《压水堆核电厂安全降压和排气系统设计准则》等3项标准通过审查

<正>能源行业核电标准化技术委员会秘书处于2014年12月22日~24日在北京组织召开了核电标准审查会,本次会议审查了由中国核动力研究设计院主编的《压水堆核电厂安全降压和排气系统设计准则》、《压水堆核电厂反应堆压力容器及反应堆冷却剂系统管道和设备保温层设计制造规范》和中广核工程有限公司主编的《核级金属波纹管膨胀节设计制造规范》。来自环境保护部核与辐射安全中心、上海核工程研究设计院、中国核电工     

工作简讯

《压水堆核电厂汽轮机转子设计制造规范第1部分：整锻转子》等3项标准通过审查
　　能源行业核电标准化技术委员会秘书处于2014年7月10日～11日,在北京组织召开了核电标准的审查会。本次会议审查了由哈尔滨汽轮机厂有限责任公司主编的《核电厂汽轮机转子设计制造规范第1部分：整锻转子》、《核电厂汽轮机转子设计制造规范第2部分：套装转子》和《核电厂汽轮机转子设计制造规范第3部分：焊接转子》3项标准。来自上海发电设备成套设计研究院、中国核电工程有限公司、上海核工程研究设计院等8家单位的18位专家和代表参会。     

《压水堆核电厂汽轮机转子设计制造规范第1部分:整锻转子》等3项标准通过审查

正能源行业核电标准化技术委员会秘书处于2014年7月10日~11日,在北京组织召开了核电标准的审查会。本次会议审查了由哈尔滨汽轮机厂有限责任公司主编的《核电厂汽轮机转子设计制造规范第1部分:整锻转子》、《核电厂汽轮机转子设计制造规范第2部分:套装转子》和《核电厂汽轮机转子设计制造规范第3部分:焊接转子》3项标准。来自上海发电设备成套设计研究院、中国核电工程有限公司、上海核工程研究设计院等8家单位的18位专家和代表参会。     

《压水堆核电厂调试大纲编写规范》通过审查

2013年1月11日,能源行业核电标准化技术委员会秘书处组织相关专家对秦山核电有限公司主持修订的《压水堆核电厂调试大纲编写规范》进行了讨论和审查。来自环境保护部核与辐射安全中心、中国核电工程有限公司调试中心、中广核工程有限公司调试中心、国核工程有限公司调试中心及工程建设和设计单位的13位专家对标准编写组提交的审查材料进行了审查,针对标准内容展开了深入讨论和分析,提出了具体的修改意见和建议。     

改进型压水堆核电站的总体安全要求——对URD和EUR的初步探讨

本文分析比较了URD和EUR对改进型压水堆核电站在安全、性能、建造和经济目标等总体方面特别是安全方面的要求。介绍了为满足这些要求在设计上所采取的措施。同时也比较了我国现行核电规范的要求并提出了对差别的看法。     

《压水堆核电厂控制棒驱动机构设计制造规范》通过审查

2012年7月4日至6日,由能源行业核电标准化技术委员会秘书处组织有关单位代表和专家在北京召开了能源行业核电标准《压水堆核电厂控制棒驱动机构设计制造规范》送审稿的审查会。来自环保部核与辐射安全中心、东方电气集团东方汽轮机有限公司、江苏核电有限公司、上海核工程研究设计院、中广核工程有限公司、中国核动力研究设计院、中核核电运行管理有限公司、上     

三维设计和协同设计平台在核电设计院的应用

结合上海核工程研究设计院(SNERDI)三维设计和协同设计平台的建立、推广和应用经验,阐述了以传统设计方法为主流的核电设计院应如何完整、有序、可控地建立满足核电工程设计需求和核电质量保证要求的协同设计平台的方法.     

《核电厂取排水隧洞结构设计规范》(送审稿)通过审查

<正>2015年6月15日,能源行业核电标准化技术委员会秘书处组织召开了能源行业核电标准《核电厂取排水隧洞结构设计规范》送审稿审查会。来自环境保护部核与辐射安全中心、中国核电工程有限公司、中广核工程有限公司、上海核工程研究设计院等9家单位的14位专家和代表参会。该标准规定了核电厂取排水隧洞的结构设计、施工监测和运行要求。适用于核电厂     

能源行业核电标准《压水堆核电厂反应堆冷却剂主管道设计制造规范》通过审查

正能源行业核电标准化技术委员会秘书处于2014年8月6日~7日,在北京组织审查了由中国核动力研究设计院主编的核电行业标准《压水堆核电厂反应堆冷却剂主管道设计制造规范》。来自环保部核与辐射安全中心、中国核电工程有限公司等9家单位的12位专家和代表参加了此次标准审查会。与会专家听取了编制单位代表对标准编制情况和相关意见处理情况的汇报,经过对标准内容的深入讨论,认为本标准(送审稿)     

2项能源行业核电标准送审稿通过审查

正2018年4月25日～26日,能源行业核电标准化技术委员会秘书处组织召开了能源行业核电标准《核电厂地脚螺栓设计技术规程》(送审稿)和《压水堆核电厂预埋件制作与安装技术要求》(送审稿)审查会。来自生态环境部核与辐射安全中心、中国核电工程有限公司、上海核工程研究设计院有限公司、中广核工程有限公司等12家单位的专家和代表参会。     

美国核电厂循环冷却水系统运行实践及借鉴

对美国不同类型核电厂所采用的循环冷却水系统进行了概括和总结,给出了直流、自然和机械通风冷却塔、冷却池及一次和二次循环混合使用等冷却方式在美国若干核电厂址的应用实践,并提出了针对我国核电厂冷却水系统的可借鉴之处。     

Thermal and gravity stresses in hyperboloidal cooling towers

Nuclear power plants usually require particularly large cooling towers, the design of which necessitates detailed analyses of the structural behavior under thermal and mechanical loadings.Thermal stresses are obtained in this paper for a reinforced hyperboloidal cooling tower for a wide range of foundation constants. The temperature field in the tower is assumed to be asymmetric with a linear variation through the thickness of the tower and an arbitrary variation along the height. The effects of the weight of the tower are included. The reinforcing material is assumed to have an equivalent orthotropic distribution along meridional and circumferential coordinate lines, and it is taken to be symmetrically located relative to the middle surface of the tower. The partial differential equations of equilibrium of the tower are based upon thin elastic shell theory and utilize the Love-Kirchhoff approximation. Fourier expansions in the circumferential variable are used to reduce these equations to ordinary differential equations in the meridional variable. Numerical solutions to the ordinary differential equations are obtained by the finite element subdomain collocation method.Numerical results are presented for thermal stresses in the Beaver Valley Station cooling tower for assumed temperature distributions on the inner and outer tower surfaces. The results show that the difference in temperature distributions on the inner and outer tower surfaces gives rise to significant thermal stresses in the tower. An example of a hot spot on a cylindrical shell is also presented. The analysis is particularly applicable to reinforced concrete cooling towers with hot spots.     

ARARAT — A computer code for thermal design of cooling towers

A computer program is presented for thermal and hydraulic design of cooling towers. Options have been provided for the evaluation of cooling tower size and performance curves by applying a basic physical model of heat and mass transfer.The solution is conducted by multiple iteration, in which iteration loops are mutually inclusive. Both film and spray-filled cooling towers are considered with either induced or natural air circulation.Numerical solutions are presented to a number of natural draft cooling towers which serve present nuclear or conventional power plants.     

Analysis of BDBA in RBMK-1500 reactor with long-term loss of heat removal from the core

The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: depressurisation of reactor cooling system (RCS) (if pressure in cooling circuit is high) and supply of water into cooling system from low pressure water sources, removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using cooling circuit of control and protection system channels, etc. The possibility to remove the heat using cooling circuit of control and protection system channels looks very attractive, because the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The heat from fuel channels, where heat is generated, through graphite columns is transferred in radial direction to cooled channels with control rods. Therefore, the heat removal from RBMK-1500 reactor core using control rods cooling circuit can be used as non-regular mean for reactor cool-down in case of BDBAs with loss of long-term heat removal from the core.     

核电厂内部水淹概率风险分析

福岛核事故后,外部灾害事件对核电厂安全的影响逐渐受到重视,而核电厂内部水淹是常见的一种重要灾害,可能导致核电厂发生严重事故,通常无法通过外部事件筛选分析筛除,需对此进行定量评估。本文在对核电厂水淹概率风险分析方法进行了大量研究的基础上,探讨了发生水淹对核电厂设计和运行的影响,提出了内部水淹风险的贡献项。结果表明：电气厂房和设备冷却水系统分别是对内部水淹导致的堆芯损坏频率贡献最大的厂房和水淹源,需在相关区域适当增加水淹防护措施并加强水淹探测及报警能力。     

Modeling of void, fast power and graphite temperature reactivity coefficients measurements for the validation of RELAP5-3D RBMK-1500 reactor model

This paper deals with the modeling of RBMK-1500 specific transients taking place at Ignalina NPP: measurements of void and fast power reactivity coefficients, as well as change of graphite cooling conditions transient. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and based on the obtained experimental results the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is unique and important from the point of view of model validation for the gap between fuel channel and the graphite bricks. The measurement results, obtained during this transient, enabled to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.     

Seismic analysis and design of hyperbolic cooling towers

The safety of hyperbolic cooling towers is important to the continuous operation of a power plant. Depending upon the site, earthquake may govern the design of the tower. Methods of seismic analysis have been presented. It is concluded that the response spectrum method of analysis is of maximum practical use. A method to construct the design response spectra for various earthquake zones is presented. An earthquake motion consists of three components; however, it is shown that designing for one horizontal component only is adequate. The use of boundary conditions and the effects of inelastic action on analysis and design are discussed.     

核电厂人因工程运行经验评审研究

本文论述了人因工程(HFE)在核电厂设计,尤其在人机接口设计中的重要性;简要介绍了核电TAN工程的范围和内容.本文的重点是核电厂人因工程运行经验评审的意义和作用,以及运行经验评审的范围、内容和具体实施过程.文章总结了运行经验评审的范围:分析了运行经验评审的具体内容,并举例予以说明;归纳了运行经验评审一般的实施 过程.最后,简单介绍了目前国内运行经验评审的形势.     

基于CIMS的核电厂集成管理系统设计研究

以计算机与现代集成制造系统（CIMS）的“集成”思想为指导,提出了核电厂集成管理的新方法。对核电厂集成管理系统的设计进行了分析。采用数据库和工程数据管理技术（PDM）实现核电厂异构环境下的信息集成。针对核电厂的特点和管理需要,先将核电厂的所有工作按性质分为不同的类型,然后在总系统下设计若干个分系统,每个分系统完成一种类型的管理工作.在基于CIMS的集成环境下．每个分系统既相对独立又相互联系,通过全局信息共享系统实现信息共享。该方法为充分利用核电厂的人力,物力与信息资源,构建核电厂科学的管理体系提供了一个新思路。     

核动力仿真技术及其发展

文章评述核动力仿真技术的发展状况及其特点，重点分析了核动力仿真机的发展，探索了核动力仿真技术发展的新动向。分析指出：模块化、集成化、数字化、可视化、虚拟化、网络化和智能化仿真是未来核动力仿真技术发展的重要趋势；核动力仿真逐渐突破传统的模式，向以三维数字化仿真设计为基础的核动力系统设计、制造方面拓展；以全寿期管理为目标的数字化核电厂设计是未来核动力仿真技术的一个重要的研究与应用领域。