西安脉冲反应堆半实物仿真系统设计研究

为了给在建中的西安脉冲堆数字化仪表与控制系统提供调试用模拟信号源、验证功率调节方法以及人员培训等,研制了一套西安脉冲反应堆半实物仿真系统。提出了半实物仿真系统的设计思想,设计了系统框架。改进了堆芯仿真物理模型,使用MATLAB编制了堆芯实时仿真程序。采用组态王软件编制了人机界面,采用可编程控制器S7-200进行棒位控制和棒位测量。研制了控制棒驱动机构模拟件、信号发生器和手动操作盘等多个硬件设备,建立了系统内部的通讯。在该半实物仿真系统上模拟了升、降控制棒的功率变化和发射脉冲后的脉冲参数,与堆上实验结果符合较好,测量了信号发生器的输出信号,与预期一致。结果表明,该半实物仿真系统能够实现设计目的且性能良好。      

CARR核测量系统设计

核测量系统是CARR仪表控制系统重要的组成部分,该系统监测CARR堆芯外中子注量率并向保护系统、ATWS缓解系统、功率调节系统等提供功率水平信号。本文阐述CARR核测量系统的设计,介绍CARR核测量系统的系统结构、堆外探测器、监测装置和技术特点。     

西安脉冲堆带核调试试验

介绍了西安脉冲堆带核调试试验内容，给出了稳态堆芯、脉冲堆芯及72小时额定功率运行试验堆芯的反应性和中子注量率等参数的测量结果。测量结果证明，西安脉冲堆性能参数达到了设计指标，并为该堆的运行提供了必不可少的运行参数。     

CPR1000核电厂反应堆功率标定系统设计及验证

简要介绍CPR1000核电厂反应堆功率标定系统的运算原理、系统结构、验证过程及结果。采用IAPWS-IF97水和蒸汽物性模型、焓差-热功率原理计算核电厂核蒸汽供应系统热功率与反应堆核功率。该套系统于岭澳核电站3号机组调试启动中成功完成了反应堆功率测量系统、控制系统的实态标定。实际应用表明功率标定系统能够可靠地完成CPR1000核电厂堆芯功率标定。     

反应堆堆外核测量系统的实时仿真

堆外核测量系统实时仿真是核电厂全范围培训模拟器的重要组成部分。本文给出一种基于测量原理的功能仿真处理方法,利用堆芯物理仿真计算出堆芯中子通量密度．建立了堆外核测量值与反应堆内三维中子通量密度分布之间的拟合公式．根据反应堆物理计算或功率刻度实验确定拟合系数．可以实时准确仿真堆外核测量系统,满足核电厂全范围培训模拟器的要求．     

宽量程核测量系统中裂变室输出信号的仿真研究

为建立一个实验室可用的宽量程核测量系统输入信号,本工作进行了宽量程核测量系统中裂变室输出信号的仿真研究,完成了裂变室输出的单个脉冲信号和脉冲叠加信号的仿真计算,并实现了仿真结果的物理输出,为宽量程核测量系统的算法设计和调试打下了基础。     

大亚湾核电站堆芯结构和与物理试验相关的重要系统

简单介绍了大亚湾核电站堆芯结构和与物理试验相关的重要系统。这些系统包括反应堆芯系统(COR)、长棒控制系统(RGL)、堆芯测量系统(RIC)、堆外核测仪表系统(RPN),失水事故监视系统(LSS)、集中数据处理系统(KIT)和试验数据采集系统(KDO)。此外,对于其他系统,例如反应堆冷却系统(RCP)、化学和容积控制系统     

移动式堆芯中子注量率测量系统概述

堆芯中子注量率测量系统是压水堆核电站核测量系统的主要组成部分,用于测量反应堆堆芯的中子注量率水平,从而提供反应堆的功率分布情况。文章介绍了中核（北京）核仪器厂国产化的移动式堆芯中子注量率测量系统,并对测量系统的概况、系统组成、工作原理及功能等进行了描述。     

用于ADS次临界反应堆Keff测量微电流放大器的设计

设计了一种基于噪声分析方法,利用反应堆堆芯中子涨落特性测量加速器驱动洁净核能系统（Accelerator driven system,ADS）次临界反应堆动态参数的微电流放大器,并且根据需要设计了微电流源,运用Protel实时电路仿真工具对放大器进行仿真调试,使其达到设计要求。     

大亚湾核电站堆芯结构和与物理试验相关的重要系统

简单介绍了大亚湾核电站堆芯结构和与物理试验相关的重要系统。这些系统包括反应堆芯系统(COR)、长棒控制系统(RGL)、堆芯测量系统(RIC)、堆外核测仪表系统(RPN),失水事故监视系统(LSS)、集中数据处理系统(KIT)和试验数据采集系统(KDO)。此外,对于其他系统,例如反应堆冷却系统(RCP)、化学和容积控制系统(RCV)、反应堆硼和水补给系统(REA)、核取样系统(REN)及反应堆控制系统(RRC)等,仅对与硼浓度的调节与估算有关内容作了说明。对这些系统的了解不仅可以熟悉它们的作用和功能,更能对堆芯物理试验的原理、步骤、注意事项和技术关键加深理解,也能对试验结果作进一步分析有所帮助。     

Applicability of neutron noise technique for diagnostics of in-core components for heavy water reactors

Neutron flux signal is composed of a steady or mean component resulting from the flux produced by power operation of the reactor and a very small fluctuating component called ‘noise’ component. Analysis of neutron noise from suitably located sensors is a proven technique to monitor the in-core components of light water reactors (LWRs). However, the use of neutron noise has been rare, if any, for heavy water reactors (HWRs) as it was generally felt that the unfavourable transfer function characteristics of the reactors would limit its applicability. To assess the applicability of technique in pressurised heavy water reactors (PHWRs), experiments were carried out using in-core and out-of-core neutron sensors in a research reactor. This paper discusses the measurement details and results of the experiment. This paper concludes that the neutron noise technique can be effectively utilised for diagnostics/characterisation of the in-core components of heavy water reactors.     

钒自给能探测器中子响应计算方法

钒自给能探测器被广泛用作核动力反应堆的堆内固定式探测器,为堆芯中子注量率分布测量连续不断地提供信息。研究钒自给能探测器的响应电流计算方法,为堆芯在线功率分布监测与探测器设计优化提供理论依据。首先描述钒自给能探测器的响应机理与特性,然后基于Warren提出的理论模型,详细介绍中子响应电流控制方程及电子逃脱概率的计算方法,最后根据公开报道的典型钒探测器规格与实验数据进行数值模拟分析。结果显示,单位长度热中子灵敏度计算值与测量值相对偏差在±5%以内,论证了该方法的有效性与计算精度。     

Monte Carlo modeling of KAMINI

Monte Carlo modeling of the Kalpakkam Mini Reactor (KAMINI) has been carried out for the first time by using Monte Carlo code (MCNP4A) and continuous energy cross-sections. The safety control plate (SCP) drop experiment is simulated and the computed integral worth of the SCPs is compared with the measured value. The measured axial neutron flux profile and foil reaction rates in one of the in-core irradiation location and the foil reaction rates at the west beam port are also compared with the predicted results. The agreement between measurements and calculations is quite satisfactory. It is confirmed from the calculation and measurement that the north thimble is having nearly 10–20% higher neutron flux as compared to the south thimble depending on the exact elevation.     

Investigation of the two-phase flow in a boiling water reactor using neutron-noise technique

In this paper, the fluctuations of the neutron flux (“neutron noise”) of the Mühleberg BWR are investigated. Above 2 Hz, the noise measured by the in-core neutron detectors is driven exclusively by local fluctuations of the void fraction. Characteristic changes of the neutron-noise signature along the axis can be attributed to changes of flow pattern. By measuring the phase lag between pairs of axially placed neutron detectors, the transit time of the steam between the detectors can be evaluated. The measured transit times are applied to the study of two-phase flow in the core. The neutron-noise method has the advantage of providing in-core information under operational conditions.     

一种用芯内中子探测器读数校准燃耗后堆芯截面参数的方法

燃耗后的反应堆堆芯截面参数偏离了原来的数值,尽管用燃耗表对其进行了修正,但由于修正过程中的近似,仍使得截面参数的可靠性受到怀疑。文中提出：在用芯内中子探测器读数重构出“测量的”全堆热群中子通量密度分布的基础上,约束堆芯截面参数,并采用节块格林函数法对其进行校准的一套完整方法。截面参数经校准后,可以使理论模型计算结果与实际测量值相一致。仿真结果说明了此方法的可行性。     

反应堆中子通量密度仿真研究

核电厂作为特殊企业,对职工的培训尤为重要,而培训中,仿真系统是必不可少的环节。强果用经典的点堆模型方程,仿真精度不够,不能实现物理仿真。本文利用因子分解方法解中子通量密度函数,在求解中子通量密度形状函数量,通过适当的模型简化,使其可以在一般的ＰＣ机上实现。     

In-Core Neutron Flux Instrumentation

A survey of the new work in the neutron monitoring of a nuclear power reactor is presented. The sensors have been moved into the reactor and the three modes of measurement necessary to cover the ten decades of the in-core neutron range from startup to rated power are described. The system specifications and life performance data are reviewed. Several innovative extensions of in-core monitoring such as the traversing probe and the Campbell method are covered in detail.     

Verification of Source Term Estimation Method against Measured Data for Decommissioning Waste from a CANDU Reactor

The method for the establishment of an equilibrium core model proposed in the previous paper and the source term calculation method proposed in this paper for the characterization of decommissioning waste were verified by comparing the nuclide inventory estimated by MCNP/ORIGEN2 simulations with the measured nuclide inventory according to a chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. At first, the time-average pseudoequilibrium full-core model of Wolsong Unit 1 was developed on the basis of the previously proposed modeling method for the activation of in-core and ex-core structural components. Then, the application level of the neutron flux and cross section in the radionuclide buildup calculation were compromised. Fourteen major actinides and fission products were considered to represent the irradiated fuel condition, and a geometry simplification was also introduced in the burned full-core model for MCNP simulation. The assumption of a constant neutron flux and capture cross section as a function of the irradiation time was applied in the radionuclide buildup calculation in ORIGEN2. As a result, the values estimated from the analysis system agreed with the measured data within a difference range of 30%. Therefore, it was found that the MCNP/ORIGEN system and source term characterization method proposed can be viable to estimate the source terms of the decommissioning waste from a CANDU reactor.     

Study of the uncertainties due to position change of the PWR aeroball measurement system

The aeroball measurement system (AMS) is an important in-core instrumentation in German pressurized water reactors. Therefore, it is essential to know the possible uncertainties of this system. One is the lack of knowledge of the positions of balls in the guide tubes. The position changes can be up to 7 mm. Since the neutron flux distribution is not constant across the guide tubes, different reaction rates can result from the displacements. Both fuel assembly and full core calculations were carried out with the Monte Carlo code MCNP5. Differences in the reaction rates of up to 2% could be determined. In most cases, differences are only up to 0.5%. The results were hardly influenced by burnup and boron concentration in the water moderator. For fuel assemblies containing gadolinium as a burnable poison, a more pronounced reduction could be observed in the direction towards the gadolinium fuel rods. Overall, it was found that the AMS measurement values are very robust with regard to possible variations of ball positions.