A. V. Bushuev

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.

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.

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.

Possibility of Activation γ-Spectrometric Measurements for Determining the Masses of Nuclear Materials in Samples

Measurements of γ-ray spectra performed on uranium and plutonium samples irradiated with 252Cf source neutrons are described. Peaks due to 138Cs and other short-lived fission products are observed in the spectra. The intensity of the peaks characterizes the content of fissioning isotopes in the samples.