Estimation and Mode Selection of Deuterium Flux Supply into Ampoule Device through Diffusion Filter in Experiments with Pb-15.7Li Eutectic at the IVG1.M Reactor

In first generation of fusion energy reactors hydrogen isotopes - deuterium and tritium will be used as a fuel. A reserve of tritium in nature does not exist, therefore, for tritium generation the design of fusion reactors involves a special device called breeder blanket, containing lithium

Methodology of Corrosion Testing of Nuclear and Fusion Reactors’ Materals Using TGA/DSC and MS Complex Techniques

Nowadays, nuclear power is the best that humanity has for the production and supply for cheap electric and thermal energy. The development of nuclear power takes place in a rigid competitive struggle, both with the traditional technologies of electricity production, and with alternative (renewable) sources

Study of Luminescene in Noble Gases and Their Binary Mixtures Excited by the Products of 6Li(n,α)T Nuclear Reaction

Direct conversion of nuclear particles’ energy into optical radiation opens up new opportunities in obtaining a large amount of light energy, including its most perfect form - coherent light [1-2]. Moreover, optical radiation study of a nuclear-excited plasma produced by products of nuclear reactions is of interest for: development of an alternative method of energy output from a nuclear reactor [3]; creation of devices to control and regulation of nuclear reactors’ parameters, creation one of the diagnostics of high-temperature plasma in fusion reactors [4]. Therefore, spectral studies of nuclear-excited plasma are relevant and are of interest for solving problems, associated with gas media selection with high efficiency of nuclear reaction energy conversion into optical radiation

Methodology of the Experiments to Study Lithium CPS Interaction with Deuterium Under Conditions Of Reactor Irradiation

Problems of plasma-facing materials degradation and in-vessel element destructions, tritium accumulation and plasma pollution can be overcome by the use of liquid metals with low atomic number. The best candidate as a material for divertor receiving plates and other in-vessel devices is lithium. Liquid lithium advantages as a plasma facing material have been confirmed by a large number of the experiments at the plasma-physical facilities being operated worldwide

Studies of two-phase lithium ceramics Li4_{4}SiO4_{4}-Li2_{2}TiO3_{3} under conditions of neutron irradiation

This work presents the preliminary experimental data on the study of gas release from two-phase lithium ceramics Li$_{4}$SiO$_{4}$-Li$_{2}$TiO$_{3}$ under neutron irradiation conditions. Experiments were carried out at the WWR-K research reactor (Almaty, Kazakhstan) for ∼4.3 days. The total neutron fluence during the irradiation was ∼1.8·10$^{19}$n/cm$^{2}$. In the course of irradiation, two experiments on ceramics heating during irradiation and two experiments with hydrogen isotopes (H$_{2}$ and D$_{2}$) supply into the experimental chamber with the sample were performed at a temperature of 680 °C and reactor power of 6 MW. During the entire irradiation, the gas composition in the continuously evacuated ampoule device with samples was recorded. The obtained dependences of the release of tritium-containing molecules and helium during the experiment were qualitatively analyzed

Investigation of hydrogen and deuterium impact on the release of tritium from two-phase lithium ceramics under reactor irradiation

In the development of fusion energy, an important task is the study and improvement of tritium production technologies. In this case, one of the most promising materials for tritium generation is lithium ceramics. Considering the importance of the task, numerous studies are aimed at solving the problem of determining the parameters and mechanisms of tritium release in lithium-containing materials. This paper presents the results of a study of tritium release processes from two-phase lithium ceramics of Li$_{4}$SiO$_{4}$/Li$_{2}$TiO$_{3}$ during reactor irradiation when hydrogen and deuterium are injected into the chamber with irradiated samples. The mechanisms regularities of the tritium yield process in the presence of these isotopes were established. The experiments were carried out in the WWR-K research reactor at a neutron flux density of 5∙10$^{13}$n/cm$^{2}$∙s and sample temperatures from 650 to 700 °C

Results of thermal stability tests of the IGR reactor HEU fuel

The use of highly enriched uranium-graphite fuel (HEU fuel) in research reactors is of great concern to the world community due to the danger of nuclear material proliferation suitable for nuclear weapons fabrication. In this regard, large-scale work is currently being carried out at the IAE NNC RK related to the conversion of two unique research reactors IVG.1M and IGR, which should be subjected to a procedure with a decrease in fuel enrichment from 90% to 19.75% U- 235 (LEU fuel). Individual solutions for the nuclear fuel design of batch reactors pose high requirements for the designers of a new low-enriched fuel. Before the IGR reactor is conversed to LEU fuel, it is required a series of tests, one of which is the high temperature stability of fuel under cyclic process of heating-cooling conditions. This paper describes the experimental procedure and thermocyclic test results of a HEU fuel sample. The results of thermocyclic tests of a HEU fuel sample will be a reference when comparing with the results of thermocyclic tests of HEU fuel samples. Therefore, the main criteria for the suitability of using LEU fuel in the IGR reactor will be mass loss of the sample during thermal cycling, which should not exceed the values obtained during the test data with HEU fuel. The tests have been carried out at the TiGrA experimental complex, created on the basis of the TGA/DSC 3+ thermogravimetric analyzer and ThermoStar mass spectrometric gas analysis system. As a result of testing the HEU fuel sample, 100 heating-cooling cycles were carried out in the temperature range from 150°C to 1100 °C with a heating rate of 100 °C/min and cooling rate of 50 °C/min. In this case, the change in the sample mass and the gas phase composition above the sample were recorded

Method of Measuring the Efficiency of the Conversion of Nuclear Energy into Optical Energy

A method of measuring the efficiency of converting nuclear energy into optical energy was developed based on correlations between intensities of the research line and the nitrogen second positive system in an Ar-N 2 mixture. In addition, the values of the coefficient of the conversion of nuclear energy into radiation at the lines of a Hg triplet in mixtures of HA-Hg and Kr-Hg were determined. The values measured correspond to a selectiveness of pumping of 7 3 S 1 that was close to 1 ( = 0.8 ± 0.2)

Method of Measuring the Efficiency of the Conversion of Nuclear Energy into Optical Energy

A method of measuring the efficiency of converting nuclear energy into optical energy was developed based on correlations between intensities of the research line and the nitrogen second positive system in an Ar-N2 mixture. In addition, the values of the coefficient of the conversion of nuclear energy into radiation at the lines of a Hg triplet in mixtures of Хе-Hg and Kr-Hg were determined. The values measured correspond to a selectiveness of pumping of 73S1 that was close to 1 (δ=0.8±0.2)