重水堆核电站主蒸汽旁排阀故障分析和应对策略

根据重水堆电站的电厂控制模式及蒸汽发生器压力控制的原理和特点,结合实际事例对主蒸汽旁排阀故障开启对机组运行的影响及人员响应进行了分析和探讨,总结了处理类似事件的原则和经验.     

秦山三期CANDU-6核电机组技术规格书的修改

描述了制定秦山三期CANDU 6核电机组技术规格书 (TS)的核安全法规依据 ,简要介绍了按照中国核安全法规的要求、参照美国NRC制定的压水堆核电站成熟的标准技术规格书的格式和应用加拿大CANDU机组采用的运行方针和政策所积累的成熟的运行经验 ,对AECL提交的TS进行修改的情况以及对TS中一些重要技术问题的修改内容 ,说明了修改后的TS基本满足了中国核安全法规的要求 ,可以在核电站的运行中使用     

Performance evaluation of a new signal processing system design to improve CANDU SDS1 trip response during large break LOCA events

Performance of a recently developed signal processing system for CANDU (Canada Deuterium Uraniu) reactor shutdown system 1 (SDS1) is evaluated in this paper. The evaluation is carried out in MATLAB/Simulink software environment as well as with an existing power measurement and signal processing system. The new signal processing algorithm is obtained based on the synthesis of several first order low pass filters with different delayed time constants. Throughout this paper, a special attention has been paid to compare the new signal processing system with the existing one. The dynamic behavior of the new signal processing system in the practical large loss of coolant accidents (LLOCA) events has also been examined. Simulation results show that during the LLOCA event, the reactor trip time, as well as the peak power, is decreased remarkably. Through the simulation studies, it has convincingly demonstrated that the new signal processing system has significant advantages over the existing system in terms of the improved trip response and accommodation of the spurious trip immunity. This advantage will significantly enhance the safety margin, or will bring economical benefits to nuclear power plants.     

TACR排管容器研究中多孔介质法的应用

以钍基先进重水堆(TACR:Thorium Based Advanced CANDU Reactor) 慢化剂系统排管容器作为研究对象,提出了一种适用于计算排管容器内流场和温度场的非结构网格多孔介质方法,并将其计算结果与精细数值计算法所获得的结果进行了比较,二者符合较好,由此验证了其方法与计算结果的正确性.     

TACR压力管与排管间辐射换热分析

辐射换热是钍基先进CANDU型反应堆(TACR)压力管和排管间换热的主要途径.本文以灰体辐射模型和电网络分析方法为基础建立了TACR压力管和排管间辐射换热的计算模型,利用该模型计算了给定温度边界和热流密度边界的情况下,压力管和排管间的辐射换热能力.计算结果表明,该模型可以用于TACR压力管和排管间辐射换热能力的计算.     

Overview and Lessons from More Than a Decade of Feeder Life Management

This paper provides an overview of the degradation in CANDU~a （Canadian Deuterium Uranium） feeders in operating plants observed from the mid 1990s onward. The degradation has been dominated by feeder wall thinning, caused by FAC （flow accelerated corrosion） and feeder cracking. This paper summarizes the industry＇s response to the discovery of these two degradation mechanisms and the methodologies, tools and technologies developed to monitor the degradation and assess the continued fitness for service to manage the plant life. This paper identifies some of the lessons gained from more than a decade of industry effort, and discusses how these lessons are being implemented in the refurbished and new CANDU plants. CANDU~a is a registered trademark of Atomic Energy of Canada Limited.     

Analysis of on-line maintenance strategies for -out-of- standby safety systems

The objective of this paper is to compare the performance of three on-line test and maintenance strategies (corrective maintenance, preventive maintenance and predictive maintenance) for standby k-out-of-n safety systems. Each channel of the k-out-of-n system is modelled by an age-dependent unavailability model to reflect the effect of maintenance on the aging process. The system unavailability, the probability of spurious operation and the overall cost under the above maintenance strategies are analyzed and compared to obtain the optimal maintenance strategy. Sensitivity analyses are performed to reveal the effect of different model parameters on the system performance. A standby safety system in Canadian Deuterium–Uranium (CANDU) Nuclear Power Plants (NPPs), the Shutdown System Number One (SDS1), is used to illustrate the proposed analysis and the procedure. It is concluded that maintenance should neither be performed too frequently nor too rarely. When the system deteriorates very slowly, the corrective maintenance is more preferable than the preventive and predictive maintenance. When the failure rate of the system is high, the preventive maintenance results in the best system performance.     

Analysis of surveillance test interval by Markov process for SDS1 in CANDU nuclear power plants

The Canadian Nuclear Safety Commission (CNSC) requires that each shutdown system (SDS) of CANDU plant should be available more than 99.9% of the reactor operating time and be tested periodically. The compliance with the availability requirement should be demonstrated using the component failure rate data and the benefits of the tests. There are many factors that should be considered in determining the surveillance test interval (STI) for the SDSs. These includes: the desired target availability, the actual unavailability, the probability of spurious trips, the test duration, and the side effects such as wear-out, human errors, and economic burdens. A Markov process model is developed to study the effect of test interval in the shutdown system number one (SDS1) in this paper. The model can provide the quantitative data required for selecting the STI. Representing the state transitions in the SDS1 by a time-homogeneous Markov process, the model can be used to quantify the effect of surveillance test durations and interval on the unavailability and the spurious trip probability. The model can also be used to analyze the variation of the core damage probability with respect to changes in the test interval once combined with the conditional core damage model derived from the event trees and the fault trees of probabilistic safety assessment (PSA) of the nuclear power plant (NPP).     

CANDU重水堆燃料管理

论述秦山三期核电站所采用的ＣＡＮＤＵ６ 反应堆的燃料管理。ＣＡＮＤＵ 堆的换料是带功率进行的， 这一特征使得它的堆内燃料管理与必须停堆换料的反应堆有明显的不同。ＣＡＮＤＵ 堆燃料管理有设计和运行两方面的内容。在设计阶段， 燃料管理涉及堆芯时均通量／ 功率分布的设计； 在运行阶段， 电厂换料工程师的职责包括选择要换料的燃料通道， 跟踪堆功率变化史， 以及确保各最大功率限值不被超越。     

秦山三期核电站进口设备质量监督

介绍了在总承包商交钥匙合同模式下，业主在进口设备质量监督方面所开展的工作，以及从中得到的经验教训。