共查询到19条相似文献,搜索用时 734 毫秒
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堆芯中子注量率测量系统是核电站监测系统的一个重要组成部分。它主要测量反应堆堆芯的中子注量率分布,监测堆芯功率畸变,积累燃耗数据,对核电站的安全运行及经济性起到重要作用。论文简单介绍了AP1000和EPR堆芯中子注量率测量系统的组成和特点,分析比较了两者之间的差异性。 相似文献
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微型裂变电离室是一种反应堆上广泛使用的堆芯中子探测器。国内CPR1000核电机组的堆芯中子注量率测量系统采用移动式微型裂变电离室作为中子探头,在反应堆运行过程中测量反应堆中子通量,提供堆芯中子通量分布图,是核电站重要的安全仪控设备。对标现役国外产品的服役条件和技术指标要求,研制了一款移动式微型裂变电离室中子探测器,并参照国家标准GB/T 7164-2022和行业标准NB/T 20215-2013,对探测器的核特性进行了测试。测试结果表明:其核特性与国外产品相当,有望实现该反应堆安全产品的“国产替代”。 相似文献
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反应堆启动初始阶段,中子注量非常低,是一般核测量系统的测量盲区。针对测量盲区的问题,设计了一种高灵敏度宽量程的中子注量率探测器。通过计算及实验表明,该探测器具有稳定的性能,能提供一种反应堆物理启动过程中盲区中子注量率测量的方法。 相似文献
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针对研究堆动态参数测量现有方法不足,基于反应堆中子噪声分析方法,设计了一套核功率测量系统。该系统通过对信号前置放大和信号调理的自适应控制测量反应堆临界后的核功率,实现反应堆中子噪声和核功率的智能化、自动化监测。试验测量结果表明:该系统测量的核功率与中子注量率分布测量的理论计算功率值一致,验证了系统测量的有效性,为反应堆核功率测量提供了一种便捷、可靠的测量手段。 相似文献
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介绍了基于OPC标准开发的用于C2工程堆芯中子注量率测量系统的测控数据管理软件(以下简称为CORE_DMIS),它是C2工程堆芯中子注量率测量系统的重要组成部分。运行于主机柜计算机的测控软件实现系统监控一体化,而运行于通道柜的测控软件实现系统监测。 相似文献
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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. 相似文献
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The design of the nuclear instrumentation system for the Pluto series of nuclear ramjet test reactors is an attempt to provide a very flexible nuclear sensing system that will be adequate for Tory II-C and following test reactors. The nuclear detectors will be exposed to the leakage neutron flux from the reactor during operation. Since the leakage flux is proportional to reactor power, the neutron detectors will give a measure of reactor power. A difficulty in providing nuclear instruments for this reactor is the uncertainty in the neutron energy spectrum of the leakage flux at the detectors. Since detector response varies with neutron energy, a large margin of flexibility is desirable. A difficulty which may be encountered is a significant shift in neutron energy spectrum at high power and temperature. This would make indicated nuclear power nonlinear with calorimetric power. A difficulty in insufficient instrument overlap was encountered with the Tory II-A experiment where a large margin of flexibility would have been useful. The detector placement for the Tory II-A experiment had the power range detectors in line with the reactor and main air pipe. At high air flows there was a much greater mass of air between the detectors and the reactor, allowing fewer neutrons to reach the detectors per unit reactor power. This is the reason for the power range detectors being placed off to the side of the Tory II-C test vehicle. Not all difficulties can be foreseen, but provision is made where possible to overcome them. 相似文献
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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. 相似文献
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O. Glckler 《Progress in Nuclear Energy》2003,43(1-4):91-96
Periodic testing of the dynamics of the shutdown systems and their instrumentation is performed in the CANDU nuclear power plants of Ontario Power Generation (OPG) and Bruce Power. Measurements of in-core flux detector (ICFD) and ion chamber (I/C) signals responding to the insertion of shut-off rods (shutdown system No. 1, SDS1), or to the injection of neutron absorbing poison (shutdown system No.2, SDS2) are regularly carried out at the beginning of planned outages. A reactor trip is manually initiated at high power and the trip response signals of ICFDs and I/Cs are recorded by multi-channel high-speed high-resolution data acquisition systems set up temporarily at various locations in the station. The sampling of the seaprate data acquisition systems are synchronized through the headset communication systems of the station. A total of 120 station signals can be sampled simultaneously up to 2500 samples per second. The effective prompt fractions of the ICFDs are estimated from the measured trip response. Effectiveness and the timeline of the trip mechanism are assessed in the measurement as well. The measurement can identify ICFDs with abnormally slow response (under-prompt) or overshooting response (over-prompt) at the beginning of the outage. The time required for the signals to drop to predefined fractions of their pre-trip values (level crossing time) is plotted as a function of detector position and compared against safety requirements. The propagating effect of shut-off rod insertion or poison injection on the flux is monitored by the level crossing times of ICFDs and ion chambers. 相似文献
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Local power density (LPD) at the hottest part of a hot nuclear fuel rod should be estimated accurately to confirm that the rod does not melt. The power peaking factor (PPF) is defined as the highest LPD divided by the average power density in the reactor core. In this paper, the PPF is calculated by support vector regression (SVR) models using numerous measured signals from the reactor cooling system. SVR models are regression analysis models using a kernel function for artificial neural networks. Their neural network weights are found by solving a quadratic programming problem under linear constraints. SVR models are trained using a training data set and then verified against another test data set. The proposed SVR models were applied to the first fuel cycle of the Yonggwang nuclear power plant unit 3. The root mean square errors of the SVR model, with and without in-core neutron flux sensor signal inputs, were 0.1113% and 0.0968%, respectively. This level of errors is sufficiently low for use in LPD monitoring. 相似文献