Calculation of neutron generation time with prompt neutron density decay method
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摘要: 采用蒙特卡罗程序MCNP计算了西安脉冲堆中子代时间。使用MCNP程序模拟了反应堆瞬发中子通量密度衰减，基于忽略缓发中子项的点堆动力学方程计算出中子代时间。在微次临界下，研究了次临界度、源的分布、计数区域等对西安脉冲堆中子代时间计算结果的影响。计算分析表明：采用瞬发中子密度衰减法计算中子代时间时，微次临界度、源分布、计数区域等对计算结果影响都很小；误差产生的主要原因是忽略缓发中子项的点堆动力学方程并不能较好地反应瞬发中子通量密度的衰减规律。Abstract: The neutron flux density plays a very important role in reactor kinetics. The Monte Carlo code MCNP is used to calculate the neutron generation time of Xian Pulsed Reactor. First the decay of neutron flux density is simulated with MCNP and then the neutron generation time is calculated based on the point-reactor kinetics equation. In the slightly subcritical reactor, the effect of subcriticality, count region and distribution of the neutron source on the results are studied. The calculation indicates that the subcriticality, count region and distribution of the neutron source affect the neutron generation time very little. The dominant error is that the decay of prompt neutron density cannot be precisely described with the point-reactor kinetics equation.
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Key words:
- kinetic parameter /
- neutron generation time /
- Monte Carlo /
- prompt neutron density decay
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