逆动态法实验数据符合处理方法研究

逆动态方法是测定反应性的重要方法。在空间效应不显著、堆芯有外源（或较强的分布源）和功率较低情况下，用两参数符合处理逆动态法实验数据，可同时得到反应性和有效源强。利用两参数符合得到有效源强，可逐点计算与每一时刻相对应的瞬时反应性值，了解反应性瞬时变化情况。     

基于逆动态法的ADS次临界度在线监测方法数值验证

加速器驱动次临界系统(ADS)的次临界度在线监测是ADS运行和安全的核心问题,目前次临界度的测量方法主要有:外推-周期法,跳源法,脉冲中子源法等。本文研究了基于逆动态法的ADS次临界度在线测量方法,并对该法进行初步数值模拟验证。基于临界堆反应性逆动态测量方法,增加外源项的特殊考虑:采用次临界稳态中子通量密度(或功率)及初始次临界度以确定外源,实现适用于次临界堆反应性计算的逆动态求解算法。本文使用欧洲小型加速器驱动的次临界系统PDS-XADS进行数值验证,与动力学程序NTC-2D的计算结果进行对比。结果表明:该方法可有效实现次临界堆的次临界度在线监测。     

固定棒位法测量控制棒总价值

控制棒价值测量的准确度与效率对核电厂的安全性与经济性具有重要影响。在动态刻棒等反应性测量工作中,本底与中子源对探测器有显著影响,致使根据实测电流计算得到的反应性显著偏离真实值。基于点堆逆动态方程,通过对本底与中子源影响的分析,利用固定棒位状态下的测量数据计算反应性并得到控制棒总价值,给出了一种不受本底与中子源影响的简便的控制棒总价值测量计算方法,并在零功率实验装置上进行验证。结果表明,该方法可有效避免本底和中子源组件对反应性探测的影响,并简化了离线理论计算,其与周期法计算结果的相对偏差在1%以内。     

脉冲堆动态引发过程中初始反应性的测量

提出了脉冲堆动态引发过程中初始反应性测量的当量外推方法,由逆动态反应性测量系统、控制棒定位监测系统和裂变脉冲波形测量系统共同获得的数据而得到系统的反应性。此方法在裂变脉冲动态引发实验中成功地得到了应用,获得了初始反应性及其分布。     

快中子临界装置逆动态法反应性测量系统研制与应用

逆动态法可实时、准确地测量反应性,在反应性部件刻度、扰动反应性效应测量、动力堆的瞬态分析等领域具有独特的优点。本文基于逆动态法研制了快中子临界装置反应性测量系统,并对其进行了算法验证与指标分析。实验结果表明,该系统具有测量精度高、响应快、分辨率高、使用方便等优点,满足实验需求。     

跳源法在启明星1#次临界装置中的应用

简要介绍了跳源法在启明星1#次临界装置上测量次临界度的原理、外源驱动的次临界中子学实验装置、堆芯布置及中子源驱动系统。主要研究了中子源在堆芯轴向中心位置、不同装载情况下的反应性变化,并给出不同的有效倍增系数keff。实验测量结果与理论计算结果进行了比较,结果符合较好。     

周期法测量启明星Ⅱ号散裂靶材料的反应性价值

为验证理论计算程序及相关核数据,将固体散裂靶材料放入ADS启明星Ⅱ号零功率装置（启明星Ⅱ号）的靶区内,采用周期法测量靶区内有、无散裂靶材料的反应性,从而获得净散裂靶材料对应的反应性价值,并与理论计算结果进行比较。结果表明,反应性价值的实验测量结果与理论计算结果符合较好,验证了理论计算的正确性。经实验验证的理论计算程序和核数据可用于ADS次临界反应堆的设计。     

反应性测量中本底与中子源修正计算方法研究

在核电厂反应堆的反应性测量中,γ本底电流、中子源引入的电流、电路噪声等干扰项在实测电流中占有显著份额,给反应性的测量计算带来很大影响,在动态刻棒等反应性测量工作中必须对其进行修正。本文分析了本底与中子源电流对反应性测量的影响,并自主开发了一种本底电流、中子源电流修正方法--双重拟合法。该方法基于点堆模型,通过对包含本底与中子源的逆动态方程的分析,计算得到准确的反应性以及本底电流、中子源电流。通过试验堆动态刻棒实验的数据对本方法进行了验证,得到了理想的结果。     

脉冲堆零功率物理实验

本文介绍了脉冲堆堆外零功率物理实验装置,给出了净堆临界实验、反应性、中子通量分布和动态参数等的测量结果。为验证脉冲堆物理计算方法提供了一套较完整的实验数据,并为大堆安全运行提供了必不可少的运行参数。     

便携式反应性测量系统设计

基于逆动态法和周期法,开发了一套适用于钍基熔盐堆(TMSR)物理启动实验的便携式反应性测量系统。测量系统在保留了原电流信号测量功能的同时,加强了对脉冲信号的处理能力。对该系统进行了堆上实验验证,结果表明:系统能够正确处理脉冲及电流信号,并得到反应性。在加入预测平滑算法后,系统能够实时,并较为准确地给出反应堆的功率倍周期。     

Neutron source strength determination for on-line reactivity measurements

A method is described to determine the effective neutron source strength in a nuclear reactor, which must be known when calculating the time-varying reactivity from inverse reactor kinetics for a reactor at low power. When for an initially subcritical reactor the reactivity is changed and kept constant after the change, the effective source strength can be determined from a linear regression of reactor power to a function proportional to the emission rate of delayed neutrons, which can be calculated from the reactor power history. In view of the relatively strong noise present in the reactor power signal at low power, a grouping method for the regression is preferred over the least-squares method.

Experiments with a reactor simulator with known source strength showed good agreement. Application to actual reactor signals gave consistent and satisfactory results.     

Subcriticality Measurement by Neutron Source Multiplication Method with a Fundamental Mode Extraction

A new neutron multiplication method has been proposed for an accurate measurement of subcriticality. The proposed method consists of two correction processes for (1) extraction of the fundamental mode from measuring data of a neutron detector that contains higher modes as well as the fundamental mode feeding from an external neutron source and (2) spatial corrections for perturbations induced by a reactivity addition in the distributions of the fundamental mode and a neutron importance field. Feasibility of the proposed method has been verified from a numerical study, although under some limitations such that the neutron multiplying system to be analyzed is small-sized and a reactivity change takes place homogeneously in a fuel loaded region. With extraction of the fundamental mode and the spatial corrections, the subcriticality can be estimated accurately even with measuring data highly contaminated with higher modes due to a detector position near to an external point neutron source. For a future application to measurement of control rod bank worth of a nuclear power plant from measuring data during a reactor physical testing, some useful guidelines have been obtained.     

基于OpenMC的瞬发中子衰减常数计算模块开发与验证

瞬发中子基波衰减常数α可定量描述反应堆内中子随时间的变化,是计算绝对反应性所需的中子动力学参数之一,对次临界（特别是较深次临界）绝对反应性的精确测量具有重要意义。本文在开源程序OpenMC基础上,基于k α迭代方法,以中子径迹长度上的平均时间吸收权重修正作为k α迭代参数因子,在输运过程中对瞬发、缓发中子分别考虑,开发了具有瞬发α本征值问题计算功能的OpenMC PA模块。以Godiva衍生基准题和MUSE 4次临界实验装置为计算对象,对程序计算瞬发α本征值问题能力进行验证。结果表明,该计算模块有优于MCNP4C的计算速度与计算范围,计算值与参考值的相对误差小于05%。OpenMC PA能满足次临界系统瞬发α本征值和中子动力学参数计算需求。     

An impact of inherent neutron source on subcriticality measurement in a highly enriched uranium core of Kyoto University Critical Assembly

In a variety of highly enriched uranium cores of Kyoto University Critical Assembly, many different subcriticality measurements have been strenuously performed. However, any influence of neutron source inherent in the highly enriched uranium fuels on these measurements has hardly been studied. This is because the influence has been expected to be negligible in the fuels. In this study, we revaluated the influence on pulsed neutron, accelerator-beam trip and rod drop measurements to reveal an unexpected impact of the weak inherent source. Especially, the inherent source was injurious to most of the beam trip and the rod drop measurements based on the integral count method. The least-squares inverse kinetics analysis also had a significant influence on the inherent source. In the area ratio analysis for a pulsed neutron measurement, a considerable number of neutrons from the inherent source was mixed into delayed-neutron area. When the influence was considered in these data analyses, the subcritical reactivity of the above measurements was in good agreement with that calculated by the continuous-energy Monte Carlo code MVP.     

ADS次临界系统中子时空动力学计算与瞬态分析

加速器驱动次临界反应堆(ADS)中子时空动力学计算需要考虑外中子源和空间分布的影响,比临界系统中子动力学计算要复杂得多。本文将改进准静态(IQS)近似与蒙特卡罗(MC)方法相结合,对于带外源的ADS次临界系统中子时空动力学过程,形状函数、动力学参数由MCNPX程序计算得到,幅度函数与集总参数热工反馈模型进行耦合计算,并开发了IQS/MC计算程序可视化操作界面。针对CIADS靶堆耦合系统参考方案物理模型,对引入束流瞬变及无保护失流工况过程进行瞬态模拟计算分析,给出了堆芯相对功率、燃料温度及冷却剂出口温度随时间的变化曲线。同时,将中子注量率进行分群计算,得到了堆芯分能群的相对中子注量率网格分布随时间的变化,模拟结果与理论分析一致。     

Design & construction of a digital real time reactivity meter for Tehran research reactor

Reactivity measurement is one of the challenges of monitoring, control and investigation of nuclear reactors. In this paper design and construction of a reactivity meter for continuous monitoring of reactivity in research reactors are described. The device receives amplified output of the fission chamber, which is in mA range, as the input. Using amplifier circuits, this current is converted to voltage and then digitalized with a microcontroller to be sent to serial port of computer. The device itself consists of software, which is a MATLAB real time programming for the computation of reactivity by the solution of neutron kinetic equations. After data processing the reactivity is calculated and presented using LCD. Tehran research reactor is selected to test the reactivity meter device. The results of applying this reactivity meter in Tehran research reactor (TRR) are compared with the experimental data of control rod worth, void coefficient of reactivity and reactivity changes during approach to full power. The maximum relative error in several experiments is calculated to be 13%.     

Determination of Large Negative Reactivity by Integral Versions of Various Experimental Methods

The large negative reactivity is measured in Semi-Homogeneous Experimental facility (SHE). Experimental methods are Sjöstrand's pulsed neutron, source multiplication and rod drop methods beside revised King-Simmons' pulsed neutron methods. Neutron detectors are placed at various points in the core region for multi-points measurement.

Usual one-point reactor model analysis resulted in the reactivity values, strongly dependent on the detector position with the increase of subcriticality. In addition, disagreements between the used experimental methods are also pointed out.

In order to overcome these difficulties due to the spatial higher harmonics and the kinetic distortion in the neutron flux distribution, an integral version analysis is applied, in which use is made of multi-points reactor model. In the analysis, space integration of the neutron counts obtained throughout the core region is made with weights of the adjoint function of fast neutrons, calculated using the two- or three-dimensional diffusion code. The negative reactivity values determined by the integral version analysis agreed well with each other within the uncertainty of ～5% in the reactivity range down to ～50 dollars.

It is concluded that all the experimental methods are adequate for precise determination of the large negative reactivity of reactor provided that the integral version analysis is utilized or that correction is made for the change of the neutron generation time using precise calculation.     

Prompt fission neutron multiplicity and spectrum evaluations for n+238U reaction

The prompt fission neutron multiplicity and spectra for n+²³⁸U reaction are calculated using an improved Los Alamos model which includes the linear relation between the average prompt gamma ray energy and the prompt neutron multiplicity and also the average fission fragment kinetic energy dependence on the incident neutron energy. The coefficients describing the quadratic variation of the fission fragment kinetic energy versus the incident energy are obtained by extrapolation of the data and procedure used for n+²³⁵U reaction. The inverse process compound nucleus cross-section of the fissioning nucleus is calculated using the coupled channel method. In the incident energy range where only the first fission chance is involved the comparison of present spectrum evaluation with spectrum calculation using multi-modal model is made too. The calculated prompt neutron multiplicity and spectra of ²³⁸U neutron induced fission are in good agreement with the experimental data for the entire incident energy range required in evaluations, proving the validity of the used procedure.     

Neutronic design of a sub-critical, source driven reactor at the RA-8 facility

We present a LEU-ADS design based on an existing Argentine experimental facility, the RA-8 pool type zero power reactor. The versatility of this reactor allows measurement of different core configurations using different fuel enrichment, burnable poison rods, water perturbations, different control rods types in critical or subcritical configurations with an external source.To assess the feasibility of the LEU-ADS, multiplication factors, kinetic parameters, spectra, and time flux evolution were computed. Two external sources were considered: an isotopic source, and a D-D pulsed neutron source.Parameters for different core configurations were calculated, and the feasibility of using continuous and pulsed neutron sources was verified.