Mif-skd Computation Code For Thermohydraulic Design Of A Reactor Core Cooled By Supercritical-pressure Water

G. P. Bogoslovskaya; A. A. Karpenko; P. L. Kirillov; A. P. Sorokin

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Mif-skd Computation Code For Thermohydraulic Design Of A Reactor Core Cooled By Supercritical-pressure Water

机译：超临界水冷却反应堆堆芯热工水力设计的Mif-skd计算代码

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Presented here is the MIF-SKD computation code for channel-by-channel thermohydraulic calculations of fuel-rod assemblies cooled by supercritical-pressure water. The code allows one to calculate distributions of temperature and velocity of the coolant in channels of fuel-rod assemblies and temperatures of fuel-rod claddings and fuel-rod assembly casings with regular and irregular geometrical characteristics at various distributions of energy release along the length and the radius of fuel-rod assemblies.

机译：此处显示的是MIF-SKD计算代码，用于逐个通道计算由超临界水冷却的燃料棒组件的热工液压。该代码允许人们计算沿燃料棒组件的通道中冷却剂的温度和速度的分布以及具有规则和不规则几何特性的燃料棒包壳和燃料棒组件外壳的温度，以及沿着长度和长度方向的各种能量释放分布。燃油杆组件的半径。

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《Thermal engineering》 |2009年第3期|214-217|共4页
作者
G. P. Bogoslovskaya; A. A. Karpenko; P. L. Kirillov; A. P. Sorokin;
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State Scientific Centre of the Russian Federation, Institute for Physics and Power Engineering (GNT's RF-FEI), pl. Bondarenko I, Obninsk, Kaluga oblast, 249033 Russia;

State Scientific Centre of the Russian Federation, Institute for Physics and Power Engineering (GNT's RF-FEI), pl. Bondarenko I, Obninsk, Kaluga oblast, 249033 Russia;

State Scientific Centre of the Russian Federation, Institute for Physics and Power Engineering (GNT's RF-FEI), pl. Bondarenko I, Obninsk, Kaluga oblast, 249033 Russia;

State Scientific Centre of the Russian Federation, Institute for Physics and Power Engineering (GNT's RF-FEI), pl. Bondarenko I, Obninsk, Kaluga oblast, 249033 Russia;

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2. CORE DESIGN OF A DIRECT-CYCLE, SUPERCRITICAL-PRESSURE, LIGHT WATER REACTOR WITH DOUBLE TUBE WATER RODS [J] . Okano Y., Oka Y., Koshizuka S. Journal of Nuclear Science and Technology . 1996,第5期

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3. The Computational-and-Experimental Investigation into the Head-Flow Characteristic of the Two-Stage Ejector for the Emergency Core Cooling System of the NPP with a Water-Moderated Water-Cooled Power Reactor [J] . Yu. V. Parfenov Thermal engineering . 2013,第9期

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5. Safety and core design of large liquid-metal cooled fast breeder reactors [D] . Qvist, Staffan Alexander 2013

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6. Fukushima-derived radionuclides in sediments of the Japanese Pacific Ocean coast and various Japanese water samples (seawater tap water and coolant water of Fukushima Daiichi reactor unit 5) [O] . Katsumi Shozugawa, Beate Riebe, Clemens Walther, -1

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机译：REPP：水冷式反应器的热液压设计代码。

Mif-skd Computation Code For Thermohydraulic Design Of A Reactor Core Cooled By Supercritical-pressure Water

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