核电厂控制区出入监测系统联网改进

以某核电厂的改造项目为例,在现役运行核电厂控制区出入监测系统的基础上,对现场运行设备进行联网改造并增加相应的集中控制管理软件,增设视频采集模块采集测量沾污人员的面部信息,实现对现场设备状态的实时监控及设备报警后的人员响应,提高现场辐射安全管控能力。为核电厂控制区出入监测系统及类似系统的联网升级优化提供参考。     

核电厂辐射控制区出入监测系统的改进优化

本文介绍了核电厂辐射控制区出入监测系统的基本情况,基于在运核电机组的应用反馈和管理经验,分析了现有设计中存在的问题,结合新技术的发展和管理要求的提升,从网络架构、清洁控制区、自动配发、性能提升、剂量大数据等方面提出了改进优化建议,对防止污染扩散、提高剂量管控水平等有一定的意义。     

核电厂辐射控制区出入控制模式优化

电子剂量计系统是核电厂辐射控制区人员出入许可控制环节中最重要的手段,目前我国核电厂中广泛使用法国MGP公司的电子剂量系统。本文探讨核电厂在电子剂量计系统与出入控制联动的基础上实现人员指纹识别的可行性和识别系统的初步设计。     

EPICS辐射控制区出入监测系统构建

构建了基于EPICS集成的控制区出入监测系统。该系统采用了以太网TCP/IP协议进行数据通信,完成了对EPICS IOC、历史数据库、OPI操作员界面等配置开发,经过了性能测试并投入运行,实现对人员进出权限管理、个人剂量统计、表面污染检测等功能,满足了人员进出辐射控制区的辐射安全监管要求。     

群堆管理模式下的核电厂实物保护系统

论述了早期核电厂实物保护系统存在的问题及缺陷,结合新形势对核电厂实物保护系统的要求,提出了核电厂实物保护系统的发展趋势。在分析了能够应用于核电厂实物保护系统的新兴技术后,提出了一种核电厂实物保护系统的架构。     

核电厂辐射控制区不同进出模式运行成本分析

本文从核电厂控制区两种进出模式在防护用品的使用量、清洗防护用品的废液产生量以及固废产生量方面对日常运行期间和大修期间进行了运行成本分析。分析可知,采用清洁控制区模式控制区进出模式无论在日常运行期间还是大修期间,其运行成本均低于传统“七件套”模式。尤其是日常运行期间,清洁控制区模式的运行成本仅占“七件套”模式的约三分之一。     

群堆模式下核电厂采购分工模式精细化管理的探索与实践

核电厂群堆管理模式已成为我国发展核电的主要趋势,由于群堆模式下堆型复杂、机组数量多、运行状态不同等因素,电厂的采购工作存在采购量大、物资种类复杂、市场供应条件差异大等难点,加之新时代下更加多元的市场环境,如何规范、高效、专业地开展群堆模式下的物资服务保障工作,将是今后采购工作的重点。采购管理的精细化是新形势下采购工作的主要目标,通过采购分工模式作为出发点,以秦山核电为研究对象,探索多机组、多堆型的群堆模式下采购的分工模式,在采购分工模式方面提出了机构重组、流程再造、发挥专业化优势、强化供应商管理等的分段式采购模式,优化了采购流程、提高了采购效率。     

核电站群堆运行模式下热释光剂量计的管理方法探讨

探讨了大亚湾核电站以及其他核电站所采取的热释光剂量计(TLD)的管理方法,对其管理模式的优缺点进行分析和评价.针对核电站多基地的群堆运行模式,提出了一种全新的管理方法-智能管理模式,它是一种既可保证测量的准确度又能大大降低生产成本、提高管理效率的最优化管理模式.     

核电厂停堆工况下事故及其处置研究

核电厂停堆工况下事故及其处置研究@李吉根@杜春@刘洋@周红     

AP1000核电机组辐射控制区出入口设计改进

介绍了AP1000核电机组辐射控制区出入口的组成、功能和主要设备,分析了设计方案存在的问题,结合我国已运行核电机组管理经验,提出AP1000核电机组辐射控制区出入口布置的改进方案。     

A proposal for an aseismic design method of equipment and piping for NPPs in a low seismicity area

Recently a regulatory code for an aseismic design of high-pressure gas facilities became effective by the order of the Ministry of International Trade and Industry (MITI) in Japan. This order includes details of the aseismic design of vessels whose “factor of importance” are relatively lower than Class A (Class I) items in nuclear power plants.The author develops his idea on an aseismic design method of equipment and piping of nuclear power plants in a Low Seismicity Area (LSA) based on his experience of the new code for petro-chemical industries and oil refinaries pertaining to high pressure gas facilities mentioned above.The definition of LSA is usually the area whose maximum intensity has never exceeded MMI VI or VII. However, there are two types of LSA, one is really such a low seismicity area, and the other type is the area which has the possibility of stronger earthquake occurrence than those mentioned above, even though it is low. One of the typical examples is the area subjected to “New Madrid Earthquake-1812”. The author develops his concept along these two lines.He briefly describes the new code for high-pressure gas facilities in Japan. This code describes the design methodology of both types of aseismic design analysis, that is, rather sophisticated dynamic methods for facilities whose potential hazard is as high as those in a nuclear power plant, such as liquified chlorine gas storage, and simplified dynamic and static methods for most of the equipment and vessels in those plants. One of the features of this code is the use of design formulae and charts to simplify their design procedure as well as the set of specific computer codes by the MITI. These computer codes are prepared by the MITI or approved by the MITI for providing equivalent capability to the practice designated in the MITI order.The author's philosophy for the code of equipment and pipings in LSA is that they must be as simple as possible, and most of the analytical work for the design should be eliminated, or at least limit the use of simplified methods, such as the static seismic coefficient method or the modified seismic coefficient method with a simplified response spectrum. The use of general design criteria or a guideline of structural details may be better than a sophisticated design analysis as a result.     

加压贮存和活性炭吸附在核电站放射性废气处理中的应用

加压贮存和活性炭吸附是目前国内处理压水堆核电站放射陛废气最常用的两种方法。以大亚湾核电站和田湾核电站的放射性废气处理系统为例,并结合国内其他核电机组同类系统,分别介绍这两种方法在国内核电站的应用情况,并分析它们各自的优缺点：加压贮存系统结构简单但设备体积庞大,适合处理流量变化较大的放射性废气;活性炭吸附具有安全性高,操作简单的特点,适合于处理流量较小的放射性废气。     

核电仪控系统的网络安全监测

仪表和控制系统作为核电厂中枢神经的重要组成部分,对于核电厂安全稳定的运行起到关键的作用。近年来,数字化仪控系统遇到网络安全的新问题,即预防、检测和应对针对仪控系统实施的运用数字化手段的恶意行为。这种网络攻击将致使电厂性能下降,实体设备受损,甚至引发事故工况。依据核电厂纵深防御的安全设计理念和国内外法规标准的要求,提出应对核电仪控系统实施网络安全监测。网络安全监测的实施方案为,在仪控系统中部署入侵检测系统,监视并分析仪控系统事件,发现未经授权访问仪控系统资源的尝试并给出报警。本文详细分析了部署入侵检测系统需要考虑的检测点部署位置、监测数据源、入侵检测方法3个方面的问题,并给出了评价方式。     

海上浮动核电站应急撤离分析

核事故应急撤离是核应急响应的重要组成部分, 目的在于快速有效地将可能受到事故影响的人员转移至安全地区。本文根据海上浮动核电站的运行场址与运行特点, 对海上浮动核电站应急响应特征进行分析, 给出了浮动核电站应急等级划分和应急计划区范围。结合陆地核电站场区撤离与海洋平台撤离疏散方法, 制定了海上浮动核电站应急撤离情景与撤离分析假设。对浮动核电站人员撤离的分析结果表明, 浮动核电站人员撤离满足客船撤离要求, 及海上浮动核电站应急撤离的时间要求。关键词: 海上浮动核电站; 核应急; 应急计划区;应急响应; 应急撤离     

多机组核电厂场内应急准备经验

福岛核事故后,我国监管部门要求核电厂开展针对多机组事故组合情况下的应急准备。本文依据国际原子能机构对多机组事故应急准备和响应的建议和国家核安全局对多堆厂址的相关要求,从多机组事故应急管理、事故组合及应急状态的确定、应急响应涉及的应急组织及其协调机制、事故时应急状态等级的确定流程等方面分析了我国核电站多机组核事故应急准备的内容,讨论了应对多机组事故需要加强的应急能力建设,为国内其他核设施的应急准备管理提供参考。     

Evaluation of gamma-ray and neutron energy for area monitoring system in the IFMIF/EVEDA accelerator building

The engineering validation of the IFMIF/EVEDA prototype accelerator, up to 9 MeV by supplying the deuteron beam of 125 mA, will be performed at the BA site in Rokkasho. A design of this area monitoring system, comprising of Si semiconductors and ionization chambers for covering wide energy spectrum of gamma-rays and ³He counters for neutrons, is now in progress. To establish an applicability of this monitoring system, photon and neutron energies have to be suppressed to the detector ranges of 1.5 MeV and 15 MeV, respectively. For this purpose, the reduction of neutron and photon energies throughout shield of water in a beam dump and concrete layer is evaluated by PHITS code, using the experimental data of neutron source spectra. In this article, a similar model using the beam dump structure and the position with a degree of leaning for concrete wall in the accelerator vault is used, and their energy reduction including the air is evaluated. It is found that the neutron and photon flux are decreased by 10⁴-order by employing the local shields using concrete and polyethylene around beam dump, and the photon energy can be suppressed in the low energy.     

Ageing management strategies for instrumentation and control equipment in NPPs

This paper aims to explain main ageing stressors and effects on instrumentation and control (I&C) components followed by ageing management strategies through developed design such as using freewheeling diode to eliminate relays contact ageing due to inductive loads, or through traditional practices and new online monitoring where I&C components can be checked and assessed remotely while the plant is operating. It also includes some recommendations for obsolescence management for special I&C components like microcontrollers and programmable logic controllers (PLCs).     

核电厂非能动安全壳冷却系统特性研究

RELAP5和MELCOR程序结合,计算安全壳和非能动安全壳冷却系统(PCS)的瞬态响应特性,并对影响系统运行的压力、冷凝器换热面积和冷热源高度差等主要参数进行分析。结果表明:PCS能够在一定时间内有效实现安全壳降温、降压,但长期阶段仍需进行补水降温的操作。降低压力可以提高换热功率,但有可能产生流动不稳定性现象。通过增加冷凝器换热面积能够提高换热功率,但冷却水箱温度较高时影响不明显。在本文给定边界条件下,无论是单相还是两相自然循环,冷热源高度差变化对换热功率的影响都不显著。