金属铀钚核系统的瞬发中子衰减常数测量

采用Rossi-a方法开展了金属铀钚核系统在缓发临界和次临界状态下的瞬发中子衰减常数测量,获得了系统在缓发临界时的瞬发中子衰减常数αc.分析了衰减常数a与反应性的关系,并将其与次临界计数率倒数外推和次临界反应性外推的结果作了分析比较.结果表明,在缓发临界下直接测量的数据为0.835±0.005/μs,数据误差是Rossi-a和外推方法的1/6左右.     

一种测量快堆α_c的方法研究

在超瞬发临界状态下,直接测量脉冲前沿的功率上升,得到瞬发中子增殖常数(α)。在超缓发临界,刻度调节棒的反应性当量,累加调节棒的反应性当量得出爆发脉冲的预加反应性。由超瞬发临界实验数据外推得到了CFBR-II堆缓发临界瞬发中子衰减常数(αc)和反应性定向差。测量得到的αc与Rossi-α方法测量得到的结果一致。     

模拟脉冲源方法计算CFBR-Ⅱ堆缓发临界瞬发中子衰减常数

通过计算国外基准装置缓发临界处的瞬发中子衰减常数,证明了用模拟脉冲源方法计算出的瞬发中子衰减常数是可信的。用该方法计算了CFBR-Ⅱ堆5个次临界状态的瞬发中子衰减常数;线性外推得到缓发临界时系统的瞬发中子衰减常数为0.55μs-1;计算值与试验结果相比存在着偏差。文中简单分析了产生偏差的原因。     

模拟脉冲源方法计算CFBR-Ⅱ堆缓发临界瞬发中子衰减常数

通过计算国外基准装置缓发临界处的瞬发中子衰减常数,证明了用模拟脉冲源方法计算出的瞬发中子衰减常数是可信的.用该方法计算了CFBR-Ⅱ堆5个次临界状态的瞬发中子衰减常数;线性外推得到缓发临界时系统的瞬发中子衰减常数为0.55μs-1;计算值与试验结果相比存在着偏差.文中简单分析了产生偏差的原因.     

内生脉冲中子源方法测量固态零功率堆的某些动态参数

用内生脉冲中子源方法测量固态零功率堆的瞬发中子衰减常数，从而得到不同装载下的反应堆次临界度，还给出了测量瞬发中子衰减常数所用的一些参量。     

Rossi—α方法在微型堆上的应用

利用Ｒｏｓｓｉ－α方法测量不同装载下的瞬发中子衰减常数α，求得装载量与α的数学表达式。利用该数学式和临界时的瞬发中子衰减常数求得的临界质量，与外推方法得到的临界质量相符。该方法适用于临界安全分析和对临界实验结果的校核。     

Rossi－α方法在微型堆上的应用

利用Ｒｏｓｓｉ－α方法测量不同装载下的瞬发中子衰减常数α，求得装载量与α的数学表达式。利用该数学式和临界时的瞬发中子衰减常数求得的临界质量；与外推方法得到的临界质量相符。该方法适用于临界安全分析和对临界实验结果的校核。     

单次脉冲中子源方法测量瞬发中子衰减常数

瞬发中子衰减常数α是核系统的重要标志性特征参数,该值的准确测量对于核临界安全具有重要意义。本工作采用单次脉冲中子源的方法测量了某次临界核系统的瞬发中子衰减常数,获得了几种不同次临界状态下的瞬发中子衰减常数,测量结果与²⁵²Cf随机脉冲源法测量结果一致。     

基于Labview技术的研究堆瞬发中子衰减常数α实验研究

利用反应堆互功率谱密度(CPSD)测试系统对300~#池式研究堆的瞬发中子衰减常数α进行实验研究。堆芯采用低富集度U燃料装载,燃料组件有一定燃耗。测试系统以PCI-6143高速同步数据采集卡(DAQ)为测试硬件,以Labview8.5程序为开发软件;利用紧靠堆芯对称布置的2路γ补偿电离室探测器获得中子在堆内的连续电流信号,该信号输入到测试系统后,再应用数据采集和数据处理分析程序得到互谱密度,最后用非线性最小二乘法拟合得到瞬发中子衰减常数α为(83.5±0.6)s~(-1),结果与实际运行的情况相符合,表明本次带燃耗的瞬发中子衰减常数α测量是成功的。     

Rossi-α方法测量CFBR-Ⅱ堆瞬发中子衰减常数

瞬发中子衰减常数α是核系统的重要特征参数。简要介绍了点堆模型的α本征值定义以及Rossi-α方法原理,并利用研制的Rossi-α测量系统在CFBR-Ⅱ堆上开展了多个状态的α实验测量,获得了系列的次临界α实验数据以及缓发临界时的瞬发中子衰减常数αc,为相关实验提供了基准数据。     

Prompt alpha and reactivity measurements on fast metal assemblies

This paper summarizes and reviews the methods of reactivity determination and measurement of the prompt-neutron decay, briefly describes the equipment requirements for such measurements for unmoderated metal assemblies, and presents experimental results to illustrate the methods. These assemblies include fast metal critical assemblies and fast pulsed reactors. The primary reactivity determination methods used have been: (1) stable reactor period measurements which are usually used near delayed criticality to obtain the reactivity or to calibrate the reactivity prior to burst initiation; (2) prompt reactor period measurements which are useful to determine the reactivity early in superprompt critical excursions; (3) inverse kinetics rod drop measurements which obtain the reactivity as a function of time after a rod or reactor component is removed from a delayed critical assembly, and (4) prompt neutron decay constant measurements from which the reactivity can be obtained if corrections are made for changes in the neutron lifetime. Inverse kinetics and decay constant measurements are usually used below delayed criticality, although decay constant measurements have been performed above delayed critical.The decay constant is usually obtained by the traditional pulsed-neutron method, using a pulsed neutron source such as a Cockcroft-Walton accelerator, or by the Rossi-α method. The recent use of ²⁵²Cf has resulted in some new techniques for determining the decay constant and reactivity; this method has some unique advantages over the traditional methods. The theory of these new measurements is reviewed, and some recent results are presented.     

近临界系统中子倍增时间行为的实验测量

为确定近临界系统内中子倍增因子随时间的变化与系统的反应性的关系,为反应堆启动提供参考,实验测量了中心中子在1个快中子临界装置3个次临界状态的中子倍增的时间行为,在反应性ρ=-2.27×10^-3和-4.79×10^-4的两个次临界状态下,阶跃加入反应性,分别等待180和450s后,中子密度（正比于中子倍增因子）才达到稳定值。与单群缓发中子假设计算的变化曲线比较表明,实验测量的中子倍增因子的变化比理论预估的变化快。即使如此,在反应堆启动中,仍需考虑中子倍增的延迟,以便推算正确的反应性。     

混装堆芯下300#池式反应堆瞬发中子衰减常数α实验研究

利用自行研制的反应堆功率谱密度（PSD）测试系统,对300#池式反应堆在混装堆芯下的瞬发中子衰减常数α进行实验研究。利用紧靠堆芯对称布置的两路电离室DL129探测器获得中子在堆内的连续电流信号,并把该信号输入到测试系统,再应用数据采集和数据处理分析程序得到功率谱密度曲线,最后用非线性最小二乘法拟合得到瞬发中子衰减常数α。经比对,该实验结果与理论计算值在误差范围内符合,满足工程实际需求。     

瞬发和缓发γ射线对堆内构件释热率影响的研究

为提高核电设计中反应堆堆内构件释热率计算的准确性,本文在原来MCNP外中子源模型计算方法的基础上,计算分析瞬发裂变γ对堆内构件释热率的贡献。计算结果显示,考虑瞬发裂变γ使得堆内构件的释热率增加9%～38%,离堆芯越近的堆内构件的增加值越大。另外,分析认为缓发γ对堆内构件释热率的贡献与瞬发裂变γ相当。因而反应堆堆内构件释热率计算中除了考虑中子及中子俘获所生γ的贡献,还应该考虑瞬发裂变γ和缓发γ的贡献。     

Application of Polarity Correlation Method to Graphite Moderated Reactor

The polarity correlation method has been applied to an experiment for measuring the prompt mode neutron decay constant of a reactor with a long neutron lifetime. Measurements have been performed in the SHE VIII-l at critical and several subcritical steady states (0~ — 1$).

The analog signal of detector output is converted to a polarity signal having only one-bit of information on the amplitude of the detected signal. The polarity correlation function is obtained as the correlation function between two polarity signals. Attention has been paid in the design of the correlator to obtain ample stability in operation. Using a digital computer, the prompt mode neutron decay constant is determined with 2% experimental accuracy from the decay curve of the polarity correlation function. A series of the prompt mode neutron decay constants is fitted to an approximated expression of inhour equation to obtain βl. The value of each prompt mode neutron decay constant was found to be in good agreement with that measured by the pulsed neutron technique. The polarity correlation method is particularly applicable to critical or near critical state where a pulsed neutron source cannot be effectively introduced.

In Appendix, an estimation is made for the error in the measured polarity correlation function due to unbalanced mean setting level in the experiment.     

液态熔盐堆运行安全特性初步研究

液态燃料反应堆与固态燃料反应堆相比,原理上有较大不同。液态熔盐堆中由于燃料流动带走缓发中子先驱核在堆外衰变导致堆芯反应性降低,且裂变产物在堆外回路中衰变也会引起一回路发热。本文使用熔盐堆中子动力学程序Cinsf1D探讨2 MW熔盐堆的临界动力学特性和安全特性,研究零功率临界下不同熔盐流速启泵和停泵导致的缓发中子先驱核流失所需改变的控制棒棒位。同时还计算了2 MW恒定功率情况下稳态运行及降低流速时一回路温度分布,并模拟了2 MW额定功率下停泵事件。停泵后由于缓发中子损失减少反应堆功率先缓慢增加,然后迅速降低到接近余热水平。停泵后堆芯温度缓慢增加后稳定在安全值以内,说明熔盐堆具有本征安全性。     

252Cf 随机脉冲源法测量深次临界瞬发中子衰减常数

运用²⁵²Cf 随机脉冲源法时幅变换(TAC)方式测量系统,实验获得高浓缩铀椭球壳核系统的瞬发中子衰减常数α。采用单指数和双指数最小二乘法拟合，α值均为100μs^-1。用蒙特卡罗方法模拟实验过程，α计算值为110μs^-1。结果表明，该系统对深次临界α的测量是有效的。     

Kinetic parameters of a low enriched uranium fuelled material test research reactor at end-of-life

The kinetic parameters at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA’s 10 MW benchmark reactor. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time and effective delayed neutron fraction. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that in comparison with the beginning-of-life values, at end-of-life, the neutron flux increased throughout the core, the prompt neutron generation time increased by 3.68% while the effective delayed neutron fraction decreased by 0.35%.