Development of neutron kinetic code for molten salt reactors

Kun Zhuang; Liangzhi Cao; Youqi Zheng; Hongchun Wu

doi:10.13832/j.jnpe.2014.S2.0183

Development of neutron kinetic code for molten salt reactors

Kun Zhuang
, Liangzhi Cao
, Youqi Zheng
, Hongchun Wu

School of Energy and Power Engineering

Xi'an Jiaotong University

Research output: Contribution to journal › Article › peer-review

1 Scopus citations

Abstract

This study establishes the suitable dynamic models for molten salt reactors considering the effects of fuel flow on the distribution of delayed neutron precursors and then develops a new code named MOREL. Some MSRE experimental data from Oak Ridge National Laboratory (ORNL) are chosen to verify the code, especially the DNP model, and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactors.

Original language	English
Pages (from-to)	183-185
Number of pages	3
Journal	Hedongli Gongcheng/Nuclear Power Engineering
Volume	35
DOIs	https://doi.org/10.13832/j.jnpe.2014.S2.0183
State	Published - 15 Dec 2014

Keywords

Molten salt reactor
Neutron kinetics
Reactivity loss

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10.13832/j.jnpe.2014.S2.0183

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abstract = "This study establishes the suitable dynamic models for molten salt reactors considering the effects of fuel flow on the distribution of delayed neutron precursors and then develops a new code named MOREL. Some MSRE experimental data from Oak Ridge National Laboratory (ORNL) are chosen to verify the code, especially the DNP model, and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactors.",

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T1 - Development of neutron kinetic code for molten salt reactors

AU - Zhuang, Kun

AU - Cao, Liangzhi

AU - Zheng, Youqi

AU - Wu, Hongchun

PY - 2014/12/15

Y1 - 2014/12/15

N2 - This study establishes the suitable dynamic models for molten salt reactors considering the effects of fuel flow on the distribution of delayed neutron precursors and then develops a new code named MOREL. Some MSRE experimental data from Oak Ridge National Laboratory (ORNL) are chosen to verify the code, especially the DNP model, and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactors.

AB - This study establishes the suitable dynamic models for molten salt reactors considering the effects of fuel flow on the distribution of delayed neutron precursors and then develops a new code named MOREL. Some MSRE experimental data from Oak Ridge National Laboratory (ORNL) are chosen to verify the code, especially the DNP model, and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactors.

KW - Molten salt reactor

KW - Neutron kinetics

KW - Reactivity loss

UR - https://www.scopus.com/pages/publications/84930519456

U2 - 10.13832/j.jnpe.2014.S2.0183

DO - 10.13832/j.jnpe.2014.S2.0183

M3 - 文章

AN - SCOPUS:84930519456

SN - 0258-0926

VL - 35

SP - 183

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JO - Hedongli Gongcheng/Nuclear Power Engineering

JF - Hedongli Gongcheng/Nuclear Power Engineering

ER -

Development of neutron kinetic code for molten salt reactors

Abstract

Keywords

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