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.     

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.     

Development of a multigrid-type microstrip gas chamber

Conventional microstrip gas chambers (MSGCs) have encountered many difficulties, such as limited gas gain and sparking damages. We propose a new multigrid-type MSGC (M-MSGC) to overcome some of these difficulties. Additional grid strips are inserted between the anode and the cathode in this new type of MSGC. Gaps between these strips are chosen to be as small as 10 μm where one can expect an efficient removal of the surface charge. With the existence of other strips with lower potentials than the anode, the field strength around the neighboring grid to the anode strip is not as high as the conventional small-gap MSGCs. The contribution of the surface streamer to the damage is greatly suppressed because the electric field parallel to the surface is screened by the intermediate grid electrodes. However, additional electrodes also screen all the electric field of the upper part of the substrate, and we cannot observe induced signals from the backside of the substrate. To overcome that difficulty, we propose another signal readout method using a patterning approach. Floating pads are placed close to the cathode strip on the surface of the M-MSGC, and the induced charges are read out via the pads. If the area of the pads is sufficiently large and the positive charges are moving toward the pads, the backside electrodes can sense the induced charge. Collected charges on the pads are leaked through the surface resistance. The backside signal through 2.3-mm-thick glass readout of the position along the cathode strips is successfully confirmed through experimental results     

Elements of comparison between different inert matrix fuels as regards plutonium utilisation and safety coefficients

The current international trend is to focus towards the utilization of plutonium. The use of composite fuels in inert matrix (U-free) is a potentially efficient solution to this problem. This document deals with the cermet fuels, selected for their excellent behaviour under irradiation and their high thermal conductivity. The emphasis is placed on the study of kinetic coefficients. Comparisons are performed with other solutions that use other composite fuels, especially the Solid Solutions and ROXs. As core control requires a heterogeneous assembly, an assembly whose characteristics are compared to the APA reference is proposed.     

Design and fabrication aspects of a plutonium-incineration experiment using inert matrices in a “Once-Through-Then-Out” mode

In the plutonium incineration experiment, named ‘Once-Through-Then-Out’ (OTTO), that is being prepared by JAERI, PSI and NRG, the use of highly stable inert matrices will be examined. The inert matrices MgAl₂O₄ spinel and ZrO₂ are insoluble in nitric acid and are considered as good storage media for final disposal. These inert matrices will be used in this experiment, which is representative for an OTTO scenario. A total of 7 Pu-containing targets were prepared for an irradiation in the High Flux Reactor in Petten. The objective of the irradiation is to reach a very high Pu-burnup. The main parameters to be studied are stability under irradiation, swelling, fission gas release and chemical interactions in the fuel. Four targets will be equipped with thermocouples for on-line monitoring of central temperature. Four of the targets contain MgAl₂O₄ as an inert matrix, 2 targets contain ZrO₂ and one target contains mixed-oxide (MOX) fuel for reference purposes. The fissile plutonium concentration is 0.32–0.44 g cm⁻³. Both particle-dispersed fuel and homogeneous dispersions were fabricated in order to test the effect of the size of the fissile inclusions. The design of the experiment and the fabrication of the samples are discussed.