Engineering Approach

Inorganic Materials - Zr1 – хUхN nitrides have been synthesized via nitridation of solid solutions of uranium in zirconium. High-temperature saturation of the solid solutions with...