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Properties of a Prototype Corium of Nuclear Reactor 时间:2018-09-19 23:13   来源:未知   作者:admin   点击: 次   Abstract:The paper studies structure, phase composition, and thermophysical properties (TPP) (specific heat capacity, thermal diffusivity, and heat conductivity) of a prototype corium of a fast nuclear reactor (melt of core materials of nuclear reactor produced under out-of-pile conditions). The obtained data will be used to get more accurate understanding of main regularities of actual interaction of core materials of a nuclear reactor under a severe accident.
1. Introduction
  In order to solve issues for studying processes occurring under in severe accidents in nuclear reactor and to research properties of generated alloys and compounds, the most efficient instrument is physical simulation, which simulates real processes at the laboratory conditions and enables to obtain data that can be applied for solving of actual problems.
As a rule, model physical objects are usually being produced in accordance with a certain similarity to full-scale ones.
At the practice, methods of physical simulation are used for
(i) studying gas fluxes and airflow of flying machines, vehicles, and others in wind tunnels;
(ii) hydrodynamical studying of hydraulic structures, ships, and others;
(iii) determining seismic resistance of structures at their mock-ups;
(iv) studying resistance of complex structures under effect of complex power loads;
(v) measuring thermal fluxes and heat dissipations in devices and systems operating under conditions of complex power loads;
(vi) studying of natural phenomena and their consequences.
   Experimental works on physical modeling of severe accidents are conducted in many research centers over the world [1]. For example, the studies of accidents within the following programs are PHEBUS [2], CORA [3], CODEX [4], KROTOS [5], «RASPLAV» [6], MACE [7], FARO/TERMOS [8], FRAG [9], Betulla [10], CORRECT [11], and VITI [12]. Most of the experiments are aimed at obtaining knowledge on the processes occurring at various stages of severe accidents [13, 14].
  Within the above studies, the Institute of Atomic Energy Branch of the National Nuclear Center of the Republic of Kazakhstan conducts both in-pile [15–17] and out-of-pile [18, 19] experiments to study behavior of energy nuclear reactor fuel under the conditions of severe accident with loss of coolant (LOCA).
The purpose of this work is experimental research of properties and structure of molten core materials of a nuclear reactor produced by induction heating at VCG-135 test bench.
2. The Research Materials and Methods
A corium ingot resulted from out-of-pile experiment through induction heating of materials in a graphite crucible with tantalum carbide protecting insert at VCG-135 test bench and was selected as research material.
The VCG-135 test bench is designed for high-temperature material studies into the processes of interaction of reactor core components when they are heated as high as melting temperature.
The VCG-135 test bench is created based on a VCG-60/0.066 high frequency electric generator and a hermetically sealed, water-cooled process chamber with an inbuilt inductor, and it is designed for high-temperature and thermophysical and materials testing on small materials samples. The bench is illustrated in Figure 1. The bench facilitates the controlled heating of any small samples to high temperature (3,000°C) with subsequent cooling using heat lakes to the water-cooled inductor when the generator is powered down [20].