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在通过测定~(137)Cs,~(144)Ce,~(148)Nd等裂变产物监测体浓度推算辐照燃料燃耗的方法中,需要裂变产物的平均裂变产额、(n,r)俘获反应的修正量、放射性裂变产物的堆内衰变修正量、可裂变核素的平均裂变能量等参数。这些参数是同燃料的辐照历史密切相关的。本文介绍一种计算这些参数的方法、计算机程序概况和计算结果。本方法有如下特点:1.采用燃耗物理计算获得的可裂变核素核密度及裂变截面作为本程序的输入数据。2.采用燃耗值的初始实验结果反推燃料辐照期间的中子通量。3.精确计算了~(137)Cs和~(148)Nd两种监测体(n—1)衰变链和n衰变链中俘获反应的修正量。从而提高了各种参数的精确度。对于浅燃耗天然铀辐照燃料的应用例,计算结果表明,~(137)Cs,~(144)Ce,~(148)Nd获得燃耗结果的修正量分别为+0.29%,+16.40%,-2.75%。本方法对燃耗结果可能引入的误差分别为±0.1%,±0.3%,±0.6%。
The average fission yield of fission products, (n, r), fission yield, fission yield, fission yield and fission yield are needed to estimate the burnup of irradiated fuel by measuring the concentration of fission products such as 137C, 144Cu and 148Nd. The amount of correction of capture reaction, the amount of internal decay correction of radioactive fission products, and the average fissile energy of fissile nuclides. These parameters are closely related to the irradiation history of the fuel. This article describes a method for calculating these parameters, a computer program overview, and calculation results. The method has the following characteristics: 1. The fissile nuclide nuclear density and fission cross section obtained by the fuel consumption physics calculation are used as the input data of the program. 2. Using the initial experimental results of burnup values to reverse the neutron flux during fuel irradiation. Accurately calculate the corrections of the capture reaction in decay chain and n decay chain of ~ (137) Cs and ~ (148) Nd. Thus improving the accuracy of various parameters. The calculated results show that the corrected results of the burnup results of ~ (137) Cs, ~ (144) Ce and ~ (148) Nd are +0.29% and + 16.40% respectively for the application of shallow uranium irradiated fuel. , -2.75%. The errors that this method may introduce into the fuel consumption results are respectively ± 0.1% ± 0.3% ± 0.6%.