ADS启明星1#次临界反应堆缓发中子有效份额的测量

本文提出了利用改进的源倍增法测量次临界系统的绝对反应性与跳源法测量的相对反应性相比获得缓发中子有效份额β_eff的方法。用改进的源倍增法测量了ADS启明星1#次临界反应堆某次临界状态下的绝对反应性为-2.235×10^-3。在相同的次临界状态下,用跳源法测量了以β_eff为单位的反应性ρ/β_eff为-0.291 5 $,二者相比得到ADS启明星1#次临界反应堆的缓发中子有效份额为0.007667。利用MCNP建模计算的结果为0.007 783,两者在2%内符合。     

一种测量缓发中子有效份额β_(eff)的方法

本工作通过实验与理论计算相结合,给出了测定缓发中子有效份额βeff的新方法。用实验方法确定反应堆临界状态,并测量次临界状态时以βeff为单位的次临界反应性,应用理论程序计算临界时的中子有效增殖因数keff,确定keff的计算偏差,然后理论计算次临界状态下的keff,并用确定keff的计算偏差对次临界状态下计算的keff进行修正,给出次临界状态的反应性。将实验测量结果与理论计算结果相比较,从而给出βeff。这种方法由于是实验确定的反应堆状态,因此,按实验结果计算的keff与理论描述反应堆状态的计算模型关系不大。分析表明,βeff测量结果的精度高于以往测量方法的精度。     

一元线性回归处理逆动态法实验数据

阐述了一元线性回归处理逆动态法实验数据的原理。采用该方法除能获得反应性外还，可得到外中子有效源强。根据逆动态法实验原理和一元线性回归方法编写了计算程序，对俄罗斯动力研究院零功率装置BFS-1的实验数据进行处理，并与俄罗斯的计算结果进行了比较。计算结果表明：利用一元线性回归方法处理逆动态法实验数据可准确获得反应性和有效源强，提高了测量结果的精度，利用计算出的反应性可方便获得中心元件的效率；在反应堆功率较低、外源较强的情况下，应考虑外源对反应性测量的影响。     

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.     

压水堆辐照后燃料中子源强计算方法研究

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²⁴²Cm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。     

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。     

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。     

跳源法测量ADS启明星Ⅱ号的次临界度及动态特性

本文主要利用²⁵²Cf外中子源驱动的ADS启明星Ⅱ号次临界装置来验证理论计算的次临界度及不同次临界度下的断束动态特性。简要介绍了利用跳源法在ADS启明星Ⅱ号上测量次临界度的原理、实验装置、测量系统、堆芯布置及实验结果等。实验通过变化堆芯燃料棒的装载来模拟3个次临界状态,即k_eff分别为0.99、0.98和0.97。实验结果与理论计算结果符合较好,验证了理论计算的正确性。经过实验验证的理论计算程序和核数据,为将来的中国科学院战略性先导科技专项--未来先进核裂变能ADS嬗变系统的次临界反应堆设计提供参考价值。     

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。     

核临界安全中的监督现场测量技术--源倍增方法的某些问题

给出了核临界安全中监督现场的测量技术——源倍增法的实验理论和实验方法。源倍增法实际测量的是有源次临界中子有效增殖系数k2而不是中子有效增殖系数Keff。在铀溶液核临界装置上进行了实验研究用源倍增法测量了次临界系统在外中子源作用下铀溶液不同液位的有源次临界中子有效增殖系数k2;用周期法测量了单位铀溶液位的反应性系数,然后用临界液位与次临界液位之差乘以单位铀溶液位的反应性系数,给出系统次临界液位时的反应性．由反应性给出传统观念上的中子有效增殖系数keff 。讨论了它们的差别及对核临界安全的影响。     

Measurement of the subcriticality of power-stressed reactors with an analog reactimeter

Conclusions A technique has been developed for measuring the effects of reactivity in a subcritical reactor with an analog reactimeter. It is based on the compensation of the current applied to the reactimeter input from the neutron detector. The compensation of current produced by neutrons of the subcritical multiplying assembly formalizes the algorithm for reactivity calculation, making it an adequate model of a reactor with a source and making it possible to determine the subcriticality without prior entry into the critical state. In this case the measurements are made in the presence of neutron sources characteristic of power-stressed reactors. The regular devices of the control and safety system could be used to produce unsteady variation of the neutron flux.All of this permits the proposed method to be extended to zero-power reactors and to power-stressed reactors. Once the subcriticality has been measured an analog of the neutron source is introduced into the reactimeter. This instrument measures the effects of reactivity in the subcritical state without the reactor being previously put into the critical state, monitors the entry of the reactor into the critical state by checking the reactivity, and makes all the measurements usually made with analog reactimeters. If the intensity of the source does not change during measurements (5–10 min) when chambers sensitive to rays (e.g., KNK-56 chambers) are used as neutron-flux detectors, then the accuracy of -ray compensation does not affect the results of the measurements.Translated from Atomnaya Énergiya, Vol. 45, No. 5, pp. 375–376, November, 1978.     

Monitoring the reactivity of extremely subcritical reactors by means of reactivity meters,corrections being made to the analog of the source

Conclusions The advantage of the instrumental methods here described lies in the fact that no additional apparatus is required for determining subcriticality; it is sufficient simply to use an analog reactivity meter allowing the source component in the solving part to be varied. By using these methods we may determine the bias voltage (constituting the analog of a steady-state source in the reactor) in the subcritical state. If the intensity of the source does not alter very sharply over the period of measurement (5–10 min), the introduction of the source function into the instrument is an on-off operation, after which the reactivity meter is able to monitor any changes in K_eff continuously without first bringing the reactor into the critical state. This offers the possibility of making quite accurate measurements to ensure nuclear safety. In cases in which it is required to secure especially accurate and reliable results, these measurements may be repeated as from the critical state as part of a total monitoring operation. The safety of operations requiring passage into the critical state is also increased, since the passage into the critical state is monitored completely.Instrumental methods of measuring subcriticality are universal; they embrace not only critical assemblies but also any energy-stressed reactors in the subcritical state, and enable us to measure temperature effects, effects of poisoning, and so on, i. e., effects characteristic of energy-stressed reactors. In addition to this, these methods may be applied both in the presence and in the absence of a strong neutron background in the reactor.Translated from Atomnaya Énergiya, Vol. 41, No. 4, pp. 238–241, October, 1976.     

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.     

深度次临界刻棒电子学实现方法研究

为实现深度次临界刻棒计算所需数据的有效采集,研究并设计了深度次临界刻棒电子学的总体架构及关键模块,通过堆上试验对关键模块特性进行了测试。结果表明,所设计的深度次临界刻棒电子学能够有效测量经过约200 m电缆传输后的探测器信号,脉冲信号波形宽度稳定,信噪比水平良好;所测得的高压坪特性曲线可以为探测器高压选取提供有效参考;所测得的甄别特性曲线稳定,能有效获取探测器信号中的中子成分。      

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。     

钠冷快堆反应性意外引入事故时停堆保护研究

在反应堆系统中,当反应堆处于异常工况时,如果运行参数超出保护限值,则由保护系统触发相关保护动作,以保证反应堆的状态符合事故验收准则的要求。本文将通过Simulink建立钠冷快堆主要系统模型,在发生反应性意外引入事故时,借鉴快堆事故分析中预期瞬态无停堆保护（ATWS）的分析方法,基于相应保护参数的测量误差和数据处理过程对反应堆一回路的保护参数及其整定值进行研究,并确保钠冷快堆的状态在整个反应性引入事故过程中符合钠冷快堆的事故验收准则。仿真结果表明,当发生补偿棒失控提升5 s和10 s时,目前的堆芯出口钠温、功率、功率流量比等保护参数的整定值、信号测量延迟及落棒时间可取其他值。当补偿棒失控提升15 s时,只要保证保护参数整定值、相应参数的信号测量延迟及落棒时间能使反应堆在36.45 s前进入深度次临界都是可以的。     

Determination of the subcriticality level using the Cf source-detector method

Measurement and monitoring of reactivity in a subcritical state, e.g. during the loading of a power reactor, has a clear safety relevance. The methods currently available for the measurement of k_eff in stationary subcritical conditions should be improved as they refer to the critical state. This is also very important in the framework of ADS (accelerator driven systems) where the measurement of a subcritical level without knowledge of the critical state is looked for. An alternative way to achieve this is by mean of the ²⁵²Cf source-detector method. The method makes use of three detectors inserted in the reactor: two “ordinary” neutron detectors and one ²⁵²Cf source-detector which contains a small amount of ²⁵²Cf that introduces neutrons in the system through spontaneous fission. By observing fissions through the detection system and correlating the signals of the three detectors, the reactivity ρ (and hence the multiplication factor k) can be determined.     

Experimental determination of subcriticality at subcritical PWRs in Korea

A direct subcriticality measurement system (SMS) based on the Feynman-α method has recently been developed by KEPRI. It was applied to six commercial pressurized water reactors in Korea. However, the obtained Feynman curves failed to give proper multiplication factors. The objective of these tests was to investigate the performance of the Feynman method to predict directly the subcriticality of a given subcritical reactor by using the neutron pulse counts only without any reactor perturbation in the large commercial reactors. Recently, two methods developed by Hokkaido University and Westinghouse Electric Corporation. These methods have a defect due to being based on the modified neutron source multiplication method. To overcome this defect and derive operational benefits is necessary to estimate the subcriticality of a subcritical core directly from the neutron pulse counts only. The performance of the developed SMS was verified in the Kyoto University Critical Assembly and applied to eight 1000 MWe Optimized Pressurized Water Reactors (OPR1000) in Korea. The obtained results show that the SMS based on the Feynman method can be a useful tool to estimate the reactivity of a subcritical power reactor. Although the discrimination level of the signal-processing unit in OPR1000 suffers from noise and gamma ray effects, SMS can provide good Feynman curves and effective multiplication factors. However, the SMS has failed to give the reactivity for the entire measured data set. Improving the SMS and investigating the effects of different discriminator level settings at SPU in OPR1000 will be topics for further study.     

252Cf源驱动噪声分析法测量α初步研究

在一球形浓缩铀临界装置上,采用²⁵²Cf源驱动噪声分析方法对其次缓发临界状态下的α进行测量,在频域内分析数据得到α。对比频域内的两种数据处理方式（功率谱实虚部相除方法和功率谱取模直接拟合方法）,用互功率谱密度函数实虚部相除方法得到在-0.1$情况下,脉冲堆的α为0.58μs^-1,与Rossi-α方法的结果相吻合。