Thermal properties of prototype corium of fast reactor

被引：13

作者：

Mukhamedov, N. ^{[1
]}

Skakov, M. ^{[2
]}

Deryavko, I. ^{[2
]}

Kukushkin, I. ^{[2
]}

机构：

[1] Shakarim State Univ, 20a Glinky St, Semey 071400, Kazakhstan

[2] Inst Atom Energy NNC RK, 10 Krasnoarmeyskaya St, Kurchatov 071100, Kazakhstan

来源：

NUCLEAR ENGINEERING AND DESIGN | 2017年 / 322卷

关键词：

Prototype corium of a nuclear reactor; Thermophysical properties; Melting crucible; Material carbonization; Uranium dioxide; Stainless steel; NUCLEAR-POWER-PLANTS; SEVERE ACCIDENTS;

D O I：

10.1016/j.nucengdes.2017.06.026

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

The paper is devoted to development and testing of a technology to manufacture the ingot of the prototype corium (resulted from out-of-pile conditions) of fast reactor followed by an experimental determination of the thermophysical properties (TP) (thermal diffusivity a, specific heat capacity C-p, and thermal conductivity lambda) of such corium at the room temperature (298 K). The data on the thermo-physical properties of corium (melt of structural and fuel materials of the reactor core) will be used to calculate the temperature fields in the modeling the processes of keeping corium inside the power reactor vessel under the conditions of a severe accident. (C) 2017 Elsevier B.V. All rights reserved.

引用

页码：27 / 31

页数：5

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