蒸汽发生器泄漏率监测

基于主蒸汽管道中＾１６Ｎ核素监测的蒸汽发生器泄漏率监测方法,目前已成为蒸汽发生器泄漏监测的主导方法。本文主要介绍了蒸汽发生器泄漏率的监测原理、蒸汽发生器泄漏率与＾１６Ｎγ计数率的换算关系以及实际应用中存在的问题,并简要介绍了秦山和大亚湾核电站年用该类监测系统的主要特性。     

核电站蒸汽发生器泄漏监测——氮-16辐射监测仪概述

蒸汽发生器是压水堆核电站核蒸汽供应系统的主要设备之一,对蒸汽发生器传热管进行泄漏监测关系到核电站的安全和经济运行。介绍了用于蒸汽发生器泄漏监测的氮-16辐射监测仪的概况、工作原理、系统组成等。     

蒸汽发生器泄漏率监测仪应用与改进

介绍了蒸汽发生器泄漏率(16N)监测仪的应用现状,分析了当前主流产品应用问题,提出设备设计、系统设计的改进建议.     

蒸汽发生器泄漏率监测仪修复经验总结

目前压水堆核电站使用辐射监测仪对主蒸汽管道中的N-16核素浓度进行连续监测,由于蒸汽发生器泄漏监测仪长期工作在高温高湿环境下,普遍存在故障率偏高的问题,影响到了核电站相关系统的正常运行。从6台SGLM201型蒸汽发生器泄漏率监测仪的故障诊断、故障原因分析、修复技术研究等方面入手,分析了总γ计数异常闪发故障、稳峰异常故障...     

压水反应堆蒸汽发生器泄漏监测系统优化研究

为了优化蒸汽发生器泄漏监测系统性能,分析了放射性物质在二回路和蒸汽发生器内的迁移规律,建立了蒸汽发生器传热管发生破损后炉水和主蒸汽管道内放射性核素的计算模型,获得了利用排污水和主蒸汽管道两测点数据计算蒸汽发生器传热管破损部位和泄漏率的方法,并利用蒙特卡罗法仿真排污水γ能谱,确定了排污水放射性探测器的类型。     

用N—16测量蒸汽发生器的泄漏

利用索伦多电子公司（ＳＥ）设计众多放射性监测应用设备的经验和西屋电气公司在蒸汽发生器方面的技术知识，这两家公司已合作研制出了一种先进的监测蒸汽发生器一次侧到二次侧的泄漏率的系统。该系统的多通道结构设计和现代化的电子设备，提供一种可靠的、精确的控制室读数，而且安装和维护的费用低。     

苏联压水堆卧式蒸汽发生器的汽水分离装置

目前,苏联压水堆核电站主要有 BBЭP-210,BBЭP-365,BBЭP-440和 BBЭP-1000型反应堆,相应的ПГВ-1,ПГВ-3,ПГВ-4和ПГВ-1000型蒸汽发生器均为卧式。运行经验表明,各型卧式蒸汽发生器的汽水分离装置均能保证饱和蒸汽出口湿度不超过规定的指标(0.2—0.25%)。文献[1]对这几种型式蒸汽发生器的结构特点,传热特性的计算与试验,二回路水化学处理和     

快堆电站蒸汽发生器中小泄漏探测系统的设计

文章描述和分析了用于快堆电站蒸汽发生器小泄漏探测的扩散型微氢探测系统,建立了钠-水反应和氢通过镍膜扩散的理论模型。在实验数据的基础上对本系统的灵敏度、响应时间、机械性能和稳定性进行了讨论和评价,结果表明国内自行研制的镍管具有良好的机械性能和透氢性能,本系统能够满足快堆电站的要求。     

一种新型蒸汽发生器结构设计

目前设计的核岛蒸汽发生器经长期运行后,分离干燥器如有零部件损坏或脱落会造成U形管破裂,并引起一回路载热体外泄导致潜在的核岛安全事故,因此提出了一个新型蒸汽发生器的设计结构,此蒸汽发生器可避免以上事故的发生。     

核电厂主蒸汽16N监测用于蒸汽发生器泄漏率测量的研究

文章介绍了CPR1000、AP1000和EPR核电厂广泛采用的法国新型16N监测设备的组成和技术特点,通过实例描述了16N探测器可用性放射源检查试验.对16N监测的量化计算方法进行了分析和研究,根据国内外已有研究成果,推导出了量化计算公式.介绍了红沿河核电站某机组的蒸汽发生器泄漏率16N监测计算系数和应用实例.     

Large leak sodium-water reaction analysis of an LMFBR steam generator using a variable temperature spherical bubble model

A simplified 1-D numerical model has been developed to study the pressure transients due to a large leak sodium-water reaction. In the initial stage the hydrogen generated in the reaction is taken to be of spherical shape and once the diameter of the bubble equals the shell diameter its growth is modelled in a columnar form. The model has been used to predict the pressure evolution in the reaction site for a typical experiment and has performed well. Also presented is a comparison with the predictions of a numerical 3-D model for the same experiment.     

Development of an acoustic leak monitoring system

The purpose, structure, and basic characteristics of an acoustic system for monitoring coolant leaks in the first loop of a VVER reactor are presented. The principle of operation of the diagnostics algorithm is described. The system has been checked on a special stand. The system satisfies all sensitivity and temporal requirements. In 2005, SAKT was put into experimental operation in the No. 3 unit of the Kalinin nuclear power plant. __________ Translated from Atomnaya énergiya, Vol. 103, No. 6, pp. 342–347, December, 2007.     

Feedback control of a large steam generator

Feedback control of a large counter current heat exchanger employed as a once-through steam generator for a breeder reactor power plant is presented. The mathematical model of the exchanger consists of a set of nonlinear partial differential equations with boundary conditions specified at three space points. The system structure does not allow the control signals to be distributed in space; they are applied only at the boundary points.The objective of the investigation is to choose a practical implementable time-invariant feedback controller which would give satisfactory regulation of the system. Although the emphasis is placed on a specific once-through steam generator system, the techniques used here are equally applicable to other heat exchanger systems.     

A new LMR steam generator free from SWR with a double tube bundle configuration

This paper describes the activities made at KAERI to develop an advanced liquid metal reactor (LMR) steam generator which is free from a sodium water reaction (SWR) to resolve the concern of the SWR possibility and improve the economic features of the LMR. The steam generator design houses two tube bundles that are functionally different and its tube bundle region is radially or vertically divided into two. The SG is equipped with hot and cold fluid tube bundles, a medium fluid and a pump. It prevents the occurrence of the sodium water reaction while sodium is still used as the coolant for the primary heat transport system. The feasibility of using the SG with a double tube bundle for an actual use in a LMR plant is evaluated by setting up the skeleton of the NSSS for various possible configurations of the SG tube bundles.Analysis was made for various types of the new steam generator with a double tube bundle. Since the heat transfer in the SG is made among three different fluids, it has unique heat transfer characteristics. The analysis showed the possibility of the occurrence of an undesirable reversed heat transfer not only in the integrated single-region bundle type but also in the integrated double-region bundle type. The performance analysis revealed practical performance characteristics for the various bundle configurations. Also the feasibility study for the various NSSS configurations with the new SG is described.     

Design and modification of steam generator safety system of FBR MONJU

MONJU is a prototype fast breeder reactor (FBR) in Japan. The sodium–water reaction in the steam generator (SG) is one of the important safety assessment items for a sodium cooled reactor like MONJU. MONJU is equipped with hydrogen gas detectors for the small water leak detection, gas pressure gauges for the medium leak and sensors of rupture discs for the large leak. As a design basis accident, one tube failure then failure propagation of neighboring three tubes is assessed to verify the structural integrity of the secondary components. A latest evaluation method on the design margin against the overheating tube rupture showed that the present SG system had not an enough margin in the worst case. For improving the margin, it needs to shorten the time of the sodium–water reaction by earlier water leak detection in the SG and sooner water ejection from the SG tubes. Therefore, MONJU is now carrying out the following modification works: (1) addition of steam relief valves, (2) addition of a gas pressure gauge with changing the interlock logic and lowering the trigger level, (3) reducing the opening of the valves on the SG gas flow line to the dump tank because of earlier detection for the pressure rise. After this modification, the design margin of the SG system will be sufficiently improved.