V. N. Zubarev

Impurity Composition and Content in Graphite from Commercial Reactors

The impurity activation products produced during prolonged neutron irradiation make the main contribution to the contamination of spent reactor graphite. Data on the content of impurity elements in reactor graphite are needed to predict its radioactivity. The results of the determination of the content of 16 impurity elements in graphite from commercial and RBMK reactors by the neutron–activation method based on the IRT at the Moscow Engineering-Physics Institute are presented and compared with the results obtained previously and by other methods.

Experimental determination of the spent graphite radioactive contamination at plutonium-production reactors of the Siberian Group of Chemical Enterprises (TOMSK-7)

Abstract Spent graphite from decommissioned plutonium-production uranium-graphite reactors is contaminated with radionuclides, and this graphite represents an important fraction of the radioactive wastes accumulated by the Russian nuclear power industry. To select proper ways and dates for the management of graphite, the information regarding the composition and level of the graphite contamination is required. In the paper, results are presented that were obtained in studies carried out at the I-1, EI-2, and ADE-3 reactors of the Siberian Group of Chemical Enterprises (Russia) in 1996-1999. The main feature of the studies is a wide-scale sampling from the graphite piles of the aforementioned reactors followed by complex assays of their radioactive contamination. The analyses performed for the large number of graphite samples made it possible to obtain a detailed picture of the pile contamination, to study radionuclide distributions over the piles, to construct schemes for evaluation of radionuclide stockpiles, and to evaluate stockpiles of several radionuclides including 14C, 3H, 90Sr, 241Am, 244Cm, 238,239,240,241Pu, 137,134Cs, and 60Co.

Experimental Study of the Radioactive Contamination of Graphite Masonry in the Commercial Reactors at the Siberian Chemical Combine

The graphite masonry from decommissioned commercial uranium–graphite reactors is contaminated with radionuclides and is a significant fraction of the radioactive wastes from the nuclear industry. Information about the composition and degree of contamination of the graphite is needed to choose methods and time periods for handling this graphite. The results of investigations performed on I-1, ÉI-2, and ADÉ-3 reactors at the Siberian Chemical Combine in 1996–2001 are presented in this paper. Analysis of a large number of samples made it possible to construct a detailed picture of the contamination of the masonry, study the distribution of the radionuclides of different origin in the masonry, construct schemes for making estimates, and estimate the content of certain radionuclides, including 14C, 3H, 90Sr, 241Am, 244Cm, 238–241Pu, 137,134Cs, and 60Co.

Possibility of salvaging spent graphite sleeves from the reactors of the Siberian chemical combine by incineration

Several thousand metric tons of spent graphite sleeves, which are being kept at solid-wastes storage facilities, have now been accumulated. The construction and state of most of the storages sites do not meet modern requirements because of the long time that these sites have been in operation and because the depths reach ground water, if the required protective barriers are absent. Consequently, there is a real danger that the radionuclides will contaminate the ground waters and the surrounding environment. To prevent this from happening, a system of measures must be instituted to put into place the required safety barriers or the sites must be liquidated.The results of an investigation of the radioactive contamination of spent graphite sleeves from the Siberian Chemical Combine are reported. These results show that the quite low contamination level of the sleeves permits salvaging them by incineration. In so doing, the risk of negative public-health effects in the region of the combine due to 14C emissions into the atmosphere will be neglible.

Possible Application of γ-Ray Spectrometers Based on CdZnTe Detectors

New semiconductor γ-ray detectors based on CdZnTe have been developed in the last few years, and they are now being widely used. Their advantages are small size and possibility of operating without cooling. In the present work, these detectors are used to measure the radiation from spent fuel assemblies in holding ponds and dry storage sites, unirradiated nuclear materials, and radioactive wastes. The results are analyzed and compared with other types of detectors. The possible applications of CdZnTe-type detectors are determined.

γ-Spectrometric Monitoring of Large Plutonium Samples

The γ-ray spectra of the samples of spontaneously fissioning 252Cf in Pu with different isotopic composition were measured. Peaks belonging to definite fission products were found in the spectrum. It was concluded that the mass and isotopic composition of large plutonium samples can be monitored on the basis of the results of only γ spectrometric measurements.

Fission Products and Actinides in Spent Graphite Masonry of the Reactors at the Siberian Chemical Combine

The characteristic feature of the investigations performed was the wide-scale and representative selection of samples of graphite masonry from reactors and the comprehensive analysis of their radioactive contamination. An analysis of a large number of samples made it possible to construct a detailed picture of the contamination. The role of incidents in the formation of the contamination of graphite by individual radionuclides was determined. The distribution of the radionuclides in the masonry was studied. Correlations between the content of different radionuclides were investigated.Schemes for estimating the store of radionuclides in graphite masonry were constructed on the basis of the distribution of the radionuclides of different origin. In the approximate schemes the masonry was represented as a collection of cells. The store of a radionuclide was determined by summing its content in individual cells, which were divided into several categories according to their degree of contamination.The estimates of the store of radionuclides in the masonry differed strongly from the predicted values. The store of fission products and actinides was approximately ten times smaller than previously assumed; this could substantially simplify the disassembly and choice of utilization technology. 2 figures, 10 tables, 3 references.

Gamma-Spectroscopy Methodology for Simultaneous Determination of Mass and Isotopic Composition of Large Plutonium Samples

Abstract This paper presents a nondestructive methodology for the determination of mass and isotopic composition of large plutonium samples. The methodology is based on experimental data from measurements of gamma radiation emitted by plutonium samples under assay. Both plutonium mass and plutonium isotopic composition are derived from one experimental gamma spectrum in two energy ranges. Gamma spectrum in the middle-energy range is used for the determination of plutonium isotopic composition from experimental intensities of photo peaks belonging to different plutonium isotopes. Gamma spectrum in the high-energy range contains photo peaks of spontaneous fission products, and these data can be used for the determination of plutonium mass. The calibration curve for dependence of the count rates in photo peaks of spontaneous fission products on effective 240Pu mass was plotted based on experimental data for the enterprise-level reference plutonium samples. When processing experimental data, some corrections were introduced to account for self-absorption of gamma radiation in the plutonium samples and for neutron-induced fission reactions. The correction factors were calculated with the application of Monte Carlo methodology. The final relative errors in the determination of plutonium dioxide mass were within the range of (4 to 10)% (1σ) for nuclear material containers with different cooling times and different isotopic compositions of plutonium.

A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

An active neutron method for measuring the residual mass of 235U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual 235U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of 238U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

Activation γ-Spectrometric Method of Determining the Mass of Samples of Nuclear Materials

The possibility of using activation γ-spectrometry to determine the mass content of nuclear materials in matter is investigated. Irradiation of samples for a short time with moderated neutrons from a ~107 sec–1 Pu–Be source is used to induce 1436 keV γ-ray emission from 138Cs. These γ-rays are suitable for measurements; the mass of the nuclear materials is determined from the intensity of the radiation. Three series of experiments are performed with sets of samples consisting of uranium and uranium dioxide with different mass and degrees of enrichment.Experiments showed that the error in determining the mass of uranium samples can reach 1–3% with 30–60 min irradiation and the same measurement duration.Special experiments were performed to investigate the influence of the experimental geometry and the self-absorption of the γ rays in the sample, which limit the possibility of γ-spectrometric measurements on samples of nuclear materials.The activation γ-spectrometric method can be used for analyzing metallic uranium samples, powder samples, samples of fuel micropellets and uranium hexafluoride, and plutonium samples.