CEFR钠空泡反应性效应试验测量与计算分析

本文介绍了中国实验快堆物理启动试验中钠空泡反应性效应测量试验的试验程序及测量结果评估,测量结果显示中国实验快堆典型位置钠空泡反应性价值皆为数值较大的负反应性,结果符合试验验证要求,验证了组件瞬间堵流事故专设监测系统的信号基础。并对试验进行了计算分析,试验前的分析为试验提供支持,试验验证了计算分析程序系统。     

先进数字式反应性仪跨量程连续测量误差补偿方法

动态刻度控制棒试验方法(DCRM)的发明极大地减小了核电厂控制棒价值测量试验的难度,压缩了试验时间。动态刻棒试验需要使用先进数字式反应性仪(ADRC)。与传统反应性仪使用环境不同,动态刻棒试验时,反应性仪需实现多量程测量以满足探测器电流快速下降多个量级的需求。进行反应性仪量程切换时产生的测量值的误差会对反应性测量结果产生影响。本文通过分析动态刻棒过程中探测器电流曲线特征,提出基于曲线拟合进行测量误差补偿的方法,核电厂试验数据表明,应用此种方法可以将反应性仪量程切换带来的计算最大误差由48%减小到2.0%。     

启明星Ⅱ号安全控制部件的实验验证

启明星Ⅱ号零功率装置（启明星Ⅱ号）所设计的安全控制部件有安全棒和调节棒,这些控制部件是反应堆安全运行的关键。本文采用逆动态反应性计测量的方法对所选定的控制部件的反应性价值进行了实验测量,并与理论计算结果进行了比较。结果表明,安全控制部件的反应性价值的实验测量结果与理论计算结果的相对偏差为4.46%,二者吻合较好。安全棒系统经力学分析评定,结果表明不会出现卡棒现象,能实现快速停闭反应堆的目的。安全棒系统、调节棒系统的机械性能经堆上反复实验验证,各系统性能稳定可靠,重复性好。     

次临界反应性测量方法研究

压水堆启动前需要进行物理启动试验,其中调临界是相对耗时的一部分。采用次临界条件下的试验,可以省去调临界的步骤,提高物理启动试验的安全性和适用性,加速物理启动试验的进程,提高压水堆的负荷因子。目前,次临界反应性测量受限于测量信号较弱,背景噪声强,误差较大,难以满足商用压水堆的工程应用要求。本文基于源倍增方法,利用空间重要性修正因子和信号线性修正提高次临界反应性测量的精度,成功地在核电厂首循环上实现了次临界反应性测量的蒙特卡罗方法计算,取得了较好的计算精度与效果。     

动态刻棒技术的应用

在前期经过认真调研的基础上,大亚湾核电厂和岭澳核电厂于2009年开始研究西屋先进的棒价值测量技术——动态刻棒技术。经过理论培训、计算培训、对比试验,大亚湾核电厂和岭澳核电厂已成功实施动态刻棒技术。动态刻棒技术在大亚湾核电厂和岭澳核电厂的正式实施表明,动态刻棒技术在保证测量精度满足要求的前提下,可以提高试验过程的安全性,简化反应性操作、减少废水排放,可以大幅缩短大修关键路径。     

铀棒栅临界实验装置安全控制部件价值的实验验证

<正>铀棒栅临界实验装置设计的安全控制部件有安全棒和调节棒,其是装置安全运行的关键。采用逆动态反应性计测量方法对所选定的安全棒和调节棒的反应性价值进行了实验测量,并与理论计算值进行比较,列于表1。表1中:理论计算值的标准差为0.012%;Δρ=实验测量值-理论计     

CEFR组件替换反应性价值试验测量与计算分析

组件替换反应性价值定义为测量位置组件替换成相应组件时引入的反应性变化。中国实验快堆物理启动试验中组件替换反应性价值测量试验方案中,试验测量了8个典型位置,其中6个位置为燃料组件替换成不锈钢组件,另外两个为不锈钢组件替换成燃料组件。测量结果显示,燃料组件替换反应性价值由内至外依次减少,内圈燃料组件替换反应性价值约-980 pcm,外圈燃料组件替换反应性价值约-470 pcm,补偿棒棒组测量和单根补偿棒测量的结果差别微小。使用CITATION程序对试验方案进行了理论计算,结果表明,计算结果与实验值符合良好,检验了CITATION程序的工程设计实用性。     

秦山核电厂反应堆控制棒价值和硼价值测量试验

本文系统地介绍了秦山核电厂反应堆控制棒组价值和硼价值在首次物理启动中的测量试验。简要介绍了测量方法、仪器装置、试验经过、实测结果和误差分析。试验结果表明:测量值与理论计算值符合得很好,达到了验收准则的要求。     

CAP1400数值反应堆系统关键技术研究及示范应用

本文系统介绍了“大型先进压水堆及高温气冷堆核电站”国家科技重大专项课题“CAP1400数值反应堆关键技术研究”的主要研究成果。课题首先分别开发了基于确定论方法和蒙特卡罗方法的高保真堆芯物理计算程序,然后开发了pin by pin先进子通道分析程序和基于精细网格的燃料棒性能分析程序,以此为基础建立了物理 热工 燃料性能多物理耦合的CAP1400数值反应堆系统。利用国际基准题VERA、AP1000启动物理实验参数对数值反应堆系统进行了验证和确认,并进一步实现了CAP1400大型先进压水堆的启动物理参数、循环模拟分析和部分功率能力分析的示范应用。数值结果表明,所开发的数值反应堆关键分析软件具有很高的计算精度,可直接服务于CAP1400的设计验证、物理启动和运行支持。     

大型先进压水堆弹棒事故模拟及敏感性分析

相对于传统堆型,大型非能动先进压水堆堆芯设计具有重大改变,这些改变对弹棒事故分析具有重要影响,进而影响反应堆的安全性。通过选取典型的四类工况（寿期初满功率、寿期初零功率、寿期末满功率和寿期末零功率）,利用中子动力学软件和燃料性能分析程序开展大型先进压水堆CAP1400的弹棒事故模拟计算,验证大型先进压水堆弹棒事故工况下的安全性,并针对弹棒事故分析关键输入参数开展敏感性分析。计算分析结果表明：大型先进压水堆发生弹棒事故时,其结果能够满足验收准则的要求,反应堆处于安全可控状态;弹棒事故分析中功率峰值对弹棒价值最敏感,事故分析结果对停堆反应性敏感性较小。     

Feasibility Study for Evaluation of Control Rod Worth in Pressurized Water Reactors using Neutron Count Rate during a Control Rod Drop Testing

Applicability of the modified Neutron Source Multiplication (NSM) method with extraction of the fundamental mode to subcriticality measurement has been proposed. Following the feasibility verification in the previous study based on numerical analyses, its applicability has been proven in a more realistic situation; in a withdrawal sequence of control rod banks during the PWR startup. Subcriticalities with various control rod insertion configurations were estimated based on the modified NSM method. The subcriticality could be evaluated with a good accuracy even with the mockup experiment where any special treatments for accurate measurement were not taken into account and furthermore the insensitivity of measured signals by reactivity changes and their large fluctuations were seen.

Based on this fact, we further investigated a feasibility to use neutron count rate data obtained during the control rod drop testing, which is carried out before the reactor physics tests at hot zero power condition. When it is proven that these data could be used for the estimation of each control rod worth, the following reactor physics tests could be performed with the advanced knowledge of each control rod worth and procedures for detailed control rod worth measurement could be simplified or eliminated from the reactor physics tests.     

New dynamic method to measure rod worths in zero power physics test at PWR startup

To measure and validate the worth of control (or shutdown) bank in zero power physics test at PWRs, a dynamic control rod reactivity measurement (DCRM) technique has been developed and applied to six startups of Westinghouse plants as well as Korea Standard Nuclear power Plants based on the Combustion Engineering System 80 NSSS. With this technique, just one test bank is inserted into the bottom of the core at maximum stepping rate and withdrawn immediately to the all rod-out position. Specially designed inverse point kinetics equations are used to determine the test bank worth from the measured ex-core detector signals, which are controlled by the neutron-to-response conversion factor and the dynamic-to-static conversion factor. These two parameters are predetermined by the three-dimensional neutron adjoint flux distribution for both the top and bottom ex-core detector and the three-dimensional steady and transient core power distribution for test bank movement. To eliminate the gamma-ray effect on ex-core detector signals, a simple method, using reactivity curve characteristics, was also developed. To verify the DCRM method, a total of 28 bank worths of six different PWRs was measured by the DCRM and compared with those of conventional method. Results show that the DCRM method has a similar accuracy as the conventional technique. However, with the DCRM method, it only takes ∼15 min per bank from the beginning of rod insertion to the determination of measured static worth. From its performance, one can conclude that the DCRM method is an effective replacement for the conventional rod worth measurement method.     

动态棒价值测量软件包LIGHT的开发与数值验证

动态棒价值测量是一种快速测量控制棒组价值的方法。基于测量过程和相关的反应堆物理数值计算方法，开发了动态棒价值测量软件包LIGHT。LIGHT可产生进行动态棒价值测量所需的参数，包括静态空间因子、动态空间因子和缓发中子参数。针对基准问题和AP1000核电厂进行了数值计算并进行了比较。分析表明，计算结果具有较高的精度，说明建立的计算模型及开发的程序是正确的。     

反应性计校核试验

反应性计是秦山核电厂物理启动和运行过程中进行物理试验的主要仪器,其测量准确性直接关系到各项堆物理参数的测量误差。为保证各项物理试验结果的准确可靠,按照物理启动程序,在反应堆达到首次临界的热态零功率水平时,先进行本试验。本文简要介绍了由上海核工程研究设计院和北京核仪器厂设计、研制的模拟-数字混合式反应性计的基本原理。用本系统在堆上测量反应性和同时用周期法测量的值进行对比校核试验,结果表明二者之间偏差在±4％以内。符合本试验验收准则。     

COSINE软件包基于三代非能动压水堆低功率物理试验的确认与评估

基于三代非能动压水堆AP1000堆型的低功率物理试验结果,对COSINE软件包核设计软件的计算功能及计算精度进行确认与评估。从低功率物理试验的控制棒价值、所有棒组全提(ARO)工况末端硼浓度、ARO工况等温温度系数对比结果可以看出,COSINE软件包核设计软件计算结果与AP1000堆型低功率物理试验中的各项实测数据符合较好,均满足工程设计要求,计算精度良好。     

铅铋堆零功率实验装置控制棒价值测量及方法优化

零功率实验装置的控制棒价值测量一般采用周期法、置换法或落棒法对刻棒实验进行简单处理。为提高刻棒效率,本文提出了无补偿的多步降棒刻棒方法,采用该方法对我国首个铅铋堆零功率实验装置控制棒价值进行了测量,与补偿刻棒方式及落棒法测量结果进行了对比,并通过理论计算验证了该方法的准确性。结果表明：本文方法有效降低了空间效应对测量值的影响,控制棒价值测量结果准确可靠,可在较短时间内完成较高精度的刻棒实验,适用于需经常更换装料方案的临界实验装置。     

Control Rod Worth Measurement and Method Optimization of Lead-bismuth Reactor Zero-power Experimental Device

The control rod worth measurement of the zero-power experimental device is generally performed by the periodic method, the substitution method or the rod drop method. In order to improve the efficiency of measurement, a multi-step rod-insertion method without compensation was proposed. The control rod worth of the first lead-bismuth reactor zero-power experimental device in China was measured by this method, and the results were compared with those obtained by the compensation method and the rod drop method. The accuracy of the method was verified by theoretical calculations. The results indicate that this method effectively reduces the space effect on measured worth, and the measurement results of control rod worth are accurate and reliable. The method can complete the higher precision rod worth measurement in a short time, and is suitable for the critical experimental device that needs to change the charging scheme frequently.     

An assessment of a VVER-1000 core during Turbo-Generator load reduction test using RELAP5/MOD3.2 and WIMSD-5B/PARCSv2.7

Reactor dynamic tests, which are categorized as one of the power start-up test groups, are the most complex tests during commissioning of the new nuclear power plants. This paper presents the results of Turbo-Generator load reduction test as one of the reactor dynamic tests for VVER-1000/V446 unit at Bushehr Nuclear Power Plant (BNPP). In this test modeling because of the need for control rod bank worth and core reactivity coefficients, the core geometry has been modeled first by using WIMSD-5B/PARCSv2.7 codes for neutronic calculations. For performing the thermal-hydraulic analysis, the RELAP5/MOD3.2 computer code has been used. The control rod bank worth and core reactivity coefficients obtained from WIMSD-5B/PARCSv2.7 are compared with BNPP FSAR that confirm the ability and reliability of the method. Also comparison of the thermal-hydraulic core parameters obtained from RELAP5/MOD3.2 against actual plant data, indicate that this code can properly predict behavior of VVER-1000 reactor for this dynamic start-up test.     

次临界反应性测量的空间修正及其应用综述

次临界下的反应性测量技术有着自身的特点,次临界下控制棒的动作、堆芯的次临界度以及外中子源的存在都会对堆芯中子通量的分布产生影响,因此通常情况下堆芯的次临界度只能"监视",无法准确测量。在堆芯模拟软件发展的基础上,国外科研人员提出了次临界下点堆模型的空间修正方法,将这种方法用于动态棒价值测量(DRWM),并在此基础上进一步发展了次临界控制棒价值测量(SRWM),这些技术有的已经被国内核电站使用,但是国内对空间修正的原理及方法鲜有介绍。本文针对这种需求,总结概括了国外商用堆次临界反应性测量的基本原理与方法,并结合反应性测量仪表技术,给出了次临界反应性仪的数据处理流程,这对于推进国内商用堆次临界反应性测量的研究和实际应用具有较为重要的意义。