压水堆辐照后燃料中子源强计算方法研究

Research on Neutron Source Intensity Calculation Method for Pressurized Water Reactor Burned Fuel

压水堆辐照后燃料中子源强在次临界状态下的堆芯反应性测量中具有重要作用。本文研究了压水堆辐照后燃料自发裂变源强和(α,n)源强的计算方法,提出了^(242)Cm近似法和比例系数拟合法两种(α,n)源强计算方法。基于自主开发核设计程序系统,开发了堆内辐照后燃料中子源强计算模块,结合微观燃耗模型可以精确考虑对辐照后燃料中子源有重要影响的反应堆空间效应和实际运行历史效应。燃料组件测试算例结果表明,辐照后燃料总中子源强最大相对偏差约5%。本文工作为次临界状态下堆芯反应性测量技术的研发奠定了基础。

The accurate calculation of pressurized water reactor burned fuel neutron source intensity plays a significant role in the subcritical reactivity measurement.The calculation methods of both spontaneous fission source and(α,n)neutron source were researched in this paper.There were multiple sources of alpha particle in reactor core,including^(238)Pu,^(239)Pu,^(240)Pu,^(241)Am,^(242)Cm,^(244)Cm,etc.Among these sources,the dominant source was coming from^(242)Cm decay.^(242)Cm approximation method as well as ratio fitting method were proposed for obtaining(α,n)neutron source intensity.In^(242)Cm approximation method,all alpha particle sources were neglected with the exception of^(242)Cm.The ORIGEN code was used to analyze the ratio of(α,n)to the spontaneous fission neutron of^(242)Cm for different enrichment assemblies at different shutdown time.It is found that the ratio of(α,n)to the spontaneous fission neutron of^(242)Cm is fixed around 0.191.Therefore,the neutron number density of spontaneous fission neutron is multiplied by the factor of 0.191 to obtain(α,n)neutron source intensity due to^(242)Cm decay.On account of neglecting all(α,n)neutron source intensity with the exception of^(242)Cm,it is foreseeable that^(242)Cm approximation method would underestimate(α,n)neutron source intensity.In ratio fitting method,the ratio of(α,n)to the total neutron source intensity was analyzed.It is found that the ratio of(α,n)to the total neutron source intensity is linearly related to the burunp when the burnup is greater than 20000 MW·d/tU.Therefore,two fitting methods were proposed to obtain the ratio of(α,n)to the total neutron source intensity.The polynomial fitting method was used with burnup less than 20000 MW·d/tU,on the other hand,the liner fitting method was adopted with burnup greater than 20000 MW·d/tU.Based on the in-house core design code package,the burned fuel neutron source calculation module was developed.In order to account for reactor core space effect and burnup history effect,the micro-depletion correction method in core design code package was used to obtain accurate number densities of actinide nuclides.The developed module was verified by calculating fuel assembly examples.Comparing with the reference code,the maximum errors of total source intensity is about 5%provided by both^(242)Cm approximation method and ratio fitting method.With the developed burned fuel neutron source intensity calculation module,both radial and axial distributions of burned fuel neutron source intensity in the reactor core are obtained.These distributions are of significant importance to ex-core detector signal measurements.The work in this paper is an important technology support for subcritical rod worth measurement development.