PRIZMA.D computational investigation of physical processes in the IGR reactor

It is shown for one experiment that the physical processes occurring in the-core of a pulsed uranium-graphite reactor IGR can be investigated by a computational method using the PRIZMA.D program and correlated results in multivariant calculations. 7 figures, 1 table, 7 references. Translated from Atomnaya énergiya, Vol. 88, No. 2, pp. 83–88, February, 2000.     

γ-X-ray-spectrometric investigation of plutonium and241Am content in the soil in regions where peaceful nuclear explosions were conducted

A brief exposition of the γ-x-ray spectrometric method and apparatus for analyzing the content of plutonium and²⁴¹Am in samples of soil and vegetation is presented. The results of an analysis of some samples from the regions where peaceful nuclear explosions were conducted in Yakutiya and Perm oblast' are presented. It is concluded that the γ-x-ray spectrometric method can be used to perform large-scale measurements in regions where peaceful nuclear explosions have been conducted, 2 figures, 3 tables, 8 references. Translated from Atomnaya énergiya, Vol. 88, No. 1, pp. 52–55, January, 2000.     

Materials-technology investigations of fuel elements, irradiated in a reactor at the obninsk nuclear power plant, after standing for 38 years in a depository

The results of materials-technology investigations of a spent fuel assembly from a reactor at the Obninsk nuclear power plant, the first nuclear power plant in the world, before the rated burnup and after prolonged dry storage (for about 40 years) were presented. It was established that the fuel elements from the fuel assembly studied are in satisfactory condition. No appreciable damage due to the prolonged storage was found: the outer diameter remains within the technological tolerance limits and the strength and the plasticity of the jackets are high. Only surface corrosion damage to 10 μm depth was found on the fuel-element jackets. The fuel composition remained whole. 6 figures, 1 table, 3 references. State Science Center of the Russian Federation—A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 88, No. 3, pp. 183–188, March, 2000.     

Estimate of the Hazard Probability in Handling Radiation-Toxic Substances

The dangers in shipping containers holding radioactive materials is analyzed. A method is proposed for estimating the hazard probability in an accident during shipment of radiation-toxic substances. The fact that the shipment includes all operations and conditions associated with the handling of radioactive materials is taken into account (design, fabrication, and servicing of the packing units, preparation, processing, shipment, and storage of radioactive materials). 9 references.     

Models of Ceramic Tritium Breeding Zones in the Blanket of a Thermonuclear Reactor

The development, performed in the 1980s–1990s, of models of tritium breeding zones for blankets of thermonuclear reactors, based on the use of ceramic lithium-containing materials, is described. 5 figures, 1 table.     

Radiation distribution sensing with normal optical fiber

The purpose of this study is to develop a radiation distribution monitor using a normal plastic optical fiber. The monitor has a long operating length and can obtain continuous radiation distributions. A principle of the position sensing is based on a time-of-flight technique. The monitor is sensitive to beta rays or charged particles, gamma rays, and fast neutrons. The spatial resolutions for beta-rays ( ⁹⁰Sr-⁹⁰Y), gamma-rays (¹³⁷Cs), and D-T neutrons are 30, 37, and 13 cm, respectively. The detection efficiencies for the beta-rays, gamma-rays, and D-T neutrons are 0.11%, 1.6×10 ^-5% and 1.2×10^-4%, respectively. The effective attenuation length of the detection efficiency is 18 m. In this paper, we describe the basic characteristics of this monitor     

Taking burnup credit into account in criticality studies: the situation as it is now and the prospect for the future

As enrichment of the fuel has become higher than the limits used at the designing stages, it seemed necessary to consider fuel depletion during irradiation to guarantee the criticality safety for relatively high enriched fuels transportation, storage or reprocessing. This burnup credit will make it possible to use the devices for spent fuels which are initially relatively high enriched. For that purpose, a method was developed considering: (i) partial Uranium-and-Plutonium burnup credit in the criticality studies, and (ii) a conservative assumption concerning the axial profile; this actinides-only method was supported by an experimental program called HTC. The method was accepted by the French Safety Authority. Moreover, in order to reduce again the calculated values of the reactivity for irradiated fuels, a French working group was set up in 1997 to define a conservative method which enables industrial companies to take burnup credit into account with some of the fission products and using a more precise profile. The work of this group has been divided into four tasks related to: the determination of (i) the composition of the fuel, (ii) a conservative profile, (iii) a conservative irradiation history, and (iv) the calculation scheme. This work is also supported by experimental programs related to the validation of the fission products effects, in terms of reactivity.     

Challenges left in the area of in-vessel melt retention

The in-vessel melt retention becomes an important safety objective for the present or future middle power nuclear plants, so care has to be taken in the evaluation of the various phenomena related to ensuring the feasibility of this objective. Since the prediction of the relevant phenomena has to be performed for the prototypical accident conditions, the applicability of the measured data or of the correlations derived from these measurements have to be established and the uncertainties determined. In this context, most uncertainties are introduced by the non-prototypicalities in the experiments. The paper describes the major findings from the OECD RASPLAV project and discusses the remaining challenges left in the area of in-vessel molten corium coolability.